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1

Monte Carlo modelling of Th-Pb fuel assembly with californium neutron source

Oettingen M., Stanisz S.

Nukleonika

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2018

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Vol. 63, No. 3

87--91

EN

This paper describes the methodology developed for the numerical reconstruction and modelling of the thorium-lead (Th-Pb) assembly available at the Department of Nuclear Energy, Faculty of Energy and Fuels, AGH University, Krakow, Poland. This numerical study is the first step towards integral irradiation experiments in the Th-Pb environment. The continuous-energy Monte Carlo burnup (MCB) code available on supercomputer Prometheus of ACK Cyfronet AGH was applied for numerical modelling. The assembly consists of a hexagonal array of ThO2 fuel rods and metallic Pb rods. The design allows for different arrangements of the rods for various types of irradiations and experimental measurements. The intensity of the fresh neutron source intended for integral experiments is about 108 n/s, which corresponds to the mass of about 43 μg 252Cf. The source was modelled in the form of Cf2O3-Pd cermet wire embedded in two stainless steel capsules.

2

Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

Oettingen M., Cetnar J.

Nukleonika

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2015

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Vol. 60, No. 3, part 2

571--580

EN

In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB) in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238) and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242). The main results were presented as a calculated-to-experimental ratio (C/E) for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55). The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

3

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.

4

Eastern-design VVER nuclear reactors in European Union

Oettingen M., Stanisz P., Malicki M.

Logistyka

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2015

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nr 4

9537--9551, CD3

EN

In total, 131 nuclear reactors of various types operates in fourteen European Union member states providing about 830 TWh electricity. The main technology used in European Union is the Western-type Pressurized Water Reactor. However, the second most popular technology is the Eastern-type Vodo-Vodyanoi Energetichesky Reaktor-VVER. The five out of fourteen nuclear countries (Bulgaria, Czech Republic, Finland, Hungary and Slovakia) operate 18 VVER reactors with total electricity output of 80 TWh, which corresponds to about 10% share in nuclear electricity generation and to 2.5% share in total electricity generation in European Union. The paper presents the background of the VVER nuclear reactor technology and analysis of its operation history in European Union member states. The analysis was performed using advanced data banks i.e. The Power Reactor Information System and The Country Nuclear Power Profiles of International Atomic Energy Agency, The Information Library of World Nuclear Association and Eurostat of European Commission. The performed research is devoted to the analysis of the electricity supply and performance indicators of VVER nuclear reactors, which also plays an important role for the logistics and transportation in European Union.

PL

Obecnie w czternastu krajach Uni Europejskie(UE) pracuje 131 reaktorów jądrowych różnego typu zapewniających ok. 830 TWh energii elektrycznej. Głównym typem reaktora jadowego używanego w krajach UE jest reaktor wodny ciśnieniowy Typu Zachodniego. Jednak, drugim najbardziej powszechnym typem reaktora jest reaktor wodny ciśnieniowy Typu Wschodniego - VVER (Vodo-Vodyanoi Energetichesky Reaktor). Pięć z czternastu krajów posiadających energetykę jądrową w UE (Bułgaria, Czechy, Finlandia, Węgry oraz Słowacja) posiada 18 reaktorów typu VVER produkujących ok. 80 TWh energii elektrycznej, co odpowiada ok. 10% produkcji energii elektrycznej z procesów jądrowych, jak również 2.5% całkowitej produkcji energii elektrycznej w Uni Europejskiej. Artykuł przedstawia ogólną charakterystykę reaktorów typu VVER jak i historię ich implementacji w krajach Uni Europejskiej. Analiza została przeprowadzona przy użyciu zasobów dostępnych w międzynarodowych bazach danych m.in.: „The Power Reactor Information System” oraz „The Country Nuclear Power Profiles” Międzynarodowej Agencji Energii Atomowej, „The Information Library” Międzynarodowego Stowarzyszenia Jądrowego oraz „Eurostat” Komisji Europejskiej. Przedstawiona analiza skupia się na charakterystyce produkcji energii elektrycznej oraz współczynników wydajności reaktorów VVER, co ma również istotne znaczenie dla logistyki i transportu w Uni Europejskiej.

5

Depletion uncertainties propagation for replica calculations in the lead cooled fast reactor

Stanisz P., Oettingen M., Kępisty G.

