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TRISO-Coated Particle Fuel Performance
3.07
TRISO-Coated Particle Fuel Performance
D. A. Petti, P. A. Demkowicz, and J. T. Maki Idaho National Laboratory, Idaho Falls, ID, USA
R. R. Hobbins RRH Consulting, Wilson, WY, USA
Published by Elsevier Ltd.
3.07.1
Introduction
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3.07.2 3.07.2.1 3.07.2.1.1 3.07.2.1.2 3.07.2.1.3 3.07.2.1.4 3.07.2.1.5 3.07.2.1.6 3.07.2.1.7 3.07.2.1.8 3.07.2.1.9 3.07.2.1.10 3.07.2.2 3.07.2.2.1 3.07.2.2.2 3.07.2.2.3 3.07.2.2.4 3.07.2.2.5 3.07.2.2.6 3.07.2.2.7 3.07.2.2.8 3.07.2.2.9 3.07.2.3 3.07.2.3.1 3.07.2.3.2 3.07.2.3.3 3.07.2.3.4 3.07.2.3.5 3.07.2.3.6 3.07.2.3.7 3.07.2.3.8 3.07.2.3.9 3.07.2.3.10 3.07.2.3.11 3.07.2.3.12 3.07.2.3.13 3.07.2.4 3.07.2.5 3.07.2.6 3.07.2.7 3.07.2.7.1 3.07.2.7.2 3.07.2.7.3 3.07.2.7.4
Irradiation Performance Overview of Irradiation Facilities and Testing BR-2 IVV-2M HFR Petten HFIR ATR SAFARI TRISO-coated particle fuel irradiation testing Thermal and physics analysis considerations Gas control system considerations FPMS considerations German Experience R2-K12 and R2-K13 BR2-P25 HFR-P4 SL-P1 HFR-K3 FRJ2-K13 FRJ2-K15 FRJ2-P27 HFR-K6 and HFR-K5 US Experience F-30 HRB-4 and HRB-5 HRB-6 OF-2 HRB-14 HRB-15B R2-K13 HRB-15A HRB-16 HRB-21 NPR-1 and NPR-2 NPR-1A AGR-1 European Experience Chinese Experience Japanese Experience Irradiation Performance Summary Heavy metal contamination In-service failures Failure mechanisms Acceleration effects
154 154 154 154 155 155 155 155 156 156 158 159 160 160 161 162 162 163 163 164 164 165 165 166 167 169 170 171 173 174 174 175 176 177 178 178 180 181 184 185 185 186 186 187 151
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TRISO-Coated Particle Fuel Performance
3.07.3 3.07.3.1 3.07.3.1.1 3.07.3.1.2 3.07.3.1.3 3.07.3.1.4 3.07.3.2 3.07.3.3 3.07.3.3.1 3.07.3.3.2 3.07.3.3.3 3.07.3.3.4 3.07.3.4 3.07.3.4.1 3.07.3.4.2 3.07.3.5 3.07.4 References
Safety Testing Facility Overview Ku¨FA at ITU INL’s FACS ORNL’s Core Conduction Cooldown Test Facility KORA German Experience European Experience AVR 73/21 AVR 74/18 HFR K6/3 HFR K6/2 US Experience and Future Plans Past experience Future plans Japanese Experience Conclusions
Abbreviations AGR ATR AVR BAF BISO BOL BR-2 CCCTF CVD DOE EFPD EOL FACS FIMA FPMS FRJ GETR HEU HFEF HFIR HFR HRB HTGRs HTR-10 HTTR IFEL IMGA
Advanced Gas Reactor Advanced Test Reactor Arbeitsgemeinschaft Versuchsreaktor Bacon anisotropy factor Bi-structural isotropic Beginning of life Belgian Reactor 2 Core Conduction Cooldown Test Facility Chemical vapor deposition Department of Energy Effective full-power day End of life Fuel accident condition simulator Fissions per initial metal atom Fission product monitoring system Research Reactor Juelich General Electric Test Reactor Highly enriched uranium Hot Fuel Examination Facility High-Flux Isotope Reactor High-Flux Reactor HFIR Removable Beryllium High-temperature gas-cooled reactors High Temperature Reactor 10 High-temperature test reactor Irradiated fuel examination laboratory Irradiated microsphere gamma analyzer
INET
189 189 189 190 192 193 193 199 199 199 200 200 202 202 205 206 209 212
Institute of Nuclear and New Energy Technology INL Idaho National Laboratory IPyC Inner pyrolytic carbon ITU Institute for Transuranium Elements JMTR Japan Material Test Reactor KuFA Cold finger apparatus (in German) LEU Low-enriched uranium LHTGR Large High Temperature Gas Reactor LTI Low temperature isotropic MOL Middle of life NE-MHTGR Commercial version of NP-MHTGR NGNP Next Generation Nuclear Plant NP-MHTGR New Production Modular High-temperature Gas-Cooled Reactor ORNL Oak Ridge National Laboratory ORR Oak Ridge Research Reactor PIE Postirradiation examination R&D Research and development R/B Release to birth ratio SiC Silicon carbide TRIGA Training research and isotope production, General Atomics TRISO Tristructural isotropic UCO Uranium oxycarbide Uranium dioxide UO2 VHTR Very-high-temperature reactors VXF Vertical experimental facility WAR Weak acid resin
TRISO-Coated Particle Fuel Performance
3.07.1 Introduction For all high temperature gas reactors (HTGRs), tristructural isotropic (TRISO)-coated particle fuel forms the heart of the concept. Such fuels have been studied extensively over the past four decades around the world, for example, in countries including the United Kingdom, Germany, Japan, United States, Russia, China, and more recently, South Africa. In early gas-cooled reactors, the coated particle fuel form consisted of layers of carbon surrounding the fissile kernels. Highly enriched uranium (HEU) and thorium carbides and oxides were used as fissile and fertile kernels. Ultimately, the carbon layer coating system (termed BISO for bistructural isotropic) was abandoned because it did not sufficiently retain fission products, leading to the development of the current three-layer coating system (termed TRISO for tristructural isotropic). In TRISO-coated fuel, a layer of silicon carbide (SiC) is sandwiched between pyrolytic carbon layers. This three-layer system is used to both provide thermomechanical strength to the fuel and contain fission products. In addition, for operational and economic reasons, the fuel kernel of choice today is low-enriched uranium (LEU) uranium dioxide (UO2) for the pebble bed design and uranium oxycarbide (UCO) for the prismatic design. In both pebble bed and prismatic gas reactors, the fuel consists of billions of multilayered TRISOcoated particles (750–830 mm in diameter) distributed within fuel elements in the form of circular cylinders (12.5 mm in diameter and 50 mm long) called ‘compacts’ or spheres called ‘pebbles’ (6 cm in diameter). The active fuel kernel is surrounded by a layer of porous carbon, termed ‘the buffer’; a layer of dense carbon, termed ‘the inner pyrolytic carbon layer’; a layer of SiC; and another dense carbon layer, termed ‘the outer pyrolytic carbon layer.’ These collectively provide for accommodation and containment of fission products generated during operation. The buffer layer is designed to accommodate fission recoils, volumetric swelling of the kernel, and fission gas released under normal operation. The inner pyrolytic carbon layer protects the kernel from reactive chlorine compounds produced during SiC deposition in the chemical vapor deposition (CVD) coater. The SiC layer provides structural strength to the particle. The outer pyrolytic carbon layer protects the particles during formation of the fuel element. Under normal operation, radiation damage causes shrinkage of the pyrolytic carbon layers, which induces compressive stresses in the SiC layer to counteract tensile stresses associated with fission gas release. All three layers of
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the TRISO coating system exhibit low permeability. These fuel constituents are extremely stable and are designed not to fail under normal operation or anticipated accident conditions, thereby providing effective barriers to the release of fission products. Figure 1 is a montage of TRISO fuel used in both prismatic and pebble bed high-temperature gas reactors. Rigorous control is applied at every step of the fabrication process to produce high-quality, very lowdefect fuel. Defect levels are typically on the order of one defect per 100 000 particles. Specifications are placed on the diameters, thicknesses, and densities of the kernel and layers; the sphericity of the particle; the stoichiometry of the kernel; the isotropy of the carbon; and the acceptable defect levels for each layer. Statistical sampling techniques are used to demonstrate compliance with the specifications usually at the 95% confidence level. For example, fuel production for German reactors in the 1980s yielded only approximately 100 defects in 3.3 million particles produced. This remains the standard for gas-cooledreactor fuel production today.1,2 Irradiation performance of high-quality, lowdefect coated particle fuels has been excellent. In Section 3.07.2, a detailed review of the state of the art in irradiation testing, capabilities of existing fission reactors worldwide to irradiate TRISO fuel, and the irradiation behavior of modern TRISO-coated particle fuel around the world will be discussed. Testing of German fuel under simulated accident conditions in the 1980s has demonstrated excellent performance. Section 3.07.3 describes the accident behavior of TRISO-coated particle fuel largely on the basis of the German database and the plans to perform similar testing for the current generation of TRISOcoated fuels. Additional limited testing of TRISOcoated particle fuel performed under air and water ingress events and under reactivity pulses has been reported elsewhere3 and will not be repeated here. The outstanding irradiation and accident simulation testing results obtained by German researchers form the basis for fuel performance specifications used in gas-cooled-reactor designs today. Specifications for in-service failure rates under irradiation and accident conditions are very stringent, typically on the order of 104 and 5 104, respectively. Significant research and development (R&D) related to TRISO-coated fuels is underway worldwide as part of the activities of the Generation IV International Forum on Very-High-Temperature Reactors (VHTRs). The focus is largely on extending the capabilities of the TRISO-coated fuel system for higher
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TRISO-Coated Particle Fuel Performance
Pyrolytic carbon Silicon carbide Uranium dioxide or oxycarbide kernel
Prismatic
Pebble
Particles
Matrix
Compacts
Fuel element
TRISO-coated fuel particles (left) are formed into fuel compacts (center) and inserted into graphite fuel elements (right) for the prismatic reactor
Kernel Buffer layer
5 mm graphite layer Coated particles imbedded in graphite matrix
Inner PyC-layer Fuel-free shell SiC-layer Fueled zone Outer PyC-layer
Fuel sphere Dia 60 mm
Half section
TRISO-coated fuel particles are formed into fuel spheres for pebble bed reactor Figure 1 TRISO-coated particle fuel and compacts and fuel spheres used in high temperature gas reactors.
burnups (10–20%) and higher operating temperatures (1250 C) to improve the attractiveness of hightemperature gas-cooled reactors as a heat source for large industrial complexes where gas outlet temperatures of the reactor would approach 950 C.4 Of greatest concern is the influence of higher fuel temperatures and burnups on fission product interactions with the SiC layer leading to degradation of the fuel and the release of fission products. Activities are also underway around the world to examine modern recycling techniques for this fuel and to understand the ability of gas reactors to burn minor actinides.5,6
3.07.2.1.1 BR-2
The Belgian Reactor 2 (BR-2) reactor is a materials test reactor in Mol, Belgium7 that produces very fast (3.5 1014 neutrons cm2 s1 [E > 1 MeV]) and thermal neutron fluxes (1012 neutrons cm2 s1). The facilities have irradiation test rigs (15 mm ID and 400 mm long) that can be used to irradiate coated-particle gas reactor fuel forms. They have adequate flux, fluence, and temperature characterization for the capsule, and have the infrastructure needed for capsule disassembly and postirradiation examination (PIE). The capsule size precludes irradiation of pebbles; however, it could handle approximately six to eight fuel compacts.
3.07.2 Irradiation Performance 3.07.2.1 Overview of Irradiation Facilities and Testing This section provides a brief overview of irradiation facilities that are available today to perform TRISOcoated particle irradiations.
3.07.2.1.2 IVV-2M
The IVV-2M is a 15-MW water-cooled reactor that has been used in Russia for a variety of coatedparticle testing.8 Four different test rigs have been used to test specimens ranging from particles, to compacts, to spheres. The coated particle ampoule
TRISO-Coated Particle Fuel Performance
is a noninstrumented rig that can hold 10–13 graphite disks (15 mm in diameter and 2 mm thick), each of which can hold 50 particles. The rig can also hold coated particles in axial holes, 1.2 mm in diameter, and a uniform volume of coated particles, 12–18 mm in diameter and 20–255 mm high, in a graphite matrix. Another rig, termed a ‘CP hole,’ is 27 mm in diameter and that can handle six to eight capsules. A third rig, identified as ASU-8, is a 60-mm hole that can handle three compacts. The largest channel available is Vostok, which is 120 mm in diameter and contains four cells. All of these rigs can irradiate fuel at representative temperatures, burnups, and fluences for HTGRs. There is a large degree of flexibility in the testing options at IVV-2M. Their rigs can handle particles, compacts, and spheres. 3.07.2.1.3 HFR Petten
The High Flux Reactor (HFR) in Petten, Netherlands, is a multipurpose research reactor with many irradiation locations for materials testing.9 The HFR has two different types of irradiation rigs/locations in the facility: one that can accommodate compacts and another that can accommodate spheres. Rigs for spheres are multicell capsules, 63–72 mm in diameter that can handle 4–5 spheres in up to 4 separate cells. For compacts rigs/locations are 32 mm in diameter and 600 mm in useful length. They can handle three or four parallel channels of compacts. For the threechannel configuration, approximately 30 compacts could be irradiated in the rig. There is a large axial flux gradient across the useable length (40% spread maximum to minimum) that must be considered in the design of any experiment. 3.07.2.1.4 HFIR
The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is a light-water-cooled, beryllium-reflected reactor that produces high neutron fluxes for materials testing and isotope production.10 Two specific materials irradiation facilities locations are available for gas reactor fuel testing: (a) the large RB positions (eight total) that are 46 mm in diameter and 500 mm long, and can accommodate capsules holding up to 24 compacts (three in each graphite body, eight bodies axially) in a single swept cell; and (b) the small vertical experimental facility (VXF) positions (16 total) that are 40 mm in diameter and 500 mm long, and can accommodate capsules holding up to 16 compacts (eight in each graphite body, two bodies axially) in a single swept cell. Capsules can be irradiated in the lower flux small
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VXF positions and then moved to the higher flux removable beryllium positions. Neither of these positions can accommodate pebbles. A third facility, the large VXF positions (six total), are farther out in the reflector (and therefore have lower fluxes), but are 72 mm in diameter and also 500 mm long. As with the HFR, there is a large axial flux gradient that must be considered in the design of any experiment in any of these facilities. 3.07.2.1.5 ATR
The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) is a light-water-cooled, berylliumreflected reactor fuel in a four-leaf clover configuration to produce high neutron fluxes for materials testing and isotope production.11 The clover leaf configuration results in nine very high flux positions, termed ‘flux traps.’ In addition, numerous other holes of varying size are available for testing. Several positions can be used to irradiate coated-particle fuel. The 89-mm-diameter medium I position (16 total) and the 100–125-mm-diameter flux traps can accommodate pebbles. Specifically, the use of a medium I position early in the irradiation, required because of the enrichment of the fuel, followed by transfer of the test train to the northeast flux trap can provide irradiation conditions representative of a pebble bed reactor. Approximately 10–12 pebbles in five or six individually swept cells can be envisioned in the test train. The large B positions in ATR (four total) are 38 mm in diameter and 760 mm in length. They can accommodate six individually swept cells, with two graphite bodies per cell, containing up to three 2-in. long compacts per body. Thus, 36 full-size US compacts can be irradiated in this location. Of special note, here is the very flat burnup and fluence profile available axially in the ATR over the 760 mm length. This allows for nearly identical irradiation of large quantities of fuel. 3.07.2.1.6 SAFARI
The SAFARI Reactor in Pelindaba, Republic of South Africa, is an isotope production and research reactor.12 The core lattice is an 8 9 array, consisting of 28 fuel assemblies, 6 control rods, and a number of aluminum and beryllium reflector assemblies. The reactor is cooled and moderated by light water and operates at a maximum power level of 20 MW. In-core irradiation positions include six high-flux isotope production positions: two hydraulic, two pneumatic, and two fast transfer systems that are accessible during operation. Several other irradiation
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TRISO-Coated Particle Fuel Performance
positions can also be accessed when the reactor is shut down. A large poolside facility allows for a variety of radiation applications. An intermediate storage pool and a transfer canal allow for easy and safe transport of activated materials to a hot cell. 3.07.2.1.7 TRISO-coated particle fuel irradiation testing
The historical experience in irradiation testing of coated particle fuels suggests that multicell capsules wherein fuel can be tested in separate compartments under different temperature, burnup, and fluence conditions allow for tremendous flexibility and can actually save time and money in an overall fuel qualification program. Although there are differences in details of the test trains used in each of the reactors, they share a number of important similarities in the state of the art with irradiation testing of this fuel form. In this section, these important similarities are presented to highlight the technical considerations in executing this type of testing. Because of the differences in neutron flux spectrum between a gas reactor and a light-water materials test reactor, simultaneous matching of both the rate of burnup and the rate of accumulation of fast neutron fluence is difficult to achieve. In addition, the traditional 3-year fuel cycle of high-temperature gas reactors makes real-time irradiation testing both timeconsuming and an expensive part of an overall fuel development effort. To overcome these shortcomings, irradiations in material test reactors have historically been accelerated relative to those in the actual reactor. Usually, the time acceleration is focused on achieving the required burnup in a shorter time than would be accomplished in the actual reactor, with the value of the fast fluence left as a secondary variable that must fall between a minimum and maximum value. The level of acceleration can also impact the potential for fuel failure during irradiation. The level of acceleration at a given test reactor power, coupled with fuel loading in the experiment, results in a power density for the fuel specimen in the experiment. The power density peaks at the beginning of the irradiation when the fissile content is highest and decreases as the fissile material is burned out of the fuel. As the level of acceleration increases, the temperatures in the fuel kernels increase above that in the fuel matrix because of the thermal resistances associated with the coatings of the particle,13 and the potential for high temperature, thermally driven failure mechanisms to play a deleterious role in fuel performance becomes more important.
As discussed in Section 3.07.2.7, the irradiation performance database suggests that modest levels of acceleration (1.5–3) appear to be acceptable without jeopardizing fuel performance in the irradiation, and should be a baseline requirement for future gas reactor irradiations. This acceleration level can be translated into a maximum power per fuel body or power per particle that can be used by experimenters in the design of the irradiation capsule. Given the limitations of materials test reactors around the world, the TRISO-coated particle irradiation database contains results from tests conducted under a range of accelerations. Successful German TRISO-coated particle fuel irradiations in the European HFR-Petten reactor were conducted using an acceleration of less than a factor of three. By comparison, other German irradiations in the Forschungzentrum Reaktor Juelich (FRJ) reactor at Ju¨lich had a neutron spectrum that was too thermalized. This resulted in the fuel receiving too little fast fluence to be prototypic of a high-temperature gas reactor. Similarly, historic US irradiations in ORNL’s HFIR reactor had too high a thermal flux resulting in significant burnup acceleration of the irradiation. On the basis of these considerations, the large B positions (38 mm diameter) in the ATR (see Figure 2) were chosen for the US Department of Energy’s (DOE) Advanced Gas Reactor (AGR) Program fuel irradiations because the rate of fuel burnup and fast neutron fluence accumulation in these positions provide an acceleration factor of less than three times that expected in the hightemperature gas reactor. 3.07.2.1.8 Thermal and physics analysis considerations
Given the complexity of the capsules currently being designed, the extensive review by safety authorities of the thermo-mechanical stresses, and the importance of each capsule in terms of irradiation data for fuel qualification, three-dimensional physics and thermal analyses are essential in irradiation capsule design. These analyses are critical to ensure that the fuel reaches the intended burnup, fluence, and temperature conditions. To achieve high burnups with these fuels requires detailed physics calculations to determine the time to reach full burnup. Given the concerns about severely accelerated irradiations, it is not uncommon for such irradiations to take approximately 2 years to reach full burnup in LEU TRISO-coated particles. In addition, because thermocouples should not be attached directly to the fuel, thermal analysis is used to calculate the fuel temperature during the irradiation.
TRISO-Coated Particle Fuel Performance
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North
ON-8
ON-9 ON-10 ON-11 ON-12
ON-3
ON-4
ON-5
ON-6
ON-1
Fuel elements
ON-7 ON-2
I-19
I-20
I-1
I-2
H positions
I-3
I-4
I-18 I-17
Small B position
I-5 I-6
I-16 I-15
I-7 I-8
I-14
I-13
I-12
I-11
I-10
OS-6
OS-1
I-9
East large B position location for AGR-1
In-pile tube OS-2
OS-3
OS-4
OS-5
OS-7
OS-8
OS-9
OS-10 OS-11 OS-12
Control drum
OS-13 OS-14 OS-15 OS-16 OS-17
I positions OS-18 OS-19 OS-20 OS-21 OS-22
Figure 2 Schematic of ATR showing fuel and select irradiation positions.
Examples of a test train for fuel compacts used in INL’s ATR and the pebbles used in HFR-Petten are shown in Figures 3 and 4 respectively. These irradiation capsules have extensive instrumentation to measure temperature, burnup, and fast fluence at multiple locations in the test train. Traditional commercial thermocouples have been used extensively in past irradiations, but thermocouples can suffer from drift and/or de-calibration in the reactor. Redundancy in thermocouple measurements is another consideration in light of the low reliability of thermocouples at high temperatures and long times in neutron fields typical of TRISO-coated particle fuel irradiations. Melt wires are inexpensive and have been used as a backup to thermocouples where space was available in the capsule. However, melt wires only indicate that a certain peak temperature has been reached, and not the time of that peak.
Direct temperature measurements of the coated particles are problematic because direct metal contact (e.g., thermocouple wires or sheaths) with the fuel element is not recommended as the metals can attack the TRISO fuel coatings. Thus, temperatures must be calculated on the basis of thermocouples located elsewhere in the capsule. Thermocouples are generally located as close as possible to the fuel body to minimize the uncertainties on the calculated fuel temperatures related to irradiationinduced dimensional change and thermal conductivity changes of the materials in the capsule. Encapsulating the fuel element in a graphite sleeve or cup and inserting thermocouples into the graphite has been used successfully in many designs. The high conductivity of graphite minimizes the effect of irradiation-induced dimensional changes on the calculated fuel temperature.
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TRISO-Coated Particle Fuel Performance
Gas line
Fuel stack
Thermocouple
SST holder
Thermocouple Purge gas pipe Radiation shields
Hafnium shield
Graphite cup Test fuel element
Capsule spacer nub
Figure 3 Schematic of capsule used in US INL AGR program.
Historically, metal sleeves have not been allowed to touch fuel elements because of past experiences in which SiC was attacked by transition metals (Fe, Cr, and Ni). Although quantitative data on transport rates of such metals into the fuel element and corrosion rates of the SiC are unknown, 2 or 3 mm thickness of graphite between the fuel element and the metallic components (e.g., graphite sleeve) has been found to be effective in minimizing the potential for interaction. These irradiation experiments typically include both thermal and fast fluence wires. A number of different fluence wires have been used successfully to measure thermal and fast neutron fluences in coated particle fuel irradiations. The specific type of wire to be used will depend on the measurement need (fast or thermal), the temperature it will experience during the irradiation, and compatibility with the material of encapsulation. Quartz encapsulation is not recommended for high-temperature, high-fluence applications. Neutronically, transparent refractories (e.g., vanadium) are a good alternative material of encapsulation. Inert gas filling of the flux wire
Figure 4 Schematic of pebble irradiation experiment used by the Germans.
encapsulation is recommended to ensure no oxygen interaction with the flux wire. Although fission chambers and self-powered neutron detectors have been used extensively in other reactor irradiations, they may not be practical in the space-constrained capsules expected for TRISO-coated particle fuel qualification tests. 3.07.2.1.9 Gas control system considerations
Automated gas control systems – designed to change the gas mixture in the experiment to compensate for the reduction in fission heat and changes in thermal conductivity with burnup – minimize human operator error and have proven to be a reliable method of thermal control during these long fuel irradiations. The temperature of each experiment capsule is controlled by varying the mixture of two gases with differing thermal conductivities in a small insulating gas jacket between the specimens and the experiment containment. A mixture of helium and argon has been used in the past and provides a wide temperature control band for the experiments. Unfortunately,
TRISO-Coated Particle Fuel Performance
argon cannot be used in fuel experiments where online fission product monitoring is used because the activated argon will reduce detectability of the system. Therefore, helium and neon are used instead. Computer-controlled mass flow controllers are typically used to automatically blend the gases (on the basis of feedback from the thermocouples) to control temperature. The gas blending approach allows for a very broad range of control. Automatic gas verification (e.g., by a thermal conductivity analyzer) has been implemented in some experiments to prevent the inadvertent connection of a wrong gas bottle. Gas purity is important and an impurity cleanup system should be implemented during each irradiation. Flow rates and gas tubing should be sized to minimize transit times between the mass-flow controllers and the experiment, as well as between the experiment and the fission product monitors. 3.07.2.1.10 FPMS considerations
In addition to thermal control, sweep gas is used to transport any fission gases released from the fuel to a fission product monitoring system (FPMS). A number of techniques have been used historically to quantify the release of fission gases from the fuel in these irradiation capsules. Techniques include gross gamma monitoring, online gamma spectroscopy, and offline gamma spectroscopy of grab samples. Online gross gamma monitoring of the effluent gas in the
experiment using ion chambers and sodium iodide detectors is an excellent means to capture any dynamic failures of the coated particles associated with the instantaneous release upon failure. Grab samples can provide excellent noble gas isotopic information. The temporal resolution and the number of isotopes that can be measured depend on the frequency of the grab samples and the delay time between acquisition of the grab sample and offline analysis. Weekly grab samples are typical in most irradiations, although daily or even hourly samples are possible if failure has occurred, assuming operation and associated analysis costs are not too high. Typical isotopes to be measured include 85mKr, 87Kr, 88 Kr, 131mXe, 133Xe, and 135Xe. Measurement of verylong-lived isotopes (e.g., 85Kr) would be useful in elucidating fission product release mechanisms from the kernel, but would also require waiting for the decay of the shorter lived isotopes in the sample. Online gamma spectroscopy, although the most expensive in terms of hardware costs, can provide the most detailed real-time information with detailed isotopic spectrums as often as necessary subject to data storage limitations of the system. An example of the system used for the US AGR program is shown in Figure 5. With such systems, transit times from the experiment to the detector should be minimized to allow measurement of short- and medium-lived isotopes, but must remain long enough to allow decay of
Temperature control gas mixing system
Vessel wall
He
Filter
6 5 Silver zeolite
4 Capsules in-core 3 2 1
Fission product monitoring system Grab sample
Figure 5 Integrated fission product monitoring system used in US AGR program irradiations.
H and V exhaust
Ne Particulate filters
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TRISO-Coated Particle Fuel Performance
any short-lived isotopes associated with the sweep gases (2–3 min). With this delay time, 89Kr, 90Kr, 135m Xe, 137Xe, 138Xe, and 139Xe should also be capable of being measured online. Measurements of xenon gas-release during reactor outages are recommended to provide information on iodine release behavior from the decay of xenon precursors. Multiple options for fission gas-release measurements should be considered for long irradiations where reliability of the overall fission gas measurement system can be a concern. Redundancy is also recommended for online systems so that failure of a spectrometer does not jeopardize the entire experiment. On the basis of the online concentration data, a release-to-birth ratio (R/B), a key parameter used in reactor fuel behavior studies,14 can be calculated and provide some insight into the nature of any particle failures. Because these instruments are online during the entire irradiation, a complete time history of gas release is available. Gas release early in the irradiation (i.e., from the start of the irradiation) is indicative of initially failed particles or contamination outside of the SiC layer. Release later during the irradiation is indicative of in situ particle failure. The timing of the failure data can then be correlated to temperature, burnup, and/or fluence that can be used when coupled with PIE to determine the mechanisms responsible for the fuel failure. 3.07.2.2
German Experience
Previously, particle fuel development was conducted by German researchers in support of various HTGR designs that employed a pebble bed core. These reactors were intended to produce process heat or electricity. The sequence of fuel development used by German researchers followed improvement in particle quality and performance and was largely independent of developments in reactor technology. German fuel development can be categorized according to the sequence of fuels tested as provided in Table 1. German irradiation test conditions generally covered projected fuel operating conditions, where fuel was to reach full burnup at fast fluences of 2.4 1025 n m2 and operate at temperatures up to 1095 C for process-heat applications and up to 830 C for electrical production applications. With the exception of irradiation duration, the various experiments performed bounded expected nominal conditions or were purposely varied to meet other test objectives. In order to obtain results in a
Table 1
German particle fuel development sequence
Date of design consideration
Fuel form
1972 1977
BISO coated (Th, U)O2 Improved BISO coated (Th, U)O2 TRISO-coated UCO fissile particles with ThO2 fertile particles TRISO-coated (Th, U)O2 LEU TRISO-coated UO2
1981
timely manner, tests conducted by German researchers were generally accelerated by factors of 2–3. The following sections present irradiation experiment summaries for fuels of ‘modern’ German design.1 For these experiments, this definition extends to high-enriched (Th, U)O2 TRISO-coated particles fabricated since 1977, and low-enriched UO2 TRISOcoated particles fabricated since 1981. Table 2 provides the physical attributes of the fuel used in these tests. Mixed oxide fuel test summaries are presented first, followed by the UO2 tests. 3.07.2.2.1 R2-K12 and R2-K13
The R2-K12 and R2-K13 cells were irradiated in the R2 reactor at Studsvik, Sweden. The main objective of the R2-K12 experiment was to test mixed oxide (Th, U)O2 and fissile UC2/fertile ThO2 fuel elements, whereas for R2-K13, the main objective was to test mixed oxide (Th, U)O2 fuel elements and supply fuel for subsequent safety tests. In R2-K12, four full-size spherical fuel elements were irradiated in four independently gas-swept cells. Two cells contained mixed oxide fuel spheres, while the other two contained fissile/fertile fuel spheres. As the German researchers did not develop the twoparticle fissile/fertile system further, only the mixed oxide results were reported. R2-K13 was a combined experiment with the United States. Four independently gas-swept cells were positioned vertically on top of one another. The top and bottom cells each contained a full-size German fuel sphere. The middle two cells contained US fuel and will be discussed in Section 3.07.2.3. Configuration and irradiation data from both experiments are given in Tables 3 and 4. Cold gas tests on each fuel sphere during PIE indicated that all the particles had remained intact in both R2-K12 and R2-K13. These tests are conducted after the fuel has been stored (for 14 days) at room temperature and a quasi-steady-state release of fission gas has been reached. The fuel is then swept with a carrier gas that is monitored for various fission
TRISO-Coated Particle Fuel Performance
Table 2
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Characteristics of modern German TRISO fuel particles
Particle batch
EUO 2308
EUO 2309
HT 354–383
EO 1607
EO 1674
Experiments irradiated in:
FRJ2-K13 FRJ2-P27 HFR-P4 HFR-K3 SL-P1 UO2 9.82 497 3% 10.81 94 41 36 40 895 1.00 [1.9] 3.20 1.88 1.053 1.019
FRJ2-P27 HFR-P4
FRJ2-K15
R2-K12 BR2-P25
R2-K13
UO2 9.82 497 3% 10.81 93 37 51 38 922 1.00 [1.9] 3.20 1.87
UO2 16.76 501 10.8% 10.85 92 14.3 38 3.4 33 1.9 41 3.8 906 28.8 1.013 [1.9] 3.20 1.88 1.029 1.020
(Th, U)O2 89.57 494 3% 10.12 85 39 37 39 888 1.09 1.93 3.20 1.93
(Th, U)O2 89.01 496 3% 10.10 89 37 33 39 890 1.06 1.90 3.19 1.90
Kernel form U enrichment (%) Kernel diameter (mm) Kernel density (g cm3) Buffer thickness (mm) IPyC thickness (mm) SiC thickness (mm) OPyC thickness (mm) Particle diameter (mm) Buffer density (g cm3) IPyC density (g cm3) SiC density (g cm3) OPyC density (g cm3) IPyC BAF OPyC BAF 235
Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990. Notes: The entries are one standard deviation. Entries in square brackets, [ ], are estimated values. BAF is the Bacon anisotropy factor for the layer, where values closer to one are more isotropic.
Table 3
R2-K12 and R2-K13 configuration
Number of cells Number of fuel spheres Spherical fuel element diameter Fuel zone diameter Fuel type Particle batch 235 U enrichment 235 U per fuel element 232 Th per fuel element Heavy metal per fuel element Number of particles per spherical fuel element Defective SiC layersa (U/U-total)
R2-K12
R2-K13
2 2 59.9 mm
2 2 59.77 mm
47 mm HEU (Th, U)O2 LTI – TRISO EO 1607 89.57% 1.002 g 4.961 g 6.076 g
47 mm HEU (Th, U)O2 LTI – TRISO EO 1674 89.01% 1.02 g 10.125 g 11.27 g
10 960
19 780
<1 105
<5 106
a Defective SiC layer fractions reported for German fuel are per pebble with the exception of loose particle experiments that are per particle batch. Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.
gases (usually 85mKr) and heated to 60 C. Sudden increases in the amount of fission gas detected indicate failed particles. The amount of increase is
proportional to the gas source, and in a calibrated system, indicates the number of failed particles. The fuel sphere from R2-K12 Cell 1 was partially deconsolidated and visual inspection revealed two kernels ‘without coating.’ Segments from each of the two fuel spheres were also metallographically examined; those examinations revealed a reaction zone on the inner side of the buffer layer, as well as tangential cracks between the buffer and the inner pyrocarbon layer. Only one particle exhibited a radial crack in the buffer layer beyond the reaction zone. All of the SiC and PyC layers examined remained intact. 3.07.2.2.2 BR2-P25
The BR2-P25 capsule was irradiated in the BR2 reactor at Mol, Belgium. The primary objective of this experiment was to test (Th, U)O2 mixed oxide fuel. One independently gas-swept cell contained 12 compacts. Each compact was cylindrical in shape and contained a small fuel sphere. Configuration and irradiation data are given in Tables 5 and 6, respectively. During PIE, Compacts 3 and 7 were electrolytically deconsolidated with no particle failures being evident. Ceramographic examination of cross-sections from Compacts 4 and 8 revealed some radial cracks in the buffer layers; however, no defective particles were found.
162 Table 4
TRISO-Coated Particle Fuel Performance
R2-K12 irradiation data
Start date End date Duration (full power days) Cell Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Center temperature ( C) Surface temperature ( C) BOL 85mKr R/B EOL (report date) 85mKr R/B
R2-K12
R2-K13
28 November 1978 12 February 1980 308 1 11.1 5.6 1100 950 3.9 109 3.0 107
22 April 1980 19 September 1982 517 1 10.2 8.5 1170 960 2.2 109 7.0 108
2 12.4 6.9 1280 1120 4.6 109 2.0 107
4 9.8 6.8 980 750 1.5 109 5.0 108
Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.
Table 5
BR2-P25 configuration
Number of cells Number of compacts Cylindrical compact diameter Cylindrical compact height Diameter of spherical fuel zone Fuel type Particle batch U enrichment 235 U per fuel compact 232 Th per fuel compact Heavy metal per fuel compact Number of particles per compact Number of particles per cell Defective SiC layers (U/U-total) 235
Table 7 1 12 26.58–27.74 mm 29.87–30.03 mm 20 mm HEU (Th, U)O2 LTI – TRISO EO 1607 89.57% 0.136 g 0.6744 g 0.8264 g 1490 17 880 <1 105
Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.
Table 6
BR2-P25 irradiation data
Start date End date Duration (full power days) Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Maximum temperature ( C) Minimum temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
30 October 1978 19 December 1981 350 13.9–15.6 6.2–8.1 1070 1010 2 107 1 106
Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.
HFR-P4 configuration
Number of cells Number of compacts per cell Cylindrical compact diameter Cylindrical compact height Diameter of spherical fuel zone Fuel type Particle batch – cells 1 and 3 Particle batch – cell 2 235 U enrichment Number of particles per compact Number of particles per capsule Defective SiC layers (U/U-total)
3 12 23–29 mm 32 mm 20 mm LEU UO2 LTI – TRISO EUO 2308 EUO 2309 9.82% 1630 19 600 <1 106
with 36- and 51-mm-thick SiC layers irradiated at 1000 C, beyond burnups of 12% fissions per initial metal atom (FIMA), and beyond fast fluences of 6 1025 n m2 (E > 0.10 MeV). The performance of the 36 mm SiC layer fuel was also to be evaluated at an irradiation temperature of 1200 C. Three independently gas-swept cells each contained 12 compacts that were cylindrical in shape and contained a small fuel sphere in each. Configuration and irradiation data are given in Tables 7 and 8, respectively. Note that the burnup and fast fluence goals were met, while the irradiation temperature goals were not. PIE revealed that the test articles remained intact. However, some failures caused by the thermocouples and gas inlet tubes were found on the upper compacts. 3.07.2.2.4 SL-P1
3.07.2.2.3 HFR-P4
The HFR-P4 capsule was irradiated at the HFR in Petten. The main objective of this experiment was to compare the fuel performance of particles
The SL-P1 experiment was irradiated at the Siloe¨ Reactor in Grenoble, France. The objective of the experiment was to test reference LEU fuel up to the potential limits for burnup and fast fluence at 800 C.
TRISO-Coated Particle Fuel Performance
Table 8
HFR-P4 irradiation data
Start date End date Duration (full power days) Capsule SiC layer thickness (mm) Maximum temperature ( C) Minimum temperature ( C) Maximum burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.10 MeV) BOL 85mKr R/B EOL 85mKr R/B
Table 9
Table 10
10 June 1982 28 November 1983 351 1 36
2 51
3 36
940
945
1075
915
920
1050
14.7
14.9
14.0
8.0
8.0
8.0
3.5 109 8 108
– 8 108
3.6 109 8 109
SL-P1 configuration
Number of cells Number of compacts Cylindrical compact diameter Cylindrical compact height Diameter of spherical fuel zone Fuel type Particle batch 235 U enrichment Number of particles per compact Number of particles per cell Defective SiC layers (U/U-total)
1 12 30.1 mm 30.8 mm 20 mm LEU UO2 LTI – TRISO EUO 2308 9.82% 1634 19 600 <1 106
One gas-swept cell contained 12 compacts. Each cylindrical compact contained one small fuel sphere. Configuration and irradiation data are provided in Tables 9 and 10, respectively. The operational objectives for this experiment were met. PIE revealed that none of the compacts showed mechanical failure. 3.07.2.2.5 HFR-K3
The HFR-K3 capsule was irradiated at the HFR in Petten. The primary objective of this experiment was to determine the performance of reference LEU fuel from an accelerated test. Four full-size spherical fuel elements were irradiated in three independently gasswept cells. The cells were vertically positioned on top of one another, with the middle cell containing two fuel spheres. To minimize flux gradient effects on the test fuel, the entire test rig was rotated 90 several times during the irradiation. Configuration
SL-P1 irradiation data
Start date End date Duration (full power days) Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Compact mean temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Table 11
24 June 1982 23 December 1983 330 8.6–11.3 5.0–6.8 743–794 5.8 107 1.2 106
HFR-K3 configuration
Number of cells Number of fuel spheres Spherical fuel element diameter Fuel zone diameter Fuel type Particle batch 235 U enrichment Number of particles per spherical fuel element Defective SiC layers (U/U-total)
Table 12
163
3 4 59.98 mm 47 mm LEU UO2 LTI – TRISO EUO 2308 9.82% 16 400 4 105
HFR-K3 irradiation data
Start date End date Duration (full power days) Cell/sphere Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Center temperature ( C) Surface temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
15 April 1982 5 September 1983 359 A/1 7.5 4.0
B/2 10.0 5.8
B/3 10.6 5.9
C/4 9.0 4.9
1200
920
920
1220
1020
700
700
1020
1 109 2 107
9 1010 1 107
9 1010 1 107
2 109 3 107
and irradiation data are given in Tables 11 and 12, respectively. Subsequent PIE reported no failures. 3.07.2.2.6 FRJ2-K13
FRJ2-K13 cells were irradiated at the DIDO reactor in Ju¨lich, Germany. The main objective of this test was to supply irradiated reference fuel for subsequent safety tests. Fuel performance was also to be examined under the controlled irradiation conditions of significant burnup with negligible fast neutron fluence. Four fullsize spherical fuel elements were irradiated in two
164
TRISO-Coated Particle Fuel Performance
independently gas-swept cells. The cells were vertically positioned on top of each other, with the fuel spheres similarly positioned within the cells. Configuration and irradiation data are given in Tables 13 and 14, respectively. Subsequent PIE reported no failures.
11 h. The 85mKr R/B ratio from each capsule increased to a maximum of 108 at the start of the transient and then dropped back to the pretransient levels after the temperature was returned to the nominal test condition. 3.07.2.2.8 FRJ2-P27
FRJ2-K15 cells were irradiated at the DIDO reactor in Ju¨lich, Germany. The main objectives of this test were to demonstrate the high burnup potential of reference fuel used in AVR reload 21-1 and to perform in-core temperature transient tests. Fuel performance was also to be examined under the controlled irradiation conditions of significant burnup with negligible fast neutron fluence. Three full-size spherical fuel elements were irradiated in three independently gas-swept cells. Configuration and irradiation data are given in Tables 15 and 16, respectively. Capsules 2 and 3 underwent a temperature transient test at a burnup of 10% FIMA. The temperature of the sphere surfaces was raised to 1100 C and held for
FRJ2-P27 cells were irradiated at the DIDO reactor in Ju¨lich, Germany. The main objectives of this test were to investigate fission product release at various cyclic temperatures and to determine the effectiveness of thicker SiC layers on the retention of 110mAg. Each of the three independently gas-swept cells contained three compacts and two coupons (trays). The compacts were cylindrical in shape and contained an (unspecified) outer fuel-free zone. The coupons were graphite disks with holes, annularly spaced, for the insertion of 34 particles. Of the two coupons that contained the thicker SiC particles (51 mm vs. 36 mm), one was placed in Cell 1, and the other in Cell 3. Configuration and irradiation data are provided in Tables 17 and 18, respectively.
Table 13
Table 15
3.07.2.2.7 FRJ2-K15
FRJ2-K13 configuration
Number of cells Number of fuel spheres Spherical fuel element diameter Fuel zone diameter Fuel type Particle batch 235 U enrichment Number of particles per spherical fuel element Defective SiC layers (U/U-total)
Table 14
2 4 59.98 mm 47 mm LEU UO2 LTI – TRISO EUO 2308 9.82% 16 400 4 105
FRJ2-K13 irradiation data
Start date End date Duration (full power days) Cell/sphere Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Center temperature ( C) Surface temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Number of cells Number of fuel spheres Spherical fuel element diameter Fuel zone diameter Fuel type Particle batch 235 U enrichment Number of particles per spherical fuel element Defective SiC layers (U/U-total)
Table 16
24 June 1982 12 February 1984 396 A/1 7.5 0.2
A/2 8.0 0.2
B/3 7.9 0.2
B/4 7.6 0.2
1125
1150
1150
1120
985
990
990
980
2 109 2 108
2 109 2 108
8 1010 7 109
8 1010 7 109
FRJ2-K15 configuration 3 3 60.04 mm 47 mm LEU UO2 LTI – TRISO HT 354–383 16.76% 9500 <5 105
FRJ2-K15 irradiation data
Start date End date Duration (full power days) Cell Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Center temperature ( C) Surface temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
4 September 1986 20 May 1990 590 1 14.1 0.2
2 15.3 0.2
3 14.7 0.1
970
1150
990
800
980
800
2.0 1010 1.0 108
2.47 1010 5.0 109
2.0 1010 3.0 109
TRISO-Coated Particle Fuel Performance
Table 17
FRJ2-P27 configuration
Number of cells Number of compacts per cell Number of coupons per cell Cylindrical compact diameter Cylindrical compact height Coupon diameter Coupon height Diameter of coupon fuel annulus Fuel type Particle batch for compacts and four coupons Particle batch for two coupons (thick SiC) 235 U enrichment Number of particles per compact Number of particles per coupon Number of particles per cell Defective SiC layers (U/U-total)
Table 18
Table 19 3 3 2 27.9–28.03 mm 29 mm 27 mm 2.2 mm 23 mm LEU UO2 LTI – TRISO EUO 2308 EUO 2309 9.82% 2424 34 7340 <3 106
17 February 1984 10 February 1985 232 1 7.6 1.4
2 8.0 1.7
3 7.6 1.3
1080
1320
1130
880
1220
1080
6
1.0 10 1.6 106
7
8.6 10 1.0 105
U enrichment Number of particles per spherical fuel element
HFR-K6
HFR-K5
4 4 60 mm
4 4 60 mm
LEU UO2 – TRISO 10.6% 14 600
LEU UO2 – TRISO 10.6% 14 600
were irradiated in four independently gas-swept cells. A typical reactor temperature history was simulated in the test with 17 temperature cycles (corresponding to 17 passes through the core). For one-third of a cycle, the fuel sphere center temperature was held at 800 C; for the other two-thirds of the cycle, the center temperature was 1000 C. In addition, three temperature transients (sphere center temperature held at 1200 C for 5 h) were performed at beginning of life (BOL), middle of life (MOL), and end of life (EOL). Limited configuration and irradiation data are given in Tables 19 and 20, respectively. There were no particle failures reported as a result of the irradiations.
