• Strategy21
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• 통권41호
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• pp.261-293
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• 2017

Nuclear power is a way of attaining an enormous amount of energy with relatively small amount of resources and after it has been introduced to the submarine since 1954, there are approximately 150 of nuclear powered submarine currently on a mission around the world. This is due to the maneuverability, mountability and covertness of nuclear submarines. However, there are other tasks, not only the high level of nuclear technology that are needed to be dealt with in order to construct nuclear powered submarine. The biggest task of all is to secure the enriched uranium. Accordingly, this research is about the way of enriching and securing the nuclear fuel that are used in the nuclear submarine with the characteristics, merits and demerits of the nuclear submarine. Due to the fact that the pressurized water reactor in South Korea is the reactor that was originally built for the development of nuclear powered submarine, many parts is designed to be suitable for the submarine propulsion. However, in order to apply this to submarine it is needed to consider additional requests such as the position of reactor, accident-coping system, radioactive covering, reactor output adjustment and ship's pitch and roll in order to apply this to submarine. Nuclear submarines have much higher speed based on the powerful propulsion in comparison with diesel-electric submarine and also have bigger loading area. Besides, there is no need to snorkel and they also have advantages in covertness with the multi-noise proof system. The nuclear technology in South Korea has seen the dramatic development since 1962 and in 1998 reached to the level that we have succeeded in the localization of nuclear plant and exported the world-class one-piece small-sized reactor (SMART) to UAE. To operate these reactors, we import the whole quantity of low-enriched uranium and having our own uranium enrich facility is not probable because of the budget and international regulations. With the ROK/US nuclear agreement revised on 2015 November, the enrichment of uranium that are available without special permission has changed up to 20%. According to the assumption that we use the 20% enrichment of Uranium on U.S. virginia class submarine, it is necessary to change the fuel after 11 years and it will cause additional cost of 1 billion dollars. But the replace period by the uranium's enrichment rate is not fixed so that it is possible to change according to the design of reactor. Therefore, I would like to make a suggestion on two types of design concepts of nuclear submarine that can be operated for 30 years without nuclear fuel change by using the 20% enriched uranium from ONNp.First of all, it is possible by increasing the size of reactor by 3 times and it results in the 1,000t increase of the weight. And secondly, it is by designing the one piece reactor to insert devices such as steam turbine, condenser into the inside of nuclear core like the Rubis class submarines of France.

• 청정기술
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• 제23권4호
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• pp.441-447
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• 2017

개념설계 단계에서 개발 공정에 대한 경제적 타당성 분석에 대한 중요성이 대두되고 있으며, 목표 경제성에 부합하는 공정개발을 위한 공정 최적화에 대한 연구도 활발히 진행되고 있다. 발전 시스템 분야에서는 전력 단가(Levelized cost of electricity, LCOE)를 예측하여 경제적 효과를 정량적으로 비교 분석하는 평가 방법이 많이 활용되고 있다. 본 연구에서는 목표 전력 단가에 부합한 발전 시스템을 설계하기 위해서 요구되는 핵심기기의 설계 조건을 역산출 할 수 있는 플랫폼을 구축하였으며 초임계 이산화탄소 발전 시스템이 석탄 화력에 적용될 경우, 목표 전력 단가(초임계 증기 랭킨 사이클 발전 단가, $ 85.4 /kWh)를 충족하기 위해 요구되는 주요 핵심기기(압축기, 터빈, 열교환기) 등의 설계 지표 기준을 도출하였다. 터빈의 등엔트로피 효율이 86%인 경우, 주압축기 효율은 88% 이상 설계되어야 한다. 만약 터빈의 등엔트로피 효율이 88%로 설계된 경우, 주압축기 효율은 82%까지 완화하여 설계가 가능하다. End seal 부분에서 누설량을 0.24% 수준으로 유지하고, 열교환기의 경우 cold side 출구측 온도가 $92{\sim}97^{\circ}C$, 열용량은 2650 ~ 2680 MWth로 설계한다면 목표 전력단가를 충족시킬 수 있을 것으로 확인되었다.

• 대한토목학회논문집
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• 제32권2A호
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• pp.123-130
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• 2012

