ML12254A378

Relief Request Number ISI-09 - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii) Component Cooling Water 1 Inch Half-Coupling Weld Leak

ML12254A378
Person / Time
Site:	Ginna
Issue date:	09/06/2012
From:	Mogren T Constellation Energy Nuclear Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RR ISI-09
Download: ML12254A378 (10)

Text

Thomas Mogren Manager, Engineering Services R.E. Ginna Nuclear Power Plant, CENG LLC Ontario, New York 14519-9364 a joint venture of 585.771.5208 O coost"r=8ý En'grgeDF Thomas.Moq ren(&cenqllc.com September 6, 2012 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION: Document Control Desk

SUBJECT:

R.E. Ginna Nuclear Power Plant Docket No. 50-244 Relief Request Number ISI-09 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)

Component Cooling Water 1 Inch Half-Coupling Weld Leak Pursuant to 10 CFR 50.55a(a)(3)(ii), Constellation Energy Nuclear Group (CENG) requests relief from the American Society of Mechanical Engineers (ASME) Code Section Xl, 2004 Edition, No Addenda. By this request CENG is seeking relief from the requirement to perform code repairs during the next scheduled refuling outage (RFO).

Relief Request ISI-09 (Enclosure 1) is being submitted to support a planned repair during the 2014 RFO. CENG requests approval by October 8, 2012 in support of the upcoming 2012 RFO.

There are no new regulatory commitments identified in this correspondence. If you have any questions or need any other clarifying information, please contact Thomas L.

Harding, at (585) 771-5219.

Sincerely, Thomas Mogren

Enclosure:

(1) Relief Request Number ISI-09 cc: M.C. Thadani, NRC Ginna Resident Inspector, NRC W.M. Dean, NRC A-4 L'7

ENCLOSURE 1 Relief Request Number ISI-09

Relief Request Number ISI-09 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)

Component Cooling Water 1 Inch Half-Coupling Weld Leak Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety.

1. ASME Code Component(s) Affected Class 3, Component Cooling Water System, 1" half-coupling fillet weld associated with Temperature Element TE-621.

2. Applicable Code Edition and Addenda

American Society of Mechanical Engineers (ASME) Code Section Xl, 2004 Edition, No Addenda.

3. Applicable Code Requirement

10CFR50.55a approved the use of ASME Section Xl Code, 2004 Edition, No Addenda including Paragraph IWA-4133. Paragraph IWA-4133 deals with "Mechanical Clamping Devices used as Piping Pressure Boundary." IWA-4133 states that mechanical clamping devices used to replace piping pressure boundary shall meet the requirements of Mandatory Appendix IX.

Mandatory Appendix IX provides the requirements for Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundaries.

4. Reason for Request

On July 11, 2012 a small leak was identified on a Class 3, Component Cooling Water (CCW)

System, 1" Half-Coupling to fillet weld associated with a temperature element (TE-621) off a 14" pipe. The TE-621 is located on a common pipe run downstream of both Service Water (SW) to Component Cooling Water (CCW) heat exchangers. The fillet weld is the second weld out from the TE-621 branch connection off of the 14" common header pipe. The leak rate at this location prior to installation of the mechanical clamping repair was 1 drop every 4 minutes. To characterize the defect, Ginna Station's NDE Level III Inspector performed a VT-1 visual examination of the leaking location, in accordance with the requirements of the ASME Section XI Code. The visual examination that was performed by the NDE Level III inspector characterized the defect as an original welding imperfection, near a weld stop/start, with some minor undercut, along with lack of fusion at the half-coupling side bottom of the joint, approximately 3/8" to Y2" long. The location of the thermo well is approximately I inch above the sleeve for pipe support CCU-138, where it passes through the floor penetration. The Page I of 8

Relief Request Number ISI-09 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)

Component Cooling Water 1 Inch Half-Coupling Weld Leak Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety.

proximity to the pipe support sleeve makes this area difficult to access for welding. This proximity would result in restricted access for the welder and is potentially a contributing cause of the lack of fusion. The base metal on either side of the weld was examined for thickness by NDE using Ultrasonic Testing (UT) prior to installing the mechanical clamp and no material loss was identified.

