• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.406-407
• /
• 2005

사용후핵연료 차세대관리공정에서는 $UO_2$ 펠렛을 균질화된 $U_{3}O_8$의 분말 형태로 되어야만 금속 전환공정에서 반응효율을 높일 수 있다. 본 연구에서는 실증용 건식분말화 장치에서 분말의 비산을 방지하면서 산화 효율을 최대한 높일 수 있는 최적 조건을 찾는데 있다고 할 수 있다. 본 연구의 실험에서 $UO_2$ 펠렛(약 90 g)에 유량별 공정 변수를 주어 최적화된 산화 조건을 조사하였다. 그 결과, $500^{\circ}C$ 에서 유량은 이론 산화량의 2배(1640 cc/min)의 조건에서 최적의 산화 효율을 보여주었다. 따라서 본 연구에서 실증용 건식분말화 장치의 비산을 방지하면서 산화 효율을 최대한 높일 수 있는 최적 유량을 결정하였다.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.161-166
• /
• 2005

In KAERI, about 3,100 drums containing soil have been stored. The soils were generated from the decommissioning process of Seoul office in 1988. Those soils occupy about $27\%$ of the capacity of the radioactive waste storage facility and make it difficult to maintain the storage facility. The major radioactive nuclides contained in the soils were expected to be Co-60 and Cs-137. As 16 years have passed, the radioactivity of those nuclides have decayed a lot. In this study, as a basis of regulatory clearance, radionuclides and radioactivity concentration of soils were analyzed. As a result, there are only Co-60 and Cs-137 in soils as ${\gamma}-emitters$. The total concentration of ${\gamma}-emitters$ in soil is analyzed as about $0.01\;{\sim}\;0.12$ Bq/g. As the soils are expected to be regulatory cleared in 2009, those concentrations will decay to be less than 0.1 Bq/g. This concentration can be meet the regulatory criteria suggested by IAEA. The regulatory clearance will be proceeded based on not only the assessment results of environmental influence but also related regulations.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2003.11a
• /
• pp.180-185
• /
• 2003

The reductions in final waste form and material costs, which were induced from an introduction of salt waste regeneration system, have been estimated and compared with those of the present pyrochemical process, which is under development in KAERI. The results calculated on the basis of published data and proper assumption showed that the final waste form of LiCl waste from the Advanced Conditioning Process would be reduced about 3.7 to#ton HM (from 5.4 to 1.7 ton/ton HM). For the case of LiCl-KCl eutectic salt waste from the electro-refining process, the final waste form would be reduced 2.3 ton/ton U. Thus, these estimation suggested that the introduction of salt waste regeneration system was essential to improve the economical efficiency of the pyrochemical process.

• Journal of Radiation Protection and Research
• /
• v.25 no.2
• /
• pp.89-96
• /
• 2000

Nuclear fuel cycle choices and costs are important in considering energy policies, fuel diversity, security of supply and associated social and environmental impacts. Particularly, the nuclear spent fuel is very important in view of high activity and the need of long term management. This study focuses on the comparison of reprocessing and direct disposal of nuclear spent fuel in terms of cost, safety and public acceptability. The results of the study show that the direct disposal is about 7% more economical than the reprocessing. In terms of safety, the results show that the risk of vitrified HLW (high-level radioactive waste) is less than directly disposed spent fuel. For the public acceptability, both of the methods are not well understood and therefore they are not accepted. In conclusion, it is necessary to guarantee the safety of the both spent fuel processing methods through continuous development of associated technology and to have a fuel cycle policy which should consider not only the economics but also social and environmental impacts.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.3
• /
• pp.299-311
• /
• 2019

Decommissioning is a critical issue in Korea. Although compared with the operation of nuclear power plants the release of radioactive materials during decommissioning is not expected to be significant, residents should always be protected from radiation exposure. To manage this effectively, Annual Release Objectives (ARO) and Annual Release Limits (ARL) were derived from dose standards in the NSSC Notice and dose limit for the public. Based on meteorological data for the three years from 2008 to 2010 in the Shin Kori nuclear power plant site, atmospheric dispersion and ground deposition factors of gaseous effluent were evaluated using the XOQDOQ computer code. The exposure dose was evaluated using the ENDOS-G computer code. Because of differences in radiological sensitivity according to age groups, the results of Annual Release Objectives (ARO) and Annual Release Limits (ARL) showed significant differences depending on the radionuclides. The evaluation methodology of this study will provide meaningful information for radioactive effluent management for decommissioning of nuclear power plants.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2003.11a
• /
• pp.331-336
• /
• 2003

