ML031340757

Request for Additional Information - Catawba, Unit 1 and 2 - Tech Spec Amendment - Steam Generators

ML031340757
Person / Time
Site:	Catawba
Issue date:	04/24/2003
From:	NRC/NRR/DLPM
To:
References
TAC MB7842, TAC MB7843
Download: ML031340757 (22)

Text

- brcs-+ - qq~/24 ,3 REQUEST FOR ADDITIONAL INFORMATION CATAWBA UNITS 1 AND 2 TECHNICAL SPECIFICATION AMENDMENT - STEAM GENERATORS 3.4.13 RCS Operational Leakage The licensee proposes to delete the 576 gallon per day leakage limit which is applicable to total primary to secondary leakage from all steam generators. Thus, the proposed specification would permit total primary to secondary leakage from all steam generators to be 600 gpd. If the safety analysis is based on a total primary to secondary leak rate of 576 gallons per day, what is the justification for allowing total operation leakage to be 600 gallons per day?

B 3.4.13 RCS Operational Leakage (BASES)

1. The licensee proposes to delete a sentence which reads:

"The volumetric calculation of primary to secondary LEAKAGE is based on a density at operating RCS temperature of 585 degrees F."

The licensee proposes to add the following statement (Insert B):

"The primary to secondary LEAKAGE measurement is based on the methodology described in Ref. 5. Currently, a correction factor is applied to account for the fact that current safety analyses take the primary to secondary leak rate at reactor coolant conditions, rather than at room temperature as described in Ref. 5."

The licensee also proposes to add the following statement (Insert D).

"The 150 gallons per day limit is based on room temperature measurements."

The statement in Insert D appears to contradict the statement in Insert B. The licensee should resolve this discrepancy. Why should the 150 gallon per day limit not be based on reactor coolant conditions assumed in the current safety analyses?

2. The first sentence of Insert B needs to be clarified as follows:

The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, "Steam Generator Program Guidelines" (Ref. 6).

The Steam Generator Program operational leakage performance criterion in NEI 97-06 states: "The RCS operational primary to secondary LEAKAGE through any one SG shall be limited to 150 gallons per day."

3. The second sentence of paragraph 4 of Insert B oversells the case based on operating experience. The staff believes the following to be a more defensible position:

"The operational leakage rate criterion in conjunction with implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures."

Attachment

3.4.18 Steam Generator (SG) Tube Integrity

1. The licensee is proposing a LCO which would require, in part, that "all SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the SG Program. This particular LCO was not part of the GLCP LCO proposal. Why is it needed? Why is not proposed SR 3.4.18.2 sufficient to require entry into the proposed action statement if not all tubes in excess of the repair limit are plugged?

2. The proposed LCO, Actions - Condition A, and SR 3.4.18. 2 create confusion by referring to plugging tubes which satisfy the tube repair criteria. This is contrary to conventional usage whereby satisfying an acceptance limit implies an acceptable condition. One better approach is to replace the word "satisfying" with the words "failing to satisfy."

3. The words "or repaired" should be deleted from the LCO, Actions - Condition A, and SR 3.4.18.2

4. Required Action A.1 uses words similar to SR 3.4.18.1 thereby creating confusion. The BASES makes it clear that "verify" in SR 3.4.18.1 refers to condition monitoring to be performed during an inspection to confirm that tube integrity existed up to that time. In contrast, "verify" in A.1 refers to a forward looking analytical assessment to verify that tube integrity will be maintained until the next inspection. We suggest that Action A.1 be clarified to as follows" "Perform assessment to verify tube integrity of the affected tube(s) will be maintained until Required Action A.2 is completed."

5. References to the "SG Program" should be revised to "Specification 5.5.9, "SG Program"."

6. An NRC notification requirement should be added if the licensee fails to plug a tube which fails to satisfy the applicable repair criteria.

7. Suppose license inadvertently fails to plug tube which does not satisfy the plugging criteria and plant is restarted in January. Suppose condition is discovered on February 1. Suppose license implements Required Action A.1 and determines on February 8 that the subject tube will continue to meet all performance criteria only until July 15, well short of the next scheduled refueling outage inspection commencing December 31. Does this mean that the plant is in Condition B on February 8? On July 15? Does this mean that the plant does not need to be in Mode 5 until 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after July 15? Revisions to this specification are necessary to ensure the intent of the specification is clear.

B 3.4.18 Steam Generator Tube Integrity (BASES)

The attached markup provides suggested changes which are of a clarifying nature. We have not commented on the BASES pertaining to the structural performance criteria until the final form of the criteria is settled.

Inserts referred to in the markup are provided below:

Insert B Specification 5.5.9 requires that a steam generator program be established and implemented to ensure that steam generator tube integrity is maintained. Pursuant to specification 5.5.9, tube integrity is maintained when the tube integrity performance criteria are met.

