CNL-14-065, Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-491)

Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-491)

ML14175A307
Person / Time
Site:	Browns Ferry
Issue date:	06/19/2014
From:	James Shea Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML14175A0306	List: CNL-14-065, Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-491)
References
BFN TS-491, CNL-14-065, L44 140619 001
Download: ML14175A307 (105)

Text

Proprietary Information Withhold Under 10 CFR 2.390(d)(1)

This letter is decontrolled when separated from Enclosure 2 L44 140619 001 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-14-065 June 19, 2014 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Unit 2 Renewed Facility Operating License No. DPR-52 NRC Docket No. 50-260

Subject:

Browns Ferry Nuclear Plant (BFN), Unit 2 - Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T)

Limits (BFN TS-491)

In accordance with the provisions of Title 10 of the Code of Federal Regulations (CFR) 50.90, Application for amendment of license, construction permit, or early site permit, the Tennessee Valley Authority (TVA) is submitting a license amendment request to revise the Browns Ferry Nuclear Plant, Unit 2, Technical Specifications (TS) for Limiting Condition for Operation (LCO) 3.4.9, RCS Pressure and Temperature (P/T) Limits. This submittal satisfies the commitment to prepare and submit revised BFN, Unit 2, P/T limits prior to the start of the period of extended operation as discussed in Section 4.2.5 provided in Browns Ferry Nuclear Plant (BFN) - Units 1, 2 and 3 - Application for Renewed Operating Licenses, dated December 31, 2003 (ADAMS Accession No. ML040060359).

This submittal satisfies the requirements of NUREG-1843, Safety Evaluation Report Related to the License Renewal of the Browns Ferry Nuclear Plant, Units 1, 2, and 3, dated April 2006 (ADAMs Accession No. ML061030032), commitment 39 that required the development and submittal of revised P/T limit curves for NRC approval prior to the period of extended operation. In the Browns Ferry Nuclear Plant - NRC Post-Approval Site Inspection for License Renewal, Inspection Report, dated October 3, 2013 (Inspection Report 005000259/2013009, 005000260/2013009, 005000296/2013009), with respect to commitment 39, it was stated that new P/T limits will be calculated and approved before the period of extended operation. The current NRC approved P/T limit curves were approved prior to the period of extended operation and are applicable for operation into, but not to the end of the period of extended operation. Therefore, the proposed P/T limit curves were developed based on analyses projected to the end of the period of extended operation as required by 10 CFR 54.21(c)(1)(ii). The proposed P/T limit curves are expected to be approved and implemented before the current P/T limit curves expire.

U.S. Nuclear Regulatory Commission Page 2 June 19, 2014 to this letter provides a description, technical evaluation, regulatory evaluation and environmental consideration of the proposed changes. Attachments 1 and 2 to provides a markup of the proposed changes to the BFN , Unit 2, TS and TS Bases (information only), respectively. Attachments 3 and 4 to Enclosure 1 provide retyped versions of the BFN, Unit 2, TS and TS Bases (information only), respectively, to show the incorporation of the proposed changes. Attachment 5 to Enclosure 1 provides a conforming markup of the BFN Updated Final Safety Analysis Report (UFSAR) that results from this proposed change. to this letter contains information that General Electric- Hitachi (GEH) and Electric Power Research Institute (EPRI) consider to be proprietary in nature and subsequently, pursuant to 10 CFR 2.390, "Public inspections, exceptions, request for withholding ," paragraph (a)(4) , it is requested that such information be withheld from public disclosure. Enclosure 4 provides the affidavits supporting the request. Enclosure 3 contains the redacted version of the proprietary attachment with the proprietary material removed ,

which is suitable for public disclosure.

TVA requests that the NRC approve this amendment by June 19, 2015, with implementation within 60 days of issuance.

TVA has determined that there are no significant hazards considerations associated with the proposed changes and that the changes qualify for a categorical exclusion from environmenfal review pursuant to the provisions of 10 CFR 51.22(c)(9).

The BFN Plant Operations Review Committee and the TVA Nuclear Safety Review Board have reviewed the proposed changes and determined that operation of BFN in accordance with the proposed changes will not endanger the health and safety of the public.

Additionally , in accordance with 10 CFR 50.91 (b)(1 ), TVA is sending a copy of this letter and the non-proprietary enclosures to the Alabama State Department of Public Health.

There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Edward D. Schrull at 423-751-3850.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 19 day of June 2014.

lly, he a President, Nuclear Licensing

1. Evaluation of Proposed Changes

2. NEDC-33854P- Pressure and Temperature Limits Report (PTLR) Up to 38 and 48 Effective Full Power Years (Proprietary)

3. NED0-33854- Pressure and Temperature Limits Report (PTLR) Up to 38 and 48 Effective Full Power Years (Non-Proprietary)

4. Affidavits cc: See Page 3

U.S. Nuclear Regulatory Commission Page 3 June 19, 2014 cc (Enclosures):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant State Health Officer, Alabama State Department of Public Health

ENCLOSURE 1 EVALUATION OF PROPOSED CHANGES TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNIT 2 BFN TS-491, Browns Ferry Nuclear Plant (BFN), Unit 2, Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes 2.2 Need for Proposed Changes

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

ATTACHMENTS

1. Proposed Technical Specifications Pages Markups

2. Proposed Technical Specifications Bases Pages Markups (for information only)

3. Proposed Retyped Technical Specifications Pages

4. Proposed Retyped Technical Specifications Bases Pages (for Information only)

5. Proposed Updated Final Safety Analysis Report Page Markups E1-1

ENCLOSURE 1 EVALUATION OF PROPOSED CHANGES 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend the Browns Ferry Nuclear Plant (BFN), Unit 2, Renewed Facility Operating License No. DPR-52 (Reference 1). The proposed changes modify the BFN, Unit 2, Technical Specification (TS) requirements related to the Reactor Coolant System (RCS) Pressure and Temperature (P/T) Limits in TS 3.4.9, RCS Pressure and Temperature (P/T) Limits.

Specifically, the proposed change replaces the current sets of TS Figures 3.4.9-1 and 3.4.9-2. The figures proposed to be replaced consist of two sets of P/T limit curves, one set valid up to and including 23 effective full power years (EFPY) of operation and another set valid from 23 EFPY to 30 EFPY of operation. The proposed change replaces the current curves with a set of figures valid for operation up to and including 38 EFPY and another set valid for operation from >38 EFPY to 48 EFPY. This proposed TS change is necessary to satisfy the commitment to develop and submit to the NRC P/T limits that have been projected to the end of the Unit 2 period of extended operation, prior to the start of the period of extended operation.

2.0 DETAILED DESCRIPTION 2.1 Proposed Changes The Tennessee Valley Authority (TVA) proposes to delete and replace TS Figures 3.4.9-1 and 3.4.9-2 that are applicable for operations up to and including 23 EFPY and for operations

> 23 EFPY and 30 EFPY, respectively, with new TS Figures 3.4.9-1 and 3.4.9-2 that are applicable for operations up to and including 38 EFPY and for operations > 38 EFPY to 48 EFPY, respectively.

The TVA proposes to revise Note 1 of TS SR 3.4.9.1 to change the vessel pressure from

>312 psig to >313 psig to conform to the modified P/T limit curves.

In addition, an associated note for each figure is changed to reflect the new operational applicability limit with respect to EFPY.

2.2 Need for Proposed Changes The License Renewal Application for BFN Units 1, 2, and 3 (Reference 2) states in Appendix A - UFSAR Supplement, Section A.3.1.5, Because of the relationship between the operating pressure-temperature limits and the fracture toughness transition of the reactor vessel, all three units will require new operating pressure-temperature limit curves to be calculated and approved for the extended period of operation.

License Condition 2.E in the BFN, Unit 2, renewed operating license (Reference 1) states, The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than June 28, 2014, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The current Unit 2 P/T limit curves were approved by the NRC through the license amendment process (Reference 3) and are valid to 30 EFPY, well into the period of extended operation.

The proposed P/T limit curves that reflect the analyses projected to the end of the period of extended operation required by 10 CFR 54.21(c)(1)(ii) use conservative fluence values calculated for a period of 60 years of operation. The current approved Unit 2 P/T limit curves E1-2

ENCLOSURE 1 EVALUATION OF PROPOSED CHANGES are valid up to and including 30 EFPY, which covers entry into the period of extended operation. The proposed Unit 2 P/T limit curves are necessary to replace the current curves and provide operational coverage beyond the current approved limit of 30 EFPY to 48 EFPY.

3.0 TECHNICAL EVALUATION

10 CFR 50 Appendix G, Fracture Toughness Requirements, requires the establishment of P/T limits for reactor coolant pressure boundary materials. Appendix G also requires an adequate margin to brittle failure be maintained during normal operation, anticipated operational occurrences, and system hydrostatic tests. The P/T limits are acceptance limits in themselves, because operation in accordance with these limitations precludes operation in an unanalyzed condition. The P/T limits are not derived from Design Basis Accident analyses.

The proposed P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and thus, the curves are composites of the most restrictive regions.

These proposed P/T limit curves are primarily dependent upon the fracture toughness of the vessel ferritic materials. The key parameters which characterize a material's fracture toughness are the reference temperature of nil-ductility transition (RTNDT) and the Upper Shelf Energy (USE). These parameters are defined in 10 CFR 50, Appendix G, and in Appendix G of the ASME Boiler and Pressure Vessel Code,Section XI. These documents also contain the requirements used to establish the P/T operating limits to avoid brittle fracture. The method used to account for irradiation embrittlement is described in Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Rev. 2.

Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," provides an acceptable method for calculating P/T limits that satisfies the requirements of 10 CFR 50 Appendix G.

The P/T limit curves for BFN, Unit 2, have been revised based on methodologies consistent with this regulatory guide using plant-specific material and fluence information. The proposed P/T limit curves reflect changes from those currently licensed. The new P/T limit curves incorporate a revised fluence calculated in accordance with NRC approved GE Licensing Topical Report NEDC-32983P-A, (Reference 4) representing BFN, Unit 2, operating conditions of up to 3952 MWt and incorporate the NRC approved methodologies described in GEH Topical Report NEDC-33178P-A (Reference 5). The operating condition of up to 3952 MWt represents the Extended Power Uprate (EPU) power level; however, the current license power level is 3458 MWt. In addition, the latest information from the BWRVIP Integrated Surveillance Program (ISP) applicable to BFN, Unit 2, has been incorporated.

Five separate curves are depicted in the revised TS Figure 3.4.9-1 and Figure 3.4.9-2 to clearly show the P/T limitations for the reactor bottom head area and the vessel beltline and upper vessel areas for all operating conditions.

Curve 1 of Figure 3.4.9-1 specifies minimum temperature for bottom head during mechanical heatup or cooldown following nuclear shutdown (i.e., reactor not critical).

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ENCLOSURE 1 EVALUATION OF PROPOSED CHANGES Curve 2 of Figure 3.4.9-1 specifies minimum temperature for upper RPV and beltline during mechanical heatup or cooldown following nuclear shutdown (i.e., reactor not critical).

Curve 3 of Figure 3.4.9-1 specifies minimum temperature for core operation (i.e., reactor critical).

Curve 1 of Figure 3.4.9-2 specifies minimum temperature for bottom head during in-service leak or hydrostatic testing.

Curve 2 of Figure 3.4.9-2 specifies minimum temperature for upper RPV and beltline during in-service leak or hydrostatic testing.

The 1998 Edition of the ASME Section XI Boiler and Pressure Vessel Code including 2000 Addenda was used for the evaluation to generate the P/T curves.

Values of initial RTNDT for the vessel materials were calculated in accordance with the methods described in GEH Topical Report NEDC-33178P-A. The values used to determine the initial RTNDT were obtained from the Certified Material Test Reports (CMTRs) for BFN, Unit 2. Initial RTNDT values for the BFN, Unit 2, weld materials were not calculated; these values were obtained from previous reports (References 6 and 7) and verified for input for this evaluation.

The adjusted reference temperature (ART) for the beltline region was determined using the methods described in Regulatory Guide 1.99, Rev. 2. These values are summarized in Tables B-4 and B-5 of Enclosures 2 and 3.

The limiting ART of 175°F for the 48 EFPY calculation remains below the 200°F criterion of Regulatory Guide 1.99, Rev. 2. The Upper Shelf Energy (USE) equivalent margin analyses values calculated for end of life (i.e., 48 EFPY) remain within the limits of Regulatory Guide 1.99, Rev. 2 and 10 CFR 50 Appendix G. A single set of P/T curves for the heatup and cooldown operating condition at a given EFPY that apply for both the 1/4T and 3/4T locations was developed. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (assumed inside surface flaw) and the 3/4T location (assumed outside surface flaw) because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4T location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness, KIR, at 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, because BFN, Unit 2, is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits.

The GEH Pressure-Temperature Limits Report for BFN, Unit 2, provided as Enclosures 2 (proprietary) and 3 (non-proprietary), demonstrates the technical methods and contains the data for producing the composite P/T curves which are proposed to be placed in the TS.

Table 1 in the GEH Report for BFN, Unit 2, contains the data for producing the composite 38 EFPY curves. In the same manner, Table 2 of the GEH Report contains the data for producing the composite 48 EFPY curves.

The proposed P/T curves have been developed utilizing the methodology of Regulatory Guide 1.190 and ASME Section XI. The regulatory guidance provides an allowance for margin to be included in the bounding values of the ART. Use of this methodology ensures E1-4

ENCLOSURE 1 EVALUATION OF PROPOSED CHANGES adequate safety margins are maintained. In addition, the analysis conforms to the requirements of 10 CFR 50, Appendix G, which ensures the most limiting material is considered in the development of the P/T curves. The vessel is in compliance with the regulatory requirements, adequate safety margins are maintained, and, therefore, BFN, Unit 2 operation to 48 EFPY, representing a 60 year license, will not have an adverse effect on reactor vessel fracture toughness.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria Pursuant to 10 CFR 50.90, TVA is submitting a request for a Technical Specifications (TS) change to renewed license DPR-52 for BFN, Unit 2. The proposed change revises the reactor vessel pressure-temperature (P/T) limits depicted in current TS Figure 3.4.9-1 and Figure 3.4.9-2.

10 CFR 50, Appendix G, contains the requirements for the P/T limit curves, and requires that P/T curves for the reactor pressure vessel be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code.

The regulatory requirements for fluence calculations are contained in General Design Criteria (GDC) 30 and 31. NRC issued Regulatory Guide 1.190 in March 2001, which provided state-of-the-art calculations and measurement procedures that are acceptable to the NRC staff for determining pressure vessel fluence. NRC has approved vessel fluence calculation methodologies that satisfy the requirements of GDCs 30 and 31 performed with approved methodologies or with methods that are shown to adhere to the guidance in Regulatory Guide 1.190. The analyses supporting this submittal were performed in accordance with Regulatory Guide 1.190 guidance.

4.2 Precedent The NRC has previously approved a License Amendment Request (LAR) that included two sets of P/T limit curves for BFN Units 2 and 3 (Amendments 288 and 247), one set for operation up to and including 23 EFPY (Unit 2) and up to and including 20 EFPY (Unit 3),

and a second set for operations up to and including 30 EFPY (Unit 2) and up to and including 28 EFPY (Unit 3) (Reference 3). TVA proposed to the NRC in the February 24, 2004, LAR letter (Reference 8) that both sets of curves be placed in the TS upon approval of the amendments. In the NRC Safety Evaluation for BFN Units 2 and 3, Amendments 288 and 247, dated March 10, 2004 (Reference 3), the NRC stated that because the applicability of each set of curves was clearly defined, the approach proposed by TVA was acceptable.

The NRC has previously approved an LAR that revised facility P/T limit curves based on the application of methodology in GE Hitachi Nuclear Energy (GEH) Licensing Topical Reports NEDC-33178P-A (Reference 5) and NEDC-32983P-A (Reference 4) for the Peach Bottom Atomic Power Station, Units 2 and 3, Amendment Nos. 286 and 289, issued by NRC letter dated April 1, 2013 (Reference 9). The Peach Bottom Atomic Power Station LAR used the 1998 edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda in its evaluation and applied calculated fluence values based on uprated conditions that conservatively bounded future operations.

E1-5

ENCLOSURE 1 EVALUATION OF PROPOSED CHANGES 4.3 No Significant Hazards Consideration The proposed change modifies the Browns Ferry Nuclear Plant (BFN), Unit 2, Technical Specification (TS) requirements related to TS 3.4.9, RCS Pressure and Temperature (P/T)

Limits.

Tennessee Valley Authority (TVA) has concluded that the changes to BFN, Unit 2, TS 3.4.9 do not involve a significant hazards consideration. TVA's conclusion is based on its evaluation in accordance with 10 CFR 50.91(a)(1) of the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

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

Response: No The proposed changes are to accepted operating parameters that have been approved in previous license amendments. The changes to P/T curves were developed based on NRC approved methodologies. The proposed changes deal exclusively with the reactor vessel P/T curves, which define the permissible regions for operation and testing.

