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Conceptual core design study for a high-flux LFR demonstrator
Progress in Nuclear Energy 54 (2012) 56e63
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Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Conceptual core design study for a high-flux LFR demonstrator Sara Bortota, *, Patrizio Console Camprinib, Giacomo Grassoc, Carlo Artiolic a
Politecnico di Milano, Department of Energy, CeSNEF-Nuclear Engineering Division, via Ponzio 34/3, 20133 Milano, Italy University of Bologna, Nuclear Engineering Laboratory (LIN) of Montecuccolino, via dei Colli, 16, 40136 Bologna, Italy c National Agency for the New Technologies, Energy and Sustainable Economic Development (ENEA), via Martiri di Monte Sole 4, 40136 Bologna, Italy b
a r t i c l e i n f o
a b s t r a c t
Article history: Received 6 January 2011 Received in revised form 31 August 2011 Accepted 1 September 2011
A preconceptual core design study for a pool-type Lead-cooled Fast Reactor (LFR) demonstrator (DEMO) has been developed in the frame of an I-NERI between the Italian National Agency for the New Technologies, Energy and Sustainable Economic Development (ENEA) and Argonne National Laboratory (ANL), based on the European Lead-cooled System (ELSY) reference concept. A demonstration reactor is expected to prove the viability of technology to be implemented in the first-of-a-kind industrial power plant. DEMO specifications as a nuclear power facility demonstrating ELSY main features and performance, besides validating design methodology and tools, have been defined. Suitable design parameters have been set to meet the foremost objective of reaching a high fast neutron flux while respecting all technological constraints. Preliminary thermal-hydraulic analyses have been carried out to verify safety limits were not exceeded. Ó 2011 Elsevier Ltd. All rights reserved.
Keywords: Core design Demonstration Generation IV Lead-cooled fast reactor LFR Neutronics
1. Introduction The Lead-cooled Fast Reactor (LFR) is under development worldwide as a very promising fast neutron system to be operated in a closed fuel cycle (Generation IV International Forum, 2002). In particular, within the 6th EURATOM Framework Program, the European LFR community proposed the ELSY (European Leadcooled SYstem) concept (Cinotti et al., 2007), an innovative 600 MWe pool-type LFR fully complying with Generation IV goals of economics, safety, proliferation resistance and sustainability, the latter being led by the feature of no net production of Minor Actinides (MAs). As recognized by the Strategic Research Agenda worked out by the European Sustainable Nuclear Energy Technology Platform (SNETP), LFR full development up to industrial deployment requires e as a fundamental intermediate step e the realization of a demonstrator (DEMO) of some hundreds MWth, aimed at validating LFR technology as well as the overall system behavior. In order to define a first reference configuration for a DEMO, an I-NERI (International Nuclear Energy Research Initiative) between the Italian National Agency for the New Technologies, Energy and * Corresponding author. Tel.: þ39 02 2399 6327; fax: þ39 02 2399 6309. E-mail addresses: [email protected] (S. Bortot), [email protected] (P. Console Camprini), [email protected] (G. Grasso), [email protected] (C. Artioli). 0149-1970/$ e see front matter Ó 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2011.09.001
Sustainable Economic Development (ENEA) and Argonne National Laboratory (ANL) has been signed in 2007, under the umbrella of the EURATOM-DOE Agreement. On the Italian side the work has been carried out in the frame of the national R&D program on “New Nuclear Fission” supported by the Italian Minister of Economic Development (MED) through a general Agreement with ENEA and the Italian University Consortium CIRTEN. In such a context, a reference configuration of a pool-type LFR DEMO has been developed under a cooperation between ENEA, Politecnico di Milano and University of Bologna. A demonstration reactor is expected to prove the viability of technology to be implemented in the first-of-a-kind industrial power plant. Therefore, the first step toward DEMO conceptual design has been the definition of its specifications as a nuclear power facility able to both validate ELSY main features and performances, and to qualify numerical codes and tools (Bortot et al., 2009). Based on the results of preliminary scoping analyses (Bortot et al., 2010), an optimization study has been performed in order to accomplish a reference core characterization. Suitable design parameters have been set so as to meet the foremost objective of reaching a high fast neutron flux while respecting all technological constraints, among which peak cladding and fuel centerline temperatures, and peak neutron-fluence. A fuel cycle hypothesis consistent with the maximum burn-up and cladding dpa allowed in ELSY has been adopted, and
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
depletion calculations have been performed in order to determine the reference Beginning of Cycle (BoC) and End of Cycle (EoC) core characteristics (i.e., effective multiplication factor keff, power profiles and neutron fluxes), as well as reactivity coefficients and kinetic parameters. Two different and independent Control Rod (CR) systems have been foreseen to guarantee the required reliability for reactor shutdown and safety, the first one addressing reactivity swing compensation over the cycle, regulation and safe shut-down, and the second one being devoted to scram only. Preliminary thermal-hydraulic analyses have finally been performed with reference to the hot channel parameters, in order to verify a posteriori that safety limits were not exceeded. 2. Core design approach 2.1. Aimed performances Two essential missions have been envisioned for DEMO as a GEN-IV LFR demonstrator: comprehensive demonstration of successful lead-cooled reactor operation and testing. Due to this broad spectrum approach, R&D priorities and planned efforts focus on neutronics, thermal-hydraulics, materials, control systems and strategies, systems and components, safety and performance requirements. The main objectives regarding neutronics are mainly related to the need of demonstrating the key issues related to the adiabaticity of a core (i.e. characterized by a unitary breeding ratio and burning its self produced MAs), which is a very crucial point for an actual sustainability with respect to both natural resources exploitation and environmental impact minimization (Sarotto et al., 2009). To this end, it is necessary to test the system behavior when at least some MA-bearing fuel assemblies (FAs) are hosted in DEMO core; this would allow to validate numerical codes outcomes in terms of both nuclear data and reactor physics methodology. Furthermore, overall core performances e such as peak burn-up e should be evaluated. Consequently, the driving criterion in designing DEMO core has been the achievement of a high neutron (fast) flux, so as to enable efficient irradiation of fuels and materials to high burn-ups over reasonable irradiation times. DEMO plant is also requested to attest the worth and operation reliability of the ELSY innovative Finger Absorber Rod (FAR) system. In particular, one has to verify that adopted calculation tools provide accurate estimations of FAR reactivity worth, as well as that the control system is suitable as for compensation, local power regulation and safe shut-down. Material extensive testing is not among the main goals of the demonstration project; nevertheless, DEMO core materials (fuel, cladding and structures, coolant) are to be the same of ELSY ones in order to guarantee technological consistency with the industrialsize LFR nuclear power plant. 2.2. Reference criteria Conventional MOX (reactor grade, RG, Pu; 1.97 stoichiometric ratio; 95% theoretical density) has been considered as the reference fuel (isotopic vectors showed in Table 1). Annular pellets have been required to both moderate centerline temperatures e enabling to afford 365 W cm1 peak linear heat rating, which is compatible with the limit of 2400 C maximum fuel temperature e and lower the average density so as to reach high burn-ups (beyond 100 GWd t1 on heavy metal, HM) without affecting the fuel pin integrity (Bortot et al., 2010). Due to the choice of T91 ferritic-martensitic (FMS) steel (elemental composition given in Table 2) as the candidate material
57
Table 1 Fuel isotopic vectors. Isotope
Depleted U [wt.%]
RG Pu [wt.%]
234
0.003 0.404 0.010 99.583 e e e e e
e e e e 2.333 56.873 26.997 6.104 7.693
U 235 U 236 U 238 U 238 Pu 239 Pu 240 Pu 241 Pu 242 Pu
for both fuel rods and structures due to its high irradiation resistance and promising ongoing R&D, 600 C maximum operational temperature has been assumed, since it is the limit expected to be affordable in the near future whether superficial treatments (namely aluminization by GESA, currently under development and testing at Karlsruhe Institute of Technology and ENEA-Brasimone) show to be reliable over time against corrosion (Weisenburger et al., 2007; Lead-bismuth Eutectic Alloy and Lead Properties, 2007). Furthermore, a limit on peak neutron fluences (w4$1023 cm2 for ferritic steels) has been considered. A coolant inlet temperature of 400 C is required in order to avoid excessive embrittlement of structural materials subjected to fast neutron flux and to provide a sufficient margin from the lead melting point (327 C). The same coolant outlet temperature as in ELSY (480 C) has been kept, since such a choice enables to reach higher values of neutron flux thanks to higher fuel power densities allowed (Bortot et al., 2010); in addition, the whole system would be more representative of ELSY, being the operational characteristic temperatures e and, in general, conditions e reproduced. Furthermore, the coolant velocity has been limited to 3 m s1 (Weisenburger et al., 2007; Muller et al., 2002) to prevent the cladding from erosion risks. As to FA specifications, a ductless design with pins arranged on a square lattice has been focused on. Every FA must be provided with four structural uprights at corners and a central FMS T91 control rod driveline allowing the fast insertion of FARs. A safety requirement demanded to be necessarily met pertains to the control system reactivity worth and redundancy: the need of two different and independent absorber sets has been assumed, guaranteeing a 3000 pcm minimum margin for safe shut-down purposes, besides a suitable insertion speed and passive reliability. A primary system would be required to have sufficient reactivity worth to bring the reactor from any operating condition (i.e. overpower together with reactivity fault) to cold sub-critical at the refueling temperature with the most reactive control element stuck at the full power operating position. In addition, the primary system should serve the purpose of compensating for the reactivity effects of fuel burn-up and axial growth, while accommodating both the criticality-uncertaintyrelated reactivity and fissile loading. A secondary system would be intended to shut the reactor down from any operating condition to the hot standby status, in turn with the most reactive assembly inoperative. Also for the secondary system design the reactivity fault is included as a requirement, since it should
Table 2 FMS T91 isotopic composition. Element
Cr
Mo
Nb
Mn
Si
Ni
V
Fe
[wt.%]
9.0
1.0
0.1
0.6
0.5
0.2
0.2
88.4
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S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
Fig. 1. Core layout (left) and 2-D RZ cylindrical section (right); dimensions in cm. A single FAR is represented in the equivalent RZ scheme to clarify the absorber set out-of-core positioning.
