05000530/LER-2015-001
| Docket Number | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| Initial Reporting | |
| ENS 50968 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| 5302015001R00 - NRC Website | |
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I. FACILITY NAME 3. PAGE 2. DOCKET 6. LER NUMBER Palo Verde Nuclear Generating Station All times are Mountain Standard Time and approximate unless otherwise indicated.
1. REPORTING REQUIREMENT(S):
This Licensee Event Report (LER) is being submitted pursuant to 10 CFR 50.73(a)(2)(ii)(A) as a degraded principal safety barrier and 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.14, RCS Operational Leakage, due to reactor coolant system (RCS) pressure boundary leakage.
At approximately 0032 on April 7, 2015, with Unit 3 in Mode 5, during a scheduled boric acid walk-down inspection of the Unit 3 RCS, engineering personnel at the Palo Verde Nuclear Generating Station (PVNGS) identified white residue on a one inch instrument nozzle on the reactor coolant pump 2A suction pipe. Isotopic analysis confirmed the white residue resulted from leakage of RCS coolant and, at 0330 on April 8, 2015, engineering personnel determined that RCS pressure boundary leakage had occurred resulting in a condition prohibited by Technical Specification 3.4.14.
This condition was reported to the NRC pursuant to 10 CFR 50.72 (b)(3)(ii)(A) at 0840 on April 8, 2015, via the Emergency Notification System (ENS 50968).
2. DESCRIPTION OF STRUCTURE(S), SYSTEM(S) AND COMPONENT(S):
A primary function of the RCS (EIIS: AB) is to provide a barrier against fission product release to the environment. In order to ensure integrity of the RCS, piping and component joints are made by welding, bolting, rolling, or pressure loading, and valves are provided to isolate connecting systems from the RCS. During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant leakage, through normal operational wear or mechanical deterioration.
The RCS piping is arranged with two flow loops connected in parallel to the reactor vessel.
Each of the two loops consists of one 42 inch outlet (hot leg) pipe, one steam generator, two 30-inch reactor coolant pump (RCP) inlet (cold leg) pipes, two RCPs, and two 30-inch RCP outlet (cold leg) pipes. The RCS piping is configured with various penetrations which serve to provide connections to interfacing systems and instrumentation. Each instrumentation penetration is provided with a nozzle which creates the interface between the pipe and the connected instrumentation. There are 47 instrument nozzles on RCS piping including 27 hot leg pipe nozzles and 20 cold leg pipe nozzles. Specific to this event, instrument nozzle 18 is ')RC FORM 366A (02-2014) Palo Verde Nuclear Generating Station the loop 2A RCP suction pipe cold leg instrument nozzle. The cold leg instrument nozzles are fabricated from Alloy 600 and are connected to the RCS piping with J-groove welds made with Alloy 82/182 weld materials. Instrument nozzle 18 was welded using Alloy 182 weld material. These weld materials and Alloy 600 have been determined to be susceptible to primary water stress corrosion cracking (PWSCC). Three factors: susceptible material, a corrosive environment, and stress must be present for PWSCC to occur. Stresses that make Alloy 600 nozzles and their Alloy 82/182 J-groove attachment welds susceptible to cracking are inherent residual stresses induced by the welding of the nozzle to the inside surface of the pipe during fabrication.
To monitor the integrity of the RCS pressure boundary, the PVNGS In-Service Inspection (ISI) Program requires the conduct of bare metal visual examinations of RCS cold leg pressure retaining welds fabricated with Alloy 600 and Alloy 82/182 once per 10 year interval in accordance with American Society of Mechanical Engineers (ASME) Code Case N-722-1 as conditioned by 10 CFR 50.55a. In addition to the ISI Program inspections, the PVNGS Boric Acid Corrosion Control Program requires the performance of boric acid walk-down inspections of the RCS in each refueling outage to identify boric acid deposits which may be indicative of leakage. Because 16 of the 20 RCS cold leg instrument nozzles are readily accessible, the boric acid walk-down inspections conducted by engineering personnel in each refueling outage also inspect these nozzles.
PVNGS TS LCO 3.4.14, RCS Operational Leakage, is applicable in Modes 1 through 4 when the RCS is capable of being pressurized and provides limitations for RCS leakage. The LCO specifies the types of leakage and provides required actions when leakage rates exceed allowable values. The LCO specifies that no RCS pressure boundary leakage, defined as non-isolable leakage (except primary to secondary leakage) through a component body, pipe wall or vessel wall, is allowed because such leakage could cause further deterioration and result in higher leakage rates.
