Xenon transient simulation of the VVER-1000 nuclear reactor using adiabatic approximation

The neutronic behavior of very slow transients like fuel burn up and xenon studies can be performed with the sequence of instantaneous criticality calculations. Such a scheme is known as the adiabatic approximation. ¹³⁵Xe as a fission product has an enormous thermal absorption cross section, on the order of a million barns, therefore the study of xenon poisoning and its effect on flux and power distribution is very important in thermal reactors. In this work xenon transient analysis of the VVER-1000 nuclear reactor and its effect on the flux and power distribution from reactor start up to xenon saturation and the change of power from nominal to 25% of nominal is carried out using WIMS and CITATION codes. We used the WIMS cell calculation code and found some relations between xenon concentration and group constants of different FA (Fuel Assemblies); in this way we bypassed the WIMS running at each time step. Also, the CITATION code is used as a core calculation code to find the effective multiplication factor as well as flux and power distributions. In order to link WIMS and CITATION codes and facilitate numerous executions, a VISUAL BASIC program has been developed. The results have a good agreement with the safety analysis report of the reference plant such that the relative differences in most cases are less than 10%.     

Control rod worth calculation for VVER-1000 nuclear reactor using WIMS and CITATION codes

This paper summarizes efforts related to developing a technically justifiable approach for investigating the control rod worth of VVER-1000 reactor. For this assessment, computer simulation of nuclear reactor was needed. In this study nuclear reactor behavior was modeled by WIMS code, which solves transport equation for fuel assemblies' modeling at first step, and CITATION code that solves diffusion equation for core modeling. From these two codes, neutronic calculation of reactor was performed and control rod worth was calculated. Results of this study are comparable with the plant's FSAR.On comparing results of this study and reference some unacceptable errors were discerned. To find out the cause of these errors, some efforts had been performed and finally was discerned that the method of cell calculation, i.e., DSN method, was the important cause of errors. Therefore, some analysis had been performed by WIMS in PIJ + PERSEUS method and was shown that the results were improved.     

Benchmark analysis of SPERT IV reactor

Benchmark calculations have been performed for SPERT IV D-12/25 core. Experimental data of the core was provided by International Atomic Energy Agency (IAEA). Combination of WIMS/D4 and CITATION codes has been used for performing the neutronic analysis of the reactor. Lattice calculations have been performed through WIMS/D4 while 3-dimensional reactor core has been modeled in CITATION. Ten energy groups were considered for these calculations. Energy wise microscopic cross-sections were generated for fuel, control absorber, control follower, guide tube, grid plate, reflector and structural regions separately of the core using WIMS/D4. Thermal neutron flux profiles at different axial and radial locations of the core were evaluated. Critical position of the control rods, excess reactivity, shut down margin, control rod worth, reactivity feed back coefficients and kinetic parameters of the core were estimated. Reasonable agreement has been found between experimentally determined and the calculated parameters.     

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.     

Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries

The aim of this study is to analyze the neutronic parameters of TRIGA Mark-II research reactor using the chain of NJOY-WIMS-CITATION computer codes based on evaluated nuclear data libraries CENDL-2.2 and JEFF-3.1.1. The nuclear data processing code NJOY99.0 has been employed to generate the 69 group WIMS library for the isotopes of TRIGA core. The cell code WIMSD-5B was used to generate the cross sections in CITATION format and then 3-dimensional diffusion code CITTATION was used to calculate the neutronic parameters of the TRIGA Mark-II research reactor. All the analyses were performed using the 7-group macroscopic cross section library. The CITATION test-runs using different cross section sets based on different models applied in WIMS calculations have shown a strong influence of those models on the final integral parameters. Some of the cells were specially treated with PRIZE options available in WIMSD-5B to take into account the fine structure of the flux gradient in the fuel-reflector interface region. It was observed that two basic parameters, the effective multiplication factor, k_eff and the thermal neutron flux, were in good agreement among the calculated results with each other as well as the measured values. The maximum power densities at the hot spot were 1.0446E02 W/cc and 1.0426E02 W/cc for the libraries CENDL-2.2 and JEFF-3.1.1 respectively. The calculated total peaking factors 5.793 and 5.745 were compared to the original SAR value of 5.6325 as well as MCNP result. Consequently, this analysis will be helpful to enhance the neutronic calculations and also be used for the further thermal–hydraulics study of the TRIGA core.     

