ML051660184

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Response to Request for Additional Information Regarding Steam Generator Replacement and Power Uprate License Amendment Request
ML051660184
Person / Time
Site: Palo Verde  Arizona Public Service icon.png
Issue date: 06/03/2005
From: Mauldin D
Arizona Public Service Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
102-05285-CDM/TNW/RAB
Download: ML051660184 (20)
v • d • e


Text

10 CFR 50.90 LAPS David Mauldin Vice President Mail Station 7605 Palo Verde Nuclear Nuclear Engineering Tel: 623-393-5553 PO Box 52034 Generating Station and Support Fax: 623-393-6077 Phoenix, Arizona 85072-2034 102-05285-CDM/TNW/RAB June 3, 2005 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

Reference:

Letter No. 102-05116-CDM/TNW/RAB, dated July 9, 2004, from C. D.

Mauldin, APS, to U. S. Nuclear Regulatory Commission, "Request for a License Amendment to Support Replacement of Steam Generators and Uprated Power Operations in Units 1 and 3, and Associated Administrative Changes for Unit 2"

Dear Sirs:

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units I and 3, Docket Nos. STN 50-528 and STN 50-530 Response to Request for Additional Information Regarding Steam Generator Replacement and Power Uprate License Amendment Request In the referenced letter, Arizona Public Service Company (APS) submitted a license amendment request to support steam generator replacement and uprated power operations for PVNGS Units 1 and 3.

The enclosure to this letter provides written responses to the questions provided by the NRC in electronic mail and discussed in telephone conversations. The response to NRC Question 3 contains the following statement in which APS commits to a degradation management program for reactor vessel internals:

Arizona Public Service Company (APS) is currently an active participant in the Electric Power Research Institute (EPRI) Materials Reliability Program research initiatives on aging related degradation of reactor vessel internals components.

APS commits to:

  • Continue its active participation in the MRP initiative to determine appropriate reactor vessel internals degradation management programs, A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
  • Comanche Peak
  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

U. S. Nuclear Regulatory Coifmmission ATTN: Document Control Desk Response to Request for Additional lnforiiatioh Regarding Steam Generator Replacement and Power Uprate License Amendment Request Page 2

  • Evaluate the recommendations resulting from this initiative and implement a reactor vessel internals degradation management program applicable to Palo Verde Units I and 3,
  • Incorporate the resulting reactor vessel internals inspections into the Palo Verde 1 and 3 augmented inspection program as appropriate.

In addition, as requested by the NRC, a description of the program, including the inspection plan, will be submitted to the NRC for review and approval. The submittal date will be within 24 months of the EPRI MRP final recommendations or within five years from the date of issuance of the uprated license, whichever comes first.

Should you have any questions, please contact Thomas N.Weber at 623-393-5764.

Sincerely, CDM/TNW/RAB/ca

Enclosures:

1. Notarized Affidavit
2. Materials and Chemical Engineering Branch Questions and APS Responses cc: B. S. Mallet NRC Region IV M. B. Fields NRC Project Manager G. G.Warnick NRC Senior Resident Inspector for PVNGS A. V. Godwin Arizona Radiation Regulatory Agency (ARRA)

ENCLOSURE 1 NOTARIZED AFFIDAVIT STATE OF ARIZONA )

) ss.

COUNTY OF MARICOPA )

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

David Mauldin Swom To Before Me This 00t Day Of.- n"Jzo

. a 2005.

No~~c Notary Commission Stamp

Enclosure 2 Materials and Chemical Engineering Branch Questions and APS Responses

Units 1 & 3 PUR RAI - RV Integrity The Attachment to this Enclosure contains ekcerpts from the referenced letters as identified below. The Attachment also contains excerpts from the NRC Reactor Vessel Integrity Database for Palo Verde highlighting the requested information.

NRC Question 1.1.- The projected neutron fluence (E>1.0 Mev) for each vessel beftline material at EOL including the impact of the proposed UPO.

APS Response As cited in the references below, the projected neutron fluence for PUR (3990 MWt) given that the projected fluence at end-of-license, 3.29E+1 9 n/cm2 , E>l MeV, is bounded by the Analysis of Record (AOR).

References

1) Letter 102-03448, dated August 17,1995 from APS to USNRC, Response to NRC Generic Letter 92-01, Rev.1, Supplement 1 reported fluence values for Palo Verde Units 1, 2 and 3.
2) USNRC Reactor Vessel Integrity Database contains the EOL fluence values for Palo Verde Units 1, 2, and 3. See the database summary in The Attachment to this Enclosure.
3) Letter 102-04641 from APS to USNRC, dated December 21, 2001, Attachment 6, Sections 5.1.2 & 7.5 discuss the impact of the proposed UPO. Excerpts from 102-04641 are provided in The Attachment to this Enclosure.
4) Letter 102-04834 from APS to USNRC, dated August 29, 2002, Attachment 2, NRC question 7 response provides discussion on the impact of the proposed UPO for Unit 2. These results are summarized for Units 1, 2, and 3 in The Attachment to this Enclosure.
5) Letter 102-04847 from APS to USNRC, dated October 11, 20C2, Attachment 2, NRC Question 17 response provides further discussion on the impact of the proposed UPO.

NRC Question 1.11.- Reactor vessel beltline material properties including initial RTNDT, Cu and Ni contents and the source of the information (generic or plant specific).

APS Response As cited in the references below, for Palo Verde Units 1, 2, and 3, the reactor vessel beltline plate and as deposited weld chemistries were controlled to low weight percentages for the following elements: copper < 0.1 wt.%, phosphorus < 0.012 wt.%,

vanadium < 0.01 wt.%, and sulfur _<0.022 wt.%. The initial RTNDT is s 40 0F. The source of this information is plant specific.

