Text

Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity and Steam Generator Tube Inspection Length Through the Tubesheet
	ML051520413
Person / Time
Site:	Palo Verde
Issue date:	05/26/2005
From:	Mauldin D; Arizona Public Service Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	102-05276-CDM/TNW/GAM
	Download: ML051520413 (98)
	v • d • e

Enclosure 4 of this letter contains proprietary information and should be withheld from public disclosure under 10 CFR 2.390 10 CFR 50.90 LAP1S David Mauldin Vice President Mail Station 7605 Palo Verde Nuclear Nuclear Engineering Tel: 623-393-5553 PO Box 52034 Generating Station and Support Fax: 623-393-6077 Phoenix. Arizona 85072-2034 1 02-05276-CDM/TNW/GAM May 26, 2005 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Dear Sirs:

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2 and 3 Docket Nos. STN 50-528, 50-529, and 50-530 Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity and Steam Generator Tube Inspection Length through the Tubesheet Pursuant to 10 CFR 50.90, Arizona Public Service Company (APS) hereby requests to amend Operating Licenses NPF-41, NPF-51, and NPF-74 for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3, respectively.

The proposed amendment would revise the Technical Specification (TS) requirements related to steam generator (SG) tube integrity. The proposed TSs are consistent with those in 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).

The proposed amendment also includes changes to the revised SG program in TS Section 5.5.9 to specify the SG tube inspection length through the SG tubesheet and establish plugging criteria in the inspected tubesheet region for the remaining original SGs containing Alloy 600 mill annealed (MA) tubes. This change is being proposed to establish conformance with the NRC position identified in Generic Letter (GL) 2004-01, as committed by APS in letter no. 102-05171 to NRC dated October 28, 2004. The A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

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U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity and Steam Generator Tube Inspection Length through the Tubesheet Page 2 proposed inspection length is based on the results of Westinghouse Electric Corporation analysis and testing described in report WCAP-1 6208-P, Revision 1, "NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions" (Enclosure 4). Westinghouse has determined that this report contains proprietary information and requests that it be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390(a)(4). Enclosure 3 contains an affidavit from Westinghouse stating the reasons that WCAP-16208-P should be considered a proprietary document. A redacted, non-proprietary version of this report, WCAP-16208-NP, is provided as Enclosure 5.

The proposed changes specify inspection requirements both for SGs containing Alloy 600 MA tubes and SGs containing Alloy 690 thermally treated (TT) tubes. The SGs in Units 1 and 3 containing Alloy 600 MA tubes are scheduled to be replaced with Alloy 690 TT-tube SGs in Fall 2005 (Unit 1) and Fall 2007 (Unit 3). The Unit 2 SGs were replaced with Alloy 690 TT-tube SGs in fall 2003. Therefore, the only remaining Alloy 600 MA-tube SG inspection will be the Unit 3 SG inspection during its spring 2006 refueling outage. Since this proposed amendment specifies inspection requirements for SGs of specific tube material, and not by unit number, it can be implemented at one time for all three PVNGS units. Each unit will comply with the inspection requirements for the specific SG tube material type at the time of the inspection. Following completion of SG replacement in all three units (after Fall 2007), APS intends to submit an amendment request to delete the inspection requirements for Alloy 600 MA-tube SGs. provides a notarized affidavit required by 10 CFR 50.30. Enclosure 2 provides a description of the proposed changes. Attachment 2A provides the existing PVNGS TS pages marked-up to show the proposed changes.

In accordance with the PVNGS Quality Assurance Program, the Plant Review Board and the Offsite Safety Review Committee have reviewed and concurred with this proposed amendment. By copy of this letter, this submittal is being forwarded to the Arizona Radiation Regulatory Agency (ARRA) pursuant to 1 OCFR 50.91 (b)(1).

APS requests approval of the proposed amendment by February 28, 2006, just prior to the Unit 3 refueling inspection when the last Alloy 600 MA SG tubes at PVNGS will be inspected before replacement. Also, an implementation period of 150 days is requested.

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity and Steam Generator Tube Inspection Length through the Tubesheet Page 3 No commitments are being made to the NRC by this letter. If you have any questions, please contact Thomas N. Weber at (623) 393-5764.

Sincerely, CDMITNW/GAM

Enclosures:

1. Notarized affidavit

2. APS' evaluation of the proposed changes A Proposed Technical Specification changes (mark-up) B Proposed Technical Specification changes (retyped) C Proposed Technical Specification Bases changes (mark-up)

3. Westinghouse affidavit requesting that WCAP-16208-P, Revision 1 be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390(a)(4)

4. WCAP-16208-P, Revision 1, UNDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions" (Proprietary)

5. WCAP-16208-NP, Revision 1 (Non-proprietary) cc:

B. S. Mallett NRC Region IV Regional Administrator M. B. Fields NRC NRR Project Manager G. G. Warnik NRC Senior Resident Inspector for PVNGS A. V. Godwin Arizona Radiation Regulatory Agency (ARRA)

ENCLOSURE I AFFIDAVIT STATE OF ARIZONA COUNTY OF MARICOPA

) ss.

1, David Mauldin, represent that I am Vice President, Nuclear Engineering and Support, Arizona Public Service Company (APS), that the foregoing document has been signed by me on behalf of APS with full authority to do so, and that to the best of my knowledge and belief, the statements made therein are true and correct.

I David Mauldin Sworn To Before Me This X9& YDay Of Q-mn6L

, 2005.

U Notary Pu SUSIE LYNN ERGISH Notary Pubic - Arizona I <,Maricopa County My Comm. Expires Jul 14,2007

,1 A

Notary Commission Stamp

ENCLOSURE 2 ARIZONA PUBLIC SERVICE COMPANY'S EVALUATION

Subject:

Application for Technical Specification Improvement Regarding Steam Generator (SG) Tube Integrity and SG Tube Inspection Length through the Tubesheet

1.0 INTRODUCTION

1.1

==

Introduction:==

CLIIP Changes 1.2

==

Introduction:==

SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

2.0 DESCRIPTION

OF PROPOSED AMENDMENT 2.1 Description of Proposed Amendment: CLIIP Changes 2.1.1 Variations from the CLIIP 2.2 Description of Proposed Amendment: SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

3.0 BACKGROUND

3.1

Background:

CLIIP Changes 3.2

Background:

SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes) 4.0 REGULATORY REQUIREMENTS AND GUIDANCE 4.1 Regulatory Requirements and Guidance: CLIIP Changes 4.2 Regulatory Requirements and Guidance: SG Tube Inspection Length through the Tubesheet

5.0 TECHNICAL ANALYSIS

5.1 Technical Analysis: CLIIP Changes 5.2 Technical Analysis: SG Tube Inspection Length through the Tubesheet

6.0 REGULATORY ANALYSIS

6.1 Regulatory Analysis: CLIIP Changes 6.1.1 Verification and Commitments: CLIIP Changes 6.2 Regulatory Analysis: SG Tube Inspection Length through the Tubesheet 7.0 NO SIGNIFICANT HAZARDS CONSIDERATION 7.1 No Significant Hazards Consideration: CLIIP Changes 7.2 No Significant Hazards Consideration: SG Tube Inspection Length through the Tubesheet 8.0 ENVIRONMENTAL EVALUATION 8.1 Environmental Evaluation: CLIIP Changes 8.2 Environmental Evaluation: SG Tube Inspection Length through the Tubesheet 9.0 PRECEDENT

9.1 Precedent

CLIIP Changes

9.2 Precedent

SG Tube Inspection Length through the Tubesheet

10.0 REFERENCES

10.1

References:

CLIIP Changes 10.2

References:

SG Tube Inspection Length through the Tubesheet APS' Evaluation

1.0 INTRODUCTION

This proposed license amendment is comprised of two parts:

1.

Incorporate steam generator (SG) integrity technical specification (TS) changes associated with the Consolidated Line Item Improvement Process (CLIIP) for TSTF-449, Revision 4; and

2.

Proposed changes in TS Section 5.5.9 to specify the SG tube inspection length through the SG tubesheet and establish plugging criteria in the inspected tubesheet region for the remaining original SGs containing Alloy 600 mill annealed (MA) tubes to establish conformance with the NRC position identified in Generic Letter (GL) 2004-01, as committed by APS in letter no. 102-05171 to NRC dated October 28, 2004.

These two parts of the proposed amendment are discussed in separate subsections in this Enclosure.

1.1

==

Introduction:==

CLIIP Changes This proposed license amendment revises the requirements in Technical Specifications (TS) related to steam generator tube integrity. The proposed TSs 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 (70 FR 24126) as part of the consolidated line item improvement process (CLIIP).

The proposed revised TSs are consistent with those in TSTF-449, Rev 4. However, since the current PVNGS Units 1, 2, and 3 TSs have variations from the current Combustion Engineering (CE) Improved Standard Technical Specifications (ISTS; NUREG-1432, Revision 3), which is the starting point for TSTF-449, there are some variations to the changes that are needed to incorporate TSTF-449 in the PVNGS TSs.

The variations to the changes in TSTF-449, Revision 4 are described in section 2.1.1 of this submittal.

1.2

==

Introduction:==

SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

The proposed change to TS 5.5.9 includes changes to specify the SG tube inspection length through the SG tubesheet and establish plugging criteria in the inspected tubesheet region for the remaining original SGs containing Alloy 600 mill annealed (MA) tubes.

The SGs in Units 1 and 3 containing Alloy 600 MA tubes are scheduled to be replaced with Alloy 690 TT-tube SGs in Fall 2005 (Unit 1) and Fall 2007 (Unit 3). The Unit 2 SGs APS' Evaluation were replaced with Alloy 690 TT-tube SGs in fall 2003. Therefore, the only remaining Alloy 600 MA-tube SG inspection will be the Unit 3 SG inspection during its spring 2006 refueling outage. This proposed amendment can be implemented at one time for all three PVNGS units, and each unit will comply with the inspection requirements for the specific SG-type at the time of the inspection. Following completion of SG replacement in all three units (after Fall 2007), APS intends to submit an amendment request to delete the tubesheet-region inspection and plugging criteria for SGs with Alloy 600 MA tubes.

2.0 DESCRIPTION

OF PROPOSED AMENDMENT 2.1 Description of Proposed Amendment: CLIIP Changes Consistent with the NRC-approved Revision 4 of TSTF-449, the proposed TS changes include:

Revised TS 1.1 definition of LEAKAGE

Revised TS 3.4.14, "RCS [Reactor Coolant System] Operational Leakage"

New TS 3.4.18, "Steam Generator (SG) Tube Integrity"

Revised TS 5.5.9, Steam Generator (SG) Program"

Revised TS 5.6.8, "Steam Generator Tube Inspection Report" Proposed revisions to the TS Bases are also included in this application. As discussed in the NRC's model safety evaluation, 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 (TS 5.5.14).

2.1.1 Variations from the CLIIP The proposed revised PVNGS TSs are consistent with the revised TSs in TSTF-449, Revision 4. However, since the current PVNGS Units 1, 2, and 3 TSs have differences from the latest revision (Rev. 3) of the Combustion Engineering (CE) Improved Standard Technical Specifications (ISTS, NUREG-1432), which was used as the markup TSs for TSTF-449, there are some variations to the changes that are needed to incorporate TSTF-449 in the PVNGS TSs.

The proposed changes are consistent with TSTF-449, Revision 4, except for the following variations due to differences between the current PVNGS TSs and the ISTS used as the mark-up TSs in TSTF-449, and clarifications in the TS Bases:

This proposed amendment does not include the TSTF-449 changes to the primary to secondary leakage limits in TS 3.4.14 since those changes have previously been made to the PVNGS TS. This submittal does not include the TSTF-449 changes to the Safety Analysis section in TS Bases B 3.4.14 since the current TS Bases reflects site-specific safety analysis. This submittal does APS' Evaluation include changes to the TS Bases B 3.4.14 LCO to be consistent with the Improved Standard Technical Specification Bases.

This proposed amendment will delete Action C of TS 3.4.14 and its TS Bases to be consistent with the changes being made for TSTF-449. Action C is not in the CE Improved Standard Technical Specifications, NUREG-1432, but is unique to PVNGS. TS 3.4.14, Action C, which currently requires entering LCO 3.0.3 (plant shutdown and cooldown) when one or more SGs are inoperable, will not be needed because the new TS 3.4.18, Action B will accomplish the same action (plant shutdown and cooldown) when SG tube integrity is not maintained.

This proposed amendment will revise the Notes in SR 3.4.14.1 to be consistent with the proposed Note to SR 3.4.14.2 and with the current revision of CE Improved Standard Technical Specifications, NUREG-1432. The changes to TS Bases B 3.4.14 for SRs 3.4.14.1 and 3.4.14.2 include discussions that are not in TSTF-449 but help clarify the purpose of the Notes.

