Text

Request for Change to Technical Specifications to Add Requirements for Steam Generator Tube Integrity, Steam Generator Program, & Steam Generator Tube Inspection Report & to Revise Reactor Coolant System Operational.
	ML050610480
Person / Time
Site:	Salem
Issue date:	02/23/2005
From:	Joyce T; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LCR S04-07, LR-N05-0018
	Download: ML050610480 (65)
	v • d • e

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 FEB 2 3 2005 LR-N05-001 8 0 PStcsErGC LCR S04-07 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS TO ADD REQUIREMENTS FOR STEAM GENERATOR TUBE INTEGRITY, STEAM GENERATOR PROGRAM, AND STEAM GENERATOR TUBE INSPECTION REPORT AND TO REVISE REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE REQUIREMENTS SALEM GENERATING STATION - UNIT 1 DOCKET NO. 50-272 FACILITY OPERATING LICENSE NO. DPR-70 In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear, LLC (PSEG) hereby transmits a request for amendment of the Technical Specifications (TS) for Salem Generating Station Unit 1.

The proposed amendment will provide a programmatic framework for monitoring and maintaining the integrity of steam generator tubes consistent with 10 CFR 50, Appendices A and B. This framework includes performance criteria that, if satisfied, provide reasonable assurance that tube integrity is being maintained.

In addition, this framework provides for monitoring and maintaining the tubes to provide reasonable assurance that the performance criteria are met at all times between scheduled inspections of the tubes.

The proposed amendment adds a new Technical Specification (TS) for steam generator tube integrity by establishing actions that are necessary should the performance criteria not be met.

The proposed amendment adds a new Steam Generator Program to meet the intent of the guidance provided in the Steam Generator Integrity Elements section of NEI 97-06, uSteam Generator Program Guidelines.' The Steam Generator Program is an improvement over existing requirements.

The proposed amendment also adds a new reporting requirement, which provides the requirements for and contents of the Steam Generator Tube Inspection Report. A oI

01-~

95-2168 REV. 7/99

Document Control Desk FEB 2 3 2005 LR-N05-0018 The revision to Reactor Coolant System operational leakage requirements revise the Limiting Conditions for Operation, ACTION requirements and surveillance requirements to clarify the requirements related to primary-to-secondary leakage.

Conforming revisions to the definitions of identified leakage and pressure boundary leakage are made for clarity.

In summary, the proposed amendment removes the detailed inspection requirements from the TS and replaces them with the essential elements of a Steam Generator Program. These changes are consistent with changes previously approved in Amendments 163 and 156 dated September 10, 2004 for the Farley Nuclear Plant, Units I and 2 and Amendments 164 and 154 dated November 24, 2004 for South Texas Project, Units I and 2; and changes proposed in Technical Specification Task Force (TSTF) Traveler TSTF-449, Revision 3, "Steam Generator Tube Integrity," which was transmitted by letter dated January 14, 2005. This will enhance the safety function of the steam generators by increasing the probability that the integrity of the steam generator tubes will be maintained between scheduled inservice inspections. provides an evaluation of the proposed changes. Attachment 2 provides the existing TS pages marked-up to show the proposed changes. summarizes the regulatory commitments made in this submittal. provides the existing TS Bases pages marked-up to show the proposed changes for informational purposes.

PSEG requests a 60-day implementation period after amendment approval.

Approval of this change is requested by September 9, 2005 to support Salem Generating Station Unit 1 refueling outage 1R1 7 currently scheduled to begin on October 16, 2005. A request to change the Salem Unit 2 TS will be submitted following an announcement in the FederalRegister as part of the consolidated line item improvement process (CLIIP).

Should you have any questions regarding this request, please contact Mr.

Courtney Smyth at (856) 339-5298.

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

Executed on Z/23 /OS. Sincerely,

~Date)A E Tom Joy e Site Vice President Salem Generating Station Attachments (4)

Document Control Desk FEB 2 3 2005 LR-N05-0018 C Mr. S. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. D. Collins, Project Manager - Salem & Hope Creek U. S. Nuclear Regulatory Commission Mail Stop 08C2 Washington, DC 20555 USNRC Senior Resident Inspector - Salem Unit 1 and Unit 2 (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P.O. Box 415 Trenton, NJ 08625

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 SALEM GENERATING STATION - UNIT I FACILITY OPERATING LICENSES NOS. DPR-70 DOCKET NO. 50-272 ADDITION OF REQUIREMENTS FOR STEAM GENERATOR TUBE INTEGRITY, STEAM GENERATOR PROGRAM, AND STEAM GENERATOR TUBE INSPECTION REPORT AND REVISION OF REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE REQUIREMENTS Table of Contents

1. DESCRIPTION ...................................... .1

2. PROPOSED CHANGE ....................................... 1

3. BACKGROUND ....................................... 2

4. TECHNICAL ANALYSIS ................................... .. 4

5. REGULATORY SAFETY ANALYSIS ...................................... 25 5.1 No Significant Hazards Consideration ...................................... 25 5.2 Applicable Regulatory Requirements/Criteria ...................................... 28

6. ENVIRONMENTAL CONSIDERATION ...................................... 31

7. REFERENCES ....................................... 31

ATTACHMENT 1 LCR S04-07 LR-N05-0018 CHANGES TO TECHNICAL SPECIFICATIONS

1. DESCRIPTION The purpose of this amendment is to replace the steam generator (SG) detailed programmatic requirements contained in Technical Specifications (TS) with a SG Tube Integrity TS and Bases, revise the TS for reactor coolant system (RCS) Operational Leakage, and add a SG Program and SG Tube Inspection Report. The proposed changes are necessary in order to implement the guidance for the industry initiative on NEI 97-06, "Steam Generator Program Guidelines," (Reference 1). The changes proposed are based on Technical Specification Task Force (TSTF)

Traveler TSTF-449, Revision 3, uSteam Generator Tube Integrity," which was transmitted by letter dated January 14, 2005.

2. PROPOSED CHANGE The detailed, prescriptive requirements in existing Salem Unit 1 TS 3/4.4.5 are replaced by requirements for a new Limiting Condition for Operation (LCO), "Steam Generator (SG) Tube Integrity," a new program 6.8.4.i, "Steam Generator (SG) Program," and a new reporting requirement 6.9.1.10, "Steam Generator Tube Inspection Report." The amendment replaces a large amount of prescriptive, outdated details on SG inspection requirements with a requirement to implement a state of the art performance-based program that is supported by a NEI SG initiative (NEI 97-06), extensive industry guidance, and an active industry Technical Advisory Group. TS 6.8.4.i requires a Steam Generator Program to be established and implemented to ensure that SG tube integrity is maintained, and to describe SG condition monitoring, performance criteria, repair methods, repair criteria, and inspection intervals. TS 6.9.1.10 requires a report within 180 days of initial entry into MODE 4 following a steam generator inspection. These changes are a significant improvement over the existing outdated TS requirements. The TS for SG Tube Integrity contains surveillance requirements for tube integrity verification and repair and actions necessary should tube integrity not be maintained. The proposed changes to Salem Unit 1 TS 3/4.4.6.2, "Reactor Coolant System Operational Leakage," reduce the allowable leakage from any one SG from 500 to 150 gallons per day. The proposed changes to Salem Unit 1 TS 3/4.4.6.2 also revise the LCO, ACTION requirements and Surveillances to clarify the requirements related to primary-to-secondary leakage. The proposed changes to TS 1.15, "IDENTIFIED LEAKAGE," and TS 1.21, "PRESSURE BOUNDARY LEAKAGE," are conforming changes to clarify primary-to-secondary leakage. An editorial change is proposed to TS INDEX page XII, TS 1.19, "OPERATIONAL MODE - MODE," and TS 6.8.4.g.9) to correct a 1

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 typographical error. TS Bases changes are made to reflect the corresponding changes proposed to the TS.

The above changes are shown on the attached marked-up TS pages (Attachment 2). Changes to be inserted in the Bases to reflect the proposed TS changes are included in Attachment 4 for informational purposes.

3. 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 (RCPB) and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the RCPB, the SG tubes are unique in that they act as a heat transfer surface between the primary and secondary systems to remove heat from the primary system.

In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system.

SG tube integrity is necessary in order to satisfy the tubing's safety functions. Maintaining tube integrity ensures that the tubes are capable of performing their intended safety functions consistent with their licensing basis, including applicable regulatory requirements.

Concerns relating to the integrity of the tubing stem from the fact that the SG tubing is subject to a variety of degradation mechanisms. SG tubes have experienced tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. When the degradation of the tube wall reaches a prescribed repair criterion, the tube is considered defective and corrective action is taken.

The criteria governing structural integrity of SG tubes were developed in the 1970s and assumed uniform tube wall thinning. This led to the establishment of a through wall SG tube repair criterion (e.g., 40%) that has historically been incorporated into most pressurized water reactor TS and has been applied, in the absence of other repair criteria, to all forms of SG tube degradation where sizing techniques are available. Since the basis of the through wall depth criterion was 3600 wastage, it is generally considered to be conservative for other mechanisms of SG tube degradation. The repair criterion does not allow licensees the flexibility to manage different types of SG tube degradation. Licensees must either use the through wall criterion for all forms of degradation or obtain 2

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 approval for use of more appropriate repair criteria that consider the structural integrity implications of the given mechanism.

For the last several years, the industry, through the Electric Power Research Institute (EPRI) Steam Generator Management Program (SGMP), has developed a generic approach to improving SG performance referred to as "Steam Generator Degradation Specific Management" (SGDSM). Under this approach, different methods of inspection and different repair criteria may be developed for different types of degradation. A degradation specific approach to managing SG tube integrity has several important benefits. These include:

Improved scope and methods for SG inspection,

Industry incentive to continue to improve inspection methods, and

Development of plugging and repair criteria based on appropriate nondestructive examination (NDE) parameters.

As a result, the assurance of SG tube integrity is improved and unnecessary conservatism is eliminated.