1

Units 1 & 3 PUR RAI - RV Integrity References

1) Letter 102-03448 from APS to USNRC, dated August 17,1995, Response to NRC Generic Letter 92-01, Rev. 1, Supplement 1 reported values for RTNDT, Cu and Ni.
2) USNRC Reactor Vessel Integrity Database contains the initial RTNDT, Cu and Ni values for Palo Verde Units 1,2 and 3. See the database summary in The Attachment to this Enclosure.
3) Letter 102-04139 from APS to USNRC, dated June 24, 1998 provides responses to NRC questions that updates information provided in the APS response to GL 92-01, Rev. 1, Supplement 1.

NRC Question 1.111.- RTPTs values at end of current licensed life including the impact of UPO for all vessel beltline materials. Also, provide the basis of RTPTS values.

APS Response Pressurized Thermal Shock (PTS) - The screening criteria in 10 CFR Part 50.61 is 270 OF for plates, forgings, and axial weld materials, and 300 OF for circumferential weld materials. The highest RTpTs value for a plate from the intermediate shell course of the RV for Palo Verde at the end of the current license was determined to be as follows:

Unit 1 122.5 0F Unit 2 78 OF Unit 3 68.1 OF The projected RTPTS value at the end of the current license for the be!tline materials are summarized in Figure 7-2 (from the NRC Reactor Vessel Integrity Database (RVID).

These values represent conditions for PUR (3990 MWt) given that the projected fluence at end-of-license, 3.29E+1 9 n/cm2 , E>l MeV, is bounded by the Analysis of Record (AOR) and the method for predicting RTPTS unchanged.

References

1) Letter 102-04641 from APS to USNRC, dated December 21, 2001, Attachment 6, Sections 5.1.2 & 7.5 discuss the impact of the proposed UPO on the reactor vessel. These results are summarized for Units 1, 2, and 3 in The Attachment to this Enclosure.
2) Letter 102-04847 from APS to USNRC, dated October 11, 2002, Attachment 2, NRC Question 17 response provides further discussion on the impact of the proposed UPO.
3) Letter 102-04834 from APS to USNRC, dated August 29, 2002, Attachment 2, NRC question 7 response provides discussion on the impact of the proposed UPO. These results are summarized for Units 1, 2, and 3 in The Attachment to this Enclosure.
4) Letter 102-04899 from APS to NRC, dated March 11, 2003, Attachment 2, NRC question 15 response also provides further discussion on the impact of the proposed UPO.

2

Units 1 & 3 PUR RAI - RV Integrity NRC Question 1.IV.- For each beltline material, provide the USE values at the end of the current licensed life including the impact of UPO. Also, provide the basis of the calculation including beltline material copper percentage, the unirradiated USE, the projected neutron fluence (E>1.0 Mev) 1A thickness. If surveillance data was used, provide the surveillance data.

APS Response Upper Shelf Energy (USE) - 10 CFR Part 50 Appendix G requires that the upper shelf energy throughout the life of the vessel be no less than 50 ft-lb. For Palo Verde, based on surveillance data, the lowest USE value at the end of the current license was determined to be as follows:

Unit 1 65.20 ft-lb Unit 2 74 ft-lb Unit 3 70.70 ft-lb The projected USE value at the end of the current license for the beltline materials is summarized in Figure 7-1 (from the NRC RVID). These values represent conditions for PUR (3990 MWt) given the projected fluence at end-of-life, 3.29E+19 n/cm2 , E>l MeV.

The RPV copper content is less than 0.1 wt.%.

The lowest unirradiated USE value was:

Unit 1 83 ft-lb Unit 2 95 ft-lb Unit 3 90 ft-lb References

1) Letter 102-03448 from APS to USNRC, dated August 17,1995, Response to NRC Generic Letter 92-01, Rev. 1, Supplement 1 reported thee requested data.
2) USNRC Reactor Vessel Integrity Data contains USE values at the end of the current licensed life.
3) Letter 102-04139 from APS to USNRC, dated June 24,1998 provides responses to NRC questions that updates information provided in the APS response to GL 92-01, Rev. 1, Supplement 1.
4) Letter 102-04641 from APS to USNRC, dated December 21, 2001, Attachment 6, Sections 5.1.2 & 7.5 discuss the impact of the proposed UPO. Excerpts from 102-04641 are provided in the Attachment to this Enclosure.
5) Letter 102-04834 from APS to USNRC, dated August 29, 2002, Attachment 2, NRC question 7 response provides further discussion on the impact of the proposed UPO. These results are summarized for Units 1, 2, and 3 in the Attachment to this Enclosure.

3

Units 1 & 3 PUR RAI - RV Integrity NRC Question 1.V.- Basis for current P-T limits (applicability in EFPY, 1/4T, and 3/4T ART values).

APS Response Current P-T limits are based on using Peak-end-of-life (i.e. 32 EFPY) fluence of 3.29E+1 9 n/cm 2 . Adjusted reference temperatures (ART) for all beltline materials at the 1/4T, and 3/4T locations were calculated using this end-of-life fluence and the methods in Regulatory Guide 1.99 Revision 02.

There are approximately 14 effective full power years (EFPY) of operation in each of the Palo Verde units, therefore the current P-T curves are more conservative since they were determined using peak end-of-life fluence.

References

1) Letter 102-04834 from APS to USNRC, dated August 29, 2002, Attachment 2, NRC question 7 response discusses the basis for P-T limits and the impact of the proposed UPO. These results are summarized for Units 1, 2, and 3 in The Attachment to this Enclosure.
2) Letter 102-04847 from APS to USNRC, dated October 11, 2002, Attachment 2, NRC Question 17 response provides further discussion on the impact of the proposed UPO.

NRC Question 1.VI.- Projected ART values for the proposed period of applicability using the UPO fluence.

APS Response The calculated ART values for Palo Verde Unit 1, 2 & 3 are based on the peak end-of-life fluence of 3.29E+19 n/cm 2 . The results and conclusions of the assessment of the impact of the proposed UPO on the Analysis of Record (AOR) fluences are that the AOR values will continue to be bounded for current and future operation of Palo Verde including the uprated power condition of 3990 MWt. Therefore, the calculated ART values is applicable for the proposed period using the UPO fluence.

