• Journal of IKEEE
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• v.26 no.3
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• pp.515-519
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• 2022

In this paper, we propose the development of a high-sensitivity entry-level nuclide analysis module. The proposed measurement sensor module consists of an electronic driving circuit for nuclide analysis resolution, prototype production with nuclide analysis function, and GUI development applied to prototypes. The electronic part driving circuit for nuclide analysis resolution is divided into nuclide analysis resolution process by the electronic part driving circuit block diagram, MCU circuit design used for radiation measurement, and PC program design for Spectrum acquisition. Prototyping with nuclide analysis function is made by adding a 128×128 pixel OLED display, three buttons for operation, a Li-ion battery, and a USB-C type port for charging the battery. The GUI development department applied to the prototype develops the screen composition such as the current time, elapsed measurement time, total count, and nuclide Spectrum. To evaluate the performance of the proposed measurement sensor module, an expert witness test was conducted. As a result of the test, it was confirmed that the calculated result by applying the resolution formula to the Spectrum (FWHM@662keV) obtained using the Cs-137 standard source in the nuclide analysis device had a resolution of 17.77%. Therefore, it was confirmed that the nuclide analysis resolution method proposed in this paper produces improved performance while being cheaper than the existing commercial nuclide analysis module.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.781-790
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• 2012

The HyperGam program was developed for the analysis of complex HPGe ${\gamma}$-ray spectra. The previous version of HyperGam was mainly limited to the analysis of ${\gamma}$-ray peaks and the manual logging of the result. In this study, it is specifically developed into a tool for the isotopic analysis of spectra. The newly developed features include nuclide identification and activity determination. An algorithm for nuclide identification was developed to identify the peaks in the spectrum by considering the yield, efficiency, energy and peak area for the ${\gamma}$-ray lines emitted from the radionuclide. The detailed performance of nuclide identification and activity determination was accessed using the IAEA 2002 set of test spectra. By analyzing the test spectra, the numbers of radionuclides identified truly (true hit), falsely (false hit) or missed (misses) were counted and compared with the results from the IAEA 2002 tests. The determined activities of the radionuclides were also compared for four test spectra of several samples. The result of the performance test is promising in comparison with those of the well-known software packages for ${\gamma}$-ray spectrum analysis.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1120-1130
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• 2018

An analysis of a pyroprocessing safeguards methodology employing the Pu-to-$^{244}Cm$ ratio is presented. The analysis includes characterization of representative used nuclear fuel assemblies with respect to computed nuclide composition. The nuclide composition data computationally generated is appropriately reformatted to correspond with the material conditions after each step in the head-end stage of pyroprocessing. Uncertainty in the Pu-to-$^{244}Cm$ ratio is evaluated using the Geary-Hinkley transformation method. This is because the Pu-to-$^{244}Cm$ ratio is a Cauchy distribution since it is the ratio of two normally distributed random variables. The calculated uncertainty of the Pu-to-$^{244}Cm$ ratio is propagated through the mass flow stream in the pyroprocessing steps. Finally, the probability of Type-I error for the plutonium Material Unaccounted For (MUF) is evaluated by the hypothesis testing method as a function of the sizes of powder particles and granules, which are dominant parameters to determine the sample size. The results show the probability of Type-I error is occasionally greater than 5%. However, increasing granule sample sizes could surmount the weakness of material accounting because of the non-uniformity of nuclide composition.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.302-307
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• 1989

Nuclide release through penetrations in radioactive waste container is analyzed. Penetrations may result from corrosion or cracking and may be through the container material or through deposits of corrosion products. The analysis deals with the resultant nuclide release, but not with the way these penetrations occur. Numerical illustrations show that mass transport from multiple holes can be significant and may approach the mass transfer rate calculated from bare waste forms. Although partially-failed containers may present an important long-term barrier to release of radionuclides, numerous small holes on a container surface have the potential of bypassing the effectiveness of these barriers.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3878-3887
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• 2022

