ML15175A439

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Duke Energy Carolinas, LLC, Application to Revise Technical Specifications to Adopt TSTF-523, Rev. 2, Generic Letter 2008-01, Managing Gas Accumulation, Using the Consolidated Line Item Improvement Process. Part 1 of 6
ML15175A439
Person / Time
Site: Oconee, Mcguire, Catawba, Harris, Brunswick, McGuire  Duke Energy icon.png
Issue date: 06/24/2015
From: Repko R
Duke Energy Carolinas
To:
Document Control Desk, Office of New Reactors
Shared Package
ML15175A438 List:
References
RA-15-0006
Download: ML15175A439 (50)
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{{#Wiki_filter:Regis T. Repko 526 South Church Street Charlotte, NC 28202 Mailing Address: Mail Code EC07H / P.O. Box 1006 Charlotte, NC 28201-1006 704-382-4126 704-382-6056 fax Serial: RA-15-0006 10 CFR 50.90 June 24, 2015 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-325 AND 50-324 / RENEWED LICENSE NOS. DPR-71 AND DPR-62 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 / RENEWED LICENSE NO. NPF-63 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 / RENEWED LICENSE NOS. NPF-35 AND NPF-52 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370 / RENEWED LICENSE NOS. NPF-9 AND NPF-17 OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-269, 50-270 AND 50-287 / RENEWED LICENSE NOS. DPR-38, DPR-47 AND DPR-55

SUBJECT:

APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-523, REVISION 2, GENERIC LETTER 2008-01, MANAGING GAS ACCUMULATION, USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS

REFERENCES:

1. Duke Energy letter, Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Schedule for Submittal of a License Amendment Request to Adopt Technical Specification Task Force Traveler 523, dated October 13, 2014 (ADAMS Accession No. ML14296A380)
2. Duke Energy letter, Shearon Harris Nuclear Power Plant, Unit No. 1, Schedule for Submittal of a License Amendment Request to Adopt Technical Specification Task Force Traveler 523, dated October 13, 2014 (ADAMS Accession No. ML14286A097)
3. Duke Energy letter, Catawba Nuclear Station, Units 1 and 2, Schedule for Submittal of a License Amendment Request to Adopt Technical Specification Task Force Traveler 523, dated October 9, 2014 (ADAMS Accession No. ML14301A335)
4. Duke Energy letter, McGuire Nuclear Station Units 1 and 2, Schedule for Submittal of a License Amendment Request to Adopt Technical Specification Task Force Traveler 523, dated October 8, 2014 (ADAMS Accession No. ML14296A384)

U.S. Nuclear Regulatory Commission RA-15-0006 Page 2 REFERENCES (CONTD):

5. Duke Energy letter, Oconee Nuclear Station, Units 1, 2 and 3, Schedule for Submittal of a License Amendment Request to Adopt Technical Specification Task Force Traveler 523, dated October 6, 2014 (ADAMS Accession No. ML14290A015)
6. Federal Register dated January 15, 2014 (79 FR 2700), TSTF-523, Generic Letter 2008-01, Managing Gas Accumulation, Using the Consolidated Line Item Improvement Process Pursuant to 10 CFR 50.90, Duke Energy Progress, Inc., and Duke Energy Carolinas, LLC, collectively referred to henceforth as Duke Energy, is submitting a request for amendments to the Technical Specifications (TSs) for Brunswick Steam Electric Plant, Unit Nos. 1 and 2 (BSEP); Shearon Harris Nuclear Power Plant, Unit 1 (HNP); Catawba Nuclear Station, Units 1 and 2 (CNS); McGuire Nuclear Station, Units 1 and 2 (MNS); and Oconee Nuclear Station, Units 1, 2, and 3 (ONS).

The proposed amendments would modify TS requirements to address Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems, as described in TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation. Duke Energy committed to submit these proposed changes in References 1 through 5. The Notice of Availability for TSTF-523 was published in the Federal Register on January 15, 2014 (Reference 6). provides a description and assessment of the proposed change. Attachment 2 provides the existing TS pages marked up to show the proposed change. Attachment 3 provides existing TS Bases pages marked up to show the proposed change. Changes to the existing TS Bases, consistent with the technical and regulatory analyses, will be implemented under the Technical Specification Bases Control Program. They are provided in Attachment 3 for information only. Attachment 4 provides the retyped TS pages. Approval of the proposed amendment is requested by June 24, 2016. Once approved, the amendment shall be implemented within one year. This extended implementation period is necessary to allow completion of various actions associated with incorporation of the proposed changes into the TS. This submittal contains no new regulatory commitments. In accordance with 10 CFR 50.91, Duke Energy is notifying the States of North Carolina and South Carolina of this license amendment request by transmitting a copy of this letter and attachments to the designated State Officials. Should you have any questions concerning this letter, or require additional information, please contact Art Zaremba, Manager - Nuclear Fleet Licensing, at 980-373-2062.

U.S. Nuclear Regulatory Commission RA-15-0006 Page3 I declare under penalty of perjury that the foregoing is true and correct. Executed on June 24, 2015. Sincerely, Regis T. Repko Sr. Vice President-Governance, Projects, and Engineering MKL Attachments: 1. Description and Assessment

2. Proposed Technical Specification Changes (Mark-Up)
3. Proposed Technical Specification Bases Changes (Mark-Up)

(For information only)

4. Retyped Technical Specification Pages cc:

USNRC Region II M. P. Catts, USNRC Resident Inspector-BSEP J. D. Austin, USNRC Resident Inspector-HNP G. A. Hutto, Ill, USNRC Resident Inspector-CNS J. Zeiler, US NRC Resident Inspector-MNS E. L. Crowe, USNRC Resident Inspector-ONS A. Hon, NRR Project Manager-BSEP M. C. Barillas, NRR Project Manager-HNP G. E. Miller, NRR Project Manager-CNS & MNS J. R. Hall, NRR Project Manager-ONS J. A. Whited, NRR Project Manager W. L. Cox, Ill, Chief, Division of Health Service Regulation, RP Section (NC) S. E. Jenkins, Manager, Radioactive and Infectious Waste Management (SC) Chair - North Carolina Utilities Commission RA-15-0006 Page 1 of 9 DESCRIPTION AND ASSESSMENT

Subject:

License amendment application to revise Technical Specifications to adopt TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation

1.0 DESCRIPTION

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation 2.2 Optional Changes and Variations

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination 4.0 ENVIRONMENTAL EVALUATON RA-15-0006 Page 2 of 9

1.0 DESCRIPTION

The proposed change revises or adds Surveillance Requirements to verify that the system locations susceptible to gas accumulation are sufficiently filled with water and to provide allowances which permit performance of the verification. The changes are being made to address the concerns discussed in NRC Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems. The proposed amendment is consistent with TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation.

