RA-19-0338, Pressure and Temperature Limits Report for Unit Nos. 1 and 2

Pressure and Temperature Limits Report for Unit Nos. 1 and 2

ML19227A307
Person / Time
Site:	Brunswick
Issue date:	08/15/2019
From:	Yodersmith S Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-19-0338
Download: ML19227A307 (57)

Text

DESIGN ANALYSES AND CALCULATIONS 0B11-0062 Rev. 004 Cover i Calculation Cover Sheet Pressure and Temperature Limits Report for 54 Effective Full

-Power Years Calculation Number: 0B11-00 62 Rev # 004 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 Scott Karriker 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.RC ISSUED 10/28/2018 Owners Review By:

Victor Bliss Date: See digital signature Approval By:

Scott Karriker Date: See digital signature

DESIGN ANALYSES AND CALCULATIONS 0B11-00 62 Rev. 00 4 LOAP ii List of Affected Pages Calculation Number:

0B11-00 62 Revision Number:

00 4 Body of Calculation (including appendices)

Supporting Documents Rev. # Pages Revised Pages Deleted Pages Added Rev. # Type Pages Revised Pages Deleted Pages Added 00 0 53 001 3 (p a g e s 2, 5 & 54) 002 1 2 (pages 1, 4, 13, 15, 18, 19, 21, 22, 31, 34, 40, 43) 003 7 (pages 19 of 47 thru 25 of 47 of Attach. 1) 004 2 (pages 6 and 8 of Attachment

1)

DESIGN ANALYSES AND CALCULATIONS 0B11-00 62 Rev. 00 4 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

DESIGN ANALYSES AND CALCULATIONS 0B11-0062 Rev. 00 4 Revision iv Revision Summary Revision Summary 00 0 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.

002 Curve B and Curve C have been updated due to a fracture mechanics analysis for the feedwater nozzle that incorporates revised plant

-specific heat transfer and fluid temperature inputs to the finite element thermal analysis, NEDC030633R1(Unit 2) and NEDC030634R1 (Unit 1) which revised SIA 1700147.301R1 and 1700147.30 2 R1. 003 Editorial correction on graph titles. Starting on Figure 4, its graph title was left on the previous page during pagination, and the error followed through on the remaining graphs. The title alignment was corrected.

004 Editorial revision Attachment 1. This revision copies a reference comment currently in section 5 (Discussion) regarding Brunswick's operating guideline limits, which were previous noted in Brunswick's Technical Specifications, and adding that reference comment to section 4 (Operating Limits). The addition is in section 4, page 6 of 47 Attachment 1, noted by a revision line. In addition, the copied comment from section 5, page 8 of 47 Attachment 1, has been modified to note its historical reference (i.e. no longer included in the Technical Specifications). The reference comment is "Normal Operational conditions, the RCS heatup and cooldown rates are in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

"

DESIGN ANALYSES AND CALCULATIONS 0B11-00 62 Rev. 00 4 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 0B11-0062 R000.

DESIGN ANALYSES AND CALCULATIONS 0B11-00 62 Rev. 004 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

9. SIA 1700147.301.R1, Feedwater Nozzle Fracture Mechanics Evaluation for Pressure

-Temperature Limit Curve Development.

10. SIA 1700147.302.R2, Brunswick Nuclear Plant Unit 1 and 2 Updated P

-T Curve Calculation for 54 EFPY 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 (ART NDT) shown in Attachment 1 into the Ref. 2 spreadsheets. Instrument uncertainty values of 10 oF 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 K IC 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 004 SIA 1700147.401 RC ISSUED 10/28/2018 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 C 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]

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 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 35 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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 3 of 47 1.0 Purpose The purpose of the Brunswick Steam Electric Plant (B SE P) 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 bolt

-up 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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 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. Revision B: Reference date correction.

Revision C: Curve B and Curve C have been updated due to a fracture mechanics analysis for the feedwater nozzle that incorporates revised plant-specific heat transfer and fluid temperature inputs to the finite element thermal analysis.

