ML042680078

Third 10-Year Inservice Inspection Interval Program Plan

ML042680078
Person / Time
Site:	Susquehanna
Issue date:	09/24/2004
From:	Richard Laufer NRC/NRR/DLPM/LPD1
To:	Shriver B PPL Generation
Guzman R, NRR/DLPM 415-1030
References
TAC MC1185, TAC MC1186, TAC MC1191, TAC MC1192, TAC MC1193, TAC MC1194, TAC MC1195, TAC MC1196, TAC MC1197, TAC MC1198, TAC MC1199, TAC MC1200
Download: ML042680078 (22)

Text

September 24, 2004 Mr. Bryce L. Shriver President, PPL Generation and Chief Nuclear Officer PPL Generation, LLC 2 North Ninth Street Allentown, PA 18101

SUBJECT:

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 - THIRD 10-YEAR INSERVICE INSPECTION (ISI) INTERVAL PROGRAM PLAN (TAC NOS. MC1185, MC1186, MC1191, MC1192, MC1193, MC1194, MC1195, MC1196, MC1197, MC1198, MC1199, MC1200)

Dear Mr. Shriver:

By letter dated September 16, 2003, as supplemented on May 27, 2004, PPL Susquehanna, LLC (PPL, the licensee), submitted 10 Requests for Relief (RR) for Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), proposing alternatives to the requirements of Title 10 of the Code of Federal Regulations, Part 50, Section 55a (10 CFR 50.55a), concerning the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for its Third 10-Year ISI Program Plan. The Nuclear Regulatory Commission (NRC) staff has reviewed PPLs regulatory and technical analysis in support of its requests for relief for 3RR-03, 3RR-07, 3RR-08, 3RR-09, 3RR-10, and 3RR-11. The remaining requests for relief, 3RR-01, 3RR-02, 3RR-04, and 3RR-06, are currently under NRC review and will be issued as a separate safety evaluation (SE).

PPL is currently in its third 10-year ISI interval which began on June 1, 2004, and will end on May 31, 2014. The ISI Code of record for the third 10-year interval for SSES 1 and 2 is the 1998 Edition with the 2000 Addenda of the ASME Code,Section XI.

Based on the information provided by PPL, the NRC staff concludes that PPLs proposed alternatives for 3RR-03, 3RR-08, 3RR-09, 3RR-10, and 3RR-11 provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes the proposed alternatives for 3RR-03, 3RR-08, 3RR-09, 3RR-10, and 3RR-11 as described in PPLs letter dated September 16, 2003, for SSES 1 and 2 for the third 10-year ISI interval. In addition, the NRC staff has determined that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) for 3RR-07 is authorized by law and will not endanger life or property, or the common defense and security and is otherwise in the public interest giving due consideration to the significant burden upon the licensee that could result if the requirements were imposed on the facility. The NRC staff concludes that PPLs proposed alternatives for 3RR-07 provides reasonable assurance of structural integrity of the subject components; therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the NRC staff also authorizes the proposed alternatives for 3RR-07 for the third 10-year ISI interval.

B. Shriver If you have any questions, please contact the project manager, Rich Guzman, at (301) 415-1030.

Sincerely,

/RA/

Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-387 and 50-388

Enclosure:

As stated cc w/encl: See next page

B. Shriver If you have any questions, please contact the project manager, Rich Guzman, at (301) 415-1030.

Sincerely,

/RA/

Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-387 and 50-388

Enclosure:

As stated cc w/encl: See next page DISTRIBUTION:

PUBLIC PDI-1 RF RLaufer DTerao ACRS ALee RGuzman MOBrien GHill (2) OGC GMatakas, Rgn-1 HAshar KManoly JJolicoeur, Rgn-1, EDO SCoffin TMcLellan ACCESSION NO.: ML042680078 *SE provided. No substantive changes made.

OFFICE PDI-1/PM PDI-2/LA EMEB/S EMEB/SC* EMCB/SC OGC PDI-1/SC C*

• NAME RGuzman CRaynor for DTerao KManoly MMitchell HMcGurren RLaufer MOBrien DATE 9/24/04 9/24/02 7/7/04 2/18/04 SE 6/21/04 9/22/04 9/24/04 SE DTD DTD SE DTD OFFICIAL RECORD COPY

Susquehanna Steam Electric Station, Units 1 and 2 cc:

Britt T. McKinney Michael H. Crowthers Vice President - Nuclear Site Operations Supervising Engineer PPL Susquehanna, LLC Nuclear Regulatory Affairs 769 Salem Blvd., NUCSB3 PPL Susquehanna, LLC Berwick, PA 18603-0467 Two North Ninth Street, GENPL4 Allentown, PA 18101-1179 Richard L. Anderson Vice President - Nuclear Operations Dale F. Roth PPL Susquehanna, LLC Manager - Quality Assurance 769 Salem Blvd., NUCSB3 PPL Susquehanna, LLC Berwick, PA 18603-0467 769 Salem Blvd., NUCSB2 Berwick, PA 18603-0467 Aloysius J. Wrape, III General Manager - Nuclear Assurance Luis A. Ramos PPL Susquehanna, LLC Special Office of the President Two North Ninth Street, GENPL4 PPL Susquehanna, LLC Allentown, PA 18101-1179 634 Salem Blvd., SSO Berwick, PA 18603-0467 Terry L. Harpster General Manager - Plant Support Bryan A. Snapp, Esq PPL Susquehanna, LLC Assoc. General Counsel 769 Salem Blvd., NUCSA4 PPL Services Corporation Berwick, PA 18603-0467 Two North Ninth Street, GENTW3 Allentown, PA 18101-1179 Robert A. Saccone General Manager - Nuclear Engineering Supervisor - Document Control Services PPL Susquehanna, LLC PPL Susquehanna, LLC 769 Salem Blvd., NUCSB3 Two North Ninth Street, GENPL4 Berwick, PA 18603-0467 Allentown, PA 18101-1179 Rocco R. Sgarro Richard W. Osborne Manager - Nuclear Regulatory Affairs Allegheny Electric Cooperative, Inc.

