ML062430402

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..

ML062430402
Person / Time
Site:	Fort Calhoun
Issue date:	08/30/2006
From:	Reinhart J A Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-06-0087, TAC MD2188
Download: ML062430402 (81)

Text

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) (ML06 15 10203)

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.

Employment with Equal Opportunity U. S. Nuclear Regulatory Commission LIC-06-0087 Page 2 Attachment 1 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.

Director Fort Calhoun Station 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 Attachment I Page 1 ATTACHMENT 1 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

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 Attachment 1 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 (CLIP). 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 Refieling 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 f?om the TS because analyses supporting the existing specification are not applicable to the MHI steam generators.

LIC-06-0087 Attachment 1 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 MH17s 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 (Preliminary Assessment of the Impact of Non- Pressure Loads 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 Attachment 1 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 staffs review of this Amendment Application:

LIC-06-0087 Attachment 1 Page 6 Plant Name, Unit No. Fort Calhoun Station Unit No.

1 Steam Generator Model(s): Mitsubishi Model:

49TT- 1 Effective Full Power Years (EFPY) of service for currently installed SGs O EFPY I Number and percentage of tubes plugged in each SG Tubing Material Number of tubes per SG 1 tube in SG RC-2B (0.019%)

Alloy 690TT 5200 1 Degradation mechanism(s) identified Number of tubes repaired in each None 0 Approved Alternate Tube Repair Criteria (ARC): Current primary-to-secondary leakage limits: Approved SG Tube Repair Methods 150 gallons per day through any one steam generator at standard temperature. Performance criteria for accident Leakage None Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40%

of the nominal tube wall thickness shall be 1.0 gallon per minute at standard temperature.

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 50.9 1 (a).

LIC-06-0087 Attachment 1 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 Attachment 1 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 staffs 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 environment. Thus, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR Part 5 1.22(~)(9), and, pursuant to 10 CFR Part 5 1.22(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 staffs 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 Attachment 1 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-0 1-0

12) 10.4 Fort Calhoun Station Updated Safety Analysis Report, Responses to 70 Criteria, Appendix G.

LIC-06-0087 Attachment 2 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)

DELETED Engineered Safety Features System Initiation lnstrumentation Settings lnstrumentation and Control Systems River Level Miscellaneous Radioactive Material Sources DELETED DELETED Steam Generator Coolant Radioactivity Post-Accident Monitoring lnstrumentation Toxic Gas Monitors Steam Generator (SG) Tube lntegrity

3.0 SURVEILLANCE

REQUIREMENTS lnstrumentation and Control Equipment and Sampling Tests Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance DELETED Containment Test Safety Injection and Containment Cooling Systems Tests Emergency Power System Periodic Tests Main Steam Isolation Valves Auxiliary Feedwater System Reactor Core Parameters DELETED Radioactive Waste Disposal System Radioactive Material Sources Surveillance DELETED DELETED Residual Heat Removal System Integrity Testing Steam Generator (SG) Tubes lntegrity 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. ? ? ,27,32,- a7 184 133 lW'l' I,--, 1 P 17C 1- 325 11I-9, 1 7 9 -"

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

5.0 ADMINISTRATIVE

CONTROLS

5.3 Facility

Staff Qualifications

5.4 Training

5.5 Not Used 5.6 Not Used 5.7 Safety Limit Violation

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) 5.1 0 Record Retention 5.1 1 Radiation Protection Program 5.12 DELETED 5.1 3 Secondary Water Chemistry 5.14 Systems Integrity 5.1 5 Post-Accident Radiological Sampling and Monitoring 5.16 Radiological Effluents and Environmental Monitoring Programs 5.16.1 Radioactive Effluent Controls Program 5.1 6.2 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 Testirrg 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. 3&3!,?2,5?, 55,57, a? ao p4 4~3 ""I ""9 ' 'I ,'-=-I 226,237 TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS . TABLES TABLE OF CONTENTS TABLE 1-1 2-1 2-2 2-3 2-4 2 . 5 2-9 2- 10 2-11 3-1 3-2 SECTION .......................................................................................... RPS LSSS ..Section 1 . 0 ............................................ ESFS Initiation Instrumentation Setting Limits Section 2.14 ................................................ Instrument Operating Requirements for RPS Section 2.1 5 Instrument Operating Requirements for Engineered Safety Features