override the uncontrollable withdrawal of one primary control assembly used for burn-up control. Finally, DEMO should have an affordable cost; this imposes a limitation on the system size. Based on preliminary economical evaluations (Cinotti, 2009), a 250e300 MWth reactor power seems to be a reasonable trade-off between DEMO plant cost and performances in terms of ELSY representativeness. Such a figure is compatible with the opposite requirement of shielding the active zone with a sufficient number of dummy elements, and of limiting the axial and radial core size so that the reactor tank does not exceed the dimensions compatible with both costs and sloshing risks minimization.
2.3. Optimization approach The starting point of DEMO neutronics design has been the work carried out in (Bortot et al., 2010): the scoping analyses performed on a 250 MWth MOX-fueled core demonstrated that a 6 mm clad outer diameter,1 associated with 365 W cm1 peak linear power, assures high power densities and, consequently, high values of neutron flux. Starting from those established fuel pin specifications, an effort has been undertaken to conceive a very compact core configuration with the main objective of lowering the required plutonium fraction under its acceptable limit. In this perspective, classical massive control rods have been forsaken: compensation, regulation and scram tasks have been assigned to FARs, since such a solution enables to significantly reduce neutron leakage, besides preserving ELSY representativeness. With the same aim, an optimization work has been carried out concerning core dimensions. An effort has been spent to strike a balance between enlarging the core axial extension e so as to accomplish a more critical
1
An attempt of enhancing the neutron flux has been pursued as far as possible by raising the fuel power density up, i.e. moving toward small values of pin diameter. Since an important constraint is that, beyond some point, small pins become increasingly expensive to fabricate, a lower limit on this parameter has been set at 6 mm.
geometry e and not losing criticality by reducing the fuel volume fraction (VF) due to the coolant flow area enhancement required to compensate any core height increase.2 Indeed, scoping analyses outcomes highlighted that material volume fractions impact on criticality more strongly than the geometrical shape of the core, in this range of scenarios. Consequently, it has been opted for increasing the coolant velocity from 2 m s1 to its maximum acceptable limit of 3 m s1 in order to leave a margin to downsize the pin pitch. Reference peak and average linear heat ratings have been preserved and it has been possible to meet both the needs through a pin pitch reduction (from 8.71 to 8.53 mm) and a concurrent core axial extension increase (from 42 to 65 cm) abiding by lead outlet temperature requirements. According to the core size constraint, the need of a fuel assembly size comparable with ELSY one3 and the above-mentioned shielding requirements, a suitable FA design has been accomplished. Then, aiming at keeping the fresh fuel plutonium fraction below the postulated limit of 35 wt.%, FAs have been arranged in a symmetrical and compact staggered configuration consistent with 300 MWth total power; such a new size has been found to comply with the above-mentioned economics constraint. A further optimization has been pursued following DEMO core geometrical settlement. Since for ELSY a power/FA distribution factor of at most 1.2 has been evaluated suitable to assure temperature limits to be respected with a proper safety margin (Artioli et al., 2008), such a reference has been borrowed for DEMO. As required by the wrapper-less solution choice,4 two radial enrichment regions have been foreseen in order to smooth the coolant outlet temperatures through a careful flattening of power distributions. A first iteration has been made to optimize both the definition of the two enrichment zones and
2 Since a pin pitch enlargement aimed at enhancing the mass flow rate would be required to reinstate the enthalpy balance between core outlet and inlet. 3 To host the same internal mechanisms without miniaturizing cost issues. 4 It does not permit to exploit different FA orificing to locally regulate the coolant flow rate.
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
59
Fig. 2. Fuel assembly outline.
the mutual ratio between the respective fuel enrichment figures pursuant to the reference distribution factor. The average plutonium fraction required for criticality has been determined accordingly. At that point it has been necessary to assume a fuel cycle strategy in order to evaluate DEMO main neutronic performances, as keff swing over the equilibrium cycle, BoC and EoC power
distributions, neutron fluxes, etc. A simplified hypothesis has been adopted: evolution calculations have been performed according to a 1-batch approximation,5 i.e. by postulating a homogeneous core
5 It has been demonstrated such an approach gives accurate predictions of the effective core behavior during the cycle (Krepel et al., 2009).
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S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
burn-up while defining ideally the BoC and EoC core configurations as the one corresponding to the core average situation immediately after refueling and the one corresponding to the core average situation immediately before refueling, respectively (Krepel et al., 2009). The fuel residence time has been set so that the neutron-fluence-related constraint was not exceeded; a firstattempt calculation has been carried out on the basis of the Beginning of Life (BoL) peak neutron flux e since it was the only figure available e , and a more accurate assessment has ensued from considering the EoC corresponding value. The consequent total time interval has been divided into four refueling sub-cycles (open 4-batch cycle, simulated again according to a 1-batch approximation) in order to allow the cycle reactivity swing to be manageable by compensation FARs; then, plutonium fractions have been readjusted so as to reach keff ¼ 1 at EoC while respecting the postulated power/FA distribution (considering that at BoC a local power modulation is granted by the compensation FARs insertion, whereas at EoC absorbers are completely withdrawn from the active core). 2.4. Calculation tools and reactor model Neutronics analyses have been performed by means of ERANOS (European Reactor ANalysis Optimized System) deterministic code ver. 2.1 (Rimpault et al., 2002), coupled with JEFF-3.1 (Joint Evaluated File, 2006) data library. Multi-group cross-sections associated to every reactor zone have been retrieved from ECCO (European Cell COde) (Rimpault, 1997) calculations.
