Nuclear Power Plant Risk Research Articles

Recognition of earthquake-prone areas (M ≥ 5.0) in the Iberian Peninsula

Seismogenic nodes capable of earthquakes with M ≥ 5.0 or I 0 ≥ VII have been identified in the Iberian Peninsula using the pattern recognition approach. Recognition objects, morphostructural nodes, have been delineated with the morphostructural zoning method. Most of the recognized seismogenic nodes (D) are scattered at the periphery of the Peninsula, while in its internal part, apart from the northern part of the Iberian Chain, there is no indication for the existence of D nodes. The performed recognition pinpoints a number of D nodes where moderate events have not been recorded to date, specifically, in the Cantabrian Mts, Portuguese basin, westernmost termination of the Betics, and in the area around Valencia. Some of the recognized D nodes are potential sources of seismic risk for nuclear and water power plants and large metropolitan areas.

Perception of people for the risk of Tianwan nuclear power plant

A questionnaire survey of residents’ risk perceptions related to Taiwan nuclear power plant in China was carried out to explore the determining factors that affect individual risk perception. This study proposed to pursue a more comprehensive understanding of factors that affected individual risk perception to nuclear power plants. Covariance structure analysis was conducted using risk perceptions of nuclear power as dependent variable and including interest and knowledge levels of nuclear power, acceptability, benefit perception, trust in nuclear power operation, and trust in government as independent variables. The use of the hypothesis of Elaboration likelihood model (ELM) was also proposed. The results showed that persons with higher levels of interest and knowledge of nuclear power had their own perceptions of risk closely associated with acceptability and potential benefits of nuclear power. In contrast, persons with no interest in and knowledge of nuclear power would have risk perceptions related to their trust in nuclear operation and the government, which partially supported the ELM hypothesis. All these results indicated that the government in China plays an important role in rational risk perceptions, and well-designed communication of risks will help the public to be involved in risk management and improve people’s rational acceptance of risk.

The Impact of Severe Accident Management Guidance on the Frequency of Containment Failure Modes of a Boiling Water Reactor

Probabilistic safety assessment (PSA) employs a systematic approach to estimate the risk associated with the operation of nuclear power plants (NPPs). Severe accident management guidance (SAMG), which delineates the mitigation actions of core meltdown accidents of NPPs, is developed to support operators and staffs in the technical support centers during the emergency responses of core melt accidents. Proper execution of SAMG could lower the failure probability of containment and reduce the amount of radionuclides released to the environment during the accident. It can be expected that the implementation of SAMG will reduce the risk of NPPs. However, SAMG is not available when most of the conventional level-2 PSA analyses are performed. In the present study, the mitigation actions of SAMG are incorporated into the level-2 PSA model of the ChinShan Nuclear Power Station of the Taiwan Power Company. The NPP analyzed employs a General Electric-designed boiling water reactor-4 with Mark I containment.The effectiveness of the mitigation actions specified in SAMG to terminate the progression of the accident is verified and validated using the MAAP4 code. The containment system event trees and containment phenomenological event trees of the level-2 PSA model are modified to incorporate the new mitigation actions specified in SAMG. The Human Cognitive Reliability (HCR) and Technique for Human Error Rate Prediction (THERP) models are used to quantify the human error probability (HEP) of all the actions in the level-2 PSA model. The MAAP4 code is used to perform thermohydraulic calculations to determine the demand time required in the HEP analysis.The results show that the total frequency of accident progression beyond vessel failure is reduced by 41% and the change in the probability of containment staying intact is not very significant because of the implementation of SAMG. After SAMG implementation, the frequency of containment early failure is reduced by 69.9%. The frequency of suppression pool venting is increased by 77.9%. The changes in the frequency of other containment failure modes are relatively insignificant. The most important human action is specified in Guideline RC/F of Severe Accident Guideline-1, i.e., In-Vessel Injection to Arrest Core Damage.

