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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%.

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.

At present time, Ukraine faces the problem of small margins of acceptance criteria in connection with the implementation of a conservative approach for safety evaluations. The problem is particularly topical conducting feasibility analysis of power up-rating for Ukrainian nuclear power plants. Such situation requires the implementation of a best-estimate approach on the basis of an uncertainty analysis. For some kind of accidents, such as loss-of-coolant accident (LOCA), the best estimate approach is, more or less, developed and established. However, for reactivity initiated accident (RIA) analysis an application of best estimate method could be problematical. A regulatory document in Ukraine defines a nomenclature of neutronics calculations and so called “generic safety parameters” which should be used as boundary conditions for all VVER-1000 (V-320) reactors in RIA analysis. In this paper the ideas of uncertainty evaluations of generic safety parameters in RIA analysis in connection with the use of the 3D neutron kinetic code DYN3D and the GRS SUSA approach are presented.

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.

The article provides an overview of the reactor dynamics code DYN3D. The code comprises various 3D neutron kinetics solvers, a thermal-hydraulics reactor core model and a thermo-mechanical fuel rod model. The implemented models and methods and the capabilities and features of the code are described. Latest developments of models and methods are delineated. An overview on the status of verification and validation is given. Code applications for selected safety analyses are described. Furthermore, multi-physics code couplings to thermal-hydraulic system codes, CFD and sub-channel codes as well as to the fuel performance code TRANSURANUS are outlined. Developments for innovative reactor concepts, in particular Molten Salt Reactor, High Temperature Gas-cooled Reactor and Sodium Fast Reactor are delineated. The management of code maintenance is briefly described. An outlook on further code development is given.

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.

This research focuses on the verification of a recently developed methodology accounting for spectral history effects in 3D full core nodal simulations. The traditional deterministic core simulation procedure includes two stages: (1) generation of homogenized macroscopic cross section sets and (2) application of these sets to obtain a full 3D core solution with nodal codes. The standard approach adopts the branch methodology in which the branches represent all expected combinations of operational conditions as a function of burnup (main branch). The main branch is produced for constant, usually averaged, operating conditions (e.g. coolant density). As a result, the spectral history effects that associated with coolant density variation are not taken into account properly. Number of methods to solve this problem (such as micro-depletion and spectral indexes) were developed and implemented in modern nodal codes. Recently, we proposed a new and robust method to account for history effects. The methodology was implemented in DYN3D and involves modification of the few-group cross section sets. The method utilizes the local Pu-239 concentration as an indicator of spectral history. The method was verified for PWR and VVER applications. However, the spectrum variation in BWR core is more pronounced due to the stronger coolant density change. The purpose of the current work is investigating the applicability of the method to BWR analysis. The proposed methodology was verified against recently developed BGCore system, which couples Monte Carlo neutron transport with depletion and thermal-hydraulic solvers and thus capable of providing a reference solution for 3D simulations. The results clearly show that neglecting the spectral history effects leads to a very large deviation (e.g. 1700 pcm in multiplication factor) from the reference solution. Application of the Pu-correction method results in a very good agreement between DYN3D and BGCore on the order of 200 pcm in k inf .

The article provides an overview on the code DYN3D which is a three-dimensional core model for steady-state, dynamic and depletion calculations in reactor cores with quadratic or hexagonal fuel assembly geometry being developed by the Helmholtz-Zentrum Dresden-Rossendorf for more than 20 years. The current paper gives an overview on the basic DYN3D models and the available code couplings. The verification and validation status is shortly outlined. The paper concludes with the current developments of the DYN3D code. For more detailed information the reader is referred to the publications cited in the corresponding chapters.

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.

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.

The reactor dynamics code DYN3D is a three-dimensional best-estimate tool for simulating steady states and transients of light-water reactors and innovative reactor designs. An overview of the DYN3D features is provided. This paper further focuses on the recently developed trigonal-geometry diffusion model DYN3D-TRIDIF including a description of the underlying nodal approach and the characteristics of trigonal geometries. Via a mesh refinement study by means of a VVER-1000-type core benchmark using a fine-mesh diffusion reference solution, DYN3D-TRIDIF shows spatial convergence. Furthermore, the performance of DYN3D-TRIDIF is verified by means of a single-assembly problem on pin-cell level. Good agreement between DYN3D-TRIDIF and the detailed-geometry transport reference is achieved with an average deviation in power of less than 1%.

The few-group cross section libraries, used by reactor dynamics codes, are affected by the spectral history effect – a dependence of fuel cross sections not only on burnup, but also on local spectral conditions during burnup. Neglecting this effect leads to an additional component of error in neutron-physical characteristics. Two solution approaches to this problem implemented in the reactor dynamic code DYN3D are described and compared in this paper: a cross section correction method based on239Pu concentration and separate cross sections treatment for each axial layer of reactor core. Steady-state and burnup characteristics of a PWR and a WWER-1000 cores, calculated by DYN3D with and without cross section corrections, are compared. An impact of the correction on transient calculations is studied for a control rod ejection example. Studies have shown a significant influence of spectral history on axial power and burnup distributions as well as on calculated cycle length. Two different correction methods have shown similar major effects.

The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.