• Title, Summary, Keyword: MCNP 4.2 코드

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Monte Carlo Calculation of Thermal Neutron Flux Distribution for (n, v) Reaction in Calandria (몬테칼로 코드를 이용한 중수로 Calandria에서의 $(n,\;{\gamma})$ 반응유발 열중성자속분포 계산)

  • Kim, Soon-Young;Kim, Jong-Kyung;Kim, Kyo-Youn
    • Journal of Radiation Protection and Research
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    • v.19 no.1
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    • pp.13-22
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    • 1994
  • The MCNP 4.2 code was used to calculate the thermal neutron flux distributions for $(n,\;{\gamma})$reaction in mainshell, annular plate, and subshell of the calandria of a CANDU 6 plant during operation. The thermal neutron flux distributions in calandria mainshell, annular plate, and subshell were in the range of $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$ which is somewhat higher than the previous estimates calculated by DOT 4.2 code. As an application to shielding analysis, photon dose rates outside the side and bottom shields were calculated. The resulting dose rates at the reactor accessible areas were below design target, $6 {\mu}Sv/h$. The methodology used in this study to evaluate the thermal neutron flux distribution for $(n,\;{\gamma})reaction$ can be applied to radiation shielding analysis of CANDU 6 type plants.

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A Study on the Comparison of HPGe Detector Response Data for Low Energy Photons Using MCNP, EGS, and ITS Codes (MCNP, EGS, ITS코드를 이용한 고순도 게르마늄 검출기의 저에너지 광자에 대한 반응 비교연구)

  • Kim, Soon-Young;Kim, Jong-Kyung;Kim, Jong-Oh;Kim, Bong-Hwan
    • Journal of Radiation Protection and Research
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    • v.21 no.2
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    • pp.125-129
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    • 1996
  • The energy response of HPGe detector for low energy Photons was determined by using three Monte Carlo codes. MCNP4A. EGS4, and CYLTRAN in ITS3. In this study. bare HPGe detector$(100 mm^2{\times}10mm)$ was used and a pencil beam was incident perpendicularly on the center of the detector surface. The photopeak efficiency, $K_{\alpha}$ and $K_{\beta}$ escape fractions were calculated as a function of incident X-ray energies ranging from 12 to 60 keV in 2-keV increments. Since the Compton. elastic. ana penetration fraction were negligible in this energy range. they were ignored in the calculation. Although MCNP. EGS, and CYLTRAN codes calculated slightly different energy response of HPGe detector for low energy Photons, it appears that the three Monte Carlo codes can Predict the low energy Photon scattering Processes accurately. The MCNP results, which are generally known as to be less accurate at low energy ranges than the EGS and ITS results. are comparable to the results of EGS and ITS and are applicable to the calculation of the low energy response data of a detector.

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MCNP-4A와 CASMO-3를 이용한 CE 16$\times$16 핵연료집합체 임계도 및 봉출력 분포 해석

  • 김교윤;김강석;박찬오
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.79-84
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    • 1995
  • 핵연료집합체 연소도 계산용 전산코드인 CASMO-3를 도입하여 한국고유핵설계체계를 개발하기 위해서는 CE형 핵연료집합체의 핵적특성을 파악하는 것은 필수적이다. 따라서, CASMO-3와 몬테칼로 전산코드인 MCNP-4A를 이용하여 CE형 16$\times$16 핵연료집합체에 대한 $K_{inf}$ 및 봉출력 분포를 비교 분석하였다. $K_{inf}$ 의 경우는 CASMO-3에 의한 계산 결과가 0.5% 이내에서 MCNP-4A의 계산 결과와 일치하였으며, 봉출력분포의 경우도 제어봉 주변이나 Gd$_2$O$_3$ 독봉을 제외하고는 CASMO-3에 의한 계산 결과가 MCNP-4A의 계산 결과와 거의 일치하는 것으로 나타났다.

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Evaluation on the Radiological Shielding Design of a Hot Cell Facility (핫셀시설의 방사선 안전성 평가)

  • 조일제;국동학;구정회;정원명;유길성;이은표;박성원
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.2 no.1
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    • pp.1-11
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    • 2004
  • The hot cell facility for research activities related to the lithium reduction of spent fuel, which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations peformed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}, 2.97{\times}10^{-3} and 1.01{\times}10{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}, 2.99{\times}10^{-3} and 7.88{\times}10^{-2}$ mSv/h, respectively, The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01 mSv/h for the operation area and 0.15 mSv/h for the service (maintenance) area.

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Dose-Rates Evaluation on a Reinforced Hot Cell facility (핫셀시설의 방사선 안전성 평가)

  • 조일제;국동학;구정회;정원명;유길성;이은표;박성원
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • pp.584-589
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    • 2003
  • The hot cell facility which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations performed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}$, $2.97{\times}10^{-2}$ and $1.01{\times}10^{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}$, $2.99{\times}10^{-3}$ and $7.88{\times}10^{-2}$ mSv/h, respectively The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources, and penetration and toboggan will be partly reinforced by lead shield.

