본 연구는 민간기업과는 다른 특성을 보이는 공공기관 HRD의 평가지표 도출이 필요하다는 문제의식에 서 출발하였으며, 공공기관 HRD 체계의 타당한 평가 준거 개발을 목적으로 한다. 이를 위해 CIPP 모형을 활용하여 공공부문 HRD체계와 과정 준거를 개발했던 연구사례를 분석하고, 국내 공공기관 인적자원개발 실무자 6인 대상의 심층 인터뷰를 실시하여 15개 평가영역, 41개 평가항목, 56개 평가지표로 구성된 평가 준거 초안을 도출하였다. 평가준거 초안에 대한 타당성을 확보하고자 학계, 실무, 컨설팅 전문가 12인을 대상으로 2차의 델파이 조사를 거쳐 평가 준거를 확정하였다. 최종적인 평가 준거는 요구분석, 기관장 리더 십, HRD 인력 전문성, 경영성과 등 15개 영역, 37개 항목, 45개 지표로 도출되었다. 본 연구는 공공기관의 양면적 성격을 반영한 공공기관 HRD 체계의 평가 준거를 제시했다는 데 의의가 있다. 본 연구는 다양한 공공기관의 사업 특성을 감안한 평가 준거 보완이 필요하다는 점을 제언하였다.

Tritium is a radioactive isotope of hydrogen with a half-life of about 12.3 years, and it is commonly found in the environment as a result of the production of Nuclear Power Plants. The World Health Organization (WHO) has established guidelines for the permissible levels of tritium in drinking water. The guideline value for tritium in drinking water is 10,000 Bq/L. It is important to note that the guideline value for tritium is not a legal limit, but rather a recommendation. National and local authorities may establish legal limits that are more restrictive than the WHO guideline value based on local conditions and risk assessments. The Australia and Finland have set a limit for tritium in drinking water at 76,103 Bq/L and 30,000 Bq/L respectively, which is more than three to seven times higher compare to guideline value of WHO. The United States Environmental Protection Agency (EPA) has set a maximum contaminant level (MCL) for tritium in drinking water at 20,000 picocuries per liter (pCi/L), which is equivalent to 740 Bq/L. The Health Canada has set a guideline value for tritium in drinking water at 7,000 Bq/L. Assuming drinking water corresponding to each tritium limit (or guideline value) for one year, the expected exposure dose is 0.01 mSv to 1 mSv. It means that the tritium in drinking water below the limits or guideline value does not pose a significant risk to human health.

The acceptance criteria for low and intermediate level radioactive waste disposal facilities in Korea to regulate that homogeneous waste, such as concentrated waste and spent resin, should be solidified. In addition, solidification requirements such as compressive strength and leaching test must be satisfied for the solidified radioactive waste solidified sample. It is necessary to develop technologies such as the development of a solidification process for radioactive waste to be solidified and the characteristics of a solidification support. Radioactive waste solidification methods include cement solidification, geopolymer solidification, and vitrification. In general, low-temperature solidification methods such as cement solidification and geopolymer solidification have the advantage of being inexpensive and having simple process equipment. As a high-temperature solidification method, there is typically a vitrification. Glass solidification is generally widely used as a stabilization method for liquid high-level waste, and when applied to low- and intermediate-level radioactive waste, the volume reduction effect due to melting of combustible waste can be obtained. In this study, the advantages and disadvantages of the solidification process technology for radioactive waste and the criteria for accepting the solidified material from domestic and foreign disposal facilities were analyzed.

Domestic NPPs had produced the paraffin-solidifying concentrate waste (PSCW) for nearly 20 years. At that time radioactive waste management policy of KHNP was to reduce the volume and to store safely in site. The PSCW has been identified not to meet the leaching index after introducing the treatment system. PSCW has to be treated to meet current waste acceptance criteria (WAC) for permanent disposal. PSCW consists of dried concentrate 75% and paraffin 25% of volume. When PSCW is separated into a dried concentrate and a paraffin by solubility, total volume separated is increased twice. Final disposal volume of dried concentrate can reach to several times when solidifying by cement even considering exemption. Application of polymer solidification technology is difficult because dried concentrate is hard to make form to pellet. When PSCW is packaged in High Integrity Container (HIC), volume of PSCW is equal to the volume before package. The packaging process of HIC is simple and is no necessary of large equipment. It is important to recognize that HIC was developed to replace solidification of waste. HIC has as design goal a minimum lifetime of 300 years under disposal environment. The HIC is designed to maintain its structural integrity over this period, to consider the corrosive and chemical effects of both the waste contents and the disposal environment, to have sufficient mechanical strength to withstand loads on the container and to be capable of meeting the requirements for a Type A transport Package. The Final waste form is required for facilitating handling and providing protection of personnel in relation to solidification, explosive decomposition, toxic gases, hazardous material, etc. Structural stability of final waste form is required also. Structural stability of the waste can be provided by the waste itself, solidifying or placing in HIC. Final waste form ensure that the waste does not structurally degrade and affect overall stability of the disposal site. The HIC package contained PSCW was reviewed from several points of view such as physicochemical, radiological and structural safety according to domestic WAC. The result of reviewing shows that it has not found any violation of WCP established for silo type disposal facility in Gyeongju city.

