QRA Considering Multi-Vessel Failure Scenarios due to a Natural Disaster – Lessons from Fukushima

PresentationThe Fukushima accident is the largest nuclear power plant accident caused by a natural disaster, which shut off the cooling system. In this accident, an initiating event from a single unit was propagated to other units at the site. Prior to the Fukushima accident, scenarios for multi-unit failures had been screened out, so that only single unit failure scenarios were taken into account in the nuclear industry. Since that accident, the nuclear industry in South Korea is now more concerned with multi-unit Probabilistic Safety Assessment (PSA) and is attempting to develop a new methodology since there are generally more than four units on one site in the country The chemical and petrochemical industries have experienced a number of incidents/accidents related to multi-units such as vessels and tanks because many of these units are usually installed on a site. The chemical industry has a wide range of experience, but most scenarios have involved domino effects, while the study of multi-vessel accidents caused by natural disasters (i.e., earthquakes) is lacking. The purpose of this paper is to review past experiences in the chemical industry and adapt appropriate process safety applications using risk analysis related to multiple vessels (tanks) at a site. Several QRA approaches have been searched and employed to compare the risks of some chemical plant complexes in South Korea regarding multi-vessel (unit) failure scenarios due to natural disasters

Issues of the Seismic Safety of Nuclear Power Plants

Seismic safety of nuclear power plants became an eminent importance after the Great Tohoku earthquake on 11th of March, 2011 and subsequent disaster of the Fukushima Dai-ichi nuclear power plant. Intensive works are in progress all over the world that include review of the site seismic hazard assessment, revision of the design bases, evaluation of vulnerability, and development of accident management capabilities of the plants. The lessons learned from the Fukushima-accident changed the paradigm of the design. Preparedness to the impossible, i.e. the development of means and procedures for ensuring the plant safety in extreme improbable situations became great importance. Main objective of the Chapter is to provide brief insight into the actual issues of seismic safety of nuclear power plants, provide interpretation of these issues, and show the possible solutions and scientific challenges. The “specific-to-nuclear” aspects of the characterisation of seismic hazard, including fault displacement are discussed. The actual design requirements, safety analysis procedures are briefly presented with main focus on the design extension situations. Operation aspects and problems for restart after earthquake are also discussed. The Chapter is more focusing on seismic safety of the inland plants, located on soil sites, in low-to-moderate (diffuse) seismicity regions

An overview of current status of carbon dioxide capture and storage technologies

AbstractGlobal warming and climate change concerns have triggered global efforts to reduce the concentration of atmospheric carbon dioxide (CO2). Carbon dioxide capture and storage (CCS) is considered a crucial strategy for meeting CO2 emission reduction targets. In this paper, various aspects of CCS are reviewed and discussed including the state of the art technologies for CO2 capture, separation, transport, storage, leakage, monitoring, and life cycle analysis. The selection of specific CO2 capture technology heavily depends on the type of CO2 generating plant and fuel used. Among those CO2 separation processes, absorption is the most mature and commonly adopted due to its higher efficiency and lower cost. Pipeline is considered to be the most viable solution for large volume of CO2 transport. Among those geological formations for CO2 storage, enhanced oil recovery is mature and has been practiced for many years but its economical viability for anthropogenic sources needs to be demonstrated. There are growing interests in CO2 storage in saline aquifers due to their enormous potential storage capacity and several projects are in the pipeline for demonstration of its viability. There are multiple hurdles to CCS deployment including the absence of a clear business case for CCS investment and the absence of robust economic incentives to support the additional high capital and operating costs of the whole CCS process

Comparison of the ranges of uncertainty captured in different seismic-hazard studies

The inclusion of epistemic uncertainties, generally via logic trees (Kulkarni et al., 1984), within probabilistic seismic‐hazard assessments (PSHAs) is becoming standard for all types of studies (commercial, governmental, or research; site specific, national, regional, or global). Consequently many studies publish expected ground motions for a given annual frequency of exceedance (AFE) or return period derived from the hazard curves for the mean, median, and various fractiles (percentiles)

