LIC-06-0087, Revised License Amendment Request, Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process and Deletion of Sleeving as a SGT Repair Method.

Revised License Amendment Request, Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process and Deletion of Sleeving as a SGT Repair Method.

ML062480074
Person / Time
Site:	Fort Calhoun
Issue date:	08/30/2006
From:	Reinhart J Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-06-0087, TAC MD2188
Download: ML062480074 (81)

Text

Omaha Public Power Distict 444 South 16th Street Mall Omaha NE 68102-2247 August 30, 2006 LIC-06-0087 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Reference:

1) Docket No. 50-285

2) Letter from Jeffrey A. Reinhart (OPPD) to Document Control Desk (NRC) dated May 30, 2006, Fort Calhoun Station Unit No. 1 License Amendment Request, "Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process and Deletion of Sleeving as a Steam Generator Tube Repair Method" (LIC-06-0002) (ML061510203)

SUBJECT:

Fort Calhoun Station Unit No. 1 Revised License Amendment Request, "Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process and Deletion of Sleeving as a Steam Generator Tube Repair Method" (TAC# MD2188)

Pursuant to 10 CFR 50.90, Omaha Public Power District (OPPD) hereby proposes to make changes to the Fort Calhoun Station Unit No. 1 (FCS) Technical Specifications (TS). 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).

OPPD also proposes to change the FCS TS by deleting the sleeving repair alternative to plugging for steam generator tubes. The FCS replacement steam generators (RSGs) to be installed during the fall of 2006 are manufactured by Mitsubishi Heavy Industries, Ltd. (MHI). The change is being requested because OPPD has determined that the sleeving repair alternative to plugging can not be used for the MHI RSGs at this time.

Reference 2 contained OPPD's original submittal with respect to TSTF-449. OPPD is resubmitting this amendment request to incorporate various revisions to the TS as a result of a Request for Additional Information (RAI) from the NRC staff. Attachment 5 provides the RAI questions and responses as discussed in a phone call on July 19, 2006. This revised amendment request does not change Attachment 1 as previously submitted, including the No Significant Hazards Consideration and the Environmental Evaluation.

Emplopnent with Equal Opportunity 4171

U. S. Nuclear Regulatory Commission LIC-06-0087 Page 2 provides a description of the proposed change, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides the existing TS pages marked up to show the proposed change. Attachment 3 provides revised (clean) TS pages. The proposed FCS TS changes are consistent with NRC approved Revision 4 of TSTF-449. Since FCS has custom TS, the numbering of the proposed TS and the location of information differ in some instances from TSTF-449. As a result, a cross reference table is provided in Attachment 4 to identify the location of TSTF-449 revisions in the FCS TS.

OPPD requests approval of the proposed amendment by November 1, 2006, to support scheduled implementation during the 2006 refueling outage. OPPD requests that the effective date for this TS change be the end of the 2006 refueling outage to allow for implementation of these proposed changes. No new commitments are made to the NRC in this letter.

I declare under penalty of perjury that the foregoing is true and correct. (Executed on August 30, 2006).

If you have any questions or require additional information, please contact Thomas R. Byrne at (402) 533-7368.

rSincer e rey A. Reinhart SiDirector Fort Calhoun Station JAR/TRB/trb Attachments:

1. Omaha Public Power District Evaluation

2. Markup of Technical Specification Pages

3. Proposed Technical Specifications (clean)

4. Location of TSTF-449 Requirements in FCS TS

5. Responses to Requests for Additional Information related to the May 30, 2006 TSTF-449 Submittal c: Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska

LIC-06-0087 Page 1 ATTACHMENT I Omaha Public Power District Fort Calhoun Station Application For Technical Specification Improvement Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process and Deletion of Sleeving as a Steam Generator Tube Repair Method

1.0 DESCRIPTION

2.0 ASSESSMENT

3.0 BACKGROUND

4.0 REGULATORY REQUIREMENTS AND GUIDANCE

6.0 REGULATORY ANALYSIS

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION 8.0 ENVIRONMENTAL EVALUATION 9.0 PRECEDENT

10.0 REFERENCES

LIC-06-0087 Page 2 Omaha Public Power District Fort Calhoun Station Application For Technical Specification Improvement Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process and Deletion of Sleeving as a Steam Generator Tube Repair Method

1.0 DESCRIPTION

The Omaha Public Power District (OPPD) proposes to revise the requirements in the Fort Calhoun Station Unit No. 1 (FCS) Technical Specifications (TS) related to steam generator tube integrity. The changes are consistent with 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). The proposed change also deletes steam generator tube sleeving as an alternative to plugging for steam generator tube repairs.

2.0 ASSESSMENT The proposed FCS TS changes are consistent with NRC approved Revision 4 of TSTF-449.

Since FCS has custom TS, however, the numbering of the proposed TS and the location of information differs in some instances from that of TSTF-449. Attachment 4 identifies the location of TSTF-449 revisions in the FCS TS.

Proposed revisions to the TS Bases are also included in this application. As discussed in the NRC's model safety evaluation, 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.

The sleeving repair method for Combustion Engineering steam generators was added to the FCS TS by Amendment 195 (Reference 10.3). During the 2006 Refueling Outage, replacement steam generators (RSGs) will be installed at FCS. Because the RSGs are manufactured by Mitsubishi Heavy Industries, Ltd. (MHI), the RSG tubes do not have an approved sleeving repair alternative to plugging. Therefore, OPPD is requesting the deletion of the sleeving capability from the TS because analyses supporting the existing specification are not applicable to the MHI steam generators.

LIC-06-0087 Attachment I Page 3 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.

The proposed TS changes to delete surveillance requirements for a steam generator tube repair alternative (sleeving) are being requested since they will no longer be utilized or credited.

Likewise, the analyses supporting the tube sleeving repair alternative are not applicable and are not planned to be applied to the RSGs. Thus, the proposed changes eliminate requirements not applicable to the RSGs. The TS will still contain the steam generator tube surveillance requirements which existed before tube sleeving was incorporated by Amendment 195 (Reference 10.3). Other TS changes, not related to tube sleeving, made by Amendment 195 are not being affected or altered. In accordance with the FCS general design criteria (Reference 10.4), the FCS TS are being revised as a result of a change in facility equipment.

5.0 TECHNICAL ANALYSIS

OPPD has reviewed the safety evaluation (SE) published on March 2, 2005 (70 FR 10298) as part of the CLIIP Notice for Comment. This included the NRC staffs SE, the supporting information provided to support TSTF-449, and the changes associated with Revision 4 to TSTF-449. OPPD has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to FCS and justify this amendment for the incorporation of the changes to the FCS TS.

The proposed changes remove provisions added to the TS in Amendment 195 (Reference 10.3) related to the surveillance requirements of leak tight sleeves, and the reference to the Electric Power Research Institute (EPRI) Pressurized Water Reactor (PWR) Steam Generator Examination Guidelines in the basis. Other TS changes, not related to tube sleeving, implemented by Amendment 195 are not being affected or altered.

The proposed changes remove the capability for repairing degraded steam generator tubes using sleeving which was approved for the original Combustion Engineering steam generators. These generators are being replaced in the fall of 2006 with steam generators manufactured by MHI which are not expected to use the tube sleeving alternative to tube plugging. The proposed changes eliminate tube sleeving as a repair for the RSG tubes. These changes are necessary, not as a result of new or enhanced analyses or evaluations, but as a consequence of MHI's steam generator operating experience which has not resulted in the need to develop the sleeving repair option.

The proposed changes do not alter, degrade, or prevent actions described or assumed in any accident analysis. They will not change any assumptions previously made in evaluating

LIC-06-0087 Attachment 1 Page 4 radiological consequences or affect any fission product barriers, nor do they increase any challenges to safety systems. They do not create any new systems interactions. Therefore, the proposed change does not increase or have any impact on the consequences of events described and evaluated in Chapter 14 of the Fort Calhoun Updated Safety Analysis Report (USAR).

5.1 Accident Induced Leakage Performance Criterion The Nuclear Energy Institute (NEI) provided members the following information and status report in a letter dated September 2, 2005. NEI stated that this was offered for use in license amendment requests. NEI stated that the following information and status report has been reviewed with the NRC.

The industry is currently evaluating a technical issue related to the Accident Induced Leakage Performance Criterion (AILPC) specified in Section 5.23 of the proposed Technical Specifications. The issue concerns the consideration of non-pressure (bending) loads on the accident induced leak rates of steam generator tubes (axial differential thermal loads are routinely considered in assessing accident induced leakage). The EPRI Steam Generator Management Program (SGMP) is conducting a study to determine if bending loads are significant, and if they are, to define how to account for the loads in steam generator tube integrity assessments. In the interim, as this study is being completed, EPRI has completed a preliminary impact assessment. The assessment (PreliminaryAssessment of the hnpact of Non-PressureLoads on Leakage Integrity of Steam Generator Tubing) found that the effect of the loads in question may, in certain circumstances, initiate primary-to-secondary leakage, or increase pre-existing primary-to-secondary leakage during and after load application. The effort also assessed the effect of such loads in combination with the applicable design basis accident.

The results indicate that these circumstances are expected to be limited to the presence of significant circumferential cracks located in high bending stress regions of tubing. As of this date, such degradation has not been observed in the industry.

The structural integrity impact of non-pressure loads on degraded steam generator tubes has been well-documented in a previous EPRI report (NRC accession number ML050760208) related to the revised Structural Integrity Performance Criterion (SIPC). Experimental results indicated that neither axial loads nor bending loads have a significant effect on the burst pressure of tubing with axial degradation. Similarly, these loads are considered inconsequential for axially oriented degradation with respect to localized pop-through conditions and corresponding accident leakage. As such, industry experience indicates that the only meaningful impact of non-pressure loads with respect to leakage is due to the application of bending moments on circumferential cracking.

The EPRI Preliminary Assessment found that high bending loads that could affect the leakage analysis are only present in the top span region in the original design of once-through steam generators (OTSGs) and in the U-bend region of large-radius tubes in some recirculating steam generators. The high bending loads in the OTSGs are a consequence of crossflow during a steam line break whereas the high bending loads in the recirculating steam generators are a result of a seismic event.

LIC-06-0087 Page 5 After review of available analysis and experimental data, the EPRI Assessment concluded that the effect of high bending loads is only noteworthy for large 100% or near through-wall circumferential degradation. From a degradation assessment perspective, the EPRI study also reported that current industry experience indicates that there have been no observed stress corrosion circumferential cracks that are both capable of leaking and located in high bending stress regions. The industry's preliminary impact assessment and the plans for further technical study and experimental testing were presented to the NRC Staff in meetings on August 12, 2005 and July 12, 2006. The NRC Staff did not have any significant comments on the results presented.

Based on the above, OPPD believes that the effect of bending loads is not safety significant for FCS with respect to leakage integrity given the expected effect and existing margins with respect to degradation type, susceptible location and allowable flaw size.

If upon completion of EPRI's technical study, it is concluded that the effect of non-pressure loads, including bending loads, should be specifically accounted for in integrity assessments, the industry will revise the applicable steam generator program guideline documents to reflect the means developed to account for the loads.

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:

LIC-06-0087 Page 6 Plant Name, Unit No. Fort Calhoun Station Unit No. 1 Steam Generator Model(s): Mitsubishi Model: 49TT-1 Effective Full Power Years 0 EFPY (EFPY) of service for currently installed SGs Tubing Material Alloy 690TT Number of tubes per SG 5200 Number and percentage of tubes 1 tube in SG RC-2B (0.0 19%)

plugged in each SG Number of tubes repaired in each 0 SG Degradation mechanism(s) None identified Current primary-to-secondary 150 gallons per day through any one steam leakage limits: generator at standard temperature.

Approved Alternate Tube Repair None Criteria (ARC):

Approved SG Tube Repair Tubes found by inservice inspection to contain Methods flaws with a depth equal to or exceeding 40%

of the nominal tube wall thickness shall be plugged.

Performance criteria for accident 1.0 gallon per minute at standard temperature.

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

LIC-06-0087 Page 7 OPPD has evaluated whether or not a significant hazards consideration is involved with the proposed amendment to delete steam generator tube sleeving as an alternative to plugging for steam generator tube repairs by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The elimination from the TS surveillance requirements of leak tight sleeves as a repair method alternative to plugging defective steam generator tubes does not introduce an initiator to any previously evaluated accident. The frequency or periodicity of performance of the remaining surveillance requirements for steam generator tubes (including plugged tubes) is not affected by this change.

Elimination of the tube repair method has no effect on the consequences of any previously evaluated accident. The proposed changes will not prevent safety systems from performing their accident mitigation function as assumed in the safety analysis.

Therefore, this change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change only affects the TS surveillance requirements. The proposed change is a result of installation of RSGs. The proposed change will eliminate a steam generator tube repair alternative which cannot be utilized or credited for the RSGs. This change will not alter assumptions made in the safety analysis and licensing bases and will not create new or different systems interactions.

Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change deletes surveillance requirements for a steam generator tube repair alternative which will no longer be necessary or applicable. The remaining

LIC-06-0087 Page 8 TS steam generator tube surveillance requirements, including inspection and plugging requirements, will continue to maintain the applicable margin of safety.

Therefore, this TS change does not involve a significant reduction in the margin of safety.

Based on the above, Omaha Public Power District concludes that the proposed amendment to delete steam generator tube sleeving as an alternative to plugging for steam generator tube repairs presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of"no significant hazards consideration" is justified.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

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

Based on the above considerations, the proposed amendment to delete steam generator tube sleeving as an alternative to plugging for steam generator tube repairs does not involve and will not result in a condition which significantly alters the impact of Fort Calhoun Station on the ehvironment. Thus, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR Part 51.22(c)(9), and, pursuant to 10 CFR Part 51.22(b), no environmental assessment need be prepared.

9.0 PRECEDENT The TS changes addressed in TSTF-449, Revision 4, are being made in accordance with the CLIIP. OPPD is not proposing variations or deviations from those TS changes, or the NRC staff's model SE published on March 2, 2005 (70 FR 10298) with the exception that the Basis for the applicable safety analyses has been changed to reflect the FCS-specific Steam Generator Tube Rupture (SGTR) assumptions.

OPPD is also requesting deletion from the FCS TS of the surveillance requirements for a steam generator tube repair alternative which will no longer be necessary or applicable due to installation of RSGs. Removal of unnecessary, inapplicable provisions from TS is an accepted practice.

LIC-06-0087 Page 9

10.0 REFERENCES

10.1 Federal Register Notice for Comment published on March 2, 2005 (70 CFR 10298) 10.2 Federal Register Notice of Availability published on May 6, 2005 (70 FR 24126) 10.3 Letter from NRC (L. R. Wharton) to OPPD (S. K. Gambhir) dated March 1, 2001, Issuance of Amendment Re: Leak Tight Sleeves as an Alternative Tube Repair Method To Plugging Defective Steam Generator Tubes (TAC No. MA9653) (NRC-01-012) 10.4 Fort Calhoun Station Updated Safety Analysis Report, Responses to 70 Criteria, Appendix G.

LIC-06-0087 Page 1 ATTACHMENT 2 Markup of Technical Specification Pages (NOTE: Additions are indicated by italic font; deletions are indicated by strikethrough.)

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued) 2.13 DELETED 2.14 Engineered Safety Features System Initiation Instrumentation Settings 2.15 Instrumentation and Control Systems 2.16 River Level 2.17 Miscellaneous Radioactive Material Sources 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 Toxic Gas Monitors 2.23 Steam Generator(SG) Tube Integrity 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 DELETED 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 Radioactive Material Sources Surveillance 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator (SG) Tubes Integrity 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. 11,27,32,38,-3,16,51,60,81n,86, 93,97,10"1,122,136,152, 160,176,183, 214,230, 236

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued) 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Training 5.5 Not Used 5.6 Not Used 5.7 Safety Limit Violation 5.8 Procedures 5.9 Reporting Requirements 5.9.1 Not Used 5.9.2 Not Used 5.9.3 Special Reports 5.9.4 Unique Reporting Requirements 5.9.5 Core Operating Limits Report 5.9.6 RCS Pressure-Temperature Limits Report (PTLR) 5.10 Record Retention 5.11 Radiation Protection Program 5.12 DELETED 5.13 Secondary Water Chemistry 5.14 Systems Integrity 5.15 Post-Accident Radiological Sampling and Monitoring 5.16 Radiological Effluents and Environmental Monitoring Programs 5.16.1 Radioactive Effluent Controls Program 5.16.2 Radiological Environmental Monitoring Program 5.17 Offsite Dose Calculation Manual (OCDM) 5.18 Process Control Program (PCP) 5.19 Containment Leakage Rate Testing Program 5.20 Technical Specification (TS) Bases Control Program 5.21 Containment Tendon Testing Program 5.22 Diesel Fuel Oil Testing Program 5.23 Steam Generator(SG) Program 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS 6.1 Deleted 6.2 Deleted 6.3 Deleted 6.4 Deleted TOC - Page 3 Amendment No. 32, 31, 12, 51, 55, 57, 7-3, 80, 86, 93, 99,111, 152, 157,181, 185, 221 236,237

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS TABLE SECTION 1- 1 R PS LSSS ............................................................................................ Sectionl.0 2- 1 ESFS Initiation Instrumentation Setting Limits ...................... Section 2.14 2-2 Instrument Operating Requirements for RPS ................................................ Section 2.15 2 -3 Instrument Operating Requirements for Engineered Safety Features ................. Section 2.15 2-4 Instrument Operating Conditions for Isolation Functions .................................. Section 2.15 2-5 Instrumentation Operating requirements for Other Safety Feature Functions ....... Section 2.15 2- 9 RCS Pressure Isolation Valves .................................................................. Section 2.1 2- 10 Post-Accident Monitoring, Instrumentation Operating Limits ............................. Section 2.21 2 - 11 Toxic Gas Monitors Operating limits ............................................................ Section 2.22 3 - 1 Minimum Frequencies for Checks, Calibrations, and Testing of RPS .................. Section 3.1 3 -2 Minimum Frequencies for Checks, Calibrations and Testing of ........................... Section 3.1 Engineered Safety Features, Instrumentation and Controls 3- 3 Minimum Frequencies for Checks, Calibrations, and Testing ............................. Section 3.1 of Miscellaneous Instrumentation and Controls 3 - 3a Minimum Frequency for Checks, Calibrations and Functional ........................... Section 3.1 Testing of Alternate Shutdown Panels (Al-1 85 and AI-212) and Emergency Auxiliary Feedwater Panel (AI-179) Instrumentation And Control Circuits 3 -4 Minimum Frequencies for Sampling Tests .................................................... Section 3.2 3- 5 Minimum Frequencies for Equipment Tests .................................................. Section 3.2 3- 6 Reactor Coolant Pump Surveillance ............................................................ Section 3.3 3 13 Steam Generator Tube Inspection................................................Sectionp 3.17 TOC - Page 4 Amendment No. **,6-,

, 4 ,60

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS -TABLES TABLE OF CONTENTS TABLE t'*L .......................

/'* * "1= L I'll I * ==_

13414 Stem '3onoraIOr iU~ 9V incpCno upo ioovo ........................................ beG!'GR 6.1 (

5.2-1 Minim um Shift Crew Com position ............................................................................................. Section 5.0 TOC- Page 5 Amendment No. 116,125,142,145,152,176, 195

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS (ALPHABETICAL ORDER)

Continued TABLE DESCRIPTION SECTION 2-10 Post-Accident Monitoring Instrumentation Operating Limits ................................................... Section 2.21 2-9 RCS Pressure Isolation Valves ................................................................................................. Section 2.1 3-6 Reactor Coolant Pum p Surveillance ......................................................................................... Section 3.3 1-1 R PS LS S S ................................................................................................................................. Section 1.0 3 :13 Steam*Geonerator Tube ,I-n.ption .. .... .. Section 3.17 3 14 Steam Generater*Tube Sleeve . nspction .SeGtin 3.17 2-11 Toxic Gas Monitoring Operating Limits ................................................................................... Section 2.22 TOC - Page 7 Amendment No. 116,145,152,176, 195

TECHNICAL SPECIFICATION DEFINITIONS E - Average Disintegration Enerqy P_is 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 gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 minutes making up at least 95% of the total non-iodine radioactivity in the coolant.

Offsite Dose Calculation Manual (ODCM)

The document(s) that contain the methodology and parameters used in the calculations of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent radiation monitoring Warn/High (trip) Alarm setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain:

1) The Radiological Effluent Controls and the Radiological Environmental Monitoring Program required by Specification 5.16.

2) Descriptions of the information that should be included in the Annual Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports required by Specifications 5.9.4.a and 5.9.4.b.

Unrestricted Area Any area at or beyond the site boundary access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

Core Operating Limits Report (COLR)

The Core Operating Limits Report (COLR) is a Fort Calhoun Station Unit No. I specific document that provides core operating limits for the current operating cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Section 5.9.5. Plant operation within these operating limits is addressed in the individual specifications.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal leakoff), that is captured and conducted to collection systems or a sump or collecting tank,

2. 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

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (&G) to the Secondary System (primaryto secondaryLEAKAGE),

Definitions - Page 8 Amendment No. 67,86,111,152,!61,221,226 Correction Letter of 06-17-2004

TECHNICAL SPECIFICATION DEFINITIONS

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal leakoff) that is not identified LEAKAGE, and

c. Pressure Boundary LEAKAGE LEAKAGE (except SG primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

RCS Pressure-Temperature Limits Report (PTLR)

The PTLR is a fluence dependent document that provides Limiting Conditions for Operation (LCO) in the form of pressure-temperature (P-T) limits to ensure prevention of brittle fracture. In addition, this document establishes power operated relief valve setpoints which provide low temperature overpressure protection (LTOP) to assure the P-T limits are not exceeded during the most limiting LTOP event. The P-T limits and LTOP criteria in the PTLR are applicable through the effective full power years (EFPYs) specified in the PTLR. NRC approved methodologies are used as the bases for the information provided in the PTLR.

References (1) USAR, Section 7.2 (2) USAR, Section 7.3 Definitions - Page 9 Amendment No. 226 Correction Letter of 06-17-2004

TECHNICAL SPECIFICATION 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable Components (Continued)

(5) DELETED (6) Both steam generators shall be filled above the low steam generator water level trip set point and available to remove decay heat whenever the average temperature of the reactor coolant is above 300 0 F. Each 6team generator shall be demonstrated operable by performaRnc of the inr.we'-

anspction program spcified in Sect,-in 3.17 prior to eXceeding a reactor coolant temperature

-o (7) Maximum reactor coolant system hydrostatic test pressure shall be 3125 psia. A maximum of 10 cycles of 3125 psia hydrostatic tests are allowed.

(8) Reactor coolant system leak and hydrostatic test shall be conducted within the limitations of the pressure and temperature limit Figure(s) shown in the PTLR.

(9) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia. A minimum measured temperature of 73 0F is required. Only 10 cycles are permitted.

(10) Maximum steam generator steam side leak test pressure shall not exceed 1000 psia. A minimum measured temperature of 73°F is required.

(11) Low Temperature Overpressure Protection (LTOP)

(a) The LTOP enable temperature and RCP operations shall be maintained in accordance with the PTLR.

(b) The unit can not be placed on shutdown cooling until the RCS has cooled to an indicated RCS temperature of less than or equal to 300'F.

(c) If no reactor coolant pumps are operating, a non-operating reactor coolant pump shall not be started while Tc is below the LTOP enable temperature stated in the PTLR unless there is a minimum indicated pressurizer steam space of at least 50% by volume.

2.1 - Page 3 Amendment No. 39,56,66,71,119,4366, 161,188, 207, 221

TECHNICAL SPECIFICATION 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.4 Reactor Coolant System Leakage Limits Applicability Applies to the leakage rates of the reactor coolant system whenever the reactor coolant temperature (Tcowd) is greater than 210 *F.

Obiective To specify limiting conditions of the reactor coolant system leakage rates.

Specifications To assure safe reactor operation, the following limiting conditions of the reactor coolant system leakage rates must be met:

(1) RCS operational LEAKAGE shall be limited to:

a. No Pressure Boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

c. 10 gpm identified LEAKAGE,

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator(SG).

(2) If RCS operationalLEAKAGE limits of (1), above, are not met for reasons other than Pressure Boundary LEAKAGE orprimaryto secondary LEAKAGE, then reduce LEAKAGE to meet limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

(3) If the Required Action and associated completion time of (2), above, is not met, OR Pressure Boundary LEAKAGE exists, or primary to secondaryLEAKAGE is not within limits, then be in MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND be in MODE 4, Cold Shutdown, within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

(4) To determine leakage to the containment, a containment atmosphere radiation monitor (gaseous or particulate) or dew point instrument, and a containment sump level instrument must be operable.

a. With no containment sump level instrument operable, verify that a containment atmosphere radiation monitor is operable, and restore the containment sump level instrument to operable status within 30 days.

b. With no containment atmosphere radiation monitor and no dewpoint instrument operable, restore either a radiation monitor or dewpoint instrument to operable status within 30 days.

c. With only the dewpoint instrument operable, or with no operable instruments, enter Specification 2.0.1 immediately.

