ML12032A224

License Amendment Request to Relocate TS Surveillance Frequencies to License Controlled Program in Accordance with TSTF-425, Revision 3

ML12032A224
Person / Time
Site:	Millstone
Issue date:	01/25/2012
From:	Price J Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
11-687
Download: ML12032A224 (141)

Text

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, Virginia 23060 OPDominioW Web Address: www.dom.com January 25, 2012 U.S. Nuclear Regulatory Commission Serial No.11-687 Attention: Document Control Desk NSSL/MLC RO Washington, DC 20555 Docket No. 50-336 License No. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2 LICENSE AMENDMENT REQUEST TO RELOCATE TS SURVEILLANCE FREQUENCIES TO LICENSEE CONTROLLED PROGRAM IN ACCORDANCE WITH TSTF-425, REVISION 3 In accordance with the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) is submitting a request for an amendment to the technical specifications (TS) for Millstone Power Station Unit 2 (MPS2). The proposed amendment would modify TSs 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 5b." Additionally, the change would add a new program, the Surveillance Frequency Control Program (SFCP), to TS Section 6, Administrative Controls. The changes are consistent with NRC-approved Industry/lTSTF Standard Technical Specifications (STS) change TSTF-425, Revision 3, (ADAMS Accession No. ML090850642). The Federal Register notice published on July 6, 2009 (74 FR 31996), announced the availability of this TS improvement.

Attachment 1 provides a description and assessment of the proposed change.

Attachment 2 includes DNC documentation with regard to Probabilistic Risk Assessment technical adequacy. Attachment 4 provides a cross-reference between the NUREG-1432 surveillances included in TSTF-425 versus the MPS2 surveillances included in this amendment request. Attachments 3 and 6 provide the MPS2 marked-up TS pages and TS Bases pages, respectively. The marked-up TS Bases pages are provided for information only. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program upon approval of this amendment request.

As detailed in Attachment 5, the proposed amendment does not involve a Significant Hazards Consideration pursuant to the provisions of 10 CFR 50.92.

The Facility Safety Review Committee has reviewed and concurred with the determinations herein.

Issuance of this amendment is requested no later than January 28, 2013 with the amendment to be implemented within 60 days.

Serial No: 11-687 Docket No. 50-336 Adoption of TSTF-425, Rev. 3 Page 2 of 3 In accordance with 10 CFR 50.91(b), a copy of this license amendment request is being provided to the State of Connecticut.

Should you have any questions in regard to this submittal, please contact Wanda Craft at (804) 273-4687.

Sincerely, VICKI L. HULL, Notary Public J. ,c Commonweaiui of Virgini Vi e siden t - Nuclear Engineering ] 140542 j My; Commission Expires May'31. 2014 COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. Alan Price, who is Vice President - Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this 25,oday of 2012.

My Commission Expires: .. /,

t,//1L10? (

Notary Public Attachments:

1. Description and Assessment of Proposed Changes

2. Documentation of PRA Technical Adequacy

3. Marked-up Technical Specifications Changes

4. Cross-References - NUREG-1432 to MPS2TS Surveillance Frequencies Removed

5. Significant Hazards Consideration Determination

6. Marked-Up Technical Specifications Bases Changes (For Information Only)

Commitments made in this letter: None

Serial No: 11-687 Docket No. 50-336 Adoption of TSTF-425, Rev. 3 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road King of Prussia, PA 19406-1415 C. J. Sanders NRC Project Manager U.S. Nuclear Regulatory Commission, Mail Stop 08B3 One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Bureau of Air Management Monitoring and Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.11-687 Docket No. 50-336 ATTACHMENT 1 Description and Assessment of Proposed Changes f(C DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No.11-687 Docket No. 50-336 Attachment 1, Page 1 of 5 DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGES

1.0 DESCRIPTION

In accordance with the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc.

(DNC) is submitting a request for an amendment to the technical specifications (TSs) for Millstone Power Station Unit 2 (MPS2). The proposed amendment would modify TSs 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 5b." Additionally, the change would add a new program, the Surveillance Frequency Control Program (SFCP), to TS Section 6, Administrative Controls. The changes are consistent with NRC-approved Industry/TSTF Standard Technical Specifications (STS) change TSTF-425, Revision 3, (ADAMS Accession No. ML090850642). The Federal Register notice published on July 6, 2009 (74 FR 31996),

announced the availability of this TS improvement.

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation DNC has reviewed the safety evaluation provided in Federal Register Notice 74 FR 31996, 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). includes DNC documentation with regard to the technical adequacy of the probabilistic risk assessment (PRA) consistent with the requirements of Regulatory Guide (RG) 1.200, Revision 1 (ADAMS Accession No. ML070240001), Section 4.2. also describes any PRA models without NRC-endorsed standards, including documentation of'the quality characteristics of those models in accordance with RG 1 200.

DNC has concluded that the justifications presented in the TSTF proposal and the safety evaluation prepared by the NRC staff are applicable to MPS2 and justify this amendment to incorporate the changes to the MPS2 TSs.

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

Serial No.11-687 Docket No. 50-336 Attachment 1, Page 2 of 5

1. Revised (typed) TS pages are not included in this amendment request given the number of TS pages affected, the straightforward nature of the proposed changes, and outstanding MPS2 amendment requests that may impact some of the same TS pages. Providing only mark-ups of the proposed TS changes satisfies the requirements of 10 CFR 50.90 in that the mark-ups fully describe the changes desired., This represents an administrative deviation from the NRC staff's model application dated July 6, 2009 (74 FR 31996) with no impact on the NRC staff's model safety evaluation published in the same Federal Register Notice. As a result of this deviation, the contents and numbering of the attachments for this amendment request differ from the attachments specified in the NRC staff's model application.

The proposed TS Bases changes are provided to the NRC for information.

2. The inserts provided in TSTF-425 are revised to fit the MPS2 TS format.

The TSTF-425 insert for each relocated surveillance frequency is changed from "in accordance with the Surveillance Frequency Control Program to "at the frequency specified in the Surveillance Frequency Control Program."

The insert provided in TSTF-425 to replace text describing the basis for each frequency relocated to the SFCP has been revised from "The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program" to read "The(se) Surveillance Frequency(ies) is/are controlled under the Surveillance Frequency Control Program." This deviation is consistent with recent NRC guidance. After NRC approval of the license amendment request (LAR) and as part of the LAR implementation, the existing MPS2 Bases information describing the basis for the relocated surveillance frequencies will also be relocated to a licensee-controlled program with the relocated surveillance frequencies.

In addition, other editorial changes to the existing TS wording and/or text inserts are being made. These administrative/editorial deviations of the TSTF-425 inserts and the existing TS wording are necessary to fit the MPS2 TS format.

3. Attachment 4 provides a cross-reference between the NUREG-1432 surveillances included in TSTF-425 versus the MPS2 surveillances included in this amendment request. Attachment 4 includes a summary description of the referenced TSTF-425 (NUREG-1432)/MPS2 TS surveillances which is provided for information purposes only and is not intended to be a verbatim description of the TS surveillances. This cross reference highlights the following:

a. NUREG-1432 surveillances included in TSTF-425 and corresponding MPS2 surveillances with plant-specific surveillance numbers,

Serial No.11-687 Docket No. 50-336 Attachment 1, Page 3 of 5

b. NUREG-1432 surveillances included in TSTF-425 that are not contained in the MPS2 TS, and

c. MPS2 plant-specific surveillances that are not contained in NUREG-1432 and, therefore, are not included in the TSTF-425 mark-ups.

Since the MPS2 TSs are custom TSs, the applicable surveillance requirements and associated Bases numbers differ from the STSs presented in NUREG-1432 and TSTF-425, but with no impact on the NRC staffs model safety evaluation dated July 6, 2009 (74 FR 31996).

For NUREG-1432 surveillances not contained in MPS2 TSs, the corresponding mark-ups included in TSTF-425 for these surveillances are not applicable to MPS2.

This is an administrative deviation from TSTF-425 with no impact on the NRC staffs model safety evaluation dated July 6, 2009 (74 FR 31996).

For MPS2 plant-specific surveillances not included in the NUREG-1432 markups provided in TSTF-425, DNC has determined that since these surveillances involve fixed periodic frequencies, relocation of these frequencies is consistent with TSTF-425, Revision 3, and with the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996), including the scope exclusions identified in Section 1.0, "Introduction," of the model safety evaluation. In accordance with TSTF-425, changes to the frequencies for these surveillances would be controlled under the SFCP.

There are several instances in the MPS2 TSs where the words 'and' and 'or' appear at the end of a surveillance requirement. In most cases, these words are not intended to be logical connectors which place the constraints of the preceding surveillance requirement (often times event-driven) on the remaining portion of the surveillance but rather are used for purposes of readability and flow. This situation applies to the following SRs: 4.1.1.2, 4.1.1.5b, 4.1.3.1.1, 4.1.3.1.4b; 4.2.3.2b, 4.5.1d, 4.9.16.1 and 4.9.17.

As currently written, SR 4.2.1.3b does not specify a surveillance frequency, however; it is performed at least once per 31 days, as required by its applicable station surveillance procedure. As a result, the markup for this SR references the SFCP in accordance with TSTF-425.

The SFCP provides the necessary administrative controls to require that surveillances related to testing, calibration, and inspection are conducted at a frequency to assure the necessary quality of systems and components is maintained, facility operation will be within safety limits, and the limiting conditions for operation will be met. Changes to frequencies in the SFCP would be evaluated using the methodology and PRA guidelines contained in NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," as approved by NRC letter dated September 19, 2007 (ADAMS Accession No. ML072570267). The NEI 04-10, Revision 1

Serial No.11-687 Docket No. 50-336 Attachment 1, Page 4 of 5 methodology includes qualitative considerations, risk analyses, sensitivity studies and bounding analyses, as necessary, and recommended monitoring of the performance of structures, systems, and components (SSCs) for which frequencies are changed to assure that reduced testing does not adversely impact the SSCs. In addition, the NEI 04-10, Revision 1 methodology satisfies the five key safety principles specified in RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," dated August 1998, relative to changes in surveillance frequencies.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration DNC has reviewed the proposed no significant hazards consideration (NSHC) determination published in the Federal Register dated July 6, 2009 (74 FR 31996).

DNC has concluded that the proposed NSHC presented in the Federal Register notice is applicable to MPS2, and is provided as Attachment 5 to this amendment request, which satisfies the requirements of 10 CFR 50.91 (a).

3.2 Applicable Regulatory Requirements A description of the proposed changes and their relationship to applicable regulatory requirements is provided in TSTF-425, Revision 3 and the NRC's model safety evaluation published in the Notice of Availability dated July 6, 2009 (74 FR 31996). DNC has concluded that the relationship of the proposed changes to the applicable regulatory requirements presented in the Federal Register notice is applicable to MPS2.

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

4.0 ENVIRONMENTAL CONSIDERATION

DNC has reviewed the environmental consideration included in the NRC staffs model safety evaluation published in the Federal Register on July 6, 2009 (74 FR 31996).

DNC has concluded that the staffs findings presented therein are applicable to MPS2, and the determination is hereby incorporated by reference for this application.

Serial No.11-687 Docket. No. 50-336.

Attachment 1, Page 5 of 5

5.0 REFERENCES

1. TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -

RITSTF Initiative 5b," March 18, 2009 (ADAMS Accession Number:

ML090850642).

2. NRC Notice of Availability of Technical Specification Improvement to Relocate Surveillance Frequencies to Licensee Control - Risk-Informed Technical Specification Task Force (RITSTF) Initiative 5b, Technical Specification Task Force - 425, Revision 3, published on July 6, 2009 (74 FR 31996).

3. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession Number: ML071360456).

4. Regulatory Guide. 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

January 2007 (ADAMS Accession Number: ML070240001).

5. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," dated August 1998 (ADAMS Accession No. ML003740176).

Serial No.11-687 Docket No. 50-336 ATTACHMENT 2 Documentation of Probabilistic Risk Assessment (PRA)

Technical Adequacy DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 1 of 33 Documentation of Probabilistic Risk Assessment (PRA)

Technical Adequacy 1.0 PURPOSE The purpose of this risk assessment is to provide the Probabilistic Risk Assessment (PRA) technical adequacy of the Millstone Power Station Unit 2 (MPS2) model, M209Aa, to support the Risk-Informed Technical Specification Initiative (RITS) 5b. This includes status of critical PRA model reviews during the PRA Peer Review and a gap assessment with respect to American Society of Mechanical Engineers (ASME) PRA Standard RA-Sb-2005 and its endorsing Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200, Rev. 1.

2.0 INTRODUCTION

The MPS2 PRA model has benefited from the comprehensive technical PRA peer review and self-assessment. These include the MPS2 internal events PRA receiving a formal industry PRA Peer Review in 1999 (Ref. 6.1) and a self-assessment/independent review of the MPS2 PRA against Addendum B of the ASME/ANS PRA Standard and RG 1.200, Revision 1 (Ref. 6.3).

3.0 ANALYSIS Documentation of the PRA technical adequacy includes the following information:

1. Proposed Risk-Informed Application

• Description of RITS 5b process

2. PRA Quality Overview

3. Technical Adequacy of the PRA Model

• PRA Maintenance and Update

" PRA Model timeline of improvements

4. Comprehensive Critical Reviews

• CEOG PRA Peer Review

" MPS2 PRA Self-Assessment

5. Status of Identified Gaps to NEI 00-02 and Capability Category II of the ASME PRA Standard

6. External Events Considerations

" Fire Risk

, Seismic Risk

" High Winds, Floods and Other External Events

7. Summary

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 2 of 33 3.1 Proposed Risk-Informed Application The implementation of the Surveillance Frequency Control Program (SFCP, also referred to as RITS 5b) at MPS2 will follow the guidance provided in NEI 04-10, Revision 1 (Ref.

6.5) in evaluating proposed surveillance test interval (STI; also referred to as "surveillance frequency") changes. The following steps of the risk-informed STI revision process are common to all proposed STI changes within the proposed licensee-controlled program.

• Each STI revision is reviewed to determine whether there are any commitments made to the NRC that may prohibit changing the interval. If there are no related commitments, or the commitments may be changed using a commitment change process based on NRC endorsed guidance, then evaluation of the STI revision would proceed. If a commitment exists and the commitment change process does not permit the change, then the STI revision would not be implemented.

Only after receiving formal NRC approval to change the commitment would a STI revision proceed.

• A qualitative analysis is performed for each STI revision that involves several considerations as explained in NEI 04-10, Revision 1.

a Each STI revision is reviewed by an expert panel, referred to as the Integrated Decision-making Panel (IDP), which is normally the same panel used for Maintenance Rule implementation, but with the addition of specialists with experience in surveillance tests and system or component reliability. If the IDP approves the' STI revision, the change is documented and implemented, and available for future audits by the NRC. If the IDP does not approve the STI revision, the STI value is left unchanged.

• Performance monitoring is conducted as recommended by the IDP. In some cases, no additional monitoring may be necessary beyond that already conducted under the Maintenance Rule. The performance monitoring helps to confirm that no failure mechanisms related to the revised test interval become important enough to alter the information provided for the justification of the interval changes.

• The IDP is responsible for periodic review of performance monitoring results. If it is determined that the time interval between successive performances of a surveillance test is a factor in the unsatisfactory performances of the surveillance, the IDP returns the STI to the previously acceptable STI.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 3 of 33 In addition to the above steps, the PRA is used, when possible, to quantify the effect of a proposed individual STI revision compared to acceptance criteria in NEI 04-10, Revision 1. Also, the cumulative impact of all risk-informed STI revisions on all PRA evaluations (i.e., internal events, external events and shutdown) is also compared to the risk acceptance criteria as delineated in NEI 04-10, Revision 1.

For those cases where the STI 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.

3.2 PRA Quality Overview The NEIl 04-10, Revision 1 methodology endorses the guidance provided in RG 1.200, Revision 1 (Ref. 6.7), "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities." The guidance in RG 1.200 indicates that the following steps should be followed when performing PRA assessments:

1. Identify the parts of the PRA used to support the application.

" Structures, systems, and components (SSCs), operational characteristics affected by the application and how these are implemented in the PRA model.

" A definition of the acceptance criteria used for the application.

2. Identify the scope of risk contributors addressed by the PRA model.

• If not full scope (i.e., internal events, external events, shutdown), identify appropriate compensatory measures or provide bounding arguments to address the risk contributors not addressed by the PRA model.

3. Summarize the risk assessment methodology used to assess the risk of the application.

• Include how the PRA model was modified to appropriately model the risk impact of the change request.

4. Demonstrate the technical adequacy of the PRA.

• Identify plant changes (design or operational practices) that have been incorporated at the site, but are not yet in the PRA model and justify why the change does not impact the PRA results used to support the application.

• Document peer review findings and observations that are applicable to the parts of the PRA required for the application, and for those that have not yet been addressed, justify why the significant contributors would not be impacted.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 4 of 33

" Document that the parts of the PRA used in the decision are consistent with applicable standards endorsed by the RG (specifically RG 1.200, Revision 1, which includes only internal events). Provide justification to show that where specific requirements in the standard are not adequately met, it will not unduly impact the results.

" Identify key assumptions and approximations relevant to the results used in the decision-making process.

(NOTE: Because of the broad scope of potential Initiative 5b applications, and the fact that the risk assessment details will differ from application to application, each of the issues encompassed in Items 1 through 3 above will be covered with the preparation of each individual PRA assessment made in support of the individual STI interval requests.

Item 3 satisfies one of the requirements of Section 4.2 of RG 1.200. The remaining requirements of Section 4.2 are addressed by Item 4, which is described in the next section.)

3.3 Technical Adequacy of the PRA Model Dominion employs a structured approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating Dominion nuclear generating sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the MPS2 PRA.

PRA Maintenance and Update The MPS2 PRA model of record, M209Aa, and associated documentation, has been maintained as a living program and the PRA is updated approximately every 3 to 5 years to reflect the as-built, as-operated plant. The M209Aa PRA model is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA model quantification process used for the MPS2 PRA is based on the event tree/fault tree methodology, which is a well-known methodology in the industry.

There are several procedures and GARDs (Guidance and Reference Documentation) that govern Dominion's PRA program. Procedure NF-AA-PRA-101 controls the maintenance and use of the PRA documentation and the associated NF-AA-PRA Procedures and GARDs. These documents define the process to delineate the types of calculations to be performed, the computer codes and models used, and the process (or technique) by which each calculation is performed.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 5 of 33 The NF-AA-PRA series of GARDs and Procedures provide a detailed description of the methodology necessary to:

0 Perform PRA for the Dominion Nuclear Fleet, including Kewaunee, Millstone, North Anna and Surry Power Stations 0 Create and maintain products to support licensing and plant operation concerns for the Dominion Nuclear Fleet

• Provide PRA model configuration control

• Create and maintain configuration risk evaluation tools for the Dominion Nuclear Fleet The purpose of the NF-AA-PRA GARDs and Procedures is to provide information and guidelines for performing PRA. Nevertheless, non-routine risk assessments are often unique, requiring departure from these guidelines and information in order to correctly perform and meet the risk assessment objectives. Such departure must be evaluated and documented in accordance with applicable regulations and Dominion policies.

An administratively controlled process is used to maintain configuration control of the MPS2 PRA models, data, and software. In addition to model control, administrative mechanisms are in place to assure that plant modifications, procedure changes, system operation changes and industry operating experience (OE) are appropriately screened, dispositioned and scheduled for incorporation into the model. These processes help assure that the MPS2 PRA reflects the as-built, as-operated plant within the limitations of the PRA methodology.

The process for performing PRA involves a periodic review and update cycle to model any changes in the plant design or operation. Plant hardware and procedure changes are reviewed on an approximate quarterly or more frequent basis to determine if they impact the PRA and if a PRA model and/or documentation change is warranted. These reviews are documented, and if any PRA changes are warranted, they are added to the PRA Configuration Control (PRACC) database for PRA implementation tracking.

