DEMARCHE DE REDUCTION DES INCERTITUDES EN NEUTRONIQUE - PowerPoint PPT Presentation

DEMARCHE DE REDUCTION DES INCERTITUDES EN NEUTRONIQUE CEA/DEN/CAD/DER/SPRC 1

Outline 1. Computer-based simulations in neutronics 2. The validation process 3. Experimental validation 4. Examples 5. Conclusion CEA/DEN/DER/SPRC 2

1. Computer-based simulations in neutronics � Objective � To provide recommendations to users on how to use a given neutronics code in order to meet the users’ needs : � A well defined parametric application domain � Recommended calculation options or procedures Errors and uncertainties ∆ C to be assigned to the code predictions C � (calculating C alone is not sufficient) for a given application domain → “ formulaire ” � To provide recommendations to physicists on how to improve the models and data CEA/DEN/DER/SPRC 3

1. Computer-based simulations in neutronics � Main “ingredients” ← nuclear data (JEFF) 1. Data libraries ← theoretical physics models of 2. Calculation codes and procedures + neutron/gamma transport v -1 ∂Ψ / ∂ t = H Ψ + S sensitivity calc. modules equations ∂ N/ ∂ t = AN ← experiments in reactors 3. Validation data CEA/DEN/DER/SPRC 4

1. Computer-based simulations in neutronics � Schematic flow diagram File of evaluated nuclear data (JEFF…) Domain System of validation Development Targeted range of methods Interpretation Recommended of application and codes of integral calculation measurements procedures Characteristics of Definition of interest and target recommended Validation Errors and accuracies procedures uncertainties Integral experiments CEA/DEN/DER/SPRC 5

1. Computer-based simulations in neutronics � CEA neutronics codes � APOLLO-2 + CRONOS-2 (SAPHYR) � PWR, BWR, HTR, … sub-assembly calculations (rods, plates…) and core calculations � Special procedures NARVAL (naval reactors), HORUS3D (RJH) � ERANOS � FR sub-assembly and core calculations � TRIPOLI-4 � Monte Carlo calculations � DARWIN � Fuel depletion, nuclide inventory and source calculations � CRISTAL � Criticality-safety calculations CEA/DEN/DER/SPRC 7

2. The validation process – Nuclear data Differential Trends/Priorities Measurements (1) Needs Sensitivity Integral analyses measurements Calculation-vs.-experiment comparisons Modelling Validation (5) & Evaluation (2) Application libraries Tests & Compilation (3) Statistical Processing Adjustment Users (4) JEFF File JEFF-3 CEA/DEN/DER/SPRC 11

2. The validation process – Nuclear data JEFF-3.0 vs. JEF-2.2 Inelastic Scattering Cross Section of Fe-56 CEA/DEN/DER/SPRC 12

2. The validation process – Nuclear data JEFF-3.0 vs. JEF-2.2 Radiative Capture Cross Section of Pu-240 CEA/DEN/DER/SPRC 13

2. The validation process – Calculation procedures � Distinguish � numerical validation = calculation-vs.- calculation comparisons using the same nuclear data Reference results may be provided by a Monte Carlo code � experimental validation = calculation-vs.- measurement comparisons � Methodology � Allows in principle to separate (and hence avoid compensations between) � Errors arising from the nuclear data � Errors arising from the methods / procedures and to suggest improvements on each of these � Has been systematically used at CEA for the past 10 years Is possible because of progress in computing power → Monte Carlo � calculations are becoming routine → method biases under control � In practice, separation is achieved to a great extent but not fully CEA/DEN/DER/SPRC 14

2. The validation process – Calculation procedures � Schematic flow diagram Nuclear data file Calculation Monte Carlo C / C ref procedure Measurements Numerical C / E in reactors validation Experimental validation CEA/DEN/DER/SPRC 15

3. Experimental validation � The needed “integral” experiments must be Specific → representative of the targeted application range � Analytic → phenomena can be analysed individually � � As simple as possible in terms of geometrical arrangement, constituents, … � Sufficiently accurate � Sufficiently diverse � The experimental validation makes it possible to establish � If the quality of the nuclear data and models are sufficient to meet the application needs � To identify those data/models that require improvements and by how much CEA/DEN/DER/SPRC 16

3. Experimental validation � Physics measurements � in zero-power critical facilities such as EOLE, MINERVE, MASURCA, AZUR at CEA Cadarache � in power reactors → e.g., irradiations experiments � Zero-power reactors � are characterised by well-known constituents, operating conditions, and a high degree of flexibility in terms of core loading, geometrical arrangements, operation � allow measurements that are difficult or impossible to do in power reactors � can be modelled with very good accuracy (systematic errors are under control) � Power reactors � provide full-scale and actual operating conditions ( → coupled phenomena) � provide information on capture cross sections and fuel inventory � require more effort and some approximations in modelling CEA/DEN/DER/SPRC 17

