ML060240111

Response to Request for Additional Information - License Amendment Request: Technical Specification Improvement Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process

ML060240111
Person / Time
Site:	Calvert Cliffs
Issue date:	01/20/2006
From:	Spina J Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML060240111 (17)

Text

James A. Spina Calvert Cliffs Nuclear Power Plant, Inc.

Vice President 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410.495.4455 410.495.3500 Fax Constellation Energiz Generation Group January 20, 2006 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant UnitNos. 1 & 2; DocketNos. 50-317 & 50-318 Response to Request for Additional Information - License Amendment Request:

Technical Specification Improvement Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process

REFERENCES:

(a) Letter from Mr. GJ. Vanderheyden (CCNPP) to Document Control Desk (NRC), dated July 13, 2005, "License Amendment Request: Technical Specification Improvement Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process" (b) Letter from Mr. P. D. Milano (NRC) to Mr. G. Vanderheyden (CCNPP),

dated December 23, 2005, "Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Supplemental Request for Additional Information (RAI) Regarding Steam Generator Tube Integrity Requirements (TAC Nos. MC8067 and MC8068)"

By letter dated July 13, 2005 (Reference a), Calvert Cliffs Nuclear Power Plant, Inc. (CCNPP) submitted a license amendment request to the Nuclear Regulatory Commission (NRC) to revise the Technical Specification requirements related to steam generator tube integrity. By letter dated December 23, 2005 (Reference b), the NRC requested additional information to support its evaluation. This letter provides the requested information.

NRC Request 1:

Discuss the licensee 's plans to:

(a) Use the industry accepted definition of accident-induced leakage and modify the proposed TS Bases to be consistent with this definition. If the industry accepted definition will not be used, discuss the plans for modifying the proposed accident-inducedleakage rate limit to accountfor operational leakage.

rbh) Modify plantprocedures to ensure that the accident-inducedleak rate limit will not be exceeded as a result of the higher leak rates that may be observed during a DBA [design basis accident](as a result of inducing new" leakage or as a result of the higher drivingforcefor leakage). Alternatively, So0(

Document Control Desk January 20, 2006 Page 2 discuss any plans (including the technical basis) for modifying the normal operating and accident-induced leakage limit to addressthese effects.

CCNPP Response:

(a) We will use the industry accepted definition of accident-induced leakage, i.e., the term accident induced leakage includes any primary-to-secondary leakage existing prior to the accident in addition to the primary-to-secondary leakage induced during the accident. The modified proposed Technical Specification Bases pages are attached (Attachment 1). The changes related to this questions are shown in bold type.

(b) We will modify plant procedures to ensure that the accident-induced leak rate limit will not be exceeded as a result of the higher leak rates that may be observed during a design basis accident.

Plant procedures will be revised to limit operational leakage to 50 gpd/SG to ensure the Technical Specification limit of 100 gpd/SG assumed in the accident analysis will not be exceeded as a result of additional leakage induced during the accident.

NRC Request 2:

In the table on Page 2 of the [July 13], 2005, application, the operationalprimary-to-secondaryleakage limit and accident-inducedprimary-to-secondaryleakage limit areprovided. Both limits are 100 gpd per SG measured at room temperature. In November 29, 2005, letter, the operationaland accident-induced primary-to-secondaryleakage limits are 100 gpd per SG measuredat hot plant conditions. Discuss this apparentdiscrepancy.

CCNPP Response:

The information provided in the July 13, 2005, application is incorrect and this issue has been entered into our corrective action program. As corrected in our November 29, 2005, letter, the operational and accident-induced primary-to-secondary leakage limits are 100 gpd/SG measured at hot plant conditions.

As shown in Attachment (1), the Technical Specification Bases have been corrected to reflect hot plant conditions also.

The information provided above does not change the previously referenced No Significant Hazards Consideration or the Environmental Assessment..

Document Control Desk January 20, 2006 Page 3 Should you have questions regarding this matter, please contact Mr. L. S. Larragoite at (410) 495-4922.

STATE OF MARYLAND  :

TO WIT:

COUNTY OF CALVERT  :

I, James A. Spina, being duly sworn, state that ][ am Vice President - Calvert Cliffs Nuclear Power Plant, Inc. (CCNPP), and that I am duly authorized to execute and file this License Amendment Request on behalf of CCNPP. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other CCNPP employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subscribed and sworn before me a Notary-1ublic in and for the State of Maryland and County of 5ib)7a r 7 is ,this day off L X , 2006.