As part of the PRA evaluation for each STI change request, a review of open items in the PRACC database will be performed and an assessment of the impact on the results of the application will be made prior to presenting the results of the risk analysis to the expert panel. If a non-trivial impact is expected, then this may include the performance of additional sensitivity studies or PRA model changes to confirm the impact on the risk analysis.

The Level 1 and Level 2 MPS2 PRA analyses were originally developed and submitted to the NRC in 1993 as the Individual Plant Examination (IPE) Summary Report (Ref.

6.10). In response to Supplement 4 of Generic Letter 88-20, the IPE External Events (IPEEE) Summary Report was submitted to the NRC in 1995 (Ref. 6.11). The MPS2 PRA has been updated many times since the original IPE.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 6 of 33 Since 1995, updates have been made to incorporate plant and procedure changes, update plant-specific reliability and unavailability data, improve the fidelity of the model, incorporate Combustion Engineering Owners Group (CEOG) Peer Review comments and support other applications, such as On-line Maintenance, Risk-Informed In-Service Inspection (RI-ISI), Maintenance Rule Risk Significance, and Mitigating System Performance Index (MSPI).

The enhancements to the MPS2 PRA model include a major internal flooding update and number of updates to the Level 2 PRA model to allow a more realistic assessment of the Large Early Release Frequency (LERF). A summary of the MPS2 PRA history is listed below.

Date Model Change 12/93 IPE submitted 05/94 Supplement regarding a potential vulnerability identified in IPE submittal 09/95 Responses to RAIs on the IPE submittal provided 12/95 IPEEE submitted 05/96 IPE approved by NRC 11/99 CEOG peer review report completed 01/00 PRA model updated - Plant-specific data incorporated 06/00 PRA model updated - Addressed significant peer review comments 01/01 IPEEE approved by NRC 04/01 PRA model updated - Incorporated design change to electrically separate from Unit 1 and connect to Unit 3 12/05 PRA model updated - Plant-specific data incorporated 10/07 Initial PRA self-assessment performed 01/11 PRA model updated - Addressed not met ASME/ANS supporting requirements 02/11 Updated PRA self-assessment based on latest PRA model and regulatory requirements 3.4 Comprehensive Critical Reviews The MPS2 PRA model has benefited from the comprehensive technical PRA Peer Reviews:

CEOG PRA Peer Review The MPS2 internal events PRA received a formal industry PRA Peer Review in 1999 (Ref. 6.1). The purpose of the PRA Peer Review process was to provide a method for establishing the technical quality of a PRA for the spectrum of potential risk-informed plant licensing applications for which the PRA may be used. The PRA Peer Review process used a team composed of industry PRA and system analysts, each with

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 7 of 33 significant expertise in both PRA development and PRA applications. This team provided both an,-o-bjective review of the PRA technical elements and a subjective assessment, based on their PRA experience, regarding the acceptability of the PRA elements. The team used a set of checklists as a framework within which to evaluate the scope, comprehensiveness, completeness, and fidelity of the PRA products available.

The MPS2 review team used the NEI-00-02 "PRA Peer Review Process Guidance" as the basis for the review.

The general scope of the implementation of the PRA Peer Review included review of eleven main technical elements, using checklist tables (to cover the elements and sub-elements), for an at-power PRA including internal events, internal flooding, and containment performance (with focus on LERF).

The findings and observations from the PRA Peer Review were prioritized into four categories (A through D) based upon importance to the completeness of the model.

With the exception of one Category B comment, all comments in Categories A and B have been addressed. The remaining Category B comment is listed in Section 3.5.

MPS2 PRA Self Assessment Reference 6.3 documents the results of a self assessment/independent review of the MPS2 PRA model, data, and documentation in accordance with the Capability Category II requirements of the ASME Standard for PRA (Ref. 6.6) and RG 1.200 (Ref. 6.7). The initial review was performed by Dominion in 2007 with support from a contracting company, MARACOR, using a team of experts with experience in performing NEI PRA Certifications and ASME PRA Standard Reviews. The assessment included a review of the Dominion PRA procedures, current documentation notebooks, and other documentation.

The intent of this independent assessment was to provide a basic assessment of the current PRA against the ASME standard and the RG to determine if each of the requirements of Capability Category II had been met and documented. The assessment team reviewed the technical adequacy of compliance with each of the requirements as compared to current PRA practices in the industry. Insights gained from recent industry programs to comply with the ASME standard were also used.

All technical areas, described in Section 4 of the ASME standard and RG 1.200, have been reviewed, with the exception of the PRA Configuration Control Program. During this review, specific "Facts and Observations" (F&Os) were not generated. However, specific recommendations were provided for each supporting requirement, which was assessed as not met by the current PRA model and documentation. These recommendations were entered into the PRACC database and will be used directly to guide future PRA enhancement activities. The PRACC database is being used to track each supporting requirement that was assessed as not met in a corresponding database

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 8 of 33 item. Of the 328 supporting requirements, the MPS2 PRA does not meet 39 Category II supporting requirements. Section 3.5 lists the supporting requirements not met after the M209Aa model update. The self-assessment was performed against the previous ASME standard (Ref. 6.7), but Section 3.5 lists the supporting requirement numbers from the current ASME/ANS standard (Ref. 6.12).

3.5 Identify Gaps between PRA Model and applicable PRA standard References 6.2 and 6.3 contain the gap analysis between the PRA capability and PRA standards (i.e., CEOG peer review and ASME standards). There are 39 ASME standard supporting requirements not met and one peer review element not met. Of the 40 total elements not met, 14 could impact the RITS 5b application while the remaining 26 pertain only to documentation requirements. Table 1 groups these 14 not met supporting requirements into eight categories and evaluates the impact of the gap on the RITS 5b application. If the gap potentially affects components that could be subject to the RITS 5b application, then a sensitivity study will be performed as part of the surveillance frequency change evaluation. Table 2 lists the gaps and provides an assessment of the potential impact on implementation of the SFCP or RITS 5b.

It is important to note that for each element in the ASME PRA Standard there is a separate high level requirement for documentation. Dominion made the decision in order to meet Category II for a supporting requirement, there had to be documented evidence that the supporting, requirement was met. Since each high level requirement of the standard has a separate documentation part, the supporting requirement could have been categorized as met with the documentation part categorized as not met.

Dominion's approach was to conservatively categorize the supporting requirement as not met due to documentation issues. Therefore, there are numerous technical supporting requirements that are "not met" for lack of documentation. For example, IE-A6 is not met due to the lack of documented evidence for plant personnel interviews. Dominion agrees that documentation is essential in maintaining PRAs and understanding the results.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 9 of 33 Table 1 - Justification for Gao not Imoactina RITS 5b Aoolication Table . . ............ tio n o ..r ...

no ... ......in ..R I, .. . . r l.... ...

Element Element Description Review Comment Importance to Application Not Met IE-C1 b CREDIT recovery actions [those implied in Recovery, actions are appropriately credited in As part of the 2009 model update, the IE

[IE-C3] IE-C4(c), and those implied and discussed the initiating event (IE) analysis and each such series notebooks were revised to address in IE-C6 through IE-C9] as appropriate. credit is justified (all credited actions are the supporting requirements not met in the JUSTIFY each such credit (as evidenced proceduralized). However, the station blackout self assessment. The SBO model changes such as through procedures or training). (SBO) initiating event fault tree logic includes recommended in the self assessment for this the potential to align to MPS3 power supporting requirement will not be made transformers or the SBO diesel. Such actions because the SBO accident sequence would occur after the SBO initiating event development would not change if a separate (available response times for the actions are node was added to the SBO event tree to approximately 100 minutes) and would appear include starting aligning the SBO diesel or to be more appropriately modeled in the post- power from the other unit.

initiator portions of the SBO logic (e.g., the power recovery function). This gap has no impact on the RITS 5b application.

AS-1 0 Dependencies among top events are Main Feedwater success criteria do not require Given that there are four steam dump valves identified and addressed. makeup to the condenser when steam dump with only one valve required to provide valves fail. No documentation of the verification adequate condenser inventory and the main that adequate volume exists in the condenser feedwater pumps rely on the same support for successful cooldown. No modeling of systems as the steam dump valves (i.e.,

makeup to the condenser was identified. Instrument Air (IA) and Main Condenser), the impact of adding the steam dump valves as a required support system for Main Feedwater has an insignificant impact on the overall model.

The steam dump valves are not required by technical specifications and are therefore, not in-scope to the RITS 5b process.

Consequently, this gap has no impact on the RITS 5b application.

AS-A7 DELINEATE the possible accident Anticipated Transient Without Scram (ATWS) These issues have been addressed with the sequences for each modeled initiating does not consider the time of adverse exception of the comment regarding throttling event, unless the sequences can be shown moderator temperature coefficient (MTC). Loss AFW after restoration of power following an to be a non-contribution using qualitative of seal cooling, loss of all AC (SBO), inadvertent SBO. Not modeling the operator action has arguments. opening of power-operated relief valves an insignificant impact based on the two (PORVs) and safety relief valves (SRVs) are consequences of not throttling AFW, which

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 10 of 33.

Table I - Justification for Gap not Impacting RITS 5b Application Element Element Description Review Comment Importance to Application Not Met included in some, but not all event tree models. are:

Assumption 8 in AS.1 states-that operator action to throttle auxiliary feedwater (AFW) after power

• Premature draining of the Condensate restoration following a SBO is assumed Storage Tank, which is mitigated with successful. No justification is provided for offsite power available by supplying fire omitting this sequence. water to the suction of the AFW pumps.

0 Potential steam generator overfill, which could lead to failure of main steam line piping and therefore, loss of secondary heat removal capability. However, with offsite power available, once through cooling would be available to remove decay heat.

The AFW throttle valves are potentially subject to the RITS 5b application.

Therefore, ifa change to the AFW throttle valve surveillance frequency is being evaluated as part of the RITS 5b process, a sensitivity study would be required to evaluate the impact of this gap.

SY-A19 IDENTIFY system conditions that cause a Room heatup calculations are planned to be During the 2009 model update, the model

[SY-A21] loss of desired system function (e.g., performed as a part of the next MPS2 model and documentation were updated to address SY-B6 excessive heat loads, excessive electrical update. However, that documentation does not the supporting requirements not met. The SY-B7 loads, excessive humidity, etc.). appear to exist currently, or is not readily failure of load shedding was added to the accessible. Also, no mention is made of electric power fault tree. The accidents that PERFORM engineering analyses to electrical load shedding or excessive humidity result in excessive humidity such as steam determine the need for support systems that conditions that could lead to a loss of function. line breaks (SLB) include failures of are plant-specific and reflect the variability equipment where the SLB occurs. Room in the conditions present during the heatup calculations have been performed for postulated accidents for which the system is the most risk significant rooms (i.e.,

required to function. switchgear rooms) and ventilation failures included in the model as appropriate.

BASE support system modeling on realistic success criteria and timing, unless a I The ventilation systems components are

Serial No. 11:-687 Docket No. 50-336 Attachment 2, Page 11 of 33 Table I - Justification for Gap not Impacting RITS 5b Application Element Element Description Review Comment Importance to Application Not Met conservative approach can be justified (i.e., potentially subject to the RITS 5b application.

iftheir use does not impact risk significant Therefore, ifa change to ventilation system contributors). components surveillance frequency is being evaluated as part of the RITS 5b process, a sensitivity study would be required to evaluate the impact of this gap.

SY-A20 TAKE CREDIT for system or component The Component Cooling (CC) notebook Room heatup calculations have been

[SY-A22] operability only if an analysis exists to mentions that a GOTHIC analysis was performed for the DC switchgear rooms and demonstrate that rated or design performed which stated that room cooling for ventilation failures included in the model as capabilities are not exceeded. the DC switchgear is needed only for equipment appropriate.

which requires DC power for more than one hour. However, the suggestion has been made The ventilation systems components are that this analysis needs to be reviewed and potentially subject to the RITS 5b application.

other room heatup calculations need to be Therefore, ifa change to ventilation system performed. components surveillance frequency is being evaluated as part of the RITS 5b process, a sensitivity study would be required to evaluate the impact of this gap.

LE-Al IDENTIFY those physical characteristics at Include steam generator (SG) characteristics As part of the 2009 model update, the PDS LE-C5 the time of core damage that can influence and containment. isolation status in the Plant tree was revised to specifically include

[LE-C6] LERF. Examples include (a) RCS pressure Damage State (PDS) binning, unless availability of feedwater, which affects SG LE-D6 (high RCS pressure can result in'high justification can be given for excluding them. level. SG pressure is addressed in the

[LE-D7] pressure melt ejection) (b) status of SG characteristics are necessary for accurate Containment Event Tree (CET), which uses emergency core coolant systems (failure in induced steam generator tube rupture (SGTR) the NUREG-1570 methodology. This injection can result in a dry cavity and and SGTR initiating event LERF calculation, methodology bases the probability on the extensive Core Concrete Interaction) (c) and containment isolation may be required ifthe failure to close probability of an atmospheric status of containment isolation (failure of valve closure has dependencies on other dump valve (ADV). Since no support isolation can result in an unscrubbed systems modeled in the Level 1 (e.g., isolation systems are required to close an ADV (i.e.,

release) (d) status of containment heat signal dependency on DC power and actuation they fail close on loss of air or power), there removal (e) containment integrity (e.g., logic). Include consideration of Emergency is no interaction with Level 1 and therefore it vented, bypassed, or failed) (f) steam Core Cooling System (ECCS) / Low Pressure is appropriate to put it in the CET. However, generator pressure and water level (PWRs) Safety Injection (LPSI) availability. containment isolation does requir#esome (g) status of containment inerting (BWRs). support systems so it should be in the PDS tree using bridge trees.

DEVELOP system models that support the accident progression analysis consistent Per FSAR Table 5.2-11, all Containment

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 12 of 33 Table I - Justification for Gap not Impacting RITS 5b Application Element Element Description Review Comment Importance to Application Not Met I I with the applicable requirements for Isolation Valves (CIVs) that are not normally Paragraph 4.5.4, as appropriate for the level locked closed and are required to close post-of detail of the analysis. accident are fail-closed valves. Therefore, the only support system required for success PERFORM containment isolation analysis is the Engineered Safeguards Actuation in a realistic manner for the significant System (ESAS), which produces the accident progression sequences resulting in Containment Isolation Actuation Signal a large early release. USE conservative or (CIAS). Penetrations with twoactive CIVs a combination of conservative or realistic are treated as separate trains and therefore, treatment for the non-significant accident receive train-specific CIAS signals.

progression sequences resulting in a large Consequently, for the containment isolation early release. INCLUDE consideration of function to fail, both trains of CIAS would both the failure of containment isolation need to fail. As a result, this is considered systems to perform properly and the status an insignificant risk contributor.

of safety systems that do not have automatic isolation provisions. There are CIVs that open post-accident, which require support systems and operator action to close. The majority of the penetrations with these CIVs contain a check valve on the inside of containment, which require no operator action or support system to close. These penetrations are considered insignificant risk contributors.

The only exceptions are the penetrations that contain the Containment Sump Isolation motor-operated valves (MOVs) as they do not contain an inside CIV. These MOVs open on a Sump Recirculation Actuation Signal (SRAS), which correspondsto low Refueling Water Storage Tank (RWST) level, to provide suction to the High Pressure Safety Injection (HPSI) and Containment Spray (CS) pumps during the sump recirculation phase. Failure of the open function is a sianificant core damaae risk

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 13 of 33 Table I - Justification for Gap not Impacting RITS 5b Application Element Element Description Review Comment Importance to Application Not Met contributor. In a core damage scenario, this penetration will either be full of water or closed (SRAS not generated) and therefore, does not represent a significant risk contributor.

The CIVs are potentially subject to the RITS 5b application. Therefore, ifa change to the CIV surveillance frequency is being evaluated as part of the RITS 5b process, a sensitivity study would be required to evaluate the impact of this gap.

LE-C2a INCLUDE realistic treatment of feasible IPE Table 4.8-3 is titled "Operator Action Basic Human Reliability Analysis (HRA)

[LE-C2] operator actions following the onset of core Events", but no values for the actions are calculations were performed for the operator -

LE-C6 damage consistent with applicable provided, and no detailed human error actions credited; ensuring dependencies with

[LE-C7] procedures, e.g., Emergency Operating probabilities (HEP) calculation appears to have other operator actions are accounted for.

Procedures (EOPs)/Severe Accident been performed. Per Table 4.8-4, a basic event Management Guidelines (SAMGs), probability of 0.1 was assigned to the probability The SAMGs have not yet been incorporated proceduralized actions, or Technical of in-vessel recovery due to recovery of reactor into the Level 2 model. However, the impact Support Center guidance. pressure vessel (RPV) injection after core of not meeting this supporting requirement is damage. No evaluation of the operator action is that the current model is conservative.

In crediting Human Failure Events (HFEs) provided (the value was based on a value used that support the accident progression in NUREG-4551). The SAMGs have not been This gap has no impact on the RITS 5b analysis, USE the applicable requirements reviewed for potential impact on the LERF, application.

of Paragraph 4.5.5, as appropriate for the while certain actions could significantly affect level of detail of the analysis. the LERF. For example, opening RCS PORV prior to core damage can significantly reduce the chance of an induced SGTR.

LE-C2b REVIEW significant accident progression IPE Section 4.8.2 considers recovery events. It The sequences have not been reviewed for

[LE-C3] sequences resulting in a large early release states that all recovery actions that involve AC options available to reduce the LERF.

LE-C8b to determine if repair of equipment can be power (HPSI, LPSI, CS, and Containment Air However, the impact of not meeting this

[LE-C10] credited. JUSTIFY credit given for repair Recirculation (CAR) fan coolers) are accounted supporting requirement is that the current

[i.e., ensure that plant conditions do not for in the Level 1 analysis. For other recoveries, model is conservative.

preclude repair and actuarial data exists Table 4.8.2-1 presents some recoveries, but the from which to estimate the repair failure text indicates that these were treated as "house This gap has no impact on the RITS 5b I probability (see SY-A22, DA-C14, and DA- gates" that were set to zero. The CET used to application.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 14 of 33 Table I - Justification for Gap not Impacting RITS 5b Application Element Element Description Review Comment Importance to Application Not Met D8)]. AC power recovery based on generic quantify the MPS2 Level 2 could not be found data applicable to the plant is acceptable. by Dominion, so the actual modeling could not be reviewed.

REVIEW significant accident progression sequences resulting in a large early release to determine if engineering analyses can support continued equipment operation or operator actions during accident progression that could reduce LERF. USE conservative or a combination of conservative and realistic treatment for non-significant accident progression sequences. I

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 15 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 6b Application Number Not Met 1 IE-A6 INTERVIEW plant personnel (e.g., No documentation of plant personnel interviews Documentation issue only, no

[IE-A8] Operations, Maintenance, Engineering, to determine if potential initiating events have impact on application.

Safety Analysis) to determine if potential been overlooked was found in the PRA initiating events have been overlooked, notebooks.

2 IE-Clb CREDIT recovery actions [those implied in Recovery actions are appropriately credited in Potential logic model issue. The

[IE-C3] IE-C4(c), and those implied and discussed the IE analysis and each such credit is justified impact of not meeting this in IE-C6 through IE-C9] as appropriate. (all credited actions are proceduralized). element on the RITS 5b JUSTIFY each such credit (as evidenced However, the SBO initiating event fault tree application is required to be such as through procedures or training, logic includes the potential to align to MPS3 reviewed.

power transformers or the SBO diesel. Such actions would occur after the SBO initiating event (available response times for the actions are approximately 100 minutes) and would appear to be more appropriately modeled in the post-initiator portions of the SBO logic (e.g., the power recovery function).