Over the course of this effort, the SGMP has developed a series of EPRI guidelines that define the elements of a successful SG program. These guidelines include:

"Steam Generator Examination Guideline" (Reference 2),

"Steam Generator Integrity Assessment Guideline" (Reference 3),

"In-situ Pressure Testing Guideline" (Reference 4),

"Primary-to-Secondary Leak Guideline" (Reference 5),

uPrimary Water Chemistry Guideline" (Reference 6), and

"Secondary Water Chemistry Guideline" (Reference 7).

These EPRI guidelines, along with NEI 97-06, "Steam Generator Program Guidelines," tie the entire SG program together, while defining a comprehensive, performance based approach to managing SG performance.

In parallel with the industry efforts, the NRC pursued resolution of SG performance issues. In December of 1998, the NRC Staff acknowledged that the Steam Generator Program described by NEI 97-06 and its referenced EPRI Guidelines provides an acceptable starting point to use in the resolution of differences between it and the staffs proposed Generic Letter and draft Regulatory Guide (DG-1074). Since then the industry and the NRC have participated in a series of meetings to resolve the differences and develop the regulatory framework necessary to implement a comprehensive Steam Generator Program.

3

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 Revising the existing regulatory framework to accommodate degradation specific management is the most appropriate way to address the issues of regulatory stability, resource expenditure, use of state-of-the-art inservice inspection techniques, repair criteria, and enforceability. The NRC staff has stated that an integrated approach for addressing SG tube integrity is essential and that materials, systems, and radiological issues that pertain to tube integrity need to be considered in the development of the new regulatory framework.

This license amendment request provides the integrated approach for addressing SG tube integrity.

4. TECHNICAL ANALYSIS The proposed changes do not affect the design of the SGs, their method of operation, or primary coolant chemistry controls. The primary coolant activity limit and its assumptions are not affected by the proposed changes to these TS. The proposed changes are an improvement to the existing SG inspection requirements and provide additional assurance that the plant licensing basis will be maintained between SG inspections.

A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of Salem's licensing basis. The analysis of a SGTR event assumes a bounding primary-to-secondary leakage rate equal to the operational leakage rate limits in the licensing basis plus the leakage rate associated with a double-ended rupture of a single tube.

For design basis accidents such as main steam line break (MSLB), rod ejection, and reactor coolant pump locked rotor, the SG tubes are assumed to retain their structural integrity (i.e., they are assumed not to rupture). These analyses typically assume that primary-to-secondary leakage for all SGs is 1 gallon per minute or increases to 1 gallon per minute as a result of accident-induced stresses. For accidents that do not involve fuel damage, the reactor coolant activity levels are at the TS allowable limits. For accidents that do involve fuel damage, the primary coolant activity values are a function of the amount of activity released from the damaged fuel.

The consequences of these design basis accidents are, in part, functions of the radioactivity levels in the primary coolant and the accident primary-to-secondary leakage rates. As a result, limits are included in the Salem TS for operational leakage and for DOSE EQUIVALENT 1-131 in primary coolant to ensure that Salem is operated within its analyzed condition.

4

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 The proposed TS change includes a reduction in the current TS Reactor Coolant System operational leakage limit from 500 gallons per day to 150 gallons per day. The new limit of 150 gallons per day of primary-to-secondary leakage through any one SG is based on operating experience as an indication of one or more tube leaks. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

The otherTS changes proposed are in general a significant improvement over current requirements. They replace an outdated prescriptive TS with one that references Steam Generator Program requirements that incorporate the latest knowledge of SG tube degradation morphologies and the techniques developed to manage them.

The requirements being proposed are more effective in detecting SG degradation and prescribing corrective actions than those required by current TS. As a result, these proposed changes will result in added assurance of the function and integrity of SG tubes. The proposed requirements are performance based and provide the flexibility to adopt new technology as it matures. These changes are consistent with the guidance in NEI 97-06. Adopting the proposed changes will provide appropriate assurance that SG tubing will remain capable of fulfilling its specific safety function of maintaining RCPB integrity.

The table below and associated sections describe in detail and provide the technical justification for the proposed changes.

5

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 Condition or Requirement Current Licensing Basis Location - Proposed Change Section Operational primary-to-secondary leakage <1 gpm total through all SGs and < 500 RCS Operational leakage TS < 150 1 gallons per day through any one SG gallons per day through any one SG.

RCS primary-to-secondary leakage Reduce LEAKAGE to within limits within 4 RCS Operational leakage TS - Be in at 2 through any one SG not within limits hours or be in at least HOT STANDBY least HOT STANDBY within the next 6 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD hours and in COLD SHUTDOWN within SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

RCS LEAKAGE determined by water Modifying notes not specified Added new Notes indicating SR not 3 inventory balance (Unit 1 SR 4.4.6.2.d) applicable to primary-to-secondary leakage and not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

SR for primary-to-secondary leakage Not specified RCS Operational leakage TS: 4 Added SR to verify primary-to-secondary leakage every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Added Note stating "Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation."

6

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 Condition or Requirement Current Licensing Basis Location - Proposed Change Section Frequency of verification of tube integrity 6 to 40 months depending on SG category SG Tube Integrity TS - Requires 5 (Unit 1 SR 4.4.5.3) defined by previous inspection results. Surveillance Frequency in accordance with TS 6.8.4.i, Steam Generator Program. Frequency is dependent on tubing material and the previous inspection results and the anticipated defect growth rate.

Steam Generator Program -

Establishes maximum inspection intervals Tube sample selection (Unit I SR 4.4.5.2) Based on SG Category, industry Steam Generator Program and 6 experience, random selection, existing implementing procedures - Dependent indications, and results of the initial sample on a pre-outage evaluation of actual set - 3% times the number of SGs at the degradation locations and mechanisms, plant as a minimum and operating experience - 20% of the active tube population as a minimum.

7

ATTACHMENT I LCR S04-07 LR-N05-001 8 Condition or Requirement Current Licensing Basis Location - Proposed Change Section Inspection techniques Not specified SG Tube Integrity TS - Unit 1 SR 7 4.4.5.1 requires that tube integrity be verified in accordance with the Steam Generator Program.

Steam Generator Program and implementing procedures - Establishes requirements for using qualified NDE techniques. Requires use of qualified techniques in SG inspections. Requires a pre-outage evaluation of potential tube degradation morphologies and locations and an identification of NDE techniques capable of finding the degradation.

Inspection Scope (Unit 1 SR 4.4.5.4.a.8) Hot leg point of entry to the top support Steam Generator Program procedures 8 plate on the cold leg side of the U-bend - Inspection scope is defined by the degradation assessment that considers existing and potential degradation morphologies and locations. Explicitly requires consideration of entire length of tube from tube-sheet weld to tube-sheet weld. (The tube-to-tubesheet weld and tube end are not part of the tube.)

8

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 Condition or Requirement Current Licensing Basis Location - Proposed Change Section Performance criteria Operational leakage < I gpm total or < 500 RCS Operational leakage TS - Unit 1 9 gallons per day through any one SG. LCO 3.4.6.2 requires Operational leakage

150 gallons per day through No criteria specified for structural integrity any one SG.

or accident induced leakage.

SG Tube Integrity TS - Unit 1 TS 3/4.4.5 requires that tube integrity be maintained.

TS 6.8.4.i - Defines structural integrity and accident induced leakage performance criteria which are dependent on design basis limits.

Provides provisions for condition monitoring assessment to verify compliance.

Repair criteria (Unit 1 SR 4.4.5.4.a.6) Plug tubes with imperfections extending 2 TS 6.8.4.i - Criteria unchanged 10 40% nominal tube wall thickness.

ACTIONS (Unit 1 LCO 3.4.6.2.c) Performance Criteria not defined. Primary- RCS Operational leakage TS and SG 11 to-secondary leakage limit and actions Tube Integrity TS - Contains primary-included in TS. to-secondary leakage limit, SG tube integrity requirements and ACTIONS u trequired upon failure to meet Plug tubes exceeding plugging limit, performance criteria.

Plug tubes satisfying repair criteria.

9

ATTACHMENT I LCR S04-07 LR-N05-001 8 Condition or Requirement Current Licensing Basis Location - Proposed Change Section Repair methods (Unit 1 SR 4.4.5.4.a.6) Methods (except plugging) require previous TS 6.8.4.i - Requirements unchanged 12 approval by the NRC. No alternate repair criteria has been approved by NRC.

Reporting requirements (Unit 1 SR 4.4.5.5) Plugging report required 15 days after each CFR - Serious SG tube degradation 13 inservice inspection, 12-month report (i.e., tubing fails to meet the structural documenting inspection results, and reports integrity or accident induced leakage in accordance with §50.72 when the criteria) requires reporting in inspection results fall into category C-3. accordance with 50.72 or 50.73.

TS 6.9.1.10 - 180 days after the initial entry into MODE 4 after performing a SG inspection Defining SG Terminology Normal TS definitions (i.e., Definitions TS 6.8.4.i, TS Bases, Steam Generator 14-Section) did not address SG Program Program procedures - Includes Steam issues. The Definitions Section uses the Generator Program terminology term "steam generator tube leakage." applicable only to SGs. The Definitions Section is revised to use the term "primary-to-secondary leakage."

10

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 Section 1: Operational Leakage The primary-to-secondary leakage limit has been reduced to

150 gallons per day through any one SG. The operational leakage rate criterion in conjunction with the implementation of the Steam. Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures. This together with the allowable accident induced leakage limit helps to ensure that the dose contribution from tube leakage will be limited to less than the 10 CFR 100 and GDC 19 dose limits for postulated faulted events.

This limit also contributes to meeting the GDC 14 requirement that the reactor coolant pressure boundary "have an extremely low probability of abnormal leakage, of rapidly propagating to failure, and of gross rupture." The proposed Surveillance ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. The Steam Generator Program uses the EPRI Primary-to-Secondary Leak Guideline (Ref. 5) to establish sampling requirements for determining primary-to-secondary leakage and plant shutdown requirements if leakage limits are exceeded. The guidelines ensure leakage is effectively monitored and timely action is taken before a leaking tube exceeds the performance criteria.

