william r. gideon (-,duke energy. - nrc: home pagewilliam r. gideon vice president brunswick nuclear...

( -, DUKE ENERGY.

April 4, 2018

Serial: BSEP 18-0041

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

Subject: Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324

William R. Gideon Vice President

Brunswick Nuclear Plant P.O. Box 10429

Southport, NC 28461

o: 910.832.3698

10 CFR 50.90

Request for License Amendment to Revise the Technical Specifications to Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report

Ladies and Gentlemen:

In accordance with 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), is requesting a license amendment for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.

The proposed amendment modifies the Technical Specifications (TS), Section 1.0 (i.e., Definitions), TS 3.4.9 (i.e., RCS Pressure and Temperature (PIT) Limits), and TS 5.6 (i.e., Reporting Requirements) to delete reference to the pressure and temperature (P-T) limit curves, currently located in TS, and to include reference to the Pressure and Temperature Limits Report (PTLR).

The enclosure of this letter provides an evaluation supporting the proposed changes. Attachments 1 and 2 of the Enclosure provide existing TS pages marked up to show the proposed changes for BSEP, Unit 1 and Unit 2, respectively. Attachment 3 provides the Unit 1 TS Bases pages marked up to show the proposed changes and are being provided for information only. Attachments 4 and 5 provide typed TS pages reflecting the proposed changes for Units 1 and 2, respectively. Attachment 6 provides the BSEP, Units 1 and 2 Pressure and Temperature Limits Report (PTLR).

The guidance of NRC Generic Letter (GL) 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, 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, which has received NRC approval, was followed in development of the proposed TS changes.

Duke Energy requests approval of the proposed amendment by April 4, 2019, with a 120-day implementation period. The current Unit 1 and 2 P-T limit curves expire at 32 EFPY, which the first BSEP unit (i.e., Unit 2) is expected to reach in May 2019. The requested approval date for the proposed amendment supports establishment of updated P-T limit curves prior to reaching this limit.

U.S. Nuclear Regulatory Commission of 3

In accordance with 1 0 CFR 50.91, Duke Energy is providing a copy of the proposed license amendment to the designated representative for the State of North Carolina.

This document contains no new regulatory commitments.

Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 832-2487.

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on April 4, 2018.

~ William R. Gideon

WRM/wrm

Enclosure:

Description and Assessment of the Proposed Change

Attachment 1: Proposed Technical Specification Changes (Mark-Up) Unit 1 Attachment 2: Proposed Technical Specification Changes (Mark-Up) Unit 2 Attachment 3: Proposed Technical Specification Bases Changes (Mark-up) Unit 1

Attachment 4: Attachment 5: Attachment 6:

(For Information Only) Revised (Typed) Technical Specification Pages Unit 1 Revised (Typed) Technical Specification Pages Unit 2 Pressure Temperature Limits Report (PTLR)

U.S. Nuclear Regulatory Commission of 3

cc: U. S. Nuclear Regulatory Commission, Region II ATTN: Ms. Catherine Haney, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257

U. S. Nuclear Regulatory Commission ATTN: Mr. Gale Smith, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869

U. S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon 11555 Rockville Pike Rockville, MD 20852-2738

Chair - North Carolina Utilities Commission (Electronic Copy Only) P.O. Box 29510 Raleigh, NC 27626-0510 [email protected]

Mr. W. Lee Cox, III, Section Chief (Electronic Copy Only) Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-1645 [email protected]

BSEP 18-0041 Enclosure

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Description and Assessment of the Proposed Change

Subject: Request for License Amendment to Revise the Technical Specifications to Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report

1. SUMMARY DESCRIPTION

2. DETAILED DESCRIPTION

2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Technical Specification Changes 2.5 Description of the Technical Specification Bases Changes

3. TECHNICAL EVALUATION

3.1 Background 3.2 Technical Analysis

4. REGULATORY EVALUATION

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

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES

ATTACHMENTS:

1. Proposed Technical Specification Changes (Mark-Up) - Unit 1 2. Proposed Technical Specification Changes (Mark-Up) - Unit 2 3. Proposed Technical Specification Bases Changes (Mark-up) - Unit 1 (For Information Only) 4. Revised (Typed) Technical Specification Pages - Unit 1 5. Revised (Typed) Technical Specification Pages - Unit 2 6. Pressure Temperature Limits Report (PTLR)


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1. SUMMARY DESCRIPTION

The proposed change revises the Technical Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2, by replacing the pressure and temperature (P-T) limit curves (i.e., TS Figures 3.4.9-1, 3.4.9-2, 3.4.9-3, 3.4.9-4, and 3.4.9-5) with references to the Pressure and Temperature Limits Report (PTLR).

The PTLR contains updates to the P-T limit curves for the beltline, bottom head, and non-beltline regions for the BSEP Unit 1 and 2 reactor pressure vessels (RPV). The P-T curves are developed for 54 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 (i.e., Structural Integrity Associates, Inc. Report SIR-05-044, Revision 1-A), Pressure Temperature Limits Report Methodology for Boiling Water Reactors (i.e., 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 (i.e., Reference 2).

The guidance of NRC Generic Letter (GL) 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, (i.e., 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 (i.e., Reference 4), which has received NRC approval, was followed in development of the proposed TS changes.

2. DETAILED DESCRIPTION

2.1 System Design and Operation

Pressure-temperature limit curves have been developed for the RPV in accordance with 10 CFR 50 Appendix G, ASME Code Section XI, Appendix G, and NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, as applicable. These P-T curves account for material property changes due to radiation and ensure that a postulated surface defect having a depth of 1/4 of the RPV material thickness can be safely accommodated in the vessel shell without promoting brittle fracture.

2.2 Current Technical Specifications Requirements

TS Limiting Condition for Operation 3.4.9 requires that Reactor Coolant System (RCS) pressure, temperature, and heatup and cooldown rates be maintained within the pressure and temperature limits specified in TS Figure 3.4.9-1 during normal operation with the core not critical, TS Figure 3.4.9-2 during normal operation with the core critical, and TS Figures 3.4.9-3 through 3.4.9-5 during hydrostatic tests and leak tests. The current P-T limit curves in the TSs are applicable to plant operation up to 32 Effective Full Power Years (EFPY).

2.3 Reason for the Proposed Change

The current Unit 1 and 2 P-T limit curves expire at 32 EFPY, which the first BSEP unit (i.e., Unit 2) is expected to reach in approximately May 2019. The requested approval date for the proposed amendment supports establishment of updated P-T limit curves prior to reaching this limit.


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2.4 Description of the Technical Specification Changes

This license amendment request revises the reactor coolant system P-T limit curves (i.e., TS Figures 3.4.9-1, 3.4.9-2, 3.4.9-3, 3.4.9-4, and 3.4.9-5) and relocates them to a PTLR, as follows:

1. Adds a definition in TS Section 1.0 for the Pressure and Temperature Limits Report. The wording of this definition is consistent with NUREG-1433, Revision 4, Standard Technical Specifications – General Electric BWR/4 Plants.

2. Revises TS 3.4.9, RCS Pressure and Temperature (P/T) Limits, to refer to the PTLR.

3. Combines existing Surveillance Requirement (SR) 3.4.9.1 and existing SR 3.4.9.2. The new SR 3.4.9.1 verifies that RCS pressure, RCS temperature, and RCS heatup and cooldown rates are within the limits specified in the PTLR.

4. Renumbers existing SR 3.4.9.3 through SR 3.4.9.8 as SR 3.4.9.2 through SR 3.4.9.7. Revises these SRs to reference the PTLR for specified limits.

5. Removes the present pressure-temperature curves, Figures 3.4.9-1, 3.4.9-2, 3.4.9-3, 3.4.9-4, and 3.4.9-5.

6. Adds a new Specification 5.6.7, Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR), in Section 5.0, Administrative Controls. The new specification is consistent in format and content with the NUREG-1433 Standard Technical Speculations (STS) and includes:

The individual TSs that address reactor coolant system pressure and temperature limits,

References the NRC approved topical report which documents the PTLR methodology, and

Requires the PTLR and any revisions or supplements to be submitted to the NRC.

7. Revises TS Bases 3.4.9 to refer to the pressure and temperature limit curves specified in the PTLR.

The marked-up TS pages for BSEP Units 1 and 2 are provided in Attachments 1 and 2, respectively.

2.5 Description of Technical Specification Bases Changes

The proposed changes to the TS Bases are provided in Attachment 3 and are provided for information only. Changes to the attached TS Bases pages will be incorporated in accordance with TS 5.5.10, Technical Specifications (TS) Bases Control Program.

3. TECHNICAL EVALUATION

3.1 Background

10 CFR 50.60 requires that light-water nuclear power reactors meet the fracture toughness requirements for the reactor coolant pressure boundary set forth in Appendix G to 10 CFR 50. Appendix G is the regulatory basis for P-T curves for light water reactors. Appendix G specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the


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reactor coolant pressure boundary to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Appendix G also requires that the reference temperature and Charpy upper-shelf energy for reactor vessel beltline materials account for the embrittlement caused by neutron fluence over the life of the vessel.

Historically, utilities have submitted license amendment requests (LARs) to update their P-T curves. Processing these LARs has caused both the NRC and licensees to expend resources that could otherwise be devoted to other activities.

In BSEP Amendments 228 and 256 for Units 1 and 2, respectively (i.e., Reference 5), the curves were calculated using Westinghouse Electric Company, LLC vessel fluence methodology and based on the methodology specified in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Code Case N-640, Alternative Reference Fracture Toughness for Development of P-T Limit Curves for ASME Section XI, Division I, the 1989 ASME Code, Section XI, Appendix G, and Appendix G of 10 CFR Part 50. Adjusted reference temperatures at the nil ductility transition values were developed for the reactor pressure vessel materials in accordance with Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials. The NRC approved revising these reactor pressure vessel pressure-temperature limits and their validity to 32 effective full-power years (EFPY). These are the current P-T limits in the Brunswick Unit 1 and Unit 2 TS.

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 GL 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 NRC agreed with the industry that the curves may be relocated outside the plant TS to a PTLR so that the licensee could efficiently maintain these limits. One of the prerequisites for having the PTLR option is that the P-T curves and limits be derived using methodologies approved by the NRC, and that the associated licensing topical reports describing the approved methodologies be referenced in the plant TS.

By letter dated November 17, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML113260534), the Boiling Water Reactor Owners' Group (BWROG) submitted LTR BWROG-TP-11-022, Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors, to the NRC (i.e., Reference 10). By letter dated May 16, 2013 (i.e., Reference 8), the NRC 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 (i.e., Reference 4). The BWROG issued the final report on September 4, 2013 (i.e., Reference 9), which contains the final SER, along with the NRC requests for additional information (RAIs), and the BWROG's responses to the NRC RAIs.


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The NRC 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.

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

3.2 Technical Analysis

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 NRC-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.

Revised P-T curves were developed for hydrostatic pressure and leak tests, and normal operation with core not critical and core critical conditions. A report describing the inputs, methodology and results for the revised curves is provided in Attachment 6. The revised curves have been developed for application up to 54 EFPY.