Failure of the reactor vessel is not considered a design basis accident. Through the design conservatisms used to calculate the P/T curves, reactor vessel failure has a low probability of occurrence and is not considered in the safety analyses. The proposed changes adjust the reference temperature for the limiting material to account for irradiation effects and provide the same level of protection as previously evaluated and approved.

The adjusted reference temperature calculations were performed in accordance with the requirements of 10 CFR 50 Appendix G using the guidance contained in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," to reflect use of the operating limits to no more than 48 Effective Full Power Years (EFPY). These changes do not alter or prevent the operation of equipment required to mitigate any accident analyzed in the BFN Final Safety Analysis Report. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed changes are to accepted operating parameters that have been approved in previous license amendments. The changes to P/T curves were developed based on NRC approved methodologies. The proposed changes to the reactor vessel P/T curves do not involve a modification to plant equipment. No new failure modes are introduced.

There is no effect on the function of any plant system, and no new system interactions are introduced by this change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

E1-6

ENCLOSURE 1 EVALUATION OF PROPOSED CHANGES

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed changes are to accepted operating parameters that have been approved in previous license amendments. The changes to P/T curves were developed based on NRC approved methodologies. The proposed curves conform to the guidance contained in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," and maintain the safety margins specified in 10 CFR 50 Appendix G. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions 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.

5.0 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 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.

6.0 REFERENCES

1. Tennessee Valley Authority Browns Ferry Nuclear Plant, Unit 2, Docket No. 50-260, Renewed Facility Operating License, Renewed License No. DPR-52 (ADAMS Accession No. ML052780020).

2. Letter from TVA to NRC, Browns Ferry Nuclear Plant (BFN), Units 1, 2 and 3 -

Application for Renewed Operating Licenses, dated December 31, 2003 (ADAMS Accession No. ML040060359).

3. Letter from NRC to TVA, Browns Ferry Nuclear Plant, Units 2 and 3 - Issuance of Amendments Regarding Pressure-Temperature Limit Curves (TAC Nos. MC0807 and MC0808)," dated March 10, 2004 (ADAMS Accession Nos. ML040480013, ML040750188, and ML040750194).

E1-7

ENCLOSURE 1 EVALUATION OF PROPOSED CHANGES

4. GE Nuclear Energy, NEDC-32983P-A, Rev. 2 General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations, dated January 2006 (TAC No. MC9891).

5. GEH Nuclear Energy, NEDC-33178P-A, Revision 1, GE Hitachi Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves, June 2009 (GEH Proprietary).

6. Evaluation of RTNDT, USE, and Chemical Composition of Core Region Electroslag Welds for Quad Cities Units 1 and 2, Framatome Technologies, Lynchburg, Virginia, January 1996 (BAW-2259).

7. Letter, TE Abney (TVA) to US NRC, "Browns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3 - Generic Letter (GL) 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity - Response to NRC Request for Additional Information (TAC Nos. MA1179, MA1180, and MA1181)," September 8, 1998.

8. Letter from TVA to NRC, "Browns Ferry Nuclear Plant (BFN) - TVA Revision to Implementation Plant Described in Units 2 and 3 - Technical Specifications (TS)

Change No. 441 Revision 1 - Pressure-Temperature (P-T) Curve Update (MC0807 and MC0808)," dated February 24, 2004 (ADAMS Accession No. ML040550496).

9. Letter from Exelon Nuclear to NRC, Peach Bottom Atomic Power Station, Units and 3

- Issuance of Amendments RE: Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report (TAC Nos. ME8535 and ME8536),

dated April 1, 2013 (ADAMS Accession No. ML13079A219).

E1-8

ATTACHMENT 1 Proposed Technical Specifications Pages Markups

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 -------------------------- NOTES-----------------------

1. Only required to be performed during RCS heatup and cooldown operations or RCS inservice leak and hydrostatic testing when the vessel pressure is > 312 psig.

313

2. The limits of Figure 3.4.9-2 may be applied during nonnuclear heatup and ambient loss cooldown associated with inservice leak and hydrostatic testing provided that the heatup and cooldown rates are < 150F/hour.

3. The limits of Figures 3.4.9-1 and 3.4.9-2 do not apply when the tension from the reactor head flange bolting studs is removed.

Verify: 30 minutes

a. RCS pressure and RCS temperature are within the limits specified by Curves No. 1 and No. 2 of Figures 3.4.9-1 and 3.4.9-2; and

b. RCS heatup and cooldown rates are

< 100F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

SR 3.4.9.2 Verify RCS pressure and RCS temperature Once within 15 are within the criticality limits specified in minutes prior to Figure 3.4.9-1, Curve No. 3. control rod withdrawal for the purpose of achieving criticality (continued)

BFN-UNIT 2 3.4-26 Amendment No.-253, 288

RCS P/T Limits 3.4.9 1400 Curve No. 1 Minimum temperature for bottom head during 1300 mechanical heatup or cooldown following ROWNS FERRY UNIT2 _2_ nuclear shutdown.

1200 CURVES 1, 2, AND 3 3 ARE VALID FOR 23 EFPY Curve No. 2 1100 OF OPERATION Minimum temperature for upper RPV and beltline during mechanical 1000 heatup or cooldown following nuclear 900 shutdown.

Curve No. 3 800 Minimum temperature for core operation (criticality).

700 Notes 600 These curves include sufficient margin to provide protection 500 against feedwater nozzle degradation.

The curves allow for 400 shifts in RT,= of the Reactor vessel beltline 300 materials, in accordance with Reg.

Guide 1.99, Rev. 2, to 200 compensate for radiation embrittlement BOTTOM for 23 EFPY.

100 HEAD FLANGE- 38 0

68 F' REGION 83FF -- The acceptable area for 0 operation is to the right of the applicable 0 50 100 150 200 25 0 curves.

MINIMUM REACTOR VESSEL METAFL TEMPERATURE (OF)

Replace with Figure 3.4.9-1 Insert 1 Pressure/Temperature Limits for Mechanical Heatup, Cooldown following Shutdown, and Reactor Critical Operations BFN-UNIT 2 3.4-29 Amendment No. 257, 275, 288

1400 BROWNS FERRY UNIT 2 1 2 3 CURVES 1, 2 AND 3 ARE 1300 VALID FOR 38 EFPY OF OPERATION 1200 1100 PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 900 800 700 600 500 480 psig 420 psig 400 360 psig 313 psig 300 200 100 BOTTOM HEAD 68°F BOLTUP 83°F 0

0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Insert 1

RCS Pff Limits 3.4.9 1400 Curve No. 1 II_ l f--

Minimum temperature for bottom head 1 II 1300 2- r- during in-service f-BROWNS FERRY UNIT 2 I I leak or hydrostatic 7

r-CURVES 1 AND 2 ARE testing.

1200 r-VALID FOR 23 EFPY OF f-OPERATION f-- ll Curve No. 2

~

1100 l Minimum temperature c.? I J for upper RPV and beltline during in-H 1:1.1 1/ 1/

...... 1000 service leak or p.

I J hydrostatic testing.

II lj a 900 I

v J

l llf 0

E-4 800 Notes p:

I I These curves include 0

E-4 700 v ,, I sufficient margin to

~ provide protection against feedwater zH 600 nozzle degradation.

E-4 The curves allow for

~ 500 I shifts in RT~ of the H

.:I Reactor vessel beltline materials, m

1:1.1

• 400 in accordance with 1:1.1 I Reg. Guide 1.99, Rev.

t1llf 2, to compensate for 300 - radiation ernbrittlement for 23 200 f-- BOTTOM EFPY. 38 HEAD  ;

FLANGE f-- 68°F. REGION 83°F The acceptable area 100 I for operation is to I I I I the right of the 0 applicable curves.

0 50 100 150 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE {°F)

Replace with Figure 3.4.9-2 Insert 2 Pressure/Temperature Limits for Reactor In-Service Leak and Hydrostatic Testing BFN-UNIT2 3.4-29a Amendment No. 288

1400 BROWNS FERRY UNIT 2 1 2 CURVES 1 AND 2 ARE VALID FOR 38 EFPY OF 1300 OPERATION 1200 1100 PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 900 800 700 690 psig 600 530 psig 500 400 313 psig 300 BOTTOM HEAD 68°F 200 FLANGE REGION 83°F 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Insert 2

RCS Prr Limits 3.4.9 1400 Curve No. 1 r- slokJsiFlR~j Jm~J! -1 Ill II I Minimum temperature

~ for bottom head 1300 r- CURVES 1, 2, AND 3 ARE VALID FOR 3 0 I 2 during mechanical r- EFPY OF OPERATION 1:- - -3 heatup or cooldown following nuclear 1200 I shutdown.

...... Curve No. 2 CJ 1100 Minimum temperature I

H tl.l

p. for upper RPV and

~

if J J beltline during a

1000 mechanical heatup or

~- I

=p. 900 I cooldown following nuclear shutdown.

0 E-t I I 1/ Curve No. 3

..:r I 1 fzl tl.l 800 I Minimum temperature tl.l

~ for core operation l I (criticality).

~

p:: 700 J ~

0 E-t if J I Notes u These curves include a 600 1/ ~ 1 sufficient margin to

z:

H 1 I 1/ provide protection against feedwater E-t 500 J J nozzle degradation.

H I

~ The curves allow for H

..:r shifts in RT~ of the 400 Reactor vessel

~ i beltline materials, I:J tl.l tl.l in accordance with r--~

300 Reg. Guide 1.99, Rev.

~

p. J 2, to compensate for If If radiation 200 embrittlement for 30

)I ,)

If EFPY. 48 BOTTOM ~

100 HEAD FLANGE The acceptable area 6B°F- ,...~ REGION

=~ for operation is to B3°F 0 IIII I the right of the applicable curves.

0 50 100 150 200 250 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (oF)

Figure 3.4.9-1 Pressure/Temperature Limits for Mechanical Replace with Heatup, Cooldown following Shutdown, and Insert 3 Reactor Critical Operations BFN-UNIT2 3.4-29b Amendment No. 288

1400 BROWNS FERRY UNIT 2 1 2 3 CURVES 1, 2 AND 3 ARE VALID FOR 48 EFPY OF 1300 OPERATION 1200 1100 PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 900 800 700 600 500 480 psig 420 psig 400 313 psig 300 200 100 BOTTOM HEAD 68°F BOLTUP 83°F 0

0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Insert 3

RCS Pff Limits 3.4.9 1400 Curve No. 1

~ I Jl Minimum temperature 1300 1 2 for bottom head BROWNS FERRY UNIT 2 during in-service f-cURVES 1 AND 2 ARE - 1/ leak or hydrostatic VALID FOR 30 EFPY 1200 OF OPERATION ~ testing.

I

..... 1100 I Curve No. 2 Minimum temperature Cl H

ttl I for upper RPV and PI 1000 1 I beltline during in-service leak or

~

a 900 1/ J hydrostatic testing.

PI 0

f-1 I

..:I rzl 800 j Notes

'(

ttl ttl J J These curves include

~ I 1/ sufficient margin to p:;

0 f-1 700 v provide protection against feedwater

)I nozzle degradation.

~ 600 The curves allow for shifts in RTmn of the 12:

H 500 Reactor vessel f-1 H beltline materials, rs

..:I

/ in accordance with 400 / Reg. Guide 1.99, Rev.

2, to compensate for

~

ttl radiation

~ 300 ernbrittlement for 30 PI EFPY. 48 I

200 The acceptable area 1- BOTTOM I for operation is to 100 HEAD the right of the 1- 68°F---...... FLANGE REGION applicable curves.

I I I 83°F 0 I I I I 0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 3.4.9-2 Pressure/Temperature Limits for Replace with Reactor In-Service Leak and Hydrostatic Testing Insert 4 BFN-UNIT 2 3.4-29c Amendment No. 288

1400 BROWNS FERRY UNIT 2 1 2 CURVES 1 AND 2 ARE VALID FOR 48 EFPY OF 1300 OPERATION 1200 1100 PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 900 800 700 690 psig 600 500 500 psig 400 313 psig 300 BOTTOM HEAD 68°F 200 FLANGE REGION 83°F 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Insert 4

ATTACHMENT 2 Proposed Technical Specifications Bases Pages Markups (for information only)

RCS P/T Limits B 3.4.9 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.

The pressure-temperature (P-T) curves included in the Technical Specifications have been developed to present 38 steam-dome pressure versus minimum vessel metal temperature, incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline. There are two sets of curves provided for Unit 2. The first set applies to and including 38 operation up to 23 effective full power years (EFPY), and the second set applies to operation greater than 23 EFPY and less than or equal to 48 than 30 EFPY, where 30 EFPY represents the end of the 48 renewed 40-year license and 23 EFPY is provided as a midpoint between the current EFPY and 30 EFPY. The P-T curves are provided in 38 Figure 3.4.9-1 and Figure 3.4.9-2, respectively. Figure 3.4.9-1 contains P-T limit curves for mechanical heatup or cooldown following nuclear shutdown (bottom head and upper RPV/beltline) and for core operation (criticality). Figure 3.4.9-2 contains P-T limit curves for inservice leakage and hydrostatic testing (bottom head and upper RPV/beltline). The maximum rate of change of reactor coolant temperature is contained in SR 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7.

(continued)

BFN-UNIT 2 B 3.4-55 Revision 0, 26 March 17, 2004

RCS P/T Limits B 3.4.9 BASES Section XI BACKGROUND The P-T curves incorporate a fluence calculated in accordance (continued) with GE Licensing Topical Report NEDC-32983P (Ref. 10),

which has been approved by the NRC and is in compliance with Regulatory Guide 1.190 (Ref. 11). The fluence represents an Extended Power Uprate (EPU) for the rated power of 3952 MWt, and is conservatively applied for the rated power of 3458 MWt.

The 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda was used in accordance with 10 CFR 50.55a.

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 is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel.

(continued)

BFN-UNIT 2 B 3.4-55a Revision 26 March 17, 2004

RCS P/T Limits B 3.4.9 BASES BACKGROUND 10 CFR 50, Appendix G (Ref. 1), requires the establishment of (continued) P/T limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, abnormal operational transients, and system hydrostatic tests. It mandates the use of the ASME Code,Section III, Appendix G (Ref. 2).

Revision 1 The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3), Appendix H of 10 CFR 50 (Ref. 4),

and BWRVIP-86-A (Ref.12), and BWRVIP-116 (Ref. 13). The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 5.

Minimum reactor vessel temperature requirements for pressure-temperature (P/T) limits depend on the reactor vessels controlling material (which is either the material in the closure flange or the material in the beltline region with the highest reference temperature) and on the reactors operating condition (i.e., hydrostatic pressure and leak tests, or normal operation including anticipated operation occurrences), vessel pressure, whether or not fuel is in the vessel, and whether the core is critical. The beltline region of the reactor vessel is the region that directly surrounds the effective height of the active core and adjacent regions that are predicted to experience sufficient radiation exposure to be considered in the selection of the most limiting material with regard to radiation exposure. The metal temperature of the controlling material, in the region of the controlling material which has the least favorable combination of stress and temperature, must exceed the appropriate minimum temperature requirements for the reactor vessel condition and pressure specified in 10 CFR 50, Appendix G.

(continued)

BFN-UNIT 2 B 3.4-56 Revision 0, 15, 26, 38 September 21, 2006

RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 (continued)

REQUIREMENTS This SR has been modified by three Notes. Note 1 requires this Surveillance to be performed only during system heatup and cooldown operations or inservice leakage and hydrostatic testing. Also, Note 1 only requires this SR to be performed 313 during inservice leakage and hydrostatic testing when reactor pressure is > 312 psig. Note 2 allows the limits of Figure 3.4.9-2 to be applied during nonnuclear heatup and ambient loss cooldown associated with inservice leak and hydrostatic testing provided that the heatup and cooldown rates are 15°F/hr.

Note 3 provides that the limits of Figures 3.4.9-1 and 3.4.9-2 do not apply when the tension from the reactor head flange bolting studs is removed.

SR 3.4.9.2 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.

(continued)

BFN-UNIT 2 B 3.4-63 Revision 0, 26 March 17, 2004

RCS P/T Limits B 3.4.9 BASES REFERENCES 12. BWRVIP-86-A: BWR Vessel and Internals Project, (continued) Updated BWR Integrated Surveillance Program (ISP)

Implementation Plan, EPRI Technical Report 1003346, October 2002.

13. BWRVIP-116: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Implementation For License Renewal, EPRI Technical Report 1007824, July 2003.

"BWRVIP-86 Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan, " EPRI Technical Report 1016575, May 2013.

BFN-UNIT 2 B 3.4-66a Revision 26, 38 September 21, 2006

ATTACHMENT 3 Proposed Retyped Technical Specifications Pages

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 --------------------------NOTES-----------------------

1. Only required to be performed during RCS heatup and cooldown operations or RCS inservice leak and hydrostatic testing when the vessel pressure is > 313 psig.