Insert C Specification 5.5.9 requires that a steam generator program be established and implemented to ensure that steam generator tube integrity is maintained. Pursuant to specification 5.5.9, tube integrity is maintained when the tube integrity performance criteria are met.

Specification 5.5.9 defines the minimum regulatory requirements for establishing an implementing the SG Program. These minimum requirements are generally performance based, with some prescriptive requirements to ensure that the performance criteria are met.

NEI 97-06, "Steam Generator Program Guidelines," provides guidelines for programmatic elements of the SG Program for ensuring the tube integrity performance criteria are met and references additional industry guidelines concerning details of these programmatic elements.

Compliance with the LCO during MODES 1 through 4 is determined by verifying that SG tube integrity is maintained in accordance with the SG Program. As part of the SG Program, specification 5.5.9 requires that the condition of the tubes be assessed during each outage during which the steam generators tubes are inspected or plugged to confirm that the performance criteria were met during the previous period of operation. In addition, an operational assessment is performed consistent with guidance in NEI 97-06 to ensure that the performance criteria will continue to be met until the next scheduled inspection.

Insert D Specification 5.5.9 defines the minimum regulatory requirements for establishing an implementing the SG Program. These minimum requirements are generally performance based, with some prescriptive requirements, including specified maximum tube inspection intervals and specified tube repair criteria, to ensure that the performance criteria are met.

Insert E In accordance with specification 5.5.9, the scope and methods of inspection are performed such as to ensure reliable detection of any flaws that are present along the length of the tube, from tube end weld location (hot) to tube end weld location (cold) that may exceed the applicable tube repair criteria. In addition, the scope, method, and frequency of inspection are such as to ensure that steam generator tube integrity is maintained.

Insert F Reference land its referenced EPRI guidelines provide detailed guidelines for defining inspection scope, methods, and frequency consistent with the objectives in specification 5.5.9.

Consistent with these guidelines, the licensee will perform operational assessments to establish that the program it has implemented reliably ensures that the performance criteria will continue to be met prior to the next scheduled inspection. In addition specification 5.5.9 contains a number of prescriptive restrictions on inspection frequency to provide added assurance that the

tube integrity performance criteria will be met between scheduled inspections.

Insert G The tube repair criteria specified in specification 5.5.9 are intended to ensure that tubes accepted for continued service satisfy the tube integrity performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria must ensure that the tube integrity performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 and the referenced EPRI guidelines provide guidelines for performing operational assessments to verify that the tubes remaining in service will continue to meet the performance criteria.

Specifications 5.5.9. "Steam Generator (SG) Program" and 5.5.10. "Steam Generator (SG)

Tube Inspection Report"

1. Editorial comment (discussed at March 27, 2003 meeting): Proposed specification 5.5.9 could be mis-construed to mean that an adequate SG Program need only consist of provisions for defining the performance criteria, conducting condition monitoring, and implementing tube repair criteria, tube repair methods, and inspection intervals not to exceed specified maximums.

The staff recommends revised wording and format, as shown in the attached sample technical specifications (consistent with that presented at March 27, 2002 meeting), to clarify that the keystone requirement is to establish and implement a program which ensures tube integrity will be maintained. The staff notes that specification 5.5.9 is not intended to be a cookbook defining all programmatic elements to be included in the program or the details of these elements. The intent of the specification is that licensee's should have the flexibility in establishing and implementing the details of the program as necessary to ensure tube integrity is maintained. Thus, only a few details of the program, of both a prescriptive and performance based nature, are included in the specification which are critical to ensuring that the SG Program will not significantly increase risk over that associated with the implementation of current requirements.

2. Structural integrity performance criteria: Staff has previously issued RAI pertaining to this issue.

3. Maximum inspection interval issue (discussed at March 27 meeting): The licensee has proposed some prescriptive criteria. The staff recommends some clarification of these prescriptive criteria as shown in the attached sample technical specification (consistent with that presented at March 27, 2002 meeting).

4. Tube repair criteria (discussed at March 27 meeting): It is the staff's understanding that the licensee is revising its amendment request to delete proposed requirements relating to tube repairs. This is acceptable to the staff.

5. Emergent issues (discussed at March 27 meeting): As discussed at the March 27 meeting, recent experiences have highlighted the importance of ensuring that staff expectations concerning the minimum actions necessary to ensure tube integrity are clearly spelled out in the technical specifications. Accordingly, at the March 27, 2003 meeting the staff identified additional changes that are needed to ensure, in conjunction with 10 CFR 50, Appendix B, that (1) tube inspection scope and methods are implemented that ensure that all flaws that exceed

the tube repair criteria are reliabty detected and (2) tube inspection scope, inspection methods, and repair criteria are implemented such as to ensure that SG tube integrity is maintained.

