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Transmutation: reducing the storage time of spent fuel

Mazgaj P., Darnowski P.

Journal of Power Technologies

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2014

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Vol. 94, nr 5

27--34

EN

Transmutation can reduce the storage time of spent fuel. Efficient transmutation requires a high flux of neutrons and can therefore be done only in nuclear reactors. The article shows the concepts of different solutions of transmutation in nuclear reactors. Knowledge of transmutation is supplemented by information on spent fuel and its radiotoxicity.

2

Simulation and analysis of pipe and vessel blowdown phenomena using RELAP5 and TRACE

Gurgacz S., Pawluczyk M., Niewiński G., Mazgaj P.

Journal of Power Technologies

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2014

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Vol. 94, nr 5

61--68

EN

This paper presents the results of a pipe and vessel theoretical blowdown analysis in order to identify and assess rapid depressurization phenomena occurring during a Loss of Coolant Accident in a nuclear power plant. Calculations were performed in conditions similar to those in a reactor pressure vessel. Due to critical two-phase flow occurrence, RELAP5 mod 3.3. and TRACE v 3.0 system codes were used for this purpose and a computational model was built on the basis of available literature and implemented in MATLABr code. Finally, pipe depressurization results were compared to experimental results taken from the literature.

3

Development of a measurement and reconstruction system for determining the phase distribution in a two-phase flow vertical tube using Electrical Impedance Tomography

Gatkowski M., Buchner T., Niewiński G., Mazgaj P.

Journal of Power Technologies

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2014

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Vol. 94, nr 5

77--84

EN

This paper is based on experience gained during six months of work on Electrical Impedance Tomography (EIT) at the Areva NP test facility in Erlangen, Germany. The aim of this work was to build an EIT system capable of providing fairly accurate imaging of the resistivity field in a vertical pipe with two-phase flow. The described system will be used to analyze two-phase flow in the PKL Integral Test Facility. EIT is an imaging method in which electrical current is injected through electrodes into a volume of interest and the potential distribution obtained is measured by surface electrodes. Using these measurements an image of the electrical conductivity/resistivity can be reconstructed in the volume of interest using numerical techniques. The reconstruction is diffcult because the mathematical problem is ill-posed and non-linear, which implicates high computational intensiveness.

4

Design of a sub-critical reactor for transmutation of higher actinides

Mazgaj P., Jimenez D., Lopez D.

Journal of Power Technologies

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2014

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Vol. 94, nr 5

35--40

EN

To reduce the storage time of spent fuel its radiotoxicity must be reduced through the transmutation of minor actinides. Generation IV reactors or sub-critical Accelerator Driven Systems (ADS) are options that appear technically feasible in view of the transmutation goal. The purpose of this work is to design a lead-cooled 400 MWth sub-critical reactor which uses inert matrix nitride fuel. Many design issues are dealt with, such as the lifetime of the cladding material, the minimum reactivity swing fuel composition, the burning rates of MA and the safety coefficients. The results obtained show that ADS performs well in transmuting minor actinides and has good safety levels in respect of reactivity perturbations.

5

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.