References

1) Letter 102-03448, dated August 17,1995 from APS to USNRC, Response to NRC Generic Letter 92-01, Rev.1, Supplement 1 reported the requested data for Palo Verde Units 1, 2 and 3.
2) Letter 102-04139, dated June 24,1998 from APS to USNRC provides responses to NRC questions that updates information provided in the APS response to GL 92-01, Rev. 1, Supplement 1.
3) Letter 102-04641 from APS to USNRC, dated December 21, 2001, Attachment 6, Sections 5.1.2 & 7.5 discuss the impact of the proposed UPO. Excerpts from 102-04641 are provided in the Attachment to this Enclosure.

4

Units 1 & 3 PUR RAI - RV Integrity

4) Letter 102-04847, dated October 11 ,2002 from APS to USNRC, Attachment 2, NRC Question 17 response provides further discussion on the impact of the proposed UPO.

NRC Question 2.- Discuss the impact of uprated power operations on surveillance capsule program developed in accordance with 10 CFR 50, Appendix H criteria.

APS Response Surveillance Capsule Withdrawal Schedule - 10 CFR Part 50, Appendix H defines the RV surveillance program that is to be used by the licensee to monitor the neutron radiation induced changes in fracture toughness of the vessel during the life of the plant.

It includes requirements to establish a surveillance capsule withdrawal schedule. The schedule was established based on the original calculation of fluence that was shown to bound conditions for PUR (3990 MWt). The detailed surveillance schedule is discussed in UFSAR Section 5.3.1.6.6 and Table 5.3-19. Therefore, the existing surveillance capsule withdrawal schedule remains applicable under conditions for PUR.

References

1) Letter 102-04641 from APS to USNRC, dated December 21, 2001, Attachment 6, Sections 5.1.2 & 7.5 discuss the impact of the proposed UPO on the reactor vessel. APS has determined that no change to the surveillance capsule program is required. Excerpts from 102-04641 are provided in The Attachment to this Enclosure.
2) Letter 102-04834 from APS to USNRC, dated August 29, 2002, Attachment 2, NRC question 7 response provides discussion on the impact of the proposed UPO. These results are summarized for Units 1, 2, and 3 in The Attachment to this Enclosure.
3) Letter 102-04847 from APS to USNRC, dated October 11, 2002, Attachment 2, NRC Question 17 response provides further discussion on the impact of the proposed UPO on the reactor vessel.

NRC Question 3.- Table Matrix-1 of NRC Review Standard RS-001, Revision 0, provides the staff's basis for evaluating the potential impacts for uprated power operations and the subsequent aging effects. In Table Matrix-1, the staff states that, in addition to the Standard Review Plan (SRP), guidance on the neutron irradiation-related threshold levels inducing irradiation assisted stress corrosion cracking (IASCC) in reactor vessel (RV) internal components are given in Westinghouse document, License Renewal Evaluation, Aging Management for Reactor Internals, WCAP-1 4577, Revision 1-A.

WCAP-1 4577, Revision 1-A establishes, a threshold of 1 X 1021 n/cm2 (E 2 0.1 MeV) for the initiation of IASCC, loss of fracture toughness, and/or void swelling in pressurized water reactor (PWR) RV internal components made from stainless steel (including cast austenitic stainless steels) or Alloy 600/82/182 materials. In Table Matrix-1 of NRC 5

Units 1 & 3 PUR RAI - RV Integrity Report RS-001, the staff established guidance that plants exceeding this threshold of neutron irradiation would either have to establish plant-specific degradation management programs for managing the aging effects associated with their RV internals or else indicate that the licensees would participate in industry programs designed for investigating and managing age-related degradation in the RV internal components. Please provide the threshold fluence values for the internals (E > 0.1 MeV) due to UPO. Also, discuss the inspection program that will be implemented by Palo Verde Nuclear Generating Station if the threshold values exceed 1 X 1021 n/cm 2 (E 2 0.1 MeV).

APS Response Arizona Public Service Company (APS) is currently an active participant in the Electric Power Research Institute (EPRI) Materials Reliability Program research initiatives on aging related degradation of reactor vessel internals components. APS commits to:

  • Continue its active participation in the MRP initiative to determine appropriate reactor vessel internals degradation management programs,
  • Evaluate the recommendations resulting from this initiative and implement a reactor vessel internals degradation management program applicable to Palo Verde 1 and 3,
  • Incorporate the resulting reactor vessel internals inspections into the Palo Verde 1 and 3 augmented inspection program as appropriate and provide the internals inspection plan to the NRC staff for information.

In addition, as requested by the NRC, a-description of the program, including the inspection plan, will be submitted to the NRC for review and approval. The submittal date will be within 24 months of the EPRI MRP final recommendations or within five years from the date of issuance of the uprated license, whichever comes first.

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Units 1 & 3 PUR RAI - RV Integrity Attachment 7

Units 1 & 3 PUR RAI - RV Integrity Excerpts from letter 102-04641 from APS to USNRC, dated December 21, 2001, Section 5.1.2 Reactor Vessel Integrity RV integrity is impacted by any changes in plant parameters including the effects of neutron fluence levels (see Section 7.5), RCS temperature, or pressure/temperature transients. The most critical area, in terms of RV integrity, is the beltline region of the RV. Therefore, the changes in neutron fluence resulting from the PUR were evaluated to determine the impact on RV integrity.

The evaluation shows that the heating rates, pressure/temperature transients, and neutron fluence estimates that were used to represent operation at 3800 MWt bound the values at the PUR power level of 3990 MWt. The neutron fluence projections on the RV for the PUR power level will not adversely affect RV integrity AOR (i.e.,

pressure/temperature limits and Pressurized Thermal Shock (PTS) screening limits) for operation at 3990 MWt. Therefore, operation at PUR condition will have no detrimental impact on the RV integrity.

