Testing of the multi-application small light water reactor (MASLWR) passive safety systems

Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the multi-application small light water reactor (MASLWR). The MASLWR is a pressurized light water reactor design with a net output of 35 MWe that uses natural circulation in both normal and transient operation. Due to its small size, portability and modularity, the MASLWR design is well suited to help fill the potential need for grid appropriate reactor designs for smaller electricity grids as may be found in developing or remote regions. The purpose of the OSU MASLWR test facility is to assess the operation of the MASLWR under normal full operating pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR test facility. These tests included one design basis accident and one beyond design basis accident. During the performance of these tests, plant operations to include start up, normal operation and shut down evolutions were demonstrated successfully.     

PSB-VVER simulation of Kozloduy NPP “loss of feed water transient”

This paper provides a comparison between the PSB test facility experimental results obtained during the simulation of loss of feed water transient (LOFW) and the calculation results received by INRNE computer model of the same test facility. Integral thermal-hydraulic PSB-VVER test facility located at Electrogorsk Research and Engineering Center on NPPs Safety (EREC) was put in operation in 1998. The structure of the test facility allows experimental studies under steady state, transient and accident conditions.RELAP5/MOD3.2 computer code has been used to simulate the loss of feed water transient in a PSB-VVER model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for simulation of loss of feed water transient.The objective of the experiment “loss of feed water”, which has been performed at PSB-VVER test facility is simulation of Kozloduy NPP LOFW transient. One of the main requirements to the experiment scenario has been to reproduce all main events and phenomena that occurred in Kozloduy NPP during the LOFW transient. Analyzing the PSB-VVER test with a RELAP5/MOD3.2 computer code as a standard problem allows investigating the phenomena included in the VVER code validation matrix as “integral system effects” and ”natural circulation“. For assessment of the RELAP5 capability to predict the “Integral system effect” phenomenon the following RELAP5 quantities are compared with external trends: the primary pressure and the hot and cold leg temperatures. In order to assess the RELAP5 capability to predict the “Natural circulation” phenomenon the hot and cold leg temperatures behavior have been investigated.This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory (ANL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Validation of CATHARE V2.5 thermal-hydraulic code against full-scale PERSEO tests for decay heat removal in LWRs

The PERSEO experimental program was performed in the framework of a domestic research program on innovative safety systems with the purpose to increase the reliability of passive decay heat removal systems implementing in-pool heat exchangers. The conceived system was tested at SIET laboratories by modifying the existing PANTHERS IC-PCC facility utilized in the past for testing a full scale module of the GE-SBWR in-pool heat exchanger. Integral tests and stability tests were conducted to verify the operating principles, the steadiness and the effectiveness of the system. Two of the more representative tests have been analyzed with CATHARE V2.5 for code validation purposes. The paper deals with the comparison of code results against experimental data. The capabilities and the limits of the code in simulating such kind of tests are highlighted. An improvement in the modeling of the large water reserve pool is suggested trying to reduce the discrepancies observed between code results and test measurements.     

Reactor cooling systems thermal-hydraulic assessment of the ASTEC V1.3 code in support of the French IRSN PSA-2 on the 1300 MWe PWRs

The French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe PWRs. This PSA-2 study is relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, a wide-ranging series of comparisons with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe-accident scenarios. The present paper details 4 out of the 14 studied scenarios: a 12-in. cold leg Loss of Coolant Accident (LOCA), a 2-tube Steam Generator Tube Rupture (SGTR), a 12-in. Steam Line Break (SLB) and a total Loss of Feed Water scenario (LFW). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and compared to the CATAHRE 2 V2.5 results. The ASTEC results of the core degradation phase are also presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results.     

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.     

