10-11-2017 | On August 6, 2017, with Unit 1 at 100% power, a Reactor Coolant System ( RCS) leak rate calculation determined the unidentified leak rate increased by 0.08 .gallons per minute. On August 8, a leak was obServed at an RCS hot leg.sample system valve,- and Unit 1 power level was reduced to investigate leakage indications. The. root isolation valve for the sample system valve was closed; however, leakage could not be verified as completely isolated. Further evaluation determined the leak to be through wall at the inlet of the sample system valve. Based upon the source of the leak and possible continued leakage, a Technical Specification shutdown clock was entered on August 9, at 13:38 hours. At 16:37 hours, Unit 1 was placed in Hot Shutdown.
The cause of the event was the RCS pressure boundary leakage at the tubing/socket weld area of the hot leg sample system valve. With the unit in Hot Shutdown, the leak was isolated and repaired, and Unit 1 was returned to power operation on August 11, 2017. An apparent cause evaluation is being conducted. The event was reported as a plant shutdown required by Technical Specifications pursuant to 10 CFR 50.72(b)(2)(i) and degraded condition pursuant to 10 CFR 50.72(b)(3)(ii)(A). This report is being provided pursuant to 10 CFR 50.73(a)(2)(i)(A) and 10 CFR 50.73(a)(2)(ii)(A). |
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Category:Letter
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[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000281/LER-2020-001, Cancellation of LER 2020-001-00 for Surry Power Station Unit 2 Re Loss of Containment Cooling Affecting Containment Partial Pressure Indication2020-11-0606 November 2020 Cancellation of LER 2020-001-00 for Surry Power Station Unit 2 Re Loss of Containment Cooling Affecting Containment Partial Pressure Indication 05000280/LER-2019-0022020-02-0404 February 2020 LER 2019-002-00 for Surry Power Station Unit 1, Items Non-Conforming to Design for Tornado Missile Protection 05000280/LER-2017-0012017-10-0606 October 2017 1 OF 3, LER 17-001-00 for Surry, Unit 1, Regarding Shutdown due to an Unisolable Leak in Reactor Coolant Pressure Boundary 05000281/LER-2016-0012016-12-0202 December 2016 Unit 2 Reactor Trip due to Generator Differential Lockout, LER 16-001-00 for Surry Power Station, Unit 2, Regarding Reactor Trip Due to Generator Differential Lockout 05000280/LER-2016-0012016-07-11011 July 2016 1 OF 4, LER 16-001-00 for Surry Power Station, Unit 1 and 2, Regarding Emergency Service Water Pump Inoperable Due to Corrosion of Valve Support ML1015505662009-03-0404 March 2009 Event Notification for Surry on Relief Valve Failure Results in Tritium and Cesium Spill ML0702501902007-01-17017 January 2007 Special Report on Qpt Not Less than Two Percent for Twenty Four Hours ML19105B1281979-08-21021 August 1979 LER 1979-011-03X for Surry Units 1 & 2 Loss of Power with Both Units at Cold Shutdown, Observed Radiation Monitors Without Power ML19093B4141978-09-26026 September 1978 Reporting Condenser Cooling Water Outlet Temperature to James River Decreased by 3.5oF as Measured at Station Discharge Structure ML19105A3001978-09-0808 September 1978 LER 78-013-03X-1, LER 78-029-03L-0, & LER 78-031-03L-0 for Surry 2 Re Critical Below Minimum Insertion Limit on D Control Bank