BAW-1543, Revision 4, Supplement to the Master Integrated Reactor Vessel Surveillance ... · 2018. 7. 5. · • BAW-1543, Revision 4, Supplement 6 reflecting updates that were made

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Supplement to the Master Integrated Reactor Vessel Surveillance Program

Supplement 7-A

March 2018

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BAW-1543 Revision 4

Supplement 7-A

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BAW-1543 Revision 4

UNITED STATES NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

March 27, 2018

Mr. W. Anthony Nowinowski, Program Manager PWR Owners Group, Program Management Office Westinghouse Electric Company 1000 Westinghouse Drive, Suite 380 Cranberry Township, PA 16066

SUBJECT: FINAL SAFETY EVALUATION FOR PRESSURIZED WATER REACTOR OWNERS GROUP TOPICAL REPORT BAW-1543, REVISION 4, SUPPLEMENT 7, "SUPPLEMENT TO THE MASTER INTEGRATED REACTOR VESSEL SURVEILLANCE PROGRAM"

Dear Mr. Nowinowski:

By letter dated January 23, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15033A086), the Pressurized Water Reactors Owners Group (PWROG), submitted Topical Report (TR) BAW-1543, Revision 4, Supplement 7, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" to the U.S. Nuclear Regulatory Commission (NRC) for review and acceptance for referencing in regulatory actions. The revisions contained in this supplement were necessary to show updates to the surveillance capsule withdrawal schedules and the change of status of Capsules A2 and A4 to "withdrawal not planned."

The NRC staff has found the revised and updated information to the withdrawal schedules and change in capsule status, as indicated in Topical Report BAW-1543, Revision 4, Supplement 7, acceptable for referencing in licensing applications for the B&W-designed 177-FA plants and the Westinghouse-designed plants with B&W-fabricated reactor vessels to the extent specified and under the limitations delineated in the TR and in the enclosed final SE. The final SE defines the basis for our acceptance of the TR.

The proposed withdrawal schedules satisfy the requirements of Appendix H to Title 10 of the Code of Federal Regulations (10 CFR) Part 50 and are consistent with XI.M31, "Reactor Vessel Surveillance," in NUREG-1801, the Generic Aging Lessons Learned Report, for the license renewal period.

Our acceptance applies only to material provided in the subject TR. We do not intend to repeat our review of the acceptable material described in the TR. When the TR appears as a reference in licensing action requests, our review will ensure that the material presented applies to the specific plant involved. Requests for licensing actions that deviate from this TR will be subject to a plant-specific review in accordance with applicable review standards.

W. Nowinowski - 2 -

In accordance with the guidance provided on the NRC website, we request that the PWROG publish approved proprietary ~nd non-proprietary versions of TR BAW-1543, Revision 4, Supplement 7, within 3 months of receipt of this letter. The approved versions shall incorporate this letter and the enclosed final SE after the title page. Also, they must contain historical review information, including NRC requests for additional information and your responses. The approved versions shall include an "-A" (designating approved) following the TR identification symbol.

As an alternative to including the request for additional information (RAls) and RAI responses behind the title page, if changes to the TR were provided to the NRC staff to support the resolution of RAI responses, and if the NRC staff reviewed and approved those changes as described in the RAI responses, there are two ways that the accepted version can capture the RAls:

1. The RAls and RAI responses can be included as an Appendix to the accepted version. 2. The RAls and RAI responses can be captured in the form of a table (inserted after the final SE) which summarizes the changes as shown in the approved version of the TR. The table should reference the specific RAls and RAI responses which resulted in any changes, as shown in the accepted version of the TR.

If future changes to the NRC's regulatory requirements affect the acceptability of this TR, PWROG will be expected to revise the TR appropriately or justify its continued applicability for subsequent referencing. Licensees referencing this TR would be expected to justify its continued applicability or evaluate their plant using the revised TR.

Docket No. 99902037

Enclosure: Final Safety Evaluation (Non-Proprietary)

Sincerely,

/RAJ

Dennis C. Morey, Chief Licensing Processes Branch Division of Licensing Processes Office of Nuclear Reactor Regulation

W. Nowinowski - 3 -

SUBJECT: FINAL SAFETY EVALUATION FOR PRESSURIZED WATER REACTOR OWNERS GROUP TOPICAL REPORT BAW-1543, REVISION 4, SUPPLEMENT 7, "SUPPLEMENT TO THE MASTER INTEGRATED REACTOR VESSEL SURVEILLANCE PROGRAM" DATED: MARCH 27, 2018

DISTRIBUTION: PUBLIC RidsResOd RidsNrrDlp RidsNrrLADHarrison RidsOgcMailCenter RidsNrrDlpPlpb RidsNrrDe RidsNrrDeEmib BBenney RidsNroOd

RidsACRS_MailCTR SRuffin DMorey

ADAMS Accession Nos.: ML 18038A347 (Letter); ML 18038A345 (Enclosure); ML 18038a332 (Package); *concurrence via e-mail NRR-106

OFFICE NRR/DLP/PLPB NRR/DLP/PLPB* NRR/DE/EMIB* NRR/DLP/PLPB

NAME BBenney DHarrison SRuffin DMorey

DATE 3/26/18 3/14/18 3/6/18 3/27/18

OFFICIAL RECORD COPY

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

TOPICAL REPORT (TR) BAW-1543 REVISION 4, SUPPLEMENT 7,

"SUPPLEMENT TO THE MASTER INTEGRATED REACTOR VESSEL

SURVEILLANCE PROGRAM"

PRESSURIZED WATER REACTOR OWNERS GROUP

PROJECT NO. 694

1.0 INTRODUCTION AND BACKGROUND

By letter dated January 23, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15033A086), and supplemented by letter dated August 4, 2016 (ADAMS Accession No. ML 17256A514), the Pressurized Water Reactor (PWR) Owners Group (PWROG), submitted for U.S. Nuclear Regulatory Commission (NRC) approval, Technical Report BAW-1543, Revision 4, Supplement 7, "Supplement to the Master Integrated Reactor Vessel Surveillance Program." The revisions contained in this supplement address changes to surveillance capsule withdrawal schedules. BAW-1543, Revision 4, Supplement 7, reported the essential features of the Master Integrated Reactor Vessel Surveillance Program (MIRVP) for all operating Babcock and Wilcox (B&W) 177-fuel assembly (FA) plants and those participating Westinghouse Electric Company (Westinghouse) plants having B&W-fabricated reactor vessels (RVs). These RVs include seven B&W-designed 177-FA plants and six Westinghouse designed plants with B&W-fabricated RVs. The program was built upon the integrated surveillance program (ISP) developed by the B&W Owner's Group (B&WOG) for the B&W 177-FA plants. All 13 reactors are of the same basic design concept: PWR, operating at about 550° F and 2250 pounds per square inch (psi) nominal inlet temperature and pressure, and with low enrichment fuel (approximately 2 to 4 percent enrichment).

The irradiation schedules for the MIRVP include the plant-specific capsules, supplementary weld metal surveillance capsules and higher neutron fluence supplementary weld metal surveillance capsules. All the irradiations, with the exception of Capsule W1 and the Westinghouse plant-specific capsules, are performed in the B&W host reactors, Crystal River, Unit 3 and Davis-Besse. The Westinghouse plant-specific capsules are irradiated in their respective plants. Capsule W1, of Westinghouse design, was irradiated in Surry, Unit 2. This capsule was created to provide a comparison between the effects of the irradiation environments on the test data between the Westinghouse-designed and the B&W-designed

. RVs to determine whether the data could be used interchangeably between the designs.

The MIRVP is an ISP that was developed to include all PWRs in the Unites States that contain Linde 80 submerged-arc welds. NRC staff found that the ISP criteria, as provided by Appendix H to Title 10 to the Code of Federal Regulations (10 CFR), Part 50, "Reactor Vessel Material Surveillance Program Requirements," were met for the MIRVP. Highlights of changes

. between the NRC-approved revisions to the MIRVP include:

• By letter dated June 11, 1991, BAW-1543, Revision 3, was approved for use by the NRC.

• BAW-1543, Revision 4, was the same as the approved Revision 3 with the exception of updated withdrawal schedules for some units.

Enclosure

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• BAW-1543, Revision 4, Supplement 1, contained quantitative information which was, in general, neutron fluence dependent and, therefore, subject to change. This revision reflected revised neutron fluence values for some units and revised some withdrawal schedules to comply with the 1973 Edition of American Society for Testing and Materials (ASTM) Standard E 185, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels" (ASTM E 185-73). It was anticipated that future updates to BAW-1543(NP) would only involve changes to the Revision 4 Supplement.

• BAW-1543(NP), Revision4, Supplement 2, reflected the revised neutron fluence values and the revised withdrawal schedules.

• Revising BAW-1543, Revision 4, Supplement 2 with Revision 4, Supplement 3, the B&WOG deleted Rancho Seco, Ginna, and Zion, Units 1 and 2, from the program. In addition, the B&WOG updated the surveillance capsule status and the peak end-oflicense neutron fluences for several plants.

• In BAW-1543, Revision 4, Supplement 4, the B&WOG incorporated the disposal plan for stored capsules, updated the status for various capsules, and incorporated current neutron fluence levels.

• BAW-1543, Revision 4, Supplement 5 modified the previous supplement to include a commitment regarding Capsules OC1-D and OC3-F; however, that commitment could not be met because these capsules could not be removed from Crystal River, Unit 3. The NRC staff approved the revised withdrawal schedules for Oconee, Units 1, 2, and 3, and Three Mile Island, Unit 1, in this submittal.

• BAW-1543, Revision 4, Supplement 6 reflecting updates that were made to neutron fluence values and to the surveillance capsule withdrawal schedules.

PWROG submitted BAW-1543, Revision 4, Supplement 7, which reflects a change in status of surveillance capsules A2 and A4 to "withdrawal not planned" and a change in status of surveillance capsule TE1-C to "withdrawn and tested."

2.0 REGULATORY REQUIREMENTS

The nuclear power plant licensees are required by Appendix H to 1 O CFR Part 50 to implement RV material surveillance programs to "monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region: .. which result from exposure of these materials to neutron irradiation and the thermal environment." Two specific alternatives are provided with regard to the design of a facility's RV surveillance program which may be used to address the requirements of Appendix H to 10 CFR Part 50. The first alternative is the implementation of a plant-specific RV surveillance program consistent with the requirements of ASTM E-185, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." In the design of a plant-specific RV surveillance program, a licensee may use the edition of ASTM E-185 that was current on the issue date of the American Society of Mechanical Engineers Code to which the RV was purchased, or later editions through the 1982 edition.· The second alternative provided in Appendix H to 1 O CFR Part 50 is the implementation of an ISP. An ISP is defined in Appendix H to 10 CFR Part 50 as occurring when, "the representative materials chosen for surveillance for a reactor are irradiated in one or more other reactors that have similar design and operating features."

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3.0 TECHNICAL EVALUATION

3.1 Changes to Capsule TE1-C

Table 1-6, "Summary Status of the B&W Surveillance Capsules," and Table 1-7, "Summary Status of the Westinghouse Surveillance Capsules," of BAW-1543, Revision 4, Supplement 7, summarizes the status of the MIRVP ISP and plant-specific surveillance capsules. In this submittal, Table 1-6 was revised to reflect a change in status of Capsule TE1-C to withdrawn and tested.- Appendix H to 10 CFR Part 50 requires that within one year of capsule withdrawal, the test results must be submitted to the NRC in a technical summary report. In BAW-1543, Revision 4, Supplement 6, Capsule TE1-C was identified as "removed, will be disposed." By letter dated March 17, 2000, the B&WOG Reactor Vessel Working Group informed the NRC staff of its plans to begin the disposal of all B&W plant-specific and intact B&W standby capsules. The letter stated that the existing B&W plant-specific standby capsules either did not contain weld materials or were not expected to contribute significantly to the Linde 80-weld metal surveillance database.

The Davis-Besse capsule, TE1-C, was in storage and was part of the disposal plan. This capsule was a standby capsule in the Davis-Besse Appendix H program for the original operating period. However, as a result of license renewal activities, the status of capsule TE1-C was changed from standby to being scheduled for withdrawal and testing to meet the recommendations in the aging management program of the Generic Aging Lessons Learned (GALL) Report. The Davis-Besse renewed operating license was received on December 8, 2015, and cited the following Appendix H summary technical report: AREVA report ANP-3339, Revision 0, "Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C." Therefore, the staff has concluded that revision to Table 1.,.6 for the change in status for Capsule TE1-C was accurate and consistent with the requirements of Appendix H to 10 CFR Part 50.

3.2 Changes to Capsules A2 and A4

BAW-1543, Revision 4, Supplement 6, Table VI, "Summary Status of the B&W Surveillance Capsules," stated that the withdrawal of Capsules A2 and A4 were scheduled for the end of cycle 29 at a target neutron fluence of 6.6 x 1019 n/cm2 (neutrons/centimeter squared) for each capsule. The scheduled capsule withdrawals in Table 1-3, "Capsule Insertion and Withdrawal

· Schedule for Crystal River Unit 3," and Table 1-7 of BAW-1543, Revision 4, Supplement 7, for . Capsules A2 and A4 were revised from "end of twenty-ninth cycle" to "withdrawc;1I not planned."

A staff request for additional information (RAI) regarding Capsules A2 and A4 was necessary to complete review of the submittal. Specifically, the NRC staff requested a list of the materials contained in Capsules A2 and A4; identification of other ISP plants with the same weld material heat numbers as the surveillance capsule materials; and a description of how the proposed revision to change the status of these capsules to "withdrawal not planned" will continue to meet the objectives of MIRVP through the period of extended operation.

By letter dated August 4, 2016, the PWROG responded with details regarding the MIRVP. There are a total of eight HUPCAP supplemental capsules in the program, which includes Capsules A2 and A4. Table 2-1 identifies the Linde 80 weld metals that are in Capsules A2 and A4.

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Table 2-1 Linde 80 Weld Metals Included in the MIRVP Supplemental A2 and A4

Capsules

Capsule Linde 80 Weld Wire Weld ID Heat

Capsule Linde 80 Weld Wire Weld ID Heat

SA-1101 71249 WF-25 299L44

SA-1135 61782 WF-67 72442 A2 A4

SA-1526 299L44 WF-70 72105

SA-1585 72445 --- ---

Table 2-2 identified a listing of the plants participating in the MIRVP with reactor vessel beltline weld materials containing the same weld wire heats found in the MIRVP Capsules A2 and A4 listed in Table 2-1.

Table 2-2

Plants Participating in the MIRVP with Reactor Vessel Beltline Welds Containing the Same Weld Wire Heats Found in MIRVP Supplemental A2 and A4 Capsules

Weld Reactor Vessel Containing Linde 80 Beltline Weld Wire Heat Linde 80 Weld ID

61782 SA-84 7 SA-1135 R.E. Ginna, Point Beach Unit 1 Oconee Unit 1

SA-1101 R.E. Ginna, Point Beach Unit 1, Turkey Point Unit 3, 71249 SA-1229 Turkey Point Unit 4

Oconee Unit 1

72105 WF-70 Oconee Unit 3, Three Mile Island Unit 1, Turkey Point Unit4

72442 SA-1484 WF-67 Point Beach Unit 2, Turkey Point Unit 3 Oconee Unit 3, Turkey Point Unit 4

72445 SA-1585 SA-1650 Oconee Unit 1, Surry Unit 1, Surry Unit 2 Surry Unit 1

299L44 , SA-1526 WF-25 Surry Unit 1, Three Mile Island Unit 1 · Oconee Unit 1, Oconee Unit 2, Three Mile Island Unit 1

Table 2-3 identified alternate sources of Linde 80 surveillance weld metals with the same weld wire heats as the weld metals included in Capsules A2 and A4.

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Table 2-3

Alternate Sources of Linde 80 Surveillance Weld Metals with Weld Wire Heats Included in MIRVP Supplemental Capsules A2 and A4

Weld Wire Heat

61782

72445

71249 72442

299L44

71249

71249

299L44 72442 72105

Note 1:

Note 2:

Linde 80 Weld ID

SA-1036

SA-1263

SA-1101 WF-67

SA-1526

SA-1101

SA-1094

WF-25 WF-67 WF-70

Alternate Source Reactor Vessel Surveillance Programs (RVSP)

RE. Ginna Plant-Specific RSVP

Point Beach Unit 1 Plant-Specific RVSP

Supplemental Capsule "A" for Point Beach Unit 1 & Unit 2 (Installed in Point Beach Unit 2)

Surry Unit 1 Plant-Specific RVSP

Turkey Point Unit 3 Plant-Specific RVSP (Note 1)

Turkey Point Unit 4 Plant-Specific RVSP

MIRVP Supplemental Capsules A1 & L2 Irradiated in Davis-Besse (Note 2)

Turkey Point Unit 3 plant-specific Capsules T, V, and X contain the Linde 80 weld metal. As documented in Table 1-8 of BAW-1543, Revision 4, Supplement 7, these three (3) capsules have been removed and tested.

MIRVP Supplemental Capsules A 1 & L2 contain the same weld metals as MIRVP Supplemental Capsule A4.

In BAW-1543, Revision 4, Supplement 7, the PWROG compared the projected 60-year neutron fluences for the MIRVP member plants to available data for the Linde 80 weld metals. Table 1, "Minimum Recommended Number of Surveillance Capsules and Their Withdrawal Schedule," in ASTM E185-82 is applicable for 32 effective full-power years (EFPY), which corresponds to the original licensed period of 40 years of operation. For the period of extended operation, from 40 to 60 years, the GALL Report (Section XI.M31) states, "The plant-specific or integrated surveillance program shall have at least one capsule with a projected neutron fluence equal to or exceeding the 60-year peak reactor vessel wall neutron fluence prior to the end of the period of extended operation. The program withdraws one

. capsule at an outage in which the capsule receives a neutron fluence of between one and two times the peak reactor vessel wall neutron fluence at the end of the period of extended operation and tests the capsule in accordance with the requirements of ASTM E 185-82." The GALL report also addresses alternative dosimetry for plants without in-vessel capsules (which contain dosimetry). All the Westinghouse-designed plants participating in the MIRVP have in-vessel capsules, and therefore do not need to address ex-vessel dosimetry. The B&W-designed plants have ex-vessel dosimetry in place. Therefore, the dosimetry aspects of the MIRVP surveillance programs are consistent with the GALL report and are adequate for the period of extended operation.

In addition to dosimetry, surveillance programs monitor changes in fracture toughness of the ferritic beltline materials through the periodic testing of the test specimens in the surveillance capsules. In the RAI response, the PWROG stated that the, "expectation of the time of

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withdrawal and target neutron fluence of the MIRVP Capsules A2 and A4 was selected to generate irradiated high copper Linde 80 weld metal data at a neutron fluence level to support

· license extension beyond 40-years for the Westinghouse-designed plants with RVs containing high copper Linde 80 weld metals." The PWROG noted that, 'The B&W-designed plants also have B&W-fabricated reactor vessels that contain high copper Linde 80 weld metals;

. however these reactor vessels are expected to reach much lower fluences than the Westinghouse-designed plants at end of license (EOL}, thus the data from the MIRVP capsules A2 and A4 is not useful to the B&W-fabricated plants (i.e., the data would be greater than two (2) times the expected 60-year fluences)."

The Westinghouse plants in the MIRVP have plant-specific reactor vessel material surveillance programs. The plant-specific surveillance programs at the Point Beach, Units 1 and 2 are the only Westinghouse units that identify Capsules A2 and A4 as part of their RV material surveillance programs. MIRVP Capsule A2 contains Linde 80 weld wire heat

. numbers 61782 and 71249, which are also found in the Point Beach, Unit 1 beltline; MIRVP Capsule A4 contains Linde 80 weld wire heat number 72442, which is in the Point Beach, Unit 2 reactor vessel beltline. Staff reviewed the availability of other capsules with matching heat numbers to those contained in Capsules A2 and A4 and identified the following: ·

• Weld wire heat number 61782 is included in the R. E. Ginna Nuclear Power Plant (Ginna) plant-specific Appendix H program,

• Weld wire heat number 71249 is included in the Point Beach, Unit 1 and Point Beach, Unit 2 plant-specific Capsule A, Turkey Point, Unit 3 and Turkey Point, Unit 4 plantspecific surveillance programs, and

• Weld wire heat number 72442 is included in the Point Beach, Unit 1 and Point Beach, Unit 2 plant-specific Capsule A.

Based on the current withdrawal schedules, the Point Beach, Unit 1 and Point Beach, Unit 2 plant-specific Capsule A and Turkey Point, Unit 4 plant-specific Capsule X are scheduled to be withdrawn at 43 EFPY (-5.07E19 n/cm2} and 38.1 EFPY (-9.297E19 n/cm2) respectively.

High neutron fluence capsule test data for weld wire heat 61782 is currently available from the Ginna Capsule N. The PWROG indicated that these capsules are.expected to have neutron fluences similar to (or exceeding) that of the neutron fluence at the planned withdrawal time reported in BAW-1543, Revision 4, Supplement 6-A for the MIRVP Capsules A2 and A4. Therefore, the NRC staff finds that the withdrawal and testing of these surveillance capsules provides the necessary data to support Appendix H programs for Point Beach, Unit 1 and Point Beach, Unit 2 through the period of extended operation and adequately replaces the testing of Capsules A2 and A4. The PWROG further states, "To provide clear direction on when the

· Westinghouse plant-specific RVSP capsules are to be removed, Table 1-7 of BAW-1543, Revision 4, Supplement 7 will be updated to reflect the target withdrawal EFP_Y of the remaining capsules\ consistent with the current NRC accepted withdrawal schedules."

Therefore, upon the aforementioned update of the target neutron fluences and corresponding EFPY for the remaining capsules, the NRC staff concludes that the weld heats in Capsules A2 and A4 exist in other MIRVP surveillance capsules and those data appropriately represent the materials in Capsules A2 and A4. In summary, the NRC staff finds that the change in status of Capsules A2 and A4 to "withdrawal not planned" is acceptable and consistent with the requirements of Appendix H to 10 CFR Part 50.

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4.0 CONCLUSION

Based on its review of BAW-1543, Revision 4, Supplement 7, the NRC staff found that the updates to the MIRVP capsule withdrawal tables and the change in status for Capsules TE1-C, A2 and A4 are acceptable for the B&W-designed 177-FA plants and the Westinghousedesigned plants with B&W-fabricated reactor vessels. The NRC staff determined that BAW-1543, Revision 4, Supplement 7, including the update of Table 1-7 as described above, complies with Appendix H to 10 CFR Part 50. Therefore, the NRC staff approves the revisions contained in BAW-1543, Revision 4, Supplement 7.

5.0 REFERENCES

ASTM Standard E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, Philadelphia, Pennsylvania, 1982.

Code of Federal Regulations, Title 10, Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements [60 FR 65476, Dec. 19, 1995; 68 FR 75390 Dec. 31, 2003; 73 FR 5723, Jan. 31, 2008].

U.S. Nuclear Regulatory Commission, NUREG-1801, Revision 2, "Generic Aging Lessons Learned (GALL) Report," dated December 2010 (ADAMS Accession No. ML 103490041).

BAW-1543, Revision 4, Supplement 7, "Supplement to the Master Integrated Reactor Vessel Surveillance Program," dated January 23, 2015 (ADAMS Accession No. ML 15033A087) (Proprietary).

