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Radionuclide neutron source trajectories in the closed nuclear fuel cycle

Stanisz Przemysław, Cetnar Jerzy, Oettingen Mikołaj

Nukleonika

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2019

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Vol. 64, No. 1

3--9

EN

The highest efficiency in the usage of nuclear energy resources can be implemented in fast breeder reactors of generation IV. It is achieved thanks to the ability of consuming minor actinides (MAs) in energy production. One of the options to use this benefit is full recycling of MAs to close the nuclear fuel cycle. Monte Carlo burn up (MCB), an integrated burn-up calculation code, deals with the complexity of the burn-up process which is applied to the European Lead-cooled Fast Reactor (ELFR). MCB uses continuous energy representation of cross section and spatial effects of full core reactor model; however, it automatically calculates nuclide production in all possible reactions or decay channels. Multi-recycling of MAs can cause an intensified build-up of curium, berkelium and californium. Some of their isotopes are strong neutron emitters from spontaneous fission, which hinders fuel recycling. The implementation of a novel methodology for trajectory period folding allows us to trace the life cycle of crucial MAs from the beginning of the reactor life towards the state of adiabatic equilibrium. The result of the analysis performed is presented, showing the sources of strong contribution to the neutron production rate. The parametric sensitivity analysis method for selected nuclide reactions is applied, revealing sensitivity of transmutation chains for the production of neutron emitter isotopes.

2

Bezpieczne sterowanie napędów pneumatycznych - zagadnienia wybrane

Trajdos M.

Utrzymanie Ruchu

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2018

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Nr 2

73--75

PL

W artykule omówiono wybrane zagadnienia dotyczące bezpiecznego sterowania napędami pneumatycznymi, tj. przygotowanie sprzężonego powietrza czy sterowanie siłownikiem.

3

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.

4

Comparison of simple design of sodium and lead cooled fast reactor cores

Mazgaj P., Darnowski P., Gurgacz S., Lipka M., Dziubanii K.

Journal of Power Technologies

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2014

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Vol. 94, nr 5

16--25

EN

This report presents the results of a numerical simulation of thermal hydraulics processes in a liquid metal cooled fast reactor core, combined with simple neutron population computing for an infinite pin cell lattice. Two types of coolant were studied: liquid sodium and liquid lead, with all requirements regarding safety conditions observed. Temperature distributions along the cooling channel and distributions in the radial direction were prepared, then criticality calculations were performed for MOX fuel using MCNP Monte Carlo code.

5

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.