The system fluid is potassium chromated water at an operating temperature of approximately 90 o F and an operating pressure of approximately 85 psig. The system design temperature and pressure at the location is 2000 F and 150 psig, respectively. Due to the location of this leak, the weld of concern cannot be isolated for performance of a welded repair without configuring the plant for a loss of all CCW operation. A mechanical clamp designed, fabricated and installed to the requirements of ASME Code Section Xl Mandatory Appendix IX was successful in stopping the leak. Since the location of this weld is un-isolatable from the common header, the performance of a corrective action welded repair/replacement activity will require a complete core defuel to establish a plant condition where CCW is not required. The upcoming 2012 Refueling Outage will start in October 2012 and is a normal fuel shuffle refueling outage and not a complete core defuel outage.

Relief Request Number ISI-09 is being submitted to provide an alternative to the Mandatory Appendix IX mechanical clamping devices for Class 2 and 3 piping pressure boundaries, Article IX-1000 (a). Article IX-lO00 (a) states that mechanical clamping devices used as a piping pressure boundary may remain in service until the next refueling outage, at which time the defect shall be removed or reduced to an acceptable size. All other applicable requirements of Mandatory Appendix IXwere satisfied.

Under Mandatory Appendix IX, a mechanical clamp that provides structural integrity could be used for a period of up to 24 months until the next refueling outage (if a plant is on a 2-year refueling cycle) where corrective action would be required. At the R. E. Ginna Nuclear Power Plant, the spring 2014 RFO will be a planned core defuel outage; the plant is currently on an 18 month refueling cycle. Postponing the corrective action activity to the 2014 RFO would result in the clamp being installed for a total time period of 21 months (from 7/11/2012 to May 2014).

The present refueling outage (RFO) cycles for R. E. Ginna Nuclear Power Plant were changed in 2009 to ensure full core offload RFOs occur in the spring and fuel shuffles occur in the fall. This was based on INPO Significant Operating Experience Report (SOER) 09-01, Shutdown Safety, in Page 2 of 8

Relief Request Number IS1-09 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)

Component Cooling Water 1 Inch Half-Coupling Weld Leak Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety.

regards to providing increased margins for time to boil considerations in the Spent Fuel Pool (SFP) for a Loss of Decay Heat Removal event.

For the current 2012 RFO, one third of Ginna Station fuel assemblies are planned to be discharged to the Spent Fuel Pool (SFP). To perform a corrective action activity on this low energy original construction defect, a full defuel to the SFP will be required. Changing the scope of the outage to a full core offload is a significant hardship due to the significant planning and scheduling resources required to expedite a normal 18 month planning process.

In addition to the hardship of implementing a full core offload, human performance associated with scheduled development and implementation is expected to be challenged by altering the outage schedule at this time. While these challenges can be managed, the measures required to perform a Repair/Replacement activity during the 2012 RFO do not provide a compensating increase in the level of quality or safety to account for the impacts that would be recognized.

5. Proposed Alternative and Basis for Use R. E. Ginna Nuclear Power Plant proposes that, instead of performing corrective actions in accordance with Mandatory Appendix IX,IX-1000(a) during the next refueling outage (2012 Refueling Outage), corrective actions shall be performed during the 2014 scheduled defueling outage.

The use of ASME Code Section XI Mandatory Appendix IXthat deals with the use of mechanical clamping devices is an approved configuration that provides reasonable assurance of structural integrity on a small diameter low energy line. The code recognizes the approved use of mechanical clamping devices for up to a full fuel cycle duration of 24 months which is a longer duration than is needed to support Ginna Station's 2014 Refueling Outage in May 2014.

Engineering evaluations were performed to (1) verify the integrity of the mechanical clamping device throughout the proposed time period, (2) identify all potential degradation mechanisms that could have propagated this original weld defect to cause a leak, (3) determine flaw tolerance conservatively assuming the mechanical clamp was not installed throughout the proposed time period, and (4) determine the consequences should the one inch line completely fail. This approach ensures adequate defense-in-depth in the highly unlikely event that the mechanical clamp should fail.

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Relief Request Number IS1-09 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)

Component Cooling Water 1 Inch Half-Coupling Weld Leak Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety.

