PWR-UO2 nuclear fuel criticality study: control rod effects on infinite neutron multiplication factor and spent fuel composition

2013 ◽  
Vol 263 ◽  
pp. 42-46
Author(s):  
R.V. Sousa ◽  
C. Pereira ◽  
C.A.M. Silva ◽  
A.L. Costa ◽  
M.A.F. Veloso ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Fuel Composition ◽  
Multiplication Factor ◽  
Spent Fuel ◽  
Control Rod ◽  
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.

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

scholarly journals A comparison of bounding approach with isotopic correction factors and Monte Carlo sampling in burnup credit method

2021 ◽  
Vol 253 ◽  
pp. 07011
Author(s):  
Pavlína Haroková ◽  
Martin Lovecký
Keyword(s):  
Monte Carlo ◽  
Computational Prediction ◽  
Monte Carlo Sampling ◽  
Fuel Composition ◽  
Correction Factors ◽  
Conservative Approach ◽  
Multiplication Factor ◽  
Spent Fuel ◽  
Realistic Result ◽  
Fuel Storage

One of the methodologies used in criticality safety analysis is burnup credit method, which allows considering fuel burnup in models with spent fuel. This removes excessive conservatism from the analysis, but it also brings new uncertainties originating from computational prediction of spent fuel composition. The burnup credit method offer several possibilities of how to deal with this problem, e.g. using bounding approach with correction factors on nuclide concentrations, which is simple, but still very conservative approach. Another option is Monte Carlo sampling, which aims at receiving the most realistic result as possible, but is very computationally demanding. In this work, we have analyzed correction factors for selected nuclides and compared the results of both methods on model of spent fuel storage pool. The results show how much conservative the bounding approach is – in this case, the multiplication factor was higher by almost 0.03 than in Monte Carlo sampling, exceeding the standard deviation by more than 5.4 times.

Release of 108Ag from Irradiated PWR Control Rod Absorbers under Deep Repository Conditions

2015 ◽  
Vol 1744 ◽  
pp. 217-222
Author(s):  
O. Roth ◽  
M. Granfors ◽  
A. Puranen ◽  
K. Spahiu
Keyword(s):  
Nuclear Fuel ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Nuclear Reactors ◽  
Reactor Irradiation ◽  
Spent Fuel ◽  
Control Rod ◽  
Potential Source ◽  
Absorber Material ◽  
Control Rods

ABSTRACTIn a future Swedish deep repository for spent nuclear fuel, irradiated control rods from PWR nuclear reactors are planned to be stored together with the spent fuel. The control rod absorber consists of an 80% Ag, 5% Cd, 15% In alloy with a steel cladding. Upon in-reactor irradiation 108Ag is produced by neutron capture. Release of 108Ag has been identified as a potential source term for release of radioactive substances from the deep repository.Under reducing deep repository conditions, the Ag corrosion rate is however expected to be low which would imply that the release rate of 108Ag should be low under these conditions. The aim of this study is to investigate the dissolution of PWR control rod absorber material under conditions relevant to a future deep repository for spent nuclear fuel. The experiments include tests using irradiated control rod absorber material from Ringhals 2, Sweden. Furthermore, un-irradiated control rod absorber alloy has been tested for comparison. The experiments indicate that the release of Ag from the alloy when exposed to water is strongly dependent on the redox conditions. Under aerated conditions Ag is released at a significant rate whereas no release could be measured after 133 days during leaching under H2.

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.

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