ML100890320

Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a License Controlled Program

ML100890320
Person / Time
Site:	Mcguire, McGuire
Issue date:	03/24/2010
From:	Repko R Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML100890320 (311)

Text

REGIS T. REPKO Vice President Pdh Energyo McGuire Nuclear Station Duke Energy MG01VP / 12700 Hagers Ferry Rd.

Huntersville, NC 28078 980-875-4111 980-875-4809 fax regis.repko@duke-energy.com March 24, 2010 10 CFR 50.90 U. S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION: Document Control Desk

Subject:

Duke Energy Carolinas, LLC McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program In accordance with the provisions of 10 CFR 50.90, Duke Energy Carolinas (Duke Energy) is submitting a request for an amendment to the Technical Specifications (TS) for McGuire Nuclear Station (McGuire) Units 1 and 2.

The proposed amendment would modify McGuire's Technical Specifications by relocating specific surveillance frequencies to a licensee controlled program with the implementation of" Nuclear Energy Institute (NEI) 04-10, Revision 1, "Risk-Informed Technical Specification Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies."

The changes are consistent with NRC approved Industry Technical Specification Task Force (TSTF) change TSTF-425, Revision 3 (ADAMS Accession No. ML080280275). The Federal Register notice published on July 6, 2009 (74 FR 31966) announced the availability of this TS improvement.

Attachment 1 provides a description of the proposed changes, the requested confirmation of applicability, and plant specific verifications. Attachment 2 provides documentation of PRA technical adequacy. Attachment 3 provides the existing TS pages marked up to show the proposed change. Attachment 4 provides the proposed TS Bases pages. Attachment 5 provides a cross reference table comparing the TSTF surveillance numbers to the McGuire surveillance numbers. Attachment 6 provides the Proposed No Significant Hazards Consideration.

Duke Energy requests NRC review and approval of the proposed license amendment within one year of this submittal. Duke Energy is requesting a 90 day implementation grace period due to the extensive document changes necessary to implement this license amendment.

www.duke-energy. corn m Q

U.S. Nuclear Regulatory Commission March 24, 2010 Page 2 In accordance with Duke Energy administrative procedures and the Quality Assurance Program Topical Report, this proposed amendment has been reviewed and approved by the McGuire Plant Operations Review Committee.

There are no new commitments being made as a result of this proposed amendment.

In accordance with 10 CFR 50.91, a copy of this proposed amendment is being forwarded to the appropriate North Carolina State officials.

Inquiries regarding this submittal should be directed to Lee A. Hentz at 980-875-4187.

Sincerely, Regis T. Repko Attachments:

1, Description and.Assessment

2. Documentation of PRA Technical Adequacy

3. Proposed Technical Specification Changes

4. Proposed Technical Specification Bases Changes,

5. Surveillance Frequency Cross Reference Table

6. Proposed No Significant Hazards Consideration

U.S. Nuclear Regulatory Commission March 24, 2010 Page 3 cc: w/Attachments L. A. Reyes Administrator, Region II U.S. Nuclear Regulatory Commission Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303 J. H. Thompson Project Manager (McGuire)

U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852-2738 Mail Stop 8-G9A J. B. Brady NRC Senior Resident Inspector McGuire Nuclear Station B. 0. Hall, Section Chief North Carolina Department of Environment and Natural Resources Division of Environmental Health Radiation Protection Section 1645 Mail Service Center Raleigh, NC 27699-1645

U.S. Nuclear Regulatory Commission March 24, 2010 Page 4 OATH AND AFFIRMATION Regis T. Repko affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

Regis T. Repko, Site Vice President Subscribed and sworn to me: mwý 24.2010

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Date 6 0;Mt2/-

Notary Public C)

My commission expires: QJ* /_ 4 '9/

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ATTACHMENT 1 DESCRIPTION AND ASSESSMENT Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Description and Assessment Page 1 1.0 Description The proposed amendment would modify Technical Specifications by relocating specific surveillance frequencies to a licensee controlled program with the adoption of Technical Specification Task Force (TSTF)-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5."

Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5.0, "Administrative Controls."

The changes are consistent with NRC approved IndustrylTSTF Standard Technical Specifications (STS) change TSTF-425, Revision 3 (ADAMS Accession No. ML080280275).

The Federal Register notice published on July 6, 2009 (74 FR 31966) announced the availability of this TS improvement.

2.0 Assessment 2.1 Applicability of Published Safety Evaluation Duke Energy has reviewed the safety evaluation dated July 6, 2009. This review included a review of the NRC staff's evaluation, TSTF-425, Revision 3, and the requirements specified in NEI 04-10, Rev. 1, (ADAMS Accession No. ML071360456).

Attachment 2 includes Duke Energy's documentation with regard to PRA technical adequacy consistent with the requirements of Regulatory Guide 1.200, Rev. 1 (ADAMS Accession No. ML070240001), Section 4.2, and describes any PRA models without NRC endorsed standards, including documentation of the quality characteristics of those models in accordance with Regulatory Guide 1.200.

Duke Energy has concluded that the justifications presented in the TSTF proposal and the safety evaluation prepared by the NRC staff are applicable to McGuire Nuclear Station, Units 1 and 2, and justify this amendment to incorporate the changes to the McGuire Technical Specifications.

2.2 Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision 3, however, Duke Energy proposes variations or deviations from TSTF-425, as identified below.

Attachment I Description and Assessment Page 2

1. Revised (clean) Technical Specification (TS) pages are not included in this proposed amendment given the number of TS pages affected, the straightforward nature of the proposed changes, and the outstanding McGuire amendment requests currently under NRC review that impact some of the same TS pages. Providing only the mark-ups of the proposed TS changes satisfies the requirements of 10 CFR 50.90 in that the mark-ups fully describe the desired changes. This is an administrative deviation from TSTF-425 with no impact on the NRC staff's model safety evaluation dated July 6, 2009. This administrative deviation is consistent with Exelon's Peach Bottom and Oyster Creek License Amendment applications dated August 31, 2009 (NRC Accession No. ML092470153) and October 30, 2009 (NRC Accession No. ML093060126) respectively.

2. Regarding the frequency for TS SR 3.1.4.2, Duke Energy would like to remove the "one time" frequency for Unit 1 (by Amendment 186) since it expired at the end of cycle 13 refueling outage (1999). This is an administrative deviation from TSTF-425 with no impact on the NRC staff's model safety evaluation dated July 6, 2009.

3. Attachment 5 provides a cross reference table between the NUREG-1431 Surveillances included in TSTF-425 versus the McGuire Surveillances included in this amendment request. This cross reference table. highlights the following:

a. TSTF-425 Surveillances with identical corresponding McGuire Surveillance numbers,

b. TSTF-425 Surveillances and corresponding McGuire Surveillances but with differing Surveillance numbers,

c. TSTF-425 Surveillances that are not contained in the McGuire Technical Specifications and therefore not applicable, and

d. McGuire plant specific Surveillances that are not contained in TSTF-425 Surveillance mark-ups but are applicable to this amendment request.

Concerning the above, McGuire Surveillances with identical corresponding TSTF-425 Surveillance numbers (a) are not deviations from TSTF-425.

McGuire Surveillance numbers that differ from the corresponding TSTF-425 Surveillance numbers (b) are administrative deviations only from TSTF-425 with no impact on the NRC Staff's model safety evaluation dated July 6, 2009.

TSTF-425 Surveillances that are not contained in the McGuire Technical Specifications (c) are not applicable to this amendment request. This also includes McGuire corresponding Surveillances that are event driven or performed in accordance with an existing program (safety evaluation scope exceptions).

Not including these TSTF-425 Surveillances is an administrative deviation from TSTF-425 with no impact on the NRC Staff's model safety evaluation dated July 6, 2009.

Attachment I Description and Assessment Page 3 For McGuire plant specific Surveillances that are not contained in TSTF-425 Surveillance mark-ups, but are applicable to this amendment request (d), Duke Energy has determined that the relocation of these Surveillance frequencies is consistent with TSTF-425, Revision 3, and the NRC Staff's model safety evaluation dated July 6, 2009. This includes the scope exceptions documented in Section 1.0, "Introduction," of the model safety evaluation, since the McGuire plant specific Surveillances involve purely fixed periodic frequencies and therefore do not meet any of the four exceptions.

A similar cross reference table comparing the TSTF and plant specific Surveillances was also provided by Exelon's Peach Bottom and Oyster Creek License Amendment applications to relocate specific Surveillances in accordance with TSTF-425 dated August 31, 2009 (NRC Accession No. ML092470153) and October 30, 2009 (NRC Accession No.ML093060126) respectively.

McGuire currently has six license amendment requests (LARs) that are pending NRC review and approval that affect surveillances modified in this LAR. A table of those LARs is provided below. Once final deposition of these LARs is known, McGuire will provide updated TS and Bases pages prior to approval of this LAR.

Date of LAR Affected Surveillances Oct. 2, 2008 SRs 3.6.13.1, 3.6.13.4, 3.6.13.5 and 3.6.13.6. Modifies Ice Condenser Door SR descriptions and deletes 3.6.13.6.

Dec. 1, 2008 SR 3.8.1.4. Modifies minimum EDG day tank level.

July 1, 2009 SR 3.3.1.11. Excore detector replacement modification Sept. 30, 2009 SR 3.6.6.7. Revise spray nozzle inspection frequency.

Dec. 14, 2009 SRs 3.8.4.2 and 3.8.4.5. Revise Battery resistance values.

Dec. 15, 2009 SRs 3.4.16.1 and 3.4.16.3, Adopt TSTF-490 (E-bar).

3.0 Regulatory Analysis 3.1 No Significant Hazards Consideration Duke Energy has reviewed the proposed no significant hazards consideration determination (NSHC) published in the Federal Register on July 6, 2009, 74 FR 31996-32006. Duke Energy has concluded that the proposed NSHC presented in the Federal Register notice is applicable to McGuire Nuclear Station, Units 1 and 2, and is provided as Attachment 6 to this amendment request which satisfies the requirements of 10 CFR 50.91(a).

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ATTACHMENT 2 DOCUMENTATION OF PRA TECHNICAL ADEQUACY Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Documentation of PRA Technical Adequacy Page 1 TABLE OF CONTENTS Section Page 2 .1 O ve rv ie w ....................................................................................................................... 2 2 .2 Histo rical S um m ary ................................................................................................ .. 3 2.3 PRA Technical Adequacy Consistent With RG 1.200, Section 4.2 ................................ 4 2.3.1 PRA Model Adequately Represents the as-built, as-operated Plant .................... 4 2.3.2 Unincorporated Changes to the Plant ............................................................. 6 2.3.3 Departures from ASME Requirements ............................................................. 6 2.3.4 Methodology to be Used for Initiative 5b ........................................................ 6 2.3.5 Identification of Key A ssum ptions ........................................................................ 7 2.3.6 Resolution of Relevant Peer Review/Self-Assessment Findings and O bservations .................................................................................................. . . . .. 7 2.3.7 Applicable Capability Category for Initiative 5b ............................................... 8 2.4 External Events Considerations ................................................................................. 8 2.4.1 Overall External Hazards Analysis Methodology .............................................. 8 2.4.2 McGuire Seismic PRA Model ......................................................................... 9 2.4.3 McG uire Fire P RA Model............................................................................... 10 2.4.3.1 McGuire Future State Fire PRA Model Initiative .... ............ 11 2.4.4 McGuire Shutdown Risk Impact Analysis .......................... .......................... 12 2 .5 S u m m a ry ......................................................... ... ...... ...................................... . . . 12 2 .6 R e fe re n c e s .................................................................................................................. 13 Table 2-1 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD THROUGH ADDENDA RA-Sc-2007 ......... 15 Documentation of PRA Technical Adequacy Page 2 2.1 Overview The technical adequacy of the probabilistic risk assessment (PRA) must be compatible with the safety implications of the proposed Technical Specification (TS) changes and the role the PRA plays in justifying the changes. The Nuclear Regulatory Commission (NRC) has developed regulatory guidance to address PRA technical adequacy, Regulatory Guide (RG) 1.200 (Ref. 1), which references the American Society of Mechanical Engineers (ASME) PRA standard RA-Sb-2005, Addenda to ASME RA-S-2002 (Ref. 2) for internal events at power. External events and shutdown risk impacts may be considered quantitatively or qualitatively. RG 1.200 also references the NEI peer review process NEI 00-02 (Ref. 3).

The industry guidance document for the implementation of Initiative 5b is NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies". The NRC issued a Final Safety Evaluation for NEI 04-10 Revision 0, on September 28, 2006 (Ref. 4). The Staff found that NEI 04-10, Revision 0, was acceptable for referencing by licensees proposing to amend their TSs to establish a Surveillance Frequency Control Program (SFCP), provided that the following conditions are satisfied:

1. The licensee submits documentation with regard to PRA technical adequacy consistent with the requirements of RG 1.200, Section 4.2.

2. When a licensee proposes to use PRA models for which NRC-endorsed standards do not exist, the licensee submits documentation, which identifies the quality characteristics of those models, consistent with RG 1.200, Sections 1.2 and 1.3. Otherwise, the licensee identifies and justifies the methods to be applied for assessing the risk contribution for those sources of risk not addressed by PRA models.

Subsequently NEI 04-10 Revision 1 was approved (Ref. 5) and is the current document of record.

The implementation of the SFCP at the McGuire Nuclear Station will follow the guidance provided in NEI 04-10, Revision 1 in evaluating proposed surveillance frequency changes.

The McGuire PRA is a full scope PRA including both internal and external events (i.e.,

flood, seismic, fire, high winds (tornado)). Having previously completed a self-assessment against the supporting requirements of ASME PRA Standard through addenda RA-Sc-2007 (Ref. 6), Duke Energy is planning to perform a self-assessment against the supporting requirements of ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 (Ref. 7) for the current McGuire PRA model of record (including fire, seismic, and tornado models) in 2010. Also there is currently significant work being performed at Duke Energy in the area of fire PRAs. This will be discussed further in the Fire PRA Model section.

The following information is submitted by Duke Energy to address the conditions of the NRC Safety Evaluation for Initiative 5b.

Documentation of PRA Technical Adequacy Page 3 2.2 Historical Summary The original McGuire PRA was initiated in March 1982 by Duke Power Company staff with Technology for Energy Corporation as a contractor. Law Engineering Testing Company and Structural Mechanics Associates provided specific input to the seismic analysis. It was a full scope Level 3 PRA with internal and external events (i.e., flood, seismic, high winds (tornado), fire). A peer review of the draft PRA was conducted by Electric Power Research Institute's (EPRI) Nuclear Safety Analysis Center (NSAC) in May 1983 (Ref. 8). The final study, which incorporated the comments of the peer review, was completed in July 1984 and resulted in an internal Duke Power Company report (Ref. 9) as Revision 0 to the PRA.

In January 1988, Duke Power Company initiated a complete review and update of the original study. On November 23, 1988, the NRC issued Generic Letter (GL) 88-20 (Ref.

10), which requested that licensees conduct an Individual Plant Examination (IPE) in order to identify potential severe accident vulnerabilities at their plants. The McGuire response to GL 88-20 was provided by letter dated November 4, 1991 (Ref. 11). In this letter Duke Power Company noted that the enclosed Revision 1 of the PRA consisted of a complete Level 3 PRA with a detailed analysis of both internal and external events. By letter dated June 30, 1994 (Ref. 12), the NRC provided a Staff Evaluation of the internal events portion of the above McGuire IPE submittal.

In response to Generic Letter 88-20, Supplement 4 (Ref. 13), Duke Power Company completed an Individual Plant Examination of External Events (IPEEE) for severe accidents. This IPEEE was submitted to the NRC by letter dated June 1, 1994 (Ref. 14).

The IPEEE report contained a detailed write-up of the McGuire seismic and fire PRA analysis methods, results and conclusions. It also addressed other events such as high winds, floods, and, transportation accidents. The IPEEE study did not identify any plant changes that would significantly reduce the risk from external events..

Duke Power Company subsequently responded to an NRC Request for Additional Information (RAI) on the IPEEE submittal November 17, 1995 (Ref. 15). Duke Power Company also submitted a Supplemental IPEEE Fire Analysis Report to the NRC July 30, 1996 (Ref. 16) in response to a request for supplemental fire investigations as described below.

1. Develop fire accident sequences for those areas that were previously screened from further review because they were considered to be subsumed by other initiators.

2. Perform sensitivity studies on fire detection and suppression parameters.

3. Re-review cable routing to confirm that potential plant trip initiators have been considered in all areas.

The supplemental fire investigations in the report produced a more complete quantification of the fire induced core damage sequences but the conclusions and results were not significantly different from those reported in the original IPEEE report.

Documentation of PRA Technical Adequacy Page 4 By letter dated February 16, 1999 (Ref. 17), the NRC provided an evaluation of the IPEEE submittal. The conclusion of the NRC letter [page 6] states:

"On the basis of the overall review findings, the staff concludes that: (1) the licensee's IPEEE is complete with regard to the information requested by Supplement 4 to GL 88-20 (andassociatedguidance in NUREG-1407), and (2) the IPEEE results are reasonablegiven the MNS design, operation, and history.

Therefore, the staff concludes that the licensee's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, and therefore, that the MNS IPEEE has met the intent of Supplement 4 to GL 88-20 and the resolution of specific generic safety issues discussed in the SER."

While the IPEEE Program was a one-time review of external hazard risk and was limited in its purpose to the identification of potential plant vulnerabilities and the understanding of associated severe accident risks, there have not been significant numbers of plant changes made since the initial NRC review that would invalidate the methodologies (discussed later) used in the existing external events models of record.

In 1997, McGuire initiated Revision 2 of the 1991 IPE and provided the results to the NRC in 1998 (Ref. 18). Revision 3 of the McGuire PRA was completed in July 2002 and Revision 3a was completed in February 2005. Revision 3 was a major comprehensive revision to the PRA models and associated documentation. Revision 3a was a minor change to merge the Containment Air Return and Hydrogen Mitigation fault trees into the simplified LERF fault tree. Revision 3a is the current model of record. Work is currently underway on Revision 4 of the McGuire PRA which is a major revision to the PRA and includes a planned revision to the fire PRA model (discussed in Section 2.4.3.1).

2.3 PRA Technical Adequacy. Consistent With RG 1.200, Section 4.2 This section addresses Condition 1 of the NRC Safety Evaluation for Initiative 5b.

2.3.1 PRA Model Adequately Represents the as-built, as-operated Plant The basis to conclude that the PRA model to be used adequately represents the as-built, as-operated plant is as follows.

The existing PRA Configuration Control Program at McGuire was assessed against Section 5 of the ASME PRA Standard to meet the requirements necessary to support risk-informed decisions. The results of the self-asses~ment concluded that the PRA fully meets the requirements for configuration control of a PRA to be used with the ASME PRA Standard to support risk-informed decisions for nuclear power plants. A summary of the program and the basis to conclude that the PRA model adequately represents the as-built, as-operated plant is provided below.

The PRA Configuration Control Program at McGuire is governed by the following workplace procedures.

Documentation of PRA Technical Adequacy Page 5

" XSAA-101, Risk-Impact Review of Nuclear Plant Changes Including Nuclear Station Modification, and Emergency or Abnormal Procedure Changes

• XSAA-106, PRA Maintenance, Update and Application XSAA-101 addresses the process for review of plant design changes, plant emergency and abnormal procedure changes, and Technical Specification (TS) changes that have been made for PRA impacts. It also describes in detail the process used to review the impact of potentially significant changes that could impact the PRA before the changes have been made.

XSAA-106 addresses the conditions when a PRA update may be required (e.g.,

cumulative risk impact of unincorporated PRA changes exceeds a threshold such that the as-built as-operated plant is not adequately represented by the PRA). It addresses a process to assess the risk of a change to the plant and a method to prioritize the implementation of a plant change based on the risk impact to the PRA. It describes a process to ensure that an annual assessment is made of the cumulative impact of PRA changes that have not yet been incorporated into the PRA and provides guidance as to when a PRA update is needed based on the results of the annual assessment. Finally, it describes the electronic tracking tool that is used to track changes that impact the PRA till they are incorporated into the PRA.

The process is as follows. Notification of any completed (and planned changes that could significantly impact the PRA model) plant modifications, Technical Specification changes, or Emergency Procedure changes are sent to the PRA Section for a review of any PRA impacts. This review is documented. If a plant change is determined to impact the PRA then is it entered into the electronic tracking tool where a risk assessment is performed on the change. The outcome of the risk assessment will "bucket" the plant change into a Low, Medium, or High risk change category based on the estimated delta Core Damage Frequency (CDF) or delta, Large Early Release Frequency (LERF) results.

Plant changes that are determined to be of a Low risk impact are tracked to completion in the electronic tracking tool and are annually assessed for their cumulative impact on the PRA model. Plant changes that are determined to be of Medium or High risk impact, are entered into the site corrective action program for further analysis as to their impact on current applications. They also are tracked to completion in the electronic tracking tool and are annually assessed for their cumulative impact on the baseline PRA model.

For any application that requires a PRA analysis (e.g., License Amendment Request (LAR) or Notice of Enforcement Discretion (NOED)) workplace procedures require that all of the outstanding PRA model changes listed in the electronic tracking tool are individually reviewed for their impact on the application. A justification is made as to why each item does not impact the PRA results used to support the application. This review is documented. If it is determined that an unincorporated change might impact an application then steps are taken to either perform sensitivity studies to demonstrate that the contributors significant to the application were not impacted or the PRA model is revised to address the impact of the change on the application. This analysis will also be performed and documented for every application of Initiative 5b.

The outstanding items in the electronic tracking tool are ultimately incorporated into a major PRA revision which is performed periodically to ensure that the overall number of Documentation of PRA Technical Adequacy Page 6 items being tracked remains manageable. This robust process, governed by written procedures, is sufficient to ensure the PRA model represents the as-built, as-operated plant.

2.3.2 Unincorporated Changes to the Plant The justification of how unincorporated changes to the plant will be addressed is provided in the response in Section 2.3.1.

2.3.3 Departures from ASME Requirements The justification for departures from the ASME Standard Capability Category II requirements, including any unresolved findings/observations is as follows.

In October 2000, the PRA at Duke Energy's McGuire Nuclear Station received a peer review by an industry team of knowledgeable PRA practitioners (Ref. 19). Since the performance of this peer review, the industry has utilized the American Society of Mechanical Engineers (ASME) process to develop a standard identifying the requirements associated with PRA. RG 1.200 endorses the ASME PRA Standard as an acceptable method for demonstrating the technical adequacy of a PRA - provided various clarifications are made as identified in the regulatory guide.

Subsequently in 2008, as noted earlier, Duke Energy conducted a self-assessment of the McGuire PRA (Ref. 6) against the ASME PRA Standard through addenda RA-Sc-2007.

The McGuire PRA self-assessment included the Risk Assessment Technical Requirements listed in Section 4 of the ASME PRA Standard.-,This self-assessment evaluated the PRA with respect to Capability Category II. For the purposes of Initiative 5b, deviations from the Capability Category II supporting requirements were identified and dispositioned to ensure that these issues do not negatively impact Initiative 5b. For those requirements of the standard that have not been met, a justification of why it is acceptable that the requirement has not been met has been provided. A summary of these items is shown in Table 2-1 for McGuire (Ref. 6). Of the 29 items, 25 are either documentation or have no expected impact on Initiative 5b applications. The remaining four could have an impact based on the specific Initiative 5b application.

Because of the broad scope of potential Initiative 5b applications, and the fact that the impact of assumptions may differ for each surveillance requirement being evaluated, Duke Energy will address each of the deviations from Capability Category II listed in Table 2-1 for the McGuire PRA respectively for each application of Initiative 5b on an application specific basis. Again, if a requirement is not met, a justification of why it is acceptable that the requirement has not been met will be provided. These results will be with the documentation package for the specific Initiative 5b application.

2.3.4 Methodology to be Used for Initiative 5b NEI 04-10 Revision 1 provides the detailed process requirements for controlling surveillance frequencies of the TS Surveillance Requirements (SRs) that have been Documentation of PRA Technical Adequacy Page 7 relocated from the TSs to the SFCP. The methodology described in NEI 04-10 Revision 1 provides a risk-informed process to support a plant expert panel (called an Integrated Decisionmaking Panel or IDP) assessment of proposed changes to surveillance frequencies, assuring appropriate consideration of risk insights and other deterministic factors, which may impact surveillance frequencies, along with appropriate performance monitoring of changes and documentation requirements.

The Duke Energy SFCP, including the methodology of assessing surveillance frequency-changes utilized at McGuire, is consistent with NEI 04-10, Revision 1 and the supporting background document TSTF-425-A, Rev. 3 (Ref. 20).

2.3.5 Identification of Key Assumptions Identification of Key Assumptions related to surveillance frequencies (if any) and how they will be addressed is given below.

The overall Initiative 5b process is a risk-informed process with the PRA model results providing one of the inputs to the IDP to determine if a surveillance frequency change is warranted. The methodology recognizes that a key area of uncertainty for this application is the standby failure rate utilized in the determination of the surveillance frequency change impact. Therefore, the methodology requires the performance of selected sensitivity studies on the standby failure rate of the component(s) of interest for the surveillance frequency change assessment.

Because of the broad scope of potential Initiative 5b applications, any key assumptions and approximations relevant to the results obtained for an application of Initiative 5b will be addressed and documented on an application specific basis. This includes not only the results of the standby failure rate sensitivity study, but the results of any additional sensitivity studies identified during the performance of the reviews as outlined in Sections 2.3.1, 2.3.2, and 2.3.3.

2.3.6 Resolution of Relevant Peer.Review/Self-Assessment Findings and Observations Section 2.3.3 discusses departures from the ASME PRA Standard Capability Category II requirements and summarizes them on Table 2-1 for McGuire. However as previously noted, because of the broad scope of potential Initiative 5b applications, and the fact that the impact of assumptions may differ for each surveillance requirement being evaluated, Duke Energy will address each of the deviations from Capability Category II listed in Table 2-1 for McGuire for each application of Initiative 5b on an application specific basis. If a requirement is not met a justification of why it is acceptable that the requirement has not been met will be provided. If the PRA model is changed for a specific application of Initiative 5b to address self-assessment findings or if a sensitivity study is performed to demonstrate contributors significant to the application were not impacted by a self-assessment finding, a discussion of the results and conclusions for resolution will be included in the documentation package. Duke Energy will maintain a current listing of deviations from ASME PRA Standard Capability Category II requirements for McGuire for review and resolution against each application of Initiative 5b.

Documentation of PRA Technical Adequacy Page 8 2.3.7 Applicable Capability Category for Initiative 5b In accordance with NEI 04-10 Revision 1, the PRA must meet Capability Category II to be used for Initiative 5b applications. Duke Energy will ensure the McGuire PRA used for Initiative 5b applications either fully meets Capability Category II or departures from Capability Category II are justified to show insignificant impact on the results of the analysis. This will be done by performing a review of all outstanding departures from Capability Category II against the specific Initiative 5b application being addressed. The results of this review will be in the documentation package for the specific Initiative 5b application.

2.4 External Events Considerations This section addresses Condition 2 of the NRC Safety Evaluation for Initiative 5b.

Specifically it identifies quality characteristics for PRA models for which NRC-endorsed Standards do not exist, consistent with RG 1.200, Sections 1.2 and 1.3, and justifies the methods to be applied for assessing the risk contribution for those sources of risk not addressed by PRA models.

NRC endorsed standards currently exist for external hazards including seismic and fire PRAs. Revision 2 of Regulatory Guide (RG) 1.200 (Ref. 21), references the ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 (Ref. 7) for internal and external hazards. An NRC endorsed standard does not currently exist for shutdown PRAs. NEI 04-10 Revision 1 references RG 1.200 Revision 1 and ASME PRA Standard RA-Sb-2005b as the governing documents for Initiative 5b.

The NEI 04-10 Revision 1 methodology allows for Surveillance Frequency (SF) change evaluations to be performed in the absence of quantifiable PRA models for all external hazards. For those cases where the SF cannot be modeled in the plant PRA (or where a particular PRA model does not exist for a given hazard group), a' qualitative or bounding

.analysis is performed to provide justification for the acceptability of the proposed test interval change. In general, it is not expected that seismic, fire, or other external hazards will play a significant role in the impact of a given surveillance frequency change.

This section discusses the McGuire overall external hazards analysis methodology, the McGuire specific seismic and fire PRAs, and describes the methodology to be used to address shutdown risk impacts for Initiative 5b consistent with the requirements of the NEI 04-10 Revision 1 methodology.

2.4.1 Overall External Hazards Analysis Methodology The general approach used to develop the external event PRA at McGuire is as follows:

1) Identify all natural and man-made credible external events that may affect the site using many reference sources.

Documentation of PRA Technical Adequacy Page 9

2) A screening analysis was conducted using defined bounding criteria in order to select those events that may require further review.

3) A scoping analysis was performed on the remaining non-screened events to determine those that warranted a detailed site and plant-specific analysis.

This approach is consistent with that previously submitted to the NRC in Section 2.3 of Reference 14 and Volume 1, Section 3.0 of Reference 11. These references provide a greater level of detail of the approach if needed.

2.4.2 McGuire Seismic PRA Model The current McGuire seismic PRA model of record was last updated as part of Revision 3 of the PRA model (Ref. 22). However, the current methodology used is the same as that described in detail in the IPE submittal (Ref. 11) and Section 3 of the IPEEE submittal (Ref. 14), both of which have already been reviewed by the NRC. The reader is referred to those references for additional details of the seismic analysis.

The plant-specific seismic PRA analysis consists of four steps each of which are described below:

1) The McGuire site was evaluated to obtain the seismic hazard in terms of the frequency of occurrence of ground motions of various magnitudes. The site-specific hazard analysis (Ref. 23) was performed using the Seismicity Owners Group (SOG) methodology developed by EPRI for seismic hazard analysis of nuclear power plant sites in the Central and Eastern United States (CEUS).

Uncertainties were addressed in the hazard analysis.

• 2) From the site-specific seismic hazard curve, the capacities of important plant

-structures and equipment to withstand seismic events were evaluated to determine Conditional probabilities of failure as a function of ground acceleration for significant contributors (i.e., SSCs). These are commonly referred to as 'fragilities' or the site-specific fragility curves. Plant walkdowns were conducted, the most recent ones consistent with the guidelines of EPRI NP-6041 (Ref. 24).

3) An event tree was developed along with supporting top logic and system fault trees to reflect plant response to seismic events. These modified logic models were then solved to obtain Boolean expressions for the seismic event sequences of interest.

4) The Boolean expressions were quantified by convolving the probabilistic site seismicity and the fragilities for the plant structures and equipment obtained in steps 1 and 2. The resulting sequence frequencies are then integrated into the overall McGuire PRA risk results, resulting in final quantitative results.

The major changes to the current seismic analysis that have been made since the IPEEE submittal are as follows:

Documentation of PRA Technical Adequacy Page 10

1. Comprehensive review and revision of the seismic analysis documentation write-up.

2. Added component/structure fragility information to support values used in analysis.

3. Updated model with new Human Reliability Analysis (HRA) data.

4. Updated model with new common cause data.

5. Changes made to the fault tree are listed below.

• Made a new top gate for the model to address Containment Safeguards Responses. Added all supporting logic for new containment safeguards responses gate. This change was made to aid in accident "binning" in the seismic analysis.

• Removed duplicate failure of 4160V ac switchgear.

• Added failure of the Low Level Intake as a failure mode for Emergency Diesel Generator (EDG) 'A' to represent a loss of Nuclear Service Water (RN) train 'A'.

• Renamed the plant level surrogate fragility events to describe the various locations in the model where they are applied for ease of identification.

• Added failure of main control boards.

• Added Loss of Reactor Coolant (NC) Pump Seal Support logic to reflect the addition of a redesigned seal package. New seals are qualified for higher temperatures that limit the amount of leakage should failure occur.

6. Updated the seismic event tree.

7. Updated the seismic analysis quantitative results table.

As noted previously, Duke Energy is planning to perform a self-assessment against the supporting requirements for seismic events of ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 for the McGuire seismic PRA in 2010. The method as described in Section 2.3.3 of this attachment will be used to justify any departures from the ASME Standard Capability Category II requirements for each application of Initiative 5b. However, in accordance with the discussion in this section above, Duke Energy considers the current seismic model of record as meeting the required quality characteristics of RG 1.200 Sections 1.2 and 1.3 and is therefore sufficient for use as is in the application of Initiative 5b surveillance frequency changes.

2.4.3 McGuire Fire PRA Model The current McGuire fire PRA model analysis and methodology (Ref. 25) used in the model of record is the same analysis and methodology as described in the IPE submittal (Ref. 11); Section 4 and Appendix B of the IPEEE submittal (Ref. 14); and as discussed in the Supplemental IPEEE Fire Analysis Report (Ref. 16), all of which have already been reviewed by the NRC. The reader is referred to those references for additional details of the fire analysis.

The plant-specific fire PRA analysis consists of four steps each of which are described below:

a. The McGuire site and plant areas were analyzed to determine critical fire areas and possible scenarios for the possibility of a fire causing one or more of a Documentation of PRA Technical Adequacy Page 11 predetermined set of initiating events. Screening criteria were defined for those fire areas excluded from the fire analysis.

b. If there was a potential for an initiating event to be caused by a fire in an area, then the area was analyzed for the possibility of a fire causing other events which would impact the ability to shutdown the plant. These were identified by reviewing the impact on the internal event analysis models.

c. Each area was examined with an event tree fire model to quantify fire damage probabilities. The event tree related fire initiation, detection suppression, and propagation probabilities to equipment damage states.

d. Fire sequences were derived and quantified based on the fire damage probabilities and the additional failures necessary for a sequence to lead to a core melt. The additional failures were quantified by the models used in the internal events analysis.

The major changes to the current fire analysis that have been made since the IPEEE submittal deal with implementation of changes from the Supplemental IPEEE Fire Analysis Report (Ref. 16) and revised base case fire initiating event frequencies.

Since the McGuire fire PRA model is integrated into the overall PRA model, quantitative fire risk insights will be obtained each time when the PRA model is exercised. When the integrated PRA model is not utilized for a quantitative assessment and modeling of the affected equipment is not feasible, the fire risk insights will be assessed qualitatively.

This approach is consistent with the accepted NEI 04-10 Revision 1 methodology.

Duke Energy is planning to perform a self-assessment against the supporting requirements for fire events of ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 for the McGuire fire PRA in 2010. The method as described in Section 2.3.3 of this attachment will be used to justify any departures from the ASME Standard Capability Category II requirements for each application of Initiative 5b. However, in accordance with the discussion in this section above, Duke Energy considers the current fire model of record as meeting the required quality characteristics of RG 1.200 Sections 1.2 and 1.3 and is therefore sufficient for use as is in the application of Initiative 5b surveillance frequency changes.

2.4.3.1 McGuire Future State Fire PRA Model Initiative In February 2005, Duke Energy notified the NRC (Ref. 26) of its intent to adopt National Fire Protection Association (NFPA) Standard 805, "NFPA 805, Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generation Plants," 2001 edition, pursuant to Section 50.48(c) of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.48(c)), at all of its nuclear stations.

In a letter dated June 8, 2005, the NRC accepted Duke Energy's intent to adopt 10 CFR 50.48(c) (NFPA 805 Rule) for all three sites with Oconee Nuclear Station beginning the transition as a pilot plant on June 1, 2005 (Ref. 27). Duke Energy was requested to inform the NRC when the transition would begin at McGuire.

Attachment 2 Documentation of PRA Technical Adequacy

,Page 12 Subsequently, Duke Energy informed the NRC in 2006 (Ref. 28) that Duke Energy had begun the transition to NFPA 805 at McGuire Nuclear Station. The NRC response on September 26, 2006 (Ref. 29) acknowledged the transition to the performance-based standard for fire protection had begun at McGuire Units 1 and 2.

The McGuire Fire PRA model being developed uses guidance contained in NUREG/CR-6850/EPRI TR-1 011989, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities (Ref. 30). This is the same methodology and approach as that being used for the Oconee pilot. The McGuire Fire PRA model received an industry peer review against the requirements of Part 4 of ASME/ANS RA-Sa-2009, Addendum to RA-S-2008 (Ref. 7) September 14 - 18, 2009. A draft peer review report has been received and a final report is expected in 2010. The departures from Capability Category II requirements and other findings are being addressed. In 2010, Duke Energy is planning to submit a License Amendment Request (LAR) to the NRC to adopt the new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a),

10 CFR 50.48(c), and the guidance in Regulatory Guide (RG) 1.205. A discussion of the peer review open items and their disposition is expected to be part of that submittal.

2.4.4 McGuire Shutdown Risk Impact Analysis Since no approved quantitative shutdown risk PRA model for shutdown events currently exists at Duke Energy, McGuire will either 1) utilize the plant shutdown safety assessment tool developed to support implementation of NUMARC 91-06 (Ref. 31) as described in Duke Energy Nuclear Station Directive (NSD) 403 (Ref. 32) or 2) perform an alternate qualitative risk evaluation process to assess the proposed surveillance frequency change that utilizes Initiative 5b. These are acceptable options to not having a quantitative shutdown PRA model in accordance with Section 4 Step 10 (and other places) of NEI 04-10 Revision 1. In either case, the. guidance of NEI 04-10 Revision 1 will be followed.

2.5 Summary In Section 2.3 of this document the McGuire PRA technical adequacy was evaluated in accordance with the requirements of RG 1.200, Section 4.2. Section 2.4 of this document submitted quality characteristics of the seismic and fire PRA models in accordance with the requirements of RG 1.200, Sections 1.2 and 1.3. A discussion of the qualitative method to address shutdown risk was also discussed in Section 2.4.

Because of the broad scope of potential Initiative 5b applications and the fact that the risk assessment details will differ from application to application, for each individual surveillance frequency interval request, a review of the unincorporated changes to the plant and remaining gaps to specific requirements in the PRA standard will be made to determine which, if any, would merit additional application-specific sensitivity studies in the final analysis.

The results of the discussions above provide a basis for concluding that the current McGuire Units 1 and 2 PRA model is sufficiently robust and suitable for use in risk-Documentation of PRA Technical Adequacy Page 13 informed processes such as that proposed for the implementation of a Surveillance Frequency Control Program.

2.6 References

1. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of ProbabilisticRisk Assessment Results for Risk-Informed Activities", Revision 1, US Nuclear Regulatory Commission, January 2007.

2. ASME RA-S-2002, "Standardfor ProbabilisticRisk Assessment for Nuclear Power Plant Applications",with Addenda ASME RA-Sa-2003 and ASME RA-Sb-2005, December 2005.

3. NEI 00-02, "ProbabilisticRisk Assessment PeerReview Process Guidance,"

Revision A3, Nuclear Energy Institute, March 20, 2000.

