• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.433-439
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• 1996

노심 외각에 설치되어 노심 외각으로 누설되는 중성자를 검출하여 노심내의 출력 변화를 지시해 주는 노외계측기(Excore Detector)는 운전중 노심의 변화를 정확히 감지하도록 정기적으로 교정되어져야 한다. 노외계측기는 노내계측기(Incore Detector)를 통하여 측정되어진 축방향 출력편차(Axial Offset)를 이용하여 교정하고 있다. 기존의 방법은 노내계측기로 최소한 4회 노심 출력을 측정하여 최소자승법(Least Square Method)으로 상수들을 구한후 노외계측기를 교정한다. 여기서 소개되는 단순 노외계측기 교정법은 노내계측기로 2회 측정되어진 자료들을 이용하는 2점 교정법과 1회 측정되어진 자료들을 이용하는 1점 교정법으로, 계측기 반응상수(Detector Response Factor)를 계산한후 교정되어진 노외계측기의 출력편차를 측정값과 비교하였다. 위의 두가지 방법을 고리 3호기 9주기, 10주기에 적용하여 노심 운전영역(~$\pm$10%)에서 2점 교정법은 최대 1.40 %, 1점 교정법은 최대 0.63 %의 오차를 보여주고 있다. 단순 노외계측기 교정법은 노심출력을 1회 또는 2회 측정하므로 교정시간을 줄이고 제어봉의 사용을 억제하여 방사성 폐기물을 감소시키는 효과와 기존의 교정 방법과 같은 정확성을 기대할수 있다.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.97-104
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• 2004

Neutron induced prompt gamma-ray spectroscopy(NIPS) system measures the prompt gamma-ray emitting by the interaction of a neutron with various materials. This system will be of great benefit to scientists worldwide, since it provides the non-destructive measurement of many element in either solid or liquid wastes. In this study, the full-energy-peak (FEP) efficiency calibration for a HPGe detector was constructed in the ${\gamma}$-ray energy range from 80 keV to 8 MeV, using $^{l33}$Ba and >TEX>$^{152}Eu$ RI sources and $ ^{35}Cl(n, ${\gamma}$)^{36}Cl$ thermal neutron captured reaction. The FEP efficiency curve for the higher energies using the $^{35}Cl(n, ${\gamma}$)^{36}Cl$ reaction was normalized with the curve obtained from the RI sources, since the accurate activity of its prompt ${\gamma}$-ray is unknown. The average thermal neutron flux was theoretically calculated using the FEP efficiency curve for the KCl standard solutions. The NIPS system equipped with a ${\gamma}$-${\gamma}$ coincidence setup with two n-type coaxial HPGe detectors was considered in order to reduce the interfering ${\gamma}$-ray background. The FEP efficiency curve for the ${\gamma}$-${\gamma}$ coincidence system was also obtained for full energy range. The performance of the normal and coincidence NIPS system was tested by comparing signal-to-noise ratio in each mode using the reference sample.e.

• Proceedings of the KIEE Conference
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• 2006.07d
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• pp.1995-1996
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• 2006

검출기 안내관은 노내 핵계측 계통의 중성자 분포 측정을 위한 이동형 검출기의 이동경로를 제공할 뿐만 아니라 원자로 냉각수 계통의 압력경계를 유지하는 안전성 등급의 중요한 설비이다. 그러나, 인출과 삽입을 위한 검출기 안내관 취급은 의외로 낙후되어 작업자의 인력에만 의존하고 있는 실정이며, 원자로 격납용기 내부에 위치한 고방사선 지역에서 작업이 수행되고 있는 실정이다. 따라서 노내 핵계측 계통의 검출기 안내관의 안정적인 관리를 위해 검출기 안내판을 일정한 힘으로 인출하고 삽입할 수 있는 자동화시스템의 개발이 이루어지고 있다. 전력연구원에서 개발한 안내관 취급기구는 롤러에 의해 안내관을 파지하고, DC 모터 구동에 의해 안내관을 인출하고 삽입하는데, 본 논문에서는 안내관 취급 기구의 제어 시스템 구성과, 롤러와 안내관 사이에 발생하는 슬립을 고려한 제어기 구조를 제안하고, 실험을 통해 구현한 제어 시스템의 효용성을 보인다.