ML19105A3011978-09-0101 September 1978 LER 1978-028-03L-0 for Surry 2 Re Containment Pressure Exceeded Allowable Limit ML19095A4971978-08-25025 August 1978 Submit Licensee Event Report No. LER 78-024/03L-0 Re No Alarm During a Planned Release of Liquid Radioactive Waste ML19095A4981978-08-22022 August 1978 Submit Licensee Event Report Nos. LER 1978-019/03L-0, LER 1978-021/03L-0, LER 1978-022/03L-0, LER 1978-023/03L-0, LER 1978-025/03L-0, LER 1978-026/03L-0 for Surry Unit No. 1 ML19105A0771978-08-16016 August 1978 Submit LER 78-027/03L-0 Re Seven Snubbers Not Meeting Inspection Criteria ML19105A0781978-08-16016 August 1978 Submit LER 78-027/03L-0 Re Seven Snubbers Not Meeting Inspection Criteria ML19105A0791978-08-11011 August 1978 LER 78-026/03L-0. Follow-up Inspection for Valve Cleanliness, Discovered That Containment Spray Isolation Valves (MOV-CS-201C and D) Contained No More That Five Milliliters. Technical Specification TS 6.6.2.b.(3) ML19095A4991978-08-0404 August 1978 Sutmit Licensee Event Report No. LER 78-018/03L-0 Re Two Main Steam Safety Valves Were Not within Plus or Minus 3% of Setpoint as Required ML19105A0801978-07-28028 July 1978 LER 1978-022-03L-0 Re Axial Power Distribution Monitoring Was Exceeded,& Three LERs 1978-021-03L-0, LER 1978-023-03L-0, & LER 1978-024-03L-0 Re Inlet Circulating Water Temperature Exceeded ML19095A5001978-07-28028 July 1978 Submit Licensee Event Report Nos. LER 78-017/03L-0 Re Not Documenting Four Electric Maintenance Procedures & Two Periodic Tests, & 78-020/03L-0 Re Suspended Air Bubbles in Reservoir Fluid ML19095A2281978-07-27027 July 1978 Reporting & Providing Additional Information on Incidence of Missing Two Steam Generator Tube Plugs Found During Inspection ML19095A2291978-07-26026 July 1978 Reporting of Cooling Water Discharge Temperature Change Rate on 7/9/1978 & Found No Evidence of Any Detrimental Effects in River or Its Inhabitants ML19095A5011978-07-21021 July 1978 Submit Licensee Event Report No. LER 78-016/03L-0 Re Valve MOV-RH-100 Was Not Opening ML19105A0821978-07-17017 July 1978 LER 1978-025-03L for Surry Unit 2 Excessive Gap Around Battery Room Ventilation Duct ML19105A0811978-07-17017 July 1978 LER 1978-025-03L for Surry Unit 2 Excessive Gap Around Battery Room Ventilation Duct ML19095A5021978-07-0707 July 1978 Submit Licensee Event Report Nos. LER 78-013/03L-0 Re Six Snubbers Not Meeting Test Criteria, & 78-014/03L-0 Re Mount Plate Was Warped, & 3 of 8 Bolts Were Insufficiently Embedded ML19095A5031978-06-27027 June 1978 Submit Licensee Event Report No. LER-78-015/03L-0 Re Valve Cleanliness Inspection Found Foreign Material on Valve Seating Surfaces ML19095A5041978-06-27027 June 1978 Submit Licensee Event Report No. LER-78-015/03L-0 Re Valve Cleanliness Inspection Found Foreign Material on Valve Seating Surfaces ML19105A0831978-06-23023 June 1978 LER 1978-017-03L, LER 1978-018-03L, LER 1978-019-03L, & LER 1978-020-03L for Surry Unit 2 Related to Snubbers ML19095A2481978-06-19019 June 1978 06/19/1978 Letter Reporting Increase of Condenser Cooling Water Discharge ML19095A5051978-06-0707 June 1978 Submit Licensee Event Report No. LER-78-012/03L-0 Re Inside Recirculation Spray Pump 1-RS-P-1A, Failure to Rotate by Hand ML19105A0851978-05-23023 May 1978 LER 1978-015-03L for Surry Unit 2 Finding of Air Leak on Header of