shearon harris nuclear power plant, unit 1, license

.,

< Cornelius J. Cannon. Jr.& Progress Energy Vice President

Harris Nuclear PlantProgress Energy Carolinas. Inc.

MAY 2 3 2006 Serial: HNP-06-06010 CFR 50.90

U.S. Nuclear Regulatory CommissionATTENTION: Document Control DeskWashington, DC 20555

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1DOCKET NO. 50-400/LICENSE NO. NPF-63LICENSE AMENDMENT REQUESTAPPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDINGSTEAM GENERATOR TUBE INTEGRITY

Ladies and Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.90, CarolinaPower and Light Company (CP&L) doing business as Progress Energy Carolinas, Inc.,requests an amendment to the Technical Specifications (TS) of the Harris Nuclear Plant(HNP).

The proposed amendment would revise the TS requirements related to steam generatortube integrity. The change is consistent with NRC-approved Revision 4 to TechnicalSpecification Task Force (TSTF) Standard Technical Specification Change Traveler,TSTF-449, "Steam Generator Tube Integrity." The'availability of this TS improvementwas announced in the Federal Register on May 6, 2005 (70 FR 24126) as part of theconsolidated line item improvement process (CLIIP).

In the HNP response (Serial: HNP-06-029 dated February 16, 2006) to NRC GenericLetter (GL) 2006-01, "Steam Generator Tube Integrity and Associated TechnicalSpecifications," HNP committed to submit a request to modify the Steam Generatorportion of the TS that will be consistent with TSTF-449, Revision 4 by May 31, 2006.

Attachment I provides a description of the proposed change and confirmation ofapplicability.

Attachment 2 provides the existing TS pages marked up to show the proposed

changes.

Attachment 3 provides the retyped TS pages.

Attachment 4 provides the existing TS Bases pages marked up to show the proposed

changes (for information only).

Harris Nuclear PlantP.D. Box 165New Hill, NC 27562

T> 919.362.2502F> 919.362.2095

HNP-06-060

HNP requests approval of the proposed license amendment by May 31, 2007, with the- - -- ?-- I - --I- ... ULL!.. ^f%~ -2 .

Attachment 1 to SERIAL: HNP-06-060

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1DOCKET NO. 50-400/LICENSE NO. NPF-63

REQUEST FOR LICENSE AMENDMENTDESCRIPTION AND ASSESSMENT

DESCRIPTION AND ASSESSMENT

1.0 INTRODUCTION

The proposed license amendment revises the requirements in Technical Specifications(TS) related to steam generator tube integrity. The changes are consistent with theNRC-approved Technical Specification Task Force (TSTF) Standard TechnicalSpecification Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision4. The availability of this technical specification improvement was announced in theFederal Register on May 6, 2005 as part of the consolidated line item improvementprocess (CLIIP).

2.0 DESCRIPTION OF PROPOSED AMMENDMENT

Consistent with the NRC-approved Revision 4 of TSTF-449, the proposed TS changes

include:

• Revised TS definition of IDENTIFIED LEAKAGE* Revised TS definition of PRESSURE BOUNDARY LEAKAGE

' New TS 3/4.4.5, "Steam Generator (SG) Tube Integrity"* Revised TS 3/4.4.6.1, "Reactor Coolant System Leakage Detection Systems"

, Revised TS 3/4.4.6.2, "Reactor Coolant System Operational Leakage"* New TS 6.8.4.1, "Steam Generator (SG) Program"• New TS 6.9.1.7, "Steam Generator Tube Inspection Report"

Proposed revisions to the TS Bases are also included in this application. As discussedin the NRC's model safety evaluation (SE), adoption of the revised TS Basesassociated with TSTF-449, Revision 4 is an integral part of implementing this TSimprovement. The changes to the affected TS Bases pages will be incorporated inaccordance with the TS Bases Control Program.

3.0 BACKGROUND

The background for this application is adequately addressed by the NRC Notice ofAvailability published on May 6, 2005 (70 FR 24126), the NRC Notice for Commentpublished on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

4.0 REGULATORY REQUIREMENTS AND GUIDANCE

The applicable regulatory requirements and guidance associated with this applicationare adequately addressed by the NRC Notice of Availability published on May 6, 2005(70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR

10298), and TSTF-449, Revision 4.

Page Al-1 of 4


SHEARON! H--ARRIS NUCLEAR POWER PLANT, UNIT NO. 1DOCKET NO. 50-400/LICENSE NO. NPF-63


5.0 TECHNICAL ANALYSIS

HNP has reviewed the model safety evaluation (SE) published on March 2, 2005 (70FR 10298) as part of the CLIIP Notice for Comment. This model SE included the NRCstaff's SE, the supporting information provided to support TSTF-449, and the changesassociated with revision 4 to TSTF-449. HNP has concluded that the justificationspresented in the TSTF proposal and the SE prepared by the NRC staff are applicableto HNP and justify this amendment for the incorporation of the changes to the HNP TS.

6.0 REGULATORY ANALYSIS

A description of this proposed change and its relationship to applicable regulatoryrequirements and guidance was provided in the NRC Notice of Availability published onMay 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005(70 FR 10298), and TSTF-449, Revision 4.

6.1 Verification and Commitments

The following information is provided to support the NRC staff's review of thisamendment application:

Request Response

Plant Name, Unit No. Shearon Harris Nuclear Power Plant, Unit 1

Steam Generator Model(s) Westinghouse D75

Effective Full Power Years 4(EFPY) of service for currentlyinstalled SGs

Tubing Material Inconel 690TT

Number of tubes per SG 6307

Number and percentage of tubes SG A SG B SG Cplugged in each SG 3 (0.05%) 1 (0.02%) 3 (0.05%)

Number of tubes repaired in each Not ApplicableSG

Degradation mechanism(s) No active degradation mechanisms have beenidentified identified.

Current primary-to-secondary - 150 gpd through any one steam generatorleakage limits - 1 gpm total from all steam generators

- Leakage is calculated at room temperature

Page A1-2 of 4

Attachment I to SERIAL: HNP-06-060



Request Response

Approved Alternate Tube Repair Not ApplicableCriteria (ARC)

Approved SG Tube Repair Not ApplicableMethods

Performance criteria for accident 1 gpm primary-to-secondary leakage isleakage assumed in the licensing basis accident

analysis. Assumed temperature condition isroom temperature.

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION

HNP has reviewed the proposed no significant hazards consideration determinationpublished on March 2, 2005 (70 FR 10298) as part of the CLIIP. HNP has concludedthat the proposed determination presented in the notice is applicable to HNP and thedetermination is hereby incorporated by reference to satisfy the requirements of10 CFR 50.91(a).

8.0 ENVIRONMENTAL EVALUATION

HNP has reviewed the environmental evaluation included in the model SE published onMarch 2, 2005 (70 FR 10298) as part of the CLIIP. HNP has concluded that the staffsfindings presented in that evaluation are applicable to HNP and the evaluation ishereby incorporated by reference for this application.

Page A1-3 of 4

Attachrient 1 to SERIAL: HNP-06-060



9.0 PRECEDENT

This application is being made in accordance with the CLIIP. HNP is not proposingvariations or deviations from the TS changes described in TSTF-449, Revision 4, or theNRC staff's model SE published on March 2, 2005 (70 FR 10298). The followingdifferences from the improved Standard Technical Specifications (ITS) changesdescribed in TSTF-449, Revision 4 are necessary due to the non-ITS format of theHNP TS:

1. The current format and terminology used in the HNP TS are retained to maintainconsistency with the current specifications. In addition, the HNP proposed TSchanges have also been compared with the proposed TS changes from severalother plants with non-ITS formatted TS for consistency. Examples include:

- The general format and numbering convention associated with the current TS forLimiting Conditions for Operation (LCOs), Actions, Surveillance Requirements(SRs) and Notes are retained. For example, the note below ACTIONS of LCO3.4.20 of TSTF-449, Revision 4 has been moved to the bottom of the page as afootnote of LCO 3.4.5 of the HNP TS, and the note has been changed to read,"Separate ACTION entry is allowed for each SG tube," rather than,'"SeparateCondition entry is allowed for each SG tube."

- Terminology used in the current TS Actions is maintained such as: HOTSTANDBY, HOT SHUTDOWN and COLD SHUTDOWN rather than MODE 3,MODE 4, and MODE 5, respectively.

2. Necessary changes regarding the proper timing and conditions for performing the

RCS water inventory balance were made to Specification 3.4.6.1 ACTION c.3.

10.0 REFERENCES

Federal Register Notices:1. Notice for Comment published on March 2, 2005 (70 FR 10298)2. Notice of Availability published on May 6, 2005 (70 FR 24126)

11.0 CONCLUSION

HNP has concluded, based on the considerations discussed above, that: (1) there isreasonable assurance that the health and safety of the public will not be endangered byoperation in the proposed manner, (2) such activities will be conducted in compliancewith the Commission's regulations, and (3) the issuance of the amendment will not beinimical to the common defense and security or to the health and safety of the public.

Page A1-4 of 4



REQUEST FOR LICENSE AMENDMENTPROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGES

PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGES

Page A2-1 of 22

INDEX

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION PAGE

3/4.4 REACTOR COOLANT SYSTEM

3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION

Startup and Power Operation ...... ............... 3/4 4-1

Hot Standby ...... ... ... ... ........... ... ... 3/4 4-2

Hot Shutdown ........ ..... ... ... ..... ........ 3/4 4-4

Cold Shutdown - Loops Filled......... . ..... .. 3/4 4-6

Cold Shutdown - Loops Not Filled ................ .. 3/4 4-7

3/4.4.2 SAFETY VALVES

Shutdown ........ ... ......................... 3/4 4-8

Operating ............. ... ........ ... ..... 3/4 4-9

3/4.4.3 PRESSURIZER......... ... ... .... ........... 3/4 4-10

3/4.4.4 RELIEF VALVES . . ....... .3/4 4-11

3/4.4.5 STEAM .......... .. ... ......... .... ... ......... 3/4 4-13TABLE 4.4- MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED

DURING INSERVICE INSPECTION . . . . 3/4 4-18,•

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION -c •__h.. 3 / 4 4-19•'

TABLE 4.4-2B (DELETED). .. ..... ... ... ... ..... ... ..... .....3/4 4-20

XTABLE 4.4-2C (DELETED) .. .. .. ..... ... ... ..... ... .........3/4 4-20a

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE

Leakage Detection Systems ..... ................ .. 3/4 4-21

Operational Leakage ....... ................... 3/4 4-23

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES . . . 3/4 4-25

3/4.4.7 CHEMISTRY ......... ........................ 3/4 4-26

TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS ......... .. 3/4 4-27

TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCEREQUIREMENTS ........ ....... ........ ... ....... 3/4 4-28

3/4.4.8 SPECIFIC ACTIVITY .............. ... ............. 3/4 4-29).

FIGURE 3.4-1 (DELETED) ........ ... ...................... 3/4 4-30e(<

TABLE 4.4-4 REACTOR COOLANT SPECIFX. ACTIVITY SAMPLE AND ANALYSISPROGRAM ........... ......................... 3/4 4-31

SHEARON HARRIS - UNIT I 1 vii Amendment No.10SI

INDEX

3.0/4.0 BASES

SECTION

3/4.0 APPLICABILITY . . . . . ... . . . . ..

