Text

Supplement to License Amendment Request to Relocate Technical Specification Surveillance Frequencies to Licensee Controlled Program
	ML11262A003
Person / Time
Site:	Millstone
Issue date:	09/12/2011
From:	Price J; Dominion, Dominion Nuclear Connecticut
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	11-474
	Download: ML11262A003 (39)
	v • d • e

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, Virginia 23060 Web Address: www.dom.com September 12, 2011 U. S. Nuclear Regulatory Commission Serial No.11-474 Attention: Document Control Desk NSSL/WDC RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNIT 3 SUPPLEMENT TO LICENSE AMENDMENT REQUEST TO RELOCATE TECHNICAL SPECIFICATION SURVEILLANCE FREQUENCIES TO LICENSEE CONTROLLED PROGRAM By letter dated July 5, 2011, 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 of Licensee Control

- 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 U.S.

Nuclear Regulatory Commission (NRC) and was announced in the Federal Register on July 6, 2009 (74 FR 31996).

In a letter dated August 22, 2011, the NRC provided DNC an opportunity to supplement the LAR identified above. This was based on the NRC staffs acceptance review which identified proposed changes that did not meet the specific relocation criteria delineated in TSTF-425 Rev. 3. Due to schedule impacts associated with Hurricane Irene, the due date for this response was extended from September 7, 2011 to September 12, 2011. provides DNC's response to NRC letter dated August 22, 2011.

New marked-up TS and TS Bases pages for the revised changes are provided in Attachments 2 and 4, respectively, to replace the corresponding pages submitted previously. A revised Significant Hazards Consideration (SHC) Determination is provided in Attachment 3. The revised changes proposed herein do not affect the conclusion of the revised SHC.

If you have any questions regarding this submittal, please contact Wanda Craft at (804) 273-4687.

Sincerely, J. aN rice VjeP esident Nuclear Engineering

Serial No.11-474 Docket No. 50-423 Supplement to SFCP Amendment Request Page 2 of 2 Commitments made in this letter: None Attachments:

1. Supplement to License Amendment Request To Relocate Technical Specification Surveillance Frequencies To Licensee Controlled Program

2. Supplemental Marked-Up TS Pages

3. Revised Significant Hazards Consideration Determination

4. Supplemental Marked-Up TS Bases Pages cc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 C. J. Sanders Project Manager - Millstone Power Station U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 08-B3 Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.11-474 Docket No. 50-423 ATTACHMENT I Supplement to License Amendment Request To Relocate Technical Specification Surveillance Frequencies To Licensee Controlled Program DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.11-474 Docket No. 50-423 Supplement to SFCP Amendment Request Attachment 1, Page 1 of 4 Supplement to License Amendment Request To Relocate Technical Specification Surveillance Frequencies To Licensee Controlled Program By letter dated July 5, 2011, 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 of Licensee Control -

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 U.S. Nuclear Regulatory Commission (NRC) and was announced in the Federal Register on July 6, 2009 (74 FR 31996).

In a letter dated August 22, 2011, the NRC provided DNC an opportunity to supplement the LAR identified above. This was based on the NRC staff's acceptance review which identified proposed changes that did not meet the specific relocation criteria delineated in TSTF-425 Rev. 3. The proposed changes identified by the NRC that do not qualify for relocation are as follows:

Item 1 TS Table 4.3-3, "RadiationMonitoring Instrumentation for Plant Operations Surveillance Requirements," includes the MODES of applicability. Only surveillance frequencies meeting the requirements discussed above [exceptions listed in TSTF-425] may be relocated. Any unique information, such as the MODES of applicability,must be maintainedin the MPS3 TSs. Currently the entire table is proposed to be removed.

DNC Response The previously proposed changes to TS Table 4.3-3 have been revised to maintain TS Table 4.3-3 in TSs, including the MODES of applicability contained in the table.

Only the surveillance frequencies specified in TS Table 4.3-3 will be relocated. The new marked TS page for Table 4.3-3 is provided in Attachment 2.

Item 2 TS Table 4.3-6, "Remote Shutdown Monitoring InstrumentationSurveillance Requirements," does not require a channel calibrationof the Reactor Trip Breaker Indication instrument. The table also includes a note on Function No. 14, Source Range Count Rate, stating that the channel check is requiredbelow P-6 (intermediaterange neutron flux interlock setpoint). Only surveillance frequencies meeting the requirements discussed above may be relocated. Any unique

Serial No.11-474 Docket No. 50-423 Supplement to SFCP Amendment Request Attachment 1, Page 2 of 4 information must be maintainedin the MPS3 TSs. Currently DNC is proposing to remove the entire table.

DNC Response The previously proposed changes to TS Table 4.3-6 have been revised to maintain TS Table 4.3-6 in TSs, including the unique information identified above. Only the surveillance frequencies specified in TS Table 4.3-6 will be relocated. The new marked TS page for Table 4.3-6 is provided in Attachment 2.

Item 3 TS Table 4.3-7, "AccidentMonitoring Instrumentation Surveillance Requirements,"

contains two notes regardingthe channel calibrations(not the frequency of the channel calibration). Only surveillance frequencies meeting the requirements discussed above may be relocated. Any unique information must be maintainedin the MPS3 TSs. Currently DNC is proposingto remove the entire table.

