基于RELAP5的中国氦冷固态包层真空室外破口瞬态特性分析

利用RELAP5/MOD3.4对中国氦冷固态包层、氦气冷却剂回路和二次侧水冷系统进行建模和系统热工水力安全评价。依据ITER事故分析制定的事故序列,对设计基准真空室外破口进行了瞬态分析,并对比了不同破口位置、面积和停堆方式对第一壁的影响。结果表明：真空室外破口发生在风机的下游较上游危险,且小破口较大破口更危险;若真空室外破口同时包层第一壁破口,也可通过自然循环和辐射换热带走衰变热冷却包层;真空室外破口事故中采用聚变停堆系统的3s停堆方式,可避免第一壁熔化。     

Thermal–hydraulic and safety analysis for Chinese helium-cooled solid breeder TBM cooling system

The thermal–hydraulic behavior and safety performance of the Chinese helium-cooled solid breeder (CH HCSB) test blanket module (TBM) with helium cooling system (HCS) has been studied using RELAP5/Mod3.4 code. According to accident analysis specification for TBM, two design basis accidents including loss of off-site power and TBM first wall (FW) ex-vessel coolant pipe break are investigated. The influences of different break locations and plasma termination behaviors are analyzed comprehensively. The results show that natural circulation is established in helium cooling circuit and the TBM can be cooled effectively after loss of off-site power. It is much more critical when the pipe break occurs at the downstream side of the circulator compared with that of upstream side of the circulator. In case of a more serious accident that the ex-vessel break extends to the TBM FW, the results reveal that TBM could be cooled down by natural circulation and radiation. In addition, at the beginning of ex-vessel loss of coolant accident (LOCA), large temperature difference between break and intact TBM FW pipes is found. The accidental results finally show that the integrity of the FW can be guaranteed if the plasma is terminated with a 3 s delay time by fusion power shutdown system (FPSS) in the case of ex-vessel LOCA.     

LOCA analysis for Korean helium cooled solid breeder TBM

One of the major ITER goals is test blanket module (TBM) program which is for the demonstration of the breeding capability that would lead to tritium self-sufficiency in a reactor and the extraction of high-grade heat suitable for electricity generation under the ITER fusion environment. While the engineering design of Korean helium cooled solid breeder (HCSB) TBM and its ancillary systems has been performed, a safety assessment on different possible accident scenarios should be carried out for the purpose of licensing. In this paper, accident analyses for several loss of coolant accident (LOCA) cases were performed in order to assess safety aspects of the TBM design using RELAP5/MOD3.2. Since the TBM forms a loop with helium cooling system (HCS) which is one of ancillary systems required for removing heat deposited in the TBM by neutron wall loading and surface heat flux from plasma, it is necessary to model the complete loop for accident analysis. In this study, the helium passage including the TBM and HCS was nodalized for each accident scenario. The TBM and HCS components were modeled as the associated heat structures provided by RELAP5 to include heat transfer across solid boundaries. Based on computational results it was found that current design of the TBM is robust from the safety point of view.     

Development of TSACO and application to Chinese HCCB TBM cooling system

In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system has been developed using Fortran 90 language, and the simulation has been performed for the cooling system of the Chinese helium cooled ceramic breeder test blanket module (CH HCCB TBM). The semi-implicit finite difference technique was adopted for the solution of the dynamic behavior of helium cooling system. Furthermore, a detailed illustration of the numerical solution for heat structures and critical model was presented. The code was verified by the comparison of RELAP5 code with the same initial condition, boundary condition, heat transfer and flow friction models. The TBM inlet/outlet temperatures and pressure drop were obtained and the results simulated by TSACO were shown in good agreement with those by RELAP5. Thereafter, the design basis accident in-vessel loss of coolant accident (LOCA), was investigated for the CH HCCB TBM cooling system. The critical flow model was also verified by comparing with RELAP5 code. The results indicated that the TBM can be cooled down effectively. The vacuum vessel (VV) pressure and the mass of helium spilled into the VV maintained below the design limits with a large margin.     

压水堆一回路卸压措施影响因素研究

研究压水堆一回路管道小小破口失水事故叠加辅助给水失效导致的高压堆芯熔化严重事故进程，对比验证不同严重事故缓解措施入口温度条件下一回路卸压缓解途径的充分性和有效性，并确认较佳的一回路冷却系统（RCS）降压途径。结果显示，以低于650℃的温度作为降压缓解措施入口条件，可及时恢复可能的堆芯冷却能力。一、二回路卸压效果分析表明，考虑了长期衰变热移出注水流量和堆芯过冷度要求，较佳的卸压配置为初期打开一列稳压器卸压阀，同时迅速恢复辅助给水并开启蒸汽发生器卸压阀。      

Analysis of the reference accidental sequence for safety assessment of LLCB TBM system

