CFD Analyses of the TN-24P PWR Spent Fuel Storage Cask

2012 ◽  
Author(s):  
Robert A. Brewster ◽  
Emilio Baglietto ◽  
Eric Volpenhein ◽  
Christopher S. Bajwa
Keyword(s):  
Porous Media ◽  
Spent Nuclear Fuel ◽  
Direct Determination ◽  
Measured Temperature ◽  
Spent Fuel ◽  
Dry Storage ◽  
Fuel Assemblies ◽  
Fuel Storage ◽  
Calculation Results ◽  
Cfd Analyses

Dry storage casks are used to store spent nuclear fuel after removal from the reactor spent fuel pool. Even prior to the Fukushima earthquake of March 2011, dry storage of spent fuel was receiving increased attention as many reactor spent fuel pools near their capacity. Many different types of cask designs are used, and one representative design is the TN-24P spent fuel cask, a non-ventilated steel cask with a shielded exterior shell and lid. The cask is typically filled with an inert gas such as helium, argon or nitrogen. In this paper, Computational Fluid Dynamics (CFD) calculation results for the thermal performance of the TN-24P cask using the commercial CFD software STAR-CCM+ are presented. Initial calculations employ a common approach of treating the fuel assemblies as conducting porous media with calibrated volume-averaged properties, and comparison to existing measured temperature data shows good agreement. One of the fuel assemblies is then replaced with a more accurate representation that includes the full geometric detail of the fuel rods, guide tubes, spacer grids and end fittings (flow nozzles), and the results shown are consistent with the initial analysis, but without the need for the assumptions inherent in the porous media approach. This hybrid modeling approach also permits the direct determination of important results, such as the precise location of peak fuel cladding temperatures (PCTs), which is not possible using the more traditional porous media approach.

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 Approaches to Safety Justification for Loading of VSC-VVER Containers in ZNPP DSFSF

2019 ◽  
pp. 82-87
Author(s):  
Ya. Kostiushko ◽  
O. Dudka ◽  
Yu. Kovbasenko ◽  
A. Shepitchak
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Operating Conditions ◽  
Spent Fuel ◽  
Fuel Assemblies ◽  
Fuel Storage ◽  
The Government

The introduction of new fuel for nuclear power plants in Ukraine is related to obtaining a relevant license from the regulatory authority for nuclear and radiation safety of Ukraine. The same approach is used for spent nuclear fuel (SNF) management system. The dry spent fuel storage facility (DSFSF) is the first nuclear facility created for intermediate dry storage of SNF in Ukraine. According to the design based on dry ventilated container storage technology by Sierra Nuclear Corporation and Duke Engineering and Services, ventilated storage containers (VSC-VVER) filled with SNF of VVER-1000 are used, which are located on a special open concrete site. Containers VSC-VVER are modernized VSC-24 containers customized for hexagonal VVER-1000 spent fuel assemblies. The storage safety assessment methodology was created and improved directly during the licensing process. In addition, in accordance with the Energy Strategy of Ukraine up to 2035, one of the key task is the further diversification of nuclear fuel suppliers. Within the framework of the Executive Agreement between the Government of Ukraine and the U.S. Government, activities have been underway since 2000 on the introduction of Westinghouse fuel. The purpose of this project is to develop, supply and qualify alternative nuclear fuel compatible with fuel produced in Russia for Ukrainian NPPs. In addition, a supplementary approach to safety analysis report is being developed to justify feasibility of loading new fuel into the DSFSF containers. The stated results should demonstrate the fulfillment of design criteria under normal operating conditions, abnormal conditions and design-basis accidents of DSFSF components.  Thus, the paper highlights both the main problems of DSFSF licensing and obtaining permission for placing new fuel types in DSFSF.

scholarly journals Simulation of the Thermal Conditions of Cask with Fuel Assemblies of BN-350 Reactor for Dry Storage

2019 ◽  
Vol 2019 ◽  
pp. 1-5
Author(s):  
Ye. T. Koyanbayev ◽  
M. K. Skakov ◽  
D. A. Ganovichev ◽  
Ye. A. Martynenko ◽  
A. A. Sitnikov
Keyword(s):  
Thermal Condition ◽  
Spent Nuclear Fuel ◽  
Thermal Conditions ◽  
Structural Elements ◽  
Fuel Temperature ◽  
Ansys Software ◽  
Dry Storage ◽  
Fuel Assemblies ◽  
Fuel Storage ◽  
Thermal Calculations