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.

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 5 of 47 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.

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): 1 [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 Sa fety Valve (SRV) Blowdown th er mal tr an sient event, Event No. 14 from the RPV

1 Interpreted as the temperature change in any 1

-hour period is less than or equal to 25°F.

Page 6 of 47 Brunswick Steam Electric Plant Units 1 and 2 PTLR , 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 thermal cycle diagram (TCD), has a maximum cooldown rate of and is t he li mi ting S ervi ce Level A/B ev ent u sed in the ca l culations of Li mit Curve B and Curve C. However, for Normal Operational conditions, t he RCS heatup and cooldown rates are in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period. RPV bottom head coolant temperature to RPV coolant temperature T limit duringRecirculation Pump startup: 145 [1].Recirculation loop coolant temperature to RPV coolant temperature T limit duringRecirculation Pump startup: 50 [1].RPV flange and adjacent shell temperature limit oBSEP Unit 1: 76 [7].oBSEP Unit 2: 70 [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.

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 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 forgin gs.

The peak RPV ID fluence values of 3.27 x 1018 n/cm 2 for Unit 1 and 3.33 x 1018 n/cm 2 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/cm 2 for BSEP Unit 1 and 2.40 x 1018 n/cm 2 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/cm 2 for Unit 1 and 1.28 x 1018 n/cm 2 for Unit 2 at 54 EFPY , obtained from Reference [

4] and calculated in accordance with RG Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 8 of 47 1.190 [3]. The fluence value for the instrument nozzle location is based upon an attenuation 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/cm 2 for Unit 1 and 9.22 x 1017 n/cm 2 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 historically, Technical Specifications limit ed 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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 9 of 47 blowdown event at Brunswick. For conservatism, the SRV blowdown event was selected for evaluation as the 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 RT NDT, 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/cm 2 for E > 1 MeV) are shown in Table 7 and Table 8 for Unit 1 and Unit 2, respectively

[6]. Use of initial RT NDT values in the determination of P-T curves for BSEP was approved by the NRC in Reference s [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:

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 10 of 47 o FW nozzle (non-beltline) through-wall thermal and pressure stress distribution s in Reference [19]. 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.

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 11 of 47 With respect to operating conditions, the thermal transient which represents th e maximum thermal ramp for the regions corresponding to the FW nozzles during normal 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, K It , 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. Th e 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 K Ip 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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 12 of 47 was analyzed [

20]. The thermal stress distribution, corresponding to the limiting time in Reference [

20], along a linear path through the nozzle corner is used. The BIE/IF methodology presented in Reference [

1] was used to calculate the thermal stress intensity factor, KIt, due to the thermal stresses by fitting a third order polynomial equation to the path stress distribution for the thermal load case. 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].

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 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. S I 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 1, "Brunswick Nuclear Plant Unit 1 and 2 Updated P-T Curve Calculation for 54 EFPY", May 29, 2018. 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).

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 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, Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 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 . 18. ANSYS Mechanical APDL, Release 14.5 (w/Service Pack 1 UP20120918

), ANSYS, Inc., September 2012.

19. SI Calculation No. 1700147.301, Revision 1, "Feedwater Nozzle Fracture Mechanics Evaluation for Pressure-Temperature Limit Curve Development," May 29, 2018.

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 Requirement s," 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.

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 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.