2. The limits of Figure 3.4.9-2 may be applied during nonnuclear heatup and ambient loss cooldown associated with inservice leak and hydrostatic testing provided that the heatup and cooldown rates are 15°F/hour.

3. The limits of Figures 3.4.9-1 and 3.4.9-2 do not apply when the tension from the reactor head flange bolting studs is removed.

Verify: 30 minutes

a. RCS pressure and RCS temperature are within the limits specified by Curves No. 1 and No. 2 of Figures 3.4.9-1 and 3.4.9-2; and

b. RCS heatup and cooldown rates are 100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

SR 3.4.9.2 Verify RCS pressure and RCS temperature Once within 15 are within the criticality limits specified in minutes prior to Figure 3.4.9-1, Curve No. 3. control rod withdrawal for the purpose of achieving criticality (continued)

BFN-UNIT 2 3.4-26 Amendment No. 253, 288, 000

RCS P/T Limits 3.4.9 1400 Curve No. 1 BROWNS FERRY UNIT 2 1 2 3 CURVES 1, 2 AND 3 ARE Minimum temperature 1300 VALID FOR 38 EFPY OF for bottom head during OPERATION mechanical heatup or cooldown following 1200 nuclear shutdown.

Curve No. 2 1100 Minimum temperature for upper RPV and beltline during mechanical heatup or PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 cooldown following nuclear shutdown.

900 Curve No. 3 Minimum temperature 800 for core operation (criticality).

700 Notes These curves include sufficient margin to 600 provide protection against feedwater nozzle degradation.

500 480 psig The curves allow for shifts in RTNDT of the 420 psig Reactor vessel 400 beltline materials, in 360 psig accordance with Reg.

313 psig Guide 1.99, Rev. 2, to 300 compensate for radiation embrittlement for 38 200 EFPY.

The acceptable area 100 BOTTOM HEAD for operation is to 68°F BOLTUP the right of the 83°F applicable curves.

0 0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 3.4.9-1 Pressure/Temperature Limits for Mechanical Heatup, Cooldown following Shutdown, and Reactor Critical Operations BFN-UNIT 2 3.4-29 Amendment No. 257, 275, 288, 000

RCS P/T Limits 3.4.9 1400 BROWNS FERRY UNIT 2 1 2 Curve No. 1 CURVES 1 AND 2 ARE Minimum temperature for 1300 VALID FOR 38 EFPY OF bottom head during OPERATION in-service leak or hydrostatic testing.

1200 Curve No. 2 Minimum temperature for 1100 upper RPV and beltline during in-service leak or hydrostatic testing.

PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 Notes 900 These curves include sufficient margin to provide protection 800 against feedwater nozzle degradation. The curves allow for shifts in RTNDT 700 690 psig of the Reactor vessel beltline materials, in accordance with Reg.

600 Guide 1.99, Rev. 2, to compensate for radiation 530 psig embrittlement for 38 500 EFPY.

The acceptable area for 400 operation is to the right of the applicable curves.

313 psig 300 BOTTOM HEAD 68°F 200 FLANGE REGION 83°F 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 3.4.9-2 Pressure/Temperature Limits for Reactor In-Service Leak and Hydrostatic Testing BFN-UNIT 2 3.4-29a Amendment No. 288, 000

RCS P/T Limits 3.4.9 1400 Curve No. 1 BROWNS FERRY UNIT 2 1 2 3 CURVES 1, 2 AND 3 ARE Minimum temperature VALID FOR 48 EFPY OF for bottom head during 1300 mechanical heatup or OPERATION cooldown following nuclear shutdown.

1200 Curve No. 2 Minimum temperature 1100 for upper RPV and beltline during mechanical heatup or 1000 PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) cooldown following nuclear shutdown.

900 Curve No. 3 Minimum temperature for core operation 800 (criticality).

Notes 700 These curves include sufficient margin to provide protection 600 against feedwater nozzle degradation.

The curves allow for 500 480 psig shifts in RTNDT of the Reactor vessel 420 psig beltline materials, in 400 accordance with Reg.

Guide 1.99, Rev. 2, to 313 psig compensate for 300 radiation embrittlement for 48 EFPY.

200 The acceptable area BOTTOM for operation is to 100 the right of the HEAD 68°F BOLTUP applicable curves.

83°F 0

0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 3.4.9-1 Pressure/Temperature Limits for Mechanical Heatup, Cooldown following Shutdown, and Reactor Critical Operations BFN-UNIT 2 3.4-29b Amendment No. 288, 000

RCS P/T Limits 3.4.9 1400 BROWNS FERRY UNIT 2 1 2 Curve No. 1 CURVES 1 AND 2 ARE 1300 VALID FOR 48 EFPY OF Minimum temperature OPERATION for bottom head during in-service leak or 1200 hydrostatic testing.

Curve No. 2 1100 Minimum temperature for upper RPV and beltline during in-1000 service leak or PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) hydrostatic testing.

900 Notes 800 These curves include sufficient margin to provide protection 700 690 psig against feedwater nozzle degradation.

600 The curves allow for shifts in RTNDT of the Reactor vessel 500 500 psig beltline materials, in accordance with Reg.

Guide 1.99, Rev. 2, to 400 compensate for radiation 313 psig embrittlement for 48 300 BOTTOM EFPY.

HEAD 68°F The acceptable area 200 FLANGE for operation is to REGION 83°F the right of the applicable curves.

100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 3.4.9-2 Pressure/Temperature Limits for Reactor In-Service Leak and Hydrostatic Testing BFN-UNIT 2 3.4-29c Amendment No. 288, 000

ATTACHMENT 4 Proposed Retyped Technical Specifications Bases Pages (for information only)

RCS P/T Limits B 3.4.9 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.

The pressure-temperature (P-T) curves included in the Technical Specifications have been developed to present steam-dome pressure versus minimum vessel metal temperature, incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline. There are two sets of curves provided for Unit 2. The first set applies to operation up to and including 38 effective full power years (EFPY), and the second set applies to operation greater than 38 EFPY and less than or equal to 48 EFPY, where 48 EFPY represents the end of the renewed license and 38 EFPY is provided as a midpoint between the current EFPY and 48 EFPY. The P-T curves are provided in Figure 3.4.9-1 and Figure 3.4.9-2, respectively. Figure 3.4.9-1 contains P-T limit curves for mechanical heatup or cooldown following nuclear shutdown (bottom head and upper RPV/beltline) and for core operation (criticality). Figure 3.4.9-2 contains P-T limit curves for inservice leakage and hydrostatic testing (bottom head and upper RPV/beltline). The maximum rate of change of reactor coolant temperature is contained in SR 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7.

(continued)

BFN-UNIT 2 B 3.4-55 Revision 0, 26, 00

RCS P/T Limits B 3.4.9 BASES BACKGROUND The P-T curves incorporate a fluence calculated in accordance (continued) with GE Licensing Topical Report NEDC-32983P (Ref. 10),

which has been approved by the NRC and is in compliance with Regulatory Guide 1.190 (Ref. 11). The fluence represents an Extended Power Uprate (EPU) for the rated power of 3952 MWt, and is conservatively applied for the rated power of 3458 MWt.

The 1998 Edition of the ASME Section XI Boiler and Pressure Vessel Code including 2000 Addenda was used in accordance with 10 CFR 50.55a.

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 is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel.

(continued)

BFN-UNIT 2 B 3.4-55a Revision 26, 00

RCS P/T Limits B 3.4.9 BASES BACKGROUND 10 CFR 50, Appendix G (Ref. 1), requires the establishment of (continued) P/T limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, abnormal operational transients, and system hydrostatic tests. It mandates the use of the ASME Code,Section III, Appendix G (Ref. 2).

The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3), Appendix H of 10 CFR 50 (Ref. 4), and BWRVIP-86-A, Revision 1 (Ref. 12). The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 5.

Minimum reactor vessel temperature requirements for pressure-temperature (P/T) limits depend on the reactor vessels controlling material (which is either the material in the closure flange or the material in the beltline region with the highest reference temperature) and on the reactors operating condition (i.e., hydrostatic pressure and leak tests, or normal operation including anticipated operation occurrences), vessel pressure, whether or not fuel is in the vessel, and whether the core is critical. The beltline region of the reactor vessel is the region that directly surrounds the effective height of the active core and adjacent regions that are predicted to experience sufficient radiation exposure to be considered in the selection of the most limiting material with regard to radiation exposure. The metal temperature of the controlling material, in the region of the controlling material which has the least favorable combination of stress and temperature, must exceed the appropriate minimum temperature requirements for the reactor vessel condition and pressure specified in 10 CFR 50, Appendix G.

(continued)

BFN-UNIT 2 B 3.4-56 Revision 0, 15, 26, 38, 00

RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 (continued)

REQUIREMENTS This SR has been modified by three Notes. Note 1 requires this Surveillance to be performed only during system heatup and cooldown operations or inservice leakage and hydrostatic testing. Also, Note 1 only requires this SR to be performed during inservice leakage and hydrostatic testing when reactor pressure is > 313 psig. Note 2 allows the limits of Figure 3.4.9-2 to be applied during nonnuclear heatup and ambient loss cooldown associated with inservice leak and hydrostatic testing provided that the heatup and cooldown rates are 15°F/hr.

Note 3 provides that the limits of Figures 3.4.9-1 and 3.4.9-2 do not apply when the tension from the reactor head flange bolting studs is removed.

SR 3.4.9.2 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.

(continued)

BFN-UNIT 2 B 3.4-63 Revision 0, 26, 00

RCS P/T Limits B 3.4.9 BASES REFERENCES 12. BWRVIP-86 Revision 1-A: BWR Vessel and Internals (continued) Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan, EPRI Technical Report 1016575, May 2013.

BFN-UNIT 2 B 3.4-66a Revision 26, 38, 00

ATTACHMENT 5 Proposed Updated Final Safety Analysis Report Page Markups

BFN-25.3 Although little corrosion of plain carbon or low-alloy steels occurs at temperatures of 500qF to 600qF, higher corrosion rates occur at temperatures around 140qF. The 0.125-inch minimum-thickness cladding provides the necessary corrosion resistance during reactor shutdown and also helps maintain water clarity during refueling operations. Since the vessel head is exposed to a saturated steam environment throughout its operating lifetime, stainless steel cladding is not required over its interior surfaces. Exterior, exposed ferritic surfaces of pressure-containing parts have a minimum corrosion allowance of 1/16 inch. The interior surfaces of the top head and all carbon and low-alloy steel nozzles exposed to the reactor coolant have a corrosion allowance of 1/16 inch. The vessel shape is designed to limit coolant retention pockets and crevices.

The nil-ductility transition (NDT) temperature is defined as the temperature below which ferritic steel breaks in a brittle, rather than ductile, manner. The NDT temperature increases as a function of neutron exposure at integrated neutron exposures greater than about 1 x 1017 nvt with neutrons of energies in excess of 1 MeV. Since the material NDT temperature dictates the minimum operating temperature at which the reactor vessel can be pressurized, it is desirable to keep the NDT temperature as low as possible. One way that this is accomplished is by selecting fine-grained steels and by using advanced fabrication techniques to minimize radiation effects. The as-fabricated initial NDT temperature for all carbon and low-alloy steel used in the main closure flanges, closure bolting material, and the shell and head materials connecting to these flanges, including the connecting circumferential weld material, is limited to a maximum of 10qF as determined by ASTM E208. For each main closure flange forging, a minimum of 1 tensile, 3 Charpy V-notch, and 2 drop weight test specimens have been tested from each of two locations about 180q apart on the flange. For all other carbon and low-alloy steel pressure-containing materials, including weld materials and the vessel support skirt material, the as-fabricated initial NDT temperature is no higher than 40qF. A grain size of 5 or finer, as determined by the method in ASTM E112, is maintained.

Another way of minimizing any changes (elevating) to the NDT temperature is by reducing the integrated neutron exposure at the inner surface of the reactor vessel.

The coolant annulus between the vessel and core shroud and the core location in the vessel limit the integrated neutron exposure of reactor vessel material to less than 1 x 1019 nvt from neutrons with energy levels greater than 1 MeV, within the 40-year design lifetime of the vessel. TLAAs have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Sections O.3.1 and O.3.2. This is not the expected exposure, nor is it the absolute limit of safe exposure; it is an exposure value that can be demonstrated to be safe and practical to maintain. The maximum calculated exposure for neutrons of 1 MeV or greater is 3.8 X 1017 nvt.

less than1x 1019 4.2-3

BFN-25.3 ambient temperatures, the insulation or enclosure provided, and the minimum temperature maintained. Further interpretations and requirements are as follows:

A. Charpy V-notch (American Society for Testing and Material Standard A370 Type A) or drop weight (per ASTM E208) tests have been performed to , for the demonstrate that all materials and weld metal meet brittle fracture surveillance requirements at test temperature. Test specimens were prepared and tested capsule pulled with minimum impact energy requirements in accordance with Table N-421 in 1994, and the general provisions of N-313, N-331, N-332, and N-511 of Section III of the ASME Boiler and Pressure Vessel Code. Prior to the Summer 1972 Addenda of the 1971 ASME Section III Boiler and Pressure Vessel Code, impact testing was not required on materials with a nominal section thickness of 1/2 inch or less. However, this 1/2 inch thickness exclusion was increased to 5/8 inch by the ASME Boiler and Pressure Vessel Code,Section III, 1971 Edition, Summer 1972 Addenda. Therefore, after issuance of the Summer 1972 Addenda, impact testing is not required on materials with a nominal section thickness of 5/8 inch or less. The welding procedures used were qualified by impact testing of weld metal and heat affected zone to the same requirements as the base metal in accordance with N-541.

B. Impact tests were not required for the following:

1. Bolting, including nuts, 1-inch nominal diameter or less,

2. Bars with a nominal cross-sectional area not exceeding 1 square inch,

3. Materials with a nominal (section) wall thickness of less than 1/2 inch or 5/8 inch (refer to paragraph 4.2.4.10.A),

4. Components including pumps, valves, piping, and fittings with a nominal inlet pipe size of 6-inch-diameter and less, regardless of thickness, and

5. Consumable insert material, austenitic stainless steel, and nonferrous materials.

C. Impact testing was not required on components or equipment pressure parts having a minimum service temperature of 250qF or more when pressured over 20 percent of the design pressure. Example: Steam line is excluded from brittle fracture test requirement since the steam temperature will be over 250qF when the steam line pressure is at the 20 percent design pressure.

for the surveillance capsule pulled in 2011, per BWRVIP-271/NP, the Charpy impact tests were conducted in accordance with ASTM Standards E185-82 and E23-02.

4.2-9

BFN-25.3 Appendix G, "Fracture Toughness Requirements," paragraph IV.A.2.C. An exemption from specific requirements of 10 CFR Part 50, Appendix G is taken by use of ASME Code Case N-640 for Unit 2 and Unit 3. ASME Code Case N-640 permits the use of an alternative reference fracture curve K1c for RPV materials for use in determining the PT limits. The PT limit curves based on the K1c fracture toughness curve enhance overall plant safety by minimizing challenges to operators since requirements for maintaining a high vessel temperature during pressure testing are lessened. ASME Code Case N-588 methodology was also used as a basis for the PT curves. This code case permits the use of an alternative procedure for calculating applied stress intensity factors during normal operation and pressure test conditions due to pressure and thermal gradients for axial flaws. This methodology is incorporated into the ASME,Section XI Code, 1995 Edition, 1996 Addenda, which is the current code of record for the Unit 2 inservice inspection program. Since Unit 3 uses an earlier code of record for the inservice inspection program, Unit 3 implements the requirements of only the 1995 Edition, 1996 Addenda of ASME Section XI, Appendix G to allow the use of the ASME Code Case N-588 methodology for PT curves. The operating limits are provided in the technical specifications for Browns Ferry. For the purpose of setting these operating limits, the initial RTNDT (nil-ductility reference temperature) was determined from the impact test data taken in accordance with the requirements of the code to which the reactor vessels were designed and manufactured. The maximum NDT temperature allowed by the vessel specifications was 40qF.

Although test data on beltline base material show lower NDT temperatures, an assumed RTNDT of 40qF was used in the vessel beltline area, as well as the areas remote from the beltline because the generally accepted NDT temperature for electroslag welds used in the beltline longitudinal seams is 40qF.

The current operating limits on the pressure/temperature (P/T) curves in the technical specifications are based on the following (RTNDT) values. Unit 1 has used 20qF for the (RTNDT) value, Unit 2 has used 22qF for the (RTNDT) value, and Unit 3 has used 10qF for the (RTNDT) value. 23.1 For power uprated conditions, the estimated fluence was conservatively increased above the UFSAR end-of-life value. This fluence increase was estimated to be greater than proportional to the power increase, considering the changes in power distribution. The higher fluence was used to evaluate the vessel against the requirements of 10 CFR 50, Appendix G in accordance with Regulatory Guide 1.99, Revision 2. The results of these evaluations indicated that:

(a) The upper shelf energy will remain greater than 50 ft-lb for the design life of the vessel and maintain the margin requirements of Appendix G.