The option to sleeve tubes is being deleted from TS 5.5.9 in the proposed TS.

There are no sleeves installed in the PVNGS SGs.

The TS Bases changes for the new TS B 3.4.18 Applicable Safety Analysis reflect the site specific safety analysis for SG tube rupture.

2.2 Description of Proposed Amendment: SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

The proposed change to TS 5.5.9 includes changes to specify the SG tube inspection length through the SG tubesheet and establish plugging criteria in the inspected tubesheet region for the remaining original SGs with Alloy 600 mill annealed (MA) tubes. This change is being proposed to establish conformance with the NRC position described in Generic Letter (GL) 2004-01, "Requirements for Steam Generator Tube Inspections."

The proposed inspection length in TS 5.5.9 is, "For the original SGs with Alloy 600 MA tubes, the section of tube from the tube-to-tubesheet weld to 12 inches below the bottom of the expansion transition for both the tube inlet and outlet may be excluded from inspection based on Westinghouse Report WCAP-1 6208-P, Revision 1, 'NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions."'

The proposed plugging criteria is, "For the original SGs with Alloy 600 MA tubes, all detected degradation in the portion of tube below the bottom of the expansion transition that is not excluded from inspection per the provisions of 5.5.9.d shall be removed from service."

APS' Evaluation

3.0 BACKGROUND

3.1

Background:

CLIIP Changes 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.

3.2

Background:

SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

APS' review of NRC Generic Letter (GL) 2004-01, "Requirements for Steam Generator Tube Inspections," identified the need to clarify the TS to identify the extent of the tubes to be inspected within the thickness of the tubesheet for SGs with Alloy 600 mill annealed (MA) tubes. This proposed amendment includes a change to address that clarification.

4.0 REGULATORY REQUIREMENTS AND GUIDANCE 4.1 Regulatory Requirements and Guidance: CLIIP Changes 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 revised TSs are consistent with those in TSTF-449, Rev 4. However, since the current PVNGS Units 1, 2, and 3 TSs have variations from the current Combustion Engineering (CE) Improved Standard Technical Specifications (ISTS, NUREG-1 432), which is the starting point for TSTF-449, there are some variations to the changes that are needed to incorporate TSTF-449 in the PVNGS TSs. The variations are described in section 2.1.1 of this enclosure.

4.2 Regulatory Requirements and Guidance: SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

NRC Generic Letter (GL) 2004-01, "Requirements for Steam Generator Tube Inspections," identified the NRC position regarding the need for TSs to specify the SG tube inspection length through the tubesheet for Alloy 600 MA tubes. This proposed change will establish conformance with the NRC position.

5.0 TECHNICAL ANALYSIS

5.1 Technical Analysis: CLIIP Changes APS has reviewed the safety evaluation (SE) published on March 2, 2005 (70 FR APS' Evaluation 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. APS has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to PVNGS Units 1, 2, and 3 and justify this amendment for the incorporation of the changes to the PVNGS TS.

The proposed revised TSs are consistent with those in TSTF-449, Rev 4. However, since the current PVNGS Units 1, 2, and 3 TSs have variations from the current Combustion Engineering (CE) Improved Standard Technical Specifications (ISTS, NUREG-1 432), which is the starting point for TSTF-449, there are some variations to the changes that are needed to incorporate TSTF-449 in the PVNGS TSs. The variations are described in section 2.1.1 of this enclosure.

5.2 Technical Analysis: SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

Westinghouse report WCAP-16208-P, Revision 1, UNDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions" (Enclosure 4), provides a technical basis for an inspection program within the tubesheet region of Alloy 600 MA-tube SGs in Combustion Engineering (CE) plants. Revision 0 of WCAP-1 6208-P was submitted to the NRC by Florida Power and Light Company (FPL) as a supporting enclosure to FPL letter L-2004-245, dated November 8, 2004, to amend the St. Lucie Unit 2 operating license to define the depth of the required SG tube inspections and clarify the plugging criteria in the tubesheet region. By letter dated March 31, 2005, FPL submitted responses to an NRC request for additional information that provided responses from Westinghouse answering questions regarding WCAP-1 6208-P, Revision 0. Westinghouse subsequently revised WCAP-1 6208-P, issued as Revision 1, to incorporate changes to reflect the responses to the NRC RAI submitted by FPL in their March 31, 2005 letter.

WCAP-16208-P, Revision 1 provides recommended inspection lengths in the tubesheet region based on results from a conservative test and analysis program to ensure that the requirements for tube structural and leakage integrity provided in NEI 97-06 are met.

The WCAP-16208-P, Revision 1 recommended minimum inspection lengths below the tube-to-tubesheet joint expansion transition are 10.4 inches for PVNGS Unit 1 and 11.6 inches for Unit 3 (specified as Plants CE1 and CE3). The proposed inspection length requirement "from the tube-to-tubesheet weld to 12 inches below the bottom of the expansion transition" bounds the WCAP-16208-P recommended inspection lengths for both Unit 1 and Unit 3.

The WCAP-16208-P, Revision 1 report supports the conclusion that the surrounding tubesheet prevents tube rupture and provides resistance against axial pullout loads during normal and accident conditions. In addition, any primary-to-secondary leakage from tube degradation below the inspection length is an inconsequential contribution to the total leakage assumed for a steam line break accident and may be considered APS' Evaluation negligible. Consequently, any tube degradation that may go undetected below the inspection length would not affect structural or leakage margins.

The PVNGS Steam Generator Inspection Program requires a degradation assessment to determine susceptible areas of the tubing to be inspected and the appropriate eddy current techniques to detect and quantify the degradation within each area. Input data needed for the subsequent Condition Monitoring and Operational Assessment are considered as part of the tube integrity assessment required by PVNGS Steam Generator Inspection Program and satisfies the intent mandated in NEI 97-06, "Steam Generator Program Guidelines."

Tube burst is precluded for cracks within the tubesheet by the constraint provided by the tubesheet. Thus, structural integrity is maintained by the tubesheet constraint.

However, a 360-degree circumferential crack or many axially oriented cracks could permit severing of the tube and tube pullout from the tubesheet under the axial forces on the tube from primary to secondary pressure differentials. However, Section 5 of WCAP-1 6208-P describes the testing that was performed to define the length of non-degraded tubing that is sufficient to compensate for the axial forces on the tube and thus prevent pullout. The operating conditions utilized in WCAP-1 6208-P included those at PVNGS (specified as Plants CE1 and CE3) and are summarized in Section 3 of the WCAP report.

Operating experience has demonstrated negligible normal operating leakage from primary water stress corrosion cracking (PWSCC) in expansion transitions. PWSCC in expansions in the tubesheet region would be even further leakage limited by the tight tube-to-tubesheet crevice and the limited crack opening permitted by the tubesheet constraint. The steam line break (SLB) accident conditions provide the most severe radiological hazards for postulated accidents involving loss of pressure or fluid in the secondary system. Section 3.2.3 in WCAP-1 6208-P provides the justification to neglect the total SLB leak rate contributed by cracks below the inspection length. Therefore, eddy current inspection in the area below the inspection length is not necessary to preclude normal operating or accident induced leakage.

Additional conservatism is shown in PVNGS operational assessments, which conservatively include postulated leakage in the inspection length without credit for the leak-limiting effect provided by the tubesheet.

6.0 REGULATORY ANALYSIS

6.1 Regulatory Analysis: CLIIP Changes 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.

APS' Evaluation The proposed revised TSs are consistent with those in TSTF-449, Rev 4. However, since the current PVNGS Units 1, 2, and 3 TSs have variations from the current Combustion Engineering (CE) Improved Standard Technical Specifications (ISTS, NUREG-1 432), which is the starting point for TSTF-449, there are some variations to the changes that are needed to incorporate TSTF-449 in the PVNGS TSs. The variations are described in section 2.1.1 of this enclosure.

6.1.1 Verification and Commitments: CLIIP Changes The following information is provided to support the NRC staff s review of this amendment application:

Plant/Unit PVNGS Unit I PVNGS Unit 2 PVNGS Unit 3 Steam Generator Model CE System 80 Replacement CE System 80 CE System 80 Effective full power years 14.8 1.2 14.1 (EFPY) of service for currently installed SGs Tubing material Alloy 600 MA")

Alloy 690 TT Alloy 600 MA(')

Number of tubes per SG 11,012 12,580 11,012 Number and percentage of SG 11: 873 (8%)

SG 21: 19 (0.15%)

SG 31: 631 (6%)

tubes plugged in each SG tubes plugged; tubes plugged; tubes plugged; SG 12: 1203 (11%)

SG 22:20 (0.16%)

SG 32: 755 (7%)

tubes plugged tubes plugged tubes plugged Number of tubes repaired in None None None each SG Degradation mechanism Upperbundle Support wear Upperbundle identified freespan axial freespan axial ODSCC, ID and OD ODSCC, ID and OD circumferential and circumferential and axial cracking at axial cracking at tubesheet transition tubesheet region, U-bend transistion region, PWSCC, axial U-bend PWSCC, ODSCC at axial ODSCC at eggcrates and eggcrates and support wear support wear Current primary-to secondary 150 gpd per SG 150 gpd per SG 150 gpd per SG leakage limits, evaluated at (TS 3.4.14) at room (TS 3.4.14) at (TS 3.4.14) at room APS' Evaluation Plant/Unit PVNGS Unit 1 PVNGS Unit 2 PVNGS Unit 3 what temperature condition?

temperature room temperature temperature Approved alternate tube None None None repair criteria Approved SG tube repair This proposed This proposed This proposed methods amendment will amendment will amendment will delete the delete the delete the previously-approved previously-previously-tube repair method approved tube approved tube repair method repair method Performance criteria for 0.5 gpm per SG 0.5 gpm per SG 0.5 gpm per SG accident leakage (1)

The SGs in Units I and 3 containing Alloy 600 MA tubes are scheduled to be replaced with Alloy 690 TT-tube SGs in Fall 2005 (Unit 1) and Fall 2007 (Unit 3). The only remaining Alloy 600 MA-tube SG inspection will be the Unit 3 SG inspection during its spring 2006 refueling outage.

6.2 Regulatory Analysis: SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

NRC Generic Letter (GL) 2004-01, Requirements for Steam Generator Tube Inspections," identified the NRC position regarding the need for TSs to specify the SG tube inspection length through the tubesheet for Alloy 600 MA tubes. This proposed change will establish conformance with the NRC position.

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION 7.1 No Significant Hazards Consideration: CLIIP Changes APS has reviewed the proposed no significant hazards consideration determination published on March 2, 2005 (70 FR 10298) as part of the CLIIP. APS has concluded that the proposed determination presented in the notice is applicable to PVNGS Units 1, 2, and 3 and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91 (a). APS is not proposing variations or deviations from the proposed no significant hazards consideration determination, except for the following:

The PVNGS SG tube rupture analysis conservatively assumes a leakage of 1 gpm (1440 gpd) from the unaffected SG instead of the operational leakage limit (150 gpd) that is described in the model SE No Significant Hazards Consideration.

The proposed revised PVNGS TSs are consistent with those in TSTF-449, Rev 4.

However, since the current PVNGS Units 1, 2, and 3 TSs have variations from the APS' Evaluation current Combustion Engineering (CE) Improved Standard Technical Specifications (ISTS, NUREG-1432), which is the starting point for TSTF-449, there are some variations to the changes that are needed to incorporate TSTF-449 in the PVNGS TSs.

The variations are described in section 2.1.1 of this enclosure.

7.2 No Significant Hazards Consideration: SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

The proposed change to the proposed Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," would specify the SG tube inspection length through the SG tube sheet for the Palo Verde Nuclear Generating Station (PVNGS) Alloy 600 mill annealed (MA) tube SGs.

1.

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

Response: No The analysis that established the inspection length through the SG tube sheet for the PVNGS Alloy 600 MA-tube SGs took into account the reinforcing effect the tubesheet has on the external surface of an expanded SG tube. Tube-bundle integrity will not be adversely affected by the implementation of the revised tube inspection scope. SG tube burst or collapse cannot occur within the confines of the tubesheet; therefore, the tube burst and collapse criteria of draft Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," are inherently met. Any degradation below the inspection length is shown by analyses and test results to be acceptable, thereby precluding an event with consequences similar to a postulated tube rupture event.

Tube burst is precluded for cracks within the tubesheet by the constraint provided by the tubesheet. Thus, structural integrity is maintained by the tubesheet constraint.

However, a 360-degree circumferential crack or many axially oriented cracks could permit severing of the tube and tube pullout from the tubesheet under the axial forces on the tube from primary to secondary pressure differentials. Analysis and testing was performed to define the length of non-degraded tubing that is sufficient to compensate for the axial forces on the tube and thus prevent pullout. That length is bounded by the inspection length proposed in this change.

