SBK-L-11186, License Amendment Request 11-06, Application to Revise the Applicability of the Reactor Coolant System Pressure - Temperature Limits and the Cold Overpressure Protection Setpoints

License Amendment Request 11-06, Application to Revise the Applicability of the Reactor Coolant System Pressure - Temperature Limits and the Cold Overpressure Protection Setpoints

ML11329A017
Person / Time
Site:	Seabrook
Issue date:	11/17/2011
From:	Freeman P NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-11186, LAR 11-006
Download: ML11329A017 (23)

Text

NEXTera EN E RGY0.4 November 17, 2011 10 CFR 50.90 SBK-L-1 1186 Docket No. 50-443 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Seabrook Station License Amendment Request 11-06 Application to Revise the Applicability of the Reactor Coolant System Pressure - Temperature Limits and the Cold Overpressure Protection Setpoints In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), NextEra Energy Seabrook, LLC (NextEra) is submitting License Amendment Request (LAR) 11-06 for an amendment to the Technical Specifications (TS) for Seabrook Station. The proposed change revises the applicability of the figures in the TS for the reactor coolant system pressure - temperature limits and the cold overpressure protection setpoints. The change revises the applicability of the figures from 20 effective full power years (EFPY) to 23.7 EFPY.

Attachment 1 to this letter provides NextEra's evaluation of the proposed change, and Attachment 2 provides a markup of the TS showing the proposed change. Changes to the TS Bases will be implemented in accordance with TS 6.7.6j, TS Bases Control Program, upon implementation of the license amendment. As discussed in the evaluation, the proposed change does not involve a significant hazards consideration pursuant to 10 CFR 50.92, and there are no significant environmental impacts associated with the change.

No new commitments are made as a result of this change.

The Station Operation Review Committee has reviewed this LAR. A copy of this LAR has been forwarded to the New Hampshire State Liaison Officer pursuant to 10 CFR 50.91 (b).

4oo3 NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

United States Nuclear Regulatory Commission SBK-L- 11186 / Page 2 NextEra requests NRC review and approval of LAR 11-06 with issuance of a license amendment by November 30, 2012 and implementation of the amendment within 60 days.

Should you have any questions regarding this letter, please contact Mr. Michael O'Keefe, Licensing Manager, at (603) 773-7745.

Sincerely, NextEra Energy Seabrook, LLC Paul Freeman Site Vice President Attachments

1. NextEra Energy Seabrook's Evaluation of the Proposed Change

2. Markup of the Technical Specifications cc: NRC Region I Administrator G. E. Miller, NRC Project Manager W. J. Raymond, NRC Senior Resident Inspector Mr. Christopher M. Pope, Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 Mr. John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399

ENERG ZI4 SEABROOK AFFIDAVIT SEABROOK STATION UNIT 1 Facility Operating License NPF-86 Docket No. 50-443 License Amendment Request 11-06 Application to Revisethe Applicability of the Reactor Coolant: System Pressure -

Temperature L.,imits andthe Cold Overpressure Protection Setpoints The following information is enclosed in support of this License Amendment Request:

" NextEra Energy Seabrook's Evaluation of the Proposed Change

" Markup of the Technical Specifications I, Paul Freeman, Site Vice President of NextEra Energy Seabrook, LLC hereby affirm that the information and statements contained within this license amendment request are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Sworn and Subscribed before me this

/ ' day of . 2011 Paul Freeman Site Vice President

Attachment 1 NextEra Energy Seabrook's Evaluation of the Proposed Change

Subject:

Application to Revise the Applicability of the Reactor Coolant System Pressure Temperature Limits and the Cold Overpressure Protection Setpoints 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Significant Hazards Consideration 4.3 Conclusion

6.0 REFERENCES

Appendix I

1.0

SUMMARY

DESCRIPTION The proposed change revises the applicability of the figures in the Technical Specifications (TS) for the reactor coolant system (RCS) pressure - temperature (PT) limits and the cold overpressure protection (COP) setpoints. The change revises the applicability of the figures from 20 effective full power years (EFPY) to 23.7 EFPY.