PPL Susquehanna, LLC 212 Locust Street Two North Ninth Street, GENPL4 P.O. Box 1266 Allentown, PA 18101-1179 Harrisburg, PA 17108-1266 Walter E. Morrissey Director - Bureau of Radiation Protection Supervising Engineer Pennsylvania Department of Nuclear Regulatory Affairs Environmental Protection PPL Susquehanna, LLC P.O. Box 8469 769 Salem Blvd., NUCSA4 Harrisburg, PA 17105-8469 Berwick, PA 18603-0467 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 35, NUCSA4 Berwick, PA 18603-0035

Susquehanna Steam Electric Station, Units 1 and 2 cc:

Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Board of Supervisors Salem Township P.O. Box 405 Berwick, PA 18603-0035 Dr. Judith Johnsrud National Energy Committee Sierra Club 443 Orlando Avenue State College, PA 16803

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST NOS. 3RR-03, 3RR-07, 3RR-08, 3RR-09, 3RR-10, AND 3RR-11 FOR THE INSERVICE INSPECTION PROGRAM PLAN FOR THE THIRD TEN-YEAR INSPECTION INTERVAL PER THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE, SECTION XI, REQUIREMENTS PPL SUSQUEHANNA, LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS AND 2 DOCKET NOS. 50-387 AND 50-388

1.0 INTRODUCTION

By letter dated September 16, 2003 (ADAMS Accession No. ML032670839), as supplemented on May 27, 2004 (ADAMS Accession No. ML041630133), PPL Susquehanna, LLC (PPL, the licensee), submitted 10 Requests for Relief (RR) for Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), proposing alternatives to the requirements of Title 10 of the Code of Federal Regulations, Part 50, Section 55a (10 CFR 50.55a), concerning the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for its Third 10-Year Interval Inservice Inspection (ISI) Program Plan. The Nuclear Regulatory Commission (NRC) staff has reviewed PPLs regulatory and technical analysis in support of its requests for relief for 3RR-03, 3RR-07, 3RR-08, 3RR-09, 3RR-10, and 3RR-11. The remaining requests for relief, 3RR-01, 3RR-02, 3RR-04, and 3RR-06, are currently under NRC review and will be issued as a separate safety evaluation (SE).

PPL is currently in its third 10-year ISI interval which began on June 1, 2004, and will end on May 31, 2014. The ISI Code of record for the third 10-year interval for SSES 1 and 2 is the 1998 Edition with the 2000 Addenda of the ASME Code,Section XI.

The SE addresses (1) 3RR-03, related to the inclusion of ISI snubbers in the Technical Requirements Manual (TRM) Snubber Visual Examination and Testing Program, (2) 3RR-07, related to the exemption from pressure testing of the reactor pressure vessel head flange seal leak detection line, (3) 3RR-08, related to the pressure monitoring of the control rod drive (CRD) accumulators, (4) 3RR-09, related to the corrective measures for leakage at bolded connections, (5) 3RR-10, related to the synchronization of Class MC and Class CC component ISIs to coincide with the 10-year ISI interval for Class 1, 2, and 3 components, and (6) 3RR-11, related to the alternative requirements for IWE-5240 visual examination.

Enclosure

2.0 REGULATORY EVALUATION

Section 50.55a(g) requires that ISI of the ASME Code, Class 1, 2, and 3 components be performed in accordance with Section XI of the ASME Code and applicable addenda, except where specific written relief as been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). According to 10 CFR 50.55a(a)(3), alternatives to the requirements of paragraph (G) may be used, when authorized by the NRC, if an applicant demonstrates that the proposed alternatives would provide an acceptable level of quality and safety or if the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that ISI of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.

The applicable Code of record for the third 10-year ISI for SSES 1 and 2 is the 1998 Edition through the 2000 Addenda of the ASME Code,Section XI. The third 10-year interval starts June 1, 2004, and will end on May 31, 2014.

3.0 TECHNICAL EVALUATION

The NRC staff has reviewed PPLs regulatory and technical analysis in support of its requests for relief for the third 10-year ISI interval program plan for SSES 1 and 2. The detailed evaluation below supports the conclusion that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by the operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the relief will not be inimical to the common defense and security or to the health and safety of the public.

3.1 Request for Relief No. 3RR-03 Code Requirement PPL proposed an alternative to the requirements of the ASME Code,Section XI, 1998 Edition up to and including the 2000 Addenda, Article IWF-5000, with regard to snubber ISI activities.

The 1998 ASME Code 2000 Addenda,Section XI, IWF-5200(a) and IWF-5300(a) require preservice and inservice examinations to be performed in accordance with ASME/American National Standards Institute (ANSI) OM, Part 4, using the VT-3 visual examination method described in IWA-2213. Paragraphs IWF-5200(a), IWF-5300(a), IWF-520-0(b) and IWF-5300(b) reference the 1988 Addenda to ASME/ANSI OM-1987, Part 4 (OMa-1988, Part 4),

for the snubber visual examination and functional testing. In addition, Paragraphs IWF-5200(c) and IWF-5300(c) require that integral and non-integral attachments for snubbers, including lugs, bolting, pins, and clamps shall be examined in accordance with the requirements of the ASME Code,Section XI, Subsection IWF.

