ML20273A215

License Amendment Request: Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report

ML20273A215
Person / Time
Site:	Limerick
Issue date:	09/29/2020
From:	David Helker Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML20273A214	List: ML20273A215
References
Download: ML20273A215 (271)

Text

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 September 29, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

Subject:

License Amendment Request - Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report In accordance with 10 CFR 50.90, Exelon Generation Company, LLC (Exelon) requests proposed changes that would modify Technical Specification (TS) Section 1.0

("DEFINITIONS"), Section 3/4.4.6 (PRESSURE/TEMPERATURE LIMITS) and Section 6.0

("ADMINISTRATIVE CONTROLS) by replacing the existing reactor vessel heatup and cooldown rate limits and the pressure and temperature (P-T) limit curves with references to the Pressure and Temperature Limits Report (PTLR) at Limerick Generating Station (LGS),

Units 1 and 2.

The proposed change has been reviewed by the Limerick Plant Operations Review Committee in accordance with the requirements of the Exelon Quality Assurance Program.

Exelon requests approval of the proposed amendments by September 29, 2021. Once approved, these amendments shall be implemented within 60 days of issuance. contains the evaluation of the proposed changes. Attachment 2 provides the marked up TS pages. Attachment 3 contains the marked up Bases pages. The Bases pages are being provided for information only. Attachments 6, 7, 11, and 12 contain the PTLRs for LGS, Units 1 and 2.

Attachments 11, 12, and 13 (BWRVIP-135, Revision 3: BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations) contain information proprietary to EPRI. EPRI requests that these documents be withheld from public disclosure in accordance with 10 CFR 2.390(b)(4). An Affidavit supporting this request is contained in Attachment 5. Non-proprietary versions of these documents are contained in Attachments 6, 7, and 8.

Attachments 14 and 15 contain information proprietary to TransWare Enterprises Inc.

TransWare Enterprises Inc. requests that these documents be withheld from public disclosure in accordance with 10 CFR 2.390(b)(4). Affidavits supporting this request are contained in Attachment 4. Non-proprietary versions of this document are contained in Attachments 9 and 10.

Attachments 11, 12, 13, 14, and 15 transmitted herewith contain Proprietary Information. When separated from attachments, this document is decontrolled.

License Amendment Request -

Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report September 29, 2020 Page 2 There are no regulatory commitments contained in this letter.

In accordance with 10 CFR 50.91, Exelon is notifying the Commonwealth of Pennsylvania of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 29th of September 2020.

Respectfully, David P. Helker Sr. Manager, Licensing Exelon Generation Company, LLC : Evaluation of Proposed Changes : Markup of Technical Specifications Pages : Markup of Technical Specifications Bases Pages : TransWare Enterprises Inc. Affidavits : Electric Power Research Institute (EPRI) Affidavit : Limerick Generating Station Unit 1 Pressure and Temperature Limits Report (PTLR) for 57 Effective Full-Power Years (EFPY) (Non-Proprietary Version) : Limerick Generating Station Unit 2 Pressure and Temperature Limits Report (PTLR) for 57 Effective Full-Power Years (EFPY) (Non-Proprietary Version) : BWRVIP-135, Revision 3: BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations (Non-Proprietary Version) : Limerick Generating Station Unit 1 Reactor Pressure Vessel Fluence Evaluation - Non-Proprietary Report, LIM-FLU-002-R-009, Rev. 1 (Non-Proprietary) 0: Limerick Generating Station Unit 2 Reactor Pressure Vessel Fluence Evaluation - Non-Proprietary Report, LIM-FLU-002-R-010, Rev. 1 (Non-Proprietary) 1: Limerick Generating Station Unit 1 Pressure and Temperature Limits Report (PTLR) for 57 Effective Full-Power Years (EFPY) (Proprietary Version) 2: Limerick Generating Station Unit 2 Pressure and Temperature Limits Report (PTLR) for 57 Effective Full-Power Years (EFPY) (Proprietary Version) 3: BWRVIP-135, Revision 3: BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations (Proprietary Version)

License Amendment Request -

Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report September 29, 2020 Page 3 4: Limerick Generating Station Unit 1 Reactor Pressure Vessel Fluence Evaluation - Licensing Report, LIM-FLU-002-R-005, Rev. 1 (Proprietary) 5: Limerick Generating Station Unit 2 Reactor Pressure Vessel Fluence Evaluation - Licensing Report, LIM-FLU-002-R-006, Rev. 1 (Proprietary) cc: USNRC Region I, Regional Administrator USNRC Project Manager, Limerick USNRC Senior Resident Inspector, Limerick Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection W. DeHass, Pennsylvania Department of Environmental Protection

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES

SUBJECT:

License Amendment Request - Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

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

6.0 REFERENCES

Page 1

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) requests an amendment to Renewed Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2. The proposed amendment modifies the Technical Specifications (TS) by replacing the pressure and temperature (P-T) limit curves with references to the Pressure and Temperature Limits Report (PTLR). Attachments 6, 7, 11, and 12 contain copies of the PTLRs.

The PTLRs contain updates to the P-T limit curves for the beltline, bottom head, and non-beltline regions for the LGS, Units 1 and 2 reactor pressure vessels (RPV). The P-T curves are developed for 57 effective full power years (EFPY) of operation. The P-T curves were prepared using the methods documented in the Boiling Water Reactor Owners' Group (BWROG)

Licensing Topical Report (LTR) BWROG-TP-11-022-A, Revision 1 (Structural Integrity Associates, Inc. Report SIR-05-044, Revision 1-A), Pressure-Temperature Limits Report Methodology for Boiling Water Reactors (Reference 1). This BWROG LTR satisfies the requirement of 10 CFR 50, Appendix G, and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, Nonmandatory Appendix G (Reference 2).

The guidance of USNRC Generic Letter (GL) 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, (Reference 3) was applied during P-T curve development. Also, Technical Specification Task Force (TSTF)

Traveler TSTF-419-A, Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR (Reference 4), which has received USNRC approval, was followed in development of the proposed TS changes.

2.0 DETAILED DESCRIPTION The proposed change includes the following TS revisions:

a) TS Section 1.0, "DEFINITIONS" - A new definition, "Pressure and Temperature Limits Report," is added. The wording for this definition is consistent with that in TSTF-419-A.

b) TS Section 3/4.4.6, PRESSURE/TEMPERATURE LIMITS - The P-T curves and the associated TS wording have been deleted and replaced with references to the PTLR.

c) TS Section 6.0, "ADMINISTRATIVE CONTROLS" - A new Section 6.9.1.13 ("REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR))

has been added. The format and content are consistent with that in TSTF-419-A and includes the full methodology citation. This new section: (1) identifies the individual TS that address reactor coolant system P-T limits; (2) references the USNRC-approved topical report that documents PTLR methodologies in a complete citation; and (3) requires that the PTLR and any revision or supplement thereto be submitted to the USNRC. provides the existing TS pages marked up to show the proposed changes.

Marked up pages showing corresponding changes to the TS Bases are provided in for information only.

Page 2

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES The attached PTLRs provide the P-T curves developed to represent steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

3.0 TECHNICAL EVALUATION

10 CFR 50, Appendix G, requires the establishment of P-T limits for material fracture toughness requirements of the Reactor Coolant Pressure Boundary materials. 10 CFR 50, Appendix G requires an adequate margin to brittle failure during normal operation, abnormal operational transients, and system hydrostatic tests.

Historically, the P-T limit curves for BWRs have been contained in the TS, necessitating the submittal of license amendment requests to update the curves. This caused both the U.S.

Nuclear Regulatory Commission (USNRC) and licensees to expend resources that could otherwise be devoted to other activities.

The pressure and temperature limits were calculated in accordance with Reference 1. The neutron fluence is calculated in accordance with USNRC Regulatory Guide 1.190 (RG 1.190)

(Reference 5) using the Radiation Analysis Modeling Application (RAMA) computer code.

Attached are the TransWare Enterprises Inc. neutron transport or neutron flux calculations used in the preparation of the LGS, Units 1 and 2 PTLRs. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with USNRC Regulatory Guide 1.99, Revision 2 (Reference 6).

Generic Letter 96-03 allows plants to relocate their P-T curves and associated numerical limits (such as heatup and cooldown rates) from the plant TS to a PTLR, which is a licensee-controlled document. As stated in Generic Letter 96-03, during the development of the improved Standard Technical Specifications (STS), a change was proposed to relocate the P-T limits currently contained in the plant TS to a PTLR. As one of the improvements to the STS, the USNRC staff agreed with the industry that the curves may be relocated outside the plant TS to a PTLR so that the licensee could maintain these limits efficiently. One of the prerequisites for having the PTLR option is that the P-T curves and limits be derived using methodologies approved by the USNRC, and that the associated licensing topical reports describing the approved methodologies be referenced in the plant TS.

By letter dated May 16, 2013 (Reference 7), the USNRC staff found that Topical Report (TR)

BWROG-TP-11-022, Revision 1, is acceptable for referencing in licensing applications for boiling water reactors 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." This Safety Evaluation Report (SER) permits licensees who use the BWROG-TP-11-022 methodology and follow the PTLR guidance in GL 96-03 to relocate their P-T curves from the facility TS to a PTLR using the guidance in Technical Specification Task Force (TSTF) Traveler No. 419-A. The BWROG issued the final report on September 4, 2013 (Reference 8), which contains the Revision 1-A of BWROG-TP-11-022, Revision 1, dated August 2013, the final SER, along with the USNRC requests for additional information (RAIs), and the BWROG's responses to the USNRC RAIs.

The USNRC SER contained one condition for future potential applicants to address in their application of this LTR to their plant-specific P-T limits or PTLR submittal:

Page 3

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES Each applicant referencing this LTR shall confirm that, in addition to the requirements in the ASME Code,Section XI, Appendix G, the lowest service temperatures for all ferritic RCPB components that are not part of the RPV, are below the lowest operating temperature in the proposed P-T limits.

This condition is discussed further in the Technical Analysis section of this LAR.

Generic Letter 96-03 provides regulatory guidance regarding relocation of P-T curves and associated numerical limits, such as heatup and cooldown rates, from plant TS to a PTLR, a Licensee-controlled document. As stated in GL 96-03, a licensee requesting such a change must satisfy the following three criteria:

1. Have USNRC-approved methodologies to reference in the TS.

2. Develop a PTLR to contain the P-T limit curves, associated numerical limits, and any necessary explanation, and

3. Modify applicable sections of the TS accordingly.

Criterion 7 of Generic Letter 96-03 specifies that the licensee should "provide supplemental data and calculations of the chemistry factor in the PTLR if the surveillance data are used in the ART

[adjusted reference temperature] calculation." Therefore, in order to ensure that the proposed PTLR is consistent with Criterion 7 in GL 96-03, Exelon is providing supporting data and calculations from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program Data Source Book (BWRVIP-135) for determining the proprietary Integrated Surveillance Program (ISP) material chemistry factor values listed in the PTLRs.

Revised P-T curves were developed for hydrostatic pressure and leak tests, and normal operation with core not critical and core critical conditions. The revised curves have been developed for application up to 57 EFPY.

The revised LGS, Units 1 and 2 P-T curves were prepared using the methods documented in the BWROG-TP-11-022-A, Revision 1 (Reference 1). This BWROG LTR provides USNRC-approved BWROG fracture mechanics methodologies for generating P-T curves/limits and allows BWR plants to adopt the PTLR option in accordance with TSTF-419-A and GL 96-03.

The LTR satisfies the requirements of 10 CFR 50, Appendix G, as augmented by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Nonmandatory Appendix G.

The LTR has four sections and four appendices, the content of which is summarized below.

• Section 1.0 describes the background and purpose for the LTR. Attachment 1 of GL 96-03 provides seven technical criteria to which a methodology should conform, in order to develop P-T limits acceptable by the USNRC staff. These seven criteria are explained in this section.

• Section 2.0 of the BWROG LTR provides the fracture mechanics methodology and its basis for developing P-T limits.

Page 4

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES

• Section 3.0 of the BWROG LTR provides a step-by-step procedure for calculating P-T limit curves. This section indicates that typically three reactor pressure vessel regions are evaluated with respect to P-T limits: (1) the beltline region; (2) the bottom head region; and (3) the non-beltline region.

• Section 4.0 of the BWROG LTR provides the references.

• Appendix A of the LTR provides guidance for evaluating surveillance data.

• Appendix B provides a template for development of an acceptable PTLR.

• Appendix C provides the Revision 0 requests for additional information, along with the respective responses.

• Appendix D provides the Revision 1 requests for additional information, along with the respective responses.

Neutron Fluence Calculations:

The neutron fluence calculations were updated using the USNRC-approved methodology and in accordance with USNRC Regulatory Guide 1.190 (Reference 5).

The fluence is based upon operation for 57 EFPY. The calculated neutron fluences at the end of 57 EFPY are provided in Table 4 of each PTLR.

10 CFR 50, Appendix G, requires reactor vessel beltline materials to be tested in accordance with the surveillance program requirements of 10 CFR 50, Appendix H.

LGS, Units 1 and 2 has replaced the original RPV material surveillance program with the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP).

LGS, Units 1 and 2 are committed to use the BWRVIP ISP during the current licensed period.

Use of the BWRVIP ISP for LGS, Units 1 and 2 was approved by the USNRC on November 4, 2003 (Reference 9). LGS, Units 1 and 2 have made a license renewal commitment to use the ISP during the period of extended operation (Reference 10). The Reactor Vessel Surveillance program is based on BWRVIP-86 Rev 1-A, BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project Updated BWR Integrated Surveillance Program (ISP) Implementation Plan, dated October 2012.

Pressure-Temperature Curve Evaluation:

Three regions of the reactor pressure vessel were evaluated to develop the revised P-T curves:

the beltline region, the bottom head region, and the feedwater nozzle/upper vessel region.

These regions bound all other regions with respect to brittle fracture.

The methodology used to generate the P-T curves in this submittal is approved by the USNRC, and uses adjusted reference temperature (ART) values determined in accordance with RG 1.99, Revision 2.

The revised P-T curves and outputs from the ISP ensure that adequate RPV safety margins against non-ductile failure will continue to be maintained during normal operations, anticipated Page 5

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES operational occurrences, and inservice leak and hydrostatic testing. Together, these measures ensure that the integrity of the reactor coolant pressure boundary (RCPB) will be maintained for the life of the plant.

Proposed revisions to applicable sections of the TS have been prepared and are provided in . These proposed changes are consistent with the guidance provided in GL 96-03, as supplemented by TSTF-419-A.

Conditions and Limitations:

The USNRC SER contained one condition for future potential applicants to address in their application of this LTR to their plant-specific P-T limits or PTLR submittal:

Each applicant referencing this LTR shall confirm that, in addition to the requirements in the ASME Code,Section XI, Appendix G, the lowest service temperatures for all ferritic RCPB components that are not part of the RPV, are below the lowest operating temperature in the proposed P-T limits.

LGS, Units 1 and 2 have confirmed the lowest service temperatures for all ferritic RCPB components that are not part of the RPV, are below the lowest operating temperature in the proposed P-T limits. This confirmation has been included in Section 4.0, Operating Limits, of the LGS, Units 1 and 2 PTLR.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c)(2), Limiting conditions for operation, states: (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

The USNRC has established requirements in 10 CFR 50, Appendix G, Fracture Toughness Requirements, in order to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. Appendix G requires that the pressure and temperature limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated if the methods and margins of safety of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code were used to generate the pressure and temperature limits. Also, Appendix G requires that applicable surveillance data from reactor pressure vessel material surveillance programs be incorporated into the calculations of plant-specific pressure and temperature limits, and that the pressure and temperature limits for operating reactors be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.

Appendix H to 10 CFR Part 50 provides requirements related to facility reactor pressure vessel material surveillance programs. LGS, Units 1 and 2 demonstrates its compliance with the requirements of 10 CFR 50, Appendix H, through participation in the BWRVIP Integrated Surveillance Program (ISP) and the latest material information was utilized in preparation of the report.

Page 6

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation.

10 CFR 50.36, Technical specifications, provides the regulatory requirements for the content required in the TSs which includes limiting conditions for operation (LCOs), surveillance requirements and administrative controls. Previously the plant-specific pressure and temperature limits had been incorporated into the TS and controls were placed on operation and testing by the associated specification. This proposed change revises the TS to relocate the pressure and temperature limit curves to a licensee-controlled document in accordance with the guidance of Generic Letter 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits and TSTF-419-A, Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR.

LGS, Units 1 and 2 have determined that the proposed change does not require any exemptions or relief from regulatory requirements, other than the TS, and does not affect conformance with the intent of any of the General Design Criteria (GDC) differently than described in the Safety Analysis Report.

4.2 Precedent The USNRC has approved similar license amendments to relocate P-T limit curves to a PTLR.

Recent examples for boiling water reactor plants include:

1) Brunswick Steam Electric Plant, Unit Nos. 1 and 2, License Amendment Nos. 289 and 317 issued by USNRC letter dated April 22, 2019, ADAMS Accession No. ML19035A006.

2) Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, License Amendment Nos. 277 and 221 issued by USNRC letter dated March 23, 2016, ADAMS Accession No. ML16062A099.

4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (Exelon) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below.

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

Response: No The proposed license amendment adopts the USNRC approved methodology described in Boiling Water Reactor Owners Group (BWROG) Licensing Topical Report (LTR)

BWROG-TP-11-022-A, Revision 1 (Structural Integrity Associates, Inc. Report SIR-05-044, Revision 1-A), Pressure-Temperature Limits Report Methodology for Boiling Water Reactors, dated August 2013. The LGS, Units 1 and 2 PTLR was developed based on the methodology and template provided in the BWROG LTR.

Page 7

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES 10 CFR 50, Appendix G, establishes requirements to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants.

Implementing this USNRC approved methodology does not reduce the ability to protect the reactor coolant pressure boundary as specified in Appendix G, nor will this change increase the probability of malfunction of plant equipment, or the failure of plant structures, systems, or components. Incorporation of the new methodology for calculating pressure and temperature limit curves, and the relocation of the pressure and temperature limit curves from the TS to the PTLR provides an equivalent level of assurance that the reactor coolant pressure boundary is capable of performing its intended safety functions.

The proposed changes do not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated and maintained. The ability of structures, systems, and components to perform their intended safety functions is not altered or prevented by the proposed changes, and the assumptions used in determining the radiological consequences of previously evaluated accidents are not affected.

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

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

Response: No Creation of the possibility of a new or different kind of accident requires creating one or more new accident precursors. New accident precursors may be created by modifications of plant configuration, including changes in allowable modes of operation.

The change in methodology for calculating pressure and temperature limits and the relocation of those limits to the PTLR do not alter or involve any design basis accident initiators. Reactor coolant pressure boundary integrity will continue to be maintained in accordance with 10 CFR 50, Appendix G, and the assumed accident performance of plant structures, systems and components will not be affected. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), and the installed equipment is not being operated in a new or different manner.

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

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No The proposed changes do not affect the function of the reactor coolant pressure boundary or its response during plant transients. Calculating the LGS, Units 1 and 2 pressure temperature limits using the USNRC-approved methodology ensures adequate margins of safety relating to reactor coolant pressure boundary integrity are maintained. The proposed changes do not alter the manner in which the Limiting Conditions for Operation Page 8

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES pressure and temperature limits for the reactor coolant pressure boundary are determined.

There are no changes to the setpoints at which protective actions are initiated, and the operability requirements for equipment assumed to operate for accident mitigation are not affected.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above evaluation, Exelon concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified.

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

5.0 ENVIRONMENTAL CONSIDERATION

Exelon has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation." However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,"

paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0 REFERENCES

1) Boiling Water Reactor Owner's Group (BWROG) Licensing Topical Report (LTR) BWROG-TP-11-022-A, Revision 1 (Structural Integrity Associates, Inc. Report SIR-05-044, Revision 1-A), Pressure-Temperature Limits Report Methodology for Boiling Water Reactors, dated August 2013.

2) American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code,Section XI, Nonmandatory Appendix G, 2007 Edition, 2008 Addenda.

3) USNRC Generic Letter 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, dated January 31, 1996.

4) Technical Specification Task Force (TSTF) Traveler TSTF-419-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR," dated August 4, 2003.

Page 9

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES

5) USNRC Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, dated March 2001.

6) NRC Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Material, Revision 2, dated May 1988.

7) Letter from Sher Bahadur (USNRC) to Frederick Schiffley (BWROG), Final Safety Evaluation for Boiling Water Reactor Owners' Group Topical Report BWROG-TP-11-022, Revision 1, November 2011, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors" (TAC No. ME7649), dated May 16, 2013 (ML13107A062).

8) Letter from F. Schiffley (BWROG) to U.S. Nuclear Regulatory Commission Document Control Desk, Submittal of Boiling Water Reactor Owners' Group Topical report BWROG-TP-11-022-A, Revision 1, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors" (TAC No. ME7649), dated September 4, 2013 (ML13277A557).

9) Letter from S. Wall (U.S. Nuclear Regulatory Commission) to J. Skolds (Exelon Generation Company, LLC), Limerick Generating Station, Units 1 and 2 - Issuance of Amendment Re.

Revision to the Reactor Pressure Vessel Material Surveillance Program (TAC Nos. MB7003 and MB7004), dated November 4, 2003 (ML032310540).

10) NUREG-2171, Safety Evaluation Report Related to the License Renewal of Limerick Generating Station, Units 1 and 2, dated September 2014 (ML14276A156).

Page 10

ATTACHMENT 2 Markup of Technical Specifications Pages Revised Pages 1-5 3/4 4-18 3/4 4-19 3/4 4-20 6-18a

DEFINITIONS OPERATIONAL CONDITION - CONDITION 1.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant tempera-ture as specified in Table 1.2.

PHYSICS TESTS 1.27 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.28 PRESSURE BOUNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, pipe wall or vessel wall.

PRIMARY CONTAINMENT INTEGRITY 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when:

INSERT A a. All primary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except for valves that are opened under administrative control as permitted by Specification 3.6.3.

b. All primary containment equipment hatches are closed and sealed.

c. The primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.

d. The primary containment leakage rates are within the limits of Specification 3.6.1.2.

e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.

f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows, or O-rings, is OPERABLE.

PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assure that the solidification or dewatering and packaging of radioactive wastes results in a waste package with properties that meet the minimum and stability requirements of 10 CFR Part 61 and other requirements for transportation to the disposal site and receipt at the disposal site.

With solidification or dewatering, the PCP shall identify the process parameters influencing solidification or dewatering, based on laboratory scale and full scale testing or experience.

LIMERICK - UNIT 1 1-5 Amendment No. 48, 66, 146

DEFINITIONS OPERATIONAL CONDITION - CONDITION 1.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant tempera-ture as specified in Table 1.2.

PHYSICS TESTS 1.27 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.28 PRESSURE BOUNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, pipe wall or vessel wall.

PRIMARY CONTAINMENT INTEGRITY 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when:

a. All primary containment penetrations required to be closed during accident conditions are either:

INSERT A

1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except for valves that are opened under administrative control as permitted by Specification 3.6.3.

b. All primary containment equipment hatches are closed and sealed.

c. The primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.

d. The primary containment leakage rates are within the limits of Specification 3.6.1.2.

e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.

f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows, or O-rings, is OPERABLE.

PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assure that the SOLIDIFICATION or dewatering and packaging of radioactive wastes results in a waste package with properties that meet the minimum and stability requirements of 10 CFR Part 61 and other requirements for trans-portation to the disposal site and receipt at the disposal site. With SOLIDIFICATION or dewatering, the PCP shall identify the process parameters influencing SOLIDIFICATION or dewatering based on laboratory scale and full scale testing or experience.

LIMERICK - UNIT 2 1-5 Amendment No. 11, 48, 107

INSERT A PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 1.28a The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current vessel fluence period. The pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.9.1.13.

REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM limits specified in the PTLR LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curve A for hydrostatic or leak testing; (2) curve B for heatup by non-nuclear means, cool-down following a nuclear shutdown and low power PHYSICS TESTS; and (3) curve C for operations with a critical core other than low power PHYSICS TESTS, with:

rate within the limits specified in the PTLR

a. A maximum heatup of 100°F in any 1-hour period, rate within the limits specified in the PTLR

b. A maximum cooldown of 100°F in any 1-hour period, within the limits specified in the PTLR

c. A maximum temperature change of less than or equal to 20°F in any 1-hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves, and within the limits specified in the PTLR

d. The reactor vessel flange and head flange temperature greater than or equal to 80°F when reactor vessel head bolting studs are under tension.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6.1-1 curve A, B, or C as applicable, in accordance with the Surveillance Frequency Control Program.

limits specified in the PTLR LIMERICK - UNIT 1 3/4 4-18 Amendment No. 36, 106, 145, 186

REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM limits specified in the PTLR LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curve A for hydrostatic or leak testing; (2) curve B for heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curve C for operations with a critical core other than low power PHYSICS TESTS, with: rate within the limits specified in the PTLR

a. A maximum heatup of 100°F in any 1-hour period, rate within the limits specified in the PTLR

b. A maximum cooldown of 100°F in any 1-hour period, within the limits specified in the PTLR

c. A maximum temperature change of less than or equal to 20°F in any 1-hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves, and within the limits specified in the PTLR

d. The reactor vessel flange and head flange temperature greater than or equal to 70°F when reactor vessel head bolting studs are under tension.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6.1-1 curves A, B or C as applicable, in accordance with the Surveillance Frequency Control Program.

limits specified in the PTLR LIMERICK - UNIT 2 3/4 4-18 Amendment No. 111, 147

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curve C within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality and in accordance with the Surveillance Frequency Control Program during system heatup.

limits specified in the PTLR 4.4.6.1.3 DELETED 4.4.6.1.4 DELETED within the limits specified in the PTLR 4.4.6.1.5 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 80°F:

a. In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:

1. 100°F, in accordance with the Surveillance Frequency Control Program.

2. 90°F, in accordance with the Surveillance Frequency Control Program.

b. Within 30 minutes prior to and in accordance with the Surveillance Frequency Control Program during tensioning of the reactor vessel head bolting studs.

LIMERICK - UNIT 1 3/4 4-19 Amendment No. 29,36,126,167, 186

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curve C within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality and in accordance with the Surveillance Frequency Control Program during system heatup.

4.4.6.1.3 DELETED limits specified in the PTLR 4.4.6.1.4 DELETED within the limits specified in the PTLR 4.4.6.1.5 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 70°F:

a. In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:

1. 100°F, in accordance with the Surveillance Frequency Control Program.

2. 90°F, in accordance with the Surveillance Frequency Control Program.

b. Within 30 minutes prior to and in accordance with the Surveillance Frequency Control Program during tensioning of the reactor vessel head bolting studs.

LIMERICK - UNIT 2 3/4 4-19 Amendment No. 94, 111, 130, 147

INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED A22 A B C PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig)

A, B, C - CORE BELTLINE AFTER ASSUMED 69°F SHIFT FROM AN INITIAL PLATE RTNDT OF 20°F A - SYSTEM HYDROTEST WITH FUEL IN THE VESSEL B - NON-NUCLEAR HEATUP/COOLDOWN LIMIT C - NUCLEAR (CORE CRITICAL) 312 PSIG LIMIT CURVES A, B, C ARE VALID UP TO 32 EFPY OF OPERATION CURVE A22 IS VALID UP TO BOLTUP 80°F 22 EFPY OF OPERATION MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE FIGURE 3.4.6.1-1 LIMERICK - UNIT 1 3/4 4-20 Amendment No. 36, 106, 145,155, 163

INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED A22 A B C 1400 1300 1200 1100 PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 900 800 700 A, B, C - CORE BELTLINE AFTER ASSUMED 82°F SHIFT FROM AN INITIAL PLATE RTNDT 600 OF 40°F A - SYSTEM HYDROTEST 500 WITH FUEL IN THE VESSEL B - NON-NUCLEAR 400 HEATUP/COOLDOWN LIMIT 312 PSIG C - NUCLEAR (CORE CRITICAL) 300 LIMIT A B C 200 CURVES A, B, C ARE VALID UP TO 32 EFPY OF OPERATION BOLTUP 100 70°F CURVE A22 IS VALID UP TO 22 EFPY OF OPERATION 0

MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE FIGURE 3.4.6.1-1 LIMERICK - UNIT 2 3/4 4-20 Amendment No. 51, 80, 111, 125

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b. MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1,

c. The MINIMUM CRITICAL POWER RATIO (MCPR) and MCPR(99.9%) for Specification 3.2.3,

d. The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3,

e. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,

f. The power biased Rod Block Monitor setpoints of Specification 3.3.6 and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.1.4.3,

g. The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6,

h. The Oscillation Power Range Monitor (OPRM) period based detection algorithm (PBDA) setpoints for Specification 2.2.1,

i. The minimum required number of operable main turbine bypass valves for Specification 3.7.8 and the TURBINE BYPASS SYSTEM RESPONSE TIME for Specification 4.7.8.c.

6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a. NEDE-24011-P-A "General Electric Standard Application for Reactor Fuel" (Latest approved revision),*

b. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions INSERT B Licensing Basis Methodology for Reload Applications, August 1996.

6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

• For Cycle 8, specific documents were approved in the Safety Evaluation dated (5/4/98) to support License Amendment No. (127).

LIMERICK - UNIT 1 6-18a Amendment No. 37,66,77,127,142,177, 200, 236

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b. MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1,

c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3,

d. The MCPR(P) and MCPR(F) adjustment factor for specification 3.2.3,

e. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,

f. The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6.

g. The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6,

h. The Oscillation Power Range Monitor (OPRM) period based detection algorithm (PBDA) setpoints for Specification 2.2.1,

i. The minimum required number of operable main turbine bypass valves for Specification 3.7.8 and the TURBINE BYPASS SYSTEM RESPONSE TIME for Specification 4.7.8.c.

6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a. NEDE-24011-P-A "General Electric Standard Application for Reactor Fuel" (Latest approved revision),

b. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, August 1996.

INSERT B 6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

LIMERICK - UNIT 2 6-18a Amendment No. 4,38,48,104,139, 161

INSERT B REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.13 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

a. Limiting Condition for Operation Section 3.4.6, "RCS Pressure/Temperature Limits"

b. Surveillance Requirement Section 4.4.6, "RCS Pressure/Temperature Limits" The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a. BWROG-TP-11-022-A, Revision 1 (SIR-05-044), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated August 2013.

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

ATTACHMENT 3 Markup of Technical Specifications Bases Pages Revised Pages B 3/4 4-5 B 3/4 4-7 B 3/4 4-8

REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued) specified in the PTLR The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code Section XI, Appendix G. The curves are based on the RTNDT and stress intensity factor information for the reactor vessel components. Fracture toughness limits and the basis for compliance are more fully discussed in FSAR Chapter 5, Para-graph 5.3.1.5, "Fracture Toughness." specified in the PTLR The reactor vessel materials have been tested to determine their initial RTNDT. The results of these tests are shown in Table B 3/4.4.6-1. Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material the PTLR in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda-tions of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials." The pressure/temperature limit curves, Figure 3.4.6.1-1, include a shift in RTNDT for conditions at 32 EFPY. The A, B and C limit curves are predicted to be bounding for all areas of the RPV until 32 EFPY. In addition, an intermediate A curve was previously provided for 22 EFPY. However, Unit 1 exceeded 22 EFPY during Cycle 14. Therefore, the A22 curve identified in Tech.

Spec. Figure 3.4.6.1-1 (Pressure/Temperature Curves) can no longer be used when performing the Reactor Vessel Pressure Test for Unit 1.

The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves C, and A, for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

specified in the PTLR specified in the PTLR include an assumed shift in RTNDT for the conditions at 57 EFPY.

LIMERICK - UNIT 1 B 3/4 4-5 Amendment No. 36,106,145,167, ECR 04-00575, Rev. 1

REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued) specified in the PTLR The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code Section XI, Appendix G. The curves are based on the RTNDT and stress intensity factor information for the reactor vessel components. Fracture toughness limits and the basis for compliance are more fully discussed in FSAR Chapter 5, Para-graph 5.3.1.5, "Fracture Toughness." specified in the PTLR The reactor vessel materials have been tested to determine their initial RTNDT. The results of these tests are shown in Table B 3/4.4.6-1. Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material the PTLR in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda-tions of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials." The pressure/temperature limit curve, Figure 3.4.6.1-1, curves A, B and C, includes an assumed shift in RTNDT for the conditions at 32 EFPY. In addition, an intermediate A curve was previously provided for 22 EFPY. However, the accumulated EFPY for Unit 2 will exceed 22 EFPY during Cycle 13 for Unit 2. Therefore, the A22 curve identified in Tech. Spec. Figure 3.4.6.1-1 (Pressure/Temperature Curves) can no longer be used when performing the Reactor Vessel Pressure Test for Unit 2. The A, B and C limit curves are predicted to be bounding for all areas of the RPV until 32 EFPY.

The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves C, and A, for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

specified in the PTLR curves specified in the PTLR include an assumed shift in RTNDT for the conditions at 57 EFPY.

LIMERICK - UNIT 2 B 3/4 4-5 Amendment No. 51,80,111,130, ECR 04-00575, Rev. 1

BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS*

HEAT/SLAB MIN.UPPER BELTLINE WELD SEAM I.D. OR STARTING SHELF COMPONENT OR MAT'L TYPE HEAT/LOT CU (%) Ni (%) RTNDT (°F) RTNDT **(°F) (LFT-LBS) ART (°F)

Plate SA-533 Gr. B,CL. 1 C 7677-1 .11 .5 +20 +35 NA +89 Weld AB (Field Weld) 640892/ .09 1.0 -60 +58 NA +54 J424B27AE NOTES:

• Based on 110% of original rated power.

• • These values are given only for the benefit of calculating the end-of-life (EOL/32 EFPY) RTNDT NON-BELTLINE MT'L TYPE OR HEAT/SLAB OR HIGHEST STARTING COMPONENT WELD SEAM I.D. HEAT/LOT RTNDT (°F)

Shell Ring SA 533, Gr. B, CL. 1 C7711-1 +20 Bottom Head Dome " C7973-1 +12 Bottom Head Torus " C7973-1 +12 Top Head Dome " A6834-1 +10 Top Head Torus " B1993-1 +10 Top Head Flange SA-508, CL. 2 123B195-289 +10 Vessel Flange " 2V1924-302 -20 Feedwater Nozzle " Q2Q22W-412 -20 Weld Non-Beltline All -12 LPCI Nozzle*** SA-508, CL. 2 Q2Q25W -6 Closure Studs SA-540, Gr. B-24 All Meet requirements of 45 ft-lbs and 25 mils Lat. Exp. at +10°F Note: *** The design of the LPCI nozzles results in their experiencing an EOL fluence in excess of 1017 N/Cm2 which predicts an EOL (32 EFPY) RTNDT of +41°F.

INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT 1 B 3/4 4-7 Amendment No. 36, 106 145

BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS*

LIMITING HEAT/SLAB MIN.UPPER BELTLINE WELD SEAM I.D. OR STARTING SHELF COMPONENT OR MAT'L TYPE HEAT/LOT CU (%) Ni (%) RTNDT (°F) RTNDT **(°F) (LFT-LBS) ART (°F)

Plate SA-533 Gr. B,CL. 1 B 3416-1 .14 .65 +40 +48 NA +122 Weld AB (Field Weld) 640892/ .09 1.0 -60 +58 NA +54 J424B27AE NOTES:

• Based on 110% of original power.

• • These values are given only for the benefit of calculating the end-of-life (EOL/32 EFPY) RTNDT NON-BELTLINE MT'L TYPE OR HEAT/SLAB OR HIGHEST STARTING COMPONENT WELD SEAM I.D. HEAT/LOT RTNDT (°F)

Top Shell Ring SA 533, Gr. B, CL. 1 C9800-2 -16 Bottom Head Dome " C9245-2 +22 Bottom Head Torus " C9362-2 +28 Top Head Torus " C9646-2 -20 Top Head Flange SA-508, CL. 2 123B300 +10 Vessel Flange " 2L2058 +10 Feedwater Nozzle " Q2Q29W 0 Weld Non-Beltline All -12 LPCI Nozzle*** SA-508, CL.

INFORMATION 2 CONTAINED ON THIS PAGEQ2Q33W HAS BEEN DELETED -4 Closure Studs SA-540, Gr. B-24 All Meet requirements of 45 ft-lbs and 25 mils Lat. Exp. at +10°F

• • • The design of the LPCI nozzles results in their experiencing an EOL fluence in excess of 1017 N/Cm2 which predicts an EOL (32 EFPY) RTNDT of +35°F.

LIMERICK - UNIT 2 B 3/4 4-7 Amendment No. 51, 111

INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED BASES FIGURE B 3/4.4.6-1 FAST NEUTRON FLUENCE (E>1 MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE*

• At 90% of Rated Thermal Power and 90% availability LIMERICK - UNIT 1 B 3/4 4-8 Amendment No. 33, 106

BASES FIGURE B 3/4.4.6-1 FAST NEUTRON FLUENCE (E> 1 MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE*

INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED

• At 90% of Rated Thermal Power and 90% availability LIMERICK - UNIT 2 B 3/4 4-8 Amendment No. 51

ATTACHMENT 4 TransWare Enterprises Inc. Affidavits

Affidavit I, Kathleen A. Jones, state as follows:

1. I am the Chief Operating Officer of TransWare Enterprises Inc. ("TWE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

2. The information sought to be withheld is contained in the attachment TransWare Enterprises Inc. Document No. LIM-FLU-002-R-005, Revision 1, "Limerick Generating Station Unit 1 Reactor Pressure Vessel Fluence Evaluation - Licensing Report," September 2020. TWE proprietary information is indicated by enclosing it in double brackets and highlighting the proprietary text in blue. Paragraph 3 of this affidavit provides the basis for the proprietary determination.

3. In making this application for withholding of proprietary information of which it is the owner or licensee, TWE relies upon the exemption of disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec.

1905, and the NRC regulations 10CFR9.l 7(a)(4) and 2.390(a)(4) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential and commercial information," and some portions also qualify under the narrower definition of "trade secret," within the meanings assigned to those terms for purposes ofFOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA. 704F2d1280 (DC Cir. 1983).

4. Some examples of categories of information that fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by TWE's competitors without license from TWE constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, could reduce the competitor's expenditure of resources or improve competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information that reveals cost or price information, production capacities, budget levels, or commercial strategies of TWE, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future TWE customer-funded development plans and programs of potential commercial value to TWE;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4a. and 4b., above.