Table 3 DEMO major core specifications, cold dimensions (20 C). Specification Reference goal parameters Thermal power Average coolant outlet temperature Coolant inlet temperature Average coolant speed Peak surface cladding temperature Core data Number of inner/outer fuel assemblies Number of shielding assemblies FA data FA pitch Pin lattice Number of pins/FA Pin pitch CR driveline inner/outer width Fuel pin data Cladding outer diameter Cladding thickness Pellet outer diameter Pellet hole diameter Fuel smear density Fuel column height Gas plenum height Total pin height Fuel Pu fraction inner/outer zone
Value 300 480 400 3.0 600 10/14 74
Units MWth C C m s1 C
e e
238.84 28 28 744 8.53 45.65/48.65
mm e e mm mm
6.00 0.34 5.14 1.71 0.95 650 650 1300 29.3/32.2
mm mm mm mm e mm mm mm wt.%
FAR data Cladding outer diameter Cladding thickness Gap B4C pellet stack height
44.0 0.60 0.10 850
mm mm mm mm
Radial reflector assembly data Assembly size Lead volume fraction Steel volume fraction
238.84 85 15
mm wt.% wt.%
Table 4 Fuel cycle hypothesis outline. Fuel Residence Time Months
¼ core
¼ core
¼ core
¼ core
0 5 10 15 20
0 5 10 15 20/0
0 5 10 15/0 5
0 5 10/0 5 10
0 5/0 5 10 15
A very refined cell description has been adopted for both FAs and sub-critical zones surrounding the active core (with the exception of radial reflector dummy assemblies) and containing FARs, for which an accurate heterogeneous geometry model has been set up; the remaining zones (namely ‘foot FA’, ‘dummy belt’, ‘barrel’ and ‘external lead’, see Fig. 1) have been modeled homogeneously. In both cases, ECCO computations have been carried out by treating the main nuclides with a fine energy structure (1968 groups) and condensing the obtained cross-sections into 33 groups. ECCO evaluations have been used to set detailed tri-dimensional core models up; then the whole system has been solved through nodal transport calculations by means of the ERANOS TGV/ VARIANT module (Ruggieri et al., 1993). 3. DEMO core neutronics characterization 3.1. Core layout In the current design, DEMO features a 300 MWth MOX-fueled core, composed by ductless FAs. Fig. 1 shows (on the left) a section of the core composed by 24 FAs surrounded by dummy assemblies, and (on the right) a bi-dimensional RZ representation of the vertical section that yields a useful scheme for the main structural zones. Ensuing from criticality assessment and power flattening, 10 FAs with 29.3 wt.% Pu fraction constitute the inner zone, and 14 FAs with 32.2 wt.% Pu fraction compose the outer zone. As far as FAs design, fuel pins are arranged in a 28 28 square lattice; every FA is provided with 4 structural uprights at corners connected with the central FMS T91 box beam replacing 6 6 central positions (Fig. 2). In Table 3 the main DEMO core specifications are summarized. 3.2. Fuel cycle hypothesis and burn-up calculations Owing to the approach described above (x 2.3), a fuel cycle of 20month fuel residence time and 5 month refueling interval has been adopted. Table 4 presents the residence time of each quarter of the core: according to the open 4-batch cycle refueling strategy, bold values highlight batches undergoing refueling, the two figures referring to the batch aging just before/immediately after reloading. Since the real fuel average residence times in the reactor are 12.5/7.56 months (i.e. 375/225 days) before/after refueling, these figures have been considered as the equilibrium cycle extremes, representing just the core conditions at EoC and BoC, in accordance with the 1-batch approximation (Krepel et al., 2009). As mentioned in x 2.3, the two radial enrichments have been tuned to obtain keff ¼ 1 and a power/FA distribution of 1.2 at EoC. In Fig. 3 the evolution of keff during irradiation is presented.
6 At EoC and BoC the mean aging of the fuel results indeed 1/4(5 þ 10 þ 15 þ 20) and 1/4(0 þ 5þ10 þ 15) months, that is: 12.5 and 7.5 months respectively.
k ef f
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
1.12 1.1 1.08 1.06 1.04 1.02 1 0.98 0.96 0.94 0.92
61
BOL
BOC 4000 pcm
B7
B8
EOL
100
enriched B4C FARs
42 at.
enriched B4C FARs
42 at.
enriched B4C FARs, extracted
EOC
A4 A5
0
90 at.
200
300
400
500
600
A1 700
B6 A2
B3
Fig. 4. ¼ core layout with control/regulation and safety FARs positioning in the core.
Time [days]
As shown in the graph, the calculated reactivity swing during the cycle is slightly greater than 4000 pcm. The overcriticality at BoC (about 4000 pcm) is expected to be compensated by FARs, whose positioning has been, in turn, optimized in order to duly flatten the radial power distribution at BoC as well (x 3.3). 3.3. Control rod system design Two different and independent CR systems have been envisaged to guarantee the required reliability for reactor shut-down and safety, for which at least 3000 pcm have been requested to be supplied by each absorber set. Four passive7 B4C (90 at.% enrichment in 10B) FARs have been foreseen exclusively for SCRAM purposes. Since in normal operation they are to be positioned outside the active core, FAs characterized by the lowest power (i.e., power distribution factor lower than 1.2) e and therefore not requiring any flattening by regulation FARs e have been considered suitable for shut-down FARs. Then, their location has been determined so as to maximize absorber efficacy while respecting the core symmetry, as indicated in Fig. 4. Furthermore, a motorized FAR system demanded for cycle reactivity swing control and safe shut-down has been foreseen. Every remaining FA has been provided with a B4C (42 at.% enriched in 10B) FAR so that the required worth (w7000 pcm, i.e. 3000 for scram and 4000 for reactivity swing compensation) was assured. A subset of 42%-enriched FARs, corresponding to the four most peripheral positions (coldest FAs), has been conceived to fulfill only scram functions and thereby they lie 1 cm above the active zone in operating conditions, whereas FARs assigned also to local regulation are partially inserted in the core over the entire equilibrium cycle. For a proper assessment of both control systems efficacy, a correction factor reckoning with the strong approximations involved in deterministic analyses has been necessarily applied to the calculated FARs reactivity worth. In fact, FARs limited dimensions introduce a large uncertainty in reactor spatial calculations, despite the associated fine cell analysis description. As a result, macroscopic cross-sections representing FAs with FARs inserted over-estimate the absorbers effectiveness, basically due to a raw treatment of the self-shielding phenomenon brought by homogenization and B4C dilution over the whole cell (Artioli et al., 2009). An attempt of quantifying the influence of cell homogenization has been pursued by comparing the kN parameter resulting from
Table 5 Compensation/regulation and safety system worth. Control System and insertion share
Worth at BoC [pcm] Aimed
Actual
Aimed
Actual
Safety FARs e complete insertion Regulation FARs e complete insertion Regulation FARs subset - 32.5 cm insertion
3000 7000 4000
4624 11,312 4083
3000 7000 e
4856 11,923 e
Worth at EoC [pcm]
homogeneous and heterogeneous cell calculations on inner/outer FAs with/without FARs. For both inner and outer FAs without FARs, discrepancies between homogeneous and heterogeneous models have turned out to be slight (some hundreds of pcms), whereas for FAs with FARs inserted they have turned out to be particularly relevant (some thousands of pcms). Hence, the respective absorber worth has been adjusted accordingly, the SCRAM FAR one being reduced by 25% and the regulation/compensation FAR one by 20%. Active and passive control systems reactivity worth has been assessed at BoC and EoC: as it is shown in Table 5, both FARs sets meet the required safety criteria. In addition to global evaluations, the negative reactivity due to the regulation FAR subset progressive penetration into the active zone has been figured so as to determine the insertion share required for criticality at BoC (Fig. 5). Different compensation/regulation FAR subset insertion strategies have been further investigated in order to derive benefit from power/FA distribution, according to the two concurring needs of complying with the stringent design requirements ensuing from the limit on maximum cladding temperature while maximizing the neutron flux. Neutronics calculations showed that no FARs differentiated insertion is needed to accomplish the aimed 1.2 power/FA distribution, since their uniform introduction suffices for the required power flattening; hence, the corresponding configuration,
Reactivity worth [pcm]
Fig. 3. keff(t) behavior from Beginning of Life (BOL, fresh core condition -singularity-) to End of Life (EOL); ERANOS/JEFF-3.1 calculations.
10000 9000 8000 7000 6000 5000 4000 3000 2000 1000 0
k eff 1 0
5
10
15
20
25
30
35
40
45
50
55
60
FARs insertion [cm] 7 Scram FARs are driven by gravity. A sudden insertion is allowed by their positioning 1 cm above the active zone.
Fig. 5. Regulation/compensation FAR set reactivity insertion curve.
65
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S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
Table 6 DEMO major neutronics results.
Table 7 Reactivity coefficients and kinetic parameters.
Parameter
BoC
EoC
Unit
Parameter
BoC
EoC
Units
Multiplication factor, keff Reactivity, r Total peaking factor Peak power/FA distribution factor Average power density Peak total neutron flux, 4 Neutron flux fast fraction (E > 0.82 MeV) Average/peak discharge burn-up
1.00033 33 1.43 1.20 327 7.27 16
1.00093 93 1.40 1.20 328 7.54 15
e pcm e e W cm3 1015 cm2 s1 %
Doppler constant Doppler coefficient (from 1173.15 to 1500 K) Lead density coefficient (active core) Axial expansion coefficient (linked) Radial expansion coefficient Delayed neutron fraction, b Prompt neutron lifetime, L
235 0.18
268 0.20
pcm pcm K1
MWd kg1(HM)
90/130
0.14 0.19 0.92 323 8.4980
pcm K1 pcm K1 pcm K1 pcm 107 s
superposition principle validity and on the assumption that the system responds linearly within the interval defined by the reference and the perturbed configurations. Consequently, the respective reactivity coefficients have been calculated according to (Sarotto et al., 2009):
with all the 16 regulation FARs inserted for half active height, has been assumed as a reference for BoC.
