Technical workshop : Dynamic nuclear fuel cycle

Transcription

1 Technical workshop : Dynamic nuclear fuel cycle Reactor description in CLASS Baptiste LENIAU* Institut d Astrophysique de Paris 6-8 July, 2016

2 Introduction Summary Summary The CLASS package : a brief overview Problematic of reactor description h i! N 0 Mean cross section «prediction» Fresh fuel construction

3 The CLASS package : a brief overview CLASS : general informations First code line wrote at Subatech in Soon after the LNC-IRSN joins the collaboration. CLASS is a C ++ library ~15000 code lines 3 Involved laboratories Subatech, Nantes IPNO, Orsay LPSC,Grenoble LNC, IRSN, Fontenay aux Roses Publications Articles : 3 (in 2015) Conferences : 7 Past and present PHD : 3 3 teaching workshops Main developers B.Mouginot (post-doc) 2011->2015 F. Courtin (PHD) 2014->2017 N.Thiollière > - B.Leniau (post-doc) 2013->2016 GIT/SVN repository, forum, documentation, manual : CeCILL License (in progress) Open source code and models. Each publication using CLASS should explicitly cite CLASS and models references

4 The CLASS package : a brief overview CLASS : a nuclear fuel cycle code As any fuel cycle code, CLASS aims to describe «units» which store and/or transform and exchange matter. The goal being to calculate isotops inventories in units and material flux between units. Mines Infinite storage Storage Store decay Reprocessing FP AM Store Waste decay Fresh fuel manufacturing Mix isotops and/or chemical elements Store Decay Enrichment plant Enrichment fresh fuel Reactors Fuel depletion Spent fuel Chemical separation Spent fuel Spent fuel pool Store decay Storage StockageStorage Décroissances Store decay 4 exemple of possible connexions

5 The CLASS package : a brief overview CLASS input : a C ++ interface Exemple of a CLASS input : Storage : Spent fuel pool : Reactor : 5

6 The CLASS package : a brief overview Decay handling (out core) Mines Infinite storage Storage Store decay Reprocessing FP AM Store Waste decay Fresh fuel manufacturing Mix isotops and/or chemical elements Store decay Enrichment plant Enrichment fresh fuel Reactors Fuel depletion Spent fuel Chemical separation Spent fuel Spent fuel pool Store decay Storage StockageStorage Décroissances Store decay 6

7 The CLASS package : a brief overview Decay handling (out core) Nucleus inventory Daughter1 inventory Daughter2 inventory Sum of pre-calculated ascii tables Time portion of a pre-calculated decay table of ONE NUCLEUS ( 241 Pu) Example : Considering a sample made of : N 1 nuclei of 241 Pu and N 2 nuclei of 238 Pu What is the composition of this sample 3 month after? portion of 241 Pu table! Daughter 1 portion of 238 Pu table! Daughter 2 N 1 N 2! Daughter 1 +! Daughter 2 Sample with 3m of decay 7

8 The CLASS package : a brief overview Reactor handling Mines Infinite storage Storage Store decay Reprocessing FP AM Store Waste decay Fresh fuel manufacturing Mix isotops and/or chemical elements Store decay Enrichment plant Enrichment fresh fuel Reactors Fuel depletion Spent fuel Chemical separation Spent fuel Spent fuel pool Store decay Storage StockageStorage Décroissances Store decay 8

9 The CLASS package : a brief overview Reactor handling : two cases Fixed fresh fuel (recipe based) Dynamic fresh fuel Infinite mine of fresh fuel always the same Storages Reactor Pool Reactor Pool Reprocessing Storage Storage 9

10 The CLASS package : a brief overview Reactor handling Fixed fresh fuel (recipe based) EvolutionData : pre-calculated fuel depletion Infinite mine of fresh fuel Reactor always the same Interpolated spent fuel at user defined cycle time Pool Storage 10