3/4.1 REACTIVITY CONTROL SYSTEMS

3/4.1.1 BORATION CONTROL ..........

3/4.1.2 BORATION SYSTEMS ..........

3/4.1.3 MOVABLE CONTROL ASSEMBLIES .....

3/4.2 POWER DISTRIBUTION LIMITS ........

3/4.2.1 AXIAL FLUX DIFFERENCE ...........

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTORAND NUCLEAR ENTHALPY RISE HOT CHANNEL

FIGURE B 3/4.2-1 (DELETED) ...........

3/4.2.4 QUADRANT POWER TILT RATIO ........

3/4.2.5 DNB PARAMETERS ...........

FACTOR

3/4.3 INSTRUMENTATION

3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATIONAND ENGINEERED SAFETY FEATURES ACTUATION SYSTEMINSTRUMENTATION ..... ................

3/4.3.3 MONITORING INSTRUMENTATION ..........

3/4.3.4 (DELETED) ...... .............. .....



3/4.4.2 SAFETY VALVES ....... ... ... ...... ......

3/4.4.3 PRESSURIZER ...... ..................

3/4.4.4 RELIEF VALVES .... __ _ _ _

3/4.4.5 STEAM GENERATO . -

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE . . . .....

3/4.4.7 CHEMISTRY ..... ....................

3/4.4.8 SPECIFIC ACTIVITY .... ......... ......

3/4.4.9 PRESSURE/TEMPERATURE LIMITS .............

PAGE

B 3/4 0-1

B 3/4 1-1

B 3/4 1-2

B 3/4 1-3

B 3/4 2-1

B 3/4 2-1

B 3/4 2-2a

B 3/4 2-:3

B 3/4 2-6

B 3/4 2-6

B 3/4 3-1

.B''3/4 3-3

B 3/4 3-6

B 3/4 4-1

B 3/4 4-1

B 3/4 4-2

B 3/4 4-2

B 3/4 4-2b

B 3/4 4-3

B 3/4 4-4

B 3/4 4-5

B 3/4 4-6

Amendment No. LlJ16SHEARON HARRIS - UNIT I xiii

INDEX

ADMINISTRATIVE CONTROLS

SECTION

6.6 REPORTABLE EVENT ACTION ....

6.7 SAFETY LIMIT VIOLATION .....

6.8 PROCEDURES AND PROGRAMS ....

PAGE

6-16

. . . . . . . . 6-16

. . . . . . . . 6-16

6.9 REPORTING REQUIREMENTS

6.9.1 ROUTINE REPORTS .... .............. . . .

Startup Report .............. .

Annual-Reports . . . . . . . . . . . . . ... . .

Annual Radiological Environmental Operating Report

Annual Radioactive Effluent Release Report ..

. . . . . . . . 6-20

. . . . . . . . 6-20

. . . . . . . . 6-20

. . . . . . .. 6-21• ~6-22 ,

Core Operating Limits Report ....... ................... 6-24

6.9.2 SPECIAL REPORTS .................................. 6-24

r 6.10 DELETED ........................................ 6-24

6.11 RADIATION PROTECTION PROGRAM ....... ...... ............ 6-26

6.12 HIGH RADIATION AREA ............ ...................... 6-26

6.13 PROCESS CONTROL PROGRAM (PCP) ...... . ...... ........... 6-27

SHEARON HARRIS - UNIT 1 xix Amendment No. d118

DEFINITIONS .I1 .1.

S- AVERAGE DISINTEGRATION ENERGY

1.12 E shall be the average, weighted in proportion to the concentrationof each radionuclide in the reactor coolant atthe time of sampling, of thesum of the average beta and gamma energies per disintegration (MeV/d) forisotopes. with half-lives greater than 15 minutes. making up at least 95% ofthe total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME

1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that timeinterval from when the monitored parameter exceeds its ESF Actuation Set ointat the channel sensor until the ESF equipment is capable of performing itssafety function (i.e., the valves travel to their required positions, pumpdischarge pressures reach their required values, etc.). Times shall includediesel generator starting and sequence loading delays where applicable. Theresponse time may be measured by means of any series of sequential.overlapping, or total steps so that the entire response time is measured. Inlieu of measurement, response time may be verified for selected componentsprovided that the components and the methodology for verification have beenpreviously reviewed and approved by the NRC.

EXCLUSION AREA BOUNDARY

1.14 The EXCLUSION AREA'BOUNDARY shall be that line beyond which the land isnot controlled by the licensee to limit access.

FREQUENCY NOTATION

1.15 The FREQUENCY NOTATION specified for the performance of SurveillanceRequirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM

1.16 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installedto reduce radioactive gaseous effluents by collecting primary coolant systemoff-gases from the primary system and providing for delay or holdup for thepurpose of reducing the total radioactivity prior to release to theenvironment.

IDENTIFIED LEAKAGE

1.17 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such aspump seal or valve packing leaks that are captured and conductedo a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are bothspecifically located and known either not to interfere with theoperation of Leakage Detection Systems or not to be PRESSUREBOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to theSecondary Coolant System.

SHEARON HARRIS - UNIT 1 1-3 Amendment No.

DEFINITIONS

MASTER RELAY TEST

1.18 A MASTER RELAY TEST shall be the energization of each master relay andverification of OPERABILITY of each relay. The MASTER RELAY TEST shallinclude a continuity check of each associated slave relay.

MEMBER(S) OF THE PUBLIC

1.19 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employeesof the licensee, its contractors, or vendors. Also excluded from thiscategory are persons who enter the site to service equipment or tomakedeliveries. This category does include persons who use portions of the sitefor recreational, occupational, or other purposes not associated with theplant.

OFFSITE DOSE CALCULATION MANUAL

1.20 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodologyand parameters used in the calculation of offsite doses resulting fromradioactive gaseous and liquid effluents, in the calculation of gaseous andliquid effluent monitoring Alarm/Trip Setpoints. and in the conduct of theEnvironmental Radiological Monitoring Program. The ODCM shall also contain(1) the Radioactive Effluent Controls and Radiological EnvironmentalMonitoring Programs required by Section 6.8.4 and (2) descriptions of'theinformation that should be included in the Annual Radiological EnvironmentalOperating and Annual Radioactive Effluent Release Reports required bySpecifications 6.9.1.3 and 6.9.1.4.

OPERABLE - OPERABILITY

1.21 A system, subsystem, train, component or device shall be OPERABLE orhave OPERABILITY when it is capable of performing its specified function(s).and when all necessary attendant instrumentation, controls, electrical power,cooling or seal water, lubrication or other auxiliary equipment that arerequired for the system, subsystem, train, component, or device to perform itsfunction(s) are also capable of performing their, related support function(s).

OPERATIONAL MODE - MODE

1.22 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusivecombination of core reactivity condition, power level, and average reactorcoolant temperature specified in Table 1.2.

PHYSICS TESTS

1.23 PHYSICS TESTS shall be those tests performed to measure the fundamentalnuclear characteristics of the reactor core and related instrumentation:(1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisionsof 10 CFR 50.59. or (3) otherwise approved by the Coii* n

PRESSURE BOUNDARY LEAKAGE rlMr o eo)4t

1.24 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steamgenerator tube-leakage) through a nonisolable fault in a Reactor Coolant tys componenbody, pipe wall. or vessel wall.

SHEARON HARRIS - UNIT 1 1-4 Amendment No. 58

Ale~rVV- ZIA A

REACTOR COOLANT SYSTEM

LMTNG-CONDITION FOR OPERATIO-N - --- • '--

3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1. 2. 3. and 4.

ACTION:

With one or more steam generators inoperable, restore the inoperablegenerator(s) to OPERABLE status prior to increasing T,,, above 200'F.

SURVEILLANCE REQUIREMENTS

4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance ofthe following augmented inservice inspection program and the requirements ofSpecification 4.0.5.

4.4.5.1 Steam Generator Sample Selection and Inspection - Each steamgenerator shall be determined OPERABLE during shutdown by selecting andinspecting at least the minimum number of steam generators specified inTable 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam'generator tube minimum sample size. inspection result classification, and'the*,corresponding action required shall be as specified in Table 4.4-2.. Theinservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall beverified acceptable per the acceptance criteria of Specification 4.4.5.4. Thetubes selected for each inservice inspection shall include at least 3% of thetotal number of tubes in all steam generators: the tubes selected for theseinspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistryindicates critical areas to be inspected, then at least 50% of thetubes inspected shall be from these critical areas:

b. The first sample of tubes selected for each inservice inspection(subsequent to the preservice inspection) of each steam generatorshall include:

1. All nonplugged tubes that previously had detectable wallpenetrations (greater than 20%).

2. Tubes in those areas where experience has indicatedpoeta roblems. and 9

SHEARON HARRIS - UNIT 1 3;4 4-13 Amendment No.&!I•

INSERT 3/4.4.5


3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY

LIMITING CONDITION FOR OPERATION

3.4.5 Steam generator tube integrity shall be maintained.

AND

All steam generator tubes satisfying the tube repair criteria shall be plugged inaccordance with the Steam Generator Program.

APPLICABILITY: MODES 1,2, 3, and 4

ACTION*:

a. With one or more steam generator tubes satisfying the tube repair criteria and notplugged in accordance with the Steam Generator Program, within 7 days verifytube integrity of the affected tube(s) is maintained until the next refueling outageor SG tube inspection, or be in HOT STANDBY within 6 hours and in COLDSHUTDOWN within the next 30 hours.

AND

b. Plug the affected tube(s) in accordance with the Steam Generator Program priorto entering HOT SHUTDOWN following the next refueling outage or steamgenerator tube inspection.


4.4.5.1 Verify steam generator tube integrity in accordance with the SteamGenerator Program.

4.4.5.2 Verify that each inspected steam generator tube that satisfies the tube repaircriteria is plugged in accordance with the Steam Generator Program prior toentering HOT SHUTDOWN following a steam generator tube inspection.

* Separate ACTION entry is allowed for each SG tube.

SURVEILLANCE REOUIREMENTS (Continued)

4.4.5.2 (Continued)

3. A tube inspection (pursuant to Specification 4.4.5.4a.8)shall be performed on each selected tube. If any selectedtube does not permit the passage of the eddy current probefor a tube inspection, this shall be recorded and anadjacent tube shall be selected and subjected to a tubeinspection.

c. The tubes selected as the second and third samples (if required byTable 4.4-2) during each inservice inspection may be subjected toa partial tube inspection provided:

1. The tubes selected for these samples include the tubes fromthose areas of the tube sheet array where tubes withimperfections were previously found, and

2. The inspections include those portions of the tubes whereimperfections were previously found.

The results of each sample inspection shall be classified into one of thefollowing three categories:

Categorv Inspection Results

C-i Less than 5% of the total tubes inspectedare degraded tubes and none of theinspected tubes are defective.

C-2 One or more tubes, but not more than 1% ofthe total tubes inspected are defective.or between 5% and 10% of the total tubesinspected are degraded tubes.

C-3 More than 10% of the total tubes inspectedare degraded tubes or more than 1% of theinspected tubes are defective.

