Text

Request for Technical Specification Amendment and Exemption from 10 CFR 50, Appendix G, to Relocate the Reactor Coolant System Pressure and Temperature Limits and Low Temperature Overpressure Protection Enable Temperatures
	ML090641014
Person / Time
Site:	Palo Verde
Issue date:	02/19/2009
From:	Mims D; Arizona Public Service Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	102-05960-DCM/GAM
	Download: ML090641014 (87)
	v • d • e

10 CFR 50.12 10 CFR 50.60(b) 10 CFR 50.90 A

A subsidiarv of Pinnacle West Capital Corporation Dwight C. Mims Mail Station 7605 Palo Verde Nuclear Vice President Tel: 623-393-5403 PO Box 52034 Generating Station Regulatory Affairs and Plant Improvement Fax: 623-393-6077 Phoenix, Arizona 85072-2034 102-05960-DCM/GAM February 1-9, 2009 ATTN: -Document Control -Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Dear Sirs:

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

-Units 1, 2, and -3

-Docket Nos. STN 528, 50-529, and 50-530 Request for Technical Specification Amendment and Exemption from 10 CFR 50, Appendix G, to Relocate the Reactor Coolant System Pressure and Temperature Limits and the Low Temperature Overpressure Protection Enable Temperatures

-Pursuant to 1.0 CFR 50.90, Arizona Public Service Company (APS) hereby requests a Technical Specification amendment to relocate the -reactor coolant system (RCS)

-pressure and temperature (P/T) -limits and the low temperature overpressure protection (LTOP) enable temperatures to a licensee-controlled document outside of the Technical Specifications (TSs). The.PIT limits -and LTOP enable temperatures would be specified

-in -a -Pressure and Temperature Limits Report (PTLR) that would be -located in the PVNGS Technical Requirements Manual (TRM) and administratively controlled by a new Specification 5.6.9. The P/T limits and LTOP enable temperatures specified in the

-proposed PTLR were determined using the NRC-approved Topical Report CE NPSD-683-A, -as -required by the proposed Specification 5.6.9. Enclosure 1 contains the

-evaluation of the proposed TS -changes.

These proposed changes are consistent with: (1) the guidance in NRC Generic Letter 03, "Relocation of the -Pressure-Temperature Limits Curves and Low Temperature Overpressure Protection System Limits"; (2) Topical Report CE NPSD-683-A,

-Revision 6, including the -related NRC safety evaluation; (3) Technical Specification Task Force (TSTF) Traveler number TSTF-408, Revision 1; and (4) Combustion Engineering Standard Technical Specifications, NUREG-1432, Revision 3.1.

A oo(

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway -

Comanche Peak

Diablo Canyon 0 Palo Verde 0 San Onofre

South Texas

Wolf Creek

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Request for Technical Specification Amendment and Exemption for Pressure-Temperature Limits Page 2 Provided in Enclosure 2 in this submittal is an application for an exemption from the requirements of 10 CFR Part 50, Appendix G. The requested exemption would allow for the application of the Combustion Engineering Nuclear Steam Supply System (CE NSSS) methods of CE NPSD-683-A, Revision 6, for calculating KIM values to the calculation of P/T limits, in lieu of the methodology cited in the ASME Code,Section XI, Appendix G. This exemption request is being submitted pursuant to the provisions of 10 CFR 50.60(b) and 10 CFR 50.12.

Approval of the proposed amendment and exemption is requested by January 31, 2010.

Once approved, the amendment shall be implemented within 90 days.

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 10 CFR 50.91 (b)(1).

No commitments are being made to the NRC by this letter. Should you need further information regarding this amendment request, please contact Russell A. Stroud, Licensing Section Leader, at (623) 393-5111.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on _______

(Date)

Sincerely, DCM/SAB/GAM/gat

Enclosures:

1.

Evaluation of the Proposed TS Change

2.

Request for Exemption cc:

E. E. Collins Jr.

NRC Region IV Regional Administrator R. Hall NRC NRR Project Manager R. I. Treadway NRC Senior Resident Inspector for PVNGS A. V. Godwin Arizona Radiation Regulatory Agency (ARRA)

T. Morales Arizona Radiation Regulatory Agency (ARRA)

ENCLOSURE 1 Evaluation of the Proposed TS Change

Subject:

Request for Technical Specification Amendment to Relocate the Reactor Coolant System Pressure and Temperature Limits and the Low Temperature Overpressure Protection Enable Temperatures to a Pressure and Temperature Limits Report (PTLR) 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

3.1 Evaluation of the Proposed Changes to the PVNGS 1, 2, and 3 TSs 3.2 Evaluation of the Proposed Methodology for the PTLR Against the Criteria for Approved Methodologies in Attachment 1 of Generic Letter 96-03 3.3 Evaluation of the PTLR Contents Against the Seven Criteria for PTLR Contents in Attachment 1 of Generic Letter 96-03 and the 26 Action Items from CE NPSD-683-A, Revision 6

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

ATTACHMENTS:

1.

Technical Specification Page Markups

2.

Technical Specification Bases Page Markups

3.

Retyped Technical Specification Pages

4.

Technical Requirements Manual Page Markups (Includes the PTLR)

5.

WCAP-16835-NP, Palo Verde Nuclear Generating Station Units 1,2, and 3; Basis for RCS Pressure and Temperature Limits Report, June 2008

6.

APS Responses to the NRC Request for Additional Information Related to the San Onofre Nuclear Generating Station (SONGS) PTLR Amendment Request 1

Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR 1.0

SUMMARY

DESCRIPTION This evaluation supports a request 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 changes would revise the Operating Licenses to relocate the reactor coolant system (RCS) pressure and temperature (P/T) limits and the low temperature overpressure protection (LTOP) enable temperatures to a licensee-controlled document outside of Technical Specifications (TSs). The P/T limits and LTOP enable temperatures would be specified in a Pressure and Temperature Limits Report (PTLR) that would be located in the PVNGS Technical Requirements Manual (TRM) and administratively controlled by a new Specification 5.6.9. The P/T limits and LTOP enable temperatures specified in the proposed PTLR were determined using the NRC-approved Topical Report CE NPSD-683-A, as required by the proposed Specification 5.6.9. The TRM is a licensee-controlled document referenced in Section 13.7 of the Updated Final Safety Analysis Report (UFSAR), and changes to the TRM are controlled in accordance with the provisions of 10 CFR 50.59.

These proposed changes are consistent with: (1) the guidance in NRC Generic Letter 96-03, "Relocation of the Pressure-Temperature Limits Curves and Low Temperature Overpressure Protection System Limits" (Ref. 1); (2) Topical Report CE NPSD-683-A, Revision 6, including the related NRC safety evaluation (Ref. 2); (3) Technical Specification Task Force (TSTF) Traveler number TSTF-408 (Ref. 3); and (4)

Combustion Engineering Standard Technical Specifications, NUREG-1432, Revision 3.1 (Ref 4).

2.0 DETAILED DESCRIPTION The following TS changes are proposed in order to relocate the P/T limits and LTOP temperatures to a licensee controlled document:

1.1, Definitions o

List and define "PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)."

(Note: The proposed PTLR definition is editorially modified from the definition in Combustion Engineering Standard Technical Specifications, NUREG-1432, by replacing "unit specific document" with "site specific document," to reflect the single PTLR that will contain the P/T limits for all three PVNGS units.)

2 Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR 3.4.3, RCS Pressure and Temperature (P/T) Limits o Relocate the P/T limit table and figures to the PTLR, and reference the PTLR.

o Relocate the maximum temperature change during hydrostatic testing operations to the PTLR.

3.4.6, RCS Loops - MODE 4 o

Relocate the LTOP enable temperatures to the PTLR, and reference the PTLR.

3.4.7, RCS Loops - MODE 5, Loops Filled o

Relocate the LTOP enable temperatures to the PTLR, and reference the PTLR.

3.4.11, Pressurizer Safety Valves - MODE 4 o Relocate the LTOP enable temperatures to the PTLR, and reference the PTLR.

3.4.13, Low Temperature Overpressure Protection (LTOP) System o Relocate the LTOP enable temperatures to the PTLR, and reference the PTLR.

5.6.9, Reactor Coolant System Pressure and Temperature Limits Report o Add this new section to TS 5.6, Reporting Requirements, to specify the PTLR content and reporting requirement. The analytical method used to determine the PTLR limits, CE NPSD-683-A, will be identified by number and title. The complete identification for this Topical Report, including the revision and date, will be specified in the PTLR.

The P/T limits and LTOP enable temperatures specified in the proposed PTLR were determined using the NRC-approved Topical Report CE NPSD-683-A, as required by the proposed Specification 5.6.9.

Technical Specification page markups are provided in Attachment 1. Associated TS Bases page markups are provided in Attachment 2. Retyped TS Pages are provided in. Technical Requirements Manual (TRM) page markups, which include the PTLR as a new Appendix, are provided in Attachment 4.

3 Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR These proposed changes are consistent with: (1) the guidance in NRC Generic Letter (GL) 96-03, "Relocation of the Pressure-Temperature Limits Curves and Low Temperature Overpressure Protection System Limits" (Ref. 1); (2) Topical Report CE NPSD-683-A, Revision 6, including the related NRC safety evaluation (Ref. 2); (3)

Technical Specification Task Force (TSTF) Traveler number TSTF-408 (Ref. 3); and (4)

Combustion Engineering Standard Technical Specifications, NUREG-1432, Revision 3.1 (Ref 4).

As described in GL 96-03, during the development of the improved standard technical specifications (STS), a change was proposed to relocate the P/T curves and LTOP limits currently contained in the TS to a licensee-controlled document. As one of the improvements to the STS, the NRC staff agreed with the industry that the curves and limits may be relocated outside the TS to a licensee-controlled document so that the licensee could maintain these limits efficiently and at a lower cost, provided that the parameters for constructing the curves and limits are derived using a methodology approved by the NRC.

This amendment request is similar to one that Southern California Edison (SCE) submitted to the NRC by letter dated January 28, 2005 (Ref. 5), and supplemented by letter dated January 12, 2006 (Ref. 6), for SONGS Units.2 and 3 to relocate the RCS P/T limits and LTOP limits from the TSs to a licensee-controlled PTLR. The NRC approved the SONGS Operating License amendments in a letter dated July 13, 2006 (Ref. 7).

3.0 TECHNICAL EVALUATION

Appendix G to Part 50 of 10 CFR requires licensees to establish limits on the allowable pressure and temperature in order to protect the reactor coolant pressure boundary against brittle failure. These limits are defined by P/T limit curves for normal operations (including reactor heatup and cooldown operations, operations with the reactor critical, and transient operating conditions) and during pressure testing conditions (i.e., inservice leak rate testing and hydrostatic testing conditions). For PWRs, the LTOP system limits ensure that the pressure remains below the applicable P/T limits.

In support of this amendment request to relocate the P/T LTOP limits from the PVNGS TSs to the PTLR, Westinghouse Electric Company LLC prepared report WCAP-1 6835, "Palo Verde Nuclear Generating Station, Units 1, 2, and 3; Basis for RCS Pressure and Temperature Limits Report," June 2008. This report is provided as Attachment 5 to'this Enclosure.

3.1 Evaluation of the Proposed Changes to the PVNGS Units 1, 2, and 3 TSs A detailed description of the proposed TS requirements related to the implementation of the PTLR for PVNGS Units 1, 2, and 3 is provided in Section 2.0 of this Enclosure. The proposed definition of the PTLR in Section 1.1 of the PVNGS TSs identifies the new 4

Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR Specification 5.6.9 which would control the PTLR content. Specification 5.6.9 directly references the individual specifications for which RCSP/T limits and LTOP system limits are established in the PTLR. Each of these specifications references the PTLR, as appropriate, in the proposed specification LCO. The proposed PTLR definition (as modified) and controlling TSs meet the technical criteria of GL 96-03 (Ref. 1) and are consistent with NUREG-1432, Revision 3.1, "Standard Technical Specifications, Combustion Engineering Plants" (Ref 4), as modified by NRC-approved Technical Specification Task Force (TSTF) traveler TSTF-408 described below. The P/T limits and LTOP enable temperatures specified in the proposed PTLR were determined using the NRC-approved Topical Report CE NPSD-683-A, as required by the proposed Specification 5.6.9.

The adoption of TSTF-408 (Ref. 3) allows NRC-approved Topical Reports (TRs) to be identified by number and title in Specification 5.6.9. This allows APS to ýuse current, approved TRs to support the calculation of parameters in the PTLR without having to submit an amendment to the Operating License every time the TR is revised. The proposed PVNGS Units 1, 2, and 3 PTLR (Attachment 4) provides the specific information (i.e., report number, title, revision, and date) identifying the particular approved TR which documents the methodology used to determine the P/T limits and LTOP system limits. This provides assurance that only the approved version of the referenced TRs is used for the determination of the P/T limits and LTOP system limits since the complete citation is provided in the PTLR, and the'PTLR methodology documented in the TR was approved by the NRC.

3.2 Evaluation of the Proposed Methodology for the PTLR Against the Criteria for Approved Methodologies in Attachment 1 of Generic Letter 96-03 As described in the proposed PVNGS Units 1, 2, and 3 PTLR (Attachment 4) and WCAP-1 6835 (Attachment 5), the proposed PVNGS P/T limits and LTOP system limits have been established in accordance with the NRC-approved methodology in Combustion Engineering Nuclear Steam Supply System (CE NSSS) topical report CE NPSD-683-A, Revision 6 (Ref. 2). Any future changes to the PVNGS PTLR would be determined in accordance with an approved version of CE NPSD-683-A, as required by Section 5.6.9 of the proposed revised TS.

The NRC has evaluated the methodology in CE NPSD-683-A for establishing in a PTLR the P/T limits and LTOP system limits for CE plants. This evaluation was documented in the NRC safety evaluation dated March 16, 2001, which was incorporated into CE NPSD-683-A, Revision 6, along with the key NRC recommendations in the SE. Thus, CE NPSD-683-A, Revision 6, represents the latest NRC-approved version of CE NPSD-683.

The methodology in CE NPSD-683-A, Revision 6, meets the minimum technical requirements for approved methodologies specified in Attachment 1 to GL 96-03. The use of CE NPSD-683-A, Revision 6, as the proposed methodology for the PVNGS Units 5

Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR 1, 2, and 3 PTLR meets the criterion in GL 96-03 specifying that an approved methodology be used for the development of the PTLR.

3.3 Evaluation of the PTLR Contents Against the Seven Criteria for PTLR Contents in Attachment 1 of Generic Letter 96-03 and the 26 Action Items from CE NPSD-683-A, Revision 6 of GL 96-03 (Ref. 1) contains seven technical criteria (PTLR criteria) that a license amendment request for relocation of TS P/T limits and LTOP system limits into a PTLR must address. In addition, Section 5.0 of the NRC safety evaluation (SE) related to CE NPSD-683-A (Ref. 2) contains 26 action items that Licensees must address in their plant-specific submittals requesting a license amendment to relocate the P/T limits and LTOP system limits.

Disposition of all 26 original action items is documented herein. The discussion is organized according to the seven PTLR criteria.

PTLR Criterion 1 (1)

Describe the methodology used to calculate the neutron fluence values for the reactor vessel materials, including a description of whether or not the methodology is consistent with the guidance of Draft Regulatory Guide DG-1053, a description of the computer codes used to calculate the neutron fluence values, and a description of how the computer codes for calculating the neutron fluence values were benchmarked.