The proposed revision to the technical specification requirement to limit primary-to-secondary leakage through any one SG to less than or equal to 150 gallons per day is significantly more conservative than the existing technical specification limit of 1 gpm total primary-to-secondary leakage through all SGs that is based on an initial condition of the safety analysis.

Section 2: Operational Leakage Actions If primary-to-secondary leakage exceeds 150 gallons per day through any one SG, a plant shutdown must be commenced. The existing technical specifications allow 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to reduce primary-to-secondary leakage to less than the limit. HOT STANDBY must be achieved within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The proposed technical specification removes this allowance.

The removal of the 4-hour period during which primary-to-secondary leakage can be reduced to avoid a plant shutdown results in a technical specification that is significantly more conservative than the existing RCS Operational Leakage specification. This change is consistent with the Steam Generator Program that also does not allow 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months before commencing a plant shutdown.

Section 3: RCS Operational Leakage Determined by Water Inventory Balance The proposed change adds Notes to Unit 1 SR 4.4.6.2.d that make the water inventory balance method not applicable to determining primary-to-secondary 11

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 leakage and allows the SR to not be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. This change is proposed because primary-to-secondary leakage as low as 150 gallons per day through any one SG cannot be measured accurately by an RCS water inventory balance. This change is necessary to make the surveillance requirement appropriate for the proposed LCO.

Section 4: SG Tube Integrity Verification Unit 1 SR 4.4.6.2.c has been added to verify the LCO requirement on primary-to-secondary leakage only. Steam generator tube integrity is verified in accordance with a SR in the SG Tube Integrity Specification.

The Steam Generator Program and the EPRI "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines" (Ref. 5) provide guidance on leak rate monitoring. During normal operation the program depends upon continuous process radiation monitors and/or radiochemical grab sampling in accordance with the EPRI guidelines. The monitoring and sampling frequency increases as the amount of detected leakage increases or if there are no continuous radiation monitors available.

Determination of primary-to-secondary leakage is required every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The SR is modified by a Note stating the SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of stable operating conditions. As stated above, additional monitoring of primary-to-secondary leakage is also required by the Steam Generator Program based upon guidance provided in Reference 5.

Section 5: Frequency of Verification of SG Tube Integrity The current technical specifications contain prescriptive inspection intervals which depend on the condition of the tubes as determined by the last SG inspection. The tube condition is classified into one of three categories based on the number of tubes found degraded and defective. The minimum inspection interval is no less than 12 and no more than 24 months unless the results of two consecutive inspections are in the best category (no additional degradation), and then the interval can be extended to 40 months.

The surveillance Frequency in the proposed Steam Generator Tube Integrity specification is governed by the requirements in the Steam Generator Program and specifically by References 2 and 3. The proposed Frequency is also prescriptive, but has a stronger engineering basis than the existing technical specification requirements. The interval is dependent on tubing material and whether any active degradation associated with cracking is found. The interval is limited by existing and potential degradation mechanisms and their anticipated growth rate. In addition, a maximum inspection interval is established in TS 6.8.4.i.

12

ATTACHMENT 1 LCR S04-07 LR-N05-0018 Salem Unit I SGs are constructed of Alloy 600 thermally treated tubing. The maximum inspection interval for Alloy 600 thermally treated tubing is "Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected."

Even though the maximum interval for Alloy 600 thermally treated tubing is slightly longer than allowed by current technical specifications, it is only applicable to SGs with advanced materials, it is only achievable early in SG life and only if the SGs are free from active degradation associated with cracking. In addition, the interval must be supported by an evaluation that shows that the performance criteria will continue to be met at the next SG inspection. Taken in total, the proposed inspection intervals provide a larger margin of safety than the current requirements because they are based on an engineering evaluation of the tubing condition and potential degradation mechanisms and growth rates, not only on the previous inspection results. As an added safety measure, the Steam Generator Program requires a minimum sample size at each inspection that is significantly larger than that required by current technical specifications (20 percent versus 3 percent times the number of SGs in the plant); thus providing added assurance that any degradation within the SGs will be detected and accounted for in establishing the inspection interval.

The proposed maximum inspection intervals are based on the historical performance of advanced SG tubing materials. Reference 8 shows that the performance of Alloy 600TT is significantly better than the performance of 600MA tubing, the material used in SG tubing at the time that the current technical specifications were written. There have been very few instances of cracking in 600TT tubes in a U.S. SG and this degradation appears to be limited to a small number of tubes in specific SGs that were left with high residual stress as a result of a problem in their manufacturing process.

In summary, the proposed change is an improvement over the current technical specification. The current technical specification bases inspection intervals on the results of previous inspections; it does not require an evaluation of expected performance. The proposed technical specification uses information from previous plant inspections as well as industry experience to evaluate the length of time that the SGs can be operated and still provide reasonable assurance that the performance criteria will be met at the next inspection. The actual interval is the shorter of the evaluation results and the requirements in Reference 3.

Allowing plants to use the proposed inspection intervals maximizes the potential 13

ATTACHMENT I LCR S04-07 LR-N05-001 8 that plants will use improved techniques and knowledge since better knowledge of SG conditions supports longer intervals.

Section 6: SG Tube Sample Selection The current technical specifications base tube selection on SG conditions and industry and plant experience. The minimum sample size is 3% of the tubes times the number of SGs in the plant. The proposed change refers to the Steam Generator Program degradation assessment guidance for sampling requirements. The minimum sample size is 20% of the active tube population inspected.

The Steam Generator Program requires the preparation of a degradation assessment. The degradation assessment is the key document used for planning a SG inspection, where inspection plans and related actions are determined, documented, and communicated. The degradation assessment addresses the various reactor coolant pressure boundary components within the SG (e.g., plugs, sleeves, tubes, and components that support the pressure boundary.) In a degradation assessment, tube sample selection is performance based and is dependent upon actual SG conditions and plant operational experience and of the industry in general. Existing and potential degradation mechanisms and their locations are evaluated to determine which tubes will be inspected. Tube sample selection is adjusted to minimize the possibility that tube integrity might degrade during an operating cycle beyond the limits defined by the performance criteria. The EPRI Steam Generator Examination Guidelines (Ref.

2) and EPRI Steam Generator Integrity Assessment Guidelines (Ref. 3) provide guidance on degradation assessment.

In general, the sample selection considerations required by the current technical specifications and the requirements in the Steam Generator Program as proposed by this change are consistent, but the Steam Generator Program provides more guidance on selection methodologies and incorporation of industry experience and requires more extensive documentation of the results.

Therefore, the sample selection method proposed by this change is more conservative than the current technical specification requirements. In addition, the minimum sample size in the proposed requirements is larger.

Section 7: SG lnspection Techniques The Surveillance Requirements proposed in the Steam Generator Tube Integrity specification require that tube integrity be verified in accordance with the requirements of the Steam Generator Program. The Steam Generator Program uses the EPRI Steam Generator Examination Guidelines (Ref. 2) to establish requirements for qualifying NDE techniques and maintains a list of qualified techniques and their capabilities.

14

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 The Steam Generator Program requires the performance of a degradation assessment and refers utilities to EPRI Steam Generator Examination Guidelines (Ref. 2) and EPRI Steam Generator Integrity Assessment Guidelines (Ref. 3) for guidance on its performance. The degradation assessment will identify current and potential degradation locations and mechanisms and NDE techniques that are effective in detecting their existence. Tube inspection techniques are chosen to reliably detect flaws that might progress during an operating cycle beyond the limits defined by the performance criteria.

The current technical specifications contain no requirements on NDE inspection techniques. The proposed change is an improvement over the current technical specifications that contained no similar requirement.

Section 8: SG Inspection Scone The current technical specifications include a definition of tube inspection that specifies the end points of the eddy-current examination of each tube. An inspection is required from the point of entry of the tube on the hot leg side to the top support plate on the cold leg side of the tube after the U-bend. This definition is overly prescriptive and simplistic and has led to interpretation questions in the past.

The Steam Generator Program states, "The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations." The Steam Generator Program provides extensive guidance and a defined process, the degradation assessment, for determining the extent of a tube inspection. This guidance takes into account industry and plant specific history to determine potential degradation mechanisms and the location that they might occur within the SG. This information is used to define a performance based inspection scope targeted on plant specific conditions and SG design.

The proposed change is an improvement over the current technical specifications because it focuses the inspection effort on the areas of concern.

15

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 Section 9: SG Performance Criteria The proposed change adds a performance-based Steam Generator Program to the Technical Specifications. A performance-based approach has the following attributes:

measurable parameters,

objective criteria to assess performance based on risk-insights,

deterministic analysis and/or performance history, and

licensee flexibility to determine how to meet established performance criteria.

The performance criteria used for SGs are based on tube structural integrity, accident induced leakage, and operational leakage. The structural integrity and accident induced leakage criteria were developed deterministically and are consistent with Salem's licensing basis. The operational leakage criterion was based on providing an effective measure for minimizing the frequency of tube ruptures at normal operating and faulted conditions. The proposed structural integrity and accident induced leakage performance criteria are new requirements. The performance criteria are specified in TS 6.8.4.i. The requirements and methodologies established to meet the performance criteria are documented in the Steam Generator Program. The current technical specifications contain only the operational leakage criterion; therefore, the proposed change is more conservative than the current requirements.

The SG performance criteria identify the standards against which performance is to be measured. Meeting the performance criteria provides reasonable assurance that the SG tubing will remain capable of fulfilling its specific safety function of maintaining RCPB integrity throughout each operating cycle.

The structural integrity performance criterion is:

"Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down 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 primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design

'basis accidents, or combination of accidents in accordance 16

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. Inthe assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

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

Adjustments to include contributing loads are addressed in the applicable EPRI guidelines.