Note: In all inspections. previously degraded tubes must exhibit significant•I (greater than 10%) further wall penetrations to be included in the

above percentage calculations._•

SHEARON HARRIS - UNIT 1 314 4-14 Amendment No. "I'

SURVEILLANCE REQUIREMENTS (Continued)

4.4.5.3 Inspection Frequencies - The above required inservice inspections ofsteam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection shall be performed after6 Effective Full Power Months but within 24 calendar months ofSteam Generator Replacement. Subsequent inservice inspectionsshall be performed at intervals of not less than 12 nor more than24 calendar months after the previous inspection. If twoconsecutive inspections, not including the preservice inspection.result in all inspection results falling into the C-1 category orif two consecutive inspections demonstrate that previouslyobserved degradation has not continued and no additionaldegradation has occurred, the inspection interval may be extendedto a maximum of once per 40 months:

b. If the results of the inservice inspection of a steam generatorconducted in accordance with Table 4.4-2 at 40-month intervalsfall in Category C-3, the inspection frequency shall be increasedto at least once per 20 months. The increase in inspectionfrequency shall apply until the subsequent inspections satisfy thecriteria of Specification 4.4.5.3a.: the interval may then beextended to a maximum of once per 40 months: and

c. Additional. unscheduled inservice inspections shall be performedon each steam generator in accordance with the first sampleinspection specified in Table 4.4-2 during the shutdown subsequent Ito any of the following conditions:

1. Reactor-to-secondary tubes.leak (not including leaksoriginating from tube-to-tube sheet welds) in excess of thelimits of Specification 3.4.6.2. or

2. A seismic occurrence greater than the Operating BasisEarthquake. or

3. A loss-of-coolant accident requiring actuation of theEngineered Safety Features. or7

4. A main steam line or feedwater line-break.

TheJe.+eA Iy Amen~bQnt

SHEARON HARRIS - UNIT 1 3/4 4-15 Amendment No.107I.

SSTEAM GENERATORSSURVEILLANCE REQUIREMENTS (Continued)

4.4.5.4 Acceptance Criteria

a. As used in this specification:

1. Imperfection means an exception to the dimensions, finish.or contour of a tube from that required by fabricationdrawings or specifications. Eddy-current testingindications below 20% of the nominal tube wall thickness.if detectable. may be considered as imperfections:

2. Degradation means a service-induced cracking, wastage, wear.or general corrosion occurring on either inside or outsideof a tube:

3. Degraded Tube means a tube containing imperfections greater Ithan or equal to 20% of the nominal wall thickness caused bydegradation:

4. % Degradation means the percentage of the tube wallthickness affected or removed by degradation:

5. Defect means an imperfection of such severity that itexceeds the plugging limit. A tube containing a defect isdefective:

6. Plugging Limit means the imperfection depth at or beyondwhich the tube shall be removed from service and is equal to40% of.the nominal tube wall thickness.

7. Unserviceable describes the condition of a tube if it leaksor contains a defect large enough to affect its structuralintegrity in the event of an'Operating Basis Earthquake. aloss-of-coolant accident, or a steam line or feedwater linebreak as specified in Specification 4.4.5.3c.. above:

8. Tube Inspection means an inspection of the steam generatortube from the point of entry (hot leg side) completelyaround the U-bend to the top support of the cold leg.

9. Preservice Inspection means an inspection of the full lengthof each tube in each steam generator performed by eddycurrent techniques prior to service to establish a baselinecondition of the tubing. This inspection shall be performedprior to POWER OPERATION with the replacement of steamgenerators using equipment and techniques expected to beused during subsequent inservice inspections.

SHEARON HARRIS - UNIT 1 314 4-16 Amendment No 07g ,

(BREACTOR COOLANT SYSTEMI-"

STEAM GENERATORS

SURVEILLANCE REQUIREMENTS (Continued)

4.4.5.4 Acceptance Criteria (Continued)

b. The steam generator shall be determined OPERABLE after completingthe corresponding actions (plug all tubes exceeding the plugging Ilimit) required by Table 4.4-2.

4.4.5.5 Reports

a. Within 15 days following the completion of each inserviceinspection of steam generator tubes, the number of tubes pluggedin each steam generator shall be reported to the Commission in aSpecial Report pursuant to Specification 6.9.2;

b. The complete results of the steam generator tube inserviceinspection shall be submitted to the Commission in a SpecialReport pursuant to Specification 6.9.2 within 12 months followingthe completion of the inspection. This Special Report shallinclude:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for eachindication of an imperfection. and

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall intoCategory C-3 shall be reported in a Special Report pursuant toSpecification 6.9.2 within 30 days and prior to resumption ofplant operation. This report shall provide a description ofinvestigations conducted to determine cause of the tubedegradation and corrective measures taken to prevent recurrence.

SS3/4- 4-17

SHEARON HARRIS -UNIT 1 3/4 4-17 Amendment No.I

f " TABLE 4.4-1:

| MINIMUM NUMBER OF STEAM GENERATORS TO BE" ~ INSPECTED DURING INSERVIC.E.iNSPECTION

No. of Steam Generators per Unit 3

First Inservice Inspection "

Second & Subsequent Inservice Inspections 1C!)(z)

TABLE NOTATIONS

(1) The inservice inspection may be limited to one steam generator on arotating schedule encompassing S% of the tubes if the results of the firstor previous inspections indicate-that.-all steam generators are performingin a like manner. Note that,'under somecircumstances, the operatingconditions in one or more steam generators may be found to be more severethan those in other steam generators. Under such circumstances the samplesequence shall be modified to inspect the most severe conditions.

(2) The other steam generator not inspected during the first inservice inspec-tion shall be inspected. The third and subsequent inspections shouldfollow the instructions described in 1. above.

SHEAON HARRIS -UNIT 1 3/4 4-18 N ''e~l~Q' ~o.

•I'

TABL 4.-2\ STEAM GENERATOR TUB INSPECTION"

1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION

Sample Size Result Action Required Result Action Required Result Action Required

A minimum of C-I None N/A N/A N/A N/AS T s C-2 Plug defective C-I None N/A N/A I.... tubes and inspect

... tu• S i S C- Plug defective C-i None J.tuein-this C-2 tubes and inspect defective tue ,.S.G. additional 4S C-2 Plug defective tubesi

tubes in this S.G. Perform action forC-3 C-3 result of first

samplePerform action

C-3 for C-3 result N/A N/Aof first sample

C-3 Inspect all tubes All otherin this S.G.. S.G.s are None N/A N/A -llug defective C-1tubes and inspect•_2S tubes in eachotherS.G. Some S.G.s Perform action for C-2 N/A N/A

C-2 but no result of second sample

Notification to additionalNRC pursuant to S.G.s are C-3Specification Additional S.G., Insect all tubes in4.4.5.5.c. is C-3 each S.G. and plug N/A N/A

defective tubes.Notification to NRCSursuant topecification 4.4.5.5.c.

S = •% where n is the number of steam generators inspected during an inspection.

SHEARON HARRIS - UNIT 1 3/4 4-19 Amendment No.



LEAKAGE DETECTION SYSTEMS

LIMITING CONDITION FOR OPERATION.

3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shallbe OPERABLE:

a. The Containment Airborne Gaseous Radioactivity Monitoring System.b. The Reactor Cavity Sump Level and Flow Monitoring System, andc. The Containment Airborne Particulate Radioactivity Monitoring

System.

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

ACT ION : T ̀I - - _ - _ -- _ -- n

a. With a. or c. of the above required Leakage Detection SystemsINOPERABLE, operation may continue for up to 30 days provided grabsamples of the containment atmosphere are obtained and analyzedfor airborne gaseous and particulate radioactivity at least onceper 24 hours when the required Airborne Gaseous or ParticulateRadioactivity Monitoring System is inoperable; otherwise, be in atleast HOT STANDBY within the next 6 hours and in COLD SHUTDOWNwithin the following 30 hours.

b. With b. of the above required Leakage Detection Systems inoperablebe in at least HOT STANDBY within the next 6 hours and in, COLDSHUTDOWN within the following 30 hours.

c. With a. and c. of the above required Leakage Detection Systemsinoperable:

1. Restore either Monitoring System (a. or c.) to OPERABLEstatus within 72 hours and

2. Obtain and analyze a grab sample of the containmentatmosphere for gaseous and particulate radioactivity atleast once per 24 hours, and

3. Perform a React r Coolant System water inventory balance atleast one per 8hours.

Otherwise, be in at least HOT STANDBYwithin the next 6 hours andin COLD SHUTDOWN within the following 30 hours.

SHEARON HARRIS - UNIT 1 3/4 4-21


OPERATIONAL LEAKAGE--

LIMITING CONDITION FOR OPERATION OpZ_.-')

3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE,

c. I gpm total re d r e through all steam. 10r g150 Ig aK fro m hpe Q eyathro u an yesteamgenerator, p o, ncry-4o- 5844n A ar,

d. 10 gpm IDENTIFIED LEAKAGE from the eactor •oo an yste -,

e. 31 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of2235 ± 20 psig, and

f. The maximum allowable leakage of any Reactor Coolant SystemPressure Isolation Valve sh~ll be as specified in Table 3.4-1 at apressure of 2235 ± 20 psig.

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

ACTION: wii W rt+ t nC,Ar- Se~coneJorP fet4lcfzue not LAWfI; it)UM{

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBYwithin 6 hours and in COLD SHUTOG e following. 30; hours.

b. With any Reactor Coolant System der than any one ofthe above limits, excluding PRESSURE BOUDAR LEAKAGE and leakagefrom Reactor Coolant System Pressure Isolation Valves, reduce t eleakage rate to within limils within 4 hours or be in at least HOTSTANDBY within the next 6 h urs and i COLD SHU DO N wi in thefollowing 30 hours. S' 0 . ,'

c. With any Reactor Coolant System ru-1 o a-on v egreater than the limit specified in Table 3.4-1, isolate the highpressure portion of the affected system from the low pressureportion within 4 hours by use of at least two closed manual ordeactivated automatic valves, or be in at least HOT STANDBY withinthe next 6 hours and in COLD SHUTDOWN within the following30 hours.

Test pressures less than 2235 psig but greater than 150 psig are allowed.Observed leakage shall be adjusted by multiplying the observed leakage by thesquare root of the quotient of 2235 divided by the test pressure.

SHEARON HARRIS - UNIT 1 3/4 4-23 Amendment No. 0


OPERATIONAL LEAKAGE

SURVEILLANCE REQUIREMENTS Of Ck+4'on C

4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be withineach of the above limits by:

a. Monitoring the containment Airborne Gaseous or ParticulateRadioactivity Monitor at least once per 12 hours:

b. Monitoring the containment sump inventory and Flow MonitoringSystem at least once per 12 hours:

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pumpseals when the Reactor Coolant System pressure is 2235 ± 20 psigat least once per 31 days with the modulating valve fully open.The provisions of Specification 4.0.4 are not applicable for entry

'into MODE 3 or 4:

d. Performance of a React Coolant System water inventory balance atleast once per 72 hour"and

e. Monitoring the Reactor Heed Flange Leakoff System at least onceper 24 hours.

