Computer trading and systemic risk: a nuclear perspective

Financial markets have evolved to become complex adaptive systems highly reliant on the communication speeds and processing power afforded by digital systems. Their failure could cause severe disruption to the provision of financial services and possibly the wider economy. In this study we consider whether a perspective from the nuclear industry can provide additional insights

Conjugating ALARA, BEPU, Safety Margins and Independent Assessment in Nuclear Reactor Safety

ALARA (As-Low-As-Reasonably-Achievable) is an early principle in Nuclear Reactor Safety, NRS (Nuclear Reactor Safety): Designers and Operators must do their best to minimize doses to the humans. BEPU (Best Estimate Plus Uncertainty) is an approach in Accident Analysis, part of NRS: one may state that BEPU implies the best use of computational tools to determine the safety of nuclear installations. Then, ALARA may be seen at the origin of BEPU, or ALARA is at the origin of BEPU. Furthermore, BEPU (and BEPU elements like V & V, Scaling, procedures of code application and code coupling, etc.) can be extended to all analytical parts of the Final Safety Analysis Report (FSAR). This brings to BEPU-FSAR. Safety Margin (SM) is an established concept in NRS: a few dozen SM values must be calculated in current safety analyses and demonstrated to be acceptable. The SM concept can be extended to everything part of the design, the operation and the environment for a Nuclear Power Plant (NPP) Unit. Here the environment includes the personnel in charge of activities connected with the NPP. The Extended SM concept, E-SM, implies the formulation of some ten-thousands SM values, which shall correspond to a similar number of monitored variables. Reasons for E-SM are the examples in section 4.1. Independent Assessment (IA) is an early requirement in NRS: data ownership and system complexity prevented so far a comprehensive application of the requirement. IA analyses conflict with industry policies to keep proprietary data. IA based BEPU-FSAR analyses are essential to finalize the E-SM design. In the paper we discuss that: a) ALARA is at the origin of BEPU; b) BEPU-FSAR analyses are the natural origin of E-SM values; c) The implementation of E-SM equals to introducing an additional physical barrier against the release of fission products

Prioritizing Human Factors in Emergency Conditions Using AHP Model and FMEA

One of the most critical issues related to safety in industrial plant is to manage accidents that occur in industries. In general, the causes of accidents are twofold: the presence of dangerous equipment and human errors. The aim of this study is to propose a novel approach to ensure safety in emergency conditions in industrial plant considering both of these factors. The proposed idea aims to integrate the human reliability analysis (HRA) and the failure modes and effects analysis (FMEA). The human errors and failure modes are categorized using a multicriteria approach based on analytic hierarchy process (AHP). The final aim is to present a novel methodological approach based on AHP to prioritize actions to carry out in emergency conditions taking into account both qualitative and quantitative factors. A real case study is analyzed. The analysis allowed to identify possible failure modes connected with human error process

Operation of Ageing Reactors: Approaches and associated Research in the European Union

Plant Life Management (PLiM) of existing nuclear power plants in EU may consider longer term operability. This becomes an option with specific advantages and is under consideration in some Member States. It however unveils number of challenges that are closely related with extending the operational life of structures and components (SC) beyond the established operational time frame at the time they were designed. Safety related aspects of long term operations are obvious; for those SC important to safety that are selected for evaluation it is necessary to demonstrate that they perform the intended safety functions with sufficient safety margins for the entire period of operations. The ageing phenomena needs to be timely and carefully considered, in particular by structural integrity assessment, accident analysis, nuclear power plant ageing assessment and mitigation, systems interactions and risk assessment with related human factor aspects. A broad and effective dissemination of related scientific results is a further objective. The European Union, which sees a large diversity of nuclear plant types, needs a targeted investigation to upgrade knowledge on their objective safety levels. Research activities therefore concentrate on providing the scientific and technical knowledge in relation to safety important issues needed for Community policy support and for helping to enhance nuclear safety in EU and beyond. The European Commission launched within the EURATOM framework programme FP7 a research and support programme related to PLiM issues under progress at Joint Research Centre and involves partners from nuclear industry and Technical Support Organizations through several dedicated networks and projects. This report aims to provide overview of approaches proposed or followed in the USA and in EU countries when longer term operability (LTO) is considered as part of PLiM. A special attention is given to discussing existing regulatory framework available, as well as requirements set for ageing reactors in the corresponding IAEA safety reports and safety guides. A comparison of the US Licence Renewal Rule and Periodic Safety Review as a tool for assessment of Structure, Systems and Components (SSC) for PLiM and LTO is provided too. The present report also discuses several current challenges, and shows some examples how the research is supporting / or can support the safety assessment of ageing nuclear power plants in the European Union.JRC.F.5-Nuclear Reactor Safety Assessmen

EJP-CONCERT. D3.7 Second joint roadmap for radiation protection research

EJP-CONCERT Work Package 3, Deliverable 3.7

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

PENILAIAN SAFETY INTEGRITY LEVEL IRADIATOR GAMMA KATEGORI-IV PADA KEGAGALAN SISTEM CRANE PENGONTROL SUMBER ZAT RADIOAKTIF

PENILAIAN SAFETY INTEGRITY LEVEL IRADIATOR GAMMA KATEGORI-IV PADA KEGAGALAN SISTEM CRANE PENGONTROL SUMBER ZAT RADIOAKTIF. Penelitian penilaian Safety Integrity Level (SIL) telah dilakukan di  fasilitas iradiasi kategori-IV pada sistem crane pengontrol pergerakan sumber zat radioaktif Cobalt-60. Tujuan dari penelitian ini adalah untuk menentukan  tingkatan SIL pada iradiator gamma kategori-IV pada simulasi kegagalan sistem pergerakan crane sumber zat radioaktif. Safety integrity level merupakan cara untuk menunjukkan tingkat kegagalan yang masih dapat diterima dari fungsi keselamatan tertentu. Metode pendekatan yang digunakan pada penelitian ini adalah metode probabilistic risk assessment (PRA) dan initial protection layer (IPL). Nilai frekuensi risiko terhadap paparan sumber zat radioaktif Cobalt-60 saat instalasi iradiator gamma beroperasi adalah bernilai 1,7 x 10-2 fail/year. Nilai sil dari simulasi kegagalan yang dilakukan adalah SIL-4 dengan nilai frekuensi kegagalan sebesar 7x10-7 dengan mengoptimalkan sistem perlindungan high alarm dan interlock systems. Presentase pengurangan risiko yang didapatkan adalah sebesar 99,99%