11 Reactor problematic Reactor problematic : fuel from reprocessed materials Determine the fresh fuel composition and predict its irradiation in reactor Fuel composition suitable for the reactor Trial composition Fuel manufacturing k e (t) ~N(t = t i ) ~N(t = 0) = ~ N 0 Boltzmann equation R h i = Bateman s equations (E) (E) (E).dE R (E).dE Storages k e (t) Cooling Chemical separation ~N(t = t f )= ~ N f h i and strongly depends on, which is unknown a-priori How to avoid calls to time consuming neutron transport code?! N 0 11

12 Reactor problematic Reactor problematic : fuel from reprocessed materials Determine the fresh fuel composition and predict its irradiation in reactor Fuel composition suitable for the reactor Storages Trial composition Fuel manufacturing dk eff (t) Cooling Chemical separation ~N(t = 0) = ~ N 0 Neutronic datas predictor Bateman s equations ~N(t = t f )= ~ N f ch i Approach used : Replace the neutron transport code by statistical predictors 12

13 Reactor problematic Methodology : Building a reactor model for CLASS dk eff (! N 0,t)! N 0 Define fresh fuel isotopic variation domain Fuel depletion calculations (Boltzmann/ Bateman) {h i (t)} Generation of a training data base Multivariate regressions (Artificial neural networks) n ch i(! N0,t)o Statistical estimators d ~ N dt =A~ N Reactor model for CLASS Existing models : PWR-UOx PWR-MOX (this talk) PWR-MOX on enriched uranium support. (Fanny COURTIN s talk, this afternoon) SFR MOX In development : PWR-MOX with Americium (A. Zakari s talk Friday morning) ADS CANDU U from reprocessing 13

14 Mean cross section prediction Example : PWR MOX Cross section predictors {h i (t)} n ch i(! N0,t)o Statistical estimators d ~ N dt =A~ N Hyperspace of isotopic composition of the fresh fuel Depletion calculations (coupled Bateman/ Boltzmann solvers) Data base generation Multivariate interpolation (Artificial neural network) Pu hyperspace The plutonium hyperspace is generated using 6 different PWR-UOx fuel depletion calculation (representative of the french fleet) at difference irradiation time. Pu content [4%;15%] 235 U variation in U vector [0.2%;0.4%] 14

15 Mean cross section prediction Example : PWR MOX Cross section predictors {h i (t)} n ch i(! N0,t)o Statistical estimators d ~ N dt =A~ N Hyperspace of isotopic composition of the fresh fuel Depletion calculations (coupled Bateman/ Boltzmann solvers) Data base generation Multivariate interpolation (Artificial neural network) Main characteristics : Assembly 17x17 lattice 100% MOX Fixed power density : 30 W/g MOX composition : Variable Main simulation parameters : Infinite calculation (mirror boundaries) Impact of UOx neighbor not taken into account Fixed boron content : 600ppm MCNP steps : 11 Multi-group : yes (17900 groups) Data Bases (XS,FPYields,S(α,β)) : JEFF Half life threshold : 1 hour Running time (1cpu) : ~2 hours 15

16 Mean cross section prediction Example : PWR MOX Cross section predictors {h i (t)} n ch i(! N0,t)o Statistical estimators d ~ N dt =A~ N Hyperspace of isotopic composition of the fresh fuel Depletion calculations (coupled Bateman/ Boltzmann solvers) Data base generation Multivariate interpolation (Artificial neural network) N different fresh fuel compositions Depletion calculations N fuel depletion calculations using the code : N sets of results: {h i (t)} ~2h per depletion calculation on 1 CPU 16

17 Mean cross section prediction Example : PWR MOX Cross section predictors {h i (t)} n ch i(! N0,t)o d ~ N dt =A~ N Hyperspace of isotopic composition of the fresh fuel Depletion calculations (coupled Bateman/ Boltzmann solvers) Data base generation Multivariate interpolation (Artificial neural network) Statistical estimators Using the depletion calculations to train ANNs to predict : ch i = f( ~ N 0,t) (Power is constant) One neural network per reaction { Number of reactions : Nuclei considered : 365 Reaction type handle : (n,γ), (n,f) and (n,2n) Number of reactions :

