• Journal of the Korean Society of Safety
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• v.30 no.4
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• pp.168-173
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• 2015

A steam generator tubing is one of the main pressure boundary of the reactor coolant system in the nuclear power plants. Structural integrity refers to maintaining adequate margins against failure of the tubing. Burst pressure of a tube at tube support plate can be higher than that for a free-span tube because failure behaviors could be interfered from the tube support plate. Alternative repair criteria for out-diameter stress corrosion cracking indications in tubes to the drilled type tube support plate were developed, however, there are very limited information to the eggcrate type tube support plate. This paper discussed reinforcement effect of steam generator tube burst pressure with axial out-diameter stress corrosion cracking within an eggcrate type tube support plate. A series of tube burst tests were performed under the room temperature and it was found out that there is no significant but marginal effects.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.43-49
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• 2010

The piping failure probability of the reactor coolant system in Kori unit 1 was evaluated considering stress corrosion cracking. The P-PIE program (Probabilistic Piping Integrity Evaluation Program) developed in this study was used in the analysis. The effect of some variables such as oxygen concentration during start up and steady state operation, and operating temperature, which are related with stress corrosion cracking, on the piping failure probabilities was investigated. The effects of leak detection capability, the size of big leak, piping loops, and reactor types on the piping failure probability were also investigated. The results show that (1) LOCA (loss of coolant accident) probability of Kori unit 1 is extremely low, (2) leak probability is sensitive to oxygen concentration during steady state operation and operating temperature, while not sensitive to the oxygen concentration during start up, and (3) the piping thickness and operating temperature play important roles in the leak probabilities of the cold leg in 4 reactor types having same inner diameter.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.04a
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• pp.216-216
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• 2014

과거 유체 유발 진동(FIV : Fluid Induced Vibration)은 배관계 설계 하중에 고려되지 않은 설계 하중이었다. 하지만, 원자력 발전소 또는 화력 발전소의 배관형상이 복잡하고 고온수가 배관 내부에서 유동하는 배관계에서 육안으로 관측이 가능한 배관진동이 발생하였다. 이에 배관 진동에 대하여 원인 분석과 배관 구조 건전성 평가에 관심을 가지게 되었다. 배관 진동은 배관 형상에 따라 배관 내부 난류 유동에 대한 압력 변동이 하나의 원인이며, 고온수가 유동하는 배관일수록 압력 변동에 대한 배관 진동이 크게 나타나는 것으로 분석되었다. 배관 내부 난류 유동에 대한 압력 변동을 불규칙 수력하중이라고 한다. 본 연구에서는 배관 내부에서 난류 유동으로 발생하는 불규칙 수력하중을 유동해석을 이용하여 PSD(Power Spectral Density)로 산출하고, PSD 하중을 이용하여 불규칙 구조 응답 해석을 수행하여 배관계 응력 분포에 대하여 연구하였다. 배관 내부 난류 유동에 대한 불규칙 수력하중은 DES 난류 모델을 사용하여 시간에 대한 배관 내부 표면의 유체 속도를 유동 해석으로 산출하였으며, 유체 속도를 동압으로 계산한 후 FFT(Fast Fourier Transform)를 수행하여 PSD 하중으로 산출하였다. 그리고 불규칙 구조 응답 해석에서 배관 내부 유체 영향에 대한 진동 감쇠를 표현하기 위하여 유체 질량을 산출하고, 배관 구조 해석 모델 표면에 질량을 입력하는 방법으로 배관 고유진동수 및 불규칙 구조 응답 해석을 수행하였다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Journal of the Korean Society of Systems Engineering
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• v.10 no.1
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• pp.49-56
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• 2014

Since steam generator (SG) tubes are located in the boundary between the primary and secondary systems of nuclear power plant (NPP), the SG is one of the most important components in the aspects of the safety of NPP. The magnetite ($Fe_30_4$) deposition, so-called fouling, is generally known as a major aging mechanism of CANDU SGs, and this aging mechanism makes the heat transfer efficiency between the primary and secondary systems of NPP reduced. Therefore, the development of SG safety assessment system which can evaluate the effect of the SG aging degradation mechanism should be needed for safety of NPP. In this study, through the suggestion of the guideline for SG safety assessment, it is possible to strengthen the basic of establishing the effective SG aging management technique. The SG safety assessment is carried out by CATHENA(Canadian Algorithm for THErmalhydraulic Network Analysis). It is possible to determine the integrity of SGs by identifying the main safety parameters which can be changed by the aging degradation of CANDU SGs.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.4
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• pp.409-416
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• 2009

