B13627, Proposed Tech Specs Supporting Cycle 4
| ML20058H328 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/01/1990 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20058H326 | List: |
| References | |
| B13627, NUDOCS 9011150174 | |
| Download: ML20058H328 (90) | |
Text
{{#Wiki_filter:. _ _ _ _ _ _ _ Docket No. 50-423-B13627
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~ Attachment 2-Millstone-Nuclear ~ Power Station,' Unit No;'3 Proposed Technical Specification Changes; '
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November 1990 9011.150174 DR 903193 p ADOCK 05000423 PDC
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U.S. Nuclear Regulatory Commission ., B13627/ Attachment 2/Page 1 November 1, 1990 : l . . . , Millstone Nuclear Power Station, Unit No. 3 y Proposed Technical Soecification Chances'. Cycle 4'
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- Section Title Paae Number j Index Index v, xiiic Definitions Allowed Power Level 4 1-7: ,!
1 Figure 2.1-1 -Reactor Core Safety Limits--Four 2 Loops In 0peration- l
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Figure 2.1-2 Reactor Core Safety Limit's--Three 2-3 Loops in Operation . Table 2.2-1 Reactor Trip System Instrumentation 2-5,;2-6, , Trip Setpoints 2-8,L2-9,- 2-11. , l l-Bases Section 2.1.1 Reactor Core , B 2-l! ( Bases Section 2.2.1 Reactor Trip System Instrumentation. 'B 2-5, i l Setpoint- B.2-7, Section 3.1.1.1 Boration Control--Shutdown Margin-- 3/4 1-1,- j Modes 1 and 2 3/4 1-2 , Section 3.1.2.5 Borated Water Source--Shutdown. -3/4.1-11. Section 3.1.2.6 ~ Borated Water Sources--Operating L3/4'l-12 . Section 3.1.3.4 Rod Drop Time =3/4 1 , Section 3.2.1.1 < Axial Flux Difference--Four Loops 3/4.2-1, ! Operating 3/4,2-2' Section 3.2.1.2 Axial Flux Difference--Three Loops' 3/4 2 r Operating 3/4 2-5 Section 3.2.2.1 Heat Flux Hot Channel Factor--- 3/4 2 7,. I Fg (z)--Four Loops Operating- 3/4.2-8, 3/4 2-9 l l Section 3.2.2.2 Heat Flux Hot Channel . Factor-- 3/4 2-11,. F (z)--Three Loops 0perating '3/4 2-12, n 3/4 2-13 j l L
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i U.S. Nuclear Regulatory Commission B13627/ Attachment 2/Page 2 1 November 1,.1990-q. Section Title. ' Page Number' { Section 3.2.3.1 RCS Flow Rate- and Nuclear Enthalpy- 3/4.2-15 Rise Hot Channel-Factor
-l Section 3.2.3.2 RCS Flow Rate and Nuclear Enthalpy 3/4.2-18; q Rise HotiChannel. Factor. j a
Table 3.2 1 DNB Parameters 3/4 2-24)
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c_t' Table 3.3-2 Reactor Trip System Instrumentation
- 3/4 328
.j Response Times ;
. y Section 3.4.1.2 Reactor Coolant System--HotcStandby, - 3/4.4;2 1 Section 3.4.1.6 Reactor Coolant ~ System--Isolated: 3/4'4-8: h I.oop StartupL
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Section 3.5.1 ~ Accumulators.- 3/4 5-1 7
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Section 3.5.'2 ECCS Subsystems--T Greater Than- '3/4 ~. 515, or Equal.;to 350*F "V9 3/45-6 ( Section 3.5.4 Refueling Water Storage Tank
. 3/4: 5-9 l t
Section 3.6.2.3 Spray Additive:Sr 'em 3/a 6-14 a l Section 3.9.1.1 Boron Concentration- '3/4 0-1" ! q Bases j Sections 3/4.1.1.1 Shutdown Margin - B'3/41-1 ! l and3/4.1.1.2 ! Bases Section 3/4.1.2 Boration Systems B3/4'l-3' b Bases Section 3/4.'1'.3 Movable Control Assemblies B 3/4il-4' , Bases Section 3/4.-2 Power Distribution Limits - B 3/4 2-1 Bases Section 3/4.2.1 Axial Flux Difference . B 3/4 2-2, B 3/4 2-3, l
. B 3/4/2-4; (
i (- Bases Sections 3/4.2.2- Heat Flux Hot Channel Factor and B 3/4 2-5, 'l E and 3/4.2.3 RCS' Flow Rate and Nuclear. Enthalpy B3/4.2-6 L Rise Hot Channel Factor-1 Bases Section 3/4.2.5 DNB Parameters B 3/4.2-7 l l 4 l1,
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i U.S. Nuclear Regulatory Commission u B13627/ Attachment 2/Page 3 November 1,~1990 Section Title h gg Number Bases Section 3/4.4.1 Reactor Coolant Loops and Coolant B 3/4 4-1 Circulation BasesSection3/4.9.3 Boron Concentration B3/49-1 Section 6.9.1.6 Core Operating Limits Report 6 21, 6" 2'la Definitions 1.26 Radiological Effluent Monitoring 1-5 and Offsite Dose Calculation Manual (REMODCM) Figure 3.4-1 Dose Equivalent 1-131 Reactor 3/44-30 ! Coolant Specific Activity Limit i Versus Percent of Rated Thermal i Power With Reactor Coolant S ccific Activity > 1pCi/g Dose Equivalent . 1-131 { Fig'.tre 3.4 2 Reactor Coolant System Heatup 3/4 4-34 i limitations--Applicable Up to 10 EFPY Figure 3.4-3 R. actor Coolant System Cooldown 3/4 4-35 Limitations Applicable Up.to 10 EFPY Figure 3.4 4a Nominal Maximum Allowable PORV 3/4 4 40 i Setpoint for Cold Overpressure l System-(Four-LoopOperation) l Figure 3.4-4b Nominal Maximum Allowable PORY 3/4 4-41 '1 Setpoint for Cold Overpressure System (Three-LoopOperation) Table 3.3-4 Engineered Safety Features Actua- 3/43-26 tion System Instrumentation Trip , Setpoints 1 I 1 y s sc ,,
1 l i I I IEQEX ) i i DEFINITIONS ! F i SECTION EAGI i 1.32 SLAVE RELAY TEST............................................. 1-6 , 1.33 SOURCE CHECK................................................. 1-6 l 1 1.34 STAGGERED TEST BAS 1S......................................... 1-6 1.35 THERMAL P0WER................................................ 1-6 1.36 TRIP ACTVATING DEVICE OPERATIONAL TEST....................... 1-6 1.37 UNIDENTIFIED LEAKAGE......................................... 1-6 ' l l.38 UNRESTRICTED AREA..................................... ...... 16 1.39 VENTING....................................................., 1-7 1.40 SPENT FUEL P0OL STORAGE DATTERNS............................. 17 a 1.41 SPENT FUEL P0OL STORAGE PniTERNS............................. 17 . I 1.42 CORE OPERATING LIMITS REP 0RT................................. 1-7 l ND 3,7 1.43 ALLOWED POWER EVEL--APL ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
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- ALLOWED POWER LEVEL--APL ................................... 1-7
'A0 1.1 FREQUENCY N0TATION...................................... 1-8 s 16.LE 1.2 OPERATIONAL M0 DES....................................... 19 r
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l l MILLSTONE - UNIT 3 ii . Amendment No. 7#, J# L . 0012 , a
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INDEX LiliLT. LNG CONDIT10NS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
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11E SECT 10ti Three Loops Operating.................................... 3/4/23 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F q (Z)...................... 3/4 2-5 Four Loops 0perating..................................... 3/4 2-5 Three Loops 0perating.................................... 3/4 2-12 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHAT!NEL FACT 0R................................................... 3/4 2-19 Four Loops 0perating..................................... 3/4 2-19 Three Loops 0perating.................................... 3/42-22 j 3/4.2.4 QUADRANT POWER TILT RAT 10................................ 3/4 2 24 l 3/4.2,5 DNB PARAMETERS........................................... 3/42-27 TABLE 3.2-1 DNB PARAMETERS........................................ 3/4 2-28 i
.I 3/4.3 INSTRUMENTATION !
3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3 1 ! TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION. . . . . . . . . . . . . . . . . . . 3/4 3-2 TABLE 3.3 2 REACTOR TRIP. SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3 i TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE 'l REQUIREMENTS............................................. 3/4 3-10 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-15 ! TABLE 3.3 3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i INSTRUMENTATION.......................................... 3/4 3-17 .i TABLE 3.3 4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM ! INSTRUMENTATION TRIP SETP0lNTS. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/43-26 j g O i i l i l MILLSTONE - UNIT 3 v -Amendment No. 50-0013 1 _a
i INDEX BASES r !ECT10N E661 3/4.0 APPLICABillTY............................................... B3/401 3/4.1 REACTIVITY-CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1 3 3/4.2 POWER DISTRIBUTION LIMITS................................... B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B3/42-1 { 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......... B 3/4 2-3 l 3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/4 2-5 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2 5 , 1 3/4.3 INSTRUMENTATION
'4 /4 . 3 .1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and i ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i INSTRUMENTATION........................................... B3/43-1 l 3/4.3.3 MONITORING INSTRUMENTATION................................ B3/43 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. B 3/4 3 7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION............. B 3/4 4 1 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-2 3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 .j 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-2 !
i 3/4.4.5 STEAM GENERATORS.............................-............ B3/44-3 l 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................ B 3/4 4-4' ! 3/4.4.7 CHEMISTRY................................................. B 3/4 4.5 3/4.4.8 SPECIFIC ACTIVITY......................................... B 3/4 4-5' , 3/4.4.9 PRESSURE / TEMPERATURE LIMITS...............................- B3/44-7: l l i MILLSTONE'- UNIT 3 xiii h 0014 . .I 1
l DEFINITIONS VENTING 1.39 VENTING shall be the controlled process of dischargwg air or gas from a confinen.cnt to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not i provided or required during VENTING. Vent, used in system names, does not' imply a VENTING process. SPENT FUEL POOL STORAGE PATTfE!G: , t 1.40 Region I spent fuel racks contain a cell blocking device in every 4th I locatien for criticality ;ontrol. This 4th location will be referred to as the blocked location. A STORAGE PATTERN refers to 'the blocked location and i all adjacent and diagonal Region I cell locations surrounding the -blocked ' location. Boundary configuration' between Region I and Region Il must have ' , cell blockers positioned in the outermost row of the Region I perimeter, as ' shown in Figure 3.9 2. 1,41 Region 11 contains no cell blockers. LORE OPERATING LIMITS REPORT (COLR) 1.42 The CORE OPERATING LIMITS REPORT that provides core operating limits for(COLR) is the operating the current unit-specific document reload cycle. These cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6. Unit Operation within these ; operating limits is addressed in individual specifications. l ALLOWED POWER LEVEL 1,43 APL ND is the minimum allowable nuclear design power level for base load ; operation and is specified in the COLR. 1.44 APL8 ' is the maximum allowable power level when transitioning into base load operation. -i i l i I
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i MILLSTONE - UNIT 3 1 Amendment No. J), J9- :l ocor 1
I l 6B0 g 1 _ .5 i i UNACCEPTABLE E 660 2425 PSIA 4 l ^ 2250 PSIA i E L8-640 - 4 d i E ~ i
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2000 PSIA i ! N 5 1860 PSIA - n " i ; 600 - [ l l 1 ACCEPTABLE E
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OPERATION 4 i - 580 I 1 560 ' ' ' ' ' ' ' ' ' ' ' 5 0 0.2 0.4 0.6 0.8 - 1.0 1.2 I FRACTION OF RATED THERMAL POWER { i FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION I WILLSTONE - UNIT 3 2-2
- . - . . - . - _ . . -.._. - - . - . - . - _ . . _- _.. .~. .- . .
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t 4 . i i t 680 - E , i 5
- 2 e UNACCEPTABLE 5 5 660 - OPERATION .E 2425 PSIA
- t 4 ;
2250 PSIA 5 !
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- ' 640 -
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w E O . E
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2000 PSIA !
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E i < ~ 620
- 1860 PSIA i m E 1 o -
a: _ 5 i 600 " i i E i ACCEPTABLE 5 OPERATION 7 : r
- e 580 " _ !
5 ? E r 560
' ' ' ' ' ' ' i J l t 0 0.2 0.4 0.6 0.8 . 1.0 l 1 l FRACTION OF RATED THERMAL POWER !
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I l FIGURE 2.1-2 i REACTOR CORE SAFETY LIMIT - TIREE LOOPS IN OPERAT]DN 3 i
.WILLST0E - UNIT 3 2-3 j
4 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS o 5 TOTAL SENSOR U ALLOWANCE ERROR (S) TRIP SETPOINT ALLOWABLE VALUE E FUNCTIONAL UNIT LTA) I Manual Reactor Trip N.A. N.A. N.A. M.A. M.A.
.. 1.
C
- Power Range, Neutron Flux
-2.
- a. High Setpoint
- 1) Four Loops Operating 7.5 4.56 0 $ 109% of RTP** $.111.1% of RTP**
- 2) Three Loops Operating 7.5 4.56 0 $ 80% of RTP** $ 82.1% of RTP**
- b. Low Setpoint 8.3 4.56 0 s 25% of RTP** 127.1% of RTP**
- 3. Power Range, Neutron Flux, 1.6 0.5 0 $ 5% of RTP** with 5 6.3% of RTP** with High Positive Rate ~ a time constant a time constant 1 2 seconds. 1 2 seconds
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- 4. Power Range, Neutron Flux, 1.6 0.5 0 5 5% of RTP** with 5 6.3% of RTP** with High Negative Rate a time constant a time constant.
1 2 seconds 1 2 seconds
- 5. Intermediate Range, 17.0' 8.41 0 1 25% of RTP** s 30.9% of RTP** .
Neutron Flux
- 6. Source Range, Neutron Flux 17.0 10.01 0 $ 10+5 cps 1 1.4 x 10+5 . cps
- 7. _0vertemperature AT Four' Loops Operating a.
1.71 + 1.33 - See Note 1 See Note 2
.['
's-
,+.
- 1) Channels I, II 10.0 6.80 (Temp + Press) z 2) Channels III, IV 10.0 5.83 1.71 + 2.60 See Note-1 See Note 2.
(Temp + Press)
.m
**RTP = RATED THERMAL POWER q
TABLE 2.2-1 (Continued)
. .g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS F
~G JZ TOTAL SENSOR ALLOWANCE ERROR
, FUNCTIONAL UNIT (TA) I (S) TRIP SETPOINT ALLOWABLE VALUE C
z b. Three Loops Operating
" 10.0 6.80 1.71 + 1.33 See Note 1
- 1) Channels I, II See Note 2 (Temp + Press).
