Absorber materials for Generation IV reactors

15

Absorber materials for Generation IV reactors D. Gosset DEN, CEA Saclay, Université Paris Saclay, Gif sur Yvette, France

15.1

Introduction: neutron absorbers for Generation IV reactors

The different reactivity control systems in a nuclear power plant allow keeping at any time the control of the nuclear fission reactions in the core: power steering, safe reactor shutdown, wear compensation of the fuel. They are also part of the neutron protection of the out-of-core components. These systems can take various forms: gas (such as helium 3 in some experimental reactors), liquid (soluble boron in pressurized water reactor (PWR) coolant to balance the reactivity evolution of the reactor), and most of the time solid (Table 15.1). In a reactor, they are most often combined [e.g., in PWR with Ag-In-Cd (AIC) plus boron carbide control rods and with boron present both as soluble boron and as boron carbide]. In all cases those materials incorporate neutron-absorbing nuclides, unlike the fuel which is a medium generally multiplier Table 15.1 Basic properties of current and potential neutron absorber materials

Material

Density (g/cm3)

Melting temperature (8C)

Thermal conductivity RT (W/m$K)

Mechanical properties

2430

28

Brittle

Boron carbide

2.52 (100% dense, 10B natural)

AIC

10.15

800

55

Metal, recrystallization temperature 275
CYoung modulus w50 MPa

Hf

13.3

2156

22

Metallic

Dy2TiO5

4.8 (100% dense)

1870

30

Ceramic, brittle

HfB2

10.5

3380

75

Metallic, brittle

HfO2

9.7

2900

2

Ceramic, brittle

Structural Materials for Generation IV Nuclear Reactors. http://dx.doi.org/10.1016/B978-0-08-100906-2.00015-X Copyright © 2017 Elsevier Ltd. All rights reserved.

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Structural Materials for Generation IV Nuclear Reactors

of neutrons. The materials under which those nuclides are present are usually called neutron absorber materials, whereas the core components in which these nuclides are present have different denominations, depending on the reactor family. The purpose of this chapter is the description of the absorber materials used in the control rods of the main types of present reactors in the perspective of their use in the Generation IV projects. Burnable poison or nonsolids absorbers are not considered here.

15.1.1

The Generation IV project: a short presentation

A comprehensive presentation of the Generation IV project is made in Chapter 1. Here, we only give the main characteristics of the planned systems in order to draw the main requirements the neutron absorber materials should fulfill. As compared to generation II and III reactors, the Generation IV projects are intended to propose breakthrough solutions to safety, competitiveness, sustainability, nonproliferation, and waste management [1]. In that frame, and taking into account the previous experience of the GIF forum participants, six different systems are considered: • • • • • •

Gas-cooled fast reactor (GFR); Lead-cooled fast reactor (LFR); Molten salt-cooled reactor (MSR); Sodium-cooled fast reactor (SFR); Supercritical water-cooled reactor (SCWR); Very-high-temperature reactor (VHTR).

Some of their characteristics (temperature, thermal volume power, neutron spectra, cooling media, etc.), which strongly impact the neutron absorber material choice and the Control Element Assemblies (CEA) design, are reported in Table 15.2. The relations between the systems characteristics and the CEA choice and design are as following: • • • •

Core power density: number and yield of the CEA; Neutron spectrum: choice of the absorber nuclides; Cooling fluid and structure materials: physicochemical compatibility with the absorber material; Core temperature: absorber material refractoriness.

15.1.2

Neutron absorbers in generation IIeIII and prototypic reactors

In PWR [2], most of the CEA use “fingers” (or pins or rods) fastened to a central cast spider assembly inserted from the top of the core in the fuel assemblies (Fig. 15.1). The number of fingers per CEA (about 20) and the number of CEA per core (about one for four fuel assemblies) depends on the core dimensions, fuel composition [due to neutron spectrum hardening, more control rods are required for mixed oxide fuel (MOX) fuels], and power. The neutron absorber materials are most often the Ag-In-Cd (AIC)

System

Electric power (MWe)

Core averaged power density (MWth/m3)

Application (plus electricity)

VHTR

250

6e10

Hydrogen, heat

SCWR

1500

70

GFR

200e1200

100

LFR

50e150 (transportable) 300e1200 (station)

SFR MSR

Neutron spectrum

Fuel

Cooling fluid primary circuit

Core temperature (8C)

Thermal

ZrC-coated Triso particles, graphite blocks or pebbles

He

Outlet >900

Thermal to fast

UO2

Supercritical water

510e625

Hydrogen, actinides burning

Fast

U-Pu carbide/SiC

He

Outlet w850

150

Hydrogen

Fast

Mixed oxide (nitrides)

Pb or Pb/Bi

550e850

300e1500

300

Actinides burning

Fast

Mixed oxide

Liquid Na

450e550

1000

5

Hydrogen, actinides burning

Epithermal

U(Th)-Li-Be fluorides

Liquid fuel, fluorides

700e800

Absorber materials for Generation IV reactors

Some characteristics of the Generation IV projects to be taken into consideration for the choice of the neutron absorber materials Table 15.2

535

536

Structural Materials for Generation IV Nuclear Reactors

Figure 15.1 Left: PWR control rod assembly inserted in a fuel assembly mock-up. Center: BWR cruciform control rod. Right: SFR control rod (not to scale).

alloy together with B4C boron carbide (as compacted powder or low-density pellets) inserted in Inconel or stainless steel tubes. When both materials are used, AIC is at the tip of the pins, B4C is at the upper part of the pins and, thanks to a very high neutron absorption cross-section, is mainly used for reactor shutdown. However, in the case of hardened neutron spectrum (e.g., MOX fuels), boron carbide has also been used for regulation. In a Russian PWR (VVER), different materials have been tested up to a semi-industrial extent: hafnium, dysprosium titanate, with claimed good results. In a boiling water reactor (BWR), the CEA are crosses inserted from the bottom of the core between fuel assemblies. Different absorber materials are used, either B4C boron carbide (compacted powder about 70% density) or hafnium, in parallel tubes or in plates in the wings of the crosses. In CANDU reactors, different systems are used to control and stop the reactor [3]. Soluble poisons (boron anhydride, gadolinium nitrate) can be diluted in the heavywater moderator circuit. Stainless-steel rods can be inserted in order to reduce the moderation efficiency. Finally, control and shutdown rods are made of cadmium tubes sandwiched between stainless steel tubes. All the control rod systems are vertically introduced in the core, when the fuel elements are horizontal. Advanced gas-cooled reactors (AGRs) have only been built in Great Britain [4]. These are thermal reactors moderated by graphite and cooled by CO2. The core structure is built by graphite bricks packing maintained in a prestressed concrete vessel in

Absorber materials for Generation IV reactors

537

which channels allow fuel assemblies and control rods insertion. A back-up safety system consists of nitrogen injection in interstitial channels, since nitrogen is a much better neutron absorber than CO2. A better efficiency can be reached by injecting boron-glass beads in some of the nitrogen interstitial channels. The control rod systems are made with borated steel rods (4.4 wt% natural boron). Depending on the control rod system (control, safety), the lower part of the rods is made either with pure steel or borated steel. In fast neutron reactors, mostly sodium-cooled (SFR), the neutron spectrum leads to a limited neutron absorber materials choice [5]. The CEA generally incorporate highdensity B4C boron carbide, most often 10B enriched to improve its efficiency, as cylindrical pellets piled in stainless steel tubes. The use of large components (size identical to fuel assemblies) leads to an improved efficiency by a self-moderating effect. Despite totally different fuel management, MSR includes rods for safe shutdown. Although the main control of the reactor consists of adjusting the concentration of fissile uranium in the fuel salt, auxiliary control is achieved by inserting corrosionresistant sheaths containing either boron carbide [6] or AIC [7]. The high-temperature gas-cooled reactors (HTGRs) can be derived according two different fuel configuration, either prismatic blocks stacking or circulating spheres (pebble bed modular reactor, PBMR) [8]. The PBMR reactor also has a quite different fuel management system as compared to other solid fuel reactors. In this case, graphite spheres about 6 cm in diameter including Triso fuel particles are made circulating in the core chamber. Control systems are generally located in the outer reflector of the core [9,10]. Due to the high temperatures encountered and the thermal neutron flux, the absorber material is B4C boron carbide. The absorber elements are distributed in two groups, control and shutdown. The control rod design consists of annular B4C rings encased between two tubes, to form sections mechanically linked to form an articulated control rod several meters long. The control rod is suspended from the drive mechanism by a chain or a cable. A shock absorber below the control rod protects it and the core structure in the event of a chain (cable) failure. A control rod guide tube allows guiding the control rod into the core. On the other hand, tubes are located in the center of the core in which boron carbide spheres are introduced for shutdown.

