Text

License Amendment Request Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity
	ML061520433
Person / Time
Site:	Harris
Issue date:	05/23/2006
From:	Gannon C; Carolina Power & Light Co, Progress Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	HNP-06-060
	Download: ML061520433 (67)
	v • d • e

.,

Progress Energy Cornelius J. Cannon. Jr.

Vice President Harris Nuclear Plant Progress Energy Carolinas. Inc.

MAY 2 3 2006 Serial: HNP-06-060 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 LICENSE AMENDMENT REQUEST APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY Ladies and Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.90, Carolina Power 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 generator tube integrity. The change is consistent with NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity." The'availability of this TS improvement was announced in the Federal Register on May 6, 2005 (70 FR 24126) as part of the consolidated line item improvement process (CLIIP).

In the HNP response (Serial: HNP-06-029 dated February 16, 2006) to NRC Generic Letter (GL) 2006-01, "Steam Generator Tube Integrity and Associated Technical Specifications," HNP committed to submit a request to modify the Steam Generator portion 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 of applicability.

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 Plant P.D.Box 165 New Hill, NC 27562 T> 919.362.2502 F> 919.362.2095

HNP-06-060 Page 2 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. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION 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 the NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision

4. The availability of this technical specification improvement was announced in the Federal Register on May 6, 2005 as part of the consolidated line item improvement process (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 discussed in the NRC's model safety evaluation (SE), adoption of the revised TS Bases associated with TSTF-449, Revision 4 is an integral part of implementing this TS improvement. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program.

3.0 BACKGROUND

The background for this application is 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.

4.0 REGULATORY REQUIREMENTS AND GUIDANCE The applicable regulatory requirements and guidance associated with this application are 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

Attachment 1 to SERIAL: HNP-06-060 SHEARON! H--ARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION AND ASSESSMENT

5.0 TECHNICAL ANALYSIS

HNP has reviewed the model safety evaluation (SE) published on March 2, 2005 (70 FR 10298) as part of the CLIIP Notice for Comment. This model SE included the NRC staff's SE, the supporting information provided to support TSTF-449, and the changes associated with revision 4 to TSTF-449. HNP has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to 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 regulatory requirements and guidance was provided in 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.

6.1 Verification and Commitments The following information is provided to support the NRC staff's review of this amendment 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 currently installed SGs Tubing Material Inconel 690TT Number of tubes per SG 6307 Number and percentage of tubes SG A SG B SG C plugged in each SG 3 (0.05%) 1 (0.02%) 3 (0.05%)

Number of tubes repaired in each Not Applicable SG Degradation mechanism(s) No active degradation mechanisms have been identified identified.

Current primary-to-secondary - 150 gpd through any one steam generator leakage 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 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION AND ASSESSMENT Request Response Approved Alternate Tube Repair Not Applicable Criteria (ARC)

Approved SG Tube Repair Not Applicable Methods Performance criteria for accident 1 gpm primary-to-secondary leakage is leakage assumed in the licensing basis accident analysis. Assumed temperature condition is room temperature.

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION HNP has reviewed the proposed no significant hazards consideration determination published on March 2, 2005 (70 FR 10298) as part of the CLIIP. HNP has concluded that the proposed determination presented in the notice is applicable to HNP and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a).

8.0 ENVIRONMENTAL EVALUATION HNP has reviewed the environmental evaluation included in the model SE published on March 2, 2005 (70 FR 10298) as part of the CLIIP. HNP has concluded that the staffs findings presented in that evaluation are applicable to HNP and the evaluation is hereby incorporated by reference for this application.

Page A1-3 of 4

Attachrient 1 to SERIAL: HNP-06-060 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION AND ASSESSMENT 9.0 PRECEDENT This application is being made in accordance with the CLIIP. HNP is not proposing variations or deviations from the TS changes described in TSTF-449, Revision 4, or the NRC staff's model SE published on March 2, 2005 (70 FR 10298). The following differences from the improved Standard Technical Specifications (ITS) changes described in TSTF-449, Revision 4 are necessary due to the non-ITS format of the HNP TS:

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

- The general format and numbering convention associated with the current TS for Limiting Conditions for Operation (LCOs), Actions, Surveillance Requirements (SRs) and Notes are retained. For example, the note below ACTIONS of LCO 3.4.20 of TSTF-449, Revision 4 has been moved to the bottom of the page as a footnote 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,'"Separate Condition entry is allowed for each SG tube."