18 Mean cross section prediction PWR MOX model testing performances Testing procedure : h ANN Error = c i N h i test h i test ch i N h i test Mean cross section predict by ANN trained with N MURE calculations results Mean cross section calculated with MURE NOT USED FOR TRAINING For each reaction and for each training sample size : Plot each (mean,σ) on a graph Mean σ mean error Example :Error distribution on 241 Pu(n,f) prediction ANN Error Std. dev. on error 18

19 Mean cross section prediction Testing performances : fission Mean 19 σ

20 Mean cross section prediction Testing performances : capture Mean (n,γ) 20 σ

21 Mean cross section prediction Testing performances : (n,2n) Mean A X(Energy threshold, MCNP statistical error) σ 21

22 Mean cross section prediction (n,2n) statistical error (E) (E) h i = R (E) (E).dE R (E).dE Error more important for (n,2n) reactions : (n,2n) : Threshold Reaction ->Low statistic in fast region of neutron spectrum (Monte Carlo!) ->High statistical error on (n,2n) -> Noisy mean cross section ->ANN can t be better than statistical error on the training sample! 22

23 Mean cross section prediction Induce errors in fuel depletion calculation Methodology : Reference Generation of 500 random fresh fuel compositions in the hyperspace previously defined. Depletion calculations (MURE) with these 500 different fresh fuels CLASS models output The same 500 fresh fuel compositions are used as input for the mean cross sections predictor These cross sections are used by the Bateman solver of CLASS The end of cycle inventories are compared 23

24 Mean cross section prediction Induce errors in fuel depletion calculation How to read? Y axis: Proportion in spent fuel X axis: R e l a t i v e d i f f e r e n c e b e t w e e n a M U R E c a l c u l a t i o n a n d a depletion calculation with predicted <σ> via ANN 24

25 Mean cross section prediction Impact of the 3 reactions assumption CLASS 3 reactions (XS from MURE) VS MURE all reactions all reaction : 3 reactions + (n,3n) (n,α) (n,p) 238 U(n,3n) 238 Pu(n,3n) 138 La(n,p) 25

26 Fresh fuel construction Example of building a MOX fuel for a PWR Pu content? } } IV1 IV1 IV2 IV2 IV3 IV3 IV4 IV4 Plutonium Dep. Uranium Storages Isotopic compositions determined during fuel cycle simulation 1 - Pu content Fresh Fuel PWR-MOX What is the Pu content for the reactor to be critical? 26

27 B. LENIAU PISE project Introduction Fresh fuel construction The averaged k 1 model The fuel management in CANDU/PWR/BWR/SFR uses batches to flatter the flux profile and to flatter the k eff evolution (fresh fuel assemblies compensate de lake of criticality of the old a fuel loading : 1/N assemblies are loaded, I/N are moved in the core,...,i/n are unloaded For PWR : N = 3 or How to mimic batch management with infinite calculation? Considering that the behavior of an assembly isn t impacted by its surrounding media, we defined the average k 1 over the batches as : <k 1 > batch (t) = 1 N NX 1 i=0 T : Irradiation time (N loading/unloading) k (t + it 1 N ) frac k k k threshold

28 B. LENIAU PISE project Introduction Fresh fuel construction The averaged k 1 model <k 1 > batch (t) = 1 N NX 1 i=0 k (t + it 1 N ) T : Irradiation time (N loading/unloading) 28 frac k k k threshold

29 B. LENIAU PISE project Introduction Fresh fuel construction The averaged k 1 model <k 1 > batch (t) = 1 N NX 1 i=0 k (t + it 1 N ) T : Irradiation time (N loading/unloading) The reactor has to verify : k eff =1 8t 2 [0,T] <k 1 > batch =1+Leaks + CRods 29 frac k k k threshold

