Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask

Kerntechnik ◽  
2020 ◽  
Vol 85 (1) ◽  
pp. 38-53
Author(s):  
M. J. Leotlela ◽  
I. Petr ◽  
A. Mathye
Keyword(s):  
Multiplication Factor ◽  
Spent Fuel ◽  
Water Ingress ◽  
Neutron Multiplication

Burnup Credit in the Criticality Safety Analysis of Spent Fuel in Storage Systems

2014 ◽  
Author(s):  
Vladyslav Soloviov
Keyword(s):  
Isotopic Composition ◽  
Nuclear Fuel ◽  
Beta Decay ◽  
Spent Nuclear Fuel ◽  
Storage Systems ◽  
Multiplication Factor ◽  
Spent Fuel ◽  
Neutron Multiplication ◽  
Initial Fuel ◽  
Effective Neutron

In this paper accounting of spent nuclear fuel (SNF) burnup of RBMK-1000 with actinides and full isotopic composition has been performed. The following characteristics were analyzed: initial fuel enrichment, burnup fraction, axial burnup profile in the fuel assembly (FA) and fuel weight. As the results show, in the first 400 hours after stopping the reactor, there is an increase in the effective neutron multiplication factor (keff) due to beta decay of 239Np into 239Pu. Further, from 5 to 50 years, there is a decrease in keff due to beta decay of 241Pu into 241Am. Beyond 50 years there is a slight change in the criticality of the system. Accounting for nuclear fuel burnup in the justification of nuclear safety of SNF systems will provide an opportunity to increase the volume of loaded fuel and thus significantly reduce technology costs of handling of SNF.

Burnup Credit in the Criticality Safety Analysis of Spent Fuel in Transportation and Storage Systems

2013 ◽  
Author(s):  
Vladyslav Soloviov
Keyword(s):  
Nuclear Fuel ◽  
Beta Decay ◽  
Spent Nuclear Fuel ◽  
Storage Systems ◽  
Multiplication Factor ◽  
Spent Fuel ◽  
Neutron Multiplication ◽  
Initial Fuel ◽  
Effective Neutron ◽  
And Storage

In this paper accounting of spent nuclear fuel (SNF) burnup of RBMK-1000 only with actinides has been performed. The following characteristics were analyzed: initial fuel enrichment, burnup fraction, axial burnup profile in the fuel assembly (FA) and fuel weight. As the results show, in the first 400 hours after stopping the reactor, there is an increase in the effective neutron multiplication factor (keff) due to beta decay of 239Np into 239Pu. Further, from 5 to 50 years, there is a decrease in keff due to beta decay of 241Pu into 241Am. Beyond 50 years there is a slight change in the criticality of the system. Accounting for nuclear fuel burnup in the justification of nuclear safety of SNF systems will provide an opportunity to increase the volume of loaded fuel and thus significantly reduce technology costs of handling of SNF.

scholarly journals Features of Burnup Calculations in the SERPENT 2 Software Package Using the Example of a BWR Fuel Assembly

2018 ◽  
Vol 3 (3) ◽  
pp. 182
Author(s):  
Pham Bui Dinh Lam ◽  
Kolesov V.V.
Keyword(s):  
Fuel Assembly ◽  
Software Package ◽  
Multiplication Factor ◽  
Spent Fuel ◽  
Detailed Model ◽  
Fuel Assemblies ◽  
Neutron Multiplication ◽  
Phase Iiib ◽  
Nuclide Composition

In this paper, we used the data from “OECD/NEA Burnup Credit Criticality Benchmark Phase IIIB: Nuclide Composition and Neutron Multiplication Factor of BWR Spent Fuel Assembly” ([1]) for the verification of the SERPENT 2 code. The results obtained which were compared with the results of other authors, which were also given in “OECD/NEA Burnup Credit Criticality Benchmark Phase IIIB: Burnup Calculations of BWR Fuel Assemblies for Storage and Transport” ([2]). Investigations of the influence of the detailed model of pins and pins with gadolinium, as well as various methods of burn-up calculations were also carried out.

Long-Term In-Package Spent Fuel Criticality Calculations

2003 ◽  
Author(s):  
Olivier Wantz ◽  
Olivier Smidts ◽  
Alain Dubus ◽  
R. Beauwens
Keyword(s):  
Multiplication Factor ◽  
Spent Fuel ◽  
Mox Fuel ◽  
Fuel Assemblies ◽  
Neutron Multiplication

This paper presents MCNP criticality calculations for both UOX and MOX disrupted fuel assemblies canisters systems in the reference Belgian disposal concept and one of its variant. We examine the influence of different parameters (water moderation and geometry alteration) on the neutron multiplication factor, keff. In all the studied cases, the reference concept does not present criticality risks. The variant concept sometimes presents criticality risks. The present results only concern fresh UOX and MOX fuel assemblies. Further developments of this work will include irradiated (UOX and MOX) fuels.

scholarly journals INTRODUCING BURNUP CREDIT FOR EXPANSION STAGE 2 OF THE EXTERNAL SPENT FUEL POOL AT GÖSGEN NPP

2021 ◽  
Vol 247 ◽  
pp. 17007
Author(s):  
Axel Hoefer ◽  
Martin Basler ◽  
Oliver Buss ◽  
Gaëtan Girardin ◽  
Fabian Jatuff ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Multiplication Factor ◽  
Spent Fuel ◽  
Statistical Confidence ◽  
Fuel Assemblies ◽  
Neutron Multiplication ◽  
Expansion Stage ◽  
Spent Fuel Pool ◽  
Storage Racks ◽  
Criticality Safety

We present a summary of the actinide-plus-fission-product burnup credit criticality safety licensing analysis for Expansion Stage 2 (ES2) of the external spent fuel pool at Gösgen nuclear power plant. In ES2, the nine Expansion Stage 1 storage racks currently installed in the external spent fuel pool are going to be supplemented by nine ES2 storage racks with a significantly reduced fuel assembly pitch. They are designed for loadings with fuel assemblies above a well-defined minimum required burnup. The objective of the criticality safety analysis is to calculate the minimum required burnup values for the uranium and MOX fuel assemblies to be stored in the ES2 storage racks. We use a methodology that allows us to take into account the reactivity effects due to variabilities and uncertainties of all relevant parameters involved in a burnup credit criticality safety assessment in a bounding manner. These include manufacturing tolerances of the fuel assemblies and storage racks, the irradiation histories and burnup profiles of the spent fuel assemblies, the bias of the depletion code used to calculate the isotopic inventories of the irradiated fuel, and the bias of the criticality code used to calculate the neutron multiplication factor of the considered storage configuration. A combination of different statistical procedures is used to evaluate and propagate the uncertainty information on the input parameters and translate it into statistical confidence statements about the neutron multiplication factor. It should be noted that the presented analysis is related to the first implementation of a significant burnup credit for wet storage of PWR fuel in Switzerland.

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