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MCNP 선원항 평가법에 의한 SMART 압력용기 중성자 조사량 예비평가

  • 김교윤;김하용;송재승
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.606-611
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    • 1998
  • 330MWt 출력의 신형 원자로인 SMART(System integrated Mod씰w Advanced ReacTor)가 전기 생산뿐만 아니라 해수의 담수화를 위한 에너지 공급을 위해 한국원자력연구소에 의해 개발되고 있다. SMART의 원자로 압력용기에서의 중성자 조사량을 기존의 각분할법 코드 대신에 몬데칼로 수송 코드인 MCNP-4A를 이용하여 평가하였다. MCNP-4A에 의한 몬데 칼로 모사는 각분할법에 비해 핵 단면적 자료, 선원항, 그리고 기하학적 모델링의 문제로부터 야기되는 불확실성을 감소시킬 수 있을 뿐만 아니라 초기 개념 설계 단계에서 상세 노심 출력 분포 자료에 의존하지 않고 선원항을 평가할 수 있는 장점이 있다. 본 연구에서는 원자로 압력 용기 내부의 원자로 노심 및 다른 구조물을 포함하는 전체 원자로 구조에 대하여 몬테 칼로 모사를 적용하였다. 1단계에서는 임계도 계산에 의해 선원항으로 이용되는 원자로 노심내의 열 출력 분포를 평가하고, 2단계에서는 노심내의 열 출력 분포를 고정 선원으로 이용하여 압력 용기에서의 중성자 조사량을평가하였다. 그 결과 SMART 압력용기의 중성자 조사량은 규제 요건을 만족하는 것으로 나타났다.

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Verification of the Radiation Shielding Analysis of Shipping Cask Using Deterministic and Probabilistic Methods (결정론적인 방법과 확률론적인 방법을 이용한 수송용기 방사선차폐해석의 비교 및 검증)

  • Yoon, Jeong-Hyoung;Lee, In-Koo;Bang, Kyoung-Sik;Choi, Byoung-Il;Kim, Chong-Kyoung
    • Journal of Radiation Protection and Research
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    • v.21 no.1
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    • pp.17-25
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    • 1996
  • In this study, to set-up the calculation method of radiation shielding of the KSC-4 shipping cask which is being used for spent fuel transportation, the pre-existing two calculation methods, deterministic and probabilistic methods were tested. For the first, the DOT4.2 computer code adopting the deterministic theory was applied for the calculation of effective neutron shielding under assumption of continuous wall thickness of the cask. To verify the first results, the probabilistic theory was used as an alternate calculation. In this case MCNP4A computer code adopting the probabilitic theory was used. And same approximation was obtained from the two different shielding calculations. From the results, it could be confirmed that the design and calculation method used for the radiation shielding of the KSC-4 was adequate and sufficiently safe to meet the design and QA requirements of 10CFR71 Appendix H.

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Analysis of Gamma Radiation Fields in the MAPLE-X10 Facility Associated with Loss-of-Pool-Water Accident Conditions (LOSS-OF-POOL-WATER 사고시 연구용 원자로 MAPLE-X10 시설에서의 감마 방사선장 해석)

  • Kim, Kyo-Youn;Ha, Chung-Woo;I.C. Gauld
    • 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.

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Assessment of Effective Doses in the Radiation Field of Contaminated Ground Surface by Monte Carlo Simulation (몬테칼로 시뮬레이션에 의한 지표면 오염 방사선장에서의 유효선량 평가)

  • Chang, Jai-Kwon;Lee, Jai-Ki;Chang, Si-Young
    • Journal of Radiation Protection and Research
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    • v.24 no.4
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    • pp.205-213
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    • 1999
  • Effective dose conversion coefficients from unit activity radionuclides contaminated on the ground surface were calculated by using MCNP4A rode and male/female anthropomorphic phantoms. The simulation calculations were made for 19 energy points in the range of 40 keV to 10 MeV. The effective doses E resulting from unit source intensity for different energy were compared to the effective dose equivalent $H_E$ of previous studies. Our E values are lower by 30% at low energy than the $H_E$ values given in the Federal Guidance Report of USEPA. The effective dose response functions derived by polynomial fitting of the energy-effective dose relationship are as follows: $f({\varepsilon})[fSv\;m^2]=\;0.0634\;+\;0.727{\varepsilon}-0.0520{\varepsilon}^2+0.00247{\varepsilon}^3,\;where\;{\varepsilon}$ is the gamma energy in MeV. Using the response function and the radionuclide decay data given in ICRP 38, the effective dose conversion coefficients for unit activity contamination on the ground surface were calculated with addition of the skin dose contribution of beta particles determined by use of the DOSEFACTOR code. The conversion coefficients for 90 important radionuclides were evaluated and tabulated. Comparison with the existing data showed that a significant underestimates could be resulted when the old conversion coefficients were used, especially for the nuclides emitting low energy photons or high energy beta particles.

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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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