국내 최대 간척지인 새만금 방조제 내부에는 간척토지 28,300ha가 조성되어 있으며 이 중 30% 용지를 농업용지로 활용하는 계획이 수립되었다. 농업용지의 토지이용 계획이 당초 수도작 운영계획에서 전작 위주 토지이용 계획으로 변경되면서 논과 밭은 영농 및 용수체계가 다르므로 이에 대한 용수체계 및 유입 부하량 변화에 연구가 필요한 실정이다. 본 연구에서는 전작화에 따라 새만금 농업용지 특성에 부합하는 작부체계를 고려하여 수도작 대비 오염부하 배출량을 비교·분석하고자 하였다. 새만금 농생명용지 토지이용계획 및 이에 따른 작부·용수체계를 반영하기 위해 농업용지의 특성에 부합하는 합리적인 작부시나리오를 제안하였으며 이를 적용하여 최신 구축된 HSPF 모델링시스템을 적용하여 평가하였다. 주요 결과로는 전작운영 여건을 고려하여 새만금 지역의 기상조건에 적합한 작부체계 시나리오를 고안하여 채소작물, 사료작물, 식량작물로 구성된 이모작 작부체계 를 제안하였다. 또한, 제안된 작부체계를 반영하여 새만금 농생명 용지 전작화에 따른 수도작 대비 오염부하 배출량을 비교한 결과 BOD와 TN 농도는 증가하지만 배출부하량은 감소하고, TP의 경우 농도와 부하량 모두 증가하는 것으로 나타났다. 본 연구의 결과는 새만금 농생명용지 토지이용 구상에 따른 오염부하량 산정을 통한 새만금호 수질환경문제와 경제성을 고려한 합리적이고 지속가능한 토지이용계획을 수립할 수 있으며, 새만금호 수질보전을 위한 유역관리대책의 정밀한 평가 보완필요성 검토를 위한 기초자료로 활용될 수 있을 것으로 기대된다.

Kori Unit 1 was permanently shut down in 2017 and is currently being prepared for decommissioning. Decommissioning waste generated during the decommissioning of a nuclear power plant has the characteristic of being generated in large quantities over a short period. Therefore, if proper management is not carried out, abnormal situations (i.e., unauthorized disposal, diversion, etc.) may occur. According to IAEA General Safety Report Part 6, radioactive waste shall be managed for all waste streams in decommissioning. This means ensuring that all waste streams are managed by the recorded inventory of all decommissioning waste and verifying that the recorded inventory is reasonable. The radioactive waste management has been managed in units such as mass and radioactivity. However, in the case of decommissioning waste, the amount is very large, so management by radioactivity is expected to have limitations. Therefore, in this study, a simple test was conducted to verify the decommissioning waste generated by a hypothetical scenario by mass. In this study, establish a scenario assuming various flows of decommissioning waste expected to be generated and calculate the expected inventory of decommissioning waste using Microsoft Excel. Specifically, using “Material Unaccounted For” (MUF), a material balance equation in IAEA Services Series 15, Nuclear Material Accounting Handbook, the error inventory was calculated as the difference between the physical inventory of decommissioning waste in the area and the ending inventory. We propose a simple test scenario to verify the flow of decommissioning waste by verifying that the error inventory reasonably matches the set allowable error. This study aims to verify the inventory of decommissioning waste using the material balance methodology used for nuclear material accounting. It is expected that the safety and reliability of the nuclear power plant decommissioning process can be secured by verifying that the total inventory of equipment before decommissioning and the inventory of remaining equipment and decommissioning waste after decommissioning are reasonably consistent.

The post-closure safety assessment of a repository is typically conducted over an extensive timescale from ten thousand to a million years. Considering that biosphere ecosystems may undergo significant changes over such lengthy periods, it is essential to incorporate the long-term evolution of the biosphere into the safety assessment. Climate change and landscape development are identified as critical drivers with the potential to impact the hydrogeological and hydrogeochemical characteristics of the biosphere. These changes can subsequently alter the migration patterns of radionuclides through the biosphere and influence human exposure doses. Therefore, this study formulates scenarios within the context of long-term biosphere evolution. We examine biosphere assessment processes employed in other countries and conduct a comparative study on scenario conditions. For example, biosphere assessment in Finland has identified sea-level changes and land-use alterations as significant factors in the long-term evolution of the biosphere. These factors are linked to Features, Events, and Processes (FEPs) associated with climate change and human activities. Sea-level changes are related to FEPs regarding climate change, land uplift, and shoreline displacement, while land-use changes are based on human activity-related FEPs (e.g., crop type, livestock and forest management, well construction, and demographics). Based on the literature review, this study has configured long-term evolution scenarios for the safety assessment of a deep geological repository for spent fuels.