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified inTable 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be: withinits limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLDSHUTDOWN for 7 days or more and if leakagetesting has not beenperformed in the previous 9 months.

c. Prior to returning the valve to service following maintenance.repair or replacement work on the valve, and

d. Within 24 hours following valve actuation due to automatic ormanual action or flow through the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3or 4.4.4.(Q.Z.*3 Prm Vf•onaleaýbe1e.xj-+v- 5cnoAr• I"X•y 56tt ýbe vtr,,• ;J _ 1.5"0 x lon Tre"-

Koti re-,•ureA -to be pex ormeA Lm4 1 I( -hAr q4 •+er b oi- +4e&" -oMt rwkxciorn. *ýfk~k~ 4o p &j~-io-se'OJOtr lektaied~ .

rt~c i oi f je-40roeA n4tz c~s e NIr ri.dc4

d54t1 opkrgedon.

SHEARON HARRIS - UNIT 1 3/4 A-24 Amendment No.


PROCEDURES AND PROGRAMS (Continued)

k. Containment Leakage Rate Testing Program

A program shall be established to implement the leakage rate testingof the containment as required by 10 CFR 50.54 (o) and 10 CFR 50Appendix J. Option B. as modified by approved exemptions. Thisprogram shall be in conformance with the NRC Regulatory Guide 1.163."Performance-Based Containment Leak-Test Program," dated September1995. with the following exceptions noted:

1) The above Containment Leakage Rate Testing Program is onlyapplicable to Type A testing. Type B and C testing shallcontinue to be conducted in accordance with the originalcommitment to 10 CFR 50 Appendix J. Option A.

2) The first Type A test performed after the May 23. 1997 Type Atest shall be performed-no later than May 23. 2012.

3) Visual examination of the containment system shall be inaccordance with Specification 4.6.1.6.1.

The calculated peak containment internal-pressure related to thedesign basis loss-of-coolant accident-is 41.8 psig. The calculatedpeak containment internal pressure related to the design basis mainsteam line break is 41.3 psig. P. will be assumed to be 41.8 psigfor the purpose of containment testing in accordance with thisTechnical Specification.

The maximum allowable containment leakage rate. L. at P., shall be0.1 % of containment air weight per day.

The containment overall leakage rate acceptance criterion is < 1.0La. During the first unit startup following testing in accordancewith this program, the leakage rate acceptance criteria are < 0.60La for the combined Type B and Type C tests, and < 0.75 L. for TypeA tests.

The provisions of Surveillance Requirement 4.0.2 do not apply to thetest frequencies specified in the Containment Leakage Rate TestingProgram. However, test frequencies specified in this Program may beextended consistent with the guidance provided in Nuclear EnergyInstitute (NEI) 94-01, "Industry Guideline for ImplementingPerformance-Based Option of 10 CFR 50Appendix J.3 as endorsed byRegulatory Guide 1.163. Specifically. NEI 94-01 has this provisionfor test frequency extension:

I

1) Consistent with standard scheduling practices for TechnicalSpecifications Required Surveillances. intervals for recommended

rType A testing may be extended by up to 15 months. This optionshould be used only in cases where refueling schedules have beenchanged to accommodate other factors.

The provisions of Surveillance Requirement 4.0.3 are applicable to, h "n nt Leakage RateTsin -qam.

SER AR -A

SHEARON HARRIS - UNIT 1 6-19c Amendment No. 122

INSERT 6.8.4.1.' . "',

I. Steam Generator (SG) Proqram

A Steam Generator Program shall be established and implemented to ensurethat steam generator (SG) tube integrity is maintained. In addition, the SteamGenerator Program shall include the following provisions:

1. Provisions for condition monitoring assessments. Condition monitoringassessment means an evaluation of the "as found" condition of the tubingwith respect to the performance criteria for structural integrity and accidentinduced leakage. The "as found" condition refers to the condition of thetubing during a SG inspection outage, as determined from the inserviceinspection results or by other means, prior to the plugging of tubes.Condition monitoring assessments shall be conducted during each outageduring which the SG tubes are inspected or plugged to confirm that theperformance criteria are being met.

2. Performance criteria for SG tube integrity. Steam generator tube integrityshall be maintained by meeting the performance criteria for tube structuralintegrity, accident induced leakage, and operational leakage.

a) Structural integrity performance criterion. All inservice SG tubes shallretain structural integrity over the full range of normal operating.conditions (including startup, operation in the power range, HOTSTANDBY, and cooldown and all anticipated transients included in thedesign specification) and design basis accidents. This includesretaining a safety factor of 3.0 (3deltaP) against burst under normalsteady state full power operation primary-to-secondary pressuredifferential and a safety factor of 1.4 against burst applied to the designbasis accident primary-to-secondary pressure differentials. Apart fromthe above requirements, additional loading conditions associated withthe design basis accidents, or combination of accidents in accordancewith the design and licensing basis, shall also be evaluated todetermine if the associated loads contribute significantly to burst orcollapse. In the assessment of tube integrity, those loads that dosignificantly affect burst or collapse shall be determined and assessedin combination with the loads due to pressure with a safety factor of 1.2on the combined primary loads and 1.0 on axial secondary loads.

b) Accident induced leakage performance criterion. The primary-to-secondary accident induced leakage-rate for any design basisaccident, other than a SG tube rupture, shall not exceed the leakagerate assumed in the accident analysis in terms of total leakage rate forall SGs and leakage rate for an individual SG. Accident inducedleakage is not to exceed I gpm total for all three SGs.

c) The operation leakage performance criterion is specified in LCO3.4.6.2, "Reactor Coolant System Operational Leakage."

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

4. Provisions for SG tube inspections. Periodic SG tube inspections shall beperformed. The number and portions of the tubes inspected and methodsof inspection shall be performed with the objective of detecting flaws ofany type (e.g., volumetric flaws, axial and circumferential cracks) that maybe present along the length of the tube, from tube-to-tubesheet weld at thetube inlet to the tube-to-tubesheet weld at the tube outlet, and that maysatisfy the applicable tube repair criteria. The tube-to-tubesheet weld isnot part of the tube. In addition to meeting the requirements of 4a, 4b, 4cbelow, the inspection scope, inspection methods and inspection intervalsshall be such as to ensure that SG tube integrity is maintained until thenext SG inspection. An assessment of degradation shall be performed todetermine the type and location of flaws to which the tubes may besusceptible and, based on this assessment, to determine which inspectionmethods need to be employed and at what locations.

a) Inspect 100% of the tubes in each SG during the first refueling outagefollowing SG replacement.

b) Inspect 100% of the tubes at sequential periods of 144, 108, 72, andthereafter, 60 effective full power months. The first sequential periodshall be considered to begin after the first inservice inspection of theSGs. In addition, inspect 50% of the tubes by the refueling outagenearest the midpoint of the period and the remaining 50% by therefueling outage nearest the end of the period. No SG shall operatefor more than 72 effective full power months or three refueling outages(whichever is less) without being inspected.

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

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


6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

n. Mechanical Design Methodologies

XN-NF-81-58(P)(A). "RODEX2 Fuel Rod Thermal-Mechanical ResponseEvaluation Model," approved version as specified in the COLR.

ANF-81-58(P)(A); "RODEX2 Fuel Rod Thermal Mechanical ResponseEvaluation Model," approved version as specified in the COLR.

XN-NF-82-06(P)(A). "Qualification of-Exxon Nuclear Fuel forExtended Burnup," approved version as specified in the COLR.

ANF-88-133(P)(A). "Qualification of Advanced Nuclear Fuels' PWRDesign Methodology for Rod Burnups of 62 GWd/MTU," approvedversion as specified in the COLR.

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/GadoliniaIrradiation Examination and Thermal Conductivity Results,"approved version as specified in the COLR.

EMF-92-116(P)(A). "Generic Mechanical Design Criteria for PWR FuelDesigns,".approved version as specified in the COLR.

(Methodologies for Specification 3.2.1 - Axial Flux Difference.3.2.2 - Heat Flux Hot Channel Factor,.and 3.2.3 - Nuclear EnthalpyRise Hot Channel Factor).

6.9.1.6.3 The core operating limits shall be determined so that allapplicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and.accident analysis limits) of the safety analysis are met.

6.9.1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cyclerevisions or supplements, shall be provided, upon issuance.for each reloadcycle, to the NRC Document Control Desk, with copies to the RegionalAdministrator and Resident Inspector.

SPECIAL REPORTS

6.9.2 Special reports shall be submitted to the NRC in accordance with10CFR50.4 within the time period specified for each report.

6.10 'DELETED

(PAGE 6-25 DELETED)

SHEARON HARRIS - UNIT 1 6-24c Amendment No.

INSERT 6.9.1.7

STEAM GENERATOR TUBE INSPECTION REPORT

6.9.1.7 A report shall be submitted within 180 days after the initial entry into HOTSHUTDOWN following completion of a steam generator tube inspection performedin accordance with Specification 6.8.4.1. 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 degradationmechanism,

d. Location, orientation (if linear), and measured sizes (if available) of serviceinduced 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, and

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



REQUEST FOR LICENSE AMENDMENTREVISED TECHNICAL SPECIFICATIONS (TS) PAGES

REVISED TECHNICAL SPECIFICATIONS (TS) PAGES

Page A3-1 of 21

INDEX

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION



PAGE

Startup and Power Operation ....

Hot Standby .. .. . .. .. . ..

Hot Shutdown ................

Cold Shutdown - Loops Filled ....

Cold Shutdown - Loops Not Filled

3/4.4.2 SAFETY VALVES

Shutdown .... ..............

Operating . . . . . . . . . . . . .

3/4.4.3 PRESSURIZER .. ....... ...

3/4.4.4 RELIEF VALVES ...........


TABLE 4.4-1 (DELETED) ....... . . . .

TABLE-4.4-2 (DELETED) ...........

. . . . . . . . . . . 3/4 4-1

S. . . . . . . . . . 3/4 4-2

. . . ... . . . . .. 3/4 4-4

• . .. . . . . . .. . 3/4 4-6

. . . . . . . . . . . 3/4 4-7

. .. . . . . . .. . . 3/4 4-8

. . . . . . . . . . . 3/4 4-9

..... . . . . . . . 3/4 4-10

..... . . . . . .. .'3/4 4-11

.. . . . . . . . .. 3/4 4-13

. . . . . . . . . . -3/4 4-18

.. . . . . . . . .. '3/4 4-19


Leakage Detection Systems ................

Operational Leakage . . . ..... .. . . . . . . . .

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES .

3/4.4.7 CHEMISTRY . . . . . . . . . . . .. . . . . . . . . . . .

TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS .......

TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCEREQUIREMENTS ....... ... ............ .........

3/4.4.8 SPECIFIC ACTIVITY . . . . . . . . . . . . . . . . . . . .

FIGURE 3.4-1 (DELETED) . . . . . . . . . . . . . . . . . . . . . .

TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY-SAMPLE AND ANALYSISPROGRAM . . . . . . . . . . . . . . . . . . . . . . . . .

3/4 4-21

3/4 4-23

3/4 4-25

3/4 4-26

3/4 4-27

3/4 4-28

3/4 4-29

3/4 4-30

3/4 4-31

SHEARON HARRIS - UNIT 1 vii Amendment No.

INDEX

3.0/4.0 BASES

SECTION

3/4.0 APPLICABILITY . . . . . . . . . . . . . . . . . . .

3/4.1 REACTIVITY CONTROL SYSTEMS

3/4.1.1 BORATION CONTROL ................

