NRC 2013-0005, License Amendment Request 252, Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

License Amendment Request 252, Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

ML13016A028
Person / Time
Site:	Point Beach
Issue date:	01/15/2013
From:	Meyer L Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2013-0005
Download: ML13016A028 (23)

Text

ENERGY@

7 POINT BEACH January 15,2013 NRC 2013-0005 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant Svstem (RCS)

Pressure and Temperature Limits Report (PTLR)

Pursuant to 10 CFR 50.90, NextEra Energy Point Beach, LLC (NextEra) hereby requests to amend renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2, respectively. The proposed amendments would revise the PBNP Technical Specifications (TS) to allow the use of two new methodologies; Framatome ANP Topical Report BAW-2308, Revisions I-A and 2-A, "Initial RTNDTof Linde 80 Weld Materials,"

and Westinghouse Owners Group (WOG) WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." The revision would add BAW-2308, Revisions 1-A and 2-A and WCAP-14040-A, Revision 4, as approved methodologies to TS 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," for determining RCS pressure-temperature (PT) limits.

In addition, pursuant to 10 CFR 50.12 and 10 CFR 50.60(b), NextEra requests an exemption to portions of the following regulations:

(1) 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," which sets forth fracture toughness requirements for protection against pressurized thermal shock (PTS); and (2) 10 CFR 50, Appendix G, "Fracture Toughness Requirements," which sets forth fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. provides a detailed description and analysis of the proposed changes. to Enclosure 1 provides the annotated TS pages showing the proposed changes. to Enclosure 1 provides clean TS pages showing the proposed changes. provides the Exemption Request to portions of 10 CFR 50.61 and 10 CFR 50, Appendix G.

NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241

Document Control Desk Page 2 Approval of the proposed amendment is requested by January 1,2014. NextEra will implement the amendment within 180 days of Commission Approval.

This letter contains no new Regulatory Commitments and no revisions to existing Regulatory Commitments.

The proposed TS changes have been reviewed by the Plant Operations Review Committee.

In accordance with 10 CFR 50.91, a copy of this letter is being provided to the designated Wisconsin Official.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on January 15,2013.

Very truly yours, NextEra Energy Point Beach, LLC Enclosures cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

EVALUATION OF PROPOSED CHANGES 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes

3.0 TECHNICAL EVALUATION

3.1 Evaluation of the Analytical Methods Used to Develop PT Limit Curves 3.2 Low Temperature Overpressure Protection (LTOP) 3.3 Conclusion

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory RequirementsICriteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 Conclusions

6.0 REFERENCES

ATTACHMENTS:

1. Proposed Technical Specification Changes (Mark-up)

2. Proposed Technical Specification Changes (Clean) 10 pages follow

1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, NextEra Energy Point Beach, LLC (NextEra) hereby requests to amend renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2, respectively. The proposed amendment would revise Technical Specification (TS) 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," to allow the use of two new methodologies for determining RCS pressure and temperature (PT) limits, as described by the following reports:

Framatome ANP Topical Report BAW-2308, Revisions I - A and 2-A, "Initial RTNDT of Linde 80 Weld Materials," and Westinghouse Owners Group (WOG), WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

In addition, the proposed amendment would require exemption, pursuant to 10 CFR 50.12 and 10 CFR 50.60(b), to portions of 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," and 10 CFR 50, Appendix G, "Fracture Toughness Requirements." Enclosure 2 provides the Exemption Request to portions of 10 CFR 50.61 and 10 CFR 50, Appendix G.

2.0 DETAILED DESCRIPTION Currently, TS 5.6.5 references the following NRC Safety Evaluations (SEs), which describe the approved PTLR methodologies for PBNP:

Point Beach Nuclear Plant, Units 1 and 2 - Exemption from the Requirements of 10 CFR 50.60 (TAC Nos. MA9680 and MA9681), dated October 6,2000 (Reference 1);

Point Beach Nuclear Plant, Units 1 and 2 -Acceptance of Methodology for Referencing Pressure Temperature Limits Report (TAC Nos. MA8459 and MA8460), dated July 23, 2001 (Reference 2); and, Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments Re: Reactor Coolant System Pressure and Temperature Limits Report Technical Specification 5.6.5 (TAC Nos. MD3800 and MD3801), dated October 18,2007 (Reference 3).

The proposed amendment would allow use of a new method, the Master Curve Methodology, as described in BAW-2308, Revision 1-A and 2-A, for determining the adjusted reference nil-ductility temperature (RTNDT).The proposed amendment would also allow use of the WOG methodology for the development of PT limit curves as described in WCAP-14040-A, Revision 4.

