Effect of mesh grid size on the accuracy of deterministic VVER-1000 core calculations

Numerical solutions based on finite-difference method require the domain in the problem to be divided into a number of nodes in the form of triangles, rectangular, and so on. To apply the finite-difference method in reactor physics for solving the diffusion equation with satisfactory accuracy, the distance between adjacent mesh-points should be small in comparison with a neutron mean free path. In this regard the effect of number of mesh points on the accuracy and computation time have been investigated using the VVER-1000 reactor of Bushehr NPP as an example, and utilizing WIMS and CITATION codes. The best results obtained in this study belong to meshing models with higher numbers of mesh-points in both radial and axial directions of the reactor core.     

HEXTRAN–SMABRE calculation of the VVER-1000 coolant transient-1 benchmark

The VVER-1000 coolant transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three-dimensional neutron kinetic core models. It concerns switching on a main coolant pump when the other three main coolant pumps are in operation. The problem is based on an experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, an extreme scenario concerning a control rod ejection after switching on a main coolant pump was calculated. At VTT the three-dimensional advanced nodal code HEXTRAN is used for the core dynamics, and the system code SMABRE as a thermal hydraulic model for the primary and secondary loop. The parallelly coupled HEXTRAN–SMABRE code has been in production use since early 1990s, and it has been extensively used for analyses of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used at VTT. The whole core calculation is performed with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation were specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Parametric studies have been performed for selected parameters.     

VVER-1000型反应堆压力容器热老化分析评估

本文系统介绍了VVER-1000型反应堆压力容器(RPV)的温度监督情况,针对田湾核电站1#机组RPV的温度监督测试结果进行分析,评价运行3年后RPV力学性能(包括拉伸、冲击、断裂韧性)变化行为及热老化脆化机理,评估了当前田湾RPV服役运行后的热老化脆化状态和温度监督的时间安排。结果表明,温度监督样品经过堆内高温环境考验后,焊缝材料表现出一定程度的脆化特征,但母材、热影响区脆化不明显。与康采恩模型的结果和俄罗斯数据相比较后,认为田湾核电站1#机组RPV热老化脆化情况在合理范围内。     

VVER-1000型机组正常运行气液态流出物源项计算

田湾3、4号机组采用俄方设计制造的VVER-1000型反应堆,其正常运行气液态流出物排放源项是检验核电厂设计是否满足国家相关环境标准的重要指标,是辐射防护最优化设计的重要内容之一。以我国压水堆核电厂源项框架体系为基础,通过分析田湾核电站相关工艺系统流程,选取合适的工艺回路部件数学模型,采用电厂设计以及实际运行经验参数,分别计算了设计与现实排放源项,并与俄方计算结果进行对比,说明采用新源项框架体系下气液态放射性流出物的变化情况。     

Modelling effects and their impact on the results of ATHLET solutions of Exercise 1 of the VVER-1000 Coolant Transient Benchmark

The VVER-1000 Coolant Transient Benchmark consists of two phases and refers to experimental data from the Kozloduy Unit 6 Nuclear Power Plant in Bulgaria. The paper describes the modelling features and their impact on the results of the Exercise 1, Phase 1 of the Benchmark obtained by two ATHLET user groups, namely GRS and NRI. The simulated transient is a main coolant pump (MCP) switching on in one loop at reduced power while three other MCPs are in operation. Particular attention is paid to the influence of the reactor vessel modelling and especially of the nodalization in the upper plenum. The comparison and discussion of the two simulation results confirm that the two solutions with the ATHLET system code achieve quite good system response of the plant transient.     

Neutronic study of nanofluids application to VVER-1000

The change in neutronic parameters of the VVER-1000 nuclear reactor core attributable to the use of nanoparticle/water (nanofluid) as coolant is presented in this paper. Optimization of type and volume fraction of nanoparticles in water that affect the safety enhancement of core primary parameters is intended in this study. Reactivity change, radial and axial local peaking factors (LPPF), and the consequence of nanoparticle deposition on fuel clad are investigated.     

