05000289/LER-2012-003, Regarding Pressurizer Heater Bundle Leak

Regarding Pressurizer Heater Bundle Leak

ML12298A035
Person / Time
Site:	Crane
Issue date:	10/22/2012
From:	Newcomer M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TMI-12-154 LER 12-003-00
Download: ML12298A035 (7)

LER-2012-003, Regarding Pressurizer Heater Bundle Leak

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 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)(iv)(A), System Actuation 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(b)(2)(ii)
2892012003R00 - NRC Website

text

A Exeton Generation Three Mile Island Unit 1 Route 441 South, P.O. Box 480 Middletown, PA 17057 October 22, 2012 TMI-12-154 Telephone 717-948-8000 10 CFR 50.73 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 THREE MILE ISLAND NUCLEAR STATION, UNIT 1 (TMI-1)

RENEWED FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 SUBJECT: LICENSEE EVENT REPORT (LER) NO. 2012-003-00 "Pressurizer Heater Bundle Leak" This report is submitted in accordance with 10 CFR 50.73 (a)(2)(i)(A). For additional information regarding this LER contact Mike Fitzwater, Sr. Regulatory Engineer, TMI Unit 1 Regulatory Assurance at (717) 948-8228.

There are no regulatory commitments contained in this LER.

Sincerely, Marrk Nevcomer Plant Manager, Three Mile Island Unit 1 Exelon Generation Co., LLC MN/mdf cc:

TMI Senior Resident Inspector Administrator, Region I TMI-1 Project Manager

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

10-2010)

, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.

13. PAGE Three Mile Island, Unit 1 05000289 1 OF 6

4. TITLE:

Pressurizer Heater Bundle Leak

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.

MONTH DAY YEAR N/A 05000 IFACILITY NAME DOCKET NUMBER 08 22 2012 2012 - 003 -

00 10 22 20121 N/A 05000

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)

El 50.73(a)(2)(i)(C)

El 50.73(a)(2)(vii)

N El 20.2201(d)

El 20.2203(a)(3)(ii)

[] 50.73(a)(2)(ii)(A)

El 50.73(a)(2)(viii)(A)

El 20.2203(a)(1)

El 20.2203(a)(4)

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

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

El 20.2203(a)(2)(i)

[] 50.36(c)(1)(i)(A)

El 50.73(a)(2)(iii)

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

10. POWER LEVEL El 20.2203(a)(2)(ii)

[: 50.36(c)(1)(ii)(A)

El 50.73(a)(2)(iv)(A)

El 50.73(a)(2)(x)

El 20.2203(a)(2)(iii)

El 50.36(c)(2)

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

El 73.71 (a)(4)

El 20.2203(a)(2)(iv)

El 50.46(a)(3)(ii)

[I 50.73(a)(2)(v)(B)

El 73.71(a)(5) 100 [1 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A)

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

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

El 50.73(a)(2)(i)(B)

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

Specify in Abstract below or in NRC Form 366A

12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (include Area Code)

Michael Fitzwater, TMI Unit 1 Regulatory Assurance Engineer (717) 948-8228MANU-REPORTABLE CAS MANU-REPORTABLE

CAUSE

SYSTEM COMPONENT FACTURER TO EPIX FACTURER TO EPIX B

AB HTR B015 Y

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED MONTH DAY EAR SUBMISSION El YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

On August 22, 2012 Three Mile Island (TMI) Unit 1 discovered an unisolable leak in the upper pressurizer heater bundle diaphragm plate and shut down the reactor. The cause of the leak was Primary Water Stress Corrosion Cracking (PWSCC). The root cause was determined to be "The use of Alloy 600 materials in high temperature locations was a design weakness in the construction of the TMI station."

The corrective actions include replacement of the upper pressurizer heater bundle (completed September 2012) and planned replacement of the remaining Alloy 600 susceptible pressurizer heater bundle. The leak was not a threat to the safety of the reactor and did not represent a reduction in the public health and safety.

A previous PWSCC condition was reported in LER 50-289/2003-003-00.

This LER is being submitted pursuant to 10 CFR 50.73(a)(2)(i)(A).

NRC FORM 366 (10-2010)U.S. NUCLEAR REGULATORY COMMISSION (10-2010)

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A.

EVENT DESCRIPTION

Plant Conditions before the event:

Babcock & Wilcox - Pressurized Water Reactor - 2568 MWth Core Power Date/Time: August 22, 2012 / 05:00 hours Power Level: 100% steady state power Mode: Power Operations There were no structures, systems, or components out of service that contributed to this event.

Background

TMI-1 was designed by Babcock and Wilcox (B&W), currently Areva. The pressurizer (PZR) was built with heater bundles that could be removed and replaced in the event that a pressurizer heater element would need to be replaced. Pressurizer heaters provide the function of maintaining RCS pressure. This function is accomplished during steady state conditions as well as transient or post trip recovery.

A diaphragm sits in a recessed channel in the stainless steel cladding on the pressurizer shell. The diaphragm is seal welded to the shell and is retained by a large bolted cover. The cover and associated bolting retain the majority of forces developed by the RCS.

