05000269/LER-2013-004
| Oconee Nuclear Station, Unit 1 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 2692013004R01 - NRC Website | |
EVALUATION:
BACKGROUND
At the time elevated Reactor Coolant System (RCS) [EIIS:AB] leakage was identified, Oconee Nuclear Station (ONS) Unit 1 was operating in Mode 1 at approximately 100 percent power. No significant structures, systems or components were out of service at the time of this event that contributed to this event. Unit 1 [EIIS:NH] experienced reactor coolant pressure boundary leakage in the High Pressure Injection (HPI) System [EIIS:BG] that required a plant shutdown. Technical Specification (TS) 3.4.13, RCS Operational Leakage, Required Actions B.1 and B.2 require that the unit be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of identification of pressure boundary leakage.
During normal operation, the HPI System controls the RCS inventory, provides the seal water for the Reactor Coolant Pumps (RCP) [EIIS:P], and recirculates RCS letdown for water quality maintenance and reactor coolant boric acid concentration control. The discharge of the HPI pumps connects to a nozzle on each of the four reactor inlet pipes downstream of the reactor coolant pumps. The reactor coolant which is letdown is normally returned to the RCS through two of these nozzles (1A1 and 1A2).
During emergency operation, the HPI System supplies borated water from the Borated Water Storage Tank (BWST) to the RCS and the RCP seals. Three parallel HPI pumps have the capability to take suction from the BWST and discharge through two redundant flow headers into the RCS, utilizing four injection lines (two per header). The stainless steel HPI injection lines terminate at injection nozzle [EIIS:NZL] assemblies located on each of the reactor inlet pipes downstream of the RCPs. Each nozzle assembly consists of a carbon steel nozzle (stainless steel clad on the inside), to which a stainless steel safe end is welded. The HPI piping is welded to the other end of the safe end. Inside the safe end is a stainless steel thermal sleeve, which extends into the main RCS flow path.
EVENT DESCRIPTION
On November 8, 2013, at approximately 1837 hours0.0213 days <br />0.51 hours <br />0.00304 weeks <br />6.989785e-4 months <br />, while Oconee Unit 1 was operating at approximately 100% Full Power (FP), the Control Room received an alarm associated with the containment atmosphere particulate radiation monitor (i.e., RIA-47 [EIIS:IL]) used for RCS leakage detection. At 2324 hours0.0269 days <br />0.646 hours <br />0.00384 weeks <br />8.84282e-4 months <br />, an RCS Leakage Calculation was performed and unidentified leakage of 0.020 gpm was noted, consistent with leakage calculation values obtained within the prior week.
Reactor building particulate sample results indicated no detected activity, and the radiation monitor counts stabilized.
On November 9, 2013, at 0323 hours0.00374 days <br />0.0897 hours <br />5.340608e-4 weeks <br />1.229015e-4 months <br />, another RIA-47 alarm was received. No immediate signs of RCS leakage were observed. At 0614 hours0.00711 days <br />0.171 hours <br />0.00102 weeks <br />2.33627e-4 months <br />, a second RCS Leakage Calculation was performed, and unidentified leakage was found to be 0.088 gpm. This value exceeded the baseline mean by three standard deviations, indicating the leakage results were valid and action was warranted. Based on the leakage calculation results, increased activity on the radiation monitor, and an observable increase in reactor building normal sump rates, a reactor building entry was conducted to identify the leakage source. Although the general location of the leak was identified, the source could not be determined.
On November 10, 2013, at 1630 hours0.0189 days <br />0.453 hours <br />0.0027 weeks <br />6.20215e-4 months <br />, a second reactor building entry was conducted and video images and robotically obtained leakage samples were evaluated. Absent conclusive evidence of the leak source, the Shift Manager initiated a power reduction to allow for a direct visual inspection.
A normal downpower to 20% was initiated on November 10, 2013, at 2141 hours0.0248 days <br />0.595 hours <br />0.00354 weeks <br />8.146505e-4 months <br />. On November 11, 2013, at 0520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br />, the approximately 0.1 gpm leakage was visually determined to be un-isolable RCS pressure boundary leakage from the 1B2 HPI Injection Line. TS 3.4.13, RCS Operational Leakage, Conditions B.1 and B.2 were entered. At 0848 hours0.00981 days <br />0.236 hours <br />0.0014 weeks <br />3.22664e-4 months <br />, an Emergency Notification System call to the NRC was made reporting the degraded condition and the TS required shutdown. On November 12, 2013, at 0244, Unit 1 entered Mode 5 and exited the TS Mode of Applicability. Upon confirmation that the unidentified leakage being investigated was RCS pressure boundary leakage, the pressure boundary leakage had existed for a time period greater than the Technical Specification allowed COMPLETION TIME.
