05000220/LER-2012-006

LER-2012-006, Technical Specification Required Shutdown Due to Containment Leakage

Nine Mile Point Unit 1
Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(iv)(A), System Actuation
Initial Reporting
ENS 48590	10 CFR 50.72(b)(3)(iv)(A), System Actuation, 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown
2202012006R01 - NRC Website

A. PRE-EVENT PLANT CONDITIONS:

Prior to the event, Nine Mile Point Unit 1 (NMP1) was operating at 100 percent of rated thermal power with no inoperable systems affecting this event.

B. EVENT:

NMP1 started up from the last outage on December 4, 2012. On December 6, 2012 nitrogen was added to the drywell in order to maintain torus pressure above its low pressure setpoint of 1.0 psig. Nitrogen was added again on December 8 and 10, 2012. This adverse trend in nitrogen makeup resulted in generating Condition Report 2012-011157 on December 10, 2012. Operator walkdowns in the Reactor Building did not detect any drywell leakage. Nitrogen was added again to the drywell on December 11, 2012. Based on the continuing adverse trend in nitrogen makeup, on December 12, 2012, extensive walkdowns, valve position verifications, and leak tests on systems connected to the drywell were performed. At the same time, engineering analyses were initiated to quantify the amount of containment leakage based on the nitrogen makeup frequency. The primary containment was still considered operable at this time based on available margin in the previously measured 10 CFR 50 Appendix J leakage rate totals. On December 12, 2012 at 17:56:33, the source of the drywell leakage was suspected to be drywell purge isolation valves IV-201-31(outboard) and IV-201-32 (inboard), as significant leakage was noted between the valves based upon displaced oxygen detected at the downstream vent. As a result, troubleshooting was initiated on these valves including valve lineups to isolate the nitrogen purge system, local leak rate tests (LLRT), pressure drop tests and leak checks through vented connections. Nitrogen was added again to the drywell on December 12, 2012. On December 13, 2012 at 15:08:27 hours, a LLRT was initiated for drywell purge isolation valves IV-201-31 and IV-201-32; however, the test was subsequently aborted when test pressure could not be achieved.

On December 13, 2012 at 16:30:10 hours, after concluding that the source and amount of the primary containment leakage could not be positively determined, the primary containment was declared inoperable due to a primary containment leakage rate that was in excess of the Technical Specification (TS) 3.3.3.a limit. TS 3.0.1 requires that in the event Limiting Condition for Operation (LCO) requirements cannot be satisfied because of circumstances in excess of those addressed in the Specification, the unit shall be placed in a condition in which the LCO is not applicable. A normal orderly plant shutdown was commenced at 16:45:00 hours and at 19:12:46 hours; a manual reactor scram was initiated from 18 percent power, in order to reduce the reactor coolant temperature to a value less than 215 degrees F within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of declaring the containment inoperable. At 23:33:00 hours, the reactor coolant temperature was below 215 degrees F.

This event involved a TS required shutdown, the manual initiation of the Reactor Protection System which resulted in a reactor scram, and an actuation of the high pressure coolant injection system. The NRC notification per 10 CFR 50.72(b)(2)(i), 10 CFR 50.72(b)(3)(iv)(A), and 10 CFR 50.72(b)(3)(v)(C) was completed on December 13, 2012, at 18:36 (Event Number 48590).

EVENT:

The degraded sealing surfaces of drywell purge isolation valves IV-201-31 and IV-201-32 contributed to this event. There were no other inoperable structures, systems, or components that contributed to the event.

D. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

December 10, 2012 21:00:00 Condition Report CR-2012-011157 was initiated after an adverse trend was identified in nitrogen makeup to primary containment.

December 12, 2012 17:56:33 Significant leakage was noted between the drywell purge isolation valves IV-201- 31(outboard) and IV-201-32 (inboard) based upon displaced oxygen detected at the downstream vent.

December 13, 2012 15:08:27 LLRT was initiated for drywell purge isolation valves IV-201-31(outboard) and IV-201- 32 (inboard). The test was subsequently declared a failure.

16:30:10 The primary containment was declared inoperable due to a primary containment leakage rate that was in excess of the TS 3.3.3.a limit.

16:45:00 A normal orderly plant shutdown was commenced.

19:12:46 A manual reactor scram was initiated from 18 percent power in order to reduce the reactor coolant temperature to a value less than 215 degrees F within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of declaring the containment inoperable.

19:13:09 The High Pressure Coolant Injection (HPCI) system automatically initiated on low Reactor Pressure Vessel (RPV) water level as expected due to RPV level shrink following the scram.

19:13:12 RPV level was restored above the HPCI system low level actuation setpoint and the IIPCI system initiation signal was reset.

23:33:00 The reactor coolant temperature was below 215 degrees F.

E. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

Other reactor and secondary systems functioned as expected.