In conclusion, incorporation of the revised inspection scope into PVNGS TS maintains existing design limits and therefore, the proposed change does not involve a significant increase in the probability or consequences of an 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 APS' Evaluation The proposed performance based requirements are an improvement over the requirements imposed by the current TS.

Implementation of the proposed Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The result of the implementation of the Steam Generator Program will be an enhancement of SG tube performance. Primary to secondary leakage that may be experienced during all plant conditions will be monitored to ensure it remains within current accident analysis assumptions.

The proposed change does not affect the design of the SGs, their method of operation, or primary or secondary coolant chemistry controls. In addition, the proposed change does not impact any other plant system or component. The change enhances SG inspection requirements.

Tube-bundle integrity is expected to be maintained during all plant conditions upon implementation of the proposed tube inspection scope. Use of this scope does not introduce a new mechanism that would result in a different kind of accident from those previously analyzed. Even with the limiting circumstances of a complete circumferential separation of a tube occurring below the inspection length into the tubesheet, SG tube pullout is precluded and leakage is predicted to be maintained within the Updated Final Safety Analysis Report limits during all plant conditions.

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

3.

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

Response: No The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change is expected to result in an improvement in the tube integrity by implementing the Steam Generator Program to manage SG tube inspection, assessment, repair, and plugging. The requirements established by the Steam Generator Program are consistent with those in the applicable design codes and standards and are an improvement over the requirements in the current TS.

- 1 1 -

APS' Evaluation Upon implementation of the revised inspection scope, operation with potential cracking below the Inspection Extent length in the expansion region of the SG tubing will meet the margin of safety as defined by Regulatory Guide (RG) 1.83 (Ref. 17), draft RG 1.121 (Ref. 11), and the requirements of General Design Criteria 14, 15, 31, and 32 of Appendix A to 10 CFR 50.

Therefore, the proposed change does not involve a significant reduction in a margin of safety and overall plant safety will be enhanced by the proposed changes.

8.0 ENVIRONMENTAL EVALUATION 8.1 Environmental Evaluation: CLIIP Changes APS has reviewed the environmental evaluation included in the model SE published on March 2, 2005 (70 FR 10298) as part of the CLIIP. APS has concluded that the staffs findings presented in that evaluation are applicable to PVNGS Units 1, 2, and 3 and the evaluation is hereby incorporated by reference for this application.

The proposed revised TSs are consistent with those in TSTF-449, Rev 4. However, since the current PVNGS Units 1, 2, and 3 TSs have variations from the current Combustion Engineering (CE) Improved Standard Technical Specifications (ISTS, NUREG-1 432), which is the starting point for TSTF-449, there are some variations to the changes that are needed to incorporate TSTF-449 in the PVNGS TSs.

8.2 Environmental Evaluation: SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

9.0 PRECEDENT

9.1 Precedent

CLIIP Changes This application is being made in accordance with the CLIIP. APS is not proposing variations or deviations from the TS changes described in TSTF-449, Revision 4. APS APS' Evaluation is not proposing variations or deviations from the NRC staffs model SE published on March 2, 2005 (70 FR 10298), except for the following:

The PVNGS SG tube rupture analysis conservatively assumes a leakage of 1 gpm (1440 gpd) from the unaffected SG instead of the operational leakage limit (150 gpd) that is described in the No Significant Hazards Consideration discussion in the model SE.

The proposed revised TSs are consistent with those in TSTF-449, Rev 4. However, since the current PVNGS Units 1, 2, and 3 TSs have variations from the current Combustion Engineering (CE) Improved Standard Technical Specifications (ISTS, NUREG-1432), which is the starting point for TSTF-449, there are some variations to the changes that are needed to incorporate TSTF-449 in the PVNGS TSs.

9.2 Precedent

SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

The proposed change to the SG Program in TS 5.5.9 includes changes to specify the SG tube inspection length through the SG tubesheet and establish plugging criteria in the inspected tubesheet region for the remaining original SGs with Alloy 600 MA tubes.

This change is being proposed to establish conformance with the NRC position described in Generic Letter (GL) 2004-01, "Requirements for Steam Generator Tube Inspections. "

This proposed change is similar to a proposed amendment submitted to the NRC by Florida Power and Light Company (FPL) in FPL letter L-2004-245, dated November 8, 2004, to amend the St. Lucie Unit 2 operating license to define the depth of the required SG tube inspections and clarify the plugging criteria in the tubesheet region. This FPL amendment request included WCAP-16208-P, Revision 0 as a supporting enclosure.

By letter dated March 31, 2005, FPL submitted responses to an NRC request for additional information that provided responses from Westinghouse answering questions regarding WCAP-1 6208-P, Revision 0. Westinghouse subsequently revised WCAP-16208-P, issued as Revision 1, to incorporate changes to reflect the responses to the NRC RAI submitted by FPL in their March 31, 2005 letter.

10.0 REFERENCES

10.1

References:

CLIIP Changes Federal Register Notices:

Notice for Comment published on March 2, 2005 (70 CFR 10298)

Notice of Availability published on May 6, 2005 (70 FR 24126)

APS' Evaluation 10.2

References:

SG Tube Inspection Length through the Tubesheet (Alloy 600 Tubes)

1.

NRC Generic Letter 2004-01, "Requirements for Steam Generator Tube Inspections."

2.

Westinghouse report WCAP-16208-P, Revision 1, "NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions," dated May 2005. A Proposed Technical Specification Changes (mark-up)

Pages:

ii 1.1-4 and 5 3.4.14-1 and 2 3.4.18-1 and 2 (new) 5.5-6 through 5.5-24 5.6-6

PALO VERDE NUCLEAR GENERATING STATION IMPROVED TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 3.3 INSTRUMENTATION 3.3.1 Reactor Protective System (RPS) Instrumentation -- Operating 3.3.2 RPS Instrumentation -- Shutdown 3.3.3 Control Element Assembly Calculators (CEACs) 3.3.4 RPS Logic and Trip Initiation 3.3.5 Engineered Safety Features Actuation System (ESFAS) Instrumentation 3.3.6 ESFAS Logic and Manual Trip 3.3.7 Diesel Generator (DG) -- Loss of Voltage Start (LOVS) 3.3.8 Containment Purge Isolation Actuation Signal (CPIAS) 3.3.9 Control Room Essential Filtration Actuation Signal (CREFAS) 3.3.10 Post Accident Monitoring (PAM) Instrumentation 3.3.11 Remote Shutdown System 3.3.12 Boron Dilution Alarm System (BDAS) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature and Flow Departure from Nucleate Boiling (DNB) Limits 3.4.2 RCS Minimum Temperature for Criticality 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4.4 RCS Loops --

MODES 1 and 2 3.4.5 RCS Loops --

MODE 3 3.4.6 RCS Loops --

MODE 4 3.4.7 RCS Loops --

MODE 5. Loops Filled 3.4.8 RCS Loops --

MODE 5. Loops Not Filled 3.4.9 Pressurizer 3.4.10 Pressurizer Safety Valves -- MODES 1. 2. and 3 3.4.11 Pressurizer Safety Valves --

MODE 4 3.4.12 Pressurizer Vents 3.4.13 Low Temperature Overpressure Protection (LTOP) System 3.4.14 RCS Operational Leakage 3.4.15 RCS Pressure Isolation Valve (PIV) Leakage AM 83.4.16 RCS Leakage Detection Instrumentation 3.5

~) [ ~sf ad f~;i-ner..+or-(5G;) Thbe Jntore+y 3.5 EMERGENCY CORE COOLING SYSTEMS CECCS) 3.5.1 3.5.2 3.5.3 3.5.4 3.5.5 3.5.6 Safety Injection Tanks (SITs) -- Operating SITs -- Shutdown ECCS -- Operating ECCS -- Shutdown Refueling Water Tank (RWT)

Trisodium Phosphate (TSP)

PALO VERDE UNITS 1.2.3 ii AMENDMENT NO. 117

Definitions 1.1 1.1 Definitions (continued)

ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.. the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays. where applicable. The response time may be measured by means of any series of sequential.

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

Kn. 1is the K effective calculated the actual CEA configuration and fully or partially inserted full highest worth is fully withdrawn.

by considering assuming that the strength CEA of I

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or 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 (SG) to the Secondgrv Syst (e1_

f t

~

r'=i)

(continued)

PALO VERDE UNITS 1.2.3 1.1-4 AMENDMENT NO. 43-5. 152

Definitions 1.1 1.1 Definitions LEAKAGE (continued)

b. Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE:

c. Pressure Boundary LEAKAGE LEAKAGE (except through a nonisolable fault in an RCS component body.

pipe wall, or vessel wall.

A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, cold leg reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

MODE NEUTRON RATED THERMAL POWER (NRTP)

OPERABLE - OPERABILITY PHYSICS TESTS The indicated neutron flux at RTP.

A system, subsystem, train, component. or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation. controls, normal or emergency electrical power, cooling and seal water.

lubrication, and other auxiliary equipment that are required for the system. subsystem. train, component. or device to perform its specified safety function(s) are also capable of performing their related support function(s).

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

These tests are:

a. Described in Chapter 14. Initial Test Program of the UFSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

(continued)

PALO VERDE UNITS 1,2.3 1.1-5 AMENDMENT NO.449, 135

RCS Operational LEAKAGE 3.4.14 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 RCS Operational LEAKAGE LCO 3.4.14 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE:

b. 1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE: and

d.

150 gallons per day primary to secondary LEAKAGE through any one n

r MODES 1. 2. 3. and 4.

APPLICABILITY:

2;1+.. se, sw Prim 6

r c. o.av1 Le4KA-C A04 W~41%in PALO VERDE UNITS 1,2.3 3.4.14-1 AMENDMENT NO. 4-14 120 l

RCS Operational LEAKAGE 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I

Qt-944e.-_ ee,4r,6k~I'ine..d SR 3.4.14.1

I-NOTE

(_9 Not required td be performed 4

until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state op&raThion er-formo sendtry Lance J(

Perform RCS water inventory balance.

N Only equired t b erformed uring ste y state o ation 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s~~'~

Ver G tube ytegrity i in accor nce>

lwith he Ste Ge nerator ube Suryefilance)

Vtm.,8-4 seco,'fdy LC-kb46C

\\

\\5 ~

lsO tAI1Oac (VCdkf tI(7ktI y

a ce I

=\\

Iw1i)the SmX

)4nerator/Tube fSurve~ance y e2oq

<~~

&)04

.11;e bI perd-rm^

Q.+;i 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months e Aiv e4 4-

-b4 r

orfe-vr 5rr dn cP~ _.4L l-'t 1~*..-b reoii PALO VERDE UNITS 1.2.3 3.4.14-2 AMENDMENT NO. 117

INSERT NEW TS 3.4.18 SG Tube Integrity 3.4.18 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.18 Steam Generator (SG) Tube Integrity LCO 3.4.18 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY:

MODES 1. 2, 3. and 4.

ACTIONS

NOTE -------------------------------------

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity 7 days satisfying the tube of the affected repair criteria and tube(s) is maintained not plugged in until the next accordance with the refueling outage or Steam Generator SG tube inspection.

Program.

AND A.2 Plug the affected Prior to tube(s) in accordance entering MODE 4 with the Steam following the Generator Program.

next refueling outage or SG tube inspection.

B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A AND not met.

OR B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SG tube integrity not maintained.

PALO VERDE UNITS 1,2,3 3.4.18-1 AMENDMENT NO.

INSERT NEW TS 3.4.18 SG Tube Integrity 3.4.18 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.18.1 Verify SG tube integrity in accordance with In accordance the Steam Generator Program.

with the Steam Generator Program.

SR 3.4.18.2 Verify that each inspected SG tube that Prior to satisfies the tube repair criteria is entering MODE 4 plugged in accordance with the Steam following a SG Generator Program.

tube inspection.

PALO VERDE UNITS 1,2,3 3.4.18-2 AMENDMENT NO.

Programs and Manuals 5.5 Pep(qe w;4-A lr)'k r 5-j 9 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (

Tube Surveillance Pro ram This program ovides controls for the Ins vice Inspection of steam genertor tubes and tube sleeves tensure that structural integrity f this portion of the RCS imaintained. The program shall i lude the following:

5.5.9.1 St Generator Sam le Selection nd Ins ection - Each steam era or sha be e

eermine uring shutdown by electing and inspecting tub and tube sleeves in at least the

/minimum number of steam ge rators specified in Table 5.5.9-1.