Section 7.5 Neutron Fluence The calculated fluences for the existing AOR assume a core power level of 4200 MWt and an out-in type fuel-loading pattern typical of first cycle operation. The PUR level of 3990 MWt and the low-leakage fuel patterns (used since PVNGS Unit 2 Cycle 2) yield a neutron flux to the shroud and vessel that is lower than considered in the existing AOR (see Section 5.1.2). Therefore, the reactor vessel integrity AOR (i.e.,

pressure/temperature limits and Pressurized Thermal Shock (PTS) screening limits) are not affected by PUR operation at 3990 MW. In addition, fuel management guidelines for PUR cycles are set to ensure that the vessel fluence is bounded by the AOR..

8

Units 1 & 3 PUR RAI - RV Integrity Letter 102-04834 from APS to USNRC, dated August 29, 2002, Attachment 2, modified for PVNGS Units 1, 2, and 3.

NRC Question 7:

The PURLR does not discuss the power-uprate-related effects on RV integrity. Discuss the effect of the PUR on the following for Unit 2: pressurized thermal shock, fluence evaluation, heat-up and cooldown pressure temperature limit curves, low temperature overpressure protection, upper shelf energy, and surveillance capsule withdrawal schedule.

APS Response:

The factors influencing Reactor Vessel (RV) integrity are the initial properties of the materials and the neutron fluence incident on the materials. PUR does not affect the initial material properties, but the neutron fluence can change. The effect of neutron fluence changes on vessel integrity is assessed below using 10 CFR Part 50, Appendices G and H, and 10 CFR Part 50.61.

a) Pressurized Thermal Shock (PTS) - The screening criteria in 10 CFR Part 50.61 is 270 OF for plates, forgings, and axial weld materials, and 300 OF for circumferential weld materials. The highest RTPTS value for a plate from the intermediate shell course of the RV for Palo Verde at the end of the current license was determined to be as follows:

Unit 1 122.50 F Unit 2 78 OF Unit 3 68.1 OF The projected RTPTS value at the end of the current license for the beltline materials are summarized in Figure 7-2 (from the NRC Reactor Vessel Integrity Database (RVID). These values represent conditions for PUR (3990 MWt) given that the projected fluence at end-of-license, 3.29E+1 9 n/cm2 , E>l MeV, is bounded by the Analysis of Record (AOR) and the method fo. predicting RTPTS unchanged.

b) Vessel Fluence Evaluation - The AOR end-of-life fluence is 3.29E+19 n/cm2 for the vessel inside surface. The AOR is based on a core power level of 4200 MWt.

The analyses for Palo Verde were issued as follows:

Unit 1 Letter 102-04500 from APS to USNRC, dated October 20, 2000 transmitted WCAP-15589 (Analysis of 38 Degree Capsule from the Arizona Public Service Company Palo Verde Unit No. 1 Reactor Vessel Radiation Surveillance Program)

Unit 2 Letter 102-02919 from APS to the USNRC, dated April 15,1994 transmitted WCAP-1 3935 (Analysis of 137 Degree Capsule from the Arizona Public Service Company Palo Verde Unit No. 2 Reactor Vessel Radiation Surveillance Program).

Unit 3 Letter 102-03340 from APS to USNRC, dated April 26, 1995 transmitted WCAP-1 4208 (Analysis of 137 Degree Capsule from the Arizona Public Service Company Palo Verde Unit No. 3 Reactor Vessel Radiation Surveillance Program) 9

Units 1 & 3 PUR RAI - RV Integrity Based on those analyses, the 32 EFPY peak azimuthal fluence for the vessel inside surface is as follows:

Unit 1 1.725E+19 n/cm2 Unit 2 2.047E+1 9 n/cm2 Unit 3 2.047E+1 9 n/cm2 The WCAP's analyses showed that the projected end-of-life (32 EFPY) fluence was approximately one-third lower than the value in the AOR (i.e., one-third more conservative than the assessment done for the PUR submittal). The large difference between the AOR and the WCAP's analyses is based on the fact that the latter did account for actual plant operation, and much of the difference is a reflection of the low leakage fuel management program employed. The PUR submittal concerning vessel fluence was based on the AOR and showed that value to be bounding.

c) Heat-up and Cool-down Pressure Temperature Limit Curves and Low Temperature Overpressure Protection - 10 CFR Part 50 Appendix G addresses the limits on pressure and temperature that are placed on heatup and cool-down during normal operation. There are no changes to the values used to establish the Appendix G normal operating limits. The limits represent conditions for PUR (3990 MWt) given that the projected fluence at end-of-license, 3.29E+1 9 n/cm2 ,

E>l MeV, is bounded by the AOR such that the predicted vessel material properties used to establish the heat-up and cool-down limits are unchanged.

The low temperature overpressure protection limits for PUR conditions are unchanged for those same reasons.

d) Upper Shelf Energy (USE) - 10 CFR Part 50 Appendix G requires that the upper shelf energy throughout the life of the vessel be no less than 50 ft-lb. For Palo Verde, the lowest USE value at the end of the current license was determined to be as follows:

Unit 1 65.20 ft-lb Unit 2 74 ft-lb Unit 3 70.70 ft-lb The projected USE value at the end of the current license for the beltline materials is summarized in Figure 7-1 (from the NRC RVID). These values represent conditions for PUR (3990 MWt) given the projected fluence at end-of-life, 3.29E+1 9 n/cm2 , E>l MeV. ;

e) Surveillance Capsule Withdrawal Schedule - 10 CFR Part 50 Appendix H defines the RV surveillance program that is to be used by the licensee to monitor the neutron radiation induced changes in fracture toughness of the vessel during the life of the plant. It includes requirements to establish a surveillance capsule withdrawal schedule. The schedule was established based on the original calculation of fluence that was shown to bound conditions for PUR (3990 MWt).