2.1 - Page 13 Amendment No. 32,165,195,

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.4 Reactor Coolant System Leakage Limits (Continued)

c. 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 unidentified LEAKAGE and is well within the capability of the RCS makeup system. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE through Any One SG The 150 gallon per day operationallimit on primary to secondary LEAKAGE through any one SG is based upon guidance in NEI 97-06, Steam Generator Program Guidelines. The Steam GeneratorProgram operationalLEAKAGE performance Criterionin NEI 97-06 states, "The RCS operationalprimary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operatingexperience with SG tube degradationmechanisms that result in tube leakage. The operational leakage rate criterionin conjunction with the implementation of the Steam GeneratorProgramis an effective measure for minimizing the frequency of steam generatortube ruptures.

APPLICABILITY The potential for RCPB LEAKAGE is greatest when the RCS is pressurized, that is, when the reactor coolant temperature (Tc.d) is greater than 210°F.

In MODES 4 and 5, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

REQUIRED ACTIONS (2).

Unidentified LEAKAGET or identified LEAKAGE, or primary to econdar'; LEAKA.GE in excess of the LCO limits must be reduced to meet limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This Completion Time 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.

REQUIRED ACTIONS (3)

If any pressure boundary LEAKAGE exists or primary to secondaryLEAKAGE is not within limits, or if unidentifiedT or identified, or primar'-' to socondary LEAKAGE cannot be reduced to meet limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. The reactor must be brought to MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4, Cold Shutdown, within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

2.1 - Page 16 Amendment No. 32,165, 226

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator(SG) Tube Integrity Applicability Applies whenever the reactorcoolant temperature (Tco1d) is greaterthan 210 0F.

Objective To ensure that SG tube integrity is maintained.

Specification NOTE: Separate Condition entry is allowed for each SG Tube.

(1) The following conditions shall be maintained:

(a) SG tube Integrityshall be maintained,and (b)All SG tubes satisfying the tube repaircriteria shall be plugged in accordance with the Steam GeneratorProgram.

(2) If the requirementsof (1)(b) above are not met for one or more SG tubes, then perform the following.

(a) Verify tube integrity of the affected tube(s) is maintaineduntil the next refueling outage or SG tube inspection within 7 days, and (b) Plug the affected tube(s) in accordance with the Steam GeneratorProgrampriorto exceeding 210°F reactorcoolant temperature(TC,,d) following the next refueling outage or SG tube inspection.

(3) If the RequiredAction and associatedcompletion time of (2), above, is not met, or if SG tube integrity is not maintained,then be in MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND be in MODE 4, Cold Shutdown, within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Basis 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. Steam generator tubes are an integralpart of the reactorcoolantpressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactivefission 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 transfersurface between the primary and secondary systems to remove heat from the primarysystem. This Specification addressesonly the RCPB integrity function of the SG. The SG heat removal function is addressedby Technical Specification 2.1.1, "OperableComponents."

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

2.23 - Page 1 Amendment No.

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator(SG) Tube Integrity (continued)

Steam generatortubing is subject to a variety of degradationmechanisms. Steam generatortubes may experience tube degradationrelated to corrosion phenomena, such as wastage, pitting, intergranularattack, and stress corrosion cracking, along with other mechanically inducedphenomena such as denting and wear. These degradationmechanisms can impair tube integrity if they are not managed effectively. The SG performance criteriaare used to manage SG tube degradation.

Specification 5.23, "Steam Generator(SG) Program,"requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuantto Specification 5.23, tube integrity is maintained when the SG performance criteriaare met. There are three SG performance criteria:structuralintegrity, accident induced leakage, and operationalLEAKAGE. The SG performance criteria are described in Specification 5.23.

Meeting the SG performance criteriaprovides reasonableassuranceof maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam GeneratorProgram Guidelines (Ref. 1).

The steam generatortube 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 secondaryLEAKAGE rate equal to the operationalLEAKAGE rate limits in Technical Specification 2.1.4, "ReactorCoolant System Leakage Limits," plus the leakage rate associatedwith a double-ended rupture of a single tube. The accident analysis for a SGTR assumes releasesof activity occur from the faulted steam generatorto the environment via the condenser air ejector and Main Steam Safety Valves (MSSVs) and Atmospheric Dump Valves (ADVs). The release via the condenserairejector starts at the initiationof the event and continues to the reactortrip, while the release via the MSSVs/ADVs starts at the reactortrip and continues for the duration of the event."

The analysis for design basis accidents and transientsother than a SGTR assume the SG tubes retain their structuralintegrity (i.e., they are assumednot to rupture.) In these analyses, the steam discharge to the atmosphere is based on the totalprimary to secondary LEAKAGE from all SGs of I gallon per minute or is assumed to increase to 1 gallon per minute as a result of accident induced conditions. Foraccidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the Technical Specification 2.1.3, "ReactorCoolant Radioactivity," limits. Foraccidents 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 GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approvedlicensing basis (e.g., a small fraction of these limits).

Steam generatortube integrity satisfies Criterion2 of 10 CFR 50.36(c)(2)(ii). The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repaircriteriabe plugged in accordancewith the Steam GeneratorProgram.

Duringan SG inspection, any inspected tube that satisfies the Steam GeneratorProgramrepaircriteriais removed from service by plugging. If a tube was determined to satisfy the repaircriteria but was not plugged, the tube may still have tube integrity.

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

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteriaare defined in Specification 5.23, "Steam GeneratorProgram,"and describe acceptable SG tube performance. The Steam GeneratorProgramalso provides the evaluationprocess for determining conformance with the SG performance criteria.

There are three SG performance criteria:structuralintegrity,accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteriais consideredfailure to meet the LCO.

2.23 - Page 2 Amendment No.

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator(SG) Tube Integrity (continued)

The structuralintegrityperformance criterionprovides a margin of safety againsttube burst or collapse under normal and accident conditions, and ensures structuralintegrity of the SG tubes under all anticipatedtransients included in the design specification. Tube burst is defined as, "The gross structuralfailure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increasedin response to constant pressure)accompaniedby ductile (plastic)tearing of the tube materialat the ends of the degradation." Tube collapse is defined as, "Forthe load displacementcurve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structuralintegrityperformance criterionprovides guidance on assessingloads that have a significant effect on burst or collapse. In that context, the term "significant"is defined as "An accident loading condition other than differentialpressure is consideredsignificant when the addition of such loads in the assessment of the structural integrity performance criterioncould cause a lower structurallimit or limiting burst/collapsecondition to be established." Fortube integrity evaluations,except for circumferentialdegradation,axial thermal loads are classified as secondaryloads. For circumferentialdegradation,the classificationof axial thermal loads as primary or secondaryloads will be evaluated on a case-by-case basis. The division between primary and secondary classificationswill be based on detailedanalysis and/ortesting.

Structuralintegrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normaloperatingconditions) and Service Level B (upset or abnormalconditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

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

The operationalLEAKAGE performance criterionprovides an observable indication of SG tube conditions during plant operation. The limit on operationalLEAKAGE is containedin Technical Specification 2.1.4. Reactor Coolant System Leakage Limits," and limits primary to secondaryLEAKAGE through any one SG to 150 gallons per day. 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 ora 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.

Steam generatortube integrity is challenged when the pressuredifferential across the tubes is large. Large differentialpressures across SG tubes can only be experienced when Tcold is > 210°F.

RCS conditions are far less challenging in MODES 4 and 5 than during MODES 1, 2, and 3. In MODES 4 and 5, primary to secondary differentialpressure is low, resultingin lower stresses and reduced potentialfor LEAKAGE.

The ACTIONS are modified by a Note clarifying that the Conditions may be entered independentlyfor each SG tube. This is acceptable because the Required Actions provide appropriatecompensatory 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 applicationof associatedRequired Actions.

Specification 2.23(2) applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repaircriteriabut were not plugged in accordance with the Steam GeneratorProgramas requiredby Technical Specification 3.17. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generatortube integrity is based on meeting the SG performance criteriadescribed in the Steam GeneratorProgram.

2.23 - Page 3 Amendment No.

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator(SG) Tube Integrity (continued)

The SG repaircriteriadefine limits on SG tube degradationthat allow for flaw growth between inspections while still providing assurancethat the SG performance criteria will continue to be met. In orderto 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 determinationis based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradationpriorto the next SG tube inspection. If it is determined that tube integrity is not being maintained,Specification 2.23(3) applies.

A Completion Time of 7 days is sufficient to complete the evaluationwhile minimizing the risk of plant operationwith a SG tube that may not have tube integrity.

If the evaluationdetermines that the affected tube(s) have tube integrity, Required Action 2.23(2)b allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operationalassessment that reflects the affected tubes. However, the affected tube(s) must be plugged priorto exceeding 210°F reactorcoolant temperature(Tecod) following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operationalassessment.

If the Required Actions and associated Completion Times of Technical Specification 2.23(2) are not met or if SG tube integrity is not being maintained,the reactormust be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable,based on operatingexperience, to reach the desiredplant conditions from full power conditions in an orderly manner and without challengingplant systems.

References

1. NEI 97-06, "Steam GeneratorProgram Guidelines."

2. 10 CFR 50 Appendix A, GDC 19.

3. 10 CFR 100.

4. ASME Boiler and Pressure Vessel Code, Section Ill, Subsection NB.

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

6. EPRI, "PressurizedWater Reactor Steam GeneratorExamination Guidelines."

2.23 - Page 4 Amendment No.

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Samplinq Tests (continued)

The Safety Injection (SI) pump room air treatment system consists of charcoal adsorbers which are installed in normally bypassed ducts. This system is designed to reduce the potential release of radioiodine in SI pump rooms during the recirculation period following a DBA. The in-place and laboratory testing of charcoal adsorbers will assure system integrity and performance.

Pressure drops across the combined HEPA filters and charcoal adsorbers, of less than 9 inches of water for the control room filters (VA-64A & VA-64B) and of less than 6 inches of water for each of the other air treatment systems will indicate that the filters and adsorbers are not clogged by amounts of foreign matter that would interfere with performance to established levels. Operation of each system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month will demonstrate operability and remove excessive moisture build-up in the adsorbers.

The hydrogen purge system provides the control of combustible gases (hydrogen) in containment for a post-LOCA environment. The surveillance tests provide assurance that the system is operable and capable of performing its design function. VA-80A or VA-80B is capable of controlling the expected hydrogen generation (67 SCFM) associated with 1) Zirconium - water reactions, 2) radiolytic decomposition of sump water and 3) corrosion of metals within containment. The system should have a minimum of one blower with associated valves and piping (VA-80A or VA-80B) available at all times to meet the guidelines of Regulatory Guide 1.7 (1971).

If significant painting, fire or chemical release occurs such that the HEPA filters or charcoal adsorbers could become contaminated from the fumes, chemicals or foreign materials, testing will be performed to confirm system performance.

Demonstration of the automatic and/or manual initiation capability will assure the system's availability.

Verifying Reactor Coolant System (RCS) leakage to be within the LCO limits ensures the integrity of the Reactor Coolant Pressure Boundary (RCPB) is maintained. Pressure boundary leakage would at first appear as unidentified leakage and can only be positively identified by inspection. Unidentified leakage is determined by performance of an RCS water inventory balance. Identified leakage is then determined by isolation and/or inspection. Since Primaryto Secondary Leakage of 150 gallons per day cannot be measured accuratelyby an RCS water inventory balance, note "..." for line item 8a on Table 3-5 states that the Reactor Coolant System Leakage surveillanceis not applicableto Primaryto Secondary Leakage.

Primary to secondary leakage is-also measured by performance of an RCS water invonto,'y ba!ance in conjunction with effluent monitoring within the secondary steam and feedwater systems.

3.2 - Page 2 Amendment No. 15,67,128,138,169 T-SI 03 006* 0

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Samplingi Tests (continued)

Table 3-5, Item 8b verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operationalLEAKAGE performance criterion in the Steam GeneratorProgram is met. If this surveillance requirementis not met, compliance with LCO 2.23, "Steam Generator(SG) Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as describedin Reference 5. The operationalLEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practicalto assign the LEAKAGE to an individual SG, all the primary to secondaryLEAKAGE should be conservatively assumed to be from one SG.

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

The Surveillance Frequency of daily is a reasonable interval to trend primary to secondaryLEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiationmonitors or radiochemicalgrab sampling in accordancewith the EPRI guidelines (Ref. 5).

References

1) USAR, Section 9.10

2) ASTM D4057-95(2000), ASTM D975-98b, ASTM D4176-93, ASTM D129-00, ASTM D2622-87, ASTM D287-82, ASTM 6217-98, ASTM D2709-96

3) ASTM D975-98b, Table 1

4) Regulatory Guide 1.137

5) EPRI, "PressurizedWater Reactor Primary-to-SecondaryLeak Guidelines."

3.2 - Page 3b Amendment No. 229

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Test Frequency Reference

1. Control Element Drop times of all full-length CEA's Prior to reactor criticality after each 7.5.3 Assemblies removal of the reactor vessel closure head

2. Control Element Partial movement of all CEA's Q 7 Assemblies (Minimum of 6 in)

3. Pressurizer Safety Verify each pressurizer safety valve R 7 Valves is OPERABLE in accordance with the Inservice Testing Program.

