RC-03-0176, Revision to License Amendment Request - LAR 02-2767 Steam Generators - One Time Exclusion on Inspection Frequency

Revision to License Amendment Request - LAR 02-2767 Steam Generators - One Time Exclusion on Inspection Frequency

ML032380053
Person / Time
Site:	Summer
Issue date:	08/20/2003
From:	Byrne S South Carolina Electric & Gas Co
To:	Cotton K Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR 02-02767, RC-03-0176
Download: ML032380053 (29)

Text

Stephen A. Byrne Senior Vice President, Nuclear Operations 803.345.4622 August 20, 2003 A SCANA COMPANY RC-03-0176 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: K. R. Cotton

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12 REVISION TO LICENSE AMENDMENT REQUEST - LAR 02-2767 STEAM GENERATORS - ONE-TIME EXCLUSION OF INSPECTION FREQUENCY

Reference:

SCE&G letter to NRC (Document Control Desk), RC-03-0007, LICENSE AMENDMENT REQUEST LAR 02-2767, STEAM GENERATORS - ONE TIME EXCLUSION OF INSPECTION FREQUENCY, January 14,2003 Pursuant to 10 CFR 50.90, South Carolina Electric & Gas Company (SCE&G), acting for itself and as agent for South Carolina Public Service Authority, hereby requests an amendment to the VCSNS Technical Specifications to incorporate the following changes:

Surveillance Requirement 4.4.5.3.a, Steam Generator Surveillance Requirements" The proposed one-time change revises the steam generator inservice inspection frequency requirements in TS 4.4.5.3.a for V. C. Summer Nuclear Station (VCSNS) immediately after refueling outage RF-12. The change would allow a 58-month maximum inspection interval after two inspections resulting in C-1 classification, rather than a 40-month maximum inspection interval. This change is proposed to eliminate premature/unnecessary steam generator inspections, due to a shortened operating cycle. This change will also result in significant dose, schedule, and cost savings.

Limiting Condition for Operation 3.4.6.2.c, Reactor Coolant System, Operational Leakage' The proposed revision is to conservatively change the primary-to-secondary leakage limit for each steam generator from 500 gallons per day to 150 gallons per day as prescribed in the VCSNS Steam Generator Management Program.

Bases 3/4.4.5, "Steam Generators" The proposed change is to revise the discussion on the primary-to-secondary leakage limits assessed for operation of VCSNS.

Bases 3/4.4.6.2, "Operational Leakage". The proposed revision is to revise the discussion on the primary-to-secondary leakage limit for each steam generator to address 150 gallons per day as prescribed in the VCSNS Steam Generator Management Program.

SCE&G I firgil CSummer Nuclear Satidon .O.Box 88. Jenkinsille, Soth Carolina 29065 .T (803) 345.5209 www.scano.com

Document Control Desk LAR 02-2767 RC-03-0176 Page 2 of 3 The VCSNS steam generators (SGs) were replaced during RF-8, Fall 1994. The replacement SGs are the Westinghouse Delta 75 design, which incorporates significant improvements, including thermally treated Alloy 690 tubing. Inspections performed pursuant to this surveillance requirement in RF-9 and RF-10 both resulted in C-1 classification. The latest 100% bobbin inspection of all three VCSNS SGs, performed in RF-12, found no indications of stress corrosion cracking or any other active damage mechanism.

During the VCSNS RF-12, October 2000, following the fourth cycle of operation after replacement, 100% of the SG tubes were inspected full-length (i.e., from hot leg tube end to cold leg tube end, including the U-bends) with eddy current bobbin coil. Additionally, a sample of 322 tubes (10%) in SG B were inspected at the hot leg top of tubesheet area (+3/-3 inches),

14 low row u-bends in SG C with a rotating coil probe, and approximately 65 areas of special interest with the rotating coil probe. No defective or degraded tubes were indicated.

Because of the identification of a crack in Loop "Am Reactor Coolant System (RCS) hot leg pipe, RF-12 was extended approximately four months for repairs. The following operating period for Cycle 13 was reduced to 13 months, with RF-13 commencing in April 2002. Cycle 14 operating period is anticipated to be 16 months, from June 4, 2002 to October 11, 2003, with RF-14 commencing on October 11, 2003. At that time, 36 calendar months will have elapsed since the previous inspection, but only 29 Effective Full Power Months (EFPM). An inspection performed at that time would be premature, based on the excellent condition of the SGs.

These inspection results, along with the shortened Cycle 13, improved SG design, and industry experience with thermally treated Alloy 690 tubing, provide the basis for proposing this one-time extension of the maximum inspection interval.

In addition to the one-time extension of the maximum inspection interval, this request includes a permanent change to the Limiting Condition for Operation 3.4.6.2.c to reduce the primary-to-secondary leakage limit from each SG to 150 gallons per day.

SCE&G is aware of ongoing industry and NRC work regarding a generic license change (GLCP) to this section of Technical specifications. This request complies with proposed Technical Specifications changes to date, and with lengthened inspection intervals proposed in EPRI Steam Generator Inspection Guidelines, Rev. 6. We believe that our site-specific thermally treated Alloy 690 experience, and overall industry experience with thermally treated Alloy 690 tubes provides an acceptable approval basis for our one-time request. to this letter provides the No Significant Hazards Determination and Attachment I provides the TS and Bases pages marked up with the proposed change. Attachment II provides the retyped TS and Bases pages.

SCE&G License Amendment Request 03-1864, submitted on July 29, 2003, by letter RC 0159 has a proposed change to Bases page B 3/4 4-5 that has an impact to the changes proposed to B 3/4 4-5 by this request.

FSAR Section 5.2.7.5.3 is affected by this change and will be updated upon approval of this request.

