Analysis of the OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation

OECD/NEA ROSA Project experiment with the large scale test facility (LSTF) in JAEA was conducted simulating a PWR 1% cold leg small break LOCA with an assumption of high-power natural circulation due to failure of scram and total failure of high pressure injection system. The core power curve for the LSTF experiment was obtained by PWR LOCA analysis using JAEA-developed coupled three-dimensional kinetics and thermal-hydraulics code SKETCH-INS/TRAC-PF1 with detailed core model. A post-test analysis was performed against the obtained data by using JAEA-modified RELAP5/MOD3.2.1.2 code to validate the code predictability. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient of 0.61 for two-phase break flow. In the experiment, flow in hot legs became supercritical during two-phase natural circulation, causing the hot leg liquid level to be quite low. Liquid accumulation in steam generator U-tube upflow-side took place during reflux condensation mode due to high vapor velocity. The RELAP5 code predicted reasonably well the overall thermal-hydraulic phenomena observed in the experiment. The code, however, overpredicted the break flow rate especially during two-phase flow discharge period probably because of the failure in the correct simulation of the cold leg liquid level due to late decrease in the primary loop flow rate.     

RIA基准题验证堆芯三维物理-热工耦合程序

以COBRA-Ⅳ和NLSANMT程序为基础,开发了堆芯三维物理-热工耦合程序C4/NK。针对两个典型的反应性引入事故（RIA）,即NEACRP弹棒基准题和提棒基准题,分别进行了验证计算。与参考值和其他程序的计算结果相比,C4/NK耦合程序具有较好的精度,能正确模拟瞬态过程中的物理-热工反馈现象。     

典型压水堆堆芯物理-热工耦合稳态计算软件的开发与验证

为能更加准确地模拟典型压水堆中强烈的物理-热工耦合现象,研制了压水堆堆芯物理 热工耦合计算软件ARMcc。其中物理计算模块基于四阶节块展开法（NEM）和格林函数节块法（NGFM）,热工计算模块基于一维的单相单通道换热模型和一维圆柱导热计算模型,在程序中采用有限体积法和有限差分法求解一维圆柱导热模型。基于典型压水堆基准题NEACRP-L-335对程序的稳态耦合计算能力进行了验证,程序计算的堆芯关键参数如临界硼浓度、堆芯多普勒温度等参数与参考结果符合良好,临界硼浓度与参考结果的相对偏差均小于0.5%。另外研究4种计算模式对模拟堆芯物理-热工耦合过程的影响,选择PARCS程序计算结果为对比,发现NGFM+DIF模式能更加准确地模拟堆芯燃料多普勒温度和堆芯功率分布;NGFM+VOL模式能更加准确地模拟临界硼浓度;NEM+VOL模式能更加准确地模拟堆芯燃料最高温度。     

基于格林函数节块法的物理与热工-水力耦合方法研究

开发了三维物理与热工-水力耦合的PWR堆芯瞬态分析程序NGFMN-K/COBRA-Ⅳ/COBRA-Ⅳ(NCC)。少群时空中子动力学计算采用格林函数节块法程序NGFMN-K,隐式耦合子通道程序COBRA-Ⅳ实现瞬态计算。采用P10H8B功率重构方法给出热组件栅元功率分布,耦合另一个COBRA-Ⅳ程序模块,进行热组件子通道分析得到安全参数。对NEACRP-L-335 C1弹棒基准问题的计算表明,NCC程序的计算结果与参考结果符合很好,说明程序计算正确,可用于评估事故结果。     

压水堆核电站堆芯物理/热工水力耦合特性研究

采用RELAP5-HD作为堆芯耦合计算程序,以秦山核电二期工程反应堆堆芯为研究对象,建立堆芯活性区的物理/热工水力耦合模型,在此基础上进行了稳态计算和掉棒事故仿真研究。结果表明,使用RELAP5-HD计算得到的结果与电厂实测值符合较好,获得的掉棒事故参数曲线能准确反映事故工况下的参数变化趋势。稳态和事故工况的计算结果均符合堆芯物理/热工水力反馈效应的理论分析,证实了所建立的堆芯耦合模型的准确性,为下一步进行核电站系统的仿真分析提供基础。     

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.     

