SBK-L-10153, License Amendment Request 10-05 to Delete Technical Specification 3/4.4.4.10, Structural Integrity

License Amendment Request 10-05 to Delete Technical Specification 3/4.4.4.10, Structural Integrity

ML110070067
Person / Time
Site:	Seabrook
Issue date:	12/29/2010
From:	Freeman P NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-10153
Download: ML110070067 (17)

Text

NEXTera December 29, 2010 10 CFR 50.90 SBK-L-10153 Docket No. 50-443 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Seabrook Station License Amendment Request 10-05 Application to Delete Technical Specification 3/4.4.10, Structural Integrity.

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) 10-05 for an amendment to the Technical Specifications (TS) for Seabrook Station. The proposed change would delete TS 3/4.4.10, Structural Integrity. TS 3/4.4.10 does not meet the criteria of 10 CFR 30.36 for inclusion in the TS, and deletion of the TS is consistent with NUREG-1431, Standard Technical Specifications Westinghouse Plants.

Attachment I to this letter provides NextEra's evaluation of the proposed change, and provides a markup of the TS showing the proposed change. The TS bases information related to TS 3/4.4.10 will be removed in accordance with TS 6.7.6j, TS Bases Control Program, upon implementation of the license amendment and is not included in this submittal. 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).

NextEra requests NRC review and approval of LAR 10-05 with issuance of a license amendment by December 30, 2011 and implementation of the amendment within 60 days.

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

United States Nuclear Regulatory Commission SBK-L-10153 / Page 2 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

NEXTera" ENER'Y SEABROOK AFFIDAVIT 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 beforje Ine this day of "g'oX*)*,2010 Notary Public Paul Freeman Site Vice President NextEra Energy Seabrook's Evaluation of the Proposed Change

Subject:

Application to Delete Technical Specification 3/4.4.10, Structural Integrity.

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

1

1.0

SUMMARY

DESCRIPTION The proposed change deletes TS 3/4.4.10, Structural Integrity from the Seabrook Station Technical Specifications (TS). Existing surveillance requirement 4.4.10 for the reactor coolant pump flywheel inspection is relocated to a new program. The proposed change is consistent with NUREG-1431, Standard Technical Specifications Westinghouse Plants [Reference 1].

2.0 DETAILED DESCRIPTION The proposed change includes the following:

" TS 3/4.4.10, including the limiting condition for operation (LCO), actions, and surveillance requirements (SR) are deleted,

• Existing surveillance requirement 4.4.10 for the reactor coolant pump flywheel inspection is relocated to a new program, Reactor Coolant Pump Flywheel Inspection Program, as TS 6.7.6.m in the administrative section of the TS, and

• The TS index is revised to reflect deletion of the requirements for structural integrity.

=

Background===

TS 3/4.4.10 The purpose of TS 3/4.4.10, Structural Integrity, is to specify the requirements for maintaining the structural integrity of American Society of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components. This specification was originally intended to provide assurance that structural integrity and operational readiness of these components are maintained at an acceptable level throughout the life of the facility. The specification is applicable in all operational modes. However, the specification does not provide actions for a plant shutdown if its LCO is not met. In addition, the specification contains no SRs other than a periodic inspection of the reactor coolant pump flywheel. This is because the specification addresses the passive pressure boundary function of ASME Code Class 1, 2, and 3 components as established under the inservice inspection (ISI) program. The ISI program is required pursuant to 10 CFR 50.55a, thereby addressing the inspections necessary to maintain structural integrity.