3 AS-10 Dependencies among top events are Main Feedwater success criteria do not require Potential logic model issue. The identified and addressed. makeup to the condenser when steam dump impact of not meeting this valves fail. No documentation of the verification element on the RITS 5b that adequate volume exists in the condenser application is required to be for successful cooldown. No modeling of reviewed.

makeup to the condenser was identified.

4 AS-A4 For each modeled initiating event, using the In general, no summary or descriptions are Documentation issue only, no success criteria defined for each key safety provided foe operator actions in either SC. 1 or impact on application.

function (in accordance with SR SC-A4), AS.1.

IDENTIFY the necessary operator actions to achieve the defined success criteria.

5 AS-A7 DELINEATE the possible accident ATWS does not consider the time of adverse Potential logic model issue. The sequences for each modeled initiating MTC. Loss of seal cooling, loss of all AC (SBO), impact of not meeting this event, unless the sequences can be shown inadvertent opening of PORVs and SRVs are element on the RITS 5b to be a non-contribution using qualitative included in some, but not all event tree models. application is required to be I arguments. Assumption 8 in AS. 1 states that operator action reviewed.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 16 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met to throttle AFW after power restoration following a SBO is assumed successful. No justification is provided for omitting this sequence.

6 AS-A10 In constructing the accident sequence While differences in system requirements for Documentation issue only, no models, INCLUDE, for each modeled each initiating event may be included in the fault impact on application.

initiating event, sufficient detail that tree models, no delineation of how these significant differences in requirements on differences impact operator actions or system systems and operator responses are responses is provided. For example, the captured. Where diverse systems and/or success criteria for bleed and feed cooling are operator actions provide a similar function, if different between the General Plant Transient choosing one over another changes the (GPT) and Main Feedwater (MFW) event requirements for operator intervention or the models, however, no discussion is provided as need for other systems, MODEL each to why.

separately.

7 AS-B3 For each accident sequence, IDENTIFY the Only a limited discussion of phenomenological Documentation issue only, no phenomenological conditions created by the conditions created by the accident progression impact on application.

accident progression. Phenomenological is provided in Section 2.3 of Volume AS.1. For impacts include generation of harsh example, the discussion provided on how a The IAcompressors, 4160V and environments affecting temperature, secondary line break outside containment 480V switchgear rooms, and pressure, debris, water levels, humidity, etc. affects the environmental conditions of AFW system are located in the that could impact the success of the system equipment needed to mitigate the accident turbine building. Following a or function under consideration [e.g., loss of discusses the loss of IA,but no discussion is secondary line break outside pump net positive suction head (NPSH), provided on the direct impact of a loss of MFW containment, the IA compressors clogging of flow paths]. INCLUDE the or any potential impact on AFW or the electrical are expected to fail since they are impact of the accident progression switchgear rooms. not rated for a High Energy Line phenomena, either in the accident Break (HELB) environment. IA is sequence models or in the system models. a required support system for MFW; therefore, this dependency is directly accounted for in the system fault trees. The switchgear rooms are housed in Class I structures equipped with HELB doors and therefore, will

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 17 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met not be affected by a secondary line break. The AFW pumps and regulating valves are rated for a HELB environment and therefore, would not be affected by a secondary line break.

8 AS-B5a If plant configurations and maintenance The MPS2 model discusses how system Documentation issue only, no

[AS-B6] practices create dependencies among configurations impact modeling in the system impact on application.

various system alignments, DEFINE and notebooks under the "Risk Monitor MODEL these configurations and Considerations" section. However, no alignments in a manner that reflects these discussion is provided on how system dependencies, either in the accident alignments and configurations are applied when sequence models or in the system models. evaluating the PRA models outside of risk monitors.

9 AS-C2 DOCUMENT the processes used to A one-to-one correlation between each initiating Documentation issue only, no develop accident sequences and treat event and the associated event tree is not impact on application.

dependencies in accident sequences, clearly provided. The system success criteria including the inputs, methods, and results. and associated basis is not clearly provided. A For example, this documentation typically discussion of the accident sequences will need includes: (a) the linkage between the to be revised pending resolution of issues modeled initiating event in the Initiating associated with other AS supporting Event Analysis section and the accident requirements. For example, the sequence model; (b) the success criteria phenomenological conditions created by the established for each modeled initiating accident. Operator actions needed are not event including the bases for the criteria clearly delineated along with any associated (i.e., the system capacities required to dependencies on system success or other mitigate the accident and the necessary operator actions (Refer to AS-B1, B3, and B6).

components required to achieve these capacities); (c) a description of, the accident progression for each sequence or group of similar sequences (i.e., descriptions of the sequence timing, applicable procedural I _guidance, expected environmental or

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 18 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met phenomenological impacts, dependencies between systems and operator actions, end states, and other pertinent information required to fully establish the sequence of events); (d) the operator actions reflected in the event trees, and the sequence-specific timing and dependencies that are traceable to the HRA for these actions; (e) the interface of the accident sequence models with plant damage states; (f) [when sequences are modeled using a single top event fault tree] the manner in which the requirements for accident sequence analysis have been satisfied.

10 SC-135 CHECK the reasonableness and While the SC.1 and SC.2 make some Documentation issue only, no acceptability of the results of the comparisons to results from other plants (e.g., impact on application.

thermal/hydraulic, structural, or other Calvert Cliffs Interim Reliability Evaluation supporting engineering bases used to Program (IREP)) for specific success criteria, support the success criteria. Examples of there is no documented comparison of the methods to achieve this include: (a) - overall set of MP2 success criteria to those of comparison with results of the same other plants. Also, as the Calvert Cliffs IREP analyses performed for similar plants, has been superseded by more recent models, accounting for differences in unique plant references to this older study may no longer be features (b) comparison with results of appropriate.

similar analyses performed with other plant-specific codes (c) check by other means appropriate to the particular analysis.

11 SY-A4 PERFORM plant walkdowns and interviews While the IPE documentation and conversations Documentation issue only, no with knowledgeable plant personnel (e.g., with the PRA engineers indicate that these impact on application.

Engineering, Operations, etc.) to confirm tasks were performed, no documentation exists that the systems analysis correctly reflects (walkdown sheets, system engineer interviews) the as-built, as-operated plant. to support this supposition.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 19 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met 12 SY-A19 IDENTIFY system conditions that cause a Room heatup calculations are planned to be Potential logic model issue. The

[SY-A21] loss of desired system function (e.g., performed as a part of the next MPS2 model impact of not meeting this excessive heat loads, excessive electrical update. However, that documentation does not element on the RITS 5b loads, excessive humidity, etc.). appear to exist currently, or is not readily application is required to be accessible. Also, no mention is made of reviewed.

electrical load shedding or excessive humidity conditions that could iead to a loss of function.

13 SY-A20 TAKE CREDIT for system or component The Component Cooling (CC) notebook Potential logic model issue. The

[SY-A22] operability only ifan analysis exists to mentions that a GOTHIC analysis was impact of not meeting this demonstrate that rated or design performed which stated that room cooling for element on the RITS 5b capabilities are not exceeded. the DC switchgear is needed only for equipment application is required to be which requires DC power for more than one reviewed.

hour. However, the suggestion has been made that this analysis needs to be reviewed and other room heatup calculations need to be performed.

14 SY-B6 PERFORM engineering analyses to As per SY-A19, room heatup calculations have Potential logic model issue. The determine the need for support systems that not been performed. Systems that could fail impact of not meeting this are plant-specific and reflect the variability based on excessive heat have not been element on the RITS 5b in the conditions present during the properly documented. application is required to be postulated accidents for which the system is reviewed.

required to function.

15 SY-B7 BASE support system modeling on realistic As per SY-A1 9, room heatup calculations have Potential logic model issue. The success criteria and timing, unless a not been performed. Systems that could fail impact of not meeting this conservative approach can be justified (i.e., based on excessive heat have not been element on the RITS 5b iftheir use does not impact risk significant properly documented. application is required to be

-contributors). reviewed.

16 SY-B12 MODEL the ability of the available The system models for CC and IA do not Documentation issue only, no

[SY-B1 1] inventories of air, power, and cooling to appear to take credit for insufficient inventories, impact on application.

support the mission time. However, documentation of that appears I_ insufficient.

17 SY-C2 DOCUMENT the system functions and No walkdown information, documentation of Documentation issue only, no boundary, the associated success criteria, operating history, or room heatup calculations impact on application.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 20 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met I I I the modeled components and failure modes exist.

including human actions, and a description of modeled dependencies including support system and common cause failures, including the inputs, methods, and results.

For example, this documentation typically includes (a) system function and operation under normal and emergency operations (b) system model boundary (c) system schematic illustrating all equipment and components necessary for system operation (d) information and calculations to support equipment operability considerations and assumptions (e) actual operational history indicating any past problems in the system operation (f) system success criteria and relationship to accident sequence models (g) human actions necessary for operation of system (h) reference to system-related test and maintenance procedures (i) system dependencies and shared component interface () component spatial information (k) assumptions or simplifications made in development of the system models (I) the components and failure modes included in the model and justification for any exclusion of components and failure modes (m) a description of the modularization process (if used) (n) records of resolution of logic loops developed during fault tree linking (if used)

(o) results of the system model evaluations (p) results of sensitivity studies (if used) (q)

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 21 of 33 Table2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met the sources of the above information (e.g.,

completed checklist from walkdowns, notes from discussions with plant personnel) (r) basic events in the system fault trees so that they are traceable to modules and to cutsets (s) the nomenclature used in the system models.

18 DA-C10 When using surveillance test data, REVIEW There is no evidence in notebook DA.2 to Documentation issue only, no the test procedure to determine whether a indicate that a review of the surveillance impact on application. During the test should be credited for each possible procedures was performed. 2009 model update, thisdata was failure mode. COUNT only completed tests obtained based on real plant data or unplanned operational demands as obtained from the station logs and success for component operation. If the the plant computer. Therefore, component failure mode is decomposed review of the procedures is not into sub-elements (or causes) that are fully necessary.

tested, then USE tests that exercise specific sub-elements in their evaluation. Thus, one sub-element sometimes has many more successes than another. [Example: a diesel generator is tested more frequently than the load sequencer. IF the sequencer were to be included in the diesel generator boundary, the number of valid tests would be significantly decreased.]

19 DA-C15 Data on recovery from loss of offsite power, The DOM IE.2 notebook presents Offsite Power Documentation issue only, no

[DA-C16] loss of service water, etc. are rare on a (OSP) frequencies with recovery presented in impact on application. There plant-specific basis. If available, for each DOM HR.3 for all Dominion plants. OSP were no plant specific LOOP recovery, COLLECT the associated Recovery is calculated in DOM HR.3, but is not events for MPS2 for the update recovery time with the recovery time being discussed (only presented in a spreadsheet).. period. Therefore, no plant-the period from identification of the system No specific assessment of the applicability of specific recovery times are or function failure until the system or the events considered to the Millstone site is available.

function is returned to service. provided.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 22 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category Il of ASME PRA Standard

_ Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met 20 IFPP-A3 For multi-unit sites with shared systems or MPS2 is physically separate from MPS3 and Documentation issue only, no.

structures, INCLUDE multi-unit areas, if shares no fluid systems or structures with impact on application.

applicable. MPS3. There is no potential for multi-unit flood scenarios; however, the documentation in notebook IF.1 should include discussion of why multi-unit flood areas (and scenarios) are not relevant for MPS2.

21 IFSO-A5 For each source and its identified failure The IF.1 and IF.2 notebooks consider leaks, Documentation issue only, no mechanism, IDENTIFY the characteristic of ruptures and spray. The analysis generally- impact on application.

release and the capacity of the source. considers sources of all sizes, which bounds the INCLUDE (a) a characterization of the range of flow rates. The capacity of each source breach, including type (e.g., leak, rupture, is considered qualitatively or quantitatively and spray) (b) flow rate (c) capacity of source -potentially large sources are considered for their (e.g., gallons of water) (d) the pressure and resulting impacts on the extent of flooding and temperature of the source. propagation. Capacities of flood sources are also considered further in the IF.2 notebook.

The documentation does not, however, discuss the pressures and temperatures of the sources.

While most of the flood sources are relatively low temperature sources (e.g., service water, fire protection, etc.), high energy fluid sources are not highlighted, nor is there any discussion of whether the special characteristics of these sources might have unique plant effects.

22 QU-B5 Fault tree linking and some other modeling The MPS2 QU.1 and QU.2 notebooks do not Documentation issue only, no approaches may result in circular logic that include any discussion of the approach used for impact on application.

must be broken before the model is solved, breaking circular logic loops. (The discussion in BREAK the circular logic appropriately. QU.1 Attachment 1 on Revision 4 does mention Guidance for breaking logic loops is' that changes were made to system fault trees to provided in NUREG/CR-2728 [Note (1)]. . correct circular logic related to consequential When resolving circular logic, AVOID Loss of Coolant Accidents (LOCAs), and introducing unnecessary conservatisms or Section 2.2.1 of QU.2 notes that logic loops non-conservatisms. related to DC ventilation changes were

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 23 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met addressed. Finally, Tables 17-19 in QU.2 identify changes in model results due to removal of logic loops.) Table I of the systems analysis assumptions notebook (SY.2) includes several specific entries regarding the creation of circular logic cut gates to break logic loops in the AC, DC, Engineered Safety Feature Actuation System (ESFAS), Heating, Ventilation and Air Conditioning (HVAC), IA, and Service Water systems, but there is no discussion on where/how these gates break the logic. Instead, these assumptions and the response to comment 7 in Attachment 2 to SY.2 refer to documentation on the necessity of these gates in the final quantification documentation (QU.2),

but no such documentation exists.

23 QU-E3 ESTIMATE the uncertainty interval of the No parametric uncertainty analysis has been Documentation issue only, no overall CDF results. ESTIMATE the performed for the MPS2 PRA. impact on application.

uncertainty intervals associated with parameter uncertainties (DA-D3, HR-D6, HR-G9, IE-C13), taking into account the "state-of-knowledge" correlation.

24 QU-E4 EVALUATE the sensitivity of the results to No evaluation of the model uncertainties and Documentation issue only, no key model uncertainties and key assumptions have been performed or impact on application.

assumptions using sensitivity analyses. documented.

25 LE-Al IDENTIFY those physical characteristics at Include SG characteristics and containment Potential logic model issue. The the time of core damage that can influence isolation status in the PDS binning, unless impact of not meeting this LERF. Examples include (a) RCS pressure justification can be given for excluding them. SG element on the RITS 5b (high RCS pressure can result in high characteristics are necessary for accurate application is required to be pressure melt ejection) (b) status of induced SGTR and SGTR initiating event LERF reviewed.

emergency core coolant systems (failure in calculation, and containment isolation may be injection can result.in a dry cavity and required ifthe valve closure has dependencies extensive Core Concrete Interaction) (c) on other systems modeled in the Level 1 (e.g., I

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 24 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met status of containment isolation (failure of isolation signal dependency on DC power and isolation can result in an unscrubbed actuation logic). Include consideration of release) (d) status of containment heat ECCS/LPSI availability.

removal (e) containment integrity (e.g.,

vented, bypassed, or failed) (f) steam generator pressure and water level (PWRs)

(g) status of containment inerting (BWRs).

26 LE-C2a INCLUDE realistic treatment of feasible IPE Table 4.8-3 is titled "Operator Action Basic Potential logic model issue. The

[LE-C2] operator actions following the onset of core Events", but no values for the actions are impact of not meeting this damage consistent with applicable provided, and no detailed HEP calculation element on the RITS 5b procedures, e.g., EOPs/SAMGs, appears to have been performed. Per Table 4.8- application is required to be proceduralized actions, or Technical 4, a basic event probability of 0.1 was assigned reviewed.

Support Center guidance. to the probability of in-vessel recovery due to recovery of RPV injection after core damage.

No evaluation of the operator action is provided (the value was based on a value used in NUREG-4551). The SAMGs have not been reviewed for potential impact on the LERF, while certain actions could significantly affect the LERF. For example, opening RCS PORV prior to core damage can significantly reduce the chance of an induced SGTR.

27 LE-C2b REVIEW significant accident progression IPE Section 4.8.2 considers recovery events. It Potential logic model issue. The

[LE-C3] sequences resulting in a large early release states that all recovery actions that involve AC impact of not meeting this to determine if repair of equipment can be power (HPSI, LPSI, CAR fan coolers and element on the RITS 5b credited. JUSTIFY credit given for repair containment sprays) are accounted for in the application is required to be

[i.e., ensure that plant conditions do not Level 1 analysis. For other recoveries, Table reviewed.

preclude repair and actuarial data exists 4.8.2-1 presents some recoveries, but the text from which to estimate the repair failure indicates that these were treated as "house probability (see SY-A22, DA-C14, and DA- gates" that were set to zero. The CET used to D8)]. AC power recovery based on generic quantify the MPS2 Level 2 could not be found data applicable to the plant is acceptable. by Dominion, so the actual modeling could not I be reviewed. I

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 25 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gao on Imolemnentation of RITS 5b

. . .. .. . .. ... Im . c. . . . a n . .. .. r . . ... . .... .. . .. . . . .. . . . .

Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met 28 LE-C5 DEVELOP system models that support the System models that affect the accident Potential logic model issue. The

[LE-C6] accident progression analysis consistent progression (e.g., sprays and containment heat impact of not meeting this with the applicable requirements for removal) were developed and documented in element on the RITS 5b paragraph 4.5.4, as appropriate for the level the applicable system analysis notebooks (SY). application is required to be of detail of the analysis. However, a containment isolation fault tree may reviewed.

also be required, as it appears from IPE Section 4.4.5 that the isolation valves may require modeling of dependencies. The containment isolation document (2-PRA-93-032) could not be located by Dominion, so the actual system requirements are not clear.

29 LE-C6 In crediting HFEs that support the accident System level operator actions are described in Potential logic model issue. The

[LE-C7] progression analysis, USE the applicable the Level 1 System Analysis notebooks. Offsite impact of not meeting this requirements of paragraph 4.5.5, as power recovery probabilities are maintained element on the RITS 5b appropriate for the level of detail of the within the Level 1 Data Analysis. SAMGs have application is required to be analysis. not been incorporated into the MPS2 Level 2 reviewed.

analysis, although credit for initiation of low pressure injection after the onset of core damage was combined with hardware failures, and assigned a total probability of 0.1. IPE Table 4.8-3 (page 4-149) shows three other operator action basic events in the Level 2, although no HEP was presented.

30 LE-C8a JUSTIFY any credit given for equipment It appears that some consideration was given, Documentation issue only, no

[LE-C9] survivability or human actions under. as seen on pages 4-140 (#29) and F-10 of the impact on application.

) adverse environments. IPE, which state consideration of containment sprays being failed by the accident progression.

Page F-21 shows a probability of 1E-2 that sprays are failed by the accident progression, although the only basis is an assumption on page 4-162. Other than the sprays, it does not appear that any other equipment survivability I was examined, except that no credit was given

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 26 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met I to operation after containment failure. The equipment survivability should be examined and explicitly discussed to meet the supporting requirement. In general, equipment is capable (and credited) of performing at levels significantly worse than the design basis conditions. For example, even though spray headers and SG equipment are credited up until containment failure (pressures and temperatures far greater than design basis),

they will be subject to worse than design basis conditions in a severe accident. Such credit should be provided in the documentation.

31 LE-C8b REVIEW significant accident progression The significant accident progression sequences Potential logic model issue. The

[LE-Cl0] sequences resulting in a large early release were not reviewed explicitly for the impact of not meeting this to determine if engineering analyses can consideration of continued equipment operation element on the RITS 5b support continued equipment operation or or operator actions to reduce the LERF. application is required to be operator actions during accident reviewed.

progression that could reduce LERF. USE conservative or a combination of conservative and realistic treatment for non-significant accident progression sequences.