증기를 발생시켜 터빈(turbine)을 회전시키는 화력 및 원자력 발전 계통에서 냉각시설은 필수적인 구조물이며, 냉각수 순환 계통은 일반적으로 해수를 취수하여 발전소 내의 복수기까지 유입시켜 증기와 열 교환 후 다시 외해로 배출시키는 형태를 취하고 있다. 최근 냉각수 취 배수 방식을 표층 취 배수 방식이 아닌 심층 취 배수 방식으로 변경하고 있는데, 기존 원전의 재순환 온도에 대한 영향을 최소화 하고, 온배수 방류시 밀도차로 인한 부력으로 온배수 혼합효과를 높여 온배수에 의한 환경피해 범위를 최소화하기 위해서이다. 특히, 하절기에 저층의 저온 냉각수를 취수할 수 있다는 이점 때문에 향후 계획되는 발전소들도 심층 취 배수 방식을 도입할 것으로 예상된다. 본 연구에서는 원자력 발전소의 냉각시설 중 심층 취 배수 구조물의 입구 주변을 3차원 전산유체역학 코드인 $FLOW-3D^{(R)}$로 모사하여 그 흐름특성을 분석하였다. 취수구(intake)의 경우 연직취수 조건에서 유속 덮개(Velocity cap), 배수구(diffuser)의 경우 방류수의 분사방향에 변화를 주어 모의하였으며, 그 결과 취수구의 경우 유속덮개에 의한 연직 유속성분의 현저한 감소로 인한 어류 유입영향을 최소화할 수 있을 것으로 판단되며, 배수구 희석효과는 Jirka 및 Harleman이 제시한 2차원 온배수 프룸(frume)과 잘 일치 하는 것으로 나타났다.

• 수산해양교육연구
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• 제5권1호
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• pp.31-44
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• 1993

Through the early 1900's, the evolution of the surface condenser was closely tied to the development of steam engine and the turbine. As the chemical and petroleum industries evolved in the 1900's, the use of surface condensers in many different processes. Today, industry uses condensers in many shapes and sizes. The actual condensation process occurs on the outside surface of tubes. The nature of this surface geometry affects the condenser's heat transfer performance. The first condensers were built with plain tubes. As tube manufacturing techniques advanced, manufacturers started making tubes with integral fins. In the 1940's, fin densities were limited to about 600 to 700 fins per meter(fpm) because of manufacturing procedure. Today new manufacturing techniques allow production of tubes with fin densities ranging from 750 to 1600 fpm. The integral-fin tubes investigated in this paper are nominally 19 mm diameter. Eight tubes have been used with trapezodially shaped integral-fins having fin density from 748 to 1654 fpm and 10, 30 grooves. For comparison, tests are made using a plain tube having the same inside diameter and an outside diameter equal to that at the root of the fins for the finned tubes. Betty and Katz's theoretical modelis is used to predict the R-11 condensation coefficient on horizontal integral-fin tubes having 748, 1024 and 1299 fpm. Experiments are carried out using R-11 as working fluid. The refrigerant condensates at a saturation state of $30^{\circ}C$ on the outside tube surface cooled by coolant. The amount of noncondensable gases present in the test loop is reduced to a negligible value by repeated purging. For a given heat input to the boiler and given cooling water flow rate, all test data are taken at steady state. The observed heat transfer enhancement for the finned and grooved tubes significantly exceeded that to be expected on grounds of increased area. For the eight fin tubes and one plain tube tested, the best performance has been obtained with a tube having a fin density of 1299 fpm, and a fin bight of 1.2mm and 30 grooves.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.603-616
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• 2007

Roy Huddle, having invented the coated particle in Harwell 1957, stated in the early 1970s that we know now everything about particles and coatings and should be going over to deal with other problems. This was on the occasion of the Dragon fuel performance information meeting London 1973: How wrong a genius be! It took until 1978 that really good particles were made in Germany, then during the Japanese HTTR production in the 1990s and finally the Chinese 2000-2001 campaign for HTR-10. Here, we present a review of history and present status. Today, good fuel is measured by different standards from the seventies: where $9*10^{-4}$ initial free heavy metal fraction was typical for early AVR carbide fuel and $3*10^{-4}$ initial free heavy metal fraction was acceptable for oxide fuel in THTR, we insist on values more than an order of magnitude below this value today. Half a percent of particle failure at the end-of-irradiation, another ancient standard, is not even acceptable today, even for the most severe accidents. While legislation and licensing has not changed, one of the reasons we insist on these improvements is the preference for passive systems rather than active controls of earlier times. After renewed HTGR interest, we are reporting about the start of new or reactivated coated particle work in several parts of the world, considering the aspects of designs/ traditional and new materials, manufacturing technologies/ quality control quality assurance, irradiation and accident performance, modeling and performance predictions, and fuel cycle aspects and spent fuel treatment. In very general terms, the coated particle should be strong, reliable, retentive, and affordable. These properties have to be quantified and will be eventually optimized for a specific application system. Results obtained so far indicate that the same particle can be used for steam cycle applications with $700-750^{\circ}C$ helium coolant gas exit, for gas turbine applications at $850-900^{\circ}C$ and for process heat/hydrogen generation applications with $950^{\circ}C$ outlet temperatures. There is a clear set of standards for modem high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a $500{\mu}m$ diameter $UO_2$ kernel of 10% enrichment is surrounded by a $100{\mu}m$ thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of $35{\mu}m$ thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum $1600^{\circ}C$ afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modem coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond $1600^{\circ}C$ for a short period of time. This work should proceed at both national and international level.