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Analysis of AP1000 radioactive material release accidents with MELCOR Accident Consequence Code System (MACCS)

Niewiński G., Stępień M., Góral K.

Journal of Power Technologies

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2017

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Vol. 97, nr 5

446--454

EN

The probable introduction in the medium term of nuclear energy into the Polish national power system has become a source of anxiety in society. While Poland already has a research nuclear reactor (acronym: MARIA) at the National Center for Nuclear Research in Świerk, near Warsaw, issues regarding safety and the possible consequences of an accident in the first baseload nuclear power plant have triggered public debate. As part of the licensing process of any newly designed reactor, scenarios for a range of accidents at the plant together with their consequences must be modeled, analyzed and presented in the licensing documentation. In this context a model was built based on a complex set of data - including data provided by the reactor manufacturer, location and environmental data, weather conditions and possible accident scenarios to perform simulations with a computational tool called MELCOR Accident Consequence Code System (MACCS). MACCS is used to perform accident-related calculations, including release of radioactive material to the atmosphere and short and longterm consequences. The analysis involved releases of radioactive material from an AP1000 nuclear reactor assumed to be located on the Polish seacoast and demonstrates that the lethality and incidence of cancer caused by radioactive release are significantly lower than natural.

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Safety analysis of coolant needed supply for Spent Fuel Pool in light of Fukushima Accident

Malicki M., Skolik K., Oettingen M.

Logistyka

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2015

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nr 4

9450--9457, CD3

EN

The Fukushima accident shows us that not only the core and reactor could make problems during unexpected events but also Spent Fuel Pool (SFP). That accident encouraged many experts to reconsider safety features in this area of Nuclear Power Plants (NPP) and to be more mindful of this potential problem. Preparing precise analysis of such accidents could provide important information about possible consequences and bring up essential solutions about how to improve SFP fuel management and safety systems related with the fuel storage process. This paper delivers analysis based on the Fukushima SFP unit 4 accident from March 11th 2011. The Fukushima type accident was caused by a lack of heat reception: water vaporization was the only way for heat to escape from SFP. Critical to avoid serious consequences in that situation is to know when and how much water must be provided by the operator to the SFP to ensure the assembly is submerged into a coolant. During this accident the SFP was almost full, 1530 of 1560 spots were taken and instruments, safety or safety-related systems like heat exchangers were not available.

3

Analysis of melcor code utility for safety assessment of high-temperature reactors

Skolik K., Malicki M., Stanisz P.

Logistyka

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2015

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nr 4

9739--9745, CD3

EN

Small Modular Reactors (SMR) are probably one of the solutions to world’s nuclear energy problems. They could be cheaper than classical Nuclear Power Plants (NPP) and they could provide diversification of power production. High Temperature Reactors (HTR) are of interest for big companies with huge energy consumption as a fairly inexpensive and relatively independent source of power. According to the designers, in the future it will be possible to place one of the SMRs inside a factory or very close to the city. Before it happens it is necessary to conduct a lot of analysis which can prove that this concept is safe. The aim of this paper is to describe one possible way to assess safety features by using one of the best computer codes for severe accident analysis, MELCOR. The authors try to assess if existing computer codes give us a tool to create proper model of HTR and simulate its failure. The next question is what are the advantages and disadvantages that characterize Small Modular Reactors.

PL

Małe reaktory modułowe mogą być przyszłością w produkcji energii. Mogą rozwiązać wiele problemów związanych zarówno ze zwiększeniem wymagań społecznych co do konsumpcji energii jak i dywersyfikacji dostaw w miejsca trudno dostępne i dla przemysłu. Ze strony przemysłu, potencjalnych użytkowników reaktorów HTR określenie warunków lokalizacji reaktora jest jednym z najistotniejszych zagadnień. Po pierwsze reaktor aby był użyteczny musi być w bezpośrednim sąsiedztwie instalacji przemysłowych, po drugie nie może stwarzać zagrożeń dla nich, ani te instalacje nie mogą zmniejszać bezpieczeństwa samego reaktora. Dlatego też celem niniejszego opracowania jest zbadanie użyteczności kodu MELCOR powszechnie wykorzystywanego w energetyce jądrowej do określenia globalnych warunków bezpieczeństwa. Służy on do analiz awarii ciężkich w reaktorach jądrowych. Ze względu iż jest on ciągle rozwijany pozwala również na modelowanie niekonwencjonalnych lub mało popularnych reaktorów jądrowych. HTR to przykład takiego właśnie nowatorskiego podejścia do energetyki jądrowej. MELCOR dzięki ciągłemu rozwojowi zarówno istniejących już modeli, jak i poszerzania jego możliwości pozwala również na analizę takiego właśnie pryzmatycznego, wysokotemperaturowego reaktora, chłodzonego helem.

4

Computer codes in the safety analysis for nuclear power plants. Computational capabilities of thermal-hydraulic tools, using the example of the RELAP5 code

Skrzypek E., Skrzypek M.

Journal of Power Technologies

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2014

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Vol. 94, nr 5

41--50

EN

Safety is a paramount concern of the Nuclear Power Program in Poland. To this end there is a need to investigate the design of the proposed reactor and its operation principles and perform multiple analyses both before the reactor start-up (The Pre-Construction Safety Report (PCSR) and during its operational life. In the worldwide nuclear community hundreds of people are involved in this complicated and complex process. Due to the sophistication of the phenomena occurring during operation and accidents, the number of analyses is increasing rapidly. Currently, much interest in this field is focused on the use of computer codes and high computational power.