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation Duke Energy Progress, Inc., and Duke Energy Carolinas, LLC, collectively referred to henceforth as Duke Energy, have reviewed the model safety evaluation, dated December 23, 2013, as part of the Federal Register Notice of Availability. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-523. Duke Energy has concluded that the justifications presented in the TSTF-523 proposal and the model safety evaluation prepared by the NRC staff are applicable to Brunswick Steam Electric Plant, Unit Nos. 1 and 2 (BSEP), Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Catawba Nuclear Station, Units 1 and 2 (CNS), McGuire Nuclear Station, Units 1 and 2 (MNS), and Oconee Nuclear Station, Units 1, 2, and 3 (ONS), and justify this amendment for the incorporation of the changes to the respective plant Technical Specifications (TS). The Traveler and model Safety Evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). BSEP and ONS were not licensed to 10 CFR 50, Appendix A, GDC. For BSEP and ONS, conformance to the applicable GDC is discussed in Section 3.1 of the respective Updated Final Safety Analysis Reports. This difference does not alter the conclusion that the proposed change is applicable to BSEP and ONS. 2.2 Optional Changes and Variations Duke Energy is proposing the following variations from the TS changes described in the TSTF-523, Revision 2, or the applicable parts of the NRC staffs model safety evaluation dated December 23, 2013. The following plant TS utilize different numbering or titles than the Standard Technical Specifications on which TSTF-523 was based: Plant TSTF-523 Reference(2) Plant TS Reference BSEP TS 3.4.8, RHR Shutdown Cooling System - Hot Shutdown TS 3.4.7, RHR Shutdown Cooling System - Hot Shutdown BSEP TS 3.4.9, RHR Shutdown Cooling System - Cold Shutdown TS 3.4.8, RHR Shutdown Cooling System - Cold Shutdown RA-15-0006 Page 3 of 9 Plant TSTF-523 Reference(2) Plant TS Reference BSEP TS 3.9.8, RHR - High Water Level TS 3.9.7, RHR - High Water Level BSEP TS 3.9.9, RHR - Low Water Level TS 3.9.8, RHR - Low Water Level HNP(1) TS 3.4.6, RCS Loops - MODE 4 TS 3/4.4.1.3, Reactor Coolant System, Hot Shutdown HNP(1) TS 3.4.7, RCS Loops - MODE 5, Loops Filled TS 3/4.4.1.4.1, Reactor Coolant System, Cold Shutdown - Loops Filled HNP(1) TS 3.4.8, RCS Loops - MODE 5, Loops Not Filled TS 3/4.4.1.4.2, Reactor Coolant System, Cold Shutdown - Loops Not Filled HNP(1) TS 3.5.2, ECCS - Operating TS 3/4.5.2, ECCS Subsystems - Tavg Greater Than or Equal to 350ºF HNP(1) TS 3.6.6, Containment Spray and Cooling Systems TS 3/4.6.2.1, Containment Spray System HNP(1) TS 3.9.5, RHR and Coolant Circulation - High Water Level TS 3/4.9.8.1, Residual Heat Removal and Coolant Circulation, High Water Level HNP(1) TS 3.9.6, RHR and Coolant Circulation - Low Water Level TS 3/4.9.8.2, Refueling Operations, Low Water Level CNS TS 3.9.5, RHR and Coolant Circulation - High Water Level TS 3.9.4, RHR and Coolant Circulation - High Water Level CNS TS 3.9.6, RHR and Coolant Circulation - Low Water Level TS 3.9.5, RHR and Coolant Circulation - Low Water Level ONS TS 3.6.6, Containment Spray and Cooling Systems TS 3.6.5, Reactor Building Spray and Cooling Systems ONS TS 3.9.5, RHR and Coolant Circulation - High Water Level TS 3.9.4, DHR and Coolant Circulation - High Water Level ONS TS 3.9.6, RHR and Coolant Circulation - Low Water Level TS 3.9.5, DHR and Coolant Circulation - Low Water Level Table Notes: (1) HNP has not converted to the NUREG-1431 improved Standard Technical Specifications (STS). The general format and numbering convention associated with the current TS has been retained. (2) The General Electric BWR/4 STS (NUREG-1433) is applicable to BSEP. The Westinghouse STS (NUREG-1431) is applicable to HNP, CNS, and MNS. The Babcock & Wilcox STS (NUREG-1430) is applicable to ONS. RA-15-0006 Page 4 of 9 These differences, including the associated differences in the numbering of Surveillance Requirements (SRs), are administrative and do not affect the applicability of TSTF-523 to the individual plant TS. Other variations include the following: The General Electric BWR/4 STS markup included in TSTF-523 for SR 3.5.1.2 includes a new note, which is a single note for this SR. The corresponding current BSEP SR 3.5.1.2 already has an existing note. Therefore, it is proposed that the existing note be retained but relabeled as Note 1, and the new note labeled as Note 2. The General Electric BWR/4 STS markup included in TSTF-523 for SR 3.5.2.4 includes a new note, which is a single note for this SR. The corresponding current BSEP SR 3.5.2.4 already has an existing note. Therefore, it is proposed that the existing note be retained but relabeled as Note 1, and the new note labeled as Note 2. The General Electric BWR/4 STS markup included in TSTF-523 includes changes for TS 3.6.2.4, RHR Suppression Pool Spray. BSEP does not have this TS; therefore, these TSTF changes are not applicable to BSEP. The proposed numbering of the SRs in CNS TS 3.6.6 is different than the numbering of the SRs in the Westinghouse markup included in TSTF-523 for TS 3.6.6, due to the existing CNS SRs being different than the STS SRs. The STS markups included in TSTF-523 for various TSs propose to insert a new SR in the middle of a string of SRs, and renumber the remaining existing SRs in the string. In lieu of this approach, it is proposed to insert new SRs at the end of the string of SRs, eliminating the need to renumber existing SRs. This proposed approach should minimize future implementation activities. ONS has separate TSs for HPI (TS 3.5.2) and LPI (TS 3.5.3), rather than a combined ECCS TS. Therefore, the Babcock & Wilcox STS markup included in TSTF-523 for SR 3.5.2.2 and SR 3.5.2.3 is captured in the proposed changes to ONS SR 3.5.2.1, SR 3.5.2.2, SR 3.5.3.1, and SR 3.5.3.2. The HPI or LPI nomenclature, as appropriate, is proposed in lieu of the ECCS nomenclature used in the TSTF. The current ONS SR 3.6.5.1 applies to both the reactor building spray system and the reactor building cooling system, whereas the Babcock & Wilcox STS markup included in TSTF-523 for SR 3.6.6.1 applies only to the containment spray system. Therefore, the new note is reworded to clarify that it applies only to the reactor building spray system, consistent with the intent of the TSTF. The Babcock & Wilcox STS markup included in TSTF-523 for SR 3.6.6.4 uses the nomenclature containment spray. Consistent with the remainder of ONS TS 3.6.5, the corresponding change that adds ONS SR 3.6.5.9 proposes the use of the nomenclature reactor building spray. None of these differences affect the applicability of TSTF-523 to the individual plant TS. Under TSTF-523, new Surveillance Requirements (SR) are added to various systems, requiring that locations susceptible to gas accumulation are sufficiently filled with water, and the Surveillance Frequency is set at 31 days, or, if the plant has adopted the Surveillance Frequency Control Program (SFCP), the intent is that the initial frequency be established at 31 days under the SFCP. An alternative is proposed for the following systems: RA-15-0006 Page 5 of 9 Plant System Surveillance BSEP ECCS - Operating (TS 3.5.1) SR 3.5.1.1 BSEP RCIC System (TS 3.5.3) SR 3.5.3.1 BSEP RHR Suppression Pool Cooling (TS 3.6.2.3) SR 3.6.2.3.3 HNP Containment Spray System (TS 3.6.2.1) SR 4.6.2.1.e CNS Containment Spray System (TS 3.6.6) SR 3.6.6.8 MNS Containment Spray System (TS 3.6.6) SR 3.6.6.8 ONS Reactor Building Spray (TS 3.6.5) SR 3.6.5.9 Gas accumulation for these systems is currently monitored in accordance with each sites response to NRC Generic Letter (GL) 2008-01. Review of plant experience over a three year period ending on April 30, 2015 shows no instances of gas accumulation which would challenge the capability of these systems to perform their safety function. Specifically: At BSEP, a review of trending data collected in the three year period ending on April 30, 2015 indicates that no gas has been detected via venting. Venting is currently being performed on a monthly basis at eleven locations on each Unit: two locations on RHR loop A, three locations on RHR loop B, two locations on each loop of the Core Spray System, one location on the High Pressure Coolant Injection (HPCI) System, and one location on the Reactor Core Isolation Cooling (RCIC) System. At HNP, a review of trending data collected in the three year period ending on April 30, 2015 indicates no gas has been detected, with two exceptions (described below). UT has been performed on a quarterly basis during this period at five locations on Containment Spray System Train A, five locations on the Containment Spray System Train B, and on one location common to both trains. o At the location common to both trains, there is a small permanent void near valve 1CT-12 that was first identified in February 2008 and entered into the Corrective Action Program. The volume of the void was originally estimated as 257 in3, but was subsequently reduced to approximately 10 in3 with the installation of a local vent valve in 2010. It is believed that the void is air that was introduced during 2007 outage test conditions that have since been corrected to preclude air intrusion. An evaluation of the original void determined that the system remained able to perform its safety-related function. A design calculation was also performed to establish the allowable void size at this location. The allowable void size was greater than 257 in3. o At the location near valve 1CT-50 on Train A, there was a void of approximately 85 in3 identified in August 2013 and entered into the Corrective Action Program. The cause of the void could not be determined conclusively. The void most likely formed from a collection of smaller voids created on the discharge of the A Containment Spray pump when the system was re-filled in May 2013. These RA-15-0006 Page 6 of 9 voids could have been too small to detect at the time, but then, over time, collectively migrated to the system high point at 1CT-50. The void was removed. An evaluation determined that the system remained capable of performing its safety-related function with this void present. At CNS, a review of trending data collected in the three year period ending on April 30, 2015 indicates that no gas has been detected via either UT or venting. UT is currently being performed on a quarterly basis at two locations on each of two trains of the Containment Spray System on Unit 1 and at three locations on each of two trains of the Containment Spray System on Unit 2. Prior to implementation of UT in July 2012 on Unit 1 and September 2012 on Unit 2, venting was performed. At MNS, a review of trending data collected in the three year period ending on April 30, 2015 indicates that no gas has been detected. UT has been performed on a quarterly basis during this period at a single location on each Containment Spray System train (two trains per unit). At ONS, a review of trending data collected in the three year period ending on April 30, 2015 indicates that no gas has been detected. UT has been performed on a monthly basis during this period (excluding outages) at seven locations on the Reactor Building Spray System of Unit 1 (two on Train 1A and five on Train 1B), eight locations on the Reactor Building Spray System of Unit 2 (two on Train 2A and six on Train 2B), and five locations on the Reactor Building Spray System of Unit 3 (two on Train 3A and three on Train 3B). Therefore, a Surveillance Frequency of 92 days is considered reasonable to provide assurance that these systems are sufficiently filled with water. For plants that have not adopted the SFCP (BSEP and HNP), the TS markups provided in Attachment 1 reflect this variation. For plants that have adopted the SFCP (CNS, MNS, and ONS), the initial frequency established under the SFCP will be 92 days, and subsequent changes will be controlled under the provisions of the SFCP. Similar variations to the Surveillance Frequency have been proposed by other licensees in LARs proposing to adopt TSTF-523. See References 1, 2, and 3. provides markups of the TS Bases pages, which correspond to the proposed TS changes. The TS Bases markups were developed using TSTF-523, but include enhancements. Specifically: For all plants, clarification is added that if the accumulated gas is eliminated or brought within the acceptance criteria limits as part of the Surveillance performance, the Surveillance is considered met and the system is OPERABLE; past operability is then evaluated under the Corrective Action program; and if it is suspected that a gas intrusion event is occurring, then this is evaluated under the Operability Determination Process. For the discussion of RHR Shutdown Cooling System SR 3.4.7.2, SR 3.4.8.2, SR 3.9.7.2, and SR 3.9.8.2 in the BSEP TS Bases, additional detail is added regarding how the surveillances are satisfied for the suction flow path, consistent with the site specific plant configuration and existing operational practices. RA-15-0006 Page 7 of 9 These TS Bases enhancements are administrative and do not affect the applicability of TSTF-523 to the individual plant TS. Changes to the existing TS Bases will be implemented under the Technical Specification Bases Control Program, and are provided in Attachment 3 for information only.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination Duke Energy Progress, Inc., and Duke Energy Carolinas, LLC, collectively referred to henceforth as Duke Energy, requests adoption of TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation, which is an approved change to the standard technical specifications (STS), into the Brunswick Steam Electric Plant, Unit Nos. 1 and 2; Shearon Harris Nuclear Power Plant, Unit 1; Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Units 1 and 2; and Oconee Nuclear Station, Units 1, 2, and 3 Technical Specifications. The proposed change revises or adds Surveillance Requirements to verify that the system locations susceptible gas accumulation are sufficiently filled with water and to provide allowances which permit performance of the verification. Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No. The proposed change revises or adds Surveillance Requirement(s) (SRs) that require verification that the Emergency Core Cooling System (ECCS), the Decay Heat Removal (DHR) / Residual Heat Removal (RHR) System, the Containment Spray / Reactor Building Spray System, and the Reactor Core Isolation Cooling (RCIC) System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. Gas accumulation in the subject systems is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The proposed SRs ensure that the subject systems continue to be capable to perform their assumed safety function and are not rendered inoperable due to gas accumulation. Thus, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No. The proposed change revises or adds SRs that require verification that the ECCS, the DHR / RHR System, the Containment Spray / Reactor Building Spray System, and the RCIC System are not rendered inoperable due to accumulated gas and to provide RA-15-0006 Page 8 of 9 allowances which permit performance of the revised verification. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the proposed change does not impose any new or different requirements that could initiate an accident. The proposed change does not alter assumptions made in the safety analysis and is consistent with the safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No. The proposed change revises or adds SRs that require verification that the ECCS, the DHR / RHR System, the Containment Spray / Reactor Building Spray System, and the RCIC System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change adds new requirements to manage gas accumulation in order to ensure the subject systems are capable of performing their assumed safety functions. The proposed SRs are more comprehensive than the current SRs and will ensure that the assumptions of the safety analysis are protected. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, Duke Energy concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified. 4.0 ENVIRONMENTAL EVALUATION The proposed change 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 change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a 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 change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change. RA-15-0006 Page 9 of 9