Historical review of fire safety at NPP and application of fire PSA to Westinghouse PWR NPP in the frame of risk-informed decision making by

The importance of fire as a potential initiator of multiple-system failures took on a new perspective after the cable-tray fire at Browns Ferry in 1975 The review have shown that the first generation Nuclear Power Plant (NPP) fire safety was not factored as high risk area that needed to be effectively assessed and quantified. This resulted in development of peculiar fire safety regulations, standards and expensive backfits. Lack of appropriate regulations and effective methods of fire risk assessment, prescriptive, difficult and expensive retrofit regulations were instituted in USA. The alternative risk-informed performance based regulation was established in USA to resolve the challenges of the prescriptive rules. The review have revealed that both the prescriptive and risk-informed performance based approaches will not represent adequate design basis for new Nuclear Power Plants. The Japanese were pulled in the path of renew fire safety regulations and risk quantification after the Fukushima accident. It has been recognized that effective fire safety assessment, and culture, in concert with countermeasures to prevent, detect, suppress, and mitigate the effect of fires if they occur, will minimized NPP fire risk. Among the numerous recommendation the fire safety at NPP must be planned and engineered before construction begin using the state-of-the-arts technology. Also, the methods of fire risk assessment must integrate the state-of-the-arts deterministic and probabilistic approaches. Two methods are presented which serve to incorporate the fire-related risk into the current practices in nuclear power plants with respect to the assessment of configurations. The first method is a fire protection systems and key safety functions Unavailability Matrix (UM) which is developed to identify structures, systems, and components significant for fire-related risk. The second method is a fire zones and key safety functions (KSFs) fire risk matrix which is useful to identify fire zones which are candidates for risk management actions. The UM is an innovative tool to communicate fire risk. The Monte Carlo method has been used to assess the uncertainty of the UM. The analysis shows that the uncertainty is sufficiently bounded. The significant fire-related risk is localized in six KSF representative components and one fire protection system which should be included in the maintenance rule. The unavailability of fire protection systems does not significantly affect the risk. The fire risk matrix identifies the fire zones that contribute the most to the fire-related risk. These zones belong to the control building and electric penetrations building. The aggregation of Internal Events PSA model and Fire PSA model have shown that the Fire PSA contributes 38.4% to the Risk increase. The feasibility of developing Fire-related Risk Monitor from the FIRE PSA for the Spanish NPP was carried out. One of the main challenges is that RiskSpectrum® fire PSA has 384 fire cases and 384 CDF but in Risk Monitor one CDF is required. However, CAFTA is unable to convert a Sequential Fault Tree structure of the internal Event tree in the Fire PSA. The conversion fails to implement neither all of the sequences leading to core damage nor the Fault Tree selection of the frequency of fire. The proposal is to suppress exchange events and introduce the alignment of the consequences so that a unique result of core damage can be quantified. The detection and fire suppression Event Trees in the reference model were replaced by detection and fire extinction Fault trees. The frequency of each Fire Case of the conversion model and the reference model are quantified and the frequencies compared. The results shows that 90% of the cases are valid, however, the rest have challenges with MCS. A unique CDF of 7.65x10-7 is quantified compared with 9.83×10-6 of the reference. The conversion of the