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 17 of 47 1: Unit 1 P-54 Minimum bolt

-up temperature = 70°F

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 18 of 47 2: Unit 1 P-- 54 Minimum bolt

-up temperature = 70°F Minimum non-beltline temperature = 76.2

°F Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 19 of 47 3: Unit 1 P-- 54 Minimum criticality temperature = 7 6°F Minimum beltline temperature = 85.5

°F Minimum non-beltline temperature = 116.2°F Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 20 of 47 4: Unit 2 P-54 Minimum bolt

-up temperature = 70°F

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 21 of 47 5: Unit 2 P- - 54 Minimum bolt

-up temperature = 70°F Minimum non

-beltline temperature = 76.2

°F Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 22 of 47 6: Unit 2 P-- 54 Minimum criticality temperature = 7 0°F Minimum bottom head temperature = 97.5°F Minimum non

-beltline temperature = 1 16.2°F Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 23 of 47 7: 3-D [19]

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 24 of 47 8: Instrument 20]

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 25 of 47 9: Vessel Path

[20]

Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 26 of 47 Table 1: Unit 1 - for 54 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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 27 of 47 Table 1: Unit 1 Pressure -54 Curve A - Pressure Test P-T Curve Temperature P-T Curve Pressure °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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 28 of 47 Table 1: Unit 1 Pressure -54 Non- Region Curve A - Pressure Test P-T Curve Temperature P-T Curve Pressure °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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 29 of 47 Table 2: Unit 1 Core Not Critical