Document Control Desk LAR 02-2767 RC-03-01 76 Page 3 of 3 The VCSNS Plant Safety Review Committee and the Nuclear Safety Review Committee have reviewed and approved the proposed change. SCE&G has notified the State of South Carolina in accordance with 10CFR50.91 (b).

SCE&G requests approval of the proposed change as soon as possible to support postponing SG tube inspections currently planned for RF-14, which is scheduled to begin on October 11, 2003.

If you have any questions or require additional information regarding this proposed amendment, please contact Mr. Ron Clary at (803) 345-4757.

I certify under penalty of perjury that the foregoing is true and correct.

Eft?/o3n Executed on Stephen A. Byrne Senior Vice President, Nuclear Operations JT/SAB/dr

Enclosures:

Evaluation of the proposed change Attachment(s): 3

1. Proposed Technical Specification Change - Mark-up

2. Proposed Technical Specification Change - Retyped

3. List of Regulatory Commitments c: N. 0. Lorick N. S. Carns T. G. Eppink (w/o Attachments)

R. J. White L. A. Reyes K. R. Cotton NRC Resident Inspector P. Ledbetter K. M. Sutton T. P. O'Kelley RTS (LAR 02-2767)

File (813.20)

DMS (RC-03-0176)

Document Control Desk Enclosure I LAR 02-2767 RC-03-0176 Page 1 of 14

Subject:

LICENSE AMENDMENT REQUEST - LAR 02-2767 STEAM GENERATORS - ONE-TIME EXCLUSION OF INSPECTION FREQUENCY

1.0 DESCRIPTION

South Carolina Electric & Gas Company (SCE&G) proposes an amendment to revise the V. C.

Summer Nuclear Station (VCSNS) Technical Specifications (TS) and Bases. Specifically, Surveillance Requirements (SR) 4.4.5.3.a., Limiting Condition for Operation 3.4.6.2.c., Bases 3/4.4.5, and Bases 3/4.4.6.2 are addressed by this request.

NEI 97-06 and the associated EPRI guidelines are the documents that provide consistency in managing steam generator (SG) programs. These guidelines have been developed through NEI and EPRI to provide methods to ensure accurate assessment of SG tube integrity, and to extend the reliable, cost effective, operating life of the SGs. To this end, VCSNS has committed to the above as part of an industry wide effort of SG maintenance.

Inspections of the SG tubing ensure that the structural integrity of this portion of the reactor coolant system is maintained. In-service inspections of SG tubes are essential in order to maintain surveillance of tube condition in the event there is mechanical damage or progressive damage due to design or manufacturing errors or in-service conditions such as wear or corrosion. In-service inspection of the tubes also provides the means to characterize the condition such that corrective action can be taken in the case of identified tube degradation.

VCSNS current Technical Specifications are more rigorous for in-service inspection frequency requirements of the SGs than the NEI guidance. Based on the superior performance of the SGs to date and the following discussions, VCSNS requests a one-time amendment to Technical Specification 4.4.5.3.a to allow a 58-month maximum inspection interval after two inspections resulting in a C-1 classification, rather than a 40-month inspection interval. This change is proposed to eliminate premature/unnecessary SG inspections due to a shortened operating cycle, which will result in significant dose, schedule, and cost savings.

Limiting Condition for Operation 3.4.6.2.c and Bases 3/4.4.6.2 are also being permanently revised to reflect the more conservative daily leakage limits imposed on the SGs by NEI 97-06 and the VCSNS Steam Generator Performance Program.

2.0 PROPOSED CHANGE

Specifically the proposed changes would revise the following:

Currently, TS 4.4.5.3.a states, in part:

If two consecutive inspections, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;

Document Control Desk Enclosure I LAR 02-2767 RC-03-0176 Page 2 of 14 The proposed change is to add a note following the paragraph cited above:

Note: A one-time inspection interval of once per 58 months is allowed for the inspection performed immediately following refueling outage RF-12.

Limiting Condition for Operation 3.4.6.2.c, The proposed revision is to conservatively change the primary-to-secondary leakage limit for each SG from 500 gallons per day to 150 gallons per day as prescribed in the VCSNS Steam Generator Management Program.

Bases 3/4.4.5 The proposed change is to revise the discussion on the primary-to-secondary leakage limits assessed for operation of VCSNS.

Bases 3/4.4.6.2, The proposed revision is to revise the discussion on the primary-to-secondary leakage limit for each SG to address 150 gallons per day as prescribed in the VCSNS Steam Generator Management Program.

3.0 BACKGROUND

VCSNS replaced three Westinghouse model D3 SGs with replacement model Delta 75 SGs in the fall of 1994. One pre-service examination and 4 inservice examinations, which are described in this submittal, have been performed since replacement and have confirmed that no degradation of tubing has occurred. Degradation assessments have further concluded that no active degradation mechanisms are present.

VCSNS has performed inspections on the replacement SGs in excess of the scope required by the referenced Technical Specification surveillance requirement 4.4.5.3.a. The VCSNS Steam Generator Management Program also complies with industry requirements outlined in NEI 97-06 and supporting EPRI guidelines.

4.0 TECHNICAL ANALYSIS

4.1 Steam Generator Inspection History The VCSNS SGs were replaced during Refuel 8 (fall 1994). The three SGs are the Westinghouse Delta 75 design. Each SG contains 6307 thermally treated Alloy 690 tubes with an outside diameter of 0.6875 inches and a nominal wall thickness of 0.040 inches.

Inspections were performed pursuant to the technical specifications surveillance requirements in the four consecutive outages following replacement, Refuel 9, 10, 11, and 12, and found no

Document Control Desk Enclosure I LAR 02-2767 RC-03-0176 Page 3 of 14 evidence of any degradation. Each resulted in Technical Specifications category C-1 classification. The latest inspection of all three SGs, performed in Refuel 12 (fall 2000), found no indications of stress corrosion cracking or any other damage mechanism. The scope of this latest inspection is described below.

