风险指引的CPR1000核电厂LBLOCA分析方法初步研究

随着核电厂安全分析方法的不断发展,结合传统确定论分析与概率风险评价（PSA）的风险指引型安全分析方法逐渐引起安审当局和核电业主的广泛关注。本文基于国际上风险指引型分析方法在其他领域的应用现状,提出了风险指引的大破口失水事故（LBLOCA）分析方法,并重新评估了CPR1000核电厂的堆芯燃料包壳峰值温度（PCT）裕量。在PSA分析中,识别并量化了LBLOCA发生后可能发生的162个事件序列,并采用确定论现实分析方法（DRM）对筛选出的18个概率较大的事件序列进行了计算分析。然后通过期望值评估法和特定序列覆盖法对LBLOCA的PCT裕量进行了评估。结果表明,本文方法下LBLOCA的PCT裕量约为36~55 ℃,相比于传统的DRM裕量提升了16~35 ℃。     

Flow analyses using RELAP5/MOD3 code for OPR1000 under the external reactor vessel cooling

External reactor vessel cooling (ERVC) is considered as one of the most promising severe accident management strategies for an in-vessel corium retention (IVR). Heat removal capacity and water availability at the vessel outer surface can be key factors determining the success of ERVC measures. In this study, for the investigations on the effect of water availability in case of ERVC, flow analyses using the RELAP5/MOD3 code were performed. The analyses were focused to examine the flow behavior inside the reactor pressure vessel (RPV) insulator of the OPR1000 (Optimized Power Reactor 1000 MWe) under a cavity flooding. The current flow analyses results show that for the accident scenarios of station black out (SBO) and 9.6 in. large break loss of coolant accident (LBLOCA) under the ERVC, steam could not ventilate through the insulator and the pressure inside the RPV insulator increased abruptly. This induced a water sweep out and steam domination in the flow path inside the insulator. These flow analyses results indicate that sufficient water ingression and steam venting through the insulator can be a key factor determining the success of the ERVC in the operating nuclear power plant, OPR1000. According to the results of the sensitivity studies for the venting area, in terms of an effective flow circulation inside the insulator, an optimal venting is to assign four holes having a diameter of 0.3 m at the upper exit (hot leg level) of the insulator. And the effect of the inlet flow area at the insulator bottom is rather minor when compared to that of the outlet flow area of a steam venting.     

Structured treatment of model uncertainty in complex thermal-hydraulics codes: Technical challenges, prospective and characterization

Code structure uncertainties (model uncertainty) are a crucial source of uncertainty quantification for thermal-hydraulics (TH) system codes, an assembly of models and correlations that simulate physical phenomena and the behavior of system. Technical challenges dealing with the subject are discussed in this paper with the prospective of TH codes model uncertainty characterization. A literature review was conducted on the subject matter to evaluate the state of the art on the topic. A key characteristic of thermal-hydraulics systems codes is complexity. This complexity has its roots in the composite structure of these systems that comprise of many different elements (or sub-systems) whose state inevitably affects the state of the whole system. Its main implication is the dynamic (i.e., many interdependent variables in time) and/or non-linear behavior. Several different situations are met dealing with the TH model uncertainty. In some cases there are alternative sub-models, or several different correlations for calculating a specific phenomenon of interest. There are also “user options” for choosing one of several models, or correlations in performing a specific code computation. Dynamic characteristics of TH add more complexity to the code calculation, meaning, for example, that specific code models and correlations invoked are sequence-dependent, and require certain (dynamic) conditions. This paper discusses the techniques developed in the Integrated Methodology for Thermal-Hydraulics Uncertainty Analysis (IMTHUA), specifically for the treatment of uncertainties due to code structure and models. The methodology comprehensively covers various aspects of complex code uncertainty assessments for important accident transients. It explicitly examines the TH code structural uncertainties by treating internal sub-model uncertainties and by propagating such model uncertainties in the code calculations, including uncertainties about input parameters. Structural uncertainty assessment (model uncertainty) for a single model will be discussed in terms of “correction factor,” “bias,” and Bayesian sub-model output updating with available experimental evidence. In case of multiple alternative models, several techniques, including dynamic model switching, user-controlled model selection, and model mixing, are discussed. Examples from different applications, including Marviken Blowdown, LOFT LBLOCA, and typical PWR LOCA scenario calculations, are provided for greater clarification of the proposed techniques.     