Framatome ANP Topical Report BAW-2308, Revisions I-A and 2-A, were approved for referencing in plant-specific license amendments in SEs dated August 4, 2005 (Reference 4) and March 24, 2008 (Reference 5), respectively. WCAP-14040-A, Revision 4, was approved by an SE dated February 27, 2004 (Reference 6). The proposed amendment would revise TS 5.6.5 to add the NRC SE approving the use of the above methodologies for determining RCS PT limits. Approval of these methodologies will allow future changes to the PTLR to be Page 1 of 10

performed by the licensee pursuant to 10 CFR 50.59, provided the methodology approved by the NRC is used to develop these PTLR changes.

The following Condition is stated in the SE for WCAP-14040-A, Revision 3 (Reference 6):

Ticensees who wish to use WCAP- 14040, Revision 3, as their PTLR methodology must provide additional information to address the methodology requirements discussed in provision 2 in the table of Attachment 1 to GL 96-03 related to the RPV material surveillance program. "

The Condition pertains to the Reactor Vessel Material Surveillance Program, which shall comply with Appendix H to 10 CFR 50. PBNP participates in the Master Integrated Reactor Vessel Surveillance Program, which is described in Pressurized Water Reactor Owners Group (PWROG) Topical Report BAW-1543(NP), Revision 4, Supplement 6-A. The NRC approved Topical Report BAW-1543(NP) for referencing in plant-specific applications via an SE dated June 28,2007 (Reference 7).

2.1 Proposed Chanqes Replace:

"The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC letters dated October 6, 2000, July 23, 2001, and October 18, 2007."

With:

"The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC letters dated October 6, 2000, July 23,2001, October 18, 2007, and [NRC SE Date]."

Basis for the change: The proposed change adds the NRC SE for this License Amendment Request as an approved analytical method to determine RCS PT limits.

3.0 TECHNICAL EVALUATION

3.1 Evaluation of the Analvtical Methods Used to Develop PT Limit Curves The analytical methods to be used to determine the revised RCS PT limits and low temperature overpressure protection (LTOP) limits for PBNP are contained in the NRC-approved WOG Topical Report WCAP-14040-A, Revision 4; the NRC-approved Framatome ANP Topical Report BAW-2308, Revisions I - A and 2-A; the NRC-accepted American Society for Mechanical Engineers (ASME) Code Cases N-588 and N-640; the NRC-approved methodology documented in WCAP-16083-NP-A, Revision 0, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," and in Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."

WOG Topical Report WCAP-14040-A, Revision 4 In order to implement the PTLR, the analytical methods used to develop the RCS PT limits must be consistent with those previously reviewed and approved by the NRC and must be referenced in the Administrative Controls section of the Technical Specifications. The purpose of WCAP-14040-A, Revision 4, is to provide the current Westinghouse methodology for developing the RCS heatup and cooldown curves and Cold Overpressure Mitigating System (COMS) setpoints. WCAP-14040-A, Revision 4, was approved by an NRC SE dated February 27, 2004 (Reference 6).

WCAP-14040-A, Revision 4, does not provide all of the methodologies which can be used to develop RCS heatup and cooldown curves and COMS setpoints, but rather methodologies that can be referenced by licensees when approved by the NRC to license the PTLR concept.

The closure head and outlet nozzle regions of the reactor vessel do not receive significant irradiation during reactor operation; therefore, the material properties do not experience a significant change over plant life. Consequently, there are no requirements to monitor their reference transition temperatures. The methods which determine the maximum acceptable pressure for these regions are described within WCAP-14040-A, Revision 4, as modified by the use of ASME Code Cases N-588 and N-640.

Since the reactor pressure vessel (RPV) beltline region does receive significant irradiation during reactor operation, the region experiences a significant change in RTNDT over the life of the plant. The methodologies described in RG 1.99, Revision 2, are used in the analyses to account for the effects of neutron embrittlement on the materials used in the RPV beltline region. The RG describes the methodologies used to calculate a material's adjusted reference temperature (ART). When credible surveillance data from the reactor is not available, then the following equation is used:

ART = Initial RTNDT + ARTNDT+ Margin The "ARTND<term is an adjustment in the reference temperature caused by irradiation, and the "Margin" term is added to provide conservatism in the values of the ART. RG 1.99, Revision 2, contains the methodologies to calculate the ARTNDTand Margin terms.

The proposed amendment uses the BAW-2308, Revisions I - A and 2-A method to determine the initial RTNDT values for the RPV beltline region welds rather than the methodology described within WCAP-14040, Revision 4, which is used to evaluate the other beltline components.

The PT limit curves and LTOP limits will use, in part, the methodologies described in WCAP-14040, Revision 4, with exception of the following:

1) The fluence values used in the development of the PTLR will be calculated using the NRC-approved methodology documented in WCAP-16083-NP-A, Revision 0. The NRC approved the use of the FERRET code in the PBNP, Units 1 and 2 PTLR, via an SE dated October 18,2007 (Reference 3). This approved methodology follows the guidance and meets the requirements of Regulatory Guide 1.I 90, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 8).