15.1.3 Present reflections and developments status From the above short review, it appears that most of the present nuclear reactors use a narrow sampling of neutron absorber materials. This, of course, first results from the neutron properties of the elements. This is also a consequence of the materials and elaboration processes availability. For example, AIC has been developed as a surrogate to hafnium and boron is mainly used as boron carbide. This is always a compromise: regarding the previous examples, AIC has the lowest melting point of all the core materials and boron carbide is a brittle ceramic enduring premature cracking. In the case of the thermal neutron reactors (PWR, BWR, HTR, etc.), AIC and to a lesser extent B4C and Hf are practically the only used materials. In fast neutron reactors, nearly only B4C is used. Regarding boron carbide, this ceramic can be used either as compacted powders or sintered pellets or diluted in metal matrices. On the other

538

Structural Materials for Generation IV Nuclear Reactors

hand, only the 10B isotope is a neutron absorber. As a consequence, modifying the relative density, the 10B enrichment and the control elements design allow scanning a wide range of neutron absorption efficiency. Those parameters are then widely used to design the control and protection components, such as lateral neutron protections, control and shutdown assemblies or even spent fuels storage casks [11]. AIC was designed as a surrogate to hafnium (mainly on a neutronic point of view [12,13]), due to low resources of hafnium when electricity-producing civilian reactors were developed in the 1960s and 1970s. But it is worth noting that this material has by far the lowest melting temperature among all the core materials, even taking into account possible eutectics. Such a parameter will have to be addressed in the frame of the Generation IV projects. Due to the large zirconium needs for the water-cooled reactors, significant hafnium quantities are nowadays available as natural zirconium ores contain about 2 wt% of hafnium, which has to be fully removed to prevent any neutron absorption in the zirconium alloys fuel cladding This then allows this material to be reconsidered. Europium oxide was sometime used, e.g., in BN-600 [14] or in icebreakers [15] or submarine [16] reactors, and shows high stability of its efficiency, absence of gas release and swelling. The main drawback, i.e., a high residual activity, substantially complicates spent rod handling and post-use management, and often leads to designers passing over to boron carbide. This problem is also encountered with AIC, this making waste management an important issue to be considered when selecting the absorber materials. As emphasized in Chapter 1, the most important evolutions from the generation IIeIII to the Generation IV reactors deal with safety issues. As a consequence, all components must have the highest reliability. This has to be accounted for both in the design and in the choice of the materials. Regarding the neutron absorber materials, they have to be chosen in order to keep their absorption efficiency during their expected lifespan (e.g., control rods, are expected to last at least as long as the fuel elements versus neutron lateral protections, ideally not to be changed during the whole core operating life) but also keep their integrity and not interact with the control rod elements or the core medium, with the risk of preventing operation of the control systems in normal or incidental conditions. Such limitations are encountered, for example, with the Ag-In-Cd alloy in PWR (creep and swelling inducing cladding cracking [2]) or boron carbide in SFR (carburation of steel cladding and fragments relocation [14]). For these two cases, this results in a lifetime much shorter than expected from the absorber efficiency evolution. This topic has then to be addressed when considering the potential materials and CEA design to be used in future reactors. The final choice of a neutron absorber material is then not only determined by its efficiency (initial and evolution) regarding the neutron spectrum of the reactor. Examining the consequences of the damage of the material under irradiation (swelling, creep, cracking, gas release, etc.) and of the subsequent interaction with the cladding (mechanical or chemical) then the potential dispersion in the primary circuit (activation, abrasive particles, etc.) is of primary importance. This analysis has to be extended beyond the use in the reactor: transportation (cladding embrittlement, activity), dismantling, potential recycling (e.g., 10B), and waste storage (activity, periods, radiotoxicity, etc.).

Absorber materials for Generation IV reactors

539

Among the six reactor systems considered in the GIF forum (VHTR, GFR, SFR, LFR, MSR, SCWR), all use control rods for regulation and shutdown of the reactors. Depending on the maturity and the background of the projects, the present description of those regulation systems spans from mere neutron considerations to detailed designs. On the other hand, the basic parameters to the choice of the neutron absorber materials (neutron flux and spectra, temperature, etc.) for the Generation IV systems span in ranges already encountered in the generation IIeIII and prototypic reactors. It then appears that the materials to be used for neutron absorption in those systems have most often to be selected or extrapolated from those already known, even if different configurations are considered (for example, ZrH moderator together with B4C absorber assemblies in SCWR [17]).

15.1.4 Materials resources and needs Large resources of boron are available, mainly located in Turkey, the USA, and Russia. Boron producers (as borax, Na2B4O7, 10H2O) are mainly in Turkey (ETI Maden), the USA (Rio Tinto), South America, and Russia. The main applications of borax are to produce borosilicate glasses then ceramics, fertilizers, or detergents. For boron carbide elaboration, boron is first refined as oxide (B2O3) or anhydride (H3BO3). Boron carbide is then mainly produced by reduction with graphite in arc furnaces at high temperature (carbothermal reduction). The main applications are for abrasives and shielding or armor plates. The nuclear needs (both control rods and soluble boron) appear low in the boron market (less than 1%). 11B, with no neutron absorption property but which is used in microelectronics, represents a larger market than 10B. Today, most of boron carbide (natural or 10B enriched) is produced in the USA (mainly from self-extraction and refining) and China (mainly from boron acid precursors imported from Turkey, Chile, and Russia). Large quantities (batches up to 5 tons) of pure, small-grain size powders that meet the requirements for nuclear applications can be delivered by Chinese producers. To be used in control rods, boron carbide is most often 10B-enriched up to 90% instead of about 19.8% in the natural element. In this case, the cost of the boron carbide reduction is a few percent of the total cost of the material (high-density cylindrical pellets). The total cost is then nearly equally distributed between enrichment and shaping (hot-pressing, machining). In thermal neutron reactors, the boron carbide quantity is less than 500 kg, and generally low density (compacted powders or sintered pellets), natural 10B content. In large fast neutron reactors, there is about 1 ton of boron carbide in the control rods and it is generally as hot-pressed, high-density pellets and enriched from about 45e90%. It can also be used as lateral or upper neutron protection; in that case, larger quantities of natural boron carbide are required. As a consequence, a large use of (enriched) boron carbide in Generation IV reactors would require an extension of the present enrichment and processing facilities but is not a problem regarding the boron resource. Apart from nuclear applications in control rods, hafnium is produced for specific uses (microelectronic, super alloys). It is a byproduct of zirconium (about 2% in natural ores). Estimates are that the two main hafnium producers, ATI Wah Chang (USA) and CEZUS (AREVA group, France), produce around 40 and 30 tons of the

