流道阻塞后的流场模型的开发及应用

在大破口失水事故进程中 ,燃料包壳可能发生的破裂将导致流道部分阻塞 ,在事故分析中必须考虑由此产生的影响。用COBRA Ⅳ Ⅰ子通道程序详细分析了流道阻塞后的流场 ,改进了大破口失水事故分析软件包中燃料棒包壳温度分析程序FRAP T6 ,对恰希玛核电厂大破口失水事故作了分析     

新型转换堆(ATR)下降段大破口失水事故的实验研究

为了分析新型转换堆(ATR)的大破口失水事故,在日本动力堆和核燃料开发集团大洗工程中心的 ATR 安全分析实验回路上,进行了三次不同破口直径的 ATR 下降段大破口失水事故实验。通过实验,对影响燃料元件安全的主要参数进行了测量和分析讨论。     

西安脉冲反应堆超设计基准事故动态特性分析

为填补以往西安脉冲反应堆(脉冲堆)超设计基准事故研究的不足,利用RELAP5/SCDAP/MOD3.4程序对脉冲堆系统进行了建模计算,给出了脉冲堆在断电ATWS事故和大破口失水ATWS事故下的瞬态响应特性。计算结果表明:发生断电ATWS事故后,在无人为干涉情况下,反应堆部分燃料可能熔毁;发生大破口失水ATWS事故后,破口位置和尺寸对事故后果的严重程度有重要影响,破口位置在堆池底部时,燃料最高温度低于1 800℃,而破口位置高于堆芯下栅板时,将导致燃料元件熔毁。根据脉冲堆在超设计基准事故下的动态响应,针对两种事故工况分别提出了相应的缓解措施。     

秦山核电二期工程失水事故分析

失水事故是核电站设计的基准事故之一,是压水堆事故分析关注的重点.本文概括介绍了秦山核电二期工程的失水事故分析及分析计算所使用的计算程序;简要地描述了MEFRA-1等计算程序的特点.重点介绍了大破口失水事故分析,给出了分析计算的主要假设条件和分析计算结果.分析计算表明,大破口失水事故工况下,燃料元件最大峰值包壳表面温度为1092.56℃,秦山核电二期工程的安全注射系统能保证该核电站在发生失水事故时的安全.     

安全壳压力和温度分析的质量及能量方程

大破口失水事故分析的目的是检验应急堆芯冷却系统在该事故条件下是否具有保证燃料元件完整性的能力。在事故期间，由于燃料元件的完整性与安全壳压力密切相关，当安全壳内的压力比较高时，燃料元件的完整性比较容易保证；当安全壳内的压力比较低时，燃料元件比较容易损坏。在相同的应急堆芯冷却系统条件下，采用不同的安全完模型可能会得到不同的堆芯燃料元件的峰值包壳温度。比较保守的大破口失水事故分析方法一般都假定在事故期间的安全完压力为认0.1MPa，这种假定从设计上有利于保证堆芯燃料元件的完整性，但是不利于提高核电厂的经济性。根据美国联邦法规SECY一83—472(最佳估算 部分保守)方法，建立一种现实的和相对保守的安全完模型与大破口失水事故程序进行堆芯热工水力特性和安全完压力耕合计算，在保证堆芯安全性的同时可提高核电厂的经济性，克服由人工输入安全完压力进行大破口失水事故计算带来的不确定性是本文的目的。     

大破口失水事故的DRM分析方法介绍

从大破口失水事故分析方法的发展过程，阐述了法国大破口失水事故分析方法DRM。该分析方法是核电厂安全评价的有效工具，可以为核电厂的燃料管理优化及提高经济效益发挥重要的作用。该方法已在大亚湾核电站18个月换料项目的提高堆芯功率因子的分析论证中应用。     

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。     

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1 204℃的限值。     

AP1000典型事故简要分析

AP1000典型事故包括失去外部电力负荷、失水事故、小破口失水事故、大破口失水事故、失水事故后的长期冷却、主蒸汽管道破裂、弹棒事故。通过对这些典型事故的分析,详细描述了事故的发生过程,讨论了事故后果及其影响。     

