• Title, Summary, Keyword: MCNP code simulation

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Application of a new neutronics/thermal-hydraulics coupled code for steady state analysis of light water reactors

  • Safavi, Amir;Esteki, Mohammad Hossein;Mirvakili, Seyed Mohammad;Arani, Mehdi Khaki
    • Nuclear Engineering and Technology
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    • v.52 no.8
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    • pp.1603-1610
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    • 2020
  • Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.

Fundamental approach to development of plastic scintillator system for in situ groundwater beta monitoring

  • Lee, UkJae;Choi, Woo Nyun;Bae, Jun Woo;Kim, Hee Reyoung
    • Nuclear Engineering and Technology
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    • v.51 no.7
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    • pp.1828-1834
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    • 2019
  • The performance of a plastic scintillator for use in an in situ measurement system was analyzed using simulation and experimental methods. The experimental results of four major pure beta-emitting radionuclides, namely $^3H$, $^{14}C$, $^{32}P$, and $^{90}Sr/^{90}Y$, were compared with those obtained using a Monte Carlo N-particle (MCNP) code simulation. The MCNP simulation and experimental results demonstrated good agreement for $^{32}P$ and $^{90}Sr/^{90}Y$, with a relative difference of 1.95% and 0.43% between experimental and simulation efficiencies for $^{32}P$ and $^{90}Sr/^{90}Y$, respectively. However, owing to the short range of beta particles in water, the efficiency for $^{14}C$ was extremely low, and $^3H$ could not be detected. To directly measure the low-energy beta radionuclides considering their short range, a system where the source could flow directly to the scintillator was developed. The optimal thickness of the plastic scintillator was determined based on the suggested diameter. Results showed that the detection efficiency decreases with an increase in the depth of the water. The detection efficiency decreased drastically to approximately 10 cm, and the tendency was gradually constant.

Dose Determination in the IR-221 Gamma Facility Using a Monte Carlo Simulation (몬테칼로 시뮬레이션을 이용한 IR-221의 선량 평가)

  • Lim, Ik-Sung;Kim, Ki-Yup;Roh, Gyu-Hong;Lee, Chung
    • Journal of Radiation Protection and Research
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    • v.32 no.1
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    • pp.21-26
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    • 2007
  • This study is performed to evaluate the dose rate and to analyze the dose distribution of the gamma irradiation facility (IR-221) by using a Monte Calro simulation, which is helpful of upgrading the radiation processing qualification. Monte Cairo simulation is performed by MCNP4B code. Dose rates were measured at total 369 points with alanine dosimeters to compare the calculation results and the measurements data. The results have shown that the MCNP4B code is very useful to determine the dose distribution of the IR-221 gamma irradiation facility, as the calculation dose rate is within about ${\pm}5%$ of the measurement data. Dosimetry about the gamma irradiation facility usually needs enormous manpower and time. However Monte Cairo calculation method can reduce the tedious dosimetry jobs and improve the irradiation processing qualification, which will probably contribute to obtain the reliability of the irradiation products.

The Performance Test of Anti-scattering X-ray Grid with Inclined Shielding Material by MCNP Code Simulation

  • Bae, Jun Woo;Kim, Hee Reyoung
    • Journal of Radiation Protection and Research
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    • v.41 no.2
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    • pp.111-115
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    • 2016
  • Background: The scattered photons cause reduction of the contrast of radiographic image and it results in the degradation of the quality of the image. In order to acquire better quality image, an anti-scattering x-ray gird should be equipped in radiography system. Materials and Methods: The X-ray anti-scattering grid of the inclined type based on the hybrid concept for that of parallel and focused type was tested by MCNP code. The MCNPX 2.7.0 was used for the simulation based test. The geometry for the test was based on the IEC 60627 which was an international standard for diagnostic X-ray imaging equipment-Characteristics of general purpose and mammographic anti-scatter grids. Results and Discussion: The performance of grids with four inclined shielding material types was compared with that of the parallel type. The grid with completely tapered type the best performance where there were little performance difference according to the degree of inclination. Conclusion: It was shown that the grid of inclined type had better performance than that of parallel one.