Safety Evaluation BAW-1543(NP), Revision 4, Supplement 6, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" (ADAMS Accession No. ML071770640).




Safety Evaluation BAW-1543(NP), Revision 4, Supplement 2, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" (ADAMS Legacy Accession No. ML970750004).

BAW-1543(NP), Revision 4, Supplement 1, "Master Integrated Reactor Vessel Surveillance Program" (ADAMS Legacy Accession No. ML9302190137).

BAW-1543(NP), Revision 4, "Master Integrated Reactor Vessel Surveillance Program" (ADAMS Legacy Accession No. ML9608090073).

Safety Evaluation BAW-1543(NP), Revision 3, "Revision 3 to Master Integrated Reactor Vessel Surveillance Program" (ADAMS Legacy Accession No. ML9105140288).

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AREVA report ANP-3339, Revision 0, "Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C (ADAMS Accession No. ML 16224A240).

Principle Contributor: C. Fairbanks

Date: March 27, 2018

January 23, 2015

OG-15-32

U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject: PWR Owners Group

Program Management Office 1000 Westinghouse Drive, Suite 380

Cranberry Township, PA 16066

BA W-1543, Revision 4, Supplement 7 Project Number 694

Submittal of BAW-1543, Revision 4, Supplement 7, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" and Transmittal of PWROG-14066-NP, Revision O, "Master Integrated Reactor Vessel Surveillance Program Long-Range Planning" to the NRC (PA-MSC-1182)

References:

1. NRC Letter from Jonathan G. Rowley to Anthony J. Mendiola, Public Pre-Submittal Conference Call With The Pressurized Water Reactor Owner's Group Regarding Topical Report BA W-l 543(NP), Revision 4, Supplement 6-A, "Supplement To The Master Integrated Reactor Vessel Surveillance Program," dated June 5, 2014, (MLl4153A581)

2. PWROG-14066-NP, Revision 0, "Master Integrated Reactor Vessel Surveillance Program Long-Range Planning," dated December 2014

The purpose, of this letter is to submit Pressurized Water Reactor Owners Group (PWROG) Topical Report, BA W-1543, Revision 4, Supplement 7, "Supplement to the Master Integrated Reactor Vessel Surveillance Program," December 2014, in accordance with the Nuclear Regulatory Commission ·(NRC) Topical Report (TR) program for review and acceptance for referencing in regulatory actions.

· This TR supplement (Enclosure 1) addresses: the adjustment of the withdrawal schedule of the remaining capsules to provide irradiated material corresponding to reactor vessel fluences for the Westinghouse designed NSSS reactor vessels between one and two times the 60 year fluence. The supplement is a non-proprietary report.

The PWROG is submitting this TR to the NRC for review and approval so that licensees can reference the NRC approved TR which licensees will use to support the B& W designed operating plants and those Westinghouse-designed operating plants having B&W-fabricated reactor vessels. PWROG requests that the NRC complete its review of the enclosed report and issue the Safety Evaluation (SE) by January 31, 2017.

U.S. Nuclear Regulatory Commission Document Control Desk OG-15-32

January 23, 2015 Page 2 of 3

The PWROG is also transmitting PWROG-14066-NP, Revision 0, "Master Integrated Reactor Vessel Surveillance Program Long-Range Planning." PWROG-14066-NP (Enclosure 2) provides the technical basis to support the review of the TR supplement (Enclosure I). PWROG-14066 is provided for information only as requested in the Reference 1 conference call and is not intended for NRC review and approval pursuant to the provisions of 10 CFR I 70.11 (a)(l )(iii)(A) and a separate TAC number should not be opened associated with the PWROG-14066-NP report.

TR Classification: As discussed above, this TR supp01ts the applicable plants in meeting the requirements of Code of Federal Regulations, Title 10, Part 50 (IO CFR 50) Appendix H, "Reactor Vessel Material Surveillance Program Requirements." NRC review and approval of this TR, as opposed to submittal of plant-specific LARs, will ensure the most efficient use of both NRC and licensee resources.

Specialized Resource Availability: This TR is being submitted to the NRC for review and approval so that the NRC approved version can be referenced in plant-specific LARs. NRC approval of the TR will reduce the impact on both licensee and NRC resources in both the preparation of, and NRC review of LARs that will be submitted by the applicable plants.

This letter transmits four copies of BAW-1543, Revision 4, Supplement 7, (Enclosure I) and four copies of PWROG-14066-NP, Revision 0, (Enclosure 2). ·

Applicability: This TR is applicable to PWROG members with operating B& W designed Nuclear Steam Supply System (NSSS) and those Westinghouse-designed operating plants having B&W-fabricated reactor vessels. TR supports life extension activities for applicable plants, especially Point Beach Units 1 and 2. TR also supports Duke Energy's efforts regarding closure of Crystal River Unit 3.

If you have any questions related to this submittal, please contact Ms. Gayle F. Elliott, AREVA Inc. Product Licensing Manager, at (434) 832-3347 or by e-mail at [email protected] or Mr. Jack Stringfellow, PWROG Chairman and COO at (205) 992-7037.

Sincerely,

Jack Stringfellow, SNOC Chairman and COO, PWR Owners Group

NJS:JPM:las

Enclosure I - Four copies of BA W-1543, Revision 4, Supplement 7 (for review and acceptance)

Enclosure 2 - Four copies of PWROG-14066-NP, Revision O (for information only)


cc: PWROG Management Committee PWROG Materials Committee PWROG Licensing Committee PWROGPMO Brian Haibach (AREY A) Ryan Hosler (AREY A) Matt De Van (AREVA) Gayle Elliott (AREVA) Tammy Natour (AREY A) Ben Grambau (AREY A) B. Hall (W) Jonathan Rowley (NRC)

January 23, 2015 Page 3 of 3

June 11, 2015


. SUBJECT: ACCEPTANCE FOR REVIEW OF THE PRESSURIZED WATER REACTOR OWNERS GROUP TOPICAL REPORT BAW-1543, REVISION 4, SUPPLEMENT 7, "SUPPLEMENT TO THE MASTER INTEGRATED REACTOR VESSEL SURVEILLANCE PROGRAM" (TAC NO. MF5701)


By letter dated January 23, 2015 (Agencywide Documents Access and Management System Accession Number ML 15033A086), the Pressurized Water Reactor Owners Group (PWROG) submitted for U.S. Nuc)ear Regulatory Commission (NRC) staff review Topical Report (TR) BAW-1543, Revision 4, Supplement 7, "Supplement to the Master Integrated Reactor Vessel Surveillance Program." TheNRC staff has performed an acceptance review of TR BAW-1543, Revision 4, Supplement?. We have found that the material presented is sufficient to begin our comprehensive review. The NRC staff expects to issue its request for additional information by October 30, 2015, and issue its draft safety evaluation (SE) by May 20, 2016. The NRC staff estimates that the review will require approximately 300 staff hours including project management time. The review schedule milestones and estimated review costs were discussed and agreed upon in a telephone conference between PWROG Project Manager, Chad Holderbaum, and the NRC staff on May 27, 2015.

Section 170.21 of Title 1 O of the Code of Federal Regulations requires that TRs are subject to fees based on the full cost of the review. You did not request a fee waiver; therefore, NRC staff hours will be billed accordingly.

As with all TRs, the SE will be reviewed by the NRC's Office of the General Counsel (OGC) to determine whether it falls within the scope of the Congressional Review Act (CRA). During the course of this review, OGC considers whether any endorsement or acceptance of a TR by the NRC amounts to a rule as defined in the CRA. If this initial review concludes that the SE, with its accompanying TR, may be a rule, the NRC will forward the package to the Office of Management and Budget (0MB) for further review and consideration. ·

W.A. Nowinowski - 2 -

Any review by 0MB would impact the schedule for the issuance of the final SE. If you have questions regarding this matter, please contact Jonathan G. Rowley at (301) 415-4053 or by e-mail at [email protected].

Project No. 694

Sincerely,

/RA by HCruz for/

Anthony J. Mendiola, Chief Licensing Processes Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

W.A. Nowinowski - 2 -

Any review by 0MB would impact the schedule for the issuance of the final SE. If you have questions regarding this matter, please contact Jonathan G. Rowley at (301) 415A053 or by e-mail at [email protected]. ·

Project No. 694

DISTRIBUTION: PUBLIC PLPB Reading File RidsNrrDpr RidsNrrDprPlpb RidsNrrLADHarrison RidsAcrsAcnwMailCenter RidsOgcMailCenter RidsNrrDe RidsNrrDeEvib JMcHale AMendiola

ADAMS Accession No: ML15147A153

Sincerely,

/RA by HCruz for/

Anthony J. Mendiola, Chief Licensing Processes Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

NRR-106 rr=======;========.======,===========r===== OFFICE PLPB/PM

NAME JRowley

DA TE · 6/9/2015

PLPB/LA

DHarrison

6/11/2015

EVIB/BC

JMcHale

6/11/2015


PLPB/BC

AMendiola (HCruz for)

6/11/2015

June 3, 2016


SUBJECT: REQUEST FOR ADDITIONAL INFORMATION RE: PRESSURIZED WATER REACTOR OWNERS GROUP TOPICAL REPORT BAW-1543, REVISION 4, SUPPLEMENT 7, "SUPPLEMENT TO THE MASTER INTEGRATED REACTOR VESSEL SURVEILLANCE PROGRAM" (TAC NO. MF5701)


By letters dated January 23, 2015 (Agencywide Documents Access and Management

System Accession No. ML 15033A086), the Pressurized Water Reactor Owners Group

submitted Topical Report BAW-1543, Revision 4, Supplement 7, "Supplement to the Master

Integrated Reactor Vessel Surveillance Program," for U.S. Nuclear Regulatory Commission

(NRC) staff review. Upon review of the information provided, the NRC staff has determined that

additional information is needed to complete the review. Mr. Chad Holderbaum, of your staff,

and I agreed that the NRC staff will receive your response to the enclosed Request for

Additional Information (RAI) questions within 60 days from the date of this letter. If you have

any questions regarding the enclosed RAI questions, please contact me at 301-415-4053.

Project No. 694

Enclosure: RAI questions

cc w/encl: See next page

Sincerely,

IRA L. Wilkins for/

Jonathan G. Rowley, Project Manager Licensing Processes Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation


SUBJECT: REQUEST FOR ADDITIONAL INFORMATION RE: PRESSURIZED WATER REACTOR OWNERS GROUP TOPICAL REPORT BAW-1543, REVISION 4, SUPPLEMENT 7, "SUPPLEMENT TO THE MASTER INTEGRATED REACTOR VESSEL SURVEILLANCE PROGRAM" (CAC NO. MF5701)


By letters dated January 23, 2015 (Agencywide Documents Access and Management

System Accession No. ML 15033A086), the Pressurized Water Reactor Owners Group

submitted Topical Report BAW-1543, Revision 4, Supplement 7, "Supplement to the Master

Integrated Reactor Vessel Surveillance Program," for U.S. Nuclear Regulatory Commission

(NRC) staff review. Upon review of the information provided, the NRC staff has determined that

additional information is needed to complete the review. Mr. Chad Holderbaum, of your staff,

and I agreed that the NRC staff will receive your response to the enclosed Request for

Additional Information (RAI) questions within 60 days from the date of this letter. If you have

any questions regarding the enclosed RAI questions, please contact me at 301-415-4053.

Project No. 694

Enclosure: RAI questions

cc w/encl: See next page

DISTRIBUTION: PUBLIC JMcHale RidsNrrDe

RidsNrrDpr RidsOgcMailCenter RidsResOd

EXTERNAL DISTRIBUTION: [email protected]

RidsNrrLADHarrison RidsNrrDeEvib RidsNroOd

ADAMS Accession No.: ML 16126A444; *concurred via e-mail

Sincerely,

IRA L. Wilkins for!

Jonathan G. Rowley, Project Manager Licensing Processes Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

RidsNrrDprPlpb RidsACRS_MailCTR CFairbanks

NRR-106

OFFICE PLPB/PM PLPB/LA* PLPB/BC PLPB/PM NAME JRowley DHarrison KHsueh JRowle LWilkins for DATE 6/2/16 06/02/2016 6/3/16 6/3/16


RAI 1

BACKGROUND

REQUEST FOR ADDITIONAL INFORMATION

. REVIEW OF BAW-1543, REVISION 4, SUPPLEMENT 7

"SUPPLEMENT TO THE MASTER INTEGRATED REACTOR

VESSEL SURVEILLANCE PROGRAM"

Title 1 O of the Code of Federal Regulations (1 O CFR) Part 50, Appendix H, requires licensees to maintain reactor vessel material surveillance programs. The purpose of these programs is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region that result from the exposure of these materials to neutron irradiation and the thermal environment. The surveillance capsules are withdrawn periodically and tested in accordance with the approved schedule and the requirements of Appendix H to 10 CFR Part 50, "Reactor Vessel Material Surveillance Program Requirements."

ISSUE

Appendix H to 10 CFR Part 50 describes the requirements for reactor vessel material surveillance programs. In addition for license renewal, NUREG-1801, Generic Aging Lessons Learned (GALL) Report Aging Management Program XI.M31 Reactor Vessel Surveillance, provides additional information regarding reactor vessel material surveillance programs for the period of extended operation (to 60 years).

The submittal states, "Table 1-3 was revised by changing the capsule status of the Master Integrated Reactor Vessel Surveillance Program supplemental capsules A2 and A4."

REQUEST

In the submittal, the status of the supplemental capsules A2 and A4 in Table 1-3 ("Capsule Insertion and Withdrawal Schedule for Crystal River Unit 3") was changed to "withdrawal not planned." TR BAW-1543, Revision 4, Supplement 6, included withdrawal and testing plans for these capsules. Provide a list of the materials contained in supplemental capsules A2 and A4; identify the plants with the same weld material heat numbers as the surveillance capsule materials; and provide a description of how the proposed revision to change the status of these capsules to "withdrawal not planned" will continue to meet the objectives of the Pressurized · Water Reactor Owners Group's Master Integrated Reactor Vessel Surveillance Program (MIRVSP) through the period of extended operation.

The response should also address the statements in the letter dated February 15, .2007 (Agencywide Documents Access and Management System Accession No. ML070510233), where Capsules A2 and A4 are discussed. Although the context of the responses were developed to support a change in withdrawal and testing from the end of the seventeenth fuel cycle to the end of the twenty-ninth fuel cycle, the role of the specimens in Capsules A2 and A4

Enclosure

-2-

to" meet the objectives of the MIRVSP is discussed in several places. The response should describe the impact of not testing Capsules A2 and A4, including the impact to the objectives met by Capsules A2 and A4 as described in the February 15, 2007, letter, and the impact to any other plants in the MIRVSP. It should be noted that although the neutron fluence accumulated by a surveillance capsule may lag the neutron fluence of the reactor vessel, that does not invalidate the data. In addition, include a full description of the phrase, "withdrawal not planned."

RAl2

BACKGROUND

Appendix H to 1 O CFR Part 50 requires licensees to maintain reactor vessel material surveillance programs. The purpose of these programs is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region that result from the exposure of these materials to neutron irradiation and the thermal environment.

ISSUE

Appendix H to 10 CFR Part 50 describes the requirements for reactor vessel material surveillance programs. In addition, for license renewal, NUREG-1801, GALL Report Aging Management Program XI.M31 Reactor Vessel Surveillance, provides additional information regarding reactor vessel material surveillance programs for the period of extended operation (to 60 years).

Appendix H, Section IV, Part A states, "Each capsule withdrawal and the test results must be the subject of a summary technical report to be submitted, as specified in [10 CFR] 50.4, within one year of the date of capsule withdrawal, unless an extension is granted by the Director, Office of Nuclear Reactor Regulation."

REQUEST

Table 1-8 (Comparison of the Plant-Specific Surveillance Capsules with ASTM E 185-82 Requirements) column heading states, "EOL [end of license] or 1-2 times EOL fluence (Capsule may be held without testing)." Staff requests revision of Table 1-8 for consistency with Appendix H, Section IV, Part A.

August 4, 2016 OG-16-261

U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject: PWR Owners Group

Program Management Office 1000 Westinghouse Drive, Suite 380

Cranberry Township, PA 16066

BAW-1543, Revision 4, Supplement 7 Project Number 694

Transmittal of Response to NRC Request for Additional Information (RAI) Related to Supplement to the Master Integrated Reactor Vessel Surveillance Program, BAW-154301-000, Revision 4 Supplement 7 (PA-MSC-1182RO)

References:

1. Submittal of BAW-1543, Revision 4, Supplement 7, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" and Transmittal of PWROG-14066-NP, Revision 0, "Master Integrated Reactor Vessel Surveillance Program Long-Range Planning" to the NRC (PA-MSC-1182), OG-15-32, dated January 23, 20i5.

2. NRC Acceptance of PWROG Report BAW-1543, Revision 4, Supplement 7, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" (PA-MSC-1182), OG-15-259, dated June 18, 2015.

3. Request for Additional lnfonilation Re: Pressurized Water Reactor Owners Group Topical Report BAW-1543, Revision 4, Supplement 7, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" (TAC No. MF5701), dated June 3, 2016

The PWROG submitted the subject report to the NRC for review and approval under Reference 1. The report was accepted in June 2015 under Reference 2. The PWROG is pleased to provide the responses to the RAis received under Reference 3 (Enclosure 1 ).

If you have any questions related to these RAI responses, please contact Ms. Gayle F. Elliott, AREVA Inc. Product Licensing Manager, at 434-832-3347 or by e-mail at e:[email protected] or Mr. Jack Stringfellow, PWROG Chairman and COO at (205) 992-7037.

Sincerely,

71~ fl,#;~~?,.__-

Jack Stringfellow, SNOC Chairman and COO, PWR Owners Group

-----------------------------··--·-.. ·•··•····•


JPM:NJS:cah

Enclosures 1 - BA W- l 543Q 1-000, Revision 4 Supplement 7

cc: PWROG Management Committee PWROG Materials Committee PWROG Licensing Committee PWROGPMO Brian Haibach (AREY A, Inc.) Ryan Hosler (AREY A, Inc.) Matt DeVan (AREVA, Inc.) Gayle Elliott (AREVA, Inc.) Tammy Natour (AREY A, Inc.) Brian Hall (Westinghouse) Jonathan Rowley (NRC)

August 4, 2016 Page 2 of2

, I ·! I

., -····•-·--·--••••·•-··-·•·-•··-··- • -·--·····•"•'""·--·--n·• ---- .... .., .. ,. "'"---~·-··· . -· -- ---------~---------··------· '

0414-12-F04 (Rev. 001, 03/10/2016)

A AREVA

Response to NRC Request for Additional Information (RAI) Related to Supplement to the Master Integrated Reactor Vessel Surveillance Program

July 2016

AREVA Inc.

(c) 2016 AREVA Inc.

BAW-1543Q1-000 Revision 4 Supplement 7

Copyright© 2016

AREVA Inc. All Rights Reserved

BAW-1543Q 1-000 Revision 4 Supplement 7

AREVA Inc. BAW-154301-000 Revision 4 Supplement 7


. Item 1

Section(s) or Page(s) All

Pae i

Nature of Changes

Description and Justification Initial Issue



Pa e ii

Contents

Page

1.0 INTRODUCTION AND SUMMARY ................................................................... 1-1

2.0 REQUESTS FOR ADDITIONAL INFORMATION (RAIS) AND RESPONSES .................................................................................................... 2-1

2.1 RAI 1 ....................................................................................................... 2-1 . 2.1.1 Statement of RAI 1 ....................................................................... 2-1 2.1.2 PWROG Response to RAI 1 ........................................................ 2-2

2.2 RAl2 .............................................................. : ...................................... 2-13 2.2.1 Statement of RAI 2 ..................................................................... 2-13 2.2.2 PWROG Response to RAI 2 ...................................................... 2-13

3.0 REFERENCES .................................................................................................. 3-1

AREVA Inc. BAW-1543Q1-000 Revision 4 Supplement 7


Pae iii

List of Tables

Table 2-1 Linde 80 Weld Metals Included in the MIRVP Supplemental A2 and A4 Capsules ................................................................................................... 2-3

Table 2-2 Plants Participating in the MIRVP with Reactor Vessel Beltline Welds Containing the Same Weld Wire Heats Found in MIRVP Supplemental A2 and A4 Capsules ........................................................... 2-3

Table 2-3 Alternate Sources of Linde 80 Surveillance Weld Metals with Weld Wire Heats Included in MIRVP Supplemental Capsules A2 arid A4 .......... 2-4

AREVA Inc. BAW-1543Q 1-000 Revision 4 Supplement 7


List of Figures

Figure 2-1 Linde 80 Weld Metal Surveillance Data vs. Projected Plant 60-Year Reactor Vessel Beltline Weld Metal Inside Surface Fluences for the Weld Wire Heats in MIRVP Supplemental Capsules A2 and A4 (Note: the AN01 and DB reactor vessels do not contain any of the

Pae iv

applicable weld wire heats in the beltline region) ..................................... 2-12

AREVA Inc. BAW-1543Q 1-000 Revision 4 Supplement 7


Acronym AN01

B&W

B&WOG

CR3

DB

EOL

EPU

GALL

HUPCAPS

ID

IRVSP

MIRV[S]P

NRC

ONS1

ONS2

ONS3

PB1

PB2

PTLR

PWR(s)

PWROG

RAl(s)

RCS

REG

RVSP(s)

S1

S2

TMl1

TP3

TP4

TR

Nomenclature

Definition Arkansas Nuclear One Unit 1

Babcock & Wilcox

Babcock & Wilcox Owners Group

Crystal River Unit 3

Davis-Besse Unit 1

End of License

Extended Power Uprate

Generic Aging Lessons Learned

Pa ev

Higher Fluence Supplementary Weld Metal Surveillance Capsules

Inside Diameter

Integrated Reactor Vessel Surveillance Program

Master Integrated Reactor Vessel Surveillance Program

Nuclear Regulatory Commission

Oconee Nuclear Station Unit 1



Point Beach Unit 1

Point Beach Unit 2

Pressure and Temperature Limits Report

Pressurized Water Reactor(s)

PWR Owners Group

Request(s) for Additional Information

Reactor Coolant System

R. E. Ginna

Reactor Vessel Surveillance Program(s)

Surry Unit 1

Surry Unit 2

Three Mile Island Unit 1

Turkey Point Unit 3

Turkey Point Unit 4

Topical Report


Response to NRC Request for Additional Information {RAI) Related to Supplement to the Master Integrated Reactor Vessel Surveillance Program

Pa evi

ABSTRACT

The B&W Owners Group t Integrated Reactor Vessel Surveillance Program was initiated

in 1976 for the Babcock & Wilcox (B&W) 177-FA plants. Its purpose was to augment

the existing unit-specific reactor vessel surveillance programs (RVSPs) and to provide a

basis for sharing information between plants. In 1988, the program was further modified

to include a series of plants with a Westinghouse Nuclear Steam Supply System for

which B&W manufactured the reactor vessels creating the Master Integrated Reactor

Vessel Surveillance Program (MIRVP). The pressurized water reactor (PWR) units that

currently participate in the MIRVP include Arkansas Nuclear One Unit 1, Davis-Besse

Unit 1, Oconee Unit 1, Oconee Unit 2, Oconee Unit 3, Three Mile Island Unit 1, R. E.