Following testing, lift settings shall be 2485 psig +/-1% and 2530 psig +/-1%

respectively.

4. Main Steam Safety Set Point R 4 Valves

5. DELETED

6. DELETED

7. DELETED 8a. Reactor Coolant Evaluate D* 4 System Leakage***

8b Primaryto Secondary Continuousprocess 4 Leakage **** radiationmonitors or radiochemicalgrab sampling 9a Diesel Fuel Supply Fuel Inventory M 8.4 9b. Diesel Lubricating Oil Lube Oil Inventory M 8.4 Inventory 9c. Diesel Fuel Oil Test Properties In accordance with the Diesel Fuel 8.4 Properties Oil Testing Program 9d. Required Diesel Air Pressure M 8.4 Generator Air Start Receiver Bank Pressure

• Whenever the system is at or above operating temperature and pressure.

3.2 - Page 6 Amendment No. 15,24,128,160,166,169,

• • • Not applicable to primaryto secondary LEAKAGE. 171,219, 229

TECHNICAL SPECIFICATIONS TABLE 3-5 (continued)

MINIMUM FREQUENCIES FOR EQUIPMENT TESTS

• Verify primary to secondary LEAKAGE is < 150 gallons perday through any one SG.

This surveillance is not requiredto be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishmentof steady state operation.

3.2- Page 7 Amendment No.

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator (SG) Tubes Integrity Applicability Applies to in-service surveillance of steam generator tubes.

Objective To ensure the integrity of the steam generator tubes.

Specifications Each steam generator shall be demonstrated OPERABLE by performance of the following: in .service inspection program.

(1) Verify SG Tube Integrity in accordance with the Steam GeneratorProgram.

(2) Verify that each inspected SG tube that satisfies the tube repaircriteria is plugged in accordancewith the Steam GeneratorProgramprior to exceeding 210°Freactorcoolant temperature(Tcold).

() te enerator SamoI- Slnetnion andl ns.;-t," Methodr h The in senMic inspection shall be performed on each steam g8eneato~ro a rotating schedule. Under som~e circums~tances, the operating conditions in one steam generator mna" be found to be more8 severe than those in the second steam generator. Under such cirFcumsance, the sample sequence shall be modified to inspect the steam generato-r wi~th the most severe conditions.

(2) S-team;r Gelnelratolr Tube Sa_;MPle Selection and lnS0ection The steam generator tube mninimum sample si;Ze, enspection result classification, and the corresponRding action required shall be as specified inTable 3413. The in-Aservice inspection of steam generator tubes shall be performed according to Specification 3.17(4)(i), "Tube Inspection," and at the frequencies specified in Specification 3.17(3). The inspected tubes shall be verified acceptablc per the acceptance criteria of Specification 3.17(4). When applying the eXceptions Of (0),(ii)and (ili) beleW, previous degradiation, im~perfections or defects in the area of the tube repaired by s:leeving are ot considered an area requiRing reinspecton or inspection Of adjacent tubes. The tubes selected for each in sev-ice insEpection 6hall include at least 3% of the total tubes in the steam gGeneatorFs and the tub-es selected for these inspections_6 shall1 be selected On a random basis, except:

(i) if the tube is recorded as a degraded tube, then an adjacent tube shall be inspected.

(i)The first s~ample inspection during each in serVice ORInspotiOn of each steam generator shall aincludo all non -plugged tubes that previously had detectable wall penetrations (>20-0,) and shall also include tubes in those areas Whore experience has indicated potential problems.

(iii) The second and third sample iRnspections, if required, Fa"y be less than an entire tube length i nspection pro.vide..d the inspection conce.ntrates on those areas of the tub.e.

3.7 age Amendment rNo. i1041u I v i-w

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE-RE-QUIR-B*MErTS 3.17- Stearn Generator Tubos (ContiURed) shaot Rrr -. 0 Rni nn those hn nortiofle v

thAe f hh r ghArA dertect woreA mroviouilv t jdltected.

(i,) To the eato practical,

" Whero ..e . .... plants with im.ilar Water .hemist,', ind'cates 6oc oinimar critical areas to be inspected, then at least 501%of the tubes inspec.ted, shall be from thero eGWelil laeas.

The results of each sample inspection shall be classified into one of the following three categorioc (this classification shall apply to the inspectio~n of tubes and is exclusive of the sleeve inspection requirements in Specification 3.17-(2a)).

Inp*ectinn Results

_ IINo mnre than 5*%of the tubes inspected are degraded and none Of the ins;pected tubes are defectiVe.

C 2 No mnre than 1 e0% of the tubes insecSted are defective, or) between 5%

and 10-0, of the tubes inspected are degraded.

C3 Mre th~an 1%-, of the tubes inspected are defective, Or mor~e than 10%

of the tubes, inspected are degraded.

NOTE: In all inspectiens, previously degraded tubes must exhibit growth of greater than 10% through wall or growth of greater than 25% of the repair limit to be incl'ded in the above calculatiens.

(2a) ý5!GaFR UGReFaMir N-IIA-eve Sample -RlnlA-.lR6PeGl'QR The teaFm generator tube seeve minimum sample size, inspection resul classific-aton, and the corrFepending action required shall be as specified in Table 3 14. The in sewrie inspectionn of steam generatoF tube sleeves shall be performed accord;ng to Specification 3.17-(4)(0), "Tube Sleeve Inspectinn," and at the frequncie secified in Sp9eification 3.17-(3). The inspected tube sleeves shall bh e;rified acpbl per the accepn . c rite.ria;;of Specification 3.17(4). The tube sleeves selected for each in serie inspection shall icdat least 20-0G of the total number of tube sleeves ;i the steam generators and the tube sleeves selected- for thaes inspectionsE hall be se*l*ted nn a randnm basis,

() if the tube sleeve is reoddas6 a degraded tube sleeve and an adjacent tube sleeve exists, then an adjae*nt tube sleeve shall be inspected.

(ii) The first amnple inspection during each in service inspection of each steamn generatoF shall include all tube sleeveG in non plugged tubes that previously had detectable wall penetratioRns

(.20%) and shall alo incIude tube sleeves in those areas where experience has indicated potential problems.

3.17 - Page 2 4 4 49 Amendment No. ,-, 5

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator Tubes (Continued) i;) ",Tothe extent practical, where . Xperienc. in similar plaRts with similar water chomisF-ty 9nRdliatGo Fritcal a*rea to be innpected, then at least 50% of the tube silees6 inpected shall be from theSe *Fr*;al aFreas. Where the number of sleeves in the cr;ti*c* areas represent less than 50% of the initial sample, all 6leeves in the critic*a areas shall be aRGpeoted.

The results of each sample i nspecntio -. shall be classified into one of the following three categories (this GcI6aification shall apply to the inspection of sleeves and is eXclursihe of the tube inspection requirement in Specific;atio-n 3.17(2)).

Gateog ' Inseto R ults Ct1 No mere than 5% of the tube sleeves inspected are degraded and none of the inspected tube sleeves are defective.

C 2 No more than 1 -0,; of the tube sleeves inspected are defective, or between 5% anRd 10% of the tuibe sleeves ;inspected are degraed.

10% of the tube sleeves, inspected are degraded.

NOTE: In all inspections, previou6!y degraded tube sleeves must exhibit growth of greater than 10% through wall or growth of greater than 25-0; of the repair limit to be included in the above caIJatiGoRns (3) ,nc*oction Freue n c-e The above required in ser.'ice inspections of steam generator tubes and tube sleeves Shall be performed at the following freqUencies (inspections shall be performed, unless cthentise speified-,

coincident with refueling outages or any scheduled cold for plant repair and

.hutdown (I1 Inn servie iRnp;etions shall be peFormend at intervAls of net less than 12 nor more than 21 calendar mnnths after the previous inspection, subject to the foll*Wing clarifiations aRnd

1. If a plant operating cycle is less than 12 months, inse*tions mFan be performed at the end- of that cGyle.

2.) lIf ilO consecutive tube inspections fo,llowing seice urder all volatile treatment conditions result in all inspection results falling into the C- category or iftwo consecutive tube ispections demoenrate that previously obsehed degradation has net continued and no additional degradation has occurred, the tube inspectio interval may be extended- to a maximum Of ence per 10 months.

A44 P-'*I[U I M^M**fnl**'l e%

Th~f lf

..... *v v

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator Tubes (Continued)

S. The inSpections Of tubo sloevos shall be c~onfigured tto ensure that each individual tube 6leeVe is6 inspected at least once in 60 monGths, with the following eXception: if the 60 mo~nth time fram falls during an operating cydo8, completion of that cycl i acceptable prior to meeting thie requirement.

(;i

\ I j

J I

• P ........

inru~u inpocIonrrFOuEncUOR1

!=--

i. if results of the in-ser'.ice inspection of the steam generator tubes conducted in acco..,rdaRnce w*i*t;h b*le 3 1 at 10_m-o.nth inter*als fall in CategoFy.C 3,the inspection frequencY shall be increased to at least once per 20 moenths. The increase in inspection frequency shall apply until a subsequent inspectifn meets the conditions specified in Section 3.17(3)(i)2 above, at which time the intcrval can be extended to a 40 month period.

2. if results of the in.e..i;. inspection of tube sleeves conducted in accordance with Table 3 -14 fall into Gatege~y C 3,the inspectio~n frequency shall be increased sc that 100-0%of the tube sleeves in the affected steam generatorF are inspected du!rig subsequent inspectin s. The increase in inspection fequeny shall apply,untLitwo consecutive tube sleeve inspections mneet the conditions for Categor,' G 4 Or tw consecutiVe tube sleeve inspections dlemonstrate that previously obsew.ed degradation has not continued and no additional degradation has, occurred, at

.:hich time the inspection frequency of Specification 3.47-(3)(i)3 shall again apply, (iii) UnRscheduled in-ser~ice inspections shall be performed On each steam generator !n accordance with the first sample inspection specified inTables 3-13 and 3 14 during th shutdown subsequent to any Of the following conditions:'

i. Primary to secondary tube leaks (not includino leaks oriGinatinG from tube to tube sheet welds) in eXcess Of the limits Of Sectio00n 2.1.4 Of the Technical Specifications-,

R.ReiWMic UGGurrenc fgraterF than the Operating Basis Earth "'ua* eI J. oss06Of GOEoiant accIGent rEoGuirie actuation OT lnc onoineerea sareauaras. or

4. Amnain steamq line or main feed-ater Fine break.

3.17 Page 1 Amnendment No. 101, 105

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator Tubes (Continued)

(4) Acceptance Critolria Mft . A; u5et* t, t s onet, tGavew Imn~erfectiGn means an exception to the dimensions, finish Or contour of a tube or sleeve fromn that roquired by fabrication drawings Or specifications. Eddy curront testing indications; below 20% of the nominal tube or sleeve wall thickness, if detectable, ma" be considered as imperfoctionS-.

Degr~adation means- a GeR'icoiduced cracking, wastage, wear OF general corros6io occrrig OR either inside or outside of a tube or sleeve.

Dagrvde-Tbe or Sleeve means a tube Or Sleeve containing imperfections >20-0, of the nominal wall thickness caused by degradation. Any tube Which does not permit the passage of the eddy currentins;pection probe through its entire length and U bend shall be deemed a degraded tube. Any tube sleeve which does not permit the passage of the odd" current inspection probe through its entire length shall be deemed a degraded sleeve.

% Degr~adation mneans the percentage of the tube or sleevefall thickness affected or removed by degradation.

Defect mneans an imperfection Of such severity that it exceeds the plugging or repair limit.

Pluqging or Repair9! Limfit means the imporfactiOn depth at or beyond which the tube shall be removed fromn Eser.ice by plugging Or repaired by sleeving in the affected area because it mnay become Unser~iceable prior to the next inspection. Plugging Or repair limit is equal to 40-0, of the nominal tube wall thickness for the original tube wall. Sleeved tubes shall be plugged upon detection of unacceptable dlegradiation in the pressure boundar,' region of th~e6eeVG.

URSeR'iceable describes the condition of a tube or sleeve if it leaks in excess of analyzed limits Or contains, a defect large enough to affect its, srutFurtal integrity in the event of an Operating Basis Earthquake, a loss6 Of coolant accident, or a steam line or feedwater line break.

Tube or Tubing mneans that portion of the tube which forc)Fthe prinarj' systemA to the secondar,' system pressure boundarye.

Tube Inspection Mean anisetion of the steam generator tube ferom the point of entry~

(hot log side) complet and the U bhend to the top sUppo.t of the cold leg, excluding any areas defined under "TTube Sleeve Inspecti!n".

3.17 Page 5 Amendment No. 104, I 95

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator Tubes (Continued)

Tube Repair o~r RSlAeyeirefers to a proceSE; that re establishes tube 6ew.iceability.