- for 54 Curve B - Core Not Critical P-T Curve Temperature P-T Curve Pressure °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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 30 of 47 Table 2: Unit 1 -54 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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 31 of 47 Table 2: Unit 1 -54 Non- Region Curve B - Core Not Critic al P-T Curve Temperature P-T Curve Pressure °F psi 76.2 0.0 87.1 38.7 97.5 84.4 106.1 130.0 113.4 175.7 119.8 221.3 125.5 267.0 130.5 312.6 136.0 312.6 136.0 367.1 140.5 416.9 144.7 466.7 148.5 516.6 152.1 566.4 155.4 616.2 158.5 666.0 161.4 715.9 164.2 765.7 166.8 815.5 169.3 865.4 171.7 915.2 174.0 965.0 176.1 1014.9 178.2 1064.7 180.2 1114.5 182.1 1164.4 184.0 1214.2 185.7 1264.0 187.5 1313.8 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 32 of 47 Table 3: Unit 1 Core -es for 54 Curve C - Core Critical P-T Curve Temperature P-T Curve Pressure °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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 33 of 47 Table 3: Unit 1 -54 Curve C - Core Critical P-T Curve Temperature 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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 34 of 47 Table 3: Unit 1 -54 Non- Region Curve C - Core Critical P-T Curve Temperature P-T Curve Pressure °F psi 116.2 0.0 121.2 16.6 133.5 65.9 143.5 115.3 151.7 164.6 158.8 213.9 165.0 263.3 170.5 312.6 195.8 312.6 195.8 622.8 198.9 672.3 201.8 721.7 204.5 771.2 207.1 820.7 209.5 870.2 211.9 919.7 214.1 969.2 216.3 1018.7 218.3 1068.1 220.3 1117.6 222.2 1167.1 224.0 1216.6 225.8 1266.1 227.5 1315.6 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 35 of 47 Table 4: Unit 2 Pressure -54 Curve A - Pressure Test P-T Curve Temperature P-T Curve Pressure °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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 36 of 47 Table 4: Unit 2 Pressure -54 Curve A - Pressure Test P-T Curve Temperature P-T Curve Pressure °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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 37 of 47 Table 4: Unit 2 Pressure -54 Non- Region Curve A - Pressure Test P-T Curve Temperature P-T Curve Pressure °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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 38 of 47 Table 5: Unit 2 Core Not Critic-54 Curve B - Core Not Critical P-T Curve Temperature P-T Curve Pressure °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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 39 of 47 Table 5: Unit 2 -54 Curve B - Core Not Critical P-T Curve Temperature 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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 40 of 47 Table 5: Unit 2 Core -54 Non- Region Curve B - Core Not Critical P-T Curve Temperature P-T Curve Pressure °F psi 76.2 0.0 86.9 38.0 97.2 82.9 105.7 127.9 113.0 172.8 119.3 217.7 124.9 262.6 130.0 307.6 130.5 312.6 130.5 312.6 134.0 346.4 138.0 388.5 142.0 434.2 146.0 483.6 150.0 537.1 154.0 595.1 158.0 657.9 162.0 726.0 166.0 799.7 170.0 879.6 174.0 966.1 178.0 1059.8 182.0 1161.4 186.0 1271.4 190.0 1390.5 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 41 of 47 Table 6: Unit 2 -54 Curve C - Core Critical P-T Curve Temperature P-T Curve Pressure °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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 42 of 47 Table 6: Unit 2 -T 54 Curve C - Core Critical P-T Curve Temperature 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 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 43 of 47 Table 6: Unit 2 -54 Non- Region Curve C - Core Critical P-T Curve Temperature P-T Curve Pressure °F psi 116.2 0.0 128.3 43.7 138.3 88.5 146.6 133.3 153.8 178.1 160.0 223.0 165.5 267.8 170.5 312.6 188.1 312.6 188.1 511.2 191.5 559.0 194.8 606.8 197.8 654.6 200.6 702.4 203.3 750.2 205.9 798.0 208.3 845.8 210.7 893.6 212.9 941.5 215.0 989.3 217.1 1037.1 219.0 1084.9 220.9 1132.7 222.7 1180.5 224.5 1228.3 226.2 1276.1 227.8 1323.9 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 44 of 47 Table 7: Unit 1 ID No. Lot No. Initial RTNDT Chemistry Chemistry Description RT NDT Margin Terms ART Cu Ni i Plates Lower Shell 201 C4535 34 0.12 0.58 83 45.6 17.0 0.0 113.6 Lower Shell 251 C4550 10 0.11 0.60 74 40.9 17.0 0.0 84.9 Lower Int. Shell 301 C4487 10 0.12 0.56 82 50.0 17.0 0.0 94.0 Lower Int. Shell 351 B8496 10 0.19 0.58 140 85.1 17.0 0.0 129.1 Upper Int. Shell 401 C4510 22 0.35 0.58 210 26.4 13.2 1.0 74.9 Upper Int. Shell 451 C4515 10 0.35 0.53 203 25.6 12.8 2.0 61.5 Welds 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 Nozzle N16A and N16B - Q2Q1VW 247P-4A 247P-4B 48 0.16 0.82 123 49.0 17.0 0.0 131.0 Welds 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 ID Attenuation, Location 2 e-0.24x 2 f- Plates 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 Welds 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 Nozzle N16A and N16B - 5.496 1.374 1.26E+18 0.719 9.06E+17 0.397 Welds Nozzle N16A and N16B - 5.496 1.374 1.26E+18 0.719 9.06E+17 0.397 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 45 of 47 Table 8: Unit ID No. Lot No. Initial RTNDT Chemistry Chemistry Description RT NDT Margin Terms RT N DT Cu Ni i Plates 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 8 3 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 10 0.35 0.56 207 26.4 13.2 0.0 62.8 Upper Int. Shell 451 C4862 10 0.35 0.58 210 26.7 13.4 0.0 63.5 Welds 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 Nozzle N16A and N16B

- Q2Q1VW 247P-3A 247P-3B 40 0.16 0.82 123 49.4 17.0 0.0 123.4 Welds 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 ID Attenuation, Location 2 e-0.24x 2 f- Plates 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 Welds 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 Nozzle N16A and N16B

- 5.466 1.367 1.28E+18 0.720 9.22E+17 0.401 Welds Nozzle N16A and N16B

- 5.466 1.367 1.28E+18 0.720 9.22E+17 0.401 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 46 of 47 Table 9: Applied Pressure, K Ip-app Thermal, K It Feedwater 73.74 50.05 Instrument (N16) 55.42 15.1 3 K I in units of ksi

-in0.5 Brunswick Steam Electric Plant Units 1 and 2 PTLR Attachment 1, 0B11-0062, Rev 004 SIA 1700147.401 RC ISSUED 10/28/2018 Page 47 of 47 Appendix A 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 weld 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 plate 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].