Ginna, Point Beach Unit 1, Point Beach Unit 2, Surry Unit 1, Surry Unit 2, Turkey Point

Unit 3, and Turkey Point Unit 4.

The overall objective of the MIRVP is to provide the data necessary to assure

compliance with Federal Regulations:): to ensure continued licensability of the B&W

fabricated reactor vessels with high copper Linde 80 weld metals that are sensitive to

fast neutron fluence exposure.

The current MIRVP is described in the base Topical Report BAW-1543, Revision 4 and

a supplement to this Topical Report provides the withdrawal schedule of the

surveillance program capsules. The current Nuclear Regulatory Commission (NRC)

approved Topical Report supplement [BAW-1543(NP), Revision 4, Supplement 6-A] has

been revised to adjust the withdrawal status of four (4) surveillance capsules.

t At the end of 2005, the B&W Owners Group (B&WOG) ceased to exist; however the MIRVP continues within the Pressurized Water Reactor Owners Group (PWROG).

* Code of Federal Regulations, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities", U.S. Nuclear Regulatory Commission, Washington, D.C.


Response to NRC Request for Additional Information (RAJ) Related to Supplement to the Master Integrated Reactor Vessel Surveillance Program

Pa evii

Topical Report BAW-1543, Revision 4, Supplement 7 was submitted to the NRC for

review and approval in January 2015. Upon review of the information provided, the

NRC Staff has determined that additional information is needed to complete their review

of the Topical Report supplement, and thus Requests for Additional Information (RAls)

on the submittal have been issued. This report provides the responses to those RAls.


. Response to NRG Request for Additional Information (RAI) Related to Supplement to the Master Integrated Reactor Vessel Surveillance Program

Pa e 1-1

1.0 INTRODUCTION AND SUMMARY

The Master Integrated Reactor Vessel Surveillance Program (MIRVP) provides the data

necessary to assure compliance with Federal Regulations to insure continued

licensability of the Babcock & Wilcox (B&W) fabricated reactor vessels with high copper

Linde 80 weld metals that are sensitive to fast neutron fluence exposures. The

pressurized water reactors (PWRs) with high copper Linde 80 weld metals that currently

participate in the MIRVP include Arkansas Nuclear One Unit 1 (AN01), Davis-Besse

Unit 1 (DB), Oconee Unit 1 (ONS1), Oconee Unit 2 (ONS2), Oconee Unit 3 (ONS3),

Three Mile Island Unit 1 (TMl1), R. E. Ginna (REG), Point Beach Unit 1 (PB1), Point

Beach Unit 2 (PB2), Surry Unit 1 (S1 ), Surry Unit 2 (S2), Turkey Point Unit 3 (TP3), and

Turkey Point Unit 4 (TP4).

AREVA Topical Report BAW-1543, Revision 4, Supplement 7, "Supplement to the

Master Integrated Reactor Vessel Surveillance Program," was prepared by AREVA Inc.

for the PWR Owners Group (PWROG) and subsequently submitted to the Nuclear

Regulatory Commission (NRC) by the PWROG [1]. Upon review of the information

provided, the NRC Staff has determined that additional information is needed to

complete the review of the Topical Report, and thus the NRC Staff has issued Requests

for Additional Information (RAls) [2] on this submittal. This report provides the

responses to those RAls.



Pa e 2-1

2.0 REQUESTS FOR ADDITIONAL INFORMATION (RAls) AND RESPONSES

The '~Issue" and "Request" sections from the NRC RAls are reproduced from Reference

[2] in Sections 2.1.1 and 2.2.1. The PWROG responses follow each reproduced RAI in

Sections 2.1.2 and 2.2.2.

2.1 RAJ 1

2.1.1 Statement of RAI 1

ISSUE

Appendix H to 10 CFR Part 50 describes the requirements for reactor vessel. material

surveillance programs. In addition for license renewal, NUREG-1801, Generic Aging

Lessons Learned (GALL) Report Aging Management Program XI.M31 Reactor Vessel

Surveillance, provides additional information regarding reactor vessel material

surveillance programs for the period of extended operation (to 60 years).

The submittal states, "Table 1-3 was revised by changing the capsule status of the

Master Integrated Reactor Vessel Surveillance Program supplemental capsules A2 and

A4."

REQUEST

In the submittal, the status of the supplemental capsules A2 and A4 in Table 1-3

("Capsule Insertion and Withdrawal Schedule for Crystal River Unit 3") was changed to

"withdrawal not planned." [Topical Report] TR BAW-1543, Revision 4, Supplement 6,

included withdrawal and testing plans for these capsules. Provide a list of the materials

contained in supplemental capsules A2 and A4; identify the plants with the same weld

material heat numbers as the surveillance capsule materials; and provide a description

of how the proposed revision to change the status of these capsules to "withdrawal not

planned" will continue to meet the objectives of the Pressurized Water Reactor Owners

Group's Master Integrated Reactor Vessel Surveillance Program (MIRVSP) through the

period of extended operation.



Pa e 2-2

· The response should also address the statements in the letter dated February 15, 2007

(Agencywide Documents Access and Management System Accession No.

ML070510233), where Capsules A2 and A4 are discussed. Although the context of the

responses were developed to support a change in withdrawal and testing from the end

of the seventeenth fuel cycle to the end of the twenty-ninth fuel cycle, the role of the

specimens in Capsules A2 and A4 to meet the objectives of the MIRVSP is discussed in

several places. The response should describe the impact of not testing Capsu1e·s A2

and A4, including the impact to the objectives met by Capsules A2 and A4 as described

in the February 15, 2007, letter, and the impact to any other plants in the MIRVSP. It

should be noted that although the neutron fluence accumulated by a surveillance

capsule may lag the neutron fluence of the reactor vessel, that does not invalidate the

data. In addition, include a full description of the· phrase, "withdrawal not planned."

2.1.2 PWROG Response to RAI 1

2.1.2.1 Linde 80 Weld Metals Contained in Supplemental Capsules A2 and A4

As detailed in the base MIRVP document, BAW-1543, Revision 4 [3], Higher Fluence

Supplementary Weld Metal Surveillance Capsules (HUPCAPS) are included in the

MIRVP to (1) provide for additional B&W-designed irradiation capsules to expand and

enlarge the compact fracture toughness database; (2) provide for an irradiation capsule

of Westinghouse-design for correlation of irradiation data in the Westinghouse neutronic

environment with the B&W 177-FA environment; and, (3) provide capsules for a weld

metal annealing response investigation. The weld metals contained in the HUPCAPS

include high copper Linde 80 weld metals taken from the B&W Owners Group

(B&WOG) inventory. There are a total of eight (8) capsules in the HUPCAP program,

which include the supplemental capsules A2 and A4. Table 2~1 identifies the Linde 80

weld metals that are in the MIRVP supplemental capsules A2 and A4.



Pa e 2-3

Table 2-1 Linde 80 Weld Metals Included in the MIRVP Supplemental A2 ~nd A4

Capsules

Capsule Linde 80 Weld Wire Weld ID Heat Capsule Linde 80 Weld Wire

Weld ID Heat

SA-1101 71249 WF-25 299L44

SA-1135 61782 WF-67. 72442 A2 A4

SA-1526 299L44 WF-70 72105

SA-1585 72445 --- ---

Table 2-2 presents a listing of the plants participating in the MIRVP with reactor vessel

beltline weld materials containing the same weld wire heats found in the MIRVP

supplemental capsules A2 and A4 listed in Table 2-1.

Table 2-2 Plants Participating in the MIRVP with Reactor Vessel Beltline Welds Containing the Same Weld Wire Heats Found in MIRVP Supplemental

A2 and A4 Capsules

Weld. Linde 80 Reactor Vessel Containing Linde 80 Beltline Weld Wire Heat Weld ID

61782 SA-847 R.E. Ginna, Point Beach Unit 1 SA-1135 Oconee Unit 1

SA-1101 R.E. Ginna, Point Beach Unit 1, Turkey Point Unit 3, Turkey 71249 Point Unit 4

SA-1229 Oconee Unit 1

72105 WF-70 Oconee Unit 3, Three Mile Island Unit 1, Turkey Poirit Unit 4

72442 SA-1484 Point Beach Unit 2, Turkey Point Unit 3 WF-67 Oconee Unit 3, Turkey Point Unit 4

72445 SA-1585 Oconee Unit 1, Surry Unit 1, Surry Unit 2 SA-1650 Surry Unit 1

299L44 SA-1526 Surry Unit 1, Three Mile Island Unit 1 WF-25 Oconee Unit 1, Oconee Unit 2, Three Mile Island Unit 1

AREVA Inc. BA W-154301-000 Revision 4 Supplement 7


Pa e 2-4

2.1.2.2 Alternate Sources of Linde 80 Surveillance Weld Metals

The irradiation properties of several Linde 80 weld wire/flux combinations are being

characterized in existing Reactor Vessel Surveillance Programs (RVSPs) [i.e., B&WOG

Integrated Reactor Vessel Surveillance Program (IRVSP) and Westinghouse plant

specific RVSPs]. The MIRVP supplemental capsules A2 and A4 were designed and

· fabricated to supplement existing plant-specific surveillance programs for plants with

high copper Linde 80 welds (i.e., "sister plant data"), with the main purpose to expand

and enlarge the irradiation property database. Table 2-3 identifies the alternate sources

of Linde 80 surveillance weld metals with the same weld wire heats as the weld metals

included in the MIRVP supplemental capsules A2 and A4.

Table 2-3 Alternate Sources of Linde 80 Surveillance Weld Metals with Weld Wire Heats Included in MIRVP Supplemental Capsules A2 and A4

Weld Wire Linde 80 Alternate Source Heat Weld ID

61782 SA-1036 R.E. Ginna Plant-Specific RVSP

72445 SA-1263 Point Beach Unit 1 Plant-Specific RVSP

71249 SA-1101 Supplemental Capsule "A" for Point Beach Unit 1 & Unit 2 72442 WF-67 (Installed in Point Beach Unit 2)

299L44 SA-1526 Surry Unit 1 Plant-Specific RVSP

71249 SA-1101 Turkey Point Unit 3 Plant-Specific RVSP (Note 1)

71249 SA-1094 Turkey Point Unit 4 Plant-Specific RVSP

299L44 WF-25

72442 WF-67 MIRVP Supplemental (HUPCAPS) Capsules A1 & L2

72105 WF-70 Irradiated in Davis-Besse (Note 2)

Note 1: Only Turkey Point Unit 3 plant-specific Capsules T, V, and X contain the Linde 80 weld

metal. As documented in Table 1-8 of BAW-1543, Revision 4, Supplement 7 [1], these

three (3) capsules have been removed and tested.

Note 2: MIRVP Supplemental Capsules A1 & L2 contain the same we_ld metals as MIRVP

Supplemental Capsules A4.



Pa e 2-5

2.1.2.3 Regulation Compliance for Reactor Vessel 60-Year Life

The high copper Linde 80 weld metals tend to be the limiting materials within the B&W

fabricated reactor vessels with respect to reactor vessel integrity. The projected 60-

year fluence for the MIRVP member plants were compared to available data for the

Linde 80 weld metals. -The types of data considered in the comparison include upper

shelf fracture toughness and Charpy impact energy. Compliance with the requirements

. in ASTM E185-82 [4] as referenced in 10 CFR 50 Appendix H [5] and the Generic Aging

Lessons Learned (GALL) Report [6] are taken into consideration.

The last two capsule withdrawals per ASTM E185-82 should be approximately (the first

capsules have already been withdrawn and tested and are not pertinent to this

evaluation):

• Fluence at inside diameter (ID) at the end of license (EOL), and

• Fluence at not-less-than once or greater than twice the peak ID at the

EOL

The surveillance capsules that meet the above requirements are listed in Table VIII of

BAW-1543, Revision 4, Supplement 6~A [7] for 40-years or 60-years. This evaluation

reconsiders this data for projected fluences with 60-years as EOL.

The GALL report states that an acceptable surveillance program will in part (Section

XI.M31):

• All pulled and tested capsules, unless discarded before August 31, 2000,

are placed in storage. (Note: These specimens are saved for future

. reconstitution use, in case the surveillance program is reestablished.)

• Plants without in-vessel capsules shall use alternative dosimetry to

monitor fluence during the period of extended operation



•

Pa e 2-6

Test a capsule per ASTM E185 (Charpy) with fluence of equal to or

greater than 60-years (one to two times the peak reactor vessel wall

neutron fluence at EOL)

The GALL report states· that if future plant operations exceed the current limitations or

bounds, such as fluence, the impact of plant operation changes on the extent of reactor

vessel embrittlement shall be evaluated and the NRC will be notified.

The GALL report requires alternative dosimetry for plants without in-vessel dosimetry

(capsules). All the B&W-designed plants have ex-vessel dosimetry that meets this

requirement [8]. (The ONS-2 ex-vessel dosimetry is used for all the ONS units based

on geometry and operational similarities.) However, since Davis-Besse Unit 1 (DB) still

has in-vessel MIRVP capsules, which contain dosimetry, these could also be used to

meet this requirement. All the Westinghouse-designed plants participating in the

MIRVP still have in-vessel capsules, and therefore are not required to have ex-vessel

dosimetry.

The expectation of the time of withdrawal and target fluence of the MIRVP supplemental

capsules A2 and A4 was to generate irradiated high copper Linde 80 weld metal data at

a fluence level to support license extension beyond 40-years for the Westinghouse

designed plants with reactor vessels containing high copper Linde 80 weld metals.

However, the time of removal of these supplemental capsules would have been well

into the license extension period of the Westinghouse-designed plants, so the

information that would be gathered from testing these capsules would only supplement

their existing plant-specific RVSPs.

The B&W-designed plants also have B&W-fabricated reactor vessels that contain high

copper Linde 80 weld metals; however these reactor vessels are expected to reach

much lower fluences than the Westinghouse-designed plants at EOL, thus the data from

the MIRVP supplemental capsules A2 and A4 is not useful to the B&W-fabricated plants

(i.e., the data would be greater than two (2) times the expected 60-year fluences).

1·' I I



Pa e 2-7

Each of the Westinghouse-designed plants that participate in the MIRVP has a plant

specific RVSP. With the exception of PB1 and PB2, none of these Westinghouse

designed plants take direct credit for the MIRVP supplemental capsules A2 and A4 as

part of their plant-specific RVSPs.

The current plant-specific RVSP capsule withdrawal schedules for S1 and S2 [9] and for

TP3 and TP4 [1 O] do not include MIRVP supplemental capsules. These withdrawal

schedules have been reviewed by the NRC Staff, whereby they concluded in separate

Safety Evaluations that the submitted capsule withdrawal schedules satisfy the

requirements of ASTM E185-82, and therefore satisfy the requirements of Appendix H

to 10 CFR 50 for the 60-year extended operating period.

The current RVSP capsule withdrawal schedule for REG was submitted to the NRC

Staff as part of their reactor coolant system (RCS) Pressure and Temperature Limits

Report (PTLR) revisions [11]. The surveillance capsule removal schedule is provided in

Table PTLR-1 within this report (extracted from the analysis report for R~G RVSP

Capsule N [12]); the withdrawal schedule does not include MIRVP supplemental

capsules as part of the RVSP.



Pa e 2-8

The current PB1 and PB2 RVSP capsule withdrawal schedules are referenced in their

license amendment regarding extended power uprate (EPU) [13], indicating the

respective withdrawal schedules were provided in Table 2.1.1-6 of their EPU application

[14]. The NRC Safety Evaluation for the EPU license amendment indicates a revised

licensee commitment with respect to the RVSP was submitted in a letter dated January

19, 2010 [15]. In accordance with this letter, it was requested that Regulatory

Commitment .11 of NUREG-1839, Appendix A be modified from, "Capsule A2 (Unit 1)

will be removed at a target EOLE f/uence of 3. 7 x 1019 nlcm" to, "During the period of

extended operation, reactor vessel surveillance capsules will be removed in accordance

with.the schedule contained in the most recently NRG-approved Pressurized Water

Reactor Owners Group (PWROG) Master Integrated Reactor Vessel Surveillance

Program (MIRVSP) document." The NRC Staff concluded that the PB1 and PB2 RVSP

capsule withdrawal schedules, when revised in accordance with the commitment noted

above, will be appropriate to ensure that the RVSP will continue to meet the

requirements of 10 CFR Part 50, Appendix H, and 10 CFR 50.60 [16].

The change in BAW-1543, Revision 4, Supplement 7, adjusts the status of the MIRVP

. supplemental capsules A2 and A4 to "no planned withdrawal." Based on the revised

licensee commitment (Regulatory Commitment 11 ), the consequence of this change on

the applicability to PB1 and PB2 was assessed.

The MIRVP supplemental capsule A2 contains the high copper Linde 80 weld wire

.heats 61782 and 71249, which are found in the PB1 reactor vessel beltline and the

MIRVP supplemental capsule A4 contains the high copper Linde 80 weld wire heat

72442, which is found in the PB2 reactor vessel beltline. Review of the alternate

·sources of Linde 80 weld metals included in the MIRVP supplemental capsules A2 and

A4 (Table 2-3) show that the weld wire heats 61782, 71249, and 72442 are included in

other plant-specific RVSPs for Westinghouse-designed PWRs where withdrawal/testing

has been completed and/or is planned in accordance with the currently approved plant

specific RVSP withdrawal schedules:


Response to NRG Request for Additional Information (RAI) Related to Supplement to the Master Integrated Reactor Vessel Surveillance Program

Pa e 2-9

• Weld wire heat 61782 is included in the REG plant-specific RVSP

•. Weld wire heat 71249 is included in the Point Beach Unit 1 and Unit 2

supplemental surveillance capsule (Capsule "A") and in the TP3 and TP4 plant

specific RVSPs

• Weld wire heat 72442 is included in the Point Beach Unit 1 and Unit 2

supplemental surveillance capsule (Capsule "A") [Note weld wire heat is also

included in standby MIRVP supplemental capsules A 1 and L2 being irradiated in

the DB reactor.]

Based on the current withdrawal schedules, the Point Beach Unit 1 and Unit 2

supplemental capsule and the TP4 plant-specific Capsule X are scheduled to be

withdrawn at 43 EFPY (-5.07E19 n/cm2) and 38.1 EFPY (-9,297E19 n/cm2

)

respectively. High fluence capsule test data for weld wire heat 61782 is currently

available from REG Capsule N. These capsules are expected to have fluences similar

to (or exceeding) that of the fluence at the planned withdrawal time reported in BAW-

1543, Revision 4, Supplement 6-A for the MIRVP supplemental capsules A2 and A4.

Therefore, the planned withdrawal of these alternate surveillance capsules provides or

will provide the necessary data t6 support the needed reactor vessel integrity validations

for PB1 and PB2 through 60-years of operation without the need to credit the MIRVP

supplemental capsules A2 and A4. In addition, because the withdrawal schedules of

, the above alternate surveillance capsules for PB1 and PB2 are included in BAW-1543,

Revision 4, Supplement 7, the revised licensee commitment (Regulatory Commitment

11), which credits the most recent NRG-approved PWROG MIRV[S]P document,

continues to be '{alid for PB1 and PB2 current fluence projections durirg their period of

extended operation.

To provide clear direction on when the Westinghouse plant-specific RVSP capsules are

to be removed, Table 1-7 of BAW-1543, Revision 4, Supplement 7 will be updated to

reflect the target withdrawal EFPY of the remaining capsules, consistent with the current

NRC accepted withdrawal schedules.


Response to NRC Request for Additional Information (RAI) Related to Supplement to the. Master Integrated Reactor Vessel Surveillance Program ·

Pa e2-10

2.1.2.4 MIRVP Participant Surveillance Data Matrix

Figure 2-1 shows a .fluence-line, similar to a timeline, for the six (6) specific heats of filler

wire of the high copper Linde 80 weld metals included in the MIRVP supplemental

capsules A2 and A4 versus the alternate sources of surveillance data. Included also

are the projected reactor vessel 60-year and twice the 60-year fluences of these weld

wire heats for the MIRVP participating member plants (reactor vessel inside surface

fluence). For comparison purposes, the planned withdrawal fluences for the MIRVP

supplemental capsules A2 and A4 (per BAW-1543, Revision 4, Supplement 6-A) are

included within the figure.

As shown in Figure 2-1, there are or will be sufficient irradiated surveillance data

available to support all the B&W-designed and Westinghouse-designed plants with

B&W-fabricated reactor vessels with high copper Linde 80 weld metals through 60-

years of operation without withdrawal and testing of the MIRVP supplemental capsules

A2 and A4.

2.1.2.5 Impact of the MIRVP Supplemental Capsules A2 and A4 Status Change on the Objectives of the MIRVP

In the-previous BAW-1543 supplement [7], additional information was requested by the

NRC Staff concerning the MIRVP supplemental capsules A2 and A4, to which the

PWROG provided written responses in the PWROG Letter OG-07-66 dated February

15, 2007 [17]. In the responses, the PWROG provided the explanation for extending

the withdrawal of the MIRVP supplemental capsules A2 and A4 from the Crystal River

· Unit 3 seventeenth fuel cycle to the twenty-ninth fuel cycle, and the assessment of the

need for the irradiated test data from these capsules to support regulatory requirements.

The PWROG Letter OG-07-66 concluded that withdrawal a~'d testing the MIRVP

supplemental capsules A2 and A4 was not needed to fulfill the requirements of ASTM

E185-82 (10 CFR 50 Appendix H) for the current licensing periods, as shown in Table

VIII of BAW-1543, Revision 4, Supplement 6-A. However, by extending their irradiation

withdrawal, the capsules could provide supplemental (but not required) data for the

Westinghouse-designed plants.



Pa e2-11

Table 1-8 of BAW-1543, Revision 4, Supplement 7 (Table VIII in previous supplement)

has been adjusted to consider the recommendations in MRP-326 [18], NRG Staff

approved plant-specific RVSP capsule withdrawal schedules, and updated reactor

vessel EOL fluences for some units. The capsules listed in Table 1-8 of BAW-1543,

Revision 4, Supplement 7, continue to meet the requirements of ASTM E185-82 and

support the licensing periods for the current plants participating in the MIRVP without

the need for withdrawal and testing of the MIRVP supplemental capsules A2 and A4. In

addition, supplemental irradiated test data (similar to data that could result from testing

the MIRVP supplemental capsules A2 and A4) for Linde 80 weld wire heats in

Westinghouse-designed plants with corresponding fluences to 60-years of operation

could be obtained upon withdrawal and testing the PB1/PB2 supplemental capsule,

which is currently scheduled for withdrawal in the early 2020s. Therefore the withdrawal

and testing of the MIRVP supplemental capsules A2 and A4 is not necessary to meet

the requirements of ASTM E185-82 (10 CFR 50 Appendix H).

Based on the above information, the PWROG is requesting to change the status of the

MIRVP supplemental capsules A2 and A4 to "withdrawal not planned," which means

that these capsules would not be removed from the idle Crystal River Unit 3 reactor

vessel surveillance capsule holder tubes, where they currently reside, and there are no

plans for testing the contents of these capsules.

