Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

Introducing and validating a new method for coupling neutronic and thermal-hydraulic calculations

In this study, a new and innovative method is introduced for analyzing neutronic and thermal-hydraulic calculation. For this aim, VVR-S research reactor was selected, and the calculation procedure was performed for it. WIMS, CITATION and COBRA-EN codes were used for investigation. Calculation model consists of two sub-models: neutronic and thermo-hydraulic. The neutronic model uses WIMS and CITATION codes for neutronic simulation of the reactor core and calculating flux and power distribution over it. WIMS code simulates the fuel assemblies and CITATION models the core. The thermal-hydraulic model uses COBRA-EN code for performing the relative calculation. In this study, FORTRAN 90 program is used for linking two sub-models and performing the calculation. The proposed procedure is performed for VVR-S analysis and finally, the obtained results are compared with the experimental results that show good agreement with it.     

物理-热工耦合对超临界水堆系统特性的影响分析

物理-热工耦合是超临界水堆系统分析的关键问题之一。以日本超临界水冷热堆Super LWR的堆芯设计为例,借助Dragon编制中子截面数据库,建立双群中子扩散方程计算模块,联系同时建立的热工计算模块,得到超临界水堆的物理-热工耦合计算模型。通过对比稳态与瞬态工况下耦合前、后的热工工况,分析物理-热工耦合条件下的超临界水堆系统热工特性。结果表明：在稳态工况下,物理-热工耦合将导致内、外组件堆芯功率峰值沿轴向发生明显偏移,使得部分节点的包壳温度升高,但包壳最高温度降低;在瞬态工况下,物理-热工耦合将导致堆芯包壳最高温度的发生位置有所改变。发生给水加热丧失瞬态后,在某一时刻,外部组件的包壳最高温度将转而超过内部组件的包壳最高温度。可见,物理-热工耦合对包壳最高温度的大小和发生位置均可能产生明显影响。计算分析可为超临界水堆瞬态及安全分析提供相应理论参考。     

中子学和热工水力学耦合计算方式的研究

目前在反应堆设计和安全分析中,大量使用由已有的中子学计算程序和热工水力学计算程序连接而成的耦合程序。耦合方式的不同直接影响计算速度和精度。本文介绍了对几种不同的耦合方式所做的数值计算研究,得到一些有启发性的结果.     

Best Estimate BWR Transient Analysis with TRACG Assessment using Data from BWR Startup Tests and LOCA Integral Tests

TRACG is a new version of the best estimate BWR transient analysis code, which utilizes a multi-dimensional two-fluid model for the thermal hydraulics and a three-dimensional neutron kinetics model. A three-dimensional neutronics, a fully implicit integration scheme and models for advanced BWR components have been implemented in the code upon TRAC-BF1.

Assessment of TRACG has been performed in this study for the predictive capability of plant transients, which include thermal-hydraulic and neutronic interactions, as affected by responses of the plant control system. Simulations were presented for BWR representative transient tests, which were done as part of a series of BWR5 startup tests. As for the capability to predict thermal hydraulics during the design basis LOCAs, simulations were presented for the LOCA integral tests conducted in the ROSA-III at JAERI and the Hitachi TBL, which had been used for assessment of the TRAC former version.

Consequently, (1)the space-dependent power flow transitions in a BWR were confirmed by TRACG simulations in which the module coupled with neutronics and thermal hydraulics during transients has been newly introduced, and (2) the characteristic thermal-hydraulic phenomena including multi-channel effects during the design basis LOCAs were confirmed, as well as the TRAC former version, by TRACG simulations on which the influence due to a fully implicit integration scheme has not extended. Capability of TRACG to predict BWR transients ranging from simple plant operational transients to design basis LOCAs was successfully demonstrated.     

A Transient Modal Analysis of a BWR Instability Event

To solve the time dependent neutron diffusion equation a modal method, based on the expansion of the neutronic flux in terms of the dominant Lambda modes of a static configuration of the reactor is presented. This method is used to analyse transients of a nuclear power reactor where an instability event can be developed. A simulation of a transient with the same conditions given for the case 9 of Ringhals stability benchmark has been analysed. It is shown that with these conditions an out of phase oscillation associated with the two first azimuthal modes can be developed. These results are corroborated using a power modal decomposition, using the local power distribution provided by RAMONA code. To complete the analysis, the modal feedback reactivities have been calculated to study the coupling mechanism among modes.     