These additional changes identified by the staff are shown in the attached sample technical specifications (consistent with that presented at March 27, 2002 meeting).

6. Reporting Requirements: The proposed SG reporting requirements in specification 5.5.10 should apply to all plants, irrespective of the number of tubes found to exceed the tube repair limit. This change is illustrated in the attached sample technical specifications (consistent with that presented at March 27, 2002 meeting). The staff is requesting this change since the proposed technical specification change package for Catawba represents a significant departure from the existing technical specifications in that it will be significantly more performance based than current technical specifications. The staff believes it important to continue to have information pertaining to SG inspection results to allow it to continue to monitor operating experience trends and how these trends may be affected by the new technical specification requirements.

f'AŽLAJT.~ 62ses 6B Lf g SG Tube Integrity B 3.4.18 B 3.4 REACTOR COOLANT SYSTEM (RC$)

B 3.4.18 Steam Generator (SG) Tube Integrity BASES BACKGROUND SG tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers in pressurized water reactors (PWRs). In the context of this Specification, tubing is defined as:

'Steam generator tubing refers to the entire length of the tube, including the tube wall and any repairs made 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."

The SG tubes 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. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, 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. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, "RCS Loops - MODES 1 and 2," LCO 3.4.5, "RCS Loops -

MODE 3,- LCO 3.4.6, RCS Loops - MODE 4," and LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled.'

[I-t i Concerns relating to the integrity of SG tubing stem from the fact rII0 F) that the tubing is subject to a variety of degradation mechanisms.

Throughout the industry, SG tubes have experienced 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. 4-means f-detepmir§ -

fiRdana egradatie4s-neede--@perfermanee-erter-were

-develped-forithis-purpose-j- L 62 The SG performance criteria identify the standards against which performance is to be measured. Meeting the performance criteria Catawba Units 1 and 2 B 3.4.18-1 Revision No. 0

SG Tube Integrity B 3.4.18 BASES BACKGROUND (continued)

..s provides reasonable assurance that the SG tubing remains capable of fulfilling its specific safety function of maintaining RCPB integrity. The-SG-performance-criteria-and-the-processes-required

-to-meet-them are-defined by the NEl-Steam-GeneratorProgram

-Guidelines-(Refk-- '

There are three SG performance criteria: accident induced leakage, structural integrity, and operational LEAKAGE. They act together to provide reasonable assurance of tube integrity at normal and accident conditionsBWGtU6Infegrifmiiiens that the tuDes are capable of performing their intended safety functions A consistent with their licensing basis, including applicable Caw L ivi cj regulatory requirements. _

The purpose of this LCO is to require compliance with the SG performance criteria. The accident induced leakage and structural integrity performance criteria apply to SG tubes and associated appurtenances considered part of the SG primary to secondary pressure boundary (e.g., plugs, sleeves, and other repairs). The accident induced leakage and structural integrity performance criteria are documented in Specification 5.5.9.

The third performance criterion, operational LEAKAGE, is addressed by LCO 3.4.13, RCS Operational LEAKAGE."

APPLICABLE Satisfying the SG structural integrity performance criterion, SAFETY ANALYSES provides reasonable assurance against tube burst and the resulting primary to secondary LEAKAGE that might occur at normal and accident conditions.

Satisfying the accident induced leakage performance criterion provides reasonable assurance of acceptable primary to secondary LEAKAGE that might occur as a result of design basis accident conditions other than a SG tube rupture. The consequences of design basis accidents that include primary to secondary LEAKAGE depend, in part, on the accident induced leakage and the radioactive source term in the primary coolant.

The design basis accidents for which the primary to secondary 'C,-

LEAKAGE is a pathway for release of activity to the environment include the main steam line break, SG tube rupture, reactor coolant pump locked rotor accident, single rod withdrawal accident, and rod ejection accident. The analysis of radiological consequences of these design basis accidents, except for a SG 3 t&

tube rupture, assumes that the total primary to secondary Catawba Units 1 and 2 B 3.4.18-2 Revision No. 0

SG Tube Integrity B 3.4.18 BASES APPLICABLE SAFETY ANALYSES (continued)

LEAKAGE from each SG initially is 150 gallons per day. Transient thermal hydraulic analyses of these design basis accidents determine the primary to secondary LEAKAGE changes (decreases or increases) that result from changing pressures and temperatures. These calculated values are used in the analyses of radiological consequences of these design basis accidents.

The source term in the primary coolant for some design basis accidents (e.g., reactor coolant pump locked rotor accident and rod ejection accident) is associated primarily with fuel rods calculated to be breached. For other design basis accidents (e.g.,

main steam line break and SG tube rupture), the source term in the primary coolant consists primarily of the levels of Dose Equivalent 11l1 radioactivity levels calculated for the design basis accident. This, in turn, is based on the limiting values in the Technical Specifications and postulated iodine spikes.

