Probabilistic risk assessment in nuclear power plant regulation

A specific program is recommended to utilize more effectively probabilistic risk assessment in nuclear power plant regulation. It is based upon the engineering insights from the Reactor Safety Study (WASH-1400) and some follow-on risk assessment research by USNRC. The Three Mile Island accident is briefly discussed from a risk viewpoint to illustrate a weakness in current practice. The development of a probabilistic safety goal is recommended with some suggestions on underlying principles. Some ongoing work on risk perception and the draft probabilistic safety goal being reviewed in Canada is described. Some suggestions are offered on further risk assessment research. Finally, some recent U.S. Nuclear Regulatory Commission actions are described.     

大型压水堆核电站换料机的强度分析

压水堆核电站换料机对保障核电站安全运行具有重要的作用,对其主要结构的动力计算和强度评定具有重要的意义。本文应用有限元分析软件ANSYS 12对1 000 MW核电站大型换料机进行了有限元建模,并分别在正常工况(启动、制动)、异常工况(OBE)和事故工况(SSE)下进行了动力计算;采用SRSS方法对3个不同方向地震反应谱下的结构响应(内力、应力)进行了工况组合,并进一步考虑了自重条件的不利影响。根据RCCM规范对换料机主要结构、螺栓、焊缝的强度和辅吊支腿的稳定性进行了评定,并在此基础上对抓取燃料组件的指形钩进行了局部强度分析。评定结果表明换料机的强度在不同工况下均满足规范要求。     

核电站风险监测系统恢复分析方法及其应用

核电站概率安全评价中一般都包含恢复分析的内容,作为概率安全评价技术高级应用的风险监测系统同样也需要恢复分析.在分析风险监测系统实时风险计算特点及要求的基础上,总结了风险监测系统恢复分析特点,并以秦山三核风险监测系统恢复分析为例进行详细介绍.研究给出一般性指导原则,并为核电站概率安全评价模型的建立提供了参考建议.     

Seismic fragilities for nuclear power plant risk studies

Seismic fragilities of critical structures and equipment are developed as families of conditional failure frequency curves plotted against peak ground acceleration. The procedure is based on available data combined with judicious extrapolation of design information on plant structures and equipment. Representative values of fragility parameters for typical modern nuclear power plants are provided. Based on the fragility evaluation for about a dozen nuclear power plants, it is proposed that unnecessary conservatism existing in current seismic design practice could be removed by properly accounting for inelastic energy absorption capabilities of structures. The paper discusses the key contributors to seismic risk and the significance of possible correlation between component failures and potential design and construction errors.     

Probabilistic assessment of seismic risk for nuclear power plants

Probabilistic approaches to the design, siting, and safety analysis of nuclear power plants have been proposed by Farmer, Wall, and Garrick. Farmer and Wall established a limit line which delineates between acceptable and unacceptable risks. To implement the method, all accidental chains are systematically analyzed to determine their probability and associated fission product release magnitude; the combination is compared to the limit line. For proper implementation, the seismic risk should be evaluated in a quantified manner. Conceptually, this evaluation is made in two stages: the probability of an earthquake occurrence as a function of its intensity and, given a seismic intensity, the conditional probability of damage. This paper reports on an initial study into the latter aspect.The effect of uncertainty in several parameters which determine the response of a nuclear reactor building to earthquake forces is assessed. Probability distributions for material properties were determined from site measurements and these distributions were utilized for determining the building response and the damage criterion. A subjective probability density function for damping was assigned from the available information and the judgment of experienced engineers. Four accelerograms, El Centro N---S 1940, and three artificial earthquakes were used to represent the variability in the forcing functions. The uncertainty in the model idealization was assessed by evaluating three alternate models. A versatile computer program was developed to compute the response of the mathematical model to the forcing functions using matrix formulation and modal method of analysis. An exact solution, rather than numerical integration, was used to obtain the dynamic response of the system in generalized coordinates.The stresses within the reactor building are similar for different earthquakes considered in this study when they are normalized to ground acceleration, indicating that the shape of the accelerogram and its frequency content are less significant than the magnitude of the maximum ground acceleration for the reactor building. The variation in modulus of elasticity for concrete had a significant effect on the building response. Damping, in general, reduced the response, but in cases where the duration of an earthquake is short the effect was not very significant.A simple failure criteria for ultimate shear stress in shear walls, τ_ult = 4.75 √ƒ′_c, where ƒ′_c is the ultimate compressive strength of concrete, is used to estimate the initiation of cracking in the walls. The normal design of the reactor building is deterministic and is based on a 0.2 g design basis earthquake. Using the results obtained by the parametric study on the variation of response, the probability of damage was estimated by a Monte Carlo analysis. It was estimated that, given the occurrence of a design basis earthquake, there is less than approximately 10⁻³ probability of cracking in the shear walls of the reactor building. The initiation of cracking in the concrete should not lead to a significant release of contained fission products.     