- Terminology used in the current TS Actions is maintained such as: HOT STANDBY, 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 is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

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Attachment 2 to SERIAL: HNP-06-060 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT PROPOSED 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 4-13 TABLE 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 SURVEILLANCE REQUIREMENTS ........ ....... ........ ... ....... 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 ANALYSIS PROGRAM ........... ......................... 3/4 4-31 SHEARON HARRIS - UNIT I1 vii Amendment No.10 SI

INDEX 3.0/4.0 BASES SECTION PAGE 3/4.0 APPLICABILITY . . . . . ... . . . . .. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL .......... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS .......... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES ..... B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS ........ B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE ........... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR B 3/4 2-2a FIGURE B 3/4.2-1 (DELETED) ........... B 3/4 2-:3 3/4.2.4 QUADRANT POWER TILT RATIO ........ B 3/4 2-6 3/4.2.5 DNB PARAMETERS ........... B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ..... ................ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION .......... .B3/4 3-3 3/4.3.4 (DELETED) ...... .............. ..... B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION B 3/4 4-1 3/4.4.2 SAFETY VALVES ....... ...... ............ B 3/4 4-1 3/4.4.3 PRESSURIZER ...... .................. B 3/4 4-2 3/4.4.4 RELIEF VALVES .... __ _ _ _ B 3/4 4-2 3/4.4.5 STEAM GENERATO . - B 3/4 4-2b 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE . . . ..... B 3/4 4-3 3/4.4.7 CHEMISTRY ..... .................... B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY .... ......... ...... B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ............. B 3/4 4-6 SHEARON HARRIS - UNIT I xiii Amendment No. LlJ16

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 Annual-Reports . . . . . . . . . . . . . ... . . . . . . . . . . 6-20 Annual Radiological Environmental Operating Report . . . . . . .. 6-21

~6-22 ,

Annual Radioactive Effluent Release Report ..

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 concentration of each radionuclide in the reactor coolant atthe time of sampling, of the sum of the average beta and gamma energies per disintegration (MeV/d) for isotopes. with half-lives greater than 15 minutes. making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Set oint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential.

overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

EXCLUSION AREA BOUNDARY 1.14 The EXCLUSION AREA'BOUNDARY shall be that line beyond which the land is not 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 installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.17 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump 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 both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary 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 and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include 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 employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or tomake deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 1.20 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints. and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of'the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.3 and 6.9.1.4.

OPERABLE - OPERABILITY 1.21 A system, subsystem, train, component or device shall be OPERABLE or have 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 are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their, related support function(s).

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

PHYSICS TESTS 1.23 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation:

(1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 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 tyscomponen body, 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 inoperable generator(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 of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 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.. The inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators: the tubes selected for these inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes 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 generator shall include:

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

2. Tubes poeta in those areas where experience has indicated roblems. and 9 SHEARON HARRIS - UNIT 1 3;4 4-13

!I*

Amendment No.&

INSERT 3/4.4.5 REACTOR COOLANT SYSTEM 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 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 repair criteria and not plugged in accordance with the Steam Generator Program, within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

AND

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

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

4.4.5.2 Verify that each inspected steam generator tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to entering 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 selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

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

1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and

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

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

Categorv Inspection Results C-i Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective.

or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected 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 of steam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of Steam Generator Replacement. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the preservice inspection.

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

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

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

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

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

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

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

TheJe.+eA Amen~bQnt Iy SHEARON HARRIS - UNIT 1 3/4 4-15 I.

Amendment No.107

SSTEAM GENERATORS SURVEILLANCE 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 fabrication drawings or specifications. Eddy-current testing indications 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 outside of a tube:

3. Degraded Tube means a tube containing imperfections greater I than or equal to 20% of the nominal wall thickness caused by degradation:

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

5. Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective:

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

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

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

9. Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to POWER OPERATION with the replacement of steam generators using equipment and techniques expected to be used during subsequent inservice inspections.

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

(BREACTOR COOLANT SYSTEMI-"

STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria (Continued)

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

4.4.5.5 Reports

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;

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

1. Number and extent of tubes inspected.

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

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation 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:

l 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 a rotating schedule encompassing S% of the tubes if the results of the first or previous inspections indicate-that.-all steam generators are performing in a like manner. Note that,'under somecircumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence 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 should follow the instructions described in 1. above.