Section 1.1 of WCAP-16835 (Attachment 5), the basis document for the PVNGS Units 1, 2, and 3 PTLR, confirms that the neutron fluence methodology is consistent with that of Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (previously issued as Draft Regulatory Guide DG-1053).

The codes and methodology used to calculate the neutron fluence for PVNGS are described in Section 1.2 of WCAP-16835. These include DORT Version 3.2, BUGLE-96 cross section library, and the FERRET code with SNLRML library.

Computer codes used to calculate the neutron fluence were benchmarked in accordance with Regulatory Guide 1.190 as described in Section 1.4 of WCAP-16835.

(2)

Provide the values of neutron fluence used for the [ART] calculations,

.including the values of neutron fluence for the inner [diameter] (ID), 1/4 T, and 3/4T locations of the RV.

Table 1-1 of WCAP-1 6835 lists the peak calculated neutron fluence values after 32 EFPY at the inner surface, 1/4T and 3/4T locations for PVNGS Units 1, 2, and 3. A 6

Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR summary of the limiting adjusted reference temperatures through 32 EFPY is given in Table 4-1 of WCAP-1 6835 and is based on the design neutron fluence value described in Section 4.2.1 of WCAP-16835.

PTLR Criterion 2 (3)

Either provide the surveillance capsule withdrawal schedule in the proposed PTLR for the amendment, or reference in the PTLR, by title and number, the documents in which the withdrawal schedule is located.

The surveillance capsule withdrawal schedules are provided in the PVNGS UFSAR Section 5.3, Tables 5.3-13, 14, 15, 18, 19, and 19A, and are summarized in Table 2-4 of WCAP-16835 (Attachment 5). Section TA4.0 of the proposed PVNGS Units 1, 2 and 3 PTLR (Attachment 4) identifies the UFSAR tables that describe the RV surveillance program and surveillance capsule withdrawal schedules for PVNGS Units 1, 2 and 3.

(4)

Reference the surveillance capsule reports by title and number if the ART values are calculated using [RV] surveillance data.

The reactor vessel surveillance capsule evaluation reports for PVNGS were submitted to the NRC in the following documents:

APS letter no. 102-05242 to NRC, Palo Verde Nuclear Generating Station Unit 1 Reactor Vessel Material Surveillance Capsule at 230',' April 5, 2005 (ADAMS Accession No. not publicly available) (transmittal of WCAP-16374-NP, "Analysis of Capsule 2300 from Arizona Public Service Company Palo Verde Unit 1 Reactor Vessel Radiation Surveillance Program," February 2005) (Ref. 8).

0 APS letter no. 102-05457 to NRC, Palo Verde Nuclear Generating Station Unit 2 Reactor Vessel Material Surveillance Capsule at 2300,' April 4, 2006 (ADAMS Accession No. ML061040586) (transmittal of WCAP-16524-NP, "Analysis of Capsule 2300 from Arizona Public Service Company Palo Verde Unit 2 Reactor Vessel Radiation Surveillance Program," February 2006) (Ref. 9).

0 APS letter no. 102-05348 to NRC, Palo Verde Nuclear Generating Station Unit 3 Analysis of Reactor Vessel Material Surveillance Capsule at 230', September 26, 2005 (ADAMS Accession No. ML053390139) (transmittal of WCAP-16449-NP, "Analysis of Capsule 230' from Arizona Public Service Company Palo Verde Unit 3 Reactor Vessel Radiation Surveillance Program," August 2005) (Ref. 10).

These surveillance capsule evaluation reports are referenced in Section TA4.0 of the proposed PVNGS Units 1, 2 and 3 PTLR (Attachment 4).

7 Evaluation of the Proposed TS Change Relocate P/T. Limits to PTLR PTLR Criterion 3 (5)

Provide a description of the analytical method used in the energy addition transient analysis.

The reactor coolant system (RCS) energy addition transient methodology is described in Section 3.2.1.3 of WCAP-1 6835 (Attachment 5).

(6)

Provide a description of the analytical method used in the mass addition transient analysis, if different from that in Section 3.3.5.of the TR.

Section 3.2.1.2 of WCAP-1 6835 (Attachment 5) describes the RCS mass addition transient methodology. The analytical methods used are consistent with those of Section 3.3.5 of CE NPSD-683-A, Revision 06.

(7)

Provide a description of the method for selection of relief valve setpoints.

Relief valve overpressure protection is described in Section 3.2.1.1 of WCAP-16835 (Attachment 5).

(8)

Provide a justification for use of subcooled water conditions or a steam volume in the pressurizer.

The pressurizer is assumed to be water-solid with no credit for a cover gas or steam space at the initiation of either the mass addition or energy addition transients as described in Section 3.2.1 of WCAP-1 6835 (Attachment 5).

(9)

Provide a justification for a less conservative method for determination of decay heat contribution if the method used is less conservative than the "most conservative method" described in the TR.

The method for determining the decay heat contribution is consistent with that described in CE NPSD-683-A as stated in Section 3.2.1.3 of WCAP-16835 (Attachment 5).

(10)

Provide justification for operator action time used in transient mitigation or termination.

Transient analyses that support PVNGS Units 1, 2, and 3 do not credit operator action for mitigation or termination of the energy or mass addition transients as stated in Section 3.2.1 of WCAP-1 6835.

(11)

Provide correlations used for developing Power Operated Relief Valve (PORV) discharge characteristics.

8 Evaluation of the Proposed TS Change Relocate PIT Limits to PTLR Low temperature overpressure protection (LTOP) transient analyses do not take credit.

for pressurizer PORVs since PORVs are not installed at PVNGS.

(12)

Provide spring relief valve discharge characteristics if different from those described in the TR or if the peak transient pressure is above the set pressure of the valve plus 10 percent.

Section 3.2.1.1 of WCAP-1 6835 (Attachment 5) describes the PVNGS relief valve discharge characteristics.

(13)

Provide a description of how the reactor coolant temperature instrumentation uncertainty was accounted for.

Pressure and temperature instrument uncertainties are described in Section 5.13 of WCAP-1 6835 (Attachment 5).

(14)

Provide a justification for the mass and energy addition transient mitigation which credit presence of nitrogen in the pressurizer.

LTOP transient analyses do not credit the presence of nitrogen in the pressurizer at PVNGS.

(15)

Identify and explain any other deviation from the methodology included in Section 3. 0 of the TR.

The LTOP methodology in Section 3.0 of WCAP-16835 is consistent with that described in CE NPSD-683-A. LTOP heatup and cooldown rate limits are given in Table 3-1 of WCAP-1 6835.

PTLR Criterion 4 (16)

Identify the limiting materials and corresponding ART values for both the quarter-thickness (1/4T) and three-quarter-thickness (3/4T) locations of the R. V. shell.

Table 4-1 of WCAP-16835 (Attachment 5) identifies the limiting materials and provides corresponding ART and RTPTS values through 32 EFPY for PVNGS Units 1, 2, and 3.

Tables 4-5 through 4-7 of WCAP-16835 list predicted adjusted reference temperatures for each of the reactor vessel beltline materials surrounding the active core in PVNGS Units 1,2, and 3.

(17)

For [PWR] facilities, identify the limiting RTpTs values for the [RV] as calculated in accordance with the methods and criteria of 10 CFR 50.61.

9 Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR The limiting materials and corresponding RTPTS values calculated in accordance with 10 CFR 50.61 are listed in Table 4-1 of WCAP-1 6835 (Attachment 5). Table TA5-1 in the proposed PVNGS PTLR (Attachment 4) provides a summary of the ART and RTpTs values.

PTLR Criterion 5 (18)

Ensure that the ferritic RV materials that have accumulated neutron fluences in excess of 1. OE+ 17 n/cm2 (E > I Me V) will be assessed according to Section 4.0 of the CE Topical Report NPSD-683, Revision 6, regardless of whether the materials are located within the region immediately surrounding the active core.

Each ferritic plate and weld material located within the region immediately surrounding the active core of PVNGS Units 1, 2, and 3 was evaluated to identify the limiting material at the 1/4T and 3/4T locations. Adjusted RTNDT values for the limiting materials were used to establish the heatup and cooldown limits. Evaluation results are provided in Tables 4-5 through 4-7 of WCAP-1 6835 (Attachment 5). Each of the limiting materials is projected to receive sufficient neutron fluence to warrant consideration in this evaluation.

Ferritic plate and weld materials located above and below the region immediately surrounding the active core receive significantly lower neutron fluence relative to the active core region. The initial RTNDT values for these ferritic materials are comparable to the initial RTNDT values for the materials assessed in Tables 4-5 through 4-7 of WCAP-16835 (Attachment 5). The adjustment to RTNDT will be smaller than for the materials surrounding the active core because the neutron fluence is significantly lower.

Therefore, the ferritic plate and weld materials located above and below the region immediately surrounding the active core at PVNGS do not become limiting. Each of the intermediate and lower shell course plates and welds are assessed in Tables 4-5 through 4-7 of WCAP-1 6835 (Attachment 5).

The initial RTNDT values for the ferritic plate and weld materials located above and below the region immediately surrounding the active core are used in conjunction with values for the other ferritic materials in the primary coolant pressure boundary to establish other aspects of the heatup and cooldown limits, including the bolt-up temperature, the lowest service temperature, and the flange limits.

(19)

Identify which method (i.e., Kic [static plane strain fracture toughness] or KIA [dynamic/crack arrest fracture toughness]) will be used to calculate the reference stress intensity factor (KIR) values for the RV as a function of temperature.

10 Evaluation of the Proposed TS Change Relocate PIT Limits to PTLR The reference stress intensity factor used for PVNGS Units 1, 2, and 3 is the equation for K1c given in Appendix G of the ASME Code Section XI Division 1, July 1999.

(20)

If ASME Code Case N-640 and Kic are being used as the basis for calculating the KIR reference fracture toughness values, submit an exemption request [pursuant to the alternative program provisions of 10 CFR 50.60(b)] to use the methods of ASME Code Case N-640 and apply them to the P-T limit calculations.

The 2000 Edition of the ASME Code is consistent with the determination of the pressure-temperature curves. This Code Edition incorporates KIc criterion for the allowable fracture toughness. The staff has approved use of this criterion in the Code of Federal Regulations Title 10, Part 50.55a Codes and Standards, adopted in Section 1.0 Background, September 26, 2002. Therefore, Code Case N-640 is not used and an exemption request is not required.

(21)

(Applicable only if the CE NSSS methods for calculating KIM [applied stress intensity, K, due to pressure loading] and KIT [applied K due to thermal loading] factors, as stated in Section 5.4 of CE NPSD-683, Revision 6, are being used as the basis for generating the P-T limits for their facilities). Apply for an exemption against requirements of Section IV.A.2. of Appendix G to Part 50 to apply the CE NSSS methods to their P-T curves. This is consistent with the "note" on page 5-15 of CE NPSD-683, Revision 6. Exemption requests to apply the CE NSSS to the generation of P-T limit curves should be submitted pursuant to the provision of 10 CFR 50.60(b) and will be evaluated on a case-by-case basis against the exemption request acceptance criteria of 10 CFR 50.12.

Westinghouse PTLR methodology for CE NSSS plants uses influence coefficients derived from finite element methods to calculate KIT, the thermal stress intensity factor and KIM, the stress intensity factors due to internal pressure loading as documented in CE NPSD-683-A, Revision 06.

Provided in Enclosure 2 in this submittal is an application for the exemption specified in action item (21) from the requirements of 10 CFR Part 50, Appendix G. The requested exemption would allow for the application of the CE NSSS methods of CE NPSD-683-A, Revision 06 for calculating KIM values to the calculation of P/T limits, in lieu of the methodology cited in the ASME Code, Section Xl, Appendix G. This exemption request is being submitted pursuant to the provisions of 10 CFR 50.60(b) and 10 CFR 50.12.

(22)

Include in their PTLRs the P-T limit curves for heatup, cooldown, criticality, and hydrostatic and leak testing of their reactors.

11 Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR RCS composite pressure-temperature curves for boltup, heatup, cooldown, hydrostatic and leak tests at PVNGS are shown on Figures 5-3 and 5-4 of WCAP-1 6835 (Attachment 5). Criticality limits remain governed by PVNGS Technical Specification 3.4.2, "RCS Minimum Temperature for Criticality," and are not applicable in Modes 3 through 6, thus, the core critical limits for PVNGS are listed in Tables 5-5 and 5-6 of WCAP-16835 but are omitted from Figures 5-3 and 5-4.

The P/T limit data for heatup and cooldown operations and hydrostatic testing are provided in Tables TA2-1 through TA2-4 of the proposed PVNGS Units 1, 2, and 3 PTLR (Attachment 4). The P/T limit curves corresponding to these data points are provided in Figures TA2-1 and TA2-2 of the proposed PVNGS Units 1, 2, and 3 PTLR.

PTLR Criterion 6 (23)

Demonstrate how the P-T curves for pressure testing conditions and normal operations with the core critical and not-critical will be in compliance with the appropriate minimum temperature requirements as given in Table 1 to Appendix G to Part 50.

Minimum temperature requirements have been incorporated into the final composite limit curves shown in Figures TA2-1 and TA2-2 of the proposed PVNGS Units 1, 2, and 3 PTLR (Attachment 4) and Figures 5-3 and 5-4 of WCAP-1 6835 (Attachment 5).

These figures demonstrate that the pressure-temperature curves comply with the requirements of Appendix G to 10 CFR 50' PTLR Criterion 7 (24)

Include in their PTLRs the supplemental surveillance data and calculations of the chemistry factors if surveillance data are used for the calculations of the adjusted reference temperatures.

Chemistry factors are determined following Regulatory Position 1.1 of Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," as part of the adjusted reference temperature calculation and are discussed in Section 4 of WCAP-1 6835 (Attachment 5). Chemistry factor determination based on analyses of the surveillance capsule data following Regulatory Position 2.1 of RG 1.99, Revision 2, is discussed in Section 7 of WCAP-16835. Surveillance capsule data are evaluated but not used for the calculations of the adjusted reference temperatures at PVNGS.

Section TA6.3 of the proposed PVNGS Units 1, 2, and 3 PTLR (Attachment 4) contains the derivation of the chemistry factors from surveillance data.

(25) Provide the evaluation of whether the surveillance data are credible in accordance with the credibility criteria of RG 1.99, Rev. 2.

12 Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR The credibility of surveillance data for PVNGS Units 1, 2, and 3 is established in Section 7 and the chemistry factor determination is detailed in Tables 7-4 through 7-6 of WCAP-16835 (Attachment 5). PVNGS surveillance data were predictable based on the derived chemistry factors using Regulatory Position 2.1 of RG 1.99, Revision 2.

Section TA6.2 of the proposed PVNGS PTLR provides the evaluation results showing that the surveillance data are credible in accordance with the credibility criteria of RG 1.99, Revision 2.

(26)

In addition, if licensees seek to use surveillance data from supplemental plant sources, licensees must:

(a)

Identify the source(s) of the data.

(b)

Either identify by title and number the SE report that approved the use of the supplement data, along with a justification of why the data is applicable; or compare the licensee's data with the data from the supplemental plants(s) for both the radiation-environments (i.e., neutron spectrums and irradiation temperatures), and the surveillance test results, and pursuant to Section III. C of Appendix H to Part 50, submit the proposed integrated surveillance program and evaluation of the data to the NRC for review and approval.

