• Title, Summary, Keyword: ANISN

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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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An Effect of Energy Group Structure and Weighting Spectrum at the Resonance Energy Region of Iron on Neutron Shielding Calculation (철의 공명에너지 영역의 에너지군구조 및 가중스펙트럼이 중성자 차폐계산에 미치는 영향)

  • Jung-Do Kim;Yukio Ishiguro
    • Nuclear Engineering and Technology
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    • v.17 no.2
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    • pp.129-135
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    • 1985
  • Effects of differences between fine- and broad-group structures and spectrum as a weighting function at the resonance energy region of iron on a neutron shielding calculation were analyzed with the ANISN code and ENDF/B-IV data. The problems analyzed are the broad-group effect, the effect for variation of iron thickness, and the effect of problem-dependent weighting spectrum. In order to verify the group data and method used, a calculational benchmark was performed with the continuous-energy Monte Carlo code VIM. The result was compared with the ANISN calculations using the fine- and broad-group data.

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Nuclear Energy Depositions in the Primary End Shields and Side Primary Shield Systems (월성 2호기 종단 및 측면 차폐체에의 핵에너지축적 해석)

  • Kim, Kyo-Youn;Kim, Jong-Kyung
    • Journal of Radiation Protection and Research
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    • v.17 no.2
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    • pp.37-48
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    • 1992
  • It was carried out to analyze the nuclear energy deposition rates for the bulk shield components including materials of the primary end shield and side primary systems of Wolsong 2 during steady state operations at 100% full power using ANISN code. This paper has been prepared to support system design of Wolsong 2.

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BENCHMARK CALCULATION OF CANDU END SHIELDING SYSTEM

  • Gyuhong Roh;Park, Hangbok
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.618-623
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    • 1998
  • A shielding analysis was performed for the end shield of CANDU 6 reactor. The one-dimensional discrete ordinate code ANISN with a 38-group neutron-gamma library, extracted from DLC-37D library, was used to estimate the dose rate for the natural uranium CANDU reactor. For comparison MCNP-4B calculation was performed for the same system using continuous, discrete and multi-group libraries. The comparison has shown that the total dose rate of the ANISN calculation agrees well with that of the MCNP calculation. However, the individual dose rate (neutron and gamma) has shown opposite trends between AMISN and MCNP estimates, which may require a consistent library generation for both codes.

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Radiation Shielding Analysis of CANDU Spent Fuel Transport Cask (CANDU 사용후핵연료 수송용기 방사선차폐 영향평가)

  • Choi, Jong-Rak;Yoon, Jung-Hyun;Kang, Hee-Young;Lee, Heung-Young;Chung, Sung-Whan
    • Journal of Radiation Protection and Research
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    • v.18 no.2
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    • pp.27-35
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    • 1993
  • A shielding analysis of the shipping cask for transporting the CANDU spent fuel bundles has been studied. Radiation source term has been calculated on spent fuel with burn-up of 7,800 MWD/MTU and 5 years cooling time by ORIGEN2 code. The shielding calculation for the cask capable of transporting 378 bundles of CANDU spent fuel has been made by use of 1-D ANISN and 2-D DOT 4.2 codes. As a result of analysis, the optimum shield thickness of cask was obtained. And it is proved that the safety in radiation shielding under normal transport and hypothetical accident conditions is confirmed to satisfy the allowable values specified in IAEA Safety Series No. 6 and the Korean Atomic Law.

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Preliminary Estimation of Activation Products Inventory in Reactor Components for Kori unit 1 decommissioning (고리1호기 해체시의 원자로 구조물에서의 방사회 생성물 재고량 예비평가)

  • Lee, Kyung-Jin;Kim, Hak-Soo;Sin, Sang-Woon;Song, Myung-Jae;Lee, Youn-Keun
    • Journal of Radiation Protection and Research
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    • v.28 no.2
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    • pp.109-116
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    • 2003
  • Based on the necessity to evaluate the activation products inventory during decommissioning lot domestic nuclear power plants, a preliminary estimation of the activation products inventory for Kori unit 1, which is getting close to the end of lifetime, was carried out with ANISN and ORIGEN2 code. In order to calculate neutron nux using ANISN code, the reactor was divided into 9 zones from core to bioshield concrete for radial direction. Also :he cross-section of main nuclides were calibrated with neutron flux in the reactor pressure vessel(RPV) region. The results showed that 95 % of tile total radioactivity in RPV from reactor shutdown to 10 years came from the nuclides of $^{55}Fe,\;^{59}Ni,\;^{63}Ni\;and\;^{60}Co$. And the total radioactivity with cooling of more than 50 years after decommissioning was no more than 0.2 % of at the time of shutdown. Considering the weight of RPV is 210 tons, the total radioactivity of RPV reached to $5.25{\times}10^{6}GBq$ at shutdown time. As compared with the total radioactivity of bioshield concrete at reactor shutdown time, the radioactivity after tooling more than 10 years was below 1 %.