32 LE-C9b REVIEW significant accident progression Although a review of the significant accident Documentation issue only, no

[LE-C12] sequences resulting in a large early release progression sequences for post-containment impact on application.

to determine if engineering analyses can failure operation might not identify any potential support continued equipment operation or for LERF reduction, the review should be operator actions after containment failure performed and documented to meet the that could reduce LERF. USE conservative supporting requirement.

or a combination of conservative and realistic treatment for non-significant accident progression sequences.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 27 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met 33 LE-Dl b EVALUATE the impact-of accident IPE Section 4.4.2 notes the consideration of Documentation issue only, no

[LE-D2] progression conditions on containment penetrations, hatch failure, etc. The complete impact on application.

seals, penetrations, hatches, drywell heads report is documented in the 1993 EQE (BWRs), and vent pipe bellows. INCLUDE Engineering calculation 52204-R-002, as these impacts as potential containment referenced in Section 4.4.1 of the IPE. However, challenges, is required. If generic analyses the EQE report was not available for review, are used in support of the assessment, and should be brought into the Dominion JUSTIFY applicability to the plant being document control.

evaluated.

34 LE-D6 PERFORM containment isolation analysis Containment isolation was discussed in IPE Potential logic model issue. The

[LE-D7] in a realistic manner for the significant Section 4.4.5, which references a detailed impact of not meeting this accident progression sequences resulting in evaluation in MPS2 calculation 2-PRA-93-032 element on the RITS 5b a large early release. USE conservative or a (July 1993). However, this calculation could not application is required to be combination of conservative or realistic be located by Dominion, so the details could not reviewed.

treatment for the non-significant accident be reviewed. The IPE states that isolation failure progression sequences resulting in a large is dominated by a 2" line (failure of three air-early release. INCLUDE consideration of operated valves (AOVs)) and two 6" hydrogen both the failure of containment isolation purge lines (two AOVs each). The analysis systems to perform properly and the status needs to consider if the AOVs require an of safety systems that do not have actuation signal; if so, then their fault tree automatic isolation provisions, solutions should be tied to the sequence logic to capture dependencies. The analysis needs to provide a basis for small vs. large containment isolation failures. Also, there are two references in the IPE citing "personal communication" with individuals. References to memoranda or something similar should be provided. Section 2.4 of the AS.1 notebook states that "Since the Containment is operated at sub-atmospheric pressure the probability of Containment bypass as a result of failure to isolate is very low for all sequences. Hence this function has been excluded from individual event trees." This

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 28 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met apparently is an attempt to justify the isolation failure not being linked to the other functions, but a quantitative evaluation would be required to justify such a statement.

35 LE-Fla PERFORM a quantitative evaluation of the QU.2 Section 2.3.2 provided LERF by initiating Documentation issue only, no

[LE-F1] relative contribution to LERF from plant event, Section 2.3.6 presented the dominant impact on application.

damage states and significant LERF LERF cutsets, Section 2.3.9 presents the LERF contributors from Table 4.5.9-3. importance analysis, and Table 15 presents the system contribution to LERF. PRAOOYQA-03015S2, Rev. 1 quantified the LERF by PDS, but the evaluation was in 2001 and the PDS quantification has not been documented for the current results. "Significant LERF contributors" have not been defined in QU.2.

36 LE-F1 b REVIEW contributors for reasonableness Section 2.4.11 of QU.2 examines some Documentation issue only, no

[LE-F2] (e.g., to assure excessive conservatisms potential plant improvements to reduce the impact on application.

have not skewed the results, level of plant CDF, but does not select potential specificity is appropriate for significant improvements based on the dominant LERF contributors, etc.). contributors. Section 2.3.6 presented the dominant LERF cutsets, but did not discuss their potential for excess conservatism.

37 LE-G2 DOCUMENT the process used to identify The PDS documentation was created in the IPE Documentation issue only, no plant damage states and accident and has not been updated even though there impact on application.

progression contributors, define accident have been many updates to the Level 1 progression sequences, evaluate accident analysis. The IPE PDS binning documentation progression analyses of containment does not provide sufficient detail about specific capability, and quantify and review the sequence binning. The CET is documented in LERF results. For example, this the IPE, but it is difficultto follow the exact logic documentation typically includes (a) the or even the exact values used for split fraction plant damage states and their attributes, as basic events for each PDS.

used in the analysis (b) the method used to bin the accident sequences into plant damage states (c) the containment failure I I

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 29 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category II of ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Appl.ication Number Not Met I I I modes, phenomena, equipment failures and human actions considered in the development of the accident progression sequences and the justification for their inclusion or exclusion from the accident progression analysis (d) the treatment of factors influencing containment challenges and containment capability, as appropriate for the level of detail of the analysis (e) the basis for the containment capacity analysis including the identification of containment failure location(s), if applicable (f) the accident progression analysis sequences considered in the containment event trees (g) the basis for parameter estimates (h) the model integration process including the results of the quantification including uncertainty and sensitivity analyses, as appropriate for the level of detail of the analysis.

38 LE-G3 DOCUMENT the relative contribution of The PDS contribution was tallied in PRAQOYQA- Documentation issue only, no contributors (i.e., plant damage states, 03015S2 in 1991, but has not been updated in impact on application.

accident progression sequences, - the current QU or LE notebooks. The QU phenomena, containment challenges, notebooks tabulate LERF by initiating event-and containment failure modes) to LERF. system contribution, but not by the contribution due to various phenomena or containment challenges.

39 LE-G4 DOCUMENT key assumptions and key The IPE Section 4.2.2 presents a list of Documentation issue only, no sources of uncertainty associated with the sensitivities to be evaluated by the Modular impact on application.

LERF analysis, including results and Accident Analysis Program (MAAP) code, but important insights from sensitivity studies. does not actually discuss their evaluation.

However, many sensitivities are mentioned in various subsections, but it would be helpful to

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 30 of 33 Table 2 - Status of Gaps to NEI 00-02 and Capability Category IIof ASME PRA Standard Potential Impact of Gap on Implementation of RITS 5b Gap Element Element Description Review Comment Impact on RITS 5b Application Number Not Met compile a list of the sensitivities performed and present their conclusions. The QU.4 document does a good job of identifying key sources of uncertainty, but does not identify the specific assumptions from the IPE. In the IPE, the assumptions were stated as they were used, but were not tabulated and only a few were selected for sensitivity analysis. QU.4 Table 10 documents sensitivities that vary the HEPs and CCF probabilities, and Table 11 identifies some sensitivities based on Level 1 assumptions.

However, per Table 12, the sensitivities have not been completed, and in any case, no sensitivities were identified based on the Level 2 analysis. The sensitivity analyses should be expanded and should be performed on the updated models.

40 LE-G5 IDENTIFY limitations in the LERF analysis Section 2.4.12 of the QU.2 notebook states that Documentation issue only, no that would impact applications, the QU.4 notebook will identify model impact on application.

I limitations, but they are not identified in QU.4.

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 31 of 33 3.6 External Events Considerations The NEI 04-10, Revision 1 methodology allows for STI change evaluations to be performed in the absence of quantifiable PRA models for all external hazards. For those cases where the STI 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.

The external event considerations were derived from the MPS2 Individual Plant Examination

- External Events (IPEEE) (Ref. 6.11). For events such as fire, seismic, extreme winds and other external events, the risk assessments from the IPEEE can be used for insights on changes to surveillance intervals.

Fire Risk The MPS2 PRA does not include a fire model. Therefore, the results of the fire risk assessment performed for the IPEEE can be qualitatively assessed for insights on changes to surveillance intervals. The IPEEE fire risk analysis quantified a core damage frequency (CDF) by using a combination of Fire Induced Vulnerability Evaluation (FIVE) methodology and Fire PRA. The CDF due to fires is 6.3E-06/yr, with the dominant risk being fires in the auxiliary building, turbine building, cable vault, and intake structure.

Seismic Risk The MPS2 PRA does not include a seismic model. Therefore, the results of the seismic risk assessment performed for the IPEEE can be qualitatively assessed for insights on changes to surveillance intervals. The IPEEE seismic risk analysis used the EPRI Seismic Margins Method to determine seismic vulnerabilities beyond design basis and therefore, did not calculate a seismic CDF. This process utilized a screening process to identify components that are considered not seismically rugged and required further evaluation. STI changes associated with these components would require investigation within the RITS 5b process.

High Winds, Floods and Other External Events The risk of other external events such as high winds, external floods, transportation accidents, and weather-related events were assessed in the MPS2 IPEEE. This process utilized a screening process to identify components that required further evaluation. STI changes associated with these components would require investigation within the RITS 5b process.

3.7 Summary The MPS2 PRA technical capability evaluations and the maintenance and update processes described above provide a robust basis for concluding that the full power internal

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 32 of 33 events MPS2 PRA is suitable for use in risk-informed processes such as that proposed for the implementation of a SFCP. In performing the assessments for the other hazard groups, the qualitative or bounding approach will be utilized in most cases. Also, in addition to the standard set of sensitivity studies required per the NEI 04-10, Revision 1 methodology, open items for changes at the site and remaining gaps to specific requirements in the PRA standard will be reviewed to determine which, if any, would merit application-specific sensitivity studies in the presentation of the application results.

4.0 RESULTS None

5.0 CONCLUSION

S The MPS2 PRA model supports the RITS 5b application.

6.0 REFERENCES

6.1. CE NPSD-1 182-P, Millstone Nuclear Station Unit 2 ProbabilisticSafety Assessment Peer Review Report, Final Report, Task 1037, Combustion Engineering Owners Group, January 2000 6.2. MPS2 Probabilistic Risk Assessment Model Notebook Part IV Support Information, Appendix A - PRA Model Reviews, Revision 2, May 2011 6.3. MPS2 Probabilistic Risk Assessment Model Notebook Part IV, Appendix A. 1, Internal Events Model Self Assessment, Revision 2, February 2011 6.4. Peach Bottom Atomic Power Station, Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance FrequencyRequirements to a Licensee Controlled Program(Adoption of TSTF-425, Revision 3), ADAMS ML092470153, August 31, 2009 6.5. Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guidance Document, NEI 04-10, Revision 1, April 2007 6.6. ASME RA-S-2002, Standard for ProbabilisticRisk Assessment for Nuclear Power Plant Applications, with ASME RA-Sa-2003 and RA-Sb-2005 Addenda, ASME, 2005 6.7. US Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, Revision 1, January 2007 6.8. MPS2 2009 PRA Model, External Release of MPS2 PRA Model M2O9Aa, MEMO-PRA-20110009, Revision 0, August 25, 2011

.6.9. MPS2 Probabilistic Risk Assessment Quality Summary Notebook Part IV Support Information, Appendix B - Quality Summary, Revision 0, May 2011

Serial No.11-687 Docket No. 50-336 Attachment 2, Page 33 of 33 6.11. Millstone Unit 2 Nuclear Power Plant, Individual Plant Examination of External Events, Summary Report, December 29, 1995 6.12. ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency ProbabilisticRisk Assessment for Nuclear Power Plant Applications and its 2009 addendum (ASME/ANS RA-Sa-2009) 6.13. North Anna Power Station, CO-NRC-000-10-122, Virginia Electric and Power Company (Dominion) North Anna Power Station Units 1 and 2 Proposed License Amendment Request Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3), March 30, 2010 6.14. Surry Power Station, CO-NRC-000-10-183, Virginia Electric and Power Company (Dominion) Surry Power Station Units 1 and 2 Proposed License Amendment Request Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee ControlledProgram(Adoption of TSTF-425, Revision 3), March 30, 2010 6.15. Millstone Power Station Unit 3, Adams Accession Number ML11193A225, Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 3 License Amendment Request to Relocate TS Surveillance Frequenciesto Licensee Controlled Programin Accordance with TSTF-425, Revision 3, July 5, 2011 r

Serial No.11-687 Docket No. 50-336 ATTACHMENT 3 Marked-up Technical Specifications Changes DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

VtbVr 27, 2008 INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS Defined Term s .................................................................................................................. 1-1 Therm al Power ................................................................................................................ 1-1 Rated Therm al Power ....................................................................................................... 1-1 Operational M ode ............................................................................................................. 1-1 Action ............................................................................................................................... 1-1 Operable - Operability ................................................................................................. 1-1 Reportable Event .............................................................................................................. 1-1 Containm ent Integrity ...................................................................................................... 1-2 Channel Calibration ........................................................................................................ 1-2 Channel Check ................................................................................................................. 1-2 Channel Functional Test ................................................................................................... 1-2 Core A lteration ................................................................................................................. 1-3 Shutdown M argin ............................................................................................................. 1-3 Leakage ............................................................................................................................ 1-3 Azim uthal Power Tilt ....................................................................................................... 1-4 Dose Equivalent 1-131 ...................................................................................................... 1-4 Dose Equivalent Xe- 133 ................................................................................................. 1-4 t d Test B a ....................................................................................................... 1-4 Frequency N otation .......................................................................................................... 1-4 Axial Shape Index ........................................................................................................... 1-5 Core Operating Lim its Report .......................................................................................... 1-5 MILLSTONE - UNIT 2 I Amendment No. 9, 38, 404, 4-1, 448, 299, a,-

S.ptsomlber 18, 2908 INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.22 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM .................... 6-28 6.23 TECHNICAL SPECIFICATION (TS) BASES CONTROL PROGRAM ....................... 6-28 6.24 DIESEL FUEL OIL TEST PROGRAM .......................................................................... 6-29 6.25 PRE-STRESSED CONCRETE CONTAINMENT TENDON SURVEILLANCE PROGRAM ..................................................................................... 6-29 6.26 STEAM GENERATOR PROGRAM ................................................................................ 6-30 6.27 CONTROL ROOM HABITABILITY PROGRAM ......................................................... 6-32 /

ý -- 6.28 Surveillance Frequency Control Program .................................................... 6-33 MILLSTONE - UNIT 2 XVIII Amendment No. 2-49, 99, -305

O... e- 27, 200 DEFINITIONS AZIMUTHAL POWER TILT - Tq 1.18 AZIMUTHAL POWER TILT shall be the difference between the maximum power generated in any core quadrant (upper or lower) and the average power of all quadrants in that half (upper or lower) of the core divided by the average power of all quadrants in that half (upper or lower) of the core.

AZIMUTHALPOWERTILT Maximum power in any core quadrant (upper or lower)1 1 Average power of all quadrants (upper or lower) _

DOSE EQUIVALENT 1-131 1.19 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (micro-curie/gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No.

11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

DOSE EQUIVALENT XE-133 1.20 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (micro-curie/gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111. 1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

rTG% G -9EDA- TEST BAS I

• 4*-DE LEfT-ED 1.21 A STAGGERED TEST BASIS sh&T -e-co-..i.t

! Of,

a. A ta.t wchdula for a ssrtems, suboy:s.taf, U'mi"n or cthcr de*,"nitd cmponcr*,

obtained by dividinig tho speeified fast 448cR,61 into " equal subinfteryal, and-

b. Thotzein yA~tz, Stligy~teff, trainf @r ether deoignjAtcd cmRponetS# Fit th61 Hfon beginning of eneh gubimtcral.

FREOUENCY NOTATION 1.22 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

MILLSTONE - UNIT 2 1-4 Amendment No. 4-04, 246, M.8, 4

orgwa: J=ugu. J I, y99 Original Jauah 1, 196 TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

w At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 6 months.

R At least once per 18 months.

S/U Prior to each reactor startup.

P Prior to each release.

N.A. Not applicable.

1' At the frequency specified in the Surveillance Frequency Control Program.

MILLSTONE - UNIT 2 1-9 Amendment No. 4-41-4

Septzmber 25, 2003 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1. 1 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - (SDM)

LIMITING CONDITION FOR OPERATION

'I 3.1.1.1 The SHUTDOWN MARGIN shall be within the limit specified in the CORE OPERATING LIMITS REPORT. /

APPLICABILITY: MODES 3(1)*, 4 and 5.

ACTION:

With the SHUTDOWN MARGIN not within the limit specified in the CORE OPERATING LIMITS REPORT, within 15 minutes, initiate and continue boration at > 40 gpm of boric acid solution at or greater than the required refueling water storage tank (RWST) concentration (ppm) until the SHUTDOWN MARGIN is restored to within limit.

'I

/

SURVEILLANCE REQUIREMENTS 4.1.1.1 Verify SHUTDOWN MARGIN is within the limit specified in the CORE OPERATING LIMITS REPORT at 24Ihus.

the frequency specified in the Surveillance Frequency Control Program

• (l)See Special Test Exception 3.10.1 MILLSTONE - UNIT 2 3/4 1-1 Amendment No. -33, 64-, 7-, -74,4-39, 448, 2f0

September 25, 2003 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 REACTIVITY CONTROL SYSTEMS REACTIVITY BALANCE

/

LIMITING CONDITION FOR OPERATION /

3.1 .1.2 The core reactivity balance shall be within +/- 1% Ak/k of predicted values.

APPLICABILITY: MODES 1 and 2.

ACTION:

With core reactivity balance not within limit:

Re-evaluate core design and safety analysis and determine that the reactor core is acceptable for continued operation and establish appropriate operating restrictions and Surveillance 'I Requirements within 7 days or otherwise be in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

/

SURVEILLANCE REQUIREMENTS 4.1.1.2 Verify*(1) overall core reactivity balance is within +/- 1% Ak/k of predicted values prior to entering MODE 1 after fuel loading and at I- - . ....................... *(2).

The provisions of Specification 4.0.4 are not plicable.

Ithe frequency specified in the Surveillance Frequency Control Program I "p

• (I) The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel bumup of 60 Effective Full Power Days after each fuel loading.

• • (2) Only required after 60 Effective Full Power Days.

MILLSTONE - UNIT 2 3/4 1-3 Amendment No. 449, *80

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.5 The Reactor Coolant System temperature (Tavg) shall be _515'F when the reactor is critical.

APPLICABILITY: MODES I and 2*.

ACTION:

With the Reactor Coolant System temperature (Tavg) < 515'F, restore Tag to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.

SURVEILLANCE REQUIREMENTS 4.1.1.5 The Reactor Coolant System temperature (Tavg) shall be determined to be _515'F.

a. Within 15 minutes prior to making the reactor critical, and

b. At ..ast eu.p.. ,.; how; when the reactor is critical and the Reactor Coolant System temperature (Tavg) is IF.

the frequency specified in the Surveillance Frequency Control Program

• With Keff> 1.0.

MILLSTONE - UNIT 2 3/4 1-7 AMENDMENT NO. -24,"0-

Septembe 25, 2003 REACTIVITY CONTROL SYSTEMS ACTION: (Continued):

C. CEA Deviation Circuit C. 1 Verify the indicated position of each CEA to be within inoperable. 10 steps of all other CEAs in its group within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 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 6 or otherwise be in MODE 3t within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. One or more CEAs untrippable. D.1 Be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

OR Two or more CEAs misaligned by

_ 20 steps.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 Verify the indicated position of each CEA to be within 10 steps of all other CEAs in its group at AND within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following any CEA movement ger than 10 steps.

4.1.3.1.2 Verify CEA fr dom of movement (trippability) by moving each individual CEA that is not fully serted into the reactor core 10 steps in either direction 1stonvu Pei 92 days.

4.1.3.1.3 Verify the CEA De iation Circuit is OPERABLE at W by a functional test of th CEA group Deviation Circui hich verifies at the circuit prevents any CEA fr being misaligned from all ther CEAs i its group by more than 10 steps (indicate position).

