|
---|
Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20211Q3361999-09-0707 September 1999 Proposed Tech Specs Removing Current Special Exception Which Precludes Applying Eighteen Month Functional Testing Surveillance to SG Hydraulic Snubbers ML20211H6471999-08-25025 August 1999 Proposed Defueled Tech Specs,Revising Sections 5.6.1,5.7.2 & 5.7.3 & Adding Proposed Section 5.6.4 to Reflect ACs Contained in NUREG-1433 ML20210Q5211999-08-0505 August 1999 Proposed Tech Specs Sections 3.8.3.2,4.6.2.1,4.6.2.2, 4.8.1.1,4.9.12 & Bases Section B 3/4.3.2,B 3/4.6.1.2 & B 3/4.8.4,incorporating Editorial Revs ML20210C6091999-07-16016 July 1999 Proposed Tech Specs Relocating Selected TS Related to Refueling Operations & Associated Bases to Plant TRM ML20206U1041999-05-17017 May 1999 Proposed Tech Specs Section 4.4.6.2.2.e,deleting Reference to ASME Code Paragraph IWV-3472(b) Re Frequency of Leakage Rate Testing for Valves Six Inches Nominal Pipe Size & Larger ML20205R2751999-04-19019 April 1999 Proposed Tech Specs,Reflecting Permanently Defueled Condition of Unit ML20205M0891999-04-0707 April 1999 Proposed Tech Specs Modifying Value for Monthly Surveillance Testing of Tdafwp ML20204J4101999-03-19019 March 1999 Proposed Tech Specs Relocating Instrumentation TSs 3.3.3.2, 3.3.3.3 & 3.3.3.4 to Mnps,Unit 2 TRM ML20204K0971999-03-19019 March 1999 Proposed Tech Specs Supporting Spent Fuel Pool Rerack to Maintain Full Core Reserve Capability Approaching End of OL ML20204J1581999-03-19019 March 1999 Proposed Tech Specs Section 6, Administrative Controls, Reflecting Certified Fuel Handler License Amend Changes, Approved on 990305 ML20204F9031999-03-17017 March 1999 Proposed Tech Specs,Revising 3.5.2,3.7.1.7 & 3.7.6.1 Re ECCS Valves,Atmospheric Steam Dump Valves & CR Ventilation Sys. Associated Bases Will Be Modified as Necessary to Address Proposed Changes ML20206K1121999-03-0505 March 1999 Proposed Tech Specs Bases Sections 3/4.7.7, CR Emergency Ventilation Sys & 3/4.7.8 CR Envelope Pressurization Sys. Changes Are Editorial in Nature ML20207H9551999-03-0505 March 1999 Proposed Tech Specs Section 6.0 Re Administrative Controls ML20207E0321999-03-0202 March 1999 Proposed Tech Specs 3/4.7.4, SW Sys, Proposing Change by Adding AOT for One SW Pump Using Duration More Line with Significance Associated with Function of Pump ML20207D4821999-02-26026 February 1999 Proposed Tech Specs Re Addl Mods Concerning Compliance Issues Number 4 ML20203E4051999-02-11011 February 1999 Proposed Tech Specs Re DG Surveillance Requirements ML20210D2121999-01-21021 January 1999 Proposed Tech Specs Sections 3/4.5.2 & 3/4.5.3, ECCS Subsystems for Info Only ML20199L2841999-01-20020 January 1999 Proposed Tech Specs & Final SAR Proposed Rev to Ms Line Break Analysis & Revised Radiological Consequences of Various Design Basis Accidents ML20199L4561999-01-18018 January 1999 Proposed Tech Specs Revising TS Table 3.7-6, Air Temp Monitoring. Proposed FSAR Pages Describing Full Core off- Load Condition as Normal Evolution Under Unit 3 Licensing Basis,Included ML20199L3271999-01-18018 January 1999 Proposed Tech Specs 3.6.1.2, Containment Sys - Containment Leakage ML20199L0801999-01-18018 January 1999 Proposed Tech Specs Change to TS 3/4.2.2 Modifies TS to Be IAW NRC Approved W Methodologies for Heat