2) The KIccritical stress intensities are used in place of the KI, critical stress intensities.

This methodology is taken from approved ASME Code Case N-641 (which covers Code Cases N-640 and N-588).

3) PT limit curves will be generated with the most limiting circumferential weld ART value, in conjunction with Code Case N-588.

Framatome ANP Topical Report BAW-2308, Revisions 1-A and 2-A This topical report was developed to provide an alternate method for determining the initial RTNDT for Linde 80 beltline region welds in Babcock and Wilcox fabricated RPVs. The alternative methodology was based on brittle-to-ductile transition range fracture toughness test data of the weld materials, in accordance with American Standard for Testing and Materials (ASTM) Standard Test Method E 1921, and using ASME Code Case N-629. This methodology would be used in lieu of the nil-ductility reference temperature parameter specified in ASME Section Ill, Paragraph NB-2331. The use of ASME Section Ill, Paragraph NB-2331, is specified in 10 CFR 50, Appendix G(II)(D)(i), and its use is also described in RG 1.99, Revision 2. Revision 1-A of Topical Report BAW-2308 was approved by the NRC via an SE dated August 4, 2005 (Reference 4). Revision 2-A, which supplements Revision 1-A of Topical Report BAW-2308, was approved by the NRC via an SE dated March 24,2008 (Reference 5).

The SE for Revision 1-A of Topical Report BAW-2308 contained six Conditions and Limitations.

Licensees had to satisfy four of these Conditions and Limitations in order to use the alternate methodology. The first three Conditions and Limitations ltems describe performance requirements for use of the methodology. These Conditions and Limitations have been satisfied, as described in Enclosure 2. The fourth Condition and Limitation Item requires the request of an exemption, pursuant to 10 CFR 50.12 and 10 CFR 50.60(b), to portions of 10 CFR 50.61 and 10 CFR 50, Appendix G. Enclosure 2 provides the Exemption Request to portions of 10 CFR 50.61 and 10 CFR 50, Appendix G.

Topical Report BAW-2308 was revised to incorporate additional information from the Pressurized Water Reactor Owner's Group (PWROG) to satisfy the remaining two Conditions and Limitations ltems contained in the SE for Revision 1-A. The SE for Revision 2-A required licensees to address Conditions and Limitations ltems one through four contained in the SE for Revision 1-A.

Renulatorv Guide 1.99, Revision 2 RG 1.99, Revision 2, contains a methodology for determining a material's ART to account for neutron embrittlement. The use of the RG provides an acceptable means of ensuring the requirements of 10 CFR 50, Appendix G are satisfied.

Since credible surveillance data for PBNP is not available, the ART was calculated by use of the following RG equation.

ART = Initial RTNDT + ARTNDT+ Margin Since the RG provides the current methodology for the determination of a material's ART, no further evaluation of the RG is required.

3.2 Low Temperature Overpressure Protection (LTOP)

As part of developing a revision to the PBNP PT limit curves, the LTOP limits will also be reanalyzed. The LTOP analysis will use the same methodologies that will be used to develop the PT limit curves, the methodologies described in the NRC-approved WOG Topical Report WCAP-14040-A, Revision 4; the NRC-approved Framatome ANP Topical Report BAW-2308, Revisions I-A and 2-A; the NRC-accepted ASME Code Cases N-588 and N-640; the NRC-approved methodologies documented in WCAP-16083-NP-A. Revision 0; and RG 1.99, Revision 2.

Since not all of the methodologies are currently approved by the NRC for use at the PBNP, the results of the revised LTOP analysis will not be implemented until this License Amendment Request and Exemption Request are approved by the NRC.

3.3 Conclusion The revised PBNP PT limit curves will be developed using the NRC-approved WOG Topical Report WCAP-14040-A, Revision 4; RG 1.99, Revision 2; and four alternate methodologies.

The alternate methodologies are the NRC-approved Framatome ANP Topical Report BAW-2308, Revisions I-A and 2-A; the NRC-approved methodology documented in WCAP-16083-NP-A, Revision 0; and the NRC-accepted ASME Code Cases N-588 and N-640.

The LTOP reanalysis will also be developed using the aforementioned methodologies. Since the alternate methodologies are either approved or accepted for use by the NRC, the alternate methodologies provide an acceptable means of satisfying 10 CFR 50, Appendix G, which governs the development of PT limit curves and LTOP limits.

TS 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),"

requires revision to reflect the use of the above mentioned alternative methodologies. Upon NRC approval of this License Amendment Request and Exemption Request, a revised PTLR will be submitted to the NRC in accordance with TS 5.6.5.