One hundred percent of all SG tubes were full length eddy current inspected using bobbin coil in Refuel 12. In detail, SG A was examined full length with a bobbin probe; SG B was examined full length bobbin with 322 tubes examined with a rotating coil probe hot leg top of tubesheet; SG C was examined full length with a bobbin probe and a U-bend study was completed on 14 SG C Row 1 tubes comparing a mid frequency to a high frequency rotating coil probe with calibration and subsequent validation of the VCSNS site technique. These inspections were performed in accordance with guidance specified by Appendix H of EPRI SG Examination Guidelines.

Results for Refuel 12 eddy current after 5.4 effective full power years were that no indications of stress corrosion cracking were present. This data supports the laboratory test results of Alloy 690 TT tubing which indicates superior performance over previous mill annealed Inconel 600 tubing. A full description of the Refuel 12 inspection results was reported to the NRC pursuant to Technical Specifications Section 4.4.5.5.

Because of the identification of a crack in A (Alpha) reactor coolant system hot leg piping, Refuel 12 was extended approximately 4 months. This resulted in Cycle 13 being reduced to 13 operating months, with Refuel 13 commencing on April 20, 2002. No SG inspections were performed in Refuel 13. Cycle 14 is expected to be 16 operating months having commenced on June 4, 2002 and expecting completion on October 11, 2003. At that time, 36 calendar months will have elapsed since the previous inspection, but only 29 effective full power months (EFPM) would be realized. A VCSNS SG inspection performed at that time would be premature based on the excellent condition of the SGs. Cycle 15 is conservatively estimated to be 17 operating months following a 1-month refuel outage. Extending the inspection interval would allow VCSNS to operate an additional 17 EFPM to the spring 2005 refueling, for a total of 46 EFPM, or 54 calendar months. This extension of the Technical Specifications limit is supported as described below.

4.2 Steam Generator Design Improvements The Delta 75 SG design employs the key characteristics of the proven Westinghouse Model F design. The most reliable features carried over from Model F include the use of previously mentioned Alloy 690 thermally treated tube material, Type 405 stainless steel tube support plate material, trifoil-shaped broached flat contact tube support holes, hydraulically expanded tube sheet joints, minimum gap U-bend construction, and foreign material exclusion in the design of the feedwater distribution headers. In addition, stress relief was required for all tubes in the 17 innermost rows.

The corrosion resistance of thermally treated Alloy 690 has not only been proven by years of extensive laboratory testing, but has been proven in actual plant operation at plants of the same SG replacement vintage as VCSNS such as D.C. Cook 2 and Indian Point 3. To date, none of these units have reported any corrosion degradation. Also, since test conditions have been more hostile than actual plant service conditions, expectations for Alloy 690 tubing in the field

Document Control Desk Enclosure I LAR 02-2767 RC-03-01 76 Page4of 14 should result in improved performance. VCSNS will continue to track the evolution of Alloy 690 tubing for degradation concerns through association with EPRI tools such as the SG Degradation Database and related communication. Industry experience with Alloy 690 will be factored into VCSNS SG inspection planning and integrity assessments.

Current industry data shows that some plants with replacement SGs have experienced mechanical degradation (wear) at tube support plates, anti-vibration bars, and wear related to loose parts. Wear is the volumetric removal of material caused by the mechanical action of one material in contact with another and is associated with secondary side flow that induces vibration and subsequent contact between the objects.

The potential for loose parts entering the SGs is minimized by the design of the feedwater ring spray nozzle assemblies, each consisting of a series of 0.25 inch diameter outlet holes, which function to trap potential foreign objects that may be introduced from feedwater systems. The 0.25 inch dimension is smaller than the tube pitch, thus, if parts are small enough to pass through these holes, they will also pass between the tubes and be transported into the low velocity regions of the bundle. There they will have the least potential to cause service induced wear. One possible loose part was reported in Refuel 11, in SG at the top of the tube sheet. The presence of the item was evaluated, and the item was left in place until the next scheduled shutdown. During Refuel 12 the loose part was identified (a piece of wire 0.5 inches long) and removed with no indications present on any of the surrounding tubes after the part was removed. No evidence of additional loose parts was identified in the tube bundle through FOSAR (foreign object research and retrieval) inspections and eddy current, and no indication of loose part damage to SG tubing was observed. As far as minimizing loose parts during maintenance, VCSNS is committed to a strict FME (foreign material exclusion) program when any SG closure is opened to the environment.

Eddy current examinations thus far have indicated that wear is not present in the VCSNS SGs.

Periodic future inspections will continue to pay close attention to this degradation mode, since it is the only potential degradation mode currently identified for VCSNS.

The operational assessment performed following Refuel 12 provides additional support for the VCSNS extension request. In this assessment, it was assumed that the largest through wall indication (9%) found in Refuel 12, was 0% at baseline inspection. (Actual analysis confirmed that this indication was, in fact, present at baseline and had not grown). The calculated growth rate was 1.74% TW/EFPY. At this growth rate, the largest indication left in service would not reach structurally significant size (57.27% TW) until 27.7 EFPY, or approximately 18.5 cycles from Refuel 12, assuming 18 month cycles.

VCSNS implements an SG program that complies with the current revision to NEI 97-06 and supporting EPRI -Steam Generator Guidelines. The current revision to the EPRI Steam Generator Examination Guideline, Revision 6, issued in October 2002, allows Alloy 690 replacement SGs to be inspected at an initial interval (following the first in-service examination) of 144 EFPM, with a 50% inspection in all SGs required at or before 72 EFPM. This request for extension of the surveillance interval (from 29 to 46 EFPM) falls well within this guideline. Also, this extension is preceded by 3 partial and 1 full inspection in 4 consecutive cycles. For information, the previous Revision 5 to the EPRI Steam Generator Guideline allowed a 60 EFPM rolling time frame to achieve 100% tube inspection of each type of repair, regardless of

Document Control Desk Enclosure I LAR 02-2767 RC-03-01 76 Page 5 of 14 material type. Revision 5, however, provided an additional limitation that no SG should operate more than two cycles between inspections.