DNC Response The previously proposed changes to TS Table 4.3-7 have been revised to maintain TS Table 4.3-7 in TSs, including the unique information identified above. Only the surveillance frequencies specified in TS Table 4.3-7 will be relocated. The new marked TS pages for Table 4.3-7 are provided in Attachment 2.

Item 4 TS Table 4.7-1, "SecondaryCoolant System Specific Activity Sample and Analysis Program,"requires Isotopic Analysis for Dose Equivalent 1-131 Concentrationat varying frequencies dependent on the Gross RadioactivityDetermination. The threshold cannot be relocatedand must be retainedin TSs. Currently DNC is proposing to remove the entire table.

DNC Response The proposed changes to TS Table 4.7-1 have been revised to maintain TS Table 4.7-1 in TSs, including the concentration thresholds in the table. Only the surveillance frequencies specified in TS Table 4.7-1 will be relocated. The new marked TS page for Table 4.7-1 is provided in Attachment 2.

Item 5 TS SR [surveillance requirement]4.1.1.1.2.1 (b), describes some of the requirements associatedwith an inoperable control rod(s). The SR is event-driven with a time component requiring that shutdown margin shall be determined acceptable, with increasedallowance for the withdrawn worth of the immovable or untrippablecontrol rod(s), with considerationfor several factors every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

CurrentlyDNC is proposing to relocate this event-driven SR.

Serial No.11-474 Docket No. 50-423 Supplement to SFCP Amendment Request Attachment 1, Page 3 of 4 DNC Response In this case, the presence of the word "and" in the preceding SR (i.e., 4.1.1.1.2.1a) is not intended to be a logical connector which places the constraints of this event-driven surveillance on the remaining portion of the surveillance (i.e., 4.1.1.1.2.lb).

Station procedures require SR 4.1.1.1.2. lb to be performed at the specified frequency while in Modes 3, 4 and 5 and is not event-driven; therefore, no change to the original markup is required.

This situation (where use of the words "and" and "or" are not intended to be used as logical connectors) also applies to the following SRs: 4.1.1.2.1, 4.1.3.5, 4.2.2.1.2d, 4.2.2.1.4d, 4.2.3.1.2, 4.2.3.1.3, 4.4.9.3.1a, 4.5.2g, 4.6.5.1.1. Accordingly, no changes to the previously submitted markups are required.

Item 6 In addition, DNC is proposing to make changes outside of the scope of the model submittal (74 FR 31996). These additionalchanges were not addressed in the Significant Hazards Consideration. Please submit a supplement to the Significant Hazards Considerationto address these additionalchanges.

DNC Response A revised Significant Hazards Consideration (SHC) Determination is provided in Attachment 3. This revised SHC addresses the MPS3 specific surveillance frequencies that are outside of the scope of the model submittal (74 FR 31996).

Along with the corrections described above, additional replacement pages are included in Attachment 2 to correct or amend the TS markups previously submitted in the July 5, 2011 LAR. As a result of the changes proposed in this supplement, certain markups previously submitted are no longer necessary. These pages are provided in Attachment 2 and labeled "For Information Only" to identify the pages which can be removed from the July 5, 2011 submittal.

A summary of the changes provided in this supplement are identified below:

TS Page Number Reason for Change vi Removed deletion of Tables 4.3-3, 4.3-6 and 4.3-7.

x Removed deletion of Table 4.7-1.

1-8 Page added to provide the frequency notation for SFCP in TSs.

3/4 3-1 Table 4.3-1 in SR 4.3.1.1 was incorrectly deleted in previous submittal.

Serial No.11-474 Docket No. 50-423 Supplement to SFCP Amendment Request Attachment 1, Page 4 of 4 TS Page Number Reason for Change 3/4 3-16 Removed deletion of Table 4.3-2 (with inserted text) in SR 4.3.2.1.

3/4 3-36 "SFCP" was truncated in insert box at top of page.

Added missing arrow to the 92-day surveillance frequency in Table 3/4 3-41 Notation 2. As discussed with NRC on 8/30/11, added change to remove reference to the 18 month surveillance frequency in Table Notation 3. Also, removed deletion of "R" in Table Notation 4.

3/4 3-42 Removed deletion of Table 4.3-3 (with inserted text) in SR 4.3.3.1.

3/4 3-45 See DNC response to Item 1 above.

3/4 3-53 Removed deletion of Table 4.3-6 (with inserted text) in SR 4.3.3.5.1.

3/4 3-58 See DNC response to Item 2 above.

3/4 3-59a Removed deletion of Table 4.3-7 (with inserted text) in SR 4.3.3.6.1.

3/4 3-62 See DNC response to Item 3 above.

3/4 3-63 See DNC response to Item 3 above.

3/4 4-21 a Removed deletion of Table 4.3-3 (with inserted text) in SR 4.4.6.1.

As discussed with the NRC on 8/30/11, added change to relocate surveillance frequency in SR 4.7.1.3.2.

3/4 7-7 Removed deletion of Table 4.7-1 (with inserted text) in SR 4.7.1.4.

3/4 7-8 See DNC response to Item 4 above.

As discussed with the NRC on 9/1/11, added change to relocate surveillance frequency in SR 4.7.5b Removed proposed changes, however, TS 3.7.14 is being separately 3/4 7-32 deleted from TSs in accordance with issuance of License Amendment 250.