According to the analysis of the Korean Radioactive Waste Society, saturation of nuclear power plant temporary storage is expected sequentially from 2031, and accordingly, the need for highlevel radioactive waste disposal facilities has emerged. In order to establish a repository for high-level radioactive waste, the performance and safety analysis of the repository must be conducted in compliance with regulatory requirements. For safety analysis, it needs a collection of arguments and evidence. and IAEA defined it as ‘Safety case’. The Systematic method, which derives scenarios by systematizing and combining possible phenomena around the repository, is widely used for developing Safety case. Systematic methods make use of the concept of Features, Events and Processes (FEP). FEP identifies features that affect repository performance, events that can affect a short period of time, and processes that can have an impact over a long period of time. Many countries, such as Finland, Sweden, Japan, United States, etc., are in process of licensing disposal facilities by using ‘Safety case’. And they then develop their own project-specified FEP lists and employ them for performing safety assessments. However, the systematic procedure for generating scenarios for safety evaluation is not clearly defined. According to the International Atomic Energy Agency (IAEA) Safety Standards Series (SSG- 23), the bottom-up method is an approach for conducting safety analysis using Features, Events, and Processes (FEPs). However, the process of how each FEP is utilized to establish a scenario for safety evaluation remains unclear. Additionally, there exists not only a bottom-up approach for generating scenarios using FEPs, but also a hybrid scenario development method that incorporates a top-down approach based on safety functions. Each country address scenario derivation in accordance with the adopted hybrid method. Nevertheless, a challenge arises in its application due to discrepancies between their approach and the hybrid approach specific which we are going on. Hence, this study introduces the FEP integration methodology for generating scenarios based on the hybrid scenario development method using the FEP list.

Nuclear power generation is expected to be enlarged for domestic electricity supply based on the 10th Basic Plan of Long-Term Electricity Supply and Demand. However, the issues on the disposal of spent nuclear fuel or high-level radioactive waste has not been solved. KBS-3 concept of the deep geological disposal and pyroprocessing has been investigated as options for disposal and treatment way of spent nuclear fuel. In other way, the radionuclide management process with 6 scenarios are devised combining chlorination treatment and alternative disposal methods for the efficient disposal of spent nuclear fuel. Various scenarios will be considered and comprehensively optimized by evaluation on many aspects, such as waste quantity, radiotoxicity, economy and so on. Level 0 to 4 were identified with the specialized nuclide groups: Level 0 (NFBC, Hull), Level 1 (Long-lived, volatile nuclides), Level 2 (High heat emitting nuclides), Level 3 (TRU/RE), Level 4 (U). The 6 options (Op.1 to 6) were proposed with the differences between scenarios, for examples, phase types of wastes, the isolated nuclide groups, chlorination process sequences. Op.1 adopts Level 0 and 1 to separate I, Tc, Se, C, Cs nuclides which are major concerns for long-term disposal through heat treatment. The rest of spent nuclear fuel will be disposed as oxide form itself. Op.2 contains Sr separation process using chlorination by MgCl2 and precipitation by K2CO3to alleviate the burden of heat after heat treatment process. U/TRU/RE will be remained and disposed in oxide form. Op.3 is set to pyroprocessing as reference method, but residual TRU/RE chlroides after electrorefining will be recovered as precipitates by K3PO4. Op.4 introduces NH4Cl to chlorinate TRU/RE from oxides after Op.2 applied and precipitates them. TRU/RE/Sr will be simultaneously chlorinated by NH4Cl without MgCl2 in Op.5. Then, chlorinated Sr and TRU/RE groups will be separated by post-chlorination process for disposal. But, chlorinated Sr and TRU/RE are designed not to be divided in disposal steps in Op.6. In this study, the mass flow analysis of radionuclide management process scenarios with updated process variables are performed. The amount and composition of wastes by types will be addressed in detail.