5.5.9 a Steam Generator Tube S: le Selection and Inspection - The steam generator tube minimu sample size, inspection result classification, and he corresponding action required shall b as specified in Tabl 5.5.9-2. The inservice inspection of s m generator tubes hall be performed at the frequencies spcified in 5.5.9.3 and t inspected tubes shall be verified acc able per the accepta e criteria of 5.5.9.4.

The tubes sele ed for each inservice nspection shall include at least 3% of he total number of tuber n all steam generators: the tubes se cted for these inspec ons shall be selected on a random b s except:

a. Where experience in similar plants ith similar water chemistry indicates critical are to be inspected, then at least 50% of the tubes inspec shall be from these critical areas.

b. The first sample of tu selected for each inservice inspection (subseque to the preservice inspection) of each steam generator sh include:

1. All nonpX gged tubes that previously had detectable w penetr ions (greater than 20% and not sleeved in t

are

2.

ubes in those areas where experience has i cated potential problems.

3. A tube inspection (pursuant to Speci Xation 5.5.9.4.a.10.) shall be performed each selected tube.

If any selected tube does not p it the passage of the eddy current probe for a tube nspection. this shall be recorded and an adjacent tp e shall be selected and subjected to a tube ins tion.

c. The tubes selected as t second and third samples (if required by Table 5.

. -2) during each inservice inspection may be subjected a partial tube inspection provided:

(continued)

PALO VERDE UNITS 1.2.3 5.5-6 AMENDMENT NO. 4-4; 120

INSERT 5.5.9 (sheet I of 3) 5.5.9 Steam Generator (SG) Pro-gram A Steam Generator (SG) 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 and 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 SG 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 0.5 gpm per SG and 1 gpm through both SGs.

INSERT 5.5.9 continued (sheet 2 of 3)

3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.14, URCS Operational LEAKAGE."

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. For the original SGs with Alloy 600 MA tubes, all detected degradation in the portion of tube below the bottom of the expansion transition that is not excluded from inspection per the provisions of 5.5.9.d shall be removed from service.

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. For the original SGs with Alloy 600 MA tubes, the section of tube from the tube-to-tubesheet weld to 12 inches below the bottom of the expansion transition for both the tube'inlet'and outlet may be excluded from inspection based on Westinghouse Report WCAP-1 6208-P, Revision 1, "NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions." In addition to meeting the requirements of d.1, d.2a, d.2b, 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.

2a.

Original SGs with Alloy 600 MA tubes: Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.

2b.

Replacement SGs with Alloy 690 TT tubes: 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

INSERT 5.5.9 continued (sheet 3 of 3) 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.

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9.2a Steam Generator Tube Sple Selection and Inspection (continued)

1. The tube Aelected for these samples include the tubes from tnspecareas of the tube sheet array where tubes with imperfections were previously found.

2. i nspections include those portions of the tubes

/hre imperfections were previously found./

The r ults of each inspection sample shall be classified Xto one

\\of e following three categories (this classification all apply the inspection of tubes and treated exclusive of e sleeve

/nspections in 5.5.9.2b):

Category Inspection Re ts C-1 Less than 5% of the a tubes inspected are degraded tubes and one of the inspected tubes are defect e.

C-2 One or more es. but not more than 1% of the total es inspected are defective. or between and 10% of the total tubes inspecf d are degraded tubes.

C-3 Mo than 10% of the total tubes inspecte e degraded tubes or more than 1% of t inspected tubes are defective.

\\/

NOTE-------,3----------

In all inspections, previously d raded tubes must exhibit significant (gre r than 10%)

further wall penetrations tg e included in the above percentage calculat pns.

(continued)

PALO VERDE UNITS 1.2,3 5.5-7 AMENDMENT NO. 4-14 120

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9.2b Steam Generator Tube Sleve Sample Selection and Inspectiop - The steam generator tu b eeve minimum sample size. inspec n result classification, and e corresponding action required hall be as specified in Tabl

.5.9-3. The inservice inspecti of steam generator tube eeves shall be performed at the ahequencies specified in 5.9.3 and the inspected tubes sh1 be verified acceptable p the acceptance criteria of 5.

..4. The tube sleeves se cted for each inservice ins ec on shall include at least 20 of the total number of tube s 1 yes in all steam generatrs; the tube sleeves selected r these inspections shall be s ected on a random basis excep Where experience in simil plants with similar water chemistry indicates cr'ical areas to be inspected, then at least 50% of the tub sleeves inspected shall be from these critical areas. W re the number of sleeves in the critical areas represents ess than 50% of the initial sample, all sleeves in the ritical areas shall be inspected.

b.

The first ample of tube sleeves selected for each inser *ce inspect'n (subsequent to the preservice inspection) o each steam enerator shall include:

1 All tube sleeves that previously had dete able wall penetrations (greater than 20%).

2.

Tube sleeves in those areas where erience has indicated potential problems.

3. A tube sleeve inspection (pu uant to Specification 5.5.9.4.a.8.) shall be per rmed on each selected tube sleeve. If any selecte ube sleeve does not permit the passage of the ed current probe for a tube sleeve inspection. this shut be recorded and an adjacent tube sleeve shall be s ected and subjected to a tube sleeve inspection.

The results of each in ction sample shall be classified into one of the following thr categories (this classification shall apply to the inspection sleeves and treated exclusive of the tube inspections in 5.9.2a):

(continued)

PALO VERDE UNITS 1.2.3 5.5-8 AMENDMENT NO. 147-120

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) theototta unbede Slesions ed ar decti ve Cateipt Inspection Results levs C

MoLess than 5% of the total tube sleeves i nsected are degraded tube sleeves and none oftha 1

inspected tube sleeves are defective 2

/

C-2 One or more tube sleeves. but more than 1t of 1

the total tube sleeves ins are defectiv t

or between 5t and c 0% of t total tube sleev 5

9 inspected are degraded isve sleeves.

co

/

C-3 More than tub sleevestotal tube sleeves inspect ed

a. Tare degridecdtubion eeves or more than 1% of thes s

inspected tu er eeves are defective.

/

NOTE -------------------

In all inspe tfns, previously degraded tube sleeves mu nhibit significant (greater than isr

) furvicer wal penetrations to be included in o

the abX percentage calculations.{

\\p5.5.9.3 Inspection nc

.es

- The above required inservice inspections

\\

~of steam g4nerator tubes and tube sleeves shall be performed at tha

\\

followl T requencies:/

a. The first inservice inspection of the steam generator t XSl l

/shall be performed after 6 Effective Full Powe r Month. 'ut (cotined within 24 calendar months of initial criticality sequent

\\/inservice inspections shall be performed at itgas of not

\\ /less than 12 nor more than 24 calendar months Afeer the

\\ /previous inspection. If two consecutive ins ifitons following l

\\service under AVT conditions, not includin 7'he preservice

\\inspection, result in all inspecto regt falling into the

\\

~~C-1 category or if two consecutiveinecos mntre

\\

that previously observed degradatio,,thas not continued and no

\\

~additional degradation has occursr, the inspection interval

\\ my b exened o amaimum once per 40 months.

(continued)

PALO VERDE UNITS 1,2,3 5.5-9 AMENDMENT NO. 117; 120 l

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) i 5.5.ivu Inspectlon irequenc es (co a

nued)

b. The inservice in s

pton of steam generator tube sleeves sh

\\be performed te following frequencies:/

1 Ste nerator tube sleeves shall be insa e w i r t o

\\itial operation. The operating period bdrethe T

5 faial inservice inspection shall r than I/

six months nor longer than 24 months, e inspections ofI tube sleeves shall be configured to nsure that each individual tube sleeve is inspected at least once in 60 months. T n

2. If the results of the in s

ce inspection of steam

\\ /generator tube sleevesnnducted in accordance with

/

Table 5.5.9-3 fall intocategory C-3. the inspection frequency shall r

ncreased to ensure that each remaining tube eve is inspected at least once in 30 months. Thincrease in inspection frequency shall apply until e subsequent inspection satisfies the\\

the bcriteria to ar uategory C-1. o

c. If the re s of the inservice inspection of a steam p

generat conducted in accordance with Tables 5.5.9-2 and 5.5.t at 40 month intervals fall into Category C-32 the in ction fretuency for the applicable tube or sleeve f i Pection shal be increased to at least once per 20 montl

\\

Lathe increase in inspection frequency shall apply until tbe'

/

sbsequent inspections satisfy the criteria of Specifigdtion l

/ 55.9.3.a (tubes) or 5.5.9.3.b.3 (sleeves); the in -rval may

/

ten be extended to a maximum of once per 40 monjt (tubes) or

/

30 months (sleeves)./

d. Additional, unscheduled inservice inpe ns shall be

\\

erformed on each steam generator in acr ance with the

\\

irst sample inspection specified i iales 5.5.9-2 andI 5.5.9-3 during the shutdown subse W n o any of the following l

t

~conditions:/

\\1.

Primary-to-secondar X es leaks (not including leaks

\\originating from tue-to-tube sheet welds) in excess ofI

\\the limits of w ification 3.4.14./

\\2.

A seismic /

rence greater than the Operating Basis/

(continued)

PALO VERDE UNITS 1.2.3 5.5-10 AMENDMENT NO. 44-7; 120

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9.3 Inspection FreQuenciRi (continued)

3. A lo -of-coolant accident requiring actuation of e

en neered safeguards.

4. A main steam line or feedwater line break 5.5.9.4 Acce ance Criteria As used in this Specification

1. Defect means an imperfect n of such severity that it

\\exceeds the plugging o epair limit. A tube or sleeve containing a defect defective.

2.

Degradation me a service-induced cracking. wastage.

wear, or gen Ial corrosion occurring on either inside or outsid f a tube.

3.

De dation means the percentage of the tube wall ness a ected or removed by degradation.

4.

Degraded Tube or Sleeve means a tube or sleeve containing imperfections greater than or equal 20%

of the nominal wall thickness caused by degr ation.

5. Imperfection means an exception to the d ensions,

/inish, or contour of a tu e from that equired by fabrication drawings or specificati Eddy-current testing indications below 20% of e nominal tube wall thickness, if detectable. may b considered as l

~imperfections.//

6. Plugging or Repair Limit eans the imperfection depth at or beyond whichthe ube shall be removed from service by plugging repaired by sleeving in the affected area. T plugging or repair imperfection depths specifie elow are in percentage of nominal wall thicknes

a. Orig a tube wall 40%

b. A E leak tight sleeve wall 35%

(continued)

PALO VERDE UNITS 1.2.3

5. 5-11 AMENDMENT NO. 44-7 120 I

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

.5.9.4 Acceptance Criteriaiao(con n)

7. Preservice Ins Acion in the context of new steam l

generators ans an inspection of the full length of each tbezn each steam generator performed by eddy

\\

current echniqyes prior to service to establish a /

basele condition of the tubing. This inspection as per rmed prior to the field hydrostatic test an prior t initial POWER OPERATION using the equipment nd echniques expected to be used during subse nt inservice inspections.

Preservice Inspection for steam gener or tubes repaired by tube sleeving means an spection of the full length of the pressure boun y portion of the sleeved area performed by eddy urrent techniques prior to service to establish a ba me condition of the sleeved area. The sleeved rea includes the pressure retaining portions of t parent tube in contact with the sleeve, the sleev 0-tube weld and the pressure retaining portion of he sleeve.

8. Sleeve Inspecti for sleeves selected in accordance wit tab le

.9-3 means an inspection of the sleeved area. incl u ng the pressure retaining portions of the parent tu in contact with the sleeve, the sleeve-to-tube we and the pressure retaining portion of the sleev A.

9. T e or Tubing means that portion of the tub hat orms he primary system to secondary syst pressure

\\

/ b~oundary.

0. Tube Inspection for tubes selected accordance with Table 5.5.9-2 means an inspectio of the steam I/generator tube from the pointut entry (hot leg side)

Icompletely around the U-benio the top support of the cold leg. excludin sl reas.

(continued)

PALO VERDE UNITS 1.2.3 5.5-12 AMENDMENT NO. 44; 120

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9.4 Acceptance Criteriav continued)

\\

11.

Tube Rz `air or Sleeve refers to welded sleeving,.

\\~T dec ed in Combu-stion Engineering. Inc. (CE or

\\

_BSCE) report CEN-630-P. "Repair of 3/!4' O.D, team \\

Gnerator Tubes Using Leak Tight Sleeves." Bvision 02.

June 1997. which is used to maintain a tu in service or to return a previously plugged tube service.

Returning a previously plugged tube service includes the removal of the tube p gs that were installed as a preventive or cor ective measure and performing a tube inspection the tube in accordance with Technical Specificatio

.5.9.4.a.8.

12.

Unserviceable describe he condition of a tube if it leaks or contains a fect large enough to affect its structural integri in the event of an Operating Basis Earthquake, a I s-of-coolant accident. or a steam line or feedwater

ne break as specified in 5.5.9.3.d..

