05000333/LER-2007-001, Safety Relief Valve Setpoints Outside of Allowable Tolerances
| ML072260420 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 08/06/2007 |
| From: | Peter Dietrich Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CR-JAF-2007-02108, [[::JAF-07-0094|JAF-07-0094]] LER 07-001-00 | |
| Download: ML072260420 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3332007001R00 - NRC Website | |
text
Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
dh James A. Fitzpatrick NPP P.O. Box 110 Ent~gyLycoming, NY 13093 Tel 315 349 6024 Fax 315 349 6480 Pete Dietrich Site Vice President - JAF August 6, 2007 JAFP-07-0094 United States Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, D.C. 20555
Subject:
Docket No. 50-333 LICENSEE EVENT REPORT: LER-07-O01 (CR-JAF-2007-02108)
Safety Relief Valve Setpoints Outside of Allowable Tolerances
Dear Sir or Madam:
This report is submitted in accordance with 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications..."
There are no commitments contained in this report.
Questions concerning this report may be addressed to Mr. Jim Costedio at (31 5) 349-6358.
Very truly yours, Pete Dietrich PD:rp Enclosure cc:
USNRC, Region 1 USNRC, Project Directorate USNRC Resident Inspector INPO Records Center
Abstract
Review of the as-found setpoints for 10 Safety Relief Valve (SRV) [SB] pilot assemblies, removed at the end of Cycle 17, determined that 7 SRVs were outside the allowable as-found tolerance of 1145 psig +/- 34.3 psig
(+/- 3%) required by Technical Specifications (TS) Surveillance Requirement (SR) 3.4.3.1. Additionally, one pilot removed at the end of Cycle 17 could not be tested as required by TS SR 3.4.3.1. This report documents the failure to meet this SR for 8 of the 11 SRVs.
The effect of these SRVs being out of tolerance was analyzed and the results of this analysis show that Reactor Pressure Vessel (RPV) overpressure protection and nuclear plant safety were not adversely affected. Consequently, the safety significance of this event was minimal. Each of the seven out of tolerance SRV setpoints was determined to have a most probable cause of corrosion bonding between the SRV pilot disc and seat, a recognized industry generic problem.
NRC FORM 366 (6-2004)U.S. NUCLEAR REGULATORY COMMISSION I(6-2004)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL I REVISION 2
OF 5
NUMBER NUMBER James A. FitzPatrick Nuclear Power Plant 05000333 07 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
EIIS Codes in [ I
Event Description
On June 6, 2007, while the plant was operating at 100 percent power, FitzPatrick was notified that seven Safety Relief Valve (SRV) [SB] pilot assemblies removed at the end of Cycle 17 (October 2006 Refueling Outage) had as-found setpoints outside the allowable tolerance of 1145 psig +/- 34.3 psig (+/- 3%).
This allowable tolerance (1110.7 to 1179.3 psig) is required per Technical Specifications (TS) Surveillance Requirement (SR) 3.4.3.1. The seven SRVs exceeded the high limit of 1179.3 psig.
The removed SRV pilots were tested at Wyle Laboratories during the period May 29, 2007 through June 4, 2007. The results from these tests were reported to FitzPatrick by Wyle Laboratories on June 6, 2007. One pilot in location 02RV-71A was damaged during removal and could not be tested. To prevent recurrence, JAF plans to enhance the SRV pilot valve removal maintenance procedure. Three pilot valves have been sent for forensic analysis to confirm cause. This LER will be updated, if required based on the forensic results.
Test Results:
Pilot Serial Number 1013 1236 1110 1218 1191 1217 1235 1045 1051 1195 Plant Valve Number 02RV-71 B 02RV-71 C 02RV-71 D 02RV-71 E 02RV-71 F 02RV-71 G 02RV-71 H 02RV-71 J 02RV-71 K 02RV-71 L Initial Lift As-Found Setpoint 1184 1180 1155 1176 1187 1177 1195 1206 1233 1190 Initial Lift > 3%
Above Setpoint Yes Yes No No Yes No Yes Yes Yes Yes TS LCO 3.4.3 requires nine operable SRVs when in Modes 1, 2 or 3. Specifically, the TS states:
- The safety function of 9 S/RVs shall be OPERABLE.
