• Proceedings of the Materials Research Society of Korea Conference
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• 2011.05a
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• pp.15-15
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• 2011

As you know, boron compounds, borax ($Na_2B_4O_5(OH)_4{\cdot}8H_2O$) etc. were known thousands of years ago. As for natural boron, it has two naturally occurring and stable isotopes, boron 11 ($^{11}B$) and boron 10 ($^{10}B$). The neutron absorption $^{10}B$ is included about 19~20% with 80~81% $^{11}B$. Boron is similar to carbon in its capability to form stable covalently bonded molecular networks. The mass difference results in a wide range of ${\beta}$ values between the $^{11}B$ and $^{10}B$. The $^{10}B$ isotope, stable with 5 neutrons is excellent at capturing thermal neutrons. For example, it is possible to decrease a thermal neutron required for the nuclear reaction of uranium 235 ($^{235}U$). If $^{10}B$ absorbs a neutron ($^1n$), it will change to $^7Li+^1{\alpha}$ (${\alpha}$ ray, like $^4He$) with prompt ${\gamma}$ ray from $^{11}B$ $^{11}B$ (equation 1). $$^{10}B+^1n\;{\rightarrow}\;^{11}B\;{\rightarrow}\; prompt \;{\gamma}\;ray (478 keV), \;^7Li+4{\alpha}\;(4He)\;\;\;\;{\cdots}\; (1)$$ If about 1% boron is added to stainless steel, it is known that a neutron shielding effect will be 3 times the boron free steel. Enriched boron or $^{10}B$ is used in both radiation shielding and in boron neutron capture therapy. Then, $^{10}B$ is used for reactivity control and in emergency shutdown systems in nuclear reactors. Furthermore, boron carbide, $B_4C$, is used as the charge of a nuclear fission reaction control rod material and neutron cover material for nuclear reactors. The $B_4C$ powder of natural B composition is used as a charge of a control material of a boiling water reactor (BWR) which occupies commercial power reactors in nuclear power generation. The $B_4C$ sintered body which adjusted $^{10}B$ concentration is used as a charge of a control material of the fast breeder reactor (FBR) currently developed aiming at establishment of a nuclear fuel cycle. In this study for new boron compound, silicon boride ceramics for capturing thermal neutrons, preparation and characterization of both silicon tetraboride ($SiB_4$) and silicon hexaboride ($SiB_6$) and ceramics produced by sintering were investigated in order to determine the suitability of this material for nuclear power generation. The relative density increased with increasing sintering temperature. With a sintering temperature of 1,923 K, a sintered body having a relative density of more than 99% was obtained. The Vickers hardness increased with increasing sintering temperature. The best result was a Vickers hardness of 28 GPa for the $SiB_6$ sintered at 1,923K for 1 h. The high temperature Vickers hardness of the $SiB_6$ sintered body changed from 28 to 12 GPa in the temperature range of room temperature to 1,273 K. The thermal conductivity of the SiB6 sintered body changed from 9.1 to 2.4 W/mK in the range of room temperature to 1,273 K.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1769-1785
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• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.

• Journal of the Korean Society of Safety
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• v.28 no.3
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• pp.95-99
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• 2013

The goal of this study was to investigate some significant factors to influence level of safety at plant construction field and analyze degree of risk by work classification. Currently, there are lots of construction fields for the nuclear power plant for electricity generation, and our government also planned constructing more nuclear power plant in near future. However, much of the safety literature neglected the degree of risk factors on the plant construction field. Safety managers participated in the brainstorming session for drawing decision criteria of the degree of risk (i.e., significant factors). Then, they were asked to answer a structured questionnaire which was developed for drawing most important factors. Finally, the analytic hierarchy process (AHP) was used to analyze level of risk by work classification. The following results were obtained. First, total twelve factors judging degree of risk were found in the brainstorming session. Second, the questionnaire showed four significant factors, including number of workers, working environments, skill of craft and accident experience. Third, the results of AHP showed Architecture work is the most dangerous work among 6 work types. The results could be used to reduce degree of risk in construction field of the nuclear power plant.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.905-913
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• 2017

Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional-integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi-Sugeno (T-S) fuzzy logic-based power distribution system. Two T-S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T-S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

• Journal of Energy Engineering
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• v.18 no.3
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• pp.191-202
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• 2009

South Korea, ranks 10th largest emitter of carbon dioxide in the world, will probably be under the obligation to reduce GHG emission from 2013. It is very important to reduce the electrical energy consumption since 30% of GHG emission in South Korea is made during electricity generation. In this study, based on "the 1st national energy master plan", the GHG emission reduction potential and the feasibility of the scenario in the electricity generation have been analyzed using LEAP(Long-range Energy Alternative Planning system). The scenario of the mater plan contains the 41% expansion of nuclear power plant facilities and the 11% diffusion of renewable energy until 2030. In result, total $CO_2$ emission reduction rate is 28.8% in 2030. Also $CO_2$ emission of unit electricity generation of bituminous coal power plant is $0.85kgCO_2/kWh$ and its LNG power plant is $0.51kgCO_2/kWh$ in BAU scenario. Therefore when existing facilities is exchanged for nuclear or renewable energy power plant, substitute of bituminous power plant is more effective than LNG power.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.87-97
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• 2016