Acceptable tube repairs will be pe~fermed using the Combhustion Engineering, 'Anc Leak Tight Sleeve as described in the prepietar' Combustion EngineeFrig, Inc. Repaot, GEN 630 P, Revision 02, "Repair of 314" O.D. Steam Generator TubeIUsing Leak Tight Sleeves," Juno 1997.

Tube repair includes the Fernova! of plugs that were previously installed as a corrective a preventive measure TOr !R9 purpoe) Of 6leoV!ng Me Wee. /Ao inspectionGU a6 Uef!noa herfein e required prior to returnxing previously plugged tbec to soecn mae.

Tubel Sleevec plection refers t inspection Of the sectionof the steam genefatior t,ue repaird by sleeving. T-his includes the presSUre retaining poFUGns of the parent tube in contact with the sleeve, the sleeve -to-tube weld, and the pressure retaining po~tioR of the sleeve.

i The steam generator shall be deteprined OPERABLE after completing the corresponding actions (plug i Or epair all tubes eXceeding the plugging or repair limit and all tu dleevesd through wall cracks, plug all tubes With containing cotaining defects) required by Tables 3 13 and 3 14-.

(3) Reporting Requirements A report shall be submitted within 180 days after exceeding 2a107 reactor coolant temperature (Tcold) following completion of an inspection performed in accordance with the Specification 5.23, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

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

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

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

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

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

() Following each in- .e'c .Inspcton of steam generator tubes, the number of tubes plugged or repaired in eac--h s-tea-4m generator shall be reported to the CorAmmiss.ionn within 30 day-s-(ii) The complete results of the steam generator tube in Se'Gce i.nspection shall be reported to the Comm*iGGsi within 6 months f*Iowing completion of the inspection. This repl t shalll

1. Number and extent of tubes and tube sleeves inspectd

TECHNICAL SPECIFICATIONS

2. Location and perc-nt of wall thi*kree pnretFration for each impeFf;ation.

3. Identification Of tUbec plugged.

4. Identificantion- of tu-bec repaired by 61eoving.

A .... J .... LLI-- JAA J*P

TECHNICAL SPECIFICATIONS X

• AI I I h IAP MMAI 3j.0 SUKVh-ILLANCE REQlK4Jt-b--N4---"-

3.17 Stea~m Geneiratoir T-ubes (Con~tinued)

(i) Resultr, of 6team genorator tube inspections Which fall into Category C 3 and roguire prom.pt notification Of the Commissio shall be .repor.ted prior to resumpti. of plan operation. The writter foll'wup of this report shall provide a description of investigatic Rn6 conducted to determnine cause of the tube degradationR and corr~ective mneasures take n toG provont recurrence.

• A* XXX 21 -Page 7- AmenaMent No. 195, 22-

TECHNICAL SPECIFICATIONS TABLE, 3 3 S':TEAM GENERATOR TUBE ,NSPE-GT4ON 1st Sample inspection 2nd Samplo Inspection 3rd Sample Inspection

- ,Sample-Sie -iRezt ^A.rtimzRequired R,,,. Action Rqire -Rel Required

-Acftion" A minimum of 300 - None -N/A N!A -N/A N/A tubeshper S.-_

-04r Plug or repair dcfctive -04WNonA -N/A.N!A tuhbes and- inpeet additional 600 tubes in G4 Plug or repair defective -04 None-thdditio ndl 1200 ~tus in - Plug or repair defective this S.G. ti hel

- G 2- Perform,action for C 3 -- A Perform action for C/A result of first sampleut f first*-ample

- 3 Inspect all tubes, in this The- None -- NIA.e.nd NWA S.G., plug or rcpair S4ig -1 defec~tive tubes and inspect 600 tubes in other The seeeind PerfonrM action for C-2 -N/A NIA 2 CZ stC4 result of s~eGcn~d sample The seGend Inspect all tubes in the N/ANA a ~iq seconmd S-.G. and plug or repair defectiive tubhes.

N/A Not applicable,'- 3.17 Pagev R Amendment No. 46,99,4041.5.I 237-

TECHNICAL SPECIFICATIONS TABLE-3-14 STEAM GE.N R TUE- SLEEVE INSPECTION 1st Sample nsecio 2nd Sample Inspection Samp*e*Sizo "Result Action Required ,Reslt Action Required A minimu-m of 20% of the -G 4 None N/A N/A installed tube sleeves Plug tubeS containing defective sleeve& G None and inspect all remnaining installed s~tDlle in this S.G. G4 Plug tubes containing defective 61eeves

- -3 Perfom, action for C 3 res-ult of first

_________mple Inspect all installed sleeves in this S.G., Thposeconnd S.GZ None plug tuber,containing defecrtive sleeves -4 and inspect a minimumn of 20-07 of the-installed sleeves in other S.G. The second S.G. Perform action for C-2 result of first Ad.the tubes Oit defe.tiv.. .e slest tor The second S.G. Inspect all sleeves in the second thenumberofdefectiver Tubles li3 for is 3 S.G. and plug tubes containing NRC pr Tble otifcaton13defective sleeves. Add the tubes-1.4:ith deAfecRtive sleeves to the number of dneefective tubelist for NRC notificatien per Table 3 4-3 N/A -Not-apliriable 3.17 Pa-ae 9 Amendment No. .95

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator Tubes (Continued)

Basis Duringshutdown periods the SGs are inspected as required by this Surveillance requirement (SR) and the Steam GeneratorProgram. NEI 97-06, Steam GeneratorProgram Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam GeneratorProgram. Use of the Steam GeneratorProgram ensures that the inspection is appropriateand consistent with accepted industry practices.

During SG inspections a condition monitoringassessment of the SG tubes is performed. The condition monitoring assessment determines the "asfound" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteriahave been met for the previous operatingperiod.

The Steam GeneratorProgramdetermines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repaircriteria. Inspection scope (i.e.,

which tubes or areas of tubing within the SG are to be inspected)is a function of existing and potential degradationlocations. The Steam GeneratorProgramalso specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradationmorphology, non-destructive examination (NDE) technique capabilities,and inspection locations.

The Steam GeneratorProgram defines the Frequencyof SR 3.17(1). The Frequency is determined by the operationalassessment and otherlimits in the SG examination guidelines (Ref. 6). The Steam GeneratorProgramuses information on existing degradationsand growth rates to determine an inspection Frequencythat provides reasonableassurancethat the tubing will meet the SG performance criteriaat the next scheduled inspection. In addition, Specification 5.23 contains prescriptive requirementsconcerning inspection intervals to provide added assurance that the SG performance criteriawill be met between scheduled inspections.

Duringan SG inspection, any inspectedtube that satisfies the Steam GeneratorProgram repaircriteria is removed from service by plugging. The tube repaircriteriadelineatedin Specification 5.23 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteriawith allowance for errorin the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with otherelements of the Steam GeneratorProgram,ensure that the SG performance criteriawill continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operationalassessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency of priorto exceeding 21 0°F reactorcoolant temperature (Tcoid) following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repaircriteriaare plugged priorto subjecting the SG tubes to significant primary to secondarypressure differential.

References

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

2. 10 CFR 50 Appendix A, GDC 19.

3. 10 CFR 100.

4. ASME Boiler and Pressure Vessel Code, Section I1l, Subsection NB.

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

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

3.17 - Page 2 Amendment No.404, 195

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator Tubes (Continued)

The s~urveilianc ruIr.mentS for inspection of the steamf generator tubes and tube sleeves ensure that the structural integrity Of this p00orto Of the RCS will be maintained. The progarm for in-servic inpection of the teamn genRer r tubes is based on a mFodificati;n of Regulatory Gui do 1.83, Revision

, dated Jul, 1975. The p*rogram for in serVice inspectin of steam generatrF tube 1leeves is based on a modification of ER!i PWR Steam Generator EXa;minFation Guiidielines, Revision 5, Dated September 1997 In serv in spection o steam generatortubing and tue sleeves is esential in rderao aintain su veolianc or the coRnitinR of tne tub*s aRnsleeves in te eveRnt that tnere isaevidence of nmcnanical damage or progressive degradation due to des~ign, mnanufacturing eroso n 'ice conditions that lead to corrosion-.

In sevice inspection of steam eneFrator tubing and tube sleeves also provides a means of charateFriz*iR the nature and cause of any tube or sleeve degFadation so that corrFetive measures can betakeR.

Tubhes.ith d*efots may be repaired by a Crombustion Engineering, Inc. Leak Tight Sleeve. The technical bases for sleeving repair are dersribed in the Prpr;etan' rCombustion Enigineering, Inc.

Report CEN 630 P, Revision 02, "Repair of 3/1" G.D. Steam Generator Tubes Using Leak Tight Sleeves," June 1997.

W/Alhenevexr the results of any steam generatrF tubing in servie ineton fall into Category C 3, these resltsi will be promptly repoF-rted to the CoGmmissioRn prier to the resumptiRn of plant operation. S!uh cases will be considered by the Commissin on a caseby case basis and ma*" result in a r*eguiremn6t for analysis, laboratory' examinations, tests, additional odd" cr-reRnt inspection, and revisin of the T-ehnical Specifications, ifnecessar'.

3.17 - Page 10 Amendment No. 104,-95,228

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.23 Steam Generator(SG) Program A Steam GeneratorProgramshallbe establishedand implemented to ensure that SG tube integrity is maintained. In addition, the Steam GeneratorProgramshallinclude the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "asfound" condition of the tubing with respect to the performance criteriafor structuralintegrity and accident induced leakage. The "asfound" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by othermeans, priorto 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 criteriaare being met.

b. Performance criteriafor SG tube integrity. SG tube integrityshall be maintainedby meeting the performance criteriafor tube structuralintegrity, accident induced leakage, and operational LEAKAGE.

1. Structuralintegrityperformance criterion: All in-service steam generatortubes shall retain structuralintegrity over the full range of normal operatingconditions (includingstartup, operationin the power range, hot standby, and cool down and all anticipatedtransients included in the design specification)and design basis accidents. This includes retaining a safety factor of 3.0 against burst undernormal steady state full power operation primary-to-secondarypressure differentialand a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondarypressure differentials.Apart from the above requirements, additionalloading conditions associatedwith the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associatedloads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantlyaffect burst or collapse shall be determined and assessedin combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondaryloads.

2. Accident induced leakage performance criterion: The primary to secondaryaccident 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 totalleakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed I gpm per SG.

3. The operationalLEAKAGE performance criterion is specified in LCO 2.1.4, "ReactorCoolant System Leakage Limits."

c. Provisionsfor SG tube repaircriteria. 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.

5.0 - Page 19 Amendment No.

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.23 Steam Generator(SG) Program (continued)

d. Provisions for SG tube inspections. PeriodicSG tube inspections shall be performed. The number and portions of the tubes inspectedand methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferentialcracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicabletube repaircriteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirementsof d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradationshall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

2. Inspect 100% of the tubes at sequentialperiods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequentialperiod shall be consideredto 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 remaining50% by the refueling outage nearestthe 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.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradationmechanism 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, diagnosticnon-destructive testing, or engineeringevaluation indicates that a crack-like indicationis not associatedwith a crack(s), then the indication need not be treatedas a crack.

e. Provisions for monitoring operationalprimary to secondaryLEAKAGE.

5.0 - Page 20 Amendment No.

