Wyniki wyszukiwania - BazTech

1

Development of a Bayesian belief network for a boiling water reactor during fault conditions

Frid W., Knochenhauer M., Bednarski M.

Computer Assisted Mechanics and Engineering Sciences

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2005

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Vol. 12, No. 2-3

133-145

EN

This paper describes briefly the development and verification of a probabilistic system for the rapid diagnosis of plant status and radioactive releases during postulated severe accidents in a Boiling Water Reactor nuclear power plant. The probabilistic approach uses Bayesian belief network methodology, and was developed in the STERPS project in the European Union 5-th Euroatom Framework program.

2

Zastosowanie modeli odwrotnych do diagnozowania stanów utraty chłodzenia w reaktorach jądrowych typu wrzącego

Bednarski M., Cholewa W., Frid W.

Diagnostyka

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2002

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Vol. 27

105-110

PL

Celem referatu jest pokazanie przykładu zastosowania odwrotnych modeli diagnostycznych. Zapobieganie wypadkom w elektrowniach jądrowych, w szczególności przypadkom uszkodzenia rdzenia, gdzie ryzyko uwolnienia produktów radioaktywnych jest największe, jest sprawą priorytetową dla bezpieczeństwa. W celu analizy potencjalnie możliwych wypadków, jak również w celu ich zapobiegania oraz zarządzania nimi, stworzono wiele programów symulacyjnych oraz systemów wspomagających podejmowanie decyzji (Computerized Decision Support Systems - CDSS) przez operatorów. Bazują one na metodach deterministycznych i probablistycznych. W przypadku reaktorów jądrowych rozwój szybkich narzędzi symulacyjnych daje możliwość zastosowania metod diagnostycznych bazujących na przykładach. Przedstawione tutaj lokalne modele odwrotne są przykładem takiego właśnie podejścia.

EN

Aim of this paper is to present example of application inverse diagnostic models. Accident prevention in nuclear plants, in particular in case of the core damage, where the risk of release radioactive products is the highest, is the priority cause for safety. In order to analyze potentially possible accidents and also to prevent and to manage them, a lot of simulation codes and Computerized Decision Support Systems (CDSS) was implemented. They base on deterministic and probabilistic methods. In case of nuclear reactors, recent progress of very fast simulation tools opens possibility of applying case-based diagnostic methods. The method described in the paper, which uses local inverse models, is an example of such approach.

3

Nuclear power plant safety assessment and accident management : Practice in Sweden and computerized decision support systems

Frid W.

Computer Assisted Mechanics and Engineering Sciences

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2002

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Vol. 9, No. 1

41-54

EN

The methods of nuclear power plant safety assessment and accident management used in Sweden and selected computerized decision support systems are briefly described. The defense-in-depth strategy, which comprises three elements, namely prevention, protection and mitigation, is essential for keeping the fission product barriers intact. The safety assessment program, which focuses on prevention of incidents and accidents, has three main components: periodic safety reviews, probabilistic safety analysis and analysis of postulated disturbances and accident progression sequences. Even if prevention of accidents is the first priority, it is recognized that accidents involving severe core damage, including core melt, may nevertheless occur. Therefore, measures are required to achieve reasonable capability to manage such accidents. Emergency Operating Procedures and Severe Accident Management Guidelines are the vital components of the Accident Management System. Computerized decision support systems, to be used during normal, disturbed and accident states of a plant are expected to play increasingly important role in safety assessment and accident management, including support in rapid evaluation of possible radioactive releases in the event of a severe accident.

4

Some thermal-hydraulic aspects of reflooding of an overheated, partly degraded nuclear reactor core

Frid W., Miettinen J., Sjöberg A.

Zeszyty Naukowe Politechniki Świętokrzyskiej. Mechanika

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1999

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Z. 67

235--244

EN

In studies of postulated, unmitigated severe accidents in LWRs it is assumed that cooling of the reactor core is lost which leads to core uncovery and heat-up. core degradation and eventually melt-down of the core. Restoring water supply to the reactor core in order to prevent core damage and to terminate accident progression is the main objective of the accident management. In some Boiling Water Reactors it is necessary to consider a possibility that the reactor may become critical during the reflooding phase. The reason is that the melting of the control rods and their relocation from the core is likely to start before the melting and relocation of the fuel. The two-phase flow and heat transfer phenomena associated with reflooding and quenching of the reactor core are complex and not completely understood, especially under severe accident conditions due to high fuel temperatures and. In the case of recriticality, strong thermal-hydraulic transients when the core reaches critical conditions. In this paper, the two-phase flow and heat transfer phenomena during reflooding of an overheated, partly degraded core are discussed. The reflooding model developed for the RECRIT code is briefly described, as well as some validation against experimental data.