2.0 DETAILED DESCRIPTION The proposed change revises the titles of the TS figures as shown below. In addition, where the value of 20 EFPY appears in other instances on Figures 3.4-2 and 3.4-3, it is replaced with 23.7 EFPY.

1. TS Figure 3.4-2 Reactor Coolant System Heatup Limitations - Applicable Up to 20 23. 7 EFPY

2. TS Figure 3.4-3 Reactor Coolant System Cooldown Limitations - Applicable Up to N0 23.7 EFPY

3. Figure 3.4-4 Valid for the first 20 23. 7 EFPY, Setpoint Contains Margin of 50% for Transient Effects

4. The TS index is revised to reflect the change in the titles of figures 3.4-2 and 3.4-3.

3.0 TECHNICAL EVALUATION

Background

The present TS figures for the RCS PT limits (Figures 3.4-2 and 3.4-3) and the figure for the COP setpoints (Figure 3.4-4) are applicable through 20 EFPY. NextEra tracks actual EFPY in a calculation, which is revised every operating cycle to track the effective degradation years of the Seabrook reactor vessel head. The cumulative EFPY at the completion of cycle 14 in the spring of 2011 was 17.85. Assuming a near perfect operating cycle consisting of 1.5 EFPY, the cumulative EFPY will be approximately 19.35 at the end of cycle 15 in October 2012, and the TS figure applicability of 20 EFPY will be reached during cycle 16. Consequently, revised TS figures are required to support operation through cycle 16.

2

Evaluation The technical justification and methodologies employed in the development of the present TS figures and their 20 EFPY applicability are documented in WCAP-15745 and Framatone Letter NFSB 02-0061, published in December 2001 and August 2002 respectively. These documents were based upon the analysis performed on reactor vessel surveillance capsules U and Y. NextEra submitted WCAP-1 5745 and Framatone Letter NFSB 02-0061 to the NRC in October 2002 with a license amendment request [Reference 1] to revise the RCS PT limits to allow operation to 20 EFPY. Amendment 89 [Reference 2] was issued in September 2003 and provided existing TS Figures 3.4-2, 3.4-3, and 3.4-4.

WCAP-15745 was prepared using the then current neutron fluence value determined from cycle 1 through 5 core power distribution and projected forward to 16 EFPY and 20 EFPY. The table below is extracted from WCAP-15745.

TABLE 5 from WCAP-15745 Calculated Neutron Fluence Projections(b) at Key Locations on the Reactor Vessel Clad/Base Metal Interface (1019 n/cm 2, E > 1.0 MeV)

EFPY Azimuthal Location

0. 19o_21.5. 290 31.50 440-450 5.572(a) 0.201 0.355 0.196 0.197 0.369 16(c) 0.577 1.019 0.562 0.565 1.059 20(c) 0.722 1.273 0.702 0.706 1.324 32 1.155 2.037 1.123 1.130 2.119 Notes:

(a) Date of last capsule removal.

(b) Determined by multiplying Best Estimate Fluences by the Calculated-to-Measured Ratio (Ratio = 0.893).

(c) Values have been interpolated between 5.572 EFPY and 32 EFPY.

3

Surveillance capsule V was removed at the end of cycle 10 in April 2005. Analysis of capsule V was performed and documented in WCAP 16526-NP, dated March 2006. NextEra reviewed the analysis results for capsule V and concluded that the TS PT curves and the COP analysis remained valid and conservative for the current 20 EFPY period of applicability. NextEra submitted the capsule V report to the NRC in April 2006 [Reference 3].

In support of developing new PT limit curves applicable to 36 EFPY and to 55 EFPY, Westinghouse Electric Company, LLC prepared WCAP-17441. Calculated fluence projections prepared in WCAP- 17441 were based upon cycle-specific calculations performed for cycles 1 to 14, where a core thermal power of 3411 MWt was used in cycles 1-10, 3587 MWt was used in cycle 11, and 3648 MWt was used in cycles 12-

14. Future projections were based on the assumption that the core power distributions and associated plant operating characteristics from cycles 11, 12, and 13 were representative of future plant operation. Calculated fast neutron (E> 1.0 MeV) fluence for reactor vessel materials were developed at future projections to 22, 28, and through to 60 EFPY. Fluence projections beyond cycle 14 incorporated a 13%

uncertainty factor based upon the guidance in Regulatory Guide 1.190. The following table is extracted from WCAP-17441.

