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The PSA analysis of PWR emergency coolant injection availability following SBLOCA

Borysiewicz M., Bronowska K., Kopka P., Kowal K., Kwiatkowski T., Prusiński A. M., Prusiński P. A., Siess G.

Nukleonika

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2013

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Vol. 58, No. 2

307--316

EN

The aim of this article is to briefly introduce the probabilistic safety assessment (PSA) of nuclear power plants (NPPs), its scope, main concepts and application to a real case. The results of analysis presented here have been obtained by the Probabilistic Safety Analysis Group (GPSA) at the National Centre for Nuclear Research (NCBJ, Otwock) as a part of the work done for the Polish National Atomic Energy Agency (NAEA). As a reference, NPP Surry Unit 1 (USA), equipped with 800 MWe Westinghouse triple-loop PWR (pressurized water reactor), has been chosen. The emergency coolant injection (ECI) function availability following the small break loss of coolant accident (SBLOCA) was thoroughly analyzed. The approach and data, which were adopted for the selected part of the SBLOCA sequences, were those used in the U.S. NRC Reactor Safety Study (WASH-1400). As a result of this study, the SBLOCA event tree, including ECI systems, i.e. high pressure injection system (HPIS) and auxiliary feedwater system (AFWS) reliability models, was developed and quantified. The probability of each accident sequence was evaluated using Saphire v.8, the PSA software by U.S. NRC. The choice of the software was based on earlier PSA software study. The failure probability of at least one of the considered safety systems – P(FAIL) is equal to 5.76E-3 and the most pessimistic accident branch (unavailability of both HPIS and AFWS) is about 0.05% of P(FAIL). These results were obtained based on assumption that the SBLOCA has occured. The most significant failure components are those corresponding to charging pumps unavailability, loss of electric power and human errors.

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An integrated risk informed decision making in the nuclear industry

Borysiewicz M., Kowal K., Prusiński P. A., Dąbrowski M.

Informatyka, Automatyka, Pomiary w Gospodarce i Ochronie Środowiska

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2013

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nr 2

22--34

EN

The regulatory body, established to ensure safety of nuclear facilities, is expected to make right decisions and provide appropriate regulations for the nuclear industry. The traditional manner of its activity has been based on a deterministic approach to safety analyses. However, increased maturity of Probabilistic Safety Assessment (PSA) makes it complementary to deterministic studies. The new IAEA concept, described in this article, is to apply an integrated approach by combining both deterministic and probabilistic insights with other requirements affecting the decision making process.

PL

Organ regulacyjny, powołany w celu zapewnienia bezpieczeństwa jądrowego, jest odpowiedzialny za podejmowanie decyzji i wprowadzanie rozporządzeń dla przemysłu jądrowego. Tradycyjny sposób jego funkcjonowania opiera się na deterministycznym podejściu do analiz bezpieczeństwa. Rozwój analiz probabilistycznych (PSA) sprawia jednak, iż są one traktowane jako podejście komplementarne. Nowa koncepcja IAEA, opisana w tym artykule, polega na zintegrowanym podejściu, uwzględniającym analizy deterministyczne, probabilistyczne i inne aspekty procesu decyzyjnego.

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An application IRIDM the decision making process on fuel conversion of the MARIA reactor

Borysiewicz M., Kowal K., Prusiński P. A., Dabrowski M.

Informatyka, Automatyka, Pomiary w Gospodarce i Ochronie Środowiska

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2013

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nr 3

3--6

EN

Poland, when acceded to GTRI (Global Threat Reduction Initiative) in 2004, has committed to convert the nuclear fuel of the Research Reactor MARIA, operated by the National Centre for Nuclear Research (NCBJ) in Świerk. The conversion means giving up of high enriched uranium fuel containing 36% of U-235, which was used so far, and replacing it with the low enriched uranium fuel (19.7% U-235). This article describes the potential usability of the Integrated Risk Informed Decision Making (IRIDM) methodology in optimization of the fuel conversion procedure.

PL

Polska, przystępując w 2004 roku do programu GTRI (Inicjatywa Redukcji Zagrożeń Globalnych), zobowiązała się do konwersji paliwa jądrowego w reaktorze badawczym MARIA, eksploatowanym przez Narodowe Centrum Badań Jądrowych (NCBJ) w Świerku. Konwersja ta oznacza rezygnację z dotychczas użytkowanego paliwa, zawierającego 36% U-235 i zastąpienie go paliwem nisko wzbogaconym (19.7% U-235). Niniejszy artykuł opisuje potencjalne zastosowanie zintegrowanego procesu decyzyjnego (IRIDM) w optymalizacji procedury konwersji paliwa.

4

CFD analysis of the safety related thermal hydraulic parameters describing a flow domain of an experimental medical installation (BNCT converter) inside of the Research Reactor MARIA

Prusiński P. A., Potempski S., Borysiewicz M., Kowal K., Kwiatkowski T., Prusiński A. M.

Journal of Power Technologies

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2012

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Vol. 92, nr 4

227--240

EN

The Boron-Neutron Capture Therapy (BNCT) is an experimental radiotherapy technique used to treat the most aggressive types of brain tumors that cannot be surgically removed from the human body. To date, clinical trials of BNCT have been initiated at only a handful of reactors around the world, but advanced studies on BNCT are still being carried out in numerous research centers where the suitable or convertible reactors are available. Construction of BNCT facilities is justified only at some existing reactors. Others can possibly be adapted for BNCT by using fission converters to modify the energy spectrum of the primary neutron beam, which makes it useful for treatment purposes. The BNCT converter, designed for use in the MARIA research reactor at the National Centre for Nuclear Research [W1] (NCBJ) in Świerk near Warsaw, Poland, consists of 99 fuel rods (containing low-enriched uranium) inside of the aluminum box. Since its installation affects the core layout and possibly may affect the normal operating regime of the reactor, additional safety analyses must be performed to prove the existence of sufficient safety margins. In this study modern Computational Fluid Dynamics (CFD) techniques have been applied to assess the maximum temperature of the rod wall surfaces, the temperature difference between the inlet and outlet of the converter channel, as well as the maximum and average velocity of the fluid and to compare them with the results presented in the reference analytical study.