FRJ2-P27 irradiation data
Start date End date Duration (full power days) Cell Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Center temperature ( C) Surface temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
HFR-K6 and HFR-K5 configurations
Number of cells Number of fuel spheres Spherical fuel element diameter Fuel type 235
165
3.07.2.3 8
2.0 10 1.2 107
PIE revealed that all specimens and components were in excellent condition. Cold gas tests of all compacts and coupons determined that there was only one defective/failed particle present. This particle was from a Capsule 2 coupon (with nominal SiC thickness). Ceramographic examination revealed that the particle was inserted in the coupon ‘without coating’ and that kernel interactions led to a compression of the inner side of the buffer to a thickness of 10 mm. 3.07.2.2.9 HFR-K6 and HFR-K5
The HFR-K6 and HFR-K5 capsules were irradiated at the HFR in Petten.1,9 These experiments were a proof test for HTR MODUL reference fuel. In each experiment, four full-size spherical fuel elements
US Experience
Historical US particle fuel development effort (through the mid 1990s), which included design and testing, coincided with the development of various HTGRs. This sequence of development is listed in Table 21, and identifies the main fuel forms under consideration at that time. US gas reactors were designed to use prismatic graphite blocks containing fuel compacts, and were primarily intended to produce electricity with the exception of the New Production Modular High-temperature Gas-Cooled Reactor, which was designed to produce tritium. Over the years, the design has also supported steam cycle, direct cycle, process heat, and weapons material disposition applications. More recently, DOE established the AGR Fuel Development and Qualification Program to provide a baseline fuel qualification data set at a peak fuel centerline temperature of 1250 C15,16 in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Irradiation test conditions employed by the United States generally covered projected fuel operating conditions. US fuel was to operate at temperatures as
166
TRISO-Coated Particle Fuel Performance
Table 20
HFR-K6 and HFR-K5 irradiation data
Start date End date Duration (full power days) Cell Burnup (% FIMA) Fast fluence (1025 m2, E > 0.10 MeV) Temperature EOL 85mKr R/B
Table 21
HFR-K6
HFR-K5
1990 4 May 1993 634
1991 16 May 1994 564
1 7.2 3.2
2 9.3 <4.8
3 9.7 4.8
4 9.2 <4.8
1 6.7 2.9
2 8.8 <4.3
3 9.1 4.3
4 8.7 <4.3
Cycled 3 107
Cycled 3 107
Cycled 3 107
Cycled 3 107
Cycled 3 107
Cycled 3 107
Cycled 3 107
Cycled 3 107
Historical US particle fuel development and testing sequence
Date of design conception
Reactor/status
Major fuel form tested
1960 1964
Peach Bottom built Fort St. Vrain built
1967
LHTGR design only
1984
NE-MHTGR commercial design only
1989 1995
NP-MHTGR government design only GT-MHR commercial design only
2005
NGNP design
BISO-coated (Th, U)C2 TRISO-coated (Th, U)C2 fissile TRISO-coated ThC2 fertile TRISO-coated UC2 fissile BISO and TRISO-coated ThO2 fertile TRISO-P coated UCO fissile TRISO-P coated ThO2 fertile TRISO-P coated UCO TRISO-coated UCO fissile TRISO-coated UCO and/or UO2 fertile fuel not yet tested TRISO-coated UCO fissile
high as 1400 C and reach full burnup (commensurate with 235U enrichment and kernel composition) at fast fluences of 4 1025 n m2. With the exception of irradiation duration, the various experiments performed either bounded expected nominal conditions or were purposely varied to meet other test objectives. In order to obtain results in a timely manner, US tests were accelerated by factors of 3–10. The particle fuel irradiation experiments and PIE results summarized below consider only selected tests of key US fuel types. These fuel types include TRISO fissile/BISO fertile particles, weak acid resin (WAR) TRISO fissile/BISO fertile particles, TRISO fissile/TRISO fertile particles, and TRISO-P fissile particles (conventional TRISO-coated particles with an additional ‘protective’ pyrolytic carbon layer above the outer pyrolytic carbon layer) as well as TRISO fissile particles. General Atomics and Babcock & Wilcox manufactured the majority of the kernel and coating batches. However, some of the batches were manufactured by ORNL. The following US
experiment summaries are listed in chronological order and are not grouped by fuel type. Listed configuration and irradiation data are actual values, not specification values or ranges. Interpretations of PIE results are from the original sources and no overt attempt has been made to reinterpret the results. 3.07.2.3.1 F-30
The F-30 experiment was irradiated in the General Electric Test Reactor (GETR) at Pleasanton, California.17 The primary objective of this experiment was to demonstrate the irradiation performance of Fort St. Vrain production fuel. Five independently gas-swept cells contained the fuel. Cells 1, 3, and 4 contained only fuel compacts, Cell 2 contained only loose particles, and Cell 5 contained both fuel compacts and loose particles. Configuration and irradiation data are given in Tables 22 and 23, respectively. Postirradiation metallographic examination of seven fuel compacts containing fissile and fertile particles was performed. In addition, five sets of
TRISO-Coated Particle Fuel Performance
167
loose fissile particles and five sets of loose fertile particles were examined. Fissile particle failure, defined as a crack completely through the SiC layer, ranged between 0% and 6.1%, while fertile particle failure ranged between 0% and 15.1%. A typical photomicrograph of SiC failure in an F-30 fissile particle is presented in Figure 6. Metallography revealed that inner pyrolytic carbon layers had remained bonded to the SiC layer throughout irradiation. Figure 7 displays a typical photomicrograph of a fissile particle with an IPyC layer crack and a densified buffer. 50 mm
3.07.2.3.2 HRB-4 and HRB-5
The HRB-4 and HRB-5 capsules were irradiated in HFIR at ORNL.18 The main objective of these experiments was to test candidate fuel materials and manufacturing processes for the proposed large HTGR. Each test involved a single gas-swept cell containing six fuel compacts vertically positioned.
Table 22
F30 configuration
Number of cells Total number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact lengths Fissile fuel type Nominal Th/U ratio 235 U enrichment Fissile particle diameter Fertile fuel type Fertile particle diameter Number of fissile particle batches Number of fertile particle batches Defective SiC layer fraction – fissile particles Defective SiC layer fraction – fertile particles
Table 23
Figure 6 A typical SiC layer crack in an F-30 fissile fuel particle. Reproduced from Scott, C .B.; Harmon, D. P. Post Irradiation Examination of Capsule F-30; GA-A13208, UC-77; General Atomics Report, 1975.
5 13 12.45 mm 18.54 and 49.28 mm HEU (Th, U)C2 TRISO 4.25 93% 429–560 mm ThC2 TRISO 648–771 mm 7 9 <5 104–1.5 103 3 104–1.0 103
100 mm Figure 7 A typical IPyC layer crack in a fissile F-30 fuel particle. Reproduced from Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsule F-30; GA-A13208, UC-77; General Atomics Report, 1975.
F30 irradiation data
Start date End date Duration (full power days) Cell Fissile burnup (% FIMA) Fertile burnup (% FIMA) Fast fluence (1025 n m2, E > 0.18 MeV) Time average peak temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
15 May 1972 5 April 1973 269 1 15 3 8 1100 2 106 8 106
2 19 4.5 10.5 1100 7 107 1 104
3 20 5 11.5 1120 8 107 1 105
4 18 4 9.5 1100 7 107 2 105
5 12 1.5 12 1200 2 106 2 105
168
TRISO-Coated Particle Fuel Performance
Table 24
HRB-4 and HRB-5 configurations
Number of cells Number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact lengths Fissile fuel type Fertile fuel type U enrichment Fissile particle diameter Fertile particle diameter Fissile particle batch Fertile particle batch Total number of fissile particles Total number of fertile particles
235
Table 25
HRB-4
HRB-5
1 6 12.4 mm
1 6 12.4 mm
25.4 mm
25.4 mm
WAR UC2 TRISO ThO2 BISO 5.99% 639 mm 805 mm OR52A T01424BIL 17 780
WAR UC2 TRISO ThO2 BISO 5.99% 639 mm 805 mm OR52A T01424BIL 17 780
4180
4180
Figure 8 Typical HRB-4 fissile particle irradiated to 27.7% FIMA and 10.5 1025 n m2 fast fluence. Reproduced from Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsules HRB-4, HRB-5, and HRB-6; GA-A13267, UC-77; General Atomics Report, 1975.
HRB-4 and HRB-5 irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Peak temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
HRB-4
HRB-5
8 October 1972 26 June 1973 244
8 October 1972 3 February 1973 107
27.7
15.7
13.4
4.3
10.5
4.7
1250 1.4 105 3.2 104
1250 3 106 1 104
Configuration and irradiation data are given in Tables 24 and 25. Metallographic examinations were performed on each fuel compact. A typical photomicrograph of an irradiated HRB-4 fissile particle is presented in Figure 8, which shows the formation of gas bubbles in the kernel and the densification of the buffer. IPyC layers of the examined fissile particles had remained bonded to the SiC. The examination indicated that the fissile particles had failed between 0% and 6% of the SiC layers. These failures consisted primarily of radial cracks through the SiC layer. Between 4% and 73% of
Figure 9 Photomicrographs of typical fission product attack in irradiated HRB-4 fissile particles. Reproduced from Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsules HRB-4, HRB-5, and HRB-6; GA-A13267, UC-77; General Atomics Report, 1975.
the OPyC layers failed during irradiation. There were no tabulations of IPyC layer failures reported. Several of the fissile particles examined displayed evidence of fission product attack. This attack mostly occurred in large concentrations at the IPyC–SiC interface and where fission products in smaller concentrations had diffused up to 25 mm into the SiC. Figure 9 presents typical photomicrographs of fission product attack in HRB-4 fissile particles. In HRB-5, IPyC layers of the examined fissile particles had remained bonded to the SiC. There
TRISO-Coated Particle Fuel Performance
Table 26
HRB-6 configuration
Number of cells Number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact length Fissile fuel type Nominal Th/U ratio 235 U enrichment Fissile particle diameter Fertile fuel type Fertile particle diameter Fissile particle batch Fertile particle batch Defective SiC layer fraction – fissile particles
Figure 10 Typical HRB-4 fissile particle irradiated to 27.7% FIMA and 10.5 1025 n m2 fast fluence. Reproduced from Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsules HRB-4, HRB-5, and HRB-6; GA-A13267, UC-77; General Atomics Report, 1975.
were no tabulations of IPyC layer failures reported. There was no evidence of fission product attack as seen in the HRB-4 fissile particles. However, the examination indicated that between 0.4% and 17% of the SiC layers in fissile particles had failed. These failures consisted primarily of radial cracks through the SiC layer. A typical photomicrograph of irradiated HRB-5 fissile particles with cracked SiC layers is presented in Figure 10. This photomicrograph is also representative of HRB-4 fissile particles with cracked SiC layers. It was reported that a large fraction of these cracked SiC layers were due to metallographic preparation and not a result of fast neutron exposure or fuel burnup effects. 3.07.2.3.3 HRB-6
The HRB-6 capsule was irradiated in HFIR at ORNL.18 Fissile fuel particles used in HRB-6 came from the same production batch as used in the first core of Fort St. Vrain and were one of the batches previously irradiated in the F-30 experiment. This test involved a single gas-swept cell containing six fuel compacts vertically positioned. During operation, the sweep gas flow rate was reduced because of high activity in the sweep lines. Because of this gas flow reduction, in-pile fission gas-release data were not obtained. The irradiation of HRB-6 in HFIR coincided with part of the HRB-4 irradiation. Configuration and irradiation data are given in Tables 26 and 27.
Table 27
169
1 6 12.4 mm 25.4 mm HEU (Th, U)C2 TRISO 4.25 93.15% 556 mm ThO2 BISO 888 mm CU6B-2427 T01451BIL-W <5 104
HRB-6 irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Peak temperature ( C) Minimum TRIGA BOL 85mKr R/B Maximum TRIGA EOL 85mKr R/B
27 February 1973 8 September 1973 183 26.6 9.3 7.9 1100 5.0 107 2.7 104
PIE included gas-release measurements of each fuel compact performed in the Training Research and Isotope Production, General Atomics (GA) (TRIGA) reactor. However, during the unloading of the HRB-6 capsule, fuel compacts 2A, 2B, and 2C were damaged and as many as 30 broken fuel particles were observed. Therefore, the TRIGA gasrelease measurements at EOL for these compacts would be higher than in-pile sweep line measurements had they been performed. A typical photomicrograph of an irradiated HRB-6 fissile particle is presented in Figure 11, which shows the formation of gas bubbles in the kernel and densification of the buffer. The photomicrograph also shows an incipient crack in the IPyC layer. No tabulations of IPyC layer failures were reported. IPyC layers of the examined fissile particles had remained bonded to the SiC, and there was no evidence of fission product attack. However, the examination indicated that the fissile particles had failed between 0% and 2% of the SiC layers. These failures do not include the fissile particles broken during capsule unloading. It was reported that a
170
TRISO-Coated Particle Fuel Performance
Table 28
OF-2 configuration
Number of cells Total number of fuel compacts Cell 1 cylindrical fuel compact dimensions (16 compacts) Cell 2 cylindrical fuel compact dimensions (48 compacts) Cell 2 cylindrical fuel compact dimensions (24 compacts) Fissile fuel type
235
Figure 11 Typical HRB-6 fissile particle irradiated to 26.5% FIMA and 7.9 1025 n m2 fast fluence. Reproduced from Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsules HRB-4, HRB-5, and HRB-6; GA-A13267, UC-77; General Atomics Report, 1975.
U enrichment Fissile particle diameter Fertile fuel type Fertile particle diameter Number of fissile particle batches Number of fertile particle batches
Table 29
large fraction of these failures were due to metallographic preparation. 3.07.2.3.4 OF-2
The OF-2 capsule was irradiated in the Oak Ridge Research Reactor (ORR).19 The main objectives of the test were to investigate the irradiation performance of various particle fuel forms (mostly WAR UCO with different stoichiometries) and to compare the performance of fuel particles fabricated from different coaters. OF-2 consisted of 88 fuel compacts (and several sets of loose inert particles) contained in a single capsule that was divided into two independently gas-swept cells. Various combinations from 15 fissile batches, 16 fertile batches, and 4 compact matrix compositions comprised the fuel compacts (each compact contained fuel from only one fissile batch and one fertile batch). Configuration and irradiation data are given in Tables 28 and 29. Postirradiation metallography was performed on three fuel compacts from Cell 1 and on 27 fuel compacts from Cell 2. A significant level of OPyC layer failures was observed in the fissile TRISOcoated particles from Cell 1. However, there were no observed SiC layer failures or any layer failures in the BISO-coated fertile and inert particles in these compacts. Examination of 11 fuel compacts from Cell 2, containing the same three fissile particle batches as in Cell 1, also indicated significant levels of OPyC layer failures. The fissile particle batch with the highest OPyC anisotropy (optical Bacon
2 88 15.75 mm diameter, 25.4 mm long 15.75 mm OD, 3.30 mm ID, 12.70 mm long 15.75 mm diameter, 50.8 mm long WAR UCxOy TRISO (Th, U)O2 TRISO UC2 TRISO Not reported 600–753 mm ThO2 BISO 806–889 mm 15 16
OF-2 irradiation data
Start date End date Duration (full power days) Cell Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.18 MeV) Maximum temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
21 June 1975 1 August 1976 352 1 2 75.9–79.6 50.0–79.5 5.86–8.91 1.94–8.36 1350 2 105 1 104
1350 1 104 5 106
anisotropy factor (BAF) ¼ 1.069) had 100% OPyC layer failure, while the other two batches with lower anisotropy (optical BAF of 1.035 and 1.030) had 0–33% OPyC layer failures. Of the 30 fuel compacts metallographically examined, only one compact (that contained WAR UCO fissile particles) displayed cracked SiC layers. Among the 27 fissile particles observed in this compact, 16 displayed cracked SiC layers. These cracks were identified as artifacts of polishing. However, no photomicrographs of these cracks were presented to support this conclusion. The metallographic examinations also revealed typical WAR UCO behavior of kernel and buffer densification. This densification was also accompanied by varying degrees of kernel migration. Figure 12 presents a typical WAR UCO photomicrograph that displays kernel and buffer densification, and OPyC layer failure. Examination of OF-2 particles also indicated several incidences of fission
TRISO-Coated Particle Fuel Performance
product accumulation at the IPyC and SiC interface. A typical photomicrograph of fission product accumulation is presented in Figure 13.
Figure 12 Photomicrograph of irradiated OF-2 fissile WAR UCO particle. Reproduced from Tiegs, T. N.; Thoms, K. R. Operation and Post Irradiation Examination of ORR Capsule OF-2: Accelerated Testing of HTGR Fuel; ORNL-5428; 1979. Courtesy of Oak Ridge National Laboratory, U.S. Department of Energy.
3.07.2.3.5 HRB-14
The HRB-14 capsule was irradiated in HFIR at ORNL.20 The main objectives of this experiment were to test LEU particles and to demonstrate reduced matrix–OPyC layer interactions by using cure-in-place fuel compacts. This test involved a single gas-swept cell equally divided among 20 fuel compacts vertically positioned and molded planchets (wafers) containing BISO-coated ThO2 fertile particles. Online fission gas-release measurements were not reported. Also, irradiation results from the BISOcoated fertile particles were reported separately and are not included in this summary. Configuration and irradiation data are given in Tables 30 and 31. Disassembly of the HRB-14 capsule after irradiation produced five fuel compacts with no remaining structure; in essence, there were five collections of loose particles, four compacts that were partially intact, nine compacts that were intact but displayed significant amounts of debonding, and only two compacts in relatively good shape.
Table 30
Lower half of HRB-14 configuration
Number of cells Total number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact length Fissile fuel type
235
U enrichment Fissile particle diameter Fertile fuel type Fertile particle diameter Number of fissile particle batches Number of fertile particle batches Defective SiC layer fraction – fissile particles
Table 31
Figure 13 Photomicrograph of irradiated OF-2 fissile fuel particles displaying fission product accumulation at IPyC–SiC interface. Reproduced from Tiegs, T. N.; Thoms, K. R. Operation and Post Irradiation Examination of ORR Capsule OF-2: Accelerated Testing of HTGR Fuel; ORNL-5428; 1979. Courtesy of Oak Ridge National Laboratory, U.S. Department of Energy.
171
1 20 12.50 mm 9.52 mm UCxOy TRISO (Th, U)O2 TRISO UO2 TRISO 19.18–19.66% 760–813 mm ThO2 TRISO 786–882 mm 5 8 7.0 107– 1.3 104
Lower half of HRB-14 irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Maximum temperature ( C) Minimum temperature ( C) Minimum TRIGA BOL 85mKr R/B Maximum TRIGA EOL 85mKr R/B
20 May 1978 4 January 1979 214 28.6 8.5 8.3 1190 895 3.8 107 3.0 104
172
TRISO-Coated Particle Fuel Performance
Metallographic examination was performed on 15 fuel compacts, and 8 of them contained fissile particles. A few fissile particles were reported to have SiC layer cracks but these cracks were attributed to metallographic preparation. It should be noted that visual inspection of each compact during capsule disassembly indicated that between 0% and 9% of the visible particles (from compact surfaces and loose particles that had fallen off) had failed SiC layers. However, this visual inspection did not distinguish between fissile and fertile particles. The metallographic examination of fissile particles revealed that between 0% and 3% of the IPyC layers had failed (cracked) and that the IPyC layers had debonded from the SiC in 0% to 7.7% of the particles. Buffer layers did not crack in the UO2 or (Th, U)O2 fuel but did crack in 10–71% of the UCO fuel particles. Kernel extrusion was reported only in UCO fuel. Figure 14 displays typical kernel extrusion, and Figure 15 presents a typical photomicrograph of kernel migration. In several particles of each fuel form, high concentrations of fission products were observed in small, localized regions at the SiC–IPyC layer interface. In addition to fission product accumulation, localized chemical attack was also observed in the SiC layers of several (Th, U)O2 and UO2 fuel particles. This localized attack, which had penetrated 2 mm into the SiC, was attributed to palladium, and was
Figure 14 Photomicrograph of a UCO particle (batch 6157-08-020) from Compact 10 irradiated at 1040 C to 27.8% FIMA and to a fast fluence (E > 0.18 MeV) of 7.1 1025 n m2 displaying kernel extrusion. Reproduced from Young, C. A. Pre- and Post Irradiation Evaluation of Fuel Capsule HRB-14; GA-A15969, UC-77; General Atomics Report, 1980.
observed in 8% of the particles. UCO fuel particles that did not display localized chemical attack, had uniform attack along the inner SiC layer (usually on one side of the particles). This uniform attack was attributed to rare earth fission products. Figure 16 displays typical uniform fission product attack in a UCO fuel particle. It should be noted that with optimized UCO stoichiometry, the kernel retains rare earth fission products and does not display kernel migration as found here with non-optimized UCO kernels containing excess UC2 leading to rare earth migration. Metallographic examination of fertile particles indicated that between 0% and 2.4% of the particles in each compact had total coating failure, defined as cracked OPyC and SiC layers. These failures were attributed to pressure vessel failure. Figure 17 displays a typical failed fertile particle. Separate tallies of particles where only the SiC layer had failed were not reported. Other fertile particle observations include the following: 1.5–29.1% of the particles had failed OPyC layers 8–70% of the particles had failed IPyC layers 11–85% of the particles had IPyC layers debonded from the SiC 6–26% of the particles had cracked buffers no kernel migration was observed a few kernels had extruded into buffer cracks.
Figure 15 Photomicrograph of a UCO particle (batch 6157-08-020) from Compact 10 irradiated at 1040 C to 27.8% FIMA and to a fast fluence (E > 0.18 MeV) of 7.1 1025 n m2. Reproduced from Young, C. A. Pre- and Post Irradiation Evaluation of Fuel Capsule HRB-14; GA-A15969, UC-77; General Atomics Report, 1980.
TRISO-Coated Particle Fuel Performance
Table 32
HRB-15B configuration
Number of cells Total number of particle trays Maximum number of loose particles per tray Particle tray outer diameter Particle tray inner diameter Fissile fuel type
235
U enrichment Fissile particle diameter Fertile fuel type Figure 16 Photomicrograph of a UCO particle (batch 6157-08-020) from Compact 10 irradiated at 1040 C to 27.8% FIMA and to a fast fluence (E > 0.18 MeV) of 7.1 1025 n m2 displaying fission product attack of the SiC layer. Reproduced from Young, C. A. Pre- and Post Irradiation Evaluation of Fuel Capsule HRB-14; GA-A15969, UC-77; General Atomics Report, 1980.
Fertile particle diameter Number of fissile particle batches Number of fertile particle batches
The primary objective of the HRB-15B experiment irradiated in HFIR at ORNL21 was to test a variety of LEU fissile fuel designs and ThO2 fertile particle designs. This test involved a single gas-swept cell
22.3–23.6 mm 11.1 mm UCO TRISO and silicon-BISO (Th, U)O2 TRISO and silicon-BISO UC2 TRISO and silicon-BISO UO2 TRISO and silicon-BISO UO2* TRISO and silicon-BISO 19.5% 742–951 mm ThO2 TRISO, BISO and silicon-BISO 773–836 mm 19 22
HRB-15B irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Time average temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
3.07.2.3.6 HRB-15B
1 184 116
Note: Two types of UO2* fuel were tested, one with ZrC dispersed in the buffer and the other with pure ZrC layer around the kernel.
Table 33
Figure 17 Photomicrograph of a ThO2 fertile particle (batch 6252-17-010) irradiated at 1130 C to 8.5% FIMA and to a fast fluence (E > 0.18 MeV) of 8.3 1025 n m2 displaying pressure vessel failure. Reproduced from Young, C. A. Pre- and Post Irradiation Evaluation of Fuel Capsule HRB-14; GA-A15969, UC-77; General Atomics Report, 1980.
173
6 July 1978 4 January 1979 169 26.7 6.0 6.6 815–915 2.9 108 5.1 106
containing 184 thin graphite trays. Each tray could accommodate up to a maximum of 116 individual, unbonded fuel particles. The loose fissile fuel particles included UC2, UCO with four different stoichiometries, (Th, U)O2, UO2, and two types of UO2* (one type had ZrC dispersed throughout the buffer layer and the other had a pure ZrC coating around the kernel). Each fissile fuel type was tested with both TRISO coating and silicon–BISO coating which consisted of the kernel surrounded by a buffer layer, an IPyC layer, and finally a silicon doped OPyC layer. The loose fertile particles tested included TRISO-, BISO-, and silicon–BISO-coated ThO2. Configuration and irradiation data are provided in Tables 32 and 33.
174
TRISO-Coated Particle Fuel Performance
Postirradiation metallography was performed on 20 different particle types, each consisting of approximately 20 particles. These examinations revealed considerable gas bubble formation in UC2 and UCO kernels, and buffer densification in TRISO-coated particles. Some SiC layer cracking was observed in each TRISO-coated fuel type, but mostly in the UCO particles. These cracks were reported to have occurred during mount preparation because of the crack orientation and because the visual examination detected no OPyC cracking. No further tabulation of layer failures was reported. 3.07.2.3.7 R2-K13
The R2-K13 capsule was irradiated in the R2 reactor at Studsvik, Sweden.22 The main objective of this experiment was to test reference UCO fissile particles and ThO2 fertile particles. Four independently gas-swept cells were positioned vertically on top of one another. The middle two cells contained US fuel. The top and bottom cells each contained a full-size German fuel sphere (discussed in the section on German irradiation results). Configuration and irradiation data are given in Tables 34 and 35. Postirradiation metallographic examination was performed on two fuel compacts. All of the 99 fissile particles examined displayed debonding between the buffer and IPyC layers. In some cases, debonding between the buffer, IPyC, and SiC layers was also observed. Likewise, all of the 68 fertile particles examined displayed debonding between the buffer, IPyC, and SiC layers. The SiC layers of all the particles examined were observed to be intact. Table 34
R2-K13 US configuration
Number of cells Total number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact length Total number of piggyback sample sets Fissile fuel type Fertile fuel type 235 U enrichment Fissile particle diameter Fertile particle diameter Fissile particle batches Fertile particle batches Defective SiC layer fraction – fissile particles Defective SiC layer fraction – fertile particles
2 12 12.52 mm 25.4 mm 31 LEU UCO TRISO ThO2 TRISO 19.61% 803 and 824 mm 781–805 mm 2 3 1.9 104 and 4.4 104 <2 106– 1.6 105
3.07.2.3.8 HRB-15A
The main objective of the HRB-15A experiment irradiated in HFIR at ORNL23 was to test several candidate fuel designs for the proposed Large High Temperature Gas Reactor (LHTGR). This test involved a single gas-swept cell containing 20 cylindrical fuel compacts positioned vertically on top of one another. Interspersed between the fuel compacts were 17 tray assemblies. Each assembly had a graphite tray holding 54 unbonded particles in separate holes, and serving as a lid, a graphite wafer containing 54 particles bonded in separate holes with carbonaceous matrix material. Configuration and irradiation data are given in Tables 36 and 37. Table 35
R2-K13 US irradiation data
Start date End date Duration (full power days) Cell Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average center temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Table 36
22 April 1980 19 September 1982 517 2 3 22.5 22.1 4.6 4.5 7.8 7.4 1190 1 105 8 105
HRB-15A configuration
Number of cells Total number of fuel compacts Cylindrical fuel compact diameter Number of short fuel compacts/length Number of long fuel compacts/length Number of bonded wafer/unbonded tray assemblies Fissile fuel type
Fertile fuel type 235
985 2 107 8 106
U enrichment Fissile particle diameter Fertile particle diameter Fissile particle batches Fertile particle batches Defective SiC layer fraction – fissile particles Defective SiC layer fraction – fertile particles
1 20 12.54 mm 3/9.53 mm 17/19.05 mm 17 UCO TRISO UC2 TRISO UC2 ZrC-TRISO UO2 TRISO UO2 ZrC-TRISO UO2* ThO2 TRISO ThO2 silicon-BISO 19.5% 736–894 mm 713–1014 mm 10 5 1.4 105– 7.4 102 6.7 105– 1.4 103
Note: Two types of UO2* fuel were tested, one with ZrC dispersed in the buffer and the other with pure ZrC layer around the kernel.
TRISO-Coated Particle Fuel Performance
Table 37
HRB-15A irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average center temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Table 38
175
HRB-16 configuration
26 July 1980 29 January 1981 174 29.0 6.4 6.5
Number of cells Total number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact length Number of loose particle trays Number of particles per tray
1150 6.96 106 3.76 104
Fissile fuel type
Fertile fuel type 235
U enrichment Fissile particle diameter Fertile particle diameter Fissile particle batches Fertile particle batches Defective SiC layer fraction – fissile particles Defective SiC layer fraction – fertile particles
1 18 12.45 mm 18.70 mm 27 110 (2 particles per hole) UCO TRISO UCO ZrC-TRISO UC2 TRISO UC2 ZrC-TRISO UO2 TRISO UO2* TRISO (Th, U)O2 TRISO ThO2 TRISO ThC2 BISO 19.20–19.61% 742–884 mm 756 and 786 mm 9 2 4.6 107– 4.4 104 1.6 105 and 5.0 104
Note: Two types of UO2* fuel were tested, one with ZrC dispersed in the buffer and the other with pure ZrC layer around the kernel.
Figure 18 Photomicrograph of a UO2 ZrC–TRISO-coated particle (batch 6162-00-010) irradiated at 1075 C to 27.2% FIMA and to a fast fluence of 6.0 1025 n m2 (E > 0.18 MeV) displaying ZrC layer cracks. Reproduced from Ketterer, J.; et al. Capsule HRB-15A Post Irradiation Examination Report; GA-A16758, UC-77; General Atomics Report, 1984.
Postirradiation metallographic examination was performed on five fuel compacts. Between 0% and 5.6% SiC (and OPyC) layer failures were reported for the UO2 particles but were attributed to sample preparation. In contrast, the ZrC layer failures observed in the UO2 ZrC–TRISO-coated particles were also attributed to sample preparation but were not tabulated. A photomicrograph of a UO2 ZrC– TRISO-coated particle displaying a cracked ZrC layer is presented in Figure 18. No SiC layer failures were reported for the UCO fuel. Between 0% and 12.5% of the SiC layers and between 83% and 92% of the IPyC layers were
reported to have failed in the fertile particles. These high layer failures for the fertile ThO2 particles were attributed to the high IPyC BAF values for these particles. The high BAF was a result of intentionally depositing the IPyC layer at low coating rates in an attempt to produce layers that were impermeable to chlorine (chlorine trapped in the particle during SiC deposition may enhance SiC degradation during irradiation). 3.07.2.3.9 HRB-16
The main objective of the HRB-16 experiment conducted in the HFIR at ORNL24 was to test a variety of LEU fissile particle fuel designs. This test involved a single gas-swept cell containing 18 fuel compacts stacked vertically and interspersed with 27 trays of unbonded particles and several encapsulated fission product piggyback transport specimens. Configuration and irradiation data are given in Tables 38 and 39. Postirradiation metallographic examination was performed on seven fuel compacts that contained particles from six different fissile batches and one fertile batch. For fuel compacts containing multiple fissile batches, the following visual criteria were used to identify fuel forms: UO2* had the conspicuous, bright ZrC layer next to the kernel
176
TRISO-Coated Particle Fuel Performance
Table 39
HRB-16 irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average center temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Table 40 21 June 1981 23 December 1981 170 28.7 6.1 6.3 1150 2.44 105 2.08 104
Figure 19 Photomicrograph of a UO2 particle (batch 6152-04-010) irradiated at 1100 C to 26.9% FIMA and to a fast fluence of 5.61 1025 n m2 (E > 0.18 MeV) displaying kernel migration. Reproduced from Ketterer, J. W.; Myers, B. F. Capsule HRB-16 Post Irradiation Examination Report; HTGR-85-053, 1985.
UC2 had very small gas bubbles (voids) in the kernel, or if present in larger form were very irregular in shape UCO had medium size, mostly circular voids in the center of the kernel and small voids at the periphery of the kernel UO2 had large, mostly circular voids evenly distributed throughout the kernel. The metallographic examinations revealed that only the UO2 particles displayed kernel migration. Kernel migration was observed in approximately 28% of the UO2 particles in fuel compacts 2 and 13 and in 60% of the UO2 particles in compact 14. A photomicrograph of a UO2 particle from compact 14 displaying kernel migration is presented in Figure 19. All of the UC2 particles examined (eight total) showed extensive buffer and IPyC layer failure and
HRB-21 configuration
Number of cells Number of fuel compacts Number of encapsulated piggyback specimens Cylindrical fuel compact diameter Cylindrical fuel compact lengths Fissile fuel type Fertile fuel type 235 U enrichment Fissile particle diameter Fertile particle diameter Fissile particle batch Fertile particle batch Total number of fissile particles Total number of fertile particles Defective SiC layer fraction – fissile particles Defective SiC layer fraction – fertile particles
1 24 24 12.27–12.51 mm 49.13–49.35 mm LEU UCO TRISO-P ThO2 TRISO-P 19.66% 904 mm 988 mm 8876-70-0 8876-58-0 42 540 106 240 5.4 106 1.7 105
significant amounts of fission product accumulation. Two of the UC2 particles, or 25% of those examined, had SiC layer failures. These SiC failures occurred next to areas of the IPyC where high concentrations of fission products were present. Examination of the UCO particles revealed significant amounts of fission product attack of the SiC. The extent of this attack ranged from slight to severe. Although not directly measured from examinations of a similar batch of UCO particles irradiated in HRB-15A, it was surmised that this fission product attack was also due to palladium. Of the total 315 fertile ThO2 particles examined, over half displayed IPyC layer failure and nearly 2% displayed SiC layer failure. 3.07.2.3.10 HRB-21
The objective of the HRB-21 capsule irradiated in HFIR at ORNL25 was to demonstrate the irradiation performance of reference NE-MHTGR fuel. A single gas-swept cell contained eight graphite bodies, each of which held three fuel compacts. Each graphite body also contained three sets of encapsulated (piggyback) specimens. These samples were sealed in niobium tubes of up to 52 mm length and 2.2 mm diameter, and each sample contained either absorptivity specimens or loose fuel particles. The test was originally scheduled to be irradiated for six reactor cycles; however, because of difficulty in maintaining control of test temperature, the experiment was terminated after five reactor cycles. Configuration and irradiation data are given in Tables 40 and 41.
TRISO-Coated Particle Fuel Performance
Table 41
HRB-21 irradiation data
Start date End date Duration (full power days) Peak burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average temperature ( C) Peak temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Table 42 20 June 1991 21 November 1991 105 22 3.5 950 1300 1 108 2 104
NPR-1 and NPR-2 configurations
Number of cells Number of fuel compacts Number of encapsulated piggyback specimens Cylindrical fuel compact diameter Cylindrical fuel compact lengths Fuel type 235
Postirradiation metallographic examination of three fuel compacts was performed. SiC layer failure for both fissile and fertile particles ranged between 0% and 5%. During irradiation, the online ionization chambers recorded several spikes that indicated the failure of approximately 130 particles. The metallographic examinations also revealed that the IPyC layer was in contact with the SiC layer. However, in some cases where the IPyC was cracked radially, the IPyC layer was debonded from the SiC. Fission product attack of the SiC layer was also observed. The chemical attack took place at the tips of cracks in the IPyC layer where fission product transport was not likely to be enhanced. However, scanning electron microscopy did not detect localized high concentrations of fission products in the SiC but did detect low levels of palladium extending 5–10 mm uniformly into the SiC. 3.07.2.3.11 NPR-1 and NPR-2
The NPR-1 and NPR-2 capsules were irradiated in HFIR at ORNL26 to demonstrate the irradiation performance of reference NP-MHTGR fuel at the upper bounds of burnup, temperature, and fast fluence. NPR-1 was irradiated one month before and then concurrently with the NPR-2 capsule in HFIR. NPR-1 consisted of a single gas-swept cell containing 16 fuel compacts in addition to 12 sets of loose particles. The loose specimens were sealed in niobium tubes, 29 mm long and 2.2 mm in diameter. NPR-2 consisted of a single gas-swept cell containing 16 fuel compacts, in addition to 16 sets of loose particles. The loose specimens were sealed in niobium tubes, 29 mm long and 2.2 mm in diameter. Configuration and irradiation data for both capsules are given in Tables 42 and 43. Postirradiation metallographic examination of two NPR-1 fuel compacts was performed. The examination indicated that 0.6% of the SiC layers had failed in one compact and that 0% had failed in the other compact. The online gas measurements recorded 526
U enrichment Fuel particle diameter Fuel particle batch Total number of fuel particles Defective SiC layer fraction
Table 43
177
NPR-1
NPR-2
1 16 12
1 16 16
12.43 mm
12.43 mm
49.42 mm
49.42 mm
HEU UCO TRISO-P 93.15% 758 mm FM19-00001 composite 77 500
HEU UCO TRISO-P 93.15% 758 mm FM19-00001 composite 77 500
3 106
3 106
NPR-1 and NPR-2 irradiation data
Start date End date Duration (full power days) Peak burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average temperature ( C) Peak compact temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
NPR-1
NPR-2
25 July 1991 29 May 1992 170
28 August 1991 29 May 1992 172
79 3.75
79 3.75
974
753
1240
1030
1 108 3 104
5 109 6 105
spikes from the ionization chamber. Assuming that each spike corresponds to a particle failure, 0.7% of the total number of particles had failure of all coatings. Postirradiation metallographic examination of one NPR-2 fuel compact was performed. This examination indicated that 3% of the SiC layers had failed. The online gas measurements recorded 135 spikes from the Geiger–Mu¨ller tube. This detector is less sensitive than ionization chambers, and may have missed some transient spikes. However, assuming each spike corresponds to a particle failure, a lower bound of 0.2% can be set for the total number of particles that failed. The metallographic examinations also revealed that the IPyC layer had remained bonded to the SiC except in the vicinity of SiC cracks where
178
TRISO-Coated Particle Fuel Performance
debonding was observed. It was also observed that between 10% and 30% of the particles with failed IPyC layers also displayed cracked SiC layers.
demonstrate the irradiation performance of reference NP-MHTGR fuel at the upper bounds of nominal operating conditions. The same reference fuel was also irradiated in the NPR-1 and NPR-2 tests. For NPR-1A, 20 fuel compacts were placed vertically in a single, gas-swept cell. Originally, the test was scheduled for 104 days of irradiation, but was terminated after 64 days because of indications of a significant number of fuel particle failures. Configuration and irradiation data are given in Tables 44 and 45. Postirradiation metallographic examination of one fuel compact was performed. This examination indicated that 1% of the SiC layers had failed. On the basis of the online gas measurements, it was estimated that approximately 48 particles had failed, which correspond to 0.06% of the total particle population.
3.07.2.3.12 NPR-1A
The NPR-1A capsule was irradiated in the ATR at the INL.27 The primary objective of the test was to
Table 44
NPR-1A configuration
Number of cells Number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact lengths Fuel type 235 U enrichment Fuel particle diameter Fuel particle batch Total number of fuel particles Defective SiC layer fraction
Table 45
1 20 12.37–12.50 mm 49.33 mm HEU UCO TRISO-P 93.15% 758 mm FM19-00001 composite 75 360 3 106
3.07.2.3.13 AGR-1
The AGR-1 experiment involves six separate capsules, each containing approximately 50 000 particles in the form of fuel compacts. It is an instrumented lead experiment, irradiated in an inert sweep-gas atmosphere with individual online temperature monitoring and control of each capsule. A horizontal capsule cross-section at the top of the test train is shown in Figure 20. A vertical section of the capsule was shown previously in Figure 3, and the experiment flow path was shown previously in Figure 5. The sweep gas also has online fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation (see Figure 5). The first of eight planned experiments, AGR-1, had
NPR-1A irradiation data
Start date End date Duration (full power days) Peak burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average temperature ( C) Peak temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
2 October 1991 3 January 1992 64 64 2.1 977 1220 4 109 1.8 105
Graphite Thermocouples
Insulating Stack 1 ATR core center
Stack 2
Stack 3
Hf shroud
Fuel compact
SST shroud Gas lines
Figure 20 Horizontal cross-section of an AGR experimental capsule.
