Refueling Outage Research Articles

Utilization of EPRI ChemWorks tools for PWR shutdown chemistry evolution modeling

Shutdown chemistry evolution is performed in nuclear power plants at each refueling outage (RFO) to establish safe conditions to open system and minimize inventory of corrosion products in the reactor coolant system (RCS). After hydrogen peroxide is added to RCS during shutdown chemistry evolution, corrosion products are released and are removed by filters and ion exchange resins in the chemical volume control system (CVCS). Shutdown chemistry evolution including RCS clean-up time to remove released corrosion products impacts the critical path schedule during RFOs. The estimation of clean-up time prior to RFO can provide more reliable actions for RCS clean-up operations and transients to operators during shutdown chemistry. Electric Power Research Institute (EPRI) shutdown calculator (SDC) enables to provide clean-up time by Co-58 peak activity through operational data from nuclear power plants (NPPs).In this study, we have investigated the results of EPRI SDC by shutdown chemistry data of Co-58 activity using NPP data from previous cycles and modeled the estimated clean-up time by EPRI SDC using average Co-58 activity of the NPP. We selected two RFO data from the NPP to evaluate EPRI SDC results using the purification time to reach to 1.3 μCi/cc of Co-58 after hydrogen peroxide addition. Comparing two RFO data, the similar purification time between actual and computed data by EPRI SDC, 0.92 and 1.74 h respectively, was observed with the deviation of 3.7–7.2%. As the modeling the estimated clean-up time, we calculated average Co-58 peak concentration for normal cycles after cycle 10 and applied two-sigma (2σ, 95.4%) for predicted Co-58 peak concentration as upper and lower values compared to the average data. For the verification of modeling, shutdown chemistry data for RFO 17 was used. Predicted RCS clean-up time with lower and upper values was between 21.05 and 27.58 h, and clean-up time for RFO 17 was 24.75 h, within the predicted time band. Therefore, our calculated modeling band was validated. This approach can be identified that the advantage of the modeling for clean-up time with SDC is that the primary prediction of shutdown chemistry plans can be performed more reliably during shutdown chemistry. This research can contribute to improving the efficiency and safety of shutdown chemistry evolution in nuclear power plants.

Passive Autocatalytic Recombiners Periodic Testing Issue

During the refueling outage (RFO) 2013 at NPP Krško, the electric hydrogen recombiners, a system for the hydrogen control during severe accidents and design basis accidents, were replaced by passive autocatalytic recombiners (PARs). During the next RFOs in 2015, 2016 and 2018, periodic tests of PAR cartridges were performed. The periodic testing of NIS-PARs shall prove that the catalytic reaction starts up as specified. According to the licenced testing procedure, each outage six cartridges were obtained and tested in the NIS PAR test device (TD). Randomly selected PAR cartridges did not pass the periodic test at room temperature (RT). Cartridges that do not pass the periodic testing have to be regenerated according to NEK procedure, followed again by a test in the TD to demonstrate the success of the regeneration. The effort of a full regeneration of all cartridges is not necessary and conservative. Due to significant unexpected workload during outages, the need for enhanced testing procedure with higher catalyst test temperature is priority. The concept for this is based on the conclusion that the PARs installed in NPP Krško were functional under accident conditions during the operational cycles (OL27, OL28 and OL29) even though PAR cartridges did not pass the periodic testing on RT (testing at higher but well below accident temperatures was successful). Revision of procedure and methodology in that direction requires development of the new TD. New TD will allow PAR cartridge measurements at defined elevated temperatures (range 40 - 70 deg C). It shall be vacuum oven type device or upgraded current NIS device. In both examples it shall provide testing of at least two cartridges at the same time, independently, however the whole cartridge shall be tested. Test gas remains 3 vol. % hydrogen in the air. For a detailed plant specific investigation of this catalytic material behavior, increased testing of cartridges was developed. Additionally, testing of different batch behavior was performed.

