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This study presents an approach to the selection of optimal energy group structures for multi-group nodal diffusion analyses of Sodium-cooled Fast Reactor cores. The goal is to speed up calculations, particularly in transient calculations, while maintaining an acceptable accuracy of the results. In Part I of the paper, possible time-savings due to collapsing of energy groups are evaluated using 24-group energy structure as a reference. Afterwards, focusing on energy structures with a number of groups leading to significant calculation speedups, optimal grid configurations are identified. Depending on a number of possible energy grid configurations to explore, the optimization is conducted by either a direct search or applying the simulated annealing method. Speedup and optimization studies are performed on a selected case of the Superphenix static neutronic benchmark by using the nodal diffusion DYN3D code. The results demonstrate noticeable improvements in DYN3D performance with a marginal deterioration of the accuracy. (C) 2021 Elsevier Ltd. All rights reserved.

An increasing interest on the development of highly accurate methodologies in reactor physics is nowadays observed, mainly stimulated by the availability of vast computational resources. As a result, an on-going development of a wide range of coupled calculation tools is observed within diverse projects worldwide. Under this framework, the McSAFE European Union project is a coordinated effort aimed to develop multiphysics tools based on Monte Carlo neutron transport and subchannel thermal-hydraulics codes. These tools are aimed to be suitable for high-fidelity calculations both for PWR and VVER reactors, with the final goal of performing pin-by-pin coupled calculations at full core scope including burnup. Several intermediate steps are to be analyzed in-depth before jumping into this final goal in order to provide insights and to identify resources requirements. As part of this process, this work presents the results for a pin-by-pin coupling calculation using the Serpent 2 code (developed by VTT, Finland) and the subchannel code SUBCHANFLOW (SCF, developed by KIT, Germany) for a full-core VVER model. For such purpose, a recently refurbished master-slave coupling scheme is used within a High Performance Computing architecture. A full-core benchmark for a VVER-1000 that provides experimental data is considered, where the first burnup step (i.e. fresh core at hot-full rated power state) is calculated. For such purpose a detailed (i.e. pin-by-pin) coupled Serpent-SCF model is developed, including a simplified equilibrium xenon distribution (i.e. by fuel assembly). Comparisons with main global reported results are presented and briefly discussed, together with a raw estimation of resources requirements and a brief demonstration of the inherent capabilities of the proposed approach. The results presented here provide valuable insights and pave the way to tackle the final goals of the on-going high-fidelity project.

The X2 VVER-1000 benchmark provides a unique set of the operational data of a VVER-1000 reactor. This includes fresh core hot zero power (HZP) experiments, operational history of first four fuel cycles, and information on the operational transients occurred on the unit during first cycles. Since a publication of the initial versions of the benchmark, numerous updates, corrections and refinements become available. The current paper is a first in a series of publications on the revised X2 VVER-1000 benchmark. It is dedicated to the fresh core HZP experiments and includes description of the fuel and core geometries, the material compositions, description and results of the measurements taken during fresh core start-up. In addition, the paper includes the reference Monte Carlo solution for the HZP experiments obtained with Serpent 2. The calculated and measured values are in a good agreement. Further extension of the benchmark definition is foreseen in the near future.

In order to compensate the excess reactivity over the fuel cycle in pressurized water reactors, boric acid is added to the reactor coolant. During certain accident conditions, the formation and the subsequent transportation towards the reactor core of a slug of coolant with reduced boron concentration can lead to a positive reactivity insertion. In this paper, a new approach dealing with such heterogeneous boron dilution scenarios is presented. This approach is based on the use of computational fluid dynamics methods for the whole reactor pressure vessel with direct coupling to a 3D neutron kinetic model of the reactor core. The application of this approach is demonstrated on a main coolant pump start scenario for hot-shutdown conditions using the coupled ANSYS CFX/DYN3D code. The calculation results show the enormous influence of the coolant mixing on the reactivity behavior of the reactor core.