5. To address 10CFR2.390 (b)(4), the information sought to be withheld is being submitted to the NRC in confidence. The information is of a sort customarily held in confidence by TWE, and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs 6 and 7 following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by TWE, no public disclosure has been made, and it is not available to public sources. All disclosures to third parties including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

6. Initial approval of proprietary treatment of a document is made by the manner of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to TWE. Access to such documents within TWE is limited on a "need-to-know" basis.

7. The procedure for approval of external release of such a document typically requires review by the project manager, principal engineer, and by the Quality Assurance department for technical content, competitive effect, and the determination of the accuracy of the proprietary designation. Disclosures outside TWE are limited to regulatory bodies, customers, and potential customers and their agents, suppliers, and licensees, and others with a legitimate need for the information and then only in accordance with appropriate regulatory provisions or proprietary agreements.

8. The information identified in paragraph 2 is classified as proprietary because it contains details of TWE' s methodologies for fluence and uncertainty analyses.

The development of the methods used in these analyses, along with the testing, development, and approval of the supporting methodology was achieved at a significant cost, on the order of several million dollars, to TWE or its licensor.

9. Public disclosure of the information sought to be withheld is likely to cause substantial harm to TWE's competitive position and foreclose or reduce the availability of profit-making opportunities. The methodologies for fluence and uncertainty analyses are part of TWE's nuclear engineering consulting base expertise and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by TWE or its licensor.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it is clearly substantial.

