scholarly journals RESIDUAL HEAT POWER REMOVAL FROM SPENT NUCLEAR FUEL DURING DRY AND WET STORAGE

2018 ◽  
Vol 19 ◽  
pp. 36
Author(s):  
Daniel Vlček
Keyword(s):  
Nuclear Fuel ◽  
Computational Models ◽  
Spent Nuclear Fuel ◽  
Thermal Analyses ◽  
Heat Removal ◽  
Regulatory Body ◽  
Channel Code ◽  
Spent Fuel ◽  
Fuel Storage ◽  
Residual Heat

This project deals with the thermal analyses of the wet and dry storages of the spent nuclear fuel. The dry spent fuel storage sub-channel code COBRA-SFS has been used in order to calculate the temperature field. In this code, the new model of residual heat removal was created for the SKODA 1000/19 cask where the spent nuclear fuel TVSA-T type from NPP Temelin will be stored. The object of calculations was to obtain the inside temperatures under maximum loads. After that, the results were compared to the requirements of the local regulatory body. Because of the absence of experimental data, the validation of the created computational models could not be accomplished. However, according to the verification scheme of the COBRA-SFS authors, the verification of the new models was implemented.

Rationalization of Radiation Shielding Design for Spent Nuclear Fuel Storage Building

2008 ◽  
Author(s):  
Yuji Nemoto ◽  
Toshihisa Tsukiyama ◽  
Shigeki Nemezawa ◽  
Hideo Nakano
Keyword(s):  
Nuclear Fuel ◽  
Seismic Design ◽  
Scattered Radiation ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Radiation Shielding ◽  
Air Supply ◽  
Exhaust Duct ◽  
Fuel Storage ◽  
Shielding Design

A spent nuclear fuel storage building is generally a structure provided with heat removal, radiation shielding functions, and a seismic design. The facility has air supply ducts and a chimney exhaust duct for removing heat through natural cooling, with the structural shapes determined by the demands of the above functions. In radiation protection design, radiation levels must be kept below acceptable levels for the general public. The radiation dose can be lowered by increasing the concrete thickness, as typically applied in radiation shielding design, or by increasing the distance. The setting of additional shielding plates also helps reduce the scattered radiation escaping from the air supply and exhaust ducts. However, such protective structures against the scattered radiation challenges in effective heat removal and seismic design. Therefore, determining a building structure that can satisfy all safety demands requires a great deal of time. This study aims to effectively achieve radiation shielding. The height and width of the exhaust duct were considered, and the correlation of these parameters was studied. In the calculation, two-dimensional Sn code (DORT) was used to examine the validity of the results. Monte Carlo N-Particle Transport Code (MCNP) calculations were also made for comparison with the DORT results. The cross-section libraries of JENDL3.3 and DLC23F were used in this calculation, with the difference being clarified.

Numerical Analysis of the Residual Heat Removal for the Spent Fuel Storage Tank of HTR-10

2013 ◽  
Author(s):  
Jinhua Wang ◽  
Bing Wang ◽  
Bin Wu ◽  
Jiguo Lui
Keyword(s):  
Storage Tank ◽  
Natural Ventilation ◽  
Heat Removal ◽  
Fuel Tank ◽  
Spent Fuel ◽  
Loading Process ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Interim Storage ◽  
Residual Heat

Chinese 10 MW High Temperature Gas Cooled Reactor (HTR-10) has inherent safety; the residual heat of the spent fuel could be removed by natural ventilation in loading process. The spent fuel storage tank could shield radiation; the outside is covered by an iron sleeve; the spent fuel tank would be stored in atmosphere after fully loaded, and the residual heat could be discharged by natural ventilation in interim storage stage. The calculation showed that, the maximum temperature locates in the middle of the fuel pebble bed in the spent fuel tank in loading process and interim storage stage, and the temperature decrease gradually with radial distance; the temperature in the tank body and sleeve is evenly; it is feasible to remove the residual heat of the spent fuel tank by natural ventilation, and in the natural ventilation condition, the temperature of the spent fuel and the tank is lower than the temperature limit, which provides theoretical evidence for the choice of the residual heat removal method in loading process and interim storage stage.

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.

