Feasibility assessment of using external neutron and gamma radiation measurements for monitoring the state of fuel assemblies in dry storage casks

2020 ◽  
Vol 135 ◽  
pp. 106975 ◽  
Author(s):  
Reuven Rachamin ◽  
Uwe Hampel
Kerntechnik ◽  
2018 ◽  
Vol 83 (6) ◽  
pp. 513-522 ◽  
Author(s):  
U. Hampel ◽  
A. Kratzsch ◽  
R. Rachamin ◽  
M. Wagner ◽  
S. Schmidt ◽  
...  

2017 ◽  
Vol 48 (1) ◽  
pp. 1-4
Author(s):  
V. V. Gorbachev ◽  
V. N. Gavrin ◽  
T. V. Ibragimova ◽  
A. V. Kalikhov ◽  
Yu. M. Malyshkin ◽  
...  

2022 ◽  
Vol 179 ◽  
pp. 109991
Author(s):  
Habila Nuhu ◽  
Suhairul Hashim ◽  
Mohamad Syazwan Mohd Sanusi ◽  
Muneer Aziz Mohammed Saleh

Author(s):  
Mile Bace ◽  
Kresimir Trontl ◽  
Dubravko Pevec

Abstract The intention was to model a dry storage facility that could satisfy the needs of a medium nuclear power plant similar to the NPP Krsko. The attention has been focused on radiation dose rate analyses and criticality calculations. Using the SCALE 4.4 code package and modified QAD-CGGP code, we modeled a facility that satisfies the basic criteria for public radiation protection. The capacity of the storage is 1,400 spent fuel assemblies which is adequate for a forty years medium NPP lifetime.


Author(s):  
Mengqi Wang ◽  
Nan Pan ◽  
Hui Li ◽  
Baojun Jia

Spent fuel dry storage technology is one of the most important intermediate storage technologies for spent fuel, because of its high security, good economic and easy to expand the scale. This article aims at designing a spent fuel dry storage cask which can contain 21 FA300 spent fuel assemblies. The spent fuel dry storage cask is designed as concrete cask structure, which has the advantages of low manufacturing cost and simple manufacturing technology. Ventilation channels are designed for heating transfer, because the concrete is not a good thermal conductivity material. And labyrinth structure is designed for the ventilation channel to reduce the cavity streaming. Radiation sources in spent fuel assemblies are mainly produced from fission products, actinides and their daughters located inside the effective fuel region, and other activation products in structure materials, which are calculated by ORIGEN. The source and geometry of this problem are complex, and this is a real world deep penetration and streaming problem. Discrete ordinate method has great advantage in solving the deep penetration problem. Based on three-dimensional discrete ordinate code TORT, radiation shielding design method for spent fuel dry storage cask is studied, including main shield cask, cover lid, and ventilation channel. The results show that this spent fuel dry storage cask containing 21 FA300 spent fuel (cooling time: 10 years) assemblies can satisfy the requirement of dose rate limits in GB18871.


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