ML13008A328

Millstone Power Station, Unit 3, Response to Request for Additional Information Regarding Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TAC No. ME9733)

ML13008A328
Person / Time
Site:	Millstone
Issue date:	01/04/2013
From:	Price J A Dominion, Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME9733
Download: ML13008A328 (16)

Text

Dominion Nuclear Connecticut, Inc.5000 Dominion Boulevard, Glen Allen, VA 23060 omIInion Web Address: www.dom.com January 4, 2013 U. S. Nuclear Regulatory Commission Serial No.12-764 Attention:

Document Control Desk NSSLNVDC RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM (TAC NO. ME9733)By letter dated October 4, 2012, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3). The proposed amendment would relocate certain technical specification (TS) surveillance frequencies to a licensee controlled program by adopting Technical Specification Task Force (TSTF)-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF [Risk-Informed Technical Specification Task Force Initiative]

5b." The proposed change would also add a new program, the Surveillance Frequency Control Program, to the TSs, in accordance with TSTF-425.

TSTF-425 is approved for use by the Nuclear Regulatory Commission (NRC). In a letter dated December 18, 2012, the NRC transmitted a request for additional information (RAI) to DNC related to the LAR.DNC agreed to respond to the RAI by January 31, 2013.Attachment 1 provides DNC's response to the NRC's RAI with the exception of Question 3, which will be addressed by January 31, 2013. Attachment 2 provides the marked-up TS page.If you have any questions regarding this submittal, please contact Wanda Craft at (804)273-4687.Sincerely, J.Price Vice President

-Nuclear Engineering 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 kn ledge and belief.Acknowledged before me this day of ,2013.My Commission Expires: 4. ,, Cx NOTARY PL y OREGISTRATIONWA

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Serial No.12-764 Docket No. 50-423 Page 2 of 2 Attachments:

1. Response to Request for Additional Information Regarding Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program 2. Marked-up Technical Specifications Page Commitments made in this letter: None cc: U.S: Nuclear Regulatory Commission Region I 2100 Renaissance Blvd Suite 100 King of Prussia, PA 19406-2713 J. S. Kim Project Manager -Millstone Power Station U.S. Nuclear Regulatory Commission' One White Flint North 11555 Rockville Pike Mail Stop 08-C2A Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Serial No.12-764 Docket No. 50-423 ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Serial No.12-764 Docket No. 50-423 Attachment 1, Page 1 of 11 By letter dated October 4, 2012, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3). The proposed amendment would relocate certain technical specification (TS) surveillance frequencies to a licensee controlled program by adopting Technical Specification Task Force (TSTF)-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF [Risk-Informed Technical Specification Task Force Initiative]

5b." The proposed change would also add a new program, the Surveillance Frequency Control Program (SFCP), to the TSs, in accordance with TSTF-425.

TSTF-425 is approved for use by the Nuclear Regulatory Commission (NRC). In a letter dated December 18, 2012, the NRC transmitted a request for additional information (RAI) to DNC related to the LAR. This attachment provides DNC's response to the NRC's RAI, with the exception of Question 3.Question I The submittal does not identify that technical specification (TS) Bases are adopted consistent with the U.S. Nuclear Regulatory Commission (NRC) letter issued April 14, 2010 Agencywide Documents Access and Management System (Accession Number ML100990099) subsequent to the adoption of (Technical Specification Task Force)TSTF-425.

In addition, the submittal's "Insert #2" is not consistent with this letter.Please provide confirmation that the TS Bases will be adopted consistent with this letter.DNC Response In the original issuance of TSTF-425, the TS Bases insert read as follows: "The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program." Several licensees submitting LARs for adoption of TSTF-425 identified a need to deviate from this statement because it only applies to frequencies that have been changed in accordance with the SFCP and does not apply to frequencies that are relocated but not changed.In NRC letter dated April 14, 2010, the NRC staff agreed that the TSTF-425 TS Bases insert applies to surveillance frequencies that are relocated and subsequently evaluated and changed, in accordance with the SFCP. The TSTF-425 TS Bases does not apply to surveillance frequencies relocated to the SFCP but not changed. Therefore, for surveillance frequencies relocated to the SFCP but not changed, existing plant-specific TS Bases descriptions remain valid for the unchanged surveillance frequency.

