Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors

Mohammadi, Ali Akbar; Hassanzadeh, M.; Gharib, Morteza

doi:10.1016/j.apradiso.2015.12.045

Cited by 13 publications

(3 citation statements)

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“…The neutron shielding performance evaluation was carried out by comparing the neutron transmittance of SS316L and Cd-SS316L which were calculated using MCNP5. This software has been widely used for gamma and neutron shielding evaluation [8]- [12].…”

Section: Introductionmentioning

confidence: 99%

Evaluation of Neutron Shielding Performance of Cd-Ss 316l as a Candidate Alloy for Dry Cask of Research Reactor Spent Fuel

Heriyanto

Sudjadi²,

Rachmadetin³

et al. 2021

urania

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EVALUATION OF NEUTRON SHIELDING PERFORMANCE OF CD-SS 316L AS A CANDIDATE ALLOY FOR DRY CASK OF RESEARCH REACTOR SPENT FUEL Development of dry casks is necessary to support the national strategy for management of spent fuels. One of the requirements for the dry cask is shielding performance for neutron emitted by the spent fuels to be stored in the dry cask. The objectives of this study are to determine the emitted neutrons by the spent fuel generated from GAS research reactor and to evaluate the neutron shielding performance of Cd-SS316L alloy as a candidate material to be used in dry cask for the spent fuels. The former was carried out using Origen 2.1 software, while the latter using MCNP5. The result shows that the emitted neutrons by a spent fuel after 5 years discharged from GAS research reactor were 2.81×103 and 3.32×106 n/s for reactor core power of 15 and 30 MW, respectively. Addition of Cd improves the neutron shielding performance of SS 316L. The evaluation of neutron shielding performance of SS 316L with addition of Cd which is the candidate material for dry cask of the spent fuels from the GAS research reactor can be evaluated using Origen 2.1 software for neutron emission, while the neutron shielding performance was evaluated by the simulation using MNCP 5 software. This study shows the Cd-SS 316L alloy can be used for further study to develop the dry cask design for the GAS research reactor.Key words: Neutron shielding, cadmium, stainless steel, spent fuel.

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Section: Introductionmentioning

confidence: 99%

Evaluation of Neutron Shielding Performance of Cd-Ss 316l as a Candidate Alloy for Dry Cask of Research Reactor Spent Fuel

Heriyanto

Sudjadi²,

Rachmadetin³

et al. 2021

urania

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“…The purpose of this study is to assess the criticality of the dry cask designs to be used for RDNK spent fuel through determination of their k eff values. The values were calculated using MCNP5 program which had already been used in analyzing the criticality both at the nuclear reactor and at the non-nuclear reactor facility [16][17][18][19]. The calculations were performed for the dry cask designs with and without air gaps in normal and abnormal conditions.…”

Section: Introduction mentioning

confidence: 99%

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

Artiani

Purwanto²,

Aisyah³

et al. 2021

Tri Dasa Mega

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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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“…Thermal power quantification of spent fuel is used as the basis for safety analysis of heat removal in the spent fuel storage tank [9,10]. Neutron production and photon release rates are also crucial for shielding analysis of spent fuel cask [11,12]. In this study, the investigation of radionuclide characteristics was performed using ORIGEN2.1 code, a well-known and widely used computer code for the analysis of spent fuel composition and radiological characteristics necessary for nuclear facility design and safety analysis [13].…”

Section: Introductionmentioning

confidence: 99%

Radionuclide Characteristics of Rde Spent Fuels

Husnayani

Udiyani

2018

Tri Dasa Mega

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RADIONUCLIDE CHARACTERISTICS OF RDE SPENT FUELS. 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 UO 2 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 10 3 neutrons/second, and 1.79 x 10 4 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 ABSTRAK 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 UO 2 yang dilapisi dengan lapisan TRISO dan dimasukkan ke dalam teras RDE menurut skema multipass (5 siklus

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Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors

Cited by 13 publications

References 1 publication

Evaluation of Neutron Shielding Performance of Cd-Ss 316l as a Candidate Alloy for Dry Cask of Research Reactor Spent Fuel

Evaluation of Neutron Shielding Performance of Cd-Ss 316l as a Candidate Alloy for Dry Cask of Research Reactor Spent Fuel

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

Radionuclide Characteristics of Rde Spent Fuels

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