Analysis of the Peach Bottom flow stability test number 3 using the coupled RELAP5/PARCS code

Nowadays, the coupled codes technique, which consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, is extensively used for carrying out best estimate (BE) simulation of complex transient in nuclear power plants (NPP). This technique is particularly suitable for transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. Such complex interactions are encountered under normal and abnormal operating conditions of a boiling water reactors (BWR). In such reactors Oscillations may take place owing to the dynamic behavior of the liquid-steam mixture used for removing the thermal power. Therefore, it is necessary to be able to detect in a reliable way these oscillations. The purpose of this work is to characterize one aspect of these unstable behaviors using the coupled codes technique. The evaluation is performed against Peach Bottom-2 low-flow stability tests number 3 using the coupled RELAP5/PARCS code. In this transient dynamically complex neutron kinetics coupling with thermal-hydraulics events take place in response to a core pressure perturbation. The calculated coupled code results are herein assessed and compared against the available experimental data.     

Development of SIRIUS-N facility with simulated void-reactivity feedback to investigate regional and core-wide stability of natural circulation BWRs

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of a natural circulation BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of thermal-hydraulic conditions, power distributions, and fuel rod time constants, including the nominal operating conditions of a typical natural circulation BWR. The results show that there is a sufficiently wide stability margin under nominal operating conditions, even when void-reactivity feedback is taken into account. The stability experiments were extended to include a hypothetical parameter range (double-void reactivity coefficient and inlet core subcooling increased by a factor of 3.6) in order to identify instability phenomena. The regional instability was clearly demonstrated with the SIRIUS-N facility, when the fuel rod time constant matches the oscillation period of density wave oscillations.     

PARCS稳态两相热工水力程序PATHS的开发与耦合

本工作开发了PARCS的先进热工水力求解器PATHS,可对沸水堆进行热工水力稳态模拟。与RELAP5的计算结果进行验证,结果表明,PATHS的计算结果与RELAP5的基本一致。将PATHS与PARCS进行耦合,对SMART反应堆及Peach Bottom 2 OECD Turbine Trip基准题进行计算,结果表明,PARCS/PATHS耦合程序计算结果准确有效,能用于沸水堆的稳态物理热工耦合计算。     

Flashing-induced density wave oscillations in a natural circulation BWR—mechanism of instability and stability map

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of a natural circulation BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of thermal-hydraulic conditions, power distributions, and fuel rod time constants, including the nominal operating conditions of a typical natural circulation BWR. The results show that there is a sufficiently wide stability margin under nominal operating conditions, even when void-reactivity feedback is taken into account. The stability experiments were extended to include a hypothetical parameter range (double-void reactivity coefficient and inlet core subcooling increased by a factor of 3.6) in order to identify instability phenomena. The regional instability was clearly demonstrated with the SIRIUS-N facility, when the fuel rod time constant matches the oscillation period of density wave oscillations.     

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.     

Valuation of BWR stability operating in natural circulation conditions

Nowadays, new concepts of nuclear reactors have been projected to work with mechanisms of natural circulation (NC). However, NC systems are very susceptible to several kinds of instabilities being necessary careful studies about such systems. In this work, a theoretical investigation about BWR stability during a transient of recirculation pump trip bringing the reactor to operate at NC conditions is presented. The simulations were performed using the RELAP5/MOD3.3 thermal-hydraulic code and the PARCS/2.4 3D neutron-kinetic code in a coupled way to predict the transient results. The power time evolution and the related thermal-hydraulic parameters were investigated during the transient to analyze the behavior of the reactor for this special operation condition of NC.     

Valuation of BWR stability operating in natural circulation conditions

Nowadays, new concepts of nuclear reactors have been projected to work with mechanisms of natural circulation (NC). However, NC systems are very susceptible to several kinds of instabilities being necessary careful studies about such systems. In this work, a theoretical investigation about BWR stability during a transient of recirculation pump trip bringing the reactor to operate at NC conditions is presented. The simulations were performed using the RELAP5/MOD3.3 thermal-hydraulic code and the PARCS/2.4 3D neutron-kinetic code in a coupled way to predict the transient results. The power time evolution and the related thermal-hydraulic parameters were investigated during the transient to analyze the behavior of the reactor for this special operation condition of NC.     

Transient and stability analysis of a BWR core with thorium–uranium fuel

The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket–seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal–hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO₂ core, even during transient conditions. The stability and transient analysis show that the thorium–uranium fuel can be operated safely in current BWRs.     