3. Experimental validation � Integral and differential measurements are complementary from the standpoint of validating nuclear data evaluations � Differential measurements provide information of high energy/angle resolution but generally inaccurate in level � Integral measurements usually provide information of very good accuracy in level but poor resolution � NB: the errors and uncertainties affecting nuclear data are still quite large today CEA/DEN/DER/SPRC 18

4. Examples Top view of the EOLE reactor core CEA/DEN/DER/SPRC 19

4. Examples FUBILA MOX-3% FUBILA MOX-5% FUBILA MOX8.5% FUBILA MOX-11.5% MOX-3.0% MOX-8.5% Water rod AG3 rods for channel box simulation ∅ 13.2 mm MOX-5.0% MOX-11.5% EPICURE MOX-7.0% pilot rod guide-tubes X-Y View of the EOLE/FUBILA-Ref. 9x9 100% MOX Core (from P. Blaise and N. Thiollay, DER/SPEx) CEA/DEN/DER/SPRC 20

4. Examples � Measurements in EOLE Measurement type Validation purpose Critical water level k eff , k ∞ , M 2 Critical boron concentration (homogeneous cores) Radial and axial traverses by γ spectrometry or fission chambers 2 → k eff B m (homogeneous cores) Isotopic cross sections, neutron Spectral indices by γ spectrometry spectrum, conversion ratio (integral or specific peaks) or fission chambers Power distribution Worths of water holes, single Reactivity variation resulting from rod absorbers, rod clusters, burnable substitutions poisons, local voids Reactivity variation resulting from a change in Isothermal moderator temp. and the water temperature and density density coefficients β eff , β eff / Λ “Pulsed source”, Neutron noise ... Etc. CEA/DEN/DER/SPRC 21

4. Examples Top view of the MINERVE reactor core CEA/DEN/DER/SPRC 22

4. Examples � Measurements in MINERVE Reactivity variation is ∆ρ = < φ *, ∆ H φ ’> / < φ *, F’ φ ’> Top view of the MINERVE/MELODIE core CEA/DEN/DER/SPRC 23

4. Examples JEF-2.2 Trends Derived from FP Sample Oscillations in MINERVE R1-UO 2 R2-UO 2 very thermal core thermal core Fission (C-E)/E 1 σ exp. (C-E)/E 1 σ exp. in % in % Product Unc. (%) Unc. (%) - 4.5 2.9 - 3.3 3.6 Sm σ Sm149 underestimated: 149 Sm - 6.0 2.9 - 4.9 3.6 - 5% ± 2% 147 Sm + 1.3 4.3 + 2.7 4.7 152 Sm - 1.6 2.9 - 1.8 3.7 σ Nd143 underestimated: Nd + 0.4 3.0 - 3.3 3.7 - 4% ± 2% 143 Nd - 7.1 3.1 - 8.5 3.8 Confirmed by fuel analyses 145 Nd + 0.4 3.8 + 1.1 4.4 of Nd144 formation 155 Gd - 2.5 2.9 - 6.1 4.0 103 Rh + 11.0 4.0 + 8.0 4.2 103 Rh - - + 14 9.0 109 Ag - 3.6 4.3 - 4.5 4.3 σ Rh103 overestimated: 109 Ag - 4.6 9.0 + 2.8 6.9 + 10% ± 3% - 4.7 4.2 + 0.3 4.7 Ag Mo + 1.5 3.2 + 2.1 3.8 133 Cs - 0.6 3.8 - 2.4 4.3 133 Cs + 4.1 8.5 + 9.1 7.3 CEA/DEN/DER/SPRC 24

4. Examples JEFF-3.0 vs. JEF-2.2 Radiative Capture Cross Section of Sm-149 (from O. Serot) Γ tot (meV) Γ n (meV) Γ g (meV) E res (eV) Spin Comment -0.285 3 62.17 0.16914 62.00 unchanged Γ n increased +0.0973 4 61.05 0.549 60.5 +0.872 4 60.54 0.7422 59.8 unchanged 140000 Recommended 3% increase in the first Capture Cross Section (b) T=293.6 K T=293.6 K resonance Γ n , compatible with the 120000 JE JEFF3.0 FF3.0 measurement performed by Pattenden 100000 80000 Thermal Resonance 60000 Value (b) Integral (b) 40000 JE JEF2.2 F2.2 JEF-2.2 40446 3487 20000 JEFF-3.0 41617 3490 (+2.9%) (+0.1%) 0 1E-3 0.01 0.1 1 En (eV) CEA/DEN/DER/SPRC 25

4. Examples Analysis of irradiated fuel rods Q P O N M L K J I H G F E D C B A 1 2 Ex: Gravelines UOX 4.7% 3 4 5 NB: Detailed modelling required 6 3 7 2 8 9 1 0 2 4 1 1 1 2 1 3 1 4 1 5 1 6 1 7 c ra y o ns a na ly s é s : N tro u d'e a u N no m bre de c y c le s Ass FF06E2BV tube guide CEA/DEN/DER/SPRC 26