非能动先进压水堆核电厂SGTR事故概率安全评价

蒸汽发生器传热管破裂(Steam Generator Tube Rupture,SGTR)事故是核电厂的重要事故之一,并具有其自身的特点。该事故的研究和评价对核电站安全具有较大意义。选取典型非能动先进压水堆核电厂AP1000的SGTR事故进行一级概率安全评价(Probabilistic Safety Assessment,PSA),采用事件树分析方法得到电厂事件发生后系统、设备和人员不同响应所产生的事故序列,然后建立相关系统的故障树模型进行可靠性分析。借助Risk Spectrum软件,计算SGTR事故导致AP1000核电厂的堆芯损伤频率(Core Damage Probability,CDF),并进行堆芯损伤的最小割集分析及重要度和敏感性分析。通过一系列分析得到导致堆芯损伤的重要基本事件,从而找到系统存在的薄弱环节。     

Probabilistic economic approach to increasing nuclear plant safety

Editors' Note. During the editing of this article, various and at times conflicting opinions were expressed by members of the editorial board both about the article as a whole and about several ideas expressed in it — for example, on the use of the profit category, the creation of an insurance fund for nuclear plant accidents, etc. In view of this, the editors would be grateful for readers' responses to this article.     

Probabilistic assessment of thermal fatigue in nuclear components

A probabilistic framework is set up to assess the fatigue life of components of nuclear power plants. It intends to incorporate all kinds of uncertainties, such as those appearing in the specimen fatigue strength (number-of-cycles-to-failure of specimens), design margin factors (taking into account the size, surface finish and environmental effects), mechanical model (precisely, the uncertainty on the model input parameters) and the thermal loading. This paper presents the global methodology and details the statistical treatment of the fatigue specimen test data. A first analytical example shows that the reliability of a structure submitted to a periodic stress cycle S changes significantly with respect to the value of S, although the codified (deterministic) design criterion is equally fulfilled. A more comprehensive example involving a mechanical model of a pipe submitted to a deterministic inner temperature loading is finally analysed. The use of the first-order reliability method (FORM) allows to compute the probability of failure as a function of the foreseen lifetime and to rank the input random variables according to their importance in response sensitivity.     

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.     

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.     

Effect of fuel — Cladding axial asymmetry on nuclear fuel element operation

Institute of Nuclear Reactors, Russian Scientific Center“Kurchatovskii institut.”Translated from Atomnaya énergiya, Vol. 76, No. 5, pp. 417–422, May, 1994.     

Probabilistic seismic safety study of an existing nuclear power plant

This study was conducted as part of an overall safety study of the Oyster Creek nuclear power plant. The earthquake hazard was considered as an initiating event that could result in radioactive release from the site as a result of core melt. The probability of earthquake initiated releases were compared with the probability of releases due to other initiating events.Three steps are necessary to evaluate the probability of earthquake initiated core melt.

1. (1) estimate the ground motion (peak ground acceleration) and uncertainty in this estimate as functions of annual probability of occurrence;

2. (2) estimate the conditional probability of failure and its uncertainty for structures, equipment, piping, controls, etc., as functions of ground acceleration; and

3. (3) combine these estimates to obtain probabilities of earthquake induced failure and uncertainties in such estimates to be used in event trees, system models, and fault trees for evaluating the probability of earthquake induced core melt.

This paper concentrates on the first two steps with emphasis on step 2. The major difference between the work presented and previous papers is the development and use of uncertainty estimates for both the ground motion probability estimates and the conditional probability of failure estimates.The ground motion capacity of a structure, component, etc., is treated for simplicity and clarity as a product random variable A given by , where is the best estimate of the median ground acceleration capacity, _R and _U are lognormal random variables with unit median and logarithmic standard deviation β_R and β_U, respectively. β_R expresses the dispersion in the ground acceleration capacity due to underlying randomness from such sources as (1) the variability of an earthquake time-history and thus of structural response when the earthquake is only defined in terms of the peak ground acceleration; and (2) the variability of structural material properties associated with strength, inelastic energy absorption and damping. Essentially, β_R represents those sources of dispersion which cannot be reduced by more detailed evaluation or more data. Uncertainty concerning the ground motion capacity is expressed by β_U which results from such things as (1) lack of complete knowledge of structural material properties; and (2) errors in calculating response due to approximate modelling. This paper presents a methodology (with examples) for estimating , β_R, and β_U for structures and components. These estimates are then used to estimate conditional probabilities of failure with confidence bounds on these estimates.The conclusion is that a rational approach exists for estimating earthquake induced probabilities of failure. Confidence bounds on such estimates can be developed to express uncertainty in the parameters used. Such an approach is preferable over one in which dispersion due to underlying randomness, and due to uncertainty in the data are combined into a single probability of failure estimate with no estimate of the uncertainty in this probability.     

