西安脉冲堆大破口失水事故分析

分析了西安脉冲堆大破口失水事故的特点,建立了适用的数学模型,编制了计算程序。结果表明：在大破口失水事故下,部分燃料芯体最高温度将超过设计限值,但不会发生燃料元件熔毁事故。     

Transient Analysis for Design of Primary Coolant Pump Adopted to JAERI Passive Safety Reactor JPSR

In a course of a design study of the JAERI passive safety pressurized water reactor (JPSR), a complete loss-of-flow transient caused by a trip of all pumps was analyzed with the RETRAN code to determine an inertia of canned-motor pump utilized as the primary coolant pump and to confirm feasibility of the design condition. This transient was selected because the pump had a low inertia rotor inducing fast flow coastdown, and among the transients in which the pump had dominant effect on the departure from nucleate boiling (DNB), the analyzed transient was severest in view of the DNB occurrence. The DNB threshold was related, based on sensitivity calculations, with the coolant density reactivity coefficient and the pump inertia. From the calculations, it was concluded that the pump inertia higher than 250 kg·m² (8% of the ordinary PWRs) was necessary for preventing the DNB occurrence for the present design of JPSR, regardless of the actuation of the reactor scram. The DNB occurrence could be prevented only by the inherent nature of the reactor core which reduced the power by insertion of negative coolant density reactivity during the transient and this was one of major features of JPSR. It was shown by a rough estimation that the necessary condition could be practically realized by incorporation of a cylindrical-type flywheel.     

小型堆破口失水事故初步研究

为验证中国广核集团小型堆方案设计,尤其是其中非能动安全注入系统的初步设计,基于RELAP/SCDAPSIM程序,建立了小型堆的一、二回路系统和非能动安全注入系统模型,模拟计算了冷管段0.04 m等效直径破口、冷管段0.2 m等效直径破口、直接注入管道双端断裂、自动卸压系统误启动等LOCA工况。计算结果表明,一回路可实现有效的冷却和降压,堆芯不会过热,验证了其非能动安全注入系统的设计合理性和反应堆系统的安全性。     

多用途一体化轻水堆初步设计方案和安全分析

为满足未来区域性核能供电、核供热、大规模制氢、海水淡化等需求,迫切需要一种结构简单、固有安全性高、经济性高的多用途反应堆.基于此,一种多用途的一体化轻水堆设计概念被提出,包括不同设备的初步设计方案和参数;根据其特点,利用最佳估算程序RELAP5对其中一个设计方案进行了稳压器汽腔破口事故和主泵断电引起的丧失流量事故的确定论安全分析.结果表明,在保守假设条件下,其固有特性和安全系统仍能保证堆芯始终处于被淹没状态,非能动余热排出系统可有效导出停堆后的长期衰变热,从而为进一步研究一体化轻水堆的设计和运行安全特性打下了基础.     

Accident management following loss-of-coolant accidents during cooldown in a Westinghouse two-loop PWR

Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In preparation for this, the reactor system is cooled down, depressurised and partially drained. Although reactor coolant pressure is lower than during full-power operation, there remains the possibility of a loss-of-coolant accident (LOCA), with a certain but low probability. While the decay heat to be removed is lower than that from a LOCA at full power, the reduced availability of safety systems implies a risk of failing to maintain core cooling, and hence of core damage. This is recognised though probabilistic safety analyses (PSA), which identify low but non-negligible contributions to core damage frequency from accidents during cooldown and shutdown. Analyses are made for a typical two-loop Westinghouse PWR of the consequences of a range of LOCAs during hot and intermediate shutdown, 4 and 5 h after reactor shutdown respectively. The accumulators are isolated, while power to some of the pumped safety injection systems (SIs) is racked out. The study assesses the effectiveness of the nominally assumed SIs in restoring coolant inventory and preventing core damage, and the margin against core damage where their actuation is delayed. The calculations use the engineering-level MELCOR1.8.5 code, supplemented by the SCDAPSIM and SCDAP/RELAP5 codes, which provide a more detailed treatment of coolant system thermal hydraulics and core behaviour. Both treatments show that the core is readily quenched, without damage, by the nominal SI which assumes operation of only one pump. Margins against additional scenario and model uncertainties are assessed by assuming a delay of 900 s (the time needed to actuate the remaining pumps) and a variety of assumptions regarding models and the number of pumps available in conjunction with both MELCOR and versions of SCDAP. Overall, the study provides confidence in the inherent robustness of the plant design with respect to LOCA during cooldown to cold shutdown, and in the validity of a two-tier calculational method. The results have been directly used in updating the plant shutdown PSA, by changing the success criteria for core cooling during cooldown of the plant and showing a reduction in overall risk.     

