Major sensitivities in FRAPCON-xt modeling of high burn-up fuel rods during dry storage to design and irradiation variables

Kerntechnik ◽  
2020 ◽  
Vol 85 (6) ◽  
pp. 413-418
Author(s):  
C. Aguado ◽  
F. Feria ◽  
L. E. Herranz
Keyword(s):  
Fuel Rods ◽  
Dry Storage

The effect of long-term annealing simulating the parameters of dry storage of VVER-1000 fuel rods on the mechanical properties of E110 alloy shells in the longitudinal direction

2021 ◽  
Vol 4 ◽  
pp. 42-49
Author(s):  
G. P. Kobylyansky ◽  
◽  
А. О. Mazaev ◽  
Е. А. Zvir ◽  
S. G. Eremin ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Yield Strength ◽  
Relative Elongation ◽  
Mechanical Test ◽  
Longitudinal Direction ◽  
Radiation Defects ◽  
Thermal Tests ◽  
Fuel Rods ◽  
Dry Storage

Presented are the results of mechanical tensile tests of longitudinal (segmental) samples cut from the midsection of claddings spent as VVER-1000 FA during one- and six-year campaigns and subject to thermal tests in helium at 480 °С during 468 full days. An average burnup of these fuel rods achieved ~ 20 and ~ 70 (MW·day)/kg U, respectively. The tests followed the examinations for cladding mechanical properties performed using the tests results for ring samples cut from the specified fuel rods. These fuel rods were tested in the experiments to determine impact of long-term thermal tests that model dry storage conditions on mechanical properties of Zr E110 claddings. Based on mechanical tests results at room temperature and at 380 °С there was determined as follows: ultimate strength sв, yield strength s0,2 and total relative elongation d0 of claddings length-wise on the fuel rod segments at the fuel column midsection. The obtained characteristics were compared to corresponding values for initial (unirradiated) cladding tubes and mechanical test results of the ring samples in the transverse direction. Long-term thermal tests have led to partial return to initial (before operation) values sв, s0,2 and d0 of radiation-hardened claddings; this return was more prominent in the longitudinal direction than in the transverse one. A radiation hardening decrease was accompanied with an increase in total relative elongation values in both cladding directions. Anisotropy of yield strength has changed as well. These changes can be explained by partial annealing of radiation defects, which are obstacles to dislocation movements during cladding strain. The morphology of above radiation defects is different in various sliding planes in texturized grains of cladding material.

High Temperature Postirradiation Materials Performance of Spent Pressurized Water Reactor Fuel Rods under Dry Storage Conditions

1982 ◽  
Vol 57 (1) ◽  
pp. 65-80 ◽  
Author(s):  
Robert E. Einziger ◽  
Sharon D. Atkin ◽  
David E. Stellrecht ◽  
V. Pasupathi
Keyword(s):  
High Temperature ◽  
Storage Conditions ◽  
Reactor Fuel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Water Reactor ◽  
Dry Storage

Dry Storage of Spent Research Reactor Fuel in Castor BR3® Casks at Belgoprocess in Belgium

2001 ◽  
Author(s):  
Marnix Braeckeveldt ◽  
Luc Ooms ◽  
Gustaaf Geenen
Keyword(s):  
Radioactive Waste ◽  
Nuclear Research ◽  
High Level Waste ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Dry Storage ◽  
Nuclear Site ◽  
Test Reactor ◽  
Intermediate Storage ◽  
High Level

