• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1951-1956
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• 2019

As a part of the preliminary decommissioning plan of KEPCO-NF fuel fabrication facility, DCGLs of three target radionuclides, ²³⁴U, ²³⁵U, and ²³⁸U, were derived using RESRAD-BUILD code and contaminated areas of the facility were classified based on contamination levels from the derived DCGLs. From code simulations, one-room modeling results showed that the grinding room in building #2 was the most restrictive (DCGL_gross = 10493.01 Bq/㎡). The DCGL_gross results in contaminated areas from one-room modeling were slightly more conservative than three-room modeling. Prior to the code simulation, field survey and measurements conducted by each survey unit. For a conservative approach, the most restrictive DCGL_gross in each survey unit was taken as a reference to classify the contaminated areas of the facility. Accordingly, seven rooms and 37 rooms in the nuclear-fuel buildings were classified as Class 1 and Class 2, respectively. As expected, fuel material handling and processing rooms such as the grinding room, sintering room, compressing room, and powder collecting room were included in the Class 1 area.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.621-626
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• 2003

A DUPIC nuclear fuel is a newly developed fuel for CANDU reactors based on the concept of refabrication of spent PWR fuel by a dry process. Because a spent PWR fuel, a highly radioactive material, is used as a starting material, the experimental verification of DUPIC nuclear fuel fabrication requires an appropriate facility which should satisfy engineering requirements and guarantees safe operation. DUPIC nuclear fuel development team modified M6 hot-cell in IMEF to construct the dedicated facility(DFDF) for tile experiment. The experiment with spent PWR fuel have been conducted since January of 2000. Environmental effects of DFDF normal operation have been investigated when DUPIC nuclear fuel is fabricated with the maximum capacity of 50kg U/yr. The analysis results of the radiological safety of DFDF facility have shown that both national regulation limit and IMEF design criteria are satisfied.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.505-510
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• 1996

A simple model for RAM (Reliability, Availability and Maintainability) analysis and its computer code are developed for application to DUPIC fuel fabrication system. The approach is obtained by linking the allocation model (top-down method) to bottom-up method for RAM analysis. As a result, the availability requirement of subsystem, as well as the buffer storage requirement between processes, are evaluated for the DUPIC facility..

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.133-134
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• 2004

DUPIC (Direct Use of spent PWR fuel In CANDU) fuel cycle technology is being developed at Korea Atomic Energy Research Institute (KAERI). All the DUPIC fuel fabrication processes are remotely conducted in the completely shielded M6 hot-cell located in the Irradiated Material Examination Facility (IMEF) at KAERI. Undesirable products such as spent nuclear fuel powder debris and contaminated wastes are inevitably created during the DUPIC nuclear fuel fabrication processes.(omitted)

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.834-835
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• 2005

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.275-279
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• 2001

In uranium enrichment plants and nuclear fuel fabrication facilities, exact measurement of fissile isotope enrichment of uranium is required for material accounting in international safeguards inspection as well as process quality control. The purpose of this study was to develop a simple measurement system which can portably be used at nuclear fuel fabrication plants especially dealing with low enriched uranium. For this purpose, a small size CZT (CdZnTe) detector was used, and the detector performance in low uranium gamma/X -rays energy range was investigated by use of various enriched uranium oxide samples. New enrichment measurement technique and analysis method for low enriched uranium oxide, so-called, 'semi-peak ratio technique' was developed. The newly developed method was considered as an alternative technique for the low enrichment and would be useful to account nuclear material in safeguarding activity at nuclear fuel fabrication facility.

• Journal of the Korean Ceramic Society
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• v.33 no.6
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• pp.631-636
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• 1996

The powder characteristics of UO2 via AUC prepared by precipitation from a UN with AC soiution produced from nuclear fuel powder conversion plant and that of the existing facility were compared. Mean particle size of AUC powder was decreased and agglomerates were much occured in case of using the AC solution that that of the gases but other properties such as particle size distribution and shape of particle are thought to be similarly. In compaction of UO2 powder the breaking pressur of agglomerated UO2 powder and the sintered density of final UO2 pellet from AC solution were measured 1.45$\times$108 N/m2 and 10.52 g/cc, These values could be used in nuclear fuel powder fabrication process.

• Journal of Radiation Protection and Research
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• v.48 no.1
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• pp.52-57
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• 2023

Background: Korea Atomic Energy Research Institute (KAERI) operates several nuclear research facilities licensed by Nuclear Safety and Security Commission (NSSC). The emergency preparedness requirements, GSR Part 7, by International Atomic Energy Agency (IAEA) request protection strategy based on the hazard assessment that is not applied in Korea. Materials and Methods: In developing the protection strategy, it is important to consider an accident scenario and its consequence. KAERI has tried the hazard assessment based on a hypothesis accident scenario for the major nuclear facilities. During the assessment, the safety analysis report of the related facilities was reviewed, the simulation using MELCOR, MACCS2 code was implemented based on a considered accident scenario of each facility, and the international guidance was considered. Results and Discussion: The results of the optimized protective actions were 300 m evacuation and 800 m sheltering for the High-Flux Advanced Neutron Application Reactor (HANARO), the evacuation to radius 50 m, the sheltering 400 m for post-irradiation examination facility (PIEF), 100 m evacuation or sheltering for HANARO fuel fabrication plant (HFFP) facility. Conclusion: The results of the optimized protective actions and its distances for the KAERI facilities for the maximum postulated accidents were considered in establishing the emergency plan and procedures and implementing an emergency exercise for the KAERI facilities.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.529-534
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• 1996

본 논문은 현재까지 수행된 신규 핵연료가공시설의 안전성 평가 내용, 안전심사의 경험을 통해 도출된 문제과 해결 현황 등을 제시한다. 이를 위해 신규 핵연료가공시설의 인허가에 관련된 국내의 관련 원자력법령체계의 특성을 고찰하고, 관련되는 국내외 규제요건 및 기술기준과의 비교를 통해 적용가능한 규제기준을 도출하여, 본 적용 기준에 따른 신규 핵연료가공시설의 안전설계 특성 평가 결과를 제시한다. 이러한 과정에서 도출된 주요 문제점과 그 해결을 위한 개선방향을 제언한다.

• Nuclear Engineering and Technology
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• v.38 no.4
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• pp.359-374
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• 2006

Since 1991, Korea, Canada and United States have performed the direct use of spent pressurized water reactor (PWR) fuel in the Canada deuterium uranium (CANDU) reactors (DUPIC) fuel development project. Unlike the Tandem fuel cycle, which requires a wet reprocessing, the DUPIC fuel technology can directly refabricate CANDU fuels from the PWR spent fuel and, therefore, is recognized as a highly proliferation-resistant fuel cycle technology, which can be adopted even in non-proliferation treaty countries. The Korea Atomic Energy Research Institute (KAERI) has fabricated DUPIC fuel elements in a laboratory-scale remote fuel fabrication facility. KAERI has demonstrated the fuel performance in the research reactor, and has confirmed the operational feasibility and safety of a CANDU reactor loaded with the DUPIC fuel using conventional design and analysis tools, which will be the foundation of the future practical and commercial uses of DUPIC fuel.