For accidents in which the source term in the primary coolant consists of the Dose Equivalent 1131 activity levels, the SG tube rupture yields the limiting values for radiation doses at offsite locations. In the calculation of radiation doses following this event, the rate of primary to secondary LEAKAGE in the intact SGs is set equal to the.operational LEAKAGE rate limits in LCO 3.4.13. For the ruptured SG, a double ended rupture of a single tube is assumed. Following the initiating event, contaminants in flashed and atomized break flow (the latter computed for time spans during which the tubes are calculated to be uncovered), as well as secondary coolant, may be released to the atmosphere.

Before reactor trip, the accident analysis for the SG tube rupture assumes that these contaminants are released to the condenser and from there to the environment with credit taken for scrubbing of iodine contaminants in the condenser. Following reactor trip (and loss of offsite power), the accident analysis assumes that these contaminants are released to the environment through the SG power operated relief valves and the main steam code safety valves until such time as the closure of these valves can be credited.

For other design basis accidents such as main steam line break, rod ejection accident, reactor coolant pump locked rotor accident, and uncontrolled rod withdrawal accident, the tubes are assumed to retain their structural integrity (i.e., they are assumed not to rupture). The LEAKAGE is assumed to be initially at the limit given in LCO 3.4.13. This is consistent with the accident induced leakage performance criterion.

Catawba Units 1 and 2 B 3.4.18-3 Revision No.

SG Tube Integrity B 3.4.1 8 BASES APPLICABLE SAFETY-ANALYSES (continued)

The three SG performance criteria and the limits included in the plant Technical Specifications for Dose Equivalent 1i31 in primary coolant and secondary coolant ensure the plant is operated within its analyzed condition. The dose consequences resulting from the most limiting design basis accident are within the limits defined in GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3), or the NRC approved licensing basis (e.g., a small fraction of these limits or 10 CFR 50.67 (Ref. 4)):

SG Tube Integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

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

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

de tube integrity is defined by the performance cgite:ia. The persatisfa criteria includedesign basis parameters thdffge accodancVperformance. The Steam Generator rogramrm provides theruation process for determining coermance ih he performance riteria.

Compliance with the LC cudng MOI)S 1 through 4 is determined by veaifyinge: s

. atlsfactory compgetlon o an ter-asessment in accordance iteiSteam Generor Programrequirements as part o e f-SG inspection, and\

• lan oeaion within the operating cycle defined by te Aoperational assessment.

Performance Criteria Accident induced leakage and structural integrity are two of the three performance criteria defined by the Steam Generator Program. These two, along with the third performance criterion, operational LEAKAGE, act together to provide reasonable assurance of tube integrity at normal and accident conditions.

Catawba Units I and 2 B 3.4.1 8-4 Revision No. 0

SG Tube Integrity B 3.4.18 BASES LCO (continued)

The structural integrity and accident induced leakage performance criteria are documented in Specification 5.5.9. The operational LEAKAGE performance criterion is included in LCO 3.4.13, "RCS Operational LEAKAGE. All three performance criteria are described below:

(i) Structural Integrity Criterion The structural integrity criterion is:

"SG tubing shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown, and all anticipated transients included in the accident analysis 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 largest primary to secondary pressure differential associated with ASME Section III, Level D service. Additional conditions identified in the design and licensing basis shall be evaluated to determine if the associated loads do not contribute to burst.

I Contributing loads that do affect burst shall be assessed with a safety factor of 1.0 and combined with the appropriate load due to the defined pressure differential."

The structural integrity criterion can be broken into two separate considerations:

• Providing a margin of safety against tube burst under normal and accident conditions, and

• Ensuring structural integrity of the SG tubes under all anticipated transients included in the design specification.

Tube Burst 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."

Catawba Units 1 and 2 B 3.4.1 8-5 Revision No. 0

SG Tube Integrity B 3.4.18 BASES LCO (continued)

The structural integrity criterion provides reasonable assurance that a SG tube will not burst during normal or accident conditions.

The structural integrity criterion requires that the tubes not burst when subjected to differential pressures equal to 3.0 times those experienced during normal steady state full power operation and 1.4 times ASME Section I, Level D accident pressure differentials. Other loadings required by the design and licensing basis shall be combined with the design basis accident loads without application of the 1.4 safety factor. The safety factors of 3.0 and 1.4 and the requirement to include applicable design basis loads are based on ASME Code Section III Subsection NB (Ref.

5) requirements and Draft Regulatory Guide 1.121 (Ref. 6) guidance.

In the context of the structural integrity criterion, 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 or sleeving levels, primary or secondary modifications, or Th,t should be assessed and their effects on differential pressure should be included if significant."

Guidance on accounting for changes in these parameters is provided in the EPRI Steam Generator Integrity Assessment Guidelines (Ref. 7).