PPLs Basis for Relief PPL stated that the current SSES 1 and 2 TRM snubber program, Section 3.7.8, includes a comprehensive program for visual examination and functional testing of safety and non-safety related snubbers. There are approximately 956 safety-related and non-safety related snubbers at SSES 1 and 2. PPL stated that the overlap of the visual examination program required by ASME Code,Section XI and the TRM snubber program for the Code class snubbers presents an unnecessary redundancy without a compensating increase in the level of quality and safety.

PPL stated that the TRM snubber visual examination program requires a sample size of 100%

of all safety-related and non-safety related snubbers and incorporates the alternative snubber visual examination schedule delineated in NRC Generic Letter (GL) 90-09, Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions, dated December 11, 1990. Unlike OMa-1988, Part 4, the alternate schedule in GL 90-09 is based on the number of unacceptable snubbers found during the previous examination in proportion to the sizes of the various snubber populations or categories. The alternative examination interval is based on a refueling cycle of up to 24 months and may be extended to as long as two fuel cycles, depending on the number of unacceptable snubbers found during the previous visual examination.

ASME Code,Section XI, paragraphs IWF-5200(a) and IWF-5300(a) require that snubber visual examinations be performed using the VT-3 visual examination method described in paragraph IWA-2213. In its September 16, 2003, letter, PPL stated that visual examiners, who are qualified to the applicable rules of ASME Code,Section XI, Article IWA-2000, Examination and Inspection, will perform the examinations and tests of safety-related and non-safety related snubbers.

PPL stated in the above letter that the TRM snubber program maintains the same level of quality and safety as that of OMa-1988, Part 4, for the visual examination and functional testing of snubbers. However, PPL did not provide specific details of its program and the staff did not find an adequate basis for concluding that the program maintains the same level of quality and safety as that of OMa-1988, Part 4. In addition, PPL did not provide an adequate description of, and the basis for, its proposed alternate examination requirements for snubber attachments.

The staff raised these concerns to PPL during a teleconference on February 4, 2004, and issued a letter dated February 20, 2004, requesting additional information. In question No. 1 of the request for additional information (RAI-1), the staff requested that PPL discuss how the TRM snubber program maintains the required confidence level for the snubber surveillance activities. The staff also requested PPL to discuss the requirements of its snubber attachment examinations and how the alternative program compares with the requirements of ASME Code,Section XI, Subsection IWF. By letter dated May 27, 2004, PPL responded by stating that the requirements for its snubber program are detailed in each units TRM Section 3.7.8 and TRM Bases Section B 3.7.8. The program requires the following surveillance be performed:

 Demonstrate that each snubber is operable by performing a visual inspection with the inspection frequency based upon the number of unacceptable visually inspected snubbers in the previous interval with a maximum of 48 months.

 Functionally test a representative sampling of all snubbers once per 24 months.

 Monitor the installation and maintenance records for each snubber to ensure that the service life has not been exceeded and will not be exceeded prior to the next snubber surveillance inspection.

 Test the snubbers that are in locations of snubbers that failed the functional test during the previous test period.

 Perform an inspection on snubbers attached to sections of systems that have experienced potentially damaging transients as determined from a review of operational data and a visual inspection of the systems within.

PPL stated that the above requirements are implemented by documented SSES 1 and 2 procedures and specifications, namely, (1) NEPM-QA-0595, Snubber Program, (2)

Surveillance Procedures ST-100-001 for Unit 1 and ST-200-001 for Unit 2, Snubber Visual Inspection, and (3) Specification M-1090, Inservice Testing of Safety-Related Mechanical Snubbers, Hydraulic Snubbers and Compensating Struts.

Surveillance Procedures ST-100-001 and ST-200-001 require the following:

 Personnel performing this surveillance (visual inspection of snubbers) shall be VT-3 certified.

 The visual inspection shall include [the following examinations]:

1. Presence of mechanical interferences which would inhibit snubber motion.

2. Loss of integrity of pinned/bolted connections as evidenced by broken connections or disengaged pins/bolts.

3. Large dents or other major deformations in the Snubber housing, support cylinder, or position indicating tube to an extent that snubber movement would be hindered.

4. Scratches and other superficial surface defects, which are [generally]

acceptable.

5. Total separation of spherical ball bushings from their socket.

6. Material coating the exterior of the snubber to an extent that snubber movement would be hindered.

7. Anchor plate loose or pulled away from support structure or embedment material.

8. Embedment material is cracked/concrete is spalled.

9. Snubber is bottomed/topped out.

10. Cracked material/welds or bent components on hanger support structures, anchor plates, etc.

11. Loose clamping, loose clamp hex nut.

12. Snubber frozen in place and/or cold set out of tolerance (+/- 1/2").

13. Check hydraulic snubber for hydraulic fluid levels.

PPL stated that the Surveillance Procedures contain a checklist to document the above areas of inspection. This checklist encompasses the checklist from ASME/ANSI OMa-1988, Part 4. PPL stated that Specification M-1090 details the following functional testing requirements for snubbers:

a. Breakaway force test in both tension and compression,

b. Activation level test,

c. Post-Activation Displacement Test, and

d. Running Drag Force Test.

These functional testing requirements are further described in TRM Bases Section B 3.7.8, under TRS 3.7.8.2, and in TRM Table 3.7.8-3, which encompasses the requirement of ASME/ANSI OMa-1988, Part 4. In addition, a complete snubber program is described in NEPM-QA-0595.

PPLs Proposed Alternative Examination PPL proposed in its letter of September 16, 2003, for the third 10-year ISI interval that, as an alternative to the requirements of ASME Code,Section XI, paragraphs IWF-5200(a) and IWF-5300(a), IWF-5200(b) and IWF-5300(b), and IWF-5200(c) and IWF-5300(c), the snubber visual examinations and functional testing of all SSES 1 and 2, Class 1, 2, and 3 snubbers be performed in accordance with SSES 1 and 2 TRM Section 3.7.8.