LIC-06-0087 Page 1 ATTACHMENT 3 Proposed Technical Specification Pages (clean)

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 2.13 DELETED 2.14 Engineered Safety Features System Initiation Instrumentation Settings 2.15 Instrumentation and Control Systems 2.16 River Level 2.17 Miscellaneous Radioactive Material Sources 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 Toxic Gas Monitors 2.23 Steam Generator (SG) Tube Integrity 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 DELETED 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 Radioactive Material Sources Surveillance 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator (SG) Tube Integrity 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. 11,27,32,38,43,46,54,60,84,846, 93,97,104,122,136,152, 160,176,183, 214,230, 236

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Training 5.5 Not Used 5.6 Not Used 5.7 Safety Limit Violation 5.8 Procedures 5.9 Reporting Requirements 5.9.1 Not Used 5.9.2 Not Used 5.9.3 Special Reports 5.9.4 Unique Reporting Requirements 5.9.5 Core Operating Limits Report 5.9.6 RCS Pressure-Temperature Limits Report (PTLR) 5.10 Record Retention 5.11 Radiation Protection Program 5.12 DELETED 5.13 Secondary Water Chemistry 5.14 Systems Integrity 5.15 Post-Accident Radiological Sampling and Monitoring 5.16 Radiological Effluents and Environmental Monitoring Programs 5.16.1 Radioactive Effluent Controls Program 5.16.2 Radiological Environmental Monitoring Program 5.17 Offsite Dose Calculation Manual (OCDM) 5.18 Process Control Program (PCP) 5.19 Containment Leakage Rate Testing Program 5.20 Technical Specification (TS) Bases Control Program 5.21 Containment Tendon Testing Program 5.22 Diesel Fuel Oil Testing Program 5.23 Steam Generator (SG) Program 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS 6.1 Deleted 6.2 Deleted 6.3 Deleted 6.4 Deleted TOC - Page 3 Amendment No. 32, 34, *2,54, 55, 57, 73, 80, 86, 93, 99,4141,152, 157, 184, 185, 221 236,237

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS TABLE SECTION 1- 1 R PS LS SS ............................................................................................ Section1.0 2- 1 ESFS Initiation Instrumentation Setting Limits ............ ................................ Section 2.14 2-2 Instrument Operating Requirements for RPS ................................................ Section 2.15 2-3 Instrument Operating Requirements for Engineered Safety Features ................. Section 2.15 2-4 Instrument Operating Conditions for Isolation Functions .................................. Section 2.15 2-5 Instrumentation Operating requirements for Other Safety Feature Functions ....... Section 2.15 2- 9 RCS Pressure Isolation Valves .................................................................. Section 2.1 2- 10 Post-Accident Monitoring, Instrumentation Operating Limits ............................. Section 2.21 2 - 11 Toxic Gas Monitors Operating limits ............................................................ Section 2.22 3- 1 Minimum Frequencies for Checks, Calibrations, and Testing of RPS .................. Section 3.1 3 -2 Minimum Frequencies for Checks, Calibrations and Testing of ........................... Section 3.1 Engineered Safety Features, Instrumentation and Controls 3 -3 Minimum Frequencies for Checks, Calibrations and Testing ............................. Section 3.1 of Miscellaneous Instrumentation and Controls 3 - 3a Minimum Frequency for Checks, Calibrations, and Functional ........................... Section 3.1 Testing of Alternate Shutdown Panels (AI-185 and AI-212) and Emergency Auxiliary Feedwater Panel (Al-1 79) Instrumentation And Control Circuits 3-4 Minimum Frequencies for Sampling Tests .................................................... Section 3.2 3- 5 Minimum Frequencies for Equipment Tests.................................................. Section 3.2 3 -6 Reactor Coolant Pump Surveillance ............................................................ Section 3.3 TOC - Page 4 Amendment No.

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS TABLE 5.2-1 M inim um Shift Crew Com position ............................................................................................. Section 5.0 TOC - Page 5 Amendment No. 4 16,125,142,145,152,176,-19*

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS (ALPHABETICAL ORDER)

Continued TABLE DESCRIPTION SECTION 2-10 Post-Accident Monitoring Instrumentation Operating Limits ................................................... Section 2.21 2-9 RCS Pressure Isolation Valves ................................................................................................. Section 2.1 3-6 Reactor Coolant Pump Surveillance ......................................................................................... Section 3.3 1-1 RPS LSSS ................................................................................................................................. Section 1.0 2-11 Toxic Gas Monitoring Operating Limits ................................................................................... Section 2.22 TOC - Page 7 Amendment No. 116,145,152,17 -5

TECHNICAL SPECIFICATION DEFINITIONS E - Averagqe Disintegration Enerqy E is 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 gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 minutes making up at least 95% of the total non-iodine radioactivity in the coolant.

Offsite Dose Calculation Manual (ODCM)

The document(s) that contain the methodology and parameters used in the calculations of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent radiation monitoring Warn/High (trip) Alarm setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain:

1) The Radiological Effluent Controls and the Radiological Environmental Monitoring Program required by Specification 5.16.

2) Descriptions of the information that should be included in the Annual Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports required by Specifications 5.9.4.a and 5.9.4.b.

Unrestricted Area Any area at or beyond the site boundary access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

Core Operating Limits Report (COLR)

The Core Operating Limits Report (COLR) is a Fort Calhoun Station Unit No. 1 specific document that provides core operating limits for the current operating cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Section 5.9.5. Plant operation within these operating limits is addressed in the individual specifications.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except rea ctor coolant pump (RCP) seal leakoff), that is captured and conducted to collection systems or a sump or collecting tank,

2. 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

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE),

Definitions - Page 8 Amendment No. 67,86,141,152,164,221,226 Correction Letter of 06-17-2004

TECHNICAL SPECIFICATION DEFINITIONS

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal leakoff) that is not identified LEAKAGE, and

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

I RCS Pressure-Temperature Limits Report (PTLR)

The PTLR is a fluence dependent document that provides Limiting Conditions for Operation (LCO) in the form of pressure-temperature (P-T) limits to ensure prevention of brittle fracture. In addition, this document establishes power operated relief valve setpoints which provide low temperature overpressure protection (LTOP) to assure the P-T limits are not exceeded during the most limiting LTOP event. The P-T limits and LTOP criteria in the PTLR are applicable through the effective full power years (EFPYs) specified in the PTLR. NRC approved methodologies are used as the bases for the information provided in the PTLR.

References (1) USAR, Section 7.2 (2) USAR, Section 7.3 Definitions - Page 9 Amendment No. 226 Correction Letter of 06-17-2004

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable Components (Continued)

(5) DELETED (6) Both steam generators shall be filled above the low steam generator water level trip set point and available to remove decay heat whenever the average temperature of the reactor coolant is above 300 0 F. I (7) Maximum reactor coolant system hydrostatic test pressure shall be 3125 psia. A maximum of 10 cycles of 3125 psia hydrostatic tests are allowed.

(8) Reactor coolant system leak and hydrostatic test shall be conducted within the limitations of the pressure and temperature limit Figure(s) shown in the PTLR.

(9) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia. A minimum measured temperature of 73°F is required. Only 10 cycles are permitted.

(10) Maximum steam generator steam side leak test pressure shall not exceed 1000 psia. A minimum measured temperature of 73°F is required.

(11) Low Temperature Overpressure Protection (LTOP)

(a) The LTOP enable temperature and RCP operations shall be maintained in accordance with the PTLR.

(b) The unit can not be placed on shutdown cooling until the RCS has cooled to an indicated RCS temperature of less than or equal to 300 0 F.

(c) If no reactor coolant pumps are operating, a non-operating reactor coolant pump shall not be started while Tc is below the LTOP enable temperature stated in the PTLR unless there is a minimum indicated pressurizer steam space of at least 50% by volume.

2.1 - Page 3 Amendment No. 39,56,66,71,149,136, 161,188, 207, 221

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.4 Reactor Coolant System Leakage Limits Applicability Applies to the leakage rates of the reactor coolant system whenever the reactor coolant temperature (TCOld) is greater than 210 OF.

Obiective To specify limiting conditions of the reactor coolant system leakage rates.

Specifications To assure safe reactor operation, the following limiting conditions of the reactor coolant system leakage rates must be met:

(1) RCS operational LEAKAGE shall be limited to:

a. No Pressure Boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

c. 10 gpm identified LEAKAGE,

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

(2) If RCS operational LEAKAGE limits of (1), above, are not met for reasons other than Pressure Boundary LEAKAGE or primary to secondary LEAKAGE, then reduce LEAKAGE to meet limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

(3) If the Required Action and associated completion time of (2), above, is not met, OR Pressure Boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within limits, then be in MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND be in MODE 4, Cold Shutdown, within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

(4) To determine leakage to the containment, a containment atmosphere radiation monitor (gaseous or particulate) or dew point instrument, and a containment sump level instrument must be operable.

a. With no containment sump level instrument operable, verify that a containment atmosphere radiation monitor is operable, and restore the containment sump level instrument to operable status within 30 days.

b. With no containment atmosphere radiation monitor and no dewpoint instrument operable, restore either a radiation monitor or dewpoint instrument to operable status within 30 days.

c. With only the dewpoint instrument operable, or with no operable instruments, enter Specification 2.0.1 immediately.

2.1 - Page 13 Amendment No. 32,165,195,

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.4 Reactor Coolant System Leakage Limits (Continued)

c. 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 unidentified LEAKAGE and is well within the capability of the RCS makeup system. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE through Any One SG The 150 gallon per day operational limit on primary to secondary LEAKAGE through any one SG is based upon guidance in NEI 97-06, Steam Generator Program Guidelines. The Steam Generator Program operational LEAKAGE performance Criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

APPLICABILITY The potential for RCPB LEAKAGE is greatest when the RCS is pressurized, that is, when the reactor coolant temperature (Tco1d) is greater than 210°F.

In MODES 4 and 5, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

REQUIRED ACTIONS (2).

Unidentified LEAKAGE or identified LEAKAGE in excess of the LCO limits must be reduced to meet limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This Completion Time 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.

REQUIRED ACTIONS (3)

If any pressure boundary LEAKAGE exists or primary to secondary LEAKAGE is not within limits, or if unidentified or identified LEAKAGE cannot be reduced to meet limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences.

The reactor must be brought to MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4, Cold Shutdown, within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

2.1 - Page 16 Amendment No. 32,165, 226

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator (SG) Tube Integrity Applicability Applies whenever the reactor coolant temperature (Tcoid) is greater than 21 0°F.

Obiective To ensure that SG tube integrity is maintained.

Specification NOTE: Separate Condition entry is allowed for each SG Tube.

(1) The following conditions shall be maintained:

(a) SG tube Integrity shall be maintained, and (b) All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

(2) If the requirements of (1)(b) above are not met for one or more SG tubes, then perform the following:

(a) Verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection within 7 days, and (b) Plug the affected tube(s) in accordance with the Steam Generator Program prior to exceeding 210°F reactor coolant temperature (Tcold) following the next refueling outage or SG tube inspection.

(3) If the Required Action and associated completion time of (2), above, is not met, or if SG tube integrity is not maintained, then be in MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND be in MODE 4, Cold Shutdown, within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Basis 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. Steam generator 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 Technical Specification 2.1.1, "Operable Components."

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.

2.23 - Page 1 Amendment No.

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator (SG) Tube Integrity (continued)

Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear.

These degradation mechanisms can impair tube integrity ifthey are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 5.23, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 5.23, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 5.23. 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 (Ref. 1).

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 the operational LEAKAGE rate limits in Technical Specification 2.1.4, "Reactor Coolant System Leakage Limits," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes releases of activity occur from the faulted steam generator to the environment via the condenser air ejector and Main Steam Safety Valves (MSSVs) and Atmospheric Dump Valves (ADVs). The release via the condenser air ejector starts at the initiation of the event and continues to the reactor trip, while the release via the MSSVs/ADVs starts at the reactor trip and continues for the duration of the event."

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 gallon per minute or is assumed to increase to 1 gallon per minute 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 Technical Specification 2.1.3, "Reactor Coolant Radioactivity," limits. 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 GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

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

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.

The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.23, "Steam Generator Program," and describe acceptable SG tube performance.

The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

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

2.23 - Page 2 Amendment No.

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator (SG) Tube Inteqrity (continued)

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero."

The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential 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 Ill, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

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

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in Technical Specification 2.1.4.

Reactor Coolant System Leakage Limits," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. 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.

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 when Tco;d is > 210°F.

RCS conditions are far less challenging in MODES 4 and 5 than during MODES 1, 2, and 3. In MODES 4 and 5, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

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.

Specification 2.23(2) 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 Technical Specification 3.17. 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.

2.23 - Page 3 Amendment No.

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator (SG) Tube Integrity (continued)

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, Specification 2.23(3) applies.

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Required Action 2.23(2)b 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 exceeding 210°F reactor coolant temperature (Tcold) following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

If the Required Actions and associated Completion Times of Technical Specification 2.23(2) are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

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

References

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

2. 10 CFR 50 Appendix A, GDC 19.

3. 10 CFR 100.

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

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

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

2.23 - Page 4 Amendment No.

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Sampling Tests (continued)

The Safety Injection (SI) pump room air treatment system consists of charcoal adsorbers which are installed in normally bypassed ducts. This system is designed to reduce the potential release of radioiodine in SI pump rooms during the recirculation period following a DBA. The in-place and laboratory testing of charcoal adsorbers will assure system integrity and performance.

Pressure drops across the combined HEPA filters and charcoal adsorbers, of less than 9 inches of water for the control room filters (VA-64A & VA-64B) and of less than 6 inches of water for each of the other air treatment systems will indicate that the filters and adsorbers are not clogged by amounts of foreign matter that would interfere with performance to established levels. Operation of each system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month will demonstrate operability and remove excessive moisture build-up in the adsorbers.

The hydrogen purge system provides the control of combustible gases (hydrogen) in containment for a post-LOCA environment. The surveillance tests provide assurance that the system is operable and capable of performing its design function. VA-80A or VA-80B is capable of controlling the expected hydrogen generation (67 SCFM) associated with 1) Zirconium - water reactions, 2) radiolytic decomposition of sump water and 3) corrosion of metals within containment. The system should have a minimum of one blower with associated valves and piping (VA-80A or VA-80B) available at all times to meet the guidelines of Regulatory Guide 1.7 (1971).