风险指引型设备分级方法研究

风险指引型设备分级是综合确定论和概率论的分析结果对设备进行重新分级的一种方法.它可以使核电厂将资源更合理地分配到安全重要度高的设备上,同时节省大量的采购及其他相关费用.文章介绍了风险指引型设备分级的分析方法和过程,并以大亚湾核电站辅助给水系统为例对这种方法进行了研究.     

Automated service tool for moving instrumentation thimbles in PWRs during refueling outages

During the refueling outage period of PWR nuclear power plants, the in-core movable detector thimbles located inside fuel assemblies are withdrawn from the seal table, and inserted again as the refueling is finished. The thimble handling process has been commonly performed manually. However, the manual handling process can give many drawbacks including unsatisfactory labor efficiency due to the narrow and high radiation working environment around the seal table, and the possibility of giving serious damage to the delicate thimbles caused by uneven force exerted on the thimbles. In order to minimize radiation exposure to the workers and maintain the thimbles safely, an easy-to-use and robust thimble service tool has been developed. The performance of the tool was verified at the test facility and Kori #3 unit (Westinghouse design, 1000 MWe).     

核电站风险监测系统数据库的设计和实现

风险监测系统是一种能辅助和优化核电站运行和维修决策的实时风险在线分析工具.数据库作为风险监测系统的重要部分,是分析计算工作的基础,而且其涉及数据量大、数据关系复杂、使用对象广泛,因此其设计和实现是风险监测系统的关键环节之一.本文采用了面向对象关系设计模式进行了数据库设计,并介绍了在数据库实现中面临的一系列问题和为解决这...     

Comparative health risk assessment of coal power and nuclear power in China

The health risk from human exposure of radioactivities and accidents in coal power and nuclear power energy chain were compared in this paper. We got the results that the health risk of coal-fired energy chain was higher than that of nuclear energy chain.     

An estimation method for basemat uplift behavior of nuclear power plant buildings

It is important to accurately estimate the effects of strong earthquake motions on the basemat uplift behavior and structural responses for the seismic design of nuclear power plant buildings. In this paper, an analysis model which describes the soil part using the 3 dimensional FEM was proposed to be used when the ground contact ratio is low, and the validity of this model was confirmed. Furthermore, investigations using the model were carried out where the attaching force under the basemat was taken into account, in order to more realistically estimate the basemat uplift behavior. The effects in the case of the building being embedded were also investigated.     

Quantitative assessment of severe accident management strategies in a nuclear power plant

Nuclear power plants in Korea are preparing improvement countermeasures for severe accidents including mobile gas turbine generators, passive autocatalytic recombiners, containment-filtered venting systems, and external injection using portable pumps. However, these improvement countermeasures have only been determined by expert judgment, and detailed validation of their effects has not been performed. In this paper, the quantitative safety impact of these improvement countermeasures was evaluated for the Westinghouse 3-loop pressurized water reactor. Our evaluation of four improvement countermeasures using the at-power internal event probabilistic safety assessment models revealed that all containment failure modes have positive effects, except for the containment isolation failure and the containment bypass. Therefore, post-Fukushima action plans for coping with a severe accident in Korea have been appropriately evaluated and established.     