The revised Brunswick P-T curves were prepared using the methods documented in the BWROG-TP-11-022-A (SIR 05 044, Revision 1-A), Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors (i.e., Reference 1). This BWROG LTR provides NRC-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 (i.e., Reference 4) and GL 96-03 (i.e., Reference 3). The LTR satisfies the requirements of 10 CFR 50, Appendix G, and the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, Nonmandatory Appendix G.

The licensing topical report 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 NRC 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.

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


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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 NRC-approved methodology and in accordance with NRC Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (i.e., Reference 6).

The fluence is based upon operation for 54 EFPY. The calculated neutron fluences at the end of 54 EFPY are provided below:

Parameter Fluence, Unit 1 Fluence, Unit 2 Peak Surface 3.27 x 1018 n/cm2

3.33 x 1018 n/cm2 Limiting Beltline Material Peak ¼ T

2.35 x 1018 n/cm2 2.40 x 1018 n/cm2

Instrument Nozzle Vessel Inside Diameter

1.26 x 1018 n/cm2 1.28 x 1018 n/cm2

Instrument Nozzle Peak ¼ T Limiting

9.06 x 1017 n/cm2 9.22 x 1017 n/cm2

BSEP Surveillance Capsule Results and Adjusted Reference Temperature:

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. In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, one surveillance capsule has been removed and tested from each of the BSEP Unit 1 and Unit 2 reactor vessels. The first Unit 1 capsule was removed in Summer 1993, at 8.67 EFPY and the first Unit 2 capsule was removed in Spring 1996, at 10.9 EFPY. The surveillance capsules contained 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.

BSEP has replaced the original RPV material surveillance program with the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). BSEP is currently committed to use the BWRVIP ISP, and has made a license renewal commitment to use the ISP for BSEP 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


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approved by the NRC. Use of the BWRVIP ISP for BSEP was approved by the NRC on January 14, 2004 (i.e., Reference 7). Under the ISP, no further capsules are scheduled for removal from the BSEP Unit 1 and 2 vessels. Representative surveillance capsule materials for the BSEP Unit 1 and 2 limiting beltline plate are contained in the River Bend and Supplemental Surveillance Program (SSP) Capsules C, F, and H. Representative materials for the BSEP Unit 1 and 2 limiting beltline weld are in the Duane Arnold and SSP-F surveillance capsules. No further SSP capsules are scheduled for withdrawal. The next River Bend surveillance capsule is scheduled to be withdrawn and tested under the ISP in approximately 2025, and the next Duane Arnold surveillance capsule is scheduled for withdrawal and testing in approximately 2027.

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 NRC, and include 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 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 Attachments 1 and 2 of this Enclosure. These proposed changes are consistent with the guidance provided in GL 96-03, as supplemented by TSTF-419-A.

Conditions and Limitations:

The NRC 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.

Duke Energy has 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 Brunswick PTLR.


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4. 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 NRC 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. BSEP demonstrates its compliance with the requirements of 10 CFR 50, Appendix H, through participation in the BWRVIP Integrated Surveillance Program (ISP) (i.e., Reference 7) and the latest material information was utilized in preparation of the report.

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.

Duke Energy has 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.


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4.2 Precedent

The NRC has approved similar license amendments to relocate the P-T limit curves to a PTLR. Recent examples for BWR plants include:

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

2. Cooper Nuclear Station, License Amendment No. 256 issued by NRC letter dated July 25, 2016, ADAMS Accession No. ML16158A022

3. Duane Arnold Energy Center, License Amendment No. 294 issued by NRC letter dated July 25, 2016, ADAMS Accession No. ML16180A086

4. Nine Mile Point Nuclear Station, Unit 1, License Amendment No. 204 issued by NRC letter dated January 21, 2010, ADAMS Accession No. ML093370002

5. Hope Creek Generating Station, Amendment No. 209 issued by NRC letter dated December 14, 2017, ADAMS Accession No. ML17324A840

4.3 No Significant Hazards Consideration Determination Analysis

Duke Energy Progress, LLC (Duke Energy), is requesting an amendment to Renewed Facility Operating License Nos. DPR-71 and DPR-62 for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed Technical Specification (TS) changes modify the Brunswick TS by replacing references to existing reactor vessel heatup and cooldown rate limits and pressure and temperature limit curves with references to the Pressure and Temperature Limits Report (PTLR).

Duke Energy has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No

The proposed license amendment adopts the NRC approved methodology described in Boiling Water Reactor Owner’s Group (BWROG) Licensing Topical Report (LTR) (BWROG-TP-11-022-A, SIR-05-044, Revision 1-A), Pressure Temperature Limits Report Methodology for Boiling Water Reactors. The BSEP PTLR was developed based on the methodology and template provided in the BWROG LTR.

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

Implementing this NRC 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


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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 amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No

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 amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change 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 Brunswick pressure temperature limits using the NRC approved structural integrity 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 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 amendment does not result in a significant reduction in the margin of safety.


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Based on the above, Duke Energy 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 Conclusion

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

5. ENVIRONMENTAL CONSIDERATION

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

6. REFERENCES

1. Boiling Water Reactor Owner's Group (BWROG) Licensing Topical Report (LTR) BWROG-TP-11-022-A, Revision 1 (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, 1989 Edition, No Addenda

3. NRC 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

5. Letter from Brenda L. Mozafari (NRC) to J. S. Keenan, Brunswick Steam Electric Plant, Units 1 and 2 – Issuance of Amendment Re: Pressure-Temperature Limit Curves (TAC Nos. MB5579 and MB5580), dated June 18, 2003

6. NRC Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence

7. Letter from Margaret H. Chernoff (NRC) to C. J. Gannon (Carolina Power & Light Company), Brunswick Steam Electric Plant, Units 1 and 2 – Issuance of Amendments Regarding the Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program (TAC Nos. MC0254 and MC0255), dated January 14. 2004, ADAMS Accession Number ML040150192

8. Letter from Sher Bahadur (NRC) to Frederick Schiffley (BWROG), Final Safety Evaluation for Boiling Water Reactor Owners' Group Topical Report


Page 12 of 12

BWROG-TP-11-022, Revision 1, November 2011, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors" (TAC No. ME7649), dated May 16, 2013, ADAMS Accession Number ML13107A062

9. Letter from Frederick 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, ADAMS Accession Number ML13277A557

10. Letter from Frederick Schiffley (BWROG) to U.S. Nuclear Regulatory Commission Document Control Desk, BWR Owners' Group Submittal of Revision 1 of Structural Integrity Associates Fracture Mechanics Methodology Licensing Topical Report, dated November 17, 2011, ADAMS Accession Number ML113260534


Attachment 1

Proposed Technical Specification Changes (Mark-Up) Unit 1

Definitions 1.1

Brunswick Unit 1 1.1-5 Amendment No. 277

1.1 Definitions (continued)

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

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 2923 MWt. REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval from SYSTEM (RPS) RESPONSE when the monitored parameter exceeds its RPS trip setpoint TIME at the channel sensor until de-energization of the scram pilot

valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is

subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is ≥ 68F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

(continued)

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PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including heatup REPORT (PTLR) and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

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RCS P/T Limits 3.4.9


3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.9 RCS Pressure and Temperature (P/T) Limits

LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within limits.

APPLICABILITY: At all times.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. --------------NOTE----------------Required Action A.2 shall be completed if this Condition is entered. --------------------------------------

Requirements of the LCO not met in MODE 1, 2, or 3.

A.1 Restore parameter(s) to within limits.

AND

A.2 Determine RCS is acceptable for continued operation.

30 minutes

72 hours

B. Required Action and associated Completion Time of Condition A not met.

B.1 Be in MODE 3.

AND

B.2 Be in MODE 4.

12 hours

36 hours

(continued)

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the limits specified in the PTLR.

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ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION

TIME

C. ---------------NOTE---------------Required Action C.2 shall be completed if this Condition is entered. --------------------------------------

Requirements of the LCO not met in other than MODES 1, 2, and 3.

C.1 Initiate action to restore parameter(s) to within limits.

AND

C.2 Determine RCS is acceptable for operation.

Immediately

Prior to entering MODE 2 or 3.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.4.9.1 -------------------------------NOTE-------------------------------- Only required to be performed during RCS heatup and cooldown operations. ------------------------------------------------------------------------

Verify:

a. RCS pressure and RCS temperature are within the applicable limits specified in Figures 3.4.9-1 and 3.4.9-2; and

b. RCS heatup and cooldown rates are 100F in any 1 hour period.

In accordance with the Surveillance Frequency Control Program

(continued)

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

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and RCS inservice leak and hydrostatic testing.



SURVEILLANCE REQUIREMENTS (continued)


SR 3.4.9.2 ------------------------------NOTE--------------------------------- Only required to be performed during RCS inservice leak and hydrostatic testing. ------------------------------------------------------------------------

Verify:

a. RCS pressure and RCS temperature are within the applicable limits specified in Figure 3.4.9-3; 3.4.9-4, or 3.4.9-5, as applicable.



SR 3.4.9.3 Verify RCS pressure and RCS temperature are within the criticality limits specified in Figure 3.4.9-2.

Once within 15 minutes prior to control rod withdrawal for the purpose of achieving criticality

SR 3.4.9.4 -------------------------------NOTE-------------------------------- Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start. ------------------------------------------------------------------------

Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is 145F.

Once within 30 minutes prior to each startup of a recirculation pump

(continued)

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the PTLR.

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within the limits specified in the PTLR.

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COMBINED WITH SR 3.4.9.1.

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3.4.9.2

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3.4.9.3





SR 3.4.9.5 ------------------------------NOTE--------------------------------- Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start. -----------------------------------------------------------------------

Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is 50F.


SR 3.4.9.6 ------------------------------NOTE--------------------------------- Only required to be performed when tensioning the reactor vessel head bolting studs. -----------------------------------------------------------------------

Verify reactor vessel flange and head flange temperatures are 70F.


SR 3.4.9.7 ------------------------------NOTE--------------------------------- Not required to be performed until 30 minutes after RCS temperature 80F in MODE 4. ----------------------------------------------------------------------



SR 3.4.9.8 ------------------------------NOTE--------------------------------- Not required to be performed until 12 hours after RCS temperature 100F in MODE 4. -----------------------------------------------------------------------



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3.4.9.4

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Figure 3.4.9-1 (page 1 of 1) RCS Pressure and Temperature Limits Normal Operation with Core Not Critical

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OPERATE TO RIGHT AND/OR BELOWLIMITING LINE

BASES:1. FUEL IN REACTOR2. ≤ 32 EFPY3. 5.50x1017 n/cm2 > 1 MEV (Limiting Location, 1/4t)4. BELTLINE LIMITING RTNDT = 120.0°F (1/4t)5. BOTTOM HEAD REGION RTNDT = 10.0°F (1/4t)6. 30 PSI, 10.0°F INSTRUMENT CORRECTION INCLUDED7. REG. GUIDE 1.99 REV. 2

BOLTUP 70°F

NORMAL OPERATIONCORE NOT CRITICAL - UNIT 1

HEATUP / COOLDOWN

283 psig

729 psig

705 psig

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Relocated to the PTLR.