4. Letter, USNRC to Nuclear Energy Institute, "Final Safety Evaluation for Nuclear Energy Institute (NEI) Industry Guidance Document NEI 04-10, Revision 0, "Risk-Informed Technical Specifications Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies"", September 28, 2006 (Adams Accession Number ML062700012).

5. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies,"April 2007.

6. DPC-1535.00-00-0013 (Cross references: CNC-1535.00-00-0094, MCC-1535.00-00-0089, OSC-9380), "PRA Quality Self-Assessment, Catawba Units 1 & 2, McGuire Units I & 2, Oconee Units 1, 2 & 3', Revision 2, November 2009.

7. ASME/ANS RA-Sa-2009, "Standardfor Level 1/Large Early Release Frequency ProbabilisticRisk Assessment for Nuclear Power Plant Applications,"Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.

8.. Nuclear Safety Analysis Center, "McGuire Unit 1 PRA Peer Review,"

May 27, 1983.

9. "McGuire Nuclear Station Unit 1 ProbabilisticRisk Assessment," Volumes 1-2, Duke Power Company, July 1984.

10. NRC Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities", US Nuclear Regulatory Commission, November 23, 1988.

11. Letter Duke Power Company to Document Control Desk (USNRC), McGuire Nuclear Station, "GenericLetter 88-20," November 4, 1991.

12. Letter USNRC to Duke Power Company, "Staff Evaluation of the McGuire Nuclear Station, Units I and.2 Individual Plant Examination - Internal Events Only,"

June 30, 1994.

13. NRC Generic Letter 88-20, "IndividualPlant Examination of External Events (IPEEE)for Severe Accident Vulnerabilities 10 CFR 50.54(f), Supplement 4,"

June 28, 1991.

14. Letter Duke Power Company to Document Control Desk (USNRC), McGuire Nuclear Station, Units 1 and 2, "IndividualPlant Examinationof External Events (IPEEE)Submittal," June 1, 1994.

15. Letter Duke Power Company to Document Control Desk (USNRC), McGuire Nuclear Station, "Request for Additional Information- Individual Plant Examinations for External Events; Response," November 17, 1995.

Documentation of PRA Technical Adequacy Page 14

16. Letter Duke Power Company to Document Control Desk (USNRC), "Supplemental IPEEEReport," Duke Power Company, McGuire Nuclear Station, Catawba Nuclear Station, July 30, 1996.

17. Letter USNRC to Duke Power Company, "Review of McGuire Nuclear Station, Units I and 2 - Individual Plant Examination of External Events Submittal',

February 16, 1999.

18. Letter Duke Energy Corporation to Document Control Desk (USNRC), McGuire Nuclear Station," 1997 Update of ProbabilisticRisk Assessment," March 19, 1998.

19. "McGuire Nuclear Station ProbabilisticSafety Assessment Peer Review Report",

Westinghouse Electric Co. for the Westinghouse Owners Group, January 2002.

20. Technical Specification Task Force Traveler number TSTF-425, Revision 3, "Relocate Surveillance Frequenciesto Licensee Control- RITSTF Initiative 5,"

July 2009.

21. Regulatory Guide 1.200, "An Approach for Determiningthe Technical Adequacy of ProbabilisticRisk Assessment Results for Risk-Informed Activities", Revision 2, US Nuclear Regulatory Commission, March 2009.

22. MCC-1 535.00-00-0049, McGuire Nuclear Station, External Events -Seismic Analysis, October 2001.

23. EPRI NP-4726-A, "Seismic Hazard Methodology for the Central and Eastern United States," July 1986.

24. EPRI NP-6041, Revision 1, "A Methodology of Assessment of Nuclear Power Plant Seismic Margin,"August 1991.

25. MCC-1535.00-00-0047, McGuire Nuclear Station, Fire Analysis Notebook, September 1997.

26. Letter Duke Energy Corporation to Document Control Desk (USNRC), "Letterof Intent to Adopt NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water Reactor GeneratingPlants, 2001 Edition," February 28, 2005 (Adams Accession Number ML050670305).

27. Letter USNRC to Duke Energy Corporation, ."NRC Response to Duke's Letter'of Intent to Adopt 10 CFR 50.48(c) (NFPA 805 Rule)," June 8, 2005 (Adams Accession Number ML051080005).

28.- Letter Duke Energy Corporation to Document Control Desk (USNRC), "Letterof Intent to Start the Transition to NFPA 805 Performance-BasedStandardfor Fire Protection for Light Water Reactor GeneratingPlants, 2001 Edition,"April 18, 2006 (Adams Accession Number ML061150375).

29. Letter USNRC to Duke Energy Corporation, "NRC Response to Letter of Intent to Adopt Title 10 of the Code of FederalRegulations, Part 50, Section 50.48(c), for McGuire Nuclear Station, Units 1 and 2," September 26, 2006 (Adams Accession Number ML062700009).

30. EPRI/NRC-RES, "FirePRA Methodology for Nuclear Power Facilities,"

NUREG/CR-6850, EPRI TR-1 011989, Final Report, September 2005.

31. NUMARC 91-06, "Guidelinesfor Industry Actions to Address Shutdown Management," December 1991.

32. NSD-403, "Shutdown Risk Management (Modes 4, 5, 6, and No-Mode) per 10CFR 50.65(a)(4)," Revision 19, April 2009.

Documentation of PRA Technical Adequacy Page 15 TABLE 2-1 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD THROUGH ADDENDA RA-Sc-2007 Title Description of Gap Applicable Current Status I Comment Importance to 5b SRs Application Gap #1 Accident sequence notebooks and AS-B3 Open. Phenomenological None - documentation system model notebooks should effects are considered in the issue.

document the environmental effects model, although these of the initiating event and the impact considerations are not always on mitigation systems. documented.

Gap #2 Revise the data calc. to discuss DA-Ala Open. SSC and unavailability None - documentation component boundaries definitions, boundaries, SSC failure modes issue.

and success criteria are used consistently across analyses; however, these need to be formally documented.

Gap #3 Revise the data calc. to segregate DA-B1 Open. Previously, generic data Partitioning the failure standby and operating component sources often did not provide rates represents a data. Segregate components by standby and operating failure refinement to the data service condition to the extent rates. NUREG/CR-6928 does analysis process, but supported by the data. provided more of this data, and is not expected to will be used going forward. impact the 5b analysis.

Gap #4 Enhance the documentation to DA-D4 Open. As part of the Bayesian None - documentation include a discussion of the specific update process, checks are issue.

checks performed on the Bayesian- performed to assure that the updated data, as required by this SR. posterior distribution is reasonable given the prior distribution and plant experience. These checks need to be formally documented.

Documentation of PRA Technical Adequacy Page 16 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application Gap #5 Provide documentation of the DA-D6 Open. Generic common cause None - documentation comparison of the component failure (CCF) probabilities are issue.

boundaries assumed for the generic considered for applicability to CCF estimates to those assumed in the plant. CCF probabilities the PRA to ensure that these are consistent with plant boundaries are consistent. experience and component boundaries, although the CCF documentation needs to be enhanced to discuss component boundaries.

Gap #6 Enhance the human reliability HR-A2 Open. Based on evaluations Relative to post-analysis (HRA) to consider the using the EPRI HRA calculator, initiator human error potential for calibration errors. calibration errors that result in probabilities (HEPs),

failure of a single channel are equipment random expected to fall in the low 10-3 failure rates and range. maintenance unavailability, calibration HEPs are not expected to contribute significantly to overall equipment unavailability.

Additionally, the next revision of the PRA will incorporate the potential for calibration errors in the HRA.

Thus there is no impact on the 5b analysis.

Gap #7 Identify maintenance and calibration HR-A3 Open. Based on evaluations Relative to post-activities that could simultaneously using the EPRI HRA calculator, initiator HEPs, latent affect equipment in either different calibration errors that result in human error trains of a redundant system or failure of multiple channels are probabilities, Documentation of PRA Technical Adequacy Page 17 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application Gap #7 diverse systems. expected to fall in the low equipment random (continued) 10-5 range. failure rates and maintenance unavailability, calibration HEPs and misalignment of multiple trains of equipment are not expected to contribute significantly to overall equipment unavailability. Thus there is no impact on the 5b analysis.

Gap #8 Develop mean values for pre-initiator HR-D6 Open. Pre-initiator HEPs are The suggested data HEPs. generally set to relatively high refinement is not screening values, which bound expected to have a the mean values. Even so, pre- significant impact on initiator HEPs are not the results. Thus significant contributors to risk. there is no impact on the 5b analysis.

Gap #9 Document in more detail the HR-G3 Open. Performance shaping None - documentation influence of performance shaping factors are accounted for in the issue.

factors on execution human error development of human error probabilities, probabilities, although detailed documentation is not always available for every HRA input.

Gap #10 Enhance HRA documentation HR-G4 Open. Thermal Hydraulic (T/H) None - documentation accordingly. analyses, simulator runs and issue.

operator interviews are used in developing the time available to complete operator actions. The time at which the cue to take action is received is specified in Documentation of PRA Technical Adequacy Page 18 Title Description of Gap Applicable Current Status I Comment Importance to 5b SRs Application Gap #10 the HEP quantification.

(continued) However, the HRA documentation needs to be enhanced to provide a traceable path to all analysis inputs.

Gap #11 Document a review of the human. HR-G6 Open. HFEs are reviewed by None - documentation failure events (HFEs) and their final knowledgeable site personnel issue.

HEPs relative to each other to to assure high quality.

confirm their reasonableness given However, this review needs to the scenario context, plant history, be better documented.

procedures,. operational practices, and experience.

Gap #12 Develop mean values for post- HR-G9 Open. The use of mean values The 5b analysis will initiator HEPs. for HEPs instead of lower include a sensitivity probability median values can study to evaluate the affect the PRA results. use of different HEPs if the calculated risk is close to the threshold.

Gap #13 Develop more detailed HR-H2 Open. Operator recovery None - documentation documentation of operator cues, actions are credited only ifthey issue.

relevant performance shaping are feasible, as determined by factors, and availability of sufficient the procedural guidance, cues, manpower to perform the action. performance shaping factors and available manpower. As noted for HR-G3, -G4, and -G6 above, the documentation of these considerations needs to be enhanced.

Gap #14 IE-A3: Consider the two events IE-Al Open. The 5b analysis will identified in PRA change form M IE-A3 include a sensitivity 0012 for inclusion in the CA pump IE-A3a study to evaluate the room flood frequency. The other IE-A4 impact of a larger CA Documentation of PRA Technical Adequacy Page 19 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application Gap #14 SRs require various enhancements IE-A4a pump room frequency.

(continued) to the initiating events analysis IE-A5 documentation. IE-A6 IE-A7 IE-B1 IE-B2 IE-B3 IE-D3 Gap #15 Various enhancements to the internal IF-B3 Open. Until the flooding flood analysis: IF-C2c analysis is upgraded,

• Discuss flood mitigative features. IF-C3 the potential for flood-

• Address the potential for spray, jet IF-C3b induced failures of impingement, and pipe whip IF-E6b SSCs will be assessed failures. IF-F2 on a case-by-case

" Provide more analysis of flood basis.

propagation flowpaths. Address potential structural failure of doors or walls due to flooding loads and the potential for barrier unavailability. Address potential indirect effects.

• Enhance the documentation to address all of the SR details.

Gap #16 Explicitly model Reactor Coolant LE-C6 Open. This issue affects No impact on the 5b System (RCS) depressurization for certainsmall LOCAs. analysis.

small Loss of Coolant Accidents However, since the small LOCA (LOCAs) and perform the contribution to LERF is small, dependency analysis on the HEPs. there is no significant impact on the PRA results.

Gap #17 Various enhancements to the. LERF LE-G3 Open. None - documentation documentation. LE-G4 issue.

LE-G5 LE-G6 Documentation of PRA Technical Adequacy Page 20 Title Description of Gap Applicable Current Status I Comment Importance to 5b SRs Application Gap #18 Perform and document a comparison LE-F3 Open. Since McGuire and None - documentation of PRA results with similar plants.: QU-D3 Catawba are sister plants, in issues.

practice, their results are often compared. Also, comparisons performed for mitigating system performance indicators (MSPI) and other programs help identify causes for significant differences. However, to fully meet this SR, the model quantification documentation needs to be enhanced to provide a results comparison.

Gap #19 Perform and document sensitivity LE-F2 Open. Perform and analyses to determine the impact of QU-E4 document sensitivity 7 the assumptions and sources of analyses to determine model uncertainty on the results. the impact of the assumptions and sources of model uncertainty on the 5b analysis results.

Gap #20 Expand the documentation of the QU-F2 Open. These SRs pertain to None - documentation PRA model results to address all QU-F6 the model quantification issues.

required items. documentation.

Gap #21 Improve the documentation on the SC-A4 Open. Success criteria are None - documentation T/H bases for all safety function developed to address all of the issue.

success criteria for all initiators, modeled initiating events.

However, the documentation of success criteria needs to be improved to include initiator information.

Gap #22 Provide evidence that an SC-B5 Open. McGuire success None - documentation acceptability review of the T/H criteria are consistent with issue.

analyses is performed. those of sister plants included Documentation of PRA Technical Adequacy Page 21 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application Gap #22 in the Pressurized Water (continued) Reactor Owners Group (PWROG) Probabilistic Safety Assessment (PSA) database.

However, to fully meet this SR, the success criteria documentation needs to be enhanced to include a results comparison.

Gap #23 Expand the documentation of the SC-Cl Open. These SRs pertain to None - documentation success criteria development to SC-C2 the success criteria issues.

address all required items. documentation.

Gap #24 Enhance the system documentation SY-A4 Open. To support system None - documentation to include an up-to-date system model development, issue.

walkdown checklist and system walkdowns and plant personnel engineer review for each system. interviews were performed.

However, documentation of an up-to-date system walkdown is not included with each system notebook.

Gap #25 Enhance systems analysis SY-A8 Open. Basic event component None - documentation documentation to discuss component boundaries utilized in the issue.

boundaries. systems analysis are consistent with those in the data analysis.

In addition, component boundaries are consistent with those defined in the generic failure rate source documents, such as NUREG/CR-6928.

Dependencies among components, such as interlocks, are explicitly modeled, consistent with the PRA Modeling Guidelines Documentation of PRA Technical Adequacy Page 22 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application Gap #25 workplace procedure. There is (continued) no evidence of a technical problem with component boundaries, just a need to improve the documentation.

Gap #26 Provide quantitative evaluations for SY-A14 Open. It is expected that There is no evidence screening. conversion to a more of a technical problem quantitative approach would not associated with the change decisions about screening of whether or not to exclude components or components or failure modes. component failure A review of our qualitative modes, just a need to screening process confirms this document a expectation. For example, quantitative screening.

transfer failure events for Thus there is no motor-operated valves (MOVs) impact on the 5b with 24 hr exposure times may analysis.

not be modeled unless probabilistically significant with respect to logically equivalent basic events. For McGuire, the MOV transfers failure probability is less than 1% of the MOV fails to open on demand failure rate. In cases like this, not including the relatively low probability failure mode in the PRA model does not have an appreciable impact on the results.

Gap #27 Per Duke Energy's PRA modeling.. SY-B8 Open. As noted for SY-A4, None - documentation guidelines, ensure that a walkdowns (which look for issue.

walkdown/system engineer interview spatial and environmental checklist is included in each system hazards) have been performed, Documentation of PRA Technical Adequacy Page 23 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application Gap #27 notebook. Based on the results of the although up-to-date walkdown (continued) system walkdown, summarize in the documentation is not included system write-up any possible spatial with each system notebook.

dependencies or environmental hazards that may impact system operation.

Gap #28 Document a consideration of. SY-B15 Open. The impact of adverse None - documentation potential SSC failure due to adverse environmental conditions on issue.

environmental conditions. SSC reliability is considered but is not always documented.

However, there is no evidence of a technical problem associated with components that may be required to operate in conditions beyond their environmental qualification, just a need to improve the

• ___ documentation.

Gap #29 Enhance system model SY-C2 Open. This SR pertains to the None - documentation documentation to comply with all systems analysis issue.

ASME PRA Standard requirements. documentation.

ATTACHMENT 5 TSTF-425 (NUREG-1431) SURVEILLANCES VERSUS MCGUIRE SURVEILLANCES CROSS REFERENCE TABLE Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Surveillance Frequency Cross Reference Table Page 1 TSTF-425 (NUREG-1431) SRs vs. McGuire SRs Cross Reference TSTF SR McGuire SR COMMENTS 3.1.1.1 3.1.1.1 3.1.2.1 3.1.2.1 3.1.4.1 3.1.4.1 3.1.4.2 3.1.4.2 Remove the expired one time change for Unit 1 3.1.5.1 3.1.5.1 3.1.6.2 3.1.6.2 3.1.6.3 3.1.6.3 3.1.8.2 3.1.8.2 3.1.8.3 3.1.8.3 3.1.8.4 3.1.8.4 3.2.1.1 3.2.1.1 McGuire is a RAOC plant (option B) 3.2.1.2 3.2.1.2 McGuire is a RAOC plant (option B) 3.2.1.3 McGuire SR not in TSTF 3.2.2.1 3.2.2.1 3.2.2.2 McGuire SR not in TSTF 3.2.3.1 3.2.3.1 McGuire is a RAOC plant (option B) 3.2.4.1 3.2.4.1 3.2.4.2 3.2.4.2 3.3.1.1 3.3.1.1 3.3.1.2 3.3.1.2 3.3.1.3 3.3.1.3 3.3.1.4 3.3.1.4 3.3.1.5 3.3.1.5 3.3.1.6 3.3.1.6 3.3.1.7 3.3.1.7 3.3.1.8 3.3.1.8 3.3.1.9 3.3.1.9 3.3.1.10 3.3.1.10 3.3.1.11 3.3.1.11 Excore detector modification LAR dated July 1, 2009 in NRC review to revise SR description 3.3.1.12 3.3.1.12 3.3.1.13 3.3.1.13 3.3.1.14 3.3.1.14 3.3.1.16 3.3.1.16 3.3.1.17 McGuire SR not in TSTF 3.3.2.1 3.3.2.1 3.3.2.2 3.3.2.2 3.3.2.3 3.3.2.3 3.3.2.4 3.3.2.4 3.3.2.5 3.3.2.5 3.3.2.6 3.3.2.6 Surveillance Frequency Cross Reference Table Page 2 TSTF SR McGuire SR COMMENTS 3.3.2.7 ---------- TSTF SR not in McGuire TS 3.3.2.8 3.3.2.7 3.3.2.9 3.3.2.8 3.3.2.10 3.3.2.9 3.3.2.11 ---------- TSTF SR not in McGuire TS 3.3.3.1 3.3.3.1 3.3.3.2 3.3.3.3 3.3.4.1 3.3.4.1 3.3.4.2 3.3.4.2 3.3.4.3 3.3.4.3 3.3.4.4 ---------- TSTF SR not in McGuire TS 3.3.5.1 ---------- TSTF SR not in McGuire TS 3.3.5.2 3.3.5.1 3.3.5.3 3.3.5.2 3.3.6 --------- McGuire deleted TS 3.3.6 by Amendment Nos. 243/224 dated July 26, 2007.

3.3.7 --------- McGuire does not have this TSTF TS 3.3.8 --------- McGuire does not have this TSTF TS 3.3.9 --------- McGuire does not have this TSTF TS 3.4.1.1 3.4.1.1 3.4.1.2 3.4.1.2 3.4.1.3 3.4.1.3 3.4.1.4 --------- TSTF SR not in McGuire TS 3.4.1.4 McGuire SR not in TSTF 3.4.2.1 ------- Event driven SR at McGuire, retain SR frequency in current Tech Spec.

3.4.3.1 3.4.3.1 3.4.4.1 3.4.4.1 3.4.5.1 3.4.5.1 3.4.5.2 3.4.5.2 3.4.5.3 3.4.5.3 3.4.6.1 3.4.6.1 3.4.6.2 3.4.6.2 3.4.6.3 3.4.6.3 3.4.7.1 3.4.7.1 3.4.7.2 3.4.7.2 3.4.7.3 3.4.7.3 3.4.8.1 3.4.8.1 3.4.8.2 3.4.8.2 3.4.9.1 3.4.9.1 3.4.9.2 3.4.9.2 3.4.9.3 TSTF SR not in McGuire TS 3.4.11.1 3.4.11.1 Surveillance Frequency Cross Reference Table Page 3 TSTF SR McGuire SR COMMENTS 3.4.11.2 3.4.11.2 3.4.11.3 TSTF SR not in McGuire TS 3.4.11.3 McGuire SR not in TSTF 3.4.11.4 TSTF SR not in McGuire TS 3.4.12.1 3.4.12.1 McGuire SR is TSTF 3.4.12.1 and 3.4.12.2 3.4.12.2 3.4.12.3 3.4.12.2 3.4.12.4 3.4.12.3 3.4.12.5 3.4.12.4 3.4.12.6 3.4.12.5 3.4.12.7 TSTF SR not in McGuire TS 3.4.12.8 3.4.12.6 3.4.12.9 3.4.12.7 3.4.13.1 3.4.13.1 3.4.13.2 3.4.13.2 3.4.14.1 3.4.14.1 3.4.14.2 3.4.14.2 3.4.14.3 TSTF SR not in McGuire TS 3.4.15.1 3.4.15.1 3.4.15.2 3.4.15.2 3.4.15.3 3.4.15.3 3.4.15.4 3.4.15.4 3.4.15.5 3.4.15.5 3.4.15.6 McGuire SR not in TSTF 3.4.16.1 3.4.16.1* TSTF-490 (E bar) LAR dated December 15, 2009 in NRC review to revise SR description.

3.4.16.2 3.4.16.2 3.4.16.3 3.4.16.3* TSTF-490 LAR above deletes this SR.

3.4.17 McGuire does not have this TSTF TS 3.4.19.1 3.4.17.1 3.5.1.1 3.5.1.1 *.

3.5.1.2 3.5.1.2 3.5.1.3 3.5.1.3 3.5.1.4 3.5.1.4 3.5.1.5 3.5.1.5 3.5.2.1 3.5.2.1 3.5.2.2 3.5.2.2 3.5.2.3 3.5.2.3 3.5.2.5 3.5.2.5 3.5.2.6 3.5.2.6 3.5.2.7 3.5.2.7 3.5.2.8 3.5.2.8 3.5.4.1 3.5.4.1 3.5.4.2 3.5.4.2 3.5.4.3 3.5.4.3 3.5.5.1 3.5.5.1 Surveillance Frequency Cross Reference Table Page 4 TSTF SR McGuire SR COMMENTS 3.5.6 McGuire does not have this TSTF TS 3.6.2.2 McGuire SR not in TSTF 3.6.2.2 3.6.2.3 3.6.3.1 3.6.3.1 3.6.3.2 McGuire deleted SR 3.6.3.2 by Amendment Nos. 243/224 dated July 26, 2007.

3.6.3.3 3.6.3.3 3.6.3.5 McGuire SR is lAW the IST Program 5.5.8 3.6.3.6 TSTF SR not in McGuire TS 3.6.3.7 McGuire SR is lAW the CLRT Program 5.5.2 3.6.3.8 3.6.3.7 3.6.3.9 ---------- TSTF SR not in McGuire TS 3.6.3.10 TSTF SR not in McGuire TS 3.6.4.1 A 3.6.4.1 3.6.5.1 B 3.6.5.1 3.6.5.2 B 3.6.5.2 3.6.6.1 C 3.6.6.1 3.6.6.3 C 3.6.6.3 3.6.6.4 C 3.6.6.4 3.6.6.5 C 3.6.6.7* Cont. Spray LAR dated Sept. 30, 2009 in NRC review to revise spray nozzle inspection frequency.

3.6.6.5 McGuire SR not in TSTF 3.6.6.6 McGuire SR not in TSTF 3.6.7 McGuire does not have this TSTF TS 3.6.8.1 TSTF SR not in McGuire TS 3.6.8.2 3.6.16.1 3.6.8.3 3.6.16.3 3.6.8.4 3.6.16.2 3.6.9.1 3.6.8.1 3.6.9.2 ----------- TSTF SR not in McGuire TS

-- 7-- 3.6.8.2 McGuire SR not in TSTF..

3.6.9.3 3.6.8.4 3.6.8.3 McGuire SR not in TSTF 3.6.10.1 3.6.9.1 3.6.10.2 3.6.9.2 3.6.10.3 3.6.9.3 3.6.13.1 3.6.10.1 3.6.13.3 3.6.10.3 3.6.13.4 3.6.10.4 3.6.13.5 3.6.10.5 3.6.14.1 3.6.11.1 3.6.14.2 3.6.11.2 3.6.14.3 3.6.11.3 3.6.14.4 TSTF SR not in McGuire TS 3.6.11.4 McGuire SR not in TSTF 3.6.11.5 McGuire SR not in TSTF

-3.6.11.6 McGuire SR not in TSTF Surveillance Frequency Cross Reference Table Page 5 TSTF SR McGuire SR COMMENTS 3.6.11.7 McGuire SR not in TSTF 3.6.15.1 3.6.12.1 3.6.15.2 3.6.12.4 3.6.15.3 3.6.12.5 3.6.15.4 3.6.12.3 3.6.15.5 3.6.12.7 3.6.15.6 3.6.12.6 3.6.16.1 3.6.13.1* Ice Condenser Door TS LAR dated Oct. 2, 2008 in NRC review to revise SR description.

3.6.16.2 3.6.13.2 3.6.16.3 3.6.13.4* Above LAR to revise description.

3.6.16.4 3.6.13.5* Above LAR to revise description.

3.6.16.5 3.6.13.6* Above LAR to delete SR.

3.6.16.6 3.6.13.7 3.6.16.7 3.6.13.3 3.6.17.2 3.6.14.2 3.6.17.4 3.6.14.4 3.6.17.5 3.6.14.5 3.6.18.1 3.6.15.2 McGuire SR is TSTF SR 3.6.18.1 c. only 3.6.18.2 3.6.15.3 3.7.2.2 McGuire SR is lAW the IST Program 5.5.8 3.7.3.2 McGuire SR is lAW the IST Program 5.5.8 3.7.4.1 3.7.4.1 3.7.4.2 3.7.4.2 3.7.5.1 3.7.5.1 3.7.5.3 3.7.5.3 3.7.5.4 3.7.5.4 3.7.6 --------- McGuire does not have this TSTF TS 3.7.7.1 3.7.6.1 3.7.7.2 .3.7.6.2 3.7.7.3 3.7.6.3 3.7.8.1 3.7.7.1 3.7.8.2 3.7.7.2 3.7.8.3 3.7.7.3 3.7.9.1 3.7.8.1 3.7.9.2 3.7.8.2 3.7.9.3 ---------- TSTF SR not in McGuire TS 3.7.9.4 TSTF SR not in McGuire TS 3.7.8.3 McGuire SR not in TSTF 3.7.10.1 3.7.9.1 3.7.10.3 3.7.9.3 3.7.10.4 ---------- McGuire SR is lAW CRH Program 5.5.16 3.7.11.1 TSTF SR not in McGuire TS 3.7.10.1 McGuire SR not in TSTF Surveillance Frequency Cross Reference Table Page 6 TSTF SR McGuire SR COMMENTS 3.7.12.1 3.7.11.1 3.7.12.3 3.7.11.3 3.7.12.4 3.7.11.4 3.7.12.5 TSTF SR not in McGuire TS 3.7.13.1 TSTF SR not in McGuire TS 3.7.12.1 McGuire SR not in TSTF 3.7.13.3 TSTF SR not in McGuire TS 3.7.13.4 3.7.12.4 3.7.13.5 3.7.12.5 3.7.14 McGuire does not have this TSTF TS 3.7.15.1 3.7.13.1 3.7.16.1 3.7.14.1 3.7.18.1 3.7.16.1 3.8.1.1 3.8.1.1 3.8.1.2 3.8.1.2 3.8.1.3 3.8.1.3 3.8.1.4 3.8.1.4* Diesel day tank level LAR dated Dec. 1, 2008 in NRC review to revise SR description.

3.8.1.5 3.8.1.5 3.8.1.6 3.8.1.6 3.8.1.7 3.8.1.7 3.8.1.8 3.8.1.8 3.8.1.9 3.8.1.9 3.8.1.10 3.8.1.10 3.8.1.11 3.8.1.11 3.8.1.12 3.8.1.12 3.8.1.13 3.8.1.13 3.8.1.14 3.8.1.14 3.8.1.15 3.8.1.15 3.8.1.16 3.8.1.16 3.8..17 3.8.1.17 3.8.1.18 3.8.1.18 3.8.1.19 3.8.1.19 3.8.1.20 3.8.1.20 3.8.3.1 3.8.3.1 3.8.3.2 TSTF SR not in McGuire TS 3.8.3.4 3.8.3.3 3.8.3.5 3.8.3.4 3.8.4.1 3.8.4.1 3.8.4.2* McGuire SR not in TSTF. Battery Resistance LAR dated Dec. 14, 2009 in NRC review to revise SR description.

3.8.4.3 McGuire SR not in TSTF 3.8.4.4 McGuire SR not in TSTF 3.8.4.5* McGuire SR not in TSTF. Battery Resistance LAR above in NRC review to revise SR description.

Surveillance Frequency Cross Reference Table Page 7 TSTF SR McGuire SR COMMENTS 3.8.4.2 3.8.4.6 3.8.4.3 3.8.4.7 3.8.6.1 --------- TSTF SR not in McGuire TS 3.8.6.2 --------- TSTF SR not in McGuire TS 3.8.6.3 TSTF SR not in McGuire TS 3.8.6.4 TSTF SR not in McGuire TS 3.8.6.5 TSTF SR not in McGuire TS 3.8.6.6 3.8.4.8 3.8.6.1 McGuire SR in Table format (TS Table 3.8.6-1) 3.8.6.2 McGuire SR in Table format (TS Table 3.8.6-1)

S3.8.6.3 McGuire SR not in TSTF but similar to TSTF SR 3.8.6.4 3.8.7.1 3.8.7.1 3.8.8.1 3.8.8.1 3.8.9.1 3.8.9.1 3.8.10.1 3.8.10.1 3.9.1.1 3.9.1.1 3.9.2.1 3.9.2.1 3.9.3.1 3.9.3.1 3.9.3.2 3.9.3.2 3.9.4.1 3.9.4.1 3.9.4.2 TSTF SR not in McGuire TS 3.9.5.1 3.9.5.1 3.9.6.1 3.9.6.1 3.9.6.2 3.9.6.2 3.9.7.1 3.9.7.1

ATTACHMENT 6 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Proposed No Significant Hazards Consideration Page 1 In accordance with the provisions of 10 CFR 50.90, Duke Energy Carolinas (Duke Energy) is submitting a request for an amendment to the Technical Specifications (TS) for McGuire Nuclear Station (McGuire) Units 1 and 2.

The proposed amendment requests the adoption of an approved change to the Standard Technical Specifications (STS) for Westinghouse Plants (NUREG-1431) to allow relocation of specific Technical Specification Surveillance frequencies to a licensee controlled program. The proposed change is described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML080280275) related to the Relocation of Surveillance Frequencies to Licensee Control, RITSTF Initiative 5b, and was described in the Notice of Availability published in the Federal Register on July 6, 2009, 74 FR 31996-32006.

The proposed changes are consistent with NRC approved TSTF-425 Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b." The proposed change relocates surveillance frequencies to a licensee controlled program, the SFCP. This change is applicable to licensees using probabilistic risk guidelines contained in NRC approved NEI 04-10, Revision 1, "Risk Informed Technical Specifications Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. 071360456).

Basis for proposed no significant hazards consideration: As required by 10 CFR 50.91(a), the Duke Energy analysis of the issue of no significant hazards consideration is presented below:

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

Response: No.

The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control. under a new Surveillance Frequency Control Program.

Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased.

The systems and components required by the Technical Specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, 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.

(1 Proposed No Significant Hazards Consideration Page 2

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

Response: No.

No new or different accidents result from utilizing the proposed change. The changes do 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 changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

Therefore, the proposed changes do 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 the margin of safety?

Response: No.

The design, operation, testing methods, and acceptance criteria for systems; structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the Updated Final Safety Analysis Report and Bases to the Technical Specifications), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Duke Energy will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Revision 1 in accordance with the TS SFCP. NEI 04-10, Revision 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177..

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

Based upon the reasoning presented above, Duke Energy concludes that the requested change does not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), Issuance of Amendment.

ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION CHANGES Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program

INSERT I (TS SR Frequency Pages)

In accordance with the Surveillance Frequency Control Program

Definitions 1.1 1.1 Definitions (continued)

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device OPERATIONAL TEST and verifying the OPERABILITY of required alarm, interlock, (TADOT) and trip functions. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the required accuracy.

McGuire Units 1 and 2 1.1-6 Amendment Nos4.

SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDM shall be within the limit specified in the COLR.

APPLICABILITY: MODE 2 with keff < 1.0, MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to restore 15 minutes SDM to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 .. Verify. SDM is within the limit specified in the COLR. 4/oy

.. . . ' * . .L J

McGuire Units 1 and 2 3.1.1-1 Amendment Nos 8

Core Reactivity 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 ------------------ NOTE --------------------

The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.

Verify measured core reactivity is within +/- 1% Ak/k of Once prior to predicted values. entering MODE 1 after each refueling AND

• 1 A" IP" McGuire Units 1 and 2 3.1.2-2 Amendment Nosetýý

Rod Group Alignment Limits 3.1.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition B not met.

D. More than one rod not D.1.1 Verify SDM is within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within alignment limit, limit specified in the COLR.

OR D.1.2 Initiate boration to restore required SDM to within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> limit.

AND D.2 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual rod positions within alignment limit. Z(ios AND i '-1 Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter when the rod position deviation monitor is inoperable (continued)

McGuire Units 1 and 2 3.1.4-3 Amendment Nos.*

Rod Group Alignment Limits 3.1.4 SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn Prior to reactor position, is < 2.2 seconds from the beginning of decay of criticality after stationary gripper coil voltage to dashpot entry, with: each removal of the reactor head

a. Tavg > 551OF; and

b. All reactor coolant pumps operating.

z6tpch ,Wli caA t on nit 1.pKIy.

McGuire Units 1 and 2 3.1.4-4 Amendment No s+Jirl s7-t4 n*

Shutdown Bank Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank is within the limits specified in the COLR.

McGuire Units 1 and 2 3.1.5-2 Amendment Nos.(!ýý

Control Bank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.6.2 Verify each control bank insertion is within the limits specified in the COLR.

AND Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter when the rod insertion limit monitor is inoperable SR 3.1.6.3 Verify sequence and overlap limits specified in the COLR 126'y&

are met for control banks not fully withdrawn from the core.

McGuire Units 1 and 2 3.1.6-3 Amendment Nos.(ii

PHYSICS TESTS Exceptions 3.1.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D. 1 Be in MODE 3. 15 minutes associated Completion Time of Condition C not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Perform a CHANNEL OPERATIONAL TEST on power Prior to initiation of range and intermediate range channels per SR 3.3.1.7, PHYSICS TESTS SR 3.3.1.8, and Table 3.3.1-1.

SR 3.1.8.2 Verify the RCS lowest loop average temperature is t

> 541 OF.

SR 3.1.8.3 Verify THERMAL POWER is < 5% RTP. -

SISR 3.1.8.4 Verify SDM is within the limit specified in the COLR.

X ZA/SE/

McGuire Units 1 and 2 3.1.8-2 Amendment Nos. fEý

FQ(X,Y,Z) 3.2.1 SURVEILLANCE REQUIREMENTS During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify FmQ(X,Y,Z) is within steady state limit. Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions after exceeding, by

> 10% RTP, the THERMAL POWER at which Fm (X,Y,Z) was last verified AND (continued)

McGuire Units 1 and 2 3.2.1-3 Amendment Nos<.ý -

Fo(X,Y,Z) 3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.2 NOTE -----------------------------------

1. Extrapolate Fm(X,Y,Z) using at least two measurements to 31 EFPD beyond the most recent measurement. If Fm(X,Y,Z) is within limits and the 31 EFPD extrapolation indicates:

0 FQ(X,Y,Z)EXTRAPOLATED > FQ(X,Y,Z) EXTRAPOLATED, and

__EXTRAPOLATED > F(XY*Z)

FQ(X,Y,z)OPEXTRAPOLATED FL(xY,Z)°p then:

a. Increase Fm(X,Y,Z) by the appropriate factor specified in the COLR and reverify Fm(X,Y,Z) < FL(X,Y,Z)°P; or

b. Repeat SR 3.2.1.2 prior to the time at which Fm(X,Y,Z) < F (X,Y,Z)°P is extrapolated to not be met.

2. Extrapolation of Fm(X,Y,Z) is not required for the initial flux map taken after reaching equilibrium conditions.

Once within Verify FMx(X,Y,Z)° FL(XyZ)P, . 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions after exceeding, by >

10% RTP, the THERMAL

.1 " - "*  : ..f- ' , POWER at which Fm (X,Y,Z) was last verified AND

-'ýýontinued)

McGuire Units 1 and 2 3.2.1-4 Amendment NosA Uj@

FQ(X,Y,Z) 3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.3 ---------------------- NOTES------ ------

1. Extrapolate FmQ(X,Y,Z) using at least two measurements to 31 EFPD beyond the most recent measurement. If Fm(X,Y,Z) is within limits and the 31 EFPD extrapolation indicates:

FQ(X,Y,Z)EXTRAPOLATED > FL(X,Y,Z)RPSEXTRAPOLATED, and FeXYZ)EXTRAPOLATED > FM(XY.Z)

FL(x,Y,Z)-PSEXTRAPOLATED FQ(X,Y,Z)RPs then: (

a. Increase FM(X,Y,Z) by the appropriate factor specified in the COLR and reverify FM,(X,Y,Z) < F (X,Y,Z)RPS; or

b. Repeat SR 3.2.1.3 prior to the time at which Fm(X,Y,Z) < FL(X,Y,Z)RPS is extrapolated to not be met.