Understanding the dispersion of xenon isotopes following a nuclear test is critical for global security and falls within the remit of both the Comprehensive Nuclear-Test-Ban Treaty (CTBT) and the International Noble Gas Experiment (INGE). This paper aims to show if it is possible to discriminate the source of xenon releases based on the atmospheric dispersion of xenon isotopes using HYSPLIT. Using ORIGEN and SERPENT simulations, four released scenarios are defined with four different fractionation times (i.e., 1 hour, 1 day, 10 days, and 30 days) after a 1kt TNT equivalent 235U explosion event. These time-delayed release scenarios were selected to certify the possibility of mis-determining xenon release source. We use the Lagrangian dispersion model for atmospheric dispersion to predict the concentration distribution of xenon isotopes under each scenario. The model allows us to better understand how these isotopes would distribute over time and space, offering valuable data for real-world detection efforts. To our knowledge, there have been no researches on the analysis of xenon isotopic ratios considering atmospheric dispersion. In this work, we focused on the atmospheric dispersion using HYSPLIT to characterize the xenon isotopic ratios from nuclear tests. In addition, we compared the xenon isotopic ratios obtained from the atmospheric dispersion with those from ORIGEN calculations, which would be helpful to discriminate the source of the xenon releases.

In order to solve the rapidly increasing domestic delivery volume and various problems in the recent metropolitan area, domestic researchers are conducting research on the development of “Urban Logistics System Using Underground Space” using existing urban railway facilities in the city. Safety analysis and scenario analysis should be performed for the safe system design of the new concept logistics system, but the scenario analysis techniques performed in previous studies so far do not have standards and are defined differently depending on the domain, subject, or purpose. In addition, it is necessary to improve the difficulty of clearly defining the control structure and the omission of UCA in the existing STPA safety analysis. In this study, an improved scenario table is proposed for the AGV horizontal transport device, which is a key equipment of an urban logistics system using underground space, and a process model is proposed by linking systematic STPA safety analysis and scenario analysis, and UCA and Control Structure Guidelines are provided to create a safety analysis.

Social welfare facilities are used by a wide range of local residents, including vulnerable populations such as the elderly, children, and people with disabilities. During emergencies like fires, confusion can arise as these individuals try to evacuate. Evacuation simulation results have shown that utilizing evacuation systems based on specific evacuation scenarios can significantly decrease the time required for evacuation compared to general evacuation procedures. By anticipating potential fires based on changes in social and facility environments, appropriate evacuation scenarios can be developed and applied to evacuation systems, thus contributing to the safety and security of individuals during emergencies. In conclusion, for social welfare facilities that serve a large number of people, it is necessary to expand the focus on performance-based design depending on the size of the facility, and to continuously develop and train for appropriate evacuation scenarios that align with changing facility environments.

RUCAS (Recycling-Underlying Computational Dose Assessment System), a dose assessment program based on the RESRAD-RECYCLE framework, is designed to evaluate dose for recycling scenarios of radioactive waste in metals and concrete. To confirm the validity of the recycling scenarios provided by RUCAS, comparative evaluations will be conducted with RESRAD-RECYCLE for metal radioactive waste recycling scenarios and with MicroShield® for concrete radioactive waste recycling scenarios. In the evaluation of metal recycling scenarios without shielding, RUCAS showed similar results when compared to both MicroShield® and RESRAD-RECYCLE. This validates the function of dose assessments using RUCAS for metal recycling scenarios. However, when shielding was present, RUCAS produced results that were comparable to MicroShield®, but differed from those of RESRAD-RECYCLE. The underestimation of dose values up to 1.66E+08 times difference by RESRAD-RECYCLE could potentially decrease reliability and safety in evaluated doses, further emphasizing the importance of RUCAS. Because validation is also necessary for the expanded calculation capabilities resulting from methodological changes of RUCAS (i.e., various radiation source geometries), based on prior validations, it was determined that additional validations are required for different radiation source materials and shielding conditions. In case where the radiation source and shielding materials were identical, RUCAS and MicroShield® produced similar results according to both the Kalos et al. (1974) and Lin and Jiang (1996) methodologies. This demonstrates that the that differences in methodology are inconsequential when considering the same source and shielding materials. However, when the atomic number of the radiation source materials was larger than that of shielding material (HZ-LZ condition), RUCAS obtained results similar to MicroShield® only for the Kalos et al. (1974) methodology. While Lin and Jiang (1996) methodology yield higher results than MicroShield®. Lastly, in case where the atomic number of the radiation source material was smaller than that of the shielding material (LZ-HZ condition,) both methodologies yielded results comparable to MicroShield®. In conclusion, the validity of RUCAS’s shielding calculations has been verified, confirming improvements in dose assessment compared to RESRAD-RECYCLE. Additionally, we observed that shielding effectiveness calculations differ depending on the methodology of build-up effect. If the validity of these methodologies is confirmed, it is expected that selecting the most advantageous methodology for each condition will enable more rational dose assessments. Consequently, in future research, we plan to evaluate the validity of Lin and Jiang (1996) methodology using particle transport codes based on the Monte Carlo method, such as MCNP and Geant 4, rather than MicroShield®.

