0-0-5000 | On March 10, 2009, during the performance of surveillance procedure 3-PT-R006A, main steam safety valves ( MSSV) MS-45-1 and MS-48-3 failed their As-Found lift set point test.T In accordance with the test, these valves must lift at +/- 3% of their required setting. Valve MS-45-1 lifted at 1112.7 psig outside its acceptance range of 1034 to 1096 psig. Valve MS-48-3 lifted at 1165.1 psig outside its acceptance range of 1077 to 1143 psig. All other MSSVs tested passed their As-Found test criteria and were left within +/- 1% of their required setting in accordance with the test procedure.
Technical Specification (TS) 3.7.1," Main Steam Safety Valves," requires the MSSVs to be operable in accordance with TS Table 3.7.1-1 and Table 3.7.1.-2.TTS Surveillance Requirement (SR)T 3.7.1.1 requires each MSSV be verified to lift per Table 3.7.1-2 in accordance with the Inservice Testing Program. Operability of the MSSVs includes the ability to open within the setpoint tolerances. As these two valves were found outside their limit they failed their As-Found testing. Section In accordance with NUREG-1022,T 3.2.2, reporting guidelines, the existence of similar discrepancies in multiple valves is an indication that the discrepancy may have arose over a period of time, and therefore existed during plant operation and is reportable.TThe apparent cause of the two MSSVs lifting greater than 3% of their nominal setpoint is indeterminate but most likely caused by setpoint drift.T Corrective actions included adjusting the valves, subsequent maintenance during the refueling outage, and re-testing with an As-Left setting within the +/- 1% As-Left set point criteria.T The event had no effect on public health and safety |
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LER-2009-002, Technical Specification Prohibited Condition Caused by Two Main Steam Safety Valves Outside Their As-Found Lift Setpoint Test Acceptance CriteriaIndian Point 3 |
Event date: |
05-11-2009 |
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Report date: |
0-0-5000 |
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Reporting criterion: |
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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2862009002R00 - NRC Website |
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Note: The Energy Industry Identification System Codes are identified within the brackets (1.
DESCRIPTION OF EVENT
On March 10, 2009, during the performance of surveillance procedure 3-PT-R006A, main steam (MS) (SB} safety valves (MSSVs) (RV} MS-45-1 and MS-48-3 failed their As-Found lift set point test. In accordance with the test, these valves must lift at +/- 3% of their required setting. Valve MS-45-1 lifted at 1112.7 psig outside its acceptance range of 1034 to 1096 psig. Valve MS-48-3 lifted at 1165.1 psig outside its acceptance range of 1077 to 1143 psig. All other MSSVs valves tested passed their As-Found test criteria and were left within +/- 1% -of their required setting in accordance with the test procedure.
Technical Specification (TS) 3.7.1,'Main Steam Safety Valves," requires the MSSVs to be operable in accordance with TS Tables 3.7.1-1 and 3.7.1.-2. TS Surveillance Requirement (SR) 3.7.1.1 requires each MSSV be verified to lift per Table 3.7.1-2 in accordance with the Inservice Testing (IST) Program.
Operability of the MSSVs is determined by periodic surveillance testing in accordance with the TS and IST program. As these two valves were found outside their limit they failed their AS-Found test criteria. MS-45-1 is associated with steam generator (SG)-31 and MS-48-3 is associated with SG-33.
There are five code safety valves (MSSVs) and one atmospheric dump valve (ADV) (RV} on each main steam (MS) line outside the Reactor Containment (NH} and upstream of the MS isolation valves (ISV}. The five code safety valves (MSSV) consist of four 6-inch by 10-inch and one 6-inch by 8-inch valve per SG on each of four MS lines. The valves are set to open at 1065, 1080, 1095, 1110, and 1120 psig. The operability of the MSSVs is defined as the ability to open within the set points tolerances, relieve SG overpressure, and reset when pressure has been reduced. The accident analysis requires five MSSVs per SG to provide overpressure protection for design basis transients occurring at 102% reactor thermal power. The MSSVs are Code relief valves, manufactured by Crosby Valve and Gauge Company {C710}. The 6-inch by 8-inch valves (e.g., MS 45-1) are Model # HC-65W-6Q8. The 6-inch by 10-inch valves (e.g., MS-48-3) are Model # HC-65W 6R10.