2. Extrapolation of Fm (X,Y,Z) is not required for the initial flux map taken after reaching equilibrium conditions.

Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after Verify FM(X,Y,Z) < FL(X,Y,Z)RPS.*

achieving equilibrium conditions after exceeding, by >

10% RTP, the THERMAL POWER at which Fm(X,Y,Z) was last verified AND McGuire Units 1 and 2 3.2.1-5 Amendment Nost

F6H(X,Y) 3.2.2 SURVEILLANCE REQUIREMENTS

---

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

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify FMH(X,Y) is within steady state limit. Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions after exceeding, by >

10% RTP, the THERMAL POWER at which FmH(X,Y) was last verified AND (continued)

McGuire Units 1 and 2 3.2.2-3 Amendment %4ý

FAH(XY) 3.2.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY i

SR 3.2.2.2

1. Extrapolate F'R(X,Y) using at least two measurements to-31 EFPD beyond the most recent measurement. If FmH(X,Y) is within limits and the 31 EFPD extrapolation indicates:

F*.(,Y)ET*PC*T[ >__ FIYH(XRFA L (,yvsuRv V EXTRAPOLATED and F mX1Y)EXTRAPOLATEO >_FaHfx_

F'H (Xy)SURVExRAPOT F(X,Y)suRV then:

a. *Increase F&H(X,Y) by the appropriate factor specified in the COLR and reverify Fm. (XY) < FL. (X Y)SURv; or-

b. Repeat SR 3.2.2.2 prior to the time at which FmH (X,Y) < FLA (X,Y)suRv is extrapolated to not be met.

2. Extrapolation of FmH (X,Y) is not required for the initial flux map taken after reaching equilibrium conditions.

Verify F'. (X,Y)<!S F L (X.Y)SURV. Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions after exceeding, by >

10% RTP, the THERMAL POWER at which Fm, (X,Y) was last verified AND McGuire Units I and 2 3.2.2-4 Amendment Nos  :

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits specified in the COLR.

---

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

The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

APPLICABILITY: MODE 1 with THERMAL POWER > 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL 30 minutes POWER to < 50% RTP.

SURVEILLANCE REQUIREMENTS_ _ _ _

SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore channel.

AND Once within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> thereafter with the AFD monitor alarm inoperable McGuire Units 1 and 2 3.2.3-1 Amendment Nosýý

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

+

SR 3.2.4.1 NOTES ------------------

1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER

<75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance.

Verify QPTR is within limit by calculation.

AND Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter with the QPTR alarm inoperable SR 3.2.4.2 ------------------- NOTES ------------------

Only required to be performed if input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP..

Verify QPTR is within limit using the movable incore. U detectors.

McGuire Units 1 and 2 3.2.4-4 Amendment Nos. C+94:ýD

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

---

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

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK.

SR 3.3.1.2 ------------------- NOTES -----------------

1. Adjust NIS channel if absolute difference is > 2%

RTP.

2. Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is > 15% RTP.

Compare results of calorimetric heat balance calculation u$

to Nuclear Instrumentation System (NIS) channel output.

SR 3.3.1.3 *NOTES ------------------

1. Adjust NIS channel if absolute difference is > 3%

AFD.

2., Not required.to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after . "

THERMAL POWER is>>15% RTP.

Compare results of the incore detector measurements to I NIS AFD. F d y/X...

(continued)

McGuire Units 1 and 2 3.3.1-9 Amendment Nos.

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.4 ----------------- NOTES ---------------

This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.

Perform TADOT.

SR 3.3.1.5 Perform ACTUATION LOGIC TEST. "-i SR 3.3.1.6 --------.... -------.------. NOTES----------

Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 75% RTP.

Calibrate excore channels to agree with incore detector measurements.

SR 3.3.1.7 ----------- NOTES---------

Not required to be performed for source range .

instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entry into MODE 3.

Perform COT.

(continued)

McGuire Units 1 and 2 3.3.1-10 Amendment Nos.

• RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

+/-

SR 3.3.1.8 ----------------- NOTES -------------

This Surveillance shall include verification that interlocks P-6 (for the Intermediate Range channels) and P-10 (for the Power Range channels) are in their required state for existing unit conditions.

Perform COT. NOTE------- -

Only required when not e- Fr & 2 " e iý  !;Ip e

'ý4 F i

FkC7. coý 411-1 ' re Prior to reactor startup OR tv,,* d,,Y;ý AND Four hours after reducing power below P-1O for power and intermediate range instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND 4FIZ T (continued)

McGuire Units 1 and 2 3.3.1-11 Amendment Nosý ,

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.9 ------------------- NOTES ------------------

Verification of setpoint is not required.

Perform TADOT.

SR 3.3.1.10 ------------------

NOTES ------------------

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

SR 3.3.1.11 --------------- NOTES---- --------------

1. Neutron detectors are excluded from CHANNEL CALIBRATION.

2. Power and Intermediate Range Neutron Flux detector plateau voltage verification is not required to be performed- prior to entry into MODE 1 or2.

Perform CHANNEL CALIBRATION. ;0 ,,vs SR 3.3.1.12 Perform CHANNEL CALIBRATION.

SR 3.3.1.13 Perform COT. j/-"

(continued)

McGuire Units 1 and 2 3.3.1-12 Amendment Nos. 2

RTS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY SR 3.3.1.14 -- NOTES---------

Verification of setpoint is not required.

Perform TADOT. rbr(h SR 3.3.1.15 ------------------ NOTES ---------------------- ---- NOTE -----

Verification of setpoint is not required. Only required when not performed within previous 31 days Perform TADOT. Prior to reactor startup SR 3.3.1.16 ------------------ NOTES -------------

Neutron detectors are excluded from response time testing.

Verify RTS RESPONSE TIME is within limits.

SR 3.3.1.17: Verify RTS RESPONSE TIME for RTDs is Within limits.

McGuire Units 1 and 2 3.3.1-13 Amendment Nos.M -

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS p.

I*ll u'*"trvr I I --- ----------

Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.

SURVEILLANCE FREQUENCY SR 3.3.2.1 Perform CHANNEL CHECK.

SR 3.3.2.2 Perform ACTUATION LOGIC TEST.

S R 3.3.2.3 Perform COT.

SR 3.3.2.4 Perform MASTER RELAY TEST.

\

i"

.'j SR 3.3.2.7 -- ---- NOTE ....--- -----------------------

Verification of setpoint not required. for manual initiation functions.

McGuire Units 1 and 2 3.3.2-8 Amendment Nos. -

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.8 NOTE ------------------

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

SR 3.3.2.9 --------- ------- NOTE -----------------

)

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after SG pressure is > 900 psig.

Verify ESFAS RESPONSE TIMES are within limit.

't McGuire Units 1 and-2 3.3.2-9 Amendment Nos.-ý

PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS

---

N -----------------------------------------------------------

SR 3.3.3.1 and SR 3.3.3.3 apply to each PAM instrumentation Function in Table 3.3.3-1.

SR 3.3.3.1 SURVEILLANCE Perform CHANNEL CHECK for each required I FREQUENCY instrumentation channel that is normally energized.

SR 3.3.3.2 Not Used SR 3.3.3.3 - --- NOTE Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION.

I McGuire Units 1 and 2 3.3.3-3 Amendment Nos.oi -

Remote Shutdown System 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized.

SR 3.3.4.2 Verify each required control circuit and transfer switch is capable of performing the intended function.

SR 3.3.4.3 Perform CHANNEL CALIBRATION for each required instrumentation channel.

McGuire Units 1 and 2 3.3.4.2 Amendment Nos.o -

LOP DG Start Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.5.1 Perform TADOT.

SR 3.3.5.2 ------------ NOTE ------------- ---------------

A NOMINAL TRIP SETPOINT associated with this SR 7- /V-5_67- T shall be set within the channel's calibartion tolerance band.

Perform CHANNEL CALIBRATION with NOMINAL TRIP SETPOINT and Allowable Value as follows:

a. Loss of voltage Allowable Value > 3122 V (Unit

1) 3108 V (Unit 2) with a time delay of 8.5 +/- 0.5 second.

Loss of voltage NOMINAL TRIP SETPOINT 3174 V (Unit 1) 3157 V (Unit 2) +/- 45 V with a time delay of 8.5 +/- 0.5 second.

b. Degraded voltage Allowable Value > 3661 V (Unit

1) > 3685.5 V (Unit 2) with a time delay of < 11 seconds with SI and < 600 seconds without SI.

Degraded voltage NOMINAL TRIP SETPOINT 3678.5 V (Unit 1) 3703 V (Unit 2) with a time delay of < 11 seconds with SI and < 600 seconds without SI.

McGuire Units 1 and 2 3.3.5-2 Amendment Nos-,ýý

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is within limits.

SR 3.4.1.2 Verify RCS average temperature is within limits.

SR 3.4.1.3 Verify RCS total flow rate is within limits.

SR 3.4.1.4 Perform CHANNEL CALIBRATION for each RCS total 1 flow indicator.

PA 5J~T£ McGuire Units 1 and 2 3.4.1-3 Amendment Nos.o*--

RCS P/T Limits 3.4.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. -------- NOTE -------- C.1 Initiate action to restore Immediately Required Action C.2 parameter(s) to within shall be completed limits.

whenever this Condition is entered. AND C.2 Determine RCS is Prior to entering Requirements of LCO acceptable for continued MODE 4 not met any time, in other operation.

than MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 ------------ NOTE -.--.-..............------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and RCS heatup and cooldown rates are within limits.

i.

CI-McGuire Units I and 2 3.4.3-2 Amendment Nosý

RCS Loops - MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops-MODES 1 and 2 LCO 3.4.4 Four RCS loops shall be OPERABLE and in operation.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of LCO A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 *Verify each RCS loop is in operation.

McGuire Units 1 and 2 3.4.4-1 Amendment Nosi -

RCS Loops - MODE 3 3.4.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation.

N SR 3.4.5.2 Verify steam generator secondary side water levels are

> 12% narrow range for required RCS loops.

SR 3.4.5.3 Verify correct breaker alignment and indicated power are available to the required pumps that are not in operation.

7~L McGuire Units 1 and 2 3.4.5-3 Amendment Nos.< tr

RCS Loops - MODE 4 3.4.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One RHR loop B.1 Be in MODE 5. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE.

AND ALL RCS loops inoperable.

C. Both required RCS or C.1 Suspend operations that Immediately RHR loops inoperable, would cause introduction of coolant into the RCS with OR boron concentration less than required to meet the No RCS or RHR loop in SDM of LCO 3.1.1 and operation. maintain Keff < 0.99.

AND C.2 Initiate action to restore Immediately one loop to OPERABLE status and operation.

" Z7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation.

SR 3.4.6.2 Verify SG secondary side water levels are > 12% narrow range for required RCS loops.

SR 3.4.6.3 Verify correct breaker alignment and indicated Power are available to the required pump that is not in operation.

McGuire Units 1 and 2 3.4.6-2 Amendment Nos4 -

RCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

+/-

SR 3.4.7.1 -Verify one RHR loop is in operation.

SR 3.4.7.2 Verify SG secondary side water level is > 12% narrow range in required SGs.

SR 3.4.7.3 Verify correct breaker alignment and indicated power are available to the required RHR pump that is not in operation.

(

McGuire Units 1 and 2 3.4.7-3 Amendment Nos-OýEý

RCS Loops - MODE 5, Loops Not Filled 3.4.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required RHR loops B.1 Suspend operations that Immediately inoperable, would cause introduction of coolant into the RCS with OR boron concentration less than required to meet SDM No RHR loop in of LCO 3.1.1.

operation.

AND B.2 Initiate action to restore Immediately one RHR loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 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.

McGuire Units I and 2 3.4.8-2 Amendment Nosýý

Pressurizer 3.4.9 McGuire Units 1 and 2 3.4.9-2 Amendment NoSq

Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 ------------------ NOTE-Not required to be met with block valve closed in accordance with the Required Action of Condition A, B, or E.

Perform a complete cycle of each block valve.

SR 3.4.11.2 ---------------- NOTE- ------------------

Required to be performed in MODE 3 or MODE 4 when the temperature of all RCS cold legs is > 300'F and the block valve closed.

Perform a complete cycle of each PORV.

SR 3.4.11.3 Verify the nitrogen supply for each PORV is OPERABLE by:

a. Manually transferring motive power from the air supply to the nitrogen supply,

b. Isolating and Venting the air supply, and C. Operating the PORV through one complete. cycle.

McGuire Units 1 and 2 3.4.11-4 Amendment Nos. 6ý

LTOP System 3.4.12

,SR 3.4.12.4 - ------------ NOTE- -

Only required to be performed when complying with LCO 3.4.12.b.

Verify RCS vent > 2.75 square inches open.

McGuire Units 1 and 2 3.4.12-5 Amendment Nos --

LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.12.6 - -------------- NOTE -------------------

Not required to be met until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing RCS cold leg temperature to < 300'F.

Perform a COT on each required PORV, excluding actuation.

SR 3.4.12.7 Perform CHANNEL CALIBRATION for each required PORV actuation channel.

McGuire Units I and 2 3.4.12-6 Amendment Nos -ý

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 ------------------ NOTES -------------------------- NOTE-Only required to

1. Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after be performed establishment of steady state operation. during steady state operation

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS Operational LEAKAGE is within limits by performance of RCS water inventory balance.

SR 3.4.13.2 ------------------ NOTE ------------------

Not required to be performed Until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

Verify primary to secondary LEAKAGE is < 135 gallons per day through any one SG and < 389 gallons per day total through all SGs.

McGuire Units I and 2 3.4.13-2 Amendment Nos.o"

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 ------------ NOTE-

1. ' Not required to be per

. formed in MODES 3 and 4.

2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.

3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

Verify leakage from each RCS PIV is equivalent to < 0.5 In accordance with gpm per nominal inch of valve size up to a maximum of 5 the Inservice gpm at an RCS pressure > 2215 psig and < 2255 psig. Testin .oPto am, and ýW)

AND Prior to entering dtF7-i MODE 2 whenever the unit has been in MODE 5 for.

7 days or more, if leakage testing has not been performed in the previous 9 months AND Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the valve (continued)

McGuire Units 1 and 2 3.4.14-3 Amendment Nos_(ýý

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.14.2 Verify RHR system interlock prevents the valves from being opened with a simulated or actual RCS pressure signal > 425 psig.

JT/EF7-1 McGuire Units 1 and 2 3.4.14-4 Amendment Nos-oý

RCS Leakage Detection instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of the containment atmosphere particulate radioactivity monitor.

SR 3.4.15.2 Perform COT of the containment atmosphere particulate radioactivity monitor.

SR 3.4.15.3 Perform CHANNEL CALIBRATION of the containment floor and equipment sump level monitors.

SR 3.4.15.4 Perform CHANNEL CALIBRATION of the containment atmosphere particulate radioactivity monitor.

SR 3.4.15.5 Perform CHANNEL CALIBRATION of the containment 08 9/t: --

ventilation unit condensate drain tank level monitor.

SR 3.4.15.6 Perform CHANNEL CALIBRATION of the incore instrument sump level alarm.

McGuire Units I and 2 3.4.15-4 Amendment Nos*

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Tavg < 500 0 F.

Time of Condition A not met.

OR DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.16-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific activity < 100/E pCi/gm.

SR .3.4.16.2 ---------------

NOTE- . , ......

Only required to be performed in MODE I.

Verify reactorcoolant DOSE EQUIVALENT 1-131 specific activity < 1.0 pCi/gm.

AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of > 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period (continued)

McGuire Units 1 and 2 3.4.16-2 Amendment Nos.-E -

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.16.3 ------------------ NOTE -------------------

Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was A'5~ TI last subcritical for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Determine E from a sample taken in MODE 1 after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

McGuire Units 1 and 2 3.4.16-3 AmendmentNo

RCS Loops - Test Exceptions 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 RCS Loops-Test Exceptions LCO 3.4.17 The requirements of LCO 3.4.4, "RCS Loops--MODES I and 2," may be suspended, with THERMAL POWER < P-7.

APPLICABILITY: MODES 1 and 2 during startup and PHYSICS TESTS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. THERMAL POWER > A.1 Open reactor trip breakers. Immediately P-7.

SURVEILLANCE REQUIREMENTS SURVEILLANCE, FREQUENCY SR 3.4.17.1 Verify THERMAL POWER is < P-7. / '

SR 3.4.17.2 Perform a COT for each power range neutron flux-low Prior to initiation of and intermediate range neutron flux channel and P-7. startup and PHYSICS TESTS McGuire Units 1 and 2 3.4.17-1 Amendment NoS -

Accumulators 3.5.1

• - SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is full, S>R 3.5.1.2 Verify borated water volume in each accumulator is

> 6870 gallons and < 7342 gallons.

SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator i*-

_>585 psig and _<639 psig.,

SR 3.5.1.4 Verify boron concentration in each accumulator is the limits specified in the COLR.

AND NOTE-------- -

Only required to be performed for affected accumulators Once within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of > 1%

of tank volume that is not the result of addition from the refueling water storage tank S R 3.5.1.5 Verify power is removed from each accumulator isolation valve operator when RCS pressure is > 1000 psig.

McGuire Units 1 and 2 3.5.1-2 Amendment Nos! -

ECCS - Operating 3.5.2

,E REQUIREMENTS SURVEILLANCE FREQUEN SR 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 SI Cold Leg Injection NI121A Closed SI Hot Leg Injection NI152B Closed SI Hot Leg Injection N1183B Closed RHR Hot Leg Injection N1173A Open RHR Cold Leg Injection N1178B Open RHR Cold Leg JJA'SC/R Injection N1100B Open SI Pump RWST Suction FW27A Open RHR/RWST Suction N1147A Open SI Pump Mini-Flow SR 3.5.2.2 Verify each ECCS manual, power operated, and automatic valve in the flow paththat is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.2.3 Verify ECCS piping is full of water.

McGuire Units 1 and 2 3.5.2-2 Amendment Nos.*-

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the Inservice developed head. Testing Program SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is I not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual or simulated actuation signal.

SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each position stop is in the correct position.

Centrifugal Charging Safety Injection Pump Injection Throttle Pump Throttle Valve Number Valve Number N1480 N1488 N1481 N1489 N1482 N1490 N1483 N1491 SR 3.5.2.8 Verify, by visual inspection, that the ECCS containment sump strainer assembly and the associated enclosure are not restricted by debris and show no evidence of structural distress or abnormal corrosion.

McGuire Units 1 and 2 3.5.2-3 Amendment Nos.-

RWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 Verify RWST borated water temperature is > 70'F and V4

< 1000 F.

SR 3.5.4.2 Verify RWST borated water volume is > 372,100 gallons.

SR 3.5.4.3 Verify RWST boron concentration is within the limits specified in the COLR.

6EE McGuire Units 1 and 2 3.5.4-2 Amendment No&s(-

Seal Injection Flow 3.5.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.5.1 ------------------ NOTE -------------

Not required to be performed until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the Reactor Coolant System pressure stabilizes at IS',r T

> 2215 psig and < 2255 psig.

Verify manual seal injection throttle valves are adjusted to give a flow within limit with centrifugal charging pump operating and the charging flow control valve full open.

McGuire Units 1 and 2 3.5.5-2 Amendment Nos r-

Containment Air Locks 3.6.2 ACTIONS (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1 NOTE- -------------- -------------------

1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.

Perform required air lock leakage rate testing in In accordance with accordance with the Containment Leakage Rate Testing the Containment Program. Leakage Rate Testing Program.

SR 3.6.2.2 Perform a pressure test on each inflatable air lock door seal and verify door seal leakage is < 15 sccm. A-1) 1V~j-X SR 3.6.2.3 Verify only one door in the air lock can be opened at a - /

time.

McGuire Units 1 and 2 3.6.2-5 Amendment Nos-ý

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each containment purge supply and exhaust valve for the lower compartment, upper compartment, and incore instrument room is sealed closed, except for one purge valve in a penetration flow path while in Condition E of this LCO.

SR 3.6.3.2 . Not Used.

SR 3.6.3.3 NOTE ----------------

Valves and blind flanges in high radiation areas may be verified by use of administrative controls.

Verify each containment isolation manual valve and blind flange that is located outside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

(continued)

McGuire Units 1 and 2 3.6.3-5 Amendment Nos 1 4

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.4 --------------- NOTE---------

Valves and blind flanges in high radiation areas may be verified by use of administrative controls.

Verify each containment isolation manual valve and blind Prior to entering flange that is located inside containment or annulus and MODE 4 from not locked, sealed, or otherwise secured and required to MODE 5 if not be closed during accident conditions is closed, except for performed within containment isolation valves that are open under the previous administrative controls. 92 days SR 3.6.3.5 Verify the isolation time of automatic power operated In accordance with containment isolation valve is within limits, the Inservice Testing Program In accordance with SR 3.6.3.6 Perform leakage rate testing for containment purge lower the Containment and upper compartment and incore Instrument room Leakage Rate valves with resilient seals. Testing Program SR 3.6.3.7 Verify each automatic containment isolation valve that .is not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal. L (continued)

McGuire Units I and 2 3.6.3-6 Amendment Nos

Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 Containment pressure shall be _>-0.3 psig and _<+0.3 psig.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment pressure A.1 Restore containment 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> not within limits, pressure to within limits.

13. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE _REQUIREMENTS ____"_"_....

-SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits.

McGuire Units 1 and 2 3.6.4-1 Amendment Nos_. -

Containment Air Temperature 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment upper compartment average air temperature is within limits.

SR 3.6.5.2 Verify containment lower compartment average air "1 temperature is within limits.

2?,V5~6P r1.

McGuire Units 1 and 2 3.6.5-2 Amendment Noazi -

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 COMPLETION TIME A. One containment spray A.1 Restore containment spray 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> train inoperable, train to OPERABLE status.

B. Required Action and B. 1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct /9, position.

(continued)

McGuire Units 1 and 2 3.6.6-1 Amendment Nos4 ..-

Containment Spray System 3.6.6 SURVEILLANCE FREQUENCY SR 3.6.6.2 Verify each containment spray pump's developed head at In accordance with the flow test point is greater than or equal to the required the Inservice developed head. Testing Program SR 3.6.6.3 Verify each automatic containment spray valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.6.4 Verify each containment spray pump starts automatically on an actual or simulated actuation signal.

SR 3.6.6.5 Verify that each spray pump is de-energized and prevented from starting upon receipt of a terminate signal and is allowed to start upon receipt of a start permissive from the Containment Pressure Control System (CPCS).

SR 3.6.6.6 Verify that each spray pump discharge valve closes or is prevented from opening upon receipt of a terminate signal and is allowed to open upon receipt of a start permissive from the Containment Pressure Control System (CPCS).

SR 3.6.6.7 Verify each spray nozzle is unobstructed.

McGuire Units 1 and 2 3.6.6-2 Amendment Nos--

HSS 3.6.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.8.1 Operate each HSS train for > 15 minutes.

SR 3.6.8.2 Verify the fan motor current is < 21.5 amps when the fan speed is Ž_3579 rpm and < 3619 rpm with the hydrogen skimmer fan operating and the motor operated suction valve closed.

SR 3.6.8.3 Verify the motor operated suction valve opens automatically and the hydrogen skimmer fans receive a start permissive signal from the Containment Pressure Control System.

SR 3.6.8.4 Verify each HSS train starts on an actual or simulated actuation signal after a delay of > 8 minutes and < 10 minutes.

McGuire Units 1 and 2 3.6.8-2 Amendment Nos-4 -

HMS 3.6.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.9.1 Energize each HMS train power supply breaker and y verify >_ 34 ignitors are energized in each train.

SR 3.6.9.2 Verify at least one hydrogen ignitor is OPERABLE in each containment region.

SR 3.6.9.3 Energize each hydrogen ignitor and verify temperature is A t-s

_ 1700 0F.

IR McGuire Units 1 and 2 3.6.9-2 Amendment Nosýý

AVS 3.6.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.10.1 Operate each AVS train for > 10 continuous hours with heaters operating.

SR 3.6.10.2 SR 3.6.10.3 Perform required AVS filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

Verify each AVS train actuates on an actual or simulated actuation signal.

4 In accordance with the VFTP J )

SR 3.6.10.4 Verify each AVS filter cooling bypass valve can be opened.

SR 3.6.10.5 Verify each AVS train flow rate is > 7200 cfm and < 8800 cfm.

t//SERT:L McGuire Units 1 and 2 .3.6.10-2 Amendment Nos, ý

ARS 3.6.11 3.6 CONTAINMENT SYSTEMS 3.6.11 Air Return System (ARS)

LCO 3.6.11 Two ARS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ARS train A.1 Restore ARS train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.11.1 Verify each ARS fan starts on an actual or simulated actuation signal, after a delay of >_8 minutes and

< 10 minutes, and operates for > 15 minutes.

2EýýV4

_fAlrZ-9 7--L (continued)

McGuire Units 1 and 2 3.6.11-1 Amendment N

ARS 3.6.11 SURVEILLANCE jFREQUENCY SR 3.6.11.2 Verify, with the ARS fan damper closed and with the.

bypass dampers open, each ARS fan motor current is

_*32.0 amps when the fan speed is _>840 rpm and < 900 rpm.

SR 3.6.11.3 Verify, with the ARS fan not operating, each ARS motor operated damper opens automatically on an actual or simulated actuation signal after a delay of 2! 9 seconds and < 11 seconds.

,)

SR 3.6.11.4 Verify the check damper is open with the air return fan operating.

SR 3.6.11.5 Verify the check damper is closed with the air return fan not operating.

SR 3.6.11.6 Verify that each ARS fan is de-energized or is prevented from starting upon receipt of a terminate signal and is allowed to start upon receipt of a start permissive from the. Containment Pressure Control System (CPCS).

SR 3.6.11.7 Verify that ARS fan motor-operated damper is allowed to

.open upon receipt of a start permissive from the Containment Pressure Control System (CPCS) and is prevented from opening in the absence of a start permissive.

McGuire Units 1 and 2 3.6.11-2 Amendment Nos(ý

Ice Bed 3.6.12 3.6 CONTAINMENT SYSTEMS 3.6.12 Ice Bed LCO 3.6.12 The ice bed shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Ice bed inoperable. A. 1 Restore ice bed to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS McGuire Units 1 and 2 3.6.12-1 Amendment Nos.A

Ice Bed 3.6.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.12.2 ---------------- NOTE -------------------------------------

The chemical analysis may be performed on either the liquid solution or on the resulting ice.

Verify, by chemical analysis, that ice added to the ice Each ice addition condenser meets the boron concentration and pH requirements of SR 3.6.12.7.

SR 3.6.12.3 Verify by visual inspection, accumulation of ice on structural members comprising flow channels through the ice bed is < 15 percent blockage of the total flow area for each safety analysis section.  !-TA P- 7 SR 3.6.12.4 Verify total mass of stored ice is > 1,890,000 lbs by calculating the mass of stored ice, at a 95 percent confidence, in each of three Radial Zones as defined below, by selecting a random sample of >_30 ice baskets in each Radial Zone, and Verify:

1. Zone A (radialrows 8, 9), has a total mass of

> 313,200 lbs

2. Zone B.(radial rows 4, 5, 6, 7), has a total mass of

> 901,000 lbs

3. Zone C (radial rows 1, 2, 3), has a total mass of

.> 675,800 lbs (continued)

McGuire Units 1 and 2 3.6.12-2 Amendment Nos4ýý

Ice Bed 3.6.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.12.5 Verify that the ice mass of each basket sampled in SR 3.6.12.4 is _>600 lbs.

SR 3.6.12.6 Visually inspect, for detrimental structural wear, cracks, corrosion, or other damage, two ice baskets from each, group of bays as defined below:

a. Group 1 - bays 1 through 8;

b. Group 2 - bays 9 through 16; and

c. Group 3 - bays 17 through 24. )

SR 3.6.12.7 --------------- NOTE -------------------

The requirements of this SR are satisfied if the boron concentration and pH values obtained from averaging the individual sample results are within the limits specified below.

Verify, by chemical analysis of the stored ice in at least one randomly selected ice basket from each ice condenser bay, that. ice bed:

a. Boron concentration is > 1800 ppm and < 2330 ppm; and

b. pH is> 9.0 and< 9.5.

McGuire Units 1 and 2 3.6.12-3 Amendment NosQý -

Ice Condenser Doors 3.6.13 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Restore ice condenser door 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> associated Completion to OPERABLE status and Time of Condition B not closed position.

met.

D. Required Action and D. 1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or C AND not met.

D.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.13.1 Verify all inlet doors indicate closed by the Inlet Door Position Monitoring System.

SR 3.6.13.2 Verify, by visual inspection, each intermediate deck door is closed and not impaired by ice, frost, .or debris..

I

.SR 3.6.13.3 Verify, by visual inspection, each top deck door:

a. Is in place; and

b. Has no condensation, frost, or ice formed on the door that would restrict its opening.

(continued)

McGuire Units 1 and 2 3.6.13-2 Amendment Nosýý'-

Ice Condenser Doors 3.6.13 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.13.4 Verify, by visual inspection, each inlet door is not n impaired by ice, frost, or debris.

SR 3.6.13.5 Verify torque required to cause each inlet door to begin to open is _<675 in-lb.

SR 3.6.13.6 Perform a torque test on each inlet door. pn'-

SR 3.6.13.7 Verify for each intermediate deck door: i~4ntps~

a. No visual evidence of structural deterioration;

b. Free movement of the vent assemblies; and 7 /2T/V-~*P 7

c. Free movement of the door.

McGuire Units 1 and 2 3.6.13-3 Amendment No4

Divider Barrier Integrity 3.6.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.14.1 Verify, by visual inspection, all personnel access doors Prior to entering and equipment hatches between upper and lower MODE 4 from containment compartments are closed. MODE 5 SR 3.6.14.2 Verify, by visual inspection, that the seals and sealing Prior to final surfaces of each personnel access door and equipment closure after each hatch have: opening

a. No detrimental misalignments; AND

b. No cracks or defects in the sealing surfaces; and NOTE .......

Only required for

c. No apparent deterioration of the seal material, seals made of resilient materials SR 3.6.14.3 Verify, by visual inspection, each personnel access door After each opening or equipment hatch that has been opened for personnel transit entry is closed.

SR 3.6.14.4 Remove two divider barrier seal test coupons and verify X both test coupons tensile strength is _>39.7 psi.

(continued)

McGuire Units 1 and 2 3.6.14-2 Amendment Nos-6 -

Divider Barrier Integrity 3.6.14 SURVEILLANCE FREQUENCY SR 3.6.14.5 Visually inspect > 95% of the divider barrier seal length, and verify:

a. Seal and seal mounting bolts are properly installed; and 5CI~T*l-

b. Seal material shows no evidence of deterioration due to holes, ruptures, chemical attack, abrasion, radiation damage, or changes in physical appearance.

McGuire Units 1 and 2 3.6.14-3 Amendment No ý9 -

Containment Recirculation Drains 3.6.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.15.1 Verify, by visual inspection, that: Prior to entering MODE 4 from

a. Each refueling canal drain valve is locked open; MODE 5 after each partial or

b. Each refueling canal drain is not obstructed by complete fill of the debris; and canal SR 3.6.15.2 Verify, by visual inspection, that no debris is present in the upper compartment or refueling canal that could obstruct the refueling canal drain.

SR 3.6.15.3 Verify for each ice condenser floor drain that the:

a. Valve opening is not impaired by ice, frost, or debris;

b. Valve seat shows no evidence of damage;

c. Valve opening force is <66 lb; and

d. Drain line from the ice condenser floor to the lower compartment is unrestricted.

McGuire Units 1 and 2 3.6.15-2 Amendment No&Aiýiýý

Reactor Building 3.6.16 3.6 CONTAINMENT SYSTEMS 3.6.16 Reactor Building LCO 3.6.16 The reactor building shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor building A.1 Restore reactor building to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS _ _ __ _

SURVEILLANCE FREQUENCY SR 3.6.16.1 Verify the door in each access opening is closed, except when the access opening is being used for normal transit entry and exit.

tT-(continued)

McGuire Units 1 and 2 3.6.1,6-1 Amendment Nos.*

Reactor Building 3.6.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.16.2 Verify each Annulus Ventilation System train produces a pressure equal to or more negative than -0.5 inch water gauge in the annulus within 22 seconds after a start signal and -3.5 inches water gauge after 48 seconds.

Verifying that upon reaching a negative pressure of -3.5 inches water gauge in the annulus, the system switches into its recirculation mode of operation and that the time required for the annulus pressure to increase to -0.5 inch water gauge is _>278 seconds.

SR 3.6.16.3 Verify reactor building structural integrity by performing a visual inspection of the exposed interior and exterior surfaces of the reactor building.

McGuire Units 1 and 2 3.6.16-2 Amendment Nos -

SG PORVs 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify one complete cycle of each SG PORV.

SR 3.7.4.2 Verify one complete cycle of each SG PORV block valve.

McGuire Units 1 and 2 3.7.4-2 Amendment Nos.-

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 --------------- NOTE Not applicable to automatic valves when THERMAL POWER is < 10% RTP.

Verify each AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not d ly locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.5.2 -------------------- NOTE ------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after > 900 psig in the steam generator.

Verify the developed head of each AFW pump at the flow In accordance test point is greater than or equal to the required with the Inservice developed head. Testing Program SR 3.7.5.3 -------------- -- NOTE&------------- ......---.....

Not applicable in MODE 4 when steam generator is relied upon for heat removal.

Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

(continued)

McGuire Units 1 and 2 3.7.5-3 Amendment Nos.*

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.5.4 ------------------ NOTE------------------

1. Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after > 900 psig in the steam generator.

21 Not applicable in MODE 4 when steam generator is relied upon for heat removal. L 7

-T-A/

Verify each AFW pump starts automatically on an actual njX70 or simulated actuation signal.

McGuire Units 1 and 2 3.7.5-4 Amendment Nos Aý .-

CCW System 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 - -------------- NOTE ------------------

Isolation of CCW flow to individual components does not render the CCW System inoperable.

Verify each CCW manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.6.2 Verify each CCW automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.7.6.3 Verify each CCW pump starts automatically on an actual or simulated actuation signal.

McGuire Units I and 2 3.7.6-2 Amendment Nos.o -

NSWS 3.7.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 NOTE- -------------

NOTE ..--------

Isolation of NSWS flow to individual components does not render the NSWS inoperable.

Verify each NSWS manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.7.2 Verify each NSWS automaticvalve in the flow path servicing safety related equipment, that is not locked,

)

sealed, or otherwise secured in position, actuates to the correct positionon an actual or simulated actuation.

signal..

SR 3.7.7.3 Verify each NSWS pump starts automatically on an actual or simulated actuation signal.

7ý,V,-<-e Z T-C Eý McGuire Units 1 and 2 3.7.7-2 Amendment Nos(ýý

SNSWP 3.7.8 SURVEILLANCE REQUIREMENTS (continued) 3.7 PLANT SYSTEMS 3.7.8 Standby Nuclear Service Water Pond (SNSWP)

LCO 3.7.8 The SNSWP shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SNSWP inoperable. A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND A.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 Verify water level of SNSWP is"> 739.5 ft mean sea level.

SR 3.7.8.2 -------------------- NOTE------------

Only required to be performed during the months of July, )

August, and September.

Verify average water temperature of SNSWP is _<82 0 F at an elevation of 722 ft. in SNSWP.

Icniud (continued)

McGuire Units I and 2 3.7.8-1 Amendment Nos(

SNSWP 3.7.8 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.8.3 Verify, by visual inspection, no abnormal degradation, /in of the SNSWP dam.

erosion, or excessive seepage McGuire Units 1 and 2 3.7.8-2 Amendment Nosý

CRAVS 3.7.9 CONDITION REQUIRED ACTION COMPLETION TIME G. One or more CRAVS G.1 Restore CRAVS train(s) 7 days train(s) heater heater to OPERABLE inoperable, status.

OR G.2 Initiate action in 7 days accordance with Specification 5.6.6.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Operate each CRAVS trainfor _ 10 continuous hours A',7iz4 with the heaters operating.

7 SR 3.7.9.2 Perform required CRAVS filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP SR. 3.7.9.3 Verify each CRAVS. train actuates on an actual or simulated actuation signal.

SR 3.7.9.4 Perform required CRE unfiltered air inleakage testing in In accordance with accordance with the Control Room Envelope Habitability the Control Room Program. Envelope Habitability Program McGuire Units 1 and 2 3.7.9-3 Amendment No-

CRACWS 3.7.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Two CRACWS trains D. 1 Suspend CORE Immediately inoperable in MODE 5 ALTERATIONS.

or 6, or during movement of irradiated AND fuel assemblies, or during CORE D.2 Suspend movement of Immediately ALTERATIONS. irradiated fuel assemblies.

E. Two CRACWS trains E.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Verify the control room temperature is < 90 0 F.

-rA'.*

.1.1 McGuire Units 1 and 2 3.7.10-2 Amendment N o*s' -

ABFVES 3.7.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

+

SR 3.7.11.1 Operate each ABFVES for >15 minutes.

TI SR 3.7.11.2 Perform required ABFVES filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.11.3 Verify each ABFVES actuates on an actual or simulated actuation signal.

SR 3.7.11.4 Verify one ABFVES can maintain a pressure

_<-0.125 inches water gauge in the ECCS pump room area relative to atmospheric pressure during the post accident mode of operation.

7-1.

McGuire Units 1 and 2 3.7.11-2 Amendmentý -

FHVES 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Fuel Handling Ventilation Exhaust System (FHVES)

LCO 3.7.12 The FHVES shall be OPERABLE and in operation.

APPLICABILITY: During movement of irradiated fuel assemblies in the fuel building.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. FHVES inoperable. ------------- NOTE ----------

LCO 3.0.3 is not applicable.

A.1 Suspend movement of Immediately irradiated fuel assemblies in the fuel building.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Verify FHVES in operation. M SR 3.7.12.2 Operate FHVES for 2! 15 minutes. Prior to movement of irradiated fuel assemblies (continued) 7~_gvTD.

McGuire Units 1 and 2 3.7.12-1 Amendment Nosýý

FHVES 3.7.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.12.3 Perform required FHVES filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.12.4 Verify FHVES can maintain an exhaust flow rate

> 8000 cfm greater than the supply flow rate.

SR 3.7.12.5 Verif',

McGuire Units 1 and 2 3.7.12-2 Amendment Nos,-ý

Spent Fuel Pool Water Level 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Spent Fuel Pool Water Level LCO 3.7.13 The spent fuel pool water level shall be > 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool water A.1 -------- NOTE--------

level not within limit. LCO 3.0.3 is not applicable.

Suspend movement of Immediately irradiated fuel assemblies in the spent fuel pool.

.SURVEILLANCE REQUIREMENTS -

SURVEILLANCE:' FREQUENCY SR 3.7.13.1 Verify the spent fuel pool water level is > 23 ft above the top of the irradiated fuel assemblies seated in the storage racks.

McGuire Units 1 and 2 3.7.13-1 Amendment No". )

Spent Fuel Pool Boron Concentration 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Boron Concentration LCO 3.7.14 The spent fuel pool boron concentration shall be within the limit specified in the COLR.

APPLICABILITY: When fuel assemblies are stored in the spent fuel pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool boron- ---------- NOTE ----------

concentration not within LCO 3.0.3.is not applicable.

lim it.

A.1 Suspend movement of fuel Immediately assemblies in the spent fuel pool.