核电厂选址及环境影响评价应关注的问题

结合近年来我国核电厂选址工作,根据核安全以及环境保护的有关法规要求,重点分析了我国核电厂选址及环境影响评价的特点,并提出了我国核电厂选址及环境影响评价应关注的问题。     

Underground nuclear power plant siting

This paper summarizes the results of investigations to define the design concepts and estimate cost penalties associated with the burial of large light-water reactor nuclear power plants in underground rock cavities. Several cavities are proposed to contain the major components of the power plant without requiring excessive spans. The cost penalty of the underground plant is estimated to be less than 10% above a similar surface plant in favorable geologic media. Preliminary analyses also indicate a potential improvement in containment of radioactive materials following a postulated accident.     

Uniterm low-capacity nuclear power plant

The results of development work on an innovative low-capacity nuclear power plant (LCNPP) Uniterm together with substantiation of the need for a technical design of a reactor system and the basic design of the power plant are presented. The results of the work are used to determine the power range of the power generating unit where the main consumer qualities are guaranteed to be attained. Economic assessments of the development and operation of a two-block LCNPP under conditions in Russia are presented. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 44–48, July, 2007.     

Transport of large nuclear power plant components: experiences in mechanical design assessment

Abstract

In the course of decommissioning of power plants in Germany large nuclear components (steam generator, reactor pressure vessel) must be transported over public traffic routes to interim storage facilities, where they are dismantled or stored temporarily. Since it concerns surface contaminated objects or low specific activity materials, a safety evaluation considering the IAEA transport regulations mainly for industrial packages (type IP-2) is necessary. For these types of industrial packages the requirements from normal transport conditions are to be covered for the mechanical proof. For example, a free drop of the package from a defined height, in dependence of its mass, onto an unyielding target, and a stacking test are required. Since physical drop tests are impossible generally due to the singularity of such 'packages', a calculation has to be performed, preferably by a complex numerical analysis. The assessment of the loads takes place on the basis of local stress distributions, also with consideration of radiation induced brittleness of the material and with consideration of recent scientific investigation results. Large nuclear components have typically been transported in an unpackaged manner, so that the external shell of the component provides the packaging wall. The investigation must consider the entire component including all penetration areas such as manholes or nozzles. According to the present IAEA regulations the drop position is to be examined, which causes the maximum damage to the package. In the case of a transport under special arrangement a drop only in an attitude representing the usual handling position (administratively controlled) is necessary. If dose rate values of the package are higher than maximum allowable values for a public transport, then it is necessary that additional shielding construction units are attached to the large component.     

Integral equation for nuclear power plant reduced-load-operation strategy

An integral equation of the Fredholm first kind is developed semi-analytically to quantify the strategy of operating a nuclear power generating unit at reduced load in the (100–600) MW (e) range so as to break even its average (over the lifetime of the plant) power generation costs with the generation costs of a nuclear plant of a smaller size and with an identical plant availability curve. For the model of plant constant availability, the integral equation may be reduced to a transcendental relation between the reduced-load-operation time, the annual fixed charge rate of plant capital costs, its availability figure, the rate of load reduction and the plant lifetime.     

Equipment response spectra for nuclear power plant systems

An analytical method is developed whereby a simple estimate can be obtained of the maximum dynamic response of light equipment attached to a structure subjected to ground motion. The natural frequency of the equipment, modeled as a single-degree-of-freedom system, is considered to be close or equal to one of the natural frequencies of the N-degree-of-freedom structure. This estimate provides a convenient, rational basis for the structural design of the equipment and its installation.The approach is based on the transient analysis of lightly damped tuned or slightly nontuned equipment-structure systems in which the mass of the equipment is much smaller than that of the structure. It is assumed that the information available to the designer is a design spectrum for the ground motion, fixed-base modal properties of the structure, and fixed-base properties of the equipment. The results obtained are simple estimates of the maximum acceleration and displacement of the equipment. The method can also be used to treat closely spaced modes in structural systems, where the square root of the sum of the squares procedure is known to be invalid.This analytical method is also applied to nontuned equipment-structure systems for which the conventional floor spectrum method is mathematically valid. A closed-form solution is obtained which permits an estimate of the maximum response of the equipment to be determined without the necessity to compute time histories as required by the floor spectrum method.