NRC Staffs Evaluation Based on the information provided above, the NRC staff finds that SSES 1 and 2 TRM 3.7.8, supplemented by NEPM-QA-0595, Surveillance Procedures ST-100-001 and ST-200-001, and Specification M-1090, provides adequate requirements for visual examination and functional testing of SSES 1 and 2 snubbers, including Class 1, 2, and 3 snubbers. On this basis, the NRC staff concludes that the SSES 1 and 2 snubber program as presented in the TRM, and the reference documents, maintains the same level of quality and safety as that of ASME Code,Section XI, paragraphs IWF-5200(a), IWF-5300(a), IWF-5300(b) and IWF-5300(b).

In regard to snubber attachments, PPL stated in its May 27, 2004, letter, that the examination of snubber attachments as required by paragraphs IWF-5200(c) and IWF-5300(c), is performed in accordance with the TRM snubber program. PPL stated that the visual examination performed for the snubber, as described above, also covers the visual inspection for snubber attachments.

The NRC staff has reviewed the visual inspection requirements delineated in the above Surveillance Procedures ST-200-001 and ST-200-002, and finds them to adequately cover the visual inspection for the snubber attachments, and hence provides an acceptable level of quality and safety as that provided in the ASME Code, Subsection IWF.

In its alternative under Proposed Alternate Examination, PPL states, It should be noted that snubber welded attachments will be performed in accordance with the ASME Section XI Subsections IWB, IWC, and IWD welded attachment examination requirements. In the manner in which it was initially presented in the alternative, this statement seemed to imply that ASME Code requirements for welded attachments were somehow related to the proposed alternate examination. In RAI-2 of the February 20, 2004, RAI letter, the staff requested PPL to clarify this statement in the proper context of the alternative. The staff requested that PPL explain how Subsections IWB, IWC, and IWD requirements were related to the requirements of the TRM program, and how they were being used as an alternative examination requirement for snubber attachments in the alternative. In its letter of May 27, 2004, PPL clarified that Subsections IWB, IWC, and IWD are not being used as alternative examination requirements, and are not a part of 3RR-03. For snubbers, these requirements are in addition to the proposed alternative to use the PPL TRM snubber program. The staff finds that PPL adequately clarified its statement made in the alternative, regarding the use of ASME Code,Section XI, Subsections IWB, IWC, and IWD for the welded attachments and finds it acceptable.

In its review of PPLs TRM 3.7.8, the staff was not able to locate PPLs documented commitment to inspect all snubbers attached to piping systems that have experienced a transient event. In RAI-3, the staff requested the licensee to provide such information. In its letter of May 27, 2004, PPL noted that Technical Requirement Surveillance (TRS) 3.7.8.5 states, An inspection shall be performed of all snubbers attached to sections of systems that have experienced unexpected potentially damaging transients as determined from a review of operational data and a visual inspection of the systems, and that this surveillance is required to be carried out within 6 months of the transient. The staff found PPLs response adequately addresses the need to perform inspections of snubbers on piping systems that experience damaging transients.

Based on the above discussions, the NRC staff finds that snubber visual examinations (including those of snubber attachments) and functional testing, conducted in accordance with TRM 3.7.8, together with the referenced specification and procedures (1) NEPM-QA-0595, Snubber Program, (2) Surveillance Procedures ST-100-001/ST-200-001, Snubber Visual Inspection, and (3) Specification M-1090, Inservice Testing of Safety-Related Mechanical Snubbers, Hydraulic Snubbers and Compensating Struts, provide visual examination and functional testing requirements equivalent to those of the ASME Code,Section XI and provide reasonable assurance of snubber operational readiness and structural integrity. Therefore, the staff finds PPLs proposed alternative provides an acceptable level of quality and safety. It should be noted that in authorizing Relief Request 3RR-03, TRM 3.7.8 as well as the specifications and procedures cited above become regulatory requirements that may be used in lieu of ASME Code,Section XI requirements for performing ISI of snubbers. Changes to these alternative requirements must be submitted to the NRC staff for authorization pursuant to 10 CFR 50.55a(a)(3) or as an exemption pursuant to 10 CFR 50.12.

On the basis of the information provided by PPL in its submittals dated September 16, 2003,

and May 27, 2004, the NRC staff concludes that PPLs proposed alternative to use the SSES 1 and 2 TRM, together with the associated specification and procedures for snubber surveillance activities, in lieu of the requirements of the ASME Code,Section XI, provides an acceptable level of quality and safety. On this basis, PPLs alternative is authorized for the third 10-year ISI interval for SSES 1 and 2 pursuant to 10 CFR 50.55a(a)(3)(i).

3.2 Request for Relief No. 3RR-07 Code Requirement The 1998 ASME Code 2000 Addenda,Section XI, Table IWC-2500-1, Examination Category C-H requires a VT-2 visual examination to be performed during a system leakage test.

PPLs Basis for Relief (as stated)

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternatives provide an acceptable level of quality and safety.

The Reactor Pressure Vessel Head Flange Leak Detection Line is separated from the reactor pressure boundary by one passive membrane, a silver plated O-ring located on the vessel flange. A second O-ring is located on the opposite side of the tap in the vessel flange (See Figure 3RR-07.1 [of PPLs submittal dated September 16, 2003.]) This line is required during plant operation in order to indicate failure of the inner flange seal O-ring. Failure of the inner 0-ring is the only condition under which this line is pressurized.

The configuration of this system precludes manual testing while the vessel head is removed because the odd configuration of the vessel tap (See Figure 3RR-07.1), combined with the small size of the tap and the high test pressure requirement (1035 psig minimum), prevents the tap in the flange from being temporarily plugged. The opening in the flange is only 3/16 of an inch in diameter and is smooth walled making a high pressure temporary seal very difficult. Failure of this seal could possibly cause ejection of the device used for plugging into the vessel.