Normal steady state full power operation is defined as the conditions existing during MODE 1 operation at the maximum steady state reactor power as defined in the design or equipment specification. Changes in design parameters such as plugging levels, primary or secondary modifications, or THOT should be assessed and included if significant.

The definition of normal steady state full power operation is important as it relates to application of the safety factor of three in the structural integrity performance criterion. The criterion requires "...retaining a safety factor of 3.0 under normal steady state full power operation primary-to-secondary pressure differential...'.

The application of the safety factor of three to normal steady state full power operation is founded on past NRC positions, accepted industry practice, and the intent of the ASME Code for original design and evaluation of inservice components. The assumption of normal steady state full power operating pressure differential has been consistently used in the analysis, testing and verification of tubes with stress corrosion cracking for verifying a safety factor of three against burst. Additionally, the 3AP criterion is measurable through the condition monitoring process.

The actual operational parameters may differ between cycles. As a result of changes to these parameters, reaching the differential pressure in the equipment specification may not be possible during plant operations. Therefore, the definition allows adjustment of the 3AP limit for changes in these parameters when necessary. Further guidance on this adjustment is provided in Appendix M of the EPRI Steam Generator Integrity Assessment Guidelines (Ref. 3).

The accident induced leakage performance criterion is:

"The primary-to-secondary accident induced leakage rate for all design basis accidents, other than a steam generator tube rupture, 17

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all steam generators and leakage rate for an individual steam generator. Leakage is not to exceed 1 gpm per SG."

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

When calculating offsite doses, the safety analysis for the limiting Design Basis Accident, other than a steam generator tube rupture, assumes a total of 1 gpm primary-to-secondary leakage as an initial condition. Recent experience with degradation mechanisms involving tube cracking has revealed that leakage under accident conditions can exceed the level of operating leakage by orders of magnitude. The NRC has concluded (Item Number 3.4 in Attachment 1 to Reference 13) that additional research is needed to develop an adequate methodology for fully predicting the effects of leakage on the outcome of some accident sequences. Therefore, a separate performance criterion was established for accident-induced leakage. The limit for accident-induced leakage is 1 gpm, which is the plant's design basis.

The operational leakage performance criterion is:

'The RCS operational primary-to-secondary leakage through any one steam generator shall be limited to 150 gallons per day."

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

The operational leakage performance criterion is documented in the Steam Generator Program and implemented in Unit 1 LCO 3.4.6.2, "RCS Operational LEAKAGE."

Proposed Administrative TS 6.8.4.i contains the performance criteria and is more conservative than the current technical specifications. The current technical specifications do not address the structural integrity and accident induced leakage criteria. In addition, the primary-to-secondary leakage limit (150 gallons per day per SG) included in the proposed changes is more conservative than the primary-to-secondary leakage limit in the current RCS operational leakage specification.

18

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 Section 10: SG Repair Criteria Repair criteria are those NDE measured parameters at or beyond which the tube must be removed from service by plugging.

Tube repair criteria are established for each active degradation mechanism.

Tube repair criteria are the standard through-wall depth-based criterion (i.e., 40%

through-wall.) This requirement is unchanged from the current technical specifications.

The surveillance requirements of the proposed Steam Generator Integrity specification require that tubes that satisfy the tube repair criteria be plugged in accordance with approved methods. SG tubes experiencing a damage form or mechanism for which no depth sizing capability exists are "plugged-on-detection" and their integrity should be assessed. It cannot be guaranteed that every flaw will be detected with a given eddy-current technique and, therefore, it is possible that some flaws will not be detected during an inspection. If a flaw is discovered and it is determined that this flaw would have satisfied the repair criteria at the time of the last inspection of the affected tube, this does not mean that the Steam Generator Program was violated. However, it may be an indication of a shortcoming in the inspection program.

Section 11: ACTIONS The RCS Operational Leakage and Steam Generator Tube Integrity specifications require the licensee to monitor SG performance against performance criteria in accordance with the Steam Generator Program.

During plant operation, monitoring is performed using the operational leakage criterion. Exceeding that criterion will lead to a plant shutdown in accordance with Unit 1 LCO 3.4.6.2. Once shutdown, the Steam Generator Program will ensure that the cause of the operational leakage is determined and corrective actions are taken to prevent recurrence. Operation may resume when the requirements of the Steam Generator Program have been met. This requirement is unchanged from the current technical specifications.

Also during plant operation the licensee may discover an error or omission that indicates a failure to implement a required plugging during a previous SG inspection. Under these circumstances, the licensee is expected to take the ACTION requirements required by the Steam Generator Tube Integrity specification. If a performance criterion has been exceeded, a principal safety barrier has been challenged and 10 CFR 50.72 (b) (3) (ii) (A) and 50.73 (a) (2) (ii)

(A) require NRC notification and the submittal of a report containing the cause and corrective actions to prevent recurrence. The Steam Generator Program additionally requires that the report contain information on the performance criteria exceeded and the basis for the planned operating cycle. The current 19

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 technical specifications only address operational leakage during operations and therefore do not include the proposed requirement.

During MODES 5 and 6,the operational leakage criterion is not applicable, and the SGs will be inspected as required by the surveillance in the Steam Generator Tube Integrity specification. A condition monitoring assessment of the "as found" condition of the SG tubes will be performed to determine the condition of the SGs with respect to the structural integrity and accident leakage performance criteria.

If the performance criteria are not met, the Steam Generator Program requires ascertaining the cause and determining corrective actions to prevent recurrence.

Operation may resume when the requirements of the Steam Generator Program have been met.

The proposed technical specification's change to the ACTION requirements required upon exceeding the operational leakage criterion is conservative with respect to the current technical specifications as explained in Section 2 above.

The current technical specifications do not address ACTION requirements required while operating if it is discovered that the structural integrity or accident induced leakage performance criteria or a repair criterion are exceeded, so the proposed change is conservative with respect to the current technical specifications.

If performance or repair criteria are exceeded while shutdown, the affected tubes must be plugged. A report will be submitted to the NRC in accordance with Technical Specification 6.9.1.10. The changes in the required reports are discussed in Section 13 below.

Section 12: SG Repair Methods Repair methods are those means used to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. Plugging a SG tube is not a repair.

The purpose of a repair is typically to reestablish or replace the RCPB. The proposed Steam Generator Tube Integrity surveillance requirements requires that tubes that satisfy the tube repair criteria be plugged in accordance with the Steam Generator Program. Salem does not have any NRC approved repair method established. A subsequent license change request will be submitted when a repair method is proposed.

Steam generator tubes experiencing a damage form or mechanism for which no depth sizing capability exists are uplugged-on-detection' and their integrity is assessed. This requirement is unchanged by the proposed technical specifications.

20

ATTACHMENT 1 LCR S04-07 LR-N05-0018 Note that SG plug designs do not require NRC review and therefore plugging is not considered a repair in the context of this requirement.

The above approach is not a change to the technical specifications.

Section 13: Reportina Requirements The current technical specifications require the following reports:

A report listing the number of tubes plugged or repaired in each SG submitted within 15 days of the end of the inspection.

A SG inspection results report submitted within 12 months after the inspection.

Reports required pursuant to 10 CFR 50.73.

The proposed changes to Technical Specifications replace the 15-day and the SG inspection reports with one report required within 180 days. The proposed report also contains more information than the current SG inspection report. This provision expands the report to provide more substantive information and will be sent earlier (180 days versus 12 months). This allows the NRC to focus its attention on the more significant conditions.

The guidance in NUREG-1022, Rev. 2, "Event Reporting Guidelines 10 CFR 50.72 and 50.73," identifies serious SG tube degradation as an example of an event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. Steam generator tube degradation is considered serious if the tubing fails to meet the structural integrity or accident induced leakage performance criteria. Serious SG tube degradation would be reportable in accordance with 10 CFR 50.72 (b) (3) (ii) (A) and 50.73 (a) (2) (ii) (A) requiring NRC notification and the submittal of a report containing the cause and corrective actions to prevent recurrence.

The proposed reporting requirements are an improvement as compared to those required by the current technical specifications. The proposed reporting requirements are more useful in identifying the degradation mechanisms and determining their effects. In the unlikely event that a performance criterion is not met, NEI 97-06 (Ref. 1) directs the licensee to submit additional information on the root cause of the condition and the basis for the next operating cycle.

The changes to the reporting requirements are performance based. The new requirements remove the burden of unnecessary reports from both the NRC and the licensee, while ensuring that critical information related to problems and significant tube degradation is reported more completely and, when required, more expeditiously than under the current technical specifications.

21

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 Section 14: SG Terminologq The proposed Steam Generator Tube Integrity specification Bases explain a number of terms that are important to the function of a Steam Generator Program. The Technical Specification Bases are controlled by the Technical Specification Bases Control Program, which appears in the Administrative Technical Specifications. Changes are proposed to the TS Definitions Section terms "IDENTIFIED LEAKAGE" and "PRESSURE BOUNDARY LEAKAGE".

The terms are described below.

1. Accident induced leakage rate means the primary-to-secondary leakage rate occurring during postulated accidents other than a steam generator tube rupture. This includes the primary-to-secondary leakage rate existing immediately prior to the accident plus additional primary-to-secondary leakage induced during the accident.

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 cause radiological dose consequences in excess of the 10 CFR Part 100 guidelines for offsite doses, or the GDC 19 requirements for control room personnel.

2. The Steam Generator Tube Integrity Bases define the term "burst" as "the gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation."

Since a burst definition is required for condition monitoring, a definition that can be analytically defined and is capable of being assessed via in situ and laboratory testing is necessary. Furthermore, the definition must be consistent with ASME Code requirements, and apply to most forms of tube degradation.

The definition developed for tube burst is consistent with the testimony of James Knight (Ref. 9), and the historical guidance of draft Regulatory Guide 1.121 (Ref. 10). The definition of burst per these documents is in relation to gross failure of the pressure boundary; e.g., "the degree of loading required to burst or collapse a tube wall is consistent with the design margins in Section IlIl of the ASME B&PV Code (Ref. 11)." Burst, or gross failure, according to the Code would be interpreted as a catastrophic failure of the pressure boundary.