3/4.1.2 BORATION SYSTEMS ................

3/4.1.3 MOVABLE CONTROL ASSEMBLIES ...........

3/4.2 POWER DISTRIBUTION LIMITS .............

3/4.2.1 AXIAL FLUX DIFFERENCE .... ...............

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTORAND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR

FIGURE B 3/4.2-1 (DELETED) ................

3/4.2.4 QUADRANT POWER TILT RATIO ..... ............

3/4.2.5 DNB PARAMETERS .............. . . .

3/4.3 INSTRUMENTATION

3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATIONAND ENGINEERED SAFETY FEATURES ACTUATIONSYSTEMINSTRUMENTATION . . . . . . . . . . . . . . . .

3/4.3.3 MONITORING INSTRUMENTATION ..... ......

3/4.3.4 (DELETED) ....... ....................



3/4.4.2 SAFETY VALVES ...... ..................

3/4.4.3 PRESSURIZER ...... ...................

3/4.4.4 RELIEF VALVES ..... ............ .......

3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY .......

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE .........

3/4.4.7 CHEMISTRY ...... ....................

3/4.4.8 SPECIFIC ACTIVITY ...... ................

3/4.4.9 PRESSURE/TEMPERATURE LIMITS ....... ......

PAGE

B 3/4 0-1

B 3/4 1-1

B 3/4 1-2

B 3/4 1-3

B 3/4 2-1

B 3/4 2-1

B 3/4 2-2a

B 3/4 2-3

B 3/4 2-6

B 3/4 2-6

B 3/4, 3-1

B 3/4' 3-3

B 3/4 3-6

B 3/4 4-1

B 3/4 4-1

B 3/4 4-2

B 3/4 4-2

B 3/4 4-2b

B 3/4 4-3

B 3/4 4-4

B 3/4 4-5

B 3/4 4-6

SHEARON HARRIS - UNIT 1 xiii Amendment No.

INDEX

ADMINISTRATIVE CONTROLS '

SECTION PAGE

6.6 REPORTABLE EVENT ACTION ......... ...................... 6-16

6.7 SAFETY LIMIT VIOLATION ....... ....................... 6-16

6.8 PROCEDURES AND PROGRAMS ....... .......... ............ 6-16

6.9 REPORTING REQUIREMENTS

6.9.1 ROUTINE REPORTS .......... .......................... 6-20

:Startup Report ....... ... .......................... 6-20

ýAnnual Reports ....... ... .......................... 6-20

Annual Radiological Environmental Operating Report ... ........ 6-21

Annual Radioactive Effluent Release Report . ..... .......... 6-22

Core Operating Limits Report ............ ............... 6-24

Steam Generator Tube Inspection Report ..... ............. 6-24c I

6.9.2 SPECIAL REPORTS .......... .......................... 6-24

6.10 DELETED ........... .............................. 6-24

6.11 RADIATION PROTECTION PROGRAM ...... ...... .............. 6-26

6.12 HIGH RADIATION AREA ..... ............. ............... 6-26

6.13 PROCESS CONTROL PROGRAM (PCP) ...... ................... 6-27

SHEARON HARRIS - UNIT 1 xix Amendment No.

DEFINITIONS

E - AVERAGE DISINTEGRATION ENERGY

1.12 E shall be the average, weighted in proportion to the concentrationof each radionuclide in the reactor coolant at the time of sampling, of thesum of the average beta and garmma energies per disintegration (MeV/d) forisotopes, with half-lives greater than 15 minutes. making up at least 95% ofthe total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME

1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that timeinterval from when the monitored parameter exceeds its ESF Actuation Setpointat the channel sensor until the ESF equipment is capable of performing itssafety function (i.e., the valves travel to their required positions, pumpdischarge pressures reach their required values, etc.). Times shall include

.diesel generator starting and sequence loading delays where applicable. Theresponse time may be measured by means of any series of sequential,overlapping, or total steps so that the entire response time is measured. Inlieu of measurement, response time may be verified for selected componentsprovided that the components and the methodology for verification have beenpreviously reviewed and approved by the NRC.

EXCLUSION AREA BOUNDARY

1.14 The EXCLUSION AREA BOUNDARY shall be that line beyond which the landjisnot controlled by the licensee to limit access.

FREQUENCY NOTATION

1.15 The FREQUENCY NOTATION specified for the performance of Surveillance"Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM

1.16 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installedto reduce radioactive gaseous effluents by collecting primary coolant systemoff-gases from the primary sýstem andtproviding for delay or holdup for thepurpose of reducing the total radioactivity prior to release to theenvironment.

IDENTIFIED LEAKAGE

1.17 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such aspump seal or valve packing leaks that are captured and conducted

o a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are bothspecifically located and known either not to interfere with theoperation of Leakage Detection Systems or not to be PRESSUREBOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam enerator to theSecondary Coolant System (primary-to-secondary leakage).


DEFINITIONSMASTER RELAY TEST

1.18 A MASTER RELAY TEST shall be the energization of each master relay andverification of OPERABILITY of each relay. The MASTER RELAY TEST shallinclude a continuity check of each associated slave relay.

MEMBER(S) OF THE PUBLIC

1.19 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employeesof the licensee, its contractors, or vendors. Also excluded from thiscategory are persons who enter the site to service equipment or to makedeliveries. This category does include persons who use portions of the sitefor recreational, occupational, or other purposes not associated with theplant.

OFFSITE DOSE CALCULATION MANUAL

1.20 The OFFSITE DOSE CALCULATION MANUAL (ODCM),shall contain the methodologyand parameters used in the calculation of offsite doses resulting fromradioactive gaseous and liquid effluents, in the calculation of gaseous andliquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of theEnvironmental Radiological Monitoring Program. The ODCM shall also contain(1) the Radioactive Effluent Controls and Radiological EnvironmentalMonitoring Programs required by Section 6.8.4 and (2) descriptions of.:theinformation that should be included in the Annual Radiological EnvironmentalOperating and Annual Radioactive Effluent Release Reports required bySpecifications 6.9.1.3 and 6.9.1.4.

OPERABLE - OPERABILITY

1.21 A system, subsystem, train, component or device shall be OPERABLE orhave OPERABILITY when it is capable of performing its specified function(s),and when all necessary attendant instrumentation, controls, electrical power,cooling or seal water, lubrication or other auxiliary equipment that arerequired for the system, subsystem, train, component, or device to perform itsfunction(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE

1.22 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusivecombination of core reactivity condition, power level, and average reactorcoolant temperature specified in Table 1.2.

PHYSICS TESTS

1.23 PHYSICS TESTS shall be those tests performed to measure the fundamentalnuclear characteristics of the reactor core and related instrumentation:(1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisionsof 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE

1.24 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondaryleakage) through a nonisolable fault in a Reactor Coolant System componentbody, pipe wall, or vessel wall.



3/4.4.5 STEAM GENERATOR-,(SG) TUBE INTEGRITY


3.4.5 Steam generator tube integrity shall be maintained.

AND

All steam generator tubes satisfying the tube repair criteriashall be plugged in accordance with the Steam Generator Program.

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

ACTION*:

a. With one or more steam generator tubes satisfying the tube repaircriteria and not plugged in accordance with the Steam GeneratorProgram, within 7 days verify tube integrity of the affectedtube(s) is maintained until the next refueling outage or SG tubeinspection, or be in HOT STANDBY within 6 hours and in COLDSHUTDOWN within the next 30 hours.

AND

b. Plug the affected tube(s) in accordance with the Steam GeneratorProgram prior to entering HOT SHUTDOWN following the nextrefueling outage or steam generator tube inspection.


4.4.5.1

4.4.5.2

Verify steam generator tube integrity in accordancewith the Steam Generator Program.

Verify that each inspected steam generator tube thatsatisfies the tube repair criteria is plugged inaccordance with the Steam Generator Program prior toentering HOT SHUTDOWN following a steam generator tubeinspection.

•CTION entry is allowed for each SG tube." Separate A


/4 4-14Deleted by Amendment


Table 4.4-1Deleted by Amendment


Table 4.4-2Deleted by Amendment




LEAKAGE DETECTION SYSTEMS


3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shallbe OPERABLE:

a. The Containment Airborne Gaseous Radioactivity Monitoring System,b. The Reactor Cavity Sump Level and Flow Monitoring System, andc. The Containment Airborne Particulate Radioactivity Monitoring

System.

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

ACTION:

a. With a. or c. of the above required Leakage Detection SystemsINOPERABLE, operation may continue for up to 30 days provided grabsamples of the containment atmosphere are obtained and analyzedfor airborne gaseous and particulate radioactivity at least.onceper 24 hours when the required Airborne Gaseous or ParticulateRadioactivity Monitoring System is inoperable; otherwise, be in atleast HOT STANDBY within the next 6 hours and in COLD SHUTDOWNwithin the following 30 hours.

b. With b. of the above required Leakage Detection Systems inoperablebe in at least HOT STANDBY within the next 6 hours and in COLDSHUTDOWN within the following 30 hours.

c. With a. and c. of the above required Leakage Detection Systemsinoperable:

1. Restore either Monitoring System (a. or c.) to OPERABLEstatus within 72 hours and

2. Obtain and analyze a grab sample of the containmentatmosphere for gaseous and particulate radioactivity atleast once per 24 hours, and

3. Perform a Reactor Coolant System water inventory balance perSurveillance Requirement 4.4.6.2.1.d at least once per 8hours

Otherwise, be in at least HOT STANDBY within the next 6 hours andin COLD SHUTDOWN within the following 30 hours.

Not required to be performed until 12 hours after establishment of steady-state operation.



OPERATIONAL LEAKAGE


3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE.

c. 150 gallons per day primary-to-secondary leakage through any onesteam generator,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 31 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of2235 ± 20 psig. and

f. The maximum allowable leakage of any Reactor Coolant System-Pressure Isolation Valve shall be as specified in Table 3.4-1 at apressure of 2235 ± 20 psig."

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

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, or with primary-to-secondaryleakage not within limit, be in at least HOT STANDBY within6 hours and in COLD SHUTDOWN within the following 30 hours.

b. With any Reactor Coolant System operational leakage greater thanany one of the above limits, excluding primary-to-secondaryleakage, PRESSURE BOUNDARY LEAKAGE and leakage from ReactorCoolant System Pressure Isolation Valves, reduce the leakage rateto within limits within 4 hours or be in at least HOT STANDBYwithin the next 6 hours and in COLD SHUTDOWN within the following30 hours.

c. With any Reactor Coolant System Pressure Isolation Valve leakagegreater than the limit specified in Table 3.4-1, isolate the highpressure portion of the affected system from the low pressureportion within 4 hours by use of at least two closed manual ordeactivated automatic valves, or be in at least HOT STANDBY withinthe next 6 hours and in COLD SHUTDOWN within the following30 hours.

"Test pressures less than 2235 psig but greater than 150 psig are allowed.Observed leakage shall be adjusted by multiplying the observed leakage by thesquare root of the quotient of 2235 divided by the test pressure.



