• Journal of Radiation Protection and Research
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• 제27권2호
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• pp.95-100
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• 2002

원자력발전소에서 주로 발생되는 금속폐기물인 탄소강을 중성염전해질인 1.7M의 황산나트륨($Na_2SO_4$)과 질산나트륨($NaNO_3$)을 이용하여 기존전해제염과 개선된 전해제염공정의 비교실험을 수행하였다. 양극은 인코넬, 음극은 티타늄으로 하여 상온에서 1시간동안 반응시켜 금속폐기물 모재의 weight loss, 두께변화. 전해질 내 침전물농도, SEM을 이용하여 제염전후의 금속폐기물 표면의 형상을 분석하였다. 실험결과 개선된 전해제염 적용시 전해질 종류별 전류밀도 변화에 따른 실험에서는 전류밀도가 $0.1{\sim}0.6A/cm^2$으로 증가함에 따라 1.7M의 황산나트륨 적용시 금속폐기물 모재의 두께변화는 $0.48{\pm}0.005{\sim}67.7{\pm}0.02{\mu}m$, 1.7M의 질산나트륨 적용시에는 $0.06{\pm}0.005{\sim}17.7{\pm}0.05{\mu}m$로 나타나 같은 전류밀도에서 황산나트륨 적용시 금속폐기물 모재의 표면 제염효율이 더욱 높은 양상을 보였다. 또한 전류밀도 $0.3A/cm^2$ 및 1.7M의 황산나트륨의 조건에서 개선된 전해제염 적용 시 $9.8{\pm}0.01{\mu}m$의 금속폐기물 두께변화를 보여 기존전해제염 적용시인 $3.7{\pm}0.03{\mu}m$의 금속폐기물 두께변화보다 2배 이상의 표면 제염효과를 보였다.

• 전기학회논문지
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• 제62권7호
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• pp.994-1002
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• 2013

This study suggests a generalized availability and safety evaluation model to evaluate the influences to the system's fault tolerant capabilities depending on automatic fault detection function such as the automatic periodic testings. The conventional evaluation model of automatic fault detection function deals only with the self diagnostics, and supposes that the fault detection coverage of self diagnostics is always constant. But all of the fault detection methods could be degraded. For example, the periodic surveillance test has the potential human errors or test equipment errors, the self diagnostics has the potential degradation of built-in logics, and the automatic periodic testing has the potential degradation of automatic test facilities. The suggested evaluation models have incorporated the loss or erroneous behaviors of the automatic fault detection methods. The availability and the safety of each module of the safety grade platform have been evaluated as they were applied the automatic periodic test methodology and the fault tolerant evaluation models. The availability and safety of the safety grade platform were improved when applied the automatic periodic testing. Especially the fault tolerant capability of the processor module with a weak self-diagnostics and the process parameter input modules were dramatically improved compared to the conventional cases. In addition, as a result of the safety evaluation of the digital reactor protection system, the system safety of the digital parts was improved about 4 times compared to the conventional cases.

• 수산경영론집
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• 제50권2호
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• pp.23-40
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• 2019