TRISO-Coated Particle Fuel Performance
Table 46
179
Fuel attributes for AGR-1
Property
Kernel diameter (mm) Kernel density (Mg M3) Buffer thickness (mm) IPyC thickness (mm) SiC thickness (mm) OPyC thickness (mm) Buffer density (Mg M3) IPyC density (Mg M3) SiC density (Mg M3) OPyC density (Mg M3) IPyC anisotropya (BAF) OPyC anisotropy (BAF) IPyC anisotropy postcompact anneal (BAF) OPyC anisotropy postcompact anneal (BAF) Sphericity (aspect ratio)
Mean uranium loading (g U per compact) Compact diameter (mm) Compact length (mm) Defective sic coating fraction Defective IPyC coating fraction Defective OPyC coating fractionb
Specified range for mean value
Actual mean value population standard deviation Baseline
Variant 1
Variant 2
Variant 3
350 10 10.4 100 15 40 4 35 3 40 4 0.95 0.15 1.90 0.05 3.19 1.90 0.05 1.035 1.035 Not specified
349.7 9.0 10.924 0.015 103.5 8.2 39.4 2.3 35.3 1.3 41.0 2.1 1.10 0.04 1.904 0.014 3.208 0.003 1.907 0.008 1.022 0.002 1.019 0.003 1.003 0.004
102.5 7.1 40.5 2.4 35.7 1.2 41.1 2.4 1.10 0.04 1.853 0.012 3.206 0.002 1.898 0.009 1.014 0.001 1.013 0.002 1.021 0.002
102.9 7.3 40.1 2.8 35.0 1.0 39.8 2.1 1.10 0.04 1.912 0.015 3.207 0.002 1.901 0.008 1.023 0.002 1.018 0.001 1.036 0.001
104.2 7.8 38.8 2.1 35.9 2.1 39.3 2.1 1.10 0.04 1.904 0.013 3.205 0.001 1.911 0.008 1.029 0.002 1.021 0.003 1.034 0.003
Not specified
1.003 0.003
1.030 0.003
1.029 0.004
1.036 0.002
1% of the particles shall have an aspect ratio 1.14 0.905 0.04
1.054 0.019
1.056 0.019
1.053 0.019
1.055 0.018
0.917
0.915
0.904
0.912
12.22–12.45 25.02–25.4 2.0 104
12.36 0.01 25.066 0.080 4.0 105
12.36 0.01 25.123 0.030 0
12.36 0.01 25.077 0.065 2.0 105
12.36 0.01 25.227 0.037 0
2.0 104
0
0
0
0
1.0 102
0
9.6 105
0
0
a
Specification does not apply to variants 1 and 2. Value is an estimate of an attribute property, not the mean of a variable property.
b
been under irradiation at the INL ATR and was completed in November, 2009; PIE is scheduled to begin in April, 2010.28,29 Table 46 presents pertinent attributes of the fuel that is being irradiated in AGR1.30 Configuration data are presented in Table 47. Irradiation of the experiment began on 24 December 2006, and will continue for approximately 2.5 years to reach a peak burnup of 19% FIMA for the fuel compacts. Table 48 provides a summary of the irradiation status of the AGR-1 experiment. Detailed as-run physics and thermal analyses are performed cycle by cycle to track fuel burnup, fast neutron fluence damage, and fuel temperatures during the irradiation. Peak burnups ranged from 16 to 19% FIMA and fast fluences were between 3.0 and 4.0 1025 n m2 (E > 0.18 MeV). On the basis of the fuel temperature distributions during each cycle, time-averaged peak and
Table 47
AGR-1 configuration data
Number of cells Number of compacts per cell Cylindrical compact diameter Cylindrical compact height Fuel type Particle batch – capsule 3 and 6 Particle batch – capsule 1 and 4 Particle batch – capsule 2 Particle batch – capsule 5 235 U enrichment Number of particles per compact Number of particles per capsule Defective SiC layers
6 12 12.34–12.36 mm 25.0–25.3 mm LEU UCO TRISO Baseline Variant 3 Variant 2 Variant 1 19.6% 4 150 49 800 <4 105
time-averaged volume-averaged temperatures are calculated as the irradiation progresses. After 514 effective full power days, the time-averaged peak fuel temperatures ranged between 1120 and 1180 C
180
TRISO-Coated Particle Fuel Performance
Table 48
AGR-1 irradiation data
Start date End date Duration (full power days) Capsule Time-averaged peak temperature ( C) Time-averaged volume-averaged temperature ( C) Maximum burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) BOL 85mKr R/B EOL 85mKr R/B
24 December 2006 30 October 2009 621 1 2 1136 1118 1029 991 17.5 18.9 3.2 3.8 8 108 1 108 9 108 4 108
and time-averaged volume-averaged temperatures were 100–150 C lower, depending on the capsule. R/B rate ratios have been calculated for many of the short-lived fission gases.31 In all cases, the R/B is less than 107, indicative of release from heavy metal contamination. (A failure of one particle in a capsule would result in an R/B of 3.5 106 based on 4150 particles per capsule and a release of 1.5% from the kernel, which is a typical value at these temperatures and burnups.)32 3.07.2.4
European Experience
The European Commission’s 7th Framework Programme has conducted recent TRISO-coated particle fuel irradiations termed ‘HFR-EU1’ and ‘HFR-EU1bis.’ The experiments share the objective of exploring the potential for high performance and high burnup of the existing German UO2 TRISO-coated particle fuel pebbles for advanced applications, such as the conceptual Generation IV very-high-temperature gas-cooled reactor. As discussed in Section 3.07.2.2, during extensive irradiation tests at and above nominal power-plant conditions in the 1980s and 1990s, not a single coated particle of ‘near-to-production’ fuel elements produced by German researchers with LEU–TRISOcoated particles failed. Irradiating this fuel under defined conditions to extremely high burnups and higher temperature would allow a better understanding of the ultimate irradiation performance of German UO2 TRISO-coated particle fuel. The goal of the HFR-EU1 was to obtain particularly high burnup (20% FIMA) at a peak temperature of 1150 C, typical of pebble bed operation, whereas HFR-EU1bis was dedicated to a particularly high central pebble temperature of up to 1250 C and up to typical pebble bed burnups (10% FIMA).33,34 The irradiated pebbles were 60 mm in diameter with LEU-TRISO-coated UO2. Details of the German
3 1118 980 19.5 4.1 6 109 1 108
4 1180 1041 19.3 4.0 9.0 109 5 108
5 1152 1005 18.3 3.7 1 108 2 107
6 1117 980 16.2 3.0 1 108 1 107
(Arbeitgemeinschaft Versuchsreaktor) and Chinese (Institute of Nuclear and New Energy Technology) fuel attributes are found in Table 49. The design of HFR-EU1 and HFR-EU1bis is on the basis of previous experience of HTR fuel irradiations within the European Union. HFR-EU1 contained five pebbles and six mini samples (ten coated particles each, packed in graphite powder and contained in a niobium tube). Five pebbles in a full-size standard high-temperature gas reactor fuel element rig were used in HFR-EU1bis. Schematic drawings for each arrangement are shown in Figures 21 and 22. Configuration and irradiation data are in Tables 50 and 51. In HFR-EU1, the upper sample holder containing the two Chinese INET fuel pebbles is equipped with 14 thermocouples, while the lower holder containing the German AVR fuel pebbles has 20. In HFR-EU1bis, the central temperature was held constant to control the experiment, whereas control was achieved in HFR-EU1 by holding the surface temperature of the pebble constant. Measured temperatures during the irradiation (without correction for thermal drift and neutron induced decalibration) ranged between 800 and 1000 C for HFR-EU1 and between 900 and 1200 C for HFR-EU1bis, consistent with the peak fuel temperature targets set for the irradiations. In HFR-EU1bis, neutronic calculations indicate that the peak pebble burnup varied between 9% and 11% FIMA and neutron fluence varied between 3.0 and 4.0 1025 n m2 (E > 0.1 MeV) depending on the axial location of the pebble. After 12 cycles in HFR-EU1, neutronic calculations indicate that the peak pebble burnup varied between 9% and 11% FIMA and neutron fluence varied between 2.7 and 3.7 1025 n m2 (E > 0.1 MeV) depending on the axial location of the pebble. Fission product monitoring in HFR-EU1 was accomplished using gas grab samples. At the end of
TRISO-Coated Particle Fuel Performance
Table 49
181
German and Chinese fuel attributes for EU1 and EU1bis
Irradiation
EU1
EU1
EU1bis
Coated particle source Particle batch Kernel composition Kernel diameter (mm) Enrichment (235U wt%) Thickness of coatings (mm) Buffer IPyC SiC OPyC Particle diameter (mm) Pebble Heavy metal loading (g per pebble) 235 U content (g per pebble) Coated particles per pebble Volume packing fraction (%) Free uranium fraction/U-total Matrix grade and density (kg m3)
INET V000802 UO2 490.3 17.08
AVR HT 384–393 UO2 502 16.76
AVR HT 354–383 UO2 501 16.75
97.7 42 37.8 40.8 926.9
92 40 35 40 916
92 38 33 41 909
5.02 0.858 8500 5.0 2.3 107 A3-3 1760 1900
6.0 1.00 9560 6.2 7.8 106 A3-3 1750 1900
6.0 1.00 9560 6.2 7.8 106 A3-3 1750 1900
Temperature of final heat treatment ( C)
-300
-200
-100
Core
-100
-200
the particle releases 1% of the measured shortlived Kr and Xe isotopes (which is not an unreasonable estimate on the basis of other irradiations), then HFR-EU1bis would have contained a few initially defective particles at the beginning of irradiation and additional tens of particle failures at the end. The particle failures may have been related to overheating of the fuel early in the irradiation when an improper gas mixture was inadvertently introduced into the capsule. After 12 irradiation cycles in HFR-EU1, R/B of 4 108 and 1.4 107 were measured for Chinese INET and German AVR fuel, respectively. These low values suggest that in HFR-EU1, no failures have been detected. Instead, the measured fission gas-release probably originates, again, from uranium and thorium impurities in the matrix graphite of the pebbles and in the graphite cups used to hold the pebbles in place. 3.07.2.5
-300 Figure 21 HFR-EU1 capsule schematic.
the HFR-EU1bis irradiation, the R/B was 4 106. In the earlier experiments, HFR-K5 and HFR-K6, R/B values of 5 107 had been measured on fresh fuel. If it is assumed that when particle coatings fail,
Chinese Experience
The Chinese fuel development effort has been performed in part to support the HTR-10 reactor.35,36 HTR-10 is a 10 MW modular high-temperature gascooled test reactor fueled with 60 mm diameter spherical fuel elements, each containing 8300 lowenriched UO2 TRISO-coated fuel particles. Over 20 000 spherical fuel elements have been manufactured for the HTR-10 in 2000 and 2001.
182
TRISO-Coated Particle Fuel Performance
To evaluate the fuel performance of the HTR-10 fuel, two fuel elements (spheres) were chosen at random from each of the first two production batches. These four fuel spheres, together with approximately
Gaslube connectors Thermocouple plug Penetration reactor top lid Dynamic seating penetration plug Second containment gas out Biological shield Refa-170 Thermocouples Second containment (Refa) Purge gas out
+270 +190
1
+120
2
+50
3
Core
0 –20
4
–90
5
–160
6
–230
First containment (capsule)
0.4 0.5 0.6 0.7 0.8 0.9 1.0
7
–310 First containment purge gas in Second containment gas in
Figure 22 HFR-EU1bis capsule schematic. Reproduced from Fu¨tterer, M. A.; et al. Nucl. Eng. Des. 2008, 238(11), 2877–2885.
Table 50
13 500 loose fuel particles, were irradiated in the Russian IVV-2M reactor. Selected physical attributes of the fuel used in this test are listed in Table 52. The irradiation test rig consisted of five capsules. Capsule 1, located at the top, contained the loose particles, while Capsules 2 through 5, in descending order, each contained a test fuel sphere. Irradiation test data are summarized in Table 53. After 3.9% FIMA, where 85mKr R/B values were in the 3 107 range, gas-release suddenly increased in Capsule 4 with R/B values exceeding 103. This result indicated a significant number of fuel particle failures. Further measurements of gas-release from this capsule were terminated at this point. PIE indicated that oxygen impurities in the helium flow gas (such as from air ingress) had reacted with the fuel element matrix material. This reaction turned the matrix into a graphite powder agglomerate and damaged the capsule thermocouple, the internal metal containment, and the fuel particles. No intact fuel particle from this capsule has been found. From beginning to end of irradiation, the remaining three capsules displayed slight variations in gas-release (other than for the in-pile heat-up tests). These variations were attributed to fluctuations in temperature control and implied that no complete (through all layers) particle failures had occurred during irradiation. The high R/B values were due to high U contamination levels and initially defective particles. Quality was improved in subsequent fuel production batches. During irradiation, after 4.1% FIMA, the temperature in Capsule 3 was raised from a nominal 1000– 1200 C for 200 h and then returned to the nominal level. Likewise, after 6.1% FIMA, Capsule 3 temperature was raised from a nominal 1000–1250 C for 200 h and again returned to the nominal level. In both cases, the 85mKr R/B increased by approximately an order of magnitude during the heat-up. When the temperatures were returned to the nominal level, R/B values also returned to those that existed prior to the heat-up. This result indicated that no particles had failed
HFR-EU1 and HFR-EU1bis configuration data
Irradiation
HFR-EU1
Number of cells Number of spheres per cell Spherical fuel element diameter Fuel type Particle batch Number of particles per sphere Number of particles per capsule
2 2 in cell 1 60 mm LEU UO2 LTI – TRISO INET V000802 8500 17 000
HFR-EU1bis
3 in cell 2 60.4 mm LEU UO2 LTI – TRISO AVR HT 384–393 9560 28 680
1 5 60.4 mm LEU UO2 LTI – TRISO AVR HT 354–383 9560 47 800
TRISO-Coated Particle Fuel Performance
Table 51
183
HFR-EU1 and HFR-EU1bis irradiation data
Irradiation
HFR-EU1
HFR-EU1bis
Start date End date Duration (full power days) Capsule Maximum temperature ( C) Minimum temperature ( C) Maximum burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) BOL 85mKr R/B EOL 85mKr R/B
29 September 2006 24 February 2008 332.8a Cell 1 1000 800 9–11 3.7 1.0 108 4.0 108
9 September 2004 18 October 2005 249.9 – 1200 900 9–11 3.0–4.0 1.0 106 4.0 106
Cell 2 1000 800 9–11 3.7 1.0 108 1.4 107
a
After 12 irradiation cycles.
Table 52
Characteristics of Chinese fuel
Batch number Coated particle Kernel form 235 U enrichment (wt%) Kernel diameter (mm) Kernel density (g cm3) Buffer thickness (mm) IPyC thickness (mm) SiC thickness (mm) OPyC thickness (mm) Buffer density (g cm3) IPyC density (g cm3) SiC density (g cm3) OPyC density (g cm3) IPyC BAF OPyC BAF Spherical fuel element Fuel element diameter (mm) Matrix density (g cm3) U per fuel element (g) U contamination U-free/U-total
Table 53
1
2
UO2 17 500.9 10.81 91 48.1 35.0 41.6 1.02 1.84 3.20 1.90 1.03 1.03
UO2 17 497.0 10.90 96 44.7 36.6 42.1 1.05 1.84 3.20 1.84 1.02 1.03
60 1.73 5.00 1.1 106 1.4 104
60 1.72 4.98 7.6 107 2.3 104
HTR-10 test fuel irradiation data
Start date End date Irradiation temperature ( C) Capsule number Fuel batch number/sphere ID Burnup (% FIMA) Fast neutron fluence (1025 n m2)a EOL 85mKr R/Bb Capsule notes
a
during the heat-up tests. Following irradiation, 2014 Capsule 3 particles, approximately one-fourth of the total number in this sphere were examined by an irradiated microsphere gamma analyzer (IMGA), which identified one failed particle (most likely an initially failed particle). Ceramographic examinations of particles from this capsule did not reveal any PyC or SiC layer cracks, fission product attack of the SiC, or kernel migration (amoeba effect). IMGA examination of 2117 Capsule 2 particles, approximately one-fourth of the total, identified two failed particles, which again, were most likely initially failed particles. Ceramography of Capsule 2 particles were similar to Capsule 3 particles where no irradiation induced defects were found. Following the irradiation phase of the experiment, a high-temperature, in-pile heat-up test was conducted with Capsule 5. This heat-up test was performed in-pile because a hot cell furnace was not available. Capsule 5 was moved to the center high flux region of the test reactor while the remaining capsules were transferred to low flux reactor regions where fuel
July 2000 February 2003 1000 50 2 1/SFE5 10.4 1.10 1.7 105
3 2/SFE12 11.4 1.31 6.7 107 In-pile heat-up test to 1200 and 1250 C for 200 h each
4 2/SFE8 10.9 1.30 Not measured Capsule failed after 3.9% FIMA
5 1/SFE7 10.1 1.06 7.8 106 In-pile heat-up test above 1600 C at end of irradiation
Cutoff energy for fast neutron fluence not specified. The tabulated end of life (EOL) R/B for capsule 5 was measured at the end of the irradiation test phase and prior to the start of the high-temperature heat-up test. b
184
TRISO-Coated Particle Fuel Performance
temperatures did not exceed 1000 C. Temperature in Capsule 5 was increased step-wise to the intended 1600 C, held for approximately 4 h, and then reduced step-wise after completion of the test. The total duration of the heat-up test was 50 h. Since the thermocouple in Capsule 5 failed during transfer, temperatures were calculated where further analyses indicated that test temperatures had exceeded the intended 1600 C. Gas release measurements during the heat-up indicated many particle failures. Postirradiation IMGA examination of 2189 particles identified that 127, or 6% of the particles sampled, had failed. Ceramography revealed that most particles had remained intact. However, the failed particles displayed several types of defects including radial and tangential SiC cracks, IPyC cracks, and thorough failures in all layers. None of the observed particles displayed fission product attack of the SiC or kernel migration. 3.07.2.6
Japanese Experience
The Japanese have conducted a number of irradiations of high quality LEU TRISO fuel for the High-Temperature Test Reactor (HTTR), including the Japanese fuel design for first and second core loadings of HTTR and a ‘high burnup’ particle design, developed for later introduction into the HTTR. The gas reactor development program in Japan has been focused on high-temperature process-heat applications from its inception. In addition, the Japanese ‘pin-in-block’ prismatic fuel design has higher calculated maximum fuel temperatures for a given reactor coolant temperature than traditional prismatic gas reactor designs. As a result, the temperatures in the HTTR and projected for commercial gas reactor applications are elevated, with time averaged peak fuel temperatures in the range of 1350 C. Under
Table 54
these conditions, phenomena such as kernel migration, metallic fission product diffusion through the SiC layer, and fission product chemical attack of the SiC layer become limiting, particularly at higher burnup. To address these service conditions with LEU SiC TRISO fuel, the Japanese particle design has been optimized for a relatively low maximum burnup (e.g., larger kernel, thinner buffer), with a maximum burnup limit for service in HTTR of 3.6% FIMA. The test data for the HTTR fuel particle design are provided in Table 54. The burnups in most of the irradiations were very low, between 0.5% and 4% FIMA. The fast fluence values were not available in all cases, but on the basis of 13 and 15 Oarai Gas Loop (OGL-1) irradiations, the fast fluence was probably less than 0.5 1025 n m2 for all the irradiations, approximately an order of magnitude below values for commercial HTGRs. Defect levels in the fuel showed continual improvement with each subsequent irradiation, and the R/B values were very low. Table 55 summarizes the conditions and results of irradiation testing of the Japanese ‘high burnup’ fuel design, plus HTTR fuel irradiated to extended burnup. The most recent irradiation was conducted in the HFIR reactor at ORNL, designated HRB-22. The range of burnup experienced by the fuel was 4.1–6.7% FIMA, with an average of 5.3% FIMA. Four particle failures (out of a total of 32 200 particles in the capsules) were observed during the irradiation phase. This would produce an estimated population failure fraction of 2.8 104 at 95% confidence at a fairly low burnup. In the high burnup fuel irradiation in 91F-1A, conducted in the HTTR, Japanese fuel was taken to burnups of between 7% and 9% FIMA at temperatures between 1120 and 1340 C. In-pile gas-release and postirradiation leaching indicated 4 particle failures, two in the higher temperature upper
Summary of low burnup irradiations of HTTR fuel
Tests
Time (EFPD) Temperature ( C) Burnup (% FIMA) Fast fluence (1025 n m2) Exposed kernels (BOL) Exposed kernels (EOL) SiC defects (BOL) SiC defects (EOL) R/B at EOL
OGL-1 6th
7th
8th
10th
12th
13th
15th
22 980–1488 0.1–0.4
58 984–1376 0.4–1.2
54 940–1370 0.3–1.0
130 879–1331 0.8–2.8
195 1300 3.0–3.9
5 105 1 104 2 103 2 103 1.5 106
3 105 1 104 4 103 4 103 4.0 107
6 106 1 104 5 104 2 103 1.1 107
3 104 5 104 7 104 1 103 3.2 106
1 104 3 104
244 1083–1342 1.3–3.7 0.11–0.49 4 106 7 105 4 106
218 1180–1344 1.1–3.1 0.1–0.44 3 106 3 107 5 105
6.8 108
3.8 107
3.1 106
TRISO-Coated Particle Fuel Performance
Table 55
185
Japanese particle attributes and irradiation conditions
Particle attributes for each irradiation
HRB-22
91F–1A
94F–9A
14th OGL-1
235 U enrichment (%) Kernel diameter (mm) Buffer thickness (mm) IPyC thickness (mm) SiC thickness (mm) OPyC thickness (mm) Irradiation conditions Effective full power days Number of compacts per capsule Number of particles per capsule Number of cells Maximum temperature ( C) Average temperature ( C) Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.1 MeV) BOL 85mKr R/B EOL 85mKr R/B
4.07 544 97 33 34 39
7.98 551 97 32 34 39
7.82 615 59 30 29 47
19.54 553 84 32 37 42
89
360 2 10 100 Upper 1120
Lower 1340
360 2 25 000 Upper 1120
Lower 1190
9.6 6.0
7.9 4.6
7.0 3.1
5.5 2.3
65 18 full and one half-size 245 000 1 1160–1339 1122–1305 0.73–1.17 0.07–0.12
32 200 1 898–1156 821–1071 4.1–6.7 1.3–2.8 1.0 106 3.0 106
Source: Minato, K.; et al. HRB-22 Capsule Irradiation Test for HTGR Fuel; JAERI-research 98-021; 2008; Sawa, K.; et al. JAERI-research 2001-043; Irradiation Test of High Burnup Coated Fuel Particles for High Temperature Gas-cooled Reactor (91F-1A Sweep Gas Capsule Irradiation Test); 2001; Sawa, K.; et al. JAERI-research 2002-012; Acceleration Irradiation Test of First-loading Fuel of HTTR up to High Burnup; 2002; Hayshi, K.; et al. JAERI-research 2000-001; Irradiation Experiments of the 13th–15th OGL-1 Fuel Assemblies; 2000.
capsule, and two in the lower capsule. Pressure vessel failure was the stated cause of the failures. The 95% confidence value for particle failures during the irradiation was 4.6 104. In irradiation 94F-9A conducted in the Japan Material Test Reactor, TRISO fuel was irradiated at 1200 C to burnups between 5.5% and 7% FIMA. No failures occurred in one capsule, and in the other capsule, only one particle out of 50 000 failed. Another irradiation of ‘high burnup’ fuel was conducted in the OGL-1 loop to a burnup of 1% FIMA with very low fast neutron fluence with a failure fraction of 1 104 observed during a temperature transient test to 1500 C. 3.07.2.7
Irradiation Performance Summary
Extensive TRISO-coated particle irradiations have been performed around the world over the past 40 years. The majority of the particles have been irradiated in the United States and Europe/Germany, with smaller amounts of fuel irradiated by Chinese and Japanese programs. A summary of salient features of all the irradiations in which R/B measurements were available is found in Table 56. The irradiation experiments reviewed here, including many others not summarized above, contained particle fuels with widely different coatings and kernels and were conducted at a variety of burnups, temperatures, and fluences.
Figure 23 plots EOL online gas-release (R/B) as a function of temperature, burnup, and fluence separately, and no trend is evident. Releases observed in the experiments are a function of (a) the initial level of heavy metal contamination in the fuel, and (b) in-service failures. 3.07.2.7.1 Heavy metal contamination
BOL 85mKr R/B is often used as a metric of as-manufactured fuel quality. Fuel developed by German researchers exhibited overall low initial defect fractions, with measured R/B values ranging between 1010 and 106. These results also demonstrated that fuel quality improved over time, as was the case for fuel used in the AVR. It should be noted that the lowest ever 85mKr R/B measured in an in-reactor irradiation is from German fuel (1010). Both the Chinese and Japanese fuel exhibited slightly higher levels of contamination, as indicated by the initial R/B ratios in their irradiations. By contrast, the high gas release from the historic US database (all experiments except AGR-1) is a combination of high initial levels of heavy metal contamination in the fuel and in-service failures. During the 1970s and 1980s, the lower fuel quality of the historic US fuel resulted in R/B ratios that varied between approximately 2 108 and 104, indicating that the initial contamination levels in US fuel displayed great variability and could be rather high. The level did not
TRISO-Coated Particle Fuel Performance
1.E-02
1.E-02
1.E-03
1.E-03
1.E-03
1.E-04
1.E-04
1.E-04
1.E-05 1.E-06
EOL 85mKr R/B
1.E-02
EOL 85mKr R/B
EOL 85mKr R/B
186
1.E-05 1.E-06
1.E-05 1.E-06
1.E-07
1.E-07
1.E-07
1.E-08
1.E-08
1.E-08
1.E-09 800
1.E-09 1000 1200 1400 Temperature ( ˚C)
1
10 100 Burnup (%FIMA)
1.E-09 0 2 4 6 8 10 12 14 Fast fluence (10E25 n m-2)
US historic
German
US historic
German
US historic
German
US new (AGR)
EU new
US new (AGR)
EU new
US new (AGR)
EU new
China
Japan
China
Japan
China
Japan
Figure 23 TRISO-coated fuel performance as a function of temperature, burnup, and fast fluence.
significantly change until the early 1990s, when serious efforts at reduction of initial contamination were undertaken in the fabrication campaigns. With changes to fabrication processes, the most recent US irradiation, AGR-1, is demonstrating very low levels of contamination with initial R/B ratios in the range of 108.31 3.07.2.7.2 In-service failures
The behavior of German fuel under irradiation is excellent. Measured EOL 85mKr R/B values ranged between 3 109 to 105 for the various German irradiation experiments. In most cases, the maximum R/B was measured at EOL; however, in some irradiations, the final portion of the experiment was conducted at lower temperatures, which caused the R/B to decrease. Nevertheless, these results indicate that no fuel particle had failed while under irradiation. PIE results confirmed the more extensive and more reliable gas-release data. Out of approximately 380 000 UO2 and 80 000 (Th, U)O2 particles tested, there were no in-pile failures and only a few ‘damaged’ particles due to experimental anomalies. Gas release was attributed to as-manufactured defects and heavy metal contamination only. The historic US fuel irradiations indicate that inreactor fuel failures occurred at the percent level. PIEs of the historic US fuel revealed in most experiments up to percent level particle failures, with very high levels of individual layer failures. These failures
resulted in the observed high gas-release (R/B), which is approximately a factor of 1000 times greater than that of the German fuel under a broad range of conditions (i.e., temperature, burnup, and fluence).37 The current US fuel in AGR-1 is the culmination of a significant effort to overcome the limitations observed in past irradiations. Fabrication processes were changed, and irradiations were not accelerated. The result is a fuel that has survived irradiation to high burnup with no failures. The smaller statistical databases associated with the fuel from China, Japan, and Europe are more mixed, with some irradiations indicating good performance, and some indicating fuel failures. In many cases, the root cause of the failures is unclear because of less extensive PIE. 3.07.2.7.3 Failure mechanisms
Review of the US irradiation database indicates that a number of different failure mechanisms in the individual layers of the TRISO coating contributed to the less-than-satisfactory US fuel performance. Failures of the coating layers were attributed to (a) pressure vessel failure, (b) kernel migration (amoeba effect), (c) fission product attack of the SiC layer, (d) irradiation-induced IPyC cracking and/or debonding leading to cracking in the SiC layer, and (e) matrix–OPyC interaction and irradiationinduced OPyC failure.
TRISO-Coated Particle Fuel Performance
Fission product and impurity attack of the SiC and kernel migration are thermally driven phenomena strongly influenced by burnup, temperature, and the temperature gradient across the particle. The temperature gradient is a strong function of the power density in the fuel body, which is directly related to the level of acceleration in the experiment. The PyC-related mechanisms are strongly related to the anisotropy and porosity in the coatings. The anisotropy has a strong influence on the shrinkage and swelling behavior of the PyC layers under irradiation. However, the anisotropy measurements, especially optical methods (OPTAF), are not reliable predictors of PyC failure under irradiation as indicated by the lack of correlation between the measured BAF and PyC failure (e.g., OF-2, HRB-5, HRB-6), and the high level of PyC failure observed in most historic US irradiations. The fabrication processes in the United States have been changed, with the introduction of more accurate methods to measure anisotropy to improve the survivability of the PyC layers. The lack of fuel failure in the US Table 56 Test/cell
Germany R2-K12/1 R2-K12/2 R2-K13/1 R2-K13/4 BR2-P25 HFR-P4/1 HFR-P4/2 HFR-P4/3 SL-P1 HFR-K3/1 HFR-K3/2 HFR-K3/3 HFR-K3/4 FRJ2-K13/1 FRJ2-K13/2 FRJ2-K13/3 FRJ2-K13/4 FRJ2-K15/1 FRJ2-K15/2 FRJ2-K15/3 FRJ2-P27/1 FRJ2-P27/1 FRJ2-P27/1 HFR-K5/1 HFR-K5/2
187
AGR-1 fuel taken to 19% FIMA demonstrates that the changes made to the process were correct and resulted in excellent in-pile fuel performance. 3.07.2.7.4 Acceleration effects
The rate of accumulation of burnup and fast fluence relative to that expected in the reactor (i.e., the degree of acceleration) is an important factor in the behavior observed in these experiments. For most of these fuels, the time to reach goal burnup and fast fluence is 1095 days (3 years), whereas the irradiation times varied greatly. For historic US irradiations, acceleration varied between factors of 2–10. One Japanese irradiation (HRB-22) was also greatly accelerated. By contrast, the German, Chinese, and European irradiations, as well as the US AGR-1 irradiation, were all generally accelerated by only factors of two to three. These low acceleration factors limited thermal gradients across the fuel, which could have jeopardized fuel integrity with enhanced fission product attack and kernel migration (Table 56).
Summary of TRISO-coated particle fuel irradiation experiments Fuel forms
Irradiation time (d)/ acceleration level
Peak temperature ( C)
Peak fissile and fertile burnup (% FIMA)
Peak fluence (1025 n m2)
HEU (Th, U)O2 TRISO
308/3
HEU (Th, U)O2 TRISO
517/2
HEU (Th, U)O2 TRISO LEU UO2 TRISO
350/3 351/3
LEU UO2 TRISO LEU UO2 TRISO
330/3 359/3
LEU UO2 TRISO
396/2.75
LEU UO2 TRISO
533/2
LEU UO2 TRISO
232/4.7
LEU UO2 TRISO
563/2
1100 1280 1170 980 1070 940 945 1075 794 1200 920 920 1220 1125 1150 1150 1120 970 1150 990 1080 1320 1130 Cycled proof test
11.1 12.4 10.2 9.8 15.6 14.7 14.9 14.0 11.3 7.5 10.0 10.6 9.0 7.5 8.0 7.9 7.6 13.2 14.6 13.9 7.6 8.0 7.6 6.7 8.8
5.6 6.9 8.5 6.8 8.1 8.0 8.0 8.0 6.8 4.0 5.8 5.9 4.9 0.2 0.2 0.2 0.2 0.2 0.2 0.1 1.4 1.7 1.3 2.9 <4.3
EOL 85m Kr R/B (106)
0.300 0.200 0.070 0.050 1.000 0.080 0.080 0.008 1.200 0.200 0.100 0.100 0.300 0.020 0.020 0.007 0.007 0.010 0.005 0.003 1.600 10.000 0.120 <0.3 <0.3 Continued
188
TRISO-Coated Particle Fuel Performance
Table 56
Continued
Test/cell
HFR-K5/3 HFR-K5/4 HFR-K6/1 HFR-K6/2 HFR-K6/3 HFR-K6/4 United States F-30/1 F-30/2 F-30/3 F-30/4 F-30/5 HRB-4 HRB-5 HRB-6 OF-2/1 OF-2/2 HRB-14 HRB-15B
R2-K13/2 R2-K13/3 HRB-15A
HRB-16
HRB-21 NPR-1 NPR-2 NPR-1A AGR-1/1 AGR-1/2 AGR-1/3 AGR-1/4 AGR-1/5 AGR-1/6 Europe EU1 China EU1 German EU1bis China Test fuel cell 2 Test fuel cell 3 Test fuel cell 5 Japan HRB-22
Fuel forms
Irradiation time (d)/ acceleration level
Peak temperature ( C)
Peak fissile and fertile burnup (% FIMA) 9.1 8.7 7.2 9.3 9.7 9.2
Peak fluence (1025 n m2)
EOL 85m
Kr R/B (106) <0.3 <0.3 <0.3 <0.3 <0.3 <0.3
4.3 <4.3 3.2 <4.8 4.8 <4.8
LEU UO2 TRISO
634/1.7
Cycled proof test
HEU (Th, U)C2 TRISO and ThC2 TRISO
269/4
LEU WAR UC2 TRISO and ThO2 BISO LEU WAR UC2 TRISO and ThO2 BISO HEU (Th, U)C2 TRISO and ThO2 BISO WAR UCO UC2 (Th, U)O2 TRISO and ThO2 BISO LEU UCO UO2 (Th, U)O2 TRISO and ThO2 BISO LEU UCO UC2 (Th, U)O2 UO2 TRISO and Si BISO and ThO2 TRISO, BISO, and Si-BISO LEU UCO TRISO and ThO2 TRISO LEU UCO UC2 UO2 TRISO and ZrC-TRISO and ThO2 TRISO and Si-BISO LEU UCO UC2 UO2 (Th, U) O2 TRISO and ZrC-TRISO and ThC2 ThO2 TRISO and BISO LEU UCO and ThO2 TRISO-P HEU UCO TRISO-P HEU UCO TRISO-P HEU UCO TRISO-P LEU UCO TRISO
244/4.5
1100 1100 1120 1100 1200 1250
15.0/3.0 19.0/4.5 20.0/5.0 18.0/4.0 12.0/1.5 27.7/13.4
8.0 10.5 11.5 9.5 12.0 10.5
8 100 10 20 20 320
107/10
1250
15.7/4.3
4.7
100
183/6
1100
26.6/9.3
7.9
270
352/3
1350 1350 1190
79.6/4.3 79.5/4.3 28.6/8.5
8.9 8.4 8.3
100 5 300
915
26.7/6.0
6.6
5
174/6.3
1190 985 1150
22.5/4.6 22.1/4.5 29.0/6.4
7.8 7.4 6.5
80 8 380
170/6.3
1150
28.7/6.1
6.3
210
105/10 170/6.3 172/6.3 64/6.3 625/1.75
1300 1240 1030 1220 1136 1118 1118 1180 1152 1117
22.0/2.2 79.0 79.0 64.0 17.5 18.9 19.5 19.3 18.3 16.2
3.5 3.8 3.8 2.1 3.2 3.8 4.1 4 3.7 3
200 300 60 18 0.09 0.04 0.01 0.05 0.2 0.1
LEU UO2 TRISO LEU UO2 TRISO LEU UO2 TRISO
333/3.3 333/3.3 250/4.4
1000 1000 1000
11 11 11
4 4 3.7
0.04 0.14 4
LEU UO2 TRISO
900/1.2 900/1.2 900/1.2
1000 1000 1000
10.4 11.4 10.1
1.1 1.31 1.06
17 0.67 7.8
LEU UO2 TRISO
89/12
1156
6.7
2.8
214/5 169/6.5
517/2
Note: US fluence is for E > 0.18 MeV, fluence is not specified by China, and fluence values of all others are for E > 0.10 MeV.
3
TRISO-Coated Particle Fuel Performance
3.07.3 Safety Testing
and fuel fabrication processes (e.g., SiC coating rate) on fission product release are investigated.
The release of fission products from TRISO-coated irradiated fuels heated at elevated temperatures to simulate accident conditions is reviewed in this section. For a small HTGR, the temperature evolution during a loss of coolant accident with complete depressurization is calculated to reach 1600 C (including a 100 C uncertainty margin) for approximately 30 h, as shown in Figure 24. The preponderance of the experimental data is from tests with fuels of German manufacture, but insights are also provided from a few experiments using Japanese and US fuels. There is some evidence of particle failure caused by internal pressure (pressure vessel failure), but of most interest is degradation of the SiC layer in the TRISO coating during accident performance testing at elevated temperatures, 1500–1800 C. Because releases of cesium are sensitive to the integrity of the SiC layer, much attention is paid to this fission product. Releases of krypton are sensitive to pressure vessel failure, but otherwise trail releases of cesium due to holdup by PyC layers. Releases of other fission products, such as strontium, europium, and cerium are treated where data are available. Silver, although not a safety concern because of its low yield, has potential consequences for reactor maintenance and diffuses readily through the SiC layer, even at 1200 C. The influences of irradiation characteristics (e.g., fuel burnup, neutron fluence, and irradiation temperature), SiC microstructure (e.g., grain size and orientation),
3.07.3.1
Facility Overview
Several operational facilities currently exist for testing high-temperature fission product retention performance of irradiated TRISO fuel. The three primary facilities discussed here (Ku¨FA, Fuel Accident Condition Simulator (FACS), and Core Conduction Cooldown Test Facility (CCCTF)) are all on the basis of the basic concept of a refractory metal (tantalum) furnace operating at maximum temperatures of 1800–2000 C with a helium atmosphere and a water-cooled cold finger intruding into the hot zone to collect condensable fission products on an exchangeable plate connected to the end of the cold finger. 3.07.3.1.1 Ku¨ FA at ITU
Initial development on Ku¨FA (from the German Ku¨hlfingerapparatur, ‘cold finger device’) began in the early 1980s at Forschungszentrum Ju¨lich.38 The system was designed to simulate reactor accident conditions by heating spherical fuel elements in helium to temperatures of up to 1800 C while measuring the release of condensable fission products, as well as noble fission gases. The system was used in the 1980s and 1990s to test high-temperature fission product retention on various irradiated particle fuel specimens.1,38,39 More recently, the system was
2000 Accident simulation tests
1800
Temperature (⬚C)
1600 1400 1200
MODUL, max. fuel element temperature in hypothetical and design basis accident *including 100 ºC margin
1000 800 Reactor operation Temperature
600 400
0
189
100
200
300
400
500
Time (h) Figure 24 Temperature evolution during a depressurized reduction cooling of a small HTR, and in heating tests with irradiated fuel elements. Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
190
TRISO-Coated Particle Fuel Performance
transferred from Ju¨lich, upgraded, and installed in the hot cells at the Institute for Transuranium Elements (ITU) in Karlsruhe for continued accident testing.40 The system consists of a double-walled, watercooled metal jacketed furnace with tantalum electricalresistance heating elements (maximum power of 70 kW at 2400 A), tantalum hot zone components, and a water-cooled cold finger that holds an exchangeable metal plate to condense fission products during operation (see Figures 25 and 26). The cooling system maintains the plate at a temperature of less than 100 C during furnace operation. The cold finger can be removed from the furnace through a sliding valve and gas lock system that maintains the purity of the gas in the furnace during exchange operations. Plates can then be packaged and removed from the facility for gamma spectroscopic counting or for acid leaching and radiochemical analysis. The entire system is operated inside an alpha containment box which itself resides inside a shielded hot cell. Fission gases in the sweep gas circuit are captured in two charcoal cryo-traps, with the second trap serving to indicate the trapping
Figure 25 The cold finger device Ku¨FA installed in the hot cells at ITU. © European Atomic Energy Community, 2011. See http://itu.jrc.ec.europa.eu.
efficiency of the first. The activity in the traps is monitored continuously with NaI detectors during a heating test. Upgrades to this system since the original work at Ju¨lich include an increased maximum temperature capability to 2000 C, updated temperature regulation system, and a redesigned cold finger to allow sufficiently low plate temperatures during operation of the furnace at up to 1800 C. The upgraded Ku¨FA system has recently been described in further detail elsewhere.40 3.07.3.1.2 INL’s FACS
The FACS furnace at the INL is designed to heat fuel specimens at temperatures up to 2000 C in inert atmospheres while monitoring the release of volatile fission metals and fission gases (see Figure 27). The furnace has tantalum hot zone components and is heated using a graphite resistance element. The fuel specimen is heated inside an 85 mm diameter tantalum flow tube and is supported in a tantalum holder that sits atop a tantalum tube inserted from the bottom of the furnace. The fuel specimen temperature is monitored with a Type C (W–Re) thermocouple housed within the tantalum tube that supports the sample holder, with the junction located 12.5 mm below the specimen. The hot zone can accommodate samples up to 75 mm in diameter. Special sample holders have been designed to accommodate the compacts or spherical pebbles, although any specimen geometry can be accommodated through design of a custom tantalum holder. The furnace main chamber is double-walled, water-cooled stainless steel. Both the top and bottom flanges are water-cooled stainless steel and can be removed from the main chamber for sample loading or maintenance operations as needed. The watercooled cold finger is of aluminum construction and holds a 35 mm diameter plate on the end by means of a mechanical lock using a rotary actuator. The temperature at the base of the cold finger, where the condensation plate is attached, is monitored using a Type J thermocouple. Figure 28 depicts a cutaway view of the furnace system. Condensate plates are introduced to the furnace through a separate transfer chamber, purged with helium prior to opening a gate valve to the main chamber. The entire condensate plate exchange operation is performed automatically by computer control. The operator must only remove the used plate once it has been retracted into the transfer chamber and replace it with a fresh plate to be placed on the cold finger during the next exchange cycle.
TRISO-Coated Particle Fuel Performance
191
Coldfinger jacket (vacuum/He filled)
Valve
Water cooled
Removable condensation plate Gas stream cylinder in Ta Heat element in Ta Steel heat sink Spherical fuel element W/RE–thermocouple
Vacuum pump connection Helium output Power connection
Helium input
Figure 26 Ku¨FA system at ITU, Karlsruhe. Reproduced from Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, USA, October 1–4, 2008.
The helium sweep gas is introduced to the furnace from the bottom and injected into the hot zone beneath the specimen and flows toward the condensation plate at the end of the cold finger as shown in Figure 28. The helium is then swept from the furnace through particulate and moisture filters and carried to the fission gas monitoring system for analysis of 85Kr, 133Xe, or other radioactive noble gases.
This system, located outside of the hot cell, consists of dual liquid nitrogen-cooled charcoal traps to contain the noble gases, each with a high purity germanium detector and adjustable collimator. A schematic diagram of the cooling water and He gas circuits is shown in Figure 29. Condensation plates that are removed from the furnace during operation will be packaged in clean
192
TRISO-Coated Particle Fuel Performance
Figure 27 FACS furnace.
plastic containers and sent to the analytical chemistry laboratory for analysis. This will consist of gamma counting of the plates to measure the inventory of major fission products (e.g. 137Cs, 110mAg), followed by acid leaching to remove fission products and for aqueous radiochemical analysis of 131I and betaemitter 90Sr. The system is currently undergoing remote qualification testing and preliminary condensate efficiency tests, after which the system will be installed in the INL Hot Fuel Examination Facility (HFEF) hot cells. The system will be used to conduct hightemperature accident testing on irradiated AGR TRISO fuel compacts from the AGR-1 experiment for the NGNP program. 3.07.3.1.3 ORNL’s Core Conduction Cooldown Test Facility
The CCCTF was developed at ORNL in the early 1990s to perform high-temperature accident simulation performance tests on coated particle fuels.41 The system is operated in a modular hot cell at the
Irradiated Fuel Examination Laboratory (IFEL) at ORNL. The system has undergone several modifications and now consists of a water-cooled graphite resistance-heated tube furnace in a vertical orientation with an internal tantalum container that isolates the test specimens from the furnace heating elements. A helium sweep gas system provides the test atmosphere and transports released fission gases to a cryogenic trap system for collection and gamma counting. Water cooling circuits provide temperature control for sensitive furnace components (including the metal furnace jacket and the cold finger). The furnace system diagram is shown in Figure 30 and a photograph of the furnace is shown in Figure 31. The fuel specimen is loaded from the top and during operation; a water-cooled cold finger is inserted and removed from the top through an air lock. Fuel specimens are held in a graphite or refractory metal holder supported by a tantalum container with a small well for a Type C thermocouple, which serves to both measure the sample temperature and control the furnace control system. The temperature of the tantalum container can be monitored from the outside also through a furnace window using an optical pyrometer. A steel deposition cup is attached to the bottom of the cold finger by means of a screw mechanism and acts as the collection surface for condensable fission products. By periodically inserting and removing the cold finger, the deposition cup can be changed and a history of the metallic releases from the fuel can be obtained; gamma counting or radiochemical analysis allows the identification of the collected elements and their quantities. Fission gases released from the heated fuel are continually swept from the furnace by the helium gas stream and collected in a bank of dual liquid-nitrogen-cooled charcoal traps monitored with NaI detectors. The fission gas traps and detectors are located outside of the hot cell in a low-background area to facilitate detection of very small quantities. The CCCTF has been used for several heating tests of irradiated fuel compacts and unbonded fuel particles.42,43 The system is currently undergoing a series of upgrades to improve performance and process control, including the installation of a redesigned cold-finger airlock, an upgraded fission sweep gas system with an automated gas supply system, a modified liquid nitrogen supply system, and upgraded gamma counting electronics. This upgraded system will be used for accident testing of AGR fuel compacts for the NGNP program, starting with fuel from the AGR-1 irradiation test.
TRISO-Coated Particle Fuel Performance
Cold finger
193
Gate valve
Transfer chamber Condensation plate cradle Heat shields Graphite heating element
Sample holder Tantalum flow tube
Sample support and helium inlet
Helium outlet
Figure 28 Three-dimensional cut away diagram of the FACS main chamber and interior components.
3.07.3.1.4 KORA
The corrosion furnace (KORA) test equipment was developed at Forschungszentrum Ju¨lich in the early 1990s to examine fuel performance under water and air ingress accident conditions.3 The system is depicted in Figure 32. The system was capable of testing fuel specimens up to the size of a spherical fuel element in helium mixed with air or water vapor (more than 80 kPa water vapor partial pressure possible) at temperatures of up to 1600 C while measuring the release of 85Kr. The system was used to test several AVR fuel elements and particles as well as compacts and particles from the HFR-B1 irradiation experiment.3 Work is currently underway to develop a KORA apparatus at ITU in Karlsruhe.44 3.07.3.2
German Experience
Safety tests carried out by German researchers were performed primarily on whole spherical fuel elements containing 16 400 fuel particles, and in some cases on fuel compacts containing 1600 fuel particles. Irradiation data from these tests3,42,45,46 for spherical fuel elements (60 mm in diameter) are presented in
Table 57, and for compacts (cylinders manufactured from spherical elements with a fuel zone 20 mm in diameter) in Table 58. Both tables also contain the maximum integral fractional releases of krypton and cesium measured during the heating tests. The spherical fuel elements listed in Table 57 experienced burnup within the historic pebble bed burnup envelope (10% FIMA) with the exception of HFR-K3/3 (10.6% FIMA). These fuel elements experienced fast fluence within or slightly greater than the historic pebble bed fluence envelope 2.5 1025 n m2, with the exception of HFR-K3/3 (6.0 1025 n m2). By contrast, all of the compacts were irradiated beyond the historic pebble bed reactor envelope. Krypton integral releases as a function of time at various temperatures are shown in Figure 33(a)– 33(c). In Figure 33(a), krypton releases are below the level of one particle failure (1/16 400 ¼ 6.1 105) at 1600 C, whereas that level of release is exceeded at 1700 and 1800 C for spherical fuel elements. The occasional vertical lines in the releases at temperatures above 1600 C are associated with pressure vessel failure of particles. Pressure vessel failure is a function of burnup (fission gas inventory and, in
194 TRISO-Coated Particle Fuel Performance
Hot cell boundary
He to transfer chamber He to condensation plate release
Cold finger
Top cover Cooling water system Manifold
Supply Building Heat cooling exchanger water
Return
Cold finger Top cover Chamber Bottom cover Electrodes Exhaust
Flow meters
Particulate filter
Manifold
FM FM FM FM FM FM
He to sample flow tube
Vacuum pump
Bottom cover Chamber Electrodes Exhaust
He to main chamber Figure 29 Schematic of FACS cooling water and helium supply system.