22

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 The above definition of burst was chosen for a number of reasons:

The burst definition supports field application of the condition monitoring process. For example, verification of structural integrity during condition monitoring may be accomplished via in situ testing. Since these tests do not have the capability to provide an unlimited water supply, or the capability to maintain pressure under certain leakage scenarios, opening area may be more a function of fluid reservoir rather than tube strength.

Additionally, in situ designs with bladders may not be reinforced. In certain cases, the bladder may rupture when tearing or extension of the defect has not occurred. This condition may simply mean the opening of the flanks of the defect was sufficient to permit extrusion of the bladder, and that the actual, or true, burst pressure was not achieved during the test. The burst definition addresses this issue.

The definition does not characterize local instability or "ligament pop-through", as a burst. The onset of ligament tearing need not coincide with the onset of a full burst. For example, an axial crack about 0.5" long with a uniform depth at 98% of the tube wall would be expected to fail the remaining ligament, (i.e., extend the crack tip in the radial direction) due to deformation during pressurization at a pressure below that required to cause extension at the tips in the axial direction. Thus, this would represent a leakage situation as opposed to a burst situation and a factor of safety of three against crack extension in the axial direction may still be demonstrated. Similar conditions have been observed for localized deep wear indications.

3. The Steam Generator Tube Integrity Bases define a SG tube as, "the entire length of the tube, including the tube wall and any repairs to it, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube."

This definition ensures that all portions of SG tubes that are part of the RCPB, with the exception of the tube-to-tubesheet weld, are subject to Steam Generator Program requirements. The definition is also intended to exclude tube ends that cannot be NDE inspected by eddy-current.

For the purposes of SG tube integrity inspection, any weld metal in the area of the tube end is not considered part of the tube. This is necessary since the acceptance requirements for tubing and weld metals are different.

4. The Steam Generator Tube Integrity Bases define the term "collapse" as "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero."

23

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 In dealing with pure pressure loadings, burst is the only failure mechanism of interest. If bending loads are introduced in combination with pressure loading, the definition of failure must be broadened to encompass both burst and bending collapse. Which failure mode applies depends on the relative magnitude of the pressure and bending loads and also on the nature of any flaws that may be present in the tube. Guidance on assessing applicable failure modes is provided in the EPRI steam generator guidelines.

5. The Steam Generator Tube Integrity Bases define the term "significant" as used in the structural integrity performance criterion as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established."

6. The Steam Generator Tube Integrity Bases describes how to determine whether thermal loads are primary or secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

7. TS Definitions "IDENTIFIED LEAKAGE" and "PRESSURE BOUNDARY LEAKAGE" are revised to clarify that steam generator tube leakage to the secondary system is referred to as primary-to-secondary leakage. An editorial change is made to the definition of OPERATIONAL MODE - MODE to correct punctuation. These changes to TS Definitions are administrative in nature and have no impact on safety.

Conclusion The proposed changes will provide greater assurance of SG tube integrity than that offered by the current technical specifications. The proposed requirements are performance based and provide the flexibility to adopt new technology as it matures. These changes are consistent with the guidance in NEI 97-06, MSteam Generator Program Guidelines," (Ref. 1).

Adopting the proposed changes will provide added assurance that SG tubing will remain capable of fulfilling its specific safety function of maintaining RCPB integrity.

24

ATTACHMENT I LCR S04-07 LR-N05-001 8

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration The proposed changes are necessary in order to implement the guidance for the industry initiative on NEI 97-06, "Steam Generator Program Guidelines." PSEG Nuclear, LLC (PSEG) has evaluated whether or not a significant hazards consideration is involved with the proposed changes to TS 1.15, "Identified Leakage," TS 1.21, "Pressure Boundary Leakage,"

Salem Unit I TS 3/4.4.6.2, "Reactor Coolant System Operational Leakage," and the additions of Salem Unit 1 TS 3/4.4.5, "Steam Generator (SG) Tube Integrity," TS 6.8.4.i, 'Steam Generator (SG) Program," and 6.9.1.10, "Steam Generator Tube Inspection Report," by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment" as discussed below:

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

Response: No.

The proposed change requires a Steam Generator Program that includes performance criteria that will provide reasonable assurance that the steam generator (SG) tubing will retain integrity over the full range of operating conditions (including startup, operation in the power range, hot standby, cool down and all anticipated transients included in the design specification). The SG performance criteria are based on tube structural integrity, accident induced leakage, and operational leakage.

The structural integrity performance criterion is:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down 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 primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. Inthe assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and 25

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

The accident induced leakage performance criterion is:

The primary-to-secondary accident induced leakage rate for any design basis accidents, 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.

The operational leakage performance criterion is:

The reactor coolant system operational primary-to-secondary leakage through any one SG shall be limited to 150 gallons per day.

A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. In the analysis of a SGTR event, a bounding primary-to-secondary leakage rate equal to the operational leakage rate limits in the licensing basis plus the leakage rate associated with a double-ended rupture of a single tube is assumed.

For other design basis accidents such as main steam line break (MSLB),

rod ejection, and reactor coolant pump locked rotor the tubes are assumed to retain their structural integrity (i.e., they are assumed not to rupture). These analyses assume that primary-to-secondary leakage for all SGs is 1 gallon per minute or increases to 1 gallon per minute as a result of accident-induced stresses. The accident induced leakage criterion retained by the proposed changes accounts for tubes that may leak during design basis accidents. The accident induced leakage criterion limits this leakage to no more than the value assumed in the accident analysis.

The SG performance criteria proposed as part of these TS changes identify the standards against which tube integrity is to be measured.

Meeting the performance criteria provides reasonable assurance that the SG tubing will remain capable of fulfilling its specific safety function of maintaining reactor coolant pressure boundary integrity throughout each operating cycle and in the unlikely event of a design basis accident. The performance criteria are only a part of the Steam Generator Program required by the proposed addition of TS 6.8.4.i. The program defined by NEI 97-06 includes a framework that incorporates a balance of prevention, inspection, evaluation, repair, and leakage monitoring.

26

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 The consequences of design basis accidents are, in part, functions of the DOSE EQUIVALENT 1-131 in the primary coolant and the primary-to-secondary leakage rates resulting from an accident. Therefore, limits are included in the Salem TS for operational leakage and for DOSE EQUIVALENT 1-131 in primary coolant to ensure the plant is operated within its analyzed condition. The Salem analysis of the limiting design basis accident assumes that primary-to-secondary leak rate after the accident is 1 gallon per minute with no more than 500 gallons per day through any one SG, and that the reactor coolant activity levels of DOSE EQUIVALENT 1-131 are at the TS values before the accident.

The proposed change does not affect the design of the SGs, their method of operation, or primary coolant chemistry controls. The proposed approach updates the current TS and enhances the requirements for SG inspections. The proposed change does not adversely impact any other previously evaluated design basis accident and is an improvement over the current TS.

Therefore, the proposed changes do not affect the consequences of a SGTR accident and the probability of such an accident is reduced. In addition, the proposed changes do not affect the probabilities or consequences of an MSLB, rod ejection, or a reactor coolant pump locked rotor event.

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 performance based requirements are an improvement over the requirements imposed by the current TS.

Implementation of the proposed Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The result of the implementation of the Steam Generator Program will be an enhancement of SG tube performance. Primary-to-secondary leakage that may be experienced during all plant conditions will be monitored to ensure it remains within current accident analysis assumptions.

The proposed changes do not affect the design of the SGs, their method of operation, or primary or secondary coolant chemistry controls. In addition, the proposed change does not impact any other plant system or component. The change enhances SG inspection requirements.

27

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change is expected to result in an improvement in the tube integrity by implementing the Steam Generator Program to manage SG tube inspection, assessment, and plugging. The requirements established by the Steam Generator Program are consistent with those in the applicable design codes and standards and are an improvement over the requirements in the current TS.

For the above reasons, the margin of safety is not changed and overall plant safety will be enhanced by the proposed changes to the TS.

Based on the above, PSEG concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria The regulatory requirements applicable to SG tube integrity are the following:

10 CFR 50.55a, Codes and Standards - Section (b), ASME Code - c)

Reactor coolantpressure boundary. (1) Components which are part of the reactor coolant pressure boundary must meet the requirements for Class 1 components in Section III of the ASME Boiler and Pressure Vessel 28

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 Code, except as provided in paragraphs (c)(2), (c)(3), and (c)(4) of this section.

The proposed change and the Steam Generator Program requirements that underlie it are in full compliance with the ASME Code. The proposed TS are more effective at ensuring tube integrity and, therefore, compliance with the ASME Code, than the current TS as described in Section 4.0 (Technical Analysis).

10 CFR 50.65 Maintenance Rule - Each holder of a license to operate a nuclear power plant under §§50.21 (b) or 50.22 shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, as defined in paragraph (b), are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety and, where practical, take into account industry-wide operating experience. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken.

Under the Maintenance Rule, PSEG has classified SGs as risk significant components because they are relied on to remain functional during and after design basis events. The performance criteria included in the proposed TS are used to demonstrate that the condition of the SG "is being effectively controlled through the performance of appropriate preventive maintenance" (Maintenance Rule §(a)(2)). If the performance criteria are not met, a root cause determination of appropriate depth is done and the results evaluated to determine if goals should be established per §(a)(1) of the Maintenance Rule.

10 CFR 50, Appendix A. GDC 14 - Reactor CoolantPressureBoundary.

The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, or rapidly propagating failure, and of gross rupture.

There are no changes to the SG design that impact this general design criterion. The evaluation performed in Section 4.0 (Technical Analysis) concludes that the proposed change will continue to comply with this regulatory requirement.

10 CFR 50. Appendix A. GDC 30 - Quality of reactorcoolantpressure boundary. Components that are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

29

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 There are no changes to the SG design that impact this general design criterion. The evaluation performed in Section 4.0 (Technical Analysis) concludes that the proposed change will continue to comply with this regulatory requirement.