Safety analysis of LBLOCA in BDBA scenarios for the Bushehr's VVER-1000 nuclear power plant

This paper presents the results of thermal-hydraulic calculations of a large break loss of coolant accident (LBLOCA) analysis for a VVER-1000/V446 unit at Bushehr nuclear power plant (BNPP). LBLOCA is analysis in two different beyond design basis accident (BDBA) scenarios using the RELAP5/MOD3.2 best estimate code. The scenarios are LBLOCA with station blackout (SBO) and LBLOCA with pump re-circulation blockage which have been evaluated in the final safety analysis report (FSAR) of BNPP. A model of VVER-1000 reactor based on Unit 1 of BNPP has been developed for the RELAP5/MOD3.2 thermal-hydraulics code consists of 4-loop primary and secondary systems with all their relevant sub-systems important to safety analysis. The analysis is performed without regard for operator's actions on accident management. The safety analysis is carried out and the results are checked against the acceptance criteria which are the possibility of using water inventory in the emergency core cooling system (ECCS) accumulators and the KWU tanks for core cooling and the available time to operators before the maximum design limit of fuel rod cladding damage is reached. These kinds of analyses are performed to provide the response of monitored plant parameters to identify symptoms available to the operators, timing of the loss of critical safety functions and timing of operator actions to avoid the loss of critical safety functions of core damage. The results of performed analyses show that the operators have 2.9 and 3.1 h for LBLOCA with SBO and LBLOCA with pump re-circulation blockage scenarios, respectively, before the fuel rod cladding rupture. The results are also compared with the BNPP FSAR data.     

Assessment of the MARS code for design of water pool-type reactor cavity cooling system in HTGR

This paper reports an experimental and numerical study on the assessment of the MARS code as a tool for analyzing the water pool-type reactor cavity cooling system (RCCS), which was developed by Seoul National University (SNU). A series of experiments were performed to determine the heat removal capability of the proposed RCCS and assess the capability of MARS code to predict the forced convective, natural convective and radiative heat transfer under normal operation conditions and boiling heat transfer during accident conditions in the RCCS. In the loss of forced convection (LOFC) accident experiment performed at the integral effect test facility called RCCS-SNU, the MARS code underestimated the vapor generation rate at the inner wall of the water pool. Therefore, the newly developed models of the bubble departure and lift-off diameters were implemented into the MARS code to make a better prediction of the vapor generation rate. The improved MARS code was assessed again using the experimental data of the LOFC accident conditions in the RCCS-SNU facility.     

基于PSA的压水堆LBLOCA不确定性分析

为了结合确定论与概率论分析开展更加真实的核反应堆事故工况安全分析,提出了一种结合概率安全分析（PSA）和最佳估算加不确定性（BEPU）分析的方法,并以典型三环路压水堆冷管段双端断裂大破口失水事故（LBLOCA）的极限事故为对象,首先基于PSA开展了应急堆芯冷却系统的事故失效分析,而后结合BEPU分析评估了事件树中各事故序列的包壳峰值温度（PCT）分布及条件堆芯损坏概率（CCDP）,最终确定了压水堆在该事故工况中的堆芯损坏频率（CDF）。分析结果表明,压水堆在冷管段双端断裂工况中应急堆芯冷却系统能够保证反应堆的安全,且一列低压安注系统足以排出堆芯余热及保证反应堆安全。      

Design and transient analyses of emergency passive residual heat removal system of CPR1000

The steam generator secondary emergency passive residual heat removal system (EPRHRs) is a new design for traditional generation II + reactor CPR1000. The EPRHRs is designed to improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the station blackout or loss of heat sink accident. The EPRHRs consists of steam generator (SG), heat exchanger (HX), emergency makeup tank (EMT), cooling water tank (CWT), and corresponding pipes and valves. In order to improve the safety and reliability of CPR1000, the model of the primary loop and the EPRHRs was developed to investigate residual heat removal capability of the EPRHRs and the transient characteristics of the primary loop affected by the EPRHRs using RELAP5/MOD3.4. The transient characteristics of the primary loop and the EPRHRs were calculated in the event of station blackout accident. Sensitivity studies of the EPRHRs were also conducted to investigate the response of the primary loop and the EPRHRs on the main parameters of the EPRHRs. The EPRHRs could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRs could take away the decay heat from the primary loop effectively, and that the single-phase and two-phase natural circulations were established in the primary loop and EPRHRs loop, respectively. The results also indicated that the effect of isolation valve open time on the transient characteristics of the primary loop was little. However, the effect of isolation valve open time on the EPRHRs condensate flow was relatively greater. The isolation valves should not be opened too rapidly during the isolation valve opening process, and the isolation valve opening time should be greater than 10 s, which could avoid the steam impact on the EPRHRs, and improve the stability of the system.     