4.1.3.1.4 Verify the CEA Motion ibit is OPERABLE a functi al test which verifies that the circuit maintains t e CEA group overl and se encing requirements of Specification 3.1.3.6 and th t the circuit prev ts regul ing CEAs from being inserted beyond the Transie Insertion Limi speci d in the CORE OPERATING LIMITS REPORT:

a. Prior to each entry into DE 2 fro M DE 3, except that such verification need not be performed mo often an ce per 31 days, and

b. At .... ..... ...... a....

the frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 1-21 Amendment No. 4-2 September,25,

' 003 REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS (Continued)

LIMITING CONDITION FOR OPERATION (Continued) b) The CEA group(s) with the inoperable indicator is fully inserted, and subsequently maintained fully inserted, while maintaining the withdrawal sequence and THERMAL POWER level required by Specification 3.1.3.6 and when this CEA group reaches its fully inserted position, the "Full In" limit of the CEA with the inoperable position indicator is actuated and verifies this CEA to be fully inserted. Subsequent operation shall be within the limits of Specification 3.1.3.6.

4. If the failure of the position indicator channel(s) is during STARTUP, the CEA group(s) with the inoperable position indicator channel must be moved to the "Full Out" position and verified to be fully withdrawn via a "Full Out" indicator within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

c. With a maximum of one reed switch position indicator channel per group or one pulse counting position indicator channel per group inoperable and the CEA(s) with the inoperable position indicator channel at either its fully inserted position or fully withdrawn position, operation may continue provided:

1. The position of this CEA is verified immediately and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter by its "Full In" or "Full Out" limit (as applicable).

2. The fully inserted CEA group(s) containing the inoperable position channel is subsequently maintained fully inserted, and

3. Subsequent operation is within the limits of Specification 3.1.3.6.

d. With one or more pulse counting position indicator channels inoperable, operation in MODES 1 and 2 may continue for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided all of the reed switch position indicator channels are OPERABLE.

SURVEILLANCE REQUIREMENTS Irequired[

4.1.3.3 Each position indicator channel shall be determined to be OPERABLE by verifying the pulse counting position indicator channels and the reed switch position indicator channels agree within 6 steps at ,. - n P- p r 1 . "1r Fthe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 1-25 Amendment No. +-§4, 20&

StMber

. 5, 003 REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual CEA drop time, from a fully withdrawn position, shall be ,2.75 4" seconds from when electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:

a. Tavg _Ž 515'F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

With the drop time of any CEA determined to exceed the above limit, restore the 4, CEA drop time to within the above limit prior to proceeding to MODE I or 2.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time shall be demonstrated through measurement with Tavg > 515 0 F, L and all reactor coolant pumps operating prior to reactor criticality:

a. For all CEAs following each removal of the reactor vessel head,

b. For specifically affected individual CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and

c. At !a--t .n.. p.r 189- men.hm.

the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 1-26 Amendment No. 38, 1-52, 90, *4-6,-484-

Septmber 25, 2003 REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdrawn to > 176 steps.

MODE 1*(1) /

APPLICABILITY:

MODE 2(I),(2)** with any regulating CEA not fully inserted.

ACTION:

INOPERABLE EQUIPMENT REQUIRED ACTION A. One or more shutdown CEAs not A. 1 Restore shutdown CEA(s) to within limit, within limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or otherwise be in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

I SURVEILLANCE REQUIREMENTS 4.1.3.5 Verify each shutdown CEA is withdrawan > 176 steps at least n.. per 12 hou.

the frequency specified in the Surveillance Frequency Control Program

• (1) This LCO is not applicable while performing Specification 4.1.3.1.2. If

• • (2)See Special Test Exceptions 3.10.1 and 3.10.2. ,r MILLSTONE - UNIT 2 3/4 1-27 Amendment No. 2-M

Setme 25, 2333 f,

............ [ I[ I XDf' i IZ

• Z Y

Z Z*ZZZ*

A~A A.4-.

ZZ*Z2ZZZ 1 I 114 Illl AA IIII Vv*lv***w -i 1*'~*

REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS (Continued)

B. Regulating CEA groups B. 1 Verify Short Term Steady State Insertion Limits as inserted between the Long Term specified in the CORE OPERATING LIMITS REPORT Steady State Insertion limit and are not exceeded within 15 minutes or otherwise be in the Transient Insertion Limit MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

specified in the CORE OPERATING LIMITS REPORT OR for intervals > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval. B.2 Restrict increases in THERMAL POWER to < 5%

RATED THERMAL POWER per hour within 15 minutes or otherwise be in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

C. Regulating CEA groups C. 1 Restore regulating CEA groups to within the Long inserted between the Long Term Term Steady State Insertion Limit specified in the CORE Steady State Insertion Limit and OPERATING LIMITS REPORT within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or '9 the Transient Insertion Limit otherwise be in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

specified in the CORE OPERATING LIMITS REPORT for intervals > 5 effective full power days (EFPD) per 30 EFPD or interval > 14 EFPD per 365 EFPD.

D. PDIL alarm circuit D.1 Perform Specification 4.1.3.6.1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and inoperable, once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter or otherwise be in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. I, SURVEILLANCE REQUIREMENTS 4.1.3.6.1 Verify each regulating CEA group position is within the Transient Insertion Limits specified in the CORE OPERATING LIMITS REPORT at least onse per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> .

The provisions of Specification 4.0.4 are not applicable entering into MODE 2 from MODE 3.

4.1.3.6.2 Verify the accumulated times during which the re lating CEA groups are inserted beyond the Steady State Insertion Limits but wi in the Transient Insertion Limits /

specified in the CORE OPERATING LIMITS PORT 1"s4 onve per-21 u.....

4.1.3.6.3 Verify PDIL alarm circuit is OPERABLE-the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 1-29 Amendment No. 449, 246, 24*

Mareh 16, 2006 REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE MECHANISMS LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod drive mechanisms shall be de-energized.

APPLICABILITY: MODES 3*, 4, 5 and 6, whenever the RCS boron concentration is less than refueling concentration of Specification 3.9.1.

ACTION:

With any of the control rod drive mechanisms energized, restore the mechanisms to their de-energized state within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or immediately open the reactor trip circuit breakers.

SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod drive mechanisms shall be verified to be de-energized t leet les+ @eer-

• 4-hes .

the frequency specified in the Surveillance Frequency Control Program The control rod drive mechanisms may be energized for MODE 3 as long as 4 reactor coolant pumps are OPERATINQ the reactor coolant system temperature is greater than 5000 F, the pressurizer pressure is greater than 2000 psia and the high power trip is OPERABLE.

MILLSTONE - UNIT 2 3/4 1-31 Amendment No. +4-6,44

Sepftrniber 25, 2003 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

[the frequency specified in the Surveillance Frequency Control Program 4.2.1.2 Excore Detector Monitoring Svstem*(l) - The e ore detector monitoring system may ,-

be used for monitoring the core power distribut

a. Verifying at e per 1 1---- that the CEAs are withdrawn to and maintained at or beyond ong Term Steady State Insertion Limits of Specification 3.1.3.6.

b. Verifying at Go per 4 ! d.,., that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the allowable limits specified in the CORE OPERATING LIMITS REPORT.

4.2.1.3 Incore Detector Monitoring System**( 2),***( 3 ) - The incore detector monitoring ,-

system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:

a. Are adjusted to satisfy the requirements of the core power distribution map which shall be updated at !cat c pcr 31 Jd,,.

pcc

b. Have their alarm se oint adjusted to less than or equal to the limits specified in the CORE OPERATING IMITS REPORT.

lat the frequency specified in the Surveillance Frequency Control Program, A(l) Only required to be met when the Excore Detector Monitoring System is being used to determine Linear Heat Rate.

+

• • (2 )Only required to be met when the Incore Detector Monitoring System is being used to determine Linear Heat Rate.

• • • (3)Not required to be performed below 20% RATED THERMAL POWER.

MILLSTONE - UNIT 2 3/4 2-2 Amendment No. 2-, -3,8, -354, 99, 439, 448,484

March 16, 2006-POWER DISTRIBUTION LIMITS TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR - FTr LIMITING CONDITION FOR OPERATION 3.2.3 The calculated value of FTr shall be within the 100% power limit specified in the CORE OPERATING LIMITS REPORT. The FTr value shall include the effect of AZIMUTHAL POWER TILT.

APPLICABILITY: MODE 1 with THERMAL POWER >20% RTP*.

ACTION:

With FTr exceeding the 100% power limit within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

a. Reduce THERMAL POWER to bring the combination of THERMAL POWER and FTr to within the power dependent limit specified in the CORE OPERATING LIMITS REPORT and withdraw the CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or

b. Be in at least HOT STANDBY.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 FTr shall be determined to be within the 100% power limit at the following intervals:

a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading,

b. At per 31 days of aetu-a~ed eper~ioa in MODE 1, and

c. Within ur hours if the AZIMUTHAL POWER TILT (Tq) is > 0.020.

4.2.3.3 FTr shall be dete *ned by using the incore detectors to obtain a power distribution map with all CEAs at or above t Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump Combinatio the frequency specified in the Surveillance Frequency Control Program

• See Special Test Exception 3,10.2.

MILLSTONE - UNIT 2 3/4 2-9 Amendment No. 398, -2, 79, 90, 99, 4-1, 4-39,4484-5, 4-64,2-30, M8,294

March 16, 2006 POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - TQ LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL POWER TILT (Tq) shall be < 0.02.

APPLICABILITY: MODE 1 with THERMAL POWER > 50% of RATED THERMAL POWER(l)*.

ACTION:

a. With the indicated Tq > 0.0 2 but <0.10, either restore T to < 0.02 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or verify the TOTAL UNRODDED INTEGRATED RAiIAL PEAKING FACTOR (FTr) is within the limit of Specification 3.2.3 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. Or otherwise, reduce THERMAL POWER to < 50% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. With the indicated Tq > 0.10, perform the following actions: (2)**

1. Verify the TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR (FTr) is within the limit of Specification 3.2.3 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; and

2. Reduce THERMAL POWER to < 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; and

3. Restore Tq _0.02 prior to increasing THERMAL POWER. Correct the cause of the out of limit condition prior to increasing THERMAL POWER.

Subsequent power operation above 50% of RATED THERMAL POWER may proceed provided that the measured T is verified < 0.02 at least once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or until verified at 99% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.4.1 Verify T is within limit at, ve..ry 12 he,. The provisions of Specification 4.0.4 are not applicable for entering inI ODE 1 with THERMAL POWER > 50% of RATED THERMAL POWER from MODE 1.

the frequencv specified in the Surveillance Frequency Control Pro qcramJ

• (1)See Special Test Exception 3.10.2.

• • (2 )All subsequent Required ACTIONS must be completed if power reduction commences prior to restoring Tq < 0.10.

MILLSTONE - UNIT 2 3/4 2-10 Amendment No. M8, 4, 90, 4-39,444, 280,2%9

October1 12, 1 990 POWER DISTRIBUTION LIMITS DNB MARGIN LIMITING CONDITION FOR OPERATION 3.2.6 The DNB margin shall be preserved by maintaining the cold leg temperature, pressurizer pressure, reactor coolant flow rate, and AXIAL SHAPE INDEX within the limits specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its specified limits, restore the parameter to within its above specified limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to < 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.6.1 The cold leg temperature, pressurizer pressure, and AXIAL SHAPE INDEX shall be determined to be within the limits specified in the CORE OPERATING LIMITS REPORT l*est

@neeFer.-lea*-s.

121 The reactor coolant flow rate shall be determined to be within the ilt1 specified in the CORE OPERATING LIMITS REPORT/at a least nte p-4.2.6.2 The provisions of Specification 4.0.4 are not 7 .

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 2-13 Amendment No. -38,90, 4-4-, +48-

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-I.

SURVEILLANCE REQUIREMENTS Irequired I V-4.3.1.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The bypass function and automatic bypass removal function shall be demonstrated OPERABLE during a CHANNEL FUNCTIONAL TEST once within 92 days prior to each reactor startup. The total bypass function shall be demonstrated OPERABLE at least @nee pe menhs during CHANNEL CALIBRATION testing of each channel affected bypass operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of ea reactor trip function shall be demonstrated to be within its limit at J-At Offee P@F 18 ffinfl"=f*lats eutron detectors are exempt from response time testing. EaIh t¢ lJ, ii di.

L at. V,;a ' ".Ifeti,,,n h,,

Suc,,'*thatl a!!i, Itheln Of lt 3.3 1.se d Ithe frequency specified in the Surveillance Frequency Control ProgramI MILLSTONE - UNIT 2 3/4 3-1 Amendment No. 74, +99, 294, 304-

lReplace each marked through surveillance frequency in the Check, Calibration, and Functional Test columns with "SFCP" REACTOR PROTECTIVE INSTRZ L IANCE ýNTAu REOUIREMENTS H

CHANNEL MODES IN WHICH zO CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. Manual Reactor Trip N.A. N.A. SUI( 1) N.A.

2. Power Level - High

a. Nuclear Power -M- 1,2, 3*

b. AT Power 1

3. Reee Coolant Flow - Low -S-- -Ni- 1,2 1,2 0 4. Pressurizer Pressure - High "S"

- i4 5. Containment Pressure - High -R- -M-- 1,2

--R-ta. 6. Steam Generator Pressure - Low -R M 1,2

7. Steam Generator Water 44 1,2 Level - Low -to-

8. Local Power Density - High N-. 4-R 1 b4-

9. Thermal Margin/Low Pressure 1,2

-R-- S4D+

10. Loss of Turbine--Hydraulic N.A. N.A.

Fluid Pressure - Low

+

lReplace each marked through surveillance frequency in the Check, Calibration, and Functional Test columns with "SFCP" I (tI REACTOR PROTECTIVE INSTUMENýATIoNLLANCE REOUIREMENTS H m C

z CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED H

11. Wide Range Logarithmic Neutron -W5) SAJ(l) 3,4,5 Flux Monitor - Shutdown

12. DELETED I

13. Reactor Protection System N.A. N.A. -M-and S/U(1) 1, 2 and*

Logic Matrices tJ. 14. Reactor Protection System N.A. N.A. -M-and S/U(1) 1, 2 and

• Logic Matrix Relays

15. Reactor Trip Breakers N.A. N.A. M 1, 2 and

• I CL E3 0D I

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The engineered safety feature actuation system instrumentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a. With an engineered safety feature actuation system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, either adjust the trip setpoint to be consistent with the value specified in the Trip Setpoint column of Table 3.3-4 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or declare the channel inoperable and take the ACTION shown in Table 3.3-3.

b. With an engineered safety feature actuation system instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS

-Jrequired 4.3.2.1.1 Each engineered safety feature actuation system instrumentation channel shall be ,'

demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The bypass function and automatic bypass removal function shall be demonstrated OPERABLE during a CHANNEL FUNCTIONAL TEST once within 92 days prior to each reactor startup. The total bypass function shall be demonstrated OPERABLE al t Me during CHANNEL CALIBRATION testing of each channelafec"y bypass operation.

per +ff 4nc Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 3-9 Amendment No. 4-98, 2G, 2-9+I, -3*

September 2,5, 2,003 INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF function shall be demonstrated to be within the limit aqtt . t .. on. 8...oths.

1 a_, test shall

, i_, .. at tfe heaW euy.

of s*pcifi ei*nteeS1llance r aluuu 3qnC*yCtrl Lthe frequency specified in the Surveillance Frequency Control Program-]

MILLSTONE - UNIT 2 3/4 3-10 Amendment No. 49, 229,24-5, 2 ,

lReplace each marked through surveillance frequency in the Check, Calibration, and Functional Test columns with "SFCP" ENGINEERED SAFETY FEATURE ACTUATIONURV IN RU XMVEILLANCE REOUIREMENTS CHANNEL MODES IN WHICH CHANNEL C"NNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. SAFETY INJECTION (SIAS)

a. Manual (Trip Buttons) N.A. N.A. N.A.

b. Containment Pressure - High -R-- 1,2,3

c. Pressurizer Pressure - Low -S -R 1,2,3

- d. Automatic Actuation Logic N.A. N.A. 4.4(l) 1,2,3

2. CONTAINMENT SPRAY (CSAS)

a. Manual (Trip Buttons) N.A. N.A. N.A.

b. Containment Pressure-- --S- -R- 1,2,3

-R1)

High - High

c. Automatic Actuation Logic N.A. N.A. 1,2,3

3. CONTAINMENT ISOLATION (CIAS)

a. Manual CIAS (Trip Buttons) N.A. N.A. N.A.

-R N.A.

b. Manual SIAS (Trip Buttons) N.A. N.A.

c. Containment Pressure - High -R 1,2,3

d. Pressurizer Pressure - Low 1,2,3

e. Automatic Actuation Logic N.A. N.A. 1,2,3

4. MAIN STEAM LINE ISOLATION

a. Manual (Trip Buttons) N.A. N.A. N.A.

b. Containment Pressure - High N-R- 1,2,3

c. Steam Generator Pressure - -R- 1,2,3 tb Low

-R- 1,2,3

d. Automatic Actuation Logic N.A. N.A.

z -M-

5. ENCLOSURE BUILDING

!)

0 -m-FILTRATION (EBFAS)

a. Manual EBFAS (Trip Buttons) N.A. N.A. N.A.

b. Manual SIAS (Trip Buttons) N.A. N.A. N.A. p

c. Containment Pressure - High -'- 1,2,3

-R'-

d. Pressurizer Pressure - Low ,-- 1,2,3

e. Automatic Actuation Logic N.A. N.A. 1,2,3

[Replace each marked through surveillance frequency in the Check, Calibration, and Functional Test columns with "SFCP" I TAfT F 4 ~-9 (CJ'~ntiniie&h TABLE 41-2 (Chnfinued)

ENGINEERED SAFETY FEATURE ACTUATION SYSJEI'I I ITR'UMý[AION S JRVFHILANCE REOT ITREMENTS

• FUNCTIONAL UNIT CHANNEL CHECK z CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE REQUIRED

• 6. CONTAINMENT SUMP RECIRCULATION (SRAS)

a. Manual SRAS (Trip Buttons) N.A. N.A. N.A.

-'M

b. Refueling Water Storage -N.- 1,2,3 Uj Tank - Low

c. Automatic Actuation Logic N.A. N.A. -M(1) 1,2,3

7. DELETED

8. LOSS OF POWER

-M

a. 4.16 kv Emergency Bus -Js 1,2,3 Undervoltage - level one

> b. 4.16 kv Emergency Bus -.R "-R- 1,2,3 Undervoltage - level two E 9. AUXILIARY FEEDWATER

a. Manual N.A. N.A. N.A.

b. -M Steam Generator Level - Low -S- 1,2,3 10.

c. Automatic Actuation Logic STEAM GENERATOR BLOWDOWN N.A. N.A. --M 1,2,3 I r

a. Steam Generator Level - Low -R -it- -v" 1,2,3 C

C I.

4-

TABLE 4.3-2 (Continued)

Ithe frequency specified in the Surveillance Frequency Control Program TABLE NOTATION (1) The coincident logic circuits shall be tested automatically or mi ally least Once per3-4ays. The automatic test feature shall be verified OPERABLE a leasmt once per 31 days.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or other specified conditions for surveillance testing of the following:

a. Pressurizer Pressure Safety Injection Automatic Actuation Logic; and

b. Pressurizer Pressure Containment Isolation Automatic Actuation Logic; and

c. Steam Generator Pressure Main Steam Line Isolation Automatic Actuation Logic; and

d. Pressurizer Pressure Enclosure Building Filtration Automatic Actuation Logic.

Testing of the automatic actuation logic for Pressurizer Pressure Safety Injection, Pressurizer Pressure Containment Isolation, and Pressurizer Pressure Enclosure Building Filtration shall be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after exceeding a pressurizer pressure of 1850 psia in MODE 3. Testing of the automatic actuation logic for Steam Generator Pressure Main Steam Line Isolation shall be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after exceeding a steam generator pressure of 700 psia in MODE 3.