540

Structural Materials for Generation IV Nuclear Reactors

metal annually [18]. India, Russia, and China produce or have the potential to produce a few tons per year in order to fulfill their own needs. It presently rates about $800e900 per kilo, with a sharp increase in 2006 from about $200/kg. Replacing AIC with Hf in commercial PWR or introducing it in Generation IV reactors instead of AIC would lead to needing about 500 kg per core loading. It then appears that the nuclear needs could rather easily unbalance the market. This point has indeed to be evaluated as soon as rare or precious elements are considered: it is worth noting that those materials, when they are used in nuclear reactors, unlike traditional applications, can no longer be recycled for any other use. The Ag-In-Cd ternary alloy (usually 80 wt% Ag, 15 wt% In, 5 wt% Cd) is based on the availability of the three metals. Cadmium and indium are byproducts of zinc production [19]. Cadmium is mostly used in batteries (Ni-Cd) but this use is drastically reducing due to regulations induced by its toxicity. Cadmium price widely varies but remains relatively low, between a few dollars to less than $1/kg. Indium is used in electronics (LCD screens, LEDs, etc.), or fusible alloys or solders. Depending on large demand (mainly electronics) and availability variations, the price of indium has shown large variations, from about $20 to $1000/kg, from 1970 to 2010; it is presently about $600/kg. The price of silver is about $600/kg, mainly coupled to the price of gold. In traditional industry, an increasing proportion of those metals is recycled [19], this leading to limit removals of the resource: this can no longer be the case when they have been used as neutron absorbers. Rare earths have been used either as neutron absorbers (Eu, Dy) or as burnable poisons (Gd, Er) either as oxides or titanates (see Section 15.3.4.1). Their main drawback (excepted for Dy) is a very high specific activity after neutron irradiation, leading to complicated management after use. Moreover, rare earths are presently in high demand for magnets, catalysis, technical ceramics, or electronic applications, etc. [20] with the demand increasing [21]. The main reserves are located in the USA, Australia, China, and Russia. Due to lower processing costs, rare earths are now mainly extracted and refined in China [22] (Fig. 15.2). Extraction and separation 150,000

Tons/year

120,000 90,000

China United States India ex-USSR/CIS/Russia Australia others

60,000 30,000

Figure 15.2 Production of rare earth oxides [22].

12 20

20 10

20 08

20 06

20 04

20 02

20 00

19 98

19 96

19

94

0

Absorber materials for Generation IV reactors

541

processes have led to ecological concerns [23,24]. Their rates (depending on the metal, in the range $200e6000/kg) have been recently drastically changing for commercial policy reasons. Depending on the applications and the elements, significant recovery rates are observed. The neutron absorber elements are among the “heavy” rare earth elements, for which the supply is already a concern for classical industrial applications. A large use for nuclear applications would then contribute to unbalance this market.

15.2

Scaling the neutron absorbers

15.2.1 Nuclear properties 15.2.1.1 Microscopic absorption cross-sections of the elements Some properties of the most used neutron absorber elements are reported in Table 15.3. The first parameter to be taken into account for the choice of a neutron absorber element is the adequacy between the local energy spectrum of the neutrons to be absorbed and the neutron absorption efficiency (the microscopic absorption cross-section) versus neutron energy of the nuclide. As an example (Fig. 15.3), cadmium has a very high absorption cross-section in the low-energy part of the neutron spectrum with a sharp decrease at about 0.5 eV (this cut-off is used to define the limit between the thermal and epithermal domains), whilst boron (10B isotope) shows a smooth decrease on the whole neutron energy range [25]. The former will possibly be used (and is actually widely used) in thermal neutron reactors (PWR, BWR); the latter can be used in any kind of reactor, both as a neutron absorber or a burnable poison (e.g., borated glasses in PWR). On the other hand, the efficiency of a neutron absorber material should not be too high, this leading to quick consumption then too short a life span (this is indeed the expected behavior of the burnable poisons) and tricky tuning of the CEA. This can be achieved either by proper choice of the absorber element or also by dilution (e.g., the 304B7 borated steel is about 15 times less efficient than boron carbide). The microscopic neutron absorption cross-sections of the most common elements are reported in Fig. 15.4. The curves are obtained from weighted sums of the different isotopes of a given element as obtained from Ref. [25]. For AIC, we have considered the classical ternary alloy 80 wt% Ag, 15 wt% In, 5 wt% Cd. They all show a monotonous decrease, due to the 1/OE first-order evolution of the neutronenuclei interaction time. Most of the elements display strong resonances in the epithermal energy range: this results from the (n,g) absorption reactions on heavy nuclei, this leading to isotopes of the same elements, possibly unstable (see section 15.2.1.3). This also leads to complex efficiency calculations (Doppler effect). The 10B isotope mainly absorbs a neutron via a (n,a) reaction producing 4He and 7Li; there is also a minority (n,2a) reaction producing 3H in the fast neutron energy range.

542

Table 15.3

Nuclear properties of the main neutron absorber elements Cross-section (barn)

Element

Atomic number

Bnat (20% B) 10

10

B

Ag

5 47

Atomic mass

Thermal

10.811

766

10.013

3830

107.87

68

Res. int.

Absorption products Main nuclides

Period

Decay mode

3

12.35 years

b

Ag* Ag*

438 years 250 days

K, b, IT b, IT

H

923

108 110

Cd

48

112.41

2349

109

Cd

461 days

K

In*

49 days

K, b, IT

Dy Ho*

144 days 1200 years

K b

177

Lu*

160 days

b

175

Hf

70 days

K

178

Hf**

31 years

IT

181

Hf

42 days

b

182

Ta

114 days

b

In

49

114.82

193

3085

114

Dy

66

162.50

1023

1961

159 166

Hf

72

178.49

111

1762

Atomic mass: for natural elements; Res. int.: Resonance integral for epithermal neutrons (for boron, integral cross section in epithermal domain); Periods: days, years; Decay modes: IT (isomeric transition), K (electronic capture). *:isomeric state

Structural Materials for Generation IV Nuclear Reactors

63

Absorber materials for Generation IV reactors

543

1.E+05

σ (barn/atom)

1.E+04 1.E+03

10B

1.E+02

Cd

1.E+01 1.E+00 1.E–01 1.E–02 1.E–03 1.E–03

1.E–01

1.E+01

1.E+03

1.E+05

1.E+07

E (eV)

Figure 15.3 Neutron absorption cross-section for natural Cd and

10

B [25].

15.2.1.2 Macroscopic cross-sections of materials The microscopic cross-sections are then used to calculate the efficiency of the actual materials (alloys, oxides, cermets, carbides, etc.) to be introduced in the CEA, taking into account the concentration of all the absorber nuclides. This leads to the macroscopic absorption cross-sections S(E) (in cm1) we have reported in Fig. 15.5 for

1.E+05

σ (barn/atom)

1.E+03

1.E+01

AIC Hf 1.E–01

Er Dy 10B

1.E–03 1.E–03

1.E–01

1.E+01 1.E+03 E (eV)

1.E+05

1.E+07

Figure 15.4 Microscopic neutron absorption cross-section for the most common absorber elements. AIC is the 80 wt% Ag, 15 wt% In, 5 wt% Cd alloy [25].

544

Structural Materials for Generation IV Nuclear Reactors

1.E+04

AIC

Hf

Dy2TiO5

B4C e0,2

Σ (/cm)

1.E+02

1.E+00

1.E–02

1.E–04 1.E–03

1.E–01

1.E+01

1.E+03

1.E+05

1.E+07

E (ev)

Figure 15.5 Macroscopic neutron absorption cross-section for some common compounds. B4C boron carbide is calculated for natural boron (10B/B ¼ 0.198) and fully dense material.

some common materials (not taking into account self-screening or moderating effects, see e.g., hafnium hydride, Section 15.2.6.2): X

ðEÞ ¼ 1024

X

si ðEÞNi

i

where si(E) (barn) is the microscopic neutron absorption cross-section for the ‘i’ absorber nuclide of the material and Ni is the concentration (at%/cm3). Some peculiar effects here have to be taken into account. When highly efficient neutron absorbers are used in control rods, the neutron absorption mainly occurs at the surface of the absorber elements and the neutron spectrum hardens with depth (self-screening effect). In that case, it can be interesting to use annular elements: this is actually made in PBMR. This also leads to a strong radial gradient consumption of the material; this can have consequences on its mechanical properties or integrity. As an example, desquamation is observed at the surface of the boron carbide cylindrical pellets used in PWR. On the other hand, introducing a moderator (low-Z element from H to C) into the absorber will increase its efficiency by locally softening the neutron spectrum. In the case of hafnium hydride HfH1.3, when used in SFR, the control rod efficiency is three times as large as that of hafnium without hydrogen [26] and then becomes equivalent to 10B-enriched boron carbide. Due to the low-Z elements constituting the material, this effect is also effective in boron carbide in fast neutron reactors.