Table 2-3 from WCAP 17441 Seabrook Unit 1 Calculated Neutron Fluence at the Reactor Vessel Clad/Base Metal Interface for cycles 1 through 14 and Future Projections ye Cycle. umulatve Fluence (n/cm 2, E>1.O MeV) Maximum zAimuiial ID im Ccle me Fluence Location of (EF"Y)

SFPY) 00 150 300 4maximum is*_ 3y<. 4 1.0 Mev) Fluence(0 1 0.91 0.91 4.04x10" 6.16x10 1" 7.25x10"7 7.92x10"7 7.92x10" 45 1 18 8 2 0.87 1.78 7.58x10 ' 1.15x10 1.27x10 1.35x10' 1.3 6 x10" 22 8 18 8 8 3 1.21 2.99 1.21x10" 1.80x10 2.04x10" 2.20x101 2.20x10" 45 1 8 4 1.21 4.20 1.63x10" 2.38x10 ' 2.63x10' 2.78x10' 2.78x10" 45 1

5 1.37 5.57 1.99x10 2.94x101' 3.31x10" 3.46x10' 3.46x10" 45 18 8 8 6 1.49 7.06 2.44x10" 3.59x10" 4.04x10 4.17x10" 4.18x10" 23 8 8 8 8 7 1.41 8.47 2.83x10" 4.17x101 4.81x10' 4.95x101' 4.95x10' 28 8 8 8 1.20 9.67 3.17x10" 4.66x10" 5. 4 0x108 5.57x10' 5.57x10' 45 8 8 9 1.34 11.01 3.55x101' 5.24x10' 6.10x10 6.31x10" 6.31x10' 45 10 1.40 12.41 3.95x10" 5.83x1018 6.82x10"8 7 .07 x108 7.07x10" 45 1 18 8 11 1.40 13.81 4.36x10 " 6.44x10 7.53x10" 7.80x101' 7.80x10' 45 12 1.34 15.15 8 18 8 18 8 4.80x101 7.10x10 8.27x10' 8.53x10 8.53x10' 45 13 1.39 16.54 5.27x10l 7.77x10" 8.98x1018 9.22x10" 9.25x101' 28 4

1 1 1 9 14 1.31 17.85 5.71x10 " 8.40x10'8 9.74x10" 9.99xi0 8 1.00x101 28 1 9 9 9 19

- - -.- 22.00 7.07x10" 1.04x10 1.19x101 1.22x10' 1.23x10 ---

--- --- 28.00 9.03x101 1.33x10" 1.51x101' 1.54x101 1.57x109 ---

9 9 19 9

--- --- 36.00 1.16x10" 1.71x10' 1.94x10 1.96x10" 2.01x109 -- -

9 9 9 19

--- --- 42.00 1.36x10" 1.99x10" 2.26x10'9 2.28x10' 2.35x10 - --

9 9 19 19

--- 48.00 1.56x10'9 2.28x10" 2.58x10" 2.59x10 2.68x10 ---

--- --- 55.00 1.79x109 2.62x1019 2.95x109 2.96x10 3.07x10 -- -

. --.- 60.00 1.95x109 2.86x10" 3.22x10" 3.23x10" 3.35x1019 - --

A comparison of the neutron fluence future projection presented in WCAP 15745, Table 5 on the previous page and the projections based upon actual cycle I through 14 data shown in WCAP-17441, Table 2-3 above indicates that, accounting for the data retrieved for the additional operating cycles since the previous fluence calculations, the actual rate of fluence accumulation is lower than that originally considered in the development of the present curves. Considering the identified conservatism in the development of the present TS figures and their 20 EFPY applicability, and the new curves developed in WCAP-17441, Westinghouse provided an applicability evaluation of the current TS Figures 3.4-2, 3.4-3 and 3.4-4.