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Figure 3.4.9-2 (page 1 of 1) RCS Pressure and Temperature Limits

Normal Operation with Core Critical

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283 psig

486 psig

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Hydrostatic and Leak Tests

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BASES:1. FUEL IN REACTOR2. ≤ 20 EFPY3. 3.10x1017 n/cm2 > 1 MEV (Limiting Location, 1/4t)4. BELTLINE LIMITING RTNDT = 103.2.0°F (1/4t)5. BOTTOM HEAD REGION RTNDT = 10.0°F (1/4t)6. 30 PSI, 10.0°F INSTRUMENT CORRECTION INCLUDED7. REACTOR NOT CRITICAL8. REG. GUIDE 1.99, REV. 2.

BOLTUP 70°F

HYDROSTATIC PRESSURE TEST - UNIT 1

283 psig

762 psig780 psig

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BASES:1. FUEL IN REACTOR2. ≤ 24 EFPY3. 3.90x1017 n/cm2 > 1 MEV (Limiting Location, 1/4t)4. BELTLINE LIMITING RTNDT = 110.9°F (1/4t)5. BOTTOM HEAD REGION RTNDT = 10.0°F (1/4t)6. 30 PSI, 10.0°F INSTRUMENT CORRECTION INCLUDED7. REACTOR NOT CRITICAL8. REG. GUIDE 1.99, REV. 2.BOLTUP

70°F


283 psig

716 psig

780 psig

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BASES:1. FUEL IN REACTOR2. ≤ 32 EFPY3. 5.50x1017 n/cm2 > 1 MEV (Limiting Location, 1/4t)4. BELTLINE LIMITING RTNDT = 120.0°F (1/4t)5. BOTTOM HEAD REGION RTNDT = 10.0°F (1/4t)6. 30 PSI, 10.0°F INSTRUMENT CORRECTION INCLUDED7. REACTOR NOT CRITICAL8. REG. GUIDE 1.99, REV. 2.

BOLTUP 70°F


283 psig

671 psig

780 psig

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Reporting Requirements 5.6


5.6 Reporting Requirements (continued)

5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

20. BAW-10247PA, Realistic Thermal-Mechanical Fuel RodMethodology for Boiling Water Reactors, Revision 0, April 2008.

21. ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation,Revision 1, March 2014.

c. The core operating limits shall be determined such that all applicablelimits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits,Emergency Core Cooling Systems (ECCS) limits, nuclear limits such asSDM, transient analysis limits, and accident analysis limits) of the safetyanalysis are met.

d. The COLR, including any midcycle revisions or supplements, shall beprovided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report

When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

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Attachment 2

Proposed Technical Specification Changes (Mark-Up) Unit 2

Definitions 1.1



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

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to (RTP) the reactor coolant of 2923 MWt. REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval from SYSTEM (RPS) RESPONSE when the monitored parameter exceeds its RPS trip setpoint TIME at the channel sensor until de-energization of the scram pilot








(continued)

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PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including heatup REPORT (PTLR) and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.





LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within limits.


ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. --------------NOTE---------------Required Action A.2 shall be completed if this Condition is entered. -------------------------------------



AND


30 minutes

72 hours


B.1 Be in MODE 3.

AND

B.2 Be in MODE 4.

12 hours

36 hours

(continued)

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the limits specified in the PTLR.

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ACTIONS (continued)


TIME

C. --------------NOTE---------------Required Action C.2 shall be completed if this Condition is entered. -------------------------------------



AND


Immediately




SR 3.4.9.1 --------------------------------NOTE------------------------------- Only required to be performed during RCS heatup and cooldown operations. ------------------------------------------------------------------------

Verify:

a. RCS pressure and RCS temperature are within the applicable limits specified in Figures 3.4.9-1 and 3.4.9-2; and



(continued)

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and RCS inservice leak and hydrostatic testing.

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SR 3.4.9.2 -------------------------------NOTE-------------------------------- Only required to be performed during RCS inservice leak and hydrostatic testing. ------------------------------------------------------------------------

Verify:

a. RCS pressure and RCS temperature are within the applicable limits specified in Figure 3.4.9-3; 3.4.9-4, or 3.4.9-5, as applicable;



SR 3.4.9.3 Verify RCS pressure and RCS temperature are within the criticality limits specified in Figure 3.4.9-2.



Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is 145F.


(continued)

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COMBINED WITH SR 3.4.9.1.

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the PTLR.

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3.4.9.2

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3.4.9.3

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Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is 50F.


SR 3.4.9.6 -------------------------------NOTE-------------------------------- Only required to be performed when tensioning the reactor vessel head bolting studs. ------------------------------------------------------------------------



SR 3.4.9.7 -------------------------------NOTE-------------------------------- Not required to be performed until 30 minutes after RCS temperature 80F in MODE 4. ------------------------------------------------------------------------



SR 3.4.9.8 -------------------------------NOTE-------------------------------- Not required to be performed until 12 hours after RCS temperature 100F in MODE 4. ------------------------------------------------------------------------



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3.4.9.4

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3.4.9.5

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3.4.9.7

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Figure 3.4.9-1 (page 1 of 1) RCS Pressure and Temperature Limits Normal Operation with Core Not Critical

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Normal Operation with Core Critical

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BASES:1. FUEL IN REACTOR2. ≤ 32 EFPY3. 5.50x1017 n/cm2 > 1 MEV (Limiting Location, 1/4t)4. BELTLINE LIMITING RTNDT = 111.9°F (1/4t)5. BOTTOM HEAD REGION RTNDT = 40.0°F (1/4t)6. 30 PSI, 10.0°F INSTRUMENTCORRECTION INCLUDED7. REG. GUIDE 1.99 REV. 2

BOLTUP 80°F

NORMAL OPERATIONCORE CRITICAL - UNIT 2HEATUP / COOLDOWN

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318 psig

749 psig

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BASES:1. FUEL IN REACTOR2. ≤ 24 EFPY3. 3.89x1017 n/cm2 > 1 MEV (Limiting Location, 1/4t)4. BELTLINE LIMITING RTNDT = 102.8°F (1/4t)5. BOTTOM HEAD REGION RTNDT = 40.0°F (1/4t)6. 30 PSI, 10.0°F INSTRUMENT CORRECTION INCLUDED7. REACTOR NOT CRITICAL8. REG. GUIDE 1.99, REV. 2.


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20. BAW-10247PA, Realistic Thermal-Mechanical Fuel RodMethodology for Boiling Water Reactors, Revision 0, April 2008.

21. ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation,Revision 1, March 2014.

c. The core operating limits shall be determined such that all applicablelimits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits,Emergency Core Cooling Systems (ECCS) limits, nuclear limits such asSDM, transient analysis limits, and accident analysis limits) of the safetyanalysis are met.

d. The COLR, including any midcycle revisions or supplements, shall beprovided upon issuance for each reload cycle to the NRC.




Attachment 3

Proposed Technical Specification Bases Changes (Mark-up) Unit 1

(For Information Only)

RCS P/T Limits B 3.4.9

Brunswick Unit 1 B 3.4.9-1 Revision No. 31

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 RCS Pressure and Temperature (P/T) Limits

BASES

BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

This SpecificationThe Reactor Coolant System PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (Ref. 8) contains P/T limit curves for heatup, cooldown, and inservice leakage and hydrostatic testing, and data for the maximum rate of change of reactor coolant temperature. The criticality curve provides limits for both heatup and cooldown during criticality. Development of the curves considered instrument uncertainty values of 10°F for temperature and 15 psig for pressure plus an additional 15 psig for pressure instrument location (Ref. 1).

Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel (including its appurtenances) is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel (including its appurtenances).

10 CFR 50, Appendix G (Ref. 2), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials. Reference 2 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section XI, Appendix G (Ref. 3).

The P/T limit curves in this Specificationthe PTLR were developed in accordance with the 1989 Edition of the ASME Code, Section XI, Appendix G (Ref. 3). These P/T limit curves were developed using the initiation fracture toughness, KIC, for the allowable material fracture toughness. The use of

(continued)



BASES (continued)

APPLICABLE The P/T limits are not derived from Design Basis Accident (DBA) SAFETY ANALYSES analyses. They are prescribed during normal operation to avoid

encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed. Reference The PTLR (Ref. 8) provides the curves and limits in this Specification. Since the P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

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

LCO The elements of this LCO are:

a. RCS pressure and temperature, and heatup or cooldown rates are within the applicable limits specified in Figures 3.4.9-1 and 3.4.9-2, and heatup or cooldown rates are 100F in any 1 hour period, during RCS heatup and cooldownthe PTLR during RCS heatup, cooldown, and inservice leak and hydrostatic testing;

b. RCS pressure and temperature are within the applicable limits in Figures 3.4.9-3, 3.4.9-4, or 3.4.9-5 and heatup or cooldown rates are 30F in any 1 hour period, during RCS inservice leak and hydrostatic testing;

cb. The temperature difference between the reactor vessel bottom head coolant and the RPV coolant is 145F within the limits specified in the PTLR during recirculation pump startup;

dc. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is within the limits specified in the PTLR 50F during recirculation pump startup;

ed. RCS pressure and temperature are within the criticality limits specified in Figure 3.4.9-2the PTLR, prior to achieving criticality; and

fe. The reactor vessel flange and the head flange temperatures are within the limits specified in the PTLR 70F when tensioning the reactor vessel head bolting studs.

(continued)



BASES

LCO These limits define allowable operating regions and permit a large (continued) number of operating cycles while also providing a wide margin to

nonductile failure.

The rate of change of temperature limits control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and inservice leakage and hydrostatic testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves.

Violation of the limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCS components. The consequences depend on several factors, as follows:

a. The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature;

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

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

APPLICABILITY The potential for violating a P/T limit exists at all times. For example, P/T limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed temperature for boltup. Therefore, this LCO is applicable even when fuel is not loaded in the core.

ACTIONS A.1 and A.2

Operation outside the P/T limits in the PTLR, while in MODES 1, 2, and 3, must be corrected so that the RCPB is returned to a condition that has been verified as safe by stress analyses.

The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.

(continued)



BASES

ACTIONS B.1 and B.2 (continued)

Pressure and temperature are reduced by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2

Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified as safe by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored. With the applicable limits of Figure 3.4.9-3, 3.4.9-4, or 3.4.9-5 exceeded during inservice hydrostatic and leak testing operations, the maximum temperature change shall be limited to 10F in any 1 hour period during restoration of the P/T limit parameters to within limits.

Besides restoring the P/T limit parameters to within limits, an engineering evaluation is required to determine if RCS operation is allowed. This engineering evaluation will determine the effect of the P/T limit violation on the fracture toughness properties of the RCS. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 212F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components. ASME Code, Section XI, Appendix E (Ref. 9), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

(continued)



BASES (continued)

SURVEILLANCE SR 3.4.9.1 and SR 3.4.9.2 REQUIREMENTS

Verification that operation is within limits is required when RCS pressure and temperature conditions are undergoing planned changes. The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.

Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.

SR 3.4.9.1 is modified by a Note that requires the Surveillance to be performed only during system heatup and cooldown operations and. SR 3.4.9.2 is modified by a Note that requires the Surveillance to be performed only during inservice leakage and hydrostatic testing.

SR 3.4.9.32

A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.

Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.

SR 3.4.9.4 3 and SR 3.4.9.54

Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.