On July 6, 2012 the Reactor Coolant System (RCS) leak rate step changed by an approximate factor of four. On August 12, 2012 the RCS leak rate changed again, this time by a factor of ten, placing the leak rate at approximately 0.25 gallons per minute (gpm). Significant efforts were undertaken to determine the source of the change in the leak rate. TMI technical specifications allow an unidentified leak rate of 1.0 gpm; however, no leakage is permitted through a primary pressure (strength) boundary.

Numerous walk down inspections were performed in the reactor building outside the D-rings but no source of this size leak was located. Camera and robotic inspections were performed inside the D-Rings and during the night shift on August 21/22, 2012 the leak was determined to be coming from the upper pressurizer heater bundle.

TMI-1 reported the event via EN 48220 on August 22, 2012 at 03:53 under 10 CFR 50.72(b)(2)(i) "the initiation of any nuclear plant shutdown required by the plant's Technical Specifications." The unit commenced a shut down in the early morning hours of August 22, 2012 and was down for approximately two weeks. During the shutdown the upper heater bundle was replaced with the stainless steel 12-element high watt density Watlow bundle which the station had purchased in 2004. This bundle is stainless steel and is not susceptible to PWSCC.

The source of the leak identified on August 22, 2012 was a crack in the diaphragm on the heater bundle.

The leak location was identified prior to the removal of the heater bundle. The relaxation of the stresses on the diaphragm allowed the crack to tighten sufficiently to prevent detection after shutdown. Very tight cracks can typically only be observed during destructive examination. Non destructive examination (NDE) such as the use of ultrasonic transducers (UT) and Dye Penetrant Testing (PT) have been shown to be unreliable in the detection of PWSCC through experience at TMI-1.

The diaphragm was not destructively examined to positively confirm PWSCC as the cause. However, the high level of susceptibility of Alloy 600 at the elevated pressurizer temperature coupled with the similar failure in the lower bundle in 2004 clearly indicate this defect was caused by PWSCC. TheU.S. NUCLEAR REGULATORY COMMISSION (10-2010)

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behavior of the defect was also noted to be similar to other verified PWSCC cracks. The amount of uncertainty is small and is considered acceptable by plant management to eliminate the added radiation dose destructive examination would have entailed. The diaphragm and its seal weld are primarily there to prevent water leakage past the pressure retaining structure.

TMI station replaced the upper bundle with the 12-element high watt density Watlow bundle that was purchased in 2004. The stainless steel diaphragm is not susceptible to PWSCC. Once the upper pressurizer bundle was replaced the station returned to power and RCS leak rate calculations were lower than prior to the event.

B.

CAUSE OF EVENT

Problem Statement (Root Cause):

The pressurizer heater bundle diaphragm was constructed of Alloy 600. 'This material is susceptible to PWSCC.

Statement of Cause (Root Cause):

The use of Alloy 600 materials in high temperature locations was a design weakness in the construction of the TMI station and the industry.

Basis for Cause Statement (Root Cause):

Alloy 600 is susceptible to cracking at the pressures and temperatures experienced at the pressurizer heater bundle location. A similar leak in 2003 resulted in a similar root cause performed in 2004. The action to resolve the 2004 root cause was to mitigate the Alloy 600, i.e., replace the heater bundles. The root cause in 2004 took appropriate actions. Efforts to obtain a heater bundle that was reliable was a continuing effort. As a result, the bundle was scheduled for replacement but had not been replaced at the time of the event.

Extent of Condition (Root Cause):

The extent of condition is limited to Alloy 600 susceptible materials. The pressurizer is considered the highest risk because the temperature is the highest. Corrective Actions to Prevent Recurrence (CAPRs) 1403278-15 and 1403278-16 will resolve the material issue for the pressurizer by replacing the middle bundle, which is the last external high temperature Alloy 600 location. Other alloy 600 locations within the TMI plant are being addressed under the Alloy 600 program.

Extent of Cause Cause Being Addressed Extent of Cause Root Cause:

New materials and designs that have not been The use of Alloy 600 materials in high proven through years of service are subject to temperature locations was a design failures that may not have been predicted nor weakness in the construction of the TMI expected. Appropriate NDE and monitoring station.

techniques in accordance with the codes and standards of construction are barriers to

Cause

degrading conditions causing material failures.

The susceptibility to cracking due toU.S. NUCLEAR REGULATORY COMMISSION (10-2010)

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high temperatures and internal stresses was not known at the time of design and construction. It took significant exposure to the temperature to indicate the material had an unfavorable crack initiation and propagation property.

This LER is being submitted pursuant to 10 CFR 50.73(a)(2)(i)(A).

C.

ANALYSIS / SAFETY SIGNIFICANCE Risk Assessment This event resulted in a leak in the diaphragm plate of a pressurizer heater bundle. The leak was not a threat to the safety of the reactor; however, the leak caused a mid-cycle shutdown that had radiological dose implications.

Actual Safety Consequences:

The actual safety consequence was minimal. The leak was approximately 0.25 gpm and the unit was shutdown in advance of significant material wastage or long term boric acid leakage. The pressurizer heater bundle diaphragm cover bolts were inspected when the heater bundle was replaced and found to have exhibited no loss of material. The diaphragm bolts were potentially exposed to boric acid from July 6, 2012 to August 23, 2012. Over this 48 day period the leak rate weighted average was 0.175 gpm.