This applicability evaluation extends the use of the present PT Limit curves from 20 to 23.7 EFPY. The applicability evaluation is provided in the Appendix.

The current Adjusted Reference Temperature (ART) values were based upon the projected reactor vessel neutron fluence used to develop the 20 EFPY PT limit curves. Since the applicability evaluation shows that the same fluence projection is reached in 23.7 EFPY, ART values are not impacted. The COP requirements are based upon the PT limits and likewise are not impacted.

A similar applicability evaluation (extension of applicability of PT limit curves from 23.6 to 35 EFPY on the basis of fluence overestimation) was approved by the NRC for Saint Lucie in September 2005 [Reference 4].

Conclusion Based upon the fluence comparison between WCAP 15745 and WCAP-17441, and the applicability evaluation, the present applicability of 20 EFPY for TS Figure 3.4-2, "Reactor Coolant System Heatup Limitations," Figure 3.4-3, "Reactor Coolant System Cooldown Limitations," and Figure 3.4-4, "RCS Cold Overpressure Protection Setpoints," can be changed to 23.7 EFPY. Actual accumulated fluence and fluence projections are less than previously projected. Using the new fluence projection, the period of applicability for the existing PT Limit curves and COP requirements can be extended from 20 EFPY to 23.7 EFPY with the same margin of safety.

5

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

• 10 CFR 50.60, Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation - requires that light-water nuclear power reactors meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in appendices G and H to this part.

• Appendix G to Part 50, Fracture Toughness Requirements - specifies 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.

4.2 Significant Hazards Consideration No SignificantHazards Consideration The proposed change revises the applicability of the figures in the Technical Specifications for the reactor coolant system (RCS) pressure - temperature limits and the cold overpressure protection setpoints from 20 effective full power years (EFPY) to 23.7 EFPY.

In accordance with 10 CFR 50.92, NextEra Energy Seabrook has concluded that the proposed change does not involve a significant hazards consideration (SHC). The basis for the conclusion that the proposed change does not involve a SHC is as follows:

1. The proposed change does not involve a significant increase in the probabilityor consequences of an accidentpreviously evaluated The proposed change does not impact the physical function of plant structures, systems, or components (SSCs) or the manner in which SSCs perform their design function. The proposed change neither adversely affects accident initiators or precursors, nor alters design assumptions.

The proposed change does not alter or prevent the ability of operable SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. The change does not 6

affect the integrity of the RCS pressure boundary. The proposed change to the applicability of the RCS pressure - temperature limits and the cold overpressure protection setpoints continues to protect the integrity of the RCS pressure boundary.

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

2. The proposed change does not create the possibility of a new or different kind of accidentftom any previously evaluated The proposed change, which revises the applicability of the RCS pressure

- temperature limits and the cold overpressure protection setpoints, will not impact the accident analysis. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a significant change in the method of plant operation, or new operator actions. The proposed change will not introduce failure modes that could result in a new accident. The RCS pressure - temperature limits and the cold overpressure protection setpoints are not accident initiators.

The change does not alter assumptions made in the safety analysis.

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

3. The proposed changes do not involve a significant reduction in the margin qf safety.

Mar2in of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed change does not involve a significant change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design basis. The proposed change does not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. The proposed change to the applicability of the RCS pressure - temperature limits and the cold overpressure protection setpoints continues to protect the integrity of the RCS pressure boundary.

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

7

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(b), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.3 Conclusions 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

NextEra has evaluated the proposed amendment for environmental considerations.

