• 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.

• Journal of the Korean Society of Radiology
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• v.15 no.4
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• pp.415-427
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• 2021

Korea Institute of Radiological and Medical Sciences has provided various beam irradiation services to researchers using a 50 MeV cyclotron beam line. In particular, since the neutron beam service uses the nuclear reaction between protons and beryllium, the possibility of activation of the irradiated sample increases by using a high current. In this study, MCNP 6.2 and FISPACT-II 4.0 were used to evaluate the possible activation during the 35 MeV 20 ㎂ neutron beam service, which is preferred by the researchers. As a result of the calculation, if the iron, copper, and tungsten samples were irradiated for more than 1 hour, long-lived radioisotopes were produced and their radioactivity exceeded the standard level for self-disposal. Under the conditions of 2 hours of daily irradiation, no activation occurred in the building materials, and the internal exposure of workers due to air activation inside the irradiation room was very insignificant. And when this air was discharged to environment, the radioactivity including this air was also satisfied the emission standard.

• Journal of Radiation Protection and Research
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• v.22 no.1
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• pp.47-58
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• 1997

The conversion coefficients of effective dose per unit air kerma and equivalent dose per unit fluence were calculated by MCNP4A code for antero-posterior(AP) and postero- anterior(PA) incidence of broad, unidirectional beams of photons into anthropomorphic MIRD phantom. Calculations have been performed for 20 monoenergetic photons of energy ranging from 0.03 to 10 MeV. The conversion coefficients showed a good agreement with the corresponding values given in the draft publication of joint task group of ICRP and ICRU within 10%. The deviations may arise from the differences of geometry in the MIRD phantom and the ADAM/EVE phantoms, and the differences in the codes and cross-section data used. Inclusion of a specific oesophagus model results in effective dose slightly different(5% at most) from the effective doses obtained by adopting the equivalent doses for the thymus or pancreas. Deletion of the ULI from the remainder organ appeared not to be significant for the cases of photon dosimetry covered in this study.

• Nuclear Engineering and Technology
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• v.31 no.5
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• pp.512-521
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• 1999

A feasibility study was performed to design an epithermal neutron beam for BNCT using the neutron of 2.45 MeV on the average produced from $^2H(d,n)^3$He reaction induced by plasma focus in the z-pinch instead of the conventional accelerator-based $^3H(d, n)^4$He neutron generator. Flux and spectrum were analyzed to use these neutrons as the neutron source for BNCT. Neutronic characteristics of several candidate materials in this neutron source were investigated Using MCNP Code, and $^7LiF$ ; 40%Al + 60%$AIF_3$, and Pb Were determined as moderator, filter, and reflector in an epithermal neutron beam design for BNCT, respectively. The skin-skull-brain ellipsoidal phantom, which consists of homogeneous regions of skin-, bone-, or brain-equivalent material, was used in order to assess the dosimetric effect in brain. An epithermal neutron beam design for BNCT was proposed by the repeated work with MCNP runs, and the dosimetric properties (AD, AR, ADDR, and Dose Components) calculated within the phantom showed that the neutron beam designed in this work is effective in tumor therapy. If the neutron source flux is high enough using the z-pinch plasma, BNCT using the neutron source produced from $^2H(d,n)^3$He reaction will be very feasible.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.110-110
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• 1998

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1505-1512
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• 2017

Because accelerator-based boron neutron capture therapy (BNCT) systems are planned for use in hospitals, entry into the medical room should be controlled as hospitals are generally assumed to be public and safe places. In this paper, computational investigation of the medical staff effective dose during BNCT has been performed in different situations using Monte Carlo N-Particle (MCNP4C) code and two voxel based male phantoms. The results show that the medical staff effective dose is highly dependent on the position of the medical staff. The results also show that the maximum medical staff effective dose in an emergency situation in the presence of a patient is ${\sim}25.5{\mu}Sv/s$.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4019-4025
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• 2023

Recently, a new monkey computational phantom, called Visible Monkey, was developed for non-ionizing radiation studies in animal research. In this study, we extended its applications to ionizing radiation studies by implementing the voxel model of the Visible Monkey into three general-purpose Monte Carlo (MC) codes: MCNP6, PHITS, and Geant4. The implementation work for MCNP and PHITS was conducted using the LATTICE, UNIVERSE, and FILL cards. The G4VNestedParameterisation class was used for Geant4. Then, organ dose coefficients (DCs) for idealized photon beams in the antero-posterior direction were calculated using the three codes and compared, showing excellent agreement (differences <3%). Additionally, organ DCs in other directions (postero-anterior, left-lateral, and right-lateral) were calculated and compared with those of the newborn and 1-year-old reference phantoms. Significant differences were observed (e.g., the stomach DC of the monkey was 5-fold greater than that of the 1-year-old phantom at 0.03 MeV) while the differences tended to decrease with increasing energy (mostly <20% at 10 MeV). The results of this study allows conducting MC simulations using the Visible Monkey to estimate organ-level doses, which should be valuable to support/improve monkey experiments involving ionizing radiation exposures.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1120-1130
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• 2020

The CREOLE experiment performed In the EOLE critical facility located In the Nuclear Center of CADARACHE - CEA have allowed us to get interesting and complete experimental information on the temperature effects in the light water reactor lattices. To analyze these experiments with accuracy an elaborate calculation scheme using the Monte Carlo method implemented in the MCNP6.1 code and the ENDF/B-VII.1 cross section library has been developed. We have used the ENDF/B-VII.1 data provided with the MCNP6.1.1 version in ACE format and the Makxsf utility to handle the data in the specific temperatures not available in the MCNP6.1.1 original library. The main purpose of this analysis is the qualification of the ENDF/B-VII.1 nuclear data for the prediction of the Reactivity Temperature Coefficient while ensuring the ability of the MCNP6.1 system to model such a complex experiment as CREOLE. We have analyzed the case of UO₂ lattice with 1166 ppm of boron in ordinary water moderator in specified temperatures. A detailed comparison of the calculated effective multiplication factors with the reference ones [1] in room temperature presented in this work shows a good agreement demonstrating the validation of our 3D calculation model. The discrepancies between calculations and the differential measurements of the Reactivity Temperature Coefficient for the analyzed configuration are relatively small: the maximum discrepancy doesn't exceed 1,1 pcm/℃. In addition to the analysis of direct differential measurements of the reactivity temperature coefficient performed in the poisoned UO₂ lattice configuration, we have also analyzed integral measurements in UO₂ clean lattice configuration using equivalency of the integral temperature reactivity worth with the driver core fuel reactivity worth and soluble boron reactivity worth. In this case both of the ENDF/B-VII.1 and JENDL.4 libraries were used in our analysis and the obtained results are very similar.

• Journal of Radiation Protection and Research
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• v.21 no.1
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• pp.41-50
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• 1996

In this study, the theoretical calculation of the air kerma-to-dose equivalent conversion factors was performed with a Monte Carlo N-Particle transport code for the two types of extremity phantom of the ANSI and the KAERI, respectively. Considering the distribution of absorbed dose due to the interaction of homogeneous Parallel broad beam of monoenergetic primary photons in the range between 15keV and 1.5MeV, the air kerma-to-dose equivalent conversion factors based on the kerma approximation were calculated. It is showed that all the theoretical conversion factors of the two types of the extremity phantom for the ANSI and the KAERI agree well with the experimental values of the ANSI N13.32 draft(1995) for each energy within 5.7%, maximum difference ratio, except for 13.6%, difference ratio in the case for the energy of less than 40keV. It is due to uncertainties of experiment occurred in the low X-ray energy range and geometry considered in the MCNP code.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.50-50
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• 1999