REA analysis of the Iranian VVER-1000 core by modification of PRORIA code for annular pellets

The PRORIA code and its recent modifications are described here. The PRORIA code analyzes the transient response of the core against the reactivity increase caused by the control rod rapid withdrawal. The code solves and analyzes neutronic and thermal–hydraulic equations simultaneously. The code is designed for western PWR-type reactor performance. The equations representing thermal–hydraulic and neutronic should be modified to use the code to analyze VVER-1000 reactor core transients, because The VVER-1000 reactor fuel has a central hole in the fuel pellet. In a cylindrical solid fuel pellet, operation of an oxide fuel material at high temperature alters its morphology and the inner region is restructured to form a void at the center surrounded by a dense fuel region. Most of the restructuring occurs within the first few days of operation with slow changes afterward. Hence, the effects of a central hole in mathematical equations and in the transient are investigated. After the code modification, three accident scenarios with control rod ejection are simulated. The results are in good agreement with those reported in the plant’s FSAR. The results show that the peak fuel temperature in the hot fuel pin is lower than what the original code predicts by 150–500 °C. Furthermore, the Doppler reactivity effect, when the fuel pellet has a central hole, is higher than the solid fuel pellet.     

Simulation of the VVER-1000 pump start-up experiment of the OECD V1000CT benchmark with CATHARE and TRAC-PF1

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark for VVER-1000 was defined. The benchmark consists of calculation of a pump start-up experiment labelled V1000CT-1 (Phase 1), as well as a vessel mixing experiment and main steam line break (MSLB) transient labelled V1000CT-2 (Phase 2), respectively. The reference nuclear plant is Kozloduy-6 in Bulgaria. The overall objective is to assess computer codes used in the analysis of VVER-1000 reactivity transients. A specific objective is to assess the vessel mixing models used in system codes. Plant data are available for code validation consisting of one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The validated codes can be used to calculate asymmetric MSLB transients involving similar mixing patterns. This paper summarizes a comparison of CATHARE and TRAC-PF1 system code results for V1000CT-1, Exercise 1, which is a full plant point kinetics simulation of a reactor coolant system (RCS) pump start-up experiment. The reference plant data include integral and sector average parameters. The comparison is made from the point of view of vessel mixing and full system simulation. CATHARE used a six-sector multiple 1D vessel thermal-hydraulic model with cross flows and TRAC used a six-sector, 18-channel coarse-mesh 3D vessel model. Good agreement in terms of integral parameters and inter-loop mixing is observed.     

AFEN method and its solutions of the hexagonal three-dimensional VVER-1000 benchmark problem

The unique features of the analytic function expansion nodal (AFEN) method in hexagonal-z geometry are described. The COREDAX code implementing the AFEN method is verified testing on the VVER-440 benchmark problem and a “simplified” VVER-1000 benchmark problem. The COREDAX code then applied to the original VVER-1000 benchmark problem exercise 2 (HZP case and HP case) provides very good results in comparison with those of other benchmark participants.     

Effects of water based Al2O3, TiO2, and CuO nanofluids as the coolant on solid and annular fuels for a typical VVER-1000 core

In this paper, the effect of nanofluids as the coolant on solid and annular fuels for a typical VVER-1000 core is analysed. The considered nanofluids are various mixture composed of water and particles of Al₂O₃, TiO₂, and CuO. The fuel rod is modeled using a CFD code. To validate the calculated results, the present results of solid fuel with nanofluid and pure water are compared with other studies which have been done with visual FORTRAN language, DRAGON/DONJON code, COBRA-EN code and the mentioned analytical approaches have been validated by comparing with the final safety analysis report (FSAR). The comparison of the calculated results shows that the results are in good agreement with other studies. Thus, the accuracy of the validation is satisfactory. Radial and axial temperature distributions in various components of fuel are illustrated. Moreover, the temperature distributions of the fuel, clad and coolant are described for water based Al₂O₃, TiO₂, and CuOnanofluids in solid fuel and annular fuel. The results are compared with base fluid and it is concluded the nanoparticles of Al₂O₃have good properties in comparison with other nanoparticles. By using the nanofluids, the central fuel temperature is reduced and the temperature of the coolant is increased. In addition, by increasing the heated surfaces in annular fuel, the heat flux on these surfaces is reduced, the minimum departure from nucleate boiling ratio (MDNBR) margin is increased, and therefore the critical heat flux can be increased. Finally, it is concluded the use of the annular fuel instead of solid fuel and also the use of the nanofluids as coolant in the core of the reactor, security and efficiency of the nuclear power plant will be increased.     