As discussed with the NRC on 8/30/11, deleted reference to surveillance 3/4 8-12 interval in second sentence of SR 4.8.2.1e for clarity. Also, removed proposed relocation of surveillance frequency in SR 4.8.2.lf.

3/4 10-4 Corrected numbering error in SR 4.10.3 (should read 4.10.3.3)

Consistent with the TS changes described above, revised TS Bases pages are provided in Attachment 4 for information only. Insert 2 for the TS Bases pages is provided in the July 5, 2011 submittal.

Serial No.11-474 Docket No. 50-423 ATTACHMENT 2 Supplemental Marked-UD TS Paaes DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

For Information Only January 29, 2001 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-5 DELETED TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ................ 3/4 3-36 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations ............................................... 3/4 3-42 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS ..................................................................... 3/4 3-43 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS ............................. 3/4 3-45 TABLE 3.3-7 DELETED TABLE 4.3-4 DELETED TABLE 3.3-8 DELETED TABLE 4.3-5 DELETED Remote Shutdown Instrumentation ............................................................ 3/4 3-53 TABLE 3.3-9 REMOTE SHUTDOWN INSTRUMENTATION ...................................... 3/4 3-54 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ....................................................... 3/4 3-58 Accident Monitoring Instrumentation ........................................................ 3/4 3-59 TABLE 3.3-10ACCIDENT MONITORING INSTRUMENTATION ............................... 3/4 3-60 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQ U IRE M ENTS ...................................................................................... 3/4 3-62 TABLE 3.3-11 DELETED TABLE 3.3-12DELETED TABLE 4.3-8 DELETED MILLSTONE - UNIT 3 vi Amendment No. 84, 94-, 188, 193

For Information Only September 15, 2006 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-3 STEAM LINE SAFETY VALVES PER LOOP ................................... 3/4 7-3 A uxiliary Feedwater System ................................................................. 3/4 7-4 Dem ineralized Water Storage Tank ....................................................... 3/4 7-6 Specific Activity .................................................................................... 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ............................................ 3/4 7-8 M ain Steam Line Isolation Valves ........................................................ 3/4 7-9 Steam Generator Atmospheric Relief Bypass Lines ............................. 3/4 7-9a 3/4.7.2 D ELETED ............................................................................................. 3/4 7-10 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM ..3/4 7-l1 3/4.7.4 SERVICE WATER SYSTEM ............................................................... 3/4 7-12 3/4.7.5 ULTIM ATE HEAT SINK ...................................................................... 3/4 7-13 3/4.7.6 D ELET ED ............................................................................................. 3/4 7-14 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ......... 3/4 7-15 3/4.7.8 D EL ET E D ............................................................................................. 3/4 7-18 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM ..................................... 3/4 7-20 3/4.7.10 SN U B B ER S .......................................................................................... 3/4 7-22 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL ....................3/4 7-27 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST .............. 3/4 7-29 3/4.7.11 D ELET E D ............................................................................................ 3/4 7-30 3/4.7.12 DELETED TABLE 3.7-4 DELETED TABLE 3.7-5 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING ........................................... 3/4 7-32 TABLE 3.7-6 AREA TEMPERATURE MONITORING ........................................... 3/4 7-33 MILLSTONE - UNIT 3 x Amendment No. 62, 84, 4-00, 4-60, 2-4-4, 232

O*be-r- 2, 1-997 TABLE 1.1 FREOUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

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 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

P Completed prior to each release.

TI FSF CP] At the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 1-8

NoVemIe, 3, 2000 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-I shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-I.

ACTION:

As shown in Table 3.3-I.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-I.

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be verified to be within its limit a ,ca- , oncc per 18 mcnths. Neutron detectors and speed sensors arequeny specifesponse tuiverfication FEquh.en Progr ificytCont alam l l ,to hain,, stetta t k oh ÷, -ri.-. "--.*

.- g._ at__

l e s once. pe 3-6*

.... month

... ...

and ch n e

... inchJ_

....

Ithe frequency specifi dt in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 3/4 3-1 Amendment No. 4-S, 49, 94-, +00,--6 Nvme 3,2000 INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME* of each ESFAS function shall be verified to be within the limit at est ..... per 18, mt.. . Eac.h vetiVii leattonce per N' tiiimes t 8 montths wltI~I a~,t speetlv ifie ESFAS JL f.,*nebttie as15 shownl E' IlI I S t toal rtiniltU, VJCI**.,

VL %111,nllal ll of iutanuant chlteftrhra l ,UffIIuni e i l e

3.3-3I ii

.*'

ii

[the frequency specified in the Surveillance Frequency Control Proqram I

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, ;9, 96, 4-00, +8

Replace each marked through surveillance frequency in the Check, Calibrate, and Test columns with "SFCP" Cjr LMASTER SLAV MODES CHANNEL CHANNE OPE IONA TION AYSTIO RELAY RELAfE FOR WHICH FUNCTIONAL UNIT (Reactor Trip,

. Safety Injection CHECK CALIB TI TESTi TET G ES TEST TESi XY SURVEILLANCE IS REQUIRED Feedwater Isolation, Control Building/

Isolation (Manual Initiation Only), Sta" '