Supplemental plant sources were not used in the determination of the chemistry factor at PVNGS.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Req uirements/Criteria 4.1.1 Title 10 CFR Part 50 Requirements for Generating P/T Limits and LTOP System Limits for Pressurized-Water Reactors (PWRs)

The NRC established requirements in 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," in order to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. Title 10 CFR Part 50, Appendix G, requires that the P/T limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated using the methods of Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code). For conditions with the core critical, P/T limits must be more conservative than the ASME Code, Section Xl, Appendix G limits. Table 1 of 10 CFR Part 50, Appendix G, provides a summary of the requirements for P/T limits relative to the ASME Code,Section XI, Appendix G criteria, as well as the minimum temperature requirements for bolting up the reactor vessel (RV) during normal and pressure testing operations. Title 10 CFR Part 50, Appendix G, also requires that

-13 Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR applicable surveillance data from RV material surveillance programs be incorporated' into the calculations of plant-specific P/T limits and that the P/T limits for operating reactors be generated using a method that accounts fdce the effects of neutron irradiation on the RCPB. The rule also establishes conservative requirements for determining the temperature and pressure setpoints for LTOP systems. P/T limits and LTOP system limits are subject to General Design Criteria (GDC) 14, "Reactor coolant pressure boundary," GDC 15, "Reactor coolant system design," GDC 30, "Quality of reactor coolant pressure boundary," and GDC 31, "Fracture prevention of reactor coolant pressure boundary," in 10 CFR Part 50, Appendix A.

Title 10 CFR Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," provides the NRC's criteria for the design and implementation of RV material surveillance programs for operating light-water reactors. The NRC's requirements for protecting the RVs of PWRs against pressurized thermal shock (PTS) events are given in 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events."

NRC regulatory guidance related to determining the change in RV material parameters and P/T limit curves due to the effects of radiation embrittlement is found in Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."

NRC guidance related to the review of P/T limit curves and PWR PTS criteria is found in Standard Review Plan (SRP) Section 5.3.2, "Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock." NRC guidance related to the review of LTOP system limits is found in SRP Section 5.2.2, "Overpressure Protection."

The regulatory requirements for RV fluence calculations are specified in GDC 14, 30, and 31 of 10 CFR Part 50, Appendix A. In March 2001, the NRC issued RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." Fluence calculations are acceptable if they are done with approved methodologies or with methods that are shown to conform to the guidance in RG 1.190.

4.1.2 Technical Specification Requirements for P/T Limits and LTOP System Limits Section 182a of the Atomic Energy Act of 1954 (Title 42 USC Section 2232) requires applicants for nuclear power plant operating licenses to include TSs as part of the operating license. The Commission's regulatory requirements related to the content of the TS are set forth in 10 CFR 50.36. That regulation requires that the TS include items in five specific categories: (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions of operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls.

Paragraph 50.36(d)(2)(ii) of 10 CFR requires that LCOs be established for the P/T limits and LTOP system limits because the parameters fall within the scope of Criterion 2 identified in the rule:

14 Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The P/T limits and LTOP system limits for PWR-designed light-water reactors fall within the scope of Criterion 2 stated above, and are, therefore, ordinarily required to be included within the TS LCOs for a plant-specific facility operating license (FOL).

On January 31, 1996, the staff issued Generic Letter (GL) 96-03, "Relocation of the Pressure-Temperature Limit Curves and Low Temperature Overpressure Protection System Limits" (Ref. 1) to inform licensees that they may request a license amendment to relocate the actual P/T limit curves and/or LTOP system limit values from the TS LCOs into a PTLR or other licensee-controlled document that would be administratively controlled through the administrative controls section of the TS. In order to permit relocation of the P/T limits and LTOP system limits, GL 96-03 indicated that licensees seeking to relocate P/T limits and LTOP system limits for their reactors would need to generate their P/T limits and LTOP system limits in accordance with an NRC-approved methodology and that the methodology used to generate the P/T limits and LTOP system limits would need to comply with the requirements of 10 CFR Part 50, Appendices G and H. Furthermore, the methodology used to generate the P/T limits and LTOP system limits would need to be incorporated by reference in the administrative controls section of the TS. The GL also stipulated that the TS administrative controls section for the PTLR would need to reference the staff's safety evaluation (SE) issued on the PTLR methodology and that the PTLR be defined in Section 1.0 of the TS. Attachment I to GL 96-03 provided a list of the criteria that the approved PTLR methodology and plant-specific PTLR license amendment application would be required to meet.

Technical Specification Task Force (TSTF) Traveler No. TSTF-408 (Ref. 3) amended the CE Standard Technical Specifications (STS) (NUREG-1432) to: (1) delete references to the TS LCO specifications for the P/T limits and LTOP system limits in the TS definition of the PTLR, and (2) revise STS 5.6.6 to identify, by number and title, the NRC-approved Topical Reports that document PTLR methodologies, or the NRC SE for a plant-specific methodology by NRC letter and date. A requirement was added to the reviewers note to specify the complete citation of the PTLR methodology in the plant-specific PTLR, including the report number, title, revision, date, and any supplements.

Only the figures, values, and parameters associated with the P/T limits and LTOP system limits are relocated to the PTLR. The methodology for their development must be reviewed and approved by the NRC. TSTF-408 did not change the requirements associated with the review and approval of the methodology or the requirement to operate within the limits specified in the PTLR. Any changes to a methodology that had not been approved by the staff would continue to require staff review and approval pursuant to the license amendment request provisions and requirements of 10 CFR 50.90, "Application for amendment of license or construction permit."

1.5 Evaluation of the Proposed TS Change Relocate PIT Limits to PTLR 4.2 Precedent Palo Verde Nuclear Generating Station Units 1, 2, and 3 and San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 were designed by Combustion Engineering.

The reactor vessel geometry, weld and plate materials.for both the PVNGS and SONGS units are similar, as is the methodology used to establish the RCS P/T and LTOP limits for these units.

Southern California Edison (SCE) submitted a license amendment request to the NRC by letter dated January 28, 2005 (Ref. 5), and supplemented this request by letter dated January 12, 2006 (Ref. 6), for SONGS Units 2 and 3 to relocate the RCS P/T limits and LTOP limits from the TSs to a licensee-controlled PTLR. The NRC approved the SONGS Operating License amendments in a letter dated July 13, 2006 (Ref. 7).

In their January 12, 2006, submittal, SCE identified and provided responses to nine NRC requests for additional information (RAIs) related to the PTLR amendment request. Provided in Attachment 6'are APS responses to the nine RAIs.

4.3 No Significant Hazards Consideration Determination The proposed Operating License amendment would revise Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 Technical Specifications (TSs) 1.1, "Definitions," 3.4, "Reactor Coolant System (RCS)," and 5.6, "Reporting Requirements,"

to relocate the RCS pressure-temperature curves and limits and the low temperature overpressure protection temperatures from the TSs to a licensee-controlled document identified as the Pressure and Temperature Limit Report (PTLR) within the Technical Requirements Manual (TRM). The proposed amendment to TS 5.6 would require that the analytical methods used to determine the RCS pressure and temperature limits in the PTLR shall be those described in the NRC-approved Topical Report CE NPSD-683-A.

Arizonra Public service Company (APS) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

This proposed change revises the Technical Specifications by relocating the reactor coolant'system (RCS) pressure and temperature limits, heatup and cooldown curves and low temperature overpressure protection (LTOP) enable temperatures from the Technical Specifications to an APS-controlled RCS 16 Evaluation of the Proposed TS Change Relocate PIT Limits to PTLR Pressure and Temperature Limits Report (PTLR), and requiring that the limits in the PTLR be determined using the analytical methods described in the NRC-approved Topical Report CE NPSD-683-A. Relocation of this information and updating it using NRC-approved methodology will not alter the requirement to update the RCS pressure and temperature curves and limits in accordance with 10 CFR 50 Appendices G and H. Updating the P/T curves and LTOP limits ensures the reactor coolant system's pressure boundary integrity is protected throughout plant life. Consequently, this proposed change is determined to not contribute to an increase in the probability of, or the initiation of, a design basis accident. Similarly, the safety analysis information presented in the Updated Final Safety Analysis Report remains unchanged.

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

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

Response: No.

The proposed change revises the Technical Specifications by relocating the RCS pressure and temperature limits, heatup and cooldown curves and LTOP enable temperatures from the Technical Specifications to a PVNGS PTLR, and requiring that the limits in the PTLR be determined using the analytical methods described in the NRC-approved Topical Report CE NPSD-683-A. The PTLR documents removal, testing and analyzing the surveillance capsules, and will be updated by APS to reflect the results of testing and analysis of surveillance specimens withdrawn in the future. Removal, testing and analysis of surveillance specimens may result in a need to implement changes to the RCS pressure and temperature limits. Such Changes are implemented to ensure the integrity of the RCS pressure boundary throughout plant lifetime. Updates to the RCS pressure and temperature curves and limits will not create a new or different kind of accident. Relocating the P/T curves, heatup and cooldown rates and LTOP limits to the PTLR has no impact on any safety analyses.

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

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

Response: No.

Pressure and temperature curves and limits are provided as limits to plant operation to ensure RCS pressure boundary integrity is maintained throughout 17 Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR the plant's lifetime. Changes to the RCS pressure and temperature curves and limits, resulting from the removal, testing and analysis of surveillance capsules, are only made within the acceptable margin limits thereby maintaining the required margin of safety. There is no change to the safety analysis.

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

APS concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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

5.0 ENVIRONMENTAL CONSIDERATION

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 a significant increase in the amounts of any effluents 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.

6.0 REFERENCES

1.

NRC Generic Letter 96-03, "Relocation of the Pressure-Temperature Limits Curves and Low Temperature Overpressure Protection System Limits,"

January 31, 1996.

2.

Combustion Engineering Owners Group Topical Report CE NPSD-683-A, Revision 6, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications," April 2001 (ADAMS Accession No. ML011350387).

18 Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR

3.

Technical Specification Task Force (TSTF) Traveler number TSTF-408, Revision 1, "Relocation of LTOP Enable Temperature and PORV Lift Setting to the PTLR," May 2001 (ADAMS Accession No. ML011630261)

4.

NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants," Revision 3.1, December 2005 (ADAMS Accession No. ML062510040).

5.

Letter from Southern California Edison Company (SCE) to the NRC, "San Onofre Nuclear Generating Station Units 2 and 3, Docket Nos. 50-361 and 50-362, Proposed Change Number NPF-1 0/15-551, License Amendment Request,

'Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)'," January 28, 2005 (ADAMS Accession No. ML050320286).

6.

Letter from Southern California Edison Company (SCE) to the NRC, "San Onofre Nuclear Generating Station, Units 2 and 3, Docket Nos. 50-361 and 50-362, Proposed Change Number NPF-10/15-551, License Amendment Request,

'Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)'," January 12, 2006 (ADAMS Accession No. ML060190101).

7.

Letter from the NRC to Southern California Edison Company (SCE), "San Onofre Nuclear Generating Station, Units 2 And 3 - Issuance of Amendments Re:

Reactor Coolant System (RCS) Pressure And Temperature Limits Report (PTLR)

(TAC NOS. MC5773 AND MC5774)," July 13, 2006 (ADAMS Accession No. ML062170006).

8.

Letter no. 102-05242 from APS to the NRC, "Palo Verde Nuclear Generating Station Unit 1 Reactor Vessel Material Surveillance Capsule at 2300, April 5, 2005 (transmittal of WCAP-16374-NP, "Analysis of Capsule 2300 from Arizona Public Service Company Palo Verde Unit 1 Reactor Vessel Radiation Surveillance Program," February 2005) (ADAMS Accession No. not publicly available).

9.

Letter no. 102-05457 from APS to the NRC, "Palo Verde Nuclear Generating Station Unit 2 Reactor Vessel Material Surveillance Capsule at 2300, April 4, 2006 (transmittal of WCAP-1 6524-NP, "Analysis of Capsule 2300 from Arizona Public Service Company Palo Verde Unit 2 Reactor Vessel Radiation Surveillance Program," February 2006) (ADAMS Accession No. ML061040586).

10.

Letter no. 102-05348 from APS to the NRC, "Palo Verde Nuclear Generating Station Unit 3 Analysis of Reactor Vessel Material Surveillance Capsule at 2300,,

September 26, 2005 (transmittal of WCAP-1 6449-NP, "Analysis of Capsule 2300 from Arizona Public Service Company Palo Verde Unit 3 Reactor Vessel Radiation Surveillance Program," August 2005) (ADAMS Accession No.

MIL053390139).

19 Evaluation of the Proposed TS Change Relocate P-T Limits to PTLR ENCLOSURE 1, ATTACHMENT 1 Technical Specification Page Markups Pages:

1.1-6 3.4.3-1 3.4.3-2 3.4.3-3 3.4.3-4 3.4.3-5 3.4.3-6 3.4.3-7 3.4.6-1 3.4.7-1 3.4.11-1 3.4.13-1 5.6-6 5.6-6 INSERT

Definitions 1.1 1.1 Definitions (continued)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

REACTOR PROTECTIVE SYSTEM (RPS)

RESPONSE

TIME SHUTDOWN MARGIN (SDM)

The PTLR is the site specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period.

These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.9.

.RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3876 MWt for Unit 1 through operating cycle 12 and Unit 3 through operating cycle 13, and 3990 MWt for Unit 1 after operating cycle 12, Unit 2, and Unit 3 after operating cycle 13.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until electrical power to the CEAs drive mechanism is interrupted.

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.

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a.

All full strength CEAs (shutdown and regulating) are fully inserted except for the single CEA of highest reactivity worth, which is assumed to be fully withdrawn.

With any full strength CEAs not capable of being fully inserted, the withdrawn reactivity worth of these CEAs must be accounted for in the determination of SDM and

b.

There is no change in part length or part strength CEA position.

PALO VERDE UNITS 1,2,3 1.1-6 AMENDMENT NO. 415ý, 157

RCS P/T Limits 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.3 RCS pressure, RCS temperature, and RCS heatup and cooldown rates shal I be maintained within the limits specified in the PTLR. limited in acc.rdance with thc limits show in 4 2

3...*

2 4 ' 2 ý d-. inA. h' ;t-p nnnl*.

An 4-4 -

4--- 4-4 I P4F W

1-1 i i '

i k MR M!

i.

[....

!....i.

H Wi th.

-1

.1

a.

Maximum hcatup and cooldow-n spec4ied in Table 3.*.3 1.

1)017S RII I.41~~

4W4I l

.4 1 4.14 I

III i.AiIJ

4.

1131.41 I31..l 1414 1.41.41 lll~

operations.

APPLICABILITY:

At all times; except when reactor vessel head is fully detensioned such that the RCS Cannot be pressurized.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE-------- A.1 Restore parameter(s) 30 minutes Required Action A.2 to within limits.

shall be completed whenever this AND Condition is entered.

A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acceptable for Requirements of LCO continued operation.

not met in MODE 1, 2, 3, or 4.

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 with 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months RCS pressure

< 500 psia.

(continued)

PALO VERDE UNITS 1,2,3 3.4.3-1 AMENDMENT NO.

117

RCS P/T Limits 3.4,3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

NOTE-------- C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed to within limits.

whenever this Condition is entered.