4.0 REGULATORY EVALUATION

4.1 Applicable Requlatorv Requirementsicriteria Point Beach Nuclear Plant (PBNP) was licensed prior to the 1971 publication of 10 CFR 50, Appendix A, General Design Criteria (GDC) (ML003674718). As such, PBNP is not licensed to Appendix A GDCs. PBNP Final Safety Analysis Report (FSAR) Section 1.3 lists the plant-specific GDCs to which the plant was licensed. The PBNP GDCs are similar in content to the draft GDCs proposed for public comment in 1967. The following discussion addresses the proposed changes with respect to meeting the requirements of the applicable draft design criteria to which PBNP is licensed.

PBNP GDC 9 - Reactor Coolant Pressure Boundary: The reactor coolant pressure boundary shall be designed, fabricated, and constructed so as to have an exceedingly low probability of gross rupture or significant uncontrolled leakage throughout its design lifetime.

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PBNP GDC 34 - Pressure Boundary Rapid Propagation Failure Prevention: The reactor coolant pressure boundary shall be designed and operated to reduce to an acceptable level the probability of rapidly propagating type failures. Consideration is given (a) to the provisions for control over service temperature and irradiation effects which may require operational restrictions, (b) to the design and construction of the reactor pressure vessel in accordance with applicable codes, including those which establish requirements for absorption of energy within the elastic strain energy range and for absorption of energy by plastic deformation and (c) to the design and construction of reactor coolant pressure boundary piping and equipment in accordance with applicable codes.

Additionally, the NRC's regulatory requirements related to the content of Technical Specifications (TS) are set forth in 10 CFR 50.36. Specifically, 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether an L C 0 is required to be included in the TS. The criteria are:

Criterion 1 - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3 - A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4 - A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The proposed amendment would revise TS 5.6.5 to add the NRC SE for this License Amendment Request as an approved analytical method for determining RCS PT limits. PBNP GDCs 9 and 34 continue to be met, as RPV integrity is assured using the Master Curve fracture toughness data. The TS described in this LAR also meet Criterion 2 of 10 CFR 50.36(c)(2)(ii).

In accordance with Conditions and Limitations Item four of the SE for Revision 1-A of Topical Report BAW-2308 (Reference 4), Licensees requesting approval to utilize the methodology of Topical Report BAW-2308, Revision I-A, must request exemption, pursuant to 10 CFR 50.12 and 10 CFR 50.60(b), to portions of 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Thermal Shock Events" and 10 CFR 50, Appendix G, "Fracture Toughness Requirements." The requested exemption would allow use of a different method, the Master Curve Methodology, as described in Framatome ANP Topical Report BAW-2308, Revisions 1-A and 2-A, "Initial RTNDTof Linde 80 Weld Materials," for determining the adjusted RTNDT.

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4.2 Precedent The proposed amendment would allow use of a new method, as described in Framatome ANP Topical Report BAW-2308, Revisions I - A and 2-A, "Initial RTNDT of Linde 80 Weld Materials,"

for determining the adjusted reference nil-ductility temperature (RTNDT).The NRC approved the use of the BAW-2308, Revision I-A and 2-A methodology for Davis-Besse Nuclear Power Station, Unit 1, by an SE dated January 28,201 1 (Reference 9).

The proposed amendment would also allow use of Westinghouse Owners Group (WOG),

WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," for the development of pressure and temperature limit curves. The NRC previously approved the WCAP-14040-A, Revision 4 methodology for the development of pressure and temperature limit curves at Vogtle Electric Generating Plant, Units 1 and 2, by an SE dated March 28, 2005 (Reference 1O),

and Comanche Peak Steam Electric Station, Units 1 and 2, by an SE dated February 22, 2007 (Reference 11).

4.3 No Siqnificant Hazards Consideration Determination The proposed amendment would allow use of a new method, the Master Curve Methodology, as described in Framatome ANP Topical Report BAW-2308, Revisions 1-A and 2-A, "Initial RTNDTof Linde 80 Weld Materials," for determining the adjusted reference nil-ductility temperature (FITNDT).The proposed amendment would also allow use of the Westinghouse Owners Group (WOG) Topical Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," for the development of Reactor Coolant System (RCS) pressure and temperature (PT) limit curves.

The proposed amendments would revise Technical Specification (TS) 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," to add the Master Curve Methodology and WCAP-14040-A, Revision 4, as approved methodologies.

NextEra has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92(c) as discussed below:

1.0 Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or the manner in which the plant is operated and maintained. The proposed change does not alter or prevent the ability of structures, systems or components from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits.

There will be no adverse change to normal plant operating parameters, engineered safety feature actuation setpoints, accident mitigation capabilities, or accident analysis assumptions or inputs. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

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

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

Response: No The proposed change does not impose any new or different requirements or eliminate any existing requirements. The proposed change is consistent with the current safety analysis assumptions and current plant operating practice. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. Equipment important to safety will continue to operate as designed. The change does not result in any event previously deemed incredible being made credible. The change does not result in adverse conditions or result in any increase in the challenges to safety systems.