- 2) Channels III, IV 10.0 5.83 1.71 + 2.60 See Note 1 See Note 2 !
(Temp + Press)
- 8. Overpower AT 4.8 1.24 1.71 See Note 3- See Note 4 i' 9. Pressurizer Pressure-Low 5.0 1.77 3.3 2 1900 psia > 1890 psia
- 10. Pressurizer Pressure-High 5.0 1.77 3.3 5 2385 psi 2 . s 2395 psia .
E' 11. Pressurizer Water Level-High 8.0 5.13 2.7 s 89% of instrument. 5 90.7% of instrument - ' span span
- 12. Reactor t.colant Flow-Low - 2.5 1.52 ' 0.78 2 90% of loop 189.1% of loop .
design flow
- design flow *
- 13. Steam Generator Water'-
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18.10 1.50 ^ 16.64 1 18.10% of narrow - > 17.11% of narrow
- - [ Level' Low-Low ' range instrument range instrument ~
g span span , R. - 3 14. General Warning Alarm N. A. - N.A. N.A. N.A. M.A. z o
- 15. Low Shaft Speed - Reactor- ' 3.8 0.5 0.. 2 95.8% of rated _ 1 92.5% of rated Coolant Pumps .. speed- speed M. _
2
-g :
Q .* Minimum Measured Flow Per Loop - %,870 gpe. (Four Loops Operating); 101,066 gym (Three Loops Operating)
, ._ q* .+.,+y - - . - - - - , , , - . . - , - . . f. - -m a s -- # e -
.--..y, , tr-n. c.. . ..%- sy e.g ,m .-,.wm. -., _ -- . _ _ , _ ..,_,-c_,,+,
i l TABLE 2.2-1 (Continued) 2 REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS 17
~G l g TOTAL SENSOR R ALLOWANCE ERROR
, FUNCTIONAL UNIT (TA) I (S) TRIP SETPOINT ALLOWABLE VALUE C
5-
- 16. Turbine Trip
" a. Low Fluid Oil Pressure E.A. N.A. N.A. 1 500 psig 1 450 psig
- b. Turbine Stop Valve N.A. N.A. N.A. 1 1% open 1 1% open Closure
- 17. Safety Injection Input N.A. M.A. M.A. N.A. N.A.
from ESF
- 18. Reactor Trip System Interlocks 2
~2
.. m a. Intermediate Range N.A. N.A. N.A. 1 1 x'10 ze amp 1 6 x 10 amp ';
a Neutron Flux, P-6
- b. Low Power Reactor Trips-Block, P-7
- 1) P-10 input N.A. N.A. M.A. 1 10% of.RTP**. 1 12.1% of RTP**
- 2) P-13 input N.A. N.A. N.A. 1 10% RTP** Turbine i 12.1% RTP** Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent
- c. Power Range' Neutron Flux,;P-8
- 1) Four Loops Operating N. A. - N.A. N.A. 1 37.5% of RTP** 139.6% of RTP** -
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- 2) Three Loops Operating. M.A. .N. A.. -N.A. 1 37.5% of RTP** - 139.6% of RTP**
**RTP - RATED THERMAL POWER M
__u _ _ . _ _ _ _ _ _ _ - __
^m -
v r- ew- -' ' "+ r - tv -"
'"fF - "'" v 1'---~~**' -' " ' ' ' " *
- v-*'~ "'~+' =w-' *'* *--<' - + ' + - ' = " ^ * *-w =
TABLE 2.2-1 (Continued) t 2 REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS i F
G 1- TOTAL SENSOR
$ ALLOWANCE ERROR
, FUNCTIONAL UNIT (TA) I (S) TRIP SETPOINT ALLOWABLE VALUE
- d. Power Range Neutron N.A. M.A. M.A. 151% of RTP** 1 53.1% of RTP** !
9 Flux, P-9 w
- e. Power Range Neutron N.A. N.A. N.A 1 10% of RTP** > 7.9% of RTP**
Flux, P-10
- 19. Reactor Trip Breakers N.A. N.A. N.A N.A. M.A.
- 20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. N.A.
Logic
- 21. Three Loop Operation N.A. N.A. M.A. N.A. N.A m -Bypass Circuitry do N
.a R.
a g. D **RTP = RATED THERNAL' POWER
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*p TABLE 2.2-1 (Continued)
- r-c: g z
TABLE NOTATIONS 7 NOTE 1: OVERTEMPERATURE AT I + 3 I
$ AT (1 + r jS) 1 O IKI~K2 4 ) [T 1 + 7 Sy - T'] + K3 (P - P') - 3f (AI)}
] (1 + 27 5) (I +3 # S) <AT II + # S) 5 6 Where: AT - Measured AT by Reactor Coolant System Instrumentation; I+7S3 - Lead-lag compensator on measured AT; I+732 7,
3 7 2 - Time constants utilized in lead-lag compensator for AT, r3-8s, 7 2 - 3 s; I - Lag compensator on measured AT; y,73 E 7 - Time constants utilized in the lag compensator for AT, 73 - O s; 3 AT - Indicated AT at RATED THERMALjPOWER;- O
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Kg - 1.20-(Four Loops Operating); 1.20 (Three Loops Operating); K. - 0.02456; 2 I+7S4 - The function generated by the lead-lag compensator for T dynamic 1+735 -compensation;
'[ 7,'7 - Time. constants utilized in the lead-lag compensator for T,,9, 74 - 20 s, 4 5 g -1 5 = 4 s; 3
$ T - Average temperature, *F;.
~E I ;
1+r36 - Lag compensator on measured T, 5. 7 6
- Time constant utilized in the measured T,,, lag compensator, 767 0 s; -
m - - . . .
.-m. . - - _ . -.m.- - . _ _ . - . . _ . - - - - . _ , _ . . -
__---.-.--~-~~----.-_---.m.
[h
- r-TABLE 2.2-1 (Continued) ,
jz TABLE NOTATIONS (Continued) 7 NOTE 1: (Continued) E T' s 587.l*F (Nominal T,yg at RATED THERMAL POWER); w K - 0.001311/ psi; 3 P - Pressurizer pressure, 9sia; P' - 2250 psia (Nominal RCS operating pressure);
'S - Laplace transform operator, s 1; and fj (AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response m . during plant startup tests such that:
(1) For q - gb between -26% and + 3%, tf (AI) = 0, where q are percent RfTED THERMAL POWER 9b in th$ top and bottom halves of the core respectively,t an *d qt b+ is9 total THERIEL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of q -q exceeds.-26%,-the AT Trip Setpoint shall be automatically reduced by 3.55% of its _vafue aI RATED THERMAL POWER; and (3) For each percent that the magnitude of q - qs exceeds +3%, the AT Trip Setpoint'shall be _ automatically; reduced by 1.98% of its vakue at RATED THERMAL POWER. The channel's-maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.7%
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NOTE 2: ' AT span (Four. Loop Operation); 2.7% AT span (Three Loop Operation).
- - - - - - - - - = - - - - .
- - - - - - -- - - - - ~"--c-^-- - - - - - " = - - - - ~ - - - - - - - + - - - -
. _ _ _ . . _ . _ _ _ _ _m.__
;ax
* *: TABLE 2.2-1 (Continued)
"G
$ TABLE NOTATIONS (Continued) i'i b
z NOTE 3: OVERPOWER A8T
) < AT, (K, - 3K I#3 ) I I I T - K [T I I ) - T"] - f, (AI)}
5 "' AT (I + #1 3) ( I 7 (1 + 72 5) (1 + T sS) (1 + r,5) (1 + r,5) (1 + r,5) Where: AT - As defined in Note 1, 1+rSi - As defined in Note 1, I + r,S 7 1, 7 2 - As defined in Note I, I - As defined in Note 1, 1+r53 73 - As defined in Note 1, AT, - As _ defined in Note I, K, '- 1.09, K5 - 0.02/*F for increasing average temperature.and O for_ decreasing average temperature, 7S 7 7 l + r,S The function generated by the rate-lag compensator for T,,9 dynamic' { compensation, Po g 77 ~= Time constants utilized in the. rate-lag compensator for T,,9, 77 _10 s,
- As denned in M e 1,
~ l + r,S
'D r, - As defined in' Note I,
-----~----m--- - - - - - - - - - ' - - - + ---p----- - - - - - - - - - - - - + - - - - - - -- + - - - - ~ ~ - - - - ~ ~ - - - -
_._m______rm_----
oZ E T* TABLE 2.2-1 (Continued)
** G g TABLE NOTATIONS (Continued)
. i'i -
h NOTE 3: . (Continued) i'i , w K 6 0.00180/*F for T > T" and K6 - O for T 1 T", ' T - As defined in Note 1, , T" - Indicated T a instrumentafI8n,t 1RATED THERMAL POWER (Calibration temperature for AT 587.18F), S - As defined in Note 1, and f 2(AI) = 0 for all AI. g NOTE'4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.7% AT span. i r k
,< A. ,
ev,s ..m, n 1.,~~, - - , . - - - , ,r , , - + . ,--~n,v
---~ev e- - - -- , - , , - vA -e-- n, ,,n.,-
l 2.1 SAFETY LIMITS l i BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and ! possible cladding perforation which would result in- the release of fission , products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel . operation to within the nucleate boiling regime where the heat transfer coefficient is large - and the cladding surface j temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could i result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.. DNB is not a directly measurable parameter during operation and ~ l therefore THERMAL POWER and reactor coolant temperature and pressure have been ' related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform- and nonuniform heat flux > distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio 1 of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB. l The DNB design basis is as follows: uncertainties in'the WRB-1 or WRB-2 correlations, plant operating parameters, nuclear and thermal parameters, fuel 's fabrication parameters, and computer codes are considered statistically such that there is at least a 95 percent probability with 95 percent confidence. { 1evel that DNBR vill not occur on the most limiting fuel rod during , Condition I and 11 tents. This establishes a design DNBR value which must be i" met in plant safby analyses using values of input parameters without uncertainties. In addition, margin has been maintained in the design by I meeting safety analysis DNBR limits in performing safety analyses. i The curves of Figures 2.1-1 and 2.1-2 show the. loci of points of THERMAL POWER, Reactor Coolant System pressure, and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid. ; These curves are based on an enthalpy hot channel factor, F H, of 1.70 (includes measurement uncertainty) and a reference. cosine with a ak ofol.55 ; for axial power shape. ' reduced power based on the Anexpression. allowance is included 'for an increase in F{H at ! N F H = 1.70 [1 + 0.3 (1-P)] where P is the fraction of RATED THERMAL POWER t These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control a rod insertion assuming axial imbalance is within the limits of F deltaI) function of the Overtemperature trip. When the axial power imballnc(e is not l within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with < core safety limits. M,,1L,L, STONE - UNIT 3 B 2 l .
_ , _ ~ __ _ . . _ __ _ . 4 l l LIMITING SAFETY SYSTEM SETTINGS BASES l Intermediate and Source Ranae. Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-l trolled rod cluster control assembly bank withdrawal from a subcritical j condition. These trips provide redundant protection to the Low Setpoint trip l of the Power Range, Neutron. Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked - when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of ~. RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the- Intermediate or Source Range Channels in the accident analyses; however, their functional ca) ability at the specified trip settings is required by this specification to , en1ance the overall reliability of the Reactor Trip System. Overtemoerature AT The Overtemperature AT trip provides core protection to prevent DNB for ! l all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient ' is slow with respect to piping ; transit delays from the core to. the temperature detectors, and pressure is i within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water.and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial lower distribution. With , normal axial power distribution, this Reactor trip ' limit is always below the ! core Safety Limit as shown in Figure 2.1-1. If axial peaks are- greater than design, as indicated by the difference between top and bottom power range i nuclear detectors, the Reactor trip is automatically reduced according to the j notations in Table 2.2-1. Operation with a reactor coolant loop out of service requires Reactor l Trip System modification. Three loop operation is ; the K1 input to the Overtemperature AT channels, reducing' permissible after resetting the Power Range- ' Neutron Flux High setpoint to a value just above the three loo) maximum i permissible power level, and resetting the P-8 setpoint to its' three loop ; value. These modifications have been chosen so that, in three loop operation ~, ' each component of the Reactor Trip System performs. its normal four loop .l function,. prevents operation outside the safety limit curves, and prevents the i DNBR from going below the design limit during normal. operational and antici- I pated transients. Overpower AT - The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel
. pellet melting ~and less than 1% cladding ' strain) under all possible overpower, conditions, ' limits the required range _ for Ovartemperature AT:
. MILLSTONE 0005 UNIT 3 B 2-5. Amendment No. U l
]j
1 j l 1 , LIMITING SAFETY SYSTEM SETTINGS ! BASES { t Steam Generator Water Level The Steam Generator Water level Low low trip protects the reactor from ; loss of heat sink in the event of. a sustained steam /feedwater flow mismatch : resulting from loss of normal feedwater. The specified .Setpoint provides allowances for starting delays of the Auxiliary Feedwater System.. ; low Shaft Speed - Reactor Coolant Pumos The Low Shaft Speed - Reactor Coolant Pumps trip provides ' core protection , to prevent DNB in the event of a sudden significant decrease in . reactor , coolant pump speed (with resulting decrease in flow) on two reactor. coolant pumps in any two operating reactor coolant loops. .The trip setpoint ensures that a reactor trip will be generated, considering instrument errors and response times, in sufficient time to allow the DNBR to be maintained greater ' than the design limit following a four-pump loss of flow event. Turbine Trio ' j A Turbine trip initiates a Reacter trip. On decreasing power the Reactor trip from the Turbine trip is automatically blocked by P-9 (a power level of j approximately 50% of RATED THERMAL POWER); and on increasing power, reinstated l automatically by P-9. Safety Iniection Inout from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic . channels will initiate a Reactor trip upon any signal which initiates .a Safety Injection. The ESF instrumentation channels which initiate a. Safety Injection signal are shown in Table 3.3-3. Reactor Trio System Interlocks The Reactor Trip System interlocks perform the following functions: l P-6 On increasing power P-6 allows the manual block of the Source Range trip (i.e., prevents premature block of Source. Range trip) and deenergizes the high voltage to the detectors. On decreasing >ower, ! Source Range Level trips are automatically reactivated and higi vol-tage restored. . P-? On increasing power P-7 automatically enables: Reactor l trips on low- ; flew in more than one reactor coolant loop, reactor coolant pump low ; shaft speea, pressurizer low pressure and pressurizer high leve1~. On ' decreasing power, the above listed trips are automatically blocked. l l MILLSTONE - UNITE 3 B 2-7 . 0006 )
E 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 1 AND 2 LIMITING CONDITION FOR OPERATION 3.1.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/k for both four loop and three loop operation. APPLICABILITY: MODES 1 and 2*. , ACTION: With the SHUTDOWN MARGIN less than 1.3% Ak/k, immediately initiate and con- ! tinue boration at greater than or equal to 33 gpm of a solution containing i greater than or equal to 6300 ppm boron or equivalent until the required ! SHUTDOWN MARGIN is restored. SURVEILLANCE RE0VIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% Ak/k: , 4
- a. Within I hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is ino)erable.
