Development of a Thermal Analysis Model for a Nuclear Spent Fuel Storage Cask and Experimental Verification With Prototype Testing

1989 ◽  
Vol 111 (4) ◽  
pp. 647-651 ◽  
Author(s):  
J. Y. Hwang ◽  
L. E. Efferding
Keyword(s):  
Thermal Analysis ◽  
Storage Period ◽  
Department Of Energy ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Analysis Model ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
Spent Fuel Storage ◽  
Fuel Storage

A thermal analysis evaluation is presented of a nuclear spent fuel dry storage cask designed by the Westinghouse Nuclear Components Division. The cask is designed to provide passive cooling of 24 Pressurized Water Reactor (PWR) spent fuel assemblies for a storage period of at least 20 years at a nuclear utility site (Independent Spent Fuel Storage Installation). A comparison is presented between analytical predictions and experimental results for a demonstration cask built by Westinghouse and tested under a joint program with the Department of Energy and Virginia Power Company. Demonstration testing with nuclear spent fuel assemblies was performed on a cask configuration designed to store 24 intact spent fuel assemblies or canisters containing fuel consolidated from 48 assemblies.

Probabilistic Risk Assessment of Bolted Dry Spent Fuel Storage Cask

2004 ◽  
Author(s):  
Jeffrey T. Mitman ◽  
Ken Canavan
Keyword(s):  
Risk Assessment ◽  
Life Cycle ◽  
Probabilistic Risk Assessment ◽  
First Year ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Probabilistic Risk

EPRI performed a Probabilistic Risk Assessment (PRA) of a dry spent fuel storage cask. The study was performed for a bolted cask at a “generic” pressurized water reactor site. A generic site was chosen so the widest variety of challenges could be considered. The study calculates the annual individual radiological risk and consequences associated with a single cask life cycle, where the life cycle is divided into three phases: loading, on-site transfer, and on-site storage. The project used standard methods of PRA with the following analysis tasks: initiating event (IE), data, human reliability, structural, thermal hydraulic, accident sequence, and consequence. The results shows that the risk is extremely low with no calculated early fatalities and a first year risk of latent cancer fatality of 3.5E−11 per year per cask. Subsequent year risk to the general public is even lower; with, again, no early fatalities and a cancer risk of 4.2E−12.

scholarly journals Criticality safety analysis of spent fuel pool for a PWR using UO2, MOX, (Th-U)O2 and (TRU-Th)O2 fuels

2019 ◽  
Vol 7 (3A) ◽  
Author(s):  
Claubia Pereira ◽  
Jéssica P. Achilles ◽  
Fabiano Cardoso ◽  
Victor F. Castro ◽  
Maria Auxiliadora F. Veloso
Keyword(s):  
Normal Operation ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Operation Conditions ◽  
Transport Code ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Criticality Safety

A spent fuel pool of a typical Pressurized Water Reactor (PWR) was evaluated for criticality studies when it uses reprocessed fuels. PWR nuclear fuel assemblies with four types of fuels were considered: standard PWR fuel, MOX fuel, thorium-uranium fuel and reprocessed transuranic fuel spiked with thorium. The MOX and UO2 benchmark model was evaluated using SCALE 6.0 code with KENO-V transport code and then, adopted as a reference for other fuels compositions. The four fuel assemblies were submitted to irradiation at normal operation conditions. The burnup calculations were obtained using the TRITON sequence in the SCALE 6.0 code package. The fuel assemblies modeled use a benchmark 17x17 PWR fuel assembly dimensions. After irradiation, the fuels were inserted in the pool. The criticality safety limits were performed using the KENO-V transport code in the CSAS5 sequence. It was shown that mixing a quarter of reprocessed fuel withUO2 fuel in the pool, it would not need to be resized 

Recommendations for Conducting Probabilistic Risk Assessment of Spent Fuel Storage Facilities

2017 ◽  
Author(s):  
Sai Zhang ◽  
Jun Zhao ◽  
Jiejuan Tong ◽  
Zhixin Xu
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Spent Fuel ◽  
Complementary Method ◽  
Fuel Assemblies ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Storage Units ◽  
Storage Facilities ◽  
Probabilistic Risk

Currently, the probabilistic risk assessments (PRA) for the nuclear power plant (NPP) sites are primarily focused on the reactor counterpart. However, evoked by the 2011 Fukushima Daiichi accident, it has been widely recognized that a complete site risk profile should not be confined to the reactor units, but should cover all the radiological sources in a site, e.g. spent fuel storage facilities. During the operation of the reactor units, the used fuel assemblies will be unloaded from the reactor core to the storage facilities in a continuous or periodical manner. Accident scenarios involving such facilities can occur with non-negligible frequencies and significant consequences, posing threat to public safety. Hence, the risk contributions from such scenarios should be carefully estimated and integrated into the safety goal evaluations. The spent fuel storage facilities can be categorized as two types: pool storage units and dry cask storage facilities. In the former type, spent fuel assemblies are stored in large pools inside or outside the reactor building, with the residual heat removed by natural or forced water circulation. The latter type, where air or inert gas circulation plays an important role, appear mostly as a complementary method, along with the pool storage units, to expand the plant’s storage capacity. For instance, at the Daiichi plant, there are several fuel pool units holding some fresh fuel and some used fuel, the latter awaiting for its transfer to the dry cask storage facilities on site. Note that, as well as in a joint manner, both storage facilities can be designed to serve the NPPs independently. As a fully developed method to identify potential risk in a logical and quantitative way, the framework of PRA can be generally applied to the spent fuel storage facilities with some special considerations. This paper is aimed at giving recommendations for the spent fuel storage facility PRAs, including (1) clarifying the analysis scope of risk from spent fuel storage facilities; (2) illustrating four key issues that determines such risk; (3) presenting three essential considerations when conducting PRAs to evaluate such risk. Also, this paper integrates the insights obtained from two representative case studies involving two NPP sites with different types of both fuel elements and storage facilities.

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.

scholarly journals Operational and Technical Results from the Spent Fuel Test-Climax

1986 ◽  
Vol 29 (1) ◽  
pp. 51-54
Author(s):  
Wesley Patrick
Keyword(s):  
Spent Nuclear Fuel ◽  
Radiation Transport ◽  
Storage Period ◽  
Test Site ◽  
Department Of Energy ◽  
Mechanical And Thermal Properties ◽  
Spent Fuel ◽  
Technical Data ◽  
Geologic Storage ◽  
Fuel Assemblies

The technical feasibility of short-term storage and retrieval of spent nuclear fuel assemblies has recently been demonstrated in a test of deep geologic storage at the U.S. Department of Energy Nevada Test Site (NTS). Handling systems and procedures developed and deployed on this test functioned safely and reliably to emplace eleven intact spent-fuel assemblies and retrieve them three years later. Three exchanges of spent fuel were conducted at regular intervals during the storage period to maintain the proficiency of personnel and the readiness of the handling system. Technical data was collected using nearly 1,000 instruments. These data show that the mechanical and thermal properties of granites are compatible with nuclear waste isolation objectives. Measured and calculated temperatures are in excellent agreement, confirming the adequacy of available heat transfer codes. Radiation transport calculations were of high quality, exceeding the accuracy of available long-term dosimetry techniques which were used on the test. We also found good agreement between measured and calculated displacements within the rock mass.

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