Simulation of an hypothetical out-of-phase instability case in boiling water reactor by RELAP5/PARCS coupled codes

A very complex type of power instability occurring in boiling water reactor (BWR) consists of out-of-phase regional oscillations, in which normally subcritical neutronic modes are excited by thermal-hydraulic feedback mechanisms. The out-of-phase mode of oscillation is a very challenging type of instability and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, simulations of out-of-phase instabilities in a BWR obtained by assuming an hypothetical continuous control rod bank withdrawal are being presented. The RELAP5/Mod3.3 thermal-hydraulic system code coupled with the PARCS/2.4 3D neutron kinetic code has been used to simulate the instability phenomenon. Data from a real BWR nuclear power plant (NPP) have been used as reference conditions and reactor parameters. Simulated neutronic power signals from local power range monitors (LPRM) have been used to detect and study the local power oscillations. The decay ratio (DR) and the natural frequency (NF) of the power oscillations (typical parameters used to evaluate the instabilities) have been used in the analysis. The results are discussed also making use of two-dimensional plots depicting relative core power distribution during the transient, in order to clearly illustrate the out-of-phase behavior.     

On certain aspects of pressure transient without scram in a BWR

This paper examines the effects of gamma and neutron heating and pressure wave propagation on the core response during an instantaneous loss of condenser vacuum transient without scram (ATWS) in a BWR. By incorporating the gamma and neutron heating, which contribute about 3% of the total power to the moderator, into the transient thermal-hydraulic analysis, the peak power was found to be 35% lower compared with the case with no direct heating. The incorporation of the two-dimensional radial and axial variations of reactivity feedbacks and moderator density into the transient analysis led to a lower power prediction than the one-dimensional model. The pressure surge was examined by a computer program based on the method of characteristics. The pressure rise calculated by this new code was found to be in good agreement with experimental data, while the results of similar calculations done by computer code RELAP4, which is based on finite differencing of the flow equations, were 50% lower.     

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.     

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.     

Similarity study of ROSA-III and fist large break cuonterpart tests to BWR large break LOCA

A large break test in a recirculation pump suction line with the assumption of LPCI-diesel generator failure was conducted at the ROSA-III test facility of Japan Atomic Energy Research Institute. A counterpart test was also performed at the FIST test facility of General Electric Company. The objective of the tests was to develop common understanding and interpretation of the controlling thermal-hydraulic phenomena during a large break LOCA of a BWR. The fundamental thermal-hydraulic phenomena in the ROSA-III and FIST tests such as the system pressure, mixture level and fuel rod surface temperatures agreed well. The FIST test had more bundle uncovery than that in ROSA-III since lower plenum steam in the FIST test flowed out of the jet pumps when they uncovered allowing more liquid to drain from the bundle. The ROSA-III and FIST tests and a BWR counterpart were analyzed with the RELAP5/MODI (cycle 018) code. The similarity of the ROSA-III and FIST large break tests to a BWR large break LOCA has been confirmed through comparison of calculated results though they are slightly different in details. It is perhaps desirable to reexamine the DNB and interphase drag correlations and the jet pump models usedin the code.     

Investigation of BWR LOCA at ROSA-III — Effect of break configuration on system transients

The ROSA-III test facility is a volumetrically scaled ( ) BWR/6 system with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA).Six loss-of-coolant experiments with a break area of 15%, 50% or 200% at the main recirculation pump inlet line were conducted at the ROSA-III test facility with a high pressure core spray failure. A sharp-edged orifice or a long throat nozzle was used as a break plane. It was found in the experiments that the break flow differences between the orifice and the nozzle break configurations with the same flow area were observed only in the subcooled break flow region. Subcooled break flow rate through the orifice was much larger than that through the nozzle. The break configuration difference had little influence on the other system responses, especially on the peak cladding temperature. The applicability of the test results to a BWR/6 has been confirmed through analyses of the 15% break ROSA-III LOCA experiments and BWR/6 LOCAs by using RELAP4/MOD6/U4/J3 code. The experimental results of the ROSA-III LOCA experiments were calculated well by the code, and the same trends were calculated in the BWR analyses.     

Development of a Transient Boiling Transition Analysis Method Based on a Film Flow Model

A new single-channel, transient boiling transition (BT) prediction method based on a film flow model has been developed for a core thermal-hydraulic code. This method could predict onset and location of dryout and rewetting under transient conditions mechanically based on the dryout criterion and with consideration of the spacer effect. The developed method was applied to analysis of steady-state and transient BT experiments using BWR fuel bundle mockups for verification. Comparisons between calculated results and experimental data showed that the developed method tended to predict occurrence of rewetting earlier, however, onset time of BT and maximum rod surface temperature were well predicted within 0.6 s and 20°C, respectively. Moreover, it was confirmed that consideration of the spacer effect on liquid film flow rate on the rod surface was required to predict dryout phenomena accurately under transient conditions.     

