scholarly journals Natural Safety Analysis of the Spent Fuel Residual Heat Removal in Loading and Storage Process of HTR-10

2013 ◽  
Vol 39 ◽  
pp. 227-239 ◽  
Author(s):  
Jin-hua Wang ◽  
Yi-fan Huang ◽  
Yong Tang ◽  
Bin Wu
Keyword(s):  
Heat Removal ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Storage Process ◽  
And Storage ◽  
Residual Heat

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.

Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask

2018 ◽  
Vol 114 ◽  
pp. 495-509 ◽  
Author(s):  
Jaerim Jang ◽  
Wonkyeong Kim ◽  
Sanggeol Jeong ◽  
Eun Jeong ◽  
Jinsu Park ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Safety Analysis ◽  
Monte Carlo Code ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
And Storage ◽  
Criticality Safety

Numerical simulation of residual heat removal from the spent fuel assemblies storage pit

2019 ◽  
Author(s):  
A. V. Kurteev ◽  
V. A. Klimova ◽  
M. M. Sevastyanov ◽  
O. L. Tashlykov
Keyword(s):  
Numerical Simulation ◽  
Heat Removal ◽  
Spent Fuel ◽  
Fuel Assemblies ◽  
Residual Heat

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.

Detailed Experimental and Analytical Study on Long Two-Phase Closed Thermosiphons Related to Passive Spent-Fuel Pool Cooling

2018 ◽  
Author(s):  
Claudia Graß ◽  
Anne Krüssenberg ◽  
Rudi Kulenovic ◽  
Fabian Weyermann ◽  
Jörg Starflinger ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
High Heat ◽  
Heat Transfer Process ◽  
Normal Operation ◽  
Working Fluid ◽  
Spent Fuel ◽  
Two Phase ◽  
Spent Fuel Pool ◽  
Residual Heat

New concepts are currently being discussed for passive residual-heat removal with heat pipes from spent-fuel pools and wet-storage facilities. Because of their high heat-transport capability and their simple design, two-phase closed thermosiphons have a great potential to satisfy the demands of a reliable and independent passive heat removal. The geometry of spent-fuel pools and the potential incorporation into existing plants requires thermosiphons of at least 10 m in length including bends. Such thermosiphons are neither available nor have they been investigated yet. Therefore, experimental and numerical investigations are being carried out. At IKE the basic operational behavior of 10-m-long thermosiphons with water — as working fluid — are being experimentally investigated. Measurements for different pipe diameters (32 mm and 45 mm) are performed at various heat sink temperatures (10 °C, 20 °C and 30 °C), heat inputs (1000 W to 4000 W), and filling ratios (50%, 70% and 100%). GRS is developing codes, such as AC2, in order to simulate all relevant phenomena within a nuclear power plant during normal operation, incidents, accidents, and severe accidents. Regarding passive residual-heat removal with thermosiphons, the models of AC2 are being improved to properly simulate the thermos-hydraulics of this heat transfer process. Starting with the module ATHLET (Analysis of Thermal-hydraulic of Leaks and Transients), the applicability of its existing models is checked for modeling long thermosiphons and calculating their operational behavior. The main model improvements are being validated against the new experiments of IKE.

Fueling Machine Model for Simulation of Spent Fuel Behavior in Postulated Fuel Handling Accidents

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
L. Y. Huang ◽  
H. Z. Fan ◽  
M. Maltchevski ◽  
A. Ranger
Keyword(s):  
Fission Product ◽  
Heat Removal ◽  
Safety Analysis ◽  
Complex Configuration ◽  
Integral Approach ◽  
Spent Fuel ◽  
Machine Model ◽  
Product Release ◽  
Fuel Behavior ◽  
In Transit

Abstract On-power fueling is an important feature of the CANDU® reactor. Fueling is a routine operation with a large number of channel fueling visits made each year with the fueling machines acting as the key system. Hence, safety analysis must consider fueling machine events typically when the fueling machine is in transit toward the spent fuel port. This paper presents a model of fueling machine containing spent fuel with complex configuration and multiprocess mechanisms. Using an integral approach with fuel and fueling machine, this model tends to improve previous modeling method, which only takes account of a slice of fuel or fueling machine. This fueling machine model is developed for simulations of the fueling machine coolant thermal hydraulics behavior, the spent fuel behavior, and potential fission product release during postulated loss of heat removal accidents. An example of its application is also presented in this paper.

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.

scholarly journals CRITICALITY SAFETY ANALYSIS OF THE DRY CASK DESIGN WITH AIR GAPS FOR RDNK SPENT PEBBLE FUELS STORAGECriticality Safety Analysis of the Dry Cask Design with Air Gaps for RDNK Spent Pebble Fuels Storage

2021 ◽  
Vol 23 (3) ◽  
pp. 123
Author(s):  
Pungky Ayu Artiani ◽  
Yuli Purwanto ◽  
Aisyah Aisyah ◽  
Ratiko Ratiko ◽  
Jaka Rachmadetin ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Thermal Power ◽  
Heat Removal ◽  
Threshold Value ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Air Gaps ◽  
Fuel Elements ◽  
Submerged Conditions ◽  
Criticality Safety

Reaktor Daya Non-Komersial (RDNK) with a 10 MW thermal power has been proposed as one of the technology options for the first nuclear power plant program in Indonesia. The reactor is a High Temperature Gas-Cooled Reactor-type with spherical fuel elements called pebbles. To support this program, it is necessary to prepare dry cask to safely store the spent pebble fuels that will be generated by the RDNK. The dry cask design has been proposed based on the Castor THTR/AVR but modified with air gaps to facilitate decay heat removal. The objective of this study is to evaluate criticality safety through keff  value of the proposed dry cask design for the RDNK spent fuel. The keff  values were calculated using MCNP5 program for the dry cask with 25, 50, 75, and 100% of canister capacity. The values were calculated for dry casks with and without air gaps in normal, submerged, tumbled, and both tumbled and submerged conditions. The results of calculated keff  values for the dry cask with air gaps at 100% of canister capacity from the former to the latter conditions were 0.127, 0.539, 0.123, and 0.539, respectively. These keff values were smaller than the criticality threshold value of 0.95. Therefore, it can be concluded that the dry cask with air gaps design comply the criticality safety criteria in the aforementioned conditions.

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