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

~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/A ST 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 for C-3 C-3 result of first sample Perform action C-3 for C-3 result N/A N/A of first sample C-3 Inspect all tubes All other in this S.G.. S.G.s are None N/A N/A -

llug defective C-1 tubes

_2S and ininspect tubes each otherS.G. Some S.G.s Perform action for C-2 N/A N/A C-2 but no result of second sample Notification to additional NRC pursuant to S.G.s are C-3 Specification Additional S.G., Insect all tubes in 4.4.5.5.c. is C-3 each S.G. and plug N/A N/A defective tubes.

Notification to NRC Sursuant to pecification 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.

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION.

3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a. The Containment Airborne Gaseous Radioactivity Monitoring System.

b. The Reactor Cavity Sump Level and Flow Monitoring System, and

c. 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 Systems INOPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed for airborne gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required Airborne Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With b. of the above required Leakage Detection Systems inoperable be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in, COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

1. Restore either Monitoring System (a. or c.) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and

2. Obtain and analyze a grab sample of the containment atmosphere for gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

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

Otherwise, be in at least HOT STANDBYwithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SHEARON HARRIS - UNIT 1 3/4 4-21

REACTOR COOLANT SYSTEM 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 g150Ig

. generator, 10r aK fromQ hpe eyathro u an yesteam p o, ncry-4o- 5844nA ar,

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

e. 31 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 +/- 20 psig, and

f. The maximum allowable leakage of any Reactor Coolant System Pressure Isolation Valve sh~ll be as specified in Table 3.4-1 at a pressure of 2235 +/- 20 psig.

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

ACTION: Wwii rt+ tnC,Ar- Se~coneJorP fet4lcfzue not LAWfI;it)UM{

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOG e following. 30; hours.

b. With any Reactor Coolant System der than any one of the above limits, excluding PRESSURE BOUDAR LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce t e leakage rate to within limils within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 h urs and i COLD SHU DO N wi in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . S' 0 . ,'

c. With any Reactor Coolant System ru-1 o a-on v e greater than the limit specified in Table 3.4-1, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Test pressures less than 2235 psig but greater than 150 psig are allowed.

Observed leakage shall be adjusted by multiplying the observed leakage by the square root of the quotient of 2235 divided by the test pressure.

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

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS Of Ck+4'on C 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment Airborne Gaseous or Particulate Radioactivity Monitor at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

b. Monitoring the containment sump inventory and Flow Monitoring System at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 +/- 20 psig at 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 at least once per 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months "and

e. Monitoring the Reactor Heed Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be: within its limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakagetesting has not been performed 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

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

+4e&"

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rt~c i oi fje-40roeA n4tz c~s e NIr ri.dc4 d54t1 opkrgedon.

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

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

k. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54 (o) and 10 CFR 50 Appendix J. Option B. as modified by approved exemptions. This program shall be in conformance with the NRC Regulatory Guide 1.163.

"Performance-Based Containment Leak-Test Program," dated September 1995. with the following exceptions noted: I

1) The above Containment Leakage Rate Testing Program is only applicable to Type A testing. Type B and C testing shall continue to be conducted in accordance with the original commitment to 10 CFR 50 Appendix J. Option A.

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

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

The calculated peak containment internal-pressure related to the design basis loss-of-coolant accident-is 41.8 psig. The calculated peak containment internal pressure related to the design basis main steam line break is 41.3 psig. P. will be assumed to be 41.8 psig for the purpose of containment testing in accordance with this Technical Specification.

The maximum allowable containment leakage rate. L. at P., shall be 0.1 %of containment air weight per day.

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

The provisions of Surveillance Requirement 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. However, test frequencies specified in this Program may be extended consistent with the guidance provided in Nuclear Energy Institute (NEI) 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50Appendix J.3 as endorsed by Regulatory Guide 1.163. Specifically. NEI 94-01 has this provision for test frequency extension:

1) Consistent with standard scheduling practices for Technical Specifications Required Surveillances. intervals for recommended rType A testing may be extended by up to 15 months. This option should be used only in cases where refueling schedules have been changed 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 ensure that steam generator (SG) tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

1. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during a SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

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

a) Structural integrity performance criterion. All inservice SG tubes shall retain structural integrity over the full range of normal operating.

conditions (including startup, operation in the power range, HOT STANDBY, and cooldown and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 (3deltaP) against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

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

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

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

4. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of 4a, 4b, 4c below, the inspection scope, inspection methods and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

a) Inspect 100% of the tubes in each SG during the first refueling 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 after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 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 inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

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

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

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

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

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

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

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"

approved version as specified in the COLR.