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.     

Validation of the CFD code fluent by post-test calculation of a density-driven ROCOM experiment

During the last years, boron dilution events with the potential of reactivity transients were an important issue of German PWR safety analyses. A coolant with a low-boron concentration could be collected in localized areas of the reactor coolant system, e.g., by separation of a borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux-condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems.During the course of follower core assessments, TÜV NORD SysTec appraises safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses demonstrate that boron dilution events cannot lead to recriticality of the core. Hence, the boron concentration at the core inlet has to be determined.TÜV NORD SysTec applies the CFD code FLUENT for the investigation of boron dilution events in pressurized water reactors. To affirm the FLUENT abilities for the simulation of boron dilution events, a validation against the ROCOM experiment T6655_21 with a density-driven coolant mixing was performed. This validation proves that FLUENT is able to appropriately simulate the effects of boron transport and dilution such as streaks of coolant with lower density in the downcomer. Deficits were identified in the simulation of fluid layering in the cold leg, which fortunately have a rather small influence on the predicted core inlet concentration. Therefore, the boron concentration in the reactor core can be determined with sufficient accuracy to solve the safety issue, regardless of the core becoming critical or not.     

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.     

Nodal kinetics model upgrade in the Penn State coupled TRAC/NEM codes

The Pennsylvania State University currently maintains and does development and verification work for its own versions of the coupled three-dimensional kinetics/thermal-hydraulics codes TRAC-PF1/NEM and TRAC-BF1/NEM. The subject of this paper is nodal model enhancements in the above mentioned codes. Because of the numerous validation studies that have been performed on almost every aspect of these codes, this upgrade is done without a major code rewrite. The upgrade consists of four steps. The first two steps are designed to improve the accuracy of the kinetics model, based on the nodal expansion method. The polynomial expansion solution of 1D transverse integrated diffusion equation is replaced with a solution, which uses a semi-analytic expansion. Further the standard parabolic polynomial representation of the transverse leakage in the above 1D equations is replaced with an improved approximation. The last two steps of the upgrade address the code efficiency by improving the solution of the time-dependent NEM equations and implementing a multi-grid solver. These four improvements are implemented into the standalone NEM kinetics code. Verification of this code was accomplished based on the original verification studies. The results show that the new methods improve the accuracy and efficiency of the code. The verification of the upgraded NEM model in the TRAC-PF1/NEM and TRAC-BF1/NEM coupled codes is underway.     

Cross-section modelling effects on pressurised water reactor main steam line break analyses

Because of the strong asymmetric overcooling effects occurring during a PWR main steam line break (MSLB) event, an accurate analysis of this transient requires the use of 3-D kinetics methods. An assessment has been made of the relative performance of the two kinetics solvers currently employed at PSI for such analyses, viz. CORETRAN and SIMULATE-3 K. For the purpose, the simulation of a hypothetical MSLB in a real operated PWR MOX cycle has been considered, employing consistent 3-D core models with specified thermal-hydraulic boundary conditions at the lower and upper plenums. Although the employed cross-section library is in both codes based on the same set of homogenised 2-group cross-sections prepared with CASMO-4, significant differences are shown to occur due to the smaller moderator reactivity coefficient calculated in CORETRAN. It is found that this stems largely from differences in the cross-section formalism, i.e. the manner in which feedback dependencies are modelled and interpolated for the cross-section sets.In particular, the CORETRAN cross-section formalism induces an inadequate treatment of coupled feedback effects, principally between boron density and moderator temperature, which renders the MSLB dynamics predictions quite sensitive to the methodology employed during the cross-section preparation. As such, transient-specific cross-section libraries need to be produced for reliable MSLB analysis in this case. The cross-section model for SIMULATE-3 K, on the other hand, is shown to be adequate for accurately capturing the coupled reactivity effects occurring during an MSLB. In this case, the sensitivity of the results to other sources of uncertainties becomes more apparent, e.g. to those related to the neutron data and/or the thermal-hydraulic boundary conditions. Considering that many other state-of-the-art advanced kinetics solvers have cross-section formalisms similar to that of CORETRAN, effects of the type currently investigated need to be taken into account while developing methodologies for assessing neutronics-related uncertainties in best-estimate transient analysis.     