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

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

Response: No The proposed change does not alter safety limits, limiting safety system settings, or limiting conditions for operation. The setpoints at which protective actions are initiated are not altered by the proposed change. There are no new or significant changes to the initial conditions contributing to accident severity or consequences. The proposed amendment will not otherwise affect the plant protective boundaries, will not cause a release of fission products to the public, nor will it degrade the performance of any other structures, systems or components important to safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, NextEra concludes that the proposed amendment does not involve a 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.

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

5.0 ENVIRONMENTAL CONSIDERATION

The proposed amendment is confined to (i) changes to surety, insurance, and/or indemnity requirements, or (ii) changes to recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(lO). 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) NRC letter to Nuclear Management Company, LLC, dated October 6,2000, Point Beach Nuclear Plant, Units 1 and 2 - Exemption from the Requirements of 10 CFR 50.60 (TAC Nos. MA9680 and MA9681) (ML003758475)

(2) NRC letter to Nuclear Management Company, LLC, dated July 23,2001, Point Beach Nuclear Plant, Units 1 and 2 -Acceptance of Methodology for Reference Pressure Temperature Limits Report (TAC Nos. MA8459 and MA8460) (ML011900127)

(3) NRC letter to FPLE-Point Beach, dated October 18, 2007, Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments RE: Reactor Coolant System Pressure and Temperature Limits Report Technical Specification 5.6.5 (TAC Nos. MD3800 and MD3801) (ML072180204)

(4) NRC letter to Framatome ANP, dated August 4, 2005, Final Safety Evaluation for Topical Report BAW-2308, Revision 1, "Initial FITNDT of Linde 80 Weld Materials" (TAC No. MB6636) (ML052070408)

(5) NRC letter to Westinghouse Electric Company, dated March 24, 2008, Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR)

BAW-2308, Revision 2, "Initial RTNDT of Linde 80 Weld Materials" (TAC No. MD4241)

(ML080770349)

(6) NRC letter to Westinghouse Electric Company, dated February 27, 2004, Final Safety Evaluation for Topical Report WCAP-14040, Revision 3, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (TAC No. MB5754) (ML040620297)

(7) NRC letter to Westinghouse Electric Company, dated June 28, 2007, Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR)

BAW-1543(NP), Revision 4, Supplement 6, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" (TAC No. MC9608) (ML071770640)

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(8) U.S. Nuclear Regulatory Commission, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Regulatory Guide 1.I 90, dated March 2001 (ML010890301)

(9) NRC letter to FirstEnergy Nuclear Operating Company, dated January 28, 201 1, Davis-Besse Nuclear Power Station, Unit 1 - lssuance of Amendment Re: Request to Incorporate the Use of Alternative Methodologies for the Development of Reactor Pressure Vessel Pressure-Temperature Limit Curves (TAC No. ME1127)

(MLI 03610148)

(10) NRC letter to Southern Nuclear Operating Company, Inc., dated March 28, 2005, Vogtle Electric Generating Plant, Units 1 and 2 - lssuance of Exemption and Amendments Re: Request to Revise Technical Specifications and Pressure Temperature Limits Report and Relocate the Cold Overpressure Protection System (COPS) Arming Temperature (TAC Nos. MC2225, MC2226, MC2227, MC2228, MC3090, and MC3091)

(ML05690228)

(11) NRC letter to TXU Power, dated February 22, 2007, Comanche Peak Steam Electric Station, Units 1 and 2 - lssuance of Amendments Re: Revise Technical Specification 5.6.6 on Reactor Coolant System Pressure and Temperature Limits Report (TAC Nos. MC9500 and MC9501) (ML070320823)

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ENCLOSURE 1 ATTACHMENT 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP) 1 page follows

Reporting Requirements 5.6

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

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC 5.6.5 Reactor Coolant Svstem (RCS) PRESSURE AND TEMPERATURE LIMITS

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, LTOP enabling, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

(1) L C 0 3.4.3, "RCS Pressure and Temperature (PIT) Limits" (2) L C 0 3.4.6, "RCS Loops-MODE 4" (3) L C 0 3.4.7, "RCS Loops-MODE 5, Loops Filled" (4) L C 0 3.4.10, "Pressurizer Safety Valves" (5) L C 0 3.4.12, "Low Temperature Overpressure Protection (LTOP)"

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC Letters dated October 6, 2000, July 23, 2001, &October 18, 2007, and INRC SE Datel.