In addition to the safety factors of 3.0 and 1.4, further adjustments may be required to ensure representative verification of tube burst integrity for various damage forms. For example, adjustments to include axial loading associated with locked tube supports in recirculating SG designs is addressed in Ref. 8 to ensure that the evaluated or tested conditions are at least as severe as those expected during operating and accident events. However, these loads are not subject to the safety factor applied to normal full power operation and accident pressure differentials.

-Tube%frucraUktegrRY Uo Y,Jce? Yi44rtc X Pursuant to the structural integrity criterion, Ref. 1 requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Secton I, Level A (normal operating conditions) and Level B (upset or abnormal conditions) transients included in the design specification.

Catawba Units 1 and 2 B 3.4.18-6 Revision No. 0

SG Tube Integrity B 3.4.18 BASES LCO (continued)

(ii) Accident Induced Leakage Criterion The accident induced leakage criterion is:

"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 150 gallons per day through each SG for a total of 600 gallons per day through all SGs."

In the context of the accident induced leakage criterion, accident induced leakage rate is defined as:

"Accident induced leakage rate means the primary to secondary LEAKAGE occurring during accidents other than a SG tube rupture when tube structural integrity is assumed. This includes the primary to secondary LEAKAGE rate existing immediately prior to the accident plus additional primary to secondary LEAKAGE induced during the accident."

The accident induced leakage criterion can be broken into two separate considerations:

• Meeting design basis conditions, and

• Limiting accident induced leakage to 150 gallons per day through each SG under all circumstances.

Design Basis Primary to secondary LEAKAGE is a factor in the activity releases outside containment resulting from a limiting design basis accident. The radiological dose consequences resulting from a potential primary to secondary leak during design basis accidents must not exceed the offsite dose limits required by Ref. 3, or the control room personnel dose limits required by Ref. 2, dr the NRC approved licensing basis. d When calculating offsite doses, the safety analysis for the limiting design basis accident, other than a SG tube rupture, sets the initial primary to secondary LEAKAGE in each SG to 150 gallons per day.

Catawba Units 1 and 2 B 3.4.18-7 Revision No. 0

SG Tube Integrity B 3.4.1 8 BASES LCO (continued)

Limiting Accident Induced Leakage to 150 Gallons per Day through Each SG 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.

Therefore, a separate performance criterion for accident induced leakage was established. The numerical limit for the accident induced leakage criterion is established at the value for operational LEAKAGE (i.e., 150 gallons per day through each SG).

The NRC has concluded (Item Number 3.4 in Attachment 1 to Ref.

8) that additional research is needed to develop an adequate methodology for fully predicting the effects of LEAKAGE on the outcome of some accident sequences. As a result, LEAKAGE greater than the accident induced leakage criterion is not allowed.

(iii) Operational LEAKAGE Criterion The operational LEAKAGE criterion and its associated Required Action and Surveillance Requirements are contained in LCO 3.4.13, "RCS Operational LEAKAGE." The operational LEAKAGE criterion is not included in the SG Tube Integrity Specification because it is one of the forms of RCS LEAKAGE that are addressed by the RCS Operational LEAKAGE Specification and because, unlike structural integrity and accident induced leakage, it is observable by the operator during MODES 1 through 4. The operational LEAKAGE criterion is presented below for completeness since all of the performance criteria act together to ensure tube integrity.

The operational LEAKAGE criterion is:

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

An explanation of the operational LEAKAGE criterion is provided 17 in the Bases for LCO 3.4.13, "RCS Operational LEAKAGE."

TeJ .ekz The Bases for SR 3.4.13.2 indicates that if this SR is not met, Po+ lu-we compliance with LCO 3.4.18 should be evaluated. If SR 3.4.13.2 is met, then compliance with LCO 3.4.18 need not be evaluated insofar as primary to secondary LEAKAGE is concerned.

5 .{ s VV%c a- rP1ouCA

(- o,A[ -2 OZ C6v/seccR Cwansadetasn Rein t No0 Catawba Units 1 and 2 B 3.4.18-8 ) Revision No 0 izVIgpI0, .

SG Tube Integrity B 3.4.18 BASES APPLICABILITY SG tubes are designed to withstand the stresses due to differential pressures as large as 3.0 times those experienced under normal full power operations or 1.4 times the largest primary to secondary pressure differential for ASME Section III, Level D (faulted) accidents. This requirement is delineated in the structural integrity criterion. This magnitude of differential pressure or the possibility of an accident impacting tube integrity is only possible during MODES 1, 2, 3, and 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1 through 4. When the plant is shut down, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE. In addition, primary coolant activity is also low. Therefore, this LCO is applicable in MODES 1 through 4 only.