................. Section 2.1 5 Instrument Operating Conditions for Isolation Functions

................................. .Section 2.15 Instrumentation Operating requirements for Other Safety Feature Functions

...... .Section 2.1 5 RCS Pressure Isolation Valves

..................................................................

Section 2.1 Post-Accident Monitoring. Instrumentation Operating Limits

............................. Section 2.21 Toxic Gas Monitors Operating limits

............................................................ Section 2.22 Minimum Frequencies for Checks. Calibrations. and Testing of RPS .................. Section 3.1 Minimum Frequencies for Checks. Calibrations and Testing of

........................... Section 3.1 Engineered Safety Features. lnstrumentation and Controls Minimum Frequencies for Checks. Calibrations. and Testing

............................. Section 3.1 of Miscellaneous lnstrumentation and Controls 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 (AI-179) lnstrumentation And Control Circuits Minimum Frequencies for Sampling Tests

.................................................... Section 3.2 Minimum Frequencies for Equipment Tests

.................................................. Section 3.2 Reactor Coolant Pump Surveillance

............................................................ Section 3.3 TOC . Page 4 Amendment No . ? ? 6. ? 36. ?a TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS TABLE n I Corn 2 17 ...............................................................................

VY ". 8 r 5.2-1 Minimum Shift Crew Composition

.............................................................................................

Section 5.0 TOC - Page 5 Amendment No. ??6,?25,??2,??5,?52,?76, 195 TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS (ALPHABETICAL ORDER) Continued TABLE DESCRIPTION SEC'l7ON 2-1 0 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 9 4 u .a C-e '2 4 7 ..............................................................................

W" 2-1 1 Toxic Gas Monitoring Operating Limits

...................................................................................

Section 2.22 TOC - Page 7 Amendment No.

W6+%,'! 52,?76, '!S TECHNICAL SPECIFICATION DEFINITIONS E - Averaae Disintearation Eneray 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 WarnIHigh (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 5.1

6. 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 bevond the site boundarv access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

Core Operatina 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 reactor coolant pump (RCP) seal leakoff), that is captured and conducted to collection systems or a sump or collectirlg 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 0 to the Secondary System (primary to secondary LEAKAGE), Definitions - Page 8 Amendment No. Correction Letter of 06-1 7-2004 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. 2.26 Correction Letter of 06-1 7-2004 TECHNICAL SPECIFICATION

2.0 LIMITING

CONDITIONS FOR OPERATION

2.1 Reactor

Coolant Svstem (Continued) 2.1 .I 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 30O0F.7 awlk (7) Maximum reactor coolant system hydrostatic test pressure shall be 31 25 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.

(1 1) 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 T, 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.

3*+!%@3,7?

,? ?8,%,

TECHIVICAL SPECIFICATION

2.0 LIMITING

CONDITIONS FOR OPERATION

2.1 Reactor

Coolant System (Continued)

2.1.4 Reactor

Coolant System Leakaqe Limits Applies to the leakage rates of the reactor coolant system whenever the reactor coolant temperature (Tcold) is greater than 21 0 OF. Obiective To specify limiting conditions of the reactor coolant system leakage rates. To assure safe reactor operation, the followiqg 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 orprimary 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, orprimaryto 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. Z%l-*%,?95,233 TECHNICAL SPECIFICATIONS

2.0 LIMITING

CONDITIONS FOR OPERATION

2.1 Reactor

Coolant System (Continued)

2.1.4 Reactor

Coolant System Leakage Limits (Continued)

c. ldentified 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. ldentified 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 LC0 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 7 50 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 (Tcold) 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 LC0 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 orprimary 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,?65,226 TECHNICAL SPECIFICATIONS

2.0 LIMITING

CONDITIONS FOR OPERATION 2.23 Steam Generator (SG) Tube lntearity Applies whenever the reactor coolant temperature (Tcold) is greater than 210°F. 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 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 (l)(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 if they 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.