10 CFR 50. Appendix A. GDC 32 - Inspection of reactorcoolantpressure boundary. Components which are part of the reactor coolant pressure boundary shall be designed to (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

There are no changes to the SG design that impact this general design criterion. The evaluation performed in Section 4.0 (Technical Analysis) concludes that the proposed change will continue to comply with this regulatory requirement.

General Design Criteria (GDC) 14. 30, and 32 of 10 CFR Part 50.

Appendix A, define requirements for the reactor coolant pressure boundary with respect to structural and leakage integrity. Steam generator tubing and tube repairs constitute a major fraction of the reactor coolant pressure boundary surface area. Steam generator tubing and associated repair techniques and components, such as plugs and sleeves, must be capable of maintaining reactor coolant inventory and pressure.

The Steam Generator Program required by the proposed TS establishes performance criteria, repair criteria, repair methods, inspection intervals and the methods necessary to meet them. These requirements provide reasonable assurance that tube integrity will be met in the interval between SG inspections.

The proposed changes provide requirements that are more effective in detecting SG degradation and prescribing corrective actions. The proposed changes result in added assurance of the function and integrity of SG tubes.

10 CFR 50. Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. "Quality assurance" comprises all those planned and systematic actions necessary to provide adequate confidence that a structure, system, or component will perform satisfactorily in service.

The SG Program required by the proposed TS establishes performance criteria, repair criteria, repair methods, inspection intervals and the methods necessary to meet them. These requirements provide 30

ATTACHMENT 1 LCR S04-07 LR-N05-001 8 reasonable assurance that the SG will perform satisfactorily in service and meet this regulatory requirement.

Therefore, based on the considerations discussed above:

1) There is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner;

2) Such activities will be conducted in compliance with the Commission's regulations; and

3) Issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6. ENVIRONMENTAL CONSIDERATION PSEG has determined the proposed amendment relates to changes in a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or relates to changes in an inspection or a surveillance requirement. The proposed amendment does not involve (i) a significant hazards consideration, (ii)a significant change in the types 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. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 .22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

an environmental impact statement or environmental assessment of the proposed change is not required.

7. REFERENCES

1. NEI 97-06, "Steam Generator Program Guidelines."

2. EPRI, "Steam Generator Examination Guideline."

3. EPRI, "Steam Generator Integrity Assessment Guideline."

4. EPRI, "Steam Generator In-situ Pressure Test Guideline."

5. EPRI, "PWR Primary-to-Secondary Leak Guideline."

6. EPRI, "Primary Water Chemistry Guideline."

7. EPRI, "Secondary Water Chemistry Guideline."

31

ATTACHMENT 1 LCR S04-07 LR-N05-0018

8. EPRI Report R-5515-00-2, "Experience of US and Foreign PWR Steam Generators with Alloy 600TT and Alloy 690TT Tubes and Sleeves," June 5, 2002.

9. Testimony of James Knight Before the Atomic Safety and Licensing Board, Docket Nos. 50-282 and 50-306, January 1975.

10. Draft Regulatory Guide 1.121, "Bases for Plugging Degraded Steam Generator Tubes," August 1976.

11. ASME B&PV Code, Section 11I, Rules for Construction of Nuclear Facility Components.

12. NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Revision 3.

13. S. C. Collins memo to W. D. Travers, "Steam Generator Action Plan Revision to Address Differing Professional Opinion on Steam Generator Tube Integrity," May 11, 2001.

14. The NRC has approved a similar license amendment for Farley Nuclear Plant, Units 1 and 2 - Amendments 163 and 156 dated September 10, 2004.

15. The NRC has approved a similar license amendment for South Texas Project, Units 1 and 2 - Amendments 164 and 154 dated November 24, 2004.

32

ATTACHMENT 2 LCR S04-07 LR-N05-001 8 TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES Salem Unit 1 Affected Page List Index Page V Index Page XII The following Technical Specifications for Salem Unit 1 Facility Operating License DPR-70 are affected by this change request:

Technical Specification Pa1e 1.15, "Identified Leakage' 1-4 1.19, "Operational MODE - MODE" 1-4 1.21, "Pressure Boundary Leakage" 1-5 3/4.4.5, "Steam Generator (SG) Tube Integrity" 3/4 4-7 through 3/4 4-13a 3/4.4.6.2, "Operational Leakage" 3/4 4-15 and 3/4 4-16 6.8.4.i, "Steam Generator (SG) Program" 6-1 9b 6.9.1.10, "Steam Generator Tube Inspection Report" 6-24a

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS Normal Operation. . . . . . . . . . . . . . . . . .3/4 4-1 Hot Standby .................. . .3/4 4-2 Hot Shutdown ................. . .3/4 4-3 Cold Shutdown ................. . .3/4 4-3b 3/4.4.2.1 SAFETY VALVES - SHUTDOWN . . . . . . . . . . . . .3/4 4-4 3/4.4.2.2 SAFETY VALVES - OPERATING . . . . . . . . . . . . .3/4 4-4a 3/4.4.3 RELIEF VALVES . . . . . . . . . . . . . . . . . . .3/4 4-5 3/4.4.4 PRESSURIZER . . . . . . . . . . . . . . . . . . . .3/4 4-6 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY . . . . . . . .3/4 4-7 I 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection System . . . . . . . . . . . .3/4 4-14 Operational Leakage . . . . . . . . . . . . . . .3/4 4-15 Primary Coolant System Pressure Isolation Valve s .3/4 4-16a 3/4.4.7 DELETED 3/4.4.8 SPECIFIC ACTIVITY . . . . . . . . . . . . . . . . .3/4 4-20 3/4.4.9 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System . . . . . . . . . . . . .. 3/4 4-24 Pressurizer . . . . . . . . . . . . . . . . . . .. 3/4 4-29 Overpressure Protection Systems . . . . . . . . .. 3/4 4-30 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2, and 3 Components . . . . . .3/4 4-32 3/4.4.11 INTENTIONALLY BLANK . . . . . . . . . . . . . . . .3/4 4-34 3/4.4.12 HEAD VENTS. . . . . . . . . . . . . . . . . . . . .3/4 4-35 SALEM - UNIT I v Amendment No. 225

INDEX BASES

== == === == === == == === == === == === == == === == === == === == ==

SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE AND 3/4.3.2 ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION . . . . . . . . . . . . . . . B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION . . . . . . . . . . B 3/4 3-la 3/4.3.4 TURBINE OVERSPEED PROTECTION . . . . . . . . . B 3/4 3-4 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION . . . . . . . . . . . . 3/4 4-1 .B 3/4.4.2 SAFETY VALVES . . . . . . . . . . . B 3/4 4-la ..

3/4.4.3 RELIEF VALVES . . . . . . . . . . . 3/4 4-la .B 3/4.4.4 PRESSURIZER . . . . . . . . . . . . B 3/4 4-2 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE B 3/4 4-3 3/4.4.7 DELETED 3/4.4.8 SPECIFIC ACTIVITY . . . . . . . . . B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS . . . . B 3/4 4-6 3/4 .4 .10 STRUCTURAL INTEGRITY . . . . . . . 3/4 4-17 .B 3/4 .4.11 BLANK . . . . . . . . . . . . . . . 3/4 4-17 .B 3/4 .4. 12 REACTOR VESSEL HEAD VENTS . . . . . 3/4 4-17 .B SALEM - UNIT 1 XII Amendment No. 225

DEFINITIONS

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage).

MEMBER(S) OF THE PUBLIC 1.16 MEMBER(S) OF THE PUBLIC shall be all those persons who are not occupationally associated with the plant. This category does not include employees of PSE&G, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent controls and Radiological Environmental Monitoring programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8 respectively.

OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified functionts), and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

SALEM - UNIT I 1-4 Amendment No. 234

DEFINITIONS PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the Updated FSAR, 2) authorized under the provisions of 10CFR50.59, or 3) otherwise by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary Steam generater tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of radioactive waste.

PURGE - PURGING 1.23 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3459 MWt.

SALEM - UNIT I 1-5 Amendment No. 243

REACTOR COOLANT SYSTEM STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 SG tube integrity Each steam generator shall be GPERABT3 maintained and all SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

'ith one or more steam generators inoperable, restore the inoperable generator(s) t 0P w LE status prior to increasing Tavg above 2000F./

SuRV ACE REQUIREMENTS/

4.4.5.0 < h steam generator shall be demonstrated OPERABLE by perfo of the following aug nted inservice inspection program and the requirements X Specification 4 .5.

4.4.5.1 Steam Gen ator Sample Selection and Inspection - Eac steam generator shall be determined 0 RABLE during shutdown by selecting and nspecting at least the minimum number of st m generators specified in Table 4 -1.

4.4.5.2 Steam Generator Tu Sample Selection and Ins ction - The steam generator tube minimum sample size, insp tion result classific ion, and the corresponding action required shall be as spec ied in Table 4.4- . The inservice inspection of steam generator tubes shall be per rmed at the equencies specified in Specification 4.4.5.3 and the inspec d tubes all be verified acceptable per the acceptance criteria of Specification 4 .5. 4 The tubes selected for each inservice inspection shall include at least 3% of t total number of tubes in all steam generators; the tubes selected for these n ections shall be selected on a random basis except:

a. Where experience in simi r plants with milar water chemistry indicates critical areas to be i pected, then at lea t 50% of the tubes inspected all be from these criti 1 areas.

b. The first inserv e inspection (subsequent to the reservice inspection) of each steam gen ator shall include:

1. All n plugged tubes that previously had detectabl all penetrations

(>2/ ), and\

2. ubes in those areas where experience has indicated poten a1 problems.

A tube inspection (pursuant to Specification 4.4.5.4.a.8) shaX be performed on each selected tube. If any selected tube does not rmit the passage of the eddy current probe for a tube inspection, this all be recorded and an adjacent tube shall be selected and subjected to tube inspection.