Flux Hot Channel factor-FQ(Z).Changes to TS Section 6.9.1.6 Are Adminstrative in Nature ML20199L0431999-01-18018 January 1999 Proposed Tech Specs Removing TS 3/4.6.4.3, Containment Systems,Hydrogen Purge Sys ML20206P5121999-01-0404 January 1999 Proposed Tech Specs 3.5.2,3.6.2.1,3.7.1.2,3.7.3.1 & 3.7.4.1, Incorporating Changes to ESF Pump Testing ML20198K6361998-12-31031 December 1998 Proposed Tech Specs Section 6.0, Administrative Controls ML20198P9751998-12-28028 December 1998 Proposed Tech Specs Pages Revising Loss of Normal Feedwater (Lonf) Analyses to TS 2.2.1,TS Bases Change to Floor Value for Thermal Margin Low Pressure Reactor Trip & Proposed FSAR Changes ML20196H6301998-12-0404 December 1998 Proposed Tech Specs Re Section 6.0, Administrative Controls ML20197G9831998-12-0404 December 1998 Proposed Tech Specs 4.7.10.e,eliminating Need to Cycle Plant & Components Through SD-startup Cycle by Allowing Next Snubber Surveillance Interval to Be Deferred Until End of RFO6 of 990910,whichever Date Is Earlier ML20195D4041998-11-10010 November 1998 Proposed Tech Specs,Modifying Sections 3.3.1.1 & 3.3.2.1 by Restricting Time That Reactor Protection or ESF Actuation Channel Can Be in Bypass Position to 48 H,From Indefinite Period of Time ML20195D8101998-11-10010 November 1998 Revised marked-up Page of Current TS 3.8.1.1 & Revised Retyped Page Re 980717 Request to Change TS ML20155B0331998-10-22022 October 1998 Proposed Tech Specs Changing TS 3.3.2.1, Instrumentation - ESFAS Instrumentation, 3.4.9.3, RCS - Overpressure Protection Sys & ECCS - ECCS Subsystems - Tavg 300 F ML20154A3701998-09-28028 September 1998 Proposed Tech Specs Sections 3.3.2.1,3.4.6.2,3.4.8,3.6.2.1, 3.6.5.1,3.7.6.1 & 3.9.15,revising Info Re Revised MSLB Analyses & Revised Determinations of Radiological Consequences of MSLB & LOCA ML20154C0491998-09-28028 September 1998 Proposed Tech Specs Revising FSAR Separation Requirement of Six Inches Which Is Applied to Redundant Vital Cables, Internal Wiring of Redundant Vital Circuits & Associated Devices ML20151V5011998-09-0909 September 1998 Proposed Tech Spec Changing TS Definitions 1.24,1.27,1.31, 3.0.2,4.0.5,3.2.3,3.3.2.1,3.4.1.1,3.4.11 & Adding TS 3.0.6 B17385, Proposed Tech Specs 6.9.1.8b,updating List of Documents Describing Analytical Methods Specified1998-08-12012 August 1998 Proposed Tech Specs 6.9.1.8b,updating List of Documents Describing Analytical Methods Specified B17341, Proposed Tech Specs Surveillance 4.4.5.3.a Re SG Tube Insp Interval1998-08-0606 August 1998 Proposed Tech Specs Surveillance 4.4.5.3.a Re SG Tube Insp Interval ML20236Y0831998-08-0404 August 1998 Proposed Tech Specs Changing TS 3.7.1.3, Plant Sys - Condensate Storage Tank & Adding TS 3.7.1.7, Plant Sys - Atmospheric Steam Dump Valves ML20236X2521998-07-30030 July 1998 Proposed Tech Specs Bases 3/4.9.1,3/4.1.1.3,3/4.7.1.6, 3/4.7.7,3/4.5.4 & 3/4.3.3.10,resolving Miscellaneous Condition Repts ML20236W0201998-07-30030 July 1998 Proposed Tech Specs Bases Section 3/4.6.1.1,clarifying Administrative Controls for RHR Isolation Valves When RHR Sys Is in Svc for Core Cooling ML20236T2681998-07-21021 July 1998 Proposed Tech Specs Re Reactor Protection & ESFs Trip Setpoints ML20236T5301998-07-17017 July 1998 Proposed Tech Specs Pages for TS Bases Section 3/4.4.9, Pressure/Temperature