A pneumatic test performed with the head installed is precluded due to the configuration of the top head. The top head of the vessel contains two grooves that hold the O-rings. The O-rings are held in place by a series of retainer clips.

The retainer clips are contained in a recessed cavity in the top head (See Figure 3RR-07.1). If a pressure test was performed from the leak-off line side with the head on, the inner O-ring would be pressurized in a direction opposite to what it would see in normal operation. This test pressure would result in a net inward force on the O-ring that would tend to push it into the recessed cavity that houses the retainer clips. The O-ring material is a thin silver plating and could very likely be damaged by this deformation into the recessed areas on the top head. In addition to the problems associated with the O-ring design that preclude this testing, it is also questionable whether a pneumatic test is appropriate for this line. Although the line will initially contain steam if the inner O-ring leaks, the system actually detects leakage rate by measuring the level of condensate in a collection chamber. This would make the system medium water

at the level switch. Finally, the use of a pneumatic test performed at a minimum of 1000 psig would represent an unnecessary risk in safety for the inspectors and test engineers in the unlikely event of a test failure, due to the large amount of stored energy contained in air pressurized to 1000 psig.

System leakage testing of this line is precluded because the line will only be pressurized in the event of a failure of the inner O-ring. It is extremely impractical to purposely fail the inner O-ring in order to perform a test.

Based on the above, SSES requests relief from the ASME [Code,] Section XI requirements for system leakage testing of the Reactor Pressure Vessel Head Flange Seal Leak Detection System.

PPLs Proposed Alternative Examination (as stated)

A VT-2 visual examination will be performed on the line during a refueling outage. The static head developed due to the water above the vessel flange during flood-up will allow for the detection of any gross indications in the line.

This examination will be performed with the frequency specified by Table IWC-2500-1 for a System Leakage Test (once each inspection period).

NRC Staffs Evaluation The ASME Code,Section XI, Table IWC-2500-1, Examination C-H requires a VT-2 visual examination to be performed during a system leakage test. PPL proposed as an alternative to the Code requirement to perform a VT-2 visual examination on the reactor pressure vessel head flange leak detection line when flooding-up the vessel during a refueling outage. The proposed VT-2 visual examination will be performed with the frequency specified by Table IWC-2500-1 for a System Leakage Test once each inspection period of 40 months. The static head developed due to the water above the vessel flange during flood-up will allow for the detection of any gross indications in the line.

The design of the reactor pressure vessel head flange leak detection line precludes PPL from performing the Code required leakage test. The subject line is separated from the reactor pressure boundary by one passive membrane, a silver plated O-ring located on the vessel flange. There is a second O-ring that is located on the opposite side of the tap in the vessel flange. Failure of the inner O-ring is the only condition under which the subject line is pressurized. Therefore, the subject line is required during plant operation in order to indicate failure of the inner flange seal O-ring.

PPL considered performing a manual test while the vessel head is removed; however, due to the configuration of the vessel tap, the small size of the tap, and the high test pressure requirement (1035 psig minimum), the tap in the flange cannot be temporarily plugged. A high pressure temporary seal is very difficult to install, because the flange opening is only 3/16 of an inch in diameter and is smooth walled. If this seal failed it could possibly cause an ejection of the plug and be a hazard to personal performing the test.

PPL has also considered a pneumatic test with the head installed; however, it is not practical due to the configuration of the top head. The top head of the vessel contains two grooves that

hold the O-rings which are held in place by a series of retainer clips. A recessed cavity in the top head is where the retainer clips are contained. With the head on, and if a pressure test was performed from the leak-off line side, the inner O-ring would be pressurized in a direction opposite to what it would see in normal operation. As a result, the net inward force on the O-ring would tend to push it into the recessed cavity that houses the retainer clips. The O-ring material is made of a thin silver plating which could be damaged by this test.

The NRC staff has determined from PPLs Figure 3RR-07.1, Flange Seal Leak Detection Line Detail, and description of the flange seal leak detection line configuration as noted above, the Code required VT-2 visual examination during a system leakage test is impractical. In order to perform the Code required examination, the subject system would have to be redesigned and modified resulting in a significant burden on the licensee.

On the basis of the information provided by PPL in its submittal dated September 16, 2003, the NRC staff concludes that the Code requirements are impractical and the subject system would have to be redesigned in order to perform the Code required examinations, placing a substantial burden on PPL. Furthermore, the NRC staff concludes that the licensees proposed alternative provides reasonable assurance of structural integrity of the subject components.

Therefore, PPLs request for relief is granted pursuant to 10 CFR 50.55a(g)(6)(i) for the third 10-year ISI interval. The NRC staff has determined that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property, or the common defense and security and is otherwise in the public interest giving due consideration to the significant burden upon the licensee that could result if the requirements were imposed on the facility.

3.3 Request for Relief No. 3RR-08 Code Requirement The 1998 ASME Code 2000 Addenda,Section XI, Table IWC-2500-1 requires a VT-2 visual examination to be performed during a system leakage test each 40-month ISI period.

PPLs Basis for Relief (as stated)

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternatives provide an acceptable level of quality and safety.

As required by the SSES Technical Specifications, the CRD Accumulator Pressure must be greater than or equal to 940 psig. Once a week, the accumulator pressure is verified for each accumulator in accordance with SSES Technical Specifications. Additionally, the accumulator pressure is continuously monitored by system instrumentation. Since the accumulators are isolated from the source of makeup nitrogen, continuous monitoring of the CRD Accumulators serves as a pressure decay type test. Should accumulator pressure fall below approximately 980 psig, an alarm is received in the control room. The pressure for the accumulator is recorded and the accumulator is recharged and checked for leaks in accordance with SSES procedures. Should a leak be detected, corrective actions are taken to repair the leak in accordance with SSES procedures.