If the inoperable control rod is immovable' or untrippable, tie above ' required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable' or untrippable control rod (s); 1
- b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at ;
least once per 12 hours by verifying that control bank withdrawal is ; within the limits of Specification 3.1.3.6; j
- c. When in MODE 2 with K,ff less - than .1, within 4- hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification-3.1.3.6;
- d. Prior to . initial operation aboye 5% RATED THERMAL POWER. after each !
fuel loading, by consideration of 'the factors of Specification ! 4.1.1.1.2, with the control banks at the maximum insertion limit of i Specification 3.1.3.6; and j
*See Special Test Exceptions Specification 3.10.1. !
MILLSTONE - UNIT 3 3/4 1-1 ! 0007 __ d
}
t REACTIVITY CONTROL SYSTEMS SURVEllLANCE REQUIREMENTS (Continued) , 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within 1% Ak/k at leastLonce per 31 Effec-4 tive Full Power Days (EFPD). This comparison shall consider at least the ' i following factors: I
- 1) Reactor Coolant System boron concentration,
, 2) Control rod position,
- 3) Reactor Coolant System average temperature, I
- 4) Fuel burnup based on gross thermal energy generation,
- 5) Xenon concentration, and
- 6) Samarium concentration.
The predicted reactivity values shall be adjusted (normalized) to correspond , to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD ' after each fuel loading. c l b
' MILLSTONE - UNIT 3 3/4 1-2 0007
REACTIVITY CONTROL SYSIDi$ BORATED WATER SOURCE - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:
- a. A Boric Acid Storage System with:
- 1) A minimum contained borated water volume of 6700 gallons,
- 2) A boron concentration between 6300 and 7175 ppm, and
- 3) A minimum solution temperature of 67'F.
- b. The refueling water storage tank (RWST) with:
- 1) A minimum contained borated water volume of 250,000 gallons,
- 2) A minimum boron concentration of 2700 ppm, and
- 3) A minimum solution temperature of 40'F. ,
APPLICABILITY: MODES 5 and 6. ACTION: i With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SVRVEILLANCE RE0VIREMENTS 4.1.2.5 The above required borated water _ source shall be demonstrated OPERABLE:
-j
- a. At least once per 7 days by:
- 1) Verifying the boron concentration of the water,
- 2) Verifying the contained borated water volume, and
- 3) Verifying the Boric Acid Transfer Pump Room -temperature and the boric acid storage tank solution temperature when it is the source of borated water.
- b. At least once per 24 hours by verifying the RWST temperature when it is the source of borated water and the:outside air. temperature is~1ess than 35'F.
MILLSTONE - UNIT 3 3/4 1-11 -Amendment'No. JJ 0009-
REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2: '
- a. A Boric Acid Storage System with:
- 1) A minimum contained borated water volume of 23,620 gallons,
- 2) A boron concentration between 6300 and 7175 ppm, and
- 3) A minimum solution temperature of 67'F.
- b. The refueling water storage tank (RWST) with:
- 1) A minimum contained ' borated water volume of- 1,166,000 gallons, '
i
- 2) A boron concentration between 2700 and 2900 ppm, l
- 3) A minimum solution temperature of 40'F, and l A maximum solution temperature of 50'F.
4) j APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With the Boric Acid Storage System inoperable, restore the system j to OPERABLE states within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHVTDOWN MARGIN I equivalent to at least 1.3% Ak/k at 200'F; restore the Boric Acid <
Storage System to OPERABLE status within the next 7 days or be in - COLD SHUTDOWN within the next 30 hours. ;
- b. With the RWST inoperable, restore the tank to OPERABLE status within I hour or be in at least HOT. STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours -
i MILLSTONE - UNIT 3 3/4 1-12 0009 Amendment No. U
1 1 REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION . 3.1.3.4 The individual full length shutdown and control rod drop time from i the fully withdrawn position shall b(e less than or equal)to 2.7 seconds from i beginning of decay of stationary gripper coil voltage to dashpot entry with:
- a. T avg greater than or equal to 551'F, and j
- b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2. r ACTION:
- a. With the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the- above limit ,
prior to proceeding to MODE 1 or 2. , 3
- b. With the rod drop times within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal. to 65% of RATED c THERMAL POWER with the roactor coolant stop valves in the nonoperat-ing loop closed. -
l SVRVEILLANCE RE0VIREMENTS l l 4.1.3.4 The rod drop time of full-length rods shall .be demonstrated through- J ( measurement prior to reactor criticality:
- a. For all rods following each removal of the reactor vessel' head,
- b. For specifically affected individual rods following any maintenance o, on or modification to the Control Rod Drive System which could l affect the drop time of those specific rods, and l
l c. At least once per 18 months. t i' MILLSTONE - UNIT 3 3/4 1-19 i 0010 l- j
. . . -. . . - .J
i l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2.1.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within: a. The limits s)ecified in the CORE OPER*. TING LIMITS REPORT (COLR) for-Relaxed Axial Offset Control (RAOC) operation, or
- b. Within the target band about the target flux difference during base load operation, specified in the COLR.
APPLICABILITY: MODE 1 above 50% RATED THERMAL POWER *. ACTION:
- a. For RAOC operation with the indicated AFD outside of the applicable limits specified in.the COLR,
- 1. Either restore the indicated AFD to within the COLR- specified limits within 15 minutes, or 2.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux--High .j Trip setpoints to less- than or equal to 55% of RATED THERMAL ~ POWER within the next 4 hours. ;
- b. For base load operation above APLND with the indicted AFD outside of the applicable target band'about the target flux differences: i
- 1. Either restore the indicated AFD to within the COLR specified target band within 15 minutes, or {
- 2. Reduce THERMAL POWER to less than APLHD of RATED THERMAL POWER and discontinue base load operation within 30 minutes.
- c. I THERMAL POWER shall not be increased above 50% of RATED THERMAL '
POWER unless the indicated AFD is within the limits specified in the , COLR. ' i
*See Special Test Exception 3.10.2 !
I I MILLSTONE - UNIT 3 3/4 2-1 Amendment No. Ep'
'0011 !
l
- _ - - _ - _ _ _ - - ._2
l 1 r POWER DISTRIBUTION LIMITS {
. SURVEILLANCE REOUIREMENTS j 4.2.1.1.1 The indicated AFD shall be determined to be within its limits ;
during POWER OPERATION above 50% of RATED THERMAL POWER by: l i
- a. Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE:
- b. Monitoring and logging the indicated AFD for each OPERABLE- excore i channel at least once per hour for the first 24 hours and at least ;
once per 30 minutes thereafter, when the AFD Monitor Alarm is- -! inoperable. The logged values of the indicated AFD shall be assumed ! to exist during the interval preceding each logging. 4.2.1.1.2 The indicated AFD shall be considered outside of its limits 'when j two or more OPERABLE excore channels are indicating the AFD to be outside the : limits. j 4.2.1.1.3 When in base load operation, the target flux difference of each j OPERABLE excore channel shall be determined by measurement at least once per ; 92 Effective Full Power Days. The - provisions of Specification 4.0.4 are not l applicable, 4.2.1.1.4 When in base load ration, the target flux . difference shall be updated at leas ; once per 31 Effective Full Power Days by either determining , the target flux difference in conjunction with the surveillance requirements ; of Specification 4.2.1.1.3 or by linear interpolation between the most ; recently measured value and the calculated value at the end of cycle life. : The provisions of Specification 4.0.4 are not applicable. ; I 1 i MILLSTONE - , NIT 3- 3/4 2-2 Amendment No. M 0011 i
., e
POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2.1.2 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:
- a. The limits specified in the CORE OPERATING LIMITS REPORT (C0i.R) for Relaxed Axial Offset Control (RAOC) operation, or
- b. Within the target band specified in the COLR about the target flux difference during base load operation.
APPLICABILITY: MODE I above 37.5% of RATED THERMAL POWER
- ACTION:
i
- a. For RAOC operation with the indicated AFD outside of the applicable limits specified in the COLR, ,
1 4
- 1. Either restore the indicated AFD to within the COLR specified limits within 15 minutes, or ,
- 2. Reduce THERMAL POWER to less than 37.5% of RATED THERMAL POWER I within 30 minutes and reduce the Power Range Neutron Flux--High Trip setpoints to less than or equal to 41% of RATED THERMAL ;
POWER within the next 4 hours,
- b. For base load operation above APLHD with the indicated AFD outside of the applicable target band-about the target flux differences: 4
- 1. Either restore the indicated AFD to within the COLR specified target band within 15 minutes, or
- 2. Reduce THERMAL POWER to less than APLND.of RATED THERMAL POWER and discontinue base load operation within 30 minutes.
l 1
- c. THERMAL POWER shall not be increased above. 37.5% .of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.
I 1
- See Special Test Exception 3.10.2.
I MILLSTONE - UNIT 3 3/4 2-3 0011 Amendment No.'JJ .
-.._______o
l P_QWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS 4.2.1.2.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 37.5% of RATED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE:
- b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD-shall be assumed i to exist during the interval preceding each logging.
4.2.1.2.2 The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside the limits. i 4.2.1.2.3 When in base load operation, the target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per j 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable. ' 1 4.2.1.2.4 When in base load operation, the target flux difference shall be u> dated at least once per 31 Effective Full Power Days by either determining tie target flux difference in conjunction witn the surveillance requirements of Specification 4.2.1.2.3 or by linear interpolation between the most recently measured value and the calcu'ated value at the end of cycle life. The provisions of Specification 4.0.4 are not applicable, i I
\
1 l i 3/4'2-4 glLSTONE-UNIT 3 Amendment No. #
I POWER DISTRI E ION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fp FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION e 3.2.2.1 Fg (Z) shall be limited by the following relationships: l Fg (Z) 1 F RTP K(Z) for P > 0.5 , P : RTP F(Z)sF g g K(Z) for P s 0.5 0.5 a RTP F the F limit at RATED THERMAL POWER (RTP) provided in f theco=reoper$tinglimitsreport(COLR).
, and-Where: P = THERMAL POWER RATED THERMAL POWER ;
K(Z) = the normalized F g (Z) as a function- of core height- . specified in the COLR. ' APPLICABILITY: MODE 1. ACTION: With F (Z) exceeding its limit: n
- a. Reduce THERMAL POWER at least 1% for each-1% Fn(Z) exceeds the l limit within 15 ~ minutes and similarly reduce 'the Pewe',' Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent i POWER OPERATION may proceed provided the Overpower AT Trip Set- ;
points have been reduced at least 1% for each 1% F g(Z) exceeds l the limit, and ,
. s
- b. Identify and correr.t the cruse of the out-of-limit condition-prioi' to increasing THERMAL POWER above the reduced'11mit re-quired by ACTION a., 'aboys;; THERMAL POWER'may then be increased provided Fn(Z) is demonstrated through incore mapping to be within its ' limit.. t MILLSTONE - UNIT 3 3/4 2 5 Amendment No. #
0011. s!
I POWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS
~)
4.7.2.1.1 Tho provisions of Specification 4.0.4 are~ not applicable. 4.2.2.1.2 For RAOC operation,qF (z) shall be evaluated t'o determine if F (z) is within its limit by: 9
- a. Using the movable incore detectors to obtain a power; distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER' .
- b. Increasing the measured Fn(z) component of the power? distribution:
map by 3% to account foT manufacturing tolerance: and further-increasing the value by 5% to account for measurement uncertainties.- Verify the requirements of Specification-3.2.2.1 are ' satisfied.-
- c. Satisfying the following relationship:
F (z) s F0 x K(z) for P > 0.5 P x W(z) M p (z) s 0 F x K(z) for P $ 0.5
.W(z) x 0.F ,
wh.re F$(z)- is the measured. n F (z) increased!by the :agances for i.wfacttring tolerances-and me'hsurement uncertainty, F is1the'F iirit, K(z) is the normalized F q(z) as a' function of, coke height,0 is the relative THERMAL POWER, and W(z) is - the cycle-dependent function that accounts for g er distribution transients encountered during normal operation. Fg . , K(z), and W(z) are specified in-the CORE GERATING LIMITS REPORT as per Specification _6.9.1.6.-
- d. Measuring ~ F (z) according .to the 'following schedule:
(1) Upon achieving' equilibrium conditions after exceeding by 10%'or 1
-more of RATED THERMAL POWER,- the THERMAL POWER at which. F (z) q was last determined,* or 1 *During power: escalation at the beginning of each cycle, power level .may be i increased until e ' power level for extended' operation has' been achieved and pow. * ;istribut4n map outlined, a
l' . q MILLSTONE UNIT 3 3/4 2 6 0011 Amendment No. EE
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) ..
~\
(2) At least once per 31 Effective Full- Power Days, ' whichever d occurs first. ! l
- e. WiththemaxImumvalueof !
l F (z) ' K(z) over the core height (z)-increasing since the: previous __ determination of F (z), either of the-following actionsishall be taken:. 1 1 d (1) F$(z) shaR La increased by~ 2% over that: specified in Specifi- ^! cation 4.2.2.1.2c, or (2) F$(z) shall be -measured at least once per 7 Effective Full: Power Days until' two successive ' maps) indicate that the maximum-value of _d 1 F (z) 1 K(z) ] over the core height Tz) is not increasing.
- f. With the relationships: specified _in -Specification 4.2.2.1.2c not being satisfied: '
q (1) Calculate the maximum percent over the core height -(z) that ! Fq (z) exceeds its limit by the following expression: ! i j Ff(z)xW(z)_ RTP
-1 x 100 for P 1 0.5-0 l
x K(z) t P
. . /j
-i
?
1 , 5
' MILLSTONE-- UNIT 3 3/4 2-7 Amendment No.: 79 ,
0011 ,
'i I
t I POWER DISTRIBUTION LIMIT $ SURVEILLANCE REQUIREMENTS (Continued) e
\,
I ' t r - I F"(z) x W(z)' RTP
-1 x 100 for P < 0.5 .j
._Q__. x K(z) 0.5
. . i (2) One of the following actions shall be taken:L i (a) Within 15 minutes, control the AFD . to within -new: AFD limits which are determined by reducing the applicable.AFD limitsby1%AFD:for-eachpercentFn(z).exceedsits-limits-as. determi.ned ^ in Specification ) 2.2.1 2f.1, Within !
