[en] In this study, the feasibility of implementing the ferrite treatment method into the LRWPS has been proved in both theoretical and experimental ways. Currently limited applications of the ferrite process into the LRWPS can be ascribed to the followings : 1) deficiency in knowledge about the reaction mechanisms, 2) absence of adequate model, and 3) difficulty in removing non-transition metal elements. First of all, the overall reaction mechanisms involved in the Co²⁺/FeO· Fe₂O₃ system were analyzed, and a new concept sorption model named ESCM were developed and proposed by extending the conventional SCM. The validity of the ESCM was verified by using a series of experimental data. As a result, two of the most important parameter values were determined as log^*K_Co2+ = -2.0 and log^*K_CoOH+ = -8.0, respectively. The optimal operating condition turned out to be above pH 8. The ion exchange of the Co²⁺ with the structural Fe²⁺ is dominant below pH 6, the surface complexations of Co²⁺ and CoOH⁺ play important parts in the range of pH 6 to 9, and the slight increase of the removal efficiency above pH 9 can be attributed to the surface or bulk precipitation of Co²⁺ into Co(OH)₂. In addition, the removal efficiency of the Co species are hardly affected by the disturbing factors such as coexistence of chelating agent or competing cation, and high ionic strength. By performing a set of experiments with varying the ferrite composition, it turned out that MnO·Fe₂O₃ has the highest selectivity toward Sr, which has been known to be scarcely removed by the conventional ferrite process. The validity of the ESCM was reassured by using the experimental data attained from the Sr²⁺/MnO·Fe₂O₃ system. The complexation constants log^*K_Sr2+ and log^*K_SrOH+ have the values of -0.5 and -13.0, respectively. The theoretical maximum Sr-removing capacity of MnO·Fe₂O₃ is about 0.28g·g^-1. However, it turned out that the optimal operating condition is above pH 9,and the decrease of the removal efficiency caused by the disturbing factors are relatively large compared to the Co²⁺/FeO·Fe₂O₃ system. In most of the ferrite treatment systems, the plots of logC_f vs. -log[H⁺] can be well-fitted with a quadratic equation. This relation can be used as a system-specific characteristic curve in order to predict the process efficiency of the large-scale ferrite process. It is anticipated that the pre-formed ferrite process can be effectively and directly adopted into pre-treatment stage of the conventional LRPWS or various alkaline (> pH 9) waste streams such as miscellaneous wastewater and floor drains. Furthermore, development of a continuous-type ferrite reactor, selection of Cs- and I-selective ferrite materials, and economical efficiency analysis over the whole process should be preceded to the full-scale application

[en] A real-time emergency dose assessment computer code called KEDA (KAIST NPP Emergency Dose Assessment) has been developed for the NPP severe accident. A new mathematical model which can calculate cloud shine has been developed and implemented in the code. KEDA considers the specific Korean situations(complex topography, orientals' thyroid metabolism, continuous washout, etc.), and provides functions of dose-monitoring and automatic decision-making. To verify the code results, KEDA has been compared with an NRC officially certified code, RASCAL, for eight hypertical accident scenarios. Through the comparison, KEDA has been proved to provide reasonable results. Qualitative sensitivity analysis also the been performed for potentially important six input parameters, and the trends of the dose v.s. down-wind distance curve have been analyzed comparing with the physical phenomena occurred in the real atmosphere. The source term and meteorological conditions are turned out to be the most important input parameters. KEDA also has been applied to simulate Kori site and a hyperthetical accident with semi-real meteorological data has been simulated and analyzed

[en] This is the result of the project 'Development of Regulatory Technology for Radioactive Waste', and provides the state-of-the-art report on the dry storage of spent nuclear fuel. Currently 28 sites in the US have been allowed to load irradiated fuel assemblies into dry storage systems. Twelve sites have site-specific licenses and the remained sixteen sites have general licensees. The design features and basic concepts of fifteen cask models granted for general license were analyzed. As standard safety factors for ensuring the safety of dry storage facilities, totally twenty one factors were derived as: (1) site characteristics, (2) major design requirements, (3) structural integrity, (4) materials, (5) decay heat, (6) fire protection, (7) shielding, (8) criticality, (9) confinement, (10) handling of spent fuel, (11) physical protection and security, (12) operating procedure, (13) test and maintenance, (14) radwaste management, (15) radiation protection, (16) accident analysis, (17) auxiliary and/or service systems, (18) technical specifications and limiting conditions for operation, (19) quality assurance, (20) decommissioning, (21) retrievability. Furthermore, a draft classification system for categorizing safety classes of structures, systems, and components important to safety, for both metal and concrete casks. Licensing procedures, requested licensing documents, and related regulatory guides of the US NRC were surveyed and analyzed. In addition similar case studies were performed for Germany and Japan. Totally nine technical issues including 'impact of temperature to long-term integrity' were derived. In addition, seventeen Interim Staff Guidances issued by the Spent Fuel Project Office were analyzed for gathering information on the United States' recent regulatory issues. In relation to BUC(Burn-up Credit) issue, national application practices and application levels of the BUC in twenty countries were analyzed for each step of spent fuel management such as transport, storage, reprocessing, and disposal. As a result of this term's research and development, at last, a series of technical issues and regulatory issues were proposed to be include in the next term's work scope