30 B. LENIAU PISE project Introduction Fresh fuel construction The averaged k 1 model <k 1 > batch (t) = 1 N NX 1 i=0 k (t + it 1 N ) T : Irradiation time (N loading/unloading) The reactor has to verify : k eff =1 8t 2 [0,T] <k 1 > batch =1+Leaks + CRods 30 frac k k k threshold <k 1 > batch = k threshold

31 B. LENIAU PISE project Introduction Fresh fuel construction The averaged k 1 model <k 1 > batch (t) = 1 N NX 1 i=0 k (t + it 1 N ) T : Irradiation time (N loading/unloading) The reactor has to verify : k eff =1 8t 2 [0,T] <k 1 > batch =1+Leaks + CRods 31 frac k k k threshold <k 1 > batch = k threshold as [boron] is kept constant in simulation the criticality conditions is : <k 1 > batch (T/N)=<k 1 > batch (2T/N)=... = k threshold

32 B. LENIAU PISE project Introduction Fresh fuel construction The averaged k 1 model The plutonium content in a fresh fuel is such as the criticality condition is verified : Criticality condition : <k 1 > batch (T/N)=k threshold 32 The time T is the maximum irradiation time (Burnup) a given fuel can reach : BUmax (Given by user) Method used to find maximum average discharged burnup in candu & pwr : Nuttin, A., et al. "Comparative analysis of high conversion achievable in thorium-fueled slightly modified CANDU and PWR reactors." Annals of Nuclear Energy 40.1 (2012): This model involves to build : a k 1 (t) predictor : an algorithm to find the Pu content according to the criticality condition b k1 (t)

33 B. LENIAU PISE project b k1 (t) Introduction Fresh fuel construction : Artificial neural network (ANN) approach 33 Total number of events : 5000 depletion calculation & 11 time step each : events. Train with half (randomly picked) and tested with the other half

34 Fresh fuel construction b k1 (t) Results of the ANN testing procedure 34 Mean error : 30 pcm Std. dev. on error: 200 pcm b k1 (t) k 1 (t) k 1 (t) MCNP Statistical error : ~ 150 pcm B. LENIAU PISE project

35 PWR-MOX Model b k1 (t) Results of the ANN testing procedure 35 Mean error : 30 pcm Std. dev. on error: 200 pcm b k1 (t) k 1 (t) k 1 (t) MCNP Statistical error : ~ 150 pcm B. LENIAU PISE project

36 LENIAU Baptiste Scénarios électronucléaires 36 PWR-MOX fresh fuel construction Model algorithm Which Pu content %Pu = f

37 LENIAU Baptiste Scénarios électronucléaires 37 PWR-MOX fresh fuel construction Model algorithm Which Pu content to reach a given burnup %Pu = f(bu target, BU target : Inputs wanted burnup

38 LENIAU Baptiste Scénarios électronucléaires 38 PWR-MOX fresh fuel construction Model algorithm Which Pu content to reach a given burnup with a given batch management, %Pu = f(bu target,n, Inputs BU target : wanted burnup N : # of batches

39 PWR-MOX fresh fuel construction Model algorithm Which Pu content to reach a given burnup with a given batch management, knowing isotopic compositions of Pu et U %Pu = f(bu target,n,! P u,! U,k th ) { BU target :! Pu :! U : Inputs wanted burnup N : # of batches Fissile composition Fertile composition } LENIAU Baptiste Scénarios électronucléaires 39

40 PWR-MOX fresh fuel construction Model algorithm Which Pu content to reach a given burnup with a given batch management, knowing isotopic compositions of Pu et U %Pu = f(bu target,n,! P u,! U,k th ) { %Pu = P0 //Initialization BU target :! Pu :! U : Inputs wanted burnup N : # of batches Fissile composition Fertile composition do { varying BU until <k>(bu/n)= kth using eq 1 & 2 : BUmax eq 1 : <k 1 > batch (t) = 1 N NX k 1 (t + it N ) i eq 2 : b k1 (BU, ~ N) ~N =%Pu ~ Pu+(1 %Pu) ~ U } LENIAU Baptiste Scénarios électronucléaires 40