I). 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 I gallon per minute or is assumed to increase to I 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 I9 (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 LC0 requires that SG tube integrity be maintained.

The LC0 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 lnteoritv (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 111, 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 111, 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 1 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 Tcold is > 210aF. 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 lntearitv (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 111, 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 Eaui~ment and Samplincl 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 LC0 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

'I***" 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. v 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 LC0 2.23, "Steam Generator (SG) 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. 223 TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Test - 1. Control Element Dro~ times of all full-lenath CEA's Frequency Prior to reactor criticalitv after each " Assemblies removal of the reactor iessel closure head

2. Control Element Partial movement of all CEA's Q Assemblies (Minimum of 6 in) 3. Pressurizer Safety Verify each pressurizer safety valve R Valves is OPERABLE in accordance with the Inservice Testing Program. Following testing, lift settings shall be 2485 psig -+I % and 2530 psig

&I % respectively. 4. Main Steam Safety Set Point Valves 5. DELETED

6. DELETED 7. DELETED 8a. Reactor Coolant Evaluate System Leakage***

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

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

3.2 - Page 6 ***Not applicable to primary to secondary LEAKAGE.

Reference

7.5.3 Amendment

No.

1&24+$28,?60,?65,W 171 34Q 92Q mILnV~ '-

TECHNICAL SPECIFICATIONS TABLE 3-5 (continued) MINIMUM FREQUENCIES FOR EQUIPMENT TESTS ****Verify primary to secondary LEAKAGE is I 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

3.0 SURVEILLANCE

REQUIREMENTS 3.1 7 Steam Generator (SG) Tubes lntearitv Applies to in-service surveillance of steam generator tubes.

Obiective 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).

TECHNICAL SPECIFICATIONS 3.1 7 - Page 2 Amendment No. 444+35 TECHNICAL SPECIFICATIONS

3.0 SURVEILLANCE

REQUIREMENTS 3.17 Steam Generator Tubes (Continued)

TECHNICAL SPECIFICATIONS

3.0 SURVEILLANCE

REQUIREMENTS 3.1 7 Steam Generator Tubes (Continued)

(panu!luo3) saqnl JoleJauaE)

Lueals L 1'~ SlN3R3tIlnd3tI 33NV1113AtInS O'E TECHNICAL SPECIFICATIONS

3.0 SURVEILLANCE

REQUIREMENTS 3.17 Steam Generator Tubes (Continued)

(3) Reporting Requirements A report shall be submitted within 180 days after exceeding 21 0°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.

TECHNICAL SPECIFICATIONS 2 Id- .. . u.

TECHNICAL SPECIFICATIONS

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.

I), 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. lnspection 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 (Tcold) 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 Ill, Subsection NB.

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.-

TECHNICAL SPECIFICATIONS

3.0 SURVEILLANCE

REQUIREMENTS 3.1 7 Steam Generator Tubes (Continued) 3.17 - Page 10 Amendment No. 444,?85,228 TECHNICAL SPECIFICATIONS

5.0 ADMINISTRATIVE

CONTROLS 5.23 Steam Generator (SGl Proqram 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 if the associated loads contribute significantly to burst or collapse.