Determination of Failure Criteria for Electric Cables Exposed to Fire for Use in a Nuclear Power Plant Risk Analysis

Abstract The vulnerability of electrical cables exposed to a fire environment is of particular concern to the nuclear power plant community. The community is interested in data that could be used for predicting cable failures during a fire situation. For this reason, a cable test program was conducted using two different types of cable insulation. Several different exposure heat fluxes were tested, as well as different test arrangements such as cable trays and conduits. The program revealed that a single failure temperature for all cable types is not recommended, but if it is necessary a reasonable temperature could be chosen for the thermosets tested in this project.

投資リスクを踏まえた電源の投資問題

In this paper, we consider the investment risk of nuclear power plants using the real options approach. It is essential that the Japanese society evaluate the investment risk, because nuclear power plants are facing definite uncertainty and Japanese governments intend to promote and assist nuclear power plants through subsidies and policy actions.We assumed that the wholesale market prices of electricity constitute the definite uncertainty and that the wholesale market prices follow the geometric Brownian motion with drift. Using the Bellman equation and a lattice framework, we evaluated the value of investment opportunity, the value of equipment, and the critical prices that are optimal prices to invest in a nuclear power plant in the finite time horizon.This analysis shows that higher volatility of the wholesale market prices would give power companies lower incentive to construct electric power plants, particularly capital-intensive power plants. In order to deliberate and hold the Japanese governments accountable for the economics of nuclear power plants, multifaceted evaluation is needed.

A genetic algorithm solution for a nuclear power plant risk–cost maintenance model

Reliability Centered Maintenance (RCM) is one of the popular maintenance optimization methods according to certain kinds of priorities. Traditional RCM usually analyzes and optimizes the maintenance strategy from the viewpoint of component instead of the whole maintenance program impact. Research presented in this paper is a pilot study using PSA techniques in RCM. How to reflect the effect on component unavailability by the maintenance activities such as surveillance testing and preventive maintenance in PSA model is discussed firstly. Based on the discussion, a maintenance risk–cost model is established for global maintenance optimization in a nuclear power plant, and a genetic algorithm (GA) is applied to solve such a model to get the global optimized maintenance strategy. Finally, the result got from a simple test case based on a risk–cost model consisting of 10 components is presented.

Generic approach for designing and implementing a passive autocatalytic recombiner PAR-system in nuclear power plant containments

PARSOAR project is one of the scientific projects on hydrogen risk in nuclear power plants co-sponsored by the European Commission in the Fifth Euratom Framework Program and the Swiss government. It is one of the components of the “Safety Accident Management” (SAM) cluster that is focusing on review of severe accident risks and on development of countermeasures for defence-in-depth. The first objective is to carry out a state-of-the-art on passive autocatalytic recombiners (PAR) to constitute a database helpful for the PAR-designers, nuclear power plant designers and utilities, and safety authorities. The second objective is to elaborate a handbook guide that defines an approach for implementing catalytic recombiners in nuclear power plants, and that proposes recommended practices to support the detailed implementation by individual nuclear actors. The third objective is to identify the needs in terms of complementary experimental qualification or computer code developments about the hydrogen risk in nuclear power plants. It is also recommended that future experimental tests and numerical modelling should concentrate on the remaining issues of (a) lowering of existing uncertainties about hydrogen sources as core reflood or the reactions between boron carbide and steam or uranium and steam, (b) completing the experimental qualification of catalytic recombiners for existing and future nuclear power plant applications, and (c) performing well-qualified couplings between computational fluid dynamics computer codes and numerical models simulating PAR-behaviour (thanks to some global and well-instrumented experiments reproducing a complete reference accident sequence in presence of catalytic recombiners). Such areas to be considered for future work could be integrated in the scope of a European hydrogen network of excellence, in the framework of the Sixth Euratom Framework Program (2002–2006).

Risk significance and safety significance

Abstract Existing measures of the risk significance of elements of risk models (such as the Fussell–Vesely, or ‘F–V’, importance of basic events) are based on the properties of cut sets containing the element. A measure of safety significance (prevention worth, or PW ) is proposed, based on the properties of path sets containing the element. A high value of F–V means that cut sets containing the element contribute significantly to top event frequency; a high value of PW means that path sets containing the element contribute significantly to top event prevention. The properties of PW as a measure of basic event significance are illustrated first with a simple block diagram example, and then with an example based on nuclear power plant risk models. PW can also be understood as a property of a set of success scenarios, and as such, can be applied more broadly than just as a measure of element significance.