The detailed surveillance schedule is discussed in UFSAR Section 5.3.1.6.6 and Table 5.3-19. Therefore, the existing surveillance capsule withdrawal schedule remains applicable under conditions for PUR.

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NRC - Reactor Vessel Integrity Database Printed 10/20/2004 12:56:2 PTS Summary Report Docket No: 50-528 Page 1 PALO VERDE 1 EOL Date: 12131/2024 Neutron Fluence Befline Identlifcation RTpts Fluence ARTndt(u) Factor Chem Chemistry Factor Type Heat ID OEOL 0 EOL RTndt(u) RTndt(u) METHOD 0 EOL 0 EOL Factor Method Margin Margin Method Cu % NI% P% S%

LOWER SHELL M-4311-1 58.1 3.290 *10.0 PLANT SPECIFIC 34.1 1.312 26.00 TABLE 34.0 POSmON 1.1 0.040 0.650 0.004 0.003 PLATE 6 62467-1 (NO S DATA)

LOWER SHELL M-4311-3 32.5 3.290 -20.0 PLANT SPECIFIC 28.2 1.312 20.00 TABLE 26.2 OVERRIDE 0.030 0.640 0.004 0.005 PLATE 62722-1 LOWER SHELL M-4311-2 12.5 3.290 -40.0 PLANT SPECIFIC 26.2 1.312 20.00 TABLE 26.2 OVERRIDE 0.030 0.620 0.005 0.007 PLATE l 62817-1 .

INTERMEDIATE SHELL M-6701-1 121.7 3.290 30.0 PLANT SPECIFIC 57.7 1.312 44.00 TABLE 34.0 POSmON 1.1 0.070 0.660 0.005 0.018 PLATE I C4142-1 (NO S DATA)

INTERMEDIATE SHELL M-6701-3 122.5 3.290 40.0 PLANT SPECIFIC 48.5 1.312 37.00 TABLE 34.0 POSITION 1.1 0.060 0.610 0.004 0.016 PLATE l -C4188-1 (NO S DATA)

INTERMEDIATE SHELL M-6701-2 122.5 3.290 40.0 PLANT SPECIFIC 48.5 1.312 37.00 TABLE 34.0 POSmON 1.1 0.060 0.610 0.004 0.017 PLATE l C418a-2 (NO S DATA)

INTERMEDIATE SHELL AXIAL WELDS 101-124A&B,C 30.6 3.290 -50.0 PLANT SPECIFIC 40.3 1.312 30.74 TABLE 40.3 OVERRIDE 0.047 0.049 0.010 0.008 WELD l . 4PO052 .

CIRC. WELD 101-171 5.4 3.290 -700 PLANT SPECIFIC 37.7 1.312 2.73 TABLE 37.7 OVERRIDE 0.031 0.098 0.013 0.009 WELD l 4P7869 _

LOWER SHELL AXIAL WELDS 101-142A.B.C

  • 3.0 3.290 -80.0 PLANT SPECIFIC 38.5 1.312 29.32 TABLE 38.5 OVERRIDE 0.035 0.079 0.005 .

WELD 90071 _ _

plant References and Beltline Material Notes NOTE: Margin method for all welds Is'override' since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99, Rev. 2.

Chemical composition data are from the June 24, 1998 letter from J.M. Levine (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1,2.3 Response to Request for Additional Information Regarding RPV Integrity at PVNGS.

Information on the beftline plate heat numbers, and weld wire heat numbers are from the August 17, 1995 letter from W.L. Stewart (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1,2 and 3, Response to NRC Generic Letter 92-01. Revision 1, Supplement 1.

Plate UUSE data are from Table 5.2-SA of the FSAR, and weld LUSE values are from Charpy Curves of FSAR.

Fluence and RTndt(u) data are from the January 31, 1989, letter from D.B. Kamer (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3, Generic Letter 88-11, Radiation Embrittlement of Reactor Vessel Materials.

Margin method for plate M-4311-3 (heat number 62722-1) is override since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99, Rev. 2.

Margin method for plate M-4311-2 (heat number 62817-1) is override since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99. Rev. 2.

UUSE value for intermediate shell axial welds 101-124A,B,C (heat number 4P6052) Is the average of three (3)Charpy Impact tests for Palo Verde 1weld wire/flux (Linda 0091) data from weld certified material test report (CMTR).

Certified material test report (CMTR) USE data for heat 4P6052 Is from letter dated July 25, 1994.

NRC - Reactor Vessel Integrity Database Upper Shelf Energy Summary Report Printed 10/20/2004 12:59:3 Docket No: 50-528 Page 1 PALO VERDE 1 EOL Date: 12/31/2024 elinIdnfcaon114 T %Drp In BettHtne IdenU Iaaatlon USE 0 EOL NeultrnFluence Unirradiated Unlrradlated USE 0 EOL %Drop In USE Type HealID Material1Type 4T 0 EOL USE USE Method 0 114T Method Cu %

LOWERSHELL M-4311-1 A 533B 105.27 1.681, 134.00 DIRECT 21.44 POSITION 1.2 (NO S 0.040 PLATE l 62467-1 DATA)

LOWER SHELL M-4311-3 A 533B 111.56 1.681 142.00 DIRECT 21.44 POSITION 1.2 (NO S 0.030 PLATE l 62722-1 . DATA)

LOWER SHELL M-4311-2 A 533B 99.77 1.681 127.00 DIRECT 21.44 POSITION 1.2 (NO S 0.030 PLATE l 62817-1 . DATA)

INTERMEDIATE SHELL M-8701-1

  • A 533B 65.20 1.681 83.00 DIRECT 21.44 POSITION 1.2 (NO S 0.070 PLATE I C4142-1 _ DATA)

INTERMEDIATE SHELL M-670t-3 A 533B 78.56 1.681 100.00 DIRECT 21.44 POSITION 1.2 (NO S 0.060 PLATE I C4188-1 DATA)

