Comparison of Uranium Analysis Methods Using Alpha Spectrometer and Thermal Ionization Mass Spectrometer (TIMS)

Taeyang Ban,Daehyeon Kim,Jongki Choi,Jungbok Lee,Hajin Song,Eunju Kim,Ahreum Lim,Soohwan Kim,Jungsuk Oh 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.1

Radionuclide analysis methods must be secured in the event of emergencies such as the discovery of unknown nuclear material or nuclear accidents in neighboring countries or Korea. Most institutions in Korea are in their early stages of radionuclide analysis method development and do not even have Radiation Controlled Areas where they can handle the samples safely. Some institutions such as the Korea Atomic Energy Research Institute have the ability to perform radionuclide analysis for nuclear facilities or verification of nuclear activities. In Korea, it is necessary to secure nuclide analysis technology to enable independent verification in times of emergency or need. This paper analyzes uranium as the target nuclide using alpha spectrometer and TIMS. Alpha spectrometer detects alpha particles emitted from uranium samples and measures the concentration of uranium isotopes. This method has a high selectivity that distinguishes it from other elements, and accurate measurements can be made even when uranium samples are mixed with other elements. In addition, there is minimal interference from other radioactive isotopes in the sample, and the sample preparation is simple, resulting in relatively short analysis times. In contrast, TIMS detects ionized uranium ions by heating the uranium sample. This method may have potential interference from other elements and may take relatively longer analysis times. However, TIMS has high sensitivity and accuracy and can detect various elements other than uranium, making it suitable for various analyses. Therefore, when analyzing uranium, it is recommended to select and use the appropriate device according to the purpose, as both alpha spectrometer and TIMS have their pros and cons. Furthermore, by using both devices in parallel, more accurate and reliable results can be obtained. This paper aims to compare the analysis methods of alpha spectrometer and thermal ionization mass spectrometry, which are widely used for nuclide analysis in unknown nuclear materials.

NEW DEVELOPMENT OF HYPERGAM AND ITS TEST OF PERFORMANCE FOR γ-RAY SPECTRUM ANALYSIS

Park, B.G.,Choi, H.D.,Park, C.S. Korean Nuclear Society 2012 Nuclear Engineering and Technology Vol.44 No.7

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.

DeCART2D/MASTER Core Follow Calculation for Hanbit Unit 3 and Comparison With McCARD Single Fuel Assembly Burnup Analyses

Jeong Woo Park,Seung-Ah Yang,Ho Jin Park 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.2

In the design of a spent-fuel (SF) storage, the consideration of burnup credit brings the benefits in safety and economic views. According to it, various SF burnup measurement systems have been developed to estimate high fidelity burnup credit, such as FORK and SMOPY. Recently, there are a few attempts to localize the SF burnup measurement system in South Korea. For the localization of SF burnup measurement systems, it is very important to build the isotope inventory data base (DB) of various kinds of SFs. In this study, we performed DeCART2D/MASTER core follow calculations and McCARD single fuel assembly (FA) burnup analyses for Hanbit unit 3 and confirmed the characteristic of the isotope inventory over burnup. Firstly, the core follow calculations for Cycles 1~7 were performed using DeCART2D/MASTER code system. The core follow calculation is very realistic and practical because it considers the design conditions from its nuclear design report (NDR). Secondly, the Monte Carlo burnup analyses for single FAs were conducted by the McCARD Monte Carlo (MC) transport code. The McCARD code can utilize continuous energy cross section library and treat complex geometric information for particle transport simulation. Accordingly, the McCARD code can provide accurate solutions for burnup analyses without approximations, but it needs huge computing resources and time burden to perform whole-core follow calculations. Therefore, we will confirm the effectiveness of the single McCARD FA burnup analyses by comparing the DeCART2D/MASTER core follow results with the McCARD solution. From the results, the use of single FA burnup analyses for the establishment of the DBs will be justified. Various FAs, that have different 235U enrichments and loading pattern of fuel rods and burnable absorbers, were considered for the burnup analyses. In addition, the results of the sensitivity analyses for power density, initial enrichment, and cooling time will be presented.