In a subsequent discussion on April 28, 2012 between the TSTF and the NRC, the NRC staff supported the following recommended changes :

Serial No.12-764 Docket No. 50-423 Attachment 1, Page 2 of 11 1. The existing Bases information describing the basis for the surveillance frequencies will be relocated to the licensee-controlled SFCP.2. The TSTF-425 TS Bases, INSERT #2, will be changed to read as follows: "The Surveillance Frequency is controlled under the Surveillance Frequency Control Program." In response to this issue, a note was added to the insert section of TSTF-425-A, Rev. 3.The note reads as follows: NOTE: The NRC and TSTF recommend a change to the Insert 2 wording and the NRC accepts it as an acceptable deviation from the Traveler.

After the "OR" and before the"Reviewer's Note," the recommended wording is, "The Surveillance Frequency is controlled under the Surveillance Frequency Control Program." Additionally, this change has been incorporated into Revision 4 of NUREG 1431, Standard Technical Specifications for Westinghouse Plants.Consistent with these actions, DNC incorporated the recommended changes into this LAR.Question 2 Please discuss how the focused peer review met the peer review guidance in the American Society of Mechanical Engineers (ASME) standard ANS[American Nuclear Society] RA-Sa-2009 and the clarifications in Regulatory Guide 1.200.DNC Response The focused peer review was performed by a diverse team of consultants from Science Applications International Corporation (SAIC) with individual experience ranging from 24 to 32 years in the nuclear industry.

SAIC met the peer review team composition and personnel qualifications outlined in Section 1-6.2 of the ASME/ANS Probabilistic Risk Assessment (PRA) Standard RA-Sa-2009.

In addition, SAIC stated in their final report, "The purpose of this report is to document the final results of the focused scope Peer Review of the Millstone Power Station Unit 3 (MPS3) Probabilistic Risk Assessment (PRA) against the requirements of the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA standard and any Clarifications and Qualifications provided in the Nuclear Regulatory Commission (NRC) endorsement of the Standard contained in Revision 2 to Regulatory Guide (RG) 1.200. This peer review was performed using .the process defined in Nuclear Energy Institute (NEI) 05-04."

Serial No.12-764 Docket No. 50-423 Attachment 1, Page 3 of 11 Question 4 Gap #1. Please clarify if the interviews with plant personnel were to determine if potential initiating events have been overlooked.

If not, describe the steps to address this supporting requirement.

DNC Response Since documentation of the interviews was not retained, specific record of the questions asked is not available.

A recent interview between PRA staff and Operations validated that potential initiating events have been identified, which satisfies the requirements of SR IE-A8.As stated in the October 4, 2012 submittal, a process has been developed and implemented to document discussions with plant personnel.

Guidance and Reference Documentation (GARD) NF-AA-PRA-101-2011, Initiating Event Identification and Grouping, provides instructions for interviewing plant personnel (e.g., Operations, Maintenance, Engineering) including identification of potential initiating events and past operating experience.

As additional interviews with plant personnel are completed, the information will be documented in the PRA notebooks.

Question 5 Gap #2. The disposition of this supporting requirements states this is a documentation issue only; however, no discussion is provided on the plant walkdowns and interviews mentioned in the supporting requirement.

Please discuss these walkdowns and interviews which address this supporting requirement (SR).DNC Response As stated in the October 4, 2012 submittal, a process has been developed and implemented to document information on plant walkdowns and discussions with plant personnel.-

Documentation of plant walkdowns and interviews are now maintained in the PRA system notebooks.

Plant walkdowns and interviews with knowledgeable plant personnel are conducted as described in GARD NF-AA-PRA-101-2040, Systems Analysis and Fault Tree Model Development, to confirm system analysis correctly reflects the as-built, as-operated plant Question 6 Gap #3 and #4. The submittal notes that human reliability analysis (HR) SRs[supporting requirements]

HR-G5 and HR-G7 were met, however, an F&O [findings and observations]

was' written for the SRs based on nonsystematic discrepancies that the Serial No.12-764'Docket No. 50-423 Attachment 1, Page 4 of 11 PRA peer review team judged to require correction.