(continued)



BASES

SURVEILLANCE SR 3.4.9.4 3 and SR 3.4.9.5 4 (continued) REQUIREMENTS

Performing the Surveillance within 30 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.

An acceptable means of demonstrating compliance with the differential temperature requirement of SR 3.4.9.4 3 is to compare the temperature of the reactor coolant in the dome to the bottom head drain temperature.

As specified in procedures, an acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.5 4 is to compare the temperatures of the operating recirculation loop and the idle loop.

SR 3.4.9.4 3 and SR 3.4.9.5 4 are modified by a Note that requires the Surveillance to be met only in MODES 1, 2, 3, and 4. In MODE 5, the overall stress on limiting components is lower. Therefore, ∆T limits are not required. The Note also states the SR is only required to be met during recirculation pump startup, since this is when the stresses occur.

SR 3.4.9.65, SR 3.4.9.76, and SR 3.4.9.87

Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.

The flange temperatures must be verified to be above the limits before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature 80F, checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature 100F, monitoring of the flange temperature is required to ensure the temperature is within the specified limits.

SR 3.4.9.8 7 applies only to the reactor vessel flange and head flange temperatures; TS 3.4.9 does not limit the temperature of any other areas of the vessel when RCS temperature is 100F in MODE 4 or MODE 5, or when defueled.

(continued)



BASES

SURVEILLANCE SR 3.4.9.65, SR 3.4.9.76, and SR 3.4.9.8 7 (continued) REQUIREMENTS

The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.

SR 3.4.9.6 5 is modified by a Note that requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs. SR 3.4.9.7 6 is modified by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature is 80F in MODE 4. SR 3.4.9.8 7 is modified by a Note that requires the Surveillance to be initiated 12 hours after RCS temperature is 100F in MODE 4. The Notes contained in these SRs are necessary to specify when the reactor vessel flange and head flange temperatures are required to be verified to be within the specified limits.

REFERENCES 1. Calculation 0B21-1029, "Instrument Uncertainty for RCS Pressure/Temperature Limits Curve," Revision 0.

2. 10 CFR 50, Appendix G.

3. 1989 Edition of the ASME Code, Section XI, Appendix G.

4. ASME Code Case N-640. "Alternate References Fracture Toughness for Development of P-T Limit Curves Section XI. Division 1."

5. EPRI Report TR-1003346, BWRVIP-86-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan, October 2002.

6. 10 CFR 50, Appendix H.

7. Regulatory Guide 1.99, Revision 2, May 1988.

8. Calculation 0B11-0005, "Development of RPV Pressure-Temperature Curves For BNP Units 1 and 2 For Up To 32 EFPY of Plant Operation," Revision 1.Calculation 0B11-0062, "Pressure and Temperature Limits Report for 54 Effective Full Power Years," (Latest Revision)

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

10. 10 CFR 50.36(c)(2)(ii).


Attachment 4

Revised (Typed) Technical Specification Pages Unit 1

Definitions 1.1



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

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates, for the current reactor vessel

fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 2923 MWt. REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval from SYSTEM (RPS) RESPONSE when the monitored parameter exceeds its RPS trip setpoint TIME at the channel sensor until de-energization of the scram pilot








(continued)





LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits specified in the PTLR.


ACTIONS


TIME

A. --------------NOTE---------------- Required Action A.2 shall be completed if this Condition is entered. --------------------------------------



AND


30 minutes

72 hours


B.1 Be in MODE 3.

AND

B.2 Be in MODE 4.

12 hours

36 hours

(continued)



ACTIONS (continued)


TIME

C. ---------------NOTE--------------- Required Action C.2 shall be completed if this Condition is entered. --------------------------------------



AND


Immediately




SR 3.4.9.1 -------------------------------NOTE-------------------------------- Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing. ------------------------------------------------------------------------



(continued)





SR 3.4.9.2 Verify RCS pressure and RCS temperature are within the criticality limits specified in the PTLR.



Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is within the limits specified in the PTLR.


(continued)






Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature iswithin the limits specified in the PTLR.



Verify reactor vessel flange and head flange temperatures are within the limits specified in the PTLR.








Reactor Steam Dome Pressure 3.4.10



3.4.10 Reactor Steam Dome Pressure

LCO 3.4.10 The reactor steam dome pressure shall be 1045 psig.

APPLICABILITY: MODES 1 and 2.

ACTIONS


TIME

A. Reactor steam dome pressure not within limit.

A.1 Restore reactor steam dome pressure to within limit.

15 minutes

B. Required Action and associated Completion Time not met.

B.1 Be in MODE 3. 12 hours



SR 3.4.10.1 Verify reactor steam dome pressure is 1045 psig. In accordance with the Surveillance Frequency Control Program





20. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, Revision 0, April 2008.

21. ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, March 2014.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.



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

a. RCS pressure and temperature limits for heat up, 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:

1. Limiting Conditions for Operation Section 3.4.9, "RCS Pressure and Temperature (P/T) Limits,"

2. Surveillance Requirement Section 3.4.9, "RCS Pressure and Temperature (P/T) Limits."

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

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

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


Attachment 5

Revised (Typed) Technical Specification Pages Unit 2

Definitions 1.1



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

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates, for the current reactor vessel

fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to (RTP) the reactor coolant of 2923 MWt. REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval from SYSTEM (RPS) RESPONSE when the monitored parameter exceeds its RPS trip setpoint TIME at the channel sensor until de-energization of the scram pilot








(continued)





LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits specified in the PTLR.


ACTIONS


TIME

A. --------------NOTE--------------- Required Action A.2 shall be completed if this Condition is entered. -------------------------------------



AND


30 minutes

72 hours


B.1 Be in MODE 3.

AND

B.2 Be in MODE 4.

12 hours

36 hours

(continued)



ACTIONS (continued)


TIME

C. --------------NOTE--------------- Required Action C.2 shall be completed if this Condition is entered. -------------------------------------



AND


Immediately




SR 3.4.9.1 -------------------------------NOTE-------------------------------- Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing. ------------------------------------------------------------------------



(continued)





SR 3.4.9.2 Verify RCS pressure and RCS temperature are within the criticality limits specified in the PTLR.



Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is within the limits specified in the PTLR.


(continued)






Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is within the limits specified in the PTLR.











Reactor Steam Dome Pressure 3.4.10



3.4.10 Reactor Steam Dome Pressure

LCO 3.4.10 The reactor steam dome pressure shall be 1045 psig.

APPLICABILITY: MODES 1 and 2.

ACTIONS


TIME

A. Reactor steam dome pressure not within limit.

A.1 Restore reactor steam dome pressure to within limit.

15 minutes

B. Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

12 hours



SR 3.4.10.1 Verify reactor steam dome pressure is 1045 psig. In accordance with the Surveillance Frequency Control Program





20. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, Revision 0, April 2008.

21. ANP-10298PA, ACE/ATRIUM 10XM Critical Power Correlation, Revision 0, March 2010.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.



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

a. RCS pressure and temperature limits for heat up, 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:

1. Limiting Conditions for Operation Section 3.4.9, "RCS Pressure and Temperature (P/T) Limits,"

2. Surveillance Requirement Section 3.4.9, "RCS Pressure and Temperature (P/T) Limits."

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

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

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


Attachment 6

Pressure Temperature Limits Report (PTLR)

Facility Code :

Applicable Facilities :

Document Number :

Document Revision Number :

Document EC Number :

Change Reason :

Document Title :

Notes :

1/10/2018ConcurrenceGrzeck, Lee

1/10/2018ConcurrenceWiemann, Richard A

1/10/2018SupervisorBecker, John D.

1/10/2018PreparerBliss, Victor M

Pressure and Temperature Limits Report for 54 Effective

Full-Power Years

Reference Date Correction

001

0B11-0062

BNP

DESIGN ANALYSES AND CALCULATIONS 0B11-0062 Rev. 001

Cover i

Calculation Cover Sheet

Pressure and Temperature Limits Report for 54 Effective Full-Power Years

Calculation Number:

0B11-0062 Rev # 001

System: 1005 DSD List: Yes No

[BNP, HNP, RNP] Sub-Type: RVI Microfiche Attachment List: Yes No

Quality Level Priority E: Yes No

All

BNP Unit ___0__________ CNS Unit _____________ HNP Unit _____________

MNS Unit ______________ ONS Unit _____________ RNP Unit _____________ WLS Unit ______________ LNP Unit _____________ HAR Unit _____________ General Office Keowee Hydro Station

Originated By Design Verification Review By Approved By

Signature Signature Signature

Verification Method 1 2 3 Other

Printed Name Printed Name Printed Name

Victor Bliss John Becker

Date Date Date

See digital signature See digital signature

YES NO Check Box for Multiple Originators or Design Verifiers (see next page)

For Vendor Calculations:

Vendor: Structural Integrity Associates, Inc. Vendor Document #: 1700147.401 RB ISS 01/09/2018

Owners Review By: Victor Bliss Date: See digital signature

Approval By: John Becker Date: See digital signature


LOAP ii

List of Affected Pages

Calculation Number: 0B11-0062

Revision Number: 001

Body of Calculation (including appendices) Supporting Documents Rev. # Pages Revised Pages Deleted Pages Added Rev. # Type Pages Revised Pages Deleted Pages Added

000 53

001 3 (pages 2, 5 & 54)


TOC iii

TABLE OF CONTENTS

1.0 PURPOSE ............................................................................................................ 1 2.0 REFERENCES ..................................................................................................... 1 3.0 BODY OF CALCULATION ................................................................................... 1 4.0 CONCLUSION ..................................................................................................... 1 Attachment 1 (47 pages) ............................................................................................ 2-48 ATTACHED SUPPORTING DOCUMENTS Calculation Cover Sheet ………………………………………………… .............................. i List of Effected Pages ...................................................................................................... ii Table of Contents ............................................................................................................ iii Revision Summary .......................................................................................................... iv Documenting Indexing Table ........................................................................................... v


Revision iv

Revision Summary

Revision Summary

000 New issue of vendor calculation defining Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (EFPY) using the new format for calculations per AD-EG-ALL-1117. Include data and conclusions from 2017 vendor provided calculations as referenced in supporting calculation 0B11-0027 Rev 001. All technical content provided by vendor, Structural Integrity Associates, Inc. This calculation submits the pressure temperature curves for use during the periods of extended operation of Brunswick Nuclear Plant Units 1 and 2 in accordance with the 10 CFR 50.90 License Amendment Process at least one year prior to the expiration of the 32 EFPY P-T Limit Curves that are currently approved in the Technical Specifications.

001 Date Correction: Reference [13] which is discussed in Appendix A of Attachment 1 on page 54 of 0B11-0062 (page 47 of 47 in Attachment 1) has an incorrectly copied date of July 23, 2004. The reference [13] as noted on page 21 of 0B11-0062 (page 14 of 47 in Attachment 1) has the correct date of January 14, 2004. The approving letter from the NRC is dated January 14, 2004. Therefore, the date on page 54 will be changed to January 14, 2004. The Revision Summary and the List of Affected pages are also revised to record the change. Added the vendor calculation number to the cover page.


Indexing v

DOCUMENT INDEXING TABLE

The purpose of this table is to list cross-references..