Design and Implementation of a Thermal Analysis Framework for Solidified Waste Form Containing Heat-Generating Nuclides

Jong Kwang Lee,Jung-Hoon Choi,Byeonggwan Lee,Hwan-Seo Park 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2

The design and fabrication of suitable waste forms with high thermal and structural stability are essential for the safe management and disposal of radioactive wastes. In particular, the thermal properties and temperature distribution of waste form containing high heat-generating nuclides such as Cs and Sr can be used to evaluate its thermal stability, but also provide useful information for the design of canisters, storage systems, and repositories. In this study, a new program code-based thermal analysis framework has been developed to facilitate the characterization, design, and optimization of the waste form. Matlab was used as a software development platform because it provides powerful mathematical computation and visualization components such as the partial differential equation (PDE) toolbox for solving heat transfer problems using finite element method, the App Designer for graphical user interface (GUI), and the MATLAB Compiler for sharing MATLAB programs as standalone applications and web applications. The thermal analysis results such as temperature distribution, heat flux, maximum/ minimum temperature, and centerline/surface temperature profile are visualized with graphs and tables. To evaluate the effectiveness of the developed program, several design and optimization studies were carried out for the SrTiO3 waste form, selected as a stable form of strontium nuclide.

고감도 보급형 핵종 분석 모듈 개발

오승진,이주현,이승호 한국전기전자학회 2022 전기전자학회논문지 Vol.26 No.3

In this paper, we propose the development of a high-sensitivity entry-level nuclide analysis module. The proposedmeasurement sensor module consists of an electronic driving circuit for nuclide analysis resolution, prototypeproduction with nuclide analysis function, and GUI development applied to prototypes. The electronic part drivingcircuit for nuclide analysis resolution is divided into nuclide analysis resolution process by the electronic partdriving circuit block diagram, MCU circuit design used for radiation measurement, and PC program design forSpectrum 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 developmentdepartment applied to the prototype develops the screen composition such as the current time, elapsed measurementtime, total count, and nuclide Spectrum. To evaluate the performance of the proposed measurement sensor module, anexpert witness test was conducted. As a result of the test, it was confirmed that the calculated result by applying theresolution formula to the Spectrum (FWHM@662keV) obtained using the Cs-137 standard source in the nuclide analysisdevice had a resolution of 17.77%. Therefore, it was confirmed that the nuclide analysis resolution method proposed inthis paper produces improved performance while being cheaper than the existing commercial nuclide analysis module. 본 논문에서는 고감도 보급형 핵종 분석 모듈 개발을 제안한다. 제안하는 측정센서 모듈은 핵종 분석 분해능을 위한 전자부 구동회로, 핵종 분석 기능이 적용된 시제품 제작, 시제품에 적용되는 GUI 개발 등으로 구성된다. 핵종 분석 분해능을 위한 전자부 구동회로는 전자부 구동 회로 블록도에 의한 핵종 분석 분해능 과정, 방사선 측정에 사용되는 MCU 회로 설계, Spectrum 취득용 PC프로그램 설계 등으로 나뉘어진다. 핵종 분석 기능이 적용된 시제품 제작은 128x128 픽셀의 OLED display, 조작을 위한 3개의버튼, Li-ion 배터리, 배터리 충전을 위한 USB-C Type 포트의 구성을 추가하여 제작한다. 시제품에 적용되는 GUI 개발부는 현재시간, 측정 경과 시간, 토탈 카운트, 핵종 Spectrum 등의 화면구성으로 개발한다. 제안된 측정센서 모듈의 성능을 평가하기 위하여공인기관 전문가 입회시험을 시행한 결과, 핵종 분석 장치에 Cs-137 표준선원을 이용하여 취득한 Spectrum(FWHM@662keV)으로 분해능 공식을 적용하여 계산한 결과가 17.77%의 분해능을 가짐이 확인되었다. 따라서, 제안된 본 논문에서 제안한 핵종 분석분해능 방법이 기존의 상용의 핵종 분석 모듈보다 저렴하면서도 향상된 성능이 산출됨이 확인되었다.