Diesel Generators, and Service W ttr) *',

a. Manual Initiation N.A. N.A. N.A. -R*- N.A. N.A. N.A. 1,2,3,4

b. Automatic Actuation N.A. N.A. N.A. N.A. NA N.A(4) 1,2,3,4 Logic and Actuation Relays

c. Containment Pressure- -S N.A. N.A. N.A. N.A. 1,2,3 High-1

d. Pressurizer Pressure- -S --Q N.A. N.A. N.A. N.A. 1,2,3 Low

e. Steam Line Pressure- --S- -R N.A. N.A. N.A. N.A. 1,.2,3 Low

2. Containment Spray CD

a. Manual Initiation N.A. N.A. N.A. --R- N.A. N.A. N.A. 1,2,3,4

b. Automatic Actuation N.A. N.A. N.A. N.A. -- l -Q (4) 1,2,3,4 Logic and Actuation Relays

c. Containment Pressure- -R --Q N.A. N.A. N.A. N.A. 1,2,3,4 High-3

Septem.~

1 ber 18, 2008 TABLE 4.3-2 (Continued)

TABLE NOTATION I Ea.h train shall be testedL at east every 62 days on a STAGGERED TEST BASIS.

2. This surveillance may be performed continuously by the emergency generator load sequencer auto test system as long as the EGLS auto test system is demonstrated OPERABLE by the

_8Y9lAt the performance.

frequency of an ACTUATION specified LOGIC TEST at Iaaon e pen 92 in the Surveillance Frequency C ntrol Program

3. Oin a ionvthly basis, a loss of voltage condition will be initiated at each u dervoltage monitoring relay to verify individual relay operation. Setpoint verificati n and actuation of the associated logic and alarm relays will be performed as part of the HANNEL CALIBRATION required onee pe, 18 1 Inthl.

4. For Engineered Safety Features Actuation System functional units wi only Potter &

Brumfield MDR series relays used in a clean, environmentally contr lied cabinet, as discussed in Westinghouse Owners Group Report WCAP- 13900, th surveillance interval for slave relay testing is R.

the frequency specified in the Surveillance Frequency Control Program I MODES 1, 2, 3, and 4.

During movement of recently irradiated fuel assemblies. #

MILLSTONE - UNIT 3 3/43-41 Amendment No. 44, 74, 9, +l0,4-29,

+98, 203, 2+9, 22-9, 24-, 2+3-

Octebcr 25, 1990 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels for plant operations 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 for plant operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Table 3.3-6.

c. The provisions of Specification 3.0.3 are not applicable. -I-SURVEILLANCE REQUIREMENTS required 4.3.3.1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST for the MODES and at the frequencies shown in Table 4.3-3.

MILLSTONE - UNIT 3 3/4 3-42 Amendment No. -54

]Replace each marked through surveillance frequency in the Check, Calibrate, and Test columns with "SFCP" I AB L 4.3-0 RADIATION MONITO NG IN TRU TATION FOR PLANT ztTI OPERATIONS 4RVEILI JNCE RE.IREMENTS ANALOG CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED

1. Containment

a. Deleted

b. RCS Leakage Detection

1) Particulate Radio- R Q 1,2,3,4 activity

2) Deleted

2. Fuel Storage Pool Area Monitors S *

a. Radiation Level R Q TABLE NOTATIONS

With fuel in the fuel storage pool area.

I z

C

(

C

(

C t

C C

-Jrauy 3, t995 INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The Remote Shutdown Instrumentation transfer switches, power, controls and monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With one or more Remote Shutdown Instrumentation transfer switches, power, or control circuits inoperable, restore the inoperable switch(s)/circuit(s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Entry into an OPERATIONAL MODE is permitted while subject to these ACTION requirements.

SURVEILLANCE REQUIREMENTS required 4.3.3.5.1 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.

4.3.3.5.2 Each Remote Shutdown Instrumentation transfer switch, power and control circuit including the actuated components, shall be demonstrated OPERABLE at least on- e per!-4 the frequency specified in the Surveillance Frequency Control Program ]

MILLSTONE - UNIT 3 3/4 3-53 Amendment No. , fO7-

IReplace each marked through surveillance frequency in the Check and Calibrate columns with "SFCP" I TABLE 4.3-. (G INSTROMUEN1AIO CI REMOTE SHUTDOWN MONITOPu IIREMENTS \ý SURVEILLANCE REOU CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION

1. Reactor Trip Breaker Indication M N.A.

00 2. Pressurizer Pressure M

3. Pressurizer Level M R-R q--

4. Steam Generator Pressure M RL

5. Steam Generator Water Level M R -I--

zD 6. Auxiliary Feedwater Flow Rate M

7. Loop Hot Leg Temperature M

8. Loop Cold Leg Temperature M R

9. Reactor Coolant System Pressure M (Wide Range)

CDJ

10. DWST Level M R

11. RWST Level M

12. Containment Pressure M R

13. Emergency Bus Voltmeters M R

14. Source Range Count Rate M* R

15. Intermediate Range Amps M R

16. Boric Acid Tank Level M R

When below P-6 (intermediate range neutron flux interlock setpoint).