Fission gas monitoring system
TRISO-Coated Particle Fuel Performance
195
Cold finger
Spool piece
Gate valve assembly
Thermal shield Sweep gas out
Deposition cup Tantalum can
Upper cooling flange Sweep gas in
Furnace insulation
Graphite heating element
Watercooled furnace jacket
Specimen assembly (graphite) Graphite holder
Dual type 'C' thermocouple
Graphite composite reflector
Lower cooling flange
Figure 30 Cross-section schematic of the CCCTF furnace.
UO2 fuel, CO inventory), fuel irradiation temperature (fission gas pressure and oxygen to fission ratio in UO2 fuel),47 and fuel particle design and properties (buffer void volume and SiC strength). Figure 34(b) shows the larger releases associated with higher burnups. 85Kr release can be used as an indicator of 131 I release, on the basis of reactivation of irradiated
Figure 31 Core Conduction Cooldown Test Facility.
fuels immediately before heating tests.48 Figure 35(c) shows that krypton release is negligible (106) for compacts with 10–12% FIMA at 1600 C, but becomes significant (104) at approximately 200 h at 1700 C, and at approximately 150 h at 1800 C. While krypton releases at levels above 6.1 105 are due to the failure of all three coating layers (IPyC, SiC, and OPyC), resulting in an exposed kernel, cesium releases above 6.1 105 measured in safety testing at 1600 C are an indication of defective SiC at the end of irradiation.46 Cesium and krypton releases presented in Figure 34 show cesium release at an inventory level of two particles (1.2 104) for fuel sphere AVR 88/33 and one particle (6.1 105) for fuel sphere 82/20, whereas the krypton releases are well below the inventory of a single particle. These results indicate that fuel sphere AVR 88/33 contains two fuel particles with defective SiC and fuel sphere 82/20 contains one fuel particle
TRISO-Coated Particle Fuel Performance
Ionization chamber H O 2 P2Os
H2O
CO
Blue gel
Tritium Hot cell
Soda CuO lime 600 ⬚C Safety dust filters
Fission gas tap (charcoal)
Liquid N2
Filter for solid fission product
Naldetector
Gas tight box
196
Compressor Blue-gel P2Os Moisture traps
Electronics
Chart recorder
H2O bath 1 Gas
Furnace
2
Heated pipe
Figure 32 Schematic diagram of the KORA facility developed at Forschungzentrum Juelich in the 1990s. Reproduced from IAEA. Fuel Performance and Fission product Behavior in Gas-cooled Reactors. IAEA TECDOC-978; IAEA: Vienna, Austria, November 1997. Table 57 Fuel element
Results of accident simulation tests with irradiated spherical fuel elements Burnup (% FIMA)
Fast fluence (1025 n m2)
Heating test Temperature ( C)
Duration (h)
1600 1600 1600 1600 1600 1600 1600 1700 1600 1800 1800 1800 1800 1800
500 500 160 50 100 50 500 185 138 100 100 200 90 175
AVR 71/22 HFR-K3/1 FRJ-K13/2 AVR 88/33 AVR 82/20 AVR 88/15 AVR 82/9 AVR 74/11 FRJ-K13/4
3.5 7.5 8.0 8.5 8.6 8.7 8.9 6.2 7.6
0.9 4.0 0.1 2.3 2.4 2.4 2.5 1.6 0.1
HFR-K3/3 AVR 76/18 AVR 74/10 AVR 70/33
10.6 7.1 5.5 1.6
6.0 1.9 0.9 0.4
Table 58 Fuel compact
HFR-P4/3/7 HFR-P4/1/8 HFR-P4/2/8 HFR-P4/1/12 SL-P1/6 SL-P1/10 SL-P1/9 HFR-P4/3/12
Fractional release
85
137
4 107 2 106 6 107 1 107 2 107 6 108 5 107 3 105 3 107 7 105 7 104 1 104 2 103 2 103
2 105 1 104 4 105 1.2 104 6 105 1 105 8 104 8 105 3 106 1 102 6 102 5 102 1 101 2 102
Kr
Cs
Results of accident simulation tests at 1600–1800 C with irradiated fuel compacts Irradiation conditions
Heating test
Fractional release
Burnup (% FIMA)
Fast fluence (1025 n m2)
Temperature ( C)
Temperature ( C)
Duration (h)
13.9 13.8 13.8 11.1 10.7 10.3 10.7 12.0
7.5 7.2 7.2 5.5 6.7 6.0 6.3 5.5
1075 940 945 940 800 800 800 1075
1600 1600 1600 1600 1600 1700 1700 1800
304 304 304 304 304 304 304 279
Source: Schenk, W.; Nabielek, H. Nucl. Technol. 1991, 96, 323–336.
85
Kr
1 103 5 105 8 105 5 107 7 107 9 105 4 105 1 103
137
Cs
4 103 2 103 1 103 3 104 4 104 6 102 1 101 5 101
TRISO-Coated Particle Fuel Performance
100
197
10–2 Compact Fuel element
2100 ⬚C 10–1 10–3
1800 ⬚C 10–4
Level of one particle failure 1700 ⬚C
fractional release
10–3
85Kr
85Kr
fractional release
10–2
10–5
10–6
10–4 14% FIMA 10–5
10–6
8–10% FIMA
1600 ⬚C 10–7
10–7
10–8 (a)
0
100
200
300
Heating time (h)
400
10–8
500
0
100
200
Heating time (h)
(b)
10–2
1800 ⬚C
85Kr
fractional release
10–3
1700 ⬚C
10–4
10–5
1600 ⬚C
10–6
10–7 (c)
0
100 200 Heating time (h)
300
Figure 33 Accumulated fractional release of 85Kr as a function of heating time at constant temperature; (a) spherical elements at 1600–1800 C, (b) compacts of 8–14% FIMA at 1600 C, and (c) compacts in the range 10–12% FIMA at 1600–1800 C. Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
300
198
TRISO-Coated Particle Fuel Performance
with defective SiC, while, in both cases, the pyrocarbon layers remain intact. Integral fractional releases of silver, cesium, krypton, and strontium are shown as a function of time at 1600 C for sphere HFR-K3/1 in Figure 35. The release of silver is on the order of 1–2% at the outset
10–03 AVR 88/33, 8.5 AVR 82/20, 8.6 AVR 71/22, 3.5 AVR 88/15, 8.7
137Cs release 85Kr release
10–04
Release fraction
One particle, 100% release 10–05
*including 100 ⬚C margin
10–06
10–07
10–08 –100
0
100
200
300
Time (h)
Figure 34 137Cs and 85Kr releases at 1600 C for AVR spherical fuel elements. Reproduced from Kendall, J.K.; et al. PBMR fuel performance envelope and test program, Pebble Bed Modular Reactor Limited, Document No. 039116, 2007.
of heating as considerable silver was released from fuel particles to the matrix during irradiation at temperatures in the range 1000–1200 C for 359 days. The release of cesium is considerably greater than the release of krypton and strontium, because krypton is held up by the PyC and strontium is retained in the UO2 kernel and graphite matrix to a greater extent than cesium. In Figure 35, the radial profiles of silver and cesium in the graphite matrix exhibit strong concentration gradients typical of materials that are diffusing, whereas the strontium profile is much more flat, indicating little diffusive release from the matrix. In this same figure, the release of cesium is observed to climb strongly after 200 h of heating, possibly indicating a deterioration of the SiC, perhaps from fission product palladium attack of SiC grain boundaries. Cesium releases as a function of time at temperature are shown in Figure 36, where it can be seen that five compacts with burnup in the range 10.7–13.9% FIMA exhibit higher releases than five spherical fuel elements with burnup in the range 3.5–8.9% FIMA at 1600 C. The results plotted at 1700 and 1800 C are from compacts in Table 58. The reason for the increased releases at 1600 C at high burnup and/or fluence is not well understood, but has been attributed to increased permeability of the SiC layer, related perhaps to fission product attack during postirradiation heating, especially at longer times (e.g.,
HFR-K3/1.75% FIMA 25 –2 100 4.0 ⫻ 10 m (E > 16 fJ) 1020–1200 ⬚C
10–1
Fractional of fission products in matrix
100
110mAg
Fractional release
10–2 10–3 10–4 10–5 137Cs
10–6 85Kr
10–7
90Sr
10–8 0
100 200 300 400 Heating time at 1600 ⬚C (h)
500
10–1 10–2
90Sr 137Cs
10–3 10–4
110mAg
After 500 h at 1600 ⬚C
10–5 10–6 Fuel zone 10–7 10–8 –30 –10 0 10 30 Fuel element radius (mm)
Figure 35 Fission product release and distribution in sphere HFR-K3/1 after irradiation at 1000–1200 C for 359 days and 1600 C heating. Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
TRISO-Coated Particle Fuel Performance
100
1800 ⬚C
the range 1900–2500 C is thermal decomposition of the SiC layer.50
10–1
1700 ⬚C
3.07.3.3
fractional release
10–2
137Cs
199
1600 ⬚C
10–3 Five compacts 10–4
1600 ⬚C
10–5 Five fuel elements
10–6
10–7 0
50
100
150 200 250 Heating time (h)
300
350
Figure 36 Cesium release during heating of spherical fuel elements (1600 C) and compacts (1600–1800 C). Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
European Experience
Several high-temperature accident tests have been performed at the new Ku¨FA installation at ITU.51 Prior to these tests, the upgraded system was tested to verify furnace operation, fission gas system performance, and condensation plate collection efficiency. It was found that the collection efficiency of the plate for Cs was 70% (16%), which is in agreement with the efficiency determined previously at Ju¨lich during earlier work with the furnace.38 The new accident tests on irradiated fuel include several fuel elements from the AVR reactor, from the HFR K6 irradiation test (proof test for HTR Modul reactor fuel element design), and from the more recent EU1bis irradiation test. Table 60 provides information and the irradiation history of the AVR and HFR K6 fuel elements tested to date. Because of the age of the fuel tested, no 110mAg data were collected in the latest accident tests. Testing on four HFR-EU1bis spheres has been completed, but data from these experiments have not yet been published. 3.07.3.3.1 AVR 73/21
cesium at times greater than 200 h in Figure 35, and temperatures above 1600 C). Ceramographic sections shown in Figure 37 show evidence of increasing degradation in the SiC layer for longer times at 1600 C and higher burnup, the most degraded being the SiC in sphere HFR-K3/1. Microprobe profiles through particles after heating, as shown in Figure 38, show the buildup of fission product palladium at the IPyC/SiC interface in spheres HFR-K3/1 and HFR-K3/3. It is hypothesized that corrosion by palladium degrades the SiC, leading to accelerated diffusion of cesium through grain boundaries.45 The distribution of metallic fission products averaged over a number of UO2 TRISO fuel element spheres examined after accident testing is shown in Table 59. At 1600 C,49 the kernel and the coatings are equally important in holding up cesium, while the kernel is the principal reservoir for strontium and silver. At 1800 C, cesium and silver are contained principally in the coatings, whereas strontium is retained in the kernel. The primary mechanism for coating failure and fission product release at extreme temperatures in
Testing of the AVR 73/21 fuel element was the first use of the upgraded Ku¨FA with irradiated fuel and was primarily conducted as a shakedown test of the equipment under hot conditions. The sphere was heated to 1600 C for 5 h, followed by a ramp to 1800 C and hold for an additional 5 h. The 85Kr release during this relatively short test was below the detection limit of the system. Two cold plates were exchanged during the test, but a problem was encountered with 137Cs contamination of the plate container from the hot cell, which interfered with analysis of released 137Cs on the plate and prompted a change in procedures to eliminate the carry-over contamination issue on subsequent tests. 3.07.3.3.2 AVR 74/18
The second accident test was performed on the AVR 74/18 sphere. The test took place in two parts: the first phase involved heating at 1600 C for 100 h, followed by a second phase of heating at 1800 C for an additional 100 h. Unplanned temporary furnace shutdowns were encountered during each heating cycle (one per cycle) due to operational issues with the furnace system, but
200
TRISO-Coated Particle Fuel Performance
100 mm
1600 ⬚C, 160 h
50 mm
FCs1374⫻10–5(FRJ2–K13/2;8%FIMA)
100 mm
1600 ⬚C, 500 h
1600 ⬚C, 500 h
20 mm
20 mm
FCs1372⫻10–5(AVR71/22;3.5%FIMA)
FCs1371⫻10–4(HFR–K3/1;7.7%FIMA)
Figure 37 Ceramographic sections through particles heated at 1600 C (complete particle followed by enlarged views of SiC layers) showing increasing degradation of the SiC layer with increasing burnup. Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
these do not appear to have significantly affected the test results. No particle failures were detected during this test. Fractional releases at the end of the test were 6 106 for 85Kr and 8 106 for 137Cs. 3.07.3.3.3 HFR K6/3
This heating test was carried out in four stages: 1600 C for 100 h, 1700 C for 100 h, 1800 C for 100 h, and 1800 C for 300 h. The temperature was returned to room temperature in between stages. 85Kr fractional release was low (<105) until the start of the fourth heating cycle, when a drastic increase in 85Kr was observed. By the end of the fourth heating cycle, the 85Kr fractional release was 4 104, indicating several particle failures. The 137Cs fractional release was low (106) after the first two heating cycles but increased dramatically during the third
cycle to a value of 1 103. The release further increased during the second 1800 C heating cycle, ending with a final value of 0.03. The results of the heating test are shown in Figure 39. SEM characterization of the particles at the end of the test revealed failed particles with cracked SiC layers. The results indicate good particle performance at up to 1700 C, with significant Cs release through intact coatings starting at 1800 C and multiple particle failures starting with the second 1800 C cycle. It is unclear if thermal cycling was responsible for the observed particle failures. 3.07.3.3.4 HFR K6/2
This heating test was performed in two stages: the first was at 1600 C for 100h followed by a second stage at 1800 C for 200 h. The temperature was not cycled to
TRISO-Coated Particle Fuel Performance
10–1 33
36
34
31
28
18
18
SiC
10 78
10–2
45
74
Parts of weight
41
21
48 36
78
50
78
77
Ba 10–3 Ru
Ag
Ru
10–4
13
26
SiC
LTI Buffer Buffer LTI layer layer
24 41
PyC Kernel PyC
201
81 26
78
47
34
30 10–5
39
30 40 50 60 70 80 10 20 Measure points radial through coating
0
(b) 10–1
SiC SiC PyC PyC Kernel PyC PyC LTI Buffer Buffer LTI LTI layer layer Cs Pd Pd
(a)
10–2 Parts of weight
Parts of weight
10–1 PyC SiC PyC Kernel PyC SiC PyC LTI LTI Buffer Buffer LTI LTI layer layer Cs
10–2 Pd
Pd
Cs Cs
10–4
10–3
10–4
10–5 0
(c)
10–3
10 20 30 40 50 60 70 80 Measure points radial through coating
0 (d)
10 20 30 40 50 60 70 80 Measure points radial through coating
Figure 38 Microprobe profiles of fission product elements through coatings of particles from HFR-K3. (a) Arrangement of sectioned particles (HFR-K3/3) for microprobe measurements. The numbers show the percentage loss of cesium from each particle after heating at 1800 C. (b) Barium, ruthenium, silver profiles in a particle with 78% cesium loss after 1800 C test. (c) Cesium, iodine, palladium profiles in a particle from HFR-K3/1 (0.01% cesium loss from sphere) after 1600 C test. (d) Cesium, iodine, palladium profiles in a particle (78% cesium loss) from HFR-K3/3. Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30. Table 59 Fuel type
UO2 TRISO
Averaged fission product distribution for spherical fuel elements after accident simulation tests Heating temperature ( C)
Time at temperature (h)
1600
Up to 500
Nuclide
137
Cs Sr 110m Ag 137 Cs 90 Sr 110m Ag 90
UO2 TRISO
1800
Up to 200
Fractional fission product content in: Kernel
Coating
Matrix
5 101 9.5 101 8 101 2 102 7 101 9 102
5 101 5 102 2 101 6 101 8 102 2 101
2 105–1 103 2 103–5 103 9 104 1.5 101 2 101 3 102
Fractional release from fuel element 5 105 1 106 1 103–3 102 5 102 3 103 7 101
Source: Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, USA, October 1–4, 2008.
202
TRISO-Coated Particle Fuel Performance
Fuel description/irradiation conditions for spheres accident tested in Ku¨FA at ITU
Table 60 Fuel element
AVR 73/21
AVR 74/18
HFR K6/3
HFR K6/2
Type Graphite Fuel particles Heavy metal (g) Fuel composition Enrichment (% 235U) Burnup (% FIMA) Irradiation time (EFPD) End of irradiation Irradiation temperature ( C) Thermal fluence (cm2) Fluence, E > 0.1 MeV (cm2)
GLE-3 A3-27 16 400 10 UO2 9.82 2.5 235 February 1984 700 2 1021 4 1020
GLE-3 A3-27 16 400 10 UO2 9.82 4.8 480 February 1985 700 4.15 1021 8 1020
GLE-4 similar A3-27 14 600 9.44 UO2 10.6 9.7 633 May 1993 1140 2.5 1021 4.8 1021
GLE-4 similar A3-27 14 600 9.44 UO2 10.6 9.3 633 May 1993 1140 2.5 1021 4.6 1021
Source: Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, Oct 1–4, 2008.
3.07.3.4
US Experience and Future Plans
0.1
2000 1800
1400
1E-3
1200 1E-4
1000 800
1E-5
600 Temperature (⬚C) Kr 137 Cs
400
85
200 0 0
100
200
300
400 500 600
700
Fractional release
0.01
1600 Temperature (⬚C)
room temperature at the end of the 1600 C test, but several interruptions of the experiment during the first stage occurred because of system malfunctions. The results of this heating test are shown in Figure 40. The 85Kr fractional release was below the detection limit of the system until 90 h into the 1800 C cycle, when the release increased up to 107. Approximately 120 h into the 1800 C cycle, another continuous release was observed that reached approximately 105 by the end of the test. 137Cs was released fairly continuously throughout the test up to a 2 103 fractional release at the end of the test. In general, these initial tests demonstrated excellent fuel performance at 1600 and 1700 C with zero particle failures. Higher 137Cs and 85Kr releases were observed at 1800 C, with some particle failures occurring for the sphere HFR K6/3.
1E-6 1E-7 800
Time (h)
Figure 39 Results of heating test HFR K6/3 at ITU. Reproduced from Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, USA, October 1–4, 2008.
3.07.3.4.1 Past experience
Very little work has been done on accident testing of US fuels as most irradiations produced significant irradiation-induced particle failure fractions. However, postirradiation anneals of long durations at temperatures up to 1500 C were performed to accelerate diffusive fission product releases from a variety of fuel types.52 UO2, UC2, UCO, UO2*(1), and UO2*(2) fuel particles were irradiated in the HRB-15B capsule in HFIR.52 In the UO2*(1) fuel, the kernel was coated with a ZrC layer, while in the UO2*(2) fuel, ZrC was dispersed in the buffer layer surrounding the kernel. These changes were intended to control free oxygen released during fission, which should improve fission product retentiveness. The fuel
burnup was in the range of 21–25% FIMA and the fast neutron fluence was in the range of 3.4 – 5.5 1025 n m2. The irradiation was accelerated with a residence time of 169 effective full power days. Only the fission product release data at 1500 C are discussed here, as cesium was not released at the lower temperatures. No fission product releases were measured at any temperature from UO2*(1) fuel particles. Ten particles of each fuel type were annealed for 11 866 h at 1500 C. Integral releases for each 10-particle batch were measured from individual particles by gamma counting each particle before and after the test and by periodic gamma monitoring of fission product collectors
TRISO-Coated Particle Fuel Performance
during the anneal as a function of time. The agreement of the integral releases from each 10-particle batch by these two methods was excellent. Cesium was released from only the UO2 and UC2 fuel particles as is shown in Figure 41. These same two fuel batches released the greatest fractions of silver as illustrated in Figure 42. The time signatures
2000
0.01
1800
Temperature (⬚C)
1400
1E-4
1200 1E-5
1000 800
1E-6
600 400
Fractional release
1E-3
1600
1E-7
200 1E-8
0 0
50
100
150 200 250 Time (h)
300
350
Temperature (⬚C) 85
Kr Cs
137
Figure 40 Results from heating test HFR K6/2 at ITU. Reproduced from Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, USA, October 1–4, 2008.
203
of the releases of cesium and silver from the UO2 fuel particles in Figures 41 and 42 indicate a diffusion release mechanism through the SiC layer. However, the release of cesium from the UC2 fuel batch is sudden in Figure 41, and the release of silver shows a rapid increase at the same time as the sudden release of cesium, as pointed out in Figure 43. The distribution of fission product releases among particles within the fuel batches in Table 61 indicates that the release of cesium from the UO2 fuel particles is from two of the ten particles and from only one particle in the UC2 fuel batch. This same table shows that the release of silver was 100% from the UO2 fuel batch, and 82% from the UC2, with 6 of the 10 UC2 particles releasing 100% of their silver inventories, two particles releasing 85–95%, one particle releasing 50%, and one particle retaining 100%. Particle-toparticle variations in fission product release dominate the behavior implied by the data shown in Table 61. The microstructures in Figure 44 show that the SiC layer in the UO2 batch exhibits large columnar grains, whereas the UCO batch exhibits a strong laminar grain structure in the SiC. The UC2 and UO2*(1) batches exhibit laminar structures in the SiC that are somewhat weaker than in the UCO batch. The results in Table 61 indicate that silver release at 1500 C is greatest (100%) in the case of columnar SiC, least (3%) for strongly laminar SiC, and intermediate (82%) for somewhat less strong laminar SiC. Although Cs was released from only three of the 50 particles annealed at 1500 C, two
Cesium release (%)
102
101
100
134Cs 137Cs
10−1 0
10
20 30 Annealing time (Ms)
40
50
Figure 41 Release of Cs from various types of TRISO-coated fuel particles at 1500 C. Reproduced from Bullock, R. E. J. Nucl. Mater. 1984, 125(3), 304–319.
204
TRISO-Coated Particle Fuel Performance
102
UO2 UC2
Silver release (%)
UO2* (2) 101
Release from UO2* (1) = 0 UCO
100
10−1 0
10
20 30 Annealing time (Ms)
50
40
Figure 42 Release of 110Ag from various types of TRISO-coated fuel particles at 1500 C. Reproduced from Bullock, R. E. J. Nucl. Mater. 1984, 125(3), 304–319.
102
Ag
Cs
Release (%)
101
100
10−1
0
10
20 30 Annealing time (Ms)
40
50
Figure 43 Abrupt 10% increase in 110Ag release from UC2 particles at 1500 C when one of the ten test particles released its entire Cs inventory.
particles had columnar SiC and one had a somewhat weak laminar SiC. The sensitivity of cesium release to SiC grain structure was recognized in Myers53 where the diffusivity of cesium through columnar SiC was given as an order of magnitude greater than through laminar SiC. The diffusion equations from Myers53 are accessible in Table A-3 of IAEA.3 As shown in Figure 45, releases of europium are greatest (37–46%) for the fuel batches containing UC2 in the kernel, compared with fuel batches
containing only UO2 in the kernel (9–16%). As shown in Table 61, cerium release is 45% in UC2, only 1% in UCO, and nil in UO2 particles. These behaviors are related to the thermodynamics of rareearth oxides and carbides, according to Homan et al.,54 where oxides formed in UO2 (such as Eu2O3 and Ce2O3) are less likely to escape from the kernel than are the more mobile rare-earth carbides formed in UC2. In UCO, europium forms a carbide and cerium forms an oxide.54
TRISO-Coated Particle Fuel Performance
Table 61
205
Distribution of fission product release within particle batches during postirradiation annealing
Annealing temperature ( C)
TRISO particle typea
1500
UC2 with laminar SiC
9 ¼ 0% 1 ¼ 99% 10 ¼ 12%
1500
UO2 with columnar SiC
1500
UC0.4O1.6 with laminar SiC
8 ¼ 0% 2 ¼ 99% 10 ¼ 24% 10 ¼ 0%
1500
UO2*(2) with laminar SiC
10 ¼ 0%
1350 1350
UC0.4O1.6 with laminar SiC UO2*(2) with laminar SiC
10 ¼ 0% 10 ¼ 0%
1200 1200
UC0.4O1.6 with laminar SiC UO2*(2) with laminar SiC
10 ¼ 0% 10 ¼ 0%
Release breakdown from the ten particles within a test batch for: 134 137
Cs Cs
110m
Ag
1 ¼ 0% 1 ¼ 50% 85% < 2 < 95% 6 ¼ 100% 10 ¼ 82% 10 ¼ 100% 7 ¼ 0% 10% < 3 < 20% 10 ¼ 3% 7 ¼ 0% 70% < 3 < 80% 10 ¼ 27% 10 ¼ 0% 7 ¼ 0% 45% < 3 < 55% 10 ¼ 19% 10 ¼ 0% 10 ¼ 2%b
154
144
15% < 5 < 25% 45% < 3 < 55% 2 ¼ 100% 10 ¼ 46% Uniform release of 16%
12% < 3 < 18% 18% < 3 < 25% 70% < 3 < 80% 1 ¼ 99% 10 ¼ 45% 10 ¼ 0%
Uniform release of 37%
10 ¼ 1%b
2 ¼ 0% 0% < 5 < 10% 15% < 3 < 25% Uniform release of 23% 10 ¼ 4%b
10 ¼ 0% 10 ¼ 0% 10 ¼ 0%
Uniform release of 6% 10 ¼ 0%
10 ¼ 0% 10 ¼ 0%
Eu
Ce
There was zero release within about 5% as determined from individual particle counting before and after annealing for all isotopes from each of the ten particles in all test combinations not listed, that is, UO2*(1) at all temperatures, and UC2, UO2, and UC0.4O1.6 at 1350 and 1200 C. As no release on collectors was detected at the 0.01% level from the combined ten particles within each of these test batches, it can be assumed that release from any one of these particles was certainly <0.01% and was probably not more than 0.001%. b These total releases from ten particles were too small and too uniformly distributed among particles to allow the determinations of individual release values. a
These results,42 with admittedly relatively few particles, indicate that under a long annealing time at 1500 C (a) silver and cesium releases are at a maximum in the case of SiC with a columnar grain structure; (b) europium releases are largest in UC2 fuels, but can be significant in UCO and, to a lesser extent, in UO2 fuels; and (c) cerium release is significant only in UC2 fuel. 3.07.3.4.2 Future plans
The NGNP/AGR fuel program is currently planning an extensive campaign of accident testing of irradiated TRISO fuel compacts to measure fission product retention at elevated temperatures and provide verification of fuel performance code predictions. The first accident tests will be performed on irradiated fuel compacts from the AGR-1 experiment, using both the FACS and CCCTF test equipment described earlier. Individual fuel compacts in each capsule will be chosen for accident tests based primarily on their irradiation history (i.e., burnup, fast fluence, and irradiation temperature). Both isothermal and transient temperature tests will
be performed on the AGR-1 fuel with maximum temperatures between 1600 and 1800 C. Hightemperature margin tests up to 2000 C may also be performed if warranted, based on the results of the lower temperature tests. The primary released radionuclides of interest are 85Kr, 137Cs, 134Cs, 110mAg, 90 Sr, and 154Eu. (131I and 133Xe are also of interest, but will require reirradiation of the fuel in a test reactor to regenerate these isotopes because of rapid decay times. Work is underway to develop this capability at INL.) Condensate plate analysis will consist of initial measurement of sample activity using gamma spectrometry. Following this step, the fission products will be chemically stripped from the plates, and the solutions analyzed for other fission products of interest. The primary fission product of interest will be 90 Sr, a beta emitter. The strontium can be separated from the rest of the analytes in an ion exchange resin, and the 90Sr activity measured using scintillation techniques. The feasibility of chemical separation of iodine from the solution has also been demonstrated, and the separated iodine solution can be analyzed for
206
TRISO-Coated Particle Fuel Performance
(a)
UCO
(b)
UO2
(c)
UC2
(d)
UO*2(1)
20 mm
Figure 44 Microstructures of etched SiC barrier layers in various types of TRISO-coated particles. Reproduced from Bullock, R. E. J. Nucl. Mater. 1984, 125(3), 304–319.
131
I using gamma spectroscopy. Alternatively, 90Sr, as well as other radioisotopes of interest, can be analyzed using mass spectrometric methods. In preparation for this accident testing campaign, the FACS furnace system will undergo remote qualification testing, fission gas system performance checkout, and integrated testing with the FACS furnace system, condensation plate collection efficiency calibrations, and installation in the hot cells at the INL HFEF. Collection efficiency calibration of the upgraded CCCTF deposition cup will also be performed, and the results compared with those of the FACS furnace so that the performance of both testing systems can be compared using known standards.
PIE of fuel compacts after accident testing will involve a number of methods designed to study fission product retention of the particles during the heating cycles in further detail and to document the condition of the fuel kernels and coating layers after the high-temperature testing. This will include microscopy on compact and particle cross-sections to examine fuel kernel and coating microstructures as well as evidence of damage; deconsolidation of compacts; leach-burn-leach analysis to determine SiC coating failure fractions and to measure fission products held up in the compact matrix material (determined during the first leach step); and gamma spectrometry of individual fuel particles to study metallic fission product retention behavior and search for failed particles which can be used in subsequent analysis to determine failure modes. Accident testing in inert atmospheres on irradiated fuel from subsequent AGR experiments (a total of eight different irradiation experiments are currently planned16) will also be performed with various objectives, including testing performance of fuel produced in large scale coaters, studying fission product transport at high temperatures, and performing proof tests to qualify the reference fuel for licensing. In addition, the program has plans to perform air and water ingress accident testing on irradiated fuel to verify fuel performance in these more reactive environments. The facilities to perform the air/water tests have not yet been developed. 3.07.3.5
Japanese Experience
Individual UO2-TRISO fuel particles deconsolidated from a compact of Japanese manufacture were heated at 1700 C for 270 h and 1800 C for 222 h at ORNL.42 The fuel had been irradiated in HFIR in the HRB-22 capsule to a burnup of 4.8% FIMA and a fast neutron fluence of 2.1 1025 n m2 for a duration of 89 effective full-power days (EFPD) and a timeaveraged maximum temperature of 1100 C. Releases of silver, cesium, europium, and krypton were measured as a function of time as shown in Figures 46 and 47 for batches of 25 particles at each temperature. The krypton release in Figure 47 shows that one particle failed early in the heating, also releasing antimony. Fission product inventories of the individual fuel particles were measured before and after the heating tests with the IMGA apparatus. Results of these measurements are shown for silver, cesium, and europium in Figure 48 (1700 C) and Figure 49 (1800 C). Both these figures exhibit large variations in fission
TRISO-Coated Particle Fuel Performance
207
102
Europium release (%)
UC2 UCO UO2 UO2* (2)
101
100 Release from UO2* (1) = 0
10−1 0
10
20 30 Annealing time (Ms)
50
40
Figure 45 Release of 154Eu from various types of TRISO-coated fuel particles of 1500 C. Reproduced from Bullock, R. E. J. Nucl. Mater. 1984, 125(3), 304–319.
100
100
110 mAg 110mAg
10–2
154Eu 137Cs
10–3
85Kr
10–4
10–5
137Cs
10–1 Fractional release
Fractional release
10–1
125Sb
10–2
85Kr
154Eu
10–3
10–4
0
50
100
150
200
250
300
Heating time (h)
10–5
0
50
100 150 Heating time (h)
200
250
Figure 46 Time-dependent fractional releases of fission products during the ACT3 heating test at 1700 C for 270 h, obtained by the online measurements of fission gas-release and intermittent measurements of metallic fission product release. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
Figure 47 Time-dependent fractional releases of fission products during the Act 4 heating test at 1800 C for 222 h, obtained by the online measurements of fission gas-release and intermittent measurements of metallic fission product release. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
product release from particle to particle. At 1700 C, silver release varies from 10% to 100%, cesium from 0% to 20%, and europium from 5% to 30%. At 1800 C, not including the failed particle, the silver release varied from 24% to 100%, cesium from 0% to 55%, and europium from 0% to 69%. Individual particles were examined at JAERI by
X-ray microradiographs, ceramography, and electron microprobe. Accumulations of fission products in the buffer show up as bright spots in X-ray microradiographs of fuel particles from which large fission product releases were measured, as shown in Figure 50 (1700 C) and Figure 51 (1800 C). In Figure 50,
TRISO-Coated Particle Fuel Performance
Inventory ratio of post-/preheating tests
208
1 0.8
0.6 0.4
144Ce 95Zr 106Ru
0.2
125Sb
0 0
5
10 15 Particle number
Inventory ratio of post-/preheating tests
(a)
20
25
20
25
1 0.8
0.6 0.4 110mAg 137Cs
0.2
154Eu
0 0
5
10 15 Particle number
(b)
Figure 48 Inventory ratios of post- to preheating tests in individual particles in ACT3 measured with the IMGA system: (a) 95Zr, 106 Ru, 125Sb, and 144Ce; (b) 110mAg, 137Cs, and 154Eu. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47. 1.2
1.0 0.8 0.6 0.4 0.2 0
(a)
Inventory ratio of post-/ preheating tests
Inventory ratio of post-/ preheating tests
1.2
144Ce
0
5
95Zr
10
106Ru
15
125Sb
20
Particle number
(b)
137Cs
154Eu
0.8 0.6 0.4 0.2 0 0
25
110mAg
1.0
5
10
15
20
25
Particle number
Figure 49 Inventory ratios of post- to preheating test in individual particles in ACT4 measured with the IMGA system: (a) 95Zr, 106 Ru, 125Sb, and 144Ce; (b) 110mAg, 137Cs, and 154Eu. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
particle ACT3–5 had relatively little fission product release and shows no evidence of fission product accumulation in the buffer, whereas particle ACT3–6 had relatively large releases and shows a
bright spot in the buffer. A similar trend is shown in Figure 51 for a relatively non-releasing particle (ACT4–3) and two strongly releasing particles (ACT 4–9 and ACT 4–13). These figures also show
TRISO-Coated Particle Fuel Performance
(a)
100 mm
(b)
100 mm
(c)
100 mm
(d)
100 mm
209
Figure 50 X-ray microradiographs and ceramographs of the particles after the ACT3 heating test: (a) and (b) show particle ACT3-5; (c) and (d) show particle ACT3-6. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
that the buffers of the low-releasing particles are intact, whereas those of the heavily releasing particles are severely cracked. In addition, the kernels of the releasing particles in Figures 50 and 51 exhibit larger pores than the kernels of the nonreleasing particles. The SiC layers of all the particles show signs of degradation as seen in Figures 50–52. Accumulations of fission products, especially palladium, were found in all particles at the IPyC/SiC interface, as shown in Figure 53, and sometimes within the SiC layer. The Japanese work corroborates results from Germany in that wide variations are measured in fission product release from particle to particle, palladium buildup at the IPyC/SiC interface is observed in both releasing and nonreleasing particles, and fission product releases increase with increasing test temperature.
3.07.4 Conclusions TRISO-coated particle fuel has been fabricated and tested around the world. A review of the fuel performance database indicates that high-quality lowdefect TRISO fuel, the heart of the high-temperature
gas-cooled reactor, is achievable if the following conditions are met: a. Disciplined and controlled fabrication of the coated particles under conditions that produce the proper microstructure of the kernels and coatings with robust quality control to demonstrate the statistical adequacy of the pedigree of the material. Much of the poor performance observed in the past is associated with improper fabrication conditions. b. Selection of normal operation conditions that will result in very low failures. For a pebble bed reactor using 500-mm UO2 TRISO fuel, an envelope of 1150 C, 10% FIMA, and 4 1025 n m2 was demonstrated by German fuel developers in the 1980s and remains the standard for current pebble bed designs. For prismatic gas-cooled reactors using 350-mm UCO TRISO fuel, an envelope of 1250 C, 19% FIMA, and 4 1025 n m2 appears achievable, on the basis of the results of the US AGR-1 irradiation. c. Well-planned series of irradiations that limit the burnup and fluence acceleration to preclude artificial failures that would not be expected in the actual reactor application, yet contain enough fuel
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TRISO-Coated Particle Fuel Performance
(a)
(c)
(e)
100 mm
100 mm
100 mm
(b)
(d)
(f)
100 mm
100 mm
100 mm
Figure 51 X-ray micrographs and ceramographs of the particles after the ACT4 heating test: (a) and (b) show particle ACT4-3; (c) and (d) show particle ACT4-9; and (e) and (f) show particle ACT4-13. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
to draw statistically significant conclusions about fuel performance. To meet an in-service failure probability of 104 at 95% confidence, the minimum number of particles to be tested is approximately 300 000, allowing for some failures during the experiment. d. Selection of an accident envelope that will result in very low failures. For both pebble bed and prismatic designs, limiting accident temperatures in depressurized conduction cooldown scenarios to 1600 C is acceptable. The large accident database developed by German fuel developers in the 1980s remains the standard for gas reactors worldwide. There has been very limited accident testing
for UCO TRISO, but work is underway to begin such testing in the near future. The current accident database at 1700 and 1800 C is inadequate to determine with statistical confidence if TRISOcoated UO2 or UCO can sustain these higher accident temperatures. Future R&D programs will investigate the fuel performance at these elevated temperatures. e. Well-planned accident safety testing that tests enough fuel to draw statistically significant conclusions (population on the order of 200 000 particles) and measures all relevant radionuclides important to the ultimate safety and licensing of the high-temperature gas-cooled reactor.
TRISO-Coated Particle Fuel Performance
211
The world TRISO-coated particle fuel database suggests that these goals are attainable. See Chapter 2.10, Graphite: Properties and Characteristics; Chapter 3.06, TRISO Fuel Production; Chapter 3.08, Advanced Concepts in TRISO Fuel; Chapter 3.24, TRISO Fuel Performance Modeling and Simulation; and Chapter 4.07, Radiation Effects in SiC and SiC-SiC.
20 mm
(a)
(b)
20 mm
(c)
20 mm
Acknowledgments
Figure 52 Ceramographs of coating layers of the particles after the ACT4 heating test: (a) shows particle ACT4-3, (b) shows particle ACT4-9, and (c) shows particle ACT4-13. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
This manuscript was authored by Battelle Energy Alliance, LLC, under Contract Number DE-AC0705ID14517 with the US DOE. The US Government retains and the publisher, by accepting the article for publication, acknowledges that the US Government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US Government purposes. This information was prepared as an account of work sponsored by an agency of the US Government. Neither the US Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. References herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise do not necessarily constitute or imply its endorsement, recommendation, or favoring by the US Government or any agency thereof. The views and opinions of the authors expressed herein do not necessarily state or reflect those of the US Government or any agency thereof.
SEI
(a)
10 mm
Pd
(b)
Figure 53 Electron probe microanalysis of coating layers of particle ACT4-3 after the ACT4 heating test: (a) secondary electron image and (b) X-ray image for palladium. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
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TRISO-Coated Particle Fuel Performance
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Gontard, R.; Nabielek, H. Performance evaluation of modern HTR TRISO fuels; Technical Report, HTA-IB-05/ 90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, Jul 1990. Verfondern, K.; Nabielek, H.; Kendall, J. M. Nucl. Eng. Technol. 2007, 39(5), 603–616. IAEA. Fuel performance and fission product behavior in gas-cooled reactors; IAEA TECDOC-978; Vienna, Austria, Nov 1997. Generation IV Technology, Systems. Very-high temperature reactor website: http://www.gen-4.org/ Technology/systems/vhtr.htm/. Versluis, R.; et al. Project deep-burn: Development of transuranic fuel for high temperature heliumcooled reactors. In 4th International Topical Meeting on High Temperature Reactor Technology, HTR-2008, Washington, DC, Sept 28–Oct 1, 2008. Kuijper, J. C. Pu and Ma management in thermal HTGRs – impact at fuel, reactor and fuel cycle levels. In 4th International Topical Meeting on High Temperature Reactor Technology, HTR-2008, Washington, DC, Sept 28–Oct 1, 2008. Verwimp, A.; et al. Irradiation and post irradiation capabilities at SCK CEN. In OECD Proceedings, First Information Exchange Meeting on Survey on Basic Studies in the Field of High Temperature Engineering, Paris, France, Sept 27–29, 1999. Koshcheev, K.; et al. Experimental capabilities of IVV-2M reactor and its research complex to determine workability of coated particles and fuel elements of high temperature gas-cooled reactors. In OECD Proceedings, First Information Exchange Meeting on Survey on Basic Studies in the Field of High Temperature Engineering, Paris, France, Sept 27–29, 1999. Conrad, R.; Bakker, K. Irradiation of fuels for high temperature gas-cooled reactors at the HFR Petten. In International HTR Fuel Seminary, Brussels, Belgium, Feb 2002. ORNL Neutron Sciences. HFIR operating status website: http://neutrons.ornl.gov/hfir/index.shtml/. Grover, S. B.; Furstenau, R. V. Capabilities and facilities available at the Advanced Test Reactor to support development of the next generation reactors. In Proceedings of GLOBAL 2005, Advanced Nuclear Fuel Cycles and Systems, Tsukuba, Japan, Oct 9–13, 2005. NTP Facilities. The SAFARI-1 nuclear reactor website: http://www.radioisotopes.co.za/. Maki, J. T.; et al. J. Nucl. Mater. 2007, 317(1–3), 270–280. Olander, D. R. Fundamental Aspects of Nuclear Reactor Fuel Elements; National Technical Information Service, US Department of Commerce: Springfield, VA, 1976; Publication No. TID-26711-P1. Bell, G. L.; et al. Technical program plan for the advanced gas reactor fuel development and qualification program; Oak Ridge National Laboratory Report, ORNL/TM-2002/ 262; Apr 2003. Petti, D. A.; et al. Technical program plan for the advanced gas reactor fuel development and qualification program; Idaho National Laboratory Report, INL/EXT-05-00465, Revision 1; Aug 2005. Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsule F-30; General Atomics Report, GA-A13208, UC-77; 1975. Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsules HRB-4, HRB-5, and HRB-6; General Atomics Report, GA-A13267, UC-77; 1975.