AND A.2 Initiate action to restore Immediately spent fuel pool boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the spent fuel pool boron concentration is within limit.

McGuire Units 1 and 2 3.7.14-1 Amendment No,,*

Secondary Specific Activity 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Secondary Specific Activity LCO 3.7.16 The specific activity of the secondary coolant shall be *50.10 gCi/gm DOSE EQUIVALENT 1-131.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> within limit.

AND A.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify. the specific..activity of the secondary coolant is d y

• 0.10 gCi/gm DOSE EQUIVALENT 1-131.

McGuire Units 1 and 2 3.7.16-1 Amendment Nos*

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i

SR 3.8.1.1 Verify correct breaker alignment and indicated power availability for each offsite circuit.

SR 3.8.1.2 --------------- NOTES ----------------

1. Performance of SR 3.8.1.7 satisfies this SR.

2. All DG starts may be preceded by an engine prelube period and followed by a warmup period

)

prior to loading.

3. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer.

When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.

Verify each DG starts from standby conditions and achieves steady state voltage > 3740 V and < 4580 V, and frequency _>58.8 Hz and *<61.2 Hz.

(continued)

McGuire Units 1 and 2 3.8.1-5 Amendment Nos<-

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.3 ------------------ NOTES ----------

1. DG loadings may include gradual loading as recommended by the manufacturer.

2. Momentary transients outside the load range do not invalidate this test.

3. This Surveillance shall be conducted on only one DG at a time.

4. This SR shall be preceded by and immediately follow without shutdown a successful performance of SR 3.8.1.2 or SR 3.8.1.7.

Verify each DG is synchronized and loaded and operates for _>60 minutes at a load _>3600 kW and < 4000 kW.

SR 3.8.1.4 Verify each day tank contains > 120 gal of fuel oil.

SR 3.8.1.5 Check for and remove accumulated water from each day tank.

SR 3..8.1.6 Verify the fuel oil transfer system operates to automatically transfer fuel oil from storage tank to the day tank.

(continued)

McGuire Units 1 and 2 3.8.1-6 Amendment Nos. -

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.7 --

NOTES All DG starts may be preceded by an engine prelube period.

Verify each DG starts from standby condition and achieves in < 11 seconds voltage of > 3740 V and frequency of > 57 Hz and maintains steady state voltage

> 3740 V and

• 4580 V, and frequency >_58.8 Hz and

• 61.2 Hz.

SR 3.8.1.8 -------------------- NOTES -----------------

This Surveillance shall not be performed in MODE 1 or 2.

Verify automatic and manual transfer of AC power sources from the normal offsite circuit to each alternate offsite circuit.

(continued)

W McGuire Units 1 and 2 3.8.1-7 Amendment Nos.@

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.9 Verify each DG, when connected to its bus in parallel with offsite power and operating with maximum kVAR loading that offsite power conditions permit, rejects a load greater than or equal to its associated single largest post-accident load, and:

a. Following load rejection, the frequency is _<63 Hz;

b. Within 3 seconds following load rejection, the voltage is > 3740 V and _<4580 V; and

c. Within 3 seconds following load rejection, the frequency is _>58.8 Hz and _< 61.2 Hz.

SR 3.8.1.10 Verify each DG does not trip and voltage is maintained

< 4784 V during and following a load rejection of

_3600 kW and

• 4000 kW.:

(continued)

McGuire Units 1 and 2 3.8.1-8 Amendment Nos =

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.11 ------------------- NOTES --------

1. All DG starts may be preceded by an engine prelube period.

2. This Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify on an actual or simulated loss of offsite power signal:

a. De-energization of emergency buses;

b. Load shedding from emergency buses;

c. DG auto-starts from standby condition and:

1. energizes the emergency bus in 11 seconds,

2. energizes auto-connected blackout loads through automatic load sequencer,

3. maintains steady state voltage

>_3740Vand.<4580V,

4. maintains steady state frequency

> 58.8 Hz and*< 61.2 Hz, and

5. supplies auto-connected blackout loads for

>5 minutes.

(continued)

McGuire Units 1 and 2 3.8.1-9 Amendment Nos4

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.12 ------------------ NOTES ----------------

All DG starts may be preceded by prelube period.

Verify on an actual or simulated Engineered Safety Feature (ESF) actuation signal each DG auto-starts from standby condition and:

a. In _<11 seconds after auto-start signal achieves voltage of _>3740 and during tests, achieves steady state voltage >_3740 V and _<4580 V;

b. In < 11 seconds after auto-start signal achieves frequency of _>57 Hz and during tests, achieves steady state frequency _>58.8 Hz and < 61.2 Hz;

c. Operates for _>5 minutes; and

d. The emergency bus remains energized from the offsite power system.

(continued)

McGuire Units 1 and 2 3.8.1-10 Amendment Nos3.

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.13 Verify each DG's non-emergency automatic trips are bypassed on actual or simulated loss of voltage signal on the emergency bus concurrent with an actual or simulated ESF actuation signal.

SR 3.8.1.14 ----------------- NOTES ----------------

1. Momentary transients outside the load range do not invalidate this test.

2. DG loadings may include gradual loading as recommended by the manufacturer.

Verify each DG, when connected to its bus in parallel with offsite power and operating with maximum kVAR loading that offsite power conditions permit, operates for

_>24 hours:

a. For _>2 hours loaded > 4200 kW and

• 4400 kW; and

b. For the remaining hours of the test loaded

_>3600 kW and _<4000 kW.

(continued)

McGuire Units 1 and 2 3.8.1-11 Amendment No(-

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.15 -------- --------- NOTES -----------------

1. This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated _> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded > 3600 kW and 4000 kW.

Momentary transients outside of load range do not invalidate this test.

2. All DG starts may be preceded by an engine prelube period.

Verify each DG starts and achieves, in < 11 seconds, voltage >_3740 V, and frequency > 57 Hz and maintains steady state voltage > 3740 V and < 4580 V and frequency _i58.8 Hz and < 61.2 Hz.

)

SR 3.8.1.16 NOTES --

This Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify each DG:

a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;

b. Transfers loads to offsite power source; and

c. Returns to standby operation.

(continued)

McGuire Units 1 and 2 3.8.1-12 Amendment Nos.c

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.17 NOTES--

This Surveillance shall not be performed in MODE 1, 2, 3, or4.

Verify, with a DG operating in test mode and connected to its bus, an actual or simulated ESF actuation signal overrides the test mode by:

a. Returning DG to standby operation; and

b. Automatically energizing the emergency load from offsite power.

SR 3.8.1.18 Verify interval between each sequenced load block is within design interval for each automatic load sequencer.

(continued)

McGuire Units 1 and 2 3.8.1-13 Amendment No<ýýýý

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.19 ------------- NOTES----------------

1. All DG starts may be preceded by an engine prelube period.

2. This Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify on an actual or simulated loss of offsite power signal in conjunction with an actual or simulated ESF actuation signal: 8--

a. De-energization of emergency buses;

b. Load shedding from emergency buses; and

c. DG auto-starts from standby condition and:

1. energizes the emergency bus in _<11 seconds,

2. energizes auto-connected emergency loads through load sequencer,

3. .achieves steady state voltage > 3740 V and _< 4580 V,

4. achieves steady state frequency > 58.8 Hz and 61.2 Hz, and

5. supplies auto-connected emergency loads for >_5 minutes.

(continued)

McGuire Units 1 and 2 3.8.1-14 Amendment Nosý

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) _______

SURVEILLANCE FREQUENCY 1~

SR 3.8.1.20 - -----------------NOTES-----------------

All DG starts may be preceded by an engine prelube period.

Verify when started simultaneously from standby condition, each DG achieves, in < 11 seconds, voltage of

> 3740 V and frequency of > 57 Hz and maintains steady state voltage > 3740 V and < 4580 V, and frequency

> 58.8 Hz and < 61.2 Hz.

McGuire Units 1 and 2 3.8.1-15 Amendment No.(ýýý

Diesel Fuel Oil and Starting Air 3.8.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify the fuel oil storage system contains >_39,500 gal of / Y , ,

fuel for each DG.

SR 3.8.3.2 Verify fuel oil properties of new and stored fuel oil are In accordance with tested in accordance with, and maintained within the the Diesel Fuel Oil limits of, the Diesel Fuel Oil Testing Program. Testing Program SR 3.8.3.3 Verify each DG air start receiver SR 3.8.3.4 Check for and remove accumulated water from the fuel oil storage tank.

McGuire Units 1 and 2 3.8.3-3 Amendment Nos.4

DC Sources - Operating 3.8.4 SURVEILLANCEREQUIREMENTS_______

SR 3.8.4.1 SURVEILLANCE Verify battery terminal voltage is -a125 V on float char I FREQUENCY SR 3.8.4.2 Verify no visible corrosion at battery terminals and connectors.

OR Verify connection resistance of these items is < 1.5 E-4 ohm.

SR 3.8.4.3 Verify battery cells, cell plates, and racks show no visual 1

II ILl4; l.J

+; H f11 yI..I 'Am aj LI LImmi k ciUI A + u iIV I

" ý.'C,,LIW" ! % J lPiy lC, ,Cl J al lclll*

I u ki; aI "%UIIII:I a Vl Ul

• U C l, UI "

that could degrade battery performance.

SR 3.8.4.4 Remove visible terminal corrosion, verify battery cell to cell and terminal connections are clean and tight, and are coated with anti-corrosion material.

SR 3.8.4.5 Verify battery connection resistance is < 1.5 E-4 ohm for inter-cell connections, and <1.5 E-4 ohm.for terminal connections.

SR 3.8.4.6 Verify each battery charger supplies > 400 amps at

Ž 125 V for >_1 hour.

(continued)

McGuire Units 1 and 2 3.8.4-2 Amendment Nos -_qz2Lý5r

DC Sources - Operating 3.8.4 SURVEILLANCE FREQUENCY i

SR 3.8.4.7 NOTE --------------------....--------

S------

The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4.7.

Verify battery capacity is adequate to supply, and maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.

SR 3.8.4.8 Verify battery capacity is >_80% of the manufacturer's rating when subjected to a performance discharge test or a modified performance discharge test. AND 12 months when battery shows degradation or has reached 85% of expected life with capacity < 100%

of manufacturer's rating AND 4*I' II uIILl II WIhen battery has reached 85% of the expected life with capacity >

100% of manufacturer's rating McGuire Units 1 and 2 3.8.4-3 Amendment Nos-ý

Battery Cell Parameters 3.8.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Declare associated battery Immediately associated Completion inoperable.

Time of Condition A not met.

OR One or more batteries with average electrolyte temperature of the representative cells

< 600 F.

OR One or more batteries with one or more battery cell parameters not within Category C values.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet Table 3.8.6-1 /

Category A limits.

(continued)

Ev~j- rJ.L McGuire Units 1 and 2 3.8.6-2 Amendment Nos(ý -

Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 Category B limits.

AND Once within 7 days after a battery discharge

<110V AND Once within 7 days after a battery overcharge

> 150V SR 3.8.6.3 Verify average electrolyte temperature of representative cells is _>601F.

McGuire Units 1 and 2 3.8.6-3 Amendment Nos-(ý

Inverters - Operating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Inverters- Operating LCO 3.8.7 The four required Channels of inverters shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One inverter inoperable. A.1 ---------- NOTE-------

Enter applicable Conditions and Required Actions of LCO 3.8.9, "Distribution Systems - Operating" with any vital bus de-energized.

Restore inverter to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

B. Required Action and- B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.11 Verify correct inverter voltage, and alignment to required AC vital buses.

McGuire Units 1 and 2 3.8.7-1 Amendment Nos4 -

Inverters - Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or required boron concentration.

AND A.2.4 Initiate action to restore Immediately required inverters to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct voltage and alignments to required AC vital buses.

McGuire Units 1 and 2 3.8.8-2 Amendment Nos*E --

Distribution Systems - Operating 3.8.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One channel of DC C.1 Restore DC channel of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> electrical power electrical power distribution distribution subsystem subsystem to OPERABLE AND inoperable, status.

16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from discovery of failure to meet LCO D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND D.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E. Two trains with E.1 Enter LCO 3.0.3. Immediately inoperable distribution subsystems that result in a loss of safety function.

SURVEILLANCE REQUIREMENTS. _ _ _

SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to AC, DC, and AC vital bus electrical power distribution subsystems.

McGuire Units 1 and 2 3.8.9-2 Amendment Nos.'ýy

Distribution Systems - Shutdown 3.8.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.4 Initiate actions to restore Immediately required AC, DC, and AC vital bus electrical power distribution subsystems to OPERABLE status.

AND A.2.5 Declare associated Immediately required residual heat removal subsystem(s) inoperable and not in operation.

AND A.2.6 Declare affected Low Immediately Temperature Overpressure Protection (LTOP) feature(s) inoperable.

SURVEILLANCE REQUIREMENTS "__""____.._" _

SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to required 4 AC, DC, and AC vital bus electrical power distribution subsystems.

McGuire Units I and 2 3.8.10-2 Amendment Nos * -

Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the COLR.

---

NOTE --------.-.--------- ............----------------------

Only applicable to the refueling canal and refueling cavity when connected to the RCS.

APPLICABILITY: MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not A.1 Suspend CORE Immediately within limit. ALTERATIONS.

AND A.2 Suspend positive reactivity Immediately additions.

AND A.3 Initiate action to restore Immediately boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit specified in / /rv COLR.

McGuire Units 1 and 2 3.9.1-1 Amendment Nos.-

Unborated Water Source Isolation Valves 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Unborated Water Source Isolation Valves LCO 3.9.2 Each valve used to isolate unborated water sources shall be secured in the closed position.

APPLICABILITY: MODE 6.

ACTIONS

--- ------------ IJL) I r-----------------------------------------------------------

-Separate Condition entry is allowed for each unborated water source isolation valve.

CONDITION REQUIRED ACTION COMPLETION TIME A. ---- NOTE - .--------

A.1 Suspend CORE Immediately Required Action A.3 ALTERATIONS.

must be completed whenever Condition A is AND entered.

---

A.2 Initiate actions to secure Immediately valve in closed position.

One or more valves not secured in closed AND position.

A.3 Perform SR 3.9.1.1. -4 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify each valve that isolates unborated water sources is secured in the closed position.

McGuire Units 1 and 2 3.9.2-1 Amendment N os_ -

Nuclear Instrumentation 3.9.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.9.3.1 Perform CHANNEL CHECK.

SR 3.9.3.2 ----------- \I OTE ---------------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION.

McGuire Units 1 and 2 3.9.3-2 Amendment Ne&ý

Containment Penetrations 3.9.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify each required containment penetration is in the /

required status.

SR 3.9.4.2 Perform required Containment Purge Exhaust System In accordance with Testing in accordance with theVentilation Filter Testing the VFTP Program (VFTP).

McGuire Units 1 and 2 3.9.4-2 Amendment Nos-ý

RHR and Coolant Circulation - High Water Level 3.9.5 CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.4 Close all containment 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> penetrations providing direct access from containment atmosphere to outside atmosphere.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify one RHR loop is in operation and circulating 4 reactor coolant at a flow rate of >_1000 gpm and RCS temperature is < 140 0 F.

L McGuire Units 1 and 2 3.9.5-2 Amendment No s4

RHR and Coolant Circulation - Low Water Level 3.9.6 CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Initiate action to restore Immediately one RHR loop to operation.

AND B.3 Close all containment 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> penetrations providing direct access from containment atmosphere to outside atmosphere.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of > 1000 gpm and RCS temperature is < 140 0 F.

SR 3.9.6.2 Verify correct breaker alignment and indicated power available to the required RHR pump that is not in operation.

McGuire Units 1 and 2 3.9.6-2 Amendment Nos- -

Refueling Cavity Water Level 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Refueling Cavity Water Level LCO 3.9.7 Refueling cavity water level shall be maintained _Ž 23 ft above the top of reactor vessel flange.

APPLICABILITY: During CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling cavity water A.1 Suspend CORE Immediately level not within limit. ALTERATIONS.

AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify refueling cavity water level is Ž23 ft above the top of reactor vessel flange.

McGuire Units 1 and 2 3.9.7-1 Amendment Nos.(ýýý

INSERT 2 (New TS Section 5 Program) 5.5.17 Surveillance Frequency Control Progqram This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Control Room Envelope Habitability Progqram (continued)

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to atmospheric pressure during the pressurization mode of operation by one train of the CRAVS, operating at a makeup flow rate of < 2200 cfm, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the periodic assessment of the CRE boundary in accordance with Regulatory Guide 1.197, Figure 1.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c.

The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensingbasis 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 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining. CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively..

f's6T i5-,*7 and 22 5.5-15 Amendment Units 11 and McGuire Units 5.5-15 Amendment NosQt4---q

ATTACHMENT 4 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES Application.for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program

INSERT 3 (TS SR Bases Pages)

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

(NOTE: TEXT REPLACED BY INSERT 3 WILL BE PLACED IN THE SFCP DOCUMENT PER TSTF-425)

/<,/O Cf-I4M 0A T /- PA SDM B 3.1.1 BASES ACTIONS (continued) storage tank, or the refueling water storage tank. The operator should borate with the best source available for the plant conditions.

In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may approach or exceed 2000 ppm. Using its normal makeup path, the Chemical and Volume Control System (CVCS) is capable of inserting negative reactivity at a rate of approximately 30 pcm/min when the RCS boron concentration is 1000 ppm and approximately 35 pcm/min when the RCS boron concentration is 100 ppm. Ifthe emergency boration path is used, the CVCS is capable of inserting negative reactivity at the rate of 65 pcm/min when the RCS boron concentration is 1000 ppm and 75 pcm/min when the RCS boron concentration is 100 ppm. Therefore, ifSDM had to be increased by 1%

Ak/k or 1000 pcm, normal makeup path at 1000 ppm could restore SDM in approximately 33 minutes. At 100 ppm, SDM could be restored in approximately 29 minutes. In the emergency boration mode at 1000 ppm, the 1% Ak/k could be restored in approximately 15 minutes. With RCS boron concentration at 100 ppm, SDM could be increased by 1000 pcm in approximately 13 minutes using emergency boration. These boration parameters represent typical values and are provided for the purpose of offering a specific example.

SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODES1 and 2 with keff > 1.0, SDM is verified by observing that the.

requirements of LCO 3.1.5 and LCO 3.1.6 are met.. In.the event.that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum worth.

In MODE 2 with keff < 1.0 and MODES 3, 4, and 5, SDM is verified by performing a reactivity balance calculation, considering the listed reactivity effects:

a. RCS boron concentration;

b. Control bank position;

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation;

e. Xenon concentration; McGuire Units 1 and 2 B 3.1.1-5 Revision N0o6

SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

iouqais býAed orjtiKe gen~i REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2.

UFSAR, Section 15.1.5.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

4. UFSAR, Section 15.4.6.

5. 10 CFR 100.

McGuire Units 1 and 2 B 3.1.1-6 Revision NO-05 I

Core Reactivity B 3.1.2 BASES ACTIONS (continued)

B.1 If the core reactivity cannot be restored to within the 1% Ak/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the SDM for MODE 3 is not met, then the boration required by SR 3.1.1.1 would occur. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable, including control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization of predicted core reactivity to the measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents A15S-T-operation for a large fraction of the fuel cycle without establi.shing a benchmark for the desigqcajn ..... atr lI

.of c nge uet e.de aesen tion dthe of en ica s(Q ,A , etc rpro t inctona REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.

2. UFSAR, Chapter 15.

3. 10 CFR 50.36, Technical Specification, (c)(2)(ii).

McGuire Units I and 2 B 3.1.2-5 Revision No.O

Rod Group Alignment Limits B 3.1.4 BASES SURVEILLANCE SR 3.1.4.1 REQUIREMENTS JVe tYiiuLo piin rihna~nme imilts t a*

r-eq crY o ho prov s a his tha ws oper r to Ld ct a ;ddthat fýbegi ng to d iatefr its e ected ition If t position deviation monitor is inoperable, a Frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> accri~~h hemt~ oal.h s cite equenytaug t ccnher"" dpin for ont iscoava' le to' ere n tconro sot duri rod.o via n c im iatel e de cted.

SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES I and 2, tripping each control rod would result in radial or axial w or oscillations. Exercising each individual control rod e9 provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 10 steps will not cause radial or axial power tilts, or oscillations, to and aligned, the control rod(s) is considered to be OPERABLE. At any time, ifcontrol a rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken. This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism. During performance of the Control Rod Movement periodic test, there have been some "Control.

Malfunctions" that prohibited a control rod bank or group from moving when selected, as evidenced by the demand counters and DRPI. In all cases, when the control malfunctions were corrected, the rods moved freely (no excessive friction or mechanical interference) and were trippable.

SR 3.1.4.3 Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Since a removal of the reactor vessel head has the potential to change component alignments affecting McGuire Units I and 2 B 3.1.4-8 Revision Nov.

Shutdown Bank Insertion Limits B 3.1.5 BASES ACTIONS (continued)

B. 1 If the shutdown banks cannot be restored to within their insertion limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the unit must be brought to a MODE where the LCO is not applicable. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the

.reactor, and the required SDM will be maintained following a reactor trip.

This SR and Frequency ensure that the shutdown banks are withdrawn S-/9 before the control banks are withdrawn during a unit startup.

Since es tdo ban are ositi ed m allyb hec trolro m op ator ?v1e catti "of s tdo bank sition a Fr uenc rs, ert reac ist encri a1, is qua o enre that y wit' the' inse .n Iim Als5'the 1 our F que takes ' o acontrnt ot herinf atusatio availa

,shut vn r e in t cntr room r the pose of REFERENCESt 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, and GDC 28.

2. 10 CFR 50.46..

3. UFSAR, Section 15.4.

4. 10 CFR 50.36, Technical Specification, (c)(2)(ii).

McGuire Units 1 and 2 B 3.1.5-4 Revision No-,Qý

Control Bank Insertion Limits B 3.1.6 BASES ACTIONS (continued)

C. 1 If Required Actions A.1 and A.2, or B.1 and B.2 cannot be completed within the associated Completion Times, the plant must be brought to MODE 3, where the LCO is not applicable. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits.

The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration. If the ECP was calculated long before criticality, xenon concentration could change to make the ECP substantially in error. Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Verifying the ECP calculation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to criticality avoids a large error from changes in xenon concentration, but allows the operator some flexibility to schedule the ECP calculation with other startup activities.

SR 3.1.6.2 con -bank insel on limits at aeq cy of hours-is .fficient to S sure O;PERAI6LITY control býsta a of th aank prahn insertoeisrinlfssn

'imit monito nd to detect

ýnl',ion theminsert limit monitor becomes inopera e, yen ication o e control bank position at a Frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is sufficient to detect control banks that may be approaching the insertion limits.

SR 3.1.6.3 When control banks are maintained within their insertion limits as checked by SR 3.1.6.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the Mcuir.ency 1.hours of isadstent wiBh3. Rision

"*~~~~ch.eck abc~e in SR 3/t*.2. -

McGuire Units I and 2 B 3.1.6-5 Revision Nq__e--

PHYSICS TESTS Exceptions B 3.1.8 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.1.8.2 Verification that the RCS lowest loop Ta,, is _541OF will ensure that the unit is not operating in a condition that could invalidate the safety 33fii e- urin/8 e r3b** cz,*.qe oshre

~h-nitiaonci SR 3.1.8.3 Verification that THERMAL POWER is *< 5% RTP will ensure that the plant is not operating in the condition that could invalidate the safety anas iaino T RML P 1[

E F r e~un-c un1the 1Fh erf o1an of Or6IPH *SCS TSTS ill e ~ere th he SR 3.1.8.4 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:

a. RCS boron concentration;

b. Control bank position;

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation;

e. Xenon concentration;.

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

2Th

,e eU 1and B 3.1.8y slo ange.isio McGuire Units I and 2 B 3.1.8-5 Revision Nol*(*

FQ(X,Y,Z)

B 3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued) verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ was last measured.

SR 3.2.1.1 Verification that FMQ(X,Y,Z) is within.its specified steady state limits involves either increasing FMQ(X,Y,Z) to allow for manufacturing tolerance, K(BU), and measurement uncertainties for the case where these factors are not included in the FQ limit. For the case where these factors are included, a direct comparison of FMQ(X,Y,Z) to the Fa limit can be performed. Specifically, FMo(X,Y,Z) is the measured value of FQ(X,Y,Z) obtained from incore flux map results. Values for the manufacturing tolerance, K(BU), and measurement uncertainty are specified in the COLR.

The limit with which FMo(X,Y,Z) is compared varies inversely with power above 50% RTP and directly with functions called K(Z) and K(BU) provided in the COLR.

If THERMAL POWER has been increased by > 10% RTP since the last determination of FMQ(X,Y,Z), another evaluation of this factor is required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions at this higher power level (to ensure that FMQ(X,Y,Z) values have decreased sufficiently with power increase to stay within the LCO limits).

The-'* of *FI eu-r.t rtec jge of. ;er .

1v c z- u I* hsde hnic/l 3 Sp' cific. ns SR 3.2.1.2 and 3.2.1.3 The nuclear design process includes calculations performed to determine that the core can be operated within the FQ(X,Y,Z) limits.

Because flux maps are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation. The maximum peaking factor increase over steady state values, is determined by a maneuvering analysis (Ref.

5).

McGuire Units I and 2 B 3.2.1-9 Revision Noos

AFQ(X,Y,Z) A/c ~'4 44LB

'/I~d2~ON 3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)

The limit with which FMo(X,Y,Z) is compared varies and is provided in the COLR. No additional uncertainties are applied to the measured FQ(X,Y,Z) because the limits already include uncertainties.

FLQ(X,Y,Z)OP and FL0 (X,Y,Z)RPS limits are not applicable for the following axial core regions, measured in percent of core height:

a. Lower core region, from 0 to 15% inclusive; and

b. Upper core region, from 85 to 100% inclusive.

The top and bottom 15% of the core are excluded from the evaluation because of the low probability that these regions would be more limiting in the safety analyses and because of the difficulty of making a precise measurement in these regions.

This Surveillance has been modified by a Note that may require that more frequent surveillances be performed. If FMa(X,Y,Z) is evaluated and found to be within the applicable transient limit, an evaluation is required to account for any increase to FMa(X,Y,Z) that may occur and cause the FQ(X,Y,Z) limit to be exceeded before the next required FQ(X,Y,Z) evaluation.

In addition to ensuring via surveillance that the heat flux hot channel factor is within its limits when a measurement is taken, there are also requirements to extrapolate trends in both the measured hot channel factor and in its operational and RPS limits. Two extrapolations are performed for each of these two limits:

1. The first extrapolation determines whether the measured. heat flux hot channel factor is. likely to exceed its limit prior to the next performance of the SR.

2. The second extrapolation determines whether, prior to the next performance of the SR, the ratio of the measured heat flux hot channel factor to the limit is likely to decrease below the value of that ratio when the measurement was taken.

Each of these extrapolations is applied separately to each of the operational and RPS heat flux hot channel factor limits. If both of the extrapolations for a given limit are unfavorable, i.e., if the extrapolated factor is expected to exceed the extrapolated limit and the extrapolated factor is expected to become a larger fraction of the extrapolated limit McGuire Units 1 and 2 B 3.2.1-10 Revision Nq4Z

FQ(X,Y,Z)

B 3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued) than the measured factor is of the current limit, additional actions must be taken. These actions are to meet the FQ(X,Y,Z) limit with the last FMQ(X,Y,Z) increased by the appropriate factor specified in the COLR or to evaluate FQ(X,Y,Z) prior to the projected point in time when the extrapolated values are expected to exceed the extrapolated limits.

These alternative requirements attempt to prevent FQ(X,Y,Z) from exceeding its limit for any significant period of time without detection using the best available data. FMa(X,Y,Z) is not required to be extrapolated for the initial flux map taken after reaching equilibrium conditions since the initial flux map establishes the baseline measurement for future trending. Also, extrapolation of FMQ(X,Y,Z) limits are not valid for core locations that were previously rodded, or for core locations that were previously within +/-2% of the core height about the demand position of the rod tip.

F0 (X,Y,Z) is verified at power levels _>10% RTP above the THERMAL POWER of its last verification, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions to ensure that FQ(X,Y,Z) is within its limit at higher power levels.

REFERENCES 1.: 10 CFR 50.46.

2. UFSAR Section 15.4.8.

3. 10 CFR 50, Appendix A, GDC 26.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

5. DPC-NE-201 1 PA "Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors".

McGuire Units 1 and 2 B 3.2.1-11 Revision N)oo"

(FAH(X,Y))

CI/O :CA/Arv olY 1T/i

>,A 6-XE_ B 3.2.2 BASES ACTIONS (continued) limit. This Action demonstrates that the FAH(X,Y) limit is within the LCO limits prior to exceeding 50% RTP, again prior to exceeding 75% RTP, and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is Ž_95% RTP.

This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action.

B.1 When Required Actions A.1 through A.4 cannot be completed within their required Completion Times, the plant must be placed in a mode in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 and SR 3.2.2.2 are modified by a Note. The Note applies REQUIREMENTS during the first power ascension after a refueling. Itstates that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution map can be obtained. This allowance is modified, however, by one of the Frequency conditions that requires verification that FMAH(X,Y) is within the specified limits after a power rise of more than 10% RTP over the THERMAL POWER at which it was last verified to be within specified limits. Because FMAH(X,Y) could not have previously been measured:in.this reload core, power may be increased to RTP prior to an equilibrium verification of FAH(X,Y) provided nonequilibrium measurements of FAH(X,Y) are performed at various power levels during startup physics testing This ensures that'some determination of FAH(X,Y) is made at a lower power level at which adequate margin is available before going to 100% RTP. The Frequency condition is not intended to require verification of the parameter after every 10% increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FAH(X,Y) was last measured.

SR 3.2.2.1 The value of FMAH(X,Y) is determined by using the movable incore detector system to obtain a flux distribution map at any THERMAL POWER greater than 5% RTP. A computer program is used to process McGuire Units 1 and 2 B 3.2.2-7 Revision Nq;K

(FAH(X,Y))

B 3.2.2 BASES SURVEILLANCE REQUIREMENTS (continued) the measured 3-D power distribution to calculate the steady state FLAH(X,Y)LCO limit which is compared against FMAH(X,Y).

FMAH(X,Y) is verified at power levels > 10% RTP above the THERMAL POWER of its last verification, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions to ensure that FMAH(X,Y) is within its limit at high power levels.

Th EFPrequeo* is acc able be,cuse the p,*er distribjý, on

,--'ýth2isequenc eq sstheortod g ii a oheFH t be eýeded any si 'icant pe od of op tion.

SR 3.2.2.2 The nuclear design process includes calculations performed to determine that the core can be operated within the FAH(X,Y) limits. Because flux maps are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation. The maximum peaking factor increase over steady state values is a limit called FLAH (X,Y)SuRV. This Surveillance compares the measured FMAH(X,Y) to the Surveillance limit to ensure that safety analysis limits are maintained.

This Surveillance. has been modified by a Note that may require that more.

frequent surveillances be performed. If FMAH(XY) is evaluated and found to be within its surveillance limit, an evaluation is required to account for any increase to FMAH(XY) that may occur and- cause the FAH(X,Y)SURV limit to be exceeded before the next required FtH(XY)SRv evaluation.

In addition to ensuring via surveillance that the enthalpy rise hot channel factor is within its steady state and surveillance limits when a measurement is taken, there are also requirements to extrapolate trends in both the measured hot channel factor and in its surveillance limit. Two extrapolations are performed for this limit:

1. The first extrapolation determines whether the measured enthalpy rise hot channel factor is likely to exceed its surveillance limit prior to the next performance of the SR.

2. The second extrapolation determines whether, prior to the next performance of the SR, the ratio of the measured enthalpy rise hot McGuire Units 1 and 2 B 3.2.2-8 Revision No.-

(FAH(XY))

B 3.2.2 BASES SURVEILLANCE REQUIREMENTS (continued) channel factor to the surveillance limit is likely to decrease below the value of that ratio when the measurement was taken.

Each of these extrapolations is applied separately to the enthalpy rise hot channel factor surveillance limit. If both of the extrapolations are unfavorable, i.e., ifthe extrapolated factor is expected to exceed the extrapolated limit and the extrapolated factor is expected to become a larger fraction of the extrapolated limit than the measured factor is of the current limit, additional actions must be taken. These actions are to meet the FMAH(X,Y) limit with the last FMAH(X,Y) increased by the appropriate factor specified in the COLR, or to evaluate FM6H(X,Y) prior to the point in time when the extrapolated values are expected to exceed the extrapolated limits. These alternative requirements attempt to prevent FMAH(X,Y) from exceeding its limit for any significant period of time without detection using the best available data. FMAH(X,Y) is not required to be extrapolated for the initial flux map taken after reaching equilibrium conditions since the initial flux map establishes the baseline measurement for future trending.

FMAH(X,Y) is verified at power levels 10% RTP above the THERMAL POWER of its last verification, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions to ensure that FMAH(X,Y) is within its limit at high power levels.

The S equueillan y of 3,FPD adeq to moý r the, c ge of r di utio lith 0o. rnu he Su p;illancep y be done re fre ently i quire y the ults of FMH(X,Y) uations.

-he F uenc 31 E is a quate to onitor th hange o ower di ibutio ecaus uch a ange is s iciently w, when e plan opera in acc ance the TS preclu adverse aking ftors be een 31 y surveil ces.

REFERENCES 1. UFSAR Section 15.4.8

2. 10 CFR 50, Appendix A, GDC 26.

3. 10 CFR 50.46.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

5. DPC-NE-2005P "Duke Power Company Thermal Hydraulic Statistical Core Design Methodology", September 1992.

6. DPC-NE-2004P-A, Rev. 1, "Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01," SER Dated February 20, 1997 (DPC Proprietary)

McGuire Units 1 and 2 B 3.2.2-9 Revision No-O

AFD

//0 OCiA1\1,1- ~f/-7 PA6L B 3.2.3 BASES LCO (continued)

Signals are available to the operator from the Nuclear Instrumentation System (NIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors. The AFD is defined as the difference in normalized flux signals between the top and bottom excore detectors in each detector well. For convenience, this flux difference is converted to provide flux difference units expressed as a percentage and labeled as %A flux or %AI.

The AFD limits are provided in the COLR. The AFD limits do not depend on the target flux difference. However, the target flux difference may be used to' minimize changes in the axial power distribution.

Violating this LCO on the AFD could produce unacceptable consequences if a Condition 2, 3, or 4 event occurs while the AFD is outside its specified limits.

APPLICABILITY The AFD requirements are applicable in MODE 1 greater than or equal to 50% RTP when the combination of THERMAL POWER and core peaking factors are of primary importance in safety analysis.

For AFD limits developed using maneuvering analysis methodology, the value of the AFD does not affect the limiting accident consequences with THERMAL POWER < 50% RTP and for lower operating power MODES.

ACTIONS, A._1 As an alternative to restoring the AFD to within its specified limits, Required Action A. 1 requires a THERMAL POWER reduction to

< 50% RTP. This places the core in a condition for which the value of the AFD is not important in the applicable safety analyses. A Completion Time of 30 minutes is reasonable, based on operating experience, to reach 50% RTP without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS The AFD is monitored on an automatic basis using the unit process computer, which has an AFD monitor alarm. The computer determines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately ifthe AFD for two or more OPERABLE excore channels is outside its specified limits.

Revision No McGuire Units McGuire and 2 Units I1 and 2 B 3.2.3-3 B 3.2.3-3 Revision No(?)

AFD B 3.2.3 BASES SURVEILLANCE REQUIREMENTS (continued)

This Surveillance verifies that the AFD, as indicated by the NIS excore channel, is within its specified limits and is consistent with the status of the AFD monitor alarm. With the AFD monitor alarm inoperable, the AFD is monitored every hour to detect operation outside its limit. The Frequency of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is based on operating experience regarding the amount of time required to vary the AFD, and the fact that the AFD is ceoseymonitored. With the AFD monit*or-alarm OPERA.LE_ te--- _

Fsisdecithat ýAFDJs REFERENCES 1. DPC-NE-201 1 PA, "Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors".

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

3. UFSAR, Chapter 7.

Revision No 4 MoGuire Units McGuire and 22 Units 1I and B 3.2.3-4 B 3.2.3-4 Revision No

QPTR B 3.2.4 BASES ACTIONS (continued) reaching RTP. As an added precaution, if the core power does not reach RTP within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but is increased slowly, then the peaking factor surveillances must be performed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the time when the more restrictive of the power level limit determined by Required Action A.1 orA.2 is exceeded. These Completion Times are intended to allow adequate time to increase THERMAL POWER to above the more restrictive limit of Required Action A.1 or A.2, while not permitting the core to remain with unconfirmed power distributions for extended periods of time.

Required Action A.7 is modified by a Note that states that the peaking factor surveillances must be done after the excore detectors have been calibrated to show zero tilt (i.e., Required Action A.6). The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors are calibrated to show zero tilt and the core returned to power.

B.1 If Required Actions A.1 through A.7 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to

• 50% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience regarding-the amount of time required to reach the reduced power level without challenging plant systems.

SURVEILLANCE SR 3.2.4.1 REQUIREMENTS..

SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is <

75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1 if more than one input from Power Range Neutron Flux channels are inoperable.

This Surveillance verifies that the QPTR, as indicated by the Nuclear Ins~trume ation System (NIS) excore channels, is within its limits.

Mc"E U 1the 3 QPTR alarm isB s Ne McGuire Units 1 and 2 B 3.2.4-5 Revision No(c

QPTR B 3.2.4 BASES SURVEILLANCE REQUIREMENTS (continued)

When the QPTR alarm is inoperable, the Frequency is increased to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This Frequency is adequate to detect any relatively slow changes in QPTR, because for those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.

SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is required only when the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is >_75% RTP.

With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts

• i For purposes of monitoring the QPTR when one power range channel is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is performed with a full incore flux map or two sets of four

• thimble locations a set of core withis quarter eight symmetry. The two Unique detector sets of four locations, These Symmetric thimbles

" locations are C-8, E-5, E-1 1, H-3, H-13, L-5, L-1e1, and N-8.

• The symmetric thimble flux map can be used to generate symmetric thimble tilt." This canbe compared to a reference symmetric thimbletilt, from the most recent full core flux map, to generate an incore tilt.

Therefore, incore tilt can be used to confirm that QPTR is within limits.

With one ormmore NIS channel inputs to QPTR inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. To confirm that no change in tilt has actually occurred, which might cause the QPTR limit to be exceeded, the incore result may be compared against previous flux maps either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent flux map data.

