05000306/LER-2004-001, Re Shutdown Required by Technical Specifications Due to Two Trains of Containment Cooling Inoperable

Re Shutdown Required by Technical Specifications Due to Two Trains of Containment Cooling Inoperable

ML050180089
Person / Time
Site:	Prairie Island  (DPR-060)
Issue date:	01/17/2005
From:	Solymossy J Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-04-127 LER 04-001-00
Download: ML050180089 (6)

LER-2004-001, Re Shutdown Required by Technical Specifications Due to Two Trains of Containment Cooling Inoperable

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(v), Loss of Safety Function
3062004001R00 - NRC Website

text

Prairie Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC JAN 1 7 2009 L-PI-04-127 10 CFR 50.73 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-000 1 Prairie Island Nuclear Generating Plant Unit 2 Docket 50-306 License No. DPR-60 LER 2-04-01, Unit 2 Shutdown Rewired by Technical SDecifications Due to Two Trains of Containment Coolina InoDerable The Licensee Event Report for this occurrence is attached. This event was reported via the Emergency Notification System in accordance with 10 CFR Part 50, Section 50.72, on November 17, 2004. Please contact us if you require additional information related to this event.

Summary of Commitments This letter contains one new commitment and no revisions to existing commitments.

Specifically, NMC commits that, following completion of additional cause investigation, Event Report.

Enclosure cc:

Administrator, Region 111, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC Glenn Wilson, State of Minnesota 171 7 Wakonade Drive East 0 Welch, Minnesota 55089-9642 Telephone: 651.388.1 121

ENCLOSURE LICENSEE EVENT REPORT 2-04-01 4 pages follow

NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 6-30-2007 FACILITY NAME (1)

Prairie Island Nuclear Generating Plant Unit 2 COMMISSION

6-2004)

LICENSEE EVENT REPORT (LER)

(See reverse for required number of digitslcharacters for each block)

DOCKET NUMBER (2)

PAGE (3) 05000 306 1 of 4

, the NRC may not d u c t or sponsor, and a person is not required to respond to, the information collection.

EVENT DATE (5)

MO DAY YEAR 11 17 04 LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

FACILITY NAME DOCKET NUMBER SEQUENTIAL REV YEAR NUMBER NO MO DAY YEAR 04

- 0 1

17 05 FACILITY NAME DOCKET NUMBER I

OPERATING THIS REPORT IS SUBYlTTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: Check all that aDDlV) f l l b MODE (9)

POWER LEVEL (10) 1 20.2201(b) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50.73(a)(2)(ix)(A) 20.2201(d) 20.2203( a )(4) 50.73(a)(2)(iii) 50.73(a)(2)(x)

O0 20.2203(a)(l) 50.36(c)(1 )(i)(A) 50.73(a)(2)(iv)(A) 73.71(a)(4) UAME Jeff Kivi I I20.2203(a)(3)(i)

I 150.73(a)(2)(ii)(A)

I 150.73(a)(2)(viii)(B) 1 LICENSEE CONTACT FOR THIS LER 112)

TELEPHONE NUMBER (Indude Area Code) 651.388.1 121

CAUSE

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

I I

I I

I I I

I I

I MANU-REPORTABLE MANU-REPORTABLE SYSTEM COMPONENT FACTURER TO EPlX

CAUSE

SYSTEM COMPONENT FA CTURER TO EPlX I

SUPPLEMENTAL REPORT EXPECTED (14) x YES (If yes, complete EXPECTED SUBMISSION DATE).

N O Repair of the leaks on the 22 and 23 CFCUs was completed on November 18,2004 and Unit 2 returned to power operation on November 19,2004.

EXPECTED MONTH DAY YEAR DATE (15) 3 21 2005 SUBMISSION

b h l

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTl N UATl ON TEXT (If more space is required. use additional copies of NRC Form 366A) (17)

EVENT DESCRIPTION

On November 17, 2004, Prairie Island Nuclear Generating Plant (PINGP) Staff entered Unit 2 containment to investigate suspected leakage of the 23 containment fan coil unit' (CFCU). The 23 CFCU was suspected to be leaking based on control room indication of condensate pot level and Sump A pump2 run times. The investigation identified two leaks on the 23 CFCU and also identified one leak on the 22 CFCU. The lack of control room indication of the leak on 22 CFCU is believed to be due to the leak spraying beyond the catch tray such that condensate pot level was not affected, but is under investigation.

The leaking CFCUs were isolated. Since the leaks affected one of two CFCUs in each of the two trains of containment cooling, and those CFCUs were isolated, both trains of containment cooling were declared inoperable and PINGP Technical Specification Limiting Condition of Operability (LCO) 3.0.3 was entered. The leaks could not be repaired in time to exit LCO 3.0.3 before a plant shutdown was required, thus, a plant shutdown was initiated and completed.