SALEM - UNIT I 3/4 4-7 Amendment No. 118

INSERT I a.* With one or more SG tubes satisfying the tube repair criteria and not plugged in accordance with the Steam Generator Program:

1. Verify tube integrity of the affected tube(s) is maintained until the next inspection within 7 days; and

2. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or SG tube inspection.

b. With SG tube integrity not maintained or the required Action and associated Completion Time of a. above not met, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with the Steam Generator Program.

4.4.5.2 Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.

Separate Action and Completion Time is allowed for each SG tube.

XEACTOR COOLANT SYSTEM S VEILLANCE REQUIREMENTS (Continued) c.

\ The tubes selected as the second and third samples (if requi ed by Table 4.4-2) during each inservice inspection may be subje ed to a partial tube inspection provided:

\1. The tubes selected for these samples include the t bes from those areas of the tube sheet array where tubes with i erfections were previously found.

2. The inspections include those portions of th tubes where mperfections were previously found.

The results of each ample inspection shall be classifie into one of the following three categories:

Category Inspection Results C-1 Less than % of the total tubes i pected are degraded tubes and none of the inspected tubes are efective.

C-2 One or more t es, but not mo than 1% of the total tubes inspected are d fective, or etween 5% and 10% of the total tubes inspected are de aded tube C-3 More than 10% of th tot I tubes inspected are degraded tubes or more than 1% of the l s ected tubes are defective.

Note: In all inspections, prev s y degraded tubes must exhibit significant

(>10%) further wall pen trati s to be included in the above percentage calculations.

4.4.5.3 Inspection Frequencies/- The above r quired inservice inspections of steam generator tubes shall be perfo med at the foll ing frequencies:

a. The first inser ice inspection shall performed after 6 Effective Full Power Mo hs but within 24 calenda months of initial criticality.

Subsequent i service inspections shall b performed at intervals of not less than 1 nor more than 24 calendar mo G hs after the previous inspectio. If two consecutive inspections ollowing service under AVT conditio s, not including the preservice ins ection, result in all inspect on results falling into the C-1 categ y or if two consecutive inspe ions demonstrate that previously observe degradation has not cont'nued and no additional degradation has occu ed, the inspection it rval may be extended to a maximum of once per 0 months.

b. f the results of the inservice inspection of a stea generator conducted in accordance with Table 4.4-2 at 40 month i tervals fall in Category C-3, the inspection frequency shall be increas d to at least once per 20 months. The increase in inspection frequency hall apply until the subsequent inspections satisfy the criteria of S ecification 4.4.5.3.a; the interval may then be extended to a maximum o once per 40 months.

S EM - UNIT 1 3/4 4-8 Amendment No. 5

REACTOR COOLANT SYSTEM S XVEILLANCE REQUIREMENTS (Continued)

Additional, unscheduled inservice inspections shall be performed/on each steam generator in accordance with the first sample inspec ion pecified in Table 4.4-2 during the shutdown subsequent to an of the llowing conditions.

1. Primary-to-secondary tubes leaks (not including leaks riginating om tube-to-tube sheet welds) in excess of the limi of Sp cification 3.4.6.2,

2. A se mic occurrence greater than the Operating sis Earthquake,

3. A loss- -coolant accident requiring actuation/of the engineered safeguar , or

4. A main stea line or feedwater line break.

4.4.5.4 Acceptance Crite a

a. As used in this Spe fication:

1. Imperfection means n exception to he dimensions, finish or contour of a tube fr that requi ed by fabrication drawings or specifications. Eddy- rrent tes ing indications below 20% of the nominal tube wall thick ess, i detectable, may be considered as imperfections.

2. Degradation means a servic induced cracking, wastage, wear or general corrosion occurri g n either inside or outside of a tube.

3. Degraded Tube means a ube contaning imperfections 0 20% of the nominal wall thickne s caused by egradation.

4. % Degradation mea the percentage o the tube wall thickness affected or remo ed by degradation.

5. Defect means imperfection of such sev rity that it exceeds the plugging lim't. A tube containing a defec is defective.

6. Pluqqinq mit means the imperfection depth t or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness.

7. Unse iceable describes the condition of a tube i it leaks or con ains a defect large enough to affect its struc ural integrity i the event of an Operating Basis Earthquake, a lo -of-coolant cident, or a steam line or feedwater line break as pecified in 4.4.5.3.c, above.

Tube Inspection means an inspection of the steam generato tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

- UNIT I 3/4 4-9 Amendment No.

REACTOR COOLANT SYSTEM SUR EILLANCE REQUIREMENTS (Continued)

9. Preservice Inspection means an inspection of the full lengt of each tube in each steam generator performed by eddy curren techniques prior to service establish a baseline conditio of the tubing. This inspection shall be performed after t field hydrostatic test and prior to initial POWER OPERATION sing the equipment and techniques expected to be used during s bsequent service inspections.

b. The ste m generator shall be determined OPERABLE aft completing the corr sponding actions (plug all tubes exceeding the plugging limit and 11 tubes containing through-wall crack required by Table 4.4-4.4.5.5 Reports

a. Following each i service inspection of stea generator tubes, the number of tubes p ugged in each steam gen ator shall be reported to the Commission with*n 15 days.

b. The complete results the steam gen ator tube inservice inspection shall be incuded in the nual Operating Report for the period in which the insp ction was ompleted. This report shall include:

1. Number and extent of tub s nspected.

2. Location and percent of A thickness penetration for each indication of an imper ctio

3. Identification of t es plugge

c. Results of steam gene ator tube inspe tions which fall into Category C-3 shall evaluated for reprtability pursuant to 10CFR50.72 and 10C 50.73. The evaluati n shall be documented, and shall provide a scription of investigat ons conducted to determine cause of the tu degradation and correcti measures taken to prevent recurr nce.

- UNIT I 3/4 4-10 Amendment 133

THIS PAGE LEFT INTENTIONALLY BLANK

- UNIT I 3/4 4 -11 Amendment 118

TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice spection No ' Yes No. of Steam GŽ>ators per Unit Two Three Four T Three Four First Inservice Insp a ion All One Two Two Second & Subsequent Inse *ce Inspections One One One One3 Tabl e Notation:

1. The inservice inspection may be limite to one stea generator on a rotating schedule encompassing 3 N% of the tubes (where N is the number stea enerators in the plant) if the results of the first or previous inspections indicate that steam generators are performing in a like manner.

Note that under some circumstances, the ope t conditions in one or more steam generators may be found to be more severe than those in ot r steam enerators. Under such circumstances the sample sequence shall be modified to inspect e most sever conditions.

2. The other steam generator not in ected during the first a ervice inspection shall be inspected.

The third and subsequent insp ions should follow the instr tions described in 1 above.

3. Each of the other two st X generators not inspected during the fi t inservice inspections shall be inspected during t9 second and third inspections. The fourth an ubsequent inspections shall follow the instruc *ons described in 1 above.

UNIT I 3/4 4-12

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE ,SPECTION Action Required Result Action Required Result Ac on Required A minimum of None N/A N/A N/A N/A S Tubes per S.G.

I II N C-1 None N/A A

C-1 None C-2 Plug defectiv C-2 Plug defective tubes tubes and in ect additional S tubes in this S C-3 Perform action for C-3 result of first sample

\ -3 rform action for N/A N/A

/C-3 result of first sample C-3 Inspect all tubes All oth in this S.G., plug S.G.s re defective tubes and C- None N/A N/A inspect 2S tubes in each other S.G.

/ome S.G.s rfo action for

'1 C-2 but no additional C-2 resu t of second sa le N/A N/A S.G. are C-3 I - I I Additional Inspect all tubes S.G. is in each S.G. and \

C-3 plug defective N/A N/A tubes. Notification to NRC pursuant to 10CFR50.72 and 10CFR50.73, as applicable.

S = 3 -N Where N is the number of steam generators in the unit, and n is the numbe of n

steam generators inspected during an inspection Altern e action may be used in accordance with paragraph (see next page)

SAL - UNIT 1 3/4 4-13 Amendment N 133

Alternate Action During the thir refueling outage surveillance exa nations, indications associated with w 11 reduction were detected on t e tubing of No. 12 and No. 14 Steam Gener ors. The condition was esta ished as occurring on the outside diamete of tubes located around t periphery of the tube sheet on the cold leg side at the intersectio s of the first, second, and third support plates.

During the third refuelin outage the fol owing action may be taken in place of that required by ble 4.4-2 w n the results of the initial sample require that an addi onal samp e or samples must be inspected and the condition for which the a ded ins ection is required is limited to the peripheral tubes:

1. The second inspection sa le shall, as a minimum, include an area of 5 tubes inwardm sured from the innermost defective or degraded tube and in ar completely around that side of the steam generator (col leg or hot leg) where the initial indications are fo d.

2. Subsequent sampl s shall be take in a similar manner if degraded or de ctive tubes are f und in the second sample inspection.

EM - UNIT 1 3/4 4-13a Amendmen o. 43

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. 1 CPH total primary to secondary lea'ag: through all steoe generatoro and 150G gallons per day primary-to-secondary leakage through any one steam generator,

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

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

a. With any PRESSURE BOUNDARY LEAKAGE or primary-to-secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

SURVEILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by;

a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Monitoring the containment sump inventory at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SALEM - UNIT 1 3/4 4-15 Amendment No. 178

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) c*. 1-T USED Verifying primary-to-secondaryleakage is

150 gallons per day through any one steam generator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation, d*. Performance of a Reactor Coolant System water inventory balance** at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The water inventory balance shall be performed with the plant at steady state conditions. The provisions of specification 4.0.4 are not applicable for entry into Mode 4, and

e. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Not required to be completed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

- Not applicable to primary-to-secondary leakage. I SALEM - UNIT I 3/4 4-16 Amendment No. 178

ADMINISTRATIVE CONTROLS

7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1,

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iondine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

10) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

6.8.4.h Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,

2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of the census, and

3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.8.4.i Steam Generator (SG) Program l INSERT 2 l SALEM - UNIT l 6-19b Amendment No. 234 SAE NT161bImnmn o 3

INSERT 2 A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the 'as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down 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 primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. Inthe assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any 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 gallon per minute per SG.