AND C.2 Determine RCS is Prior to Requirements of LCO acceptable for entering MODE 4 not met any time in continued operation.

other than MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 NOTE----------------

Only required to be performed during RCS.

heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and 30 minutes RCS heatup and cooldown rates within limits specified in the PTLR. Tab!ee 3.-4.3--

1, and Fiu4 3.31ad..3 2.

PALO VERDE UNITS 1,2,3 3.4.3-2 AMENDMENT NO,'117

RCS P/T Limits 3.4.3 DELETE THIS TABLE TABLE 3.A.3 1 Maxim um Al ý.a...

e Lin÷ an. d rgeAE4 d... ni÷-n 89 E~ffcti).cp Full Powr~q-Yc;;rs Heatup

~~1Rate 200 /HR4 1<8oF 1F t*oO/WIR 181° 230°F 50°QF'/HR

-]RIO 75" Fo I_

Ra-te--(O2FH See Figure 3.-.3 3 lF ll-l4 100 o F"/HR l15OF 1,41o 24notLHR

>,,l D,-R-10V00F/4R E"..1 1 D-

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~2160 F750PF/HR Cool1 ownn 1o',p See Figr 2 3.

'1 1090 126 *F 10 Fl/HR 127 0F 147OF F-49no/4R 148°F 1624F 40°F/HR 1620F 1000 F'oR/

II T'A I '...LA AAJ

~J'.J I 1.4 L.I.>~

I I.AII~.J1...I 1.4 L.LAI 1..

PALO VERDE UNITS 1,2,3 3.4.3-3 AMENDMENT NO.

4-I--I7 158

RCS P/T Limits 3.4.3 DELETE THIS FIGURE Figurce3.4.3 1 Reactor Coolant SystmPcsr/cprtr L MitationO fo LP Tha4 81 A

[ ffPccti Y&

Full PDncrpr Years of npp-ation INDICATED RCS TEMPERATURE (OF)

Tc PALO VERDE UNITS 1,2,3 3.4.3-4 Amendment No. 117

RCS P/T Limits 3.4.3 DELETE THIS FIGURE F igure 3.4.3 2 Reactor Coolant SYSetm Dr-iTessure,'/mr'-r-Limitations for 8 to 32 Effective Full Pewcr Ycars of Operation 2 20,5 LOWEST JKERýVICE 0P.

1 0OF 0

1500 Ucc (n

L.

1000 1

0-,00 3"::

HEATUP z

INDICATED RCS TEMPERATURE ((F)

PALO VERDE UNITS 1,2,3 3.4.3-5 Amendment No. 117

RCS P/T Limits 3.4.3 DELETE THIS FIGURE Figure 3.4.3 3 Maximvm AI lcwIable Cogold*ow Rates S8E IlPY 10 II L-0 LU i-80 I

]-----------

7 -----

i el L.

........ *............ i......

4-----

A4CEýTAOLE:

K 90 93 T--

- INDICATED RCS TEMPERATURE (oF)

PALO VERDE UNITS 1,2,3 3.4.3-6 Amendment No. 117

RCS P/T Limits 3.4.3 DELETE THIS FIGURE Figure 3.4.34 Maximum Allowable Goodown Rate--&

8 32 EFPY

%3 I

N 20 0

c~10

..........~.6....* 6 4..4.....

,6

-*... 9

...... S Ip

j

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6.*

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..... 6...

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+

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~L....

._L _.L.L

_'-.L -.

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I

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6 6,,.6 6

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6.

.2 I

i 66-16 i

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6 9

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6 S

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n 80 90 100 I

ICATED RCS TEMPERATURE, 108 0

I PALO VERDE UNITS 1,2,3 3.4.3-7 Amendment No. 117

RCS Loops-MODE 4 3.4.6 3,4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops -

MODE 4 LCO 3.4.6 Two loops or trains consisting of any combination of RCS loops and shutdown cooling (SDC) trains shall be OPERABLE and at least one loop or train shall be in operation.

NOTES

1.

All reactor coolant pumps (RCPs) and SDC pumps 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, provided:

a.

No operations are permitted that would cause reduction of the RCS boron concentration; and

b. Core outlet temperature is maintained at least 10OF below saturation temperature.

2.

No RCP shall be started with any RCS cold leg temperature less than or equal to the LTOP enable temperature specified in the PTLR

-214'° durin coold.wn, op ! 291oF dbrpi h

,up unless the secondary side water temperature in each Steam Generator (SG) is < 1000F above each of the RCS cold leg temperatures.

3.

No more than 2 RCPs may be in operation with RCS cold leg temperature

2000 F.

No more than 3 RCPs may be in operation with RCS Cold leg temperature > 200°F but

5000 F.

APPLICABILITY:

MODE 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One required RCS loop A.1 Initiate action to Immediately inoperable, restore a second loop or train to OPERABLE AND status.

Two SDC trains inoperable.

(continued)

PALO VERDE UNITS 1,2,3 3.4.6-1 AMENDMENT NO. 117

RCS Loops -

MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops -

MODE 5, Loops Filled LCO 3.4.7 One Shutdown Cooling (SDC) train shall be OPERABLE and in operation, and either:

a.

One additional SDC train shall be OPERABLE; or

b.

The secondary side water level of each Steam Generator (SG) shall be Ž 25%.

*----NOTES --------------------

1.

The SDC pump of the train in operation 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 provided:

a.

No operations are permitted that would cause reduction of the RCS boron concentration:

and

b. Core outlet temperature is maintained at least 10OF below saturation temperature.

2. One required SDC train may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other SDC train is OPERABLE and in operation.

3.

No Reactor Coolant Pump (RCP) shall be started with one or more of the RCS cold leg temperatures less than or equal to the LTOP enable temperature specified in the PTLR

214°F during cooldown, or ! 2910 during h

,atup unless the secondary side water temperature in each SG is < 100°F above each of the RCS cold leg temperatures.

4.

No more than 2 RCPs may'be in operation with RCS cold leg temperature

2000 F.*

No more than 3 RCPs may be in operation with RCS cold leg temperature > 200'F but 5000 F.

5.

All SDC trains may be removed from operation'during planned heatup to MODE 4 when at least one RCS loop is in operation.

APPLICABILITY:

MODE 5 with RCS loops filled.

PALO VERDE UNITS 1,2,3

3.4.7-1 AMENDMENT NO. 117

Pressurizer Safety Valves-MODE 4 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 Pressurizer Safety Valves-MODE 4 LCO 3.4.11 APPLICABILITY:

One pressurizer safety valve shall be OPERABLE with a lift setting _> 2450.25 psia and _< 2549.25 psia.

MODE 4 with all RCS cold leg temperatures greater than the LTOP enable temperature specified in the PTLR. > 2!4°F duringq cooldown, o M~~fl4 t,

,~-

-1 fi rc,-.-.1 1,-,,

+'4~,,

0. O~CI 0

NOTE The lift settings are not required to be within LCO limits during MODES 3 and 4 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions.

This exception is allowed for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following entry into MODE 3 provided a preliminary cold setting was made prior to heatup.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.. All pressurizer safety A.1 Be in MODE 4 with Immediately valves inoperable.

one Shutdown Cooling.

System suction line relief valve in service.

AND A.2 Perform SR 3.4.11.2 Immediately and SR 3.4.11.3 for the required Shutdown Cooling System suction line relief valve to comply with Action AlI.

AND A.3 Be in MODE 4 with 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months any RCS cold leg temperatures less than or equal to the LTOP enable temperature specified in the PTLR. < 2!4°4 during coodoWn or

~291°F during PALO VERDE UNITS 1,2,3 3.4.11-1 AMENDMENT NO. 117

LTOP System 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 Low Temperature Overpressure Protection (LTOP) System LCO 3.4.13 An LTOP System shall be OPERABLE consisting of:

a.

Two OPERABLE Shutdown Cooling System suction valves with lift settings

467 psig aligned.

overpressure protection for the RCS; or line relief to provide

b.

The RCS depressurized and an RCS vent of 16 square inches.

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

No RCP shall be started unless the secondary side water temperature in each steam generator (SG) is

100'F above each of the RCS cold leg temperatures.

MODE 4 when any RCS cold leg temperature is less than or equal to the LTOP enable temperature specified in the PTLR. !* 214OF durin9olon APPLICABILITY:

MODE 5, MODE 6 when the reactor vessel head is on.

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

i,.k A-ri-inri rlT!.I1C f c1-c1A\\I Ci+-nt ttflWmtnV'tl

,r' conditi unt.l all rC cold lg tcmp...*e..t.....

h 0!

0 k

+ký t'r V

nnnli-k

-+ýI 11 D(C

-IA Inn

,k

1)

F

- Z

,,,i-,

2-- LCO 3.0.4.b is not applicable when entering MODE 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.1 Restore required 7 days A.

One required Shutdown Shutdown Cooling Cooling System suction System suction line line relief valve relief valve to inoperable in MODE 4.

OPERABLE status.

(continued)

PALO VERDE UNITS 1,2,3 3.4.13-1 AMENDMENT NO.

4-14-,

165

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 Areport 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.

5.6.9 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

ADD INSERT 5.6.9 PALO VERDE UNITS 1,2,3 5.6-6 AMENDMENT NO. 13-ý, 161

INSERT 5.6.9

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following Technical Specifications (TSs):

1.

TS 3.4.3, RCS Pressure and Temperature (P/T) Limits;

2.

TS 3.4.6, RCS Loops - Mode 4;

3.

TS 3.4.7, RCS Loops - Mode 5 Loops Filled;

4.

TS 3.4.11, Pressurizer Safety Valves - Mode 4; and

5.

TS 3.4.13, Low Temperature Overpressure Protection (LTOP)

System.

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

CE NPSD-683-A, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications."

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Evaluation of the Proposed TS Change Relocate P-T Limits to PTLR ENCLOSURE 1, ATTACHMENT 2 Technical Specification Bases Page Markups Pages:

B 3.4.3-1 B 3.4.3-2 B 3.4.3-3 (For information - no changes)

B 3.4.3-4 B 3.4.3-5 B 3.4.3-6 (For information - no changes)

B 3.4.3-7 B 3.4.3-8 B 3.4.6-3 B 3.4.6-5 B 3.4.7-3 B 3.4.7-7 B 3.4.11-4 B 3.4.13-4 B 3.4.13-5 B 3.4.13-7 B 3.4.13-8 B 3.4.13-11

RCS P/T Limits B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes.

These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips.

This LCO limits the pressure-and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

The Pressure and Temperature Limits Report (PTLR) contains P/T limit curves for heatup, cooldown, and inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature (Ref. 1).

Each P/T limit curve defines an acceptable region for normal operation.

The usual use of the curves is.operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB).

The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel.

The limits do not apply to the pressurizer, which has different design characteristics and operating functions.

10 CFR 50, Appendix G (Ref. 1 2), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials.

Reference 4 2 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.

It mandates the use of the ASME Code,Section III, Appendix G (Ref. 2 3).

The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3 4) and Appendix.H of 10 CFR 50 (Ref. 4 5).

The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 2 3.

(continued)

PALO VERDE UNITS 1,2,3 B 3.4.3-1 REVISION 0

RCS P/T Limits B 3.4.3 BASES BACKGROUND (continued)

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive.

At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit.

Across the span of the P/T limit curves, different locations are more restrictive, and, thus,. the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed.

The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The criticality limit includes the Reference 1 2 requirement that the limit be no less than 40°F above the heatup curve.

or the cooldown curve and not less than the minimum permissible temperature for inservice leak and hydrostatic (ISLH) testing.

However, the criticality limit is not operationally limiting; a more restrictive limit exists in LCO 3.4.2, "RCS Minimum Temperature for Criticality."

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident.

In the.event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.

The ASME Code,Section XI, Appendix E (Ref. - 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES

'(DBA) Analyses.

They are.prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition.

(continued)

PALO VERDE UNITS 1,2,3 B 3.4.3-2 REVISION 0

RCS P/T Limits B 3.4.3 NO CHANGES TO THIS PAGE BASES APPLICABLE Since the P/T limits are not derived from SAFETY ANALYSES any DBA, there are no acceptance limits related to the P/T (continued) limits.

Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

The RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The two elements of this LCO are:

a.

The limit curves for heatup, cooldown, and ISLH testing; and

b.

Limits on the rate of change of temperature.

The LCO limits apply to all components of the RCS, except the pressurizer.,

These limits define allowable operating regions and permit a large number of operating cycles while providing a wide margin to nonductile failure.

The limits for the rate of change of temperature control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and ISLH testing P/T limit curves.

Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal.

gradients and also ensures the validity of the P/T limit curves.

Violating the LCO limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCPB components.

The consequences depend on several factors, as follows:

a.

The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature;

b.

The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and (continued)

NO CHANGES TO THIS PAGE PALO VERDE UNITS 1,2,3 B 3.4.3-3 REVISION 0

RCS P/T Limits B 3.4.3 BASES LCO (continued)

c.

The existences, sizes, and orientations of flaws in the vessel material.

The limi curVcS (Figures 3.1.3 1, 344.

2, 3.1.3 3, and 3.4.3 '1) and the tabula;ted limits (Týable 3.1.3 1) onH the

-r-ate o-f c-hange of temperature do not account forP all instumet uceraint.

Idictedtemperatur-e limits, 20F greater and indicated pressure limis approximately 9 psi less tlhanp the-corres-ponding9 LC9O M

lmts inclGIude approepriate in-str.ument-uncer-t-ainty and ensure that the pr-essure&,,

temperature and rate of temper-ature czhange of the limiting RCS ~

~

i4 copnnt r 'tinteacullmits that proevide the r-equir-ed mar-gin to br-ittle failure These Ialues, Whic include approGpr-iate instrument ucrany r salse wi.thin the applicaable plant pro~eduires.

APPLICABILITY The RCS P/T limits Specification provides a definition of acceptable operation for prevention of nonductile failure in accordance with 10 CFR 50, Appendix G (Ref. 2 3).

Although the P/T limits were developed to provide guidance for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, their Applicability is at all times, except when reactor vessel head is fully detensioned such that the RCS cannot be pressurized, in keeping with the concern for nonductile failure.

The limits do not apply to the pressurizer.

During MODES 1 and 2, other Technical Specifications provide limits for operation that can be more restrictive than or can supplement these P/T limits.

LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits"; LCO 3.4.2, "RCS Minimum Temperature for Criticality"; and Safety Limit 2.1, "Safety Limits," also provide operational restrictions for pressure and temperature and maximum pressure.

Furthermore, MODES 1 and 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent.

The actions of this LCO consider the premise that a violation of the limits occurred during normal plant maneuvering.

Severe violations caused by abnormal transients, at times accompanied by equipment failures, may also require additional actions from emergency operating procedures.

(continued)

REVISION 2 PALO VERDE UNITS 1,2,3 B 3.4.3-4

RCS P/T Limits B 3.4.3 BASES ACTIONS A.1 and A.2 Operation outside the P/T limits must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.

The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range.

Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.

Besides restoring operation to within limits, an evaluation is required to determine if RCS operation can continue.

The evaluation must verify the RCPB integrity remains acceptable and must be completed before continuing operation.

Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.

ASME Code,Section XI, Appendix E (Ref. 5 6), may be used to support the evaluation.

However, its use is restricted to evaluation of the vessel beltline.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable to accomplish the evaluation.

The evaluation for a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections.

A favorable evaluation must be completed before continuing to operate.

Condition A is modified by a Note requiring Required Action A.2 to be completed whenever the Condition is entered.

The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.

Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

B.1 and B.2 If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because:

a.

The RCS remained in an unacceptable P/T region for an extended period of increased stress; or (continued)

PALO VERDE UNITS 1,2,3 B 3.4.3-5 REVISION 2

RCS P/T Limits B 3.4.3 NO CHANGES TO THIS PAGE BASES ACTIONS B.1 and B.2 (continued)

b.

A sufficiently severe event caused entry into an unacceptable region.

Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature.

With reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased.

Pressure and temperature are reduced by placing the plant in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 with RCS pressure

< 500 psia within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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

C.1 and C.2 The actions of this LCO, anytime other than in MODE 1, 2, 3, or 4, consider the premise that a violation of the limits occurred during normal plant maneuvering.

Severe violations caused by abnormal transients, at times accompanied by equipment failures, may also require additional actions from emergency operating procedures.

Operation outside the P/T limits must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.

The Completion Time of "immediately" reflects the urgency of restoring the parameters to within the analyzed range.

Most violations will not be severe, and the activity can be accomplished in a short period of time.in a controlled manner.

(continued)

NO CHANGES TO THIS PAGE PALO VERDE UNITS 1,2,3 B 3.4.3-6 REVISION 0

RCS P/T Limits B 3.4.3 BASES ACTIONS C.1 and C.2 (continued)

Besides restoring operation to within limits, an evaluation is required to determine if RCS operation can continue.

The evaluation must verify that the RCPB integrity remains acceptable and must be completed before continuing operation.

Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.

ASME Code,Section XI, Appendix E (Ref. - 6), may be used to support the evaluation.

However, its use is restricted to evaluation of the vessel beltline.

The Completion'Time of prior to entering MODE 4 forces the evaluation prior to entering a MODE where temperature and pressure can be significantly increased.

The evaluation for a mild violation is possible within several days, but more severe violations may require special, event specific stress analyses or inspections.

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered.

The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.

Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the PTLR limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes.

This Frequency is considered reasonable in view of the control room indication available to monitor RCS status.

Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time.

Surveillance for heatup, cooldown, or ISLH testingmay be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

(continued)

PALO VERDE UNITS 1,2,3 B 3.4.3-7 REVISION 0

RCS P/T Limits B 3.4.3 BASES SURVEILLANCE SR 3.4.3.1 (continued)

REQU I REMENTS This SR is modified by a Note that requires this SR be performed only during RCS system heatup, cooldown, and ISLH testing.

No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.

REFERENCES

1.

TRM Appendix TA, Reactor Coolant System Pressure and Temperature Limits Report (PTLR); (limits determined using methods described in Topical Report CE NPSD-683-A, Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications).

12. 10 CFR 50, *Appendix G.

2-

3.

ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.

3 4. ASTM E 185-82, July 1982.

4 5. 10 CFR 50, Appendix H.

- 6. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

6.

PVNGS Calcu-lation 13 N991 6.02 650 2.

PALO VERDE UNITS 1,2,3 B 3.4.3-8 REVISION 2

RCS Loops -

MODE 4 B 3.4.6 BASES LCO (continued)

Note 2 requires that secondary side water temperature in each SG is < 100°F above each of the RCS cold leg temperatures before an RCP may be started with any RCS cold leg temperature less than or equal to the LTOP enable temperature specified in the PTLR. Wotc 2 re.u. rres, that h*for.e an RP may be startcd wi th any RCS col1d leg tem~perature

2110F dur~ing coldn or< 2914F duPrin heatulp, that secondar-y si 1de PwqA t er tempera"ture (sa~turation temperature? corresponding to SG9 pre;ssure) in each SG is 100F aboVe e~ach of the RCS cold eg tmperture.

Th numrica val es o RCS cold leg temperaiture ait ',ihthis Note is.

applicable do not accoun AI At for all instrument uncer-tainty.

Use of an indica;ted value of 211 0F or below during coldoWn -and 29402-AP b&-ew duPRqn heatup ensures that the actu'al limiS will --e~t-beý exceede~d.

These values, whichý inclu1de approeprite instrment uncertainty, are established within the applicable plant proGe~durPes.

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 ! 2000 F, and no more than 3 RCPs with RCS cold leg temperature >200'F but

500 0 F.

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 OPERABLE and has the minimum water level specified in SR 3.4.6.2.

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 38

RCS Loops -

MODE 4 B 3.4.6 BASES SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that one required loop or train is in operation and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm.

This ensures forced flow is providing heat removal.

Verification includes flow rate, temperature, or pump status monitoring.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency has been shown by operating practice to be sufficient to regularly assess RCS loop status.

In addition, control room indication and alarms will normally indicate loop status.

SR 3.4.6.2 This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of secondary side water level in the required SG(s) Ž 25% wide range.

An adequate SG water level is required in order to have a heat sink for removal of the core decay heat from the reactor coolant.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess degradation and verify operation within safety analyses assumptions.

SR 3.4.6.3 Verification that the required pump is OPERABLE ensures that an additional RCS loop or SDC train can be placed in operation, if needed to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to the required pumps.

The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

REFERENCES

1.

PVNGS Operating License Amendments 52, 38 and 24 for Units 1, 2 and 3, respectively, and associated NRC Safety Evaluation dated July 25, 1990.

2.

PVNGS Calculation 13 NO90 6.02 652 2. Not used.

3.

PVNGS Calculation 13-JC-SH-0200, Section 2.9.

PALO VERDE UNITS 1,2,3 B 3.4.6-5 REVISION 6

RCS Loops -

MODE 5, Loops Filled B 3.4.7 BASES LCO in order to use the provisions of the Note allowing the (continued) pumps to be de-energized.

In this MODE, the SG(s) can be used as the backup for SDC heat removal.

To ensure their availability, the RCS loop flow path is to be maintained with subcooled liquid.

In MODE 5, it is sometimes necessary to stop all RCP or SDC forced circulation.

This is permitted to change operation from one SDC train to the other, perform surveillance or startup testing, perform the transition to and from the SDC, or to avoid operation below the RCP minimum net positive suction head limit.

The time period is acceptable because natural circulation is acceptable for decay heat removal the reactor coolant temperature can be maintained subcooled, and boron stratification affecting reactivity control is not expected.

Note 2 allows one SDC train to be inoperable for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided that the other SDC train is OPERABLE and in operation.

This permits periodic surveillance tests to be performed on the inoperable train during the only time when such testing is safe and possible.

Note 3 requires that secondary side water temperature in each SG is < 100°F above each of the RCS cold leg temperatures before an RCP may be started with any RCS cold leg temperature less than or equal to the LTOP enable temperature specified in the PTLR. Note 3 requ1 rH s that before an RCP may be utarted With any RCi cold

'lnceg tenty Ue < 21n 0i duitng a coodo7n.

or < 291'.

during a heatu-,n th 2econdary ieow water empertur-e (satur-ation temper-ature corresponding to SG pressure) i n each SG muost be < pre ablode each of to h

RCS cold leg temperatures.

Thep-nu cal n mr-ae fra 3 RCS w

cold leg temperature At 20i0° cuh ' this-ote appl ble do nt Saccutisforn alls ionditrumnt uncertainty.n Use ofanalyindcae valsuet of t7F oP helow dingud cesldwR aRd 29coF ort below duringQ heatup Hensue that the ac-tual jrlmts Will not be eedued Thesoalues, whicCh incluode appo instrui1--men-t uncerrtainty, areestablishedwithi applicable plant proe-dures.

Satisfying the above condition will preclude a low temperature overpressure event due t~o a thermal transient when the RCP is started.

Note 4 restricts RCP operation to no more than 2 RCPs with RCS cold leg temperature *ý 200 0F, and no more than 3 RCPs with RCS cold leg temperature > 200OF but *ý 500 9F.

Satisfying these conditions will maintain the analysis assumptions of the flow induced pressure correction factors due to RCP operation (Ref. 3).

(conti nued)

PALO VERDE UNITS 1,2,3 B 3. 4. 7-.3 REVISION 6

RCS Loops -

MODE 5, Loops Filled B 3.4.7 BASES (continued)

REFERENCES

1.

Not Used

2.

CE NPSD-770, GEN' PS-D 77 Analysis for Lower.Mode Functional Recovery Guidelines.

3.

PVNGS Operating License Amendments 52, 38, and 24 for Units 1, 2 and 3, respectively, and associated NRC Safety Evaluation dated July 25, 1990.

4 WJNI(ý CiAl cii

.ti on 13 1n49 6.02 2n2 2. Not used.

5.

PVNGS Calculation 13-JC-SH-0200, Section 2.9.

PALO VERDE UNITS 1,2,3.

B 3.4.7-7 REVISION 27

Pressurizer Safety Valves-MODE 4 B 3.4.11 BASES (continued)

ACTIONS A.1, A.2, and A.3 If all pressurizer safety valves are inoperable, the plant must be brought to a condition where overpressure protection is provided, then to a MODE in which the requirement does not apply.

To achieve this status, one Shutdown Cooling System suction line relief must be placed in service immediately, then the plant must be brought to at least MODE 4 with any RCS cold leg temperature; less than or equal to the LTOP enable temperature specified in the PTLR 4 2142,F diu*ring.cldown or !ý 2912F du1rig heatup within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , so that LCO 3.4.13 (LTOP System) would apply.

It is reasonable to pursue the ACTION to place a shutdown cooling system suction relief valve in service immediately (without delay) because the plant is already within the shutdown cooling system entry temperature of less than 350 0 F.

The Completion Time of immediately requires that the required action be pursued without delay and.in a controlled manner, and reflects the importance of maintaining the RCS overprotection system.

The 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months allowed to be in MODE 4 wi th any RCS temperature; less than or equal to the LTOP enable temperature specified in the PTLR.4 2,,42-cooldown o-r<_ 2910F durin~gjh u is reasonable, based on operating experience, to reach this condition without challenging plant systems.

For the Shutdown Cooling System suction line relief valve that is required to be in service in accordance with Required Action A.1, SR 3.4.11.2 and SR 3.4.11.3 must be performed or verified performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

This ensures that the required Shutdown Cooling System suction line relief valve is OPERABLE.

A Shutdown Cooling System suction line relief valveis OPERABLE when its isolation valves are open, its lift s.etpoint is set at 467 psig or less, and testing has proven its ability to open at that setpoi nt.

If the Required Actions and associated Completion Times are not met, overpressurization is possible.

The 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time to be in MODE 4 with any RCS cold leg temperature; less than or equal to the LTOP enable temperature specified in the PTLR _211A d1urFin*o-lNdoWn or 291°F during heatup places the unit in a condition where the LCO does not apply.

(continued)

PALO VERDE UNITS 1,2,3 B 3.4. 11-4 REVISION 0

LTOP System B 3.4.13 BASES (continued)

APPLICABLE SAFETY ANALYSES Safety analyses (Ref. 3) demonstrate that the reactor vessel is adequately protected against exceeding the Reference 1 P/T limits during shutdown.

In MODES 1, 2, and 3, and in MODE 4 with any RCS cold leg temperature greaterthan the LTOP enable temperature specified in the PTLR exedi-R 2;4oP dH,,9 0

A*1,.

hatp, the pressurizer.safety valves prevent RCS pressure from exceedi ng the Reference 1 1 i mi ts.

At the LTOP enable temperature specified in the PTLR and below about 2140F and be I GIA, R4 g E)E 1 nlAnP.n 2*QlO2 8nA4

-hPnI.* A-4r-in R9 hetA4P overpressure prevention falls to the OPERABLE Shutdown Cooling System suction line relief valves or to a depressurized RCS and a sufficient sized RCS vent.

Each of these means has a limited overpressure relief capability.

The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement.

Each time the P/T limit curves are revised, the LTOP System will be re-evaluated to ensure its functional requirements can still be satisfied using the Shutdown Cooling System suction line relief valve method or the depressurized and vented RCS condition.

Reference. 3 contains the acceptance limits that satisfy the LTOP requirements.

Any change to the RCS must be evaluated against these analyses to determine the impact of the change on the LTOP acceptance limits.

Transients that are capable of overpressurizing the RCS are categorized as either mass or heat input transients, examples of which follow:

Mass Input Type Transients

a.

Inadvertent safety injection: or

b.

Charging/letdown flow mismatch.

Heat Input Type Transients

a.

Inadvertent actuation of pressurizer heaters;

b.

Loss of shutdown cooling (SDC); or

c.

Reactor coolant pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.

(continued)

PALO VERDE UNITS 1,2,3 B 3.4.13-4 REVISION 0

LTOP System B 3.4.13 BASES APPLICABLE References 3, 7, 8 and 9 analyses demonstrate that either SAFETY ANALYSES one Shutdown Cooling System suction line relief valve or the (continued)

RCS vent can maintain RCS pressure below limits for the two most limiting analyzed events:

a.

The start of an idle RCP with secondary water temperature of the SG

100'F above RCS cold leg temperatures.

b.

An inadvertent SIAS with two HPSI pumps injecting into a water solid RCS, three charging pumps injecting, and letdown isolated.

Fracture mechanics analyses established the temperature of LTOP Appl i cabi 1 i ty at less than or equal to the LTOP enable temperature specified in the PTLR 214°F and below during Gnnldn,,R *a~

29I°E B= hpl~....

R9*n het,,.

Above these temperatures, the pressurizer safety valves provide the reactor vessel pressure protection.

The vessel materials were assumed to have a neutron irradiation accumulation equal to the effective full power years of operation specified in the PTLR 32 effecti ve full power years of operation.

The consequences of a small break Loss Of Coolant Accident (LOCA) in LTOP MODE 4 conform to 10 CFR 50.46 and 10 CFR 50, Appendix K (Refs. 4 and 5).

The fracture mechanics analyses show that the vessel is protected when the Shutdown Cooling System suction line relief valves are set to open at or below 467 psig.

The setpoint is derived by modeling the performance of the LTOP System, assuming the limiting allowed LTOP transient.

The Shutdown Cooling System suction line relief valves setpoints at or below the derived limit ensure the Reference 1 limits will be met.

The Shutdown Cooling System suction line relief valves setpoints will be re-evaluated for compliance when the revised P/T limits conflict with the LTOP analysis limits.

The P/T limits are periodically modified as the reactor vessel material toughness decreases due to embrittlement caused by neutron irradiation.

Revised P/T limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens.

The Bases for LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," discuss these examinations.

(continued)

PALO VERDE UNITS 1,2,3 B 3.4.13-5 REVISION 0

LTOP System B 3.4.13 BASES LCO Each of these methods of overpressure prevention is capable (continued) of mitigating the limiting LTOP transient.

The Note requires that, before an RCP may be started, the secondary side water temperature (saturation temperature corresponding to SG pressure) in each SG is

100°F above each of the RCS Cold leg temperatures.

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

APPLICABILITY This LCO is applicable in MODE 4 when-the temperature of any RCS col d 1 eg is less than or equal to the LTOP enable temperature specified in the PTLR 4 24°4OP during cooGdown or 291 0F during heatup, in MODE 5, and in MODE 6 when the reactor vessel head is on.

The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T 1 i mi ts above the LTOP enable temperature 2!4°F4 d-ung coolG.d.Wn

-and 29-1-271 during heatup. The requirements for overpressure protection in MODES 1, 2 and 3, and in MODE 4 above the LTOP System temperatures are covered by LCO 3.4.10, "Pressurizer Safety Valves - MODES 1, 2, and 3," and LCO 3.4.11, "Pressurizer Safety Valves - MODE 4." When the reactor vessel head is off overpressurization cannot occur.