Limits ML20236T7331998-07-17017 July 1998 Proposed Tech Specs Modifying DG Testing Requirements ML20249A2811998-06-10010 June 1998 Proposed Tech Specs Re Post Accident Access to Vital Areas (Plar 3-98-6) ML20249A3121998-06-0606 June 1998 Proposed Tech Specs Re SLCRS Bypass Leakage (Plar 3-98-5) ML20249A2681998-06-0505 June 1998 Proposed Tech Specs Re Revised Steam Generator Tube Rupture Analysis (Plar 3-98-4) ML20248M2221998-06-0404 June 1998 Revised Tech Specs Pages,Changing TS Bases Section 3/4.7.1.5 to Reword Section Which Describes Limiting Temperature Case for Containment Analysis ML20247G6841998-05-14014 May 1998 Proposed Tech Specs,Modifying TSs 3.3.1.1 & 3.3.2.1 to Restrict Time Most Reactor Protection or Esfa Channels Can Be in Bypass Position to 48 Hours,From Indefinite Period of Time B17211, Proposed Tech Specs Re Refueling Water Storage Tank Back Leakage1998-05-0707 May 1998 Proposed Tech Specs Re Refueling Water Storage Tank Back Leakage ML20247B9411998-05-0101 May 1998 TS Change Pages for TS Bases Section 3/4.5.4,modifying Wording Associated W/Refueling Water Storage Tank Minimum Boron Concentration ML20217D5941998-04-30030 April 1998 Proposed Tech Specs Re Change to Basis 3/4.6.4 Which Modifies Accuracy Range Associated W/Measured Std Cubic Feet Per Minute & Corrects Listed Component Number ML20217N9441998-04-29029 April 1998 Proposed Tech Specs Replacing Two low-range Pressurizer Pressure transmitters,PT-103 & PT-103-1,which Will Identify That Two low-range Pressurizer Pressure Instrument Channels Are Independent & Redundant Only 1999-09-07
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARB17848, Startup Test Rept Cycle 7. with1999-09-30030 September 1999 Startup Test Rept Cycle 7. with ML20211Q3361999-09-0707 September 1999 Proposed Tech Specs Removing Current Special Exception Which Precludes Applying Eighteen Month Functional Testing Surveillance to SG Hydraulic Snubbers ML20211H6471999-08-25025 August 1999 Proposed Defueled Tech Specs,Revising Sections 5.6.1,5.7.2 & 5.7.3 & Adding Proposed Section 5.6.4 to Reflect ACs Contained in NUREG-1433 ML20210Q5211999-08-0505 August 1999 Proposed Tech Specs Sections 3.8.3.2,4.6.2.1,4.6.2.2, 4.8.1.1,4.9.12 & Bases Section B 3/4.3.2,B 3/4.6.1.2 & B 3/4.8.4,incorporating Editorial Revs ML20210C6091999-07-16016 July 1999 Proposed Tech Specs Relocating Selected TS Related to Refueling Operations & Associated Bases to Plant TRM ML20206U1041999-05-17017 May 1999 Proposed Tech Specs Section 4.4.6.2.2.e,deleting Reference to ASME Code Paragraph IWV-3472(b) Re Frequency of Leakage Rate Testing for Valves Six Inches Nominal Pipe Size & Larger ML20206M8221999-05-10010 May 1999 Restart Assessment Plan Millstone Station ML20206D1761999-04-27027 April 1999 Rev 1 to Millstone Unit 3 ISI Program Manual,Second Ten-Yr Interval ML20205R2411999-04-19019 April 1999 Rev 3 to CP2804L, Unit 2 Rx Coolant & Liquid Waste Pass ML20205R2501999-04-19019 April 1999 Rev 0 to CP2804M, Unit 2 Vent & Containment Air Pass ML20205R2751999-04-19019 April 1999 Proposed Tech Specs,Reflecting Permanently Defueled Condition of Unit ML20205S5611999-04-16016 April 1999 Rev 5 to Epop 4426, On-Site Emergency Radiological Surveys ML20205M0891999-04-0707 April 1999 Proposed Tech Specs Modifying Value for Monthly