The main purpose of this study is to analyze and examine the problems of the law systems of the safety and maintenance of nuclear facilities and to propose the improvements with respect to the related problems especialy focused on safety management of aquatic products. Therefore, the results of the paper would be helpful to build an effective management law system of safety and maintenance of nuclear facilities and fisheries products. The research methods are longitudinal and horizontal studies. This study compares domestic policies with foreign policies of nuclear plants and aquatic products. Using the above methods, examining the current system of nuclear-related laws and regulations, we have found that there exist 13 Acts including "Nuclear Safety Act", etc. Safety laws related on nuclear facilities have seven Acts including "Nuclear Safety Act", "the Act on Physical Protection and Radiological Emergency", "Radioactive waste control Act", "Act on Protective Action Guidelines against Radiation in the Natural Environment", "Special Act on Assistance to the locations of facilities for disposal low and intermediate level radioactive waste", "Korea Institute of Nuclear Safety Act". "Act on Establishment and Operation of the Nuclear Safety and Security Commission". The seven laws are composed of 119 legislations. They have 112 lower statute of eight Presidential Decrees, six Primeministrial Decrees and Ministrial Decrees, 92 administrative rules (orders), 6 legislations of local self-government aself-governing body. The concluded proposals of this paper are as follows. Firstly, we propose that the relationship between the special law and general law should be re-established. Secondly, the terms with respect to law system of safety and maintenance of nuclear plants should be redefined and specified. Thirdly, it is advisable to re-examine and re-establish the Law System for Safety and Maintenance of Nuclear Facilities. and environmental rights like the French Nuclear Safety Legislation. Lastly, inadequate legislation on the aquatic pollution damage should be re-established. It is necessary to ensure sufficient transparency as well as environmental considerations in the policy decisions of the Korean government and legislation of the National Assembly. It is necessary to further study the possibilities of accepting the implications of the French legal system as a legal system in Korea. In conclusion, the safety management of nuclear facilities is not only focused on the secondary industry and the tertiary industry centering on power generation and supply, but also on the primary industry, which is the food of the people. It is necessary to prevent damage to be foreseen. Therefore, it is judged that there should be no harm to the people caused by contaminated marine products even if the "Food Safety Law for Prevention of Radiation Pollution Damage" is enacted.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.921-926
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• 2001

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. In this study, a thorough investigation on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. A design of the control rod position indication system using reed switch for the CEDM on the system-integrated reactor SMART was developed based on the position indicator technology identified through the investigation. The feasibility of the design was evaluated by test of manufactured control rod position indicator using reed switch for SMART.

• Journal of Radiation Protection and Research
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• 제22권4호
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• pp.309-317
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• 1997

비파괴적인 방법으로 방사성폐기물드럼에 대한 핵종분석을 수행할 때 드럼내 매질에 의한 방사선의 감쇠에 의해 핵종분석장치로 측정한 계수값은 실제 드럼내 방사능에 의한 계수값보다 작게 나타나 결과적으로 방사능의 측정결과가 과소평가된다. 그러므로 드럼내 매질에 의한 감쇠를 보정해 주어야 하는데 감쇠 보정방법은 드럼내 매질의 분포나 매질의 밀도에 따라 달리 적용해야 한다. 본 연구에서는 원자력발전소에서 발생하는 드럼 종류별로 모델드럼을 제작한 후 모델드럼에 표준감마선원을 넣고 핵종분석장치를 이용해 측정을 하여 드럼내 매질의 밀도를 구하였고, 이 값을 실제 매질의 밀도와 비교해 드럼종류에 따라 매질에 의한 최적의 감쇠 보정방법을 결정하였다. 그 결과 잡고체드럼의 경우에는 전송선원 보정방법과 평균밀도 보정방법, 차폐잡고체드럼의 경우 전송선원 보정방법과 두 감마선피크비교 보정방법이 최적의 감쇠 보정방법임을 알 수 있었고, 고밀도드럼인 폐수지, 농축폐액 및 폐필터드럼의 경우에는 평균밀도 보정방법과 두 감마선피크비교 보정방법을 사용해 드럼내 매질의 감쇠를 보정할 수 있다.

• Corrosion Science and Technology
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• 제14권4호
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• pp.177-183
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• 2015

Prestressed Concrete steel Cylinder Pipe (PCCP) is extensively used as seawater pipes for cooling in nuclear power plants. The internal surface of PCCP is exposed to seawater, while the external surface is in direct contact with underground soil. Therefore, materials and strategies that would reduce the corrosion of its cylindrical steel body and external steel wiring need to be employed. To prevent against the failure of PCCP, operators provided a cathodic protection to the pre-stressing wires. The efficiency of cathodic protection is governed by the anodic performance of the system. A mixed metal oxide (MMO) electrode was developed to meet criteria of low over potential and high corrosion resistance. Increasing coating cycles improved the performance of the anode, but cycling should be minimized due to high materials cost. In this work, the effects of $RuCl_3$ concentration on the electrochemical properties and lifespan of MMO anode were evaluated. With increasing concentration of $RuCl_3$, the oxygen evolution potential lowered and polarization resistance were also reduced but demonstrated an increase in passive current density and oxygen evolution current density. To improve the electrochemical properties of the MMO anode, $RuCl_3$ concentration was increased. As a result, the number of required coating cycles were reduced substantially and the MMO anode achieved an excellent lifespan of over 80 years. Thus, we concluded that the relationship between $RuCl_3$ concentration and coating cycles can be summarized as follows: No. of coating cycle = 0.48*[$RuCl_3$ concentration, $M]^{-0.97}$.

