4. GROUP DISCUSSIONS
4.2. Validation and criticality safety criteria
4.2.2. Codes and Nuclear Data Libraries
4.2.2.1. Summary of codes and nuclear data used by participants
A summary of information on codes used by the participants is presented in Table 4.1. It is seen that there is a wide range of techniques including deterministic and Monte Carlo methods and covering a range of energy group schemes.
4.2.2.2. Verification – Validation – Qualification Process
The first step in establishing the accuracy of a code/data package is usually based on numerical testing of the code (reliable algorithms, no bugs and regression in new code version) and its nuclear data library. This “Verification” phase normally includes checks to ensure that the processed cross-sections accurately represent the information contained in the basic nuclear data file.
“Validation” of the code may then be made by comparison with reference methods, such as continuous energy Monte Carlo. This validation process which aims for calibrating calculation biases (multigroup assumption, resonance self-shielding model, anisotropic scattering, etc) is generally carried out on representative simplified geometries (numerical benchmarks). Both standard route calculation and reference calculation must use the same nuclear data file.
Once this validation process has been performed, the global accuracy of the code/nuclear-data package is validated through benchmark experiments. This third phase is called
“Experimental Validation” or “Qualification”. In the context of BUC the qualification process
generally consists of analyzing PIE data (to validate depletion calculations) as well as critical
benchmark experiments and Spent Nuclear Fuel (SNF) worth measurements (to validate
criticality, i.e. k
effcalculations). Some participants pointed out that, in the specific case of
criticality safety, numerical and experimental validation are often gathered in a unique phase,
as the general goal remains to compare Calculated (C) to Experimental (E) k
effvalues and to
acquire thus knowledge of the C/E in k
effof the code scheme, according to the considered
application.
Table 4.1: CODES USED FOR BURNUP CREDIT CALCULATIONS
Depletion Criticality Country/Organisation
Code Data Code Data
Belgium/Tractebel LWRWIMS UKNDL MCNP ENDF-BV CE*
Belgium/BN WIMS8 JEF2.2 KENO5A DANTSYS
JEF2.2 172g
Bulgaria/VVER-440 SCALE4.4 ENDF-BV SCALE4.4 ENDF-BV 44g
Bulgaria/VVER-1000 NESSEL/NUKO ENDF-BIV SCALE4.4 ENDF-BV 44g Czech Republic SCALE4.4a/5 ENDF-BV SCALE4.4a/5 ENDF-BV 44g
France
Germany/GRS KENOREST-98 JEF2.2 KENOREST-98 JEF2.2
Germany/WTI (PWR)
SCALE4.4,
HELIOS/SNF ENDF-BV SCALE4.4,
MCNP ENDF-BV/VI
Japan/JAERI SRAC JENDL3..2 VIM JENDL3..2 CE
Japan/JNES SRAC
Russia/Kurchatov KASSETA BROND MCU BROND/ENDF
Russia/IPPE CORE, ORIGEN FOND-2.2 file MMK, KENO-5a
ABBN 299g (FOND-2.2 file)
Slovakia SCALE4.4/5 ENDF-BV SCALE4.4/5 ENDF-BV 44g
Spain/ENUSA (PWR) PHOENIX ENDF-BV SCALE4.3 ENDF-BV 44g
Spain/ENUSA (BWR) TGBLA ENDF-BV SCALE4.4 ENDF-BV 44g
Spain/CSN CASMO4 ENDF-BV SCALE5 ENDF-BV 238g
Spain/SEA MONTEBURNS ENDF-BVI MCNP ENDF-BVI
Sweden SCALE5 ENDF-BV/VI KENO5a ENDF-BV
Switzerland/PSI BOXER JEF1 BOXER JEF1 20g
UK/BNFL WIMS8 FISPIN
JEF2.2 WIMS8 MONK8
JEF2.2 CE
USA/Westinghouse PHOENIX ENDF-BV KENO ENDF-BV
USA SCALE4.4
Verification - Validation of the codes is usually carried out by the code developer, often in the
context of a formal Quality Assurance (QA) framework. Typically the QA programme will
include recommended procedures for optimal code utilization and the methodology for
identifying, correcting faults, notifying the user community of such faults and any resulting
corrections/improvements.
