• 해양환경안전학회지
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• 제20권2호
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• pp.130-142
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• 2014

정부는 훼손된 해양생태계의 현황 및 원인을 파악하고 생태계 기능 복원 및 손실 방지 전략을 수립하기 위해 해양생태계 복원기술개발 사업의 시행을 고려하고 있다. 사업 시행 여부에 대한 판단을 위해서는 이 사업에 대한 경제적 타당성 분석이 필수적으로 요구된다. 이에 본 연구에서는 조건부 가치측정법(CVM, contingent valuation method)을 적용하여 사업 수행의 경제적 타당성을 분석하고자 한다. 지불의사 유도방법으로 유인일치적인 양분선택형 모형을 이용하되, 지불의사액 추정모형으로 영(0)의 응답을 명시적으로 다룰 수 있는 스파이크 모형을 적용한다. CVM 적용을 위한 설문조사는 미국 해양대기청의 지침에 따라 전국 1,000가구를 대상으로 일대일 개별면접을 통해 2013년에 시행되었다. 분석결과 연간 가구당 평균 지불의사액은 5,414원으로 추정되었다. 이 값을 전국으로 확장하면 향후 5년 동안 연간 약 986억원에 달한다. 이 값과 해양생태계 복원기술개발 사업의 투자비 정보를 이용하여 경제성을 분석한 결과, 순현재가치, 편익/비용 비율, 내부수익률은 각각 3,378억원, 5.20, 65.9 %로 산정되어 각각 0, 1.0, 5.5 %를 상회하므로 이 사업은 비용-편익 분석을 통과한다.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.678-699
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• 2015

Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for nearterm human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of $100MW_{th}$ and an electricity generation mode of $100MW_{th}$, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and thermohydraulics was carried out. The result indicates that the innovative design has great potential for high propellant efficiency and thrust-to-weight of engine ratio, compared with the existing NTR designs. However, the build-up of fission products in fuel has a significant impact on the bimodal operation of the moderated reactor such as xenon-induced dead time. This issue can be overcome by building in excess reactivity and control margin for the reactor design.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.274-284
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• 2016

Purpose: The purpose of this work is to describe the spectrometric analysis of gaseous cloud formation over reactor mixed uranium-and-plutonium (UP) fuel $(U_{0.8}Pu_{0.2})O_2$ samples heated to a temperature $>2,000^{\circ}C$, and thus forecast and evaluate radiation hazards threatening humans who cope with the consequences of any accident at a fission reactor loaded by UP mixed oxide $(U_{0.8}Pu_{0.2})O_2$, such as a mixture of 80% U and 20% Pu in weight. Materials and methods: The UP nuclear fuel samples were heated up to a temperature of over $2,000^{\circ}C$ in a suitable assembly (apparatus) at out-of-pile experiments' implementation, the experimental in-depth study of metabolism of active materials in living organisms by means of artificial irradiation of pigs by plutonium. Spectrometric measurements were carried out on the different exposed organs and tissues of pigs for the further estimation of human internal exposure by nuclear materials released from the core of a fission reactor fueled with UP mixed oxide. Results: The main results of the research described are the following: (1) following the research on the influence of mixed fuel fission products (radioactive isotopes being formed during reactor operation as a result of nuclear decay of elements included into the fuel composition) on living organisms, the authors determined the quantities of plutonium dioxide ($PuO_2$) that penetrated into blood and lay in the pulmonary region, liver, skeleton and other tissues; and (2) experiments confirmed that the output speed of plutonium out of the basic precipitation locations is very small. On the strength of the experimental evidence, the authors suggest that the biological output of plutonium can be disregarded in the process of evaluation of the internal irradiation doses.

• Nuclear Engineering and Technology
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• 제44권5호
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• pp.523-534
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• 2012