TWE's competitive advantage will be lost if its competitors are able to use the results of the TWE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to TWE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall and deprive TWE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed at Sycamore, Illinois, this 17th day of September, 2020.

~~~AO~

KathleehA.Jot;es Trans Ware Enterprises Inc.

11 QFFICIAL SEAL" DAMARIS KEEF NOTARY PUBLIC. STATE OF ILLINOIS MY COMMISSION EXPIRES 8/3/2021

~~

ntjl._ ~~o~a

Affidavit I, Kathleen A. Jones, state as follows:

1. I am the Chief Operating Officer of Trans Ware Enterprises Inc. ("TWE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

2. The information sought to be withheld is contained in the attachment Trans Ware Enterprises Inc. Document No. LIM-FLU-002-R-006, Revision 1, "Limerick Generating Station Unit 2 Reactor Pressure Vessel Fluence Evaluation - Licensing Report," September 2020. TWE proprietary information is indicated by enclosing it in double brackets and highlighting the proprietary text in blue. Paragraph 3 of this affidavit provides the basis for the proprietary determination.

3. In making this application for withholding of proprietary information of which it is the owner or licensee, TWE relies upon the exemption of disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec.

1905, and the NRC regulations 10CFR9.17(a)(4) and 2.390(a)(4) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential and commercial information," and some portions also qualify under the narrower definition of "trade secret," within the meanings assigned to those terms for purposes ofFOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA. 704F2d1280 (DC Cir. 1983).

4. Some examples of categories of information that fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by TWE's competitors without license from TWE constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, could reduce the competitor's expenditure of resources or improve competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information that reveals cost or price information, production capacities, budget levels, or commercial strategies of TWE, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future TWE customer-funded development plans and programs of potential commercial value to TWE;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4a. and 4b., above.

5. To address 10CFR2.390 (b)(4), the information sought to be withheld is being submitted to the NRC in confidence. The information is of a sort customarily held in confidence by TWE, and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs 6 and 7 following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by TWE, no public disclosure has been made, and it is not available to public sources. All disclosures to third parties including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

6. Initial approval of proprietary treatment of a document is made by the manner of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to TWE. Access to such documents within TWE is limited on a "need-to-know" basis.

7. The procedure for approval of external release of such a document typically requires review by the project manager, principal engineer, and by the Quality Assurance department for technical content, competitive effect, and the determination of the accuracy of the proprietary designation. Disclosures outside TWE are limited to regulatory bodies, customers, and potential customers and their agents, suppliers, and licensees, and others with a legitimate need for the information and then only in accordance with appropriate regulatory provisions or proprietary agreements.

8. The information identified in paragraph 2 is classified as proprietary because it contains details of TWE' s methodologies for fluence and uncertainty analyses.

The development of the methods used in these analyses, along with the testing, development, and approval of the supporting methodology was achieved at a significant cost, on the order of several million dollars, to TWE or its licensor.

9. Public disclosure of the information sought to be withheld is likely to cause substantial harm to TWE' s competitive position and foreclose or reduce the availability of profit-making opportunities. The methodologies for fluence and uncertainty analyses are part of TWE's nuclear engineering consulting base expertise and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by TWE or its licensor.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it is clearly substantial.

TWE' s competitive advantage will be lost if its competitors are able to use the results of the TWE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to TWE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall and deprive TWE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed at Sycamore, Illinois, this 17th day of September, 2020.

~~i\afJyy--,

KathitenA. Jones TransWare Enterprises Inc.

ATTACHMENT 5 Electric Power Research Institute (EPRI) Affidavit

BWRVIP 2020-069, Attachment 1 EI~121 -

1 ELECTRIC POWER RESEARCH INSTITUTE Ref EPRI Docket No. 99902016 August26,2020 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Request for Withholding of the following Proprietary Information Included in:

Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRG Docket Nos. 50-352 and 50-353 License Amendment Request - Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Reports To Whom It May Concern:

This is a request under 10 C.F.R. §2.390(a)(4) that the U.S. Nuclear Regulatory Commission ("NRG")

withhold from public disclosure the report identified in the enclosed Affidavit consisting of the proprietary information owned by Electric Power Research Institute, Inc. ("EPRI") identified above in the attached report. Proprietary and non-proprietary versions of the Report and the Affidavit in support of this request are enclosed.

EPRI desires to disclose the Proprietary Information in confidence to assist the NRG review of the enclosed submittal to the NRG by Exelon. The Proprietary Information is not to be divulged to anyone outside of the NRG or to any of its contractors, nor shall any copies be made of the Proprietary Information provided herein. EPRI welcomes any discussions and/or questions relating to the information enclosed.

If you have any questions about the legal aspects of this request for withholding, please do not hesitate to contact me at (704) 595-2630. Questions on the content of the Report should be directed to Nathan Palm of EPRI at (724) 288-4043.

Sincerely, d-'{~

Attachment(s)

To9ether . .. Shaping the Fu ture of El ectr ici ty 1300 West W.T. Horris lloulevord, Chorlo11e, NC 28262*8550 USA

• 704.595.2732

• Mobile 704.490.2653

• n\#ilmshurit@epri.com

~~121 1 ELECTRIC POWER

~*- RESEARCH INSTITUTE AFFIDAVIT RE: Request for Withholding of the Following Proprietary Information Included In:

Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRG Docket Nos. 50-352 and 50-353 License Amendment Request - Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Reports I, Steven Swilley, being duly sworn, depose and state as follows:

I am the Senior Director and Deputy Chief Nuclear Officer at Electric Power Research Institute, Inc.

whose principal office is located at 3420 Hillview Avenue, Palo Alto, California ("EPRI") and I have been specifically delegated responsibility for the above-listed Report which contains EPRI Proprietary Information that is sought under this Affidavit to be withheld "Proprietary Information". I am authorized to apply to the U.S. Nuclear Regulatory Commission ("NRG") for the withholding of the Proprietary Information on behalf of EPRI.

EPRI Proprietary Information is identified in the above referenced report with text inside double brackets. Examples of such identification is as follows:

((This sentence is an example(El))

Tables containing EPRI Proprietary Information are identified with double brackets before and after the object. In each case the superscript notation (Elrefers to this affidavit and all the bases included below, which provide the reasons for the proprietary determination.

EPRI requests that the Proprietary Information be withheld from the public on the following bases:

Withholding Based Upon Privileged And Confidential Trade Secrets Or Commercial Or Financial Information (see e.g. 10 C.F.R. §2.390(a)(4)): :

a. The Proprietary Information is owned by EPRI and has been held in confidence by EPRI. All entities accepting copies of the Proprietary Information do so subject to written agreements imposing an obligation upon the recipient to maintain the confidentiality of the Proprietary Information. The Proprietary Information is disclosed only to parties who agree, in writing, to preserve the confidentiality thereof.

b. EPRI considers the Proprietary Information contained therein to constitute trade secrets of EPRI. As such, EPRI holds the information in confidence and disclosure thereof is strictly limited to individuals and entities who have agreed, in writing, to maintain the confidentiality of the Information.

c. The information sought to be withheld is considered to be proprietary for the following reasons. EPRI made a substantial economic investment to develop the Proprietary Information and, by prohibiting public disclosure, EPRI derives an economic benefit in the form of licensing royalties and other additional fees from the confidential nature of the Proprietary Information. If the Proprietary Information were publicly available to consultants and/or other businesses providing services in the electric and/or nuclear power industry, they would be able to use the Proprietary Information for their own commercial benefit and profit and without expending the substantial economic resources required of EPRI to develop the Proprietary Information.

d. EPRl's classification of the Proprietary Information as trade secrets is justified by the Uniform Trade Secrets Act which California adopted in 1984 and a version of which has been adopted by over forty states. The California Uniform Trade Secrets Act, California Civil Code §§3426 - 3426.11, defines a "trade secret" as follows:

"'Trade secret' means information, including a formula, pattern, compilation, program device, method, technique, or process, that:

(1) Derives independent economic value, actual or potential, from not being generally known to the public or to other persons who can obtain economic value from its disclosure or use; and (2) Is the subject of efforts that are reasonable under the circumstances to maintain its secrecy."

e. The Proprietary Information contained therein are not generally known or available to the public. EPRI developed the Information only after making a determination that the Proprietary Information was not available from public sources. EPRI made a substantial investment of both money and employee hours in the development of the Propretary Information. EPRI was required to devote these resources and effort to derive the Proprietary Information. As a result of such effort and cost, both in terms of dollars spent and dedicated employee time, the Proprietary Information is highly valuable to EPRI.

f. A public disclosure of the Proprietary Information would be highly likely to cause substantial harm to EPRl's competitive position and the ability of EPRI to license the Proprietary Information both domestically and internationally. The Proprietary Information and Report can only be acquired and/or duplicated by others using an equivalent investment of time and effort.

I have read the foregoing and the matters stated herein are true and correct to the best of my knowledge, information and belief. I make this affidavit under penalty of perjury under the laws of the United States of America and under the laws of the State of North Carolina.

Executed at 1300 W WT Harris Blvd, Charlotte, NC being the premises and place of business of Electric Power Research Institute, Inc.

Date: _ _._d. . . .:-. .i.vJ_,_1_

_ : .~ _R_~_),_.__2_0_'2._o_ _ __

,4---7 ~'\

Steven Swilley

(State of North Carolina)

(County of Mecklenburg)

My Commission Expires _L,lday of

ATTACHMENT 6 Limerick Generating Station Unit 1 Pressure and Temperature Limits Report (PTLR) for 57 Effective Full-Power Years (EFPY)

(Non-Proprietary Version)

Non-Confidential Information Submitted Under 10 CFR 2.390 Exelon Corporation Limerick Generating Station Unit 1 Pressure and Temperature Limits Report (PTLR) for 57 Effective Full-Power Years (EFPY)

Revision 0-NP

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 2 of 37 Table of Contents Section Page 1.0 Purpose 3 2.0 Applicability 3 3.0 Methodology 4 4.0 Operating Limits 5 5.0 Discussion 7 6.0 References 14 Figure 1 LGS Unit 1 P-T Curve A (Hydrostatic Pressure and Leak Tests) for 57 17 EFPY Figure 2 LGS Unit 1 P-T Curve B (Normal Operation - Core Not Critical) for 57 18 EFPY Figure 3 LGS Unit 1 P-T Curve C (Normal Operation - Core Critical) for 57 EFPY 19 Figure 4 LGS Unit 1 Overall Composite Curves A, B, and C, 57 EFPY 20 Figure 5 LGS Feedwater Nozzle 3-D Finite Element Model [22] 21 Figure 6 LGS LPCI Nozzle Finite Element Model [20] 22 Table 1 LGS Unit 1 Pressure Test (Curve A) P-T Curves for 57 EFPY 23 Table 2 LGS Unit 1 Core Not Critical (Curve B) P-T Curves for 57 EFPY 25 Table 3 LGS Unit 1 Core Critical (Curve C) P-T Curves for 57 EFPY 28 Table 4 LGS Unit 1 ART Table for 57 EFPY 31 Table 5 Nozzle Stress Intensity Factors 34 Table 6 LGS Unit 1 P-T Curve Input Parameters 34 Appendix A Limerick Unit 1 Reactor Vessel Materials Surveillance Program 36

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 3 of 37 1.0 Purpose The purpose of the Limerick Generating Station (LGS) Unit 1 Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

1. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heat-up, Cool-down and Hydrostatic/Class 1 Leak Testing;

2. RCS Heat-up and Cool-down rates;

3. RPV head flange boltup temperature limits.

This report has been prepared in accordance with the requirements of Licensing Topical Reports SIR-05-044, Revision 1-A, contained within BWROG-TP-11-022-A, Revision 1 [1], and 0900876.401, Revision 0-A, contained within BWROG-TP-11-023-A, Revision 0 [2].

2.0 Applicability This report is applicable to the LGS Unit 1 RPV for up to 57 Effective Full-Power Years (EFPY).

The following LGS Unit 1 Technical Specifications (TS) are affected by the information contained in this report:

TS 3/4.4.6 RCS Pressure/Temperature (P-T) Limits

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 4 of 37 3.0 Methodology The limits in this report were derived as follows:

1. The methodology used is in accordance with Reference [1] and Reference [2],

incorporating the NRC Safety Evaluations in References [3] and [4], respectively.

2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [5], using the RAMA computer code, as documented in Reference [6].

3. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 (RG 1.99) [7], as documented in Reference [8].

4. The pressure and temperature limits, which were calculated in accordance with Reference

[1], are documented in Reference [9].

5. This revision of the pressure and temperature limits report is to incorporate the following changes:

• Revision 0: Initial issue of PTLR.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59 [11], provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPV or other plant design assumption modifications in the UFSAR, cannot be made without prior NRC approval. Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 5 of 37 4.0 Operating Limits The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operating limits for pressure and temperature are required for three categories of operation:

(a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 57 EFPY for LGS Unit 1 as documented in Reference

[9], and are provided in Figure 1 through Figure 4. A tabulation of the curves is included in Table 1 through Table 3. The adjusted reference temperature (ART) table for 57 EFPY for the LGS Unit 1 vessel beltline materials is shown in Table 4 [8]. Inputs to the P-T curves are summarized in Table 5 and Table 6.

The resulting P-T curves are based on the geometry, design and materials information for the LGS Unit 1 vessel with the following conditions:

• Heat-up/Cool-down rate limit during Hydrostatic Class 1 Leak Testing (Figure 1: Curve A): 25F/hour 1 [9].

• Normal Operating Heat-up/Cool-down rate limit (Figure 2: Curve B - non-nuclear heating, and Figure 3: Curve C - nuclear heating): 100°F/hour 2 [9].

• RPV bottom head coolant temperature to RPV coolant temperature T limit during Recirculation Pump startup: 145°F [1].

1 Interpreted as the temperature change in any 1-hour period is less than or equal to 25°F.

2 Interpreted as the temperature change in any 1-hour period is less than or equal to 100°F.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 6 of 37

• Recirculation loop coolant temperature to RPV coolant temperature T limit during Recirculation Pump startup: 50°F [1].

• RPV flange and adjacent shell temperature limit 80°F [9].

Minimum temperature limits are set in accordance with 10CFR50, Appendix G [10, Table 1].

The minimum moderator temperature used in the plant shutdown margin evaluation, 68°F [14],

is also included as a minimum bolt-up temperature requirement. An additional 60°F margin above the requirements in Table 1 of 10CFR50, Appendix G, has been commonly applied in the BWR industry. For LGS Unit 1, the limiting RTNDT for the closure flange region is 20°F, and the minimum temperature is 80°F (i.e. 20°F + 60°F) [9]. For Curves A and B, this 60°F margin is a recommendation, but for Curve C, the 60°F margin is required. For consistency with prior work, the minimum temperature for Curves A, B, and C for LGS Unit 1 is set to 80°F.

These values are consistent with the minimum temperature limits and minimum bolt-up temperature in the previous docketed P-T curves [12] (approved by the NRC in Reference [13]).

These values also bound the minimum temperature in the first set of P-T limits approved for initial operation (i.e. initial licensed curves in Technical Specifications Figure 3.4.6.1-1) [14],

thereby addressing the NRC condition in Reference [3, Section 4.0].

The composite P-T curves are extended below 0 psig to -14.7 psig based on the evaluation in Reference [15], which demonstrates that the P-T curves are applicable to negative gauge pressures. A pressure of -14.7 psig bounds the maximum expected vacuum pressure as well as externally applied pressures the RPV may experience. Since the P-T curve calculation methods used do not specifically apply to negative values of pressure, the tabulated results start at 0 psig.

However, the minimum analyzed RPV pressure is -14.7 psig.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 7 of 37 5.0 Discussion 5.1 Adjusted Reference Temperatures The adjusted reference temperature (ART) of the limiting beltline material is used to adjust beltline P-T curves to account for irradiation effects. RG 1.99 [7] provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the LGS Unit 1 vessel plate, weld, and forging materials [8]. The Cu and Ni values were used with Table 1 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds.

The Cu and Ni values were used with Table 2 of RG 1.99 to determine a CF per Paragraph 1.1 of RG 1.99 for plates and forgings. For materials where surveillance data exists, a fitted CF has been used in the calculation of ART for those heats, in accordance with Regulatory Position 2.1 in RG 1.99. Use of surveillance data from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) for LGS Unit 1 was approved by the NRC in Reference [13].

The peak RPV ID fluence value of 1.09 x 1018 n/cm2 at 57 EFPY used in the P-T curve evaluations was obtained from Reference [6]. Fluence values in Reference [6] were calculated in accordance with RG 1.190 [5]. These fluence values apply to the limiting beltline lower intermediate shell plates (heat no. C7677-1). A plant-specific damage assessment, in terms of displacements per atom (dpa) was performed in Reference [6] to determine through-wall fluence for LGS Unit 1, as permitted by RG 1.99. The resulting attenuation factor for beltline plates and welds is 0.67-0.69 for a postulated 1/4T flaw. Consequently, the 1/4T fluence for 57 EFPY for the limiting lower intermediate shell plate is 7.35 x 1017 n/cm2. The limiting value for ART for beltline plates and welds is 72.3°F [8].

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 8 of 37 The P-T limits are developed to bound all ferritic materials in the RPV, including the consideration of stress levels from structural discontinuities such as nozzles. LGS Unit 1 has two sets of nozzles in the RPV beltline: the instrument (N16) nozzles and the low pressure coolant injection (LPCI, N17) nozzles [8]; the N16 and N17 nozzles are in the lower-intermediate shell beltline plates. There are no additional forged or partial penetration nozzles in the extended beltline at LGS Unit 1 [8]. The feedwater (FW) nozzle is considered in the evaluation of the non-beltline (upper vessel) region P-T limits.

The LPCI (N17) nozzles and welds have a limiting fluence at the RPV ID of 4.68 x 1017 n/cm2 at 57 EFPY, obtained from Reference [6] and calculated in accordance with RG 1.190 [5]. Similar to the RPV beltline plates and welds described above, through-wall fluence for the LPCI nozzles was attenuated using the dpa methodology in Reference [6]. The resulting attenuation factor is 0.71 for a postulated 1/4T flaw in the LPCI nozzle blend radius. Consequently, the 1/4T fluence for 57 EFPY for the limiting LPCI nozzle location is 3.30 x 1017 n/cm2. The limiting 57 EFPY ART value for the LPCI nozzles and welds is 59.9°F [8].

The instrument (N16) nozzle inserts and welds at LGS Units 1 and 2 are fabricated from non-ferritic materials and do not require evaluation for loss of fracture toughness [8]. However, the effect of the penetration on the adjacent shell is considered in the development of bounding beltline P-T limits, according to the methodology in Reference [2]. The instrument nozzles have a limiting fluence at the RPV ID of 3.34 x 1017 n/cm2 at 57 EFPY, obtained from Reference [6]

and calculated in accordance with RG 1.190 [5]. This fluence value applies to the limiting N16 nozzle location (lower-intermediate shells). Similar to the RPV beltline plates and welds described above, through-wall fluence for the N16 nozzles was attenuated using the dpa methodology in Reference [6]. The resulting attenuation factor is 0.70 for a postulated 1/4T flaw in the instrument nozzle corner. Consequently, the 1/4T fluence for 57 EFPY for the limiting instrument nozzle location is 2.33 x 1017 n/cm2. The limiting 57 EFPY ART value for the instrument nozzles is 47.6°F [8].

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 9 of 37 5.2 P-T Curve Development The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cool-down and is in the outer wall during heat-up. However, as a conservative simplification, the thermal gradient stresses at the 1/4T location are assumed to be tensile for both heat-up and cool-down. This results in the approach of applying the maximum tensile stresses at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness at the 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, and for a given pressure, the coolant saturation temperature is well above the P-T curve limiting temperature. Consequently, the material toughness at a given pressure would exceed the allowable toughness.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heat-up and cool-down temperature rate of 100F/hour for which the curves are applicable. However, the core not critical and core critical curves were also developed to bound Service Level A/B RPV thermal transients. P-T curves are developed for anticipated operational occurrences, and Technical Specifications limit operation to 100°F/hour. For the hydrostatic pressure and leak test curve (Curve A), a coolant heat-up and cool-down temperature rate of 25°F/hour must be maintained. The P-T limits and corresponding limits of either Curve A or B may be applied, if necessary, while achieving or recovering from test conditions.

So, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heat-up/cool-down rate limits cannot be maintained.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 10 of 37 5.3 Material Data The initial RTNDT, chemistry (weight-percent copper and nickel), and ART at the 1/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 1017 n/cm2 for E >

1 MeV) are shown in Table 4 [8]. Use of initial RTNDT values in the determination of P-T curves for LGS Unit 1 was approved by the NRC in Reference [16].

Per Reference [8] and in accordance with Appendix A of Reference [1], the LGS Unit 1 representative weld and plate surveillance materials data were reviewed from the BWRVIP ISP

[18]. The representative plate material heat for LGS Unit 1 (C2761-2) in the ISP is not the same as the target plate heat in LGS Unit 1 (C7677-1). The representative weld material heat for LGS Unit 1 (5P6756) is not the same heat number as the target vessel weld in LGS Unit 1 (640892).

However, the representative weld heat 5P6756 does match a weld heat in the LGS Unit 1 beltline, and there are four irradiated data sets for this weld. The data were determined to be credible, and a reduced margin term ( = 28°F/2 = 14°F) was used to calculate ART for weld heat 5P6756 in the LGS Unit 1 beltline per RG 1.99 Position 2.1 [7].

For all LGS Unit 1 beltline materials other than weld heat 5P6756, ART values are calculated using the table CF values in accordance with RG 1.99 Position 1.1 [7].

5.4 Plant-Specific Evaluation of Nozzles The only computer code used in the determination of the LGS Unit 1 P-T curves was the ANSYS finite element computer program:

• ANSYS Mechanical APDL and Workbench Release 18.1 [19] for:

o FW nozzle (non-beltline) and LPCI nozzle (beltline) thermal and pressure stress distributions in Reference [20].

ANSYS finite element analyses were used to develop the stress distributions through the FW and LPCI nozzles, and these stress distributions were used in the determination of the stress intensity factors for the FW and LPCI nozzles [20]. At the time that each of the analyses above was

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 11 of 37 performed, the ANSYS program was controlled under the vendors 10 CFR 50 Appendix B [24]

Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 [25] was performed as a part of the computer program verification by comparing the solutions produced by the computer code to hand calculations for several problems.

The plant-specific LGS Unit 1 FW nozzle analyses were performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients.

Detailed information regarding the analyses can be found in References [20-23]. The following summarizes the development of the thermal and pressure stress intensity factors for the FW nozzle:

• A one-quarter symmetric, three-dimensional (3-D) finite element model (FEM) of the FW nozzle was constructed and is shown in Figure 5. A single model was developed that represents both units, as the FW nozzle geometry is identical for both units. Details of the model are provided in Reference [22]. Temperature-dependent material properties were taken from the ASME Code,Section III, 1968 Edition with 1969 Summer Addenda and Article 4 of the 1969 Winter Addenda [26] and from the ASME Code,Section II, Part D, 2001 Edition with Addenda through 2003 [27] and are tabulated in Reference [20].

• Heat transfer coefficients were calculated in Reference [21] and are a function of FW temperature and flow rate.

• With respect to operating conditions, the bounding thermal transients during Normal and Upset operating conditions were tabulated in Reference [21] and analyzed in Reference

[20]. The thermal stress distributions, corresponding to the limiting times presented in Reference [20], along a linear path through the nozzle corner is used. The boundary integral equation/influence function (BIE/IF) methodology presented in Reference [1] is used in Reference [20] to calculate the thermal stress intensity factor, KIt, due to the thermal stresses by fitting a third order polynomial equation to the path stress distribution for the thermal load case.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 12 of 37

• With respect to pressure stress, a unit pressure of 1000 psig was applied to the internal surfaces of the 3-D model in Reference [20]. The pressure stress distribution was taken along the same path as the thermal stress distribution. The BIE/IF methodology presented in Reference [1] was used to calculate the applied pressure stress intensity factor, KIp, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting KIp may be linearly scaled to determine the KIp for various RPV internal pressures.

The plant-specific LGS Unit 1 LPCI nozzle analysis was performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients.

Detailed information regarding the analysis can be found in Reference [20]. The following summarizes the development of the thermal and pressure stress intensity factors for the LPCI nozzle:

• A one-quarter symmetric, 3-D FEM of the LPCI nozzle was constructed and is shown in Figure 6. A single model was developed that represents both units, as the LPCI nozzle geometry is identical for both units. Details of the model are provided in Reference [20].

Temperature-dependent material properties were taken from the ASME Code,Section III, 1968 Edition with 1969 Summer Addenda and Article 4 of the 1969 Winter Addenda

[26] and from the ASME Code,Section II, Part D, 2001 Edition with Addenda through 2003 [27] and are tabulated in Reference [20].

• Heat transfer coefficients were calculated in Reference [20] and are a function of LPCI temperature and flow rate.

• With respect to operating conditions, the bounding thermal transient for the region corresponding to the LPCI nozzles during normal and upset operating conditions was analyzed [20]. The thermal stress distribution, corresponding to the limiting time in Reference [20], along a linear path through the nozzle corner is used. The BIE/IF methodology presented in Reference [1] was used to calculate the thermal stress intensity factor, KIt, due to the thermal stresses by fitting a third order polynomial equation to the path stress distribution for the thermal load case.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 13 of 37

• With respect to pressure stress, a unit pressure of 1000 psig was applied to the internal surfaces of the FEM [20]. The pressure stress distribution was taken along the same path as the thermal stress distribution. The BIE/IF methodology presented in Reference [1] is used to calculate the pressure stress intensity factor, KIp, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting KIp can be linearly scaled to determine the KIp for various RPV internal pressures.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 14 of 37 6.0 References

1. BWROG-TP-11-022-A, Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors, August 2013. (ADAMS Accession No. ML13277A557)

2. BWROG-TP-11-023-A, Revision 0, Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations, May 2013. (ADAMS Accession No. ML13183A017

3. U.S. NRC Letter to BWROG dated May 16, 2013, Final Safety Evaluation for Boiling Water Reactor Owners Group Topical Report BWROG-TP-11-022, Revision 1, November 2011, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors (TAC NO. ME7649, ADAMS Accession No. ML13277A557).

4. U.S. NRC Letter to BWROG dated March 14, 2013, Final Safety Evaluation for Boiling Water Reactor Owners Group Topical Report BWROG-TP-11-023, Revision 0, November 2011, Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations (TAC NO. ME7650, ML13183A017)

5. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001.

6. TransWare Report No. LIM-FLU-002-R-001, Revision 0, Limerick Generating Station Unit 1 Reactor Pressure Vessel Fluence Evaluation, Excerpt provided in Exelon TODI No. 04139609-04, Rev. 0, dated May 29, 2020. SI File No. 1701104.201.

7. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

8. SI Calculation No. 1701104.301, Revision 1, Limerick Unit 1 and 2 RPV Beltline ART and USE Evaluation, August 18, 2020.

9. SI Calculation No. 1701104.302, Revision 1, Limerick Unit 1 and 2 P-T Curve Calculation for 57 EFPY, August 18, 2020.

10. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix G, Fracture Toughness Requirements, December 12, 2013.

11. U.S. Code of Federal Regulations, Title 10, Part 50, Section 59, Changes, tests and experiments, August 28, 2007.

12. Attachment to PECO Nuclear Letter dated May 19, 2000, GE Nuclear Energy Report No.

GE-NE-B11-00836-00-01a NP, Revision 0, Pressure-Temperature Curves For PECO Energy Company, Limerick Unit 1, April 2000 (ADAMS Accession No. ML003718786).

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 15 of 37

13. Limerick Generating Station, Unit 1, License Amendment No. 145, Issuance of Amendment Re. Update Pressure-Temperature (P-T) Limit Curves (TAC No. MA8953, ADAMS Accession No. ML003745510), September 15, 2000.

14. Exelon TODI No. 04139609-04, Revision 0, May 29, 2020. SI File No. 1701104.201.

15. SI Calculation No. 1701104.304, Revision 0, Limerick Units 1 and 2 RPV Vacuum Assessment, June 15, 2020.

16. NRC acceptance of initial RTNDT values for LGS Unit 1:

a. Letter from F. Rinaldi (USNRC) to G.A Hunger (PECO), dated April 21, 1994, Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, and Limerick Generating Station (LGS), Units 1 and 2, (TAC Nos. M83495, M83496, M83477, and M83478). (ADAMS Accession No. 9405040314). SI File No. 1701104.210.

b. Letter from F. Rinaldi (USNRC) to G.A. Hunger (PECO), dated December 9, 1996, Closeout for Philadelphia Electric Company (PECO) Response to Generic Letter 92-01, Revision 1, Supplement 1, Limerick Generating Station (LGS),

Units 1 and 2 (TAC Nos. M92691 and M92692). (ADAMS Accession No.

9612230225). SI File No. 1701104.210.

17. Limerick Generating Station, Units 1 License Amendment No. 167 and Unit 2 License Amendment No. 130, Issuance of Amendment Re. Revision to the Reactor Pressure Vessel Material Surveillance Program (TAC Nos. MB7003 and MB7004, ADAMS Accession No. ML032310540), November 4, 2003.

18. BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014.

3002003144. EPRI PROPRIETARY INFORMATION.

19. ANSYS Mechanical APDL (UP20170403) and Workbench (March 31, 2017), Release 18.1, SAS IP, Inc.

20. SI Calculation No. 1701104.303, Revision 0, Low Pressure Coolant Injection Nozzle and Feedwater Nozzle Fracture Mechanics Evaluation for Pressure-Temperature Limit Curve Development, June 17, 2020.

21. SI Calculation No. 1000818.301, Revision 0, Feedwater Nozzle FEA Loads Calculation, November 16, 2010.

22. SI Calculation No. 1000818.302, Revision 0, Feedwater Nozzle Finite Element Model, November 18, 2010.

23. SI Calculation No. 1000818.303, Revision 0, Feedwater Nozzle Stress Analysis, November 19, 2010.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 16 of 37

24. U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Appendix B, Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants.

25. U.S. Nuclear Regulatory Commission, Generic Letter 83-11, Supplement 1, License Qualification for Performing Safety Analyses, June 24, 1999.

26. ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Vessels, 1968 Edition with 1969 Summer Addenda and Article 4 of the 1969 Winter Addenda.

27. ASME Boiler and Pressure Vessel Code,Section II, Part D, Material Properties, 2001 Edition with Addenda through 2003.

28. License Renewal Application, Limerick Generating Station Units 1 and 2, Facility Operating Licenses Nos. NPF-39 and NPF-85. (ADAMS Accession No. ML11179A101)

29. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, January 31, 2008.

30. BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2012. 1025144.

EPRI PROPRIETARY INFORMATION.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 17 of 37 Figure 1: LGS Unit 1 P-T Curve A (Hydrostatic Pressure and Leak Tests) for 57 EFPY

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 18 of 37 Figure 2: LGS Unit 1 P-T Curve B (Normal Operation - Core Not Critical) for 57 EFPY

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 19 of 37 Figure 3: LGS Unit 1 P-T Curve C (Normal Operation - Core Critical) for 57 EFPY

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 20 of 37 Figure 4: LGS Unit 1 Overall Composite Curves A, B, and C, 57 EFPY

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 21 of 37 Figure 5: LGS Feedwater Nozzle 3-D Finite Element Model [22]

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 22 of 37 Figure 6: LGS LPCI Nozzle Finite Element Model [20]

With Thermal Sleeve Without Thermal Sleeve

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 23 of 37 Table 1: LGS Unit 1 Pressure Test (Curve A) P-T Curves for 57 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 80.0 -14.7 80.0 498.9 88.7 547.2 96.0 595.6 102.5 644.0 108.1 692.3 113.3 740.7 117.9 789.1 122.1 837.5 126.0 885.8 129.7 934.2 133.0 982.6 136.2 1030.9 139.2 1079.3 142.0 1127.7 144.7 1176.0 147.2 1224.4 149.6 1272.8 151.9 1321.2

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 24 of 37 Table 1: LGS Unit 1 Pressure Test (Curve A) P-T Curves for 57 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 80.0 -14.7 80.0 1042.9 83.0 1090.2 85.9 1137.5 88.5 1184.8 91.1 1232.0 93.5 1279.3 95.8 1326.6 Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 80.0 -14.7 80.0 312.6 110.0 312.6 110.0 857.9 113.7 904.9 117.2 951.9 120.4 998.9 123.5 1045.9 126.3 1092.9 129.1 1139.9 131.6 1186.9 134.1 1233.9 136.4 1280.9 138.6 1328.0

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 25 of 37 Table 2: LGS Unit 1 Core Not Critical (Curve B) P-T Curves for 57 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 82.5 -14.7 82.5 0.0 93.3 48.9 102.2 97.7 109.8 146.5 116.3 195.4 122.1 244.2 127.3 293.1 132.0 341.9 136.3 390.8 140.3 439.6 144.0 488.4 147.4 537.3 150.6 586.1 153.6 635.0 156.4 683.8 159.1 732.7 161.7 781.5 164.1 830.4 166.4 879.2 168.6 928.0 170.7 976.9 172.8 1025.7 174.7 1074.6 176.6 1123.4 178.4 1172.3 180.1 1221.1 181.8 1269.9 183.5 1318.8

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 26 of 37 Table 2: LGS Unit 1 Core Not Critical (Curve B) P-T Curves for 57 EFPY (continued)

Bottom Head Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 80.0 -14.7 80.0 710.1 84.0 757.5 87.7 804.9 91.1 852.3 94.3 899.6 97.4 947.0 100.2 994.4 102.9 1041.8 105.5 1089.2 107.9 1136.6 110.2 1183.9 112.4 1231.3 114.5 1278.7 116.6 1326.1

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 27 of 37 Table 2: LGS Unit 1 Core Not Critical (Curve B) P-T Curves for 57 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 80.0 -14.7 80.0 240.3 86.7 276.5 92.7 312.6 140.0 312.6 140.0 820.7 142.9 870.2 145.6 919.7 148.2 969.1 150.7 1018.6 153.0 1068.1 155.3 1117.6 157.4 1167.1 159.5 1216.6 161.5 1266.1 163.4 1315.6

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 28 of 37 Table 3: LGS Unit 1 Core Critical (Curve C) P-T Curves for 57 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 122.5 -14.7 122.5 0.0 133.3 48.9 142.2 97.7 149.8 146.5 156.3 195.4 162.1 244.2 167.3 293.1 172.0 341.9 176.3 390.8 180.3 439.6 184.0 488.4 187.4 537.3 190.6 586.1 193.6 635.0 196.4 683.8 199.1 732.7 201.7 781.5 204.1 830.4 206.4 879.2 208.6 928.0 210.7 976.9 212.8 1025.7 214.7 1074.6 216.6 1123.4 218.4 1172.3 220.1 1221.1 221.8 1269.9 223.5 1318.8

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 29 of 37 Table 3: LGS Unit 1 Core Critical (Curve C) P-T Curves for 57 EFPY (continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 80.0 -14.7 80.0 395.6 88.7 444.3 96.1 492.9 102.5 541.6 108.2 590.2 113.3 638.8 118.0 687.5 122.2 736.1 126.1 784.8 129.8 833.4 133.2 882.0 136.3 930.7 139.3 979.3 142.1 1028.0 144.8 1076.6 147.3 1125.2 149.7 1173.9 152.0 1222.5 154.2 1271.2 156.3 1319.8

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 30 of 37 Table 3: LGS Unit 1 Core Critical (Curve C) P-T Curves for 57 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 80.0 -14.7 80.0 102.6 95.9 144.6 107.9 186.6 117.6 228.6 125.7 270.6 132.7 312.6 180.0 312.6 180.0 820.7 182.9 870.2 185.6 919.7 188.2 969.1 190.7 1018.6 193.0 1068.1 195.3 1117.6 197.4 1167.1 199.5 1216.6 201.5 1266.1 203.4 1315.6

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 31 of 37 Table 4: LGS Unit 1 ART Table for 57 EFPY Initial Chemistry Fluence Fluence Adjustments For 1/4t Chemistry Fluence RTNDT (wt%) (1) at ID at 1/4T Description ID Heat No. Lot No. Factor, CF Factor, RTNDT Margin Terms ART

(°F) (n/cm2) (n/cm2)

(°F) FF (°F) (°F)

(1) Cu Ni (2) (2) (°F) i (°F)

Lower Shell #1 14-1 C7688-1 - 10 0.12 0.51 81 8.08E+17 5.50E+17 0.308 25.0 12.5 0.0 59.9 Lower Shell #1 14-2 C7698-2 - 10 0.11 0.48 73 8.08E+17 5.50E+17 0.308 22.5 11.2 0.0 55.0 Plates Lower Shell #1 14-3 C7688-2 - 10 0.12 0.51 81 8.08E+17 5.50E+17 0.308 25.0 12.5 0.0 59.9 Lower-Int Shell #2 17-1 C7689-1 - 10 0.11 0.48 73 1.09E+18 7.35E+17 0.358 26.1 13.1 0.0 62.3 Lower-Int Shell #2 17-2 C7677-1 - 20 0.11 0.50 73 1.09E+18 7.35E+17 0.358 26.1 13.1 0.0 72.3 Lower-Int Shell #2 17-3 C7698-1 - 10 0.11 0.48 73 1.09E+18 7.35E+17 0.358 26.1 13.1 0.0 62.3 Axial (Lower-Int Shell) BE 411A3531 H004A27A -50 0.02 0.96 27 8.88E+17 6.05E+17 0.324 8.7 4.4 0.0 -32.5 Axial (Lower Shell) BA 06L165 F017A27A -50 0.03 0.99 41 5.77E+17 3.96E+17 0.257 10.6 5.3 0.0 -28.9 Axial (Lower Shell) BB 06L165 F017A27A -50 0.03 0.99 41 5.80E+17 3.97E+17 0.258 10.6 5.3 0.0 -28.9 Axial (Lower-Int Shell) BD 06L165 F017A27A -50 0.03 0.99 41 8.89E+17 6.05E+17 0.324 13.3 6.6 0.0 -23.4 Axial (Lower-Int Shell) BF 06L165 F017A27A -50 0.03 0.99 41 7.90E+17 5.41E+17 0.305 12.5 6.3 0.0 -25.0 Axial (Lower Shell) BA 662A746 H013A27A -20 0.03 0.88 41 5.77E+17 3.96E+17 0.257 10.6 5.3 0.0 1.1 Axial (Lower-Int Shell) BD 662A746 H013A27A -20 0.03 0.88 41 8.89E+17 6.05E+17 0.324 13.3 6.6 0.0 6.6 Axial (Lower-Int Shell) BE 662A746 H013A27A -20 0.03 0.88 41 8.88E+17 6.05E+17 0.324 13.3 6.6 0.0 6.6 Axial (Lower-Int Shell) BF 662A746 H013A27A -20 0.03 0.88 41 7.90E+17 5.41E+17 0.305 12.5 6.3 0.0 5.0 Axial (Lower Shell) BA (3) 3P4000 3932-989 -50 (( . (E)

)) 27 5.77E+17 3.96E+17 0.257 7.0 3.5 0.0 -36.1 Welds Axial (Lower Shell) BB (3) 3P4000 3932-989 -50 (( . (E)

)) 27 5.80E+17 3.97E+17 0.258 7.0 3.5 0.0 -36.1 Axial (Lower Shell) BC (3) 3P4000 3932-989 -50 (( . (E)

)) 27 7.32E+17 4.99E+17 0.292 7.9 3.9 0.0 -34.2 Axial (Lower-Int Shell) BF (3) S3986 RUN934 -42 (( . (E)

)) 79 7.90E+17 5.41E+17 0.305 24.1 12.1 0.0 6.2 Axial (Lower Shell) BA (3) 1P4218 3929-989 -50 (( . (E)

)) 79 5.77E+17 3.96E+17 0.257 20.3 10.2 0.0 -9.3 Axial (Lower Shell) BB (3) 1P4218 3929-989 -50 (( . (E)

)) 79 5.80E+17 3.97E+17 0.258 20.4 10.2 0.0 -9.3 Axial (Lower-Int Shell) BE (3) 1P4218 3929-989 -50 (( . (E)

)) 79 8.88E+17 6.05E+17 0.324 25.6 12.8 0.0 1.2 Axial (Test Plate) (1) N/A 421A6811 F022A27A -50 0.09 0.81 122 1.09E+18 7.35E+17 0.358 43.7 21.8 0.0 37.4 Girth (Lower to Lower-Int) AB 07L857 B101A27A -6 0.03 0.97 41 8.08E+17 5.50E+17 0.308 12.6 6.3 0.0 19.3 Girth (Lower to Lower-Int) AB 402C4371 C115A27A -50 0.02 0.92 27 8.08E+17 5.50E+17 0.308 8.3 4.2 0.0 -33.4 Girth (Lower to Lower-Int) AB 411A3531 H004A27A -50 0.02 0.96 27 8.08E+17 5.50E+17 0.308 8.3 4.2 0.0 -33.4 Girth (Lower to Lower-Int) AB 09M057 C109A27A -36 0.03 0.89 41 8.08E+17 5.50E+17 0.308 12.6 6.3 0.0 -10.7 Girth (Lower to Lower-Int) AB 412P3611 J417B27AF -80 0.03 0.93 41 8.08E+17 5.50E+17 0.308 12.6 6.3 0.0 -54.7

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 32 of 37 Table 4. LGS Unit 1 ART Table for 57 EFPY (continued)

Adjustments For 1/4t Initial Chemistry (wt%) Fluence at Chem. Fluence at Fluence Margin RTNDT (1) 1/4T Description ID Heat No. Lot No. Factor, ID (n/cm2) Factor, RTNDT Terms ART

(°F) (n/cm2)

CF (°F) (2) FF (°F) i (°F)

(1) Cu Ni (2)

(°F) (°F)

Girth (Lower to Lower-Int) AB 03M014 C118A27A -34 0.01 0.94 20 8.08E+17 5.50E+17 0.308 6.2 3.1 0.0 -21.7 Welds, contd.

Girth (Lower to Lower-Int) AB L83355 S411B27AD -70 0.03 1.08 41 8.08E+17 5.50E+17 0.308 12.6 6.3 0.0 -44.7 Girth (Lower to Lower-Int) AB 640892 J424B27AE -60 0.09 1.00 122 8.08E+17 5.50E+17 0.308 37.6 18.8 0.0 15.2 Girth (Lower to Lower-Int) AB 401P6741 S419B27AG -60 0.03 0.92 41 8.08E+17 5.50E+17 0.308 12.6 6.3 0.0 -34.7 Girth (Lower to Lower-Int) AB (3) 5P6756 N/A -60 (( . (E)

)) 108 8.08E+17 5.50E+17 0.308 33.3 16.6 0.0 6.5 LPCI (N17) N17-45° Q2Q25W - -6 0.18 0.85 142 4.68E+17 3.30E+17 0.232 33.0 16.5 0.0 59.9 LPCI (N17) N17-135° Q2Q35W - -8 0.18 0.78 140 4.68E+17 3.30E+17 0.232 32.5 16.2 0.0 57.0 LPCI (N17) N17-225° Q2Q25W - -6 0.18 0.85 142 4.68E+17 3.30E+17 0.232 33.0 16.5 0.0 59.9 Nozzles LPCI (N17) N17-315° Q2Q35W - -8 0.18 0.78 140 4.68E+17 3.30E+17 0.232 32.5 16.2 0.0 57.0 N16-0°,

100°,

Instrument (N16) (4) (SB-166) - 20 0.11 0.50 73.0 3.34E+17 2.33E+17 0.189 13.8 6.9 0.0 47.6 200°,