MILLSTONE - UNIT 2 3/4 3-22 Amendment No. 6-, 240,48- k-

Mereh 16, 2006 INSTRUMENTATION ENGINEERED SAFETY FEATURE ACTUATION SYSTEM SENSOR CABINET POWER SUPPLY DRAWERS LIMITING CONDITION FOR OPERATION 3.3.2.2 The engineered safety feature actuation system Sensor Cabinets (RC02A I, RC02B2, RC02C3 & RC02D4) Power Supply Drawers shall be OPERABLE and energized from the normal power source with the backup power source available. The normal and backup power sources for each sensor cabinet is detailed in Table 3.3-5a:

CABINET NORMAL POWER BACKUP POWER RC02A1 VA-10 VA-40 RC02B2 VA-20 VA-30 RC02C3 VA-30 VA-20 RC02D4 VA-40 VA-10 Table 3.3-5a APPLICABILITY: MODES 1, 2,3 and 4 ACTION:

With any of the Sensor Cabinet Power Supply Drawers inoperable, or either the normal or backup power source not available as delineated in Table 3.3-5a, restore the inoperable Sensor Cabinet Power Supply Drawer to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS lat the frequency specified in the Surveillance Frequency Control Program 4.3.2.2.1 The engineered safe feature actuation system Sensor Cabinet Power Supply Drawers shall be determined OPERABtLe.-p.-hift by visual inspection of the power supply drawer indicating lamps.

4.3.2.2.2 Verify the OPERABILITY of the Sensor Cabinet Power Supply auctioneering circuit J/

at ..........

k 1 onths.A Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 3-23 Amendment No. 479, 2-G, 291

slay 200

• +

INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING LIMITING CONDITION FOR OPERATION 3.3.3,1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

a. With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or declare the channel inoperable.

b. With the number of OPERABLE channels less than the number of MINIMUM CHANNELS OPERABLE in Table 3.3-6, take the ACTION shown in Table 3.3-6.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS requred 4.3.3.1.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-3.

4.3.3.1.2 DELETED 4.3.3.1.3 Verify the response time of the control room isolation channel at t ........per 18 the frequency specified in the Surveillance Frequency Control Programn MILLSTONE - UNIT 2 3/4 3-24 Amendment No. 4-5+, 24-, 242, 294, 4-,-2,98

IReplace each marked through surveillance frequency in the Check, Calibration, and Functional Test columns with "SFCP" I RADIATION MONITORING INSTRUM AILLANCE REOUIREMENTS cz~

CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST REQUIRED

1. AREA MONITORS t'~)

a. Deleted

b. Control Room Isolation -M ALL MODES

c. Containment High Range -8* -M 1, 2,3,&4 t*

d* 2. PROCESS MONITORS

a. Containment Atmosphere- -- t- 1,2,3, & 4 Particulate

b. Deleted I,

c. Noble Gas Effluent -"M 1, 2,3, & 4 Monitor (high range)

(Unit 2 Stack)

I,

• Calibration of the sensor with a radioactive source need only be performed on the lowest range. Higher ranges may be calibrated electronically.

z 0

S.LPI.b.i 25, 200-INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the number of OPERABLE remote shutdown monitoring instrumentation channels less than required by Table 3.3-9, either:

a. Restore the inoperable channel to OPERABLE status within 7 days, or

b. Be in HOT SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS

-reuired 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.

MILLSTONE - UNIT 2 3/4 3-28 Amendment No. G- ,,

IReplace each marked through surveillance frequency in the Check and Calibration columns with "SFCP" I REMOTE SHUTDOWN MONITORING INSTRUMENTA ONSURVEIL AE REOUIREMENTS H

0 ztrl CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION P,*

1. Wide Range Logarithmic Neutron Flux M H

t'J

2. Reactor Trip Breaker Indication M N.A.

3. Reactor Cold Leg Temperature M R

4. Pressurizer Pressure

a. Low Range M

b. High Range M R 0

5. Pressurizer Level M R

6. Steam Generator Level M R

7. Steam Generator Pressure M R

• Neutron detectors are excluded from the CHANNEL CALIBRATION.

1L~

March 16, 2006 INSTRUMENTATION e-ACCIDENT MONITORING LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. ACTIONS per Table 3.3-11.

SURVEILLANCE REQUIREMENTS required 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.

MILLSTONE - UNIT 2 3/4 3-31 Amendment No. 66,4-1-, 28K,-394-

lReplace each marked through surveillance frequency in the Check and Calibration columns with "SFCP" TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SS VEILLANCE MENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION

1. Pressurizer Water Level M

2. Auxiliary Feedwater Flow Rate M

3. Reactor Coolant System Subcooled/Superheat Monitor M R*

4. PORV Position Indicator M 4-0 5. PORV Block Valve Position Indicator N.A.

R*

6. Safety Valve Position Indicator M R*

7. Containment Pressure M

8. Containment Water Level (Narrow Range) M

9. Containment Water Level (Wide Range) M

10. Core Exit Thermocouples M

11. Main Steam Line Radiation Monitor M

12. Reactor Vessel Coolant Level M

• Electronic calibration from the ICC cabinets only.

March 16, 2006 REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant loops shall be OPERABLE and in operation.

APPLICABILITY: MODES 1 and 2.

ACTION:

With the requirements of the above specification not met, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation at 4eeest QQ1;e por !Q h'.:*.

[the frequency specified in the Surveillance Frequency Control Program A'

MILLSTONE - UNIT 2 3/4 4-1 Amendment No. -0, 69, 2-3, 49,29 Reissue.d.* NP.. I".f'eTr d*atnei Septemb.r 27, 2006

REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 Two reactor coolant loops shall be OPERABLE and one reactor coolant loop shall be in operation.

NOTE All reactor coolant pumps may not be in operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period provided:

4/

a. no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1.1; and

b. core outlet temperature is maintained at least 10°F below saturation temperature.

APPLICABILITY: MODE 3.

ACTION: a. With one reactor coolant loop inoperable, restore the required reactor coolant loop to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With no reactor coolant loop OPERABLE or in operation, immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1 and immediately initiate corrective action to return one required reactor coolant 4,

loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS

-at the frequency specified in the Surveillance Frequency Control Program 4.4.1.2.1 T uired reactor coolant pump, if not in operation, shall be determined to be OPERABLE ,epe.T-.7. by verifying correct breaker alignment and indicated power available.

4.4.1.2.2 One reactor coolant loop shall be verified to be in operation at lceat C*Cf,pcr 12 hourz.

4.4.1.2.3 Each steam generator secondary side water level sha e verified to be > 10% narrow range at Jat once par 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.s.

MILLSTONE - UNIT 2

.. ptemet 14, 2000 REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION HOT SHUTDOWN SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required pump, if not in operation, shall be determined OPERAB oLefp.1 pr A by verifying correct breaker alignment and indicated power available.

lat the frequency specified in the Surveillance Frequency Control Program 4.4.1.3.2 The required steam generator(s) shall be determined OPERABLE, by verifying the secondary side water level to be > 10% narrow range at eas;t, ene per 12 he-ar.

4.4.1.3.3 One reactor coolant loop or shutdown c oling train shall be verified to be in operation L at l...t onc er 1r. hcumr. A rthe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 4-1c Amendment No. 69,.249

REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION COLD SHUTDOWN - REACTOR COOLANT SYSTEM LOOPS FILLED LIMITING CONDITION FOR OPERATION (continued) ,

APPLICABILITY: MODE 5 with Reactor Coolant System loops filled.

ACTION: a. With one shutdown cooling train inoperable and any steam generator secondary water level not within limits, immediately initiate action to either restore a second shutdown cooling train to OPERABLE status or restore steam generator secondary water levels to within limit.

b. With no shutdown cooling train OPERABLE or in operation, immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1 and immediately initiate action to restore one shutdown cooling train to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS lat the frequency specified in the Surveillance Frequency Control Program 4.4.1.4.1 The required shutdown cooling pumpjf.*t in operation, shall be determined OPERABLE e pe.pr. ; daysq. e ýcorrect breaker alignment and indicated power available.

4.4.1.4.2 The required steam generators shall be determined OPERABLE, by verifying the secondary side water level to be > 10% narrow range at least @cne per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.1.4.3 One shutdown cooling train shall be ye jed to be in operati at least onee.@ -F-12

. tdour Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 4-1 e Amendment No. 249, 293

September- 14, 2000 REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION COLD SHUTDOWN - REACTOR COOLANT SYSTEM LOOPS NOT FILLED SURVEILLANCE REQUIREMENTS lat the frequency specified in the Surveillance Frequency Control Program I 4.4.1.5.1 The required shutdo cooling pump, if not in operation, shall be determined OPERABLE @nee per:7 ,,ys4 verifying correct breaker alignment and indicated power l, available.

4.4.1.5.2 One shutdown cooling train shall be verified to be in operation-at lea..e.ee.pe...

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 4-1g Amendment No.-249-

SIep.tIlInbet 14, 2000 REACTOR COOLANT SYSTEM REACTOR COOLANT PUMPS COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.6 A maximum of two reactor coolant pumps shall be OPERABLE.

APPLICABILITY: MODE 5 ACTION:

With more than two reactor coolant pumps OPERABLE, take immediate action to comply with 0 Specification 3.4.1.6.

SURVEILLANCE REQUIREMENTS 4.4.1.6 Two reactor coolant pumps shall be demonstrated inoperable at 4.as@ po hetaos by verifying that the motor circuit breakers have been disconnet m their electrical power supply circuits. n Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 4-1h Amendment No. "-8, 249

Febmary 12, 2008 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.3.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE:

a. by performance of a CHANNEL FUNCTIONAL TEST, peration, and

b. hs&by performance of a CHANNEL CALIBRATION.

C. Eh& by operating the PORV through one complete cycle of full s representative of MODES 3 or 4. /4 4.4.3.2 Each block val sh I e demonstrated OPERABLF, one Fp, "ir 92-day, by operating the valve through one complec of full travel. ThisAons6tration is not required if a PORV block valve is closed and power ved to meet ication 3.4.3 b or c.

lat the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 4-3a Amendment No. 66,68, +", 3 REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with:

a. Pressurizer water level < 70%, and 4-

b. At least two groups of pressurizer heaters each having a capacity of at least 130 kW.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.4.1 The pressurizer water level shall be determined to be within its limits 4" 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.4.2 Verify at least two groups of pressurizer heaters ea e a capacity at least 130 k W at t.; t ,ncc vcr 9 2 Jp;' ja.

MILLSTONE - UNIT 2 3/4 4-4 Amendment No. 66, 74, 94, 4-30, 21-9, 2"I, 196-

September 30, 2008 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

2. Appropriate grab samples of the containment atmosphere are obtained and analyzed for particulate radioactivity within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at least once per 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> thereafter, and

3. A Reactor Coolant System water inventory balance is performed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at least once per 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> thereafter.

Otherwise, be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:

a. Containment atmosphere particulate monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and

b. Containment sump level monitoring system-performance of CHANNEL CALIBRATION TEST at lcast @znce pcr 18 mcmths.

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 4-8a Amendment 306-

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System Operational LEAKAGE shall be limited to: -'r

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE, t1

c. 75 GPD primary to secondary LEAKAGE through any one steam generator, and

d. 10 GPM IDENTIFIED LEAKAGE.

APPLICABILITY: MODES 1, 2,3 and 4.

ACTION:

a. With any RCS operational LEAKAGE not within limits for reasons other than PRESSURE BOUNDARY LEAKAGE or primary to secondary LEAKAGE, reduce LEAKAGE to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. With ACTION and associated completion time of ACTION a. not met, or PRESSURE BOUNDARY LEAKAGE exists, or primary to secondary LEAKAGE not within limits, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1

---

NOTES -.-.-.------------------ -- --

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 establishment of 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 at'a7t ......par Q hours.

Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 4-9 Amendment No. S, -3Z, K-, 8-S, 404-,

424-, 4-31,2 -S, 228, m)9-

-May 3 1, 0@;

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.2.2

---

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 < 75 gallons per day through any one SG at leatenee Per ;2helts.

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 4-10 Amendment No. 266, 299

Oetaber:27, 2008 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.8.1 Verify the specific activity of the primary coolant < 1100 .tCi/gram DOSE

/

EQUIVALENT XE- 133 *

/

4.4.8.2 Verify the specific activity the primary coolant <1.0 JtCi/gram DOSE EQUIVALENT 1-131 ,* 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 c ge o > 15% RATED THERMAL POWER within a one /

hour period.

lat the frequency specified in the Surveillance Frequency Control Program I

• Surveillance only required to be performed for MODE I operation, consistent with the provisions of Specification 4.0.1.

MILLSTONE - UNIT 2 3/4 4-14 Amendment No. 44-5, 3&7

St-areh- 30 2000 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENT 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at.p

.34ys.thereafter when the PORV is required OPERABLE./a

b. Performance of a CHANNEL CALIBRATION on the channel a least civ 18 mon. /

C.

d.

4.4.9.3.2 Verify no injecting into the RC: 4, 4.4.9.3.3 Verify no injecting into the R 4-4.4.9.3.4 erify is open least enee per 31 days when the vent pathway is provided by yen ocke*c-Aeled, or otherwise secured in the open position, otherwise, verify t MILLSTONE - UNIT 2 3/4 4-21b Amendment No. 60, 4-4-, 4--8, 24-8, 224,243.

EMERGENCY CORE COOLING SYSTEMS SAFETY INJECTION TANKS (Continued)

SURVEILLANCE REQUIREMENTS 4.5.1 Each SIT shall be demonstrated OPERABLE:

a. Verify each SIT isolation valve is fully open at least onse per. 12a, hw. *()

b. Verify borated water volume in each SIT is > 080 cubic feet and < 1190 cubic feet at per 12 hearS.**(2)

c. Veri itrogen

""\ "-rfy -- cover-pressure

• h... ***(3)/ in each SIT s 2! 200 psig and < 250 psig at letst

d. Verify boron ncentration in each SIT i 1720 ppm .

and once within hours after each solu on volume 1i rase* 1%of tank volume****( 4 ) tha *s not the result of ddition fro the ueling water storage tank.

e. Verify that the closing coi *n the val bre ubicle is removed Ithe frequency specified in the Surveillance Frequency Control Program

• (1) If one SIT is inoperable, except as a result of boron concentration not within limits or inoperable level or pressure instrumentation, surveillance is not applicable to the affected SIT.

• • (2) If one SIT is inoperable due solely to inoperable water level instrumentation, surveillance is not applicable to the affected SIT.

• • • (3) If one SIT is inoperable due solely to inoperable pressure instrumentation, surveillance is not applicable to affected SIT.

• • • • (4)Only required to be performed for affected SIT.

MILLSTONE - UNIT 2 3/4 5-2 Amendment No. 45,-202, 22-, Q68

Septe* ber 9, 20U4 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At by verifying each Emergency Core Cooling System ma ual, power operated, and automatic valve in the flow path servicing safety rela d equipment, that is not locked, sealed, or otherwise secured in position, is in /

the c ect position. /

b. At per--34-days by verifying that the following valves are in the indicated pd power to the valve operator removed:

Valve u Valve Function Valve Position 2-SI-3 Shutdown Cooling Open*

Flow Control 2-SI-659 SRAS Recirc. Open**

2-SI-660 SRAS Recirc. Open**

/

locked at preset throttle open position.

• • To be

• prior to recirculation following LOCA.

c. By verifying the de el ped head of each high pressure safety injection pump at the flow test point is gre t than or equal to the required developed head when tested pursuant to Specifica o 4.0.5.

d. By verifying the develo e head of each low pressure safety injection pump at the flow test point is greater n or equal to the required developed head when tested pursuant to Specification . .5.

e. By verifying the delivered of each charging pump at the required discharge pressure is greater than or eq to the required flow when tested pursuant to Specification 4.0.5.

f.

A F-t 1b -r- by ifying each Emergency Core Cooling System auto c valve in flow path t is not locked, sealed, or otherwise secured in position, ates to the correct po tion on an actual or simulated actuation signal.

g. At

• AMPby verin ng each high pressure safety injection pump and lo ssure safe ection pu starts automatically on an actual or .I simulated actua ignal.

MILLSTONE - UNIT 2 3/4 - Amendment No. -, 59, -36,2t" Ithe frequency specified in the Surveillance Frequency Control Program :

September- 1s, 2007 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

h. At per 18 men by verifying each low pressure safety injection pump sto automatically on an actual or simulated actuation signal.

By veri ing the correct position of each electrical and/or mechanical position stop for each1 *ection valve in Table 4.5-1:

1. Withi 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after completion of valve operations.

2. At ,ca.. ,'*,per  !°8cmend".

j18o°' by verifying through visual inspection of the n su tha each Emergency Core Cooling System subsystem suction ent inlet is n.restrict by ebris and the suction inlet strainers show no evidence of ~1-structural d ss or no al corrosion.

k. At 15*:verifying vltAJALUthe Shutdown Cooling System open pe interloc p en t Shutdown Cooling System inlet isolation valves from being o d with a c or simulated Reactor Coolant System pressure signal of > 300 psia.

Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 5-5 Amendment No. -t, 4-5, 6-5,6*-, 4-0+,

4-9, 46+, 24-7, 244, 2-3-,28-

EMERGENCY CORE COOLING SYSTEMS REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank shall be OPERABLE with:

a. A minimum contained volume of 370,000 gallons of borated water,

b. A minimum boron concentration of 1720 ppm,

c. A minimum water temperature of 50°F when in MODES I and 2, and

d. A minimum water temperature of 35 0 F when in MODES 3 and 4.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the refueling water storage tank inoperable, restore tank 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 /> or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:

a. At t --fl.e r 7 day by:

1. Verifying the water level in the tank, and

2. Verifying the boron concentration of the water.

b. When in ODES 3 and 4, aea szt ,nzc pcr 21 hour- by verifying the RWST temperatu e is > 35F whe he RWST ambient air temperature is < 35'F.

c. When in M DES I an , a least once per 24 hou, by verifying the RWST temperature i Ž 5OTF h he RWST ambient air temperature is < 50F.

the frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 5-8

EMERGENCY CORE COOLING SYSTEMS TRISODIUM PHOSPHATE (TSP)

LIMITING CONDITION FOR OPERATION 3.5.5 The TSP baskets shall contain Ž282 ft3 of active TSP.

APPLICABILITY: MODES 1, 2, and 3 ACTION:

With the quantity of TSP less than required, restore the TSP quantity within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.5.1 Verify that the TSP baskets contain >282 ft3 of TSP at least once per 18 moaths.

4.5.5.2 Verifv that a samnle from the TSP baskets orovides a(eauate oH adjustment of borated water at .... m , .1..u..........

MILLSTONE - UNIT 2 3/4 5-9 Amendment No. 2-t-7,-29()

Mre1 16, 2006 3/4,6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At k!,. -.- F,, 31iday by verifying that all penetrations0') not capable of being dodby OPERABLE containment automatic isolation valves(2) and required to be Co* d during accident conditions are closed by valves, blind flanges, or deactiv d automatic valves secured in their positions,(3) except for valves that are open der administrative control as permitted by Specification 3.6.3.1.

b. Ar ncr. days

--. 7 by verifying the equipment hatch is closed and sealed.

c. By ven' ng the c tainment air lock is in compliance with the requirements of Specificati 3.6.1.3.

d. After each closin a petration subject to type B testing (except the containment air lock), op ed following a Type A or B test, by leak rate testing in accordance with the a ent Leakage Rate Testing Program.

e. By verifying Containment struc I integrity in accordance with the Containment Tendon Surveillance Program.

1the frequency specified in the Surveillance Frequency Control Program (1) Except valves, blind flanges, and deactivated automatid valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position.

These penetrations shall be verified closed prior to entering MODE 4 from MODE 5, if not performed within the previous 92 days.

