• Journal of Energy Engineering
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• v.17 no.4
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• pp.233-240
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• 2008

Conventionally, the second-order self-adjoint neutron transport equations have been studied using the even parity and the odd parity equations. Recently, however, the SAAF(self-adjoint angular flux) form of neutron transport equation has been introduced as a new option for the second-order self-adjoint equations. In this paper we validated the SAAF equation mathematically and clarified how it relates with the existing even and odd parity equations. We also developed a second-order SAAF differencing formula including DSA(diffusion synthetic acceleration) from the first-order difference equations. Numerical result is attached to show that the proposed methods increases accuracy with effective computational effort.

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.371-379
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• 1990

The Xenon spatial oscillation may give rise to operational difficulties in a nuclear power plant. In this study, in order to investigate the Xenon instability for a PWR, the frequency-domain technique is adopted by using Generalized Nyquist Criterion, which is more general and suitable for the multi-input/multi-output system. Also linearized modal fluxes are obtained by a modal expansion. This model has been implemented to test the axial Xenon stability of YGN-1 unit against the changes in plant operating parameters ; power level, control rod position, and core average burnup. The results show that the increase of power level and the deeper insertion of control rod have the destabilizing effect, and that the burnup progress makes the core less stable. Also the results show that the overestimation due to modal interaction was found not to be significant.

• Nuclear Engineering and Technology
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• v.15 no.1
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• pp.57-69
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• 1983

The LWR fuel performance analysis computer code, FRAPCON-1, are evaluated to investigate the performance of CANDU fuel elements loaded in Wolsung-1 reactor. The FRAPCON-1 models of neutron flux depression in fuel and of fuel-to-cladding heat transfer are modified, and the validity of fission gas release model for CANDU fuel is evaluated. And the heavy water properties are provided in calculating the heat transfer coefficient between cladding and coolant. By using the modified code, FRAPCON-1-CSK, the sensitivity studies are carried out for Wolsung-1 fuel element design parameters. The performance analysis is also performed for Wolsung-l fuel elements. The calculated results are discussed in terms of. LWR fuel design criteria because of unavailability of CANDU fuel design criteria.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.205-211
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• 1994

The best method is selected among the 3 digital dynamic compensation methods which are developed or applied for the Rhodium self-powered neutron detector. The three digital dynamic compensation methods are the existing Dominant Pol Tustin method of the COLSS(Core Operating Limit Supervisory System), the Direct Inversion method and Kalman Filter method. The Direct Inversion method is an improved method of D. Hoppe and R. Maletti and the Kalman Filter method is developed using the Kalman Filter. Response times of the compensated signals to achieve 90% of a step input are 28.1, 17.2 and 6.5 seconds respectively for the same noise gain telling that the Kalman Filter method is the best amens the 3 methods.

• Nuclear Engineering and Technology
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• v.19 no.3
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• pp.180-185
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• 1987

A computer code, KAFEPA, for analysing in-reactor behavior of a PHWR-fuel rod under reactor normal operating condition was developed. This code, KAFEPA, corresponds to the ELESIM code that was developed for the same purpose by AECL. Even though the KAFEPA originated from the ELESIM, it contains more accurate and theoretical models in comparison with the ELESIM, such as fission gas release model, in-reactor densification model and a new database for neutron flux depression across the radial direction in a fuel pellet. The KAFEPA code was verified by comparing the predictions with 22 measurements of fission product gas release. The predictions of the KAFEPA was well agreed with the experimental data.