[en] Spent nuclear fuel (SNF) generated from domestic pressurized water reactors (PWRs) has been stored in onsite At-Reactor (AR) storage facilities. Due to the limited existing SNF storage capacity, however, onsite Away-From-Reactor (AFR) storage has been considered as an attractive alternative. In this regard, staffs in KINS (Korea Institute of Nuclear Safety) have investigated to establish a preliminary safety evaluation framework for a hypothetical AFR PWR-SNF storage facility. At first, a reference fuel storage system was assumed based upon public domain information on commercial PWR-SNF storage casks. And then, potential safety cases for the reference storage system were analyzed and standard evaluation procedures for major safety factors were established. Finally, preliminary evaluations were performed for a few representative scenarios which may affect the safety functions of the reference storage system

[en] The Korean nuclear industry has more than 10 years experience of releasing slightly contaminated radioactive wastes, which have been generated during operation of nuclear facilities and use of radioisotopes, from regulatory control. Domestic legislations and regulations adopted the concept of clearance in 1994, based upon internationally endorsed dose criteria for clearance. A single concentration limit was set forth for short-lived beta/gamma emitting radionuclides. For waste streams containing other radionuclides, the feasibility of release has been determined on a case-by-case basis. Korea Institute of Nuclear Safety (KINS) derived draft clearance levels for other major radionuclides and proposed draft guidance for clearance in 2002. Since then, the case-by-case regulatory reviews have been performed based on the draft guidance. This paper introduces Korean legislations, regulations, regulatory review process, the characteristics of cleared waste, and safety research for implementing the concept of clearance

[en] The first national radioactive waste management (RWM) policy and strategy was established in 1984, and then the RWM policy and strategy have been revised several times until recently. Though some of the objectives in the RWM policy and strategy have been attained, many of them have not been successfully implemented due to oppositions from residents and misconducts in communication with interested parties. However, no specific studies on the advantages and disadvantages of the Korean government's past policy framework for RWM have been openly reported. Accordingly, fundamental principles for establishing the RWM policy and strategy are reviewed, the past experiences in Korea are analyzed, and then the desirable future direction of the RWM policy framework is proposed in this paper

[en] This is a partial result of the research project 'Development of Regulatory Technology for Radioactive Waste', and provides safety verification methodology on the interim dry storage of spent nuclear fuel generated from CANDU reactors. The main objective of this report is to support safety reviewers to perform systematic and detailed safety verification of the CANDU SNUFI-D (CANDU Spent NUclear Fuel Interim storage - Dry type) to be proposed by domestic applicant(s) in the near future. With this regard, related safety factors were categorized into three parts (i.e. verification of radiological safety and non-radiological safety, and accident analysis), and then basic considerations and detailed verification methodology were provided for each part. In addition, preliminary case-studies performed based upon the proposed verification methodology were also provided, so as to maximize the applicability of this report. In order to make the methodology applicable to various types of storage systems, design-specific consideration was excluded as possible

[en] In order to evaluate the potential radiological risks for the reuse of the site after decommissioning of nuclear facilities, a mathematical model was developed and materialized into the Microsoft spreadsheets frame. A set of input parameter values was proposed, which is useful in the preliminary risk screening step before the detailed evaluation with the site-specific data. It appeared that the screening levels calculated by the present model was agreed with the derived concentration guideline limits resulted from RESRAD Ver.6.2 and the German dose criteria for releasing a nuclear site from regulatory control.

[en] A conventional single compartment model cannot simulate reasonably the migration phenomenon of contaminants through unsaturated zone, due to the intrinsic unrealistic assumption of the compartment model that contaminants entering a compartment are immediately and uniformly mixed. Although, a multi-compartments model, in which even physically identical layer is divided into multiple compartments, may be used for explaining the retardation of contaminant mass flux along with increasing number of compartments, its numerical modeling is usually time-consuming and appropriate analytical solutions have not been reported yet. In order to improve the conventional compartment models on contaminant migration through unsaturated zone, a series of analytical solutions for multi-compartments model were derived and a generalized constraint under which the results from multi-compartments model can be simply approximated by single compartment model was proposed. The simplified approximation method was verified by a simple numerical analysis on the constraint under hypothetical conditions. It was also proved that the influent contaminant transfer rate from the bulk unsaturated zone can be generally represented into a time-dependent nominal transfer rate rather than a constant. In addition, the nominal transfer rate turned out to be very sensitive to the contaminant transfer rate between compartments in unsaturated zone, but to be almost insensitive to the transfer rate from contaminated zone

[en] Recently, International Atomic Energy Agency and major leading countries in radioactive waste management tend to subdivide the categories of radioactive waste based upon risk-graded approach. In this context, the category of very low level waste has been newly introduced, or optimized management options for this kind of waste have been pursued in many countries. The application of engineered surface landfill type facilities dedicated to dispose of very low level waste has been gradually expanded, and it was analyzed that their design concept of isolation has been much advanced than those of the old fashioned surface trench-type disposal facilities for low and intermediate level waste, which were usually constructed in 1960's. In addition, the management options for very low level waste in major leading countries are varied depending upon and interfaced with the affecting factors such as: national framework for clearance, legal and practical availability of low and intermediate level waste repository and/or non-nuclear waste landfill, public acceptance toward alternative waste management options, and so forth. In this regard, it was concluded that optimized long-term management options for very low level waste in Korea should be also established in a timely manner through comprehensive review and discussions, in preparation of decommissioning of large nuclear facilities in the future, and be implemented in a systematic manner under the framework of national policy and management plan for radioactive waste management