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.79-86
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• 2015

$^3He$ gas has been used for neutron monitors as the neutron converter owing to its advantages such as high sensitivity, good ${\gamma}$-discrimination capability, and long-term stability. However, $^3He$ is becoming more difficult to obtain in last few years due to a global shortage of $^3He$ gas. Accordingly, the cost of a neutron monitor using $^3He$ gas as a neutron converter is becoming more expensive. Demand on a neutron monitor using an alternative neutron conversion material is widely increased. $^{10}B$ has many advantages among various $^3He$ alternative materials, as a neutron converter. In order to develop a neutron converter using $^{10}B$ (actually $B_4C$), we calculated the optimal thickness of a neutron converter with a Monte Carlo simulation using MCNP6. In addition, a neutron converter was fabricated by the Ar sputtering method and the neutron signal detection efficiencies were measured with respect to various thicknesses of fabricated a neutron converter. Also, we developed a 2-dimensional multi-wire proportional chamber (MWPC) for neutron beam profile monitoring using the fabricated a neutron converter, and performed experiments for neutron response of the neutron monitor at the 30 MW research reactor HANARO at the Korea Atomic Energy Research Institute. The 2-dimensional MWPC with boron ($B_4C$) neutron converter was proved to be useful for neutron beam monitoring, and can be applied to other types of neutron imaging.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11b
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• pp.655-662
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• 1996

기술적으로는 다짐장비의 개선으로 성토다짐의 시공속도가 매우 빨라지고 있으나 국내에서 적용하고 있는 현장다짐 밀도 및 수분함량 측정 방법(KS F 2311 KS F 2306)$^{(1)}$ 은 신속한 측정을 어렵게 하고 있다. 본 연구는 외국에서 개발 현재 활발히 적용되고있는 방사성동위원소를 이용한 함수량측정기개발의 시작단계로 기본설계 결정을 목적으로 한다. 함수량 측정 RI계기의 원리를 먼저 살펴본 후 실험실 내에서 자연건조된 성토용 흙으로 다짐을 하여 공시체를 제작한 뒤, 실험실용 함수량측정 RI계기를 공시체위에 놓고 폴리에틸렌과 중성자 검출기의 개수를 변화시켜가며 일정시간동안 측정개수를 측정한 값을 분석한 결과 목표측정 시간을 1분으로 하였을 때 신뢰측정개수인 10,000개이상을 계측하면서도 경제적으로 최소인 중성자검출기의 개수는 2개, 폴리에틸렌의 두께는 7cm로 결정되었다.

• Nuclear Engineering and Technology
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• v.16 no.1
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• pp.18-20
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• 1984

The detection limit of boron has been lowered further in the capture gamma-ray measurement after preconcentration of boron by placing natural lithium brick in front of Ge(Li) detector. The experimental detection limit is found to be 0.30ppm, 0.18ppm, 0.045ppm and 0.090ppm for the samples of aluminum, steel, uranium dioxide and graphite, respectively. An alternate counting technique kas been also used for neglecting the error caused by the fluctuation of neutron flux during counting.

• Analytical Science and Technology
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• v.16 no.1
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• pp.12-24
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• 2003

For the design and set-up of neutron induced prompt ${\gamma}$-ray spectroscopy system using $^{252}Cf$ neutron source, the effects of shielding and moderator materials have been examined. The $^{252}Cf$ source being used for TLD badge calibration in Korea Atomic Energy Research Institute was utilized for this preliminary experiment. The ${\gamma}$-ray background and prompt ${\gamma}$-ray spectrum of the sample containing Cl were measured using HPGe (GMX 60% relative efficiency) located at the inside of the system connected to notebook PC at the outside of the system (about 20 meter distance). The background activities of neutron and ${\gamma}$-rays were measured with neutron survey meter as well as ${\gamma}$-ray survey meters, respectively and the system was designed to minimize the activities. Prompt ${\gamma}$-ray spectrum was measured using ${\gamma}$-${\gamma}$ coincident system for reduce the background and the continuum spectrum. The optimum system was designed and set up using the experimental data obtained.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.58-65
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• 1990

A PC-based system for measuring and analysing random neutron process in the thermal reactor is developed and applied to TRIGA Mark-II reactor at KAERI. It is confirmed that this system has several advantages compared to conventional methods. So far, two techniques, autocorrelation and variance to mean ratio (VTMR), have been applied for analysing the count data collected from the single detector by using this system. The results of the two techniques agree within acceptable difference, but VTMR's results show much superior statistical reliability than those of autocorrelation especially when it is near critical. The $\beta$/Λ of TRIGA Mark-II reactor is measured to be about 125/sec when the reactivity is within -3＄ and about 150/sec when it is below -4＄.

• Journal of Sensor Science and Technology
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• v.25 no.4
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• pp.264-270
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• 2016

Self-powered neutron detector (SPND) is a sensor to monitor a neutron flux proportional to a reactor power of the nuclear power plants. Since an SPND is usually installed in the reactor core and does not require additional outside power, it generates electrons itself from interaction between neutrons and a neutron-sensitive material called an emitter, such as rhodium and vanadium. This paper presents the simulations of the depletion sensitivity evaluations based on MCNP models of rhodium and vanadium SPNDs and light water reactor fuel assembly. The evaluations include the detail geometries of the detectors and fuel assembly, and the modeling of rhodium and vanadium emitter depletion using MCNP and ORIGEN-S codes, and the realistic energy spectrum of beta rays using BETA-S code. The results of the simulations show that the lifetime of an SPND can be prolonged by using vanadium SPND than rhodium SPND. Also, the methods presented here can be used to analyze a life-time of those SPNDs using various emitter materials.