3. The operational leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld and tube end are not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 120, 90, and thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary-to-secondary leakage.

ADMINISTRATIVE CONTROLS

2. WCAP-8385, Power Distribution Control and Load Following Procedures -

Topical Report, September 1974 (W Proprietary) Methodology for Specification 3/4.2.1 Axial Flux Difference. Approved by Safety Evaluation dated January 31, 1978.

3. WCAP-10054-P-A, Rev. 1, Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP Code, August 1985 (W Proprietary), Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved for Salem by NRC letter dated August 25, 1993.

4. WCAP-10266-P-A, Rev. 2, The 1981 Version of Westinghouse Evaluation Model Using BASH Code, Rev. 2. March 1987 (W Proprietary) Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved by Safety Evaluation dated November 13, 1986.

5. WCAP-12472-P-A, BEACON - Core Monitoring and Operations Support System, Revision 0, (W Proprietary). Approved February 1994.

6. CENPD-397-P-A, Rev. 1, Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology, May 2000.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

_ ISERT 4 SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with a copy to the Administrator, USNRC Region I within the time period specified for each report.

6.9.3 Violations of the requirements of the fire protection program described in the Updated Final Safety Analysis Report which would have adversely affected the ability to achieve and maintain safe shutdown in the event of a fire shall be submitted to the U. S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the Regional Administrator of the Regional Office of the NRC via the Licensee Event Report System within 30 days.

6.9.4 When a report is required by ACTION 8 or 9 of Table 3.3-11 "Accident Monitoring Instrumentation", a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring for inadequate core cooling, the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.

SALEM - UNIT 1 6-24a Amendment No. 243

INSERT 4 6.9.1.10 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 6.8.4.1, "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. The effective plugging percentage for all plugging in each SG.

ATTACHMENT 3 LCR S04-07 LR-N05-0018 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by PSEG in this document. Any other statements in this submittal are provided for information only purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Mr. Courtney Smyth at (856) 339-5298.

1 Regulatory Commitment Due DatelEvent PSEG will implement the Steam Generator Program in Concurrent with accordance with NEI 97-06, "Steam Generator Program implementation of Guidelines" the amendment ,

ATTACHMENT 4 LCR S04-07 LR-N05-001 8 PROPOSED CHANGES TO TS BASES PAGES The following Technical Specifications Bases for Salem Unit 1, Facility Operating License No. DPR-70, are affected by this change request:

Salem Unit 1 Technical SDecification Page Bases 3/4.4.5 B 3/4 4-2 and B 3/4 4-3 Bases 3/4.4.6.2 B 3/4 4-4

REACTOR COOLANT SYSTEM BASES 3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance is sufficient to assure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE assures that the plant will be able to establish natural circulation.

3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY The Surveillance Requirements for inspection of the steam generator tubes enNre that the structural integrity of this portion of the RCS will be maint *ned.

The p ram for inservice inspection of steam generator tubes is based on modifica n of Regulatory Guide 1.83, Revision 1. Inservice inspectio f steam generator t *ng is essential in order to maintain surveillance of conditions of the tubes in e event that there is evidence of mechanical d age or progressive degrad on due to design, manufacturing errors, inservice conditions that lead t corrosion. Inservice inspection of team generator tubing also provides a means of aracterizing the nature and use of any tube degradation so that correcti measures can be take The plant is expected to be o ated in anner such that the secondary coolant will be maintained within thos c stry limits found to result in negligible corrosion of the steam gene tubes. If the secondary coolant chemistry is not maintained within se limi , localized corrosion may likely result in stress corrosion crack . The extent X cracking during plant operation would be limited by the limit ion of steam generat tube leakage between the primary coolant system and e secondary coolant syste rimary-to-secondary leakage = 500 gallons p day per steam generator). Cracks aving a primary-to-secondary leakage 1e than this limit during operation will ye an adequate margin of safety withstand the loads imposed during normal ope tion and by postulated ac c ents. Operating plants have demonstrated that prima -to-secondary leakage of 0 gallons per day per steam generator can readily be detec d by radiati monitors of steam generator blowdown. Leakage in excess of this it will -equire plant shutdown and an unscheduled inspection, during which the 1 ing t es will be located and plugged.

I ZNSERT3 l SALEM - UNIT 1 B 3/4 4-2 Amendment No. 39

REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY (Continued) tazage-type defects are unlikely with proper chemistry treatment of the secondary c nt. However, even if a defect should develop in service will be found during sch nservice steam generator tube examinati lugging will be required for all tube imperfections exceeding ugging limit of 40% of the tube nominal wall thickness generat oe inspections of operating plants have demonstrated the capabilit iably detect degradation that has penetrated 20% of the original ll thickn Whenever the of any steam generator tubing inser ection fall into Catse~y C-3, these results will be evaluated for reportability csn_~ssion pursuant to the applicable sections of 10CFR50.72 and IOCFR50.73-.-

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

SALEM - UNIT I B 3/4 4-3 Amendment No. 133

INSERT 3 The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld and tube end are not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.8.4.i, "Steam Generator (SG) Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, 'The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that significantly affect burst or collapse. Inthat context, the term "significantly" is defined as, "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." The determination of whether thermal loads are primary or secondary loads is based on the ASME definition in which secondary loads are self-limiting and will not cause failure under single load application. For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code, Section 111, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB.

The accident induced leakage performance criterion ensures that the primary-to-secondary leakage caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 1 gpm per SG. The accident induced leakage rate includes any primary-to-secondary leakage existing prior to the accident in addition to primary-to-secondary leakage induced during the accident.

The ACTION requirements are modified by a Note clarifying that the Actions and allowed outage times may be entered independently for each SG tube. This is acceptable because the ACTION requirements provide appropriate compensatory actions for each affected SG tube.

Complying with the ACTION requirements may allow for continued operation, and subsequent affected SG tubes are governed by subsequent ACTION requirements and allowed outage times.

If it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program, an evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. An allowed outage time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity. If the evaluation determines that the affected tube(s) have tube integrity, plant operation is allowed to continue until the next SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to entering HOT SHUTDOWN following the next refueling outage or SG inspection. This allowed outage time is acceptable since operation until the next inspection is supported by the operational assessment.

If SG tube integrity is not being maintained or the ACTION requirements and associated allowed outage times of ACTION requirements are not met, the reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed outage times are reasonable based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period. The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations.

The Steam Generator Program also specifies the inspection methods to be used to find existing and potential degradation. Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities and inspection locations. The Frequency is determined by the operational assessment and other limits inthe SG examination guidelines. The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.8.4.A contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 6.8.4.i are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in size measurement and future growth.

In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). The Frequency of prior to entering HOT SHUTDOWN following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary-to-secondary pressure differential.

REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

The total steam generator tube leakage limit of 1 GPM for all ste erators (but not more gpd for any steam generator) ensures that the d contribution from the tube leakage will ed to a small fraction of Part 10g +/-ts in the event of either a steam generator tube ruptur team line break. Th M limit is consistent with the assumptions used in the analy idents. The 500 gpd leakage limit per steam generator ensures that steam integrity is maintained in the event of a main steam line rupture or LOCA conditions.

OUNDARY LEAKAGE of any magnitude is unaccept it may be indicative of

+/-mpending gross failure of the pressure boundary. Therefore, the p any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

The primary-to-secondary leakage rate limit applies to leakage through any one steam generator. The limit of 150 gallons per day per steam generator is based on the operational leakage performance criterion in NEI 97-06, Steam Generator Program Guidelines. The Steam Generator Program operational leakage performance criterion in WEI 97-06 states, "The RCS operational primary-to-secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with steam generator tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Proqram is an effective measure for minimizing the frequency of steam generator tube ruptures. Unidentified leakage or identified leakage in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This allowed outage time allows time to verify leakage rates and either identif, unidentified leakage or reduce leakage to within limits before the reactor must be shut down.

This action is necessary to prevent further deterioration of the reactor coolant pressure boundary (RCPB). If any pressure boundary leakage exists, or primary-to-secondaryleakage is not within limit, or if unidentified or identified leakage cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the leakage and its potential consequences. It should be noted that leakage past seals and gaskets is not pressure boundary leakage. The reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reduces the leakage and also reduces the factors that tend to degrade the pressure boundary. The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In COLD SHUTDOWN, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely. Satisfying the primary-to-secondary leakage limit ensures that the operational leakage performance criterion in the Steam Generator Program is met. The 150 gallons per day limit is measured at room temperature (in accordance with EPRI PWR Primary-to-SecondaryLeak Guidelines). If it is not practical to assign the leakage to an individual steam generator, all the primary-to-secondary leakage should be conservatively assumed to be from one steam generator. The Surveillance is modified by a Note which states that the surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. For RCS primary-to-secondaryleakage determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and Reactor Coolant Pump seal injection and return flows. The Surveillance Frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary-to-secondarv leakage and recognizes the importance of early leakage detection in the prevention of accidents. The primary-to-secondaryleakage is determined using continuous process radiation monitors or radiochemical grab sampling.

3/4.4.7 THIS SECTION DELETED SALEM - UNIT 1 B 3/4 4-4 Amendment No. 180

LR-N05-0018, Request for Change to Technical Specifications to Add Requirements for Steam Generator Tube Integrity, Steam Generator Program, & Steam Generator Tube Inspection Report & to Revise Reactor Coolant System Operational.

Text

Navigation menu

LR-N05-0018, Request for Change to Technical Specifications to Add Requirements for Steam Generator Tube Integrity, Steam Generator Program, & Steam Generator Tube Inspection Report & to Revise Reactor Coolant System Operational.

Text

Navigation menu

Search