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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.

This paper is dedicated to the strength assessment of powered axle of shunting locomotive MDD-04, manufactured for Swiss market by Bulgarian Rolling stock manufacturer Express Service Ltd. Like many other components of the newly built locomotives, the axles are also subject to strict approval procedures. Based on geometrical characteristics of the axle combined with loads acting on the axle, as defined in European standard EN 13104:2009+A2:2012, strength assessment according to calculation method defined in this standard was performed. In this way obtained safety factor values were not satisfying, so it was decided to use Finite Elements Method for validation of first calculation results. Since the results obtained with first calculation method did not meet the requirements, and with the FEM they were satisfactory, an attempt was made to approve the design only with FEM. Although the results were satisfactory, the notified body demanded a change in the original design to be made, because of approval procedures prescribed in EN 13104. The results of the new calculations were more satisfying, so the powered axle could be approved by notified body. The new design of the axle was released for commissioning and use in new produced locomotives.

For the multiple sources of error introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily to quantify the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the “two-step” method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the two-step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor and in the core power distribution was examined in the framework of phase I-3 of the OECD Uncertainty Analysis in Modeling benchmark. With the Three Mile Island Unit 1 core as a base model for analysis, the XSUSA and two-step methods were applied with certain limitations, and the results were compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions were drawn as to the method that is currently more viable for computing uncertainties in burnup and transient calculations.

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.

The Organization for Economic Cooperation and Development (OECD) Pressurized Water Reactor Main Steam Line Break (MSLB) Benchmark has been calculated for all three exercises by the coupled code system ATHLET-QUABOX/CUBBOX developed by Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). The results obtained are presented, and a detailed comparison with other solutions of the benchmark is discussed. An attempt is made to explain the differences observed in the solutions by the different modeling of physical processes in the codes. The sensitivity of results on modeling features is also investigated. In addition, the effect of different mapping schemes between fuel assemblies of the core loading and the thermal-fluid dynamics on the accuracy of three-dimensional (3-D) neutronics solutions is studied. The results for the MSLB transient are also evaluated to compare 3-D neutronics and point-kinetics solutions in view of integral and local parameters. Thus, the experiences with the coupled code system ATHLET-QUABOX/CUBBOX during the MSLB benchmark activity are summarized.

Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.