Nuclear Energy and Technology

Calculation and experimental studies for the spent nuclear fuel shipping cask sealing assembly
Calculation and experimental studies for the spent nuclear fuel shipping cask sealing assembly

Artem Z. Gayazov, Oleg Z. Gaiazov, Viacheslav Yu. Kozlov, Sergey V. Pavlov, Aleksandr A. Samsonov

Nuclear Energy and Technology, № 9(3), 2023

One of the safety requirements regarding the shipping cask for spent nuclear fuel is that its leak-tightness should be maintained by preserving the cask body structural integrity and the sealing system tightness under normal and accident transportation conditions. The cask under design has a cylindrical process penetration (port) in its bottom which is sealed using a plug with a radial seal composed of two rubber O-rings. The cask sealing assembly design was justified by the ANSYS LS-DYNA code calculation results. In particular, the strains of the cask components were calculated when dropped from a height of 1 m with the sealing assembly hitting a vertical bar. The cask was concluded to be leak-tight or leaky based on the strain nature and amount. To verify the adequacy of the results, computer-aided and realistic simulations were undertaken with a 1/2.5 scale mockup cask dropped on a bar from a height of 1 m. The computational and experimental results show a good agreement in terms of the impact response accelerations (overloads) for the mockup cask and bar collision and in terms of the plastic strains for the key components of the mockup bottom port sealing assembly. This proves the adequacy of the numerical cask model that has been developed and the efficiency of the LS-DYNA simulations. The inner rubber O ring compression is reduced by the plastic strains in the cask’s bottom port area, leading to a loose inner radial seal, as shown by the calculations. But the outer seal remains leak-tight, ensuring so the mockup cask tightness. The physical test results have also confirmed that the mockup cask remains leak-tight.

Study of hydrogen generation and radionuclide release during wet damaged oxide spent fuel storage
Study of hydrogen generation and radionuclide release during wet damaged oxide spent fuel storage

A. Gayazov, S. Komarov, A. Leshchenko, K. Revenko, V. Smirnov (Sosny R&D Company), E. Zvir , P. Ilyin, V. Teplov (JSC “SSC RIAR”)

Nuclear Energy and Technology 5 (1), 2019

The paper describes the outcomes of the experiments to study hydrogen and gaseous fission products accumulation during simulations of the wet damaged VVER-440 SNF storage in air-tight canisters with the water drained and no drying conducted. Physical and chemical processes occurring during the damaged oxide SNF storage in wet environment are discussed. The experiments were carried out in two stages: 1) preliminary soaking of fine fuel particles in water in an air-tight canister, 2) water draining and keeping the wet SNF in the air-tight canister.

The experiments were conducted one after another using the same SNF canister and differing only in the SNF soaking temperature, i.e. 25 and 80 °С.

The radionuclide release into the liquid phase during the SNF storage under water was studied. Uranium and cesium isotopic concentrations were found to reach steady values when the SNF is kept under water for more than a month. The kinetics of hydrogen and gaseous fission product accumulation in the gaseous phase during wet storage of the spent fuel in the air-tight canister with the water drained coincide for both experiments. The kinetics demonstrate an abrupt decrease of the hydrogen and gaseous fission product accumulation rate in 46 hours. The data obtained can be applied for development and verification of the damaged SNF behavior models during SNF storage in wet environment under radiolysis.

 

In-service Change in the Flexural Rigidity of the VVER-1000 Fuel Assemblies
In-service Change in the Flexural Rigidity of the VVER-1000 Fuel Assemblies

S.V. Pavlov

Nuclear Energy and Technology, № 4, 2016

This paper describes a method and a facility for the hot cell testing of the irradiated VVER-1000 FA flexural rigidity. The method is based on measurements of the FA bowing induced by cross-sectional loading. The load applied to the spacer grids is perpendicular to the grid face, and the FA bowing is measured optically using a TV camera. The facility can also be used to test the flexural rigidity of the FA skeleton after all of the fuel rods are removed. Several tens of VVER-1000 FAs with a burnup of ∼4 to ∼65 МW day/kg U were tested by Dimitrovgrad Research Institute of Atomic Reactors. The generalization and an analysis of the test results have made it possible to identify the major factors that contribute to the in-service change in the flexural rigidity of the VVER-1000 FAs and to determine the experimental dependence of its change on burnup.

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