Development of an uncertainty quantification method of the best estimate large LOCA analysis

A statistical uncertainty quantification methodology for evaluation of the emergency core cooling system (ECCS) performance is proposed and applied in assessing the best-estimate peak cladding temperature (PCT). In the proposed methodology, the Latin hypercube sampling method is adopted, and separate model uncertainties are used as input variables. The independency of the input variables is verified through a correlation coefficient test for statistical treatment of their uncertainties. Next, the PCT response distribution is determined through a goodness-of-fit test. Finally, the PCT with a one-sided 95% probability and a confidence level of 0.95 is estimated.This methodology is applied to the large-break loss-of-coolant accident (LBLOCA) of Kori Nuclear Units 3 and 4. This study shows that the proposed methodology is a useful one.     

Modeling of a confinement bypass accident with CONSEN,a fast-running code for safety analyses in fusion reactors

The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code.     

Creep rupture assessment for Level-2 PSA of a 2-loop PWR: accounting for phenomenological uncertainties

The Level-2 probabilistic safety assessment (PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences.The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA.A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach.This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR,which is the representative case for high pressure sequences.MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident.In addition,a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty.The probability for steam generator tube creep rupture is estimated at 0.17.     

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.     

An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the ‘ASME PRA standard’ also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using the MARS platform. The typical example case is a small break loss of coolant accident without the high pressure safety injection system, which is one of the most important accident sequences in the PSA of the KSNP. Under the given accident sequence, the operator has to perform a recovery action known as a fast cooldown operation. This study focuses on two aspects regarding an operator's action; one is how they can operate it under some restrictions; the other is how much time is available to mitigate this accident sequence. To assess these aspects, this study considered: (1) the operator's action model and (2) the starting time of the operation. To show an effect due to an operator's action, three kinds of control models (the best-fitting, the conservative, and the proportional-integral) have been assessed. This study shows that the developed method and the platform are useful tools for this type of problem and they can provide a valuable insight related to an operator's actions.     

Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m² using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX 11 calculation results.     

基于先进程序+保守评价模型的300MW压水堆核电站大破口失水事故分析

大破口失水事故(LBLOCA)是决定核电站运行功率的设计基准事故之一,本文利用最佳估算系统分析程序RELAP5/MOD3,通过修改其相关模型或关系式,结合有关分离效应与整体效应试验数据验证,形成满足10CFR50附录K中保守评价模型要求的LOCA分析工具——先进程序+保守评价模型程序及分析方法。在此工具与方法开发基础上,对300MW压水堆核电站进行了一回路冷管段双端剪切断裂LBLOCA计算分析,计算的包壳峰值温度(PCT)与应急堆芯冷却系统(ECCS)验收准则及相应最终安全分析报告对比表明：应用该工具与分析方法,可望获得进一步的PCT裕量。     

CPR1000核电厂LBLOCA裕量提升研究

为进一步提升核电厂安全性,核电厂用户提出了15%安全裕量的要求。为提升CPR1000核电厂大破口失水事故(LBLOCA)安全裕量,从改动最小、收益最大的角度出发,提出了两种改进措施:增加安注箱水体积和采用热棒统计分析方法(HRSM)。利用CATHARE程序,对安注箱水体积增加进行敏感性计算,以得到水体积增加量的最优值;热棒统计法中,对3个很大程度上影响计算结果的输入参数进行了抽样,并对计算结果进行统计分析,得到95%95%值(95%置信度下95%概率值)。结果表明,在采用上述两种改进措施后,CPR1000核电厂的LBLOCA安全裕量提升了6.5%。     

CTOF measurements and Monte Carlo analyses of neutron spectra for the backward direction from an iron target irradiated with 400–1200 MeV protons

A calorimetric-time-of-flight (CTOF) technique was used for real-time, high-precision measurement of the neutron spectrum at an angle of 175° from the initial proton beam direction, which hits a face plane of a cylindrical iron target of 20 cm in diameter and 25 cm thick. A comparison was performed between the neutron spectra predicted by the MARS and the MCNPX codes and that measured for 400, 600, 800, 1000 and 1200 MeV protons.     

