VALMOX VALIDATION OF NUCLEAR DATA FOR HIGH BURNUP MOX FUELS

Transcription

1 VALMOX VALIDATION OF NUCLEAR DATA FOR HIGH BURNUP MOX FUELS B. Lance, S. Pilate (Belgonucléaire Brussels), R. Jacqmin, A. Santamarina (CEA Cadarache), B. Verboomen (SCK-CEN Mol), J.C. Kuijper (NRG Petten) SUMMARY VALMOX, an acronym for Validation of Nuclear Data for High Burnup MOX Fuels, is one of the projects of the cluster EVOL (Evolutionary Fuel Concepts : High Burnup and MOX Fuels). It covers 30 months, from October 2001 to March It considers the evaluation of the actinide inventory of MOX fuel at high burnup (typically 60 GWd/t) in Light Water Reactors, with special attention to the helium production. Calculated values for the spent fuel isotopic masses are compared to the measured ones, with sensitivity analyses made in support. The JEF 2.2 nuclear data file is taken as a basis for calculation. The resulting recommendations on nuclear data should be employed in the preparation and testing of the next JEFF3 file. So far, the major effort was placed on the evaluation of MOX fuel irradiations in Pressurised Water Reactors, and first results will be presented and compared. A. INTRODUCTION VALMOX is a follow-up of the former project of the Fourth Framework Programme ( ) : Nuclear Data for Advanced MOX Fuels, the results of which have been reported in ref.1, with some salient issues presented in ref.2. While this project aimed at improving the knowledge of actinide cross-sections and their transmutation properties, the present project is based on irradiations of commercial MOX fuels at higher burnups and aims at better identifying the burnup limitations, related in particular to the helium buildup. Increasing the MOX fuel burnup has clear economical and ecological advantages. B. WORK PROGRAMME In France, MOX fuel is being irradiated in 20 different reactors of the 900-MWe class, loaded with 30% MOX. Well characterised pins have been extracted for CEA analysis from 2 of them : Saint-Laurent-des-Eaux B1 (SLB1) and Dampierre 1 (DA). In SLB1, a very large number of MOX fuel pins were discharged after 1, 2 and 3 cycles of reactor operation, reaching a maximum burnup of about 45 GWd/t. The results of chemical analysis are available. The MOX pins of the central zone, where the spectrum is asymptotic, had a total Pu enrichment of 5.6%. In DA, MOX pins have been extracted after 1, 2, 3, 4 and 5 cycles, reaching about 65 GWd/t. The chemical analysis at Cadarache recently became available for the 3- and 5-cycle pins, and the recalculations are just starting. In Belgium, while MOX fuel is loaded in the 900-MWe reactors of Tihange-2 and Doel-3, BN and SCK are conducting the ARIANE programme of spent fuel analysis for

2 different MOX-loaded reactors ; this includes irradiation experiments which took place from 1990 to 1996 in the 400-MWe PWR of Beznau-1, in the assemblies M109 (10 rods analysed) and M308 (6 rods analysed). Chemical analyses were made by 3 laboratories in parallel, one of which is SCK. The total plutonium enrichment was respectively 5.9% in M109 and 5.5% in M308. Three Work Packages have been defined. WP1 and WP2 group the evaluation of the MOX fuel analyses described above : WP1 corresponds to the Belgian group (BN, SCK) and WP2 to the French group (CEA). WP1 covers in addition MOX fuel in BWRs, also irradiated up to more than 60 GWd/t. WP3 comprises intercomparisons, sensitivities, and uncertainties, with all 4 partners active. The basic nuclear data file used by all partners is JEF 2.2 (CEA intends to also perform some recalculations with a preliminary, working JEFF3 file.) The programme package at CEA is APOLLO-2, see ref.3, in its validated version APOLLO2.5/CEA93.V6, ref.4. The calculation scheme is the one recommended to EDF and Framatome-ANP, and described in ref.5 ; it implies : - in the MOX region, 23 different pin types (i.e. each pin) individualised for the depletion calculations ; - the UOX region decomposed into 3 pin types : the outer pin row, the second outer row, and the rest ; group cross-sections to finely model the resonance region (especially near 1 ev for Pu240, Pu239) ; - a burnup chain with 20 actinides, 83 fission products (i.e. 95% of all fission products). BN and NRG use the WIMS-8a code package, ref.6. At BN, starting from a routine calculation scheme, which had been used in the preceding FP4 study contract, about 20 different variants of the model were calculated, to determine the best practical methods of calculation. The standard scheme recommended by WIMS authors, in 6 energy groups, using a transport code based on the characteristic lines method (CACTUS, see ref.7) could still be kept. At SCK they performed calculations in parallel to BN ones, with the Monte Carlo code MCNP coupled to a burnup routine, to test this method with respect to BN one. The work in WP3 is essential to identify the uncertainties of the whole evaluation work ; CEA and NRG are active there to qualify the recommendations on actinide crosssections, as well as on helium production in MOX fuels. C. MAIN PRELIMINARY RESULTS C.1 Comparison of Actinide Inventories Table I presents calculation-to-experiment ratios (C/E) for the actinide isotopes, with U238, the bulk of the fuel, taken as reference ; the first column ( SLB1 ) contains CEA results while the 2 other columns ( BM1 ) refer to BN work, for 2 schemes differing in the number of energy groups, 6 or 21. The latter should be more accurate ; in reality they hardly differ. The main observations on the CEA (SLB1) results are as follows.