280° LPCI Nozzle-to-shell KA 411A3531 H004A27A -50 0.02 0.96 27 4.68E+17 3.30E+17 0.232 6.3 3.1 0.0 -37.5 Nozzle Welds LPCI Nozzle-to-shell KA 662A746 H013A27A -20 0.03 0.88 41 4.68E+17 3.30E+17 0.232 9.5 4.8 0.0 -1.0 LPCI Nozzle-to-shell KA (3) 3P4000 3932-989 -50 (( . (E)

)) 27 4.68E+17 3.30E+17 0.232 6.3 3.1 0.0 -37.5 LPCI Nozzle-to-shell KA (3) S3986 RUN934 -42 (( . (E)

)) 79 4.68E+17 3.30E+17 0.232 18.3 9.2 0.0 -5.3 LPCI Nozzle-to-shell KA 07L669 K004A27A -50 0.03 1.02 41 4.68E+17 3.30E+17 0.232 9.5 4.8 0.0 -31.0 LPCI Nozzle-to-shell KA 401Z9711 A022A27A -50 0.02 0.83 27 4.68E+17 3.30E+17 0.232 6.3 3.1 0.0 -37.5 ISP Surveillance Weld (5) - 5P6756 - -60 0.06 0.93 (( (E)

)) 8.08E+17 5.50E+17 0.308 47.4 14.0 0.0 15.4 Notes:

1. Initial RTNDT and chemistry data for as-fabricated RPV materials are obtained from Tables 4.2.3-1 and 4.2.3-2 of Reference [28], unless otherwise noted.

The test late reported in Table 4.2.3-1 of Reference [28] is included and assumed to have fluence equal to the maximum fluence plates.

2. Fluence values for 57 EFPY are obtained from Reference [6]. Fluence values for nozzles are reported for the 1/4T location along the nozzle extraction path, based on a plant-specific damage assessment (i.e. dpa) methodology.

3. For the noted vessel welds, best-estimate chemistry values from Appendix D of BWRVIP-135, Rev. 3 [18], are used and assumed to supersede original plant-specific chemistry values.

4. The N16 nozzle inserts and welds are fabricated from non-ferritic material and do not require evaluation for loss of fracture toughness. The corresponding ART value applies to the penetration in the surrounding shell, based on the initial RTNDT and CF corresponding to the limiting adjacent lower-intermediate shell plate and the fluence corresponding to the limiting nozzle location.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 33 of 37

5. The ISP representative weld heat is not the same heat as the target vessel weld but does exist in the LGS Unit 1 beltline. A separate ART calculation is performed for the surveillance weld 5P6756 using the most recent ISP surveillance data. The fitted CF is obtained from Table B-11-6 in BWRVIP-135, Rev. 3 [18]. The fitted CF is adjusted to account for differences in chemistry between the surveillance weld and vessel weld, per Eq. 3-5 in BWRVIP-135, Rev. 3 [18]. Best-estimate chemistry data for the surveillance weld is provided in Table B-11-2 in BWRVIP-135, Rev. 3 [18], and is used to calculate ART for the surveillance weld, in preference to the industry-wide best-estimate chemistry data for the vessel weld from Table D-1 in BWRVIP-135, Rev. 3 (the latter is used to calculate ART for the vessel weld). The adjusted CF is: CFadj = (Table CFvessel / Table CFsurv) * (CFFitted) = (108 / 82) * (( (E))) = (( (E))).

Because the adjusted surveillance CF is higher than the table CF, the adjusted surveillance CF must be used in the ART calculation. Because the data are credible, a reduced margin term ( = 28/2 = 14°F) is used for the surveillance weld according to Regulatory Position 2.1 in RG1.99 [7].

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 34 of 37 Table 5: Nozzle Stress Intensity Factors Nozzle Applied Pressure, KIp-app Thermal, KIt Feedwater 78.6 29.8 LPCI (N17) 80.7 60.7 Instrument (N16) 86.3 26.2 KI in units of ksi-in0.5 Table 6: LGS Unit 1 P-T Curve Input Parameters General Parameters Values Unit System for Tables and Plots English Temperature Instrument Uncertainty Adjustment (°F) 0 Pressure Instrument Uncertainty Adjustment (psig) 0 Water Density (lbm/ft3) 62.4 Full-Vessel Water Height (in) 870.5 Safety Factor for Curve A 1.5 Safety Factor for Curves B and C 2 Bolt-up Temperature (°F) 80 ART of Closure Flange Region (°F) 20 Starting Pressure for Curves (psig) -14.7 Atmospheric Pressure Adjustment (psi) 14.7 Preservice hydrotest pressure (psi) 1563 In-service hydrotest pressure (psi) 1150 Minimum in-service hydrotest temperature (°F) 144 Beltline Parameters Values Adjusted Reference Temperature (°F) 72.3 Vessel Radius (in) 126.69 Vessel Thickness (in) 6.19 Heat-up / Cool-down Rate (°F/hr) 100

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 35 of 37 Table 6: LGS Unit 1 P-T Curve Input Parameters (continued)

Instrument (N16) Nozzle Parameters Values Adjusted Reference Temperature (°F) 47.6 Vessel Radius (in) 126.69 Vessel Thickness (in) 6.19 Nozzle Outer Diameter (in) 3.5 Nozzle Inner Diameter (in) 1.938 Coefficient of Thermal Expansion (in/in/°F) 7.70E-06 Reference Pressure for Thermal Transient (psig) 1000 LPCI (N17) Nozzle Parameters Values Adjusted Reference Temperature (°F) 59.9 Applied Pressure Stress Intensity Factor (ksi*in^0.5) 80.7 Applied Thermal Stress Intensity Factor (ksi*in^0.5) 60.7 Scale KIT based on Saturation Temperature? No Reference Pressure for Thermal Transient (psig) 1000 Bottom Head Parameters Values Adjusted Reference Temperature (°F) 12 Vessel Radius (in) 126.69 Vessel Thickness (in) 6.19 Heat-up / Cool-down Rate (°F/hr) 100 Stress Concentration Factor 3 Upper Vessel (Feedwater Nozzle) Parameters Values Adjusted Reference Temperature (°F) 48 Applied Pressure Stress Intensity Factor (ksi*in^0.5) 78.6 Applied Thermal Stress Intensity Factor (ksi*in^0.5) 29.8 Minimum Thermal Stress Intensity Factor (ksi*in^0.5) 0 Scale KIT based on Saturation Temperature? No Reference Pressure for Thermal Transient (psig) 1000

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 36 of 37 Appendix A LIMERICK REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements [29], three surveillance capsules are installed in the LGS Unit 1 reactor vessel. The surveillance capsules contain flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region. None of the LGS Unit 1 surveillance capsules have been withdrawn from the reactor vessel, and all remain on standby for use as backup capsules by the BWRVIP Integrated Surveillance Program (ISP), or as otherwise needed.

LGS Unit 1 has replaced the original RPV material surveillance program with the BWRVIP ISP

[30]. LGS Unit 1 is currently committed to use the BWRVIP ISP and has made a licensing commitment to use the ISP for LGS Unit 1 during the period of extended operation. The BWRVIP ISP meets the requirements of 10 CFR 50, Appendix H, for Integrated Surveillance Programs, and has been approved by the NRC [30]. LGS Unit 1 committed to use the ISP in place of its existing surveillance programs in the license amendment issued by the NRC regarding the implementation of the BWRVIP ISP, dated November 4, 2003 [17].

Under the ISP, LGS Unit 1 is not a host plant, and no capsules are scheduled for removal from the LGS Unit 1 reactor vessel. Representative surveillance capsule materials for the LGS Unit 1 target beltline weld are contained in the River Bend Capsules and Supplemental Surveillance Program (SSP) Capsules C, F, and H. Representative materials for the LGS Unit 1 target beltline plate are in the Peach Bottom 2 Capsules.

The BWRVIP ISP program and capsule withdrawal schedule are administered in accordance with BWRVIP-86, Revision 1-A [30]. The capsule withdrawal schedule is described here for information only. As of this writing, the next ISP surveillance capsule from River Bend is

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 1 PTLR Revision 0-NP Page 37 of 37 scheduled to be withdrawn and tested in approximately 2025, and one additional capsule will be withdrawn during the license renewal period in approximately 2030 [30]. As of this writing, one ISP capsule is scheduled to be withdrawn from Peach Bottom 2 during the license renewal period in approximately 2030 [30]. No further SSP capsules are scheduled for withdrawal.

ATTACHMENT 7 Limerick Generating Station Unit 2 Pressure and Temperature Limits Report (PTLR) for 57 Effective Full-Power Years (EFPY)

(Non-Proprietary Version)

Non-Confidential Information Submitted Under 10 CFR 2.390 Exelon Corporation Limerick Generating Station Unit 2 Pressure and Temperature Limits Report (PTLR) for 57 Effective Full-Power Years (EFPY)

Revision 0-NP

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 2 of 38 Table of Contents Section Page 1.0 Purpose 3 2.0 Applicability 3 3.0 Methodology 4 4.0 Operating Limits 5 5.0 Discussion 7 6.0 References 14 Figure 1 LGS Unit 2 P-T Curve A (Hydrostatic Pressure and Leak Tests) for 57 17 EFPY Figure 2 LGS Unit 2 P-T Curve B (Normal Operation - Core Not Critical) for 57 18 EFPY Figure 3 LGS Unit 2 P-T Curve C (Normal Operation - Core Critical) for 57 EFPY 19 Figure 4 LGS Unit 2 Overall Composite Curves A, B, and C, 57 EFPY 20 Figure 5 LGS Feedwater Nozzle 3-D Finite Element Model [22] 21 Figure 6 LGS LPCI Nozzle Finite Element Model [20] 22 Table 1 LGS Unit 2 Pressure Test (Curve A) P-T Curves for 57 EFPY 23 Table 2 LGS Unit 2 Core Not Critical (Curve B) P-T Curves for 57 EFPY 26 Table 3 LGS Unit 2 Core Critical (Curve C) P-T Curves for 57 EFPY 29 Table 4 LGS Unit 2 ART Table for 57 EFPY 32 Table 5 Nozzle Stress Intensity Factors 35 Table 6 LGS Unit 2 P-T Curve Input Parameters 35 Appendix A Limerick Reactor Vessel Materials Surveillance Program 37

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 3 of 38 1.0 Purpose The purpose of the Limerick Generating Station (LGS) Unit 2 Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

1. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heat-up, Cool-down and Hydrostatic/Class 1 Leak Testing;

2. RCS Heat-up and Cool-down rates;

3. RPV head flange boltup temperature limits.

This report has been prepared in accordance with the requirements of Licensing Topical Reports SIR-05-044, Revision 1-A, contained within BWROG-TP-11-022-A, Revision 1 [1], and 0900876.401, Revision 0-A, contained within BWROG-TP-11-023-A, Revision 0 [2].

2.0 Applicability This report is applicable to the LGS Unit 2 RPV for up to 57 Effective Full-Power Years (EFPY).

The following LGS Unit 2 Technical Specifications (TS) are affected by the information contained in this report:

TS 3/4.4.6 RCS Pressure/Temperature (P-T) Limits

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 4 of 38 3.0 Methodology The limits in this report were derived as follows:

1. The methodology used is in accordance with Reference [1] and Reference [2],

incorporating the NRC Safety Evaluations in References [3] and [4], respectively.

2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [5], using the RAMA computer code, as documented in Reference [6].

3. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 (RG 1.99) [7], as documented in Reference [8].

4. The pressure and temperature limits, which were calculated in accordance with Reference

[1], are documented in Reference [9].

5. This revision of the pressure and temperature limits report is to incorporate the following changes:

• Revision 0: Initial issue of PTLR.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59 [11], provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPV or other plant design assumption modifications in the UFSAR, cannot be made without prior NRC approval. Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 5 of 38 4.0 Operating Limits The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operating limits for pressure and temperature are required for three categories of operation:

(a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 57 EFPY for LGS Unit 2, as documented in Reference

[9], and are provided in Figure 1 through Figure 3 for LGS Unit 2. A tabulation of the curves is included in Table 1 through Table 3. The adjusted reference temperature (ART) table for 57 EFPY for the LGS Unit 2 vessel beltline materials is shown in Table 4 [8]. Inputs to the P-T curves are summarized in Table 5 and Table 6.

The resulting P-T curves are based on the geometry, design and materials information for the LGS Unit 2 vessel with the following conditions:

• Heat-up/Cool-down rate limit during Hydrostatic Class 1 Leak Testing (Figure 1: Curve A): 25F/hour 1 [9].

• Normal Operating Heat-up/Cool-down rate limit (Figure 2: Curve B - non-nuclear heating, and Figure 3: Curve C - nuclear heating): 100°F/hour 2 [9].

• RPV bottom head coolant temperature to RPV coolant temperature T limit during Recirculation Pump startup: 145°F [1].

1 Interpreted as the temperature change in any 1-hour period is less than or equal to 25°F.

2 Interpreted as the temperature change in any 1-hour period is less than or equal to 100°F.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 6 of 38

• Recirculation loop coolant temperature to RPV coolant temperature T limit during Recirculation Pump startup: 50°F [1].

• RPV flange and adjacent shell temperature limit 70F [9].

Minimum temperature limits are set in accordance with 10CFR50, Appendix G [10, Table 1].

The minimum moderator temperature used in the plant shutdown margin evaluation, 68°F [14],

is also included as a minimum bolt-up temperature requirement. An additional 60°F margin above the requirements in Table 1 of 10CFR50, Appendix G, has been commonly applied in the BWR industry. For LGS Unit 2, the limiting RTNDT for the closure flange region is 10°F, and the minimum temperature is 70°F (i.e. 10°F + 60°F) [9]. For Curves A and B, this 60°F margin is a recommendation, but for Curve C, the 60°F margin is required. For consistency with prior work, the minimum temperature for Curves A, B, and C for LGS Unit 2 is set to 70°F.

These values are consistent with the minimum temperature limits and minimum bolt-up temperature in the previous docketed P-T curves, [12] (approved by the NRC in Reference [13]).

These values also bound the minimum temperature in the first set of P-T limits approved for initial operation (i.e. initial licensed curves in Technical Specifications Figure 3.4.6.1-1) [14],

thereby addressing the NRC condition in Reference [3, Section 4.0].

The composite P-T curves are extended below 0 psig to -14.7 psig based on the evaluation in Reference [15], which demonstrates that the P-T curves are applicable to negative gauge pressures. A pressure of -14.7 psig bounds the maximum expected vacuum pressure as well as externally applied pressures the RPV may experience. Since the P-T curve calculation methods used do not specifically apply to negative values of pressure, the tabulated results start at 0 psig.

However, the minimum analyzed RPV pressure is -14.7 psig.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 7 of 38 5.0 Discussion 5.1 Adjusted Reference Temperatures The adjusted reference temperature (ART) of the limiting beltline material is used to adjust beltline P-T curves to account for irradiation effects. RG 1.99 [7] provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the LGS Unit 2 vessel plate, weld, and forging materials [8]. The Cu and Ni values were used with Table 1 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds.

The Cu and Ni values were used with Table 2 of RG 1.99 to determine a CF per Paragraph 1.1 of RG 1.99 for plates and forgings. For materials where surveillance data exists, a fitted CF has been used in the calculation of ART for those heats, in accordance with Regulatory Position 2.1 in RG 1.99. Use of surveillance data from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) for LGS Unit 2 was approved by the NRC in Reference [13].

The peak RPV ID fluence values of 1.05 x 1018 n/cm2 at 57 EFPY for the lower-intermediate shell plates and 7.69 x 1017 n/cm2 for the lower shell plates were obtained from Reference [6].

Fluence values in Reference [6] were calculated in accordance with RG 1.190 [5]. A plant-specific damage assessment, in terms of displacements per atom (dpa) was performed in Reference [6] to determine through-wall fluence for LGS Unit 2, as permitted by RG 1.99. The resulting attenuation factor for beltline plates and welds is 0.67-0.69 for a postulated 1/4T flaw.

Consequently, the 1/4T fluence for 57 EFPY is 7.08 x 1017 n/cm2 for the lower intermediate shell plates and 5.23 x 1017 n/cm2 for the lower shell plates. The limiting value for ART for beltline plates and welds, corresponding to lower shell beltline plate heat no. B3416-1 is 100.6°F [8].

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 8 of 38 The P-T limits are developed to bound all ferritic materials in the RPV, including the consideration of stress levels from structural discontinuities such as nozzles. LGS Unit 2 has two sets of nozzles in the RPV beltline: the instrument (N16) nozzles and the low pressure coolant injection (LPCI, N17) nozzles [8]; the N16 and N17 nozzles are in the lower-intermediate shell beltline plates. There are no additional forged or partial penetration nozzles in the extended beltline at LGS Unit 2 [8]. The feedwater (FW) nozzle is considered in the evaluation of the non-beltline (upper vessel) region P-T limits.

The LPCI (N17) nozzles and welds have a limiting fluence at the RPV ID of 4.54 x 1017 n/cm2 at 57 EFPY, obtained from Reference [6] and calculated in accordance with RG 1.190 [5]. Similar to the RPV beltline plates and welds described above, through-wall fluence for the LPCI nozzles was attenuated using the dpa methodology in Reference [6]. The resulting attenuation factor is 0.70 for a postulated 1/4T flaw in the LPCI nozzle blend radius. Consequently, the 1/4T fluence for 57 EFPY for the limiting LPCI nozzle location is 3.19 x 1017 n/cm2. The limiting 57 EFPY ART value for the LPCI nozzles and welds is 60.2°F [8].

The instrument (N16) nozzle inserts and welds at LGS Units 1 and 2 are fabricated from non-ferritic materials and do not require evaluation for loss of fracture toughness [8]. However, the effect of the penetration on the adjacent shell is considered in the development of bounding beltline P-T limits, according to the methodology in Reference [2]. The instrument nozzles have a limiting fluence at the RPV ID of 3.29 x 1017 n/cm2 at 57 EFPY, obtained from Reference [6]

and calculated in accordance with RG 1.190 [5]. This fluence value applies to the limiting N16 nozzle location (lower-intermediate shells). Similar to the RPV beltline plates and welds described above, through-wall fluence for the N16 nozzles was attenuated using the dpa methodology in Reference [6]. The resulting attenuation factor is 0.70 for a postulated 1/4T flaw in the instrument nozzle corner. Consequently, the 1/4T fluence for 57 EFPY for the limiting instrument nozzle location is 2.29 x 1017 n/cm2. The limiting 57 EFPY ART value for the instrument nozzles is 37.7°F [8].

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 9 of 38 5.2 P-T Curve Development The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cool-down and is in the outer wall during heat-up. However, as a conservative simplification, the thermal gradient stresses at the 1/4T location are assumed to be tensile for both heat-up and cool-down. This results in the approach of applying the maximum tensile stresses at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness at the 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, and for a given pressure, the coolant saturation temperature is well above the P-T curve limiting temperature. Consequently, the material toughness at a given pressure would exceed the allowable toughness.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heat-up and cool-down temperature rate of 100F/hour for which the curves are applicable. However, the core not critical and core critical curves were also developed to bound Service Level A/B RPV thermal transients. P-T curves are developed for anticipated operational occurrences, and Technical Specifications limit operation to 100°F/hour. For the hydrostatic pressure and leak test curve (Curve A), a coolant heat-up and cool-down temperature rate of 25°F/hour must be maintained. The P-T limits and corresponding limits of either Curve A or B may be applied, if necessary, while achieving or recovering from test conditions.

So, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heat-up/cool-down rate limits cannot be maintained.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 10 of 38 5.3 Material Data The initial RTNDT, chemistry (weight-percent copper and nickel), and ART at the 1/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 1017 n/cm2 for E >

1 MeV) are shown in Table 4 [8]. Use of initial RTNDT values in the determination of P-T curves for LGS Unit 2 was approved by the NRC in Reference [16].

Per Reference [8] and in accordance with Appendix A of Reference [1], the LGS Unit 2 representative weld and plate surveillance materials data were reviewed from the BWRVIP ISP

[18]. The representative plate material heat for LGS Unit 2 (B0673-1) in the ISP is not the same as the target plate heat in LGS Unit 2 (B3416-1). The representative weld material heat for LGS Unit 2 (5P6756) is not the same heat number as the target vessel weld in LGS Unit 2 (640892).

No other surveillance heats are present in the LGS Unit 2 beltline. Therefore, the table CF values from RG1.99 Position 1.1 [7] were used in the determination of ART for all materials in the LGS Unit 2 beltline.

5.4 Plant-Specific Evaluation of Nozzles The only computer code used in the determination of the LGS Unit 2 P-T curves was the ANSYS finite element computer program:

• ANSYS Mechanical APDL and Workbench Release 18.1 [19] for:

o FW nozzle (non-beltline) and LPCI nozzle (beltline) thermal and pressure stress distributions in Reference [20].

ANSYS finite element analyses were used to develop the stress distributions through the FW and LPCI nozzles, and these stress distributions were used in the determination of the stress intensity factors for the FW and LPCI nozzles [20]. At the time that each of the analyses above was performed, the ANSYS program was controlled under the vendors 10 CFR 50 Appendix B [24]

Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 [25] was performed as a part of the computer program

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 11 of 38 verification by comparing the solutions produced by the computer code to hand calculations for several problems.

The plant-specific LGS Unit 2 FW nozzle analyses were performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients.

Detailed information regarding the analyses can be found in References [20-23]. The following summarizes the development of the thermal and pressure stress intensity factors for the FW nozzle:

• A one-quarter symmetric, three-dimensional (3-D) finite element model (FEM) of the FW nozzle was constructed and is shown in Figure 5. A single model was developed that represents both units, as the FW nozzle geometry is identical for both units. Details of the model are provided in Reference [22]. Temperature-dependent material properties were taken from the ASME Code,Section III, 1968 Edition with 1969 Summer Addenda and Article 4 of the 1969 Winter Addenda [26] and from the ASME Code,Section II, Part D, 2001 Edition with Addenda through 2003 [27] and are tabulated in Reference [20].

• Heat transfer coefficients were calculated in Reference [21] and are a function of FW temperature and flow rate.

• With respect to operating conditions, the bounding thermal transients during Normal and Upset operating conditions were tabulated in Reference [21] and analyzed in Reference

[20]. The thermal stress distributions, corresponding to the limiting times presented in Reference [20], along a linear path through the nozzle corner is used. The boundary integral equation/influence function (BIE/IF) methodology presented in Reference [1] is used in Reference [20] to calculate the thermal stress intensity factor, KIt, due to the thermal stresses by fitting a third order polynomial equation to the path stress distribution for the thermal load case.

• With respect to pressure stress, a unit pressure of 1000 psig was applied to the internal surfaces of the 3-D model in Reference [20]. The pressure stress distribution was taken along the same path as the thermal stress distribution. The BIE/IF methodology presented

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 12 of 38 in Reference [1] was used to calculate the applied pressure stress intensity factor, KIp, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting KIp may be linearly scaled to determine the KIp for various RPV internal pressures.

The plant-specific LGS Unit 2 LPCI nozzle analysis was performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients.

Detailed information regarding the analysis can be found in Reference [20]. The following summarizes the development of the thermal and pressure stress intensity factors for the LPCI nozzle:

• A one-quarter symmetric, 3-D FEM of the LPCI nozzle was constructed and is shown in Figure 6. A single model was developed that represents both units, as the LPCI nozzle geometry is identical for both units. Details of the model are provided in Reference [20].

Temperature-dependent material properties were taken from the ASME Code,Section III, 1968 Edition with 1969 Summer Addenda and Article 4 of the 1969 Winter Addenda

[26] and from the ASME Code,Section II, Part D, 2001 Edition with Addenda through 2003 [27] and are tabulated in Reference [20].

• Heat transfer coefficients were calculated in Reference [20] and are a function of LPCI temperature and flow rate.

• With respect to operating conditions, the bounding thermal transient for the region corresponding to the LPCI nozzles during normal and upset operating conditions was analyzed [20]. The thermal stress distribution, corresponding to the limiting time in Reference [20], along a linear path through the nozzle corner is used. The BIE/IF methodology presented in Reference [1] was used to calculate the thermal stress intensity factor, KIt, due to the thermal stresses by fitting a third order polynomial equation to the path stress distribution for the thermal load case.

• With respect to pressure stress, a unit pressure of 1000 psig was applied to the internal surfaces of the FEM [20]. The pressure stress distribution was taken along the same path

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 13 of 38 as the thermal stress distribution. The BIE/IF methodology presented in Reference [1] is used to calculate the pressure stress intensity factor, KIp, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting KIp can be linearly scaled to determine the KIp for various RPV internal pressures.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 14 of 38 6.0 References

1. BWROG-TP-11-022-A, Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors, August 2013. (ADAMS Accession No. ML13277A557)

2. BWROG-TP-11-023-A, Revision 0, Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations, May 2013. (ADAMS Accession No. ML13183A017

3. U.S. NRC Letter to BWROG dated May 16, 2013, Final Safety Evaluation for Boiling Water Reactor Owners Group Topical Report BWROG-TP-11-022, Revision 1, November 2011, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors (TAC NO. ME7649, ADAMS Accession No. ML13277A557).

4. U.S. NRC Letter to BWROG dated March 14, 2013, Final Safety Evaluation for Boiling Water Reactor Owners Group Topical Report BWROG-TP-11-023, Revision 0, November 2011, Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations (TAC NO. ME7650, ML13183A017)

5. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001.

6. TransWare Report No. LIM-FLU-002-R-002, Revision 0, Limerick Generating Station Unit 2 Reactor Pressure Vessel Fluence Evaluation, Excerpt provided in Exelon TODI No. 04139609-04, Rev. 0, dated May 29, 2020. SI File No. 1701104.201.

7. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

8. SI Calculation No. 1701104.301, Revision 1, Limerick Unit 1 and 2 RPV Beltline ART and USE Evaluation, August 18, 2020.

9. SI Calculation No. 1701104.302, Revision 1, Limerick Unit 1 and 2 P-T Curve Calculation for 57 EFPY, August 18, 2020.

10. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix G, Fracture Toughness Requirements, December 12, 2013.

11. U.S. Code of Federal Regulations, Title 10, Part 50, Section 59, Changes, tests and experiments, August 28, 2007.

12. Attachment 7 to PECO Nuclear Letter dated November 20, 2000, GE Nuclear Energy Report No. GE-NE-B11-00836-00-02a NP, Revision 0, Pressure-Temperature Curves For PECO Energy Company, Limerick Unit 2, July 2000 (ADAMS Accession No. ML003772248).

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 15 of 38

13. Limerick Generating Station, Unit 2, License Amendment No. 111, Issuance of Amendment Re. Update The Pressure-Temperature Limit Curves (TAC No. MB0590, ADAMS Accession No. ML010540068), March 23, 2001.

14. Exelon TODI No. 04139609-04, Revision 0, May 29, 2020. SI File No. 1701104.201.

15. SI Calculation No. 1701104.304, Revision 0, Limerick Units 1 and 2 RPV Vacuum Assessment, June 15, 2020.

16. NRC acceptance of initial RTNDT values for LGS Unit 2:

a. Letter from F. Rinaldi (USNRC) to G.A Hunger (PECO), dated April 21, 1994, Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, and Limerick Generating Station (LGS), Units 1 and 2, (TAC Nos. M83495, M83496, M83477, and M83478). (ADAMS Accession No. 9405040314). SI File No. 1701104.210.

b. Letter from F. Rinaldi (USNRC) to G.A. Hunger (PECO), dated December 9, 1996, Closeout for Philadelphia Electric Company (PECO) Response to Generic Letter 92-01, Revision 1, Supplement 1, Limerick Generating Station (LGS),

Units 1 and 2 (TAC Nos. M92691 and M92692). (ADAMS Accession No.

9612230225). SI File No. 1701104.210.

17. Limerick Generating Station, Unit 1 License Amendment No. 167 and Unit 2 License Amendment No. 130, Issuance of Amendment Re. Revision to the Reactor Pressure Vessel Material Surveillance Program (TAC Nos. MB7003 and MB7004, ADAMS Accession No. ML032310540), November 4, 2003.

18. BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014.

3002003144. EPRI PROPRIETARY INFORMATION.

19. ANSYS Mechanical APDL (UP20170403) and Workbench (March 31, 2017), Release 18.1, SAS IP, Inc.

20. SI Calculation No. 1701104.303, Revision 0, Low Pressure Coolant Injection Nozzle and Feedwater Nozzle Fracture Mechanics Evaluation for Pressure-Temperature Limit Curve Development, June 17, 2020.

21. SI Calculation No. 1000818.301, Revision 0, Feedwater Nozzle FEA Loads Calculation, November 16, 2010.

22. SI Calculation No. 1000818.302, Revision 0, Feedwater Nozzle Finite Element Model, November 18, 2010.

23. SI Calculation No. 1000818.303, Revision 0, Feedwater Nozzle Stress Analysis, November 19, 2010.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 16 of 38

24. U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Appendix B, Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants.

25. U.S. Nuclear Regulatory Commission, Generic Letter 83-11, Supplement 1, License Qualification for Performing Safety Analyses, June 24, 1999.

26. ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Vessels, 1968 Edition with 1969 Summer Addenda and Article 4 of the 1969 Winter Addenda.

27. ASME Boiler and Pressure Vessel Code,Section II, Part D, Material Properties, 2001 Edition with Addenda through 2003.

28. License Renewal Application, Limerick Generating Station Units 1 and 2, Facility Operating Licenses Nos. NPF-39 and NPF-85.

29. Limerick Generating Station Updated Final Safety Analysis Report, Chapter 5 - Reactor Coolant System and Connected Systems, Revision 19, September 2018. SI File No.

1701104.203 (ADAMS Accession No. ML18285A600).

30. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, January 31, 2008.

31. BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2012. 1025144.

EPRI PROPRIETARY INFORMATION.

32. Letter from L. Regner (USNRC) to A. McGehee (EPRI), dated July 23, 2019, Duane Arnold Energy Center - Approval of Change in the BWRVIP Integrated Surveillance Program Capsule Test Schedule to Accommodate Early Closure (EPID L-2019-LLL-0022). (ADAMS Accession No. ML19198A010).

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 17 of 38 Figure 1: LGS Unit 2 P-T Curve A (Hydrostatic Pressure and Leak Tests) for 57 EFPY

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 18 of 38 Figure 2: LGS Unit 2 P-T Curve B (Normal Operation - Core Not Critical) for 57 EFPY

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 19 of 38 Figure 3: LGS Unit 2 P-T Curve C (Normal Operation - Core Critical) for 57 EFPY

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 20 of 38 Figure 4: LGS Unit 2 Overall Composite Curves A, B, and C, 57 EFPY

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 21 of 38 Figure 5: LGS Feedwater Nozzle 3-D Finite Element Model [22]

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 22 of 38 Figure 6: LGS LPCI Nozzle Finite Element Model [20]

With Thermal Sleeve Without Thermal Sleeve

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 23 of 38 Table 1: LGS Unit 2 Pressure Test (Curve A) P-T Curves for 57 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 70.0 -14.7 70.0 451.2 80.2 498.5 88.7 545.7 96.0 593.0 102.3 640.3 107.9 687.6 113.0 734.8 117.5 782.1 121.7 829.4 125.6 876.7 129.2 923.9 132.6 971.2 135.7 1018.5 138.7 1065.7 142.5 1115.5 146.0 1165.2 149.3 1214.9 152.4 1264.6 155.4 1314.4

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 24 of 38 Table 1: LGS Unit 2 Pressure Test (Curve A) P-T Curves for 57 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 70.0 -14.7 70.0 734.2 75.1 782.9 79.8 831.7 84.0 880.4 87.9 929.2 91.5 978.0 94.9 1026.7 98.1 1075.5 101.1 1124.2 103.9 1173.0 106.5 1221.7 109.1 1270.5 111.5 1319.2

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 25 of 38 Table 1: LGS Unit 2 Pressure Test (Curve A) P-T Curves for 57 EFPY (continued)

Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 70.0 -14.7 70.0 312.6 100.0 312.6 100.0 768.0 104.6 817.7 108.8 867.4 112.6 917.1 116.2 966.8 119.6 1016.4 122.7 1066.1 125.7 1115.8 128.5 1165.5 131.1 1215.2 133.6 1264.9 136.0 1314.6

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 26 of 38 Table 2: LGS Unit 2 Core Not Critical (Curve B) P-T Curves for 57 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 82.8 -14.7 82.8 0.0 93.6 48.9 102.5 97.7 110.1 146.5 116.6 195.4 122.4 244.2 127.6 293.1 132.3 341.9 136.6 390.8 140.6 439.6 144.3 488.4 147.7 537.3 150.9 586.1 153.9 635.0 156.7 683.8 159.4 732.7 162.0 781.5 164.4 830.4 166.7 879.2 168.9 928.0 171.0 976.9 173.1 1025.7 175.0 1074.6 176.9 1123.4 178.7 1172.3 180.4 1221.1 182.1 1269.9 183.8 1318.8

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 27 of 38 Table 2: LGS Unit 2 Core Not Critical (Curve B) P-T Curves for 57 EFPY (continued)

Bottom Head Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 70.0 -14.7 70.0 478.6 76.8 527.8 82.7 577.1 88.1 626.4 92.9 675.7 97.3 725.0 101.3 774.3 105.1 823.6 108.5 872.9 111.8 922.2 114.8 971.5 117.7 1020.8 120.4 1070.1 123.0 1119.4 125.5 1168.7 127.8 1217.9 130.0 1267.2 132.2 1316.5

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 28 of 38 Table 2: LGS Unit 2 Core Not Critical (Curve B) P-T Curves for 57 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 70.0 -14.7 70.0 203.3 77.9 239.8 84.7 276.2 90.7 312.6 130.0 312.6 130.0 697.9 133.3 745.9 136.4 794.0 139.3 842.1 142.0 890.1 144.6 938.2 147.1 986.3 149.4 1034.3 151.7 1082.4 153.8 1130.4 155.9 1178.5 157.9 1226.6 159.8 1274.6 161.6 1322.7

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 29 of 38 Table 3: LGS Unit 2 Core Critical (Curve C) P-T Curves for 57 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 122.8 -14.7 122.8 0.0 133.6 48.9 142.5 97.7 150.1 146.5 156.6 195.4 162.4 244.2 167.6 293.1 172.3 341.9 176.6 390.8 180.6 439.6 184.3 488.4 187.7 537.3 190.9 586.1 193.9 635.0 196.7 683.8 199.4 732.7 202.0 781.5 204.4 830.4 206.7 879.2 208.9 928.0 211.0 976.9 213.1 1025.7 215.0 1074.6 216.9 1123.4 218.7 1172.3 220.4 1221.1 222.1 1269.9 223.8 1318.8

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 30 of 38 Table 3: LGS Unit 2 Core Critical (Curve C) P-T Curves for 57 EFPY (continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 70.0 -14.7 70.0 291.6 83.9 340.5 94.8 389.4 103.7 438.3 111.3 487.2 117.8 536.1 123.6 585.0 128.8 633.9 133.5 682.8 137.9 731.7 141.8 780.6 145.5 829.5 148.9 878.4 152.1 927.3 155.1 976.2 158.0 1025.1 160.6 1074.0 163.2 1122.9 165.6 1171.8 167.9 1220.7 170.1 1269.6 172.3 1318.5

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 31 of 38 Table 3: LGS Unit 2 Core Critical (Curve C) P-T Curves for 57 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 70.0 -14.7 70.0 86.0 89.4 131.3 103.3 176.6 114.2 221.9 123.1 267.3 130.7 312.6 170.0 312.6 170.0 697.9 173.3 745.9 176.4 794.0 179.3 842.1 182.0 890.1 184.6 938.2 187.1 986.3 189.4 1034.3 191.7 1082.4 193.8 1130.4 195.9 1178.5 197.9 1226.6 199.8 1274.6 201.6 1322.7

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 32 of 38 Table 4: LGS Unit 2 ART Table for 57 EFPY Initial Chemistry Fluence Fluence Adjustments For 1/4t Chemistry Fluence RTNDT (wt%) (1) at ID at 1/4T Description ID Heat No. Lot No. Factor, CF Factor, RTNDT Margin Terms ART

(°F) (n/cm2) (n/cm2)

(°F) FF (°F) (°F)

(1) Cu Ni (2) (2) (°F) i (°F)

Lower Shell #1 14-1 B3312-1 - 10 0.13 0.58 90 7.69E+17 5.23E+17 0.300 27.0 13.5 0.0 64.0 Lower Shell #1 14-2 B3416-1 - 40 0.14 0.65 101 7.69E+17 5.23E+17 0.300 30.3 15.1 0.0 100.6 Plates Lower Shell #1 14-3 C9621-2 - 22 0.15 0.60 110 7.69E+17 5.23E+17 0.300 33.0 16.5 0.0 88.0 Lower-Int Shell #2 17-1 C9569-2 - 10 0.11 0.51 73 1.05E+18 7.08E+17 0.351 25.7 12.8 0.0 61.3 Lower-Int Shell #2 17-2 C9526-1 - 10 0.11 0.56 74 1.05E+18 7.08E+17 0.351 26.0 13.0 0.0 62.0 Lower-Int Shell #2 17-3 C9526-2 - 10 0.11 0.56 74 1.05E+18 7.08E+17 0.351 26.0 13.0 0.0 62.0 Axial (Lower Shell) BA 432A2671 H019A27A -12 0.04 1.08 54 5.36E+17 3.67E+17 0.247 13.3 6.7 0.0 14.6 Axial (Lower Shell) BB 432A2671 H019A27A -12 0.04 1.08 54 5.71E+17 3.91E+17 0.256 13.8 6.9 0.0 15.6 Axial (Lower-Int Shell) BD 432A2671 H019A27A -12 0.04 1.08 54 8.71E+17 5.91E+17 0.320 17.3 8.6 0.0 22.6 Axial (Lower-Int Shell) BE 432A2671 H019A27A -12 0.04 1.08 54 8.69E+17 5.91E+17 0.320 17.3 8.6 0.0 22.6 Axial (Lower-Int Shell) BF 432A2671 H019A27A -12 0.04 1.08 54 7.46E+17 5.11E+17 0.296 16.0 8.0 0.0 20.0 Axial (Lower Shell) BA 03R728 L910A27A -50 0.03 0.92 41 5.36E+17 3.67E+17 0.247 10.1 5.1 0.0 -29.8 Axial (Lower Shell) BC 03R728 L910A27A -50 0.03 0.92 41 7.09E+17 4.83E+17 0.287 11.8 5.9 0.0 -26.4 Axial (Lower Shell) BA (3) 3P4000 3933 -50 (( . (E)

)) 27 5.36E+17 3.67E+17 0.247 6.7 3.3 0.0 -36.7 Axial (Lower Shell) BB (3) 3P4000 3933 -50 (( . (E)

)) 27 5.71E+17 3.91E+17 0.256 6.9 3.5 0.0 -36.2 Axial (Lower Shell) BC (3) 3P4000 3933 -50 (( . (E)

)) 27 7.09E+17 4.83E+17 0.287 7.8 3.9 0.0 -34.5 Welds Axial (Lower-Int Shell) BD (3) 3P4000 3933 -50 (( . (E)

)) 27 8.71E+17 5.91E+17 0.320 8.6 4.3 0.0 -32.7 Axial (Lower-Int Shell) BE (3) 3P4000 3933 -50 (( . (E)

)) 27 8.69E+17 5.91E+17 0.320 8.6 4.3 0.0 -32.7 Axial (Lower-Int Shell) BF (3) 3P4000 3933 -50 (( . (E)

)) 27 7.46E+17 5.11E+17 0.296 8.0 4.0 0.0 -34.0 Axial (Lower Shell) BB 401Z9711 A022A27A -50 0.02 0.83 27 5.71E+17 3.91E+17 0.256 6.9 3.5 0.0 -36.2 Axial (Lower Shell) BC 662A746 H013A27A -20 0.03 0.88 41 7.09E+17 4.83E+17 0.287 11.8 5.9 0.0 3.6 Axial (Lower Shell) BC 402A0462 B023A27A -50 0.02 0.90 27 7.09E+17 4.83E+17 0.287 7.8 3.9 0.0 -34.5 Axial (Lower Shell) BC 07L669 K004A27A -50 0.03 1.02 41 7.09E+17 4.83E+17 0.287 11.8 5.9 0.0 -26.4 Axial (Lower-Int Shell) BD 07L669 K004A27A -50 0.03 1.02 41 8.71E+17 5.91E+17 0.320 13.1 6.6 0.0 -23.8 Axial (Lower-Int Shell) BE 07L669 K004A27A -50 0.03 1.02 41 8.69E+17 5.91E+17 0.320 13.1 6.6 0.0 -23.8 Axial (Lower-Int Shell) BF 07L669 K004A27A -50 0.03 1.02 41 7.46E+17 5.11E+17 0.296 12.1 6.1 0.0 -25.7 Axial (Lower-Int Shell) BD 09L853 A111A27A -50 0.03 0.86 41 8.71E+17 5.91E+17 0.320 13.1 6.6 0.0 -23.8 Axial (Lower-Int Shell) BE 09L853 A111A27A -50 0.03 0.86 41 8.69E+17 5.91E+17 0.320 13.1 6.6 0.0 -23.8

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 33 of 38 Table 4: LGS Unit 2 ART Table for 57 EFPY (continued)

Initial Chemistry Fluence Fluence Adjustments For 1/4t Chemistry Fluence RTNDT (wt%) (1) at ID at 1/4T Margin Terms Description ID Heat No. Lot No. Factor, CF Factor, RTNDT ART

(°F) (n/cm2) (n/cm2)

(°F) FF (°F) (°F)

(1) Cu Ni (2) (2) i (°F)

(°F)

Girth (Lower to Lower-Int) AB 07L857 B101A27A -6 0.03 0.97 41 7.69E+17 5.23E+17 0.300 12.3 6.1 0.0 18.6 Girth (Lower to Lower-Int) AB L83355 S411B27AD -70 0.03 1.08 41 7.69E+17 5.23E+17 0.300 12.3 6.1 0.0 -45.4 Girth (Lower to Lower-Int) AB 402C4371 C115A27A -50 0.02 0.92 27 7.69E+17 5.23E+17 0.300 8.1 4.0 0.0 -33.8 Welds, contd.

Girth (Lower to Lower-Int) AB 03M014 C118A27A -34 0.01 0.94 20 7.69E+17 5.23E+17 0.300 6.0 3.0 0.0 -22.0 Girth (Lower to Lower-Int) AB 411A3531 H004A27A -50 0.02 0.96 27 7.69E+17 5.23E+17 0.300 8.1 4.0 0.0 -33.8 Girth (Lower to Lower-Int) AB 09M057 C109A27A -36 0.03 0.89 41 7.69E+17 5.23E+17 0.300 12.3 6.1 0.0 -11.4 Girth (Lower to Lower-Int) AB 640892 J424B27AE -60 0.09 1.00 122 7.69E+17 5.23E+17 0.300 36.6 18.3 0.0 13.2 Girth (Lower to Lower-Int) AB 401P6741 S419B27AG -60 0.03 0.92 41 7.69E+17 5.23E+17 0.300 12.3 6.1 0.0 -35.4 Girth (Lower to Lower-Int) AB 412P3611 J417B27AF -80 0.03 0.93 41 7.69E+17 5.23E+17 0.300 12.3 6.1 0.0 -55.4 Surveillance Weld (4) N/A (3) CTY538 A027A27A -50 (( . (E)

)) 41 1.05E+18 7.08E+17 0.351 14.4 7.2 0.0 -21.2 LPCI (N17) 892L-1 Q2Q33W - -20 0.18 0.83 141 4.54E+17 3.19E+17 0.228 32.1 16.0 0.0 44.2 LPCI (N17) 892L-2 Q2Q33W - -6 0.18 0.81 141 4.54E+17 3.19E+17 0.228 32.1 16.0 0.0 58.2 LPCI (N17) 892L-3 Q2Q33W - -4 0.18 0.82 141 4.54E+17 3.19E+17 0.228 32.1 16.0 0.0 60.2 Nozzles LPCI (N17) 892L-4 Q2Q33W - -20 0.18 0.82 141 4.54E+17 3.19E+17 0.228 32.1 16.0 0.0 44.2 N16-0°,

100°,

Instrument (N16) (5) (SB-166, SB-167) - 10 0.11 0.56 74 3.29E+17 2.29E+17 0.187 13.8 6.9 0.0 37.7 200°,

280° LPCI Nozzle-to-shell KA C3L46C J020A27A -20 0.02 0.87 27 4.54E+17 3.19E+17 0.228 6.1 3.1 0.0 -7.7 LPCI Nozzle-to-shell KA 422B7201 L030A27A -40 0.04 0.90 54 4.54E+17 3.19E+17 0.228 12.3 6.1 0.0 -15.4 Nozzle Welds LPCI Nozzle-to-shell KA 4P4784 (single) 3930 -50 0.06 0.87 82 4.54E+17 3.19E+17 0.228 18.7 9.3 0.0 -12.7 LPCI Nozzle-to-shell KA 4P4784 (tandem) 3930 -20 0.06 0.87 82 4.54E+17 3.19E+17 0.228 18.7 9.3 0.0 17.3 LPCI Nozzle-to-shell KA 07L669 K004A27A -50 0.03 1.02 41 4.54E+17 3.19E+17 0.228 9.3 4.7 0.0 -31.3 LPCI Nozzle-to-shell KA 09L853 A111A27A -50 0.03 0.86 41 4.54E+17 3.19E+17 0.228 9.3 4.7 0.0 -31.3 LPCI Nozzle-to-shell KA 432A2671 H019A27A -12 0.04 1.08 54 4.54E+17 3.19E+17 0.228 12.3 6.1 0.0 12.6 Notes:

1. Initial RTNDT and chemistry data for as-fabricated RPV materials are obtained from Tables 4.2.3-3 and 4.2.3-4 of Reference [28], unless otherwise noted.

2. Fluence values for 57 EFPY are obtained from Reference [6]. Fluence values for nozzles are reported for the 1/4T location along the nozzle extraction path, based on a plant-specific damage assessment (i.e. dpa) methodology.

3. For the noted vessel welds, best-estimate chemistry values from Appendix D of BWRVIP-135, Rev. 3 [18], are used and assumed to supersede original plant-specific chemistry values.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 34 of 38

4. Weld heat CTY538 is identified in Table 5.3-5 (page 5.3-58) of the LGS UFSAR [29] as a surveillance weld for LGS Unit 2, although UFSAR page 5.3-11 states that heat CTY538 was not used for production beltline welds. This weld heat was included in Table 4.2.3-4 of the LGS License Renewal Application [28] and is included for completeness here. The peak fluence for lower-intermediate beltline plates was used.

5. The N16 nozzle inserts and welds are fabricated from non-ferritic material and do not require evaluation for loss of fracture toughness. The corresponding ART value applies to the penetration in the surrounding shell, based on the initial RTNDT and CF corresponding to the limiting adjacent lower-intermediate shell plate and the fluence corresponding to the limiting nozzle location. The Unit 2 N16D (280°) nozzle was repaired with a half-nozzle repair [14]. The replacement half-nozzle material is SB-167, Alloy 690. A new weld pad and J-groove weld attaching the replacement nozzle to the weld pad at the RPV OD are ERNiCrFe-7A, Alloy 52M.

The half-nozzle repair materials are non-ferritic, and ART for the N16D nozzle is calculated based on the material properties of the adjacent shell plate as described above.

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 35 of 38 Table 5: Nozzle Stress Intensity Factors Nozzle Applied Pressure, KIp-app Thermal, KIt Feedwater 78.6 29.8 LPCI (N17) 80.7 60.7 Instrument (N16) 86.3 26.2 KI in units of ksi-in0.5 Table 6: LGS Unit 2 P-T Curve Input Parameters General Parameters Values Unit System for Tables and Plots English Temperature Instrument Uncertainty Adjustment (°F) 0 Pressure Instrument Uncertainty Adjustment (psig) 0 Water Density (lbm/ft3) 62.4 Full-Vessel Water Height (in) 870.5 Safety Factor for Curve A 1.5 Safety Factor for Curves B and C 2 Bolt-up Temperature (°F) 70 ART of Closure Flange Region (°F) 10 Starting Pressure for Curves (psig) -14.7 Atmospheric Pressure Adjustment (psi) 14.7 Preservice hydrotest pressure (psi) 1563 In-service hydrotest pressure (psi) 1150 Minimum in-service hydrotest temperature (°F) 145 Beltline Parameters Values Adjusted Reference Temperature (°F) 100.6 Vessel Radius (in) 126.69 Vessel Thickness (in) 6.19 Heat-up / Cool-down Rate (°F/hr) 100