Boiling Water Reactor Loss of Coolant Tests Single Failure Tests with ROSA-III

The single failure tests with the ROSA-III were simulated BWR LOCA experiments by the scaled BWR test facility resulting from a 200% double-ended break at the recirculation pump suction line to evaluate the core cooling capability of a BWR ECCS under the single failure condition.

The experimental results showed that the loss of LPCS and one LPCI resulted in the highest PCT of 870 K of the single failure series tests, yet a core cooling capability by the ECCS was maintained. The REALP4/Mod 6 code was used to evaluate the predictive capability of the LOCA analysis code. The calculated results showed that the RELAP4/Mod 6 code was able to predict occurrences and sequence of major events anticipated to occur during a BWR LOCA correctly. However it was found that the code still needs to be improved in a CCFL model to better describe thermohydraulic behavior in the core.

The analyses presented in this paper are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a BWR during a LOCA.     

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。     

Investigation of Break Location Effects on Thermal-Hydraulics during Intermediate Break Loss-of-Coolant Accident Experiments at ROSA-III

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25% main recirculation pump suction line break (MRPS-B) experiments, the 21% single-ended jet pump drive line break (JPD-B) experiment and the 15% main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests.

In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop.     

失水事故排放过程中的不稳定性

叙述了低温供热堆发生上空腔小破口失水事故后，自然循环系统的不稳定性，揭示了在排放过程中，由于冷却剂闪蒸现象引起的系统两相流不稳定性，以及在排放不同阶段中流量振荡特性。     

大型非能动压水堆核电厂安全注射系统旋启式止回阀阻力变化对长期堆芯冷却影响分析

大型非能动压水堆核电厂在发生失水事故(LOCA)后的长期堆芯冷却阶段依靠重力向堆芯注入应急冷却水,其注射管线上设置的旋启式止回阀的阻力可随流量变化,管线的阻力可能将非预期地增加。根据旋启式止回阀阻力特性,为失水事故最佳估算系统分析程序添加相应的计算功能,对压力容器直接注射(DVI)管线双端断裂事故后长期堆芯冷却工况进行了计算分析。结果表明:安全注射管线上旋启式止回阀阻力变化对大型非能动压水堆核电厂LOCA后长期冷却的影响较小;在安全裕量不足的情况下,旋启式止回阀的阻力特性将影响到非能动注射管线的安全注射功能的执行。     

自然循环反应堆NHR-200Ⅱ失水事故初步研究

破口事故是压水堆最为关注的一类重要事故,其失水量与事故后果严重程度密切相关。NHR-200Ⅱ是由清华大学核能与新能源技术研究院经过多年研究和不断改进,设计的一种全功率自然循环低温供热反应堆,其设计中采用了多种先进的非能动和固有安全设计。本研究针对NHR-200Ⅱ反应堆,选取后果最为严重的控制棒引水管断裂且无法隔离事故,利用系统热工瞬态分析程序对事故过程进行了模拟和分析。结果表明,即使在最严重的破口失水事故下,NHR-200Ⅱ主回路中剩余的冷却剂始终能覆盖反应堆堆芯,并有效通过非能动余热载出系统带走堆芯热量,从而保证反应堆堆芯不会因裸露造成烧毁,这表明NHR-200Ⅱ具有很好的安全特性。     

Experimental Assessment of Evaluation Model for Safety Analysis on Reflood Phase of PWR-LOCA

A model named Evaluation Model (EM) is used for the safety evaluation analysis on a PWR-LOCA. The model is conservatively constructed to severely evaluate the core cooling during the LOCA.

In order to assess the model for the reflood phase of the LOCA, a test named “EM Test” was conducted with Cylindrical Core Test Facility which modeled a PWR in 1/21 volumetric scale. The test conditions simulated the reflood phase predicted in the safety evaluation analysis. The test results were compared with the blindfold results predicted by EM codes which are programmed based on the EM. The main conclusions are as follows:

(1) The core heat transfer model built in the EM codes gives conservative results.

(2) The system models in the present EM codes predict the flooding rate conservatively in early transient and realistically in later period.

(3) Conservative items and items to be improved are pointed out. The downcomer slow water accumulation observed in a test at lower flow rate of emergency core cooling water did not appear in the EM test.     

