• Nuclear Engineering and Technology
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• 제48권3호
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• pp.624-634
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• 2016

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

• Nuclear Engineering and Technology
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• 제50권5호
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• pp.709-716
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• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.3044-3050
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• 2021

This study offers a Monte Carlo alternative for computing mass energy absorption coefficients of any material through calculation of photon energy deposited per mass of the sample and the energy flux obtained inside a sample volume. This approach is applied in this study to evaluate mass energy absorption coefficients of some amino acids found in human body at twenty-eight different photon energies between 10 keV and 20 MeV. The simulations involved a pencil beam source modeled to emit a parallel beam of mono-energetic photons toward a 1 mean free path thick sample of rectangular parallelepiped geometry. All the components in the problem geometry were surrounded by a 100 cm vacuum sphere to avoid any interactions in materials other than the absorber itself. The results computed using the Monte Carlo radiation transport packages MCNP6.2 and GAMOS5.1 were checked against the theoretical values available from the tables of XMUDAT database. These comparisons indicate very good agreement and support the conclusion that Monte Carlo technique utilized in this fashion may be used as a computational tool for determining the mass energy absorption coefficients of any material whose data are not available in the literature.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3151-3157
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• 2021

Copper is an important structural material in various nuclear energy applications, therefore the correct knowledge of copper cross sections is crucial. The presented paper deals with a validation of different copper transport libraries by means of activation of selected samples. An intense ²⁵²Cf(sf) source with a reference neutron spectrum was used as a neutron source. After irradiation, the samples were measured using a high purity germanium detector and the dosimeter reaction rates were inferred. These experimental data were compared with MCNP6 calculations using CENDL-3.1, JENDL-4.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.2 and JEFF-3.3 evaluated Cu transport libraries. The experiment specifically focuses on ⁵⁸Ni(n,p)⁵⁸Co, ⁹³Nb(n,2n)^92mNb, ¹⁹⁷Au(n,g)¹⁹⁸Au and ⁵⁵Mn(n,g)⁵⁶Mn dosimetry reactions. Evaluated activation cross sections of these dosimetric reactions were taken from the IRDFF-II library. The best library performance depends on the energy region of interest.

• 방사성폐기물학회지
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• 제17권4호
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• pp.375-387
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• 2019

본 연구에서는 최근 개발중인 360 다발 장전용량의 중수로 사용후핵연료 운반용기에 대한 설계기준연료의 방사선원항 평가와 용기외부에서의 방사선량률 계산을 수행하였다. 그리고 국·내외 방사선적 안전성평가와 관련한 기술기준 부합여부를 판단하고 결과의 적합성을 제시하였다. 방사선원항으로 작용하는 설계기준연료 선정을 위해 월성원전에서 운영중인 운반 용기 및 두 가지 방식의 건식저장시설에 적용된 설계기준연료의 사양 및 특성을 조사하였다. 각 운반·저장 시스템 별 설계 기준연료의 연소도, 최소 냉각기간 및 중간저장시설로의 운반시점 등을 바탕으로 연소도 7,800 MWD/MTU와 최소 냉각기간 6년을 설계기준연료로 설정하였다. 설계기준연료의 방사선원항은 SCALE 전산코드의 ORIGEN-ARP모듈을 이용하여 평가하였다. 운반용기의 방사선차폐평가는 MCNP6 전산코드를 이용하였으며, 기술기준에서 요구하는 운반용기 외부에서의 방사선량률 평가를 정상 및 사고조건으로 구분하여 수행하였다. 방사선량률 평가결과, 정상운반조건의 운반용기 표면 및 운반용기 표면 2 m 이격지점에서 계산된 최대 방사선량률은 각각 0.330 mSv·h^-1와 0.065 mSv·h^-1로 도출되어 선량률 제한치인 2.0 mSv·h^-1와 0.1 mSv·h^-1를 모두 만족하는 결과를 도출하였다. 또한 운반사고조건하 운반용기 표면 1 m 지점에서의 최대 방사선량률은 0.321 mSv·h^-1로서 기술기준인 10.0 mSv·h^-1 미만으로 평가되어, 대용량 중수로 사용후핵연료 운반용기는 방사선적 안전성을 확보하는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1060-1067
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• 2018

A feasibility of reusing option of spent nuclear fuel in a fusion-fission hybrid system without partitioning was checked as an alternative option of pyro-processing with critical reactor system. Neutronic study was performed with MCNP 6.1 for this option, direct reuse of spent PWR fuel (DRUP). Various options with DRUP fuel were compared with the reference design concept; transmutation purpose blanket with (U-TRU)Zr fuel loading connected with pyro-processing. Performance parameters to be compared are transmutation performance of transuranic (TRU) nuclides, required fusion power and tritium breeding ratio (TBR). When blanket part is loaded only with DRUP, initial $k_{eff}$ level becomes too low to maintain a practical subcritical system, increasing the required fusion power. In this case, production rate of TRU nuclides exceeds the incineration rate. Design optimization is done for combining DRUP fuel with (U-TRU)Zr fuel. Reactivity swing is reduced to about 2447 pcm through fissile breeding compared to (U-TRU)Zr fuel option. Therefore, a required fusion power is reduced and tritium breeding performance is improved. However, transmutation performance with TRU nuclides especially $^{241}Am$ is degraded because of softening effect of spectrum. It is known that partitioning and transmutation should be accompanied with fusion-fission hybrid system for the effective transmutation of TRU.