Modelling the Activation of H2 on Spent Fuel Surface and Inhibiting Effect of UO2 Dissolution

2013 ◽  
Vol 1518 ◽  
pp. 133-138 ◽  
Author(s):  
L. Duro ◽  
O. Riba ◽  
A. Martínez-Esparza ◽  
J. Bruno
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Pellet ◽  
Dissolution Behavior ◽  
Dissolution Mechanism ◽  
Water Radiolysis ◽  
Spent Fuel ◽  
Secondary Phases ◽  
The Matrix ◽  
Fuel Storage

ABSTRACTThe dissolution of spent nuclear fuel is defined in two different time steps, i) the Instant Release Fraction (IRF) occurring shortly after water contacts the solid spent fuel and responsible of the fast release of those radionuclides that have been accumulated in the zones of the spent fuel pellet with low confinement, such as gap and grain boundaries and ii) the long term release of radionuclides confined in the spent fuel matrix, much slower and dependent on the conditions of the water that contacts the spent fuel.Several models have been developed to date to explain the dissolution behavior of spent nuclear fuel under disposal conditions. The Matrix Alteration Model (MAM) is one of the most evolved radiolytic models describing the dissolution mechanism in which an Alteration/Dissolution source term model is based on the oxidative dissolution of spent fuel. Under deep repository conditions and at the expected of water contacting time (after 1000 years of spent fuel storage), α radiation will be the main contributor to water radiolysis. In the current study, simulations evaluating the effect of surface area on the alteration/dissolution of spent fuel matrix are performed considering different particle sizes of spent fuel and simulations integrating the actinides dissolution have been performed considering the precipitation of secondary phases.

Japanese Codes and Standards for Dual Purpose Metal Casks for Spent Nuclear Fuel

2014 ◽  
Author(s):  
Toshiari Saegusa ◽  
Makoto Hirose ◽  
Norikazu Irie ◽  
Masashi Shimizu
Keyword(s):  
Nuclear Fuel ◽  
Power Plants ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Spent Fuel ◽  
Dual Purpose ◽  
Service Conditions ◽  
Interim Storage ◽  
And Storage

The first Japanese spent fuel interim storage facility away from a reactor site is about to be commissioned in Mutsu City, Aomori Prefecture. In designing, licensing and construction of the Dual Purpose Casks (DPCs, for transport and storage) for this facility, codes and standards established by the Atomic Energy Society of Japan (AESJ) and by the Japan Society of Mechanical Engineers (JSME) have been applied. The AESJ established the first standard for DPCs as “Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facilities” in 2002 (later revised in 2010). The standard provides the design requirements to maintain the basic safety functions of DPCs, namely containment, heat removal, shielding, criticality prevention and the structural integrity of the cask itself and of the spent fuel cladding during transport and storage. Inspection methods and criteria to ensure maintenance of the basic safety functions and structural integrity over every stage of operations involving DPCs including pre-shipment after storage are prescribed as well. The structural integrity criteria for major DPC components refer to the rules provided by the JSME. JSME completed the structural design and construction code (the Code) for DPCs as “Rules on Transport/Storage Packagings for Spent Nuclear Fuel” in 2001 (later revised in 2007). Currently, the scope of the rules cover the Containment Vessel, Basket, Trunnions and Intermediate Shell as major components of DPCs. Rules for these components are based on those for components of nuclear power plants (NPP) with similar safety functions, but special considerations based on their shapes, loading types and required functions are added. The Code has differences from that for NPP components with considerations to DPC characteristics; - The primary stress and the secondary stress generated in Containment Vessels shall be evaluated under Service Conditions A to D (from ASME Sec III, Div.1). - Stress generated in the seal region lid bolts of Containment Vessels shall not exceed yield strength under Service Conditions A to D in order to maintain the containment function. - Fatigue analysis on Baskets is not required, and Trunnions can be designed only for Service Conditions A and B with special stress limits consistent with conventional assessment methods for transport packages. - Stress limits for earthquakes during storage are specified. - Ductile cast iron with special fracture toughness requirements can be used as a material for Containment Vessels. DPC specific considerations in standards and rules will be focused on in this paper. Additionally, comparison with the ASME Code will be discussed.