Nuclear heat storage and recovery for the APR1400

Detailed here is the proposed design of the tertiary side of a Thermal Energy Storage (TES) System to be interfaced with the steam cycle of a Light Water Reactor (LWR) plant. This study extends the understanding provided by many previous investigations (i.e., beyond thermodynamic considerations) by specifically addressing real world constraints associated with the backfit of such a system to an operating LWR. These constraints relate to the feasibility of design modifications during an extended refueling outage, and to licensing, operational, and maintenance considerations. The Korean designed and built APR1400 plant is selected for integration with the TES tertiary system, resulting in a design which can be readily be adapted to most LWR plant. Heat transfer and transport from and to the nuclear steam cycle is by synthetic oil with high temperature capabilities. Heat storage is in the form of packed beds consisting of crushed rock. While the rates of storage and recovery are constrained by the design of the nuclear reactor and steam plant (i.e., 20% of reactor thermal power during storage, and ~11% during recovery) the total stored energy component of the system can be readily and economically scaled to any desired capacity (at the marginal cost of carbon steel vessels filled with rock). Finally, it is proposed that any nuclear installation with water access could employ bulk thermal energy storage built into an Ultra Large Barge for turnkey delivery with a concomitant reduction in installed cost and risk.

Optimization of hydrogen mitigation systems in CPR1000 during a refueling outage

This study deals with the identification of hydrogen risk and optimization of current hydrogen mitigation systems in the CPR1000 nuclear power plant (NPP) during a refueling outage. A detailed MELCOR input is developed, which includes detailed containment nodalization, accurate configuration of the passive autocatalytic recombiner (PAR) systems, and realistic boundary conditions based on design data and technical specifications. Two cases that represent different plant operation states (POS) and specific hydrogen release paths are selected, and the accident progression and hydrogen risk are analyzed subsequently. The analysis indicates that the hydrogen risk is still of great importance during a refueling outage, although the accident progression is relatively slow. PAR systems with the current configuration are added to the plant model to verify the effectiveness of the hydrogen mitigation system in the CPR1000 during a refueling outage. The results indicate that the current configuration of PAR systems is still effective in mitigating the hydrogen risk during an accident in POS E but is insufficient in mitigating the hydrogen risk during an accident in POS H. Therefore, two strategies to improve the hydrogen mitigation systems are proposed, and their effectiveness is compared. It is concluded that mobile ignitors should be employed in the refueling pool region to avoid severe hydrogen risk during a refueling outage. The results of this study can support the development of the low power and shutdown (LPSD) probabilistic safety assessment (PSA) model and improve the capacity of NPPs to reduce the hydrogen risk during a refueling outage.

APPLICATION OF NONLINEAR PRINCIPAL COMPONENT ANALYSIS TECHNIQUE TO NUCLEAR POWER PLANTS

Traditionally, manual calibration of sensors is required and performed during each refueling outage. If the traditional time-directed calibration is replaced by an online monitoring technique, the ...

Development of ToSPACE for Pipe Wall Thinning Management in Nuclear Power Plants

A number of piping components in the secondary system of nuclear power plants are exposed to aging mechanisms such as FAC (Flow-Accelerated Corrosion), cavitation, flashing, SPE (Solid Particle Erosion), LDIE (Liquid Droplet Impingement Erosion), etc. Those mechanisms may lead to thinning, leak, or rupture of the components. Due to the pipe ruptures caused by wall thinning in Surry unit 2 of USA in 1986 and in Mihama unit 3 of Japan in 1994, the pipe wall thinning management has emerged as one of the most important issues in nuclear power plants. To manage the pipe wall thinning in the secondary system, Korea has used a foreign program since 1996. As using the foreign country’s program for long term, it was necessary to improve from the perspective of the users. Accordingly, KEPCO-E & C has started to develop the 3D-based pipe wall thinning management program (ToSPACE, Total Solution for Piping And Component Engineering management) from eight years ago, and the development was successful. This paper describes the major functions included in ToSPACE program, such as 3D-based DB (Database) buildup, development of FAC and erosion evaluation theories, UT (Ultra-sonic Test) data reliability analysis, field connection with 3D, automatic establishment of long-term inspection plan, etc. ToSPACE program was developed to allow site engineers performing the selection of inspection quantity at each refueling outage, UT data reliability analysis, UT evaluation, determination of next inspection timing, identification of the inspecting and replacing components in 3D drawings, etc., to access easily.