The analysis of the thermal condition of spent FA (fuel assembly) of BN-350 reactor in a six-place cask for dry storage is presented. Simulation of the thermal condition of the cask is conducted with finite elements method using ANSYS software. Calculations of fuel temperature, fuel cladding, and assembly structural elements are the part of the safety analysis for storage of spent FA. In conclusion, the results of the thermal calculations in the cases of filling cask with argon and atmospheric air are given when the thickness of the insulation cask with concrete is 0.5 and 1 m. As a result of the calculated studies, the parameters of SNF (spent nuclear fuel) storage are determined, under which the fuel temperatures will have minimum and maximum values.

scholarly journals Selection of Conservative Assumptions in Nuclear Safety Justification of SNF Storage Systems

2017 ◽  
pp. 24-28
Author(s):  
V. Borysenko ◽  
V. Goranchuk ◽  
Yu. Pionkovskyi ◽  
M. Sapon
Keyword(s):  
Spent Nuclear Fuel ◽  
Storage Systems ◽  
Storage System ◽  
Operation Mode ◽  
Nuclear Safety ◽  
Spent Fuel ◽  
Maximum Value ◽  
Fuel Storage ◽  
Calculation Results ◽  
Selection Of

The paper addresses the description of computer model for the spent fuel assemblies storage system in SCALE and MCNP codes, as well as the results in selection of conservative assumptions made to justify the nuclear safety of moving, transport and storage operations with the VVER-1000 spent nuclear fuel (SNF) in designed Centralized Spent Fuel Storage Facility (CSFCF). When justifying the nuclear safety, it is necessary to confirm that the maximum value of the effective multiplication coefficient K eff in SNF storage systems is kept below specified limit of 0.95 in any design-basis operation mode. The paper presents calculation results and analysis on the selection of the most conservative conditions of neutron multiplication leading to the maximum value of Keff.

scholarly journals The Three-­Dimensional Neutron-­Physical Model of Spent Nuclear Fuel Storage System

2021 ◽  
Vol 20 ◽  
pp. 51-59
Author(s):  
О. R. Trofymenko ◽  
◽  
І. M. Romanenko ◽  
М. І. Holiuk ◽  
C. V. Hrytsiuk ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Storage System ◽  
Three Dimensional ◽  
Monte Carlo Code ◽  
Storage Facility ◽  
Spent Fuel ◽  
Wide Range ◽  
Fuel Assemblies ◽  
Fuel Storage

The management of spent nuclear fuel is one of the most pressing problems of Ukraine’s nuclear energy. To solve this problem, as well as to increase Ukraine’s energy independence, the construction of a centralized spent nuclear fuel storage facility is being completed in the Chornobyl exclusion zone, where the spent fuel of Khmelnytsky, Rivne and South Ukrainian nuclear power plants will be stored for the next 100 years. The technology of centralized storage of spent nuclear fuel is based on the storage of fuel assemblies in ventilated HI-STORM concrete containers manufactured by Holtec International. Long-term operation of a spent nuclear fuel storage facility requires a clear understanding of all processes (thermohydraulic, neutron-physical, aging processes, etc.) occurring in HI-STORM containers. And this cannot be achieved without modeling these processes using modern specialized programs. Modeling of neutron and photon transfer makes it possible to analyze the level of protective properties of the container against radiation, optimize the loading of MPC assemblies of different manufacturers and different levels of combustion and evaluate biological protection against neutron and gamma radiation. In the future it will allow to estimate the change in the isotopic composition of the materials of the container, which will be used for the management of aging processes at the centralized storage of spent nuclear fuel. The article is devoted to the development of the three-dimensional model of the HI-STORM storage system. The model was developed using the modern Monte Carlo code Serpent. The presented model consists of models of 31 spent fuel assemblies 438MT manufactured by TVEL company, model MPC-31 and model HISTORM 190. The model allows to perform a wide range of scientific tasks required in the operation of centralized storage of spent nuclear fuel.

scholarly journals SCIENTIFIC BASIS OF THERMAL SAFETY ANALYSIS OF DRY STORAGE OF SPENT NUCLEAR FUEL ON ZAPORIZHSKA NPP

2020 ◽  
pp. 81-84
Author(s):  
S. Alyokhina ◽  
A. Kostikov ◽  
I. Koriahina
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Thermal State ◽  
Thermal Safety ◽  
Spent Fuel ◽  
Storage Container ◽  
Dry Storage ◽  
Multi Stage ◽  
Thermal Investigations ◽  
Fuel Assemblies