McGuire Units 1 and 2 B 3.2.4-6 Revision Nliis

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)

Channel Ill, and Channel IV (if applicable). The CHANNEL CALIBRATION and COTs are performed in a manner that.is consistent with the assumptions used in analytically calculating the required channel accuracies.

Performing the Neutron Flux Instrumentation surveillances meets the License Renewal Commitments for License Renewal Program for Neutron Flux Instrumentation Circuits per UFSAR Chapter 18, Table 18-1 and License Renewal Commitments Specification MCS-1274.00-00-0016, Section 4.44.

SR 3.3.1.1 Performance of the CHANNEL CHECK e ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff based on a combination of the channelinstrument uncertainties, including indication aýnd readability, If a channel is outside. the criteria, it may be an indication tha ensor or the signal processing equipment has drifted outside its limit.

SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output . If the calorimetric exceeds the NIS channel output by > 2% RTP, the NIS is not declared inoperable, but must be adjusted. If the NIS channel output cannot be properly adjusted, the channel is declared inoperable.

Two Notes modify SR 3.3.1.2. The first Note indicates that the NIS channel output shall be adjusted consistent with the calorimetric results if the absolute difference between the NIS channel output and the calorimetric is > 2% RTP. The second Note clarifies that this Surveillance McGuire Units 1 and 2 B 3.3.1-41 Revision No(;6

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) is required only if reactor power is -> 15% RTP and that 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed for completing the first Surveillance after reaching 15% RTP. At lower power levels, calorimetric data are inaccurate.

i!Xequateit is SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel outputý-D 4&?AD. If the absolute difference in AFD is> 3%, the NIS channel is still OPERABLE, but must be readjusted.

If the NIS channel cannot be properly readjusted, the channel is declared inoperable. This Surveillance is performed to verify the f(AI) input to the overtemperature AT Function and overpower AT Function.

Two Notesmodify SR 3.3.1.3. Note 1 indicates that the excore NIS channel shall be adjusted if the absolute difference between the inCore and excore AFD is._> 3%. Note 2 clarifies that the Surveillance is required only if reactor power is > 15% RTP and that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed for completing the first Surveillance after reaching 15% RTP.

is adeauate. It i,ý SR 3.3.1.4 SR 3.3.1.4 is th ce of a TADOTde ý ý BAS This test shall verify OPERABILITY by actuation of the end devices.

The RTB test shall include separate verification of the undervoltage and shunt trip mechanisms. Independent verification of RTB undervoltage McGuire Units 1 and 2 B 3.3.1-42 Revision Noo

RTS Instrumentation

.B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) and shunt trip Function is not required for the bypass breakers. No capability is provided for performing such a test at power. The independent test for bypass breakers is included in SR 3.3.1.14. The bypass breaker test shall include a local shunt trip. A Note has been added to indicate that this test must be performed on the bypass breaker prior to placing it in service.

SR 3.3.1.6 V SR 3.3.1.6 is a calibration of the excore channels to the incore channels.

If the measurementsdo not agree, the excore channelsare not declared

• inoperable but must beexcorecalibrated-to agree with the incore detector measurementsb If the channels cannot be adjusted, the channels are declared inoperable. This Surveillance is performed to verify the f(AI) input to the overtemperaturel ATFunction and overpower AT Function..

At Beginning of Cycle (BOC), the excore channels are compared to the incore detector measurements. This comparison is typically performed prior to exceeding 75% power. Excore detectors are adjustederformance as necessary. This low power surveillance satisfies the initial At BOC, after reaching full power steady state conditions, additional incore and excore measurements are taken at various Al conditions to determine the Mfactors. The M factors are normally only determined at BOC, but they may priobe 5%tochanged ower xceeingExcre points in the at otherdtecors fuel cycle aif the re ajustd McGuire Units 1 and 2 B 3.3.1-43 Revisionrmed

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) relationship between excore and incore measurements changes significantly.

A Note modifies SR 3.3.1.6. The Note states that this Surveillance is required only if reactor power is > 75% RTP and that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed for completing the first surveillance after reaching 75% RTP.

(Th e ncy 2 D ' de ate. isb~ do ustry ýratiin~g-pe nce onsi rin strint ~il tdo =ary f instr ent ift.

SR 3.3.1.7 SR 3.3.1.7 is the performance of a COT4 A COT is performed on each required channel to ensure the channel will perform the intended Function.

The tested portion of the Loop must trip within the Allowable Values specified in Table 3.3.1-1.

The setpoint shall be left set consistent with the assumptions of the 3 setpoint methodology.

SR 3.3.1.7 is modified by a Note that provides a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> delay in the requirement to perform this Surveillance for source range instrumentation when entering MODE 3 from MODE 2. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for a short time in MODE 3 until the RTBs'are open and SR 3.3.1.7 is no longer required to be performed., If the unit is to be in MODE 3 with the.

RTBs closed for > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> this Surveillance must be completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entry into MODE 3. The surveillance shall include verification of the high flux at shutdown alarm setpoint of less than or equal to the average CPS Neutron Level reading (most consistent value between highest and lowest CPS Neutron Level reading) at five times background.

S R 3.3.1.8 SR 3.3.1.8 is the performance of a COT as described in SR 3.3.1.7, except it is modified by a Note that this test shall include verification that the P-6, during the Intermediate Range COT, and P-10, during the Power Range COT, interlocks are in their required state for the existing unit condition. The verification is performed by visual observation of the McGuire Units 1 and 2 B 3.3.1-44 Revision No.06

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS continued) permissive status light i he unit control room. The Frequency is S-modified by a Note that a s this surveillance to be satisfied if it has

/- /A been performed within of the Frequencies prior to reactor startup and four hours a er reducing power below P-10 and P-6. The I- "c Frequency of "prior to startup" ensures this surveillance is performed prior 4~~

-/ p ,_,A to critical operations and applies to the source, intermediate and power Rcs'4 /..t4.,,-. j/ range low instrument channels. The Frequency of "4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reducing power below P-10" (applicable to intermediate and power range low channels) and "4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reducing power below P-6" (applicable to source range channels) allows a normal shutdown to be completed and the unit removed from the MODE of Applicability for this surveillance without a de to erform the testing required by this surveillance. The Frequencyncyýhereafter applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup and four hours after reducing power below P-10 or P-6. The MODE of Applicability for this surveillance is < P-10 for the power range low and intermediate range channels and < P-6 for the source range channels. Once the unit is in MODE 3, this surveillance is no longer required. If power is to be maintained < P-1 0 or < P-6 for more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, then the testing required by this surveillance must be performed prior to the expiration of the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limit. Four hours is a reasonable time to complete the required testing or place the unit in a MODE where this surveillance is no longer required. This test ensures that the NIS source, intermediate, and power range low channels are OPERABLE prior to taking the reactor critical and after reducing power in othe applicable The SR is modified by a Note that K~',. (

the TADOT. Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification is accomplished during the CHANNEL CALIBRATION.

SR 3.3.1.10 Was t,*RAT}N iv-erf7---nec-vepl--8 The CHANNEL may be performed at power or during refueling based on testing capability. Channel unavailability evaluations in McGuire Units 1 and 2 B 3.3.1-45 Revision NC

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)

References 10 and 11 have conservatively assumed that the CHANNEL CALIBRAITON is performed at power with the channel in bypass.

CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint methodology.

verification that 5Ithe time,*constants 7-r" aeadJsSad are adjusted to°he the prescribed re ibned*//*

values where applicable. The applicable time constants are shown in Table SR 3.3.1.11 SR 3.3.1.11 is the performance of a CHANNEL CALIBRATION, as described 3.0 in SR Two notes modify this SR.

Notei states that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the power range neutron detectors consists of a normalization of the detectors based on a power calorimetric and flux map performed above 15% RTP. The CHANNEL CALIBRATION for the source range neutron detectors consists of two methods. Method 1 consists of obtaining the discriminator curves for source range, evaluating those curves, and comparing the curves to the manufacturer's data (adjustments to the discriminator voltage are performed as required). Method 2 consists of performing waveform analysis. This analysis process monitors the actual number and amplitude of the Neutron/Gamma pulses being generated by the SR detector. The high voltage is adjusted to optimize the amplitude of the pulses while maintaining as low as possible high voltage value in order to prolong the detector life. The discriminator voltage is then adjusted, as required, to reasonably ensure that the neutron pulses are being counted by the source range instrumentation and the unwanted gamma pulses are not being counted as neutron pulses.

The CHANNEL CALIBRATION for the intermediate range neutron detectors consists of the high voltage detector plateau for intermediate range, evaluating those curves, and comparing the curves to the manufacturer's data. Note 2 states that this Surveillance is not required McGuire Units 1 and 2 B 3.3.1-46 Revision N96

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) for the NIS power range detectors for entry into MODE 2 or 1, and is not required for the NIS intermediate range detectors for entry into MODE 2, because the unit must be in at least MODE 2 to perform the test for the intermediate range detectors and MODE 1 for the power range detectors.

SR 3.3.1.12 SR 3.3.1.12 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.10 e. Calibration of the AT channels is required at the beginning of each cycle upon completion of the precision heat balance. RCS loop AT values shall be determined by precision heat balance measurements at the beginning of each cycle.

The eqýue Is ied b e ss an 1~rot 6raio erva te ermi on of;Pmag;ude of;-G__ ri~ft irjýl1 SR 3.3.1.14 SR 3.3.1.14 is the performance of a TADOT of the Manual Reactor Trip ade SI Input from ESFAS.

0 The test shall independently verify the OPERABILITY of the undervoltage and shunt trip mechanisms for the Manual Reactor Trip Function forthe Reactor Trip Breakers and Reactor Trip Bypass Breakers. The Reactor Trip Bypass include testing of the automatic undervoltage trip. ('0 F,6 o, f+_)

McGuire Units 1 and 2 B 3.3.1-47 Revision No(%

RTS Instrumentation BASE * (-Y /** E_*'*B 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

The SR is modified by a Note that excludes verification of setpoints from the TADOT. The Functions affected have no setpoints associated with them.

SR 3.3.1.15 SR 3.3.1.15 is the performance of a TADOT of Turbine Trip Functions.

This TADOT is as described in SR 3.3.1.4,.except that this test is performed prior to reactor startup. A Note states that this Surveillance is not required if it has been performed within the previous 31 days.

Verification of the Trip Setpoint does not have to be performed for this Surveillance. Performance of this test will ensure that the turbine trip Function is OPERABLE prior to taking the reactor critical. This test cannot be performed with the reactor at power and must therefore be performed prior to reactor startup.

SR 3.3.1.16 and SR 3.3.1.17 SR 3.3.1.16 and SR 3.3.1.17 verify that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analysis. Response time testing acceptance criteria are included in the UFSAR (Ref. 1). Individual component response times are not modeled in the analyses.

The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the equipment reaches the required functional state (i.e.,

control and shutdown rods. fully inserted in the reactor core).

For channels that include dynamic transfer Functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer Function set to one, with the resulting measured response time compared to the appropriate UFSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value, provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.

Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from:

McGuire Units 1 and 2 B 3.3.1-48 Revision No.-

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)

(1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) in place, onsite, or offsite (e.g.,

vendor) test measueements, or (3) utilizing vendor engineering '

specifications. WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements" provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be either demonstrated by test, or their equivalency to those listed in WCAP-13632-P-A, Revision 2. Any demonstration of equivalency must have been determined to be acceptable by NRC staff review.

WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests' provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time.

The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.

c;qceptab46 from a reliability stap~lpoint."

SR 3.3.1.16 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. The response time of the neutron flux signal portion of the channel shall be measured from detector output or input of the first electronic component in the channel.

McGuire Units 1 and 2 B 3.3.1-49 Revision N<)ý)

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)

Similarly, train A and train B must be examined when testing channel II, channel Ill, and channel IV (if applicable). The CHANNEL CALIBRATION and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accuracies.

SR 3.3.2.1 Performance of the CHANNEL CHECK(nsures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and reliability. If a sschannel is outsidhe criteria, it may be an indication that the sensor or the sin eqimnuasditdousdstimit. h

• _. ' ... SR 3.3.2.2 is the performance of an ACTUATION LOGIC TEST *~7~

/**,./ --i-j F- semiautomatic tester. The train being tested is placed in the bypass condition, T "-pncy*thus preventing inadvertent actuation. Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function. In addition, the master relay coil is pulse tested for continuity. This verifies that the logic modules are OPERABLE and Mc~~~~~uire~yps ..- 9Rvsonditon,~

Untc n SR 3.3.2.3

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.2.3 is the performance of a COT on the RWST level and Containment Pressure Control Start and Terminate Permissives.

A COT is performed on each required channel to ensure the entire channel will perform the intended Function. Setpoints must be found within the Allowable Values specified in Table 3.3. 2-1.lhigist is p orm.ad every S R 3.3.2.4 SR 3.3.2.4 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage fý is insufficient to pick u the slave relay, but large enough to dem nstrate signal path continuity.

r >The time allowed for the testin (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) is justified in Reference 7.

SR 3.3.2.5

"* SR 3.3.2.5 is the performance of a COT.

A COT is performed on each. required channel to ensure the channel wi.ll perform the intended Function. The tested portion of the loop must trip within the Allowable Values specified in Table 3.3. 2-1.

The setpoint shall be left set consistent with the assumptions'of the setpoint methodology.

SR 3.3.2.6 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function, or is placed in a condition where the relay contact operation can be verified without operation of the equipment.

Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing McGuire Units 1 and 2 B 3.3.2-40 Revision No. 99

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.2.7 isthe performance of a TADOT. This test is a check of the Manua Actuation Functions, AFW pump start, Reactor Trip (P-4) Interlock and Do ue ae Level - High High feedwater isolation. <hjviis ~

ý Each Manua] Actuation Function is tested up o, an including, the master relay coils. In some instances, the test includes actuation of the end dLevice (La.,_ýump starts. valve cycles, etc. . e qu yIt3 q e, t%" 0 0 i xp ce-and ' on ' ent fu n e. he SR is modified by a Note that excludes verification of setpoints during the TADOT for manual initiation Functions. he manual initiation Functions have no associated setpoints.

Of S R 3.3.2.8 SR 3.3.2.8 is the performance of a CHANNEL CALIBRATION.

TI s e d I The CHANNEL CALIBRATION may be performed at power or during refueling based on bypass testing capability. Channel unavailability evaluations in References 10 and 11 have conservatively assumed that the CHANNEL CALIBRATION is performed at power with the channel in bypass.

CHANNEL CALIBRATION is a complete check. of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.

th Feqrcyo ont is ba~d o 4supiIof an~ot intmýethaiin t deter nt6 ation he ma ritude of m rift in

,eimbr

\..heeto lo This SR is modified by a Note stating that this test should include verification that the time constants are adjusted to the prescribed values where applicable.

The applicable time constants are shown in Table 3.3.2-1.

SR 3.3.2.9 McGuire Units 1 and 2 B 3.3.2-41 Revision N<0

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.

ESF RPONCE TIMEe ts arenducte-don an 18-fnonth ý5rTAGGERF) not cý&a~nel failde, are infreduent occurreeinces.

This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 900 psig in the SGs.

REFERENCES 1. UFSAR, Chapter 6.

2. UFSAR, Chapter 7.

3. UFSAR, Chapter 15.

4. IEEE-279-1971.

5. 10 CFR 50.49.

6. 10 CFR'50.36, Technical Specifications, (c)(2)(ii).

7. WCAP-10271-P-A, Supplement 1 and Supplement 2, Rev. 1, May 1986 and June 1990.

8. WCAP 13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements" Sep., 1995.

9. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests" Oct., 1998.

10. WCAP-14333-P-A, Revision 1, October 1998.

11. WCAP-15376-P-A, Revision 1, March 2003.

McGuire Units 1 and 2 B 3.3.2-43 Revision Ndo

PAM Instrumentation B 3.3.3 BASES ACTIONS (continued)

H.1 Alternate means of monitoring Containment Area Radiation have been developed and tested. These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. If these alternate means are used, the Required Action is not to shut down the unit but rather to follow the directions of Specification 5.6.7, in the Administrative Controls section of the TS. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.

SURVEILLANCE A Note has been added to the SR Table to clarify that REQUIREMENTS SR 3.3.3.1 and SR 3.3.3.3 apply to each PAM instrumentation Function in Table 3.3.3-1.

Performing the Neutron Flux Instrumentation and Containment Atmosphere Radiation (High-Range) surveillances meets the License Renewal Commitments for License Renewal Program for High-Range Radiation and Neutron Flux Instrumentation Circuits per UFSAR Chapter 18, Table 18-1 and License Renewal Commitments Specification MCS-1274.00-00-0016, Section 4.44.

SR 3.3.3.1 Performance of the CHANNEL.CHECK K ensures that a gross instrumentation failure has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar unit instruments located throughout the unit.

McGuire Units 1 and 2 B 3.3.3-14 Revision No."

PAM Instrumentation B 3.3.3 BASES SURVEILLANCE REQUIREMENTS (continued)

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE.

As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.

SR 3.3.3.2 Not Used SR 3.3.3.3 CA FFAe A L LI T ,is,ýP ordevff1,-o~f, efu HANýNEL CALIBRAiTIONis a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the i I

necessary range andaccuracy. This SR is modified by a Note that excludes neutron detectors. The calibration method for neutron detectors ii is specified in the Bases of LCO 3.3.1, "Reactor Trip System (RTS_ [

.i

[.

Instrumentation." ed

'-Fre-uen-Js oD tinýexieriecee apN oi*

REFERENCES 1. UFSAR Section 1.8.

2. Regulatory Guide 1.97, Rev. 2.

3. NUREG-0737, Supplement 1, "TMI Action Items."

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.3.3-15 Revision No-0

Remote Shutdown System B 3.3.4 BASES SURVEILLANCE SR 3.3.4.1 REQUIREMENTS Performance of the CHANNEL CHEC ensures that a gross failure of instrumentation has not occurre . A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

As specified in the Surveillance, a CHANNEL CHECK is only required for those channels which are normally energized.

SR 3.3.4.2 SR 3.3.4.2 verifies each required Remote Shutdown System control circuit and transfer switch performs the intended function. This verification is performed from the remote shutdown panel and locally, as appropriate. Operation of the equipment from the remote shutdown panel is not necessary. The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the unit can be placed and maintained in MO om the remote shutdown panel and the locatcontrol station,!The,,18ymonth McGuire Units 1 and 2 B 3.3.4-4 Revision N1.6

Remote Shutdown System B 3.3.4 BASES SURVEILLANCE REQUIREMENTS (continued) duorr at tou* m rai Fre en c _

e/ ffin t r SR 3.3.4.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

reld ori co jesie w yofthe mpca Its5 du asy ref popiu erýii l

c *scy REFERENCES 1. 10 CFR 50, Appendix A, GDC 19.

McGuire Units 1 and 2 B 3.3.4-5 Revision No$

LOP DG Start Instrumentation B 3.3.5 BASES SURVEILLANCE SR 3.3.5.1 REQUIREMENTS SR 3.3.5.1 is the performance of a TADOT.e

.. ýThe test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment. For these tests, the relay

.. 4~M1NL TIP SETPOINT p erified and adjusted as necessa T~F Fque ts ed th~e own eliab'g of th relays .

7 4d ontr an emicha -I re ndan availa ,and s ow be ceptVe thr gha erti exeine Testing consists of voltage sensor relay testing only. Actuation of load shedding and time delay timers is not required.

SR 3.3.5.2 r SR 3.3.5.2 is the performance of a CHANNEL CALIBRATION.

}

The setpoints, as well as the response to a loss of voltage and a degraded voltage test, shall include a single point verification that the trip occurs within the required time delay, as shown in Reference 1.

,2 (at e f i HANNLCLBAINi opeeceko h instrument loop, including the sensor. The test verifies that the channel respondtsU a-maured Pxaeter within the necessaryL ranqe and accuracy.

The F* u:of18 ntps* base on ope a ingheeen

? ~ iste t t t ndus.* refueli cyl

• f i s inc 18 hhca ration if~rval i! Win nýK m iueeuipýent drin the tpoint.,wi ss, If plant conditions warrant, the definition of NOMINAL TRIP SETPOINT.

provides an option for setting a trip setpoint in plant hardware outside the NOMINAL TRIP SETPOINT calibration tolerance band as long as the trip setpoint is conservative with respect to the NOMINAL TRIP SETPOINT.

Application of that provision to this SR could result in premature separation of safety related equipment from offsite power during switchyard voltage fluctuations. Consequently, this SR has been modified by a Note stating that a NOMINAL TRIP SETPOINT shall be set within the channel's calibration tolerance band.

REFERENCES 1. UFSAR, Section 8.3.

2. UFSAR, Chapter 15.

3. Loss of Voltage Relay Setting Calculation, MCC-13811.05-00-0094.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.3.5-5 Revision Noe

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS (continued)

C.1, C.2.1, C.2.2, and C.2.3 If the indicated RCS total flow rate is less than 99% of the value specified in the COLR, then RCS total flow must be restored to greater than or equal to 99% of the value specified in the COLR within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or power must be reduced to less than 50% RTP. The Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is consistent with Required Action A.I. If THERMAL POWER is reduced to less than 50% RTP, the Power Range Neutron Flux - High Trip Setpoint must also be reduced to _<55% RTP. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to reset the trip setpoints is consistent with Required Action B.2. This is a sensitive operation that may inadvertently trip the Reactor Protection System. Operation is permitted to continue provided the RCS total flow is restored to greater than or equal to 99% of the value specified in the COLR within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is reasonable considering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint.

D. 1 If the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable to reach the required plant conditions in an orderly manner.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms.and other indications are availle to alert operators ifthis limit is approached or exceede . hefr-equency'of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> McGuire Units 1 and 2 B 3.4.1-4 Revision Ndo

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.1.2 This surveillance demonstrates that the average RCS temperature remains within the required limits. Alarms and other indications are avaibeto op rs is s litisiaproachedhor eixceededns SR 3.4.1.3 This surveillance demonstrates that the RCS total flow rate remains within the required limits. Alarms and other indications are a a _lert operators if this limit is approached or exceeded hejpequency.of 1 SR 3.4.1.4 Calibration of the installed RCS flow instrumentation permits verification that the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.

REFERENCES 1. UFSAR, Section 15.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.4.1-5 Revision NOO

RCS P/T Limits B 3.4.3 BASES ACTIONS (continued)

Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains acceptable and must be completed prior to entry into MODE 4. Several methods may be used, including comparison' with pre-analyzed transients in the stress analyses, or inspection of the components.

ASME Code,Section XI, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the specified 4ýýfien RCS pressure and tem eratu undergoing planned changes,hisybequ.*,fc' Surveillance for heatup, cooldown, or ISLH testing may be discontinued.

when the definition given in the relevant plant procedure for ending the activity is satisfied..

'1ý This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.

McGuire Units 1 and 2 B 3.4.3-6 Revision Noýý

RCS Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABILITY (continued)

Operation in other MODES is covered by:

LCO 3.4.5, "RCS Loops-MODE 3";

LCO 3.4.6, "RCS Loops-MODE 4";

LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled";

LCO 3.4.17, "RCS Loops-Test Exceptions";

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6).

ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the plant to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.

The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.

SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verificationghat each RCS loop isin operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure.that forced flow is providing heat removal while maintaininq .the marqin to DNB, ýh2houms Wi-REFERENCES 1. UFSAR, Section 15.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.4.4-3 Revision No

RCS Loops - MODE 3 B 3.4.5 BASES ACTIONS (continued) to OPERABLE status and operation must be initiated. Boron dilution requires forced circulation for proper mixing, and opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to criticality.

The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verificatio hat the required loops are in operation. Verification includes flow rate, temperature, and pump status monitoring, which help ensure that forced flow is providing heat removal.

-Fr2end SR 3.4.5.2 SR 3.4.5.2 requires.verification of SG.OPERABILITY. SG OPERABILITY.

is verified by ensuring that the secondary side narrow range water level is

> 12% for. required RCS loops. If the SG secondary side narrow range water level is < 12%, the tubes may become uncovered and the associated loop may not be capable of rovidin the heat sink for removal of the decay hea I h Fr enc d a quate verin ion aiaa in thn,,ntro om 6_m ertUl&ae

.,td a s of leve.

SR 3.4.5.3 Verification that the required RCPs are OPERABLE ensures that safety analyses limits are met. The requirement also ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power availability to the required 2U 1 an Bhe3.sid4.5-5reason ein REFERENCES ,,Xcepta.bl y operati ~experienc McGuire Units 1 and 2 B 3.4.5-5 Revision No$'

RCS Loops - MODE 4 B 3.4.6 BASES ACTIONS (continued)

C.1 and C.2 If no loop is OPERABLE or in operation, except during conditions permitted by Note 1 in the LCO section, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1 and maintain Keff < 0.99 must be suspended and action to restore one RCS or RHR loop to OPERABLE status and operation must be initiated. The required margin to criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3:1.1 and maintains Keff < 0.99 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the minimum SDM and Keff requirements maintains acceptable margin to criticality. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification( ý zsthat one RCS or RHR loop is in operation. Verification includes flow rate, temperature, or pump status

"* y SR 3.4.6.2 requires verification of SG OPERABILITY. SG OPERABILITY verified nis by ensuring that the secondary side narrow range water level is RSG levpeT. r c S>12%. Ifthe SG secondary side narrow range water level is < 12%, the tubes may become uncovered and the associated loop may not be ca bl_ f roiinte heat sink necessary for removal of decay heat.

SR 3.4.6.3 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to McGuire Units 1 and 2 B 3.4.6-4 Revisionbe

RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE REQUIREMENTS (continued) maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker aliqnment and power available to the required um he Fr quency of 7 Wys iscoj~idered re is:onablej~

view othe dministr e co~ntrol ailable has bee n shown pv6e a by ope ing expri ce.

REFERENCES 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.4.6-5 Revision N9ýý

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES ACTIONS A.1 and A.2 If one RHR loop is inoperable and the required SGs have secondary side narrow range water levels < 12%, redundancy for heat removal is lost.

Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

B.1 and B.2 If no RHR loop is in operation, except during conditions permitted by Note 1, or if no loop is OPERABLE, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to criticality. The immediate Completion Times reflect the importance of maintaining operation for heat removal.

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS.

This SR requires verification(o~ that the required loop is in operation. Verification. includes flow rate, temperature, or pump status.

tat at least two SGs are OPERABLE by ensuring their secondary side narrow range water levels are > 12% ensures an alternate decay heat removal method in the event that the second RHR loop is not OPERABLE. If both RHR looas are OPERABLE this Surveillance is not McGuire Units 1 and 2 B 3.4.7-4 Revision Ný

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.7.3 Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the RHR pump.

If secondary side narrow range water level is _>12% in at least two SGs, this Surveillance is not neede The Fr uency' o as is c nsid re 9aena n vieo othe- dmine ative c ols avail an as

-ee owptbe a able operati eiperie REFERENCES 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.4.7-5 Revision N00

RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES ACTIONS (continued) initiated immediately to restore an RHR loop to OPERABLE status and operation. The required margin to criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to criticality.

The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must continue until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification that one loop is in operation.

Verification includes flow rate, temperature, or pump status monitoring, whjiLhhetJp ensure that foioed flow is providinci heat rernovLfh_-.-

SR 3.4.8.2 Verification that the required number of pumps are OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker ali nment and ower available to' the required pum s he REFERENCES 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.4.8-3 Revision N$K

Pressurizer B 3.4.9 BASES ACTIONS (continued)

C.1 and C.2 If one group of pressurizer heaters are inoperable and cannot be restored in the allowed Completion Time of Required Action B. 1, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The S rveillance is erformed b observin the indicated le-J ýh-Freauencv of I Thours ýcorresoondsto verifina the SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the associated. pressurizer heaters are verified to be at their design rating. This may be done by testing the power supply output and by performing an electrical check on heater element continuity and resistance. e eque o 9 ays i

ýý:ered, equt -o cons r deg atio ad*ha een ,hwn obpe;jaing ex5rie to b ccep REFERENCES 1. UFSAR, Section 15.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

3. NUREG-0737, November 1980.

McGuire Units 1 and 2 B 3.4.9-4 Revision No

Pressurizer PORVs B 3.4.11 BASES ACTIONS (continued)

H.1 and H.2 If the Required Actions of Condition F or G are not met, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4 and 5, maintaining PORV OPERABILITY may be required. See LCO 3.4.12.

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS

..... Block a

vav or yln e eiistattevles F o 92 *a ' t aecosed ifneeI*

ME ode, ec" e lc avei lsd slt PORV that is capable of valve is closed to isolate an otherwise inoperable PORV, the maximum Completion Time to restore the PORV and o en the bloc v Ive is PHt 9--" ti o to~extenQK 2 d urthermore, these test requirements would be completed by the reopening of a recently closed block valve upon restoration of the PORV to OPERABLE status (i.e., completion of the Required Actions fulfills the SR).

The Note modifies this SR by stating that it is not required to be met with the block valve closed, in accordance with the Required Action of this LCO.

SR 3.4.11.2 SR 3.4.11.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be The SR is modified by a Note which states that the SR is required to be performed in MODE 3 or 4 when the temperature of the RCS cold legs is

> 300OF consistent with Generic Letter 90-06 (Ref. 5).

McGuire Units 1 and 2 B 3.4.11-6 Revision Nq

Pressurizer PORVs B 3.4.11 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.11.3 The Surveillance demonstrates that the emergency nitrogen supply can be provided and is performed by transferring power from normal air supI to emer ency nitrogen supply and cydingthe valve The

• " k,._..*k, FrEccep r ncof 18

'te pra c e nths ibased ; typicajpefue-rin-ycle a 41ýduýity REFERENCES 1. Regulatory Guide 1.32, February 1977.

2. UFSAR, Section 15.4.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

4. ASME Code for Operation and Maintenance of Nuclear Power Plants.

5. Resolution of Generic Issue 70, "Power-Operated Relief Valve and Block Valve Reliability," and Generic Issue 94, "Additional Low-Temperature Overpressure Protection for Light-Water Reactors,"

Pursuant to 10 CFR 50.54(f) (Generic Letter 90-06).

McGuire Units 1 and 2 B 3.4.11-7 Revision No. 102

LTOP System B 3.4.12 BASES ACTIONS (continued)

a. Both required PORVs are inoperable; or

b. A Required Action and associated Completion Time of Condition C, D, E, or F is not met; or

c. The LTOP System is inoperable for any reason other than Condition A, B, C, D, E, or F.

The vent must be sized > 2.75 square inches to ensure that the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.

The Completion Time considers the time required to place the plant in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.

SURVEILLANCE SR 3.4.12.1 and SR 3.4.12.2 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, all but one centrifugal charging pump or one safety injection pump are verified incapable of injecting into the RCS

-and the accumulator discharge isolation valves are verified-closed and power removed (See Ref. 10).

The centrifugal charging pump and .safety injection -pump are rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control. An alternate method of LTOP control may be employed using at least two independent means to prevent a pump start such that a single failure or single action will not result in an injection into the RCS. This may be

~ -r accomplished through two valves in the discharge flow path being closed.

a9,nyth e,/5ý ro- rom ~i

'Tlh~em`F uency of hours is s cient, con ring other joationsand~N al ms availa to the op -4orin the crol roomt iyter~iie Zstatus of equipme SR 3.4.12.3 The RHR suction relief valve shall be demonstrated OPERABLE by verifying the RHR suction isolation valves are open and by testing it in McGuire Units 1 and 2 B 3.4.12-10 Revision No*"

LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued) accordance with the Inservice Testing Program. This Surveillance is only required to be performed ifthe RHR suction relief valve is being used to meet the Required Actions of this LCO.

ie RHR suction valves are verified to be 7V" a-s valve~tatus !pdlcations/,dvailableX

ý6nta*~iriyt" n az R sueton valvaremain theýý-insuqh opperr in the q-titrol The ASME OM Code (Ref. 9), test per Inservice Testing Program, verifies OPERABILITY by proving proper relief valve mechanical motion-and by measuring and, if required, adjusting the lift setpoint.

SR 3.4.12.4 The RCS vent of >2.75 square inches is proven OPERABLE by verifying its open condition The passive vent arrangement must only be open to be OPERABLE.

This Surveillance is required to be performed ifthe vent is being used to satisfy the pressure relief requirements of the LCO 3.4.12b.

SR 3.4.12.5 The PORV block valve must be verified openz e I" to provide the flow path for each required PORV to perform its function when actuated. The valve must be remotely verified open in the main control room. This Surveillance is performed if the PORV satisfies the LCO.

The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required removed, and the manual operator is not required locked in the inactive position. Thus, the block

'valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation.

The 7 oure co dd in viewn other ad nistrativ ontrols av able to th6perato in control ro , such s valve sition mdi ion, that vprffy that the RV block v ye remains open.

McGuire Units I and 2 B 3.4.12-11 Revision Noý6

LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.12.6 Performance of a COT is require _ithin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing RCS temperature to < 3001F and s n each required PORV to verify and, as necessary, adjus its lift setpoint. The COT will verify the setpoint is within the allowed maximum limits. PORV actuation could depressurize the RCS and is not required.

The c.* 3idersay xI frequ ility.

y The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> event during overpressure Frequency considers the unlikelihood of a low temperature this time.