OECD activities in fire protection / OECD Aktivitäten im Bereich Brandschutz

Numerous fire PSAs (probabilistic safety assessments) have shown that fire can be a major contributor to nuclear power plant risk. However, there are considerable uncertainties in the results of these assessments, due to significant gaps in current abilities to perform realistic assessments. These gaps involve multiple aspects of fire PSA, including the estimation of the probability of important fire scenarios, the modeling of fire growth and suppression, the prediction qf fire-induced damage to equipment (including the effects of smoke), and the treatment of plant and operator responses to the fire. Therefore, the Nuclear Energy Agency (NEA) undertook a study among the Countries of the Organisation for Economic Cooperation and Development (OECD) with the focus on some special issues such as fire simulation, fire spreading, and impact of smoke and heat on instrumentation electronics. The results of this study are reported here.

Risk-Based Inspection of Pressure Components in Power and Process Plants

An important application of reliability and risk assessment in pressure vessels and piping engineering is risk-based ranking of pressure components for inservice inspection. The higher-ranked components may then be inspected more often or by more rigorous nondestructive examination procedures than the lower-ranked components in the plants. The development of risk-based ranking of components has its origin in the nuclear power industry. More recently, it has been employed in fossil power and process plants. One major differences between nuclear power plant risk assessment and fossil-power/process plant risk assessment is that the former has only one dominant risk (undesired event), namely, core melt accident, whereas the latter may have more than one dominant risk, for example, explosion, catastrophic rupture, leakage of toxic chemicals, etc. This paper presents a ranking procedure that can be used in cases of possible multiple undesired events, and thus applicable to fossil power and process plants.

Fire Safety Problems with Older Nuclear Power Plants

The fire risk of older nuclear power plants i.e. plants being more than 15 or 20 years under operation is a concern since in several plants serious fires occurred. Due to the rapid increase of knowledge in the field of fire science during the last three decade on the hand and with respect to public discussion about nuclear safety on the other, the responsible authorities permanently exert for the safety of older plants whereby the fire safety plays an increasing role. The report discusses experiences and method which were gathered and developed during the investigation of several older plants in Germany and the CSFR in the last decade. It related especially to the results of a survey in two German boiling water reactors (BWR 640 and BWR 670) which operate since 1969 and 1972 and a Russian pressure water reactor (WWER 440-V230) in the CSFR which came under operation in 1979. Conclusion of the report are that the fire safety of the Russian WWER 440-V230 is comparatively low. The proposed basic objectives of fire safety can not be fulfilled by this reactor type. It is mentioned that a Russian reactor of the same type in East-Germany was shut-down because of safety considerations immediately after German reunification. Inadequacy of fire safety was one of the main reasons which led to this decision.

Assessment of the potential applicability of fuzzy set theory to accident progression event trees with phenomenological uncertainties

Abstract Expert opinion is frequently used in the risk analysis of nuclear power plant systems to assess, in particular, the probabilities of rare events. However, this procedure is always accompanied by imprecision and uncertainty that characterise the experts judgment. Since the fuzzy set theory provides a framework for dealing with such judgmental imprecision and uncertainty, the potential applicability of the fuzzy set theory to the uncertainty analysis of accident progression event trees with imprecise and uncertain branch probabilities and/or with a number of phenomenological uncertainty issues are examined as a possible alternative procedure to that used in the current probabilistic risk assessments. The main purpose of this paper is to demonstrate a potential use of fuzzy set theory and provide its formal procedure in the quantification of the uncertainties of accident progression event trees. First, an example application of the fuzzy set theory is made to the simple portion of a given accident progression event tree with typical imprecise and uncertain type of input data, and thereby computational algorithms suitable for application of the fuzzy set theory to the accident progression event tree analysis are identified and illustrated. Secondly, to show the merits of the fuzzy set theory model in real application, the procedure used in the simple example is extended to extremely complex accident progression event trees with a number of phenomenological uncertainty issues, i.e. a typical plant damage state ‘SEC’ of the Zion nuclear power plant risk assessment, and the results are compared with the one obtained by current probabilistic methods. In addition, discussions and answers for five major questions on its real application are given.

