05000220/LER-2012-006, Technical Specification Required Shutdown Due to Containment Leakage

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Technical Specification Required Shutdown Due to Containment Leakage
ML13050A022
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 02/11/2013
From: Philippon M
Constellation Energy Nuclear Group, EDF Group, Nine Mile Point
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 12-006-00
Download: ML13050A022 (8)


LER-2012-006, Technical Specification Required Shutdown Due to Containment Leakage
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown

10 CFR 50.73(a)(2)(iv)(A), System Actuation

10 CFR 50.73(a)(2)(v), Loss of Safety Function

10 CFR 50.73(a)(2)(i)

10 CFR 50.73(a)(2)(vii), Common Cause Inoperability

10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded

10 CFR 50.73(a)(2)(viii)(A)

10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition

10 CFR 50.73(a)(2)(viii)(B)

10 CFR 50.73(a)(2)(iii)

10 CFR 50.73(a)(2)(ix)(A)

10 CFR 50.73(a)(2)(x)

10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor

10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat

10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2202012006R00 - NRC Website

text

Michel A. Philippon P.O. Box 63 Plant General Manager Lycoming, New York 13093 315.349.5205 315.349.1321 Fax CENG a joint venture of Costellation eF NINE MILE POINT NUCLEAR STATION February 11, 2013 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION:

Document Control Desk

SUBJECT:

Nine Mile Point Nuclear Station Unit No. 1; Docket No. 50-220 Licensee Event Report 2012-006, Technical Specification Required Shutdown Due to Containment Leakage In accordance with 10 CFR 50.73(a)(2)(i)(A), 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(2)(v)(C),

please find attached Licensee Event Report 2012-006, Technical Specification Required Shutdown Due to Containment Leakage.

There are no regulatory commitments in this submittal.

Should you have questions regarding the information in this submittal, please contact John J. Dosa, Director Licensing, at (315) 349-5219.

Very truly yours, MAP/GvN

Attachment:

Licensee Event Report 2012-006, Technical Specification Required Shutdown Due to Containment Leakage cc:

NRC Regional Administrator NRC Project Manager NRC Resident Inspector

ATTACHMENT LICENSEE EVENT REPORT 2012-006 TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO CONTAINMENT LEAKAGE Nine Mile Point Nuclear Station, LLC February 11, 2013

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Nine Mile Point Unit 1 05000220 1 OF 6
4. TITLE Technical Specification Required Shutdown Due to Containment Leakage
5. EVENT DATE
6. LER NUMBER
7. REPORT DATE
8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR NA NA NUMBER NO.NAA FACILITY NAME DOCKET NUMBER 12 13 2012 2012 006 00 2

11 2013 1NA NA

9. OPERATING MODE
11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)

El 20.2201(b)

El 20.2203(a)(3)(i)

[E 50.73(a)(2)(i)(C)

El 50.73(a)(2)(vii)

N 0l 20.2201(d)

El 20.2203(a)(3)(ii)

El 50.73(a)(2)(ii)(A)

[I 50.73(a)(2)(viii)(A)

El 20.2203(a)(1)

[E 20.2203(a)(4)

El 50.73(a)(2)(ii)(B)

El 50.73(a)(2)(viii)(B)

10. POWER LEVEL I] 20.2203(a)(2)(i)

El 50.36(c)(1)(i)(A)

El 50.73(a)(2)(iii)

El 50.73(a)(2)(ix)(A)

El 20.2203(a)(2)(ii)

El 50.36(c)(1)(ii)(A)

Z 50.73(a)(2)(iv)(A)

[I 50.73(a)(2)(x)

[I 20.2203(a)(2)(iii)

El 50.36(c)(2)

El 50.73(a)(2)(v)(A)

El 73.71(a)(4) 100 [1 20.2203(a)(2)(iv)

[I 50.46(a)(3)(ii)

El 50.73(a)(2)(v)(B)

El 73.71(a)(5)

El 20.2203(a)(2)(v)

[D 50.73(a)(2)(i)(A)

[

50.73(a)(2)(v)(C)

El OTHER El 20.2203(a)(2)(vi)

[E 50.73(a)(2)(i)(B)

El 50.73(a)(2)(v)(D)

Specify in Abstract below or in C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE 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 HPCI 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.

NRC FORM 366 (10-2010)

4

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-201-3 land 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.

III. 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.

NRC FORM 366 (10-2010)

I

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 IV-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 next refueling outage, remove 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.
2. During the next refueling outage, coat the internal piping and associated attachments to prevent corrosion for the four horizontally-oriented containment vent and purge isolation valves IV-201-09, IV-201-10, IV-201-31 and IV-201-32.
3.

During the next refueling outage, inspect 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 and, based on the inspection findings, develop a cleaning/coating strategy (if required).

V. ADDITIONAL INFORMATION

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.

NRC FORM 366 (10-2010)

C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION IDENTIFIER AND SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS LER:

COMPONENT Reactor Protection System High Pressure Coolant Injection System Reactor Pressure Vessel Containment Isolation Valve Nitrogen Injection System Turbine Bypass Valve D. SPECIAL COMMENTS:

None IEEE 803 FUNCTION IDENTIFIER N/A N/A RPV ISV N/A N/A IEEE 805 SYSTEM IDENTIFICATION JC BJ AD NA NA JI NRC FORM 366 (10-2010)