The objective of this study was to experimentally investigate the effect of heat curing methods on the temperature history and strength development of slab concrete exposed to $-10^{\circ}C$. The goal was to determine proper heat curing methods for the protection of nuclear power plant structures against early-age frost damage under adverse (cold) conditions. Two types of methods were studied: heat insulation alone and in combination with a heating cable. For heat curing with heat insulation alone, either sawdust or a double layer bubble sheet (2-BS) was applied. For curing with a combination of heat insulation and a heating cable, an embedded heating cable was used with either a sawdust cover, a 2-BS cover, or a quadruple layer bubble sheet (4-BS) cover. Seven different slab specimens with dimensions of $1200{\times}600{\times}200$ mm and a design strength of 27 MPa were fabricated and cured at $-10^{\circ}C$ for 7 d. The application of sawdust and 2-BS allowed the concrete temperature to fall below $0^{\circ}C$ within 40 h after exposure to $-10^{\circ}C$, and then, the temperature dropped to $-10^{\circ}C$ and remained there for 7 d owing to insufficient thermal resistance. However, the combination of a heating cable plus sawdust or 2-BS maintained the concrete temperature around $5^{\circ}C$ for 7 d. Moreover, the combination of the heating cable and 4-BS maintained the concrete temperature around $10^{\circ}C$ for 7 d. This was due to the continuous heat supply from the heating cable and the prevention of heat loss by the 4-BS. For maturity development, which is an index of early-age frost damage, the application of heat insulation materials alone did not allow the concrete to meet the minimum maturity required to protect against early-age frost damage after 7 d, owing to poor thermal resistance. However, the combination of the heating cable and the heat insulating materials allowed the concrete to attain the minimum maturity level after just 3 d. In the case of strength development, the heat insulation materials alone were insufficient to achieve the minimum 7-d strength required to prevent early-age frost damage. However, the combination of a heating cable and heat insulating materials met both the minimum 7-d strength and the 28-d design strength owing to the heat supply and thermal resistance. Therefore, it is believed that by combining a heating cable and 4-BS, concrete exposed to $-10^{\circ}C$ can be effectively protected from early-age frost damage and can attain the required 28-d compressive strength.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.54-70
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• 2017

The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility-the advanced core-cooling mechanism experiment (ACME)-was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups-a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break-were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative effects on the passive core cooling performance caused by nitrogen injection during the SBLOCA transient.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2196-2203
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• 2020

A dryout mechanism model for rectangular narrow channels at high pressure conditions is developed by assuming that the Kelvin-Helmholtz instability triggered the occurrence of dryout. This model combines the advantages of theoretical analysis and empirical correlation. The unknown coefficients in the theoretical derivation are supported by the experimental data. Meanwhile, the decisive restriction of the experimental conditions on the applicability of the empirical correlation is avoided. The expression of vapor phase velocity at the time of dryout is derived, and the empirical correlation of liquid film thickness is introduced. Since the CHF value obtained from the liquid film thickness should be the same as the value obtained from the Kelvin-Helmholtz critical stability under the same condition, the convergent CHF value is obtained by iteratively calculating. Comparing with the experimental data under the pressure of 6.89-13.79 MPa, the average error of the model is -15.4% with the 95% confidence interval [-20.5%, -10.4%]. And the pressure has a decisive influence on the prediction accuracy of this model. Compared with the existing dryout code, the calculation speed of this model is faster, and the calculation accuracy is improved. This model, with great portability, could be applied to different objects and working conditions by changing the expression of the vapor phase velocity when the dryout phenomenon is triggered and the calculation formula of the liquid film.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2379-2386
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• 2020

This paper presents the results of the first intercomparison exercise performed by the Korea retrospective dosimetry (KREDOS) working group using electron paramagnetic resonance (EPR) spectroscopy. The intercomparison employed the alanine dosimeter, which is commonly used as the standard dosimeter in EPR methods. Four laboratories participated in the dose assessment of blind samples, and one laboratory carried out irradiation of blind samples. Two types of alanine dosimeters (Bruker and Magnettech) with different geometries were used. Both dosimeters were blindly irradiated at three dose levels (0.60, 2.70, and 8.00 Gy) and four samples per dose were distributed to the participating laboratories. Assessments of blind doses by the laboratories were performed using their own measurement protocols. One laboratory did not participate in the measurements of Magnettech alanine dosimeter samples. Intercomparison results were analyzed by calculating the relative bias, E_n value, and z-score. The results reported by participating laboratories were overall satisfactory for doses of 2.70 and 8.00 Gy but were considerably overestimated with a relative bias range of 10-95% for 0.60 Gy, which is lower than the minimum detectable dose (MDD) of the alanine dosimeter. After the first intercomparison, participating laboratories are working to improve their alanine-EPR dosimetry systems through continuous meetings and are preparing a second intercomparison exercise for other materials.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2204-2220
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• 2020