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

5.6.6 PAM Report When a report is required by Condition B or F of L C 0 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Point Beach 5.6-5 Unit I- Amendment No. 234 Unit 2 - Amendment No. 229 1

ENCLOSURE I ATTACHMENT 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

PROPOSED TECHNICAL SPECIFICATION CHANGES (CLEAN) 1 page follows

Reporting Requirements 5.6

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

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC 5.6.5 Reactor Coolant Svstem (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, LTOP enabling, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

(1) L C 0 3.4.3, "RCS Pressure and Temperature (PIT) Limits" (2) L C 0 3.4.6, "RCS Loops-MODE 4" (3) L C 0 3.4.7, "RCS Loops-MODE 5, Loops Filled" (4) L C 0 3.4.10, "Pressurizer Safety Valves" (5) L C 0 3.4.12, "Low Temperature Overpressure Protection (LTOP)"

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC Letters dated October 6, 2000, July 23, 2001, October 18, 2007, and

[NRC SE Date].

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

5.6.6 PAM Report When a report is required by Condition B or F of L C 0 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Point Beach 5.6-5 Unit 1 - Amendment No.

Unit 2 - Amendment No. 1

ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

REQUEST FOR EXEMPTION

1.0 INTRODUCTION

In accordance with the provisions of 10 CFR 50.12 and 10 CFR 50.60(b), NextEra Energy Point Beach, LLC (NextEra) is submitting a request for exemption from certain requirements of 10 CFR 50, Appendix G, "Fracture Toughness Requirements," and 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," The requested exemption would allow use of an alternate method, the Master Curve Methodology, as described in Framatome ANP Topical Report BAW-2308, Revisions I - A and 2-A, "Initial RTNDTof Linde 80 Weld Materials," for determining the initial, unirradiated material reference temperatures of the Linde 80 weld materials present in the reactor pressure vessel (RPV) beltline region of Point Beach Nuclear Plant (PBNP), Units 1 and 2.

2.0 BACKGROUND

10 CFR 50.61 (a)(5) and 10 CFR 50, Appendix G(II)(D)(i), require that the pre-service or unirradiated condition reference nil-ductility temperature (RTNDT)be evaluated according to the procedures in the American Society for Mechanical Engineers (ASME) Code, Section Ill, Paragraph NB-2331, which requires Charpy V-notch impact tests and drop weight tests.

Topical Report BAW-2308, Revisions 1-A and 2-A, provide an NRC-approved alternate method for determining the initial, unirradiated material reference temperatures of the Linde 80 weld materials present in the beltline region of the PBNP RPV. BAW-2308, Revisions I - A and 2-A, were approved by the NRC for referencing in plant-specific license amendments in NRC Safety Evaluations (SEs) dated August 4, 2005 (Reference I ) and March 24, 2008 (Reference 2),

respectively. BAW-2308, Revision 2-A, is a supplement to Revision 1-A, and incorporated additional test data and a re-evaluation of the reference temperature, To, determination, as requested by the NRC in the SE for Revision I-A. As stated in the SE for Revision 2-A of BAW-2308, the Conditions and Limitations, Items ( I ) through (4), contained in the SE for Revision I-A of BAW-2308 must still be addressed in all future plant-specific applications referencing Topical Report BAW-2308, Revisions 1-A and 2-A.

Page 1 of 6

Conditions and Limitations ltem (4) contained in the SE for Revision I-A of BAW-2308 states:

Any licensee who wants to utilize the methodology of TR BA W-2308, Revision 1 as outlined in items (I) through (3) above, must request an exemption, per 10 CFR 50.12, from the requirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61 to do so. As part of a licensee's exemption request, the NRC staff expects that the licensee will also submit information which demonstrates what values the licensee proposes to use for ARTNDT and the margin term for each Linde 80 weld in its RPV through the end of its facility's current operating license.

Conditions and Limitations ltem (1) contained in the SE for Revision 1-A of BAW-2308 provides criteria associated with the use of the NRC-accepted values of initial (unirradiated) reference temperature, IRTTo,and the corresponding uncertainty term, ol, to define the initial heat-specific or generic properties of its facility's Linde 80 welds.

Conditions and Limitations ltem (2) contained in the SE for Revision 1-A of BAW-2308 requires that a minimum chemistry factor of 167.0°F be applied when the methodology of Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," is used to assess the shift in nil-ductility transition temperature due to irradiation.

Conditions and Limitations ltem (3) contained in the SE for Revision I-A of BAW-2308 requires that a value of o~ = 28.0°F be used to determine the margin term, as defined in BAW-2308, Revision 1-A, and RG 1.99, Revision 2.

NextEra is planning to develop and implement revised pressure-temperaturelimit curves and low pressure overpressure protection (LTOP) system limits for operation beyond 35.9 Effective Full Power Years and desires to use the appropriate BAW-2308 initial RTNDT values in the revision. Hence, exemption from certain requirements of 10 CFR 50, Appendix G is required.