5.0 REFERENCES

1. Wolf Creek Nuclear Operating Corporation letter, Wolf Creek Generating Station, Application to Revise Technical Specifications to Adopt TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation, Using the Consolidated Line Item Improvement Process, dated November 20, 2014 (ADAMS Accession No. ML14330A247)
2. Virginia Electric and Power Company letter, Surry Power Station Units 1 and 2, Proposed License Amendment Request, Technical Specifications Surveillance Requirement and Basis, Revisions for Generic Letter 2008-01 (Gas Accumulation),

dated January 14, 2015 (ADAMS Accession No. ML15021A130)

3. Dominion Nuclear Connecticut, Inc. letter, Millstone Power Station Units 2 and 3, Proposed License Amendment Requests to Adopt TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation, dated January 15, 2015 (ADAMS Accession No. ML15021A128)

RA-15-0006 Proposed Technical Specification Changes (Mark-up)

RHR Shutdown Cooling System-Hot Shutdown 3.4.7 SURVEILLANCE REQUIREMENTS SR 3.4.7.1 SURVEILLANCE

N 0 TE ---------------------------------

Not required to be met until 2 hours after reactor steam dome pressure is less than the RHR shutdown cooling isolation pressure. FREQUENCY Verify one required RHR shutdown cooling subsystem 12 hours or recirculation pump is operating. s* R 3. +. 1, z - - - - --- r-..J ore- - - _ _ _ _ -, - Brunswick Unit 1 Not ~"e.iul;-c..L -h> be. fer.J.or-Yl"fed LJn-f,*l IZ. ho~.:>rf.. AU.- rcet c.~r ~ +-c~NI ~ f/""tJJure- ,,s ic.S.5 Th&rYV ~ F-Ht-.s4ufJ.,,wYI/ L..O o/1 ~ I~ o Jc, ft drt_, f.A~.S.J vre., v e r**, [J R-H /!.. s ~ viJc*wn t.-c?O I,.~ l ~ v!:,s )'.s ~ I() c~.ttOr:-S SvJ (er+;rle. '-/-o... j tf ~. ~ c c. v/J1 ": J?, +w~ a /' t.- s u-fl-ie-le/l-f/'f.f:d f eJ w,ft.v w/t+c r, 3.4-16 Amendment No.~