The 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 effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set for in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. NextEra letter NYN-02093, "License Amendment Request 02-04, Revisions to Technical Specification Associated with Pressure/Temperature Curves and Low Temperature Overpressure Protection," October 11, 2002 (ADAMS Accession No. ML022940024)

2. NRC letter "Seabrook Station Unit 1 - Issuance of Amendment Re: Changes to Technical Specifications Associated with Pressure/Temperature Limits and Overpressure Protection Systems (TAC No. MB6613)," September 11, 2003 (ADAMS Accession No. ML032250621) 8

3. NextEra letter SBK-L-06070, "Reactor Vessel Surveillance Capsule Report,"

April 10, 2006 (ADAMS Accession No. ML061030087)

4. NRC letter "St. Lucie Unit 1 - Issuance of Amendment Regarding Extension of the Reactor Coolant System Pressure and Temperature Curve Limits to 35 EFPY (TAC No. MC5580)," September 21, 2006 (ADAMS Accession No. ML052450426) 9

APPENDIX Seabrook Unit I Heatup and Cooldown Limit Curves Applicability Evaluation I

Westinghouse Non-Proprietary Class 3 O Westinghouse To: C. E. Meyer Date: August 3, 2011 cc: A. E. Leicht M. G. Semmler From: B. A. Rosier Ref: LTR-AMLRS-1 1-50 Phone: 412-374-2549 Revision: 0 Fax: 724-940-8559

Subject:

Seabrook Unit 1 Heatup and Cooldown Limit Curves Applicability Evaluation Per the request of Nextera Energy, an applicability evaluation of their current Pressure-Temperature (P-T) limit curves has been performed. The attachment to this letter summarizes the evaluation and conclusion. Please transmit to Nextera Energy.

Do not hesitate to contact the undersigned regarding the content of the letter attachment.

Benjamin A. Rosier Electronically Approved*

Author Date Amy E. Leicht Electronically Approved*

Reviewer Date Michael G. Semmler Electronically Approved*

Acting Manager, Aging Management Date and License Renewal Services

Attachment:

Seabrook Unit 1 Heatup and Cooldown Limit Curves Applicability Evaluation

• Electronically Approved Records are Authenticated in the Electronic Document Management System

Westinghouse Non-Proprietary Class 3 Attachment to LTR-AMLRS-1 1-50, Revision 0 Seabrook Unit 1 Heatup and Cooldown Limit Curves Applicability Evaluation (4 Pages)

Attachment to LTR-AMLRS-1 1-50, Revision 0 Page 1 of 4

Westinghouse Non-Proprietary Class 3

Background

Westinghouse has previously calculated Pressure-Temperature (P-T) limit curves for Seabrook Unit 1 under normal operating conditions through 20 Effective Full Power Years (EFPY) in WCAP-15745 [Ref. 1]. These P-T limit curves are currently implemented by Seabrook Unit 1 and are contained in Figures 3.4-2 and 3.4-3 of the Seabrook Unit 1 Technical Specifications (TS). The evaluation below determines the applicability term with consideration of updated fluence information by comparing the fluence value used in the analysis of record with the updated fluence values that were calculated in support of the 'Seabrook Pressure-Temperature Limit Curves with Optional Scope' project, which is summarized in Westinghouse Project Plan PP-ES-10-0671 [Ref. 2]. The fluence comparison is done only for the limiting reactor vessel material. The new term of applicability is calculated based on a comparison of the fluence values via linear interpolation using the updated fluence projections documented herein.

Acceptance Criteria The acceptance criteria for the P-T limit curves applicability evaluation are not explicitly defined by any regulatory documents. However, the general acceptance criterion for the evaluation herein may be stated as follows:

If the fluence values used in the previous analysis are higher than the fluence values documented herein based on updated conditions, then the applicability date of the existing 20 EFPY P-T limit curves will be increased. If the fluence values used in the previous analysis are lower than the fluence values documented herein using updated conditions, then the applicability date of the existing 20 EFPY P-T limit curves will be decreased.

Fluence Proiection for the Limiting Material Used in Analysis of Record (WCAP-1 5745)

The limiting material used in development of the P-T limit curves in WCAP-15745 [Ref. 1] was Lower Shell (LS) Plate R1808-1. The 20 EFPY surface fluence projection for this material in WCAP-1 5745 is 1.324 x 1019 n/cm 2 (E > 1.0 MeV).