Temporal adaptive algorithm for TRAC-BF1/NEM/COBRA-TF coupled calculations in BWR safety analysis

A temporal adaptive algorithm was developed to perform the synchronization and optimization of the performance of TRAC-BF1/NEM/COBRA-TF three-dimensional neutron/thermal-hydraulics sub-channel analysis coupled code system. The multi-level coupling scheme for time synchronization of the TRAC-BF1/NEM and COBRA-TF under PVM is developed considering the different time-step selection algorithms of TRAC-BF1, NEM and COBRA-TF codes. The developed methodology allows one to synchronize the codes in time without doing significant code modifications to the time-step selection logic of the involved codes. The advantage of this approach is that COBRA-TF can capture the nature of a given transient, without losing any time-dependent data. Results for steady state and transient calculations that show how the implemented temporal adaptive algorithm works are presented. In addition selected results are presented to illustrate dynamic behavior and the type of thermal-hydraulic boundary conditions provided by the system code.     

The inherent boron dilution mechanism in pressurized water reactors

The purpose of this paper is to describe a mechanism that inherently causes boron dilution in pressurized water reactors (PWRs). The phenomenon is due to the fact that boric acid does not markedly dissolve into steam. This is relevant for transient and accident situations in PWRs where decay heat removal is accomplished by coolant vapourization and condensation, which inherently leads to formation of dilute plugs in the primary. In particular, it is found that inherent dilution will be inevitable for a range of small break loss of coolant accidents (SB LOCAs), with maximum amount of total diluted coolant mass exceeding 20 tonnes for a modern 1300 MWe PWR equipped with U-tube steam generators. A simple analysis of dilute plug motion during the late phases of a SB LOCA and core response to boron dilution shows that the damaging potential might extend to widespread fuel failures. Other transients and accidents are also discussed from the point of view of inherent dilution. Some possible remedies to the problem, as well as suggestions for further research, are presented.     

A 10% cold leg break test at ROSA-IV large scale test facility

The Large Scale Test Facility (LSTF) in the ROSA-IV Program is an integral test facility for investigation of pressurized water reactor (PWR) thermal-hydraulic behavior during small break loss-of-coolant accidents (SBLOCAs) and operational transients.A 10% cold leg break test was conducted in the facility shakedown phase to assess and confirm the facility capability and to collect code assessment data. The test conditions, test procedures and test results are described. The test results are compared with a pretest analysis obtained using RELAP5/MOD1 Cy18.     

Development of a novel 1D coupled neutronics/thermal-hydraulics code and its verification on PWR rod ejection accident benchmark

This paper presents a new 1D Neutronics/Thermal-hydraulics code ATAC-1D based on the advanced Jacobian-Free Newton-Krylov (JFNK) method and the low dimensional equivalent strategy. Conventional operator-splitting (OS) strategies are used to maintain its accuracy with small time steps and linearization of the nonlinear problem, which leads to slow computation speed and linearization error. The JFNK method solves the troubles in the coupled neutronics/thermal-hydraulics problems mentioned above. Furthermore, a core-wide three dimension to one dimension equivalent method has been developed to provide variable few-group parameters. Finally, the performance of the coupled neutronics/thermal-hydraulics code ATAC-1D is studied by simulating four OECD/NEA CRP PWR rod ejection benchmark problems. The simulation results are compared to the reference ones, which proves that the developed 1D code has a good accuracy and practicability in nuclear reactor transient calculation.     