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。     

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。     

田湾核电站长周期换料大破口失水事故分析

田湾核电站拟采用长周期换料策略,堆芯设计的改变需对设计基准事故进行重新分析。本文对反应堆入口主管道大破口失水事故进行了计算分析,在保守的初始输入及计算假设的基础上,通过对轴向功率分布及应急堆芯冷却系统的保守性分析,得出基于燃料包壳温度的最保守的计算工况,并进行了计算。计算结果表明,实施长周期策略后,大破口失水事故仍可满足验收准则的要求,堆芯设计具有足够的安全裕量。     

Experimental analysis of the power curve sensitivity test series 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. Seven recirculation pump suction line break LOCA experiments were conducted at the ROSA-III facility in order to examine the effect of the initial stored heat of a fuel rod on the peak cladding temperature (PCT). The break size was changed from 200% to 5% in the test series and a failure of a high pressure core spray (HPCS) diesel generator was assumed. Three power curves which represented conservative, realistic and zero initial stored heat, respectively, were used.In a large break LOCA such as 200% or 50% breaks, the initial stored heat in a fuel rod has a large effect on the cladding surface temperature because core uncovery occurs before all the initial stored heat is released, whereas in a small break LOCA such as a 5% break little effect is observed because core uncovery occurs after the initial stored heat is released. The maximum PCTs for the conservative initial stored heat case was 925 K, obtained in the 50% break experiment, and that for the realistic initial stored heat case was 835 K, obtained in the 5% break experiment.     

Multi-Channel Thermal Hydraulic Model for LOCA Analysis of Heterogeneous BWR Core

A multi-channel thermal hydraulic model for LOCA analysis of a heterogeneous core such as a HCBWR has been developed. This model solves integral formulations for basic equations based on a one-dimensional drift flux model. The core region is divided into several fuel channel groups which differ in their thermal power or geometry. The various flow patterns in the core are determined by calculating the redistribution of vapor generated in the lower plenum into the fuel channel groups. In order to verify the multi-channel model, a computer program FLORA was developed based on the multi-channel model and large and small break LOCA experiments conducted in the Two Bundle Loop (TBL) facility were analyzed by the FLORA program. As a result, the difference in thermal hydraulic behavior between two bundles with different power in the various break LOCA experiments were well simulated.     

Similarity study of ROSA-III and fist large break cuonterpart tests to BWR large break LOCA

A large break test in a recirculation pump suction line with the assumption of LPCI-diesel generator failure was conducted at the ROSA-III test facility of Japan Atomic Energy Research Institute. A counterpart test was also performed at the FIST test facility of General Electric Company. The objective of the tests was to develop common understanding and interpretation of the controlling thermal-hydraulic phenomena during a large break LOCA of a BWR. The fundamental thermal-hydraulic phenomena in the ROSA-III and FIST tests such as the system pressure, mixture level and fuel rod surface temperatures agreed well. The FIST test had more bundle uncovery than that in ROSA-III since lower plenum steam in the FIST test flowed out of the jet pumps when they uncovered allowing more liquid to drain from the bundle. The ROSA-III and FIST tests and a BWR counterpart were analyzed with the RELAP5/MODI (cycle 018) code. The similarity of the ROSA-III and FIST large break tests to a BWR large break LOCA has been confirmed through comparison of calculated results though they are slightly different in details. It is perhaps desirable to reexamine the DNB and interphase drag correlations and the jet pump models usedin the code.     