• Proceedings of the Korea Information Processing Society Conference
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• 2001.04b
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• pp.841-844
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• 2001

원자력 발전소의 에너지를 생성하는 원자로의 동특성(Reactor Dynamic Characteristics)은 반응도(Reactivity) 영향인자가 변하여 원자로에 가해지는 외란(disturbances)에 의한 반응도 궤환(feedback)과 부가적으로 변하는 각종 원자로 계통의 설계변수 들에 대한 복잡한 노물리 현상을 통해 결정되며 이러한 현상을 이해하는 것은 원자력 발전소 종사자에게는 무척이나 중요한 일이다. 본 고에서는 가상현실 등 첨단기법을 이용하여 전력연구원에서 개발한 교육지원시스템(VRCATS)의 일환으로 강의실 이론 교육 시에 원격으로 시뮬레이터에 접속하여 각종 반응도 변화를 통한 원자로 노심상태 즉, 노심 내 전체 중성자속분포, 각종 온도 분포를 실시간으로 3 차원으로 보여주며, 시간에 따른 제논, 보론농도 등 반응도 변화 인자 및 총 반응도 변화추이 등을 감시 할 수 있는 컴퓨터 지원 노심설계 및 훈련 시스템(PREMARK) 개발 내용 및 특징을 기술하였다.

• Proceedings of the Korean Vacuum Society Conference
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• 2012.02a
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• pp.454-454
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• 2012

Magnetron sputtering에서, 영구자석의 자속은 target 표면 가까이에 전자를 구속한다. 구속된 전자는 Ar중성기체와 충돌하여 Ar이온을 발생시킬 수 있으므로, target 근처에서의 플라즈마 밀도를 높여, 자석이 없을 때보다 낮은 압력 또는 낮은 전압에서 방전할 수 있다. 구속 전자가 밀집된 공간에서 sputtering 현상이 주로 발생하기 때문에, planar target을 사용할 경우에는 target이 불균일하게 식각되어 target의 사용효율이 좋지 못하다. 이에 대한 한 가지 대안은 target을 원통형으로 만들어 회전시키는 것이다. Cylindrical target 의 내부에 위치한 영구자석은 고정시키고, target만을 회전시키면 비교적 균일하게 식각되므로 target의 사용효율을 높일 수 있다. 본 연구에서는 기존의 planar target에 대한 Particle-In-Cell Simulation을 Cylindrical target 에 적용시키기 위한 방법을 알아본다. 또한, 개발된 Simulator를 이용하여, Sputtering 조건의 변화에 대한 I-V curve의 변화를 살펴본다.

• 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.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.74-83
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• 1995

The accurate determination of the fast neutron flux/fluence onto the pressure vessel is an essential part of the reactor pressure vessel surveillance program. It has been reported recently that the iron cross section data in ENDF/B versions III through V might underestimate the flux/fluence of fast neutrons in steel structures such as reactor pressure vessel. In this study, for the comparison of iron data of ENDF/B-IV and VI we produced two 47-group cross section sets, CXFe-IV and CXFe-Ⅵ, which are based on Yonggwang nuclear unit-3/4 model and the iron data of ENDF/B-IV and VI, respectively. A comparison was made of the results obtained from DOT4.3 calculation using CXFe-IV and CXFe-VI. From the results, it was found that the fast flux(E 〉 1.0 MeV), which is important for the pressure vessel embrittlement analysis, increases by about 7.6％ at the inner wall and 20％ at the outer wall of the vessel, if the iron data are used from ENDF/B-VI instead of ENDF/B-IV.

• Journal of the Korean Society of Radiology
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• v.4 no.4
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• pp.29-32
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• 2010

The neutron capture cross section of $^{197}Au$ has been measured relative to the $^{10}B(n,{\gamma})$ standard cross section by the neutron time-of-flight(TOF) method using a 46-MeV electron linear accelerator(linac) at the Research Reactor Institute, Kyoto University(KURRI). In order to experimentally prove the result obtained, the supplementary cross section measurement has been made from 0.1 eV to 10 keV using the Kyoto University Lead slowing-down spectrometer (KULS) coupling to the linac. The relative measurement by the TOF method has been normalized to the reference value(24.5 b) at 1 eV. The evaluated capture cross sections in JENDL/D-99 Dosimetry have been compared with the current measurements by the KULS experiments.