A Note has been added indicating that this SR is required to be met 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing RCS cold leg temperature to < 3000 F. The test, must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering the LTOP MODES.

~~~~~SR .3.4.12.7 . . . .

3:7 A/ ,e_-'- ) "Performance of a CHANNEL CALIBRATION on each required PORV actdaJtion channel is requiredm to adjust the whole channel so that it responds and the valve opens within the required range REFERENCES Il 0 F 0,Apndix (G. ee* ience wi* .

2. Generic Letter 88-11. an m es theitypic

3. ASME, Boiler and Pressuir refu ing Qutage scdule.

5. 10 CFR 50, Section 50.46.

6. 10 CFR 50, Appendix K.

7. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

8. Generic Letter 90-06.

9. ASME Code for Operation and Maintenance of Nuclear Power Plants.

10. Duke letter to NRC, "Cold Leg Accumulator Isolation Valves", dated September 8, 1987.

McGuire Units 1 and 2 B 3.4.12-12 M Revision No.(0

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

The RCS water inventory balance must be performed with the reactor at steady state operating conditions and near operating pressure. The surveillance is modified by two Notes. Note 1 states that this SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful and a Note requires the Surveillance to be met when steady state is established. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past sealsand gaskets is not pressure.

boundary LEAKAGE. These leakage detection systems. are specified in LCO 3.4.15, "RCS Leakage- Detection Instrumentation.'

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 135 gallons per day cannot be.

measured accurately by an RCS water inventory balance.

SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 135 gallons per day through any one SG and less than or equal to 389 gallons per day total through all SGs. Satisfying the primary to secondary LEAKAGE limit ensures that the assumptions of the safety analyses are McGuire Units 1 and 2 B 3.4.13-5 Revision No_

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE (continued) met (Ref. 3). If this SR is not met, compliance with this LCO, as well as LCO 3.4.18, "Steam Generator Tube Integrity," should be evaluated. The 135 and 389 gallons per day limits are measured at a temperature of 585 0 F as described in Ref. 3. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure,

,, . temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 8).

REFERENCES 1. 10 CFR 50, Appendix A, GDC 30.

2. Regulatory Guide 1.45, May 1973.

3. UFSAR, Section 15.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

5. UFSAR, Table 18-1.

6. McGuire License Renewal Commitments MCS-1274.00-00-0016, Section 4.29, RCS Operational Leakage Monitoring Program.

7. NEI 97-06, "Steam Generator Program Guidelines".

8. EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

9. UFSAR, Table 15-24.

McGuire Units 1 and 2 B 3.4.13-6 Revision N

RCS PIV Leakage B 3.4.14 BASES ACTIONS (continued)

B.1 and B.2 If leakage cannot be reduced, or the other Required Actions accomplished, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 The RHR interlock prevents the RHR suction isolation valves inadvertent opening at RCS pressures in excess of the RHR systems design pressure. Ifthe RHR interlock is inoperable, operation may continue as long as the affected RHR suction penetration is closed by at least one closed manual or deactivated automatic valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This Action accomplishes the purpose of the interlock function.

SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each RCS PIV or isolation valve used to satisfy Required Action A.1 is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve diameter up-to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure

-condition.

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

Testin" to be p ormed evee 18 mnith , a ttypical r ýheling cycle if 2the ant does t go into M E5foat ast'dy The 18mo h McGuire Units I and 2 B 3.4.14-4 Revision No.(

RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE REQUIREMENTS (continued) a lor ,/a

/ lo otential rormed ellance wereat pd thin

/u )lanned trapsent if the Sury anrin with the'16eactor-g finiutage tsh hat ows cendition othe In addition, testing must be performed once after the valve has been opened by flow or exercised to ensure tight reseating. PIVs disturbed in the performance of this Surveillance should also be tested unless documentation shows that an infinite testing loop cannot practically be avoided. Testing must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the valve has been reseated. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is a reasonable and practical time limit for performing this test after opening or reseating a valve.

The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance. The Note that allows this provision is complementary to the Frequency of prior to entry into MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months. In addition, this Surveillance is not required to be performed on the RHR System when the RHR System is aligned to the RCS in the shutdown cooling mode of operation. PIVs contained in the RHR shutdown cooling flow path must be leakage rate tested after RHR is secured and stable unit conditions and the necessary differential pressures are established.

SR 3.4.14.2 Verifying that the RHR interlock is OPERABLE ensures that RCS pressure will not pressurize the RHR system beyond its design pressure of 600 psig. The interlock setpoint that prevents the valves from being opened is set so the actual RCS pressure must be < 425 psig to open the valves. This sebpoint ensures the RHR design pressure will not be Ms aed 2 Uniotsthe B e.14Ri Unit 2ur Mcpdr1ef~ Ba3..1- the Revision Ng.n

RCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS (continued)

E._1 With the incore sump level alarm inoperable, a water inventory balance, in accordance with SR 3.4.13.1, must be performed at an increased frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to provide alternate periodic information that is adequate to detect leakage. Required Action E.1 is modified by a Note that states the RCS water inventory balance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation in accordance with SR 3.4.13.1. This Note allows exceeding the 24-hour completion time during non-steady state operation.

F.1 and F.2 If a Required Action of Condition A, B, C, or D cannot be met, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, toreach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

G.1 With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required. The required monitors during MODE I for LCO 3.0.3 entry are defined as the simultaneous inoperability of one CFAE level monitor, the containment atmosphere particulate radioactivity monitor, and the CVUCDT level monitor. The required monitors during MODES 2,'3, and 4 for LCO 3.0.3 entry are defined as the simultaneous inoperability of.one CFAE level monitor and the CVUCDT level monitor.

This Condition does not apply to the incore instrument sump level alarm..

SURVEILLANCE REQUIREMENTS SR 3.4.15.1 SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the containment atmosphere particulate radioactivity monitor. The check

-2 gves reasonable confidence that the channel is operating proper sThe.

McGuire Units I and 2 B 3.4.15-8 Revision N

RCS Leakage Detection Instrumentation B 3.4.15 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the containment atmosphere particulate radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. The test verifies ecothe alar laetpection instr eintstrument strin e cFra cy of tahe in sint i lratingthe Pstrument

/..h*_Y*'- ' / SR 3.4.15.3, SR 3.4.15.4, SR 3.4.15.5, and SR 3. 4.

15.6 REFERENCES

1. 10 CFR 50, Appendix A, Section IV, GDC 30.

2. Regulatory Guide 1.45.

3. UFSAR, Section 5.2.7.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).. t i5. t UFSAR, Table i18-1.. .. ..

6. McGuire License Renewal Commitments MCS-i274.00-00-00.6, Section 4.29, RCS Operational Leakage Monitoring Program.

7. McGuire Safety Evaluation Report, Section 5.2.5.

8. UFSAR, Table 5-30.

McGuire Units 1 and 2 B 3.4.15-9 Revision N4

RCS Specific Activity B 3.4.16 BASES ACTIONS (continued)

B. 1 With the gross specific activity in excess of the allowed limit, the unit must be placed in a MODE in which the requirement does not apply.

The change within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to MODE 3 and RCS average temperature

< 500OF lowers the saturation pressure of the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderlymanner and without challenging plant systems.

C._1 If a Required Action and the associated Completion Time of Condition A is not met or if the DOSE EQUIVALENT 1-131 is in the unacceptable region of Figure 3.4.16-1, the reactor must be brought to MODE 3 with RCS average temperature < 500OF within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolantl A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 10 minutes and all radioiodines.

The total specific activity shall be the sum of the beta-gamma activity in the sample within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the sample is taken and extrapolated back to when the sample was taken. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level. The latest available data may be used for pure beta-emitting radionuclides. This Surveillance provides an indication of any increase in gross specific activity.

McGuire Units 1 and 2 B 3.4.16-4 Revision Nc(o

RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued)

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The Surveillance is applicable in MODES 1 and 2, and in MODE 3 with Tav at least 5000F .. e ay F..,requen0ý~ cojýsider

• uS *ali 3ýss!Vf.failr ngtem SR 3.4.16.2 This Surveillance is performed in MODE 1 only to ensure iodine remains within limit during normal operation and following fast power changes when fuel failure is more apt to occur. jhe4-day Frequency is equ

• e*4n th i di e ~ ravt~ e~v el onj*!* f g gr,2 aactiv

• der s7 heFrequency, between 2 an-6 hours after a power change >_15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.

"SR 3.4.16.3 A radiochemical analysis for E determination is requiredqýiýý Swith the plant operating in MODE 1 equilibrium conditions.

The E determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for E is a measurement of the average energies perdisintegration for L isotopes. with half lives longer than 10 minutesexcluding iodines(-

F---iepey1 of)184tdays.elo*--g'- E dp~not Sbange rap dry.

This SR has been modified by a Note that indicates sampling is required.

to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This ensures that the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event.

REFERENCES 1. 10 CFR 100.11, 1973.

2. UFSAR, Section 15.6.3.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.4.16-5 Revision NCW

RCS Loops - Test Exceptions B 3.4.17 BASES SURVEILLANCE SR 3.4.17.1 REQUIREMENTS Verification that the power level is < the P-7 interlock setpoint (10%) will ensure that the fuel design criteria are not violated during the performance of the PHYSICS TEST . he Fr~equency of o[ce per hour SR 3.4.17.2 The power range and intermediate range neutron detectors and the P-7 interlock setpoint must be verified to be OPERABLE and adjusted to the proper value. A COT is performed prior to initiation of the PHYSICS TESTS. This will ensure that the RTS is properly aligned to provide the required degree of core protection during the performance of the PHYSICS TESTS.

REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50, Appendix A, GDC 1, 1988.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.4.17-3 Revision N

Accumulators B 3.5.1 BASES SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Each accumulator valve should be verified to be fully openjýý

ý This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve c-Qud result in not

" SR 3.5.1.2 and SR 3.5.1.3 k2 ic u e-y o m/*h 'miewcf d staJi*esg oftdb t/eqmeeora accteuaOulatore te d water volume and nitrogen cover pressu ssweratroetf re verified for ea ccumulator. Thisi ca per "N- Fre--qu"- cyh--

to b ppr*pctnn and2 corr it]kon of off SR 3.5.1.4 The boron concentration should be verified to be within required limits fore

• / acmlt" tipdeach a ince the static design of the accumulators limits the ways in which the concentration can be changed.

identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration ifthe added water inventory is from the refueling water storage tank (RWST), because the water contained in the RWST is within the accumulator boron concentration requirements. This is consistent with the recommendation of NUREG-1366 (Ref. 7).

SR 3.5.1.5 terhftcaoiwi**ats removed from each accumulator isolation valve operator (see Ref. 8) when the RCS pressure ise> 1000 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to Veifcain

ý thtpoeri 6 rmoedfomeah cumlao McGuire Units 1 and 2 B 3.5.1-7 Revision N3

Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued) occur, only two accumulators would be available for injection given a single failure coincident with a LOG . Si~ pow!er* rem:ri1ieýun e~r a~~ nra ~ ~ ~ ~ ecwWrovjdi

~ vt ptai adeq

• srnet-hat p rw is remo Tihis SR allows power to be supplied to the motor operated isolation T

valves when RCS pressure is < 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. Even with power supplied to the valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves.

Should closure of a valve occur in spite of the interlock, the SI signal provided to the valves would open a closed valve in the event of a LOCA.

REFERENCES 1. IEEE Standard 279-1971.

2. UFSAR, Chapter 6.

3. 10 CFR 50.46.

4. DPC-NE-3004.

5. 10 CFR 50.36, Technical Specification, (c)(2)(ii).

6. WCAP - 15049-A, Rev. 1, April 1999,

7. NUREG-1366, .February 1990.

8. Duke letter to NRC, "Cold Leg Accumulator Isolation Valves", dated September 8, 1987 McGuire Units 1 and 2 B 3.5.1-8 Revision NIP

ECCS-Operating B 3.5.2 BASES ACTIONS (continued)

An event accompanied by a loss of offsite power and the failure of an EDG can disable one ECCS train until power is restored. A reliability analysis (Ref. 6) has shown that the impact of having one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Reference 7 describes situations in which one component, such as an RHR crossover valve, can disable both ECCS trains. With one or more component(s) inoperable such that 100% of the flow equivalent to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered.

B.1 and B.2 If the inoperable trains cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves using the power disconnect switches in the correct position ensures that they' cannot change position as a result of an active failure or be inadvertently misaligned. These valves are-of the type;, described in Reference 7, that K jz-rcan disable th e function of both ECCS, trains and invalidate the accident an ~11Ju eyiy o r esp enjw of o ýr iita'controlskait will enswr'a mispositdn-ed va lve-isunlik y.

SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, McGuire Units 1 and 2 B 3.5.2-7 Revision No9

ECCS-Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued) or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that th se valves capable of being mispositioned are in the correct position.m/The`."

accesslelocatios scptibeto ga accumu e valves aAoperated marn u

ad . strati ontro / /

S This Fr u nyh be ns oý o b a c le through 9perating

• may be used to verify water-filled conditions (e.g., ultrasonic testing or high point sightglass observation). Maintaining the piping from the ECCS pumps to the RCS full of water ensures that the system will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent water hammer, pump cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an Sl signal or during shutdown coolin . "e 3 a-

"pe. a s ip't considerat0 thegofas /

( a~cpu'rmulation/juythe ECCS -nrg and the/ o~cedura~l co ;ols gov ng

"*'* -*2* ,'* ystem oP ation .i ]*

SR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME OM Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis.

SRs are specified in the Inservice Testing Program, which encompasses the ASME OM Code. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements.

McGuire Units 1 and 2 B 3.5.2-8 Revision N(fl5

ECCS-Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.5 and SR 3.5.2.6 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the re uired on u ea inistrative co ols e 18 mont Frquenc is based on th*- need to perforr e

,, ellns nerecodti ta ly rng a plant otage and

/ thlfrplne he o t rnint-Te Surveill arces were /

pe rmed w thereact atpower. e 18 month F quency is a ccptb -)sdo sdrantedsgn rfiability (and coi

• i g operat xei~c h qi t he actu i'on logic is tt as part of SF Actuati ystem testi , and equip nt pe ormance i monitored a part of the Ins rvice Testing rogram.

SR 3.5.2.7 i

The position of throttle valves in the flow path on an SI signal is necessary for proper ECCS performance. These valves have mechanical locks to ensure proper positioning for restricted flow to a ruptured cold leg, ensuring that -- t e other cold legs receive at least the reIuired V SR..528 2 . . .8 Periodic inspections of the ECCS containment sump strainer assembly (consisting of modular tophats, grating, plenums and waterboxes) and the

..associated enclosure (the stainless steel structure surrounding the.

strainer assembly located inside the crane wall) ensure they are unrestricted and stay in proper operating condition. Inspections will consist of a visual examination of the exterior surfaces of the strainer assembly and interior and exterior surfaces of the enclosure for any evidence of debris, structural distress, or abnormal corrosion. The intent of the surveillance is to ensure the absence of any condition which could adversely affect strainer functionality. Surveillance performance will not require removal of any tophat modules, but the strainer assembly exterior shall be visually inspected. This inspection will necessarily entail opening the top of the enclosure to allow access for inspection of the strainers, and to verify cleanliness of the enclosure interior space. A detailed inspection of the enclosure and exterior strainer assembly surfaces is required to establish a hiqh confidence that no adverse conditions are present. 8 r:nrrth Freqfe-ncy is-b:ed o e nee e Revision Units 1 McGuire Units and 2 1 and 2 B 3.5.2-9 B 3.5.2-9 Revision NGT

ECCS-Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 35.

2. 10 CFR 50.46.

3. UFSAR, Section 6.2.1.

4. UFSAR, Chapter 15.

5. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

6. NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

7. IE Information Notice No. 87-01.

B 3.5.2-10 Revision N McGuire Units 1 McGuire Units and 2 1 and 2 B 3.5.2-10 Revision N

RWST B 3.5.4 BASES ACTIONS (continued) restore the RWST to OPERABLE status is based on this condition simultaneously affecting redundant trains.

C.1 and C.2 If the RWST cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.4.1 REQUIREMENTS The RWST borated water temperature should be verifie h r

_tob within the limits assumed in the accident anal ses band. hi

(*"TA*A iý- Fr>uency is--sýWcient to id'e if a temper r chanp-that ho

_,,-1cp ro* i tferileim cit n s e n shj;¢ to be aedep' be o g SR 3.5.4.2 The RWST water volume should be verifiedcrw7 to be above the required minimum level in order to ensure-that a sufficient initial supply is available for injection and to support continued ECCS and Containment

- S Ira stempumnp operation on recircjlation. Sincný& the .S volume L*"T" ) /"is n~rally--st . Fe and is jrtectecJa an m dayl* n~cy i1"..

IA:

• // (*propri and has en so*wKto be acc able thr pr SR 3.5.4.3 The boron concentration of the RWST should be verified c6i,7 to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA and that the boron content assumed for the injection water in the MSLB analysis is available. Further, it assures that the resulting sump pH will be maintained in an acceptable.range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be Mcuireince andy2B54y RWS ume is nor stable, a sai

(" r/e erency t rify boro 6ncentratio viappropriate

• ardhas b~i

• -sown t acceptablpthrough operating experie4e McGuire Units 1 and 2 B 3.5.4-5 Revision No~_

Seal Injection Flow B 3.5.5 BASES ACTIONS (continued) operator has 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from the time the flow is known to be above the limit to correctly position the manual valves and thus be in compliance with the accident analysis. The Completion Time minimizes the potential

• exposure of the plant to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow within limits.

This time is conservative with respect to the Completion Times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.

B.1 and B.2 When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators. Continuing the plant shutdown begun in Required Action B.1, an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.

SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Verification that the manual seal injection throttle valves are adjusted to give a flow within the limit ensures that proper manual seal injection throttle valve position, and hence, proper seal injection flow, is maintained h-e u I r *1*a ay sed -on n e ug a nie.

is e!'esistent c iother a

S val oven t[o b64cetabl rveillancrog.

As noted, the Surveillance is required to be performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after

> -the RCS pressure has stabilized within a +/- 20 psig range of normal operating pressure. The RCS pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly. The exception is limited to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to ensure that the Surveillance is timely.

McGuire Units 1 and 2 B 3.5.5-3 Revision N$"S

Containment Air Locks B 3.6.2 BASES Additionally, the affected air lock(s) must be restored to OPERABLE status within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time. The specified time period is considered reasonable for restoring an inoperable air lock to OPERABLE) status, assuming that at least one door is maintained closed in each affected air lock.

D.1 and D.2 If the inoperable containment air lock cannot be restored to OPERABLE status within the required Completion time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.2.1 REQUIREMENTS Maintaining containment air locks OPERABLE requires compliance with the leakage rate test requirements of the Containment Leakage Rate Testing Program. This SR reflects the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The acceptance criteria were established during initial air lock and containment OPERABILITY testing. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by the Containment Leakage Rate Testing Program.

The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous: successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria which are applicable to SR 3.6.1.1. This ensures that air lock leakage is properly accounted for in determining the combined Type B and C containment leakage rate.

SR 3.6.2.2 Door seals must be tested Eýý3o verify the integrity of the inflatable door seal. The measured leakage rate must be less than 15 standard cubic centimeters per minute (sccm) per door seal when the seal is inflated to approximately 85 psig. This ensures that the seals will remain inflated for at least 7 days should the instrument air supply to the McGuire Units I and 2 B 3.6.2-6 Revision No-

Containment Air Locks B 3.6.2 BASES SURVEILLANCE REQUIREMENTS (continued) seals be.6 th, eugh operati ac,£ ptable exper"stings been dofonstratedte.

SR 3.6.2.3 S

The air lock interlock is designed to prevent simultaneous opening of both

/ doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY. Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates thatthe of theinner interlock will function and outer as not doors will designed and thatoccr inadvertently simultaneous u t he opening

,diverse permissiv e ogic arrangemen o this inter ock, and given that the interlock me anism is not mally challenge en the contain t air lock do -Is used for e and exit (proce es require strict a erence to gile door opeg), this test is onl equired to be perf ed every 18 onths. The 1 month Frequenc based on the nee o perform this surveillanc nder the conditio that apply during a ant outage, and the pot al for loss it ofheco co inment tpw OPERABILI h8 if the surveillance, month Freunfo t ne l c s t a e o e e i atn *e~xperi,,.

lence.* *he Frequency is b qd onsenierigconsidlerd adeqat

• given. that the inelc sntcalenge* during the use of the interlock,.j

.REFERENCES 1. 10 CFR'50, Appendix J, Option B.

2. UFSAR, Section 6.2.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.6.2-7 Revision NCO

Containment Isolation Valves B 3.6.3 BASES ACTIONS (continued) necessary to ensure that containment penetrations required to be isolated following an accident, which are no longer capable of being automatically isolated, will be in the isolation position should an event occur. This Required Action does not require any testing or valve manipulation.

Rather, it involves verification, through a system walkdown or computer status indication, that those isolation devices outside containment capable of being mispositioned are in the correct position. For the isolation devices inside containment, the time period specified as "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the isolation devices and other administrative controls that will ensure that isolation device misalignment is an unlikely possibility.

For the containment purge valve with resilient seal that is isolated in accordance with Required Action E.1, SR 3.6.3.6 must be performed at least once every 92 days. This assures that degradation of the resilient seal is detected and confirms that the leakage rate of the containment purge valve does not increase during the time the penetration is isolated.

F.1 and F.2 If the Required Actions and associated Completion Times are not met, the plant must be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner' and Without challenging plant systems.

SURVEILLANCE SR 3.6.3.1 REQUIREMENTS Each containment purge supply and exhaust valve for the lower compartment, upper compartment, andent room is required to be verified sealed closed]n3

• This Surveillance is designed to ensure that a gross breach of containment is not caused by an inadvertent or spurious opening of a containment purge valve. These valves are required to be in the sealed closed position during MODES 1, 2, 3, and 4. A valve that is sealed closed must have motive power to the valve operator. removed. This can McGuire Units 1 and 2 B 3.6.3-9 Revision N

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued) be accomplished by de-energizing the source of electric power or by removing has "sealed" the no air connotation supply to theofvalve leak operator.

tiht es In he this -application, uency i" the term ,_

r-esuL.

agage requires entry into Condition E, the Surveillance permits opening one purge valve in a penetration flow path to perform repairs.

SR 3.6.3.2 Not Used

,SR 3.6.3.3 This SR requires verification that each containment isolation manual valve and blind flange located outside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. This SR does not require any testing or

-valve manipulation. Rather, it involves verification, through a system walkdown or computer status indication, that those containment isolation valves outside containment and caple of being mispositioned are in the correct positionl vSce yetrificatiareof underositoaqmor containmten

\~i iv to - "

e

• • Cfisythat

~~containment isolation valves that are open. under administrative controls ..

McGuire Units I and 2 B 3.6.3-10 Revision No.9

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)

OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses.

The isolation time is specified in the UFSAR and Frequency of this SR.

are in accordance with the Inservice Testing Program.

SR 3.6.3.6 For containment purge valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J, Option B is required to ensure OPERABILITY. The measured leakage rate for containment purge lower compartment and incore instrument room valves must be < 0.05 La when pressurized to Pa. The measured leakage rate for containment purge upper compartment valves must be < 0.01 La when pressurized to Pa. Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types. Based on this observation and the importance of maintaining this penetration leak tight (due to the direct path between containment and the environment), these valves will not be placed on the maximum extended test interval, but tested on the nominal test interval in accordance with the Containment Leakage Rate Testing Program.

SR 3.6.3.7 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment..

isolation valve will actuate to its isolation position on a containment isolation signal. The isolation signals involved are Phase A, Phase B, and Safety Injection. This surveillance is not required for.valves that are.

locked, sealed, or otherwise, secured in the required position under:

Units1

..... administrative pefrmtiurvilace controls.62e edr the conditio sthat apply RonthFqenteeo durl'inag.ant

/outage ad the potenti or an unplann edtransient if the S .eillance .

/ we rore w e reactor at p.,er. Operating e xrien ce ha*

/ l*own that the *ecomponents us I~ly pass this Surveflance we/*

pe*rfollrmed e 18 month Fr uency. Therefor , the Freque~y was rocundtoee 18ceptnlro mU renliabilityetf°pont McGuire Units 1 and 2 B 3.6.3-12 Revision Nd-(S

Containment Pressure B 3.6.4 BASES ACTIONS A.1 When containment pressure is not within the limits of the LCO, it must be restored to within these limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment," which requires that containment be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ensures that unit

• o £_eration remains within the limits assumed in the containment analysis.

ITh-ie h r:Fequency f this was eveoped basednoperai

..expe nce related t rending of con ment pressure nations dun.

t applicable ES.. Further re, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> quency is considered quate in view other indicatlo available in t control room, in ing alarms, to ert the operator o an ab ral ontainment.

press econdition.

REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50, Appendix K.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.6.4-3 Revision N

Containment Air Temperature B 3.6.5 BASES SURVEILLANCE SR 3.6.5.1 and SR 3.6.5.2 REQUIREMENTS Verifying that containment average air temperature is within the LCO limits ensures that containment operation remains within the limits assumed for the containment analyses. In order to determine the containment average air temperature, a weighted average of ambient air temperature monitoring stations is calculated using measurements taken at locations within the containment selected to provide a representative sample of the overall containment atmosphere. The weighted average is the sum of each temperature multiplied by its respective containment volume fraction. In the event of inoperative temperature sensor(s), the weighted average shall be taken as the reduced total divided by one minus the volume fraction represented by the sensor(s) out of service.

The upper compartment measurements should be taken at elevation 826 feet at the inlet of each upper containment ventilation unit. The lower compartment measurements should be taken at elevation 745 feet at the

• /fE *-) /so inleth of each lower is concontainment reccep be based ventilation unit.o'n ohe 24 ed hourslow ratp o Frequenc

- temp ature inc ase within c ainment as a r It of environm etal a rasourceotiue t the pe volume of coninment). Fu.t rmore, t'e I J24 hour F quency is *sidered adequet in view of otherindicati as abrmal contain fent temperature ondition. " .

REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50.49.

3. 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.6.5-4 Revision NCO

Containment Spray System B 3.6.6 BASES ACTIONS (continued)

B.1 and B.2 Ifthe affected containment spray train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.

SURVEILLANCE SR 3.6.6.1 REQUIREMENTS Verifying the correct alignment of manual, power operated, and automatic valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray.System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since they were verified in the correct position prior to being secured. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown or computer status

'* * .ii indication, potentially that beingthose valves outside mispositionedi, containment are in the correct and capable of position.:

nier ugmeVs consi nte"

_T h*ee a qe *sb~dq theo.e  ! iand en rues*

• ,or *valve p*itions. ..

the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 6). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on bypass flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.

The Frequency of this SR is in accordance with the Inservice Testing Program.

McGuire Units 1 and 2 B 3.6.6-5 Revision N<'(g

Containment Spray System B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.6.3 and SR 3.6.6.4 These SRs require verification that each automatic containment spray valve actuates to its correct position and each containment spray pump starts upon receipt of an actual or simulated Containment Pressure High-High signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required osition under administrative controls- e'montrequency is based on the need to Iperform these.

Surveill es under e conditions t t apply durinn plant outage nd the tential for unplanned tr sient ifthe Su eillances wer erformed wit he reactor at wer. Operati experience s shown these com nents usuallyass the Surveii nces when pformed at the 18 mon Frequency.ency was cocluded to accep able from a r iability standpoint.

The surveillance of containment sump isolation valves is also required by SR 3.6.6.3. A single surveillance may be used to satisfy both requirements.

SR 3.6.6.5 and SR 3.6.6.6 These SRs require verification that each containment spray pump discharge valve opens or is prevented from opening and each containment spray pump starts or is de-energized and prevented from

. starting upon receipt of Containment Pressure Control System start and terminate* signals. The CPCS is described in the Bases for LCO3.3.2, "ESFAS." he* mon. re y ased on the ed. to pe these Sui*ces under t ondition.s that al0fy during a p)anitoutage.

SR 3.6.6.7 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. The spray nozzles can also be periodically tested using a vacuum blower to induce air flow through each nozzle to verify unobstructed flow. This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an M uitest at ear intera s conside-k~~~te Sl:ay nozzle../*

ade uate tetect obstruio*n of 5 McGuire Units 1 and 2 B 3.6.6-6 Revision NOO '

HSS B 3.6.8 BASES SURVEILLANCE SR 3.6.8.1 REQUIREMENTS Operating each HSS train for _>15 minutes ensures that each train is OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan and/or motor failure or excessive vibratincnb eetdfr corrciv cto.he 2djay Freq,,ency i fconsse ihrise e i-n- P zragm Surv i nc Frequgericies, , ]

ope . xeie h rlaiiyo efnotors-and coptrols,/

atwotr th *redudanavailable.

SR 3.6.8.2 Verifying HSS fan motor current at rated speed with the motor operateo suction valves closed is indicative of overall fan motor performance and system flow. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indcatin abnormal performance. he IFrejecy--oM2 dqys was based on operati-n-g exp/renci hich ha shown this Ffequency Wo-b acceptab,61 SR 3.6.8.3 This SR verifies the operation of the motor operated suction valves and

> *HSS fans in response to a start permissive from the Containment Pressure Control System CPCS). The CPCS is described in the Bases

- * ~~for LCO 3.3.2,'.SAF. The F r46qency ofi- day as ased o SR 3.6.8.4 This SR ensures that each HSS train responds properly to a containment pressure high-high actuation signal. The Surveillance verifies that each fan starts after a delay of > 8 minutes and _<10 minutes.Te Frequenc o92 as conforms the testing requirements fo similar ESF equip nt and co iders the kn9W reliability o n motors an controls an ue two trý* redundancyvailable. Th fore, the Fre ency was

" ncluded tdbe acceptab16 from a reliabity standpoint.

McGuire Units 1 and 2 B 3.6.8-4 Revision No.0

HMS B 3.6.9 BASES ACTIONS (continued)

C. 1 The unit must be placed in a MODE in which the LCO does not apply if the HMS subsystem(s) cannot be restored to OPERABLE status within the associated Completion Time. This is done by placing the unit in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.9.1 REQUIREMENTS This SR confirms that > 34 of 35 hydrogen ignitors can be successfully energized in each train. The ignitors are simple resistance elements.

Therefore, energizing provides assurance of OPERABILITY. The allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise redundancy in that region, the containment regions are interconnected so that ignition in one region would cause burning to progress to the others (i.e., there is overla in each hydrogen ignitor's effectiveness between 1re~gions),ý-he F enc y .- 2 day has b i hoyto'---t-o--b-e--)

11 ahting experec.

SR 3:6.9.2 This SR confirms that the two inoperable hydrogen ignitors allowed by tSR 3.6.9.1 (i.e., one in each train) are not in the same containment

• r ion he Fnrequency o9*92 days is acep a e base o .te quencyý .

of, which provides the infefmation for pe rfmT-ming this ,3.6 SR 3.6.9.3 A more detailed functional test is performed to verify system OPERABILITY. Each glow plug is visually examined to ensure that it is clean and that the electrical circuitry is energized. All ignitors (glow plugs), including normally inaccessible ignitors, are visually checked for a glow to verify that they are energized. Additionally, the surface temperature of each glow plug is measured to be > 1700'F to mdemonstrate that a temperature sufficient for ignition is achievedj./f-e, 1.g8 m h Freency is based-on the n-e- to pýorm thyi uyIace McGuire Units 1 and 2 B 3.6.9-4 Revision NT

HMS B 3.6.9 BASES SURVEILLANCE REQUIREMENTS (continued) under the contions that appl& during a plant outage and t o ential fo an unplan d transient if Surveillance w performed wit e react t power. Op ating experienc ýasshown that th e c ponents usuaý pass the SR w n performed at 18 month requency, w is based on t refueling cycle herefore, th reque cy was concluded t e acceptable from a r biiQ t.

REFERENCES 1. 10 CFR 50.44.

2. 10 CFR 50, Appendix A, GDC 41.

3. UFSAR, Section 6.2.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.6.9-5 Revision N

AVS B 3.6.10 BASES ACTIONS (continued) 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.10.1 REQUIREMENTS Operating each AVS train from the control room with flow through the HEPA filters and activated carbon adsorbers ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for

>_ 10 continuous hours eliminates moisture on the adsorbers and HEPA filters. Experience from filter testing at operating units indicates that the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period is adequate for moisture elimination on the adsorbers and HEPA filters. Inoperable heaters are addressed by Required Actions B.1 and B.2. The inoperability of heaters between required performances of this surveillance does not affect OPERABILITY tsein of each AVS traifilter accordac y was dvelaoner in th Pg (VFTP).oThreliabiS' tofe, and th phscoaols, the rtieofeheativaed ca air ( , al*ue sti s y Sy S Cond/e ser. /.

fite i ccrdne ess~r it egltoyude1.2(Rf 5) wit SR 3.6.10.2

.. * . * .. This SR verifies that the required AVS filter testing is performed in.

_ * "*. additional infomaton accordance aredicuseon with the Ventilation dtaiin~~

Filter Testing VFTP.

thaySye Program r (VFTP). .The.AVS.

• -* .- filter tesosare y,/-.,exceptions as in accordance noted with ReCulatory

-in the UFSAR. The VFTP Guide 1.52 (Ref. 5) with includes testinig HEPA'.-

,.*/*/* .filter %performance,charcoal aidso~rb~er efficiency, minimum system flow-rate, and the physical properties of the activated charcoal (general use Sand

/* following additional specific operations).

information are discussedSpecific in detailtest frequencies in the VFTP. and

~SR 3.6.10.3 I' The automatic startup on a Containment Phase B Isolation signal ensures Sthat each AVS train responds properlyi he 1 moniit r:eK uency is ._

\ bse n te neec I per~form this Surveillance under" ch-onditions that\

/ apply i**n~g a pla utage an tnta o p,)dhlanned tran 'ent if .....

S the/ urveillan/ ewere perforo, with the react/orlat power. 0O ating

\/ere.. assont o mpone/p., usually pas/ e /,

Suvilcwerfermed at the 18 rndnth Frequency:.*--J McGuire Units 1 and 2 B 3.6.10-4 Revision NocfW

AVS B 3.6.10 BASES SURVEILLANCE REQUIREMENTS (continued)

Therefo the Frequ ncy was conc ded to be .cce*pable om a reli ity stand t. FurtherT21e, the SR iprval was eveloped nsiderin at the AVS e ipment O IABILITY demonstrad at a 31 day Frequency by S 3.6.10.1.

SR 3.6.10.4 The AVS filter cooling electric motor-operated bypass valves are tested to verify OPERABILITY. The valves are normally closed and may need to be opened to initiate miniflow cooling through a filter unit that has been shutdown following a DBA LOCA. Miniflow cooling may be necessary to limit temperature increase in the idle filter train due to decay heat from ca tured fission products. e mon requency is coesidered tob1 ,h acceptab a ve reliabilit d design, and> fact thq- /

oper g expere has shown at the valves ally pass the

-$.uveillance en performed the 18 month Aqueny.

SR 3.6.10.5 The proper functioning of the fans, dampers, filters, adsorbers, etc., as a system is verified b the ability of each train to produce the required s st !ow rate. The nonth Freuency is cs jnsistent with gulato

" Guidp4.5 5) g dance fo-46nctional)sting,<ý ---- )

REFERENCES 1. 10 CFR 50, Appendix A, GDC 41.

2. UFSAR, Section 6.2..

3. UFSAR, Chapter 15,

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

5. Regulatory Guide 1.52, Revision 2.

Revision N and 22 Units 11 and McGuire Units B 3.6.10-5 B 3.6.10-5 Revision N

ARS B 3.6.11 BASES ACTIONS A.1 If one of the required trains of the ARS is inoperable, it must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was developed taking into account the redundant flow of the OPERABLE ARS train and the low probability of a DBA occurring in this period.

B.1 and B.2 If the ARS train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.11.1 REQUIREMENTS Verifying that each ARS fan starts on an actual or simulated actuation signal, after a delay >_8.0 minutes and _ 10.0 minutes, and operates for

>_15 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly. It also ensures that blockage, fan and/or motor failure, or excessive vibration can be detected for correcti te actirThe9say Frequ cy rai dampe c on one opera c of the fand hs ts SR 3.6.11.2 indicative of ovrallfn moto ri perorane. Sc nevc et ofr Verifying ARS fanmotorcurrent falrsb.idctn abora to be at ratedaspeed erom*e eF withurc the return~air dampers closed confirms one operating condition of the fan. This test is

=- //indicative of overall fan motor performance. Such inservice tests confirm suvellcomponent OPERABILITY, trend performancel and detect incipient failures by indicating abnormal performanceo -I'-eve dy confr~ms ih tsi ngrequiregnents for simil SF equi ent a/lxf

k. ( coperders tJreknowD 'lia bzility/effan motors,,and contro sand the/N0

-*--*k*,Afin reýV~d ancy/ ailableJ*--

~SR 3.6.11.3 Verifying the OPERABILITY of the return air damper provides assurance that the proper flow path will exist when the fan is started. This surveillance also tests the cicir, nldn time delayato ensure the system operates properly. (Thj Fre-*Hy;odaysWs w-d-eve*__ped McGuire Units 1 and 2 B 3.6.11-4 Revision N

ARS B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued) consider" the impo ce of the dam s, teir locaf pyia en'! nment, an robability of fail . Operatin peience h~I o Wown this Fr uency to be aceptable.

SR 3.6.11.4 and SR 3.6.11.5 Verifying the OPERABILITY of the check damper in the air return fan discharge line to the containment lower compartment provides assurance that the proper flow path will exist when the fan is started and that-reverse flow can notR!occuri when the i efan is not operating.J/The-frýwncy eimportanceo pers, of

_m t* 92* /,

loaion yia nvrneiadpoai O ng OP_.e.

ex.Z i nc hs .ls.sw~n this Frequenc o6be acpal SR 3.6.11.6 and SR 3.6.11.7 These SRs require verification that each ARS motor operated damper opens or is prevented from opening and each ARS fan is allowed to start or is prevented from starting upon receipt of Containment Pressure Control System start permissive and terminate si nals. The CPCS is described in the Bases for LCO 3.3.2, "ESFASY Th8 month equency is d oaleratin.xperience ic as s2 w-n it tobbbbcceptWb Y<

REFERENCES 1. UFSAR, Section 6.2.

2. 10'CFR 50, Appendix K.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.6.11-5 Revision No

Ice Bed B 3.6.12 BASES SURVEILLANCE SR 3.6.12.1 REQUIREMENTS Verifying that the maximum temperature of the ice bed is < 27 0 F ensures

• that the ice is kept well below the melting poin Tzhe 12 ho rFeur n

• ' //wa5s~ed on.pWferating ex~priernce, whi hconfirmed tb, due to the

(//[age masf*,,is stsOrednr,&1 not po $lle for the ice/ed temperatupe* to/

-,/ degr iniiaW ithin .a 12z1 u r period and as also bas eon [

as ssing the oximity of th C limit to th elting temp ature.

(*Furth oe the 12 h rFrequency is *nsidered a q~uate in vie o ind tin n h orol room, nl h lrm, o~alert the-mbrator to a omliebdtmprtr dto.Tis- Rmay be satisfied by use e Ice Bed Temperature Monitoring System.

SR 3.6.12.2 This SR ensures that initial ice fill and any subsequent ice additions meet the boron concentration and pH requirements of SR 3.6.12.7. The SR is modified by a NOTE that allows the chemical analysis to be performed on either the liquid or resulting ice of each sodium tetraborate solution prepared. If ice is obtained from offsite sources, then chemical analysis data must be obtained for the ice supplied.

SR 3.6.12.3 This SR ensures that the air/steam flow channels through the ice bed have not accumulated ice blockage that exceeds 15 percent of the total flow area through the ice bed region. The allowable 15 percent. buildup of ice is based on the analysis of the sub-compartment response to a design basis LOCA with partial blockage of the ice condenser flow channels.

The analysis did not perform detailed flow area modeling, but rather lumped the ice condenser bays into six sections ranging from 2.75 bays.

to 6.5 bays. Individual bays are acceptable with greater than 15 percent blockage, as long as 15 percent blockage is not exceeded for any analysis section.

To provide a 95 percent confidence that flow blockage does not exceed the allowed 15 percent, the visual inspection must be made for at least 54 (33 percent) of the 162 flow channels per ice condenser bay. The visual inspection of the ice bed flow channels is to inspect the flow area, by looking down form the top of the ice bed, and where view is achievable up from the bottom of the ice bed. Flow channels to be inspected are determined by random sample. As the most restrictive ice bed flow passage is found at a lattice frame elevation, the 15 percent blockage criteria only applies to "flow channels" that comprise the area:

McGuire Unit 1 and 2 B 3.6.12-5 Revision No