This average leak rate had no effect on the bolts, and very little boric acid buildup collected around the bolts.

Safety Consequences if Occurred with a Design Basis Accident:

A design basis accident requires maximum injection flow from High Pressure Injection (HPI), Low Pressure Injection (LPI), and Core Flood. This large volume of makeup flow will pass through the core and exit the RCS via the break location (cold leg double ended guillotine shear). A small leak in the pressurizer diaphragm would not contribute to this event. The pressurizer would be empty in this event resulting in no contribution to the total accident response. Also, the RCS depressurizes during a design basis accident which would eliminate the diaphragm leak path as well.

System Functional Failure:

This event affected the pressurizer system function to maintain the integrity of the reactor coolant system pressure boundary. The small amount of leakage from the pressurizer heater bundle diaphragm was classified as a pressurizer system functional failure. The reactor coolant system leakage made a step change on July 6, 2012, and remained elevated until the plant was shutdown and depressurized (August 23, 2012). The system was fully capable of maintaining RCS pressure during the 48 day period of leakage. The risk of a leak that could affect reactor safety was small because the leakage potential for a cracked diaphragm is controlled by the strength of the cover which contains the majority of the forces developed by RCS pressure, not the diaphragm. The diaphragm and its seal weld are there to prevent water leakage past the pressure retaining structure, the cover, although the interior part of the diaphragmU.S. NUCLEAR REGULATORY COMMISSION (10-2010)

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plate where it attaches to the heater sheath serves a pressure boundary function.

As part of the inspection during the maintenance outage the remaining alloy 600 middle bundle was inspected. There was no evidence of leakage or observable degradation.

D.

CORRECTIVE ACTIONS

0 Perform replacement of the middle pressurizer heater bundle.

E.

PREVIOUS OCCURRENCES

Similar industry occurrences were reported over the last 30 years for PWSCC. Pressurizer heaters and nozzles were reported leaking after boric acid residues were found. The table below identifies TMI occurrences. The base materials were Alloy 600 and the failure mechanism was PWSCC. The TMI pressurizer design conditions (2500 psi and 670 degrees F) and normal operating conditions (2155 psig and 648 degrees F) meet the thresholds at which PWSCC has been shown to occur.

Previous Events Previous Event Reviiew TMI IR 184753, Pressurizer Heater Bundle Diaphragm Plate Leakage, November 4, 2003 The pressurizer lower heater bundle was identified as leaking during Outage T1 R1 5.. Weld repairs were made to the diaphragm to pressurizer seal weld and to the diaphragm plate itself. The root cause was determined to be inadequate design of the diaphragm plate (material selection) and weld materials. The materials were Alloy 600 and 82/182 which is known to be susceptible to PWSCC. The other two bundles were determined to be susceptible to PWSCC.

The lower bundle was repaired during one outage prior to the scheduled replacement (2005 refueling outage). The OE identified numerous instances of PWSCC cracking specific to pressurizer heater locations. CAPR 184753-25 was issued to revise engineering design standards addressing Alloy 600 and 82/182 material and their susceptibility to PWSCC. The lower bundle was subsequently replaced during T1 R1 5 when the bundle was identified leaking during start-up.

Corrective actions were issued to inspect the bundles during T1 R1 6 with insulation removed. The risk to the plant from this failure was considered low because the damage mechanism has a slow growth rate and RCS leakage is closely monitored. This IR has direct applicability to TMI and the decision to replace rather than repair the bundle during the 2012 outage was a direct lesson learned from the 2003 leak.

TMI IR 187903, Leak on the Lower Heater Bundle of Pressurizer, November 23, 2003 This IR documents that the seal weld, repaired as described under IR 184753, was the most probable root cause for this leak that was identified during start-up following the 2005 refueling outage. This IR was used in proceeding directly to replacement of the bundle in 2012.U.S. NUCLEAR REGULATORY COMMISSION (10-2010)

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Previous Events Previous Event Review INPO Level 3 Event This OE discusses significant damage and leakage to stainless steel heater Report 12-10, "Primary elements at a foreign plant. The initial leakage was due to stress corrosion Coolant Leak Caused cracking in a pressurizer heater sheath that allowed water ingress into the heater by Swelling and element. This water ingress resulted in swelling of the magnesium oxide Mechanical Failure of insulation used for the heaters which then swelled the heater element sheath Pressurizer Heaters" outside the pressurizer and resulted in a significant leak. This OE supports concerns over failed heater elements that goes beyond heater capacity alone and emphasizes the PWR fleet vulnerability to pressurizer heater performance.

Reference IR 1321322 for Exelon review of this OE. This event is not applicable to this root cause due to the mechanism and the material, although it is applicable to TMI.

• Energy Industry Identification System (EIIS), System Identification (SI) and Component Function Identification (CFI) Codes are included in brackets, [SI/CFI] where applicable, as required by 10 CFR 50.73 (b)(2)(ii)(F).