우리나라의 스쿠버다이빙 중 입·출수시 선박의 스크류에 신체를 부딪쳐 사망하는 사고가 매년 반복적으로 발생하고 있다. 이에 우리나라는 관련 사고 저감과 안전관리방안 마련에 대한 필요성이 대두됨에 따라 스크류망의 설치와 관리에 관한 규정을 마련하였으나, 국내 법규상 명확한 제작 규정이 없어 설치기준이 모호하고, 점검항목이 외관·고정상태에 국한되어 있어 다이버의 스크류 사고의 안전이 담보되지 않은 상황이다. 이에 따라 본 연구는 현재 스크류망의 설치 및 점검에 대한 법적 요건을 구체화하기 위한 연구를 진행하였다. 이를 위하여 국내·외 스크류망의 현황과 안전기준과 관련된 실태와 법적 요소를 검토를 통해 문제점을 진단하였고, 이를 바탕으로 그에 대한 구체적인 개선점을 발굴하여 설치와 점검을 위한 제도개선(안)을 제시하였다. 스크류망의 설치기준은 크기와 재질에 관한 내용을 명 확히 제시하였으며, 점검기준은 외관, 고정상태, 재질상태 등을 제안하였고, 이를 판단하기 위한 명확한 지표 등을 제안하였다.

Highly radioactive waste is solidified to restrict leaching, retain its shape, and maintain its structural stability to prevent it from affecting humans and the environment as much as possible. This operation should be performed consistently regardless of whether the waste is homogeneous or heterogeneous. However, currently, there are no specific performance requirements for heterogeneous waste in Korea. This study reviewed domestic research results and the status of overseas applications, and proposed immobilization requirements for heterogeneous waste to be applied in Korea. IAEA safety standards, domestic laws, and waste acceptance criteria were reviewed. The status of heterogeneous waste immobilization in countries such as the United States, France, and Spain was reviewed. Most countries treat heterogeneous waste by encasing it in concrete, and impose immobilization requirements on this concrete. Based on these data, safety standards for the thickness, compressive strength, and diffusion limit of this concrete material were proposed as immobilization requirements for heterogeneous waste disposal in Korea. Quantitative values for the above requirements need to be derived through quantitative assessments based on the characteristics of domestic heterogeneous waste and disposal facilities.

해양수산부는 2016년부터 2020년까지 국제사회의 이내비게이션 도입에 선제적으로 대응하고 어선 등 소형선박의 해사안전 증 진을 위해 “초고속 해상무선통신망(LTE-M)” 구축을 포함한 한국형 이내비게이션 구축사업을 추진하였으나, 초고속 해상무선통신망의 활용 관점에서 특정 목적에 한정하는 등의 한계점이 식별되었다. 이에 따라 통신망의 활용성 증대를 위해 사용자를 대상으로 설문 조사 및 인터뷰를 수행한 결과, 망 활용의 범위 확대, 망 활용 대상 확장, 망 활용 방식 다각화, 그리고 규제 완화 측면에서의 법·제도적 개선 사항을 확인할 수 있었다. 본 연구에서 도출한 사용자 요구사항을 기반으로 하여 향후 관련 법제 정비방안에 기여할 수 있을 것으로 기 대한다.

Korea Atomic Energy Research Institute (“KAERI”) has been developing pyroprocess technology for the sustainable use of nuclear energy and radioactive waste reduction, and is conducting design studies for a Pyroprocess Commercializing Research Facility (PCRF). High-level radioactive materials such as spent nuclear fuel, which are handled in the hot cell of the PCRF, physically change materials directly or cause chemical changes through ionization or excitation depending on the energy and types of radiation. Therefore, all facilities, including process equipment and remote handling equipment, installed into the hot cell must be evaluated for radiation hardness to be maintained in the radiological environmfent so that processes can proceed throughout the design life of the facility. In addition, as the maintenance paradigm has recently shifted from corrective maintenance to predictive maintenance, it is necessary to know in advance the condition of the equipment or facility in the radiological environment. In this study, an analysis of the radiation environment of the hot cell in the PCRF was conducted through source term, and the radiological dose impact was evaluated through the results of irradiation experiments of major components by reference data. Then, the actual dose contribution was identified through dose assessment using the MCNP code based on the pyroprocess equipment, and the radiation hardness requirements for the facility and equipment in the hot cell were derived by the above results.