Experimental validation of the TASS/SMR code for an integral type pressurized water reactor

The transient and setpoint simulation small and medium reactor (TASS/SMR) code has been applied to perform the safety analysis and performance evaluation of an integral type pressurized water reactor. Till now, the code has only been verified by using simplified and analytical problems as well as a reliable system code due to the lack of available experimental data. Recently, several kinds of experiments have been performed by focusing on an identification of the heat transfer characteristics at a heat sink and source, and the thermal hydraulic characteristics and the natural circulation performance in an integral effect test facility. In this paper, the TASS/SMR code has been validated by using the experimental data obtained from a separate effect test facility by focusing on the heat transfer characteristics and an integral effect test facility by focusing on the thermal hydraulic characteristics and the natural circulation performance. According to the validation results of the TASS/SMR code against the separate effect test and the integral effect test, the code predicts the overall variation of the thermal hydraulic parameters well, including the system pressure, fluid temperature, mass flow rate, etc., and it is applicable for the safety analysis and performance evaluation of an integral type pressurized water reactor.     

Assessment of the models in RELAP/SCDAPSIM with QUENCH-06 analysis

A fundamental principle of accident management in a nuclear power plant is the injection of water to cool the core. In this framework, a series of QUENCH tests have been conducted at Karlsruhe Institute of Technology (formerly Forschungszentrum Karlsruhe). The test results constitute a significant experimental database not only for further understanding of reflooding behavior, but also for code validation and improvement. The RELAP/SCDAPSIM code is a system code that is used to model reactor behavior and is widely used around the world. To date, assessment and validation have been performed with numerous experiments, including QUENCH tests. In the previous studies, the results of QUENCH simulations were referred to be sensitive to two main parameters: the electrical resistance and the thermal conductivity of the shroud insulator, which are subject to relatively large uncertainty. It is important to investigate these two parameters in detail, because this would enable identification of those SCDAP models that require further improvement. In this study, the uncertainty of the electrical resistance was reduced by modification of the code and subsequent validation with experimental data. In addition, modification of the thermal properties of the shroud insulator is suggested with consideration of the argon atmosphere in the facility. Finally, upcoming problems and questions are discussed. A rather good agreement was obtained than those of previous studies. As a result, more accurate modeling of the electrical resistance and the thermal properties of the shroud insulator was conducted and the importance of these parameters was evaluated.     

A PANDA integral test on the effect of light gas on a Passive Containment Cooling System (PCCS)

As part of the Euratom project TEMPEST (Testing and Enhanced Modelling of Passive Evolutionary Systems Technology for Containment Cooling), a series of five tests was performed in the PANDA facility to experimentally investigate the distribution of hydrogen inside the containment and its impact on the performance of the Passive Containment Cooling System (PCCS) designed for the Economic Simplified Boiling Water Reactor (ESBWR). In a postulated severe accident a large amount of hydrogen could be released in the Reactor Pressure Vessel (RPV) as a consequence of the cladding Metal-Water (M-W) reaction and discharged together with steam to the Drywell (DW) compartment. The retention of hydrogen in the DW, instead to be vented in the Wetwell (WW), has a positive effect toward the mitigation of the system pressure build-up. Hydrogen retention in the DW is a consequence of the stratification phenomena driven by the steam-hydrogen density difference. The paper presents the experimental results of the integral Test T1.2 performed in the PANDA facility. Helium was used to simulate hydrogen and the specific PANDA facility configuration included a dead-end volume, allowing for retaining a portion of the released helium in the DW compartment. The results from Test T1.2 showed that the containment end pressure is mainly determined by the redistribution of non-condensable gas inside the containment system and the temporary deterioration of the PCCS performance during the helium release phase plays a minor role.     

TACR热传输系统热工水力瞬态分析

钍燃料的利用对于缓解核燃料资源短缺具有重要意义,坎杜型反应堆(Canadian Deuterium Uranium,CANDU)在堆芯布置、中子利用效率及先进燃料循环方面具有较高的灵活性,使得其在CANDU反应堆中引入钍燃料循环更具现实意义。CANDU型反应堆中钍基燃料应用关键基础技术研究是加拿大与我国正在开展的合作课题,其中开发自主的CANDU堆堆芯热工水力设计和安全分析程序是钍基燃料应用必不可少的设计工作之一。本文针对CANDU型反应堆热传输系统结构特点,采用FORTRAN程序设计语言开发了适用于CANDU型反应堆热传输系统的热工水力瞬态分析程序CANTHAC(CANDU Thermal-Hydraulic Analysis Code)。利用CANTHAC对钍基先进CANDU堆(Thorium-based Advanced CANDU Reactor,TACR)进行了瞬态分析,计算工况包括满功率稳态、无保护蒸汽发生器(Steam Generator,SG)二次侧给水温度降低事故及完全失流事故。其中,满功率稳态计算结果与清华大学设计的钍基先进CANDU堆TACR设计值吻合较好,相对误差不超过2%,在可接受范围内;无保护SG二次侧给水温度降低事故及完全失流事故在计算条件下所得的燃料温度及系统压力等关键热工水力参数均在安全限值内,满足安全准则要求。程序为模块化编程,便于移植和改进,具有一定的通用性,为进一步研究工作奠定了基础。     

