• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2006.09a
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• pp.474.1-474.1
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• 2006

• 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 Nuclear Society Conference
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• 2004.10a
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• pp.1069-1070
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• 2004

• 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)

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.595-595
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• 2016

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.441-446
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• 1996

A preliminary conceptual design of the DUPIC fuel fabrication plant with production capacity of 400 MTHE/yr is presented. Capital and operating costs are also included. The levelized unit fabrication cost (LUC) for a reference mode was estimated at $509/kgHE, and sensitivity of some variable parameters to this reference was analysed.

• 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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• 2004.10a
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• pp.895-896
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• 2004

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.432-442
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• 2023

Characterizing the behavior of nuclear reactor plate fuels is vital to the progression of advanced fuel systems. The states of pre- and post-irradiation plates need to be determined effectively and efficiently prior to and following irradiation. Due to the hostile post-irradiation environment, characterization must be completed remotely. Laser-based characterization techniques enable the ability to make robust measurements inside a hot-cell environment. The Laser Shock (LS) technique generates high energy shockwaves that propagate through the plate and mechanically characterizes cladding-cladding interfaces. During an irradiation campaign, two Idaho National Laboratory (INL) fabricated MP-1 plates had a fuel breach in the cladding-cladding interface and trace amounts of fission products were released. The objective of this report is to characterize the cladding-cladding interface strengths in three plates fabricated using different fabrication processes. The goal is to assess the risk in irradiating future developmental and production fuel plates. Prior LS testing has shown weaker and more variability in bond strengths within INL MP-1 reference plates than in commercially produced vendor plates. Three fuel plates fabricated with different fabrication processes will be used to bound the bond strength threshold for plate irradiation insertion and assess the confidence of this threshold value.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.565-572
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• 2013

The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade $SiC_f/SiC$ composites are briefly reviewed. A CVI-processed $SiC_f/SiC$ composite with a PyC or $(PyC-SiC)_n$ interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.