To construct and operate nuclear power plants (NPPs), it is mandatory to submit a radiation environmental impact assessment report in accordance with Article 10 and Article 20 of the Nuclear Safety Act. Additionally, in compliance with Article 136 of the Enforcement Regulations of the same law, KHNP (Korea Hydro & Nuclear Power) annually assesses radiation environmental effects and publishes the results for operating NPPs. Furthermore, since the legalization of emission plans submission in 2015, KHNP has been submitting emission plans for individual NPPs, starting with the Shin-Hanul 1 and 2 units in 2018. These emission plans specify the emission quantities that meet the dose criteria specified by the Nuclear Safety and Security Commission. Before 2002, KHNP used programs developed in the United States, such as GASPAR and LADTAP, for nearby radiation environmental impact assessments. Since then, KHNP has been using K-DOSE60, developed internally. K-DOSE60 incorporates environmental transport analysis models in line with U.S. regulatory guidance Regulatory Guide 1.109 and dose assessment models reflecting ICRP-60 recommendations. K-DOSE60 is a stand-alone program installed on individual user PCs, making it difficult to manage comprehensively when program revisions are needed. Additionally, during the preparation of emission plans and the licensing phase, improvements to KDOSE60’ s dose assessment methodology were identified. Furthermore, in 2022, regulatory guidelines regarding resident dose assessments were revised, leading to additional improvement requirements. Currently, E-DOSE60, being developed by KHNP, is a network-based program allowing for integrated configuration management within the KHNP network. E-DOSE60 is expected to be developed while incorporating the identified improvements from K-DOSE60, in response to emission plan licensing and regulatory guideline revisions. Key improvements include revisions to dose assessment methodologies for H-13 and C-14 following IAEA TRS-472, expansion of dose assessment points, and changes in socio-environmental factors. Furthermore, data such as site meteorological information and releases of radioactive substances in liquid and gaseous forms can be linked through a network, reducing the potential for human errors caused by manual data entry. Ultimately, E-DOSE60 is expected to optimize resident exposure dose assessment and enhance public trust in NPP operation.

Currently, Korea is planning to use various equipments and technologies for cutting, decontamination, compression, solidification, and packaging at decommissioning site of Kori unit 1 and Wolsung unit 1. In particular, Korea is considering to apply new technologies like inorganic acid decontamination, spent resion treatment technology not only to localize various decommissioning technology, but to meet the limit of 14,500 drums of the decommissioning waste per unit. However, before the techniques applied to decommissioning, it is necessary to demonstrate the effectiveness and the safety of the techniques. Because unlike the industrial fields, the failure of the decommissioning technique in nuclear power plants can cause the spread or leakage of radioactive materials. In the「Regulation on Technical Standards for Nuclear Reactor Facilities, Etc.」is stated that the licensee shall apply proven technology to decommissioning and if the licensee apply new technology, he must provide resonable evidence and prove its safety. In accordance with this approach, Nuclear Safety and Security Commission (NSSC) Notice No. 2021-24 can be applied to the decommissioning technology as one of a technical standard related with the demonstration of it. And it states 9 kinds of elements related to the radioactive waste management facilites and components like the management of radioactive effluents, prevention of contamination and overflow by radioactive materials, etc. But, they are mixed with the radioactive material considerations and industrial considerations, and these considerations are usually for the facilities, not equipments or techniques. On the other hand, in the IAEA Safety Standards Series No. WS-G-2.1 Section 6.15 to 6.20, it recommended to evaluate 12 considerations for the decontamination technique and 7 considerations for the dismantling techniques. The Decommissioning Guide in Germany recommends to consider 3 conditions for radiation protection of decontamination techniques and 4 conditions for dismantling techniques. Therefore, it is necessary to compare the safety requirements or recommendations related to the demonstration of decommissioning technology with the other countries to check there is something to learn from it.