ACTIONS The Actions Table is modified by a Note to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for each affected tube. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each affected SG tube. The Completion Times of each affected tube evaluation will be tracked separately, starting from the time the Condition was entered.

A.1 and A.2 Condition applies if it is discovered that one or more inspected SG tubes satisfy the tube repair criteria but were not plugged e-repa4ad in accordance with the Steam Generator Program as required by SR 3.4.18.2. An evaluation of SG tube integrity must be made. SG tube integrity is based on meeting the structural integrity and accident induced leakage performance criteria. In general, an affected tube is one with an indication that at.ies-. tl.I-C the repair criteria. More information on repair limits is provided in Ref. 8.

If it is discovered that a required plugging or repair was not implemented during a previous inspection, the affected SG tube(s) may have SG tube integrity. In this situation, the SGs were returned to service after the last inspection with a tube .aready P T. /

satisfyi%g the repair criteria. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections and still provide assurance that the performance criteria will continue to be met. In order to determine SG tube integrity, an evaluation must be completed that demonstrates that Catawba Units 1 and 2 B 3.4.1 8-9 Revision No. 0

SG Tube Integrity B 3.4.18 BASES ACTIONS (continued) the performance criteria will continue to be met at the time of the next SG inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discoveredxc -es e<<cgr- X f)-ss /Itcr K~ t 9P fAC pt.X SY mf s; 4 A Completion Time of 7 days allows sufficient time to complete the evaluation. If it is determined that tube integrity is not being maintained, Condition B must be entered.

If the evaluation determines that tube integrity is maintained for the affected tube(s), Required Action A.2 allows plant operation to continue until the next outage as lng as the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged or repaired prior to entering MODE 4 after the outage.

This Completion Time is acceptable since the condition will be corrected no later than at the next inspection of the affected SG and the time to the next inspection is supported by the Steam Generator Program as part of the evaluation completed upon entering Condition A. The timing of the next inspection is based on continuing to meet the structural integrity and accident induced leakage performance criteria.

B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or f SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action reduces the factors that tend to challenge tube integrity.

The allowed Completion 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. In MODE 5, the pressure stresses acting on the RCPB are much lower and further deterioration is much less likely.

SURVEILLANCE SR 3.4.18.1 -i-J )

REQUIREMENTS During shutdown periods the S s will be inspected as required by the Steam Generator Program. The Steam Generator Program is required by Specification 5.5.9. Ref. I and its referenced EPRI Guidelines establish the 'onten of the Steam Generator Program. y Catawba Units 1 and 2 13 3.4.18-10 Revision No. 0

SG Tube Integrity B 3.4.18 BASES SURVEILLANCE REQUIREMENTS (continued)

A, ef-he-teare Senerat-r -Pregram-ensumes that the inspection is appropriate and consistent with accepted industry practices.

6At ,

dh5rC J During SG inspections the licensee will erform a condition 5C9 monitoring assessment of the SG tube. The condition monitoring assessment determines the "as found condition of the SG tubes following inspection with respect to the structural integrity and accident induced leakage performance criteria. The purpose of the condition monitoring assessment is to ensure that the performance criteria have been met for the previous operating period.

The team'G qneratorogr determines the.scope of th~ - £ Yt inspfctio, and.the mh L eu>,to de6j2ine com@an6e j 5e the perormance criteria.

• The inspection scope defines which tubes or areas of tubing within the SG are to be inspected. Inspection scope is a function of existing and potential degradation locations ar4 safaty/pessL1reboundaFy-eonsideratioTrs.

• Inspection methods are those Non-Destructive Examination (NDE) techniques used to find potential degradation.

Inspection methods are a function of degradation morphology, NDE technique capabilities, and inspection locations.

I he Steam uenerator Program defines the Frequency of SR.

3. 1. The Frequency is determined by the operational" assessnieot and other limitations in the PWR Steam Generator Examination3 ijidelines (Ref. 9). The limitations in'Ref. 9 and the operational assesbment determine the length-dthe surveillance period by using information on existing,degradations and growth rates to define a cycle lengthAhpatyro1ides reasonable assurance _-

that the tubing will meet the peobrmance criteria at the next z t scheduled inspection. .-

The maximum'pte etween SG nspectorisisimited.

Catawba will,perform required SG inspections of tubirqg and/or sleeve,atintervals no greater than those documentedln Scification 5.5.9.

SR 3.4.18.2 .

{(t 4 ss, During a SG inspection, any inspected tube that satisfies Steam Generator Program repair criteria is r er removed from Catawba Units 1 and 2 B 3.4.18-11 Revision No.

SG Tube Integrity B 3.4.18 BASES SURVEILLANCE REQUIREMENTS (continued) service by plugging. Repair criteria are defined as:

"Repair criteria are those NDE measured parameters at or beyond which a tube must be repaiFed-tsi-rinF-approvedrepair-method-ocremoved from service by plugging.