\\

~~above

//\\

b. The steam g Xrator shall be determined OPERABLE after completin the corresponding actions required by Ta s

5.5.9-nd 5.5.9-3. including the following:

1.

Plug or repair all defective tubes and tubes containing through-wall cracks.

2. Plug all tubes containing a (e

sleeves and all ubs otanigany sleeves trugh-wall cracks.

I (continued)

PALO VERDE UNITS 1.2.3 5.5-13 AMENDMENT NO. 4441 120

TABLE5..1 MINIMUM NUMBER OF STEAM GENERA TO BE INSPECTED DURING INSERVICE INSPE ION Preservice Inspection No Yes No. of Steam Generators per Unit Two Two First Inservice Inspection All

\\

ne S and & Subsequent Inservice Inspection One*

OneX TABLE N ATION

The inservi e inspection may be limited to one steam gene ator on a rotating schedule encompassing 3 N of the tubes (wh e N is the number of steam generators in the lant) if the results of the first or previous inspections in cate that all steam generators are performing in a like manner.

Note that under some circumstances. t operating conditions in one or more steam g erators may be found to be more severe than those in other st m generators.

Under such circumstances the suple sequence shall be modified to inspect the most severe con itions.

PALO VERDE UNITS 1.2.3 5.5-14 AMENDMENT NO. 120 PALO VERDE UNITS 1.2.3 5.5-14 AMENDMENT NO. AT;- 120 l

A mimnirn of C-oeN.A.

N.A.N...A S Tubes per S.G c-1 None N.A.____N.A.

C-2 Plug or repair defec tubes and inspect C-2 Plug or repair Nn additional 2S tubes in defective tubes and tiSG.inspect additional C-2 Plug or repair 4$ tubes in this defective tubes C-3 rfmaction C-3 Perform action N.A.

for C-3 result of C-3 inspect all tubes in All other S.G.s\\

None N.A.

NA this S.G.. plug or repair are c-I defective tubes and inspect 2S tubes in each other S.G.

Some S.G.s C-2 Pe rm action N.A.

N.A.

but no for -2 result of Notification to NRC additional S.G.

secon sample pursuant to 10 CFR 50.72 are C-3 Additional S.G.

Inspect a tubes N.A.

N.A.

\\

4 tube in tis

\\defective tube

\\

S.G.C-3Notificaactio

\\~~~R

\\pursuanrsulto i

C-3 in PeachS.

a nd n

ne P VE3 UinS pec.3 5u1r5i AE NDMENT N..

\\

1 PALO VERDE UNITS 1.2.3 5.5-15 AMENDMENT NO. -46 120 I

PALO VERDE UNITS 1,2,3 5.5-16 AMENDMENT NO. 120

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include:

a.

Identification of a sampling schedule for the critical variables and control points for these variables:

b.

Identification of the procedures used to measure the values of the critical variables:

c.

Identification of process sampling points which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage:

d.

Procedures for the recording and management of data:

e.

Procedures defining corrective actions for all off control point chemistry conditions: and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5:5.11 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52.

Revision 2. and in accordance with Regulatory Guide 1.52, Revision 2 and ANSI N510-1980 at the system flowrate specified below +/- 10%.

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1.0 Z when tested in accordance with Regulatory Guide 1.52, Revision 2. and ANSI N510-1980, at the system flowrate specified as follows

+ 10X:

(continued)

PALO VERDE UNITS 1.2,3 5.5-AMENDMENT NO. 14 120

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 Ventilation Filter Testing Program (VFTP) (continued)

ESF Ventilation System Flowrate Control Room Essential Filtration System (CREFS)

Engineered Safety Feature (ESF)

Pump Room Exhaust Air Cleanup System (PREACS) 28.600 CFM 6.000 CFM

b.

Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass

1.0 % when tested in accordance with Regulatory Guide 1.52. Revision 2, and ANSI N510-1980 at the system flowrate specified as follows +/- 10%:

ESF Ventilation System Flowrate CREFS 28.600 CFM ESF PREACS 6.000 CFM

c. Demonstrate for each of the ESF systems that a charcoal adsorber sample. when obtained in accordance with the application of Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 2. March 1978, as described in Section 1.8 of the UFSAR. shows the methyl iodide penetration less than or equal to the value specified below, when tested in accordance with ASTM D3803-1989. at a temperature of 300C and to the relative humidity specified as follows:

ESF Ventilation System CREFS Penetration

2.5%

< 2.5%

RH 70%

ESF PREACS 70%

(continued) 5.5 @

PALO VERDE UNITS 1.2,3 AMENDMENT NO. 4 130

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 Ventilation Filter Testing Program (VFTP)

(continued)

d.

For each of the ESF systems. demonstrate the pressure drop across the combined HEPA filters. the prefilters. and the charcoal adsorbers is less than or equal to the value specified below when tested in accordance with Regulatory Guide 1.52. Revision 2. and ANSI N510-1980 at the system flowrate specified as follows +/- 10%:

ESF Ventilation System Delta P Flowrate CREFS 4.8 inches water gauge 28.600 CFM ESF PREACS 5.2 inches water gauge 6.000 CFM

e.

Demonstrate that the heaters for each of the ESF systems dissipate the following specified value when tested in accordance with ANSI N510-1980:

ESF Ventilation System Wattage ESF PREACS

> 19 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides control for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure". The liquid radwaste quantities shall be determined in accordance with Standard Review Plan. Section 15.7.3.

"Postulated Radioactive Release due to Tank Failures".

(continued)

PALO VERDE UNITS 1.2.3 5.5 AMENDMENT NO. 42in 142

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e.. whether or not the system is designed to withstand a hydrogen explosion):

b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of Ž 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents:

and

c. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners. dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR Part 20, Appendix B. Table 2. Column 2. at the nearest potable water supply and the nearest surface water supply in an unrestricted area. in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.13 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements. and acceptance criteria, all in accordance with applicable ASTM Standards as referenced in the UFSAR.

The purpose of the program is to establish the following:

(continued)

PALO VERDE UNITS 1.2.3 5.5 R AMENDMENT NO. 447; 120

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.13 Diesel Fuel Oil Testing Program (continued)

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific gravity within limits.

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil. and

3. Water and sediment within limits when tested in accordance with ASTM D1796:

b. Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks: and

c. Total particulate concentration of the stored fuel oil is

< 10 mg/l when tested every 92 days in accordance with ASTM D-2276. Method A-2 or A-3.

5.5.14 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license: or

2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

(continued)

PALO VERDE UNITS 1.2.3 5.50 AMENDMENT NO. 4,29. 137

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.14 Technical Specifications (TS) Bases Control Program (continued)

d. Proposed changes that meet the criteria of Specification 5.5.14b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.15 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.

The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected:

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists:

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities: and

d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed.

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to system(s) supported by the inoperable support system is also inoperable: or (continued)

PALO VERDE UNITS 1,2.3 5.5 AMENDMENT NO -.20, 137

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.15 Safety Function Determination Program (continued)

b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable: or

c. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.16 Containment Leakage Rate Testing Program

a.

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J. Option B. as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163. "Performance-Based Containment Leak-Test Program," dated September, 1995.

as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50.

Appendix J. Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Code Section XI. Subsection IWL.

except where relief has been authorized by the NRC. The containment concrete visual examination may be performed during either power operation. e.g., performed concurrently with other containment inspection-related activities such as tendon testing. or during a maintenance/refueling outage.

2.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50. Appendix J. Option B testing. will be performed in accordance with the requirements of and frequency specified by ASME Code Section XI, Subsection IWE, except where relief has been authorized by the NRC.

(continued)

PALO VERDE UNITS 1 AND 3 5.5 AMENDMENT NO. 43-,

151 PALO VERDE UNIT 2 AMENDMENT NO. 449. 151

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.16 Containment Leakage Rate Testing Program (continued)

b. The peak calculated containment internal pressure for the design I

basis loss of coolant accident. Pa. is 52.0 psig for Units 1 and 3.

and 58.0 psig for Unit 2. The containment design pressure is 60 psig.

c. The maximum allowable containment leakage rate. La. at Pa. shall I

be 0.1 % of containment air weight per day.

d.

Leakage Rate acceptance criteria are:

I

1. Containment leakage rate acceptance criterion is < 1.0 La.

I During the first unit startup following testing in accordance with this program, the leakage rate acceptance are < 0.60 La for the Type B and C tests and 5 0.75 La for Type A tests.

2. Air lock testing acceptance criteria are:

I a)

Overall air lock leakage rate is

0.05 La when tested I

at 2 Pa.

b)

For each door, leakage rate is 5 0.01 La when I

pressurized to 2 14.5 psig.

e. The provisions of SR 3.0.2 do not apply to the test frequencies in I

the Containment Leakage Rate Testing Program.

f. The provisions of SR 3.0.3 are applicable to the Containment I

Leakage Rate Testing Program.

PALO VERDE UNITS 1.2.3 5.50 AMENDMENT NO. 4-3.

151

Reporting Requirements 5.6 5.6 Reporting Requirements (continued)

I 5.6.6 PAM Report When a report is required by Condition B or G of LCO 3.3.10. "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability. and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Tendon Surveillance Report 5.6.7 Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.

5.6.8 e 5CrA f I r

.se Steam Generator Tube Inspection Report Within 15 days follow' g the completion of each inservice inspection of stea generator tubes. the number of tubes plugged and/or repaired i each steam generator shall be reported to Commission in Special Report.

The comp e results of the steam generator tube inser ce inspec on shall be submitted to the Commission in Special Rep within 12 months following completion of e inspection.

T s Special Report shall include:

/

a.

Number and extent of tubes inspecte

b. Location and percent of wall-t 'ckness penetration for each indication of an imperfecti

c.

Identification of tub lugged and/or repaired.

Results of steam gener r tube and sleeve inspections w fall into Category C-3 sh be reported in a Special Repor o the Commission within days and prior to resumption o plant operation and I1 provide a description of iv igations conducted t etermine cause of the tube degr tion and correctiv measures taken to prevent recurr ce.

PALO VERDE UNITS 1,2.3 5.6-6 AMENDMENT NO. 4-24, 137

INSERT 5.6.8 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, 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.

B Proposed Technical Specification Changes (retyped)

Pages:

ii 1.1-4 and 5 3.4.14-1 and 2 3.4.18-1 and 2 (new) 5.5-6 through 5.5-16 5.6-6

PALO VERDE NUCLEAR GENERATING STATION IMPROVED TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 3.3 INSTRUMENTATION 3.3.1 Reactor Protective System (RPS) Instrumentation -- Operating 3.3.2 RPS Instrumentation -- Shutdown 3.3.3 Control Element Assembly Calculators (CEACs) 3.3.4 RPS Logic and Trip Initiation 3.3.5 Engineered Safety Features Actuation System (ESFAS) Instrumentation 3.3.6 ESFAS Logic and Manual Trip 3.3.7 Diesel Generator (DG) -- Loss of Voltage Start (LOVS) 3.3.8 Containment Purge Isolation Actuation Signal (CPIAS) 3.3.9 Control Room Essential Filtration Actuation Signal (CREFAS) 3.3.10 Post Accident Monitoring (PAM) Instrumentation 3.3.11 Remote Shutdown System 3.3.12 Boron Dilution Alarm System (BDAS) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure. Temperature and Flow Departure from Nucleate Boiling (DNB) Limits 3.4.2 RCS Minimum Temperature for Criticality 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4.4 RCS Loops -- MODES 1 and 2 3.4.5 RCS Loops -- MODE 3 3.4.6 RCS Loops -- MODE 4 3.4.7 RCS Loops -- MODE 5. Loops Filled 3.4.8 RCS Loops -- MODE 5. Loops Not Filled 3.4.9 Pressurizer 3.4.10 Pressurizer Safety Valves -- MODES 1. 2. and 3 3.4.11 Pressurizer Safety Valves -- MODE 4 3.4.12 Pressurizer Vents 3.4.13 Low Temperature Overpressure Protection (LTOP) System 3.4.14 RCS Operational Leakage 3.4.15 RCS Pressure Isolation Valve (PIV) Leakage 3.4.16 RCS Leakage Detection Instrumentation 3.4.17 RCS Specific Activity 3.4.18 Steam Generator (SG) Tube Integrity 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Safety Injection Tanks (SITs) -- Operating 3.5.2 SITs -- Shutdown 3.5.3 ECCS -- Operating 3.5.4 ECCS -- Shutdown 3.5.5 Refueling Water Tank (RWT) 3.5.6 Trisodium Phosphate (TSP)

PALO VERDE UNITS 1.2.3 ii AMENDMENT NO. 444.

