• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.197-201
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• 2012

Annual dose at the boundary of the interim storage facility at normal condition was calculated to estimate the site area of the facility of PWR spent nuclear fuel. In this work, source term was generated by ORIGEN-ARP for 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facilities and radiation shielding evaluations were conducted by MCNP code depending on the storage capacity. In the case of the centralized storage system, the required site area was found to have the radius of more than 700 m.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.207-214
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• 2001

The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.

• Nuclear Engineering and Technology
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• v.21 no.2
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• pp.63-72
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• 1989

An analysis for the gamma radiation fields in the research reactor MAPLE-X10 facility has been peformed under the assumption of partial loss of reactor and service pool water to assess the safety from the view point of design. Four photon source terms considered in the analysis were calculated using the ORIGEN-S code. Gamma dose rate calculations over the reactor and service pools during the water-loss accident conditions were performed using QAD-CG code. MCNP code (Monte Carlo Neuron and Photon Transport code), also, was used to assess the scattered radiation fields away from the pools, which is appropriate for calculating the scattered photon dose rates outside of the solid angle subtended by the source and pool walls.

• Radiation Oncology Journal
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• v.19 no.1
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• pp.66-73
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• 2001

Purpose : For the research of Boron Neutron Capture Therapy (BNCT), fast neutrons generated from the MC-50 cyclotron with maximum energy of 34.4 MeV in Korea Cancer Center Hospital were moderated by 70 cm paraffin and then the dose characteristics were investigated. Using these results, we hope to establish the protocol about dose measurement of epi-thermal neutron, to make a basis of dose characteristic of epi-thermal neutron emitted from nuclear reactor, and to find feasibility about accelerator-based BNCT. Method and Materials : For measuring the absorbed dose and dose distribution of fast neutron beams, we used Unidos 10005 (PTW, Germany) electrometer and IC-17 (Far West, USA), IC-18, ElC-1 ion chambers manufactured by A-150 plastic and used IC-l7M ion chamber manufactured by magnesium for gamma dose. There chambers were flushed with tissue equivalent gas and argon gas and then the flow rate was S co per minute. Using Monte Carlo N-Particle (MCNP) code, transport program in mixed field with neutron, photon, electron, two dimensional dose and energy fluence distribution was calculated and there results were compared with measured results. Results : The absorbed dose of fast neutron beams was $6.47\times10^{-3}$ cGy per 1 MU at the 4 cm depth of the water phantom, which is assumed to be effective depth for BNCT. The magnitude of gamma contamination intermingled with fast neutron beams was $65.2{\pm}0.9\%$ at the same depth. In the dose distribution according to the depth of water, the neutron dose decreased linearly and the gamma dose decreased exponentially as the depth was deepened. The factor expressed energy level, $D_{20}/D_{10}$, of the total dose was 0.718. Conclusion : Through the direct measurement using the two ion chambers, which is made different wall materials, and computer calculation of isodose distribution using MCNP simulation method, we have found the dose characteristics of low fluence fast neutron beams. If the power supply and the target material, which generate high voltage and current, will be developed and gamma contamination was reduced by lead or bismuth, we think, it may be possible to accelerator-based BNCT.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1563-1570
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• 2017

In order to perform dry interim storage and transportation of the spent-fuel assemblies of the Bushehr Nuclear Power Plant, dual-purpose casks can be utilized. The effectiveness of different neutron-shield materials for the dual-purpose cask was analyzed through a set of calculations carried out using the Monte Carlo N-Particle (MCNP) code. The dose rate for the dual-purpose cask utilizing the recently developed materials of $epoxy/clay/B_4C$ and $epoxy/clay/B_4C/carbon$ fiber was less than the allowable radiation level of 2 mSv/h at any point and 0.1 mSv/h at 2 m from the external surface of the cask. By utilization of $epoxy/clay/B_4C$ instead of an ethylene glycol/water mixture, the dose rates on the side surface of the cask due to neutron sources and consequent secondary gamma rays will be reduced by 17.5% and 10%, respectively. The overall dose rate in this case will be reduced by 11%.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.1-5
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• 2017

The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor ($K_{eff}$) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of $D_2O$, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.79-86
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• 2015

$^3He$ gas has been used for neutron monitors as the neutron converter owing to its advantages such as high sensitivity, good ${\gamma}$-discrimination capability, and long-term stability. However, $^3He$ is becoming more difficult to obtain in last few years due to a global shortage of $^3He$ gas. Accordingly, the cost of a neutron monitor using $^3He$ gas as a neutron converter is becoming more expensive. Demand on a neutron monitor using an alternative neutron conversion material is widely increased. $^{10}B$ has many advantages among various $^3He$ alternative materials, as a neutron converter. In order to develop a neutron converter using $^{10}B$ (actually $B_4C$), we calculated the optimal thickness of a neutron converter with a Monte Carlo simulation using MCNP6. In addition, a neutron converter was fabricated by the Ar sputtering method and the neutron signal detection efficiencies were measured with respect to various thicknesses of fabricated a neutron converter. Also, we developed a 2-dimensional multi-wire proportional chamber (MWPC) for neutron beam profile monitoring using the fabricated a neutron converter, and performed experiments for neutron response of the neutron monitor at the 30 MW research reactor HANARO at the Korea Atomic Energy Research Institute. The 2-dimensional MWPC with boron ($B_4C$) neutron converter was proved to be useful for neutron beam monitoring, and can be applied to other types of neutron imaging.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.174-174
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• 2004

It is necessary to perform the radionuclide inventory assessment for the disposal of low and intermediate radioactive waste containers. The $\gamma$ nuclide analyzer can be used for the assessment of containers. However, if the radioactivity in the containers is extremely low or high, radionuclide inventory of the containers can not be evaluated properly. Also, gamma scanning method is time consuming and has economical burden to the utilities.(omitted)

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.149-154
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• 2001

Calculational models to evaluate radiation doses resulting from the medical use of high energy beta-ray sources are presented. The radioactive sources considered are Sr-90/Y-90 used as ophthalmic applicator, Re-188 used for treating restenosis of coronary artery, and Ho-166 used for treating hepatic tumors. Typical therapeutic situations which might induce relatively high radiation doses the medical person involved were considered to compute by using MCNP-4C Monte Carlo code the radiation doses. The calculation results suggest that for all of the cases considered, the evaluated radiation doses are negligible compared to the dose limits. It is also found that the effect of Bremsstrahlung radiations on the total dose is insignificant, and hence the conventional lead gown is also effective in shielding beta-rays.