Virtual reality development and simulation of BMI nozzle inspection system for use during regular refueling outage of APR1400 family of reactors

Reactor vessel bottom mount instrumentation nozzle is J-groove welded to reactor vessel inside. This location has shown leaks in some NPPs in the past such as pinhole leaks at the South Texas Project in 2003 and a similar trace of leak at Palo Verde Unit 3 in 2013. Inspection of reactor vessel bottom mounted instrumentation nozzle weld crack requires removal of reactor internals in the case of APR1400 family of reactors without special tools. In Korea, the removal of reactor internals is prohibited during regular refueling outage following regulatory guidelines, hence the inspection of defects in reactor vessel bottom mounted instrumentation nozzle from inside of reactor vessel has not been performed during a regular refueling outage. In this study, a remote weld inspection system for reactor vessel bottom mounted instrumentation nozzles is developed in a virtual environment, and the operation of the inspection system was simulated to prove the validity of the system. The aim of the reactor vessel bottom mounted instrumentation nozzle inspection system was to inspect reactor vessel bottom mounted instrumentation nozzle during regular refueling outage without removing RI. This allows inspection of reactor vessel bottom mounted instrumentation nozzle weld for every refueling outage and allows finding defect at an early stage of defect formation and prevent reactor vessel crack caused by primary water stress corrosion cracking which lead to the bottom mounted instrumentation nozzle leak.

Effects of shutdown period extension on core damage frequency

The main objective of this study is to investigate the reduction possibilities of the CDF of LPSD conditions in a typical Westinghouse type PWR. For this, a comprehensive LPSD PSA model has been developed for a typical nuclear power plant (NPP). Total LPSD CDF is estimated to be 2.27E−06/year, assuming refueling outage as shutdown representative. To decrease it, CDF of LPSD conditions is re-evaluated for sensitivity case with shorter shutdown period. The results represent 13% decrease in total CDF of LPSD conditions. The contribution of CDF of the most dominant POSs changes from 66.24 to 10.45%, for POS 9 and POS 7, to 70.20 and 1.99%, respectively. The study also shows the proportionality of the change in percent contributions of POS_CDF in total CDF of LPSD conditions with that of POS duration in total duration of LPSD conditions. In this regards, percent change in CDF of POS 7 and POS 5 whose durations undergoes −4.76 and 3.31%, respectively, is −83.38 and 65.5%. Changes in CDF of POSs are also compared. Change in CDF of POSs is proportional to change in POS duration. However, for POS 9 and POS 7, with the higher primary contribution in total CDF of LPSD conditions, the percent change is more conspicuous. The study also shows the proportionality of the change in percent contributions of POS_CDF in total CDF of LPSD conditions with that of POS duration in total duration of LPSD conditions. Therefore, it is concluded that shortening of POSs with higher contributions (yet might not with longer duration) in total duration of LPSD conditions, leads to more decrease in total CDF of LPSD conditions. The results are useful for outage planning and management decision-making.