Cause of Event
The apparent cause of the two MSSVs lifting greater than 3% of their nominal setpoint is indeterminate but most likely caused by setpoint drift. MS-45-1 and MS-48-3 were disassembled and inspected and identified to have some scoring on their valve spindles. Assessment with Original Equipment Manufacturer (OEM) could not directly relate the indications discovered on the valves' spindles to the As-Found test results.
Corrective Actions
The following corrective actions have been performed under Entergy's Corrective Action Program to address the cause and prevent recurrence:
- Adjusted and tested each valve (MS45-1, MS-48-3) within the +/- As-Left set point criteria upon initial discovery of failures.
- Performed Preventive Maintenance (PM) and replaced the valve spindle on each valve during the refueling outage.
- Adjusted and tested each valve (MS-45-1, MS-48-3) with As-Left settings within the +/-1% As-Left setpoint criteria following the PM.
Event Analysis
The event is reportable under 10CFR50.73(a)(2)(i)(B). The licensee shall report any operation or condition which was prohibited by the plant TS. TS 3.7.1,"Main Steam Safety Valves," requires the MSSVs to be operable in accordance with TS Tables 3.7.1-1 and 3.7.1-2. TS Surveillance Requirement (SR) 3.7.1.1 requires each MSSV be verified to lift per Table 3.7.1-2 in accordance with the Inservice Testing Program. Operability of the. MSSVs includes the ability to open within the setpoint tolerances. As these two valves were found outside their limit they failed their As-Found testing criteria. In accordance with NUREG-1022, Section 3.2.2, reporting guidelines, the existence of similar discrepancies in multiple valves is an indication that the discrepancy may have arose over a period of time, and therefore existed during plant operation and is reportable.
Past Similar Events
A review was performed of Licensee Event Reports (LERs) for the past three years for any events reporting TS prohibited conditions due to multiple test valve failures and none were identified.
Safety Significance
This event had no effect on the health and safety of the public. There were no actual safety consequences for the event because there were no accidents or transients requiring the MSSVs.
There was no significant potential safety impact of the condition under reasonable and credible alternate conditions. Had an accident or transient occurred during the condition of the two out of tolerance MSSVs, the condition is judged to have not significantly affected accident mitigation capability and the MSSVs overpressure-function would have been adequate. The design basis of-the MSSVs is to limit the secondary system pressure to 110% of design pressure when passing 100% of design steam flow. The combined MSSVs are sufficient to relieve 108% of design steam flow. Each MS line has an ADV capable of releasing steam to the atmosphere. The ADVs have the capability to relieve approximately 10% of total steam. The combined pressure relief capability of the MSSVs and ADVs is approximately 118% of rated steam flow. Engineering judgment concluded adequate 20 MSSVs lifting at a higher pressure set point than specified.
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05000247/LER-2009-001 | Technical Specification Prohibited Condition Due to a Surveillance Requirement Never Performed for the Atmospheric Steam Dump Valve Local Nitrogen Controls | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000286/LER-2009-001 | 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249Entergy Tel (914) 734-6700 J. E. Pollock Site Vice President NL-09-114 October 30, 2009 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001 Subject:M Licensee Event Report # 2009-001-01, "Automatic Actuation of an Emergency Diesel Generator and Two Auxiliary Feedwater Pumps During Surveillance Testing due to Inadvertent De-Energization of the Normal Supply Breaker to 480 Volt Safeguards Bus 6A" Indian Point Unit No. 3 Docket No. 50-286 DPR-64 Dear Sir or Madam: Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides revised Licensee Event Report (LER) 2009-001-01. The attached revised LER identifies an event where there was an automatic actuation of an emergency diesel generator and two auxiliary feedwater pumps, systems listed in 10 CFR 50.73(a)(2)(iv)(B), which is reportable under 10 CFR 50.73(a)(2)(iv)(A) . The revised LER incorporates changes as a result of an evaluation of troubleshooting and testing performed during the Unit 3 refueling outage. This event was recorded in the Entergy Corrective Action Program as Condition Report CR-I P3-2009-00011. There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Licensing at (914) 734-6710. Sincerely, JEP/cbr cc:M Mr. Samuel J Collins, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 3 Mr. Paul Eddy, New York State Public Service Commission LEREvents@inpo.org NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104DEXPIRES: 8f31/2010 (9-2007) Estimated burden per response to comply with this mandatory collection request:D50 hours.DReportedDlessonsDlearnedDareDincorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by InternetLICENSEE EVENT REPORT (LER) e-mail to infocollects@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection. 