Surveillance Testing of Tdafwp ML20205E4411999-03-29029 March 1999 Rev 2 to CP 2804L, Unit 2 Rx Coolant & Liquid Waste Pass ML20196K5771999-03-24024 March 1999 Rev 1 to Chemistry Procedure CP2804L, Unit 2 Rx Coolant & Liquid Waste Pass ML20205D5321999-03-22022 March 1999 Rev 3 to RPM 2.3.5, Insp & Inventory of Respiratory Protection Equipment ML20204J1581999-03-19019 March 1999 Proposed Tech Specs Section 6, Administrative Controls, Reflecting Certified Fuel Handler License Amend Changes, Approved on 990305 ML20204J4101999-03-19019 March 1999 Proposed Tech Specs Relocating Instrumentation TSs 3.3.3.2, 3.3.3.3 & 3.3.3.4 to Mnps,Unit 2 TRM ML20204K0971999-03-19019 March 1999 Proposed Tech Specs Supporting Spent Fuel Pool Rerack to Maintain Full Core Reserve Capability Approaching End of OL ML20204F9031999-03-17017 March 1999 Proposed Tech Specs,Revising 3.5.2,3.7.1.7 & 3.7.6.1 Re ECCS Valves,Atmospheric Steam Dump Valves & CR Ventilation Sys. Associated Bases Will Be Modified as Necessary to Address Proposed Changes ML20207H9551999-03-0505 March 1999 Proposed Tech Specs Section 6.0 Re Administrative Controls ML20206K1121999-03-0505 March 1999 Proposed Tech Specs Bases Sections 3/4.7.7, CR Emergency Ventilation Sys & 3/4.7.8 CR Envelope Pressurization Sys. Changes Are Editorial in Nature ML20207F6211999-03-0303 March 1999 Rev 2,change 1 to Communications - Radiopaging & Callback Monthly Operability Test ML20207E0321999-03-0202 March 1999 Proposed Tech Specs 3/4.7.4, SW Sys, Proposing Change by Adding AOT for One SW Pump Using Duration More Line with Significance Associated with Function of Pump ML20207D4821999-02-26026 February 1999 Proposed Tech Specs Re Addl Mods Concerning Compliance Issues Number 4 ML20207J0001999-02-22022 February 1999 Rev 7 to Millstone Unit 2,IST Program for Pumps & Valves ML20206D1991999-02-11011 February 1999 Change 7 to Rev 5 to ISI-3.0, Inservice Testing Program. Pages 2 of 3 & 3 of 3 in Valve Relief Request Section 6.1 of Incoming Submittal Not Included ML20203E4051999-02-11011 February 1999 Proposed Tech Specs Re DG Surveillance Requirements ML20210D2121999-01-21021 January 1999 Proposed Tech Specs Sections 3/4.5.2 & 3/4.5.3, ECCS Subsystems for Info Only ML20199L2841999-01-20020 January 1999 Proposed Tech Specs & Final SAR Proposed Rev to Ms Line Break Analysis & Revised Radiological Consequences of Various Design Basis Accidents ML20199L0801999-01-18018 January 1999 Proposed Tech Specs Change to TS 3/4.2.2 Modifies TS to Be IAW NRC Approved W Methodologies for Heat Flux Hot Channel factor-FQ(Z).Changes to TS Section 6.9.1.6 Are Adminstrative in Nature ML20199L4561999-01-18018 January 1999 Proposed Tech Specs Revising TS Table 3.7-6, Air Temp Monitoring. Proposed FSAR Pages Describing Full Core off- Load Condition as Normal Evolution Under Unit 3 Licensing Basis,Included ML20199L3271999-01-18018 January 1999 Proposed Tech Specs 3.6.1.2, Containment Sys - Containment Leakage ML20199L0431999-01-18018 January 1999 Proposed Tech Specs Removing TS 3/4.6.4.3, Containment Systems,Hydrogen Purge Sys ML20199E0931999-01-13013 January 1999 Rev 2 to Health Physics Support Procedure RPM 2.3.4, Insp & Maint Process for Respiratory Protection Equipment ML20206P5121999-01-0404 January 1999 Proposed Tech Specs 3.5.2,3.6.2.1,3.7.1.2,3.7.3.1 & 3.7.4.1, Incorporating Changes to ESF Pump Testing B17501, 1998 - 2000 Performance Plan - Work Environ Focus Area Update1998-12-31031 December 1998 1998 - 2000 Performance Plan - Work Environ Focus Area Update ML20198K6361998-12-31031 December 1998 Proposed Tech Specs Section 6.0, Administrative Controls ML20199A7531998-12-31031 December 1998 Restart Backlog Mgt Plan Commitments ML20198P9751998-12-28028 December 1998 Proposed Tech Specs Pages Revising Loss of Normal Feedwater (Lonf) Analyses to TS 2.2.1,TS Bases Change to Floor Value for Thermal Margin Low Pressure Reactor Trip & Proposed FSAR Changes ML20196H6301998-12-0404 December 1998 Proposed Tech Specs Re Section 6.0, Administrative Controls ML20197G9831998-12-0404 December 1998 Proposed Tech Specs 4.7.10.e,eliminating Need to Cycle Plant & Components Through SD-startup Cycle by Allowing Next Snubber Surveillance Interval to Be Deferred Until End of RFO6 of 990910,whichever Date Is Earlier ML20196A2181998-11-20020 November 1998 Restart Assessment Plan Millstone Station ML20195D4041998-11-10010 November 1998 Proposed Tech Specs,Modifying Sections 3.3.1.1 & 3.3.2.1 by Restricting Time That Reactor Protection or ESF Actuation Channel Can Be in Bypass Position to 48 H,From Indefinite Period of Time ML20195D8101998-11-10010 November 1998 Revised marked-up Page of Current TS 3.8.1.1 & Revised Retyped Page Re 980717 Request to Change TS ML20195H8681998-11-0404 November 1998 Rev 4 to Millstone Unit 2 Operational Readiness Plan ML20196H5921998-10-29029 October 1998 Rev 0 to TPD-7.088, Millstone 1 Certified Fuel Handler/ Equipment Operator Continuing Training Program ML20196H5861998-10-29029 October 1998 Rev 0 to TPD-7.087, Millstone 1 Certified Fuel Handler Training Program B17548, Rev 0 to TPD-7.089, Millstone 1 Equipment Operator Training Program1998-10-29029 October 1998 Rev 0 to TPD-7.089, Millstone 1 Equipment Operator Training Program ML20155B0331998-10-22022 October 1998 Proposed Tech Specs Changing TS 3.3.2.1, Instrumentation - ESFAS Instrumentation, 3.4.9.3, RCS - Overpressure Protection Sys & ECCS - ECCS Subsystems - Tavg 300 F 1999-09-07
[Table view] |
Text
- . - . - - -- . _ - - . -
l '
I l
Docket No. 50-423 l B14823 Attachment 1 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Reactor Trip System Instrumentation Surveillance Requirement Marked-Up Pages of Technical Specifications May 1994 9405130144 940506 l PDR ADOCK 05000423 i P PDR
-=1-2/22/4 ..
.c TABLE 3.3-1 (Continued) ,-
jf, ou y REACTOR TRIP SYSTEM INSTRUMENTATION so~.
MINIMUM c TOTAL NO. CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE H0 DES - ACTION h FUNCTIONAL UNIT
- 17. Reactor Trip System Interlocks (Continued)
- c. Power Range Neutron Flux, P-8 4 2 3 1 8
- d. Power Range Neutron 4 2 3 I 8 Flux, P-9
- e. Power Range Neutron Flux, P-10 4 2 3 1,2 8 -
- 18. Reactor Trip Breakers 2 1 2 1, 2 10, 13 2 1 2 3*, 4*, 5* 11 ,
- 19. Automatic Trip and Interlock 2 1 2 1, 2 13A 2 2 3*, 4*, 5* 11 Logic 1 8 8 1, 2 1 b
- 20. Three Loop Operation 2 .
Bypass Circuitry (1 switch per (From differ-
[ loop in each ent loop g switches in g train) a bypass) 0 2 39, 4, 5 5 f 21. Shutdown Margin Monitor 2
?
e b
M
Jarch-4&,-199F i
TABLE 3.3-1 (Continued)
TABLE NOTATIONS
- When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.