铀氢锆堆物理计算及燃料管理软件包

建立了一套铀氢锆堆物计算软件包，首先考虑氢化锆中的热化特殊性，按ＷＭＳ格式制作 了氢化锆 氢的６９群群常数并入ＷＩＭＳ－Ｄ／４数据库中，形成了ＷＩＭＳ－Ｎ１库和ＷＩＭＳ－Ｎ２库；应用ＷＩＭＳ－Ｎ２库和国际通用的ＷＩＭＳ－Ｄ／４程序包计算了铀氢锆堆各类栅元的群常数，应用差分程序ＣＩＴＡＴＩＯＮ和六角形节块和ＳＩＸＴＵＳ进行扩散计算，同时在ＳＩＸＴＵＳ－２程序的基础上编制了燃料管理程序和ＸＰＲ－ＩＣＦ     

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.     

Comparison between triangular and hexagonal modeling of a hexagonal-structured reactor core using box method

A hexagonal-structured reactor core (e.g. VVER-type) is mostly modeled by structured triangular and hexagonal mesh zones. Although both the triangular and hexagonal models give good approximations over the neutronic calculation of the core, there are some differences between them that seem necessary to be clarified. For this purpose, the neutronic calculations of a hexagonal-structured reactor core have to be performed using the structured triangular and hexagonal meshes based on box method of discretisation and then the results of two models should be benchmarked in different cases.In this paper, the box method of discretisation is derived for triangular and hexagonal meshes. Then, two 2-D 2-group static simulators for triangular and hexagonal geometries (called TRIDIF-2 and HEXDIF-2, respectively) are developed using the box method. The results are benchmarked against the well-known CITATION computer code in case of a VVER-1000 reactor core. Furthermore, the relative powers calculated by the TRIDIF-2 and HEXDIF-2 along with the ones obtained by the CITATION code are compared with the verified results which have been presented in the Final Safety Analysis Report (FSAR) of the aforementioned reactor.Different benchmark cases revealed the reliability of the box method in contrast with the CITATION code. Furthermore, it is shown that the triangular modeling of the core is more acceptable compared with the hexagonal one.     

Revised calculation route for two group macroscopic cross-section library for RBMK-1500 type reactor

To increase the accuracy of predicted reactivity effects and coefficients for the unit equipped with a RBMK-1500 type reactor at Ignalina NPP, the calculation route used to generate the library of nuclear data constants applied in the neutronic/thermal hydraulic analysis has been updated with a modern version of the WIMS lattice code, WIMS8. The previously available two group library used with the QUABOX/CUBBOX-HYCA code, adapted to model the physical and nuclear processes in a RBMK-1500 reactor core, was created using the freely available WIMSD reactor physics cell code and its associated nuclear data library. In this article, the results that are obtained under the performance of the new two group cross-section library generated with WIMS8 for RBMK-1500 design core are presented. This discussion is mostly concentrated on the prediction of the key physics parameter for the RBMK type reactor, the void reactivity coefficient, as this parameter has been underestimated, especially at higher fuel irradiation.     

基于CITATION-ORIGEN2球床堆平衡态计算程序的实现

基于确定论中子扩散软件CITATION和点燃耗软件ORIGEN2,编写了球床堆分析程序COBBLE,以实现指定燃料球加载策略下的球床堆平衡态燃耗计算。COBBLE程序采用谱区能谱修正方法,通过迭代求解得到球床堆堆芯平衡态参数。本文选取简化的球床模块高温气冷堆(PBMR)堆芯进行建模,计算其功率分布及燃耗分布,并使用基于蒙特卡罗方法的球床堆燃耗计算程序PBRE进行了验证与分析。结果表明,COBBLE程序适用于球床堆的平衡态燃耗计算。     

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.     