Control Room Emergency Air Supply Bank ML19095A5061978-05-23023 May 1978 Submit Licensee Event Report Nos. LER-78-008/03L-0, LER-78-009/03L-0, LER-78-010/03L-0, LER-78-011/03L-0 for Surry Unit No. 1 ML19105A0861978-05-12012 May 1978 LER 1978-013-03L for Surry Unit 2 Reactor Brought Critical Below Minimum Insertion Limit on D Control Bank ML19105A0871978-05-0505 May 1978 LER 1978-014-03L for Surry Unit 2 Charging Pump Service Water Pump a Tripped on High Current ML19095A5071978-05-0505 May 1978 Submit Licensee Event Report No. LER-78-007/03L-0 Re Leak on Body to Bonnet Joint Diaphragm Valve 1-CH-98 ML19105A0881978-04-19019 April 1978 LER 1978-006-01T for Surry Unit 2 Review of Fire Protection Yard Piping Revealed That Two Pipe Extensions May Not Have Been Seismically Qualified ML19095A5081978-04-19019 April 1978 Submit Licensee Event Report No. LER-78-006/01T-0 Re Seismically Disqualified Two Pipe Extensions on Fire Protection Yard Piping ML19095A5091978-04-0707 April 1978 Forwarding Licensee Event Report No. LER-78-005/01T-0 Re Westinghouse'S Error in LOCA-ECCS Evaluation Model, Which Will Result in Higher Calculated Peak Clad Temperature ML19105A0901978-04-0606 April 1978 LER 1978-012-01T for Surry Unit 2 Failure to Meet Containment Integrity ML19105A0921978-04-0404 April 1978 LER 1978-011-03L for Surry Unit 1 Loss of Service Water Flow Through Charging Pump Intermediate Seal Cooler ML19095A5101978-04-0404 April 1978 Submit Licensee Event Report No. LER-78-004/03L-0 Re Loss of Service Water Flow Through Charging Pump Intermediate Seal Cooler ML19105A0931978-03-30030 March 1978 LER 1978-008-03L Re Snubber Body in Contact with Structure; LER 1978-009-03L Re Trip of Boric Acid Transfer Pump; & LER 1978-010-03L Re Failure of Pressurizer Channel ML19095A5111978-03-28028 March 1978 Submit Licensee Event Report No. LER-78-003/03L-0 Re Failed Heat Tracing, Panel 8 Circuit 24 ML19095A5121978-03-21021 March 1978 Submit Licensee Event Report No. LER-78-002/01T-0 Re Additional Radial Peaking Factor Surveillance ML19105A0941978-02-27027 February 1978 LER 1978-006-03L Re Failure of Sola Transformer, & LER 1978-007-03L Re Trip of Boric Acid Transfer Pump on Thermal Overload ML19095A5131978-02-24024 February 1978 Submit Licensee Event Report No. LER-78-001/03L-0 Re Hydraulic Shock Suppressor Found Inverted with Empty Reservoir ML19105A0951978-02-16016 February 1978 LER 1978-003-03L Re Erroneous Readings of Loop a Hot Leg Temperature Protection Channel; LER 1978-004-03L Re Unqualified Welding of Two Diaphragm Welds on Heat Exchanger; & LER 1978-005-03L Re Low Current Condition from Heat Tracing ML19105A0971978-02-0202 February 1978 LER 1978-001-03L for Surry Unit 2 Inoperability of Condenser Circulating Water Outlet Valve ML19095A5141978-01-0909 January 1978 Submit Licensee Event Report Nos. LER-77-021/03L-0 & LER-77-022/03L-0 ML19105A0991978-01-0909 January 1978 LER 1977-021-03L Re Routine Sampling of Bast a Revealed Tank Was at Concentration of 11.5%, & LER 1977-022-03L Re Failure of Comparator to Trip During Routine Test of a Steam Generator Level Protection Channel II 2020-02-04
[Table view] |
Reported lessons leamed are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S.
Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, 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.