8 > > > > <
2
6 dr vr vr vr 6 !þ zlT91 ðTinlet Þ 26 > 4 d d dT radial R P > absorber dPfuel > v > v :v R Pabsorber Pfuel 9 3> > > > 7= vr 1 vr 7 þ 1 5 þ ð1Þ dPsteel dPcoolant > VFcoolant > > v v > ; Psteel Pcoolant
3.4. Neutronics characterization at BoC and EoC Following the design rationales and coherent core specifications outlined in the previous paragraphs, a final BoC and EoC characterization has been accomplished. DEMO foremost neutronics results are resumed in Table 6. Both BoC and EoC configurations fulfill the aimed requirements (x 2.1): the driving purpose of designing a high power density core so as to reach high fast neutron fluxes has been achieved (in particular, about three times as much as ELSY one) while complying with all the reference criteria (x 2.2). In Fig. 6 BoC and EoC core radial power density profiles are shown.
2
0.12 þ 0.01 0.87 319 8.0659
and
13 6 Z Tclad; out C7 B 6 vr dr 1 vr vr vr dLinsertion C7 B ! z lclad ðTÞ dT 6 þ C7 B 6 dZ 7 þ vL A @ d dT axial Tclad; out Tclad; in Tclad; in dT P insertion dPfuel steel 4v 5 v v Z P steel P 0
(2)
fuel
3.5. Reactivity coefficients and kinetic parameters Reactivity coefficients and kinetics parameters have been calculated with ERANOS by employing both direct methods and the perturbation theory formalism for the steady-state conditions. Elementary perturbations have been introduced in order to figure the radial (1) and axial (2) expansions through a “partial derivative” approach. Such a method is based on the main hypothesis of
where the last term in Eq. (2) represents the reactivity contribution due to the differential expansion of FARs with respect to the active core (r being reactivity, P material density, l linear expansion coefficient, L length, T temperature, R radial coordinate, Z axial coordinate, and VFcoolant coolant volume fraction). The axial expansion coefficient has been evaluated for the linked case, i.e. when dilation is assumed to be driven by the cladding. As far as the calculation of the Doppler coefficient is concerned, a temperature enhancement by 326.85 K has been considered with respect to the initial average MOX temperature of 1173.15 K.
500 BoC
Power density [W cm-3 ]
450
Table 8 Preliminary T/H evaluations (EoL, hot channel).
EoC
400 350 300 250 200 150 100 0
10
20
30
40 X [cm]
50
60
70
Fig. 6. Radial power density distribution at BoC and EoC (core midplane).
Parameter
Value
Unit
Lead inlet temperature Lead outlet temperature Lead velocity Sub-channel mass flow rate Hydraulic diameter Peak linear power Fuel internal surface peak temperature Fuel external surface peak temperature Cladding internal surface peak temperature Cladding external surface peak temperature
400 501 3.0 1.4 9.4 362 2182 716 574 550
C C m s1 kg s1 mm W cm1 C C C C
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
2400
Temperature [ C]
2000 Fuel centre temperature Fuel surface temperature Clad surface temperature Coolant temperature
1600 1200 800 400 0 0
10
20 30 40 Active core elevation [cm]
50
60
Fig. 7. Hot channel temperature axial profiles evaluated at EoL.
For the coolant density coefficient, a 5% reduction of lead density has been applied. In Table 7 the relative results are summarized, together with the main kinetic parameters.
4. Preliminary T/H evaluation A preliminary thermal-hydraulic analysis has been performed in order to verify a posteriori that safety limits were not exceeded (in steady-state, nominal power). Calculations have been carried out with reference to the End of Life (EoL) core configuration; furthermore, hot channel parameters8 have been taken into account in order to consider the most critical conditions. A clad superficial coating and degradation of the fuel thermal conductivity with burn-up (Thetford and Sobolev, 2005) have been considered (assuming a 20-month-long fuel pin life, i.e. some 130 MWd kg1 peak burn-up). Conditions and results are provided in Table 8. The postulated safety limits (2400 C for fuel (Konno and Hirosawa, 2002) and 600 C for clad) are respected with fairly good margins,9 being the maximum fuel centerline temperature of the order of 2180 C, and the peak cladding outer surface temperature around 550 C. Temperature axial profiles in the hot pin at EoL nominal conditions are shown in Fig. 7. 5. Conclusions In this paper the static neutronics characterization of a GEN-IV LFR DEMO has been presented, along with preliminary thermohydraulic core analysis. Suitable design parameters have been set so as to meet the foremost objective of reaching a high fast neutron flux while respecting all technological constraints. A 300 MWth MOX-fueled core, composed by ductless less FAs with pins arranged on a square lattice has been investigated. Given the open FA solution and, consequently, the impossibility to duly
8 The peak linear heat rating has been retrieved from the hot FA maximum power density calculated by ERANOS. The latter has been assumed as an indicative value of the hot pin/spot, since 15 (axial) and 3 3 (XY directions) calculation points have been considered for each FA. 9 Allowing to accommodate uncertainties ensuing from both deterministic calculations and the approximations required by the lack of punctual results.
63
tune the coolant flow rate according to each FA power, two radial enrichment regions have been foreseen in order to smooth the coolant outlet temperatures through a careful flattening of power distributions. Satisfactory results have been obtained: a maximum distribution factor among FAs of 1.2 has been achieved at EoC, and preliminary T/H analyses have showed that the postulated safety limits are respected with fairly good margins, being the maximum fuel centerline temperature of the order of 2180 C, and the peak cladding outer surface temperature around 550 C at EoC (corresponding to the maximum burn-up scenario). The foremost goal of designing a high power density LFR DEMO assuring a high fast neutron flux so as to enable efficient irradiation of fuels and materials has been successfully accomplished: indeed, BoC/EoC reference configurations feature average power densities of 327/328 W cm3, and peak total neutron fluxes of 7.27/ 7.54$1015 cm2 s1 (nearly three times as much as ELSY ones) with corresponding average values of 5.1/5.4$1015 cm2 s1 and 16/15% fast fractions (E > 0.82 MeV). References Artioli, C., et al., 2008. Open Square Fuel Assembly Design and Drawings, ELSY Project, Technical Report Deliverable D6. EURATOM. Artioli, C., et al., 2009 ELSY Neutronic Analysis by deterministic and Monte Carlo methods: an innovative concept for the control rod systems, Proceedings of the 2009 International Congress on Advances in Nuclear Power Plants (ICAPP ’09), Tokyo, Japan, May 10e14. Bortot, S., et al., 2009. Preliminary Core Characterization for an ELSY-Oriented Demonstrative Reactor. Proceeding of Nuclear 2009, Pites¸ti, Romania. May 27e29. Bortot, S., et al., 2010. Preliminary core characterization of a generation IV lead fast reactor DEMO: goals, design rationales and options. Energy Conversion and Management 51, 1806e1812. Cinotti, L., et al., 2007. The Potential of the LFR and the ELSY Project, Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP 2007), Nice Acropolis, France, May 13e18. Cinotti, L., 2009. Private Communication. Generation IV International Forum e GIF 002-00. 2002. Joint Evaluated File (JEF) project, 2006. The JEFF-3.1 Nuclear Data Library. Technical Report JEFF Report 21. OECD/NEA. Konno, K., Hirosawa, T., 2002. Melting temperature of mixed oxide fuels for fast reactors. Journal of Nuclear Science and Technology 39, 771e777. Krepel, J., et al., 2009. EQL3D: ERANOS based equilibrium fuel cycle procedure for fast reactors. Annals of Nuclear Energy 36, 550e561. Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, 2007. Materials Compatibility, Thermalhydraulics and Technologies. OECD-NEA. Muller, G., et al., 2002. Results of steel corrosion tests in flowing LBE at 420e600 C after 200 h. Journal of Nuclear Materials. 301, 40e46. Rimpault, G., et al., 2002. The ERANOS code and data system for Fast Reactor neutronic analyses, In: Proceedings of the International Conference on the PHYSics Of Reactors 2002 (PHYSOR2002), Seoul, Korea. Rimpault, G., 1997. Physics documentation of Eranos: The Ecco Cell Code. Technical Report DER-SPRC-LEPh-97e001. CEA. Ruggieri, J.M., et al., 1993. TGV: a Coarse Mesh 3 Dimensional Diffusion-transport Module for the CCRR/ERANOS Code System. Technical Report DRNR-SPCILEPh-93e209. CEA. Sarotto, M., et al., 2009. ELSY Core Design, Static Dynamic and Safety Parameters with the Open Square FA, Technical Report Deliverable D8. EURATOM. Thetford, R., Sobolev, V., 2005. Recommended Properties of Fuel, Cladding and Coolant for EFIT, Report Deliverable 3.1.4. EUROTRANS. Weisenburger, A. et al., 2007. T91 Cladding Tubes with and without Modified FeCrAl Coatings Exposed in LBE at Different Flow, Stress and Temperature Conditions, IV International Workshop on Materials for HLM Cooled Reactors and Related Technologies, Rome, Italy, May 21e23.