RHR Shutdown Cooling System-Cold Shutdown 3.4.8 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown cooling B.1 Verify reactor coolant 1 hour from subsystem in operation. circulating by an alternate discovery of no method. reactor coolant AND circulation No recirculation pump in AND operation. Once per 12 hours thereafter AND B.2 Monitor reactor coolant Once per hour temperature. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one required RHR shutdown cooling subsystem 12 hours or recirculation pump is operating. S(t. 3.4. 2?. l... Vcr.,{l !2..1-1~ shv-l~w Loolt"'J .s Jia.s 7.s -t /YV I tP (A -f I ov: ~ s u.S (._ e r' I ~I(., --& j A.J: 4 c c. v/Vt u /tit --It o>V 4,....1:..-- s;. ~.rl+ ;. '- 1 r ~-t i '{.f-111 ~d. w.-1*~ ""'"' --/r r. Brunswick Unit 1 3.4-18 Amendment No. -~

ACTIONS (continued) CONDITION J. Two or more low pressure J.1 ECCS injection/spray subsystems inoperable for reason_s other than Condition A or B. HPCI System and two or more required ADS valves inoperable. SURVEILLANCE REQUIREMENTS REQUIRED ACTION Enter LCO 3.0.3. SURVEILLANCE SR 3.5.1.1 Brunswick Unit 1 Verify, for each ECCS injection/spray subsystem,.Ahe- -f.Uping is.filled \\otith water from the-p~R=Ip di~\\ -valve tg tR& injeetion vai*Je. J /o~fiCJV\\J S~S.C-ep-/.&G._-/-, J.q. .S. a~ c.. vM u I~~ t!f re..s vf{-;L, t>.A f /'{ +; II e J w ~*ftv w,'kr.. 3.5-4 ECC5-0perating 3.5.1 COMPLETION TIME Immediately FREQUENCY (continued) Amendment No. F

f'Jd re;v;rr=...J..f.v b.t.- wu+ ~:'" Sl.!.-/:rM v~n'-f' -?lrJw f"'fl-,_s o(Jt::M-e J u11ck.r CiJ.MtVIt.rl:r.,-/,V'f., 4M-f-.rc,f, ECCS-Operating 3.5.1 SURVEILLANCE REQUIREMENTS _(continued) SURVEILLANCE FREQUENCY .{ SR 3 5 1 2 CJL~~~;~~~~-r-~-~~~j~~;i~j~~~;(LPC~}-~~b~~~;~-~~---- may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) shutdown cooling isolation pressure in MODE 3, if capable of being manually realigned and -:oo.. not otherwise inoperable.

+--~ ----------------------------------------------------------------------

SR 3.5.1.3 SR 3.5.1.4 SR 3.5.1.5 Brunswick Unit 1 Verify each ECCS injection/spray subsystem manual, 31 days power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position. Verify ADS pneumatic supply header pressure is ~ 95 psig. Verify the RHR System cross tie valve is locked closed.

NOTE-----------------------------

Not required to be performed if performed within the previous 31 days. 31 days 31 days Verify each recirculation pump discharge valve and Once each startup bypass valve cycles through one complete cycle of full prior to exceeding travel or is de-energized in the closed position. 25% RTP (continued) 3.5-5 Amendment No.~

SR 3.5.2.2 SR 3.5.2.3 SURVEILLANCE ECCS-Shutdown 3.5.2 FREQUENCY Verify, for each required core spray (CS) subsystem, 12 hours the:

a.

Suppression pool water level is~ -31 inches; or

b.

NOTE----------------------------

Only one required CS subsystem may take credit for this option during OPDRVs. Condensate storage tank water volume is ~ 228,200 gallons. s sR 3 5 2 4 8 0~~-LPci-~~b~~~~~;-~~~~;-~~~-~id~;~d-O_P_E-RABLi during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable. ~~~-----------j~----------------------------------------------------------------------- Brunswick Unit 1 Verify each required ECCS injection/spray subsystem 31 days manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position. (continued) J loc.ct.-l-tovr.s. svsc.e..f-f, ~It:...f-v J4.S. 4Ll.-V!VI u L,f ioi'V' are_ sv..f+t~, -~,.tly ..t. ;u~J w;-/1-....- Wtt.~r-~ (\\J('+ r~rnre cl..fv ~V'I'Ut-4r-s j.S-1-t ~ Vf/1-t -f: I r) w ~ h s. /)ft'vt~d Vl1ckl' ae:/_MiY'IIS..Jre:.-1-w*e. L-ov,f-ro/. 3.5-10 Amendment No.~

SURVEILLANCE REQUIREMENTS SR 3.5.3.1 SURVEILLANCE Verify the RCIC System ~piRg is filled with v11ater from tl:la pl:lmp discharge valva te tl:le iAjeetion ¥alt,*e. RCIC System 3.5.3 FREQUENCY Verify each RCIC System manual, power operated, 31 days SR 3.5.3.3 Brunswick Unit 1 and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

NC>TE--------------------------------

1.

Use of auxiliary steam for the performance of the SR is not allowed.

2.

Not required to be performed until 24 hours after reactor steam pressure is adequate to perform the test. Verify, with reactor pressure;::: 945 psig and~ 1045 psig, the RCIC pump can develop a flow rate 2:: 400 gpm against a system head corresponding to reactor pressure. ioe-~~.-ft'o.,.,~ ~v~Leff,-.ble.. -fo ja.S acc.uVI-1u lerf1c>YV are svff,.c.../~111-t// t..v,fl-v w~+t!!.r, NO"T7:::-- ----- N o t-r e 1 v u-e.- d. tJ b.L- ~ f- -{,..- & 'f..!~wv "V(vt f- +low pe.-.ff,.,_s Of e vt-e J u VI.J~ r &:~ dM,*VIi.s-1-ra f, ve.. Cc>VJ fro I~ 3.5-13 92 days (continued) Amendment No.~

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling subsystem 31 days manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position. SR 3.6.2.3.2 Verify each RHR pump develops a flow rate 92 days

7700 gpm through the associated heat exchanger while operating in the suppression pool cooling mode.

SR. 3. ~

  • z,:L3 v(_.:...fl ~1--lf... svtfr-t..SiV/1 rcl71 Cfl. cL.;s c~.,ol,~ ~ ub.J;...J.f.t/YI. lo£. ~*fro/l.J'.

.$V.SUf~'~{~ ~. f')fj e:c ~ umvla-lro!L-cdc!.- ,~~J-1/ i c ; PI +II ! I Yt (' J 41 ' *ft.- L.V "~ Brunswick Unit 1 3.6-25 Amendment No.r

SURVEILLANCE REQUIREMENTS SURVEILLANCE RHR-High Water Level 3.9.7 FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem is 12 hours operating. 5/t ;s,q,1,Z. \\/e~,.f7 rc;vtr-<:.J R-1-/J.... s~v+Jows'\\/ 31 dP.'/5 C..ooJ1~.Svbs,f~ Jo ~flo~A.S. S l.).j (."(,,+,~~e._ ~ J 4,J ~ C.C UM vi~ ftC)/1...- are._ sv..ff. /c.i"-r;')f/J.J,"IIed 1-NI~ w,+er, Brunswick Unit 1 3.9-12 Amendment No.~