Now only one Dry Storage Facility of Spent Nuclear Fuel (DSFSNF) is operated in Ukraine. It is the facility on Zaporizhska NPP. Many different thermal investigations were done for ventilated containers of DSFSNF. In this study the generalization of scientific approaches to the thermal safety assessment are carried out. The multi-stage approach to the definition of thermal state of containers' group, single container, spent fuel assemblies and fuel rods was developed. Detailed thermal profiles of spent fuel assemblies inside storage container were calculated. With usage of multi-stage approach the thermal simulations of the influence of outer factors onto thermal state of containers was carried out. Results of thermal investigations were generalized and factors, which are influence on thermal state of containers, are detected. The method of spent nuclear fuel thermal state prediction and suggestion for improving the system of thermal monitoring were proposed.

scholarly journals A Methodology for Optimizing the Management of Spent Fuel of Nuclear Power Plants Using Dry Storage Casks

2019 ◽  
Vol 2019 ◽  
pp. 1-13
Author(s):  
Ian B. Gomes ◽  
Pedro L. Cruz Saldanha ◽  
Antonio Carlos M. Alvim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Real Life ◽  
Final Decision ◽  
Spent Fuel ◽  
Dry Storage ◽  
Fuel Assemblies ◽  
Intermediate Storage

The management of spent nuclear fuel assemblies of nuclear reactors is a priority subject among member states of the International Atomic Energy Agency. For the majority of these countries, the destination of such fuel assemblies is a decision that is yet to be made and the “wait-and-see” policy is thus adopted by them. In this case, the irradiated fuel is stored in on-site spent fuel pools until the power plant is decommissioned or, when there is no more racking space in the pool, they are stored in intermediate storage facilities, which can be another pool or dry storage systems, until the final decision is made. The objective of this study is to propose a methodology that, using optimization algorithms, determines the ideal time for removal of the fuel assemblies from the spent fuel pool and to place them into dry casks for intermediate storage. In this scenario, the methodology allows for the optimal dimensioning of the designed spent fuel pools and the casks’ characteristics, thus reducing the final costs for purchasing new Nuclear Power Plants (NPP), as the size and safety features of the pool could be reduced and dry casks, that would be needed anyway after the decommissioning of the plant, could be purchased with optimal costs. To demonstrate the steps involved in the proposed methodology, an example is given, one which uses the Monte Carlo N-Particle code (MCNP) to calculate the shielding requirements for a simplified model of a concrete dry cask. From the given example, it is possible to see that, using real-life data, the proposed methodology can become a valuable tool to help making nuclear energy a more attractive choice costwise.

scholarly journals The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

MRS Advances ◽  
2018 ◽  
Vol 3 (19) ◽  
pp. 991-1003 ◽  
Author(s):  
Evaristo J. Bonano ◽  
Elena A. Kalinina ◽  
Peter N. Swift
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
United States Of America ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
The United States ◽  
Spent Fuel ◽  
Dry Storage ◽  
The Us

ABSTRACTCurrent practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.

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.

scholarly journals Temperature conditions in the RBMK spent fuel pool in the event of disturbances in its cooling mode

2021 ◽  
Vol 7 (1) ◽  
pp. 9-13
Author(s):  
David A. Hakobyan ◽  
Victor I. Slobodchuk
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Maximum Temperature ◽  
Spent Fuel ◽  
Cooling Mode ◽  
Temperature Conditions ◽  
Fuel Assemblies ◽  
Spent Fuel Pool

The problems of reprocessing and long-term storage of spent nuclear fuel (SNF) at nuclear power plants with RBMK reactors have not been fully resolved so far. For this reason, nuclear power plants are forced to search for new options for the disposal of spent fuel, which can provide at least temporary SNF storage. One of the possible solutions to this problem is to switch to compacted SNF storage in reactor spent fuel pools (SFPs). As the number of spent fuel assemblies (SFAs) in SFPs increases, a greater amount of heat is released. In addition, no less important is the fact that a place for emergency FA discharging should be provided in SFPs. The paper presents the results of a numerical simulation of the temperature conditions in SFPs both for compacted SNF storage and for emergency FA discharging. Several types of disturbances in normal SFP cooling mode are considered, including partial loss of cooling water and exposure of SFAs. The simulation was performed using the ANSYS CFX software tool. Estimates were made of the time for heating water to the boiling point, as well as the time for heating the cladding of the fuel elements to a temperature of 650 °С. The most critical conditions are observed in the emergency FA discharging compartment. The results obtained make it possible to estimate the time that the personnel have to restore normal cooling mode of the spent fuel pool until the maximum temperature for water and spent fuel assemblies is reached.

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