INTERMEDIATE SHELL M-6701-2 A 5338 75A2 1.681 96.00 DIRECT 21.44 POSITION 1.2 (NO S 0.060 PLATE I C4188-2 DATA)

INTERMEDIATE SHELLAXIALWELDS 101-124AB.C LINDE0091 157.12 1.681 200.00 DIRECT 21.44 POSITION 1.2 (NOS 0.047 i WELD lT - 7-4P6052 DATA)

CIRC. WELD 101-171 UNDE 124 70.70 1.681 90.00 DIRECT 21.44 POSION 1.2 (NO S 0.031 I

1. i WELD l 4P7869 .  ; DATA) I LOWER SHELL AXIAL WELDS 101-142ABC LINDEOO9I 109.98 1.681 140.00 DIRECT 21.44 POSmON 1.2 (NO S 0.035 I WELD l 90071 DATA) I I

Plant References and Beltilne Material Notes NOTE: Margin method for all welds Is override' since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99, Rev. 2.

Chemical composition data are from the June 24. 1998 letter from J.M. Levine (APS) to the USNRC Document Control Desk, subjoct: Palo Verde Nuclear Generating Station (PVNGS) Units 1,2,3 Response to Request for Additional Information Regarding RPV Integrity at PVNGS.

Information on the beltflne plate heat numbers, and weld wire heat numbers are from the August 17 1995 letter from W.L. Stewart (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1,2 and 3, Response to NRC Generic Letter 92-01. Revision 1, Supplement 1.

Plate UUSE data are from Table 5.2-5A of the FSAR. and weld LUSE values are from Charpy Curves of FSAR.

Fluence and RTndt(u) data are from the January 31. 1989, letter from D.B. Kamer (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1.

2, and 3, Generic Letter 88-11, RadiatIon Embritlement of Reactor Vessel Materials.

Margin method for plate M-4311-3 (heat number 62722-1) Is 'override since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99. Rev. 2.

Margin method for plate M-4311-2 (heat number 62817.1) is override since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99. Rev. 2.

UUSE value for intermediate shell axial welds 101 .124A,8,C (heat number 4P6052) Is the average of three (3)Charpy Impact tests for Palo Verde 1 weld wire/flux (Linde 0091) data from weld certified material test report (CMTR).

Certified material test report (CMTR) USE data for heat 4P6052 is from letter dated July 25 1994.

NRC - Reactor Vessel Integrity Database Printed 10127/2004 9:18:09 PTS Summary Report Docket No: 50-529 Page 1 PALO VERDE 2 EOL Date: 12109/2025 Neutron Fluence Beffline Identificatlon RTpts Fluence ARTndt(u) Factor Chem Chemistry Factor i

Type Heat ID OEOL 0 EOL RTndt(u) RTrdt(u) METHOD 0 EOL 0 EOL Factor Method Margin Margin Method Cu% Ni % P% S% I INTERMEDIATE SHELL F-765-4 32.5 3.290 -20.0 PLANT SPECIFIC 26.2 1.3 12 20.00 TABLE 26.2 OVERRIDE 0.030 0.670 0.003 0.005 PLATE l 63427-1 l _ _ _ I INTERMEDIATE SHELL F-765-5 62.5 3.290 10.0 PLANT SPECIFIC 26.2 1.312 20.00 TABLE 26.2 OVERRIDE 0.030 0.650 0.004 0.007 i PLATE l 63464-1_

INTERMEDIATE SHELL F-765-6 78.1 3.290 10.0 PLANT SPECIFIC 34.1 1.312 26.00 TABLE 34.0 POSmON 1.1 0.040 0.670 0.002 0.004 PLATE 63716-1 (NO S DATA)

LOWER SHELL F-773-3 14.7 3.290 -60.0 PLANT SPECIFIC 40.7 1.312 31.00 TABLE 34.0 POSmON 1.1 0.050 0.660 0.004 0.009 PLATE l 639ff-1 l l (NO S DATA) i LOWER SHELL F-773-2 68.1 3.290 0.0 PLANT SPECIFIC 34.1 1.312 26.00 TABLE 34.0 POSITION 1.1 0.040 0.640 0.003 0.008 i

PLATE l 64065-1 l (NO S DATA)

LOWER SHELL FP773-1 62.5 3.290 10.0 PLANT SPECIFIC 26.2 1.312 20.00 TABLE 26.2 OVERRIDE 0.030 0.670 0.003 0.008 i

PLATE l 64071-1 .I _ _ _

LOWER SHELL AXIALWELDS 101-142A.9,C 28.0 3.290 -80.0 PLANT SPECIFIC 54.0 1.312 41.17 TABLE 54.0 OVERRIDE 0.074 0.067 0.009 0.011 WELD l37317 _ ___

CIRC WELD 101-171 . 45.4 3.290 -30.0 PLANT SPECIFIC 37.7 1.312 28.73 TABLE 37.7 OVERRIDE 0.031 0.096 0.012 0.009 :71 WELD T~4F`7869 _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _

"-- 1 i INTERMEDIATE SHELL AXIALIWELDS 101-124ABC 22.6 3.290 -60.0 PLANT SPECIFIC 41.3 1.312 31.51 TABLE 41.3 OVERRIDE 0.046 0.059 0.008 0.012 .1 - I-':.: I WELD l 89833 .

Plant References and Beltline Material Notes NOTE: Margin method for all beltline welds Is 'override' since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99. Rev. 2.

Chemical composition data are from the June 24, 1998 letter from J.M. Levine (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1,2,3 Response to Request for Additional Information Regarding RPV Integrity at PVNGS.

Information on tho boltlino plate heat numbers, and weld wire heat numbers are from the August 17, 1995 letter from W.L. Stewart (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1.2 and 3, Response to NRC Generic Letter 92-01. Revision 1, Supplement 1.