Cost Benefit Analysis of Treatment Process for Spent Filter

Narae Lee,Jisoo Yoon,Kyung Rok Park,Moonoh Kim 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2

The treatment process for Spent Filter(SF) of Kori-1 was developed that includes the following : 1) Taking out by robot system 2) Screening by ISOCS 3) Collection of representative samples using a sampling machine 4) Compression 5) Immobilization 6) Packaging and nuclide analysis and 7) Delivery/disposal. Although the robot system, ISOCS, sampling machine and immobilization facility are essentially required for building the above processing but decision to build the compression system and nuclide analysis system must be made after reviewing the need and cost benefit for their construction. In addition, for effcient SF treatment, it is necessary to determine the nuclide concentration range of the SF to which immobilization will be applied. In this study, a cost benefit analysis was performed on existing and alternative methods for processes related to compression treatment, nuclide analysis and immobilization methods, which are greatly affected by economics and efficiency according to the design. First, although the disposal cost is reduced with reducing the number of packaging drums by compressed and packaged but the expected benefits not be equal to or greater than the cost invested in building a compression system. As a result, non-compressed treatment of SF is expected to be economical because the construction cost of compression system is more expensive than the benefits of reducing disposal costs by compression. Second, a cost benefit analysis of direct and indirect nuclide analysis methods was performed. For indirect analysis, scaling factors should be developed and the drum scanner suitable for the analysis for DAW should be improved. As a result, direct analysis applied grouping options is expected to be more economical than indirect analysis requiring the cost for developing scaling factors and improving the scanner. Third, it is timeconsuming and inefficient to distinguish and collect filters that are subject to be immobilized according to the waste acceptance criteria among the disorderly stored SFs in the filter rooms. If the benefits of immobilization of the SFs selectively are not greater than the benefits of immobilization of all SFs, it can be economical to immobilize all SFs regardless of the nuclide concentration of them. As a result, it is more economical to immobilize all SFs with various nuclide concentrations than to selectively immobilize them. The conclusion of this study is that it is not only cost-effective but also disposal-effective to design the treatment process of SF to adopt non-compressed processing, direct analysis and immobilization of all SFs.

A Preliminary Study on Non-Fuel Radioactive Waste Stored in Spent Fuel Pool of PWR

Kyungho Roh,Younghwan Hwang,Beomgyu Kim,Sukwon Jung,Mihyun Lee 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.1

The treatment of waste generated during operation as a part of preparation for decommissioning is coming to the fore as a pending issue. Non-fuel waste stored in the spent fuel pool (SFP) of PWRs in Korea includes Dummy fuel, damaged fuel rod storage container, reactor vessel specimen cask, spent in-core instrumentation, spent control element assemblies, spent neutron source assemblies, burnable poison rods, etc. In order to treat such waste, it is necessary to classify radioactive waste level and analyze kinds of nuclide in accordance with legal requirements. In order to solve the problem, the items that KHNP-CRI is trying to conduct like followings. First, KHNP-CRI will identify the current status of non-fuel waste stored in the SFP of all domestic nuclear power plants. In order to consider the treatment of non-fuel waste, it is essential to know what kind of items and how many items are stored in the SFP. Second, to identify the dimension and characteristics of non-fuel waste stored in the SFP would be conducted. The configuration of non-fuel waste is important information to handle them. Third, the way to handle non-fuel waste would be deduced including analysis of their dimension, whether the equipment should be developed to handle each kind of non-fuel waste or not, how to transport them. In order to classify radioactive waste level and analyze the nuclide for the non-fuel waste, handling tools and the cask to transport them into the facility which nuclide analysis is able to be performed would be required. Fourth, the nuclide analysis technology would be identified. Also, domestic holding technology would be identified and which technology should be developed to classify the radioactive waste level for the non-fuel waste would be deduced. This preliminary study will provide KHNP-CRI with the insight for the nuclide analysis technology and future work which is following action for the non-fuel waste. Based on the result of above preliminary study, the feasibility of the research for the treatment of non-fuel waste would be evaluated and research plan would be established. In conclusion, the treatment of non-fuel waste stored in the spent fuel pool of domestic PWR should be considered to prepare the decommissioning. KHNP-CRI will identify the quantity, the dimension and kinds of non-fuel waste in the SFP of domestic PWR. Also, the various nuclide analysis technology would be identified and the technology which should be developed would be defined through this preliminary study.