Thermal-hydraulic transient analysis of light water reactors using moving boundaries

This paper examines the applicability of a mathematical dynamic model developed here for the simulation of the thermal-hydraulic transient analysis for light water reactors (LWRs). The thermal-hydraulic dynamic modeling of the fuel pin and adjacent coolant channel in LWRs is based on the moving boundary concept. The fuel pin model (FUELPIN) with moving boundaries is developed to accommodate the core thermal-hydraulic model, with detailed thermal conduction in fuel elements. Some results from transient calculations are examined for the first application of the thermal-hydraulic model and the fuel pin model with moving boundaries in a boiling water reactor (BWR). An accurate minimum departure from nucleate boiling ratio (MDNBR) and its axial MDNBR boundary versus time within the fuel channel are predicted during transients. Transient analysis using a known thermal-hydraulic code, COBRA and FUELPIN linked with a PWR systems analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a pressurized water reactor (PWR). The studies of the overall nuclear reactor system show that moving boundary formulation provides an efficient and suitable tool for thermal transient analysis of LWRs.     

GASFLOW程序液膜模型开发及初步验证

本文基于三维CFD安全壳程序GASFLOW开发了热构件壁面上的液膜覆盖与蒸发模型。通过选定AP1000大破口事故序列,采用耦合了液膜模型的GASFLOW程序分析了AP1000核电厂安全壳内温度压力响应及其非能动安全壳冷却系统(PCS)的性能,并与相同事故序列下WGOTHIC、MELCOR、CONTAIN等程序的计算结果进行比较。结果表明,耦合了液膜模型的GASFLOW程序可用于分析PCS的热工水力行为,其基本功能满足计算需要。     

Boron transient effects on the behavior of IRIS reactor using dynamic modeling

Small Modular Reactors (SMR) are considered as having several advantages over typical nuclear reactors under various specific conditions. They are thought to be installed in countries with small or medium power grid, in which a large power plant is not necessary or in isolated communities far from distribution centers. A plenty of developing countries are in this situation, so that a significant demand on this type of reactor is expected in a near future. The IRIS reactor is the top-front of SMRs, making its complete development very attractive, since it can fulfill the essential requirements for a future nuclear power plant: better economics, safety-by-design, low proliferation risk and environmental sustainability. IRIS reactor is an integral type PWR in which all primary components are arranged inside the pressure vessel. This configuration involves important changes when compared with a conventional PWR. These changes require several studies to comply with the safe operational limits for the reactor. In light water reactors, a solution of boric acid is used in the coolant of the primary loop to absorb neutrons, aiming to adjust the reactivity of the reactor. A significant decrease in the boron concentration in the core might lead to a considerable power excursion. Several studies on PWR have established correlations between power excursions and deficiencies in homogenization of boric acid diluted in the coolant. The IRIS reactor, due to its integral configuration, does not possess a spray system for boron homogenization which may cause power transients. In this paper, a study has been conducted to develop a dynamic model (named MODIRIS) for transient analysis, implemented in the MATLAB'S software SIMULINK, allowing the analysis of IRIS behavior by considering the neutron point kinetics model for power generation. The methodology is based on generating a set of differential equations of neutronic and thermal-hydraulic balances which describes the dynamics of the primary circuit, as well as a set of differential equations describing the dynamics of secondary circuit. The equations and initialization parameters at full power were inserted into the SIMULINK and the code was validated by comparing with RELAP simulations for a transient of feedwater reduction in the steam generators. Furthermore, the current paper looks for studying and developing a dynamic model for calculating the variations in the boric acid concentration. Then, a simplified model for boron dispersion was implemented into the code MODIRIS to simulate power transients which occur due to variations in the boron concentration in the primary loop of the IRIS reactor. The results for boron concentration, inserted reactivity and steam production showed a good precision and represented the expected behavior very well in the range of operational transients.     

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.