• Nuclear Engineering and Technology
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• v.12 no.2
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• pp.127-134
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• 1980

Shielding problems associated with neutron streaming through the reactor vessel cavity of pressurized water reactors are discussed to a certain extent with the actual examples in the currently operating reactors. Various remedial techniques are proposed herein to mitigate the tedious neutron streaming phenomena including piling up in heaps of temporary boron-containing bags and the installation of permanent shield structure making use of a certain refractory materials. In conclusion, optimum cavity shielding design concepts are presented with special emphasis on such major factors as the identification of major neutron streaming path, selection of necessary shielding materials with acceptable constraints, detailed design characteristics and physical configuration as well as the formulation of dependable mathematical tools to predict the final outcome of each design concept proposed in the context.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.99-101
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• 2014

A Comprehensive Vibration Assessment Program (CVAP) for steam generator internals (SGI) of Advanced Power Reactor 1400 (APR1400) is being performed in accordance with the United States Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.20 (RG 1.20) revision 3. This paper studies the vibration characteristics of moisture separator assembly as part of the vibration and stress analysis program for APR1400 SGI CVAP. The natural frequencies, mode shapes, and structural behavior of moisture separator assembly were investigated through modal analysis using finite element method and experimental measurement. Since the moisture separator consists of several items with complicated shape, an idealized shell model was used in the finite element analysis. Group of local modes caused by moisture separators and significant modes of shroud and separator support plate were identified. The results of this paper are to be utilized in the structural response analysis of moisture separator assembly.

• Fire Science and Engineering
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• v.30 no.4
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• pp.20-26
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• 2016

When the fire takes place in Nuclear Powr Plants(NPPs), the reactor should achieve and maintain safe shut-down conditions and minimize the radioactive material released to the environment. The U.S. Nuclear Regulatory Commission (NRC) has issued numerous generic communications related to fire protection over the past 20 years, after it issued its requirements in the Fire Protection Rule set forth in Title 10, Section 50.48 of the Code of Federal Regulations (10 CFR 50.48) and Appendix R to the 10 CFR 50. The and Nuclear Energy Institute (NEI) has developed a Methodology for Risk Informed Fire Safe-Shutdown Analysis, which is related to the Deterministic Method for Multiple Spurious Operations solutions. The aim of this study was to identify, achieve, and maintain Post-Fire Safe-Shutdown of the Main Control Room (MCR) of the CANDU reactor, even though one train of the multiple Safety Structures, Systems, and Components (SCCs) fail by the technical specification and analysis method.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.803-816
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• 2014

Defense-in-depth is a fundamental safety principle for the design and operation of nuclear power plants. Despite its general appeal, defense-in-depth is not without its drawbacks, which include its potential for concealing the occurrence of hazardous states in a system, and more generally rendering the latter more opaque for its operators and managers, thus resulting in safety blind spots. This in turn translates into a shrinking of the time window available for operators to identify an unfolding hazardous condition or situation and intervene to abate it. To prevent this drawback from materializing, we propose in this work a novel safety principle termed "observability-in-depth". We characterize it as the set of provisions technical, operational, and organizational designed to enable the monitoring and identification of emerging hazardous conditions and accident pathogens in real-time and over different time-scales. Observability-in-depth also requires the monitoring of conditions of all safety barriers that implement defense-in-depth; and in so doing it supports sensemaking of identified hazardous conditions, and the understanding of potential accident sequences that might follow (how they can propagate). Observability-in-depth is thus an information-centric principle, and its importance in accident prevention is in the value of the information it provides and actions or safety interventions it spurs. We examine several "event reports" from the U.S. Nuclear Regulatory Commission database, which illustrate specific instances of violation of the observability-in-depth safety principle and the consequences that followed (e.g., unmonitored releases and loss of containments). We also revisit the Three Mile Island accident in light of the proposed principle, and identify causes and consequences of the lack of observability-in-depth related to this accident sequence. We illustrate both the benefits of adopting the observability-in-depth safety principle and the adverse consequences when this principle is violated or not implemented. This work constitutes a first step in the development of the observability-in-depth safety principle, and we hope this effort invites other researchers and safety professionals to further explore and develop this principle and its implementation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.455-472
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• 2018

It is important to study how to manage dry storage casks of spent nuclear fuels (SNF), because wet storage spaces for SNF will shortly be at full capacity in the Republic of Korea. The US has operated a dry storage cask system for several decades, and has carried out significant studies into how to successfully manage dry storage cask for SNF. This type of expertise and experience is currently lacking in the Republic of Korea. The degradation of dry casks is an important issue that must be considered. In particular, chloride-induced stress corrosion cracking (CISCC) is known to lead to the release of radioisotopes from canisters. The U.S. Department of Energy, U.S. Nuclear Regulatory Commission, and the Electric Power Research Institute have undertaken research into the CISCC mechanism. In addition, Sandia National Laboratories (SNL) has extensively researched CISCC and how to manage it in dry storage canisters. In this review paper, the probabilistic model proposed by the SNL is analyzed and, based on this model, US-based CISCC research is reviewed in detail. This paper will inform the management of dry cask storage of SNF from light water reactors in austenite stainless steel canisters in the Republic of Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.297-308
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• 2007