15.2.1.3 Neutron absorption products As the seat of nuclear reactions, the fate of absorbing elements has to be considered. In Figs. 15.6 and 15.7, we have reported the simplified transmutation chains in AIC and

K 107

51.35%

K

Cd

1b 1.22%

γ +n

107

γ

Ag

74 b

106

γ

Pd

10 b

+n γ

Cd

108m

+n

108

Pd

Ag

γ 109

Ag γ

200 b

418 d 2.3 mn β– 108

γ

Cd

48.65%

+n

110

Ag

Bold-line frames: main absorptions

24 s 249 d

β– 110

γ γ

Cd

0.2 b 6.5 h

0.89%

Broken-line frames: unstable isotopes Absorption cross section for thermal neutrons (barn) Period (2 values: isomers)

Solid-line frames: stable isotopes Absorption cross section for thermal neutrons (barn) Initial natural concentration (%)

14 h

β-

γ

Ag

109

12.43%

111

Cd

2100 b

+n

12.86%

γ +n

112

Cd

30 b 23.79%

γ +n

113m

C

113

20800 b 14 y 12.34%

+n

γ

β

–

113

In

58 b

e.g.: Cd : 113mCd transmutes in 113In, 113Cd is stable and is present in natural cadmium 114In transmutes 98% in 114Sn and 2% in 114Cd

γ

Cd

4.16%

114

+n

γ

114

115

In

β

In

95.84%

114

Sn

+n

7.66%

γ

207 b

γ

98% β–

γ

Cd

1.5 b

44 d –

115

49 d

116

Cd

53 h

+n

28.81%

2% K γ

γ

Cd

1.2 b

117

Cd

3.4 h 2.5 h

γ

β– γ +n

116

117

In

Absorber materials for Generation IV reactors

108

AIC

In

54 mn 14 s

γ

β– 116

Sn

6 mb

γ

117

Sn

+n

Figure 15.6 Simplified transmutation chains for AIC. Absorption cross-sections for thermal neutrons in barns. Periods in years, days, hours, minutes, and seconds. The main absorber nuclides are 107Ag, 109Ag, 111Cd, 113Cd, and 115In. Sn is produced from In. 108Ag and 109mAg are strong g-emitters.

545

546

175

Hf

174

Hf

γ

1500 b +n 0.2%

+n

γ

K 175

γ

Lu

35 b

Hf

70 d

γ

+n

176

Lu

4000 b

176

Hf

γ +n

γ

15 b +n 5.2 %

177

Lu

6.7 d

β-

177

Hf

γ γ

380 b +n 18.4%

178

Hf

γ

75 b +n 27.1 %

179

Hf

γ

65 b +n 13.8%

180

Hf

γ

23 b +n 35.4%

181

Hf

42 d

β–

181

Ta

Bold-line frames: main absorptions

γ +n

γ

182

Ta

17000 b

114 d

β–

182

W

19 b

+n

γ γ +n

183

Ta

5d

β– 183

W

11 b

γ γ +n

184

W

4b

γ +n

185

W

75 d

β– 185

Re

Broken-line frames: unstable isotopes Absorption cross section for thermal neutrons (barn) Period

220 b

γ γ +n

186

Re

3.7 d

β–

186

γ

Os

Figure 15.7 Simplified transmutation chains for hafnium. Same notations as in Fig. 15.6. The main absorber nuclide is 177Hf. The main g-emitter is 182Ta.

Structural Materials for Generation IV Nuclear Reactors

Solid-line frames: stable isotopes Absorption cross section for thermal neutrons (barn) Initial natural concentration (%)

70 b

γ

Absorber materials for Generation IV reactors

547

Hf. A large number of isotopes of the initial and of new elements are created, possibly stable or radioactive with periods ranging from seconds to centuries, leading to significant changes of both the isotopic compositions of the elements and of the composition of the material. This, of course, leads to modifications of the neutron absorption efficiency of the material but also of its chemical or thermomechanical properties (e.g., precipitation of a hexagonal, tin-enriched phase in AIC). For some elements, the nuclides resulting from the neutron absorption [from (n,g) reactions] are also efficient absorbers, such as hafnium (Fig. 15.7) or dysprosium. In that case, the life span of the control rods may be significantly increased. The radiotoxicity of the materials and their descendants both in the course of operating the reactor and for waste management has continued to take more and more importance since the 1990s. Thus the development of a new AIC-based CEA could be hampered by the presence of radioactive isotopes with relatively long periods (108Ag: 438 years) in the control rods at the end of life. The presence of descendants of long periods may also be a barrier to the use of rare earths as well as their cost. In this regard, Hf or Dy are much less critical than AIC. Similarly, the main reaction absorption in boron is 10B(n,a)7Li, but the minority (yield about 103 in SFR) 10B(n,2a)3H reaction in the fast neutron energy range complicates the end-of-life management.

15.2.2 Materials properties 15.2.2.1 Boron carbide Boron carbide is a light, brittle, very hard, refractory ceramic. It is most often obtained by carbothermal reduction at high temperature of purified boron oxide [27,28]. The material has then to be ground to powder. In France, a magnesothermal process was used leading directly to micronic powders. The powders are then most often sintered or hot-pressed, depending on the density, in other to obtain the cylindrical pellets constituting the absorber pins. In actual materials, the grain size usually ranges from 5 to 20 mm, the density ranges from around 60% (vibro-compacted powders, e.g., in BWR) to 70% (sintered pellets, e.g., in PWR) up to about 95% (in SFR). Industrial materials all have a composition very close to B4C. Due to the bad behavior of graphite under irradiation (swelling, sodium intercalation in SFR), materials with as low as possible excess carbon (referred to as “free carbon”) are required. The thermal conductivity is mainly phonon-like, with a 1/T variation (for a 95% dense material, around 27 W/m$K at RT and w12 W/m$K at 1000
C). The mechanical behavior is purely brittle (for a 90% dense material at RT, KIC w2.5 MPaOm, Young modulus w250 GPa, yield w400 MPa). Those properties show low modifications up to high temperatures and decrease as the porosity increases. Boron carbide has a complex crystal structure [29,30] constituted of nearly regular icosahedra (mean composition B11C, C located on polar sites) interconnected according to a rhombohedral network, Fig. 15.8. The bonding is mainly covalent, this conferring its thermomechanical properties. The center of the cells is filled with a linear chain, most often C-B-C, allowing additional tight bondings between the icosahedra. It is then isostructural to some borides (e.g., B6Si, B6O), mainly differing by the composition of the

548

Structural Materials for Generation IV Nuclear Reactors

Figure 15.8 Cell structure of B4C boron carbide [30].

central element. Depending on the actual composition of the cell elements, boron carbide exists in a large composition range, from about B4C to approximately B10C.

15.2.2.2 Ag-In-Cd Silver had for long been used as a neutron absorber [31]. Due to bad water corrosion resistance (to be taken into account in the case of cladding cracking) and mechanical properties, the ternary alloy Ag-In-Cd was developed (Fig. 15.9), with the supplementary objective to build a material with absorbing properties close to hafnium. Adding cadmium alone shows no metallurgical or corrosion improvements and insufficient neutron efficiency (fast burning of Cd). Various additions were tested (Pd, Au, Cu, Ni, Al, etc.). Indium eventually shows significant improvements of both the corrosion and metallurgical properties of the alloy. Taking further into account the ternary phase diagram [32] leads to the standard composition 80 wt% Ag, 15 wt% In, 5 wt%. Cd. Increasing the grain size (from ASTM 6e8 to 2e3) leads to improved creep resistance. It is worth noting that tin additives still increase the water corrosion resistance; however, this is paradoxically no longer the case when tin results from indium transmutation under irradiation. Such an effect is also encountered in hafnium where tantalum addition decreases the corrosion resistance but is of no consequence when it results from Hf transmutations [33]. Further improvements of the mechanical properties (yield

Absorber materials for Generation IV reactors

Ag-In-Cd

549

1.0

350°C

0.9 0.8 L

Mo

le_

fra c

tio n

IN

0.7 0.6

0.5 0.4

0.3

hcp

0.2 0.1

fcc

0 0

0.2

0.4

0.6

γ

0.8

1.0

Mole_fraction CD

Figure 15.9 Ag-In-Cd phase diagram at 350
C, as calculated by C. Desgranges [32].

strength up to 200 MPa at 315
C) were obtained by dispersion of fine oxides of the elements in the matrix ([31], Section 5.7). AIC is a monophasic solid solution with a face-centered cubic cell structure (cell parameter 4.16 Å). The initial density is 10.17 g/cm3. The melting temperature (slightly depending on the atmosphere) is 800  15
C (this making AIC the most fusible material in a reactor core, with possible safety issues [34]), probably too low to be considered in Generation IV systems (Table 15.2). The thermal conductivity is metal-like, about 55 W/m$K at 20
C and 90 W/m$K at 600
C. The yield strength increases from 70 to 100 MPa and the ultimate strength decreases from 300 to 100 MPa in the 300e600
C temperature range.