In order to reduce the area of the high-level radioactive waste (HLW) repository, a buffer material with high thermal conductivity is required. This is because if the thermal conductivity of the buffer material is high, the distance between the disposal tunnels and the deposition holes can be reduced. Sand, which is a natural material and has higher thermal conductivity than bentonite, is added to bentonite to develop an enhanced buffer material. For the sand-bentonite mixture, it is important which sand to use and how much to add because an enhanced buffer material should satisfy both hydraulic (H) and mechanical (M) performance criteria while improving thermal conductivity (T). In this study, we would like to show what type of sand and how much sand should be added to develop an enhanced buffer material by adding sand to Gyeongju bentonite, a representative bentonite in Korea. For this purpose, the thermal conductivity, hydraulic conductivity, and swelling pressure of the sand-Gyeongju bentonite mixture according to the sand addition rate were measured. It is more efficient to use silica sand with smaller particles than Jumunjin sand which is a representative sand in Korea as an additive for an enhanced buffer material than using the Jumunjin sand. In order for the sand-Gyeongju bentonite buffer material to satisfy both the hydraulic and mechanical performance criteria as a buffer material while increasing the thermal conductivity, it is judged that the optimum dry density is 1.7 g/cm3 at least and the optimum sand addition rate is 10% at most.

국내외 매년 지속적으로 증가하는 대기오염의 상당량이 선박에서 배출되고 있으며, 이러한 선박에서의 배출량을 감축하기 위 해 정부는 환경규제를 강화하고 해양대기환경 개선을 위해 「환경친화적 선박의 개발 및 보급 촉진에 관한 법률」을 제정 및 시행하여 친환경 에너지를 동력원으로 사용하는 환경친화적 선박을 보급하여 해양오염 저감 기술이 개발되도록 하고 있다. 환경친화적 선박의 기 술은 최근 순수 전기추진 내지는 전기복합 추진체계까지 진화하고 있으나, 선박의 안전한 운용을 위한 검사 및 설비기준 등 관련제도의 정비는 이를 못 따라가는 실정으로, 특히 국내 연안선박의 다수를 차지하는 어선에 대한 제도는 정비가 시급한 시점이다. 이번 연구를 통 해 어선에의 전기복합 추진체계 등의 적용을 위해 핵심기자재인 배터리, 전력변환장치 등을 정의하고 이에 필요한 안전기준을 진단, 마련 하여 친환경 어선의 도입과 보급을 위해 관련된 산업을 지원할 수 있도록 최소한의 안전기준안을 마련하고자 한다.

이 연구는 국가 위기관리체계 중 통합방위사태와 경계태세의 발령기준 을 개선하기 위해 분석하였다. 우리 정부와 군은 평시 북한의 침투 및 도발과 같은 상황에 대응하기 위해 경계태세 및 통합방위사태와 같은 국 가 위기관리체계를 지속 발전시켜 왔다. 그러나 북한의 도발 상황을 고 려하여 통합방위사태와 경계태세가 시기적절하게 선포 및 발령되어야 함 에도 불구하고 그렇지 못한 사례가 여럿 있다. 따라서 통합방위사태 선 포절차와 경계태세의 발령조건을 반드시 개선할 필요성이 있다. 이를 위 해 기존 선행연구 사례를 통해 공통적인 연구의 경향들을 분석하였고, 위와 같은 문제가 발생하는 원인을 관련 법령 분석을 통해 연구문제를 도출하였다. 또 과거 위기대응사례를 분석하여 연구문제에 대한 결론을 내림과 동시에 발전 방향을 제시하였다. 이를 통해 본 논제가 주는 함의 는 첫째, 북한의 태도가 급변하여 예측할 수 없는 군사적 도발을 하더라 도 유연하게 대응하고 신속하게 대처할 수 있는 위기관리태세를 갖추어 야 한다는 것이며 둘째, 이러한 적시적인 위기관리를 통해 정부는 우리 국민들의 안전을 수호하는 역할을 수행해야 한다는 점을 시사하고 있다.

