• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.315-325
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• 1990

The analytical repository source term (REPS) computer code is developed for the safety assessment of radioactive waste geologic repository. For reliable prediction of the leach rates for various radionuclides, degradation of concrete structures, corrosion rate of waste container, degree of corrosion on the container surface, and the characteristics of radionuclides are considered in this REPS code. For the validation of the radionuclide leach rates predicted by the REPS model, the calculated leach rates of Cs-137, Sr-85, and Co-60 are compared with two reported leaching test results. Cesium and strontium leach congruently, and the leaching test results of these species can be reproduced by the congruent leaching model included in the REPS model. In case of cobalt, the solid diffusion model is in good agreement with the leaching test results.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.899-904
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• 1995

원자로 수명기간동안 압력용기의 중성자 조사량 계산은 사용된 핵단면적자료, 모델링시의 기하학적인 단순화 및 가정, 그리고 선원항 선정에 있어서의 가정 등에 의한 불확실성을 포함하고 있다. 이중 핵단면적자료는 이론 및 실험의 발전에 따라 계속 개선되고 있으며 Regulatory Guide(1)에서는 압력용기에서의 중성자 조사량 계산시 가장 최근의 핵자료를 적용할 것을 명시하고 있다. 특히 기존의 ENDF/B-IV나 ENDF/B-V에 포함된 철 핵단면적이 중성자 투과를 작게 평가하고 있음이 밝혀지면서[2] 새로운 핵단면적의 채택이 필요하게되었다. ENDF/B-Vl 핵자료는 개선된 철의 핵단면적을 포함하여 여러 가지 최근의 계산 및 실험치를 바탕으로 생산되었다. 따라서 ENDF/B-Vl를 근거로 하고 있는 BUGLE93[3]을 이용하여 원자로 내부구조물 및 압력용기에서의 고속중성자속 계산을 수행하였다. 그리고 기존의 핵자료를 근거로 예측한 울진 3,4호기 원자로의 수명기간 중 압력용기 중성자 조사량 계산의 타당성을 검토하였다.

• Journal of Radiation Protection and Research
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• v.23 no.4
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• pp.211-218
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• 1998

The distributional effects of radioactive materials on external gamma exposure have been analyzed. An approximate method for estimating external gamma dose given from an arbitrary distribution of radioactive material has been developed. The minimum gamma exposure given from a point source is shown at 0.07 MeV if the source to receptor distance is shorter than 10 m. But if the receptor to point source distance is longer than 20 m, gamma exposure rate increases monotonously according to the average gamma energy. For the analysis of the effects of volume source, we estimated the gamma dose given from different size of hemisphere in which radioactive materials are distributed uniformly. When the radius of hemisphere is longer than 40 m, external gamma dose rate increases monotonously. The gamma dose rate given from the radioactive materials deposited on the ground shows the minimum value at 0.07 MeV in any case. The analysis shows that external gamma exposure is strongly dependent on the distribution of radioactive materials in the environment and gamma energy.

• Proceedings of the Korea Air Pollution Research Association Conference
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• 1999.10a
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• pp.409-410
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• 1999

여천산업단지는 1969년 총면적 583만평 규모로 조성되어 96년 현재 96개 업체가 입주하여 연간 생산액이 13조원에 이르는 대규모 공단으로 성장하였다(한국과학기술원, 1996). 특히, 1996년 대기보전 특별대책지역으로 지정되어 엄격한 배출기준이 적용되는 지역이며, 현재 인근지역인 광양 등을 대기환경 규제지역으로 지정하고자 논의되고 있는 실정이다.(중략)

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.965-970
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• 1995

원자력병원의 MC-50 싸이클로트론을 이용하여 표준선원용으로 사용되는 $^{54}$Mn를 $^{59}$ Co(p, $\alpha$pn) 핵반응으로 생산하고 이온교환수지법을 통하여 무담체의 $^{54}$Mn를 11.85$\mu$Ci/ $\mu$Ah의 수율로 분리하였다. 또한 $^{59}$ Co(p,$\alpha$pn)$^{54}$Mn 핵반응에 대한 여기함수를 stacked foil 방법으로 측정하였고, 그 결과 threshold energy는 27.3 MeV이었으며 41.2MeV에서 최대치의 핵반응단면적 47.4mb를 나타내었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.45-52
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• 1996

