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Within the framework of the IAEA Coordinated Research Project on HTGR Uncertainty Analysis in Modeling, criticality calculations of the Very High Temperature Critical Assembly experiment were performed as the validation reference to the prismatic MHTGR-350 lattice calculations. Criticality measurements performed at several temperature points at this Japanese graphite-moderated facility were recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and represent one of the few data sets available for the validation of HTGR lattice physics. This work compares VHTRC criticality simulations utilizing the Monte Carlo codes Serpent and SCALE/KENO-VI. Reasonable agreement was found between Serpent and KENO-VI, but only the use of the latest ENDF cross section library release, namely the ENDF/B-VII.1 library, led to an improved match with the measured data. Furthermore, the fourth beta release of SCALE 6.2/KENO-VI showed significant improvements from the current SCALE 6.1.2 version, compared to the experimental values and Serpent.

The model implemented into coupled neutron-thermohydraulic code system ATHLET + QUABOX/CUBBOX for RBMK-1000 vapor reactivity coefficient measurements computer modeling is described. The results of ATHLET + QUABOX/CUBBOX modeling of Kursk NPP unit 1 vapor reactivity coefficient measurements are presented. Reasonable agreement between measurements and modeling results was shown.

The development of multi-physics multi-scale coupled methodologies for Light Water Reactor (LWR) analysis requires comprehensive validation and verification procedures, which include well-established benchmarks developed in international cooperation. The Nuclear Energy Agency (NEA) of the Organization for Economic Co-operation and Development (OECD) has provided such framework, and over the years a number of LWR benchmarks have been developed and successfully conducted. The first set of NEA/OECD benchmarks that permits testing of the neutronics/thermal–hydraulics coupling, and verifying the capability of the coupled codes to analyze complex transients with coupled core/plant interactions have been completed and documented. These benchmarks provided a validation basis for the new generation of coupled “best-estimate” codes. The above mentioned OECD/NEA LWR benchmark activities have also stimulated follow up developments and benchmarks to test these developments. The models utilized have been improved when moving from one benchmark to the next and this created a need to validate them using high-quality experimental data. Second set of the NEA/OECD benchmarks have been initiated by the Expert Group on Uncertainty Analysis in Modelling (EGUAM) at the Nuclear Science Committee (NSC), NEA/OECD to address the current trends in the development of LWR multi-physics and multi-scale modeling and simulation. These benchmarks include the following common features, which address some of the issues identified in the first set of OECD/NEA benchmarks: (a) utilization of high-quality experimental data; (b) refined local scale modeling in addition to global predictions; (c) more detailed comparisons and analysis; (d) including uncertainty and sensitivity analysis of modeling predictions. The paper presents each of these new benchmarks by providing description and discussion of comparative analysis of obtained results. Special attention is devoted to uncertainty propagation in LWR multi-physics and multi-scale simulations for design and safety evaluations.

One of the most interesting technologies recognized as promising for the future of nuclear energy is the Accelerator Driven System (ADS). The purpose of the ADS is to function as an element of an integrated nuclear power enterprise comprising of conventional and advanced power reactors for energy production, and for reducing the radiotoxicity of the nuclear waste produced by these power reactors before entombment in a geologic repository ( Stanculescu, 2000 ). The use of new types of coolants such as molten lead or lead–bismuth eutectic alloy (LBE) represents another particular feature in the design concept of these reactors, since it permits one to take advantage of higher Heavy Liquid Metals (HLM) boiling temperature compared to water, leading to an improvement in thermal efficiency. Furthermore, it allows the reactor to be operated at a lower pressure, thus reducing the probability of a Loss-Of-Coolant Accident (LOCA) and, consequently, increasing reactor safety. The proposed use of relatively new types of coolants, especially in combination with new fuel types and cladding materials, demands specific attention to the thermal–hydraulics and core mechanics in normal and abnormal conditions. For that reason, a detailed analysis using system codes is necessary. In this paper, a new version of the ATHLET system code is tested, in which the physical properties of liquid metals like sodium, lead and LBE are implemented. The object of this study is the spallation loop of the MYRRHA facility, in which LBE is circulated by forced convection to remove the heat deposited by a proton beam. A detailed nodalization is set up for performing thermal–hydraulic calculations for both nominal conditions and accident scenarios in order to have a good characterization of the entire loop. The start-up transient has verified the correct removal of the heat generated in the target by the foreseen heat exchanger. During accidental transients, it was noted that the level of LBE in the beam line changes in agreement with preliminary studies on the target device. The pump failure test represents the most dangerous scenario, as the temperature in the target area will reach a very high value within a few seconds after the blockage of the pump impeller. The results obtained have subsequently been compared to the ones achieved with previous numerical simulations which were performed using a version of RELAP5/mod 3.3 system code that was modified at the University of Pisa to account for LBE properties. The comparison has confirmed the capability of the new version of the ATHLET code for the analysis of hydraulic circuits cooled by liquid metals, although both codes use different heat transfer correlations.

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 a 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.