PURPOSES : Unlike European standards, domestic performance assessment standards for truck mounted attenuators (TMAs) was first stipulated in 2014 using the NCHRP Report 350 of 1993 as the standard instead of the 2009 MASH of the United States. The purpose of this study is to present an improvement in the domestic performance evaluation criteria for TMAs.. METHODS : Considering the latest TMA performance evaluation standards in the U.S. and Europe, domestic performance evaluation criteria must improve stipulations related to impact speeds, impact conditions, impact cars, and support trucks. The performance change in the TMAs according to the variation in the impact speed, impact condition, impact vehicle, and support vehicle was investigated using finite element analysis (FEA). RESULTS : The TMA for an impact speed of 100 km/h showed a limit to the safety of the occupants of the collision vehicle and workers on the road for a collision speed of 120 km/h. The safety of the workers on the road was also not guaranteed for the collision of the remaining 73.8% of vehicles that exceeds the maximum impact car weight of 1,300 kg, the lower 26.2% of the total mass composition of domestic passenger cars. In addition, a TMA that satisfied only the conditions under which the vehicle was hit head-on to the center of the TMA did not reduce the risk of a secondary collision of the impact vehicle. Furthermore, the safety of workers on the road was not guaranteed when a travel distance of a support truck of 10 tons or more was applied to a work vehicle of less than 10 tons. CONCLUSIONS : To improve the safety of road traffic, a performance level corresponding to an impact speed of 120 km/h was added to the domestic TMA performance evaluation standard, and the eccentricity and oblique collision conditions were mandatory. Furthermore, the maximum impact vehicle weight of 1300 kg was raised to 2000 kg, and the test requester had to present support trucks of lower and upper weights such that TMA mounting trucks of various weights could be used.

본 연구는 10톤 미만 어선의 내항성능 기반의 해양활동 기준 마련을 위한 기초 연구이다. 10톤 미만의 어선은 우리나라 등록어 선의 약 95%를 차지하고 있고 항해, 조업 등의 해양활동 중에 사고와 인명손실도 많이 발생하고 있다. 이에 따라 해양수산부에서는 어선 출항통제기준을 정하여 풍랑주의보 발효 시 어선의 운항을 제한하고 있지만 선박톤급과 파고에 따른 기준 없이 동일하게 적용하고 있어 해양활동 시 파고에 의한 선박의 동요에 많은 차이가 있을 수 있다. 파고에 의한 선박의 동요는 승선감 및 장비의 성능을 떨어뜨려 해양 사고의 요인이 될 수 있으므로 항해, 조업 등 안전한 해양활동 확보를 위해서 파랑 중 내항성능 검토가 필요하다. 하지만 어선에 대한 내 항성능 기반 기준 마련 검토는 부족한 실정이다. 이에 따라 우리나라 연안조업어선 10톤급(G/T 9.77톤) 어선을 대상으로 내항성능을 평가 하였고 설정된 내항성능 평가 기준의 Operation과 Survival 기준을 적용하여 유의파고와 선속에 따른 해양활동 충족 정도를 해석하였다. 해 석 결과 횡동요는 유의파고 0.4m부터 Operation 기준을 초과하였고 유의파고 2.2m부터 Survival 기준을 초과하는 것으로 나타났다. 종동요 는 유의파고 1.7m부터 Operation 기준을 초과하였고 유의파고 3.0m까지 Survival 기준은 넘지는 않았으나 횡동요가 유의파고 2.2m부터 Survival 기준을 초과한 상태로 안전하다고 할 수 없다. 따라서 10톤 미만 어선은 풍랑주의보 발효 전까지 출항은 가능하나 해양활동 관련 하여 내항성능 평가 기준에는 충족하지 못한다고 볼 수 있다. 본 연구에서는 10톤급 어선을 대상으로 제한적으로 평가 되었으나 해양활 동 기준 마련에 큰 도움이 될 것으로 판단된다.

국내에서 일어난 해양사고 발생 현황을 보면 평균적으로 약 8.5% 정도 사고 발생건수가 증가하는 추세를 보이고 있으며, 부선, 예인선, 유·도선 및 부유식해상구조물에서도 동일한 경향을 나타내고 있다. 본 연구에서는 국내 부유식해상구조물을 종류에 따라 분류하 고, 선박안전법, 어선법, 낚시 관리 및 육성법 등 관련 국내법을 참고하여 부유식 구조물의 종류에 따른 기준체계와 적용 범위를 검토하고 자 하였다. 아울러 해상 환경을 고려하여 국내 규정을 적용함에 있어 구조적으로 취약한 부분과 안전상의 사각지대에 놓인 위험 요소를 조기 발굴하고, 발굴된 위험 요소에 대해서 효과적인 개선 방안을 도출하고자 하였다.