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Tiegs, T. N.; Thoms, K. R. Operation and post irradiation examination of ORR capsule OF-2: Accelerated testing of HTGR fuel; ORNL-5428; 1979. Young, C. A. Pre- and Post-Irradiation Evaluation of Fuel Capsule HRB-14; General Atomics Report, GA-A15969, UC-77; 1980. Ketterer, J. W.; Bullock, R. E. Capsule HRB-15B Post irradiation Examination Report; General Atomics Report, GA-A15940, UC-77; 1981. Brodda, B. G.; et al. The German–US cooperative experiment R2-K13, Part I: Irradiation of UCO and ThO2 TRISO particles in prismatic block segments; KFA-HBK-IB-09/85; 1985. Ketterer, J.; et al. Capsule HRB-15A Post Irradiation Examination Report; General Atomics Report, GA-A16758, UC-77; 1984. Ketterer, J. W.; Myers, B. F. Capsule HRB-16 Post Irradiation Examination Report; HTGR-85-053; 1985. Baldwin, C. A.; et al. Interim Post Irradiation Examination Data Report for Fuel Capsule HRB 21; ORNL/M-2850; 1993. Baldwin, C. A.; et al. The New Production Reactor Fuel Post Irradiation Examination Data Report for Capsules NPR 1, NPR 2, and NPR 1A; ORNL/M-2849; 1993. Maki, J. T.; et al. NP-MHTGR Fuel Development Program Results; INEEL/EXT-2002-1268; Oct 2008. Grover, S. B. Final assembly and initial irradiation of the first advanced gas reactor fuel development and qualification experiment in the advanced test reactor. In Proceedings of the International Congress on Advances in Nuclear Power Plants, Paper 7035, Nice, France, May 13–18, 2007. Petti, D. A.; et al. Status of the first advanced gas reactor fuel irradiation experiment in the advanced test reactor. In 4th International Topical Meeting on High Temperature Reactor Technology, HTR-2008, Washington, DC, Sept 28–Oct 1, 2008. Maki, J. T. AGR-1 irradiation experiment test plan; Idaho National Laboratory Report, INL/EXT-05-00593; Aug 2005. Scates, D. M.; et al. Fission product monitoring of TRISOcoated fuel for the advanced gas reactor-1 experiment. In Paper HTR2008-58202, 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, Sept 28–Oct 1, 2008. Martin, R. C. Compilation of fuel performance and fission product transport models and database for MHTGR design; ORNL/NPR-91/6; Oct 1993. Fu¨tterer, M. A.; et al. Nucl. Eng. Des. 2008, 238(11), 2877–2885. Marmier, A.; et al. Preliminary results of the HFREU1 fuel irradiation of INET and AVR pebbles in the HFR Petten. In Paper HTR2008-58049, 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, Sept 28–Oct 1, 2008. Tang, C.; et al. The behavior of HTR-10 fuel under irradiation. In Paper #B11, Proceedings of the 2nd International Topical Meeting on High Temperature Reactor Technology, Beijing, China, Sept 22–24, 2004. Tang, C.; et al. Nucl. Eng. Des. 2006, 236(1), 107–113. Petti, D. A.; et al. Nucl. Eng. Des. 2003, 222(2–3), 281–297. Schenk, W.; Pitzer, D.; Nabielek, H. Fission product release profiles from spherical HTR fuel elements at accident temperatures; Ju¨l-2234; Sept 1988. Hantke, H.-J. Performance of high quality HTR – LEU fuel elements with TRISO-coated particles; HTA-IB-7/92; Dec 1992.
TRISO-Coated Particle Fuel Performance 40. Kostecka, H.; Ejton, J.; de Weerd, W.; Toscano, E. Post irradiation testing of HTR-fuel elements under accident conditions: Part II. In 2nd International Topical Meeting on High Temperature Reactor Technology, Beijing, China, Sept 22–24, 2004. 41. Morris, R. N.; et al. The core conduction cooldown test facility: Current status and issues; ORNL/NPR-91/7; Jan 1992. 42. Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47. 43. Morris, R. N.; et al. HRB-17 and HRB-18 HEU TRISO UCO unbonded irradiated core conduction cooldown tests; ORNL/NPR-92/9; Jun 1992. 44. Kellerbauer, A.; Schenk, W.; Pott, G.; Nabielek, H. Private communication; Institute for Transuranium Elements; Sept 2, 2009. 45. Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
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Kendall, J. K.; et al. PBMR fuel performance envelope and test program, Pebble Bed Modular Reactor Limited, Document No. 039116; 2007. Proksch, E.; Strigl, A.; Proksch, E.; Nabielek, H. J. Nucl. Mater. 1982, 107(2–3), 280–285. Verfondern, K.; et al. Nucl. Technol. 1990, 91, 235–246. Schenk, W.; Nabielek, H. Nucl. Technol. 1991, 96, 323–336. Nabielek, H.; et al. Nucl. Technol. 1989, 84(1), 62–81. Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, Oct 1–4, 2008. Bullock, R. E. J. Nucl. Mater. 1984, 125(3), 304–319. Myers, B. F. Cesium diffusion in silicon carbide during post irradiation anneals; Technical Note: KFA-HBK-TN-01183; Research Center Ju¨lich; 1984. Homan, F. J.; et al. Nucl. Technol. 1977, 35(2), 428–441.
TRISO-Coated Particle Fuel Performance
D. A. Petti, P. A. Demkowicz, and J. T. Maki Idaho National Laboratory, Idaho Falls, ID, USA
R. R. Hobbins RRH Consulting, Wilson, WY, USA
Published by Elsevier Ltd.
3.07.1
Introduction
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3.07.2 3.07.2.1 3.07.2.1.1 3.07.2.1.2 3.07.2.1.3 3.07.2.1.4 3.07.2.1.5 3.07.2.1.6 3.07.2.1.7 3.07.2.1.8 3.07.2.1.9 3.07.2.1.10 3.07.2.2 3.07.2.2.1 3.07.2.2.2 3.07.2.2.3 3.07.2.2.4 3.07.2.2.5 3.07.2.2.6 3.07.2.2.7 3.07.2.2.8 3.07.2.2.9 3.07.2.3 3.07.2.3.1 3.07.2.3.2 3.07.2.3.3 3.07.2.3.4 3.07.2.3.5 3.07.2.3.6 3.07.2.3.7 3.07.2.3.8 3.07.2.3.9 3.07.2.3.10 3.07.2.3.11 3.07.2.3.12 3.07.2.3.13 3.07.2.4 3.07.2.5 3.07.2.6 3.07.2.7 3.07.2.7.1 3.07.2.7.2 3.07.2.7.3 3.07.2.7.4
Irradiation Performance Overview of Irradiation Facilities and Testing BR-2 IVV-2M HFR Petten HFIR ATR SAFARI TRISO-coated particle fuel irradiation testing Thermal and physics analysis considerations Gas control system considerations FPMS considerations German Experience R2-K12 and R2-K13 BR2-P25 HFR-P4 SL-P1 HFR-K3 FRJ2-K13 FRJ2-K15 FRJ2-P27 HFR-K6 and HFR-K5 US Experience F-30 HRB-4 and HRB-5 HRB-6 OF-2 HRB-14 HRB-15B R2-K13 HRB-15A HRB-16 HRB-21 NPR-1 and NPR-2 NPR-1A AGR-1 European Experience Chinese Experience Japanese Experience Irradiation Performance Summary Heavy metal contamination In-service failures Failure mechanisms Acceleration effects
154 154 154 154 155 155 155 155 156 156 158 159 160 160 161 162 162 163 163 164 164 165 165 166 167 169 170 171 173 174 174 175 176 177 178 178 180 181 184 185 185 186 186 187 151
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3.07.3 3.07.3.1 3.07.3.1.1 3.07.3.1.2 3.07.3.1.3 3.07.3.1.4 3.07.3.2 3.07.3.3 3.07.3.3.1 3.07.3.3.2 3.07.3.3.3 3.07.3.3.4 3.07.3.4 3.07.3.4.1 3.07.3.4.2 3.07.3.5 3.07.4 References
Safety Testing Facility Overview Ku¨FA at ITU INL’s FACS ORNL’s Core Conduction Cooldown Test Facility KORA German Experience European Experience AVR 73/21 AVR 74/18 HFR K6/3 HFR K6/2 US Experience and Future Plans Past experience Future plans Japanese Experience Conclusions
Abbreviations AGR ATR AVR BAF BISO BOL BR-2 CCCTF CVD DOE EFPD EOL FACS FIMA FPMS FRJ GETR HEU HFEF HFIR HFR HRB HTGRs HTR-10 HTTR IFEL IMGA
Advanced Gas Reactor Advanced Test Reactor Arbeitsgemeinschaft Versuchsreaktor Bacon anisotropy factor Bi-structural isotropic Beginning of life Belgian Reactor 2 Core Conduction Cooldown Test Facility Chemical vapor deposition Department of Energy Effective full-power day End of life Fuel accident condition simulator Fissions per initial metal atom Fission product monitoring system Research Reactor Juelich General Electric Test Reactor Highly enriched uranium Hot Fuel Examination Facility High-Flux Isotope Reactor High-Flux Reactor HFIR Removable Beryllium High-temperature gas-cooled reactors High Temperature Reactor 10 High-temperature test reactor Irradiated fuel examination laboratory Irradiated microsphere gamma analyzer
INET
189 189 189 190 192 193 193 199 199 199 200 200 202 202 205 206 209 212
Institute of Nuclear and New Energy Technology INL Idaho National Laboratory IPyC Inner pyrolytic carbon ITU Institute for Transuranium Elements JMTR Japan Material Test Reactor KuFA Cold finger apparatus (in German) LEU Low-enriched uranium LHTGR Large High Temperature Gas Reactor LTI Low temperature isotropic MOL Middle of life NE-MHTGR Commercial version of NP-MHTGR NGNP Next Generation Nuclear Plant NP-MHTGR New Production Modular High-temperature Gas-Cooled Reactor ORNL Oak Ridge National Laboratory ORR Oak Ridge Research Reactor PIE Postirradiation examination R&D Research and development R/B Release to birth ratio SiC Silicon carbide TRIGA Training research and isotope production, General Atomics TRISO Tristructural isotropic UCO Uranium oxycarbide Uranium dioxide UO2 VHTR Very-high-temperature reactors VXF Vertical experimental facility WAR Weak acid resin
TRISO-Coated Particle Fuel Performance
3.07.1 Introduction For all high temperature gas reactors (HTGRs), tristructural isotropic (TRISO)-coated particle fuel forms the heart of the concept. Such fuels have been studied extensively over the past four decades around the world, for example, in countries including the United Kingdom, Germany, Japan, United States, Russia, China, and more recently, South Africa. In early gas-cooled reactors, the coated particle fuel form consisted of layers of carbon surrounding the fissile kernels. Highly enriched uranium (HEU) and thorium carbides and oxides were used as fissile and fertile kernels. Ultimately, the carbon layer coating system (termed BISO for bistructural isotropic) was abandoned because it did not sufficiently retain fission products, leading to the development of the current three-layer coating system (termed TRISO for tristructural isotropic). In TRISO-coated fuel, a layer of silicon carbide (SiC) is sandwiched between pyrolytic carbon layers. This three-layer system is used to both provide thermomechanical strength to the fuel and contain fission products. In addition, for operational and economic reasons, the fuel kernel of choice today is low-enriched uranium (LEU) uranium dioxide (UO2) for the pebble bed design and uranium oxycarbide (UCO) for the prismatic design. In both pebble bed and prismatic gas reactors, the fuel consists of billions of multilayered TRISOcoated particles (750–830 mm in diameter) distributed within fuel elements in the form of circular cylinders (12.5 mm in diameter and 50 mm long) called ‘compacts’ or spheres called ‘pebbles’ (6 cm in diameter). The active fuel kernel is surrounded by a layer of porous carbon, termed ‘the buffer’; a layer of dense carbon, termed ‘the inner pyrolytic carbon layer’; a layer of SiC; and another dense carbon layer, termed ‘the outer pyrolytic carbon layer.’ These collectively provide for accommodation and containment of fission products generated during operation. The buffer layer is designed to accommodate fission recoils, volumetric swelling of the kernel, and fission gas released under normal operation. The inner pyrolytic carbon layer protects the kernel from reactive chlorine compounds produced during SiC deposition in the chemical vapor deposition (CVD) coater. The SiC layer provides structural strength to the particle. The outer pyrolytic carbon layer protects the particles during formation of the fuel element. Under normal operation, radiation damage causes shrinkage of the pyrolytic carbon layers, which induces compressive stresses in the SiC layer to counteract tensile stresses associated with fission gas release. All three layers of
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the TRISO coating system exhibit low permeability. These fuel constituents are extremely stable and are designed not to fail under normal operation or anticipated accident conditions, thereby providing effective barriers to the release of fission products. Figure 1 is a montage of TRISO fuel used in both prismatic and pebble bed high-temperature gas reactors. Rigorous control is applied at every step of the fabrication process to produce high-quality, very lowdefect fuel. Defect levels are typically on the order of one defect per 100 000 particles. Specifications are placed on the diameters, thicknesses, and densities of the kernel and layers; the sphericity of the particle; the stoichiometry of the kernel; the isotropy of the carbon; and the acceptable defect levels for each layer. Statistical sampling techniques are used to demonstrate compliance with the specifications usually at the 95% confidence level. For example, fuel production for German reactors in the 1980s yielded only approximately 100 defects in 3.3 million particles produced. This remains the standard for gas-cooledreactor fuel production today.1,2 Irradiation performance of high-quality, lowdefect coated particle fuels has been excellent. In Section 3.07.2, a detailed review of the state of the art in irradiation testing, capabilities of existing fission reactors worldwide to irradiate TRISO fuel, and the irradiation behavior of modern TRISO-coated particle fuel around the world will be discussed. Testing of German fuel under simulated accident conditions in the 1980s has demonstrated excellent performance. Section 3.07.3 describes the accident behavior of TRISO-coated particle fuel largely on the basis of the German database and the plans to perform similar testing for the current generation of TRISOcoated fuels. Additional limited testing of TRISOcoated particle fuel performed under air and water ingress events and under reactivity pulses has been reported elsewhere3 and will not be repeated here. The outstanding irradiation and accident simulation testing results obtained by German researchers form the basis for fuel performance specifications used in gas-cooled-reactor designs today. Specifications for in-service failure rates under irradiation and accident conditions are very stringent, typically on the order of 104 and 5 104, respectively. Significant research and development (R&D) related to TRISO-coated fuels is underway worldwide as part of the activities of the Generation IV International Forum on Very-High-Temperature Reactors (VHTRs). The focus is largely on extending the capabilities of the TRISO-coated fuel system for higher
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TRISO-Coated Particle Fuel Performance
Pyrolytic carbon Silicon carbide Uranium dioxide or oxycarbide kernel
Prismatic
Pebble
Particles
Matrix
Compacts
Fuel element
TRISO-coated fuel particles (left) are formed into fuel compacts (center) and inserted into graphite fuel elements (right) for the prismatic reactor
Kernel Buffer layer
5 mm graphite layer Coated particles imbedded in graphite matrix
Inner PyC-layer Fuel-free shell SiC-layer Fueled zone Outer PyC-layer
Fuel sphere Dia 60 mm
Half section
TRISO-coated fuel particles are formed into fuel spheres for pebble bed reactor Figure 1 TRISO-coated particle fuel and compacts and fuel spheres used in high temperature gas reactors.
burnups (10–20%) and higher operating temperatures (1250 C) to improve the attractiveness of hightemperature gas-cooled reactors as a heat source for large industrial complexes where gas outlet temperatures of the reactor would approach 950 C.4 Of greatest concern is the influence of higher fuel temperatures and burnups on fission product interactions with the SiC layer leading to degradation of the fuel and the release of fission products. Activities are also underway around the world to examine modern recycling techniques for this fuel and to understand the ability of gas reactors to burn minor actinides.5,6
3.07.2.1.1 BR-2
The Belgian Reactor 2 (BR-2) reactor is a materials test reactor in Mol, Belgium7 that produces very fast (3.5 1014 neutrons cm2 s1 [E > 1 MeV]) and thermal neutron fluxes (1012 neutrons cm2 s1). The facilities have irradiation test rigs (15 mm ID and 400 mm long) that can be used to irradiate coated-particle gas reactor fuel forms. They have adequate flux, fluence, and temperature characterization for the capsule, and have the infrastructure needed for capsule disassembly and postirradiation examination (PIE). The capsule size precludes irradiation of pebbles; however, it could handle approximately six to eight fuel compacts.
3.07.2 Irradiation Performance 3.07.2.1 Overview of Irradiation Facilities and Testing This section provides a brief overview of irradiation facilities that are available today to perform TRISOcoated particle irradiations.
3.07.2.1.2 IVV-2M
The IVV-2M is a 15-MW water-cooled reactor that has been used in Russia for a variety of coatedparticle testing.8 Four different test rigs have been used to test specimens ranging from particles, to compacts, to spheres. The coated particle ampoule
TRISO-Coated Particle Fuel Performance
is a noninstrumented rig that can hold 10–13 graphite disks (15 mm in diameter and 2 mm thick), each of which can hold 50 particles. The rig can also hold coated particles in axial holes, 1.2 mm in diameter, and a uniform volume of coated particles, 12–18 mm in diameter and 20–255 mm high, in a graphite matrix. Another rig, termed a ‘CP hole,’ is 27 mm in diameter and that can handle six to eight capsules. A third rig, identified as ASU-8, is a 60-mm hole that can handle three compacts. The largest channel available is Vostok, which is 120 mm in diameter and contains four cells. All of these rigs can irradiate fuel at representative temperatures, burnups, and fluences for HTGRs. There is a large degree of flexibility in the testing options at IVV-2M. Their rigs can handle particles, compacts, and spheres. 3.07.2.1.3 HFR Petten
The High Flux Reactor (HFR) in Petten, Netherlands, is a multipurpose research reactor with many irradiation locations for materials testing.9 The HFR has two different types of irradiation rigs/locations in the facility: one that can accommodate compacts and another that can accommodate spheres. Rigs for spheres are multicell capsules, 63–72 mm in diameter that can handle 4–5 spheres in up to 4 separate cells. For compacts rigs/locations are 32 mm in diameter and 600 mm in useful length. They can handle three or four parallel channels of compacts. For the threechannel configuration, approximately 30 compacts could be irradiated in the rig. There is a large axial flux gradient across the useable length (40% spread maximum to minimum) that must be considered in the design of any experiment. 3.07.2.1.4 HFIR
The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is a light-water-cooled, beryllium-reflected reactor that produces high neutron fluxes for materials testing and isotope production.10 Two specific materials irradiation facilities locations are available for gas reactor fuel testing: (a) the large RB positions (eight total) that are 46 mm in diameter and 500 mm long, and can accommodate capsules holding up to 24 compacts (three in each graphite body, eight bodies axially) in a single swept cell; and (b) the small vertical experimental facility (VXF) positions (16 total) that are 40 mm in diameter and 500 mm long, and can accommodate capsules holding up to 16 compacts (eight in each graphite body, two bodies axially) in a single swept cell. Capsules can be irradiated in the lower flux small
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VXF positions and then moved to the higher flux removable beryllium positions. Neither of these positions can accommodate pebbles. A third facility, the large VXF positions (six total), are farther out in the reflector (and therefore have lower fluxes), but are 72 mm in diameter and also 500 mm long. As with the HFR, there is a large axial flux gradient that must be considered in the design of any experiment in any of these facilities. 3.07.2.1.5 ATR
The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) is a light-water-cooled, berylliumreflected reactor fuel in a four-leaf clover configuration to produce high neutron fluxes for materials testing and isotope production.11 The clover leaf configuration results in nine very high flux positions, termed ‘flux traps.’ In addition, numerous other holes of varying size are available for testing. Several positions can be used to irradiate coated-particle fuel. The 89-mm-diameter medium I position (16 total) and the 100–125-mm-diameter flux traps can accommodate pebbles. Specifically, the use of a medium I position early in the irradiation, required because of the enrichment of the fuel, followed by transfer of the test train to the northeast flux trap can provide irradiation conditions representative of a pebble bed reactor. Approximately 10–12 pebbles in five or six individually swept cells can be envisioned in the test train. The large B positions in ATR (four total) are 38 mm in diameter and 760 mm in length. They can accommodate six individually swept cells, with two graphite bodies per cell, containing up to three 2-in. long compacts per body. Thus, 36 full-size US compacts can be irradiated in this location. Of special note, here is the very flat burnup and fluence profile available axially in the ATR over the 760 mm length. This allows for nearly identical irradiation of large quantities of fuel. 3.07.2.1.6 SAFARI
The SAFARI Reactor in Pelindaba, Republic of South Africa, is an isotope production and research reactor.12 The core lattice is an 8 9 array, consisting of 28 fuel assemblies, 6 control rods, and a number of aluminum and beryllium reflector assemblies. The reactor is cooled and moderated by light water and operates at a maximum power level of 20 MW. In-core irradiation positions include six high-flux isotope production positions: two hydraulic, two pneumatic, and two fast transfer systems that are accessible during operation. Several other irradiation
156
TRISO-Coated Particle Fuel Performance
positions can also be accessed when the reactor is shut down. A large poolside facility allows for a variety of radiation applications. An intermediate storage pool and a transfer canal allow for easy and safe transport of activated materials to a hot cell. 3.07.2.1.7 TRISO-coated particle fuel irradiation testing
The historical experience in irradiation testing of coated particle fuels suggests that multicell capsules wherein fuel can be tested in separate compartments under different temperature, burnup, and fluence conditions allow for tremendous flexibility and can actually save time and money in an overall fuel qualification program. Although there are differences in details of the test trains used in each of the reactors, they share a number of important similarities in the state of the art with irradiation testing of this fuel form. In this section, these important similarities are presented to highlight the technical considerations in executing this type of testing. Because of the differences in neutron flux spectrum between a gas reactor and a light-water materials test reactor, simultaneous matching of both the rate of burnup and the rate of accumulation of fast neutron fluence is difficult to achieve. In addition, the traditional 3-year fuel cycle of high-temperature gas reactors makes real-time irradiation testing both timeconsuming and an expensive part of an overall fuel development effort. To overcome these shortcomings, irradiations in material test reactors have historically been accelerated relative to those in the actual reactor. Usually, the time acceleration is focused on achieving the required burnup in a shorter time than would be accomplished in the actual reactor, with the value of the fast fluence left as a secondary variable that must fall between a minimum and maximum value. The level of acceleration can also impact the potential for fuel failure during irradiation. The level of acceleration at a given test reactor power, coupled with fuel loading in the experiment, results in a power density for the fuel specimen in the experiment. The power density peaks at the beginning of the irradiation when the fissile content is highest and decreases as the fissile material is burned out of the fuel. As the level of acceleration increases, the temperatures in the fuel kernels increase above that in the fuel matrix because of the thermal resistances associated with the coatings of the particle,13 and the potential for high temperature, thermally driven failure mechanisms to play a deleterious role in fuel performance becomes more important.
As discussed in Section 3.07.2.7, the irradiation performance database suggests that modest levels of acceleration (1.5–3) appear to be acceptable without jeopardizing fuel performance in the irradiation, and should be a baseline requirement for future gas reactor irradiations. This acceleration level can be translated into a maximum power per fuel body or power per particle that can be used by experimenters in the design of the irradiation capsule. Given the limitations of materials test reactors around the world, the TRISO-coated particle irradiation database contains results from tests conducted under a range of accelerations. Successful German TRISO-coated particle fuel irradiations in the European HFR-Petten reactor were conducted using an acceleration of less than a factor of three. By comparison, other German irradiations in the Forschungzentrum Reaktor Juelich (FRJ) reactor at Ju¨lich had a neutron spectrum that was too thermalized. This resulted in the fuel receiving too little fast fluence to be prototypic of a high-temperature gas reactor. Similarly, historic US irradiations in ORNL’s HFIR reactor had too high a thermal flux resulting in significant burnup acceleration of the irradiation. On the basis of these considerations, the large B positions (38 mm diameter) in the ATR (see Figure 2) were chosen for the US Department of Energy’s (DOE) Advanced Gas Reactor (AGR) Program fuel irradiations because the rate of fuel burnup and fast neutron fluence accumulation in these positions provide an acceleration factor of less than three times that expected in the hightemperature gas reactor. 3.07.2.1.8 Thermal and physics analysis considerations
Given the complexity of the capsules currently being designed, the extensive review by safety authorities of the thermo-mechanical stresses, and the importance of each capsule in terms of irradiation data for fuel qualification, three-dimensional physics and thermal analyses are essential in irradiation capsule design. These analyses are critical to ensure that the fuel reaches the intended burnup, fluence, and temperature conditions. To achieve high burnups with these fuels requires detailed physics calculations to determine the time to reach full burnup. Given the concerns about severely accelerated irradiations, it is not uncommon for such irradiations to take approximately 2 years to reach full burnup in LEU TRISO-coated particles. In addition, because thermocouples should not be attached directly to the fuel, thermal analysis is used to calculate the fuel temperature during the irradiation.
TRISO-Coated Particle Fuel Performance
157
North
ON-8
ON-9 ON-10 ON-11 ON-12
ON-3
ON-4
ON-5
ON-6
ON-1
Fuel elements
ON-7 ON-2
I-19
I-20
I-1
I-2
H positions
I-3
I-4
I-18 I-17
Small B position
I-5 I-6
I-16 I-15
I-7 I-8
I-14
I-13
I-12
I-11
I-10
OS-6
OS-1
I-9
East large B position location for AGR-1
In-pile tube OS-2
OS-3
OS-4
OS-5
OS-7
OS-8
OS-9
OS-10 OS-11 OS-12
Control drum
OS-13 OS-14 OS-15 OS-16 OS-17
I positions OS-18 OS-19 OS-20 OS-21 OS-22
Figure 2 Schematic of ATR showing fuel and select irradiation positions.
Examples of a test train for fuel compacts used in INL’s ATR and the pebbles used in HFR-Petten are shown in Figures 3 and 4 respectively. These irradiation capsules have extensive instrumentation to measure temperature, burnup, and fast fluence at multiple locations in the test train. Traditional commercial thermocouples have been used extensively in past irradiations, but thermocouples can suffer from drift and/or de-calibration in the reactor. Redundancy in thermocouple measurements is another consideration in light of the low reliability of thermocouples at high temperatures and long times in neutron fields typical of TRISO-coated particle fuel irradiations. Melt wires are inexpensive and have been used as a backup to thermocouples where space was available in the capsule. However, melt wires only indicate that a certain peak temperature has been reached, and not the time of that peak.
Direct temperature measurements of the coated particles are problematic because direct metal contact (e.g., thermocouple wires or sheaths) with the fuel element is not recommended as the metals can attack the TRISO fuel coatings. Thus, temperatures must be calculated on the basis of thermocouples located elsewhere in the capsule. Thermocouples are generally located as close as possible to the fuel body to minimize the uncertainties on the calculated fuel temperatures related to irradiationinduced dimensional change and thermal conductivity changes of the materials in the capsule. Encapsulating the fuel element in a graphite sleeve or cup and inserting thermocouples into the graphite has been used successfully in many designs. The high conductivity of graphite minimizes the effect of irradiation-induced dimensional changes on the calculated fuel temperature.
158
TRISO-Coated Particle Fuel Performance
Gas line
Fuel stack
Thermocouple
SST holder
Thermocouple Purge gas pipe Radiation shields
Hafnium shield
Graphite cup Test fuel element
Capsule spacer nub
Figure 3 Schematic of capsule used in US INL AGR program.
Historically, metal sleeves have not been allowed to touch fuel elements because of past experiences in which SiC was attacked by transition metals (Fe, Cr, and Ni). Although quantitative data on transport rates of such metals into the fuel element and corrosion rates of the SiC are unknown, 2 or 3 mm thickness of graphite between the fuel element and the metallic components (e.g., graphite sleeve) has been found to be effective in minimizing the potential for interaction. These irradiation experiments typically include both thermal and fast fluence wires. A number of different fluence wires have been used successfully to measure thermal and fast neutron fluences in coated particle fuel irradiations. The specific type of wire to be used will depend on the measurement need (fast or thermal), the temperature it will experience during the irradiation, and compatibility with the material of encapsulation. Quartz encapsulation is not recommended for high-temperature, high-fluence applications. Neutronically, transparent refractories (e.g., vanadium) are a good alternative material of encapsulation. Inert gas filling of the flux wire
Figure 4 Schematic of pebble irradiation experiment used by the Germans.
encapsulation is recommended to ensure no oxygen interaction with the flux wire. Although fission chambers and self-powered neutron detectors have been used extensively in other reactor irradiations, they may not be practical in the space-constrained capsules expected for TRISO-coated particle fuel qualification tests. 3.07.2.1.9 Gas control system considerations
Automated gas control systems – designed to change the gas mixture in the experiment to compensate for the reduction in fission heat and changes in thermal conductivity with burnup – minimize human operator error and have proven to be a reliable method of thermal control during these long fuel irradiations. The temperature of each experiment capsule is controlled by varying the mixture of two gases with differing thermal conductivities in a small insulating gas jacket between the specimens and the experiment containment. A mixture of helium and argon has been used in the past and provides a wide temperature control band for the experiments. Unfortunately,
TRISO-Coated Particle Fuel Performance
argon cannot be used in fuel experiments where online fission product monitoring is used because the activated argon will reduce detectability of the system. Therefore, helium and neon are used instead. Computer-controlled mass flow controllers are typically used to automatically blend the gases (on the basis of feedback from the thermocouples) to control temperature. The gas blending approach allows for a very broad range of control. Automatic gas verification (e.g., by a thermal conductivity analyzer) has been implemented in some experiments to prevent the inadvertent connection of a wrong gas bottle. Gas purity is important and an impurity cleanup system should be implemented during each irradiation. Flow rates and gas tubing should be sized to minimize transit times between the mass-flow controllers and the experiment, as well as between the experiment and the fission product monitors. 3.07.2.1.10 FPMS considerations
In addition to thermal control, sweep gas is used to transport any fission gases released from the fuel to a fission product monitoring system (FPMS). A number of techniques have been used historically to quantify the release of fission gases from the fuel in these irradiation capsules. Techniques include gross gamma monitoring, online gamma spectroscopy, and offline gamma spectroscopy of grab samples. Online gross gamma monitoring of the effluent gas in the
experiment using ion chambers and sodium iodide detectors is an excellent means to capture any dynamic failures of the coated particles associated with the instantaneous release upon failure. Grab samples can provide excellent noble gas isotopic information. The temporal resolution and the number of isotopes that can be measured depend on the frequency of the grab samples and the delay time between acquisition of the grab sample and offline analysis. Weekly grab samples are typical in most irradiations, although daily or even hourly samples are possible if failure has occurred, assuming operation and associated analysis costs are not too high. Typical isotopes to be measured include 85mKr, 87Kr, 88 Kr, 131mXe, 133Xe, and 135Xe. Measurement of verylong-lived isotopes (e.g., 85Kr) would be useful in elucidating fission product release mechanisms from the kernel, but would also require waiting for the decay of the shorter lived isotopes in the sample. Online gamma spectroscopy, although the most expensive in terms of hardware costs, can provide the most detailed real-time information with detailed isotopic spectrums as often as necessary subject to data storage limitations of the system. An example of the system used for the US AGR program is shown in Figure 5. With such systems, transit times from the experiment to the detector should be minimized to allow measurement of short- and medium-lived isotopes, but must remain long enough to allow decay of
Temperature control gas mixing system
Vessel wall
He
Filter
6 5 Silver zeolite
4 Capsules in-core 3 2 1
Fission product monitoring system Grab sample
Figure 5 Integrated fission product monitoring system used in US AGR program irradiations.
H and V exhaust
Ne Particulate filters
159
160
TRISO-Coated Particle Fuel Performance
any short-lived isotopes associated with the sweep gases (2–3 min). With this delay time, 89Kr, 90Kr, 135m Xe, 137Xe, 138Xe, and 139Xe should also be capable of being measured online. Measurements of xenon gas-release during reactor outages are recommended to provide information on iodine release behavior from the decay of xenon precursors. Multiple options for fission gas-release measurements should be considered for long irradiations where reliability of the overall fission gas measurement system can be a concern. Redundancy is also recommended for online systems so that failure of a spectrometer does not jeopardize the entire experiment. On the basis of the online concentration data, a release-to-birth ratio (R/B), a key parameter used in reactor fuel behavior studies,14 can be calculated and provide some insight into the nature of any particle failures. Because these instruments are online during the entire irradiation, a complete time history of gas release is available. Gas release early in the irradiation (i.e., from the start of the irradiation) is indicative of initially failed particles or contamination outside of the SiC layer. Release later during the irradiation is indicative of in situ particle failure. The timing of the failure data can then be correlated to temperature, burnup, and/or fluence that can be used when coupled with PIE to determine the mechanisms responsible for the fuel failure. 3.07.2.2
German Experience
Previously, particle fuel development was conducted by German researchers in support of various HTGR designs that employed a pebble bed core. These reactors were intended to produce process heat or electricity. The sequence of fuel development used by German researchers followed improvement in particle quality and performance and was largely independent of developments in reactor technology. German fuel development can be categorized according to the sequence of fuels tested as provided in Table 1. German irradiation test conditions generally covered projected fuel operating conditions, where fuel was to reach full burnup at fast fluences of 2.4 1025 n m2 and operate at temperatures up to 1095 C for process-heat applications and up to 830 C for electrical production applications. With the exception of irradiation duration, the various experiments performed bounded expected nominal conditions or were purposely varied to meet other test objectives. In order to obtain results in a
Table 1
German particle fuel development sequence
Date of design consideration
Fuel form
1972 1977
BISO coated (Th, U)O2 Improved BISO coated (Th, U)O2 TRISO-coated UCO fissile particles with ThO2 fertile particles TRISO-coated (Th, U)O2 LEU TRISO-coated UO2
1981
timely manner, tests conducted by German researchers were generally accelerated by factors of 2–3. The following sections present irradiation experiment summaries for fuels of ‘modern’ German design.1 For these experiments, this definition extends to high-enriched (Th, U)O2 TRISO-coated particles fabricated since 1977, and low-enriched UO2 TRISOcoated particles fabricated since 1981. Table 2 provides the physical attributes of the fuel used in these tests. Mixed oxide fuel test summaries are presented first, followed by the UO2 tests. 3.07.2.2.1 R2-K12 and R2-K13
The R2-K12 and R2-K13 cells were irradiated in the R2 reactor at Studsvik, Sweden. The main objective of the R2-K12 experiment was to test mixed oxide (Th, U)O2 and fissile UC2/fertile ThO2 fuel elements, whereas for R2-K13, the main objective was to test mixed oxide (Th, U)O2 fuel elements and supply fuel for subsequent safety tests. In R2-K12, four full-size spherical fuel elements were irradiated in four independently gas-swept cells. Two cells contained mixed oxide fuel spheres, while the other two contained fissile/fertile fuel spheres. As the German researchers did not develop the twoparticle fissile/fertile system further, only the mixed oxide results were reported. R2-K13 was a combined experiment with the United States. Four independently gas-swept cells were positioned vertically on top of one another. The top and bottom cells each contained a full-size German fuel sphere. The middle two cells contained US fuel and will be discussed in Section 3.07.2.3. Configuration and irradiation data from both experiments are given in Tables 3 and 4. Cold gas tests on each fuel sphere during PIE indicated that all the particles had remained intact in both R2-K12 and R2-K13. These tests are conducted after the fuel has been stored (for 14 days) at room temperature and a quasi-steady-state release of fission gas has been reached. The fuel is then swept with a carrier gas that is monitored for various fission
TRISO-Coated Particle Fuel Performance
Table 2
161
Characteristics of modern German TRISO fuel particles
Particle batch
EUO 2308
EUO 2309
HT 354–383
EO 1607
EO 1674
Experiments irradiated in:
FRJ2-K13 FRJ2-P27 HFR-P4 HFR-K3 SL-P1 UO2 9.82 497 3% 10.81 94 41 36 40 895 1.00 [1.9] 3.20 1.88 1.053 1.019
FRJ2-P27 HFR-P4
FRJ2-K15
R2-K12 BR2-P25
R2-K13
UO2 9.82 497 3% 10.81 93 37 51 38 922 1.00 [1.9] 3.20 1.87
UO2 16.76 501 10.8% 10.85 92 14.3 38 3.4 33 1.9 41 3.8 906 28.8 1.013 [1.9] 3.20 1.88 1.029 1.020
(Th, U)O2 89.57 494 3% 10.12 85 39 37 39 888 1.09 1.93 3.20 1.93
(Th, U)O2 89.01 496 3% 10.10 89 37 33 39 890 1.06 1.90 3.19 1.90
Kernel form U enrichment (%) Kernel diameter (mm) Kernel density (g cm3) Buffer thickness (mm) IPyC thickness (mm) SiC thickness (mm) OPyC thickness (mm) Particle diameter (mm) Buffer density (g cm3) IPyC density (g cm3) SiC density (g cm3) OPyC density (g cm3) IPyC BAF OPyC BAF 235
Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990. Notes: The entries are one standard deviation. Entries in square brackets, [ ], are estimated values. BAF is the Bacon anisotropy factor for the layer, where values closer to one are more isotropic.
Table 3
R2-K12 and R2-K13 configuration
Number of cells Number of fuel spheres Spherical fuel element diameter Fuel zone diameter Fuel type Particle batch 235 U enrichment 235 U per fuel element 232 Th per fuel element Heavy metal per fuel element Number of particles per spherical fuel element Defective SiC layersa (U/U-total)
R2-K12
R2-K13
2 2 59.9 mm
2 2 59.77 mm
47 mm HEU (Th, U)O2 LTI – TRISO EO 1607 89.57% 1.002 g 4.961 g 6.076 g
47 mm HEU (Th, U)O2 LTI – TRISO EO 1674 89.01% 1.02 g 10.125 g 11.27 g
10 960
19 780
<1 105
<5 106
a Defective SiC layer fractions reported for German fuel are per pebble with the exception of loose particle experiments that are per particle batch. Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.
gases (usually 85mKr) and heated to 60 C. Sudden increases in the amount of fission gas detected indicate failed particles. The amount of increase is
proportional to the gas source, and in a calibrated system, indicates the number of failed particles. The fuel sphere from R2-K12 Cell 1 was partially deconsolidated and visual inspection revealed two kernels ‘without coating.’ Segments from each of the two fuel spheres were also metallographically examined; those examinations revealed a reaction zone on the inner side of the buffer layer, as well as tangential cracks between the buffer and the inner pyrocarbon layer. Only one particle exhibited a radial crack in the buffer layer beyond the reaction zone. All of the SiC and PyC layers examined remained intact. 3.07.2.2.2 BR2-P25
The BR2-P25 capsule was irradiated in the BR2 reactor at Mol, Belgium. The primary objective of this experiment was to test (Th, U)O2 mixed oxide fuel. One independently gas-swept cell contained 12 compacts. Each compact was cylindrical in shape and contained a small fuel sphere. Configuration and irradiation data are given in Tables 5 and 6, respectively. During PIE, Compacts 3 and 7 were electrolytically deconsolidated with no particle failures being evident. Ceramographic examination of cross-sections from Compacts 4 and 8 revealed some radial cracks in the buffer layers; however, no defective particles were found.
162 Table 4
TRISO-Coated Particle Fuel Performance
R2-K12 irradiation data
Start date End date Duration (full power days) Cell Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Center temperature ( C) Surface temperature ( C) BOL 85mKr R/B EOL (report date) 85mKr R/B
R2-K12
R2-K13
28 November 1978 12 February 1980 308 1 11.1 5.6 1100 950 3.9 109 3.0 107
22 April 1980 19 September 1982 517 1 10.2 8.5 1170 960 2.2 109 7.0 108
2 12.4 6.9 1280 1120 4.6 109 2.0 107
4 9.8 6.8 980 750 1.5 109 5.0 108
Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.
Table 5
BR2-P25 configuration
Number of cells Number of compacts Cylindrical compact diameter Cylindrical compact height Diameter of spherical fuel zone Fuel type Particle batch U enrichment 235 U per fuel compact 232 Th per fuel compact Heavy metal per fuel compact Number of particles per compact Number of particles per cell Defective SiC layers (U/U-total) 235
Table 7 1 12 26.58–27.74 mm 29.87–30.03 mm 20 mm HEU (Th, U)O2 LTI – TRISO EO 1607 89.57% 0.136 g 0.6744 g 0.8264 g 1490 17 880 <1 105
Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.
Table 6
BR2-P25 irradiation data
Start date End date Duration (full power days) Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Maximum temperature ( C) Minimum temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
30 October 1978 19 December 1981 350 13.9–15.6 6.2–8.1 1070 1010 2 107 1 106
Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.
HFR-P4 configuration
Number of cells Number of compacts per cell Cylindrical compact diameter Cylindrical compact height Diameter of spherical fuel zone Fuel type Particle batch – cells 1 and 3 Particle batch – cell 2 235 U enrichment Number of particles per compact Number of particles per capsule Defective SiC layers (U/U-total)
3 12 23–29 mm 32 mm 20 mm LEU UO2 LTI – TRISO EUO 2308 EUO 2309 9.82% 1630 19 600 <1 106
with 36- and 51-mm-thick SiC layers irradiated at 1000 C, beyond burnups of 12% fissions per initial metal atom (FIMA), and beyond fast fluences of 6 1025 n m2 (E > 0.10 MeV). The performance of the 36 mm SiC layer fuel was also to be evaluated at an irradiation temperature of 1200 C. Three independently gas-swept cells each contained 12 compacts that were cylindrical in shape and contained a small fuel sphere in each. Configuration and irradiation data are given in Tables 7 and 8, respectively. Note that the burnup and fast fluence goals were met, while the irradiation temperature goals were not. PIE revealed that the test articles remained intact. However, some failures caused by the thermocouples and gas inlet tubes were found on the upper compacts. 3.07.2.2.4 SL-P1
3.07.2.2.3 HFR-P4
The HFR-P4 capsule was irradiated at the HFR in Petten. The main objective of this experiment was to compare the fuel performance of particles
The SL-P1 experiment was irradiated at the Siloe¨ Reactor in Grenoble, France. The objective of the experiment was to test reference LEU fuel up to the potential limits for burnup and fast fluence at 800 C.
TRISO-Coated Particle Fuel Performance
Table 8
HFR-P4 irradiation data
Start date End date Duration (full power days) Capsule SiC layer thickness (mm) Maximum temperature ( C) Minimum temperature ( C) Maximum burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.10 MeV) BOL 85mKr R/B EOL 85mKr R/B
Table 9
Table 10
10 June 1982 28 November 1983 351 1 36
2 51
3 36
940
945
1075
915
920
1050
14.7
14.9
14.0
8.0
8.0
8.0
3.5 109 8 108
– 8 108
3.6 109 8 109
SL-P1 configuration
Number of cells Number of compacts Cylindrical compact diameter Cylindrical compact height Diameter of spherical fuel zone Fuel type Particle batch 235 U enrichment Number of particles per compact Number of particles per cell Defective SiC layers (U/U-total)
1 12 30.1 mm 30.8 mm 20 mm LEU UO2 LTI – TRISO EUO 2308 9.82% 1634 19 600 <1 106
One gas-swept cell contained 12 compacts. Each cylindrical compact contained one small fuel sphere. Configuration and irradiation data are provided in Tables 9 and 10, respectively. The operational objectives for this experiment were met. PIE revealed that none of the compacts showed mechanical failure. 3.07.2.2.5 HFR-K3
The HFR-K3 capsule was irradiated at the HFR in Petten. The primary objective of this experiment was to determine the performance of reference LEU fuel from an accelerated test. Four full-size spherical fuel elements were irradiated in three independently gasswept cells. The cells were vertically positioned on top of one another, with the middle cell containing two fuel spheres. To minimize flux gradient effects on the test fuel, the entire test rig was rotated 90 several times during the irradiation. Configuration
SL-P1 irradiation data
Start date End date Duration (full power days) Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Compact mean temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Table 11
24 June 1982 23 December 1983 330 8.6–11.3 5.0–6.8 743–794 5.8 107 1.2 106
HFR-K3 configuration
Number of cells Number of fuel spheres Spherical fuel element diameter Fuel zone diameter Fuel type Particle batch 235 U enrichment Number of particles per spherical fuel element Defective SiC layers (U/U-total)
Table 12
163
3 4 59.98 mm 47 mm LEU UO2 LTI – TRISO EUO 2308 9.82% 16 400 4 105
HFR-K3 irradiation data
Start date End date Duration (full power days) Cell/sphere Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Center temperature ( C) Surface temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
15 April 1982 5 September 1983 359 A/1 7.5 4.0
B/2 10.0 5.8
B/3 10.6 5.9
C/4 9.0 4.9
1200
920
920
1220
1020
700
700
1020
1 109 2 107
9 1010 1 107
9 1010 1 107
2 109 3 107
and irradiation data are given in Tables 11 and 12, respectively. Subsequent PIE reported no failures. 3.07.2.2.6 FRJ2-K13
FRJ2-K13 cells were irradiated at the DIDO reactor in Ju¨lich, Germany. The main objective of this test was to supply irradiated reference fuel for subsequent safety tests. Fuel performance was also to be examined under the controlled irradiation conditions of significant burnup with negligible fast neutron fluence. Four fullsize spherical fuel elements were irradiated in two
164
TRISO-Coated Particle Fuel Performance
independently gas-swept cells. The cells were vertically positioned on top of each other, with the fuel spheres similarly positioned within the cells. Configuration and irradiation data are given in Tables 13 and 14, respectively. Subsequent PIE reported no failures.