There is also an ongoing industry effort through NEI to revise this portion of Technical Specifications via a generic license change (GLCP). As stated earlier, VCSNS is requesting an extension of the more restrictive Technical Specification frequency of SG inspection criteria.

While this is a relaxation of certain current requirements, the commitment to other performance criteria such as more restrictive primary to secondary leakage and rigorous condition monitoring and operational assessments that are currently not required by Technical Specifications provide added support for this request. Although the entire GLCP effort is not complete, the proposed change also falls within the criteria addressed in the GLCP.

For additional assurance that appropriate actions will be taken in response to a primary to secondary leak, VCSNS is including, in this amendment, a permanent reduction in Technical Specification LCO 3.4.6.2.c, primary to secondary leakage limit from each SG. This reduction from 500 gallons per day to 150 gallons per day Is in compliance with the proposed GLCP and industry guidelines.

In summary, the following bases provide support for assuring that VCSNS SG performance criteria will be maintained through the extended operating period.

• Improved quality of materials.

• Improved design features.

• Results of examinations performed to date.

• Reduced operating time due to extended outages.

• Operational Assessment analysis.

• VCSNS Steam Generator Management Program controls.

• Changes in industry examination guidelines for Alloy 690 replacement SGs.

• Reduction in Technical Specification 3.4.6.2.c primary to secondary leakage limit from each SG.

4.3 Steam Generator Performance

Background

The SG tubes in pressurized water reactors have a number of important safety functions. SG tubes are an integral part of the reactor coolant pressure boundary and as such, are relied upon to maintain the primary system pressure and inventory. As part of the RCS boundary, the tubes are unique in that they are also relied upon as a heat transfer medium between the primary and secondary systems such that heat may be removed from the primary system. Additionally, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system.

Document Control Desk Enclosure I LAR 02-2767 RC-03-01 76 Page 6 of 14 SG tube integrity is necessary in order to satisfy the above safety functions. Tube degradation related to corrosion, denting, and wear can impair tube integrity if they are not managed effectively. Maintaining this integrity ensures that the tubes are capable of performing their safety function consistent with the VCSNS licensing basis.

Freguency of Verification of Tube Intecritv The existing Technical Specification base inspection intervals classifies the condition of the SG into one of three categories based on the overall results of the previous inspection. The surveillance frequency described by the current VCSNS Steam Generator Management Program is performance based and is dependent on the location and severity of active degradation mechanisms and their anticipated growth rate.

Tube Samjle Selection The existing Technical Specification bases tube selection on SG conditions and industry and plant experience. This change to a performance-based program as defined in the VCSNS Steam Generator Management Program requires that a degradation assessment be performed before inspections and references EPRI Guidelines for details on tube selection. This approach is dependent on actual SG conditions and operational experience. This method is more rigorous than the existing Technical Specification.

Inspection Techniques The VCSNS Steam Generator Management Program requires the performance of a degradation assessment prior to SG inspections to identify active and potential degradation modes and mechanisms and NDE techniques that are effective in detecting their existence.

Tube inspection techniques are chosen to reliably detect flaws that may progress during an operating cycle beyond performance criteria limits. Identification of damage not predicted in the Degradation Assessment requires revision to the Degradation Assessment and initiation of corrective actions to address the identified condition.

Performance Criteria The performance criteria used for VCSNS SGs are based on structural integrity, accident induced leakage, and operational leakage. The structural and accident induced leakage criteria were developed deterministically and are consistent with the VCSNS licensing basis. The operational leakage criterion was based on providing added assurance against tube rupture at normal operating and accident conditions.

The Structural Integrity performance criterion is:

SG tubing shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation and a safety factor of 1.4 under all design basis

Document Control Desk Enclosure I LAR 02-2767 RC-03-0176 Page 7 of 14 accidents, including any additional loading combinations required by existing design and licensing basis.

The structural integrity performance criterion is based on ensuring that there is reasonable assurance that a SG tube will not burst during normal or postulated accident conditions.

The Accident-induced Leakage performance criterion is:

The primary to secondary accident induced leakage rate for the limiting design basis accident, other than .a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG, except for specific types of degradation at specific locations where the NRC has approved greater accident-induced leakage as part of the VCSNS licensing basis. Exception to the 1-gpm limit can be applied if approved by the NRC in conjunction with approved Alternate Repair Criteria.

Primary to secondary leakage is a factor in the dose releases outside containment resulting from a limiting design basis accident. The potential primary to secondary leak rate during postulated design basis accidents must not exceed the offsite radiological dose consequences required by 10 CFR Part 100 guidelines or the radiological consequences to control room personnel required by GDC-19.

The Operational Leakage performance criterion is:

The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day.

Plant shutdown should commence if primary to secondary leakage exceeds 150 gallons per day at room temperature conditions from any one SG.

The above Performance Criterion are controlled through the VCSNS Steam Generator Management Program. In addition, the operational leakage performance criteria is being incorporated into Technical Specification LCO 3.4.6.2.c by this amendment. Any future change in any of the performance criteria or Technical Specifications will require prior NRC approval.

Repair Criteria VCSNS SG tubes will be repaired as follows:

• Tubes that do not meet the performance criteria will be plugged.

• Tubes that are pressure tested or the location of tubes removed for metallurgical examination will be plugged.

• Tubes identified with a damage form or mechanism for which no depth sizing capability exists shall be plugged. In certain cases (e.g., loose parts wear), technical justification may be developed to justify leaving some small indications in place.