OPERATIONAL LEAKAGE

SURVEILLANCE REQUIREMENTS'

4.4.6.2.1 Reactor Coolant System operational leakages shall be demonstratedto be within each of the above limits by:

a. Monitoring the containment Airborne Gaseous or ParticulateRadioactivity Monitor at least once per 12 hours:

b. Monitoring the containment sump inventory and Flow MonitoringSystem at least once per 12 hours:

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pumpseals when the Reactor Coolant System pressure is 2235 ± 20 psigat least once per 31 days with the modulating valve fully open.The provisions of Specification 4.0.4 are not applicable for entryinto MODE 3 or 4:

d. Performance of a Reactor Coolant System water inventory balance atleast once per 72 hours*; and

e. Monitoring the Reactor Head Flange Leakoff System at least onceper 24 hours.

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified inTable 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be withinits limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLDSHUTDOWN for 7 days or more and if leakage testing has not beenperformed in the previous 9 months,

c. Prior to returning the valve to service following maintenance,repair or replacement work on the valve, and

d. Within 24 hours following valve actuation due to automatic ormanual action or flow through the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3or 4.

4.4.6.2.3 Primary-to-secondary leakage shall be verified < 150 gallons perday through any one steam generator at least once per 72 hours

Not required to be performed until 12 hours after establishment of steady-state operation. Not applicable to primary-to-secondary leakage.

Not required to be performed until 12 hours after establishment of steady-state operation.




1. Steam Generator. (SG) Proqram

A Steam Generator Program shall be established and implemented toensure that steam generator (SG) tube integrity is maintained. Inaddition, the Steam Generator Program shall include the followingprovisions:

1. Provisions for condition monitoring assessments. Conditionmonitoring assessment means an evaluation of the "as found"condition of the tubing with respect to the performance criteriafor structural integrity and accident induced leakage. The "asfound" condition refers to the condition of the tubing during aSG inspection outage, as determined from the inserviceinspection results or by other means, prior to the plugging oftubes. Condition monitoring assessments shall be conductedduring each outage during which the SG tubes are inspected orplugged to confirm that the performance criteria are being met.

2. Performance criteria for SG tube integrity. Steam generatortube integrity shall be maintained by meeting the performancecriteria for tube structural integrity, accident inducedleakage, and operational leakage.

a) Structural integrity performance criterion. All inserviceSG tubes shall retain structural integrity over the fullrange of normal operating conditions (including startup,operation in the power range, HOT STANDBY, and cooldown anda anticipated transients included in the designspecification) and design basis accidents. This includesretaining a safety factor of 3.0 (3deltaP) against burstunder normal steady-state full power operation primary-to-secondary pressure differential and a safety factor of 1.4against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the aboverequirements, additional loading conditions associated withthe design basis accidents, or combination of accidents inaccordance with the design and licensing basis, shall alsobe evaluated to determine if the associated loads contributesignificantly to burst or collapse. In the assessment oftube integrity, those loads that do significantly affectburst or collapse shall be determined and assessed incombination with the loads due to pressure with a safetyfactor of 1.2 on the combined primary loads and 1.0 on axialsecondary loads.

b) Accident induced leakage performance criterion. Theprimary-to-secondary accident induced leakage rate for anydesign basis accident, other than a SG tube rupture, shallnot exceed the leakage rate assumed in the accident analysisin terms of total leakage rate for all SGs and leakage ratefor an individual SG. Accident induced leakage is not toexceed 1 gpm total for all three SGs.

c) The operation leakage performance criterion is specified inLCO 3.4.6.2, "Reactor Coolant System Operational Leakage."

SHEARON HARRIS - UNIT 1 6-19d Amendment No.



3. Provisions for SG tube repair criteria. Tubes found byinservice inspection to contain flaws with depth equal to orexceeding 40% of the nominal tube wall thickness shall beplugged.

4. Provisions for SG tube inspections. Periodic SG tubeinspections shall be performed. The number and portions of thetubes inspected and methods of inspection shall be performedwith the objective of detecting flaws of any type (e.g.,volumetric flaws, axial and circumferential cracks) that may bepresent along the length of the tube, from tube-to-tubesheetweld at the tube inlet to the tube-to-tubesheet weld at the tubeoutlet, and that may satisfy the applicable tube repaircriteria. The tube-to-tubesheet weld is not part of the tube.In addition to meeting the requirements of 4a, 4b, and 4c below,the ins ection scope, inspection methods and inspectionintervals shall be such as to ensure that SG tube integrity ismaintained until the next SG inspection. An assessment ofdegradation shall be performed to determine the type andlocation of flaws to which the tubes may be susceptible and.based on this assessment, to determine which inspection methodsneed to be employed and at what locations.

a) Inspect 100% of the tubes in each SG during the firstrefueling outage following SG replacement.

b) Inspect 100% of the tubes at sequential periods of 144, 108,72. and thereafter, 60 effective full power months. The-first sequential period shall be considered to begin afterthe first inservice inspection of the SGs. In addition.inspect 50% of the tubes by the refueling outage nearest themidpoint of the period and the remaining 50% by therefueling outage nearest the end of the period. No SG shalloperate for more than 72 effective full power months orthree refueling outages (whichever is less) without beinginspected.

SHEARON HARRIS - UNIT 1 6-19e Amendment No.


PROCEDURES AND PROGRAMS'(Cbntinued)

c) If crack indications are found in any SG tube, then the nextinspection for each SG for the degradation mechanism thatcaused the crack indication shall not exceed 24 effectivefull power months or one refueling outage (whichever isless). If definitive information, such as from examinationof a pulled tube, diagnostic non-destructive testing, orengineering evaluation indicates that a crack-likeindication is not associated with a crack(s), then theindication need not be treated as a crack.

5. Provisions for monitoring operational primary-to-secondaryleakage.

SHEARON HARRIS - UNIT 1 6-19f Amendment No.


6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

n. Mechanical Design Methodologies

XN-NF-81-58(P)(A). "RODEX2 Fuel Rod Thermal-Mechanical ResponseEvaluation Model," approved version as specified in the COLR.

ANF-81-58(P)(A). "RODEX2 Fuel Rod Thermal Mechanical ResponseEvaluation Model," approved version as specified in the COLR.

XN-NF-82-06(P)(A). "Qualification of Exxon Nuclear Fuel forExtended Burnup," approved version as specified in the COLR.

ANF-88-133(P)(A). "Qualification of Advanced Nuclear Fuels' PWRDesign Methodology for Rod Burnups of 62 GWd/MTU," approvedversion as specified in the COLR.

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/GadoliniaIrradiation Examination and Thermal Conductivity Results."approved version as specified in the COLR.

EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR FuelDesigns," approved version as specified in the COLR.

(Methodologies for Specification 3.2.1 - Axial Flux Difference.3.2.2 - Heat Flux Hot Channel Factor. and 3.2.3 - Nuclear EnthalpyRise Hot Channel Factor).

6.9.1.6.3 The core operating limits shall be determined so that allapplicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient andaccident analysis limits) of the safety analysis are met.

6.9.1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cyclerevisions or supplements, shall be provided, upon issuance for each reloadcycle, to the NRC Document Control Desk, with copies to the RegionalAdministrator and Resident Inspector.

6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT

A report shall be submitted within 180 days after the initial entry into HOTSHUTDOWN following completion of a steam generator tube inspection performedin accordance with Specification 6.8.4.1. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for eachdegradation mechanism,

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

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

SHEARON HARRIS - UNIT 1 6-24c Amendment No.


6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT (Continued)

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

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

SPECIAL REPORTS

6.9.2 Special reports shall be submitted to the NRC in accordance with1OCFR50.4 within the time period specified for each report.

6.10 DELETED

(PAGE 6-25 DELETED)

SHEARON HARRIS - UNIT 1 6-24d Amendment No. I



REQUEST FOR LICENSE AMENDMENTPROPOSED TECHNICAL SPECIFICATIONS (TS) BASES CHANGES

(FOR INFORMATION ONLY)

PROPOSED TECHNICAL SPECIFICATIONS (TS) BASES CHANGES(FOR INFORMATION ONLY)

Page A4-1 of 18


BASES

RELIEF VALVES (Continued)

Surveillance Requirement 4.4.4.3 provides assurance of operability of theaccumulators and that the accumulators are capable of supplying sufficient airto operate the PORV(s) if they are needed for RCS pressure control and normalair and nitrogen systems are not available.

Surveillance Requirement 4.4.4.2 addresses the block valves. The block valvesare exempt from the surveillance requirements to cycle the valves when theyhave been closed to comply with ACTION statements "b" or "c". This precludesthe need to cycle the valves with a full system differential pressure or whenmaintenance is being performed to restore an inoperable PORV to OPERABLEstatus.

3/4.4.5 STEAM GENERATO

The Surveillance Requirements for inspection of the steam generator tubes en-sure that the structural integrity of this portion of the RCS will bemaintained. The program for inservice inspection of steam generator tubes isbased on a modification of Regulatory Cuide 1.83, Revision 1. Inserviceinspection of steam generator tubing is essential in order to maintainsurveillance of the conditions of the tubes in the event that there is evidenceof mechanical damage or progressive degradation due to design, manufacturingerrors, or inservice conditions that lead to corrosion. Inservice inspectionof steam generator tubing also provides a means of characterizing the natureand cause of any tube degradation so that corrective measures can be taken.

SHEARON HARRIS - UNIT 1 B 3/4 4-2b Amendment No.E73

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INSERT 3/4.4.5 BASES


BASES


Background

Steam generator (SG) tubes are small diameter, thin walled tubes that carry primarycoolant through the primary to secondary heat exchangers. The SG tubes have a numberof important safety functions. SG tubes are an integral part of the reactor coolant pressureboundary (RCPB) and, as such, are relied on to maintain the primary system's pressureand inventory. The SG tubes isolate the radioactive fission products in the primarycoolant from the secondary system. In addition, as part of the RCPB, the SG tubes areunique in that they act as the heat transfer surface between the primary and secondarysystems to remove heat from the primary system. This Specification addresses only theRCPB integrity function of the SG. The SG heat removal function is addressed by LCO3.4.1.1, "Reactor Coolant Loops and Coolant Circulation, Startup and Power Operation,"'LCO 3.4.1.2, "Reactor Coolant System, Hot Standby," LCO 3.4.1.3, "Reactor CoolantSystem, Hot Shutdown," and LCO 3.4.1.4.1, "Reactor Coolant System, Cold Shutdown-Loops Filled."

SG tube integrity means that the tubes are capable of performing their intended RCPBsafety function consistent with the licensing basis, including applicable regulatoryrequirements.

SG tubing is subject to a variety of degradation mechanisms. SG tubes may experiencetube degradation related to corrosion phenomena, such as wastage, pitting, intergranularattack, and stress corrosion cracking, along with other mechanically phenomena such asdenting and wear. These degradation mechanisms can impair tube integrity if they are notmanaged effectively. The SG performance criteria are used to manage SG tubedegradation.

Specification 6.8.4.1, "Steam Generator Program," requires that a program be establishedand implemented to ensure that SG tube integrity is maintained. Pursuant to Specification6.8.4.1, tube integrity is maintained when the SG performance criteria are met. There arethree SG performance criteria: structural integrity, accident induced leakage, andoperational leakage. The SG performance criteria are described in Specification 6.8.4.1.Meeting the SG performance criteria provides reasonable assurance of maintaining tubeintegrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the SteamGenerator Program Guidelines (Reference 1).