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 36 of 38 Table 6: LGS Unit 2 P-T Curve Input Parameters (continued)

Instrument (N16) Nozzle Parameters Values Adjusted Reference Temperature (°F) 37.7 Vessel Radius (in) 126.69 Vessel Thickness (in) 6.19 Nozzle Outer Diameter (in) 3.5 Nozzle Inner Diameter (in) 1.938 Coefficient of Thermal Expansion (in/in/°F) 7.70E-06 Reference Pressure for Thermal Transient (psig) 1000 LPCI (N17) Nozzle Parameters Values Adjusted Reference Temperature (°F) 60.2 Applied Pressure Stress Intensity Factor (ksi*in^0.5) 80.7 Applied Thermal Stress Intensity Factor (ksi*in^0.5) 60.7 Scale KIT based on Saturation Temperature? No Reference Pressure for Thermal Transient (psig) 1000 Bottom Head Parameters Values Adjusted Reference Temperature (°F) 28 Vessel Radius (in) 126.69 Vessel Thickness (in) 6.19 Heat-up / Cool-down Rate (°F/hr) 100 Stress Concentration Factor 3 Upper Vessel (Feedwater Nozzle) Parameters Values Adjusted Reference Temperature (°F) 46 Applied Pressure Stress Intensity Factor (ksi*in^0.5) 78.6 Applied Thermal Stress Intensity Factor (ksi*in^0.5) 29.8 Minimum Thermal Stress Intensity Factor (ksi*in^0.5) 0 Scale KIT based on Saturation Temperature? No Reference Pressure for Thermal Transient (psig) 1000

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 37 of 38 Appendix A LIMERICK REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements [30], three surveillance capsules were installed in the LGS Unit 2 reactor vessel.

The surveillance capsules contain flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region. The 120° surveillance capsule was removed in October 2017 when the capsule holder was found damaged during a maintenance outage; the capsule was not reinstalled. Therefore, two surveillance capsules remain on standby for use as backup capsules by the BWRVIP Integrated Surveillance Program (ISP), or as otherwise needed.

LGS Unit 2 has replaced the original RPV material surveillance program with the BWRVIP ISP

[31]. LGS Unit 2 is currently committed to use the BWRVIP ISP and has made a licensing commitment to use the ISP for LGS Unit 2 during the period of extended operation. The BWRVIP ISP meets the requirements of 10 CFR 50, Appendix H, for Integrated Surveillance Programs, and has been approved by the NRC [31]. LGS Unit 2 committed to use the ISP in place of its existing surveillance programs in the license amendment issued by the NRC regarding the implementation of the BWRVIP ISP, dated November 4, 2003 [17].

Under the ISP, LGS Unit 2 is not a host plant, and no further capsules are scheduled for removal from the LGS Unit 2 reactor vessel. Representative surveillance capsule materials for the LGS Unit 2 target beltline weld are contained in the River Bend Capsules and Supplemental Surveillance Program (SSP) Capsules C, F, and H. Representative materials for the LGS Unit 2 target beltline plate are in the Duane Arnold and SSP-F surveillance capsules.

The BWRVIP ISP program and capsule withdrawal schedule are administered in accordance with BWRVIP-86, Revision 1-A [31]. The capsule withdrawal schedule is described here for

Non-Confidential Information Submitted Under 10 CFR 2.390 Limerick Generating Station Unit 2 PTLR Revision 0-NP Page 38 of 38 information only. As of this writing, the next ISP surveillance capsule from River Bend is scheduled to be withdrawn and tested in approximately 2025, and one additional capsule will be withdrawn during the license renewal period in approximately 2030 [31]. As of this writing, one ISP capsule will be withdrawn from Duane Arnold upon plant closure in late 2020 or early 2021

[32]. No further SSP capsules are scheduled for withdrawal.

ATTACHMENT 8 BWRVIP-135, Revision 3: BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations (Non-Proprietary Version)

BWRVIP 2020-069, Attachment 3 r=:::!11 1-1 ELECTRIC POWER RESEARCH INSTITUTE 2014 TECHNICAL REPORT BWRVIP-135, Revision 3: BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations lJ A,.* \.0

he Export Control Agreement on the bock cover.

NOTICE: This report contains proprietary inlormotion thot is the intellectual property of EPRI. Accordingly, it is available only under license from EPRI and may not be reproduced or disclosed, wholly or in port, by any licensee to any other person or organization.

PZant-Specific Evaluations Limerick 1 Representative Surveillance Materials The ISP Representative Surveillance Materials for the Limerick 1 vessel target weld and plates are shown in the following table.

Table 2-58 Target Vessel Materials and ISP Representative Materials for Limerick 1 Target Vessel Materials ISP Representative Materials Weld 640892 5P6756 Plate C7677-1 C2761-2 Summary of Available Surveillance Data: Plate The representative plate material C2761-2 is contained in the following ISP capsules:

Peach Bottom 2 Capsules Specific surveillance data related to plate heat C2761-2 are summarized in Appendix A-9. One capsule containing this plate heat has been tested. The Charpy V-notch surveillance results are as follows:

Table 2-59 T30 Shift Results for Plate Heat C2761-2 Cu Ni Fluence Capsule AT30 (° F)

(wt%) (wt%) (10 11 n/cm2 , E > 1 MeV)

Peach Bottom 2 120 ° 0.10 0.54 1.8 -3.2 No surveillance based chemistry factor will be available until a second capsule is tested (see reference [l] for capsule test schedule).

Conclusions and Recommendations Because the representative plate material is not the same heat number as the target plate in the Limerick 1 vessel, the utility should use the chemistry factor from the Regulatory Guide 1.99, Rev. 2 tables (Regulatory Position 1.1) to determine the projected ART value for the target vessel plate. Recommended guidelines for evaluation of ISP surveillance data are provided in Section 3 of this Data Source Book.

2-40

Plant-Specific Evaluations Summary of Available Surveillance Data: Weld The representative weld material 5P6756 is contained in the following ISP capsules:

River Bend Capsules SSP Capsules C, F, and H Specific surveillance data related to weld heat 5P6756 are presented in Appendix B-11 and the results are summarized below. Four capsules containing weld heat 5P6756 have been tested. The Charpy V-notch surveillance results are as follows:

Table 2-60 T30 Shift Results for Weld Heat 5P6756 Cu Ni Fluence Capsule 11 2 AT30 (° F)

(wt%) (wt%) (10 n/cm , E > 1 MeV)

River Bend 183 ° 11.6 53.7 SSP F 19.364 61.9 0.06 0.93 SSPH 15.766 63.7 SSPC 2.93 23.6 The results given in Appendix B-11 show a fitted chemistry factor (CF) of (( (E))), as compared to a value of 82.0°F from the chemistry tables in Reg. Guide 1.99, Rev. 2. The maximum scatter in the fitted data is well within the I-sigma value of 28°F for welds as given in the Reg. Guide.

Conclusions and Recommendations Because the representative weld material is not the same heat number as the target weld in the Limerick 1 vessel, the utility should use the chemistry factor from the Regulatory Guide 1.99 Rev. 2 tables to determine the projected ART value for the target vessel weld, 640892.

However, this surveillance data should be used to evaluate the Limerick 1 beltline weld 5P6756.

Recommended guidelines for evaluation of ISP surveillance data are provided in Section 3 of this Data Source Book.

2-41

PZant-Specific Evaluations Limerick 2 Representative Surveillance Materials The ISP Representative Surveillance Materials for the Limerick 2 vessel target weld and plates are shown in the following table.

Table 2-61 Target Vessel Materials and ISP Representative Materials for Limerick 2 Target Vessel Materials ISP Representative Materials Weld 640892 5P6756 Plate 83416-1 80673-1 Summary of Available Surveillance Data: Plate The representative plate material B0673-1 is contained in the following ISP capsules:

Duane Arnold and SSP Capsule F Specific surveillance data related to plate heat B0673-1 are summarized in Appendix A-3. Four capsules containing this plate heat have been tested. The Charpy V-notch surveillance results are as follows:

Table 2-62 T30 Shift Results for Plate Heat 80673-1 Cu Ni Fluence Capsule 11 2 AT30 ( ° F)

(wt%) (wt%) (10 n/cm , E > 1 MeV)

Duane Arnold 288 ° 5.09 41.8 Duane Arnold 36° 11.7 77.0 0.15 0.65 SSP F 18.699 73.4 Duane Arnold 108 ° 26.3 94.3 The results given in Appendix A-3 show a fitted chemistry factor (CF) of (( (E))), as

° compared to a value of lll.25 F from the chemistry tables in Reg. Guide 1.99, Rev. 2. The maximum scatter in the fitted data is (( (E)))which is well within the 1-sigma value of

° l7 F for plates as given in the Reg. Guide.

2-42

Plant-Specific Evaluations Conclusions and Recommendations Because the representative plate material is not the same heat number as the target plate in the Limerick 2 vessel, the utility should use the chemistry factor from the Regulatory Guide 1.99 Rev. 2 tables (Regulatory Position 1.1) to determine the projected ART value for the target vessel plate. Recommended guidelines for evaluation of ISP surveillance data are provided in Section 3 of this Data Source Book.

Summary of Available Surveillance Data: Weld The representative weld material 5P6756 is contained in the following ISP capsules:

River Bend Capsules SSP Capsules C, F, and H Specific surveillance data related to weld heat 5P6756 are presented in Appendix B-11 and the results are summarized below. Four capsules containing weld heat 5P6756 have been tested. The Charpy V-notch surveillance results are as follows:

Table 2-63 T30 Shift Results for Weld Heat 5P6756 Cu Ni Fluence Capsule 11 2 AT30 (° F)

(wt%) (wt%) (10 n/cm , E > 1 MeV)

River Bend 183 ° 11.6 53.7 SSP F 19.364 61.9 0.06 0.93 SSPH 15.766 63.7 SSPC 2.93 23.6 The results given in Appendix B-11 show a fitted chemistry factor (CF) of (( (E))), as, as

° compared to a value of 82.0 F from the chemistry tables in Reg. Guide 1.99, Rev. 2. The maximum scatter in the fitted data is well within the I-sigma value of 28°F for welds as given in the Reg. Guide.

Conclusions and Recommendations Because the representative weld material is not the same heat number as the target weld in the Limerick 2 vessel, the utility should use the chemistry factor from the Regulatory Guide 1.99, Rev. 2 tables to determine the projected ART value for the target vessel weld. Recommended guidelines for evaluation of ISP surveillance data are provided in Section 3 of this Data Source Book.

2-43

ISP Weld Heat Evaluations B-11 Weld Heat: 5P6756 Summary of Available Charpy V-Notch Test Data The available Charpy V-notch test data sets for weld heat 5P6756 are listed in Table B-11-1. The source documents for the data are provided, and the capsule designation and fluence values are also provided for irradiated data sets.

Table B-11-1 ISP Capsules Containing Weld Heat 5P6756 Capsule Fluence (E> 1 MeV, 10" n/cm') Reference Unirradiated Baseline Data - Reference B-11-1 River Bend 183° 11.6 Reference B-11-2 SSP Capsule F 19.364 Reference B-11-3 SSP Capsule H 15.766 Reference B-11-1 SSP Capsule C 2.93 Reference B-11-12 The CVN test data for each set taken from the references noted above are presented in Tables B-11-7 through B-11-11. The BWRVIP ISP uses the hyperbolic tangent (tanh) function as a statistical curve-fit tool to model the transition temperature toughness data. Tanh curve plots for each data set have been generated using CVGRAPH, Version 5 [Reference B-11-4] and the plots are provided in Figures B-11-1 through B-11-5.

Best Estimate Chemistry Table B-11-2 details the best estimate average chemistry values for weld heat 5P6756 surveillance material. Chemical compositions are presented in weight percent. If there are multiple measurements on a single specimen, those are first averaged to yield a single value for that specimen, and then the different specimens are averaged to determine the heat best estimate.

Table B-11-2 Best Estimate Chemistry of Available Data Sets for Weld Heat 5P6756 Cu(wt%) NI (wt%) P(wt%) s (wt%) Sl(wt%) Specimen ID Source 0.06 0.93 0.013 0.015 0.37 TP2-72 0.04 0.92 0.009 - - TP2-72 Reference B-11-5 0.05 0.93 0.011 0.015 0.37 Average TP2-72 0.067 0.93 0.00659 - 0.42 W-6 Reference B-11-2 0.06 0.93 0.009 0.015 0.40 + Best Estimate Average B-11-1

ISP Weld Heat Evaluations Calculation of Chemistry Factor (CF):

The Chemistry Factor (CF) associated with the best estimate chemistry, as determined from U.S.

NRC Regulatory Guide 1.99, Revision 2 [Reference B-11-6], Table 1 (weld metal), is:

CF<...,56) = 82.0°F Effects of Irradiation The radiation induced transition temperature shifts for heat 5P6756 are shown in Table B-11-3.

The T30 [30 ft-lb Transition Temperature], T,.[50 ft-lb Transition Temperature], and T,,mil

[35 mil Lateral Expansion Temperature] index temperatures have been determined for each Charpy data set, and each irradiated set is compared to the baseline (unirradiated) index temperatures. The change in Upper Shelf Energy (USE) is also shown. The unirradiated and irradiated values are taken from the CVGRAPH fits presented at the back of this sub-appendix (only CVN energy fits are presented).

Comparison of Actual vs. Predicted Embrittlement A predicted shift in the 30 ft-lb transition temperature (dT30) is calculated for each irradiated data set using the Reg. Guide 1.99, Rev. 2, Regulatory Position 1.1 method. Table B-11-4 compares the predicted shift with the measured dT30 ( 0 F) taken from Table B-11-3.

Decrease in USE Table B-11-5 shows the actual percent decrease in upper shelf energy (USE). The measured percent decrease is calculated from the values presented in Table B-11-3.

B-11-2

ISP Weld Heat Evaluations Table B-11-3 Effect of Irradiation (E>1.0 MeV) on the Notch Toughness Properties of Weld Heat 5P6756 T,., 50 ft-lb T,., 30 ft-lb T""'"' 35 mil Lateral CVN Upper Shelf Energy Transition Material capsule Transition Temperature Expansion Temperature (USE)

Temperature Identity ID Unlrrad lrrad AT,. Unlrrad lrrad AT,. Unlrrad lrrad AT,.., Unlrrad lrrad Change

("F) ("F) ("F) ("F) ("F) ("F) ("F) ("F) ("F) (ft-lb) (ft-lb) (ft-lb) 183° -67.1 -13.4 53.7 -21.3 33.1 54.4 -20.3 11.4 31.7 104.4 84.4 -20.0 RB1 and SSPF -67.1 -5.2 61.9 -21.3 39.5 60.8 -20.3 31.7 52.0 104.4 79.3 -25.1 SSP 5P6756 SSPH -67.1 -3.4 63.7 -21.3 22.3 43.6 -20.3 12.5 32.8 104.4 84.6 -19.8 SSPC -67.1 -43.5 23.6 -21.3 0.7 22.0 -20.3 -17.4 2.9 104.4 110.7 6.3 Table B-11-4 Comparison of Actual Versus Predicted Embrittlement for Weld Heat 5P6756 RG 1.99 Rev. 2 RG 1.99 Rev. 2 Capsule Fluence Measured Shift' Predicted Material Predicted Shift' Identity (x1 O" n/cm') "F Shlft+Margln'*'

"F "F

RB 183° Weld Heat 5P6756 in River Bend 1.16 53.7 36.7 73.3 SSP Capsule F Weld Heat 5P6756 in SSP Capsule F 1.9364 61.9 46.1 92.2 SSP Capsule H Weld Heat 5P6756 in SSP Capsule H 1.5766 63.7 42.2 84.3 SSP Capsule C Weld Heat 5P6756 in SSP Capsule C 0.293 23.6 17.8 35.5 Notes:

1. SeeTableB-11-3.~T~.

2. Predicted shift= CF x FF, where CF is a Chemistry Factor taken from tables from USN RC Reg. Guide 1.99, Rev. 2, based on each material's Cu/Ni content, and FF is Fluence Factor, f28-0. 10 IDll', where f = fluence (101' n/cm2 , E > 1.0 MeV).

3. Margin= 2...J(a12 +a}, where 0 1 =the standard deviation on initial RTNDT (which is taken to be C>°F), and a is the standard deviation on ARTNDT (28°F for welds and 17°F for base materials, except that a need not exceed 0.50 times the mean value of ARTNDT). Thus, margin is defined as 34°F for plate materials and 56°F for weld materials, or margin equals shift (whichever is less), per Reg. Guide 1.99, Rev. 2.

B-11-3

ISP Weld Heat Evaluations Table B-11-5 Percent Decrease in Upper Shelf Energy (USE) for Weld Heat 5P6756 Measured RG 1.99 Rev. 2 Capsule Fluence Cu Content Material Decrease in Predicted Decrease in Identity (x1018 n/cm 2) (wt%)

USE1 (%) USE2 (%)

RB 183 ° Weld Heat 5P6756 in River Bend 1.16 0.06 19.2 12.0 SSP Capsule F Weld Heat 5P6756 in SSP Capsule F 1.9364 0.06 24.0 13.6 SSP Capsule H Weld Heat 5P6756 in SSP Capsule H 1.5766 0.06 19.0 12.9 SSP Capsule C Weld Heat 5P6756 in SSP Capsule C 0.293 0.06 --3

8.7 Notes

1. See Table B-11-3, (Change in USE)/(Unirradiated USE).

2. Calculated using equations in Regulatory Guide 1.162 [B-11-7] that accurately model the Charpy upper shelf energy decrease curves in Regulatory Guide 1.99, Revision 2

3. Value less than zero.

B-11-4

ISP Weld Heat Evaluations Credibility of Surveillance Data The credibility of the surveillance data is determined according to the guidance of Regulatory Guide 1.99, Rev. 2 and 10 CFR 50.61, as supplemented by the NRC staff [Ref. B-11-8]. The following evaluation is based on the available surveillance data for irradiated weld heat 5P6756.

The applicability of this evaluation to a particular BWR plant must be confirmed on a plant-by-plant basis to verify there are no plant-specific exceptions to the following evaluation.

Per Regulatory Guide 1.99, Revision 2 and 10 CFR 50.61, there are 5 criteria for the credibility assessment.

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

In order to satisfy this criterion, the representative surveillance material heat number must match the material in the vessel.

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

Plots of Charpy energy versus temperature for the unirradiated and irradiated condition are presented in this sub-appendix. Based on engineering judgment, the scatter in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper shelf energy. Hence, this criterion is met.

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice that value. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82

[Reference B-11-9].

For weld material 5P6756, there are 4 surveillance capsule data sets currently available. The functional form of the least squares fit method as described in Regulatory Position 2.1 is utilized to determine a best-fit line for this data and to determine ifthe scatter of these ARTNDT values about this line is less than 28°F for welds. Figure B-11-6 presents the best-fit line as described in Regulatory Position 2.1 utilizing the shift prediction routine from CVGRAPH, Version 5.0.2.

The scatter of ARTNDT values about the functional form of the best-fit line drawn as described in Regulatory Position 2.1 is presented in Table B-11-6.

B-11-5

ISP Weld Heat Evaluations Table B-11-6 Best Fit Evaluation for Surveillance Weld Heat 5P6756

<17 ° F Measured Best Fit Fitted Scatter of (Base Metal)

Material Capsule FF ARTNDT ARTNDT CF (° F) ° O

ARTNDT ( F) <28 ° F (30 ft-lb) ( F) ( O F)

(Weld metal)

RB 183 0.447 53.70 (( (E))) (( (E)))

Yes SSP F 0.562 61.90 (( (E)))

(( (E)))

Yes 5P6756 (( (E)))

SSPH 0.514 63.70 (( (E)))

(( (E)))

Yes SSPC 0.217 23.6 (( (E))) (( (E)))

Yes Table B-11-6 indicates that the scatter is within acceptable range for credible surveillance data. Therefore, weld heat 5P6756 meets this criterion.

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within+/ - 25 ° F.

BWRVIP-78 [Reference B-11-10] established the similarity of BWR plant environments in the BWR fleet. The annulus between the wall and the core shroud in the region of the surveillance capsules contains a mix of water returning from the core and feedwater. Depending on feedwater temperature, this annulus region is between 525 ° F and 535 ° F. This location of specimens with respect to the reactor vessel beltline is designed so that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperature will not differ by more than 25 ° F. Any plant-specific exceptions to this generic analysis should be evaluated.

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

Few ISP capsules contain correlation monitor material. Generally, this criterion is not applicable.

For weld heat 5P6756, these criteria are satisfied (or not applicable). The surveillance data are nominally credible because the scatter criterion is met. Prior to application of the data, a plant should verify that no plant-specific exceptions to these criteria exist.

B-11-6

ISP Weld Heat Evaluations Table B-11-7 Unirradiated Charpy V-Notch Results for Surveillance Weld 5P6756 SoeclD Temo C°F CVN Cft-lb' LE Cmilsl %Shear 1 -100 7.5 0.0 14 2 -80 22.0 13.5 16 3 -60 43.0 27.5 26 4 -60 32.5 23.0 29 5 -40 47.0 30.5 34 6 -20 54.5 40.5 28 7 0 53.5 35.5 51 8 20 72.5 52.0 69 9 40 75.5 56.0 72 10 60 70.0 55.0 66 11 60 88.0 66.0 90 12 100 102.0 78.0 100 13 180 102.0 77.0 100 14 300 106.0 78.5 100 15 400 107.5 78.0 100 Table B-11-8 Charpy V-Notch Results for 5P6756 in RB 183° Capsule Spec ID Temp(0 F) CVN {ft-lb) LE {mils) %Shear W2 -42.52 21.43 20 21.5 W5 -41.98 36.74 33 28.7 W8 11.3 25.95 24 31.6 W6 11.48 27.43 23 31.8 W1 37.04 68.18 52.5 66.1 W10 37.22 48.05 41 55.6 W4 69.62 63.52 53.5 62.7 W9 69.98 66.41 58 79.3 W11 127.76 79.78 72 100 W3 131.18 80.22 71 93.3 W12 159.26 77.2 65.5 97.1 W7 199.76 96.33 78 100 B-11-7

ISP Weld Heat Evaluations Table B-11-9 Charpy V-Notch Results for 5P6756 in SSP Capsule F Spec ID Temp (°FJ CVN (ft-lb) LE (mils) %Shear FP272G -60 15.0 8.0 15 FP272H -30 7.5 2.0 15 FP272C 0 37.5 26.0 45 FP272J 0 39.5 29.0 45 FP272I 50 55.0 44.0 65 FP272A 70 55.5 45.0 70 FP272B 150 78.0 66.0 95 FP272D 200 79.5 63.0 100 FP272E 300 80.0 74.0 100 FP272F 400 79.5 69.0 100 Table B-11-10 Charpy V-Notch Results for 5P6756 in SSP Capsule H Spec ID Temp (°FJ CVN (ft-lb) LE (mils) %Shear 30004 -25 9.5 10.0 10 30005 300 88.0 64.0 100 30006 200 82.0 73.0 98 30007 0 38.5 31.0 15 30008 50 66.5 53.0 80 30009 100 77.0 54.0 98 30010 150 81.0 56.0 99 30011 250 89.5 65.0 100 30012 25 53.0 43.0 50 30013 400 90.0 68.0 100 Table B-11-11 Charpy V-Notch Results for 5P6756 in SSP CapsuleC Spec ID Temp (°Fl CVN (ft-lbl LE (mils) %Shear CP2-72-6 -80.14 14.11 12.0 14.2 CP2-72-5 -50.44 22.46 21.0 22.8 CP2-72-9 -40.36 38.92 32.0 26.5 CP2-72-8 -20.20 36.38 32.0 33.8 CP2-72-10 0.14 54.52 44.0 47.9 CP2-72-7 19.58 60.79 49.0 54.8 CP2-72-1 67.28 77.81 59.0 76.8 CP2-72-2 149.18 102.51 71.5 100 CP2-72-3 300.38 112.54 81.0 100 CP2-72-4 399.74 117.01 70.0 100 B-11-8

ISP Weld Heat Evaluations Tanh Curve Fits of CVN Test Data for Weld Heat 5P6756 WELD HEAT 5P6756 (RB AND SSP)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/07/2003 04:46 PM Page I Coefficients of Curve I A= 53.45 B = 50.95C=106.52 TO= -14.17 D = O.OOE+OO Equation is A+ B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= I 04.4(Fixed) Lower ShelfEnergy=2.5(Fixed)

Temp@30 ft-lbs=-67 .1 Deg F Temp@SO ft-lbs=-21.3 Deg F

-r- -i---1, Plant: RIVER BEND AND SSP Material: SAW Heat: 5P6756 300r--1_1___ 1_*-*-r_1- ---,--_l_T_

Orientation: NA Capsule: UNIRRA Fluence: 0.0 n/cm"2 250 t----t- ' ' ---+-- -- --+-- -~

i J-i--t --r--T---1 i.

I I I I, I I I , I I I 200 W~ t-1. --+ .---T-------l-----1----t---t-\

- r--~:-

150 1 1

~ 100 1--i----~---- .--

I '  ! I

---~------- --i I '

1 50 r--t--

t I I

I  !

I

---

1' 0

_J_

--- I I I 1, I

• I 0 ~=::t== -~-+-~-L

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential

-100.00 7. 50 19. 4 5 -11.95

- 8 0. 00 22. 00 25. 44 - 3. 4 4

- 60. 00 43. 00 3 2. 79 I 0. 2 I

- 60. 00 3 2. 50 3 2. 7 9 - . 29

- 40. 00 47. 00 41. 3 3 5. 6 7

- 20. 00 54. 50 50.66 3. 8 4

. 00 5 3. 50 60. 19 - 6. 69

20. 0 0 72. 50 69. 2 5 3. 2 5

40. 0 0 7 5. 50 77. 3 3 - l. 8 3 Figure B-11-1 Charpy Energy Data for Weld 5P6756 Unirradiated B-11-9

ISP Weld Heat Evaluations WELD HEAT 5P6756 (RB AND SSP}

Page 2 Plant: RIVER BEND AND SSP Material: SAW Heat: 5P6756 Orientation: NA Capsule: UNlRR.A. Fluence: 0.0 n/cmA2 CharpyV-Notch Data Temperature InputCVN Computed CVN Differential

60. 00 70. 00 84. 12

• 14. 12

60. 00 88. 00 84. 12 3. 8 s I 00, 00 102. 00 93. 71 8. 29 I 80. 00 1-02. 00- 10 t . g 1 . 19 300.00 106. 00 i 04. 1 2 l. S8 400. 00 I 07. 50 I 04. 3 6 3. 14 Correlation C()efficient"" .977 Figure B-11-1 Charpy Energy Data for Weld 5P6756 Unirradiated (Continued)

B-11-10

ISP Weld Heat Evaluations IRRADIATED WELD HEAT 5P6756 (RB-183)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2003 02:44 PM Page 1 Coefficients of Curve l A = 43.45 B = 40.95 C = 92.61 TO = 18.09 D = O.OOE+oO Equation is A+ 8 * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=84.4(Fixed) Lower Shelf Energy=2.5(Fixed)

Temp@30 ft-lbs=-13.4 Deg F Temp@50 ft-lbs=33.l Deg F Plant: River Bend Material: SAW Heat: 5P6756 Orientation: NA Capsule: RB-183 Fluence: l.l 6E+ 18 n/cm"2

~ I 200 +---- ----+----- -+

0 if El 150 w

! I z

() 100 --- -J- - --1--------*--------'---.-- ___ ,

50 I

I 0 k==+/-=-:-::::J l-

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential

-42.52 2 1. 43 19. 92 1. 51

- 4 1. 9 8 3 6. 74 20.08 16. 66

11. 3 0 2 5. 95 4 0. 4 5 - 14. 50

11. 4 8 2 7. 43 4 0. 5 3 - 13. 10 3 7. 04 6 8. 18 5 I. 71 16. 47 3 7. 22 4 8. 05 51. 79 - 3. 74

69. 62 6 3. 52 64. 1 4 -. 62

69. 9 8 66. 41 64. 26 2. l5 12 7. 7 6 79. 78 7 7. 3 9 2. 39 1 3 !. 1 8 8 0. 22 7 7. 8 5 2. 37 159. 2 6 7 7. 20 8 0. 6 9 - 3. 49 Figure B-11-2 Charpy Energy Data for Weld 5P6756 in RB 183° Capsule B-11-11

ISP Weld Heat Evaluations IRRADIATED WELD HEAT 5P6756 (RB-183)

Page 2 Plant: River Bend Material: SAW Heat: 5P6756 Orientation'. NA Capsule: RB-183 Fluence: 1.16E+18 n/cml'-2 Charpy V-Notch Data Temperature lnputCVN Com11uted CVN Differential 199. 76 96. 33 s 2. 81 1 3. 5 2 Con*elation Coefficient"" .914 Figure B-11-2 Charpy Energy Data for Weld 5P6756 in RB 183' Capsule (Continued)

B-11-12

ISP Weld Heat Evaluations IRRADIATED WELD HEAT 5P6756 (SSP-F)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 0412212003 01 :22 AM Page l Coefficients of Curve 1 A= 40.9 B = 38.4 C = 83.86 TO= 19.2 D = O.OOE+OO Equation is A+ B

• fTanh((T-To)l(C+DI)}]

Upper ShetfEnergy=79.3(Fixed) Lower ShelfEnergy=2.5(Fixed)

Temp@30 ft-lbs~S.2 Deg f Temp@50 ft-lbs=39.5 Deg f Heat: 5P6756 r-1**

Plant: RIVER BEND AND SSP Material: SAW Orientation; NA Capsule: SSP-F Fluence: 1.9364E+l 8 n/cm11.2 300 II I Ii I

I I 250 '! ' '

I r

~

i 200 -*-------

tT

' I I

I

~ 150 i i

~ I I I

z I

() 100 -

I

~

I~

50 -- .r6' I i

'i v I i

I 0

I

~  !

i

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature lnputCVN Computed CVN Differential

• 60.00 15. 00 12. 59 2. 41

- 30. 00 7. 50 20. 64 - l3. I4

. 00 37. 50 3 2' 26 5.24

' 00 39. 50 3 2' 26 7. 24

50. 00 55' 00 54. 40 . 60 7{), 00 5 5. 50 6 l. 68

• 6. 18

! 50. 00 78' 00 76. 05 I. 95 200. 00 79' 50 7 8' 28 I. 22 3 00' 00 8 0' 0 0 79, 2 I ' 79 Figure B-11-3 Charpy Energy Data for Weld 5P6756 In SSP Capsule F B-11-13

ISP Weld Heat Evaluations IRRADIATED WELD BEAT SP6756 (SSPF)

Page 2 Ph,mt: RIVER BENO AN'D SSP Materi'alz SAW Heat: 5P6756 Orientation: NA Capsule: SSP-F Fluence: 1.93ME+1S n/cmA2 Charpy V-NQtch Data Temperature lnputCVN Compqted CVN Differmuial 400.00 79.50 79.29 . 21

.Correlation Coeffieient = .977 Figure B-11-3 Charpy Energy Data for Weld 5P6756 in SSP Capsule F (Continued)

B-11-14

ISP Weld Heat Evaluations Weld heat 5P6756 in SSP-H CV GRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2-006 03:59 PM Page t Coefficients; of Curve l A= 43.55 B ~ 41.05 C = 51.24 TO= 14.12 D = O.OOE+OO Equation ls A. ** 13 * [fanh((f*To)/(C_,,..DT})j t;pper ShdfEnerro-=84.6(Fixed) Lower ShelfEnergy---o25(Fixed)

Tt-mp~JO ft-lhs.-"-3.4 Deg F Temp@50 ft-Jbs,=22.3 Deg .F Plant: RIVER BE'SD AND SSP M;it~riat SA\V Heat: 5P6756 Orientation: 1\A Capsule: SSP-H Fluence: l.5766E+18nfcmA2 300 - - - ---- ,--

250

~ 200 - -----

ff.

e; 150 w

71' z

(,)

100 i 0 50 i 0 ~~-~~--cc: - 0 j j __ _

• 300 *200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Tempernture lnputCVN Computed CVK Differemial

• :25. 00 300. 00

9. 50

&8. 00

17. l 5

84. 60 ' 65

3. 40 200. 00 S2 00 84. 54 2. 54

.00 3 8' ::.o 32' 52 5' 98 5 0. GO 66. 50 68. 37 - l. 87 100. 00 77. 00 81 - S2

• 4 - 82 l 5 0' 00 8 1. 00 84. 19 - 3. l 9 2 5 0. 00 89. 50 84. 59 4. 91

25. 00 5 3. 00 52 I 4 - 86 Figure B-11-4 Charpy Energy Data for Weld 5P6756 In SSP Capsule H B-11-15

ISP Weld Heat Evaluations Weld heat 5P6756 in SSP"'.H Page 2 Plant: RIVER BEND AND SSP Material: SAW Heat: 5P6756 Orientation: NA Capsule: SSPH Fluence: 1.5766E+ 18 nforn"2 Charpy V-Notch Data Temperature InputCVN Computed CVN Differential 400, 00 90,00 84. 60 5.40 Correlation Coefficient= .985 Figure B-11-4 Charpy Energy Data for Weld 5P6756 in SSP Capsule H (Continued)

B-11-16

ISP Weld Heat Evaluations Weld Heat 5P6756 in SSP Capsule C CV GRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 06/07/2006 01 :26 PM Page 1 Coefficients of Curve 1 A= 56.6 B = 54.1C=106.54 TO= 13.75 D = O.OOE+OO Equation is A+ B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=] 10. ?(Fixed) Lower Shelf Encfb'Y=2.5(Fixed)

Temp@30 ft-lbs=-43.5 Deg F Temp@50 ft-lbs=.7 Deg F Plant: COOPER Material: SAW Heat: 5P6756 Orientation: NA Capsule: SSP C Fluence: NIA n/cm"2 300 250 U)

£? 200 0

0 u..

2l 150 (1) c:

w 0 z

> 100 u

50 0

0 0

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Ternperaturc lnputCVN Computed CVN Differential

- 8 0. 14 14. I1 1 8. 35 - 4. 24

- 5 0. 44 22. 46 2 7. 45 - 4. 99

- 40. 36 3 8. 92 3 l. 26 7. 66

- 20. 20 3 6. 38 3 9. 92 - 3 . 54

. 14 54. 52 4 9. 72 4. 80

19. 58 60. 79 5 9. 55 1. 24

67. 28 7 7. 8I 8 l. 70 - 3. 89 149. l8 102. 51 I 0 2. 80 29 3 0 0. 38 I 12. 54 1 1 0. 19 2. 3 5 Figure B-11-5 Charpy Energy Data for Weld 5P6756 in SSP Capsule C B-11-17

ISP Weld Heat Evaluations Weld Heat 51'6756 in SSP Capsule C Puge 2 Plant: COOPER ~tuteria1: S1\ \V Rem: 5P6756 Orientation: .\A Capsule: SSP C Fluence: N/A Charpy V-Notch Data Tcmpi;rati.m::. lnpu;CV\ Cmnpu\i:J CV'\ f)itforential I 1 7. 0 I 1 ro 6 l 6 4(1 Ct>rrda1.i1m C,;efficien1 - .993 Figure B-11-5 Charpy Energy Data for Weld 5P6756 In SSP Capsule C (Continued)

B-11-18

ISP Weld Heat Evaluations

((

(E)))

B-11-19

ISP Weld Heat Evaluations References B-11-1. BWRVIP-87NP, Revision 1: BWR Vessel and Internals Project Testing and Evaluation of BWR Supplemental Surveillance Program Capsules D, G, and H. EPRI, Palo Alto and BWRVIP: 2010. 1021553.

B-11-2. "River Bend 183 Degree Surveillance Capsule Report," M.P. Manahan Sr., MPM Report Number MPM-1202971, January 2003.

B-11-3. BWRVIP-11lNP, Revision 1: BWR Vessel and Internals Project, Testing and Evaluation of BWR Supplemental Surveillance Program Capsules E, F and I. EPRI, Palo Alto, CA: 2010. 1021554.

B-11-4. CVGRAPH, Hyperbolic Tangent Curve Fitting Program, Developed by ATI Consulting, Version 5.0.2, Revision 1, 3/26/02.

B-11-5. "Progress Report on Phase 2 of the BWR Owners' Group Supplemental Surveillance Program," T.A. Caine, S. Ranganath, and S.J. Stark, GE Nuclear Energy, GE-NE-523-99-0792, January 1992.

B-11-6. "Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.

B-11-7. "Format and Content of Report for Thermal Annealing of Reactor Pressure Vessels,"

USNRC Regulatory Guide 1.162, February 1996.

B-11-8. K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.

I B-11-9. ASTM E-185, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, July 1982.

B-11-10. BWR Vessel and Internals Project: BWR Integrated Surveillance Program Plan (BWRVIP-78). EPRI, Palo Alto, CA and BWRVIP: 1999. TR-114228.

B-11-11. Not used.

B-11-12. BWRVIP-169NP: BWR Vessel and Internals Project, Testing and Evaluation of BWR Supplemental Surveillance Program (SSP) Capsules A, B, and C. EPRI, Palo Alto, CA: 2010. 1021556.

B-11-20

Best Estimate Chemistry Values for Selected BWR Vessel Plates and Welds

((

(E)))

D-14

Best Estimate Chemistry Values for Selected BWR Vessel Plates and Welds

((

(E)))

D-15

Best Estimate Chemistry Values for Selected BWR Vessel Plates and Welds

((

(E)))

D-22

Best Estimate Chemistry Values for Selected BWR Vessel Plates and Welds

((

(E)))

D-23

Best Estimate Chemistry Values for Selected BWR Vessel Plates and Welds

((

(E)))

D-25

ATTACHMENT 9 Limerick Generating Station Unit 1 Reactor Pressure Vessel Fluence Evaluation - Non-Proprietary Report, LIM-FLU-002-R-009, Rev. 1 (Non-Proprietary)

LlM-FLU-002-R-009 Revision I Page i of xi Topical Report LIMERICK GENERATING STATION UNIT 1 REACTOR PRESSURE VESSEL FLUENCE EVALUATION - NON-PROPRIETARY REPORT Document Number: LIM-FLU-002-R-009 Revision 1 September 2020 Prepared by: TransWare Enterprises Inc.

1565 Mediterranean Drive Sycamore, Illinois 60178 Prepared for: Exelon Generation Company, LLC I

Limerick Generating Station Unit 1 3146 Sanatoga Road Pottstown, Pennsylvania 19464 Project Manager: Michelle Karasek Controlled Copy Number: 2 This document represents the non-proprietary version of the Trans Ware Enterprises Inc. document number LIM-FLU-00 l-R-005, Revision l. Proprietary Information removed from LIM-FLU-00 l-R-005, Revision I is identified in this report by enclosure in double brackets.

trans * - -

ENTERPRISES 1565 Mediterranean Dr Sycamore, Illinois 601 78-3141 815 -895-4 700

• www .transware.net

LIM-FLU-002-R-009 Revision 1 Page ii of xi

[ This page intentionally left blank. ]

LIM-FLU-002-R-009 Revision l Page iii of xi Topical Report LIMERICK GENERATING STATION UNIT 1 REACTOR PRESSURE VESSEL FLUENCE EVALUATION - NON-PROPRIERARY REPORT Document Number: LIM-FLU-002-R-009 Revision 1 September 2020 Preparing Organization: TransWare Enterprises Inc.

Prepared By: 'th:fuzo 1

K. E. Watkins, Project Engineer Oat Reviewed By:

Cfj~M>

Dae 9/ (I- (2~'2~

K. A. Jo;QASPecialist Date Approved By: ~All;z.ll I te Prepared For: Exelon Generation Company, LLC Limerick Generating Station Unit 1 3146 Sanatoga Road Pottstown, Pennsylvania 19464 Project Manager: Michelle Karasek This document represents the non-proprietary version of the TransWare Enterprises Inc.

document number LIM-FLU-001-R-005, Revision 1. Proprietary Information removed from LIM-FLU-00 l-R-005, Revision l is identified in this report by enclosure in double brackets.

TransWare Enterprises Inc.* 1565 Mediterranean Dr.* Sycamore, Illinois 60178-3141 *USA

+1-815-895-4700

• www.transware net

LIM-FLU-002-R-009 Revision 1 Page iv of xi DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THE INFORMATION CONTAINED IN THIS REPORT IS BELIEVED BY TRANSWARE ENTERPRISES INC. TO BE AN ACCURATE AND TRUE REPRESENTATION OF THE FACTS KNOWN, OBTAINED OR PROVIDED TO TRANSWARE ENTERPRISES INC. AT THE TIME THIS REPORT WAS PREPARED. THE USE OF THIS INFORMATION BY ANYONE OTHER THAN THE CUSTOMER OR FOR ANY PURPOSE OTHER THAN THAT FOR WHICH IT IS INTENDED, IS NOT AUTHORIZED; AND WITH RESPECT TO ANY UNAUTHORIZED USE, TRANSWARE ENTERPRISES INC. MAKES NO REPRESENTATION OR WARRANTY AND ASSUMES NO LIABILITY AS TO THE COMPLETENESS, ACCURACY OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS DOCUMENT. IN NO EVENT SHALL TRANSWARE ENTERPRISES INC. BE LIABLE FOR ANY LOSS OF PROFIT OR ANY OTHER COMMERCIAL DAMAGE, INCLUDING BUT NOT LIMITED TO SPECIAL, CONSEQUENTIAL OR OTHER DAMAGES.

((

))

QUALITY REQUIREMENTS This document has been prepared in accordance with the requirements of 10CFR50 Appendix B, 10CFR21, and TransWare Enterprises Inc.s 10CFR50 Appendix B quality assurance program.

LIM-FLU-002-R-009 Revision 1 Page v of xi ACKNOWLEDGEMENTS TransWare Enterprises Inc. wishes to acknowledge Michelle Karasek of Exelon Generation Company, LLC for her management of the project. TransWare Enterprises Inc. also wishes to acknowledge Marcus Gergar and Scott Foster of Exelon Generation Company, LLC for their support and assistance in providing the mechanical design and operating data for this reactor pressure vessel fluence evaluation.

LIM-FLU-002-R-009 Revision 1 Page vi of xi

[ This page intentionally left blank. ]

LIM-FLU-002-R-009 Revision 1 Page vii of xi CONTENTS 1 Introduction.................................................................................................................. 1-1 1.1 Regulatory Requirements ..................................................................................... 1-2 1.2 Limitations of the Fluence Evaluation ................................................................... 1-3 1.3 Quality Assurance ................................................................................................ 1-3 2 Summary of Results .................................................................................................... 2-1 3 Description of the Reactor System............................................................................. 3-1 3.1 Overview of the Reactor System Design .............................................................. 3-1 3.2 Reactor System Mechanical Design Inputs........................................................... 3-3 3.3 Reactor System Material Compositions ................................................................ 3-3 3.4 Reactor Operating Data Inputs ............................................................................. 3-5 3.4.1 Core Configuration and Fuel Design ......................................................... 3-5 3.4.2 Reactor Power History .............................................................................. 3-5 3.4.3 Reactor Statepoint Data ............................................................................ 3-7 3.4.4 Reactor Coolant Properties ....................................................................... 3-9 4 Methodology ................................................................................................................ 4-1 4.1 Computational Method.......................................................................................... 4-1 4.2 Fluence Model ...................................................................................................... 4-2 4.2.1 Geometry Model ....................................................................................... 4-5 4.2.2 Reactor Core and Core Reflector .............................................................. 4-6 4.2.3 Reactor Core Shroud ................................................................................ 4-6 4.2.4 Downcomer Region .................................................................................. 4-7 4.2.4.1 Jet Pumps................................................................................. 4-7 4.2.4.2 Surveillance Capsules .............................................................. 4-7 4.2.5 Reactor Pressure Vessel .......................................................................... 4-8

(( ))

4.2.6 Thermal Insulation .................................................................................... 4-8 4.2.7 Inner and Outer Cavity Regions ................................................................ 4-8 4.2.8 Biological Shield Model ............................................................................. 4-8

LIM-FLU-002-R-009 Revision 1 Page viii of xi 4.2.9 Above-Core Components .......................................................................... 4-9 4.2.9.1 Top Guide ................................................................................. 4-9 4.2.9.2 Core Spray Spargers and Piping............................................... 4-9 4.2.10 Below-Core Components .......................................................................... 4-9 4.2.10.1 Core Support Plate and Rim Bolts ............................................ 4-9 4.2.10.2 Fuel Support Pieces.................................................................. 4-9 4.2.10.3 Control Blades and Guide Tubes ............................................ 4-10 4.2.11 Summary of the Geometry Modeling Approach ....................................... 4-10 4.3 Particle Transport Calculation Parameters.......................................................... 4-11 4.4 Fission Spectrum and Neutron Source ............................................................... 4-11 4.5 Parametric Sensitivity Analyses .......................................................................... 4-12 5 Reactor Pressure Vessel Fast Neutron Fluence ........................................................ 5-1 6 Reactor Pressure Vessel Fluence Uncertainty Analysis ........................................... 6-1 6.1 Comparison Uncertainty ....................................................................................... 6-1 6.1.1 Operating Reactor Comparison Uncertainty .............................................. 6-1 6.1.2 Benchmark Comparison Uncertainty ......................................................... 6-2 6.2 Analytic Uncertainty .............................................................................................. 6-2 6.3 Combined Uncertainty .......................................................................................... 6-3 7 References ................................................................................................................... 7-1 7.1 References ........................................................................................................... 7-1 7.2 Glossary ............................................................................................................... 7-3

LIM-FLU-002-R-009 Revision 1 Page ix of xi LIST OF FIGURES Figure 3-1 Planar View of the Limerick 1 Reactor at the Core Mid-Plane Elevation .............. 3-2 Figure 4-1 Planar View of the Limerick 1 Fluence Model at the Core Mid-Plane Elevation in Quadrant Symmetry ......................................................................... 4-3 Figure 4-2 Axial View of the Limerick 1 Fluence Model ......................................................... 4-4 Figure 5-1 Limerick 1 RPV Beltline Region at 57 EFPY ........................................................ 5-2 Figure 5-2 Nozzle Fluence Edit Locations for Sample Nozzle ............................................... 5-3

LIM-FLU-002-R-009 Revision 1 Page x of xi

[ This page intentionally left blank. ]

LIM-FLU-002-R-009 Revision 1 Page xi of xi LIST OF TABLES Table 2-1 Maximum Fast Neutron Fluence for Limerick 1 RPV Beltline Welds, Nozzles, and Shell Plate Locations ...................................................................... 2-2 Table 2-2 RPV Beltline Elevation Range for Limerick 1 ....................................................... 2-3 Table 3-1 Summary of Material Compositions by Component Region for Limerick 1 ........... 3-4 Table 3-2 Summary of Limerick 1 Core Loading Inventory .................................................. 3-6 Table 3-3 Statepoint Data for Limerick 1 per Cycle Basis .................................................... 3-8 Table 5-1 Maximum Fast Neutron Damage Fluence for Limerick 1 RPV Beltline Welds at EOC 17 (28.6 EFPY) ....................................................................................... 5-4 Table 5-2 Maximum Fast Neutron Fluence for Limerick 1 RPV Beltline Welds at 57 EFPY .............................................................................................................. 5-5 Table 5-3 Maximum Fast Neutron Damage Fluence for Limerick 1 RPV Beltline Shell Plates .................................................................................................................. 5-6 Table 5-4 Maximum Fast Neutron Damage Fluence for Limerick 1 RPV Beltline Nozzles at EOC 17 (28.6 EFPY) .......................................................................... 5-6 Table 5-5 Maximum Fast Neutron Damage Fluence for Limerick 1 RPV Beltline Nozzles at 57 EFPY............................................................................................. 5-7 Table 5-6 Reactor Beltline Elevation Range for Limerick 1 .................................................. 5-7 Table 6-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements ....... 6-2 Table 6-2 Limerick 1 Combined RPV Uncertainty for Energy > 1.0 MeV ............................. 6-3

LIM-FLU-002-R-009 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-009 Revision 1 Page 1-1 of 1-3 1

INTRODUCTION This report presents the results of the reactor pressure vessel fast neutron fluence evaluation that was performed for the Limerick Generating Station Unit 1 Unit 1 (Limerick 1) reactor. The Limerick 1 reactor is owned and operated by Exelon Generation Company, LLC (Exelon).

The fast neutron fluence presented in this report was determined in accordance with the guidelines and requirements presented in U. S. Nuclear Regulatory Commission (NRC)

Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [1]. Fluence is presented for the 0T, 1/4T and 3/4T depths in the reactor pressure vessel (RPV) plates, welds, and nozzles throughout the RPV extended beltline region determined at 60 years of reactor operation. The damage fluence determined at the 1/4T and 3/4T depths in the RPV wall were determined using the displacements-per-atom (dpa) attenuation method prescribed in U. S. NRC Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials [2].

The Limerick 1 fluence evaluations provided in this report are based on historical operating conditions for the reactor and projected operation through the end of licensed operation. The RPV fluence evaluations are presented at the End of Cycle (EOC) 17 and projected to 57 Effective Full Power Years (EFPY) of reactor operation.

