ML18095A271

Submits Info in Support of 900522 Verbal Request for Relief from Requirements of ASME Section XI

ML18095A271
Person / Time
Site:	Salem
Issue date:	06/06/1990
From:	Labruna S Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLR-N90112, NUDOCS 9006150008
Download: ML18095A271 (5)

Text

Public Service Electric and Gas Company Stanley LaBruna Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4800 Vice President - Nuclear Operations June 6, 1990 NLR-N90112 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

REQUEST FOR ASME SECTION XI RELIEF SALEM GENERATING STATION UNIT NO. 1 DOCKET NO. 50-272 on May 22, 1990, Public Service Electric and Gas Company (PSE&G) contacted Messrs. R. Herman and s. Lee of your staff to discuss the acceptability of a non-ASME Code temporary repair which had been performed on a section of piping located in the Salem Unit No. 1 Service Water System. As a result of that conversation, PSE&G forwards the following information in support of our verbal request for relief from the requirements of ASME Section XI.

Background

In January 1990, PSE&G performed a temporary modification (TMOD) to a section of Service Water piping for the purpose of terminating a through wall leak. The TMOD consisted of welding a patch over the defect. The TMOD was conducted in accordance with existing procedures controlling these activities and in accordance with a program which had been previously reviewed and concurred to by our Authorized Nuclear Inspector. Under that program, piping defects are first reviewed to determine the potential impact on the structural integrity of the system. Upon completion of the structural review, an appropriate temporary modification is specified and implemented in order to support continued system operation until the next scheduled refueling outage. At that time the TMOD is removed and a Code acceptable repair is completed.

This TMOD was reviewed during an NRC Integrated Performance Assessment Program (IPAT) Team Inspection conducted recently at Salem. The !PAT team questioned the acceptability of the program implemented by PSE&G and recommended that approval for the temporary modification be obtained from the NRC off ice of Nuclear Reactor Regulation (NRR). Based on subsequent discussions with both NRR and NRC Region I, PSE&G is reviewing the basis for its present program. The following paragraphs address those items of additional information requested by the NRR during the May 22nd, telecon. __

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Document Control Desk 2 6-6-90 NLR-N90112 Impracticality of Code Repair Upon identification of the leak in the llSW header, the impact on continued plant operation was evaluated. This evaluation included Ultrasonic Examination (UT) of the affected area to determine the extent of the defect, followed by a structural analysis to assess adequacy of the piping subsystem in the as-found condition. The structural analysis used fracture mechanics techniques to evaluate the as-found defect. On the basis of that analysis, an immediate repair was recommemded.

In order to perform a Code acceptable repair, access to the interior surface of the header would be required. This could only be accomplished by complete isolation of the header and removal of 11SW22. This approach was identified as a high risk approach as a large number of safe shutdown systems would have been without cooling water for an extended period.

Installation of the TMOD required the draining of the header from the Intake Structure to valve 11SW22, and access to the defect from the external side of the piping only. In this case, only the buried portion (i.e. yard piping) of 11 Header was rendered inoperable and all equipment served by both trains remained operable during the repair.

The completion of a TMOD instead of a code acceptable repair was determined to be the safer of the two options as it did not render any safety related equipment inoperable. A Code acceptable repair is scheduled for completion during the upcoming Salem Unit 1 ninth refueling outage.

Repair Material Compatibility with Code Regµirements The existing piping material is ASTM A-333, grade 1, and the temporary patch was fabricated from a piece of 24 11 diameter x.

1/2" wall thickness SA106 Grade B pipe. Both material ,

specifications fall within the P No. 1 group, No. 1 grouping of base metals for qualification (QW-422) in Section IX of the ASME Boiler and Pressure Vessel Code. The P No. assignments of Section IX are based essentially on comparable base metal characteristics such as composition, weldability, and mechanical properties. A 3/8" fillet weld using 3/32" diameter E6010 .

welding electrodes for the first pass, followed by 3/32 11 diameter E7018 (low hydrogen coating) welding electrodes for the remainder of the weld joint was used for the plate to pipe attachment.

Document Control Desk 3 6-6-90 NLR-N90112 The above welding electrodes were selected in accordance with Nuclear Department Welding procedure 35 Rev.3, which is compatible with the welding procedure specifications which would have been chosen to weld piping in the Service Water System during original installation.

Piping Stress Review Upon identification of the leak in Service Water piping spool No. l-SW-116, a review of the affected piping stress calculation was performed in order to assess the structural integrity of the piping system. The following code criteria was used to determine compliance with the design basis calculations;

1. Minimum thickness of the pipe wall required for design pressure and temperature was calculated per section 101.1.2 of the ANSI B31.1 Power Piping Code 1977 Edition.

2. Stresses due to sustained and occasional loads, thermal expansion stresses and sustained plus thermal expansion stresses were calculated per section 104.8 of the ANSI B31.1 Power Piping Code 1977 Edition.

Based on the review of previously calculated stresses, a defect of .65 in. was determined to be limiting. The application of a welded patch was recommended based on the as-found defect

(.8 in.) exceeding the limiting defect.

As a result of the above, the temporary patch with fillet welds was evaluated with a stress intensification factor of 2.1 applied at this location in accordance with the above code. The calculated piping stress levels in the vicinity of the identified leak were as follows:

Stresses Due to Sustained Loads 4332 psi (Equation 11)

Stresses Due to Occasional Loads 5809 psi (Equation 12)

Thermal Expansion Stress 1232 psi (Equation 13) sustained Plus Thermal Expansion Stress 5564 psi (Equation 14)

Limiting Code Allowable Stress (Sh) 13,100 psi As can be seen, as-left stress levels for all loading conditions are much less than the allowable levels.

Document Control Desk 4 6-6-90 NLR-N90112 Safety Analysis of As-Found Condition A hole size of 0.8 inch diameter was evaluated and determined to result in a break flow of 160 gpm. The floor drain system can carry away 80 gpm. There are three sumps inside the service water piping rooms. These sumps are connected by drain lines to an external sump at 78 foot elevation. The most likely flooding scenario is that an equilibrium flood level (probably a few feet) would be reached in the break room. Half of the break flow (80 gpm) is carried away by the service water piping drain system and the remaining 80 gpm goes to the external sump which will overflow. The excess 80 gpm will be carried away by the drains at the 78 foot elevation with no flooding external to the service water piping rooms.

Although complete flooding of the affected service water piping room is not likely, it is assumed that one train of the service water system is lost. The three watertight doors Aux c11-1, C12-1 and C13-1 are maintained closed and would prevent multiple room flooding and the loss of more than one train. Loss of one train would cause LCO TSAS 3.7.4.1 to be entered and the inoperable train restored to operable status or the Unit shut down after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Root Cause of Defect Although the root-cause of this specific defect (Spool-1SW-116) has not yet been determined, similar failures in the service water system have been determined to be the result of erosion/corrosion due to the high degree of silt entrainment, and the corrosive characteristics of the saline brackish water taken from the Delaware River.

Should you have any questions, feel free to contact us.

Sincerely,

... -~ 1 Document Control Desk 5 6-6-90 NLR-N90112 c Mr. J. c. Stone Licensing Project Manager Mr. T. Johnson Senior Resident Inspector Mr. T. Martin, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625