Sun Yeong Choi

Safety evaluation of socket weld integrity in nuclear piping

The purposes of this paper are to evaluate the integrity of socket weld in nuclear piping and prepare the technical basis for a new guideline on radiographic testing (RT) for the socket weld. Recently, the integrity of the socket weld is regarded as a safety concern in nuclear power plants because lots of failures and leaks have been reported in the socket weld. The root causes of the socket weld failure are known as unanticipated loadings such as vibration or thermal fatigue and improper weld joint during construction. The ASME Code Sec. III requires 1/16 inch gap between the pipe and fitting in the socket weld. Many failure cases, however, showed that the gap requirement was not satisfied. The Code also requires magnetic particle examination (MT) or liquid penetration examination (PT) on the socket weld, but not radiographic examination (RT). It means that it is not easy to examine the 1/16 inch gap in the socket weld by using the NDE methods currently required in the Code. In this paper, the effects of the requirements in the ASME Code Sec. III on the socket weld integrity were evaluated by using finite element method. The crack behavior in the socket weld was also investigated under vibration event in nuclear power plants. The results showed that the socket weld was very susceptible to the vibration if the requirements in ASME Code were not satisfied. The constraint between the pipe and fitting due to the contact significantly affects the integrity of the socket weld. This paper also suggests a new guideline on the RT for the socket weld during construction stage in nuclear power plants.

Evaluation of Nuclear Piping Failure Frequency in Korean Pressurized Water Reactors

The purpose of this paper is to evaluate the piping failure frequency based on the piping failure events in Korean pressurized water reactors (PWRs) until the end of 2003. Two types of the piping failure frequencies including the piping damage frequency and the piping rupture frequency are considered in this study. The piping damage frequency for the failed piping system was estimated by using the piping population data such as the weld count or the base metal count. The piping rupture frequency related to the initiating event in a probabilistic safety assessment (PSA) was evaluated by using both the Bayesian approach (Method 1) and the conditional rupture probability approach (Method 2). In the Bayesian approach, two methods using Jeffreys noninformative prior (Method 1-1) and prior distributions based on the results in NUREG/CR-5750 (Method 1-2) were considered. Thirty piping failure events in ASME safety class pipings of Korean PWRs were identified and analyzed in this study. The results showed that the piping damage frequency for the events ranged from 5.42E-3/cr.yr to 2.77E-5/cr.yr. Three kinds of initiating events including the very small LOCA, the feedwater line break, and the flood are evaluated for Korean PWRs. The results for the piping rupture frequency in Korean PWRs were as follows: 1) The mean piping rupture frequency of the very small LOCA event ranged from 3.6E-3/cr.yr to 1.2E-2/cr.yr, the feedwater line break event from 3.6E-3/cr.yr to 2.5E-2/cr.yr, and the flood event from 7.8E-4/cr.yr to 3.6E-3/cr.yr. The mean piping rupture frequencies of the very small LOCA and feedwater line break events were higher than that of the flood event by one order of a magnitude. 2) Method 2 gave conservative results in the very small LOCA and feedwater line break events compared to Method 1-1 or Method 1-2, while Method 1-1 gave conservative results in the flood event. 3) The order of magnitudes in the mean piping rupture frequencies of the very small LOCA, the feedwater line break, and the flood in Korean PWRs were similar to those in the U.S. PWRs.

A New Strategy for In-Service Inspection of Nuclear Piping Considering Piping Failure Frequency

The current in-service inspection (ISI) strategy for the nuclear piping in many countries consists of both the code requirements such as ASME B & PV Code Sec. XI and the country-specific regulatory requirements, so called as the enhanced ISI. The enhanced ISI reflects the operating experience of piping failure, while the ASME Code Sec. XI requirement is based on random sampling for the inspection points. In this study, a new strategy for ISI of nuclear piping was proposed based on piping failure frequency. This strategy basically reflects the operating experience because the piping failure frequency is based on the piping failure database. The new concept of minimum inspection rate was also introduced in this new ISI strategy. As pilot study, the new ISI strategy was applied to the Class 1 piping system such as reactor coolant system and safety injection system of Ulchin Unit 5 which is the 1,000 MWe Korean Standard PWR. The results from the proposed new strategy were compared to those from the ASME Code Sec. XI. The results show that the new ISI strategy reasonably reflects the operating experience. The results also show that the concept of the minimum inspection rate can compensate the unbalance in the number of inspection points between the very large differences in the piping failure frequency.