Effect of Prompt Fission Neutron Spectral Formulae on Nuclear Criticality (핵분열(核分裂) 중성자(中性子)스펙트럼이 핵임계도(核臨界度)에 미치는 효과(效果))

  • Ro, Seung-Gy;Min, Duck-Kee;Youk, Geun-Uck;Oh, Hi-Peel
    • Journal of Radiation Protection and Research
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    • v.7 no.1
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    • pp.56-60
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    • 1982
  • A calculation of the effective multiplication factor has been made for GODIVA and JEZEBEL critical assemblies by using a computer code, ANISN, with having the Watt's, Cranberg's and Maxwellian formulae for the prompt fission neutron spectrum as a fission source. Then the calculated values have been compared with experimental data obtained by others. The Maxwellian formula seems to be the best one for representing the prompt fission neutron spectrum since the effective multiplication factor based on it shows a better agreement with the experimental value compared to the rest formulae.

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Neutron Reflecting Effects by Water and Concrete (물과 콘크리트에 의한 중성자(中性子)의 반사효과(反射效果))

  • Min, Duck-Kee;Ro, Seung-Gy
    • Journal of Radiation Protection and Research
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    • v.8 no.1
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    • pp.33-37
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    • 1983
  • Neutron reflecting effects in terms of effective multiplication factor have been calculated with varying water or concrete thickness, and gap distance between concrete reflector and a fissile solution system. A numerical calculation of effective multiplication factors has been carried out by using the discrete ordinates method with the help of the computer code, ANISN. It is revealed that the reflecting .effect by thin concrete is lower than that of the identical thickness of water while the effect by thick water is low compared to that of the identical thickness of concrete. It seems that the effective multiplication factors are first decreasing rapidly with gap distance, which is filled with water, between concrete reflector and the fissile solution system, and then decrease slowly over the distance of about 15cm.

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Radiation Streaming in KNU-1 Reactor Cavity (고리 1호기 원자로 공동에서의 방사선 흐름 현상 해석)

  • Kun-Woo Cho;Chang-Soon Kang
    • Nuclear Engineering and Technology
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    • v.18 no.1
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    • pp.27-37
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    • 1986
  • The neutron fluxes and dose rates due to radiation streaming from reactor cavities were evaluated at the KNU-1 reactor pressure vessel (RPY) head flange elevation. To find a suitable cross section data set for the evaluation, a benchmark test was performed for three data sets; DLC-23/CASK, DLC-31/FEWG, and DLC-47/BUGLE. The leakage fluxes from the KNU-1 RPV outer surface were calculated with two different methods: 1-D calculation with ANISN, and 2-D calculation with DOT3.5. The Monte Carlo procedures as embodied in the MORSE-CG code combined with the albedo option were applied to predict the radiation distributions in the cavity region. Finally, the activation analysis of the stud bolts was performed to identify the major activation products.

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Neutron Streaming Analysis in 1300 MWe Pressurized Water Reactor Cavity (1,300 MWe 가압경수로 공동내에서의 중성자 흐름해석)

  • Kwon, Seog-Guen;Kim, Kyung-Eung
    • Journal of Radiation Protection and Research
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    • v.10 no.1
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    • pp.41-49
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    • 1985
  • Neutron Streaming analysis in 1300 MWe pressurized water reactor cavity was performed. In this calculation, the discrete ordinates transport codes, ANISN and DOT 3.5, and the Monte Carlo code, TRIPOLI-02 were used with the coupling code, DOTTRI. In this study IBM 3033 type computer was used. The calculated neutron fluxes and dose rates were compared with the measured data in a 900MWe pressurized water reactor cavity to show a good agreement, although some deviations in the results for each energy group were noticed. These results will be applied in the radiation shielding design of high capacity nuclear power reactors and, to the means of radiation protection in case of the reactor maintenance and the access of the reactor cavity.

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