Please correct the impacted human error probabilities.

DNC Response Gap #3 (HR-G5) is based on outdated (circa 2006) talk-through documentation with Operations.

Operator survey information has not been updated to support the basis for revised or new human failure events (HFEs). Therefore, a bounding sensitivity study will be performed by increasing the human error probability (HEP) values by a factor of 10 for the new and revised HEPs without talk-through documentation with Operations.

The increas ed HEP values will be quantified in the PRA model to determine the impact of the sensitivity study. Gap #3 will be addressed in accordance, with the SFCP. Gap#3 will be tracked in the PRA Configuration Control (PRACC) database and assessed for inclusion in the nextMPS3.PRA model update..Gap # 4 (HR-G7) is based on several numerical inconsistencies in the supOporting HEP dependency analysis.DNC performed a sensitivity study associated with the, dependent HEPs calculation errors identified by the peer review team. Three errors out of approximately 80 dependent HEPs were identified.

Dependent HEP Previous Updated HEP-DEP-63 1.1OE-02 7.OOE-03 HEP-DEP-64 5.70E-04 5.80E-04 ,HEP-DEP-67 3.30E-04 5.OOE-04 The following table provides a cross-reference between the dependent HEPs and the individual HEPs.Dependent HEP Individual HEPs*HEP-DEP-63 HEP-C-BAF HEP-C-MFW HEP-DEP-64 HEP-C-SGI HEP-C-MFW HEP-C-BAF HEP-DEP-67 HEP-C-FTSAFW HEP-C-BAF HEP-C-MFW* Following is description of the individual HEPs: HEP-C-BAF Operators Fail to Initiate Bleed and Feed Following a Lossof Auxiliary Feedwater*(AFW)HEP-C-MFW Operators Fail'to Re-Establish Main Feedwater and Condensate to the Steam Generators (SGs)HEP-C-SGI Operators Fail to Isolate Faulted SG During Steam ,Generator Tube Rupture (SGTR)HEP-C-FTSAFW Operators Fail to Start the Affected AFW Pump Given Auto Failure of Auto Actuate Serial No.12-764 Docket No. 50-423 Attachment 1, Page 5 of 11 The model, M310A, was then requantified with the updated dependent HEPs.following table provides the changes in core damage frequency (CDF) and large release frequency (LERF)..The early Model Results CDF (/yr) LERF (/yr)Current (M31OA) 4.19E-06 6.89E-08 Updated 4.13E-06 6.63E-08 Delta -6.OOE-08

-2.60E-09 In addition, the individual HEP importance measures were compared to evaluate any significant change. The following tables provide the change in the Fussell-Vesely (FV)and Risk Achievement Worth (RAW) importance measures.

associated with core damage and large early release, frequencies.

HEP M310A CDF Updated CDF FV RAW FV, RAW HEP-C-BAF 1.22E-02 2.10 1.24E-02 2.11T HEP-C-MFW 5.56E-03 1.12 5.64E-03 1.12 HEP-C-SGI 8.90E-06 1 9.03E-06 1 HEP-C-FTSAFW was truncated (i.e., not included in the model results)HEP M310A LERF Updated LERF FV RAW FV RAW HEP-C-BAF 2.99E-02 3.68 3.1OE-02 3.79 HEP-C-MFW 5.18E-03 1.11 5.38E-03 1.11 HEP-C-SGI 6.02E-04 1.16 6.25E-04 1.16 HEP-C-FTSAFW was truncated (i.e., not included in the model results)As shown above, updating the numerical inconsistencies for the three dependent HEPs results in a negligible impact on the overall risk (CDF and LERF) and the individual HEP importance measures.

However, until these inconsistencies have been incorporated into the MPS3 PRA model, a sensitivity study will be performed as stated in the submittal of October 4, 2012. Gap #4 will be tracked in the PRACC database and assessed for inclusion in the next MPS3 PRA model update.

Serial No.12-764 Docket No. 50-423 Attachment 1, Page 6 of 11 Question 7 Gap #11. Describe the process for identifying human-induced flooding scenarios, and discuss its applicability to TSTF-425 application.