安注方式对设计基准失水事故下元件包壳破损份额影响的分析

以某船用压水堆为研究对象,采用RELAP5/MOD32程序,分析了发生在主管道冷端的极限中破口失水事故中,采取冷端、热端安注方式时不同的事故过程。引入临界管概念,确定了包壳破损临界功率因子。对全堆进行精细功率重构,确定每根燃料元件功率因子,最终确定不同安注方式下的元件包壳破损份额,并指出：对破口出现在主管道冷段的设计基准事故,热端安注能减轻事故后果,减少破损份额。     

西安脉冲堆满功率运行工况内部始发事件一级概率安全评价

针对西安脉冲堆(XAPR)2 MW满功率运行工况,建立了内部始发事件一级概率安全评价(PSA)模型,对始发事件识别、事故序列分析及可靠性数据处理等进行了研究。应用小事件树-大故障树方法,在Risk Spectrum平台上完成XAPR堆芯损伤事故序列的定量分析。结果表明,XAPR内部事件导致的堆芯损伤频率(CDF)为4.14×10~(-6)/(堆·年),对CDF贡献最大的为堆水池堆芯高度处大破口失水事故,支配性事故序列是大破口失水事故后紧急排水系统失效。研究结果证明XAPR具有较高的安全性。     

Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs.     

燃料热导率降级对CAP1000大破口失水事故的影响分析

在高燃耗情况下,燃料芯块的热导率随燃耗降低,该现象被称之为热导率降级(TCD)现象。TCD现象影响失水事故(LOCA)前稳态工况的燃料平均温度和燃料储能,进而影响大破口LOCA过程中的包壳峰值温度(PCT)。本研究采用大破口LOCA分析程序WCOBRA/TRAC对CAP1000冷段双端剪切断裂事故进行了不同燃耗的敏感性分析,并获得了不同工况下的PCT。分析中采用美国核燃料研究所(NFI)修正的TCD模型对降级后的燃料热导率进行模拟,同时考虑了燃耗大于30GW·d/tU后FQ和FΔh峰值因子的降低。敏感性分析表明,考虑TCD和峰值因子降低的影响,PCT极限工况不再出现在低燃耗区间,而出现在燃耗为29GW·d/tU附近。与其他燃耗水平相比,该燃耗点的PCT第1峰值和第2峰值均处于最高水平。本研究结果可为高燃耗情况下非能动电厂大破口LOCA的分析评估提供参考。     

Core Safety Characteristics of High Flux Safe Reactor (HFSR)

Abstract

Transient analyses are performed for graphite moderated helium-cooled high flux reactor to obtain the high flux safe reactor design. In order to promote the safety of the high flux reactor, the present design adopts the pebble bed reactor and its fuel technology. In the transient analyses, among the postulated off-normal events and accidents, the reactivity accident followed by a loss of helium forced circulation with system depressurization is found to be the severest potential event which may threaten the reactor safety from fission products release point of view. Several neutronic and thermal-hydraulic design parameters are indicated and exploited to promote the reactor safety. Neutronic and core thermal-hydralic models are proposed and used to simulate the reactor responses to the off-normal events and accidents. As the results of the transient analyses and accident simulation, safe and optimal design parameters are obtained which provide high thermal neutron flux with a desirable spectrum and large usable volume constrained by safety limitations.     

压水堆核电站大破口失水事故分析

压水堆核电站安全分析报告是核安全监管部门对其进行安全审查的重要文件,大破口失水事故是核电站运行的设计基准事故,是安全分析报告中的重要内容。本文使用RELAP5/MOD3.2进行压水堆冷管段大破口失水事故的计算,对比发现一回路冷管段发生双端断裂大破口时燃料元件包壳温度峰值（PCT）最高,且长时间维持在较高温度,此条件下反应堆最危险。计算结果表明,事故发生后,一回路压力迅速下降,堆芯冷却剂的流动性变差,导致堆芯裸露,燃料包壳温度又重新回升。通过安注系统和辅助给水系统等一系列动作,能保证燃料元件包壳温度不超过1204 ℃的限值。     

船用堆小破口失水事故处置的影响因素分析

针对船用堆小破口失水事故处置复杂的特点,利用运行安全分析平台对事故进行了仿真研究,探讨了补水系统、危急冷却系统、二回路设备等对事故处置过程和后果的影响,为运行人员的处理和操作提供了参考,有助于失水事故应急处置规程的制定。     

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。