Some Applications of Piping Failure Database to Nuclear Safety Issues in Korea

Korean applications of a piping failure database to some nuclear safety issues related to nuclear piping such as a domestic initiating event frequency for a probabilistic safety assessment (PSA), an enhanced in-service inspection (ISI) program, a new strategy for the ISI program, leak before break (LBB) screening criteria, a risk-informed ISI (RI-ISI) program, and aging management for a periodic safety review (PSR) and/or a life extension are introduced in this study. By using a piping rupture frequency and a piping damage frequency obtained from a piping failure database, a domestic initiating frequency for a PSA, an enhanced ISI program, and a new strategy for an ISI are established. Some safety issues such as the LBB screening criteria, ISI program, RI-ISI program, and the aging management of nuclear piping are reviewed based on the information and insights obtained from a case-by-case analysis of the piping failure events in the piping failure database.Copyright

Surface Crack Behavior in Socket Weld of Nuclear Piping under Fatigue Loading Condition

The ASME B & PV Code Sec. allows the socket weld for the nuclear piping in spite of the weakness on the weld integrity. Recently, the integrity of the socket weld is regarded as a safety concern in nuclear power plants because many failures and leaks have been reported in the socket weld. OPDE (OECD Piping Failure Data Exchange) database lists 108 socket weld failures among 2,399 nuclear piping failure cases during 1970 to 2001. Eleven failures in the socket weld were also reported in Korean NPPs. Many failure cases showed that the root cause of the failure is the fatigue and the gap requirement for the socket weld given in ASME Code was not satisfied. The purpose of this paper is to evaluate the fatigue crack behavior of a surface crack in the socket weld under fatigue loading condition considering the gap effect. Three-dimensional finite element analysis was performed to estimate the fatigue crack behavior of the surface crack. Three types of loading conditions such as the deflection due to vibration, the pressure transient ranging from P=0 to 15.51MPa, and the thermal transient ranging from T=25oC to 288oC were considered. The results are as follows; 1) The socket weld is susceptible to the vibration where the vibration levels exceed the requirement in the ASME Operation and Maintenance (OM) Code. 2) The effect of pressure or Temperature transient load on the socket weld integrity is not significant. 3) No-gap condition gives very high possibility of the crack initiation at the socket weld under vibration loading condition. 4) For the specific systems having the vibration condition to exceed the requirement in the ASME Code OM and/or the transient loading condition from P=0 and T=25oC to P=15.51MPa and T=288oC, radiographic examination to examine the gap during the construction stage is recommended.

Piping Failure Analysis for the Korean Nuclear Piping Including the Effect of In-Service Inspection

The purposes of this paper are to perform piping failure analysis for the failed safety class piping in Korean nuclear power plants(NPPs) and evaluate the effect of an in-service inspection(ISI) on the piping failure probability. For data collection, a database for piping failure events was constructed with 135 data fields including population data, event data, and service history data. A total of 6 kinds of events with 25 failure cases up to June 30, 2003 were identified from Korean NPPs. The failed systems were main feedwater system, CVCS, primary sampling system, essential service water system, and CANDU purification system. Piping failure analyses such as evaluation of the impact on nuclear safety and piping integrity and the root cause analysis were performed and the piping failure frequencies for the failed piping were calculated by using population data. The result showed that although the integrity was not maintained in the failed piping, the safety of the plants was maintained for all the events. And the root causes of the events were analyzed as FAC, vibration, thermal fatigue, corrosion, and/or an improper weld joint. The piping failure frequencies ranged from 6.08E-5/Cr-Yr to 1.15E-3/Cr-Yr for the events. According to the ASME Code Sec. XI requirements, the small bore piping less than the nominal diameter of 4 inch is exempt from ISI. There, however, were many piping failures reported in the small bore piping. The effect of ISI considering the pipe size on the piping failure probability was investigated by using the Win-PRAISE program based on probabilistic fracture mechanics. The results showed that there is no significant difference between the small and large bore piping from the viewpoint of the ISI effect on the piping failure probability. It means that ISI for a small bore piping is recommended as well as the large bore piping. Introduction During the past decades, lots of piping failures have been reported world-wide in nuclear power plants (NPPs).[1~5] A total of 4,064 piping failure cases in the United States were reported during 1961 to 1997.[4] OPDE(OECD Piping Failure Data Exchange) database(rev. 0) shows 1,181 cases of nuclear piping failures during 1970 to 2001 in 28 countries except the US.[5] Some piping failures were also reported in Korean NPPs. In order to prevent further similar piping failure, piping failure analyses are generally required. The results obtained from the piping failure analyses can be used in some applications for plant design and maintenance such as in-service inspection(ISI), leak before break concept, aging management, re-evaluation of the initiating event frequency for probabilistic safety analysis and risk-informed regulation. [6,7] In this study, a database was developed for piping failure events in Korean NPPs, and the piping failure events up to June 30, 2003 in Korean NPPs were collected. The impact of the piping failure on the component integrity and plant safety was evaluated including the root cause analysis. Based on the plant population data, the piping failure frequency was estimated. According to the ASME Code Sec. XI requirements, the small bore piping less than the nominal diameter of 4 inch is exempt from ISI.[8] However, many piping failures in the small bore piping were reported.[1~5] Because an additional ISI for small bore piping is one of the current regulatory safety issues,[9] the effect of ISI including Key Engineering Materials Online: 2004-08-15 ISSN: 1662-9795, Vols. 270-273, pp 1731-1736 doi:10.4028/www.scientific.net/KEM.270-273.1731