Similarity analysis of thermal-fluid flow for thermal testing using scaled-down model of spent fuel storage cask

2020 ◽  
Vol 149 ◽  
pp. 107791
Author(s):  
Ju-Chan Lee ◽  
Kyung-Sik Bang ◽  
Seung-Hwan Yu ◽  
Woo-Seok Choi ◽  
Sungho Ko
Keyword(s):  
Fluid Flow ◽  
Similarity Analysis ◽  
Thermal Testing ◽  
Spent Fuel ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Thermal Fluid Flow

Thermal-fluid flow analysis and demonstration test of a spent fuel storage system

2009 ◽  
Vol 239 (3) ◽  
pp. 551-558 ◽  
Author(s):  
J.C. Lee ◽  
W.S. Choi ◽  
K.S. Bang ◽  
K.S. Seo ◽  
S.Y. Yoo
Keyword(s):  
Fluid Flow ◽  
Storage System ◽  
Flow Analysis ◽  
Spent Fuel ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Thermal Fluid Flow ◽  
Fluid Flow Analysis

Influence of Gap Size on Added Mass for Spent Fuel Storage Rack

2018 ◽  
Author(s):  
Daogang Lu ◽  
Yu Liu ◽  
Shu Zheng
Keyword(s):  
Vibration Frequency ◽  
Input Parameter ◽  
Added Mass ◽  
Water Tank ◽  
Spent Fuel ◽  
Free Standing ◽  
Fluid Force ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Spent Fuel Pool

Free standing spent fuel storage racks are submerged in water contained with spent fuel pool. During a postulated earthquake, the water surrounding the racks is accelerated and the so-called fluid-structure interaction (FSI) is significantly induced between water, racks and the pool walls[1]. The added mass is an important input parameter for the dynamic structural analysis of the spent fuel storage rack under earthquake[2]. The spent fuel storage rack is different even for the same vendors. Some rack are designed as the honeycomb construction, others are designed as the end-tube-connection construction. Therefore, the added mass for those racks have to be measured for the new rack’s design. More importantly, the added mass is influenced by the layout of the rack in the spent fuel pool. In this paper, an experiment is carried out to measure the added mass by free vibration test. The measured fluid force of the rack is analyzed by Fourier analysis to derive its vibration frequency. The added mass is then evaluated by the vibration frequency in the air and water. Moreover, a two dimensional CFD model of the spent fuel rack immersed in the water tank is built. The fluid force is obtained by a transient analysis with the help of dynamics mesh method.

scholarly journals Antineutrino background from spent-fuel storage in sensitive searches for θ 13 at reactors

2006 ◽  
Vol 69 (2) ◽  
pp. 185-188 ◽  
Author(s):  
V. I. Kopeikin ◽  
L. A. Mikaelyan ◽  
V. V. Sinev
Keyword(s):  
Spent Fuel ◽  
Spent Fuel Storage ◽  
Fuel Storage

LWR Spent Fuel Storage Behaviour

1986 ◽  
Vol 137 (3) ◽  
pp. 190-202 ◽  
Author(s):  
M. Peehs ◽  
J. Fleisch
Keyword(s):  
Spent Fuel ◽  
Spent Fuel Storage ◽  
Fuel Storage

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.

Study on Radiation Dose Calculation of PWR Spent Fuel Storage and Transportation

2021 ◽  
Author(s):  
Wen Yang ◽  
Xing Li ◽  
Jinrong Qiu ◽  
Lun Zhou
Keyword(s):  
Radiation Dose ◽  
Normal Condition ◽  
Radiation Safety ◽  
Spent Fuel ◽  
Dose Rates ◽  
Storage Pool ◽  
Transport Container ◽  
Photon Dose ◽  
Spent Fuel Storage ◽  
Fuel Storage

Abstract With the rapid development of nuclear energy, spent fuel will accumulate in large quantities. Spent fuel is generally cooled and placed in a storage pool, and then transported to a reprocessing plant at an appropriate time. Because spent fuel is content with a high level of radiation, spent fuel storage and transportation safety play important roles in the nuclear safety. Radiation dose safety are checked and validated using source analysis and Monte Carlo method to establish a radiation dose rate calculation model for PWR spent fuel storage pool and transport container. The calculation results show that the neutron and photon dose rates decrease exponentially with increase of water level under normal condition of storage pool. The attenuation multiples of neutron and photon dose rates are 4.64 and 1.59, respectively. According to radiation dose levels in different water height situations, spent fuel pool under loss of coolant accident can be divides into five workplaces. They are supervision zone, regular zone, intermittent zone, restricted zone and radiation zone. Under normal condition of transport container, the dose rates at the surface of the container and at a distance of 1 m from the surface are 0.1759 mSv/h and 0.0732 mSv/h, respectively. The dose rates decrease with the increasing radius of break accident, and dose rate at the surface of the transport container is 0.278 mSv/h when the break radius is 20 cm. Transport container conforms to the radiation safety standards of International Atomic Energy Agency (IAEA). This study can provide some reference for radiation safety analysis of spent fuel storage and transportation.