15.2.2.3 Hafnium Hafnium has been one of the first absorbers to be used (e.g., in the Nautilus submarine) and has been widely used in water reactors, leading to great feedback [33]. As compared to AIC, it has very good resistance to corrosion and friction when it is used in water reactors without cladding tubes, due to the formation of a protective and hard oxide layer [31]. It can then be directly used as bars or tubes (to be potentially filled with other absorber materials [35]). This is a very dense metal (13.3 g/cm3, compared to 10.2 g/cm3 for the AIC and 6.5 g/cm3 for Zr): such density may require changes of the mechanisms moving the absorbing rods to reflect their extra weight. The mechanical properties strongly depend on metallurgical properties [35] and residual contents of impurities. Resistance to cracking requires low oxygen (<300 ppm) and is enhanced by addition elements (Zr between 3% and 5%, Nb). The addition of Fe or Si allows to maintain a low grain size (<20 mm) during the forming operations. Additions of Sn and O increase the tensile and creep strength; Fe, Cr, and Nb

550

Structural Materials for Generation IV Nuclear Reactors

improve the corrosion resistance; Mo improves hardness, wear resistance, and machinability [36]. Its oxidation resistance (excluding neutron flux), due to the passivating effect of the hafnia layer, is much greater than that of zirconium, which motivated its use in nonwrapped rods. Hafnium is a refractory metal (melting temperature 2233
C). Regarding the compatibility with potential cladding materials, it forms a solid solution with Zr and shows a eutectic at 1300
C with iron. The crystal structure is hexagonal (hcp) below 1760
C and cubic (bcc) above. The yield strength is about 250 MPa and ultimate strength 350 MPa at 300
C.

15.3

Behavior under irradiation of neutron absorber materials

15.3.1

Boron carbide

Most of the postirradiation examinations of irradiated boron carbide have been done on control rods in SFR: very few results are available for boron carbide irradiated in water reactors. The main neutron absorption reaction is the 10B(n,a)7Li: 10

B þ 1n / 7Li þ 4He þ 2.6 MeV

Boron carbide has very high structural stability under irradiation [37]. However, when used as a neutron absorber, very high absorption rates are obtained (the volume density of neutron captures by 10B is most often called burnup): in SFR, up to 1022 He/cm3 is produced, leading to drastic modifications of the microstructure and of the composition and high energy release (about 100 W/cm3).

15.3.1.1 Thermal water reactors In PWR, the neutron absorption cross-section of 10B is very high (Fig. 15.5). As a result, the neutron captures mainly occur in the outskirts of absorbing elements (several hundred microns). The low temperature (<400
C) leads to high retention of the helium generated within the material, leading to a locally significant swelling (of the order of 0.4% by volume for 1020 captures/cm3). It should be noted that, contrary to SFR, very few studies have been performed aiming at analyzing the helium behavior (release, clustering) in the material. The fragile nature of the material then causes rapid erosion of the surface of the absorbent pellets and desquamation from the periphery (Fig. 15.10(a)). This damage has two consequences. First the produced particles progressively fill the gap between the pellets and the cladding. The subsequent swelling may cause deformations of the cladding incompatible with the conduct of the reactor (induced absorber swelling cladding cracking, IASCC [2]). Moreover, the cracked material becomes very sensitive to radiolysis (dissolution in water induced by free radicals created by radiation). A water inlet permitted by IASCC then leads to rapid dissolution of the absorbent, again unacceptable. Both effects severely limit the

Absorber materials for Generation IV reactors

(a)

551

(b)

2 mm

5 mm

Figure 15.10 Postirradiation observation of absorber pins [38]. (a) Low-density (70%), natural 10 B B4C pellet in thermal neutron reactor (French Osiris reactor). (b) High-density (95%), 10Benriched (48%) B4C pellet in SFR (French Phenix reactor). In both cases, the maximum burnup is about 5.1021/cm3. The primary damage in Osiris comes from strong radial capture gradients, whilst in Phenix it comes from high-temperature gradients.

use of B4C as the main control material and should confine it as a component of the shutdown clusters. In this case, only maximum absorption efficiency is sought, and the absorbent remains mostly in areas of the core with very low neutron flux. The life of the CEA is thus limited by that of the primary absorbent, AIC, or other. This point should be reconsidered when MOX fuels are used: in that case, the neutron spectrum is hardened; this can lead to inserting boron carbide even for regulation. In BWR, the CEA are comprised of crosses inserted between the fuel elements. Boron carbide is present in the form of vibro-compacted powder; the relative density can reach 70%. Despite its complete lack of cohesion, this material also shows a macroscopic swelling due to microscopic swelling and the hooping of the grains, which prohibits their rearrangement. Swelling (and to a lesser extent helium release) can also lead to IASCC, resulting, as in PWR, in a risk for quick solution of the absorbent in the primary coolant. Helium accumulation in the grains also tends to their fracturing, this exacerbating helium release and slightly delaying swelling.

15.3.1.2 Fast neutron reactors In fast neutron reactors (SFR), the absorption cross-section in the B4C is low. Sufficient efficiency is obtained with high-density materials (>90%) and highly enriched 10 B, used as cylindrical pellets about 2 cm diameter. The radial profile of neutron captures in the absorber material is rather flat (ratio of less than 2 between the periphery and the heart of the control rods), even considering the self-moderating effect of the light elements B and C constituting the material. It follows that the thermal power in the neutron capture is uniformly distributed in the absorber. In normal use, this power can, in the most stressed areas, be higher than 100 W/cm3, comparable to that generated by the fuel elements. The low thermal conductivity of B4C then leads to a veryhigh-temperature gradient, up to 1000
C/cm. The stresses induced by the differential dilation between the periphery and the heart of the pellets exceeds the strength of the

552

Structural Materials for Generation IV Nuclear Reactors

30 k init.

25

12 29

k (W/m K)

20

49 15

67 82

10

5

0 0

200

400

600

800

T (°C)

Figure 15.11 Thermal conductivity of high-density B4C irradiated in Phenix versus burnup (in 1020/cm3). kinit, nonirradiated B4C [39].

material, inducing a radial fracture (Fig. 15.10(b)). This fracturing therefore occurs at the very first use of the CEA and it remains active during the whole life of the absorber element, due to the sharp deterioration of the thermal conductivity of B4C under irradiation (Fig. 15.11, [39]). A second source of fracturing is local and comes from the accumulation of helium produced within the material. In medium-temperature ranges (between 500 and 1200
C), the helium release rate is low (Fig. 15.12: it is worth noting that most of the results on irradiated boron carbide have been obtained in control rods for which the effects of the actual irradiation parametersdtemperature and fluxdcould not really be deconvoluted, this leading to a poor analytical description of its behavior). It accumulates in the form of flat, parallel, lenticular bubbles, both within the grains and at the grain boundaries (Fig. 15.13). This subjects the material to very high internal stresses (shear stresses at grain boundaries), which exceed its strength for burnup at about 1021/cm3. The average swelling associated with the retention of helium is about 0.15 vol% for 1020/cm3, lower than in water-cooled reactors. Cracking is initially intergranular (burnup from 1021 to 5  1021/cm3), then mixed inter- and intragranular. At high temperatures (about 1500
C), different mechanisms are activated (defect diffusion, plastic transition) and helium bubbles become three-dimensional, leading to accelerated swelling. The combination of fracturing, swelling, and fragment relocation rapidly induces IASCC. This severely limits the life of the absorber elements, long before 10B exhaustion, for a noncracking criterion of the sheath is required. The evacuation of the thermal power produced by the absorbent is most often achieved thanks to a slow circulation of the coolant (liquid sodium) inside the absorber rods. For this, the steel tubes are provided with porous vents at both ends, the first

Absorber materials for Generation IV reactors

553

300 T irrad. (±100°C) He release (cm3 TPN/cm3 B4C)

250 500 700

200

900 1100

150

100

50

0 0

50

100

150

200

Burnup (1020/cm3)

Figure 15.12 Helium release rate in irradiated boron carbide. Full line, total release; 500e1100, estimated mean temperatures in reactor [41].