(2) In MODE 4, the requirement for an OPERABLE containment automatic isolation valve system is satisfied by use of the containment isolation trip pushbuttons (3) Isolation devices in high radiation areas may be verified by use of administrative means.

MILLSTONE - UNIT 2 3/4 6-1 . .,R ,,.-D,40.P,.* N -

Amendment No. 2-S, 9-, 293, 24-0, 24-S, 279,2~91-

, ei ,,20 CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS SURVEILLANCE REQUIREMENTS 4.6.1.3.1 Each containment air lock shall be demonstrated OPERABLE in accordance with the Containment Leakage Rate Testing Program. Containment air lock leakage test results shall be evaluated against the leakage limits of Technical Specification 3.6.1.2. (An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test).

4.6.1.3.2 Each containment air lock shall be demonstrated OPERABLE least p.24 /ga~

-ntoths by verifying that only one door in each air lock can be opene a ime.

Ithe frequency specified in the Surveillance Frequency Control Programl MILLSTONE - UNIT 2 3/4 6-6a Amendment No. 4-4, 203, CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between -12 inches Water Gauge and +1.0 PSIG.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the containment internal pressure in excess of or below the limits above, restore the internal pressure to within the limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in HOT STANDBY within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; go to COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to within the limits

[the frequency specified in the Surveillance Frequency Control Programl MILLSTONE - UNIT 2 3/4 6-8 Amendment No. 2499-

A..gust 21, 1998 CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 120'F.

APPLICABILITY: MODES 1, 2,3 and 4.

ACTION:

With the containment average air temperature > 120'F, reduce the average air temperature to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be determined to be <

least ,nee per 24 hoers.

• 10°F a

/ F Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 6-9 Amendment No. e+9-

Mareh 16, 2006 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY AND COOLING SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.2.1 Two containment spray trains and two containment cooling trains, with each cooling train consisting of two containment air recirculation and cooling units, shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3*.

ACTION: 'F Inoperable Equipment Required ACTION 41

a. One containment a. 1 Restore the inoperable containment spray train to spray train OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1750 psia within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. One containment b. 1 Restore the inoperable containment cooling train to cooling train OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. One containment c. 1 Restore the inoperable containment spray train or the spray train inoperable containment cooling train to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT SHUTDOWN within the next AND 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

One containment cooling train

d. Two containment d. I Restore at least one inoperable containment cooling train to cooling trains OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

e. All other e. I Enter LCO 3.0.3 immediately.

combinations SURVEILLANCE REQUIREMENTS 4.6.2.1.1 Each containment spray train shall be demonstrated OPERABLE:

a. At l*,*anc pr 31 d-yc yerifying each containment spray manual, power operated, and automatic vilveF4e spray train flow path, that is not locked, sealed, or otherwise secured in positip, is in the correct position.

Ithe frequency specified in the Surveillance Frequency Control ProgramI The Containment Spray System is not required to be OPERABLE in MODE 3 if pressurizer pressure is < 1750 psia.

MILLSTONE - UNIT 2 3/4 6-12 Amendment No. 24-1, 2-28, 236, 283, 291

Mach 31, 2008 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

[the frequency specified in the Surveillance Frequency Control Program

b. By verifying the developed head of each cont& ent spray pump at the flow test point is greater than or equal to th eloped head when tested pursuant to Specification 4.0.5 C. At n Pe l b-stveri lby each automatic containment spray valve in the flow path t s not locked d, or otherwise secured in position, actuates to orrect position actual or simulated actuation signal.

d. t-- erifying each containment spray pump starts automatically on an ac I mulated actuation signal.

e. By verifying eac pr zzle is unobstructed following activities that could cause nozzle bca4.

4.6.2.1.2 Each contain ent r irculation and cooling unit shall be demonstrated OPERABLE:

a. At t by operating each containment air recirculation and coolin uni n slow speed for > 15 minutes.

b. Ator- 9 by verifying each containment air recirculation and cooli unit cooling water flow rate is > 500 gpm.

c. At lat wie@ pe. 18 mon by verifying each containment air recirculation and cooling unit starts automatically on an actual or simulated actuation signal.

MILLSTONE - UNIT 2 3/4 6-13 Amendment No. 24-S, 2*3, -303-

CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3.1 Each containment isolation valve shall be OPERABLE.(1) (2) *,

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more of the isolation valve(s) inoperable, either:

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or

b. Isolate the affected penetration(s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of a deactivated automatic valve(s) secured in the isolation position(s), or

c. Isolate the affected penetration(s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of a closed manual valve(s) or blind flange(s); or

d. Isolate the affected penetration that has only one containment isolation valve and a closed system within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> by use of at least one closed and deactivated automatic valve, closed manual valve, or blind flange; or

e. Be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.3.1 Each containment isolation valve shall be demonstrated OPERABLE:

a. By verifying the isolation time of each power operated automatic containment isolation valve when tested pursuant to Specification 4.0.5.

b. At least once per 18 mont*, verifying each automatic containment isolation valve that is not locked, seale'd, o ise secured in position, actuates to the isolation position on an actual or simulated ac signal.

the frequency specified in the Surveillance Frequency Control Program (1) Containment isolation valves may be opened on an intermittent basis under administrative controls. 4 (2) The provisions of this Specification in MODES 1, 2 and 3, are not applicable for main steam line isolation valves. However, provisions of Specification 3.7.1.5 are applicable for main steam line isolation valves.

MILLSTONE - UNIT 2 3/4 6-15 LH T*N DITIC,*ERATIGN Amendment No. 6, 2M0, 2743,-2

Rine. 16, 199 CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.3.2 The containment purge supply and exhaust isolation valves shall be sealed closed.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one containment purge supply and/or one exhaust isolation valve open, close the open valve(s) within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

I, SURVEILLANCE REQUIREMENTS 4.6.3.2 The containment purge supply and exhaust isolation valves shall be determined sealed closed at l, t ..... pe.r 31 day .

Ni the frequency specified in the Surveillance Frequency Control Programl MILLSTONE - UNIT 2 3/4 6-19 Amendment No. 6+, 2+6-

CONTAINMENT SYSTEMS POST-INCIDENT RECIRCULATION SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.4.4 Two separate and independent post-incident recirculation systems shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTION:

With one post-incident recirculation system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS Ithe frequency specified in the Surveillance Frequency Control Program 4.6.4.4 Each post-incident recirculation system shall be demonstrated OPERABLE aIeoet on.e per. 92 days on a STGGERDtE TEST BASIS by:

a. Verifying that the system can be started on operator action in the control room, and

b. Verifying that the system operates for at least 15 minutes.

MILLSTONE - UNIT 2 3/4 6-24

september 30, 199:7 CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT ENCLOSURE BUILDING FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.1 Two separate and independent Enclosure Building Filtration Trains shall be OPERABLE.

APPLICABILITY: MODES 1,2,3 and 4.

ACTION:

With one Enclosure Building Filtration Train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 4 SURVEILLANCE REQUIREMENTS

[the frequency specified in the Surveillance Frequency Control Program 4.6.5.1 Each Enclosure Building Filtra s e demonstrated OPERABLE:

a. At i ,,, p-* PCERE) TEST-BASI by initiating, from the control room, flb ough'the HEPA filter and charcoal absorber train and verifyie train operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

b. At lea.t cncc pzr 1 onths or (1) after any structural maintenance on the HEPA filter or charcoal absorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the train by: ,4 MILLSTONE - UNIT 2 3/4 6-25 Amendment No. -200

March 10, 1999 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

1. Verifying that the cleanup train satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the train flow rate is 9000 cfm +/- 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

3. Verifying a train flow rate of 9000 cfm +/- 10% during train operation when tested in accordance with ANSI N510-1975.

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

d. At ....... r 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 2.6 inches Water Gauge while operating the amn at a flow rate of 9000 cfmn +/- 10%.

2. ifying that the train starts on an Enclosure Building Filtration Actuation Si al (EBFAS).

e. After each co plete or partial replacement of a HEPA filter bank by verifying that the HEPA filt banks remove greater than or equal to 99% of the DOP when they are tested in-pI e in accordance with ANSI N510-1975 while operating the train ataflowrateof 000 cfm+/- 10%.

[the frequency specified in the Surveillance Frequency Control Program]

ASTM D3803-89 shall be used in place of ANSI N509-1976 as referenced in table 2 of Regulatory Guide 1.52. The laboratory test of charcoal should be conducted at a temperature of 30'C and a relative humidity of 95% within the tolerances specified by ASTM D3803-89.

Additionally, the charcoal sample shall have a removal efficiency of _ 95%.

MILLSTONE - UNIT 2 3/4 6-26 Amendment No. -2, ;2, 4-74, 28, 28

Septembe 39, 19971 CONTAINMENT SYSTEMS ENCLOSURE BUILDING LIMITING CONDITION FOR OPERATION 3.6.5.2 The Enclosure Building shall be OPERABLE. -

APPLICABILITY: MODES 1, 2,3 and 4.

ACTION:

With the Enclosure Building inoperable, restore the Enclosure Building to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 1 SURVEILLANCE REQUIREMENTS 4.6.5.2.1 OPERABILITY of the Enclosure Building shall be demonstrated at leas -pe3 L days by verifying that each access opening is closed except when the access o ing is being used for normal transit entry and exit.

4.6.5.2.2. At !e_.t .nceper 1_ monthoerify each Enclosure Buildin iltration Train produces a negative pressure of greater than or equa o 0.25 inches W.G in th nclosure Building Filtration Region within 1 minute after an En sure Building Filtr ion Actuation Signal.

the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 6-28 Amendment No. M

Mwitety 31 20 PLANT SYSTEMS AUXILIARY FEEDWATER PUMPS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

Inoperable Equipment Required ACTION

e. Three auxiliary feedwater e.

pumps in MODE 1, 2, or 3.

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

'F

'I.

LCO 3.0.3 and all other LCO required ACTIONS requiring MODE changes are suspended until one AFW pump is restored to OPERABLE status.

Immediately initiate ACTION to restore one auxiliary feedwater pump to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. At k1aa, pie , 31 Jday by verifying each auxiliary feedwater manual, power op ted, and automatic valve in each water flow path and in each steam supply flow p to the steam turbine driven pump, that is not locked, sealed, or otherwise secured position, is in the correct position.

b. By verifying e developed head of each auxiliary feedwater pump at the flow test point is greater an or equal to the required developed head when tested pursuant to Specification 4. 5. (Not required to be performed for the steam turbine driven auxiliary feedwater p until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 800 psig in the steam generators. The provis ns of Specification 4.0.4 are not applicable to the steam turbine driven auxiliary fdwater pump for entry into MODE 3.)

Ithe frequency specified in the Surveillance Frequency Control Programl MILLSTONE - UNIT 2 3/4 7-5 Amendment No. 32, 6-3, 2-83, 24-

PLANT SYSTEMS AUXILIARY FEEDWATER PUMPS SURVEILLANCE REQUIREMENTS (Continued)

C. At nper 18 men by verifying each auxiliary feedwater automatic valve th *snot locked, sealed, or otherwise secured in position, actuates to the correct positi , as designed, on an actual or simulated actuation signal.

d. At I 1by verifying each auxiliary feedwater pump starts aut atical as designed, on an actual or simulated actuation signal.

e. By verifyi th roper alignment of the required auxiliary feedwater flow paths by verifying w om the condensate storage tank to each steam generator prior to entering MO whenever the unit has been in MODE 5, MODE 6, or defueled for a cum *e period of greater than 30 days.

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 7-5a Amendment No. 294- ,4

Deeember 31, 1998 PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank shall be OPERABLE with a minimum contained volume L of 165,000 gallons. 4 APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With less than 165,000 gallons of water in the condensate storage tank, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

a. Restore the water volume to within the limit or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or

b. Demonstrate the OPERABILITY of the fire water system as a backup supply to the auxiliary feedwater pumps and restore the condensate storage tank water volume to within its limits within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.3 The condensate storage tank shall be demonstrated OPERABLE least onee-pe h outb y v erify in g the w ater le v el.

Ithe frequency specified in the Surveillance Frequency Control Programi MILLSTONE - UNIT 2 3/4 7-6 Amendment No. 2'12ý

Attgust 2, 19 85 TABLE 4.7-2 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT MINIMUM AND ANALYSIS FREQUENCY

1. Gross Activity Determination 3 timesppor. 4s 'with a maximum ime of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples.

2. Isotopic Analysis for DOSE a) -l-per 31 daya, whenever the EQUIVALENT 1-131 gross activity determination Concentration indicates iodine concentrations greater than 10% of the allowable limit

,whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit.

lAt the frequency specified in the Surveillance Frequency Control Program I I'

MILLSTONE - UNIT 2 3/4 7-8 Amendment No. 43, 444

Mareh 16, 20,6 PLANT SYSTEMS MAIN FEEDWATER ISOLATION COMPONENTS (MFICs)

LIMITING CONDITION FOR OPERATION (Continued)

b. With two or more of the feedwater isolation components inoperable in the same flow path, either:

I. Restore the inoperable component(s) to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until ACTION 'a' applies, or

2. Isolate the affected flow path within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and verify that the inoperable feedwater isolation components are closed or isolated/secured once per 7 days, or

3. Be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS the frequency specified in the Surveillance Frequency Control Program 4.7.1.6 Each/feedwater isolation valve/feedwater pump trip circuitry shall be demonstrated OPERABLE atL . ence p' r 18 months by:

a. Verifying that on 'A' main steam isolation test signal, each isolation valve actuates to its isolation position, and

b. Verifying that on 'B' main steam isolation test signal, each isolation valve actuates to its isolation position, and

c. Verifying that on 'A' main steam isolation test signal, each feedwater pump trip circuit actuates, and

d. Verifying that on 'B' main steam isolation test signal, each feedwater pump trip circuit actuates.

MILLSTONE - UNIT 2 3/4 7-9b Amendment No. 41882,94 Reissued by NTRC Lt--- datcd September 27, 2006

A...... 1 1 PLANT SYSTEMS ATMOSPHERIC DUMP VALVES 7 LIMITING CONDITION FOR OPERATION 3.7.1.7 Each atmospheric dump valve line shall be OPERABLE. k APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one atmospheric dump valve line inoperable, restore the inoperable line to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With more than one atmospheric dump valve line inoperable, restore one L inoperable line 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 /> or be in MODE 3 within the 4 next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.7 Verify the OPERABILITY of each atmospheric dump valve line by local manual operation of each valve in the flowpath through one complete cycle of operation at le@as $e eG@p

]the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 7-9c Amendment No. 2-2-, 2M

Fvbittry 8, 999 PLANT SYSTEMS STEAM GENERATOR BLOWDOWN ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.8 Each steam generator blowdown isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3 ACTION:

With one or more steam generator blowdown isolation valves inoperable, either:

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or

b. Isolate the affected steam generator blowdown line within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or

c. Be in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.8 Verify the closure time of each steam generator blowdown isolation valve is ___

10 seconds on an actual or simulated closure signal at I Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 7-9d Amendment No. 2 PLANT SYSTEMS 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.1 Two reactor building closed cooling water loops shall be OPERABLE.

• APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one reactor building closed cooling water loop inoperable, restore the inoperable loop to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.3.1 Each reactor building closed cooling water loop shall be demonstrated OPERABLE:

a. At by verifying each reactor building closed cooling water mal, power operated, and automatic valve in the flow path servicing safety relate the corcequipment, that is not locked, sealed, or otherwise secured in position, is in

b. p t A

swate *t --

d du~to ion, actuates by verifying each reactor building tic valve in the to flow thepath correct closed cooling

b. thatposition is not locked, on an sealed, At per imen actual ororsimulated otherwise by verifying each reactor s al. building closed actuation cooling wft matically on an actual or simulated actuation signal.

lat the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 7-11 Amendment No. 236, 2-3

F..b.A..y 13, 2003 PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1 Two service water loops shall be OPERABLE. x" APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one service water loop inoperable, restore the inoperable loop to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.4.1 Each service water loop shall be demonstrated OPERABLE:

a. At -*'e.e

--- - 3 1 &"ysby verifying each service water manual, power operated, and' tomatic valve in the flow path servicing safety related equipment, that is not locked,Nealed, or otherwise secured in position, is in the correct position.

b. At I- 0mmzihr, by verifying each service water automatic valve in the flow t locked, sealed, or otherwise secured in position, actuates to the c4 on an actual or simulated actuation signal.

C. At swath. by verifying each service water pump starts aIutýol t o4 r simulated actuation signal.

lat the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 7-12 Amendment No. , 2-36, 24

StaLrc 10, 1999 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS

[the frequency specified in the Surveillance Frequency Control Program 4.7.6.1 Each Control Room Emergency Ventilation Trai shall be demonstrated OPERABLE:

a. At least @ ,. .. -that the control room air temperature is <

I000 F.

b. At hofit8AGRETETBS by initiating from the control room, fl through the HEPA filters and charcoal absorber train and ver5 g he train operates for at least 15 minutes.

c. At st efte per 8,en,,, or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the train by:

1. Verifying that the cleanup train satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the train L flow rate is 2500 cfm +/- 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978.* The carbon sample shall have a removal efficiency of > 95 percent.

3. Verifying a train flow rate of 2500 cfm + 10% during train operation when tested in accordance with ANSI N510-1975.

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

• ASTM D3803-89 shall be used in place of ANSI N509-1976 as referenced in table 2 of Regulatory Guide 1.52. The laboratory test of charcoal should be conducted at a temperature of 30'C and a relative humidity of 95% within the tolerances specified by ASTM D3803-89.

MILLSTONE - UNIT 2 3/4 7-17 Amendment No. 2-, =2, 400, 14-9, 42-,

449, 4;4, 2H8

Marj- 10, 199 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

-- the frequency specified in the Surveillance Frequency Control Program

e. At .......

to t^ by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3.4 inches Water Gauge while operating the train at a flow rate of 2500 cfm +/- 10%.

2. Verifying that on a recirculation signal, with the Control Room Emergency I Ventilation Train operating in the normal mode and the smoke purge mode, the train automatically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks.

MILLSTONE - UNIT 2 3/4 7-17a Amendment No. 4, 2, 4-00, 4-1-9, --- 2,

-144, 7-5,41,28

M~y 3, 0 PLANT SYSTEMS 3/4.7.11 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.11 The ultimate heat sink shall be OPERABLE with a water temperature of less than or equal to 75'F.

APPLICABILITY: MODES 1, 2,3, AND 4 ACTION:

a. With the ultimate heat sink water temperature > 75'F and < 77'F, operation may continue provided the water temperature averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is verified <75°F at least once per hour. Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. /,

b. With the ultimate heat sink water temperature > 77'F, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.11 The ultimate heat sink shall be determined OPERABLE:

a. At lceat @cne pcr 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, by verifying the water temperature to be within limits.

b. At lcat c@ne pr 6 hc-w-r*y erifying the water temperature to be within limits when the water tem a e exceeds 70'F.

Ithe frequency specified in the Surveillance Frequency Control Program t1 MILLSTONE - UNIT 2 3/4 7-34 Amendment No. 4-4-, 4-62, 4-94-, 24-3, 2*4-, 4, Maireh 16, 2006 ELECTRICAL POWER SYSTEMS ACTION (Continued)

Inoperable Equipment Required ACTION

e. Two diesel e.1 Perform Surveillance Requirement 4.8.1.1.1 for the generators offsite circuits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.

AND e.2 Restore one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

AND e.3 Following restoration of one diesel generator restore remaining inoperable diesel generator to OPERABLE status following the time requirements of ACTION 4 Statement b above based on the initial loss of the remaining inoperable diesel generator.

SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Verify correct breaker alignment and indicated power available for each required offsite circuit at leact o r 21 heurc.