• 한국산업융합학회 논문집
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• 제21권6호
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• pp.337-344
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• 2018

Food irradiation is important not only in ensuring safety but also improving antioxidant activity of peaches. Our objective was to establish the best irradiation treatment for peaches by calculating dose distribution using Monte Carlo simulation. 3-D geometry and component densities of peaches, extracted from CT scan, were entered into MCNP to obtain simulated dose distribution. Radiation energies for electron beam were 1.35 MeV (low energy) and 10 MeV (high energy). Co (1.25 MeV) and the Husman irradiator, containing three sealed Cs source rods in an annular array, were used for gamma irradiation. At 1.35 MeV electron beam simulation, electrons penetrated well beyond the peach skin, enough for surface treatment for microorganisms and allergens. At 10 MeV electron beam simulation, for top-beam only treatment, doses at the core were the highest and for double beam treatment, the electron energy was absorbed by the entire sample. At Co source, the radiation doses were presented on the whole area. At Cs source, the dose uniformity ratios were 2.78 for one source and 1.48 for three ones at 120 degrees interval. Proper control of irradiation treatment is critical to establish confidence in the irradiation process.

• Journal of Radiation Protection and Research
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• 제42권1호
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• pp.48-55
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• 2017

Background: Simultaneous detection of neutrons and gamma rays have become much more practicable, by taking advantage of good gamma-ray discrimination properties using pulse shape discrimination (PSD) technique. Recently, we introduced a commercial CLYC system in Korea, and performed an initial characterization and simulation studies for the CLYC detector system to provide references for the future implementation of the dual-mode scintillator system in various studies and applications. Materials and Methods: We evaluated a CLYC detector with 95% $^6Li$ enrichment using various gamma-ray sources and a $^{252}Cf$ neutron source, with validation of our Monte Carlo simulation results via measurement experiments. Absolute full-energy peak efficiency values were calculated for gamma-ray sources and neutron source using MCNP6 and compared with measurement experiments of the calibration sources. In addition, behavioral characteristics of neutrons were validated by comparing simulations and experiments on neutron moderation with various polyethylene (PE) moderator thicknesses. Results and Discussion: Both results showed good agreements in overall characteristics of the gamma and neutron detection efficiencies, with consistent ~20% discrepancy. Furthermore, moderation of neutrons emitted from $^{252}Cf$ showed similarities between the simulation and the experiment, in terms of their relative ratios depending on the thickness of the PE moderator. Conclusion: A CLYC detector system was characterized for its energy resolution and detection efficiency, and Monte Carlo simulations on the detector system was validated experimentally. Validation of the simulation results in overall trend of the CLYC detector behavior will provide the fundamental basis and validity of follow-up Monte Carlo simulation studies for the development of our dual-particle imager using a rotational modulation collimator.

• Journal of Radiation Protection and Research
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• 제49권1호
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• pp.50-64
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• 2024

Background: The reference dose coefficients (DCs) of the International Commission on Radiological Protection (ICRP) have been widely used to estimate organ doses of individuals for risk assessments. This approach has been well accepted because individual anatomy data are usually unavailable, although dosimetric uncertainty exists due to the anatomical difference between the reference phantoms and the individuals. We attempted to quantify the individual variation of organ doses for photon external exposures by calculating and comparing organ DCs for 30 individuals against the ICRP reference DCs. Materials and Methods: We acquired computed tomography images from 30 patients in which eight organs (brain, breasts, liver, lungs, skeleton, skin, stomach, and urinary bladder) were segmented using the ImageJ software to create voxel phantoms. The phantoms were implemented into the Monte Carlo N-Particle 6 (MCNP6) code and then irradiated by broad parallel photon beams (10 keV to 10 MeV) at four directions (antero-posterior, postero-anterior, left-lateral, right-lateral) to calculate organ DCs. Results and Discussion: There was significant variation in organ doses due to the difference in anatomy among the individuals, especially in the kilovoltage region (e.g., <100 keV). For example, the red bone marrow doses at 0.01 MeV varied from 3 to 7 orders of the magnitude depending on the irradiation geometry. In contrast, in the megavoltage region (1-10 MeV), the individual variation of the organ doses was found to be negligibly small (differences <10%). It was also interesting to observe that the organ doses of the ICRP reference phantoms showed good agreement with the mean values of the organ doses among the patients in many cases. Conclusion: The results of this study would be informative to improve insights in individual-specific dosimetry. It should be extended to further studies in terms of many different aspects (e.g., other particles such as neutrons, other exposures such as internal exposures, and a larger number of individuals/patients) in the future.