05000456/LER-2001-001-01, Three Main Steam Safety Valves Exceeded the Technical Specification Limit by Greater than 3% on September 19 and 20, 2001
| ML021710077 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 06/07/2002 |
| From: | Vonsuskil J Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BW020053 LER 01-001-01 | |
| Download: ML021710077 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 4562001001R01 - NRC Website | |
text
Exelon.
Exelon Generation Company, LLC wwwexeloncorp.com Niudaclear Braidwood Station 35100 South Rt 53, Suite 84 Braceville, IL 60407-9619 Tel. 815-417-2000 June 7, 2002 BW020053 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 1 Facility Operating License No. NPF-72 NRC Docket No. STN 50-456
Subject:
Submittal of Supplemental Licensee Event Report Number 2001-001-01, "Three Main Steam Safety Valves Exceeded The Technical Specification Limit By Greater Than 3%"
The enclosed Licensee Event Report (LER) is being submitted in accordance with 10 CFR 50.73, "Licensee event report system", paragraph (a)(2)(i)(B). This LER is a Supplement to LER 2001-001-00 which was submitted in November, 2001. At the time of LER 2001-001-00 submittal, the root cause for the Main Steam Safety Valve (MSSV) failures had not yet been determined. A failure analysis has been completed on one of the failed MSSV disks. The results of that failure analysis and additional corrective actions are discussed in the supplemental LER.
Should you have any questions concerning this letter, please contact Amy Ferko, Regulatory Assurance Manager, at (815) 417-2699.
Respectfully, 4mes D. von Suskil Site Vice President Braidwood Station
Enclosure:
LER Number 2001-001-01 cc:
Regional Administrator - Region IlIl NRC Braidwood Senior Resident Inspector
bcc: Braidwood Station Project Manager - NRR Nicholas Reynolds - Winston & Strawn Regulatory Assurance Manager - Braidwood Station Regulatory Assurance Manager - Byron Station Vice President - Licensing and Regulatory Affairs Director, Licensing - Midwest Regional Operating Group Manager, Licensing - Braidwood and Byron Stations Braidwood Nuclear Licensing Administrator Exelon Document Control Desk Licensing (Hard Copy)
Exelon Document Control Desk Licensing (Electronic Copy)
Abstract
During testing of the Unit 1 Main Steam Safety Valves (MSSVs) on September 19 and September 20, 2001, three valves lifted in excess of their setpoints by greater than the 3% Technical Specification (TS) tolerance. For each case where the MSSV lifted greater than the 3%
tolerance, the appropriate TS Condition was entered. The valves were tested sequentially.
Therefore only one valve was known to be inoperable at any time. The valves were adjusted as necessary and retested to place them in the required tolerance range prior to exiting the TS Condition. An evaluation has been performed by Nuclear Fuels Management with the results indicating the limits of the UFSAR accident scenarios impacted by the MSSVs have not been exceeded.
The results of an industry survey conducted by Braidwood Station engineers and a failure analysis performed by Altran Corporation indicate that a thicker oxide layer may decrease oxide bonding and minimize the sticking characteristics of the MSSVs.
However, Altran Corporation summarized that based on what was observed through laboratory investigation, no specific conclusions could be drawn or root cause identified as to why the X-750 disk from Braidwood exhibited the high lift setpoint phenomenon attributed to oxide bonding.
The results of the industry survey suggest that long continuous runs contribute to valve sticking.
Prior to the testing on September 19 and 20, 2001, Braidwood Unit 1 had completed a 535 day continuous run.
The oxide bond appears to be broken when valves are subjected to a thermal cycle between tests.
This event is being reported pursuant to 10CFR50.73(a)(2)(i)(B).
(Ifimore space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366X) (17) analysis was inconclusive, a correlation between oxide layer thickness and oxide bonding appears to exist.
As part of the failure analysis, a comparison was made between the disks from a Diablo Canyon valve and the disk removed from Braidwood's lMS014D.
Both disks were of the same Inconel X-750 disk material.
The analysis showed that the Diablo Canyon disk had an oxide layer approximately three times as thick as the disk from the lMS014D.
Diablo Canyon has 40 valves with the X-750 disk material installed and has not had test failures during insitu testing.
However, Diablo Canyon's practice of periodic lifts of the safety valves at power along with a number of plant thermal cycles makes it difficult to correlate data from the two valves.
The cause of the lMS017B to lift in excess of the +3% criteria is attributed to setpoint drift.
The lMS017B did not show signs of sticking and responded to adjustments in a predictable fashion.
The 1MS014D, lMS016B and lMS017B were all successfully left within 1% of setpoint as required per the Technical Specification Surveillance, BwVSR 3.7.1.1.
D.
Safety Consequences
The primary purpose of the MSSVs is to provide overpressure protection for the secondary system.
These valves also provide protection against over-pressurizing the reactor coolant (AB) pressure boundary by providing a heat sink for the removal of energy from the Reactor Coolant System if the preferred heat sink, provided by the Condenser (SD) and Circulating Water (KE) System, is not available.
The design basis for the MSSVs is to limit the secondary system pressure to
</=110% of design pressure for any Anticipated Operational Occurrence (AOO), or accident considered in the Design Basis Accident and transient analysis. The events that challenge the relieving capacity of the MSSVs, and thus RCS pressure, are those characterized as decreased heat removal events, which are presented in the Updated Final Safety Analysis Report (UFSAR), Section 15.2, "Decrease in Heat Removal by the Secondary System."
The LOCA Analyses and the Non-LOCA and Containment Analyses were evaluated by Exelon Nuclear Fuels Management with the results indicating the limits of the UFSAR accident scenarios impacted by the MSSVs remain bounding.
The event did not result in a Safety System Functional Failure.
E.
Corrective Actions
The high lift phenomenon has been an industry wide issue with respect to Consolidated Dresser 3700 series steam safety valves.
Based on the multiple industry wide events, a consortium of utilities along with EPRI funded a project to investigate the root cause of this phenomenon.
The findings of the root cause are published in EPRI Technical Report, TR-113560, and have been utilized as the basis for the corrective actions that have been implemented at Braidwood.
(Ifmore space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17) acceptance criteria and that the magnitude of the valves that lifted relatively low when compared to AlR08 data.
This data indicates the material provided improvement but has not completely eliminated seat high were X-750 disk bonding.
Failure analysis was performed on the disk that was removed from the lMS014D during refueling outage AlR09.
Based on the results of that analysis, a corrective action to prevent recurrence is implementation of an improved pre-oxidation process.
Additional corrective actions include biasing the setpoint to the low side of the li tolerance for valves refurbished with X-750 disks and to perform mid-cycle insitu testing on the five valves that were refurbished during refueling outage AlRO9.
Additionally, lMS017B will be tested prior to refueling outage AlR10.
If the valve lift setpoint is high, the valve will be refurbished.
F.
Previous Occurrences
Two previous events of MSSV high lifts have occurred at Braidwood Station.
These events were reported under LERs 98-004-00 and 2000-002-00.
G.
Component Failure Data
Manufacturer Dresser Nomenclature Main Steam Safety Valve Model 3707R Mfg. Part Number N/A