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 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.

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.

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.

Dry Storage of the BR3 Spent Fuel in the CASTOR® BR3 Cask

2003 ◽  
Author(s):  
Luc Ooms ◽  
Vincent Massaut ◽  
L. Noynaert ◽  
M. Braeckeveldt ◽  
G. Geenen
Keyword(s):  
Spent Fuel ◽  
Dual Purpose ◽  
Shock Absorbers ◽  
Commercial Plant ◽  
Long Term Storage ◽  
Fuel Assemblies ◽  
Test Reactor ◽  
And Storage ◽  
Term Storage

The BR3 reactor was the first PWR plant installed in Europe. Started in 1962, BR3 was definitely shut down on June 30th, 1987. Used at the beginning of its life as a training device for commercial plant operators, it was also used during its whole life as test-reactor for new fuel types and assemblies. Most of the spent fuel was stored in the deactivation pool of the plant for more than 15 years. The reactor being now in decommissioning, it was decided to remove the spent fuel from the plant. After comparison of different solutions, the long term storage in dual purpose storage casks was selected in 1997. The selected CASTOR-BR3® cask is designed as a transport and storage cask for accommodating 30 spent fuel assemblies. As a type B(U) cask fitted with shock absorbers, it meets the transport requirements according to the IAEA guidelines and fulfils also the conditions for cask storage.

Mixed Reload Design Using MOX and UOX Fuel Assemblies

2002 ◽  
Author(s):  
J. Ramo´n Rami´rez Sa´nchez ◽  
R. T. Perry
Keyword(s):  
Conceptual Design ◽  
Cycle Length ◽  
Multiplication Factor ◽  
The Core ◽  
Mox Fuel ◽  
Fuel Assemblies ◽  
Current Cycle ◽  
Full Power ◽  
Control Rods ◽  
Mixed Load

As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10×10 BWR fuel assemblies but different fisil material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fisil plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO2 fresh fuel were tested to verify the shutdown margin, the UO2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper.

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