Plate LUSE data are from Table 5.2-5A of FSAR. and weld UUSE values are from Charpy Curves of FSAR.

Fluence and RTndt(u) data are from the January 31 1989, letter from D. B. Kamer (APS) to the USNRC Document Control Desk, subject Palo Verde Nuclear Generating Station (PVNGS) Units 1 2, and 3, Generic Letter 88-11, Rndlatfcn Embrittlement of Reactor Vessel Materials.

Margin method for plate F-765-4 (heat number 63427-1) Is 'overrlde since sigma delta need not be greater than 1/2 delta RTndt per FIG 1.99. Rev. 2.

Margin method for plate F-765-5 (heat number 63464-1) Is 'override' since sigma delta need not be greater than 1/2 delta RTndt per FIG 1.99, Rev. 2.

Margin method for plate F-773-1 (heat number 64071-1) is 'override' since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99, Rev. 2.

NRC - Reactor Vessel Integrity Database Printed 10/27/20049:18:31 Upper Shelf Energy Summary Report Docket No: 50-529 Page 1 PALO VERDE 2 EOL Date: 12/09/2025 114T %Drop In Bettitne Identification USE 0 EOL NeuroaFnuence UnIrradated Unlrmrdlated USE 0 EOL %Drop In USE Type Heat ID Material Type 0 1/4T 0 EOL USE USE Method 0 114T Method Cu %

INTERMEDIATE SHELL FP765-4 A 5330 89.56 1.681 114.00 DIRECT 21.44 POSMON 1.2 (NO S 0.030 PLATE 63427.1 ] _ DATA)

INTERMEDIATE SHELL F-765-5 A 533B 95.06 1.681 121.00 DIRECT 21.44 POSTON 1.2 (NO S 0.030 PLATE I 63464-1 DATA)

INTERMEDIATE SHELL F-7658 A 533B 98.99 1681 126.00 DIRECT 21.44 POSITION 1.2(NOS 0.040 PLATE l 63716-1 DATA)

LOWER SHELL F-773-3 A 5338 101.34 1.681 129.00 DIRECT 21.44 POSITION 1.2 (NO S 0.050 PLATE 8 i3s9871 DATA)

LOWER SHELL F-773-2 A 533B 99.77 1.681 127.00 DIRECT 21.44 POSITION 1.2 (NOS 0.040 PLATE 64065-1 DATA)

LOWER SHELL F-773-1 A 533B 82A9 1.681 105.00 DIRECT 21.44 POSION 1.2 (NO S 0.030 PLATE 64071-1 _ _ DATA)

LOWER SHELLAXIAL WELDS 101-142AB.C LINDE 124 76.06 1.681 100.00 DIRECT 23.94 POSTIN 1.2 (NO S 0.074 W ELD 3P 3 7_ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _DATA)

CIROWELD 101-17t UNDE 124 74.63 .1.681 95.00 DIRECT 21.44 POSITION 1.2 (NOS 0.031 WELD 4P7869 D.ATA)

INTERMEDIATE SHELL AXIAL WELDS t01-124AB,C UNDE 124 78.56 1.681 100.00 DIRECT 21.44 POSITION 1.2 (NO S 0.046 WELD l 89833 . DATA)

Plant References and Beltilne Material Notes NOTE: Margin method for all beltllne welds Is 'overrfde' since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99, Rev. 2.

Chemical composition data are from the June 24. 1998 letter from J.M. Levine (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1,2,3 Response to Request for Additional Information Regarding RPV Integrity at PVNGS.

Information on the beltline plate heat numbers, and weld wire heat numbers are from the August 17, 1995 letter from W.L. Stewart (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1,2 and 3, Response to NRC Generic Letter 92-01. Revision 1, Supplement 1.

Plate UUSE data are from Table 5.2-5A of FSAR, and weld UUSE values are from Charpy Curves of FSAR.

Fluence and RTndt(u) data are from the January 31, 1989, letter from D. B. Kamer (APS) to the USNRC Document Control Desk. subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1.

2, and 3. Generic Letter 88-11, PFadlation Embrititement of Reactor Vessel Materials.

Margin method for plate F-765-4 (heat number 63427-1) Is override' since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99. Rev. 2.

Margin method for plate F-765-5 (heat number 63464-1) Is override since sigma delta need not be greater than 1/2 de'ta RTndt per RG 1.99, Rev. 2.

Margin method for plate F-773-1 (heat number 64071-1) Is 'override since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99, Rev. 2.

NRC - Reactor Vessel Integrity Database Printed 10/20/2004 2:45:29 PTS Summary Report Docket No: 50-530 Page 1 PALO VERDE 3 EOL Date: 03/25/2027 Neutron Fluence Bentine Identification RTpts Fluence 6RTndt(u) Factor Chem Chemistry Factor Type Heat ID a EOL 0 EOL RTndt(u) RTndt(u) METHOD 0 EOL 0 EOL Factor Method Margin Margin Method Cu % Ni % P % S%

INTERMEDIATE SHELL F-6407-4 38.1 3.290 -30.0 PLANT SPECIFIC 34.1 1.312 26.00 TABLE 34.0 POSITION 1.1 0.040 0.620 0.002 0.005 PLATE l 65202-1 (NO SDATA)

INTERMEDIATE SHELL F-6407-5 54.7 3.290 -20.0 PLANT SPECIFIC 40.7 1.312 31.00 TABLE 34.0 POSITION 1.1 0.050 0.610 0.002 0.005 PLATE l 65219-1 . (NO S DATA)

INTERMEDIATE SHELL F-6407-8 48.1 3.290 -20.0 PLAN1 SPECIFIC 34.1 1.312 26.00 TABLE 34.0 POSITION 1.1 0.040 0.610 0.002 0.004 PLATE l 79011-1 (NO SDATA)