Since monitoring the nitrogen side of the accumulators is continuous, any leakage from the accumulator would be detected by normal system instrumentation. An additional VT-2 visual examination performed once per inspection period would not provide an increase in safety, system reliability, or structural integrity. In addition, performance of a VT-2 visual examination would require applying a leak detection solution to 185 accumulators per Unit resulting in additional radiation exposure without any added benefit in safety. This inspection would not be consistent with ALARA [as low as is reasonably achievable] practices.

Relief is requested from the VT-2 visual examination requirements specified in Table IWC-2500-1 for the nitrogen side of the CRD Accumulators on the basis that continuous monitoring of the accumulator pressure and a Technical Specification required walkdown of each accumulator exceed the ASME [Code,]

Section XI requirement for a VT-2 visual examination.

PPLs Proposed Alternative Examination (as stated)

As an alternate to the VT-2 visual examination requirements of Table IWC-2500-1, SSES will perform continuous pressure decay monitoring and a weekly Technical Specification required walkdown for the nitrogen side of the CRD accumulators including the attached piping.

NRC Staffs Evaluation The ASME Code,Section XI, Table IWC-2500-1 requires a VT-2 visual examination to be performed during a system leakage test each ISI period (40 months). PPL proposed to perform a continuous pressure decay monitoring and a weekly TS required walkdown for the nitrogen side of the CRD accumulators including the attached piping in lieu of the Code requirements.

In order for the licensee to perform a VT-2 visual examination on the nitrogen side of the CRD accumulators, including the attached piping, they would be required to apply a leak detection solution to 185 accumulators per Unit resulting in additional radiation exposure. As clarified in a phone call with PPL on December 18, 2003, the radiation exposure would be minimal, however, it would be inconsistent with ALARA practices.

PPLs TSs for SSES 1 and 2 requires the plant staff to verify at a frequency of once a week that the pressure is greater than or equal to 940 psig for 185 CRD accumulators for each unit.

Additionally, the accumulators pressure is continuously monitored by system instrumentation.

PPL noted that the accumulators are isolated from the source of makeup nitrogen and continuous monitoring of the CRD accumulators serves as a pressure decay type test. An alarm is activated in the control room if the accumulator pressure falls below approximately 980 psig. If the accumulator pressure falls below approximately 980 psig, the pressure for the accumulator is recorded. The accumulator is then recharged and checked for leaks in accordance with the plants procedures.

The NRC staff has determined that when comparing the Code requirements to PPLs proposed alternative, the subject system will be monitored at a frequency greater than that of the Code of each inspection period of 40 months. Therefore, PPLs proposed alternative to perform

continuous pressure decay monitoring and a weekly TS required walkdown for the nitrogen side of the CRD accumulators, including the attached piping, provides reasonable assurance of quality and safety.

On the basis of the information provided by PPL in its submittal dated September 16, 2003, the NRC staff concludes that PPLs proposed alternative to perform continuous pressure decay monitoring and a weekly TS required walkdown for the nitrogen side of the CRD accumulators, including the attached piping, in lieu of the requirements of the ASME Code,Section XI, provides an acceptable level of quality and safety. Therefore, PPLs alternative is authorized for the third 10-year ISI interval for SSES 1 and 2 pursuant to 10 CFR 50.55a(a)(3)(i).

3.4 Request for Relief No. 3RR-09 Code Requirement The 1998 ASME Code 2000 Addenda,Section XI, IWA-5250(a)(2) stated that if leakage occurs at a bolted connection, one of the bolts shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100. The bolt selected shall be the one closest to the source of leakage. When the removed bolt has evidence of degradation, all remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100.

PPLs Basis for Relief (as stated)

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety.

Removal of pressure retaining bolting at mechanical connections for VT-3 visual examination and subsequent evaluation in locations where leakage has been identified is not always the most prudent course of action to determine condition of the bolting and/or the root cause of the leak. The requirement to remove, examine and evaluate bolting in this situation does not allow consideration of other factors which may indicate the condition of mechanical joint bolting. Other factors which should be considered in an evaluation of bolting condition when leakage has been identified at a mechanical joint include, but should not be limited to:

• Bolting materials

• Corrosiveness of process fluid

• Service age of joint bolting materials

• Leakage location

• Leakage history at connection

• Visual evidence of corrosion at connection (connection assembled)

• Plant/Industry studies of similar bolting materials in a similar environment

• Condition and leakage history of adjacent components An example at SSES is the complete replacement of bolting materials (e.g.,

studs, bolts, nuts, washers, etc.) at mechanical joints during plant outages. In some cases, when the associated system process piping is pressurized during

plant start-up, leakage is identified at these joints. The cause of this leakage is often due to thermal expansion of the piping and bolting materials at the joint and subsequent process fluid seepage at the joint gasket. In most of these cases, proper re-torquing of the joint bolting stops the leakage. Removal of any of the joint bolting to evaluate for corrosion would be unwarranted in this situation.

ASME Code,Section XI Code Interpretation XI-1-92-01 has recognized that this situation exists, and has clarified that the requirements of IWA-5250(a)(2) do not apply.

Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety.

PPLs Proposed Alternative Examination (as stated)

SSES proposes the following alternative, consistent with the methodology of Code Case N-566-2, to the requirements of IWA-5250(a)(2), which will provide an equivalent level of quality and safety when evaluating leakage and bolting material conditions at Class 1, 2, and 3 bolted connections.

As an alternative to the [. . .] requirements of Subparagraph IWA-5250(a)(2), one of the following requirements will be met for leakage at bolted connections:

(a) The leakage will be stopped, and the bolting and component material will be reviewed for joint integrity as described in (c) below.