Contents lists available at SciVerse ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Conceptual core design study for a high-flux LFR demonstrator Sara Bortota, *, Patrizio Console Camprinib, Giacomo Grassoc, Carlo Artiolic a
Politecnico di Milano, Department of Energy, CeSNEF-Nuclear Engineering Division, via Ponzio 34/3, 20133 Milano, Italy University of Bologna, Nuclear Engineering Laboratory (LIN) of Montecuccolino, via dei Colli, 16, 40136 Bologna, Italy c National Agency for the New Technologies, Energy and Sustainable Economic Development (ENEA), via Martiri di Monte Sole 4, 40136 Bologna, Italy b
a r t i c l e i n f o
a b s t r a c t
Article history: Received 6 January 2011 Received in revised form 31 August 2011 Accepted 1 September 2011
A preconceptual core design study for a pool-type Lead-cooled Fast Reactor (LFR) demonstrator (DEMO) has been developed in the frame of an I-NERI between the Italian National Agency for the New Technologies, Energy and Sustainable Economic Development (ENEA) and Argonne National Laboratory (ANL), based on the European Lead-cooled System (ELSY) reference concept. A demonstration reactor is expected to prove the viability of technology to be implemented in the first-of-a-kind industrial power plant. DEMO specifications as a nuclear power facility demonstrating ELSY main features and performance, besides validating design methodology and tools, have been defined. Suitable design parameters have been set to meet the foremost objective of reaching a high fast neutron flux while respecting all technological constraints. Preliminary thermal-hydraulic analyses have been carried out to verify safety limits were not exceeded. Ó 2011 Elsevier Ltd. All rights reserved.
Keywords: Core design Demonstration Generation IV Lead-cooled fast reactor LFR Neutronics
1. Introduction The Lead-cooled Fast Reactor (LFR) is under development worldwide as a very promising fast neutron system to be operated in a closed fuel cycle (Generation IV International Forum, 2002). In particular, within the 6th EURATOM Framework Program, the European LFR community proposed the ELSY (European Leadcooled SYstem) concept (Cinotti et al., 2007), an innovative 600 MWe pool-type LFR fully complying with Generation IV goals of economics, safety, proliferation resistance and sustainability, the latter being led by the feature of no net production of Minor Actinides (MAs). As recognized by the Strategic Research Agenda worked out by the European Sustainable Nuclear Energy Technology Platform (SNETP), LFR full development up to industrial deployment requires e as a fundamental intermediate step e the realization of a demonstrator (DEMO) of some hundreds MWth, aimed at validating LFR technology as well as the overall system behavior. In order to define a first reference configuration for a DEMO, an I-NERI (International Nuclear Energy Research Initiative) between the Italian National Agency for the New Technologies, Energy and * Corresponding author. Tel.: þ39 02 2399 6327; fax: þ39 02 2399 6309. E-mail addresses: [email protected] (S. Bortot), [email protected] (P. Console Camprini), [email protected] (G. Grasso), [email protected] (C. Artioli). 0149-1970/$ e see front matter Ó 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2011.09.001
Sustainable Economic Development (ENEA) and Argonne National Laboratory (ANL) has been signed in 2007, under the umbrella of the EURATOM-DOE Agreement. On the Italian side the work has been carried out in the frame of the national R&D program on “New Nuclear Fission” supported by the Italian Minister of Economic Development (MED) through a general Agreement with ENEA and the Italian University Consortium CIRTEN. In such a context, a reference configuration of a pool-type LFR DEMO has been developed under a cooperation between ENEA, Politecnico di Milano and University of Bologna. A demonstration reactor is expected to prove the viability of technology to be implemented in the first-of-a-kind industrial power plant. Therefore, the first step toward DEMO conceptual design has been the definition of its specifications as a nuclear power facility able to both validate ELSY main features and performances, and to qualify numerical codes and tools (Bortot et al., 2009). Based on the results of preliminary scoping analyses (Bortot et al., 2010), an optimization study has been performed in order to accomplish a reference core characterization. Suitable design parameters have been set so as to meet the foremost objective of reaching a high fast neutron flux while respecting all technological constraints, among which peak cladding and fuel centerline temperatures, and peak neutron-fluence. A fuel cycle hypothesis consistent with the maximum burn-up and cladding dpa allowed in ELSY has been adopted, and
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
depletion calculations have been performed in order to determine the reference Beginning of Cycle (BoC) and End of Cycle (EoC) core characteristics (i.e., effective multiplication factor keff, power profiles and neutron fluxes), as well as reactivity coefficients and kinetic parameters. Two different and independent Control Rod (CR) systems have been foreseen to guarantee the required reliability for reactor shutdown and safety, the first one addressing reactivity swing compensation over the cycle, regulation and safe shut-down, and the second one being devoted to scram only. Preliminary thermal-hydraulic analyses have finally been performed with reference to the hot channel parameters, in order to verify a posteriori that safety limits were not exceeded. 2. Core design approach 2.1. Aimed performances Two essential missions have been envisioned for DEMO as a GEN-IV LFR demonstrator: comprehensive demonstration of successful lead-cooled reactor operation and testing. Due to this broad spectrum approach, R&D priorities and planned efforts focus on neutronics, thermal-hydraulics, materials, control systems and strategies, systems and components, safety and performance requirements. The main objectives regarding neutronics are mainly related to the need of demonstrating the key issues related to the adiabaticity of a core (i.e. characterized by a unitary breeding ratio and burning its self produced MAs), which is a very crucial point for an actual sustainability with respect to both natural resources exploitation and environmental impact minimization (Sarotto et al., 2009). To this end, it is necessary to test the system behavior when at least some MA-bearing fuel assemblies (FAs) are hosted in DEMO core; this would allow to validate numerical codes outcomes in terms of both nuclear data and reactor physics methodology. Furthermore, overall core performances e such as peak burn-up e should be evaluated. Consequently, the driving criterion in designing DEMO core has been the achievement of a high neutron (fast) flux, so as to enable efficient irradiation of fuels and materials to high burn-ups over reasonable irradiation times. DEMO plant is also requested to attest the worth and operation reliability of the ELSY innovative Finger Absorber Rod (FAR) system. In particular, one has to verify that adopted calculation tools provide accurate estimations of FAR reactivity worth, as well as that the control system is suitable as for compensation, local power regulation and safe shut-down. Material extensive testing is not among the main goals of the demonstration project; nevertheless, DEMO core materials (fuel, cladding and structures, coolant) are to be the same of ELSY ones in order to guarantee technological consistency with the industrialsize LFR nuclear power plant. 2.2. Reference criteria Conventional MOX (reactor grade, RG, Pu; 1.97 stoichiometric ratio; 95% theoretical density) has been considered as the reference fuel (isotopic vectors showed in Table 1). Annular pellets have been required to both moderate centerline temperatures e enabling to afford 365 W cm1 peak linear heat rating, which is compatible with the limit of 2400 C maximum fuel temperature e and lower the average density so as to reach high burn-ups (beyond 100 GWd t1 on heavy metal, HM) without affecting the fuel pin integrity (Bortot et al., 2010). Due to the choice of T91 ferritic-martensitic (FMS) steel (elemental composition given in Table 2) as the candidate material
57
Table 1 Fuel isotopic vectors. Isotope
Depleted U [wt.%]
RG Pu [wt.%]
234
0.003 0.404 0.010 99.583 e e e e e
e e e e 2.333 56.873 26.997 6.104 7.693
U 235 U 236 U 238 U 238 Pu 239 Pu 240 Pu 241 Pu 242 Pu
for both fuel rods and structures due to its high irradiation resistance and promising ongoing R&D, 600 C maximum operational temperature has been assumed, since it is the limit expected to be affordable in the near future whether superficial treatments (namely aluminization by GESA, currently under development and testing at Karlsruhe Institute of Technology and ENEA-Brasimone) show to be reliable over time against corrosion (Weisenburger et al., 2007; Lead-bismuth Eutectic Alloy and Lead Properties, 2007). Furthermore, a limit on peak neutron fluences (w4$1023 cm2 for ferritic steels) has been considered. A coolant inlet temperature of 400 C is required in order to avoid excessive embrittlement of structural materials subjected to fast neutron flux and to provide a sufficient margin from the lead melting point (327 C). The same coolant outlet temperature as in ELSY (480 C) has been kept, since such a choice enables to reach higher values of neutron flux thanks to higher fuel power densities allowed (Bortot et al., 2010); in addition, the whole system would be more representative of ELSY, being the operational characteristic temperatures e and, in general, conditions e reproduced. Furthermore, the coolant velocity has been limited to 3 m s1 (Weisenburger et al., 2007; Muller et al., 2002) to prevent the cladding from erosion risks. As to FA specifications, a ductless design with pins arranged on a square lattice has been focused on. Every FA must be provided with four structural uprights at corners and a central FMS T91 control rod driveline allowing the fast insertion of FARs. A safety requirement demanded to be necessarily met pertains to the control system reactivity worth and redundancy: the need of two different and independent absorber sets has been assumed, guaranteeing a 3000 pcm minimum margin for safe shut-down purposes, besides a suitable insertion speed and passive reliability. A primary system would be required to have sufficient reactivity worth to bring the reactor from any operating condition (i.e. overpower together with reactivity fault) to cold sub-critical at the refueling temperature with the most reactive control element stuck at the full power operating position. In addition, the primary system should serve the purpose of compensating for the reactivity effects of fuel burn-up and axial growth, while accommodating both the criticality-uncertaintyrelated reactivity and fissile loading. A secondary system would be intended to shut the reactor down from any operating condition to the hot standby status, in turn with the most reactive assembly inoperative. Also for the secondary system design the reactivity fault is included as a requirement, since it should
Table 2 FMS T91 isotopic composition. Element
Cr
Mo
Nb
Mn
Si
Ni
V
Fe
[wt.%]
9.0
1.0
0.1
0.6
0.5
0.2
0.2
88.4
58
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
Fig. 1. Core layout (left) and 2-D RZ cylindrical section (right); dimensions in cm. A single FAR is represented in the equivalent RZ scheme to clarify the absorber set out-of-core positioning.
