AP1000小破口失水始发严重事故的源项研究

建立AP1000的事故分析模型,选取小破口失水始发的严重事故,在研究事故进程的基础上,分析计算事故下裂变产物释放和迁移的特性,重点关注惰性气体、挥发性裂变产物和非挥发性裂变产物在核电厂的分布,并选择破口位置、破口尺寸和安全壳泄漏率进行源项敏感性分析.本文分析结果可为严重事故管理和厂外放射性后果评价提供支持.     

船用堆大破口失水事故放射性后果分析

本文以严重事故分析程序MELCOR为计算工具,建立了某型船用堆的计算模型,研究了某型船用堆发生冷段双端断裂大破口失水事故的源项行为及放射性后果。分析了惰性气体Xe与挥发性气体CsI的释放、迁移和舱室分布规律,并对通风系统投入时机进行研究。结果表明:为保证堆舱临舱的剂量辐射在剂量限值内,应于事故发生后10min内投入全船通风。否则,应于全身剂量和甲状腺剂量达到剂量限值前及时采取防护措施。     

低温堆上空腔失水事故模拟实验研究

叙述了位于低温堆上空腔位置的中小尺寸管道破裂引起的小破口失水事故研究。在核供热堆热工水力学实验系统ＨＲＴＬ－５上，对停堆后堆内有剩余功率的上空腔小破口失水事故进行了模拟实验，分析了小破口失水事故发生后，系统运行重要参数的变化，给出了上空腔小破口失水事故对低温安全性的影响。     

小破口失水事故研究综述

对小破口失水事故（ＳＢＬＯＣＡ）及其研究状况进行了综述。描述了典型的压水堆和沸水堆小破口失水事故过程和破口位置、破口尺寸及反应堆冷却泵对失水过程的影响，对现有文献按实验和数值模拟两大类进行了归纳，给出了目前世界上用于小破口失水事故研究的主要设备，对小破口失水事故的研究进行了总结。     

小型动力堆码头中破口失水事故大气扩散研究

采用高斯分段烟羽模型估算了某小型动力堆在码头内发生破口尺寸为29.4%当量直径的设计基准事故时,放射性核素在码头20 km区域范围内的大气扩散规律。源项采用严重事故计算程序MELCOR仿真获得,并将计算结果输入到大气扩散分析软件MACCS进行分析计算。计算结果表明：中破口失水事故会造成码头区域的放射性污染,风速越小、气象条件越稳定,放射性的影响范围越大。     

中国先进研究堆事故源项分析

研究建立了中国先进研究堆（CARR）在事故工况下放射性核素从燃料芯块向环境释放的数学模型。根据CARR初步事故分析结果，对可能导致放射性向外界释放的5种事故工况（小破口失水事故、换热器传热板破裂事故、重水回路管道破裂事故、燃料操作事故、冷却剂流道堵塞事故）以及假想的3盒组件燃料板熔化超设计基准事故进行了源项分析，分别给出了不同事故和释放途径下释放到环境的放射性核素的量，以防止事故情况下公众和环境遭受过量放射性损伤。     

破口大小和二回路运行状态对核供热堆自然循环断流瞬态的影响

为研究一体化布置的核供热堆在发生破口失水事故中破口大小和从中间回路排出热量减少对断流过程的影响，选用不同的破口尺寸和不同的二回路工作状态，在５ＭＷ核供热堆热工水力模拟回路ＨＲＴＬ－５上进行了实验研究。稳态运行工况的系统压力为１．５ＭＰａ，在发生小破口失水事故后，加热功率维持为额定功率的５％以模拟剩余发热情况。实验研究并比较了不同条件下压力、温度、循环流量、液位和失水量等重要参数的变化。这些实验数据为核供热堆的安全分析提供了实验依据。     

