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

2019 ◽  
Vol 5 (1) ◽  
pp. 61-66
Author(s):  
Artem Z. Gayazov ◽  
Sergey V. Komarov ◽  
Anton Yu. Leshchenko ◽  
Ksenia E. Revenko ◽  
Valery P. Smirnov ◽  
...  

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.

1965 ◽  
Vol 20 (12) ◽  
pp. 1566-1568 ◽  
Author(s):  
J. Takagi

The evaporation behavior of non-gaseous fission products from UO2 was studied. The release rate of non-gaseous fission products during the post-irradiation annealing of UO2 was found to be controlled by the proportionality constant of evaporation, a, as well as the diffusion constant. The values of α for Ru, Ce, La, Mo and Te were determined at temperatures ranging 750 to 1600 °C and the general method of treating non-gaseous fission product release from UO2 was discussed.


1987 ◽  
Vol 112 ◽  
Author(s):  
Shirley A. Rawson ◽  
William L. Neal ◽  
James R. Burnell

AbstractThe Basalt Waste Isolation Project has conducted a series of hydrothermal experiments to characterize waste/barrier/rock interactions as a part of its study of the Columbia River basalts as a potential medium for a nuclear waste repository. Hydrothermal tests of 3–15 months duration were performed with light water reactor spent fuel and simulated groundwater, in combination with candidate container materials (low-carbon steel or copper) and/or basalt, in order to evaluate the effect of waste package materials on spent fuel radionuclide release behavior. Solutions were filtered through 400 and 1.8 nm filters to distinguish colloidal from dissolved species. In all experiments, 14C, 129I, and 137Cs occurred only as dissolved species, whereas the actinides occurred in 400 nm filtrates primarily as spent fuel particles. Actinide concentrations in 1.8 nm filtrates were below detection in steel-bearing experiments. In the system spent fuel + copper, apparent time-invariant concentrations of 14C and 137Cs were obtained, but in the spent fuel + steel system, the concentrations of 14C and 137Cs increased gradually throughout the experiments. In experiments containing basalt or steel + basalt, 137Cs concentrations decreased with time. In tests with copper + basalt, 14C and 129I concentrations attained time-invariant values and 137Cs concentrations decreased. Concentrations for the actinides and fission products measured in these experiments were below those calculated from Federal regulations governing radionuclide release.


1987 ◽  
Vol 112 ◽  
Author(s):  
L. H. Johnson ◽  
D. W. Shoesmith ◽  
S. Stroes-Gascoyne

AbstractThe concept of disposal of unreprocessed spent fuel has now been under study internationally for over ten years. Considerable progress has been made in understanding the factors that will control radionuclide release from spent fuel in an underground disposal vault. This progress is reviewed and the research areas of significance in providing further data for source term models are discussed. Key areas for future research are identified; these include improved characterization of spent fuel to determine the inventories of fission products at grain boundaries, together with their release kinetics; and a better understanding of the effects of solution chemistry on spent fuel dissolution, in particular the effects of salinity, redox chemistry, and radiolysis of groundwater. Approaches to modelling the dissolution of spent fuel are discussed, and a possible approach for developing an oxidative dissolution model is outlined.


1986 ◽  
Vol 75 (3) ◽  
pp. 326-331 ◽  
Author(s):  
Shlomo Ron ◽  
Zeev B. Alfassi ◽  
Michael Baer

1984 ◽  
Vol 44 ◽  
Author(s):  
John O. Barner ◽  
J. L. Daniel

AbstractThe Materials Characterization Center (MCC) provides characterized light-water reactor (LWR) spent fuel approved testing materials (ATMs) for use by experimenters in radionuclide release studies in support of repository licensing efforts. The characterization completed to date is described for MCC ATM-101, a Zircaloy-clad UO2 fuel from the H. B. Robinson, Unit 2, Assembly BO-5. ATM-101 is a moderate burnup, low-releasing (fission products in-reactor) spent fuel typical of the fuel that is expected to initially be placed in a repository.


2018 ◽  
Vol 2018 (3) ◽  
pp. 125-136
Author(s):  
Artem Zuferovich Gaiazov ◽  
Sergey Vladimirovich Komarov ◽  
Leshchenko A.Yu. Leshchenko ◽  
Ksenia Evgenievna Revenko ◽  
Smirnov V.P. Smirnov ◽  
...  

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