SURVEILLANCE REQUIREMENTS SR 3.9.8.1 SURVEILLANCE Verify one RHR shutdown cooling subsystem is operating. ( _________ _____ RHR-Low Water Level 3.9.8 FREQUENCY 12 hours .SR.. 3, ~.. 3. Z. V ( r' {'7 !2.. '1-1ft-. s ~ v-.1-J,:n.V 11 c..oo/1~ 3 1 cler7.S S vb ~~.s +-r NV f bt:..q ; cw1 s SvJ c.. e..;.f,*b It:-- ~... / ~..s 4CC v~ vl~<+l or-t..- c1.-"-e__ svf.:f:/r_,;,.,-t/1 -fl/1-eJ wd*4-t-U.-.icr. Brunswick Unit 1 3.9-15 Amendment No.J93'

RHR Shutdown Cooling System-Hot Shutdown 3.4.7 SURVEILLANCE REQUIREMENTS SR 3.4.7.1 Sfl.. 3.4.r.z. Brunswick Unit 2 SURVEILLANCE

N 0 TE -------------------------------

Not required to be met until 2 hours after reactor steam dome pressure is less than the RHR shutdown cooling isolation pressure. FREQUENCY Verify one required RHR shutdown cooling subsystem 12 hours or recirculation pump is operating. - - t-J 071:..- - N,)+ rc7v/rc J.. --h, k p-~rfvrWf<cl LJ 11-/- ~ / i Z.. h I:)U r.S. A'.f.. --k..- r r" A c ~ r- ~ ~~~,.ht. J OYJA-<..- f> n.,s.Jv r~ t~ I c-J.{ +£, Cf #'\\/ -!tv._ !<-H /L s 4 v + ~wvv Vev-,{f P-HI- .S~u+c:4wJ/l ~ 00 /,'"'j .Sv1.5/'S../-r-vl.. loc..n-f/ OY1S.sv..s '-~ r+,l:.ft... 1-.J ~ 11,S acc..v111v IAf, oYV errc_ .S v+f.. t-~, ~.II fly J, 1 J c J.. i..vl~ W-t ht",. 3.4-16 Amendment No. }33

RHR Shutdown Cooling System-Cold Shutdown 3.4.8 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown cooling B.1 Verify reactor coolant 1 hour from subsystem in operation. circulating by an alternate discovery of no method. reactor coolant AND circulation No recirculation pump in AND operation. Once per 12 hours thereafter AND B.2 Monitor reactor coolant Once per hour temperature. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one required RHR shutdown cooling subsystem 12 hours or recirculation pump is operating. ~----------------------------------~--------------~ .St. J,-1.8-l v~/;{7 f-_;*1"-.. s4v!Jt>,~l~ t-c?olt-::J. 31 d~75 .s \\)~ s rs--1-t wv I 0 C;o. +/o~.S S"vJc..-et { {;, fe. ~ Jtr..i .ttC.C.v/1.1v (&tftOi/1./ ef'/'C- ,S~,rff-1-C../;/i+lf /;l[t:J. VI I ~ WA..f.crJ --~---~--~------- Brunswick Unit 2 3.4-18 Amendment No.J213

ACTIONS (continued) CONDITION J. Two or more low pressure J.1 ECCS injection/spray subsystems inoperable for reasons other than Condition A or B. HPCI System and two or more required ADS valves inoperable. SURVEILLANCE REQUIREMENTS REQUIRED ACTION Enter LCO 3.0.3. SURVEILLANCE SR 3.5.1.1 Brunswick Unit 2 Verify, for each ECCS injection/spray subsystem~ pi{;AAg-is-fmed-witR-Wilter-fr-em-tf.te-pt:ffl'l1Tdisehar-ge- -va~eet1eA--velve-:- foe~ -/-,*0~* S\\JJL e_/~' k /<- ..f-o j4...S } 4 c c VM vI a fw!V ~rc_.Sv-f.l..,.. e-,*r11f/y +

n~ J U/ ( +~ w lt. +r,..,

3.5-4 ECCS-Operating 3.5.1 COMPLETION TIME Immediately FREQUENCY (continued) Amendment No. p3

ECCS-Operating re rJI/'~d.. --1-o h ~-r.._k,... $7..s HM 1/-(A-{- 3.5.1 f./.J'-f/ p~ 1-i.,.s 0f't!'"' ~ J v/1 ~r if c!f"1/1.',,*.J-Ira-1-, ~e- ~'ira/, SURVEILLANCE REQUIREMENTS _(continued) SURVEILLANCE FREQUENCY SR 3 5 1 C) L~~;;.~~~-.;;,~~~~7~£-(t.PCii~~b~;~;;;;;--- may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) shutdown cooling isolation pressure in MODE 3, if capable of being manually realigned and

s~::... not otherwise inoperable.

SR 3.5.1.3 SR 3.5.1.4 SR 3.5.1.5 Brunswick Unit 2 Verify each ECCS injection/spray subsystem manual, 31 days power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position. Verify ADS pneumatic supply header pressure is ~ 95 psig. Verify the RHR System cross tie valve is locked closed.

NOTE--------------------------------

Not required to be performed if performed within the previous 31 days. 31 days 31 days Verify each recirculation pump discharge valve and Once each startup bypass valve cycles through one complete cycle of full prior to exceeding travel or is de-energized in the closed position. 25% RTP (continued) 3.5-5 Amendment No. ~

SR 3.5.2.2 SR 3.5.2.3 SURVEILLANCE ECCS-Shutdown 3.5.2 FREQUENCY Verify, for each required core spray (CS) subsystem, 12 hours the:

a.

Suppression pool water level is 2::-31 inches; or

b.

NOTE-----------------------------

Only one required CS subsystem may take credit for this option during OPDRVs. Condensate storage tank water volume is 2:: 228,200 gallons. sR 3 5 2*(0 o-~;-LPci-~~b~;~~~~-~~~N~T~~~~~~~;;~-0-PE-~ABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

'17"'"

Brunswick Unit 2 Verify each required ECCS injection/spray subsystem 31 days manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position. ~ (continued) )iec.q-/;".,..s !v.J {.-el..f,!,(e__.../o Je~..l a c~ VIY' vi~ ~I f);"l.,..- et re.. .S *.. ) J/..; 'c__; ~/Iffy f-, lf~J ""'.,-ft.._ w 4.fr r. t.. * /0,-/- re. rJ,;.C-j h ~ ~+ fr,.-- s '/' 4-t-NV v r11t +ivw f#f +h..s o;el/1 <" d v11 ~J.~,- ~ d""' ;11 I~ f,.q~hve.. LoVJ+rol, 3.5-10 Amendment No. ~

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.5.3.1 Verify the R C I C System ~Jotetnet-ts-iffffeG-IAAm--w:dl4ef-ffeflV the pump discharge \\(alve to the injec;tion-va~ve:' RCIC System 3.5.3 FREQUENCY Verify each RCIC System manual, power operated, 31 days SR 3.5.3.3 Brunswick Unit 2 and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

N()TES----------------------------

1.

Use of auxiliary steam for the performance of the SR is not allowed.

2.

Not required to be performed until 24 hours after reactor steam pressure is adequate to perform the test. Verify, with reactor pressure~ 945 psig and~ 1045 psig, the RCIC pump can develop a flow rate ~ 400 gpm against a system head corresponding to reactor pressure. /oc.'f-ltoVI--J. Sv.SLe-/~' !:,lc.-..Jo J_"' ~ 4'~c.. vMvLet+IOI/V ~Ire.. SuH-ic_l~.,tff +1 J1<"d w;tk.. WA*r, f'-..J olt - N o"t rci',.nrc J ~ ~ Vl'\\..2+ -h,,. ~~J --1-rwt Vf,.,-(- +ltJw f""+IA_s. Df'f!'V/ ~ d_ vl7 d.t /" <::~ ~ .M '~ 1..sf~+l v"e.- L oo/lf ro /. 3.5-13 92 days (continued) Amendment No. ~

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling subsystem 31 days manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position. SR 3.6.2.3.2 Verify each RHR pump develops a flow rate 92 days

7700 gpm through the associated heat exchanger while operating in the suppression pool cooling mode.

