원자력발전소(원전) 내부에 설치되어 있는 주요 기기는 원전의 안정적인 운영을 돕는 주요 2차 구조 물이다. 경주 지진, 포항 지진과 같은 강한 지진이 발생하였을 때, 원전 주요 기기의 손상은 원전의 안정한 정지에 문제를 초래할 수 있다. 따라서, 원전 주요 기기의 지진응답을 저감시키기 위한 연구가 필수적으로 요구된다. 이러한 배경 아래, 본 연구에서는 원전 주요 기기의 내진성능 향상을 위하여 동 흡진장치(Dynamic Absorber)를 활용하였다. 연구에서 사용된 동흡진장치는 스프링, 댐퍼, 및 질량체로 구성된다. 이러한 동흡진장치를 설계하기 위하여 기존에 제안된 방법론들을 활용하였으며, 각 방법론 들을 기반으로 설계된 동흡진장치의 지진응답 저감효과를 비교 및 분석하였다. 구체적으로, 진동대 시 험 결과를 바탕으로 유한요소 모델을 검증하였다. 또한, 이를 기반으로 기존 동흡진장치의 설계방법론 에 따른 원전 주요 기기의 지진응답 저감 효과를 비교 및 분석하였다. 결과적으로 각 방법론들은 원전 주요기기의 가속도, 변위, 응력 응답을 평균적으로 약 30% 정도 감소시키는 효과를 보였다.

As remote sensing measures, satellite imagery has played an essential role in verifying nuclear activities for decades. Starting with the first artificial satellite, Sputnik 1, in 1957, thousands of satellites are currently missioning in space. Since the 2000s, the level of detail in pixels of an image (spatial resolution) has been significantly improving, thereby identifying objects less than one meter, even tens of centimetres. The more things are identifiable, the wider regions become targets for observation. With the increasing number of satellites, computer vision technology is required to explore the applicability of algorithm-based automation. This paper aims to investigate the R&D publications worldwide from the 1990s to the present, which have tried to apply algorithms to verify any clandestine nuclear activities or detect anomalies at the site. The versatile open-source publications, including the IAEA, ESARDA, US-DOE national laboratories, and universities, are extensively reviewed from the perspective of nuclear nonproliferation (or counter-proliferation). Thus, target objects for applications are essentially located in nuclearrelated sites, and the source type of satellite sensors focuses on electro-optical images with high spatial resolution. The research trend over time by groups is discussed with limitations at the time in order to contemplate the role of algorithms in the field and to present recommendations on a way forward.

Any type of nuclear arms control or disarmament agreement requires some form of verification measure. Existing nuclear arms control treaties drew upon previous agreements such as the INF treaty, START, and IAEA nuclear safeguards inspections. However, previous treaties focused on limiting specific types of nuclear weapons and their delivery vehicles or reducing the total number of nuclear weapons rather than eliminating the nuclear enterprise as a whole. A potential nuclear disarmament verification treaty or agreement will depend on the geopolitical environment of the time as well as the national policies and priorities of each signatory state. Although research on the gradual reduction and eventual elimination of nuclear weapons is still ongoing, several states have cooperated to conduct experiments, exercises, and simulations on the procedures and technologies required for nuclear disarmament verification. Three of these efforts are the LETTERPRESS simulation conducted by the Quadrilateral Nuclear Verification Partnership (QUAD), NuDiVe Exercise conducted by the International Partnership for Nuclear Disarmament Verification (IPNDV), and the Menzingen experiment organized by the UNIDIR in partnership with the Swiss Armed Forces, Spiez Laboratory, Princeton University’s Program on Science and Global Security, and the Open Nuclear Network. These contain aspects for the development of a potential nuclear disarmament verification. The LETTERPRESS exercise conducted in 2017 tested potential activities and equipment inspectors might utilize in a nuclear weapon facility. The IPNDV NuDiVe exercises conducted in 2021 and 2022 tested the activities and equipment required for the verified dismantlement of a warhead within a dismantlement facility. Finally, the Menzingen experiment conducted in 2023 tested the practical procedures for the verification of a nuclear weapon’s absence at a storage site. This paper will analyze the three cases to offer considerations on the procedures and technologies future nuclear disarmament verification might include.

One of the most important factors in the delivery and acceptance requirements for dry storage of spent fuel is the burnup of spent fuel. Here, burnup has a unit of MWD/MTU and is used as a measure of how much nuclear fuel is depleted in a nuclear reactor. In addition, since it is one of the most basic characteristic information for the soundness evaluation of spent nuclear fuel, it is a required item not only by regulatory agencies but also by KORAD, the acquiring agency. The burnup of spent nuclear fuel is the burnup calculated through flux mapping using signals measured from in-reactor instruments during nuclear power plant operation (hereinafter: actual burnup) and the burnup calculated using the core design code (hereinafter: design burnup). In this paper, the design burnup of spent nuclear fuel discharged from OPR100 NPPs (Nuclear Power Plants) in Korea was recalculated to confirm the reliability of the actual burnup currently managed at the nuclear power plant. Basically, since spent nuclear fuel must maintain subcriticality under wet storage or dry storage, a burnup error of about 5% is considered as a conservative approach when evaluating the criticality safety of wet storage tanks and dry storage systems. Therefore, in this paper, we tried to verify whether the difference between actual burnup and design burnup for all spent nuclear fuel released from domestic OPR100 type light water reactor nuclear power plants is within 5%. As a result of the evaluation, the largest deviation between actual burnup and design burnup was about 1,457 MWD/MTU, and when converted into a percentage, it was about 3.3%. Therefore, it was confirmed that the actual burnup managed by OPR1000 NPPs in Korea has sufficient reliability. In the future, we plan to check the reliability of the performance burnup managed in WH NPPs, and some of them will be verified through measurement.