- Above the P-7 (At Power) Setpoint.
- Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
- Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) 5etpoint. >
0 The Shutdown Margin monitor may be blocked during reactor startup in /
\f
[
accordance with approved procedure.
(1) The applicable MODES and ACTION statements for these channels noted in Table 3.3-3 are more restrictive and, therefore, applicable.
(MLah 3 mm ruuh.y 4Q byp<4u b.<, dens Nk,n%-vcuked way Mp m W d*M ACTION STATEMENTS %v by p.
h ,'t"
ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel 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 STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 2 - With the number of bPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1, and
- c. Either, THERMAL POWER is restricted to less than or equal to 757. of RATED THERMAL POWER for four loop operation or 507. of RATED THERMAL POWER for three loop operation and the Power Range Neutron Flux Trip Setpoint is raduced to less than or equal to 85Y. of RATED THERMAL POWER for four loop operation or 607. of RATED THERMAL POWER for three loop operation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.
MILLSTONE - UNIT 3 3/4 3-5 Amendment No. E7 60.
0034* ,
-- .' a .
TABLE 4.3-1 (Continued) 33 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTE OF
$ TRIP e ANALOG tCTUATING MODES FOR R CHANNEL DEVICE WHICH CHANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCI g FUNCTIONAL UNIT CHECK CALIBRATION TEST JEST LOGIC TEST IS RE0UIRED
[ 18. Reactor Trip Breaker N.A. N.A. N.A. M(7. 11) N.A. 1, 2, 3* ,
4*, 5*
- 19. Automatic Trip and N.A. N.A. N.A. N.A. M(7) 1, 2, 3*,
Interlock Logic .
4*, 5*
- 20. Three Loop Operation N.A. N.A. N . /. . R(20) N.A. 1, 2 Bypass Circuitry )
w 21. Reactor Trip Bypass N.A. N.A. N.A. M( 5) N.A. 1, 2, 3* ,
1 5*
Breake f :1(16) 4*,
- 22. Shutdown Margin Monitor N.A. N.A Q(19) 1.A. N.A. 3,4,5 N
a 2
a
?
i i
=-
April 9, 1987
[.[
3 1
F I[T/
INSTRUMENTATION BASES REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUA SYSTEM INST M ENTATION (Continued)
The Engineered Safety Features Acutation System senses selected plant parameters anc determines whether or not predetermined limits are being exceeded.
If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents, events, and transients. Once the required logic ccmcination is completed, the system sends actuation signals to those Engineered Safety Features, components whose aggregate function oest serves the receirements of the condition. As an example, the following actions may be ir.itiated by the Engineered Safety Features Actuation System to mitigate tne consequences of a steam line break or loss-of-coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) feed-water isolation, (4) startup of the emergency diesel generators. (5) quench spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start, (10) sarvice water pumps start and automatic valves position, and (11) Control noom i,solates.
Rent &w ~Tvip hnkeys,--
% e. >e athy hig, h keg (,p.Tg)
% 5 % p Ewc%n cyeW e 5, 9 Ex chaw e_ o t i wcuv', dx *\ 4 vi e me ch s4 im s . Tk4. Lc.c w p aes 4w o O PE tA B LE. Arc <b, s. o@- As e b>eskers . -A M p ben Lee-co nMsts o f gi \ N p b>cske.vs. cs ss e ea caed u h q s.iv,g e c\osed cw d ccQ c4b\c RTS io@c tycdn -\be4 CLrc vac.ked N f s
hm.
o9 Supp\ying pw ex do ee. conb\ ,vod ci ,1v e. Ccctt0 sys bx skey ht ; %e. kvcun mg. ce w&isk
by g%,sa bxaker , ov mctie breske v cd by p^5 4 baskey)M- ~
U ?E ME citpe utin g 4 ow e t syJ;Aevw ce k %ve,\v./w. ccm cus Yoetbl o e, 4meim s e.w s o e. wo diq\ c. Tawclom 4edluve,
& P T S, &A p ces pW % .
mw d be 6 pef A B LF. S
~~Pnc.se &aie &webms Mced L M oD E 3/
M o PE d. or '2. w% We web isvne Wbe. 69E9 ABLE 4 , cer 5 , A kes c FLT S 4 s's p bcb s. cN CA M >
b >whx s toQ.m w e, BTB,s o v cusacicdecA by pcts t.