There are generally two kinds of spent filter; one is spent filter media for mainly gaseous purification such as HEPA filter, the other is spent filter cartridge for liquid purification such as CVCS BRS cartridge type filter. The spent filter cartridge from liquid purification system has been storing in special shielding space in auxiliary building in NPPs since the beginning of 2006 according to the long term storage strategy for decaying short lived radionuclide and gaining the time for selecting practical treatment technology before final packaging. The spent filter cartridges generated Kori-1 reactor vary in their sizes as in length from 913 mm to 290 mm and range in radiation level from several hundred mSv per hour to below mSv per hour . It is high time that the spent filter cartridge is treated and packaged because LILW repository in Wolsung area is operating and Kori-1 reactor is scheduled to decommission. The spent filter cartridge is one of the wet solid wastes required of solidification. It is difficult for the spent filter cartridge to solidify because of their shape, structure, physical and chemical characteristics in addition to having high radiation level. NSSC notice defines that solidification of wet solid wastes include that solid material such as spent filter is encapsulated with cement, etc. as a form of macro-encapsulation. The radioactive waste acceptance criteria describes that non-homogeneous waste having above 74,000 Bq/g such as spent filter, dry active waste should be encapsulated with qualified material. Homogeneous waste such as spent resin, sludge, concentrated waste (liquid waste evaporator bottoms), etc. should be solidified complied with requirements except that spent filter which is allowed to encapsulate. It is needed to guide to the practice of these two requirements for spent filter. The sampling and test method is different between homogeneous solidification waste form and spent filter cartridge encapsulation waste form. For example, how core sample can be taken and how void space can be measured among spent filter cartridge in encapsulation waste form. The technical evaluation report for spent filter cartridge polymer encapsulation by US NRC has been reviewed and the technical position of US NRC was identified. As a result of review, improvement fields of waste acceptance criteria for spent filters are pointed out, and the technical position of US NRC for spent filter cartridge solidification is summarized. The recommendation on improvement directions for spent filter cartridge encapsulation is suggested.

Currently, as the saturation capacity of wet storage pool for spent nuclear fuel (SNF) of PWR in Korea has reached approximately 75%, Dry Storage Facilities (DSF) are necessary for sustainable operation of nuclear power plants. It is necessary to develop acceptance requirements for the delivery of SNF from reactor storage site to Centralized DSF. To do this end, the mechanical integrity of SNF is directly related to its repacking, retrieving, and transporting/handling performances. And also, this integrity is a key factor associated with the criticality safety that is connected to the damaged status of SNF. According to the NUREG/CR-6835, the NRC expects that the potential for nuclear fuel failures will increase because of the increase of the fuel discharge burnup and the degradation of fuel and clad material properties. Due to such damages and/or degradation, the fuel rods in the fuel assembly may be extracted and empty for following treatments (transportation, storage, handling etc). This condition can have a detrimental effect on the criticality safety of SNF. Thus, this study investigated whether extracted and empty of damaged SNF rod affects criticality safety. In this analysis, it is assumed that up to four fuel rods are missed. As a result of the analysis, As the number of fuel rods miss up to a certain number, the value of multiplication factor value of the fuel assembly increases. In addition, since the fuel rods located at the outermost layer contained relatively less fissile material than the fuel rods located center of the lattice, and neutrons were lost by the absorption material, the effective multiplication factor value gradually decreased. Nevertheless, the criticality safety was assessed to be maintained.

In Malaysia, there are several industries processing mineral ores generate residues containing naturally occurring radioactive material (NORM) with activity concentrations above the control limits established by the Malaysian Atomic Energy Licensing Board (AELB). These industries use mineral ores or concentrated ores as their feed materials to produce or extract valuable sand minerals or rare earth compounds for use in another industries. The control limits for activity concentrations of Uranium-238 (U-238) and Thorium-232 (Th-232) and their decay series is 1.0 Becquerel per gram (Bq·g−1) while activity concentration of Potassium 40 (K-40) is 10.0 Bq·g−1. The management of residue containing NORM radioactivity above the control limits must be done in accordance with current rules and regulations including proper handling, storage, transportation and/or disposal. Where possible, appropriate mixture process with other non-radiological material would reduce the activity concentrations to below the control limits. Depending on specific characteristics of residue, appropriate approach to reuse or recycle should be encouraged as part of special waste management. For this case, an exemption to release it from radiological controls can be applied but require scrutiny review and approval process by AELB. In addition, the health and safety aspects and environmental issues should be assessed which to be done in accordance with the relevant rules and regulations. As a last resort, a disposal of residue containing NORM radioactivity shall be done at the landfill disposal facility approved by AELB and other relevant Authorities.