For the future energy-mix policy for carbon neutrality, demand for the capability of load-follow operation has emerged in nuclear power plants in order to accommodate the intermittency of renewable energy. The short-term decay heat analysis is also required to evaluate the decay heat level varied by the power level change during the load-follow operation, which is a very important parameter in terms of short-term decay heat removal during a grace time. In this study, the short-term decay heat level for 10 days after the shutdown was evaluated for both seasonal and daily load-follow cases. Additionally, the nuclide-wise contribution to the accumulated decay heat for 10 days was analyzed for further understanding of the short-term decay heat behavior. The result showed that in the seasonal case, the decay heat level was mainly determined by the power level right before the shutdown and the amount of each nuclide was varied with the power variation due to the long variation interval of 90 days. Whereas, in the daily case, the decay heat level was strongly impacted by the average power level during operation and meaningful mass variations for those nuclides were not observed due to the short variation interval of 0.5 days.

• Journal of radiological science and technology
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• v.44 no.4
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• pp.351-358
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• 2021

This research project is a program promoted to seek diversification of domestic radioactive waste analysis institutions, and seeks public development, win-win cooperation, and cooperation between the entrusted institution and the entrusted institution. Accordingly, the entrusted institution established a standard analysis procedure for establishing a quality control system for radioactivity analysis, establishing a radiation control zone, obtaining KOLAS accreditation, and performing proficiency tests, which are the performance ranges requested by the entrusted institution, and intersecting the radioactive isotope waste generated at the actual site. Verification was performed to confirm the analysis quality. In addition, facilities and equipment for radioactivity analysis were supplemented and expanded, and the basic technology foundation and technical skills were secured through securing professional technicians and education/training. It is judged that the entrusted institution will contribute to securing radiation safety through the smooth execution of treatment, disposal, and transportation through value creation and analysis of radioactive waste generated by radioactive isotope-using institutions (research institutes, hospitals, industries, etc.) by succeeding in this research project do.

• Journal of the Korean Society of Radiology
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• v.5 no.1
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• pp.5-10
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• 2011

Based on basic concept of detection limit, sample measurement time & background measurement time was considered, and MDA values according to background measurement time and sample measurement time in land samples(river soil, surface soil, drinking water, underground water, surface water, pine leaf, mugwort) analysis among environmental samples were compared. Seeing the water sample analysis result, it was shown that most of the samples were not detected, and most of the samples in land specimen analysis showed to be below the detection limit of "Ministry of Education, Science and Technology Announcement Je-2008-28-ho", but $^{137}Cs$ which is one of artificial radioactive nuclide was detected in some samples. It can be traced back to 1950s and 1960s when nuclear tests were carried out in atmosphere and catastrophic Chernobyl atomic power station accident that caused fallouts in the sky, and this is common level of detection that can be observed worldwide. Seeing the result that the $^{134}Cs$(which is a isotope of $^{137}Cs$, and it has relatively short half life) was not detected in all samples, it can be considered it doesn't affect to the operation of atomic power station.

• Journal of radiological science and technology
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• v.43 no.6
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• pp.481-487
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• 2020

Radiation measurement technology has steadily improved and its usage is expanding in various industries such as nuclear medicine, security search, satellite, nondestructive testing, environmental industries and the domain of nuclear power plants (NPPs). Especially, the simultaneous measurements of gamma rays and neutrons can be even more critical for nuclear safety management of spent nuclear fuel and monitoring of the nuclear material. A semiconductor detector comprising cadmium, zinc, and tellurium (CZT) enables to detect gamma-rays due to the significant atomic weight of the elements via immediate neutron and gamma-ray detection. Semiconductor sensors might be used for nuclear safety management by monitoring nuclear materials and spent nuclear fuel with high spatial resolution as well as providing real-time measurements. We aim to introduce a portable nuclide-analysis device that enables the simultaneous measurements of neutrons and gamma rays using a CZT sensor. The detector has a high density and wide energy band gap, and thus exhibits highly sensitive physical characteristics and characteristics are required for performing neutron and gamma-ray detection. Portable nuclide-analysis device is used on NPP-decommissioning sites or the purpose of nuclear nonproliferation, it will rapidly detect the nuclear material and provide radioactive-material information. Eventually, portable nuclide-analysis device can reduce measurement time and economic costs by providing a basis for rational decision making.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.621-622
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• 2004

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.313-342
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• 2014

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: $^{238}U$ radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of $^{239}Pu$).