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。     

脉冲堆失水事故计算程序的开发和应用

由于西安脉冲堆的特点，致使国际上通用的瞬时堆芯裸露模型不能使用。中国核动力研究设计院建立了反映西安脉冲堆失水事故机理和过程的真实真芯裸露模型，开发了相应的计算机程序，用于分析和评价西安脉冲堆的安全特性。分析结果表明，真实堆芯裸露模型具有广泛的实用性，可用于计算全部侧面破口和底部破口的失水事故。在破口直径相同的条件下，西安脉冲堆侧面破口失水事故后果比底部破口失水事故严重。在目前的设计条件下，即使发生失水事故，西安脉冲堆也能满足安全准则的要求。     

小型动力堆大破口事故下安全壳内气载放射性物质浓度计算

事故是压水堆固有属性之一,在众多导致核事故的初因事件中,大破口事故现象复杂,后果特别严重。基于此,本文以小型动力堆为研究对象,针对最重要的设计基准事故——大破口事故,计算了50、150、320满功率燃耗天冷端安注、双端安注条件下安全壳内放射性源项,并将部分计算结果与安全分析报告计算结果进行了对比。结果表明：假设合理、结果正确,对于保障反应堆运行安全、及时采取合理应急措施,意义重大。     

The effects of break location on PWR small break LOCA: Experimental study at the ROSA-IV LSTF

This paper discusses the effect of break location on the break flow rate and break flow quality transitions during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). Results from five experiments conducted at the ROSA-IV Large Scale Test Facility (LSTF) are compared for this purpose. These experiments simulated a 2-inch break at the lower plenum, upper head, pressurizer top, cold leg, and hot leg, respectively. The controlling phenomena for the break flow quality transitions in cold-leg and hot-leg break experiments are described.     

Investigation of fluid mixing with direct vessel injection

The object of this work is to investigate fluid mixing phenomena as they related to pressurized thermal shock (PTS) in a pressurized water reactor vessel downcomer during transient cooldown with direct vessel injection (DVI), using test models. The test model designs were based on ABB Combustion Engineering (CE) System 80+ reactor geometry. A cold-leg, small-break loss-of-coolant accident (LOCA) and a main steam line break were selected as the potential PTS events for the ABB-CE System 80+. This work consists of two parts. The first part provides the visualization tests of the fluid mixing between DVI fluids and existing coolant in the downcomer region, and the second part presents the results of thermal mixing tests with DVI in the other test model. Flow visualization tests with DVI have clarified the physical interaction between DVI fluid and primary coolant during transient cooldown. A significant temperature drop was observed in the downcomer during the tests of a small-break LOCA. The measured transient temperature profiles compare well with the predictions from the REMIX code for a small-break LOCA, and with the calculations from the COMMIX-1B code for a stream line break event.     

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.     

船用堆失水事故元件包壳破损温度阈值及气隙释放后果计算研究

针对船用堆特殊安全性要求,对船用堆失水事故包壳破损温度阈值进行研究。摒弃以往的保守假设,采用最佳估算模型,得到合理的温度阈值,并采用MELCOR程序对典型破口事故下包壳破损份额及气隙释放的放射性后果进行了计算。计算结果为评估舱室剂量、保障运行人员安全提供了依据。     

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。      

Application of PSA Methodology to Design Improvement of JAERI Passive Safety Reactor (JPSR)

A probabilistic safety assessment (PSA) technique was applied to the design of JAERI Passive Safety Reactor (JPSR). A PSA was performed to clarify safety features and identify vulnerabilities of the original design. Based on the PSA results and considering thermal-hydraulic analyses and experiments, the JPSR design was improved to enhance plant safety. The improved design was re-evaluated with the PSA. Initiating events selected in this study were: large-break LOCA, medium- and small-break LOCAs, SGTR, main steam line break, loss of offsite power, loss of feed water, and other transients. Fault tree analyses were used to evaluate the system unavailabilities. The total core damage frequency due to internal events was estimated to be less than 10^?7/RY. The contribution of high frequency non-LOCA events could be significantly reduced by the design modification. The dominant initiating event was the small break LOCA and the dominant sequence was the failure of residual heat removal system. The present study indicated that the improved JPSR design has sufficient safety margin and the PSA methodology is very effective to improve reactor safety systems in a conceptual design phase.     

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。     

Commissioning of the ATLAS thermal-hydraulic integral test facility

KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.     

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。