LIMITING CONDITION FOR OPERATION (Continued) action taken, the cause of the inoperability, and the plans and schedule for restoring the channel to OPERABLE status.

f. With the number of OPERABLE channels for the reactor vessel water level monitor less than the minimum channels OPERABLE requirements of Table 3.3-.10, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repairs are feasible without shutting down or:

1. Initiate an alternate method of monitoring the reactor vessel inventory;

2. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the channel(s) to OPERABLE status; and

3. Restore the channel(s) to OPERABLE status at the next scheduled refueling.

g. Entry into an OPERATIONAL MODE is permitted while subject to these ACTION requirements.

SURVEILLANCE REQUIREMENTS frequired 4.3.3.6.1 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3-7.

4.3.3.6.2 Deleted MILLSTONE - UNIT 3 3/4 3-59a Amendment No. 44, 54 , 44-2-,4

lReplace each marked through surveillance frequency in the Check and Calibrate columns with "SFCP" zC- TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTAI0 SURV *L CE REQUIRE-S CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION

1. Containment Pressure

a. Normal Range M-

b. Extended Range M-

2. Reactor Coolant Outlet Temperature - THOT (Wide Range)

3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) M-C-

t.,J P 4. Reactor Coolant Pressure - Wide Range M- -"

CD

5. Pressurizer Water Level 4"-

P

6. Steam Line Pressure -NI- 4-

7. Steam Generator Water Level - Narrow Range -1*- 4-

8. Steam Generator Water Level - Wide Range -NI- z -1*-

0, 9. Refueling Water Storage Tank Water Level

10. Demineralized Water Storage Tank Water Level -MI-

11. Auxiliary Feedwater Flow Rate t-

12. Reactor Coolant System Subcooling Margin Monitor

¢--

13. Containment Water Level (Wide Range) -NI-

14. Core Exit Thermocouples

15. DELETED

IReplace each marked through surveillance frequency in the Check and Calibrate columns with "SFCP" I K

C,H TABLE 4.3-7 (Conued ACCIDENT MONITORING INSTRUMENTATION SURVE LANCE REQUQOTS zrn CHANNEL CHANNEL z~INSTRUMENT CHECK CALIBRATION q 16. Containment Area - High Range Radiation Monitor

17. Reactor Vessel Water Level

18. Deleted -I-

19. Neutron Flux -M- -R 0L'

CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/h and a one point calibration check of the detector below 10 R/h with an installed or portable gamma source.

- Electronic calibration from the ICC cabinets only.

z0 I

Septmiiiei 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and at least once per 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months thereafter, and

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

Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

a. Containment Atmosphere Particulate Radioactivity Monitoring System-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and

b. Containment Drain Sump Monitoring System-performance of CHANNEL -<

CALIBRATION at least cc -^- 24 .... hs Ithe frequency specified in the Surveillance Frequency Control Proqram I MILLSTONE - UNIT 3 3/4 4-21 a Amendment No. 244- X'

Septembcr 11, 1997 PLANT SYSTEMS DEMINERALIZED WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The demineralized water storage tank (DWST) shall be OPERABLE with a water volume of at least 334,000 gallons.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With the DWST inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a. Restore the DWST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , or

b. Demonstrate the OPERABILITY of the condensate storage tank (CST) as a backup supply to the auxiliary feedwater pumps and restore the DWST to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The DWST shall be demonstrated OPERABLE at !cast once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the water volume is within its limits when the tank is the sup y source for the auxiliary feedwater ,/

pumps.

4.7.1.3.2 The CST shall be demonstrated OPERABLE/, - V2a by verifying that the combined volume of both the DWST and CST is t least 3 000 gallons of water whenever the CST and DWST are the supply source for he aux" ary feedwater pumps.

Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 3 3/4 7-6 Amendment No. 44ft-

For Information Only]

January 31, 1986 PLANT SYSTEMS SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microCurie/gram DOSE EQUIVALENT 1-131.

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

ACTION:

With the specific activity of the Secondary Coolant System greater than 0.1 microCurie/gram DOSE EQUIVALENT 1-131, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.1.4 The specificd activity of the Secondary Coolant System shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-I.

MILLSTONE - UNIT 3 3/4 7-7

TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY

1. Gross Radioactivity At least enee per 72 heur-s.

Determination

2. Isotopic Analysis for DOSE a) 9nee per 31 daldys, when-EQUIVALENT 1-131 ever the gross radio-Concentration /activity determination indicates concentrations greater than 10% of the allowable limit for radioiodines.

b) Once per 6 n , when-ever the gross radio-activity determination indicates concentrations less than or equal to 10%

of the allowable limit for radioiodines.

P JAt the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 3/4 7-8

Agst 28, 995-PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with an average water temperature [,

of less than or equal to 75 0 F.

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

ACTION:

If the UHS temperature is above 75°F, monitor the UHS temperature once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If the UHS temperature does not drop below 75 0 F during this period, place the plant in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . During this period, if the UHS temperature increases above 77°F, place the plant in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.5 The UHS shall be determined OPERABLE:

a. At aby verifying the average water temperature to be within lim'i*

b. At 9,by verifying the average water temperature to be within water temperature exceeds 70'F.

MILLSTONE - UNIT 3 3/4 7-13 Amendment No. ff9-

I For Information Only June 24, 1997 PLANT SYSTEMS 3/4.7.14 AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION 3.7.14 The temperature limit of each area shown in Table 3.7-6 shall not be exceeded.

APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE.