Comparison of RELAP5 calculations of VVER-1000 coolant transient benchmark phase 1 at different power

This paper provides comparisons between experimental data of “MCP switching on when the other three MCPs are in operation” and RELAP5 calculations with different initial levels of the reactor power 29.45% and 27.47% from the nominal.

The reference power plant for this analysis is Unit 6 at the Kozloduy nuclear power plant (NPP) site. RELAP5/MOD3.2 computer code has been used to simulate the investigated transient. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation.

This investigation was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Reloading pattern optimization of VVER-1000 reactors (in transient cycles) using imperialist competitive algorithm

In this paper, the Imperialist Competitive Algorithm was for the first time used for reloading pattern optimization of Bushehr's VVER-1000 reactor in the second cycle. Since the diversity of loadable fuels in the reactor core is at its highest level in the second cycle as compared to other operational cycles, it was decided to test optimization calculations in the most complicated state. To estimate the fuel compositions remained from the first cycle, and to precisely calculate the objective parameters of each of the arrangements examined in the optimization process, a program was designed based on the coupling of WIMS-D5B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermohydraulic part. The process of reloading pattern optimization was carried out in two states. In the first state, it was tried to obtain an arrangement with the maximum effective multiplication factor and the safe maximum power peaking factor. The objective of the second state was to obtain a reloading pattern with the flattest distribution of radial power peaking factor. In both of the optimization states, to ensure the optimality and safety of the proposed arrangements during the cycle, the behavior of thermo-neutronic parameters of the reactor core in the second cycle was studied through time-dependent calculations. The comparison between the results of this study and the arrangement proposed by the Russian contractor for a similar VVER-1000 reactor (Balakovo) revealed that the objective parameters of the arrangement proposed in this research provide more optimality. Finally, considering the innovative use of the imperialist competitive algorithm for optimizing reactor's reloading pattern and in view of the high speed of this algorithm, the present research can seemingly be a new step toward optimization of reloading patterns of nuclear reactors.     

Calculation of radial burnup and nuclides atom density distributions in a VVER-1000 fuel rod using Monte Carlo method

A fundamental knowledge of fuel behavior in different situations is required for safe and economic nuclear power generation. Due to the importance of a fuel rod behavior modelling in high burnup, in this paper, the radial distribution of burnup, fission products, and actinides atom density and their variations by increasing burnup and other factors such as temperature, enrichment and power density are studied in a fuel pellet of a VVER-1000 reactor in an operational cycle using the MCNPX 2.7 Monte Carlo code. A benchmark including a Uranium-Gadolinium (UGD) fuel assembly is used for verification of the developed model in the MCNPX code for radial burnup calculation. A sensitivity study is carried out to investigate the effect of different parameters such as the number of particles per cycle, the number of geometrical radial nodes in the fuel pellet, the number of burnup steps and the selection of different fission-product contents (i.e. those isotopes that are used for particle transport) on the MCNPX model for speed and accuracy compromising. To calculate the radial temperature profiles and to analyze the effect of temperature on the radial burnup distribution and vice versa, the HEATING 7.2 code, which is a general-purpose conduction heat transfer program, and the MCNPX code are applied together. The results show the accuracy and capability of the proposed model in the MCNPX and HEATING codes for radial burnup calculation.     