------- - ---------------- ------------------

AREVA Inc. BA W-1543Q 1-000 Revision 4 Supplement 7

Response to NRC Request for Additional Information (RAI) Related to Supplement to the Master Integrated Reactor Vessel Surveillance Program Pa e 2-12

60 60 60x2 60x2 72445 ••• • - •1- -11• •-

60~ ~ 60

60

OI - x -60x

60x2 60

72442 1--- -1..J t--- ----tr--------~- · -alH9--------; _______________ _,

6060 60 60x2 60x2 60x2

12105 • ---~-IIIIICll-----M ------

60 60x2 60 60x2 60 60 60x2 60x2

11249 - 11-•••i----••-----• - • 1--------111-0 1----,0 1----, ,-----------~ a- •--

060x2 60 60 60x2 61782 ...__. _ _ __________ ,. - -------• - •---9(-----<>---------

60 60 60x2 80 60 80 60x2 60x2 60x2 60x2

299L44 •11--11~~---•o-~ - • -A-------o-- -0-----10 - x---------------,

0 1 2 3 4 5 6 7 8 9 10

Fluence, n/cm2 (x 1011)

•Capsule Tested O CapsuleTest Planned <> Standby Capsule .A. ONSl .A.ONS2 ll. ONS3 LI. TMll 8 REG · • PBl a PB2 0 51 • s2 OTP3/TP4

X BAW-1543 Suppl. 6-A Planned Withdraw for A2/A4

Figure 2-1 Linde 80 Weld Metal Surveillance Data vs. Projected Plant 60-Year Reactor Vessel Beltline Weld

Metal Inside Surface Fluences for the Weld Wire Heats in MIRVP Supplemental Capsules A2 and A4 (Note: the AN01 and DB reactor vessels do not contain any of the applicable weld wire heats in the

beltline region)



Pa e 2-13

2.2 RA/ 2

2.2.1 Statement of RAI 2

ISSUE

Appendix H to 10 CFR Part 50 describes the requirements for reactor vessel material

surveillance programs. In addition, for license renewal, NUREG-1801, GALL Report

Aging Management Program XI.M31 Reactor Vessel Surveillance, provides additional

information regarding reactor vessel material surveillance programs for the period of

extended operation (to 60 years).

Appendix H, Section IV, Part A states, "Each capsule withdrawal and the test results

must be the subject of a summary technical report to be submitted, as specified in [10

CFR Part] 50.4, within one year of the date of capsule withdrawal, unless an extension

is granted by the Director, Office of Nuclear Reactor Regulation."

REQUEST

Table 1-8 (Comparison of the Plant-Specific Surveillance Capsules with ASTM E 185-

82 Requirements) column heading states, "EOL [end of license] or 1-2 times EOL

fluence (Capsule may be held without testing)." Staff requests revision of Table 1-8 for

consistency with Appendix H, Section IV, Part A.

2.2.2 PWROG Response to RAI 2

The last column heading in Table 1-8 (Comparison of the Plant-Specific Surveillance

Capsules with ASTM E 185-82 Requirements) will be revised by deleting the

parenthetical "(Capsule may be held without testing)." The new last column heading in

Table 1-8 (Comparison of the Plant-Specific Surveillance Capsules with ASTM E 185-

• 82 Requirements) will state "EOL or 1-2 Times EOL Fluence."



Pa e 2-14

In addition, the words "(I/NT)= Irradiated and not tested" will be removed from the

Legend to Table 1-8 because "(I/NT)" no longer appears in the table and Note 3 will be

added to Table 1-8 stating Capsule TE1-C was tested to support License Renewal for

Davis-Besse.



Pa e 3-1

3.0 REFERENCES

1. Letter from Jack Stringfellow, PWROG to NRC Document Control Desk,

OG-15-32, Subject: PWR Owners Group Submittal of BAW-1543,

Revision 4, Supplement 7, "Supplement to the Master Integrated Reactor

Vessel Surveillance Program" and Transmittal of PWROG-14066-NP,

Revision 0, "Master Integrated Reactor Vessel Surveillance Program

Long-Range Planning" to the NRC (PA-MSC-1182), dated January 23,

2015. (NRC Accession No. ML 15033A086)

2. Letter from Jonathan G. Rowley, NRC Project Manager; "Request for

Additional Information RE: Pressurized Water Reactor Owners Group

Topical Report BAW-1543, Revision 4, Supplement 7, 'Supplement to the

Master Integrated Reactor Vessel Surveillance Program' (TAC NO.

MF5701)"; dated June 3, 2016. (NRC Accession No. ML 16126A444)

3. AREVA Document 43-1543-04 (BAW-1543 Revision 4), "Master

Integrated Reactor Vessel Surveillance Program," February 1993.

4. ASTM Standard E185-82, "Standard Practice for Conducting Surveillance

Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American

Society for Testing and Materials, Philadelphia, Pennsylvania, 1982.

5. Code of Federal Regulations, Title 10, Part 50 (10 CFR 50), Appendix H,

Reactor Vessel Material Surveillance Program Requirements [60 FR

65476, Dec. 19, 1995; 68 FR 75390, Dec. 31, 2003; 73 FR 5723, Jan. 31,

2008].

6. U.S. Nuclear Regulatory Commission, NUREG-1801, Revision 2, "Generic

Aging Lessons Learned (GALL) Report," dated December 2010. (NRC

Accession No. ML 103490041)

7. AREVA Inc. Document 43-1543S-11 (BAW-1543(NP), Revision 4,

Supplement 6-A), "Supplement to the Master Integrated Reactor Vessel

Surveillance Program," June 2007. (NRC Accession No. ML072570104)



Pa e 3-2

8. AREVA Document, 47-1159862-00 (BAW-1875), "The B&W Owners

Group Cavity Dosimetry Program," August 1985.

9. Letter from Karen Cotton, NRG to David Heacock, Virginia Electric and

Power Company, Subject: Surry Power Station, Units Nos. 1 and 2 -

Safety Evaluation for Revision to Reactor Vessel Surveillance capsule

Withdrawal Schedule (TAC Nos. ME4133 and ME4134), dated January

31, 2001. (NRG Accession No. ML 103000386)

10. ~etter from Farideh Saba, NRG to Mano Nazar, NextEra Energy, Subject:

Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Review of Reactor

Vessel Material Surveillance Program - Revised Surveillance Capsule

Withdrawal Schedule (TAC Nos. ME9564 and ME9565), dated September

4, 2013. (NRG Accession No. ML 13191 A090)

11. R.E. Ginna Nuclear Power Plant Transmittal of RCS Pressure and

Temperature Limits Report (PTLR), dated September 22, 2010. (NRG

Accession No. ML 102730623)

12.Westinghouse ReportWCAP-17036-NP, Revision 1, "Analysis of Capsule

N from the R. E. Ginna Reactor Vessel Radiation Surveillance Program,"

dated September 2010.

13. Letter from Terry Beltz, NRG to Larry Meyer, NextEra Energy Point Beach,

LLC, Subject: Point Beach Nuclear Plant (PBNP), Units 1 and 2 -

Issuance of License Amendments Regarding Extended Power Uprate

(TAC Nos. ME1044 and ME1045), dated May 3, 2011. (NRG Accession

Nos. ML 110880039 and ML 110450159)

14. Letterfrom Larry Meyer, FPL Energy Point Beach, LLC to NRG Document

Control Desk, Subject: Point Beach Nuclear Plant, Units 1 and 2 License

Amendment Request 261 Extended Power Uprate, dated April 7, 2009.

(NRG Accession Nos. ML091250564, ML091250566, and ML091250569)



Pa e 3-3

15. Letter from Larry Meyer, NextEra Energy Point Beach, LLC to NRC

Document Control Desk, Subject: Point Beach Nuclear Plant Units 1 and 2

Reactor Vessel Surveillance Program Request to Change Reactor Vessel

Surveillance Specimen Withdrawal Schedule, dated January 19, 2010.

(NRC Accession No. ML 100190182)

16. Code of Federal Regulations, Title 10, Part 50.60 (10 CFR 50.60) 50.60,

Acceptance Criteria for Fracture Prevention Measures for Lightwater

Nuclear Power Reactors for Normal Operation [48 FR 24009, May 27,

1983, as amended at 50 FR 50777, Dec. 12, 1985; 61 FR 39300, July 29,

1996].

17. Letter from Frederick P. Schiffley, PWROG to NRC Document Control

Desk, OG-07-06, Subject: Pressurized Water Reactor Owners Group

Responses to the NRC Request for Additional Information (RAI) on PWR

Owners Group (PWROG) Report BAW-1543 (NP), Revision 4,

Supplement 6 "Supplement to the Master Integrated Reactor Vessel

Surveillance Program" (TAC No. MC9608) PA-MSC-0230, dated February

15, 2007. (NRC Accession No. ML070510233)

18. Materials Reliability Program: Coordinated PWR Reactor Vessel

Surveillance Program (CRVSP) Guidelines (MRP-326). EPRI, Palo Alto,

CA: 2011. 1022871.

A AREVA


Supplement 7

December 2014

AREVA Inc.

(c) 2014 AREVA Inc.

BAW-1543 Revision 4

AREVA Inc. BAW-1543 Revision 4

Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7 Page ii

Section(s) or Item Page(s)

1 General

2 Executive Summary

3 Section 1.0

4 Table 1-1

5 Table 1-2

6 Table 1-3

7 Table 1-4

8 Table 1-5

9 Table 1-6

Nature of Changes

Description and Justification

• Updated document format to the current AREVA Inc. Licensing Document template.

• Updated to reflect current revision.

• Addition of footnote (5) for reference to RAI responses (-A version).

·• Included revision statements for Supplement 7 changes.

• Deleted changes made in Supplement 6.

• Added descriptions of Supplement 7 changes.

• Updated to Supplement 7-A (-A version) .

. • Removed the word "current" when describing Table 1-9 (-A version).

• Deleted the Crystal River Unit 3 capsule detailed summary information.

• Added information on the TE1-C capsule for Davis-Besse.

• Added the R. E. Ginna capsule detailed summary information.

• Updated capsule status for various capsules.

• Updated capsule status for various capsules.

• Updated the Westinghouse plant-specific surveillance program status based on the CRVSP and their respective NRC approved plant-specific RVSP withdrawal schedules.

• Added the R. E. Ginna capsule insertion and withdrawal schedule.

• Updated the TE1-C capsule status.

• Deleted the Crystal River Unit 3 plant-specific surveillance capsule summary status information.


Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7 Page iii

.10 Table 1-7

11 Table 1-8

12 Table 1-9

• Updated the time of removal of the supplemental capsules A2 and A4 to "Not Planned."

• Deleted Expected/Received fluence values for the QC 1-D and OC3-F capsules. ·

• Updated Expected/Received fluences for TMl1-A, TE1-E, A3, and L 1 capsules.

• Added Note 5 for TMl1-C capsule (-A version).


• Added the R. E. Ginna plant-specific surveillance capsule summary status information.

• Added Expected/Received fluence for the S2-W1 capsule.

• Updated time of removal for capsules to EFPY from RAI 1 response (-A version).

• Deleted the Crystal River Unit 3 plant-specific surveillance information.

• Added the R. E. Ginna plant-specific surveillance information.

• Updated the Westinghouse plant-specific surveillance program information based on the CRVSP and their respective NRC approved plant-specific RVSP withdrawal schedules.

• Updated the end-of-license or 1-2 times end-of-license fluence capsules for some units based on fluence values in Table 1-9.

• Updated last column heading from RAI 2 response (-A version).

• Removed (I/NT) from legend from RAI 2 response (-A version).

• Added Note 3 from RAI 2 response (-A version).

• Updated licensing dates and projected reactor vessel peak end-of-license fluences for some units.



Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7 Page iv

13 Section 2.0

· • Added the R. E. Ginna plant-specific surveillance capsule information.

• Updated licensing date and projected reactor vessel peak end-of-license fluence for Davis-Besse (-A version)

• Updated (including renumbering) references to include reports for the CRVSP and R. E. Ginna and deletion of references for the Crystal River Unit 3 plant-specific surveillance capsules.

AREVA Inc.

Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7

Contents

BAW-1543 Revision 4

Pagev

Page

1.0 INTRODUCTION ............................................................................................... 1-1

2.0 REFERENCES .................................................................................................. 2-1


Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7 Page vi

List of Tables

Table 1-1: B&W 177-FA Plant-Specific Reactor Vessel Surveillarice Program-Detailed Summary .................................................................................. 1-3

. Table 1-2: Westinghouse Plant-Specific Reactor Vessel Surveillance Program -Detailed Summary .................................................................................. 1-5

Table 1-3: Capsule Insertion and Withdrawal Schedule for Crystal River Unit 3 ......... 1-7

Table 1-4: Capsule Insertion and Withdrawal Schedule for Davis-Besse .................. 1-10

Table 1-5: Capsule Insertion and Withdrawal Schedule for Westinghouse Plant-. Specific RVSPs with B&W Fabricated Reactor Vessels ....................... 1-13

Table 1-6: Summary Status of the B&W Surveillance Capsules ............................... 1-16

Table 1-7: Summary Status of the Westinghouse Surve.illance Capsules ................. 1-19

Table 1-8: Comparison of the Plant-Specific Surveillance Capsules with ASTM E 185-82 Requirements ........................................................................... 1-22

Table 1-9: Peak End-of-Life Inside Surface Fluences and Significant Licensing Dates .................................................................................................... 1-23

AREVA Inc.

· Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7

Executive Summary

BAW-1543 Revision 4

Page vii

The Master Integrated Reactor Vessel Surveillance Program (MIRVP) was initiated in

1976 for the Babcock & Wilcox (B&W) 177-FA Plants. Its purpose was to augment the

existing reactor vessel surveillance programs and to provide a basis for sharing

information between plants. All of the early vintage B&W manufactured reactor vessels

were fabricated using the submerged arc welding process and particular consumables

which resulted in welds that are sensitive to fast neutron fluence. exposures. The welds

in these early vintage B&W manufactured reactor vessels are referred to as the Linde

80 class of materials. In 1988, the MIRVP was further modified to include a series of

plants with the Westinghouse Nuclear Steam Supply System (NSSS) for which B&W

manufactured the reactor vessels. These vessels have virtually identical welds as those

used in the B&W 177-FA plants. The overall objective of the MIRVP is to provide the

data necessary to assure compliance with Federal Regulations.(1) Individual

Pressurized Water Reactor Owners Group (PWROG) members may cite this document,

as needed, in support of their reactor vessel surveillance program (RVSP), which must

meet the requirements of Code of Federal Regulations, Title 10, Part 50 (10 CFR 50)

Appendix H, "Reactor Vessel Material Surveillance Program Requirements."

This document is a supplement to the base document, "Master Integrated Reactor

Vessel Surveillance Program," BAW-1543, Revision 4_(2) Both the base document and

the current supplement are used to document the progress of the MIRVP, especially the

withdrawal schedule of the RVSP capsules. The last full revision to the base document

reviewed and approved by the NRC is Revision 4_(3) This document is being revised to

adjust the withdrawal schedule of four (4) surveillance capsules to provide irradiated

<1> Code of Federal Regulations, Title 10, Part 50, "Domestic Licensing of Production and Utilization

Facilities", U.S. Nuclear Regulatory Commission, Washington, D.C.

<2> B&W Nuclear Technologies, Inc. Document, BAW-1543, Revision 4 (43-1543-04), "Master Integrated

Reactor Vessel Surveillance Program," February 1993. (Available through AREVA Inc.)

<3> Nuclear Regulatory Commission Safety Evaluation Report, "Babcock & Wilcox Owners Group

(B&WOG) Reactor Vessel Working Group Report," BAW-1543, Revision 4, Supplement 2, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" (TAC No. M98089), July 11, 1997.

AREVA Inc.


BAW-1543 Revision 4

Page viii

material data corresponding to reactor vessel fluences for the Westinghouse-designed

NSSS reactor vessels between 1 and 2 times the 60 year fluence. The last supplement

to this document reviewed and approved by the NRC is BAW-1543(NP), Revision 4,

Supplement 6-A.(4) Changes reflected in the NRG-approved version (-A) of this

document are in response to requests for additional information (RAls)_(5)

<4> AREVA Inc. Document, "Supplement to the Master Integrated Reactor Vessel Surveillance Program,"

BAW-1543(NP), Revision 4, Supplement 6-A (43-1543S-11), June 2007.

<5> Letter OG-16-261 from PWR Owner's Group to U.S. Nuclear Regulatory Commission, "Transmittal of

Response to Supplement to the Master Integrated Reactor Vessel Surveillance Program, BAW-1543Q1-000, Revision 4 Supplement 7 (PA-MSC-1182RO)," August 4, 2016. (NRC Accession No. ML 17256A514.)

AREVA Inc.


1.0 INTRODUCTION

BAW-1543 Revision 4

Page 1-1

The AREVA Inc. document, BAW-1543, Revision 4, reports the essential features of a

· Master Integrated Reactor Vessel Surveillance Program (MIRVP) for all operating B&W

177-FA plants and those Westinghouse-designed plants having B&W-fabricated reactor

vessels. 1 This supplementary document to BAW-1543, Revision 4, contains

information on the surveillance capsule insertion and withdrawal schedules for the B&W

177-FA plants and the Westinghouse-designed plants with reactor vessel containing

high copper Linde 80 weld metals. In addition, the insertion and withdrawal schedules

for the MIRVP supplementary capsules are provided. This document, Supplement 7-A,

is a revision to and replaces Supplement 6-A in its entirety. Table 1-1 and Table 1-2 are

listings of plant-specific surveillance capsules and direct the reader to the appendices of

BAW-1543, Revision 4, where additional information can be found on material and

capsule specifications. These tables also provide a listing of surveillance capsule

reports available as of the date of this document. Table 1-1 provides information for the

B&W 177-FA plant-specific capsules and Table 1-2 provides information on the plant

specific capsules for the Westinghouse-designed plants having B&W-fabricated reactor

vessels with high copper Linde 80 weld metals.

Table 1-3 and Table 1-4· provide capsule insertion and withdrawal schedules for B&W

177-FA host plants Crystal River Unit 3 and Davis-Besse, respectively. Table. 1-3 was

revised by changing the capsule status of the OC1-D capsule, the OC3-F capsule and

the MIRVP supplemental capsules A2 and A4. The status of the TE1-C capsule has

been updated in Table 1-4 to reflect that testing· of the irradiated specimens has been

performed.

Table 1-5 provides capsule insertion and withdrawal schedules for the Westinghouse

designed plants with reactor vessel containing high copper Linde 80 weld metals.

These listed plant-specific schedules are not MIRVP commitments, but merely reflect

the current capsule withdraw plans for these plants in accordance with their respective

. NRC approved plant-specific RVSP schedules. Table 1-5 was revised to update the

AREVA Inc.


BAW-1543 Revision 4

Supplement 7 Page 1-2

plant-specific surveillance program status based on the Coordinated Reactor Vessel

Surveillance Program (CRVSP)2 and their respective NRC approved plant-specific

RVSP withdrawal schedules.

Table 1-6 and Table 1-7 summarize the status of all MIRVP and plant-specific RVSP

capsules for B&W 177-FA and Westinghouse-designed plants with reactor vessels

containing high copper Linde 80 weld metals, respectively. These tables state whether

the capsules have.been withdrawn or are still being irradiated. For capsules that have

been withdrawn and tested, the appropriate surveillance capsule report number has

been listed. For those capsules that are being irradiated, the target and expected

fluences are listed along with the insertion and/or withdrawal date. Table 1-6 was

revised by changing the status of the TE1-C capsule. In addition, the time of withdrawal

of the supplemental capsules A2 and A4 was changed to "Not Planned." Table 1-7 was

revised to update the Westinghouse plant-specific surveillance program status based on

the CRVSP and their respective NRC approved plant-specific RVSP withdrawal

schedules. Note that the Westinghouse-designed plant-specific withdrawal schedules,

listed in Table 1-7, are not MIRVP commitments, but merely reflect the current capsule

withdraw plans for these plants in accordance with their respective NRC approved

plant-specific RVSP schedules.

Table 1-8 .shows the conformance of the current PWROG MIRVP member plant-specific

surveillance programs to the requirements of ASTM E185-82.3

Table 1-9 lists licensing dates and projected reactor vessel peak end-of-license

fluences.

AREVA Inc.


BAW-1543 Revision 4

Page 1-3

Table 1-1: B&W 177-FA Plant-Specific Reactor Vessel Surveillance Program -Detailed Summary

Table of Material Table of Capsule Irradiation Report Capsule Type Specifications Specifications Site Date Report

[Note 1] [Note 2]

Oconee Unit 1 Topical Report BAW-10006A, Revision 34

A I A-1 D-1 OC1/CR3 Aug. 84 BAW-18375

B II A-1 D-1 OC1/CR3 ----- -----

C I A-1 D-1 OC1/CR3 Oct. 88 BAW-20506

D II A-1 D-1 OC1/CR3 ----- -----E I A-1 D-1 OC1 Sept. 77 BAW-14367

F II A-1 D-1 OC1 Sept. 75 BAW-1421 Rev. 18

Oconee Unit 2 Topical Report BAW-10006A, Revision 3

A I A-2 D-2 OC2/CR3 Dec. 81 BAW-16999

B II A-2 D-2 OC2/CR3 ----- -----C I A-2 D-2 OC2 May77 BAW-143?1° D II A-2 D-2 OC2/CR3 ----- -----E I A-2 D-2 OC2/CR3 Oct. 88 BAW-2051 11

F II A-2 D-2 OC2/CR3 ----- -----

Oconee Unit 3 Topical Report BAW-10100A12 [Note 3]

A V A-3 D-3 OC3 Jul. 77 BAW-143813

B VI A-3 D-3 OC3/CR3 Oct 81 . BAW-169?14

C V A-3 D-3 OC3/CR3 ----- -----D VI A-3 D-3 OC3/CR3 May92 BAW-2128 Rev. 115

E V A-3 D-3 OC3/CR3 ----- -----F VI A-3 D-3 OC3/CR3 ----- -----

Three Mile Island Unit 1 Topical Report BAW-10006A, Revision 3

A I A-4 D-4 TMl1/TMl2 [Note 4] BAW-204i6

B II A-4 D-4 TMl1/CR3 ----- -----C I A-4 D-4 TMl1/CR3 Mar. 86 BAW-1901 17

D II A-4 D-4 TMl1/CR3 ----- -----E I A-4 D-4 TMl1 Jan. 77 BAW-143918

F II A-4 D-4 TMl1/CR3 ----- -----

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BAW-1543 Revision 4

Page 1-4

Table 1-1 (cont.): B&W 177-FA Plant-Specific Reactor Vessel Surveillance Program - Detailed Summary

Table of Material Table of Capsule Irradiation Report

Capsule Type Specifications Specifications Report [Note 1] [Note 2]

Site Date

Arkansas Nuclear One· Unit 1 : Topical Report BAW-10006A, Revision 3

A I A-6 D-6 AN01/D81 Jul. 84 BAW-183619

B II A-6 D-6 AN01/D81 Nov. 81 BAW-169820

C I A-6 D-6 AN01/DB1 Oct. 89 BAW-2075 Rev. 121

D II A-6 D-6 AN01/D81 ----- -----E I A-6 D-6 AN01 Apr. 77 BAW-144022

F II A-6 D-6 AN01/D81 ----- -----

Davis-Bess·e ··

Topical Report BAW-10100A

A Ill .A-8 D-8 D81 Jun.89 BAW-1882 Rev. 123

B IV A-8 D-8 D81 May84 BAW-183424 & Jun.85 BAW-186725

C Ill A-8 D-8 D81 Oct. 14 ANP-333926

D IV A-8 D-8 D81 Dec. 90 BAW-212527 & Oct. 93 BAW-220828

E Ill A-8 D-8 D81 ----- -----F IV A-8 D-8 D81 Jan.82 BAW-1701 29 &

Mar. 82 BAW-171930

Notes for Table 1-1:

1. Refer to BAW-1543, Revision 4, Appendix A.

2. Refer to BAW-1543, Revision 4, Appendix D.

3. The Oconee Unit 3 capsules were fabricated before BAW-101 OOA was published; however, it is the

Oconee Unit 3 program that is described in BAW-10100A. ·

4. Capsule used for Three Mile Island Unit 2 capsule requalification.

AREVA Inc.