3 Among the plutonium isotopes, the major ones Pu239, 240 and 241 are correctly predicted, though with an average overestimate for Pu239 (1.03). Pu238 and Pu242 are underestimated, and this requires more verifications. Am241 is quite correctly predicted, what is important for helium production. The other Am isotopes and the Cm isotopes all exhibit underprediction, which is very pronounced for 3 of them. The extensive parametric study engaged at BN on the sample BM1 ( 45 GWd/t) led to compare 20 different variants of the model. 12 of them are illustrated in Table II. The following major model refinements were retained : A differentiation of all MOX pins for the depletion and spectrum calculations ; A finer subdivision of the WIMS time steps (likely to better represent Am241) ; A geometrical representation of the MOX assembly with its three UOX neighbours, simulated by a 4-assembly cluster. Simplifications could be accepted, as they had nearly no influence on the results. The recalculation of sample BM5, irradiated up to 57 GWd/t, gave similar trends. On the other hand, calculations will be repeated with the new WIMS-9 version, just received. C2. Discussion of the Differences When comparing BN results with CEA ones, one observes the same trends for U235, Pu238 (same underestimate), Pu240 and 241 (C/E close to unity), Pu242 (same underestimate) and for Am242m-Am243 and all Cm isotopes (general underestimate). For Pu239 the overestimate is significantly larger at BN (1.08) than at CEA (1.03). For Am241, which is correctly predicted at CEA, BN finds a large overestimate : 1.08 at measurement time (4 to 7 years after end of irradiation, EoI) and 1.24 at EoI. An independent evaluation of ARIANE results by ORNL, ref.8, based on the same power history and experimental data, showed trends similar to BN ones. ORNL employed HELIOS and ENDF/B VI. They even found higher C/E ratios for Am241. To help resolve the discrepancies, all 4 VALMOX partners decided to recalculate the OECD/NEA benchmark BUC phase 4, see ref.9. The pin cell here is fairly simple, and any difference specifically due to the code package, (e.g. WIMS vs APOLLO) can better be revealed. For what concerns the discrepancy on Pu239, it is observed that the calculation of self-shielding in APOLLO2 by collision probabilities takes into account the anisotropy of the current at outer boundaries, whereas the corresponding module of WIMS (CACTUS) does not so. It is suspected that this could notably influence the U238 captures and the Pu239 build-up. BN has established close contacts with the WIMS developers to clarify this. Concerning Am241, the general underestimate observed by BN and CEA on Am242m, Cm242(-Cm243) and also Pu242 is consistent with a too small capture rate in Am241 and thus, a too large final content in Am241 (as observed at BN only). Efforts are devoted at CEA to evaluate other experimental results and also to re-examine the basic Am241 data, including the branching ratio from neutron capture.

4 C3. Work ongoing at SCK and NRG The work at SCK consists of using the BN data on geometry and irradiation conditions for the sample BM1, to perform calculations based on the Monte-Carlo code MCNP, applied in a 2D model (like in BN WIMS-8), coupled to a burnup routine. The code coupling MCB1C (Monte Carlo continuous energy Burnup Code), an integration of MCNP4C and a novel Transmutation Trajectory Analysis code, was implemented (ref. 10). MCB1C uses continuous energy cross-sections for transport and reaction rates prepared for various material temperatures on the basis of JEF2.2, JENDL3.2, ENDF/B- VI-8 and EAF99 files. Current results are not satisfactory, as the methods and data used did not allow to gain much more information with respect to the BN WIMS evaluations. In a trial to improve these Monte-Carlo results, 3 alternate routes are being explored. Cross-section sensitivity studies are performed at CEA and NRG. At NRG, a code CSS1SMAT has been developed, see ref.11, which calculates final nuclide densities for a given process time step, assuming a constant one-group flux and constant one-group microscopic cross-sections. It also solves the analytical derivatives of the Bateman equations with respect to initial densities, one-group microscopic crosssections and decay constants, hereby yielding the respective relative sensitivity matrix elements for the burnup time step under study. Changes in the neutron spectrum during the burnup history are taken into account by combining the information from consecutive steps. NRG takes as reference cases for the present work : - the analysis of the BM1 sample, as transmitted by BN ; - one or two reference irradiation cases selected by CEA. C4. Helium Production Helium production is just being calculated ; no special problem of method is appearing. Its evolution during cooling and storage needs to be determined, as helium production continues long after irradiation. Contacts were established with the experts in charge of modelling from gas production to gas release ; gas comprises helium as well as the usual fission gases Xe and Kr ; one aims at defining an indicator relating in a simplified way the pin internal pressure to the helium production. D. DISSEMINATION OF THE RESULTS Results of the VALMOX work will be transmitted to the group of evaluators of the JEF data files, working for the OECD/NEA ; this group can use VALMOX results to improve and qualify the new data file JEFF3. Indirectly, all users of the JEF data files will benefit from the VALMOX work. Among end users are : - electricity utilities, for which any increase in fuel burnup will immediately result in an economic gain (fuel cycle component of the electricity cost reduced accordingly) ; - safety regulators, who examine such applications and authorise burnup increases : the final VALMOX report may be used as reference in their verification work. E. CONCLUSIONS The VALMOX work is based on the best available analyses of spent, commercial MOX fuel, for burnups of 60 GWd/t and more. Increased MOX fuel burnups have clear

