• Nuclear Engineering and Technology
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• 제28권6호
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• pp.576-582
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• 1996

The integrity of reactor vessel support system of the Korean Standard Nuclear Power Plant (KSNPP) is investigated for a postulated reactor vessel closure head drop event. The closure head is disassembled from the reactor vessel during refueling process or general inspection of reactor vessel and internal structures, and carried to proposed location by the head lift rig. A postulated closure head drop event could be anticipated during closure head handling process. The drop event may cause an impact load on the reactor vessel and supporting system. The integrity of the supporting system is directly relevant to that of reactor vessel and reactor internals including fuels. Results derived by elastic impact analysis, linear and non-linear buckling analysis and elasto-plastic stress analysis of the supporting system implied that the integrity of the reactor vessel supporting system is intact for a postulated reactor vessel closure head drop event.

• 한국항해항만학회:학술대회논문집
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• 한국항해항만학회 2006년도 International Symposium on GPS/GNSS Vol.1
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• pp.417-421
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• 2006

ICAO defines performance requirements of navigation system such as accuracy, integrity, continuity and availability. The integrity is most significant performance requirement in aviation where safety of life is crucial. Many researches on this topic anticipate that GPS with SBAS or Galileo can meet APV requirements and GPS with GBAS or Galileo with GBAS will meet CAT II and III requirements. These performance expectations are based on global analysis. In this paper regional integrity analysis in Korea using various combinations of modernized GPS, Galileo and SBAS is given. The simulation results show that CAT I requirement can be met using modernized GPS and Galileo alone, however, CAT II and III are not met even augmenting SBAS because of VPL. A more efficient augmentation such as GBAS which can reduce VPL dramatically is required to meet CAT II and III in Korean region.

• ETRI Journal
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• 제41권5호
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• pp.670-683
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• 2019

In this paper, we analyzed the signal integrity of a system interconnection module for a proposed high-density server. The proposed server integrates several components into a chassis. Therefore, the proposed server can access multiple computing resources. To support the system interconnection, among the highly integrated computing resources, the interconnection module, which is based on Serial RapidIO, has been newly adopted and supports a bandwidth of 800 Gbps while routing 160 differential signal traces. The module was designed for two different stack-up types on a printed circuit board. Each module was designed into 12- (version 1) and 14-layer (version 2) versions with thicknesses of 1.5T and 1.8T, respectively. Version 1 has a structure with two consecutive high-speed signal-layers in the middle of two power planes, whereas Version 2 has a single high-speed signal placed only in the space between two power planes. To analyze the signal integrity of the module, we probed the S-parameters, eye-diagrams, and crosstalk voltages. The results show that the high-speed signal integrity of Version 2 has a better quality than Version 1, even if the signal trace length is increased.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.1307-1308
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.1301-1302
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• 2004

• Journal of Mechanical Science and Technology
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• 제17권7호
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• pp.947-957
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• 2003

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle. In order to complete the integrity evaluation of pressure tube, expert knowledge, iterative calculation procedures and a lot of input data are required. More over, results of integrity assessment may be different according to the evaluation method. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached database, was developed. The present system was built on the basis of 3D FEM results, ASME Sec. XI, and Fitness For Service Guidelines for CANDU pressure tubes issued by the AECL (Atomic Energy Canada Limited). The present system also covers the delayed hydride cracking and the blister evaluation, which are the characteristics of pressure tube integrity evaluation. In order to verify the present system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.39-46
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• 2014

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

• 방사성폐기물학회지
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• 제20권3호
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• pp.357-364
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• 2022

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62℃, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446℃ lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

• Nuclear Engineering and Technology
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• 제42권3호
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• pp.237-248
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• 2010

The safety of metal cask and concrete cask storage technology has been verified by CRIEPI through several research programs on demonstrative testing for the interim storage of spent fuel. The results have been reflected in the safety requirements for dry casks issued by NISA/METI (Nuclear and Industrial Safety Agency, Ministry of Economy, Trade and Industry) of the Japanese government. On top of that, spent fuel integrity has been studied by the Japan Nuclear Energy Safety Organization (JNES). This paper reviews these research programs. Future issues include the long-term integrity of cask components and high burn-up spent fuel.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.264-267
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• 2004

Wall thinning of carbon steel pipe components due to Flow-Accelerated Corrosion (FAC) is one of the most serious threats to the integrity of steam cycle piping systems in Nuclear Power Plants (NPP). Since the mid-1990s, secondary side piping systems in Korean NPPs have experienced wall thinning, leakages and ruptures caused by FAC. Korea Electric power Research Institute (KEPRI) and Korea Hydro & Nuclear Power Co., LTD. (KHNP) have conducted a study to develop the methodology for systematic pipe management and established the Korean Thinned Pipe Management Program (TPMP). To effectively maintain the integrity of piping system, FAC engineer should understand the criterions of the structural integrity evaluation and the safety margin assessment for the thinned pipe component. This paper describes the technical items of TPMP, and shows the example of the integrity evaluation and safety margin assessment for three thinned pipe component of a NPP.