ctw ct 4k. cRD syh 4 c q w e. e 4 ved Jk d w.M -
MILLSTONE - UNIT 3 8 3/4 3-2a 'Amenament No. 3
~>
a-m ,- , ,-e, ---r4 --n-'
.}
l Docket No. 50-423
.B14823 Attachment 2 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Reactor Trip System Instrumentation Surveillance Requirement Retyped Pages of Technical Specifications May 1994
og TABLE 3.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATI0ff M MINIMUM
, TOTAL NO. CHANNELS CHANNELS APPLICABLE c FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTIm 5
- 17. Reactor Trip System Interlocks (Continued)
]
- c. Power Range Neutron Flux, P-8 4 2 3 1 8
- d. Power Ranga Neutron 4 2 3 1 8 Flux, P-9
- e. Power Range Neutron Flux, P-10 4 2 3 1,2 8 R.
to 18. Reactor Trip Breakers (2) 2 1 2 1, 2 10, 13 A 2 1 2 3* , 4*, 5* 11
- 19. Automatic Trip and Interlock 2 1 2 1, 2 12?
Logic 2 1 2 3*, 4*, 5* 11
- 20. Three Loop Operation 8 2 1, 2 1
_g Bypass Circuitry (1 switch per (From differ-g loop in each ent loop g train) switches ira g bypass) e g 21. Shutdown Margin Monitor 2 0 2 30, 4, 5 5 4
4
TABLE 3.3-1 (Continuedl TABLE NOTATIONS
- When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.
- Above the P-7 (At Power) Setpoint.
- Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
- Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
O The Shutdown Margin monitor may be blocked during reactor startup in accordance with approved procedure.
(1) The applicable MODES and ACTION statements for these channels noted in -
Table 3.3-3 are more restrictive a..J, therefore, applicable.
(2) Including any reactor trip bypass breakers that are racked in and closed for bypassing a reactor trip breaker.
ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel 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 H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed
. *vided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, i
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1, and
- c. Either, THERMAL POWER is restricted to less than or equal 1 to 75% of RATED THERMAL POWER for four loop operation or i 50% of RATED THERMAL F0WER for three loop operation and .
the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85 % of RATED THERMAL POWER for four loop operation ar 60% of RATED THERMAL POWER for three loop operation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.
MILLSTONE - UNIT 3 3/4 3-5 Amendment No. # , JP 0219
gx O
TABLE 4.3-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5 TRIP i ANALOG ACTUATING MODES FOR
- c. CHANNEL DEVICE WHICH z CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE Q FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
- 18. Reactor Trip Breaker N.A. N.A. N.A. M(7, 11) N.A.
g,2,3*,
- 19. Automatic Trip and N.A. N.A. N.A. N.A. M(7) 1 2 Interlock Logic 44 , $*3*,
- 20. Three Loop Operation N.A. N.A. N.A. R(20) N.A. 1, 2 Bypass Circuitry
- 21. Reactor Trip Bypass N.A. N.A. N.A. M(7 N.A. 1, 2 Breaker R(IE)15) 4*,$*3*,
m 22. Shutdown Margin Monitor N.A. N.A Q(19) N.A. N.A. 3,4,5 i
Y M
!e a
n M.
[1 , .
INSTRUMENTATION BASES REAC. TOR TRIP SYSTEM jnSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION l SYSTEM INSTRUMENTATIDH (Continued) l l
The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermincd limits are being exceeded.
If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) Safety injection pumps start and automatic valves position, (2) Reactor trip, (3) feed-water isolation, (4) startup of the emergency diesel generators, (5) quench spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start, (10) service water pumps start and automatic valves position, and (11) Control Room isolates.
REACTOR TRIP BREAKERS 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 RT3 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 H0DE 3, 4, or 5, these RTS trip functions must be OPERABLE when the RTBs or associated bypass breakers are closed, and the CR0 system is capable of rod withdrawal.
i MILLSTONE - UNIT 3 B 3/4 3-2a Amendment No. 7, i 0221
_ --__- __ _