The end-of-life upper shelf energy was evaluated by an equivalent (a) The results of the upper shelf energy EMA for margin analysis (EMA). limiting welds and plates for the three vessels remain 4.2-12 less than the acceptance criterion in all cases.

...license renewal, fluences were conservatively calculated for licensed operating periods of 54 EFPY for Unit 1 and 52 EFPY for Units 2 and 3.

Proprietary Information Withhold Under 10 CFR 2.390 ENCLOSURE 2 NEDC-33854P - Pressure and Temperature Limits Report (PTLR) Up to 38 and 48 Effective Full Power Years (Proprietary)

This report contains GEH and EPRI proprietary information. Affidavits supporting the proprietary nature of this report are provided in Enclosure 4.

ENCLOSURE 3 NEDO-33854 - Pressure and Temperature Limits Report (PTLR) Up to 38 and 48 Effective Full Power Years (Non-Proprietary)

GE Hitachi Nuclear Energy NEDO-33854 Revision 0 April 2014 Non-Proprietary Information-Class I (Public)

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNIT 2 Pressure and Temperature Limits Report (PTLR) up to 38 and 48 Effective Full Power Years Copyright 2014 GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

NON-PROPRIETARY INFORMATION NOTICE This is a non-proprietary version of NEDC-33854P, Revision 0 which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here (( )).

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY The design, engineering, and other information contained in this document are furnished for the purposes of supporting a License Amendment Request by Tennessee Valley Authority (TVA) for pressure temperature limits in proceedings before the U.S. Nuclear Regulatory Commission.

The only undertakings of the GEH respecting information in this document are contained in the contract between TVA and GEH, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than TVA, or for any purpose other than that for which it is intended, is not authorized; and, with respect to any unauthorized use, GEH makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

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Table of Contents Page Abbreviations & Acronyms ................................................................................................................... iv 1.0 Purpose ............................................................................................................................. 1 2.0 Applicability ..................................................................................................................... 1 3.0 Methodology .................................................................................................................... 1 4.0 Operating Limits............................................................................................................... 5 5.0 Discussion ........................................................................................................................ 6 6.0 References ........................................................................................................................ 9 Table of Figures Figure 1. BFNP Unit 2 Composite Curve A Pressure Test P-T Curves Effective for up to 38 EFPY ......................................................................................................................... 11 Figure 2. BFNP Unit 2 Composite Curve B Core Not Critical including Bottom Head and Curve C Core Critical P-T Curves Effective for up to 38 EFPY ................................... 12 Figure 3. BFNP Unit 2 Composite Curve A Pressure Test P-T Curves Effective for up to 48 EFPY ......................................................................................................................... 13 Figure 4. BFNP Unit 2 Composite Curve B Core Not Critical including Bottom Head and Curve C Core Critical P-T Curves Effective for up to 48 EFPY ................................... 14 Table of Tables Table 1 - BFNP Unit 2 Tabulation of Curves - 38 EFPY ................................................................... 15 Table 2 - BFNP Unit 2 Tabulation of Curves - 48 EFPY ................................................................... 20 Table of Appendices Appendix A. Reactor Vessel Material Surveillance Program ............................................................. 25 Appendix B. BFNP Unit 2 Reactor Pressure Vessel P-T Curve Supporting Plant-Specific Information ..................................................................................................................... 26 Appendix C. BFNP Unit 2 Reactor Pressure Vessel P-T Curve Checklist ......................................... 37 Appendix D. Sample P-T Curve Calculations ..................................................................................... 42 iii

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Abbreviations & Acronyms Short Form Description

%Cu Weight percent Copper

%Ni Weight percent Nickel 1/4T 1/4 depth into the vessel wall from the inside diameter 3/4T 3/4 depth into the vessel wall from the inside diameter ASME American Society of Mechanical Engineers ART Adjusted Reference Temperature BFNP Browns Ferry Nuclear Plant BWR Boiling Water Reactor BWR/6 BWR Product Line 6 BWRVIP BWR Vessel and Internals Project CF Chemistry Factor CMTR Certified Material Test Report CRD Control Rod Drive EFPY Effective Full Power Years EPRI Electric Power Research Institute ESW Electroslag Weld FW Feedwater GEH GE-Hitachi Nuclear Energy Americas LLC GL Generic Letter ID Inside Diameter ISP Integrated Surveillance Program LTR Licensing Topical Report n/cm2 neutrons per square centimeter (measure of fluence)

N16 BFNP Unit 1 Water Level Instrumentation Nozzle NDT Nil Ductility Transition NRC Nuclear Regulatory Commission P/T Pressure and Temperature P-T Pressure-Temperature PTLR Pressure and Temperature Limits Report RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RG Regulatory Guide RPV Reactor Pressure Vessel RTNDT Reference Temperature of Nil Ductility Transition RVID Reactor Vessel Integrity Database (by NRC)

SSP Supplemental Surveillance Program TS Technical Specification TVA Tennessee Valley Authority UFSAR Updated Final Safety Analysis Report WLI Water Level Instrumentation WRC Welding Research Council iv

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public) 1.0 Purpose The purpose of the Browns Ferry Nuclear Plant (BFNP) Unit 2 Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

1. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heatup, Cooldown and Hydrostatic/Class 1 Leak Testing;

2. RCS Heatup and Cooldown rates;

3. Reactor Pressure Vessel (RPV) to RCS coolant T requirements during Recirculation Pump startups;

4. RPV bottom head coolant temperature to RPV coolant temperature T requirements during Recirculation Pump startups;

5. RPV head flange bolt-up temperature limits.

This report has been prepared in accordance with the requirements defined in Reference 6.2.

2.0 Applicability This report is applicable to the BFNP Unit 2 RPV for up to 38 and 48 Effective Full Power Years (EFPY), representing a 60-year license.

The following Technical Specification (TS) is affected by the information contained in this report:

TS 3.4.9 RCS Pressure and Temperature (P/T) Limits 3.0 Methodology The limits in this report were derived from the Nuclear Regulatory Commission (NRC)-approved methods listed in the specific revisions listed below:

1. The neutron fluence was calculated per Licensing Topical Report (LTR), General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation, NEDC-32983P-A, Revision 2, January 2006, approved in Reference 6.1.

2. The pressure and temperature limits were calculated per GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves, NEDC-33178P-A, Revision 1, June 2009, approved in Reference 6.2.

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

• Application of GEH Topical Report for Pressure-Temperature (P-T) Curves

• The Water Level Instrumentation (WLI) nozzle in the beltline region was fabricated from ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °° ° ° ° ° ° )) material and has been considered in the Adjusted Reference Temperature (ART) evaluation. The material properties of the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) material have been considered with the fluence for the nozzle location.

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• Application of new Integrated Surveillance Program (ISP) testing and analysis results from the BFNP Unit 2 surveillance capsule that are applicable to BFNP Unit 2.

3.1 Chemistry The N16 WLI nozzle is defined as being within the beltline region, and is evaluated for ART.

This nozzle is fabricated from ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) materials. Therefore, the chemistry for the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) to evaluate the ART to represent this nozzle forging.

Chemistry for all other materials remains unchanged from those used in the development of the currently licensed P-T curves.

Surveillance materials are evaluated using the chemistries obtained from Boiling Water Reactor (BWR) Vessel and Internals Project (BWRVIP)-135, Revision 2 and Reference 6.5, as presented in Section 3.4. It is noted that there are no best estimate chemistries for the BFNP Unit 2 beltline materials described in BWRVIP-135.

The chemistry factors (CF) for all materials are calculated based upon the requirements of Regulatory Guide (RG) 1.99, Revision 2.

3.2 Initial Reference Temperature of Nil-Ductility Transition The N16 WLI nozzle is evaluated for ART. As the nozzle forging is fabricated ((° ° ° ° ° ° ° ° ° ° ° ° ° ° °

°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

Surveillance materials are evaluated using the limiting initial RTNDT of the beltline plate or weld material.

Initial RTNDTs for all other beltline materials remain unchanged from those used in the development of the currently licensed P-T curves.

3.3 Adjusted Reference Temperature The ART values for 38 and 48 EFPY included in Appendix B are developed considering the latest BWRVIP ISP surveillance data available that is representative of the applicable materials in the BFNP Unit 2 RPV (Reference 6.3). The surveillance data used in the BFNP Unit 2 ART calculations are obtained from actual BFNP Unit 2 RPV test specimens. The ISP plate materials are not limiting with respect to the ART. The ISP weld is the limiting material; this value is considered in the development of the P-T curves because the ISP material is considered to be the identical heat to the material in the BFNP Unit 2 RPV.

The N16 nozzle ART is determined considering the initial RTNDT ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

material together with the fluence at the nozzle elevation.

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NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public) 3.4 Surveillance Program As discussed in Appendix A, BFNP Unit 2 participates in the ISP. Of the three surveillance capsules installed at plant startup, one remains in the vessel. BFNP Unit 2 is a host plant in the ISP. The specimens from the first surveillance capsule that was removed at approximately 8.2 EFPY were reconstituted and placed in the vessel in the same location as those removed.

Therefore, two capsules remain in the vessel, and are slated for removal at approximately 20 and 26 EFPY per Reference 6.4.

BWRVIP-135, Revision 2 and Reference 6.5 provide the surveillance data considered in determining the chemistry and any adjusted CF for the beltline materials.

Excerpt from Reference 6.5:

((

))

}

For BFNP Unit 2, the ISP representative plate, heat (( )) target plate. This heat (( )). The resultant chemistry is (( )) Cu and (( )) Ni. The CF from RG 1.99, Revision 2 is

(( )). The fitted CF is (( )); as the ISP material (( ))

BFNP Unit 2 vessel, and the (( )) the RG 1.99 CF, the ART table evaluated the ISP plate material (( )). This material (( )) in determining the limiting ART for the P-T curves, but (( )).

Excerpt from Reference 6.5:

((

))

For BFNP Unit 2, the ISP representative weld, heat ((

)). This heat was ((

} }

)). The resultant chemistry is (( )) Cu and

} }

(( )) Ni. The CF from RG 1.99 is (( )). The fitted CF is (( )),

calculated using surveillance data. The scatter in this data ((

)). Since the ISP material ((

}

)) in the ART calculation. In addition, the 3

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(( )). This material (( )) the limiting ART for the P-T curves, ((

)).

3.5 Reactor Coolant Pressure Boundary American Society of Mechanical Engineers (ASME) Code Section XI, Appendix G, Article G-3000, paragraph G-3100 states that for materials used for piping, pumps, and valves for which impact tests are required (NB-2311), the tests and acceptance standards of Section III, Division 1 are considered to be adequate to prevent non-ductile failure under the loadings and with the defect sizes encountered under normal, upset, and testing conditions. Level C and Level D Service Limits should be evaluated on an individual case basis (G-2300). As described in Section 4.3 of Reference 6.2, ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) the development of all non-beltline P-T limits.

ASME Section III, Paragraph NB-2332 states that Pressure-retaining materials (other than bolting) with nominal thickness over 2.5 inches for piping (pipe and tubes) and materials for pumps, valves and fittings with any pipe connection of nominal wall thickness greater than 2.5 inches shall meet the requirements of NB-2331. The lowest service temperature shall be not lower than RTNDT + 100°F unless a lower temperature is justified by following the methods similar to those contained in Article G-2000. All BFNP Unit 2 ferritic Reactor Coolant Pressure Boundary (RCPB) piping has nominal wall thicknesses less than 2.5 inches. Other Class 1 RCPB components are significantly smaller with nominal wall thicknesses well below 2.5 inches, including all of the ferritic RCPB components. The lowest service temperatures may be less than 250°F in some cases; however the methods of Appendix G have been followed to justify lower temperatures. Therefore, the requirements of NB-2332 have been met, and there are no ferritic RCPB piping components that require consideration in the RPV P-T curves for BFNP Unit 2.

The ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )), are more limiting relative to stress than any of the Class 1 ferritic branch piping in the RCPB.

With respect to concern regarding irradiation effects on RCPB piping, a qualitative fluence assessment was performed. With a 48 EFPY peak surface inside diameter (ID) fluence of 1.93e18 n/cm2 as the maximum fluence of concern, accrual of fluence greater than 1.0e17 n/cm2 outside the vessel for 48 EFPY is not expected, based on historical calculations of flux vs. vessel thickness.

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As discussed in the BFNP Updated Final Safety Analysis Report (UFSAR), it is noted that the manner in which the RCPB was designed and constructed was to ensure a high degree of integrity with adequate toughness throughout the plant life. The RCPB components were designed and fabricated, and are maintained and tested such that adequate assurance is provided that the boundary will behave in a non-brittle manner throughout the life of the plant.

3.6 Future Changes 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 UFSAR, can be made pursuant to 10 CFR 50.59, provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

4.0 Operating Limits The P-T curves provided 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) non-nuclear heatup/cooldown (core not critical), referred to as Curve B; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 38 and 48 EFPY. The P-T curves are provided in Figures 1 through 4, and a tabulation of the curves is included in Table 1 (38 EFPY) and Table 2 (48 EFPY).

Other temperature limits applicable to the RPV and controlled by the TS are:

• Heatup and Cooldown rate limit during Hydrostatic and Class 1 Leak Testing: 15

°F/hour.

• Normal Operating Heatup and Cooldown rate limit: 100 °F/hour.

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

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

• RPV flange and adjacent shell temperature limit: 80 °F.

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NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public) 5.0 Discussion The procedures described in References 6.1 and 6.2 were used in the development of the P-T curves for BFNP Unit 2.

The method for determining the initial Reference Temperature of Nil-Ductility Transition (RTNDT) for all vessel materials is defined in Section 4.1.2 of Reference 6.2. Initial RTNDT values for all vessel materials considered are presented in tables in Appendix B of this report.

In order to ensure that the limiting vessel discontinuity has been considered in the development of the P-T curves, the methods in Sections 4.3.2.1 and 4.3.2.2 of Reference 6.2 for the non-beltline and beltline regions, respectively, are applied.

In order to determine how much to shift the P-T curves, an evaluation is performed using Tables 4-4a and 4-5a from NEDC-33178P-A. These tables define the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) Each component listed in these tables is ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) for each component. The required temperature is then determined by ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )), thereby resulting in the required T for the curve. As the upper vessel curve is initially based on the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

°° ° ° ° )) T-RTNDT, all resulting T values are compared to the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° )). The

((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) the upper vessel curve. The same method is applied for the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) curve. In this manner, it is assured that each curve bounds the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) that is represented.

For the BFNP Unit 2 upper vessel curve, the maximum T value for ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° )) from the method described above is ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° )). The initial required T-RTNDT for the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ));

this is then adjusted by the BFNP Unit 2-specific ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

°°°°°°°° ° ° ° ° )), resulting in ((° ° ° ° ° ° ° ° )). Comparing this to the other components bounded by the upper vessel curve, the limiting value is for the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). The required T-RTNDT for the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )), which is ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). It is seen that the resulting T required for the ((° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). As ((° ° ° ° ° ° ° ° )) is limiting, the BFNP Unit 2 upper vessel curve is based on an RTNDT of ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). As noted above, this calculation was performed for each component shown in Table 4-4a of NEDC-33178P-A; only the limiting case is presented here.

For the BFNP Unit 2 bottom head or CRD ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) the maximum T value from the method described above ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). The required T-RTNDT for the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° )); this is adjusted by the BFNP Unit 2-specific maximum ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° )), resulting in ((° ° ° ° ° ° ° ° )). Comparing this to the limiting value, the required T-RTNDT is

((° ° ° ° ° ° ° ° )), which is added to the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). It is seen that the resulting T required for the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). As ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° 6

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° ° ° ° ° ° ° ° ° ° ° ° ° ° )), the BFNP Unit 2 bottom head (CRD) curve is based on an ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° )). As noted above, this calculation was performed for each component shown in Table 4-5a of NEDC-33178P-A; only the limiting case is presented here.

Appendix H of NEDC-33178P-A contains the details of an analysis performed to determine the baseline requirement (non-shifted) for the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °° ° ° ° ° ° ° ° ° ° ° )). It can be seen in Section H.5 of Appendix H that the stresses developed in this finite element analysis demonstrated that the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )), resulting in a baseline non-shifted required T-RTNDT of ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )).

Therefore, considering the determination of the required shift from the paragraph above for ((° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° )), calculations for all components listed in Table 4-5a of NEDC-33178P-A were compared to the CRD T, which is ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) (where ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°° ° ° ° ° ° ° )) materials). Therefore, the shift for the bottom head ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )).

For BFNP Unit 2, the limiting surveillance material, (( )), was considered using (( )) as defined in Appendix I of Reference 6.2. This procedure was used because the target vessel material and the surveillance material ((

)).