CAUSAL FACTORS
A Failure Investigation Team was created to develop a repair plan, and a Root Cause team was created to investigate the causal factors for this event. Two Root Causes for this event have been identified. This report has been revised to reflect the results of the completed root cause evaluation.
RC-1: The un-isolable reactor coolant system leak on the 1B2 HPI line was caused by mechanical, high-cycle fatigue which resulted in a through wall crack in the butt weld between the safe end and HPI piping.
The crack initiated at two separate locations at the weld root due to an unspecified high vibration event, likely associated with the 2008 1B2 reactor coolant pump seal failure.
These cracks then merged into a single crack and continued to propagate through wall in the intervening years. Crack propagation is attributed to HPI Full Flow testing in subsequent refueling outages (i.e., for test purposes, flow is directed through a single header, rather than splitting into multiple headers) until normal reactor coolant system operating conditions grew the crack thru-wall. For each unit, HPI line vibration is greatest on the B2 line and is bounded by the 1B2 HPI line. The 1B2 HPI line is dedicated for emergency injection only, and it is generally stagnant during normal plant operation.
RC-2: Ownership and oversight of augmented examination program was inadequate.
Commitment fidelity was lost resulting in incomplete examination of weld 1RC-201-105 required to meet commitments to GL 85-20 and IEB 88-08.
The Engineering Programs Inservice Inspection group did not clearly define examination volumes and areas in the ISI plan to be examined to meet the B&W examination requirements that were committed to the NRC. GL 85-20 commitments were to perform ultrasonic examinations of HPI Nozzle, Nozzle inside radius, safe end-to-nozzle weld, and safe end-to-pipe weld including base metal to first block valve. IEB 88-08 included much of the area examined in GL 85-20 plus some additional upstream piping base material and butt welds. The level of detail in the ISI plan does not aid the examiner in understanding the committed examination volumes.
The Engineering programs NDE group failed to provide adequate guidance to the examiners for actions to be taken when full volume coverage could not be achieved. NDE procedure control processes failed to maintain regulatory commitments within the procedures associated with GL 85-20.
CORRECTIVE ACTIONS
Numerous actions were taken to investigate, determine extent of condition, and repair the HPI line.
Those most relevant are included below.
Immediate:
1. Inspected 1B2 injection line pipe and pipe support configuration to verify conformance with design. The piping and supports were installed per design requirements and tolerances.
There were no indications of damage or pipe rubbing due to vibration.
2. Performed Radiography Testing (RT) of 1B2 High Pressure Injection (HPI) nozzle and associated original Babcock & Wilcox 2-ply thermal sleeve assembly. RT images were compared to previous, historical RTs of the thermal sleeve and no visible cracks were found on the thermal sleeve. In addition, the position of the thermal sleeve and visible gaps within the expansion area remained unchanged.
3. Removed 1B2 injection line safe end-to-pipe butt weld and adjacent piping for metallurgical analysis. Metallurgical analysis identified high cycle mechanical fatigue (vibration) as the failure mechanism.
4. Performed surface conditioning to achieve maximum coverage of weld volume to be examined and Non-Destructive Examination (NDE) Evaluations on 1A1, 1A2, and 1B1 HPI nozzle safe end-to-piping stainless welds, including penetrant testing (PT), conventional UT, and phased array encoded UT. UT, PT, and phased array encoded UT results were acceptable for the 1A1, 1A2, and 1B1 HPI nozzle-to-safe end welds.
5. As part of an extent of condition review, performed surface conditioning to achieve maximum coverage of weld volume to be examined and NDE Evaluations on 2A1, 2A2, 2B1, and 2B2 HPI nozzle safe end-to-piping stainless welds, including PT, conventional UT, and phased array encoded UT during the scheduled refueling outage. UT, PT, and phased array encoded UT results were acceptable for each of the Unit 2 HPI safe end-to-pipe welds.
6. Performed dimensional checks of Unit 1 and Unit 2 HPI nozzle safe end and stainless steel pipe to determine stress and fatigue sensitivity to any observed discrepancies from design conditions. Of these eight lines, 1B2 safe end-to-pipe dimensions had the largest safe end variation and highest offset angle. The worst case geometry was analyzed and found to meet Code requirements.
7. Replaced 1B2 HPI safe end-to-pipe butt weld and adjacent piping. Replacement allowed for correction of the offset angle that was identified with the previous weld.