ACTION:

With one or more areas exceeding the temperature limit(s) shown in Table 3.7-6:

a. By less than 207F and for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , record the cumulative time and the amount by which the temperature in the affected area(s) exceeded the limit(s).

b. By less than 20'F and for greater than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area(s) exceeded the limit(s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment. The provisions of Specification 3.0.3 are not applicable.

c. With one or more areas exceeding the temperature limit(s) shown in Table 3.7-6 by greater than or equal to 20'F, prepare and submit a Special Report as required by ACTION b. above and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either restore the area(s) to within the temperature limit(s) or declare the equipment in the affected area(s) inoperable.

SURVEILLANCE REQUIREMENTS 4.7.14 The temperature in each of the'areas shown in Table 3.7-6 shall be determined to be within its limits:

a. At least once per seven days when the alarm is OPERABLE, and;

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the alarm is inoperable.

MILLSTONE - UNIT 3 3/4 7-32 Amendment No. 7, 9-5, +-00,141

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At !et - "--r 92 day: and within 7 days after a battery discharge with battery t inal voltage below 110 volts, or battery overcharge with battery terminal v itage above 150 volts, by verifying that:

) The parameters-in Table 4.8-2a meet the Category B limits,

2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10-6 ohm, and

3) The average electrolyte temperature of six connected cells is above 60'F.

C. At'a ucby verifying that:

the frequency I) The cells, cell plates, and battery racks show no visual indication of specified in the physical damage or abnormal deterioration, Surveillance

2) The cell-to-cell and terminal connections are clean, tight, and coated with Frequency anticorrosion material, Control Program

3) The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10- 6 ohm, and

) Each battery charger will supply at least the amperage indicated in Table 4.8-2b at greater than or equal to 132 volts for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . k"

d. t least p 18 ... t.', during shutdown, by verifying that the battery on apacity is adequate to supply and maintain in OPERABLE status all of the actual o simulated emergency loads for the design duty cycle when the battery is su 'jectedto a battery service test;

e. At ,at oee per 60 month , during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. On,,per 6 m..onth,nt- , r va- this performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1d.; and

f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.

MILLSTONE - UNIT 3 3/4 8-12 Amendment No. 64, 79, 00, 49--

juy31, t986 SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,

b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and

c. The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 541°F.

APPLICABILITY: MODE 2.

ACTION:

a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers.

b. With a Reactor Coolant System operating loop temperature (Tavg) less than 541 'F, restore Tavg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes.

SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at lcast ,ncz per hour during PHYSICS TESTS.

4.10.3.2 Each Intermediate and P er Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST w in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.

4.10.3. The Reactor Coolant System te erature (Tavg) shall be determined to be greater than or equal to 541°F at uring PHYSICS TESTS.

Ithe freciuencv specified in the Surveillance Freauencv Control Procaramr MILLSTONE - UNIT 3 3/4 10-4

Serial No.11-474 Docket No. 50-423 ATTACHMENT 3 Revised Significant Hazards Consideration Determination DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.11-474 Docket No. 50-423 Supplement to SFCP Amendment Request Attachment 3, Page 1 of 2 REVISED NO SIGNIFICANT HAZARDS CONSIDERATION Description of Amendment Request:

This amendment request involves the adoption of approved changes to the standard technical specifications (STS) for Westinghouse Pressurized Water Reactors (NUREG-1431), to allow relocation of specific technical specification (TS) surveillance frequencies to a licensee controlled program. The proposed changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML090850642), "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b" and are described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996).

The proposed changes are consistent with NRC-approved Industry/lTSTF Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b." For MPS3 plant-specific surveillances not included in the NUREG-1431 mark-ups provided in TSTF-425 (identified in Attachment 4), DNC has determined that since these surveillances involve fixed periodic frequencies, 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 proposed changes relocate surveillance frequencies to a licensee controlled program, the Surveillance Frequency Control Program (SFCP). The changes are applicable to licensees using probabilistic risk guidelines contained in NRC-approved NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. 071360456). In addition, administrative/editorial deviations of the TSTF-425 inserts and the existing TS wording are being proposed to fit the custom TS format.

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

1. Do the proposed changes involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No.

The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the TSs for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Serial No.11-474 Docket No. 50-423 Supplement to SFCP Amendment Request Attachment 3, Page 2 of 2

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

Response: No.

No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in the margin of safety?

Response: No.

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

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

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

Serial No.11-474 Docket No. 50-423 ATTACHMENT 4 Supplemental Marked-Up TS Bases Pacies DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

For Information Only LBDCR No. 06-MP3-014 June 22, 2006 POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued)

(2) APLND (for base load operation). Penalty deviation minutes for base load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than +/-12 steps, indicated, from the group demand position;

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

FNAH will be maintained within its limits provided Conditions a. through d. above are maintained. The relaxation of FNAH as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

The FNAH as calculated in Specification 3.2.3.1 is used in the various accident analyses where FNAH influences parameters other than DNBR, e.g., peak clad temperature, and thus is the maximum "as measured" value allowed.

The RCS total flow rate and FNAH are specified in the CORE OPERATING LIMITS REPORT (COLR) to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow rate, that is based on 10% steam generator tube plugging, is retained in the Technical Specifications.