Manufacturing and testing of W/Cu mono-block small scale mock-up for EAST by HIP and HRP technologies

ITER-like W/Cu mono-block plasma-facing components (PFCs) will be used in vertical target regions of the experimental advanced superconducting tokamak (EAST) divertor. The first W/Cu mono-block small scale mock-up with five W mono-blocks has been manufactured successfully by technological combination of hot isostatic pressing (HIP) and hot radial pressing (HRP). The joining of a W mono-block and a pure copper interlayer was achieved by means of HIP technology and the bonding strength was over 150 MPa. The good bonding between the pure copper interlayer and a CuCrZr cooling tube was obtained by means of HRP technology. In order to understand deeply the process of HRP, the stress distribution of the mock-up during HRP process was simulated using ANSYS code. Ultrasonic Nondestructive Testing (NDT) of the W/Cu and Cu/CuCrZr interfaces was performed, showing that excellent bonding of the W/Cu and Cu/CuCrZr interfaces. The thermal cycle fatigue testing of the mock-up has been carried out by means of an e-beam device in Southwest Institute of Physics, Chengdu (SWIP) and the mock-up withstood 1000 cycles of heat loads up to 8.4 MW/m² with the cooling water of 2 m/s, 20 °C, 0.2 MPa.     

R&D activities of the liquid breeder blanket in Korea

A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5–1.0 MW/m². Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and circulator, a 9 MPa He loop was constructed, and it supplies high temperature (500 °C) and pressure (8 MPa) He to the high heat flux test facility. For an electromagnetic (EM) pump development for circulating the liquid breeder, magnetohydrodynamic (MHD) experiment, and flow corrosion test, a PbLi breeder loop was constructed. From the performance test, the EM pump and magnet show their capability, and flow and static corrosion tests including oxide coating for corrosion protection were performed. For tritium extraction from the liquid breeder, a gas–liquid contact method was adopted and a tritium extraction chamber was constructed. For measurement of the tritium amount in the liquid breeder, permeation sensors have been developed.     

Numerical analysis of RBHT reflood experiments using MARS 1D and 3D modules

The Rod Bundle Heat Transfer (RBHT) program was performed experimentally to analyze the reflood heat transfer phenomena under the conditions of reflood phase following a hypothesized loss of coolant accident (LOCA) by the team of Penn State University. In order to verify the experimental data using a numerical analysis, the Multi-dimensional Analysis of Reactor Safety (MARS) assessment of the RBHT experimental data was carried out for the flooding rates of 0.0254 and 0.1524 m/sec with the upper plenum pressure of 276 kPa. The RBHT experimental data of Tests 1285 and 1383 were compared with the calculation results of the MARS 1D and 3D modules. The MARS code shows a good agreement in the general trend of the peak cladding temperatures although there are limitations in predicting accurate quenching time for both modules. However, in comparison to the MARS 1D module simulation, the MARS 3D module shows the improved calculation capability in that the code can capture local enhanced heat transfer with implication of spacer grids. Moreover, the temperature profiles simulated by the 3D module show the accurate prediction at which the local peak temperatures occur. For more enhanced simulations, local flow parameters such as cross flow and vortex flow need to be analyzed for a more accurate prediction of quenching behavior.     

Shut-down margin study for the next generation VVER-1000 reactor including 13 × 13 hexagonal annular assemblies

Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 × 13 assemblies are calculated as the main aim of the present research. We have applied the MCNP-5 code for many cases with different values of core burn up at various core temperatures, and therefore their corresponding coolant densities and boric acid concentrations. There is a substantial drop in SDM in the case of annular fuel for the same power level. Specifically, SDM for our proposed VVER-1000 annular pins is calculated when the average fuel burn up values at the BOC, MOC, and EOC are 0.531, 11.5, and 43 MW-days/kg-U, respectively.