new model in CAFTA was not successful due to software incompatibility.La importància del incendi com un potencial iniciador de sistema múltiples fallides van agafar una nova perspectiva després del incendi al cable-safata de Browns Ferry el 1975. La revisió ha mostrat que la primera generació de seguretat contra incendis de centrals d'Energia Nuclear (NPP) no va ser àrea de alt risc, àrea que necessitava ser efectivament avaluada i quantificada. Això va resultar en el desenvolupament de normes de seguretat de incendi peculiar, estàndards i cares revisions. La manca d'una reglamentació adequada i mètodes eficaços d'avaluació de risc d'incendi, va fer que als USA foren instituïts mètodes d'adaptació de normativa preceptius, difícils i costós. L'alternativa de regulació informada per el risc es va establir als USA per resoldre els reptes de la regulació preceptiva. La revisió ha mostrat que tant als enfocaments de normativa preceptiva i regulació informada per el risc no representen bases de disseny adequades per a noves NPP. Ha estat reconeguda que la efectiva avaluació de seguretat al incendi i la cultura en concert amb mesures per prevenir, detectar, suprimir i mitigar l'efecte d'incendis, si es produeixen, minimitzarà el risc d'incendi en una NPP. Entre les nombroses recomanacions la seguretat contra incendis a una NPP s'hauran previst i dissenyat abans de començar la construcció i utilitzant estat del art de la tecnologia. També, els mètodes d'avaluació del risc d'incendi tindran que integrar el estat del art en els enfocaments de determinista i probabilístics. Dos mètodes són presentats que serveixen per incorporar el risc relacionats amb el foc a les pràctiques actuals en centrals nuclears en respecte a l'avaluació de configuracions. El primer mètode és un sistema de protecció contra incendis i una matriu de indisponiblitats de les funcions clau de seguretat (MU) que es desenvolupa per a identificar estructures, sistemes i components significatius per riscos relacionats amb els incendis. El segon mètode és zones de focs i matriu de risc d'incendi i funcions (KSFs) clau de seguretat que és útil identificar les zones de foc que són candidats per a les accions de gestió de risc. La MU és una eina innovadora per comunicar el risc d'incendi. El risc significatiu relacionats amb el incendi està localitzat en sis components representatius KSF i un sistema de protecció de foc que cal que figuri en la regla de manteniment. La manca de sistemes de protecció contra incendis no afecta significativament al risc. La matriu de risc d'incendi identifica les zones de foc que mes contribueixen al risc relacionats amb el incendi. Aquestes zones pertanyen a l'edifici de control i edifici de penetracions elèctriques. L'agregació del model de PSA de esdeveniments interns i model de incendis PSA han demostrat que el PSA de incendis aporta 38.4% a l'augment de risc. S'ha desenvolupat la viabilitat del Monitor de risc de incendis a partir del PSA de incendis per a una central nuclear espanyola. Un dels reptes principals és que RiskSpectrum® incendis PSA te 384 casos de incendis i te 384 CDF però en risc Monitor és necessària una CDF. Tanmateix, el CAFTA és incapaç de convertir una estructura seqüencial de arbre de fallida de l'arbre esdeveniment interna en el PSA de incendis. La conversió fracassa al posar en pràctica totes les seqüències de danys al nucli i la selecció de l'arbre de fallida de la freqüència de incendi. La descoberta i supressió de arbres de l'esdeveniment de incendi en el model de referència es van substituir per detecció i els arbres de fallades d'extinció d'incendi. La freqüència de cada cas de incendi del model de conversió i el model de referència son quantificades i les freqüències son comparades. Els resultats demostra que el 90% dels casos són vàlid, no obstant això, la resta té reptes amb MCS. Un únic CDF de 7.65x10-7 s'ha quantificat en comparació amb 9.83 × 10-6 de la referència. La conversió del nou model a CAFTA no va tenir èxit a causa de la incompatibilitat del programari

