• Nuclear Engineering and Technology
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• 제49권3호
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• pp.651-661
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• 2017

The Korea Institute of Nuclear Safety, as a representative organization of Korea, in February 2012 participated in an international Program to Assess the Reliability of Emerging Nondestructive Techniques initiated by the U.S. Nuclear Regulatory Commission. The goal of the Program to Assess the Reliability of Emerging Nondestructive Techniques is to investigate the performance of emerging and prospective novel nondestructive techniques to find flaws in nickel-alloy welds and base materials. In this article, Korean round-robin test results were evaluated with respect to the test blocks and various nondestructive examination techniques. The test blocks were prepared to simulate large-bore dissimilar metal welds, small-bore dissimilar metal welds, and bottom-mounted instrumentation penetration welds in nuclear power plants. Also, lessons learned from the Korean round-robin test were summarized and discussed.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.317-326
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• 2005

The burning-out of excess plutonium from the reprocessing of spent nuclear fuel and from the dismantlement of nuclear weapons is recently emphasized due to the difficulties in securing the final repository for the spent fuel and the necessity to consume the ex-weapons plutonium. An irradiation test in the Halden reactor was launched by the OECD Halden Reactor Project (HRP) to investigate the in-pile behavior of plutonium-embedded fuel as a form of mixed oxide (MOX) and of inert matrix fuel (IMF). The first cycle of irradiation was successfully accomplished with good integrity of test fuel rods and without any undesirable fault of instrumentations. The test results revealed that the MOX fuel is more stable under irradiation environments than IMF. In addition, MOX fuel shows lower thermal resistance due to its better thermal conductivity than IMF. The on-line measured in-pile performance data of attrition milled MOX fuel are used in the analysis of the in-pile performance of the fuel with the fuel performance code, COSMOS. The COSMOS code has been developed for the analysis of MOX fuel as well as $UO_2$ fuel up to high burnup and showed good capability to analyze the in-reactor behavior of MOX fuel even with different instrumentation.

• 한국정밀공학회지
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• 제30권2호
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• pp.223-230
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• 2013

When a new nuclear fuel is developed, an irradiation test needs to be carried out in the research reactor to analyze the performance of the new nuclear fuel. In order to check the performance of a nuclear fuel during the irradiation test in the test loop of a research reactor, sensors need to be attached in and out of the fuel rod and connect them with instrumentation cables to the measuring device located outside of the reactor pool. In particular, to check the temporary temperature change at the center of a nuclear fuel during the irradiation test, a thermocouple should be instrumented at the center of the fuel rod. Therefore, a hole needs to be made at the center of fuel pellet to put in the thermocouple. However, because the hardness and the density of a sintered $UO_2$ pellet are very high, it is difficult to make a small fine hole on a sintered $UO_2$ pellet using a simple drilling machine even though we use a diamond drill bit made by electro deposition. In this study, an automated drilling machine using a CVD diamond drill has been developed to make a fine hole in a fuel pellet without changing tools or breakage of workpiece. A sintered alumina ($Al_2O_3$) block which has a higher hardness than a sintered $UO_2$ pellet is used as a test specimen. Then, it is verified that a precise hole can be drilled off without breakage of the drill bit in a short time.

• Journal of Radiation Protection and Research
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• 제33권1호
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• pp.1-12
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• 2008

최근 국제 표준은 글로벌 패권확보 수단으로, 또 국제 무역의 중요 수단으로 여겨질 정도로 중요성이 커지고 있다. 특히 표준과 규제정책을 연계하며 국제 표준의 구속력을 강화하고 있는 추세에 있어, 나라마다 국제표준 선점을 두고 치열한 경쟁을 벌이고 있다. 세계 각국은 점차 규격을 IEC에 부응하는 방향으로 바꾸고 있으며 국내에서는 원자력 분야의 국제전기위원회 (International Electrotechnical Commission)기술위원회인 TC45(Nuclear Instrumentation) 총회를 2005년에 유치하여 국제적 활동을 강화하고 있으며 수 년 전부터 IEC 규격을 적용한 국가규격으로 대체하고 있는 실정이다. 여기에서는 원자력 중장기사업의 일환으로 "방사선계측기 성능평가 기술개발" 연구를 수행하여 물리적 및 전자기적 환경에서의 성능평가 기술 및 방사선장특성 평가 기술 개발을 수행하였으며 IEC61526(Radiation protection instrumentation)의 규격을 적용하여 개발된 방사선 방호용 계측기의 성능평가 기술을 소개하고 이러한 성능시험에서 얻어진 결과를 분석하여 방사선계측기를 개발할 때 고려하여야 할 사항과 현재 국내에서 사용 중인 방사선계측기의 성능평가 결과를 분석하여 봄으로서 나타난 문제점 등을 토의하여 보았다.

• 전기학회논문지
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• 제65권5호
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• pp.836-842
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• 2016

When an accident occurs in the nuclear power plant, the faulted information might mislead to the high possibility of aggravating the accident. At the Fukushima accident, the operators misunderstood that there was no core exposure despite in the processing of core damage, because the instrument information of the reactor water level was provided to the operators optimistically other than the actual situation. Thus, this misunderstanding actually caused to much confusions on the rapid countermeasure on the accident, and then resulted in multiplying the accident propagation. It is necessary to be equipped with the function that informs operators the status of instrument integrity in real time. If plant operators verify that the instruments are working properly during accident conditions, they are able to make a decision more safely. In this study, we have performed various tests for the fault detection sensitivity of an data-driven empirical model to review the usability of the model in the accident conditions. The test was performed by using simulation data from the compact nuclear simulator that is numerically simulated to PWR type nuclear power plant. As a result of the test, the proposed model has shown good performance for detecting the specified instrument faults during normal plant conditions. Although the instrument fault detection sensitivity during plant accident conditions is lower than that during normal condition, the data-drive empirical model can be detected an instrument fault during early stage of plant accidents.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1181-1190
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• 2023

