• Proceedings of the KIEE Conference
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• 2001.07b
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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.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2306-2312
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• 2020

Here we report on the first results of sensitivity evaluation of the gadolinium-aluminum-gallium- garnet (GAGG) scintillation detector with SiPM readout to fast and slow neutrons and, to the natural background and Co-60 γ-radiation as well. Data on sensitivity were obtained using certified dosimetry benches, so it can be utilized in the calculation of detection limits of neutron flux with such type of detectors. It was concluded that use of GAGG scintillator has a good prospect for neutron monitoring in different parts of nuclear research reactors and power plants.

• Journal of IKEEE
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• v.21 no.4
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• pp.408-411
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• 2017

In this paper, we propose an integrated control system that measures neutrons, gamma ray, and x-ray. The proposed system is able to monitor and control the data measured and analyzed on the remote or network, and can monitor and control the status of each part of the system remotely without remote control. The proposed system consists of a gamma ray/x-ray sensor part, a neutron sensor part, a main control embedded system part, a dedicated display device and GUI part, and a remote UI part. The gamma ray/x-ray sensor part measures gamma ray and x-ray of low level by using NaI(Tl) scintillation detector. The neutron sensor part measures neutrons using Proportional Counter Detector(low-level neutron) and Ion Chamber Type Detector(high-level neutron). The main control embedded system part detects radiation, samples it in seconds, and converts it into radiation dose for accumulated pulse and current values. The dedicated display device and the GUI part output the radiation measurement result and the converted radiation amount and radiation amount measurement value and provide the user with the control condition setting and the calibration function for the detection part. The remote UI unit collects and stores the measured values and transmits them to the remote monitoring system. In order to evaluate the performance of the proposed system, the measurement uncertainty of the neutron detector was measured to less than ${\pm}8.2%$ and the gamma ray and x-ray detector had the uncertainty of less than 7.5%. It was confirmed that the normal operation was not less than ${\pm}15$ percent of the international standard.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1022-1028
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• 2020

One of the possible options for spent-fuel management in Korea is pyroprocessing, which is a process for electrochemical recycling of spent nuclear fuel. Nuclear material accountancy is considered to be a safeguards measure of fundamental importance, for the purposes of which, the amount of nuclear material in the input and output materials should be measured as accurately as possible by means of chemical analysis and/or non-destructive assay. In the present study, a neutron measurement system based on the fast-neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) techniques was designed for nuclear material accountancy of a spent-fuel assembly (i.e., the input accountancy of a pyroprocessing facility). Various parameters including inter-detector distance, source-to-detector distance, neutron-reflector material, the structure of a cadmium sleeve around the close detectors, and an air cavity in the moderator were investigated by MCNP6 Monte Carlo simulations in order to maximize its performance. Then, the detector responses with the optimized geometry were estimated for the fresh-fuel assemblies with different ²³⁵U enrichments and a spent-fuel assembly. It was found that the measurement technique investigated here has the potential to measure changes in neutron multiplication and, in turn, amount of fissile material.

• Journal of Radiation Protection and Research
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• v.37 no.2
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• pp.70-74
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• 2012

Silicon carbide (SiC) is a promising material for neutron detection at harsh environments because of its capability to withstand strong radiation fields and high temperatures. Two PIN-type SiC semiconductor neutron detectors, which can be used for nuclear power plant (NPP) applications, such as in-core reactor neutron flux monitoring and measurement, were designed and fabricated. As a preliminary test, MCNPX simulations were performed to estimate reaction probabilities with respect to neutron energies. In the experiment, I-V curves were measured to confirm the diode characteristic of the detectors, and pulse height spectra were measured for neutron responses by using a $^{252}Cf$ neutron source at KRISS (Korea Research Institute of Standards and Science), and a Tandem accelerator at KIGAM (Korea Institute of Geoscience and Mineral Resources). The neutron counts of the detector were linearly increased as the incident neutron flux got larger.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2419-2431
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• 2023