混合堆、聚变堆包层热工水力与中子学综合分析模拟及FEB混合堆包层综合分析

为了混合堆及聚变堆包层分析的需要，开发了ＢＩＴＨ程序，以对包层的热工水力学及中子学进行综合分析。简述球床的热工水力特征及其数学物理模型，介绍编制的包层热工水力分析程序ＴＨＰＢＨＲ，对ＢＩＳＯＮ１．５全面改造，考虑了共振自屏效应，并与热工水力计算相耦合，并更换ＢＩＳＯＮ１．５自带数据库，修改燃耗计算方法，以适应放射性废物处理、辐照损伤等计算需要。还对ＦＥＢ混合堆外包层用ＢＩＴＨ程序进行了分析。     

浅谈核电领域中的热工水力分析程序

比较了几种典型热工水力分析程序的功能和应用范围,指出了保守估算方法与最佳估算方法的特点以及二者之间的差异,阐述了热工水力分析程序与堆芯物理计算程序及计算流体力学程序耦合的应用和意义,并分析了我国热工水力分析程序的现状和发展。     

Calculations of reactivity-initiated transients in gas-cooled fast reactors using the code system fast

The FAST code system is a general tool for analyzing advanced reactors from the viewpoint of the static and dynamic behavior of the whole reactor system. It includes an integrated three-dimensional representation of the core neutronics, appropriate modeling of the core thermal-hydraulics and fuel pin behavior, coupled to models of the reactor primary and secondary systems. Use is made largely of well-established individual neutronic, thermal-hydraulic and fuel behavior modules. Clearly, it is important to verify the individual parts of the code, including the links between them. The paper is focused on this detailed verification procedure. Steady-state conditions, as well as the transient behavior of hypothetical reactivity-initiated accidents, are investigated for two specific gas-cooled fast reactors. While the first system, a CO₂-cooled CAPRA-CADRA core, is loaded with Superphénix-like MOX fuel, the second system being analyzed, a He-cooled Generation IV-like core, uses ceramic (U,Pu)C fuel dispersed in a silicon-carbide matrix. In the current study, the TRAC/PARCS elements of FAST are compared with the 3D-kinetics stand-alone ERANOS/KIN-3D code, which is considered state-of-the-art, using as far as possible equivalent options. A new methodology is proposed to improve a diffusion-theory, coarse-group PARCS-solution by scaling the original cross-section derivatives and input kinetic parameters.     

10 MW固态燃料钍基熔盐堆稳态物理-热工耦合

固态燃料钍基熔盐堆(Thorium Molten Salt Reactor-Solid Fuel,TMSR-SF1)作为第四代先进核反应堆堆型之一,继承了熔盐冷却剂和球形燃料元件的许多优点和技术基础,具有良好的经济性、设计上的固有安全性、钍铀燃料的可持续性和防核扩散性。本文以10 MW固态燃料钍基熔盐堆为模型,利用MCNP(Monte Carlo N Particle Transport Code)和ANSYS Fluent等模拟程序对其进行多物理耦合分析,同时利用C++语言编写了堆芯活性区的物理-热工耦合计算程序,实现了MCNP计算结果与Fluent程序的对接,并且通过对比耦合前后结果,分析了堆芯功率密度分布、有效增殖因子、温度分布等主要参数,为熔盐堆的设计、安全性评估和操作运行提供了参考依据。     

Review of thermal-hydraulic calculations for Calvert Cliffs and H.B. Robinson PTS study

The thermal-hydraulic calculations for the USNRC pressurized thermal shock study, which were performed by the Los Alamos National Laboratory for the Calvert Cliffs nuclear power plant using the TRAC-PF1/MODI code and by the Idaho National Engineering Laboratory for the H.B. Robinson Unit 2 nuclear power plant using the RELAP5/MOD1.6 code, were reviewed at Brookhaven National Laboratory.To quantitatively review these calculations, a simple method based on mass and energy balances was developed at BNL to predict the primary system temperature. In this approach the entire reactor system was lumped into a single volume and the energy balance was applied to that volume. Because significant nonequilibrium effects made it difficult to estimate the pressures, the upper and lower bounds of the pressure were calculated using adiabatic and equilibrium assumptions.In general, the temperatures and pressures of the primary system calculated by both codes were reasonable. The secondary pressures calculated by TRAC indicated it had some difficulty with the condensation model. However, it is not expected that this uncertainty would affect the transient calculations significantly.Review of one typical transient calculation for each plant is discussed in this paper.     

Homogeneous zones definition in deterministic codes and effect on computed neutronic parameters

The design or modification and in general the analysis and control of nuclear reactors require complex calculations, which are carried out by numerical codes including neutronic and thermal-hydraulic components. Among the neutronic codes, the deterministic ones which solve the neutron transport/diffusion equation simulate the reactor core by dividing it into homogenized zones, i.e. volumes within which the macroscopic nuclear properties are considered uniform. These codes have been extensively used and tested for several decades and are shown to perform well when they analyze reactor cores containing regions with relatively homogeneous distributions of fuel, moderator and absorbing materials. In this work, the sensitivity of computed key neutronic parameters to the partitioning of the reactor core in homogenized zones is examined. Application is made for a configuration of the Greek Research Reactor (GRR-1) core, which is pool type, fueled by slab-type fuel elements. For the calculations, the neutronic code system consisting of XSDRNPM (cell-calculations) and CITATION (core analysis) is used with two different definitions of homogeneous zones for the special/control fuel assemblies. The effect on computations of neutron flux distribution, void-induced reactivity and total control rod worth is examined based on corresponding measurements. It is shown that with a more appropriate partition in homogeneous zones, the agreement of computed results with measurements can be remarkably improved concerning mainly the neutron flux, while the control rods worth is the less affected quantity.     