EMF-92-116(P)(A). "Generic Mechanical Design Criteria for PWR Fuel Designs,".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 Enthalpy Rise Hot Channel Factor).

6.9.1.6.3 The core operating limits shall be determined so that all applicable 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-cycle revisions or supplements, shall be provided, upon issuance.for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC in accordance with 10CFR50.4 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 HOT SHUTDOWN following completion of a steam generator tube inspection performed in 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 degradation mechanism,

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

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

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

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

Attachment 3 to SERIAL: HNP-06-060 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT REVISED TECHNICAL SPECIFICATIONS (TS) PAGES REVISED TECHNICAL SPECIFICATIONS (TS) PAGES Page A3-1 of 21

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 .. .. . .. .. . .. S. . . . . . . . . . 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 ........... ..... . . . . . .. 4-11 .'3/4 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY .. . . . . . . . .. 3/4 4-13 TABLE 4.4-1 (DELETED) ....... .. . . . . . . . . . . . . -3/4 4-18 TABLE-4.4-2 (DELETED) ........... .. . . . . . . . .. '3/4 4-19 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 SURVEILLANCE REQUIREMENTS ....... ... ............ ......... 3/4 4-28 3/4.4.8 SPECIFIC ACTIVITY . . . . . . . . . . . . . . . . . . . . 3/4 4-29 FIGURE 3.4-1 (DELETED) . . . . . . . . . . . . . . . . . . . . . . 3/4 4-30 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY-SAMPLE AND ANALYSIS PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-31 SHEARON HARRIS - UNIT 1 vii Amendment No.

INDEX 3.0/4.0 BASES SECTION PAGE 3/4.0 APPLICABILITY . . . . . . . . . . . . . . . . . . . B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL ................ B 3/4 1-1 3/4.1.2 BORATION SYSTEMS ................ B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES ........... B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS ............. B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE .... ............... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR B 3/4 2-2a FIGURE B 3/4.2-1 (DELETED) ................ B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATIO ..... ............ B 3/4 2-6 3/4.2.5 DNB PARAMETERS .............. . . . B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATIONSYSTEM INSTRUMENTATION . . . . . . . . . . . . . . . . B 3/4, 3-1 3/4.3.3 MONITORING INSTRUMENTATION ..... ...... B 3/4' 3-3 3/4.3.4 (DELETED) ....... .................... B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION B 3/4 4-1 3/4.4.2 SAFETY VALVES ...... .................. B 3/4 4-1 3/4.4.3 PRESSURIZER ...... ................... B 3/4 4-2 3/4.4.4 RELIEF VALVES ..... ............ ....... B 3/4 4-2 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY ....... B 3/4 4-2b 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ......... B 3/4 4-3 3/4.4.7 CHEMISTRY ...... .................... B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY ...... ................ B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ....... ...... 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 concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and garmma energies per disintegration (MeV/d) for isotopes, with half-lives greater than 15 minutes. making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge

.diesel pressures reach their required values, etc.). Times shall include generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

EXCLUSION AREA BOUNDARY 1.14 The EXCLUSION AREA BOUNDARY shall be that line beyond which the landjis not 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 installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary sýstem andtproviding for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.17 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump 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 both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

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

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 and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include 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 employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 1.20 The OFFSITE DOSE CALCULATION MANUAL (ODCM),shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1)the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of.:the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.3 and 6.9.1.4.

OPERABLE - OPERABILITY 1.21 A system, subsystem, train, component or device shall be OPERABLE or have 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 are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

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

PHYSICS TESTS 1.23 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation:

(1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 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-secondary leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

SHEARON HARRIS - UNIT 1 1-4 Amendment No.

REACTOR COOLANT SYSTEM 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 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 repair criteria and not plugged in accordance with the Steam Generator Program, within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

AND

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

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

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

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

SHEARON HARRIS - UNIT 1 3/4 4-13 Amendment No.

Page 3/4 4-14 Deleted by Amendment SHEARON HARRIS - UNIT 1 3/4 4-14 Amendment No.

Page 3/4 4-15 Deleted by Amendment SHEARON HARRIS - UNIT 1 3/4 4-15 Amendment No.