A review of research at Sandia National Laboratories associated with the problem of smoke corrosivity

This paper presents certain results of fire safety research at Sandia National Laboratories (SNL). The work presented here is related to the issue of the development of standardized tests for determining the corrosive potential of materials when burned. This paper will focus on information about five specific aspects of the corrosivity issue that has been gathered as a part of several individual experimental and analytical studies. These five topics are (1) the current perception of fire risk for nuclear power plants and the role of corrosivity in that risk, (2) the composition of smoke particulate from large-scale enclosure cable fire tests and observations of chloride scavenging behavior by soot, (3) the aging behavior of smoke particulate, (4) the effect of fire size on the physical characteristics of generated smoke particulate, and (5) electrical equipment fire exposure test results.

Honest debate on nuclear energy

D Infield (Physics Bulletin January 1987 p3) has mistaken my objective. I was not intending to compare in one sentence the overall risks of coal-fired and nuclear power plants, the first having a certainty of chronic risks and the second a low probability of acute risks.

A Methodology for Analyzing the Detection and Suppression of Fires in Nuclear Power Plants

The assessment of the fire risk in nuclear power plants requires the analysis of fire scenarios within specified rooms. A methodology that integrates the fire protection features of a given room in...

Bayesian Updating of Seismic Probability

For those nuclear power plants for which a seismic probabilistic risk assessment has been conducted, it is possible to infer a prediction for the recurrence rate of the safe shutdown earthquake, and then to compare it with the historic seismicity at the site. Using the Bayesian algorithm, it is then possible to update the prediction in such a way as to quantify the degree of conservatism. By using a sample of eight plants, and other assumptions that are reasonable but by no means unique, the conservatism is estimated to be a factor of the order of 2 or 3. The uncertainty is also reduced, though there are caveats on this point. This suggests that earthquakes are somewhat overrated as sources of risk for nuclear power plants.

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.

Risk Criteria for Nuclear Power Plants: A Pragmatic Proposal1

Criteria are proposed for both an acceptable upper bound of nuclear power plant risk and a lower bound as a design target. Recognizing that the public risk associated with a power plant can be estimated only by probabilistic analysis of the design features, the spread between the lower design target and the upper bound provides a margin for uncertainty in th probabilistic estimate. The combination of a low probabilistic design target and this margin provides a reasonable expectation that the overall performance will be in the domain of an acceptable risk level. Because the exposure to potential risk is chiefly in the locality of the nuclear station, it is also proposed that compensatory benefits should be provided locally and that these be included as a cost of operation. It is suggested that the upper bound be set at a risk level equivalent to those risks of routine living which are normally accepted, i.e., about 10‐4 deaths per year per person (100 deaths/yr/million). The proposed lower design target is 10‐8 (0.1 deaths/yr/million), about one‐hundredth of the minimal risk from the natural hazards all people are exposed to.

Component probabilistic seismic analysis — What could be done now?

General outline on the direction of a probabilistic analysis of the seismic risk for nuclear power plants, including as far as possible the statistics of the seismic events, and of the structural or functional failures.

A complex study on the reliability assessment of the containment of a PWR Parts I–III

The intention of this paper is to contribute to the development of methods to be used for the quantification of the risk of nuclear power plants. For this purpose a reliability analysis of a structural component, i.e. a reactor containment structure is carried out. Detailed information in various fields had to be developed and compiled. The project consists of three parts: Part I concentrates mainly on the analysis of the load condition of the steel hull following a Loss of Coolant Accident (LOCA). Part II deals with the material aspects of the design properties of containment steels and furthermore the behaviour of concrete under impact load conditions are discussed. Part III of the paper is concerned on the one hand with external load conditions, and on the other hand with assembling the information of the previous parts to a reliability analysis. The methodology is exemplified by applying the general and theoretical results to the containment of the PWR-plant “Biblis B”.