국내 가압경수로형 원전 계획예방정비기간 주요 방사선작업에 대한 납 차폐복 선량저감효과 분석

국내 가압경수로형 원전 계획예방정비기간 주요 방사선작업에 대한 납 차폐복 선량저감효과 분석

원전 증기발생기 와전류검사 장치의 전기적 특성 측정

원전 증기발생기는 원자로 냉각재 계통에서 발생한 열에너지를 터빈 계통의 주급수에 전달하여 터빈을 회전시키기 위한 증기를 생산하는 일종의 열교환기이다. 증기발생기 전열관의 손상은 증기발생기의 구조적 및 누설 건전성 유지 능력을 저해시키기 때문에 주기적으로 와전류검사를 수행하여 전열관의 건전성을 평가한다. 증기발생기 전열관의 건전성 평가는 보통 원자로 연료 재장전 기간 중에 수행된다. 현재 국내 증기발생기 전열관에 적용되는 와전류검사는 KEPIC 및 ASME 코드 요건에 따라 수행되며, 와전류검사 수행에 필요한 검사 시스템은 와전류검사 장치와 수집된 신호를 평가하기 위한 평가 프로그램으로 구성된다. 검사에 적용되는 와전류검사 시스템을 구성하는 핵심기기인 와전류검사 장치는 ASME 및 KEPIC 코드에서 총 고조파 왜곡율, 입출력 임피던스, 증폭기 직선성 및 안정성, 위상 직선성, 대역폭 및 복조필터 응답, 디지털 변환, 채널 간섭 등과 같은 전기적 특성을 측정하도록 규정하고 있다. 이에 따라 본 논문에서는 국내 최초로 개발한 원전 증기발생기 와전류검사 장치의 전기적 특성 측정을 위한 ASME 및 KEPIC 코드 요건을 설명하고, 이 요건에 따른 증기발생기 와전류검사 장치의 전기적 특성의 측정 결과를 제시하였다. A steam generator in nuclear power plant is a heatexchager which is used to convert water into steam from heat produced in a nuclear reactor core, and the steam produced in steam generator is delivered to the turbine to generate electricity. Because of damage to steam generator tubing may impair its ability to adequately perform required safety functions in terms of both structural integrity and leakage integrity, eddy current testing is periodically performed to evaluate the integrity of tubes in steam generator. This assessment is normally performed during a reactor refueling outage. Currently, the eddy current testing for steam generator of nuclear power plant in Korea is performed in accordance with KEPIC & ASME Code requirements, the eddy current testing system is consists of remote data acquisition unit and data analysis program to evaluate the acquired data. The KEPIC & ASME Code require that the electrical properties of remote data acquisition unit, such as total harmonic distortion, input & output impedance, amplifier linearity & stability, phase linearity, bandwidth & demodulation filter response, analog-to-digital conversion, and channel crosstalk shall be measured in accordance with the KEPIC & ASME Code requirements. In this paper, the measurement requirements of electrical properties for eddy current testing instrument described in KEPIC & ASME Code are presented, and the measurement results of newly developed eddy current testing instrument by KHNP(Korea Hydro & Nuclear Power Co., LTD) are presented.

Ultrasonic phased array examination of circumferential weld joint in reactor pressure vessel of BWR

The weld joints in the reactor pressure vessel (RPV) of Boiling Water Reactors (BWR) are required to be examined periodically for assurance of structural integrity. Ultrasonic phased array examination technique has been developed in authors’ laboratory for inspection of the top flange to shell circumferential weld joint in RPV of BWRs, which are in operation in India since the late 1960s. The development involved detailed simulation studies for computation of focal laws followed by validation on mock-up. The paper brings out the limitations of the conventional ultrasonic technique and how this can be overcome by the phased array approach for the weld joint under consideration. The phased array technique was successfully employed for field examination of this weld joint in RPV during the re-fuelling outage.

원전 연료집합체의 손상, 변형 및 이물질 검사시스템 개발에 관한 연구

Screen inspection system applied vision and laser scan technology which detect foreign materials caused fuel rod to be damaged, and which inspect fuel rod damage, bowing, distortion and grid damages, was developed to secure reliability and reproductivity of inspection method for nuclear fuel assembly during outage. In further, datum of inspection results will be continuously monitored and given understand the pattern of bowing and distorting for fuel assembly in reactor. Understanding of the pattern will be key technical information to avoid grid demage might be happened during refueling outage and provides important data base for safe operation of nuclear power plant in Korea and world wide.