scholarly journals RADIONUCLIDE CHARACTERISTICS OF RDE SPENT FUELS

2018 ◽  
Vol 20 (2) ◽  
pp. 69 ◽  
Author(s):  
Ihda Husnayani ◽  
Pande Made Udiyani
Keyword(s):  
Thermal Power ◽  
Neutron Production ◽  
Cooling Time ◽  
Spent Fuel ◽  
Release Rates ◽  
Radiation Dose Rate ◽  
The Core ◽  
Spent Fuel Storage ◽  
Fuel Storage

Reaktor Daya Eksperimental (RDE) is a 10 MWth pebble-bed High Temperature Gas-cooled Reactor that is planned to be constructed by National Nuclear Energy Agency of Indonesia (BATAN) in Puspiptek complex, Tangerang Selatan. RDE utilizes low enriched UO2 fuel coated by TRISO layers and loaded into the core by means of multipass loading scheme. Determination of radionuclide characteristics of RDE spent fuel; such as activity, thermal power, neutron and photon release rates; are very important because those characteristics are crucial to be used as a base for evaluating the safety of spent fuel handling system and storage tank. This study is aimed to investigate the radionuclide characteristics of RDE spent fuel at the end of cycle and during the first 5 years cooling time in spent fuel storage. The method used to investigate the radionuclide characteristics is burnup calculation using ORIGEN2.1 code. In performing the ORIGEN2.1 calculation, one pebble fuel was assumed to be irradiated in the core for 5 cycles and then decayed for 5 years. At the end of the fifth cycle, it is obtained that the total activity, thermal power, neutron production, and photon release rates from all radionuclides inside one spent fuel are approximately 105.68 curies, 0.41 watts, 2.65 x 103 neutrons/second, and 1.79 x 104 photons/second, respectively. The results for the radionuclides characteristics during the first 5 years cooling time in the spent fuel storage show that the radioactivity characteristics from all radionuclides are rapidly decreasing at the first year and then slowly decreasing at the second until the fifth year of cooling time. The results obtained in this study can provide data for safety evaluation of fuel handling and spent fuel storage, such as the calculation of sourceterm, radiation dose rate, and the determination of radiation shielding.Keywords: RDE, spent fuel, radionuclide activity, thermal power, neutron production, photon releaserates KARAKTERISTIK RADIONUKLIDA DI DALAM BAHAN BAKAR RDE. Reaktor Daya Eksperimental (RDE) adalah reaktor tipe Reaktor Temperatur Tinggi Berpendingin Gas dengan daya termal 10MW yang akan dibangun oleh BadanTenagaNuklirNasional (BATAN) di kawasanPuspiptek, Tangerang Selatan. RDE menggunakan bahan bakar UO2 yang dilapisi dengan lapisan TRISO dan dimasukkan ke dalam teras RDE menurut skema multipass (5 siklus). Penentuan karakteristik radionuklida di dalam bahan bakar RDE; seperti aktivitas, daya termal, laju produksi neutron dan pelepasan foton; adalah sangat penting karena informasi karakteristik ini diperlukan sebagai dasar untuk melakukan evaluasi keselamatan system penanganan dan penyimpanan bahan bakar bekas. Penelitian ini bertujuan untuk menganalisis karakteristik radionuklida bahanbakar RDE setelah 5 siklus dan pada 5 tahun pertama pendinginan ditempat penyimpanan bahan bakar bekas. Metode yang digunakan dalam menghitung karakteristik radionuklida adalah menggunakan program ORIGEN2.1. Satu bola bahan bakar RDE diasumsikan diiradiasi selama 5 siklus dan kemudian meluruh selama 5 tahun. Pada akhir siklus, diperoleh hasil aktivitas total, daya termal, laju produksi neutron dan pelepasan foton dari seluruh radionuklida di dalam satu bola bahan bakar RDE sebesar 105,68 curies, 0,41 watts, 2,65 x 103 neutron/detik, dan 1,79 x 104 foton/detik. Hasil untuk karakteristik radionuklida selama 5 tahun penyimpanan menunjukkan bahwa karakteristik radioktivitas radionuklida menurun dengan cepat pada tahun pertama dan kemudian menurun lebih lambat pada tahun kedua hingga tahun kelima. Hasil perhitungan karakteristik radionuklida dari penelitian ini dapat digunakan sebagai basis untuk analisis keselamatan penanganan dan penyimpanan bahan bakarbekas RDE.Kata kunci:RDE, bahan bakar bekas, aktivitas radionuklida, daya termal, produksi neutron, laju foton

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