Document Control Desk Enclosure I LAR 02-2767 RC-03-0176 Page 8 of 14

• Tubes with defects (indications of degradation which exceed the repair limits) as defined in the Degradation Assessment shall be plugged.

• New repair methods shall be reviewed and approved by the NRC.

Exceeding Performance or Repair Criteria VCSNS will monitor SG performance as described in the VCSNS Steam Generator Management Program. This includes meeting the performance criteria during all MODES of plant operation as well as during SG inspections and subsequent assessments.

Failure to meet a Performance Criterion or Failure to implement a plugging or repair will require operational assessments and NRC reporting as described in the VCSNS Steam Generator Management Program. VCSNS will follow NEI 97-06 Guidelines for reporting requirements, which is more rigorous than the existing Technical Specification reporting requirement.

Failure to meet the primary to secondary leakage limit of 150 gallons per day from each SG will require shutdown action in accordance with revised Technical Specification LCO 3.4.6.2.c.

4.4 Conclusion This Technical Specification change does not affect the design of the SGs, their method of operation, or primary coolant chemistry control. The change does not adversely impact any previously evaluated design basis accident.

The maximum allowed SG tube leakage of 450 GPD (3 SGs with 150 GPD each) for all SGs not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a SG tube rupture or steam line break. The 150 GPD per SG limit preserves the assumptions used in the analysis of these accidents and ensures that SG tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

For accidents that do not involve fuel damage, the reactor coolant specific activity levels are assumed at the Technical Specification values. For accidents that do involve fuel damage, the primary coolant specific activity values are a function of the accident conditions. None of the accident assumptions are affected by this change.

Based on the above evaluation, the proposed change does not affect the consequences of a SGTR or any other design basis accident.

Document Control Desk Enclosure I LAR 02-2767 RC-03-01 76 Page 9 of 14 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration South Carolina Electric & Gas Company (SCE&G) has evaluated the proposed changes to the VCSNS TS described above against the significant Hazards Criteria of 10CFR50.92 and has determined that the change does not involve any significant hazard. The proposed change is part of a SG program that defines a performance-based approach to maintaining SG tube integrity. The VCSNS Steam Generator Management Program includes performance criteria that define the basis for SG operability and provide reasonable assurance that the SG will remain capable of fulfilling its safety function of maintaining reactor coolant pressure boundary integrity. Also, the operational leakage performance criteria is incorporated into the Technical Specifications by this amendment. The existence of the VCSNS Steam Generator Management Program, and inclusion of the operational leakage performance criteria in Technical Specifications is more rigorous than existing Technical Specifications. The following is provided in support of this conclusion.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed one-time extension of the Technical Specification management interval and reduction of operational leakage limits does not involve changing any structure, system or component or affect plant operations. It is not an initiator of any accident and does not change any FSAR safety analyses. As such, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

Probability of an Accident The VCSNS Steam Generator Management Program includes provisions that are more rigorous than existing Technical Specification requirements. The topics addressed by the program include:

• SG performance criteria, including a reduced operational leakage limit.

• SG repair criteria and repair methods.

• SG inspections that include Degradation Assessments, Condition Monitoring Assessments, and Operational Assessments.

• NDE technique requirements.

The results of the above program requirements demonstrated that all performance requirements were met during Refuel 12.

Document Control Desk Enclosure I LAR 02-2767 RC-03-0176 Page 10 of 14 The inclusion of reduced SG operational leakage limits will provide more conservative limits for operating VCSNS under primary to secondary leakage conditions.

Consequences of an Accident The consequences of design basis accidents are, in part, functions of the specific activity in the primary coolant and the primary to secondary leakage rates resulting from an accident. Therefore, limits are included in the Technical Specifications for operational leakage and for specific activity in the reactor coolant to ensure the plant is operated in its analyzed condition.

The VCSNS program requires a 150-gallon per day per SG limit for leakage prior to an accident. This limit is incorporated into Technical Specification LCO 3.4.6.2.c by this amendment. Since the new operational leakage limit is more conservative than the existing value, it -will not increase the likelihood or consequences of an accident.

In consideration of the above, past 100% eddy current results after 5.4 EFPY of operation, and the current leak free condition of the SGs, extending the tube inspection frequency does not involve a significant increase in the consequences of a previously evaluated accident.

Summary The proposed changes do not affect the design of the SGs, their method of operation, or primary coolant chemistry controls. The change does not adversely impact any other previously evaluated design basis accident.

Therefore, the changes do not affect the consequences of a SGTR or any other accident.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed Technical Specification changes do not involve changing any structure, system or component or affect plant operations. They are not an initiator of any accident and do not change any FSAR safety analyses.

Primary to secondary leakage that may be experienced during plant conditions is expected to remain within current accident analysis assumptions.

The proposed changes do not affect the design of the SGs, their method of operation, or primary coolant chemistry controls. In addition, the changes do not impact any other plant systems or components.

Document Control Desk Enclosure I LAR 02-2767 RC-03-0176 Page 11 of 14 Therefore, the changes do not create the possibility of a new or different type of accident or malfunction from any accident previously evaluated.

3. Does this change involve a significant reduction in margin of safety?

Response: No.

The SG tubes are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system pressure and inventory.

As part of the RCS boundary, the tubes are unique in that they are also relied upon as a heat transfer medium between the primary and secondary systems such that heat may be removed from the primary system. Additionally, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of the SG is maintained by ensuring the integrity of its tubes.

SG tube integrity is a function of the design, environment, and the physical condition of the tube. Extending the tube inspection frequency and reducing operational leakage limits will not alter the design function of the SGs. Previous inspections conducted during Refuel 12 demonstrate that there is no active tube damage mechanism. The Improved design of the Model Delta 75 generator also provides reasonable assurance that leakage is not likely to occur over the next operating period. Also, the reduced operational leakage limits will increase the margin of safety for primary to secondary leakage conditions.