0.25 μm

Figure 15.13 Intragranular helium bubbles in irradiated boron carbide. Scale: 0.25 mm [42].

554

Structural Materials for Generation IV Nuclear Reactors

function of which is to allow evacuation of released helium. These vents have a mesh which prohibits the training of B4C particles in the primary circuit. A drawback is that the liquid sodium promotes very effective diffusion of carbon into the B4C cladding. This induces an extensive carburization of the cladding, inducing embrittlement then shortening its lifespan. Boron carbide has shown good compatibility with liquid sodium. It is worth noting that very few studies have been devoted to lithium behavior. Some authors mention grain boundary embrittlement. A thermal diffusion coefficient has been determined [40]. Some measurements have shown retention rates much higher than deduced from this diffusion coefficient: it is then assumed that sodium inhibits lithium release. As a result, the life of the absorbent elements is not primarily limited by 10B exhaustion (burnup up to 2.5  1022/cm3 has been achieved in the Phenix reactor [14]), but especially by the degradation of the cladding. Several solutions have been developed to limit the effects of these impairments. For more advanced absorbent elements [43], a “liner” (or shroud), thin metal tube, is placed around the stack of pellets, preventing the dispersion of fragments, maintaining the sodium flow, and slowing carburization of the sheath. Moreover, reducing the diameter of the control rod elements results in a decrease of thermal gradients and the resulting fracturing. In the frame of the GIF forum, analytical studies are performed aiming at a better description of the behavior of boron carbide [44,37]. At last, due to the low activity of irradiated boron carbide, 10 B recycling can be performed, either by direct crushing and resintering the pellets, or by oxidation and carbothermal reduction, then again crushing and sintering [14].

15.3.2

Ag-In-Cd alloy

Chains of transmutation of the elements Ag, In, and Cd (Fig. 15.6, Table 15.4) show the formation of tin, mainly from indium. Beyond a concentration of about 2% [45], a second phase precipitates. This precipitation is triggered by complex thermodynamic mechanisms, leading to a phase distribution different from that obtained with the classical lever rule [46]. This phase is hexagonal, with an atomic density lower than that of the original face-centered-cubic (fcc) AIC structure; this causes a significant swelling of the alloy (Fig. 15.14). This second phase has a melting temperature still below that of AIC [32]. The swelling is heterogeneous Table 15.4 Mean composition of AIC at the feet of a pin used eight cycles in a PWR (at%) Element (atom %)

Ag

In

Cd

Sn

Nonirradiated

80.8

14.3

4.9

d

Irradiated

68.1

7.0

17.8

7.1

From J. Bourgoin, D. Gosset, F. Couvreur, F. Defoort, M. Monchanin, X. Thibault, The behaviour of control rod absorber under irradiation, J. Nucl. Mat. 275 (1999), 296e304.

Absorber materials for Generation IV reactors

555 3 A3, hcp

ΔVat/Vat(%)

2.5

Rim

A3, fcc

2 hcp single phase

1.5 1

fcc + hcp

0.5 0 0

300 μm

0.5

1 1.5 Depth (mm)

2

2.5

Figure 15.14 Left: microstructure of irradiated AIC (four cycles PWR): the dark areas correspond to the tin-rich hexagonal phase, in greater proportion in the rim of the rod. Right: local swelling (atomic density variation) of the hcp ( ) and fcc ( ) phases as compared to the initial AIC (eight cycles PWR) [45].

due to the radial profile of neutron absorption (self-screening effect), larger in the periphery of the rod. The poor mechanical properties of AIC induce rapid deformation in operation [2]. This results, on the one hand, in a conventional creep and, on the other hand, a compaction due to large accelerations (on the order of tens of g) sustained during the vertical displacements of the control clusters. These effects are amplified by the temperature of the reactor material which, under the effect of heating due to the absorption of radiation, reaches 350e450
C and exceeds the recrystallization temperature (about 275
C). The deformations quickly lead to a closing of the initial clearance at the foot of the rods between the rod and the sheath, resulting in high mechanical stress that can lead to cracking of the sheath (IASCC). This results in drastically shortening the lifetime of the AIC. Moreover, the direct contact with water of the primary circuit induces degradation due essentially to the internal oxidation of the indium and the hydrolysis of the AIC: one part is dragged away in the coolant, leading to a strong contamination of the primary circuit in 108mAg and 110Ag. As a result of these difficulties, modifications of the absorber pins have been performed [2,12,31]: • •

Microstructure: a larger and homogeneous grain size and the addition of finely dispersed oxides improve the creep resistance; Geometry of the pins: an increase of the rodesheath clearance delays the time of the mechanical interaction.

The isotopes produced by transmutation (mainly silver isotopes) have a radioactivity that complicates their management after irradiation. After a stay in plant storage pools, their radiological characteristics make them suitable for the waste category B. However, given the high solubility of indium and indium oxide In2O3 in cements (aqueous alkaline medium), it should be kept in mind that the chemical stability of AIC is not guaranteed on geological periods.

556

Structural Materials for Generation IV Nuclear Reactors

15.3.3

Hafnium

The neutron absorption of hafnium leads to the formation of hafnium isotopes, themselves absorbers (Fig. 15.7) and then to isotopes of heavier elements: Ta, and W (up to a few percent; Lu is produced in very small amounts). Among these, 182Ta is the main source of the radiotoxicity of the material, but with a 115-day period, making it a shortlived waste. The degradation of mechanical strength and the swelling crucially depend on the formation of hydrides [2]: they are due to the diffusion of hydrogen formed by radiolysis of water, which diffuses into the material; the solubility of hydrogen in Hf is very low. These hydrides induce unacceptable embrittlement (lamellar precipitates) and swelling for the resistance and functioning of the absorber rods. They have been the cause of a partial abandonment of hafnium as an absorber material. This hydride formation can however be easily inhibited by the formation of a hafnia (HfO2) oxide film about 10 microns thick and which totally passivate the material surface. This layer is compact, does not crack, and has a good wear resistance. The use of unsheathed hafnium rods thus maintains this oxide layer. Under irradiation, hafnium weakens until fast fluences (fast neutrons, E > 0.1 MeV) of the order of 5  1021 n/cm2. Its properties (hardness, tensile stress, ultimate tensile strength) then remain nearly constant. Damage is due to the formation of a high concentration of dislocation loops. Although neutron captures lead to the formation of isotopes of heavier elements, the density decreases slightly under irradiation by formation of vacancy clusters. The solubility limit of Ta in Hf is low, inducing the formation of nanometer-sized precipitates for the most irradiated materials [33]: this precipitation may be delayed by adding niobium to the original metal. Hafnium has a hexagonal structure, this inducing anisotropic deformation under irradiation (“growth”) similar to that observed on the Zircaloy sheaths (Fig. 15.15). This is due to a combination of two effects: • •

The fabrication process of the absorber rods or tubes leads to an initially anisotropic material (crystallographic basal planes, normal to the axis, preferably perpendicular to the axis of the rods or claddings); Interstitial and vacancy defects accumulate preferentially in the form of dislocation loops located respectively on basal and prismatic planes.