The nuclear power plant (NPP) decommissioning market is expected to expand not only domestically but also overseas. Proven technologies must be applied to decommission NPP. This is based on Article 41-2, Paragraph 2 of the domestic ‘Enforcement Decree Of The Nuclear Safety Act’. Proven technology refers to technology that has verified that it can be applied in the field through demonstration. In other words, in order to carry out NPP decommissioning, verification must be done. Demonstration refers to reducing technological uncertainty and directly verifying services implemented in the field. From a technology commercialization perspective, demonstration requires an approach based on technology readiness level (TRL) from a technology perspective and market readiness level (MRL) from a market perspective. The characteristics of demonstration also differ depending on the characteristics of each field. The demonstration in the field of nuclear energy is the demonstration of demand matching. This is to confirm the feasibility of the technology in the company’s required environment. In order to perform demonstration, a scenario must be derived by reflecting demonstration design considerations. After evaluating the derived scenario, an actual assessment is conducted using lab-based demonstration/virtual environment demonstration/real environment demonstration. What must be preceded by an actual assessment is confirming the consumer’s requirements. In this study, the necessary environment and requirements of consumer’s to perform NPP decommissioning were reviewed. The domestic decommissioning procedure requirements management system presents decommissioning procedures, potential worker accidents, and worker requirements. In the case of foreign countries, it was confirmed that complex wide need, cost benefit, risk reduction, waste generation, operation, reliability and maintenance (RAM) improvement and quantitative measures were evaluated for the technology to be demonstrated. Also the requirements for demonstrating decommissioning need to a detailed review of actual decommissioning cases. Therefore, a comparison must be made between the requirements based on actual NPP decommissioning cases and the requirements derived from this research process. Afterwards, the empirical research approach proposed by the Ministry of Trade, Industry and Energy was applied. The empirical research approach proposed by the Ministry of Trade, Industry and Energy is to secure a track record over a certain period of time and performance under conditions similar to the actual environment in the final research stage at the TRL level 6 to 8. Through this, it will be possible to confirm the suitability of overseas technology for domestic application.

In general, radioactive waste with high radioactivity is made into a solid form with performance such as leaching restriction, shape retention, and structural stability so that radioactive waste does not affect humans and the environment as much as possible. This should be applied equally to radioactive waste, whether homogeneous or heterogeneous. The requirements are stipulated in the “Low and Intermediate Level Radioactive Waste Delivery Regulations” notice of the Korea Nuclear Safety and Security Commission. On the other hand, the waste acceptance criteria for domestic disposal facilities require immobilization of heterogeneous waste when the activity concentration is above a certain level, but do not provide specific immobilization performance requirements. In this study, the immobilization requirements applied to heterogeneous radioactive waste in various overseas countries operating low and intermediate-level radioactive waste disposal facilities were studied. First, the IAEA’s safety standards for radioactive waste immobilization, domestic regulations, and disposal facility waste acceptance criteria were reviewed. Countries operating surface disposal facilities such as the United States, France, Spain, and Japan and countries operating underground disposal facilities such as Sweden and Finland were divided to review the current status of immobilization application to heterogeneous waste in overseas countries. When reviewing overseas cases, each country’s disposal methods, types of disposal waste, and waste treatment criteria were also reviewed. It was found that the immobilization requirements for heterogeneous radioactive waste vary depending on the disposal method and the type of barrier used to ensure disposal safety in each country. The common point is to surround heterogeneous radioactive waste within a concrete lining of a certain thickness, and to apply the thickness, compressive strength, and diffusion coefficient of the concrete lining as immobilization performance requirements. Through this study, the immobilization performance requirements for heterogeneous radioactive waste in various overseas countries that stably operate low- and intermediate-level radioactive waste were confirmed, which is expected to contribute to specifying the performance requirements for immobilization of heterogeneous radioactive waste in domestic disposal facilities.

According to the second high-level radioactive waste management national basic plan announced in December 2021, the reference geological disposal concept for spent nuclear fuels (SNF) in Korea followed the Finnish concept based on KBS-3 type. Also, the basic plan required consideration of the development of the technical alternatives. Accordingly, Korea Atomic Energy Research Institute is conducting analyses of various alternative disposal concepts for spent nuclear fuels and is in the final selection stage of an alternative disposal concept. 10 disposal concepts including reference concept were considered for analysis in terms of disposal efficiency and safety. They were reference concept, mined deep borehole matrix, sub-seabed disposal, deep borehole disposal, multi-level disposal, space disposal, sub-sea bed disposal, long-term storage, deep horizontal borehole disposal, and ice-sheet disposal. Among them, first 4 concepts, mined deep borehole matrix, sub-seabed disposal, deep borehole disposal, multi-level disposal, were selected as candidate alternative disposal concepts by the evaluation of qualitative items. And then, by the evaluation of quantitative and qualitative items with specialists, multi-level disposal concept was being selected as a final alternative disposal concept. Design basis and performance requirements for designing alternative disposal systems were laid in the previous stage. Based on this, the design strategy and main design requirements were derived, and the engineered barrier system of a high-efficiency disposal concept was preliminary designed accordingly. In addition, as an alternative disposal concept, performance targets and related requirements were established to ensure that the high-efficiency repository system and its engineered barrier system components, such as disposal containers, buffer bentonites, and backfill perform the safety functions. Items that qualitatively describe safety functions, performance goals, and related requirements at this stage and items whose quantitative values are changed according to future test results will be determined and updated in the process of finalizing and specifically designing an alternative highefficiency disposal system.