EVENT ANALYSIS

The containment cooling system3 is a closed system in containment. Integrity of the tubes in the CFCUs is required for containment integrity. The CFCU tubes are American Society of Mechanical Engineers (ASME) Code Class 2 and support two safety functions: (1) circulating cooling water for containment cooling in the event of a loss of coolant accident (LOCA) or main steam line break in containment, and (2) serving as a containment boundary during the first hours of a LOCA when containment pressure will exceed the pressure of the cooling water in the CFCUs.

The leakage was identified at H-bends in the faces of the affected CFCUs. The faces are multi-pass heat exchangers. The H-bends are brazed copper elbows which redirect cooling water flow through the CFCU faces.

Each CFCU has four faces and each containment has four CFCUs. Thus, there are 32 faces between Unit 1 and Unit 2. The original faces were supplied by Westinghouse. Subsequently, 24 of the faces have been replaced with a new design Aerofin coil. The eight remaining faces are original design, but are replacement faces that were installed in the 1988 to 1993 time frame. The leaks on the 22 and 23 CFCUs were all on Westinghouse design faces (the Aerofin coils do not have H-bends).

(EIIS Component Identifier: FCU)

(ENS Component Identifier: P)

(EIIS System Identifier: BK) 1 2

3

LICENSEE EVENT REPORT (LER)

TEXT (if more space is required, use additional copies of NRC Form 366A) (17)

Impact on Safety System Functional Failure Performance Indicator The CFCUs contain an air-to-water heat exchanger supplied with water from the cooling water system in order to remove containment heat. In this event, three of the tubes in the heat exchanger portion of 22 and 23 CFCUs were found to have pin-hole leaks - one leak on 22 CFCU and two leaks on 23 CFCUs. The leaks would not have impacted the ability of the fan coil unit to provide its containment heat removal function following any accident. However, the leaks would adversely impact the ability of the CFCUs to maintain the containment boundary following an accident. The design accident pressure for the containment structure is 46 psig. Due to head losses in the piping and high elevation of the FCU cooling water lines, the cooling water system can experience negative pressures in the return piping from the CFCUs. As a result, the leaks in the 22 and 23 CFCU H-bends represent a potential failure to maintain the containment boundary function following a postulated accident. To assure that the CFCU maintained their containment boundary function, the leaking CFCUs were isolated, rendering them incapable of performing their heat-removal function.

However, following initial identification of the leak and prior to isolating the CFCUs, the containment boundary function could not be assured; therefore, this event is reportable per 10 CFR 50.73(a)(2)(v)(C) as a Unit 2 Safety System Functional Failure.

SAFETY SIGNIFICANCE

As noted above, prior to isolating the leaking CFCUs, the containment heat removal function of each train was not adversely impacted by the leaks in the 22 and 23 CFCUs; however, the containment function was adversely affected during the short time prior to isolating the affected CFCUs. After isolation, the system design is such that containment leakage is precluded -the pressure on outboard side of the isolation valves for both the inlet and the outlet of the CFCUs is greater than containment pressure. There is no current analysis that quantifies the potential impact of this bypass leakage on the offsite or control room dose analyses, however by comparison with a previous analysis and engineering judgment this leakage is viewed as not significant. Based on the above, the health and safety of the public was not affected.

CAUSE

Nuclear Management Company, LLC, (NMC) conducted an apparent cause evaluation (ACE) to determine why the CFCU leaks occurred. The ACE concluded that the apparent cause of the CFCU leaks was the design of the H-bends, which causes turbulent flow, which, in turn, causes erosion in the areas where the joint brazing of the H-bend appears flawed. The presence of erosion was indicated by smooth, deposit free grooves. The erosion occurred around braze material that had overflowed from the brazed joints onto the internal surfaces of the H-bends. The buildup of material caused flow disruptions in the H-bends, resulting in grooving and eventual leak formation in the copper base material. Corrosion may also have contributed to the failures. Some corrosion of the

(1-2001)

A LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION TEXT (If more space is required, use additional copies of NRC Form 3664 (17)

I I braze material that was exposed to the inside diameter surface was observed. Any evidence of corrosion of the copper in the damaged areas was removed by the erosion. Additional evaluation of 1

the cause is in progress. Following completion of this additional evaluation, NMC will issue a I

supplement to this report.

CORRECTIVE ACTION

Immediate:

1. Unit 2 was shut down per Technical Specification 3.0.3.

Subsequent:

2. The leaking CFCUs were repaired per the ASME Code and Unit 2 was returned to power operation.

Planned:

3. Complete CFCU face replacement,

4. Analyze CFCU flow rates to determine if existing flow rates are contributing to erosion.

PREVIOUS SIMILAR EVENTS

No previous unit shutdowns have been completed due to CFCU leakage.