Study of spent fuel elements from the first atomic electric station

1964 ◽  
Vol 14 (5) ◽  
pp. 482-485
Author(s):  
Sh. Sh. Ibragimov ◽  
L. A. Syshchikov ◽  
I. M. Voronin ◽  
V. G. Kudryashov
Keyword(s):  
Spent Fuel ◽  
Electric Station ◽  
Fuel Elements

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

Study of a spent fuel element from the first atomic electric station

Atomic Energy ◽  
1962 ◽  
Vol 11 (2) ◽  
pp. 758-761
Author(s):  
A. P. Smirnov-Averin ◽  
V. I. Galkov ◽  
V. I. Ivanov ◽  
V. P. Meshcheryakov ◽  
I. G. Sheinker ◽  
...  
Keyword(s):  
Fuel Element ◽  
Spent Fuel ◽  
Electric Station

Parametric Investigation of Rock Salt Behaviour Resulting from Disposal of High Level Radioactive Waste

1990 ◽  
Vol 212 ◽  
Author(s):  
D. Caramelle ◽  
M.T. Gaudez ◽  
J. Monig ◽  
G. Ouzounian ◽  
G. Simonet
Keyword(s):  
Grain Size ◽  
Rock Salt ◽  
Absorbed Dose ◽  
Gas Production ◽  
High Level Waste ◽  
Spent Fuel ◽  
Salt Composition ◽  
Gas Atmosphere ◽  
High Level ◽  
Fuel Elements

ABSTRACTThe liberation and generation of gases from rock salt due to heat and gamma irradiation is investigated in order to obtain some of the data needed for the development of the concept for the disposal of high level waste in rock salt.Our work is concerned with the influence of various parameters on gas production, e.g. salt composition and grain size, total absorbed dose, dose rate, temperature and gas atmosphere. Some of these parameters have not been studied previously in detail.The original gamma irradiator employing spent fuel elements and capable of exposing samples at temperatures up to 250°C will be described. Experimental results from some 150 experiments will be given. The two major gases found were CO2 and N2O.CO, H2, CH4, Hydrocarbons, CI2, HCl and SO2 were also detected. The dependence of the gas yields on the various parameters will be presented and discussed.

Determining burn-up in spent fuel elements

Atomic Energy ◽  
1962 ◽  
Vol 11 (5) ◽  
pp. 1102-1104
Author(s):  
A. P. Smirnov-Averin ◽  
V. I. Galkov ◽  
I. G. Sheinker ◽  
V. P. Meshcheryakov ◽  
L. A. Stabenova ◽  
...  
Keyword(s):  
Spent Fuel ◽  
Fuel Elements

Hydrothermal interactions of basalts with Cs and Sr of spent fuel elements

1981 ◽  
Vol 43 (9) ◽  
pp. 1967-1975 ◽  
Author(s):  
Sridhar Komarneni ◽  
Barry E. Scheetz
Keyword(s):  
Spent Fuel ◽  
Fuel Elements

VALIDATION OF A MONTE CARLO MODEL OF THE FORK DETECTOR WITH A CALIBRATED NEUTRON SOURCE

2014 ◽  
Vol 27 ◽  
pp. 1460151
Author(s):  
ALESSANDRO BORELLA ◽  
LIVIU-CRISTIAN MIHAILESCU
Keyword(s):  
Nuclear Power ◽  
Monte Carlo Model ◽  
Measurement Data ◽  
Nuclear Research ◽  
Research Centre ◽  
Spent Fuel ◽  
The Past ◽  
Research Areas ◽  
Fuel Elements ◽  
Comparison Exercise

The investigation of experimental methods for safeguarding spent fuel elements is one of the research areas at the Belgian Nuclear Research Centre SCK•CEN. A version of the so-called Fork Detector has been designed at SCK•CEN and is in use at the Belgian Nuclear Power Plant of Doel for burnup determination purposes. The Fork Detector relies on passive neutron and gamma measurements for the assessment of the burnup and safeguards verification activities. In order to better evaluate and understand the method and in view to extend its capabilities, an effort to model the Fork detector was made with the code MCNPX. A validation of the model was done in the past using spent fuel measurement data. This paper reports about the measurements carried out at the Laboratory for Nuclear Calibrations (LNK) of SCK•CEN with a 252Cf source calibrated according to ISO 8529 standards. The experimental data are presented and compared with simulations. In the simulations, not only was the detector modeled but also the measurement room was taken into account based on the available design information. The results of this comparison exercise are also presented in this paper.

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.

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
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