Heat Transfer from Transport Cask Storage Facilities for Spent Fuel Elements

1983 ◽  
Vol 62 (1) ◽  
pp. 62-70 ◽  
Author(s):  
Wolfgang von Heesen ◽  
Heinz Malmström ◽  
Rüdiger Detzer ◽  
Werner Loew
Keyword(s):  
Heat Transfer ◽  
Spent Fuel ◽  
Fuel Elements ◽  
Storage Facilities

scholarly journals Design Of Dry Cask Storage For Serpong Multipurpose Reactor Spent Nuclear Fuel

2018 ◽  
Vol 24 (1) ◽  
Author(s):  
Dyah Sulistyani Rahayu ◽  
Yuli Purwanto ◽  
Zainus Salimin
Keyword(s):  
Heat Transfer ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Construction Material ◽  
Trial And Error ◽  
Spent Fuel ◽  
Concrete Wall ◽  
Transfer Resistance ◽  
Steel Liner ◽  
Fuel Elements

DESIGN OF DRY CASK STORAGE FOR SERPONG MULTI PURPOSE REACTOR SPENT NUCLEAR FUEL. The spent nuclear fuel (SNF) from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF). At present there are a remaining of 245 elements of SNF on the ISSF,198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decayat a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg2. The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectedly. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method basedon SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister.  The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 oC and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i.e. rack canister no 3, 8 and 10. The value of I0 from the rack no 3, 8 and 10 are 434.307; 446.344; and 442.375 mrem/h respectively. The total heat decay from rack canister no 3,8 and 10 are 179.640 ; 335.2; and 298.551 watts. The result of the trial and error calculation indicates that the rack canister no 3, 8 and 10 need the thickness of concrete shielding of 0.1912, 0.1954 and 0.1940 m respectively.Keywords: heat and radiation decay, spent fuel , storage cask.

scholarly journals Methodology for Determining the Thermal and Thermal-Stress States of a Concrete Storage Container for Spent Nuclear Fuel for Assessment of Its Service Life

2021 ◽  
pp. 33-39
Author(s):  
S. Alyokhina ◽  
А. Kostikov ◽  
N. Smetankina ◽  
P. Gontarovskyi ◽  
N. Garmash ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Thermal Stress ◽  
Service Life ◽  
Conjugate Heat Transfer ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Fuel Storage ◽  
Stress States ◽  
Transfer Problems ◽  
Storage Facilities

The work is devoted to the development of methodologies for determining the thermal and thermal-stress states of the main equipment in dry container storage facilities for spent nuclear fuel. Storage facilities of this type are most common for spent fuel of nuclear power reactors. The safety of storage equipment in terms of assessing its service life is not covered widely enough in the world scientific literature. In particular, there are no effective methods for calculating the thermal and thermal-stress states of the equipment that would take into account the influence of many external factors throughout the life of a storage facility. To assess the thermal state of the containers, forward conjugate heat transfer problems, accounting for the mutual heat transfer in both a solid body and in the fluid environment (air), are proposed to be solved. Based on the solution of the conjugate heat transfer problems, the boundary conditions are to be determined to further assess the thermal-stress state of storage containers using inverse heat transfer problems. The proposed approach to determining the thermal and thermal-stress states of a concrete spent fuel container will promote more effective methods for assessing the service life of dry spent fuel storage facilities, which is, in turn, necessary in the development of ageing management programs for storage equipment and long-term safe operation.

In-core heat transfer in MTR plate type fuel elements

1970 ◽  
Vol 12 (2) ◽  
pp. 231-248 ◽  
Author(s):  
P.J. Kreyger ◽  
W.A. Essler ◽  
W. Dellmann
Keyword(s):  
Heat Transfer ◽  
Fuel Elements ◽  
Type Fuel ◽  
Plate Type

Recovery of Uranium and Plutonium from Spent Fuel Elements of Nuclear Reactors

Atomic Energy ◽  
2005 ◽  
Vol 99 (5) ◽  
pp. 823-828
Author(s):  
V. T. Gotovchikov ◽  
V. I. Makarov ◽  
V. T. Orekhov ◽  
A. G. Rybakov ◽  
V. A. Seredenko
Keyword(s):  
Nuclear Reactors ◽  
Spent Fuel ◽  
Fuel Elements

Paper 12: The Development of an Apparatus to Measure the Normal Emissivity of Small Plane Samples in a Controlled Atmosphere

1967 ◽  
Vol 182 (8) ◽  
pp. 120-129 ◽  
Author(s):  
B. N. Furber ◽  
R. Broderick
Keyword(s):  
Heat Transfer ◽  
Carbon Dioxide ◽  
Fuel Element ◽  
Geometrical Factor ◽  
Controlled Atmosphere ◽  
Effective Emissivity ◽  
General Application ◽  
Safety Assessments ◽  
Fuel Elements ◽  
Normal Emissivity

The heat transfer from fuel elements in magnox reactors under all normal conditions is predominantly by forced convection. However, in safety assessments a burst in a bottom main coolant duct is postulated, a reversal of the carbon dioxide coolant flow takes place, and heat transfer from the fuel elements at this instant could be by radiation only. The effective emissivity of the fuel element or the normal emissivity of plane specimens of fuel element material and a geometrical factor are therefore required to enable the maximum fuel element temperature to be determined. The paper is mainly concerned with the development and calibration of an apparatus suitable for measuring the normal emissivity of small plane samples at temperatures up to 650°C. Though the design of the apparatus has been influenced by the special requirements involved in testing magnox specimens, the apparatus has a general application and the normal emissivities of other materials are also given.

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