C:::---------------- .S-t 3.l.. c?.:J.J Ve~*'l t.-H). s.{/'_-~.Jiu/1 /' ol co~..,i;J .SJJ:,.s!J k.v,.. loc~f,v,..J.sv.Jc~rhJ-,4.., h ( 1-'1 j. ~' c.. t..NV' J i ~~*-h,;'YV &rl' <,.... ~viJ.,*c.*, -r:"'-1.// *ft. II r-J wr-ft..... Wtt*r* Brunswick Unit 2 3.6-25 Amendment No.~

SURVEILLANCE REQUIREMENTS SURVEILLANCE RHR-High Water Level 3.9.7 FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem is 12 hours operating. .SR.. 3.9~7-l. Vc~'J.., reJvv'*cd.R...I--1).. slt0cloJWVL/ 31 J...l.s ~oo/,~..sv.b.s 1 s~""'V I o c..~ *hl)~~-r .sv..s~*~-t..J-'b k ~ J4.t ace. vM,.;/ ~-holl'\\./ er /' c.. s u-1-f-t "c._; ; "' -1 / y -f.;~, / ~ J L-V 1 '/0 w-t-1-cr, Brunswick Unit 2 3.9-12 Amendment No.~

SURVEILLANCE REQUIREMENTS SURVEILLANCE RHR-Low Water Level 3.9.8 FREQUENCY SR 3.9.8.1 Verify one RHR shutdown cooling subsystem is 12 hours operating. S). 3, 9,'fr.l V~;,.rr ~4ft-sh~~'</vt.. Coo/':] 3 I cltx./,5 ~,/:::.~ y..s+rNV /oc&~--/,cJ 1..s Sv.ice,-1-;b/-e., .f.o J tt.s 4 cc vM v /.:. +~iJ""'-" a.r ~ S v-lf-t~l*,/11// fl'll~cl (/-.)1-/4 w.:ofrr.. Brunswick Unit 2 3.9-15 Amendment No.~

REACTOR COOLANT SYSTEM HOT SHUTDOWN SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump(s). if not in operation. shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.3.2 The required steam generator(s) shall be determined OPERABLE by verifying wide range CWR) secondary side water level is greater than 74% or narrow range (NR) secondary side water level is greater than 30% at least once per 12 hours. 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours. -4.f./.. 3 ~ 4-Ver',{7 re1u;rcd P-1-1-(L /oof /oce><-1-to;.,,s S v.i c. e p--1-' b I~ -/-o J a.s ~c C.VM u l ~ ft on._.- a l"c__ s 1..) +f...i Lt ~r. t/{" .f,-/1.-d w1ff..v wpff~,- t\\+ ~~sf OI1C--!... /J-<'l ]/ cJ_.... '(S. ~ ~ NoT f' e_ JV I r'c J ~ ~ f t" r..f-a f' W1 ~ J u/1-f< / I 2-l1 c? U/' s a...+~r O\\ +~r ;~ Mob£. SHEARON HARRIS - UNIT 1 3/4 4-5 Amendment No. } *6

REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation*. and either:

a.

One additional RHR loop shall be OPERABLE**. or

b.

The secondary side water level of at least two steam generators shall be greater than 74% wide range <WR) or greater than 30% narrow range <NR). APPLICABILITY: MODE 5 with reactor coolant loops filled***. ACTION:

a.

With one of the RHR loops inoperable and with less than the required steam generator water level. immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water level as soon as possible.

b.

With no RHR loop in operation. suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours. 4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and ~c lating reactor coolant at least once per 12 hours. /uJe!-1 SP-4, 4-- I. 4, J _) {&<fi-ne-~ J "The RHR pump may be deenergized for up to 1 hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration. and (2) core outlet temperature is maintained at least 10°F below saturation temperature.

    • one RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.

~.

      • A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 325°F unless the secondary water temperature OPeach steam generator is less than 50°F above each of the Reactor Coolant System cold leg temperatures.

SHEARON HARRIS - UNIT 1 3/4 4-6 Amendment No. Jl6

5 l-f E" AIL OJ-,) l-.1 A R-/L I.s *- u j..J IT I AJJet-1 Sfl-4~ 4-I ~ 4-, I. J 4.f. I' *L I. 3 v ;__,.' +r r-e.l'/'rf.J f!-NI-I <>Of I 0 <~ -J., ~n.J -Su..s L.e f..f-,.b/e.. ~ Ja~ llf cc v.111 v /"' +;o/'1./ .1'/~ .. sv-JI-/c.. 1~A~~ -{:, 1/-.: j w ".fk. 4.1, fer- "' f /..,_. J t-on u... j>-<-r

5 I c!. *7 s

REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two residual heat removal (RHRl loops shall be OPERABLE* and at least one RHR loop shall be in operation. APPLICABILITY: MODE 5 with reactor coolant loops not filled. ACTION:

a.

With less than the above required RHR loops OPERABLE. immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.

b.

With no RHR loop in operation. suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. SURVEILLANCE REQUIREMENTS 4.4.1.4.~least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.

4. 4, I, + * 'Z.. 2.. Ve ;~ f: 7 I'~) *.. " r ~ cJ. :-1-R 1 *~p / o '-*.f..~" J sv.< t..L/ f, ~*!....

+o Je!-1 "fCC.vMI.)[Pf\\O"V.(,re__.SI..)-/+1'.. 1~/It/f +}tlt:J WI~ We>*/ o+-- /us+ ovpu.. r~r.31 c:l-.!.5.

  • One RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.

HThe RHR pump may be deenergized for up to 1 hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration. and (2) core outlet temperature is maintained at least l0°F below saturation temperature. SHEARON HARRIS - UNIT 1 3/4 4-7

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) CP&L Valve No. lSI-107 lSI-86 lSI-52 lSI-340 lSI-341 lSI-359

b.
c.

EBASCO Valve No *. Valve Function Valve Position 2SI-V500SA-l High Head Safety Injection to Closed Reactor Coolant System Hot Legs 2SI-V501SB-l High Head Safety Injection to Closed Reactor Coolant System Hot Legs 2SI-V502SA-l High Head Safety Injection to Closed Reactor Coolant System Cold Legs 2SI-Y579SA-l Low Head Safety Injection to Open Reactor Coolant System Cold Legs 2SI-V578SB-l Low Head Safety Injection to Open Reactor: Coolant System Cold Legs 2SI-V587SA-1 Low Head Safety Injection to Closed Reactor: Coolant System Hot Legs At least once £"'- 5 lcc..dlo"..f.5UJ.t:."'/-/.,!::,/c.. -/.v Jq.S.., a o,#'lt tJI" 1-' CA.,. by """Yet\\£ i Ag

2.

Verifying that ~e-&GGS pipiAI ia fwll of water -acc*ssi~le -ieahaF,I JPipinl high poinra, and ef/'~ Sv/fi 'C-1 -t-.dfr..f=i I < J,VI-1£;.. w.. f-<.r Vertfying that each valve {manual, power-operated, or automatic) in the flow path that is not locked, sealed, or:/~~ise secured in position, is in its correct positio~ By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is pre1ent in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1.

For all acces1ible areas of the containment prior to establish* ing COITAIHMENT INTEGRITY, and

2.

Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGIITY is established. )F: Nv-r.,-e i~;*.,.< J ¥.u ~ ~ ~~ s1s+r-¥- Vf/!t +I c)~ 1.,. -fh; IJf~ll-e d v1.Ju- ~ J/11,~,' 5 +-r~.-+, vc.... (.....rJVfi-n/, SHEARON HARRIS - UHIT 1 3/4 5-4 Alllendment No *,,)'(

CONTAJNMFNT SYSTEMS 3/4 6.2 DEPRESSURIZAT10N AND COOLlNG S YSTE ~S CONTA I NMCN r SPRAY S Y STE~1 LIMITING CONDITION FOR OPERATION

3. 6. 2 1 Two 1 ndependent Conta 1 nment Spr*ay Systems stla 11 be OPERAS E Wl t h each Spray System capabl e of taking sucti on from the R~JST and transfer-ring suction to t!1e conta r nment sump.

