Enrico Sartori

OECD/NRC BWR Turbine Trip Transient Benchmark as a Basis for Comprehensive Qualification and Studying Best Estimate Coupled Codes

Abstract An Organisation for Economic Co-operation and Development (OECD)/U.S. Nuclear Regulatory Commission (NRC)-sponsored coupled-code benchmark has been initiated for a boiling water reactor (BWR) turbine trip (TT) transient. Turbine trip transients in a BWR are pressurization events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. In addition, the available real plant experimental data make this benchmark problem very valuable. Over the course of defining and coordinating the BWR TT benchmark, a systematic approach has been established to validate best-estimate coupled codes. This approach employs a multilevel methodology that not only allows for a consistent and comprehensive validation process but also contributes to the study of different numerical and computational aspects of coupled best-estimate simulations. This paper provides an overview of the OECD/NRC BWR TT benchmark activities with emphasis on the discussion of the numerical and computational aspects of the benchmark.

Validation of Coupled Thermal-Hydraulic and Neutronics Codes for Safety Analysis by International Cooperations

Incorporating full three-dimensional models of the reactor core into system transient codes allows for a “best-estimate” calculation of interactions between the core behavior and plant dynamics. Considerable efforts have been made in various countries and organizations on the development of coupled thermal-hydraulic and neutronics codes. Appropriate benchmarks have been developed in international cooperations led by the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) that permit testing of the neutronics–thermal-hydraulics coupling and verification of the capability of the coupled codes to analyze complex transients with coupled core-plant interactions. Three such benchmarks are presented in this paper—the OECD/U.S. Nuclear Regulatory Commission (NRC) pressurized water reactor main steam line break benchmark, the OECD/NRC boiling water reactor turbine trip benchmark, and the OECD/U.S. Department of Energy/Commissariat à l’Energie Atomique V1000 coolant transient benchmark. To meet the objectives of the validation of best-estimate coupled codes, a systematic approach has been introduced to evaluate the analyzed transients employing a multilevel methodology. Since these benchmarks are based on both code-to-code and code-to-data comparisons, further guidance for presenting and evaluating results has been developed. During the course of the benchmark activities, a professional community has been established, which allowed our carrying out in-depth discussions of different aspects considered in the validation process of the coupled codes. This positive output has certainly advanced the state of the art in the area of coupling research.

Using the OECD/NRC Pressurized Water Reactor Main Steam Line Break Benchmark to Study Current Numerical and Computational Issues of Coupled Calculations

Abstract Incorporating full three-dimensional (3-D) models of the reactor core into system transient codes allows for a “best-estimate” calculation of interactions between the core behavior and plant dynamics. Recent progress in computer technology has made the development of coupled thermal-hydraulic (T-H) and neutron kinetics code systems feasible. Considerable efforts have been made in various countries and organizations in this direction. Appropriate benchmarks need to be developed that will permit testing of two particular aspects. One is to verify the capability of the coupled codes to analyze complex transients with coupled core-plant interactions. The second is to test fully the neutronics/T-H coupling. One such benchmark is the Pressurized Water Reactor Main Steam Line Break (MSLB) Benchmark problem. It was sponsored by the Organization for Economic Cooperation and Development, U.S. Nuclear Regulatory Commission, and The Pennsylvania State University. The benchmark problem uses a 3-D neutronics core model that is based on real plant design and operational data for the Three Mile Island Unit 1 nuclear power plant. The purpose of this benchmark is threefold: to verify the capability of system codes for analyzing complex transients with coupled core-plant interactions; to test fully the 3-D neutronics/T-H coupling; and to evaluate discrepancies among the predictions of coupled codes in best-estimate transient simulations. The purposes of the benchmark are met through the application of three exercises: a point kinetics plant simulation (exercise 1), a coupled 3-D neutronics/core T-H evaluation of core response (exercise 2), and a best-estimate coupled core-plant transient model (exercise 3). In this paper we present the three exercises of the MSLB benchmark, and we summarize the findings of the participants with regard to the current numerical and computational issues of coupled calculations. In addition, this paper reviews in some detail the sensitivity studies on exercises 2 and 3 performed by the benchmark team using the coupled code TRAC-PF1/NEM. The purpose of these supporting studies was to aid participants in developing their models.

Plutonium management in the medium term

Abstract For many years various countries with access to commercial reprocessing services have been routinely recycling plutonium as UO2/PuO2 mixed oxide (MOX) fuel in light water reactors (LWRs). This LWR MOX recycle strategy is still widely regarded as an interim step leading to the eventual establishment of sustainable fast reactor fuel cycles. The OECD/NEA Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles (WPPR) has recently completed a review of the technical options for plutonium management in what it refers to as the “medium term.” For the purpose of the review, the WPPR considers the medium term to cover the period from now up to the point at which fast reactor fuel cycles are established on a commercial scale. The review identified a number of different designs of innovative plutonium fuel assemblies intended to be used in current LWR cores, in LWRs with significantly different moderation properties, as well as in high-temperature gas reactors. The full review report describes these various options and highlights their respective advantages and disadvantages. This paper briefly summarizes the main findings of the review.

Organization for economic cooperation and development/Nuclear Energy Agency international benchmark on the VENUS-2 MOX core measurements

Abstract Within the framework of the Nuclear Energy Agency Nuclear Science Committee, theoretical physics benchmarks and multiple recycling issues related to various mixed-oxide (MOX)–fueled systems have been studied. Many improvements and clarifications in nuclear data libraries and calculation methods have been achieved from the results of the theoretical benchmarks performed. However, it was felt that there was also a need to link these findings to data from experiments. Hence, a blind international benchmark exercise based on the two-dimensional VENUS-2 MOX core measurement data was carried out. Twelve participants from ten countries participated in the benchmark. Both the deterministic and the Monte Carlo methods were applied with different nuclear data sets. This technical note provides a comparative analysis between calculated and measured results. Comparison with experimental results identified the origins of discrepancies between calculations and measurements and enabled the quantitative comparison of the relative merits of the different calculation methods.

The Physics of Plutonium Fuels - A Review of Organization for Economic Cooperation and Development/Nuclear Energy Agency Activities

In 1993, the Organization for Economic Cooperation and Development/Nuclear Energy Agency first convened the Working Group on the Physics of Plutonium Recycle (WPPR) (now renamed the Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles). Since its inception, the WPPR (whose task has now been expanded to include innovative fuel cycles) has published six volumes of detailed results from analyses of plutonium fuel in pressurized water reactors and fast reactors. A seventh volume on the physics of plutonium fuel in boiling water reactors is in preparation. The analyses have been mostly in the form of theoretical benchmark exercises for situations beyond current experience, for which multinational contributions provide a basis for comparison of diverse calculational methods and nuclear data libraries. The overall activities of the WPPR are reviewed and summarized.