PALO VERDE UNITS 1.2.3 ii AMENDMENT NO. A4dd,

Definitions 1.1 1.1 Definitions (continued)

ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values. etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

Kn.lis the K effective calculated by considering the actual CEA configuration and assuming that the fully or partially inserted full strength CEA of highest worth is fully withdrawn.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or 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 (SG) to the Secondary System (primary to secondary LEAKAGE).

I (continued)

PALO VERDE UNITS 1,2.3 1.1-4 AMENDMENT NO. 4752,

Definitions Li1 1.1 Definitions LEAKAGE (continued)

b. Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE:

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

A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, cold leg reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

I MODE NEUTRON RATED THERMAL POWER (NRTP)

OPERABLE - OPERABILITY PHYSICS TESTS The indicated neutron flux at RTP.

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

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

These tests are:

a. Described in Chapter 14. Initial Test Program of the UFSAR:

b. Authorized under the provisions of 10 CFR 50.59: or

c. Otherwise approved by the Nuclear Regulatory Commission.

(continued)

PALO VERDE UNITS 1,2,3 1.1-5 AMENDMENT NO. 3S9,

RCS Operational LEAKAGE 3.4.14 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 RCS Operational LEAKAGE LCO 3.4.14 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE:

b. 1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE: and

d.

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

i APPLICABILITY:

MODES 1. 2. 3. and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE not within within limits.

limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

I PALO VERDE UNITS 1.2.3 3.4.14-1 AMENDMENT NO. F4-9,

RCS Operational LEAKAGE 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1

NOTES------------------

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

2. Not applicable to primary to secondary LEAKAGE Perform RCS water inventory balance.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SR 3.4.14.2

NOTE-------------------

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

Verify primary to secondary LEAKAGE is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s 150 gallons per day through any one SG.

PALO VERDE UNITS 1.2.3 3.4.14-2 AMENDMENT NO. az

SG Tube Integrity 3.4.18 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.18 Steam Generator (SG) Tube Integrity LCO 3.4.18 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY:

MODES 1. 2. 3. and 4.

ACTIONS

NOTE-------------------------------------

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity 7 days satisfying the tube of the affected repair criteria and tube(s) is maintained not plugged in until the next accordance with the refueling outage or Steam Generator SG tube inspection.

Program.

AND A.2 Plug the affected Prior to tube(s) in accordance entering MODE 4 with the Steam following the Generator Program.

next refueling outage or SG tube inspection.

B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A AND not met.

OR B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SG tube integrity not maintained.

PALO VERDE UNITS 1.2.3 3.4.18-1 AMENDMENT NO.

SG Tube Integrity 3.4.18 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.18.1 Verify SG tube integrity in accordance with In accordance the Steam Generator Program.

with the Steam Generator Program.

SR 3.4.18.2 Verify that each inspected SG tube that Prior to satisfies the tube repair criteria is entering MODE 4 plugged in accordance with the Steam following a SG Generator Program.

tube inspection.

PALO VERDE UNITS 1,2,3 3.4.18-2 AMENDMENT NO.

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Program A Steam Generator (SG) 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 and 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 SG 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, (continued)

PALO VERDE UNITS 1.2,3 5.5-6 AMENDMENT NO. 4BOB

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Program (continued) 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 0.5 gpm per SG and 1 gpm through both SGs.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.14, "RCS Operational LEAKAGE."

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.

For the original SGs with Alloy 600 MA tubes, all detected degradation in the portion of the tube below the bottom of the expansion transition that is not excluded from inspection per the provisions of 5.5.9.d shall be removed from service.

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. For the original SGs with Alloy 600 MA tubes, the section of tube from the tube-to-tubesheet weld to 12 inches below the bottom of the expansion transition for both the tube inlet and outlet may be excluded from inspection based on Westinghouse Report WCAP-16208-P, Revision 1, 'NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions." In addition to meeting the requirements of d.1, d.2a, d.2b, 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.

2a.

Original SGs with Alloy 600 MA tubes:

Inspect 100% of the tubes at seguentia1 periods of 60 effective full power months. The first sequential period shall be (continued)

PALO VERDE UNITS 1,2,3 5.5-7 AMENDMENT NO. 42t,

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Program (continued) considered to begin after the first inservice inspection of the SGs.

No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.

2b.

Replacement SGs with Alloy 690 TT tubes:

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 LEAKAGE.

monitoring operational primary to secondary (continued)

PALO VERDE UNITS 1.2,3 5.5-8 AMENDMENT NO. CAB.

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;

c. Identification of process sampling points which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage:

d. Procedures for the recording and management of data:

e. Procedures defining corrective actions for all off control point chemistry conditions; and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5.5.11 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2. and in accordance with Regulatory Guide 1.52, Revision 2 and ANSI N510-1980 at the system flowrate specified below +/- 10%.

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1.0 % when tested in accordance with Regulatory Guide 1.52, Revision 2. and ANSI N510-1980, at the system flowrate specified as follows

+ 10%:

(continued)

PALO VERDE UNITS 1.2.3 5.5-9 AMENDMENT NO. 424,

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 Ventilation Filter Testing Program (VFTP)

(continued)

ESF Ventilation System Control Room Essential Filtration System (CREFS)

Engineered Safety Feature (ESF)

Pump Room Exhaust Air Cleanup System (PREACS)

Flowrate 28,600 CFM 6,000 CFM

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 1.0 % when tested in accordance with Regulatory Guide 1.52, Revision 2. and ANSI N510-1980 at the system flowrate specified as follows +/- 10%:

ESF Ventilation System Flowrate 28,600 CFM CREFS ESF PREACS 6.000 CFM

c.

Demonstrate for each of the ESF systems that a charcoal adsorber sample, when obtained in accordance with the application of Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 2. March 1978, as described in Section 1.8 of the UFSAR, shows the methyl iodide penetration less than or equal to the value specified below, when tested in accordance with ASTM D3803-1989. at a temperature of 300C and to the relative humidity specified as follows:

ESF Ventilation System CREFS Penetration

< 2.5%

RH 70%

ESF PREACS 70%

(continued)

PALO VERDE UNITS 1.2,3 5.5-10 AMENDMENT NO. -13G,

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 Ventilation Filter Testing Program (VFTP) (continued)

d.

For each of the ESF systems. demonstrate the pressure drop across the combined HEPA filters. the prefilters. and the charcoal adsorbers is less than or equal to the value specified below when tested in accordance with Regulatory Guide 1.52. Revision 2, and ANSI N510-1980 at the system flowrate specified as follows +/- 10%:

ESF Ventilation System Delta P Flowrate CREFS 4.8 inches water gauge 28.600 CFM ESF PREACS 5.2 inches water gauge 6.000 CFM

e.

Demonstrate that the heaters for each of the ESF systems dissipate the following specified value when tested in accordance with ANSI N510-1980:

ESF Ventilation System Wattage ESF PREACS

> 19 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides control for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5. "Postulated Radioactive Release due to Waste Gas System Leak or Failure". The liquid radwaste quantities shall be determined in accordance with Standard Review Plan. Section 15.7.3.

"Postulated Radioactive Release due to Tank Failures".

(continued)

PALO VERDE UNITS 1.2.3 5.5-11 AMENDMENT NO. 1421,

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e.. whether or not the system is designed to withstand a hydrogen explosion);

b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of 2 0.5 rem to any individual in an unrestricted area. in the event of an uncontrolled release of the tanks' contents:

and

c. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to-the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR Part 20. Appendix B. Table 2. Column 2. at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.13 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established.

The program shall include sampling and testing requirements. and acceptance criteria. all in accordance with applicable ASTM Standards as referenced in the UFSAR. The purpose of the program is to establish the following:

(continued)

PALO VERDE UNITS 1.2.3 5.5-12 AMENDMENT NO. apt9,

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.13 Diesel Fuel Oil Testing Program (continued)

a. Acceptability of new fuel oil for use prior to storage tanks by determining that the fuel oil addition to has:

1. An API gravity or an absolute specific gravity within limits.

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. Water and sediment within limits when tested in accordance with ASTM D1796:

b. Other properties for ASTM 2D fuel oil within 31 days following sampling and tanks; and are within limits addition to storage

c. Total particulate concentration of the stored fuel oil is

< 10 mg/l when tested every 92 days in accordance with ASTM D-2276. Method A-2 or A-3.

Technical Specifications (TS) Bases Control Program 5.5.14 This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1.

A change in the TS incorporated in the license; or

2.

A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

(continued)

PALO VERDE UNITS 1,2.3 5.5-13 AMENDMENT NO. 44;.

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.14 Technical Specifications (TS) Bases Control Program (continued)

d.

Proposed changes that meet the criteria of Specification 5.5.14b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.15 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6. an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.

The SFDP shall contain the following:

a.

Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected:

b.

Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists:

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities: and

d.

Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed.

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to system(s) supported by the inoperable support system is also inoperable: or (continued)

PALO VERDE UNITS 1.2.3 5.5-14 AMENDMENT NO FW

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.15 Safety Function Determination Program (continued)

b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable: or

c. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program.

the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.16 Containment Leakage Rate Testing Program

a.

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J. Option B. as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163. "Performance-Based Containment Leak-Test Program." dated September. 1995, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50.

Appendix J. Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Code Section XI, Subsection IWL.

except where relief has been authorized by the NRC. The containment concrete visual examination may be performed during either power operation, e.g., performed concurrently with other containment inspection-related activities such as tendon testing, or during a maintenance/refueling outage.

2.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50. Appendix J. Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Code Section XI, Subsection IWE, except where relief has been authorized by the NRC.

(continued)

PALO VERDE UNITS 1. 2 AND 3 5.5-15 AMENDMENT NO. Abe,

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.16 Containment Leakage Rate Testing Program (continued)

b. The peak calculated containment internal pressure for the design basis loss of coolant accident. Pa. is 52.0 psig for Units 1 and 3.

and 58.0 psig for Unit 2. The containment design pressure is 60 psig.

c. The maximum allowable containment leakage rate, La, at Pa. shall be 0.1 Z of containment air weight per day.

d. Leakage Rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is < 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests.

2. Air lock testing acceptance criteria are:

a)

Overall air lock leakage rate is < 0.05 La when tested at 2 Pa.

b)

For each door, leakage rate is < 0.01 La when pressurized to 2 14.5 psig.

e. The provisions of SR 3.0.2 do not apply to the test frequencies in the Containment Leakage Rate Testing Program.

f. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

PALO VERDE UNITS 1,2.3 5.5-16 AMENDMENT NO. 44z,

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.6 PAM Report When a report is required by Condition B or G of LCO 3.3.10. "Post Accident Monitoring (PAM) Instrumentation." a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability. and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days.

The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.

5.6.8 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9. 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.

PALO VERDE UNITS 1.2.3 5.6-6 AMENDMENT NO. 44X, C

Proposed Technical Specification Bases Changes (mark-up)

Pages:

B 3.4.4-2 B 3.4.5-2 and 3 B 3.4.6-3 B 3.4.7-4 B 3.4.14-4 through B 3.4.14-7 B 3.4.18-1 through B 3.4.18-8 (new)

RCS Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABLE aspect for this LCO is the reactor coolant forced flow rate, SAFETY ANALYSES which is represented by the number of RCS loops in service.

(continued)

The reactor coolant pumps provide sufficient forced circulation flow through the reactor coolant system to assure adequate heat removal from the reactor core during power operation. The plant is designed to operate with both reactor coolant loops and associated reactor coolant Pumps in operation, and maintain a departure from nucleate boiling ratio (DNBR) above the DNBR Safety Limit during all normal operations and anticipated transients. The safety analyses that are of most importance to RCP operation are the total loss of reactor coolant flow. single pump locked rotor.

single pump (broken shaft or coastdown). and rod withdrawal events (Ref. 1).

RCS Loops - MODES 1 and 2 satisfy Criteria 2 and 3 of 10 CFR 50.36 (C)(2)(ii).

LCO The purpose of this LCO is to require adequate forced flow for core heat removal.

Flow is represented by having both RCS loops with both RCPs in each loop in operation for removal of heat by the two SGs. To meet safety analysis acceptance criteria for DNB, four pumps are required at rated power.

Each OPERABLE loop consists of two RCPs providing forced flow for heat trauEsonrt to an SG that is OPERABLE.J-;>

.4it le tea;e-iT ator TuSt PFe nram

an e

oop.

RABILITY with regard to S wa Wer level is ensured by the Reactor Protection System (RPS) in MODES 1 and 2.