• 에너지공학
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• 제25권1호
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• pp.15-22
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• 2016

월성 2,3,4호기의 비정상 운전 중 열수송계통의 과도변화해석이 수행되었다. 중수로에 대한 캐나다의 규제문서인 AECB R-77 요구조건에 대한 만족성을 평가하였다. 해석 결과 여러 비정상 운전시 과도변화에 의한 원자로 모관의 최고압력값은 ASME 코드의 제한치 이내로 만족되었다. 고압시 보호장치인 LRV의 영향은 미미한 것으로 나타났다.

• Journal of Radiation Protection and Research
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• 제34권3호
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• pp.129-136
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• 2009

국내 원전의 계획예방정비기간 중에 원자로계통의 개방과정에서 원자로건물내 공기 중으로 누설된 $^{131}I$의 체내 흡입으로 원전종사자의 내부피폭이 발생하였다. 이에 따라 원전에서 보유하고 있는 전신계측기(Whole body counter)를 이용하여 내부방사능을 측정하였다. 이들 측정값을 근거로 국제방사선방호위원회(ICRP)의 내부피폭 선량평가 지침을 적용하여 섭취량을 산정하고, 내부 피폭 방사선량을 평가하였다. $^{131}I$은 체내에서 섭취와 배설이 빠르고 갑상선으로 재축적이 일어나기 때문에 섭취 후 측정시점에 따라 섭취량이 차이를 보였다. 또한 ICRP 간행물에서 $^{131}I$의 전선에 대한 섭취잔류분율 자료를 제공하고 있지 않아 갑상선 섭취잔류분율 자료를 이용함으로써 섭취량 평가에서 오차를 나타내었다. 이에 따라 수계산과정으로 섭취량을 산정하고 예탁유효선량을 평가하였다. 한편 전선에 대한 섭취잔류분율을 새로 계산하였으며, 이 결과를 검증하였다. 또한 국제적으로 이용되고 있는 내부 피폭 선량평가 전신코드들 이용하여 섭취량 산정과 내부피폭 선량평가 평가결과에 대한 비교 계산이 병행하여 이루어졌다.

• 한국안전학회지
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• 제16권3호
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• pp.125-134
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• 2001

The Advanced Pressurized rater Reactor 1400(APR1400) system is advanced of the successful Korean Nuclear Power Plants(KSNP) design which meets functional needs for safety enhancement reliability improvement, and control in the human-computer monitoring system. Therefore this paper describes the scoring model in order to justify the reliability and safety in APR 1400 under uncertainty. The structure of this paper consists of the human engineering, risk safety, quality function, safety organization management factors of the qualitative factors in chapter 2, and the expectation results of the normalized scoring model in chapter 3. Finally, the proposed reliability model have provided the technical flexibility not only for functional control fields but also for accidents protection systems in APR 1400 under uncertainty.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.110-116
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• 2020

As the management plan for domestic spent nuclear fuel is delayed, the storage of the operating nuclear power plant is approaching saturation, and the Kori 1 Unit that has reached its end of operation life is preparing for the dismantling plan. The first stage of dismantling is the transfer of spent nuclear fuel stored in storage at plants. The spent fuel management process leads to temporary storage, interim storage, reprocessing and permanent disposal. In this paper, the technical issues to be considered when transporting spent fuel in this process are summarized. The spent fuels are treated as high-level radioactive waste and strictly managed according to international regulations. A series of integrity tests are performed to demonstrate that spent fuel can be safely stored for decades in a dry environment before being transferred to an intermediate storage facility. The safety of spent fuel transport container must be demonstrated under normal transport conditions and virtual accident conditions. IAEA international standards are commonly applied to the design of transport containers, licensing regulations and transport regulations worldwide. In addition, each country operates a physical protection system to reduce and respond to the threat of radioactive terrorism.