• 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.19 no.1
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• pp.13-22
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• 1994

The MCNP 4.2 code was used to calculate the thermal neutron flux distributions for $(n,\;{\gamma})$reaction in mainshell, annular plate, and subshell of the calandria of a CANDU 6 plant during operation. The thermal neutron flux distributions in calandria mainshell, annular plate, and subshell were in the range of $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$ which is somewhat higher than the previous estimates calculated by DOT 4.2 code. As an application to shielding analysis, photon dose rates outside the side and bottom shields were calculated. The resulting dose rates at the reactor accessible areas were below design target, $6 {\mu}Sv/h$. The methodology used in this study to evaluate the thermal neutron flux distribution for $(n,\;{\gamma})reaction$ can be applied to radiation shielding analysis of CANDU 6 type plants.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4426-4430
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• 2023

In this article, Exposure Buildup Factor(EBF) and the Energy Absorption Buildup Factor(EABF) have been determined for blood, brain, and muscle using the Monte Carlo method which is represented by MCNP5 codes and compared with geometric progression(G-P) fitting method which is represented by Phy-X/PSD online platform. The novelty of the present work is used an energy source of less than 0.1 MeV to determine buildup factors using MCNP5 and using Phy-X/PSD for some human tissues. thus, the energy range used in this case study was 0.06-3 MeV for penetration depths covered 0.5-3 MFP. Results of MCNP5 and Phy-X/PSD are validated against reference values of water that were reported at ANS-6.4.3. present results of EABFs and EBFs for the previously mentioned human tissues appeared good agreement between MCNP5 in comparison with Phy-X/PSD, whereas, the maximum average relative deviation did not exceed 2.37%. results of our article can be used in different medical applications, such as brachytherapy, radiotherapy, and diagnostics.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.735-740
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• 1998

방사성동위원소를 이용한 아스팔트함량 측정장비의 실험적인 방법에 의한 설계는 많은 시간과 비용이 소요되므로, 코드모사를 통해 설계할 경우 이러한 노력을 줄일 수 있다. 본 연구에서는 장비의 활용성을 증대시키기 위해 법적 규제 면제치인 100 $\mu$Ci이하의 방사성동위원소를 이용하며, 6%의 아스팔트함량을 갖는 혼합물을 5분간 측정하였을 경우 0.2%이내의 함량측정오차를 갖는 장비를 MCNP 코드를 이용하여 설계하였다 또한 코드 모사를 통한 설계를 바탕으로 장비를 제작한 후 5개의 시료에 대한 함량을 측정하고 그 결과를 비교하여 코드의 적용가능성을 검증하였다 실험결과 6.03% 아스팔트 함량을 가진 시료를 5분간 측정하여 5.85%의 함량을 얻을 수 있었다.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4447-4464
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• 2023

This paper focuses on the development of a new computational model of the CNESTEN's TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor k_eff. For that purpose, the T4 model was used to calculate the sensitivity profiles of the k_eff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the k_eff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the k_eff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the k_eff is not negligible and should be taken into account when performing S/U analysis.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.841-845
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• 2020

An airborne alpha beta detection system using passivated implanted planar silicon (PIPS) detector was modeled with the MCNP6 code and its resolution and detection efficiency were analyzed. Simulation of the resolution performed using the Gaussian energy broadening (GEB) function showed that the full width at half maximum (FWHM) of 35.214 keV for alpha particles was within 34-38 KeV, which is the FWHM range of the actual detector, and the FWHM of 15.1 keV for beta particles was constructed with a similar model to 17 keV, which is the FWHM range of an actual detector. In addition, the detection efficiency and the resolution were simulated according to the distance between the detector and the air filter. When the distance was decreased to 0.2 cm from 0.8 cm, the efficiency of the alpha and beta particles detection decreased from 5.33% to 4.89% and from 5.64% to 4.27%, respectively, and the FWHM of the alpha and beta particles improved from 40.9 KeV to 29.84 keV and 25.76 keV-13.27 keV, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.91.2-91
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• 2001

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.446-452
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• 2018

To improve the accuracy and safety of irradiation tests in High flux Advanced Neutron Application ReactOr (HANARO), the nuclear energy deposition rate, which is called nuclear heating, was estimated for an irradiation capsule with an iridium sample in the irradiation hole in order. The gamma rays emitted from the radioisotopes (RIs) of the structural materials such as flow tubes of fuel assemblies and heavy water reflector tank were considered as radiation source. Using the ORIGEN2.1 code, emission rates of delayed gamma rays were calculated in consideration of the activation procedure for 8 years and 2 months of HANARO operation. Calculated emission rates were used as a source term of delayed gamma rays in the MCNP6 code. By using the MCNP code, the nuclear heating rates of the irradiation capsules in the inner core, outer core, and heavy water reflector tank were estimated. Calculated nuclear heating in the inner core, outer core, and heavy water reflector tank were 200-260 mW, 80-100 mW, and 10 mW, respectively.