For the above reasons, the margin of safety is unchanged and overall plant safety will be maintained by the proposed Technical Specification revision.

Pursuant to 10 CFR 50.91, the preceding analyses provide a determination that the proposed Technical Specification change poses no significant hazard as delineated by 10 CFR 50.92.

5.2 Applicable Reaulatory Requirements/Criteria 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants," 1, 2,4,14, 30,31, and 32, define requirements for the reactor coolant pressure boundary (RCPB) with respect to structural and leakage integrity. SG tubing and tube repairs constitute a major fraction of the RCPB surface area. SG tubing and associated repair techniques and components, such as tube plugs, must be capable of maintaining reactor coolant inventory and pressure.

Document Control Desk Enclosure I LAR 02-2767 RC-03-01 76 Page 12 of 14 General Design Criteria 19 of 10 CFR 50, Appendix A, defines requirements for the control room and for the radiation protection of the operators working within it. Accidents involving the leakage or burst of SG tubing comprise a challenge to the habitability of the control room. SG tubing and associated repair techniques and components, such as tube plugs, must be capable of maintaining reactor coolant inventory and pressure in order to prevent excessive leakage and the resulting radiation doses to the control room operator.

10 CFR 50, Appendix B, establishes quality assurance requirements for the design, construction, and operation of safety related components. The pertinent requirements of this appendix apply to all activities affecting the safety related function of SGs that include inspection, testing, operating, and maintaining.

10 CFR 50.65, the Maintenance Rule, classifies SGs as risk significant components because they are relied upon to remain functional during and after design basis events. SGs are monitored under a (2) of the Maintenance Rule against industry established performance criteria. If performance criteria are not met, a cause determination shall be done and the results evaluated to determine if goals should be established per a (1)of the Maintenance Rule.

10 CFR 100, Reactor Site Criteria, establishes the reactor-siting criteria with respect to the risk of public exposure to the release of radioactive fission products. Accidents involving the leakage or burst of SG tubing may comprise a challenge to containment and therefore involve increased risk of radioactive release. SG tubing and associated repair techniques and components, such as tube plugs, must be capable of maintaining reactor coolant inventory and pressure in order to prevent excessive leakage.

VCSNS Technical Specifications include a requirement to shut the plant down when primary to secondary leakage exceeds an established threshold.

VCSNS complies with the above regulatory requirements.

5.2.1 Regulations The regulatory basis for TS 4.4.5, Steam Generators," is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of SG tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of SG tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

Document Control Desk Enclosure I LAR 02-2767 RC-03-01 76 Page 13 of 14 5.2.2 Desian Bases (FSAR)

FSAR Chapter 5 describes the SG tubes and tube sheet as forming the pressure boundary across which heat is transferred and which is designed to prevent or control the transfer of activity within the core to the secondary system.

FSAR Appendix 3A discusses the SCE&G position on Regulatory Guide 1.83 and the exceptions applied to the recommendations of the Regulatory Guide.

5.2.3 Approved Methodologies Regulatory Guide 1.83, "Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes,' Revision 1, July 1976, discusses programs that are acceptable for determining SG tube integrity. This guide was utilized in the formation of the SG management program at VCSNS.

The VCSNS steam generator management program has been updated in accordance with the guidelines of NEI 97-06, an industry initiative related to SG tube integrity.

5.2.4 Analysis The Acceptance Criteria for the SG tube inspections are described in Regulatory Guide 1.83, Revision 1,July 1976.

These criteria were established to reduce the probability and consequences of SG tube failures through periodic inservice inspection for early detection of defects and deterioration.

5.2.5 Conclusion The technical justification performed by SGE&G demonstrates that the proposed amendment has no impact on the design basis function of the SG and its associated tubes. The change represents a preference of extending an inspection based on a shorter operating interval and previous inspection results. Therefore, the proposed License amendment is in compliance with previously established regulatory requirements.

The one-time extension to the inspection frequency requirements allows synchronization with respect to the schedule prescribed by NEI 97-06 guidelines. The reduced operational leakage will increase the margin of safety for primary-to-secondary leakage conditions. Therefore, the proposed License amendment is in compliance with GDC 32.

Document Control Desk Enclosure I LAR 02-2767 RC-03-0176 Page 14 of 14

6.0 ENVIRONMENTAL CONSIDERATION

SCE&G has determined that the proposed amendment would change requirements with respect to the installation or use of a facility component located within the restricted area, as defined in -10 CFR 20 (Reference 7), or would change an inspection or surveillance requirement. SCE&G has evaluated the proposed change and has determined that the change does not involve, (i) a significant hazards consideration, (ii) a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. As discussed above, the proposed changes do not involve a significant hazards consideration and the analysis demonstrates that the consequences of a SG tube rupture are well within design boundaries. Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 (Reference 6), specifically 10 CFR 51.22(c)(9). Therefore, pursuant 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

7.0 REFERENCES

1. FSAR Section(s) 5 and Appendix 3A

2. 10 CFR 50, Appendix A, GDC

3. 10 CFR 50.46

4. 10 CFR 50, Appendix B

5. 10 CFR 100

6. 10 CFR 51

7. 10 CFR 20

Document Control Desk Attachment I LAR 02-2767 RC-03-0176 Page 1 of 6 ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

Attachment to License Amendment No. XXX To Facility Operating License No. NPF-12 Docket No. 50-395 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages 3/4 4-13 3/4 4-13 3/4 4-19 3/4 4-19 B 314 4-3 B 3/44-3 B 314 4-5 B 3/44-5 SCE&G - EXPLANATION OF CHANGES Page Affected Bar Description of Change Reason for Change Section #

3/4 4-13 4.4.5.3.a 1 Add

• at end to identify To note exemption statement footnote. to be applied.