5 economic and ecological advantages. The close co-operation between the 4 partners allows a thorough level of method comparisons. Cross-checks are still being done to clarify the discrepancies concerning two important isotopes : - Pu239, the major Pu isotope, for which a difference by 5% is too large ; - Am241, one important source of helium gas, with Pu238, Cm242 and Cm244. Due to the late availability of the some of the experimental analyses, calculations are just starting for the irradiation programmes in the Dampierre PWR (at CEA) and in the Gundremmingen BWR (at BN). In fact 60% only of the work has been done, and not 80% as anticipated. Sensitivity studies are needed to confirm which actinide cross-sections are to be improved, before sending final results to the evaluators of the new data file JEFF3, who are the prime users. REFERENCES [1] Nuclear Data for Advanced MOX Fuels, Final Report of Contract n FI4I-CT , by S. Pilate et al., EUR EN, 2000 [2] Validation of Nuclear Data for Transmutation from the Evaluation of MOX Fuel Irradiations, by S. Pilate, R. Jacqmin et al, PHYSOR2000 Conf., Pittsburgh, May2000 [3] APOLLO Twelve Years After, by S. Loubière, R. Sanchez et al, Int. Top. Meeting on Mathematics and Computation, M&C99, Madrid, Sept [4] Qualification of the APOLLO2.5/CEA93.V6 Code for UOX and MOX fuelled PWR, by A. Santamarina, C. Chabert et al, PHYSOR2002 Conf., Seoul (Korea), October 2002 [5] Elaboration and Experimental Validation of the APOLLO2 Depletion Transport Route for PWR Pu Recycling, by C. Chabert, A. Santamarina and P. Bioux, PHYSOR2000 Conf., Pittsburgh, May 2000 [6] WIMS, the ANSWERS Software Package, A General Purpose Neutronics Code, AEA Technology, 1999 [7] CACTUS, a Characteristics Solution to the Neutron Transport Equations in Complicated Geometries, by M.J. Halsall, AEEW-R1291, Winfrith, April 1980 [8] Simulation of Mixed-Oxide ( ) Analysis Results, B.D. Murphy and R.T. Primm, Nucl. Sci. Eng. 142, (2002) [9] OECD/NEANSC Burnup Credit Benchmark Phase IV-B : Mixed Oxide (MOX) Fuels (updated February 2002) by P.R. Thorne, G.J. O Connor, R.L. Boden. [10] Methods devised by J. Cetnar, W. Gudowski and J. Wallenius at KTH Stockholm. [11] Sensitivity and uncertainty assessment associated to burnup calculations, by J.C. Kuijper et al., PHYSOR2000 Conf., Pittsburgh, May 2000.

6 Table I MOX Fuel Irradiated in SLB1 (central zone) and in Beznau1 (sample BM1) C/E Ratios for U, Pu, Am and Cm Isotope Masses Isotopes SLB1 BM1 Scheme A (6 groups) BM1 Scheme C (21 groups) U235 U236 1 Np Pu238 Pu239 Pu240 Pu241 Pu Am241 Am242m Am Cm243 Cm244 Cm Table II : Parametric study for the reference BM1 sample using the WIMS8a code package Case Main tested parameters Level of fuel temperature Single value of fuel temperature for the whole irradiation Refined XY meshing in the main transport calculation Energy groups structure Power and spectrum differentiation of H-MOX and M-MOX rods Radial subdivision (3 meshes) of the MOX pellets in the resonance treatment and in the multicell collision probability calculation Single moderator temperature and composition for the whole irradiation Single power rating for the whole irradiation Schemes A / B/ C/ D (see text) Burnup of the neighbouring UO2 fuel (20 or 30 GWd/t) Data libraries : 172 groups 69 groups (WIMS'97) Finer condensed groups structure (21 groups vs 6 groups) 1 After correction, taking into account the Leal, Derrien, Larson evaluation. 2 Values in italics correspond to measurements done by less than 3 laboratories.