8. Collected vibration data on the 1B2 line during HPI Full Flow testing. Also collected vibration data on all four Unit 2 HPI lines during HPI Full Flow test during the refueling outage. Based on this review and historical data, the vibration levels observed for the HPI Lines during the Full Flow test are largest for the B2 lines on all three units, with the 1B2 line having the highest reported vibration levels of all 12 HPI Lines. The current vibration levels on the 1B2 line are within the design acceptance criteria of the piping.
9. Reviewed the population of Oconee Unit 3 welds that are inspected by the same NDE UT procedure as the 1B2 HPI safe end-to-pipe weld and determined they were adequately examined in 2010 to detect the presence of cracking. This includes each Unit 3 HPI safe end-to-pipe weld. The only UT inspections with coverage limitations noted (i.e., valve-to- valve butt welds) were further dispositioned by RTs at that time. The Unit 3 HPI thermal sleeve RTs from the spring of 2012, taken to assess thermal sleeve tightness and position, were also reviewed. Where visible in the RTs, the safe end-to-pipe weld showed no crack- like indications. Based on the reviews performed and the current negligible Unit 3 RCS unidentified leakage, there are no concerns with the operation of Unit 3.
Subsequent:
1. Increased the frequency of the ISI augmented examination (UT) to every refueling outage (1R) for HPI safe end-to pipe butt weld for the HPI lines that have higher vibration potential (six total). Examinations for HPI safe end-to-pipe butt welds that have lower vibration potential will remain at current 151 augmented plan 2R (every other refueling outage) frequency.
2. Revised NDE procedures to provide prescriptive guidance for maximizing examination coverage when performing augmented examinations, including entry into the corrective action program for evaluation by functional owner when limitations or indications of degradation are detected.
Planned:
1. Separate the Augmented Inspection portion of the ISI program from the ASME Section XI portion so that the Augmented Inspection portion can function as an independent Engineering Program. This action is intended to ensure that the process for managing Augmented Inspections is formally managed using the controls typical of formally designated engineering programs to ensure sustainability.
2. Revise NDE procedure for Level III oversight of routine NDE examination activities to utilize a "graded approach" relative to the significance of the NDE examination activity and operating experience.
SAFETY ANALYSIS
While at 100% power on November 8, 2013, RCS leakage of Unit 1. The leak was later found to be from a circumferential crack located at the safe end-to-pipe butt weld (1-RC-201-105) located between the 1B2 HPI injection nozzle and valve 1HP-152. The total circumferential extent of the aggregate crack was about 1.2 inches on the pipe inside diameter and about 0.1 inch (-1/8") on the outside surface.
The 1B2 HPI line is dedicated for emergency injection only, and it is generally stagnant during normal plant operation. The leak in this line remained small, and an orderly shutdown was performed. The leak was much less than what is considered in the Probabilistic Risk Analysis (PRA). However, an un-isolable leak in the reactor coolant system (RCS) pressure boundary constitutes degradation of a principal safety barrier and is reportable to the NRC. The leak was entirely within the reactor building containment and no radioactive releases were made.
Based on the metallurgical evidence and comparison of this approximately 0.1 gpm leak to the 1997 ONS Operating Experience associated with a similarly located and larger HPI safe end-to-pipe weld leak, the leak before break capacity of this material was demonstrated. There is reasonable assurance that the line would not have catastrophically broken, even during a design basis event, based on a comparison of the materials (same), loading (similar) and flaw extent (smaller) for these two leaks. Additionally, had the leak location failed catastrophically, the 2 1/2 inch pipe break (approximately 0.025 square feet), would have constituted a small break Loss Of Coolant Accident.
Breaks at this location are bounded by analyses in the Oconee UFSAR which concludes that this break can be handled without core damage.
ADDITIONAL INFORMATION
A review of the ONS corrective action program data was conducted to include all Root Causes for the last five years with similar event and cause codes as well as appropriate key word searches for this event. Selected apparent cause evaluations were reviewed as well. A broader search, however, revealed an ONS HPI event in 1997 and an issue with evaluating limited examination results in 2006. In both cases, Oconee did not adequately control the examination program, and the required examinations were not being performed effectively. The event in 1997 associated with the HPI nozzle thermal fatigue issue (PIP 0-97-1324) is a similar occurrence with related actions; therefore, a similar event does exist.
Energy Industry Identification System (El IS) codes are identified in the text as M. This event is considered INPO Consolidated Events System (ICES) Reportable. There were no releases of radioactive materials, radiation exposures or personnel injuries associated with this event.