Neutronic simulation of water-based nanofluids as a coolant in VVER-1000 reactor

The main goal of this study is to perform the neutronic simulation of nanofluids application to reactor core. The variation of the Bushehr VVER-1000 reactor primary neutronics parameters is investigated with using different nanofluids as coolant. In the present neutronic simulation, water-based nanofluids containing various volume fractions of Al₂O₃, Si, Zr, TiO₂, CuO, Ti, Cu and Ag nanoparticles are investigated. Optimization of type and volume fraction of nanoparticles affects the reactor neutronic characteristics. The results achieved by using WIMS and CITATION codes, show that below 0.1 percent volume fraction of Al₂O₃ is the optimum nanoparticle for normal operation and Ag/water nanofluid is suggested to use as a reactor safety enhancement tool.     

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。     

A methodology for the coupling of RAMONA-3B neutron kinetics and TRAC-BF1 thermal-hydraulics

The initial objective of this project was to directly couple the RAMONA and TRAC codes running on different PCs. The idea was to use the best part of each one and eliminate some of their limitations and widen the applicability of these codes to simulate different BWR and system components. It was required to try to minimize the amount of changes to present code subroutines and calculation procedures. If possible, just substitute values obtained in the parallel code. Preliminary results indicated that using a CHAN component of TRAC to model thermal-hydraulic phenomena for each neutronic channel modeled in RAMONA is rather difficult. Large amounts of CPU time consumption are obtained and lots of PCs would make this solution difficult, besides considerable large transients are introduced by the differences in thermal-hydraulic results of these codes. The substitution of the thermal-hydraulics of RAMONA, by the TRAC channel calculations, is possible but simulation of a null transient on both codes must be planed and a gradual change must be controlled by an additional supervisory subroutine. An indirect coupling of these codes, it is therefore proposed, in order to eliminate most of these limitations. In this indirect coupling, a thermal-hydraulic model of the average tube in a bundle and the global channel cooling fluid dynamics is programmed for each neutronic channel while the global reactor vessel and core is modeled by TRAC with just four channels and four rings. Results are more reliable, coupling is simpler and faster simulations are possible.     

Automation of the modeling and some neutronic calculations of the Syrian miniature neutron source reactors

An automated software, BMAC, for modeling and performing the neutronics calculations of MNSRs and similar reactors (TRIGAs) has been developed. Calculation of initial excess reactivity, flux and power distributions, and all other neutronic parameters of the reactor, full core representation, can be made automatically using a 3-D model, by coupling WIMSD-4 and CITATION codes, in a very quick and simple way. No preliminary CITATION input file is needed. All required data are read from an external input file simply prepared. Accurate results for the parameters of the reactor, in the framework of Diffusion Theory, can be obtained.     

Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).     

A spatial kinetic model for simulating VVER-1000 start-up transient

An accurate prediction of reactor core behavior in transients depends on how much it could be possible to exactly determine the thermal feedbacks of the core elements such as fuel, clad and coolant. In short time transients, results of these feedbacks directly affect the reactor power and determine the reactor response. Such transients are commonly happened during the start-up process which makes it necessary to carefully evaluate the detail of process. Hence this research evaluates a short time transient occurring during the start up of VVER-1000 reactor. The reactor power was tracked using the point kinetic equations from HZP state (100 W) to 612 kW. Final power (612 kW) was achieved by withdrawing control rods and resultant excess reactivity was set into dynamic equations to calculate the reactor power. Since reactivity is the most important part in the point kinetic equations, using a Lumped Parameter (LP) approximation, energy balance equations were solved in different zones of the core. After determining temperature and total reactivity related to feedbacks in each time step, the exact value of reactivity is obtained and is inserted into point kinetic equations. In reactor core each zone has a specific temperature and its corresponding thermal feedback. To decrease the effects of point kinetic approximations, these partial feedbacks in different zones are superposed to show an accurate model of reactor core dynamics. In this manner the reactor point kinetic can be extended to the whole reactor core which means “Reactor spatial kinetic”. All required group constants in calculations are prepared using the WIMS code. In addition CITATION code was used to calculate the flux, power distribution and core reactivity inside the core. To update the last change in group constants and resultant reactivity in point kinetic equations, these neutronic codes were coupled with a developed dynamic program. This study is applied on a typical VVER-1000 reactor core to show the reactor response in short time transients caused during start-up procedure.     

