For safeguarding dry storage facilities, a tomography system based on fast and thermal neutron detection was studied in Korea Institute of Nuclear Nonproliferation and Control. The study was conducted laboratory-scale experiments based on a custom built 1/10th scale model cask, He-4 gas scintillation detector array, and multiple 252Cf sources. A filtered back projection (FBP) was utilized to obtain the cask image via MATLAB. The Ram-Lak filter (ramp filter) was employed in FBP for improved the reconstructed image quality. The Ram-Lak filter is the increasing amplitude filter due to the increasing spatial frequency of the image. In spatial frequency, the frequency of brightness change in the low-frequency region is relatively low, and the frequency of brightness change in the highfrequency region is large. Thus, the high-frequency region in the neutron tomographic image is near the neutron sources and the cask, and the low-frequency region is outside of the cask and/or between the source and cask in the study. In order to apply the ramp filter, a Fourier transform is initially performed on projection data, and image reconstruction is performed with the corrected projection data. In this case, the filter is linearly changed. Therefore, a small filter value is applied at lower spatial frequencies to reduce the projection data, and a large filter value is applied at high spatial frequencies to reduce the projection data. The filter scale is a fraction of frequency amplitude, and the filter value applied to the projection data is determined according to the filter scale. This study was conducted for discussion of the image quality due to the effect of the filter scale used for image reconstruction of a neutron tomography system. The results show that in the experiment with one source, the source location was founded when we used the frequency scale of 0.5 and over. In the double or triple source experiment, the source locations and relative activities were found when we used a filter scale of 0.4 to 0.6. When the filter frequency scale of 0.7 to over, the relative activities are hard to know. It can be found that if the filter value is too large or too small, distortion may occur in the reconstruction results. Therefore, it seems reasonable to set a value between 0.4 and 0.6 as the scaling factor for the neutron tomography system. In the future, additional comparative studies will perform validation of the frequency scaling methods.

Since 2015, the IAEA has been implementing inspections by changing the paradigm from a nuclear facility level perspective to state level approach in order to enhance effectiveness/efficiency, and random inspections have become a major means of paradigm conversion. From this point of view, Safeguards Division analyzed the implementation status of random inspection conducted in ROK over the past five years by type, facility, and year, and predict the future implementation direction. A total of 106 random inspections have been implemented in ROK over the past five years (2017- 2021). When classified by random inspection type, 20 Short Notice Random Inspections (SNRI), 9 Short Notice Inspections (SNI), 34 Unannounced Inspections (UI), and 43 Random Interim Inspections (RII) were performed, of which RII accounted for the largest proportion with about 40.6%. IAEA conducts customized random inspections according to the characteristics of each nuclear facility. In the past five years, 34 cases of LWR, 20 cases of Bulk Handling Facility, 22 cases of CANDU, 30 cases of random inspections were implemented at R&D facilities, with the largest proportion of LWR. In connection with this, as a result of analyzing the current status of random inspections by year, 19 cases in 2017, 21 cases in 2018, 20 cases in 2019, 19 cases in 2020, and 27 cases in 2021 were implemented, especially in 2021, an increase of about 40% compared to the previous year. In conclusion, the IAEA is increasing the proportion of random inspections every year in line with the updated SLA, and this stance is expected to be maintained in the future. This can be seen as proof that randomized tests are showing an increasing trend even in the COVID-19 pandemic situation that has occurred since December 2019. and, In the case of LWR, it is expected that the form of random inspection may vary depending on the direction of IAEA’s SLA improvement. Through these analyses, it is expected that the direction of future safeguards policy can be established.

Even though it is emphasized to apply safeguards-by-design (SBD) concept in the early phase of the design of a new nuclear facilities, there is no clear guideline or tools for the practical SBD implementation. Generally known approach is trying to review whether there is any conflicts or shortcomings on a conceptual safeguards components in a design information. This study tries to build a systematic tools which can be easily applied to safeguards analysis. In evaluating the safeguards system or performance in a facility, it is essential to analyze the diversion path for nuclear materials. Diversion paths, however, can be either extremely simplified or complicated depending on the level of knowledge and purpose of specific person who do analyze in the field. In the context, this study discusses the applicability of an event tree and fault tree method to generating diversion paths systematically. The essential components constituting the diversion path were reviewed and the logical flow for systematically creating the diversion path was developed. The path generation algorithm based on the facility design components and logical flow as well as the initial information of the nuclear materials and material flows was test using event tree and fault tree analysis tools. The usage and limitation of the applicability of this two logic methods are discussed and idea to incorporate the logic algorithm into the practical program tools is suggested.The results will be used to develop a program module which can systematically generate diversion paths using the event tree and fault tree method.