8 hours, reset the. AFD alarm setpoints - to these modified i limits, or r (b) Comply with the requirements.'of _ Specification'3.2.2.1 for Fg (z) exceeding its limit by the percent ' calculated, or. (c) Verify that the ! requirements - of - Specification 4.2.2;1.3 for base load operation are satisfied and enter' base' load ; operation. ,
- g. The limits specified in Specifications 4.2.2.1.2c, '4.2.2.1'.2e, and' 4.2.2.1.2f above are not applicable in .theJ following core plane regions: -
F (1) Lower core region from 0% toL15%, inclusive. (2) Upper. core region.from 85% to 100%, inclusive. . s 4.2.2.1.3 Base load operation '.is permitted .at' powers above ~ APL$ if ~ the-following conditions are satisfied: *
- a. Prig to entering base load operation, maintain THERMAL POWER above APL and less than or equal; to - that allowedL by Specifica-tion 4'.2.2.1.2; for at least the ' previous : 24_-hours.- : Maintain base.
-load operation surveillance - (AFD within; the.. target t band limit; the.- ,
target flux difference ofc Specification 3.2.1'.1). during . this1 time ' period. Base 1oad' operation -.gthen. pgmitted providig THERMAli '
. POWER:is maintained between APL and'APL or between APL and a
[
.5 MILLSTONE - UNIT 3- 3/4 2-8 LAmendment No. M 0011' '
. ~ _ . . - _ _
1 POWER DISTRIBUTION LIMITS I 1 SURVEILLANCE RE0VIREMENTS (CQDtinued) ; 1 4 100%-(whichever is'most-1_imiting) and F sugeillance is maintained pursuant to Specification 4.2;2.1.4. qAPL lis defined as the minimum value of: < l RTP F -xK(z)- APLBL , 0 I x 100% N F (z) r W(z)BL h over the core height '(z)_ whe're: FN(z) is -- the _ measured . F ! increased by the allowances for manuficgping' tolerancesiand mea- 1 0(z surement uncertainty.- The F 11mit is F isfthe cycle-
-dependent - function thataccS3unts for 13mited .c.W(z)B pow H
istribution j transient encountered during> base load -operation. F , L K(z)', g and_ W(z)9L arespecifiedintheCOLRasperSpecificationk.9.1.6.
- b. Durgg base load operation, if the ' THERMAL' POWERiis l decreased below-APL then the conditions of 4.2.2.1.3.a shall be satisfied - before< ;
reentering base load operation. , 4.2.2.1.4 During base load operation Fq(z) shall _be'evaluat'edito determinevifi Fn (z) is within its limit by: a. Using the movable incore detectorg to obtain a power distribution. q map at any THERMAL POWER above APL _. y\
- b. Increasing the measured Fn (z) component of' therpower distr;ibuticn map by 3% to account f6Y' manufacturing 1 tolerances and further 4 increasing the value by 5% to. account for measurement uncertainties. !
Verify the requirements of Specification 3L2.2.1 are; satisfied - ' 4
- c. Satisfying:the following relationship:
F (z) 1 x K(z) for P > APL" Px-W(z)BL i ' M where: F z) is the measured. F FRTP is the F L11mit, ithe normalizeh(F n (z) . as- a function ok(.z). core P is hSight. khe relative .1 THERMAL' POWER. W(z)BL is_the cycle-dependent function that-accounts 1 MILLSTONE - UNIT 3 -3/42-9 Amendment No. JS 0011- e-
.i
]
-\
i POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- for limited power Rgstribution. transients encountered 'during base' ,
load operation. F o , K(z), and W(z)BL:are specified in-the COLR as per Specification 6.9.1.6. 1 4 a
- d. Measuring F (z) in conjunction with target flux-difference determi- <
nation acco ding to-the following' schedule: .j , (1) Prior to entering base load operation. ~after . satisfying Sec-- i tion 4.2.2.1.3 unless a full core flux" map has been taken -inc d the - previous 31 EFPD witg theJ relative : thermal power having> j! been maintained above.APL for' the 24 hours prio'r-to mapping,. and ;
' 3 i
(2) At least once per 31 Effective Fu11' Power Days.: r
- e. With the maximum value of <
K(z) over the core height (z) increasing since the' previous determination ' i of.F (z), either of the following actions shall be taken: 1 (1). F (z) shall be increased by 2% over that- specified in 4 4.2.2.1.4.c, or (2) FM (z) shall be measured at least
- once per .7 Effective Full" '
P wer Days until 2 successive maps .-indicate? that' the maximum. ! value of s F (z) K(z) -l over the core height (z)'is not increasing. ; 1 4
. MILLSTONE - UNIT-3 3/4 2-10 - Amendment :No,'- ##-- l 0011'. .1 1-
]
r l-POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) !
- f. With the. relationship specified in 4.2.2.1.4.c not being satisfied,_ .l either of the following actions shall be-taken:- 1 (1) P1 ace core in - anL- equilibrium- conditioywhere . the limit in- .
4.2.2.1.2.0 is satisfied, and remeasure Fg (z),;or.: 1 (2) Comply with the requirements of Specification 3.2.2.1 for Fg(z) , exceeding its limit by.the maximum percent calculated over the core height--(z)_with the fcilowing expression:: ll 1 F (z) x W(z)BL ND RTP
-.1 x 100 for Pi:APL l 0 xK(z)- !
P s :; i
- g. The limits .specified in - 4.2.2.1.4.c, 4.2.2.1.4.e,-J and 4.2;2.1.4.f .!
are not applicable in the following core plane regions:- 1 (1)~ Lower core region 0%~to-15%, inclusive. (2) Upper core region 85% to 100%, -inclusive. .i 4.2.2.1.5 When F0(z) is measured for_ reasons other'than meeting the require-ments of Specification 4.2.2.1.2, an overallimeasured. Fn(z)ishallL be obtained ! from a power distribution map and increased by 3% to acc' bunt for manufacturing a tolerances and furtherLincreased by 5% to account /,or measurement uncertainty. , t l J
-i i
)
i gSTONE-UNIT 3 3/4.2-11 Amendment No.- Jp- - i
i I POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR - F Q (Z)
-THREE LOOPS OPERATING' LIMITING CONDITION FOR OPERATION _
i- . 3.2.2.2 F (Z) shall be limited by the following-relationships: 4 n RTP. . . F Fn (Z) s 0 (K(Z)]-for P > 0.375 ! P < r RTP. . Fn(Z) 1 (F0 ) (K(Z)] for P s 0.375 l 0.375 RTP F = The F limit' at RATED THERMAL' POWER (RTP) specified in - theCOREOPEbTINGLIMITSREPORT(COLR). THERMAL ~ POWER ,-and Where: P = RATED THERMAL POWER K(2) = the normalized F (Z) as 'a function. of core height speci- l fied in the COLR. 0 l APPLICABILITY: MODE 1. ACTION: With Fg (Z) exceeding its limit: 1 l a. Reduce THERMAL POWER.at least 1% for each 1% Fn(Z) exceeds the l limit within 15. minutes and- similarly reduce 'theLPower Range Neutron Flux-High Trip:Setpoints within the'next'4 h'ours; POWER_ OPERATION may proceed for:up to a? total of 72. hours;' subsequent i POWER OPERATION may proceed provided the Overpower AT Trip' Set-L points have been reduced at=least:1% for each.1% F (Z): exceeds l the limit. The Overpower AT Trip: Setpoint' reductbn- shall be performed with the' rea; tor in at'least ' HOT STANDBY.
- b. Identify. and correct. the cause of the. out-of-limit condition' i prior to increasing THERMAL ~ POWER above,the reduced limit re-quired by ACTION a., above; THERMALLPOWER'may then be increased provided Fn(Z) is- demonstrated through incore imapping to be ;
within its ' limit. I . Millstone - Unit 3 3/4 2-12 -Amendment No. JS - 0011 s li__ '_____._._____.____ + 4-
POWER DISTRIBUTION LIMITS i SURVEILLANCE RE0VIREMENTS 4.2.2.2.1 The provisions of Specification 4.0.4 are not applicable. j i 4.2.2.2.? For RAOC operation,q F (z) shall be-evaluated to-determine if F q (z) is within its limit by: i
- a. Using the movable incore detectors to obtain'a power distribution map ]
at any THERMAL POWER greater than 5% of RATED THERMAL POWER. j i
- b. Increasing the measured F (z) component of the power distrib'ution . map . l by 3% to account for . manhfacturing ' tolerances - and - further - increasing 1
the value by 5% to account for measurement uncertainties. Verify the- ; requirements of Specification 3.2.2.2'are satisfied. (
- c. Satisfy the following relationship: -4 i
* *I for P > 0.375-F (z) 1 !
P x W(z) l i' RTP * *) F (z) 1 0 for P $ 0.375 1 W(z) x 0.375 where F$(z)is the measured Fn (z)- increased- by the agwances for i manufacttring tolerances and me'&surement- uncertainty, F is the F limit, K(z) is the' normalized Fn(z) as a function of corh: height, P 10 the relative THERMAL POWER,. and. W(z) is the cycle-dependent function that accounts for gr . distribution transients, encountered during normal operation. F , K(z),' and W(z) are specified in the COLR as perSpecification6.h.l.6.
- d. Measuring F (z) according to the following schedule:
(1) - Upon achieving equilibrium conditions after exceeding by 10%- or i more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was q last determined,
- or i
*During power escalation at the beginning of each cycle, the power level may be increased until a power level for extended operation-has been achieved and power distribution map obtained.
1
. Millstone - Unit-3 3/4 2-13 Amendment No. fEl 0011
l l> POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) i (2) At least once per 31 Effective Full Power Days ~, whichecer occurs-first.
- e. With the maximum value ofl F (z) I K(z) -
over the core height-(z) increasing since the previous determination: ! of F (z), either of the following actions shall be taken: j i (1) F (z) shall be increased by 2% over that specifiedLin Specifica- ) t on 4.2.2.2.2c, or -i (2) F (z) shall be measured at least once per 7 Effective Full Power i D ys until two successive maps indicate that the maximum value of F (z) , K(z) over the core height (z) is not increasing,
- f. With the relationships specified in Specification 4.2.2.2.2c not being satisfied:
(1) Calculate the maximum percent over the core height (z) that Fq (z) exceeds its limit by the following expression:l F (z) x W(z)- l RTP 1 x 100 for P 2 0.375 0 x K(z) P '
?
M]lstone-Unit 3 3/4 2-14 Amendment No J7, 59
l 1 POWER DISTRIBUTION LIMITS SVRVEILLANCE RE0VIREMENTS (Continued) 'I
~ '
,1 F (z) x W(z)
RTP
-1 x 100 for P <.0.375 0 x K(z) '
O.375 i (2) One of the following actions shall.be taken. i (a) Within 15 minutes, control .the AFD to within new AFD limits ! which are determined by reducing the applicable AFD limits' ! by 1% AFD for each percent - F z) exceeds ' its ' limits; as determined 'in Specification 4.0.(2.2.2f l. Within 8, hours,-
]
reset the AFD alarm setpoints to these modified limits, or j (b) Comply with- the requirements of. Specification 3.2.2.2 for - j F9 (z) exceeding its limit by the percent calculated, or ] (c) Verify that the requirements of Specification!4.2.2.2.3 for base load operation are satisfied and enter base load ; operation. , 1
- g. The limits specified in Specifications 4.2.2.2.2c, ,4.2.2'.2.2e, and-4.2.2.2.2f are not applicable in the following core plane regions:- '
(1) Lower core region from 0% to 15%, inclusive. (2) Upper core region from 85% to 100%, inclusive. 4.2.2.2.3 Base load operation is - permitted at powers _ above APLND.47.the following conditions are ' satisfied:
- a. Prig to entering base load operation, maintain . THERMAL POWER .above APL and less than or equal to that allowed .by Specifica-tion 4.2.2.2.2 for at least the previous 24 hours. Maintain base load i operation -surveillance (AFD within the target band limit about - the-target flux difference of Specification 3.2.1.2), during this time-period.
Base load _ operation g then pgmitted providig THERMAL POWER is ! maintained between APL and APL or between APL and 100% (whichev-er is most limiting) and F surveillance is maintained pursuant to-Specification 4.2.2.2.4. nBL APE is defined as the minimum value of: Millstone - Unit 3 3/4_2-15 Amendment No. JJ, JS 0011-l a
i POWER DISTRIBt) TION LIMITS SURVEILLANCE REQUIREMENTS (Continued) i Bl
- APL = x 100%
F (z) x W(z)BL j over the core height (z) where: Ff(z)isthemeasuredF(z) o increased -l by the allowances for manufacg ing tolerances and measurement- l uncertainty. The Fn limit is F 0 W(z)BL is the cycle dependent - ft:nction that- accollnts for limited poweETPdistribution transient ]i encountered during base load operation. F g , K(z), and W(z)g are specified in the COLR as per Specification 6'.9.1.6.
- b. Durgg base load operation, if the THERMAL POWER is decreased 'below l '
APL then the conditions of 4~.2.2.2.3.a shall be satisfied before reentering base load operation. 4.2.2.2.4 During base load operation F (z) shalil be evaluated to determine if F9 (z) is within its limit by: 9 a. Using the movable incore detergrs to obtain a power distribution map at any THERMAL POWER above APL ] a
- b. Increasing the measured F n (z) component of the power distribution map by 3% to account for mant7facturing . tolerances and further increasing the value by 5% to account' for measurement uncertainties. Verify-the '
requirements of Specification 3.2.2.2 are satisfied. ,
- c. Satisfying the following relationship:
F Ff(z)1 0RTP
- EIZ) for P >-APL HD P x W(z)BL M RTP where: F The F is the F limit, the 0
normalizeh(z) n is the measured F (z).F (z)relative P is khe as a function 0f ' c THERMAL PCWER. W(z)' is +.he cycle-dependent function that accounts for limited pgr diskribution transients encountered during base load. operation. F g , K(z), and W(z)BL are specified in' the COLR as per Specification 6.9.1.6.
- d. Measuring F$(z) in conjunction with target flux difference determina-tion accordYng to the following schedule:.
Millstone - Unit 3 3/4 2-16 Amendment No. A7, R 0011
l; POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) ! (1) Prior - to entering base load operation . after satisfying Sec-tion 4.2.2.2.3, unless a full core flux map has been taken in the . previous 31 Effective Full Power Daysgth -the relative: THERMAL l POWER having been maintained above APL for the24 hours; prior to mapping, and a (2) At least once per 31 Effective Full Power Days.- 1
- e. With the maximum value of M
F0 5*)' K(z)- { over the core height (z) increasing since the previous determination' of F (z), either of the following. actions shall be taken: i (1) F (z) shall be increased by 2 percent over : that specified in: 4 2.2'2.4.c, or (2) F$(z) shall be measured at least once per 7 Effective Full Power ! Days until 2 successive maps.' indicate that-the maximum value of N FO III K(z) overthecoreheight(z)isnotincreasing. *
- f. With the relationship specified in 4.2.2.2.4.c.;not being satisfied,u either of the following actions shall be taken:: ,
(1) Place core in an equilibrium,conditi where the limit 'in 4.2.2.2.2.c is satisfied, and remeasure F (z), or (2) Comply with the requirements of Specification 3.2.2.2 for F (z)- exceeding its limit by the maximum percent calculated over9ths core height (z) with the-followin'g expression: Millstone - Unit 3 3/4 2-17 Amendment No. A7, 59 > 0011 .
