scholarly journals Assessment of the integrity of spent fuel assemblies used in dry storage demonstrations at the Nevada Test Site

10.2172/60183 ◽  
1987 ◽  
Author(s):  
A.B. Jr. Johnson ◽  
J.C. Dobbins ◽  
F.R. Zaloudek
Keyword(s):  
Test Site ◽  
Spent Fuel ◽  
Nevada Test Site ◽  
Dry Storage ◽  
Fuel Assemblies

Model of a Dry Storage Facility for a Medium Nuclear Power Plant

2001 ◽  
Author(s):  
Mile Bace ◽  
Kresimir Trontl ◽  
Dubravko Pevec
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Radiation Dose ◽  
Dose Rate ◽  
Nuclear Power ◽  
Storage Facility ◽  
Spent Fuel ◽  
Radiation Dose Rate ◽  
Dry Storage ◽  
Fuel Assemblies

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.

Radiation Shielding Analysis of a Spent Fuel Dry Storage Cask for FA300 Spent Fuel Assemblies

2017 ◽  
Author(s):  
Mengqi Wang ◽  
Nan Pan ◽  
Hui Li ◽  
Baojun Jia
Keyword(s):  
Design Method ◽  
Three Dimensional ◽  
Fission Products ◽  
Radiation Shielding ◽  
Deep Penetration ◽  
Manufacturing Cost ◽  
Spent Fuel ◽  
Discrete Ordinate ◽  
Dry Storage ◽  
Fuel Assemblies

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.

Validation of Thermal Analysis of Dry Storage Casks at Diablo Canyon

2015 ◽  
Author(s):  
Jie Li ◽  
Yung Y. Liu
Keyword(s):  
Thermal Analysis ◽  
Ambient Temperature ◽  
Convergence Criterion ◽  
Measured Temperature ◽  
Spent Fuel ◽  
Ansys Fluent ◽  
Pressurized Water ◽  
Dry Storage ◽  
Long Term Storage ◽  
Fuel Assemblies

This paper is a continuation of previous work; it focuses on validating the thermal analysis of a vertical dry storage cask by using the measured temperature data and the results obtained by others in thermal modeling of a HI-STORM 100 storage cask at Diablo Canyon’s independent spent fuel storage installation (ISFSI). The cask chosen for thermal analysis contains a welded canister for 32 pressurized water reactor (PWR) used fuel assemblies in a stainless-steel basket with a total decay heat load of 17.05 kW. An effective thermal conductivity model was used to represent the used fuel assemblies with non-uniform assembly heat loads. The pressure of the canister’s helium fill gas was assumed to be 5 atm, and the ambient temperature was assumed to be 10°C. The results showed reasonably good agreement between the calculated and measured canister axial surface temperatures. The results of ANSYS/FLUENT simulations showed that a tighter convergence criterion yielded slightly better agreement with the data; however, improvement could be obtained by adjusting the assumed ambient temperature value in the simulation. Validating the results of ANSYS/FLUENT simulation against the data (as well as the experience and additional insights gained from the validation exercise) is important to our future simulation and analysis of the thermal performance of dry storage casks, particularly for aging management and monitoring the condition and performance of dry casks during extended long-term storage at ISFSIs.

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

2021 ◽  
Author(s):  
Davor Grgic ◽  
Mario Matijevic ◽  
Paulina Duckic ◽  
Radomir Jecmenica
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Variance Reduction ◽  
Spent Fuel ◽  
Dose Rates ◽  
Dry Storage ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
The Impact

Abstract In this paper shielding analysis was performed to determine neutron and gamma dose rates around the transfer cask HI-TRAC VW loaded with Spent Fuel Assemblies (SFA) from Nuclear Power Plant (NPP) Krsko Spent Fuel Dry Storage (SFDS) Campaign one. The HI-TRAC VW is a multi-layered cylindrical vessel designed to accept a Multi Purpose Canister (MPC) during loading, unloading and transfer to dry storage building. The MPC can contain up to 37 spent fuel assemblies. The analysis was divided into two steps. The first step was the source term generation using ORIGEN-S module of the SCALE code package. The source was calculated based on the operating history of spent fuel assemblies currently located in the NPP Krsko spent fuel pool. The obtained particle intensities and source spectra of the SFA were used in the second step to calculate the dose rates around the transfer cask. A comprehensive hybrid shielding analysis included the calculation of dose rates resulting from fuel neutrons and gammas, neutron induced gammas (n-g reaction), and hardware activation gammas under normal conditions and during accident scenario. To obtain the dose rates within the acceptable uncertainties, FW-CADIS variance reduction scheme, as implemented in ADVANTG code, was adopted for accelerating final MCNP6 calculations. The dose rates around HI-TRAC VW cask were calculated using MCNP6 code for all 16 casks loading belonging to Campaign one in order to illustrate the impact of fuel assembly selection schemes proposed by company responsible for project realization (Holtec International).

Thermal Performance Investigation of Hexagonal Spent Fuel Dry Storage Facilities

2017 ◽  
Author(s):  
Bing Ren ◽  
Chenxiao Ni ◽  
Yu Dang ◽  
Jiazheng Liu
Keyword(s):  
Heat Transfer ◽  
Storage System ◽  
Parameter Method ◽  
Spent Fuel ◽  
Decay Heat ◽  
Dry Storage ◽  
Lumped Parameter ◽  
Fuel Assemblies ◽  
Cfd Method ◽  
Lumped Parameter Method

A new type of dry storage system is designed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI), which can efficiently remove the decay heat of the hexagonal spent fuel assemblies such as VVER fuel assemblies. The dry storage system includes a Ventilated Concrete Cask (VCC) and a Multi-assembly sealed basket (MSB). Decay heat is removed by natural circulation with helium and air, heat conduction and thermal radiation heat transfer. Thermal performance of the dry storage system has been investigated by two different numerically methods, i.e., the Computational Fluid Dynamics (CFD) method and the lumped parameter method. The CFD method is utilized based on the commercial software STAR-CCM+, and fuel assemblies are modeled as a porous medium characterized by effective conductivity and the permeability and inertial resistance factor, while other geometry including the lids, base plates, inner and outer shell are modeled explicitly with necessary simplifications. The lumped parameter method is utilized based on the system code GOTHIC, the geometry and the fuel assemblies are divided and represented by 44 volumes. The flow of the air and helium are modeled by flow path which connects the related volumes, and the heat transfer between fluid and solid structures are modeled by thermal conductor models. Heat transfer by convection, conduction and thermal radiation is modeled in both of the two methods. The maximum temperature of spent fuel assembly can be obtained by both of the two methods, which can be a design basis for investigations attempting to improve the performance of the dry storage system. It is found that the simulation results calculated by the lumped parameter method are more conservative than those calculated by the CFD method. Both methods indicate that after the storage of 7.5 years, the dry storage system is able to remove the decay heat from the hexagonal spent fuel assemblies, keeping maximum cladding temperature below the design limit. Besides, detailed flow characteristic are obtained by CFD simulation. Furthermore, effects of MSB normal operating pressure and the ambient temperature are studied.

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