11 h. The 85mKr R/B ratio from each capsule increased to a maximum of 108 at the start of the transient and then dropped back to the pretransient levels after the temperature was returned to the nominal test condition. 3.07.2.2.8 FRJ2-P27
FRJ2-K15 cells were irradiated at the DIDO reactor in Ju¨lich, Germany. The main objectives of this test were to demonstrate the high burnup potential of reference fuel used in AVR reload 21-1 and to perform in-core temperature transient tests. Fuel performance was also to be examined under the controlled irradiation conditions of significant burnup with negligible fast neutron fluence. Three full-size spherical fuel elements were irradiated in three independently gas-swept cells. Configuration and irradiation data are given in Tables 15 and 16, respectively. Capsules 2 and 3 underwent a temperature transient test at a burnup of 10% FIMA. The temperature of the sphere surfaces was raised to 1100 C and held for
FRJ2-P27 cells were irradiated at the DIDO reactor in Ju¨lich, Germany. The main objectives of this test were to investigate fission product release at various cyclic temperatures and to determine the effectiveness of thicker SiC layers on the retention of 110mAg. Each of the three independently gas-swept cells contained three compacts and two coupons (trays). The compacts were cylindrical in shape and contained an (unspecified) outer fuel-free zone. The coupons were graphite disks with holes, annularly spaced, for the insertion of 34 particles. Of the two coupons that contained the thicker SiC particles (51 mm vs. 36 mm), one was placed in Cell 1, and the other in Cell 3. Configuration and irradiation data are provided in Tables 17 and 18, respectively.
Table 13
Table 15
3.07.2.2.7 FRJ2-K15
FRJ2-K13 configuration
Number of cells Number of fuel spheres Spherical fuel element diameter Fuel zone diameter Fuel type Particle batch 235 U enrichment Number of particles per spherical fuel element Defective SiC layers (U/U-total)
Table 14
2 4 59.98 mm 47 mm LEU UO2 LTI – TRISO EUO 2308 9.82% 16 400 4 105
FRJ2-K13 irradiation data
Start date End date Duration (full power days) Cell/sphere Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Center temperature ( C) Surface temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Number of cells Number of fuel spheres Spherical fuel element diameter Fuel zone diameter Fuel type Particle batch 235 U enrichment Number of particles per spherical fuel element Defective SiC layers (U/U-total)
Table 16
24 June 1982 12 February 1984 396 A/1 7.5 0.2
A/2 8.0 0.2
B/3 7.9 0.2
B/4 7.6 0.2
1125
1150
1150
1120
985
990
990
980
2 109 2 108
2 109 2 108
8 1010 7 109
8 1010 7 109
FRJ2-K15 configuration 3 3 60.04 mm 47 mm LEU UO2 LTI – TRISO HT 354–383 16.76% 9500 <5 105
FRJ2-K15 irradiation data
Start date End date Duration (full power days) Cell Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Center temperature ( C) Surface temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
4 September 1986 20 May 1990 590 1 14.1 0.2
2 15.3 0.2
3 14.7 0.1
970
1150
990
800
980
800
2.0 1010 1.0 108
2.47 1010 5.0 109
2.0 1010 3.0 109
TRISO-Coated Particle Fuel Performance
Table 17
FRJ2-P27 configuration
Number of cells Number of compacts per cell Number of coupons per cell Cylindrical compact diameter Cylindrical compact height Coupon diameter Coupon height Diameter of coupon fuel annulus Fuel type Particle batch for compacts and four coupons Particle batch for two coupons (thick SiC) 235 U enrichment Number of particles per compact Number of particles per coupon Number of particles per cell Defective SiC layers (U/U-total)
Table 18
Table 19 3 3 2 27.9–28.03 mm 29 mm 27 mm 2.2 mm 23 mm LEU UO2 LTI – TRISO EUO 2308 EUO 2309 9.82% 2424 34 7340 <3 106
17 February 1984 10 February 1985 232 1 7.6 1.4
2 8.0 1.7
3 7.6 1.3
1080
1320
1130
880
1220
1080
6
1.0 10 1.6 106
7
8.6 10 1.0 105
U enrichment Number of particles per spherical fuel element
HFR-K6
HFR-K5
4 4 60 mm
4 4 60 mm
LEU UO2 – TRISO 10.6% 14 600
LEU UO2 – TRISO 10.6% 14 600
were irradiated in four independently gas-swept cells. A typical reactor temperature history was simulated in the test with 17 temperature cycles (corresponding to 17 passes through the core). For one-third of a cycle, the fuel sphere center temperature was held at 800 C; for the other two-thirds of the cycle, the center temperature was 1000 C. In addition, three temperature transients (sphere center temperature held at 1200 C for 5 h) were performed at beginning of life (BOL), middle of life (MOL), and end of life (EOL). Limited configuration and irradiation data are given in Tables 19 and 20, respectively. There were no particle failures reported as a result of the irradiations.
FRJ2-P27 irradiation data
Start date End date Duration (full power days) Cell Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.10 MeV) Center temperature ( C) Surface temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
HFR-K6 and HFR-K5 configurations
Number of cells Number of fuel spheres Spherical fuel element diameter Fuel type 235
165
3.07.2.3 8
2.0 10 1.2 107
PIE revealed that all specimens and components were in excellent condition. Cold gas tests of all compacts and coupons determined that there was only one defective/failed particle present. This particle was from a Capsule 2 coupon (with nominal SiC thickness). Ceramographic examination revealed that the particle was inserted in the coupon ‘without coating’ and that kernel interactions led to a compression of the inner side of the buffer to a thickness of 10 mm. 3.07.2.2.9 HFR-K6 and HFR-K5
The HFR-K6 and HFR-K5 capsules were irradiated at the HFR in Petten.1,9 These experiments were a proof test for HTR MODUL reference fuel. In each experiment, four full-size spherical fuel elements
US Experience
Historical US particle fuel development effort (through the mid 1990s), which included design and testing, coincided with the development of various HTGRs. This sequence of development is listed in Table 21, and identifies the main fuel forms under consideration at that time. US gas reactors were designed to use prismatic graphite blocks containing fuel compacts, and were primarily intended to produce electricity with the exception of the New Production Modular High-temperature Gas-Cooled Reactor, which was designed to produce tritium. Over the years, the design has also supported steam cycle, direct cycle, process heat, and weapons material disposition applications. More recently, DOE established the AGR Fuel Development and Qualification Program to provide a baseline fuel qualification data set at a peak fuel centerline temperature of 1250 C15,16 in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Irradiation test conditions employed by the United States generally covered projected fuel operating conditions. US fuel was to operate at temperatures as
166
TRISO-Coated Particle Fuel Performance
Table 20
HFR-K6 and HFR-K5 irradiation data
Start date End date Duration (full power days) Cell Burnup (% FIMA) Fast fluence (1025 m2, E > 0.10 MeV) Temperature EOL 85mKr R/B
Table 21
HFR-K6
HFR-K5
1990 4 May 1993 634
1991 16 May 1994 564
1 7.2 3.2
2 9.3 <4.8
3 9.7 4.8
4 9.2 <4.8
1 6.7 2.9
2 8.8 <4.3
3 9.1 4.3
4 8.7 <4.3
Cycled 3 107
Cycled 3 107
Cycled 3 107
Cycled 3 107
Cycled 3 107
Cycled 3 107
Cycled 3 107
Cycled 3 107
Historical US particle fuel development and testing sequence
Date of design conception
Reactor/status
Major fuel form tested
1960 1964
Peach Bottom built Fort St. Vrain built
1967
LHTGR design only
1984
NE-MHTGR commercial design only
1989 1995
NP-MHTGR government design only GT-MHR commercial design only
2005
NGNP design
BISO-coated (Th, U)C2 TRISO-coated (Th, U)C2 fissile TRISO-coated ThC2 fertile TRISO-coated UC2 fissile BISO and TRISO-coated ThO2 fertile TRISO-P coated UCO fissile TRISO-P coated ThO2 fertile TRISO-P coated UCO TRISO-coated UCO fissile TRISO-coated UCO and/or UO2 fertile fuel not yet tested TRISO-coated UCO fissile
high as 1400 C and reach full burnup (commensurate with 235U enrichment and kernel composition) at fast fluences of 4 1025 n m2. With the exception of irradiation duration, the various experiments performed either bounded expected nominal conditions or were purposely varied to meet other test objectives. In order to obtain results in a timely manner, US tests were accelerated by factors of 3–10. The particle fuel irradiation experiments and PIE results summarized below consider only selected tests of key US fuel types. These fuel types include TRISO fissile/BISO fertile particles, weak acid resin (WAR) TRISO fissile/BISO fertile particles, TRISO fissile/TRISO fertile particles, and TRISO-P fissile particles (conventional TRISO-coated particles with an additional ‘protective’ pyrolytic carbon layer above the outer pyrolytic carbon layer) as well as TRISO fissile particles. General Atomics and Babcock & Wilcox manufactured the majority of the kernel and coating batches. However, some of the batches were manufactured by ORNL. The following US
experiment summaries are listed in chronological order and are not grouped by fuel type. Listed configuration and irradiation data are actual values, not specification values or ranges. Interpretations of PIE results are from the original sources and no overt attempt has been made to reinterpret the results. 3.07.2.3.1 F-30
The F-30 experiment was irradiated in the General Electric Test Reactor (GETR) at Pleasanton, California.17 The primary objective of this experiment was to demonstrate the irradiation performance of Fort St. Vrain production fuel. Five independently gas-swept cells contained the fuel. Cells 1, 3, and 4 contained only fuel compacts, Cell 2 contained only loose particles, and Cell 5 contained both fuel compacts and loose particles. Configuration and irradiation data are given in Tables 22 and 23, respectively. Postirradiation metallographic examination of seven fuel compacts containing fissile and fertile particles was performed. In addition, five sets of
TRISO-Coated Particle Fuel Performance
167
loose fissile particles and five sets of loose fertile particles were examined. Fissile particle failure, defined as a crack completely through the SiC layer, ranged between 0% and 6.1%, while fertile particle failure ranged between 0% and 15.1%. A typical photomicrograph of SiC failure in an F-30 fissile particle is presented in Figure 6. Metallography revealed that inner pyrolytic carbon layers had remained bonded to the SiC layer throughout irradiation. Figure 7 displays a typical photomicrograph of a fissile particle with an IPyC layer crack and a densified buffer. 50 mm
3.07.2.3.2 HRB-4 and HRB-5
The HRB-4 and HRB-5 capsules were irradiated in HFIR at ORNL.18 The main objective of these experiments was to test candidate fuel materials and manufacturing processes for the proposed large HTGR. Each test involved a single gas-swept cell containing six fuel compacts vertically positioned.
Table 22
F30 configuration
Number of cells Total number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact lengths Fissile fuel type Nominal Th/U ratio 235 U enrichment Fissile particle diameter Fertile fuel type Fertile particle diameter Number of fissile particle batches Number of fertile particle batches Defective SiC layer fraction – fissile particles Defective SiC layer fraction – fertile particles
Table 23
Figure 6 A typical SiC layer crack in an F-30 fissile fuel particle. Reproduced from Scott, C .B.; Harmon, D. P. Post Irradiation Examination of Capsule F-30; GA-A13208, UC-77; General Atomics Report, 1975.
5 13 12.45 mm 18.54 and 49.28 mm HEU (Th, U)C2 TRISO 4.25 93% 429–560 mm ThC2 TRISO 648–771 mm 7 9 <5 104–1.5 103 3 104–1.0 103
100 mm Figure 7 A typical IPyC layer crack in a fissile F-30 fuel particle. Reproduced from Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsule F-30; GA-A13208, UC-77; General Atomics Report, 1975.
F30 irradiation data
Start date End date Duration (full power days) Cell Fissile burnup (% FIMA) Fertile burnup (% FIMA) Fast fluence (1025 n m2, E > 0.18 MeV) Time average peak temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
15 May 1972 5 April 1973 269 1 15 3 8 1100 2 106 8 106
2 19 4.5 10.5 1100 7 107 1 104
3 20 5 11.5 1120 8 107 1 105
4 18 4 9.5 1100 7 107 2 105
5 12 1.5 12 1200 2 106 2 105
168
TRISO-Coated Particle Fuel Performance
Table 24
HRB-4 and HRB-5 configurations
Number of cells Number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact lengths Fissile fuel type Fertile fuel type U enrichment Fissile particle diameter Fertile particle diameter Fissile particle batch Fertile particle batch Total number of fissile particles Total number of fertile particles
235
Table 25
HRB-4
HRB-5
1 6 12.4 mm
1 6 12.4 mm
25.4 mm
25.4 mm
WAR UC2 TRISO ThO2 BISO 5.99% 639 mm 805 mm OR52A T01424BIL 17 780
WAR UC2 TRISO ThO2 BISO 5.99% 639 mm 805 mm OR52A T01424BIL 17 780
4180
4180
Figure 8 Typical HRB-4 fissile particle irradiated to 27.7% FIMA and 10.5 1025 n m2 fast fluence. Reproduced from Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsules HRB-4, HRB-5, and HRB-6; GA-A13267, UC-77; General Atomics Report, 1975.
HRB-4 and HRB-5 irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Peak temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
HRB-4
HRB-5
8 October 1972 26 June 1973 244
8 October 1972 3 February 1973 107
27.7
15.7
13.4
4.3
10.5
4.7
1250 1.4 105 3.2 104
1250 3 106 1 104
Configuration and irradiation data are given in Tables 24 and 25. Metallographic examinations were performed on each fuel compact. A typical photomicrograph of an irradiated HRB-4 fissile particle is presented in Figure 8, which shows the formation of gas bubbles in the kernel and the densification of the buffer. IPyC layers of the examined fissile particles had remained bonded to the SiC. The examination indicated that the fissile particles had failed between 0% and 6% of the SiC layers. These failures consisted primarily of radial cracks through the SiC layer. Between 4% and 73% of
Figure 9 Photomicrographs of typical fission product attack in irradiated HRB-4 fissile particles. Reproduced from Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsules HRB-4, HRB-5, and HRB-6; GA-A13267, UC-77; General Atomics Report, 1975.
the OPyC layers failed during irradiation. There were no tabulations of IPyC layer failures reported. Several of the fissile particles examined displayed evidence of fission product attack. This attack mostly occurred in large concentrations at the IPyC–SiC interface and where fission products in smaller concentrations had diffused up to 25 mm into the SiC. Figure 9 presents typical photomicrographs of fission product attack in HRB-4 fissile particles. In HRB-5, IPyC layers of the examined fissile particles had remained bonded to the SiC. There
TRISO-Coated Particle Fuel Performance
Table 26
HRB-6 configuration
Number of cells Number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact length Fissile fuel type Nominal Th/U ratio 235 U enrichment Fissile particle diameter Fertile fuel type Fertile particle diameter Fissile particle batch Fertile particle batch Defective SiC layer fraction – fissile particles
Figure 10 Typical HRB-4 fissile particle irradiated to 27.7% FIMA and 10.5 1025 n m2 fast fluence. Reproduced from Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsules HRB-4, HRB-5, and HRB-6; GA-A13267, UC-77; General Atomics Report, 1975.
were no tabulations of IPyC layer failures reported. There was no evidence of fission product attack as seen in the HRB-4 fissile particles. However, the examination indicated that between 0.4% and 17% of the SiC layers in fissile particles had failed. These failures consisted primarily of radial cracks through the SiC layer. A typical photomicrograph of irradiated HRB-5 fissile particles with cracked SiC layers is presented in Figure 10. This photomicrograph is also representative of HRB-4 fissile particles with cracked SiC layers. It was reported that a large fraction of these cracked SiC layers were due to metallographic preparation and not a result of fast neutron exposure or fuel burnup effects. 3.07.2.3.3 HRB-6
The HRB-6 capsule was irradiated in HFIR at ORNL.18 Fissile fuel particles used in HRB-6 came from the same production batch as used in the first core of Fort St. Vrain and were one of the batches previously irradiated in the F-30 experiment. This test involved a single gas-swept cell containing six fuel compacts vertically positioned. During operation, the sweep gas flow rate was reduced because of high activity in the sweep lines. Because of this gas flow reduction, in-pile fission gas-release data were not obtained. The irradiation of HRB-6 in HFIR coincided with part of the HRB-4 irradiation. Configuration and irradiation data are given in Tables 26 and 27.
Table 27
169
1 6 12.4 mm 25.4 mm HEU (Th, U)C2 TRISO 4.25 93.15% 556 mm ThO2 BISO 888 mm CU6B-2427 T01451BIL-W <5 104
HRB-6 irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Peak temperature ( C) Minimum TRIGA BOL 85mKr R/B Maximum TRIGA EOL 85mKr R/B
27 February 1973 8 September 1973 183 26.6 9.3 7.9 1100 5.0 107 2.7 104
PIE included gas-release measurements of each fuel compact performed in the Training Research and Isotope Production, General Atomics (GA) (TRIGA) reactor. However, during the unloading of the HRB-6 capsule, fuel compacts 2A, 2B, and 2C were damaged and as many as 30 broken fuel particles were observed. Therefore, the TRIGA gasrelease measurements at EOL for these compacts would be higher than in-pile sweep line measurements had they been performed. A typical photomicrograph of an irradiated HRB-6 fissile particle is presented in Figure 11, which shows the formation of gas bubbles in the kernel and densification of the buffer. The photomicrograph also shows an incipient crack in the IPyC layer. No tabulations of IPyC layer failures were reported. IPyC layers of the examined fissile particles had remained bonded to the SiC, and there was no evidence of fission product attack. However, the examination indicated that the fissile particles had failed between 0% and 2% of the SiC layers. These failures do not include the fissile particles broken during capsule unloading. It was reported that a
170
TRISO-Coated Particle Fuel Performance
Table 28
OF-2 configuration
Number of cells Total number of fuel compacts Cell 1 cylindrical fuel compact dimensions (16 compacts) Cell 2 cylindrical fuel compact dimensions (48 compacts) Cell 2 cylindrical fuel compact dimensions (24 compacts) Fissile fuel type
235
Figure 11 Typical HRB-6 fissile particle irradiated to 26.5% FIMA and 7.9 1025 n m2 fast fluence. Reproduced from Scott, C. B.; Harmon, D. P. Post Irradiation Examination of Capsules HRB-4, HRB-5, and HRB-6; GA-A13267, UC-77; General Atomics Report, 1975.
U enrichment Fissile particle diameter Fertile fuel type Fertile particle diameter Number of fissile particle batches Number of fertile particle batches
Table 29
large fraction of these failures were due to metallographic preparation. 3.07.2.3.4 OF-2
The OF-2 capsule was irradiated in the Oak Ridge Research Reactor (ORR).19 The main objectives of the test were to investigate the irradiation performance of various particle fuel forms (mostly WAR UCO with different stoichiometries) and to compare the performance of fuel particles fabricated from different coaters. OF-2 consisted of 88 fuel compacts (and several sets of loose inert particles) contained in a single capsule that was divided into two independently gas-swept cells. Various combinations from 15 fissile batches, 16 fertile batches, and 4 compact matrix compositions comprised the fuel compacts (each compact contained fuel from only one fissile batch and one fertile batch). Configuration and irradiation data are given in Tables 28 and 29. Postirradiation metallography was performed on three fuel compacts from Cell 1 and on 27 fuel compacts from Cell 2. A significant level of OPyC layer failures was observed in the fissile TRISOcoated particles from Cell 1. However, there were no observed SiC layer failures or any layer failures in the BISO-coated fertile and inert particles in these compacts. Examination of 11 fuel compacts from Cell 2, containing the same three fissile particle batches as in Cell 1, also indicated significant levels of OPyC layer failures. The fissile particle batch with the highest OPyC anisotropy (optical Bacon
2 88 15.75 mm diameter, 25.4 mm long 15.75 mm OD, 3.30 mm ID, 12.70 mm long 15.75 mm diameter, 50.8 mm long WAR UCxOy TRISO (Th, U)O2 TRISO UC2 TRISO Not reported 600–753 mm ThO2 BISO 806–889 mm 15 16
OF-2 irradiation data
Start date End date Duration (full power days) Cell Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.18 MeV) Maximum temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
21 June 1975 1 August 1976 352 1 2 75.9–79.6 50.0–79.5 5.86–8.91 1.94–8.36 1350 2 105 1 104
1350 1 104 5 106
anisotropy factor (BAF) ¼ 1.069) had 100% OPyC layer failure, while the other two batches with lower anisotropy (optical BAF of 1.035 and 1.030) had 0–33% OPyC layer failures. Of the 30 fuel compacts metallographically examined, only one compact (that contained WAR UCO fissile particles) displayed cracked SiC layers. Among the 27 fissile particles observed in this compact, 16 displayed cracked SiC layers. These cracks were identified as artifacts of polishing. However, no photomicrographs of these cracks were presented to support this conclusion. The metallographic examinations also revealed typical WAR UCO behavior of kernel and buffer densification. This densification was also accompanied by varying degrees of kernel migration. Figure 12 presents a typical WAR UCO photomicrograph that displays kernel and buffer densification, and OPyC layer failure. Examination of OF-2 particles also indicated several incidences of fission
TRISO-Coated Particle Fuel Performance
product accumulation at the IPyC and SiC interface. A typical photomicrograph of fission product accumulation is presented in Figure 13.
Figure 12 Photomicrograph of irradiated OF-2 fissile WAR UCO particle. Reproduced from Tiegs, T. N.; Thoms, K. R. Operation and Post Irradiation Examination of ORR Capsule OF-2: Accelerated Testing of HTGR Fuel; ORNL-5428; 1979. Courtesy of Oak Ridge National Laboratory, U.S. Department of Energy.
3.07.2.3.5 HRB-14
The HRB-14 capsule was irradiated in HFIR at ORNL.20 The main objectives of this experiment were to test LEU particles and to demonstrate reduced matrix–OPyC layer interactions by using cure-in-place fuel compacts. This test involved a single gas-swept cell equally divided among 20 fuel compacts vertically positioned and molded planchets (wafers) containing BISO-coated ThO2 fertile particles. Online fission gas-release measurements were not reported. Also, irradiation results from the BISOcoated fertile particles were reported separately and are not included in this summary. Configuration and irradiation data are given in Tables 30 and 31. Disassembly of the HRB-14 capsule after irradiation produced five fuel compacts with no remaining structure; in essence, there were five collections of loose particles, four compacts that were partially intact, nine compacts that were intact but displayed significant amounts of debonding, and only two compacts in relatively good shape.
Table 30
Lower half of HRB-14 configuration
Number of cells Total number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact length Fissile fuel type
235
U enrichment Fissile particle diameter Fertile fuel type Fertile particle diameter Number of fissile particle batches Number of fertile particle batches Defective SiC layer fraction – fissile particles
Table 31
Figure 13 Photomicrograph of irradiated OF-2 fissile fuel particles displaying fission product accumulation at IPyC–SiC interface. Reproduced from Tiegs, T. N.; Thoms, K. R. Operation and Post Irradiation Examination of ORR Capsule OF-2: Accelerated Testing of HTGR Fuel; ORNL-5428; 1979. Courtesy of Oak Ridge National Laboratory, U.S. Department of Energy.
171
1 20 12.50 mm 9.52 mm UCxOy TRISO (Th, U)O2 TRISO UO2 TRISO 19.18–19.66% 760–813 mm ThO2 TRISO 786–882 mm 5 8 7.0 107– 1.3 104
Lower half of HRB-14 irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Maximum temperature ( C) Minimum temperature ( C) Minimum TRIGA BOL 85mKr R/B Maximum TRIGA EOL 85mKr R/B
20 May 1978 4 January 1979 214 28.6 8.5 8.3 1190 895 3.8 107 3.0 104
172
TRISO-Coated Particle Fuel Performance
Metallographic examination was performed on 15 fuel compacts, and 8 of them contained fissile particles. A few fissile particles were reported to have SiC layer cracks but these cracks were attributed to metallographic preparation. It should be noted that visual inspection of each compact during capsule disassembly indicated that between 0% and 9% of the visible particles (from compact surfaces and loose particles that had fallen off) had failed SiC layers. However, this visual inspection did not distinguish between fissile and fertile particles. The metallographic examination of fissile particles revealed that between 0% and 3% of the IPyC layers had failed (cracked) and that the IPyC layers had debonded from the SiC in 0% to 7.7% of the particles. Buffer layers did not crack in the UO2 or (Th, U)O2 fuel but did crack in 10–71% of the UCO fuel particles. Kernel extrusion was reported only in UCO fuel. Figure 14 displays typical kernel extrusion, and Figure 15 presents a typical photomicrograph of kernel migration. In several particles of each fuel form, high concentrations of fission products were observed in small, localized regions at the SiC–IPyC layer interface. In addition to fission product accumulation, localized chemical attack was also observed in the SiC layers of several (Th, U)O2 and UO2 fuel particles. This localized attack, which had penetrated 2 mm into the SiC, was attributed to palladium, and was
Figure 14 Photomicrograph of a UCO particle (batch 6157-08-020) from Compact 10 irradiated at 1040 C to 27.8% FIMA and to a fast fluence (E > 0.18 MeV) of 7.1 1025 n m2 displaying kernel extrusion. Reproduced from Young, C. A. Pre- and Post Irradiation Evaluation of Fuel Capsule HRB-14; GA-A15969, UC-77; General Atomics Report, 1980.
observed in 8% of the particles. UCO fuel particles that did not display localized chemical attack, had uniform attack along the inner SiC layer (usually on one side of the particles). This uniform attack was attributed to rare earth fission products. Figure 16 displays typical uniform fission product attack in a UCO fuel particle. It should be noted that with optimized UCO stoichiometry, the kernel retains rare earth fission products and does not display kernel migration as found here with non-optimized UCO kernels containing excess UC2 leading to rare earth migration. Metallographic examination of fertile particles indicated that between 0% and 2.4% of the particles in each compact had total coating failure, defined as cracked OPyC and SiC layers. These failures were attributed to pressure vessel failure. Figure 17 displays a typical failed fertile particle. Separate tallies of particles where only the SiC layer had failed were not reported. Other fertile particle observations include the following: 1.5–29.1% of the particles had failed OPyC layers 8–70% of the particles had failed IPyC layers 11–85% of the particles had IPyC layers debonded from the SiC 6–26% of the particles had cracked buffers no kernel migration was observed a few kernels had extruded into buffer cracks.
Figure 15 Photomicrograph of a UCO particle (batch 6157-08-020) from Compact 10 irradiated at 1040 C to 27.8% FIMA and to a fast fluence (E > 0.18 MeV) of 7.1 1025 n m2. Reproduced from Young, C. A. Pre- and Post Irradiation Evaluation of Fuel Capsule HRB-14; GA-A15969, UC-77; General Atomics Report, 1980.
TRISO-Coated Particle Fuel Performance
Table 32
HRB-15B configuration
Number of cells Total number of particle trays Maximum number of loose particles per tray Particle tray outer diameter Particle tray inner diameter Fissile fuel type
235
U enrichment Fissile particle diameter Fertile fuel type Figure 16 Photomicrograph of a UCO particle (batch 6157-08-020) from Compact 10 irradiated at 1040 C to 27.8% FIMA and to a fast fluence (E > 0.18 MeV) of 7.1 1025 n m2 displaying fission product attack of the SiC layer. Reproduced from Young, C. A. Pre- and Post Irradiation Evaluation of Fuel Capsule HRB-14; GA-A15969, UC-77; General Atomics Report, 1980.
Fertile particle diameter Number of fissile particle batches Number of fertile particle batches
The primary objective of the HRB-15B experiment irradiated in HFIR at ORNL21 was to test a variety of LEU fissile fuel designs and ThO2 fertile particle designs. This test involved a single gas-swept cell
22.3–23.6 mm 11.1 mm UCO TRISO and silicon-BISO (Th, U)O2 TRISO and silicon-BISO UC2 TRISO and silicon-BISO UO2 TRISO and silicon-BISO UO2* TRISO and silicon-BISO 19.5% 742–951 mm ThO2 TRISO, BISO and silicon-BISO 773–836 mm 19 22
HRB-15B irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Time average temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
3.07.2.3.6 HRB-15B
1 184 116
Note: Two types of UO2* fuel were tested, one with ZrC dispersed in the buffer and the other with pure ZrC layer around the kernel.
Table 33
Figure 17 Photomicrograph of a ThO2 fertile particle (batch 6252-17-010) irradiated at 1130 C to 8.5% FIMA and to a fast fluence (E > 0.18 MeV) of 8.3 1025 n m2 displaying pressure vessel failure. Reproduced from Young, C. A. Pre- and Post Irradiation Evaluation of Fuel Capsule HRB-14; GA-A15969, UC-77; General Atomics Report, 1980.
173
6 July 1978 4 January 1979 169 26.7 6.0 6.6 815–915 2.9 108 5.1 106
containing 184 thin graphite trays. Each tray could accommodate up to a maximum of 116 individual, unbonded fuel particles. The loose fissile fuel particles included UC2, UCO with four different stoichiometries, (Th, U)O2, UO2, and two types of UO2* (one type had ZrC dispersed throughout the buffer layer and the other had a pure ZrC coating around the kernel). Each fissile fuel type was tested with both TRISO coating and silicon–BISO coating which consisted of the kernel surrounded by a buffer layer, an IPyC layer, and finally a silicon doped OPyC layer. The loose fertile particles tested included TRISO-, BISO-, and silicon–BISO-coated ThO2. Configuration and irradiation data are provided in Tables 32 and 33.
174
TRISO-Coated Particle Fuel Performance
Postirradiation metallography was performed on 20 different particle types, each consisting of approximately 20 particles. These examinations revealed considerable gas bubble formation in UC2 and UCO kernels, and buffer densification in TRISO-coated particles. Some SiC layer cracking was observed in each TRISO-coated fuel type, but mostly in the UCO particles. These cracks were reported to have occurred during mount preparation because of the crack orientation and because the visual examination detected no OPyC cracking. No further tabulation of layer failures was reported. 3.07.2.3.7 R2-K13
The R2-K13 capsule was irradiated in the R2 reactor at Studsvik, Sweden.22 The main objective of this experiment was to test reference UCO fissile particles and ThO2 fertile particles. Four independently gas-swept cells were positioned vertically on top of one another. The middle two cells contained US fuel. The top and bottom cells each contained a full-size German fuel sphere (discussed in the section on German irradiation results). Configuration and irradiation data are given in Tables 34 and 35. Postirradiation metallographic examination was performed on two fuel compacts. All of the 99 fissile particles examined displayed debonding between the buffer and IPyC layers. In some cases, debonding between the buffer, IPyC, and SiC layers was also observed. Likewise, all of the 68 fertile particles examined displayed debonding between the buffer, IPyC, and SiC layers. The SiC layers of all the particles examined were observed to be intact. Table 34
R2-K13 US configuration
Number of cells Total number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact length Total number of piggyback sample sets Fissile fuel type Fertile fuel type 235 U enrichment Fissile particle diameter Fertile particle diameter Fissile particle batches Fertile particle batches Defective SiC layer fraction – fissile particles Defective SiC layer fraction – fertile particles
2 12 12.52 mm 25.4 mm 31 LEU UCO TRISO ThO2 TRISO 19.61% 803 and 824 mm 781–805 mm 2 3 1.9 104 and 4.4 104 <2 106– 1.6 105
3.07.2.3.8 HRB-15A
The main objective of the HRB-15A experiment irradiated in HFIR at ORNL23 was to test several candidate fuel designs for the proposed Large High Temperature Gas Reactor (LHTGR). This test involved a single gas-swept cell containing 20 cylindrical fuel compacts positioned vertically on top of one another. Interspersed between the fuel compacts were 17 tray assemblies. Each assembly had a graphite tray holding 54 unbonded particles in separate holes, and serving as a lid, a graphite wafer containing 54 particles bonded in separate holes with carbonaceous matrix material. Configuration and irradiation data are given in Tables 36 and 37. Table 35
R2-K13 US irradiation data
Start date End date Duration (full power days) Cell Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average center temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Table 36
22 April 1980 19 September 1982 517 2 3 22.5 22.1 4.6 4.5 7.8 7.4 1190 1 105 8 105
HRB-15A configuration
Number of cells Total number of fuel compacts Cylindrical fuel compact diameter Number of short fuel compacts/length Number of long fuel compacts/length Number of bonded wafer/unbonded tray assemblies Fissile fuel type
Fertile fuel type 235
985 2 107 8 106
U enrichment Fissile particle diameter Fertile particle diameter Fissile particle batches Fertile particle batches Defective SiC layer fraction – fissile particles Defective SiC layer fraction – fertile particles
1 20 12.54 mm 3/9.53 mm 17/19.05 mm 17 UCO TRISO UC2 TRISO UC2 ZrC-TRISO UO2 TRISO UO2 ZrC-TRISO UO2* ThO2 TRISO ThO2 silicon-BISO 19.5% 736–894 mm 713–1014 mm 10 5 1.4 105– 7.4 102 6.7 105– 1.4 103
Note: Two types of UO2* fuel were tested, one with ZrC dispersed in the buffer and the other with pure ZrC layer around the kernel.
TRISO-Coated Particle Fuel Performance
Table 37
HRB-15A irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average center temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Table 38
175
HRB-16 configuration
26 July 1980 29 January 1981 174 29.0 6.4 6.5
Number of cells Total number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact length Number of loose particle trays Number of particles per tray
1150 6.96 106 3.76 104
Fissile fuel type
Fertile fuel type 235
U enrichment Fissile particle diameter Fertile particle diameter Fissile particle batches Fertile particle batches Defective SiC layer fraction – fissile particles Defective SiC layer fraction – fertile particles
1 18 12.45 mm 18.70 mm 27 110 (2 particles per hole) UCO TRISO UCO ZrC-TRISO UC2 TRISO UC2 ZrC-TRISO UO2 TRISO UO2* TRISO (Th, U)O2 TRISO ThO2 TRISO ThC2 BISO 19.20–19.61% 742–884 mm 756 and 786 mm 9 2 4.6 107– 4.4 104 1.6 105 and 5.0 104
Note: Two types of UO2* fuel were tested, one with ZrC dispersed in the buffer and the other with pure ZrC layer around the kernel.
Figure 18 Photomicrograph of a UO2 ZrC–TRISO-coated particle (batch 6162-00-010) irradiated at 1075 C to 27.2% FIMA and to a fast fluence of 6.0 1025 n m2 (E > 0.18 MeV) displaying ZrC layer cracks. Reproduced from Ketterer, J.; et al. Capsule HRB-15A Post Irradiation Examination Report; GA-A16758, UC-77; General Atomics Report, 1984.
Postirradiation metallographic examination was performed on five fuel compacts. Between 0% and 5.6% SiC (and OPyC) layer failures were reported for the UO2 particles but were attributed to sample preparation. In contrast, the ZrC layer failures observed in the UO2 ZrC–TRISO-coated particles were also attributed to sample preparation but were not tabulated. A photomicrograph of a UO2 ZrC– TRISO-coated particle displaying a cracked ZrC layer is presented in Figure 18. No SiC layer failures were reported for the UCO fuel. Between 0% and 12.5% of the SiC layers and between 83% and 92% of the IPyC layers were
reported to have failed in the fertile particles. These high layer failures for the fertile ThO2 particles were attributed to the high IPyC BAF values for these particles. The high BAF was a result of intentionally depositing the IPyC layer at low coating rates in an attempt to produce layers that were impermeable to chlorine (chlorine trapped in the particle during SiC deposition may enhance SiC degradation during irradiation). 3.07.2.3.9 HRB-16
The main objective of the HRB-16 experiment conducted in the HFIR at ORNL24 was to test a variety of LEU fissile particle fuel designs. This test involved a single gas-swept cell containing 18 fuel compacts stacked vertically and interspersed with 27 trays of unbonded particles and several encapsulated fission product piggyback transport specimens. Configuration and irradiation data are given in Tables 38 and 39. Postirradiation metallographic examination was performed on seven fuel compacts that contained particles from six different fissile batches and one fertile batch. For fuel compacts containing multiple fissile batches, the following visual criteria were used to identify fuel forms: UO2* had the conspicuous, bright ZrC layer next to the kernel
176
TRISO-Coated Particle Fuel Performance
Table 39
HRB-16 irradiation data
Start date End date Duration (full power days) Peak fissile burnup (% FIMA) Peak fertile burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average center temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Table 40 21 June 1981 23 December 1981 170 28.7 6.1 6.3 1150 2.44 105 2.08 104
Figure 19 Photomicrograph of a UO2 particle (batch 6152-04-010) irradiated at 1100 C to 26.9% FIMA and to a fast fluence of 5.61 1025 n m2 (E > 0.18 MeV) displaying kernel migration. Reproduced from Ketterer, J. W.; Myers, B. F. Capsule HRB-16 Post Irradiation Examination Report; HTGR-85-053, 1985.
UC2 had very small gas bubbles (voids) in the kernel, or if present in larger form were very irregular in shape UCO had medium size, mostly circular voids in the center of the kernel and small voids at the periphery of the kernel UO2 had large, mostly circular voids evenly distributed throughout the kernel. The metallographic examinations revealed that only the UO2 particles displayed kernel migration. Kernel migration was observed in approximately 28% of the UO2 particles in fuel compacts 2 and 13 and in 60% of the UO2 particles in compact 14. A photomicrograph of a UO2 particle from compact 14 displaying kernel migration is presented in Figure 19. All of the UC2 particles examined (eight total) showed extensive buffer and IPyC layer failure and
HRB-21 configuration
Number of cells Number of fuel compacts Number of encapsulated piggyback specimens Cylindrical fuel compact diameter Cylindrical fuel compact lengths Fissile fuel type Fertile fuel type 235 U enrichment Fissile particle diameter Fertile particle diameter Fissile particle batch Fertile particle batch Total number of fissile particles Total number of fertile particles Defective SiC layer fraction – fissile particles Defective SiC layer fraction – fertile particles
1 24 24 12.27–12.51 mm 49.13–49.35 mm LEU UCO TRISO-P ThO2 TRISO-P 19.66% 904 mm 988 mm 8876-70-0 8876-58-0 42 540 106 240 5.4 106 1.7 105
significant amounts of fission product accumulation. Two of the UC2 particles, or 25% of those examined, had SiC layer failures. These SiC failures occurred next to areas of the IPyC where high concentrations of fission products were present. Examination of the UCO particles revealed significant amounts of fission product attack of the SiC. The extent of this attack ranged from slight to severe. Although not directly measured from examinations of a similar batch of UCO particles irradiated in HRB-15A, it was surmised that this fission product attack was also due to palladium. Of the total 315 fertile ThO2 particles examined, over half displayed IPyC layer failure and nearly 2% displayed SiC layer failure. 3.07.2.3.10 HRB-21
The objective of the HRB-21 capsule irradiated in HFIR at ORNL25 was to demonstrate the irradiation performance of reference NE-MHTGR fuel. A single gas-swept cell contained eight graphite bodies, each of which held three fuel compacts. Each graphite body also contained three sets of encapsulated (piggyback) specimens. These samples were sealed in niobium tubes of up to 52 mm length and 2.2 mm diameter, and each sample contained either absorptivity specimens or loose fuel particles. The test was originally scheduled to be irradiated for six reactor cycles; however, because of difficulty in maintaining control of test temperature, the experiment was terminated after five reactor cycles. Configuration and irradiation data are given in Tables 40 and 41.
TRISO-Coated Particle Fuel Performance
Table 41
HRB-21 irradiation data
Start date End date Duration (full power days) Peak burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average temperature ( C) Peak temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
Table 42 20 June 1991 21 November 1991 105 22 3.5 950 1300 1 108 2 104
NPR-1 and NPR-2 configurations
Number of cells Number of fuel compacts Number of encapsulated piggyback specimens Cylindrical fuel compact diameter Cylindrical fuel compact lengths Fuel type 235
Postirradiation metallographic examination of three fuel compacts was performed. SiC layer failure for both fissile and fertile particles ranged between 0% and 5%. During irradiation, the online ionization chambers recorded several spikes that indicated the failure of approximately 130 particles. The metallographic examinations also revealed that the IPyC layer was in contact with the SiC layer. However, in some cases where the IPyC was cracked radially, the IPyC layer was debonded from the SiC. Fission product attack of the SiC layer was also observed. The chemical attack took place at the tips of cracks in the IPyC layer where fission product transport was not likely to be enhanced. However, scanning electron microscopy did not detect localized high concentrations of fission products in the SiC but did detect low levels of palladium extending 5–10 mm uniformly into the SiC. 3.07.2.3.11 NPR-1 and NPR-2
The NPR-1 and NPR-2 capsules were irradiated in HFIR at ORNL26 to demonstrate the irradiation performance of reference NP-MHTGR fuel at the upper bounds of burnup, temperature, and fast fluence. NPR-1 was irradiated one month before and then concurrently with the NPR-2 capsule in HFIR. NPR-1 consisted of a single gas-swept cell containing 16 fuel compacts in addition to 12 sets of loose particles. The loose specimens were sealed in niobium tubes, 29 mm long and 2.2 mm in diameter. NPR-2 consisted of a single gas-swept cell containing 16 fuel compacts, in addition to 16 sets of loose particles. The loose specimens were sealed in niobium tubes, 29 mm long and 2.2 mm in diameter. Configuration and irradiation data for both capsules are given in Tables 42 and 43. Postirradiation metallographic examination of two NPR-1 fuel compacts was performed. The examination indicated that 0.6% of the SiC layers had failed in one compact and that 0% had failed in the other compact. The online gas measurements recorded 526
U enrichment Fuel particle diameter Fuel particle batch Total number of fuel particles Defective SiC layer fraction
Table 43
177
NPR-1
NPR-2
1 16 12
1 16 16
12.43 mm
12.43 mm
49.42 mm
49.42 mm
HEU UCO TRISO-P 93.15% 758 mm FM19-00001 composite 77 500
HEU UCO TRISO-P 93.15% 758 mm FM19-00001 composite 77 500
3 106
3 106
NPR-1 and NPR-2 irradiation data
Start date End date Duration (full power days) Peak burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average temperature ( C) Peak compact temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
NPR-1
NPR-2
25 July 1991 29 May 1992 170
28 August 1991 29 May 1992 172
79 3.75
79 3.75
974
753
1240
1030
1 108 3 104
5 109 6 105
spikes from the ionization chamber. Assuming that each spike corresponds to a particle failure, 0.7% of the total number of particles had failure of all coatings. Postirradiation metallographic examination of one NPR-2 fuel compact was performed. This examination indicated that 3% of the SiC layers had failed. The online gas measurements recorded 135 spikes from the Geiger–Mu¨ller tube. This detector is less sensitive than ionization chambers, and may have missed some transient spikes. However, assuming each spike corresponds to a particle failure, a lower bound of 0.2% can be set for the total number of particles that failed. The metallographic examinations also revealed that the IPyC layer had remained bonded to the SiC except in the vicinity of SiC cracks where
178
TRISO-Coated Particle Fuel Performance
debonding was observed. It was also observed that between 10% and 30% of the particles with failed IPyC layers also displayed cracked SiC layers.
demonstrate the irradiation performance of reference NP-MHTGR fuel at the upper bounds of nominal operating conditions. The same reference fuel was also irradiated in the NPR-1 and NPR-2 tests. For NPR-1A, 20 fuel compacts were placed vertically in a single, gas-swept cell. Originally, the test was scheduled for 104 days of irradiation, but was terminated after 64 days because of indications of a significant number of fuel particle failures. Configuration and irradiation data are given in Tables 44 and 45. Postirradiation metallographic examination of one fuel compact was performed. This examination indicated that 1% of the SiC layers had failed. On the basis of the online gas measurements, it was estimated that approximately 48 particles had failed, which correspond to 0.06% of the total particle population.
3.07.2.3.12 NPR-1A
The NPR-1A capsule was irradiated in the ATR at the INL.27 The primary objective of the test was to
Table 44
NPR-1A configuration
Number of cells Number of fuel compacts Cylindrical fuel compact diameter Cylindrical fuel compact lengths Fuel type 235 U enrichment Fuel particle diameter Fuel particle batch Total number of fuel particles Defective SiC layer fraction
Table 45
1 20 12.37–12.50 mm 49.33 mm HEU UCO TRISO-P 93.15% 758 mm FM19-00001 composite 75 360 3 106
3.07.2.3.13 AGR-1
The AGR-1 experiment involves six separate capsules, each containing approximately 50 000 particles in the form of fuel compacts. It is an instrumented lead experiment, irradiated in an inert sweep-gas atmosphere with individual online temperature monitoring and control of each capsule. A horizontal capsule cross-section at the top of the test train is shown in Figure 20. A vertical section of the capsule was shown previously in Figure 3, and the experiment flow path was shown previously in Figure 5. The sweep gas also has online fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation (see Figure 5). The first of eight planned experiments, AGR-1, had
NPR-1A irradiation data
Start date End date Duration (full power days) Peak burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) Average temperature ( C) Peak temperature ( C) BOL 85mKr R/B EOL 85mKr R/B
2 October 1991 3 January 1992 64 64 2.1 977 1220 4 109 1.8 105
Graphite Thermocouples
Insulating Stack 1 ATR core center
Stack 2
Stack 3
Hf shroud
Fuel compact
SST shroud Gas lines
Figure 20 Horizontal cross-section of an AGR experimental capsule.