override the uncontrollable withdrawal of one primary control assembly used for burn-up control. Finally, DEMO should have an affordable cost; this imposes a limitation on the system size. Based on preliminary economical evaluations (Cinotti, 2009), a 250e300 MWth reactor power seems to be a reasonable trade-off between DEMO plant cost and performances in terms of ELSY representativeness. Such a figure is compatible with the opposite requirement of shielding the active zone with a sufficient number of dummy elements, and of limiting the axial and radial core size so that the reactor tank does not exceed the dimensions compatible with both costs and sloshing risks minimization.
2.3. Optimization approach The starting point of DEMO neutronics design has been the work carried out in (Bortot et al., 2010): the scoping analyses performed on a 250 MWth MOX-fueled core demonstrated that a 6 mm clad outer diameter,1 associated with 365 W cm1 peak linear power, assures high power densities and, consequently, high values of neutron flux. Starting from those established fuel pin specifications, an effort has been undertaken to conceive a very compact core configuration with the main objective of lowering the required plutonium fraction under its acceptable limit. In this perspective, classical massive control rods have been forsaken: compensation, regulation and scram tasks have been assigned to FARs, since such a solution enables to significantly reduce neutron leakage, besides preserving ELSY representativeness. With the same aim, an optimization work has been carried out concerning core dimensions. An effort has been spent to strike a balance between enlarging the core axial extension e so as to accomplish a more critical
1
An attempt of enhancing the neutron flux has been pursued as far as possible by raising the fuel power density up, i.e. moving toward small values of pin diameter. Since an important constraint is that, beyond some point, small pins become increasingly expensive to fabricate, a lower limit on this parameter has been set at 6 mm.
geometry e and not losing criticality by reducing the fuel volume fraction (VF) due to the coolant flow area enhancement required to compensate any core height increase.2 Indeed, scoping analyses outcomes highlighted that material volume fractions impact on criticality more strongly than the geometrical shape of the core, in this range of scenarios. Consequently, it has been opted for increasing the coolant velocity from 2 m s1 to its maximum acceptable limit of 3 m s1 in order to leave a margin to downsize the pin pitch. Reference peak and average linear heat ratings have been preserved and it has been possible to meet both the needs through a pin pitch reduction (from 8.71 to 8.53 mm) and a concurrent core axial extension increase (from 42 to 65 cm) abiding by lead outlet temperature requirements. According to the core size constraint, the need of a fuel assembly size comparable with ELSY one3 and the above-mentioned shielding requirements, a suitable FA design has been accomplished. Then, aiming at keeping the fresh fuel plutonium fraction below the postulated limit of 35 wt.%, FAs have been arranged in a symmetrical and compact staggered configuration consistent with 300 MWth total power; such a new size has been found to comply with the above-mentioned economics constraint. A further optimization has been pursued following DEMO core geometrical settlement. Since for ELSY a power/FA distribution factor of at most 1.2 has been evaluated suitable to assure temperature limits to be respected with a proper safety margin (Artioli et al., 2008), such a reference has been borrowed for DEMO. As required by the wrapper-less solution choice,4 two radial enrichment regions have been foreseen in order to smooth the coolant outlet temperatures through a careful flattening of power distributions. A first iteration has been made to optimize both the definition of the two enrichment zones and
2 Since a pin pitch enlargement aimed at enhancing the mass flow rate would be required to reinstate the enthalpy balance between core outlet and inlet. 3 To host the same internal mechanisms without miniaturizing cost issues. 4 It does not permit to exploit different FA orificing to locally regulate the coolant flow rate.
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
59
Fig. 2. Fuel assembly outline.
the mutual ratio between the respective fuel enrichment figures pursuant to the reference distribution factor. The average plutonium fraction required for criticality has been determined accordingly. At that point it has been necessary to assume a fuel cycle strategy in order to evaluate DEMO main neutronic performances, as keff swing over the equilibrium cycle, BoC and EoC power
distributions, neutron fluxes, etc. A simplified hypothesis has been adopted: evolution calculations have been performed according to a 1-batch approximation,5 i.e. by postulating a homogeneous core
5 It has been demonstrated such an approach gives accurate predictions of the effective core behavior during the cycle (Krepel et al., 2009).
60
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
burn-up while defining ideally the BoC and EoC core configurations as the one corresponding to the core average situation immediately after refueling and the one corresponding to the core average situation immediately before refueling, respectively (Krepel et al., 2009). The fuel residence time has been set so that the neutron-fluence-related constraint was not exceeded; a firstattempt calculation has been carried out on the basis of the Beginning of Life (BoL) peak neutron flux e since it was the only figure available e , and a more accurate assessment has ensued from considering the EoC corresponding value. The consequent total time interval has been divided into four refueling sub-cycles (open 4-batch cycle, simulated again according to a 1-batch approximation) in order to allow the cycle reactivity swing to be manageable by compensation FARs; then, plutonium fractions have been readjusted so as to reach keff ¼ 1 at EoC while respecting the postulated power/FA distribution (considering that at BoC a local power modulation is granted by the compensation FARs insertion, whereas at EoC absorbers are completely withdrawn from the active core). 2.4. Calculation tools and reactor model Neutronics analyses have been performed by means of ERANOS (European Reactor ANalysis Optimized System) deterministic code ver. 2.1 (Rimpault et al., 2002), coupled with JEFF-3.1 (Joint Evaluated File, 2006) data library. Multi-group cross-sections associated to every reactor zone have been retrieved from ECCO (European Cell COde) (Rimpault, 1997) calculations.