Two sets of analyses for the cases of groundwater release to well and sea ecosystems were conducted for the environmental impact assessment of high-level radioactive waste disposal facilities. After obtaining the respective BDCF (Biosphere Dose Conversion Factor) results for the scenarios of well-farming and marine water fishing using different biosphere assessment conceptual models implemented in ECOLEGO, they were compared each other. The purposes of these analyses are to identify reference generic biosphere conceptual models and to get insight on model uncertainty. In this study, the endpoint used for the comparison of the ECOLEGO biosphere models was the socalled Biosphere Dose Conversion Factor (BDCF), which is defined as the maximum value of the total dose to the exposed group, in Sv/yr, resulting from a continuous unit release of 1 Bq/yr during the whole simulation time either to the well compartment (BDCF_Well) or to the marine water compartment (BDCF_Sea). The radionuclides considered in the comparison were Cs-137, I-129, Nb-94, Ni-59, Ni- 63, Sr-90 and Tc-99. The conceptual models used in the biosphere assessment of the releases to a well are based on models that have been used by the DOE (simple-soil model) and SKB (complex-soil model) in safety assessments of radioactive waste repositories, respectively. Difference between two conceptual models used in the assessment of the releases to a sea is the number of compartments representing the sea; i.e., one model represents the sea with one compartment for the water and one for the sediment (singlecompartment model), whereas the alternative model uses two compartments for the water and the sediments: one for the inner coast and one for the outer coast (double-compartment model). The results of the BDCF_Well to a farmer obtained with the DOE and SKB models are shown to be very close to each other. Despite the differences in conceptual models and parameters, the results are within a maximum difference of a factor of 4. The results from the SKB model were higher for all radionuclides. The values of the BDCF_Sea obtained with the single- and double-compartment models are shown to be larger differences with a maximum order of 2. For all studied radionuclides, the double-compartment model produces higher BDCFs than does the single-compartment model. The differences would be due to activity concentrations in both water and sediments. Since the hydrodynamic behavior assumed for flow in the sea could significantly influence the dilution volumes and hence the concentrations, it is found that site-specific investigations are necessary to establish an appropriate marine biosphere conceptual model.

To obtain a license for a deep geological disposal repository for spent nuclear fuel, it is necessary to perform a safety assessment that quantifies the radiological impact on the environment and humans. One of the key steps in the safety assessment of a deep geological repository is the development of scenarios that describe how the repository evolves over the performance period and how events and processes affect performance. In the field of scenario development, demonstrating comprehensiveness is critical, which describes whether all factors that are expected to have a significant impact on the repository's performance have been considered. Mathematical proof of this is impossible. However, If the scenario development process is logical and systematic, it can support the claim that the scenario is comprehensive. Three primary approaches are being considered for scenario development: ‘Bottomup’, ‘Top-down’, and ‘Hybrid’. Hybrid approach provides a more systematic and structured process by considering both the FEPs (Features, Events, Processes) and safety functions utilized in the bottomup and top-down approaches. Many countries that develop recent scenarios prefer demonstrating scenario comprehensiveness using a hybrid approach. In this study, a systematic and structured scenario development process of a hybrid approach was formulated. Based on this, sub-scenarios were extracted that describe the phenomena occurring in the repository over the performance period, categorized by period. By integrating and screening the extracted sub-scenarios, a scenario describing the phenomena occurring over the entire period of disposal was developed.

