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2 2015

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Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

Stanisz P., Cetnar J., Domańska G.

Nukleonika

2015

Vol. 60, No. 3, part 2

581--590

The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR) was defi ned and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System) and LEADER (Lead-cooled European Advanced Demonstration Reactor) projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA) are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs), and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB) code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

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

Stanisz P., Oettingen M., Kępisty G.

Logistyka

2015

nr 4

9794--9804, CD3

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.