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

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 LC0 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 (SGI Procrram (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. lnspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. lnspect 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 Attachment 3 Page 1 ATTACHMENT 3 Proposed Technical Specification Pages (clean)

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued)

DELETED Engineered Safety Features System Initiation lnstrumentation Settings lnstrumentation and Control Systems River Level Miscellaneous Radioactive Material Sources DELETED DELETED Steam Generator Coolant Radioactivity Post-Accident Monitoring lnstrumentation Toxic Gas Monitors Steam Generator (SG) Tube Integrity

3.0 SURVEILLANCE

REQUIREMENTS lnstrumentation and Control Equipment and Sampling Tests Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance DELETED Containment Test Safety Injection and Containment Cooling Systems Tests Emergency Power System Periodic Tests Main Steam Isolation Valves Auxiliary Feedwater System Reactor Core Parameters DELETED Radioactive Waste Disposal System Radioactive Material Sources Surveillance DELETED DELETED Residual Heat Removal System Integrity Testing Steam Generator (SG) Tube lntegrity

4.0 DESIGN

FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No.

4-? ,27,32,-

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 (P'TLR) 5.1 0 Record Retention 5.1 1 Radiation Protection Program 5.12 DELETED 5.1 3 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 5.1 6.2 Radiological Environmental Monitoring Program 5.1 7 Offsite Dose Calculation Manual (OCDM) 5.1 8 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. ZG% 42, 54, FF, 57, TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS . TABLES TABLE OF CONTENTS TABLE SECTION . 1-1 RPS LSSS ..........................................................................................

..Section 1 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. lnstrumentation and Controls 3-3 Minimum Frequencies for Checks. Calibrations and Testing

.............................

Section 3.1 of Miscellaneous lnstrumentation and Controls 3 . 3a Minimum Frequency for Checks. Calibrations. and Functional

........................... Section 3.1 Testing of Alternate Shutdown Panels (Al-185 and AI-212) and Emergency Auxiliary Feedwater Panel (AI-179) lnstrumentation 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 I 5.2-1 Minimum Shift Crew Composition

............................................................................................

Section 5.0 TOC - Page 5 Amendment No. W*1,~?5,-+42+%%2,?76,*

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

Continued TABLE DESCRIPTION SECTION 2-1 0 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 2-1 1 Toxic Gas Monitoring Operating Limits

...................................................................................

Section 2.22 TOC - Page 7 Amendment No. -, 4%

TECHNICAL SPECIFICATION DEFINITIONS E - Averaqe Disintegration Enerqv 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 makiqg 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 WarnIHigh (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 5.1

6. 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 bevond the site boundarv 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 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 to the Secondary System (primary to secondary LEAKAGE), Definitions - Page 8 Amendment No.

-,22? ,226 Correction Letter of 06-1 7-2004 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 I 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 06-1 7-2004 TECHNICAL SPECIFICATIONS

2.0 LIMITING

CONDITIONS FOR OPERATION

2.1 Reactor

Coolant System (Continued)

2.1.1 Operable

Corn~onents (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°F. (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@) 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°F. (c) If no reactor coolant pumps are operating, a non-operating reactor coolant pump shall not be started while T, 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.

3&56$6,7?

,? ? S?, W, TECHNICAL SPECIFICATIONS

2.0 LIMITING

CONDITIONS FOR OPERATION

2.1 Reactor

Coolant System (Continued)

2.1.4 Reactor

Coolant System Leakaqe 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).

I (2) If RCS operational LEAKAGE limits of (I), 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 />.

I (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.

TECHNICAL SPECIFICATIONS

2.0 LIMITING

CONDITIONS FOR OPERATION 2.1 Reactor Coolant Svstem (Continued)

2.1.4 Reactor

Coolant Svstem Leakaqe Limits (Continued)

c. ldentified 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 LC0 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 (Tcold) 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 LC0 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+%4%

TECHNICAL SPECIFICATIONS

2.0 LIMITING

CONDITIONS FOR OPERATION 2.23 Steam Generator (SG) Tube lntearity Applies whenever the reactor coolant temperature (Tcold) is greater than 210°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 (l)(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 .l, "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 lntesrity (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 if they 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 MSSVsIADVs 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 1-1 31 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 LC0 requires that SG tube integrity be maintained.