DNC Response Maintenance induced flood scenarios are covered in GARD NF-AA-PRA-101-2073, Flood-Induced Initiating Events. In addition, maintenance-induced flooding scenarios are documented in MPS3 PRA Internal Flooding notebook IF.2 Revision 4, which states: "Maintenance that is conducted while the plant is online in which a closed system is opened is generally limited to cleaning and inspection of heat exchangers and air compressors.

It is possible that, when performing these maintenance activities, that either the wrong component is isolated or maintenance opens the wrong component.

Both of these situations result in a flood initiating event. It is assumed that as the component is opened, water begins to leak out of the component boundary in the form of spray or jets, at which time the mistake is noticed, and recovery actions are taken. Thus, only equipment within the area that may be affected by spray or jet impingement damage are assumed to fail, and there is no propagation to other areas and no damage due to submersion.

Each flood area was then reviewed, analyzed and documented for maintenance-induced flooding events." GAP #11 is a documentation issue only. In accordance with NEI 04-10, only open gaps that would impact the results of the SFCP PRA assessment require sensitivity studies.Question 8 Please identify the plan for closing findings that are identified as "Documentation, issues only." DNC Response"Documentation, issues only" findings are tracked in the PRACC database.

GARD NF-AA-PRA-101-4040, PRA Model Update Planning Guidelines, provides instructions to ensure the items contained in the PRACC database are assessed for closure and model incorporation during the nominal update cycle (approximately every 3 to 5 years).

Serial No.,12-764 Docket No. 50-423 Attachment 1, Page 7 of 11 Question 9 Table 3, "Status of Identified Gaps to Capability Category II of the ASME/ANS PRA Standard," the submittal notes that sensitivity studies will be performed in accordance with PRA procedures for certain gap items. It is not sufficient to perform sensitivity studies in accordance with PRA procedures.

The sensitivity studies must be performed in accordance with Nuclear Energy Institute (NEI) 04-10 guidance.

Please make this clarification as appropriate for the gap items.DNC Response Sensitivity studies will be performed in accordance with NEI 04-10, Rev. 1. The PRA procedures were developed using the guidance in NEI 04-10, Rev. 1.Question 10 The submittal indicates that fire risk and seismic risk would be qualitatively assessed.However, NEI 04-10 guidance for fire events mentions quantitative assessment methods (fire PRA or Fire Induced Vulnerability Evaluation (FIVE)), or qualitative screening.

For seismic events, it mentions quantitative assessment (seismic PRA), qualitative assessment (seismic margins analysis), or qualitative screening.

Please describe in more detail how fire and seismic events would be assessed in terms of NEI 04-10 guidance.DNC Response DNC is committed to evaluating changes to surveillance frequencies in accordance with the guidance provided in NEI 04-10, Rev. 1. NEI 04-10 methodology allows a qualitative screening or bounding analysis to provide justification for acceptability of proposed surveillance frequency changes. Since the MPS3 PRA model does not currently include external events, the NEI 04-10 guidance will be used to evaluate the potential risk impact of external events associated with surveillance frequency changes.Specifically, fire and seismic information from the Millstone Individual Plant Examination of External Events (IPEEE) report will be screened.

The information will be reviewed and qualitatively assessed based on engineering judgment to determine the impact of the external events on surveillance frequency changes.Question 11 The submittal provides no discussion on assessing high winds, floods and other external events. Please describe how these events would be assessed in terms of NEI 04-10 guidance.

Serial No.12-764 Docket No. 50-423 Attachment 1, Page 8 of 11 DNC Response DNC is committed to evaluating changes to surveillance frequencies in accordance with guidance provided in NEI 04-10, Rev. 1. NEI 04-10 methodology allows a qualitative screening or bounding analysis to provide justification for acceptability of proposed surveillance frequency changes. Since the MPS3 PRA model does not currently include external events, the NEI 04-10 guidance will be used to evaluate the potential risk impact of external events associated with surveillance frequency changes.Specifically, flooding and other external event information from the Millstone IPEEE report will be screened.