,IlJNESSmy Hand and Notarial Seal:

Notary Public My Commission Expires: Dat'e Dae JAS/GT/bjd

Attachment:

(1) Modified Proposed Technical Specification Bases Changes cc: P. D. Milano, NRC Resident Inspector, NRC S. J. Collins, NRC R. I. McLean, DNR

ATTACHMENT (1)

MODIFIED PROPOSED TECHNICAL SPECIFICATION BASES CHANGES Mark-up Technical Specification Bases Pages B 3.4.13-2 through B 3.4.13-5 NEW B 3.4.18-1 through B 3.4.18-8 Calvert Cliffs Nuclear Power Plant, Inc.

January 20, 2006

INSERT B 3.4.13 A

d. Primary to Secondary LEAKAGE through Any One Steam Generator The limit of 100 gallons per day per SG is based on safety analysis assumption. Plant procedures further limit operational LEAKAGE to 50 gpd/SG to ensure the TS operational limit of 100 gpd/SG assumed in the accident analysis will not be exceeded as a result of additional LEAKAGE induced during the accident. This limit is more conservative than the 150 gpd/SG operational LEAKAGE performance criterion in Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines (Reference 3). 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." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage.

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.

INSERT B 3.4.13 B The surveillance is modified by two Notes. Note 1 states that this SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

INSERT B 3.4.13 C Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 100 gallons per day cannot be measured accurately by an RCS water inventory balance.

INSERT B 3.4.13 D This SR verifies that primary to secondary LEAKAGE is less or equal to 100 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.18, "Steam Generator Tube Integrity," should be evaluated. The 100 gallons per day limit is measured at hot plant condition as described in Reference 4. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady-state operation. For RCS primary to secondary LEAKAGE determination, steady-state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, and makeup and letdown.

The Surveillance Frequency of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the Electric Power Research Institute (EPRI) guidelines (Reference 4).

INSERT B 3.4.13 E

3. NEI 97-06, Steam Generator Program Guidelines

4. EPRI, Pressurized Water Reactor Primary-to-Secondary Leak Guidelines

RCS Operational LEAKAGE B 3.4.13

.BASES Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for a LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes a 100 gpd/SG primary to secondary LEAKAGE as the initial condition.

Primary to secondary LEAKAGE is a factor in the dose releases outside the Containment Structure resulting from a steam line break accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a SGTR. The leakage contaminates the secondary fluid.

Reference 1, Section 14.15 analysis for SGTR assumes the contaminated secondary fluid is released via the atmospheric dump valves and main steam safety valves. Most of the released radiation is due to the ruptured tube. The 70 ig64~+'100 gpd/SG primary to secondary LEAKAGE is relatively inconsequential..

The steam line break is more limiting for site radiation releases. The safety analysis for the steam line break accident assumes 100 gpd/SG primary to secondary LEAKAGE as an initial condition. The dose consequences resulting from the steam line break accident are described in Reference 1, Section 14.14.

Reactor Coolant System operational LEAKAGE satisfies 10 CFR 50.36(c)(2)(ii), Criterion 2.

LCO Reactor Coolant System operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE.

Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

CALVERT CLIFFS - UNITS 1 & 2 B 3.4. 13-2 Revision dire

RCS Operational LEAKAGE B 3.4.13 BASES

b. Unidentified LEAKAGE One gpm of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment, can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with the detection of unidentified LEAKAGE and is well within the capability of the RCS makeup system. Identified LEAKAGE includes LEAKAGE to the Containment Structure from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled RCP seal leakoff (a normal.

function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d.-LEt ar to Secondary LEAKAGE through Any One Steamn The 100 gallon per mit o to secondary LEAKAG ny one SG is consistent i leevingoicommitments.

APPLICABILITY In MODEs 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODEs 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

ACTIONS A.1 EAKAGE dentified LEAKAG Bin excess of the LCO limits must be reduced to within limits within four hours. This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits CALVERT CLIFFS - UNITS 1 & 2 133.4. 13-3 Rev i si on.0?

RCS Operational LEAKAGE

-B 3.4.13 BASES before the reactor must be shut down. This action is necessary to prevent further deterior tion of the RCPI B.1 and B.2 If any pressure boundary LEAKAG xists or if unidentified):;,

identified oLEAKAGE cannot be reduced tow i within four s, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. 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 to 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. 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 conditions from full power conditions in an orderly manner and without challenging plant Systems. In MODE 5, the pressure stresses acting on the RCF'B are much lower, and further deterioration is much less likely.

SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would first appear as unidentified LEAKAGE and can only be positively identified by inspection. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

191In'9 Ia .Ie .. Em AWR in'e:ntery balaneE: i~aWn coe .tineffluc t men;iteron w:ith wihn the seeendaff. s An

<.eam edate s a.A---

The RCS water inventory balance must be perf rmed with the reactor at steady-state operating conditionstaA-nd-4ear operating pres-s4e:

Steady-state operation is required to perform a proper water inventory balance; calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady-state is defined as stable RCS CALVERT CLIFFS - UNITS 1 & 2 3.4. 13-4 E13 Revision.,2-

RCS Operational LEAKAGE B 3.4.13 BASES pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal leakoff flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems

-that monitor the containment atmosphere radioactivity and the containment sump level. These leakage detection systems are specified in LCO 3.4.14.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. I SR 3.4.13.2 S~~ide~lhe means necessary to determine S OPER TY in an operational MODE. The requirement demonstra tube integrity in accordance wi Steam Generator Tube eillance Program emph es the importance of SG tub egrity, e ough this surveillance test cannot ormed at normal operating conditions.