1. FACILITY NAME: INDIAN POINT 3 2. DOCKET NUMBER 13. PAGE 05000-286 1 OF 5 4. TITLE: Automatic Actuation of an Emergency Diesel Generator and Two Auxiliary Feedwater Pumps During Surveillance Testing due to Inadvertent De-Energization of the Normal Supply Breaker to 480 Volt Safeguards Bus 6A | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000247/LER-2009-002 | Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused by a Loss of Main Feedwater Pump 21 and Failure of the Main Turbine to Automatically Runback | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000286/LER-2009-002 | Technical Specification Prohibited Condition Caused by Two Main Steam Safety Valves Outside Their As-Found Lift Setpoint Test Acceptance Criteria | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000247/LER-2009-003 | Loss of Single Train 21 Pressurizer Backup Heater Required for Remote Shutdown From the Control Room Due to an Inoperable Breaker | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000247/LER-2009-004 | Loss of Single Train 23 Charging Pump Required for Remote Plant Shutdown From the Control Room Due to a Failure of a Pump Internal Check Valve | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000247/LER-2009-005 | 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249Entergy Tel (914) 734-6700 J. E. Pollock Site Vice President NL-09-159 January 4, 2010 U.S. Nuclear Regulatory Commission
Attn: Document Control Desk
Mail Stop 0-P1-17
Washington, D.C. 20555-0001
SUBJECT:MLicensee Event Report # 2009-005-00, "Automatic Reactor Trip Due to a Turbine-Generator Exciter Protective Trip Caused by a Loss of the Generrex Power Supply Monitored Voltage Due to a High Resistance Ground Connection" Indian Point Unit No. 2 Docket No. 50-247 DPR-26 Dear Sir or Madam: Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2009-005-00. The attached LER identifies an event where the reactor was automatically tripped, which is reportable under 10 CFR 50.73(a)(2)(iv)(A) . As a result of the reactor trip, the Auxiliary Feedwater System was actuated and the Main Steam Isolation Valves (MSIVs) were closed which is also reportable under 10 CFR 50.73(a)(2)(iv)(A). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-IP2-2009-04530. There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Licensing at (914) 734-6710. Sincerely, -qrsuer-Pc,a JEP/cbr cc:MMr. Samuel J Collins, Regional Administrator, NRC Region I
NRC Resident Inspector's Office, Indian Point 2
Mr. Paul Eddy, New York State Public Service Commission
LEREvents@inpo.org
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES: 8/31/2010 (9-2007)D• Estimated burden per response to comply with this mandatory collection request: 50 hours.DReported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internetLICENSEE EVENT REPORT (LER) e-mail to infocollects@ nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection. 1. FACILITY NAME: INDIAN POINT 2 2. DOCKET NUMBER 1 3. PAGE 05000-247 1TOF 5 4. TITLE: Automatic Reactor Trip Due to a Turbine-Generator Exciter Protective Trip Caused by a Loss of the Generrex Power Supply Monitored Voltage Due to a High Resistance Ground Connection | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000286/LER-2009-005 | | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000286/LER-2009-006 | Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by Actuation of the Generator Protection System Lockout Relay During a Severe Storm with Heavy Lightning | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000286/LER-2009-007 | Automatic Reactor Trip Due to a Turbine Trip As a Result of Turbine Autostop Oil Actuation Caused by a Failed Autostop Oil Fitting | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000286/LER-2009-008 | Technical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for an Inoperable Over Power Delta Temperature (OPDT) Bistable | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000286/LER-2009-009 | Loss of Single Train Neutron Flux Detector N-38 Required for Plant Shutdown Remote From the Control Room Due to a Power Supply Failure | 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
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