The information will be reviewed and qualitatively assessed based on engineering judgment to determine the impact of the external events on surveillance frequency changes.Question 12 The submittal provides no discussion on assessment for shutdown events. Please describe how these events would be assessed in terms of NEI 04-10 guidance.DNC Response The MPS3 PRA model does not include shutdown modes. Changes in the surveillance frequencies will be qualitatively assessed using guidance from NEI 04-10, Rev. 1. In particular, the impact of a proposed surveillance frequency change(s) will be evaluated for each of the following plant key safety functions during shutdown: " Decay Heat Removal" Inventory Control" Reactivity Control" Containment Control* Electric Power Availability Question 13 Table 4.3-2 shows the Emergency Generator Load Sequencer marked in the TS under the actuation logic test column, as well as other functional units with the corresponding notation (1). The notation (1) is shown as deleted. Notation (1) is "Each train shall be tested at least every 62 days on a STAGGARD [sic] TEST BASIS." This is also the case for notation (7) in Table 4.3-1 for different functional units. Please discuss why the requirement to test each train is being deleted and why it is consistent with the TSTF-425 program which relocates the frequency of the surveillance to the surveillance frequency control program (SFCP).

Serial No.12-764 Docket No. 50-423 Attachment 1, Page 9 of 11 DNC Response As described in TSTI-425, Rev.3, Section 4.0, Technical Analysis, the phrase "on a STAGGERED TEST BASIS"- is also relocated to licensee control under the SFCP and the defined term, which would no longer be used in TSs, is removed from the Definition section. Relocating the frequency requirement to perform surveillances on a STAGGERED TEST BASIS, along with the periodicity, allows licensees the flexibility to adjust the frequency based on operational experience and risk assessment results. NEI 04-10 contains information to support the correct risk modeling of surveillance frequencies with and without a requirement to perform the surveillance on- a STAGGERED TEST BASIS. Therefore, MPS3 TS notes/notations that describe testing of trains, functional units, etc. on a STAGGERED TEST BASIS, can be relocated to the SFCP in accordance with TSTF-425.The definition of STAGGERED TEST BASIS is being retained in MPS3 TS Definition Section 1 since this terminology is mentioned in Administrative TS Section 6.8.4.h,"Control Room Envelope Habitability Program," which is not the subject of this amendment request and is not proposed to be changed. This represents an administrative deviation from TSTF-425 with no impact on the NRC staffs model safety evaluation dated July 6, 2009 (74 FR 31996).Question 14 Similarly, surveillance requirement 4.3.2.2 deletes requirements on what to test.Discuss why it is consistent with TSTF-425.DNC Response The TS mark-up for Surveillance Requirement 4.3.2.2 has been revised to relocate the surveillance frequency to the SFCP while still maintaining the requirements onwhat to test. A revised marked-up TS page reflecting these changes is provided in Attachment 2.Question 15 It is noted that TS 4.8.2. 1.f is shown in the matrix table as belonging to the SFCP, but it has no markup. Since the surveillance has -an age-related component it would not meet the criteria to be in the SFCP. Please clarify why the submittal proposes to include TS 4.8.2.1. f in the SFCP.

Serial No.12-764 Docket No. 50-423 Attachment 1, Page 10 of 11 DNC Response DNC is not proposing to relocate the surveillance frequency in Surveillance, Requirement 4.8.2.1.f.

to the SFCP. DNC did not provide a mark-up for this surveillance since it was recognized that it contains an age-related component and does not meet the criteria to be in the SFCP. Reference to this surveillance requirement, which appears in Attachment 4, Page 12 of 12 of DNC's October 4, 2012 submittal, should not have been included in the matrix table.Question 16 The following MPS3-specific systems identified in the cross-reference table to NUREG-1431 have no corresponding standard technical specification section:,* Shutdown Margin Monitoring MPS3* pH Tri-sodium Phosphate Storage Baskets 0 Steam Jet Air Ejectors* Secondary Containment MPS3 In addition, the following surveillance requirement identified in the cross-reference table is not in the TSTF-425 scope for the section "RCS Specific Activity":