Ice Bed B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)

a. Between ice baskets, and

b. Past lattice frames and wall panels.

Due to a significantly larger flow area in the regions of the upper deck grating and the lower inlet plenum support structures and turning vanes, it would require a gross buildup of ice on there structures to obtain a degradation in air/steam flow. Therefore, these structures are excluded as part of a flow channel for application of the 15 percent blockage criteria. Plant and industry experience have shown that removal of ice from the excluded structures during the refueling outage is sufficient to ensure they remain operable throughout the operating cycle. Thus, removal of any gross ice buildup on the excluded structures is performed following outage maintenance activities.

Operating experience has demonstrated that the ice bed is the region that is the most flow restrictive, due to the normal presence of ice accumulation on lattice frames and wall panels. The flow area through the ice basket support platform is not a more restrictive flow area because it is easily accessible from the lower plenum and is maintained clear of ice accumulation. There is not a mechanistically credible method for ice to accumulate on the ice basket support platform during plant operation.

Plant and industry experience has shown that the vertical flow area I through the ice basket support platform remains clear of ice accumulation that could produce blockage. Normally only a glaze may develop or exist on the ice basket support platform which is not significant to blockage of flow area. Additionally, outage maintenance practices provided measures to clear the ice basket support platform following maintenance activities of any accumulation of ice that could block flow areas.

Activities that have a potential for significant degradation of flow channels

..should be limited to outage periods. Performance of this SR following completion of these maintenance activities assures the ice bed is in an acceptable condition for the duration of the operating cycle.

ph-asis that, Frost buildup or lose ice is not to be considered as flow channel blockage, mas frequenc, whereas attache.d ice is considered blockage of a flow channel. Frost is m9Rthrequncy, the solid form of water that is loosely adherent, and can be brushed off iefý of ie ons_ . tal/with the open hand.

will not co romise _ A Ice mass determination methodology is designed to verify the total as-found (pre-maintenance) mass of ice in the ice bed, and the appropriate distribution of that mass, using a random sampling of individual baskets.

The random sample will include at least 30 baskets from each of three McGuire Unit 1 and 2 B 3.6.12-6 Revision NCO

Ice Bed

/lý-10:0@//AIv4< 0/N,/ -T/-1/5 AB 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued) defined Radial Zones (at least 90 baskets total). Radial Zone A consists of baskets located in rows 8, and 9 (innermost rows adjacent to the Crane Wall), Radial Zone B consists of baskets located in rows 4, 5, 6, and 7 (middle rows of the ice bed), and Radial Zone C consists of baskets located in rows 1, 2, and 3 (outermost rows adjacent to the Containment Vessel).

The Radial Zones chosen include the row groupings nearest the inside and outside walls of the ice bed and the middle rows of the ice bed.

These groupings facilitate the statistical sampling plan by creating sub-populations of ice baskets that have similar mean mass and sublimation characteristics.

Methodology for determining sample ice basket mass will be either by direct lifting or by alternative techniques. Any method chosen will include procedural allowances for the accuracy of the method used. The number of sample baskets in any Radial Zone may be increased once by adding 20 or more randomly selected baskets to verify the total mass of that Radial Zone.

In the event the mass of a selected basket in a sample population (initial or expanded) cannot be determined by any available means (e.g., due to surface ice accumulation or obstruction), a randomly selected representative alternate basket may be used to replace the original selection in that sample population. If employed, the representative alternate must meet the following criteria:

ai.* Alternate selection must be from the same bay-Zone (i.e., same bay, same Radial Zone) as the original sele(tron, and

b. Alternate selection cannot. be a repeated selection (original:or.

alternate) in the current Surveillance, and cannot have been used as an analyzed alternate selection in the three most recent Surveillances.

The complete basis for the methodology used in establishing the 95%

confidence level in the total ice bed mass is documented in Ref. 5.

The total ice mass and individual Radial Zone ice mass requirements defined in this Surveillance, and the minimum ice mass per basket requirement defined by SR 3.6.12.5, are the minimum requirements for McGuire Unit 1 and 2 B 3.6.12-7 Revision No{

Ice Bed B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)

OPERABILITY. Additional ice mass beyond the SRs is maintained to address sublimation. This sublimation allowance is generally applied to baskets in each Radial Zone, as appropriate, at the beginning of an operating cycle to ensure sufficient ice is available at the end of the operating cycle for the ice condenser to perform its intended design function.

The Frequea y of 18 monthz das based on iceto ge tests, and the thypi caS*R imation allo*erae maintained anyice massoover an ove er th eiinimum satqir ice mlag i m nin theo a l assumed thteRi wni bsety Z analysesg. a apdngeand

~maintaine. he minimu ass and distributn requirements in t e ice be d ae

.. * */.---/SR 3.6.12.5 i siniianl affqectingthe conain ent prsue r espo containment~

paprsuedrnaDBi

~

anVerifying that each selectedi ~sample ~ basket~ from no e seRtef.contains SR 3.6.12.4

~nssinianl ) at least 600 bs of ice in the as-found (proe-maintenance) condition ensures affctdb that a significant cnainiengn les t localized degraded ehan60ls mass ofiereqires condition is avoided.

tappropiatelhntrn the T8Sýr~ieuny This SR establishes a per basket limit to ensure'any ice mass was~~~~~~ bsoicStgtets, degradation is consistent mindiviua iebstmass with the th anddsrbov initial conditions of the requiredmaenty analysiseDBA medanfre by not significantly affecting the containment pressure response. Ref. 5 provides insights through sensitivity runs that demonstrate that the containment peak pressure during a DBA is not significantly affected by theice mass inra lame nocalizedmregiongot ocree ofindivdual baskelowt ineib a m d w iredsaetanlyss require eah.te saethe Radial Zone.and total ice.ys. .

" .. mass requirements containing less than.600 of SRIbs 3.6.12M4 are satisfied. Any basket identified as of ice requires.appropriately entering the T8

• .. ~Required Action for~an inoperable ice bed due.to the potential.th at it may"

... -.*.. . '. .. . .. . * *representi a. significant condition adverse to quality. *.. .. *.-

Th reqno 18 rPmS As documented in Ref. 5, maintenance practices actively manage was bas on ice st g~e tests, ')individual ice basket mass above the required safety analysis mean for and typical s imation ) each Radial Zone. Specifically, each basket is serviced to keep its ice a wnc m anedinthe *mass*. above 725 Ibs for Radial Zone A, 1043 Ibs for Radial Zone B, and ic ms e adabove th 1043 Ibs for Radial Zone C. If a basket sublimates below the safety mii ic/asase n analysis mean value, this instance is identified within the plant's the safety analyses. perating ,_ .

dmaintenance xperience has orret._ action program, including evaluating maintenance practices to I tht,*

veifie he 8*enh identify the cause and correct~any deficiencies. These maintenance Frequenc , the minin*fmas) practices provide defense in depth beyond compliance with the ice bed and di ributio eirem~ents /"surveillance requirements by limiting the occurrence of individual baskets in th ice bed a* maintained./ with ice mass less than the required safety analysis mean.

McGuire Unit 1 and 2 B 3.6.12-8 Revision No.(

Ice Bed B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.12.6 This SR ensures that a representative sampling of accessible portions of ice baskets, which are relatively thin walled, perforated cylinders, have not been degraded by wear, cracks, corrosion, or other damage. The SR is designed around a full-length inspection of a sample of baskets, and is intended to monitor the effect of the ice condenser environment of ice baskets. The groupings defined in the SR (two baskets in each azimuthal third of the ice bed) ensure that the sampling of baskets is reasonably distribu 1c s for visual i3sp.e6..I of the sc'ud-rljrsetndness o f4e ice baskets *ased on en~ering judgme t an sies u acos ste ea f h ket walls rel

• _to" ktherlngt ice storage =ting, SR 3.6.12.7 Verifying the chemical composition of the stored ice ensures that the stored ice has a boron concentration > 1800 ppm and < 2330 ppm as sodium tetraborate and a high pH, > 9.0 and < 9.5 at 20°C, in order to meet the requirement for borated water when the melted ice is used in the ECCS recirculation mode of operation. Additionally, the minimum boron concentration setpoint is used to assure reactor subcriticality in a post LOCA environment, while the maximum boron concentration is used as a bounding value in the hot leg switchover timing calculation (Ref. 4). This is accomplished by obtaining at least 24 ice samples. Each sample is taken approximately one foot from the top of the ice of each randomly selected ice basket in each ice condenser bay. The SR is modified by a NOTE that allows the boron concentration and pH valueiobtained from

-averaging the indiViduat samples analysis results to satisfy-the requirements of the SR. If either.the average boron concentration or" average pH value is outside their prescribed limit, then entry into ACTION Condition A is required. Sodium tetraborate has been proven effective in maintaining the boron content for long storage periods, and it also enhances the ability of the solution to remove and retain fission product iodine. The high pH is required to enhance the effectiveness of the ice and the melted ice in removing iodine from the containment atmosphere.

This pH range also minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS o rto.Te and Containment Spray System fluids in the recirculation mode of Frý.uency of 54 mordhs is intended to be consistent w the expected gth of three fue des, and was devel considering these fac F Long term =ic orage tests have d rmined that tjhý emica comr~ositi of the) stored ice is tremelv stable:

McGuire Unit 1 and 2 B 3.6.12-9 Revision N Z5'

Ice Bed B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)

/ b. There ea ~the oboron

~~ch normal operatinion of concent/r echanisms the storedthat i

sianadfficantly iiiii, pH remains

/ /ithin a 9.0-95rn hn oo crtions are above

\

• approximately12 pm; and

c. Operatin xperience has dem strated that meeting boron conce ration and pH requir ents has not been roblem.

REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50, Appendix K.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

4. UFSAR, Section 6.3.3.10.

5. Topical Report ICUG-001, Application of the Active Ice Mass Management Concept to the Ice Condenser Ice Mass Technical Specification, revision 2.

6. UFSAR, Table 18-1 and Section 18.2.14.

7. McGuire License Renewal Commitments MCS-1274.00-00-0016, Section 4.19, Ice Condenser Inspections.

McGuire Unit 1 and 2 B 3.6.12-10 Revision NC(P

Ice Condenser Doors B 3.6.13 BASES ACTIONS (continued) ice bed because of the large mass of ice involved. The 14 day Completion Time is based on long term ice storage tests that indicate that if the temperature is maintained below 27 0 F, there would not be a significant loss of ice from sublimation. If the maximum ice bed

.temperature is > 27 0 F at any time or ifthe doors are not closed and restored to OPERABLE status within 14 days, the situation reverts to Condition C and a Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is allowed to restore the inoperable door to OPERABLE status or enter into Required Actions D.1 and D.2. Ice bed temperature must be verified within the specified Frequency as augmented by the provisions of SR 3.0.2.

C.1 If Required Actions B.1 or B.2 are not met, the doors must be restored to OPERABLE status and closed positions within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Completion Time is based on the fact that, with the very large mass of ice involved, it would not be possible for the temperature to increase to the melting point and a significant amount of ice to melt in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.

D.1 and D.2 If the ice condenser doors cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply.. To achieve this status, the plant must-be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on

- operating experience, to reach the required plant conditions fromifull power conditions in an orderly manner and without challenging plant -

systems'.

SURVEILLANCE SR 3.6.13.1 REQUIREMENTS Verifying, by means of the Inlet Door Position Monitoring System, that the inlet doors are in their closed positions mak the operator aware of an inrning(*e sn h ~ e at of r~sone oh eor a chmore

• door3J s Rv The Fr31uency re a wth ýo

-atts t hof -ior-McGuire Units 1 and 2 B 3.6.13-5 Revision NoO "

Ice Condenser Doors B 3.6.13 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.13.2 Verifying, by visual inspection, that each intermediate deck door is closed and not impaired by ice, frost, or debris provides assurance that the intermediate deck doors (which form the floor of the upper plenum where frequent maintenance on the ice bed is performed) have not been left open or obstructed. In determining if a door is impaired by ice, the frost accumulation on the doors, joints, and hinges are to be considered in coiunction with the lifting force limits of SR 3.6.13.7..*_e -Frequencyof 7day s, 'absed~p eng ine~erngugmt ataeinto 0codeato in oh e nt m d ce o de, r

" freque ncyfenr

- rýs a the s, ato deck, the *de required f significant fr dst buildup, the prob q~ity that a A will occur.."

SR 3.6.13.3 Verifying, by visual inspection, that the top deck doors are in place and not obstructed provides assurance that the doors are performing their function of keeping warm air out of the ice condenser during normal operation, and would not be obstructed if called upon to open in response to a DBA. Tfre-n-cy o---92 days is ba sedn engineerng j dgriien ,

whic consid red such facto as the fol lo :

a. T relative ina essibility and ck of traffic in th icinity of the doors make. unlikely that a or would be in ertentIy left op SR 3.6.13.4 Verifying, by visual inspection, that the ice condenser inlet doors are not impaired by ice, frost, or debris provides free to open in the event of a 1],16 assurance 1:iBA.i ni, thethat the doors Frequency pf-are 8 mon s is based on door design, which s not allow w r condeneion to freezend operating eeience, which icates tha the i t doors very rely fail to me eir SR accept ce criteria.

B ause of high diation in the inity of the inlet oors during er peration, thi urveillance is ormally perform during a sh own. .

McGuire Units 1 and 2 B 3.6.13-6 Revision N

Ice Condenser Doors B 3.6.13 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.13.5 Verifying the opening torque of the inlet doors provides assurance that no doors have become stuck in the closed position. The value of 675 in-lb is based on the design opening pressure on the doors of 1.0 lb/ft2. or this renc of1 mon s'is ased on the ssive na reof the closin echa m (i.e., ce adjust there arno known actors that wo chan the settil, except ssibly a ildup of ice ce buildu s likel owever, cause of e door d ign, which, ,es not al wate ondensati tofreez . Operati experiencrndicates at the doors us ly meet t Ir SR acccl tance cnrter . Becaus of high dir tio in e iint eile rsdwn er oper iton, this-S.urrveillrace is nor nnny prorm~e during a S ýItown."

SR 3.6.13.6 The torque test Surveillance ensures that the inlet doors have not developed excessive friction and that the return springs are producing a door return torque within limits. The torque test consists of the following:

1. Verify that the torque, T(OPEN), required to cause opening motion at the 400 open position is < 195 in-lb; v~-/T-A

2. Verify that the torque, T(CLOSE), required to hold the door stationary (i.e., keep it from closing) at the 400 open position is

_ 78 in-lb but < 250.6 in-Ib; and

,3. Calculate the frictional torque, T(FRICT) 0.5{ T(OPEN)-

- T(CLOSE)}, and verify that the T(FRICT) is > - 40 in-lb but_! + "

40 in-lb.

The purpose of the friction and return torque Specifications is to ensure that, in the event of a small break LOCA or SLB, all of the 24 door pairs open uniformly. This assures that, during the initial blowdown phase, the steam and water mixture entering the lower compartment does not pass through part of the ice condenser, depleting the ice there, while b assing theiceinotherbays. e nFrem ueCnY o. °onths is bas doo the pasv ntro e closin mechanis i~e., o~nce ad* sted, there Xe n own f tors that 2 change e setting, ex pt possiblya-uildup

~fie cbuildup is likely, ho ver, becaus f the door sign, whic oes not all water coo ensation to freeze). Oper ing

__z McGuire Units 1 and 2 B 3.6.13-7 Revision No'e Z

Ice Condenser Doors B 3.6.13 BASES SURVEILLANCE REQUIREMENTS (continued) exe*peidiasthtte oosv rarely fail meet aa acc a c B ca e of high r a iation in th e icinity of I linlet .,

,/,6r

/ ur owerop ao,n this. *;rveillance i nhormally pejrormed

• SR 3.6.13.7 Verifying the OPERABILITY of the intermediate deck doors provides assurance that the intermediate deck doors are free to open in the event of a DBA. The verification consists of visually inspecting the intermediate doors for structural deterioration, verifying free movement of the vent assemblies, and ascertaining free movement of each door when lifted with the applicable force shown below:

Door Lifting Force

a. Adjacent to crane wall < 37.4 lb

b. Paired with door adjacent to crane wall _ 33.8 lb

c. Adjacent to containment wall _ 31.8 lb

d. Paired with door adjacent to containment < 31.0 lb wall REFERENCES 1. UFSAR, Chapter 6.

2. 10 CFR 50, Appendix K.

3. 10CFR 50.36, Technical Specifications, (c)(2)(ii).

4. DPC-1201.17-00-0006 "Design and Licensing Basis for Ice Condenser Lower Inlet Doors Technical Specification Surveillance Requirements, 4w0 Opening, Closing and Frictional Torques".

5. MCS-1558.NF-00-0001 "Design Basis Specification for the NF System".

McGuire Units 1 and 2 B 3.6.13-8 Revision No.

Divider Barrier Integrity B 3.6.14 BASES SURVEILLANCE REQUIREMENTS (continued) inspection cannot be made when the door or hatch is closed. Therefore, SR 3.6.14.2 is required for each door or hatch that has been opened, prior Iong to the final closure. Some doors and hatches

.9periods may innot

?516refl-ials thebesealsmur-o ened for beopene nd insp eted at I~a' once evp 1year to*vde

• • d ote~~lnt of deg rade~i assur e thata ,seal m rial has n aged to the nt of degradeA-pe rmanc . The Fre enc ears is base n the known esilien( of the m rials used f seals, the fa that the ope gs have not n opene to cause w r), and opera*g experienc hat confirms t the seal.4 spected at is Frequency have been found to e SR 3.6.14,3 Verification, by visual inspection, after each opening of a personnel access door or equipment hatch that it has been closed makes the operator aware of the importance of closing it and thereby provides additional assurance that divider barrier integrity is maintained while in applicable MODES.

SR 3.6.14.4 Conducting periodic physical property tests on divider barrier seal test coupons provides assurance that the seal material has not degraded in the containment environment, including the effects of irradiation with the reactor at power. The required tests include a tensile,strength test¢,,, h Frequen of 18 ths was eveloped c sidering such factors as 0e know esilien of.the se material us the.inacc ibility.of th seaI a abs.en of traffi& heir vicinity, ndthe uni onditions n Idedto.

perfor e SR. 0 rating exper' nce has shn thatthes compone usu y pass the urveillance en performed at the 18 onth F eouencv. erefore. the reauencv w concluded t be acc table SR 3.6.14.5 McGuire Units 1 and 2 B 3.6.14-5 Revision Ný

Divider Barrier Integrity B 3.6.14 BASES SURVEILLANCE REQUIREMENTS (continued) fFreq cy. The oe, the equ ency*"a fre a reliab concludedtobeaetbe standpirt--" -

REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.6.14-6 Revision N096

Containment Recirculation Drains B 3.6.15 BASES . -

APPLICABILITY (continued)

As such, the containment recirculation drains are not required to be OPERABLE in these MODES.

ACTIONS A.1 If one ice condenser floor drain is inoperable, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment," which requires that containment be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

B._1 If one refueling canal drain is inoperable, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of LCO 3.6.1, which requires that containment be restored to OPERABLE status in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C.1 and C.2 If the affected-drain(s).cannot be restored to OPERABLE status within the required Completion.Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this.status, the plant must be-

.,brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.15.1 and SR 3.6.15.2 REQUIREMENTS Verifying the OPERABILITY of the refueling canal drains ensures that they will be able to perform their functions in the event of a DBA. SR 3.6.15.1 confirms that the refueling canal drain valves have been locked open and that the drains are clear of any obstructions that could impair their functioning. In addition to debris near the drains, SR 3.6.15.2 requires attention be given to any debris that is located where it could be McGuire Units 1 and 2 B 3.6.15-3 Revision NY_!ý

Containment Recirculation Drains B 3.6.15 BASES SURVEILLANCE REQUIREMENTS (continued) moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains. SR 3.6.15.1 must be performed before entering MODE 4 from MODE 5 after every filling of the canal to ensure that the valves have been locked open and that no debris that could impair the drains was deposited durn the ti e the canal was filled SR .6.15.2 i 11erformed e ery ays-r the uppereompartm ýt

-and re eling can areas. Th 2 day Fre"ency was eloped co deing s factors a e inacce ility of theains, the sence of traffic i e vicinity he drains dthdrains.

t d SR 3.6.15.3 Verifying the OPERABILITY of the ice condenser floor drains ensures that they will be able to perform their functions in the event of a DBA.

Inspecting the drain valve disk ensures that the valve is performing its function of sealing the drain line from warm air leakage into the ice condenser during normal operation, yet will open if melted ice fills the line condenser.YThe are not o th 18 moensures following a DBA.to Verifying drain water thatfr omthe the drain icelines ucted their readiness u B a t o, rs a s t h e in a s es ib ilit y )

'v a d vlo p e d c o n s i d r ice cdner -y of~~~uige the drj un oeeacent f~~~~o cause erform ed atan 18 month, requency. operaion, this eillance is dr th the S reqen the of *an dinit r in lation of hig h " . .. ...

ly done du nir as ud

• g ....

Suv

• a c i n r REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.6.15-4 Revision No

Reactor Building B 3.6.16 BASES APPLICABILITY Maintaining reactor building OPERABILITY prevents leakage of radioactive material from the reactor building. Radioactive material may enter the reactor building from the containment following a LOCA.

Therefore, reactor building OPERABILITY is required in MODES 1, 2, 3, and 4 when a steam line break, LOCA, or rod ejection accident could release radioactive material to the containment atmosphere.

In MODES 5 and 6, the probability and consequences of these events are low due to the Reactor Coolant System temperature and pressure limitations in these MODES. Therefore, reactor building OPERABILITY is not required in MODE 5 or 6.

ACTIONS A. 1 In the event reactor building OPERABILITY is not maintained, reactor building OPERABILITY must be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Twenty-four hours is a reasonable Completion Time considering the limited leakage design of containment and the low probability of a Design Basis Accident occurring during this time period.

B.1 and B.2 If the reactor building cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.16.1 REQUIREMENTS Maintaining reactor building OPERABILITY requires maintaining the door in usedeachforaccess normal opening closed, transit entry andlexcept when ex_*tAh the access 311* ;reue'opening is being

• of ths.! -"**-

a*gisd ng ering judg --ft and is c idered ade ate in v' of th(

,0*r indica/hns of doorýfatus that a available.

McGuire Units 1 and 2 B 3.6.16-2 Revision No.,

Reactor Building B 3.6.16 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.16.2 The ability of a AVS train to produce the required negative pressure within the required times provides assurance that the building is adequately sealed. The negative pressure prevents leakage from the building, since outside air will be drawn in by the low pressure. The negative pressure must be established within the time limit to ensure that no significant quantity of radioactive material leaks from the reactor building prior to developing the negative pressure.

The AVS-frains are-tested evervA* months opra STAGGEFýD TEST 10 CFR 50.36, Technical Specifications, McGuire Units 1 and 2 B 3.6.16-3 Revision NoO

SG PORVs B 3.7.4 BASES ACTIONS A.1 With one required SG PORV line inoperable, action must betaken to restore OPERABLE status within 7 days. The 7 day Completion Time allows for the redundant capability afforded by the remaining OPERABLE SG PORV lines, a nonsafety grade backup in the Steam Dump System, and MSSVs.

B.1 With two or more SG PORV lines inoperable, action must be taken to restore all but one SG PORV line to OPERABLE status. Since the block valve can be closed to isolate an SG PORV, some repairs may be possible with the unit at power. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is reasonable to repair inoperable SG PORV lines, based on the availability of the Steam Dump System and MSSVs, and the low probability of an event occurring during this period that would require the SG PORV lines.

C.1 and C.2 If the SG PORV lines cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4, without reliance upon steam generator for heat removal, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonablej based on operating experience, to reach the required unit conditions from full power conditions in an orderly.

",,man,,ner a.nd without challenging unit systems.

SURVEILLANCE SR 3.7.4.1 .

REQUIREMENTS To perform a controlled cooldown of the RCS, the SG PORVs must be able to be opened manually using the handwheel and throttled through their full range. This SR ensures that the SG PORVs are tested through a full cycle at least once per fuel cycle. Performance of inservice testing or use of an SG PORV during a unit cooldown may satisfy this requirement. pr Kng experin--ce has shoiw t-hat these e)mponents/-

usuallass th urveillan when perfred at the onth Fpouency. he Frequ cy is accep le from a rlability stan oint.

McGuire Units 1 and 2 B 3.7.4-3 Revision No

SG PORVs B 3.7.4 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.4.2 Th'e function of the block valve is to isolate a failed open SG PORV.

Cycling the block valve manually using the handwheel both closed and 7-- open demonstrates its capability to perform this function. Performance of inservice testing or use of the block valve during unit cooldown ma satisfy this requirement. erien as shown that these compo nts usally pass th urveillanc hen perfo d at th,ý, '

18§ronth F quency. T Frequency s acceptabtefrom a rp ability REFERENCES 1. UFSAR, Section 10.3.

2. UFSAR, Chapter 15.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.7.4-4 Revision NCO

BASES Ic//,A 61c' 0/V ,B FWSystem A 3.7.5 ACTIONS (continued) accordance with the Note that modifies the LCO. Althoughnot required, the unit may continue to cool down and initiate RHR.

D. 1 If all three AFW trains are inoperable in MODE 1, 2, or 3, the unit is in a seriously degraded condition with no safety related means for conducting a cooldown, and only limited means for conducting a cooldown with nonsafety related equipment. In such a condition, the unit should not be perturbed by any action, including a power change, that might result in a trip. The seriousness of this condition requires that action be started immediately to restore one AFW train to OPERABLE status.

Required Action D.1 is modified by a Note indicating that all required MODE changes or power reductions are suspended until one AFW train is restored to OPERABLE status. In this case, LCO 3.0.3 is not applicable because it could force the unit into a less safe condition.

E. 1 In MODE 4, either the reactor coolant pumps or the RHR loops can be used to provide forced circulation. This is addressed in LCO 3.4.6, "RCS Loops-MODE 4." With one required AFW train with a motor driven pump inoperable, action must be taken to immediately restore the inoperable train to OPERABLE status. The immediate Completion Time is consistent with LCO 3.4.6.

SURVEILLANCE SR 3.7,5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the AFW System water and steam supply flow paths provides assurance that the proper flow paths will exist for AFW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. The SR is also modified by a note that excludes automatic valves when THERMAL POWER is <

10% RTP. Some automatic valves may be in a throttled position to support low power operation.

McGuire Units 1 and 2 B 3.7.5-6 Revision No.Aý

BASES AFW System B 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

The 31 ay Frequcy is based engineerinýudgment, consiste*

wit ,te proceral controls verning val*operation d ensur5-'l

\rrect val positions.

Verifying that each AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref 3). Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.

Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. Performance of inservice testing discussed in the ASME OM Code (Ref. 3) (only required at 3 month intervals) satisfies this requirement.

The Frequency for this SR is in accordance with the Inservice Testing Program.

This SR is modified by a Note indicating that the SR should be deferred until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test. The test should be conducted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the steam oressure exceedino 900 psig.,

SR 3.7.5.3 .

This SR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.

This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

SThoe 18 mon Ir1ncy I base on the nee to perform this Su i lance und the conditio that apply ring a unit outaaand the ential for a nplanned nsient if the urveillance wererformed with the re tor at powe .Te 18 mo Frequency is -ceptable sed I on oper ing experie e an the d ign reliability ofte equipm nt.

This SR is modified by a Note that states the SR is not required in MODE 4. In MODE 4, the required AFW train may already be aligned and operating.

McGuire Units 1 and 2 B 3.7.5-7 Revision Noo

BASES AFW System B 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.5.4 This SR verifies that the AFW pumps will start in the event of any accident or transient that generates an ESFAS by demonstrating that each AFW pump starts automatically on an actual or simulated actuation signal in MODES 1, 2, and 3. In MODE 4, the required pump may already be operating and the autostart function is not reqeise The 18 m sth InMOque 4 is rased eunee Ea pemfay I

/ con ions that aapy during a uit outage an l'e potential frudan r the rveillance

( plplanned t sient if the S:rveillance we perform ed th the reac* a

./power.

SThis SR is modified by two Notes. Note I indicates that the SR can be deferred until suitable test conditions are established. This deferral is

, required because there is insufficient steam pressure to perform the test.

The test should be conducted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the steam pressure exceeding 900 psig. Note 2 states that the SR is not required in MODE 4.

In MODE 4, the required pump may already be operating and the autostart function is not required. In MODE 4, the heat removal requirements would be less providing more time for operator action to manually start the required AFW pump ifit were not in operation.

REFERENCES 1. UFSAR, Section 10.4.7.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

3. ASME Code for Operation and Maintenance of Nuclear Power Plants. I McGuire Units 1 and 2 B 3.7.5-8 Revision No

CCW System B 3.7.6 BASES ACTIONS A.1 Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4,"

be entered if an inoperable CCW train results in an inoperable RHR loop.

This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.

If one CCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining OPERABLE CCW train is adequate to perform the heat removal function.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period.

B.1 and B.2 If the CCW train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE .in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.6.1 REQUIREMENTS This SR is modified by a Note indicating that the isolation of the CCW flow to individual components may render those components inoperable but does not affect the OPERABILITY of the CCW System.

Verifying the correct alignment for manual, power operated, and automatic valves in the CCW flow path provides assurance that the proper flow paths exist for CCW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the 5 correct position.

Mcueday 1rig 2 FUs B 37-R judgma isionN yA,*f~the pr eduralc sgoennave opera>t7f, atn~diens~xes

..... dv ositio McGuire Units 1 and 2 B 3.7.6-3 Revision NoO*

CCW System B 3.7.6 BASES SURVEILLANCE REQUIREMENTS (continued)

ýSR 3.7.6.2 This SR verifies proper automatic operation of the CCW valves on an actual or simulated actuation safety injection signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation. This Surveillance is not required for valves that are locked, sealed, or oth ise secured in the re uired position under administrative controls./'he 18 m eth Freq uen*, is based on the n to perfo this Surv ance under e condition hat apply durin unit out e and the ential for a nplanned tr sient ifthe S eillance re perform with the re or at power. perating experienc as shown t these com onents usual pass the Surveillc hen p d at th 8 ont Fr uency. Ther ore, the

,Fre ency is acce able from a r iability stan oint.

SR 3.7.6.3 This SR verifies proper automatic operation of the CCW pumps on an actual or simulated actuation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testin, durinqg normaloeain f mnth ~euenyi based on the ne"-

to'perform* s Survei a nder the c ditions that a y during a u outa the potent* for an unpl ned transient/ e Surveilla e wer erformed wi the reactor power. Oper ng experien has own that the component sually pass t urveillance hen performed he 18 month requency. T refore,.the Frequency is REFERENCES 1. UFSAR, Section 9.2..

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.7.6-4 Revision N1W

NSWS B 3.7.7 BASES ACTIONS (continued) indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources-Operating," should be entered if an inoperable NSWS train results in an inoperable emergency diesel generator. The second Note indicates that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," should be entered if an inoperable NSWS train results in an inoperable decay heat removal train.

This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period.

B.1 and B.2 If the NSWS train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.7.1 REQUIREMENTS This SR is modified by a Note indicating that the isolation of the NSWS components or systems may render those components inoperable, but does not affect the OPERABILITY of the NSWS.

Verifying the correct alignment for manual, power operated, and automatic valves in the NSWS flow path provides assurance that the proper flow paths exist for NSWS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position- prior to being locked, sealed, or secured. This SR does not require any testing or valve manipulation;

/ -£ rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

ce iray 1 an 2 B 37-nt, FUreneecme is co N sten

• k (' wolre ppua ~ osg ~ ~ g vl e rajtion,an McGuire Units I and 2 B, 3.7 .7-4 Revision N.O4*/

NSWS B 3.7.7 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.7.2 This SR verifies proper automatic operation of the NSWS valves on an actual or simulated actuation safety injection signal. The NSWS is a normally operating system that cannot be fully actuated as part of normal testing. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required osition under..

administrative controls.ns based n-the need9.2 5 10Chis Surv 5, Aance unA the c"S ns that o p rturin d au outa and e the .ential fon unplanntransient ife Surveilla 6 e penforiLwith the actor at poer. Opera tiog experience as hown th nhese cof 1 ass tie uerv*ilance watenSt nents usuey prFa lur ese, the FreqM-ency isS cu Uaccittable fro1ad2reliability sRndpoiei A.*/

• ._..S R 3.7.7.3

"* This SR verifies proper automatic operation of the NSWS pumps on an

~actual or simulated actuation signal. The NSWS is a normally operating system that cannot be fully actuated as part of normal testing du ring normal operat~ion. heA month~requency*-;rbasedoF n eh,

• /-*rfpeornm"tis Survei Uence und5eAhe condi 'ns tat appl pduring a unit.

/outage~nd the tential foa -n unplan/0' transient if e Surveilla0 r.h m ths c oet ual as h u illa nce v/en *J

.. ... *":.prfo e'd at tth'el8 month Fr euency. The'fore,'the Freuency is _/

REFERENCES 1. UFSAR, Section 9.2.

2. UFSAR, Section 6.2.

3. UFSAR, Section 5.4.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

5. 10 CFR 50, Appendix A, GDC 5, "Sharing of Structures, Systems, and Components".

6. Generic Letter 91-13, "Request for Information related to the resolution of Generic Issue 130, Essential Service Water System Failures at Multi-Unit Sites".

McGuire Units 1 and 2 B 3.7.7-5 Revision No(A*

SNSWP B 3.7.8 BASES APPLICABLE SAFETY ANALYSES (continued) decay heat, and worst case single active failure (e.g., single failure of a manmade structure). The SNSWP is designed in accordance with Regulatory Guide 1.27 (Ref. 2), which requires a 30 day supply of cooling water in the SNSWP.

The SNSWP satisfies Criterion 3 of 10 CFR 50.36 (Ref. 3).

LCO The SNSWP is required to be OPERABLE and is considered OPERABLE if it contains a sufficient volume of water at or below the maximum temperature that would allow the NSWS to operate for at least 30 days following the design basis LOCA without the loss of net positive suction head (NPSH), and without exceeding the maximum design temperature of the equipment served by the NSWS. To meet this condition, the SNSWP temperature should not exceed 82 0 F at 722 ft mean sea level and the level should not fall below 739.5 ft mean sea level during normal unit operation.

APPLICABILITY In MODES 1, 2, 3, and 4, the SNSWP is required to support the OPERABILITY of the equipment serviced by the SNSWP and required to be OPERABLE in these MODES.

In MODE 5 or 6, the requirements of the SNSWP are determined by the systems it supports.

ACTIONS A.1 If the SNSWP is inoperable the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.8.1 REQUIREMENTS This SR verifies that adequate long term (30 day) cooling can be maintained. The specified level also ensures that sufficient NPSH is McGuire Units 1 and 2 B 3.7.8-2 Revision NO

SNSWP B 3.7.8 BASES SURVEILLANCE REQUIREMENTS (continued) duringeppliýble MOlES./ t'his SR verifies that the SNSWP water level is > 739.5 ft mean sea level.

SR 3.7.8.2 This SR verifies that the NSWS is available to cool the CCW System to at least its maximum design temperature with the maximum accident or normal design heat loads for 30 days following a Design Basis Accident.

/ The*24-hour Fruency ised onerating exp:tence d ndig....tnding o-tite pararmet~er variatieriI during teapp lica bleMO _S./This SR veifies that the average water temperature of the SNSWP is < 82 0 F at an elevation of 722 ft. The SR is modified by a Note that states the Surveillance is only required to be performed during the months of July, August, and September. During other months, the ambient temperature is below the surveillance limit.

SR 3.7.8.3 This SR verifies dam integrity by inspection to detect degradation, REFERENCES 1. UFSAR, Section. 9.2.

2. Regulatory Guide 1.27.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units I and 2 B 3.7.8-3 Revision N1ý5

CRAVS B 3.7.9 BASES ACTIONS (Continued) radiation doses are within 10 CFR 100 limits during a DBA LOCA under these conditions.

SURVEILLANCE SR 3.7.9.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing each train once every month provides an adequate check of this system. Monthly heater operations

.dry out any moisture accumulated in the charcoal from humidity in the ambient air. Systems with heaters must be operated from the control room for _>10 continuous hours with the heaters energized and flow through the HEPA filters and charcoal adsorbers. Inoperable heaters are addressed by Required Actions G.1 and G.2. The inoperability of heaters between required performances of this surveillance does not affect of each CR rAVS trainhe 3ay Frequen. a-the;ehdbility of theqipment and the o train redundkfi*cy.

SR 3.7.9.2 This SR verifies that the required CRAVS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The 3 CRAVS filtertests are in accordance with Regulatory Guide 1.52 (Ref. 4).

The VFTP includes testing the performance of the HEPAfilter, charcoal adsorber efficiency, minimum flow rate, and the physical properties of the activated charcoal. Specific test Frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.9.3 This SR verifies that each CRAVS train starts and operates with flow through the HEPA filters smuaeacuation signaIandTe charcoaluFadsorbers y of 18 on an nt actual s baeor n--

id, oeating*qexperience.

SR 3.7.9.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.

McGuire Units 1 and 2 B 3.7.9-7 Revision N6.)

CRACWS B 3.7.10 BASES SURVEILLANCE SR 3.7.10.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to maintain the temperature in the control room at or below ho-.,frK'Fequept:y is apperopriat isnc-LmI" * (rsijigrificatd da of theFACi8sslwni t

.e pet eo6v e r t t im e po-bod.j REFERENCES 1. UFSAR, Section 6.4.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.7.10-4 Revision N

ABFVES B 3.7.11 BASES ACTIONS (continued) based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.11.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not severe, testing each train once a month provides an adequate check on this system. Systems without heaters need only be operated from the control room for

> 15 minutes with flow through the HEPA filters and chlaoal adsorbers to demonstrate the function of the systen_3/The-3*day)

FrSe* baseýlethe *n reli of eup* .

SR 3.7.11.2 This SR verifies that the required ABFVES testing is performed in accordance with the Ventilation Filter Testing Program (VFTP).

The ABFVES filter tests are in accordance with Reference 4. The EIz -T- VFTP includes testing HEPA filter performance, carbon adsorbers efficiency, minimum system flow rate, and the physical properties of the carbon (general use and following specific operations).

Specific test Frequencies and additional information are discussed in detail in the VFTP. '

SR 3.7.11.3 This SR. verifies that.ABFVES starts.and operates with flow through the HEPA filters and charcoal adsorbers on an actual or

~simulated actuation signaJ/'e8 Mon -- requpee iýýs c

• ** 7.1aol.cifie";q Reguyaty Gui .52,v)e.jj This SR verifies the integrity of the ECCS pump room enclosure.

The ability of the ECCS pump room to maintain a negative pressure, with respect to potentially uncontaminated adjacent areas, is periodically tested to verify proper functioning of the ABFVES. During the post accident mode of operation, the ABFVES is designed to maintain a slight negative pressure in the ECCS pump room area, with respect to adjacent areas, to prevent unfiltered LEAKAGE. The ABFVES is designed to maintain a

< -0.125 inches water gauge relative to atmospheric pressure.

McGuire Units 1 and 2 B 3.7.11-5 Revision No.4'

ABFVES B 3.7.11 BASES SURVEILLANCE REQUIREMENTS (continued)

This SR is required to be performed for each fan combination (1A and 1B, 2A and 2B, 1A and 2A, 1B and 2B) described in the LCO Bases he of 18 onthsj>,consis wi e 9 jifan , ýXý . 7).

An 18 month Frequency on a STAGGERED TEST BASIS is consistent with that specified in Reference 6.

REFERENCES 1. UFSAR, Section 9.4.

2. UFSAR, Section 12.2.

3. UFSAR, Section 15.6.5.

4. 10 CFR 100.11.

5. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

6. Regulatory Guide 1.52 (Rev. 2).

7. NUREG-0800, Section 6.5.1, Rev. 2, July 1981.

8. UFSAR, Table 9-38.

McGuire Units 1 and 2 B 3.7.11-6 Revision No 9K

FHVES B 3.7.12 BASES APPLICABILITY The FHVES is required to be OPERABLE and in operation in filtered mode during the following evolutions:

1. Movement of irradiated fuel in the fuel handling building;

2. Movement of loads in excess of 100 lbs. over irradiated fuel stored in the spent fuel pool; and

3. Movement of a loaded dry storage cask in the fuel handling building with the 125 ton overhead crane.

ACTIONS A.1 With the FHVES inoperable, action must be taken to immediately suspend the movement of irradiated fuel in the fuel handling building.

This does not preclude movement of a fuel assembly to a safe position.

his action ensures a release to the environment will be within the limits of 10 CFR 100 limits (Ref. 5), if a fuel handling accident were to occur.

Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

SURVEILLANCE SR 3.7.12.1 REQUIREMENTS With the FHVES in service, a periodic monitoring of the system for operation is required to ensure that the system functions proper h e2 is fficie"to ensurepoper opera, on through FE15A and <harcoafilters~afd is baseeon the knpd reliabilitv.4rf i.....

SR 3.7.12;2 Systems should be checked periodically to ensure that they function properly. As the environmental and normal operating conditions on this system are not severe, testing prior to movement of irradiated fuel will ensure an adequate check on this system.

Systems without heaters need only be operated for _>15 minutes to demonstrate the function of the system.

McGuire Units I and 2 B 3.7.12-3 Revision Ncýg

FHVES B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.12.3 This SR verifies that the required FHVES testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The FHVES filter tests are in accordance with Regulatory Guide 1.52 (Ref. 6).

The VFTP includes testing HEPA filter performance, carbon adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.12.4 This SR verifies the integrity of the fuel building enclosure. The ability of the fuel building to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically verified by ensuring the exhaust flow rate of the FHVES is 8000 cfm greater than the supply flow rate. During the post accident mode of operation, the FHVES is designed to maintain a slight negative pressure in the fuel building, to prevent unfiltered LEAKAGE.

The Fseciuency.if 18 morlt's is consisent with-th'e guidancel-rovid ein An month Fer-ency is-,

