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Recent developments of the reactor dynamics code DYN3D have introduced the micro-depletion model which allows for explicit calculation of radioactive decay heat. Such a unique combination of nodal diffusion, thermal hydraulic (T/H) and depletion solvers allows DYN3D to perform fuel cycle depletion and obtain detailed core isotopic concentration and decay heat distributions. More importantly, this capability allows for the generation of time-dependent spatial decay heat distribution, which is essential to predict the behavior of the reactor under off-nominal scenarios. The new sequence utilizes considerably less computational resources than coupled Monte Carlo-T/H-depletion systems, but with comparable accuracy. This capability was recently tested on a limited number of simple unit cell models. The main objective of this work is to further verify the decay heat calculation capabilities of DYN3D by applying it to a considerably more realistic and detailed full core model. For the purpose of the current analysis a 3D full core model of Advanced Burner Reactor (ABR) was adopted from the OECD/NEA Benchmark for Neutronic Analysis of Sodium-Cooled Fast Reactor Cores with Various Fuel Types and Core Sizes. In this work, the Monte Carlo code Serpent was used to generate macro- and microscopic parameters, and the neutron diffusion code, DYN3D, was used to perform neutronic and depletion analyses. Detailed spatial isotopic and decay heat distributions obtained with DYN3D were verified against the equivalent Serpent reference 3D full core solution. Results indicate very good agreement between the Serpent-DYN3D code sequence and the reference Serpent solutions, with a discrepancy in total decay heat on the order of 0.5%.

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.

The paper presents a general hybrid method that combines the micro-depletion technique with correction of micro- and macro-diffusion parameters to account for the spectral history effects. The fuel in a core is subjected to time- and space-dependent operational conditions (e.g. coolant density), which cannot be predicted in advance. However, lattice codes assume some average conditions to generate cross sections (XS) for nodal diffusion codes such as DYN3D. Deviation of local operational history from average conditions leads to accumulation of errors in XS, which is referred as spectral history effects. Various methods to account for the spectral history effects, such as spectral index, burnup-averaged operational parameters and micro-depletion, were implemented in some nodal codes. Recently, an alternative method, which characterizes fuel depletion state by burnup and 239Pu concentration (denoted as Pu-correction) was proposed, implemented in nodal code DYN3D and verified for a wide range of history effects. The method is computationally efficient, however, it has applicability limitations. The current study seeks to improve the accuracy and applicability range of Pu-correction method. The proposed hybrid method combines the micro-depletion method with a XS characterization technique similar to the Pu-correction method. The method was implemented in DYN3D and verified on multiple test cases. The results obtained with DYN3D were compared to those obtained with Monte Carlo code Serpent, which was also used to generate the XS. The observed differences are within the statistical uncertainties.

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.

Аналізуючи ядерну безпеку сховищ відпрацьованого ядерного палива з урахуванням вигоряння палива (burnup credit), необхідно враховувати розподіл вигоряння по довжині касети. Найпростіше та водночас консервативно можна прийняти вигоряння по довжині відпрацьованих ТВЗ однаковим й рівним середньому вигорянню на найменш вигорілих кінцевих ділянках. Однак це призводить до заниження вигоряння по касеті порівняно з реальним середнім значенням у 1,5—2,5 раза. Для зняття надлишкового консерватизму, закладеного в такому підході, пропонується метод консервативного врахування аксіального профілю розподілу вигоряння по довжині ТВЗ. Заснований на результатах аналізу аксіальних профілів розподілу вигоряння відпрацьованих ТВЗ, цей метод може використовуватися в обґрунтуванні ядерної безпеки систем поводження й зберігання відпрацьованого палива.

Як показав аналіз ядерної безпеки основних систем поводження з відпрацьованим паливом, не всі вони задовольняють вимогам нормативних документів. Проте подібну невідповідність не слід розглядати тільки як недоліки систем. Значною мірою це наслідок надлишкового консерватизму, закладеного в нормативних вимогах, який, в свою чергу, є наслідком недостатнього розвитку засобів моделювання процесів у ядерному паливі, може бути значно зменшений за рахунок більш точного моделювання паливних систем без зниження рівня ядерної безпеки цих систем.