The fluence evaluations are performed based on the RAMA Fluence Methodology [3]. Under funding from Electric Power Research Institute, Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP), the RAMA Fluence Methodology was developed by TransWare Enterprises Inc. for the purpose of calculating fast neutron fluence in nuclear reactor pressure vessels and reactor vessel internal components. The RAMA Fluence Methodology (hereafter referred to as RAMA) has received generic approval [4] from the U.S. NRC for determining fast neutron fluence in BWRs and PWRs in accordance with the requirements of Regulatory Guide 1.190. ((

))

In compliance with Regulatory Guide 1.190, TransWare Enterprises Inc. (TransWare) has benchmarked the RAMA Fluence Methodology against industry standard benchmarks and plant-specific dosimetry measurements for boiling water reactors and pressurized water reactors. The results of the benchmarking show that the fluence methodology implemented by TransWare predicts specimen activities with no discernable bias in the computed fluence. ((

))

LIM-FLU-002-R-009 Revision 1 Page 1-2 of 1-3 1.1 Regulatory Requirements Part 50 of Title 10 of the Code of Federal Regulations provides requirements for establishing irradiated material monitoring programs that serve to ensure the integrity of the reactor coolant pressure boundary of light water nuclear power reactors. Two appendices to Part 50 present the requirements that guide the fluence determinations presented in this report: Appendix G, Fracture Toughness Requirements [6], and Appendix H, Reactor Vessel Material Surveillance Program Requirements [7].

Appendix G specifies fracture toughness requirements for the carbon and low-alloy ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary to ensure adequate margins of safety during any condition of normal operation, including anticipated conditions for system hydrostatic testing, to which the pressure boundary may be subjected over its service lifetime. These requirements apply to base metal, welds and weld heat-affected zones in the materials within the reactor pressure vessel beltline region.

Appendix H specifies the requirements for a material surveillance program that serves to monitor changes in the fracture toughness properties of the ferritic materials in the reactor beltline region.

The changes in fracture toughness properties of ferritic materials are attributed to the exposure of the materials to neutron irradiation and the thermal environment.Section III of Appendix H specifies that a material surveillance program is required for light water nuclear power reactors if the peak fast neutron fluence with energy greater than 1 MeV (E > 1 MeV) at the end of the design life of the vessel is expected to exceed 1017 n/cm2.

In compliance with the Appendix H requirements, fracture toughness test data are obtained from material specimens that are exposed to neutron irradiation in surveillance capsules installed at or near the inner surface of the reactor pressure vessel. These capsules are withdrawn periodically from the reactor for measurement and analysis. Fast neutron fluence is not a measurable quantity and must be determined using analytical methods. It must be demonstrated that the analytical method used to determine the fast neutron fluence provides a conservative prediction over the beltline region of the pressure boundary when compared to the measurement data with allowances for all uncertainties in the measurements.Section III of Appendix H also allows for an Integrated Surveillance Program (ISP) in which representative materials for the reactor are irradiated in one or more other reactors of sufficiently similar design and operating features to permit accurate comparisons of the predicted amount of radiation damage.

Implementing guidelines addressing the requirements of Appendices G and H are provided in U. S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials [2], and Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [1]. Regulatory Guide 1.99 addresses the requirements of Appendix G for determining the damage fluence that is used in the evaluation of fracture toughness in light water nuclear reactor pressure vessel ferritic materials. Regulatory Guide 1.190 addresses the requirements for determining the fast neutron fluence and uncertainty in the fluence predictions that are used in fracture toughness evaluations.

The fast neutron fluence evaluations described in this report meet the requirements of Appendices G and H of Part 50 of Title 10 of the Code of Federal Regulations and U. S. NRC Regulatory Guides 1.190 and 1.99 Revision 2.

LIM-FLU-002-R-009 Revision 1 Page 1-3 of 1-3 1.2 Limitations of the Fluence Evaluation The fast neutron fluence presented in this report is based on historical and projected operating conditions of the reactor. The RPV fast neutron fluence that is based on historical operating conditions is determined to meet the requirements of Regulatory Guide 1.190 with no discernable bias in the results. It is determined, therefore, that the RPV fast neutron fluence presented in this report is suitable for use in evaluating material embrittlement conditions of reactor pressure vessel materials in accordance with Regulatory Guide 1.99. Use of the results for other purposes is not demonstrated.

Fluence projections are determined using the most up-to-date cycle operating data available at the time of this report. The projection cycle is assumed to be an equilibrium cycle representative of how the reactor will operate until the end of the reactors operating license. Continued use of the projected fluence presented in this report must be demonstrated as applicable as new operating history data from the reactor becomes available. Deviations from the design basis of this analysis, including changes in future fuel designs, core loadings, and operating strategies that result in a significant change to the core power shapes relative to the projected data may require re-evaluation to determine the impact of the altered flux profiles on the projected RPV and RVI component fluences.

It is cited in Regulatory Position 1.2 of Regulatory Guide 1.190 that a best-estimate power distribution may be used for reactor vessel neutron fluence calculations. The best-estimate fluence presented in this report meets the requirements of Regulatory Position 1.2. Regulatory Position 1.2 further states that if a best-estimate is used, the power distribution must be updated if changes in core loadings, surveillance measurements, or other information indicate a significant change in projected fluence values. Under this requirement, the other information that can necessitate an update of the fluence model can include: implementation of power uprates/derates, introduction of new fuel designs, changes in projected cycle lengths, changes in core loading and/or operational strategies, changes in reactor flow, or other changes that could alter the power/flux profiles used in the fluence projections and uncertainty analysis.

1.3 Quality Assurance The fluence evaluations presented in this report were performed in compliance with the quality assurance requirements of Appendix B to Part 50 of Title 10 of the Code of Federal Regulations, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants (10CFR50 Appendix B) [8], and to Part 21, Reporting of Defects and Noncompliance (10CFR21) [9].

LIM-FLU-002-R-009 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-009 Revision 1 Page 2-1 of 2-3 2

SUMMARY

OF RESULTS This section provides a summary of the fast neutron fluence determined for the Limerick 1 reactor pressure vessel (RPV). Details of the RPV fluence evaluation are presented in Section 5, Reactor Pressure Vessel Fast Neutron Fluence. Section 6, Reactor Pressure Vessel Fluence Uncertainty Analysis, provides a determination of the uncertainty in the RPV fluence evaluation.

Limerick 1 is a BWR/4 class plant with a core loading of 764 fuel assemblies. The fluence evaluation for this plant is based on historical operating data through Cycle 17 (28.6 EFPY).

Fluence evaluations are also performed at 57 EFPY. Cycle 18 was provided as a partial projection cycle. A projection cycle using an advanced fuel design, GNF3, was also provided and was used for the fluence projection to 57 EFPY.

Table 2-1 presents a summary of the maximum fast neutron fluence determined for the RPV shell plates, welds and nozzles at EOC 17 (28.6 EFPY) and 57 EFPY. The significant fluence for the evaluated areas of the RPV occur at the inside surface of the RPV base metal, which is denoted as the 0T depth in the pressure vessel wall. With the exception of the N2 nozzles, Shell 3, and welds AC, BG, BH, and BJ, all other evaluated areas of the RPV have exceeded the fluence threshold of 1.0E+17 n/cm2. It is shown in Table 2-1 that the maximum fluence is determined to occur at the 0T location of the intermediate shell plate with a value of 1.09E+18 n/cm2 at 57 EFPY. Note in Table 2-1 that all fluence that has exceeded the fluence threshold of 1.0E+17 n/cm2 are shown in red font and that the maximum fluences in the RPV are additionally shown in bold font.

Table 2-2 shows the axial span of the RPV beltline region that was determined for Limerick 1 at EOC 17 (28.6 EFPY) and 57 EFPY. The reactor beltline region is defined in Appendices G [6]

and H [7] of 10CFR50 to include those regions that directly surround the effective height of the reactor core, as well as those adjacent areas of the RPV that are predicted to experience sufficient neutron irradiation damage. This definition of the RPV beltline is considered to include all materials that exceed a fast neutron fluence of 1.0E+17 n/cm2. At 57 EFPY the RPV beltline covers 421.7 cm, or approximately 13.8 ft of the reactor vessel. The scope of the fluence model was developed to provide an evaluation of the reactor pressure vessel over the full height of the RPV extended beltline region.

LIM-FLU-002-R-009 Revision 1 Page 2-2 of2-3 Table 2-1 Maximum Fast Neutron Fluence for Limerick 1 RPV Beltline Welds, Nozzles, and Shell Plate Locations Maximum Fast Neutron Fluence (n/cm2)

Component EOC 17 (28.6 EFPY) 57 EFPY RPV Beltline Welds AB 4.81E+17 8.08E+17 AC 9.40E+15 2.05E+16 BA 3.59E+17 5.77E+17 BB 3.22E+17 5.80E+17 BC 4.20E+17 7.32E+17 BD 4.75E+17 8.89E+17 BE 4.77E+17 8.88E+17 BF 4.54E+17 7.90E+17 BG 7.98E+15 1.81E+16 BH 9.31E+15 2.01E+16 BJ 9.31E+15 2.01E+16 Nozzle Forging-to-Base-Metal Welds Nozzle Weld N2 1.32E+16 2.26E+16 Nozzle Weld N16 1.59E+17 3.34E+17 Nozzle Weld N17 2.37E+17 4.68E+17 Shell Plates Shell 1 4.81E+17 8.08E+17 Shell 2 6.04E+17 1.09E+18 Shell 3 9.40E+15 2.05E+16

LIM-FLU-002-R-009 Revision 1 Page 2-3 of2-3 Table 2-2 RPV Beltline Elevation Range for Limerick 1 Lower Elevation Upper Elevation Axial Span of the RPV Reactor Lifetime

[in (cm)] [in (cm)] Beltline [in (cm)]

EOC 17 (28.6 EFPY) 215.0 (546.1 cm) 369.3 (938.0 cm) 154.3 (391.9 cm) 57 EFPY 210.1 (533.6 cm) 376.1 (955.3 cm) 166.0 (421.7 cm)

Section 5 provides detailed results for the RPV fast neutron fluence evaluation. RPV damage fluence is reported at the OT, 1/4T, and 3/4T depths of the RPV wall for each horizontal (circlllllferential) weld, ve1iical (axial) weld, shell plate, and nozzle in the RPV beltline.

Figure 5-1 illustrates the location of the welds, shell plates, and nozzles in the RPV. Fluence damage through the thickness of the RPV wall is detennined using the displacements-per-atom (dpa) attenuation method prescribed in Regulato1y Guide 1.99 [2].

Section 6 provides an evaluation of the combined unce1iainty detennined for the Limerick 1 fluence evaluation. The combined unce1iainty is detennined by combining the measurement and analytic unce1iainties and is dete1mined to be 11.2%. In accordance with Regulato1y Guide 1.190, there is no discemable bias in the computed fluence; therefore, the computed fluence is the best-estimate fluence.

LIM-FLU-002-R-009 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-009 Revision 1 Page 3-1 of 3-9 3

DESCRIPTION OF THE REACTOR SYSTEM This section provides an overview of the reactor design and operating data inputs that were used to develop the computational fluence model for the Limerick 1 reactor. All reactor design and operating data inputs used to develop the model are plant-specific and were provided by Exelon.

The inputs for the fluence geometry model were developed from nominal and as-built drawings for the reactor pressure vessel, vessel internals, fuel assemblies, and containment regions.

Several modifications were made to the Limerick 1 RAMA geometry model since the previous vessel internal component fluence evaluation performed by TransWare in 2010 [10]. These modifications include:

• Addition of RPV nozzle forgings for the recirculation outlet (N1), recirculation inlet (N2), and LPCI (N17) nozzles.

• Improved shroud head and steam separator standpipe model.

• Improved jet pump rams head, hold down beam, slip join clamp, and auxiliary spring wedge models.

• Improved below core model, including the following discrete components:

o Core support plate and rim bolts, o Orificed fuel support piece and peripheral fuel support piece, and o Control rod drive tubes.

The reactor operating history was also updated to provide a historical accounting of how the reactor operated for Cycles 1 through 17 with projections to end of license.

3.1 Overview of the Reactor System Design Limerick 1 is a General Electric BWR/4 class reactor with a core loading of 764 assemblies. The unit began commercial operation in 1986 with a design rated power of 3293 MWt. In Cycle 7 the rated power was increased to 3458 MWt and during Cycle 14 a rated power of 3515 MWt was achieved. At the time of this fluence evaluation, Limerick 1 has completed 17 cycles of operation.

Figure 3-1 illustrates the basic planar configuration of the Limerick 1 reactor at an axial elevation near the reactor core mid-plane. All the radial regions of the reactor that are required for fluence evaluations are shown. Beginning at the center of the reactor and projecting outward, the regions include: the core region; core reflector region (bypass water); central shroud wall; downcomer water region including the jet pumps; RPV wall; cavity region between the RPV wall and insulation; insulation; cavity region between the insulation and biological shield; and the biological shield wall. Cladding is included on the inner RPV surface as well as the inner and

LIM-FLU-002-R-009 Revision 1 Page 3-2 of 3-9 outer surfaces of the biological shield wall. Also represented in Figure 3-1 are notations indicating the control rod and fuel assembly locations within the core. Note that the fuel locations are shown only for the northeast quadrant of the core region.

Figure 3-1 Planar View of the Limerick 1 Reactor at the Core Mid-Plane Elevation

LIM-FLU-002-R-009 Revision 1 Page 3-3 of 3-9 3.2 Reactor System Mechanical Design Inputs The mechanical design inputs used to construct the Limerick 1 fluence geometry model are based upon nominal design and as-built dimensional information. As-built data is always preferred when constructing plant-specific reactor fluence models; however, as-built data is not always available and nominal dimensions are used.

For the Limerick 1 fluence model, the predominant dimensional information used to construct the fluence model is nominal design data. ((

))

An important component of a computational reactor pressure vessel fluence model is the accurate description of the surveillance capsules installed in the pressure vessel. Figure 3-1 shows that the Limerick 1 reactor was initially equipped with three surveillance capsules. The capsules were installed at an elevation around the reactor core mid-plane. Each capsule was mounted radially near the inside surface (0T) of the RPV wall. The surveillance capsules were distributed around the circumference of the pressure vessel at the 30°, 120° and 300° azimuths relative to the reactor north 0° angular direction. The importance of surveillance capsules in fluence analyses is that they contain flux wires that are irradiated during reactor operation. When a capsule is removed from the reactor, the irradiated flux wires are evaluated to obtain activity measurements. These measurements are used to validate the fluence model. At this time, it is noted that no surveillance capsules have been removed from the Limerick 1 reactor.

3.3 Reactor System Material Compositions Each region of the reactor is comprised of materials that can include reactor fuel, metal, water, insulation, concrete, and air. Accurate material information is essential for the fluence evaluation as the material compositions determine the scattering and absorption of neutrons throughout the reactor system and, thus, affect the determination of neutron fluence in the RPV, surveillance capsules, vessel internal components, and ex-vessel structures.

Table 3-1 provides a summary of the materials for the principal components and regions of the Limerick 1 reactor. The material attributes for the metal, insulation, concrete, and air compositions (i.e., material densities and isotopic concentrations) are assumed to remain constant for the operating life of the reactor. The bulk water coolant properties throughout the reactor system, except for the core region, are determined assuming rated power and flow conditions.

The coolant properties remain constant unless there is a reported change in system heat balance conditions that affect the water properties in the reactor. The nuclear fuel compositions and coolant properties in the reactor core region change continuously during reactor operation. The

LIM-FLU-002-R-009 Revision 1 Page 3-4 of3-9 fuel and coolant prope1iies in the core region are updated for each reactor statepoint condition based on the actual or predicted operating states of the reactor. Water prope1iies immediately above and below the core region are updated on a cycle-by-cycle basis based on average cycle operating conditions.

Table 3-1 Summary of Material Compositions by Component Region for Limerick 1 Region Material Composition Biological Shield Clad Low-Alloy Steel Biological Shield Wall Reinforced Concrete Capsule Low-Alloy Steel Cavity Regions Air Cruciform Control Rods Stainless Steel, B4C Capsule Flux Wire Holder Pipe Stainless Steel Control Rod Guide Tubes Stainless Steel Coolant Water Core 2 , 2 , 2 9Pu, 24DPu, 241 Pu, 242Pu, 01ue1 , Zircaloy 3 5LJ 3BLJ 3 Core Hardware Zircaloy, Stainless Steel, lnconel, B4C Core Support Plate & Hardware Stainless Steel Fuel Support Pieces Stainless Steel Insulation Stainless Steel, Aluminum, Air Jet Pump Piping Stainless Steel Jet Pump Riser Brace Leaves Stainless Steel Jet Pump Riser Brace Pad Stainless Steel Jet Pump Riser Brace Yoke Stainless Steel Jet Pump Rams Head Beam lnconel Jet Pump Rams Head Bracket Stainless Steel LPCI coupler Stainless Steel LPRM Tube Water Nozzle Forgings Low-Alloy Steel RPV Clad Stainless Steel RPVWall Low-Alloy Steel Shroud Wall Stainless Steel Shroud Head Stainless Steel Slip Joint Clamp Stainless Steel Sparger Piping Stainless Steel Steam Separator Standpipes Stainless Steel Core Exit Steam Top Guide Stainless Steel Top Guide Pads Stainless Steel

LIM-FLU-002-R-009 Revision 1 Page 3-5 of 3-9 3.4 Reactor Operating Data Inputs An accurate evaluation of reactor vessel and component fluence requires an accurate accounting of the reactor's operating history. The principal operating parameters that affect the determination of neutron fluence in light water reactors include: core configurations and fuel assembly designs, power history, exposure and isotopic distributions, and water density distributions. The following subsections provide additional information on the characterization of reactor operating data for fluence evaluations.

3.4.1 Core Configuration and Fuel Design The reactor core configuration and the fuel assembly designs loaded in the reactor determine the neutron source and spatial source distribution contributing to the irradiation of the pressure vessel, vessel internals and ex-vessel supporting structures. The Limerick 1 core is comprised of 764 fuel assemblies in a fixed configuration. Several designs of fuel assemblies may be loaded in the reactor core in any given operating cycle. In order to determine accurate spatial fluence profiles throughout the reactor system, it is important to account for the different fuel designs loaded in the reactor over the operating lifetime of the reactor, especially those designs that reside in the peripheral locations of the core region.

Table 3-2 provides a summary of the many fuel assembly designs that have been loaded in the Limerick 1 reactor core for each operating cycle evaluated in this report. Table 3-2 also identifies the dominant fuel design loaded on the core periphery for each cycle and indicates the dominant (most numerous) assembly present in bold font.

3.4.2 Reactor Power History Reactor power history is the measure of reactor power levels, reactor power spatial distributions, fuel exposure distributions, and fuel isotopic distributions that a reactor experiences over its operating life. The power history data used in the Limerick 1 fluence evaluation includes daily power levels for Cycles 1 through 17 and part of Cycle 18. Projected operating data is used for the remainder of Cycle 18. A projected equilibrium cycle comprised of GNF3 fuel is used to project operating conditions to reactor end of license. The power history for Limerick 1 also accounts for periods of reactor shutdown due to refueling outages and other events that affect the activation and decay of dosimetry data.

Table 3-3 provides a summary of the operating history of the Limerick 1 reactor for each operating cycle. The number of statepoints used to represent the operating history and core power distributions of the reactor are listed for each cycle. Table 3-3 also shows that the reactor operated with a design rated thermal power of 3293 MWt for Cycles 1 through 6. The reactor implemented an extended power uprate at the beginning of Cycle 7 bringing the rated thermal power to 3458 MWt. The reactor implemented another power uprate in mid Cycle 14 bringing the rated thermal power to 3515 MWt, the power level at which the reactor is still operating at the time of this report.

Table 3-3 also shows the EFPY accumulated at the end of each cycle. The accumulated EFPY is computed from the operating data provided by Exelon and is verified against power production and exposure data obtained separately from plant records.

LIM-FLU-002-R-009 Revision 1 Page 3-6 of3-9 Table 3-2 Summary of Limerick 1 Core Loading Inventory 9x9 Fuel Domina 8x8 Fuel Designs 1Ox10 Fuel Designs Designs nt Cycle Periphe GE6 GE7 GE8 GE9 GE11 GE13 GE14 GNF2 GNF3 ral Fuel Design 1 764 GE6 2 424 72 268 GE6 3 200 72 492 GE6 4 52 296 304 112 GE6 5 108 304 112 240 GE?

6A 56 64 8 112 524 GE?

6B 56 64 9 112 523 GE?

7 140 340 284 GE6 8 112 113 539 GE6 9 80 56 628 GE6 10 484 280 GE13 11 221 543 GE13 12 764 GE14 13 764 GE14 14A 764 GE14 14B 764 GE14 15 488 276 GE14 16 212 552 GE14 17 764 GNF2 18(1) 764 GNF2 19+(2) 764 GNF3 (1) Cycle 18 is provided as partial historical operating data with projections to end ofcycle.

(2) Cycle 19+ is assumed to be an equilibrium cycle for projecting fluence to the end oflicense.

LIM-FLU-002-R-009 Revision 1 Page 3-7 of 3-9 3.4.3 Reactor Statepoint Data Statepoints break up operating history into ranges of operation based on similar power, exposure, and isotopic distributions. Typically, several statepoints are chosen for each cycle to represent the different operating conditions experienced by the reactor over the course of that cycle.

Table 3-3 shows that (( )) statepoints are used to represent the first 18 cycles of operating history for the Limerick 1 reactor, and another (( )) statepoints for the projected operating cycle, for a total of (( )) statepoints.

It is also shown in Table 3-3 that the number of statepoints varies between the cycles. This variation is due to several factors but is mostly related to the availability of data to represent the operating conditions of the reactor for any given operating cycle.

Core simulator data was provided by Exelon to characterize the operating conditions of Limerick 1 for Cycles 1 through 17, partial projection Cycle 18, and the GNF3 equilibrium projection cycle. The data calculated with core simulator codes represents the best-available information about the reactor core's operating history over the reactor's operating life. In this analysis, the core simulator data provided by Exelon was processed by TransWare to generate statepoint data files for input to the fluence model.

Because core simulator codes are used for a variety of core analysis functions, 10s to 100s of core calculations may be performed to track and monitor the operation of a reactor over the course of an operating cycle. Not all core calculations are suitable for use in fluence evaluations.

Therefore, each cycle of operating data is investigated to select the statepoints that are suitable for use in fluence evaluations. When all reactor conditions are considered, the number of core simulator statepoints selected for a fluence evaluation can vary from cycle to cycle.

A separate neutronics transport calculation is performed for each selected statepoint. The neutron fluxes calculated for each statepoint are then combined with the appropriate daily power history data described in Section 3.4.2 to provide an accurate accounting of the neutron fluence for the reactor pressure vessel, reactor vessel internals, and surveillance capsules. The periods of reactor shutdown are also accounted for in this process, particularly to allow for an accurate calculation of irradiated surveillance capsule activities.

LIM-FLU-002-R-009 Revision 1 Page 3-8 of3-9 Table 3-3 Statepoint Data for Limerick 1 per Cycle Basis Number of Reactor Rated Thermal Power 11, Cycle Number Accumulated EFPY Statepoints (MWt) 1 (( 3293 1.3 2 3293 2.3 3 3293 3.5 4 3293 4.6 5 3293 6.1 6 3293 7.9 7(2) 3458 9.8 8 3458 11.6 9 3458 13.4 10 3458 15.3 11 3458 17.2 12 3458 19.1 13 3458 21.1 14A(3l 3458 22.2 14g(3) 3515 22.9 15 3515 24.8 16 3515 26.7 17 3515 28.6 1814> 3515 30.5 19 +(5) )) 3515 57.0 (1) The rated tlmmal power for the reactor is listed for each cycle. Note, however, that tl1e actual daily power levels are used in deteimination of fluC11ce.

(2) The power uprate to 3458 MWt occun-ed at the sta1t ofCycle 7.

(3) The power uprate to 3515 MWt occun-ed on May 2, 2011, in Cycle 14, which is split to reflect the uprate.

(4) Cycle 18 is provided as partial historical operating data with projections to Clld ofcycle.

(5) Cycle 19+ is assumed to be the equilibrium cycle for projecting fluence to the plant end oflicense.

LIM-FLU-002-R-009 Revision 1 Page 3-9 of 3-9 3.4.4 Reactor Coolant Properties The reactor coolant water densities used in the fluence model are determined using combinations of core simulator codes and reactor heat balance data.

((

))

LIM-FLU-002-R-009 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-009 Revision 1 Page 4-1 of 4-12 4

METHODOLOGY This section provides an overview of the methodology and modeling approach used to determine fast neutron fluence for the Limerick 1 reactor pressure vessel (RPV), fast neutron fluence for the reactor vessel internals (RVI), and the fast neutron fluence and activations for the reactor surveillance capsules. The fluence model for Limerick 1 is a plant-specific model that is constructed from the design inputs described in Section 3, Description of the Reactor System.

The computational tools used in the fluence and activation analyses are based on the RAMA Fluence Methodology (RAMA) software. ((

))

4.1 Computational Method The RAMA Fluence Methodology is a system of computer codes, a data library, and an uncertainty methodology that determines best-estimate fluence and activations in light water reactor pressure vessels and vessel internal components. The primary software that comprises the methodology includes model builder codes, a particle transport code, and a fluence calculator code.

The primary inputs for the fluence methodology are mechanical design parameters and reactor operating history data. The mechanical design inputs are obtained from plant-specific design drawings, which include as-built measurements when available. The reactor operating history data is obtained from multiple sources, such as core simulator software, system heat balance calculations, daily operating logs, and cycle summary reports. A variety of outputs are available from the fluence methodology that include neutron flux, fast neutron fluence, dosimetry activation, and an uncertainty analysis.

The model builder codes consist of geometry and material processor codes that generate input for the RAMA transport code. The geometry model builder code uses mechanical design inputs and meshing specifications to generate three-dimensional geometry models of the reactor. The material processor code uses reactor operating data and material property inputs to process fuel materials, structural materials, and water densities that are consistent with the geometry meshing generated by the geometry model builder code.

The RAMA transport code performs three-dimensional neutron flux calculations using a deterministic, multigroup, particle transport theory method with anisotropic scattering ((

)) The transport solver is coupled with a general geometry modeling capability based on combinatorial geometry techniques. The coupling of general (arbitrary) geometry with a deterministic transport solver provides a flexible, efficient, and stable method for calculating neutron flux in light water reactor pressure vessels, vessel components, and structures. The primary inputs for the transport code include the geometry and material data generated by the model builder codes and numerical integration and convergence

LIM-FLU-002-R-009 Revision 1 Page 4-2 of 4-12 parameters for the iterative transport calculation. The primary output from the transport code is the neutron flux in multigroup form for every material region mesh in the fluence model.

The fluence calculator code determines fluence and activation in the reactor pressure vessel, surveillance specimens, and vessel components over specified periods of reactor operation. The fluence calculator also includes treatments for isotopic production and decay that are required to calculate specific activities for irradiated materials, such as the dosimetry specimens in the surveillance capsules. The primary inputs to the fluence calculator include the multigroup neutron flux from the transport code, response functions for the various materials in the reactor, reactor power levels for the operating periods of interest, specification of which components to evaluate, and the energy ranges of interest for evaluating neutron fluence. The reactor operating history is generally represented with several reactor statepoints that represent the core power and core power distributions of the reactor over the operating life of the reactor. These statepoints are integrated with the daily variations in reactor power levels to predict the fluence and activations accumulated throughout the reactor system.

The RAMA nuclear data library contains atomic mass data, nuclear cross-section data, response functions, and other nuclear constants that are needed for each of the code tools. The structure and contents of the data contained within the nuclear data file are based on the BUGLE-96 nuclear data library [13], with extended data representations derived from the VITAMIN-B6 data library [14].

The uncertainty methodology provides an assessment of the overall accuracy of the fluence and activation calculations. Variations in the ((

)) are evaluated to determine if there is a statistically significant bias in the calculated results that might affect the determination of the best-estimate fluence for the reactor. The plant-specific results are also weighted with comparative results from experimental benchmarks and other plant analyses and analytical uncertainties pertaining to the methodology to determine if the plant-specific model under evaluation is statistically acceptable as defined in Regulatory Guide 1.190 [1].

4.2 Fluence Model Section 3 describes the design inputs that were provided by Exelon for constructing the Limerick 1 reactor fluence model. These design inputs are used to develop a plant-specific, three-dimensional computer model of the Limerick 1 reactor for determining fast neutron fluence in the RPV components and for determining activation and fluence in reactor dosimetry for validating the RPV fluence predictions.

Figure 4-1 and Figure 4-2 provide general illustrations of the primary components, structures, and regions developed for the Limerick 1 fluence model. Figure 4-1 shows the planar configuration of the reactor model at an elevation corresponding to the reactor core mid-plane elevation. Figure 4-2 shows an axial configuration of the reactor model. Note that the figures are not drawn to scale and do not include representations of the meshing developed for this evaluation. The figures are intended only to provide a perspective for the layout of the model, and specifically how the various components, structures, and regions lie relative to the reactor core region (i.e., the neutron source). Additional detail is beyond the scope of this document.

LIM-FLU-002-R-009 Revision 1 Page 4-3 of 4-12

((

))

Figure 4-1 Planar View of the Limerick 1 Fluence Model at the Core Mid-Plane Elevation in Quadrant Symmetry

LIM-FLU-002-R-009 Revision 1 Page 4-4 of 4-12

((

))

Figure 4-2 Axial View of the Limerick 1 Fluence Model

LIM-FLU-002-R-009 Revision 1 Page 4-5 of 4-12 4.2.1 Geometry Model The Limerick 1 fluence model is constructed on a Cartesian coordinate system using a generalized three-dimensional geometry modeling technique based on combinatorial geometry.

The axial plane of the reactor model is defined by the (x,y) coordinates of the modeling system and the axial elevation at which a plane exists is defined along a perpendicular z-axis of the modeling system. This allows any point in the reactor model to be referenced by specifying the (x,y,z) coordinates for that point.

((

)) This modeling approach permits a model to be developed in any level of high-definition detail, such as is necessary for fluence and activation evaluations.

Figure 3-1 illustrates a planar cross-section view of the Limerick 1 reactor design at an axial elevation corresponding to the reactor core mid-plane. It is shown for this one elevation that the reactor design is a complex geometry ((

)) When the reactor is viewed in three dimensions, the varying heights of the different components, structures, and regions create additional geometry modeling complexities. An accurate representation of these geometrical complexities in a predictive computer model is essential for calculating accurate, best-estimate fluence in the reactor pressure vessel, surveillance capsules, vessel internals, and the supporting structures inside and outside of the reactor vessel.

Figure 4-1 and Figure 4-2 provide general illustrations of the planar and axial geometry complexities that are represented in the fluence model. For comparison purposes, the planar view illustrated in Figure 4-1 corresponds to the core elevation illustrated in Figure 3-1. ((

))

As previously noted, Figure 4-1 and Figure 4-2 are not drawn precisely to scale and are intended only to provide a perspective of how the various components, structures, and regions of the reactor are positioned relative to the reactor core region. The following subsections provide additional information on the constituent models developed for the individual components, structures, and regions of the fluence model.

LIM-FLU-002-R-009 Revision 1 Page 4-6 of 4-12 4.2.2 Reactor Core and Core Reflector The reactor core contains the nuclear fuel that is the source of the neutrons that irradiate all components and structures of the reactor. The core is surrounded by a shroud structure that serves to channel the reactor coolant through the core region during reactor operation. The coolant-containing region between the core and the core shroud is the core reflector. The reactor core geometry is rectangular in design ((

))

4.2.3 Reactor Core Shroud The core shroud is a canister-like structure that surrounds the reactor core. It channels the reactor coolant and steam produced by the core into the steam separators. Axially the shroud extends almost the entire height of the model and is divided into three sections: lower, central, and upper.

The lower shroud extends from the bottom of the model to the core support plate flange, the central shroud extends from the core support plate flange to the top guide flange, and the upper shroud extends from top guide flange to the top of the shroud head rim.

((

))

LIM-FLU-002-R-009 Revision 1 Page 4-7 of 4-12 Above the shroud wall is the shroud head which is penetrated by numerous steam separator standpipes. ((

))

4.2.4 Downcomer Region The downcomer region lies between the core shroud and the reactor pressure vessel. The downcomer is effectively cylindrical in design, but with geometrical complexities created by the presence of jet pumps and surveillance capsules in the region. ((

))

4.2.4.1 Jet Pumps Limerick 1 has ten jet pump assemblies in the downcomer region, which provide the main recirculation flow for the core. ((

))

4.2.4.2 Surveillance Capsules Section 3 describes the surveillance capsules installed in the Limerick 1 reactor. The three (3)

OEM surveillance capsules installed in the Limerick 1 reactor are positioned in close proximity to the RPV inner wall surface at the 30°, 120° and 300° azimuthal angles around the pressure vessel. ((

))

LIM-FLU-002-R-009 Revision 1 Page 4-8 of 4-12

((

))

4.2.5 Reactor Pressure Vessel The reactor pressure vessel and vessel cladding lie outside the downcomer region, ((

))

4.2.6 Thermal Insulation The reactor vessel thermal insulation lies in the cavity region outside the pressure vessel wall.

((

))

4.2.7 Inner and Outer Cavity Regions There are effectively two cavity regions represented in the model. The inner cavity region lies between the outer surface of the pressure vessel wall and the inner surface of the vessel insulation. The outer cavity region lies between the outer surface of the vessel insulation and inner surface of the biological shield wall cladding. ((

))

4.2.8 Biological Shield Model The biological shield (concrete) defines the outermost region of the fluence model. ((

))

LIM-FLU-002-R-009 Revision 1 Page 4-9 of 4-12 4.2.9 Above-Core Components Figure 4-2 includes illustrations of other components and regions that lie above the reactor core region. The predominant above-core components represented in the model include the top guide, core spray spargers, upper core shroud wall, shroud head, and steam separator standpipes. The shroud regions and standpipes are mentioned in further detail in Section 4.2.3.

4.2.9.1 Top Guide The top guide component lies above the core region and is appropriately modeled to include discrete representations of the top guide plates and accounting for the fuel assembly parts, top guide pads, and coolant between the plates. ((

))

4.2.9.2 Core Spray Spargers and Piping The core spray spargers include upper and lower sparger annulus pipes and a vertical inlet pipe.

((

))

4.2.10 Below-Core Components Figure 4-2 includes illustrations of other components and regions that lie below the reactor core region. The predominant below core (i.e., below active fuel) components represented in the fluence model include the lower fuel assembly parts, fuel support pieces, core support plate, core support plate rim bolts, cruciform control rods, control rod guide tubes, and lower shroud wall.

The lower shroud wall and fuel assembly components are described in previous sections, with the remaining components described in the following subsections.

4.2.10.1 Core Support Plate and Rim Bolts The core support plate includes appropriate penetrations for the fuel support pieces, control rod guide tubes, cruciform control rods, and the core support plate rim bolts. Core support plate rim bolts protrude from the top of the core support plate and traverse through the plate, rim, and core shroud lower flange. ((

))

4.2.10.2 Fuel Support Pieces The nuclear fuel assemblies loaded in the reactor are seated on fuel support pieces, which then rest in the core support plate and control blade guide tubes. ((

))

LIM-FLU-002-R-009 Revision 1 Page 4-10 of 4-12

((

))

4.2.10.3 Control Blades and Guide Tubes The fluence model allows for the representation of cruciform-shaped control blades and tubular control blade guide tubes in the below-core regions of the reactor. Coolant flow paths are included in the model ((

))

4.2.11 Summary of the Geometry Modeling Approach To summarize the reactor modeling process, there are several key features that allow the reactor design to be accurately represented for RPV fluence evaluations. ((

))

LIM-FLU-002-R-009 Revision 1 Page 4-11 of 4-12

((

))

4.3 Particle Transport Calculation Parameters The accuracy of the transport method is based on a numerical integration technique ((

))

4.4 Fission Spectrum and Neutron Source Modern core simulator software is capable of providing three-dimensional core power distributions and fuel isotopics in high-definition detail, viz., on a pin-by-pin basis. This allows fluence models to be constructed with a high-level of modeling detail for representing unique fission spectrum and neutron source terms for the transport calculation. ((

))

LIM-FLU-002-R-009 Revision 1 Page 4-12 of 4-12

((

))

4.5 Parametric Sensitivity Analyses Several plant-specific sensitivity analyses are performed to evaluate the accuracy and predictability of the neutral particle transport methodology for determining RPV fluence.

((

))

LIM-FLU-002-R-009 Revision 1 Page 5-1 of 5-7 5

REACTOR PRESSURE VESSEL FAST NEUTRON FLUENCE This section presents the predicted best-estimate fast neutron fluence (energy > 1.0 MeV) for the Limerick 1 reactor pressure vessel (RPV) at EOC 17 (28.6 EFPY) and 57 EFPY. It is reported in Section 6, Reactor Pressure Vessel Fluence Uncertainty Analysis, that the calculated fluence does not require a bias adjustment; therefore, the calculated fluence is the best-estimate fluence for Limerick 1.

The reactor pressure vessel fast neutron fluence is determined at the interface of the RPV base metal and cladding, which is denoted as the 0T location of the RPV wall. Damage fluence through the RPV wall is reported at the 1/4T and 3/4T depths in the wall. These values are determined based on the as-built RPV wall thickness of 16.823 cm (6.623 in).

The fast neutron fluence that is used in material embrittlement evaluations should be determined using an appropriate damage function (such as displacements-per-atom of iron) rather than the computed fast neutron fluence obtained from transport calculations. Two acceptable methods for estimating the damage fluence are prescribed in Regulatory Guide 1.99 [2]. ((

))

LIM-FLU-002-R-009 Revision 1 Page 5-2 of 5-7 Plant-specific fast neutron damage fluence for Limerick 1 is presented in this report for the RPV horizontal (circumferential) welds, vertical welds, shell plates, and nozzles that reside in the RPV beltline region. The location and naming convention of the shell plates, welds, and nozzles are shown in Figure 5-1. Also shown in Figure 5-1 is the calculated RPV beltline region for the Limerick 1 reactor at 57 EFPY.

Figure 5-1 Limerick 1 RPV Beltline Region at 57 EFPY Table 5-1 through Table 5-5 report the maximum fast neutron damage fluence that is determined for each of the RPV welds, shell plates, and nozzles residing in the RPV beltline. Damage fluence is reported at the 0T, 1/4T, and 3/4T depths in the RPV wall for the reporting periods of interest. In all tables, the damage fluence that exceeds the threshold fluence of 1.0E+17 n/cm2 is shown in red font. The maximum damage fluence determined for the welds, shells and nozzles is also shown in bold font. It is observed in the tables that many welds will have exceeded the fluence threshold of 1.0E+17 n/cm2 before 57 EFPY.

Table 5-1 and Table 5-2 report the maximum damage fluence that is determined for the RPV horizontal (circumferential) and vertical welds at EOC 17 (28.6 EFPY) and 57 EFPY, respectively. The maximum damage fluence for each weld is determined to occur at the 0T depth, with the maximum fluence occurring in vertical weld BD of Shell 2 with a value of 8.89E+17 n/cm2 at 57 EFPY.

Table 5-3 reports the maximum damage fluence that is determined for each RPV shell plate at EOC 17 (28.6 EFPY) and 57 EFPY. The maximum damage fluence for each shell plate is determined to occur at the 0T depth, with the maximum fluence occurring in Shell 2 with a value of 1.09E+18 n/cm2 at 57 EFPY.