Document Type (e.g., CALC, DWG,

Procedure, Tag, Software)

Document Number (e.g., Calculation Number, Equipment Tag Number,

Procedure Number, Software Name and

Version)

Function IN for Design Inputs;

OUT for Affected Documents

Relationship to this Calculation

(e.g., Design Input, Assumption Basis,

Reference, Document affected by results)

CALC 0B11-0012, Rev. 1 Input Provides fluence information used in the development of the P-T curves.

CALC 0B21-1029, Rev. 0 Input Provides instrument uncertainty values used in the development of the P-T curves.

CALC 0B11-0005, Rev. 1 Input Methodology documented in 0B11-0005, Rev. 1 was used to develop the P-T curves in this calculation package.

CALC 0B11-0027R001 Input Revision 001 includes additional calculations which developed the final PTLR document


Page 1

1.0 Purpose

The purpose of this calculation package is to document the formulation of Reactor Pressure Vessel (RPV) pressure-temperature curves for the licensing renewal period, i.e. up to 54 effective full power years (EFPY). Curves have been developed for core critical, core not critical, and pressure test conditions. The curves were developed using the same methodology that was used to develop the pressure-temperature curves for extended power uprate (EPU).

This revision submits the pressure temperature curves for use during the periods of extended operation of Brunswick Nuclear Plant Units 1 and 2 in accordance with the 10 CFR 50.90 License Amendment Process at least one year prior to the expiration of the 32 EFPY P-T Limit Curves that are currently approved in the Technical Specifications. 2.0 References

1. Code of Federal Regulations 10 CFR 50, Appendix G.

2. BNP Calculation 0B11-0005, Rev. 1, “Development of RPV Pressure-Temperature Curves For BNP Units 1 & 2 For Up To 32 EFPY of Plant Operation.”

3. BNP Calculation 0B21-1029, Rev. 0, “Instrument Uncertainty for RCS Pressure/Temperature Limits Curve.”

4. BNP Calculation 0B11-0012, Rev. 1, “Neutron Exposure Evaluations for the Core Shroud and Pressure Vessel Brunswick Units 1 and 2.”

5. ASME Boiler and Pressure Vessel Code, Code Case N-640, “Alternative Reference Fracture Toughness for Development of P-T Limits Curves,” Section XI, Division 1, Approved February 26, 1999.

6. USNRC Regulatory Guide 1.99, Revision 2, “Radiation Embrittlement of Reactor Vessel Materials,” U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, (Talk ME 305-4), May 1988.

7. ASME Boiler and Pressure Vessel Code, Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components, Nonmandatory” Appendix G, “Fracture Toughness Criteria for Protection Against Failure,” 1989 Edition, No Addenda.

8. 0B11-0027, Rev. 001, Development of RPV Pressure-Temperature Curves for License Renewal

3.0 Body of Calculation

The development of extended power uprate (EPU) pressure-temperature (PT) curves for the Brunswick reactor pressure vessels are documented in calculation package 0B11-0005, Rev. 1 [2]. These curves are valid for up to 32 effective full power years (EFPY) and were approved by the Nuclear Regulatory Commission (NRC) on June 18, 2003.

This calculation package documents the development of RPV PT curves for up to 54 EFPY of plant operation to address the license renewal period of plant operation. The methodology utilized in the development of the curves in this calculation package is the same as utilized for the EPU pressure-temperature curves (see Ref. 2 for discussion of methodology). The curves were developed by inserting the applicable updated adjusted reference temperature values (ARTNDT) shown in Attachment 1 into the Ref. 2 spreadsheets. Instrument uncertainty values of 10oF and 15 psig [3] (plus an additional 15 psig for static water head) are also included in the development of the curves. The 54 EFPY limiting material in Unit 1 is plate heat number B8496 which is located in the lower intermediate shell. The limiting material in Unit 2 are the N16 nozzles, heat number Q2Q1VW. See Attachment 2 for a comparison of the beltline region and the N16 nozzles.

The curves were developed in accordance with the 1989 ASME Code Section XI, Appendix G [7], and ASME Code Case N-640 [5], which allows the use of KIC for the allowable material fracture toughness.

4.0 Conclusions

This calculation provides pressure-temperature curves for core critical, core not critical, and pressure test conditions for Unit 1 and Unit 2 for up to 54 effective full power years of operation.

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 001 SIA 1700147.401 RB ISSUED 01/09/2018

Page 1 of 47

Duke Energy

Brunswick Steam Electric Plant Units 1 and 2

Pressure and Temperature Limits Report (PTLR)

for 54 Effective Full-Power Years (EFPY)

Revision B ISSUED 01/09/2018

Prepared by: see digital signature Date: see digital signature

Victor Bliss [Program Engineer]

Approved by: see digital signature Date: see digital signature

John Becker [Program Engineering Manager]

Concurred by: see digital signature Date: see digital signature

Rich Wiemann [Director, Engineering]

Concurred by: see digital signature Date: see digital signature

Lee Grzeck [Manager-BNP Regulatory Affairs]


Page 2 of 47

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 6 6.0 References 13

Figure 1 BSEP Unit 1 P-T Curve A (Hydrostatic Pressure and Leak Tests) for 54

EFPY 17

Figure 2 BSEP Unit 1 P-T Curve B (Normal Operation – Core Not Critical) for 54 EFPY

18

Figure 3 BSEP Unit 1 P-T Curve C (Normal Operation – Core Critical) for 54 EFPY 19 Figure 4 BSEP Unit 2 P-T Curve A (Hydrostatic Pressure and Leak Tests) for 54

EFPY 20

Figure 5 BSEP Unit 2 P-T Curve B (Normal Operation – Core Not Critical) for 54 EFPY

21

Figure 6 BSEP Unit 2 P-T Curve C (Normal Operation – Core Critical) for 54 EFPY 22 Figure 7 BSEP Feedwater Nozzle 3-D Finite Element Model [19] 23 Figure 8 BSEP Instrument Nozzle Finite Element Model [20] 24

Table 1 BSEP Unit 1 Pressure Test (Curve A) P-T Curves for 54 EFPY 26 Table 2 BSEP Unit 1 Core Not Critical (Curve B) P-T Curves for 54 EFPY 29 Table 3 BSEP Unit 1 Core Critical (Curve C) P-T Curves for 54 EFPY 32 Table 4 BSEP Unit 2 Pressure Test (Curve A) P-T Curves for 54 EFPY 34 Table 5 BSEP Unit 2 Core Not Critical (Curve B) P-T Curves for 54 EFPY 38 Table 6 BSEP Unit 2 Core Critical (Curve C) P-T Curves for 54 EFPY 41 Table 7 BSEP Unit 1 ART Table for 54 EFPY 44 Table 8 BSEP Unit 2 ART Table for 54 EFPY 45 Table 9 Nozzle Stress Intensity Factors 44

Appendix A Brunswick Reactor Vessel Materials Surveillance Program 47


Page 3 of 47

1.0 Purpose

The purpose of the Brunswick Steam Electric Plant (BSEP) 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].

2.0 Applicability

This report is applicable to the BSEP Unit 1 and Unit 2 RPVs for up to 54 Effective Full-Power

Years (EFPY).

The following BSEP Technical Specifications (TS) are affected by the information contained in

this report:

TS 3.4.9 RCS Pressure/Temperature (P-T) Limits


Page 4 of 47

3.0 Methodology

The limits in this report were derived as follows:

1. The methodology used is in accordance with Reference [1], “Pressure – Temperature

Limits Report Methodology for Boiling Water Reactors,” August 2013, incorporating the

NRC Safety Evaluation in Reference [2].

2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG

1.190) [3] as documented in Reference [4].

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) [5], as

documented in Reference [6].

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

[1], are documented in Reference [7].

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

changes:

• Revision A: 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 [9], 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.


Page 5 of 47

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 54 EFPY for BSEP, as documented in Reference [7],

and are provided in Figure 1 through Figure 3 for BSEP Unit 1 and in Figure 4 through Figure 6

for BSEP Unit 2. A tabulation of the curves is included in Table 1 through Table 3 for BSEP

Unit 1 and in Table 4 through Table 6 for BSEP Unit 2. The adjusted reference temperature

(ART) tables for 54 EFPY for the BSEP Unit 1 and Unit 2 vessel beltline materials are shown in

Table 7 and Table 8, respectively [6].

The resulting P-T curves are based on the geometry, design and materials information for the

BSEP Unit 1 and Unit 2 vessels with the following conditions:

• Heat-up/Cool-down rate limit during Hydrostatic Class 1 Leak Testing (Figures 1 and 4:

Curve A): ≤ 25˚F/hour1 [7].

• Normal Operating Heat-up/Cool-down rate limit (Figures 2 and 5: Curve B – non-

nuclear heating, and Figures 3 and 6: Curve C – nuclear heating): The Single Relief or

Safety Valve (SRV) Blowdown thermal transient event, Event No. 14 from the RPV

thermal cycle diagram (TCD), has a maximum cooldown rate of 954°F/hr and is the

limiting Service Level A/B event used in the calculations of Limit Curve B and Curve C.

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


Page 6 of 47

• RPV bottom head coolant temperature to RPV coolant temperature ∆T limit during

Recirculation Pump startup: ≤ 145°F [1].

• Recirculation loop coolant temperature to RPV coolant temperature ∆T limit during

Recirculation Pump startup: ≤ 50°F [1].

• RPV flange and adjacent shell temperature limit

o BSEP Unit 1: > 70°F (Curve A) or ≥ 76˚F (Curves B and C) [7].

o BSEP Unit 2: ≥ 70˚F [7].

Minimum temperature limits are set in accordance with 10CFR50, Appendix G [8, Table 1]. An

additional 60°F margin above the requirements in Table 1 of 10CFR50, Appendix G, has been

commonly applied in the BWR industry. For the BSEP closure flange material, the minimum

temperature would be 76°F for Unit 1 (i.e. RTNDT,max of 16°F + 60°F) and 70°F for Unit 2 (i.e.

RTNDT,max of 10°F + 60°F) [7]. For Curves A and B, this 60°F margin is a recommendation.

Consequently, for Curves A and B, the minimum temperature for Unit 1 was set to 70°F for

consistency with Unit 2 and with past work. These values are consistent with the minimum

temperature limits and minimum bolt-up temperature in the current docketed P-T curves [10]

(approved by the NRC in Reference [11]). These values also bound the lowest service

temperatures (LST) for ferritic non-RPV components of the reactor coolant pressure boundary

(RCPB), per the component design specifications [12], thereby addressing the NRC condition in

Reference [2, Section 4.0].

5.0 Discussion

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 [5] 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.


Page 7 of 47

The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the

BSEP vessel plate, weld, and forging materials [6]. 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.

The peak RPV ID fluence values of 3.27 x 1018 n/cm2 for Unit 1 and 3.33 x 1018 n/cm2 for Unit 2

at 54 EFPY used in the P-T curve evaluations were obtained from WCAP-17660-NP [4].