4.4.5.3.a 2 Add footnote To identify a one-time extension of inservice inspection of steam

_______ _ _______ _________________________ generator tubing.

3/4 4-19 3.4.6.2.c 1 Remove 1 GPM total leakage To reflect the conservative limit. Change 500 gallons limits of the VCSNS Steam per day to 150 gallons per Generator Management day. Program.

B 3/4 4-3 314.4.5 1 Change 500 gallons per day To reflect the conservative 2 to 150 gallons per day. limits of the VCSNS Steam Generator Management Program.

B 3/4 4-5 3/4.4.6.2 1 Discussion rewritten to clarify To reflect the conservative 150 gallons per day limit. limits of the VCSNS Steam Generator Management Program.

REACTOR C02LAN SYSTEM SRVEJLLAŽ40E REMIEMMNS12Contnued).

4.4.5.3 lIspectson Freonaencies - The ove requred inservice inspectios of steam giwi tuesili~1epesormed at hes followingfrequencies:

a. Ite firt inservice inspection the seam ge tor replacement shll be perfrmed e lest8Effective FullPowerMonths fom the time of the replacement but within 24 calendar O f i ions hall be performed at fntervals of not less than 12 nor morethan 24 e mths after the previous econ Iftwo eonsecutive inspections f i rvice under A conditions, not Indng thePreseice inspection, result in an inspection fling Ito the C-3 category or iftwo consecutive inpecs demonstrate that previously observed grda has not continued and no aitnal dgradio has occurred, the inspection interval MYbe extended to ama of once per 40 montbs.

b. Iffhe results ofthe inservice ecion. of a steam geerator Conducted in accordancev itie 4.4-2 at 40 month intervals fall in CdegryC-3 tie ispw= reue6hyll beince~dtafi leas once per 20 months. The increase in inspect= frequency until the subsequent inspections iati* the criteria of 8pecifibon 4.4.5.3.a; the interval may then be extended to a maxmm of one per 40 moths.

c. Additional, unscheduled insevice insections shall be performed on eachsteam erator in acordance with the irst sample inspection Ied i Pab i442 duing the shutdown subsequent to any of fimblowirngendtionm

1. P - nd tube leaks (not including leaks riting om tbtube sheet welds) in excess ofthe limits of Speccation 3.4.62.

2. A seismic occurrence greater than the Operting Basis Earthquake.

3. A loss-of-coolant accident requiring atuation of the engineered safeguards.

4. Amain steam line or feedwater line break.

+ rie +i ng.ieffvAt riteVA OV Once yes i sQI...4 4 SUM - UN I 14 4-13 Amendment No. 119

REOR COLNSYSTM OPERATIONAL LEAKAGE 3.4.6.2 Reactor Coolant System leakage shal be limited to-

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPMA UNIDETIFIED LEAKAGE, C. 1 GP'M total primar t o a;cig hu olooam genertror~

isoatodem-lhe-Reator-Coolant-Syseimena6 alons per day ough any one steam generator not isolated from the Reactor Coolant System,

d. 10 (PM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 33 3P3GM CONTRO LFD LEAKAGE at a Reactor Coolant System pressue of 2235

• 20 pelg.

f. The leakae rate specified for each Reactor Coolant System Pressure isolabtor Valve In Table 3.4-1 at a Reactor Coolant System pressure of 2235 i20 psig.

APPIJCABiIllY: MODES 1. 2, 3 and 4 ADTiON:

a. With any PRESSURE BOUNDARY LEAKAGE, be hI at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and hI COL) SHUTDOWN within Ow follovwng 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any Reactor Cooant System leakage greater than any one of the above rlmit excuding PRESSURE BOUNDARY LEAKAGE and Leacage from Reactor Coolant System Pressure Isolation Vahes, reduce the leakage rate to within limits vAthin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be In at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and In COLW SHUTDOWN within the followhng 30 hous

c. With any Reactor Coolant System Pressure Isolatfo Valve Leakage greater ,

than the limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two cosed muanual or deactivated automatic valves, or be Inat least HOT STANDBY witlin the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and In COLD SHUTDOWN within the followIng 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

aEIL LE UFIIREMENTS 4.4.6.2.1 The Reactor Coolant System leakages shalf be demonstrated to be within each of the above limits by:

a. Monitoring the reactor building atmosphere (gaseous or particulate) radioactivity monitor at least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s-SUMMER - UNIT 1 3/4 4-19 Amendment No. 154

REACTOR COOLANT SYSTEM RA8ES 3W4.4 A STEAM gENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion ofte RC8 will be maintained.

The program for inservice inspection of steam generator tubes is based on a modiication fReulatory Guide 2.83, Revision 1. Iservie inspection of steam generator bin is essential in order to maintain surveillance cfbe conditions of the tubes i the event that there is evidence of mechanical damage or prressive degradation due to desmanufacturinjerrors, or nservice conitions that lead toeorrosion v spection fteam geneat tubin also provides a mean ofcharacterizing the nature and eause of any tube degradation sO that corrective measures can be taken.

The lant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the ste erator tubes. If the secondary coolant chemisry isnot maintained i ese limits, localized corrosion may SkSres ins stress corrosion emakig. e etent of eackn durig a oerationwould be limited by the limitation of steam orator tube J

/-J6C?

eksztbetwen t ihe coolant system and te sec soola ystem t~rimaZY-io secon aage = allons per teageneator).

haig a primaecon Crads a e less fathsLnt g operation will!ve an adequate marTin of sey to withstand the loas ipsed during normal operation an by postulated accidents. erating pants ae demonst rated by radiation myp ofsteam generator b Leage inecess of tisli lOdO~ wlrequire plant utd and an tmsieduled Inspection, during which the leakig tubes will b located mud plugged.e Wastage-type defects are unlikely with prper chemistry treatment ofthe secondary coolant. However, even if a defect should develop in service, It willibe found during scheduled inservice steam generator tibe examinations.