This results in elongation (less than percent, lower than in the case of zirconium) of the claddings or rods, which may be limited by controlling the metallurgical state of the initial metal.

15.3.4

Other materials

15.3.4.1 Dysprosium titanate In the Dy2O3-TiO2 system there are two defined compounds, Dy2TiO5 and Dy2Ti2O7 (Fig. 15.16 [47]). They are obtained by conventional methods of blending and sintering Dy2O3 and TiO2 precursors. The materials used in Russia are usually multiphasic, but the predominantly present compound is Dy2TiO5. Depending on the temperature,

Absorber materials for Generation IV reactors

557

0.4 0.35 0.3 ΔL/L DL/L

ε (%)

0.25

DJ/J ΔΦ/Φ

0.2 0.15 0.1 0.05 0 0

1

2

Flux (1026 m–2, E > 0.1 MeV)

Figure 15.15 Growth (anisotropic swelling) of Hf rods [33].

L

2400 Dy2TiO5

Dy2Ti2O7 (P)

T (°C)

2000

C+L C + F (fluorine)

F+P

P+L

R+L

1600 C

β+P

C + β (hex.)

P+R (rutile) 1200

Dy2O3

C + α (orthor.) 20

α+P

40

60

Mol (%)

Figure 15.16 Phase diagram of Dy2O3-TiO2 system [47].

80

TiO2

558

Structural Materials for Generation IV Nuclear Reactors

the latter has three different structures, low-temperature orthorhombic, hexagonal at intermediate temperature, and cubic (analogous to the fluorite structure) at high temperature. Unlike zirconia, small volume changes are associated with these phase transformations. Under irradiation, the orthorhombic and hexagonal phases undergo phase changes to the fluorite one; the material then shows a behavior similar to that of zirconia [48]. It is then desirable to obtain directly the fluorite structure. As for zirconia, this phase can be easily stabilized down to room temperature, by adding a few percent of molybdenum [49] (this material slightly activates under irradiation). The theoretical densities of the titanate (cubic) and the bititanate are respectively 7.34 and 6.86 g/cm3. Few thermomechanical data are available; the properties are basically those of an oxide ceramic (low thermal conductivity, brittle). The material has then to be used with a cladding. The neutron absorption cross-section of dysprosium is relatively high in the range of thermal neutrons (Fig. 15.5), which gives the titanate a comparable efficiency to that of conventional materials, AIC, B4C (natural boron), Hf. The succession of absorbing isotopes in the transmutation chain induces the efficiency decreases very slowly under irradiation (Fig. 15.17), resulting in a design life greater than that of boron carbide and equivalent to that of AIC or Hf. The unstable isotopes produced by the transmutation reactions are in low quantities (166mHo) and mostly with short periods (165Dy), causing low concern stress for the end of the cycle, in contrast to AIC. Under irradiation, the material shows little damage. The swelling of the fluorite phase is low [less than 0.5% by volume for a neutron fluence of 1022/cm2 (E > 0.1 MeV)]. Orthorhombic and hexagonal phases show a greater swelling, mainly 100 90

Efficiency (%)

80 70 60

B4C nat. Hf

50

AIC Dy2TiO5

40 30 0

10,000

20,000

30,000

40,000

Irrad. time (h)

Figure 15.17 Dependence of the absorber material efficiency (7-mm diameter pellets) in the VVER-1000 [47].

Absorber materials for Generation IV reactors

559

due to phase changes, hence the importance of using the stabilized cubic phase. The material is used either as vibro-compacted powders or as sintered pellets. For powders, the low swelling results in the absence of deformation of the sheaths containing them. In the pellets, cracking appears in the irradiated materials at high fluence due to constraints caused by differential expansion induced by the high thermal gradients (>60
C/mm). These gradients result from the power dissipated in the neutron absorption (g heating) and the low thermal conductivity of the material. In temperature ranges encountered in PWR (320e350
C), there is no significant chemical interaction after 15 years between the titanate and sheaths (06Cr18Ni10Ti steel). In 2005 the material had already benefited from an experience return of more than 20 years in the research reactor MIR and nearly 10 years in VVER-1000 without incident. A low fluence experiment in a pool reactor in France (Osiris) also showed good behavior of this material. In the same material family, dilutions of Dy2O3 in a ZrO2 matrix are considered as burnable poisons in the Indian AHWR project [50]. Dysprosium hafnate has also been evaluated [51]. As compared to dysprosium titanates, this material incorporates two efficient absorbers. As a result, its efficiency is improved by about 8% in the thermal and epithermal ranges. A single-phase domain (from about 15e55% Dy2O3, fluorite structure) exists in the HfO2-Dy2O3 system. Preliminary irradiation testing has been performed in BOR60 in small capsules up to a 1022/cm2 fluence. The material shows good behavior, with capsule and pellet integrity and low swelling.

15.3.4.2 Hafnium compounds Hafnium hydride Hafnium can no longer be used as a neutron absorber in fast neutron reactors, due to insufficient absorption efficiency. However, due to the strong moderation effect of hydrogen (concentration in metal hydrides is comparable to water), hafnium hydride has an initial absorption efficiency comparable to enriched boron carbide, with much slower decrease during operation (Fig. 15.18, [52]). Metal hydrides have been considered for many applications in nuclear reactors (moderators, shielding, reflector, and control; [53]). Few studies have been carried out on the Hf-H system. However, it shows properties very similar to Zr-H, which is extensively used as a moderator. As in other hydrides, the equilibrium pressure of hydrogen depends on the temperature and the stoichiometry of the hydride. The main issue is then hydrogen release and diffusion through the metal cladding (stainless steel). This can be limited by additives or inhibited by surface coatings, either on the hydride or on the cladding. Moreover, the thermal conductivity of HfHx is good, leading to low heating in reactor then low hydrogen equilibrium pressure. Irradiation tests of HfH1.5 samples have been performed in the BOR-60 reactor [26,54]. The samples were irradiated for 4 years up to a fluence of 2.6  1022 (n/cm2, E > 0.1 MeV) at temperatures from 500 to 600
C. As hydrides show good stability with liquid sodium, Na-filled capsules were also tested. The latter show a very low hydrogen release rate. Postirradiation experiments show integrity of the capsules, no structural modifications, low hydrogen release, and low swelling.

560

Structural Materials for Generation IV Nuclear Reactors

2

Control rod worth (%dk/k)

1.75

1.5

HfH1.0 80% 10B

1.25

40% 10B B4C nat

1

0.75

0.5 0

2

4

6

8

10

Operation time (years)

Figure 15.18 Decreases in efficiency of HfH1:0 as compared to B4C (natural and 10B enriched to 40% and 80%) during operation [52].

Hafnium dioxide (hafnia) Different elaboration processes have been developed in order to produce hafnia artefacts in the perspective of nuclear applications [55e57]. Pure hafnia has a monoclinic structure at room temperature and undergoes a phase transition to a fluorite structure at high temperature, leading to cell distortion and possible microcracking. The fluorite structure can however be stabilized down to room temperature by additives, such as Y or Mn. High-density pellets can be obtained by sintering a powder at about 1600
C. This material shows many advantages, such as very high melting temperature, good inertness either with cooling fluids (water) or cladding materials (zircaloy, steel). As with other materials with the fluorite structure, it also shows good stability under irradiation. The main drawback is a relatively low hafnium concentration, about 2.9  1022/cm3 in HfO2 to be compared to 4.5  1022/cm3 in hafnium: this would confine this material to neutron protections rather than control rods. However, its density is consequently significantly lower (10 g/cm3 vs. 13.3 g/cm3), this leading to lower concern on the mechanisms of the CEA (cf. Section 15.2.2.3).