SHTV,as a technique for core calculation using spatial homogenization and temperature variation

The accuracy of static neutronic parameters in the nuclear reactors depends upon the determination of group constants of the diffusion equation in the desired geometry. Although several methods have been proposed for calculating these parameters, there is still the need for more reliable methods. In this paper a powerful and innovative method based on Spatial Homogenization and Temperature Variation (SHTV) of physical properties of a WWER-1000 nuclear reactor core for calculating the relative power distribution of Fuel Assemblies (FA) and the hot fuel rod, is presented. The method is based on replacing the heterogeneous lattices of materials with different properties by an equivalent homogeneous mixture of these material for determining the few group constants, while the effect of temperature variation in the fuel and coolant density along the axial core direction is considered. All calculations are performed using WIMS and CITATION codes. The obtained results are compared with the results of Final Safety Analysis Report (FSAR) prepared by the designer, and good agreement between the two results is shown.     

Design studies of a typical PWR core using advanced computational tools and techniques

This study deals with the design and development of calculational techniques and evaluation of key neutronic parameters of a typical PWR core having a total reactor power of 2652 MW_t (890 MW_e). The PWR core consists of 157 fuel assemblies containing a total of ∼72 tons of uranium arranged vertically in a concentric square array within the core shroud. Each fuel assembly contains 264 UO₂ fuel pins with various enrichments (2.1, 2.6 and 3.1%), 24 control rods of Gd₂O₃ and one central water channel and all are arranged in a 17 × 17 array of matrix. Different computer codes including WIMS, TWOTRAN, CITATION and MCNP have been employed to develop a versatile and accurate reactor physics model of the PWR core. The computational methods, tools and techniques, customization of cross section libraries, various models for cells and super cells, and a lot of associated utilities have been standardized and established/validated for the overall core analysis. The analyses were performed in 3 steps: firstly for fuel pincells, then for the fuel assemblies and finally for the whole core. The WIMS and MCNP calculated infinite multiplication factors for fuel pincells having 2.1% enriched ²³⁵U were found to be 1.23393 and 1.23654, for 2.6% enrichment 1.28635 and 1.28887, and finally for 3.1% enrichment 1.32481 and 1.32812, respectively. For fuel assembly, WIMS and MCNP calculated infinite multiplication factors having 2.1% enrichment were found to be 1.24853 and 1.25445, for 2.6% enrichment 1.30372 and 1.30992, and for 3.1% enrichment 1.34424 and 1.35041, respectively. The effective multiplication factor calculated by CITATION, TWOTRAN and MCNP for whole core were found to be 1.25580, 1.25909 and 1.26382, respectively. The peak thermal neutron flux in the core calculated by MCNP was found to be 5.0298 × 10¹⁴ neutrons/cm² s and the average core power density was 17.1 kW/cm³. The calculated results from different codes were found to be very good agreement for different moderator conditions. The choice of computer codes like WIMSD, TWOTRAN, CITATION and MCNP which are being used in nuclear industry for many years were selected to identify and develop new capabilities needed to support PWR analysis. The ultimate goal of the validation of the computer codes for PWR applications is to acquire and reinforce the capability of these general purpose computer codes to perform the core design and optimization study.