Table 3 DEMO major core specifications, cold dimensions (20 C). Specification Reference goal parameters Thermal power Average coolant outlet temperature Coolant inlet temperature Average coolant speed Peak surface cladding temperature Core data Number of inner/outer fuel assemblies Number of shielding assemblies FA data FA pitch Pin lattice Number of pins/FA Pin pitch CR driveline inner/outer width Fuel pin data Cladding outer diameter Cladding thickness Pellet outer diameter Pellet hole diameter Fuel smear density Fuel column height Gas plenum height Total pin height Fuel Pu fraction inner/outer zone
Value 300 480 400 3.0 600 10/14 74
Units MWth C C m s1 C
e e
238.84 28 28 744 8.53 45.65/48.65
mm e e mm mm
6.00 0.34 5.14 1.71 0.95 650 650 1300 29.3/32.2
mm mm mm mm e mm mm mm wt.%
FAR data Cladding outer diameter Cladding thickness Gap B4C pellet stack height
44.0 0.60 0.10 850
mm mm mm mm
Radial reflector assembly data Assembly size Lead volume fraction Steel volume fraction
238.84 85 15
mm wt.% wt.%
Table 4 Fuel cycle hypothesis outline. Fuel Residence Time Months
¼ core
¼ core
¼ core
¼ core
0 5 10 15 20
0 5 10 15 20/0
0 5 10 15/0 5
0 5 10/0 5 10
0 5/0 5 10 15
A very refined cell description has been adopted for both FAs and sub-critical zones surrounding the active core (with the exception of radial reflector dummy assemblies) and containing FARs, for which an accurate heterogeneous geometry model has been set up; the remaining zones (namely ‘foot FA’, ‘dummy belt’, ‘barrel’ and ‘external lead’, see Fig. 1) have been modeled homogeneously. In both cases, ECCO computations have been carried out by treating the main nuclides with a fine energy structure (1968 groups) and condensing the obtained cross-sections into 33 groups. ECCO evaluations have been used to set detailed tri-dimensional core models up; then the whole system has been solved through nodal transport calculations by means of the ERANOS TGV/ VARIANT module (Ruggieri et al., 1993). 3. DEMO core neutronics characterization 3.1. Core layout In the current design, DEMO features a 300 MWth MOX-fueled core, composed by ductless FAs. Fig. 1 shows (on the left) a section of the core composed by 24 FAs surrounded by dummy assemblies, and (on the right) a bi-dimensional RZ representation of the vertical section that yields a useful scheme for the main structural zones. Ensuing from criticality assessment and power flattening, 10 FAs with 29.3 wt.% Pu fraction constitute the inner zone, and 14 FAs with 32.2 wt.% Pu fraction compose the outer zone. As far as FAs design, fuel pins are arranged in a 28 28 square lattice; every FA is provided with 4 structural uprights at corners connected with the central FMS T91 box beam replacing 6 6 central positions (Fig. 2). In Table 3 the main DEMO core specifications are summarized. 3.2. Fuel cycle hypothesis and burn-up calculations Owing to the approach described above (x 2.3), a fuel cycle of 20month fuel residence time and 5 month refueling interval has been adopted. Table 4 presents the residence time of each quarter of the core: according to the open 4-batch cycle refueling strategy, bold values highlight batches undergoing refueling, the two figures referring to the batch aging just before/immediately after reloading. Since the real fuel average residence times in the reactor are 12.5/7.56 months (i.e. 375/225 days) before/after refueling, these figures have been considered as the equilibrium cycle extremes, representing just the core conditions at EoC and BoC, in accordance with the 1-batch approximation (Krepel et al., 2009). As mentioned in x 2.3, the two radial enrichments have been tuned to obtain keff ¼ 1 and a power/FA distribution of 1.2 at EoC. In Fig. 3 the evolution of keff during irradiation is presented.
6 At EoC and BoC the mean aging of the fuel results indeed 1/4(5 þ 10 þ 15 þ 20) and 1/4(0 þ 5þ10 þ 15) months, that is: 12.5 and 7.5 months respectively.
k ef f
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
1.12 1.1 1.08 1.06 1.04 1.02 1 0.98 0.96 0.94 0.92
61
BOL
BOC 4000 pcm
B7
B8
EOL
100
enriched B4C FARs
42 at.
enriched B4C FARs
42 at.
enriched B4C FARs, extracted
EOC
A4 A5
0
90 at.
200
300
400
500
600
A1 700
B6 A2
B3
Fig. 4. ¼ core layout with control/regulation and safety FARs positioning in the core.
Time [days]
As shown in the graph, the calculated reactivity swing during the cycle is slightly greater than 4000 pcm. The overcriticality at BoC (about 4000 pcm) is expected to be compensated by FARs, whose positioning has been, in turn, optimized in order to duly flatten the radial power distribution at BoC as well (x 3.3). 3.3. Control rod system design Two different and independent CR systems have been envisaged to guarantee the required reliability for reactor shut-down and safety, for which at least 3000 pcm have been requested to be supplied by each absorber set. Four passive7 B4C (90 at.% enrichment in 10B) FARs have been foreseen exclusively for SCRAM purposes. Since in normal operation they are to be positioned outside the active core, FAs characterized by the lowest power (i.e., power distribution factor lower than 1.2) e and therefore not requiring any flattening by regulation FARs e have been considered suitable for shut-down FARs. Then, their location has been determined so as to maximize absorber efficacy while respecting the core symmetry, as indicated in Fig. 4. Furthermore, a motorized FAR system demanded for cycle reactivity swing control and safe shut-down has been foreseen. Every remaining FA has been provided with a B4C (42 at.% enriched in 10B) FAR so that the required worth (w7000 pcm, i.e. 3000 for scram and 4000 for reactivity swing compensation) was assured. A subset of 42%-enriched FARs, corresponding to the four most peripheral positions (coldest FAs), has been conceived to fulfill only scram functions and thereby they lie 1 cm above the active zone in operating conditions, whereas FARs assigned also to local regulation are partially inserted in the core over the entire equilibrium cycle. For a proper assessment of both control systems efficacy, a correction factor reckoning with the strong approximations involved in deterministic analyses has been necessarily applied to the calculated FARs reactivity worth. In fact, FARs limited dimensions introduce a large uncertainty in reactor spatial calculations, despite the associated fine cell analysis description. As a result, macroscopic cross-sections representing FAs with FARs inserted over-estimate the absorbers effectiveness, basically due to a raw treatment of the self-shielding phenomenon brought by homogenization and B4C dilution over the whole cell (Artioli et al., 2009). An attempt of quantifying the influence of cell homogenization has been pursued by comparing the kN parameter resulting from
Table 5 Compensation/regulation and safety system worth. Control System and insertion share
Worth at BoC [pcm] Aimed
Actual
Aimed
Actual
Safety FARs e complete insertion Regulation FARs e complete insertion Regulation FARs subset - 32.5 cm insertion
3000 7000 4000
4624 11,312 4083
3000 7000 e
4856 11,923 e
Worth at EoC [pcm]
homogeneous and heterogeneous cell calculations on inner/outer FAs with/without FARs. For both inner and outer FAs without FARs, discrepancies between homogeneous and heterogeneous models have turned out to be slight (some hundreds of pcms), whereas for FAs with FARs inserted they have turned out to be particularly relevant (some thousands of pcms). Hence, the respective absorber worth has been adjusted accordingly, the SCRAM FAR one being reduced by 25% and the regulation/compensation FAR one by 20%. Active and passive control systems reactivity worth has been assessed at BoC and EoC: as it is shown in Table 5, both FARs sets meet the required safety criteria. In addition to global evaluations, the negative reactivity due to the regulation FAR subset progressive penetration into the active zone has been figured so as to determine the insertion share required for criticality at BoC (Fig. 5). Different compensation/regulation FAR subset insertion strategies have been further investigated in order to derive benefit from power/FA distribution, according to the two concurring needs of complying with the stringent design requirements ensuing from the limit on maximum cladding temperature while maximizing the neutron flux. Neutronics calculations showed that no FARs differentiated insertion is needed to accomplish the aimed 1.2 power/FA distribution, since their uniform introduction suffices for the required power flattening; hence, the corresponding configuration,
Reactivity worth [pcm]
Fig. 3. keff(t) behavior from Beginning of Life (BOL, fresh core condition -singularity-) to End of Life (EOL); ERANOS/JEFF-3.1 calculations.
10000 9000 8000 7000 6000 5000 4000 3000 2000 1000 0
k eff 1 0
5
10
15
20
25
30
35
40
45
50
55
60
FARs insertion [cm] 7 Scram FARs are driven by gravity. A sudden insertion is allowed by their positioning 1 cm above the active zone.
Fig. 5. Regulation/compensation FAR set reactivity insertion curve.
65
62
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
Table 6 DEMO major neutronics results.
Table 7 Reactivity coefficients and kinetic parameters.
Parameter
BoC
EoC
Unit
Parameter
BoC
EoC
Units
Multiplication factor, keff Reactivity, r Total peaking factor Peak power/FA distribution factor Average power density Peak total neutron flux, 4 Neutron flux fast fraction (E > 0.82 MeV) Average/peak discharge burn-up
1.00033 33 1.43 1.20 327 7.27 16
1.00093 93 1.40 1.20 328 7.54 15
e pcm e e W cm3 1015 cm2 s1 %
Doppler constant Doppler coefficient (from 1173.15 to 1500 K) Lead density coefficient (active core) Axial expansion coefficient (linked) Radial expansion coefficient Delayed neutron fraction, b Prompt neutron lifetime, L
235 0.18
268 0.20
pcm pcm K1
MWd kg1(HM)
90/130
0.14 0.19 0.92 323 8.4980
pcm K1 pcm K1 pcm K1 pcm 107 s
superposition principle validity and on the assumption that the system responds linearly within the interval defined by the reference and the perturbed configurations. Consequently, the respective reactivity coefficients have been calculated according to (Sarotto et al., 2009):
with all the 16 regulation FARs inserted for half active height, has been assumed as a reference for BoC.