F. METHOD OF DISCOVERY:

This event was discovered by the operators when the LLRT on drywell purge isolation valves IV-201- 31 (outboard) and IV-201-32 (inboard) was declared a failure.

G. MAJOR OPERATOR ACTION:

Operators initiated the manual reactor scram per plant procedure. The HPCI system initiation signal was reset after the RPV level was restored above 53 inches. Pressure control was established on the Turbine Bypass Valves, which is the preferred system.

H. SAFETY SYSTEM RESPONSES:

Following initiation of the reactor scram, all control rods fully inserted. The HPCI system automatically initiated on low RPV level as expected due to RPV level shrink following the scram. No other operational conditions requiring the response of safety systems occurred as a result of this event.

II. CAUSE OF EVENT:

The cause for this event was excessive leakage through drywell purge isolation valves IV-201-31and IV-201- 32. The general corrosion layer that developed on the purge line carbon steel piping over time became disturbed causing portions of the corrosion to break loose and collect inside the containment vent and purge piping. Containment vent and purge events carried loose corrosion from within the piping through the containment vent and purge isolation valves causing increased leakage due to uneven valve seating. This event was entered into the Nine Mile Point Nuclear Station corrective action program as condition report number CR-2012-011247.

HI. ANALYSIS OF THE EVENT:

This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(A) as a plant shutdown required by Technical Specifications, with 10 CFR 50.73(a)(2)(iv)(A) as an actuation of the high pressure coolant injection system, and with 10 CFR 50.73(a)(2)(v)(C) as a condition that could have prevented the fulfillment of a safety function of a system needed to control the release of radioactive material.

There were no actual nuclear safety consequences associated with this event. All control rods fully inserted following the reactor scram. The HPCI system automatically initiated on low RPV level as expected, due to RPV level shrink following the scram. There were no other automatic initiations of safety systems, and immediate actions performed by the operators were adequate and appropriate in placing and maintaining the reactor in a safe shutdown condition. The reactor scram was without complications and was not risk significant.

The closest related transient described in the NMP1 Updated Final Safety Analysis Report (UFSAR) is the Turbine Trip with Partial Bypass (Low Power) event described in UFSAR Section XV-B.3.14. The maximum reactor pressure and peak neutron flux reached during the December 13, 2012 event were both less than the calculated values presented in the UFSAR analysis for a low power turbine trip with partial bypass flow. In addition, this transient event does not challenge the Minimum Critical Power Ratio safety limit and, therefore, is not evaluated on a reload cycle basis.

Based on the above discussion, it is concluded that the safety significance of this event is low and the event did not pose a threat to the health and safety of the public or plant personnel.

This event affects two NRC Regulatory Oversight Process (ROP) performance indicators (PIs). The PI for Unplanned Power Changes per 7000 Critical Hours is projected to rise from 0.85 to 1.73, compared to the Green-to-White threshold value of 6. Also, the PI for Safety System Functional Failures rises from 3 to 4, compared to the Green-to-White threshold value of 6. Both PIs will remain green.

IV. CORRECTIVE ACTIONS:

A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:

1. Disassembled, inspected and cleaned containment vent and purge isolation valves IV-201-09, IV- 201-10, IV-201-31 and IV-201-32 and associated piping. All retrievable loose corrosion was removed.

2. Performed local leak rate testing on valves IV-201-07, IV-201-08, IV-201-09, IV-201-10, IV-201-16, IV-201-17, IV-201-31 and TV-201-32.

3. Pressurized containment to approximately 1.5 psig and trended containment pressure to verify there was no significant leakage.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

1. During the 2013 refueling outage, removed corrosion from the internal surfaces of the four horizontally-oriented containment vent and purge isolation valves IV-201-09, IV-201-10, IV-201-31 and IV-201-32 and associated piping. A visual inspection of internal piping between the primary containment and IV-201-09, IV-201-10, IV-201-31 and IV-201-32 will be performed every refueling outage.

2. During the 2013 refueling outage, inspected the internal piping and associated attachments of the four vertically-oriented containment vent and purge isolation valves IV-201-07, IV-201-8, IV-201-16 and IV-201-17. The inspections did not reveal any internal degradation that required cleaning.

A. FAILED COMPONENTS:

There were no other failed components that contributed to this event.

B. PREVIOUS LERs ON SIMILAR EVENTS:

There were no previous LERs on similar events.

C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION

THIS LER:

IEEE 803 FUNCTION IEEE 805 SYSTEM

COMPONENT IDENTIFIER IDENTIFICATION

Reactor Protection System N/A JC High Pressure Coolant Injection System Reactor Pressure Vessel Containment Isolation Valve Nitrogen Injection System Turbine Bypass Valve N/A BJ

ISV NA

N/A NA N/A JI

D. SPECIAL COMMENTS:

None