APPL ICAB1LITY. MODES 1. 2. 3. and 4. ACTION: W1th one Conta1nment Spray System 1noperable. restore the inoperable Spray System to OPERABLE status w1t11111 72 l'1ours or' be nr at 'least HOl STANDBY with1n t he next 6 hour*s. restore the inoperable Spray System to OPERABLE status withi n t ile next 48 hollrs or* be in COLD SHUTDOWN w1 thin the fo 11 ow1 ng 30 hours Refer also to Spec1 f1cat10n 3 6.2.3 Act 1on SURVEILLANCE REQUIREMENTS 4.6.2 1 Each Containment Spray System s11all be demonstrated OPERABLE: a At 1 east once per 31 days by ven fy1 ng that each va 1 ve Crnanua 1. power-operated. or automat1c) 111 the flow path that 15 not locked. sealed. ~ o~~~~ secured 1 pos1tion. 1s 1n its cor*rect pos1t1orr: ~ b By ver1fy1r19 t hat. on an inq1cated rec1rculat10n flow of at least 1832 gpm. each pump develops a d1fferential pressure of greater than or equal to 186 ps1 when tested pursuant to the Inservice Test1ng Program: c At least once per 18 months by. 1 Ver1fy1ng that each automat1c valve 1n the flow path actuates to 1ts correct posit10n on a conta1nment spray actuat1on test s1gnal and

2.

Verifying that each spray pump starts automat1ca1ly on a conta1nment spray actuat10n test s1gnal 3 Yer1fy1ng that. colllCldent w1th an 1ndtcat10n of conta1nment spray pump runmng. each automat1c Valve from the sump and RWST actuates to 1ts appropnate pos1t1on following an RWST Lo-Lo test s1gnal d At least once per 10 years by perfor'mlllg an air or smoke flow test through each spr*ay header and ven fy1 ng each spray nozz 1 e 1 s unobstructed SHEARON HARRIS UNIT 1 3/4 6 11 Amendment No ~

1 t0J cC/ sf-. 4. &. z. 1 e. A+ I e-c..s + o VI u f-er 9 z_ J. c. '( s j, 7 v< r;..t: 7 1~ -{-{,0'-f C<Vl-hrt/i IMf'/1 t.Sf r~ r I o Lt:l --4 o-"~..S ~ v.s c. e f-1-' *~:.I e.- -f.o J,., s. tl c c. VM u { ""ft o -1\\....- a rc_ S vff, ~~ r/1-f ( f +, -/{ c.1 {.;v-i -fC..., L-V lA --f..c r-, ~ t-Jo+ r-e7v1~cJ -fv !....._ ~+ f-or .. '(s-1-<""--' if-( /1 + f-1 ~fA.) f,., + l-?.s 0 f< VI '('./_ V 11 j_ f:.. r PI J. IMI/1 ~-~+n-f,'v'- [dvl{ ro(..

REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.* APPLICABILITY: MODE 6, with irradiated fuel in the vessel when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet. ACTION: With no RHR loop OPERABLE and in operation. suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. SURVEILLANCE REQUIREMENTS \\ 4.9.8. ~ t least one RHR loop shall be verified in operation and circulating reactor oolant at a flow rate of greater than or equal to 2500 gpm at least e er 12 hours.

4. 9.. ~, I, l-

\\1 ~;' /'( r~ J :.nr-e.l P-HI-loof I oc.--r+tov'I.J .SUS c ~If, ~I e.. /o JerJ accv.M vlt:oito/1.- pr-e.. Svffc"c..,'fY1+/.1 -h 1/c J wi-lh w~-hl' ,. f /-t>cA s+ o 11 c.A- ~~ r 3 I cJ., (.s, ~---------------- The RHR loop may be removed from operation for up to 1 hour per 2-hour period during the performance of CORE ALTERATIONS and core loading verification in the vicinity of the reactor vessel hot legs. SHEARON HARRIS - UNIT 1 3/4 9-9

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent residual heat removal CRHR> loops shall be OPERABLE.and at least one RHR loop shall be in operation.* APPLICABILITY: MODE 6. with irradiated fuel in the vessel when the water level above the top of the reactor vessel flange is less than 23 feet. ACTION:

a.

With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status or to establish greater than or equal to 23 feet of water above the reactor vessel flange as soon as possible.

b.

With no RHR loop in operation. suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. SURVEILLANCE REQUIREMENTS 4.9.8.2.1 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2500 gpm at least once per 12 hours whenever the water level is at or above the reactor vessel flange. 4.9.8.2.2 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 900 gpm at least once per 12 hours whenever the water level is below the reactor vessel fl nge. ~----------------------------------------~~ Ver*, 17 !2-H JL f oof [ e:, c.-q..J--i o/1-J sv.S{ i 1-t; ~~ ~ J.If f

  • otCC V ll*i£/ l 01~1/Jt'\\_...

eft l'l. SJJf I ;_ I*Ot f [ f + ~ -{ ( f d i.,V1 f£, W u/.c r-c;._ f / t:As-f o 11 (...(_ 1-<r 31 c1e.75,

  • The operating RHR loop may be removed from operation for up to 1 hour per 2-hour period during the performance of CORE ALTERATIONS and core loading verification in the vicinity of the reactor vessel hot legs.

SHEARON HARRIS - UNIT 1 3/4 9-10 Amendment No. Y

SURVEILLANCE RCS Loops - MODES 4 3.4.6 FREQUENCY SR 3.4.6.3 Verify correct breaker alignment and indicated power are available to the required pump that is not in operation. In accordance with the Surveillance Frequency Control Program .5 R 3, +I'.4-NOTE - Not-re-7 u1"rc..-J -h b.t. peri-CJr~ e cl lJ 11fi 1 12.. i-t ovr.r "-f -hr-t'l'l c r ':} ji/10bt.f-, V<"r~.[ r-cJ'-';r~ J P-1-lf.- loor

r

{oc~~+,di1.J s.use-ep-1-ibf..t., i-e Jtt.s ~ CC UM l) l t1-i I 0 t'V a /'C-.S\\/-fft*C. t -1"/1" ( f + i -, I c J w.-+ ~ w A..J.r,-. T11 4.'Co('d lf/lc.-t..-- wtf/,.__ ~ .S V~"V ~;-J/ 111/1 C-<...-- 1-reJ'J~ncf {._o,., l r" I P rcJJ /'.; M/ Catawba Units 1 and 2 3.4.6-3 Amendment Nos.~

RCS Loops - MODES 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS (continued) SR 3.4.7.3 SURVEILLANCE Verify correct breaker alignment and indicated power are available to the required RHR pump that is not in operation. Sit 3,4,7. 4-vf:/' .~.r.7 r-r:l~.,c J f'.l-IR-loor Joc...~+roll'l..S _ sv.s.c~r+.bk,.../., J ~..s &i cc VA'f v t~ +* (H'\\..- 6f ;e._..Sv +F L~t ~"'t: I r f;I(~J w-~~ IA.J~~r-FREQUENCY In accordance with the Surveillance Frequency Control Program .L t1 "" ( c c,..cJ *l 11 C-<.- w,1tv ~ .Su rv-r i /I Jt'l fA-; ~-~ r.lt'/1 c 'f (,J,*i-ra I P r~ r 01 M-- Catawba Units 1 and 2 3.4.7-3 Amendment Nos.~