TXhe tube repair creia establish limits for tube degradation that

- pr vide reasonae ssurance that aIl tubes,leftin service,(e.g.,

wit egrad to not atisfyingthe repair criteria) will meet the e r '

performancvcriteria at theJlext sch'ecyl1d inspectiorlby allowing for anti ip ed growth during the intervening time interval.

Tube repair crteria are either the standard through wall (TW) depth based criterion (e.g., 40% TW for Catawba), or TW depth based criteria for repair techniques approved by the NRC, or other <- fvje4-  !

Alternate Repair Criteria (ARC) approved by the NRC such as a voltage based repair limit per Generic Letter 95-05 (Ref. 10).

The depth based criterion, approved for use at all plants by the NRC, was established when the most frequent form of degradation was general wastage corrosion. This type of nIrinLJrtmtin cnr i eii i II%

JI hi n IPr'fhnrmc ric f'%rnf rnrirnrlnf;ti nsnrl ic characterized by a volumetric loss of the tube wall. This criterion was established to allow for NDE uncertainties and growth and still provide a reasonable assurance that all tubes with degradation not exceeding the criterion will exhibit acceptable structural integrity and accident induced leakage. Additional basis information is provided in Ref. 8.

Since not all forms of tube degradation can be accurately measured for flaw depth in terms of percentage of tube wall thickness, some tubes are "plugged or repaired on detection" to ensure that detected flaws that exceed the depth based criterion are not left in service.

In ddition, since the.probability of detecting a flaw is not a ce aintyfbr a given'edd current te6hnique, it is probable that sone f)awskwill not be detected during an\ sppcf1on. This condition does not meanthat-plug on deteclon' has not been folloded or that the depth based criterion has been violated.

In recent years, improved inspection techniques, knowledge of I corrosion mechanisms, and experience have revealed additional types of tube degradation in the form of cracks in the tube wall. In UO 1) e Cej some instances, a reliable method of cliaracterizing specific types 's 11"r) of cracks at defined locations within certain SG designs has been Catawba Units I and 2 B .4.18-12 Revision No. 0

SG Tube Integrity B3.4.18 BASES SURVEILLANCE REQUIREMENTS (continued) developed. In these cases, the industry has developed, and the NRC has approved ARC to permit leaving a tube in service (as opposed to plugging) when the tube has indications that fall within N the limits established by the ARC. "Plug or repair on detection" is not an ARC.

The NRC must approve all repair criteria prior to use. The repair criteria approved for use at Catawba are listed in Specification 5.5.9.

Due to technique and analyst uncertainties, sampling plans, and probability of detection, there is a possibility that tube(s) satisfying the repair criteria will not be detected during a particular SG inspection. If the flaw(s) is detected during a subsequent inspection, the condition is not considered a reportable event unless it is determined that the performance criteria are not met.

'NS tube repairs are only performed using approved repair met s. Repair methods are defined as:

'Repair methods are those means used to reestablish the RCS pressure,boundary integrity of SG tubes wJhyut removing the tube from service. Plugging a SG tube is not a repair.' /

Repair methods are appr(oved by the.,RC either by license amendment or as part of the NRC'sXapproval of applicable ASME Code requirements. The repairymethods approved by license amendment (if any) are listedi-Specification 5.5.9. The repair methods approved by t NRC J thr'ough the ASME Code are those specifically listed in4S` E Section XlI,',IWA-4720 (Ref. 11) of Code editions and addenda listed in 10 CFR 50.55a (Ref. 12). New repair methods,designed in accordance witt general Code requirements"as opposed to being specifically listed in the Code article ci¢e4 above) may not be implemented without prior NRC appro tl.\

T ere are no repair methods presently approved by license

/amendment for use at Catawba.

Inspected SG tubes that satisfy the repair criteria are repaired or removed from service bytplugging prior to entry into MODE 4.

This is necessary in order to provide reasonable assurance that tube integrity will be maintained until the next scheduled inspection.

Catawba Units 1 and 2 B 3.4.18-13 Revision No. 0

SG Tube Integrity B 3.4.18 BASES REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."

2. 10 CFR 50 Appendix A, GDC 19, "Control Room."

3. 10 CFR 100, "Reactor Site Criteria."

4. 10 CFR 50.67, "Accident Source Term."

5. ASME Boiler and Pressure Vessel Code, Section 1II, Subsection NB, "Rules for Construction of Nuclear Facility Components, Class 1 Components."

6. Draft Regulatory Guide 1.121, Basis for Plugging Degraded Steam Generator Tubes," August 1976.

7. EPRI Steam Generator Integrity Assessment Guidelines.

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

9. EPRI PWR Steam Generator Examination Guidelines.

10. Generic Letter 95-05, "Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," August 3, 1995.

11. ASME Section Xi, IWA-4720, "Sleeving.'

12. 10 CFR 50.55a, "Codes and Standards."