Since seven pilot valves exceeded the allowable setpoint range, this report is being made under 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications..."
Cause of Event
The most probable cause of each of the seven high out of tolerance pilot setpoints was determined to be corrosion bonding between the SRV pilot disc and seat [Cause Code B]. With a bond forming between the pilot disc and seat, more pressure is needed to raise the pilot disc off the seat. Since the normal balance of pilot assembly spring force and steam pressure force necessary to lift the pilot disc corresponds to the nominal setpoint of the SRV, the pilot disc to seat bond results in a higher pilot lift setpoint.U.S. NUCLEAR REGULATORY COMMISSION
!6-2 004)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL I REVISION 3
OF 5
NUMBER NUMBER James A. FitzPatrick Nuclear Power Plant 05000333 07 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
==Cause of Event==An oxygen rich environment in the pilot assembly, due to the radiolytic breakdown of water to hydrogen and oxygen, causes corrosion bonding. Oxygen accumulates in the area of the pilot disc because the pilot assembly is a high point on the main steam [SBI line.
A contributing cause for corrosion bonding of the pilot disc to seat may be related to SRV insulation. The installation of insulation on the Target Rock 7567F SRVs has proven to be critical in the industry. Currently JAF is investigating the configuration of installed SRV insulation to determine recommended improvements. Based on the results of the investigation of insulation configuration this LER will be updated if required.
Event Analysis
The SRVs provide overpressure protection for the Reactor Coolant Pressure Boundary (RCPB) as required by the ASME Boiler and Pressure Vessel Code. SRV pilots actuating at pressures higher than the required setpoint may be significant if adequate overpressure protection is not available. The RCPB Overpressure Analysis is performed each fuel cycle based on the worst case anticipated transient with nine SRVs opening at an analyzed Upper Limit pressure of 1195 psig, and two SRVs out of service.
The current Anticipated Transient Without Scram (ATWS) analysis was performed using the worst case ATWS with two SRVs out of service and the other nine opening at the upper end of the uncertainty range for the Electric Lift trip setpoints. This analysis is not affected by as-found setpoint testing unless Electric Lift is inoperable during the cycle. During Cycle 17, SRV D experienced an electrical actuation failure due to a loose electrical connector (CRs-JAF-2006-02384 and 04108), which rendered Electric Lift inoperable for that one valve. Since the analysis assumes two SRVs out of service, the actual performance is enveloped by the analysis. Also, SRV D as-found lift setpoint (1155 psig) was less than the associated lift setpoint from the analysis (1157 psig). Accordingly, operation during Cycle 17 complied fully with the current ATWS analysis.
In comparing as-found SRV lift setpoints to the RCPB Overpressure Analysis, SRV A must be considered as not opening based on the inability to perform as-found testing (see CR-JAF-2007-01944). Two other pilots lifted at greater than 1195 psig; taking the higher as the second out of service valve, one of the nine lifted at 1206. However, eight of the pilots lifted at or below the 1195 analytical value, some by significant margins; also, the "out of service" valve lifted at 1233 psig, well below the peak transient pressure. The effect of the early lifts of several SRVs more than overcomes the effect of the slightly late lift of one SRV. Therefore, the peak pressure resulting from a limiting overpressure transient with the as-found SRV setpoints would be less than the peak pressure of 1307.4 psig from the cycle reload analysis, which met the safety limit of 1325 psig.
Additionally, the Electric Lift system installed in 2000 was operable throughout the cycle, except for SRV D as noted above. Electric Lift is not credited in the RCPB Overpressure Analysis. This system actuates the SRVs at the specified setpoints regardless of corrosion bonding, further limiting the peak pressure in the event of a pressurization transient.