For BFNP Unit 2, there are thickness discontinuities in the vessel: (1) between the bottom head upper and lower torus, and (2) between the bottom head torus and the support skirt attachment.

The 38 EFPY beltline curves are based on an ART of 161°F, and the 48 EFPY beltline curves are based on an ART of 175°F. Curves based on these temperatures bound the requirements due to the beltline thickness discontinuities.

The ART of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. RG 1.99, Revision 2 (RG 1.99) provides the methods for determining the ART. The basis for the difference in the margin terms in Tables B-4 and B-5 is due, in part, to the effective fluence associated with 38 and 48 EFPY. For many of the BFNP Unit 2 materials, the margin term is dependent on the RTNDT. This is consistent with Position 1.1 of RG 1.99, Revision 2, which provides the methodology for determining ART. The final paragraph of this section of the RG states that (standard deviation for RTNDT) is 17°F for plates and 28°F for welds, but that need not exceed 0.5*RTNDT. The BFNP Unit 2 ART calculation has incorporated the use of 0.5*RTNDT for all materials, where applicable. A reduced margin term was used for the (( )) as permitted by RG 1.99.

Similarly, a reduced margin term was used for the (( )) as permitted by RG 1.99.

The vessel beltline copper and nickel values were obtained from plant-specific vessel purchase order records, Certified Material Test Reports (CMTRs), or are values previously approved by the NRC, and remain unchanged from previous submittals.

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The pressure head for the beltline hydrostatic test curve (Curve A) for BFNP Unit 2 is

(( )). This is determined using the height of the vessel and the elevation of the bottom of active fuel. The full vessel pressure head is ((° ° ° ° ° ° ° ° ° ° ° ° )). This pressure is used in the P-T curve analysis, considered in the determination of KI for the bottom head curve as discussed in Sections 4.3.2.1.1 and 4.3.2.2.2 of the LTR.

The P-T curves for the non-beltline region were conservatively developed for a BWR ((° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) with nominal inside diameter of ((° ° ° ° ° ° ° ° ° ° ° ° ° )). As discussed in Section 4.3.2.1.1 of the LTR, if the plant-specific result of Equation 4-3 is greater than that of the

((° ° ° ° ° ° ° ° ° ° ° ° ° )) from Equation 4-2, the user is directed to perform a plant-specific evaluation.

The plant-specific BFNP Unit 2 geometry demonstrates that it is bounded by the Equation 4-2 result:

BFNP Unit 2: R / t1/2 = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

The analysis is therefore considered appropriate for BFNP Unit 2. The ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) was adapted to the conditions at BFNP Unit 2 using plant-specific RTNDT values for the RPV.

The peak RPV inside diameter (ID) fluence used in the P-T curve evaluation for BFNP Unit 2 at 48 EFPY is 1.93e18 n/cm2 and at 38 EFPY is 1.49e18 n/cm2. These values were calculated using methods that comply with the guidelines of RG1.190, (as discussed in Reference 6.1).

This fluence applies to the lower-intermediate shell plates. The fluence is adjusted for the lower plates, associated longitudinal welds, and the girth weld based upon the axial fluence distribution calculated as part of the fluence evaluation; hence, the peak ID surface fluence for these components is 1.56e18 n/cm2 for 48 EFPY and 1.20e18 n/cm2 for 38 EFPY. Similarly, the fluence is adjusted for the N16 nozzle based upon the axial fluence distribution; hence the peak ID surface fluence used for this component is 5.84e17 n/cm2 for 48 EFPY and 4.50e17 n/cm2 for 38 EFPY. Conservatism was removed from the fluence applied to the lower-intermediate shell axial welds using azimuthal locations. The maximum azimuthal location of these welds has a fluence of 1.32e18 n/cm2 for 48 EFPY and 1.01e18 n/cm2 for 38 EFPY.

The P-T curves for the heatup and cooldown 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 cooldown and the outer wall during heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4T location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness, KIr, at 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, well above the heatup/cooldown curve limits.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heatup and cooldown temperature rate of 100°F/hr for which the curves are 8

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public) applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram and the nozzle thermal cycle diagrams. For the hydrostatic pressure and leak test curve (Curve A), a coolant heatup and cooldown temperature rate of 15°F/hr must be maintained. The P-T limits and corresponding heatup/cooldown rates of either Curve A or B may be applied while achieving or recovering from test conditions. Curve A applies during pressure testing and when the limits of Curve B cannot be maintained.

As shown in Tables B-4 and B-5, the ((

)). The initial RTNDT for electroslag weld heat material is 23.1°F. The generic pressure test P-T curve is applied to the BFNP Unit 2 beltline curve by shifting the P vs. (T - RTNDT) values to reflect the ART values.

Using the fluence discussed above, the P-T curves are beltline limited for Curves A, B, and C, for 38 and 48 EFPY. For 38 EFPY, Curve A is beltline limited above 530 psig, Curve B is beltline limited between 290 and 313 psig and above 360 psig, and Curve C is beltline limited above 290 psig. For 48 EFPY, Curve A is beltline limited above 500 psig, Curve B is beltline limited between 270 and 313 psig and above 330 psig, and Curve C is beltline limited above 260 psig. For Curve C at 38 EFPY, the upper vessel region is bounding at pressures between 50 and 290 psig. For Curve C at 48 EFPY, the upper vessel region is bounding at pressures between 50 psig and 260 psig.

The N16 WLI nozzle is a partial penetration design similar to that shown in Figure 1 in Appendix J of the LTR, fabricated with ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). Reference to this status is contained in the BFNP UFSAR. Therefore, the evaluation is performed, consistent with the statement in Appendix J, ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). Appendix J of the LTR provides detailed results of an analysis performed for the WLI nozzle that provides the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °° ° ° ° ° ° ° ° ° ° ° ° )) a specific curve applicable for the WLI nozzle to ensure that this location is bounded in the P-T curves. The nozzle curve ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

for BFNP Unit 2 Curves A, B, or C. Sample calculations are provided in Appendix D.

6.0 References 6.1 Final Safety Evaluation Regarding Removal of Methodology Limitations for NEDC-32983P-A, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC NO. MC3788), November 17, 2005.

6.2 Final Safety Evaluation for Boiling Water Reactors Owners Group Licensing Topical Report NEDC-33178P, General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves (TAC NO. MD2693), April 27, 2009.

6.3 BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations, BWRVIP-135, Revision 2, Electric Power Research Institute (EPRI), Palo Alto, CA, October 2009 (EPRI Proprietary).

9

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public) 6.4 BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP)

Implementation Plan, BWRVIP-86, Revision 1, EPRI, Palo Alto, CA: September 2008. 1016575 (EPRI Proprietary).

6.5 Letter 2013-050, Bob Carter (EPRI) to Victor Schiavone (TVA), Evaluation of the Browns Ferry Unit 2 120° Surveillance Capsule Data, EPRI, Palo Alto, CA, April 10, 2013 (EPRI Proprietary).

10

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public) 1400 1300 INITIAL RTNDT VALUES ARE 23°F FOR BELTLINE, 1200 44°F FOR UPPER VESSEL, AND 1100 49°F FOR BOTTOM HEAD PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 900 38 138 800 HEATUP/COOLDOWN RATE OF COOLANT

< 15°F/HR 700 690 psig 600 530 psig ACCEPTABLE REGION OF OPERATION IS TO 500 THE RIGHT OF THE APPLICABLE CURVE 400 313 psig 300 BOTTOM HEAD 68°F UPPER VESSEL 200 FLANGE AND BELTLINE REGION LIMITS 83°F BOTTOM HEAD 100 CURVE 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 1. BFNP Unit 2 Composite Curve A Pressure Test P-T Curves Effective for up to 38 EFPY 11

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

B C 1400 INITIAL RTNDT VALUES 1300 ARE 23°F FOR BELTLINE, 44°F FOR UPPER 1200 VESSEL, AND 49°F FOR BOTTOM HEAD 1100 BELTLINE CURVES PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 38 138 900 HEATUP/COOLDOWN RATE OF COOLANT 800 < 100°F/HR FOR CURVES B&C 700 600 ACCEPTABLE REGION OF OPERATION IS TO THE RIGHT OF THE 500 480 psig APPLICABLE CURVE 420 psig 400 360 psig 313 psig 300 Bottom Head Curve B Composite Curve B 200 Composite Curve C 100 BOTTOM HEAD 68°F BOLTUP 83°F 0

0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 2. BFNP Unit 2 Composite Curve B Core Not Critical including Bottom Head and Curve C Core Critical P-T Curves Effective for up to 38 EFPY 12

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public) 1400 1300 INITIAL RTNDT VALUES ARE 23°F FOR BELTLINE, 1200 44°F FOR UPPER VESSEL, AND 1100 49°F FOR BOTTOM HEAD PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 900 48 152 800 HEATUP/COOLDOWN RATE OF COOLANT

< 15°F/HR 700 690 psig 600 ACCEPTABLE REGION 500 psig OF OPERATION IS TO 500 THE RIGHT OF THE APPLICABLE CURVE 400 313 psig 300 BOTTOM HEAD 68°F UPPER VESSEL 200 FLANGE AND BELTLINE REGION LIMITS 83°F BOTTOM HEAD 100 CURVE 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 3. BFNP Unit 2 Composite Curve A Pressure Test P-T Curves Effective for up to 48 EFPY 13

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

B C 1400 INITIAL RTNDT VALUES 1300 ARE 23°F FOR BELTLINE, 44°F FOR UPPER 1200 VESSEL, AND 49°F FOR BOTTOM HEAD 1100 BELTLINE CURVES PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig) 1000 ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 48 152 900 HEATUP/COOLDOWN RATE OF COOLANT 800 < 100°F/HR FOR CURVES B&C 700 600 ACCEPTABLE REGION OF OPERATION IS TO THE RIGHT OF THE 500 480 psig APPLICABLE CURVE 420 psig 400 313 psig 300 Bottom Head Curve B Composite Curve B 200 Composite Curve C 100 BOTTOM HEAD 68°F BOLTUP 83°F 0

0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 4. BFNP Unit 2 Composite Curve B Core Not Critical including Bottom Head and Curve C Core Critical P-T Curves Effective for up to 48 EFPY 14

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

Table 1 - BFNP Unit 2 Tabulation of Curves - 38 EFPY UPPER RPV UPPER RPV BOTTOM & BOTTOM & LIMITING HEAD BELTLINE HEAD BELTLINE 38 EFPY PRESSURE AT AT (PSIG)

CURVE A 38 EFPY CURVE B 38 EFPY CURVE C

(°F) CURVE A (°F) CURVE B (°F)

(°F) (°F) 0 68.0 83.1 68.0 83.1 83.1 10 68.0 83.1 68.0 83.1 83.1 20 68.0 83.1 68.0 83.1 83.1 30 68.0 83.1 68.0 83.1 83.1 40 68.0 83.1 68.0 83.1 83.1 50 68.0 83.1 68.0 83.1 83.1 60 68.0 83.1 68.0 83.1 84.0 70 68.0 83.1 68.0 83.1 91.2 80 68.0 83.1 68.0 83.1 97.2 90 68.0 83.1 68.0 83.1 102.3 100 68.0 83.1 68.0 83.1 106.8 110 68.0 83.1 68.0 83.1 110.9 120 68.0 83.1 68.0 83.1 114.7 130 68.0 83.1 68.0 83.1 118.2 140 68.0 83.1 68.0 83.1 121.4 150 68.0 83.1 68.0 84.2 124.2 160 68.0 83.1 68.0 86.9 126.9 170 68.0 83.1 68.0 89.5 129.5 180 68.0 83.1 68.0 91.9 131.9 190 68.0 83.1 68.0 94.2 134.2 200 68.0 83.1 68.0 96.3 136.3 210 68.0 83.1 68.0 98.3 138.3 220 68.0 83.1 68.0 100.3 140.3 230 68.0 83.1 68.0 102.1 142.1 240 68.0 83.1 68.0 103.9 143.9 250 68.0 83.1 68.0 105.6 145.6 260 68.0 83.1 68.0 107.2 147.2 270 68.0 83.1 68.0 108.8 148.8 280 68.0 83.1 68.0 110.3 150.3 290 68.0 83.1 68.0 111.8 151.8 300 68.0 83.1 68.0 117.4 157.4 310 68.0 83.1 68.0 122.5 162.5 313 68.0 83.1 68.0 123.8 163.8 313 68.0 113.1 68.0 143.1 200.6 15

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

UPPER RPV UPPER RPV BOTTOM & BOTTOM & LIMITING HEAD BELTLINE HEAD BELTLINE 38 EFPY PRESSURE AT AT (PSIG)

CURVE A 38 EFPY CURVE B 38 EFPY CURVE C

(°F) CURVE A (°F) CURVE B (°F)

(°F) (°F) 320 68.0 113.1 68.0 143.1 200.6 330 68.0 113.1 68.0 143.1 200.6 340 68.0 113.1 68.0 143.1 200.6 350 68.0 113.1 68.0 143.1 200.6 360 68.0 113.1 68.0 143.1 200.6 370 68.0 113.1 68.0 145.7 200.6 380 68.0 113.1 68.0 148.7 200.6 390 68.0 113.1 68.0 151.5 200.6 400 68.0 113.1 68.0 154.2 200.6 410 68.0 113.1 68.0 156.7 200.6 420 68.0 113.1 68.0 159.1 200.6 430 68.0 113.1 68.0 161.4 201.4 440 68.0 113.1 68.0 163.6 203.6 450 68.0 113.1 68.0 165.8 205.8 460 68.0 113.1 68.0 167.8 207.8 470 68.0 113.1 68.0 169.7 209.7 480 68.0 113.1 68.0 171.6 211.6 490 68.0 113.1 69.4 173.4 213.4 500 68.0 113.1 71.6 175.2 215.2 510 68.0 113.1 73.8 176.8 216.8 520 68.0 113.1 75.8 178.5 218.5 530 68.0 113.1 77.8 180.0 220.0 540 68.0 115.2 79.7 181.6 221.6 550 68.0 119.3 81.5 183.1 223.1 560 68.0 123.1 83.3 184.5 224.5 570 68.0 126.6 85.0 185.9 225.9 580 68.0 129.9 86.6 187.3 227.3 590 68.0 133.0 88.2 188.6 228.6 600 68.0 135.9 89.8 189.9 229.9 610 68.0 138.6 91.3 191.1 231.1 620 68.0 141.2 92.7 192.4 232.4 630 68.0 143.7 94.1 193.6 233.6 640 68.0 146.0 95.5 194.7 234.7 650 68.0 148.3 96.8 195.9 235.9 660 68.0 150.4 98.1 197.0 237.0 16

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

UPPER RPV UPPER RPV BOTTOM & BOTTOM & LIMITING HEAD BELTLINE HEAD BELTLINE 38 EFPY PRESSURE AT AT (PSIG)

CURVE A 38 EFPY CURVE B 38 EFPY CURVE C

(°F) CURVE A (°F) CURVE B (°F)

(°F) (°F) 670 68.0 152.5 99.4 198.1 238.1 680 68.0 154.5 100.7 199.2 239.2 690 68.0 156.4 101.9 200.2 240.2 700 69.2 158.2 103.0 201.2 241.2 710 70.7 160.0 104.2 202.2 242.2 720 72.1 161.7 105.3 203.2 243.2 730 73.5 163.3 106.4 204.2 244.2 740 74.8 164.9 107.5 205.1 245.1 750 76.1 166.5 108.6 206.1 246.1 760 77.4 168.0 109.6 207.0 247.0 770 78.6 169.5 110.6 207.9 247.9 780 79.8 170.9 111.6 208.8 248.8 790 81.0 172.3 112.6 209.6 249.6 800 82.2 173.6 113.5 210.5 250.5 810 83.3 174.9 114.5 211.3 251.3 820 84.4 176.2 115.4 212.2 252.2 830 85.5 177.4 116.3 213.0 253.0 840 86.5 178.6 117.2 213.8 253.8 850 87.6 179.8 118.0 214.5 254.5 860 88.6 181.0 118.9 215.3 255.3 870 89.6 182.1 119.7 216.1 256.1 880 90.5 183.2 120.6 216.8 256.8 890 91.5 184.3 121.4 217.6 257.6 900 92.4 185.4 122.2 218.3 258.3 910 93.4 186.4 123.0 219.0 259.0 920 94.3 187.4 123.7 219.7 259.7 930 95.1 188.4 124.5 220.4 260.4 940 96.0 189.4 125.3 221.1 261.1 950 96.9 190.3 126.0 221.8 261.8 960 97.7 191.3 126.7 222.5 262.5 970 98.6 192.2 127.5 223.1 263.1 980 99.4 193.1 128.2 223.8 263.8 990 100.2 194.0 128.9 224.4 264.4 1000 101.0 194.9 129.6 225.0 265.0 1010 101.7 195.7 130.2 225.7 265.7 17