Development of Safety Factors for the UT Data Analysis Method in Plant Piping

There are several thousand piping components in a nuclear power plant. These components are affected by degradation mechanisms such as FAC (Flow-Accelerated Corrosion), cavitation, flashing, and LDI (Liquid Droplet Impingement). Therefore, nuclear power plants implement inspection programs to detect and control damages caused by such mechanisms. UT (Ultrasonic Test), one of the non-destructive tests, is the most commonly used method for inspecting the integrity of piping components. According to the management plan, several hundred components, being composed of as many as 100 to 300 inspection data points, are inspected during every RFO (Re-Fueling Outage). To acquire UT data of components, a large amount of expense is incurred. It is, however, difficult to find a proper method capable of verifying the reliability of UT data prior to the wear rate evaluation. This study describes the review of UT evaluation process and the influence of UT measurement error. It is explored that SAM (Square Average Method), which was suggested as a method for reliability analysis in the previous study, is found to be suitable for the determination whether the measured thickness is acceptable or not. And, safety factors are proposed herein through the statistical analysis taking into account the components’ type.

배관감육 평가를 위한 UT 측정 신뢰도 분석

UT(Ultrasonic Test), one of the non-destructive tests, is the most common thickness measurement method for evaluating the wear rate in NPPs(Nuclear Power Plants). UT is used widely because it is easy and safe for use. However some amount of error inevitably occurs in attempting to measure the thickness. The error, that could make the thickness data thicker or thinner, may affect estimation of wear rate in pipes. NPPs are composed of a lot of pipes and components. Some of them are tested to check the current status during RFO(Re-Fueling Outage). Reliability analysis of UT is essential for evaluating pipe wear rate and establishing the long-term management plan in NPPs. This paper reviewed the cause of error occurrence and presented the UT data reliability analysis method. Also, this paper shows the application result of reliability analysis to the UT data acquired in NPPs.

Financial Impact of Nuclear Power Plants: Firstenergy and Davis-Besse

In March 2002, during a scheduled refueling outage, workers at FirstEnergy's Davis-Besse nuclear power plant found a football-sized hole in the reactor head caused by boric acid corrosion. Redundant shielding prevented any radiation leakage, but FirstEnergy projected the outage would last two to three months and cost from $15 million to $30 million. Instead, the outage lasted 26 months and cost $588.9 million. Students will learn the financial impact of the event, how investors reacted, and lessons in financial transparency.

In-Reactor Performance of an Advanced PWR Fuel, PLUS7, for OPR1000s in Korea

An advanced Pressurized Water Reactor (PWR) Fuel, PLUS7, for Optimized Power Reactors (OPR1000s) in Korea has been developed to achieve peak rod average burnup of 72GWD/MTU, overpower margin increase of 10% with respect to a reference fuel used in the OPR1000s, fuel integrity maintenance even with 0.3 g of seismic load, zero defect against debris, and grid-to-rod fretting wear. All the mechanical and thermal hydraulic performances of PLUS7 had been verified through a wide spectrum of out-of-pile verification tests, whereas the burnup-dependent performances have been verified through in-reactor verification tests of four PLUS7 Lead Test Assemblies (LTAs) in one of the OPR1000s. During each refueling outage period, burnup-dependent parameters were measured, which include fuel assembly, fuel rod and spacer grid dimensional changes, fuel assembly bow and twist, fuel rod bow, and fuel rod oxide layer thickness. Based on the measured data, it is said that the PLUS7 fuel will maintain integrity at least up to a peak rod average burnup of 58 GWD/MTU. In addition, from the coolant activity analyses and ultrasonic tests, the PLUS7 fuel was found to be free from any fuel damage including debris-induced and grid-to-rod fretting-wear-induced fuel failures. This successful in-reactor operation of the PLUS7 LTAs has made the Korea Nuclear Fuel (KNF) initiate the supply of PLUS7 fuel on a region basis for all OPR1000s from the year 2006.