BAW-1543 Revision 4

Page 1-5

Table 1-2: Westinghouse Plant-Specific Reactor Vessel Surveillance Program -Detailed Summary

Table of Material Table of Capsule Report

Capsule Type Specifications Specifications Report [Note 1] [Note 2] Date

Point Beach Unit 1 WCAP-751331

N IV A-10 D-10 ----- -----p IV A-10 D-10 ----- -----R Ill A-10 D-10 Aug. 78 WCAP-935732

s IV A-10 D-10 Nov. 76 WCAP-873933

.T Ill A-10 D-10 Dec. 84 WCAP-1 073634

V Ill A-10 D-10 Jun. 73 BCL Report35

Point Beach Unit 2 WCAP-771236 & WCAP-1585637

N IV A-11 D-11 ----- -----p IV A-11 D-11 ----- -----R V A-11 D-11 Dec. 79 WCAP-963538

s V A-11 D-11 Aug. 91 BAW-214039

T IV A-11 D-11 Aug. 78 WCAP-9331 40

V V A-11 D-11 Jun. 75 BCL Report41

Suppl. Suppl. [Note 3] [Note 3] ----- -----R. E. Ginna WCAP-725442

N II A-9 D-9 Sept. 10 WCAP-17036NP Rev. 143

p II A-9 D-9 ----- -----R I A-9 D-9 Nov. 74 WCAP-8421 44

s II A-9 D-9 Dec. 93 WCAP-1390245

T I A-9 D-9 Apr. 82 WCAP-1008646

V I A-9 D-9 Mar. 73 FR-RA-1 47

Surry Unit 1 WCAP-772348

s VI A-12 D-12 ----- -----T VII A-12 D-12 Jun. 75 BCL Report49

u VI A-12 D-12 ----- -----

V VII A-12 D-12 Feb. 87 WCAP-1141550

w VI A-12 D-12 Mar. 79 BCL-585-8R51

X VII A-12 D-12 Apr. 98 BAW-232452

y VI A-12 D-12 ----- -----z VII ·A-12 D-12 ----- -----

AREVA Inc.


BAW-1543 Revision 4

Page 1-6

Table 1-2 (cont.): Westinghouse Plant-Specific Reactor Vessel Surveillance Program - Detailed Summary

Table of Material Table of Capsule Report Capsule Type Specifications Specifications Date

Report [Note 1] [Note 2]

Surry Unit 2 WCAP-808553

s VIII A-13 D-13 Dec. 96 WCAP 1481054

T VIII A-13 D-13 ----- -----u VIII A-13 D-13 ----- -----V VIII A-13 D-13 Jun.87 WCAP-1149955

w VIII A-13 D-13 Feb. 81 BCL-585-02656

X VIII A-13 D-13 Sept. 75 BCL Report57

y IX A-13 D-13 Feb.03 WCAP-16001 58

z IX A-13 D-13 ----- -----

Turkey Point Unit 3 WCAP-765659

s VI A-14 D-14 May 79 SwRl-02-5131 60

T VII A-14 D-14 Dec. 75 WCAP-8631 61

u VI A-14 D-14 ----- -----V VII A-14 D-14 Aug. 86 SwRl-06-857562

w VI A-14 D-14 ----- ------

X VII A-14 D-14 Sept. 02 WCAP-1591663

y VI A-14 D-14 ----- -----z VI A-14 D-14 ----- -----

Turkey Point Unit 4 WCAP-766064

s VI A-15 D-15 May 79 SwRl-02-538060

T VII A-15 D-15 Dec. 75 SwRl-02-4221 65

u VI A-15 D-15 ----- -----V VII A-15 D-15 ----- -----w VI A-15 D-15 ----- -----X VII A-15 D-15 ----- -----y VI A-15 D-15 ----- -----z VI A-15 D-15 ----- -----

Notes for Table 1-2: .

1. Refer to BAW-1543, Revision 4, Appendix A.

2. Refer to BAW-1543, Revision 4, Appendix D.

3. This Westinghouse-designed supplemental capsule contains Charpy, compact fracture toughness

and tensile specimens fabricated from Linde 80 welds SA-1101 (Heat No. 71249), WF-67 (Heat No.

72442), and WF-182-1 (Heat No. 821T44).

AREVA Inc.


BAW-1543 Revision 4

Page 1-7

Table 1-3: Capsule Insertion and Withdrawal Schedule for Crystal River Unit 3

Holder Tube Location in Insert Withdraw Capsule Status Holder Tube [Note 1]

Installed at Initial Fuel Load

xw Top CR3-B (WC) [Note 2] ----- ---xw Bottom CR3-D (WC) ----- ---End of First Fuel Cycle (1A)

wz Top CR3-LG1 (WC) ----- ---wz Bottom CR3-LG2 (WC) ----- ---ZY Top CR3-C (W) [Note 3] ----- ---ZY Bottom CR3-A (W) ----- ---

YZ Top OC2-A (W) ----- ---YZ Bottom OC1-A (W) ----- ---YX Top OC2-E (W) ----- ---

YX Bottom OC3-D (W) ----- ---xw Top CR3-E (W) CR3-B (WC) Tested wx Top OC3-B (W) ----- ---

wx Bottom CR3-F (WC) ----- ---End of First Fuel Cycle (1 B)

No Changes.

End of Second Fuel Cycle

YZ Top OC1-C (W) OC2-A (W) Tested wx Top TMl1-C (W) OC3-B (W) Tested

End of Third Fuel Cycle

No Changes.

End of Fourth Fuel Cycle

YZ Bottom OC1-B OC1-A (W) Tested

wz Top None CR3-LG1 (WC) Tested

wz Bottom None CR3-LG2 (WC) Tested (WZ now empty)

End of Fifth Fuel Cycle

wx Top OC3-C (W) TM11-C (W) Tested

xw Bottom TMl1-B CR3-D (WC) Tested ZY Top OC3-F (W) CR3-C (W) Tested wz Top OC2-B None wz Bottom CR3-LG2 (WC) None

(WZ no longer empty)

End of Sixth Fuel Cycle

YX Top TMl2-D [Note 4] OC2-E (W) Tested wx Bottom TMl1-F CR3-F (WC) Tested YZ Top TMl2-LG1 (WC) OC1-C (W) Tested YZ Bottom TMl2-LG2 (WC) OC1-B (a)

AREVA Inc.


BAW-1543 Revision 4

Page 1-8

Table 1-3 (cont.): Capsule Insertion and Withdrawal Schedule for Crystal River Unit 3

Holder Tube Location in Insert Withdraw Capsule Status Holder Tube [Note 1]

End of Seventh Fuel Cycle

xw Bottom TMl2-D [Note 4] TMl1-B (a) from YX top

YX · Top A2 (WC) TMl2-D [Note 4] ---

to XW bottom YX Bottom A4 (WC) OC3-D (W) Tested wz Top OC3-E (W) OC2-B (a)

End of Eighth Fuel Cycle

ZY Bottom OC1-D CR3-A (W) (a) xw Top None CR3-E (W) (a) xw Bottom None TMl2-D [Note 4] ---

(XW now empty) wx Top OC2-F OC3-C (W) (a) wx Bottom TMl1-D TMl1-F (a)

End of Ninth Fuel Cycle

YZ Top OC2-D TMl2-LG1 (WC) Tested wz Bottom TMl2-D [Note 4] CR3-LG2 (WC) Tested

End of Tenth Fuel Cycle

No Changes.

End of Eleventh Fuel Cycle

wx Top None OC2-F (a) wx Bottom None TMl1-D (a)

(WX now empty)

End of Twelfth Fuel Cycle

YZ Top None OC2-D (a) YZ Bottom None TMl2-LG2 (WC) Tested

(YZ now empty) wz Top None OC3-E (W) (a) wz Bottom None TMl2-D [Note 4] (a)

(WZ now empty)

End of Thirteenth through Twenty-eighth Fuel Cycles

No Changes.

Withdrawal Not Planned

YX Top None A2 (WC) (b) YX Bottom None A4 (WC) (b) ZY Top None OC3-F (W) (b) ZY Bottom None OC1-D (b)

AREVA Inc.



1. (a) = Capsule has been disposed of in accordance with Reference 66.

(b) = Capsule resides in the reactor vessel with no planned withdrawal.

2. (WC) = Capsule contains weld metal and compact fracture toughness specimens.

3. (W) = Capsule contains weld metal specimens.

4. Dummy capsule.

BAW-1543 Revision 4

Page 1-9

AREVA Inc.


BAW-1543 Revision 4

Page 1-10

Table 1-4: Capsule Insertion and Withdrawal Schedule for Davis-Besse

Location in - Capsule Status Holder Tube Holder Tube Insert Withdraw [Note 1]

Installed at Initial Fuel Load

wz Top AN1-B ----- ---wz Bottom RS1-B (WC) [Note 2] ----- ---ZY Top TE1-B (WC) ----- ---

ZY Bottom TE1-F (WC) ----- ---YZ Top AN 1-A (W) [Note 3] ----- ---YZ Bottom AN1-C (W) ----- ---YX Top RS1-D (WC) ----- ---YX Bottom TE1-C (W) ----- ---xw Top TE1-D (WC) ----- ---xw Bottom RS1-C (W) ----- ---wx Top TE1-A (W) ----- ---wx Bottom RS1-F (WC) ----- ---

End of First Fuel Cycle

wz Top DB1-LG1 (WC) AN1-B Tested wz Bottom RS1-E (W) RS1-B (WC) Tested ZY Bottom DB1-LG2 (WC) TE1-F (WC) Tested

End of Second Fuel Cycle

YX Top RS1-A (W) RS1-D (WC) Tested

End of Third Fuel Cycle

YZ Top AN1-D AN1-A (W) Tested ZY Top TE1-E (W) TE1-B (WC) Tested

End of Fourth Fuel Cycle

YX Top AN1-F RS1-A (W) (a) wz Top RS1-F from WX bottom DB1-LG1 (WC) Tested wx Top None TE1-A (W) Tested wx Bottom None . RS1-F to WZ top ---

(WX now empty)

End of Fifth Fuel Cycle

wz Top None RS1-F (WC) Tested wz Bottom None RS1-E (W) (a)

(WZ now empty) YZ Top TMl2-C [See Note 4] AN1-D to XW- ---

YZ Bottom TMl2-E [See Note 4] bottom

xw Bottom AN1-D from YZ top AN1-C (W) Tested RS1-C (W) (a)

AREVA Inc.


BAW-1543 Revision 4

Page 1-11

Table 1-4 (cont.): Capsule Insertion and Withdrawal Schedule for Davis-Besse

Location in Capsule Status Holder Tube Holder Tube Insert Withdraw [Note 1]

End of Sixth Fuel Cycle

xw Top None TE1-D (WC) Tested xw Bottom None AN1-D (a)

(XW now empty) YZ Top A3 (WC) TMl2-C [Note 4] ---YZ Bottom A1 (WC) TMl2-E [Note 4] ---wz Top L2 (WC) ----- ---

wz Bottom L 1 (WC) ----- ---End of Seventh Fuel Cycle

YX Top EPRI Cap. [Note 4] AN1-F (a) YX Bottom A5 TE1-C (W) Tested wx Top IBSP-2 [Note 4] ----- ---wx Bottom IBSP-1 [Note 4] ----- ---

(WX no longer empty)

End of Eighth through Tenth Fuel Cycles

No Changes.

End of Eleventh Fuel Cycle

ZY Top None TE1-E (W) (a) ZY Bottom None DB1-LG2 (WC) Tested

(ZY now empty) YX Top None EPRI Cap. [Note 4] ---YX Bottom None A5 (WC) Tested

(YX now empty)

End of Twelfth Fuel Cycle

YZ Top Dummy-L2 (WC) A3 (WC) Tested wz Top None L2 (WC) to YZ top ---wz Bottom None L 1 (WC) Tested

(WZ now empty) wx Top None IBSP-2 [Note 4] ---wx Bottom None IBSP-1 [Note 4] ---

(WX now empty)

End of Thirteenth Cycle through Current Operation

No Changes.

Withdrawal Not Planned

YZ Top None Dummy-L2 (WC) (a); (b) YZ Bottom None A1 (WC) (b)

(all holder tubes empty)

AREVA Inc.



1. (a) = Capsule has been disposed of in accordance with Reference 66.

BAW-1543 Revision 4

Page 1-12

(b) = Capsule A1 and L2 duplicate the materials contained in Capsule A4. Capsule A1 and L2 are

considered standby capsules.

2. (WC) = Capsule contains weld metal and compact fracture toughness specimens.


4. Not part of the MIRVP.


Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7 Page 1-13

Table 1-5: Capsule Insertion and Withdrawal Schedule for Westinghouse PlantSpecific RVSPs with B&W Fabricated Reactor Vessels

Plant

Point Beach Unit 1

Point Beach Unit2

R.E. Ginna

Capsule Location [Note 1]

13°

Capsule Identification

[Note 2]

V(WC)

13° R (WC) .............. ______ .... _. __ ... _ ............................................................................ .

23° T (WC) ........................................................................ .................................... ..

23° p (W)

33°

33°

13°

S(W)

N(W)

V(WC)

Insert Withdraw

EOC-1

EOC-5

Capsule Status [Note 3]

Tested

Tested

EOC-11 Tested ................................................................................................ _ ...................................................................... _,_ ........................................ ..

EOC-21 Removed/Stored

EOC-3

EOL

EOC-1

Tested

(b); (d); (e); (i)

Tested

13° R (WC) EOC-5 Tested OHOHH-MOMOHHH0--0-HH,-HHOHOH OMOH HHHOHH,HHH-000-HOO••-•H-O•H-•HOOO•OHOH-HMOHHHOHOH •OOOHHH,H 000-H••HH••HO-mHO-OHHMOH ••• ••••o••-•-•••oH••HH•H•OoH,o•HMOHOMO, .. OH,•oo .. •o•ooHOO• .. • .. H••••O•••••••• 0

23° T (W) EOC-3 Tested 0HHH0-0H _______ -----.. H--·--·--·--------· - -HH ____________ HH0-···---···---····-······-·---H·-·---··--·-H••HH ••••• _____ H ____________ H __________ H•HOOHOHOHHHH••

__ ?.~~------- ________ P (W) ----·-·· _ -··------· ---~9-s;_::_?.? __________________ R~~~~:.~~!~~:.~-33° S(WC) EOC-16 Tested

.... HOHOHHOHHmOHHHOH ...... ·-·- .... H ........ - .......... _H ___________________ • ___________ HHO ··--···-·-HOHHHOHH-•oHOHOHOHHH- ..................... H ...................... H ••••••••••••••••••••••••••••••••••••••••••••••••••••••••••• --·····-·-·HHHOHH••••••-•HHHO•O•HHOHH•••••·····•H••·•····-·········

33° N(W) EOL (b); (d); (e) ••••••••HOHH••••••HOHH•O••••H•H••••••••HOH•••l•·········••••••••······················•••••••••••••••••••••••••···················l················••••••••••••••••••••••••••••••••••••••·I·········································································································•·

13° Suppl. (WC) EOC-25 Year 2022 (a)

13° V (WC) EOC-1 Tested. ····-····-·--·-·- ·-·--·------·-- ·----- ·----····-----·------·--·--······--········-···--·-· ----·--·····--········--·-··············

13° R(WC)

T(WC)

p (WC)

EOC-3

EOC-9

-EOC 41

Tested

Tested

(g)

33° S (WC) EOC-22 Tested -·-----·- ·-·-··-··-·-··-·HHH•H•HOOH•H•-····----··· •••••••••••••••··-·- •••••••OHOHOOOH••--OOHOOOOH-0H•••---••••• ••••••••••••••••••••·····-···--·-·-·-----····--·--·--····---·-··-·--·--·- ··-··--·····-·-·····--········--··-·---··----·---·--···-·

33° N (WC) EOC-33 Tested

Surry Unit 1 15° T (WC) EOC-1 Tested •o•OOH•H•-•-•o••OHOH•O••OOOOO•OH •OHOOOoOHOH•O•OO•••HOHoOOHHoHOo•••HOHooH••••H ••••-••••••OOHHOOH00Ho-OOHHOO• H OOHOOOHOHOHHO .. HHHHOHO-OH-HHHHHHHHOH HHHHOHOHO HOOHHHOHHHO-OHOHOOHHOHHOHOHHOHHOHOHOHHH

15° V (WC) EOC-8 Tested OOHOHOHOHHH-OHOHOHHOOO-H "HHOHOH OHOHHHHO-HOHOOOHHOOHHHOHHOHHOHHHOHOHHHHHH HHHOHOHOOHOO• •H ... OOHOHHOOHO-OHOH ... OOHOHOHHOH-HOOHO• •••••• ••••• OH•HH•••-•H00-000•00HOH000HOOOH•H0H•OOOOOHH0HHOOOOH•HHOH ... OHHHO-HOHO

35° W EOC-4 Tested [Note 4] •••OOHOHO•••HOO•OOH•-•••••••-•- oo00HOH••• oHOHOOHO••-•••----••••-•-HOOHOHO-HOHoOH_O ____ OHOO•O-OH oHHoO---•••••-•--••-••H•-•-••••••••H• ••• ••••OOoOH•••-•••-OHOHOOO .. _OH-HH_O ________ HHHHOOOHHH--OHHHHO OHO--HOO-OOH•HHOOHOOH_H ______ OHO•OOHOHH-HoH••-••

25° S EOL (d); (h) OH ____ HHHOH0-00-00-••--- -OHHHO-HHHHHO ___________ HOH_H_OHH-OH ·-·--------·-HOMO HH-OHHOOHHHOHOH ______ HHOH--HOHHOH-OHO-HOHOHH _____ , ------·-•HH-HHOHHOH_O_O_O_H_O_O ________ _

25° X (WC) EOC-12 (c) 15° X (WC) EOC-12 EOC-14 Tested

OHOOHOHOOHHHHHHOHHHOHOHHOHHOOH0HH HOOH00000H00•o0HOHOHH•00000HoH00•oH00HHO>o .. 0000H000H0000>HoOOH00• .. •ooH0HOHOOOHOOOHOH00•H000H0H00 .. ,n,m .. ,,., ... 000o0000o0,0HHo000•H••••--••--•" ................................................................................ ,,,, .......................................................................... .

25° Z (WC) EOC-12 (c) 15° z (WC) EOC-12 Year 2025 (a); (e)

-------•-••---·-•--••-••H•H••• •- •••••••-•••-•-••••HH-•-·-··-·-•-•HHOHO-HOHOH••HOH•-••---••·-··----•HO-HHH------------------------·---··-·-·-·-·--- -----·--·-·-·---·-·

35° Y EOC-14 (c) 15° Y EOC-14 EOL (d); (h)

45° 25°

u u EOC-12

EOC-12 EOL

(c) (d); (h)



Table 1-5 (cont.): Capsule Insertion and Withdrawal Schedule for Westinghouse Plant-Specific RVSPs with B&W Fabricated Reactor Vessels

Plant

Surry Unit 2

Turkey Point Unit 3

Turkey Point Unit4

Capsule Location [Note 1]

Capsule Identification

[Note 2] Insert Withdraw Capsule Status

[Note 3]

15° Tested X (W) EOC-1 .................................................. --.-····-············--·'·-·······--·····-···-+·--·········-······-····················l···································-········-···········-····················-··············l······-·-··········-······-··-··················

15° V (W) EOC-8 Tested --... - .. -·----·-·-··-····-·-···t················· .. ·········-·····-···--·········-······-········l·······-··--·-··-···-·--········l·····-··-·················-·········-·························-···················t·····-·-·-·····-·-··· .. ···-·-··----·-·-.. -.. ---·-·--··-··

25° W (W) EOC-4 Tested [Note 4]

25° Y (WC) EOC-12 15° Y (WC) EOC-12 EOC-17

.......................... _ ................................................................................................ . 25° U(W) Year2009

·---~-~-:....... U (W) Year 2009 Year 2027 35° Z (WC) EOC-12 25° Z (WC) EOC-12 EOL

·····-· .. ··-·-·····-·····-·-·····----·-··- -----···--·--------.. -·--·-·--·-· ..... --.. -·-·-·-·--·---·--·--.... ·································-·--··"'-·····

.................... -·-·-···-.. -·-( c)

Tested

(c) (a); (e)

(c) (d); (h)

35° T (W) EOC-17 (c) 15° T (W) EOC-17 EOL (d); (h)

····-··-·-.. --.. ····-· .•.• • ................. - ................. --,-·-"-"'"·-·--.. ,--.. -·····-------.... - .• -.,---·-----.. , ............. ______ ._ .. , ...... j ..................... -,_· .. --·-···-·------1 45° S (W) EOC-13 (f)

15° W1 (WC) [Note 5] EOC-10 EOC-14 Tested oo T(WC) EOC-1 Tested

10° s EOC-4 Tested ........................ _........... .. .......... _________ .... , ................... _______ ., ____ ,.___________ .................... ·-··-·--·-· ... ·-·-·-·-.. ·•· ............................................... -............. . ..... _._,.,_,_, .......... -........ - ... -..................... --.. ···-·-· .. ····-·····-·-·-·· .............. ..

20° V (WC) EOC-9 Tested

Year 1990 40° oo

X(WC) X(WC) Year 1990 Year 2002

30° U EOL ......................................................................................................... -.................................................................. -....................................................... ..

30° Y EOL

40°

40° oo

10°

w z T(WC)

s ..... -............................... ····-·-··· ·-·--··-"-····-··--·-··--··-·-·--·"--"·"--·-·-.. ---- ·-.. ---.. -·-·--·-·-·-· .. ··-·-

40° X (WC) oo X (WC) Year 1990

"'""··-·-········-·-··· ................................................................. -...................................................................................................... .

20° V(WC)

30° u 30° y

EOL

EOL

EOC-1

EOC-3

Year1990 -Year 2021

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

EOL

EOL

EOL

(c) Tested

(d); (h) . ........................... -........................... _ .............. .

(d); (h

(d); (h)

(d); (h)

Tested

Tested

(c) (a); (g)

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

(d); (h) ··········-··-.. ···········-· ········-·-·-··-·-.. ··--·-----·-·--·-----

( d); (h) ...........................................................................................

(d); (h) ----·-·-·-·-··-·······-··· ·-·---···-······-·--·--·-·-·--·-·-·-·--- ··---·----·-··· ······--······-········-····--·--····---.. ··--·--------··--·-· ·-----·-·-·--···-·-········-

400 w EOL (d); (h) ........................... -... , ................... -, ................................................. .