Experimental validation of the code is more commonly carried out by the users, although some differences in approach were noted. In particular, the French code scheme for BUC is developed, verified and validated by a team of specialists. Following this process a closely defined code scheme is released to the user community. No ad hoc changes to the code package are permitted; therefore, bias and associated uncertainty can be generated in a “semi-automatic” way from the validation dataset provided to the user. In other countries the responsibility for validation is entirely placed on the code user; and it is worthy of note that some regulatory bodies require the carrying out of the code validation of the user as a part of demonstrating the user’s competence.
The following conclusions were also drawn:
•
As a result of the increased validation requirement for BUC, there are significant advantages in the use of modern codes and data packages which have been verified and validated through coordinated programmes.
•
For some BUC application areas validation data is limited, so that the use of approximate methods, which may contain compensating errors can lead to significant extrapolation uncertainties.
•
It could be dangerous to use automated calculations of numerous ICSBEP benchmarks2), based on code data proposed in the Appendix of the ICSBEP evaluation, because on the one hand these code data could be wrong and on the other hand the physical analysis of the experimental information is missing.
•
Only few validation guides exist: one ANS standard (USA), the German safety code KTA 3101.2 for validation of depletion calculation codes, and the German standard DIN 25478 for validation of criticality calculation codes.
4.2.2.3. BUC validation experiments
In general PIE data is used to validate depletion calculations, and critical experiments including SNF and BUC nuclide worth measurements are used to validate the criticality calculation code. One observes that, on the contrary to criticality experiments involving fresh fuel, few PIE and criticality experiments with spent fuel exist and are available for BUC validation. Participants also noted that evidence of code accuracy may be deduced from comparisons with LWR BOC measurements (CRC – Commercial Reactor Criticals) and boron let down during the fuel cycle in PWRs, although it is recognized that there are difficulties in applying reactivity effects in a hot reactor core to an accurate derivation of code bias for spent fuel environment.
In most countries depletion calculations have been validated against PIE data from public domain (OECD SFCOMPO database) or proprietary programmes. An important exception arises in Eastern Europe where there is an acute shortage of PIE data for validation of VVER depletion calculations; however, recent Russian results from the ISTC programme just became available. The participants noticed that the useful SFCOMPO database addresses only PWR and BWR fuels; furthermore this database should be enlarged to increased enrichments and larger burnup range. These new experiments should be more documented, with a detailed irradiation history.
2) Benchmarks from the "International Handbook of Evaluated Criticality Safety Benchmark Experiments", NEA/NSC/DOC(95)03/I through VIIa, Nuclear Energy Agency, OECD Paris
Participants consider that the ICSBEP is a useful experiment database but not completely satisfying for BUC validation, due to the lack of BUC nuclide worth measurements as well as SNF experiments.
BUC nuclide worth measurements are very valuable because they allow direct validation of each BUC component. Furthermore these experiments enable actinide and Fission Product (FP) cross section improvements. Unfortunately, these BUC nuclide measurements are only performed in MINERVE reactor by oscillation of fuel samples containing one separated isotope. Some of the separated FP samples were also measured in the DIMPLE reactor in the frame of the CERES CEA-UKAEA collaborative programme. Six natural elements (Sm, Cs, Rh, Gd, Nd, and Mo containing the major BUC FP isotopes) were investigated in the Valduc Appareillage-B.
SNF worth measurements are required for the validation of BUC calculations because they supply the total burnup credit that allows the demonstration that a proposed BUC methodology is conservative. Two techniques are used in these experiments:
•
Introduction of a SNF bundle at the center of the driver core of a zero power reactor:
REBUS experiment in VENUS reactor, PWR/BWR assembly insertion at the center of PROTEUS.
•
Oscillation of SNF rod sample at the center of the MINERVE reactor: This technique allows investigating the effect of different initial enrichments and spanning the total burnup range. Moreover the SNF samples are oscillated in various spectra (such as UO2-LWR, MOX-LWR and High Conversion LWR).
4.2.2.4. Recent Improvements in International Data Files
BUC experimental programmes have heavily contributed to the improvement of international datafiles ENDF-B, JEFF and JENDL.
In the European library particularly, the nuclear data evaluation of main actinides and FPs were modified from the previous JEF2.2 to the current JEFF3.1 file:
•
U235 and Pu241 evaluations were modified with a significant +6% increase of their resonant (n,γ) cross section. Am241 evaluation was strongly modified with a +15%
augmentation of the thermal-epithermal capture. A huge work on U238 resonance range reevaluation was performed in the framework of the OECD-WPEC-sg22 Group;
a new JEFF3.1 evaluation was adopted with a 0.7% reduction of the U238 resonance integral. All these new evaluations modify the reactivity of the main BUC actinide nuclides.
•