• Nuclear Engineering and Technology
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• 제56권2호
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• pp.575-600
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• 2024

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%-8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ~4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ~6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.617-626
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• 2022

To effectively monitor the variety of distributions of neutron flux, fuel power or temperatures in the reactor core, usually the ex-core and in-core neutron detectors are employed. The thermocouples for temperature measurement are installed in the coolant inlet or outlet of the respective fuel assemblies. It is necessary to reconstruct the measurement information of the whole reactor position. However, the reading of different types of detector in the core reflects different aspects of the 3D power distribution. The feasibility of reconstruction the core three-dimension power distribution by using different combinations of in-core, ex-core and thermocouples detectors is analyzed in this paper to synthesize the useful information of various detectors. A comparison of multilayer perceptron (MLP) network and radial basis function (RBF) network is performed. RBF results are more extreme precision but also more sensitivity to detector failure and uncertainty, compare to MLP networks. This is because that localized neural network could offer conservative regression in RBF. Adding random disturbance in training dataset is helpful to reduce the influence of detector failure and uncertainty. Some convolution neural networks seem to be helpful to get more accurate results by use more spatial layout information, though relative researches are still under way.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1533-1540
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• 2023

Self-Powered Neutron Detectors(SPNDs) are used to calculate core power distributions, an essential factor in the safe operation of nuclear power plants. Some detectors may fail during normal operation, and signals from failed detectors are isolated from intact signals. The calculated detailed power distribution accuracy depends on the number of available detector signals. Failed detectors decrease the operating margin by enlarging the power distribution measurement error. Therefore, a thorough reconstruction of the failed detector signals is critical. This note suggests a compressive sensing based methodology that rationally reconstructs the readings of failed detectors. The methodology significantly improves reconstruction accuracy by sorting signals and removing high-frequency components from conventional compressive sensing methodology.

• 전기학회논문지
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• 제58권2호
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• pp.356-362
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• 2009

The ELectromagnetic Core Imperfection Detector (EL-CID) test was performed on a small generator in the laboratory and a gas turbine generator in the field to assess the fault condition of generator stator core. Artificial defects with six different sizes were introduced in the small generator. The scan results on six defects show a very large increase in the magnitude of fault current compared to that obtained with a healthy core. After the stator core heats up, a thermal imaging camera was used to detect hot spot on the inner surface of the core for comparison. Several faults were found during inspection of the gas turbine generator with the EL-CID. It has been shown that the existence of a fault can be determined by monitoring the magnitude of fault current.

• 한국방사선학회논문지
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• 제9권2호
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• pp.73-77
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• 2015

검출기의 자세한 구조를 알고자 CT 스캐닝을 하였으며 크리스털 형상과 사층에 관한 세부적인 구조를 전산모사 계산법을 이용해 재현하였다. 낮은 에너지의 감마선에 대한 피크 효율이 거리가 작아질수록 감소, 보다 높은 에너지(400 keV) 아상에서의 전체 효율성은 검출기 코어를 조정함으로써 불확도를 줄일 수 있었다. PENELOPE 계산법을 이용해 얻은 공간적 의존성 사이에 적절한 일치점이 달성되었음을 확인 하였다. 이는 크리스털 코어, 모서리와 크리스털 코어의 라운딩을 설명해 주는 매개변수들을 조정함으로써 달성되었다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2001년도 하계학술대회 논문집 B
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• pp.675-677
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• 2001

The neutron controls a nuclear fission in the core of reactor. In-core neutron detector drive system is a equipment that drives detector and cable to survey neutron flux in the reactor. The program introduced by this paper governs proto-type drive system. The basic function of drive system is the insert and the withdraw of a cable, and the control of the movement speed. Also this program have a special function, test, auto operation, to increase the capacity of drive system.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.58-63
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• 1997

This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore defector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place.

• Journal of Radiation Protection and Research
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• 제30권2호
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• pp.55-60
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• 2005

SMART 연구로의 노외계측기 설계를 위하여 고온 전출력 조건과 중성자 계수율이 최소가 되는 조건에 대해서 중성자속 분포 평가를 수행하였다. 고온 전출력 조건에서 IST 영역의 에너지 구간별 중성자속 분포 계산은 DORT와 MCNP코드를 이용하였으며, 계산 결과 IST 내의 첫 번째 물 영역에서 최대의 열중성자속을 보였고 두 코드 결과는 대략 10% 이내에서 일치하는 것으로 나타났다. 그리고 중성자 계수율이 최소가 되는 조건에서 노외계측기 설치 영역에서의 중성자속을 계산한 결과, 선원의 세기가 $1.0{\times}10^8(n/sec)$이라고 가정한 경우 최대 열중성자속의 크기는 $6.99{\times}10^{-2}(n/cm^2-sec)$로 전체 중성자속의 80% 이상을 차지하는 것으로 나타났는데 이는 IST 철 구조물을 통과한 속중성자가 감속능이 큰 물 영역에서 에너지를 잃고 열중성자로 변하였기 때문이다. 그러므로 노외계측기 설계시 계측기를 둘러싸는 계측기 안내관 충전물질, 설치위치 그리고 각 계측기 Segment들의 길이 등을 최적화하여 중성자 계수율을 증가시키는 방안을 모색할 필요가 있겠으며, 이러한 중성자속 평가 결과는 노외계측기가 IST 영역에 설치될 경우 노외계측기 선속 요건으로 이용될 수 있다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 학술대회 논문집 정보 및 제어부문
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• pp.225-227
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• 2005

The in-core neutron Flux Mapping System in a pressurized water reactor yields information on the neutron flux distribution in the reactor core at selected core locations by means of movable detectors. The obtained data are used to verify the reactor core design parameters. The detector cables run through guide tubes(thimbles), and typically thirty-six to fifty-eight thimbles are allocated in the reactor depending on the number of fuel assemblies. These thimbles are inserted into nuclear fuel assemblies through conduits connected from the bottom of the reactor vessel to a seal table. During the plant refueling outage period, the thimbles are withdrawn up to 4m from the seal table, the height of a nuclear fuel. In spite of their importance, however, the thimble handling work has been performed by only human operators. In addition, its efficiency is very low due to narrow working environments on the seal table, thereby resulting in the excessive radiation exposure of maintenance personnel. To solve these problems, a new thimble handling equipment for in-core flux mapping system was developed, and we confirmed its effectiveness through experiments.