TRISO-Coated Particle Fuel Performance
Table 46
179
Fuel attributes for AGR-1
Property
Kernel diameter (mm) Kernel density (Mg M3) Buffer thickness (mm) IPyC thickness (mm) SiC thickness (mm) OPyC thickness (mm) Buffer density (Mg M3) IPyC density (Mg M3) SiC density (Mg M3) OPyC density (Mg M3) IPyC anisotropya (BAF) OPyC anisotropy (BAF) IPyC anisotropy postcompact anneal (BAF) OPyC anisotropy postcompact anneal (BAF) Sphericity (aspect ratio)
Mean uranium loading (g U per compact) Compact diameter (mm) Compact length (mm) Defective sic coating fraction Defective IPyC coating fraction Defective OPyC coating fractionb
Specified range for mean value
Actual mean value population standard deviation Baseline
Variant 1
Variant 2
Variant 3
350 10 10.4 100 15 40 4 35 3 40 4 0.95 0.15 1.90 0.05 3.19 1.90 0.05 1.035 1.035 Not specified
349.7 9.0 10.924 0.015 103.5 8.2 39.4 2.3 35.3 1.3 41.0 2.1 1.10 0.04 1.904 0.014 3.208 0.003 1.907 0.008 1.022 0.002 1.019 0.003 1.003 0.004
102.5 7.1 40.5 2.4 35.7 1.2 41.1 2.4 1.10 0.04 1.853 0.012 3.206 0.002 1.898 0.009 1.014 0.001 1.013 0.002 1.021 0.002
102.9 7.3 40.1 2.8 35.0 1.0 39.8 2.1 1.10 0.04 1.912 0.015 3.207 0.002 1.901 0.008 1.023 0.002 1.018 0.001 1.036 0.001
104.2 7.8 38.8 2.1 35.9 2.1 39.3 2.1 1.10 0.04 1.904 0.013 3.205 0.001 1.911 0.008 1.029 0.002 1.021 0.003 1.034 0.003
Not specified
1.003 0.003
1.030 0.003
1.029 0.004
1.036 0.002
1% of the particles shall have an aspect ratio 1.14 0.905 0.04
1.054 0.019
1.056 0.019
1.053 0.019
1.055 0.018
0.917
0.915
0.904
0.912
12.22–12.45 25.02–25.4 2.0 104
12.36 0.01 25.066 0.080 4.0 105
12.36 0.01 25.123 0.030 0
12.36 0.01 25.077 0.065 2.0 105
12.36 0.01 25.227 0.037 0
2.0 104
0
0
0
0
1.0 102
0
9.6 105
0
0
a
Specification does not apply to variants 1 and 2. Value is an estimate of an attribute property, not the mean of a variable property.
b
been under irradiation at the INL ATR and was completed in November, 2009; PIE is scheduled to begin in April, 2010.28,29 Table 46 presents pertinent attributes of the fuel that is being irradiated in AGR1.30 Configuration data are presented in Table 47. Irradiation of the experiment began on 24 December 2006, and will continue for approximately 2.5 years to reach a peak burnup of 19% FIMA for the fuel compacts. Table 48 provides a summary of the irradiation status of the AGR-1 experiment. Detailed as-run physics and thermal analyses are performed cycle by cycle to track fuel burnup, fast neutron fluence damage, and fuel temperatures during the irradiation. Peak burnups ranged from 16 to 19% FIMA and fast fluences were between 3.0 and 4.0 1025 n m2 (E > 0.18 MeV). On the basis of the fuel temperature distributions during each cycle, time-averaged peak and
Table 47
AGR-1 configuration data
Number of cells Number of compacts per cell Cylindrical compact diameter Cylindrical compact height Fuel type Particle batch – capsule 3 and 6 Particle batch – capsule 1 and 4 Particle batch – capsule 2 Particle batch – capsule 5 235 U enrichment Number of particles per compact Number of particles per capsule Defective SiC layers
6 12 12.34–12.36 mm 25.0–25.3 mm LEU UCO TRISO Baseline Variant 3 Variant 2 Variant 1 19.6% 4 150 49 800 <4 105
time-averaged volume-averaged temperatures are calculated as the irradiation progresses. After 514 effective full power days, the time-averaged peak fuel temperatures ranged between 1120 and 1180 C
180
TRISO-Coated Particle Fuel Performance
Table 48
AGR-1 irradiation data
Start date End date Duration (full power days) Capsule Time-averaged peak temperature ( C) Time-averaged volume-averaged temperature ( C) Maximum burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) BOL 85mKr R/B EOL 85mKr R/B
24 December 2006 30 October 2009 621 1 2 1136 1118 1029 991 17.5 18.9 3.2 3.8 8 108 1 108 9 108 4 108
and time-averaged volume-averaged temperatures were 100–150 C lower, depending on the capsule. R/B rate ratios have been calculated for many of the short-lived fission gases.31 In all cases, the R/B is less than 107, indicative of release from heavy metal contamination. (A failure of one particle in a capsule would result in an R/B of 3.5 106 based on 4150 particles per capsule and a release of 1.5% from the kernel, which is a typical value at these temperatures and burnups.)32 3.07.2.4
European Experience
The European Commission’s 7th Framework Programme has conducted recent TRISO-coated particle fuel irradiations termed ‘HFR-EU1’ and ‘HFR-EU1bis.’ The experiments share the objective of exploring the potential for high performance and high burnup of the existing German UO2 TRISO-coated particle fuel pebbles for advanced applications, such as the conceptual Generation IV very-high-temperature gas-cooled reactor. As discussed in Section 3.07.2.2, during extensive irradiation tests at and above nominal power-plant conditions in the 1980s and 1990s, not a single coated particle of ‘near-to-production’ fuel elements produced by German researchers with LEU–TRISOcoated particles failed. Irradiating this fuel under defined conditions to extremely high burnups and higher temperature would allow a better understanding of the ultimate irradiation performance of German UO2 TRISO-coated particle fuel. The goal of the HFR-EU1 was to obtain particularly high burnup (20% FIMA) at a peak temperature of 1150 C, typical of pebble bed operation, whereas HFR-EU1bis was dedicated to a particularly high central pebble temperature of up to 1250 C and up to typical pebble bed burnups (10% FIMA).33,34 The irradiated pebbles were 60 mm in diameter with LEU-TRISO-coated UO2. Details of the German
3 1118 980 19.5 4.1 6 109 1 108
4 1180 1041 19.3 4.0 9.0 109 5 108
5 1152 1005 18.3 3.7 1 108 2 107
6 1117 980 16.2 3.0 1 108 1 107
(Arbeitgemeinschaft Versuchsreaktor) and Chinese (Institute of Nuclear and New Energy Technology) fuel attributes are found in Table 49. The design of HFR-EU1 and HFR-EU1bis is on the basis of previous experience of HTR fuel irradiations within the European Union. HFR-EU1 contained five pebbles and six mini samples (ten coated particles each, packed in graphite powder and contained in a niobium tube). Five pebbles in a full-size standard high-temperature gas reactor fuel element rig were used in HFR-EU1bis. Schematic drawings for each arrangement are shown in Figures 21 and 22. Configuration and irradiation data are in Tables 50 and 51. In HFR-EU1, the upper sample holder containing the two Chinese INET fuel pebbles is equipped with 14 thermocouples, while the lower holder containing the German AVR fuel pebbles has 20. In HFR-EU1bis, the central temperature was held constant to control the experiment, whereas control was achieved in HFR-EU1 by holding the surface temperature of the pebble constant. Measured temperatures during the irradiation (without correction for thermal drift and neutron induced decalibration) ranged between 800 and 1000 C for HFR-EU1 and between 900 and 1200 C for HFR-EU1bis, consistent with the peak fuel temperature targets set for the irradiations. In HFR-EU1bis, neutronic calculations indicate that the peak pebble burnup varied between 9% and 11% FIMA and neutron fluence varied between 3.0 and 4.0 1025 n m2 (E > 0.1 MeV) depending on the axial location of the pebble. After 12 cycles in HFR-EU1, neutronic calculations indicate that the peak pebble burnup varied between 9% and 11% FIMA and neutron fluence varied between 2.7 and 3.7 1025 n m2 (E > 0.1 MeV) depending on the axial location of the pebble. Fission product monitoring in HFR-EU1 was accomplished using gas grab samples. At the end of
TRISO-Coated Particle Fuel Performance
Table 49
181
German and Chinese fuel attributes for EU1 and EU1bis
Irradiation
EU1
EU1
EU1bis
Coated particle source Particle batch Kernel composition Kernel diameter (mm) Enrichment (235U wt%) Thickness of coatings (mm) Buffer IPyC SiC OPyC Particle diameter (mm) Pebble Heavy metal loading (g per pebble) 235 U content (g per pebble) Coated particles per pebble Volume packing fraction (%) Free uranium fraction/U-total Matrix grade and density (kg m3)
INET V000802 UO2 490.3 17.08
AVR HT 384–393 UO2 502 16.76
AVR HT 354–383 UO2 501 16.75
97.7 42 37.8 40.8 926.9
92 40 35 40 916
92 38 33 41 909
5.02 0.858 8500 5.0 2.3 107 A3-3 1760 1900
6.0 1.00 9560 6.2 7.8 106 A3-3 1750 1900
6.0 1.00 9560 6.2 7.8 106 A3-3 1750 1900
Temperature of final heat treatment ( C)
-300
-200
-100
Core
-100
-200
the particle releases 1% of the measured shortlived Kr and Xe isotopes (which is not an unreasonable estimate on the basis of other irradiations), then HFR-EU1bis would have contained a few initially defective particles at the beginning of irradiation and additional tens of particle failures at the end. The particle failures may have been related to overheating of the fuel early in the irradiation when an improper gas mixture was inadvertently introduced into the capsule. After 12 irradiation cycles in HFR-EU1, R/B of 4 108 and 1.4 107 were measured for Chinese INET and German AVR fuel, respectively. These low values suggest that in HFR-EU1, no failures have been detected. Instead, the measured fission gas-release probably originates, again, from uranium and thorium impurities in the matrix graphite of the pebbles and in the graphite cups used to hold the pebbles in place. 3.07.2.5
-300 Figure 21 HFR-EU1 capsule schematic.
the HFR-EU1bis irradiation, the R/B was 4 106. In the earlier experiments, HFR-K5 and HFR-K6, R/B values of 5 107 had been measured on fresh fuel. If it is assumed that when particle coatings fail,
Chinese Experience
The Chinese fuel development effort has been performed in part to support the HTR-10 reactor.35,36 HTR-10 is a 10 MW modular high-temperature gascooled test reactor fueled with 60 mm diameter spherical fuel elements, each containing 8300 lowenriched UO2 TRISO-coated fuel particles. Over 20 000 spherical fuel elements have been manufactured for the HTR-10 in 2000 and 2001.
182
TRISO-Coated Particle Fuel Performance
To evaluate the fuel performance of the HTR-10 fuel, two fuel elements (spheres) were chosen at random from each of the first two production batches. These four fuel spheres, together with approximately
Gaslube connectors Thermocouple plug Penetration reactor top lid Dynamic seating penetration plug Second containment gas out Biological shield Refa-170 Thermocouples Second containment (Refa) Purge gas out
+270 +190
1
+120
2
+50
3
Core
0 –20
4
–90
5
–160
6
–230
First containment (capsule)
0.4 0.5 0.6 0.7 0.8 0.9 1.0
7
–310 First containment purge gas in Second containment gas in
Figure 22 HFR-EU1bis capsule schematic. Reproduced from Fu¨tterer, M. A.; et al. Nucl. Eng. Des. 2008, 238(11), 2877–2885.
Table 50
13 500 loose fuel particles, were irradiated in the Russian IVV-2M reactor. Selected physical attributes of the fuel used in this test are listed in Table 52. The irradiation test rig consisted of five capsules. Capsule 1, located at the top, contained the loose particles, while Capsules 2 through 5, in descending order, each contained a test fuel sphere. Irradiation test data are summarized in Table 53. After 3.9% FIMA, where 85mKr R/B values were in the 3 107 range, gas-release suddenly increased in Capsule 4 with R/B values exceeding 103. This result indicated a significant number of fuel particle failures. Further measurements of gas-release from this capsule were terminated at this point. PIE indicated that oxygen impurities in the helium flow gas (such as from air ingress) had reacted with the fuel element matrix material. This reaction turned the matrix into a graphite powder agglomerate and damaged the capsule thermocouple, the internal metal containment, and the fuel particles. No intact fuel particle from this capsule has been found. From beginning to end of irradiation, the remaining three capsules displayed slight variations in gas-release (other than for the in-pile heat-up tests). These variations were attributed to fluctuations in temperature control and implied that no complete (through all layers) particle failures had occurred during irradiation. The high R/B values were due to high U contamination levels and initially defective particles. Quality was improved in subsequent fuel production batches. During irradiation, after 4.1% FIMA, the temperature in Capsule 3 was raised from a nominal 1000– 1200 C for 200 h and then returned to the nominal level. Likewise, after 6.1% FIMA, Capsule 3 temperature was raised from a nominal 1000–1250 C for 200 h and again returned to the nominal level. In both cases, the 85mKr R/B increased by approximately an order of magnitude during the heat-up. When the temperatures were returned to the nominal level, R/B values also returned to those that existed prior to the heat-up. This result indicated that no particles had failed
HFR-EU1 and HFR-EU1bis configuration data
Irradiation
HFR-EU1
Number of cells Number of spheres per cell Spherical fuel element diameter Fuel type Particle batch Number of particles per sphere Number of particles per capsule
2 2 in cell 1 60 mm LEU UO2 LTI – TRISO INET V000802 8500 17 000
HFR-EU1bis
3 in cell 2 60.4 mm LEU UO2 LTI – TRISO AVR HT 384–393 9560 28 680
1 5 60.4 mm LEU UO2 LTI – TRISO AVR HT 354–383 9560 47 800
TRISO-Coated Particle Fuel Performance
Table 51
183
HFR-EU1 and HFR-EU1bis irradiation data
Irradiation
HFR-EU1
HFR-EU1bis
Start date End date Duration (full power days) Capsule Maximum temperature ( C) Minimum temperature ( C) Maximum burnup (% FIMA) Peak fast fluence (1025 n m2, E > 0.18 MeV) BOL 85mKr R/B EOL 85mKr R/B
29 September 2006 24 February 2008 332.8a Cell 1 1000 800 9–11 3.7 1.0 108 4.0 108
9 September 2004 18 October 2005 249.9 – 1200 900 9–11 3.0–4.0 1.0 106 4.0 106
Cell 2 1000 800 9–11 3.7 1.0 108 1.4 107
a
After 12 irradiation cycles.
Table 52
Characteristics of Chinese fuel
Batch number Coated particle Kernel form 235 U enrichment (wt%) Kernel diameter (mm) Kernel density (g cm3) Buffer thickness (mm) IPyC thickness (mm) SiC thickness (mm) OPyC thickness (mm) Buffer density (g cm3) IPyC density (g cm3) SiC density (g cm3) OPyC density (g cm3) IPyC BAF OPyC BAF Spherical fuel element Fuel element diameter (mm) Matrix density (g cm3) U per fuel element (g) U contamination U-free/U-total
Table 53
1
2
UO2 17 500.9 10.81 91 48.1 35.0 41.6 1.02 1.84 3.20 1.90 1.03 1.03
UO2 17 497.0 10.90 96 44.7 36.6 42.1 1.05 1.84 3.20 1.84 1.02 1.03
60 1.73 5.00 1.1 106 1.4 104
60 1.72 4.98 7.6 107 2.3 104
HTR-10 test fuel irradiation data
Start date End date Irradiation temperature ( C) Capsule number Fuel batch number/sphere ID Burnup (% FIMA) Fast neutron fluence (1025 n m2)a EOL 85mKr R/Bb Capsule notes
a
during the heat-up tests. Following irradiation, 2014 Capsule 3 particles, approximately one-fourth of the total number in this sphere were examined by an irradiated microsphere gamma analyzer (IMGA), which identified one failed particle (most likely an initially failed particle). Ceramographic examinations of particles from this capsule did not reveal any PyC or SiC layer cracks, fission product attack of the SiC, or kernel migration (amoeba effect). IMGA examination of 2117 Capsule 2 particles, approximately one-fourth of the total, identified two failed particles, which again, were most likely initially failed particles. Ceramography of Capsule 2 particles were similar to Capsule 3 particles where no irradiation induced defects were found. Following the irradiation phase of the experiment, a high-temperature, in-pile heat-up test was conducted with Capsule 5. This heat-up test was performed in-pile because a hot cell furnace was not available. Capsule 5 was moved to the center high flux region of the test reactor while the remaining capsules were transferred to low flux reactor regions where fuel
July 2000 February 2003 1000 50 2 1/SFE5 10.4 1.10 1.7 105
3 2/SFE12 11.4 1.31 6.7 107 In-pile heat-up test to 1200 and 1250 C for 200 h each
4 2/SFE8 10.9 1.30 Not measured Capsule failed after 3.9% FIMA
5 1/SFE7 10.1 1.06 7.8 106 In-pile heat-up test above 1600 C at end of irradiation
Cutoff energy for fast neutron fluence not specified. The tabulated end of life (EOL) R/B for capsule 5 was measured at the end of the irradiation test phase and prior to the start of the high-temperature heat-up test. b
184
TRISO-Coated Particle Fuel Performance
temperatures did not exceed 1000 C. Temperature in Capsule 5 was increased step-wise to the intended 1600 C, held for approximately 4 h, and then reduced step-wise after completion of the test. The total duration of the heat-up test was 50 h. Since the thermocouple in Capsule 5 failed during transfer, temperatures were calculated where further analyses indicated that test temperatures had exceeded the intended 1600 C. Gas release measurements during the heat-up indicated many particle failures. Postirradiation IMGA examination of 2189 particles identified that 127, or 6% of the particles sampled, had failed. Ceramography revealed that most particles had remained intact. However, the failed particles displayed several types of defects including radial and tangential SiC cracks, IPyC cracks, and thorough failures in all layers. None of the observed particles displayed fission product attack of the SiC or kernel migration. 3.07.2.6
Japanese Experience
The Japanese have conducted a number of irradiations of high quality LEU TRISO fuel for the High-Temperature Test Reactor (HTTR), including the Japanese fuel design for first and second core loadings of HTTR and a ‘high burnup’ particle design, developed for later introduction into the HTTR. The gas reactor development program in Japan has been focused on high-temperature process-heat applications from its inception. In addition, the Japanese ‘pin-in-block’ prismatic fuel design has higher calculated maximum fuel temperatures for a given reactor coolant temperature than traditional prismatic gas reactor designs. As a result, the temperatures in the HTTR and projected for commercial gas reactor applications are elevated, with time averaged peak fuel temperatures in the range of 1350 C. Under
Table 54
these conditions, phenomena such as kernel migration, metallic fission product diffusion through the SiC layer, and fission product chemical attack of the SiC layer become limiting, particularly at higher burnup. To address these service conditions with LEU SiC TRISO fuel, the Japanese particle design has been optimized for a relatively low maximum burnup (e.g., larger kernel, thinner buffer), with a maximum burnup limit for service in HTTR of 3.6% FIMA. The test data for the HTTR fuel particle design are provided in Table 54. The burnups in most of the irradiations were very low, between 0.5% and 4% FIMA. The fast fluence values were not available in all cases, but on the basis of 13 and 15 Oarai Gas Loop (OGL-1) irradiations, the fast fluence was probably less than 0.5 1025 n m2 for all the irradiations, approximately an order of magnitude below values for commercial HTGRs. Defect levels in the fuel showed continual improvement with each subsequent irradiation, and the R/B values were very low. Table 55 summarizes the conditions and results of irradiation testing of the Japanese ‘high burnup’ fuel design, plus HTTR fuel irradiated to extended burnup. The most recent irradiation was conducted in the HFIR reactor at ORNL, designated HRB-22. The range of burnup experienced by the fuel was 4.1–6.7% FIMA, with an average of 5.3% FIMA. Four particle failures (out of a total of 32 200 particles in the capsules) were observed during the irradiation phase. This would produce an estimated population failure fraction of 2.8 104 at 95% confidence at a fairly low burnup. In the high burnup fuel irradiation in 91F-1A, conducted in the HTTR, Japanese fuel was taken to burnups of between 7% and 9% FIMA at temperatures between 1120 and 1340 C. In-pile gas-release and postirradiation leaching indicated 4 particle failures, two in the higher temperature upper
Summary of low burnup irradiations of HTTR fuel
Tests
Time (EFPD) Temperature ( C) Burnup (% FIMA) Fast fluence (1025 n m2) Exposed kernels (BOL) Exposed kernels (EOL) SiC defects (BOL) SiC defects (EOL) R/B at EOL
OGL-1 6th
7th
8th
10th
12th
13th
15th
22 980–1488 0.1–0.4
58 984–1376 0.4–1.2
54 940–1370 0.3–1.0
130 879–1331 0.8–2.8
195 1300 3.0–3.9
5 105 1 104 2 103 2 103 1.5 106
3 105 1 104 4 103 4 103 4.0 107
6 106 1 104 5 104 2 103 1.1 107
3 104 5 104 7 104 1 103 3.2 106
1 104 3 104
244 1083–1342 1.3–3.7 0.11–0.49 4 106 7 105 4 106
218 1180–1344 1.1–3.1 0.1–0.44 3 106 3 107 5 105
6.8 108
3.8 107
3.1 106
TRISO-Coated Particle Fuel Performance
Table 55
185
Japanese particle attributes and irradiation conditions
Particle attributes for each irradiation
HRB-22
91F–1A
94F–9A
14th OGL-1
235 U enrichment (%) Kernel diameter (mm) Buffer thickness (mm) IPyC thickness (mm) SiC thickness (mm) OPyC thickness (mm) Irradiation conditions Effective full power days Number of compacts per capsule Number of particles per capsule Number of cells Maximum temperature ( C) Average temperature ( C) Burnup (% FIMA) Fast fluence (1025 n m2, E > 0.1 MeV) BOL 85mKr R/B EOL 85mKr R/B
4.07 544 97 33 34 39
7.98 551 97 32 34 39
7.82 615 59 30 29 47
19.54 553 84 32 37 42
89
360 2 10 100 Upper 1120
Lower 1340
360 2 25 000 Upper 1120
Lower 1190
9.6 6.0
7.9 4.6
7.0 3.1
5.5 2.3
65 18 full and one half-size 245 000 1 1160–1339 1122–1305 0.73–1.17 0.07–0.12
32 200 1 898–1156 821–1071 4.1–6.7 1.3–2.8 1.0 106 3.0 106
Source: Minato, K.; et al. HRB-22 Capsule Irradiation Test for HTGR Fuel; JAERI-research 98-021; 2008; Sawa, K.; et al. JAERI-research 2001-043; Irradiation Test of High Burnup Coated Fuel Particles for High Temperature Gas-cooled Reactor (91F-1A Sweep Gas Capsule Irradiation Test); 2001; Sawa, K.; et al. JAERI-research 2002-012; Acceleration Irradiation Test of First-loading Fuel of HTTR up to High Burnup; 2002; Hayshi, K.; et al. JAERI-research 2000-001; Irradiation Experiments of the 13th–15th OGL-1 Fuel Assemblies; 2000.
capsule, and two in the lower capsule. Pressure vessel failure was the stated cause of the failures. The 95% confidence value for particle failures during the irradiation was 4.6 104. In irradiation 94F-9A conducted in the Japan Material Test Reactor, TRISO fuel was irradiated at 1200 C to burnups between 5.5% and 7% FIMA. No failures occurred in one capsule, and in the other capsule, only one particle out of 50 000 failed. Another irradiation of ‘high burnup’ fuel was conducted in the OGL-1 loop to a burnup of 1% FIMA with very low fast neutron fluence with a failure fraction of 1 104 observed during a temperature transient test to 1500 C. 3.07.2.7
Irradiation Performance Summary
Extensive TRISO-coated particle irradiations have been performed around the world over the past 40 years. The majority of the particles have been irradiated in the United States and Europe/Germany, with smaller amounts of fuel irradiated by Chinese and Japanese programs. A summary of salient features of all the irradiations in which R/B measurements were available is found in Table 56. The irradiation experiments reviewed here, including many others not summarized above, contained particle fuels with widely different coatings and kernels and were conducted at a variety of burnups, temperatures, and fluences.
Figure 23 plots EOL online gas-release (R/B) as a function of temperature, burnup, and fluence separately, and no trend is evident. Releases observed in the experiments are a function of (a) the initial level of heavy metal contamination in the fuel, and (b) in-service failures. 3.07.2.7.1 Heavy metal contamination
BOL 85mKr R/B is often used as a metric of as-manufactured fuel quality. Fuel developed by German researchers exhibited overall low initial defect fractions, with measured R/B values ranging between 1010 and 106. These results also demonstrated that fuel quality improved over time, as was the case for fuel used in the AVR. It should be noted that the lowest ever 85mKr R/B measured in an in-reactor irradiation is from German fuel (1010). Both the Chinese and Japanese fuel exhibited slightly higher levels of contamination, as indicated by the initial R/B ratios in their irradiations. By contrast, the high gas release from the historic US database (all experiments except AGR-1) is a combination of high initial levels of heavy metal contamination in the fuel and in-service failures. During the 1970s and 1980s, the lower fuel quality of the historic US fuel resulted in R/B ratios that varied between approximately 2 108 and 104, indicating that the initial contamination levels in US fuel displayed great variability and could be rather high. The level did not
TRISO-Coated Particle Fuel Performance
1.E-02
1.E-02
1.E-03
1.E-03
1.E-03
1.E-04
1.E-04
1.E-04
1.E-05 1.E-06
EOL 85mKr R/B
1.E-02
EOL 85mKr R/B
EOL 85mKr R/B
186
1.E-05 1.E-06
1.E-05 1.E-06
1.E-07
1.E-07
1.E-07
1.E-08
1.E-08
1.E-08
1.E-09 800
1.E-09 1000 1200 1400 Temperature ( ˚C)
1
10 100 Burnup (%FIMA)
1.E-09 0 2 4 6 8 10 12 14 Fast fluence (10E25 n m-2)
US historic
German
US historic
German
US historic
German
US new (AGR)
EU new
US new (AGR)
EU new
US new (AGR)
EU new
China
Japan
China
Japan
China
Japan
Figure 23 TRISO-coated fuel performance as a function of temperature, burnup, and fast fluence.
significantly change until the early 1990s, when serious efforts at reduction of initial contamination were undertaken in the fabrication campaigns. With changes to fabrication processes, the most recent US irradiation, AGR-1, is demonstrating very low levels of contamination with initial R/B ratios in the range of 108.31 3.07.2.7.2 In-service failures
The behavior of German fuel under irradiation is excellent. Measured EOL 85mKr R/B values ranged between 3 109 to 105 for the various German irradiation experiments. In most cases, the maximum R/B was measured at EOL; however, in some irradiations, the final portion of the experiment was conducted at lower temperatures, which caused the R/B to decrease. Nevertheless, these results indicate that no fuel particle had failed while under irradiation. PIE results confirmed the more extensive and more reliable gas-release data. Out of approximately 380 000 UO2 and 80 000 (Th, U)O2 particles tested, there were no in-pile failures and only a few ‘damaged’ particles due to experimental anomalies. Gas release was attributed to as-manufactured defects and heavy metal contamination only. The historic US fuel irradiations indicate that inreactor fuel failures occurred at the percent level. PIEs of the historic US fuel revealed in most experiments up to percent level particle failures, with very high levels of individual layer failures. These failures
resulted in the observed high gas-release (R/B), which is approximately a factor of 1000 times greater than that of the German fuel under a broad range of conditions (i.e., temperature, burnup, and fluence).37 The current US fuel in AGR-1 is the culmination of a significant effort to overcome the limitations observed in past irradiations. Fabrication processes were changed, and irradiations were not accelerated. The result is a fuel that has survived irradiation to high burnup with no failures. The smaller statistical databases associated with the fuel from China, Japan, and Europe are more mixed, with some irradiations indicating good performance, and some indicating fuel failures. In many cases, the root cause of the failures is unclear because of less extensive PIE. 3.07.2.7.3 Failure mechanisms
Review of the US irradiation database indicates that a number of different failure mechanisms in the individual layers of the TRISO coating contributed to the less-than-satisfactory US fuel performance. Failures of the coating layers were attributed to (a) pressure vessel failure, (b) kernel migration (amoeba effect), (c) fission product attack of the SiC layer, (d) irradiation-induced IPyC cracking and/or debonding leading to cracking in the SiC layer, and (e) matrix–OPyC interaction and irradiationinduced OPyC failure.
TRISO-Coated Particle Fuel Performance
Fission product and impurity attack of the SiC and kernel migration are thermally driven phenomena strongly influenced by burnup, temperature, and the temperature gradient across the particle. The temperature gradient is a strong function of the power density in the fuel body, which is directly related to the level of acceleration in the experiment. The PyC-related mechanisms are strongly related to the anisotropy and porosity in the coatings. The anisotropy has a strong influence on the shrinkage and swelling behavior of the PyC layers under irradiation. However, the anisotropy measurements, especially optical methods (OPTAF), are not reliable predictors of PyC failure under irradiation as indicated by the lack of correlation between the measured BAF and PyC failure (e.g., OF-2, HRB-5, HRB-6), and the high level of PyC failure observed in most historic US irradiations. The fabrication processes in the United States have been changed, with the introduction of more accurate methods to measure anisotropy to improve the survivability of the PyC layers. The lack of fuel failure in the US Table 56 Test/cell
Germany R2-K12/1 R2-K12/2 R2-K13/1 R2-K13/4 BR2-P25 HFR-P4/1 HFR-P4/2 HFR-P4/3 SL-P1 HFR-K3/1 HFR-K3/2 HFR-K3/3 HFR-K3/4 FRJ2-K13/1 FRJ2-K13/2 FRJ2-K13/3 FRJ2-K13/4 FRJ2-K15/1 FRJ2-K15/2 FRJ2-K15/3 FRJ2-P27/1 FRJ2-P27/1 FRJ2-P27/1 HFR-K5/1 HFR-K5/2
187
AGR-1 fuel taken to 19% FIMA demonstrates that the changes made to the process were correct and resulted in excellent in-pile fuel performance. 3.07.2.7.4 Acceleration effects
The rate of accumulation of burnup and fast fluence relative to that expected in the reactor (i.e., the degree of acceleration) is an important factor in the behavior observed in these experiments. For most of these fuels, the time to reach goal burnup and fast fluence is 1095 days (3 years), whereas the irradiation times varied greatly. For historic US irradiations, acceleration varied between factors of 2–10. One Japanese irradiation (HRB-22) was also greatly accelerated. By contrast, the German, Chinese, and European irradiations, as well as the US AGR-1 irradiation, were all generally accelerated by only factors of two to three. These low acceleration factors limited thermal gradients across the fuel, which could have jeopardized fuel integrity with enhanced fission product attack and kernel migration (Table 56).
Summary of TRISO-coated particle fuel irradiation experiments Fuel forms
Irradiation time (d)/ acceleration level
Peak temperature ( C)
Peak fissile and fertile burnup (% FIMA)
Peak fluence (1025 n m2)
HEU (Th, U)O2 TRISO
308/3
HEU (Th, U)O2 TRISO
517/2
HEU (Th, U)O2 TRISO LEU UO2 TRISO
350/3 351/3
LEU UO2 TRISO LEU UO2 TRISO
330/3 359/3
LEU UO2 TRISO
396/2.75
LEU UO2 TRISO
533/2
LEU UO2 TRISO
232/4.7
LEU UO2 TRISO
563/2
1100 1280 1170 980 1070 940 945 1075 794 1200 920 920 1220 1125 1150 1150 1120 970 1150 990 1080 1320 1130 Cycled proof test
11.1 12.4 10.2 9.8 15.6 14.7 14.9 14.0 11.3 7.5 10.0 10.6 9.0 7.5 8.0 7.9 7.6 13.2 14.6 13.9 7.6 8.0 7.6 6.7 8.8
5.6 6.9 8.5 6.8 8.1 8.0 8.0 8.0 6.8 4.0 5.8 5.9 4.9 0.2 0.2 0.2 0.2 0.2 0.2 0.1 1.4 1.7 1.3 2.9 <4.3
EOL 85m Kr R/B (106)
0.300 0.200 0.070 0.050 1.000 0.080 0.080 0.008 1.200 0.200 0.100 0.100 0.300 0.020 0.020 0.007 0.007 0.010 0.005 0.003 1.600 10.000 0.120 <0.3 <0.3 Continued
188
TRISO-Coated Particle Fuel Performance
Table 56
Continued
Test/cell
HFR-K5/3 HFR-K5/4 HFR-K6/1 HFR-K6/2 HFR-K6/3 HFR-K6/4 United States F-30/1 F-30/2 F-30/3 F-30/4 F-30/5 HRB-4 HRB-5 HRB-6 OF-2/1 OF-2/2 HRB-14 HRB-15B
R2-K13/2 R2-K13/3 HRB-15A
HRB-16
HRB-21 NPR-1 NPR-2 NPR-1A AGR-1/1 AGR-1/2 AGR-1/3 AGR-1/4 AGR-1/5 AGR-1/6 Europe EU1 China EU1 German EU1bis China Test fuel cell 2 Test fuel cell 3 Test fuel cell 5 Japan HRB-22
Fuel forms
Irradiation time (d)/ acceleration level
Peak temperature ( C)
Peak fissile and fertile burnup (% FIMA) 9.1 8.7 7.2 9.3 9.7 9.2
Peak fluence (1025 n m2)
EOL 85m
Kr R/B (106) <0.3 <0.3 <0.3 <0.3 <0.3 <0.3
4.3 <4.3 3.2 <4.8 4.8 <4.8
LEU UO2 TRISO
634/1.7
Cycled proof test
HEU (Th, U)C2 TRISO and ThC2 TRISO
269/4
LEU WAR UC2 TRISO and ThO2 BISO LEU WAR UC2 TRISO and ThO2 BISO HEU (Th, U)C2 TRISO and ThO2 BISO WAR UCO UC2 (Th, U)O2 TRISO and ThO2 BISO LEU UCO UO2 (Th, U)O2 TRISO and ThO2 BISO LEU UCO UC2 (Th, U)O2 UO2 TRISO and Si BISO and ThO2 TRISO, BISO, and Si-BISO LEU UCO TRISO and ThO2 TRISO LEU UCO UC2 UO2 TRISO and ZrC-TRISO and ThO2 TRISO and Si-BISO LEU UCO UC2 UO2 (Th, U) O2 TRISO and ZrC-TRISO and ThC2 ThO2 TRISO and BISO LEU UCO and ThO2 TRISO-P HEU UCO TRISO-P HEU UCO TRISO-P HEU UCO TRISO-P LEU UCO TRISO
244/4.5
1100 1100 1120 1100 1200 1250
15.0/3.0 19.0/4.5 20.0/5.0 18.0/4.0 12.0/1.5 27.7/13.4
8.0 10.5 11.5 9.5 12.0 10.5
8 100 10 20 20 320
107/10
1250
15.7/4.3
4.7
100
183/6
1100
26.6/9.3
7.9
270
352/3
1350 1350 1190
79.6/4.3 79.5/4.3 28.6/8.5
8.9 8.4 8.3
100 5 300
915
26.7/6.0
6.6
5
174/6.3
1190 985 1150
22.5/4.6 22.1/4.5 29.0/6.4
7.8 7.4 6.5
80 8 380
170/6.3
1150
28.7/6.1
6.3
210
105/10 170/6.3 172/6.3 64/6.3 625/1.75
1300 1240 1030 1220 1136 1118 1118 1180 1152 1117
22.0/2.2 79.0 79.0 64.0 17.5 18.9 19.5 19.3 18.3 16.2
3.5 3.8 3.8 2.1 3.2 3.8 4.1 4 3.7 3
200 300 60 18 0.09 0.04 0.01 0.05 0.2 0.1
LEU UO2 TRISO LEU UO2 TRISO LEU UO2 TRISO
333/3.3 333/3.3 250/4.4
1000 1000 1000
11 11 11
4 4 3.7
0.04 0.14 4
LEU UO2 TRISO
900/1.2 900/1.2 900/1.2
1000 1000 1000
10.4 11.4 10.1
1.1 1.31 1.06
17 0.67 7.8
LEU UO2 TRISO
89/12
1156
6.7
2.8
214/5 169/6.5
517/2
Note: US fluence is for E > 0.18 MeV, fluence is not specified by China, and fluence values of all others are for E > 0.10 MeV.
3
TRISO-Coated Particle Fuel Performance
3.07.3 Safety Testing
and fuel fabrication processes (e.g., SiC coating rate) on fission product release are investigated.
The release of fission products from TRISO-coated irradiated fuels heated at elevated temperatures to simulate accident conditions is reviewed in this section. For a small HTGR, the temperature evolution during a loss of coolant accident with complete depressurization is calculated to reach 1600 C (including a 100 C uncertainty margin) for approximately 30 h, as shown in Figure 24. The preponderance of the experimental data is from tests with fuels of German manufacture, but insights are also provided from a few experiments using Japanese and US fuels. There is some evidence of particle failure caused by internal pressure (pressure vessel failure), but of most interest is degradation of the SiC layer in the TRISO coating during accident performance testing at elevated temperatures, 1500–1800 C. Because releases of cesium are sensitive to the integrity of the SiC layer, much attention is paid to this fission product. Releases of krypton are sensitive to pressure vessel failure, but otherwise trail releases of cesium due to holdup by PyC layers. Releases of other fission products, such as strontium, europium, and cerium are treated where data are available. Silver, although not a safety concern because of its low yield, has potential consequences for reactor maintenance and diffuses readily through the SiC layer, even at 1200 C. The influences of irradiation characteristics (e.g., fuel burnup, neutron fluence, and irradiation temperature), SiC microstructure (e.g., grain size and orientation),
3.07.3.1
Facility Overview
Several operational facilities currently exist for testing high-temperature fission product retention performance of irradiated TRISO fuel. The three primary facilities discussed here (Ku¨FA, Fuel Accident Condition Simulator (FACS), and Core Conduction Cooldown Test Facility (CCCTF)) are all on the basis of the basic concept of a refractory metal (tantalum) furnace operating at maximum temperatures of 1800–2000 C with a helium atmosphere and a water-cooled cold finger intruding into the hot zone to collect condensable fission products on an exchangeable plate connected to the end of the cold finger. 3.07.3.1.1 Ku¨ FA at ITU
Initial development on Ku¨FA (from the German Ku¨hlfingerapparatur, ‘cold finger device’) began in the early 1980s at Forschungszentrum Ju¨lich.38 The system was designed to simulate reactor accident conditions by heating spherical fuel elements in helium to temperatures of up to 1800 C while measuring the release of condensable fission products, as well as noble fission gases. The system was used in the 1980s and 1990s to test high-temperature fission product retention on various irradiated particle fuel specimens.1,38,39 More recently, the system was
2000 Accident simulation tests
1800
Temperature (⬚C)
1600 1400 1200
MODUL, max. fuel element temperature in hypothetical and design basis accident *including 100 ºC margin
1000 800 Reactor operation Temperature
600 400
0
189
100
200
300
400
500
Time (h) Figure 24 Temperature evolution during a depressurized reduction cooling of a small HTR, and in heating tests with irradiated fuel elements. Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
190
TRISO-Coated Particle Fuel Performance
transferred from Ju¨lich, upgraded, and installed in the hot cells at the Institute for Transuranium Elements (ITU) in Karlsruhe for continued accident testing.40 The system consists of a double-walled, watercooled metal jacketed furnace with tantalum electricalresistance heating elements (maximum power of 70 kW at 2400 A), tantalum hot zone components, and a water-cooled cold finger that holds an exchangeable metal plate to condense fission products during operation (see Figures 25 and 26). The cooling system maintains the plate at a temperature of less than 100 C during furnace operation. The cold finger can be removed from the furnace through a sliding valve and gas lock system that maintains the purity of the gas in the furnace during exchange operations. Plates can then be packaged and removed from the facility for gamma spectroscopic counting or for acid leaching and radiochemical analysis. The entire system is operated inside an alpha containment box which itself resides inside a shielded hot cell. Fission gases in the sweep gas circuit are captured in two charcoal cryo-traps, with the second trap serving to indicate the trapping
Figure 25 The cold finger device Ku¨FA installed in the hot cells at ITU. © European Atomic Energy Community, 2011. See http://itu.jrc.ec.europa.eu.
efficiency of the first. The activity in the traps is monitored continuously with NaI detectors during a heating test. Upgrades to this system since the original work at Ju¨lich include an increased maximum temperature capability to 2000 C, updated temperature regulation system, and a redesigned cold finger to allow sufficiently low plate temperatures during operation of the furnace at up to 1800 C. The upgraded Ku¨FA system has recently been described in further detail elsewhere.40 3.07.3.1.2 INL’s FACS
The FACS furnace at the INL is designed to heat fuel specimens at temperatures up to 2000 C in inert atmospheres while monitoring the release of volatile fission metals and fission gases (see Figure 27). The furnace has tantalum hot zone components and is heated using a graphite resistance element. The fuel specimen is heated inside an 85 mm diameter tantalum flow tube and is supported in a tantalum holder that sits atop a tantalum tube inserted from the bottom of the furnace. The fuel specimen temperature is monitored with a Type C (W–Re) thermocouple housed within the tantalum tube that supports the sample holder, with the junction located 12.5 mm below the specimen. The hot zone can accommodate samples up to 75 mm in diameter. Special sample holders have been designed to accommodate the compacts or spherical pebbles, although any specimen geometry can be accommodated through design of a custom tantalum holder. The furnace main chamber is double-walled, water-cooled stainless steel. Both the top and bottom flanges are water-cooled stainless steel and can be removed from the main chamber for sample loading or maintenance operations as needed. The watercooled cold finger is of aluminum construction and holds a 35 mm diameter plate on the end by means of a mechanical lock using a rotary actuator. The temperature at the base of the cold finger, where the condensation plate is attached, is monitored using a Type J thermocouple. Figure 28 depicts a cutaway view of the furnace system. Condensate plates are introduced to the furnace through a separate transfer chamber, purged with helium prior to opening a gate valve to the main chamber. The entire condensate plate exchange operation is performed automatically by computer control. The operator must only remove the used plate once it has been retracted into the transfer chamber and replace it with a fresh plate to be placed on the cold finger during the next exchange cycle.
TRISO-Coated Particle Fuel Performance
191
Coldfinger jacket (vacuum/He filled)
Valve
Water cooled
Removable condensation plate Gas stream cylinder in Ta Heat element in Ta Steel heat sink Spherical fuel element W/RE–thermocouple
Vacuum pump connection Helium output Power connection
Helium input
Figure 26 Ku¨FA system at ITU, Karlsruhe. Reproduced from Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, USA, October 1–4, 2008.
The helium sweep gas is introduced to the furnace from the bottom and injected into the hot zone beneath the specimen and flows toward the condensation plate at the end of the cold finger as shown in Figure 28. The helium is then swept from the furnace through particulate and moisture filters and carried to the fission gas monitoring system for analysis of 85Kr, 133Xe, or other radioactive noble gases.