Surry Power Station, Unit 1 05000- 280
3. LER NUMBER
2017 001 00 1.0 DESCRIPTION OF THE EVENT On August 6, 2017 at 18:55 hours, with Unit 1 operating at 100% power, containment air particulate radiation monitor alert annunciator was received in the main control room. A Reactor Coolant System (RCS) leak rate calculation was performed and indicated that the unidentified leak rate had increased by 0.08 gallons per minute (GPM). Actions were taken to investigate and evaluate the source of leakage. On August 8, at 10:55 hours, steam was identified coming from an RCS hot leg sample system valve in an RCS loop room. At 16:32 hours, power reduction was initiated to facilitate personnel entry into the loop room to investigate and isolate the leakage. The root isolation valve was closed, however, leakage could not be verified as completely isolated. On August 9, 2017, at approximately 13:38 hours, with Unit 1 at 32% power, a through wall pressure boundary leak was identified at the inlet to a 3/8 inch RCS sample system valve (EIIS-AB-HCV). Since the source of the leakage could not be verified as drainage from the isolated tubing, Unit 1 was shut down for repairs at 16:37 hours. On August 10, 2017, additional isolation valves were closed and the sample system tubing was cut and capped on each side of the RCS sample system valve to repair the leak.
Unit 1 was returned to power operation on August 11, 2017.
A four-hour report was made to the NRC pursuant to 10 CFR 50.72(b)(2)(i) due to a plant shutdown required by Technical Specifications and an eight-hour report was made pursuant to 10 CFR 50.72(b)(3)(ii)(A) due to degradation of a principal safety barrier. This report is being made pursuant to 10 CFR 50.73(a)(2)(i)(A) and 10 CFR 50.73(a)(2)(ii) (A), as a plant shutdown required by Technical Specifications and as a condition resulting in the degradation of a principal safety barrier.
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS Technical Specification Limiting Condition for Operation (LCO) 3.1.C, RCS Operational Leakage, does not permit pressure boundary leakage through a unisolable fault in an RCS component body, pipe wall, or vessel wall. Since it could not be verified that the pressure boundary leak was completely isolated after closing the root isolation valve, Unit 1 shutdown was required due to the degradation of a principal safety barrier. Leakage was confined to small bore 3/8 inch tubing. No reactor protection actuation occurred, nor was actuation necessary. A complete rupture of this sample line would have resulted in RCS leakage within the capacity of the normal makeup system, and therefore, Emergency Core Cooling System (ECCS) actuation would not have been expected to occur. Therefore, this event is of minimal safety significance.
3.0 CAUSE OF THE EVENT An apparent cause evaluation is being conducted. The equipment that was leaking is located in a high radiation area in an RCS loop room inside containment. Therefore, component removal, failure analysis, and permanent repair will be performed during the next Unit 1 refueling outage.
4.0 IMMEDIATE CORRECTIVE ACTIONS While Unit 1 was at Hot Shutdown, additional isolation valves were closed, and the area of leakage was isolated from the RCS. The tubing was cut and capped on each side of the RCS hot leg sample system valve.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S.
Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, means used to impose an informa on 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.
Surry Power Station, Unit 1 05000- 280
3. LER NUMBER
001 00 2017 5.0 ADDITIONAL CORRECTIVE ACTIONS An American Society of Mechanical Engineers (ASME) Code repair will be completed, and the tubing and the RCS hot leg sample system valve will be restored to design specifications and functionality during the next Unit 1 refueling outage.
6.0 ACTIONS TO PREVENT RECURRENCE Corrective actions determined by the apparent cause evaluation will be implemented in accordance with the corrective action program.
7.0 SIMILAR EVENTS None 8.0 MANUFACTURER/MODEL NUMBER NUPRO SS-6UG-TN3-4C 9.0 ADDITIONAL INFORMATION Unit 2 was at 100 percent power at the time of this event and was unaffected by the Unit 1 shutdown. There were no other structures, systems, or components that were inoperable at the start of the event that contributed to the event.