The numerical. val. s forR.

C cod leg temper.at..ure at hich thi is.

applicable de dopent aco T

for al l

MinDtrume unw c eraintu Use of an i rdicated iVale of 217 0 m

ori beoW during cootdown ad 2Qn 0

PR below dursing heait end msures that the actua limis will not be exc-eeded.

These values, which~--

inld h proraeisrment unetitare established in the appropriate plant procedurPs.

LCD 3.4.3 provides the operational P/T limits for all MODES.

Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event.

The Appli*cabili*ti i mosRdified by a Note stating* w*en nh e cr*,

more cold legs reach 214 0 P, this IL 2erairns applicable during perids of steady stat-temperature coditions, until all RCS cold leg temperaturpes re.ac.h

.29....

.lso, if, c dn s terminated prior to reaching 21 (F

and a heatup is comm~fenced, this LCD is applicableP unilal RCS cold leg (continued)

PALO VERDE UNITS 1,2,3 B 3.4. 13-7 REVISION 2

LTOP System B 3.4.13 BASES (continued)

APPLICABILITY temperaturcs reac.

h 291 0F. This Note pro-.

Vides*

clar-ifiration (continued) about Appli*aýb4lt, intent.

Since PIINI'GSse tWo different tempratresat which the Shutdown Cooling System SUcto linel relie Mave mSt be Pl aced inR sepjrvic there is some possiility1., of confusion.

ThisNoe cIrlalrifiaes thos'ea circumstances whee~ the ShutdAow Coln Syst suctio line r e l i e f v a l v e s m u s t b e6 p l a c e d i R e..

A Note prohibits the application of LCO 3.0.4.b to an inoperable LTOP system.

There is an increased risk associated with entering MODE 4 from MODE 5 with LTOP inoperable and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of the risk assessment addressing inoperable the systems and components, should not be applied in this circumstance.

.ACTIONS A.1 In MODE 4 when any RCS cold leg temperature is lessthan or equal to the LTOP enable temperature specified in the PTLR _

214O' during c-P9 ld-Wn or !-"

291 0 P durng heatup with one Shutdown Cooling System suction line relief valve inoperable, two Shutdown Cooling System suction line relief valves must be restored to OPERABLE status within a Completion Time of 7 days.

Two valves are required to meet the LCO requirement and to provide low temperature overpressure mitigation while withstanding a single failure of an active component.

The Completion Time is based on the facts that only one Shutdown Cooling System suction line relief valve is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low.

B.1 The consequences of operational events that will overpressure the RCS are more severe at lower temperature (Ref. 6).

Thus, one required Shutdown Cooling System suction line relief valve inoperable in MODE 5 or in MODE 6 with the head on, the Completion Time to restore inoperable valve to OPERABLE status is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time to restore two Shutdown Cooling System suction line relief valves OPERABLE in MODE 5 or in (continued)

PALO VERDE UNITS 1,2,3 B 3.4.13-8 REVISION 42

LTOP System B 3.4.13 BASES REFERENCES (continued)

8.

Pressure Transient Analyses

a.

V-PSAC-009 (3876 MWt w/Original Steam Generators)

b.

MN725-00118 (Unit 2, 4070 MWt w/Replacement Steam Generators)

c.

MN725-00562 (Units 31, 4070 MWt w/Replacement Steam Generators)

9.

Mass Input Pressure Transient in Water Solid RCS

a.

V-PSAC-010 (3876 MWt w/Original Steam Generators)

b. MN725-00117 (Unit 2, 4070 MWt w/Replacement Steam Generators)

c. MN725-01495 (Units 31,4070 MWt w/Replacement Steam Generators)

10.

ASME, Boiler and Pressure Vessel Code,Section XI.

11.

13-COO-93-016, Sensitivity Study on Pressurizer Vent Paths vs. Days Post Shutdown.

10 D%/iNI(c rf i,-

1

ý --

10 hinifl a~ nlo a0 0 PALO VERDE UNITS 1,2,3 B 3.4.13-11 REVISION 42 Evaluation of the Proposed TS Change Relocate P-T Limits to PTLR ENCLOSURE 1, ATTACHMENT 3 Retyped Technical Specification Pages Pages:

1.1-6 3.4.3-1 3.4.3-2 3.4.6-1 3.4.7-1 3.4.11-1 3.4.13-1 5.6-6 5.6-7

Definitions 1.1 1.1 Definitions (continued)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED (RTP)

THERMAL POWER REACTOR PROTECTIVE SYSTEM (RPS)

RESPONSE

TIME SHUTDOWN MARGIN (SDM)

The PTLR is the site specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period.

These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.9.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3876 MWt for Unit 1 through operating cycle 12 and Unit 3 through operating cycle 13, and 3990 MWt for Unit 1 after operating cycle 12, Unit 2, and Unit 3 after operating cycle 13.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until electrical power to the CEAs drive mechanism is interrupted.

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.

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a.

All full strength CEAs (shutdown and regulating) are fully inserted except for the single CEA of highest reactivity worth, which is assumed to be fully withdrawn.

With any full strength CEAs not capable of being fully inserted, the withdrawn reactivity worth of these CEAs must be accounted for in the determination of SDM and

b.

There is no change in part length or part strength CEA position.

PALO VERDE UNITS 1,2,3 1.1-6 AMENDMENT NO. ;1-4ý,

RCS P/T Limits 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO

3.4.3 APPLICABILITY

RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR.

At all times; except when reactor vessel head is fully detensioned such that the RCS cannot be pressurized.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. --------- NOTE --------

A.1 Restore parameter(s) 30 minutes Required Action A.2 to within limits.

shall be completed whenever this AND Condition is entered.

A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acceptable for Requirements of LCO continued operation.

not met in MODE 1, 2, 3, or 4.

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 with 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months RCS pressure

< 500 psia.

(continued)

PALO VERDE UNITS 1,2,3 3.4.3-1 AMENDMENT NO. 44-7-,

RCS P/T Limits 3.4.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

NOTE --------

C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed to within limits.

whenever this Condition is entered.

AND C.2 Determine RCS is Prior to Requirements of LCO acceptable for entering MODE 4 not met any time in continued operation.

other than MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1

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

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and 30 minutes RCS heatup and cooldown rates within limits specified in the PTLR.

PALO VERDE UNITS 1,2,3 3.4.3-2 Amendment No. 4-14,

RCS Loops -

MODE 4 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops -

MODE 4 LCO 3.4.6 Two loops or trains consisting of any combination of RCS loops and shutdown cooling (SDC) trains shall be OPERABLE and at least one loop or train shall be in operation.

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

1..

All reactor coolant pumps (RCPs) and SDC pumps 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, provided:

a.

No operations are, permitted that would cause reduction of the RCS boron concentration; and

b. Core outlet temperature is maintained at least 10OF below saturation temperature.

2.

No RCP shall be started with any RCS cold leg temperature less than or equal to the LTOP enable temperature specified in the PTLR unless the secondary side water temperature in each Steam Generator (SG) is

< 100°F above each of the RCS cold leg temperatures.

3.

No more than 2 RCPs may be in operationwith RCS cold leg temperature

200 0F.

No more than 3 RCPs may be in operation with RCS cold leg temperature > 200°F but

! 5000 F.

APPLICABILITY:

MODE 4.

ACTIONS.

CONDITION REQUIRED ACTION COMPLETION'TIME A.

One required RCS loop A.1 Initiate action to Immediately inoperable, restore a second loop or train to OPERABLE AND status.

Two SDC trains inoperable.

(continued)

PALO VERDE UNITS 1,2,3 3.4.6-1 AMENDMENT NO. 4-14,

RCS Loops -

MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops -

MODE 5, Loops Filled LCO 3.4.7 One Shutdown Cooling (SDC) train shall be OPERABLE.and in operation, and either:

a.

One additional SDC train shall be OPERABLE: or b.'

The secondary side water level of each Steam Generator (SG) shall be Ž 25%.

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

1. The SDC pump of the train in operation 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 provided:

a.

No operations are permitted that would cause reduction of the RCS boron concentration; and

b. Core outlet temperature is maintained at least 100F below saturation temperature.

2.

One required SDC train may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other SDC train is OPERABLE and in operation.

3.

No Reactor Coolant Pump (RCP) shall be started with one or more of the RCS cold leg temperatures less than or equal to the LTOP enable temperature specified in the PTLR unless the secondary side water temperature in each SG is < 100°F above each of the RCS cold leg temperatures.

4.

No more than 2 RCPs may be in operation with RCS cold leg temperature

200 0 F.

No more than 3 RCPs may be in operation with RCS cold leg temperature > 200°F but 500 0F.

5.

All SDC trains may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.

APPLICABILITY:

MODE 5 with RCS loops filled.

PALO VERDE UNITS 1,2,3 3.4.7-1 AMENDMENT NO. 44-ý,

Pressurizer Safety Valves-MODE 4 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 Pressurizer Safety Valves-MODE 4 LCO 3.4.11 One pressurizer safety valve shall be OPERABLE with a lift setting Ž 2450.25 psia and

2549.25 psia.

APPLICABILITY:

MODE 4 with all RCS cold leg temperatures greater than the LTOP enable temperature specified in the PTLR.

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

The lift settings are. not required to be within LCO limits during MODES 3 and 4 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions.

This exception is allowed for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following entry into MODE 3 provided a preliminary cold setting was made prior to heatup.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

All pressurizer safety A.1 Be in MODE 4 with Immediately valves inoperable.

one Shutdown Cooling System suction line relief valve in service.

AND A.2 Perform SR 3.4.11.2 Immediately and SR 3.4.11.3 for the required Shutdown Cooling System suction line relief valve to comply with Action A.1.

AND A.3 Be in MODE 4 with 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months any RCS cold leg temperature less than or equal to the LTOP enable temperature specified in the PTLR.

PALO VERDE UNITS 1,2,3 3.4.11-1 AMENDMENT NO. 1-ý;,

LTOP System 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 Low Temperature Overpressure Protection (LTOP)

System LCO 3.4.13 An LTOP System shall be OPERABLE consisting of:

a.

Two OPERABLE Shutdown Cooling System suction line relief valves with lift settings : 467 psig aligned to provide overpressure protection for the RCS; or

b.

The RCS depressurized and an RCS vent of

Ž 16 square inches.

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

No RCP shall be started unless the secondary side water temperature in each steam generator (SG) is ! 100'F above each of the RCS cold leg temperatures.

MODE 4 when any RCS cold leg temperature is less than or equal to the LTOP enable temperature specified in the PTLR.

MODE 5, MODE 6 when the reactor vessel head is on.

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

LCO 3.0.4.b is not applicable when entering MODE 4.

APPLICABILITY:

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.1 Restore required 7 days A.

One required Shutdown Shutdown Cooling Cooling System suction System suction line line relief valve relief valve to inoperable in MODE 4.

OPERABLE status.

(continued)

PALO VERDE UNITS 1,2,3

.3.4.13-1 AMENDMENT NO. 4-65,

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.

(continued)

PALO VERDE UNITS 1,2,3 5.6-6 AMENDMENT NO. 4,

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.9 Reactor Coolant System (RCS)

PRESSURE AND. TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following Technical Specifications (TSs):

1.

TS 3.4.3, RCS Pressure and Temperature (P/T) Limits;

2.

TS 3.4.6, RCS Loops - Mode 4:

3.

TS 3.4.7, RCS Loops - Mode 5 Loops Filled;

4.

TS 3.4.11, Pressurizer Safety Valves - Mode 4; and

5.

TS 3.4.13, Low Temperature Overpressure Protection (LTOP)

System.

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

CE NPSD-683-A, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications."

c.

The PTLR shall be provided to the NRCupon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

PALO VERDE UNITS 1,2,3 5.6-7 AMENDMENT NO.

Evaluation of the Proposed TS Change Relocate P/T Limits to PTLR ENCLOSURE 1, ATTACHMENT 4 Technical Requirements Manual Page Markups (includes the Pressure-Temperature Limits Report [PTLR])

Pages:

iv T3.4.200-1 TA-i TA-ii TA-iii TA-1 TA-2 TA-3 TA-4 TA-5 TA-6 TA-7 TA-8 TA-9 TA-10 TA-11 TA-12 TA-13 TA-14 TA-15 TA-16

TABLE OF CONTENTS (continued)

T7.0 COMPONENT T7.0.100 T7.0.200 T7.0.300 T7.0.400 T7.0.500 Appendix TA LISTS Remote Shutdown Disconnect Switches............

T7.0.100-1 Remote Shutdown Control Circuits..................

T7.0.200-1 Containment Isolation Valves..................

T7.0.300-1 MOV Thermal Overload Protection and Bypass Devices T7.0.400-1 Containment Penetration Overcurrent Prot. Devices. T7.0.500-1 Reactor Coolant System Pressure and Temperature Limits Report (PTLR)

TA-i PALO VERDE UNITS 1, 2, 3 iv Rev 33 4/20/05

RCS Pressure and Temperature (P/T) Limits TRM 3.4.200 T3.4 REACTOR COOLANT SYSTEM (RCS)

T3.4.200 RCS Pressure and Temperature (P/T) Limits TLCO 3.4.200 APPLICABILITY:

Refer to PVNGS Improved Technical Specifications 3.4.3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of TSR A.1 Document the condition Immediately 3.4.200.1 not met.

on a CRDR and initiate an operability determination, as necessary, to determine the impact on equipment in the technical specifications.

NOTE-Changes to the reactor vessel surveillance specimen withdrawal schedule that meet the applicable ASTM standard must be submitted to the NRC with technical justification for approval prior to implementation (the NRC must verify compliance with the ASTM standard) in accordance with 10 CFR 50, Appendix H, paragraph III.B.3.

Changes to the withdrawal schedule that do not meet the applicable ASTM standard must be submitted to the NRC for approval as a license amendment with information required by 10 CFR 50.91 and 50.92 (see NRC Administrative Letter 97-04 dated September 30, 1997).

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.4.200.1 The reactor vessel material irradiation Refer to PVNGS surveillance specimens shall be removed and UFSAR Section examined to determine changes in material 5.3.1.6.6 properties at the intervals required by 10 "Withdrawal CFR 50, Appendix H in accordance with PVNGS Schedule" UFSAR section 5.3.1.6.6 "Withdrawal Schedule".

The results of these examinations shall be used to update the PTLR ITS Figurcs in specificatien 3.4.3.

PALO VERDE UNITS 1, 2, 3 T3.4. 200-1 Rev 0 8/13/98

Technical Requirements Manual APPENDIX TA REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

Palo Verde Nuclear Uni ts 1, Generating Station 2, and 3 Summary of PTLR:

This reactor coolant system pressure and temperature. limits report (PTLR) has been prepared in accordance with the reporting requirements of Technical Specification 5.6.9.

NRC letter dated March 16, 2001, accepted report CE NPSD-683-A, Rev. 6, which provides the methodology for developing this PTLR.

Application of CE NPSD-683 to PVNGS is documented in report WCAP-16835, Rev. 0.

Palo Verde Units 1, 2, 3 TA-i Rev X X/XX/08

TRM Appendix TA PTLR TABLE OF CONTENTS TA1.0 Reactor Coolant System Pressure and Temperature Limits Report TA-1 TA2.0 Operating Limits..............