고리 1 호기 원자로 감시？슐에서의 고속중성자 플루언스를 계산하기 위하여 MCNP코드로 3차원 모델링을 하였다. 중성자선원은 MCNP에 의해 계산된 핵연료봉출력분포를 사용하였고 원자로심부터 원자로 압력용기 밖까지 1 MeV이상의 중성자를 수송시켰다. 감시？슐은 실제의 같은 크기로 모델링하였고 감시？슐안의 시편은 원자로 압력용기와 같은 재질의 직육면체로 가정하였다. 그리고 MCNP에 의해 감시시편내의 방사화 시료의 핵반응단면적을 계산하였다. 또한 MCNP에 의해 이론적으로 계산된 감시？슐에서 중성자 플루언스와 기존의 감시시험에서 측정된 포화방사능으로 부터 계산된 실험적 감시？슐 중성자 플루언스를 비교하였다. 이론적 ？슐플루언스와 실험적 ？슐플루언스의 비는 대체로 1.0에서 크게 벗어나지 않았으나 감시시험과 시편에 따라 크게 벗어나는 경우도 있었다. MCNP에 의한 유효반응단면적의 계산방법이 기존의 방법보다 모델링 및 계산의 불확실성을 최소화 할 수 있으므로 이번 연구에서 고려하지 못한 원자로심의 연소도를 고려한다면 매우 신뢰성이 높은 결과를 얻을 수 있다.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.63-70
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• 1993

A simplified safety assessment is carried out on rock-cavern type disposal of LLW using the analytical repository source term (REPS) model. For reliable prediction of the leach rates for various radionuclides, degradation of concrete structures, corrosion rate of waste container, degree of corrosion on the container surface, and the characteristics of radionuclides are considered in the REPS model. The results of preliminary assessment show that Cs-137, Ni-63, and Sr-90 are dominant. For the parametric uncertainty and sensitivity analysis, Latin hypercube sampling technique and rank correlation technique are applied. The results of the potential public health impacts show that radiological dose to intruder in the worst case scenario will be negligible and that more attention should be given to near-field performance.

• Analytical Science and Technology
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• v.9 no.3
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• pp.262-269
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• 1996

The muiltiwire proportional counter for the measurement of low-level and environmental $\alpha$ particles emitting nuclides was developed. External dimension of the devloped multiwire proportional counter is $350{\times}290{\times}30mm$ and the sensitivity area is $250{\times}200mm$. The wall material of the detector was selected the stainless steel to prevent the deformation by external impact and to obtain minimum background. The anode and cathode wires were used the stainless steel material of diameter $50{\mu}m$. The spacing of each wires are 10.0mm, 5.0mm and the numbers of total wire are 21, 42 lines, respectively. The multiwire proportional counter was designed that the measurement source is placed within the detector to prevent the wall absorption effect and the efficiency variation by various source heights. The characteristics of the developed detector have been investigated to obtain the plateau, operating voltage, background, counting efficiency, position sensitivity and energy resolution etc. For the $^{241}Am$ nuclide, the calculated LLD(Lower Limit of Detection) is 5.0mBq/L which is lower than 40mBq/L of recommended LLD value by ISO(International Organization for Standardization).

• Journal of Radiation Protection and Research
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• v.8 no.1
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• pp.26-32
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• 1983

A somewhat detailed energy spectra in terms of the track length resulting from coupled electron-photon slowing down are calculated throughout the ICRU standard tissue with uniformly distributed gamma sources of $^{60}Co\;and\;^{137}Cs$, respectively. The calculation was accomplished by utilizing the latest available cross-section data as input to a carefully optimized computer code. In this report, the calculational method is described in detail. Furthermore, results of calculations are given in graphical form. The results show that the energy spectrum defined in terms of differential track length has about same shape although the energies of gamma source are different. The discontinuity in the energy spectrum appears at the energy of $T=(1/T_0+2/m_0c^2)^{-1}$, because a primary photon can not be degrade to a point below this energy.

• Journal of Radiation Protection and Research
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• v.32 no.1
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• pp.9-13
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• 2007

From HPGe calibration spectrum of liquid mixed source in cylindrical vial, we developed simulated spectrum for spectrum analysis education. It is the spectrum that combine peaks separated from measured spectrum. After that, spectrum removed statistical variation of channel counts. Statistical fluctuation of the spectrum is made by Box-Muller function. The spectrum contains 18 peaks. The peak's centroid and area were defined exactly. Developed spectra are calibration spectrum, sample spectrum, background spectrum and spectra for efficiency correction for geometry and cascade coincidence.