3.0 PROPOSED EXEMPTION The exemption requested by NextEra addresses portions of the following regulations:

(1) 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Thermal Shock Events," which sets forth fracture toughness requirements for protection against pressurized thermal shock (PTS); and (2) 10 CFR 50, Appendix G, "Fracture Toughness Requirements," which sets forth fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the system may be subjected over its service lifetime.

The exemption from Appendix G to 10 CFR 50 is to replace the required use of the existing Charpy V-notch and drop-weight-based methodology with the use of an alternate methodology that incorporates the use of fracture toughness test data for evaluating the integrity of the Linde 80 weld materials present in RPV beltline region of PBNP, Units 1 and 2. The alternate methodology employs direct fracture toughness testing per the Master Curve methodology based on use of American Standard for Testing and Materials (ASTM) Standard Test Method E 1921 (1997 and 2002 editions), and ASME Code Case N-629. The exemption is Page 2 of 6

required since 10 CFR 50, Appendix G, requires that for the pre-service or unirradiated condition, RTNDT be evaluated according to the procedures in the ASME Code, Section Ill, Paragraph NB-2331.

The exemption from 10 CFR 50.61 is to use an alternate methodology to allow the use of fracture toughness test data for evaluating the integrity of the Linde 80 weld materials present in the PBNP RPV beltline regions, based on the use of the ASTM Standard Test Method E 1921 (1997 and 2002 editions), and ASME Code Case N-629. The exemption is required because the methodology for evaluating RPV material fracture toughness in 10 CFR 50.61 requires that the pre-service or unirradiated condition be evaluated according to the procedures in the ASME Code, Section Ill, Paragraph NB-2331.

Additionally, the SE for BAW-2308, Revision I-A, concludes that an exemption is required to address issues related to 10 CFR 50.61 inasmuch as the BAW-2308, Revision 1-A, methodology, as modified and approved by the NRC staff, represents a significant change to the 10 CFR 50.61 methodology for determining the RTpTsvalue for Linde 80 weld material. The changes in the BAW-2308, Revision 1-A, methodology with respect to the 10 CFR 50.61 methodology, include the requirements for use of a minimum chemistry factor of 167°F and a value of o~= 28.0°F for Linde 80 weld materials.

10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that: I ) the exemption is authorized by law; 2) the exemption will not result in an undue risk to public health and safety; 3) the exemption is consistent with the common defense and security; and 4) special circumstances, as defined in 10 CFR 50.12(a)(2) are present. The requested exemption to allow the use of BAW-2308, Revisions 1-A and 2-A, as the basis for the Linde 80 weld material initial properties at the PBNP satisfy these requirements, as described below.

1. The requested exemption is authorized bv law.

No law exists which precludes the activities covered by this exemption request.

10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50, Appendix G, when an exemption is granted by the Commission under 10 CFR 50.12.

In addition, 10 CFR 50.61 permits other methods for use in determining the initial material properties provided such methods are approved by the Director, Office of Nuclear Reactor Regulation.

2. The requested exemption does not present an undue risk to the ~ u b l i chealth and safetv.

The proposed material initial properties basis described in BAW-2308, Revisions 1-A and 2-A, represent an NRC-approved methodology for establishing lRTTovalues for Linde 80 welds. BAW-2308, Revisions 1-A and 2-A, includes conservatisms to ensure that use of the proposed initial material properties basis does not increase the probability of occurrence or the consequences of an accident at the PBNP and will not create the possibility for a new or different type of accident that could pose a risk to public health and safety.

The use of this proposed approach ensures that the intent of the requirements specified in 10 CFR 50, Appendix G, and 10 CFR 50.61, are satisfied.

Page 3 of 6

The requested exemption is consistent with the NRC requirements specified in the SE for the approved BAW-2308, Revisions 1-A and 2-A; consequently, the exemption does not present an undue risk to the public health and safety.

3. The requested exemption will not endanuer the common defense and securitv.

The requested exemption is specifically concerned with RPV material properties and is consistent with NRC requirements specified in the SE for the approved BAW-2308, Revisions I-A and 2-A. Consequently, the requested exemption will not endanger the common defense and security.

4. Special circumstances are present which necessitate the request for an exemption to the requlations of 10 CFR 50, Appendix G, and 10 CFR 50.61.

Pursuant to 10 CFR 50.12(a)(2), the NRC will not consider granting an exemption to the regulations unless special circumstances are present. The requested exemption meets the special circumstances of 10 CFR 50.12(a)(2)(ii), since application of these regulations in this particular circumstance is not necessary to achieve the underlying purpose of the regulations.