Plugging will be re d for all tubes with imperfections exceeding 40% ofthe tube nominal wall thicknesstteam generator tube inspections ofoperating plants have demonstrated the capabilit to reliably detect wastage-type degradation that has penetrated 20% of the original tbe wall thickness.

Whenever the results of any steam generator tubinguinservIce inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to IOCFR5.72(b)2(D prior to resumption of pant uch Seratio.

ca will be considered by the Commission on a case-ky-case basis arid may result in a requiremet for analysis, laborator eainins, tests, additional eddy-current inspection, and revision of the Technical Specfication, fnecessary.

SUMMER - UNIT I B 3/4 4-3 Amendment No. 3.k ,,S

%,N,119 t

REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued)

The PtVlleakage mit s 0.5 GPM per nominal inch oFvalve size wilh a maximum rlnit of 5 GPM. lbs NRC, through NUREG-1431. has endorsed this PIY leakage rate mit The surveilance requirements for RCS Presswe Isolatlon Vales provide added assurance of vae Integrity thereby reducing the probabilltod gross vuv fainure and consequent intersytem LOCA. Leakage from the ROS Pressure Isolation Valves s IDENTIED LEAKAGE and wili be considered as a portion of the allowed imIt Leakage from the RCS Pressure Isolation Valves may be Identified by surveiilance testing perfonned during plant heatup or cooldown above 2000 psig and may be adjusted to L r obtain Ihe leakage value at 2235* 20 psg using calculation guidance provided by ASME Code.

Section XI, Part OM-10.

PRSUE BOUNDARY LEAKAGE of any magnide Is unacceptable since It may be lickatime of an Wnpencing gross failure of t pressur boundary. Therefore, te presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed In COLD SHUTDOWN.

3/4.4.7 CEMISTR The Imitations on Reactor Coolant System chemistryensure that corrosion of the Reato Coolant System Is minimized and reueWh oeta for Reactor Coolant System eakage or falure due to stress corro Maintaining the cmistry within the Steady State Limits provides adequate corrosion protection lo ensure the structural frtegrlty of the Reactor Coolant System over the ltc of the plant The associated effects of exeding the oxygen, chloride and fluoride limits are time and temperature dependent Corrsion studies show that operation may be continued with contaminant concentration levelsIn excess of the Steady State Limits, up to the Transient Umitis, for the specified liited time Intervals without having a gnifcant effect on the stritural Integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Tursient Units provides tM for taking correctlys actions to restore the contaminant concentratifos to witin the Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations In excess enthe limits will be detected i sufficient time to take corrective action.

SUMMER - UNIT I B 34 4-5 Amendment No. 154

Document Control Desk Attachment I LAR 02-2767 RC-03-01 76 Page 6 of 6 INSERT 1-3/4.4.6 The maximum allowed steam generator tube leakage of 450 GPD (3 steam generators with 150 GPD each) for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 150 GPD per steam generator limit preserves the assumptions used in the analysis of these accidents and ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

Document Control Desk Attachment II LAR 02-2767 RC-03-0176 Page 1 of 5 ATTACHMENT II PROPOSED TECHNICAL SPECIFICATION CHANGES (RETYPED)

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection after the steam generator replacement shall be performed after at least 6 Effective Full Power Months from the time of the replacement but within 24 calendar months of initial criticality after the steam generator replacement. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.*

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a maximum of once per 40 months.

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

1. Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2.

2. A seismic occurrence greater than the Operating Basis Earthquake.

3. A loss-of-coolant accident requiring actuation of the engineered safeguards.

4. A main steam line or feedwater line break.

• A one time inspection interval of once per 58 months is allowed for the inspection performed immediately following refueling outage RF-12.

SUMMER - UNIT 1 3/4 4-1 3 Amendment No449,

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day through any one steam generator not isolated from the Reactor Coolant System,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 33 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 +/- 20 psig.

f. The leakage rate specified for each Reactor Coolant System Pressure Isolation Valve in Table 3.4-1 at a Reactor Coolant System pressure of 2235 +/- 20 psig.

APPLICABILITY: MODES 1, 2,3 and 4 ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and Leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With any Reactor Coolant System Pressure Isolation Valve Leakage greater than the limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 The Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the reactor building atmosphere (gaseous or particulate) radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SUMMER - UNIT 1 3/4 4-19 Amendment No. 464,

REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 150 gallons per day per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 150 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage-type degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to 1 CFR50.72(b)2(i) prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

SUMMER - UNIT 1 B 314 4-3 Amendment No. 354-59

- 96, t 1A4Q

REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued)

The PIV leakage limit is 0.5 GPM per nominal inch of valve size with a maximum limit of 5 GPM. The NRC, through NUREG-1431, has endorsed this PIV leakage rate limit.

The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

Leakage from the RCS Pressure Isolation Valves may be identified by surveillance testing performed during plant heatup or cooldown above 2000 psig and may be adjusted to obtain the leakage value at 2235 +/- 20 psig using calculation guidance provided by ASME Code, Section Xl, Part OM-10.

The maximum allowed steam generator tube leakage of 450 GPD (3 steam generators with 150 GPD each) for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 150 GPD per steam generator limit preserves the assumptions used in the analysis of these accidents and ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

SUMMER - UNIT 1 B 34 4-5 Amendment No. 464

Document Control Desk Attachment IlIl LAR 02-2767 RC-03-01 76 Page 1 of 1 ATTACHMENT II LIST OF REGULATORY COMMiTMENTS There are no regulatory commitments created due to this License Amendment Request.

The proposed change allows a frequency extension to an existing inspection commitment.