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

UPPER RPV UPPER RPV BOTTOM & BOTTOM & LIMITING HEAD BELTLINE HEAD BELTLINE 38 EFPY PRESSURE AT AT (PSIG)

CURVE A 38 EFPY CURVE B 38 EFPY CURVE C

(°F) CURVE A (°F) CURVE B (°F)

(°F) (°F) 1015 102.1 196.1 130.6 226.0 266.0 1020 102.5 196.6 130.9 226.3 266.3 1030 103.3 197.4 131.6 226.9 266.9 1035 103.6 197.8 131.9 227.2 267.2 1040 104.0 198.2 132.2 227.5 267.5 1050 104.7 199.0 132.9 228.1 268.1 1055 105.1 199.4 133.2 228.4 268.4 1060 105.4 199.8 133.5 228.7 268.7 1070 106.2 200.6 134.1 229.3 269.3 1080 106.9 201.4 134.8 229.9 269.9 1090 107.6 202.1 135.4 230.4 270.4 1100 108.2 202.9 136.0 231.0 271.0 1105 108.6 203.2 136.3 231.3 271.3 1110 108.9 203.6 136.6 231.6 271.6 1120 109.6 204.3 137.2 232.1 272.1 1130 110.2 205.0 137.8 232.6 272.6 1140 110.9 205.7 138.3 233.2 273.2 1150 111.5 206.4 138.9 233.7 273.7 1160 112.1 207.1 139.5 234.2 274.2 1170 112.8 207.8 140.0 234.8 274.8 1180 113.4 208.4 140.6 235.3 275.3 1190 114.0 209.1 141.1 235.8 275.8 1200 114.6 209.7 141.7 236.3 276.3 1210 115.2 210.4 142.2 236.8 276.8 1220 115.8 211.0 142.8 237.3 277.3 1230 116.3 211.6 143.3 237.8 277.8 1240 116.9 212.2 143.8 238.3 278.3 1250 117.5 212.9 144.3 238.8 278.8 1260 118.0 213.5 144.8 239.2 279.2 1270 118.6 214.0 145.3 239.7 279.7 1280 119.1 214.6 145.8 240.2 280.2 1290 119.7 215.2 146.3 240.6 280.6 1300 120.2 215.8 146.8 241.1 281.1 1310 120.7 216.4 147.3 241.6 281.6 1320 121.3 216.9 147.8 242.0 282.0 18

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

UPPER RPV UPPER RPV BOTTOM & BOTTOM & LIMITING HEAD BELTLINE HEAD BELTLINE 38 EFPY PRESSURE AT AT (PSIG)

CURVE A 38 EFPY CURVE B 38 EFPY CURVE C

(°F) CURVE A (°F) CURVE B (°F)

(°F) (°F) 1330 121.8 217.5 148.2 242.5 282.5 1340 122.3 218.0 148.7 242.9 282.9 1350 122.8 218.6 149.2 243.3 283.3 1360 123.3 219.1 149.6 243.8 283.8 1370 123.8 219.6 150.1 244.2 284.2 1380 124.3 220.1 150.5 244.6 284.6 1390 124.8 220.7 151.0 245.1 285.1 1400 125.3 221.2 151.4 245.5 285.5 19

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

Table 2 - BFNP Unit 2 Tabulation of Curves - 48 EFPY UPPER RPV UPPER RPV BOTTOM & BOTTOM & LIMITING HEAD BELTLINE HEAD BELTLINE 48 EFPY PRESSURE AT AT (PSIG)

CURVE A 48 EFPY CURVE B 48 EFPY CURVE C

(°F) CURVE A (°F) CURVE B (°F)

(°F) (°F) 0 68.0 83.1 68.0 83.1 83.1 10 68.0 83.1 68.0 83.1 83.1 20 68.0 83.1 68.0 83.1 83.1 30 68.0 83.1 68.0 83.1 83.1 40 68.0 83.1 68.0 83.1 83.1 50 68.0 83.1 68.0 83.1 83.1 60 68.0 83.1 68.0 83.1 84.0 70 68.0 83.1 68.0 83.1 91.2 80 68.0 83.1 68.0 83.1 97.2 90 68.0 83.1 68.0 83.1 102.3 100 68.0 83.1 68.0 83.1 106.8 110 68.0 83.1 68.0 83.1 110.9 120 68.0 83.1 68.0 83.1 114.7 130 68.0 83.1 68.0 83.1 118.2 140 68.0 83.1 68.0 83.1 121.4 150 68.0 83.1 68.0 84.2 124.2 160 68.0 83.1 68.0 86.9 126.9 170 68.0 83.1 68.0 89.5 129.5 180 68.0 83.1 68.0 91.9 131.9 190 68.0 83.1 68.0 94.2 134.2 200 68.0 83.1 68.0 96.3 136.3 210 68.0 83.1 68.0 98.3 138.3 220 68.0 83.1 68.0 100.3 140.3 230 68.0 83.1 68.0 102.1 142.1 240 68.0 83.1 68.0 103.9 143.9 250 68.0 83.1 68.0 105.6 145.6 260 68.0 83.1 68.0 107.2 147.2 270 68.0 83.1 68.0 111.6 151.6 280 68.0 83.1 68.0 119.1 159.1 290 68.0 83.1 68.0 125.6 165.6 300 68.0 83.1 68.0 131.4 171.4 310 68.0 83.1 68.0 136.5 176.5 313 68.0 83.1 68.0 137.8 177.8 313 68.0 113.1 68.0 143.1 214.6 20

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

UPPER RPV UPPER RPV BOTTOM & BOTTOM & LIMITING HEAD BELTLINE HEAD BELTLINE 48 EFPY PRESSURE AT AT (PSIG)

CURVE A 48 EFPY CURVE B 48 EFPY CURVE C

(°F) CURVE A (°F) CURVE B (°F)

(°F) (°F) 320 68.0 113.1 68.0 143.1 214.6 330 68.0 113.1 68.0 145.5 214.6 340 68.0 113.1 68.0 149.4 214.6 350 68.0 113.1 68.0 153.1 214.6 360 68.0 113.1 68.0 156.5 214.6 370 68.0 113.1 68.0 159.7 214.6 380 68.0 113.1 68.0 162.7 214.6 390 68.0 113.1 68.0 165.5 214.6 400 68.0 113.1 68.0 168.2 214.6 410 68.0 113.1 68.0 170.7 214.6 420 68.0 113.1 68.0 173.1 214.6 430 68.0 113.1 68.0 175.4 215.4 440 68.0 113.1 68.0 177.6 217.6 450 68.0 113.1 68.0 179.8 219.8 460 68.0 113.1 68.0 181.8 221.8 470 68.0 113.1 68.0 183.7 223.7 480 68.0 113.1 68.0 185.6 225.6 490 68.0 113.1 69.4 187.4 227.4 500 68.0 113.1 71.6 189.2 229.2 510 68.0 114.5 73.8 190.8 230.8 520 68.0 119.9 75.8 192.5 232.5 530 68.0 124.8 77.8 194.0 234.0 540 68.0 129.2 79.7 195.6 235.6 550 68.0 133.3 81.5 197.1 237.1 560 68.0 137.1 83.3 198.5 238.5 570 68.0 140.6 85.0 199.9 239.9 580 68.0 143.9 86.6 201.3 241.3 590 68.0 147.0 88.2 202.6 242.6 600 68.0 149.9 89.8 203.9 243.9 610 68.0 152.6 91.3 205.1 245.1 620 68.0 155.2 92.7 206.4 246.4 630 68.0 157.7 94.1 207.6 247.6 640 68.0 160.0 95.5 208.7 248.7 650 68.0 162.3 96.8 209.9 249.9 660 68.0 164.4 98.1 211.0 251.0 21

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UPPER RPV UPPER RPV BOTTOM & BOTTOM & LIMITING HEAD BELTLINE HEAD BELTLINE 48 EFPY PRESSURE AT AT (PSIG)

CURVE A 48 EFPY CURVE B 48 EFPY CURVE C

(°F) CURVE A (°F) CURVE B (°F)

(°F) (°F) 670 68.0 166.5 99.4 212.1 252.1 680 68.0 168.5 100.7 213.2 253.2 690 68.0 170.4 101.9 214.2 254.2 700 69.2 172.2 103.0 215.2 255.2 710 70.7 174.0 104.2 216.2 256.2 720 72.1 175.7 105.3 217.2 257.2 730 73.5 177.3 106.4 218.2 258.2 740 74.8 178.9 107.5 219.1 259.1 750 76.1 180.5 108.6 220.1 260.1 760 77.4 182.0 109.6 221.0 261.0 770 78.6 183.5 110.6 221.9 261.9 780 79.8 184.9 111.6 222.8 262.8 790 81.0 186.3 112.6 223.6 263.6 800 82.2 187.6 113.5 224.5 264.5 810 83.3 188.9 114.5 225.3 265.3 820 84.4 190.2 115.4 226.2 266.2 830 85.5 191.4 116.3 227.0 267.0 840 86.5 192.6 117.2 227.8 267.8 850 87.6 193.8 118.0 228.5 268.5 860 88.6 195.0 118.9 229.3 269.3 870 89.6 196.1 119.7 230.1 270.1 880 90.5 197.2 120.6 230.8 270.8 890 91.5 198.3 121.4 231.6 271.6 900 92.4 199.4 122.2 232.3 272.3 910 93.4 200.4 123.0 233.0 273.0 920 94.3 201.4 123.7 233.7 273.7 930 95.1 202.4 124.5 234.4 274.4 940 96.0 203.4 125.3 235.1 275.1 950 96.9 204.3 126.0 235.8 275.8 960 97.7 205.3 126.7 236.5 276.5 970 98.6 206.2 127.5 237.1 277.1 980 99.4 207.1 128.2 237.8 277.8 990 100.2 208.0 128.9 238.4 278.4 1000 101.0 208.9 129.6 239.0 279.0 1010 101.7 209.7 130.2 239.7 279.7 22

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

UPPER RPV UPPER RPV BOTTOM & BOTTOM & LIMITING HEAD BELTLINE HEAD BELTLINE 48 EFPY PRESSURE AT AT (PSIG)

CURVE A 48 EFPY CURVE B 48 EFPY CURVE C

(°F) CURVE A (°F) CURVE B (°F)

(°F) (°F) 1015 102.1 210.1 130.6 240.0 280.0 1020 102.5 210.6 130.9 240.3 280.3 1030 103.3 211.4 131.6 240.9 280.9 1035 103.6 211.8 131.9 241.2 281.2 1040 104.0 212.2 132.2 241.5 281.5 1050 104.7 213.0 132.9 242.1 282.1 1055 105.1 213.4 133.2 242.4 282.4 1060 105.4 213.8 133.5 242.7 282.7 1070 106.2 214.6 134.1 243.3 283.3 1080 106.9 215.4 134.8 243.9 283.9 1090 107.6 216.1 135.4 244.4 284.4 1100 108.2 216.9 136.0 245.0 285.0 1105 108.6 217.2 136.3 245.3 285.3 1110 108.9 217.6 136.6 245.6 285.6 1120 109.6 218.3 137.2 246.1 286.1 1130 110.2 219.0 137.8 246.6 286.6 1140 110.9 219.7 138.3 247.2 287.2 1150 111.5 220.4 138.9 247.7 287.7 1160 112.1 221.1 139.5 248.2 288.2 1170 112.8 221.8 140.0 248.8 288.8 1180 113.4 222.4 140.6 249.3 289.3 1190 114.0 223.1 141.1 249.8 289.8 1200 114.6 223.7 141.7 250.3 290.3 1210 115.2 224.4 142.2 250.8 290.8 1220 115.8 225.0 142.8 251.3 291.3 1230 116.3 225.6 143.3 251.8 291.8 1240 116.9 226.2 143.8 252.3 292.3 1250 117.5 226.9 144.3 252.8 292.8 1260 118.0 227.5 144.8 253.2 293.2 1270 118.6 228.0 145.3 253.7 293.7 1280 119.1 228.6 145.8 254.2 294.2 1290 119.7 229.2 146.3 254.6 294.6 1300 120.2 229.8 146.8 255.1 295.1 1310 120.7 230.4 147.3 255.6 295.6 1320 121.3 230.9 147.8 256.0 296.0 23

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UPPER RPV UPPER RPV BOTTOM & BOTTOM & LIMITING HEAD BELTLINE HEAD BELTLINE 48 EFPY PRESSURE AT AT (PSIG)

CURVE A 48 EFPY CURVE B 48 EFPY CURVE C

(°F) CURVE A (°F) CURVE B (°F)

(°F) (°F) 1330 121.8 231.5 148.2 256.5 296.5 1340 122.3 232.0 148.7 256.9 296.9 1350 122.8 232.6 149.2 257.3 297.3 1360 123.3 233.1 149.6 257.8 297.8 1370 123.8 233.6 150.1 258.2 298.2 1380 124.3 234.1 150.5 258.6 298.6 1390 124.8 234.7 151.0 259.1 299.1 1400 125.3 235.2 151.4 259.5 299.5 24

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Appendix A. Reactor Vessel Material Surveillance Program Two of the BFNP Unit 2 surveillance capsules have been removed from the reactor vessel. One capsule was reconstituted and placed into the vessel. Therefore, one capsule remains that has been in the reactor vessel since plant startup, and one capsule remains that contains reconstituted test specimens and has been in place since the first capsule was removed and tested.

As described in the BFNP Unit 2 UFSAR Section 4.2.6, Inspection and Testing, the BWRVIP ISP will determine the removal schedule for the remaining BFNP Unit 2 surveillance capsules.

At this writing, the ISP capsule test plan shows the next BFNP Unit 2 capsule scheduled for removal and testing at 40 EFPY in 2026.

The BFNP Unit 2 material surveillance program is administered in accordance with the BWRVIP ISP. The ISP combines the United States BWR surveillance programs into a single integrated program. This program uses similar heats of materials in the surveillance programs of BWRs to represent the limiting materials in other vessels. It also adds data from the BWR Supplemental Surveillance Program (SSP). Per the BWRVIP ISP, BFNP Unit 2 is a host plant; the remaining surveillance capsules are slated for removal and testing.

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Appendix B.

BFNP Unit 2 Reactor Pressure Vessel P-T Curve Supporting Plant-Specific Information 26

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Note: The WLI nozzle centerline is at 366 elevation, at the top of active fuel.

Figure B-1: BFNP Unit 2 Reactor Pressure Vessel 27

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Table B-1: BFNP Unit 2 Initial RTNDT Values for RPV Plate and Flange Materials Drop Test Plate, Heat or Charpy Energy (T 50T-60) Weight RT NDT Component Temp Forging, Heat / Flux / Lot (ft-lb) (°F) NDT (°F)

(°F) or Weld

(°F)

PLATES & FORGINGS:

Top Head & Flange Shell Flange (MK 48) 48-127-2 ARZ 76 10 F 105 115 61 -20 10 10 Top Head Flange (MK 209) 209-127-2 AKU75 10 F 108 111 106 -20 10 10 Top Head Dollar (MK201) 201-122-1 B5524-2 10 P 42 50 43 -4 10 10 Top Head Side Plates (MK202) 202-127-5 C2426-2 10 P 53 80 79 -20 10 10 202-127-6 C2426-2 10 P 80 69 75 -20 10 10 202-127-7 C2426-3 10 P 91 71 98 -20 10 10 202-127-9 C1717-3 10 P 74 69 95 -20 10 10 202-127-10 C1717-3 10 P 93 51 90 -20 10 10 202-127-11 C1722-3 10 P 81 80 101 -20 10 10 Shell Courses Upper Shell Plates (MK 60) 6-127-11 C2559-2 10 P 31 60 54 18 10 18 6-127-21 C2791-1 10 P 63 29 54 22 10 22 6-127-22 C2660-1 10 P 67 70 67 -20 10 10 Transition Shell Plates (MK 16) 15-127-1 C2533-1 10 P 58 58 72 -20 10 10 15-127-3 C2533-3 10 P 57 67 47 -14 10 10 15-127-4 B5842-3 10 P 63 81 46 -12 10 10 Upper Intermediate Shell Plates (MK 59) 6-127-18 C2528-1 10 P 66 79 71 -20 10 10 6-127-23 C2463-2 10 P 69 60 72 -20 10 10 6-127-24 C2605-2 10 P 56 71 57 -20 10 10 Lower Intermediate Shell Plates (MK 58) 6-127-6 A0981-1 10 P 72 68 65 -20 -10 -10 6-127-16 C2467-1 10 P 65 81 74 -20 -10 -10 6-127-20 C2849-1 10 P 61 59 74 -20 -10 -10 Lower Shell Plates (MK 57) 6-127-14 C2467-2 10 P 59 78 65 -20 -20 -20 6-127-15 C2463-1 10 P 85 74 54 -20 -20 -20 6-127-17 C2460-2 10 P 59 60 40 0 -20 0 Bottom Head Bottom Head Dollar (MK1) 1-139-1 C2669-2 40 P 34 40 34 42 40 42 Bottom Head Upper Torus (MK 2) 2-139-1 B6747-1 40 P 78 80 75 10 40 40 2-139-2 B6747-1 40 P 60 54 77 10 40 40 2-139-3 B6776-2 40 P 90 88 92 10 40 40 2-139-4 B6776-2 40 P 53 64 50 10 40 40 2-127-11 C2369-1 40 P 81 80 88 10 40 40 2-127-12 C2369-1 40 P 101 101 100 10 40 40 Bottom Head Lower Torus (MK 4) 4-127-5 C2412-1 40 P 57 66 68 10 40 40 4-127-6 C2412-1 40 P 67 45 76 20 40 40 4-127-7 C2412-2 40 P 74 79 60 10 40 40 4-127-8 C2412-2 40 P 60 55 43 24 40 40 Note: Minimum Charpy values are provided for all materials.