Updated Fluence Prouection for the Limiting Material A neutron fluence update was performed in support of PP-ES-10-0671 [Ref. 2]. The neutron fluence update documents fluence projections for each material at various EFPY. The updated neutron fluence projections for LS Plate R1808-1 at levels which bound the original fluence value used in Reference 1 are summarized below in Table 1:

Table I Calculated Neutron Fluence Projections at the Reactor Vessel Clad/Base Metal Interface for Lower Shell Plate R1808-1 EFPY Fluence (xl019 n/cm 2, E > 1.0 MeV) 22 1.23 28 1.56 Attachment to LTR-AMLRS-1 1-50, Revision 0 Page 2 of 4

Westinghouse Non-Proprietary Class 3 Fluence Comparison to Determine Applicability EFPY Since LS Plate R1808-1 does not have surveillance data, only a fluence comparison is needed in order to determine the revised applicability date of the existing P-T limit curves contained in Figures 3 and 4 of WCAP-1 5745 [Ref. 1]. As stated above, the surface fluence value used in the analysis of record was 1.324 x 1019 n/cm 2 (E > 1.0 MeV). To determine the revised applicability EFPY, the updated fluence values in Table 1 are interpolated to determine the EFPY at which 1.324 x 1019 n/cm 2 (E > 1.0 MeV) is achieved.

Fluence @ 22 EFPY = 1.23 x 1019 n/cm 2 (E > 1.0 MeV)

Fluence @ 28 EFPY = 1.56 x 1019 n/cm 2 (E > 1.0 MeV)

The following interpolation determines the EFPY at which 1.324 x 1019 n/cm 2 (E > 1.0 MeV) is achieved:

2 1.324n/cm 2 -1.23n/cm X EFPY-22 EFPY 2 2 1.56n/cm -1.23n/cm (k28 EFPY-22 EFPYI X EFPY FP ==\1.56n/cm X 2_l.231n/cm2

("324n/cm2_-1"23/c2* (28 EFPY- 22 EFPY) + 22 EFPY X EFPY = 23.7 Therefore, the existing 20 EFPY P-T limit curves that are contained in Figures 3 and 4 of WCAP-15745 [Ref. 1] and in Figures 3.4-2 and 3.4-3 of the Seabrook Unit 1 TS [Ref. 3] are now applicable through 23.7 EFPY. This revised applicability date (23.7 EFPY) also applies to the Reactor Coolant System (RCS) Cold Overpressure Protection Setpoints that are contained in Figure 3.4-4 of the Seabrook Unit 1 TS [Ref. 3].

End-of-cycle (EOC) 14 pertains to 17.85 EFPY and occurred in April 2011. No single cycle time in the operating history of Seabrook Unit 1 has reached or exceeded 1.50 EFPY. Therefore, an assumption of 1.50 EFPY per cycle for future cycles is a conservative assumption. The cycle and calendar date pertaining to 23.7 EFPY are calculated below.

17.85 EFPY + (1.5 EFPY/cycle

• X cycles) = 23.7 EFPY X cycles = 3.9 Conservatively, 23.7 EFPY will pertain to three additional cycles beyond EOC 14. Therefore, the Seabrook Unit 1 P-T limit curves and RCS Cold Overpressure Protection Setpoints are applicable through 23.7 EFPY, pertaining to approximately EOC 17 and a calendar date of October 2015. New P-T limit curves and RCS Cold Overpressure Protection Setpoints should be implemented by Seabrook Unit 1 prior to the above date.

Attachment to LTR-AMLRS-1 1-50, Revision 0 Page 3 of 4

Westinghouse Non-Proprietary Class 3 References

1. WCAP-1 5745, Revision 0, "Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," T. J. Laubham, December 2001.

2. Westinghouse Project Plan PP-ES-10-0671, Revision 0, "Seabrook Pressure-Temperature Limit Curves with Optional Scope," March 2011.

3. "Seabrook Station Technical Specifications," Section 3/4.4.9, Pressure/Temperature Limits, Amendment No. 115. [Note that the Amendment No. is specific to Section 3/4.4.9]

Attachment to LTR-AMLRS-1 1-50, Revision 0 Page 4 of 4

Attachment 2 Mark-up of the Technical Specifications (TS)

The attached markups reflect the currently issued version of the TS and Facility Operating License. At the time of submittal, the Facility Operating License was revised through Amendment No. 127.