Monte Carlo Studies on Mammography System

  • Ho, Dong-Su;Lee, Hyoung-Koo;Suh, Tae-Suk;Choe, Bo-Young;Kim, Song-Hyun;Kim, Do-Il
    • Proceedings of the Korean Society of Medical Physics Conference
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    • pp.485-488
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    • 2002
  • In order to understand and quantitatively analyze the physical phenomena and behavior of each component of mammography system during the breast imaging, we simulated mammography imaging using Monte Carlo simulation codes. MCNP4B code was used for our simulation purpose, and we investigated the effect of target material, anode angle, filtration, peak voltage and exposure on the image quality of mammograms. From the simulation results we expect that optimized operation condition of mammography system can be found.

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A study on slim-hole neutron logging based on numerical simulation (소구경 시추공에서의 중성자검층 수치모델링 연구)

  • Ku, Bonjin;Nam, Myung Jin
    • Geophysics and Geophysical Exploration
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    • v.15 no.4
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    • pp.219-226
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    • 2012
  • This study provides an analysis on results of neutron logging for various borehole environments through numerical simulation based on a Monte Carlo N-Particle (MCNP) code developed and maintained by Los Alamos National Laboratory. MCNP is suitable for the simulation of neutron logging since the algorithm can simulate transport of nuclear particles in three-dimensional geometry. Rather than simulating a specific tool of a particular service company between many commercial neutron tools, we have constructed a generic thermal neutron tool characterizing commercial tools. This study makes calibration chart of the neutron logging tool for materials (e.g., limestone, sandstone and dolomite) with various porosities. Further, we provides correction charts for the generic neutron logging tool to analyze responses of the tool under various borehole conditions by considering brine-filled borehole fluid and void water, and presence of borehole fluid.

Verification of MCNP/ORIGEN-2 Model and Preliminary Radiation Source Term Evaluation of Wolsung Unit 1 (월성 1호기 MCNP/ORIGEN-2 모델 검증 및 예비 선원항 계산)

  • Noh, Kyoungho;Hah, Chang Joo
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.13 no.1
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    • pp.21-34
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    • 2015
  • Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.

Evaluation of Neutron Shielding Performance of Polyethylene Coated Boron Carbide-Incorporated Cement Paste using MCNP Simulation (MCNP 시뮬레이션을 통한 폴리에틸렌 코팅 탄화붕소 혼입 시멘트 페이스트의 중성자 차폐 성능 평가)

  • Park, Jae-Yeon;Jee, Hyeon-Seok;Bae, Sung-Chul
    • Proceedings of the Korean Institute of Building Construction Conference
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    • pp.114-115
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    • 2018
  • To develop an effective shielding material for spent fuel that emits fast neutrons is necessary. In this study, thermal neutron and fast neutron shielding performance of polyethylene coated boron carbide-incorporated cement paste was quantitatively analyzed by Monte Carlo N-Particle transport code (MCNP) simulations. As the results of the simulations, fast neutrons were effectively shielded through large quantity of hydrogen and boron elements in polyethylene and boron carbide.

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Development of an MCNP-Based Cone-Beam CT Simulator (MCNP 기반의 CBCT 전산모사 시스템 개발)

  • Lim, Chang-Hwy;Cho, Min-Kook;Han, Jong-Chul;Youn, Han-Bean;Yun, Seung-Man;Cheong, Min-Ho;Kim, Ho-Kyung
    • Journal of the Korean Society for Nondestructive Testing
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    • v.29 no.4
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    • pp.351-359
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    • 2009
  • We have developed a computer simulator fur cone-beam computed tomography (CBCT) based on the commercial Monte Carlo code, MCNP. All the functions to generate input files, run MCNP, convert output files to image data, reconstruct tomographs were realized in graphical user-interface form. The performance of the simulator was demonstrated by comparing with the experimental data. Although some discrepancies were observed due to the ignorance of the detailed physics in the simulation, such as scattered X-rays and noise in image sensors, the overall tendency was well agreed between the measured and simulated data. The developed simulator will be very useful for understanding the operation and the better design of CT systems.