• 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.

• The Science & Technology
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• v.6 no.2 s.45
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• pp.28-31
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• 1973

원자로제어를 위한 중성자열전대의 응답시간 단축을 목적으로 진공증착된 박모열전대를 이용하여 중성자 열전대를 시작하였다. 이의 실험결과를 선열전대의 것과 비교하였으며, 열중성자동범위 2x(10에 8승)x8x10¹³ neutrons/cm²/sec에서 좋은 선형특성을 가지고 있었다. 시작된 박모중성자열전대를 사용하여 TRIGA MARK-Ⅱ 원자로 로필에서의 열중성자속분포를 측정하였다.

• Journal of the Korean Institute of Telematics and Electronics
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• v.9 no.5
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• pp.1-5
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• 1972

In order to improve the response time of nelltron theromopile for reactor control a neutron thermopile made use of a vacuunl evaporated thin film thor mocouple was fablicated and tested. The test results were compared with a wire-type neutron thermopile. Good linearities between the response of the neutron thermopile and the thermal flux has been shown in the ranges from n/$\textrm{cm}^2$/sec. Thermal neutron flux distributions in the core of TRIGA Mark-II reactor were measured using the fabricated neutron thermopile, and the results were conpared with data obtained by the acrivatin foil measurement.

• Journal of Radiation Protection and Research
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• v.30 no.2
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• pp.55-60
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• 2005

The evaluation of neutron flux distribution was performed for the ex-core detector design of SMART-P. DORT and MCNP code were used for the calculation of energy-dependent neutron flux distribution at 100% full power condition. Two code results show that maximum thermal flux appears at the $1^{st}$ water region in IST region and agree within 10% difference. In addition, another evaluation was performed code with assumptions that cote was composed of fission source and control rod without fuel assemblies. These assumptions make neutron count rate to be minimized. As a results, maximum thermal flux showed $6.99{\times}10^{-2}(n/cm^2-sec)$, when the strength of initial fission source was assumed as $1.0{\times}10^8(n/sec)$. The main reason of these results is due to the thermalization of fast neutrons in the water region and thermal flux is proportional to 80% of total neutron flux. Therefore, optimization of filler material of detector guide tube, position of installation and axial length of detector segments is necessary for the design of ex-core detector to enhance the neutron count rate and above results could be used in ex-core detector design as a fluence requirement.

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.145-154
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• 1988

This paper is a study on an accurate and computationally efficient method for reconstructing pointwise power distributions from coarse-mesh nodal calculations. The modern nodal codes can calculate global reactor power shapes and criticality very efficiently and accurately. But inherent in the nodal procedures, there is inevitable loss of information on local heterogeneous quantities. In this study, an improved form function method which reflects the exponential transition of the thermal flux near the assembly surface is developed for the reconstruction of the heterogeneous fluxes. Use of the new form function method in several pressurized water reactor (PWR) benchmark problems reduces the maximum errors in the reconstructed thermal flux to those in the reconstructed fast flux. Even for assemblies adjacent to the steel baffle in realistic PWR cores, use of this method also results in improved pointwise power reconstruction.

• Nuclear Engineering and Technology
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• v.5 no.4
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• pp.334-338
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• 1973

The horizontal distribution of the fast neutron flux density in the Bulk Shielding Experimental Tank of the TRIGA Mark-II reactor at the steady power of 250 KW has been measured using a solid state track detector which is natural mica placed in contact with $^{232}$ Th fissile foil. The neutron flux density was calculated on the assumption that the fast neutron spectrum is similar to that from the thermal-induced $^{235}$ U fission. The resulting flux density distribution along the horizontal line from the center of the thermalizing column door is presented in tabular and graphical forms.

• Proceedings of the Korea Society for Simulation Conference
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• 2004.05a
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• pp.89-93
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• 2004

본 MMI를 통해 개발된 웨스팅하우스 950 Mwe 최적 NPA는 기존의 단순한 Point Kinetics 모델이 아닌 정교한 3D 실시간 노심모델과 RETRAN 코드를 기반으로 하는 실시간 NSSS 열수력 모델(ARTS)이 통합된 모델을 갖추었으며, 해당형식 발전소(Westinghouse 3 Loop PWR Plant)의 여러 가지 과도사고를 실시간으로 정상, 비정상, 비상운전 모의할 수 있도록 개발되었다. 이 NPA는 기존의 유닉스 환경이 아닌 일반 범용 PC 서버와 윈도우즈 환경(Operating System)이라는 개방형 서버-클라이언트 구조를 채택하여 저렴하고 실용적인 시스템을 추구하였다. 다양한 색상 표현이 가능한 GUI 툴을 이용하여 노심 내부의 3D 열중성자 속 분포등 사용자가 직관적으로 알 수 있는 쉬운 구성의 클라이언트 제어 시스템을 개발, 연계하여 사용자의 편의성을 도모하였다.

• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.444-455
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• 1991

An accurate and fast running NEDAR model for calculating radial power profile throughout fuel life in both solid and annular pellets for existing and advanced CANDU-PHWR-fuel was developed in this work. This model contains resultant flux depression equations and neutron depression data tables which have been developed for CANDU-PHWR fuel of pellet with the diameter 8.0 to 19.5 mm and enrichment 0.71-6.0 wt % U-235, over a bumup range of 0 to 840 MWh /kgU (35000 MWD/T). In order to obtain the neutron flux distribution in the fuel pellet, the CE-HAMMER physics code was run for a neutron flux spectrum appropriate to a CANDU-PHWR to give predictions of radial power profile for several ranges of fuel design parameters. The results, which were calculated by the CE-HAMMER physics code, were fitted to an equation which is solved in terms of Bessel and exponential functions in order to obtain the parameters, $textsc{k}$, $\beta$ and λ in the resultant equation. The present NEDAR model produce a radial profile which, when normalized to unity at the pellet surface, is slightly higher than the profile of the original ELESIM data table. The predictions of the fission gas release by KAFEPA-NEDAR are in slightly better agreement with the experiments than those of ELESIM. The NEDAR model described in this study has been shown to provide an effective, reliable, and accurate method for determining radial power profiles in CANDU-PHWR fuel rods without incurring a significant increase in computing time.