(1) Integrity of the mechanical clamping device:

The mechanical clamping device, which is currently installed at this location, was designed and approved as a Safety Related Temporary Modification. The Temporary Modification was designed in accordance with all site design requirements, system pressure and temperature and material requirements, and incorporated all requirements of ASME Section XI Mandatory Appendix IX. The clamp and sealant material were selected for chemical compatibility with the potassium chromate in the system, and was analyzed to ensure that the sealant material would not have adverse effects on the water chemistry of the system. The sealant material is a silicon-elastomer material with fiber reinforcement, which has good chemical compatibility with potassium chromate. The compatibility with potassium chromate was determined based on published literature, and by performing accelerated laboratory experimentation with elevated temperatures and elevated chemical concentration. The clamp and the half-coupling are made of carbon steel, which addresses the potential for galvanic corrosion. The thermal well side of the clamp has a stainless steel/ carbon steel interface; however there is no source of fluid or electrolytes in that location to initiate galvanic corrosion. Therefore, both the clamp material and the sealant material are chemically compatible with the system piping and fluids.

The impact of installation includes the seismic impacts of the additional weight, the thermal growth of the pipe, and the internal stresses of the clamp. These impacts were all analyzed and approved in the temporary modification. The clamp is designed to mechanically clamp on the exposed pipe on either side of the effected weld utilizing setscrews on the header side, and mechanical crunch teeth to create a mechanical seal on both ends. In accordance with ASME Section XI Appendix IX,the friction force that the set screws impart on the half-coupling exceeds the maximum longitudinal load by a safety factor of 5. As additional conservatism, the maximum longitudinal load occurs during the injection of the sealant material (-394 lbf) rather than the result of a postulated full circumferential severance of the pipe at normal operating pressure (-89 lbf). Although no particular stressors have been identified, the ejection forces of the worst-case failure were evaluated and determined that the clamp would maintain pressure-boundary integrity of the branch connection in the event of a worst-case circumferential failure.

The clamp is designed to mechanically clamp on the exposed pipe outside of the weld area; such that the affected weld area is entirely covered by the clamp. This design ensures that any potential propagation of the welding defect would be contained within the clamp.

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Relief Request Number ISI-09 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)

Component Cooling Water 1 Inch Half-Coupling Weld Leak Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety.

In the improbable event that the flaw grew beyond the value provided in the bounding evaluation, the clamp has been designed and evaluated to accommodate the design forces of the system. In addition to the required frictional forces to prevent ejection, the clamp shell and enclosure fasteners were evaluated for all loading conditions including pressure, deadweight, thermal expansion, and seismic.

The engineering evaluation also addressed the impacts of the additional weight of the clamp on the seismic loading, and determined there were no aggregate impacts on the 1 inch pipe, half-coupling, the main 14 inch pipe, or the nearest lateral and vertical pipe supports. To analyze the seismic impacts, the entire weight of the clamp (which includes sealant and hardware), was applied as a lump sum on the far edge of the clamp to bound the Center of Gravity effects. In addition to a safety factor of 2, the evaluation applied a highly conservative 5g orthogonal loading condition. This conservative analysis determined that the resulting stresses due to seismic loading were significantly below code allowable stresses for all components.

(2) Potential Degradation Mechanisms:

An engineering evaluation (failure modes and effects analysis (FMEA)) of all potential degradation mechanisms was performed. Based on the results of the FMEA, the following were the most probable failure mechanisms that were identified that could not immediately be eliminated based on available data:

" Vibration induced fatigue crack of the existing defect

" Corrosion propagated the existing defect

" External force propagated the existing defect The cause of the leak cannot be conclusively determined without removal of the 1 inch pipe.

Vibration and corrosion were included in the flaw tolerance engineering evaluation. External forces could not be completely eliminated as a potential cause. However, the flaw is located at the bottom of the pipe which would require upward force for a tensile failure. It is therefore unlikely that this caused the existing defect to propagate. The area has been clearly marked to recognize the hazard and any potential fall hazards were eliminated. The flaw tolerance evaluation does not include this potential failure mechanism.

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Relief Request Number IS1-09 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)

Component Cooling Water 1 Inch Half-Coupling Weld Leak Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety.