In the eve neor more SGs are determine not meet the reqW ents of the Steam Generator Tube Survei e ofrgram at anytime in MODEs 1 through 4, action to co

\_Wih LCQO q. mly_be REFERENCES 1. UFSAR

2. Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection Systems, May 1973

.0 S 3. .13 CALVERT CLIFFS - UNITS 1 & 2 EB3.4.13-5 I Revision 8

SG Tube Integrity B 3.4.18 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.18 Steam Generator (SG) Tube Integrity BASES BACKGROUND Steam Generator tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the RCPB and, as such, are relied on 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 act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This specification addresses only the RCPB integrity function of the SG. The St; heat removal function is addressed by LCO 3.4.4, LCO 3.4.5, LCO 3.4.6, and LCO 3.4.7.

Steam generator tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience 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. The SG performance criteria are used to manage SG tube degradation.

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

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 5.5.9. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

CALVERT CLIFFS - UNITS 1 & 2 B 3.4.18-1 Revision XX

SG Tube Integrity B 3.4.18 BASES The processes used to meet the SG performance criteria are defined by Reference 1.

APPLICABLE The SGTR accident is the limiting design basis event for SG SAFETY ANALYSIS tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is released to the atmosphere via safety valves.

The analysis for design basis accidents and transients other than a SGTR assume SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 100 gpd/SG or is assumed to increase to 100 gpd/SG as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.15 limits assuming the relevant Iodine spiking factors. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of General Design Criteria (GDC) 19 (Reference 2), 10 CFR Part 100 (Reference 3), or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO 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 Revision XX UNITS 11&& 2 B 3.4.18-2 B ,

CLIFFS - UNITS CALVERT CLIFFS -

2 3.4.18-2 Revision XX I

SG Tube Integrity B 3.4.18 BASES 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 is 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 5.5.9, and described 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 have a significant effect on burst or collapse. In that context, the term "significant" 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." For tube integrity evaluations, except for circumferential eiinX CAVR LFS-UIS1&2 B341-CALVERT CLIFFS - UNITS 1 & 2 B 3.4.18-3 Revision XX

SG Tube Integrity B 3.4.18 BASES 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 III, 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 References 4 and 5.

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 the total accident leakage does not exceed 100 gpd/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 operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13 and limits primary to secondary LEAKAGE through any one SG to 100 gpd. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

Reactor Coolant System conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, CALVERT CLIFFS - UNITS 1 & 2 B 3.4.18-4 Revision XX I

SG Tube Integrity B 3.4.18 BASES resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube.

This is acceptable because the required Actions provide appropriate compensatory actions for each affected SG tube.

Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

A.1 and A.2 Condition A applies 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 as required by SR 3.4.18.2. 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 refueling outage or 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. If it is determined that tube integrity is not being maintained, Condition B applies.

A Completion 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, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected Revision XX

& 2 B 3.4.18-5 B ,

CALVERT CLIFFS -

UNITS 1 CLIFFS - UNITS 1 & 2 3.4.18-5 Revision XX I

SG Tube Integrity B 3.4.18 BASES tubes. However, the affected tube(s)must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if 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 />.

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.

SURVEILLANCE SR 3.4.18.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. Reference 1 and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

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 potential degradation. Inspection methods are a function of degradation morphology, non-destructive CALVERT CLIFFS - UNITS 1 & 2 B 3.4.18-6 Revision XX I

SG Tube Integrity B 3.4.18 BASES examination technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.4.18.1. The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Reference 6). 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 5.5.9 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SR 3.4.18.2 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 5.5.9 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw 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 util the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessment to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency prior to entering MODE 4 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.

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

2. 10 CFR Part 50, Appendix A, GDC 19

3. 10 CFR Part 100

4. ASME Boiler and Pressure Vessel Code, section III, Subsection NB CALVERT CLIFFS - UNITS 1 & 2 B 3.4.18-7 Revision XX

SG Tube Integrity B 3.4.18 BASES

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

6. EPRI, Pressurized Water Reactor Steam Generator Examination Guidelines Revision XX

& 2 1 &

UNITS 1 B 3.4.18-8 B ,

CLIFFS - UNITS CALVERT CLIFFS -

2 3.4.18-8 Revision XXI