Page 3/4 4-16 Deleted by Amendment SHEARON HARRIS - UNIT 1 3/4 4-16 Amendment No.

Page 3/4 4-17 Deleted by Amendment SHEARON HARRIS - UNIT 1 3/4 4-17 Amendment No.

Table 4.4-1 Deleted by Amendment SHEARON HARRIS - UNIT 1 3/4 4-18 Amendment No.

Table 4.4-2 Deleted by Amendment SHEARON HARRIS - UNIT 1 3/4 4-19 Amendment No.

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a. The Containment Airborne Gaseous Radioactivity Monitoring System,

b. The Reactor Cavity Sump Level and Flow Monitoring System, and

c. 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 Systems INOPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed for airborne gaseous and particulate radioactivity at least.once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required Airborne Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With b. of the above required Leakage Detection Systems inoperable be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

1. Restore either Monitoring System (a. or c.) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and

2. Obtain and analyze a grab sample of the containment atmosphere for gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

3. Perform a Reactor Coolant System water inventory balance per Surveillance Requirement 4.4.6.2.1.d at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

SHEARON HARRIS - UNIT 1 3/4 4-21 Amendment No.

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 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 one steam generator,

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

e. 31 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 +/- 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 a pressure of 2235 +/- 20 psig."

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

ACTION:

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

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

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit specified in Table 3.4-1, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

"Test pressures less than 2235 psig but greater than 150 psig are allowed.

Observed leakage shall be adjusted by multiplying the observed leakage by the square root of the quotient of 2235 divided by the test pressure.

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

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS' 4.4.6.2.1 Reactor Coolant System operational leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment Airborne Gaseous or Particulate Radioactivity Monitor at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

b. Monitoring the containment sump inventory and Flow Monitoring System at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 +/- 20 psig at 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 Reactor Coolant System water inventory balance at least once per 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s*; and

e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

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

4.4.6.2.3 Primary-to-secondary leakage shall be verified < 150 gallons per day through any one steam generator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady-state operation. Not applicable to primary-to-secondary leakage.

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

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

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

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

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

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

a) Structural integrity performance criterion. All inservice SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, HOT STANDBY, and cooldown and a anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 (3deltaP) against burst under normal steady-state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

b) Accident induced leakage performance criterion. The primary-to-secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Accident induced leakage is not to exceed 1 gpm total for all three SGs.

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

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

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

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

4. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g.,

volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube.

In addition to meeting the requirements of 4a, 4b, and 4c below, the ins ection scope, inspection methods and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and.

based on this assessment, to determine which inspection methods need to be employed and at what locations.

a) Inspect 100% of the tubes in each SG during the first refueling 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 after the first inservice inspection of the SGs. In addition.

inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

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

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

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

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

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

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

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

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

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

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results."

approved version as specified in the COLR.

EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," 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 Enthalpy Rise Hot Channel Factor).

6.9.1.6.3 The core operating limits shall be determined so that all applicable 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-cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator 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 HOT SHUTDOWN following completion of a steam generator tube inspection performed in 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 degradation mechanism,

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

e. Number of tubes plugged during the inspection outage for each active degradation mechanism, SHEARON HARRIS - UNIT 1 6-24c Amendment No.

ADMINISTRATIVE CONTROLS 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 tube pulls and in-situ testing.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC in accordance with 10CFR50.4 within the time period specified for each report.

6.10 DELETED (PAGE 6-25 DELETED)

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

Attachment 4 to SERIAL: HNP-06-060 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT PROPOSED TECHNICAL SPECIFICATIONS (TS) BASES CHANGES (FOR INFORMATION ONLY)

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

Page A4-1 of 18

REACTOR COOLANT SYSTEM BASES RELIEF VALVES (Continued)

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

Surveillance Requirement 4.4.4.2 addresses the block valves. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply with ACTION statements "b" or "c". This precludes the need to cycle the valves with a full system differential pressure or when maintenance is being performed to restore an inoperable PORV to OPERABLE status.

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 be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Cuide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

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

Page 1 of 7 INSERT 3/4.4.5 BASES REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY

Background

Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. SG tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.1.1, "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 Coolant System, 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 RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

SG tubing is subject to a variety of degradation mechanisms. SG tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 6.8.4.1, "Steam Generator Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.8.4.1, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. The SG performance criteria are described in Specification 6.8.4.1.

Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

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

Page 2 of 7 REACTOR COOLANT SYSTEM 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 for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary-to-secondary leakage rate equal to 1 gpm, plus the leakage rate associated with a double-ended rupture of a single tube. The accident radiological analysis for a SGTR assumes the ruptured SG secondary fluid is released directly 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 SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses the steam discharge to the atmosphere is based on the total primary-to-secondary leakage from all SGs of 1 gpm, or is assumed to increase to 1 gpm as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the limits in LCO 3.4.8, "Reactor Coolant System Specific Activity." For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of 10 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 all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

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

Page 3 of 7 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY (continued)

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.8.4.1 and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

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

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load verses displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A, (normal operating conditions) and Service Level B, (upset conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Reference 3) and Draft Regulatory Guide 1.121 (Reference 4).

Page 4 of 7 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY (continued)

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

The operational leakage performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational leakage is contained in LCO 3.4.6.2 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 not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break.

If this amount of leakage is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in 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 ACTIONS provide appropriate compensatory actions for each affected SG tube. Complying with the required ACTIONS may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated required ACTIONS.

Page 5 of 7 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY (continued)

a. The condition applies if it is discovered that one or more SG tubes examined in an Inservice Inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by Surveillance Requirement 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 SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, condition (b) applies.

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

If the evaluation determines that the affected tube(s) have tube integrity, the ACTION statement allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes.

However, the affected tube(s) must be plugged prior to entering HOT SHUTDOWN following the next refueling outage or SG inspection. This allowed completion time is acceptable since operation until the next inspection is supported by the operational assessment.

b. If the required actions and associated completion times of condition (a) are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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

Page 6 of 7 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY (continued)

Surveillance Requirements 4.4.5.1 During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Reference 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections, a condition monitoring assessment of the SG tubes is performed.

The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Program determines the scope of the inspection and the method used to determine whether the tubes contain flaws satisfying the tube repair criteria.

Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.4.5.1. The frequency is determined by the operational assessment and other limits-in the SG examination guidelines (Reference 5). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.8.4.1 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

Page 7 of 7 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY (continued) 4.4.5.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 6.8.4.1 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference I provides guidance for performing operational assessments to verify that 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 criteria are plugged prior to subjecting the SG tubes to significant primary-to-secondary pressure differential.

References

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

2. 10 CFR 50.67

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

4. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," 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 secondary coolant will be maintained within those chemistry limits found to resull in negligible corrosion of the steam generator tubes_. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of crackin9 during plant operation would be limited by the limitation of steam genera or tube 1eakage 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 during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 150 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of-this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastaqe-type defects are unlikely with proper chemistry treatment of the secondary coolant. However even if a defect should develop in service. it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator ube inspections of operating plants have demonstrated the capabilitbto reliably detect degradation that has penetrated 20% of the original ude wall thickness.

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

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

These Detection .Systems are consistent with the-recommendations of Regulatory Guide 1.45. "Reactor Coolant Pressure Boundary Leakage Detection Systems,.

May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE

..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 promptlyl HplaceA in COLD SHUT1DOWN.

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

REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued)

Slow to ensure early detection of additional leakage.

The total steam generator tube leakage limit of 1 .1mfor all steam generators ensures that the dosage contribution from the tube eakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The l~gpm limit is consistent with the assumptions used in the analysis of these accidents. The 150 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection-of UNIDENTIFIED 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 modulating valve 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 flow will not be less than assumed in the safety analyses.

The maximum allowable leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure. It is 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 tested periodically to ensure ow probability of gross failure.

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

3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show SHEARON HARRIS - UNIT 1 B 3/4 4-4 Amendment No

Page 1 of 7 INSERT 3/4.4.6.2 BASES REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE

Background

Components that contain or transport the coolant to or from the reactor core make up the reactor 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 reactor coolant leakage, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational leakage LCO is to limit system operation in the presence of leakage from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of leakage.