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BASES

3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY (continued)

Applicable Safety Analyses

The steam generator tube rupture (SGTR) accident is the limiting design basis event forSG tubes and avoiding an SGTR is the basis for this Specification. The analysis of aSGTR event assumes a bounding primary-to-secondary leakage rate equal to 1 gpm, plusthe leakage rate associated with a double-ended rupture of a single tube. The accidentradiological analysis for a SGTR assumes the ruptured SG secondary fluid is releaseddirectly to the atmosphere due to a failure of the PORV in the open position.

The analysis for design basis accidents and transients other than a SGTR assume the SGtubes retain their structural integrity (i.e., they are assumed not to rupture). In theseanalyses the steam discharge to the atmosphere is based on the total primary-to-secondaryleakage from all SGs of 1 gpm, or is assumed to increase to 1 gpm as a result of accidentinduced conditions. For accidents that do not involve fuel damage, the primary coolantactivity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the limits in LCO3.4.8, "Reactor Coolant System Specific Activity." For accidents that assume fueldamage, the primary coolant activity is a function of the amount of activity released fromthe damaged fuel. The dose consequences of these events are within the limits of10 CFR 50.67 (Reference 2).

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

Limiting Condition for Operation (LCO)

The LCO requires that SG tube integrity be maintained. The LCO also requires that allSG tubes that satisfy the repair criteria be plugged in accordance with the SteamGenerator Program.

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

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 ofthe tube.

of 7


BASES


A SG tube has tube integrity when it satisfies the SG performance criteria. The SGperformance criteria are defined in Specification 6.8.4.1 and describe acceptable SG tubeperformance. The Steam Generator Program also provides the evaluation process fordetermining 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 tomeet the LCO.

The structural integrity performance criterion provides a margin of safety against tubeburst or collapse under normal and accident conditions, and ensures structural integrity ofthe SG tubes under all anticipated transients included in the design specification. Tubeburst is defined as, "The gross structural failure of the tube wall. The condition typicallycorresponds to an unstable opening displacement (e.g., opening area increased inresponse to constant pressure) accompanied by ductile (plastic) tearing of the tubematerial at the ends of the degradation." Tube collapse is defined as, "For the loaddisplacement curve for a given structure, collapse occurs at the top of the load versesdisplacement curve where the slope of the curve becomes zero." The structural integrityperformance criterion provides guidance on assessing loads that have a significant effecton burst or collapse. In that context, the term "significant" is defined as "An accidentloading condition other than differential pressure is considered significant when theaddition of such loads in the assessment of the structural integrity performance criterioncould cause a lower structural limit or limiting burst/collapse condition to beestablished." For tube integrity evaluations, except for circumferential degradation, axialthermal loads are classified as secondary loads. For circumferential degradation, theclassification of axial thermal loads as primary or secondary loads will be evaluated on acase-by-case basis. The division between primary and secondary classifications will bebased on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube notexceed the yield strength for all ASME Code, Section III, Service Level A, (normaloperating conditions) and Service Level B, (upset conditions) transients included in thedesign specification. This includes safety factors and applicable design basis loads basedon ASME Code, Section III, Subsection NB (Reference 3) and Draft Regulatory Guide1.121 (Reference 4).

of 7


BASES


The accident induced leakage performance criterion ensures that the primary-to-secondaryleakage 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 total from all SGs. The accident induced leakage rate includes anyprimary-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 SGtube conditions during plant operation. The limit on operational leakage is contained in

LCO 3.4.6.2 and limits primary-to-secondary leakage through any one SG to 150 gpd.This limit is based on the assumption that a single crack leaking this amount would notpropagate 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

MODES 1,2, 3, or 4.

RCS 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, 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 ACTIONSprovide 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.

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BASES


a. The condition applies if it is discovered that one or more SG tubes examined inan Inservice Inspection satisfy the tube repair criteria but were not plugged inaccordance with the Steam Generator Program as required by SurveillanceRequirement 4.4.5.2. An evaluation of SG tube integrity of the affected tube(s)must be made. Steam generator tube integrity is based on meeting the SGperformance criteria described in the Steam Generator Program. The SG repaircriteria define limits on SG tube degradation that allow for flaw growthbetween inspections while still providing assurance that the SG performancecriteria will continue to be met. In order to determine if a SG tube that shouldhave been plugged has tube integrity, an evaluation must be completed thatdemonstrates that the SG performance criteria will continue to be met until thenext refueling outage or SG tube inspection. The tube integrity determinationis based on the estimated condition of the tube at the time the situation isdiscovered and the estimated growth of the degradation prior to the next SGtube inspection. If it is determined that tube integrity is not being maintained,condition (b) applies.

An allowed completion time of seven days is sufficient to complete theevaluation while minimizing the risk of plant operation with a SG tube thatmay not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, theACTION statement allows plant operation to continue until the next refuelingoutage or SG inspection provided the inspection interval continues to besupported by an operational assessment that reflects the affected tubes.However, the affected tube(s) must be plugged prior to entering HOTSHUTDOWN following the next refueling outage or SG inspection. Thisallowed completion time is acceptable since operation until the next inspectionis supported by the operational assessment.

b. If the required actions and associated completion times of condition (a) are notmet or if SG tube integrity is not being maintained, the reactor must be broughtto HOT STANDBY within 6 hours and COLD SHUTDOWN within 36 hours.

The allowed completion times are reasonable, based on operating experience,to reach the desired plant conditions from full power conditions in an orderlymanner and without challenging plant systems.

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BASES


Surveillance Requirements

4.4.5.1 During shutdown periods the SGs are inspected as required by this SR and theSteam Generator Program. NEI 97-06, Steam Generator Program Guidelines(Reference 1), and its referenced EPRI Guidelines, establish the content of the SteamGenerator Program. Use of the Steam Generator Program ensures that the inspection isappropriate 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 SGtubes. The purpose of the condition monitoring assessment is to ensure that the SGperformance criteria have been met for the previous operating period.

The Steam Generator Program determines the scope of the inspection and the methodused 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) isa function of existing and potential degradation locations. The Steam Generator Programalso specifies the inspection methods to be used to find potential degradation. Inspectionmethods are a function of degradation morphology, nondestructive examination (NDE)technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.4.5.1. The frequency isdetermined by the operational assessment and other limits-in the SG examinationguidelines (Reference 5). The Steam Generator Program uses information on existingdegradations and growth rates to determine an inspection frequency that providesreasonable assurance that the tubing will meet the SG performance criteria at the nextscheduled inspection. In addition, Specification 6.8.4.1 contains prescriptiverequirements concerning inspection intervals to provide added assurance that the SGperformance criteria will be met between scheduled inspections.

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BASES


4.4.5.2 During an SG inspection, any inspected tube that satisfies the Steam GeneratorProgram repair criteria is removed from service by plugging. The tube repair criteriadelineated in Specification 6.8.4.1 are intended to ensure that tubes accepted for continuedservice satisfy the SG performance criteria with allowance for error in the flaw sizemeasurement and for future flaw growth. In addition, the tube repair criteria, inconjunction with other elements of the Steam Generator Program, ensure that the SGperformance criteria will continue to be met until the next inspection of the subjecttube(s). Reference I provides guidance for performing operational assessments to verifythat the tubes remaining in service will continue to meet the SG performance criteria.

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 criteriaare plugged prior to subjecting the SG tubes to significant primary-to-secondary pressuredifferential.

References

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

2. 10 CFR 50.67

3. ASME Boiler and Pressure Vessel Code, Section III, Subsection NB

4. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam GeneratorTubes," August 1976

5. EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines"

REACTOR COOLANT SYSTEM!

BASES

STEAM G lI[,Rlý 1

The plant is expecte~d to be operated in a manner such that the secondarycoolant will be maintained within those chemistry limits found to resull innegligible corrosion of the steam generator tubes_. If the secondary coolantchemistry is not maintained within these limits, localized corrosion maylikely result in stress corrosion cracking. The extent of crackin9 duringplant operation would be limited by the limitation of steam genera or tube1eakage between the Reactor Coolant System and the Secondary Coolant System(reactor-to-secondary leakage = 150 gallons per day per steam generator).Cracks having a reactor-to-secondary leakage less thanUthis limit duringoperation will have an adequate margin of safety to withstand the loadsimposed during normal operation and by postulated accidents. Operating plantshave demonstrated that reactor-to-secondary leakage of 150 gallons per day persteam generator can readily be detected by radiation monitors of steamgenerator blowdown. Leakage in excess of-this limit will require plantshutdown and an unscheduled inspection, during which the leaking tubes will belocated and plugged.

Wastaqe-type defects are unlikely with proper chemistry treatment of thesecondary coolant. However even if a defect should develop in service. itwill be found during scheduled inservice steam generator tube examinations.Plugging will be required for all tubes with imperfections exceeding theplugging limit of 40% of the tube nominal wall thickness. Steam generatorube inspections of operating plants have demonstrated the capabilitb to

reliably detect degradation that has penetrated 20% of the original ude wallthickness.

Whenever the results of any steam generator tubing inservice inspection fallinto Category C-3. these results will be reported to the Commission in aSpecial Report pursuant to Specification 4.4.5.5.c within 30 days and prior toresumption of plant operation. Such cases will be considered by theCommission on a case-by-case basis and may result in a requirement foranalysis, laboratory examinations, tests, additional eddy-current inspection.and revision of the Technical Specifications, if nec ary."


3/4.4.6.1 LEAKAGE DETECTION SYSTEMS

The RCS Leakage Detection Systems required by this specification are providedto monitor and detect leakage from the reactor coolant pressure boundary.These Detection .Systems are consistent with the-recommendations of RegulatoryGuide 1.45. "Reactor Coolant Pressure Boundary Leakage Detection Systems,.May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE..PPRESSURE-BOUDAYLEAKAGE of-any-magnit eis un'acceptable-sinc-e it may be--N/(indicative of an impen~ding gross failu~re of the'pressure, boundary. ,Thyer'efore,.Sthe presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly|

HplaceA in COLD SHUT1DOWN.

SHEARON HARRIS - UNIT 1 B ,3/4 4-3 Amendment No.e

,,',


BASES

OPERATIONAL LEAKAGE (Continued)

Slow to ensure early detection of additional leakage.

The total steam generator tube leakage limit of 1 .1m for all steam generatorsensures that the dosage contribution from the tube eakage will be limited toa small fraction of 10 CFR Part 100 dose guideline values in the event ofeither a steam generator tube rupture or steam line break. The l~gpm limit isconsistent with the assumptions used in the analysis of these accidents. The150 gpd leakage limit per steam generator ensures that steam generator tubeintegrity is maintained in the event of a main steam line rupture or underLOCA conditions.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limitedamount of leakage from known sources whose presence will not interfere withthe detection-of UNIDENTIFIED LEAK by the Leakage Detection Systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow sup-plied to the reactor coolant pump seals exceeds 31 gpm with the modulatingvalve in the supply line fully open at a nominal RCS pressure of 2235 psig.This limitation ensures that in the event of a LOCA, the safety injection flowwill not be less than assumed in the safety analyses.