Dry storage cask facilities are considered for temporary storage of spent nuclear fuels before their final disposal. According to relevant domestic laws and regulations, the integrity and gross defects of the PWR spent fuel must be inspected before they are transferred to the dry cask from a wet storage pool of a nuclear power plant. To meet nuclear safeguards requirements for a spent fuel transportation, the KINAC has been working to develop a simple and convenient Non-destructive Testing (NDT) equipment to verify the integrity and gross defects of the spent fuel assembly. This study was conducted in two processes. The first stage is to review the current NDT techniques conducted in the nuclear fuel manufacturing process. During the manufacturing process, the Ultrasonic testing (UT) and Eddy Current Testing (ECT) technique are used for detecting the cracks or foreign materials in a cladding of a fresh fuel. During an over-haul period after an end of one fuel cycle, the sipping test of the spent fuel is performed for detecting the failed fuel assemblies. If it is determined through the sipping test whether any fuel assembly contains a failed fuel rod, the failed fuel rod of lots of fuel rods in the assembly is found out using the UT instrument. The ECT is used for detecting the internal defects and oxide layer thickness of a fuel cladding. Because the UT and ECT are the wellknown technique and has already been employing for the spent fuel inspection, we adopted the UT and ECT technique for development of a new instrument for nuclear safeguards verification. The second stage is to design the UT and ECT equipment in consideration of nuclear safeguards activities in the spent fuel pool. For nuclear safeguards inspection, irradiated fuel or non-fuel items are distinguished. Thus, verification equipment newly designed using the UT and ECT should detect not only a failed rod, but also a false tube, or a false rod, or a different material from a cladding. New probe and signal processing methods are developed to achieve these goals. The design of UT and ECT probes are preferentially carried out according to technical requirements – the probe thickness including a damper material should be less than 1.0 mm - and the study on analyzing signal distortion caused from material difference will be conducted for development of the safeguards inspection equipment. Detailed results of our study will be discussed in this conference.

IAEA, in preparation for possible commercial operation of small modular reactors (SMR), is pursuing the early development of safeguards approaches for these reactors along with the publication of the safeguards technical report (STR) to make verification activities easier for inspectors. For this purpose, the IAEA, through the MSSP, is developing various approaches for the application of safeguards by design for SMRs in collaboration with five countries that along with the ROK includes Russia, China, and U.S. etc. In order to review the specific design information of the ROK’s only SMR facility, SMART, and to establish safeguards methods from the initial design stage, collaborations were made with the Korea Atomic Energy Research Institute, which researched and developed the SMART. As a result, the design information questionnaire (DIQ) and STR was created and sent to the IAEA. The DIQ is a collection of crucial questions regarding the application of safeguards to understand the characteristics of the reactor facility, and STR is a document referred to by IAEA inspectors during safeguards verification activities. The main contents of the STR consist of introduction, technological description of SMR, safeguards approach, conclusions and annexes. Through this study, it is anticipated to understand the technical requirements for safeguards implementation of SMRs in the design stage, and through the completion of the final report applying SBD with regards to the design of the SMART facility, it could be used as information material for future safeguard verification activities by IAEA inspectors.

In order to effectively and efficiently apply safeguards to new nuclear facilities, it is recommended to apply safeguards-by-design concept. In evaluating the safeguards in the early stage of the design of a facility, it is essential to analyze the diversion path for nuclear materials. This study suggests a simple method which can generate diversion paths. The essential components constituting the diversion path were reviewed and the logical flow for systematically creating the diversion path was developed. The path generation algorithm is based on this components and logical flow as well as the initial information of the nuclear materials and material flows in a planned facilities. The results will be used to develop a program module which can systematically generate diversion paths using the event tree and fault tree method.

For the peaceful use of nuclear energy, the international community has devoted itself to fulfilling its obligations under the Safeguards Agreement with IAEA. In this regard, uranium in a radioactive waste drum should be analyzed and reported in terms of mass and 235U enrichment. In order to characterize radioactive wastes, gamma spectroscopy techniques can be effectively applied. In the case of high-resolution gamma spectroscopy, because an HPGe detector can provide excellent energy resolution, it can be applied to analyze a mixture having a complicated isotopic composition. However, other substances such as wood, concrete, and ash are mixed in radioactive waste with various form factors; hence, the efficiency calibration is difficult. On the other hand, In Situ Object Counting System (ISOCS) has a capability of efficiency calibration without standard materials, making it possible to analyze complex radioactive wastes. In this study, the analysis procedure with the ISOCS was optimized for quantification of radioactive waste. To this end, a standard radioactive waste drum at KEPCO NF and low-level radioactive waste drums at Korea Radioactive Waste Agency (KORAD) were measured. The performance of the ISOCS was then evaluated by Monte Carlo simulations, Multi-Group Analysis for Uranium (MGAU) code, and destructive analysis. As a result, the ISOCS showed good performance in the quantification of uranium for a drum with the homogenized simple geometry and long measurement time. It is confirmed that the ISOCS gamma spectroscopy technique could be used for control and accountancy of nuclear materials contained in a radioactive waste drum.

Pyroprocessing is a promising technology for managing spent nuclear fuel. The nuclear material accounting of feed material is a challenging issue in safeguarding pyroprocessing facilities. The input material in pyroprocessing is in a solid-state, unlike the solution state in an input accountability tank used in conventional wet-type reprocessing. To reduce the uncertainty of the input material accounting, a double-stage homogenization process is proposed in considering the process throughput, remote controllability, and remote maintenance of an engineering-scale pyroprocessing facility. This study tests two types of mixing equipment in the proposed double-stage homogenization process using surrogate materials. The expected heterogeneity and accounting uncertainty of Pu are calculated based on the surrogate test results. The heterogeneity of Pu was 0.584% obtained from Pressurized Water Reactor (PWR) spent fuel of 59 WGd/tU when the relative standard deviation of the mass ratio, tested from the surrogate powder, is 1%. The uncertainty of the Pu accounting can be lower than 1% when the uncertainty of the spent fuel mass charged into the first mixers is 2%, and the uncertainty of the first sampling mass is 5%.