1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) F(z)'xW(z)BL ND
-1 x 100 for P-1 APL 2221 p xK(z)_
- g. The limits specified in 4;2.2.2.4.c, 4.2.'2.2.4.e, and. 4.2.2.4.f are not applicable in the following core plane regions:
(1) Lower core region 0% to '15%, inclusive. (2) Upper core. region 85%!to 100%, inclusive. 4.2.2.2.5 When F0 (z) is measured for : reasons other than meeting _ the _ require-ments of Specification 4.2.2.2.2, an overall _ measured Fn(:) shalli be- obtained from a power distribution map and increased by 3%;to acclunt for. manufacturing-tolerances and further increased by 5% to account for measurement uncertainty. i 1 i
)
Millstone - Unit 3 3/4 2-18 . Amendment No. 77, ES; 0011-
POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR-
-FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2.3.1 The indicated Reactor Coolant- System (RCS) total _ flow rate and F q ,
shall be mainta'.1ed as follows:
- a. RCS total flow rate 2 387,480 gpm, and
- b. Ffg1FRTP Ag (1.0 + PFAH (1.0 - P)]
Where: THERMAL POWER ,
=
- 1) P RATED THERMAL POWER
- 2) F g- Measured values of Ffg obtained by using the movable -
incore detectors to ptain a power distribution map.- The measured value of F should be - used since Specifica-tion 3.2.3.lb.takesinfo"considerationa-measurementuncertainty of 4% for incore measurement, RTP N
- 3) F = The F limit at RATED THERMAL POWER in the CORE OPERATING lblTS REPORfH(COLR),
- 4) PF anhH - The power factor multiplier for- F Hprovided in the COLR,
- 5) The measured value of RCS total flow rate shall be used since-uncertainties of 2.4% for flow measurement have been included in Specification 3.2.3.la.
APPLICABILITY: MODE 1. ACTION: With the RCS total flow rate or F H outside the region of acceptable operation:
- a. Within 2 hours either:
- 1. to. within the above Restore limits, or the RCS total flow rate and FlH
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED - THERMAL ' POWER within the next 4 hours.
MILLSTONE - UNIT 3 3/4 2-19 Amendment U , # 0011
I l. ! l l . l POWER DISTRIBUTION LIMITS ! I LIMITING ~ CONDITION FOR OPERATION l ACTION (Continued)
-{
b. Within 24 hours of initially being outside the above limigs, verify through incore flux mapping and RCS total flow. rate that F and RCS-total flow rate are restored to within .the above limits, AHor reduce j THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours.- l
- c. Identify and correct the cause of the out-of-limit condition prior to .
increasing THERMAL POWER .above the reduced THERMAL POWER' limit required by ACTION a.2. andpr b~., above; subsequent POWER OPERATION may proceed provided that F AH and indicated RCS total flow rate are. i demonstrated, through incorli flux mapping and RCS total flow rate comparison, to-be within the region of acceptable operation prior to ~ exceeding the following THERMAL POWER levels:
- 1. A nominal 50% of RATED THERMAL POWER,
- 2. A nominal -75% of-RATED THERMAL POWER, and
- 3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER.- -)
SVRVEILLANCE RE0VIREMENTS
, j 4.2.3.1.1 The provisions of Specification 4.0.4 are not applicable.
4.2.3.1.2 RCS total flow rate and F g shall be determined to be within the acceptable range:
- a. Prior to operation above 75% of RATED: THERMAL POWER after each fuel:
loading, and- 4
- b. At least once per 31' Effective Full Power Days. ,
4.2.3.1.3 The indicated RCS total flow rate shall be verified to be within the acceptable H range at least once per 12 hours when the most recently obtained value of F AH, btained per Specification 4.2.3.1.2, is assumed-to exist. 4.2.3.1.4 The RCS total flow rate indicators shall be subjected to a CHANNEL
~
CALIBRATION at least once per 18 months.- The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow
~
measurement. l l MILLSTONE - UNIT 3 3/4.2-20 0011
i l POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.3.1.5 The RCS total flow rate shall be determined by precision heat balance i measurement at least once per 18 months. Within 7 days prior to performing the ' precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated. 4.2.3.1.6 If the feedwater venturis are not inspected at least once per 18 months, an additional 0.1% will be added to the total RCS flow measurement uncertainty. I l l 1 4 MILLSTONE - UNIT 3 3/4 2-21 Amendment No. //
-0011
i l 1 EqWER DISTRIBUTION LIMITS l3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ! THREE LOOPS OPERATING i LIMITING CONDITION FOR OPERATION ,_ j 3.2.3.2 Tha indicated Reactor Coolant System (RCS) total flow rate and' FfH shall be maintained as follows: l
- a. RCS total flow rate 2.303,200 gpm, and N
fi
- b. F g gpRTP [1.0 & PFAH (1.0 - P)] [
s Where:
- 1) p THERMAL POWER RATED THERMAL POWER i
N N
- 2) F Measured values of F A obtained by using the movable' iNo=redetectorstoobtainYpowerdistributionmap.
The measured value of. F N should be . used . since Speci- 1 fication3.2.3.2b. takes:intNconsiderationameasure- ! ment uncertainty of -4% for incore measurement;.' RTP
- 3) F = The F limit at RATED THERMAL POWER in' the CORE 'l ONRATINGLIMINREPORT(COLR),- i
- 4) PFAH - The power factor multiplier for Fgg in the COLR, and- _
]
1
- 5) The measured value of RCS. total flow rate shall ~'be used since -;
uncertainties of 2.8% for flow measurement have,been included " in Specification 3.2.3.2a. APPLICABILITY: MODE 1. ACTION: N With the RCS total flow rate or F AH outside the r'egion :of acceptable operation':
- a. Within 2 hours either: '
1. Restore limits, or the RCS total flow rate and FlH. to within- the above , 1 2. Reduce THERMAL POWER to less than 32% of RATED THERMAL-POWER an'd reduce the Powe'r Range Neutron Flux - High Trip Setpoint to less i than or : equal to '37% of RATED THERMAL POWER within the next 4 hours. -MILLSTONE'- UNIT 3 3/4 2 . Amendment No. # , J# '0011 4
l 1 i POWER DISTRIBUTION LIMITS j i LIMITING CONDITION FOR OPERATION ACTION (Continued)
- b. Within 24 hours. of initially being outside~ the above_ limits,. verify-j through incore flux mapping and RCS total flow rate that- Fhand RCS.
total flow-rate are restored = to within the above limits, or reduce; 1 ' THERMAL POWER to less-than 5% of: RATED THERMAL POWER.within the next. 2 hours. l
- c. Identify and correct the cause of the- ou't-of-limit conditio'n to increasing THERMAL POWER above the reduced THERMAL . POWER prior limit required by ACTION a.2. and/or b.', above; subsequent' POWER OPERATION' may proceed provided that F g and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within-the-region of acceptable operation prior to-exceeding the following THERMAL POWER levels. -
- 1. A nominal 32% of' RATED-THERMAL POWER, ~and !
- 2. A nominal 50% of RATED . THERMAL POWER. q SURVEILLANCE RE0VIREMENTS
)
4.2.3.2.1 The provisions of Specification 4.0.4 are not applicable.- 4.2.3.2.2 RCS total flow rate and FN shall ~be- determined to be within the 1 acceptable range at least opre per 31 Effective Full Power Days. 4.2.3.2.3 The indicated Rt,S total flow rate shall ber verified to be within the acceptable rangg at least once per 12- hours-. when the. most; recently ' obtained value of F AH, btained per Specification 4.2.3.2.2, isi assumed to exist, t 4.2.3.2.4 .The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at -least once per 18 months. . The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement. : 4.2.3.2.5 The RCS total flow rate. shall be -determined by precision h' eat balance measurement at least; once' per 18 months. Within' 7 days prior ' to performing the precision heat bal ance, -the ' instrumentation used for , determination of steam pressure, feedwater pressure, - feedwater temperature, and feedwater ventt.: , AP in the calorimetric calculations shall be -calibrated. > 4.2.3.2.6 If the feedwater venturis. are not inspected at least. once per 18 months, an additional 0.1% will be added to the total RCS flow measurement uncertainty. . l MILLSTONE-- UNIT 3 3/4 2 Amendment No. E/' 0011
.i I
POWER DISTRIBUTION LIMITS i i 3/4.2.4 -OUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION E 3.2.4 The Qt'ADRANT POWER TILT RATIO shall' not exceed l'.02. j APPLICABILITY: MODE-1, above 50% of RATED THERMAL: POWER *. ACTION:
- a. With the QUADRANT POWER TILT: RATIO determined to exceed 1.02 but -
g less than or equal to 1.09: .
- 1. Calculate the QUADRANT POWER TILT RATIO. at least once per . hour until either: :
a) The QUADRANT POWER TILT RATIO is reduced .to withih its j limit, or E- b) THERMAL POWER is reduced to-less than 50% of RATED THERMAL POWER.
- 2. Within 2 hours either:
1 a) Reduce the QUADRANT' POWER TILT RATIO to.within its ~ limit, or b) Reduce THERMAL POWER at:least 3% from RATED THERMAL = POWER for each 1% of indicated -QUADRANT POWER : TILT RATIO in - excess of 1 and similarly reduce the Power. Range Neutron Flux-High Trip Setpoints within the next.4 hours.
- 3. Verify that the-QUADRANT POWER TILT RATIO .~1s within; its limit -
within 24 hours after. exceeding , the ' limit' or reduce ; THERMAL POWER to less than 50%- of RATED- THERMAL POWER within the next-2 hours and reduce the Power Range' Neutron Flux-High Trip Setpoints to less than' or equal to 55%' of- RATED THERMAL POWER within the next 4 hours; and
- 4. Identify and correct the c'ause of the out-of-limit -condition prior to . increasing THERMAL' POWER; - subsequent 1 POWER OPERATION ;
above 50% of RATED THERMAL-POWER may proceed provided that the t QUADRANT POWER TILT RATIO is verified:within its limit. at least once per hour for 12 hours or until verified' acceptable at 9$% or greater RATED THERMAL POWER.
*See Special. Test Exceptions Specification 3.10.2.
l ; MILLSTONE - UNIT 3 3/42-24 0011
i POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ; i ACTION ' Continued 1
- b. With the QUADRANT POWER TILT RATIO determined to exceed l'.09 due to misalignment of either a shutdown or control rod:
- 1. Calculate the-QUADRANT POWER TILT RATIO- at least once per hour i until either- i i
a) The QUADRANT POWER TILT RATIO is reduced to within its i limit, or j b) THERMAL POWER is reduced.to less than 50%'of RATED THERMAL POWER. a
- 2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for f each 1% of indicated QUADRANT POWER TILT RATIO in l excess of .1, within 30 minutes; j
- 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit ors reduce . THERMAL ,
POWER to less than 50% of . RATED. THERMAL ' POWER 'within the. next 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55%-of RATED-THERMAL POWER within the next 4 hours; and
- 4. Identify and correct the cause of the out-of-limit condition !
prior to increasing THERMAL POWER; subsequent POWER OPERATION t above 50% of RATED THERMAL POWER may proceed provided.that the QUADRANT POWER TILT RATIO is verified within its limit 'at least _ : once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.
- c. With the QUADRANT POWER TILT RATIO determined.to exceed 1.09 due.to causes other than the misalignment of either a shutdown or control '
rod:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is reduced to within - its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. MILLSTONE - UNIT 3 3/4 2-25 ' 0011
POWER DISTRIBUTION LIMITS 1]MITING CONDITION FOR OPERATION ACTION (Continued) ., 2. Reduce THERMAL POWER to less than- 50% of RATED THERMAL 1POWERL i within 2 hours and reduce the Power Range Neutron Flux-High.
- Trip Setpoints to less than or. equal to 55% of _ RATED ~ THERMAL _- ,
POWER within the next 4 hours; and J
- 3. Identify _ and correct the - cause of 'the out-of-limit condition-
, prior to increasing THERMAL POWER; subsequent POWER OPERATION , above 50% of RATED THERMAL POWER.may-proceed provided'that the QUADRANT POWEF. TILT RATIO is ' verified within its limit at least ' once per hour for 12 hours or until verified at 95% or-_ greater-. . l RATED THERMAL POWER. l l d. The provisions of Specification 3.0.4 are not applicable. l SURVEILLANCE RE0VIREMENTS I-t 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
- a. Calculating the ratio at least once per 7 days when_ the alarm is OPERABLE, and
- b. Calculating the ratio at~ least- once per 12 hours during. steady-state operation when the alarm is inoperable, 4.2.4.2 The QUADRANT POWER TILT RATI0'shall' be' determined to be within the 1imit when above 75% of RATED THERMAL POWER 'with one Power- Range ? channel inoperable by using the movable incore detectors - t'o ' confirm thatthe 't normalized ' symmetric power distribution, obtained from two sets of - four symmetric thimble locations or - full-core ~ flux map, is -consistent with the i
L indicated QUADRANT POWER; TILT RATIO at least once_per 12 hours.. 1 t l 1^ l l MILLSTONE . UNIT 3
- 0011- 3/4 2-26
'!g POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS i LIMITING CONDITION FOR OPERATION 3.2.5- The following DNB-related parameters shall be maintained within the limits shown on Table 3.2-1: 1
- a. Reactor Coolant System T,yg, andl b, Pressurizer Pressure.
]
APPLICABILITY: MODE 1. l
.j.
ACTION: j i With any of the above parameters exceeding its limit, restore the parameter to - i within its-limit within 2 hours or' reduce THERMAL POWER:to less than 5% of-RATED THERMAL POWER within the next 4 hours. SURVEILLANCE RE0VIREMENTS 3 4.2.5 Each of the parameters of Table'3.2-1 shall be verified- to be within j its limits at least once per 12 hours. . 1 s b f k 1 MILLSTONE - UNIT 3 3/4'2 0011
~_.,
TABLE 3.2-1 oZ l
;- f ~ DNB PARAMETERS I' !
o- > LIMITS c- Three Loops in Opera-
'5 Four Loops in tion & Loop Stop
-* PARAMETER Operation Valves Closed w
Indicated ^ Reactor Coolant System-T,yg s 591.l*F' s 583.3*F
- Indicated Pressurizer Pressure > 2218 psia
- i 2218 psia
- k '. .
7 -l g i
._ +.
X
- .a
~ z.