Standard procedures for a fracture toughness testing require very severe restrictions for the specimen geometry to eliminate a size effect on the measured properties. Therefore, the used standard fracture toughness data results in the integrity assessment being irrationally conservative. However, a realistic fracture in general structures, such as in nuclear power plants, may develop under the low constraint condition of a large scale yielding with a shallow surface crack. In this paper, cleavage fracture toughness tests have been made on side-grooved PCVN (precracked charpy V-notch) type specimens (10 by 10 by 55 mm) with various crack depths. The constraint effects on the crack depth ratios were evaluated quantitatively by the developed scaling method using the 3-D finite element method. After the fracture toughness correction from scaling model, the statistical size effects were also corrected according to the standard ASTM E 1921 procedure. The results were evaluated through a comparison with the $T_0$ of the standard CT specimen. The corrected $T_0$ for all of the PCVN specimens showed a good agreement to within $5.4^{\circ}C$ regardless of the crack depth, while the averaged PCVN $T_0$ was $13.4^{\circ}C$ higher than the real CT test results.

• KSII Transactions on Internet and Information Systems (TIIS)
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• v.8 no.11
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• pp.3937-3954
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• 2014

With the increasing electricity consumption and the wide application of renewable energy sources, energy auction attracts a lot of attention due to its economic benefits. Many schemes have been proposed to support energy auction in smart grid. However, few of them can achieve range query, ranked search and personalized search. In this paper, we propose an efficient multi-keyword range query (EMRQ) scheme, which can support range query, ranked search and personalized search simultaneously. Based on the homomorphic Paillier cryptosystem, we use two super-increasing sequences to aggregate multidimensional keywords. The first one is used to aggregate one buyer's or seller's multidimensional keywords to an aggregated number. The second one is used to create a summary number by aggregating the aggregated numbers of all sellers. As a result, the comparison between the keywords of all sellers and those of one buyer can be achieved with only one calculation. Security analysis demonstrates that EMRQ can achieve confidentiality of keywords, authentication, data integrity and query privacy. Extensive experiments show that EMRQ is more efficient compared with the scheme in [3] in terms of computation and communication overhead.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.54-60
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• 2010

Piping failures due to thermal fatigue have been widely reported in normally stagnant non-isolable reactor coolant branch lines. Since the thermal fatigue due to thermal stratification was not considered in the piping fatigue design in old NPPs, it is important to evaluate the effect of thermal stratification on the integrity of branch lines. In this study, geometrical screening criteria for Up-horizontal branch lines in MRP-132 were applied to SI(Safety Injection) lines of KSNP 2-loop and WH 3-loop. Some computational fluid dynamic(CFD) analyses on the Reactor Coolant System(RCS) branch lines were also performed to develop the regulatory guidelines for screening criteria. As a result of applying MRP-132 screening criteria, KSNP 2-loop and WH 3-loop SI lines are determined to need further detailed evaluation. Results of CFD analyses show that both valve isolation and amount of leakage through valve can be used as technical bases for the screening criteria on the thermal fatigue analysis.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.61-68
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• 2017

Reactor pressure vessel (RPV) under severe accident conditions accompanied by core melting is exposed to direct high-temperature thermal loads. Understanding the creep behavior of the material is one of the most important factors for evaluating the structural integrity at these conditions. While damage evaluation studies have been conducted on critical structures of nuclear power plants through finite element (FE) analyses considering creep behavior, for accurate creep damage evaluation, constitutive equations considered in the FE analyses may have different results depending on the time hardening and strain hardening models as well as the tertiary creep consideration. The purpose of this study is to evaluate the creep damage under severe accident conditions by using FE method for a representative domestic RPV material, SA508 Gr.3. The effect of material hardening models and constitutive equations which are the main variables were also investigated.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.236-245
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• 2016

In a reactor coolant system of a nuclear power plant (NPP), an overpressure protection system keeps pressure in the loop within 110% of design pressure. However if the system does not work properly, pressure in the loop could elevate hugely in a short time. It would be seriously disastrous if a weak point in the pressure boundary component bursts and releases radioactive material within the containment; and it may lead to a leak outside the containment. In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM) analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization. It revealed which component was the weakest and which component had the highest margin to bursting failure. This information is valuable in severe accident progression prediction. The burst pressures of APR-1400, AP1000 and VVER-1000 reactor coolant systems were evaluated and compared to give relative margins of safety.