Analisis Kuat Sumber Neutron Dan Perhitungan Laju Dosis Neutron Teras Awal Rde

Suwoto, Suwoto; Adrial, Hery; Zuhair, Zuhair

doi:10.17146/urania.2017.23.1.3119

Cited by 13 publications

(4 citation statements)

References 17 publications

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“…The graphite reflector and other reactor geometry, such as carbon layer around the system, helium channels and control rods, were modelled in detail and comprehensively. The modelling procedure in this study was first developed by Lebenhaft [16] and utilized in some publications [17][18][19][20][21][22][23][24][25][26][27][28][29][30]. Figure 4 illustrates the MCNP6 model for the second heterogeneity of the pebble bed reactor.…”

Section: Fuel Kernelmentioning

confidence: 99%

Study on McNp6 Model in the Calculation of Kinetic Parameters for Pebble Bed Reactor

et al. 2020

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When conducting a nuclear reactor transient analysis, the most important parameter, called the kinetic parameter, is required. The calculation of kinetic parameters can be conducted using several methods. The deterministic method is one possible method that relies on the forward and adjoint neutron fluxes to provide the kinetic parameters calculation based on the perturbation theory. In this study, the Monte Carlo transport code MCNP6 was utilized to perform the exact prediction of the kinetic parameters of a pebble bed reactor. The core was modelled with a different fuel composition of uranium loading per pebble, 235U enrichment and H/D ratio. It was found that keff strongly depends on the uranium loading, uranium enrichment and H/D ratio while the βeff dependence is insignificant. The increase in the prompt neutron lifetime (ℓ) and mean generation time (Ʌ) as a function of H/D ratio are insignificant as compared to the decrease of those parameters in the case of uranium loading or uranium enrichment. These results conclude that the selection of uranium loading per pebble, 235U enrichment and H/D ratio should be considered carefully for the control and inherent safety performances

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Section: Fuel Kernelmentioning

confidence: 99%

Study on McNp6 Model in the Calculation of Kinetic Parameters for Pebble Bed Reactor

et al. 2020

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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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“…Despite both of them use solid oxide fuel, the former shaped the fuel into microencapsulated particles called TRISO (tri-isotropic) dispersed within a graphite ball (pebble). The latter, however, have the fuel pelletised and stacked within zirconium-alloy fuel rod, then arranged into a fuel bundle [1][2][3].…”

Section: Introductionmentioning

confidence: 99%

Characteristics of Radionuclides on Thorium-Cycle Experimental Power Reactor Spent Fuel.

Dwijayanto

Husnayani²,

Zuhair³

2019

urania

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CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL. There are several options of nuclear fuel utilisation in the HTGR-based Experimental Power Reactor (Reaktor Daya Eksperimental/RDE). Although mainly RDE utilises low enriched uranium (LEU)-based fuel, which is the most viable option at the moment, it is possible for RDE to utilise other fuel, for example thorium-based and possibly even plutonium-based fuel. Different fuel yields different spent fuel characteristics, so it is necessary to identify the characteristics to understand and evaluate their handling and interim storage. This paper provides the study on the characteristics of thorium-fuelled RDE spent fuel, assuming typical operational cycle. ORIGEN2.1 code is employed to determine the spent fuel characteristics. The result showed that at the end of the calculation cycle, each thorium-based spent fuel pebble generates around 0,627 Watts of heat, 28 neutrons/s, 8.28x1012 photons/s and yield 192.53 curies of radioactivity. These higher radioactivity and photon emission possibly necessitate different measures in spent fuel management, if RDE were to use thorium-based fuel. Tl-208 activity, which found to be emitting potentially non-negligible strong gamma emission, magnified the requirement of proper spent fuel handling especially radiation shielding in spent fuel cask.Keywords: RDE, spent fuel, thorium, HTGR, Tl-208.

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Analisis Kuat Sumber Neutron Dan Perhitungan Laju Dosis Neutron Teras Awal Rde

Cited by 13 publications

References 17 publications

Study on McNp6 Model in the Calculation of Kinetic Parameters for Pebble Bed Reactor

Study on McNp6 Model in the Calculation of Kinetic Parameters for Pebble Bed Reactor

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

Characteristics of Radionuclides on Thorium-Cycle Experimental Power Reactor Spent Fuel.

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