This paper describes the derivation of the nodal flux expansion method HEXNEM3, its implementation into the nodal diffusion code DYN3D and the corresponding testing versus benchmarks. As in the earlier versions of expansion method HEXNEM1 and HEXNEM2, the neutron flux in a hexagonal node is expanded into superposition of orthogonal polynomials and exponential functions. The main difference of the HEXNEM3 method is the additional use of tangentially weighted exponential functions and the coupling of neighboring nodes by tangentially weighted fluxes and currents on node surfaces. The HEXNEM3 method was tested in several benchmark problems, including numerical benchmarks with given cross sections set and reference solutions by fine-mesh finite difference diffusion and a real plant benchmark with Monte Carlo reference solution. The test results demonstrate good agreement with reference solutions and improvement of method accuracy in comparison with HEXNEM1 and HEXNEM2.

The residual radioactive decay heat plays an important role in some accident scenarios and, therefore, needs to be accurately calculated when performing accident analyses. The current reactor simulation codes used for accident analysis account for the residual decay heat by means of simplified models. Typically, these models rely on semi-empirical correlations which are defined over a limited range of burnup and fuel types. Therefore, the applicability of such correlations is limited and any deviation from the definition range may lead to high uncertainties, which is detrimental in evaluating safety margins. Reactor dynamic code DYN3D was originally developed for transient and accident analysis of LWRs. In DYN3D, the residual radioactive decay heat calculation is based on the German national standard DIN Norm 25463 model. The applicability of this model is limited to a low enriched uranium dioxide fuel for light water reactors. This paper describes a new general decay heat calculation model implemented in DYN3D. The radioactive decay rate of each nuclide in each spatial node is calculated by recently implemented depletion module and the cumulative released heat is used to obtain the spatial distribution of the decay power for every time step. Such explicit approach is based on first principles and is free from approximations and, thus, can be applied to any reactor system (e.g. thermal and fast) and fuel type. The proposed method is verified through code-to-code comparison with the Serpent 2 Monte Carlo code results.

The use of best-estimate approach for WWER safety analysis in RIA is considered. The relevance of this problem is concerned with small margin to acceptance criteria under the conservative approach and becomes stronger under power uprate of nuclear power plants. Previous experience in this area for WWER-1000 reactor types is overviewed. The necessity to extend these activities for successful implementation of the best-estimate approach is noticed and areas of further work are discussed. Рассмотрены вопросы использования и реализации подходов наилучшей оценки для анализа безопасности ВВЭР в реактивностных авариях. Актуальность проблемы связана с малыми запасами до критериев приемлемости при реализации консервативного подхода, особенно усиливается в условиях повышения номинального уровня мощности реакторной установки. Представлен краткий обзор предыдущего опыта в этой области для реакторов ВВЭР-1000. Указана необходимость расширения работ для успешной реализации подходов наилучшей оценки, обсуждены направления дальнейшей деятельности. Розглянуто питання використання та реалізації підходів найкращої оцінки для аналізу безпеки ВВЕР у реактивнісних аваріях. Актуальність проблеми пов’язана з малими запасами до критеріїв прийнятності при реалізації консервативного підходу, що особливо посилюється в умовах підвищення номінального рівня потужності реакторної установки. Представлено короткий огляд попереднього досвіду в цій галузі для реакторів ВВЕР-1000. Зазначено необхідність розширення робіт для успішної реалізації підходів найкращої оцінки, обговорено напрями подальшої діяльності.

High conversion LWRs concepts typically rely on a heterogeneous core configuration, where fissile zones are interspersed with fertile blanket zones in order to achieve a high conversion ratio. Modeling such a heterogeneous structure of these cores represents a significant challenge to the conventional reactor analysis methods. It was recently suggested to overcome such difficulties, in particular, for the case of axially heterogeneous reduced moderation BWRs, by introducing an additional set of discontinuity factors in axial direction at the interfaces between fissile and fertile fuel assembly zones. However, none of the existing nodal diffusion core simulators have the capability of accounting for discontinuity of homogeneous nodal fluxes in axial direction since the fuel composition of conventional LWRs is much more axially uniform. In this work, we modified the nodal diffusion code DYN3D by introducing such a capability. The new version of the code was tested on a series of reduced moderation BWR cases with Th–U233 and U–Pu–MA fuel. The library of few-group homogenized cross sections and the data required for the calculation of discontinuity factors were generated using the Monte Carlo transport code Serpent. The results obtained with the modified version of DYN3D were compared with the reference Monte Carlo solutions and were found to be in good agreement. The current analysis demonstrates that high conversion LWRs can in principle be modeled using existing nodal diffusion core simulators.