LOWER SHELL F-6411-1 28.1 3.290 -40.0 PLANT SPECIFIC 34.1 1.312 26.00 TABLE 34.0 POSmON 1.1 0.040 0.640 0.004 0.007 PLATE l 79545-1 (NO S DATA)

LOWER SHELL F-6411-3 8.1 3.290 -60.0 PLANT SPECIFIC 34.1 1.312 26.00 TABLE 34.0 POSITION 1.1 0.040 0.660 0.007 0.018 PLATE 79659-1 LOWER SHELL F-6411-2 68.1 3.290 0.0 PLANT SPECIFIC 34.1 1.312 26.00 TABLE 34.0 POSITION 1.1 0.040 0.650 0.004 0.013 PLATE l 79745-1 (NO SDATA)

CIRC. WELD 101-171 5.4 3.290 -70.0 PLANT SPECIFIC 37.7 1.312 28.73 TABLE 37.7 OVERRIDE 0.031 0.096 0.008 0.011 WELD l 4P7869 INTERMEDIATE SHELL AXIAL WELDS 101-124A.B,C 25.4 .3.290 -50.0 PLANT SPECIFIC 37.7 1.312 28.73 TABLE 37.7 OVERRIDE 0.031 0.096 0.010 0.008 " I -,-

WELD 4P7869 _ ___

LOWER SHELL AXIAL WELDS 101-142A.B.C 25.4 3.290 -50.0 PLANT SPECIFIC 37.7 1.312 28.73 TABLE 37.7 OVERRIDE 0.031 .06 0.008 0.012 WELD 4P7869 .____0 .

Plant References and Beltilne Material Notes NOTE: Margin method for all beltline welds Is 'override since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99, Rev. 2.

Chemical composition data are from the June 24, 1998 letter from J.M. Levine (APS) to the USNRC Documenl Control Desk, subject: Palo Verdo Nuclear Generating Station (PVNGS) Units 1,2,3 Response to Request for AdditIonal Information Regarding RPV Integrity at PVNGS.

Information on the beltline plate heat numbers, and weld wire heat numbers are from the August 17, 1995 letter from W.L. Stewart (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1,2 and 3, Response to NRC Generic Letter 92-01. Revision 1, Supplement 1.

Plate UUSE data are from Table 5.2-5A of FSAR, and weld UUSE values are from Charpy Curves of FSAR.

Fluence and RTndt(u) data are from the January 31, 1989, letter from D. B. Kamer (APS) to the USNRC Docurrent Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3, Generic Letter 88-11, Radiation Embrittlement of Reactor Veseel Materials.

NRC - Reactor Vessel Integrity Database Printed 1Upper Shelf Energy Summary Report Pdtd10/20/2004 2:51:46 Docket No: 50-530 Page 1 PALO VERDE 3 EOL Date: 0312512027 1/4 T %Drop In Beltline Identification USE 0 EOL NeutronFluence UnIrradlated UnIrradlated USE 0 EOL %Drop inUSE Type Heat ID Materal Type 0 1/4T 0 EOL USE USE Method 0 114T Method Cu %

INTERMEDIATE SHELL F-6407-4 A 5338 101.34 1.681 129.00 DIRECT 21.44 POSmON 1.2 (NOS 0.040 PLATE l 65202-1 _ _ _ _ DATA)

INTERMEDIATE SHELL F-6407-5 A 5338 89.56 1.681 114.00 DIRECT 21A4 POSITION 1.2 (NO S 0.050 PLATE l 65219-1 DATA)

INTERMEDIATE SHELL F-6407.6 A 5331 104.48 1.681 133.00 DIRECT 21.44 POSmON 1.2 (NO S 0.040 PLATE l 79011-1 DATA)

LOWER SHELL F-64tt-t A 5330 122.55 1.681 156.00 DIRECT 21.44 POSION 1.2 (NO S 0.040 PLATE l 79545-1 DATA)

LOWER SHEUL F-6411-3 A 5338 84.08 1.681 107.00 DIRECT 21.44 POSITION 1.2 (NO S 0.040 PLATE i 79659.1 DATA)

LOWER SHELL F-6411-2 A 5330 87.20 1.681 111.00 DIRECT 21.44 POSTON 1.2 (NO S 0.040 PLATE 79745-1 DATA)

CIRC. WELD 101-171 UNDE 124 SAW 70.70 1.681 90.00 DIRECT 21.44 POSmON 1.2 (NO S 0.031 WELD 4p790169 ______________ DATA)

INTERMEDIATE SHELLAXIAL WELDS 101-124ABC UNDE 124 SAW 78.56 1.681 100.00 DIRECT 21.44 POSMON 1.2 (NO S 0.031 WELD I 4P7869 . DATA):.

LOWER SHELL AXIAL WELDS 101-142A.BC UNDE 124 SAW 78.56 1.681 100.00 DIRECT 21.44 POSITION 1.2 (NO S 0.031 WELD 4P7869 DATA) I Plant References and Beltifne Material Notes NOTE: Margin method for all beltline welds Is 'override since sigma delta need not be greater than 1/2 delta RTndt per RG 1.99, Rev. 2.

Chemical composition data are from the June 24, 1998 letter from J.M. Levine (APS) to the USNRC Document Cortrol Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1,2,3 Response to Request for Additional Information Regarding RPV Integrity at PVNGS.

Information on the belttine plate heat numbers, and weld wire heat numbers are from the August 17, 1995 letter from W.L. Stewart (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1,2 and 3, Response to NRC Generic Letter 92-01. Revision 1. Supplement 1.

Plate UUSE data are from Table 5.2-5A of FSAR, and weld UUSE values are from Charpy Curves of FSAR.

Fluence and RTndt(u) data are from the January 31, 1989, letter from D. B. Kamer (APS) to the USNRC Document Control Desk, subject: Palo Verde Nuclear Generating Station (PVNGS) Units 1,

2. and 3, Generic Letter 88-11. RadiatIon Embrittlement of Reactor Vessel Materials.

i

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