• Title, Summary, Keyword: MCNP코드

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ANISN-MCNP 코드를 이용한 월성2호기 반응도제어기구 방사선흐름해석

  • 김용일;진영권;김교윤
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.269-274
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    • 1996
  • 월성원자력발전소 2호기와 같은 CANDU 6형 원자로의 반응도제어기구 설치대에는 여러 반응도제어기구가 삽입되기때문에 원자로심으로부터의 방사선흐름현상으로 인한 방사선피폭이 예상될 수 있는 위치이다. 좁고 긴 반응도제어기구 도관에서의 방사선 흐름으로 인한 반응도제어기구 설치대에서의 방사선량을 예측하기 위해 몬테 칼로 MCNP 코드를 1차원 각분할법 코드인 ANISN과 연계하여 사용하였다. 월성원자력2호기의 상단차폐해석을 위한 ANISN 계산, 도관의 방사선흐름을 평가하기 위한 MCNP 계산, 그리고 반응도제어기구 설치대에서의 방사선량율 평가를 위한 MCNP 계산등 3단계 계산 기법의 적응이 시도되었다.

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MCNP코드를 이용한 수분 측정계기의 기하학적 배치

  • 최원철;이석근;황주호;전홍배;양세학;권정광
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.139-146
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    • 1996
  • 도로건설시 다짐조절은 안정성과 내구성 향상에 중요한 의미를 가지며 이러한 다짐조절에 있어서 수분함량의 측정은 매우 중요하다. 이전에는 흙의 수분함량을 측정하기 위한 계기를 설계하기 위하여 주로 실험에 의한 방법을 사용하였으나 본 연구에서는 3차원 모델링이 가능한 MCNP코드$^{(1)}$ 를 이용하여 계측기 설계에 있어서 중요한 설계변수인 방사선원의 위치와 측정계기 사이의 거리 그리고 계기구성요소인 검출기의 위치, 개수, 흡수재, 감속재의 기하학적 구조 등을 계산하여 설정하였다.

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Neutron Shielding Analysis for a Spent Fuel Container Using XSDRN, ONEDANT and MCNP Codes (XSDRN, ONEDANT및 MCNP에 의한 사용후 핵연료 용기의 중성자 차폐 해석)

  • Kim, Kyo-Youn;Lee, Tae-Young;Ha, Chung-Woo;Kim, Jong-Kyung
    • Journal of Radiation Protection and Research
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    • v.14 no.1
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    • pp.46-55
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    • 1989
  • Neutron shielding for a spent fuel container was analized using the Monte Carlo code MCNP coupled with discrete ordinates codes, XSDRN and ONEDANT. The ORIGEN-S code was used to determine the fixed neutron source, and the spectral source information for MCNP were obtained from a 10 group XSDRN calculation and a 27 group ONEDANT calculation. When the depleted uranium shield was used, the results from ONEDANT and XSDRN calculations agreed with the MCNP results within 10% and 20%, respectively. Depleted uranium appears more effective than lead or steel as a neutron shield.

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A Study on Calibration of Neutron Moisture Gauge Using MCNP4A (MCNP4A 전산코드를 이용한 중성자 수분함량 측정기의 교정식 및 교정상수 도출방법 연구)

  • Whang, Joo-Ho;Lim, Chun-Il;Song, Jung-Ho
    • Journal of Radiation Protection and Research
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    • v.22 no.4
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    • pp.289-298
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    • 1997
  • Time-consuming experiments have been required in the development of neutron moisture gauge to induce a relation between the water content in soil and the neutron counts. Applying a monte carlo computer code to simulate the experiments of neutron moisture gauging may contribute to reduce time and efforts for experiments and produce a calibration equation which is more applicable to soil in general. In this study MCNP4A, a monte carlo computer code, was employed to simulate soil experiments and the simulated results were compared with experimental ones. The comparative study showed that MCNP4A is applicable to simulate the experiments and calibration equation can be obtained through simulations. Effects of dry density changes were also studied.

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Verification of the PMCEPT Monte Carlo dose Calculation Code for Simulations in Medical Physics (의학물리 분야에 사용하기 위한 PMCEPT 몬테카를로 도즈계산용 코드 검증)

  • Kum, O-Yeon
    • Progress in Medical Physics
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    • v.19 no.1
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    • pp.21-34
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    • 2008
  • The parallel Monte Carlo electron and photon transport (PMCEPT) code [Kum and Lee, J. Korean Phys. Soc. 47, 716 (2006)] for calculating electron and photon beam doses has been developed based on the three dimensional geometry defined by computed tomography (CT) images and implemented on the Beowulf PC cluster. Understanding the limitations of Monte Carlo codes is useful in order to avoid systematic errors in simulations and to suggest further improvement of the codes. We evaluated the PMCEPT code by comparing its normalized depth doses for electron and photon beams with those of MCNP5, EGS4, DPM, and GEANT4 codes, and with measurements. The PMCEPT results agreed well with others in homogeneous and heterogeneous media within an error of $1{\sim}3%$ of the dose maximum. The computing time benchmark has also been performed for two cases, showing that the PMCEPT code was approximately twenty times faster than the MCNP5 for 20-MeV electron beams irradiated on the water phantom. For the 18-MV photon beams irradiated on the water phantom, the PMCEPT was three times faster than the GEANT4. Thus, the results suggest that the PMCEPT code is indeed appropriate for both fast and accurate simulations.

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MCNP CODE를 이용한 아스팔트함량 측정장비의 설계 및 검증

  • 임천일;황주호
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.735-740
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    • 1998
  • 방사성동위원소를 이용한 아스팔트함량 측정장비의 실험적인 방법에 의한 설계는 많은 시간과 비용이 소요되므로, 코드모사를 통해 설계할 경우 이러한 노력을 줄일 수 있다. 본 연구에서는 장비의 활용성을 증대시키기 위해 법적 규제 면제치인 100 $\mu$Ci이하의 방사성동위원소를 이용하며, 6%의 아스팔트함량을 갖는 혼합물을 5분간 측정하였을 경우 0.2%이내의 함량측정오차를 갖는 장비를 MCNP 코드를 이용하여 설계하였다 또한 코드 모사를 통한 설계를 바탕으로 장비를 제작한 후 5개의 시료에 대한 함량을 측정하고 그 결과를 비교하여 코드의 적용가능성을 검증하였다 실험결과 6.03% 아스팔트 함량을 가진 시료를 5분간 측정하여 5.85%의 함량을 얻을 수 있었다.

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Conversion Factors for Calibration of Personnel Dosimeters (개인선량계 교정을 위한 환산인자 계산)

  • Lee, Won-Koo;Lee, Tae-Young;Ha, Chung-Woo
    • Journal of Radiation Protection and Research
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    • v.16 no.1
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    • pp.25-32
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    • 1991
  • MCNP code was used to calculate conversion factor H(d)ma at the depths of 0.07 and 10mm within a water phantom recommended by IAEA and within a PMMA phantom required by the US dosimeter proficiency testing programmes. The calculations were performed for an expanded parrallel beam of monoenergetic photons of perpendicular incidence on one faces of the phantom. The results can be used as conversion factor in calibrating individual dosemeters in terms of the dose equivalent quantities defined directly in the phantom.

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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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