Fluence values in Reference [4] were calculated in accordance with RG 1.190 [3]. These fluence

values apply to the limiting beltline lower intermediate shell plates (heat nos. C4487-1 and

B8496-1 for Unit 1; C4489 and C4521 for Unit 2). The fluence values for the limiting beltline

materials in each unit are based upon an attenuation factor of 0.719 for Unit 1 and 0.720 for

Unit 2 for a postulated 1/4T flaw. Consequently, the 1/4T fluence for 54 EFPY for the limiting

lower intermediate shell plates are 2.35 x 1018 n/cm2 for BSEP Unit 1 and 2.40 x 1018 n/cm2 for

BSEP Unit 2. The limiting values for ART for beltline plates and welds are 129.1°F for Unit 1

and 103.4°F for Unit 2 [6].

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. BSEP Unit 1 and

Unit 2 have a set of instrument (N16) nozzles, which are located in the lower intermediate shell

beltline plates [14]. The feedwater (FW) nozzle is considered in the evaluation of the non-

beltline (upper vessel) region P-T limits.

The instrument (N16) nozzle material at BSEP Units 1 and 2 is a ferritic forged nozzle design,

which is welded to the RPV using a full penetration weld rather than the partial penetration

nozzle design used in other plants. The effect of the penetration on the adjacent shell is

considered in the development of bounding beltline P-T limits as described in Reference [7]. The

instrument nozzles have an RPV ID fluence of 1.26 x 1018 n/cm2 for Unit 1 and 1.28 x 1018

n/cm2 for Unit 2 at 54 EFPY, obtained from Reference [4] and calculated in accordance with RG

1.190 [3]. The fluence value for the instrument nozzle location is based upon an attenuation


Page 8 of 47

factor of 0.719 for Unit 1 and 0.720 for Unit 2 for a postulated 1/4T flaw in the instrument

nozzle blend radius. Consequently, the 1/4T fluence for 54 EFPY for the limiting instrument

nozzle location is 9.06 x 1017 n/cm2 for Unit 1 and 9.22 x 1017 n/cm2 for Unit 2. The limiting

value for ART for the instrument nozzles is 131.0°F for Unit 1 and 123.4°F for Unit 2 at 54

EFPY [6]. There are no additional forged or partial penetration nozzles in the extended beltline at

BSEP Units 1 and 2.

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.

The core not critical curve (Curve B) and the core critical curve (Curve C) are prepared

considering a coolant heat-up and cool-down temperature rate corresponding to the limiting

Service Level A/B transient on the RPV thermal cycle diagram, the SRV Blowdown event,

which has a maximum cool-down rate during the transient of 954°F/hr, although for the majority

of the transient, the cool-down rate is 100°F/hr. P-T curves are developed for normal operating

conditions, and Technical Specifications limit operation to ≤100°F/hr. Additionally, for some

BWRs, the SRV blowdown event is explicitly classified in the RPV thermal cycles as an

Emergency condition. However, this was not the case for the SRV blowdown event at

Brunswick. For conservatism, the SRV blowdown event was selected for evaluation as the


Page 9 of 47

limiting Service Level A/B transient. For the hydrostatic pressure and leak test curve (Curve A),

a coolant heat-up and cool-down temperature rate of ≤ 25°F/hr 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.

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 7 and Table 8 for Unit 1 and Unit 2, respectively [6]. Use of initial

RTNDT values in the determination of P-T curves for BSEP was approved by the NRC in

References [15, 16].

Per Reference [6] and in accordance with Appendix A of Reference [1], the BSEP representative

weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel

and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [17]. The representative

heat of the plate material for both BSEP Unit 1 and Unit 2 (B0673-1) in the ISP is not the same

as the target plate material in BSEP Unit 1 (B8496-1) or BSEP Unit 2 (C4500-2), nor does the

representative plate heat exist in the BSEP Unit 1 or Unit 2 beltlines. The representative heat of

the weld material for both BSEP Unit 1 and Unit 2 (5P6756) is not the same heat number as the

target vessel weld in BSEP Unit 1 (1P4218) or BSEP Unit 2 (S3986), nor does the representative

weld heat exist in the BSEP Unit 1 and Unit 2 beltlines. Therefore, for all BSEP Unit 1 and Unit

2 beltline materials, the CF values are calculated using table values from R.G. 1.99, Revision 2,

Position 1.1 [5].

The only computer code used in the determination of the BSEP P-T curves was the ANSYS

finite element computer program:

• ANSYS, Release 14.5. (w/Service Pack 1) [18] for:

o FW nozzle (non-beltline) through-wall thermal and pressure stress distributions in

Reference [19].


Page 10 of 47

o Instrument nozzle thermal and pressure stress distributions in Reference [20].

ANSYS finite element analyses were used to develop the stress distributions through the FW and

instrument nozzles as well as the vessel shell, and these stress distributions were used in the

determination of the stress intensity factors for the FW and instrument nozzles [19, 20] and

vessel shell. At the time that each of the analyses above was performed, the ANSYS program

was controlled under the vendor’s 10 CFR 50 Appendix B [21] Quality Assurance Program for

nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 [22]

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 BSEP 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 Reference [19]. 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 7. Details of the model and material

properties are provided in Reference [19]. Temperature-dependent material properties

were based on the ASME Code, Section II, Part D, 2007 Edition through 2008 Addenda

[23].

• A single model was developed to bound both units. The only difference between the

Unit 1 and Unit 2 FW nozzle geometries is that Unit 1 has a single welded thermal

sleeve, while Unit 2 has a single thermal sleeve interference fit to the FW nozzle safe

end.

• Heat transfer coefficients were calculated in Reference [19] and are a function of FW

temperature and flow rate.

• With respect to operating conditions, the thermal transient which represents the

maximum thermal ramp for the regions corresponding to the FW nozzles during normal


Page 11 of 47

and upset operating conditions were analyzed [19]. The thermal stress distributions,

corresponding to the limiting times presented in Reference [19], 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 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 [19]. The pressure stress distribution was taken

along a linear path through the nozzle corner. 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 BSEP instrument 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

instrument nozzle:

• A one-quarter symmetric, 3-D FEM of the instrument nozzle was constructed and is

shown in Figure 8. Temperature-dependent material properties, taken from the ASME

Code, Section II, Part D, 2007 Edition through 2008 Addenda [23], were used in the

evaluation and are described in Reference [20].

• A single model was developed to bound both units. The instrument nozzle geometry is

identical between Unit 1 and Unit 2.

• With respect to operating conditions, the bounding thermal transient for the region

corresponding to the instrument 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


Page 12 of 47

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.

• Boundary conditions and heat transfer coefficients used for the thermal analysis are

described in Reference [20].

• 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

The thermal stress intensity factor for the RPV shell in the beltline region was calculated from

the stress distribution output of a plant-specific analysis in Reference [20], using the LEFM

solution shown in Equation 2.5.1-6 of Reference [1]. Figure 9 shows the FE model and the path

through the beltline RPV shell used to extract the thermal stress distribution. Detailed

information regarding the analysis can be found in Reference [20].


Page 13 of 47

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. 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).

3. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.190, “Calculational and

Dosimetry Methods for Determining Pressure Vessel Neutron Fluence”, March 2001.

4. Westinghouse Report, WCAP-17660-NP, Revision 0, “Neutron Exposure Evaluations for

Core Shroud and Pressure Vessel Brunswick Units 1 and 2,” November 2012. SI File No.

1501581.201.

5. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, “Radiation

Embrittlement of Reactor Vessel Materials”, May 1988.

6. SI Calculation No. 1501581.301, Revision 0, “Brunswick Nuclear Plant Unit 1 and 2

RPV Beltline ART Evaluation,” November 17, 2017.

7. SI Calculation No. 1700147.302, Revision 0, “Brunswick Nuclear Plant Unit 1 and 2

Updated P-T Curve Calculation for 54 EFPY”, November 17, 2017.

8. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix G, “Fracture Toughness

Requirements,” December 12, 2013.

9. U.S. Code of Federal Regulations, Title 10, Part 50, Section 59, “Changes, tests and

experiments,” August 28, 2007.

10. Attachment 2 to Carolina Power & Light Letter No. BSEP 02-0121, dated June 26, 2002,

SI Calculation No. CPL-54Q-303, Revision 1, “Development of Updated P-T Curves For

32 EFPY,” May 13, 2002 (ADAMS Accession No. ML021890061 and ML021890087).


Page 14 of 47

11. U.S. NRC License Amendments and Safety Evaluation Report, “Brunswick Steam

Electric Plant, Units 1 and 2 – Issuance of Amendment Re: Pressure-Temperature Limit

Curves (TAC Nos. MB5579 and MB5580),” dated June 18, 2003 (ADAMS Accession

No. ML031690683).

12. Piping Design Specifications:

a. United Engineers & Constructors Inc. Specification No. 9527-01-248-3,

“Specification for Reactor Coolant Pressure Boundary Piping for Carolina Power

& Light Company, Brunswick Steam Electric Plant – Units 1 and 2,” dated

December 2, 1974. SI File No. 1700147.201.

b. GE Data Sheet No. 22A1295AJ, “Design Requirements for Pressure Integrity of

Piping and Equipment Pressure Parts – Data Sheet,” Rev. 0. SI File No.

1700147.201.

c. GE System Design Specification No. 22A1295, “Pressure Integrity of Piping and

Equipment Pressure Parts,” Rev. 4. SI File No. CPL-40Q-205.

13. Brunswick Steam Electric Plant Unit 1 License Amendment No. 229 and Unit 2 License

Amendment No. 257, Issuance of Amendments Regarding the Boiling Water Reactor

Vessel and Surveillance Program, dated January 14, 2004 (TAC No. Nos. MC0254 and

MC0255, ADAMS Accession No. ML040150192).

14. Chicago Bridge & Iron Company Drawing, Contract No. 68-2471/72, DWG No. 1,

Revision 8, “General Plan – Nozzles 18’-4 ID x 69’-1-3/4 Ins Heads Nuclear Reactor

Vessel – Gen. Electric Co. for Carolina Power Co. at Southport, North Carolina,”

October 20, 1971, SI File No. 1501581.201.

15. U.S. NRC Letter, “Closeout for Carolina Power & Light Company’s Response to Generic

Letter 92-01, Revision 1, Supplement 1 for the Brunswick Steam Electric Plant, Units 1

and 2 (TAC Nos. M92652 and M92653),” December 23, 1996, SI File No. 1700147.201.

(ADAMS Public Legacy Library Accession No. 9612260127)

16. U.S. NRC Letter, “Closeout of TAC Nos. MA1182 and MA1183 – Response to the

Requests for Additional Information to Generic Letter 92-01, Revision 1, Supplement 1,


Page 15 of 47

‘Reactor Vessel Structural Integrity.’ for Brunswick Steam Electric Plant, Units 1 and 2,”

August 5, 1999, SI File No. 1700147.201. (ADAMS Public Legacy Library Accession

No. 9908120115)

17. 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.

18. ANSYS Mechanical APDL, Release 14.5 (w/Service Pack 1 UP20120918), ANSYS,

Inc., September 2012.

19. SI Calculation No. 1700147.301, Revision 0, “Feedwater Nozzle Fracture Mechanics

Evaluation for Pressure-Temperature Limit Curve Development,” November 17, 2017.

20. SI Calculation No. 1501581.303, Revision 0, “Water Level Instrument Nozzle and Vessel

Fracture Mechanics Evaluation for Pressure-Temperature Limit Curve Development,”

November 17, 2017.

21. U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Appendix B, “Quality

Assurance for Nuclear Power Plants and Fuel Reprocessing Plants”.