LIM-FLU-002-R-009 Revision 1 Page 5-3 of 5-7 Table 5-4 and Table 5-5 report the maximum damage fluence that is determined for the RPV nozzles at EOC 17 (28.6 EFPY) and 57 EFPY, respectively. Damage fluence for the nozzles is presented along two paths: one along the nozzle forging-to-base-metal weld and the other along a 45-degree extraction path that extends from the inside corner of the forging to the outside surface of the RPV wall. Both paths are illustrated in Figure 5-2. It is noted that the extraction path is evaluated around the full circumference of the nozzle forging to determine the maximum fluence. The maximum damage fluence for the beltline nozzles is determined to occur at the 0T depth in the forging-to-base metal weld of the N17 nozzle with a value of 4.57E+17 n/cm2 at 57 EFPY.

Figure 5-2 Nozzle Fluence Edit Locations for Sample Nozzle Table 5-6 reports the elevations that define the RPV beltline at EOC 17 (28.6 EFPY) and 57 EFPY. It is shown that the RPV beltline at 57 EFPY covers 421.7 cm, or approximately 13.8 ft, of the reactor vessel.

LIM-FLU-002-R-009 Revision 1 Page 5-4 of 5-7 Table 5-1 Maximum Fast Neutron Damage Fluence for Limerick 1 RPV Beltline Welds at EOC 17 (28.6 EFPY)

Fast Neutron Fluence (n/cm2)

Weld Azimuth Elevation [in (cm)]

OT 1/4T 3/4T Horizontal Welds AB 64° 263.5 (669.3cm) 4.81E+17 3.28E+17 1.33E+17 AC 23° 400.5 (1017.3cm) 9.40E+15 6.93E+15 4.49E+15 Shell 1 Vertical Welds BA 317.5 ° 263.5 (669.3cm) 3.59E+17 2.46E+17 1.03E+17 BB 77.5 ° 263.5 (669.3cm) 3.22E+17 2.21E+17 9.23E+16 BC 197.5 ° 263.5 (669.3cm) 4.20E+17 2.86E+17 1.17E+17 Shell 2 Vertical Welds BO 255 ° 312.1 (792.8cm) 4.75E+17 3.24E+17 1.32E+17 BE 15° 312.1 (792.8cm) 4.77E+17 3.25E+17 1.33E+17

° BF 135 317.1 (805.4cm) 4.54E+17 3.12E+17 1.30E+17 Shell 3 Vertical Welds BG oo 400.5 (1017.3cm) 7.98E+15 5.94E+15 3.90E+15

° BH 120 400.5 (1017.3cm) 9.31E+15 6.85E+15 4.40E+15 BJ 240° 400.5 (1017.3cm) 9.31E+15 6.85E+15 4.40E+15 (1) Azimuth and elevation values are listed for the OT location only.

LIM-FLU-002-R-009 Revision 1 Page 5-5 of 5-7 Table 5-2 Maximum Fast Neutron Fluence for Limerick 1 RPV Beltline Welds at 57 EFPY Fast Neutron Fluence (n/cm2)

Weld Azimuth Elevation [in (cm)]

OT 1/4T 3/4T Horizontal Welds AB 65° 263.5 (669.3 cm) 8.08E+17 5.50E+17 2.24E+17 AC 67° 400.5 (1017.3 cm) 2.05E+16 1.50E+16 9.31E+15 Shell 1 Vertical Welds BA 317.5 ° 263.5 (669.3 cm) 5.77E+17 3.96E+17 1.67E+17 BB 77.5 ° 263.5 (669.3 cm) 5.80E+17 3.97E+17 1.66E+17 BC 197.5 ° 263.5 (669.3 cm) 7.32E+17 4.99E+17 2.04E+17 Shell 2 Vertical Welds BD 255 ° 323.0 (820.5 cm) 8.89E+17 6.05E+17 2.46E+17 BE 15° 323.0 (820.5 cm) 8.88E+17 6.05E+17 2.47E+17

° BF 135 323.0 (820.5 cm) 7.90E+17 5.41E+17 2.25E+17 Shell 3 Vertical Welds BG oo 400.5 (1017.3 cm) 1.81E+16 1.33E+16 8.31E+15

° BH 120 400.5 (1017.3 cm) 2.01E+16 1.47E+16 9.08E+15

° BJ 240 400.5 (1017.3 cm) 2.01E+16 1.47E+16 9.08E+15 (1) Azimuth and elevation values are listed for the OT location only.

LIM-FLU-002-R-009 Revision 1 Page 5-6 of 5-7 Table 5-3 Maximum Fast Neutron Damage Fluence for Limerick 1 RPV Beltline Shell Plates Fast Neutron Fluence Damage (n/cm2)

Shell Plate Azimuth Elevation [in (cm)]

OT 1/4T 3/4T EOC 17 (28.6 EFPY)

Shell 1 64 ° 263.5 (669.3 cm) 4.81E+17 3.28E+17 1.33E+17

° Shell 2 24 323.0 (820.5 cm) 6.04E+17 4.07E+17 1.64E+17 Shell 3 23° 400.5 (1017.3 cm) 9.40E+15 6.93E+15 4.49E+15 57 EFPY Shell 1 65° 263.5 (669.3 cm) 8.08E+17 5.50E+17 2.24E+17 Shell 2 67° 323.0 (820.5 cm) 1.09E+18 7.35E+17 2.94E+17 Shell 3 67° 400.5 (1017.3 cm) 2.05E+16 1.50E+16 9.31E+15 (1) Azimuth and elevation values are listed for the OT location only.

Table 5-4 Maximum Fast Neutron Damage Fluence for Limerick 1 RPV Beltline Nozzles at EOC 17 (28.6 EFPY)

Fast Neutron Fluence at (n/cm2)

Location Azimuth Elevation [in (cm)]

OT 1/4T 3/4T N2 Nozzles Weld 30° 1.32E+16 1.04E+16 7.35E+15

° Forging 30 1.88E+15 2.06E+15 3.78E+15

° Weld 60 1.32E+16 1.04E+16 7.31E+15 181.0 (459.7 cm)

Forging 60° 1.88E+15 2.06E+15 3.77E+15 Weld goo 6.45E+15 5.40E+15 4.42E+15 Forging goo 1.04E+15 1.23E+15 2.57E+15 N16 Nozzle Weld oo 6.22E+16 4.57E+16 2.31E+16 Weld 20° 366.0 (929.6 cm) 1.59E+17 1.11E+17 4.93E+16

° Weld 80 8.98E+16 6.41E+16 3.04E+16 N17 Nozzles Weld 45° 372.5 (946.2 cm) 2.37E+17 1.68E+17 7.25E+16

LIM-FLU-002-R-009 Revision 1 Page 5-7 of 5-7 Table 5-5 Maximum Fast Neutron Damage Fluence for Limerick 1 RPV Beltline Nozzles at 57 EFPY Fast Neutron Fluence at (n/cm2)

Location Azimuth Elevation [in (cm)]

OT 1/4T 3/4T N2 Nozzles

° Weld 30 2.23E+16 1.75E+16 1.24E+16 Forging 30° 3.21E+15 3.53E+15 6.44E+15 Weld 60° 2.26E+16 1.77E+16 1.24E+16

° 181.0 (459.7 cm)

Forging 60 3.27E+15 3.58E+15 6.47E+15 Weld goo 1.28E+16 1.05E+16 8.27E+15 Forging goo 2.08E+15 2.40E+15 4.80E+15 N16 Nozzle Weld oo 1.37E+17 1.00E+17 4.95E+16

° Weld 20 366.0 (929.6 cm) 3.34E+17 2.33E+17 1.02E+17

° Weld 80 1.97E+17 1.40E+17 6.49E+16 N17 Nozzles Weld 45° 372.5 (946.2 cm) 4.68E+17 3.30E+17 1.41E+17 Table 5-6 Reactor Beltline Elevation Range for Limerick 1 Lower Elevation Upper Elevation Axial Span of the RPV Reactor Lifetime

[in (cm)] [in (cm)] Beltline [in (cm)]

EOC 17 (28.6 EFPY) 215.0 (546.1 cm) 369.3 (938.0 cm) 154.3 (391.9 cm) 57 EFPY 210.1 (533.6 cm) 376.1 (955.3 cm) 166.0 (421.7 cm)

LIM-FLU-002-R-009 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-009 Revision 1 Page 6-1 of 6-3 6

REACTOR PRESSURE VESSEL FLUENCE UNCERTAINTY ANALYSIS This section presents the combined uncertainty analysis and the determination of bias for the Limerick 1 reactor pressure vessel (RPV) fluence evaluation. The combined uncertainty is comprised of two components. One component is the uncertainty factors developed from plant-specific measurements and the other is an analytic uncertainty factor. At this time, it is noted that no surveillance capsules have been removed from the Limerick 1 reactor. The plant-specific measurements used in this analysis are based on a database of plant measurements of similar design. When combined, the components provide a basis for determining the combined uncertainty (1) and bias in the computed RPV fluence.

The requirements for determining the combined uncertainty and bias for light water reactor pressure vessel fluence evaluations are provided in Regulatory Guide 1.190 [1]. ((

))

For pressure vessel fluence evaluations, two uncertainty factors are considered: comparison factors and uncertainty introduced by the measurement process. After analysis of these factors, it was determined that the combined uncertainty for the Limerick 1 RPV fluence was 11.2%, and that no adjustment for bias was required for the RPV fast neutron fluence presented in Section 5, Reactor Pressure Vessel Fast Neutron Fluence.

6.1 Comparison Uncertainty Comparison uncertainty factors are determined by comparing calculated activities with activity measurements. For pressure vessel fluence evaluations, two comparison uncertainty factors are considered: operating reactor comparison factors and benchmark comparison factors.

6.1.1 Operating Reactor Comparison Uncertainty TransWare has evaluated activation measurements for several BWR plants ranging from BWR/2-class plants to BWR/6-class plants. Each class of BWRs can have one or several variations of reactor core configurations, each having different radial diameters for the core shroud, reactor pressure vessel, and biological shield components. In addition, each can have different placements of the jet pumps and surveillance capsules in the reactor vessels.

The Limerick 1 reactor is a BWR/4 class design. ((

))

LIM-FLU-002-R-009 Revision 1 Page 6-2 of 6-3

(( )) The overall comparison ratio for all BWR class plants evaluated as of the date of this repo1t is 1.00 +/- 0.10. ((

))

6.1.2 Benchmark Comparison Uncertainty The benchmark comparison unce1tainty is based on a set of industiy standard simulation benchmark comparisons. fu accordance with the guidelines provided in Regulato1y Guide 1.190, it is appropriate to include comparisons of vessel simulation benchmark measurements in the overall fluence uncertainty evaluation. Two vessel simulation benchmarks are evaluated: the Pool Critical Assembly (PCA) and VENUS-3 experimental benchmarks.

The PCA experimental benchmark includes (( )) activation measurements at the mid-plane elevation in various simulated reactor components. The VENUS-3 expe1imental benchmark includes (( )) activation measurements at a range of elevations in various simulated reactor components. Table 6-1 summarizes the calculated-to-measurement (C/M) results detennined for these vessel simulation benchmarks.

Table 6-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements Average Number of Benchmark Calculated-to- St. Dev. (1a)

Measurements Measured (C/M)

Pool Critical Assembly ((

VENUS-3 Total Simulated Vessel Comparisons ))

6.2 Analytic Uncertainty The calculational models used for fluence analyses are comprised of numerous analytical parameters that have associated uncertainties in their values. The uncertainty in these parameters needs to be tested for its contribution to the overall fluence uncertainty.

((

))

LIM-FLU-002-R-009 Revision 1 Page 6-3 of 6-3

((

))

6.3 Combined Uncertainty The combined uncertainty for the reactor pressure vessel fluence evaluation is detennined with a weighting function ((

)). Table 6-2 shows that the combined uncertainty (lcr) detennined for the Limerick 1 reactor pressure vessel fluence is 11.2% for neutron energy exceeding 1.0 MeV.

It is shown in Table 6-2 that the combined unce1iainty is well below the 20% unce1iainty limit specified in Regulato1y Guide 1.190. In accordance with Regulato1y Guide 1.190, there is no discemable bias in the computed RPV fluence. Therefore, no adjustment to the RPV fast neutron fluence that is presented in Section 5, is required.

Table 6-2 Limerick 1 Combined RPV Uncertainty for Energy > 1.0 MeV Uncertainty Term Value Combined Uncertainty (1o) 11.2%

Bias None<1l

1) The bias term is less than its constituent uncertainty values, concluding that no statistically significant bias exists.

LIM-FLU-002-R-009 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-009 Revision 1 Page 7-1 of 7-4 7

REFERENCES 7.1 References

1. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.190: Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Washington, D.C.: Office of Nuclear Regulatory Research, 2001.

2. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.99: Radiation Embrittlement of Reactor Vessel Materials. Washington, D.C.: Office of Nuclear Regulatory Research, 1988.

3. ((

))

4. U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Safety Evaluation Report with Open Items Related to the License Renewal of Seabrook Station.

Docket Number 50-443. Washington, D.C.: Office of Nuclear Reactor Regulation, 2012.

5. ((

))

6. U.S. National Archives and Records Administration. Code of Federal Regulations. Title 10, Appendix G to Part 50 - Fracture Toughness Requirements. 2013.

7. U.S. National Archives and Records Administration. Code of Federal Regulations. Title 10, Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements. 2008.

8. U.S. National Archives and Records Administration. Code of Federal Regulations. Title 10, Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. 2007.

9. U.S. National Archives and Records Administration. Code of Federal Regulations. Title 10, Part 21 - Reporting of Defects and Noncompliance. 2015.

10. Limerick Generating Station Unit 1 Reactor Pressure Vessel Fluence Evaluation, TransWare Enterprises Inc., LIM-FLU-001-R-003, Rev 0, June 2010.

11. ((

))

12. ((

))

LIM-FLU-002-R-009 Revision 1 Page 7-2 of 7-4

13. Oak Ridge National Laboratory. Radiation Safety Information Computational Center.

BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. RSICC Data Library Collection, DLC-185. Oak Ridge, TN, 1996.

14. Oak Ridge National Laboratory. Radiation Safety Information Computational Center.

VITAMIN-B6: A Fine-Group Cross Section Library Based on ENDF/B-VI Release 3 for Radiation Transport Applications. RSICC Data Library Collection, DLC-184. Oak Ridge, TN, 1996.

15. E. N. Jones, Comparison of Regulatory Guide 1.99 Fluence Attenuation Methods, Journal of ASTM International. Vol. 9, No. 4, pp. 1-7 (2012).

16. ((

))

LIM-FLU-002-R-009 Revision 1 Page 7-3 of 7-4 7.2 Glossary AZIMUTHAL QUADRANT SYMMETRY - A type of core and pressure vessel azimuthal representation that represents a single quadrant of the reactor that can be rotated and mirrored to represent the entire 360-degree geometry. For example, the northeast quadrant can be mirrored to represent the northwest and southeast quadrants and can be rotated to represent the southwest quadrant.

BEST-ESTIMATE NEUTRON FLUENCE - See Neutron Fluence.

BOC - An acronym for beginning-of-cycle.

CALCULATED NEUTRON FLUENCE - See Neutron Fluence.

CALCULATIONAL BIAS - A calculational adjustment based on comparisons of calculations to measurements. If a bias is determined to exist, it may be applied as a multiplicative correction to the calculated fluence to produce the best-estimate neutron fluence.

CORE BELTLINE - The axial elevations corresponding to the active fuel height of the reactor core.

DAMAGE FLUENCE - See Neutron Fluence.

DPA - An acronym for displacements per atom which is typically used to characterize material damage in ferritic steels due to neutron exposure.

EFFECTIVE FULL POWER YEARS (EFPY) - A unit of measurement representing one full year of operation at the reactors rated power level. For example, if a reactor operates for 12 months at full rated power, this represents 1.0 EFPY. If the reactor operates for 10 months at full rated power, then goes into a power uprate and continues operating for another 2 months at the new full rated power, this also represents 1.0 EFPY.

EOC - An acronym for end-of-cycle.

EXTENDED BELTLINE REGION - See RPV beltline.

FAST NEUTRON FLUENCE - Fluence accumulated by neutrons with energy greater than 1.0 MeV (>1.0 MeV).

NEUTRON FLUENCE - Time-integrated neutron flux reported in units of n/cm2. The term best-estimate fluence refers to the fast neutron fluence that is computed in accordance with the requirements of U. S. Nuclear Regulatory Commission Regulatory Guide 1.190. The term damage fluence, which is required for material embrittlement evaluations, refers to an adjusted fast neutron fluence that is determined using damage functions specified in U. S. Nuclear Regulatory Commission Regulatory Guide 1.99.

NOZZLE EXTRACTION PATH - The path, or trajectory through the nozzle blend radius along which fluence is determined for the nozzle.

NOZZLE FORGING-TO-BASE-METAL WELD - The weld between the nozzle forging and the RPV base metal materials. This is sometimes referred to as the Nozzle-to-Shell Weld.

OEM - An acronym for original equipment manufacturer.

LIM-FLU-002-R-009 Revision 1 Page 7-4 of 7-4 RPV - An acronym for reactor pressure vessel. Unless otherwise noted, the reactor pressure vessel refers to the base metal material of the RPV wall (i.e., excluding clad/liner).

RPV BELTLINE - The RPV beltline is defined as that portion of the RPV adjacent to the reactor core that attains sufficient neutron radiation damage that the integrity of the pressure vessel could be compromised. For purposes of this evaluation, the fast neutron fluence threshold used to define the traditional RPV beltline is 1.0E+17 n/cm2. The axial span of the RPV that can exceed this threshold includes the RPV shells, welds, and heat-affected zones. An extended beltline is also defined to include lower fluence regions of the pressure vessel but with higher stresses than the traditional beltline region, such as RPV nozzles. The combination of fluence and stress may result in a limiting location in the pressure vessel for determining pressure-temperature limits.

RPV ZERO ELEVATION - The RPV zero elevation is defined at the inside surface of the lowest point in the vessel bottom head, which is typically the bottom drain plug location. Axial elevations presented in this report are relative to RPV zero.

RVI - An acronym for reactor vessel internals.

ATTACHMENT 10 Limerick Generating Station Unit 2 Reactor Pressure Vessel Fluence Evaluation - Non-Proprietary Report, LIM-FLU-002-R-010, Rev. 1 (Non-Proprietary)

LIM-FLU-002-R-O I 0 Revis ion I Page i of xi Topical Report LIMERICK GENERATING STATION UNIT 2 REACTOR PRESSURE VESSEL FLUENCE EVALUATION - NON-PROPRIETARY REPORT Document Number: LIM-FLU-002-R-010 Revision 1 September 2020 Prepared by: TransWare Enterprises Inc.

1565 Mediterranean Drive I

Sycamore, Illinois 60178 Prepared for: Exelon Generation Company, LLC Limerick Generating Station Unit 2 3146 Sanatoga Road Pottstown , Pennsylvania 19464 Project Manager: Michelle Karasek Controlled Copy Number: __2__

T his document represents the non-proprietary version of Trans Ware Enterprises Inc.

document number LfM-FLU-001-R-006, Revis ion I. Proprietary information removed from LI M-FLU-00 l-R-006, Revision I is identi fied in this report by enclosure in double brackets.

trans-E N T ERP R 1 SES 1565 Mediterranean Dr Sycamore, Illinois 601 78-3141 815-895-4 700

• www.t ransware.net

LIM-FLU-002-R-010 Revision 1 Page ii of xi

[ This page intentionally left blank. ]

LIM-FLU-002-R-O 10 Revision 1 Page iii of xi Topical Report LIMERICK GENERATING STATION UNIT 2 REACTOR PRESSURE VESSEL FLUENCE EVALUATION-NON-PROPRIETARY REPORT Document Number: LIM-FLU-002-R-010 Revision 1 September 2020 Preparing Organization: TransWare Enterprises Inc.

Prepared By: '!/;'} (2.:GZO K. E. Watkins, Project Engineer Date Reviewed By:

n11r tli te

~}11 t~~?.o Date Approved By: q6tfv.>w

/ate Prepared For: Exelon Generation Company, LLC Limerick Generating Station Unit 2 3146 Sanatoga Road Pottstown, Pennsylvania 19464 Project Manager: Michelle Karasek This document represents the non-proprietary version ofTransWare Enterprises Inc. document number LIM-FLU-001-R-006, Revision 1. Proprietary information removed from LIM-FLU-001-R-006, Revision 1 is identified in this report by enclosure in double brackets.

TransWare Enterprises Inc.* 1565 Mediterranean Dr.* Sycamore, Illinois 60178-3141 *USA

+1-815-895-4700

• www .transware.net

LIM-FLU-002-R-010 Revision 1 Page iv of xi DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THE INFORMATION CONTAINED IN THIS REPORT IS BELIEVED BY TRANSWARE ENTERPRISES INC. TO BE AN ACCURATE AND TRUE REPRESENTATION OF THE FACTS KNOWN, OBTAINED OR PROVIDED TO TRANSWARE ENTERPRISES INC. AT THE TIME THIS REPORT WAS PREPARED. THE USE OF THIS INFORMATION BY ANYONE OTHER THAN THE CUSTOMER OR FOR ANY PURPOSE OTHER THAN THAT FOR WHICH IT IS INTENDED, IS NOT AUTHORIZED; AND WITH RESPECT TO ANY UNAUTHORIZED USE, TRANSWARE ENTERPRISES INC. MAKES NO REPRESENTATION OR WARRANTY AND ASSUMES NO LIABILITY AS TO THE COMPLETENESS, ACCURACY OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS DOCUMENT. IN NO EVENT SHALL TRANSWARE ENTERPRISES INC. BE LIABLE FOR ANY LOSS OF PROFIT OR ANY OTHER COMMERCIAL DAMAGE, INCLUDING BUT NOT LIMITED TO SPECIAL, CONSEQUENTIAL OR OTHER DAMAGES.

((

))

QUALITY REQUIREMENTS This document has been prepared in accordance with the requirements of 10CFR50 Appendix B, 10CFR21, and TransWare Enterprises Inc.s 10CFR50 Appendix B quality assurance program.

LIM-FLU-002-R-010 Revision 1 Page v of xi ACKNOWLEDGEMENTS TransWare Enterprises Inc. wishes to acknowledge Michelle Karasek of Exelon Generation Company, LLC for her management of the project. TransWare Enterprises Inc. also wishes to acknowledge Marcus Gergar and Scott Foster of Exelon Generation Company, LLC for their support and assistance in providing the mechanical design and operating data for this reactor pressure vessel fluence evaluation.

LIM-FLU-002-R-010 Revision 1 Page vi of xi

[ This page intentionally left blank. ]

LIM-FLU-002-R-010 Revision 1 Page vii of xi CONTENTS 1 Introduction.................................................................................................................. 1-1 1.1 Regulatory Requirements ..................................................................................... 1-2 1.2 Limitations of the Fluence Evaluation ................................................................... 1-3 1.3 Quality Assurance ................................................................................................ 1-3 2 Summary of Results .................................................................................................... 2-1 3 Description of the Reactor System............................................................................. 3-1 3.1 Overview of the Reactor System Design .............................................................. 3-1 3.2 Reactor System Mechanical Design Inputs........................................................... 3-3 3.3 Reactor System Material Compositions ................................................................ 3-3 3.4 Reactor Operating Data Inputs ............................................................................. 3-5 3.4.1 Core Configuration and Fuel Design ......................................................... 3-5 3.4.2 Reactor Power History .............................................................................. 3-5 3.4.3 Reactor Statepoint Data ............................................................................ 3-7 3.4.4 Reactor Coolant Properties ....................................................................... 3-9 4 Methodology ................................................................................................................ 4-1 4.1 Computational Method.......................................................................................... 4-1 4.2 Fluence Model ...................................................................................................... 4-2 4.2.1 Geometry Model ....................................................................................... 4-5 4.2.2 Reactor Core and Core Reflector .............................................................. 4-6 4.2.3 Reactor Core Shroud ................................................................................ 4-6 4.2.4 Downcomer Region .................................................................................. 4-7 4.2.4.1 Jet Pumps................................................................................. 4-7 4.2.4.2 Surveillance Capsules .............................................................. 4-7 4.2.5 Reactor Pressure Vessel .......................................................................... 4-8

(( ))

4.2.6 Thermal Insulation .................................................................................... 4-8 4.2.7 Inner and Outer Cavity Regions ................................................................ 4-8

LIM-FLU-002-R-010 Revision 1 Page viii of xi 4.2.8 Biological Shield Model ............................................................................. 4-9 4.2.9 Above-Core Components .......................................................................... 4-9 4.2.9.1 Top Guide ................................................................................. 4-9 4.2.9.2 Core Spray Spargers and Piping............................................... 4-9 4.2.10 Below-Core Components .......................................................................... 4-9 4.2.10.1 Core Support Plate and Rim Bolts ............................................ 4-9 4.2.10.2 Fuel Support Pieces................................................................ 4-10 4.2.10.3 Control Blades and Guide Tubes ............................................ 4-10 4.2.11 Summary of the Geometry Modeling Approach ....................................... 4-10 4.3 Particle Transport Calculation Parameters.......................................................... 4-11 4.4 Fission Spectrum and Neutron Source ............................................................... 4-11 4.5 Parametric Sensitivity Analyses .......................................................................... 4-12 5 Reactor Pressure Vessel Fast Neutron Fluence ........................................................ 5-1 6 Reactor Pressure Vessel Fluence Uncertainty Analysis ........................................... 6-1 6.1 Comparison Uncertainty ....................................................................................... 6-1 6.1.1 Operating Reactor Comparison Uncertainty .............................................. 6-1 6.1.2 Benchmark Comparison Uncertainty ......................................................... 6-2 6.2 Analytic Uncertainty .............................................................................................. 6-2 6.3 Combined Uncertainty .......................................................................................... 6-3 7 References ................................................................................................................... 7-1 7.1 References ........................................................................................................... 7-1 7.2 Glossary ............................................................................................................... 7-3

LIM-FLU-002-R-010 Revision 1 Page ix of xi LIST OF FIGURES Figure 3-1 Planar View of the Limerick 2 Reactor at the Core Mid-Plane Elevation .............. 3-2 Figure 4-1 Planar View of the Limerick 2 Fluence Model at the Core Mid-Plane Elevation in Quadrant Symmetry ......................................................................... 4-3 Figure 4-2 Axial View of the Limerick 2 Fluence Model ......................................................... 4-4 Figure 5-1 Limerick 2 RPV Beltline Region at 57 EFPY ........................................................ 5-2 Figure 5-2 Nozzle Fluence Edit Locations for Sample Nozzle ............................................... 5-3

LIM-FLU-002-R-010 Revision 1 Page x of xi

[ This page intentionally left blank. ]

LIM-FLU-002-R-010 Revision 1 Page xi of xi LIST OF TABLES Table 2-1 Maximum Fast Neutron Fluence for Limerick 2 RPV Beltline Welds, Nozzles, and Shell Plate Locations ...................................................................... 2-2 Table 2-2 RPV Beltline Elevation Range for Limerick 2 ....................................................... 2-3 Table 3-1 Summary of Material Compositions by Component Region for Limerick 2 ........... 3-4 Table 3-2 Summary of Limerick 2 Core Loading Inventory .................................................. 3-6 Table 3-3 Statepoint Data for Limerick 2 per Cycle .............................................................. 3-8 Table 5-1 Maximum Fast Neutron Damage Fluence for Limerick 2 RPV Beltline Welds at EOC 14 (25.1 EFPY) ....................................................................................... 5-4 Table 5-2 Maximum Fast Neutron Fluence for Limerick 2 RPV Beltline Welds at 57 EFPY .............................................................................................................. 5-5 Table 5-3 Maximum Fast Neutron Damage Fluence for Limerick 2 RPV Beltline Shell Plates .................................................................................................................. 5-6 Table 5-4 Maximum Fast Neutron Damage Fluence for Limerick 2 RPV Beltline Nozzles at EOC 14 (25.1 EFPY) .......................................................................... 5-6 Table 5-5 Maximum Fast Neutron Damage Fluence for Limerick 2 RPV Beltline Nozzles at 57 EFPY............................................................................................. 5-7 Table 5-6 Reactor Beltline Elevation Range for Limerick 2 .................................................. 5-7 Table 6-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements ....... 6-2 Table 6-2 Limerick 2 Combined RPV Uncertainty for Energy > 1.0 MeV ............................. 6-3

LIM-FLU-002-R-010 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-010 Revision 1 Page 1-1 of 1-3 1

INTRODUCTION This report presents the results of the reactor pressure vessel fast neutron fluence evaluation that was performed for the Limerick Generating Station Unit 2 Unit 2 (Limerick 2) reactor. The Limerick 2 reactor is owned and operated by Exelon Generation Company, LLC (Exelon).

The fast neutron fluence presented in this report was determined in accordance with the guidelines and requirements presented in U. S. Nuclear Regulatory Commission (NRC)

Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [1]. Fluence is presented for the 0T, 1/4T and 3/4T depths in the reactor pressure vessel (RPV) plates, welds, and nozzles throughout the RPV extended beltline region determined at 60 years of reactor operation. The damage fluence determined at the 1/4T and 3/4T depths in the RPV wall were determined using the displacements-per-atom (dpa) attenuation method prescribed in U. S. NRC Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials [2].

The Limerick 2 fluence evaluations provided in this report are based on historical operating conditions for the reactor and projected operation through the end of licensed operation. The RPV fluence evaluations are presented at the End of Cycle (EOC) 14 and projected to 57 Effective Full Power Years (EFPY) of reactor operation.

The fluence evaluations are performed based on the RAMA Fluence Methodology [3]. Under funding from Electric Power Research Institute, Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP), the RAMA Fluence Methodology was developed by TransWare Enterprises Inc. for the purpose of calculating fast neutron fluence in nuclear reactor pressure vessels and reactor vessel internal components. The RAMA Fluence Methodology (hereafter referred to as RAMA) has received generic approval [4] from the U.S. NRC for determining fast neutron fluence in BWRs and PWRs in accordance with the requirements of Regulatory Guide 1.190. ((

))

In compliance with Regulatory Guide 1.190, TransWare Enterprises Inc. (TransWare) has benchmarked the RAMA Fluence Methodology against industry standard benchmarks and plant-specific dosimetry measurements for boiling water reactors and pressurized water reactors. The results of the benchmarking show that the fluence methodology implemented by TransWare predicts specimen activities with no discernable bias in the computed fluence. ((

))

LIM-FLU-002-R-010 Revision 1 Page 1-2 of 1-3 1.1 Regulatory Requirements Part 50 of Title 10 of the Code of Federal Regulations provides requirements for establishing irradiated material monitoring programs that serve to ensure the integrity of the reactor coolant pressure boundary of light water nuclear power reactors. Two appendices to Part 50 present the requirements that guide the fluence determinations presented in this report: Appendix G, Fracture Toughness Requirements [6], and Appendix H, Reactor Vessel Material Surveillance Program Requirements [7].

Appendix G specifies fracture toughness requirements for the carbon and low-alloy ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary to ensure adequate margins of safety during any condition of normal operation, including anticipated conditions for system hydrostatic testing, to which the pressure boundary may be subjected over its service lifetime. These requirements apply to base metal, welds and weld heat-affected zones in the materials within the reactor pressure vessel beltline region.

Appendix H specifies the requirements for a material surveillance program that serves to monitor changes in the fracture toughness properties of the ferritic materials in the reactor beltline region.

The changes in fracture toughness properties of ferritic materials are attributed to the exposure of the materials to neutron irradiation and the thermal environment.Section III of Appendix H specifies that a material surveillance program is required for light water nuclear power reactors if the peak fast neutron fluence with energy greater than 1 MeV (E > 1 MeV) at the end of the design life of the vessel is expected to exceed 1017 n/cm2.

In compliance with the Appendix H requirements, fracture toughness test data are obtained from material specimens that are exposed to neutron irradiation in surveillance capsules installed at or near the inner surface of the reactor pressure vessel. These capsules are withdrawn periodically from the reactor for measurement and analysis. Fast neutron fluence is not a measurable quantity and must be determined using analytical methods. It must be demonstrated that the analytical method used to determine the fast neutron fluence provides a conservative prediction over the beltline region of the pressure boundary when compared to the measurement data with allowances for all uncertainties in the measurements.Section III of Appendix H also allows for an Integrated Surveillance Program (ISP) in which representative materials for the reactor are irradiated in one or more other reactors of sufficiently similar design and operating features to permit accurate comparisons of the predicted amount of radiation damage.

Implementing guidelines addressing the requirements of Appendices G and H are provided in U. S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials [2], and Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [1]. Regulatory Guide 1.99 addresses the requirements of Appendix G for determining the damage fluence that is used in the evaluation of fracture toughness in light water nuclear reactor pressure vessel ferritic materials. Regulatory Guide 1.190 addresses the requirements for determining the fast neutron fluence and uncertainty in the fluence predictions that are used in fracture toughness evaluations.

The fast neutron fluence evaluations described in this report meet the requirements of Appendices G and H of Part 50 of Title 10 of the Code of Federal Regulations and U. S. NRC Regulatory Guides 1.190 and 1.99 Revision 2.

LIM-FLU-002-R-010 Revision 1 Page 1-3 of 1-3 1.2 Limitations of the Fluence Evaluation The fast neutron fluence presented in this report is based on historical and projected operating conditions of the reactor. The RPV fast neutron fluence that is based on historical operating conditions is determined to meet the requirements of Regulatory Guide 1.190 with no discernable bias in the results. It is determined, therefore, that the RPV fast neutron fluence presented in this report is suitable for use in evaluating material embrittlement conditions of reactor pressure vessel materials in accordance with Regulatory Guide 1.99. Use of the results for other purposes is not demonstrated.

Fluence projections are determined using the most up-to-date cycle operating data available at the time of this report. The projection cycle is assumed to be an equilibrium cycle representative of how the reactor will operate until the end of the reactors operating license. Continued use of the projected fluence presented in this report must be demonstrated as applicable as new operating history data from the reactor becomes available. Deviations from the design basis of this analysis, including changes in future fuel designs, core loadings, and operating strategies that result in a significant change to the core power shapes relative to the projected data may require re-evaluation to determine the impact of the altered flux profiles on the projected RPV and RVI component fluences.

It is cited in Regulatory Position 1.2 of Regulatory Guide 1.190 that a best-estimate power distribution may be used for reactor vessel neutron fluence calculations. The best-estimate fluence presented in this report meets the requirements of Regulatory Position 1.2. Regulatory Position 1.2 further states that if a best-estimate is used, the power distribution must be updated if changes in core loadings, surveillance measurements, or other information indicate a significant change in projected fluence values. Under this requirement, the other information that can necessitate an update of the fluence model can include: implementation of power uprates/derates, introduction of new fuel designs, changes in projected cycle lengths, changes in core loading and/or operational strategies, changes in reactor flow, or other changes that could alter the power/flux profiles used in the fluence projections and uncertainty analysis.

1.3 Quality Assurance The fluence evaluations presented in this report were performed in compliance with the quality assurance requirements of Appendix B to Part 50 of Title 10 of the Code of Federal Regulations, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants (10CFR50 Appendix B) [8], and to Part 21, Reporting of Defects and Noncompliance (10CFR21) [9].

LIM-FLU-002-R-010 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-010 Revision 1 Page 2-1 of 2-3 2

SUMMARY

OF RESULTS This section provides a summary of the fast neutron fluence determined for the Limerick 2 reactor pressure vessel (RPV). Details of the RPV fluence evaluation are presented in Section 5, Reactor Pressure Vessel Fast Neutron Fluence. Section 6, Reactor Pressure Vessel Fluence Uncertainty Analysis, provides a determination of the uncertainty in the RPV fluence evaluation.

Limerick 2 is a BWR/4 class plant with a core loading of 764 fuel assemblies. The fluence evaluation for this plant is based on historical operating data through Cycle 14 (25.1 EFPY).

Fluence evaluations are also performed at 57 EFPY. Cycle 15 was provided as a partial projection cycle. A projection cycle using an advanced fuel design, GNF3, was also provided and was used for the fluence projection to 57 EFPY.

Table 2-1 presents a summary of the maximum fast neutron fluence determined for the RPV shell plates, welds and nozzles at EOC 14 (25.1 EFPY) and 57 EFPY. The significant fluence for the evaluated areas of the RPV occur at the inside surface of the RPV base metal, which is denoted as the 0T depth in the pressure vessel wall. With the exception of the N2 nozzles, Shell 3, and welds AC, BG, BH and BJ, all other evaluated areas of the RPV have exceeded the fluence threshold of 1.0E+17 n/cm2. It is shown in Table 2-1 that the maximum fluence is determined to occur at the 0T location of the intermediate shell plate with a value of 1.05E+18 n/cm2 at 57 EFPY. Note in Table 2-1 that all fluence that has exceeded the fluence threshold of 1.0E+17 n/cm2 are shown in red font and that the maximum fluences in the RPV are additionally shown in bold font.

Table 2-2 shows the axial span of the RPV beltline region that was determined for Limerick 2 at EOC 14 (25.1 EFPY) and 57 EFPY. The reactor beltline region is defined in Appendices G [6]

and H [7] of 10CFR50 to include those regions that directly surround the effective height of the reactor core, as well as those adjacent areas of the RPV that are predicted to experience sufficient neutron irradiation damage. This definition of the RPV beltline is considered to include all materials that exceed a fast neutron fluence of 1.0E+17 n/cm2. At 57 EFPY the RPV beltline covers 420.0 cm, or approximately 13.8 ft of the reactor vessel. The scope of the fluence model was developed to provide an evaluation of the reactor pressure vessel over the full height of the RPV extended beltline region.

LIM-FLU-002-R-010 Revision 1 Page 2-2 of2-3 Table 2-1 Maximum Fast Neutron Fluence for Limerick 2 RPV Beltline Welds, Nozzles, and Shell Plate Locations Maximum Fast Neutron Fluence (n/cm2)

Component EOC 14 (25.1 EFPY) 57 EFPY RPV Beltline Welds AB 4.05E+17 7.69E+17 AC 7.68E+15 2.02E+16 BA 2.92E+17 5.36E+17 BB 2.84E+17 5.71E+17 BC 3.61E+17 7.09E+17 BD 4.10E+17 8.71E+17 BE 4.12E+17 8.69E+17 BF 3.71E+17 7.46E+17 BG 6.76E+15 1.81E+16 BH 7.52E+15 1.96E+16 BJ 7.52E+15 1.96E+16 Nozzle Forging-to-Base-Metal Welds Nozzle Weld N2 1.11E+16 2.16E+16 Nozzle Weld N16 1.35E+17 3.29E+17 Nozzle Weld N17 1.96E+17 4.54E+17 Shell Plates Shell 1 4.05E+17 7.69E+17 Shell 2 5.12E+17 1.05E+18 Shell 3 7.68E+15 2.02E+16

LIM-FLU-002-R-010 Revision 1 Page 2-3 of2-3 Table 2-2 RPV Beltline Elevation Range for Limerick 2 Lower Elevation Upper Elevation Axial Span of the RPV Reactor Lifetime

[in (cm)] [in (cm)] Beltline [in (cm)]

EOC 14 (25.1 EFPY) 216.9 (551.0 cm) 367.6 (933.8 cm) 150.7 (382.8 cm) 57 EFPY 210.4 (534.5 cm) 376.0 (955.0 cm) 165.6 (420.5 cm)

Section 5 provides detailed results for the RPV fast neutron fluence evaluation. RPV damage fluence is reported at the OT, 1/4T, and 3/4T depths of the RPV wall for each horizontal (circumferential) weld, ve1iical (axial) weld, shell plate, and nozzle in the RPV beltline.

Figure 5-1 illustrates the location of the welds, shell plates, and nozzles in the RPV. Fluence damage through the thickness of the RPV wall is detennined using the displacements-per-atom (dpa) attenuation method prescribed in Regulato1y Guide 1.99 [2].

Section 6 provides an evaluation of the combined uncertainty detennined for the Limerick 2 fluence evaluation. The combined unce1iainty is detennined by combining the measurement and analytic unce1iainties and is dete1mined to be 11.2%. In accordance with Regulato1y Guide 1.190, there is no discemable bias in the computed fluence; therefore, the computed fluence is the best-estimate fluence.

LIM-FLU-002-R-010 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-010 Revision 1 Page 3-1 of 3-9 3

DESCRIPTION OF THE REACTOR SYSTEM This section provides an overview of the reactor design and operating data inputs that were used to develop the computational fluence model for the Limerick 2 reactor. All reactor design and operating data inputs used to develop the model are plant-specific and were provided by Exelon.

The inputs for the fluence geometry model were developed from nominal and as-built drawings for the reactor pressure vessel, vessel internals, fuel assemblies, and containment regions.

Several modifications were made to the Limerick 2 RAMA geometry model since the previous vessel internal component fluence evaluation performed by TransWare in 2010 [10]. The principal mechanical modifications include:

• Addition of RPV nozzle forgings for the recirculation outlet (N1), recirculation inlet (N2), and LPCI (N17) nozzles

• Improved shroud head and steam separator standpipe model

• Improved jet pump rams head, hold down beam, slip joint clamp, and auxiliary spring wedge models

• Improved below core model, including the following discrete components:

o Core support plate and rim bolts o Orificed fuel support and peripheral fuel support pieces o Control rod drive tubes The reactor operating history was also updated to provide a historical accounting of how the reactor operated for Cycles 1 through 14 with projections to end of license.

3.1 Overview of the Reactor System Design Limerick 2 is a General Electric BWR/4 class reactor with a core loading of 764 assemblies. The unit began commercial operation in 1990 with a design rated power of 3293 MWt. In Cycle 4 the rated power was increased to 3458 MWt and during Cycle 12 a rated power of 3515 MWt was achieved. At the time of this fluence evaluation, Limerick 2 has completed 14 cycles of operation.

Figure 3-1 illustrates the basic planar configuration of the Limerick 2 reactor at an axial elevation near the reactor core mid-plane. All the radial regions of the reactor that are required for fluence evaluations are shown. Beginning at the center of the reactor and projecting outward, the regions include: the core region; core reflector region (bypass water); central shroud wall; downcomer water region including the jet pumps; RPV wall; cavity region between the RPV wall and insulation; insulation; cavity region between the insulation and biological shield; and the biological shield wall. Cladding is included on the inner RPV surface as well as the inner and

LIM-FLU-002-R-010 Revision 1 Page 3-2 of 3-9 outer surfaces of the biological shield wall. Also represented in Figure 3-1 are notations indicating the control rod and fuel assembly locations within the core. Note that the fuel locations are shown only for the northeast quadrant of the core region.

Figure 3-1 Planar View of the Limerick 2 Reactor at the Core Mid-Plane Elevation

LIM-FLU-002-R-010 Revision 1 Page 3-3 of 3-9 3.2 Reactor System Mechanical Design Inputs The mechanical design inputs used to construct the Limerick 2 fluence geometry model are based upon nominal design and as-built dimensional information. As-built data is preferred when constructing plant-specific reactor fluence models; however, as-built data is not always available and nominal dimensions are used.

For the Limerick 2 fluence model, the predominant dimensional information used to construct the fluence model is nominal design data. ((

))

An important component of a computational reactor pressure vessel fluence model is the accurate description of the surveillance capsules installed in the pressure vessel. Figure 3-1 shows that the Limerick 2 reactor was initially equipped with three surveillance capsules. The capsules were installed at an elevation around the reactor core mid-plane. Each capsule was mounted radially near the inside surface (0T) of the RPV wall. The surveillance capsules were distributed around the circumference of the pressure vessel at the 30°, 120° and 300° azimuths relative to the reactor north 0° angular direction. The importance of surveillance capsules in fluence analyses is that they contain flux wires that are irradiated during reactor operation. When a capsule is removed from the reactor, the irradiated flux wires are evaluated to obtain activity measurements. The measurements are used to validate the fluence model. At this time, it is noted that no surveillance capsules have been removed and measured from the Limerick 2 reactor.