An open-pool type research reactor is designed and operated considering the accessibility around the pool top area to enhance the reactor utilization. The reactor structure assembly is placed at the bottom of the pool and filled with water as a primary coolant for the core cooling and radiation shielding. Most radioactive materials are generated from the fuel assemblies in the reactor core and circulated with the primary coolant. If the primary coolant goes up to the pool surface, the radiation level increases around the working area near the top of the pool. Hence, the hot water layer is designed and formed at the upper part of the pool to suppress the rising of the primary coolant to the pool surface. The temperature gradient is established from the hot water layer to the primary coolant. As this temperature gradient suppresses the circulation of the primary coolant at the upper region of the pool, the radioactive primary coolant rising up directly to the pool surface is minimized. Water mixing between these layers is reduced because the hot water layer is formed above the primary coolant with a higher temperature. The radiation level above the pool surface area is maintained as low as reasonably achievable since the radioactive materials in the primary coolant are trapped under the hot water layer. The key to maintaining the stable hot water layer and keeping the radiation level low on the pool surface is to have a stable flow of the primary coolant. In the research reactor with a downward core flow, the primary coolant is dumped into the reactor pool and goes to the reactor core through the flow guide structure. Flow fields of the primary coolant at the lower region of the reactor pool are largely affected by the dumped primary coolant. Simple, circular, and duct type discharge headers are designed to control the flow fields and make the primary coolant flow stable in the reactor pool. In this research, flow fields of the primary coolant and hot water layer are numerically simulated in the reactor pool. The heat transfer rate, temperature, and velocity fields are taken into consideration to determine the formation of the stable hot water layer and primary coolant flow. The bulk Richardson number is used to evaluate the stability of the flow field. A duct type discharge header is finally chosen to dump the primary coolant into the reactor pool. The bulk Richardson number should be higher than 2.7 and the temperature of the hot water layer should be 1 ℃ higher than the temperature of the primary coolant to maintain the stability of the stratified thermal layer.

• 한국습지학회지
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• 제17권4호
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• pp.332-338
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• 2015

월악산국립공원에 발달되어 있는 소나무림에서 2013년 1월부터 2013년 12월까지 유기탄소 분포와 순환을 통한 생태계 서비스 가치를 평가하였다. 지상부와 지하부 생물량에 분포되어 있는 유기탄소량은 각각 32.17 및 8.04 ton C $ha^{-1}$이었으며, 낙엽층과 토양의 유기탄소량은 각각 5.55 ton C $ha^{-1}$ 및 58.62 ton C $ha^{-1}$ 50cm-$depth^{-1}$로 조사되었다. 조사지 소나무림의 전체 유기탄소량은 104.38 ton C $ha^{-1}$이었으며, 이중 37.9%가 식물체에 분포하였다. 소나무림의 전체 유기탄소량을 원화로 환산하면 약 1,044 만원 $ha^{-1}$의 가치를 갖는 것으로 추정되었다. 조사기간 동안 토양호흡을 통하여 방출되는 탄소량은 4.44 ton C $ha^{-1}yr^{-1}$으로 이중 미생물호흡과 뿌리호흡을 통해 방출되는 탄소량은 각각 2.18 및 2.27 ton C $ha^{-1}yr^{-1}$이었다. 유기탄소 순 생산량과 미생물호흡량의 차이로 추정했을 때 본 소나무림에서 연간 대기로부터 흡수하는 순 유기탄소는 0.44 ton C $ha^{-1}yr^{-1}$로서, 이를 원화로 환산하면 약 44,000원 $ha^{-1}$의 가치를 갖는 것으로 추정되었다.

• Korean Journal of Acupuncture
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• 제27권3호
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• pp.129-142
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• 2010

Background : The Bonghan theory is a hypothesis on the anatomical structure of the acupuncture point and meridian system. It has been regarded as a misunderstanding of the lymphatic system, or as a made-up story, for the past 40 years. Since 2002, Many studies have been published either to support the theory or to refute the old viewpoint. Objective : The purpose of this study was to find out the validity of the refutation by reviewing the publications. Methods : Searches were made from online databases (Riss4u.net, Google.com, Sciencedirect.com, Pubmed.com, baidu.com, and ci.nii.ac.jp) from 1960 to 2009. The search terms that were used were "Bonghan," "Bong han," "봉한," "thread-like structure," "KИM БOH XaHOM", "CИCTEMA KEHPAK," "鳳漢," "鳳漢管," and "鳳漢学說." References from the searched publications were also used. Results : Since the 1960s, 107 publications were identified as related works, but only 11 publications sought to refute the Bonghan theory. Two publications were researches, and nine were reviews. In the analysis of the correlation of the arguments with the publication contents, it was found that the research of G. Kellner reviewed the Bonghan system properly but that V. V. Kosmatov did not classify the ear-reflex zone as a traditional acupuncture point. For the review publications, only two reviews contained proper arguments, but seven publications were identified as broad interpretations, wrong translations, etc. Moreover, the two proper reviews were grounded on the research of G. Kellner. Conclusions : We found that the scientific origin of the refutation of the Bonghan theory is only one research by G. Kellner. This result suggest that Bonghan theory was not discussed enough to determine the invention.