22. U.S. Nuclear Regulatory Commission, Generic Letter 83-11, Supplement 1, “License

Qualification for Performing Safety Analyses”, June 24, 1999.

23. ASME Boiler and Pressure Vessel Code, Section II, Part D, Material Properties, 2007

Edition with Addenda through 2008.

24. ASME Boiler and Pressure Vessel Code, Section II, Part D, Material Properties, 2001

Edition with Addenda through 2003.

25. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix H, “Reactor Vessel

Material Surveillance Program Requirements,” January 31, 2008.

26. 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.


Page 16 of 47

27. CP&L Summary Report No. SR-BSEP1-1005-001, “Brunswick Steam Electric Plant

Unit 1 Reactor Pressure Vessel Surveillance Program, First Capsule Removal After 8

Fuel Cycles, Test Results and Projections,” August 1994. SI File No. CPL-35Q-203.

28. Enclosure 1 to CP&L Letter BSEP 97-0051 dated February 25, 1997, “Analysis of the

300 Deg Capsule from the Carolina Power and Light Company, Brunswick Unit 2

Reactor Vessel Radiation Surveillance Program,” Westinghouse Report No. WCAP-

14774, November 1996. (ADAMS Accession No. 9703040242). SI File No.

1501581.202.


Page 17 of 47

Figure 1: BSEP Unit 1 P-T Curve A (Hydrostatic Pressure and Leak Tests) for 54 EFPY

Minimum bolt-up temperature = 70°F


Page 18 of 47

Figure 2: BSEP Unit 1 P-T Curve B (Normal Operation – Core Not Critical) for 54 EFPY


Minimum non-beltline temperature = 103.2°F


Page 19 of 47

Figure 3: BSEP Unit 1 P-T Curve C (Normal Operation – Core Critical) for 54 EFPY

Minimum criticality temperature = 76°F

Minimum beltline temperature = 85.5°F



Page 20 of 47

Figure 4: BSEP Unit 2 P-T Curve A (Hydrostatic Pressure and Leak Tests) for 54 EFPY



Page 21 of 47

Figure 5: BSEP Unit 2 P-T Curve B (Normal Operation – Core Not Critical) for 54 EFPY




Page 22 of 47

Figure 6: BSEP Unit 2 P-T Curve C (Normal Operation – Core Critical) for 54 EFPY

Minimum criticality temperature = 70°F

Minimum bottom head temperature = 97.5°F



Page 23 of 47

Figure 7: BSEP Feedwater Nozzle 3-D Finite Element Model [19]


Page 24 of 47

Figure 8: BSEP Instrument Nozzle Finite Element Model [20]


Page 25 of 47

Figure 9: Vessel Path [20]


Page 26 of 47

Table 1: BSEP Unit 1 Pressure Test (Curve A) P-T Curves for 54 EFPY

Beltline Region

Curve A - Pressure Test P-T Curve

Temperature P-T Curve Pressure

°F psi 70.0 0.0 70.0 414.7

100.2 464.6 118.8 514.6 132.4 564.5 143.1 614.4 151.9 664.3 159.3 714.3 165.8 764.2 171.6 814.1 176.7 864.0 181.4 914.0 185.7 963.9 189.6 1013.8 193.3 1063.7 196.7 1113.7 199.8 1163.6 202.8 1213.5 205.7 1263.4 208.3 1313.4


Page 27 of 47

Table 1: BSEP Unit 1 Pressure Test (Curve A) P-T Curves for 54 EFPY (continued)

Bottom Head Region



°F psi 70.0 0.0 70.0 858.3 74.0 905.2 77.6 952.2 81.1 999.2 84.3 1046.2 87.3 1093.2 90.1 1140.2 92.8 1187.1 95.4 1234.1 97.8 1281.1

100.1 1328.1


Page 28 of 47


Non-Beltline Region



°F psi 70.0 0.0 70.0 312.6

106.0 312.6 106.0 640.3 111.9 688.9 117.2 737.4 122.0 786.0 126.4 834.6 130.4 883.1 134.2 931.7 137.6 980.2 140.9 1028.8 143.9 1077.4 146.8 1125.9 149.5 1174.5 152.1 1223.1 154.5 1271.6 156.9 1320.2


Page 29 of 47

Table 2: BSEP Unit 1 Core Not Critical (Curve B) P-T Curves for 54 EFPY

Beltline Region

Curve B - Core Not Critical P-T Curve


°F psi 70.0 0.0 70.0 23.1 99.7 71.5

118.3 120.0 131.8 168.4 142.4 216.9 151.1 265.3 158.6 313.7 165.0 362.2 170.8 410.6 175.9 459.1 180.6 507.5 184.8 556.0 188.8 604.4 193.5 652.3 197.8 700.3 201.7 748.2 205.4 796.1 208.8 844.0 212.0 892.0 215.0 939.9 217.8 987.8 220.5 1035.8 223.1 1083.7 225.5 1131.6 227.8 1179.6 230.0 1227.5 232.1 1275.4 234.2 1323.3


Page 30 of 47

Table 2: BSEP Unit 1 Core Not Critical (Curve B) P-T Curves for 54 EFPY (continued)

Bottom Head Region Curve B - Core Not Critical

P-T Curve Temperature

P-T Curve Pressure

°F psi 70.0 0.0 70.0 285.6 74.9 334.7 79.3 383.9 83.4 433.0 87.2 482.1 90.7 531.3 94.0 580.4 97.1 629.5

100.0 678.7 102.7 727.8 105.3 776.9 107.8 826.0 110.2 875.2 112.4 924.3 114.6 973.4 116.6 1022.6 118.6 1071.7 120.5 1120.8 122.4 1170.0 124.1 1219.1 125.8 1268.2 127.5 1317.3


Page 31 of 47


Non-Beltline Region



°F psi 103.2 0.0 110.4 42.2 116.7 84.4 122.2 126.6 127.3 168.9 131.8 211.1 136.0 253.3 138.7 282.9 141.3 312.6 141.3 312.6 198.5 1563.0


Page 32 of 47

Table 3: BSEP Unit 1 Core Critical (Curve C) P-T Curves for 54 EFPY

Beltline Region

Curve C - Core Critical P-T Curve


°F psi 85.5 0.0

126.6 46.5 148.8 93.0 164.1 139.5 175.9 186.0 185.4 232.5 193.3 279.0 200.2 325.4 206.2 371.9 211.6 418.4 216.5 464.9 220.9 511.4 225.0 557.9 228.8 604.4 233.5 652.3 237.8 700.3 241.7 748.2 245.4 796.1 248.8 844.0 252.0 892.0 255.0 939.9 257.8 987.8 260.5 1035.8 263.1 1083.7 265.5 1131.6 267.8 1179.6 270.0 1227.5 272.1 1275.4 274.2 1323.3


Page 33 of 47

Table 3: BSEP Unit 1 Core Critical (Curve C) P-T Curves for 54 EFPY (continued)

Bottom Head Region Curve C - Core Critical


P-T Curve Pressure

°F psi 76.0 0.0 76.0 49.7 85.2 98.5 92.9 147.4 99.6 196.2

105.6 245.0 110.8 293.8 115.6 342.6 120.0 391.4 124.0 440.3 127.7 489.1 131.2 537.9 134.4 586.7 137.5 635.5 140.3 684.3 143.0 733.1 145.6 782.0 148.0 830.8 150.4 879.6 152.6 928.4 154.7 977.2 156.8 1026.0 158.7 1074.8 160.6 1123.7 162.5 1172.5 164.2 1221.3 165.9 1270.1 167.5 1318.9


Page 34 of 47


Non-Beltline Region Curve C - Core Critical


P-T Curve Pressure

°F psi 143.2 0.0 150.8 44.7 157.3 89.3 163.1 134.0 168.4 178.6 173.1 223.3 177.4 267.9 181.3 312.6 195.8 312.6 195.8 509.0 198.8 556.9 201.6 604.8 204.2 652.7 206.8 700.6 209.2 748.5 211.5 796.4 213.6 844.4 215.7 892.3 217.8 940.2 219.7 988.1 221.6 1036.0 223.4 1083.9 225.1 1131.8 226.8 1179.7 228.4 1227.6 230.0 1275.5 231.5 1323.5

Table 4: BSEP Unit 2 Pressure Test (Curve A) P-T Curves for 54 EFPY


Page 35 of 47

Beltline Region



°F psi 70.0 0.0 70.0 424.6 96.7 474.1

114.0 523.6 126.8 573.1 137.0 622.6 145.5 672.1 152.7 721.6 159.1 771.1 164.7 820.6 169.7 870.1 174.3 919.6 178.5 969.0 182.4 1018.5 186.0 1068.0 189.3 1117.5 192.5 1167.0 195.4 1216.5 198.2 1266.0 200.8 1315.5


Page 36 of 47


Bottom Head Region



°F psi 70.0 0.0 70.0 604.6 77.1 652.6 83.3 700.5 88.9 748.4 93.9 796.3 98.4 844.2

102.6 892.2 106.4 940.1 110.0 988.0 113.3 1035.9 116.4 1083.8 119.3 1131.7 122.1 1179.7 124.7 1227.6 127.2 1275.5 129.6 1323.4


Page 37 of 47


Non-Beltline Region



°F psi 70.0 0.0 70.0 312.6

100.0 312.6 100.0 596.7 106.6 645.1 112.5 693.4 117.7 741.7 122.4 790.0 126.7 838.3 130.7 886.6 134.4 934.9 137.8 983.2 141.1 1031.6 144.1 1079.9 146.9 1128.2 149.6 1176.5 152.2 1224.8 154.6 1273.1 156.9 1321.4


Page 38 of 47

Table 5: BSEP Unit 2 Core Not Critical (Curve B) P-T Curves for 54 EFPY

Beltline Region



°F psi 70.0 0.0 70.0 63.1 88.9 108.7

102.5 154.4 113.2 200.0 122.1 245.7 129.6 291.3 142.0 340.2 151.9 389.1 160.2 438.0 167.3 486.9 173.6 535.8 179.1 584.8 184.1 633.7 188.6 682.6 192.8 731.5 196.6 780.4 200.2 829.3 203.5 878.2 206.6 927.1 209.6 976.1 212.3 1025.0 215.0 1073.9 217.5 1122.8 219.8 1171.7 222.1 1220.6 224.3 1269.5 226.3 1318.4


Page 39 of 47


Bottom Head Region Curve B - Core Not Critical


P-T Curve Pressure

°F psi 70.0 0.0 70.0 58.4 78.4 106.9 85.6 155.4 91.9 204.0 97.5 252.5

102.5 301.1 107.1 349.6 111.3 398.1 115.2 446.7 118.7 495.2 122.1 543.7 125.2 592.3 128.2 640.8 131.0 689.3 133.6 737.9 136.1 786.4 138.5 835.0 140.8 883.5 143.0 932.0 145.0 980.6 147.0 1029.1 149.0 1077.6 150.8 1126.2 152.6 1174.7 154.3 1223.2 156.0 1271.8 157.6 1320.3