41 PWR-MOX fresh fuel construction Model algorithm Which Pu content to reach a given burnup with a given batch management, knowing isotopic compositions of Pu et U %Pu = f(bu target,n,! P u,! U,k th ) { %Pu = P0 //Initialization BU target :! Pu :! U : Inputs wanted burnup N : # of batches Fissile composition Fertile composition do { varying BU until <k>(bu/n)= kth using eq 1 & 2 : BUmax eq 1 : <k 1 > batch (t) = 1 N NX k 1 (t + it N ) i } if(bumax > BUtarget) decrease %Pu else increase %Pu //%Pu too high //%Pu too low eq 2 : b k1 (BU, ~ N) ~N =%Pu ~ Pu+(1 %Pu) ~ U } LENIAU Baptiste Scénarios électronucléaires 41

42 PWR-MOX fresh fuel construction Model algorithm Which Pu content to reach a given burnup with a given batch management, knowing isotopic compositions of Pu et U %Pu = f(bu target,n,! P u,! U,k th ) { %Pu = P0 //Initialization BU target :! Pu :! U : Inputs wanted burnup N : # of batches Fissile composition Fertile composition do { varying BU until <k>(bu/n)= kth using eq 1 & 2 : BUmax eq 1 : <k 1 > batch (t) = 1 N NX k 1 (t + it N ) i } if(bumax > BUtarget) decrease %Pu else increase %Pu } while ( BUmax!= BUtarget ) return %Pu //%Pu too high //%Pu too low //accept %Pu or continue looping eq 2 : b k1 (BU, ~ N) ~N =%Pu ~ Pu+(1 %Pu) ~ U LENIAU Baptiste Scénarios électronucléaires 42

43 Benchmark Benchmark with COSI PWR-MOX Model Work accomplished in the framework of a CEA/CNRS collaboration (2014) Pu content predicted by the two codes for 5 different Pu isotopy : N =3 k threshold = [1.01 ; 1.05] BU target = 50 GW d/t! Pu = variable! U =0.25% 235 U k threshold { FreshFuel #1 36.5% 53.6% FreshFuel #2 24.6% 5.8% 4.6% 1.6% 1.3% 6.5% 14.5% 14% 28.8% 8.2% FreshFuel #3 45.6% 55.2% FreshFuel #4 3.9% 0.2% 7.6% 31.3% 11.3% 2.9% 2.8% 9% 20.1% 10.1% FreshFuel #5 71% 1.9% 0.7% 3.4% 6.1% 238Pu 239Pu 240Pu 241Pu 242Pu 241Am 16.8% 43 B. LENIAU PISE project

44 Conclusions & outlooks Mean cross sections interpolation : Good precision of the ANN approach on <σ> prediction. For PWR-MOX : -(n,f) : std. dev. on error < 1 % -(n,γ) : std. dev. on error < 2 % -(n,2n) : std. dev. on error < 5 % (for actinides) Discrepancy due to Monte Carlo statistical error Performances in a depletion calculation: Error (compared as MURE calc.) ~3% EOC for main actinides Computational time ~30 s This methodology is used for all reactor description in CLASS 44

45 Conclusions & outlooks Fresh fuel construction PWR-MOX model : Based on batch averaged kinfinite. kinfinite(t) prediction using MLP (std. dev. 200pcm) Flexible : possibility to change kthreshold and number of batches Pu content prediction close to COSI(CEA) with kthreshold = 1.03 List of CLASS equivalence models : 45

46 Mean cross section prediction Testing performances : Fission prediction Proportion of error due to statistical error Mean ANN error Std. Dev. on error 46

47 Mean cross section prediction Testing performances : (n,2n) prediction Proportion of error due to statistical error Mean ANN error Std. Dev. on error 47