(continued)

PALO VERDE UNITS 1.2.3 B 3.4.4-2 REVISION 7

RCS Loops -

MODE 3 B 3.4.5 BASES LCO The purpose of this LCO is to require two RCS loops to be available for heat removal, thus providing redundancy. The LCO requires the two loops to be OPERABLE with the intent of requiring both SGs to be capable (2 25% wide range water level) of transferring heat from the reactor coolant at a controlled rate. Forced reactor coolant flow is the required way to transport heat, although natural circulation flow provides adequate removal. A minimum of one running RCP meets the LCO requirement for one loop in operation.

The Note permits a limited period of operation without RCPs.

All RCPs may be de-energized for < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

This means that natural circulation has been established.

When in natural circulation, a reduction in boron concentration is prohibited because an even concentration distribution throughout the RCS cannot be ensured.

The intent is to stop any known or direct positive reactivity additions to the RCS due to dilution. Core outlet temperature is to be maintained at least 10F below the saturation temperature so that no vapor bubble may form and possibly cause a natural circulation flow obstruction. The 10 degrees F is considered the actual value of the necessary difference between RCS core outlet temperature and the saturation temperature associated with RCS pressure to be maintained during the time the pumps would be de-energized.

The instrument error associated with determining this difference is 27 degrees F. (The only restriction for instrumentation use is with pressurizer pressure less than or equal to 350 psia. and in that situation the narrow range pressurizer pressure instrumentation must be used.)

Therefore, the indicated value of the difference between RCS core outlet temperature and the saturation temperature associated with RCS pressure must be greater than or equal to 37 degrees F in order to use the provisions of the Note allowing the pumps to be de-energized.

In MODE 3 it is sometimes necessary to stop all RCPs (e.g.,to perform surveillance or startup testing, or to avoid operation below the RCP minimum net positive suction head limit). The time period is acceptable because natural circulation is adequate for heat removal, or the reactor coolant temperature can be maintained subcooled and boron stratification affecting reactivity control is not expected.

An OPERABLE RCS loop (loop 1 or loop 2) consists of at least one associated OPERABLE RCP and an associated SG that is OPERABLE, n cor a wi eam e o

3 (cninued)

PALO VERDE UNITS 1.2.3 B 3.4.5-2 REVISION 30

RCS Loops - MODE 3 B 3.4.5 BASES LCO EuPera P An RCP is OPERABLE if it is (continued) capa eing powered and is able to provide forced flow if required.

APPLICABILITY In MODE 3. the heat load is lower than at power; therefore.

one RCS loop in operation is adequate for transport and heat removal. A second RCS loop is required to be OPERABLE but not in operation for redundant heat removal capability.

Operation in other MODES is covered by:

LCO 3.4.4 "RCS Loops-MODES 1 and 2":

LCO 3.4.6. "RCS Loops - MODE 4":

LCO 3.4.7. "RCS Loops - MODE 5. Loops Filled";

LCO 3.4.8, "RCS Loops - MODE 5.

Loops Not Filled":

LCO 3.9.4. "Shutdown Cooling (SDC) and Coolant Circulation - High Water Level" (MODE 6): and LCO 3.9.5. "Shutdown Cooling (SDC) and Coolant Circulation - Low Water Level" (MODE 6).

ACTIONS A.1 If one required RCS loop is inoperable. redundancy for forced flow heat removal is lost. The Required Action is restoration of the required RCS loop to OPERABLE status within a Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This time allowance is a justified period to be without the redundant.

nonoperating loop because a single loop in operation has a heat transfer capability greater than that needed to remove the decay heat produced in the reactor core.

B.1 If restoration is not possible within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the unit must be placed in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In MODE 4. the plant may be placed on the SDC System. The Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is compatible with required operation to achieve cooldown and depressurization from the existing plant conditions in an orderly manner and without challenging plant systems.

(continued)

PALO VERDE UNITS 1.2.3 B 3.4.5-3 REVISION 30

RCS Loops - MODE 4 B 3.4.6 BASES LCO (continued)

Note 2 requires, that before an RCP may be started with any RCS cold leg temperature s 2140F during cooldown. or s 2910F during heatup. that secondary side water temperature (saturation temperature corresponding to SG pressure) in each SG is < 100OF above each of the RCS cold leg temperatures. The numerical values for RCS cold leg temperature at which this Note is applicable do not account for all instrument uncertainty. Use of an indicated value of 2170F or below during cooldown and 2940F or below during heatup ensures that the actual limits will not be exceeded.

These values, which include appropriate instrument uncertainty, are established within the applicable plant procedures.

Satisfying the above condition will preclude a large pressure surge in the RCS when the RCP is started.

Note 3 restricts RCP operation to no more than 2 RCPs with RCS cold leg temperature S 2000F. and no more than 3 RCPs with RCS cold leg temperature >200OF but

5000F.

Satisfying these conditions will maintain the analysis assumptions of the flow induced pressure correction factors due to RCP operation (Ref. 1)

An OPERABLE RCS loop consists of at least one OPERABLE RCP and an SG that is-O-PERABLE fihad-ORMneuth-h Sem water level specified in SR 3.

..2.

Similarly, for the SDC System, an OPERABLE SDC train is composed of an OPERABLE SDC pump (CS or LPSI) capable of providing flow to the SDC heat exchanger for heat removal.

RCPs and SDC pumps are OPERABLE if they are capable of being powered and are able to provide flow, if required.

APPLICABILITY In MODE 4. this LCO applies because it is possible to remove core decay heat and to provide proper boron mixing with either the RCS loops and SGs or the SDC System.

Operation in other MODES is covered by:

LCO 3.4.4 "RCS Loops-MODES 1 and 2":

LCO 3.4.5, "RCS Loops - MODE 3":

(continued)

PALO VERDE UNITS 1.2.3 B 3.4.6-3 REVISION 6

RCS Loops - MODE 5. Loops Filled B 3.4.7 BASES LCO Note 5 provides for an orderly transition from MODE 5 to (continued)

MODE 4 during a planned heatup by permitting removal of SDC trains from operation when at least one RCP is in operation.

This Note provides for the transition to MODE 4 where an RCP is permitted to be in operation and replaces the RCS circulation function provided by the SDC trains.

An OPERABLE SDC train is composed of an OPERABLE SDC pump (CS or LPSI) capable of providing flow to the SDC heat exchanger for heat removal.

SDC pumps are OPERABLE if they are capable of being powered and are able

r i e flow (current Section XI). if required.

SG can perform as a heat sink when it is OPERABLE litSSnce *olluzthe5dlE~:DlL~iQ (ptkallp§7i tfte-Eminim-~ water level specified in The RCS loops may not be considered filled until two conditions needed for operation of the steam generators are met.

First, the RCS must be intact.

This means that all removable portions of the primary pressure boundary (e.g.,

manways, safety valves) are securely fastened. Nozzle dams are removed.

All manual drain and vent valves are closed, and any open system penetrations (e.g.. letdown, reactor head vents) are capable of remote closure from the control room.

An intact primary allows the system to be pressurized as needed to achieve the subcooling margin necessary to establish natural circulation cooling. When the RCS is not intact as described, a loss of SDC flow results in blowdown of coolant through boundary openings that also could prevent adequate natural circulation between the core and steam generators.

Secondly. the concentration of dissolved or otherwise entrained gases in the coolant must be limited or other controls established so that gases coming out of solution in the SG U-tubes will not adversely affect natural circulation. With these conditions met, the SGs are a functional method of RCS heat removal upon loss of the operating SOC train. The ability to feed and steam SGs at all times is not re uired when RCS temperature is less than 2100F because significant loss of SG inventory through boiling will not occur during time anticipated to take corrective action. The required SG level provides sufficient time to either restore the SDC train or implement a method for feeding and steaming the SGs (using non-class components if necessary).

(continued)

PALO VERDE UNITS 1.2.3 B 3.4.7-4 REVISION 2

RCS Operational LEAKAGE B 3.4.14 BASES i

LCO (continued)

LCO 3.4.15, "RCS Pressure Isolation Valve (PIV)

Leakage." measures leakage through each individual PIV and can impact this LCO.

Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leaktight. If both valves leak and result in a loss of mass from the RCS. the loss must be included in the allowable identified LEAKAGE.

d.

Primary to Secondary LEAKAGE through Any One SG he maximum allo ble operational primary o secondary LEA through any one SG of 50 gpd is R l based on ope ting experience as an i ication of one or more pr agating tube leak mech nms. This operatio limit is significantl less than the initia conditions assumed in t safety analyses.

Inset The Seam Generator Tube Surv llance Program de ribed in TS Section 5.5 ensures that the ructural integrity of SG tubes is maintained

,/

he 150 gpd leakage rat limit provides addition assurance against tub rupture at normal and fa ted conditions and prov es additional assurance at cracks will not p pagate to burst prior to etection by leakage moni ring methods and commenc ent of

(

plant shutdo Primary to secondary L GE must be included i he total allowable limit or identified LEAKAGE.

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

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

(continued)

PALO VERDE UNITS 1.2.3 B 3.4.14-4 REVISION 7

TS Bases B 3.4.14, LCO, Section d (INSERT)

The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4). 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.

RCS Operational LEAKAGE B 3.4.14 BASES ACTIONS A.1 UdLE GE ldentified LEAKAG e

LE n excess of the LCO limitm iUs mu te 7 rediced~to~Tthii limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 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.

B.1 and B.2 I

a f ^.e ov *d$s,)

If any pressure boundary LEAKAGE existsor if unidentifiedfcX 9 iepfe atpar, seedo LEAKAGE cannot be reduced to WiiT Wimis within 4-hours. the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

If on r more SGs are operable, due to i 3.4.14.2. the uni is in a conditio outside the accideft analyses J refore. LCO 3.0. must be entered)'iiediately.

v (continued)

PALO VERDE UNITS 1.2.3 B 3.4.14-5 REVISION 2

RCS Operational LEAKAGE B 3.4.14 BASES SURVEILLANCE REQUIREMENTS SR 3.4.14.1 Verifying RCS LEAKAGE to bewithin the LCO limits ensures the integrity of the RCP is maintained.

Pressure boundary LEAKAGE would at first ipear as unidentified LEAKAGE and can only be positivel identified by inspection.

Unidentified LEAKA and identified LEAKAGE are determined by performance of n RCS water inventory balance.

Primary to secondary LE GE is also measured by performance of.an RCS water invtory balance in conjunction with efflupnt monitoring thin the secondary steam and feedwater,.systems.

/

The RCS ater inventory balance must be performed' with the reacto at steady state operating conditions ayrd near oper ing pressure. Therefore, this SR is got required to be erformed in MODES 3 and 4. until 12 heufs of steady spdte operation near operating pressure ybve elapsed. This eans that once steady state operatin onditions are established. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is allowed for/completing the

!XSurveillance if the Surveillance F equency interval wta Survilla in MODE 5 or 6. Further discussion of SR note format is found in Section 1.4-. Frequency.

The Note in the Frequency olumn allows for SR 3.4.14.1 nonperformance due to p nned or unplanned power manipulations. This Noe is not intended to allow power/

§ manipulations solely for the purpose of avoiding SRa4.14.1 performance. Stedy state operation is required td perform a proper water jnventory balance: calculations during maneuvering are not useful and a Note requires'the Surveillance-to be met when steady state is-established.

)

For RCS operational LEAKAGE determination'by water inventory balance.-steady state is defined as stable RCS pressure.

temperature. power level, pressurizer and makeup tank levels, makeup and letdown. and RCP seal injection and return flows.

v'--.

(continued)

PALO VERDE UNITS 1.2.3 B 3.4.14-6 REVISION 2

TS Bases SR 3.4.14.1 (REVISION)

Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance. Pgimary-te 6eoRnda'r LEAKAGE is also measurod by performance of an RGS watcr invcn~teO balaRnoe in con;ejunctin with effluent eoniteorng within the sceondar; ctcm anRd fcdwateFr Systems.

The RCS water inventory balance must be performed with the reactor at steady state operating conditions (stable power level, urizer andrnakeup tank levels, makeup andletdown, and 'RCP seal injection and return flows) and-near operating pressure. Thero -This surveillance is modified by'two~notes..Notei, states that this SR is not required to be performed in MODES 3-and 4, until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of after establishing steady state operation. neareoperating pressurc have elapsed. This means that once steady state operating conditions are established, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is allowed for completing the Surveillance if the Sureillanco Frequennc interal was cxccded in MODE 5 er 6.

When required by the7Frequency, anda'fe steady state operating conditions are established,1the surveillance must b Iecompteted prior to the end of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation. If steady state' operating conditions have not been established for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , this surveillance is not required until steavdy.~'tate operation is established for.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This SR is not required to be 6om~pleted'priortoichan'gi~ng'MODES if,steady.'state 6peration has not been'established for.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowa nce provide's sufficient time to collect and process all necessary.data afterlstable plant conditions are established. Further discussion of SR note format is found in Section 1.4, Frequency.