8 > > > > <
2
6 dr vr vr vr 6 !þ zlT91 ðTinlet Þ 26 > 4 d d dT radial R P > absorber dPfuel > v > v :v R Pabsorber Pfuel 9 3> > > > 7= vr 1 vr 7 þ 1 5 þ ð1Þ dPsteel dPcoolant > VFcoolant > > v v > ; Psteel Pcoolant
3.4. Neutronics characterization at BoC and EoC Following the design rationales and coherent core specifications outlined in the previous paragraphs, a final BoC and EoC characterization has been accomplished. DEMO foremost neutronics results are resumed in Table 6. Both BoC and EoC configurations fulfill the aimed requirements (x 2.1): the driving purpose of designing a high power density core so as to reach high fast neutron fluxes has been achieved (in particular, about three times as much as ELSY one) while complying with all the reference criteria (x 2.2). In Fig. 6 BoC and EoC core radial power density profiles are shown.
2
0.12 þ 0.01 0.87 319 8.0659
and
13 6 Z Tclad; out C7 B 6 vr dr 1 vr vr vr dLinsertion C7 B ! z lclad ðTÞ dT 6 þ C7 B 6 dZ 7 þ vL A @ d dT axial Tclad; out Tclad; in Tclad; in dT P insertion dPfuel steel 4v 5 v v Z P steel P 0
(2)
fuel
3.5. Reactivity coefficients and kinetic parameters Reactivity coefficients and kinetics parameters have been calculated with ERANOS by employing both direct methods and the perturbation theory formalism for the steady-state conditions. Elementary perturbations have been introduced in order to figure the radial (1) and axial (2) expansions through a “partial derivative” approach. Such a method is based on the main hypothesis of
where the last term in Eq. (2) represents the reactivity contribution due to the differential expansion of FARs with respect to the active core (r being reactivity, P material density, l linear expansion coefficient, L length, T temperature, R radial coordinate, Z axial coordinate, and VFcoolant coolant volume fraction). The axial expansion coefficient has been evaluated for the linked case, i.e. when dilation is assumed to be driven by the cladding. As far as the calculation of the Doppler coefficient is concerned, a temperature enhancement by 326.85 K has been considered with respect to the initial average MOX temperature of 1173.15 K.
500 BoC
Power density [W cm-3 ]
450
Table 8 Preliminary T/H evaluations (EoL, hot channel).
EoC
400 350 300 250 200 150 100 0
10
20
30
40 X [cm]
50
60
70
Fig. 6. Radial power density distribution at BoC and EoC (core midplane).
Parameter
Value
Unit
Lead inlet temperature Lead outlet temperature Lead velocity Sub-channel mass flow rate Hydraulic diameter Peak linear power Fuel internal surface peak temperature Fuel external surface peak temperature Cladding internal surface peak temperature Cladding external surface peak temperature
400 501 3.0 1.4 9.4 362 2182 716 574 550
C C m s1 kg s1 mm W cm1 C C C C
S. Bortot et al. / Progress in Nuclear Energy 54 (2012) 56e63
2400
Temperature [ C]
2000 Fuel centre temperature Fuel surface temperature Clad surface temperature Coolant temperature
1600 1200 800 400 0 0
10
20 30 40 Active core elevation [cm]
50
60
Fig. 7. Hot channel temperature axial profiles evaluated at EoL.
For the coolant density coefficient, a 5% reduction of lead density has been applied. In Table 7 the relative results are summarized, together with the main kinetic parameters.
4. Preliminary T/H evaluation A preliminary thermal-hydraulic analysis has been performed in order to verify a posteriori that safety limits were not exceeded (in steady-state, nominal power). Calculations have been carried out with reference to the End of Life (EoL) core configuration; furthermore, hot channel parameters8 have been taken into account in order to consider the most critical conditions. A clad superficial coating and degradation of the fuel thermal conductivity with burn-up (Thetford and Sobolev, 2005) have been considered (assuming a 20-month-long fuel pin life, i.e. some 130 MWd kg1 peak burn-up). Conditions and results are provided in Table 8. The postulated safety limits (2400 C for fuel (Konno and Hirosawa, 2002) and 600 C for clad) are respected with fairly good margins,9 being the maximum fuel centerline temperature of the order of 2180 C, and the peak cladding outer surface temperature around 550 C. Temperature axial profiles in the hot pin at EoL nominal conditions are shown in Fig. 7. 5. Conclusions In this paper the static neutronics characterization of a GEN-IV LFR DEMO has been presented, along with preliminary thermohydraulic core analysis. Suitable design parameters have been set so as to meet the foremost objective of reaching a high fast neutron flux while respecting all technological constraints. A 300 MWth MOX-fueled core, composed by ductless less FAs with pins arranged on a square lattice has been investigated. Given the open FA solution and, consequently, the impossibility to duly
8 The peak linear heat rating has been retrieved from the hot FA maximum power density calculated by ERANOS. The latter has been assumed as an indicative value of the hot pin/spot, since 15 (axial) and 3 3 (XY directions) calculation points have been considered for each FA. 9 Allowing to accommodate uncertainties ensuing from both deterministic calculations and the approximations required by the lack of punctual results.
63
tune the coolant flow rate according to each FA power, two radial enrichment regions have been foreseen in order to smooth the coolant outlet temperatures through a careful flattening of power distributions. Satisfactory results have been obtained: a maximum distribution factor among FAs of 1.2 has been achieved at EoC, and preliminary T/H analyses have showed that the postulated safety limits are respected with fairly good margins, being the maximum fuel centerline temperature of the order of 2180 C, and the peak cladding outer surface temperature around 550 C at EoC (corresponding to the maximum burn-up scenario). The foremost goal of designing a high power density LFR DEMO assuring a high fast neutron flux so as to enable efficient irradiation of fuels and materials has been successfully accomplished: indeed, BoC/EoC reference configurations feature average power densities of 327/328 W cm3, and peak total neutron fluxes of 7.27/ 7.54$1015 cm2 s1 (nearly three times as much as ELSY ones) with corresponding average values of 5.1/5.4$1015 cm2 s1 and 16/15% fast fractions (E > 0.82 MeV). References Artioli, C., et al., 2008. Open Square Fuel Assembly Design and Drawings, ELSY Project, Technical Report Deliverable D6. EURATOM. Artioli, C., et al., 2009 ELSY Neutronic Analysis by deterministic and Monte Carlo methods: an innovative concept for the control rod systems, Proceedings of the 2009 International Congress on Advances in Nuclear Power Plants (ICAPP ’09), Tokyo, Japan, May 10e14. Bortot, S., et al., 2009. Preliminary Core Characterization for an ELSY-Oriented Demonstrative Reactor. Proceeding of Nuclear 2009, Pites¸ti, Romania. May 27e29. Bortot, S., et al., 2010. Preliminary core characterization of a generation IV lead fast reactor DEMO: goals, design rationales and options. Energy Conversion and Management 51, 1806e1812. Cinotti, L., et al., 2007. The Potential of the LFR and the ELSY Project, Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP 2007), Nice Acropolis, France, May 13e18. Cinotti, L., 2009. Private Communication. Generation IV International Forum e GIF 002-00. 2002. Joint Evaluated File (JEF) project, 2006. The JEFF-3.1 Nuclear Data Library. Technical Report JEFF Report 21. OECD/NEA. Konno, K., Hirosawa, T., 2002. Melting temperature of mixed oxide fuels for fast reactors. Journal of Nuclear Science and Technology 39, 771e777. Krepel, J., et al., 2009. EQL3D: ERANOS based equilibrium fuel cycle procedure for fast reactors. Annals of Nuclear Energy 36, 550e561. Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, 2007. Materials Compatibility, Thermalhydraulics and Technologies. OECD-NEA. Muller, G., et al., 2002. Results of steel corrosion tests in flowing LBE at 420e600 C after 200 h. Journal of Nuclear Materials. 301, 40e46. Rimpault, G., et al., 2002. The ERANOS code and data system for Fast Reactor neutronic analyses, In: Proceedings of the International Conference on the PHYSics Of Reactors 2002 (PHYSOR2002), Seoul, Korea. Rimpault, G., 1997. Physics documentation of Eranos: The Ecco Cell Code. Technical Report DER-SPRC-LEPh-97e001. CEA. Ruggieri, J.M., et al., 1993. TGV: a Coarse Mesh 3 Dimensional Diffusion-transport Module for the CCRR/ERANOS Code System. Technical Report DRNR-SPCILEPh-93e209. CEA. Sarotto, M., et al., 2009. ELSY Core Design, Static Dynamic and Safety Parameters with the Open Square FA, Technical Report Deliverable D8. EURATOM. Thetford, R., Sobolev, V., 2005. Recommended Properties of Fuel, Cladding and Coolant for EFIT, Report Deliverable 3.1.4. EUROTRANS. Weisenburger, A. et al., 2007. T91 Cladding Tubes with and without Modified FeCrAl Coatings Exposed in LBE at Different Flow, Stress and Temperature Conditions, IV International Workshop on Materials for HLM Cooled Reactors and Related Technologies, Rome, Italy, May 21e23.