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Table B-2: BFNP Unit 2 Initial RTNDT Values for RPV Nozzle and Weld Materials Drop Test Plate, Heat or Heat / Flux Charpy Energy (T 50T-60) Weight RT NDT Component Temp Forging,

/ Lot (ft-lb) (°F) NDT (°F)

(°F) or Weld

(°F)

Nozzles:

N1 Recirc Outlet Nozzle (MK 8) 8-127-3 E31VW 431H-3 40 F 86 81 100 10 40 40 8-127-4 E30VW 431H-4 40 F 84 98 97 10 40 40 N2 Recirc Inlet Nozzle (MK 7) 7-127-11 E25VW 433H-11 40 F 97 106 86 10 40 40 7-127-12 E25VW 433H-12 40 F 63 105 94 10 40 40 7-127-13 E25VW 433H-13 40 F 95 79 101 10 40 40 7-127-14 E25VW 433H-14 40 F 93 103 107 10 40 40 7-127-15 E25VW 433H-15 40 F 105 113 98 10 40 40 7-127-16 E25VW 433H-16 40 F 100 97 86 10 40 40 7-127-17 E25VW 433H-17 40 F 101 94 99 10 40 40 7-127-18 E25VW 433H-18 40 F 112 96 85 10 40 40 7-127-19 E25VW 433H-19 40 F 107 87 110 10 40 40 7-127-20 E25VW 433H-20 40 F 96 99 116 10 40 40 N3 Steam Outlet Nozzle (MK 14) 14-127-5 E26VW 435H-5 40 F 106 113 107 10 40 40 14-127-6 E26VW 435H-6 40 F 112 99 116 10 40 40 14-127-7 E26VW 435H-7 40 F 87 99 106 10 40 40 14-127-8 E26VW 435H-8 40 F 113 112 95 10 40 40 N4 Feedwater Nozzle (MK10) 10-127-7 E25VW 436H-7 40 F 108 94 112 10 40 40 10-127-8 E25VW 436H-8 40 F 112 105 113 10 40 40 10-127-9 E25VW 436H-9 40 F 103 112 93 10 40 40 10-127-10 E25VW 436H-10 40 F 89 78 91 10 40 40 10-127-11 E25VW 436H-11 40 F 112 104 119 10 40 40 10-127-12 E25VW 436H-12 40 F 111 94 94 10 40 40 N5 Core Spray Nozzle (MK 11) 11-139-1 EV9964 N7H6029B 40 F 38 42 44 34 0 34 11-127-3 E26VW 437H-3 40 F 73 96 87 10 40 40 N6 Top Head Instrumentation Nozzle (MK 206) 206-145-1& -2 BT2615-4 40 F 123 143 144 10 40 40 N7 Top Head Vent Nozzle (MK 204) 204-127-2 ZT3043-3 40 F 102 130 117 10 40 40 N8 Jet Pump Instrumentation (MK 19) 19-122-3 & -4 214484 40 F 37 35 23 65 40 65 N9 CRD HYD System Return Nozzle (MK 13) 13-127-2 E23VW 438H 40 F 120 112 114 10 40 40 N10 Core P & Liquid Control Nozzle (MK17) 17-139-1 ZT3043-1 40 F 155 154 156 10 40 40 N11, N12, N16 Instrumentation Nozzle (MK 12) Inconel 12-127-13 through 16 8601 12-139-11& -12 8601 N13,N14 High & Low Pressure Seal Leak (MK139)

F 40*

N15 Drain Nozzle (MK22) 22-139-1 7478 40 F 136 160 172 10 40 40 WELDS:

Cylindrical Shell Axial Welds Electroslag Welds ESW 23.1**

Girth Welds Shell 1 to Shell 2 (MK57 to MK58) D55733 -40**

Notes: Minimum Charpy values are provided for all materials.

• No NDT value is available on the CMTR; obtained from the purchase specification.

• • Weld initial RTNDT values were obtained from previously-approved submittals.

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Table B-3: BFNP Unit 2 Initial RTNDT Values for RPV Appurtenance and Bolting Materials Drop Test Heat or Charpy Energy (T 50T-60) Weight RT NDT Component Temp Heat / Flux / Lot (ft-lb) (°F) NDT (°F)

(°F)

(°F)

Miscellaneous Appurtenances:

Support Skirt (MK 24 )

24-127-5 through -8 A4846-5 40 109 96 104 10 40 40 Shroud Support (MK 51, MK52, MK53) Alloy 600 51-127-5 through 8 65952-5 52-127-3 and 4 65952-5 53-127-5 332273-4 53-127-10 through -12 & -14 56782-5 53-127-13 & -16 56782-6 53-127-15 56825-1 Steam Dryer Support Bracket (MK131) Stainless Steel 131 00431 Core Spray Bracket (MK132) Stainless Steel 132 3342230 Dryer Hold Down Bracket (MK133) 133 EV8446 40 94 110 113 10 40 40 Guide Rod Bracket (MK134) Stainless Steel 134 00431 Feedwater Sparger Brackets - MK135 Stainless Steel 135 00431 Stabilizer Bracket (MK 196) 196 C6458-1 10 60 59 56 -20 40 40 Surveillance Brackets (MK199 & MK200) Stainless Steel 199, 200 342633-2 Lifting Lugs (K210) 210 A1210-3 10 98 108 72 -20 0 0 CRD penetrations (MK101 - MK128) Alloy 600 101 through 128 Refueling Containment Skirt (MK71) 71-127-5 through -8 B7478-4B 10 110 89 102 -20 10 10 Test Charpy Energy Lateral LST Component Heat Temp (ft-lb) Expansion (°F)

(°F)

(mils)

STUDS:

Closure (MK61) 6730502 10 34 52 68 n/a 70 6780382 10 42 42 42 n/a 70 6720443 10 35 38 37 n/a 70 NUTS:

Closure (MK62) 6730502 10 34 52 68 n/a 70 BUSHINGS:

Closure (MK63) T3798 10 61 68 73 51 10 M2513 10 64 65 67 40 10 M3232 10 65 65 68 45 10 WASHERS:

Closure (MK64) 6790156 10 n/a n/a n/a n/a 70 Closure (MK64 and MK65) 6730502 10 34 52 68 n/a 70 Closure (MK64) 6780278 10 40 43 43 n/a 70 Notes: Minimum Charpy values are provided for all materials.

• No NDT value is available on the CMTR; obtained from the purchase specification.

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Table B-4: BFNP Unit 2 Adjusted Reference Temperatures for up to 38 EFPY Lower-Intermediate Plates Thickness in inches = 6.125 38 EFPY Peak I.D. fluence = 1.49E+18 n/cm2 38 EFPY Peak 1/4 T fluence = 1.03E+17 n/cm2 Lower Plates & Lower to Lower-Intermediate Girth Weld Thickness in inches = 6.125 38 EFPY Peak I.D. fluence = 1.20E+18 n/cm2 38 EFPY Peak 1/4 T fluence = 8.31E+17 n/cm2 Axial Welds Thickness in inches = 6.125 38 EFPY Peak I.D. fluence = 1.01E+18 n/cm2 38 EFPY Peak 1/4 T fluence = 6.99E+17 n/cm2 WLI Nozzle Thickness in inches = 6.125 38 EFPY Peak I.D. fluence = 4.05E+17 n/cm2 38 EFPY Peak 1/4 T fluence = 3.12E+17 n/cm2 38 38 38 Initial 1/4T EFPY EFPY EFPY Heat or Adjusted Margin Component %Cu %Ni CF RTNDT Fluence 1/4T I 1/4T 1/4T Heat/Lot CF °F

°F n/cm2 RTNDT Shift ART

°F °F °F PLATES:

Lower Shell 6-127-1 C2467-2 0.16 0.52 112 -20 8.31E+17 43 0 17 34 77 57 6-127-15 C2463-1 0.17 0.48 117 -20 8.31E+17 45 0 17 34 79 59 6-127-17 C2460-2 0.13 0.51 88 0 8.31E+17 34 0 17 34 67 67 Lower-Intermediate Shell 6-127-6 A0981-1 0.14 0.55 98 -10 1.03E+18 41 0 17 34 75 65 6-127-16 C2467-1 0.16 0.52 112 -10 1.03E+18 47 0 17 34 81 71 6-127-20 C2849-1 0.11 0.50 73 -10 1.03E+18 31 0 15 31 62 52 WELDS:

Axial Welds ESW - 0.24 0.37 141 23.1 6.99E+17 49 13 25 56 105 128 Lower to Lower-Intermediate Girth Weld D55733 0.09 0.65 117 -40 8.31E+17 45 0 22 45 89 49 31

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public) 38 38 38 Initial 1/4T EFPY EFPY EFPY Heat or Adjusted Margin Component %Cu %Ni CF RTNDT Fluence 1/4T I 1/4T 1/4T Heat/Lot CF °F

°F n/cm2 RTNDT Shift ART

°F °F °F BEST ESTIMATE CHEMISTRIES:

None - - - - - - - - - - - -

NOZZLES:

N16 WLI Forging ((° ° ° ° ° ° ° ° ° ° ° °°°° °°°° °°° ((° ° ° °)) 3.12E+17 ((° ° ° °° °° °° ° ° ° ° ° ))

Weld °°°°°°° ° ° ° ))

INTEGRATED SURVEILLANCE PROGRAM:

Plate(2,3) (( 1.03E+18 ((

Weld(4,5) )) 6.99E+17 ))

Notes:

(1) ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

(2) ((

))

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Table B-5: BFNP Unit 2 Adjusted Reference Temperatures for up to 48 EFPY Lower-Intermediate Plates Thickness in inches = 6.125 48 EFPY Peak I.D. fluence = 1.93E+18 n/cm2 48 EFPY Peak 1/4 T fluence = 1.34E+18 n/cm2 Lower Plates & Lower to Lower-Intermediate Girth Weld Thickness in inches = 6.125 48 EFPY Peak I.D. fluence = 1.56E+18 n/cm2 48 EFPY Peak 1/4 T fluence = 1.08+18 n/cm2 Axial Welds Thickness in inches = 6.125 48 EFPY Peak I.D. fluence = 1.32E+18 n/cm2 48 EFPY Peak 1/4 T fluence = 9.14E+17 n/cm2 WLI Nozzle Thickness in inches = 6.125 48 EFPY Peak I.D. fluence = 5.84E+17 n/cm2 48 EFPY Peak 1/4 T fluence = 4.04E+17 n/cm2 48 48 48 Initial 1/4T EFPY EFPY EFPY Heat or Adjusted Margin Component %Cu %Ni CF RTNDT Fluence 1/4T I 1/4T 1/4T Heat/Lot CF °F

°F n/cm2 RTNDT Shift ART

°F °F °F PLATES:

Lower Shell 6-127-14 C2467-2 0.16 0.52 112 -20 1.08E+18 48 0 17 34 82 62 6-127-15 C2463-1 0.17 0.48 117 -20 1.08E+18 51 0 17 34 85 65 6-127-17 C2460-2 0.13 0.51 88 0 1.08E+18 38 0 17 34 72 72 Lower-Intermediate Shell 6-127-6 A0981-1 0.14 0.55 98 -10 1.34E+18 47 0 17 34 81 71 6-127-16 C2467-1 0.16 0.52 112 -10 1.34E+18 53 0 17 34 87 77 6-127-20 C2849-1 0.11 0.50 73 -10 1.34E+18 35 0 17 34 69 59 WELDS:

Axial Welds ESW - 0.24 0.37 141 23.1 9.14E+17 56 13 28 62 118 141 Lower to Lower-Intermediate Girth Weld D55733 0.09 0.65 117 -40 1.08E+18 51 0 25 51 101 61 33

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public) 48 48 48 Adjuste Initial 1/4T EFPY EFPY EFPY Heat or Margin Component %Cu %Ni CF d RTNDT Fluence 1/4T I 1/4T 1/4T Heat/Lot °F CF °F n/cm2 RTNDT Shift ART

°F °F °F BEST ESTIMATE CHEMISTRIES:

None - - - - - - - - - - - -

NOZZLES:

N16 WLI Forging ((° ° ° ° ° ° ° ° ° ° ° °°°° °°°° °°° ((° ° ° ° )) 4.04E+17 ((° ° ° °° °° °° ° ° ° ° ° ))

Weld °°°°°°° ° ° ° ))

INTEGRATED SURVEILLANCE PROGRAM:

Plate(2,3) (( 1.34E+18 ((

Weld(4,5) )) 9.14E+17 ))

Notes:

(1) ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

(2) ((

))

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Table B-6: BFNP Unit 2 RPV Beltline P-T Curve Input Values for 48 EFPY A = (( ))°F Adjusted RTNDT = Initial RTNDT + Shift (Based on ART values in Table B-5)

Vessel Height H = 875.13 inches Bottom of Active Fuel Height B = 216.3 inches Vessel Radius (to base metal) R = 125.7 inches Minimum Vessel Thickness (without clad) t = 6.125 inches Note: The ART for 38 EFPY is (( )) as shown in Table B-4.

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Table B-7: BFNP Unit 2 Definition of RPV Beltline Region[1]

Elevation Component (inches from RPV 0)

Shell # 2 - Top of Active Fuel 366.3 Shell # 1 - Bottom of Active Fuel 216.3 Shell # 2 - Top of Extended Beltline Region (48 EFPY) 381.4 Shell # 1 - Bottom of Extended Beltline Region (48 EFPY) 205.0 Circumferential Weld Between Shell #2 and Shell #3 391.5 Centerline of Recirculation Outlet Nozzle in Shell # 1 161.5 Top of Recirculation Outlet Nozzle N1 in Shell # 1* 188.0 Centerline of Recirculation Inlet Nozzle N2 in Shell # 1 181.0 Top of Recirculation Inlet Nozzle N2 in Shell # 1* 193.5 Centerline of 2 WLI Nozzle in Shell # 2 366.0 Bottom of WLI Nozzle in Shell #2* 364.6 Note:

[1] The beltline region is defined as any location where the peak neutron fluence is expected to exceed or equal 1.0e17 n/cm2.

• The elevation identified as the top or bottom of the nozzle represents the nozzle cut-out. In all cases, the elevation considered represents the location where the fluence is greatest.

Based on the above, it is concluded that none of the BFNP Unit 2 reactor vessel plates, nozzles, or welds, other than those included in the ART Table, are in the beltline region.

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Appendix C.

BFNP Unit 2 Reactor Pressure Vessel P-T Curve Checklist 37

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Table C-1 provides a checklist that defines pertinent points of interest regarding the methods and information used in developing the BFNP Unit 2 PTLR. This table demonstrates that all important parameters have been addressed in accordance with the P-T curve LTR (Reference 6.2), and includes comments, resolutions, and clarifications as necessary.

Table C-1: BFNP Unit 2 Checklist Parameter Completed Comments/Resolutions/Clarifications Initial RTNDT Initial RTNDT has been determined for all The N16 WLI nozzle is considered vessel materials including plates, flanges, within the beltline region. This forging forgings, studs, nuts, bolts, welds. was ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°° Include explanation (including ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

methods/sources) of any exceptions, are considered.

resolution of discrepant data (e.g.,

deviation from originally reported values). All other information remains unchanged from previous submittals.

Appendix B contains tables of all Initial RTNDT values.

Has any non-plant-specific initial RTNDT Plate heat (( )) information information (e.g., ISP, comparison to other was obtained from the ISP database.

plant) been used? This material is ((

)) to the target vessel plate material, however, ((

)). In accordance with the ISP guidance; this data was

((

)).

Weld heat (( )) information was obtained from the ISP database.

This material is ((

)) and, in accordance with ISP guidance, this data was ((

)).

If deviation from the P-T curve LTR No deviations from the P-T curve LTR process occurred, sufficient supporting process were applied.

information has been included (e.g.,

Charpy V-Notch data used to determine an Initial RTNDT).