TRAC-PF1/MOD1 calculations and data comparisons for mist small-break LOCA, feed-and-bleed and steam-generator tube-rupture experiments

Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents, loss of feedwater and other transients in Babcock and Wilcox (B&W) nuclear plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 × 4 [two hot legs and steam generators (SGs), four cold legs and reactor coolant pumps] representation of B&W lowered-loop reactor systems. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Efforts are under way at Los Alamos to assess TRAC-PF1/MOD1 against data from the MIST facility.Calculations and data comparisons for TRAC-PF1/MOD1 assessment are presented for three transients run in the MIST facility. The energy removal and depressurization mechanisms in these tests are identified and the phenomena occurring in these tests compared. The tests analyzed are MIST Test 3109AA, the nominal small-break LOCA, Test 330302, a feed and bleed test with delayed high-pressure injection; and Test 3404AA, an SG tube-rupture test with the affected SG isolated. TRAC was able to predict these phenomena although the timing and magnitude of events were not always in good agreement.The MIST test have demonstrated the thermal-hydraulic phenomena expected to occur during transients in B&W nuclear plants. Because of scaling atypicalities, test results cannot be extrapolated directly to plant conditions. Although the phenomena were demonstrated in the MIST tests, there may be differences in the timing, magnitude and sequences of events in plant transients. Assessment calculations, three of which are presented here, have shown that the TRAC computer code can predict the major trends and phenomena occurring during the MIST tests with reasonable qualitative agreement. This includes complex sequences of events. Reasonable qualitative agreement is defined as meaning that major trends are predicted correctly, although TRAC values are frequently outside the range of data uncertainty. These assessment results, taken with assessment results from other facilities at a wide range of scales, provide us with confidence that the TRAC code can adequately simulate the transient phenomena possible in B&W nuclear plants.     

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.     

TMSR-SF2全厂断电事故分析

钍基熔盐堆-固态燃料二号堆(Thorium Molten Salt Reactor-Solid Fuel 2,TMSR-SF2)是基于球床熔盐堆SF1(Solid Fuel 1)的小型模块化升级堆型,这种新概念堆结合了两者的诸多优点,目前已经完成了预概念设计,对其进行典型事故的分析与安全特性的评估成为当前重要研究内容。本文基于Relap5/MOD4.0程序,建立了反应堆事故模型,进行了全厂断电事故的模拟,分析了反应性、反应堆功率、冷却剂温度和燃料温度等关键参数的变化规律。结果表明,SF2在全厂断电事故中具备高度安全性,其中固有安全性发挥了重要作用。此外还进行了全厂断电事故伴生不同事件的后果比对以及不同温度反应性系数的敏感性分析,证明了直接反应堆辅助冷却系统(Direct Reactor Auxiliary Cooling System,DRACS)在事故前期余热排出能力的局限性,而依靠主泵可以最大限度利用熔盐堆的热惰性从而显著缓解熔盐堆堆芯过热。     

Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 °C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 °C). Additionally, the peak drywell temperature of 155.3 °C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 °C, which is below the pool temperature used for evaluating the net positive suction head of pumps of the RHR system and the Emergency Core Cooling Systems (96.7 °C). The peak containment pressure and temperature are well below the design value (386.1 kPaG and 171.1 °C). Containment integrity of Chinshan Plant can be maintained under the SPU condition.     