RCS Loops - MODES 5, Loops Not Filled 3.4.8 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Required RHR loops inoperable. B.1 Suspend operations that Immediately would cause introduction of No RHR loop in operation. SURVEILLANCE REQUIREMENTS coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1. B.2 Initiate action to restore one RHR loop to OPERABLE status and operation. SURVEILLANCE SR 3.4.8.1 Verify one RHR loop is in operation. SR 3.4.8.2 Verify correct breaker alignment and indicated power are available to the required RHR pump that is not in operation. SP.. J. 4.8.;, Vt!;,J7 R.~R._ loor I ~C-t;{+IIJ/\\-S..sv..sc e r"l ~ ~ i>.J e~J e:(CCI..) 'M\\} L&t-/,,Y'V al'c_ svfft~;-</7-1/r J1'II(J vu~flv 4/et..frr. Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Wl-/fv ~

  • ~ Jrve, lf~/1(.;(../

r=:rerveA (r C-o /1-/r.., I P r~, "~ M.;- Catawba Units 1 and 2 3.4.8-2 Amendment Nos.~

SURVEILLANCE REQUIREMENTS SURVEILLANCE SA 3.5.2.1 Verify the following valves are in the listed position with power to the valve operator removed. Number Position Function N1162A Open Sl Cold Leg Injection N1121A Closed Sl Hot Leg Injection NI152B Closed Sl Hot Leg Injection NI183B Closed RHR Hot Leg Injection NI173A Open RHR Cold Leg Injection N11788 Open RHR Cold Leg Injection N11008 Open Sl Pump Suction from RWST Nl1478 Open Sl Pump Mini-Flow SA 3.5.2.2 :a Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position. ~------------------~--~ SA 3.5.2.3 .Verify EGGS pipiRg is fyll ef water. r v~.. -'lfl EC.C....S loc... f,o.-,.s.su.sc: ~ r+,Jie..f.o ~14.5 4£.C.l..l~ vLQ-ftol'v etl'~ Sv~ic., ~..., tlr + ,1/~ J w1-f~ wet+-~r, SA 3.5.2.4 Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head. .,... ---.--- N oT'E' - t0 o i-r- ~ 7 v 1*, I! J.. k ~'/.{:.,,.... s ;sh,.,_ vr/lr f-Low p~+hs or~vr-< c1 v,dc..- a cJ M 111 t.1 +r.. ~ \\I'C-L. tJ vt fr<> J, ECCS - Operating 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the lnservice Testing Program (continued) Catawba Units 1 and 2 3.5.2-2 Amendment Nos.~

Containment Spray System 3.6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System LCO 3.6.6 Two containment spray trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. One containment spray A.1 Restore containment spray train inoperable. train to OPERABLE status. B. Required Action and B.1 Be in MODE 3. associated Completion Time not met. AND B.2 Be in MODE 5. SURVEILLANCE REQUIREMENTS SURVEILLANCE r SR 3.6.6.1 4 Verify each containment spray manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position. ,.._j O-r-£" - N~ r r c. i VIr('.I -I, !;u_ W\\...t+.h.~ S'!JkM_, \\rUij- +trj7-AI p~fl-,j t>ff'"'t!"J V"1~r 01 J /111111,....tt-r.. +t 'vc. (....,(f"'ttr-1-ru{

  • COMPLETION TIME 72 hours 6 hours 84 hours FREQUENCY In accordance with the Surveillance Frequency Control Program (continued}

Catawba Units 1 and 2 3.6.6-1 Amendment Nos. ~

Containment Spray System 3.6.6 SR 3.6.6.2 SURVEILLANCE Verify each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head. SR 3.6.6.3 Deleted. SR 3.6.6.4 Deleted. SR 3.6.6.5 SR 3.6.6.6 SR 3.6.6.7 Verify that each spray pump is de-energized and prevented from starting upon receipt of a terminate signal and is allowed to manually start upon receipt of a start permissive from the Containment Pressure Control System (CPCS). Verify that each spray pump discharge valve closes or is prevented from opening upon receipt of a terminate signal and is allowed to manually open upon receipt of a start permissive from the Containment Pressure Control System (CPCS). Verify each spray nozzle is unobstructed. v~r:~l ~.;*l!+*fln~ spra7 ~{)C"~.f .$(.)J'~f4--.fo~.~,rll lf_CC VM V~~~

o. ('-{_

~vi-f.*,~; 1/1-t/T.f-, II c J w ~..flv wu-/--rr-; FREQUENCY In accordance with the lnservice Testing Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Following activities which could result in nozzle blockage L1.iffC or d..,/7c...<.- t.v.~ --rtu.- Sv.rve

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L.t?/11rul pro r~;/VV Catawba Units 1 and 2 3.6.6-2 Amendment Nos. ~

RHR and Coolant Circulation-High Water Level 3.9.4 ACTIONS CONDITION A. (continued) REQUIRED ACTION A.4 Close all containment penetrations providing direct access from containment atmosphere to outside atmosphere. SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.9.4.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of 2! 1 000 gpm and RCS temperature is.:s. 140°F. COMPLETION TIME 4 hours FREQUENCY In accordance with the Surveillance Frequency Control Program c------~----~-- .Srt 3.'1.4-,Z. V-<:r.,f/. rr:J~Jr-e J P-HA.. loof /ccil.. ;,O:W s v.S(. er+ 1 b/ t- -+o J 4J ~ cc... vAf ui... --!, ~/V .art: S v-l.J-1 c...*, ~/If-17 ..f-t'J/~ d WJ~ f..>> ~+er, ~ Wl1k.. ~ s v ""'; II et VI i..A--- 1-I'Y: 7 Ut:..vr t'/ Co /li r" / P rrlj /'a.,._,- Catawba Units 1 and 2 3.9.4-2 Amendment Nos. ~9

RHR and Coolant Circulation-Low Water Level 3.9.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) 8.2 Initiate action to restore Immediately one RHR loop to operation. SURVEILLANCE REQUIREMENTS 8.3 Close all containment penetrations providing direct access from containment atmosphere to outside atmosphere. SURVEILLANCE SR 3.9.5.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of 2:. 1 000 gpm and RCS temperature is.=:. 140°F. SR 3.9.5.2 Verify correct breaker alignment and indicated power available to the required RHR pump that is not in operation. 5.!Z. 3. CJ. 5~ 3 Vel'*, J7 P-ti!L /o 'f I t>t. a '-ttor~J SvJC.('J-f, /,. ~ -k JRJ &'fCC l-J/141.)[,..--/-t~I'V dN... Sv/fl~tV/flf fl I I(" d w r1 lr_ w ~ a*, 4 hours FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program /,oj 1 ~ -tt...,__ S v/' 1/c; /{ t!"'U.- Fn Jc_re, ( 7 Cr.,'fro( P/'JmiM./ Catawba Units 1 and 2 3.9.5-2 Amendment Nos. ~

ACTIONS (continued) CONDITION REQUIRED ACTION B. One AHA loop B.1 Be in MODE 5. OPERABLE. AND ALL RCS loops inoperable. C. Both required RCS or C.1 Suspend operations that AHA loops inoperable. would cause introduction of coolant into the RCS with OR boron concentration less than required to meet the No RCS or AHA loop in SDM of LCO 3.1.1 and operation. maintain Keff < 0.99. AND C.2 Initiate action to restore one loop to OPERABLE status and operation. SURVEILLANCE REQUIREMENTS SURVEILLANCE SA 3.4.6.1 Verify one AHA or RCS loop is in operation. RCS Loops - MODE 4 3.4.6 COMPLETION TIME 24 hours Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program SA 3.4.6.2 Verify SG secondary side water levels are ~ 12% narrow In accordance with range for required RCS loops. the Surveillance Frequency Control Program SA 3.4.6.3 Verify correct breaker alignment and indicated power are In accordance with available to the required pump that is not in operation. the Surveillance Frequency Control Program McGuire Units 1 and 2 3.4.6-2 Amendment Nos.~ 1

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