The LC0 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 Intearity (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 Ill, Subsection NB (Ref.

4) and Draft Regulatory Guide 1 .A21 (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 1 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 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 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 3.1

7. 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 lntenrity (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 Ill, 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 Equi~ment and Sam~linq 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-646) 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-806 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-806) 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 LC0 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.

4-5,67,1 B,? 38,M v TECHNICAL SPECIFICATIONS

3.0 SURVEILLANCE

REQUIREMENTS

3.2 Equipment

and Samplin~ 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 LC0 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 9.1 0

2) ASTM D4057-95(2000), ASTM D975-98b, ASTM D4176-93, ASTM D129-00, ASTM D2622-87, ASTM D287-82, ASTM 621 7-98, ASTM D2709-96

3) ASTM D975-98b, Table 1 4) Regulatory Guide 1 .I37 5) EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines." 3.2 - Page 3b Amendment No.

2%

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

1. Control Element Drop times of all full-length CEA's Prior to reactor criticality after each Assemblies removal of the reactor vessel closure head 2. Control Element Partial movement of all CEA's Q Assemblies (Minimum of 6 in) 3. Pressurizer Safety Verify each pressurizer safety valve R Valves is OPERABLE in accordance with the Inservice Testing Program. Following testing, lift settings shall be 2485 psig k1% and 2530 psig k1 % respectively. 4. Main Steam Safety Set Point Valves 5. DELETED

6. DELETED 7. DELETED 8a. Reactor Coolant Evaluate System Leakage***

USAR Section Reference 7.5.3 8b Primary to Secondary Continuous process D* Leakage **** radiation monitors or radiochemical grab sampling 9a Diesel Fuel Supply Fuel Inventory M 8.4 9b. Diesel Lubricating Oil Lube Oil lnventory lnventory 9c. Diesel Fuel Oil Test Properties Properties In accordance with the Diesel Fuel Oil Testing Program 9d. Required Diesel Air Pressure M Generator Air Start Receiver Bank Pressure

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

3.2 - Page 6 ***Not applicable to primary to secondary LEAKAGE. Amendment No.

15,2?,?28,~

171 310 3s ' ' 1,- '"1 '- I TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS ****Verify primary to secondary LEAKAGE is 5 150 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.

Amendment No TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Test - 9e. Check for and Check for Water and Remove Remove Accumulated Water from Each Fuel Oil Storage Tank 10a. Charcoal and HEPA 1. In-Place Testinq** Filters for Control Charcoal adsorbers and HEPA Room filter banks shall be leak tested and show ~99.95% Freon (R-1 1 or R-112) and cold DOP particulates removal, respectively.

Laboratory Testinq** Verify, within 31 days after removal, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows methyliodide penetration less than 0.1 75% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%. Frequency Q USAR Section Reference On a refueling frequency or every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation or after each complete or partial replacement of the charcoal adsorber1HEPA filter banks, or after any major structural maintenance on the system housing or following significant painting, fire or chemical releases in a ventilation zone communicating with the system On a refueling frequency every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation or after any structural maintenance on the HEPA filter or charcoal adsorber housing following significant painting, fire chemical release in a ventilation zone communicating with the system. **Tests shall be performed in accordance with applicable section(s) of ANSI N510-1980.

3.2 - Page 8 Amendment No. 4-624 ,?28,? 5WQ8S3 I TECHNICAL SPECIFICATIONS 10a. (continued)

TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Test - Frequencv

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 the system shall be demonstrated.

1 0b. Charcoal Adsorbers

1. In-Place Testing** for Spent Fuel Charcoal adsorbers shall be Storage Pool Area leak tested and shall show

>99% Freon (R-I I or R-112) - removal. Laboratory Testing Verify, within 31 days after removal, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows methyliodide penetration less than 10% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 95%. USAR Section Reference On a refueling frequency or every 720

6.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s

of system operation, or after 9.10 each complete or partial replacement of 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.