Logistyka

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2015

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nr 4

9794--9804, CD3

EN

Most Monte Carlo codes are used to determine certain values with their uncertainty accompanying through stochastic process. Those estimations are crucial information to determine the logistics of frontend and the back-end of nuclear chain supply management. Monte Carlo method simulate physics interactions, where correct results can be obtained if users is running a sufficient number of neutron histories adequately to sample all significant regions of the problem. The code by using internal random walks of neutrons is able to estimate a nuclear parameter k-eff (multiplication factor) and fission source distribution responsible for the ratio of new neutrons generation in the following step. Each neutron generation converges to the fix distribution, which can be characterized by Shannon entropy. Tallies of k-eff and spatial reaction rates starts accumulated information after adjusted cut-off step. However, convergence can stop at some level causing neutron distribution tilt and introducing influence to the reaction rate. Locally slightly different power distribution can occurs resulting in slightly different density evolution of the isotopes. In this paper we apply technics of multi “independent replicas” calculations. The ide based on many simulations of the same system using different random sequences to obtain slightly various solutions which will allows us to build any probability density function. Statistical analysis of the results would allow assessing the uncertainties in the calculated isotopes densities. In this work we examine multi recycle scheme in the fast neutron spectrum based on The Lead-cooled Fast Reactor (LFR) defined and studied at the level of technical design in order to demonstrate its propagation of isotopes evolutions together with uncertainties and highlight systematic errors, due to the number of simulated particles. All simulated aspect has to be considered while performing Monte Carlo burnup simulations.

6

The MCB code for numerical modeling of Fourth Generation nuclear reactors

Oettingen M., Cetnar J., Mirowski T.

Computer Science

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2015

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Vol. 16 (4)

329--350

EN

R&D in the nuclear reactor physics demands state-of-the-art numerical tools that are able to characterize investigated nuclear systems with high accuracy. In this paper, we present the Monte Carlo Continuous Energy Burnup Code (MCB) developed at AGH University’s Department of Nuclear Energy. The code is a versatile numerical tool dedicated to simulations of radiation transport and radiation-induced changes in matter in advanced nuclear systems like Fourth Generation nuclear reactors.We present the general characteristics of the code and its application for modeling of Very-High-Temperature Reactors and Lead-Cooled Fast Rectors. Currently, the code is being implemented on the supercomputers of the Academic Computer Center (CYFRONET) of AGH University and will soon be available to the international scientific community.

7

Use of Monte Carlo Simulation to Optimize the Loading Pattern Using an Ant Colony Algorithm

Stanisz P, Oettingen M.

Logistyka

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2014

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nr 4

4927--4933, CD6

EN

This paper deals with loading pattern optimization that is logistic domain in nuclear reactors. To find the best distribution we created algorithm based on a recent method the Ant Colony Optimization (ACO) algorithm, which is used in transport networks. In our work we used the Monte Carlo methods witch the SERPENT code. This method provided well estimated multi-group cross sections. Our model, which was described by a cross section representation, was handled by the ACO algorithm coupled with the PARCS code. The final result shows convergence of our calculations. Cooperation of these three methods have been determined and presage more detailed study in future. This paper describes the methodology, with some final results obtained by the ACO algorithm through Monte Carlo calculations and Core simulation.

PL

Artykuł dotyczy optymalizacji załadunku paliwa bazującego na obliczeniach Monte Carlo. Optymali-zacja załadunku jest zagadnieniem logistycznym. Do znalezienia najlepszego rozkładu, stworzyliśmy własny algorytm bazujący na obiecujących wynikach uzyskiwanych za pomocą Algorytmu Kolonii Mrówek (AKM) szeroko wykorzystywanego w transporcie. Do otrzymania odpowiednich wyników wspiera-liśmy się programem Monte Carlo SERPENT, dzięki któremu otrzymaliśmy wielogrupowe przekroje czynne dla określonych kaset paliwowych. Następnie nasz model, był wykorzystywany przez program PARCS i zarządzającym nim, napisanym przez nas programem. Ostateczne rezultaty potwierdzają asymptotyczną zbieżność naszych wyników. Została osiągnięta współpraca trzech metod obliczeniowych. Artykuł przedstawia metodologię wraz z niektórymi rezultatami otrzymanymi za pomocą wyżej wymienionymi programami.

8

Monte Carlo Modelling of Loosely Coupled Fissionable Systems

Oettingen M., Kępisty G.

Logistyka

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2014

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nr 4

4731--4739, CD6

EN

In the paper we describe problems related to the convergence diagnostics in the Monte Carlo modeling of the loosely coupled fissionable systems like arrays of the spent fuel elements. The logistics and trans-portation of the spent nuclear fuel is a complex process due to its high radioactivity and thus contamination risk for the biosphere and human beings. The detection of convergence is important for the accurate estimation of system multiplication factor and associated standard deviation. The multiplication factor is the main safety parameter used for design and optimization of the equipment for spent fuel handling and transportation in order to fulfill the international safety and security standards.

PL

Niniejsza praca przedstawia metody detekcji zbieżności w symulacjach Monte Carlo niezwiązanych systemów jądrowych takich jak zestawy kaset z wypalonym paliwem jądrowym. Logistyka oraz transport zużytego, radioaktywnego paliwa jądrowego jest skomplikowanym procesem charakteryzującym się potencjalnym ryzykiem skarżenia ekosfery i populacji ludzkiej. Detekcja zbieżności jest niezbędna przy estymowaniu współczynnika mnożenia neutronów oraz jego niepewności. Współczynnik mnożenia neu-tronów jest podstawowym parametrem wykorzystywanym do optymalizacji urządzeń do transportu oraz manewrowania zużytym paliwem jądrowym w celu spełnienia międzynarodowych wymagań bezpieczeństwa.