Catawba Units 1 and 2 B 3.4.18-14 Revision No. 0

Sample Administrative Technical Specifications

- Steam Generators with Alloy 600 TT Tubing 5.5.9 Steam Generator (SG) Pro-gram

a. A Steam Generator Program shall be established and implemented to ensure that steam generator tube integrity is maintained during operation in Modes 1, 2, 3, 4.

Steam generator tube integrity is maintained by meeting the following tube integrity performance criteria:

1. Structural Integrity Performance Criteria: All 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 cooldown and all /

anticipated transients included in the design specification). This includes/

retaining a safety factor of 3.0 against burst under the normal steady state jull power primary to secondary pressure differential and a'afety fact6or'\f ' ,

against burst applied to the largest primary to secondary pressure djfferential, \N associated with ASME Section 111, Level D service. d6itional conditionsition:;

identified in the design and licensing basis shall be valuated t determine if the associated loads do not contribute to burst. Contributirigsloads that do affe burst shall be assessed with a safety factor of 1.0 andco'mbined with the appropriate load due to the defined pressure differenal [This criterion is under staff review. The staff issued RA.] / \

2. Accident Leakage Integrity Performance CriteriaWffie primary -try accident induced leakage rate for the limitindsign accide ther than a steam generator tube rupture, shall not exceed 150'gallori's per (total from all tubes) for any individual steam generator and 600o6aion'prday (total from all tubes) through all steam generators.';

3. Operational Leakage Integrity Pefformance Criteria: This criterion is specified in LCO 3.4.13, "RCS OperationafLeakage.i* /

b. Condition monitoring assessmeho an e fe "as found" condition of the tubing with respect to the strctujal integrity and accident leakage performance criteria.

The "as found" condition refert the condition of the tubing during a steam generator tube inspection outage',as determlne-frori, inservice inspection results or by other means, prior to the' Igg tub ontion monitoring assessments shall be conducted during each outageduring vich the steam generators tubes are inspected, plugged, or repaird 'toTco firm'tlhat the structural integrity and accident leakage performance criteria are being t

c. Periodic steam generator tube spections shall be performed. The scope of inspection (i.e., thumber of tubes a rtions of tubes inspected) and method of inspection shall be such as to ensure e reliable detection of any flaws that are present along the length of t^he tube, fromAtibe end weld location (hot) to tube end weld loaction (cold) that may the,applicble tube repair criteria. In addition, the scope, method, and frequ nsctio shall be such as to ensure that the steam generator tube integri 'ismaintaihed. In addition,

1. 100% of the tubes in each steam generator shall be inspected in the first refueling outage following installation.

2. Except as provided for in 5.5.9.c.3, inspect 100% of tubes at sequential intervals of 120, 90, and, thereafter, 60 EFPM. The first sequential interval shall be considered to begin at the first inservice inspection of the steam generators.

In addition, inspect 50% of the tubes by the refueling outage nearest the mid point of the interval and the remaining 50% by the refueling outage near the end of the interval. No steam generator can operate for more than 48 EFPM or two fuel cycles without being inspected.

3. If a crack-like indication is found in any steam generator tube, then the inspection for each steam generator for the degradation mechanism,that c the crack-like indication shall not exceed 24 EFPM or one fuel cycle6:Qf information, such from examination of a pulled tube, indicates that he mdi is not associated with crack(s), then the indication need not'be tr"ed as like.) //

Extension of these maximum inspection interval requiremens..throug.pliain Surveillance Requirement 3.0.2 is not permissible.

d. Tube Repair Criteria (i.e., tube plugging limits): Tubes should,kbplu'gged*.(or repa such that steam generator tube integrity is maintained for Jie period of tie,betwE inspections. In addition, tubes found by inservice inspection to conYarflawswith a depth equal to or exceeding 40% of the nominal tubeAwa1iickness shlib plu prior to plant restart (entry into Mode 4). I

e. Application of Surveillance Requirement 3.0.

Reporting Requirements:

If the results of the steam generator inspection indicate greater than 1% of the inspected tubes in any steam generator exceed the steam generator tube repair criteria specified in Specification 5.5.9, Steam Generator Program," a A report shall be submitted within 120 days after the initial entry into MODE 4 following completion of the each steam generator inspection.

The report shall include:

a. The scope of inspections performed on each SG.

b. Nondestructive examination techniques used for each degradation mechanism.

c. Location, orientation, and measured sizes (if available) of all indications. ,

d. Number and location of tubes plugged or repaired during the ins reason for plugging/repair.

e. Repair method utilized and the number of tubes repaired by eac

f. The effective plugging percentage for all plugging and tube repaI

g. The results of condition monitoring, including the results of tube

h. The planned operating interval until the next inspection 4