The-Note A1 in the Frequency column allows for SR 3.4.14.1 nonperformance due to planned or unplanned power manipulations transients. This Note is not intended to allow power manipulations transients solely for the purpose of avoiding SR 3.4.14.1 performance.

Steady state operation is required to perform a proper water inventory balance isince calculations during maneuvering are not useful. and a Note requires the Survoillance to be met when steady state is established. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. These leakage detection systems are specified in LCO 3.4.16, "RCS Leakage Detection Instrumentation."

2 that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day'cannot be measured accurately by an RCS water inventory balance. H;;

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. A Note under the Frequency column states that this SR is required to be performed during steady state eperation.

RCS Operational LEAKAGE B 3.4.14 BASES SURVEILLANCE REQUIREMENTS SR 3.4.14.1 (continued)

An early warning 9, pressure boundary LEAKAGE ors, unidentified L AGE is provided by the automat'systems that monitor t6e containment atmosphere radio tivity and the contal nt sump level. These leakage tection systems are spec' jed in LCO 3.4.16. "RCS Leakag etection Instru ntation."

) lTh 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Frequency is a reason e interval to trend AKAGE and recognizes the impo ance of early leakage""'

detection in the prevention accidents. A Note under the tFrequency column states thS this SR is required to be C

SR 3.4.14.2 is R povde the means nece ry to determineS leer fOERABILPITrY i zan operationa ODE. The require to

\\

g 3 4 1,.Z demonstrat G tube integr' y in accordance wi the Steam sS **Generato ou e Surveillcume Program empai~ the A5 impor ce of SG tube

'tegrity.

even thou this g {eva

( SurM illance cannot< e performed at nor operating REFERENCES

1.

10 CFR 50. Appendix A. GOC 30.

2. Regulatory Guide 1.45, May 1973.

3. UFSAR. Section 15.

\\\\44, (JEX q)-O6 l eqpi 6w v,,.4ov

=reg s

)

Pr\\

&;y~fM-condtu'y Lec6k Gc(3A1-1ee.} "

PALO VERDE UNITS 1.2.3 B 3.4.14-7 REVISION 2

TS Bases SR 3.4.14.2 (INSERT)

This SR verifies that primary to secondary LEAKAGE is less than 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 SR is not met, compliance with LCO 3.4.18, "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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. This means that once steady state operating conditions are established, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is allowed for completing the Surveillance. When required by the Frequency, and after steady state operating conditions are established, the surveillance must be completed prior to the end of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation. If steady state operating conditions have not been established for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , this surveillance is not required until steady state operation is established for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This SR is not required to be completed prior to changing MODES if steady state operation has not been established for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Further discussion of SR note format is found in Section 1.4, Frequency.

The Note allows for SR 3.4.14.2 nonperformance due to planned or unplanned transients.

This Note is not intended to allow transients solely for the purpose of avoiding SR 3.4.14.2 performance. 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 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 5).

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

B 3.4.18 Steam Generator (SG) Tube Integrity BASES BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. SG tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such. are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system.

This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, "RCS Loops - MODES 1 and 2," LCO 3.4.5. "RCS Loops - MODE 3."

LCO 3.4.6. "RCS Loops - MODE 4." and LCO 3.4.7, "RCS Loops -

MODE 5. Loops Filled."

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

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

Specification 5.5.9. "Steam Generator (SG) Program." requires that a program be established and implemented to ensure that SG tube integrity is maintained.

Pursuant to Specification 5.5.9.

tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE.

The SG performance criteria are described in Specification 5.5.9. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

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

(continued)

PALO VERDE UNITS 1,2,3 B 3.4.18-1 REVISION

INSERT New Bases B 3.4.18 SG Tube Integrity B 3.4.18 BASES APPLICABLE SAFETY ANALYSES The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes,,a bounding Drimary

,to secondaryvLEAKAGE..

rate. eaual to onegallon per minute (1440 gallons per day) in the unaffected SG-plusfthe lekaag'e 'rateassoc-iated with a' '

double'-ended 'rdpiture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.

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 of 0.5 gallon per minute (gpm) from each SG or 1 gpm from both SGs. or is assumed to increase to those levels as a result of accident induced conditions.

For accidents that do not involve fuel damage, the primary coolant activity level is assumed to be equal to the LCO 3.4.17. "RCS Specific Activity."

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).

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

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

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 (continued)

PALO VERDE UNITS 1,2,3 B 3.4.18-2 REVISION

INSERT New Bases B 3.4.18 SG Tube Integrity B 3.4.18 BASES LCO tube-to-tubesheet weld at the tube outlet. The tube-to-(continued) tubesheet weld is not considered part of the tube.

(NOTE:S n

As~~~~~~~~

SG-(OE pciiain559 Stuea Genseration rg Provision (d).--'secifies a SG tube inspection length for the original SGs with Alloy:600 MA tubes as the length of-the I-tube from 12-,inchies bels w the hot-leg expansion transition at the tube inlet 't'o'12in'ches below the cold-leg expansion

' -transition atthtjube outlet.)

An SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.5.9. "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.

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 (continued)

PALO VERDE UNITS 1,2,3 B 3.4.18-3 REVISION

INSERT New Bases B 3.4.18 SG Tube Integrity B 3.4.18 BASES LCO (continued) classified as secondary loads.

For circumferential degradation. the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III. Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification.

This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref.5).

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident. other than a SGTR. is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 0.5 gpm from each SG or 1 gpm total from both SGs. The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.14. "RCS Operational LEAKAGE." 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 main steam line break. If this amount of LEAKAGE is due to more than one crack. the cracks are very small. and the above assumption is conservative.

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

(continued)

PALO VERDE UNITS 1,2,3 B 3.4.18-4 REVISION

INSERT New Bases B 3.4.18 SG Tube Integrity B 3.4.18 BASES APPLICABILITY RCS conditions are far less challenging in MODES 5 and 6 (continued) than during MODES 1. 2. 3. and 4. In MODES 5 and 6. primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

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

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

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

A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.18.2. An evaluation of SG tube integrity of the affected tube(s) must be made.

Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met.

In order to determine if a SG tube that should have been plugged has tube integrity. an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection.

If it is determined that tube integrity is not being maintained, Condition B applies.

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

If the evaluation determines that the affected tube(s) have tube integrity. Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected (continued)

PALO VERDE UNITS 1,2,3 B 3.4.18-5 REVISION

INSERT New Bases B 3.4.18 SG Tube Integrity B 3.4.18 BASES ACTIONS A.1 and A.2 (continued) tube(s). However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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

SURVEILLANCE SR 3.4.18.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1). and its referenced EPRI Guidelines, establish the content of the Steam Generator Program.

Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

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

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

Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations.

The Steam Generator Program also specifies the inspection methods to be used to find potential degradation.

Inspection methods are a (continued)

PALO VERDE UNITS 1,2,3 B 3.4.18-6 REVISION

INSERT New Bases B 3.4.18 SG Tube Integrity B 3.4.18 BASES SURVEILLANCE SR 3.4.18.1 (continued)

REQUIREMENTS function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

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

SR 3.4.18.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 5.5.9 are intended to ensure that the 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 entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

(continued)

PALO VERDE UNITS 1,2,3 B 3.4.18-7 REVISION

INSERT New Bases B 3.4.18 SG Tube Integrity B 3.4.18 BASES REFERENCES

1.

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

2.

10 CFR 50 Appendix A, GDC 19.

3.

10 CFR 100.

4.

ASME Boiler and Pressure Vessel Code,Section III.

Subsection NB.

5.

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

6.

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

PALO VERDE UNITS 1,2,3 B 3.4.18-8 REVISION

ENCLOSURE 3 Westinghouse Affidavit Requesting That WCAP-16208-P, Revision 1 Be Withheld From Public Disclosure In Accordance With The Provisions Of 10 CFR 2.390(A)(4)

(Westin'house U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Westinghouse Electric Company Nuclear Services P. O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA Direct tel: 412-374-4643 Direct fax: 412-374-4211 e-mail: greshaja@westinghouse.com Our ref: CAW-05-1997 May 18,2005 APPLICATION FOR WITHHOLDING PROPRIFTARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Revision 1 to WCAP-16208-P, "INDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions"

Reference:

Westinghouse Report WCAP-16208-P Revision 1, "NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions," May 2005.

Westinghouse, via the Arizona Public Service Company (APS), transmits the above-referenced proprietary document for which withholding is requested. Affidavit CAW-05-1997, which accompanies this letter and is signed by the owner of the proprietary information, Westinghouse Electric Company LLC, sets forth the basis on which this information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR Section 2.390(b)(4) of the Commission's regulations. This letter also authorizes use of the accompanying affidavit by APS.

In conformance with the requirements of 10 CFR 2.390, Westinghouse confirms that the information contained within the referenced document is proprietary. The justification for claiming this document as proprietary is identified in Sections (4)(ii)(a) through (4)(ii)(f) of the enclosed affidavit.

Communication with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-05-1997, and be addressed to the undersigned.

Very truly yours, James A. Gresham Manager Regulatory Compliance and Plant Licensing

Enclosure:

A BNFL Group Company

CAWV-05 1997 Page I CAW-05-1997 Page 1 AFFIDAVIT STATE OF CONNECTICUT

)

) ss: WINDSOR, CT COUNTY OF HARTFORD

)

Before me, the undersigned authority, personally appeared James A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse"), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

ames A. Gresham, Manager Regulatory Compliance and Plant Licensing Westinghouse Electric Company, LLC Sworn to and subscribed before me this 18'h day of May 2005.

y commission expires May 31, 2008.

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CAW-05-1997 Page 2 CAW-05-1997 Page 2 (1)

I, James A. Gresham, depose and say that I am Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse Electric Company LLC ("Westinghouse"), and as such I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Electric Company LLC.

(2)

I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.

(3)

I have personal knowledge of the criteria and procedures utilized by the Westinghouse Electric Company LLC in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4)

Pursuant to the provisions of paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i)

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence.

The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a)

The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b)

It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c)

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

CAW-05-1997 Page 3 (e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f)

It contains patentable ideas, for which patent protection may be desirable.

(iii) There are sound policy reasons behind the Westinghouse system for classification of proprietary information, which include the following:

(a)

The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b)

It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c)

Use of this information by our competitors would put Westinghouse at a competitive disadvantage by reducing their expenditure of resources at our expense.

(d)

Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e)

Unrestricted disclosure of this information would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f)

The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390; it is to be received in confidence by the Commission.

(v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(vi) The proprietary information sought to be withheld in this submittal is that which is contained in Westinghouse Report WCAP-16208-P Revision 1, "NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions," May 2005.

CANV-05-1997 Page 4 CAW-05-1997 Page 4 The information is part of a model that will enable Westinghouse to support utilities with CE NSSS plants in the identification and application of a steam generator tubesheet inspection model, including:

(a)

The identification of important factors relevant to the determination of the recommended steam generator tubesheet inspection length, and (b)

Development of a generic methodology for the applicability of the inspection length model to utilities with CE NSSS plants.

(vii) Further this information has substantial commercial value as follows:

(a)

Westinghouse plans to sell the use of similar information to its customers for purposes of meeting NRC requirements for licensing documentation.

(b) The information requested to be withheld reveals the distinguishing aspects of a methodology that was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar advanced nuclear power plant designs and to provide licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

CAWV-05-1997 Page 5 CAV.O5-1997 Pane 5 PROPRIETARY INFORMATION NOTICE In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, Westinghouse confirms that the information in Westinghouse Report WCAP-16208-P Revision 1, "NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions," May 2005 is proprietary. The justification for claiming this information as proprietary is indicated in Sections (4)(ii)(a) through (4)(ii)(f) of affidavit CAW-05-1997 accompanying this transmittal.

COPYRIGHT NOTICE The information transmitted herewith is copyright by Westinghouse. The NRC is permitted to make the number of copies of the information that are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

v t e Letter Sequence Request
TAC:MC8850, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request, Request, Request Acceptance... Supplement, Supplement, Supplement Administration Withholding Request Acceptance Questions 23 March 2006 21 June 2006 Answers Results Approval Other: ML063130496
MONTHYEAR2006-11-09 [Table View]

ML051520413

Text

Dear Sirs:

Subject:

Enclosures:

Subject:

1.0 INTRODUCTION

2.0 DESCRIPTION

3.0 BACKGROUND

Background:

Background:

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

9.1 Precedent

9.2 Precedent

10.0 REFERENCES

References:

References:

1.0 INTRODUCTION

2.0 DESCRIPTION

3.0 BACKGROUND

Background:

Background:

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

9.1 Precedent

9.2 Precedent

10.0 REFERENCES

References:

References:

Subject:

Reference:

Enclosure:

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