((

))

3.3 Reactor System Material Compositions Each region of the reactor is comprised of materials that can include reactor fuel, metal, water, insulation, concrete, and air. Accurate material information is essential for the fluence evaluation as the material compositions determine the scattering and absorption of neutrons throughout the reactor system and, thus, affect the determination of neutron fluence in the RPV, surveillance capsules, vessel internal components, and ex-vessel structures.

Table 3-1 provides a summary of the materials for the principal components and regions of the Limerick 2 reactor. The material attributes for the metal, insulation, concrete, and air compositions (i.e., material densities and isotopic concentrations) are assumed to remain constant for the operating life of the reactor. The bulk water coolant properties throughout the reactor system, except for the core region, are determined assuming rated power and flow conditions.

LIM-FLU-002-R-010 Revision 1 Page 3-4 of 3-9 Table 3-1 Summary of Material Compositions by Component Region for Limerick 2 Region Material Composition Biological Shield Clad Low-Alloy Steel Biological Shield Wall Reinforced Concrete Capsule Low-Alloy Steel Cavity Regions Air Cruciform Control Rods Stainless Steel, B4C Capsule Flux Wire Holder Pipe Stainless Steel Control Rod Guide Tubes Stainless Steel Coolant Water Core 235LJ , 23BLJ, 239Pu, 24DPu, 241 Pu, 242Pu, 01ue1, Zircaloy Core Hardware Zircaloy, Stainless Steel, lnconel, B4C Core Support Plate & Hardware Stainless Steel Fuel Support Pieces Stainless Steel Insulation Stainless Steel, Aluminum, Air Jet Pump Piping Stainless Steel Jet Pump Riser Brace Leaves Stainless Steel Jet Pump Riser Brace Pad Stainless Steel Jet Pump Riser Brace Yoke Stainless Steel Jet Pump Rams Head Beam lnconel Jet Pump Rams Head Bracket Stainless Steel LPCI coupler Stainless Steel LPRM Tube Water Nozzle Forgings Low-Alloy Steel RPV Clad Stainless Steel RPVWall Low-Alloy Steel Shroud Wall Stainless Steel Shroud Head Stainless Steel Slip Joint Clamp Stainless Steel Sparger Piping Stainless Steel Steam Separator Standpipes Stainless Steel Core Exit Steam Top Guide Stainless Steel Top Guide Pads Stainless Steel

LIM-FLU-002-R-010 Revision 1 Page 3-5 of 3-9 The coolant properties remain constant unless there is a reported change in system heat balance conditions that affect the water properties in the reactor. The nuclear fuel compositions and coolant properties in the reactor core region change continuously during reactor operation. The fuel and coolant properties in the core region are updated for each reactor statepoint condition based on the actual or predicted operating states of the reactor. Water properties immediately above and below the core region are updated on a cycle-by-cycle basis based on average cycle operating conditions.

3.4 Reactor Operating Data Inputs An accurate evaluation of reactor vessel and component fluence requires an accurate accounting of the reactor's operating history. The principal operating parameters that affect the determination of neutron fluence in light water reactors include: core configurations and fuel assembly designs, power history, exposure and isotopic distributions, and water density distributions. The following subsections provide additional information on the characterization of reactor operating data for fluence evaluations.

3.4.1 Core Configuration and Fuel Design The reactor core configuration and the fuel assembly designs loaded in the reactor determine the neutron source and spatial source distribution contributing to the irradiation of the pressure vessel, vessel internals and ex-vessel supporting structures. The Limerick 2 core is comprised of 764 fuel assemblies in a fixed configuration. Several designs of fuel assemblies may be loaded in the reactor core in any given operating cycle. In order to determine accurate spatial fluence profiles throughout the reactor system, it is important to account for the different fuel designs loaded in the reactor over the operating lifetime of the reactor, especially those designs that reside in the peripheral locations of the core region.

Table 3-2 provides a summary of the many fuel assembly designs that have been loaded in the Limerick 2 reactor core for each operating cycle evaluated in this report. Table 3-2 also identifies the dominant fuel design loaded on the core periphery for each cycle and indicates the dominant (most numerous) assembly present in bold font.

3.4.2 Reactor Power History Reactor power history is the measure of reactor power levels, reactor power spatial distributions, fuel exposure distributions, and fuel isotopic distributions that a reactor experiences over its operating life. The power history data used in the Limerick 2 fluence evaluation includes daily power levels for Cycles 1 through 14. Cycle 15 is provided as partial historical operating data with projections to end of cycle. An assumed equilibrium cycle comprised of GNF3 fuel is used to project operating conditions to reactor end of license. The power history for Limerick 2 also accounts for periods of reactor shutdown due to refueling outages and other events that affect the activation and decay of dosimetry data.

LIM-FLU-002-R-010 Revision 1 Page 3-6 of 3-9 Table 3-2 Summary of Limerick 2 Core Loading Inventory 9x9 Fuel Dominant 8x8 Fuel Designs 1Ox10 Fuel Designs Cycle Designs Peripheral Fuel GE6 GE7 GE9 GE11 GE13 GE14 GNF2 GNF3 Design 1 764 GE7 2 540 220 4 GE7 3 256 220 288 GE7 4 80 120 564 GE9 5 156 324 284 GE6 6 116 104 544 GE6 7 92 404 268 GE6 8 216 548 GE13 9 764 GE14 10 764 GE14 11 764 GE14 12 488 276 GE14 13 216 548 GE14 14 764 GNF2 15<1) 764 GNF2 16(2)+ 764 GNF3 (1) Cycle 15 is provided as partial historical operating data with projections to end of cycle.

(2) Cycle 16+ is assumed to be an equilibrium cycle for projecting fluence to the reactor end oflicense.

LIM-FLU-002-R-010 Revision 1 Page 3-7 of 3-9 Table 3-3 provides a summary of the operating history of the Limerick 2 reactor for each operating cycle. The number of statepoints used to represent the operating history and core power distributions of the reactor are listed for each cycle. Table 3-3 also shows that the reactor operated with a design rated thermal power of 3293 MWt for Cycles 1 through 4. The reactor implemented an extended power uprate at the beginning of Cycle 4 bringing the rated thermal power to 3458 MWt. The reactor implemented another power uprate in mid Cycle 12 bringing the rated thermal power to 3515 MWt, the power level at which the reactor is still operating at the time of this report.

Table 3-3 also shows the EFPY accumulated at the end of each cycle. The accumulated EFPY is computed from the operating data provided by Exelon and is verified against power production and exposure data obtained separately from plant records.

3.4.3 Reactor Statepoint Data Statepoints break up operating history into ranges of operation based on similar power, exposure, and isotopic distributions. Typically, several statepoints are chosen for each cycle to represent the different operating conditions experienced by the reactor over the course of that cycle.

Table 3-3 shows that (( )) statepoints are used to represent the first 15 cycles of operating history for the Limerick 2 reactor, with another (( )) statepoints for the projected operating cycle, for a total of (( )) statepoints.

It is also shown in Table 3-3 that the number of statepoints varies appreciably between the cycles. This variation is due to several factors but is mostly related to the availability of data to represent the operating conditions of the reactor for any given operating cycle.

Core simulator data was provided by Exelon to characterize the operating conditions of Limerick 2 for Cycles 1 through 14, partial projection Cycle 15, and the GNF3 equilibrium projection cycle. The data calculated with core simulator codes represents the best-available information about the reactor core's operating history over the reactor's operating life. In this analysis, the core simulator data provided by Exelon was processed by TransWare to generate statepoint data files for input to the fluence model.

Because core simulator codes are used for a variety of core analysis functions, 10s to 100s of core calculations may be performed to track and monitor the operation of a reactor over the course of an operating cycle. Not all core calculations are suitable for use in fluence evaluations.

Therefore, each cycle of operating data is investigated to select the statepoints that are suitable for use in fluence evaluations. When all reactor conditions are considered, the number of core simulator statepoints selected for a fluence evaluation can vary from cycle to cycle.

A separate neutronics transport calculation is performed for each selected statepoint. The neutron fluxes calculated for each statepoint are then combined with the appropriate daily power history data described in Section 3.4.2 to provide an accurate accounting of the neutron fluence for the reactor pressure vessel, reactor vessel internals, and surveillance capsules. The periods of reactor shutdown are also accounted for in this process, particularly to allow for an accurate calculation of irradiated surveillance capsule activities.

LIM-FLU-002-R-010 Revision 1 Page 3-8 of 3-9 Table 3-3 Statepoint Data for Limerick 2 per Cycle Number of Reactor Rated Thermal Power 11, Cycle Number Accumulated EFPY Statepoints (MWt) 1 (( 3293 1.2 2 3293 2.7 3 3293 4.4 4 (2) 3458 6.2 5 3458 8.2 6 3458 10.0 7 3458 11.9 8 3458 13.7 9 3458 15.7 10 3458 17.6 11 3458 19.5 12(3) 3515 21.3 13 3515 23.3 14 3515 25.1 15(4) 3515 27.0 16+(5) )) 3515 57.0 (1) The rated tlmmal power for the reactor is listed for each cycle. Note, however, that tl1e actual daily power levels are used in deteimination offluC11ce.

(2) TI1e power uprate to 3458 MWt occw1*ed in Cycle 4.

(3) TI1e power uprate to 3515 MWt occw1*ed in Cycle 12.

(4) Cycle 15 is provided as partial historical operating data with projections to Clld ofcycle.

(5) Cycle 16+ is assumed to be !lie equilibriwu cycle for projecting fluence to !lie plant end oflicense.

LIM-FLU-002-R-010 Revision 1 Page 3-9 of 3-9 3.4.4 Reactor Coolant Properties The reactor coolant water densities used in the fluence model are determined using combinations of core simulator codes and reactor heat balance data.

((

))

LIM-FLU-002-R-010 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-010 Revision 1 Page 4-1 of 4-12 4

METHODOLOGY This section provides an overview of the methodology and modeling approach used to determine fast neutron fluence for the Limerick 2 reactor pressure vessel (RPV), fast neutron fluence for the reactor vessel internals (RVI), and the fast neutron fluence and activations for the reactor surveillance capsules. The fluence model for Limerick 2 is a plant-specific model that is constructed from the design inputs described in Section 3, Description of the Reactor System.

The computational tools used in the fluence and activation analyses are based on the RAMA Fluence Methodology (RAMA) software. ((

))

4.1 Computational Method The RAMA Fluence Methodology is a system of computer codes, a data library, and an uncertainty methodology that determines best-estimate fluence and activations in light water reactor pressure vessels and vessel internal components. The primary software that comprises the methodology includes model builder codes, a particle transport code, and a fluence calculator code.

The primary inputs for the fluence methodology are mechanical design parameters and reactor operating history data. The mechanical design inputs are obtained from plant-specific design drawings, which include as-built measurements when available. The reactor operating history data is obtained from multiple sources, such as core simulator software, system heat balance calculations, daily operating logs, and cycle summary reports. A variety of outputs are available from the fluence methodology that include neutron flux, fast neutron fluence, dosimetry activation, and an uncertainty analysis.

The model builder codes consist of geometry and material processor codes that generate input for the RAMA transport code. The geometry model builder code uses mechanical design inputs and meshing specifications to generate three-dimensional geometry models of the reactor. The material processor code uses reactor operating data and material property inputs to process fuel materials, structural materials, and water densities that are consistent with the geometry meshing generated by the geometry model builder code.

The RAMA transport code performs three-dimensional neutron flux calculations using a deterministic, multigroup, particle transport theory method with anisotropic scattering ((

)). The transport solver is coupled with a general geometry modeling capability based on combinatorial geometry techniques. The coupling of general (arbitrary) geometry with a deterministic transport solver provides a flexible, efficient, and stable method for calculating neutron flux in light water reactor pressure vessels, vessel components, and structures. The primary inputs for the transport code include the geometry and

LIM-FLU-002-R-010 Revision 1 Page 4-2 of 4-12 material data generated by the model builder codes and numerical integration and convergence parameters for the iterative transport calculation. The primary output from the transport code is the neutron flux in multigroup form for every material region mesh in the fluence model.

The fluence calculator code determines fluence and activation in the reactor pressure vessel, surveillance specimens, and vessel components over specified periods of reactor operation. The fluence calculator also includes treatments for isotopic production and decay that are required to calculate specific activities for irradiated materials, such as the dosimetry specimens in the surveillance capsules. The primary inputs to the fluence calculator include the multigroup neutron flux from the transport code, response functions for the various materials in the reactor, reactor power levels for the operating periods of interest, specification of which components to evaluate, and the energy ranges of interest for evaluating neutron fluence. The reactor operating history is generally represented with several reactor statepoints that represent the core power and core power distributions of the reactor over the operating life of the reactor. These statepoints are integrated with the daily variations in reactor power levels to predict the fluence and activations accumulated throughout the reactor system.

The RAMA nuclear data library contains atomic mass data, nuclear cross-section data, response functions, and other nuclear constants that are needed for each of the code tools. The structure and contents of the data contained within the nuclear data file are based on the BUGLE-96 nuclear data library [13], with extended data representations derived from the VITAMIN-B6 data library [14].

The uncertainty methodology provides an assessment of the overall accuracy of the fluence and activation calculations. Variations in the ((

)) are evaluated to determine if there is a statistically significant bias in the calculated results that might affect the determination of the best-estimate fluence for the reactor. The plant-specific results are also weighted with comparative results from experimental benchmarks and other plant analyses and analytical uncertainties pertaining to the methodology to determine if the plant-specific model under evaluation is statistically acceptable as defined in Regulatory Guide 1.190 [1].

4.2 Fluence Model Section 3, Description of the Reactor System, describes the design inputs that were provided by Exelon for constructing the Limerick 2 reactor fluence model. These design inputs are used to develop a plant-specific, three-dimensional computer model of the Limerick 2 reactor for determining fast neutron fluence in the RPV components and for determining activation and fluence in reactor dosimetry for validating the RPV fluence predictions.

Figure 4-1 and Figure 4-2 provide general illustrations of the primary components, structures, and regions developed for the Limerick 2 fluence model. Figure 4-1 shows the planar configuration of the reactor model at an elevation corresponding to the reactor core mid-plane elevation. Figure 4-2 shows an axial configuration of the reactor model. Note that the figures are not drawn to scale and do not include representations of the meshing developed for this evaluation. The figures are intended only to provide a perspective for the layout of the model, and specifically how the various components, structures, and regions lie relative to the reactor core region (i.e., the neutron source). Additional detail is beyond the scope of this document.

LIM-FLU-002-R-010 Revision 1 Page 4-3 of 4-12

((

))

Figure 4-1 Planar View of the Limerick 2 Fluence Model at the Core Mid-Plane Elevation in Quadrant Symmetry

LIM-FLU-002-R-010 Revision 1 Page 4-4 of 4-12

((

))

Figure 4-2 Axial View of the Limerick 2 Fluence Model

LIM-FLU-002-R-010 Revision 1 Page 4-5 of 4-12 4.2.1 Geometry Model The Limerick 2 fluence model is constructed on a Cartesian coordinate system using a generalized three-dimensional geometry modeling technique based on combinatorial geometry.

The axial plane of the reactor model is defined by the (x,y) coordinates of the modeling system and the axial elevation at which a plane exists is defined along a perpendicular z-axis of the modeling system. This allows any point in the reactor model to be referenced by specifying the (x,y,z) coordinates for that point.

((

)) This modeling approach permits a model to be developed in any level of high-definition detail, such as is necessary for fluence and activation evaluations.

Figure 3-1 illustrates a planar cross-section view of the Limerick 2 reactor design at an axial elevation corresponding to the reactor core mid-plane. It is shown for this one elevation that the reactor design is a complex geometry ((

)). When the reactor is viewed in three dimensions, the varying heights of the different components, structures, and regions create additional geometry modeling complexities. An accurate representation of these geometrical complexities in a predictive computer model is essential for calculating accurate, best-estimate fluence in the reactor pressure vessel, surveillance capsules, vessel internals, and the supporting structures inside and outside of the reactor vessel.

Figure 4-1 and Figure 4-2 provide general illustrations of the planar and axial geometry complexities that are represented in the fluence model. For comparison purposes, the planar view illustrated in Figure 4-1 corresponds to the core elevation illustrated in Figure 3-1. ((

))

As previously noted, Figure 4-1 and Figure 4-2 are not drawn precisely to scale and are intended only to provide a perspective of how the various components, structures, and regions of the reactor are positioned relative to the reactor core region. The following subsections provide additional information on the constituent models developed for the individual components, structures, and regions of the fluence model.

LIM-FLU-002-R-010 Revision 1 Page 4-6 of 4-12 4.2.2 Reactor Core and Core Reflector The reactor core contains the nuclear fuel that is the source of the neutrons that irradiate all components and structures of the reactor. The core is surrounded by a shroud structure that serves to channel the reactor coolant through the core region during reactor operation. The coolant-containing region between the core and the core shroud is the core reflector. The reactor core geometry is rectangular in design ((

))

4.2.3 Reactor Core Shroud The core shroud is a canister-like structure that surrounds the reactor core. It channels the reactor coolant and steam produced by the core into the steam separators. Axially the shroud extends almost the entire height of the model and is divided into three sections: lower, central, and upper.

The lower shroud extends from the bottom of the model to the core support plate flange, the central shroud extends from the core support plate flange to the top guide flange, and the upper shroud extends from top guide flange to the top of the shroud head rim.

((

))

Above the shroud wall is the shroud head which is penetrated by numerous steam separator standpipes. (( ))

LIM-FLU-002-R-010 Revision 1 Page 4-7 of 4-12

((

))

4.2.4 Downcomer Region The downcomer region lies between the core shroud and the reactor pressure vessel. The downcomer is effectively cylindrical in design, but with geometrical complexities created by the presence of jet pumps and surveillance capsules in the region. ((

))

4.2.4.1 Jet Pumps Limerick 2 has ten jet pump assemblies in the downcomer region, which provide the main recirculation flow for the core. ((

))

4.2.4.2 Surveillance Capsules Section 3 describes the surveillance capsules installed in the Limerick 2 reactor. The three (3)

OEM surveillance capsules installed in the Limerick 2 reactor are positioned in close proximity to the RPV inner wall surface at the 30°, 120° and 300° azimuthal angles around the pressure vessel. ((

))

LIM-FLU-002-R-010 Revision 1 Page 4-8 of 4-12

((

))

4.2.5 Reactor Pressure Vessel The reactor pressure vessel and vessel cladding lie outside the downcomer region, ((

))

4.2.6 Thermal Insulation The reactor vessel thermal insulation lies in the cavity region outside the pressure vessel wall.

((

))

4.2.7 Inner and Outer Cavity Regions There are effectively two cavity regions represented in the model. The inner cavity region lies between the outer surface of the pressure vessel wall and the inner surface of the vessel insulation. The outer cavity region lies between the outer surface of the vessel insulation and inner surface of the biological shield wall cladding. ((

))

LIM-FLU-002-R-010 Revision 1 Page 4-9 of 4-12 4.2.8 Biological Shield Model The biological shield (concrete) defines the outermost region of the fluence model. ((

))

4.2.9 Above-Core Components Figure 4-2 includes illustrations of other components and regions that lie above the reactor core region. The predominant above-core components represented in the model include the top guide, core spray spargers, upper core shroud wall, shroud head, and steam separator standpipes. The shroud regions and standpipes are mentioned in further detail in Section 4.2.3.

4.2.9.1 Top Guide The top guide component lies above the core region and is appropriately modeled to include discrete representations of the top guide plates and accounting for the fuel assembly parts, top guide pads, and coolant between the plates. ((

))

4.2.9.2 Core Spray Spargers and Piping The core spray spargers include upper and lower sparger annulus pipes and a vertical inlet pipe.

((

))

4.2.10 Below-Core Components Figure 4-2 includes illustrations of other components and regions that lie below the reactor core region. The predominant below core (i.e., below active fuel) components represented in the fluence model include the lower fuel assembly parts, fuel support pieces, core support plate, core support plate rim bolts, cruciform control rods, control rod guide tubes, and lower shroud wall.

The lower shroud wall and fuel assembly components are described in previous sections, with the remaining components described in the following subsections.

4.2.10.1 Core Support Plate and Rim Bolts The core support plate includes appropriate penetrations for the fuel support pieces, control rod guide tubes, cruciform control rods, and the core support plate rim bolts. Core support plate rim bolts protrude from the top of the core support plate and traverse through the plate, rim, and core shroud lower flange. ((

))

LIM-FLU-002-R-010 Revision 1 Page 4-10 of 4-12 4.2.10.2 Fuel Support Pieces The nuclear fuel assemblies loaded in the reactor are seated on fuel support pieces, which then rest in the core support plate and control blade guide tubes. ((

))

4.2.10.3 Control Blades and Guide Tubes The fluence model allows for the representation of cruciform-shaped control blades and tubular control blade guide tubes in the below-core regions of the reactor. Coolant flow paths are included in the model ((

))

4.2.11 Summary of the Geometry Modeling Approach To summarize the reactor modeling process, there are several key features that allow the reactor design to be accurately represented for RPV fluence evaluations. ((

))

LIM-FLU-002-R-010 Revision 1 Page 4-11 of 4-12

((

))

4.3 Particle Transport Calculation Parameters The accuracy of the transport method is based on a numerical integration technique ((

))

4.4 Fission Spectrum and Neutron Source Modern core simulator software is capable of providing three-dimensional core power distributions and fuel isotopics in high-definition detail, viz., on a pin-by-pin basis. This allows fluence models to be constructed with a high-level of modeling detail for representing unique

LIM-FLU-002-R-010 Revision 1 Page 4-12 of 4-12 fission spectrum and neutron source terms for the transport calculation. ((

))

4.5 Parametric Sensitivity Analyses Several plant-specific sensitivity analyses are performed to evaluate the accuracy and predictability of the neutral particle transport methodology for determining RPV fluence.

((

))

LIM-FLU-002-R-010 Revision 1 Page 5-1 of 5-7 5

REACTOR PRESSURE VESSEL FAST NEUTRON FLUENCE This section presents the predicted best-estimate fast neutron fluence (energy > 1.0 MeV) for the Limerick 2 reactor pressure vessel (RPV) at EOC 14 (25.1 EFPY) and 57 EFPY. It is reported in Section 6, Reactor Pressure Vessel Fluence Uncertainty Analysis, that the calculated fluence does not require a bias adjustment; therefore, the calculated fluence is the best-estimate fluence for Limerick 2.

The reactor pressure vessel fast neutron fluence is determined at the interface of the RPV base metal and cladding, which is denoted as the 0T location of the RPV wall. Damage fluence through the RPV wall is reported at the 1/4T and 3/4T depths in the wall. These values are determined based on the as-built RPV wall thickness of 16.959 cm (6.668 in).

The fast neutron fluence that is used in material embrittlement evaluations should be determined using an appropriate damage function (such as displacements-per-atom of iron) rather than the computed fast neutron fluence obtained from transport calculations. Two acceptable methods for estimating the damage fluence are prescribed in Regulatory Guide 1.99 [2]. ((

))

LIM-FLU-002-R-010 Revision 1 Page 5-2 of 5-7 Plant-specific fast neutron damage fluence for Limerick 2 is presented in this report for the RPV horizontal (circumferential) welds, vertical welds, shell plates, and nozzles that reside in the RPV beltline region. The location and naming convention of the shell plates, welds, and nozzles are shown in Figure 5-1. Also shown in Figure 5-1 is the calculated RPV beltline region for the Limerick 2 reactor at 57 EFPY.

Figure 5-1 Limerick 2 RPV Beltline Region at 57 EFPY Table 5-1 through Table 5-5 report the maximum fast neutron damage fluence that is determined for each of the RPV welds, shell plates, and nozzles residing in the RPV beltline. Damage fluence is reported at the 0T, 1/4T, and 3/4T depths in the RPV wall for the reporting periods of interest. In all tables, the damage fluence that exceeds the threshold fluence of 1.0E+17 n/cm2 is shown in red font. The maximum damage fluence determined for the welds, shells and nozzles is also shown in bold font. It is observed in the tables that many welds will have exceeded the fluence threshold of 1.0E+17 n/cm2 before 57 EFPY.

Table 5-1 and Table 5-2 report the maximum damage fluence that is determined for the RPV horizontal (circumferential) and vertical welds at EOC 14 (25.1 EFPY) and 57 EFPY, respectively. The maximum damage fluence for each weld is determined to occur at the 0T depth, with the maximum fluence occurring in vertical weld BD of Shell 2 with a value of 8.71E+17 n/cm2 at 57 EFPY.

Table 5-3 reports the maximum damage fluence that is determined for each RPV shell plate at EOC 14 (25.1 EFPY) and 57 EFPY. The maximum damage fluence for each shell plate is

LIM-FLU-002-R-010 Revision 1 Page 5-3 of 5-7 determined to occur at the 0T depth, with the maximum fluence occurring in Shell 2 with a value of 1.05E+18 n/cm2 at 57 EFPY.

Table 5-4 and Table 5-5 report the maximum damage fluence that is determined for the RPV nozzles at EOC 14 (25.1 EFPY) and 57 EFPY, respectively. Damage fluence for the nozzles is presented along two paths: one along the nozzle forging-to-base-metal weld and the other along a 45-degree extraction path that extends from the inside corner of the forging to the outside surface of the RPV wall. Both paths are illustrated in Figure 5-2. It is noted that the extraction path is evaluated around the full circumference of the nozzle forging to determine the maximum fluence. The maximum damage fluence for the beltline nozzles is determined to occur at the 0T depth in the forging-to-base metal weld of the N17 nozzle with a value of 4.54E+17 n/cm2 at 57 EFPY.

Figure 5-2 Nozzle Fluence Edit Locations for Sample Nozzle Table 5-6 reports the elevations that define the RPV beltline at EOC 14 (25.1 EFPY) and 57 EFPY. It is shown that the RPV beltline at 57 EFPY covers 420.6 cm, or approximately 13.8 ft, of the reactor vessel.

LIM-FLU-002-R-010 Revision 1 Page 5-4 of 5-7 Table 5-1 Maximum Fast Neutron Damage Fluence for Limerick 2 RPV Beltline Welds at EOC 14 (25.1 EFPY)

Fast Neutron Fluence (n/cm2)

Weld Azimuth Elevation [in (cm)]

OT 1/4T 3/4T Horizontal Welds AB 25° 263.5 (669.3cm) 4.05E+17 2.75E+17 1.11E+17 AC 23° 400.5 (1017.3cm) 7.68E+15 5.64E+15 3.69E+15 Shell 1 Vertical Welds BA 317.5 ° 263.5 (669.3cm) 2.92E+17 2.00E+17 8.38E+16 BB 77.5 ° 263.5 (669.3cm) 2.84E+17 1.94E+17 8.06E+16 BC 197.5 ° 263.5 (669.3cm) 3.61E+17 2.46E+17 1.00E+17 Shell 2 Vertical Welds BO 255 ° 312.1 (792.8cm) 4.10E+17 2.79E+17 1.13E+17 BE 15° 312.1 (792.8cm) 4.12E+17 2.80E+17 1.14E+17

° BF 135 317.1 (805.4cm) 3.71E+17 2.55E+17 1.05E+17 Shell 3 Vertical Welds BG oo 400.5 (1017.3cm) 6.76E+15 5.03E+15 3.30E+15

° BH 120 400.5 (1017.3cm) 7.52E+15 5.54E+15 3.59E+15

° BJ 240 400.5 (1017.3cm) 7.52E+15 5.54E+15 3.59E+15 (1) Azimuth and elevation values are listed for the OT location only.

LIM-FLU-002-R-010 Revision 1 Page 5-5 of 5-7 Table 5-2 Maximum Fast Neutron Fluence for Limerick 2 RPV Beltline Welds at 57 EFPY Fast Neutron Fluence (n/cm2)

Weld Azimuth Elevation [in (cm)]

OT 1/4T 3/4T Horizontal Welds AB 65 ° 263.5 (669.3 cm) 7.69E+17 5.23E+17 2.12E+17 AC 68° 400.5 (1017.3 cm) 2.02E+16 1.47E+16 9.04E+15 Shell 1 Vertical Welds BA 317.5 ° 263.5 (669.3 cm) 5.36E+17 3.67E+17 1.55E+17 BB 77.5 ° 263.5 (669.3 cm) 5.71E+17 3.91E+17 1.62E+17 BC 197.5 ° 263.5 (669.3 cm) 7.09E+17 4.83E+17 1.96E+17 Shell 2 Vertical Welds BD 255 ° 323.0 (820.5 cm) 8.71E+17 5.91E+17 2.39E+17 BE 15 ° 323.0 (820.5 cm) 8.69E+17 5.91E+17 2.40E+17

° BF 135 323.0 (820.5 cm) 7.46E+17 5.11E+17 2.11E+17 Shell 3 Vertical Welds BG oo 400.5 (1017.3 cm) 1.81E+16 1.33E+16 8.17E+15 BH 120° 400.5 (1017.3 cm) 1.96E+16 1.43E+16 8.77E+15

° BJ 240 400.5 (1017.3 cm) 1.96E+16 1.43E+16 8.77E+15 (1) Azimuth and elevation values are listed for the OT location only.

LIM-FLU-002-R-010 Revision 1 Page 5-6 of 5-7 Table 5-3 Maximum Fast Neutron Damage Fluence for Limerick 2 RPV Beltline Shell Plates Fast Neutron Fluence Damage (n/cm2)

Shell Plate Azimuth Elevation [in (cm)]

OT 1/4T 3/4T EOC 14 (25.1 EFPY)

Shell 1 25 ° 263.5 (669.3 cm) 4.05E+17 2.75E+17 1.11E+17 Shell 2 24° 323.0 (820.5 cm) 5.12E+17 3.45E+17 1.38E+17 Shell 3 23° 400.5 (1017.3 cm) 7.68E+15 5.64E+15 3.69E+15 57 EFPY Shell 1 65 ° 263.5 (669.3 cm) 7.69E+17 5.23E+17 2.12E+17 Shell 2 67° 323.0 (820.5 cm) 1.05E+18 7.08E+17 2.82E+17 Shell 3 68 ° 400.5 (1017.3 cm) 2.02E+16 1.47E+16 9.04E+15 (1) Azimuth and elevation values are listed for the OT location only.

Table 5-4 Maximum Fast Neutron Damage Fluence for Limerick 2 RPV Beltline Nozzles at EOC 14 (25.1 EFPY)

Fast Neutron Fluence at (n/cm2)

Location Azimuth Elevation [in (cm)]

OT 1/4T 3/4T N2 Nozzles Weld 30° 1.11E+16 8.71E+15 6.15E+15 Forging 30° 1.57E+15 1.74E+15 3.19E+15 Weld 60° 1.10E+16 8.65E+15 6.10E+15 181.0 (459.74 cm)

Forging 60° 1.57E+15 1.74E+15 3.16E+15 Weld goo 6.11E+15 5.07E+15 4.00E+15 Forging goo 9.79E+14 1.14E+15 2.32E+15 N16 Nozzle Weld oo 5.51E+16 4.04E+16 2.02E+16 Weld 20° 366.0 (929.64 cm) 1.35E+17 9.41E+16 4.16E+16 Weld 80° 7.81E+16 5.56E+16 2.62E+16 N17 Nozzles Weld 45 ° 372.5 (946.15 cm) 1.96E+17 1.39E+17 5.97E+16

LIM-FLU-002-R-010 Revision 1 Page 5-7 of 5-7 Table 5-5 Maximum Fast Neutron Damage Fluence for Limerick 2 RPV Beltline Nozzles at 57 EFPY Fast Neutron Fluence at (n/cm2)

Location Azimuth Elevation [in (cm)]

OT 1/4T 3/4T N2 Nozzles Weld 30° 2.12E+16 1.67E+16 1.17E+16 Forging 30° 3.07E+15 3.39E+15 6.15E+15 Weld 60° 2.16E+16 1.69E+16 1.18E+16 181.0 (459.74 cm)

Forging 60° 3.13E+15 3.44E+15 6.18E+15 Weld 90° 1.32E+16 1.08E+16 8.27E+15

° Forging 90 2.15E+15 2.46E+15 4.80E+15 N16 Nozzle Weld oo 1.39E+17 1.01E+17 4.92E+16

° Weld 20 366.0 (929.64 cm) 3.29E+17 2.29E+17 9.94E+16

° Weld 80 1.97E+17 1.40E+17 6.42E+16 N17 Nozzles Weld 45° 372.5 (946.15 cm) 4.54E+17 3.19E+17 1.35E+17 Table 5-6 Reactor Beltline Elevation Range for Limerick 2 Lower Elevation Upper Elevation Axial Span of the RPV Reactor Lifetime

[in (cm)] [in (cm)] Beltline [in (cm)]

EOC 14 (25.1 EFPY) 216.9 (551.0 cm) 367.6 (933.8 cm) 150.7 (382.8 cm) 57 EFPY 210.4 (534.5 cm) 376.0 (955.0 cm) 165.6 (420.5 cm)

LIM-FLU-002-R-010 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-010 Revision 1 Page 6-1 of 6-3 6

REACTOR PRESSURE VESSEL FLUENCE UNCERTAINTY ANALYSIS This section presents the combined uncertainty analysis and the determination of bias for the Limerick 2 reactor pressure vessel (RPV) fluence evaluation. The combined uncertainty is comprised of two components. One component is the uncertainty factors developed from plant-specific measurements and the other is an analytic uncertainty factor. At this time, it is noted that no surveillance capsules have been removed from the Limerick 2 reactor. The plant-specific measurements used in this analysis are based on a database of plant measurements of similar design. When combined, the components provide a basis for determining the combined uncertainty (1) and bias in the computed RPV fluence.

The requirements for determining the combined uncertainty and bias for light water reactor pressure vessel fluence evaluations are provided in Regulatory Guide 1.190 [1]. ((

))

For pressure vessel fluence evaluations, two uncertainty factors are considered: comparison factors and uncertainty introduced by the measurement process. After analysis of these factors, it was determined that the combined uncertainty for the Limerick 2 RPV fluence was 11.2%, and that no adjustment for bias was required for the RPV fast neutron fluence presented in Section 5, Reactor Pressure Vessel Fast Neutron Fluence.

6.1 Comparison Uncertainty Comparison uncertainty factors are determined by comparing calculated activities with activity measurements. For pressure vessel fluence evaluations, two comparison uncertainty factors are considered: operating reactor comparison factors and benchmark comparison factors.

6.1.1 Operating Reactor Comparison Uncertainty TransWare has evaluated activation measurements for several BWR plants ranging from BWR/2-class plants to BWR/6-class plants. Each class of BWRs can have one or several variations of reactor core configurations, each having different radial diameters for the core shroud, reactor pressure vessel, and biological shield components. In addition, each can have different placements of the jet pumps and surveillance capsules in the reactor vessels.

The Limerick 2 reactor is a BWR/4 class design. ((

))

LIM-FLU-002-R-010 Revision 1 Page 6-2 of 6-3

(( )) The overall comparison ratio for all BWR class reactors evaluated as of the date of this repo1t is 1.00 +/- 0.10. ((

))

6.1.2 Benchmark Comparison Uncertainty The benchmark comparison unce1tainty is based on a set of industiy standard simulation benchmark comparisons. fu accordance with the guidelines provided in Regulato1y Guide 1.190, it is appropriate to include comparisons of vessel simulation benchmark measurements in the overall fluence uncertainty evaluation. Two vessel simulation benchmarks are evaluated: the Pool Critical Assembly (PCA) and VENUS-3 experimental benchmarks.

The PCA experimental benchmark includes (( )) activation measurements at the mid-plane elevation in various simulated reactor components. The VENUS-3 expe1imental benchmark includes (( )) activation measurements at a range of elevations in various simulated reactor components. Table 6-1 summarizes the calculated-to-measurement (C/M) results detennined for these vessel simulation benchmarks.

Table 6-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements Average Number of Benchmark Calculated-to- St. Dev. (1a)

Measurements Measured (C/M)

Pool Critical Assembly ((

VENUS-3 Total Simulated Vessel Comparisons ))

6.2 Analytic Uncertainty The calculational models used for fluence analyses are comprised of numerous analytical parameters that have associated uncertainties in their values. The uncertainty in these parameters needs to be tested for its contribution to the overall fluence unce1tainty.

((

))

LIM-FLU-002-R-010 Revision 1 Page 6-3 of 6-3

((

))

6.3 Combined Uncertainty The combined uncertainty for the reactor pressure vessel fluence evaluation is detennined with a weighting function ((

)). Table 6-2 shows that the combined uncertainty (lcr) detennined for the Limerick 2 reactor pressure vessel fluence is 11.2% for neutron energy exceeding 1.0 MeV.

It is shown in Table 6-2 that the combined unce1iainty is well below the 20% unce1iainty limit specified in Regulato1y Guide 1.190. In accordance with Regulato1y Guide 1.190, there is no discemable bias in the computed RPV fluence. Therefore, no adjustment to the RPV fast neutron fluence that is presented in Section 5, Reactor Pressure Vessel Fast Neutron Fluence, is required.

Table 6-2 Limerick 2 Combined RPV Uncertainty for Energy > 1.0 MeV Uncertainty Term Value Combined Uncertainty (1o) 11.2%

Bias None< 1 l

1) The bias term is less than its constituent uncertainty values, concluding that no statistically significant bias exists.

LIM-FLU-002-R-010 Revision 1

[ This page intentionally left blank. ]

LIM-FLU-002-R-010 Revision 1 Page 7-1 of 7-4 7

REFERENCES 7.1 References

1. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.190: Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Washington, D.C.: Office of Nuclear Regulatory Research, 2001.

2. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.99: Radiation Embrittlement of Reactor Vessel Materials. Washington, D.C.: Office of Nuclear Regulatory Research, 1988.

3. ((

))

4. U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Safety Evaluation Report with Open Items Related to the License Renewal of Seabrook Station.

Docket Number 50-443. Washington, D.C.: Office of Nuclear Reactor Regulation, 2012.

5. ((

))

6. U.S. National Archives and Records Administration. Code of Federal Regulations. Title 10, Appendix G to Part 50 - Fracture Toughness Requirements. 2013.

7. U.S. National Archives and Records Administration. Code of Federal Regulations. Title 10, Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements. 2008.

8. U.S. National Archives and Records Administration. Code of Federal Regulations. Title 10, Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. 2007.

9. U.S. National Archives and Records Administration. Code of Federal Regulations. Title 10, Part 21 - Reporting of Defects and Noncompliance. 2015.

10. Limerick Generating Station Unit 2 Reactor Pressure Vessel Fluence Evaluation, TransWare Enterprises Inc., LIM-FLU-001-R-004, Rev 0, June 2010.

11. ((

))

12. ((

))

LIM-FLU-002-R-010 Revision 1 Page 7-2 of 7-4

13. Oak Ridge National Laboratory. Radiation Safety Information Computational Center.

BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. RSICC Data Library Collection, DLC-185. Oak Ridge, TN, 1996.

14. Oak Ridge National Laboratory. Radiation Safety Information Computational Center.

VITAMIN-B6: A Fine-Group Cross Section Library Based on ENDF/B-VI Release 3 for Radiation Transport Applications. RSICC Data Library Collection, DLC-184. Oak Ridge, TN, 1996.

15. E. N. Jones, Comparison of Regulatory Guide 1.99 Fluence Attenuation Methods, Journal of ASTM International. Vol. 9, No. 4, pp. 1-7 (2012).

16. ((

))

LIM-FLU-002-R-010 Revision 1 Page 7-3 of 7-4 7.2 Glossary AZIMUTHAL QUADRANT SYMMETRY - A type of core and pressure vessel azimuthal representation that represents a single quadrant of the reactor that can be rotated and mirrored to represent the entire 360-degree geometry. For example, the northeast quadrant can be mirrored to represent the northwest and southeast quadrants and can be rotated to represent the southwest quadrant.

BEST-ESTIMATE NEUTRON FLUENCE - See Neutron Fluence.

BOC - An acronym for beginning-of-cycle.

CALCULATED NEUTRON FLUENCE - See Neutron Fluence.

CALCULATIONAL BIAS - A calculational adjustment based on comparisons of calculations to measurements. If a bias is determined to exist, it may be applied as a multiplicative correction to the calculated fluence to produce the best-estimate neutron fluence.

CORE BELTLINE - The axial elevations corresponding to the active fuel height of the reactor core.

DAMAGE FLUENCE - See Neutron Fluence.

DPA - An acronym for displacements per atom which is typically used to characterize material damage in ferritic steels due to neutron exposure.

EFFECTIVE FULL POWER YEARS (EFPY) - A unit of measurement representing one full year of operation at the reactors rated power level. For example, if a reactor operates for 12 months at full rated power, this represents 1.0 EFPY. If the reactor operates for 10 months at full rated power, then goes into a power uprate and continues operating for another 2 months at the new full rated power, this also represents 1.0 EFPY.

EOC - An acronym for end-of-cycle.

EXTENDED BELTLINE REGION - See RPV beltline.

FAST NEUTRON FLUENCE - Fluence accumulated by neutrons with energy greater than 1.0 MeV (>1.0 MeV).

NEUTRON FLUENCE - Time-integrated neutron flux reported in units of n/cm2. The term best-estimate fluence refers to the fast neutron fluence that is computed in accordance with the requirements of U. S. Nuclear Regulatory Commission Regulatory Guide 1.190. The term damage fluence, which is required for material embrittlement evaluations, refers to an adjusted fast neutron fluence that is determined using damage functions specified in U. S. Nuclear Regulatory Commission Regulatory Guide 1.99.

NOZZLE EXTRACTION PATH - The path, or trajectory through the nozzle blend radius along which fluence is determined for the nozzle.

NOZZLE FORGING-TO-BASE-METAL WELD - The weld between the nozzle forging and the RPV base metal materials. This is sometimes referred to as the Nozzle-to-Shell Weld.

OEM - An acronym for original equipment manufacturer.

LIM-FLU-002-R-010 Revision 1 Page 7-4 of 7-4 RPV - An acronym for reactor pressure vessel. Unless otherwise noted, the reactor pressure vessel refers to the base metal material of the RPV wall (i.e., excluding clad/liner).

RPV BELTLINE - The RPV beltline is defined as that portion of the RPV adjacent to the reactor core that attains sufficient neutron radiation damage that the integrity of the pressure vessel could be compromised. For purposes of this evaluation, the fast neutron fluence threshold used to define the traditional RPV beltline is 1.0E+17 n/cm2. The axial span of the RPV that can exceed this threshold includes the RPV shells, welds, and heat-affected zones. An extended beltline is also defined to include lower fluence regions of the pressure vessel but with higher stresses than the traditional beltline region, such as RPV nozzles. The combination of fluence and stress may result in a limiting location in the pressure vessel for determining pressure-temperature limits.

RPV ZERO ELEVATION - The RPV zero elevation is defined at the inside surface of the lowest point in the vessel bottom head, which is typically the bottom drain plug location. Axial elevations presented in this report are relative to RPV zero.

RVI - An acronym for reactor vessel internals.