Safety analysis of the reference accidental sequence has been carried out for Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) system; India's prototype of DEMO blanket concept for testing in International Thermonuclear Experimental Reactor (ITER). The accidental event analyzed starts with a Postulated Initiating Event (PIE) of ex-vessel loss of first wall helium coolant due to guillotine rupture of coolant pipe with simultaneous assumed failure of plasma shutdown system. Three different variants of the sequences analyzed include simultaneous additional failures of TBM and ITER first wall, failure of TBM box resulting in to spilling of lead lithium liquid metal in to vacuum vessel and reactor trip on Loss of Coolant Accident (LOCA) signal from TBM system. The analysis address specific reactor safety concerns, such as pressurization of confinement buildings, vacuum vessel pressurization, release of activated products and tritium during these accidental events and hydrogen production from chemical reactions between lead–lithium liquid metal and beryllium with water. An in-house customized computer code is developed and through these deterministic safety analyses the prescribed safety limits are shown to be well within limits for Indian LLCB-TBM design and it also meets overall safety goal for ITER. This paper reports transient analysis results of the safety assessment.     

Experimental Study on Heat Removal during Cold Leg Small Break LOCAs in PWRs

Experiments which simulated small break loss-of-coolant accidents (SBLOCAs) resulting from 2.1–0.13% break in the cold leg of a PWR were conducted with an apparatus of 1/270 scale in volume. In the large break size case, the decay heat was mainly removed by the break flow and in the case of a small break, the steam generator played an important role. In this case, thermal hydraulic behaviors such as natural circulation and reflux condensation cooling were important during the transient. Depressurization in the secondary system due to bleeding steam from the steam generator by an operator action was so effective to make the accident to come to an end. The operation to depressurize the secondary system was also efficient to rewet the core which had been uncovered due to a loop seal formation in a cross-over leg.

No effects of initial 200 ppm dissolved gas in the coolant were observed on the cooling performance of the steam generator. It was considered that it was because the gas which came from the coolant into the steam during the depressurization transient did not remain in the tubes of the steam generator.     

AC600非能动安全壳冷却系统冷凝传热系数评价

用ＡＣ６００非能动安全壳冷却系统三维热工水力分析程序ＰＣＣＳＡＣ－ＭＤ,对几种常用的冷凝传热系数结构关系式进行了比较。这些结构关系式包括Ｕｃｈｉｄａ关系式,Ｇｉｄｏ－Ｋｏｅｓｔｌ关系式,Ｔａｇａｍｉ关系式和基于传热传质相似原理的关系式。     

基于SimCont的AP1000非能动安全壳冷却系统仿真实现

应用SimCont程序对AP1000核电厂非能动安全壳冷却系统进行仿真建模,并以主蒸汽管道破裂、冷却剂环路冷管段双端断裂等最严重的设计基准事故为研究对象,仿真分析非能动安全壳冷却系统的响应和主要设备的功能实现,对仿真程序进行综合评价。结果表明：SimCont程序仿真模型能很好地反映出非能动安全壳冷却系统的功能,计算结果与安全分析报告基本吻合。     

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER.     

Analyses for Passive Safety of Fusion Reactor during Ex-Vessel Loss of Coolant Accident

Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down a t about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time Period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident.     

非能动堆芯冷却系统LOCA下冷却能力分析

本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统（PXS）中非能动余热排出系统（PRHRS）、堆芯补水箱（CMT）、安注箱（ACC）、自动卸压系统（ADS）和安全壳内置换料水箱（IRWST）等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1 477 K,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。     

Thermal-hydraulic analysis on the effect of operator action in the loss of shutdown cooling system accident in a PWR

It is necessary to develop PSA methodology and integrated accident management technology during low power/shutdown operations. To develop this technology, thermal-hydraulic analysis is necessarily required to access the trend of plant process parameters and operator's grace time after initiation of the accident. In this study, the thermal-hydraulic behavior in the loss of shutdown cooling system accident during low power/shutdown operations at the Korean standard nuclear power plant was analyzed using the best-estimate thermal-hydraulic analysis code, MARS2.1. The effects of operator's action and initiation of accident mitigation system, such as safety injection and gravity feed on mitigation of the accident during shutdown operations are also analyzed.When steam generators are unavailable or vent paths with large cross-sectional area are open in the accident, the core damage occurs earlier than the cases of steam generators available or vent paths with small cross-sectional area. If an operator takes an action to mitigate the accident, the accident can be mitigated considerably. To mitigate the accident, high-pressure safety injection is more effective in POS4B and gravity feed is more effective in POS5. The results of this study can contribute to the plant safety improvement because those can provide the time for an operator to take an action to mitigate the accident by providing quantitative time of core damage. The results of this study can also provide information in developing operating procedure and accident management technology.     

Loss of residual heat removal during mid-loop operation: BETHSY experiments

Loss of residual heat removal has occurred in several PWRs during mid-loop operation after plant shutdown, and is now recognized as an accident situation whose relevant physical phenomena require improved understanding. Several cases have been selected for investigation in the frame of the BETHSY program, with the main purpose of providing an experimental basis for the assessment of safety codes which have been so far especially involved with accident transients initiated during full-power operation.The major results from four experiments are presented in this paper, which addresses two main kinds of physical problem: mechanisms for, and rate of, primary mass inventory reduction in the case of manways opened at various locations in the primary coolant system; cooling of the plant when the primary is only half-open through vent paths and one steam generator is available to remove decay heat in reflux condenser mode with the presence of non-condensable gas.     