This system, located outside of the hot cell, consists of dual liquid nitrogen-cooled charcoal traps to contain the noble gases, each with a high purity germanium detector and adjustable collimator. A schematic diagram of the cooling water and He gas circuits is shown in Figure 29. Condensation plates that are removed from the furnace during operation will be packaged in clean
192
TRISO-Coated Particle Fuel Performance
Figure 27 FACS furnace.
plastic containers and sent to the analytical chemistry laboratory for analysis. This will consist of gamma counting of the plates to measure the inventory of major fission products (e.g. 137Cs, 110mAg), followed by acid leaching to remove fission products and for aqueous radiochemical analysis of 131I and betaemitter 90Sr. The system is currently undergoing remote qualification testing and preliminary condensate efficiency tests, after which the system will be installed in the INL Hot Fuel Examination Facility (HFEF) hot cells. The system will be used to conduct hightemperature accident testing on irradiated AGR TRISO fuel compacts from the AGR-1 experiment for the NGNP program. 3.07.3.1.3 ORNL’s Core Conduction Cooldown Test Facility
The CCCTF was developed at ORNL in the early 1990s to perform high-temperature accident simulation performance tests on coated particle fuels.41 The system is operated in a modular hot cell at the
Irradiated Fuel Examination Laboratory (IFEL) at ORNL. The system has undergone several modifications and now consists of a water-cooled graphite resistance-heated tube furnace in a vertical orientation with an internal tantalum container that isolates the test specimens from the furnace heating elements. A helium sweep gas system provides the test atmosphere and transports released fission gases to a cryogenic trap system for collection and gamma counting. Water cooling circuits provide temperature control for sensitive furnace components (including the metal furnace jacket and the cold finger). The furnace system diagram is shown in Figure 30 and a photograph of the furnace is shown in Figure 31. The fuel specimen is loaded from the top and during operation; a water-cooled cold finger is inserted and removed from the top through an air lock. Fuel specimens are held in a graphite or refractory metal holder supported by a tantalum container with a small well for a Type C thermocouple, which serves to both measure the sample temperature and control the furnace control system. The temperature of the tantalum container can be monitored from the outside also through a furnace window using an optical pyrometer. A steel deposition cup is attached to the bottom of the cold finger by means of a screw mechanism and acts as the collection surface for condensable fission products. By periodically inserting and removing the cold finger, the deposition cup can be changed and a history of the metallic releases from the fuel can be obtained; gamma counting or radiochemical analysis allows the identification of the collected elements and their quantities. Fission gases released from the heated fuel are continually swept from the furnace by the helium gas stream and collected in a bank of dual liquid-nitrogen-cooled charcoal traps monitored with NaI detectors. The fission gas traps and detectors are located outside of the hot cell in a low-background area to facilitate detection of very small quantities. The CCCTF has been used for several heating tests of irradiated fuel compacts and unbonded fuel particles.42,43 The system is currently undergoing a series of upgrades to improve performance and process control, including the installation of a redesigned cold-finger airlock, an upgraded fission sweep gas system with an automated gas supply system, a modified liquid nitrogen supply system, and upgraded gamma counting electronics. This upgraded system will be used for accident testing of AGR fuel compacts for the NGNP program, starting with fuel from the AGR-1 irradiation test.
TRISO-Coated Particle Fuel Performance
Cold finger
193
Gate valve
Transfer chamber Condensation plate cradle Heat shields Graphite heating element
Sample holder Tantalum flow tube
Sample support and helium inlet
Helium outlet
Figure 28 Three-dimensional cut away diagram of the FACS main chamber and interior components.
3.07.3.1.4 KORA
The corrosion furnace (KORA) test equipment was developed at Forschungszentrum Ju¨lich in the early 1990s to examine fuel performance under water and air ingress accident conditions.3 The system is depicted in Figure 32. The system was capable of testing fuel specimens up to the size of a spherical fuel element in helium mixed with air or water vapor (more than 80 kPa water vapor partial pressure possible) at temperatures of up to 1600 C while measuring the release of 85Kr. The system was used to test several AVR fuel elements and particles as well as compacts and particles from the HFR-B1 irradiation experiment.3 Work is currently underway to develop a KORA apparatus at ITU in Karlsruhe.44 3.07.3.2
German Experience
Safety tests carried out by German researchers were performed primarily on whole spherical fuel elements containing 16 400 fuel particles, and in some cases on fuel compacts containing 1600 fuel particles. Irradiation data from these tests3,42,45,46 for spherical fuel elements (60 mm in diameter) are presented in
Table 57, and for compacts (cylinders manufactured from spherical elements with a fuel zone 20 mm in diameter) in Table 58. Both tables also contain the maximum integral fractional releases of krypton and cesium measured during the heating tests. The spherical fuel elements listed in Table 57 experienced burnup within the historic pebble bed burnup envelope (10% FIMA) with the exception of HFR-K3/3 (10.6% FIMA). These fuel elements experienced fast fluence within or slightly greater than the historic pebble bed fluence envelope 2.5 1025 n m2, with the exception of HFR-K3/3 (6.0 1025 n m2). By contrast, all of the compacts were irradiated beyond the historic pebble bed reactor envelope. Krypton integral releases as a function of time at various temperatures are shown in Figure 33(a)– 33(c). In Figure 33(a), krypton releases are below the level of one particle failure (1/16 400 ¼ 6.1 105) at 1600 C, whereas that level of release is exceeded at 1700 and 1800 C for spherical fuel elements. The occasional vertical lines in the releases at temperatures above 1600 C are associated with pressure vessel failure of particles. Pressure vessel failure is a function of burnup (fission gas inventory and, in
194 TRISO-Coated Particle Fuel Performance
Hot cell boundary
He to transfer chamber He to condensation plate release
Cold finger
Top cover Cooling water system Manifold
Supply Building Heat cooling exchanger water
Return
Cold finger Top cover Chamber Bottom cover Electrodes Exhaust
Flow meters
Particulate filter
Manifold
FM FM FM FM FM FM
He to sample flow tube
Vacuum pump
Bottom cover Chamber Electrodes Exhaust
He to main chamber Figure 29 Schematic of FACS cooling water and helium supply system.
Fission gas monitoring system
TRISO-Coated Particle Fuel Performance
195
Cold finger
Spool piece
Gate valve assembly
Thermal shield Sweep gas out
Deposition cup Tantalum can
Upper cooling flange Sweep gas in
Furnace insulation
Graphite heating element
Watercooled furnace jacket
Specimen assembly (graphite) Graphite holder
Dual type 'C' thermocouple
Graphite composite reflector
Lower cooling flange
Figure 30 Cross-section schematic of the CCCTF furnace.
UO2 fuel, CO inventory), fuel irradiation temperature (fission gas pressure and oxygen to fission ratio in UO2 fuel),47 and fuel particle design and properties (buffer void volume and SiC strength). Figure 34(b) shows the larger releases associated with higher burnups. 85Kr release can be used as an indicator of 131 I release, on the basis of reactivation of irradiated
Figure 31 Core Conduction Cooldown Test Facility.
fuels immediately before heating tests.48 Figure 35(c) shows that krypton release is negligible (106) for compacts with 10–12% FIMA at 1600 C, but becomes significant (104) at approximately 200 h at 1700 C, and at approximately 150 h at 1800 C. While krypton releases at levels above 6.1 105 are due to the failure of all three coating layers (IPyC, SiC, and OPyC), resulting in an exposed kernel, cesium releases above 6.1 105 measured in safety testing at 1600 C are an indication of defective SiC at the end of irradiation.46 Cesium and krypton releases presented in Figure 34 show cesium release at an inventory level of two particles (1.2 104) for fuel sphere AVR 88/33 and one particle (6.1 105) for fuel sphere 82/20, whereas the krypton releases are well below the inventory of a single particle. These results indicate that fuel sphere AVR 88/33 contains two fuel particles with defective SiC and fuel sphere 82/20 contains one fuel particle
TRISO-Coated Particle Fuel Performance
Ionization chamber H O 2 P2Os
H2O
CO
Blue gel
Tritium Hot cell
Soda CuO lime 600 ⬚C Safety dust filters
Fission gas tap (charcoal)
Liquid N2
Filter for solid fission product
Naldetector
Gas tight box
196
Compressor Blue-gel P2Os Moisture traps
Electronics
Chart recorder
H2O bath 1 Gas
Furnace
2
Heated pipe
Figure 32 Schematic diagram of the KORA facility developed at Forschungzentrum Juelich in the 1990s. Reproduced from IAEA. Fuel Performance and Fission product Behavior in Gas-cooled Reactors. IAEA TECDOC-978; IAEA: Vienna, Austria, November 1997. Table 57 Fuel element
Results of accident simulation tests with irradiated spherical fuel elements Burnup (% FIMA)
Fast fluence (1025 n m2)
Heating test Temperature ( C)
Duration (h)
1600 1600 1600 1600 1600 1600 1600 1700 1600 1800 1800 1800 1800 1800
500 500 160 50 100 50 500 185 138 100 100 200 90 175
AVR 71/22 HFR-K3/1 FRJ-K13/2 AVR 88/33 AVR 82/20 AVR 88/15 AVR 82/9 AVR 74/11 FRJ-K13/4
3.5 7.5 8.0 8.5 8.6 8.7 8.9 6.2 7.6
0.9 4.0 0.1 2.3 2.4 2.4 2.5 1.6 0.1
HFR-K3/3 AVR 76/18 AVR 74/10 AVR 70/33
10.6 7.1 5.5 1.6
6.0 1.9 0.9 0.4
Table 58 Fuel compact
HFR-P4/3/7 HFR-P4/1/8 HFR-P4/2/8 HFR-P4/1/12 SL-P1/6 SL-P1/10 SL-P1/9 HFR-P4/3/12
Fractional release
85
137
4 107 2 106 6 107 1 107 2 107 6 108 5 107 3 105 3 107 7 105 7 104 1 104 2 103 2 103
2 105 1 104 4 105 1.2 104 6 105 1 105 8 104 8 105 3 106 1 102 6 102 5 102 1 101 2 102
Kr
Cs
Results of accident simulation tests at 1600–1800 C with irradiated fuel compacts Irradiation conditions
Heating test
Fractional release
Burnup (% FIMA)
Fast fluence (1025 n m2)
Temperature ( C)
Temperature ( C)
Duration (h)
13.9 13.8 13.8 11.1 10.7 10.3 10.7 12.0
7.5 7.2 7.2 5.5 6.7 6.0 6.3 5.5
1075 940 945 940 800 800 800 1075
1600 1600 1600 1600 1600 1700 1700 1800
304 304 304 304 304 304 304 279
Source: Schenk, W.; Nabielek, H. Nucl. Technol. 1991, 96, 323–336.
85
Kr
1 103 5 105 8 105 5 107 7 107 9 105 4 105 1 103
137
Cs
4 103 2 103 1 103 3 104 4 104 6 102 1 101 5 101
TRISO-Coated Particle Fuel Performance
100
197
10–2 Compact Fuel element
2100 ⬚C 10–1 10–3
1800 ⬚C 10–4
Level of one particle failure 1700 ⬚C
fractional release
10–3
85Kr
85Kr
fractional release
10–2
10–5
10–6
10–4 14% FIMA 10–5
10–6
8–10% FIMA
1600 ⬚C 10–7
10–7
10–8 (a)
0
100
200
300
Heating time (h)
400
10–8
500
0
100
200
Heating time (h)
(b)
10–2
1800 ⬚C
85Kr
fractional release
10–3
1700 ⬚C
10–4
10–5
1600 ⬚C
10–6
10–7 (c)
0
100 200 Heating time (h)
300
Figure 33 Accumulated fractional release of 85Kr as a function of heating time at constant temperature; (a) spherical elements at 1600–1800 C, (b) compacts of 8–14% FIMA at 1600 C, and (c) compacts in the range 10–12% FIMA at 1600–1800 C. Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
300
198
TRISO-Coated Particle Fuel Performance
with defective SiC, while, in both cases, the pyrocarbon layers remain intact. Integral fractional releases of silver, cesium, krypton, and strontium are shown as a function of time at 1600 C for sphere HFR-K3/1 in Figure 35. The release of silver is on the order of 1–2% at the outset
10–03 AVR 88/33, 8.5 AVR 82/20, 8.6 AVR 71/22, 3.5 AVR 88/15, 8.7
137Cs release 85Kr release
10–04
Release fraction
One particle, 100% release 10–05
*including 100 ⬚C margin
10–06
10–07
10–08 –100
0
100
200
300
Time (h)
Figure 34 137Cs and 85Kr releases at 1600 C for AVR spherical fuel elements. Reproduced from Kendall, J.K.; et al. PBMR fuel performance envelope and test program, Pebble Bed Modular Reactor Limited, Document No. 039116, 2007.
of heating as considerable silver was released from fuel particles to the matrix during irradiation at temperatures in the range 1000–1200 C for 359 days. The release of cesium is considerably greater than the release of krypton and strontium, because krypton is held up by the PyC and strontium is retained in the UO2 kernel and graphite matrix to a greater extent than cesium. In Figure 35, the radial profiles of silver and cesium in the graphite matrix exhibit strong concentration gradients typical of materials that are diffusing, whereas the strontium profile is much more flat, indicating little diffusive release from the matrix. In this same figure, the release of cesium is observed to climb strongly after 200 h of heating, possibly indicating a deterioration of the SiC, perhaps from fission product palladium attack of SiC grain boundaries. Cesium releases as a function of time at temperature are shown in Figure 36, where it can be seen that five compacts with burnup in the range 10.7–13.9% FIMA exhibit higher releases than five spherical fuel elements with burnup in the range 3.5–8.9% FIMA at 1600 C. The results plotted at 1700 and 1800 C are from compacts in Table 58. The reason for the increased releases at 1600 C at high burnup and/or fluence is not well understood, but has been attributed to increased permeability of the SiC layer, related perhaps to fission product attack during postirradiation heating, especially at longer times (e.g.,
HFR-K3/1.75% FIMA 25 –2 100 4.0 ⫻ 10 m (E > 16 fJ) 1020–1200 ⬚C
10–1
Fractional of fission products in matrix
100
110mAg
Fractional release
10–2 10–3 10–4 10–5 137Cs
10–6 85Kr
10–7
90Sr
10–8 0
100 200 300 400 Heating time at 1600 ⬚C (h)
500
10–1 10–2
90Sr 137Cs
10–3 10–4
110mAg
After 500 h at 1600 ⬚C
10–5 10–6 Fuel zone 10–7 10–8 –30 –10 0 10 30 Fuel element radius (mm)
Figure 35 Fission product release and distribution in sphere HFR-K3/1 after irradiation at 1000–1200 C for 359 days and 1600 C heating. Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
TRISO-Coated Particle Fuel Performance
100
1800 ⬚C
the range 1900–2500 C is thermal decomposition of the SiC layer.50
10–1
1700 ⬚C
3.07.3.3
fractional release
10–2
137Cs
199
1600 ⬚C
10–3 Five compacts 10–4
1600 ⬚C
10–5 Five fuel elements
10–6
10–7 0
50
100
150 200 250 Heating time (h)
300
350
Figure 36 Cesium release during heating of spherical fuel elements (1600 C) and compacts (1600–1800 C). Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
European Experience
Several high-temperature accident tests have been performed at the new Ku¨FA installation at ITU.51 Prior to these tests, the upgraded system was tested to verify furnace operation, fission gas system performance, and condensation plate collection efficiency. It was found that the collection efficiency of the plate for Cs was 70% (16%), which is in agreement with the efficiency determined previously at Ju¨lich during earlier work with the furnace.38 The new accident tests on irradiated fuel include several fuel elements from the AVR reactor, from the HFR K6 irradiation test (proof test for HTR Modul reactor fuel element design), and from the more recent EU1bis irradiation test. Table 60 provides information and the irradiation history of the AVR and HFR K6 fuel elements tested to date. Because of the age of the fuel tested, no 110mAg data were collected in the latest accident tests. Testing on four HFR-EU1bis spheres has been completed, but data from these experiments have not yet been published. 3.07.3.3.1 AVR 73/21
cesium at times greater than 200 h in Figure 35, and temperatures above 1600 C). Ceramographic sections shown in Figure 37 show evidence of increasing degradation in the SiC layer for longer times at 1600 C and higher burnup, the most degraded being the SiC in sphere HFR-K3/1. Microprobe profiles through particles after heating, as shown in Figure 38, show the buildup of fission product palladium at the IPyC/SiC interface in spheres HFR-K3/1 and HFR-K3/3. It is hypothesized that corrosion by palladium degrades the SiC, leading to accelerated diffusion of cesium through grain boundaries.45 The distribution of metallic fission products averaged over a number of UO2 TRISO fuel element spheres examined after accident testing is shown in Table 59. At 1600 C,49 the kernel and the coatings are equally important in holding up cesium, while the kernel is the principal reservoir for strontium and silver. At 1800 C, cesium and silver are contained principally in the coatings, whereas strontium is retained in the kernel. The primary mechanism for coating failure and fission product release at extreme temperatures in
Testing of the AVR 73/21 fuel element was the first use of the upgraded Ku¨FA with irradiated fuel and was primarily conducted as a shakedown test of the equipment under hot conditions. The sphere was heated to 1600 C for 5 h, followed by a ramp to 1800 C and hold for an additional 5 h. The 85Kr release during this relatively short test was below the detection limit of the system. Two cold plates were exchanged during the test, but a problem was encountered with 137Cs contamination of the plate container from the hot cell, which interfered with analysis of released 137Cs on the plate and prompted a change in procedures to eliminate the carry-over contamination issue on subsequent tests. 3.07.3.3.2 AVR 74/18
The second accident test was performed on the AVR 74/18 sphere. The test took place in two parts: the first phase involved heating at 1600 C for 100 h, followed by a second phase of heating at 1800 C for an additional 100 h. Unplanned temporary furnace shutdowns were encountered during each heating cycle (one per cycle) due to operational issues with the furnace system, but
200
TRISO-Coated Particle Fuel Performance
100 mm
1600 ⬚C, 160 h
50 mm
FCs1374⫻10–5(FRJ2–K13/2;8%FIMA)
100 mm
1600 ⬚C, 500 h
1600 ⬚C, 500 h
20 mm
20 mm
FCs1372⫻10–5(AVR71/22;3.5%FIMA)
FCs1371⫻10–4(HFR–K3/1;7.7%FIMA)
Figure 37 Ceramographic sections through particles heated at 1600 C (complete particle followed by enlarged views of SiC layers) showing increasing degradation of the SiC layer with increasing burnup. Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30.
these do not appear to have significantly affected the test results. No particle failures were detected during this test. Fractional releases at the end of the test were 6 106 for 85Kr and 8 106 for 137Cs. 3.07.3.3.3 HFR K6/3
This heating test was carried out in four stages: 1600 C for 100 h, 1700 C for 100 h, 1800 C for 100 h, and 1800 C for 300 h. The temperature was returned to room temperature in between stages. 85Kr fractional release was low (<105) until the start of the fourth heating cycle, when a drastic increase in 85Kr was observed. By the end of the fourth heating cycle, the 85Kr fractional release was 4 104, indicating several particle failures. The 137Cs fractional release was low (106) after the first two heating cycles but increased dramatically during the third
cycle to a value of 1 103. The release further increased during the second 1800 C heating cycle, ending with a final value of 0.03. The results of the heating test are shown in Figure 39. SEM characterization of the particles at the end of the test revealed failed particles with cracked SiC layers. The results indicate good particle performance at up to 1700 C, with significant Cs release through intact coatings starting at 1800 C and multiple particle failures starting with the second 1800 C cycle. It is unclear if thermal cycling was responsible for the observed particle failures. 3.07.3.3.4 HFR K6/2
This heating test was performed in two stages: the first was at 1600 C for 100h followed by a second stage at 1800 C for 200 h. The temperature was not cycled to
TRISO-Coated Particle Fuel Performance
10–1 33
36
34
31
28
18
18
SiC
10 78
10–2
45
74
Parts of weight
41
21
48 36
78
50
78
77
Ba 10–3 Ru
Ag
Ru
10–4
13
26
SiC
LTI Buffer Buffer LTI layer layer
24 41
PyC Kernel PyC
201
81 26
78
47
34
30 10–5
39
30 40 50 60 70 80 10 20 Measure points radial through coating
0
(b) 10–1
SiC SiC PyC PyC Kernel PyC PyC LTI Buffer Buffer LTI LTI layer layer Cs Pd Pd
(a)
10–2 Parts of weight
Parts of weight
10–1 PyC SiC PyC Kernel PyC SiC PyC LTI LTI Buffer Buffer LTI LTI layer layer Cs
10–2 Pd
Pd
Cs Cs
10–4
10–3
10–4
10–5 0
(c)
10–3
10 20 30 40 50 60 70 80 Measure points radial through coating
0 (d)
10 20 30 40 50 60 70 80 Measure points radial through coating
Figure 38 Microprobe profiles of fission product elements through coatings of particles from HFR-K3. (a) Arrangement of sectioned particles (HFR-K3/3) for microprobe measurements. The numbers show the percentage loss of cesium from each particle after heating at 1800 C. (b) Barium, ruthenium, silver profiles in a particle with 78% cesium loss after 1800 C test. (c) Cesium, iodine, palladium profiles in a particle from HFR-K3/1 (0.01% cesium loss from sphere) after 1600 C test. (d) Cesium, iodine, palladium profiles in a particle (78% cesium loss) from HFR-K3/3. Reproduced from Schenk, W.; et al. J. Nucl. Mater. 1990, 171(1), 19–30. Table 59 Fuel type
UO2 TRISO
Averaged fission product distribution for spherical fuel elements after accident simulation tests Heating temperature ( C)
Time at temperature (h)
1600
Up to 500
Nuclide
137
Cs Sr 110m Ag 137 Cs 90 Sr 110m Ag 90
UO2 TRISO
1800
Up to 200
Fractional fission product content in: Kernel
Coating
Matrix
5 101 9.5 101 8 101 2 102 7 101 9 102
5 101 5 102 2 101 6 101 8 102 2 101
2 105–1 103 2 103–5 103 9 104 1.5 101 2 101 3 102
Fractional release from fuel element 5 105 1 106 1 103–3 102 5 102 3 103 7 101
Source: Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, USA, October 1–4, 2008.
202
TRISO-Coated Particle Fuel Performance
Fuel description/irradiation conditions for spheres accident tested in Ku¨FA at ITU
Table 60 Fuel element
AVR 73/21
AVR 74/18
HFR K6/3
HFR K6/2
Type Graphite Fuel particles Heavy metal (g) Fuel composition Enrichment (% 235U) Burnup (% FIMA) Irradiation time (EFPD) End of irradiation Irradiation temperature ( C) Thermal fluence (cm2) Fluence, E > 0.1 MeV (cm2)
GLE-3 A3-27 16 400 10 UO2 9.82 2.5 235 February 1984 700 2 1021 4 1020
GLE-3 A3-27 16 400 10 UO2 9.82 4.8 480 February 1985 700 4.15 1021 8 1020
GLE-4 similar A3-27 14 600 9.44 UO2 10.6 9.7 633 May 1993 1140 2.5 1021 4.8 1021
GLE-4 similar A3-27 14 600 9.44 UO2 10.6 9.3 633 May 1993 1140 2.5 1021 4.6 1021
Source: Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, Oct 1–4, 2008.
3.07.3.4
US Experience and Future Plans
0.1
2000 1800
1400
1E-3
1200 1E-4
1000 800
1E-5
600 Temperature (⬚C) Kr 137 Cs
400
85
200 0 0
100
200
300
400 500 600
700
Fractional release
0.01
1600 Temperature (⬚C)
room temperature at the end of the 1600 C test, but several interruptions of the experiment during the first stage occurred because of system malfunctions. The results of this heating test are shown in Figure 40. The 85Kr fractional release was below the detection limit of the system until 90 h into the 1800 C cycle, when the release increased up to 107. Approximately 120 h into the 1800 C cycle, another continuous release was observed that reached approximately 105 by the end of the test. 137Cs was released fairly continuously throughout the test up to a 2 103 fractional release at the end of the test. In general, these initial tests demonstrated excellent fuel performance at 1600 and 1700 C with zero particle failures. Higher 137Cs and 85Kr releases were observed at 1800 C, with some particle failures occurring for the sphere HFR K6/3.
1E-6 1E-7 800
Time (h)
Figure 39 Results of heating test HFR K6/3 at ITU. Reproduced from Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, USA, October 1–4, 2008.
3.07.3.4.1 Past experience
Very little work has been done on accident testing of US fuels as most irradiations produced significant irradiation-induced particle failure fractions. However, postirradiation anneals of long durations at temperatures up to 1500 C were performed to accelerate diffusive fission product releases from a variety of fuel types.52 UO2, UC2, UCO, UO2*(1), and UO2*(2) fuel particles were irradiated in the HRB-15B capsule in HFIR.52 In the UO2*(1) fuel, the kernel was coated with a ZrC layer, while in the UO2*(2) fuel, ZrC was dispersed in the buffer layer surrounding the kernel. These changes were intended to control free oxygen released during fission, which should improve fission product retentiveness. The fuel
burnup was in the range of 21–25% FIMA and the fast neutron fluence was in the range of 3.4 – 5.5 1025 n m2. The irradiation was accelerated with a residence time of 169 effective full power days. Only the fission product release data at 1500 C are discussed here, as cesium was not released at the lower temperatures. No fission product releases were measured at any temperature from UO2*(1) fuel particles. Ten particles of each fuel type were annealed for 11 866 h at 1500 C. Integral releases for each 10-particle batch were measured from individual particles by gamma counting each particle before and after the test and by periodic gamma monitoring of fission product collectors
TRISO-Coated Particle Fuel Performance
during the anneal as a function of time. The agreement of the integral releases from each 10-particle batch by these two methods was excellent. Cesium was released from only the UO2 and UC2 fuel particles as is shown in Figure 41. These same two fuel batches released the greatest fractions of silver as illustrated in Figure 42. The time signatures
2000
0.01
1800
Temperature (⬚C)
1400
1E-4
1200 1E-5
1000 800
1E-6
600 400
Fractional release
1E-3
1600
1E-7
200 1E-8
0 0
50
100
150 200 250 Time (h)
300
350
Temperature (⬚C) 85
Kr Cs
137
Figure 40 Results from heating test HFR K6/2 at ITU. Reproduced from Freis, D.; et al. Post irradiation test of high temperature reactor spherical fuel elements under accident conditions. In 4th International Topical Meeting on High Temperature Reactor Technology, Washington, DC, USA, October 1–4, 2008.
203
of the releases of cesium and silver from the UO2 fuel particles in Figures 41 and 42 indicate a diffusion release mechanism through the SiC layer. However, the release of cesium from the UC2 fuel batch is sudden in Figure 41, and the release of silver shows a rapid increase at the same time as the sudden release of cesium, as pointed out in Figure 43. The distribution of fission product releases among particles within the fuel batches in Table 61 indicates that the release of cesium from the UO2 fuel particles is from two of the ten particles and from only one particle in the UC2 fuel batch. This same table shows that the release of silver was 100% from the UO2 fuel batch, and 82% from the UC2, with 6 of the 10 UC2 particles releasing 100% of their silver inventories, two particles releasing 85–95%, one particle releasing 50%, and one particle retaining 100%. Particle-toparticle variations in fission product release dominate the behavior implied by the data shown in Table 61. The microstructures in Figure 44 show that the SiC layer in the UO2 batch exhibits large columnar grains, whereas the UCO batch exhibits a strong laminar grain structure in the SiC. The UC2 and UO2*(1) batches exhibit laminar structures in the SiC that are somewhat weaker than in the UCO batch. The results in Table 61 indicate that silver release at 1500 C is greatest (100%) in the case of columnar SiC, least (3%) for strongly laminar SiC, and intermediate (82%) for somewhat less strong laminar SiC. Although Cs was released from only three of the 50 particles annealed at 1500 C, two
Cesium release (%)
102
101
100
134Cs 137Cs
10−1 0
10
20 30 Annealing time (Ms)
40
50
Figure 41 Release of Cs from various types of TRISO-coated fuel particles at 1500 C. Reproduced from Bullock, R. E. J. Nucl. Mater. 1984, 125(3), 304–319.
204
TRISO-Coated Particle Fuel Performance
102
UO2 UC2
Silver release (%)
UO2* (2) 101
Release from UO2* (1) = 0 UCO
100
10−1 0
10
20 30 Annealing time (Ms)
50
40
Figure 42 Release of 110Ag from various types of TRISO-coated fuel particles at 1500 C. Reproduced from Bullock, R. E. J. Nucl. Mater. 1984, 125(3), 304–319.
102
Ag
Cs
Release (%)
101
100
10−1
0
10
20 30 Annealing time (Ms)
40
50
Figure 43 Abrupt 10% increase in 110Ag release from UC2 particles at 1500 C when one of the ten test particles released its entire Cs inventory.
particles had columnar SiC and one had a somewhat weak laminar SiC. The sensitivity of cesium release to SiC grain structure was recognized in Myers53 where the diffusivity of cesium through columnar SiC was given as an order of magnitude greater than through laminar SiC. The diffusion equations from Myers53 are accessible in Table A-3 of IAEA.3 As shown in Figure 45, releases of europium are greatest (37–46%) for the fuel batches containing UC2 in the kernel, compared with fuel batches
containing only UO2 in the kernel (9–16%). As shown in Table 61, cerium release is 45% in UC2, only 1% in UCO, and nil in UO2 particles. These behaviors are related to the thermodynamics of rareearth oxides and carbides, according to Homan et al.,54 where oxides formed in UO2 (such as Eu2O3 and Ce2O3) are less likely to escape from the kernel than are the more mobile rare-earth carbides formed in UC2. In UCO, europium forms a carbide and cerium forms an oxide.54
TRISO-Coated Particle Fuel Performance
Table 61
205
Distribution of fission product release within particle batches during postirradiation annealing
Annealing temperature ( C)
TRISO particle typea
1500
UC2 with laminar SiC
9 ¼ 0% 1 ¼ 99% 10 ¼ 12%
1500
UO2 with columnar SiC
1500
UC0.4O1.6 with laminar SiC
8 ¼ 0% 2 ¼ 99% 10 ¼ 24% 10 ¼ 0%
1500
UO2*(2) with laminar SiC
10 ¼ 0%
1350 1350
UC0.4O1.6 with laminar SiC UO2*(2) with laminar SiC
10 ¼ 0% 10 ¼ 0%
1200 1200
UC0.4O1.6 with laminar SiC UO2*(2) with laminar SiC
10 ¼ 0% 10 ¼ 0%
Release breakdown from the ten particles within a test batch for: 134 137
Cs Cs
110m
Ag
1 ¼ 0% 1 ¼ 50% 85% < 2 < 95% 6 ¼ 100% 10 ¼ 82% 10 ¼ 100% 7 ¼ 0% 10% < 3 < 20% 10 ¼ 3% 7 ¼ 0% 70% < 3 < 80% 10 ¼ 27% 10 ¼ 0% 7 ¼ 0% 45% < 3 < 55% 10 ¼ 19% 10 ¼ 0% 10 ¼ 2%b
154
144
15% < 5 < 25% 45% < 3 < 55% 2 ¼ 100% 10 ¼ 46% Uniform release of 16%
12% < 3 < 18% 18% < 3 < 25% 70% < 3 < 80% 1 ¼ 99% 10 ¼ 45% 10 ¼ 0%
Uniform release of 37%
10 ¼ 1%b
2 ¼ 0% 0% < 5 < 10% 15% < 3 < 25% Uniform release of 23% 10 ¼ 4%b
10 ¼ 0% 10 ¼ 0% 10 ¼ 0%
Uniform release of 6% 10 ¼ 0%
10 ¼ 0% 10 ¼ 0%
Eu
Ce
There was zero release within about 5% as determined from individual particle counting before and after annealing for all isotopes from each of the ten particles in all test combinations not listed, that is, UO2*(1) at all temperatures, and UC2, UO2, and UC0.4O1.6 at 1350 and 1200 C. As no release on collectors was detected at the 0.01% level from the combined ten particles within each of these test batches, it can be assumed that release from any one of these particles was certainly <0.01% and was probably not more than 0.001%. b These total releases from ten particles were too small and too uniformly distributed among particles to allow the determinations of individual release values. a
These results,42 with admittedly relatively few particles, indicate that under a long annealing time at 1500 C (a) silver and cesium releases are at a maximum in the case of SiC with a columnar grain structure; (b) europium releases are largest in UC2 fuels, but can be significant in UCO and, to a lesser extent, in UO2 fuels; and (c) cerium release is significant only in UC2 fuel. 3.07.3.4.2 Future plans
The NGNP/AGR fuel program is currently planning an extensive campaign of accident testing of irradiated TRISO fuel compacts to measure fission product retention at elevated temperatures and provide verification of fuel performance code predictions. The first accident tests will be performed on irradiated fuel compacts from the AGR-1 experiment, using both the FACS and CCCTF test equipment described earlier. Individual fuel compacts in each capsule will be chosen for accident tests based primarily on their irradiation history (i.e., burnup, fast fluence, and irradiation temperature). Both isothermal and transient temperature tests will
be performed on the AGR-1 fuel with maximum temperatures between 1600 and 1800 C. Hightemperature margin tests up to 2000 C may also be performed if warranted, based on the results of the lower temperature tests. The primary released radionuclides of interest are 85Kr, 137Cs, 134Cs, 110mAg, 90 Sr, and 154Eu. (131I and 133Xe are also of interest, but will require reirradiation of the fuel in a test reactor to regenerate these isotopes because of rapid decay times. Work is underway to develop this capability at INL.) Condensate plate analysis will consist of initial measurement of sample activity using gamma spectrometry. Following this step, the fission products will be chemically stripped from the plates, and the solutions analyzed for other fission products of interest. The primary fission product of interest will be 90 Sr, a beta emitter. The strontium can be separated from the rest of the analytes in an ion exchange resin, and the 90Sr activity measured using scintillation techniques. The feasibility of chemical separation of iodine from the solution has also been demonstrated, and the separated iodine solution can be analyzed for
206
TRISO-Coated Particle Fuel Performance
(a)
UCO
(b)
UO2
(c)
UC2
(d)
UO*2(1)
20 mm
Figure 44 Microstructures of etched SiC barrier layers in various types of TRISO-coated particles. Reproduced from Bullock, R. E. J. Nucl. Mater. 1984, 125(3), 304–319.
131
I using gamma spectroscopy. Alternatively, 90Sr, as well as other radioisotopes of interest, can be analyzed using mass spectrometric methods. In preparation for this accident testing campaign, the FACS furnace system will undergo remote qualification testing, fission gas system performance checkout, and integrated testing with the FACS furnace system, condensation plate collection efficiency calibrations, and installation in the hot cells at the INL HFEF. Collection efficiency calibration of the upgraded CCCTF deposition cup will also be performed, and the results compared with those of the FACS furnace so that the performance of both testing systems can be compared using known standards.
PIE of fuel compacts after accident testing will involve a number of methods designed to study fission product retention of the particles during the heating cycles in further detail and to document the condition of the fuel kernels and coating layers after the high-temperature testing. This will include microscopy on compact and particle cross-sections to examine fuel kernel and coating microstructures as well as evidence of damage; deconsolidation of compacts; leach-burn-leach analysis to determine SiC coating failure fractions and to measure fission products held up in the compact matrix material (determined during the first leach step); and gamma spectrometry of individual fuel particles to study metallic fission product retention behavior and search for failed particles which can be used in subsequent analysis to determine failure modes. Accident testing in inert atmospheres on irradiated fuel from subsequent AGR experiments (a total of eight different irradiation experiments are currently planned16) will also be performed with various objectives, including testing performance of fuel produced in large scale coaters, studying fission product transport at high temperatures, and performing proof tests to qualify the reference fuel for licensing. In addition, the program has plans to perform air and water ingress accident testing on irradiated fuel to verify fuel performance in these more reactive environments. The facilities to perform the air/water tests have not yet been developed. 3.07.3.5
Japanese Experience
Individual UO2-TRISO fuel particles deconsolidated from a compact of Japanese manufacture were heated at 1700 C for 270 h and 1800 C for 222 h at ORNL.42 The fuel had been irradiated in HFIR in the HRB-22 capsule to a burnup of 4.8% FIMA and a fast neutron fluence of 2.1 1025 n m2 for a duration of 89 effective full-power days (EFPD) and a timeaveraged maximum temperature of 1100 C. Releases of silver, cesium, europium, and krypton were measured as a function of time as shown in Figures 46 and 47 for batches of 25 particles at each temperature. The krypton release in Figure 47 shows that one particle failed early in the heating, also releasing antimony. Fission product inventories of the individual fuel particles were measured before and after the heating tests with the IMGA apparatus. Results of these measurements are shown for silver, cesium, and europium in Figure 48 (1700 C) and Figure 49 (1800 C). Both these figures exhibit large variations in fission
TRISO-Coated Particle Fuel Performance
207
102
Europium release (%)
UC2 UCO UO2 UO2* (2)
101
100 Release from UO2* (1) = 0
10−1 0
10
20 30 Annealing time (Ms)
50
40
Figure 45 Release of 154Eu from various types of TRISO-coated fuel particles of 1500 C. Reproduced from Bullock, R. E. J. Nucl. Mater. 1984, 125(3), 304–319.
100
100
110 mAg 110mAg
10–2
154Eu 137Cs
10–3
85Kr
10–4
10–5
137Cs
10–1 Fractional release
Fractional release
10–1
125Sb
10–2
85Kr
154Eu
10–3
10–4
0
50
100
150
200
250
300
Heating time (h)
10–5
0
50
100 150 Heating time (h)
200
250
Figure 46 Time-dependent fractional releases of fission products during the ACT3 heating test at 1700 C for 270 h, obtained by the online measurements of fission gas-release and intermittent measurements of metallic fission product release. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
Figure 47 Time-dependent fractional releases of fission products during the Act 4 heating test at 1800 C for 222 h, obtained by the online measurements of fission gas-release and intermittent measurements of metallic fission product release. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
product release from particle to particle. At 1700 C, silver release varies from 10% to 100%, cesium from 0% to 20%, and europium from 5% to 30%. At 1800 C, not including the failed particle, the silver release varied from 24% to 100%, cesium from 0% to 55%, and europium from 0% to 69%. Individual particles were examined at JAERI by
X-ray microradiographs, ceramography, and electron microprobe. Accumulations of fission products in the buffer show up as bright spots in X-ray microradiographs of fuel particles from which large fission product releases were measured, as shown in Figure 50 (1700 C) and Figure 51 (1800 C). In Figure 50,
TRISO-Coated Particle Fuel Performance
Inventory ratio of post-/preheating tests
208
1 0.8
0.6 0.4
144Ce 95Zr 106Ru
0.2
125Sb
0 0
5
10 15 Particle number
Inventory ratio of post-/preheating tests
(a)
20
25
20
25
1 0.8
0.6 0.4 110mAg 137Cs
0.2
154Eu
0 0
5
10 15 Particle number
(b)
Figure 48 Inventory ratios of post- to preheating tests in individual particles in ACT3 measured with the IMGA system: (a) 95Zr, 106 Ru, 125Sb, and 144Ce; (b) 110mAg, 137Cs, and 154Eu. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47. 1.2
1.0 0.8 0.6 0.4 0.2 0
(a)
Inventory ratio of post-/ preheating tests
Inventory ratio of post-/ preheating tests
1.2
144Ce
0
5
95Zr
10
106Ru
15
125Sb
20
Particle number
(b)
137Cs
154Eu
0.8 0.6 0.4 0.2 0 0
25
110mAg
1.0
5
10
15
20
25
Particle number
Figure 49 Inventory ratios of post- to preheating test in individual particles in ACT4 measured with the IMGA system: (a) 95Zr, 106 Ru, 125Sb, and 144Ce; (b) 110mAg, 137Cs, and 154Eu. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
particle ACT3–5 had relatively little fission product release and shows no evidence of fission product accumulation in the buffer, whereas particle ACT3–6 had relatively large releases and shows a
bright spot in the buffer. A similar trend is shown in Figure 51 for a relatively non-releasing particle (ACT4–3) and two strongly releasing particles (ACT 4–9 and ACT 4–13). These figures also show
TRISO-Coated Particle Fuel Performance
(a)
100 mm
(b)
100 mm
(c)
100 mm
(d)
100 mm
209
Figure 50 X-ray microradiographs and ceramographs of the particles after the ACT3 heating test: (a) and (b) show particle ACT3-5; (c) and (d) show particle ACT3-6. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
that the buffers of the low-releasing particles are intact, whereas those of the heavily releasing particles are severely cracked. In addition, the kernels of the releasing particles in Figures 50 and 51 exhibit larger pores than the kernels of the nonreleasing particles. The SiC layers of all the particles show signs of degradation as seen in Figures 50–52. Accumulations of fission products, especially palladium, were found in all particles at the IPyC/SiC interface, as shown in Figure 53, and sometimes within the SiC layer. The Japanese work corroborates results from Germany in that wide variations are measured in fission product release from particle to particle, palladium buildup at the IPyC/SiC interface is observed in both releasing and nonreleasing particles, and fission product releases increase with increasing test temperature.
3.07.4 Conclusions TRISO-coated particle fuel has been fabricated and tested around the world. A review of the fuel performance database indicates that high-quality lowdefect TRISO fuel, the heart of the high-temperature
gas-cooled reactor, is achievable if the following conditions are met: a. Disciplined and controlled fabrication of the coated particles under conditions that produce the proper microstructure of the kernels and coatings with robust quality control to demonstrate the statistical adequacy of the pedigree of the material. Much of the poor performance observed in the past is associated with improper fabrication conditions. b. Selection of normal operation conditions that will result in very low failures. For a pebble bed reactor using 500-mm UO2 TRISO fuel, an envelope of 1150 C, 10% FIMA, and 4 1025 n m2 was demonstrated by German fuel developers in the 1980s and remains the standard for current pebble bed designs. For prismatic gas-cooled reactors using 350-mm UCO TRISO fuel, an envelope of 1250 C, 19% FIMA, and 4 1025 n m2 appears achievable, on the basis of the results of the US AGR-1 irradiation. c. Well-planned series of irradiations that limit the burnup and fluence acceleration to preclude artificial failures that would not be expected in the actual reactor application, yet contain enough fuel
210
TRISO-Coated Particle Fuel Performance
(a)
(c)
(e)
100 mm
100 mm
100 mm
(b)
(d)
(f)
100 mm
100 mm
100 mm
Figure 51 X-ray micrographs and ceramographs of the particles after the ACT4 heating test: (a) and (b) show particle ACT4-3; (c) and (d) show particle ACT4-9; and (e) and (f) show particle ACT4-13. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
to draw statistically significant conclusions about fuel performance. To meet an in-service failure probability of 104 at 95% confidence, the minimum number of particles to be tested is approximately 300 000, allowing for some failures during the experiment. d. Selection of an accident envelope that will result in very low failures. For both pebble bed and prismatic designs, limiting accident temperatures in depressurized conduction cooldown scenarios to 1600 C is acceptable. The large accident database developed by German fuel developers in the 1980s remains the standard for gas reactors worldwide. There has been very limited accident testing
for UCO TRISO, but work is underway to begin such testing in the near future. The current accident database at 1700 and 1800 C is inadequate to determine with statistical confidence if TRISOcoated UO2 or UCO can sustain these higher accident temperatures. Future R&D programs will investigate the fuel performance at these elevated temperatures. e. Well-planned accident safety testing that tests enough fuel to draw statistically significant conclusions (population on the order of 200 000 particles) and measures all relevant radionuclides important to the ultimate safety and licensing of the high-temperature gas-cooled reactor.
TRISO-Coated Particle Fuel Performance
211
The world TRISO-coated particle fuel database suggests that these goals are attainable. See Chapter 2.10, Graphite: Properties and Characteristics; Chapter 3.06, TRISO Fuel Production; Chapter 3.08, Advanced Concepts in TRISO Fuel; Chapter 3.24, TRISO Fuel Performance Modeling and Simulation; and Chapter 4.07, Radiation Effects in SiC and SiC-SiC.
20 mm
(a)
(b)
20 mm
(c)
20 mm
Acknowledgments
Figure 52 Ceramographs of coating layers of the particles after the ACT4 heating test: (a) shows particle ACT4-3, (b) shows particle ACT4-9, and (c) shows particle ACT4-13. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
This manuscript was authored by Battelle Energy Alliance, LLC, under Contract Number DE-AC0705ID14517 with the US DOE. The US Government retains and the publisher, by accepting the article for publication, acknowledges that the US Government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US Government purposes. This information was prepared as an account of work sponsored by an agency of the US Government. Neither the US Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. References herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise do not necessarily constitute or imply its endorsement, recommendation, or favoring by the US Government or any agency thereof. The views and opinions of the authors expressed herein do not necessarily state or reflect those of the US Government or any agency thereof.
SEI
(a)
10 mm
Pd
(b)
Figure 53 Electron probe microanalysis of coating layers of particle ACT4-3 after the ACT4 heating test: (a) secondary electron image and (b) X-ray image for palladium. Reproduced from Minato, K.; et al. Nucl. Technol. 2000, 131(1), 36–47.
212
TRISO-Coated Particle Fuel Performance
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