Abstract The BR3 reactor (10.5 MWe) at the Nuclear Research Center SCK•CEN was the first PWR plant installed in Europe and has been shut down in 1987. The BR3 reactor is from 1989 in a decommissioning stage and most of the spent fuel is presently still stored in the deactivation pool of the BR3 plant and has to be evacuated. The BR3 was used as a test-reactor for new fuel types and assemblies (Mixed Oxide (MOX) fuel, fuel rods containing burnable poison (Gd2O3) and other types of fuels). Some fuel rods, having undergone a destructive analysis, are stored in different laboratories at the SCK•CEN. In total, the BR3 spent fuel comprises the equivalent of almost 200 fuel assemblies corresponding to some 5000 fuel rods. Beside the spent BR3 fuel, a limited number of spent fuel rods, with equivalent characteristics as the BR3 fuel but irradiated in research reactors outside Belgium and stored in other buildings at the SCK•CEN nuclear site, were added to the inventory of spent fuel to be evacuated. Various options such as reprocessing and intermediate storage awaiting final disposal were evaluated against criteria as available techniques, safety, waste production and overall costs. Finally the option of an AFR (away-from-reactor) intermediate dry storage of the BR3 and other spent fuel in seven CASTOR BR3® casks was adopted. As the SCK•CEN declared this spent fuel as radioactive waste, NIRAS/ONDRAF, the Belgian radioactive waste management agency became directly involved and the decision was taken to construct a small building at the Belgoprocess nuclear site for storing the CASTOR BR3® casks. Loading at the SCK•CEN followed by transport to Belgoprocess and storage is scheduled to take place at the end of 2001. The CASTOR BR3® cask weighing some 25 tonnes, consists of a monolithic body and has two independent lids with metal seals guaranteeing the long term leak-tightness of the cask. The CASTOR BR3® cask is designed for transport and the intermediate storage of at least 50 years. Although a defect of the leaktightness of a CASTOR BR3® cask is very unlikely to occur, an intervention scenario had to be developed. As no pool is present at the Belgoprocess nuclear site to unload the fuel, an innovative procedure is developed that consists of transferring the basket, containing the spent fuel, into another CASTOR BR3® cask. This operation can be performed in the hot cell of the existing storage building for high level waste at the Belgoprocess site.

Accelerated High-Temperature Tests with Spent PWR and BWR Fuel Rods Under Dry Storage Conditions

1986 ◽  
Vol 74 (3) ◽  
pp. 287-298 ◽  
Author(s):  
Gerd Porsch ◽  
Joachim Fleisch ◽  
Bernd Heits
Keyword(s):  
High Temperature ◽  
Storage Conditions ◽  
Fuel Rods ◽  
Dry Storage

scholarly journals Investigation of water-logged spent fuel rods under dry storage conditions

1986 ◽  
Author(s):  
R. Kohli ◽  
V. Pasupathi
Keyword(s):  
Storage Conditions ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Dry Storage

scholarly journals Recommended temperature limits for dry storage of spent light water reactor Zircaloy-clad fuel rods in inert gas

1987 ◽  
Author(s):  
I.S. Levy ◽  
B.A. Chin ◽  
E.P. Simonen ◽  
C.E. Beyer ◽  
E.R. Gilbert ◽  
...  
Keyword(s):  
Light Water Reactor ◽  
Inert Gas ◽  
Fuel Rods ◽  
Light Water ◽  
Water Reactor ◽  
Dry Storage

Commercial Spent Nuclear Fuel Integrity During Long-Term Dry Storage

2001 ◽  
Author(s):  
Leroy Stewart ◽  
Mikal A. McKinnon
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Dry Storage ◽  
Environmental Laboratory ◽  
Cooperative Project

Abstract The United States Department of Energy (DOE) Office of Civilian Radioactive Waste Management conducted spent nuclear fuel integrity and cask performance tests from 1984–1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). Between 1994 and 1998, DOE also initiated a Spent Fuel Behavior Project that involved enhanced surveillance, monitoring, and gas-sampling activities for intact fuel in a GNS CASTOR V/21 cask and for consolidated fuel in a Sierra Nuclear VSC-17 cask. The results of these series of tests are reported in this paper. Presently, DOE is involved in a cooperative project to perform destructive evaluations of fuel rods that have been stored in the CASTOR V/21 cask. The results of those evaluations are presented elsewhere in these proceedings in a paper entitled “Examination of Spent PWR Fuel Rods after 15 years in Dry Storage”.

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