The new generation of nuclear power plants (NPPs) will likely make use of state-of-the-art technologies in many areas of the plant. The analysis, design, and selection of advanced human-system interfaces (HSIs) constitute an important part of power plant engineering. Designers need to consider the new capabilities afforded by these technologies in the context of current regulations and new operational concepts, which is why they need a more rigorous method by which to plan the introduction of advanced HSIs in NPP work areas. Much of current human factors research stops at the user interface and fails to provide a definitive process for integration of end user devices with instrumentation and control and operational concepts. The current lack of a clear definition of HSI technology, including the process for integration, makes characterization and implementation of new and advanced HSIs difficult. This paper describes how new design concepts in the nuclear industry can be analyzed and how HSI technologies associated with new industrial processes might be considered. It also describes a basis for an understanding of human as well as technology characteristics that could be incorporated into a prioritization scheme for technology selection and deployment plans.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1786-1795
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• 2021

Printed Circuit Heat Exchanger (PCHE) is a widely used heat exchanger in the supercritical carbon dioxide (sCO₂) Brayton cycle because it can work under high temperature and pressure, and has been a hot topic in Next Generation Nuclear Plant (NGNP) projects for use as recuperators and condensers. Most previous studies focused on channel structures or shapes. However, no clear advancement has so far been seen in the allover size of the PCHE. In this paper, we proposed an optimal size of the PCHE with a fixed volume. Two boundary conditions of PCHE were simulated, respectively. When the volume of PCHE was fixed, the heat transfer rate and pressure loss were picked as the optimization objectives. The Pareto front was obtained by the Multi-objective optimization procedure. We got the optimized number of PCHE channels under two different boundary conditions from the Pareto front. The comprehensive performance can be increased by 5.3% while holding in the same volume. The numerical results from this study can be used to improve the design of PCHE with straight channels.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3324-3335
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• 2022

Several countries show interest in the Generation-IV power reactor innovative small module (PRISM), including: Canada, Japan, Korea, Saudi Arabia and the United Kingdom. Generation IV International Forum (GIF) has recommended the utilizing of probabilistic safety assessment (PSA) in evaluating the safety of Generation-IV reactors. This paper reviews the PSA performed for PRISM using SAPHIRE 7.27 code. This work shows that the core damage frequency (CDF) of PRISM for a single module is estimated by 8.5E-8/year which is lower than the Generation-IV target that is 1E-6 core damage per year. The social risk of PRISM (likelihood of latent cancer fatality) with evacuation is estimated by 9.0E-12/year which is much lower than the basic safety objective (BSO) that is 1E-7/year. The social risk without evacuation is estimated by 1.2E- 11/year which is also much lower than the BSO. For the individual risk (likelihood of prompt fatality), it is concluded that it can be considered negligible with evacuation (1.0E-13/year). Assuming no evacuation, the individual risk is 2.7E-10/year which is again much lower than the BSO. In comparison with other PSAs performed for similar sodium fast reactors (SFRs), it shows that PRISM concept has the lowest CDF.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1658-1668
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• 2019

Nuclear power generates a large portion of the energy used today and plays an important role in energy development. To ensure safe nuclear power generation, it is essential to conduct an accurate analysis of reactor structural integrity. Accordingly, in this study, a methodology for obtaining accurate structural responses to the combined seismic and reactor coolant loads existing prior to the shutdown of a nuclear reactor is proposed. By applying the proposed analysis method to the reactor vessel internals, it is possible to derive the seismic responses considering the influence of the hydraulic loads present during operation for the first time. The validity of the proposed methodology is confirmed in this research by using the finite element method to conduct seismic and hydraulic load analyses of the advanced APR1400 1400 MWe power reactor, one of the commercial reactors. The structural responses to the combined applied loads are obtained using displacement-based and stress-based superposition methods. The safety of the subject nuclear reactor is then confirmed by analyzing the design margin according to the American Society for Mechanical Engineers (ASME) evaluation criteria, demonstrating the promise of the proposed analysis method.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.199-206
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• 1997

For the Korean Next Generation Reactor(KNGR) development, LBB is considered for the Main Steam Line(MSL) piping inside its containment to achieve cost and safety Improvement. To apply LBB concept to MSL, leak sensors highly sensitive to humidity is required. In this paper, a ceramic material, MgCr$_2$O$_4$-TiO$_2$ has been developed as a humidity sensor for MSL applications. Experiments peformed to characterize the electrical conductivity shows that the conductivity of MgCr$_2$O$_4$-TiO$_2$ responds sensitively to both temperature and humidity changes. At a constant temperature below 10$0^{\circ}C$, the conductivity increases as the relative humidity increases, which makes the sensor favorable for application to the outside of MSL insulation layer But as temperature increases beyond 10$0^{\circ}C$, the sensor composition should be adjusted for the application to KNGR is to be made at temperature above 10$0^{\circ}C$.