(b) If the leakage is not stopped, the SSES will evaluate the structural integrity and consequences of continuing operation, and the effect on the system operability of continued leakage. This engineering evaluation will include the considerations listed in (c) below.

(c) The evaluation of (a) and (b) above is to determine the susceptibility of the bolting to corrosion and failure. This evaluation will include the following:

(1) the number and service age of the bolts; (2) bolt and component material; (3) corrosiveness of process fluid; (4) leak location and system function; (5) leakage history at the connection or other system components; (6) visual evidence of corrosion at the assembled connection.

If any of the above parameters indicates a need for further examination, the corrective action will be taken in accordance with IWA-5250(a)(2).

NRC Staffs Evaluation The 1998 ASME Code 2000 Addenda,Section XI, IWA-5250(a)(2) requires that if leakage

occurs at a bolted connection, one of the bolts shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100. The bolt selected shall be the one closest to the source of leakage. When the removed bolt has evidence of degradation, all remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100. The Code requirements provide assurance that bolting corroded by system leakage will be detected and that corrective actions will be taken. However, the Code requirements are often unnecessarily conservative since corrosion is dependent on other factors beyond system leakage. Additionally, removal and examination of all bolts may not be necessary to assure continued integrity of the bolted connection.

In lieu of these requirements, PPL has proposed to implement an alternative which requires, in part, an engineering evaluation to determine the need for additional examinations of the bolts considering the elements listed in the licensees proposed alternative.

PPL noted that when an evaluation of the proposed elements is concluded and the evaluation determines that the leaking condition has not degraded the fasteners, then no further action is necessary. In addition, if the leakage is not stopped, PPL will evaluate the structural integrity and consequences of continuing operation, and the effect on the system operability of continued leakage. The engineering evaluation will include six attributes as noted in the licensees proposed alternative and are as follows:

(4) the number and service age of the bolts; (5) bolt and component material; (6) corrosiveness of process fluid; (7) leak location and system function; (8) leakage history at the connection or other system components; (9) visual evidence of corrosion at the assembled connection.

If any of the above parameters indicates a need for further examination, corrective action will be taken in accordance with IWA-5250(a)(2) to remove one of the bolts and preform a visual VT-3 examination and evaluate the bolt in accordance with IWA-3100. IWA-3100 requires that the evaluation of flaws are in accordance with IWB-3000, IWC-3000, and IWD-3000 Acceptance Standards for Class 1, 2, and 3 pressure retaining components, respectively.

The NRC staff finds PPLs proposed alternative as stated above an acceptable alternative to the Code requirements. Based on the items included in the evaluation process, the NRC staff concludes that the evaluation proposed by PPL presents a sound engineering approach and provides reasonable assurance of quality and safety. In addition, if the initial evaluation indicates the need for a more detailed analysis, the bolt closest to the source of leakage will be removed, VT-3 visually examined, and evaluated in accordance with IWA-3100(a) as PPL stated in its September 16, 2003, submittal. Therefore, PPLs alternative is authorized for the third 10-year ISI interval for SSES 1 and 2 pursuant to 10 CFR 50.55a(a)(3)(i).

All other requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

3.5 Request for Relief No. 3RR-10 Code Requirement Paragraph IWA-2432 requires that each inspection interval consist of 10-year duration; except as modified by IWA-2430(d) which permits the inspection interval to be extended or decreased by as much as 1 year, provided that successive intervals are not altered by more than 1 year from the original pattern of intervals.

PPLs Basis for Relief PPL is requesting relief from the 10-year containment ISI interval required by IWA-2430 of the ASME Code, pursuant to 10 CFR 50.55a(a)(3)(i). PPL asserts that the proposed alternative will provide an acceptable level of quality and safety.

Specifically, PPL is seeking this relief, so that it can reduce the duration of the first containment ISI interval that will permit synchronization of the subsequent containment ISI interval dates with future ISI intervals for the Class 1, 2, and 3 components.

PPL provided the following background information in support of the relief request:

Containment ISI Programs were initially required by regulation (10 CFR 50.55a) as amended within a Final Rule (61FR41303) issued on August 8, 1996.

Accordingly, the SSES 1 and 2 containment ISI Program was prepared and has been implemented to the 1992 Edition through the 1992 Addenda of Subsections IWE and IWL of ASME Section XI, "Rules for Inservice Inspection of Nuclear Plant Components," Division 1 as modified by the regulation at that time.

All examinations and tests required by the containment ISI Program have been implemented in accordance with the established schedule. For the IWE portion of the program, all the examinations scheduled for the first and second inspection periods have been completed. Additionally, the first five-year examination has been completed for the IWL portion of the program. All of these examinations and tests performed to date have satisfied the acceptance standards contained within Articles IWE-3000 and IWL-3000, without exception.

Currently, the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b)(2) of the regulation is the 1998 Edition through the 2000 Addenda. PPL is planning to use this Code version, as required by 10 CFR 50.55a(g)(4)(ii), for development of the SSES 1 and 2 ISI Program update for the third inspection interval of ISI Class 1, 2, and 3 components and their component supports, subject to the limitations and modifications within the regulation. The 1998 Edition through the 2000 Addenda of Section XI will also be used for development and implementation of the second containment ISI Interval, subject to the limitations and modifications within 10 CFR 50.55a(b)(2) of the regulation.

PPL further explains that the current containment ISI Program requires performance of several VT-3 visual examinations in accordance with Table IWE-2500-1, Examination Category E-A, Items E1.12 and E1.20 at the end of the first inspection interval. Accordingly, these examinations have not yet been performed. Table IWE-2500-1, Examination Category E-A,

Items E1.12 and E1.20 of the 1998 Edition through the 2000 Addenda of Section XI do not limit performance of these examinations to the end of the interval. Therefore, PPL has scheduled performance of these examinations for the first inspection period in the 2nd containment ISI interval. Scheduling these examinations for this inspection period will retain the originally scheduled point in time, thus meeting the intent for performance of these examinations.