40° z EOL (d); (h)

AREVA Inc.



BAW-1543 Revision 4

Page1-15

1. Capsule locations are relative with regard to quadrant; e.g., 0° is equivalent to 90°, 180°, or 270°.


(WC) = Capsule contains weld metal and compact fracture toughness/wedge opening loading (WOL)

specimens.

3. (a) = Capsule to be removed, specimens will be tested, dosimetry evaluated, and thermal monitors

evaluated.

(b) = Capsule to be removed and placed in storage. Dosimetry may be evaluated at this time.

(c) = Capsule reinserted in higher lead factor location.

(d) = Capsule to be maintained in location to EOL.

(e) = Standby capsule to be removed at 1-2 times the vessel EOL fluence.

(f) = Capsule was evaluated for dosimetry and placed in storage.

(g) = Capsule to be removed at projected 80-year peak RPV fluence.

(h) = Standby EOL capsule, as needed.

(i) = At current lead factor, capsule reaches -8.4x1019 n/cm2 at-42 EFPY of vessel operation

(equivalent to Surry Unit 1 projected 80-year vessel fluence). Movement to a capsule location with

higher lead factor reduces time of vessel operation to reach this fluence.

4. Only dosimetry was evaluated.

5. MIRVP HUPCAP, not a plant-specific capsule.

AREVA Inc.


Table 1-6: Summary Status of the B&W Surveillance Capsules

Capsule ID

Capsule Contents

Weld Metal

Fracture Toughness Specimens

Status/ Location

Fluence, n/cm2 (x1019)

Target Expected/ Received

Time of Removal Comments

BAW-1543 Revision 4

Page 1-16

OC1-F Tested 0.057 Reported in BAW-1421, Rev. 18; fluence

corrected in BAW-14367

OC1-E X Tested 0.150 Reported in BAW-14367

----·- - ·--·----·--.. ·--·-.. -· ........ -....... ---·-·---....... -..................... - .. --, .... - .............. -............. ··-... --... --............ --.. -· ·-·-.. --·-··· .. ---·-·--......................... ,_ ... OC1-B Removed 0.700 [Note 1] Has been disposed66

······--.. -·············-·············· ···························--···· ··················-···· ..........•.....•.................. ································•················ ···························•···•·•·········· ································-····························· .... ··············································-·····-·-········-·-········-········-····l·-·····-·············-··························-····························-·····················--·-········-·-··············-·-········-··-··············-·····-·-·-·-········-·-·······1

OC1-A X Tested 0.895 Reported in BAW-18375

OC1-C X Tested 0.986 .................. _,_, ........... -... , ..... ·-··-.............................. . ............................................................................................. - ....... ,_ .. ................................................. . .................................... _ .... , .............. .. OC1-D CR3-ZY N/A Not Planned

OC2-C Tested 0.102 [Note 3] -----·-·-- ·-

X

X . ···--··-·--·---.. - .. --............................ . ................. -........ _, __ , ............ --.. ···--··-·-·····-·····-·-.......... _ ....... ·-·-·-· .... -·-·--··-··-·-.............. -............. . ......... - ............ -............ -.... -·-·-··-·-·--··-·-·-··-·-··

OC2-A Tested 0.337

OC2-B Removed 0.562 [Note 4]

Reported in BAW-20506


.............................. -... -............................ ,_ .. _ .. ,_,_ ... , .................. ·--·--·-.. -·-·- ····-·-·-···-· ............................ ---·-.. -·-·-·-·--Reported in BAW-16999

---·----- -.. - ............................. --· .... --·----·-.................. _ ...................................... ,---·-.... -,,-........... _ .. _,, ____ .... _,, ___ ................. -·-·-·-·-.. -.... ·----.... -,-.... , ...... -.............. --·-.. -· ... -................................................. ,_ ..................... .. Has been disposed66

Reported in BAW-2051 11 OC2-E X Tested 1.210

OC2-D Removed 0.803 [Note 2] Has been disposed66

·-----·----- ·-·--.. -·--.. --.. -·-· ....... - ........ --........ _ .. __ , .... -.............................................. _ ................................. -.... -..... _ .. ___ ,, __ ... ,_ ...................... _. __ ... __ ..................... -.......... _ .. ,--.. -·--··--·-·-.. --·-·-................................... ---·--·-· .................. ·--·-.. -···--.. --, ..... , .. ,_ .................... ,_ .............................. -._ .. ___ .... _,_,_,_, ___ ,_ .. __ ,._, _____ ··---.. --OC2-F Removed 0.803 [Note 2] Has been disposed66

OC3-A X Tested 0.081 [Note 3] Reported in BAW-143813

-·-·--·-·-· _ ....... ,-·-·-·-·-·-·-· ... -............. ,---·--.. ·--· .... -... --.......... ·-.. --,-......... _ ..... _ .. __ .. , .................... -.... -,--.. --.............. , __ ................... _ .. _, .................. ,-.................. -.................. ,-·-·-·-"--·--.. ---···-·--·· .... -.................... _. __ .. _,_

OC3-B X Tested 0.312 Reported in BAW-169?14

OC3-C X Removed 0.783 [Note 2] Has been disposed66

·-·--·--······-·-............. _.,_ ·-·-.... ,-.... ----·-· _, ............ - .. - ..................................... _ ................ _ .. ___ ,_, ...... -......... ... ···-.. -·-·-·-·-.. --·-.. ·-· ............................. , ... -....................... , ___ ..................................... -,_ ....... - ..... , __ ..... , ... -.... --·---··-·-·--· .. --·-·-··-·-·-····· .. ····-·-·-...... ,_, ___ ,,,_ .. ___ ,, ______ ,_,_,,,., ___ ,_,,_,. .... ,-.. .

OC3-D X Tested 1.45 Reported in BAW-2128, Rev. 115

-·-·-·-·-·-"·"-·-·- ................ -...... - ....................... _,_ .. _, __ ... , .... _ ...................................................... -.............. _,,_ .......................................................... --·-·---·--........................................................ ,_ .. _ .. __ ,_ .. __ .... _, .. , ....... -·-·-· .. ·-·-·····-· ............................ -·-·-·--.. --.. - .... -, .......... _._ ..... _. __ .... , ........ .. OC3-E X Removed 1.262 [Note 2] Has been disposed66

OC3-F X CR3-ZY N/A Not Planned

AREVA Inc.


BAW-1543 Revision 4

Page 1-17

Table 1-6 (cont.): Summary Status of the B&W Surveillance Capsules·

Capsule ID

TMl1-E

TMl1-B

TMl1-C

TMl1-A

TMl1-D

TMl1-F

AN1-E

Capsule Contents

Weld Metal

X

X

X


Status/ Location

Tested

Removed

Tested

Removed

Removed

Removed

X Tested



0.107

0.444 [Note 4]

0.882 [Note 5]

0.158

0.816 [Note 2]

0.631 [Note 2]

0.0727


_Reported in BAW-143918

Has been disposed66

Reported in BAW-1901 17

Held in storage - reported in BAW-204i6

Has been disposed66

Has been disposed66


AN1-B Tested 0.428 Reported in BAW-169820

-·--·---·--·-···· -··--·-·-·-·-··--·-···· -·--·-··--·-··-·······--··-·--- ··---·-··········-·-·-··-··· ·····--··-······························ ·········-···-··-··-··--·-···--·-·-······· ············-·-·--·-··-···-···-··-·-··-·····t···-····-·················-···············-···-··-·--··--·--······-·-·····-c~--·---·-·----------1 AN1-A X Tested 1.03 Reported in BAW-1 '19

AN1-C X Tested 1.46 Reported in BAW-2075, Rev. 121

........................ - ................................................................................................................................... -, ........................................... " ............................................................................................................................ ..

AN1-D Removed 0.760 [Note 4] Has been disposed66

-----·------· , .. ____ ,. ____ ,, ........... , ... _._,,. ··-···-·-·····-·····-·····-············-··- .. , .... _ .. ___________ .. _____ -----------.. --·--··-··-··-.. .,_ .. ,_ .. ___________________________ .. _ ...... ____ .. _______________ , ............................... ···········-········--·-········-·-·····-----------------

AN1F Removed 0.783 [Note 2] Has been disposed66

TE1-F X X Tested 0.196 Reported in BAW-1701 29 and BAW-171930

····--·---····---·-··-··--------·--- ·------------··---·----------· ·--------------------------------------------- --------------------·---------- .. ·----------------·-········-·-···---·-··· . . ..................................... __________ . ·--------------------------------------------------------------··------ ·------·-···········-----····---·--········•·-·-····-··-·-------------------····---------------------------·----------------------· TE1-B X X Tested 0.592 Reported in BAW-183424 and BAW-186725

--------------· ·-------------------- --------------·----------- ---------------·-·· ---------------------------------------- -·-----··-·--·---·-··-·-·-·-·-·----- ··------------------------------------------- .. -----------------------------------·-·-··----------------------- -------·----------·-··--------·--··------------------.. ---.. ,---------------.. -----·----------.. -· .. -·----·--·------·---·-·----·-------TE1-A X

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

TE1-D X X

TE1-C X

TE1-E X

Tested

Tested

Tested

Removed

1.29 Reported in BAW-1882, Rev. 123

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

0.962

1.88

1.267 [Note 2]

Reported in BAW-212527 and BAW-220828

Reported in ANP-333926

Has been disposed66

AREVA Inc.

Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7 ·

Table 1-6 (cont.): Summary Status of the B&W Surveillance Capsules

Capsule ID

Capsule Contents

Weld Metal

Fracture Status/ Toughness Location Specimens




CR3-LG1 X X Tested 0.500-0.779 Reported in BAW-1910P67

BAW-1543 Revision 4

Page 1-18

..... -... , .................. -................................................................................................................................................. ,-.................................... ......................................................................................................................... .. ............................................................................................................................................................................................................................................................ ..

CR3-LG2 X X Tested 1.19-1.95 Reported in BAW-2254P68

·-·-·-·-·-·-·--···-·-·-· ............ ·-·---·····-.. ···-··---·-""' ...... _, ... , .................................................................. ---·-·-·-.... -,,---·- ... - .. -·-·· .. -, .. , .... -.............. ,_ ........................ -.... -.... - .......... -......................................... -.................... -....... -........................ -.................... _ ................................................................... , ... -... -........................ - ...... ___ ................ -..... - ..... -·-·-.. -·--.. --·-· D81-LG1 X X Tested 0.661-1.03 Reported in BAW-1920P69

---·--------.. ··--·--·--·- ··--.. -·-·-.. ----.. --.. -· .......... _ .. __ ... _ .. _ ...... -............. , .... _ ....................... -·--·-·-·-..... -·---·-· ..................................... -......... .. __ .......... _._ .. _ ................... _._.,_ ....... ···-·-.. -··--···-·-........ _ ................. --.-· .. ·--·--·-···"·--· .............. , .... -............... -.......... --........................................................ --.. --·-·-·-...................... -.--.. -·--··--·--·-·-·-·-·-DB 1-L G2 X X Tested 1.10-1.65 Reported in BAW-248670

........................................................................................................................................................................... ............................................................ . .. ............................... .. .......................................................................................................... -............ . ............................................................................................................................................................................. ,-................................. ..

TMl2-LG1 X X Tested 0.585-0.992 Reported in BAW-2253P71

........ --.... ·--·--..... -, .... --·-·- .................... --.. -·-·-·- ..... , ... _ .. __ .. _ .. _ ... , ... - ............. ,-.............................................. ---·-· ......... , ... -.................. , .... -............. -·--... - ........ -.... , ................. -,................ .. .. -................. ___ .. __ ............................... --.. --............ _ TMl2-LG2 X X Tested 1.17-2.01 Reported in BAW-243972

-------.. --· .. ---.. --·-·-· ··--·-......................................... -.. --.. -- ·--.. - .. --··-·-·-·-· .... ---· ·-·---·---·-·-.... _,_ ..... - ................................................................................................................................... ,-............ _ . .,_,_ ............................. -........... -.... - .......... ,--.--.. ---·-·-·-·-.. -·--··-··-.... -... -.... ·-·-··-·-··---·-···-·-·-....... _ ...... . A1 X X D81-YZ 5.1 5.1 Standby

A2 X X CR3-YX N/A Not Planned ·--·-·-·------·-- ·----............................ .

A3 X X Tested 0.881-1.450 ...................................................... _ ... _ ..... _ ..................................... +-.. --.......................................................... _ ...................................................................... _ ............................. __ ........ - ....... - .... -._·I

Reported in BAW-241 '73

c-----·-··-•-•H"HOH .. _••"-•"••-••-•,.••- , .,,,,.fi.HOHHH> .. H-HOHHHHH,.,_>H-H>H•HHH-" ·--.. ---·--·-·- -H-••--•HoH•H•"••-••.,•-•H ••"'""'"" ... •"OH,HHHOHHHH-•--·----H •-•H>HH-·-·-N-HHHHHHM,HHH-•--·-H-HH>H_O_ mH,H>H ______ .. ,_,_H,Hm,H>-,OH--HH>H•H-->HH_O_>_M ____ ,H,H>H•-•H->HH_H_O_MO-->M __ _

A4 X X CR3-YX N/A Not Planned ................................ _ .. _, __ , ................................................................................................................................................ ,-... , .. , ... , .. --·-··- ............ ,-................................................................ ...................................... ..................................................................................................... ..................................................................................................................................................... . ................................................. .

A5 X X Tested 0.637-1.042 Reported in BAW-2360P74

--·---··----.. · .. --·-·-·-.. -·--··--...... __ , ............................ -...................................................... --..... -, .... _ .. ___ ·--·--·-............................... , .. --........................................... , .... -...... .. . .......................... -........................................................................................................................ . ................................................. .

L1 X X Tested 1.169-1.624 Reported in BAW-240075

-·---·--·-·-·--·---·- ··-.. -·--·-..................... .. .. ·--·-·-........ -................................. . ................. -...... ,_, ......... ---·-· ......... ,_ .. _, __ ,,_ .............................. -........................................................ . ............................ -............................................................. _._ .... , ... -, ........................................ _ ........ -, .. ·-·-·"·-·-·- ...................... -..... -............................................ --·-.. ---·-·-.... ·-·-·--L2 X x D81-YZ 4.5 4.5 Standby


1. BAW-1543, Revision 376


3. NUREG CR-4816, Volumes 1 & 278

4. BAW-210879

5. Recalculated in 2003.

AREVA Inc.


Table 1-7: Summary Status of the Westinghouse Surveillance Capsules

Capsule Contents Capsule

ID Weld Metal

PB1-N X


Status/ Location [Note 1]

33°



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

PB1-P X Removed


27.2-51.5 EFPY Standby [Note 2]

Held in Storage

BAW-1543 Revision 4

Page 1-19

---·-·--·--·--·-·----·- ......................................... ................... ........................................... . .. , .... _ .. ___ ,.., ____ , ..... , _______ ··-··--·-···""'''"'""'''" . ................................................................. .. ...................................................... -................................................ ·····-····-·---.. ---·-··'"·-·-·--········-------.. ------··-··---·-·-----------.. ------------------.. ---------------------------.. ·-----PB1-R X X Tested Reported in WCAP-935732

Reported in WCAP-873933 PB1-S

PB1-T

PB1-V

X

X

X

Tested

Tested

Tested

· ···································-·····························t···········-············································································+······································································-·············································c···,·····································································I Reported in WCAP-1073634 X

X ··············---·-------------.. ------·-·· ............. ______ , .......... _____________ ., ______ ....... 1·····--··-····························--··---·-······--···············-1·····-·--········-·-·-····-·-·-·-·-·····-·-·---·--····-·-··-·····--·······-···-·-·-··--··-·-··-·--·-·······-··-·---··········-··-···-··-·--··-··--1

Suppl. X

PB2-N X __________ ,. _________________________ -----------·-·

PB2-P X

X 13°

33°

Removed

5.000 5.000 43 EFPY

27.2-51.5 EFPY

Reported in BCL report dated 6/7335

Supplemental Capsule

Standby

Held in Storage

PB2-R X X Tested Reported in WCAP-963538

__ ,, ____ ,, __ ., ______ ., ________ , ....................... -... ,......................... . ....................... , .... _, __ ,, ___ , . ____ ,,,, ...... -.......................................... ,-.................................................. '""'""'"'"'""""""'''"'""""""'"""

PB2-S X X Tested Reported in BAW-214039 ·-·-·-·--------------·-- ......... _ ........... ,_ .. _ .................................... _ .. ---------------·-·--·--------· ,. ______ .,_, .. ,,... .. .......................... , .... _____ ,_, ......... _, ____ ,,____ __ ___________ .... ____ ,, ________ ,, ___ ,, ______ ,, ____ ,_ ............................................................ _ .......................... ,_, ....................... _, ______ ,. _________ ,, ________________________ ,,,. ______ _

PB2-T X Tested Reported in WCAP-9331 40

--------·-·-·------ ·--------------------.. _________ ,, __ ,.,_,_,, ___ ,,,,.......................... ·-·----.. ---··-·-·-·· .. --------·- ______________ , .. , ......... , .. ,_, ___ ,. ______ ... _______ ,, ______ ,, ________ ............................ .

PB2-V X X Tested Reported in BCL report dated 6/7541

REG-N X X Tested Reported in WCAP-17036 Rev. 143

,, ______ ,,_, __ ,, ___ .. -, ... _ ...................................... -..................................................... _ ..................................................... ,-....... , .. , _____ ,_ ........................................................ ,-.............................................................. , .. ,-................................................................ _,, _______________ ,, __ ,, ... , .... ,-. ··-·-·-·""'"""'""""'""' ___ ,, __ ............................... -........... _,_,,_,,, ..................................... , _______ .... , .... , ____ ,, ____ ,, _________ ,,_

REG-P X X 23° 33.9-39.9 EFPY Standby -----------·-·-·------·-·-·---- ____________________ ,. ________ ·--------------- ................... ,_, .............................. ______ .......... ____________ _ REG-R X X Tested Reported in WCAP-8421 44

................................................................................................................................................................................... -.............................. _,, ................................................................................................................................................................................................................................................................................................................................................................................................ ___ _

REG-S X X Tested Reported in WCAP-1390245

··--·--·----------------- ______ ,.____________ . ·-------·----------------------------------· ·----------·--.. -,.......................... .. ............. , ____________________ ,, __________ ·-----------·-·--·------·-·-.. --------.. --..

REG-T X X Tested Reported in WCAP-1008646

REG-V X X Tested Reported in FR-RA-1 47

AREVA Inc.


BAW-1543 Revision 4

Page 1-20

Table 1-7 (cont.): Summary Status of the Westinghouse Surveillance Capsules

Capsule ID

S1-S

S1-T

S1-U

S1-V

Capsule Contents

Weld Metal

X

X


X

X


25°

Tested

45°125°

Tested



Time of Removal

End of License

Comments

Standby

Reported in BCL report dated 6/7549

....................................................................................................................................................................................................................... , .. _ ................................................................................................................... . End of License Standby, transferred to 25° at EOC-12

Reported in WCAP-1141550

!--·--· ..... -..................... . ...................................... - ........................................................................... - ........................ _,, .. _, ................. , ___ ......................................................................... ,, ____ , .. , .................. .

S1-W Dosimetry Evaluated for dosimetry only. Results reported in BCL-585-8R51

s=t=x - · X X Tested ----- ~· ----- ----- Reported in BAW-232452 .

·····-·-·-···-··-·-·--.... ·--·--.... ·--·-·- .............................................................................................................. ___ , ____ .. , ......... --.. --.-- ................................................... .. ........................................................... .. .................................... , ____________ , ............. , .. ,_..... ........................................ . ......................................................................................................................... '"'-·"··-·····-·····"

S1-Y 35°115° End of License Standby, transferred to 15° at EOC-14

S1-Z X X 25°115° 6.31

S2-S X Dosimetry

S2-T X 35°115°

6.31 43.2 EFPY Standby, transferred to 15° at EOC-12

Evaluated for dosimetry oni. Results reported in WCAP-1481 O 4

......................................... , ... - ........ , ... -.... , .. ,_,_ ................... _ ............................................... ,_ .... ,-------·--·-.. ----·---·-·-·-·-·-·-.. ···-·-··-···

End of License Standby, transferred to 15° at EOC-17 ._ ...................... -.......... -............................................... -......... ............................................................... . ................ -....................... --.... --. ........................................................... ..... .. ...................................................................................................................................................................................................................................................................................................................... -............................................................... ..

S2-U X 25°/15° 5.95 5.95 45.0 EFPY Standby, transferred to 15° in Year 2009 -------·-··-···"······-··-···---··· -·----··-··-····-········-·····-···· ·-··-·--·--·-·--·---·-········ ·--·-··-·······-··-··--··-·-··-·-·- · ··········-·-·-·--·--·-·-·--·--·-··-··-· -·-···-·- ·····-···-·····-·--·-----·-··-·-··-··--·--+-·-·-·-·-··--·-···-·-·-···········-············--··--·-·-·······-·-··----·-··-,.-c--··--····----···············-·-··--·-·--1

Reported in WCAP-11 1955 S2-V X Tested

-~-?:_'!!_ __ ___ _!. __ . __ .. --·---·-·-·--·- ·-·Dosimetry . --·--·-·· .. ···········-·-··-······- _ -··-·-·-··-·······-·-·······- ····--·-·- ··-·-··-·--··--··--·-····-·····-·-······--·-----··-•·--R. ___ e __ po __ rt .. _e .. _d .. _._i._n·-·-··B·······C··-·-L·····-···5··-··8_ .. 5 .. _-. __ o_.2._6_5

_

6

--·-·--···········-······-····---·--,

S2-X X

S2-Y X

S2-Z X .. --·-·-·-·-...................................................... - ......... ..

S2-W1 X

X

X

X

Tested Reported in BCL report dated ·557

Tested

35°125°

Tested 0.602-0.802

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

Reported in WCAP-16001 58

End of License Standby, transferred to 25° at EOC-12 ...... ............................. ................................................ .. ................................... -···-·-·-................................... .

Reported in BAW-2350P80

AREVA Inc.