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Parameter Completed Comments/Resolutions/Clarifications All previously published Initial RTNDT RVID was reviewed for the beltline values from sources such as the Generic materials; all initial RTNDT values agree; Letter (GL) 88-01, Reactor Vessel Integrity no further review was performed Database (RVID), and FSAR have been reviewed.

Adjusted Reference Temperature Sigma I (I, standard deviation for Initial Sigma I is equal to 0°F for all materials RTNDT) is 0°F unless the RTNDT was except the weld heats. The Electroslag obtained from a source other than CMTRs. weld material uses I of 13°F, consistent If I is not equal to 0°F, reference/basis has with previous NRC submittals.

been provided.

Sigma (, standard deviation for RTNDT) is determined per RG 1.99, Revision 2 Chemistry has been determined for all The N16 WLI nozzle is ((° ° ° ° ° ° ° ° ° ° ° ° ° ° vessel beltline materials including plates, °°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°° forgings (if applicable), and welds. °°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°°

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) is Include explanation (including considered.

methods/sources) of any exceptions, resolution of discrepant data (e.g., ISP data is obtained from BWRVIP-135, deviation from originally reported values). Revision 2.

No deviations from previously reported values.

Non-plant-specific chemistry information Plate heat (( )) has been (e.g., ISP, comparison to other plant) used evaluated using best estimate chemistry has been adequately defined and described. from the ISP.

For any deviation from the P-T curve LTR No deviations from the P-T curve LTR process, sufficient information has been process.

included.

All previously published chemistry values RVID was reviewed; all chemistry from sources such as the GL88-01, RVID, values agree; no further review was and FSAR have been reviewed. performed The fluence used for determination of ART and any extended beltline region was obtained using an NRC-approved methodology.

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Parameter Completed Comments/Resolutions/Clarifications The fluence calculation provides an axial distribution to allow determination of the vessel elevations that experience fluence of 1.0e17 n/cm2 both above and below active fuel.

The fluence calculation provides an axial The fluence calculation also provided an distribution to allow determination of the azimuthal distribution for the axial weld fluence for intermediate locations such as locations in the lower-intermediate shell.

the beltline girth weld (if applicable) or for any nozzles within the beltline region.

All materials within the elevation range where the vessel experiences a fluence 1.0e17 n/cm2 have been included in the ART calculation. All initial RTNDT and chemistry information is available or explained.

Discontinuities The discontinuity comparison has been There are no deviations.

performed as described in Section 4.3.2.1 of the P-T curve LTR. Any deviations have been explained.

Discontinuities requiring additional All discontinuities are bounded by either components (such as nozzles) to be the Upper Vessel, Bottom Head, or considered part of the beltline have been Beltline curve; those bounded by the adequately described. It is clear which Upper Vessel and/or Bottom Head are in curve is used to bound each discontinuity. accordance with Tables 4-4a and 4-5a of the LTR.

Appendix G of the P-T curve LTR The thickness discontinuity evaluation describes the process for considering a demonstrates that no additional thickness discontinuity, both beltline and adjustment is required; the curves bound non-beltline. If there is a discontinuity in the discontinuity stresses.

the vessel that requires such an evaluation, the evaluation was performed. The affected curve was adjusted to bound the discontinuity, if required.

Appendix H of the P-T curve LTR defines the basis for the CRD Penetration curve discontinuity and the appropriate transient application. The plant-specific evaluation bounds the requirements of Appendix H.

40

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

Appendix J of the P-T curve LTR defines the basis for the WLI Nozzle curve discontinuity and the appropriate transient application. The plant-specific evaluation bounds the requirements of Appendix J.

41

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

Appendix D. Sample P-T Curve Calculations 42

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

Beltline WLI Nozzle Pressure Test (Curve A) for 48 EFPY KI for the discontinuity is determined considering the KI obtained from Table 7 of Appendix J (for hydrotest). For 1070 psig, this KI is ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) as follows:

KI = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

T-RTNDT is calculated in the following manner:

T-RTNDT = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

The ART is added to T-RTNDT to obtain the required T:

T = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

This temperature is not obvious from the P-T curve as it is bounded by the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° )).

Core Not Critical (Curve B) for 48 EFPY KI for the discontinuity is determined considering the KI obtained from Table 5 of Appendix J.

((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

The transient used for the WLI nozzle, defined in Appendix J, is used in determination of KI.

The total KI is therefore:

KI = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

T-RTNDT is calculated in the following manner:

T-RTNDT = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

The ART is added to T-RTNDT to obtain the required T:

T = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

This temperature is not obvious from the P-T curve as it is bounded by the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° )).

43

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

Correction Factor The total stress for the WLI exceeds the yield stress; therefore, the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according the following equation based on the assumptions and recommendation of Welding Research Council (WRC) Bulletin 175 as shown in Equation 4-7 of Reference 6.2.

R = [ys - pm + ((total - ys)/30)]/(total - pm)

Applied to the WLI:

R = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ))

44

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

Beltline Calculations Excluding WLI Nozzle Pressure Test (Curve A) at 1070 psig for 48 EFPY The limiting beltline material is the bounding component for Curve A; therefore, a sample calculation for this material, not including the WLI nozzle is provided for 1070 psig.

The limiting ART applied to the beltline P-T curves is (( )) for the ((

)), which is also in the (( )).

Pressure is calculated to include hydrostatic pressure for a full vessel:

P = 1070 psig + (H - B)*0.0361 psi/inch (H=vessel height; B=elevation of bottom of active fuel)

= 1070 + (875.13 - 216.3)

• 0.0361

= 1094 psig Pressure Stress:

= PR/t (P=pressure; R=vessel radius; t=vessel thickness)

= 1.094

• 125.7 / 6.125

= 22.45 ksi Mm = 0.926t

= 0.9266.125

= 2.29 The stress intensity factor, KIt, is calculated as described in Section 4.3.2.2.4 of the LTR, except that G is 15°F/hr instead of 100°F/hr.

Mt = 0.2914, from ASME Appendix G, Figure G-2214-1 T = GC2 / 2 G = coolant heatup/cooldown rate of 15°F/hr C = minimum vessel thickness including clad = 6.125+0.1875=6.313=0.526 ft

= thermal diffusivity at 550°F = 0.354 ft2/hr

= (15*(0.526)2) / (2*0.354)

= 5.86°F KIt = Mt

• T

= 0.2914

• 5.86

= 1.71 45

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

KIm =

• Mm

= 22.45

• 2.29

= 51.4 T-RTNDT = ln[(1.5*KIm + KIt - 33.2)/20.734]/0.02

= ln[(1.5*51.4 + 1.71 - 33.2)/20.734]/0.02

= 39.4°F T is calculated by adding the ART:

T = 39.4 + (( ))

= (( )) for P = 1070 psig at 48 EFPY This temperature represents the limiting point on Curve A and is cited as 214.6°F in Table 2 of this document.

Core Not Critical (Curve B) at 1070 psig for 48 EFPY The WLI nozzle ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) component at any point in the pressure range; therefore, a sample calculation for the limiting beltline material, which bounds the WLI nozzle, is provided for 1070 psig.

As discussed above and in Section 5.0 and Table B-5, the limiting ART applied to the beltline Curve B is (( )) for ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )).

The T term is calculated as shown above for the Pressure Test case, but the temperature rate change is 100°F/hr instead of 15°F/hr. Therefore, T equals 39°F.

P = 1070 psig + (H - B)*0.0361 psi/inch (H=vessel height; B=elevation of bottom of active fuel)

= 1070 + (875.13 - 216.3)

• 0.0361

= 1094 psig Pressure Stress:

= PR/t (P=pressure; R=vessel radius; t=vessel thickness)

= 1.094

• 125.7 / 6.125

22.45 ksi KIm

• Mm

= 22.45

• 2.29

= 51.41 KIt = Mt

• T (for the 100°F/hr case)

= 0.2914

• 39 46

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

= 11.36 T-RTNDT = ln[(2.0*KIm + KIt - 33.2)/20.734]/0.02

= ln[(2.0*51.41 + 11.36 - 33.2)/20.734]/0.02

= 68.12°F T is calculated by adding the ART:

T = 68.1 + (( ))

= (( )) for P = 1070 psig at 48 EFPY This temperature represents the limiting point on Curve B and is cited in Table 2 as 243.3°F.

47

NEDO-33854 Revision 0 Non-Proprietary Information-Class I (Public)

FW Nozzle Calculations An evaluation was performed for the FW nozzle as described in Section 4.3.2.1.3 of the LTR.

The first part of the evaluation is as described earlier, where it is assured that the limiting component that is represented by the upper vessel nozzle curve is bounded by the ((° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° )). A second evaluation was performed using the BFNP Unit 2-specific FW nozzle dimensions; this evaluation is shown below to demonstrate that the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) curve is applicable to BFNP Unit 2:

Vessel radius to base metal, Rv ((° ° ° ° ° ° ° ° ° ° ° ° Vessel thickness, tv °°°°°°°°°°°° Vessel pressure, Pv °°°°°°°°° Pressure stress = PR/t = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) °°°°°°°°°° Dead Weight + Thermal Restricted Free End stress °°°°°°°°° Total Stress = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) ° ° ° ° ° ° ° ° ° ° ° ° ))

The factor F (a/rn) from Figure A5-1 of PVRC Recommendations on Toughness Requirements for Ferritic Materials, Welding Research Council Bulletin 175, August 1972 (WRC-175) is determined where:

a = 1/4 (tn2 + tv2) 1/2 ((° ° ° ° ° ° ° ° ° ° ° tn = thickness of nozzle °°°°°°°°°°° tv = thickness of vessel °°°°°°°°°°°° rn = apparent radius of nozzle = ri + 0.29*rc °°°°°°°°°° ri = actual inner radius of nozzle °°°°°°°°°° rc = nozzle radius (nozzle corner radius) ° ° ° ° ° ° ° ° ° ° ° ° ° ))

Therefore, a/rn = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). The value F (a/rn), taken from Figure A5-1 of WRC-175 for ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). Including the safety factor of 1.5, the stress intensity factor, KI, is 1.5 (a)1/2

• F(a/rn):

BFNP Unit 2 Plant-Specific Nominal KI = 1.5 * ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ))

A detailed upper vessel example calculation for core not critical conditions is provided in Section 4.3.2.1.4 of the LTR. Section 4.3.2.1.3 of the LTR presents the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

°°°°°°°°°°°°°°°°° ° ° ° ° ° ° ° ° )) for the FW nozzle evaluation upon which the baseline non-shifted upper vessel P-T curve is based. It can be seen that the nominal KI from this BFNP Unit 2 evaluation is ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). Therefore, it has been shown that the nominal KI for the BFNP Unit 2-specific FW nozzle is less than the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) KI, demonstrating applicability of the FW nozzle curve for BFNP Unit 2.

48

ENCLOSURE 4 Enclosed are the affidavits supporting the request to withhold proprietary information (included in Enclosure 2) from the public.

GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Peter M. Yandow, state as follows:

(1) I am the Vice President, Nuclear Plant Projects/Services Licensing, Regulatory Affairs, of GE-Hitachi Nuclear Energy Americas LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in GEH proprietary report, NEDC-33854P, Tennessee Valley Authority Browns Ferry Nuclear Plant Unit 2 Pressure and Temperature Limits Report (PTLR) up to 38 and 48 Effective Full-Power Years, dated April 2014. The GEH proprietary information in NEDC-33854P is identified by a dotted underline inside double square brackets. ((This sentence is an example.{3})) In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

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

Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F.2d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F.2d 1280 (D.C. Cir. 1983).

(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:

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

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

c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d. Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, Page 1 of 3

GE-Hitachi Nuclear Energy Americas LLC and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7).

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

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

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains the detailed GEH methodology for pressure-temperature curve analysis for the GEH Boiling Water Reactor (BWR). These methods, techniques, and data along with their application to the design, modification, and analyses associated with the pressure-temperature curves were achieved at a significant cost to GEH.

The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience databases that constitute a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its Page 2 of 3

GE-Hitachi Nuclear Energy Americas LLC competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

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

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

Executed on this 7th day of April 2014.

Peter M. Yandow Vice President, Nuclear Plant Projects/Services Licensing, Regulatory Affairs GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Rd.

Wilmington, NC 28401 Peter.Yandow@ge.com Page 3 of 3

2014-047 BWR Vessel & Internals Project (BWRVIP)

April 4, 2014 Victor D. Schiavone Tennessee Valley Authority Browns Ferry Nuclear Plant P.O. Box 2000, Decatur, AL 35609 Mail Stop: SAB 1C-BFN

Subject:

Transmittal of EPRI Proprietary Affidavit to the NRC The purpose of this letter is to transmit proprietary affidavit for transmittal of the following document to the NRC:

Report by GE Hitachi Nuclear Energy, NEDC-33854P, Revision 0, dated April 2014 and entitled, Tennessee Valley Authority Browns Ferry Nuclear Plant Unit 2, Pressure and Temperature Limits Report (PTLR) up to 38 and 48 Effective Full Power Years Please note that the enclosed document contains EPRI proprietary information. A letter requesting that the report be withheld from public disclosure and an affidavit describing the basis for withholding this information are provided as Attachment 1.

If you have any questions on this subject, please contact me by telephone at 704-502-6440 or by e-mail at amcgehee@epri.com Sincerely, Andrew McGehee EPRI, BWRVIP Program Manager

-=*-

~~1211 ELECTRIC POWER RESEARCH INSTITUTE Attachment 1 Kurt Edsinger Director, PWR &

BWR Materials April 3, 2014 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Request for Withholding of the following Proprietary Information Included in:

Report by GE Hitachi Nuclear Energy, NEDC-33854P, Revision 0, dated April2014 and entitled, "Tennessee Valley Authority Browns Ferry Nuclear Plant Unit 2, Pressure and Temperature Limits Report (PTLR) up to 38 and 48 Effective Full Power Years" To Whom It May Concern:

This is a request under 10 C.F.R. §2.390(a)(4) that the U.S. Nuclear Regulatory Commission ("NRC") withhold from public disclosure certain identified proprietary information, which is owned by Electric Power Research Institute, Inc. ("EPRI") contained in the report that is described in the enclosed Affidavit and referenced above

("Report").

Tennessee Valley Authority ('TVA") desires to disclose the Report to NRC and EPRI desires NRC to handle the Proprietary Information contained in the Report in confidence and only for the purpose of assisting NRC review TVA's submittal. The Proprietary Information is not to be divulged to anyone outside of the NRC or to any of its contractors, nor shall any copies be made of the Proprietary Information provided herein. EPRI welcomes any discussions and/or questions relating to the information enclosed.

If you have any questions about the legal aspects of this request for withholding, please do not hesitate to contact me at (704) 704-595-2732. Questions on the content of the Report should be directed to Andy McGehee of EPRI at (704) 502-6440.

Sincerely, Together ... Shaping the Future of Electricity PALO ALTO OFFICE 3420 Hillview Avenue, Palo Alto, CA 94304*1395 USA

• 650.855.2000

• Customer Service 800.313.3774

• www.epri.com

~~~II ELECTRIC POWER t=l-lc;;;;;; RESEARCH INSTITUTE AFFIDAVIT RE: Request for Withholding of the Following Proprietary Information Included In:

Report by GE Hitachi Nuclear Energy, NEDC-33854P, Revision 0, dated April2014 and entitled, "Tennessee Valley Authority Browns Ferry Nuclear Plant Unit 2, Pressure and Temperature Limits Report (PTLR) up to 38 and 48 Effective Full Power Years" I, Kurt Edsinger, being duly sworn, depose and state as follows:

I am the Director of PWR and BWR Materials at Electric Power Research Institute, Inc. whose principal office is located at 3420 Hillview Avenue, Palo Alto, CA ("EPRI") and I have been specifically delegated responsibility for the above-listed report that contains EPRI Proprietary Information that is sought under this Affidavit to be withheld "Proprietary Information". I am authorized to apply to the U.S. Nuclear Regulatory Commission ("NRC") for the withholding of the Proprietary Information on behalf of EPRI.

EPRI Proprietary Information is identified in the above referenced report by a solid underline inside double square brackets. An example of such identification is as follows:

HJhis sentence is an example.{El))

Tables containing EPRI Proprietary Information are identified with double square brackets before and after the object. In each case, the superscript notation {El refers to this affidavit as the basis for the proprietary determination.

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

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

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

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

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

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

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

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

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

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

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

Executed at 3420 Hillview Avenue, Palo Alto, CA. being the premises and place of business of Electric Power

• zb~

Kurt Edsinger (State of California) before me on this 34£._day of iJ~&.£/ , 20/~ by

-,t:-+6444--6-4::4-~~4-'!~......,...------' proved to me on the basis cltsat1sfactory ev1dence to be llie peffinn~::

Signature~ ~ * (Seal)

My Commission Expires~ay of ~ 20#1;

• BERTE A. DAHL COMM. # 1926383 l~

II) NOTARY PUBLIC* CALIFORNIA ~

) SANTA CLARA COUNTY

'l M~ COMII. EXP. MAR. 20, 20151