TA-1 TA2.1 RCS Pressure and Temperature Limits (LCO 3.4.3)........ TA-2 TA2.2 RCS Loops - Mode 4 (LCO 3.4.6)...................

TA-2 TA2.3 RCS Loops - Mode 5, Loops Filled (LCO 3.4.7).......... TA-2 TA2.4 Pressurizer Safety Valves - Mode 4 (LCO 3.4.11)........ TA-2 TA2.5 Low Temperature Overpressure Protection System (LCO 3.4.13 ).............................

TA -2 TA3.0 Neutron Fluence........................................

TA-8 TA4.0 Reactor Vessel Material Surveillance Program.................

TA-8 TA5.0 Adjusted Reference Temperature...............................

TA-8 TA6.0 Application of Reactor Vessel Surveillance Data..............

TA-9 TA6.1 Applicability to Adjusted Reference Temperature........ TA-9 TA6.2 Evaluation of Surveillance Data Credibility.........

TA-IO TA6.3 Derivation of Chemistry Factors from Surveillance Data TA-15 TA7.0 References.................

TA-15 Palo Verde Units 1, ?, 3 TA-u Rev X X/XX/08 Palo Verde Units 1, 2, 3 TA -i i Rev X X/XX/08

TRM Appendix TA PTLR TABLE OF CONTENTS List of Figures Page TA2-1 RCS Heatup Limits through 32 EFPY...........................

TA-6 TA2-2 RCS Cooldown Limits through 32 EFPY.........................

TA-7 List of Tables Page TA2-1 RCS Heatup and. Cooldown Rate Limits through 32 EFPY......... TA-3 TA2-2 Limiting RCS Temperatures through 32 EFPY.................... TA-3 TA2-3 RCS Heatup P/T Limits through 32 EFPY......................

TA-4 TA2-4 RCS Cooldown P/T Limits through 32 EFPY.....................

TA75 TA3-1 Summary of Fluence and Fluence Factors.......

I.......... TA-8 TA5-1 Summary of Limiting ART and RTPTS Values

................ TA-9 TA6-1 Base Metal Materials Selected for Surveillance Program.....

TA-11 TA6-2 Unit 1 Credibility of Surveillance Measurements............

TA-13 TA6-3 Unit 2 Credibility of Surveillance Measurements............

TA-14 TA6-4 Unit 3 Credibility of Surveillance Measurements.........

TA-14 Palo Verde Units 1, 2, 3 TA-iii Rev X X/XX/08 Palo Verde Units 1, 2, 3 TA-ii i Rev X X/XX/08

TRM Appendix TA PTLR TA1.0 Reactor Coolant System Pressure and Temperature Limits Report (PTLR)

This Pressure and Temperature Limits Report (PTLR) has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.9.

The following TSs are addressed in this report:

TS 3.4.3, RCS Pressure and Temperature Limits; TS 3.4.6, RCS Loops -

Mode 4; TS 3.4.7, RCS Loops -

Mode 5, Loops Filled; TS 3.4.11, Pressurizer Safety Valves - Mode 4;.and TS 3.4.13, Low Temperature Overpressure Protection System.

TA2.0 Operating Limits Parametric limits for the above LCOs were developed using NRC-approved methods specified in Technical Specification 5.6.9 (Ref.

1).

Application of the method6logy approved for developing P/T limits, i.e., report CE NPSD-683-A (Ref. 2), to the Palo Verde Nuclear Generating Station is detailed in WCAP-16835 (Ref. 3).

The initial PTLR was submitted to the NRC along with the Technical Specification (TS) amendment request to.relocate P/T limits to the PTLR (Ref. 4).

The NRC approved the relocation of the P/T limits from TS to the PTLR in amendment no. L b (Ref. 5).

Subsequent changes to the PTLR are controlled in accordance with TS 5.6.9b and 10 CFR 50.59, and the PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto as required by TS 5.6.9c.

The pressure-temperature limit curves comply with Appendix G to 10 CFR Part 50 requirements that the temperature of the closure head flange and vessel flange regions must be at least 120°F higher than the limiting RTNDT for these regions.

This RTNDT limit applies during normal operation, including heatup and cooldown, when the core is not critical and the vessel pressure exceeds 625 psia, (20% of the pre-service hydrostatic test pressure of 3125 psia).

Refer to the Technical Specifications for LCOs and surveillance requirements applicable to RCS pressure and temperature limits.

Specific TS LCO limits relocated from the-Technical Specifications into this PTLR are given below.

(continued)

Palo Verde Units 1, 2, 3 TA-I Rev X X/XX/08

. TRM Appendix TA PTLR TA2.1 RCS Pressure and Temperature Limits (LCO 3.4.3)

RCS heatup and cooldown rates for Units 1, 2 and 3 through 32 EFPY shall be equal to or less than the values shown in Table TA2-1.

Limiting RCS temperature requirements through 32 EFPY are specified in Table TA2-2. The RCS pressure and temperature for vessel head boltup, inservice hydrostatic and leak testing through 32 EFPY shall be limited as specified on Figure TA2-1 (or Table TA2-3) for RCS heatup, and Figure TA2-2 (or Table TA2-4) for RCS cooldown.

A gradual change in reactor coolant system temperature of +/-100F in any 1-hour period is the maximum.perinitted during inservice hydrostatic and leak testing.

TA2.2 RCS Loops - Mode 4 (LCO 3.4.6)

The LTOP 221'F as enable temperature for RCS heatup and cooldown through 32 EFPY is specified in Table TA2-2.

TA2.3 RCS Loops - Mode 5, Loops Filled (LCO 3.4.7)

The LTOP 221°F as enable temperature for RCS heatup and cooldown through 32 EFPY is specified in Table TA2-2.

TA2.4 Pressurizer Safety Valves - Mode 4 (LCO 3.4.11)

The LTOP 221°F as enable temperature for RCS heatup and cooldown specified in Table TA2-2.

through 32 EFPY is TA2.5 Low Temperature Overpressure Protection System (LCO 3.4.13)

The LTOP 221°F as enable temperature specified in Table for RCS heatup and cooldown TA2-2.

through 32 EFPY is (continued)

Rev X X/XX/08 Palo Verde Units 1, 2, 3 TA-2

TRM Appendix TA PTLR Table TA2-1 PVNGS Units 1, 2 and 3 RCS Heatup and Cooldown Rate Limits through 32 EFPY (Formerly TS Table 3.4.3-1)

Indicated RCS Cold Leg Heatup Rate Cooldown Rate Temperature (OF)()*

(°F/hr)

(°F/hr) 800 to 920

< 75 5 30

> 920 to 1000

75

50

> 1000 to *2210 75 S 100

> 2210

75

100 (1) Corrected for instrument uncertainty.

Table TA2-2 PVNGS Units 1, 2 and 3 Limiting RCS Temperatures through 32 EFPY Requirement RCS Temperature"1 )

Minimum Boltup Temperature 800F Minimum Hydrostatic Test Temperature 181.4 0 F Lowest Service Temperature 153.2 0 F Minimum Flange Limit (Hydrostatic Test) 163.2 0 F Minimum Flange Limit (Normal Operation) 193.2 0 F LTOP Heatup and Cooldown Enable Temperature 221°F (1)

Corrected for instrument uncertainty.

(continued)

Rev X X/XX/08 Palo Verde Units 1, 2, 3 TA-3

TRM Appendix TA PTLR Table TA2-3 PVNGS Unit 1, 2 and 3 RCS Heatup P/T Limits through 32 EFPY Indicated Pressure Indicated RCS Pressure (psia)(1) @ Heatup Rate Hydrostatic Temperature Isothermal Test"2 '

(oF)(1)

(psia)

@10F/hr

@20 0F/hr

@30°F/hr

@40°F/hr

@50°F/hr

@75°F/hr (psia) 80 680.6 680.6 680.6 671.1 650.2 622.2 602.2.

954.4 83.2 690.2 690.2 690.2 676.2 650.2 622.2 602.2 967.2 93.2 727.2

'727.2 705.2 676.2 650.2 622.2 602.2 1016.2 103.2 772.2 772.2 710.2 676.2 650.2 622.2 602.2 1075.2 113.2 826.2 826.2 735.2 681.2 650.2 622.2 602.2 1148.2 123.2 893.2 893.2 778.2.

700.2 653.2 622.2 602.2 1237.2 133.2 974.2 974.2 839.2 738.2 672.2 627.2 602.2 1346.2 143.2 1074.2 1074.2 918.2 790.2 705.2 645.2 602.2 1478.2 153.2

.1195.2 1195.2 1018.2 862.2 754.2 676.2 604.2 1640.2 163.2 1344.2 1335.2 1142.2 954.2 819.2 721.2 617.2 1838.2 171.5 1494.8 1467.5 1269.5 1049.0 889.9 772.8 638.0 2039.1 172.1 1507.0 1478.3 1279.9 1057.0 896.0 777.3 598.0 2053.6 173.2 1525.2 1494.2 1295.2 1068.2 904.2 783.2 600.2 2080.2 183.2 1747.2 1689.2 1484.2 1213.2 1014.2 865.2 637.2 2375.2 186.7 1841.7 1772.5 1565.4 1275.5 1062.2 902.0 655.4 2500.0 193.2 2017.2 1927.2 1716.2 1391.2 1151.2 970.2 689.2 203.2 2347.2 2217.2 1998.2 1610.2 1320.2 1101.2 757.2 207.0 2500.0 2351.5 2129.3 1713.2 1399.2 1162.4 790.6 211.2 2500.0

2274.2 1827.0 1486.6 1230.0 827.6 213.2 2343.2 1881.2 1528.2 1262.2 845.2 213.2 2327.2 1865.2 1512.2 1246.2 829.2 217.3 2500.0 1998.9 1616.3 1327.8 874.7 223.2 2191.2 1766.2 1445.2 940.2 230,8 2500.0 2008.6 1634.4 1045.8 233.2 2085.2 1694.2 1079.2 (1) Corrected for instrument uncertainty and for RCS pressure and elevation effects.

(2) A gradual change.in reactor coolant system temperature of +/-100F in any 1-hour period is the maximum permitted during inservice hydrostatic and leak testing. *

(conti nued)

Palo Verde Units 1, 2, 3 TA-4 Rev X X/XX/08

TRM Appendix TA PTLR Table TA2-4 PVNGS Unit 1, 2 and 3 RCS Cooldown P/T Limits through 32 EFPY Indicated Indicated RCS Pressure (psia)(1) @ Cooldown Rate Temperature Tepeatr Isothermal

@10 0F/hr

@20°F/hr

@30 0F/hr

@40°F/hr

@50°F/hr

@75°F/hr

@100°F/hr (OF) (1 80 680.6 612.3 589.0 527.1 469.5 416.6 329.2 237.6 83.2 690.2 623.2 601.2 541.2 485.2 433.2 329.2 272.2 90.9 718.6 655.4 638.0 583.4 533.5 492.2 402.8 372.6 91.3 720.1 657.2 598.0 585.7 536.1 495.4 406.8 378.1 93.2 727.2 665.2 607.2 596.2 548.2 510.2 425.2 403.2 99.6 756.1 698.0 644.5 638.0 597.1

.559.7 501.1 493.2 99.9 757.5 699.6 646.3 598.0 599.4 562.1 504.7 497.5 103.2 772.2 716.2 665.2 619.2 624.2 587.2 543.2 543.2 104.7 780.4 725.6 676.1 631.3 638.0 604.8 565.0 565.0 104.9 781.6 727.0 677.7 633.1 598.0 607.3 568.2 568.2 107.6 795.8 743.4 696.7 654.2 622.1 638.0 606.3 606.3 107.8 796.8 744.4 698.0 655.6 623.6 598.0 608.7 608.7 109.8 807.8 757.0 712.6 671.9 642.1 621.6 638.0 638.0 109.9 808.5 757.9 713.6 673.0 643.4 623.2 598.0 598.0 113.2 826.2 778.2 737.2 699.2 673.2 661.2 645.2 645.2 123.2 893.2 854.2 823.2 798.2 781.2 776.2 776.2 776.2 133.2 974.2 947.2 929.2 918.2 918.2 918.2 918.2 918.2 143.2 1074.2 1060.2 1057.2 1057.2 1057.2 1057.2, 1057.2 1057.2 153.2 1195.2 1195.2 1195.2 1195.2 1195.2 1195.2 1195.2 1195.2 163.2 1344.2 1344.2 1344.2 1344.2 1344.2 1344.2 1344.2 1344.2 173.2 1525.2 1525.2 1525.2 1525.2 !

1525.2 1525.2 1525.2 1525.2 183.2 1747.2 1747.2 1747.2 1747.2 1747.2 1747.2 1747.2 1747.2 193.2 2017.2 2017.2 2017.2 2017.2 2017.2 2017;2 2017.2 2017.2 203.2 2347.2 2347.2 2347.2 2347.2 2347.2 2347.2 2347.2 2347.2 (1) Corrected for instrument uncertainty and for RCS pressure and elevation effects.

(continued)

Palo Verde Units 1, 2, 3 TA-5 Rev X X/XX/08

TRM Appendix TA PTLR Figure TA2-1 PVNGS Units 1, 2 and 3 RCS Heatup Limits(')(2) through 32 EFPY (Formerly TS Figure TS 3.4.3-2) 2500 2000 1500 CL1000 500 0

T T r4 -. r. -

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(2)

A gradual change in reactor coolant system temperature of +/-10 0F in any 1-hour period is the maximum permitted during inservice hydrostatic and leak testing.

(continued)

TA-6 Rev X X/XX/08 Palo Verde Units 1, 2, 3

TRM Appendix TA PTLR Figure TA2-2 PVNGS.Units 1, 2 and 3 RCS Cooldown Limits"""(

2) through 32 EFPY (Formerly TS Figure TS 3.4.3-2) 2500 2000 C.

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0 50 100 150 200 250 300 Indicated RCS Temperature Tc (0F)

(1)

Corrected for instrument uncertainty and for RCS pressure and elevation effects.

(2)

A gradual change in reactor coolant system temperature of +/-10 0F in any 1-hour period is the maximum permitted during inservice hydrostatic and leak testing.

(continued)

Rev X X/XX/08 Palo Verde Units 1, 2, 3 TA-7

TRM Appendix TA PTLR TA3.0 Neutron Fluence The design value of peak fast neutron fluence through 32 EFPY for determining the limiting reactor vessel beltline material adjusted reference temperature is 3.29E+19 n/cm2 (E > 1.0 MeV) and corresponds to the fluence at the vessel clad-to-base metal interface.

For conservatism, this peak fluence is assumed to apply to each of the PVNGS reactor vessel beltline plates and welds; i.e.,

no reduction factor is applied to account for axial or azimuthal variations from the peak value.

A summary of fast neutron fluence and fluence factors through 32 EFPY determined at the 1/4T and 3/4T locations in the vessel wall is given in Table TA3-1.

These fluence values are usedto calculate the adjusted reference temperature at PVNGS Units 1, 2 and 3.

Table TA3-1 Summary of Fluence and Fluence Factors Location 1/4T f (n/c&)"l) 1/4T ffC

2) 3/4T. f (n/cm2)(I) 3/4T ff(2)

Intermediate Shell 1.681E+19 1.1431 4.390E+18 0.7711 Lower Shell 1.910E+19 1.1770 6.438E+18 0.8766 (1) f = fast neutron fluence (2) ff = fluence factor TA4.0 Reactor Vessel Material Surveillance Program The PVNGS reactor vessel material surveillance program, as described in Section 5.3 of the PVNGS Updated Final Safety Analysis Report (UFSAR),

is in compliance with 10 CFR Part 50, Appendix H "Reactor Vessel Material Surveillance Program Requirements."