Thermal Analysis of Cylindrical Model of Waste Form Containing High Heat-generating Nuclides Using Analytical and Numerical Methods

Jong Kwang Lee,Jung-Hoon Choi,Byeonggwan Lee,Hwan-Seo Park 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1

The fabrication of waste forms with high thermal and structural stability is an essential technology for the safe management and disposal of radioactive wastes. In particular, the thermal characteristics of waste forms containing high heat-generating nuclides such as Cs and Sr can be used for the optimized design of the waste form to secure its thermal safety, and they also provide basic design data for the safe management of canisters, storage systems, and repositories. The Korea Atomic Energy Research Institute is actively developing processes and equipment for fabricating various types of high-level wastes into a stable glass or ceramic waste form. In previous research related to the thermal analysis of the waste form, a relatively simple analysis was performed by using the analytic solution of the one-dimensional steady-state heat conduction equation considering the decay heat properties of the waste. As a specific application study, the optimized diameter of the cylindrical glass waste form was proposed by evaluating the centerline temperature of the waste form. In this study, we extended previous research by introducing a more complicated model, and the main results are summarized as follows. First, an analytical solution was derived by applying the temperaturedependent thermal conductivity expressed in the general form of polynomial function to the onedimensional heat conduction problem previously studied. Second, the two-dimensional axisymmetric steady-state heat conduction problem with a more realistic cylinder model with finite length was modeled and solved by using the finite element method via Matlab’s PDE (partial differential equation) toolbox. Third, thermal analysis was performed on the SrTiO3 waste form, selected as a stable form of strontium nuclide, using the developed analytical and numerical methods. The differences in the temperature distribution and computation time were evaluated through a comparative study of both solutions. Although the problem considered in this study could easily be solved by using commercial CFD software such as ANSYS or SolidWorks, a code-based program was developed to facilitate parametric design study in conjunction with optimization algorithms. The analysis results could be used to evaluate the thermal stability of waste form and to optimize the shape and size of the waste form in consideration of the design constraints of storage systems or repositories.

환경 시료 중 신뢰도 검증을 위한 방사능 분석

강태우 ( Tae Woo Kang ),홍경애 ( Kyung Ae Hong ) 한국환경농학회 2007 한국환경농학회지 Vol.26 No.2

The objective of this research was to assess the reliability of data and to improve nuclear analytical techniques concerning the Domestic Radioactivity Intercomparison program for environmental radioactivity monitoring of Jeju from 1998 to 2006. Gross beta for filter papers and water samples was determined, and gamma nuclides for natural and artificial nuclides in soil and water samples were analyzed. The gross beta activity of all samples except for the water samples of 1998 and 1999 showed a good agreement within the confidence intervals. In gamma nuclides, 40K and 137Cs of soil samples and most nuclides in the water samples, with the exception of several nuclides, were evaluated to be reliable. Based on these results, it is considered that a reliable method for the analysis and monitoring of environmental radioactivity were established, which may play an important role in case of emergency radiation accident.

Investigation and Analysis of Cs-137/I-129 Nuclide Concentrations and MDA Data Characteristics for I-129 Nuclide in Radioactive Waste

Sejun Pak,Soyeon Min,Dae Soo Lee,Gwang Tae Joo,Seung Jae Han 한국방사성폐기물학회 2024 한국방사성폐기물학회 학술논문요약집 Vol.22 No.1