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The wording of TS 3.4.10 could be misconstrued to conflict with normal outage-related activities, including removal of RCS manways and the reactor vessel head in preparation for refueling, which make the pressure boundary no longer structurally intact. Maintaining a program-type requirement within an LCO creates significant interpretation issues for Operations personnel. The RCS structural integrity TS was part of the original TS, and the TS basis history regarding its intent is not documented. TS 3.4.10 appears to have been included to help ensure that plant heat up and startup would not occur until all required portions of the RCS were verified to meet ISI acceptance criteria following inspections performed during a plant outage. Meeting these acceptance criteria helps ensure the integrity of the RCS pressure boundary during all modes of operation, including accident events. Further, TS 3.4.10.1 contains no action suggesting it was designed to accommodate integrity concerns once plant heat up has commenced. Most RCS structural integrity ISI activities are performed only during plant outages when conditions exist that permit access to the RCS pressure boundary and are not monitored or controlled through application of the ISI program during the operational cycles.

Other TS are designed to monitor the structural integrity of the RCS during operation and provide actions to shut down the unit if compliance is not maintained. For example, limits on RCS heat up and cool down rates protect agdinst applying undue stresses as a result of pressure/temperature transients on RCS components and piping.

The RCS leakage TS provides a means of evaluating the RCS structural integrity by detecting and monitoring leakage. Therefore, applying TS 3.4.10 when integrity issues become evident during plant operation above cold shutdown is not necessary.

Because TS 3.4.10 is redundant to other regulations, it is acceptable to remove TS 3.4.10 from the TS. Finally, removal of this specification does not reduce the controls that are necessary to ensure compliance with the ASME Code or the need to maintain the RCS pressure boundaries. Structural integrity is maintained by compliance with 10 CFR 50.55a, as implemented through the Seabrook Station ISI program.

SR 4.4.10 This change proposes relocating the requirements in SR 4.4.10 to a new program in the administrative section of the TS. This relocation does not alter the current inspection requirements for the reactor coolant pump flywheels. Therefore, this change is administrative in nature.

Criteria for Technical Specifications Section 50.36c(2)(ii) of Title 10 of the Code of Federal Regulations (10 CFR 50.36c(2)(ii)) contains the requirements for items that must be in TS. This regulation provides four criteria that can be used to determine the requirements that must be 3

included in the TS. A TS limiting condition for operation (LCO) of a nuclear reactor must be established for each item meeting one or more of the following criteria:

Criterion 1:

Criterion 2:

Criterion 3:

Criterion 4:

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

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.

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.

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

Items not meeting any of these four criteria can be relocated from the TS to a licensee controlled document. Relocated requirements can then be changed, if necessary, in accordance with 10 CFR 50.59.

Evaluation Following is an evaluation of the proposed change with regard to the criteria of 10 CFR 50.36c(2)(ii).

Criterion 1 addresses installed instrumentation that is used to detect and indicate excessive reactor coolant system leakage. TS 3/4.4.10, which addresses structural integrity of ASME Code Class 1, 2, and 3 components, does not involve installed instrumentation that is used to detect, and indicate in the control room, a significant degradation of the reactor coolant pressure boundary. Therefore, TS 3/4.4.10 does not meet Criterion 1.

The purpose of Criterion 2 is to capture those process variables that have initial values assumed in the design basis accident and transient analyses and that are monitored and controlled during power operation. This criterion also includes active design features (e.g., high-pressure/low-pressure system valves and interlocks) and operating restrictions (pressure/temperature limits) needed to preclude unanalyzed accidents and transients. TS 3/4.4.10 does not involve process variables that have initial values assumed in the design basis accident and transient analyses and that are 4

monitored and controlled during power operation. Although this TS is related to the integrity of ASME Code Class 1, 2, and 3 components, compliance with 10 CFR 55a through implementation of the Seabrook Station ISI program maintains integrity of these components. Structural integrity is verified during periodic inspections rather than through specific monitoring and control during plant operations. Therefore, TS 3/4.4.10 does not meet Criterion 2.