核反馈对并联双通道自然循环系统流动不稳定性的影响

将热工水力系统分析程序RELAP5与三维物理瞬态输运程序TDOT T采用并行方式耦合,对并联双通道自然循环系统内核热耦合不稳定性进行分析,得到系统的不稳定边界。分别以燃料时间常数差异较大的板型元件及棒型元件为对象,讨论了核反馈对系统稳定性的影响。对于板型元件,核反馈作用对低含汽率区的第1类密度波振荡（DWO）有明显的抑制作用,而对高含汽率区的第2类DWO基本无影响。对于棒型元件,计算分析结果表明核反馈对系统稳定性几乎无影响。     

超临界水冷堆耦合条件下水棒热工特性分析

超临界水冷堆中需要单独设计水棒结构,水棒中流过慢化剂水使得堆芯得到充分慢化。本文采用日本设计堆型作为研究对象,自主设计S型、D1型、D2型3种不同水棒结构,并编制物理热工耦合程序,得到不同水棒结构及D2型水棒结构不同内层水棒外边长条件下慢化剂密度、冷却剂和慢化剂的平均密度及功率的轴向分布。结果表明：D2型双层水棒具有更均匀的慢化剂温度分布和轴向功率分布,随着内层水棒外边长的增大,轴向慢化剂密度均值有所提高。     

Analysis of TRAC-PF1 calculated core heat transfer for a CCTF test

TRAC-PF1 posttest calculation for CCTF test C1-5 (Run 14) was performed to assess the core thermal-hydraulic models in the TRAC-PF1 code during the reflood phase of a PWR LOCA. TRAC showed good agreement with data for heater rods turnaround temperatures and turnaround times in the lower half of the core. However, TRAC overpredicted turnaround times and underpredicted quench times in the upper part of the core. Even though heat transfer correlations have a strong dependency on the local void fraction in TRAC, TRAC-predicted void fraction profiles showed poor agreement with CCTF data that have been inferred from differential pressure measurements. From these comparisons with CCTF data, the following areas for future improvements of TRAC-PF1 should be considered: (1) the core hydraulic modeling used to calculate the void fraction profile in the core. (2) the method for evaluating heat transfer within the core.     

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis.     

石墨慢化通道式熔盐堆的稳态热工水力计算模型

针对石墨慢化通道式熔盐堆的堆芯结构,基于COMSOL Multiphysics程序和MATLAB程序建立了堆芯稳态热工水力学计算模型。该模型对堆芯内固体区域的温度分布采用三维热传导方程进行模拟,对通道内熔盐温度采用一维单相流体模型进行计算。固体区域与熔盐通过熔盐通道壁面的对流换热边界建立热耦合。该模型基于平行通道压力损失相等的原则,分配堆芯内各熔盐通道的流量。通过对比RELAP5程序的计算结果,验证了模型对温度和流量分配计算的正确性。针对2 MWt 液态燃料熔盐堆的一种概念设计,分析了堆芯内三维温度分布和通道间流量分配。该模型可精确计算通道式熔盐堆堆芯内稳态温度分布和流量分配,对堆芯的热工水力学设计具有重要意义。     

An assessment of a VVER-1000 core during Turbo-Generator load reduction test using RELAP5/MOD3.2 and WIMSD-5B/PARCSv2.7

Reactor dynamic tests, which are categorized as one of the power start-up test groups, are the most complex tests during commissioning of the new nuclear power plants. This paper presents the results of Turbo-Generator load reduction test as one of the reactor dynamic tests for VVER-1000/V446 unit at Bushehr Nuclear Power Plant (BNPP). In this test modeling because of the need for control rod bank worth and core reactivity coefficients, the core geometry has been modeled first by using WIMSD-5B/PARCSv2.7 codes for neutronic calculations. For performing the thermal-hydraulic analysis, the RELAP5/MOD3.2 computer code has been used. The control rod bank worth and core reactivity coefficients obtained from WIMSD-5B/PARCSv2.7 are compared with BNPP FSAR that confirm the ability and reliability of the method. Also comparison of the thermal-hydraulic core parameters obtained from RELAP5/MOD3.2 against actual plant data, indicate that this code can properly predict behavior of VVER-1000 reactor for this dynamic start-up test.     

Development of a thermal-hydraulic subchannel analysis code for motion conditions

Since the conventional subchannel analysis codes are designed for the land-based reactor core, a thermal-hydraulic subchannel analysis code was developed to evaluate thermal-hydraulic characteristics of the reactor core under motion conditions. The verification of the code was performed with experimental data and commercial codes. The ISPRA 16-rod tests were used to evaluate the steady-state prediction performance of the code, and the simulation results agree well with the test data. COBRA-EN code was applied to check the transient prediction performance of the code, and there is a good agreement between the predictions with both codes. An additional forces model for motion conditions was proposed in the code, and CFX-14.0 code was applied to verify the model. The results show that the code can be used in the thermal-hydraulic analysis of the reactor core under motion conditions. To illustrate the capabilities of the code, a fuel bundle under a complex motion condition was simulated, and the results are reasonable.