The surveillance capsule withdrawal schedules are presented in UFSAR Tables 5.3-13 through 5.3-19A and summarized in WCAP-16835 (Ref.. 3).. Test results and analyses of withdrawn surveillance specimens were reported in References 6, 7, and 8.

TA5.0 Adjusted Reference Temperature A summary of limiting adjusted reference temperatures associated with PVNGS beltline materials at the 1/4T and 3/4T locations along with RTPTS values through 32 EFPY is given in Table TA5-1.

Conservatively, themost limiting (highest) adjusted reference temperature value from the three PVNGS units is (continued)

Palo Verde Units 1, 2, 3 TA-8 Rev X X/XX/08

TRM Appendix TA PTLR applied to all three units.

Chemistry factors and adjusted reference temperatures for PVNGS are determined in accordance with Regulatory Guide 1.99 (Ref. 9).

Table TA5-1 Summary of Limiting ART and RTPTs Values PVNGS Location Material 1/41 ART 3/4T ART (OF)

RTPTs (0F)

(OF)

Unit i Inter. Shell Plate M-6701-2 11.6 103

.123 Unit 2 Inter. Shell Plate F-765-6 74 64.

78 Unit 3 Lower Shell Plate F-6411-2 65 57 68 Limiting adjusted reference temperatures are incorporated into the calculation of pressure-temperature curves and limits for heatup, cooldown, LTOP, and hydrostatic and leak tests.

Seven reactor vessel surveillance capsules have been removed.from PVNGS Units 1, 2 and 3 through December 2007, with a minimum of two credible data sets available for each PVNGS unit.

Even though WCAP-16835 shows that the post-irradiation surveillance capsule test results for PVNGS units are credible, the calculation of ART takes no credit for those credible results.

TA6.0 Application of Reactor Vessel Surveillance Data TA6.1 Applicability to Adjusted Reference Temperature Data from the reactor vesselsurveillance program or from other supplemental sources were not used to determine the adjusted reference temperature (ART) values for the PVNGS beltline materials described in Section 5.

The surveillance program data from each of the three Palo Verde units were evaluated for credibility; chemistry factors were also derived for those surveillance materials.

(This assessment is further detailed in Report WCAP-16835 [Ref. 3].)

Chemistry factors determined following Position 1.1 of Regulatory Guide 1.99 are shown to be conservative relative to those derived from surveillance plate and weld measurements for each of the PVNGS units.

Therefore, no credit is taken for those credible results in the calculation of ART.

(continued)

Palo Verde Units 1, 2, 3 TA-9 Rev X X/XX/08

TRM Appendix TA PTLR Each new set of surveillance results, as new data becomes available, will be evaluated to ascertain that the Position 1.1 chemistry factors remain conservative relative to the surveillance results.

This will ensure that the existing RCS P/T limits remain conservative for continued plant operation or will be revised as needed to provide conservative RCS pressure-temperature limitsý TA6.2 Evaluation of Surveillance Data Credibility Regulatory Guide 1.99 describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of low-alloy steels used in the PVNGS reactor vessels when credible surveillance capsule data is available.

Position C.2 of Regulatory Guide 1.99 describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data.

The methods of Position C.2 apply when two or more credible surveillance data sets become available from each unit.

Post-irradiation surveillance capsule test results for PVNGS were evaluated with respect to the credibility criteria of Regulatory Guide 1.99 Revision 2.

Results of the credibility, assessment found that:

The surveillance program plates and welds are those judged to be most likely controlling with regard to radiation-induced embrittlement,

Charpy data scatter does not cause. ambiguity*in the determination of the 30 ft-lb shift, Measured RTNDT shifts are consistent with the predicted shifts,

Capsule irradiation temperature matches that of the vessel wall, and

Correlation monitor data falls within the scatter band for that material and therefore meets the credibility test.

Revision 2 of Regulatory Guide 1.99 defines five requirements that must be met for surveillance data to be judged credible.

The purpose of the following discussion is to apply these credibility requirements to PVNGS to show that the reactor vessel surveillance data are credible.

Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement..

(continued)

Palo Verde Units 1, 2, 3 TA-IO*

Rev X X/XX/08

TRM Appendix TA PTLR The beltline region of the reactor vessel is defined in Appendix G(II)(F) to 10 CFR Part 50,

".Fracture Toughness Requirements," as:

"Beltline or beltline region ofthe reactor vessel means the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage."

Table TA6-1 identifies the Units 1, 2, and 3 reactor vessel base metal beltline, plate materials selected for the PVNGS surveillance program.

Table TA6-1 Base Metal Materials Selected for Surveillance Program PVNGS Plate ID Number Plate Location Unit 1(il M-4311-1 Lower Shell Unit 1ll)

M-6701-2 Intermediate Shell.

Unit 2 F-773-1 Lower Shell Unit 3 F-6411-2 Lower Shell (1)Unit 1 has two different base metal surveillance materials.

The weld materials for the PVNGS Units 1, 2, and 3 surveillance programs are selected to duplicate the materials in the lower shell axial weld seams.

Test specimens from these materials are heat-treated to a condition representative of the final metallurgical condition of the weld metal in the completed reactor vessel.

These surveillance weld metals were made-with the same weld wire heat. as that of the vessel beltline weld seams and are, therefore, representative of all beltline weld seams.

Intermediate shell plate M-6701-2 in Unit 1 has the highest initial RTNDT and the highest ART of all PVNGS plate materials in the beltline region.

Therefore, the PVNGS Units 1, 2, and 3 surveillance material meets the intent of this criterion.,

Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper shelf energy unambiguously.

(continued).

Palo Verde Units 1, 2, 3 TA-11 Rev X X/XX/08

TRM Appendix TA PTLR Evaluation of Charpy energy versus temperature for the unirradiated and irradiated condition is presented in References 6, 7, and 8.

Based on engineering judgment, the scatter in the data presented is small enough to permit the unambiguous determination of the 30 ft-lb temperature and the upper shelf energy of the PVNGS Units 1, 2, and 3 surveillance materials:

Thus, the PVNGS surveillance program meets this criterion.

Criterion 3:

When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17'F for base metal.

Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values.

Even if the data fail this criterion for use in shift calculations, they maybe credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82.

The surveillance program for PVNGS Units 1, 2, and 3 is based on ASTM E185-79, which presents criteria for monitoring changes in the fracture toughness properties of reactor vessel beltline materials.

References 6, 7, and 8 describe the post-irradiation evaluations of PVNGS surveillance materials.

The credibility results shown in Tables TA6-2, TA6-3 and TA6-4 for Units 1, 2, and 3, respectively, present the shift measurements available to date.

Those values are compared to predictions based on a chemistry factor determined following Position 1.1 of Regulatory Guide 1.99.

In all cases, the difference between the measured and predicted shift is less than 170F for the surveillance plates and less than 280F for the surveillance welds.

Therefore, this criterion is met for the PVNGS Units 1, 2, and 3 surveillance program plate and weld materials.

(continued)

Palo Verde Units 1, 2,.3 TA-12 Rev.X X/XX/08

TRM Appendix TA PTLR Table TA6-2 PVNGS Unit 1 Credibility of Surveillance Measurements Plate, Weld and Correlation Monitor Materials Mate rial Capsule CF ff(2 Measured Predicted Shift Difference Shift. (0F)

(CF*ff) (OF)

(OF)

Plate 137 27.5 0.7216 34.2 19.8

+ 14.4 M-6701-2 Longi tudi nal 230 27.5 0.9629 15.3 26.5 11.2 Plate 137 27.5 0.7216.

13.0 19.8

- 6.8 M-6701-2 Transverse 230 27.5 0.9629 31.9 26.5

+ 5.4 137 4.9 0.7216

- 2.8 3.5

- 6.3 Wel d Heat 38 4.9 0.8697 6.7 4.3

+ 2.4 (Heat 90071) 230 4.9 0.9629 5.1 4.7

+ 0.4 Correlation 137 131.7(l) 0.7216 101.3 95.0

+ 6.3 Monitor 38 131.7(l) 0.8697 114.1 114.5

- 0.4 Material 230 131.7(l)

.0.9629 129.2 126.8

+ 2.4 (1)Chemistry factor based on 0.174 Cu and 0.665 Ni using Table 2 of Regulatory Guide 1.99, Rev. 2.

(2)ff = fluence factor = f(O.28 - 0.1*log f)

.(continued)

Rev X X/XX/08 Palo Verde Units 1, 2, 3 TA-13

TRM Appendix TA PTLR Table TA6-3 PVNGS Unit 2 Credibility of Surveillance Measurements Plate, Weld and Correlation Monitor Materials Material Capsule CF ff(2)

Measured Predicted Shift Difference Shift (OF)

(CF*ff) (OF)

(OF)

Plate F-773-1 137 17.5 0.7372 13.3 12.9

+ 0.4 Longitudinal 230 17.5 0.9978 17.7 17.5

+ 0.2 Plate F-773-1 137 17.5 0.7372 9.5 12.9

- 3.4 Transverse 230 17.5 0.9978 19.3 17.5

+ 1.8 Weld 137 1.6 0.7372 0

1.2 1.2 (Heat 3P7317) 230.

1.6 0.9978 2.5 1.6

+ 0.9 Correlation 137 131.7'1) 0.7372 116.0 97.1

+ 18.9 Monitor Material 230 131.7 (1) 0.9978 132.4 131.4

+ 1.0 (1) Chemistry Factor based on 0.174 Cu and 0.665 Ni using Table 2 of Regulatory Guide 1.99, R02.

(2) ff = fl uence factor = f(O.28 -

.-,1og f)

Table TA6-4 PVNGS Unit 3 Credibility of Surveillance Measurements Plate, Weld and Correlation Monitor Materials Measured Predicted Shift Difference Material Capsule CF ff(2) hf 0 )

(F1f 0 )(F Shift (OF)

(CF*ff)

(OF)

(°F)

Plate F-6411-2 230 10.2 0.9726 6.3 9.9

- 3.6 Longi tudi nal Plate F-6411-2 142 10.2 0.7090 13.1 7.2

+ 5.9 Transverse 230 10.2 0.9726 9.2 9.9

- 0.7 Weld 142 29.6 0.7090 27.5 21.0

+ 6.5 (Heat 4P7869) 230 29.6 0.9726 24.1 28.8

- 4.7 Correlation 142 131.7(l) 0.7090 82.5 93.4 10.9 Monitor Material 230 131.7(l) 0.9726 141.8 128.1

+ 13.7 (1)Chemistry Factor based on 0.174 Cu and 0.665 Ni using Table 2 of Regulatory Guide 1.99, R02.

(2)ff = fluence factor = f(O.28 - 0,1*og f)

(continued)

Rev X X/XX/08 Palo Verde Units 1, 2, 3 TA-14

TRM Appendix TA PTLR Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/-,250F.

All reactor vessel surveillance specimen capsule holders are attached to the inside vessel wall cladding in the beltline region at PVNGS.

This capsule holder attachment method meets the design and inspection requirements of the ASME Code, Sections III and XI.

The location of the specimens relative to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the; temperatures will not differ by more than 250F.

Hence this criterion is met.

Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for that material.

As shown in Tables TA6-2, TA6-3, and TA6-4, the correlation monitor materials from PVNGS Units 1, 2,*and 3 meet the credibility test to be within the scatter band of the database for that material.

TA6.3 Derivation of Chemistry Factors from Surveillance Data The derived chemistry factor (CF) values for each of the Palo Verde Units 1, 2, and 3 surveillance materials is provided in Tables TA6-2, TA6-3, and TA6-4.

The derived-chemistry factor for the Unit 1 surveillance plate (M-6701-2) and surveillance weld (Heat 90071) are 27.5°F and 4.9 0 F, respectively.

These chemistry factor values compare favorably with their respective values of 370F and 27,8 0F determined following Position 1.1 of Regulatory Guide 1.99.

Similar conservative results are found for Uni-t 2 plate (F-773-1) and surveillance weld (Heat 3P7317),

and for Unit 3 plate (F-6411-2) and surveillance weld (Heat 4P7869).

Therefore, chemistry factors determined following Position 1.1 of Regulatory Guide 1.99 are shown to be conservative relative to those derived from surveillance plate and weld measurements for each of the PVNGS units.

TA7.0 References

1.

PVNGS Units 1, 2, and 3 Technical Specification 5.6.9, "Reactor Coolant System Pressure and Temperature Limits Report."

2.

CE Owners Group Topical Report CE NPSD-683-A, Revision 6, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P/T (continued)

Palo Verde Units 1, 2, 3 TA-15 Rev X X/XX/08

TRM AppendixTA PTLR Limits and LTOP Requirements from the Technical Specifications,"

April 2001.

3.

Westinghouse Report WCAP-16835, Revision 0, "Palo Verde Nuclear Generating Station Units 1, 2 and 3, Basis for RCS Pressure-Temperature Limits Report," June 2008.

4.

APS letter No.

205 to NRC, "Request for Technical Specification Amendment to Relocate the Reactor Coolant System Pressure and Temperature Limits and the Low Temperature OverpressureProtection Enable Temperatures," dated a

-/z xX.

[l~iTITh eeeanytsuppent]

5.

Letter, NRC to APS, "Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments Re: [Relocation of RCS Pressure and Temperature Limits] (TAC NOS.

_xxxxTxxxxAN

_da'-t-d Txxxx

6.

APS letter no. 102-05242 to NRC, Palo Verde Nuclear Generating Station Unit 1 Reactor Vessel Material Surveillance Capsule at 230o,,, April 5, 2005 (transmittal of WCAP-16374-NP, "Analysis of Capsule 2300 from Arizona Public Service Company Palo Verde Unit 1 Reactor Vessel Radiation Surveillance Program," February 2005).

7.

APS letter no. 102-05457 to NRC, Palo Verde Nuclear Generating Station Unit 2 Reactor Vessel Material Surveillance Capsule at 2300,' April 4, 2006 (transmittal of WCAP-16524-NP, "Analysis of Capsule 2300 from Arizona Public Service Company Palo Verde Unit 2 Reactor Vessel Radiation Surveillance Program." February 2006).

8.

APS letter no. 102-05348 to NRC, Palo Verde Nuclear Generating Station Unit 3 Analysis of Reactor Vessel Material Surveillance Capsule at 2300,,, September 26, 2005 (transmittal of WCAP-16449-NP, "Analysis of Capsule 2300 from Arizona Public Service Company Palo Verde Unit 3 Reactor Vessel Radiation Surveillance Program," August 2005).

9.

Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission,. May 1998.

Palo Verde Units 1, 2, 3 TA-16 Rev X X/XX/08

v t e Letter Sequence Request
TAC:ME0698, Relocation of Ltop Enabletemperature and PORV Lift Setting to the PTLR - CE NPSD-683 (Approved, Closed) TAC:ME0703, (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance Administration Withholding Request Acceptance Questions 25 August 2009 Answers 22 December 2009 Results Approval Other: ML092380303
MONTHYEAR2009-04-29 [Table View]

ML090641014

Text

Dear Sirs:

Subject:

Enclosures:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

RESPONSE

RESPONSE

3.4.3 APPLICABILITY

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