On a refueling frequency gr every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation or after any structural maintenance on the HEPA filter or charcoal adsorber housing following significant painting, fire chemical release in a ventilation zone communicating with the system. **Tests shall be performed in accordance with applicable section(s) of ANSI N510-1980.

3.2 - Page 9 Amendment No.

%%24,52,E8,?

54 ,lW&Q%Z%

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS lob. (continued) - 3. Overall Svstem Operation

a. Operation of each circuit Frequency Ten hours everv month USAR Section Reference 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 system shall be demon-strated. 10c. Charcoal Adsorbers

1. In-Place Testing**

On a refueling frequency or every 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 leak tested and shall show after each complete or partial

>99% Freon (R-1 1 or R-I 12) - 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 Testing 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 30°C (86°F) and a relative humidity of 95%.

3. Overall Svstem 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. 45&&2,?28,Z 59,+9%229 TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Test - Frequency 10c. (continued)

4. Automatic and/or manual initi-R ation of the system shall be demonstrated.

1 1. Containment

1. Demonstrate damper action Ventilation System Fusible Linked Dampers 2. Test a spare fusible link.

12. Diesel Generator Calibrate Under-Voltage Relays

13. Motor Operated Verify the contactor pickup value at Safety Injection - <85% of 460 V. Loop Valve Motor Starters (HCV-311, 314, 317, 320, 327, 329, 331, 333, 312, 31 5, 318,321) 1 year, 2 years, 5 years, and every 5 years thereafter.

USAR Section Reference

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 Gas Vent System valves in each vent path are in the open position. During each refueling outage just prior to plant start-up.

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 coolant vent system vent paths.

R 3.2 - Page 11 Amendment No.

4+WWiW&%,'!

55,!- I TECHNICAL SPECIFICATIONS 17. Hydrogen Purge System Test - TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS

1. Verify all manual valves are operable by completing at least one cycle.

2. Cycle each automatic valve through at 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 VA8OA and VA-806 blowers, HEPA filter, and charcoal adsorbers and verify that the system operates for at least (a) 30 minutes with suction from the 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 containment

4. Verify the pressure drop across the 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.

Frequency 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.

G&%&I#%

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 2-1 0. 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 2-1 1. 3.2 - Page 13 Amendment No. 4433,221,239 TECHNICAL SPECIFICATIONS

3.0 SURVEILLANCE

REQUIREMENTS 3.1 7 Steam Generator (SG) Tubes Integrity Applicability Applies to in-service surveillance of steam generator tubes.

Obiective 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 21 0°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.

4-04+6 TECHNICAL SPECIFICATIONS

3.0 SURVEILLANCE

REQUIREMENTS 3.1 7 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.

I), 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. lnspection 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 3.1 7(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 210°F reactor coolant temperature (T,,,,) 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 Ill, Subsection NB. 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. 444-#5 TECHNICAL SPECIFICATIONS

5.0 ADMINISTRATIVE

CONTROLS 5.23 Steam Generator (SG) Pro~ram 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 if the associated loads contribute significantly to burst or collapse.

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads. 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 LC0 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) Proclram (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 Attachment 4 Page 1 ATTACHMENT 4 Location of TSTF-449 Requirements in FCS TS LIC-06-0087 Attachment 4 Page 2 LEAKAGE Revised TS [3.4.13], RCS [Reactor Coolant System] Operational Leakage I New TS 13.4.1 81, Steam Generator Tube Integrity Revised TS [5.5.9], Steam s Generator (SG) Program Revised TS [5.6.9], Steam Revised TS 2.1.4, Reactor Coolant System Operational Leakage Limits, addresses the LC0 and ACTION revisions of TS 3.4.1 3. Revised TS 3.2 addresses the SURVEILLANCE REQUJREMENTS 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 2.23, Steam Generator (SG)

Tube Integrity, addresses the 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. New Administrative Controls TS 5.23, Steam Generator (SG)

Program Revised TS 3.17, Steam Generator Tube Integrity, Specification (5). 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 Insert B 3.4.13 B Insert B 3.4.13 C Insert B 3.4.13 D CEOG) (asis Insert B 3.4.13 E CEOG STS page B 3.4.4-2 CEOG STS page ~3.4.612 CEOG STS page B 3.4.7-3 intact steam generator.