9

Symulacje Monte Carlo fizyki rdzenia reaktora jądrowego typu PWR

Oettingen M.

Logistyka

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2013

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nr 4

PL

Tematem niniejszej pracy jest modelowanie Monte Carlo fizyki rdzenia rektora jądrowego na poziomie kasety paliwowej. Symulacje numeryczne transportu neutronów oraz zmiany składu paliwa na skutek transmutacji i rozpadów promieniotwórczych zostały przeprowadzone za pomocą kodu MCB (The Monte Carlo Continuous Energy Burnup Code). Model numeryczny opracowany w celu przeprowadzenia symulacji Monte Carlo został zbudowany na podstawie geometrii oraz składu materiałowego kasety paliwowej typu 17x17 używanej w rektorach wodnych ciśnieniowych PWR (Pressurized Water Reactor). Kaseta zawiera czyste paliwo uranowe jak i paliwo z dodatkiem wypalającej się trucizny – Gd2O3. Obecność wypalającej się trucizny istotnie wpływa na charakterystyki kasety paliwowej w polu neutronów, co zostało poddane analizie w zaprezentowanym artykule. Głównymi parametrami otrzymanymi w symulacji numerycznej są: reaktywność układu, wypalenie paliwa oraz ewolucje wybranych nuklidów, takich jak U235, Pu239, Pu241, Gd155 oraz Gd157. Wyniki symulacji numerycznej przeprowadzonej przy pomoc kodu MCB są zgodne z prawidłowościami fizyki rdzenia rektorów jądrowych typu PWR. Opracowana metodologia symulacji rdzenia reaktora jądrowego PWR na poziomie kasety paliwowej z dużą wiarygodnością odzwierciedla rzeczywiste zachowanie systemu.

EN

The Study focuses on the Monte Carlo modelling of the nuclear reactor core at the level of the fuel subassembly. The simulations of neutron transport and fuel depletion due to the nuclear transmutations and decays were performed using The Monte Carlo Continuous Energy Burnup Code – MCB. The numerical model developed for the Monte Carlo simulation was built using the engineering geometry and material composition of the 17x17 fuel subassembly for Pressurized Water Reactor (PWR). The 17x17 fuel subassembly contains pure uranium fuel as well as Gd2O3 bearing fuel. The Gd2O3 burnable poison significantly influences the characteristics of the fuel subassembly in the neutron field, which was shown in the paper. The main parameters obtained in the simulation are: system reactivity, fuel burnup and evolutions of the U235, Pu239, Pu241, Gd155 and Gd157. The results of the numerical simulation performed with the MCB code show good agreement with the theoretical predictions of the nuclear reactor physics. The developed methodology for the numerical simulation of the PWR core at the level of the fuel subassembly with high accuracy reflects the reality.

10

Depletion analysis of the HELIOS experiment using the MCB code

Oettingen M., D'Agata E., Döderlein C., Tuček K., Cetnar J.

Nukleonika

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2012

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Vol. 57, No. 4

435-441

EN

The focus of our studies is to present an advanced depletion analysis of the HELIOS experiment by means of the Monte Carlo continuous energy burn-up code (MCB). The MCB was used mainly to calculate nuclide density evolution in nuclear reactor cores. We present the capability of the MCB to investigate the depletion of nuclear fuel samples irradiated in the HELIOS experiment. In our studies we traced the behaviour of the main fissile isotopes, 242mAm and 239Pu, respectively. We also perform a sensitivity analysis to the choice of JEF2.2 and JEFF3.1 cross section libraries in terms of the released fission power and the evolution of actinide inventories. The amount of He produced at the end of irradiation, as well as Am and Pu depletion, were also considered.

11

On the neutronics of European lead-cooled fast reactor

Cetnar J., Oettingen M., Domańska G.

Nukleonika

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2010

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Vol. 55, No. 3

317-322

EN

The perspective of nuclear energy development in the near future imposes a new challenge on a number of sciences over the world. For years, the European Commission (EC) has sponsored scientific activities through the framework programmes (FP). The lead-cooled fast reactor (LFR) development in the European Union (EU) has been carried out within European lead-cooled system (ELSY) project of the 6th FP of EURATOM. This paper concerns the reactor core neutronic and burn-up design studies. We discuss two different core configurations of ELSY reactor; one loaded with the reference – mixed oxide fuel (MOX), whereas the second one with an advanced fuel – uranium- -plutonium nitride. Both fuels consist of reactor grade plutonium, depleted uranium and additionally, a fraction of minor actinides (MA). The fuel burn-up and the time evolution of the reactor characteristics has been assessed using a Monte Carlo burn-up code (MCB). One of the important findings concerns the importance of power profile evolution with burn-up as a limiting factor of the refuelling interval.