Therefore, the safety significance of this event is considered low and does not decrease the effectiveness of plant barriers providing safety to the public.
Consequently, the safety significance of this event was minimal.U.S. NUCLEAR REGULATORY COMMISSION
.(6-2004)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL I REVISION 4
OF 5
NUMBER NUM3ER James A. FitzPatrick Nuclear Power Plant 05000333 07 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
Extent of Condition:
All of the SRVs are susceptible to setpoint drift due to pilot disc to seat corrosion bonding. This is a recurring industry issue that has been the subject of both NRC and BWROG generic assessments.
As part of FitzPatrick's efforts to improve the performance of the SRV pilots, Stellite 21 discs were installed in all of the eleven SRVs at the beginning of Cycle 17.
In addition, the BWROG recommended modification to provide pressure switch actuation of the SRVs has been installed and was operational during Cycle 17. This modification provides an electric actuation of SRV pilot valves based upon a pressure switch actuating at a predetermined setpoint. This provides a diverse, redundant method of SRV actuation, which overcomes the pilot disc-seat corrosion bonding effect.
Corrective Actions
Corrective Actions Completed Prior to this Report:
- 1.
All SRV Pilots were removed from the plant during Refuel Outage 17 (October 2006) and replaced with newly refurbished and test certified pilots (using Stellite 21 discs) for Cycle 18.
- 2.
The BWROG recommended modification to provide pressure switch actuation of the SRVs was operational during Cycle 17 when these valves were in service.
- 3.
All SRV pilot assemblies are tested and replaced each operating cycle, however as stated above, 02RV-71A was not tested due to a maintenance error.
- 4.
"D" SRV loose electrical connector was repaired during Refuel Outage 17.
Corrective Actions for this Event:
- 1. Revise the maintenance procedure for SRV pilot valve removal (MP-002-04) to prevent damage of the pilot valve during removal.
(Due 08/01/08)
Safety System Functional Failure Review:
This event did not result in a safety system functional failure as defined by NEI 99-02, Revision 5.
Similar Events
- 1. JAF LER-05-002 "Safety Relief Valve Setpoint Drift," June 6, 2005.
- 2. JAF LER-03-002 "Safety Relief Valve Setpoint Drift," October 16, 2003.
- 3. JAF LER-01-005 "Safety Relief Valve Setpoint Drift," August 17, 2001.
- 4. JAF LER-99-003 "Safety Relief Valve Setpoint Drift," March 16, 1999.
- 5. JAF LER-98-002 "Safety Relief Valve Setpoint Drift," April 9, 1998.
NRC'FORM 366A U.S. NUCLEAR REGULATORY COMMISSION
,(6-2004)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION 5
OF 5
NUMBER NUMBER James A. FitzPatrick Nuclear Power Plant 05000333 07 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
Failed Component Identification:
Manufacturer:
Target Rock Corporation Model Number:
7567F-10 NPRDS Manufacturer Code:
T020 NPRDS Component Code:
Valve FitzPatrick Component ID:
02RV-071A, B, C,F, H, J, K, & L
References:
- 1. JAF Condition Report CR-JAF-2007-02108, Root Cause Analysis Report, Seven of ten SRV pilots failed as-found testing (testing high out of tolerance).
- 2. JAF Condition Report CR-JAF-2007-01944, Method of removal of A SRV (02RV-71A) pilot assembly (serial number 1087) invalidates as-found testing.
- 3. JAF Condition Reports CR-JAF-2006-02384 and 04108, D SRV electric lift inoperable due to loose electrical connector.
- 4. JAF-RPT-04-00441, Supplemental Reload Licensing Report for James A. FitzPatrick Reload 16 Cycle 17.
- 5. NEDC-33087P, Rev. 1, J. A. FitzPatrick Nuclear Power Plant APRM/RBM/Technical Specifications /
Maximum Extended Operating Domain (ARTS/MEOD).