15.3.4.3 Transition metal diborides As compared to boron carbide, transition metal or rare earth borides offer many advantages. They are most often metallic, this conferring good thermomechanical properties. They have very high melting temperatures (e.g., 3250
C for HfB2), good thermal conductivity (80 W/m$K for HfB2). They are brittle; but additions of

Absorber materials for Generation IV reactors

Table 15.5

561

Helium release and swelling in metal borides Burnup

cm3/cm3

Swelling (vol%)

T max (8C)

He 10008C (%)

0.9

15

250

31.3

Helium release

Sample

Density (%)

%10B

1020/cm3

%

HfB2

83

87

10.1

0.3

þ

TiB2

97

67*

9.8

1.1

d

40

250

33.6

VB2

90

50

8.1

7.9

0.4

19

200

35.0

ZrB2

85

83*

9.1

3.6

d

25

200

d

36

250

33.3

þ

B4C

100

37*

8.1

11

YB4

95

62

9.5

4.9

3.5

23

250

51.0

DyB4

86

44*

5.5

39

d

Frag.

300

d

DyB6

87

43*

6.3

82

d

Frag.

300

d

EuB6

87

37

5.7

19

4.3

7

250

51.3

SmB6

86

44

6.4

18

d

15

250

82.0

YB6

92

42

6.7

14

d

12

200

86.0

*, calculated; þ, extrapolated from pellet diameter; Frag., fragmented; He 1000
C, helium release (%) after annealing at 1000
C. From E.W. Hoyt, D.L. Zimmerman (part I-II), W.V. Cummings, W.I. Clark (part III), Radiation Effects in Borides, USAEC, GEAP-3743, 1962.

SiC improve their mechanical properties [58]. As mentioned above, some transition metals (Hf) and rare earths (Dy, Eu, Gd, Er) are efficient neutron absorbers, leading to quite high absorption efficiency. As such, they have for a long time been considered as potential neutron absorbers [59]. Among different compounds (diborides, tetraborides, hexaborides), the metal diborides present the best behavior under irradiation [60]. It is noted that the more the structure is closed (that is, from hexaborides to diborides), the better is the helium retention (Table 15.5). However, most of the materials cracked this requiring efficient containment with strong mechanical resistance. The behavior of hafnium diboride has also been studied when irradiated with helium and lithium ions beams [61]. Transmission electron microscope observations show the formation of dislocations loops, leading to anisotropic swelling as in the metal (Section 15.2.5). Helium clusters were observed only after annealing at high temperature, this suggests helium is trapped mainly as isolated interstitials or small clusters.

15.3.4.4 Composites materials Up to now, we have discussed only monophasic materials. However, composites have been used as neutron absorbers for a long time and different concepts have been tested in order to improve some limitations of the classical materials.

562

Structural Materials for Generation IV Nuclear Reactors

400 μm

Figure 15.19 Crack propagation in a HfB2-B4C composite (20 vol% HfB2, white granules) [65].

Borated steels have been used as control rods from the dawn of nuclear reactors [31]. They presently are widely used for spent fuel storage [62]. They consist of a stainless steel matrix incorporating boron-rich particles (e.g., the 304B range [63]), mainly as chromium borides. The mean boron concentration is low, less than 10% of boron carbide. They have good mechanical and corrosion resistance properties. Produced helium is trapped in the particles and at the particleematrix interfaces, with very low release rates: this is obtained because the boride cluster concentration is lower than the percolation limit. However, few results are available regarding their behavior in high-irradiation conditions, such as in an SFR core with high temperature and fast neutron flux. One of the main drawbacks of boron carbide flows from its low thermomechanical properties, inducing premature and progressive cracking. This issue can be addressed according two routes. The first is derived from the borated steel design, then taking benefit of the properties of the metals as compared to those of the borides. In this frame, cermets, for example, B4C particles in a hafnium matrix, have been elaborated [38]: such materials would have properties similar to those of the borated steels (thermomechanical stability, helium retention) but with a much higher absorption efficiency. Another route has also been considered: in that case, a cerecer composite is elaborated (here, a dispersion of HfB2 spherules in a B4C matrix, both components are highly efficient absorbers: Fig. 15.19), the microstructure of which is designed to prevent crack propagation (R-curve behavior) [64e66].

15.4

Conclusion: for a better definition of the needs

As emphasized in the Introduction, the Generation IV projects are deep evolutions of present or past reactors with the highest requirements regarding, for example, the safety and sustainability of the systems. Those generation IIeIII and prototypic reactors all have control elements, meaning the different designs and materials have met the efficiency and safety requirements for efficient and safe working of the plants.

Absorber materials for Generation IV reactors

563

However, the control elements range appears quite narrow since only three (and mainly two) materials are used in power reactors. The reasons that led to these choices are mentioned above. However, they have significant limitations. In the case of the AIC alloy, the low thermomechanical properties lead to shortened life span, the low melting temperature could certainly be a safety concern and the daughter isotopes of silver lead to waste management problems. As for boron carbide, the conjunction of cracking, swelling, helium release, and possible cladding carburation also leads to lifespans shortened much before the 10B exhaustion. As a result, these materials could certainly be used in the Generation IV projects but in-depth discussions are required beforehand both on the material choice and the control element designs. We have shown that challenger materials can be considered. The most promising is certainly hafnium, due to a large feedback. The main drawback is its susceptibility to hydriding in water reactors, for which effective solutions are known. Russian experience has provided evidence for the potential of the dysprosium titanate: this material has the same absorption efficiency as AIC or Hf, shows good behavior under irradiation, and creates few waste management problems. Now, those materials are to be used in thermal or epithermal neutron flux. In fast neutron flux, apart from rare earths, boron is the only element with sufficient efficiency. It is currently used only as boron carbide, with strong defects (cracking, swelling), but other compounds should be considered, such as diborides or cermets. As for rare earths (except Dy), they do not appear as a realistic option, first because of high costs and tight market, second because of the high activity of the daughter elements. More exotic materials should also be evaluated: for example, hafnium hydride appears to have an absorption efficiency as high as enriched boron carbide in fast neutron flux. The second axis on which action must be taken is the design of absorber elements. This can be made first at the material level. For example, increasing the grain size and adding fine oxides has shown significant improvement of the resistance of AIC. Likewise, inserting the boron carbide pellets in a shroud leads to efficient protection of the sheath, inducing an increase in the life duration of the control elements. Furthermore, the absorption efficiency can be improved by local modifications of the neutron energy spectrum. This is directly obtained with hafnium hydride, but also with heterogeneous control elements, for example with external moderators (e.g., zirconium hydride [67]) around the absorber material: in this configuration, hafnium could be used in a fast neutron reactor. This design can be extrapolated at the core scale: in the ASTRID project, the neutron lateral protections are located behind moderator rings [68]. Economy must also be a primary concern. For example, the life duration of the control elements has to be considered, together with the initial cost of the materials and CEA. Those parameters will have different weights depending on the actual function and expected life of the components. For example, different weights should be applied for the side neutron protections in an SFR, supposed to stay in the reactor for its whole life, and the upper neutron protections, possibly changed at the same pace as the fuel elements. Also, the supply sources should be examined carefully. For example, large boron resources are present on earth, but very few places provide enriched boron. Recycling the irradiated but still highly 10B-enriched boron carbide could then be

564

Structural Materials for Generation IV Nuclear Reactors

considered. Due to tensions on the market for rare earths, these (except Dy) can be considered only for very specific applications. As a conclusion, it appears neutron absorber materials do exist that are able to meet the needs of the Generation IV projects. However, much has still to be made to define for each reactor type the best set of materials and concepts. This certainly requires a fine cooperation between designers and materials specialists and refinements of the core simulations down to a very local scale. However, this also requires material irradiation tests in dedicated reactors in order to obtain possibly missing behavior data.

Abbreviations AGR B4C BWR CANDU CEA GFR GIF HTGR IASCC LFR MOX MSR PWR SCWR SFR VHTR VVER

Advanced gas-cooled reactor Boron carbide, composition of industrial materials Boiling water reactor Canadian uranium-deuterium reactor Control element assembly Gas-cooled fast reactor Generation IV International Forum High-temperature gas-cooled reactor Induced absorber swelling cladding cracking Lead-cooled fast reactor Mixed oxide fuel Molten salt-cooled reactor Pressurized water-cooled reactor Supercritical water-cooled reactor Sodium-cooled fast reactor Very-high-temperature reactor Russian PWR

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