MILLSTONE - UNIT 3 B 3/4 2-3 Amendment No. -50,60, 24-7,

LBDCR?4e. 08 NP3 014 Oetuabr 21, 2008 POWER DISTRIBUTION LIMITS in accordance with the Surveillance BASES Frequency Control Program 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and C FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Co mu d)

Margin is maintained between the safety analysis limit DNB nd t design limit DNBR. This margin is more than sufficient to offset the effect of rod bow nd an other DNB penalties that may occur. The remaining margin is available for plant de gn flex* ility.

When an FQ measurement is taken, an allowance for bo experim ntal error and manufacturing tolerance must be made. An allowance of 5% is appr nate for a ull core map taken with the incore detector flux mapping system and a 3% allow nce is appr riate for manufacturing tolerance.

The heat flux hot channel factor, FQ(Z), is measued periodically sing the incore detector system.

These measurements are generally taken with e core at or near steady state conditions. Using the measured three dimensional power distri tions, it is possible to derive FQM(Z), a computed value of FQ(Z). However, because this val represents a steady state condition, it does not include the variations in the value of FQ( that are present during nonequilibrium situations.

To account for these possible variation the steady state limit of FQ(Z) is adjusted by an elevation dependent factor appropriate to either OC or base load operation, W(Z) or W(Z)BL, that accounts for the calculated worst ca e transient conditions. The W(Z) and W(Z)BL, factors described above for normal operat' n are specified in the COLR per Specification 6.9.1.6. Core monitoring and control under no steady state conditions are accomplished by operating the core within the limits of the appropr' te LCOs, including the limits on AFD, QPTR, and control rod insertion. Evaluation of the s ady state FQ(Z) limit is performed in Specification 4.2.2.1.2.b and 4.2.2.1.4.b while evaluation onequilibrium limits are performed in Specification 4.2.2.1.2.c and 4.2.2.1.4.c.

When RCS flow rate a d FNAH are measured, no additional allowances are necessary prior to comparison with the imits of the Limiting Condition for Operation. Measurement errors for RCS total flow rate nd for FNAH have been taken into account in determination of the design DNBR value.

The measureme t error for RCS total flow rate is based upon performing a precision heat balance and using the suit to calibrate the RCS flow rate indicators. To perform the precision heat balance, the strumentation used for determination of steam pressure, feedwater pressure, feedwater perature, and feedwater venturi AP in the calorimetric calculations shall be calibrated at lcaet cn.- per I8 menths. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner.

Any fouling which might bias the RCS flow rate measurement can be detected by monitoring and 4 trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

MILLSTONE - UNIT 3 B 3/4 2-4 Amendment No. 2, 60, 70, 2-4-7,

For Information Only LBDCR No. 04-MP3-015 February 24, 2005 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

For slave relays, or any auxiliary relays in ESFAS circuits that are of the type Potter & Brumfield MDR series relays, the SLAVE RELAY TEST is performed at an "R" frequency (at least once every 18 months) provided the relays meet the reliability assessment criteria presented in WCAP-13878, "Reliability Assessment of Potter and Brumfield MDR series relays," and WCAP-13900, "Extension of Slave Relay Surveillance Test Intervals." The reliability assessments performed as part of the aforementioned WCAPs are relay specific and apply only to Potter and Brumfield MDR series relays. Note that for normally energized applications, the relays may have to be replaced periodically in accordance with the guidance given in WCAP-13878 for MDR relays.

REACTOR TRIP BREAKER This trip function applies to the reactor trip breakers (RTBs) exclusive of individual trip mechanisms. The LCO requires two OPERABLE trains of trip breakers. A trip breaker train consists of all trip breakers associated with a single RTS logic train that are racked in, closed, and capable of supplying power to the control rod drive (CRD) system. Thus, the train may consist of the main breaker, bypass breaker, or main breaker and bypass breaker, depending upon the system configuration. Two OPERABLE trains ensure no single random failure can disable the RTS trip capability.

These trip functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip functions must be OPERABLE when the RTBs or associated bypass breakers are closed, and the CRD system is capable of rod withdrawal.

BYPASSED CHANNEL* - Technical Specifications 3.3.1 and 3.3.2 often allow the bypassing of instrument channels in the case of an inoperable instrument or for surveillance testing.

A BYPASSED CHANNEL shall be a channel which is:

Required to be in its accident or tripped condition, but is not presently in its accident or tripped condition using a method described below; or Prevented from tripping.

MILLSTONE - UNIT 3 B 3/4 3-2b Amendment No. 2-4-9, Acknowledged by NRC letter dated 08/25/05

LBDCR Nfu. 06-MP3-02"3 August 10, 2.,6 EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued) flush upon heat exchanger return to service and procedural compliance is relied upon to ensure that gas is not present within the heat exchanger u-tubes.

Surveillance Requirement 4.5.2.C.2 requires that the visual inspection of the containment be performed at least once daily if the containment has been entered that day and when the final containment entry is made. This will reduce the number of unnecessary inspections and also reduce personnel exposure. Insert 2 Surveillance Requirement 4.5.2.d.2 addresses periodic inspectio ,-the containment sump to ensure that it is unrestricted and stays in proper operating condition. Ic -4 month fr.quancy is-,,j-baIcJ*

seddedotl o hc11 e t, L t;. cl o 1111 H -I v¥...

e,11

1 1 under tlhe k , IIat app l -1l5Mu ,- ld the~. need to have aeeess to the loeationf. This frequc1 1ey is suffic 1 1 t t detcc bonaldgaao and is confliried by upeating experience.