Page 40 of 47


Non-Beltline Region Curve B - Core Not Critical


P-T Curve Pressure

°F psi 103.2 0.0 111.3 48.4 118.4 96.9 124.5 145.3 130.0 193.8 134.1 233.4 137.8 273.0 141.3 312.6 141.3 312.6 198.5 1563.0


Page 41 of 47

Table 6: BSEP Unit 2 Core Critical (Curve C) P-T Curves for 54 EFPY

Beltline Region

Curve C - Core Critical P-T Curve


°F psi 70.0 0.0 70.0 8.3

106.0 55.4 126.7 102.6 141.3 149.8 152.6 196.9 161.8 244.1 169.6 291.3 182.0 340.2 191.9 389.1 200.2 438.0 207.3 486.9 213.6 535.8 219.1 584.8 224.1 633.7 228.6 682.6 232.8 731.5 236.6 780.4 240.2 829.3 243.5 878.2 246.6 927.1 249.6 976.1 252.3 1025.0 255.0 1073.9 257.5 1122.8 259.8 1171.7 262.1 1220.6 264.3 1269.5 266.3 1318.4


Page 42 of 47


Bottom Head Region Curve C - Core Critical


P-T Curve Pressure

°F psi 97.5 0.0

108.2 48.8 116.9 97.7 124.4 146.5 130.9 195.4 136.6 244.2 141.8 293.1 146.4 341.9 150.7 390.8 154.6 439.6 158.3 488.4 161.7 537.3 164.8 586.1 167.8 635.0 170.7 683.8 173.3 732.7 175.9 781.5 178.3 830.3 180.6 879.2 182.8 928.0 184.9 976.9 186.9 1025.7 188.9 1074.6 190.7 1123.4 192.5 1172.3 194.3 1221.1 196.0 1269.9 197.6 1318.8


Page 43 of 47


Non-Beltline Region Curve C - Core Critical


P-T Curve Pressure

°F psi 143.2 0.0 150.8 44.7 157.3 89.3 163.1 134.0 168.4 178.6 173.1 223.3 177.4 267.9 181.3 312.6 188.1 312.6 188.1 397.4 191.6 445.9 194.9 494.5 197.9 543.1 200.8 591.7 203.6 640.2 206.1 688.8 208.6 737.4 211.0 785.9 213.2 834.5 215.3 883.1 217.4 931.6 219.4 980.2 221.3 1028.8 223.1 1077.3 224.9 1125.9 226.6 1174.5 228.2 1223.0 229.8 1271.6 231.4 1320.2


Page 44 of 47

Table 7: BSEP Unit 1 ART Table for 54 EFPY

ID No. Heat No. Lot No. Initial RTNDT (°F)

Chemistry Chemistry Factor

(°F)

Adjustments For 1/4t

Description ΔRTNDT Margin Terms ART

Cu

(wt %) Ni

(wt %) (°F) σΔ (°F)

σi (°F) (°F)

Plat

es

Lower Shell 201 C4535-2 - 34 0.12 0.58 83 45.6 17.0 0.0 113.6 Lower Shell 251 C4550-1 - 10 0.11 0.60 74 40.9 17.0 0.0 84.9

Lower Int. Shell 301 C4487-1 - 10 0.12 0.56 82 50.0 17.0 0.0 94.0 Lower Int. Shell 351 B8496-1 - 10 0.19 0.58 140 85.1 17.0 0.0 129.1 Upper Int. Shell 401 C4510-2 - 22 0.35 0.58 210 26.4 13.2 1.0 74.9 Upper Int. Shell 451 C4515-2 - 10 0.35 0.53 203 25.6 12.8 2.0 61.5

Wel

ds

Lower Int. Vertical F1 F2 S3986 3876 Run 934 10 0.05 0.96 68 35.0 17.5 0.0 80.0

Lower Vertical G1 G2 S3986 3876 Run 934 10 0.05 0.96 68 30.9 15.5 0.0 71.8

Upper Int. to Lower Int. Girth EF S3986 3876

Run 934 10 0.05 0.96 68 8.6 4.3 0.0 27.1

Lower to Lower Int. Girth FG 1P4218 3929

Run 989 -50 0.06 0.87 82 47.0 23.5 0.0 43.9

Noz

zles

Nozzle N16A and N16B - Q2Q1VW 247P-4A

247P-4B 48 0.16 0.82 123 49.0 17.0 0.0 131.0

Noz

zle

Wel

ds

Nozzle N16A and N16B

- 977987 - -50 0.03 1.04 41 16.3 8.1 0.0 -17.4

- 650x006 J807A27A -50 0.03 0.96 41 16.3 8.1 0.0 -17.4

Wall Thickness (in.) Fluence at ID

Attenuation, 1/4t Fluence @ 1/4t Fluence Factor, FF

Location Full 1/4t (n/cm2) e-0.24x (n/cm2) f(0.28-0.10log f)

Plat

es

Lower Shell 201 5.496 1.374 2.59E+18 0.719 1.86E+18 0.553 Lower Shell 251 5.496 1.374 2.59E+18 0.719 1.86E+18 0.553

Lower Int. Shell 301 5.496 1.374 3.27E+18 0.719 2.35E+18 0.609 Lower Int. Shell 351 5.496 1.374 3.27E+18 0.719 2.35E+18 0.609 Upper Int. Shell 401 5.496 1.374 1.71E+17 0.719 1.23E+17 0.126 Upper Int. Shell 451 5.496 1.374 1.71E+17 0.719 1.23E+17 0.126

Wel

ds

Lower Int. Vertical F1F2 5.496 1.374 2.20E+18 0.719 1.58E+18 0.515 Lower Vertical G1 G2 5.496 1.374 1.67E+18 0.719 1.20E+18 0.454

Upper Int. to Lower Int. Girth EF 5.496 1.374 1.71E+17 0.719 1.23E+17 0.126

Lower to Lower Int. Girth FG 5.496 1.374 2.82E+18 0.719 2.03E+18 0.573

Noz

zles

Nozzle N16A and N16B - 5.496 1.374 1.26E+18 0.719 9.06E+17 0.397

Noz

zle

Wel

ds



Page 45 of 47

Table 8: BSEP Unit 2 ART Table for 54 EFPY

ID No. Heat No. Lot No. Initial RTNDT (°F)

Chemistry Chemistry

Factor (°F)

Adjustments For 1/4t

Description ΔRTNDT Margin Terms

ΔRTN

DT

Cu

(wt %) Ni

(wt %) (°F) σΔ (°F)

σi (°F) (°F)

Plat

es

Lower Shell 201 C4500 - 10 0.15 0.54 107 59.4 17.0 0.0 103.4 Lower Shell 251 C4550 - 10 0.11 0.60 74 41.2 17.0 0.0 85.2

Lower Int. Shell 301 C4489 - 10 0.12 0.60 83 50.9 17.0 0.0 94.9 Lower Int. Shell 351 C4521 - 10 0.12 0.57 82 50.6 17.0 0.0 94.6 Upper Int. Shell 401 C4854-2 - 10 0.35 0.56 207 26.4 13.2 0.0 62.8 Upper Int. Shell 451 C4862-2 - 10 0.35 0.58 210 26.7 13.4 0.0 63.5

Wel

ds

Lower Int. Vertical F1 F2 S-3986 3876 Run 934 10 0.05 0.96 68 34.8 17.4 0.0 79.5

Lower Vertical G1 G2 S-3986 3876 Run 934 10 0.05 0.96 68 30.7 15.3 0.0 71.3

Upper Int. to Lower Int. Girth EF S-3986 3876

Run 934 10 0.05 0.96 68 8.7 4.3 0.0 27.3

Lower to Lower Int. Girth FG 3P4000 3932 Run 989 -50 0.02 0.90 27 15.6 7.8 0.0 -18.7

Noz

zles

Nozzle N16A and N16B - Q2Q1VW 247P-3A 247P-3B 40 0.16 0.82 123 49.4 17.0 0.0 123.4

Noz

zle

Wel

ds

Nozzle N16A and N16B - 01R496 -50 0.02 0.92 27 10.8 5.4 0.0 -28.4

- 82D913 -50 0.03 0.80 41 16.4 8.2 0.0 -17.1

Wall Thickness (in.) Fluence at ID

Attenuation, 1/4t Fluence @ 1/4t Fluence Factor, FF

Location Full 1/4t (n/cm2) e-0.24x (n/cm2) f(0.28-0.10log f)

Plat

es

Lower Shell 201 5.466 1.367 2.63E+18 0.720 1.89E+18 0.557 Lower Shell 251 5.466 1.367 2.63E+18 0.720 1.89E+18 0.557

Lower Int. Shell 301 5.466 1.367 3.33E+18 0.720 2.40E+18 0.614 Lower Int. Shell 351 5.466 1.367 3.33E+18 0.720 2.40E+18 0.614 Upper Int. Shell 401 5.466 1.367 1.74E+17 0.720 1.25E+17 0.128 Upper Int. Shell 451 5.466 1.367 1.74E+17 0.720 1.25E+17 0.128

Wel

ds

Lower Int. Vertical F1F2 5.466 1.367 2.16E+18 0.720 1.56E+18 0.511 Lower Vertical G1 G2 5.466 1.367 1.64E+18 0.720 1.18E+18 0.451

Upper Int. to Lower Int. Girth EF 5.466 1.367 1.74E+17 0.720 1.25E+17 0.128

Lower to Lower Int. Girth FG 5.466 1.367 2.89E+18 0.720 2.08E+18 0.579

Noz

zles


Noz

zle

Wel

ds


Table 9: Nozzle Stress Intensity Factors


Page 46 of 47

Nozzle Applied Pressure, KIp-app Thermal, KIt

Feedwater 73.74 66.83 Instrument (N16) 55.42 15.13

KI in units of ksi-in0.5


Appendix A

BRUNSWICK REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM

In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program

Requirements [25], one surveillance capsule has been removed and tested from each of the BSEP Unit 1

and Unit 2 reactor vessels. The first BSEP Unit 1 capsule was removed in Summer 1993, at 8.67 EFPY

[26, 27] and the first BSEP Unit 2 capsule was removed in Spring 1996, at 10.9 EFPY [26, 28]. The

surveillance capsules contained 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 methods and results of testing are presented in References [27, 28]. In BSEP

Units 1 and 2, there are two remaining capsules in each vessel which will remain in place to serve as

backup surveillance material for the BWRVIP program, or as otherwise needed.

BSEP has replaced the original RPV material surveillance program with the BWRVIP ISP [25]. BSEP is

currently committed to use the BWRVIP ISP, and has made a licensing commitment to use the ISP for

BSEP 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 NRC. BSEP 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 January 14, 2004 [13]. Under the ISP, no

further capsules are scheduled for removal from the BSEP Unit 1 and 2 vessels. Representative

surveillance capsule materials for the BSEP Unit 1 and 2 limiting beltline plate are contained in the River

Bend and Supplemental Surveillance Program (SSP) Capsules C, F, and H. Representative materials for

the BSEP Unit 1 and 2 limiting beltline weld are in the Duane Arnold and SSP-F surveillance capsules.

No further SSP capsules are scheduled for withdrawal. The next River Bend surveillance capsule is

scheduled to be withdrawn and tested under the ISP in approximately 2025, and the next Duane Arnold

surveillance capsule is scheduled for withdrawal and testing in approximately 2027 [25].