- Limit'not applicable during either a THERMAL POWER ramp in excess of 5% of; RATED' THERMAL n- & .. . POWER:per, minute or a THERMAL POWER: step in. excess,of 10% of RATED-THERMAL? POWER.
w 1
/ +
; e -.,.._-s_ _.
~
_. .GE.__23a_____z_-_c_
;__,_ ._ ._ - !, , - , L : ,_. ;-4 1.,., ,. ,14_ y m% ., % ;,,- !.',.i. , 4 4_ .- - 4. a ,', 'en. . .._.-si.s _c Z,._.. , . . _ , . . . . . , ,._4..
TABLE 3.3-2
.h
..g REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES o RESPONSE-TIME 2 FUNCTIONAL-UNIT
' N.A.
)_
- l. Nanual Reactor Trip
-' 2.- Power Range, Neutron Flux 5 0.5 second*
- 3. Power Range,. Neutron Flux, High Positive Rate N.A.
- 4. Power Range, . Neutron Flux,.
High Negative Rate: _ $ 0.5 second*
- 5. Intermediate Range, Neutron Flux N. A. .
- 6. Source Range, Neutron Flux . N. A. - ~
.m:
1 . Overtemperature AT 1;7 seconds
- w -7.
- 8. Overpower AT' - 5 7 seconds *
- 9. Pressurizer Pressure--Low 1 2 seconds ~
- 10. Pressurizer Pressure--High . s 2 seconds.
- 11. Pressurizer Water Level--High' $ 2 seconds
.F -
a i e
- Neutron detectors are' exempt from-response ftimetesting. Response time.of. thelneutron flux. signal' portion of the -
~ ._ y, '
- channel . shall ' be measured from . detector output: or. input ;of- first electronicicomponenttin _ channel . - ..
.z -
p _ 2 -W _
...c,,
I ' * " ' * **" "'""
"'*^^'"##'" - - * - ""'""4* * " " ' '
4 F + 'i'-"*8- - ^'
, ^ - -" - *:P,N-+ E -
.w
l REACTOR COOLANT SYSTEM l HOT STANDBY l LIMITING CONDITION FOR OPERATION 3.4.1.2 At least three of the reactor coolant loops listed below shall' be- - OPERABLE, with at least three reactor coolant loops in _ operation when the t Reactor Trip System breakers are closed or-with at least one reactor coolant-loop in operation when the Reactor Trip System breakers are open:* , t . I
- a. Reactor Coolant' Loop 1 and its associated steam generator and q reactor coolant pump,
- b. Reactor Coolant Loop 2 and its associated steam -generator _ and reactor coolant pump,
- c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump, and ,
- d. . Reactor Coolant Loop ~4 and its associated steam generator and reactor coolant pump.. 3 APPLICABILITY: MODE 3. f
, ACTION: 4 l lt l a. With less than the above required reactor coolant . loops OPERABLE, l restore the required loops to OPERABLE status within 72 hours or be _ 1 in HOT SHUTDOWN within the next 12 hours.. i
- b. With less than the'above required reactor: coolant loops in operation 4 and the Reactor Trip System. breakers in the closed position, within !
I hour open the Reactor Trip. System breakers.
~
- c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant-System and immediately initiate corrective action .to' return the !
required reactor coolant loop to operation. 4 SVRVEILLANCE RE0VIREMENTS r 4.4.1.2.1 At least the above required -reactor coolant ' pumps, -if not -in operation, shall be determined .0PERABLE once per 7 days. by verifying correct breaker alignments and indicated power availability.- i 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying - , secondary side water level to be greater than or equal .to 17% at least once 1 per 12 hours. ; a 4.4.1.2.3 The required reactor coolant loops shall be verified ' in operation 1 and circulating reactor coolant' at least.Once per: 12 hours.
;i
*All reactor coolant pumps may be deenergized for up to 1 hour providedi (1) no operations are1 permitted that would cause dilution of the Reactor s Coolant System boron concentration, and (2) core outlet temperature is maintained at .least 100F below saturation temperature. !
MILLSTONE - UNIT 3 - 3/4 4-2 ~l 0019'
-. ![
l REACTOR COOLANT SYSTEM ISOLATED LOOP STARTUP 1 LIMITING CONDITION FOR OPERATION 3.4.1.6 A reactor coolant loop shall' remain-isolated with power' removed from j the associated RCS loop stop valve operators until: ( q
- a. The ten.perature at the cold leg 'of the isolated loop 'is within~
20*F of the highest cold leg temperature of the operating. loops, ' q
- b. - The boron concentration of the isolated loop.xis' greater than- or i equal to the boron concentration of the operating loops,'or-greater j than 2600' ppm whichever is less, 1 c :
- c. The isolated portion of the _ loop has been drained -and is refilled, and
- d. The reactor is subcritical' by at least 1.6% Ak/k. !
APPLICABILITY: MODES 5 and 6. ACTION: '
-r
- a. With the requirements of the above specification not satisfied, do not open the isolated loop stop: valves,
- b. The provisions-of Specification.3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.4.1.6.1 The isolated loop cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of the operating loops 'within' 30 minutes prior to opening the cold leg stop valve. t 4.4.1.6.2 The reactor shall be' determined to : be suberitical by at least j 1.6% Ak/k within 30 minutes prior to opening the cold leg stop valve, j 4.4.1.6.3' Within 4 hours prior to opening the loop stop valves, the-isolated I loop shall be determined to: 1
- a. Be drained and . refilled, and
- b. Have a boron concentration greater than or equal to the boron "
concentration of the operating loops, - or greater than 2600 ppm - whichever is less. ' MILLSTONE - UNIT 3 3/4 4-8 Amendment No. # - i 0020
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor _ Coolant System (RCS) accumulator shall- be. OPERABLE with:
- a. The isolation valve open and power remcved,
- b. A contained borated water volume of between 6618 and 7030 gallons,
- c. A boron concentratton of between 2600 and 2900 ppm, and-
- d. A nitrogen cover-pressure of between_636 and 694 psia.-
APPLICABILITY: MODES 1, 2, and 3*. ACTION:
- a. With one accumulator inoperable, except as a result of a - closed isolation valve, restore the inoperable accumulator to ' 0PERABLE status within 8 hours or be in at-least H0T STANDBY within the next 6 hours and reduce pressurizer- pressure to' less than 1000 psig within the following 6 hours,
- b. With one accumulator inoperable due to the. isolation valve: being closed, either immediately open the: isolation . valve ;or - be? in at least HOT STANDBY within 6 hours-and reduce pressurizer pressure to less than 1000 psig within the following 6 hours; SURVEILLANCE RE0VIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
- a. .At least once per 12 hours by:
1)- Verifying .the contained borated water volume and nitrogen cover-pressure in the tanks to be within the above limits, and
- 2) . Verifying that each accumulator isolation valve is open. ,
- b. At least once per 31 days and 'within-6 hours after each solution volume increase of greater than or. equal to'.1% .of tank volume by ' ,'
verifying the boron concentration of the accumulator solution; and
- Pressurizer pressure above 1000 psig.
MILLSTONE - UNIT 3 3/4 5-1 Amendment No. JE, JJ. 0021 l
_ _ _ ~ - _ . _ . ._._ _ _ _ 4 L FMERGENCY CORE COOLING SYSTEMS l
.SVRVEILLANCE RE0VIREMENTS~(Continued)
- 2) A visual inspection of-ths containment sump and' verifying;that; the subsystem suction inlets' are not restricted by debris : and - 71 that the sump: components (trash racks, screens, etc.) show no. J evidence of structural distress or abnormal corrosion,
- e. At least once per 18 months, during: shutdown,-by: -
l
, 1
- 1) . Verifying that eaciautomatic valve-i A.the flow path' actuates; to l
-its correct: position on a'SafetyjInjection3 actuation test signal,-
~~ '
and
- 2) Verifying-that each of the following pumps stirt automatically :l upon receipt of'a Safety; Injection actuation test _ signal:-
q a) Centrifugal charging pump, 1 J i b) Safety (Injection pump,?and u c) RHR pump.
- 3) Verifying that the Residual, Heat Removal- pumps:stop automatica1_1y upon. receipt of a Low-Low RWST Level test signal.,
- f. By verifying that each of the following. pumps-develops theLindicated differential pressure.on recirculation flow when tested pursuant to
- Specification 4.0.5: '
J
- 1) Centrifugal charging pump 2 2411 psid, ;
1
- 2) Safety Injection pump 2 1348 psid,
,1
- 3) RHR pump 2 165 psid, and. 1
- 4) Containment recirculation pump 2130 psid. ,
- g. By verifying the correct position;of each electricalcand/or mechanical position-stop for the following ECCS throttle, valves:t
~
- 1) Within 4 hours following completion of each valve' stroking operation or maintenance' on the valve when the .ECCS ' subsystems- ,
are required to be:0PERABLE, and- i
- 2) At least once per'18 months.
ECCS Throttle Valves Valve Number Valve Number 3SIH*V6 3SIH*V25 , 3SIH*V7 -3SIH*V27 , i MILLSTONE-- UNIT 3 3/4 5-5 0022
n EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS fContinued) f.CCS Throttle Valves Olve Ni=her Valve Numher
<SlH*V8 3SlH*V107 351H*V9 351H*V108 3!!H*V21 3SIH*V109 3SfH*V23 3SIH* Vill
- h. By performing a flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
- 1) For centrifugal charging pump lines, with a single pump running:'
a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 339 gpm, and b) The total pump flow rate is less than or equal to 560 gpm. 4
- 2) For Safety Injection pump lines, with a single pump running:
a) The sum of the injection line flow rates,- excluding the highest flow rate, is greater than or equal to 442.5 gpm, and b) The total pump flow rate is less than' or equal to 670 gpm for the A pump and 650 gpm for the B pump.
- 3) For RHR pump lines, with a single pump running, the sum of the-injection line flow rates is greater than or equal to 3976 gpm.
MILLSTONE . UNIT 3 3/4 5 6 0022-S _ . 1
EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE witn:
- a. A contained borated water volume between 1,166,000 and 1,207,000 -
gallons, (
- b. A boron concentration between 2700 and 2900 ppm of boron, I
- c. A minimum solution temperature of 40'F, and
- d. A maximum solution temperature of 50'F.
APPLICARILITY: MODES 1, 2 5 3, and 4. ACTION -i With the RWST inoperable, restore the tank to OPERABLE status within I hour or , be in at least 'l0T STANDBY within 6 hours and in COLD SHUTDOWN within the ! following 30 hours. H EMByflLLANCE REOUIREMENTS ' i 4.5.4 The RWST shall be demonstrated OPERABLE:
- a. At least once per ' days by: !
- 1) Verifying the contained borated water volume in the tank, and
- 2) Verifying the boron concentration of the water.
1
- b. At least once per 24 hours by verifying the RWST temperature, i
l l l MILLSTONE.- UNIT 3 3/4 5-9 Amendment No.'JJ. 0023 4
O TAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 9 3.6.2.3 The Spray Additive System shall be OPERABLE with:
- a. A chem.a1 addition tank containing a volume of between '17,760- and 18,760 gallons of between 3.4 and 4.1% by weight NaOH solution, and -
- b. Two gravity feeo paths each capable of adding NaOH solution from the chemical addition tank to each Containment- Quench Spray subsystem pump suction.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: l-With the Spray Additive System inoperable, restore the system to' 0PERABLE i status within 72 hours or be in at least HOT STANDBY within ine next 6 hours; restore the Spray Additive System to.0PERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.2.3 The Spray Additive System shall be demonstrated OPERABLE: I
- a. At least once per 31 days by verifyi g that each valve (manual, power-operated, or automatic) in the f$ow path that is not locked, i sealed, or otherwise secured in position, is' in its correct position;
- b. At least once per 6 months by:
- 1) Verifying the contained solution volume in the tank, and
- 2) Verifying the con ?ntration of the Na0H solution by chemical !
analysis is within the above limits,
- c. At least once per 18 months, during shutdown, by verifying that each ;
automatic valve in the flow path actuates to its correct position on ' a CDA test signal. I 1
)
i MILLSTONE - UNIT 3 -3/4614 00t4 Amendment No. J2
3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1.1 The boron concentration of all filled portions of the Reactor Coolant E System and the refueling canal shall be iraintained uniform and sufficient to , ensure that the more restrictive of the following' reactivity conditions is met; either:
- a. A K,ff of 0.95 or less, or
- b. A boron concentration of greater than or equal to 2600 ppm.
APPLICABILITY: MODE 6.* ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or- positive reactivity changes and initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6300 ppm boron or its is reduced to less than or equal to 0.95 or the boron equivalent until concentration is rI K ((ored to greater than or equal' to 2600 ppm, whichever is the more restrictive. e SURVElllANCE RE0VIREMENTS 4.9.1.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to: \
- a. Removing or unbolting the reactor vessel head, and
- b. Withdrawal of any full-length control rod-in-excess of 3 feet from its fully inserted position within the reactor vessel.
4.9.1.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours. 4.9.1.1.3 per 31 dhys.Valve 3CHS-V305 shall be verified closed and locked at least once
*The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
MILLSTONE - UNIT 3 3/4 9-1 00th Amendment No. JJ
e 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: 1 the reactor can be made suberitical from all operating conditions, (2) the(re) activity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T . The most restrictive condition occurs at EOL, with T at no load opeN@ing temperature, and is associated with a postulated stilR line break accident- and resulting uncon-trolled RCS cooldown. . In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% Ak/k is required to control the reactivity . transient. 1 Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T less than 200'F, the reactivity transients resulting from a postulated sillE i line break cooldown are minimal. A 1.6% Ak/k SHUTDOWN MARGIN is- required to ! provide protection against a boron dilution accident.
]
3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient .(MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses. . The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison. The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC. used in the FSAR analyses to nominal operating conditions. k l i l gSTONE-UNIT.3 B 3/4 1-1 Amendment No..f)
i ll REACTIVITY CONTROL SYSTEMS l BASES ! BORATIONSYSTEMS(Continued) MARGIN from expected operating conditions of 1.6% Ak/k after xenon decay and 1 cooldown to 200'F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires a usable volume of 17,610 gallons of 6300 ppe borated water from the boric acid storage tanks or 1,166,000 gallons of 2700 ppm borated water from the-refueling water storage tank (RWST). A minimum: RWST volume of 1,166,000 gallons is specified to be consistent with ECCS requirement. With the RCS temperature below 200'F, one Boron Injection . System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable. The limitation for a maximum of one centrifugal charging pump to be OPER-ABLE and the Surveillance Requirement to verify all charging pumps except the required OPEkABLE pump to be inoperable below 350'F provides assurance that a j mass addition pressure transient can be relieved by the operation of a single l PORV. l The boron capability required below 200'F is sufficient to provide a SHVIDOWN MARGIN of 1.6% Ak/k after xenon decay and cooldown from 2000F to 1400F, This condition requires either a usable volume of 1680 gallons of 6300 ppm borated water from the boric acid storage tanks or a usable volume .of 85,840 gallons of 2700 ppm borated water from the RWST. The contained water volume limits include allowance for water not available because of discharge line location and other physical
- characteristics.