Good Guys

Good Guys Maureen Stanton (bio) The Beautiful Plant My brother Patrick and I speed down Route 9, a curvy, one-lane, luge track of a road, toward Buchanan, New York, where the Indian Point III Nuclear Generating Station rises from the east bank of the Hudson River, thirty-five miles north of Manhattan. Through a friend, Patrick and I have landed temporary jobs as painters for a refueling outage and “plant beautification” project.1 To work here, Patrick and I must join the International Brotherhood of Painters and Allied Trades. For this, we need a Good Guy letter from a union member vouching for our character. I have no idea who has signed for us—our friend has arranged that—but someone is willing to say we are Good Guys. (Incidentally, the Good Guy letter is an actual form with the words “Good Guy Letter” printed across the top.) We pass a baseball diamond along the Hudson, then turn into a sylvan glade enclosed in chain link. At the guard station, Patrick offers the name of our employer, NPS Energy Systems, then we cruise past hunks of rusting metal and neolithic concrete forms in wire-mesh pens to a four-story, L-shaped training building. From the parking lot, we see only a half-dozen modular buildings and two warehouses surrounded by woods. I’m amazed at how easy it is to hide a nuclear power plant. [End Page 389] Inside the training center, next to a mock control room that visitors peer into as if looking upon a nursery, is the windowless conference room where we will spend the next five days in Fitness for Duty training. Our group of trainees includes thirty construction workers—pipe fitters, electricians, insulators, carpenters, painters—and one executive who looks pink and priggish in his white shirt and tight tie. Patrick and I stand out too, with our uncallused hands, our fingernails evenly trimmed white quarter moons, and Patrick’s fleece pullover, evidently not a masculine fabric like denim or the brown Carhartt jackets everyone wears. Patrick’s hair is resolutely short, unlike the “mullet” that is popular here, clipped bangs and sideburns with long hair in back, a schizophrenic style that shouts, “I may look like a citizen, but I’m a rebel at heart.”2 Patrick would blend more easily if he weren’t with me. Of the nearly three hundred construction workers hired for the outage, only eight are female. “It’s nice that your brother brought you into the trade,” a sixtyish electrician says to me before class. Then adds, “I just don’t think women belong in construction.” Our instructor, Jack, a thin, ruddy-faced retiree who trains part-time, begins by administering a psychological exam, a series of statements to which we must respond strongly agree, strongly disagree, or neutral. Jack hands out number two pencils and answer sheets with rows of tiny blank ovals and instructs us to select neutral as seldom as possible.3 The problem is that some statements I 4agree with slightly, but not strongly. For example: “Threats never bother me.” Well, idle threats don’t bother me. Serious threats do. The fact is threats sometimes bother me. I’m momentarily stymied, but finally choose strongly disagree. Now I worry they will interpret my answer as demonstrating paranoia. Haven’t I just implied that “Threats always bother me?” Perhaps I am being paranoid. I move on. “The pomp and splendor of any big state ceremony are things which [End Page 390] should be preserved.” I typically don’t go for ceremonies, but who doesn’t like a parade? I think I agree, but certainly not strongly. What choice do I have? “If plates are the least bit dirty, I feel too disgusted to eat.” I do have standards, but I don’t want to appear neurotic. I decide to strongly disagree. “I have disturbing thoughts that keep me awake at night.” I’ve been an insomniac most of my life, and yes, thoughts keep me awake. Are they disturbing? Sometimes, otherwise they wouldn’t keep me awake. I’m beginning to feel anxious. I can’t respond to any of the statements without...