PWR再淹没骤冷前沿温度场分析程序

为满足压水堆大破口LOCA分析的需要,在移植和开发TRAC-PF1程序中,应用了一种新颖的进行再淹没骤冷前沿处燃料元件温度场分析的方法。本文对这种方法及与之相关的燃料元件热传导数值模型、锆水反应和气隙传热计算进行了简要的描述。     

TRAC—PF1在大破口LOCA分析中的应用

文章简述了TRAC-PF1与大破口LOCA分析有关的功能和特点。针对大破口LOCA分析做出了秦山核电厂核蒸汽系统的适用于TRAC-PF1的模型。给出了对系统的稳态模拟结果和大破口LOCA分析的基本假设、事故过程及瞬态曲线。最后对结果进行了分析,指出为实际得到秦山核电厂大破口LOCA分析结果,在此基础上尚需获得并核实的关键数据。本文的意义在于介绍了一种应用TRAC-PF1进行大破口LOCA分析的方法。     

Thermal–hydraulic evaluation of the RBMK-1500 accident confinement system using CONTAIN 11AF

The response of the RBMK Accident Confinement System to a large break LOCA, medium break LOCA and small break LOCA is analyzed using the CONTAIN 11AF code. The effect of Condenser Tray Cooling System failure is investigated for the large break LOCA case. The analysis employs a best estimate mass/energy source and considers both short and long-term responses of the Accident Confinement System. Parametric studies are performed to evaluate the effects of water deposition on the short-term pressure peak and of by-pass leakage on long-term pressure increases.     

Development of Operator Action Time for SBLOCA during Plant Shutdown Operation in Power-uprated Kori 3 and 4

The ECCS performance, which mitigates a postulated catastrophic failure of the main reactor coolant piping during the full power operation, is judged to cover the consequences of LOCA occurring in other plant operational states. During Mode 3 with an accumulator isolated and Mode 4, since the normal alignment of ECCS equipments is changed from that which is available during the power operation, a potential safety issue, which involved the performance of ECCS for LOCA during Mode 3 with the accumulator isolated and Mode 4, was identified in 1985. This study is performed as the plant specific shutdown LOCA program for the power uprated Kori-3 and 4, of which the nominal core power is planned to increase by 4.5%. We determine and verify the operator action time to initiate SI following a small break LOCA in order that the peak clad temperature of fuel does not exceed the 10CFR50.46 limit of 1,477.6 K.

We evaluate the 0.1524 m (6 inches) pipe break in the cold leg to develop the SI initiation time. There is a considerable margin to the 10CFR50.46 limit of 1,477.6 K in the case that the SI is manually initiated at 25 min after an operator identifies the symptom of a small break LOCA. However, in respect of the safe plant operation, we decide the operator SI initiation time as 15 min in order that the SI water is supplied to prevent the fuel heat-up during the blowdown phase of a small break LOCA. After then, we evaluate the applicability of the pre-determined SI initiation time to other small break LOCAs, which have a smaller break size, a lower initial decay heat level or a different break location. Since the peak clad temperatures of applicability evaluation cases are lower than those of the umbrella case, we confirm that the pre-determined SI initiation time can be applied to mitigate the small break LOCAs during the plant shutdown operation. The SI initiation time developed in this study will be used in the Abnormal Operating Procedure of the power uprated Kori-3 and 4 for the small break LOCAs during the plant shutdown operation.     

Steam leak detection in advance reactors via acoustics method

Prediction of LOCA (loss of coolant activity) plays very important role in safety of nuclear reactor. Coolant is responsible for heat transfer from fuel bundles. Loss of coolant is an accidental situation which requires immediate shut down of reactor. Fall in system pressure during LOCA is the trip parameter used for initiating automatic reactor shut down. However, in primary heat transport system operating in two phase regimes, detection of small break LOCA is not simple. Due to very slow leak rates, time for the fall of pressure is significantly slow. From reactor safety point of view, it is extremely important to find reliable and effective alternative for detecting slow pressure drop in case of small break LOCA. One such technique is the acoustic signal caused by LOCA in small breaks. In boiling water reactors whose primary heat transport is to be driven by natural circulation, small break LOCA detection is important. For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault. The detection technique is reliable and plays a very important role in ensuring safety of the reactor. Methodology developed for steam leak detection is discussed in present paper. The methods to locate the leak is also developed and discussed in present paper which is based on analysis of the signal.