• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.65-70
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• 1997

Computational benchmark calculations have been performed for CANDU DUPIC fuel lattice and core using a Monte Carlo code MCNP-4B with ENDF/B-V library. The eigenvalues of the DUPIC fuel lattice have been predicted by an integral transport code WIMS-AECL using ENDF/B-V library for different burnup steps and lattice conditions. The comparison has shown that the eigenvalues match those of MCNP-4B within 0.20% $\Delta$k difference between WIMS-AECL and MCNP-4B results. The calculation of a 2-dimensional CANDU core loaded with DUPIC fuel has shown that the eigenvalue predicted by a diffusion code RFSP using lattice parameters generated by WIMS-AECL matches that of MCNP-4B within 0.12%Δk and the largest bundle power prediction error is around 7.2%.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.363.1-363.1
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• 2003

• Journal of Radiation Protection and Research
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• v.32 no.1
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• pp.21-26
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• 2007

This study is performed to evaluate the dose rate and to analyze the dose distribution of the gamma irradiation facility (IR-221) by using a Monte Calro simulation, which is helpful of upgrading the radiation processing qualification. Monte Cairo simulation is performed by MCNP4B code. Dose rates were measured at total 369 points with alanine dosimeters to compare the calculation results and the measurements data. The results have shown that the MCNP4B code is very useful to determine the dose distribution of the IR-221 gamma irradiation facility, as the calculation dose rate is within about ${\pm}5%$ of the measurement data. Dosimetry about the gamma irradiation facility usually needs enormous manpower and time. However Monte Cairo calculation method can reduce the tedious dosimetry jobs and improve the irradiation processing qualification, which will probably contribute to obtain the reliability of the irradiation products.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.191-195
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• 2016

Background: The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Materials and Methods: Validation calculations with recent nuclear data evaluations ENDF/BV-II. 0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. Results and Discussion: It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and ${chi}^2$ values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the $k_{eff}$ values. Conclusion: It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1610-1616
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• 2017

To validate the new Evaluated Nuclear Data File $(ENDF)/B-VIII.0{\beta}4$ library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016). The results obtained with the $ENDF/B-VIII.0{\beta}4$ library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X)) code. All the MCNP(X) calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.

• Korean Journal of Environmental Biology
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• v.24 no.4
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• pp.387-391
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• 2006

Computational and experimental dosimetry of Henschke applicator with respect to high dose rate brachytherapy using the MIRD phantom and a remote control afterloader were performed. A comparison of computational dosimetry was made between the simulated Monte Carlo dosimetry and GAMMADOT brachytherapy Planning system's dosimetry. Dose measurements was performed using ion chamber in a water phantom. Dose rates are calculated using Monte Carlo code MCNP4B and the GAMMADOT. Thecomputational models include the detailed geometry of Ir-192 source, tandem tube, and shielded ovoids for accurate estimation. And transit dose delivered during source extension to and retraction from a given dwell position was estimated by Monte Carlo simulations. Point doses at ICRU bladder/rectal pointswhich have been recommened by ICRU 38 was assessed. Calculated and measured dose distribution data agreed within 4% each other. The shielding effect of ovoids leads to 19% and 20% dose reduction at bladder surface and rectal points.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2008.04a
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• pp.69-75
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• 2008

기포반응도는 월성발전소를 비롯한 CANDU형 원자로의 주된 안전성 쟁점사안으로 끊임없이 논의되어 왔다. 이는 설계기준사고가 노심에서 열에너지 불균형이 원인이 되어 기준이상의 핵연료 파손과 방사성물질 누출로 발전할 위험이 있는 사건들로 정의될 때, 사건 진행 과정에 기포반응도 증가는 조기에 운전중단을 실패할 경우 출력폭주로 이어지므로 사건의 결말이 중대사고로 전환될 위험이 크기 때문이다. 본 연구는 공개된 최신 핵자료인 ENDF/B-VII.0를 NJOY.99로 처리한 연속에너지 반응단면적 라이브러리를 구축하고 MCNP-4C에 접속하여 37봉 천연우라늄 핵연료다발의 표준노심격자에 대한 기포반응도를 시뮬레이션하여, 지금까지 각종문헌에 제시된 값들과 비교, 종합하므로 내제된 불확실성을 추정하는 내용이다. ENDF/B-VII.0 기반 MCNP-4C의 CANDU 노심격자 모델은 동일한 핵자료와 핵종농도를 사용한 WIMS-IAEA 모델과 비교할 때, 초기 노심의 임계도 오차 약 3.51mk가 연소 진행에 따라 $7.5\times10^{-4}mk$/MWD/teU의 비율로 감소하는 것으로 나타났다. 또한 MCNP-4C 예측기포반응도는 초기노심에서 기포율 50% 및 100%에 대해 각각 8.38 및 15.96mk, 평형노심에서 7.68 및 14.72mk로 계산된다. 이는 월성 2, 3, 4 FSAR의 초기노심 및 평형노심에서 100% 기포상태에 대한 값, 약15.0 및 10.6mk와 비교할 때, 초기노심은 약 1.0mk 평형노심은 약4, 1mk 보수적이지만, 다른 연구결과들과는 최대오차 ${\pm}1{\sim}2mk$ 이내에서 잘 일치하는 것으로 평가되었다. 본 연구는 CANDU 노심의 기포반응도 불확실성 요인의 규명 및 영향평가를 위한 노력의 일부로서 앞으로 감속재의 붕산농도 변화, 감속재 및 냉각재의 중수 순도 변화, 기기노화에 의한 격자 구조 및 물성 변화, 중성자속 및 출력 분포 불균형, 반응도조절장치의 위치, 등 주요 설계변수의 변화에 대한 반응도영향 분석연구를 계속할 계획이다.

• 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 the Korean Society of Radiology
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• v.3 no.3
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• pp.23-25
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• 2009

Gd-157 material has very large neutron capture cross section in the thermal region. So it is very useful to shield material for thermal neutrons. Futhermore, in the neutron capture experiment and calculation, the neutron absorption and scattering are very important. Especially these effects are conspicuous in the resonance energy region and below the thermal energy region. In the case of very narrow resonance, the effect of scattering is to be more considerable factor. In the present study, we obtained energy dependent neutron absorption ratios of natural indium in energy region from 0.003 to 100 keV by MCNP-4B Code. The coefficients for neutron absorption was calculated for circular type and 1 mm thickness. In the lower energy region, neutron absorption is larger than higher region, because of large capture cross section (1/v). Furthermore it seems very different neutron absorption in the large resonance energy region. These results are very useful to decide the thickness of sample and shielding materials.

• Proceedings of the Korean Society of Medical Physics Conference
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• 2002.09a
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• pp.485-488
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• 2002

In order to understand and quantitatively analyze the physical phenomena and behavior of each component of mammography system during the breast imaging, we simulated mammography imaging using Monte Carlo simulation codes. MCNP4B code was used for our simulation purpose, and we investigated the effect of target material, anode angle, filtration, peak voltage and exposure on the image quality of mammograms. From the simulation results we expect that optimized operation condition of mammography system can be found.