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.

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.

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 COMPARATIVE ANALYSIS OF NUMERICAL METHODS USED FOR THERMAL MODELING OF SPENT NUCLEAR FUEL DRY STORAGE SYSTEMS

2019 ◽  
pp. 75-79
Author(s):  
S. Alyokhina ◽  
Y. Matsevity ◽  
V. Dudkin ◽  
R. Poskas ◽  
A. Sirvydas ◽  
...  
Keyword(s):  
Comparative Analysis ◽  
Nuclear Fuel ◽  
Nuclear Energy ◽  
Storage Time ◽  
Spent Nuclear Fuel ◽  
Storage Systems ◽  
Storage Technology ◽  
Dry Storage ◽  
Fuel Storage ◽  
Interim Storage

Management of spent nuclear fuel is a very important part in the whole cycle of nuclear energy generation. Usually “dry” storage technology in casks is selected for the interim storage of spent nuclear fuel for up to 50 years after pre-storage time in water pools. In this paper, two case studies were carried out to highlight the differences and similarities between Ukraine and Lithuania in spent nuclear fuel storage.

scholarly journals Vibrational modes for the internal characterization of a full-scale Transnuclear-32 dry storage cask for spent nuclear fuel assemblies

2019 ◽  
Vol 460 ◽  
pp. 114881 ◽  
Author(s):  
Kevin Yi-Wei Lin ◽  
Joel Mobley ◽  
Wayne E. Prather ◽  
Zhiqu Lu ◽  
Gautam Priyadarshan ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Full Scale ◽  
Vibrational Modes ◽  
Dry Storage ◽  
Fuel Assemblies

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 VERIFICATION AND VALIDATION OF BACK-END CYCLE SOURCE TERM CALCULATION OF THE NODAL CODE RAST-K

2021 ◽  
Vol 247 ◽  
pp. 10025
Author(s):  
Jaerim Jang ◽  
Bamidele Ebiwonjumi ◽  
Wonkyeong Kim ◽  
Jinsu Park ◽  
Deokjung Lee
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
Interpolation Method ◽  
Prediction Method ◽  
Computation Time ◽  
Verification And Validation ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Decay Heat ◽  
Fuel Assemblies

Verification and validation (V&V) results of source term calculation capability implemented in the nodal diffusion code RAST-K are presented in this paper. An isotope inventory prediction method is presented in this work which is implemented with RAST-K and the lattice code STREAM. STREAM generates cross-section and provides number density information by history branch calculations. RAST-K supplies the power history and three history indexes (boron concentration, moderator temperature and fuel temperature). The main feature of the newly implemented spent nuclear fuel (SNF) characterization is the direct consideration of three-dimensional (3D) core simulation conditions by using operation history information. As a result of this, it could reduce the computation time. The implemented SNF analysis capability have two main functions. The first is to predict isotope inventory by Lagrange non-linear interpolation method, using power history correction factors. The second is to calculate the radiological response activity, decay heat, and neutron/gamma source strengths. The V&V of these two functions are thus presented herein. The isotope inventory prediction is validated with measured data from ten SNF samples of Takahama-3 and six samples of Calvert Cliffs-1 pressurized water reactors (PWR). Eighteen decay heat measurements of Ringhals Unit 3 PWR fuel assemblies are then employed to validate the decay heat calculation results. In addition, STREAM is employed in a code-to-code comparison for verification. The fuel assemblies cover the burnup range 14.3 - 47.25 GWd/tU, initial enrichment of 2.1 - 4.11 235U w/o and cooling time of 3.96 to 20.01 years. The comparison to STREAM shows the accuracy of the RAST-K SNF and prediction of the decay heat is within 4%. Overall, this paper demonstrates that RAST-K SNF calculation can be applied to the back-end cycle source term analysis.

scholarly journals Time-saving method of orbital thermal regime calculations of nanosatellites as exemplified by a 3U CubeSat

2018 ◽  
Vol 158 ◽  
pp. 01012
Author(s):  
Vasily Gorev ◽  
Vitaly Prokopyev ◽  
Yury Prokopyev ◽  
Alexey Sidorchuk
Keyword(s):  
Thermal Conductivity ◽  
Simple Structure ◽  
Orbital Motion ◽  
Thermal Conditions ◽  
Thermal Regulation ◽  
Structural Elements ◽  
Time Saving ◽  
Spectrally Selective ◽  
Thermal Calculations ◽  
Structure Geometry

A time-saving approach to perform technical calculations of thermal conditions of orbital motion of 3U CubeSat nanosatellite was applied, which made it possible to make the thermal calculations of a satellite with simple structure geometry using MatLab and SolidWorks Simulation. Passive thermal regulation facilities are sufficient for a 3U CubeSat to provide thermal conductivity of the case’s structural elements and to remove heat from the lighted surface and internal components to the satellite’s shadowed surface. Application of spectrally selective coatings allows narrowing the range of surface temperatures of 3U CubeSat.

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.

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