Analyses of Passive Plasma Shutdown during Ex-vessel Loss of Coolant Accident in the First Wall/Shield Blanket of Fusion Reactor

An ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of a fusion reactor has been analyzed by a hybrid code consisting of plasma dynamics and heat transfer analysis of in-vessel components. We investigated possibility of passive plasma shutdown scenario during the accident in International Thermonuclear Experimental Reactor (ITER). The safety analysis code which we developed can treat impurity concentration from the first wall and the divertor with a transport probability into the main plasma and a time delay given as input. It was found that the plasma is passively shutdown by a density limit disruption due to beryllium release from heated first wall surfaces about 168 seconds after the LOCA, when the transport probability of beryllium from the first wall into the plasma and the time delay were assumed to be 10^?2 and the energy confinement time, respectively. At that time, the surface temperature of the outboard center (plasma facing component (PFC) with beryllium) and the temperature of the coolant tube in the first wall (stainless steel 316) reach about 1,120°C and about 1,080°C, respectively. Although the coolant tube does not melt, the copper heat sink between the PFC and the coolant tube melts before the passive shutdown. The heat sink of copper in the outboard baffle also melts before the passive shutdown, though the PFC surface of tungsten does not melt. Consequently, we have a possibility of passive plasma shutdown before the cooling tubes melt during the ex-LOCA of the first walllshield blanket in ITER, however, further studies are needed on the effects on plasma burn control, impurity release and emission of implanted D-T fuel.     

严重事故条件下安全壳响应模拟研究

压水堆核电厂发生严重事故期间,从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象,采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列：热管段中破口叠加设备冷却水失效和再循环高压安注失效,堆芯因冷却不足升温熔化导致压力容器失效,熔融物与混凝土发生反应（MCCI）,安全壳超压失效;冷管段大破口叠加再循环失效,安全壳内蒸汽不断聚集,发生超压失效。通过对两种事故工况的分析,证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。     

Response of gas-cooled fast breeder reactors to depressurization accidents

The gas-cooled fast breeder reactor (GCFR) under design by Gulf General Atomic is cooled with helium pressurized to 85 atm and has the reactor core, the steam generators and their associated steam turbine-driven helium circulators, and auxiliary core cooling loops all contained within a massive prestressed concrete reactor vessel (PCRV).The response of the GCFR to coolant depressurization accidents has been investigated and it has been shown that this class of accidents can be safely handled with considerable safety margin. Rapid depressurization is assumed to be caused by a seal failure in a large concrete plug closing one of the large PCRV cavities and the depressurization rate is controlled by a flow restrictor incorporated within the closure plug. Continued core cooling is provided by the main core cooling loops. The plant transient reponse following a depressurization accident has been calculated with a computer code developed at GGA. The results obtained indicate rather mild increases in peak clad temperature for a depressurization accident with the leak area defined by the flow restrictor.Additional cases investigating larger leak areas to explore safety margins indicate that the peak cladding temperature does not increase rapidly with increasing leak area. Secondary containment conditions in a depressurization accident have also been evaluated.     

ACME台架PRHR管线破口位置敏感性试验研究

为研究先进非能动（AP）型核电厂在非能动系统失效条件下的安全性能,利用我国先进堆芯冷却机理整体试验台架（ACME）开展了非能动余热排出（PRHR）管线破口失水试验研究,分析了主要的试验进程和破口位置对事故过程各阶段关键参数的影响。结果表明,ACME PRHR管线破口试验进程与冷管段小破口失水事故（SBLOCA）进程基本一致,再现了非能动核电厂自然循环阶段、自动卸压系统（ADS）喷放阶段和安全壳内置换料水箱（IRWST）安注阶段的安全特性;在不同破口位置的试验中,非能动堆芯冷却系统（PXS）均可保证堆芯得到补水,堆芯活性区始终处于混合液位以下;破口位置对ACME LOCA事故进程、反应堆冷却剂系统（RCS）初期降压速率、PRHR热交换器（HX）流量、喷放流量、堆芯液位、IRWST安注流量等参数具有显著影响,对堆芯补水箱（CMT）和蓄压安注箱（ACC）安注流量的影响较小。      

ACP600主蒸汽管道破裂事故的应对措施研究

针对ACP600取消高压安注系统和浓硼箱、使用一体化钆为可燃毒物、采用Mode-C运行与控制模式等设计改进导致主蒸汽管道破裂（MSLB）事故安全裕量降低的不利情况,对先进三代核电厂ACP600的MSLB事故进行分析研究。为避免MSLB事故下反应堆重返临界后堆芯功率峰值过高导致偏离泡核沸腾比（DNBR）低于限制值,分别从快速注入硼溶液和减缓堆芯冷却率的角度,评价不同的安注系统配置以及停运故障环路主泵对于缓解MSLB事故的作用。研究最佳的缓解方案,并提出增设“蒸汽管线压力低-3”信号停运故障环路主泵的设计优化建议。