Simulation of a Symbiotic Nuclear Scenario including Argentina and Brasil using CLASS F. Alderete Tommasi a,b, A. Bidauda a, B. Mouginot c, B. Leniau d, N. Thiollière d, X. Doligez e, F. Courtin d, A. Somani e, J.B. Clavel f, Z. Issoufou e, S. David e A Laboratoire de Physique Subatomique et Cosmologie, Université Grenoble Alpes, CNRS/IN2P3, 53 Avenue des Martyrs, Grenoble, France, 38000, bidaud@lpsc.in2p3.fr B Balseiro Institute, Bariloche, Argentina, C University of Wisconsin, Madison, Wisconsin, USA, D Subatech, Nantes, Ecole des Mines de Nantes, CNRS/IN2P3, France E Institut de Physique Nucléaire d'orsay, Université d'orsay, CNRS/IN2P3, France F Institut de Radioprotection et de Sureté Nucléaire, Fontenay aux Roses, France 1
Introduction Brasilian Nuclear Fleet - Angra I (PWR) 600 MWe - Angra II (PWR) 1200 MWe - Angra III (PWR) 1200 MWe (in construction) 2
Introduction Argentinian Nuclear fleet - Atucha I (PHWR) 300MWe - Atucha II (PHWR) 700MWe - Embalse (PHWR - CANDU) 600MWe 3
The CANDU reactor Very efficient neutron economy uranium A large spectrum of fuels can be used If Brazil to recycle, probable best destination for reprocessed see http://www-pub.iaea.org/mtcd/publications/pdf/te_1630_cd/pdf/iaea- TECDOC-1630.pdf Already done in PWR (ex in Cruas in France) and CANDU Limited extra complexity of Uranium fuel fabrication No need to enriched hot reprocessed uranium 4
The case of study : current situation Waste UOX Enriched UOX Brasilian PWR s Natural Uranium Argentinian CANDU 5
The case of study : proposal Vitrified Waste Enriched UOX Waste UOX Brasilian PWR s Mox Reprocessing Argentinian CANDU Natural Uranium 6
The case of study Cons Reprocessing UOX needed More complex Candu fuel fabrication Different used Candu fuel wastes (ex high Np) Pro Less Uranium sent to waste in Brasil Less need to buy or mine for Natural Uranium in Argentina 7
Core Library for Advanced Scenario Simulation Fabrication Plant Pool Reactor Storage https://forge.in2p3.fr/projects/classforge 8
The Physics Model «Point reactor» approach : the evolution of 1 reactor simulated by solving 1 Bateman Equation Fuels are recycled -> XS and fissile needs depends on the scenario => Need for reactor models Equivalence Model How to build the fresh fuel for this reactor Cross Section Model How will the average cross sections evolve as this fuel is burned? Irradiation Model Numerical method used to solve the Bateman Equations 9
The Physics Models Before this work Physics Models for UOX and MOX PWR and SFR BUT X X No models for Natural U fueled PHWR No models for Reprocessed U fueled PHWR 10
Composition of the used fuel of brasilian reactors 45GWd/ton and initial enrichment of 4,25% Uranium Isotope Mass [kg] Percentage [%] 230 U 1,59 10-28 3,4 10-31 231 U 0 0 232 U 7,30 10-5 1,6 10-7 233 U 6,49 10-5 1,4 10-7 234 U 0.46 0,00099 235 U 413,968 0,89 236 U 260,061 0,56 237 U 2,30 10-6 4,9 10-9 238 U 45926,5 98,55 240 U 3,14 10-14 6,73805 10-17 11
Creation of a Fuel Management File with MURE Assembly calculations prepared with MCNP Utility for Reactor Evolution (MURE *) Two files were generated: CANDU_Unat CANDU_Urep Uranium Isotope Percentage [%] 235 U 0,7 238 U 99,3 Uranium Isotope Percentage [%] 235 U 0,89 236 U 0,56 238 U 98,55 see NEA-1845 or http://lpsc.in2p3.fr/mure/html/mure/mure.html 12
Estimation of Burnup for CANDU Time Average assumption Final burnup calculated so as average kinf hits some target Constant leakage Assumption <Kinf> for Reprocessed uranium taken from equivalent natural uranium case Keff Keff 1.25 1.25 1.2 1.2 1.15 1.1 1.1 1.05 1.05 1 1 0.95 0.95 k = eff 1 t t 0 k eff (t')dt' Kinf evolution is dependent on UOX BU (and then initial inventory) BU Unat = 7,2 MWd/t = 288 FEPD BU UREP (@45 GWd/t)= 14,5 MWd/t = 490 FEPD 0.9 0.9 k k eff eff 0.85 0.85 0 100 200 300 400 500 600 700 800 900 0 100 200 300 400Time 500 [days] 600 700 800 900 Time [days] 13
Estimation of Load Factors (Brazil) ANGRA I ANGRA II 14
Estimation of Load Factors (Argentina) EMBALSE 15
Estimation of Load Factors Reactor Slope from the linear fit [MWd/year] Load factor [%] Pre vio us Yea rs Ext rap ola tio n ANGRA I 1982-1994 1.24 10 6 22.6 ANGRA I 1994-2009 3.29 10 6 61.6 ANGRA I 2009-2016 4.45 10 6 83.4 ANGRA II 2000-2016 9.13 10 6 81.6 EMBALSE 1983-2055 4.51 10 6 85.8 ANGRA I 2016-2055 -- 83,4 ANGRA II 2016-2055 -- 81,6 ANGRA III 2016-2055 -- 81,6 16
Brasilean alone «Waste» Uranium Uranium Mass [kg] 3,00E+06 2,25E+06 1,50E+06 7,50E+05 0,00E+00 1990 2007,5 2025 2042,5 2060 Mass Produced in Brasil Time [years] 17
Symbiotic Scenario Results Uranium Balance 8,00E+05 6,00E+05 4,00E+05 2,00E+05 140 0,00E+00 120 1990 2007,5 2025 2042,5 2060 U Produced by Brasil - Urep consumed 100by Embalse P 0 P 0 Storage2. StockMOX. 237 Np Inside 237 Np Inside Mass [kg] 80 60 40 20 0 1990 2000 2010 2020 2030 Time [year] Np 237 stocks 18
Uranium Mass [kg] 4,00E+06 3,00E+06 2,00E+06 1,00E+06 Uranium Savings 0,00E+00 Time [years] 2010 2022,5 2035 2047,5 2060 Reprocessed Uranium Natural Uranium Uranium Consumed in Embalse Difference mainly due to the change in burnup U235 entering nuclear reactors 19
Results Recycled uranium PHWR physics better understood PHWR natural and recycled uranium simplified models done Decrease of 47tonU/year in the uranium sent directly to waste from the Brazilian PWR s. Embalse can be fed only with reprocessed Uranium Decrease of 81tonU/year in the consumption of natural Uranium in Argentina. SYNERGETIC COLLABORATIONS ARE EFFICIENT! 20
Future work Improve Equivalence model to cope with dynamically evolving incoming uranium vectors Make models of the other Argentinian specific reactors Simulate Pu recycling of the separated spent fuel in fuelling Brazilian reactor with MOX. SYNERGETIC COLLABORATIONS ARE EFFICIENT! Increase international internship exchanges > 100 nuclear engineering master students / y in Grenoble INP! 21
Are there any questions? Thanks for listening 22