Incorporated in the Basis for FCS TS 2.1.4. Incorporated in the Basis for FCS TS 3.2. for FCS TS 3.2. Reference 4 is incorporated in the Basis for FCS TS 3.17. Reference 5 is incorporated in the Basis for FCS TS 3.2. Not applicable because the Bases markup deletes text in the Standard 1 CEOG STS page B 3.4.5-2 TS ~kes that is not included in the FCS-TS Bases. Not applicable because the Bases markup deletes text in the Standard TS Bases that is not included in the FCS TS Bases. Not applicable because the Bases markup deletes text in the Standard TS Bases that is not included in the FCS TS Bases. 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. 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.

TS Bases that is not included in the FCS-TS Bases. Not applicable because the Bases markup deletes text in the Standard )

LIC-06-0087 Attachment 4 Page 3 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 Location of TSTF-449 Revisions in PCS TS CEOG STS page B 3.4.13-5 TSTF-449 CEOG STS page B 3.4.13-4 CEOG STS pages B 3.4.18-1 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, beginning with the SURVEILLANCE REQUIREMENTS discussion, FCS TS The revisions in the ACTION section were incorporated in FCS TS the Basis for FCS TS 3.2. Inserts B 3.4.13 C, B 3.4.13 D (CEOG) and B 3.4.13 E on CEOG STS page 3.4.13-4 are addressed above.

Incorporated in the Basis for FCS TS 2.23. Incorporated in the Basis for FCS TS 3.17 through page B 3.4.1 8-6. CEOG STS page B 3.4.18-7 The cited references are incorporated in the Bases for FCS TS 2.23 LIC-06-0087 Attachment 5 Page 1 ATTACHMENT 5 Responses to Requests for Additional Information Related to the May 30,2006 TSTF-449 Submittal LIC-06-0087 Attachment 5 Page 2 Responses to Requests for Additional Information Related to the May 30,2006 TSTF-449 Submittal NRC Request #I In Technical Specification (TS) 2.1 .I (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°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°F. 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 .I (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. OPPD Response: 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 generator tube integrity as part of the reactor coolant 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°F), TS 3.17(2) (cold-leg temperature of 210°F), and TSTF-449 (average coolant temperature of 200°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 Tcold 210°F requirement is used in other technical specifications.

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 OPPD Response:

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

NRC Request #4 In Table 3-5, the frequency is labeled as "D." Please discuss where "Dl' is defined within the specification.

Alternatively, discuss your plans for specifying that the frequency is "daily." OPPD Response:

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. OPPD Response:

The technical basis is provided in the last paragraph of TS 3.2: "The primary to secondary LEAKAGE is determined using the continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRl 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.

OPPD Response:

OPPD agrees and Page 5 of TS 3.2, will be modified to reference the limiting condition for operation TS 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 Attachment 5 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 last paragraph in 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, "If the requirements of (l)(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). OPPD Response:

OPPD agrees and TS 2.23(2) will be reworded to: "If the requirements of (l)(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°F 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.

OPPD Response:

OPPD agrees and Bases for TS 2.23(2) will be reworded to indicate that large differential pressures across SG tubes can only be experienced when the Tcold is > 210°F. NRC Request #I0 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 maximurr~

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). OPPD Response:

OPPD agrees to truncate the sentence to delete the phrase to match the TSTF-449.

LIC-06-0087 Attachment 5 Page 5 NRC Request #I I NRC Request #I2 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 #I 3 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).

OPPD Response: This change will be made.

NRC Request #I4 In TS 5.23.d.2, the parentheses

("I") 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.