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 8-2a Amendment No. 4-31, 2*, 247,4

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) the frequency specified in the Surveillance Frequency Control Program 4.8.1.1.2 Each r Luired diesel generator shall be demonstrated OPERABLE:*

a. At , per .t...ays by: ,

1. Verifying the fuel level in the fuel oil supply tank, 2.

NOTES

/

1. A modified diesel generator start involving idling and gradual acceleration to synchronous speed may be used as recommended by the manufacturer. When modified start procedures are not used, the requirements of SR 4.8.1.1.2.d. I must be met.

2. Performance of SR 4.8.1.1.2.d satisfies this Surveillance Requirement.

Verifying the diesel generator starts from standby conditions and achieves /

steady state voltage > 3740 V and <4580 V, and Frequency > 58.8 Hz and

< 61.2 Hz.

3.

NOTES

1. Diesel generator loading may include gradual loading as (

recommended by the manufacturer.

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

3. This test shall be conducted on only one diesel generator at a time.

4. This test shall be preceded by and immediately follow without shutdown a successful performance of SR 4.8.1.1.2.a.2, or SRs 4.8.1.1.2.d.1 and 4.8.1.1.2.d.2.

5. Performance of SR 4.8.1.1.2.d satisfies this Surveillance Requirement.

Verifying the diesel generator is synchronized and loaded, and operates for

> 60 minutes at a load Ž 2475 kW and _2750 kW.

• All diesel starts may be preceded by an engine prelube period.

MILLSTONE - UNIT 2 3/4 8-3 Amendment No. 4-7;, -2+,247

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. The diesel fuel oil supply shall be checked by:

0

1. Checking for and removing accumulated water from each fuel oil storage tank at 1--- t Per 92 days.

2. Verifying fue il properties of new and stored fuel oil are tested in accordance with, d maintained within the limits of, the Diesel Fuel Oil Testing Program in a rdance with the Diesel Fuel Oil Testing Program.

c. At least etee per 18 menth 1

1. DeletedI

2. Ithe frequency specified in the Surveillance Frequency Control Program NOTE This surveillance shall not normally be performed in MODE 1, 'I 2, 3, or 4. However, portions of the surveillance may be /

performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verifying that the automatic time delay sequencer is OPERABLE with the following settings:

Sequence Time After Closing of Diesel Generator Step Output Breaker (Seconds)

Minimum Maximum 1 (TI) 1.5 2.2 2 (T2 ) T, + 5.5 8.4 3 (T 3) T2 + 5.5 14.6 4 (T4 ) T3 + 5.5 20.8 MILLSTONE - UNIT 2 3/4 8-3a Amendment No. 4-3+, 2-34, 249, 24

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENT (Continued)

,the frequency specified in the Surveillance Frequency Control Program

d. At-aQ11 cnc cr 18 d ays by:

1. Verifying the diesel starts from standby conditions and accelerates to

> 90% of rated speed and to > 97% of rated voltage within 15 seconds after the start signal.

2. Verifying the generator achieves steady state voltage > 3740 V and

<4580 V, and frequency > 58.8 Hz and < 61.2 Hz.

3.

/

NOTES /

1. Diesel generator loading may include gradual loading as recommended by the manufacturer.

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

3. This test shall be conducted on only one diesel generator at a time.

4. This test shall be preceded by and immediately I follow without shutdown a successful performance /

of SRs 4.8.1.1.2.d.1 and 4.8.1.1.2.d.2, or SR 4.8.1.1.2.a.2.

Verifying the diesel generator is synchronized and loaded, and operates for

> 60 minutes at a load > 2475 kW and < 2750 kW.

MILLSTONE - UNIT 2 3/4 8-4 Amendment No. 23-,-2-77

4ttne 16, 199 ELECTRICAL POWER SYSTEMS 3/4 8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 The following A.C. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generators with tie breakers open between redundant busses:

4160 volt Emergency Bus # 24 C 4160 volt Emergency Bus #24 D 480 volt Emergency Load Center #22 E 480 volt Emergency Load Center #22 F 120 volt A.C. Vital Bus # VA-10 120 volt A.C. Vital Bus # VA-20 A' 120 volt A.C. Vital Bus # VA-30 120 volt A.C. Vital Bus # VA-40 APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With less than the above complement of A.C. busses OPERABLE, restore the inoperable bus and/

or associated load center to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from normal A.C. sources with tie breakers open between redundant busses at least e ,. 7 &y; by verifying correct breaker alignment and indicated power availabili Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 8-6 Amendment No. -244

4Jn- 16, 1988 ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION (Continued) 3.8.2.1A Inverters 5 and 6 shall be OPERABLE and available for automatic transfer via static switches VS1 and VS2 to power busses VA-10 and VA-20, respectively. It APPLICABILITY: MODES 1, 2 & 3 ACTION:

a. With inverter 5 or 6 inoperable, restore the inverter to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With inverter 5 or 6 unavailable for automatic transfer via static switch VSl or VS2 to power bus VA-10 or VA-20, respectively, restore the automatic transfer capability within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. With inverters 5 and 6 inoperable or unavailable for automatic transfer via static switches VSI and VS2 to power busses VA-10 and VA-20, respectively, restore the inverters to OPERABLE status or restore their automatic transfer capability within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.8.2.1A a. Verify correct inverter voltage, frequency, and alignment for automatic transfer via static switches VS 1 and VS2 to power busses VA- 10 and VA-20, respectively, at i..st - days.. 40

b. Verify that busses VA-1 and VA-20 automatically transfer to their alternate power sources, nverters 5 and 6, respectively, 4.ei'e p* r-.fuceizag during shutu w Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 8-6a Amendment No. 4-88,46

September 18, 2008 ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERABLE and energized from sources of power other than a diesel generator but aligned to an OPERABLE diesel generator:

1 - 4160 volt Emergency Bus 1 - 480 volt Emergency Load Center 2 - 120 volt A.C. Vital Busses APPLICABILITY: MODES 5 and 6.

ACTION:

With less than the above complement of A.C. busses OPERABLE and energized, suspend all operations involving CORE ALTERATIONS and positive reactivity additions that could result in loss of required SDM or boron concentration, and movement of recently irradiated fuel assemblies.

SURVEILLANCE REQUIREMENTS 4.8.2.2 The specified A.C. busses shall be determined OPERABLE and energized from normal A.C. sources at by verifying correct breaker alignment and indicated power availability. "*

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 8-7 Amendment No. +W-9, 29-3, 45

ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.3 125-volt D.C. bus Train A and 125-volt D.C. bus Train B electrical power subsystems-shall be OPERABLE.

APPLICABILITY: MODES 1,2, 3 and 4.

ACTION:

With one 125-volt D.C. bus train inoperable, restore the inoperable 125-volt D.C. bus train to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.8.2.3.1 Each 125-volt D.C. bus train shall be determined OPERABLE atat ,,d by verifying correct breaker alignment and indicated power availability./K 4.8.2.3.2 Each 125-volt D.C. battery bank and charger of Train A a Train B shall be demonstrated OPERABLE: 1

a. By verifyingg'.--0,. eme- per.7 days that that the attery cell parameters meet Table 4.8-1 C A limits.

b. By verifying p the b ery cell parameters meet Table 4.8-1 Category B limits.

Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 8-8 Amendment No. -08,180, 249

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. A 'east ence per 1ofm1nth by verifying that:

1. The cells, cell plates and battery racks show no visual indication of physical damage or deterioration that could degrade battery performance,

2. The cell-to-cell and terminal connections are clean, tight, free of corrosion nd coated with anti-corrosion material, and

3. e battery charger will supply at least 400 amperes at a minimum of 130 v its for at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

d. A . .I m e n t,, d ur ing sh u td o wn , by v e rify ing th at the b attery ca city is dequate to supply and maintain in OPERABLE status all of the actual emer Incy I ads for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the battery is subjected to a battery service test.

e. At ICAnm onths, during shutdown, by verifying that the battery cap i I ast 80% of the manufacturer's rating when subjected to a perform c di harge test. This performance discharge test may be performed in lieu of the service test.

{the frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 8-9 Amendment No. 4--8', 4-80, -9

September 18, 2008 ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 One 125 - volt D.C. bus train electrical power subsystem shall be OPERABLE:

APPLICABILITY: MODES 5 and 6.

ACTION:

With no 125-volt D.C. bus trains OPERABLE, suspend all operations involving CORE ALTERATIONS and positive reactivity additions that could result in loss of required SDM or boron concentration, and movement of recently irradiated fuel assemblies.

SURVEILLANCE REQUIREMENTS 4.8.2.4.1 The above required 125-volt D.C. bus train shall be determined OPERABLE at 4emt

-n La by verifying correct breaker alignment and indicated power availabili 4.8.2.4.2 The above required 125-volt D.C. bus train battery bank and char shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2.

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 8-10 Amendment No. 48-, 4.9*, -N9, -2,

Jly 9,2o93 ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION SYSTEMS (TURBINE BATTERY) - OPERATING LIMITING CONDITION FOR OPERATION 4/

3.8.2.5 The Turbine Battery 125-volt D.C. electrical power subsystem shall be OPERABLE.

APPLICABILITY: MODES 1, 2 & 3 ACTION:

a. With the Turbine Battery 125-volt D.C. electrical power subsystem inoperable, restore the subsystem to OPERABLE status within 7 days or be in HOT

• l-U TTfTlf*lUJ xixthmn thf, novt ileAL 1I 1

• hnulre llIJUl L1A A,,

iencv sDecified in the Surveillance 4.8.2.5.1 Verify 125-volt D.C.

4.8.2.5.2 125-volt D.C. battery OPERABLE:

/

a. By verifying at i battery cell parameters meet Table 4.8-2 Category A

'I

b. By verifyingji the battery cell parameters meet Table 4.8-2 /

c. At verifying that:

1. Oes, and battery racks show no visual indication of or deterioration that could degrade battery performance,

2. -cell and terminal connections are clean, tight, free of corrosion, with anti-corrosion material.

d. At WeA -- c- _r 18 months, during shutdown, by verifying that the battery capaci s adequate to supply and maintain in OPERABLE status all of the actual loads r 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when the battery is subjected to a battery service test.

e. At caS 60 muudiS, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. This performance discharge test may be performed in lieu of the battery service test.

MILLSTONE - UNIT 2 3/4 8-11 Amendment No. 4-99, 2W

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATIONS LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained sufficient to ensure that the more restrictive of following reactivity conditions is met:

a. Either a Keff of 0.95 or less, or

b. A boron concentration of greater than or equal to 1720 ppm.

APPLICABILITY: MODE 6.

NOTE Only applicable to the refueling canal when connected to the Reactor Coolant System ACTION:

With the requirements of the above specification not satisfied, within 15 minutes suspend all operations involving CORE ALTERATIONS and positive reactivity additions and initiate and 4V continue boration at greater than or equal to 40 gpm of boric acid solution at or greater than the required refueling water storage tank concentration (ppm) until Keff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 1720 ppm, whichever is the more restrictive.

SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any CEA in excess of 3 feet from its fully inserted position within the reactor pressure vessel.

4.9.1.2 The boron concentration of all filled portions of the reactor coolant system and the refueling canal shall be determined by chemical analysis a earm t .ef Q hem".

4.9.1.3 Deleted Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 9-1 Amendment No. 24ý, 2.63,, 240,493--

itme 28, 2006-REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 Two source range neutron flux monitors shall be OPERABLE, each with continuous visual indication in the control room and one with audible indication in the containment, and control room.

APPLICABILITY: MODE 6.

ACTION:

a. With one of the above required monitors inoperable, immediately suspend all operations involving CORE ALTERATIONS and operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

b. With both of the above required monitors inoperable, immediately initiate action to restore one monitor to OPERABLE status. Additionally, determine that the boron concentration of the Reactor Coolant System satisfies the requirements of LCO 3.9.1 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

SURVEILLANCE REQUIREMENTS.

4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. Deleted

b. A CHANNEL CALIBRATION n.er 18 months*

c. A CHANNEL CHECK and verification udible cou at lGt once per 12 lat the frequency specified in the Surveillance Frequency Control Program

• Neutron detectors are excluded from CHANNEL CALIBRATION.

MILLSTONE - UNIT 2 3/4 9-2 Amendment No. -64,4

Septembe~r ",20flA REFUELING OPERATIONS CONTAINMENT PENETRATIONS SURVEILLANCE REQUIREMENTS 4.9.4.1 Verify each required containment penetration is in the required status t lrest onee-per 4.9.4.2 Deleted Ithe frequency specified in the Surveillance Frequency Control Program NULLSTONE - UNIT 2 3/4 9-5 Amendment No. 240, 2a84-

june 28 206-REFUELING OPERATIONS SHUTDOWN COOLING AND COOLANT CIRCULATION - HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION ACTION:

With no shutdown cooling train OPERABLE or in operation, perform the following actions:

a. Immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1 and the loading of irradiated fuel assemblies in the core; and

b. Immediately initate action to restore one shutdown cooling train to OPERABLE status and operation; and

c. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> place the containment penetrations in the following status:

1. Close the equipment door and secure with at least four bolts; and

2. Close at least one personnel airlock door; and

3. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be closed with a manual or automatic isolation valve, blind flange, or equivalent.

SURVEILLANCE REQUIREMENTS 4.9.8.1 One shutdown cooling train shall be verified to be in operation and circulating reactor coolant at a flow rate greater than or equal to 1000 gpm at !eect pzr pnec12 hour .

[the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 9-8a Amendment No. -, 4-85., -249,284,

-Jun 28, 2066 REFUELING OPERATIONS SHUTDOWN COOLING AND COOLANT CIRCULATION - LOW WATER LEVEL LIMITING CONDITION FOR OPERATION (continued)

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be closed with a manual or automatic isolation valve, blind flange, or equivalent.

SURVEILLANCE REQUIREMENTS

[the frequency specified in the Surveillance Frequency Control Program 4.9.8.2.1 One shutdown cooling train shall be verified to biin operation and circulating reactor coolant at a flow rate greater than or equal to 1000 gpm at @neePer 12 het,-h,.

4.9.8.2.2 The required shutdown cooling pump, if not in operation, shall be determined OPERABLE eef yverifying correct breaker alignment and indicated power available.

t lat the freauencv specified in the Surveillance Frequency Control Prowaram MILLSTONE - UNIT 2 3/4 9-8c Amendment No. 49 1

Jamaary 11, 20 REFUELING OPERATIONS WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.11 As a minimum, 23.0 feet of water shall be maintained over the top of the 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.

ACTION:

With the water level less than that specified above, immediately suspend CORE ALTERATIONS and immediately suspend movement of irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS 4.9.11 The water level shall be determined to be within its minimum depth at leaest.nc,.epe Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 9-11 Amendment No.-26-9 ,/

REFUELING OPERATIONS STORAGE POOL WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.12 As a minimum, 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: WHENEVER IRRADIATED FUEL ASSEMBLIES ARE IN THE STORAGE POOL.

ACTION:

With the requirement of the specification not satisfied, suspend all movement of fuel and spent fuel pool platform crane operations with loads in the fuel storage areas.

SURVEILLANCE REQUIREMENTS 4.9.12 The water level in the storage pool shall be determined to be within its minimum depth at ans-,g-- -a; 7days when irradiated fuel assemblies are in the fuel storage pool.

the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 9-12

REFUELING OPERATIONS SHIELDED CASK LIMITING CONDITION FOR OPERATION 3.9.16.1 All fuel within a distance L from the center of the spent fuel pool cask laydown area shall have decayed for at least 1 year. The distance L equals the major dimension of the shielded cask.

APPLICABILITY: Whenever a shielded cask is on the refueling floor.

ACTION:

With the requirements of the above specification not satisfied, do not move a shielded cask to the refueling floor. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.16.1 The decay time of all fuel within a distance L from the center of the spent fuel pool cask laydown area shall be determined to be __1 year within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to moving a shielded cask to the refueling floor and at 1--.+ -He@ .er. 72 hws thereafter.

Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 9-19 Amendment No. -2, 409, 4-2, -245

REFUELING OPERATIONS SPENT FUEL POOL BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.17 The boron concentration in the spent fuel pool shall be greater than or equal to 1720 parts per million (ppm).

APPLICABILITY: Whenever any fuel assembly or consolidated fuel storage box, is stored in /

the spent fuel pool. /

ACTION:

With the boron concentration less than 1720 ppm, suspend the movement of all fuel, consolidated fuel storage boxes, and shielded casks, and immediately initiate action to restore the spent fuel pool boron concentration to within its limit.

The provisions of specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS I,

4.9.17 Verify that the boron concentration is greater than or equal to 1720 ppW avw' days and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the initial movement of a fuel assembly or consi d fuel storage box in the Spent Fuel Pool, or shielded cask over the cask lay lat the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 2 3/4 9-21 Amendment No. 4-09, 44-4, 4"8, 24-*,

September- 25, 2003 3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The requirement of Specifications 3.1.1.1, 3.1.3.5 and 3.1.3.6 may be suspended for 4 measurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth (of those CEAs actually withdrawn) is available for trip insertion ,

from OPERABLE CEA(s).

APPLICABILITY: MODES 2 and 3(1) during PHYSICS TESTS.

ACTION:

a. With any CEA not fully inserted and with less than the above reactivity equivalent /

available for trip insertion, within 15 minutes initiate and continue boration at > 40 gpm of boric acid solution at or greater than the required refueling water storage tank (RWST) concentration (ppm) until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all CEAs inserted and the reactor subcritical by less than the above reactivity  %

equivalent, immediately initiate and continue boration at > 40 gpm of boric acid /

solution at or greater than the required refueling water storage tank (RWST) concentration (ppm) until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTSS 4.10.1.1 The position of each CEA required either partially or fully withdrawn shall be 4, determined at 4.10.1.2 Each CE ot fully inserted shall be demonstrated capable of full insertion when tripped from at least the withdrawn position once within 7 days prior to reducing the SHUTDOWN MARGIN to le an the limits of Specification 3.1.1.1 (2).

Ithe frequency specified in the Surveillance Frequency Control Program (1) Operation in MODE 3 shall be limited to 6 consecutive hours. ,'

(2) Not required to be performed during initial power escalation following a refueling outage if I/

SR 4.1.3.4 has been met MILLSTONE - UNIT 2 3/4 10-1 Amendment No. -52:,6-7, -, 4-54, 240-

SPECIAL TEST EXCEPTIONS GROUP HEIGHT AND INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.3 and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and

b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2 below.

APPLICABILITY: MODES 1 and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.3 and 3.2.4 are suspended, immediately:

a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1 or

b. Be in HOT STANDBY within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at least enee per het, during PHYSICS TESTS in which the requirements of Specifications 3.1.1.4 .1.3.1, 3.1.3.5, 3.1.3.6, 3.2.3 or 3.2.4 are suspended and shall be verified to be within the test p wer plateau.

4.10.2.2 The linear heat rate shall be determined to within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detect Monitoring System pursuant to the requirements of Specifications 4.2.1.3 during PHYS S TESTS above 5% of RATED THERMAL POWER in which the requirements of pecifications 3.1.1.4, 3.1.3.1,3.1.3.5, 3.1.3.6, 3.2.3 or 3.2.4 are suspended.

Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 2 3/4 10-2 Amendment No. 38, -54,4-39, 25

,September. 18, 2008 ADMINISTRATIVE CONTROLS 6.27 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM (Continued)

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 licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of Surveillance Requirement 4.0.2 are applicable to the frequencies for assessing CRE habitability and determining CRE unfiltered inleakage as required by paragraph c.

Insert 1 6.28 SURVEILLANCE FREQUENCY CONTROL PROGRAM This program provides controls for surveillance frequencies. The program shall ensure that surveillance requirements specified in the technical specification 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 4.0.2 and 4.0.3 are applicable to the frequencies established in the Surveillance Frequency Control Program.

MILLSTONE - UNIT 2 6-33 Amendment No. 305 4r