The limits on contained water volume and boron concentration of the RWST l also ensure a pH value of between 7.0 and 7.5 for the solution recirculated l within containment after a LOCA. This pH band' minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The minimum RWST solution temperature for MODES 5 and 6 is based on analysis assumptions in addition to freeze protection considerations. The minimum / maximum RWST solution temperatures for MODES 1, 2, 3 and 4 are based on analysis assumptions. i The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while. in MODE 6. , l MILLSTONE - UNIT 3 B 3/4 1-3 , Amendment No. JJ 0011
l REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued) rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within 112 steps at 24, 48,120, and fully withdrawn position for the Control Banks and 18, 210, and fully withdrawn position for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position Indication System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only j points in the indicated ranges are picked for verification of agreement with-
- demanded position. !
The ACTION statements which permit limited variations from the basic l requirements are accompanied by additional restrictions which ensure that the ' original design criteria are met. Misalignment of a rod requires measurement i of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation. { The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T greater than or ' equal to 5510F and with all reactor coolant pumps operatiS9 ensures that the 1 measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions. Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours with more fre-quent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable t ( LCOs are satisfied. For Specification 3.1.3.1 ACTIONS b. and c... it is incumbent upon the plant to verify the trippability of the inoperable control rod (s). Trippability is defined in Attachment C to a letter dated December 21, 1984, from E. P. Rahe (Westinghouse) to C. 0.' Thomas- (NRC). This may be by , verification of a control system -failure, usually electrical in' nature, or ! that the failure is associated with the control rod stepping mechanism.- In the event the plant is unable to verify the rod (s) trippability, it must be assumed to be untrippable and thus falls under the requirements. of ACTION a. Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately 4 hours for this verification. l MILLSTONE - MIT 3 B 3/4 1-4 0027
( i 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) ' events by: (1) maintaining the minimum DNBR in the core greater than or equal-to the design limit during normal operation and.in short-term transients, 'and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within - assumed design criteria. In_ addition, limiting the peak linear power density _ during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met i and the ECCS acceptance criteria limit of 2200*F is not exceeded, ; 1 The definitions of certain hot channel and peaking factors as used in these specifications are as follows. 1 F(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat i n flux on the surface of a fuel rod at core elevation Z divided by the i average fuel rod heat flux, allowing for manufacturing tolerances on i fuel pellets and rods; and- ! Ffg Nuclear Enthalpy Rise Hot Channel Factor, is defined as' the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. ! 3/4.2.1 AXIAL FLUX DIFFERENCE , The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fn(Z)- upper bound envelope of the Fn limit specified in the Core Operating Lim'Its Report (COLR) times the normali' led axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. Target flux difference is determined at equilibrium xenon conditions. The full-length rods may be positioned within the core in accordance with-their respective insertion limits and should be inserted -near their normal position for steady-state operation at high power . levels. The value of the target flux difference obtained under these conditions divided by~ the fraction of RATED THERMAL POWER is the target- flux difference at RATED THERMAL POWER ' for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value ; by the appropriate fractional THERMAL _ POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup ', considerations. q l l l MILLSTONE - UNIT 3 B 3/4 2-1 Amendment No. JS 0026
1 POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued) At power levels below APL ND , the limits on AFD are defined in the COLR consistent with the Relaxed Axial Offset Control (RAOC) ope ating procedure and limits. These limits were calculated in a manner such that expected operational transients, e.g., load follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the short period of time allowed outside of the limits at reduced power levels will not result in significant xenon redistribution such that the envelope of peaking facgrs would change sufficiently to prevent operation in the vicinity of the APL power level. At power level s greater than APL ND , two modes of operation are permissible: (1) RA00, the AFD limit of which are defined in the C0lR, and (2) base load operation, which is defined as the maintenance of the AFD within COLR specgications band about a target value. The RAOC operatgg procedure above APL is the same as that defined for operation below APL However, it is possible when following extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee operation with0F (z) less than its limiting value. To allow operation at the maximum permissible power level, the base load operating procedure restricts the indicated AFD to rgatively small gargat band (as specified in the COLR) and power swings (APL 1 power s APL or 100% Rated Thermal Power, whichever is lower). For base load operation, it is expected that the plant will operate within the target band. Operation outside of the target baad for the short time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above. To assure there is no residual xenon redistribution impact from past operation on tg base load operation, a 24-hour waiting period at a power level above APL and allowed by RAOC is necessary. During this time period load changes and rod motion are restricted to that allowed by the base load procedure. After the waiting period, extended base load operation is permissible. The computer determines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are: (1) outside the allowed delta-1 power operating space (for RAOC operation), or (2) outside the allowed delta-I target band (for base load operation). These alarms are active when power is greater than (1) 50% of RATED THERMAL POWER (for RA0C operation), or MILLSTONE - UNIT 3 8 3/4 2-2 Amendment No. ES 0028
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, q 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 i acceptance criteria limit. 3 Each of these is measurable but will nor;nally 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: I
- 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; j
- b. Control rod groups are sequenced with overlapping broups as described in Specification 3.1.3.6; 1
- c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and i i
- d. The axial power distribution, expressed in terms of AXIAL FLUX- ;
DIFFERENCE, is maintained within the limits. i FfH will be maintained within its limit [ provided Conditions a. _through !
- d. above are maintained. The relaxation of Fgg as a function'of THERMAL POWER l allows changes in the radial power shape for all permissible rod insertion '
limits. N The F as calculated .in Specifgcations 3.2.3.1 and 3.2.3.2 are used in the various accident analyses where F influences parameters other than DNB:t, AH e.g., peak clad temperature, and trius is ~ the maximum "as measured"~ value , allowed. I i MILLSTONE - UNIT 3 B 3/4 2 Amendment No. # l 0078 ; I
POWER DISTRIBUTION LIMITS BASES 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 (Continued) Margin is maintained between the safety ' analysis limit DNBR and the design limit DNBR. This margin is more than sufficient to offset any rod bow-penalty and transition core penalty. The remaining margin is available for plant design flexibility. When an oF measurement is taken, an allowance for both experimental error ! and manufacturlng tolerance must be made. An allowance of 5% is appropriate ' for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance. l M The hot channel factor Fo (z) is measured' periodically and increased by a I cycle and height-dependent power factor apprapriate to either RA00 or base l loadoperation,W(z)orW(z)B to provide assurance that the limit on the hot l channel factor, Fn(z) is met.L, W(z) accounts for the effects of normal ~ opera- ' tion transients ahd was determined from expected power control maneuvers over 4 the full range of burnup conditions in the core. W(z)m accounts for the more ! restrictive operating limits allowed by base load opWation which result in ' less severe transient values. The W(z) and W(z)gt actions described above for normal operation are specified in the COLR per Specification 6.9.1.6. When RCS flow rate and F N are measured, no additional allowances are necessary prior to comparison w$th the limits of the Limiting Condition for Operation. Measurement errors of 2.4% for four goop flow and 2.8% for three ! loop flow for RCS total flow rate and 4% for Fgg have been allowed for in ' determination of the design DNBR value. The measurement error for RCS total ' flow rate is based upon performing a I precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be _ l detected could bias the result from the precision heat balance in a non-conservative manner. Therefore, a penalty of 0.1% for undetected . fouling .of ' the feedwater venturi will be added if. venturis- arc inspected and cleaned j at least once for 18 months. Any fouling which might bias the RCS flow rate i measurement greater than 0.1% can be detected by monitoring and trending i 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 ' 0028 Amendment No. JJ
POWER DISTRIBUTION LIMITS i BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) The 12 hour periodic surveillance of indicated RCS flow is sufficient to . detect only flow degradation which could lead to operation outside the accept-able region of operation defined in Specifications 3.2.3.1 and 3.2.3.2. i 3/4.2.4 OUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power ; distribution satisfies the design values used in the power capability l analysis. Radial power distribution measurements are made during STARTUP-testing and periodically during power operation. The limit of 1.02, at which corrective action is required, provides DNB l and linear heat generation rate protection with x-y plane power tilts. A L limiting tilt of 1.025 can be tolerated before the. margin.for uncertainty in i F0 is depleted. A limit of 1.02 was selected to provide an allowance for the ; uncertainty associated with the indicated power tilt. ' The 2-hour time allowance for operation with a tilt condition greater ! than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod. In the ovent such, action does not correct the tilt, the margin for uncertainty on ' F is reinstated by < reducing the maximum allowed power by 3% for each percent hf tilt in excess of
- 1. '
For purpose. of monitoring QUADRANT POWER TILT RATIO when one excore 1 detector is inoperable, the moveable incore detectors are used to confirm that j the normalized symmetric power distribution is consistent with the QUADRANT ; POWER TILT RATIO. The incore detector monitoring is done with a full _. incore i flux map or two sets of four symmetric _ thimbles. . The two sets of four ! symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-II, H 3, H-13, L-5, L-11, N-B. 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the ; parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent , with .the initial FSAR assumptions and have been analytically demonstrated ! a6 equate to maintain a minimum DNBR greater than the- design limit throughout each analyzed transient. .The indicated T,yg value of'591.l*F (four. loop t MILLSTONE - UNIT 3 8 3/4 2-5 Amendment No. 27, # - 0028 .
POWER DISTRIBUTION LIMITS BASES i i DNBPARAMETERS(Continued) ! operations) or 583.3*F (three loops operating) and the indicated pressurizer pressure value is 2218 psia (four loop' or three loop operation). The calculated values of the DNB related. parameters will be an average of the indicated values for the operable channels. The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation'. i Measurement uncertainties have been accounted for in determining the. parameter limits. < 1 i y e e i MILLSTONE - UNIT 3 B 3/4 2-6 Amendment No. JJ l 9on __-_2__--____
(- 1 i 3/4.4 REACTOR COOLANT SYSTEM BASES $ 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate in MODES 1 and 2 with three or four reactor coolant loops in operation and maintain DNBR greater than the design limit during all normal operations and anticipated transients. With less than ' the required reactor coolant loops in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours. ! In MODE 3, three reactor coolant loops, and in Mode 4, two reactor coolant loops provide sufficient heat removal capability for removing core , decay heat even in the event of a bank withdrawal accident; however, a single 4 reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers. - l l In MODE 4, and in MODE 5 with reactor coolant loops filled, a single , t reactor coolant loop or RHR loop provides sufficient heat removal capability - for removing decay heat; but sin le failure considerations require that at least two loops (either RHR or RCS be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of.the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE. The operation of one reactor coolant pump (RCP) or one RHR pump provides ' adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. The restrictions on startin an RCP with one or more RLS cold legs less than or equal to 350'F are provi ed to prevent RCS pressure transients, caused - . by energy additions from the Secondary Coolant System, which could exceed the ' limits of- Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by , either: (1) restricting the water volume in the pressurizer and thereby providing a volume for the reactor coolant to expand into, or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures. The requirement to maintain the isolated loop stop valves shut with power
- removed ensures that no reactivity addition to the core could occur due to the~
startup of an isolated loop. Verifl cation of the boron concentration in an idle loop prior to opening the stop valves provides a reassurance of the-adequacy of. the boron concentration in the isolated loop. The 2600 ppm 'is ' sufficient to bound shutdown-margin requirements' and provide for boron concentration measurement uncertainty between the loop and the RWST. Draining and refilling _ the isolated loop within 4 hours prior to opening its stop-valves ensures adequate mixing of the coolant in this loop and prevents' any
- reactivity effects due to boron concentration stratifications.
1 MILLSTONE - UNIT 3 8 3/4 4-1 Amendment Nw T cons -
3/4.9 REFUELINS OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that: (1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The value of 0.95 or less for K conservative allowance for uncertainties. Similarly, t#dI. boron includes a 1% Ak/k concentration value of 2600 ppm or greater includes a conservative uncertainty allowance of-50 ppm boron. The 2600 ppm provides for boron concentration reasurement uncertainty between the spent fuel pool and the RWST. The lockin closed of the required valves during refueling o>erations precludes the pos,sibility of uncontrolled boron dilution of the fi' led portion of the RCS. This action ! prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water. - 3/4.9.1.2 Boron Concentration in Spent Fuel Pool The limitations of this specification ensure that in the event of a fuel assembly handling accident involving either a misplaced or dropped fuel [ assembly, the K,ff of the spent fuel storage racks will remain less than or equal to .95. 3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monit/ ring ca> ability is available to detect changes in the reactivity condition of tie core. 3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission i i products. This decay time is consistent with th: assumptions used h the safety analyses. I 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment bui. ding penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to tae environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release i from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. 3/4.9.5 COMMUNICATIONS i The requirement for communications capability ensures that refueling ! station personnel can be promptly informed of significant' changes in the . , facility status or core reactivity conditions during CORE ALTERATIONS. i MILLSTONE - UNIT 3 B 3/4 9-1 I 0030 Amendment No. JJ
ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT
- 6.9.1.4 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.
A supplemental report containing dose assessments for the previous year shall be submitted annually within 90 days after January 1. The report shall include that information delineated in the REMODCM.- Any changes to the REMODCM shall be submitted in the Semiannual Radioac-tive Effluent Release Report. MONTHtY OPERATING REPORTS 6.9.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission - Cocument Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator Region I, and one copy to the NRC Resident Inspector, no later than the 15th of each month following the calendar month covered by the ' report. CORE OPERATING LIMITS REPORT 6.9.1.6.a Core operating limits shall be established and documented in the l CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of ? a reload cycle for the following: l 1
- 1. Moderator Temperature Coefficient BOL and EOL -limits and 300 ppm -
-i surveillance limit for Specification 3/4.1.1.3, -
- 2. Shutdown Rod Insertion limit for Specification 3/4.1.3.5,
- 3. Control Rod Insertion limits for Specification 3/4.1.3.6,
- 4. Axial Flux Difference Limits, target band, and APL ND for Specifica-tions 3/4.2.1.1 and 3/4.2.1.2, !
- 5. Heat Flux Hot Channel Factor, K(z), W(z), APLND ,
and W(z)gt. for Specifications 3/4.2.2.1 and 3/4.2.2.2. j
- 6. Nuclear Enthalpy Rise Hot Channel Factor, Power Factor Multiplier i for Specification 3/4.2.3. !