The maximum allowable leakage from any RCS pressure isolation valve issufficiently low to ensure early detection of possible in-series check valvefailure. It is apparent that when pressure isolation is provided by twoin-series check valves and when failure of one valve in the pair can goundetected for a substantial length of time, verification of valve integrityis required. Since these valves are important in preventingoverpressurization and rupture of the ECCS low pressure piping which couldresult in a LOCA that bypasses containment, these valves should be testedperiodically to ensure ow probability of gross failure.

The Surveillance Requirements for RCS pressure isolation valves provide addedassurance of valve integrity thereby reducing the probability of gross valvefailure and consequent intersystem LOCA. Leakage from the RCS pressureisolation valve is IDENTIFIED LEAKAGE and will be considered as a portion ofthe allowed limit.

3/4.4.7 CHEMISTRY

The limitations on Reactor Coolant System chemistry ensure that corrosion ofthe Reactor Coolant System is minimized and reduces the potential for ReactorCoolant System leakage or failure due to stress corrosion. Maintaining thechemistry within the Steady-State Limits provides adequate corrosionprotection to ensure the structural integrity of the Reactor CoolantSystem over the life of the plant. The associated effects ofexceeding the oxygen, chloride, and fluoride limits are time andtemperature dependent. Corrosion studies show

SHEARON HARRIS - UNIT 1 B 3/4 4-4 Amendment No

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INSERT 3/4.4.6.2 BASES


BASES

3/4.4.6.2 OPERATIONAL LEAKAGE

Background

Components that contain or transport the coolant to or from the reactor core make up thereactor coolant system (RCS). Component joints are made by welding, bolting, rolling,or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactorcoolant leakage, through either normal operational wear or mechanical deterioration. Thepurpose of the RCS Operational leakage LCO is to limit system operation in the presenceof leakage from these sources to amounts that do not compromise safety. This LCOspecifies the types and amounts of leakage.

10 CFR 50, Appendix A, GDC 30 (Reference 1), requires means for detecting and, to theextent practical, identifying the source of reactor coolant leakage. Regulatory Guide 1.45(Reference 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS leakage varies widely depending on its source, rate, andduration. Therefore, detecting and monitoring reactor coolant leakage into thecontainment area is necessary. Quickly separating the identified leakage from theunidentified leakage is necessary to provide quantitative information to the operators,allowing them to take corrective action should a leak occur that is detrimental to thesafety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems thatcannot be made 100% leaktight. Leakage from these systems should be detected, located,and isolated from the containment atmosphere, if possible, to not interfere with RCSleakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) fromdegradation and the core from inadequate cooling, in addition to preventing the accidentanalyses radiation release assumptions from being exceeded. The consequences ofviolating this LCO include the possibility of a loss of coolant accident (LOCA).

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3/4.4.6.2 OPERATIONAL LEAKAGE (continued)

Applicable Safety Analyses

Except for primary-to-secondary leakage, the safety analyses do not address operationalleakage. However, other operational leakage is related to the safety analyses for aLOCA; the amount of leakage can affect the probability of such an event. The safetyanalysis for an event resulting in steam discharge to the atmosphere assumes thatprimary-to-secondary leakage from all steam generators is I gpm. The LCO requirementto limit primary-to-secondary leakage through any one steam generator to less than orequal to 150 gpd is significantly less than the conditions assumed in the safety analysis.

Primary-to-secondary leakage is a factor in the dose releases outside containmentresulting from a steam line break (SLB) accident or a steam generator tube rupture(SGTR). The leakage contaminates the secondary fluid.

The FSAR analysis for a SGTR assumes the contaminated secondary fluid is releaseddirectly to the atmosphere due to a failure of the PORV in the open position and willcontinue atmospheric release until the time that the PORV can be isolated. The FSARanalysis for the SLB assumes that the SG with the failed steam line boils dry releasing allof the iodine directly to the environment and that iodine carried over to the faulted SG bytube leaks are also released directly to the environment until the RCS has cooled to below212T. The dose consequences resulting from the SGTR and the SLB accidents arewithin the limits defined in 10 CFR 50.67.

The RCS operational leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

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Limiting Condition for Operation (LCO)

Reactor Coolant System operational leakage shall be limited to:

a. PRESSURE BOUNDARY LEAKAGE

No PRESSURE BOUNDARY LEAKAGE is allowed, being indicative of materialdeterioration. Leakage of this type is unacceptable as the leak itself could causefurther deterioration, resulting in higher leakage. Violation of this LCO could resultin continued degradation of the Reactor Coolant Pressure Boundary. Leakage pastseals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b. UNIDENTIFED LEAKAGE

One gallon per minute (gpm) of UNIDENTIFED LEAKAGE is allowed as areasonable minimum detectable amount that the containment air monitoring andcontainment sump level monitoring equipment can detect within a reasonable timeperiod. Violation of this LCO could result in continued degradation of the ReactorCoolant Pressure Boundary, if the leakage is from the pressure boundary.

C. Primary-to-Secondary Leakage Through Any One Steam Generator

The limit of 150 gpd per steam generator is based on the operational leakageperformance criterion in NEI 97-06, Steam Generator Program Guidelines(Reference 3). The Steam Generator Program operational leakage performancecriterion in NEI 97-06 states, "The RCS operational primary-to-secondary leakagethrough any one steam generator shall be limited to 150 gallons per day." The limitis based on operating experience with steam generator tube degradation mechanismsthat result in tube leakage. The operational leakage rate criterion is conjunction withthe implementation of the Steam Generator Program is an effective measure forminimizing the frequency of steam generator tube ruptures.

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d. IDENTIFIED LEAKAGE

Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because leakageis from known sources that do not interfere with detection of UNIDENTIFED

LEAKAGE and is well with the capability of the Reactor Coolant System Makeup

System. IDENTIFIED LEAKAGE includes leakage to the containment fromspecifically known and located sources, but does not include PRESSUREBOUNDARY LEAKAGE or CONTROLLED LEAKAGE. Violation of this LCO

could result in continued degradation of a component or system.

e. CONTROLLED LEAKAGE

The CONTROLLED LEAKAGE limitation restricts operation when the total flow

supplied to the reactor coolant pump seals exceeds 31 gpm with the modulating

valve in the supply line fully open at a nominal RCS reassure of 2235 psig. This

limitation ensures that in the event of a LOCA, the safety injection flow will not be

less than assumed in the accident analysis.

f. Reactor Coolant System Pressure Isolation Valve Leakage

The maximum allowable leakage from any RCS pressure isolation valve is

sufficiently low to ensure early detection of possible in-series check valve failure. Itis apparent that when pressure isolation is provided by two in-series check valves

and when failure of one valve in the pair can go undetected for a substantial length

of time, verification of valve integrity is required. Since these valves are important

in preventing overpressurization and rupture of the ECCS low pressure piping which

could result in a LOCA that bypasses containment, these valves should be testedperiodically to ensure low probability of gross failure.

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Applicability

In MODES 1, 2, 3, and 4, the potential for RCPB leakage is greatest when the RCS ispressurized.

In MODES 5 and 6, leakage limits are not required because the reactor coolant pressure isfar lower, resulting in lower stresses and reduced potentials for leakage.

ACTIONS

a. If any PRESSURE BOUNDARY LEAKAGE exists, or primary-to-secondaryleakage is not within limit, the reactor must be brought to HOT STANDBY within 6hours and COLD SHUTDOWN within the next 30 hours. This action reduces theleakage and also reduces the factors that tend to degrade the pressure boundary.

The allowed outage times are reasonable, based on operating experience, to reachthe required plant conditions from full power conditions in an orderly manner andwithout challenging plant systems. In COLD SHUTDOWN, the pressure stressesacting on the RCPB are much lower, and further deterioration is much less likely.

b. UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or CONTROLLEDLEAKAGE in excess of the LCO limits must be reduced to within the limits within4 hours. This allows time to verify leakage rates and either identifyUNIDENTIFIED LEAKAGE or reduce leakage to within limits before the reactormust be shut down. This action is necessary to prevent further deterioration of theRCPB.

Surveillance Requirements

4.4.6.2.1 Verifying RCS leakage to be within the LCO limits ensures the integrity ofthe RCPB is maintained. PRESSURE BOUNDARY LEAKAGE would at first appear asUNIDENTIFIED LEAKAGE and can only be positively identified by inspection. Itshould be noted that leakage past seals and gaskets is not PRESSURE BOUNDARYLEAKAGE. UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determinedby performance or a RCS water inventory balance.

I-. I

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The RCS water inventory balance must be met with the reactor at steady-state operatingconditions (stable pressure, temperature, power level, pressurizer and makeup tank levels,makeup letdown, and RCP seal injection and return flows). The surveillance is modifiedby a note. The note states that this SR is not required to be performed until 12 hours afterestablishing steady-state operation. The 12 hour allowance provides sufficient time tocollect and process all necessary data after stable plant conditions are established.

Steady-state operation is required to perform a proper water inventory balance sincecalculations during maneuvering are not useful, so the SR is not required to be performeduntil 12 hours after establishing steady-state operation. For RCS operational leakagedetermination by water inventory balance, steady-state is defined as stable RCS pressure,temperature, power level, pressurizer and makeup tank levels, makeup and letdown, andRCP seal injection and return flows.

An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIEDLEAKAGE is provided by the automatic systems that monitor containment atmosphereradioactivity and reactor cavity sump level. It should be noted that leakage past seals andgaskets is not PRESSURE BOUNDARY LEAKAGE. These leakage detection systemsare specified in LCO 3.4.6.1, "Reactor Coolant System Leakage Detection Systems."

Part (d) notes that this SR is not applicable to primary-to-secondary leakage. This isbecause leakage of 150 gallons per day cannot be measured accurately by an RCS waterinventory balance.

The 72-hour frequency is a reasonable interval to trend leakage and recognizes theimportance of early leakage detection in the prevention of accidents.

4.4.6.2.2 The Surveillance Requirements for RCS Pressure Isolation Valves provideadded assurance of valve integrity thereby reducing the probability of gross valve failureand consequent intersystem LOCA. Leakage from the RCS pressure isolation valve isIDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

4.4.6.2.3 This SR verifies that primary-to-secondary leakage is less than or equal to150 gpd through any one steam generator. Satisfying the primary-to-secondary leakagelimit ensures that the operational leakage performance criterion in the Steam GeneratorProgram is met. If this Surveillance Requirement is not met, compliance with LCO 3.4.5should be evaluated. The 150-gpd limit is measured at room temperature as described inReference 2. The operational leakage rate limit applies to leakage through any one steamgenerator. If it is not practical to assign the leakage to an individual steam generator, allthe primary-to-secondary leakage should be conservatively assumed to be from one steamgenerator.

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The surveillance is modified by a note, which states that the Surveillance is not requiredto be performed until 12 hours after establishment of steady-state operation. For RCSprimary-to-secondary leakage determination, steady-state is defined as stable RCSpressure, temperature, power level, pressurizer and makeup tank levels, makeup andletdown, and RCP seal injection and return flows.

The frequency of 72 hours is a reasonable interval to trend primary-to-secondary leakageand recognizes the importance of early leakage detection in the prevention of accidents.The primary-to-secondary leakage is determined using continuous process radiationmonitors or radiochemical grab sampling in accordance with the EPRI guidelines(Reference 4).

References

1. 10 CFR 50, Appendix A, GDC 30

2. Regulatory Guide 1.45, May 1973

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

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

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