Thermal shock in a nuclear fuel element with cladding

The object of the present paper is to look into the evolution of thermoelastic state in a nuclear pin with cladding, when a step-variation occurs in coolant temperature.For both fuel and cladding the transient stress components are obtained: they must be added to the steady stress components in the case of step-decrease (ΔT < 0), and subtracted in the case of step-growth (ΔT > 0). Some numerical results are finally reported with reference to a fast breeder reactor cooled by sodium and to a pressurized water reactor.     

Development of safety assessment methodology on fuel element failure propagation in SFR and its application to Monju

Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified from a state-of-the-art review of open papers. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation are modeled into a safety assessment code which is applicable to arbitrary SFRs by developing some needed but missing methods. Furthermore, an assessment on FEFPA of Japanese prototype fast breeder reactor (Monju) was performed using this methodology. It was clarified that FEFPA was highly unlikely and limited at most within one subassembly in Monju owing to its redundant and diverse detection and shutdown systems for FEFPA even assuming the propagation. These results also suggested future possibility of run-beyond-cladding-breach operation which would enhance the economic efficiency in Monju.     

Additional studies of criticality safety of failed used nuclear fuel

Commercial used nuclear fuel (UNF) in the USA is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high burnup values (>45 GWd t^?1) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF are not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on the criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories, and specific configurations were evaluated to understand trends and quantify the consequences of worst case potential reconfiguration progressions. These results are summarised here and indicate that the potential impacts on subcriticality can be rather significant for certain configurations (e.g. >20% Δk_eff). However, for credible fuel failure configurations from ES or transportation following ES, the consequences are judged to be manageable (e.g. <5% Δk_eff). The current work expands on the previous efforts by including part length rods in fresh boiling water reactor fuel assemblies and studying the effect of damage in varying numbers of fuel assemblies.     

Probabilistic assessment of the seismic hazard for Eastern United States nuclear power plants

The purpose of the seismic hazard characterization of the Eastern United States project, for the Nuclear Regulatory Commission, was to develop a methodology and data bases to estimate the seismic hazard at all the plant sites east of the Rocky Mountains. A summary of important conclusions reached in this multi year study is presented. The magnitude and role of the uncertainty in the hazard estimates is emphasized in regard to the intended final use of the results.     

Some aspects of nuclear and radiological safety of operating a spent nuclear fuel repository

A model of an irregular situation in a spent nuclear fuel repository with the introduction of excess reactivity into the system, consisting of containers with spent fuel assemblies and water, is examined. The neutron kinetics of a critical system is calculated taking account of the thermohydraulics of the system. The character of the flow of a short-time self-sustained chain reaction — “neutron burst” — is described. It is found that an excursion of the system in the range of reactivity introduction rates examined will result in heating of the system and self-quenching of the chain reaction by negative reactivity effects with respect to fuel temperature. Intense fluxes of fission neutrons and prompt gamma rays, accompanying a self-sustained chain reaction, are formed in the excursion process. A mixed neutron and gamma ray field near the system considered is investigated. __________ Translated from Atomnaya énergiya,Vol. 104, No. 3, pp. 141–147, March, 2008.     

一种用于退役核燃料元件包壳的破裂检测技术

本文阐述了基于活性炭对85Kr吸附特性来检测退役核燃料元件包壳破裂的技术,给出了测量原理及测量系统。方法是通过测量活性炭所吸收的85Kr所放出β射线的计数率的大小,来判断退役核燃料元件包壳是否破裂,该测量系统对β射线的探测效率大于40%,对β射线的本底计数率为0.84Counts?min–1。使用该方法的实际测量结论与取水样用化学分析方法所得结论一致。     

Micro-structured nuclear fuel and novel nuclear reactor concepts for advanced power production

Many applications (e.g. terrestrial and space electric power production, naval, underwater and railroad propulsion and auxiliary power for isolated regions) require a compact-high-power electricity source. The development of such a reactor structure necessitates a deeper understanding of fission energy transport and materials behavior in radiation dominated structures. One solution to reduce the greenhouse-gas emissions and delay the catastrophic events' occurrences may be the development of massive nuclear power. The actual basic conceptions in nuclear reactors are at the base of the bottleneck in enhancements. The current nuclear reactors look like high security prisons applied to fission products. The micro-bead heterogeneous fuel mesh gives the fission products the possibility to acquire stable conditions outside the hot zones without spilling, in exchange for advantages – possibility of enhancing the nuclear technology for power production. There is a possibility to accommodate the materials and structures with the phenomenon of interest, the high temperature fission products free fuel with near perfect burning. This feature is important to the future of nuclear power development in order to avoid the nuclear fuel peak, and high price increase due to the immobilization of the fuel in the waste fuel nuclear reactor pools.