Listed below are the license amendment requests that are awaiting NRC approval and may impact the currently issued version of the Facility Operating License affected by this LAR.

LAR Title NextEra Energy Date Seabrook Letter Submitted 10-02 Application for Change to the Technical Specifications SBK-L-10074 05/14/2010 for the Containment Enclosure Emergency Air Cleanup System 11-01 Application to Revise the Technical Specifications for SBK-L- 11066 4/21/2011 Reactor Coolant System Leakage Detection Instrumentation 11-03 License Amendment Request Regarding Containment SBK-L- 11130 7/14/2011

_ Spray Nozzles Surveillance Requirement 11-05 Cold Leg Injection Permissive SBK-L- 11181 9/30/2011 The following TS pages are included in the attached markup:

Technical Title Page Specification Index iv Figure 3.4-2 Reactor Coolant System Heatup Limitations 3/4 4-23 Figure 3.4-3 Reactor Coolant System Cooldown Limitations 3/4 4-24 Figure 3.4-4 RCS Cold Overpressure Protections Setpoints 3/4 4-30 1

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/gram DOSE EQUIVALENT 1-131 3/4 4-20 TABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM 3/4 4-21 3/4.4.9 PRESSURE/TEMPERATURE LIMITS General .................................................... 3/4 4-22 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -

APPLICABLE UP TO EFPY.. ................................................................. 3/44-23 FIGURE 3.4-3 REACTOR COOLANT S STEM COOLDOWN LIMITATIONS -

APPLICABLE UP TO E Y............. .3/4 4-24 Pressurizer 3/4 4-25 Overpressure Protection Systems ............................................................... 3/4 4-26 FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS 3/4 4-29 3/4.4.10 3441 DELETED DE E E .................................................................................... 3/4 3/. 4-30 43 3/4.4.11 REACTOR COOLANT SYSTEM VENTS 3/4 4-31 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Hot Standby, Startup, and Power Operation .......................................................... 3/4 5-1 Shutdown S ud w ...................................................................................

3/4 5-3 3/45-3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350OF ............ 3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350OF .................................................. 3/4 5-8 ECCS SYBSYSTEMS - Tavg Equal To or Less Than 200OF ................................ 3/4 5-10 3/4.5.4 REFUELING WATER STORAGE TANK 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT C ontainm ent Integrity ................................................................................................. 3/4 6-1 C ontainm ent Leakage ................................................................................................ 3/4 6-2 SEABROOK- UNIT 1 vi Amendment No. 70, 89, 115,4 26 Corrected By Letter Dated May 17, 2007

MATERIAL PROPERTY BASIS Limiting material: LOWER SHELL PLATE R-1808-1 Limiting ART values atOEFPY.I/4T, 109°F 3/4T, 88-F Curves applicable for the firs EFPY and contain margins of 20'F and 100 psig for possible instrument errors 2800 r- r- - I - r -- - T - r - - -r - _ T-

- - -I. _ _ I 2600 -

Si i i - -L J i I

i. i J L -1 1, t

e ra O L J, I I T I - 1 - I " 1, -- - - r I- i I 2400 -

LeakTest Limit

. . .L..4 , .L ... .. . -... .. ... .. . .. . .

2200 -

Z 2000 -

1 i i1-- 2..1 - -...., __,_._ J.....

.... (1-3-F'--ort,

-.... - -Operationr -...

i I 2 I I tserice perio up I t 2/

I 2 I 2 2 I 2 i I I i i2 I I i I I I i i C 1800 -

0~

... .. . - . .--4.-... . . .

---

i r - n, -r - - -I r n-i r - - i - ,--

F 1600 -

0~

0 T1400,-

0 2-Heatup 'I- Rate r2 4i- -2 2 I I2 I 1 - 2 22I,- I Critical Limit 22 22 2-...2 2211 .

i r 10ODeg.

