• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.173-180
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• 2017

In this study, air sampling locations in the stack of the Advanced Fuel Science Building (AFSB) at the Korea Atomic Energy Research Institute (KAERI) were assessed according to the ANSI/HPS N13.1-1999 specification. The velocity profile, flow angle and $10{\mu}m$ aerosol particle profile at the cross-section as functions of stack height L and stack diameter D (L/D) were assessed according to the sampling location criteria using COMSOL. The criteria for the velocity profile were found to be met at 5 L/D or more for the height, and the criteria for the average flow angle were met at all locations through this assessment. The criteria for the particle profile were met at 5 L/D and 9 L/D. However, the particle profile at the cross-section of each sampling location was found to be non-uniform. In order to establish uniformity of the particle profile, a static mixer and a perimeter ring were modeled, after which the degrees of effectiveness of these components were compared. Modeling using the static mixer indicated that the sampling locations that met the criteria for the particle profile were 5-10 L/D. When modeling using the perimeter ring, the sampling locations that met the criteria for particle profile were 5 L/D and 7-10 L/D. The criteria for the velocity profile and the average flow angle were also met at the sampling locations that met the criteria for the particle profile. The methodologies used in this study can also be applied during assessments of air sampling locations when monitoring stacks at new nuclear facilities as well as existing nuclear facilities.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.114-129
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• 2017

Background: The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. Materials and Methods: The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012-018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. Results and Discussion: The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. Conclusion: The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3 cores is well within the range estimated by the reverse or inverse method, which are summarized in UNSCEAR 2013 report. But this does not necessarily mean that these two approaches are consistent.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.311-318
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• 2010

U. S. NRC Regulation Guide 1.52 regulating ESF ACS in nuclear power plants has been revised to revision 3. To apply reduction of operability test time, allowance of alternative challenge agents for in-place leak test of HEPA filters, and upgrade of Methyl Iodide penetration acceptance criterion in activated carbon performance test suggested in Reg. Guide 1.52(Rev.3) on Yonggwang units 5 and 6 ESF ACSes, technical feasibility study was carried out with on-site experiments as well as experiments with a lab-scale model. It was confirmed that the moisture in the system returned to the level before the test in 1 or 4 days even though the moisture was removed during the operability test lasting more than 10 hours. Therefore, it is appropriate to perform monthly operability test in 15 minutes just long enough to check the operability of equipment. To change challenge material for in-place HEPA filter leak test, size of aerosol, production rate, and leak detection capability were compared for DOP and PAO. It was concluded that PAO can be substituted for DOP in nuclear power plants. The upgrade of Methyl Iodide penetration acceptance criterion from 0.175 % to 0.5 % in active carbon filter bed deeper than 4 inches was to conform to the change of activated carbon performance test method to ASTM D3803(1989). It was confirmed that Methyl Iodide penetration acceptance criterion of 0.5 % under $30^{\circ}C$, relative humidity 95 % condition was conservatively good enough for testing performance of active carbon insitu. The licence change of Yonggwang units 5 and 6 has been completed based on this study.