Semi 2D modeling of the horizontal steam generator PGV-1000 using the RELAP5 code

Developing a reliable thermal-hydraulic model of the steam generator is an essential process in the steady state and transient analysis for the Pressurized Water Reactor type of the Nuclear Power Plants. This paper provides a semi two dimensional thermal-hydraulic model of the PGV-1000 horizontal steam generator using the RELAP5 code. Applying the qualified nodalization and the cross-flow effects are some of the advantages in the present model. The obtained results from the RELAP5 steady state analysis showed a reasonable agreement with the Bushehr NPP Final Safety Analysis Reports (FSAR).     

WER-1000型反应堆停堆及临界模式优化研究

在确保安全的前提下,经济性是核电厂的重要目标之一.VVER-1000型反应堆某些非并网运行的工况,如换料后重新临界、热停堆及临界、试验后返临界等操作,在操作所占用的时间、原材料的消耗量以及产生的废水量等方面可作优化.笔者对影响停堆及临界操作的重要因素,即控制棒和硼酸浓度的配置进行定性和定量的分析,得出优化的一般步骤和基本原则,并对3个案例实施了优化.     

Effect of burnup on the neutronic parameters of ITU TRIGA Mark II research reactor

This study aims to estimate burnup of the fuel elements for the Istanbul Technical University TRIGA Mark II Research and Training Reactor using a Monte Carlo-based burnup-depletion code. Effect of burnup on the core neutronic parameters, effective core multiplication factor, fast/epithermal/thermal neutron fluxes, and core-average neutron spectrum, and incoming neutron spectrum of the piercing beam port (PBP), is investigated at the Beginning of Life (BOL) and End of Life (EOL). Operational data peculiar to a selected operation sequence, which contains positions of CRs, power level of the reactor, material temperatures and latest core map, are used to determine the current fuel burnup of fuel elements at the time under consideration. A specific operation sequence is selected for the analysis. Furthermore, all control rods are considered fully withdrawn to assess the excess reactivity. Results are obtained using MONTEBURNS2 with ENDFB/V-II.1 neutron/photon library for a full power of 250 kW. Neutron cross-section libraries at the full-power operating temperatures are generated using NJOY. From the results, the calculated burnup values of the core at the sequence considered and EOL are found to be 420 MWh and 560 MWh, respectively. Remaining excess reactivity is calculated to be less than 0.3 $. It is observed that core average thermal neutron flux reduces by 1 % while the fast and epithermal neutron fluxes remain almost unchanged.     

AP1000反应堆主泵屏蔽套制造工艺浅析

简要地从材料、成形、焊接、热处理几个方面对我国引进的第三代核电站AP1000反应堆主泵屏蔽套的制造工艺进行了浅析,阐明了在屏蔽套制造过程中应该注意的问题,对于实现我国反应堆主泵的国产化具有一定的积极意义。     

Effect of mechanical properties in the weld zone on the structural analysis results of a plate-type heat exchanger prototype and pressurized water reactor spacer grid

The mechanical properties in a weld zone are different from those in the base material owing to their different microstructures. A process heat exchanger in a nuclear hydrogen system is a key component to transfer high heat generated in a very high-temperature reactor to a chemical reaction that yields a large quantity of hydrogen. A spacer grid in pressurized water reactor (PWR) fuel is a structural component with an interconnected and welded array of slotted grid straps. Previous research on the strength analyses of these components was performed using base material properties owing to a lack of mechanical properties in the weld zone. In this study, based on the mechanical properties in the weld zone of components recently obtained using an instrumented indentation technique, strength analyses considering the mechanical properties in the weld zone were performed, and the analysis results are compared with previous research.     

Experimental and numerical simulation of air-water two-phase flow in the rod bundle with grid spacer

1 Introduction Grid spacer is the key part of reactor fuel assem-bly. The presence of spacers in fuel assemblies affectsvarious thermal-hydraulic characteristics of the reactorcore. The grid spacer with fine performance can im-prove thermal-hydraulic performance of the core fuelassembly and enhance the critical heat flux withouttoo much augment of the pressure loss. As a result,the implementation of grid spacer with high thermalperformance provides more thermal margin, then in-creases s…