This study is about the production of radiation sources of simulated concrete and soil reference materials to verify the validity of the quality establishment and measurement of the detector (HPGe) of the radioactive soil and concrete waste classification system, which is being developed to quickly and accurately classify nuclear decommissioning waste. Specific activity of gamma nucleus among radioactive wastes is evaluated using gamma spectroscopy. At this time, in order to verify the validity and reliability of measuring equipment, it shall be a standardized substance of the same medium as nuclear decommissioning waste (chemical ingredients, particles, density, etc.) in order to correct the energy and efficiency of gamma nuclide analysis equipment. The CRM used for the detector’s energy correction used a 1 L Marinelli beaker standard correctional radiation source consisting of 10 radioactive isotopes. In order to correct efficiency, in accordance with the production and certification process of the Korea Standards and Research Institute, it has produced artificial simulated radioactive concrete similar to nuclear decommissioning waste (30% for cement, 60% for regulation and 10% for bentonite). The radioactive homogeneity of the simulated concrete reference materials was evaluated using dispersion analysis (ANOVA) in accordance with ISO Guide 35, while 137Cs and 60Co of concrete reference materials were able to obtain homogeneous measurements both in and between bottles. The self-absorption rate of the simulated concrete reference material was determined by the MCNP computer simulation measurement method, and the self-absorption correction coefficients of 137Cs and 60Co were assessed at 0.995 and 0.996, respectively, and the standard value for the radiation of the simulated concrete reference material was calculated on the weighted average of the measurements of 20 samples. The uncertainty about the reference value was calculated by combining measurement uncertainty (Type B evaluation), bottle to bottle standard deviation, and uncertainty within bottle by modifying the formula suggested in ISO Guide 35. The concentration of 137Cs and 60Co of reference materials was divided into high-speed measurement mode and precision measurement mode in consideration of the self-disposal standard. The reference value and uncertainty of expansion among reference materials for high-speed measurement mode were rated at 1,032.7 ± 64.0 Bq·kg−1and 1,083.7 Bq·kg−1, respectively. The standard value and expansion uncertainty for 137Cs and 60Co among reference materials for precision measurement mode were rated at 113.7 ± 10.0 Bq·kg−1 and 122.3 ± 10.3 Bq·kg−1, respectively.

The denuclearization of North Korea was unpredictable and resulted in radical changes. Despite the skepticism and disappointment surrounding denuclearization, it is important for certain verification technologies to establish what is technically possible or practically impossible, and how reliable these technical means are. This article presents the technical hurdles in nuclear verification by systematically categorizing them into issues of correctness and completeness. Moreover, it addresses the safety and security risks during the denuclearization process, including the radiological impact on humans, environmental effects, and the illegal transfer of material, information, and technologies.

최근 미-북간 관계 개선 등으로 인하여 북핵 문제가 핵 검증단계로 진전될 것이라는 예상이 나오고 있다. 이제부터 우리 정부는 북핵 프로그램의 성공적 검증·폐기를 위해, 북한의 신고 후 전개될 상황에 대비하여 철저히 준비해야 한다. 본 논문에서는 구소련과 이라크의 대량살상무기 검증·폐기 사례로부터 두 나라의 대량살상무기 검증·폐기 과정에서 발생한 문제점을 조사·분석하여, 북핵 검증·폐기 과정 시 발생할 수 있는 문제점을 파악하고 이를 방지하기 위한 정책적 고려사항을 도출하는데 목적을 두었다.

원전의 정상운전이나 해체시 발생될 수 있는 토양의 제염을 위한 토양제염장치를 개발하였으며 실증 실험을 수행하였다. 제염장치를 이용한 제염실험을 종합해본 결과 제염조건에 큰 상관없이 이상의 제염율을 얻을 수 있었다. 방사능 준위 및 토양입도에 의한 실험결과를 보면 낮은 방사능 농도 및 고입도의 제염율이 다소 높음을 알 수 있었다. 제염용액과 토양질량의 비에 따른 제염율은 제염제 부피를 두배로 높였을 경우 방사능 농도가 높은 경우에 큰 것으로 나타났다. 반복 제염은 의 다소 작은 입자에 더욱 효과적으로, 제염이 어려운 작은 입자의 반복제염시 방사능 저감 효과가 비교적 크게 나타났다. 본 오염토양 제염장치를 활용하면 원전에서 발생되는 오염토양의 방사능 농도를 줄일 뿐 아니라 처분양을 줄여 저장공간의 확보에 기여할 뿐만 아니라 향후 원전의 해체시에도 유용하게 활용될 수 있으리라 생각된다.