Flaw Tolerance Evaluation:

The installed mechanical clamp was not credited for any positive contributions to mitigate the degradation. The weight of the mechanical clamp was included in the evaluation. The degradation mechanisms evaluated included vibration and corrosion. UT thickness measurements confirm that surrounding material has not experienced unexpected wear. The flaw tolerance evaluation concluded that fatigue cracking cannot be initiated nor can it propagate due to vibration, given the small stress field present in this portion of the CCW system. The applied alternating stress intensity of 6.7 ksi is below the material endurance limit of 12.5 ksi for the carbon steel base metal. Therefore, vibration induced stress levels are insufficient to allow crack initiation to occur. The fracture mechanics evaluation also concluded that the maximum applied stress intensity factor of 2.9 ksi/in 2 does not exceed the code allowable stress intensity factor of 14.1 ksi/in 2 . In addition, the maximum stress intensity factor range (AK) of 3.77 ksi/in 2 is below the AIK threshold for fatigue crack growth of 4.0 ksi/in 2 .

One possible cause of the defect propagating is due to internal corrosion. In this scenario, the weld defect has grown to a pinhole leak by a form of corrosion, such as galvanic corrosion at the carbon-stainless fillet weld. Based on industry studies, typical galvanic corrosion rates of 10-30 mil/year have been noted at carbon steel/stainless steel dissimilar metal welds. Using a bounding corrosion rate of 30 mil/year results in a metal loss of 0.053 inches over the 21 month period which was added to the crack growth calculations noted above.

To address possible undetected subsurface cracks along the weld fusion line, hypothetical circumferential flaws were assumed with maximum allowable flaws determined. The crack growth evaluations have shown that a 173-degree through-wall circumferential flaw and a 360-degree allowable flaw with a depth per thickness ratio of 48% can be tolerated for the 21 months of operation. Both of these analyses include metal loss due to the calculated bounding corrosion rate.

(3) Consequences of the 1 Inch Line Failure:

Although the flaw propagation analysis has determined the pipe will not fail, and the clamp is designed to withstand the worst-case postulated failure, the impacts of a highly improbable circumferential pipe failure have been analyzed and determined to be acceptable to the plant.

If the 1 inch line were to fail, the low-pressure and low-temperature water would spray onto Page 6 of 8

Relief Request Number ISI-09 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)

Component Cooling Water 1 Inch Half-Coupling Weld Leak Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety.

the nearby pipes, heat exchanger, and possibly spray the "drip-proof" CCW pump motor casing.

Based on the existing CCW Surge Tank level setpoints, a volume of 150 gallons could leak from the system prior to receiving a Main Control Room alarm. Upon confirmation of a lowering Surge Tank level, Operations would secure both CCW pumps and commence a plant shutdown in accordance with approved operating procedures. Upon securing the CCW pumps, the maximum volume of water which could leak out of this location is approximately 2000 gallons, which is significantly less than the values assumed in the Auxiliary Building flooding analysis.

Therefore the maximum flooding from a highly improbable circumferential pipe failure is bounded by the existing internal flood analysis.

The originating defect occurred during original plant construction. The original plant construction code at Ginna Station is B31.1 1955 Edition. Per this code, the fabrication inspection requirement for this line was a workmanship visual examination. Due to the small line size, no Radiographic (RT) examination was originally required to identify this original construction defect. A final Hydrostatic test was required to Paragraph 121 of B31.1 1955 Edition.

In accordance with the requirements of ASME Section XI, Appendix IX, Article 6000 the site has initiated the required monitoring of the defect and clamp. Engineering is currently performing a weekly inspection of the clamp for leakage. Engineering is also performing the quarterly volumetric examinations in accordance with Article IX-6000 (a), unless precluded by the clamping device configuration. This monitoring will be in effect until the mechanical clamp is removed and permanent repair is performed.

6. Duration of Proposed Alternative This relief request is applicable to R. E. Ginna Nuclear Power Plant's Fifth Interval ISI Program.

The duration of this relief request is through the 2014 Refueling Outage, Cycle #38 (May 2014).

The proposed alternative provides an acceptable level of quality and safety as a result of using the approved ASME Code Section XI Mandatory Appendix IX, Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundaries that provides structural integrity and could be used for a period of up to 24 months.

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Relief Request Number ISI-09 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)

Component Cooling Water 1 Inch Half-Coupling Weld Leak Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety.

7. Precedents None.

8. References

1. 10 CFR 50.55a, USNRC Codes and Standards

2. 2004 ASME Boiler and Pressure Vessel Code, 2004 Edition, No Addenda

3. 2004 ASME Boiler and Pressure Vessel Code, 2004 Edition, No Addenda, Mandatory Appendix IX, Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundary.

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