I I F/HrI i 2 2100Deg. 2 F/Hr1 1 1I, i - r i - 2 ,

C,) Unacceptabl hyrs atcteptal e.*-1200 -

t 2 i i i i 2 I I 2 2 i I i 2 i 2 i I 2

.-. "--- -4 " -4 " .- A- "1 ~-- T-7-J --- I.--4 - - -- 4 - W -T-7--

-*-2S .- I-*-

U) 1000 2 2 2 2 *2I 2 2 I 2 2 2 2 1 2 i2 1 i2 2 t i i i ii LE 2 I I I 2 I 2 2l2 1I 2 1 2 I I 2 I l I I I 2 0-Ci) 800 2-.-I-- - - .-.- ---. -

.- -,--I- --. 22

.-..- 2 1 _1_2 2 2 _._2 2 2 _2_

Cc)

.... L.. ... '.2....... ..,- L , ,L .Critica L im .itb sedI- L II 600

........ 2 . ...2 2 inserviehydrostatic test * ._i...

400 ' ' ' "2* 2 , temperature(13Ffrthe , , , 2

. . . .. olup .. -  :~ service period upt EFPY 200

.. 22........2.

...... 44......J_........... ............ ... _____________

0- I I I I " I " " ",5,5+ , i , ,

0 100 200 300 400 500 RCS TEMPERATURE (Deg. F, 20 Deg. F PER DIVISION)

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP T EFPY SEABROOK - UNIT 1 3/4 4-23 Amendment No. 1-9,89, 115

MATERIAL PROPERTY BASIS Limiting material: LOWER SHELL PLATE R-1808-1 Limiting ART values a FPY:I/4T, 109°F 3-3-/4T, 88°F Curves applicable for the firs FPY and contain margins of 20°F and 100 psig for possible instrument errors 2800 -

-, -, -- - . - -.- . - , --- , . - - ý-. , - - , - --- --- - . . -

2600 -

. . . . .. .. . - i -

T- r - - T ii IT r 2400 -

1 i T-i - i r - -i - T -i - F - 1 -i - - -i - - r - " - i - T - - - r T , -

2200 -

z SUnacceptable  :  ::::Acceptable~

O peration j- --------------------- Operation 0 2000 -

iF - I T T i - ir - r i - - -

CC1800 - ---- Cooldown Rate w ---------- uptol100Deg. F/Hr CL 1600 -

-- ~~~~~~~

-- _i-J I - _ - rIJ-

. n, r .J . ._1r LJ L *J 1

-1_ _.

- -_ -L1- - 1 71400 -

01200 cc Tempp i i ,

• Z) co 1000 ELi 800 Cooldown Ratesj LJ-L-LJ .

0

<60ODeg. F/I-r . --- -* - - I4 600 -

400- 80,lOO0Deg. F/hr -1L}J L J- LJ IJ--LLJ 200 I I 1 Boltup . ,- - --- - - - -,-- -

0 0 100 200 300 400 500 RCS TEMPERATURE (Deg. F, 20 Deg. F PER DIVISION) *7 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO FPY SEABROOK - UNIT 1 3/4 4-24 Amendment No. 1-49-89, 115

VALID FOR THE FIRSI&FPY, SETPOINT CONTAINS MARGIN OF 50-F FOR TRANSIENT EFFECTS T s 200.0 0F, P = 561.0 PSIG; 0

200.0 F < T < 230.5 0F, P = 12.1*(T-200.0) + 926.0 PSIG; 230.50F <T T_ 255.0 0F, P = 23.15*(T-230.5) + 1295.05 PSIG; T > 255.0 0F, P = 34.5*(T-255.0) + 1862.225 PSIG 250 0 225 i0 0.-

200 175 i0

-. " 150 0...................................

)0

.... i..... ......................

..................... i..........

i.............

0 I.

125 100 0...

0 0.

75 i0 0 .. ._

50 25 0. .......

v 50 100 150 200 250 300 350 RCS TEMPERATURE (DEG. F)

FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS SEABROOK - UNIT 1 3/4 4-30 Amendment No. 8 9 ,--415, 441-6