钠冷快堆棒状燃料堆芯子通道分析程序开发及验证

热工水力分析软件的验证是安全审查重点关注的问题。为了实现不同设计软件间的对比验证,本工作开发出具有自主知识产权的钠冷快堆堆芯子通道分析程序SSCFR,进行中国实验快堆（CEFR）全堆芯稳态分析、子通道稳态分析及全堆芯瞬态分析,并将分析结果与CEFR运行和设计值进行对比。结果表明,SSCFR程序的计算结果与CEFR运行值及安全分析报告中的设计计算值符合较好,可用于钠冷快堆后续的软件对比验证及设计计算工作。     

RELAP5/FLUENT耦合程序的开发

热工水力数值模拟是反应堆系统设计和安全分析的重要内容,以RELAP5为代表的系统程序可对瞬态或事故工况进行快速分析,同时以FLUENT为代表的计算流体动力学(CFD)程序对堆芯局部三维现象的分析也越来越重要。为综合利用两者的优点,以RELAP5/FLUENT为基础,利用对RELAP5程序源代码的二次开发和FLUENT的用户自定义函数(UDF)进行编程,开发了RELAP5/FLUENT耦合程序。利用flibe熔盐在水平圆管流动问题验证了程序耦合的正确性;针对2 MW熔盐堆进行了稳态模拟,耦合程序能详细分析熔盐堆的热工水力行为;模拟了2 MW熔盐堆功率突变的瞬态热工水力行为,相对于单独的RELAP5,耦合程序能更好地揭示熔盐堆系统和堆芯的三维物理现象。该耦合程序可用于解决熔盐堆热工水力分析中存在的显著三维混合现象的问题。     

Spray Cooling Test

A steady separate effects test on BWR spray cooling was performed at relatively high system pressures using the ROSA-III test vessel. These tests were conducted in order to promote a better understanding of the thermal-hydraulic phenomena in LOCA experiments and to obtain information necessary for improvement of analytical codes. The fraction of entrainment or overflow for various spray conditions was obtained and the data of CCFL at the upper tie-plate were compared with correlations. It was shown that the occurrence of CCFL significantly diminished core cooling effects and that rod quench by fall back water was quite irregular and unstable. Reflood core cooling was also studied.     

Simulation tests of BWR non-loca transients using the tbl

In the last decade, a large number of experiments have been performed in order to understand the thermal-hydraulic response in a boiling water reactor (BWR) under postulated loss of coolant accident (LOCA) conditions. These experimental results showed that the core cooling effect under the LOCA conditions was significantly affected by three-dimensional and multi-bundle phenomena after emergency core cooling systems (ECCSs) started. Also, the peak cladding temperature (PCT) during the LOCA was kept below a specific value of the licensing acceptance criteria, 1473 K (1200°C). These key results of the experiments were incorporated into a computer code, SAFER, which was developed for the BWR LOCA/ECCS analyses under the cooperative studies of Hitachi Ltd, Toshiba Co., and General Electric Co. (GE).

In a couple of years, the experimental study of multi-bundle phenomena was extended into the area of off-normal and non-LOCA transients. Thermal-hydraulic responses during boiling transition were studied using the TBL (Two Bundle Loop) test facility with two full-length bundles. The experimental results showed that interaction and feedback effects between the bundles were expected to be unaffected by core cooling during the typical off-normal and non-LOCA transients. Also, the SAFER showed good predictions for hydraulic responses in the bundles and temperature transients of the rod surfaces.     

Assessment study of the coupled code RELAP5/PARCS against the Peach Bottom BWR turbine trip test

The modeling of complex transients in nuclear power plants (NPP) remains a challenging topic for best estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used for simulating transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal–hydraulics. In this framework, the Peach Bottom BWR turbine trip experiment 2 is considered. The test involves a rapid positive reactivity addition into the core generated by a water hammer load. To perform a numerical simulation of such phenomenon a reference case was calculated using the coupled code RELAP5/PARCS. An overall data comparison shows good agreement between calculated and measured pressure wave trend in the core region. However, the predicted power response during the excursion phase did not match correctly the experimental tendency. For this purpose, a series of sensitivity analyses have been carried out to identify the most probable reasons of such discrepancy. It was found out that the uncertainties related to the cross-sections modeling and to the thermal–hydraulic closure relationships are the main source of the incorrect power feedback response during the transient.