SR 3.7.12.5 Operating the FHVES filter bypass damper, is necessary to ensure that the system functions properly. The OPERABILITY of the FHVES filter bypass dam er is verified if it can be manually clos d AndJ.&--rno Referenr ._ P ency is.onsistent q-Fre~q McGuire Units 1 and 2 B 3.7.12-4 Revision Nou I/-

Spent Fuel Pool Water Level B 3.7.13 BASES APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel pool, since the potential for a release of fission products exists.

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the spent fuel pool water level is lower than the required level, the movement of irradiated fuel assemblies in the spent fuel pool is immediately suspended to a safe position. This action effectively precludes the occurrence of a fuel handling accident. This does not preclude movement of a fuel assembly to a safe position.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.13.1 REQUIREMENTS This SR verifies sufficient spent fuel pool water is available in the event of a fuel handling accident. The water-level in the spentfuel pool must be-checked eriodicall The 7 ay Frequncy is appr rate becaue the Vol n the oo is no ly stab . Water le5 changes e contr9d pant c n areeptable b d on oper ng expeir.nce, During refueling operations, the level in the spent fuel, pool is in.. .

equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.7.1.

McGuire Units 1 and 2 B 3.7.13-2 Revision N1;X

Spent Fuel Pool Boron Concentration B 3.7.14 Bases APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.

ACTIONS A. 1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.14.1 REQUIREMENTS

• This SR verifies that the concentration of boron in the spent fuel pool is j.-//r -e"-/ within the required limit. As Iong__s this SR is m~t the analyzed 37" idt e fully addressed-A:he 71a r. ency is/pproyriate*'

.. /,/.bica~e*3 mJ~eplerjjs, eat .of ol wat is expecf to ta*.kdplace ,

REFERENCES 1. UFSAR, Section 9.1.2.

2. Issuance of Amendments, McGuire Nuclear Station, Units 1 and 2 (TAC NOS. MC0945 and MC0946), March 17, 2005.

3. 10 CFR 50.68, "Criticality Accident Requirements"

4. American Nuclear Society, "American National Standard Design Requirements for Light Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSI/ANS-57.2-1983, October 7, 1983.

5. Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19, 1998.

McGuire Units 1 and 2 B 3.7.14-3 Revision N

Secondary Specific Activity B 3.7.1.6 BASES ACTIONS A.1 and A.2 DOSE EQUIVALENT 1-131 exceeding the allowable value in the secondary coolant, is an indication of a problem in the RCS and contributes to increased post accident doses. If the secondary specific activity cannot be restored to within limits within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS This SR verifies that the secondary specific activity is within the limits of the accident analysis. A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE. he 31ay Frequ a'cy is base on the eection of increami!re

.appro s o the lebe taken,6 if action of DOSmaintai QUIVA!

n.-tvels Tbelow'th 1-131,jpn e LCallowmit /

REFERENCES 1. 10 CFR 100.11.

2. UFSAR, Section 15.1.5. .

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.7.16-3 Revision NqýRD

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that distribution buses and loads are connected to their preferred power source, and that appropriate SR 3.8.1.2 and SR 3.8.1.7 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition.

To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs are modified by a Note (Note 2 for SR 3.8.1.2) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and followed by a warmup period prior to loading.

For the purposes of SR 3.8.1.2 and SR 3.8.1.7 testing, the DGs are

-started from standby conditions using a manual start, loss of offsite power signal, safety injection signal, orlossof offsite power coincident with a safety injection signal. Standby conditions for a DG mean that the diesel engine coolant and oil are being continuously circulated alnd temperature is being maintained consistent with manufacturer recommendations.

In order to reduce stress and wear, the manufacturer recommends a modified start in which the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 3, which is only applicable when such modified start procedures are recommended by the manufacturer.

SR 3.8.1.7 requires that,3.41.u theiDG starts from standby conditions and achieves required voltage and frequency within 11 seconds. The 11 second start requirement supports the assumptions of the design basis LOCA analysis in the UFSAR, Chapter 15 (Ref. 5).

McGuire Units 1 and 2 B 3.8.1-16 Revision NOoc

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) "

The 11 second start requirement is not applicable to SR 3.8.1.2 (see Note 3) when a modified start procedure as described above is used. If a modified start is not used, the 11 second start requirement of SR 3.8.1.7 applies.

Since SR 3.8.1.7 requires a 11 second start, it is more restrictive than SR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2. This is the intent of Note 1 of SR 3.8.1.2.

The nrq31ncy r SR 3.8. . and the/"84 da F 5equenc fo,38../r fosst it egtory Gui 1. 9(Rety") Ta bley jt s Frq*4is r eae s u ra n/o IDG ZE RABILT-, ./

whlnprmzn egadation r I£ting frorýrestingj._**.

SR 3.8.1.3 This Surveillance verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting loads greater than or equal to the equivalent of the maximum expected accident loads. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.

Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0.

The 0.8 value is the design rating of the machine, while the 1.0 is an operational limitation to ensure circulating currents are minimized. The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in accordanceWith vendor recommendations in order to maintain DG' OPERABILITY.

This SR is modified by four Notes. Note 1 indicates that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized. Note 2 states that momentary transients, because of changing bus loads, do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test. Note 3 indicates that this Surveillance should be conducted on only one 'DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations. Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.

McGuire Units 1 and 2 B 3.8.1-17 Revision Nc(_*

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.8.1.4 This SR provides verification that the level of fuel oil in the day tank is at or above the level at which fuel oil is automatically added. The level is expressed as an equivalent volume in gallons, and is adequate for approximately 30 minutes of DG operation at full load.

TheWday Frqbency isadequate to sure that a ificient su y of l*aisilable, s e l rms are p vided and f ility operatpr' would aware of large use fuel oil dd g this period.

SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel oil day tanksq cý ýiliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling andd rvides data regarding the waterti ht inte1 .tyof thfuel oil system. The uurveila np, Fr ens re eofablis

/ g uj*u atGuie 1,1 (Ref. his SR is for eventative maintenance. Thepresence of water does not nec ssarily represent failure of this SR, provided the accumulatedwater is moved during the nirfnrmln" nf fhi .i iniqllnri1nni-i SR 3.8.1.6 This Surveillance demonstrates that each required fuel oil transfer pump operates and transfers fuel oil from its associated storage tank to its associated day tank. This is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE.

The ign of Eel transwfsystem such th umps opeBate a ati *or ma be starte anualy I -rd eeru To mao aman adequ volume fuel oil

• the day ta s during ollowi DG test;g. Ther re, a 31 ay Freque is appr late.

McGuire Units 1 and 2 B 3.8.1-18 Revision NoO6

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.8.1.7 See SR 3.8.1.2.

SR 3.8.1.8 Transfer of each 4.16 kV ESF bus power supply from the normal offsite circuit to the alternate offsite circuit demonstrates the OPERABILITY of the alternate circuit d" ib "on etwork to power the shutdown loads.

The 1S m dth Fre ency f rseaon fr t e a rNte.The isbtnht edrg_

oper ato wintot cotoreritic pror unmn o- tions r uired terform Sth uvellc, a inendeo be co* tsent wi expect/" fuel cy Ce prnthu Je pass sSR to the eletric n perfrd dis al at the ýon s that e comlents uc ally monthequen/C/

continued steady state operation' and, as a result, unit safety systems.

SR 3.8.1.9 Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss. of a large load could cause diesel engine overspeed, which, ifexcessive, might result in a trip of the engine. This. Surveillance demonstrates the DG load response*characteristics and capability to reject the largest single load without exceeding predetermined voltage and frequency and.while .. .

maintaining a specified margin to the overspeed trip. For this unit, the single load for each DG and its kilowatt rating is as follows: Nuclear Service Water Pump which is a. 576 kW motor. This Surveillance may be accomplished by:

a. Tripping the DG output breaker with the DG carrying greater than or equal to its associated single largest post-accident load while paralleled to offsite power, or while solely supplying the bus; or

b. Tripping its associated single largest post-accident load with the DG solely supplying the bus.

As required by Regulatory Guide 1.9 (Ref. 3), the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the McGuire Units 1 and 2 B 3.8.1-19 Revision Nog

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) difference between synchronous speed and the overspeed trip setpoint, or 15% above synchronous speed, whichever is lower.

The time, voltage, and frequency tolerances specified in this SR are derived from Regulatory Guide 1.9 (Ref. 3) recommendations for response during load sequence intervals. The 3 seconds specified is equal to 60% of a typical 5 second load sequence interval associated with sequencing of the largest load. The voltage and frequency specified are consistent with the design range of the equipment powered by the DG.

SR 3.8.1.9.a corresponds to the maximum frequency excursion, while SR 3.8.1.9.b and SR 3.8.1.9.c are steady state voltage and frequenc values to which the system must recover following load rejecti2hTe,,"'

j:*on~ ~~~ sc niTt _-Feueny wthJ recommendaýtýen-o--R"tCatoj" This Surveillance is performed with the DG connected to its bus in parallel with offsite power supply. The DG is tested under maximum kVAR loading, which is defined as being as close to design basis conditions as practical subject to offsite power conditions. Design basis conditions have been calculated to be greater than 0.9 power factor. During DG testing, equipment ratings are not to be exceeded (i.e., without creating an overvoltage condition on the DG or 4 kV emergency buses, over-excitation in the generator, or overloading the DG emergency feeder while maintaining the power factor greater than or equal to 0.9).

This Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit. or grid perturbations.

SR 3.8.1.10.

This Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the predetermined voltage limits.

The DG full load rejection may occur because of a system fault or inadvertent breaker tripping. This Surveillance ensures proper engine generator load response under the simulated test conditions. This test simulates the loss of the total connected load that the DG experiences following a full load rejection and verifies that the DG does not trip upon loss of the load. These acceptance criteria provide for DG damage protection. While the DG is not expected to experience this transient during an event and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated.

McGuire Units 1 and 2 B 3.8.1-20 Revision NO*$

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)

Although not representative of the design basis inductive loading that the DG would experience, a power factor of approximately unity (1.0) is used for testing. This power factor is chosen in accordance with manufacturer's recommendations to minimize DG overvoltage during testing.

This Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.

SR 3.8.1.11 As required by Regulatory Guide 1.9 (Ref. 3), paragraph 2.2.4, this

.Surveillance demonstrates the as designed operation of the standby power sources during loss of the offsite source. This test verifies the de-energization of the emergency buses, load shedding from the emergency buses and energization of the emergency buses and blackout loads from the DG. Tripping of non-essential loads is not verified in this test. It further demonstrates the capability of the DG to automatically achieve the required voltage and frequency within the specified time.

The DG autostart time of 11 seconds is derived from requirements of the accident analysis Surveillance shouldto be respond to a-design continued basis large for a minimum of 5break LOCA.

minutes Theto in order demonstrate that all starting transients have decayed and stability is achieved.:

The requirement to verify the connection and power supply of the emergency bus and autoconnected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation. For instance, Emergency Core Cooling Systems (ECCS) injection valves are not desired to be stroked open, or high pressure injection systems are not capable of being operated at full flow, or residual heat removal (RHR) systems performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG systems to perform these functions is acceptable.

This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

McGuire Units 1 and 2 B 3.8.1-21 Revision N9

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) f Re~try Gud _E9(Rf. 3)T5e1 a osidr nt

~This SR is modified by two Notes. The reason for Note 1 is to minimize

/ wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine

/ coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations. The reason for Note 2 is

( that performing the Surveillance would remove a required offsite circuit

~from Ssafetyservice, perturb the electrical distribution system, and challenge systems.

SR 3.8.1.12 This Surveillance demonstrates that the DG automatically starts and achieves the required voltage and frequency within the specified time (11 seconds) from the design basis actuation signal (LOCA signal) and operates for >_5 minutes. The 5 minute period provides sufficient time to demonstrate stability. SR 3.8.1.12.d ensures that the emergency bus remains energized from the offsite electrical power system on an ESF signal without loss of offsite power. This Surveillance also verified the tripping of non-essential loads. Tripping of non-essential loads is verified only once, either in this SR or in SR 3.8.1.19, since the same circuitry is tested in each SR.

The Freoar-encv of.1- 4 onths is4*nsistent1ii'fi RecdulatorvGuide 1.9

-expectKfuel-co~~nets cycleji~idgths. .:Opperting~experig~e su~y pass the,,'R when peofumed athas sho*/y'that theX month/ th*'e. ,

Fequency. T/lerefore, the*Frequency wat concluded t65 be acc al-

-fkom a re-ia~kity standpont /his SR is modified by a Noe The reason for the Note is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.

SR 3.8.1.13 This Surveillance demonstrates that DG non-emergency protective functions (e.g., high jacket water temperature) are bypassed on a loss of voltage signal concurrent with an ESF actuation test signal.

McGuire Units 1 and 2 B 3.8.1-22 Revision NOO

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)

The non-emergency automatic trips are all automatic trips except:

a. Engine overspeed;

b. Generator differential current;

c. Low lube oil pressure; and

d. Generator voltage - controlled overcurrent.

The non-emergency trips are bypassed during DBAs and provide an alarm on an abnormal engine condition. This alarm provides the operator with sufficient time to react appropriately. The DG availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the DG.

FrThe 1ncy waeco ncl is cbsistent wiaa"begulao bIt Guides (Ref.p3)

/Tablt takingzjlo consyid/eation unit pnditions quired to rfor the tded eillacv~ gnd is inter to be ensistent ith expect* fuel cycle//

This SR is not normally performed in MODE 1 or 2, but it may be performed in conjunction with periodic preplanned preventative maintenance*activity that causes the DG to be'inoperable. This is acceptable provided that performance of the SR does not increase the time the DG would be inoperable for the preplanned preventative maintenance activity.

SR 3.8.1.14 Regulatory Guide 1.9 (Ref. 3), paragraph 2.2.9, requires demonstration iC: pe-r18-mo s that the DGs can start and run continuously at full load capability for an interval of not less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, >_2 hours of which is at a load equivalent from 105% to 110% of the continuous duty rating and the remainder of the time at a load equivalent to the continuous duty rating of the DG. The DG starts for this Surveillance can be performed either from standby or hot conditions. The provisions for prelubricating and warmup, discussed in SR 3.8.1.2, and for gradual loading, discussed in SR 3.8.1.3, are applicable to this SR.

McGuire Units 1 and 2 B 3.8.1-23 Revision No.'

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)

This Surveillance is performed with the DG connected to its bus in parallel with offsite power supply. The DG is tested under maximum kVAR loading, which is defined as being as close to design basis conditions as practical subject to offsite power conditions. Design basis conditions have been calculated to be greater than 0.9 power factor. During DG testing, equipment ratings are not to be exceeded (i.e., without creating an overvoltage condition on the DG or 4 kV emergency buses, over-excitation in the generator, or overloading the DG emergency feeder while maintaining the power factor greater than or equal to 0.9).

The load band is provided to avoid routine overloading of the DG.

Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.

Th o8nth Fre9Ancy is con *etwt h eatoryj Gi 1.9 (Ref be ,tk ons(ider at n unit

/*onditions r uired to pe rm the Surv* nce, and is, fnded t~o9.l f"cons ;Vwt expec/te fuel cycle eigths.

This Surveillance is modified by two Notes. Note 1 states that momentary transients due to changing bus loads do not invalidate this test. Note 2 allows gradual loading of the DG in accordance with recommendation from the manufacturer.

This Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.

SR 3.8.1.15 This Surveillance demonstrates that the diesel engine can restart from. a hot condition, such as subsequent to shutdown from normal Surveillances, and achieve the required voltage and frequency within 11 seconds. The 11 second time is derived from the requirementso l~e accident analysis to respond to a design ai a reak LOC ./T "

('f~18 rp::onth F Frquency/is consis nt witb the reeommepdtinso g u la t

• k , * ,'ý G u i d/e / . g( R e/f 3 T a 1. 1 This SR is modified by two Notes. Note 1 ensures that the test is performed with the diesel sufficiently hot. The load band is provided to avoid routine overloading of the DG. Routine overloads may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY. The requirement that the diesel has operated for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at full load McGuire Units 1 and 2 B 3.8.1-24 Revision N.K

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) conditions prior to performance of this Surveillance is based on manufacturer recommendations for achieving hot conditions. Momentary transients due to changing bus loads do not invalidate this test. Note 2 allows all DG starts to be preceded by an engine prelube period to minimize wear and tear on the diesel during testing.

SR 3.8.1.16 As required by Regulatory Guide 1.9 (Ref. 3), paragraph 2.2.11, this Surveillance ensures that the manual synchronization and automatic load transfer from the DG to the offsite source can be made and the DG can be returned to standby operation when offsite power is restored. It also ensures that the autostart logic is reset to allow the DG to reload if a subsequent loss of offsite power occurs. The DG is considered to be in standby operation when the DG is at rated speed and voltage, the output breaker is open and can receive an autoclose signal on bus undervoltage, and the load sequence timers are reset.

ehmonsrain of t8h test modis co ertent ws e recomthth davilail Nder Thacieason for the Note is that performing the Surveillance would

3-ASdistribution andmhe a system, wequired offsite circuit from and challenge service, safety perturb the electrical systems.

• ~~SIR 3.8.1.17 .. .*

• ~Demonstration of'the tes§t mode override ensures that the DG availlibility"

..... *under *accident conditions will. hot be compromised as the result of testing*

and the DO. will automatically reset to standby operation if a. LOCA" -

actuation signal is received during operation in the test: mode. Standby* .

operation is defined as the DG running at rated speed and voltage with the DG output breaker open. These provisions for automatic switchover are required by Regulatory Guide 1.9 (Ref. 3), paragraph 2.2.13. The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.12. The intent in the requirement associated with SR 3.8.1.17.b is to show that the emergency loading was not affected by the DG operation in test mode. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

McGuire Units 1 and 2 B 3.8.1-25 Revision NqY)

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)

Te 1 nt F uency is onsisten ith the rec 1mnda s Of Re atory '3) takes i onside on oitIo[nequired perform e Surveill e, and i itendeod be consint with pected f cycle leng s.

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

SR 3.8.1.18 Under accident and loss of offsite power conditions loads are sequentially connected to the bus by the automatic load sequencer. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents. The load sequence time interval tolerance in Table 8-16 of Reference 2 ensures that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated.

Table 8-1 of Reference 2 provides a summary of the automatic loading of ESF buses. The sequencing times of Table 8-16 are committed and required for OPERABILITY. The typical 1 minute loading duration seen by the accelerated sequencing scheme is NOT required for OPERABILITY.

SR 3.8.1.19 In the event of a DBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded.

This Surveillance verifies the de-energization of the emergency buses, load shedding from the emergency buses, tripping of non-essential loads and energization of the emergency buses and ESF loads from the DG.

Tripping of non-essential loads is verified only once, either in this SR or in SR 3.8.1.12, since the same circuitry is tested in each SR. In lieu of actual demonstration of connection and loading of loads, testing that McGuire Units 1 and 2 B 3.8.1-26 Revision No. 92

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the iDGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations for OGs. The reason for

/f2 Note 2 is that the performance of the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

SR 3.8.1.20 This Surveillance demonstrates that the DG starting independence has not been compromised. Also, this Surveillance demonstrates that each engine can achieve proper speed within the specified time when the DGs are started simultaneously.

Th year Freguecy is cons t wiheJcm

.J;*~gulatory Gvie 1.9 (Rej,-3jTable This SR is modified by a Note. The reason for the Note is to minimize wear on the DG during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant.

and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.

McGuire Units 1 and 2 B 3.8.1-27 Revision No

(

Diesel Fuel Oil and Starting Air B 3.8.3 BASES ACTIONS (continued) restored to the required limit. A period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is considered sufficient to complete restoration to the required pressure prior to declaring the DG inoperable. This period is acceptable based on the remaining air start capacity, the fact that most DG starts are accomplished on the first attempt, and the low probability of an event during this period.

E. 1 With a Required Action and associated Completion Time not met, or one or more DG's fuel oil or starting air subsystem not within limits for reasons other than addressed by Conditions A through D, the associated DG may be incapable of performing its intended function and must be immediately declared inoperable.

SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 5 days at full load.

The 4 day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.

/The ay Freguency is adquate to sure tha sufficient *ly of oilis av*3 itable, sinow leve arms are rovided an nit ope tors

- would biware of y large u s of fuel o/ during thi5.p~riod.

SR 3.8.3.2 The tests listed below are a means of determining whether new fuel oil is:

of the appropriate grade and has not been contaminated with substances that would have an immediate, detrimental impact on diesel engine combustion. If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. These tests are to be conducted prior to adding the new fuel to the storage tank(s). The tests, limits, and applicable ASTM Standards are as follows:

a. Sample the new fuel oil in accordance with ASTM D4057 (Ref. 7);

b. Verify in accordance with the tests specified in ASTM D975 that the sample has a kinematic viscosity at 401C of > 1.9 centistokes and

• 4.1 centistokes, and a flash point of > 125°F; and McGuire Units 1 and 2 B 3.8.3-4 Revision N0%*

Diesel Fuel Oil and Starting Air B 3.8.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.8.3.3 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available. The system design requirements provide for a minimum of five engine start cycles without recharging. A start cycle is defined as the period of time required to reach 95% speed from' standby prelubed condition. The pressure specified in this SR is intended to reflect a conservative value at which a single fast start and five total starts can be accomplished.

dayrnmn For bancytaerinto srial.o Thsihe mosteffctiv , meaability SR 3.8.3.4 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel storage tanksc(ýý eliminates the necessary environment for bacterial survival. This isthe most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, and contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of

. . .the fuel oil system , he , *'eillape -t e i*

-- ii re eý bslis d

.*gu r Guýi d e 1.. Ref.(his',SR is for p eventive maintenance.

"M" " uireThe . presence of

... provided the accumulated waterneis removeld water does2 ly repr Uni gsent failureBof this performance SR,..

of the Surveillance.

McGuire Units I and 2 B 3.8.3-6 Revision NO*-

DC Sources-Operating B 3.8.4 BASES ACTIONS (continued)

If one of the required channels of DC is inoperable (e.g., inoperable battery, inoperable battery charger(s), or inoperable battery charger and associated inoperable battery), the remaining DC channels have the capacity to support a safe shutdown and to mitigate an accident condition. If the channel of DC cannot be restored to OPERABLE status, Action A.2 must be entered and the DC channel must be energized from an OPERABLE channel, from the same train, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The capacity of the redundant channel is sufficient to supply its normally supplied channel and cross tied channel for the required time, in case of a DBA event. The inoperable channel of DC must be returned to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and the cross ties to the other channel open. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time reflects a reasonable time to assess unit status as a function of the inoperable channel of DC and, ifthe DC channel is not restored to OPERABLE status, to prepare to effect an orderly and safe unit shutdown.

B.1 and B.2 If the inoperable channel of DC cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.

The Completion Time to bring the unit to MODE 5 is consistent with the time required in Regulatory Guide 1.93 (Ref. 9).

SURVEILLANCE. SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or a battery cell) in a fully charged state. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations Th Uay 1qu as n i anu rer McGuire Units 1 and 2 B 3.8.4-4 Revision N4* 0

DC Sources-Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.8.4.2 Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each intercell, interrack, intertier, and terminal connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.

SThe S illan~c requenc or these ins ctions, whi can detec co itions t , can cau power 1 d ce heatin , is days. his Frequ cy is consiered accept e based o perating exper' ce relate o detectingxorrosion tren s.

SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and battery racks provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance. The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an evaluation determines that the physical damage or

- -- deterioration does not affect the OPERABILITY of the battery (its abilit to pefrmis design function). ea n exp e~ience has/shown that t~hes~e com nen s usu y pass R when rformed ate 18 mont

• ( F~quency. ýKref,.,,or^e, t~h requen qy~as conc ued to be accptable/

• "-----?k, zfrom a rel~ia~ility standp int. - --- - - " -

SR 3.8.4.4 and SR 3.8.4.5 Visual inspection and resistance measurements of intercell, interrack,...

intertier, and terminal connections provide an indication of physical damage or abnormal deterioration that could indicate degraded battery condition. The anticorrosion material is used to help ensure good electrical connections and to reduce terminal deterioration. The visual inspection for corrosion is not intended to require removal of and inspection under each terminal connection. The removal of visible corrosion is a preventive maintenance SR. The presence of visible corrosion does not necessarily represent a failure of this SR provided c visible corrosion is removed during performance of SR 3.8.4.4Operati

• /e a s/wntattees component~s-usua ypa e SR eh*n ,f

• perogr~l'ed a JJ~t'18 mnt~eqency. T :efore, the e~quen wfas k,,*cluded~5 be accep4*ole froma r!ýeji~alrlity standlin* -- --

McGuire Units 1 and 2 B 3.8.4-5 Revision Ncý

DC Sources-Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.8.4.6 This SR requires that each battery charger be capable of supplying 400 amps and 125 V for _>1 hour. These requirements are based on the design requirements of the chargers. According to Regulatory Guide 1.32 (Ref. 11), the battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensures that these requirements can be satisfied.

The Surv ance Frequ cy is accept e, given the unit cSpditions requir to perform e test and the ther administrativ ontrols existi to sure adequ e charger pe mance during the 18 month int als.

n addition, t i Frequency is~tended to be co rstent with expe ed fuel /

cycle len s.

SR 3.8.4.7 A battery service test is a special test of battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> should correspond to the design duty cycle requirements as specified in Reference 4.

The Surveille Frequency** months is co stent with the=

recomm ations of Re gatory Guide 1.3 ef. 11) with the ception e.

that j allowable tpil erform the batte ervice test with nit in any.

This SR is modified by a Note. The Note allows the performance of a modified performance discharge test in lieu of a service test.

The modified performance discharge test, as defined by IEEE-450 (Ref.

12) is a simulated duty cycle consisting of just two rates; the one minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a rated one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance McGuire Units 1 and 2 B 3.8.4-6 Revision NoO

DC Sources-Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued) discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.

A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service test.

SR 3.8.4.8 A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.

A battery modified performance discharge test is described in the Bases for SR 3.8.4.7 and in IEEE-450 (Ref. 12). Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy.SR 3.8.4.8 while satisfying the requirements of SR 3.8.4.7 at the same time.

The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 12). These references recommend that the battery be replaced if its capacity is below 80% of.the manufacturers rating. A capacity of 80%

shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.

,nurveirance FrPency foromtest is 6rmalJY-'6monts.I.f the shows degradation, dattery or ifthe battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity _

100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 10), when the battery capacity drops by more than 10%

relative to its capacity on the previ us performance test or when it is > 10%

Mcr Ui below the ma B3ufacturers. ratingRse ciesvarec istentwi the_-

McGuire Units 1 and 2 B 3.8.4-7 Revision No.-?)Q

Battery Cell Parameters B 3.8.6 BASES ACTIONS (continued) completing the Required Actions of Condition A within the required Completion Time or average electrolyte temperature of representative cells falling below 60 0 F, are also cause for immediately declaring the associated DC electrical power subsystem inoperable.

SURVEILLANCE SR 3.8.6.1 REQUIREMENTS This SR verifies that Category A battery cell parameters are consistent with IEEE-450 (Ref. 4), which recommends regular battery inspectionsw I e including voltage, specific gravity, and electrolyte temperature of pilot cells.

SR 3.8.6.2 qu ySins on s f ic ra vit y-lcvolt a e s *asitrPt

/ /IEE V~efi n addition, within 7 days of a battery discharge

< 110 V or a bat ery overcharge > 150 V, the battery must be demonstrated to meet Category B limits. Transients, such as motor starting transients, which may momentarily cause battery voltage to drop to _<110 V, do not constitute a battery discharge provided the battery terminal voltage and float current return to pre-transient values. This 5 -- inspection is also consistent with IEEE-450 (Ref. 4), which recommends special inspections following a severe discharge or overcharge, to ensure that no significant degradation of the battery occurs as a consequence of such discharge or overcharge.

SR 3.8.6.3

,-f'TsT i uS iance veiriication that t eage tempera/,wre of

( rej~ ntatie91 s>6 tn ih a mmendatiL~ of'

• / .,,,'EE-50 P* 4, ha sate, at t-he temper dre of electr ~es in J represept{ive cells shoLA be determined eia quarterly basis.

Lower than normal temperatures act to inhibit or reduce battery capacity.

This SR ensures that the operating temperatures remain within an acceptable operating range. This limit is based on manufacturer recommendations.

The term "representative cells" replaces the fixed number of "six connected cells", consistent with the recommendations of IEEE-450 (Ref.

4) to provide a general guidance to the number of cells adequate to McGuire Units 1 and 2 B 3.8.6-3 Revision N

Inverters-Operating B 3.8.7 BASES ACTIONS (continued)

Required Action A.1 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to fix the inoperable inverter and return it to service. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit is based upon engineering judgment, taking into consideration the time required to repair an inverter and the additional risk to which the unit is exposed because of the inverter inoperability. This has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems such a shutdown might entail. When the AC vital bus is powered from its regulated voltage transformer, it is relying upon interruptible AC electrical power sources (offsite and onsite).

The uninterruptible inverter source to the AC vital buses is the preferred source for powering instrumentation trip setpoint devices.

B.1 and B.2 If the inoperable devices or components cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning

... properly with all, required, circuit breakers closed and AC vital bus energized from the inverter. The verification of proper. voltage output ensures that the required power is readily available for the instrumentation of the RPS. and ESFAS connected to the AC vital buses. he.7 day Frequency takes.nto account the dundarn 1aof the i ert'ers and er indications i aebl e i'e c--=ol roo at alert theperator to invert r malfun ions.

REFERENCES 1. UFSAR, Chapter 8.

2. UFSAR, Chapter 6.

3. UFSAR, Chapter 15.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.8.7-3 Revision No(

Inverters - Shutdown B 3.8.8 BASES ACTIONS (continued) sources that have a boron concentration greater than that what would be required in the RCS for minimum SDM or refueling boron concentration.

This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.

Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of required SDM.

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required inverters and to continue this action until restoration is accomplished in order to provide the necessary inverter power to the unit safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required inverters should be completed as quickly as possible in order to minimize the time the unit safety systems may be without power or powered from a regulated voltage transformer.

SURVEILLANCE SR 3.8.8.1 REQUI REMENTS This. Surveillance verifies that the power sources are functioning properly with all required circuit breakers closed and AC Vital buses energized.

/ *,V~z*-*/-- * .from the *required power source. The verification of proper voltage.

(.* /// "ensures that the required power is readily available forth

'** . ./ ..

• instrumentation connected to the AC vital buses.Aihe 7 da-y'requency'.

• ,. '

• es~I ccount t '11 ldudnrcs .

ot indicationsailable in the ntrol roomth ert the op ator to

"------) ,Kfverter malf "tions,.

REFERENCES 1. UFSAR, Chapter 6.

2. UFSAR, Chapter 15.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.8.8-3 Revision NCF

Distribution Systems-Operating B 3.8.9 BASES ACTIONS (continued)

D.1 and D.2 If the inoperable distribution subsystem cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a

• MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.

E.1 Condition E corresponds to a level of degradation in the electrical power distribution system that causes a required safety function to be lost. When more than one inoperable electrical power distribution subsystem results in the loss of a required function, the plant is in a condition outside the accident analysis. Therefore, no additional time is justified for continued operation.

LCO 3.0.3 must be entered immediately to commence a controlled shutdown.

SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This Surveillance verifies that the AC, DC, and AC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment. The correct breaker alignment ensures the appropriate. separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus. The verification of proper voltage availability on the buses ensures that the required, voltage is, readily available for motive as well as control functions for critical system.

loadsConnected. to these buses /he 71 Frequenc a es*in oac ount the Sredun t capa-t 4 of the.X DC, AC. vital b electrical er d._

d bution systems nd oth endications ailable in t control ro iij 4hat alert.ýhe operator/to subs em malfu ions.

REFERENCES 1. UFSAR, Chapter 6.

2. UFSAR, Chapter 15.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

4. Regulatory Guide 1.93, December 1974.

McGuire Units 1 and 2 B 3.8.9-7 Revision No

Distribution Systems-Shutdown B 3.8.10 BASES SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the AC, DC, and AC vital bus electrical N* power distribution subsystems are functioning properly, with all the buses

?7---7-ft energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buse§ .T 7 day I equenc takes into a Pount the ca ility of the ectrical p er dis *rution s Z.systems,,r0 other indi ions availapin the corol om that ert the operator to subs em malfurfons.

REFERENCES 1. UFSAR, Chapter 6.

2. UFSAR, Chapter 15.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.8.10-4 Revision N

Boron Concentration B 3.9.1 BASES SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This SR ensures that the coolant boron concentration in the RCS, and connected portions of the refueling canal and the refueling cavity, is within the COLR limits. The boron concentration of the coolant in each required volume is determined periodically by chemical analysis. Prior to re-connecting portions of the refueling canal or the refueling cavity to the RCS, this SR must be met per SR 3.0.4. If any dilution activity has occurred while the-cavity or canal were disconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS. One sample from the refueling canal or refueling cavity is sufficient to determine the boron concentration in that volume of water.

An additional sample is taken from the RCS.

A minim requenc once eve ours is a reaso le amount f ime verify the ron concentr n of representa-samples.

equencyis sed on oper ing experience, vvch has show 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to be ade ate.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.9.1-4 Revision NC59)

Unborated Water Source Isolation Valves B 3.9.2 BASES SURVEILLANCE SR 3.9.2.1 REQUIREMENTS These valves are to be secured closed to isolate possible dilution paths.

The likelihood of a significant reduction in the boron concentration during MODE 6 operations is remote due to the large mass of borated water in the refueling cavity and the fact that all unborated water sources are isolated, precluding a dilution. The boron concentration is checked every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during MODE 6 under SR 3.9.1.1. This Surveillance demonstrates that the valves are closed through a system walkdown.

The 31 Frequency is based on-engineering jud ent an*d

( co ered reafable in vie)4t other admi nsure th ,ffe valve openifg is an unlik ative co possibilit s that wi ")

REFERENCES 1. UFSAR, Section 15.4.6.

2. NUREG-0800, Section 15.4.6.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.9.2-3 Revision Nq0

Nuclear Instrumentation B 3.9.3 BASES SURVEILLANCE SR 3.9.3.1 REQUIREMENTS SR 3.9.3.1 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.

The F uency ofllours is ent with the NNELC* K ency sp ied similar rthe same in ments in d2 3.3.1.

1A -IJ

• T-- SR 3.9.3.2 SR 3.9.3.2 is the performance of a CHANNEL CALlBRATION! e (J -f The CHANNEL CALIBRATION ensures that the monitors ar(

calibrated. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL ALIBRAI I he 18 month requenc ased on the eed to perfor this Surveilla e under the conditi s that apply ing a plant o ge. Operatin xperience sh n these com nents usuall ss the Surveil ce when pHormed

.*the 18 montt requency./-

REFERENCES 1. 10 CFR 50, AppendixA, GDC 13, GDC 26, GDC 28, and GDC 29.

2. UFSAR, Sections 4.2,15.4.6.

3. 10 CFR 50.36, Technical Specifications, .(c)(2)(ii).

McGuire Units 1 and 2 B 3.9.3-4 Revision N -0

Containment Penetrations B 3.9.4 BASES SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. The Surveillance on the open purge and exhaust valves will demonstrate that the valves are exhausting through an OPERABLE Containment Purge Exhaust System HEPA filter and charcoal adsorber.

7d dring moveme of recenty*

r*u Iirad* tul as les within..co .ptaffiment. The §plq lanceý i l is,,

C e-tdt ý at h omldurafi~on of time t coplt fdlhnln pe*!~s ucti uvilance .ensures.

that a postulated fuel han ing accident involving recently irradiated fuel "z-"e

• that releases fission product radioactivity within the containment will not result in a release of significant fission product radioactivity to the 3 environment.

SR 3.9.4.2 This SR verifies that the required testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The Containment Purge Exhaust System filter tests are in accordance with Reference 3. The VFTP includes testing HEPA filter performance, charcoal adsorbers efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test Frequencies and additional information are discussed in detail in the VFTP.

REFERENCES 1. UFSAR, Section 15.7.4.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

3. Regulatory Guide 1.52 (Rev. 2).

4. 10 CFR 50.67, Accident Source Term.

5. Regulatory Guide 1.183, Rev 0.

6. PIP M-05-1608.

McGuire Units 1 and 2 B 3.9.4-4 Revision No

RHR and Coolant Circulation - High Water Level B 3.9.5 BASES ACTIONS (continued)

A.4 If RHR loop requirements are not met, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded.

The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on the low probability of the coolant boiling in that time.

SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate

/necessary to provide sufficient decay heat removal capability and to 3prevent thermal and boron stratification in the core. The RCS temperature is determined to ensure the appropriate deca heat removal ismittni*e ýreni'Iuency-,v fT1 2 hhwrs is sufficientx* siern fflow.lemperature mp control* alarm indicns available he)

"*,..*_ pre._

ator in th ontrol room-fdr monitoring theýRHR System.- J REFERENCES 1. UFSAR, Section 5.57.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii)..

3. NUMARC 91-06, "'Guidelines for Industry Actions to Assess Shutdown Management".

McGuire Units 1 and 2 B 3.9.5-4 Revision N(Ig

RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES ACTIONS (continued)

B.2 If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore one RHR loop to operation. Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.

B.3 If no RHR loop is in operation, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded. The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is appropriate for the majority of time during refueling operations, based on time to coolant boiling, since water level is not routinely maintained at low levels.

SURVEILLANCE SR 3.9.6.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability, prevent vortexing in the suction of the RHR pumps, and .to prevent thermal and boron stratification in the core. The RCS temperature is determined to ensure the

. appropriate decay heat removal is maintained. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vesse nozzles 1'2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> the RHR pump

• 'sufficient, suction requirements co~drn must h lwmeaue*~ he ot*

Frequen.of!

and

/ ~~~alarr *dications av ~~ile to the op -[r for moniýýr"g the RH,R"Sstlem in

,..f*""*L_*'thcontrol rooy/.7

• ,..... . J - -

SR 3.9.6.2 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and Power available to the required pump.

( a~fhstrative co s available"ýas been s/w' iB to be accep-tab*" by

• ,,.*~perating ex f 7 dU is ience, 3isi considerand .

r 2asonable hey e M?

McGuire Units 1 and 2 B 3.9.6-3 Revision N(6?*

Refueling Cavity Water Level B 3.9.7 BASES APPLICABILITY LCO 3.9.7 is applicable during CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, and is also applicable when moving irradiated fuel assemblies within containment. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel assemblies are not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident.

Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.13, "Spent Fuel Pool Water Level."

ACTIONS A.1 and A.2 With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.

The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.7.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met.

Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2).

The Frepncy of 24 rs is based*oneineering judgrert and is cOne.ered adequ in view of the I e volume of w*,*r and the nor cedural co ols of valve pos' ns, which ma significant un9Ial ned level chan s unlikely.

McGuire Units 1 and 2 B 3.9.7-2 Revision NOý95