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 Prevention of Damage to Spent Nuclear Fuel during Handling Operations

2020 ◽  
pp. 62-71
Author(s):  
M. Sapon ◽  
O. Gorbachenko ◽  
S. Kondratyev ◽  
V. Krytskyy ◽  
V. Mayatsky ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Long Term Storage ◽  
Fuel Storage ◽  
Storage Facilities

According to regulatory requirements, when carrying out handling operations with spent nuclear fuel (SNF), prevention of damage to the spent fuel assemblies (SFA) and especially fuel elements shall be ensured. For this purpose, it is necessary to exclude the risk of SFA falling, SFA uncontrolled displacements, prevent mechanical influences on SFA, at which their damage is possible. Special requirements for handling equipment (in particular, cranes) to exclude these dangerous events, the requirements for equipment strength, resistance to external impacts, reliability, equipment design solutions, manufacturing quality are analyzed in this work. The requirements of Ukrainian and U.S. regulatory documents also are considered. The implementation of these requirements is considered on the example of handling equipment, in particular, spent nuclear fuel storage facilities. This issue is important in view of creation of new SNF storage facilities in Ukraine. These facilities include the storage facility (SFSF) for SNF from water moderated power reactors (WWER): a Сentralized SFSF for storing SNF of Rivne, Khmelnitsky and South-Ukraine Nuclear Power Plants (СSFSF), and SFSF for SNF from high-power channel reactors (RBMK): a dry type SFSF at Chornobyl nuclear power plant (ISF-2). After commissioning of these storage facilities, all spent nuclear fuel from Ukrainian nuclear power plants will be placed for long-term “dry” storage. The safety of handling operations with SNF during its preparation for long-term storage is an important factor.

scholarly journals Nuclear and Radiation Safety of the Centralized Spent Fuel Storage Facility in Ukraine

2017 ◽  
pp. 3-10
Author(s):  
O. Hryhorash ◽  
O. Dybach ◽  
S. Kondratiev ◽  
O. Horbachenko ◽  
A. Panchenko ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Radiation Safety ◽  
Storage Facility ◽  
Spent Fuel ◽  
Regulatory Documents ◽  
Thermal Regimes ◽  
Fuel Storage ◽  
Nuclear And Radiation Safety ◽  
Biological Protection

The paper presents the analysis of ensuring nuclear and radiation safety in the management of spent nuclear fuel at the Centralized SFSF and activities planned for Centralized SFSF lifecycle stages. There are results of comparing requirements of U.S. regulatory documents used by the HOLTEC Company to design Centralized SFSF equipment staff with relevant requirements of Ukrainian regulations, results based on analysis of the most important factors of Centralized SFSF safety (strength and reliability, nuclear safety, thermal regimes and biological protection) and verified expert calculations of the SSTC NRS. The paper includes issues to be considered in further implementation of Centralized SFSF project.

A Comparative Evaluation of Licensing Requirements for Dry Storage of Spent Nuclear Fuel in the United States, Germany, Canada and the Russian Federation

2004 ◽  
Author(s):  
Edward Wonder ◽  
David S. Duncan ◽  
Eric A. Howden
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
The United States ◽  
Approval Process ◽  
Spent Fuel ◽  
Regulatory Requirements ◽  
Fuel Storage ◽  
The Russian Federation ◽  
The Republic

Technical activities to support licensing of dry spent nuclear fuel storage facilities are complex, with policy and regulatory requirements often being influenced by politics. Moreover, the process is often convoluted, with numerous and diverse stakeholders making the licensing activity a difficult exercise in consensus-reaching. The objective of this evaluation is to present alternatives to assist the Republic of Kazakhstan (RK) in developing a licensing approach for a planned Dry Spent Fuel Storage Facility. Because the RK lacks experience in licensing a facility of this type, there is considerable interest in knowing more about the approval process in other countries so that an effective, non-redundant method of licensing can be established. This evaluation is limited to a comparison of approaches from the United States, Germany, Russia, and Canada. For each country considered, the following areas were addressed: siting; fuel handling and cask loading; dry fuel storage; and transportation of spent fuel. The regulatory requirements for each phase of the process are presented, and a licensing approach that would best serve the RK is recommended.

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