Hydraulic performance and flow resistance tests were performed to confirm the main parameters of the hydraulic instrumentation that can affect the pump performance of the reactor coolant pump. The flow resistance test offers important experimental data, which are necessary to predict the behavior of the primary coolant when the circulation of the reactor coolant pump is stopped. Moreover, the shape of the hydraulic section of the pump, which was considered in the test, was prepared to compare the mixed-flow- and axial-flow-type models, the difference in the number of blades of the impeller and diffuser, the difference in the shape of the impeller blade and its thickness, and the effect of coating at the suction bell. Additionally, five models of the hydraulic part were manufactured for the experiments. In this study, the differences in performance owing to the design factors were confirmed through the experimental results.

• 대한기계학회논문집A
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• 제37권12호
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• pp.1573-1579
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• 2013

핵연료의 연소성능을 시험하기 위해서는 시험 루프에 설치된 조사리그 내에 냉각수가 순환되도록 설계되어야 한다. 이때, 조사리그 내 냉각수는 $300^{\circ}C$, 15.5 MPa 의 고온 고압으로 순환시키기 때문에 냉각수의 밀봉은 핵연료 조사리그를 제작할 때 가장 중요한 공정 중 하나이다. 특히 15 개의 계장선이 조사리그의 압력경계부위를 통과하게 되는데, 이의 밀봉을 위해 일반적으로 브레이징 공정이 적용된다. 본 연구에서는 조사리그 브레이징용 진공챔버 및 고주파 유도가열기를 포함하는 유도 브레이징 시스템을 개발하고, 다양한 실험을 통해 산화막이 발생하지 않는 공정변수를 검토하였으며, 브레이징 제품의 인장시험, 단면검사, 밀봉성능검사 등을 통해 브레이징 공정의 건전성과 밀봉성능을 검증하였다.

• 전기학회논문지P
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• 제63권2호
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• pp.61-68
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• 2014

Function of battery bank stores energy for DC load in general, and DC power system of the nuclear power plant is used to supply DC loads for safety- featured instrumentation and control such as inverter, class 1E power system control and indication, and station annunciation. Class 1E DC power system must provide a power for the design basis accident conditions, and adequate capacity must be available during loss of AC power and subsequent safe shutdown of the plant. In present, batteries of Class 1E DC power system of the nuclear power plant uses lead-acid batteries. Class 1E batteries of nuclear power plants in Korea are summarized in terms of specification, such as capacity, discharge rate, bank configuration and discharge end voltage, etc. This paper summarizes standards of determining battery size for the nuclear power plant, and analyzes duty cycle for the class 1E DC power system of nuclear power plant. Then, battery cell size is calculated as 2613Ah according to the standard. In addition, this paper analyzes performance test results during past 13 years and shows performance degradation in the battery bank. Performance tests in 2001 and 2005 represent that entire battery cells do not reach the discharge-end voltage. Howeyer, the discharge-end voltage is reached in 14.7% of channel A (17 EA), 13.8% of channel B (16 EA), 5.2% of channel C (6 EA) and 16.4% of channel D (19 EA) at 2011 performance test. Based on the performance test results analysis and size calculation, battery capacity and degradation by age in Korearn nuclear power plant is discussed and would be used for new design.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4422-4430
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• 2022

Studying the effects of gamma radiation on the instrumentation and control (I&C) system of a nuclear power plant is critical to the successful and reliable operation of the plant. In the accidental scenario, the adverse environment of ionizing radiation affects the performance of the I&C system and it leads to inaccurate and incomprehensible results. This paper reports the effects of gamma radiation on the AT89C51RD2, a commercial-off-the-shelf 8-bit high-performance flash microcontroller. The microcontroller, selected for the device under test for this study is used in the remote terminal unit for a nuclear power plant. The custom circuits were made to test the microcontroller under different gamma doses using a ⁶⁰Co gamma source in both ex-situ and in-situ modes. The device was exposed to a maximum dose of 1.5 kGy. Under this hostile environment, the performance of the microcontroller was studied in terms of device current and voltage changes. It was observed that the microcontroller device can operate up to a total absorbed dose of approximately 0.6 kGy without any failure or degradation in its performance.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.242-242
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• 2002

Two MIMAS MaX fuel rods base-irradiated in a commercial PWR have been reinstrumented and irradiated at a test reactor. The fabrication data for two MOX roda are characterized together with base irradiation information. Both Rods were reinstrumented to be fitted with thermocouple to measure centerline temperature of fuel. One rod was equipped with pressure transducer for rod internal pressure whereas the other with cladding elongation detector. The post irradiation examinations for various items were performed to determine fuel and cladding in-pile behavior after base irradiation. By using well characterized fabrication and re-instrumentation data and power history, the fuel performance code, COSMOS, is verified with measured in-pile and PIE information. The COMaS code shows good agreement for the cladding oxidation and creep, and fission gas release when compared with PIE dad a after base irradiaton. Based on the re-instrumention information and power history measured in-pile, the COSMOS predicts re-instrumented in-pile thermal behaviour during power up-ramp and steady operation with acceptable accuracy. The rod internal pressure is also well simulated by COSMOS code. Therfore, with all the other verification by COSMOS code up to now, it can be concluded that COSMOS fuel performance code is applicable for the design and license for MaX fuel rods up to high burnup.