In the wake of the Fukushima NPP accident, research on the safety evaluation of spent fuel storage facilities for natural disasters such as earthquakes and tsunamis has been continuously conducted, but research on the protection integrity of spent fuel storage facilities is insufficient in terms of physical protection. In this study, accident scenarios that may occur structurally and thermally for spent fuel storage facilities were investigated and safety assessment cases for such scenarios were analyzed. Major domestic and international institutions and research institutes such as IAEA, NEA, and NRC provide 13 accident scenario types for Spent Fuel Pool, including loss-of-coolant accidents, aircraft collisions, fires, earthquakes. And 10 accident scenario types for Dry Storage Cask System, including transportation cask drop accidents, aircraft collisions, earthquakes. In the case of Spent Fuel Pool, the impact of the cooling function loss accident scenario was mainly evaluated through empirical experiments, and simulations were performed on the dropping of spent nuclear fuel assembly using simulation codes such as ABAQUS. For Dry Storage Cask System, accident scenarios involving structural behavior, such as degradation and fracture, and experimental and structural accident analyses were performed for storage cask drop and aircraft collision accidents. To evaluate the safety of storage container drop accidents, an empirical test on the container was conducted and the simulation was conducted using the limited element analysis software. Among the accident scenarios for spent fuel storage facilities, aircraft and missile collisions, fires, and explosions are representative accidents that can be caused by malicious external threats. In terms of physical protection, it is necessary to analyze various accident scenarios that may occur due to malicious external threats. Additionally, through the analysis of design basis threats and the protection level of nuclear facilities, it is necessary to derive the probability of aircraft and missile collision and the threat success probability of fire and explosion, and to perform protection integrity evaluation studies, such as for the walls and structures, for spent fuel storage facilities considering safety evaluation methods when a terrorist attack occurs with the derived probability.

A person who performs or plans to conduct a physical protection inspection as stipulated by the law, the act on physical protection and radiological emergency, should obtain an inspector’s ID card certified and authorized by Nuclear Safety and Security Commission Order No.137 (referred to as Order 137). In addition, according to Order 137, KINAC has been operating some training courses for those with the inspector’s ID card or intending to acquire it. Also, strenuous efforts have been put to incrementally elevate their inspection related expertise. Since Republic of Korea has to import uranium enriched less than 20% in order to manufacture fuels of nuclear reactors in domestic and abroad, the physical protection for categorization III nuclear material in transit is significantly important along with an increase in transport. The expertise of inspectors should be constantly needed to strengthen as the increase in transport leads to an increase in inspection of nuclear material in transit. We have suggested a special way to improve the inspector’s capacities through Virtual Reality technology (VR). A 3-Dimensional virtual space was designed and developed using a 3-axis simulator and VR equipment for practical training. HP’s Reverb G2 product, which was developed in collaboration with VALVE Corporation and MicroSoft, was used as VR equipment, and the 3-axis motion simulator was developed by M-line STUDIO corp. in Korea for the purpose of realizing virtual reality. The training scenarios of transport inspection consist of three parts: preparation at the shipping point, transport in route including stops and handover at the receiving point. At the departure point, scenario of the transport preparation is composed with the contents of checking the transport-related documents which should be carried by shipper and/or carrier during transport and confirming who the shipper and/or carrier is. Second, scenario is designed for inspector to experience how carrier and/or shipper protect the nuclear material during transport or stops for rests or contingency and how they communicate with each other during transport. Lastly, scenario is developed focusing on key check items during handover of responsibilities to the facility operator at the destination. Those training scenarios can be adopted to strengthen the capabilities of those with inspector’s ID card of physical protection in accordance with Order 137 and to help new inspectors acquire inspectionrelated expertise. In addition, they can be used for domestic education to promote understanding of nuclear security, or may be used for education for people overseas for the purpose of export of nuclear facilities.

According to the “Law on protection and response measures for nuclear facilities and radiation”, Nuclear Power Plant (NPP) licensees should conduct periodic exercises based on hypothetical cyberattack scenarios, and there is a need to select significant and probable ones in a systematic manner. Since cyber-attacks are carried out intentionally, it is difficult to statistically specify the sequences, and it is not easy to systematically establish exercise scenarios because existing engineering safety facilities can be forcibly disabled. To deal with the above situation, this paper suggests a procedure using the Probabilistic Safety Assessment (PSA) model to develop a cybersecurity exercise scenario. The process for creating cyber security exercise scenarios consists of (i) selecting cyber-attack-causing initiating events, (ii) identifying digital systems, (iii) assigning cyber-attack vectors to a digital system, (iv) determining and adding type for operator’s response, (v) modifying a baseline PSA model, and (vi) extracting top-ranked minimal cut sets, and (vii) selecting a representative scenario. This procedure is described in detail through a case study, an expected cyber-attack scenario General Transient-Anticipated Transient Without Scram (GTRN-ATWS). It refers to an accident scenario for ATWS induced by GTRN. Since ATWS is targeted for cyber training in some NPPs, and GTRN is one of the most common accidents occurring in NPPs, GTRN-ATWS was chosen as an example. As for the cyber-attack vector, portable media and mobile devices were selected as examples based on expert judgment. In this paper, only brief examples of GTRN-ATWS events have been presented, but future studies will be conducted on an analysis of all initiating events in which cyber-attacks can occur.