As the first step of risk management, risk identification is inevitable to understand the degree of work safety. However, the safety requirements can be divided in necessary factors and additional factors. Thus, we propose a safety requirements assessment model using Kano model derived from Herzberg’s two-factor theory, classifying safety requirements into ideal elements and must-be elements. The Kano model is usually applied to evaluate customer satisfaction divided into three major requirements in the fields of product development and marketing: attractive, must-be, and one-dimensional requirements. Among them, attractive requirement and must-be requirement are matched with ideal element and must-be element for safety requirement classification, respectively. The ideal element is defined as preventive safety elements to make systems more safe and the must-be element is referred to as fatal elements to be essentially eliminated in systems. Also, coefficients of safety measurement and safety prevention are developed to classify different class of safety requirements. The positioning map is finally visualized in terms of both coefficients to compare the different features. Consequently, the proposed model enables safety managers to make a decision between safety measurement and prevention.

In order to use nuclear energy stably, high level radioactive waste including spent nuclear fuel that is inevitably discharged from nuclear power plants after electricity generation must be managed safely and isolated from the human living area for a long period of time. In consideration of the accumulated amount of spent nuclear fuel anticipated according to the national policy for HLW management, the area required for the deep geological repository facility is expected to be very large. Therefore, it is essential to conduct various studies to optimize the area required for the disposal of spent nuclear fuel in cases where the nationally available land is extremely limited, such as in Korea. In this study, as part of such research, the strategies and the requirements for the preliminary design of a high efficiency repository concept of spent nuclear fuel were established. For PWR spent nuclear fuel, seven assemblies of spent nuclear fuel can be accommodated in a disposal canister, and high burnup of spent nuclear fuel was taken into consideration, and the source terms such as the amount and time of discharge and disposal were based on the 2nd national basic plan. By evaluating the characteristics, the amount of decay heat that can be accommodated in the disposal canister was optimized through the combination of seven assemblies of spent nuclear fuel. The cooling period of the radiation source for the safety assessment of the repository system was set at 55 years, and the operation of the repository would start from 2070 and then the disposal schedule would be conducted according to the disposal scenario based on the national basic plan. With these disposal strategies described above, the main requirements for setting up the conceptual design of the high efficiency repository system to be carried out in this study were described below. • A combination of seven spent nuclear fuels with high heat and spent nuclear fuels with low heat was loaded into a disposal canister, and the thermal limit per disposal canister was 1,600 W. • In order to maintain the long-term performance of the repository, the maximum temperature design limit in the buffer material was set to 130°C. • In the deep disposal environment, the safety factor [yield strength/maximum stress] required to maintain the structural stability of the disposal canister should be maintained at 2.0 or higher so that integrity of the canister can be maintained even under long-term hydrostatic pressure and buffer swelling pressure in the deep disposal environment. • The repository should have a maximum exposure dose of 10 mSv/yr or less, which is the legal limit in case of a single event such as an earthquake, and the risk level considering natural phenomena and human intrusion, which is less than the legal limit of 10-6/yr. These strategies and requirements can be used to develop the high-efficiency geological disposal concept for spent nuclear fuels as an alternative disposal concept.

Spent fuel from the Wolsong CANDU reactor has been stored in above-ground dry storage canisters. Wolsong concrete dry storage canisters (silos) are around 6 m high, 3 m in outside diameter, and have shielding comprised of around 1 m of concrete and 10 mm of steel liner. The storage configuration is such that a number of fuel bundles are placed inside a cylindrical steel container known as a Fuel Basket. The canisters hold up to 9 baskets each that are 304 L stainless steel, around 42” in diameter, 22” in height, and hold 60 fuel bundles each. The operating license for the dry storage canisters needs to be extended. It is desired to perform in-situ inspections of the fuel baskets to very their condition is suitable for retrieval (if necessary) and that the temperature within the fuel baskets is as predicted in the canister’s design basis. KHNP-CNL (Canadian Nuclear Lab.) has set-up the design requirements to perform the in-situ inspections in the dry storage canisters. This Design Requirements applies to the design of the dry storage canister inspection system.