Critical Infrastructures: Enhancing Preparedness & Resilience for the Security of Citizens and Services Supply Continuity: Proceedings of the 52nd ESReDA Seminar Hosted by the Lithuanian Energy Institute & Vytautas Magnus University

Critical Infrastructures Preparedness and Resilience is a major societal security issue in modern society. Critical Infrastructures (CIs) provide vital services to modern societies. Some CIs’ disruptions may endanger the security of the citizen, the safety of the strategic assets and even the governance continuity. The European Safety, Reliability and Data Association (ESReDA) as one of the most active EU networks in the field has initiated a project group on the “Critical Infrastructure/Modelling, Simulation and Analysis – Data”. The main focus of the project group is to report on the state of progress in MS&A of the CIs preparedness & resilience with a specific focus on the corresponding data availability and relevance. In order to report on the most recent developments in the field of the CIs preparedness & resilience MS&A and the availability of the relevant data, ESReDA held its 52nd Seminar on the following thematic: “Critical Infrastructures: Enhancing Preparedness & Resilience for the security of citizens and services supply continuity”. The 52nd ESReDA Seminar was a very successful event, which attracted about 50 participants from industry, authorities, operators, research centres, academia and consultancy companies.JRC.G.10-Knowledge for Nuclear Security and Safet

Initiating Event Analysis of a Lithium Fluoride Thorium Reactor

The primary purpose of this study is to perform an Initiating Event Analysis for a Lithium Fluoride Thorium Reactor (LFTR) as the first step of a Probabilistic Safety Assessment (PSA). The major objective of the research is to compile a list of key initiating events capable of resulting in failure of safety systems and release of radioactive material from the LFTR. Due to the complex interactions between engineering design, component reliability and human reliability, probabilistic safety assessments are most useful when the scope is limited to a single reactor plant. Thus, this thesis will study the LFTR design proposed by Flibe Energy. An October 2015 Electric Power Research Institute report on the Flibe Energy LFTR asked “what-if?” questions of subject matter experts and compiled a list of key hazards with the most significant consequences to the safety or integrity of the LFTR. The potential exists for unforeseen hazards to pose additional risk for the LFTR, but the scope of this thesis is limited to evaluation of those key hazards already identified by Flibe Energy. These key hazards are the starting point for the Initiating Event Analysis performed in this thesis. Engineering evaluation and technical study of the plant using a literature review and comparison to reference technology revealed four hazards with high potential to cause reactor core damage. To determine the initiating events resulting in realization of these four hazards, reference was made to previous PSAs and existing NRC and EPRI initiating event lists. Finally, fault tree and event tree analyses were conducted, completing the logical classification of initiating events. Results are qualitative as opposed to quantitative due to the early stages of system design descriptions and lack of operating experience or data for the LFTR. In summary, this thesis analyzes initiating events using previous research and inductive and deductive reasoning through traditional risk management techniques to arrive at a list of key initiating events that can be used to address vulnerabilities during the design phases of LFTR development

Development of Approaches to Common Cause Dependencies with Applications to Multi-Unit Nuclear Power Plant

The term “common cause dependencies” encompasses the possible mechanisms that directly compromise components performances and ultimately cause degradation or failure of multiple components, referred to as common cause failure (CCF) events. The CCF events have been a major contributor to the risk posed by the nuclear power plants and considerable research efforts have been devoted to model the impacts of CCF based on historical observations and engineering judgment, referred to as CCF models. However, most current probabilistic risk assessment (PRA) studies are restricted to single reactor units and could not appropriately consider the common cause dependencies across reactor units. Recently, the common cause dependencies across reactor units have attracted a lot of attention, especially following the 2011 Fukushima accident in Japan that involved multiple reactor unit damages and radioactive source term releases. To gain an accurate view of a site's risk profile, a site-based risk metric representing the entire site rather than single reactor unit should be considered and evaluated through a multi-unit PRA (MUPRA). However, the multi-unit risk is neither formally nor adequately addressed in either the regulatory or the commercial nuclear environments and there are still gaps in the PRA methods to model such multi-unit events. In particular, external events, especially seismic events, are expected to be very important in the assessment of risks related to multi-unit nuclear plant sites. The objective of this dissertation is to develop three inter-related approaches to address important issues in both external events and internal events in the MUPRA. 1) Develop a general MUPRA framework to identify and characterize the multi-unit events, and ultimately to assess the risk profile of multi-unit sites. 2) Develop an improved approach to seismic MUPRA through identifying and addressing the issues in the current methods for seismic dependency modeling. The proposed approach can also be extended to address other external events involved in the MUPRA. 3) Develop a novel CCF model for components undergoing age-related degradation by superimposing the maintenance impacts on the component degradation evolutions inferred from condition monitoring data. This approach advances the state-of-the-art CCF analysis in general and assists in the studies of internal events of the MUPRA