10 CFR 50, Appendix A, GDC 30 (Reference 1), requires means for detecting and, to the extent 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, and duration. Therefore, detecting and monitoring reactor coolant leakage into the containment area is necessary. Quickly separating the identified leakage from the unidentified leakage is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

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

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

Page 2 of 7 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (continued)

Applicable Safety Analyses Except for primary-to-secondary leakage, the safety analyses do not address operational leakage. However, other operational leakage is related to the safety analyses for a LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary-to-secondary leakage from all steam generators is I gpm. The LCO requirement to limit primary-to-secondary leakage through any one steam generator to less than or equal 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 containment resulting 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 released directly to the atmosphere due to a failure of the PORV in the open position and will continue atmospheric release until the time that the PORV can be isolated. The FSAR analysis for the SLB assumes that the SG with the failed steam line boils dry releasing all of the iodine directly to the environment and that iodine carried over to the faulted SG by tube leaks are also released directly to the environment until the RCS has cooled to below 212T. The dose consequences resulting from the SGTR and the SLB accidents are within the limits defined in 10 CFR 50.67.

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

,* ';t Page 3 of 7 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (continued)

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 material deterioration. Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage. Violation of this LCO could result in continued degradation of the Reactor Coolant Pressure Boundary. Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b. UNIDENTIFED LEAKAGE One gallon per minute (gpm) of UNIDENTIFED LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the Reactor Coolant 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 leakage performance criterion in NEI 97-06, Steam Generator Program Guidelines (Reference 3). The Steam Generator Program operational leakage performance criterion in NEI 97-06 states, "The RCS operational primary-to-secondary leakage through any one steam generator shall be limited to 150 gallons per day." The limit is based on operating experience with steam generator tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion is conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

Page 4 of 7 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (continued)

d. IDENTIFIED LEAKAGE Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because leakage is 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 from specifically known and located sources, but does not include PRESSURE BOUNDARY 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. It is 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 tested periodically to ensure low probability of gross failure.

Page 5 of 7 REACTOR COOLANT SYSTEM BASES , ,,

3/4.4.6.2 OPERATIONAL LEAKAGE (continued)

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

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

ACTIONS

a. If any PRESSURE BOUNDARY LEAKAGE exists, or primary-to-secondary leakage is not within limit, the reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This action reduces the leakage and also reduces the factors that tend to degrade the pressure boundary.

The allowed outage times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In COLD SHUTDOWN, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

b. UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or CONTROLLED LEAKAGE in excess of the LCO limits must be reduced to within the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This allows time to verify leakage rates and either identify UNIDENTIFIED LEAKAGE or reduce leakage to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

Surveillance Requirements 4.4.6.2.1 Verifying RCS leakage to be within the LCO limits ensures the integrity of the RCPB is maintained. PRESSURE BOUNDARY LEAKAGE would at first appear as UNIDENTIFIED LEAKAGE and can only be positively identified by inspection. It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determined by performance or a RCS water inventory balance.

I-. I

- Page 6 of 7 REACTOR COOLANT SYSTEM BASES -

3/4.4.6.2 OPERATIONAL LEAKAGE (continued)

The RCS water inventory balance must be met with the reactor at steady-state operating conditions (stable pressure, temperature, power level, pressurizer and makeup tank levels, makeup letdown, and RCP seal injection and return flows). The surveillance is modified by a note. The note states that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady-state operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Steady-state operation is required to perform a proper water inventory balance since calculations during maneuvering are not useful, so the SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady-state operation. For RCS operational leakage determination by water inventory balance, steady-state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor containment atmosphere radioactivity and reactor cavity sump level. It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. These leakage detection systems are 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 is because leakage of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

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

4.4.6.2.2 The Surveillance Requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valve is IDENTIFIED 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 to 150 gpd through any one steam generator. Satisfying the primary-to-secondary leakage limit ensures that the operational leakage performance criterion in the Steam Generator Program is met. If this Surveillance Requirement is not met, compliance with LCO 3.4.5 should be evaluated. The 150-gpd limit is measured at room temperature as described in Reference 2. The operational leakage rate limit applies to leakage through any one steam generator. If it is not practical to assign the leakage to an individual steam generator, all the primary-to-secondary leakage should be conservatively assumed to be from one steam generator.

Page 7 of 7 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (continued)

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

The frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary-to-secondary leakage and recognizes the importance of early leakage detection in the prevention of accidents.

The primary-to-secondary leakage is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the 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"

HNP-06-060, License Amendment Request Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

Contents

Text

1.0 INTRODUCTION

2.0 DESCRIPTION

3.0 BACKGROUND

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

10.0 REFERENCES

11.0 CONCLUSION

Background

Background

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HNP-06-060, License Amendment Request Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

Text

1.0 INTRODUCTION

2.0 DESCRIPTION

3.0 BACKGROUND

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

10.0 REFERENCES

11.0 CONCLUSION

Background

Background

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