3. INITIAL PLANT CONDITIONS:
At approximately 0032 on April 7, 2015, PVNGS Unit 3 was stable in Mode 5 with RCS temperature at approximately 97 degrees Fahrenheit and RCS pressure at approximately 40 pounds per square inch absolute. Unit 3 had been shutdown at 0000 on April 4, 2015, to commence the 18th Unit 3 refueling outage (3R18). There were no structures, systems, or components inoperable that contributed to the event.
4. EVENT DESCRIPTION:
At approximately 0032 on April 7, 2015, during a scheduled boric acid walk-down inspection of the Unit 3 RCS, engineering personnel identified white residue on a one inch instrument nozzle on the Unit 3 RCP 2A suction pipe (nozzle 18). This residue appeared to originate at the penetration for nozzle 18 and was in a pattern consistent with leakage of high pressure steam throttled through a very small opening. Visual inspections were performed in the immediate area and directly above nozzle 18, but no other evidence of leakage was found.
Isotopic analysis confirmed the white residue resulted from leakage of RCS coolant that was deposited within the previous six to ten months. At 0330 on April 8, 2015, engineering personnel notified operations personnel that the white residue on nozzle 18 resulted from RCS pressure boundary leakage. Based on the observed condition, engineering concluded the most likely source was leakage from a flaw in the J-groove weld that attached the nozzle to the RCS pipe. The degraded condition of a principle safety barrier was reported to the NRC pursuant to 10 CFR 50.72 (b)(3)(ii)(A) at 0840 on April 8, 2015, via the Emergency Notification System (ENS 50968).
An investigation was initiated to evaluate the condition and to establish necessary corrective actions. Scheduled refueling outage activities were continued and Unit 3 transitioned from Mode 6 to a Defueled condition on April 11, 2015, when all fuel was removed from the reactor vessel. The RCS piping was drained as necessary to allow repairs to the leaking nozzle. To characterize the problem, engineering personnel conducted additional inspections of the instrument nozzle and piping penetration. On April 13, 2015, straight beam ultrasonic testing (UT) of instrument nozzle 18 was performed and no indications of circumferential fatigue cracking were identified. Repairs were performed per an ASME code approved and NRC endorsed half-nozzle repair method which installed a new Alloy 690 nozzle with Alloy 52M weld material.
5. ASSESSMENT OF SAFETY CONSEQUENCES:
This event did not result in a potential transient more severe than those analyzed in chapters 6 and 15 of the Updated Final Safety Analysis Report or result in the release of radioactive materials to the environment. There were no actual safety consequences as a result of this event and the event did not adversely affect the health and safety of the public.
The noted leakage from the instrument nozzle is bounded by the Probabilistic Risk Assessment (PRA) analysis for Very Small Loss of Coolant Accident or Leak events (NUREG/CR-5750 category G1) which is defined as pressure boundary leakage within the capacity of three charging pumps. These events are modeled as Miscellaneous Events in the PVNGS PRA model. The conditional probability of core damage given a Miscellaneous Event is 1.2E-7 in the PVNGS PRA. This risk impact is characterized as "very small" per NRC Regulatory Guide 1.174. Therefore, this condition had minimal safety significance.
The condition would not have prevented the fulfillment of a safety function; and, the condition did not result in a safety system functional failure as defined by 10 CFR 50.73 (a)(2)(v).
6. CAUSE OF THE EVENT:
The cause of the event was determined' to be primary water stress corrosion cracking of the Alloy 600 instrument nozzle.
7: CORRECTIVE ACTIONS:
To correct the instrument nozzle leakage an ASME code approved and NRC endorsed half- nozzle repair was performed on nozzle 18. In support of this repair, relief request number 53 was submitted and verbally approved by the NRC for one cycle of operation of Unit 3.
To address the extent of condition the following actions were performed or are planned:
- During 3R18, bare metal visual inspections were performed for the remaining 19 RCS cold leg instrument nozzles. No problems were identified during these inspections.
- Bare metal visual inspections will be performed on each of the 20 RCS cold leg instrument nozzles during the next refueling outage in Unit 1 and Unit 2.
No change to inspection activities was needed for the RCS hot leg instrument nozzles because all RCS hot leg instrument nozzles are inspected during each refueling outage in accordance with the PVNGS In-Service Inspection program.
8. PREVIOUS SIMILAR EVENTS:
A similar event occurred in October 2013 when RCS pressure boundary leakage was discovered on Unit 3 reactor vessel bottom mounted instrument (BMI) nozzle number three.
PVNGS reported this problem in licensee event report number 50-530/2013-001-00 on December 6, 2013. The BMI nozzle leakage was corrected with an ASME code approved and NRC endorsed weld pad/half-nozzle repair method.