The purpose of Criterion 3 is to capture only those structures, systems, and components that are part of the primary success path of the safety analysis (the actions required to mitigate the consequences of the design basis accidents and transients). The primary success path of a safety analysis consists of the combinations and sequences of equipment needed to operate so that the plant responses to the design basis accident and the transients limit the consequences of these events within the appropriate acceptance criteria. Also captured by this criterion are those support and actuation systems that are necessary in the primary success path, but this criterion does not include backup and diverse equipment. No specific TS-related structure, system, or component (SSC) is being revised or removed from the TS. Each SSC must continue to meet the requirements of 10 CFR 55a through the Seabrook Station ISI program. Therefore, TS 3/4.4.10 does not meet Criterion 3.

The purpose of Criterion 4 is to capture only those structures, systems, and components that operating experience and probabilistic safety assessment has shown to be significant to the public health and safety. As discussed in the evaluation of Criterion 3, no specific TS-related structure, system, or component (SSC) is being revised or removed from the TS. Each SSC must continue to meet the requirements of 10 CFR 55a through the Seabrook Station ISI program. Therefore, TS 3/4.4.10 does not meet Criterion 4.

Precedent In response to the Commission's Interim Policy Statement on Technical Specification Improvements, published in February 1987, the nuclear steam supply system owners groups submitted for NRC review a report that proposed relocating certain TS (Split Report). Following a review of the reports, the NRC staff published its conclusion in May 1988 [Reference 2], which concurred with the proposal that the TS for structural integrity does not meet the criteria for inclusion in the TS.

NUREG-1431, Standard Technical Specifications Westinghouse Plants, does not include a LCO for structural integrity. As discussed previously, during development of NUREG 1431, these requirements were determined not to meet the criteria of 10 CFR 50.36c(2)(ii) for inclusion in the TS.

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The NRC previously approved similar requests to remove the requirements for structural integrity from the TS in the following:

Limerick Generating Station, Amendments 199 and 160 (ML100130562)

[Reference 3]

Turkey Point, Amendments 242 and 238 (ML100210321)

[Reference 4]

Arkansas Nuclear One, Amendment 270 (ML070570506)

[Reference 5]

Conclusion The evaluation of TS 3/4.4.10 against the criteria of 10 CFR 50.36(c)(2)(ii) and determined that none of these criteria require retaining the structural integrity controls specified in TS 3/4.4.10 in the TS. In addition, the requirement in TS 3/4.4.10 are redundant to the requirements already covered under 10 CFR 55a with which NextEra must comply. Therefore, removal of TS 3/4.4.10 from the TS is acceptable.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36c(2)(ii) contains the criteria for items that must be included in the TS. The requirements in TS 3.8.4.2 do not meet any of the criteria for inclusion in the TS; therefore, NextEra concludes that relocation of TS 3.8.4.2 is acceptable.

4.2 Significant Hazards Consideration No Significant Hazards Consideration 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 probability or consequences of an accident previously 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 6

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.

This proposed change removes from the Technical Specifications the requirements associated with structural integrity. Removing these requirements will have no adverse effect on plant operation, the availability or operation of any accident mitigation equipment, or plant response to a design basis accident. The change has no impact on the ability of ASME Code Class 1, 2, and 3 components to perform their safety functions since these components remain under the control of 10 CFR 55a.

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 accident from any previously evaluated.

The proposed change 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 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 of safety.

Margin 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.

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Therefore, these proposed changes do 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(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. NUREG-1431, Standard Technical Specifications Westinghouse Plants, Revision 3

8

2. NRC letter from T. Murley to W. Wilgus, "NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications," May 9, 1988

3. NRC letter "Limerick Generating Station, Units 1 and 2 - Issuance of Amendments Re: Removal of Technical Specification Concerning Structural Integrity Requirements (TA) Nos. ME0740 and ME0741)," February 24, 2010 (ML100130562)

4. NRC letter "Turkey Point Units 3 and 4 - Issuance of Amendments Regarding Technical Specification Change Associated with Removal of Structural Integrity Requirements and Technical Specification Improvement to Extend the Inspection Interval for Reactor Coolant Pump Flywheels (TAC Nos. ME0701 and ME0702)," February 23, 2010 (ML100210321)

5. NRC letter "Arkansas Nuclear One, Unit 2 - Issuance of Amendment Re:

Removal of Reactor Coolant System Structural Integrity Requirements (TAC No.