The advanced spent fuel conditioning process(ACP) was proposed to reduce considerably the overall volume and radioactivity for effective management of the PWR spent fuel in respects on safety and economy. The ACP is under research and development, and have scheduled to perform hot test for demonstration of the ACP after several years. For hot test, hot cell facility of ${\alpha}{\gamma}$ type possess conservative safety is required essentially. A existing hot cell of ${\beta}{\gamma}$ type will be refurbished to minimize construction expenditures of hot cell facility. In this study, the design requirements are established, and the process detail work flow was analysed for the optimum arrangement to ensure effective process operation in hot cell. And also, the basic and detail design of hot cell facility and process and safety analysis was peformed to secure conservative safety of hot cell facility and process.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.9 no.2
• /
• pp.73-80
• /
• 2011

The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source ($^{60}Co$ radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.2 no.2
• /
• pp.113-124
• /
• 2004

A computer software program, called BIDAS (BIoassay Data Analysis Software) is developed to interpret the bioassay measurement data in terms of intakes and the committed effective dose using the human respiratory tract model (HRTM), gastrointestinal tract (GI-tract) model and biokinetic models currently recommended by the International Commission on Radiological Protection (ICRP) to describe the behavior of the radioactive materials within the body. The program consists of three modules; first, a database module to manage the bioassay data, second, another databasee module to store the predicted bioassay quantities of each radionuclide and finally, a computational module to estimate the intake and committed effective dose calculated with the bioassay quantity measurement values from either an acute or chronic exposure of the radionuclies within the body. This paper describes the features of the program as well as the quality assurance check results of the BIDAS software program.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.6 no.1
• /
• pp.45-54
• /
• 2008

A new proposed repository has a final capacity of 800,000 drums radioactive waste. Most of foreign repositories have a general practice of segregating control zones which mainly contributes to classification of degree of control, whether it is called buffer zone or not. Domestic regulatory requirements of establishment of buffer zone in a repository are not much different from those of nuclear power plants for operation period, in which satisfactory design objective or performance objective is the most important factor in determination of the buffer zone. The meaning of buffer zone after closure is a minimum requested area which can prevent inadvertant intruders from leading to non-allowable exposure during institutional control period. Safety assessment with drinking well scenario giving rise to the highest probability of exposure among the intruder's actions can verify fulfillment of the buffer zone which is determined by operational safety of the repository. At present. for the repository to be constructed in a few years, the same procedure and concept as described in this paper are applied that can satisfy regulatory requirements and radiological safety as well. However, the capacity of the repository will be stepwise extended upto 800,000 drums, consequently its layout will be varied too. Timely considerations will be necessary for current boundary of the buffer zone which has been established on the basis of 100,000 drums disposal.

• Tunnel and Underground Space
• /
• v.34 no.2
• /
• pp.104-126
• /
• 2024

Buffer and backfill play an essential role in isolating high-level radioactive waste and retard the migration of leaked radionuclides in deep geological disposal system. A bentonite mixture, which exhibits a swelling property, is considered for buffer and backfill materials, and excessive groundwater inflow from surrounding rock mass may affect stability and efficiency of their role as an engineered barrier. Therefore, stringent quality control as well as in-situ installation management and inflow water constrol for buffer and backfill are required to ensure the safety of deep disposal facilities. In this study, we analyzed the design requirements of buffer and backfill by examining various laboratory tests and a field study of the Steel Tunnel Test at the Äspö Hard Rock Laboratory in Sweden. We introduced how to control the quality of buffer and backfill construction in-field, and also presented how to handle excessive groundwater inflow into disposal caverns, validating the groundwater retention capacity of bentonite pellets and the effectiveness of geotexile use.