• Verify XE[sic]-1*33 Justify why these are included in the scope of TSTF-425, or withdraw their proposed inclusion in the SFCP.DNC Response For MPS3 plant-specific surveillances that do not have a corresponding

'surveillance included in' the NUREG-1431 mark-ups provided in TSTF-425, DNC evaluated these surveillance frequencies against the four exclusion criteria delineated in TSTF-425, Rev., 3. The four criteria which exclude' surveillance' frequencies from being relocated are:* Frequencies that reference other approved programs for the specific interval (such as the Inservice Testing Program or the Primary Containment Leakage Rate Testing Program);-Frequencies that are purely event driven (e.g., "Each time the control rod is withdrawn'to the 'full out' position"), Frequencies that are event-driven but have a time component for performing the surveillance on a onetime basis once the event occurs (e'.g., "within '24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after thermal power reaching > 95% RTP"); and Serial No.12-764 Docket No. 50-423 Attachment 1, Page 11 of 11 Frequencies that are related to specific conditions (e.g., battery degradation, age, and capacity) or conditions for the performance of a surveillance requirement (e.g., "drywell to suppression chamberdifferential pressure decrease").

For the MPS3-specific surveillances listed in Question 16 above, DNC determined that these surveillances involve fixed periodic frequencies and do not meet any of the exclusion criteria of TSTF-425, Rev. 3. Therefore, relocation of these frequencies is consistent with TSTF-425, Rev. 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.

The surveillance requirement for Xe-133 was added to the MPS3 TSs under License.Amendment 246 for adoption of TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec." Although not specifically included in the scope of TSTF-425, its surveillance involves a fixed periodic frequency that does not meet any of the exclusion criteria listed above. Therefore, relocation of this frequency is consistent with TSTF-425, Rev. 3, as described above.Question 17 Do the failure probabilities'of structures, systems, and components modeled in the MPS3 PRA include a standby time-related contribution and a cyclic demand-related contribution?

If not, please describe how standby time-related contribution is addressed for extended intervals.

DNC Response The current MPS3 PRA model does not -distinguish between time-related failure contribution (i.e.; the standby time-related failure rate) and cyclic demand-related failure contribution (i.e., the demand stress failure probability).

Since this distinction is not made, DNC, in accordance with NEI 04-10 Rev. 1, will assume all failures are time-related. If a further breakdown of failure probability is required to remove conservatism from the risk impact calculation of a proposed surveillance frequency change, DNC will abide by the cautionary sentence in NEI 04-1,0, Rev. 1, Step 8, third paragraph, which states, "...caution should be taken in dividing the failure probability into time-related and cyclic demand-related contributions because the test-limited risk can be underestimated when only part of the failure rate is considered as being time-related while this may not be the case."

Serial No.12-764 Docket No. 50-423 ATTACHMENT 2 MARKED-UP TECHNICAL SPECIFICATIONS PAGE DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Novem.ber 3, 2000-* INSTRUMENTATION the frequency specified in the Surveillance Frequency Control Program SURVEILLANCE REQUIREMENTS J 4.3.2.1 Each ESFAS instrumentation channel and interloc and the automatic actuation logic and relays shall be demonstrated OPERABLE by performa e of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.4.3.2.2 The ENGINEERED SAFETY FEAT S RESPONSE TIME* of each ESFAS function shall be verified to be within the limit at ca.-.t ccc pcr 1 8 mcnthsi. Each verification shall include at least one train 9,t.h tht both trains are 4v. mvAAL,; least @nee per 36 months and one channel (to include input relays to both trains) per function su'ch that 0ll c-hannckl arc crified at east oncc pzr N tiff 1e -mcnths whcrc N i the total .....r A .f red..d..t eh..AA .in a spocifie ESF.A. .Aiu.... .. tin ....the "T No. of Gha...l." elu... ofab .3..{* The provisions of Specification 4.0.4 are not applicable for response time verification of steam line isolation for entry into MODE 4 and MODE 3 and turbine driven auxiliary feedwater pump for entry into MODE 3.MILLSTONE

-UNIT 3 3/4 3-16 Amendment No. 4-5, 79, 96, 4-00, 4-8-