Experiments related to Boron Neutron Capture Therapy (BNCT) accomplished at the Institute of Nuclear Techniques (INT), Budapest University of Technology and Economics (TUB) are presented. Relevant investigations are required before designing BNCT for vivo applications. Samples of relevant boron compounds (H₃BO₃, BDTPA) usually employed in BNCT were investigated with neutron beam. Channel #5 in the research reactor (100 kW) of INT-TUB provides the neutron beam. Boron samples are mounted on a carrier for neutron irradiation. The particle attenuation of several carrier materials was investigated, and the one with the lowest attenuation was selected. The effects of boron compound type, mass, and compound phase state were also investigated. To detect the emitted charged particles, a traditional ZnS(Ag) detector was employed. The neutron beam's interaction with the detector-detecting layer is investigated. Graphite (as a moderator) was employed to change the neutron beam's characteristics. The fast neutron beam was also thermalized by placing a portable fast neutron source in a paraffin container and irradiating the H₃BO₃. The obtained results suggest that the direct measurement approach appears to be insufficiently sensitive for determining the radiation dose committed by the Alpha particles from the ¹⁰B (n,α) reaction. As a result, a new approach must be used.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2004.10a
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• pp.60-63
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• 2004

An innovative in-core neutron flux mapping system has been developed and applied successfully for service in a commercial pressurized water reactor. With the benefit of double indexing path selector (Dip $s^{ⓡ}$) mechanism, the reliability of the detector drive system has been improved five times higher than that of conventional systems, and the problems caused by the serious friction generated between the detector cable and guide tubing has been solved completely because the Dip $s^{ⓡ}$ architecture allows the detector guide tubings to have larger curvature and shorter length in nature. The simple and fast maintenance is particularly emphasized in the detector drive system to secure minimum radiation exposure to the maintenance personnel by optimizing the number of components and providing easy access to the components. The programmable logic controller based digital controller with Window $s^{ⓡ}$ based operator s console provides fully automated and user friendly operation and maintenance support means.

• Nuclear Engineering and Technology
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• v.2 no.2
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• pp.85-95
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• 1970

An attempt has been made to do interpretation of the fast neutron dose with two threshold detectors incorporated with the Harwell criticality locket. This method is based on the assumption that the spectral distribution of fission neutrons in criticality accidents may be governed by one spectral parameter. The surface-absorbed dose for a unit fission neutron fluence seems to be insensitive to spectral shifts of the fission neutron spectrum. The average cross-sections for the activation detectors, however, are considerably changed with the neutron spectral shape, which may lead to a large error in calculating the dose from the reaction rate if one uses a fixed value for the average cross sections regardless of the neutron spectral distribution. Besides, the doses calculated from three representative formulae for fission neutron spectra have been compared : these formulae are Watt, Cranberg at al. and Maxwellian forms. The results obtained front the Maxwellian formula show a departure from the Watt and Cranberg's, both being similarly close.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2712-2722
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• 2023

For an accurate and efficient time-dependent Monte Carlo (TDMC) neutron transport analysis, several advanced methods are newly developed and implemented in the Seoul National University Monte Carlo code, McCARD. For an efficient control of the neutron population, a dynamic weight window method is devised to adjust the weight bounds of the implicit capture in the time bin-by-bin TDMC simulations. A moving geometry module is developed to model a continuous insertion or withdrawal of a control rod. Especially, the history-based batch method for the TDMC calculations is developed to predict the unbiased variance of a bin-wise mean estimate. The developed methods are verified for three-dimensional problems in the C5G7-TD benchmark, showing good agreements with results from a deterministic neutron transport analysis code, nTRACER, within the statistical uncertainty bounds. In addition, the TDMC analysis capability implemented in McCARD is demonstrated to search the optimum detector positions for the pulsed-neutron-source experiments in the Kyoto University Critical Assembly and AGN201K.

• 대한방사선방어학회:학술대회논문집
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• 2011.04a
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• pp.94-95
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• 2011