Continuous Monitoring and Calibration of UTSG Process Sensors by Autoassociative Artificial Neural Network

The control and operation of complex power-generating systems, such as nuclear power plants, rely on the measurements of several sensors that monitor the process and the system state. On the basis of the sensor measurements, the system is operated for maximum economic efficiency and safety. Out-of-calibration sensors can lead to misinterpretation of the system state and problems with control and operation of the process, with possible economic losses, equipment damage, and safety consequences. To avoid such occurrences, periodic sensor calibrations are scheduled to ensure that sensors are operating correctly. These calibrations are performed manually and involve all sensors, independent of the actual need for calibration of each sensor. Continuous sensor calibration monitoring would then be most desirable both to ensure correct process control and system operation and to reduce maintenance costs associated with performing unnecessary manual sensor calibrations. This latter issue is of great relevance in nuclear power plants due to the large number of sensors employed, which are tested for calibration at each refueling outage. In this paper, the artificial neural network–based sensor calibration monitoring system is proposed to provide continuous sensor status information and virtual estimates for faulty sensors. In particular, we illustrate the design of an autoassociative artificial neural network for sensor fault detection and validation. The efficiency of the proposed method is verified through its application to eight critical transient signals coming from a U-tube steam generator of a pressurized water reactor modeled by means of a validated simulation code.

Mitigation of SCC Initiation on BWR Core Internals by Means of Hydrogen Water Chemistry during Start-Up

Hydrogen injection has been applied as a preventive measure against the stress corrosion cracking (SCC) phenomenon in many boiling water reactors. However, it can be applied only during normal plant operation since hydrogen is usually injected into the feedwater and this system is in standby mode during start-up operations. It is estimated that the core internals are subjected to the strain rate that may cause susceptibility to SCC initiation during start-up. Therefore, it is beneficial to perform hydrogen injection during start-up as well in order to suppress SCC initiation.For this purpose, we installed an additional hydrogen injection system to be used during plant start-up at the Tokai-2 power station. This trial Hydrogen water chemistry During Start-up (HDS) system was applied following the 19th refueling and maintenance outage in December 2002. By comparing results obtained during this start-up with HDS to previous start-up data using normal water chemistry, we made the following observations. First, as the reactor water temperature increased from initial conditions up to 180°C via nuclear heating, dissolved oxygen and hydrogen peroxide concentrations decreased to levels lower than previously observed. Second, during subsequent nuclear heating, up to 250°C, the dissolved oxygen concentration did not exceed 1 ppb, and the electrochemical corrosion potential was maintained in a low range near –400 mV versus the standard hydrogen electrode.

Metallurgical Evaluation of a Feedwater Nozzle to Safe-End Weld

Weld cracks in safety class systems are a serious concern, because these systems are part of the primary barrier providing containment of radioactive coolant. Loss of weld integrity yields leaks, or, under catastrophic failure, can be the basis for a severe loss of coolant accident. A circumferential indication was found by ultrasonic examination (UT) in the N4A-2 inlet feedwater nozzle to safe-end weld during the second refueling outage of River Bend Station Unit 1 in March 1989. The indication, approximately 15cm (6in) long with a reported maximum depth of 0.5cm (0.1in), was located in the Alloy 182 weld butter on the safe-end side of the weld. (The safe-end base metal was ASME SA 508 Class 1 carbon steel.) The reported characteristics of the UT indication were indicative of intergranular stress corrosion cracking. This indication was reexamined during the second and third fuel cycles in March 1990 and September 1991, respectively, and during the third refuel outage in November 1990. Crack growth was reported during each examination. The safe-end was replaced during the fourth refueling outage in the summer of 1992. The US Nuclear Regulatory Commission (NRC) subsequently contracted with Brookhaven National laboratory (BNL) to conduct a confirmatory investigation to establishmore » the failure mode and determine the root causes of cracking in the safe-end weld.« less