PPLs Proposed Alternative As an alternative to the full 10-year interval duration requirement for the first containment ISI interval, PPL proposes to reduce the length of the first containment ISI interval by approximately 3 years. This will permit the subsequent containment ISI interval to coincide with the dates for the third and subsequent inspection intervals for the ISI Program for Class 1, 2, and 3 components. Therefore, the start date of the 3rd interval for the ISI Program and the 2nd interval for the containment ISI Program will be effective on June 1, 2004, for SSES 1 and 2.

NRC Staffs Evaluation In the supplementary information contained within Section 2.2 of the Final Rule (67 FR 60520) dated September 26, 2002, the NRC staff stated that 10 CFR 50.55a(g)(4)(ii) does not prohibit licensees from updating to a later edition and addenda of the ASME Code midway through a 10-year IWE or 5-year IWL examination interval. This facilitates licensees in synchronizing their ISI intervals for the containment ISI with that of Class 1, 2, and 3 components.

Additionally, the staff advised that licensees wishing to synchronize their 120-month intervals may submit a request in accordance with 10 CFR 50.55a(a)(3) to obtain authorization to extend or reduce 120-month intervals.

In the last paragraph of the Basis for Relief, PPL states, scheduling these examinations for this inspection period will retain the originally scheduled point in time, thus meeting the intent for performance of these examinations. The staff finds PPLs approach in synchronizing the containment ISI with that of the Class 1, 2, and 3 components acceptable. Based on this assessment, the staff concludes that the proposed request for permitting the shorter (than 10 years) interval for ISIs of metallic containment and concrete containment components in the 2nd 10-year containment ISI interval will provide acceptable level of quality and safety.

3.6 Request for Relief No. 3RR-11 Code Requirement Subsubarticle IWE-5240 requires performance of a visual examination using the Detailed Visual examination method following repair/replacement (R/R) activities.

PPLs Basis for Relief PPL states that the requirement to perform a detailed visual examination following an R/R activity essentially conflicts with the regulation for performance based on local leakage rate testing as found within Option B of Appendix J to 10 CFR Part 50, and the associated documents, Nuclear Energy Institute 94-01 and Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Testing Program, for specific conditions where the pressure test is either deferred or not required. Moreover, PPL suggests that in order to reduce

unnecessary burden of performing the detailed visual examinations when a pressure test is performed following an R/R activity, ASME Code Case N-649 permits the performance of a general visual examination of the affected pressure boundary either during or following the pressure test to ensure overall integrity of the R/R component within the containment.

PPL asserts that the detailed visual examination will still be performed for those R/R activities affecting the containment pressure boundary where a pressure test is deferred. When the deferred pressure test is conducted, an additional general visual examination shall be performed of the affected pressure boundary either during or following the deferred pressure test to ensure overall integrity of the R/R component within the containment.

Based on the above rationale, PPL concludes that with the use of the alternative requirements of Code Case N-649, there is reasonable assurance that the structural integrity of the containment pressure retaining components will be assured, and an acceptable level of quality and safety will be maintained during the third 10-year inspection interval.

PPLs Proposed Alternate Examinations PPL requests the use of the alternative examination requirements of Code Case N-649 titled "Alternative Requirements for IWE-5240 Visual Examination."

NRC Staff Evaluation

The NRC staff finds that the IWE-5240 (1998 Edition through the 2000 Addenda) requirement for detailed visual examination after containment R/R activities could be graded to satisfy the following conditions: (1) when the general visual examination indicates flaws or defects, or (2) when the required pressure test is deferred. Though ASME Code Case N-649 has not been accepted in NRC RG 1.147, Inservice Inspection Code Acceptability, ASME Section XI, Division 1, Revision 13, the graded process identified in the code case is acceptable. The staff finds that the use of the code case after R/R activities will provide an acceptable level of quality and safety, and as pointed out in the RR, will reduce unnecessary burden on the licensee.

Therefore, pursuant to 10 CFR 50.55a(a)(3), the proposed request for reducing the length of the first containment ISI interval by approximately 3 years is acceptable.

The following clarifications are implicit in the staffs approval of 3RR-11:

(1) The use of the 1998 Edition through the 2000 Addenda for the containment ISI shall be incorporated by the reference in 10 CFR 50.55a (67FR60520) (the rule). Specifically, the terms general visual examination, and detailed visual examination shall be construed as VT-3 and VT-1 examinations respectively, as defined in the rule.

(2) When Code Case N649 is accepted in RG 1.147, any modification or limitation delineated in RG 1.147 shall apply.

4.0 CONCLUSION

The NRC staff concludes that PPLs proposed alternatives for 3RR-03, 3RR-08, 3RR-09, 3RR-10, and 3RR-11 provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes the proposed alternatives for 3RR-03,

3RR-08, 3RR-09, 3RR-10, and 3RR-11 as described in PPLs letter dated September 16, 2003, for SSES 1 and 2 for the third 10-year ISI interval. In addition, the NRC staff has determined that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) for 3RR-07 is authorized by law and will not endanger life or property, or the common defense and security and is otherwise in the public interest giving due consideration to the significant burden upon the licensee that could result if the requirements were imposed on the facility. The NRC staff concludes that PPLs proposed alternatives for 3RR-07 provides reasonable assurance of structural integrity of the subject components; therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the NRC staff also authorizes the proposed alternatives for 3RR-07 for the third 10-year ISI interval.

All other ASME Code,Section XI, requirements for which relief has not been specifically requested remain applicable, including third party review by the authorized nuclear inservice inspector.

Principal Contributors: A. Lee T. McLellan H. Ashar Date: September 24, 2004