• Title/Summary/Keyword: Neutron absorber

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A Study on the Neutron Detection by change of Asphalt Content (아스팔트 함량 변화에 따른 중성자 검출에 관한 연구)

  • Kim, Ki-Joon
    • Journal of the Korea Computer Industry Society
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    • v.8 no.1
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    • pp.9-16
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    • 2007
  • In this study, the change of neutron detection can be use the basic data of asphalt content detector under the influence of the jurisdiction and usage of radioisotopes are limited of $100[{\mu}Ci]$ or less. To obtain neutron detector's properties using design materials in first step phase, the change of neutron detection is to be calculated how can be increase or decrease due to the change of asphalt content, also it look over the change results which is installed absorber(cadmium plate) around moderator(polyethylene) using MCNP Code.

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Burnable Absorber Design Study for a Passively-Cooled Molten Salt Fast Reactor

  • Nariratri Nur Aufanni;Eunhyug Lee;Taesuk Oh;Yonghee Kim
    • Nuclear Engineering and Technology
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    • v.56 no.3
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    • pp.900-906
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    • 2024
  • The Passively-Cooled Molten Salt Fast Reactor (PMFR) is one of the advanced design concepts of the Molten Salt Fast Reactor (MSFR) which utilizes a natural circulation for the primary loop and aims to attain a long-life operation without any means of fuel reprocessing. For an extended operation period, it is necessary to have enough fissile material, i.e., high excess reactivity, at the onset of operation. Since the PMFR is based on a fast neutron spectrum, direct implementation of a burnable absorber concept for the control of excess reactivity would be ineffective. Therefore, a localized moderator concept that encircles the active core has been envisioned for the PMFR which enables the effective utilization of a burnable absorber to achieve low reactivity swing and long-life operation. The modified PMFR design that incorporates a moderator and burnable absorber is presented, where depletion calculation is performed to estimate the reactor lifetime and reactivity swing to assess the feasibility of the proposed design. All the presented neutronic analysis has been conducted based on the Monte Carlo Serpent2 code with ENDF/B-VII.1 library.

An approach to minimize reactivity penalty of Gd2O3 burnable absorber at the early stage of fuel burnup in Pressurized Water Reactor

  • Nabila, Umme Mahbuba;Sahadath, Md. Hossain;Hossain, Md. Towhid;Reza, Farshid
    • Nuclear Engineering and Technology
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    • v.54 no.9
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    • pp.3516-3525
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    • 2022
  • The high capture cross-section (𝜎c) of Gadolinium (Gd-155 and Gd-157) causes reactivity penalty and swing at the initial stage of fuel burnup in Pressurized Water Reactor (PWR). The present study is concerned with the feasibility of the combination of mixed burnable poison with both low and high 𝜎c as an approach to minimize these effects. Two considered reference designs are fuel assemblies with 24 IBA rods of Gd2O3 and Er2O3 respectively. Models comprise nuclear fuel with a homogeneous mixture of Er2O3, AmO2, SmO2, and HfO2 with Gd2O3 as well as the coating of PaO2 and ZrB2 on the Gd2O3 pellet's outer surface. The infinite multiplication factor was determined and reactivity was calculated considering 3% neutron leakage rate. All models except Er2O3 and SmO2 showed expected results namely higher values of these parameters than the reference design of Gd2O3 at the early burnup period. The highest value was found for the model of PaO2 and Gd2O3 followed by ZrB2 and HfO2. The cycle burnup, discharge burnup, and cycle length for three batch refueling were calculated using Linear Reactivity Model (LRM). The pin power distribution, energy-dependent neutron flux and Fuel Temperature Coefficient (FTC) were also studied. An optimization of model 1 was carried out to investigate effects of different isotopic compositions of Gd2O3 and absorber coating thickness.