The Emergency Core Cooling System (ECCS) has several piping cross connection points for use during the post-LOCA recirculation phase of operation. These cross-connection points allow the Recirculation Spray System (RSS) to supply water from the containment sump to the safety injection and charging pumps. The RSS has the capability to supply both Train A and B safety injection pumps and both Train A and B charging pumps. Operator action is required to position valves to establish flow from the containment sump through the RSS subsystems to the safety injection and charging pumps since the valves are not automatically repositioned. The quarterly stroke testing (Technical Specification 4.0.5) of the ECC/RSS recirculation flowpath valves discussed below will not result in subsystem inoperability (except due to other equipment manipulations to support valve testing) since these valves are manually aligned in accordance with the Emergency Operating Procedures (EOPs) to establish the recirculation flowpaths. It is expected the valves will be returned to the normal pre-test position following termination of the surveillance testing in response to the accident. Failure to restore any valve to the normal pre-test position will be indicated to the Control Room Operators when the ESF status panels are checked, as directed by the EOPs. The EOPs direct the Control Room Operators to check the ESF status panels early in the event to ensure proper equipment alignment. Sufficient time before the recirculation flowpath is required is expected to be available for operator action to position any valves that have not been restored to the pretest position, including localmanual valve operation. Even if the valves are not restored to the pre-test position, sufficient capability will remain to meet ECCS post-LOCA recirculation requirements. As a result, stroke testing of the ECCS recirculation valves discussed below will not result in a loss of system independence or redundancy, and both ECCS subsystems will remain OPERABLE.

When performing the quarterly stroke test of 3SIH*MV8923A, the control switch for safety injection pump 3SIH*PIA is placed in the pull-to-lock position to prevent an automatic pump start with the suction valve closed. With the control switch for 3SIH*PIA in pull-to-lock, the Train A ECCS subsystem is inoperable and Technical Specification 3.5.2, ACTION a., applies. This ACTION statement is sufficient to administratively control the plant configuration with the automatic start of 3SIH*P1A defeated to allow stroke testing of 3SIH*MV8923A. In addition, the EOPs and the ESF status panels will identify this abnormal plant configuration, if not corrected following the termination of the surveillance testing, to the plant operators to allow restoration of the normal post-LOCA recirculation flowpath. Even if system restoration is not accomplished, sufficient equipment will be available to perform all ECCS and RSS injection and recirculation functions, provided no additional ECCS or RSS equipment is inoperable, and an additional single failure does not occur (an acceptable assumption since the Technical Specification ACTION statement limits the plant configuration time such that no additional equipment failure need be postulated). During the injection phase the redundant subsystem (Train B) is fully functional, as is a significant portion of the Train A subsystem. During the recirculation phase, the Train A RSS subsystem can supply water from the containment sump to the Train A MILLSTONE - UNIT 3 B 3/4 5-2b Amendment No. +-00, 4-7, 445-7

i,,8 LBDCR 07-MfvF3-033 J.n. 25, 2007 PLANT SYSTEMS BASES SURVEILLANCE REQUIREMENTS For the surveillance requirements, the UHS temperature is measured at the locations described in the LCO write-up provided in this section.

Surveillance Requirement 4.7.5.a verifies that the UHS is capable of providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature. -he dut iti, t!I, appl.,iakbl MODES. This surveillance requirement verifies that the 2avera Insert water to 75°F.

temperature of the UHS is less than or equal Surveillance Requirement 4.7.5.b requires that the UHS temperatur e monitored on an increased frequency whenever the UHS temperature is great~er t n 707F during the applicable MODES. The intent of this Surveillance Requirement is to, rease the awareness of plant personnel regarding UHS temperature trends above 70°F. , ,i- is based a expcicne reiated to trendin~g of the parameter.~ vatiatiuJIs dating th appliable fvtfDES.

3/4.7.6 DELETED 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM BACKGROUND The control room emergency ventilation system provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke. Additionally, the system provides temperature control for the control room envelope (CRE) during normal and post-accident operations.

The control room emergency ventilation system is comprised of the CRE emergency air filtration system and a temperature control system.

The control room emergency air filtration system consists of two redundant systems that recirculate and filter the air in the CRE and a CRE boundary that limits the inleakage of unfiltered air. Each control room emergency air filtration system consists of a moisture separator, electric 4' heater, prefilter, upstream high efficiency particulate air (HEPA) filter, charcoal adsorber, downstream HEPA filter, and fan. Additionally, ductwork, valves or dampers, and instrumentation form part of the system.

The CRE is the area within the confines of the CRE boundary that contains the spaces that control room occupants inhabit to control the unit during normal and accident conditions. This area encompasses the control room, and other non-critical areas including adjacent support offices, MILLSTONE - UNIT 3 B 3/4 7-10 Amendment No. 44-9, 446, 444, 244,

v t e Letter Sequence Supplement
TAC:ME6695, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Withdrawn, Closed)
Initiation Acceptance, Acceptance Supplement, Supplement Administration Questions 8 December 2011 Answers 19 January 2012 Results Withdrawal Other:
MONTHYEAR2011-09-30 [Table View]

ML11262A003

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