MD0700)," March 1, 2007 (ML070570506) 9 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. 124.

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 Revision to Technical Specification SBK-L-09118 05/28/2009 LAR 09-03 6.7.6.k, "Steam Generator (SG)

Program," for Permanent Alternate Repair Criteria (H*)

LAR 10-02 Application for Change to the SBK-L-10074 05/14/2010 Technical Specifications for the Containment Enclosure Emergency Air Cleanup System LAR 10-03 Relocation of Technical SBK-L-10097 06/28/2010 Specification 3.8.4.2, Containment Penetration Conductor Overcurrent Protective Devices and Protective Devices for Class 1E Power Sources Connected To Non-Class 1E Circuits LAR 10-04 Amendment to the Facility SBK-L-10119 07/26/2010 Operating License and Submittal of the Seabrook Station Cyber Security Plan The following TS pages are included in the attached markup:

1

Technical Page Specification Title TS Index vi TS 3.4.10 Structural Integrity 3/4 4-31 6.7.6 Procedures and Programs 6-14a 2

REACTOR COOLANT SYSTEM QSTALeUXAL TE6T 9

~

L

~

6 F'LIMITING CONDITIONFOR"6PERATION,

"/0010!

3.4.10 T utrlitgiy M oeC a r 2,

ad3 components sh'allbe mana naccordance wi ecificto 4.4 1

PPLICABILITY:

MODES.

a.

With the St uctural integrity 3jany ASME Code Class 1 component(s) not conf ng to the abov et*uirements, restore the stýtuaral integrity of the

.aected component(,)Ito within its limit or isolate the affected componen 1(5 prior to increasingthie Reactor Coolant Systemtteperature more thanrT50 0F above the mini'mum temperature requiredby NDT considerationse's.

b.

With~the-structural integrity of a y MAME Code Class 2 com1'ponent(s) not conforming to the above reqqpirements, restore thestncural integrity of the affected component(s) too1thin its limit or isolate he affected compongnt(s) prior to increasing theReactor Coolant System temperature above2OOF.

c.

With the structural integrity of any ASME Code Class 3.coim'ponent(s) not conforming to the above requirements, restore the,structural integrity of the affe"cted component(s) to within its limit or isolate the affected component(p) om service.

SURVEILLANCE REQUIREMENTS 4.4.10 In addition to the requirements of Specification 4.0.seac-h reactor coolant pum, iflwheel shall'be inspectedatleast once every 10 years. tflis inspection shall be byeithr of thefdllowing examinations:

a.

Anin-pilace examination, utilizrg ultrasonic testing, over the volume from the inner bore of the flywheel toethe circle of one-half the oCter radius; or 2"b.

A surface examination, utilizing magnetic particle testing and/or penetrant testing, of thýeexposed surfaces of thedisassembled flywheel.

SEABROOK - UNIT 1 3/4 4-31 Amendment No. 7-9,4--6,-446-

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)

d.

Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREMAFS, operating at a flow rate of less than or equal to 600 CFM at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

e.

The quantitative limits on unfiltered air in-leakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air in-leakage measured by the testing described in paragraph c. The unfiltered air in-leakage limit for radiological challenges is the in-leakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f.

The provisions of SR 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered in-leakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

SEABROOK - UNIT 1 6-14a Amendment No.4-1 INSERT

m.

Reactor Coolant Pump Flywheel Inspection Progam In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected at least once every 10 years. This inspection shall be by either of the following examinations:

a. An in-place examination, utilizing ultrasonic testing, over the volume from the inner bore of the flywheel to the circle of one-half the outer radius; or

b. A surface examination, utilizing magnetic particle testing and/or penetrant testing, of the exposed surfaces of the disassembled flywheel.