BAW-1543 Revision 4

Page 1-21


Capsule ID

TP3-S

Capsule Contents

Weld Metal



Tested

TP3-T X X Tested



·-·-·-----.. --,-------·- --.... -, ...... ,_ .. ____ .. _,i--,---·-·-·-·-·-·--· ,_ . .,, ..................................... __________ ,, __ ,,_____ --·-··-· .............................................. , ... -....................... , .. , __ ,,_ .... . TP3-U 30°

TP3-V X X Tested

TP3-W 40°


Reported in SwRl-02-5131 60 ................................................................................. -..................................... ,, ______ .. ,, _____ ., __ ., __________ _ Reported in WCAP-8631 61

. ..................................................... ,_, ______ ..... _, __ ., ___ , ....... -, .......... , .... - .................... , .... ___ ., ___ .. _ .. ,_, ___ .. _______ ... , ..................... --.. ------·-·-·· .. -·-···-·-·-·--.... -....... _, __

End of License Standby

Reported in SwRl-06-857562 -

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

End of License Standby -·--·-···-·····-·· .. ·-·-····· .............. . ........ ·····································-··········-····· ·································-····· ········-·-······································ .... ·····························································1···--·--·-··································-··········-·······-·············+············································-···········-···································································-········-·······--·-·····--·····-1

TP3-X X X Tested Reported in WCAP-1 I 663

······-·---------------1----.. --- -----------·-""'''''···-·····-·-·-·-··- __ .. ____ ,. __________ _ ·······•········-········-·-··--··- .. _ .... ______ .. ____________________ ····-··-········-········-·-········-·-··· .. -·-········-·--·····-···-··--·······-· ................. ,----·-·--------------------... ---------··-·--·---····-·-·-TP3-Y 30° End of License Standby

TP3-Z 40° End of License Standby

TP4-S Tested Reported in SwRl-02-538060

TP4-T X X Tested Reported in SwRl-02-4221 65

'""""''"""'"'""'"""'"""''"'"'"''''"'''''•-"••• •••H-HH•--•--•- ,mH-HH_n_, __ ,_, .. ,, .... ,,.,,,_,,,,,,,, .. ,,,,,, •H•-••••••••--•-·•--•-H-m•-- ,n--•-n-•-•-m--•m•--••••·-•- • ••••••••••••••••••-•-••··-••--•-•-••••-m•m•--•----H n-•---•-•--•-n-m•--•H--•-•-•-·---•-•••••-•--•••••-· •••HH•-·-••••••••-••••-••••••••-·-•••••••m-•-•••h••••••--H-••--•--n-•-•--H-HH--•---•-•-•--•--.. -H-•-•-•-•-·-••••••••-•-•-••••••••-

TP4-U 30° End of License Standby -----·-·-·-·- -----.. ----·-··- ··---.. -·-···-··-· .... -, .. -..... _. ____ .. ___ ... -·-.. -· ... -----·--····-·---·-· ···-····-·--··-····-.......................... _,_ .. _ ·-.. - .. -·-·-·-·· ............ -.................................. ·····-·--·--·-·-·-·-········-·-·· ... -·-·-........ -.. _ .. ____________ .. __ .... ____ .. _ .... _. __ .................. _ .. ___ .......... ,_ ........... -·-·-.. ···--·-·-······-·-·-·-·· .. -·-· ... ·--·-·--.. -·-·--·-·-·--·-·-·-TP4-V X X 20° End of License Standby

---·--·----·-.. - ---·-.. --·-·-·-·-.. -· ...... ·-·-·-·-····-··-····-········-·····-·-····-··-·· ·-·-.. ··-··---··-·-·-·-.... ·---- ... -.. -·-·-.. -................................... -......................................... -.............. _______ ..... ·-·-.. -·--·-·-.. ··-·-.................. ---····· ········-·-········-··· ··-·-·-·-····· .. ·-·--··· ..... -.............. __ ........... _ .. _____ .... ·-·-· .... ·-·-·-·-·-.. -·-·-.. ··-.. -·-·······--·-·····-·-·-·-TP4-W 40° End of License Standby

....................................................................................... -, .......................................................................................................... _, ... , ........... -, ......................... -........................................................................................................................................................ , ... -......................................................... -.................................................................................................................................................................................... .

TP4-X X X 40°/0° 9.297

TP4-Y

TP4-Z


30°

40°

9.297 38.1 EFPY

End of License

End of License

1. All locations are relative with regard to quadrant; e.g.,. 0° is equivalent to 90°, 180°, or 270°.

Standby

Standby

2. At current lead factor, capsule reaches -8.4x1019 n/cm2 at -42 EFPY of vessel operation (equivalent to Surry Unit 1 projected 80-year vessel fluence). Movement to a capsule location with higher lead factor reduces time of vessel operation to reach this fluence.

AREVA Inc.


BAW-1543 Revision 4

Page 1-22

Table 1-8: Comparison of the Plant-Specific Surveillance Capsules with ASTM E 185-82 Requirements

ASTM E 185-82 Five (5) Capsule Program Requirements

1.5 EFPY 3 EFPY

Plant or 6 EFPY 15 EFPY EOL or Fluence Midway or or or

Fluence > 5E18 ART NOT :::: 50°F

Between First and T/4 EOL Fluence IS EOL Fluence 1-2 Times EOL Fluence Third Capsule

Oconee Unit 1 F - (1/T) [Note 1] E - (1/T) A- (1/T) CR3-LG2 - (1/T)

Oconee Unit 2 C - (1/T) A- (1/T) TMl2-LG1 - (1/T) E - (1/T) A5 - (1/T)

Oconee Unit 3 A- (1/T) B - (1/T) L 1 - (1/T) D - (1/T) CR3-LG2 - (1/T)

TMI Unit 1 E - (1/T) W1 - (1/T) CR3-LG1 - (1/T) C - (1/T) A5- (1/T)

ANO Unit 1 E - (1/T) [Note 1] B - (1/T) A- (1/T) C - (1/T)

Davis-Besse F - (1/T) [Note 1] B - (1/T) D - (1/T) A - (1/T); C - (1/T) [Note 3]

Point Beach Unit 1 V- (1/T) S - (1/T) R- (1/T) T - (1/T) N - (IR)

Point Beach Unit 2 V- (1/T) T - (1/T) R- (1/T) S - (1/T) N - (IR)

R. E. Ginna V- (1/T) R - (1/T) T - (1/T) S - (1/T) N - (1/T); P - (JR)

Surry Unit 1 T - (1/T) W - (1/T) [Note 2] V- (1/T) X- (1/T) Z- (IR)

Surry Unit 2 X- (1/T) W - (1/T) [Note 2] V- (1/T) y - (1/T) U - (IR)

Turkey Point Unit 3 T - (1/T) S - (1/T) V- (1/T) X- (1/T) TP4-X- (IR)

Turkey Point Unit 4 T - (1/T) S - (1/T) TP3-V - (1/T) TP3-X - (1/T) X-(IR)

Legend: (1/T) = Irradiated and tested, (IR) = Currently in reactor vessel


1. Only 4 capsules required per ASTM E185-82.

2. Only dosimetry evaluated.

3. Capsule TE1-C was tested to support license renewal for Davis-Besse.

AREVA Inc.


BAW-1543 Revision 4

Page 1-23

Table 1-9: Peak End-of-Life Inside Surface Fluences and Significant Licensing Dates

Max Projected EOL Date Construction Date Operating License Inside Surface (IS)

Plant Permit Issued License Issued Expiration Fluence, n/cm2

(E > 1 MeV)

Oconee Unit 1 November 6, 1967 February 6, 1973 February 6, 2033 [Note 1] 1.29E+19

Oconee Unit 2 November 6, 1967 October 6, 1973 October 6, 2033 [Note 1] 1.44E+19

Oconee Unit 3 November 6, 1967 July 19, 1974 July 19, 2034 [Note 1] 1.38E+19

TMI Unit 1 May 18, 1968 April 19, 1974 April 19, 2034 [Note 1] 1.52E+19

ANO Unit 1 December 6, 1968 May 21, 1974 May 20, 2034 [Note 1] 1.35E+19

Davis-Besse March 24, 1971 April 22, 1977 April 22, 2037 [Note 1] 1.70E+19

Point Beach Unit 1 July 19, 1967 October 5, 1970 October 5, 2030 [Note 1] 5.09E+19

Point Beach Unit 2 July 25, 1968 March 8, 1973 March 8, 2033 [Note 1] 5.07E+19

R. E. Ginna April 25, 1966 September 19, 1969 September 18, 2029 [Note 1] 5.45E+19

Surry Unit 1 June 25, 1968 May 25, 1972 May 25, 2032 [Note 1] 5.66E+19

Surry Unit 2 June 25, 1968 January 29, 1973 January 29, 2033 [Note 1] 5.38E+19

Turkey Point Unit 3 April 27, 1967 July 19, 1972 July 19, 2032 [Note 1] 6.38E+19

Turkey Point Unit 4 April 27, 1967 April 10, 1973 April 10, 2033 [Note 1] 6.38E+19


1. Includes 20 year license renewal term.

AREVA Inc.


2.0 REFERENCES

BAW-1543 Revision 4

Page 2-1

1. B&W Nuclear Technologies, Inc. Document, BAW-1543, Revision 4 (43-

1543-04), "Master Integrated Reactor Vessel Surveillance Program,"

February 1993. (Available through AREVA Inc.)

2. Materials Reliability Program: Coordinated PWR Reactor Vessel

Surveillance Program (CRVSP) Guidelines (MRP-326). EPRI, Palo Alto,

CA: 2011: 1022871. (NRCAccession Nos. ML12040A314 &

ML 12040A315)

3. ASTM Standard E 185-82, "Standard Practice for Conducting Surveillance

Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American

Society for Testing and Materials, Philadelphia, Pennsylvania.

4. Babcock & Wilcox Document, BAW-10006A, Revision 3, "Reactor Vessel

Material Surveillance Program," January 1975. (Available through AREVA

Inc.)

5. Babcock & Wilcox Document, BAW-1837 (77-1173973-00), "Analysis of

Capsule OCl-:A Duke Power Company Oconee Nuclear Station Unit 1

Reactor Vessel Materials Surveillance Program," August 1984. (Available

through AREVA Inc.)


Capsule OCI-C Duke Power Company Oconee Nuclear Station Unit 1

· Reactor Vessel Materials Surveillance Program," October 1988. (Available

through AREVA Inc.)

7. Babcock & Wilcox Document, BAW-1436, "Analysis of Capsule OCI-E

Duke Power Company Oconee Unit 1 Reactor Vessel Materials

Surveillance Program," September 1977. (Available through AREVA Inc.)

8. Babcock & Wilcox Document, BAW-1421, Revision 1 "Analysis of Capsule

OCI-F Duke Power Company Oconee Unit 1 Reactor Vessel Materials

Surveillance Program," September 1975. (Available through AREVA Inc.)

-- --- ----- --~

AREVA Inc.


BAW-1543 Revision 4

Page 2-2

.9. Babcock & Wilcox Document, BAW-1699 (77-1130549-00), "Analysis of

Capsule OC/1-A Duke Power Company Oconee Nuclear Station, Unit 2

Reactor. Vessel Materials Surveillance Program," December 1981.

(Available through AREVA Inc.)

10. Babcock & Wilcox Document, BAW-1437, "Analysis of Capsule OC/1-C

Duke Power Company Oconee Nuclear Station, Unit 2 Reactor Vessel

Materials Surveillance Program," May 1977. (Available through AREVA

Inc.)

11 . Babcock & Wilcox Document, BAW-2051 (77-1173071-00), "Analysis of

Capsule OC/1-E Duke Power Company Oconee Nuclear Station Unit 2

Reactor Vessel Materials Surveillance Program," October 1988. (Available

through AREVA Inc.)

12. Babcock & Wilcox Document, BAW-10100A (43-10100A-OO) "Compliance ' '

with 10CFR50, Appendix H, for Oconee Class Reactors," February 1975.


13. Babcock & Wilcox Document, BAW-1438, "Analysis of Capsule OCIII-A

Duke Power Company Oconee Nuclear Station, Unit 3 Reactor Vessel

Materials Surveillance Program," July 1977. (Available through AREVA

Inc.)


Capsule OC/11-B Duke Power Company Oconee Nuclear Station, Unit 3

Reactor Vessel Materials Surveillance Program," October 1981. (Available

through AREVA Inc.)

15. B&W Nuclear Service Company Document, BAW-2128, Revision 1 (77-

2128-01), "Analysis of Capsule OCI 11-0 Duke Power Company Oconee

Nuclear Station Unit 3 Reactor Vessel Materials Surveillance Program,"

May 1992. (Available through AREVA Inc.)

AREVA Inc.


BAW-1543 Revision 4

Page 2-3

16. Babcock & Wilcox Document, BAW-2042 (77-1172064-00), "Requalifi

cation of the TMl-2 Reactor Vessel Surveillance Program Capsules,"

August 1988. (Available through AREVA Inc.)


Capsule TM/1-C GPU Nuclear Three Mile Island Nuclear Station Unit 1

Reactor Vessel Materials Surveillance Program," Mar:ch 1986. (Available

through AREVA Inc.)

18. Babcock & Wilcox Document, BAW-1439, "Analysis of Capsule TMl-1E

from Metropolitan Edison Company Three Mile Island Nuclear Station Unit

1 Reactor Vessel Materials Surveillance Program," January 1977.



Capsule ANI-A Arkansas Power & Light Company Arkansas Nuclear One,

Unit 1 Reactor Vessel Materials Surveillance Program," July 1984:


20. · Babcock & Wilcox Document, BAW-1698 (77-1130526-00), "Analysis of

Capsule ANI-B Arkansas Power & Light Company's Arkansas Nuclear

One, Unit 1 Reactor Vessel Materials Surveillance Program," November

1981. (Available through AREVA Inc.)

21. Babcock & Wilcox Document, BAW-2075, Revision 1 (77-2075-01),

"Analysis of Capsule ANI-C Arkansas Power & Light Company Arkansas

Nuclear One, Unit 1 Reactor Vessel Materials Surveillance Program,"

October 1989. (Available through AREVA Inc.)

22. Babcock & Wilcox Document, BAW-1440, "Analysis of Capsule ANI-E

Arkansas Power & Light Company Arkansas Nuclear One, Unit 1 Reactor

Vessel Materials Surveillance Program," April 1977. (Available through

AREVA Inc.)

AREVA Inc.


BAW-1543 Revision 4

Page 2-4


"Analysis of Capsule TE1-A The Toledo Edison Company Davis-Besse

Nuclear Power Station Unit 1 Reactor Vessel Materials Surveillance

Program," June 1989. (Available through AREVA Inc.)


Capsule TE1-B The Toledo Edison Company Davis-Besse Nuclear Power

Station Unit 1 Reactor Vessel Materials Surveillance Program," May 1984.


25. Babcock & Wilcox Document, BAW-1867 (77-1156433-00), "Fracture

Toughness Test Results from Capsule TE1-B The Toledo Edison

Company Davis-Besse Nuclear Power Station Unit 1 Reactor Vessel

Materials Surveillance Program," June 1985. (Available through AREVA

Inc.)

26. AREVA Inc. Document, ANP-3339 (103-3339-000), "Davis-Besse Unit 1

Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-

C," December 2014.

27. B&W Nuclear Service Company Document, BAW-2125 (77-2125-00),

"Analysis of Capsule TE1-D The Toledo Edison Company Davis-Besse

Nuclear Power Station Unit 1 Reactor Vessel Materials Surveillance

Program," December 1990. (Available through AREVA Inc.)

28. B&W Nuclear Service Company Document, BAW-2208, "Fracture

Toughness Test Results from Capsule TE1-D The Toledo Edison


Materials Surveillance Program," October 1993. (Available through

AREVA Inc.)


Capsule TE1-F The Toledo Edison Company Davis-Besse Nuclear Power

Station Unit 1 Reactor Vessel Materials Surveillance Program," January


AREVA Inc.


BAW-1543 Revision 4

Page 2-5

30. Babcock & Wilcox Document, BAW-1719 (77-1132385-00), "Fracture

Toughness Test Results from Capsule TE1-F The Toledo Edison


Materials Surveillance Program," March 1982. (Available through AREVA

Inc.)

31. Westinghouse Electric Corporation Report, WCAP-7513, "Wisconsin

Michigan Power Company Point Beach Unit No. 1 Reactor Vessel

Radiation Surveillance Program," June 1970.

32. Westinghouse Electric Corporation Report, WCAP-9357, "Analysis of

Capsule R from the Wisconsin Electric Power Company Point Beach Unit

No. 1 Reactor Vessel Radiation Surveillance Program," August 1978.


Capsule S from the Wisconsin Electric Power Company and Wisconsin


Radiation Surveillance Program," November· 1976.


Capsule T from the Wisconsin Electric Power Company and Wisconsin


Radiation Surveillance Program," December 1984.

35. Battelle Memorial Institute Report, "Point Beach Nuclear Plant Unit No. 1

Pressure Vessel Surveillance Program: Evaluation of Capsule \I, Final

Report to Wisconsin Electric Power Company," June 15, 1973.

36. Westinghouse Electric Corporation Report, WCAP-7712, "Wisconsin

Michigan Power Company and The Wisconsin Electric Power Company

Point Beach Unit No. 2, Reactor Vessel Radiation Surveillance Program,"

WCAP-7712, June 1971.

37. Westinghouse Electric Corporation Report, WCAP-15856, "Supplemental

Reactor Vessel Surveillance Capsule A for the Point Beach Units 1 and 2

AREVA Inc.


BAW-1543 Revision 4

Page 2-6

Reactor Vessel to be Installed in the Point Beach Unit 2 Reactor Vessel,"

May 2002.


Capsule R from the Wisconsin Electric Power Company Point Beach Unit

No. 2 Reactor Vessel Radiation Surveillance Program," December 1979.


"Analysis of Capsule S Wisconsin Electric Power Company Point Beach

Unit No. 2 Reactor Vessel Materials Surveillance Program," August 1991.



Capsule T from the Wisconsin Electric Power Company Point Beach Unit

No. 2 Reactor Vessel Radiation Surveillance Program," August 1978.

41. Battelle Columbus Laboratories Report, "Point Beach Nuclear Plant Unit

No. 2 Pressure Vessel Surveillance Program: Evaluation of Capsule V,

Final Report to Wisconsin Electric Power Company," June 10, 1975.

42. Westinghouse Electric Corporation Report, WCAP-7254, "Rochester Gas

and Electric Robert E. Ginna Unit No. 1 Reactor Vessel Radiation

Surveillance Program," May 1969.

43. Westinghouse Electric Corporation Report, WCAP-17036-NP, Revision 1,

"Analysis of Capsule N from the R. E. Ginna Reactor Vessel Radiation

Surveillance Program," September 2010.


Capsule R from the Rochester Gas and Electric Corporation R. E. Ginna

Unit 1 Reactor Vessel Radiation Surveillance Program," November 197 4.


Capsule S from the Rochester Gas and Electric Corporation R. E. Ginna

Reactor Vessel Radiation Surveillance Program," December 1993.

AREVA Inc.


BAW-1543 Revision 4

Page 2-7


Capsule T from the Rochester Gas and Electric Corporation R. E. Ginna

Nuclear Plant Reactor Vessel Radiation Surveillance Program," April

1982.

47. Westinghouse Electric Corporation Report, FP-RA-1, "Analysis of Capsule

V from the Rochester Gas and Electric R. E. Ginna Unit 1 Reactor Vessel

Radiation Surveillance Program," March 1973.

48. Westinghouse Electric Corporation Report, WCAP-7723, "Virginia Electric

and Power Company Surry Unit No. 1 Reactor Vessel Radiation

Surveillance Program," July 1971.

49. Battelle Columbus Laboratories Report, "Final Report on Surry Unit No. 1

Pressure Vessel Irradiation Capsule Program: Examination and Analysis

of Capsule T," June 24, 1975.


Capsule V from the Virginia Electric and Power Company Surry Unit 1

Reactor Vessel Radiation Surveillance Program," February 1987.

51. Battelle Columbus Laboratories Report, BCL-585-8R, "Final Report on

Surry Unit No. 1 Nuclear Plant Reactor Pressure Vessel Surveillance

Program:. Examination and Analysis of Capsule W to Virginia Electric and

Power Company," March 30, 1979.

52. Framatome Technologies Inc. Document, BAW-2324, (77-2324-00),

"Analysis of Capsule X Virginia Power Surry Unit No. 1 Reactor Vessel

Material Surveillance Program," April 1998. (Available through AREVA

Inc.)

53. Westinghouse Electric Corporation Report, WCAP-8085, "Virginia Electric

and Power Company Surry Unit No. 2 Reactor Vessel Radiation

Surveillance Program," June 1973.




Dosimetry from Capsule S from the Virginia Electric andPower Company

Surry Unit 2 Reactor Vessel Surveillance Program," December 1996.


Capsule V from the Virginia Electric and Power Company Surry Unit 2

Reactor Vessel Radiation Surveillance Program," June 1987.

56. Battelle Columbus Laboratories Report, BCL-585-026, "Final Report on

. Surry Unit No. 2 Nuclear Plant Reactor Pressure Vessel Surveillance

Program: Examination and Analysis of Capsule W to Virginia Electric and

Power Company," February 27, 1981.

57. Battelle Columbus Laboratories Report, "Final Report on Surry Unit No. 2

Pressure Vessel Irradiation Capsule Program: Examination and Analysis

of Capsule X to Virginia Electric and Power Company," September 2,

1975.


Capsule Y from Dominion Surry Unit 2 Reactor Vessel Radiation

Surveillance Program," February 2003. (NRC Accession Nos.

ML030990253 & ML030990293)

· 59. Westinghouse Eledric Corporation Report, WCAP-7656, "Florida Power

and Light Company Turkey Point Unit No. 3 Reactor Vessel Radiation


60. Southwest Research Institute Report, Final Report SwRI Project No. 02-

5131 SwRI Project No. 02-5380, "Reactor Vessel Material Surveillance

Program for Capsule S - Turkey Point Unit No. 3 Capsule S - Turkey Point

Unit No. 4," May 1979.


Capsule T from the Florida Power and Light Company Turkey Point Unit

No. 3 Reactor Vessel Radiation Surveillance Program," December 1975.

AREVA Inc.


BAW-1543 Revision 4

Page 2-9

62. Southwest Research Institute Report, Final Report SwRI Project No. 06-

8575, "Reactor Vessel Material Surveillance Program for Turkey Point Unit

No .. 3: Analysis of Capsule V," August 1986.


Capsule X from the Florida Power and Light Company Turkey Point Unit 3

Reactor Vessel Radiation Surveillance Program," September 2002. (NRG

Accession Nos. ML022940497 & ML022940133)

64. Westinghouse Electric Corporation Report, WCAP-7660, "Florida Power

and Light Company Turkey Point Unit No. 4 Reactor Vessel Radiation


65. Southwest Research Institute, Final Report SwRI Project No. 02-4221,

"Reactor Vessel Material Surveillance Program for Turkey Point Unit No. 4

Analysis of Capsule T," June 14, 1976.

66. The B&W Owners Group Letter OG-1783 from D. L. Howell, Project

Manager to Document Control Desk, U.S. Nuclear Regulatory

Commission on behalf of the B&W Owners Group Reactor Vessel

Working Group, Project No. 693, March 17, 2000. (NRG Accession No.

ML003693967)

67. Babcock & Wilcox Document, BAW-1910P (77-1164769-00), "Analysis of

Capsule CR3-LG1 Babcock & Wilcox Owners Group Integrated Reactor

Vessel Materials Surveillance Program," August 1986. (Available through

AREVA Inc.)

68. B&W Nuclear Technologies, Inc. Document, BAW-2254P (77-2254P-OO),

"Test Results of Capsule CR3-LG2 B&W Owners Group Master Integrated

. Reactor Vessel Surveillance Program," October 1995. (Available through

AREVA Inc.)

.69. Babcock & Wilcox Document, BAW-1920P (77-1164769-00), "Analysis of

Capsule DB1-LG1 Babcock & Wilcox Owners Group Integrated Reactor



Vessel Materials Surveillance Program," October 1986. (Available through

AREVA Inc.)

70. Framatome ANP, Inc. Document, BAW-2486 (43-2486-00), "Analysis of

the B&W Owners Group Capsule DB1-LG2," November 2005. (Available

through AREVA Inc.)