Development of the analysis tool for the water cooling type passive residual heat removal system of Chinese pressurized reactor

An innovative design for Chinese pressurized reactor is the steam generator (SG) secondary side water cooling passive residual heat removal system (PRHRS). The new design is expected to improve reliability and safety of the Chinese pressurized reactor during the event of feed line break or station blackout (SBO) accident. The new system is comprised of a SG, a cooling water pool, a heat exchanger (HX), an emergency makeup tank (EMT) and corresponding valves and pipes. In order to evaluate the reliability of the water cooling PRHRS, an analysis tool was developed based on the drift flux mixture flow model. The preliminary validation of the analysis tool was made by comparing to the experimental data of ESPRIT facility. Calculation results under both high pressure condition and low pressure condition fitted the experimental data remarkably well. A hypothetical SBO accident was studied by taking the residual power table under SBO accident as the input condition of the analysis tool. The calculation results showed that the EMT could supply the water to the SG shell side successfully during SBO accident. The residual power could be taken away successfully by the two-phase natural circulation established in the water cooling PRHRS loop. Results indicate the analysis tool can be used to study the steady and transient operating characteristics of the water cooling PRHRS during some accidents of the Chinese pressurized reactor. The present work has very important realistic significance to the engineering design and assessment of the water cooling PRHRS for Chinese NPPs.     

Code assessment of ATLAS integral effect test simulating main steam-line break accident of an advanced pressurized water reactor

KAERI has been operating an integral effect test facility, Advanced Thermal–Hydraulic Test Loop for Accident Simulation (ATLAS), for accident simulations of advanced pressurized water reactors. As an integral effect test database for major design basis accidents has been accumulated, a domestic standard problem (DSP) exercise using ATLAS was proposed in order to transfer the database to domestic nuclear industries and to contribute to improving the safety analysis technology for pressurized water reactors (PWRs). As the third DSP exercise, a double-ended guillotine break of the main steam-line at an 8% power without loss of off-site power was decided as a target scenario. Seventeen domestic organizations joined this DSP exercise. They include universities, government, and nuclear industries. The participants of DSP-03 were classified into three groups and each group has focused on the specific subject related to the enhancement of the code assessment; (1) scaling capability of the ATLAS test data by comparing with the code analysis for a prototype, (2) multi-dimensional thermal–hydraulic phenomena anticipated during the steam-line break transient, (3) effect of various models in the one-dimensional safety analysis codes.     

Safety analysis of the IAEA reference research reactor during loss of flow accident using the code MERSAT

Using the thermal hydraulic code MERSAT detailed model including primary and secondary loop was developed for the IAEA's reference research reactor MTR 10 MW. The developed model enables the simulation of expected neutronic and thermal hydraulic phenomena during normal operation, reactivity and loss of flow accidents.Two different loss of flow accident (LOFA) have been simulated using slow and fast decrease time of core mass flow. In both cases the expected flow reversal from downward forced to upward natural circulation has been successfully simulated. The results indicate that in both accidents the limit of onset of subcooled boiling was not arrived and consequently no exceed of design limits in term of thermal hydraulic instability or DNB is observed. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermal hydraulic system codes.     

基于ECC安注热混合试验的LOCUST 1.2分析与验证

在失水事故（LOCA）工况下安注系统投入使用时,蒸汽与安注冷却剂会发生流体热力学混合,热混合过程中冷腿段的冷却是直接影响堆芯再淹没与否的重要因素。中国广核集团有限公司自主研发了一款两相流热工水力系统分析软件LOCUST,可用于压水堆核电厂事故工况的分析计算。基于西安交通大学堆芯应急冷却系统（ECCS-XJTU）试验台架进行的堆芯应急冷却（ECC）安注热混合试验,本文使用LOCUST软件对ECC热混合试验进行了几何建模及计算分析。ECC热混合试验工况主要为不同流量下主管纯蒸汽与安注管过冷水的混合,蒸汽流量为25~125 kg/h,过冷水流量为100~500 kg/h。模拟计算结果和试验结果的对比分析表明：试验段出口质量流量计算值的最大相对误差在13.8%以内,混合后温度计算值的最大相对误差在8%以内,LOCUST在计算高温蒸汽和过冷水混合时的计算结果相对保守,总体上验证了LOCUST在LOCA下两相热混合安注计算的可靠性和准确性。