If significant painting, fire or chemical release occurs such that the HEPA filters or charcoal adsorbers could become contaminated from the fumes, chemicals or foreign materials, testing will be performed to confirm system performance.

Demonstration of the automatic and/or manual initiation capability will assure the system's availability.

Verifying Reactor Coolant System (RCS) leakage to be within the LCO limits ensures the integrity of the Reactor Coolant Pressure Boundary (RCPB) is maintained. Pressure boundary leakage would at first appear as unidentified leakage and can only be positively identified by inspection. Unidentified leakage is determined by performance of an RCS water inventory balance. Identified leakage is then determined by isolation and/or inspection. Since Primary to Secondary Leakage of 150 gallons per day cannot be measured accurately by an RCS water inventory balance, note .***" for line item 8a on Table 3-5 states that the Reactor Coolant System Leakage surveillance is not applicable to Primary to Secondary Leakage.

Primary to secondary leakage is measured by performance of effluent monitoring within the secondary steam and feedwater systems.

3.2 - Page 2 Amendment No. 15,67,128,138,164 TSBC-03 006 0

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Sampling Tests (continued)

Table 3-5, Item 8b verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this surveillance requirement is not met, compliance with LCO 3.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

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

The Surveillance Frequency of daily 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 (Ref. 5).

References

1) USAR, Section 9.10

2) ASTM D4057-95(2000), ASTM D975-98b, ASTM D4176-93, ASTM D129-00, ASTM D2622-87, ASTM D287-82, ASTM 6217-98, ASTM D2709-96

3) ASTM D975-98b, Table 1

4) Regulatory Guide 1.137

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

3.2 - Page 3b Amendment No. 229

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Test Frequency Reference

1. Control Element Drop times of all full-length CEA's Prior to reactor criticality after each 7.5.3 Assemblies removal of the reactor vessel closure head

2. Control Element Partial movement of all CEA's Q 7 Assemblies (Minimum of 6 in)

3. Pressurizer Safety Verify each pressurizer safety valve R 7 Valves is OPERABLE in accordance with the Inservice Testing Program.

Following testing, lift settings shall be 2485 psig +/-1% and 2530 psig +/-1%

respectively.

4. Main Steam Safety Set Point R 4 Valves

5. DELETED

6. DELETED

7. DELETED 8a. Reactor Coolant Evaluate D* 4 System Leakage***

8b Primary to Secondary Continuous process 4 Leakage **** radiation monitors or radiochemical grab sampling 9a Diesel Fuel Supply Fuel Inventory M 8.4 9b. Diesel Lubricating Oil Lube Oil Inventory M 8.4 Inventory 9c. Diesel Fuel Oil Test Properties In accordance with the Diesel Fuel 8.4 Properties Oil Testing Program 9d. Required Diesel Air Pressure M 8.4 Generator Air Start Receiver Bank Pressure

• Whenever the system is at or above operating temperature and pressure.

3.2 - Page 6 Amendment No. 15,24,128,160,166,169, 471,249, 229

'Not applicable to primary to secondary LEAKAGE.

I

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS

• • • • Verify primary to secondary LEAKAGE is < 150 gallons per day through any one SG.

This surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

3.2-Page 7 Amendment No

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Test Frequency Reference 9e. Check for and Check for Water and Remove Q 8.4 Remove Accumulated Water from Each Fuel Oil Storage Tank 10a. Charcoal and HEPA 1. In-Place Testinq** 9.10 Filters for Control Charcoal adsorbers and HEPA On a refueling frequency or every 720 Room filter banks shall be leak hours of system operation or after each tested and show >99.95% complete or partial replacement of the Freon (R-11 or R-112) and charcoal adsorber/HEPA filter banks, or cold DOP particulates after any major structural maintenance on removal, respectively. the system housing or following significant painting, fire or chemical releases in a ventilation zone communicating with the system.

2. Laboratory Testin"**

Verify, within 31 days after removal, On a refueling frequency or every 720 that a laboratory test of a sample of hours of system operation or after any the charcoal adsorber, when obtained structural maintenance on the HEPA filter or in accordance with Regulatory charcoal adsorber housing or following Position C.6.b of Regulatory Guide significant painting, fire or chemical release in 1.52, Revision 2, March 1978, shows a ventilation zone communicating with the methyliodide penetration less than system.

0.175% when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C (86 0 F) and a relative humidity of 70%.

• • Tests shall be performed in accordance with applicable section(s) of ANSI N510-1980.

3.2 - Page 8 Amendment No 1.5 ,128,169,198,229

.. I

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Test Frequency Reference 10a. (continued) 3. Overall System Operation

a. Each circuit shall be operated. Ten hours every month.

b. The pressure drop across the R combined HEPA filters and charcoal adsorber banks shall be demonstrated to be less than 9 inches of water at system design flow rate.

c. Fan shall be shown to operate R within + 10% design flow.

4. Automatic and manual initiation of R the system shall be demonstrated.

10b. Charcoal Adsorbers 1. In-Place Testing*

for Spent Fuel Charcoal adsorbers shall be On a refueling frequency or every 720 6.2 Storage Pool Area leak tested and shall show hours of system operation, or after 9.10

>99% Freon (R-1 1 or R-1 12) each complete or partial replacement of removal. the charcoal adsorber bank, or after any major structural maintenance on the system housing or following significant painting, fire or chemical release in a ventilation zone communicating with the system.

2. Laboratory Testing Verify, within 31 days after removal, On a refueling frequency or every 720 that a laboratory test of a sample of hours of system operation or after any the charcoal adsorber, when obtained structural maintenance on the HEPA filter or in accordance with Regulatory charcoal adsorber housing or following Position C.6.b of Regulatory Guide significant painting, fire or chemical release in 1.52, Revision 2, March 1978, shows a ventilation zone communicating with the methyliodide penetration less than system.

10% when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (86°F) and a relative humidity of 95%.

• • Tests shall be performed in accordance with applicable section(s) of ANSI N510-1980.

3.2 - Page 9 Amendment No. 15,24,52,128,15,9,169, 198, 229

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Test Frequency Reference 10b. (continued) 3. Overall System Operation

a. Operation of each circuit Ten hours every month.

shall be demonstrated.

b. Volume flow rate through R charcoal filter shall be shown to be between 4500 and 12,000 cfm.

4. Manual initiation of the R system shall be demon-strated.

10c. Charcoal Adsorbers 1. In-Place Testinq** On a refueling frequency or every 9.10 for S.I. Pump Room Charcoal adsorbers shall be 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation, or 6.2 leak tested and shall show after each complete or partial

>99% Freon (R-11 or R-112) replacement of the charcoal adsorber bank, removal. or after any major structural maintenance on the system housing or following significant painting, fire or chemical release in any ventilation zone communicating with the system.

2. Laboratory Testinq Verify, within 31 days after removal, On a refueling frequency or following 720 that a laboratory test of a sample of hours of system operation or after any the charcoal adsorber, when obtained structural maintenance on the HEPA filter or in accordance with Regulatory charcoal adsorber housing or following Position C.6.b of Regulatory Guide significant painting, fire or chemical release in 1.52, Revision 2, March 1978, shows a ventilation zone communicating with the system.

methyliodide penetration less than 10% when tested in accordance with ASTM D3803-1989 at a temperature of 300 C (86 0 F) and a relative humidity of 95%.

3. Overall System Operation

a. Operation of each circuit Ten hours every month.

shall be demonstrated.

b. Volume flow rate shall be R shown to be between 3000 and 6000 cfm.

• • Tests shall be performed in accordance with applicable section(s) of ANSI N510-1980.

3.2 - Page 10 Amendment No. 15,24,52,128,169,198,229 1

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Test Frequency Reference 10c. (continued) 4. Automatic and/or manual initi- R ation of the system shall be demonstrated.

11. Containment 1. Demonstrate damper action. 1 year, 2 years, 5 years, and every 5 9.10 Ventilation System years thereafter.

Fusible Linked Dampers 2. Test a spare fusible link.

12. Diesel Generator Calibral R 8.4.3 Under-Voltage Relays

13. Motor Operated Verify the contactor pickup value at R Safety Injection <85% of 460 V.

Loop Valve Motor Starters (HCV-311, 314, 317, 320, 327, 329, 331,333, 312, 315, 318,321)

14. Pressurizer Heaters Verify control circuits operation R for post-accident heater use.

15. Spent Fuel Pool Test neutron poison samples for 1, 2, 4, 7, and 10 years after Racks dimensional change, weight, neutron installation, and every 5 years attenuation change and specific thereafter.

gravity change.

16. Reactor Coolant 1. Verify all manual isolation During each refueling outage just Gas Vent System valves in each vent path are prior to plant start-up.

in the open position.

2. Cycle each automatic valve in the R vent path through at least one complete cycle of full travel from the control room. Verification of valve cycling may be determined by observation of position indicating lights.

3. Verify flow through the reactor R coolant vent system vent paths.

3.2 -Page 11 Amendment No. 41,54,60,75,77,80,155,169,182,218,229

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Test Frequency

17. Hydrogen Purge 1. Verify all manual valves are operable by R System completing at least one cycle.

2. Cycle each automatic valve through at R least one complete cycle of full travel from the control room. Verification of the valve cycling may be determined by the observation of position indicating lights.

3. Initiate flow through the VA-80A and VA-80B blowers, HEPA filter, and charcoal adsorbers and verify that the system operates for at least (a) 30 minutes with suction from the a) M auxiliary building (Room 59)

(b) 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with suction from the b) R containment

4. Verify the pressure drop across the R VA-82 HEPAs and charcoal filter to be less than 6 inches of water. Verify a system flow rate of greater than 80 scfm and less than 230 scfm during system operation when tested in accordance with 3b. above.

18. Shutdown Cooling 1. Verify required shutdown cooling loops are S (when shutdown cooling is required by TS 2.8).

OPERABLE and one shutdown cooling loop is IN OPERATION.

2. Verify correct breaker alignment and indicated W (when shutdown cooling is required by TS 2.8).

power is available to the required shutdown cooling pump that is not IN OPERATION.

3.2-Page 12 Amendment No. 3 9

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Test Frequency

19. Refueling Water Level Verify refueling water level is 3 23 ft. above Prior to commencing, and daily during CORE ALTERATIONS the top of the reactor vessel flange. and/or REFUELING OPERATIONS inside containment.

20. Spent Fuel Pool Level Verify spent fuel pool water level is 3 23 ft. Prior to commencing, and weekly during REFUELING above the top of irradiated fuel assemblies seated OPERATIONS in the spent fuel pool.

in the storage racks.

21. Containment Penetrations Verify each required containment penetration is Prior to commencing, and weekly during CORE ALTERATIONS in the required status. and/or REFUELING OPERATIONS in containment.

22. Spent Fuel Assembly Verify by administrative means that initial Prior to storing the fuel assembly in Region 2 (including Storage enrichment and burnup of the fuel assembly is in peripheral cells).

accordance with Figure 2-10.

23. P-T Limit Curve Verify RCS Pressure, RCS temperature, and This test is only required during RCS heatup and cooldown RCS heatup and cooldown rates are within operations and RCS inservice leak and hydrostatic testing.

the limits specified by the P-T limit Figure(s) While these operations are occurring, this test shall be performed shown in the PTLR. every 30 minutes.

24. Spent Fuel Cask Loading Verify by administrative means that initial Prior to placing the fuel assembly in a spent fuel cask in enrichment and burnup of the fuel assembly the spent fuel pool.

is in accordance with Figure 2-11.

1 88 22 1 3.2 - Page 13 Amendment No. , , 2 39

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator (SG) Tubes Integrity Applicability Applies to in-service surveillance of steam generator tubes.

Objective To ensure the integrity of the steam generator tubes.

Specifications Each steam generator shall be demonstrated OPERABLE by performance of the following:

(1) Verify SG Tube Integrity in accordance with the Steam Generator Program.

(2) Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to exceeding 210°F reactor coolant temperature (Tcold).

(3) Reportinq Requirements A report shall be submitted within 180 days after exceeding 210°F reactor coolant temperature (Tcold) following completion of an inspection performed in accordance with the Specification 5.23, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

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

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

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

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

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

3.17 - Page 1 Amendment No. 404-,495

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator Tubes (Continued)

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

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

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.17(1). The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.23 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 5.23 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 1 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 exceeding 21 0°F reactor coolant temperature (Tcod) 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 Appendix A, GDC 19.

3. 10 CFR 100.

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

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

August 1976.

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

3.17 - Page 2 Amendment No. 104,4§5

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.23 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

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

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

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine ifthe 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.

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

3. The operational LEAKAGE performance criterion is specified in LCO 2.1.4, "Reactor Coolant System Leakage Limits."

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

5.0 - Page 19 Amendment No.