71. B&W Nuclear Technologies, Inc. Document, BAW-2253P (77-2253P-OO),

"Test Results of Capsule TMl2-LG1 B&W Owners Group Master

Integrated Reactor Vessel Surveillance Program," October 1995.



the B&W Owners Group Capsule TMl2-LG2," May 2003. (Available

through AREVA Inc.)


the B&W Owners Group Capsule A3 Master Integrated Reactor Vessel

Surveillance Program," April 2002. (Available through AREVA Inc.)

74. Framatome Technologies, Inc. Document, BAW-2360P (77-2360P-OO),

"Analysis of A5 Capsule, B&W Owners Group Master Integrated Reactor

Vessel Surveillance Program," June 1999. (Available through AREVA Inc.)


the B&W Owners Group Capsule L 1 Reactor Vessel Material Surveillance

Program," March 2002. (Available through AREVA Inc.)


"Master Integrated Reactor Vessel Surveillance Program," September


77. B&W Nuclear Service Company Document, BAW-2108, Revision 1 (77-

2108-01 ), "Fluence Tracking System," May 1992. (Available through

AREVA Inc.)

AREVA Inc.


78. NUREG/CR-4816, Volumes 1 and 2, "The Pressure Vessel Steel

Embrittlement Data Base (EDB)," June 1988.

BAW-1543 Revision 4

Page 2-11


"Fluence Tracking System," March 1990. (Available through AREVA Inc.)

80. Framatome Technologies, Inc. Document, BAW-2350P (77-2350P-OO),

"Test Results of W1 Capsule, B&W Owners Group Master Integrated

Reactor Vessel Surveillance Program," April 1999.

Original pages for BAW-1543 Revision 4 Supplement 7:

Original Page Number

ii through iii

vii

1-1 through 1-2

1-17 through 1-23

The original pages for BAW-1543 Revision 4 Supplement 7 follow.


Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7 Page ii

1

2

3

4

5

6

7

8

9

Section(s) or Page(s)

Table 1-1

Table 1-2

Table 1-3

Table 1-4

Table 1-5

Table 1-6

Nature of Changes

Description and Justification

• Updated document format to the current AREVA Inc. Licensing Document template.

• Updated to reflect current revision.

• Included revision statements for Supplement 7 changes.

• Deleted changes made in Supplement 6.

• dded descriptions of Supplement 7 changes.

• Del ed the Crystal River Unit 3 capsule detailed summary infer tion.

• Added in rmation on the TE1-C capsule for Davis-Besse.

. Ginna capsule detailed summary

• Updated the Westinghouse !ant-specific suNeillance program status based on the RVSP and their respective NRC approved plant-specific R SP withdrawal schedules.

• Added the R. E. Ginna capsule ins ion and withdrawal schedule.

• Updated the TE1-C capsule status.

• Deleted the Crystal River Unit 3 plant-speci capsule summary status information .

• Updated the time of removal of the supplemental apsules A2 and A4 to "Not Planned."

• Deleted Expected/Received fluence values for the OC -D and OC3-F capsules.

• Updated Expected/Received fluences forTMl1 -A, TE1-E, A3, and L 1 capsules.


Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7 Page iii

1 Table 1-7

11

12 Table 1-9

13 Section 2.0


• Added the- R. E. Ginna plant-specific surveillance capsule summary status information .

• Added Expected/Received fluence for the S2-W1 capsule.


• Added the R. E. Ginna plant-specific surveillance information.

• pdated the Westinghouse plant-specific surveillance p ram information based on the CRVSP and their res ctive NRC approved plant-specific RVSP withdrawal sche les.

• Updated e end-of-license or 1-2 times end-of-license fluence ca ules for some units based on fluence values in Table 1-9.

• Updated licensin dates and projected reactor vessel peak end-of-license flue es for some units .

• Deleted the Crystal Ri r Unit 3 plant-specific surveillance information.

• Added the R. E. Ginna plan information .

• Updated (including renumbering) ferences to include reports for the CRVSP and R. E. Gi a and deletion of references for the Crystal River Unit 3 surveillance capsules.

AREVA Inc.


BAW-1543 Revision 4

Page vii

material data corresponding to reactor vessel fluences for the Westinghouse-designed

SS reactor vessels between 1 and 2 times the 60 year fluence . The last supplement

document reviewed and approved by the NRC is BAW-1543(NP) , Revision 4,

(4l AREVA Inc. Document, "Supplement to the Master Integrated Reactor Vessel Surveillance Program,"

BAW-1543(NP), Revision 4, Supplement 6-A (43-1543S-11 ), June 2007.

AREVA Inc.


INTRODUCTION

BAW-1543 Revision 4

Page 1-1

REVA Inc. document, BAW-1543, Revision 4, reports the essential features of a

Master tegrated Reactor Vessel Surveillance Program (MIRVP) for all operating B&W

ts and those Westinghouse-designed plants having B&W-fabricated reactor

vessels .1 Th supplementary document to BAW-1543, Revision 4, contains

information on th surveillance capsule insertion and withdrawal schedules for the B&W

Westinghouse-designed plants with reactor vessel containing

Id metals. In addition , the insertion and withdrawal schedules

for the MIRVP supplemen ry capsules are provided. This document, Supplement 7, is

a revision to and replaces Su lement 6-A in its entirety. Table 1-1 and Table 1-2 are

listings of plant-specific surveilla e capsules and direct the reader to the appendices of

BAW-1543, Revision 4, where addi · nal information can be found on material and

capsule specifications. These tables a o provide a listing of surveillance capsule

reports available as of the date of this doc ment. Table 1-1 provides information for the

B&W 177-FA plant-specific capsules and Ta 1-2 provides information on the plant-

specific capsules for the Westinghouse-designe lants having B&W-fabricated reactor

vessels with high copper Linde 80 weld metals.

Table 1-3 and Table 1-4 provide capsule insertion and · hdrawal schedules for B&W

177-FA host plants Crystal River Unit 3 and Davis-Besse, r pectively. Table 1-3 was

revised by changing the capsule status of the OC1-D capsule , e OC3-F capsule and

the MIRVP supplemental capsules A2 and A4. The status of the 1-C capsule has

been updated in Table 1-4 to reflect that testing of the irradiated spe

performed.

Table 1-5 provides capsule insertion and withdrawal schedules for the Westi

designed plants with reactor vessel containing high copper Linde 80 weld metal

These listed plant-specific schedules are not MIRVP commitments , but merely refle

the current capsule withdraw plans for these plants in accordance with their respective

NRC approved plant-specific RVSP schedules. Table 1-5 was revised to update the

AREVA Inc.


BAW-1543 Revision 4

Supplement 7 Page 1-2

!ant-specific surveillance program status based on the Coordinated Reactor Vessel

eillance Program (CRVSP)2 and their respective NRC approved plant-specific

Table 1-6 nd Table 1-7 summarize the status of all MIRVP and plant-specific RVSP

W 177-FA and Westinghouse-designed plants with reactor vessels

containing high pper Linde 80 weld metals, respectively. These tables state whether

en withdrawn or are still being irradiated. For capsules that have

been withdrawn and te ted , the appropriate surveillance capsule report number has

been listed . For those ca ules that are being irradiated , the target and expected

fluences are listed along witti he insertion and/or withdrawal date. Table 1-6 was

revised by changing the status the TE1-C capsule . In addition, the time of withdrawal

of the supplemental capsules A2 a A4 was changed to "Not Planned." Table 1-7 was

revised to update the Westinghouse p nt-specific surveillance program status based on

the CRVSP and their respective NRC ap oved plant-specific RVSP withdrawal

schedules. Note that the Westinghouse-des ned plant-specific withdrawal schedules,

listed in Table 1-7, are not MIRVP commitment , but merely reflect the current capsule

withdraw plans for these plants in accordance with eir respective NRC approved

plant-specific RVSP schedules.

Table 1-8 shows the conformance of the current PWROG RVP member plant-specific

surveillance programs to the requirements of ASTM E185-82.

Table 1-9 lists current licensing dates and projected reactor vessel

fluences.

AREVA Inc.


BAW-1543 Revision 4

Paoe 1-17

Table 1-6 (cont.): Summary Status of the B&W Surveillance Capsules

Capsule Contents Fluence, n/cm 2 (x1019) Comments/ Capsule Fracture Status/ Time of Removal

ID Weld Toughness Location Target

Expected/ Metal Specimens

Received

TMl1-E X --- Tested ----- 0.107 ----- Reported in BAW-1439.Y

TMl1 -B --- --- Removed ----- 0.444 [Note 4] ----- Has been disposeg.7'

TMl1-C X --- Tested ----- 0.882 ----- Reported in s.¢"1901 17

TMl1-A X --- Removed ----- 0.158 ----- Held in stg;;ege - reported in BAW-204i6

TMl1-D --- --- Removed ----- 0.816 [Note 2] ----- Has ~ disposed66

TMl1-F --- --- Removed ----- 0.631 [Note 2] ----- ¥ been disposed66

AN1 -E X --- Tested ----- 0.0727 ----- / ....-Reported in BAW-1 44022

AN1 -B --- --- Tested ----- 0.428 ----- / Reported in BAW-169820

AN1-A X --- Tested ----- 1.03 --/ Reported in BAW-183619

AN1-C X --- Tested ----- 1.46 / ---- Reported in BAW-2075, Rev. 121

AN1-D --- --- Removed ----- 0.760 [Note 4] / ----- Has been disposed66

,,r Has been disposed66 AN1F --- --- Removed ----- 0.783 [Note 2Y -----

TE1 -F X X Tested ------ 0.196 / ----- Reported in BAW-1 701 29 and BAW-171 930

TE1-B X X Tested ----- o.s9r ----- Reported in BAW-183424 and BAW-186725

TE1-A X --- Tested ----- y.19 ----- Reported in BAW-1882, Rev. 123

TE1-D X X Tested ----- / 0.962 ----- Reported in BAW-212527 and BAW-220828

TE1-C X --- Tested /- 1.88 ----- Reported in ANP-333926

TE1-E X --- Removed / ------ 1.267 [Note 2] ----- Has been disposed66

AREVA Inc. BAW-1543 Rz Supplement to the Master Integrated Reactor Vessel Surveillance Program Sunnlement 7 P. e 1-18

Table 1-6 (cont.): Summary Status of the B&W Surveillance Capsules / Capsule Contents Fluence, n/cm 2 (x1019

) r Capsule Fracture Status/ Time of Removal ID Weld Toughness Location Target Expected/

Metal Specimens Received

CR3-LG1 X X Tested ----- 0.500-0.779 ----- Report~ BAW-1910P67 -·-----·--·-·-· .. - ------·-··-·----·-----

CR3-LG2 X X Tested ----- 1.19-1 .95 ----- Re_pdrted in BAW-2254P68 ..... .............. .... ................. ·-·-·-···-·--·--·-·····-·-·-·-······-·-·-······-

DB1-LG1 X X Tested ----- 0.661-1 .03 ----- l.-f(eported in BAW-1920P69 ........................................ .................. ............................. .....................

DB1-LG2 X X Tested ----- 1.10-1 .65 ----- / Reported in BAW-248670 ----- _____ ., ________ ___ ......... ............................

TMl2-LG1 X X Tested ----- 0.585-0.992 ----/ Reported in BAW-2253P71 . _______ ., ___ , .......... - ·-·---------·--·-·------

TMl2-LG2 X X Tested ----- 1.17-2.01 /--- Reported in BAW-243972

··- ---.. ----, .. -... -- .. A1 X X DB1-YZ 5.1 5.1 / standby -----

·-·. V A2 X X CR3-YX N/A ----- Not Planned -----

. -

·~ A3 X X Tested ----- ----- Reported in BAW-241273 ___ ,,,_ ........ ,_ .

A4 X X CR3-YX N/A Not Planned -----_____ .... , ............. ,_, ____ .

A5 X X Tested ----- 2 ----- Reported in BAW-2360P74 . -···-· -······-··-·-·-·····- -............................................. .. .... ..........

L1 X X Tested -----/ 1.169-1.624 ----- Reported in BAW-240075 . ................................... ·-·--.. , ..... , .. ,--,-.....................................

L2 X X DB1-YZ fi 4.5 Standby -----




3. NUREG CR-4816, Vol es 1 & 278

4. BAW-210879

AREVA Inc. BAW-1543

R~ Supplement to the Master Integrated Reactor Vessel Surveillance Program Sunnlement 7 P, e 1-19

Table 1-7: Summary Status of the Westinghouse Surveillance Capsules / Capsule Contents

Status/ Fluence, n/cm 2 (x10 19

)

~ ents Capsule

Weld Fracture Location Expected/ Time of Removal ID

Metal Toughness [Note 1] Target Received Specimens

PB1 -N X --- 33° ----- ----- End of License Sta_pdby [Note 2] ·--... - .. -·-.. -·-·-.. ··--

PB1 -P X --- Removed ----- ----- ----- _Aeld in Storage ·-·-···-·-··· .. ·-······-· .......... -

PB1 -R X X Tested ----- ----·- ----- / Reported in WCAP-935732

......... , .... -, .......... ,_,,_.,_

PB1 -S X - -- Tested ----- ----- -----/ Reported in WCAP-873933

-·--·---·--·· .. --PB1 -T X X Tested ----- ----- /-- Reported in WCAP-1073634

PB1 -V X X Tested ----- ----- / ----- Reported in BCL report dated 6/7335

Suppl. X X 13° 5.000 5.000 I/ Year2022 Supplemental Capsule

PB2-N X --- 33° ----- ----- / End of License Standby -- ----

PB2-P X --- Removed ----- --/ ----- Held in Storage -·--·-·-·--·-·--.. --

PB2-R X X Tested ----- /----- ----- Reported in WCAP-963538

------···-·-··-··--···-- ----------PB2-S X X Tested ----- ,,/ ----- ----- Reported in BAW-214039

·--···-·-···-·--··-·-·-··- ---·--· PB2-T X --- Tested -----/ ----- ----- Reported in WCAP-9331 40

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

PB2-V X X Tested /--- ----- ----- Reported in BCL report dated 6/7541

REG-N X X Tested V ----- ----- ----- Reported in WCAP-1 7036 Rev. 143

REG-P X X 23° / ----- ----- -EOC 41 Standby --REG-R X X T.¢ed ----- ----- ----- Reported in WCAP-8421 44

REG-S X X / 0 ested ----- ----- ----- Reported in WCAP-1390245

REG-T X x/ Tested ----- ----- ----- Reported in WCAP-1008646

-·--·----.. -·-·-·· REG-V X A Tested ----- ---- - ----- Reported in FR-RA-147

AREVA Inc.

Supplement to the Master Integrated R eactor V esse Suoolement 7

IS ·11 urve1 ance p rogram

BAW-1543 Revision 4

Paae 1-

Table 1-7 (cont.): Summary Status of the Westinghouse Surveillance Capsules / Capsule Contents

Status/ Fluence, n/cm2 (x1019

)

~ Capsule Weld

Fracture Location Expected/ Time of Removal ID

Metal Toughness [Note 1] Target Received Specimens

S1-S --- --- 25° ----- ----- End of License Standby V S1-T X X Tested ----- ----- ----- Reporte~ BCL report dated 617549

-------·--S1-U --- --- 45°/25° ----- ----- End of License Staryt{y , transferred to 25° at EOC-12

-MHONONmO-"N<_O_ ......

~ orted in WCAP-1141550 S1-V X X Tested ----- ----- -----·-·------·

----- / S1-W --- --- Dosimetry ----- ----- Evaluated for dosimetry only. Results reported in BCL-585-8R51

S1-X X X Tested ----- ----- --/ Reported in BAW-232452

S1-Y --- --- 35°115° ----- ----- E~ f License Standby, transferred to 15° at EOC-14

S1-Z X X 25°/15° 6.31 6.31 / Year2025 Standby, transferred to 15° at EOC-12

S2-S X --- Dosimetry ----- ----/ ----- Evaluated for dosimetry onli . Results reported in WCAP-14810 4

S2-T X --- 35°115° ----- /-- End of License Standby, transferred to 15° at EOC-17

S2-U X --- 25°/15° 5.95 V 5.95 Year2027 Standby, transferred to 15° in Year 2009 ------

S2-V X --- Tested ----- / ----- ----- Reported in WCAP-1149955

S2-W X --- Dosimetry /- ----- ----- Reported in BCL-585-02656

·-----S2-X X --- Tested V ----- --·--- ----- Reported in BCL report dated 917557

-·----S2-Y X X Tested / ----- ----- ----- Reported in WCAP-16001 58

S2-Z X X 35°¢ ----- ----- End of License Standby, transferred to 25° at EOC-12

S2-W1 X X l;rested ----- 0.602-0.802 ----- Reported in BAW-2350P80

AREVA Inc.



BAW-1543 Revis ion 4

Paqe 1 21

/ Capsule Contents

Status/ Fluence, n/cm2 (x1 019

) Comm•;/ Capsule Weld Fracture Location Expected/ Time of Removal

ID Metal Toughness [Note 1] Target Received

Specimens

TP3-S --- --- Tested ----- ----- ----- Reported in SwRl-97-'5131 60

-·---·-----TP3-T X X Tested ----- ----- ----- Reported in w.¢'-8631 61

--·--·-TP3-U --- -·-- 30° ----·- ----- End of License Standby /

-·-·-·-.. ·--·-·--·-TP3-V X X Tested ----- ----- ----- Repo~ in SwRl-06-857562

-·-·-·-, ................. ....... TP3-W --- --- 40° ----- ----- End of License ~ dby

·-·-···-·-·····-······-·--······

TP3-X X X Tested ----- ----- ----- / " Reported in WCAP-1591663 ........ - .... ,,.,_, ____ ..

TP3-Y --- --- 30° ----- ----- End of Licen¢ Standby ----·-----·--

TP3-Z --- --- 40° ----- ----- End of l.Je'{nse Standby

TP4-S --- --- Tested ----- ----- /----- Reported in SwRl-02-538060

TP4-T X X Tested ----- ----- / ----- Reported in SwRl-02-4221 65

- / TP4-U --- --- 30° ----- ----- / End of License Standby ____ .,._

----- / TP4-V X X 20° ----- End of License Standby ________ ................

TP4-W --- --- 40° ----- /-- End of License Standby ---·-·---·--

TP4-X X X 40°/0° 9.297 / 9.297 Year2021 -----·-·-···--······-······-·--····- -

TP4-Y --- --- 30° ----- / ----- End of License Standby --........................... - -------------/ TP4-Z --- --- 40° ----- End of License Standby


1. All locations are relative with regard to adrant ; e.g. , 0° is equivalent to 90°, 180°, or 270°.

2. At current lead factor, capsule r ches -8.4x1019 n/cm2 at -42 EFPY of vessel operation (equ ivalent to Surry Unit 1 projected 80-year vessel fluence) . Movement to a ca le location with higher lead factor red uces time of vessel operation to reach this fluence.

AREVA Inc. BAW-1543 Rz:: Supplement to the Master Integrated Reactor Vessel Surveillance Program Suoolement 7 P. e 1-22

Table 1-8: Comparison of the Plant-Specific Surveillance Capsules with ASTM E 185-82 Req~ ts

ASTM E 185-82 Five (5) Capsule Program Requirements / 1.5 EFPY

3 EFPY ~ - EOL

Plant or 6 EFPY 15 EFPY or or

Fluence > 5E18 Fluence Midway or or Times EOL Fluence

LlRT NOT::::: 50°F Between First and T/4 EOL Fluence 1SEOLF1u7 (Capsule may be

Third Capsule held w/o testing)

Oconee Unit 1 F - (1/T) [Note 1) E - (1/T) A-~ ) CR3-LG2 - (1/T)

Oconee Unit 2 C - (1/T) A- (1/T) TMl2-LG1 - (1/T) 7- (1/T) AS - (1/T)

Oconee Unit 3 A- (1/T) B - (1/T) L 1 - (1/T) / D-(1/T) CR3-LG2 - (1/T)

TMI Unit 1 E - (1/T) W1 - (1/T) CR3-LG1 - (1/T)/ C - (1/T) AS - (1/T)

ANO Unit 1 E-(1/T) [Note 1) B- (1/TV A - (1/T) C - (1/T)

Davis-Besse F- (1/T) [Note 1) B-~ ) D - (1/T) A - (1/T); C - (1/T)

Point Beach Unit 1 V- (1/T) S - (1/T) / -(1/T) T - (1/T) N - (IR)

Point Beach Unit 2 V- (1/T) T - (1/T) / R - (1/T) S - (1/T) N - (IR)

R. E. Ginna V - (1/T) R - (1/T) /' T - (1/T) S - (1/T) N - (1/T); P - (IR)

Surry Un it 1 T - (1/T) W - (1/T) [Not~ V- (1/T) X - (1/T) Z- (IR)

Surry Un it 2 X - (1/T) W- (1/T) (P'6te 2) V - (1/T) y - (1/T) U-(IR)

Turkey Point Unit 3 T - (1/T) ¥(1/T) V- (1/T) X- (1/T) TP4-X - (IR)

Turkey Point Unit 4 T - (1/T) / S- (1/T) TP3-V - (1/T) TP3-X - (1/T) X- (IR)

Legend: (1/T) = Irradiated and tested , (I/ ) = Irradiated and not tested , (IR) = Currently in reactor vessel


1. Only 4 capsules requir per ASTM E185-82.

2. On ly dosimetry aluated.

AREVA Inc. BAW-1543

~L Supplement to the Master Integrated Reactor Vessel Surveillance Program Supplement 7 P. e 1-23

Table 1-9: Peak End-of-Life Inside Surface Fluences and Significant Licensing Date~

~ rojected EOL Date Construction Date Operating License e Surface (IS)

Plant Permit Issued License Issued Expiration

/ Fluence, n/cm2

(E > 1 MeV)

Oconee Unit 1 November 6, 1967 February 6, 1973 February 6, 2033 [Noj,e1'] 1.29E+19

Oconee Unit 2 November 6, 1967 October 6, 1973 October 6, 2033 ¥ate 1) 1.44E+19

Oconee Unit 3 November 6, 1967 July 19, 1974 July 19, 203~ ote 1) 1.38E+19

TMI Unit 1 May18, 1968 April 19, 1974 April 19yl'634 [Note 1) 1.52E+19

ANO Unit 1 December 6, 1968 May21 , 1974 Ma~ , 2034 [Note 1) 1.35E+19

Davis-Besse March 24, 1971 April 22 , 1977 vPril 22 , 2017 1.04E+19

Point Beach Unit 1 July 19, 1967 October 5, 1970 / October 5, 2030 [Note 1) 5.09E+19

Point Beach Unit 2 July 25, 1968 March 8, 1973 / March 8, 2033 [Note 1) 5.07E+19

R. E. Ginna April 25, 1966 September J.S(1 969 September 18, 2029 [Note 1) 5.45E+19

Surry Unit 1 June 25, 1968 May25~ 72 May 25, 2032 [Note 1) 5.66E+19

Surry Unit 2 June 25, 1968 Ja~ 29, 1973 January 29, 2033 [Note 1) 5.38E+19

Turkey Point Unit 3 Apri l 27, 1967 Vu1y 19, 1972 July 19, 2032 [Note 1) 6.38E+19

Turkey Point Unit 4 April 27, 1967 / April 10, 1973 April 10, 2033 [Note 1) 6.38E+19


1. Includes 20 year license renewal te