The underlying purpose of 10 CFR 50, Appendix G, and 10 CFR 50.61, is to protect the integrity of the reactor coolant pressure boundary by ensuring that each reactor vessel material has adequate fracture toughness. Application of ASME Code, Section Ill, Paragraph NB-2331, in the determination of initial material properties was developed based on the level of knowledge existing in the early 1970's concerning RPV materials.

Since the early 1970's, the level of knowledge concerning these topics has greatly expanded. This increased knowledge level permits relaxation of the ASME Code, Section Ill, Paragraph NB-2331 requirements, via application of BAW-2308, Revisions 1-A and 2-A, while maintaining the underlying purpose of the ASME Code and NRC regulations to ensure an acceptable margin of safety is maintained.

The proposed change in reactor vessel material initial properties will continue to satisfy the intent of 10 CFR 50, Appendix G, and 10 CFR 50.61, thus justifying the exemption request.

Issuance of an exemption from the criteria of these regulations to permit the use of BAW-2308, Revisions I-A and 2-A, for the PBNP will not compromise the safe operation of the reactor, and will ensure that RPV integrity is maintained.

4.0 LlNDE 80 WELD VALUES As described in the NRC SE for BAW-2308, Revision I-A, Conditions and Limitations Item (4),

the licensee is required to provide the values proposed for ARTNDTand the margin term for each Linde 80 weld in its RPV through the end of its facility's current operating license. Table 1 provides these values.

Page 4 of 6

Table 1: PTS Summary Based on 53 EFPY Fluences and Linde 80 Weld Material Properties per BAW-2308, Revisions 1-A and 2-A Facility: PBNP, Unit 1 Vessel Manufacturer: B&W Plate and Weld Thickness (witl 3ut cladding)

-. 6.5", without clad I Lt$)z:: I I

Component Description %cu  % ~ i ID Fluence X

(I 019)

CF Determining CF Initial RTNDT (" F) 00 Margin ARTNDT (OF)

ART

("F)

%-T ART

("F) 36-T ART

("F) 122P237 Nozzle Belt Forging 0.1 1 1 0.82 1 0.38 1 77.0 I Table 1 50 1

1 1 88.0 Table I 1 A981 1-1 I lntermediateShell Plate 0.20 0.06 5.09 79.3 Surveillance 1 1 C1423-1 I Lower Shell Plate 0.12 0.07 4.65 55.3 35.8 Table Surveillance 1

1 8T1762 Nozzle Belt to Intermed.

(SA-1426) 0.19 0.57 0.38 167.0 Table -48.6 Shell Circ Weld (100%)

1PO815 (SA-812) lntermediate Shell Long Seam (ID 27%) 0.17 1 0.52 1 3.33 1 167.0 1 Table 1 -48.6 1PO661 (SA-775) lntermediate Shell Long Seam (OD 73%)

0.17 1 0.64 1 3.33 1 167.0 1 Table 1 -48.6 I 71249 (SA-1101) 61782 I Intermed. to Lower Shell Circ. Weld (100%)

Lower Shell Long Seam 0.23 1 0.59 1 4.54 1 167.6 167.0 1 Table Table 1 -53.5

-58.5 (SA-847) 0.23 0.52 3.14 (100%) 163.3 Surveillance -5 Facility: PBNP, Unit 2 Vessel Manufacturer: B&W and CE Plate and Weld Thickness (without claddinal

.,, 6.5". without

- clad I 1 u Component Description %Cu ID Fluence Method Of Initial RTNDT 8

%-T ART A-T ART

%Ni CF Determining DO on Margin ART (XI oi9) ("F) (OF) (OF) ("F)

CF (OF) 123V352 Nozzle Belt Forging 0.11 0.73 0.52 76 Table 40 0 17 34 62.2 136 130 113 1 123V500 1 lntermediate Shell Forging 0.09 0.70 5.07 58 Table 40 0 17 34 81.5 155 152 139 I 122W195 I Lower Shell Forging 0.05 0.72 4.88 31 42.8 Table Surveillance 40 40 0

0 17 8.5 34 17 43.3 59.8 117 117 115 114 109 105 Shell Circ Weld (100%) 0.18 0.70 0.52 170 Table -56 17 28 65.51 139 148 131 98 (SA-1484) Circ. Weld (100%) 0.26 0.60 4.63 180 Table -33.2 12.2 28 61.08 250 278 262 226 Page 5 of 6

5.0 REFERENCES

(1) NRC letter to Framatome ANP, dated August 4, 2005, Final Safety Evaluation for Topical Report BAW-2308, Revision I , "Initial RTNDT of Linde 80 Weld Materials" (TAC No. MB6636) (ML052070408)

(2) NRC letter to Westinghouse Electric Company, dated March 24, 2008, Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR)

BAW-2308, Revision 2, "Initial FITNDT of Linde 80 Weld Materials" (TAC No. MD4241)

(ML080770349)

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