Expert group on BUC criticality

The main objective of the activities of the OECD/NEA Expert Group on BUC Criticality is to demonstrate that the available criticality safety calculational tools are appropriate for application to burned fuel systems and that a reasonable safety margin can be established. For this purpose the Expert Group established a suite of BUC criticality benchmarks that assess the capability to calculate both spent fuel composition and reactivity of spent fuel.

Nuclear Science Committee

WP on Nuclear Criticality

Safety

WP on International Evaluation

Co-operation

ICSBEP Project

Sub-Critical Measurements

Experimental Needs

Minimum Critical Values

Burn-up Credit Studies WP on the Physics of

Plutonium Fuels and Innovative Fuel Cycles

Figure 1. Existing relationship between working parties reporting to the OECD/NEA Nuclear Science Committee and the criticality safety expert groups

The benchmarks were carefully specified to allow a comparison of results using a wide variety of calculational tools and nuclear data sets. Throughout the tenure of the activities of the Expert Group on BUC Criticality, experts from 17 countries (Belgium, Canada, Czech Republic, Denmark, Finland, France, Germany, Hungary, Italy, Japan, Korea, The Netherlands, Spain, Sweden, Switzerland, United Kingdom and the United States) have participated in various phases of the benchmark exercises. Participants used a wide variety of codes and methods based on transport theory, using SN, nodal and Monte Carlo techniques.

Nuclear data (both cross-section and decay data) were taken from a variety of sources:

multiple versions of the Evaluated Nuclear Data Files (ENDF/B), the Japan Evaluated Nuclear Data Libraries (JENDL) and the Joint Evaluated Files (JEF). Both multi-group and continuous energy cross-section data were used in the study.

Table I is a summary of the benchmark problems addressed noting both the primary objective and current status of each.

Table I. Summary of benchmark problems addressed by the OECD/NEA Expert Group on BUC Criticality

Benchmark Primary Objective Status

Phase I-A Examine effects of seven major actinides and 15 major fission products for an infinite array of PWR rods. Isotopic composition specified at 3.6 wt.% U-235 at 0, 30 and 40 GWd/MTU and at

one- and five-year cooled.

Completed (Ref. 1)

Phase I-B Compare computed nuclide concentrations for depletion in a simple PWR pin-cell model, comparison to actual

measurements at three burnups (27.34, 37.12 and 44.34 GWd/MTU).

Completed (Ref. 2)

Phase II-A Examine effect of axially distributed burnup in an array of PWR pins as a function of initial enrichment, burnup and cooling time. Effects of fission products independently examined.

Completed (Ref. 3)

Phase II-B Repeat study of Phase II-A in 3-D geometry representative of a conceptual BUC transportation container. Isotopic

compositions specified.

Completed (Ref. 4) Phase II-C Key sensitivities in criticality safety to burnup profiles. In progress Phase III-A Investigate the effects of moderator void distribution in

addition to burnup profile, initial enrichment, burnup and cooling time sensitivities for an array of BWR pins.

Completed (Ref. 5) Phase III-B Compare computed nuclide concentrations for depletion in a

BWR pin-cell model.

Completed (Ref. 6) Phase IV-A Investigate BUC for MOX spent fuel pin-cell for three

plutonium vectors (first recycle, fifth recycle, weapons-grade)

Final Draft Approved 2002 Phase IV-B Compare computed nuclide concentrations for depletion in a

MOX super-cell. In progress

Phase V WWER BUC. Similar to Phases I and II for PWRs but with hexagonal geometry and WWER fuel specification

Independent/

Parallel Study Phase I and Phase II included both criticality and depletion benchmarks for pressurized water reactors (PWRs). A set of selected nuclides including 7 major actinides (U-234, 235, 236, and 238; Pu-239, 240 and 241), 5 minor actinides (Pu-238 and 242; Am-241 and 243; Np-237) and 15 fission products (Mo-95; Tc-99; Ru-101; Rh-103; Ag-109; Cs-133; Sm-147, 149, 150, 151 and 152; Nd-143 and 145; Eu-153; and Gd-155) were used in these studies. The results showed no trends in standard deviation among participants with burnup or cooling time in the criticality analyses. Consistently the largest deviations among participants were for the fresh fuel cases. In the depletion analyses, there was evidence of a significant trend in the standard deviation among participants for the residual U-235 (the trend was small for most other isotopes). A number of nuclides have been identified for additional study based on the sensitivity of k to the observed standard deviations: Pu-239, Gd-155, U235, Pu-241, Pu-240 and Sm-151. Much of the differences are assumed to be in the basic nuclear data. Both 2-D

and 3-D models have been used to evaluate the impact of axially distributed burnup. It was determined that 70% of the total fissions occur in the upper 40cm of fuel that illustrates the potential importance of this parameter. Good agreement was seen among the participants relative to the calculated “end effect”. It has been noted by the group that the effect on k is strongly a function of the system being evaluated and may be even more important under postulated accident conditions that result in axial heterogeneity. Two remaining issues associated with the axial effect continue to be investigated in the expert group: (1) limited availability of measured axial profile data and detailed power history data in the open literature, and (2) defining/performing analyses to determine the sensitivities due to different axial burnup profiles across the full range of burnups.

Phase III included both criticality and depletion benchmarks for boiling water reactors (BWRs). For the most part the results are consistent with those for PWRs: the largest deviations among participants are for the fresh fuel cases, and deviations are higher for distributed burnups versus modeling the average burnup. Larger void fractions (i.e., use of a 70% uniform void distribution) tended to increase the deviation among participants. The complex geometry of the BWR fuel assemblies added complexity to the depletion calculation.

These results are in final review and should be published shortly.

Mixed oxide (MOX) fuels in PWRs were investigated in Phase IV. The problems included a MOX pincell calculation to identify sensitivities specific to MOX. The primary result of this early benchmark was to identify the need to include curium isotopes in both the criticality and depletion calculations, as Cm contributes up to 1.5% in k. The reports for these analyses are in preparation/review and should be forthcoming this year. Phase V is a completely parallel study being led by L. Markova and addressed in another presentation at this meeting.

Since the objective of the Expert Group on BUC Criticality thus far has been to assess code capabilities, the results are most often presented as the standard deviation among participants.

There has been no attempt to make a safety case for licensing or to provide bounding values on the observed trends or physical phenomena (e.g. the effect of axially distributed burnup).

However, the group does discuss specific or suspected sources of discrepancies, leading to the identification of further studies.

Information about the current activities and links to publications of the OECD/NEA Expert Group on BUC Criticality may be found at http://www.nea.fr/html/science/buc. The OECD/NEA Secretariat for this work is Mr. Ali Nouri (Ali.NOURI@oecd.org) who may be contacted for additional information.