• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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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.

• 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.

• 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.

• 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 %.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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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.

• 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.

• 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.

• 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.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.319-328
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• 1992

The Galerkin formulation of the finite element method is applied to the integral law of the first-order form of the one-group neutron transport equation in one-dimensional spherical geometry. Piecewise linear or quadratic Lagrange polynomials are utilized in the integral law for the angular flux to establish a set of linear algebraic equations. Numerical analyses are performed for the scalar flux distribution in a heterogeneous sphere as well as for the criticality problem in a uniform sphere. For the criticality problems in the uniform sphere, the results of the finite element method, with the use of continuous finite elements in space and angle, are compared with the exact solutions. In the heterogeneous problem, the scalar flux distribution obtained by using discontinuous angular and spatical finite elements is in good agreement with that from the ANISN code calculation.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.65-70
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• 1988

A Shielding analysis of the shipping cask designed conceptually, of which shielding material are lead and resin, for containing 4 PWR spent fuel assemblies, has been made with the help of a computer code, ANISN. The shielding materials being used in the cask have been selected and arranged to minimize cask weight while maintaining an overall shielding effectiveness. Radiation source terms have been calculated by means of ORIGIN-2 code under the assumptions of 38,000 MWD/MTU burnup and 3-year cooling time. A calculation of gamma-ray and neutron dose rates on the cask surface and 1m from the surface has been done. It is revealed that the total dose rates under the normal transport and hypothetical accident conditions meet the standards specified.