The granulation equipment of concentrated wastes is manufactured for the polymer solidification of concentrated wastes. It uses liquid sodium silicate as a granulating agent for the granulating of dried powder containing boric acid. The granulating agent is sprayed in the form of droplet and mean size of dried granules is $2{\sim}4mm$. The new technology which has been used for the polymer solidification of spent resin in U.S. and certified by Nuclear Regulatory Commission (NRC) is successfully applied to concentrated wastes. This uses in-situ solidification process within drum without mechanical mixing. Maximum loading of waste can be achieved without increasing of waste volume. Polymer waste forms were evaluated with several test such as fire test, compressive strength test, leaching test, immersion test, irradiation test, and thermal cycling test according to standard test procedures.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2418-2426
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• 2022

The grouping analysis is a method guided by the Korea Radioactive Waste Agency for efficient analysis of radioactive waste for disposal. In this study, experiments to verify the adequacy of grouping analysis were conducted with radioactive soil, concrete, and dry active waste in similar environments. First, analysis results of the major radionuclide concentrations in individual waste samples were reviewed to evaluate whether wastes from similar environments correspond to a single waste stream. As a result, the soil and concrete waste were identified as a single waste stream because the distribution range of radionuclide concentrations was "within a factor of 10", the range that meet the criterion of the U.S. Nuclear Regulatory Commission for a single waste stream. On the other hand, the dry active waste was judged to correspond to distinct waste streams. Second, after analyzing the composite samples prepared by grouping the individual samples, the population means of the values of "composite sample analysis results/individual sample analysis results" were estimated at a 95% confidence level. The results showed that all evaluation values for soil and concrete waste were within the set reference values (0.1-10) when five-package and ten-package grouping analyses were conducted, verifying the adequacy of the grouping analysis.

• Journal of Radiation Protection and Research
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• v.39 no.2
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• pp.89-95
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• 2014

In potential accident consequence assessments for the licensing approval of LWRs, the ground deposition of radionuclides released into the environment is not allowed into the models, as recommended in the U. S. Nuclear Regulatory Commission's regulatory guide. Meanwhile, it is allowed into the assessment models for the licensing approval of PHWRs with consideration of more detailed physical processes of radionuclides in the atmosphere. Under these backgrounds, importance of exposure dose by ground deposition was quantitatively evaluated and comprehensively discussed. For potential accidental releases of $^{137}Cs$ and $^{131}I$, total exposure doses were more conservative in case of without consideration of ground deposition than in case of with its consideration. It was because of that the depletion of air concentration resulting from ground deposition is more influential in the contribution to total exposure doses than additional doses from contaminated ground. The exposure doses by the inhalation of contaminated air showed the contribution of more than 90% in total exposure doses, depending on atmospheric stability, release period of radionuclides and distance from a release point. The exposure doses from contaminated ground showed less than 10% at most in contribution of total exposure doses. The ratios of total exposure doses in case of with consideration of deposition to without its consideration for $^{131}I$ were distinct than those for $^{137}Cs$. As the atmosphere is more stable, release duration of radionuclides is longer, distance from a release point is longer, it was more distinct.

• Journal of the Korean Geographical Society
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• v.47 no.5
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• pp.755-774
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• 2012

The main objective of the study is to propose GIS-based methods to assess the population distribution criteria for undesirable facilities such as nuclear power plants. First of all, a review of the relevant criteria was conducted for the official documents compiled by such institutions as IAEA (International Atomic Energy Agency), U.S. NRC (Nuclear Regulatory Commission), and some national institutes including the Korea Institute of Nuclear Safety. It is informed from the review that the fundamental principle underlying the various criteria is to maximize the distance between a plant and the nearest population center. It is realized that two interrelated GIS-based techniques need to be devised to put the principle into practice; sophisticated ways of representing population distribution and identifying population centers. A dasymetric areal interpolation is proposed for the former and cell-based and area-based critical density methods are introduced. Grid-based population distributions at various spatial resolutions are created by means of the dasymetric areal interpolation. By applying the critical density methods to the gridded population distribution, some population centers satisfying the population size and density criteria can be identified. These methods were applied to the case of the Gori-1 nuclear power plant and their strengths and limitations were discussed. It was revealed that the assessment results could vary depending upon which method was employed and what values were chosen for various parameters. This study is expected to contribute to foster the applications of methods and techniques developed in geospatial analysis and modeling to the site selection and evaluation.