Design of Neutron Shielder for Reducing Background of Low Level Gamma Ray Spectrometer (극저준위 감마선 분광시스템의 백그라운드 저감화를 위한 중성자 차폐체 설계)

  • Kim, Tae-Wook;Park, Jong-Mook;Park, Jong-Gil;Shin, Sang-Woon;Jun, Jae-Shik
    • Journal of Radiation Protection and Research
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    • v.26 no.2
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    • pp.67-71
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    • 2001
  • In order to shield the neutrons affecting the background of Low Level Gamma Ray Spectrometer, a neutron shielder was designed. The method used in this study for neutron shielding was the deceleration of fast neutrons by high density polyethylene(HDPE) and the absorption of those slowing-down neutrons by $B_4C$. The calculation results of neutron Interaction in HDPE using Monte Carlo simulation code MCNP4B showed that the thermal-neutron flux was maximum at 10 cm thickness of HDPE. The results also showed that 95% of the thermal neutrons were absorbed by 2 mm thickness of $B_4C$ absorber Consisted of 30 w% $B_4C$ and 70 w% polymer. The results of the Monte Carlo calculation were in good agreement with the experimental value obtained by a neutron shielding apparatus designed for this purpose.

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Neutron Shielding Performance of Mortar Containing Synthetic High Polymers and Boron Carbide (합성 고분자 화합물 및 탄화붕소 혼입에 따른 모르타르의 중성자 차폐성능 분석)

  • Min, Ji-Young;Lee, Bin-Na;Lee, Jong-Suk;Lee, Jang-Hwa
    • Journal of the Korea Concrete Institute
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    • v.28 no.2
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    • pp.197-204
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    • 2016
  • Concrete walls of neutron generating facilities such as fusion reactors and fission reactors become radioactive by neutron irradiation. Both low-activation and neutron shielding are a critical concern at the dismantling stage after the shutdown of facilities with a requirement of radioactive waste management. To tackle this, two types of additives were investigated in fabricating mortar specimens: synthetic high polymers and boron carbide. It is well known that a hydrogen atom is effective in neutron shielding by an elastic scattering because its mass is almost the same as that of the neutron. And boron is an effective neutron absorber with a big neutron absorption cross section. In this study, the effect of the type, shape, and size of polymers were investigated as well as that of boron carbide. Total 16 mix designs were prepared to reveal the effect of polymers on mechanical properties and neutron shielding performance. The neutron does equivalent of polymers-based mortar for fast neutrons decreased by 36 %, and the count rate of boron carbide-based mortar with regard to thermal neutrons decreased by 90 % compared to conventional mortar. These results showed that a combination of polymers and boron carbide compounds has potential to reduce the thickness of neutron shields as well as radioactive waste from reactors.

PARTICLE SIZE-DEPENDENT PULVERIZATION OF B4C AND GENERATION OF B4C/STS NANOPARTICLES USED FOR NEUTRON ABSORBING COMPOSITES

  • Kim, Jaewoo;Jun, Jiheon;Lee, Min-Ku
    • Nuclear Engineering and Technology
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    • v.46 no.5
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    • pp.675-680
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    • 2014
  • Pulverization of two different sized micro-$B_4C$ particles (${\sim}10{\mu}m$ and ${\sim}150{\mu}m$) was investigated using a STS based high energy ball milling system. Shapes, generation of the impurities, and reduction of the particle size dependent on milling time and initial particle size were investigated using various analytic tools including SEM-EDX, XRD, and ICP-MS. Most of impurity was produced during the early stage of milling, and impurity content became independent on the milling time after the saturation. The degree of particle size reduction was also dependent on the initial $B_4C$ size. It was found that the STS nanoparticles produced from milling is strongly bounded with the $B_4C$ particles forming the $B_4C$/STS composite particles that can be used as a neutron absorbing nanocomposite. Based on the morphological evolution of the milled particles, a schematic pulverization model for the $B_4C$ particles was constructed.

SENSITIVITY ANALYSES OF THE USE OF DIFFERENT NEUTRON ABSORBERS ON THE MAIN SAFETY CORE PARAMETERS IN MTR TYPE RESEARCH REACTOR

  • Kamyab, Raheleh
    • Nuclear Engineering and Technology
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    • v.46 no.4
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    • pp.513-520
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    • 2014
  • In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, $B_4C$, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations. The well-known WIMS-D4 and CITATION reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the $B_4C$; also the lowest PPF is gained using the Ag-In-Cd alloy. The maximum point power densities belong to the inside fuel regions surrounding the central flux trap (irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the graphite reflectors. The greatest and least fluctuation of the point power densities are gained by using $B_4C$ and Ag-In-Cd alloy, respectively.