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.23 Steam Generator (SG) Program (continued)

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

2. Inspect 100% of the tubes at sequential periods of 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.

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

e. Provisions for monitoring operational primary to secondary LEAKAGE.

5.0 - Page 20 Amendment No.

LIC-06-0087 Page 1 ATTACHMENT 4 Location of TSTF-449 Requirements in FCS TS

LIC-06-0087 Page 2 Location of TSTF-449 Revisions in FCS TS TSTF-449 FCS TS Revised TS definition of Revised TS definition of LEAKAGE.

LEAKAGE Revised TS [3.4.13], RCS Revised TS 2.1.4, Reactor Coolant System Operational Leakage

[Reactor Coolant System] Limits, addresses the LCO and ACTION revisions of TS 3.4.13.

Operational Leakage Revised TS 3.2 addresses the SURVEILLANCE REQUIREMENTS revisions of TS 3.4.13. Specifically, line 8 of Table 3-5 has been replaced by lines 8a and 8b of Table 3-5, and new footnotes designated "***" and "****" have been added.

New TS [3.4.18], Steam Generator New TS 2.23, Steam Generator (SG) Tube Integrity, addresses the Tube Integrity LCO, APLICABILITY, and ACTION portions of TS 3.4.18.

Revised TS 3.17, Steam Generator Tube Integrity, Specifications 1 and 2, address the SURVEILLANCE REQUIREMENTS portion of TS 3.4.18.

Revised TS [5.5.9], Steam New Administrative Controls TS 5.23, Steam Generator (SG)

Generator (SG) Program Program Revised TS [5.6.9], Steam Revised TS 3.17, Steam Generator Tube Integrity, Specification (5).

Generator Tube Inspection Report Insert B 3.4.13 A First sentence was incorporated in the Basis for TS 2.23. Second sentence was not incorporated because the licensing basis main steam line break analysis for FCS assumes 0 gpm leakage into the intact steam generator.

Insert B 3.4.13 B Incorporated in the Basis for FCS TS 2.1.4.

Insert B 3.4.13 C Incorporated in the Basis for FCS TS 3.2.

Insert B 3.4.13 D (CEOG) Incorporated in the Basis for FCS TS 3.2.

Insert B 3.4.13 E Reference 4 is incorporated in the Basis for FCS TS 3.17. Reference 5 is incorporated in the Basis for FCS TS 3.2.

CEOG STS page B 3.4.4-2 Not applicable because the Bases markup deletes text in the Standard TS Bases that is not included in the FCS TS Bases.

CEOG STS page B 3.4.5-2 Not applicable because the Bases markup deletes text in the Standard TS Bases that is not included in the FCS TS Bases.

CEOG STS page B 3.4.6-2 Not applicable because the Bases markup deletes text in the Standard TS Bases that is not included in the FCS TS Bases.

CEOG STS page B 3.4.7-3 Not applicable because the Bases markup deletes text in the Standard TS Bases that is not included in the FCS TS Bases.

CEOG STS page B 3.4.13-2 Insert B 3.4.13 A on CEOG STS page B 3.4.13-2 is addressed as noted above. The remaining markups are not applicable because they revise text that is not in the FCS TS Bases.

CEOG STS page B 3.4.13-3 Deletion of item "d." is not applicable because the deleted text is not included in the FCS TS Bases. Insert B 3.4.13 B on CEOG STS page B 3.4.13-3 is addressed as described above.

LIC-06-0087 Page 3 T.np2tinn nfT~TF-d4Q flpvidnnq in FC~ Th TSTF-449 FCS TS CEOG STS page B 3.4.13-4 The revisions in the ACTION section were incorporated in FCS TS Basis 2.1.4.

In the SURVEILLANCE REQUIREMENTS section, deletion of the text regarding measurement of primary to secondary leakage using a water balance method is not applicable because the FCS TS do not include that method for determining primary to secondary leakage.

The essence of the markups explaining the notes is incorporated in the Basis for FCS TS 3.2.

CEOG STS page B 3.4.13-5 Inserts B 3.4.13 C, B 3.4.13 D (CEOG) and B 3.4.13 E on CEOG STS page 3.4.134 are addressed above.

CEOG STS pages B 3.4.18-1 Incorporated in the Basis for FCS TS 2.23.

through B 3.4.18-5, excluding the SURVEILLANCE REQUIREMENTS discussion on page B 3.4.18-5.

CEOG STS page B 3.4.18-5, Incorporated in the Basis for FCS TS 3.17.

beginning with the SURVEILLANCE REQUIREMENTS discussion, through page B 3.4.18-6.

CEOG STS page B 3.4.18-7 The cited references are incorporated in the Bases for FCS TS 2.23 and 3.17, as appropriate.

LIC-06-0087 Page 1 ATTACHMENT 5 Responses to Requests for Additional Information Related to the May 30, 2006 TSTF-449 Submittal

LIC-06-0087 Page 2 Responses to Requests for Additional Information Related to the May 30, 2006 TSTF-449 Submittal NRC Request #1 In Technical Specification (TS) 2.1.1(6), you proposed that "each steam generator shall be demonstrated operable by the performance of the requirements specified in Section 3.17 prior to exceeding a reactor coolant temperature of 300 0 F."

In addition, you indicate that steam generator operability can be achieved by verifying tube integrity in accordance with the Steam Generator Program and by verifying that each inspected tube that satisfies the tube repair criteria is plugged prior to exceeding a cold leg temperature of 210 0F. As currently written, steam generator operability can be demonstrated by simply performing the surveillance requirement.

TSTF-449 expanded the definition of steam generator operability by deleting the phrase that operability is in accordance with the steam generator tube surveillance program. Please discuss your plans for making your proposal consistent with TSTF-449 (in TS 2.1.1 and TS 3.17).

In addition, given the unique structure of your TS, please clarify why TS 2.1.1(6) is still needed in light of the proposed addition of TS 2.23 which will establish new limiting conditions for operation for the steam generators.

OPPDResponse:

The sentence in TS 2.1.1(6), concerning demonstrating that each SG is operable by performance of an inservice inspection program, is duplicate information to TS 2.23, which specifies the requirements for steam generatortube integrity as part of the reactorcoolant pressure boundary in more detail,and will be deleted.

NRC Request #2 There are inconsistencies between the temperature in TS 2.1.1(6) (temperature of 300 0F), TS 3.17(2)

(cold-leg temperature of 210 0F), and TSTF-449 (average coolant temperature of 200 0 F). In addition, there are no limits on the reactivity condition in your proposed TS requirements; unlike TSTF-449.

Please discuss your plans for modifying your proposal to make it consistent with TSTF-449.

OPPD Response:

The Tc*Id 210OF requirementis used in other technicalspecifications. TS 2.1.1(6) will be deleted in response to question 1.

NRC Request #3 It would appear that the title for entry 8b in Table 3-5 would more appropriately be "primary- to-secondary leakage" rather than steam generator tube integrity. Please discuss why steam generator tube integrity was chosen rather than primary-to-secondary leakage. Alternatively, modify your proposal to indicate primary-to-secondary leakage for item 8b. The staff notes that the leakage limit will not ensure steam generator tube integrity.

LIC-06-0087 Attachment 5 Page 3 OPPDResponse:

The table entry will be revised to labeled primary-to-secondaryleakage per the Staff request.

NRC Request #4 In Table 3-5, the frequency is labeled as "D." Please discuss where "D" is defined within the specification. Alternatively, discuss your plans for specifying that the frequency is "daily."

OPPDResponse:

TS 3.0.2 defines the letter designations.

NRC Request #5 In Table 3-5, the "Test" for "primary to secondary leakage" is listed as "evaluate". The meaning of this term is not clear. Isn't the "Test" for primary to secondary leakage, continuous monitoring of the effluent (steam and feedwater systems) for radioactive isotopes or performing radiochemical analyses of grab samples of the steam and feedwater systems? Similarly, isn't the "Test" for reactor coolant system leakage, a water inventory balance? Please clarify.

OPPDResponse:

The technical basis is provided in the last paragraphof TS 3.2: "The primary to secondary LEAKAGE is determined using the continuous process radiationmonitors or radiochemicalgrab sampling in accordancewith the EPRI guidelines." The table 3.5 test will be modified to change "evaluate" by replacing it with "Continuous process radiation monitors or radiochemical grab sampling".

NRC Request #6 In the second paragraph on page 5 of TS 3.2, it would appear that the reference to the limiting condition for operation should be TS 2.23 rather than TS 3.17 since TS 3.17 are the surveillance requirements.

Please discuss your plans for modifying this reference.

OPPDResponse:

OPPD agrees and Page 5 of TS 3.2, will be modified to reference the limiting condition for operationTS 2.23.

NRC Request #7 Proposed TS 2.23(2)(b) indicates that a tube should be plugged prior to entering Mode 4 following the next refueling outage or SG tube inspection. Given that Mode 4 is Cold Shutdown at your plant, this would require you to be in a refueling outage (i.e., prior to entering Mode 4). To be more consistent with the applicability section of TS 2.23 and TSTF-449, would it be more appropriate to indicate that the tube(s) should be plugged "prior to exiting Mode 4." Please discuss. If this specification is changed, the bases on 2.23-Page 4 will also need to be modified.

LIC-06-0087 Page 4 OPPD Response:

OPPD agrees and Page 5 of TS 2.23(b) including the bases on 2.23-Page 4 will be modified to indicate that the tube(s) should be plugged prior to a cold-leg temperature of 210'F. This change is also consistent with the statement in the lastparagraphin the TS 3.17 basis.

NRC Request #8 In proposed TS 2.23, it would appear that TS 2.23(2) would permit you to elect not to plug a tube provided the conditions in TS 2.23(2)(a) and TS 2.23(2)(b) were met. This is not consistent with TSTF-449. In TSTF-449, the required actions are intended to apply only in the event a tube was inadvertently identified as not being plugged rather than electing not to plug a tube. TS 2.23 should be worded such that the plugging of SG tubes that meet the repair criteria cannot be interpreted as an elective action.

Suggested wording for TS 2.23(2) that is also consistent with TS 2.1.4 would be, "Ifthe requirements of (1)(b) above are not met for one or more SG tubes, then perform the following." Please discuss your plans to clarify your technical specifications in this regard. In addition, discuss your plans to clearly indicate that "separate condition entry" is only allowed for TS 2.23(2).

OPPDResponse:

OPPD agrees and TS 2.23(2) will be reworded to: "ifthe requirements of (1)(b) above are not met for one or more SG tubes, then perform the following."

NRC Request #9 In your Basis for TS 2.23, you indicate that "large differential pressures across SG tubes can only be experienced in MODE 1, 2, or 3". It is not clear that this is a true statement for your facility since MODE 3 is when the average temperature is greater than 515 0F and MODE 4 is when the cold-leg temperature is less than 210°F. It would appear that large differential pressures could occur when the temperature is 510°F (which is before MODE 4). Please discuss whether large differential pressures could occur when the reactor coolant temperatures are "between" those defined in MODES 3 and 4. If so, discuss your plans to modify your basis.

OPPDResponse:

OPPD agrees and Bases for TS 2.23(2) will be reworded to indicate that large differential pressuresacross SG tubes can only be experienced when the Tcold is > 210°F.

NRC Request #10 TS 3.17(1) indicates that tube integrity will be verified "at a frequency defined by the Steam Generator Program." The phrase "at a frequency defined by the Steam Generator Program" is not included in TSTF-449. The frequency specified in the Steam Generator Program (TS 5.23) is a maximum inspection interval and the actual intervals may need to be less to ensure tube integrity is maintained. Please discuss your plans to remove this phrase from your proposal (to avoid the possibility that this phrase could be referring to the maximum interval specified in TS 5.23).

OPPDResponse:

OPPDagrees to truncate the sentence to delete the phrase to match the TSTF-449.

LIC-06-0087 Page 5 NRC Request #11 TS 3.17(3) and TS 3.17(4) are "not used." Please discuss why proposed TS 3.17(5) was not included as TS 3.17(3) and why TS 3.17(4) was not deleted.

OPPDResponse:

This change will be made.

NRC Request #12 In the last sentence of TS 5.23a, there appears to be a typographical error. The sentence should read,

"...tubes are inspected or plugged to confirm .....".

OPPD Response:

This change will be made.

NRC Request #13 In TS 5.23.b.3, TS 2.1.4 is referred to as "Reactor Coolant System Operational Leakage.". The actual title is "Reactor Coolant System Leakage Limits." Please discuss your plans for correcting this typographical error. The staff notes that the correct title is used in TS 2.23 - page 3 (although without the opening quote marks).

OPPDResponse:

This change will be made.

NRC Request #14 In TS 5.23.d.2, the parentheses ("]") at the end of the sentence should be deleted. Please discuss your plans to correct this apparent typographical error.

OPPD Response:

This change will be made.