- A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station;-
however, for units with separate radwaste systems, the submittal shall , specify the releases of radioactive material from each unit' . I MILLSTONE - UNIT 3 6 21 Amendment No. # , 27; # 0031
1 i 6.9.1.6.b The analytical meP % used to determine the core operating limits shall be those previously r d - d and approved by the NRC in:
- 1. WCAP-9272 P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"
July 1985 (W Proprietary). (Methodology for Specifications 3.1.1.3 -- Moderator Temperature Coefficient, 3.1.3.5--Shutdown Bank Insertion Limit. 3.1.3.6 -Control Bank Insertion Limits, '3.2.1--Axial Flux Difference, 3.2.2--Heat Flux Hot Channel Factor, 3.2.3--Nuclear Enthalpy Rise Hot Channel Factor.)
- 2. WCAP-8385, " Power Distribution Control and Load Following Procedures - Topical Report," September 1981 (W Proprietary).
- 3. T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC),
January 31, 1980--
Attachment:
Operation and Safety-Analysis Aspects of an Improved Load Follow Package.
- 4. NUREG 800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear. Design, July 1981 Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC),
Revision 2, July 1981.
- 5. WCAP-10216 P-A,
" RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," June 1983 (W Proprietary).
(Methodology for Specifications 3.2.1--Axial Flux Difference [ Relaxed Axial Offset Control) and 3.2.2--Heat Flux Hot Channel Factor [W(z) surveillance requirements for FgMethodology).)
- 6. WCAP-9561-P-A, ADD. 3 Rev. 1, "BART A 1: '
A COMPUTER CODE FOR THE BEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS--SPECIAL REPORT: ' THIMBLE MODELING W ECCS EVALUATION MODEL," July 1986 (W Proprie-tary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel s Factor.) '
- 7. WCAP-10266-P-A, Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE," March 1987 (W Proprietary).
for Specification 3.2.2--Heat Flux Hot Channel Factor.) (Methodology
- 8. WCAP-11946, " Safety Evaluation Supporting r. More Negative EOL Moderator Temperature Coefficient Technical Specification for the 3 Millstone Nuclear Power Station Unit 3," September 1988 (W Proprie-tary). '
.)'
6.9.1.6.c The core operating limits shall be determined so that all applica- )j ble limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits,. nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. < 6.9.1.6.d The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions ! or supplements thereto, shall be provided upon issuance, for each reload - ' cycle, to the NRC Document Control Desk with copies to the Regional Adminis-trator and-Resident Inspector. ' l MILLSTONE - UNIT 3 6-21a con Amendment'No. )) y
-..L.__-_-_L_.----_----s
DEFINITIONS OVADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average. RADI0 ACTIVE WASTE TREATMENT SYSTEMS 1.25 RADI0 ACTIVE WASTE TREATMENT SYSTEMS are those liquid, gaseous and solid waste systems which are required to maintain control over radioactive material in order to meet the Limiting Conditions for Operation (LCOs) set forth in L these specifications. i RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REMODCM) 1.26 A RADIOLOGICAL EFFLUENT MONITORING MANUAL shall be a manual containing the site and environmental sampling and analysis programs for measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential . radiation exposures to in- j dividuals from station operation. An 0FFSITE DOSE CALCULATION MANUAL shall be : a manual containing the methodology and parameters to be used in the calcula- ; tion of offsite doses due to radioactive gaseous and liquid effluents and in 1 the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setponts. Requirements of the REM 00CM are provided in Specifica-tion 6.13. RATED THERMAL P0FB . 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt. REACTOR TRIP SYSTEM RESPONSE TIME t 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time -interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. REPORTABLE EVEN! 1,29 A REPORTABLE EVENT shall be any of those conditions specified in Section 50,73 of 10 CFR Part 50.
$1]UT00WN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactidty by which the reactor is subtritical or would be subcritical from itt preseht condition assuming all fulblength rod cluster assemblies (shutdown and control) are :
fully inserted except for the single rod cluster assembly of highest i reactivity worth whith is assumed to be fully withdrawn. plfLSTONE-tJNIT3~ 1-5
- 6. .
a i
r E O kD 3 : I h 3 250 - i H m . E : b
< 200 -
UNACCEPTABLE ; , OPERATION l 2 th : 8 - s 150 - 2 5
.a 8
U a : y 100 - - E - 0- : g - i y ACCEPTABLE : OPERATION . H 50 - 6 :
- i -
t b : 8 w 0 ''''''''''''''''''''''''''''''''''' 8 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER-FIGURE 3.4-l' DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL- POWER WITH THE REACTOR COOLANT SPECIFIC - ACT]VlY >1 uCt/ gram DOSE EQUIVALENT 1-131
' WILLSTONE - UNIT 3 3/4 4-30' J
MATERIAL PROPERTY- BASIS : CONTROLLING MATERIAL : PLATE METAL 1 COPPER CONTENT : CONSERVATIVELY ASSUMED TO BE 0.10 WT % PHOSPHORUS CONTENT : 0.010 WT % RTNDT INITIAL : BOT RTNDT t.FTER 10 EFPY : 1/4T,122T 3/4T.101T CURVE APPLICABLE FOR HEATUP RATES UP TO BOT /HR FOR TE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARGINS OF,10*F AND B0 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000.0 l l LEAK l TEST T t LIMIT 2000,0 ' a ,
- 1 5 l w
I
- , '{
o e y 1 O M .
$ 1000.0 HEATUP '
CURVE CRITICALITY LIMIT.
' BASED ON INSERVICE !
HYDROSTATIC ~ TEST I TEMPERATURE (2BB T) FOR THE SERVICE PERIOD UP TO 10 EFPY 0.0 - 0.0 100.0 200.0' 300.0 ( 400.0- 500.0 l INDICATES TEMPERATURE (DEG. F) FIGLRE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 10 EFPY WILLST0E - UNIT 3 3/4 4-34
. . _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ . _ _ _ . _ . . . _ _ _ . - _ ._. _ . _ _ _ _ _ . _ _ . _ ~ _ _ _ _ _ _ _ - _ _ _ _
i MATEPIAL PROPERTY BASIS , :
' CONTROLLING MATERIAL 8 PLATE METAL [
COPPER CONTENT : CONSERVATIVELY. ASSUMED 10 BE 0.10 WT % , l PHOSPHORUS CONTENT : 0.010 WT % RTNDT INITIAL : 60*F RTNDT AFTER 10 EFPY :1/4T,122T . 3/4T,101*F I i CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100T/HR FOR THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARG 1NS OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS l 3000.0
- l
$ 2000.0 2
/
Eb w p 1 W o. U y i
- 1000.0 COOLDOWN 7
~
RATES (*F/HR) 20 - 40 l ' 100 0.0 ! 0.0 100.0- 200.0 300.0 400.0: 500.0
?
' INDICATES' TEMPERATURE (DEG. F)
FIGURE 3.4-3 , REACTOR COOLANT _ SYSTEM COOLDOWN = LIMITATIONS ' APPLICABLE UP TO 10 EFPY 'l WILLSTONE - UNIT 3 - 3/4 4-35'
., ._ . , , _ , . _ , , , , , _ - _. - , _ + _ . , . . , -
s., 800 S E w E 2
} 700 ,
i h I s E 5 k 600
-l 3
.i i
a j i z 500 i D s E i 50 100 200 300 400 TRTD - AUCTONEERED LOW MEASURED RCS TEMPERATURE (*F) FIGURE 3.4-4o NOMINAL MAXIMUM ALLOWABLE PORV , SETPOINT FOR THE COLD DVERPRESSURE SYSTEM j (FOUR LOOP OPERATION) {
%dlLLS10E - UNIT- 3 3/4 4-40
)
I t l 1 800 2 5 i s t E O O 700 g , i 2 E 1 9
^
h 600
!= .
2 A E E 500 i U 2 e 50 100 200' 300 400 TRTD - AUCTONEERED LOW MEASURED RCS TEMPERATURE (*F). FIGURE 3.4-4b 1 NOMINAL MAXIMUM ALLOWABLE' PORV
' SETPOINT' FOR THE COLD OVERPRESSURE SYSTEM (THREE LOOP OPERATION)
G WILLST0E - UNIT' 3 -3/4 4-41
TABLE 3.3-4 ax 1P o r-ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION TRIP SETPOINTS
$ SENSOR E .- TOTAL ERROR c- FUNCTION"il UNIT ALLOWANCE (TA) I (S) TRIP SETPOINT ALLOWABLE VALUE i'i
- 1. -Safety Injection (Reactor Trip,
" Feedwater Isolation, Control Building Isolation (Nanual Initiation Only), Start Diesel Generators, and Service Water)
- a. Manual Initiation M.A. N.A. N.A. M.A. N.A.
b.= Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. R c. Containment Pressure--High 1 3.3 1.01 1.75 5 3.0 psig s 3.8 psig
. s.
Y d. Pressurizer Pressure--Low
% 22.16 20.1 . 1.5 1 1877.3 psig 1 1868.5 psig
- 1) Channels I and II 1 1863.3 psig 2)' Channel III and IV 22.16 15.6 3.3 1 1877.3 psig
- e. ' Steam Line Pressure--Low 17.7 15.6 2.2 2 658.6 psig* 1 648.3 psig*
- 2. Containment Spray (CDA)
- a. Manual Initiation- N.A. M.A. N.A. N.A. N. A.-
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- c. Containment Pressure--High-3 3.3 1.01: 1.75 5 8.0 psig s 8.8 psig.
- 3. IContainment Isolation
- a. . Phase "A" Isolation
- 1) Manual Initiation N.A. M.A. N.A. M.A.. N.A..
= .- -
.i i
i f i Docket No. 50 423 i B13627 ; i l
'i i
i
. . . 1 Attachment 3 1 Millstone Nucleer Power Station. Unit No. 3- !
Responses to Conditions Contained.in : NRC SER on WCAP 10444 ! I l l
- 1 l- !
i l l 1 i 1 t 1
-November.1990 r
'I e -. - . , - --'
U.S. Nuclear Regulatory Commission B13627/ Attachment 3/Page 1 November 1, 1990 Millstone Nuclear _ Power Station, Unit No. 3 Responses to Conditions Contained in NRC SER on WCAP-10444 , i l The NRC Staff reviewed WCAP-10444, " Reference Core Report VANTAGE SH Fuel Assembly " and concluded that the Westinghouse topical report was an accept-able reference to support plant-specific application of. VANTAGE 5 provided certain conditions were addressed. The conditions are addressed for .the Millstone Unit No. 3 application requesting the use of' VANTAGE SH fuel assem- { blies. These conditions are listed below with comments addressing- their j relevance to the VANTAGE SH fuel assemblies' upgrade licensing submittal , material in the report titled *The Plant Safety Evaluation for Millstone Unit' ! No. 3, VANTAGE 5H Fuel Upgrades, August 1990," and the Technical Specification changes (Attachments 1 and 2). I Condition 1 The statistical convolution method described in WCAP-10125 for the evaluation of initial fuel rod to nozzle growth gap has not been approved. This method should not be used in VANTAGE 5._ y i Resoonse : I Worst case fabrication tolerances and VANTAGE SH fuel rod and assembly growth ? are used to determine the initial fuel rod to nozzle growth gaps in .the l evaluation of fuel rod performance summarized in Section 2 of the Plant Safety Evaluation (PSE) as per Section M of WCAP 10125-P A. l Condition 2 For each plant application, it must be demonstrated that the LOCA/ seismic loads considered in WCAP-9401 bound the plant in question; otherwise addi- , tional analysis will be required to demonstrate the fuel assembly structural integrity.
Response
Millstone Unit No. 3 is not completely bounded by the LOCA/ seismic; loads considered in WCAP-9401. Thus, in accordance with Condition 2 of- the VANTAGE 5 NRC SER, additional plant-specific analyses have been performed to ! demonstrate fuel assembly structural integrity. An evaluation of 17 x 17 VANTAGE SH (with IFMs) fuel . assembly structural i integrity considering the lateral effects of seismic and LOCA accidents has been performed, in accordance with NRC requirements 'in Appendix A, Standard !l Review Plan (SRP) 4.2, the results show that the 17 x 17 VANTAGE SH is struc- j turally acceptable for an all VANTAGE 5H core and a transition core consisting i of both VANTAGE 5H and LOPAR fuel. The' grid load comparative: study results- ! i l I
U.S. Nuclear Regulatory Commission B13627/ Attachment 3/Page 2 November 1, 1990 i l show that the LOPAR assembly loads bound the transition core composed of both VANTAGE SH and LOPAR assemblies. The grids will not buckle due to combined impact forces of a seismic /LOCA event. The core coolable geometry is main- i tained. The stresses resulting from seismic and LOCA-induced deflections are within acceptable limits. The reactor can be shut down under the combined. i faulted condition loads, i Condition 3 An irradiation demonstration program should be performed to provide early. confirmation performance data for the VANTAGE 5 cesign. Resoonse The VANTAGE 5H fuel assembly design is :very similar to the VANTAGE 5 design , for which a demonstration program was performed to determine early performance ! data on the VANTAGE 5 fuel assembly design features. The VANTAGE 5 Demonstra-tion Program is described in Section 3.2.3 of Reference 2. i Condition 4 For those plants using the Improved Thermal Design Procedure (ITDP), the I restrictions enumerated in Section 4.1 of this report must be addressed and ; information regarding measurement uncertainties must be provided. l 4 Resportsa L Westinghouse has addressed (via Reference 3) the ' restrictions enumerated in 'j Section 4.1 of the VANTAGE 5 NRC~SER. The RTDP instrument uncertainty method- i ology used for Millstone Unit No. 3 with. VANTAGE 5H fuel is similar to that ; presented in WCAP-11656 (Reference 4) for ITDP. RTDP measurement uncertain-ties were provided to the NRC in Reference 5, Condition 5 i The WRB-2 correlation with a DNBR limit of 1.17 is acceptable for application to 17 x 17 VANTAGE 5 fuel. Additional data and analysis are . required when i applied to 14 x'14 or 15 x 15 fuel with an appropriate DNBR limit. The l applicability range of WRB-2 is specified in Section 4.2. i Resoonse i 17 x 17 VANTAGE SH fuel will be utilized at the Millstone Unit No. 3- plant. As described in Section 4 of Reference 1, .the WRB-2 correlation with a.DNBR i limit of 1.17 is used for the VANTAGE SH fuel with RTDP methodology. l
U.S. Nuclear Regulatory Commission B13627/ Attachment 3/Page 3 November 1, 1990 Condition 6 For 14 x 14 and 15 x 15 VANTAGE 5 fuel designs, separate analyses will be required to determine a transitional mixed core penalty.- The mixed core penalty and plant specific safety margin to compensate for the penalty should-be addressed in the plant Technical Specification bases. Resoonse - 17 x 17 VANTACE SH fuel will be utilized at Millstone Unit No. 3. As stated-in Section 4.6 of the PSE, " Transition cores are analyzed as if they were full cores of one assembly type (full STD or full VANTAGE SH), applying the appli-cable transition core penalty. For VANTAGE 5 fuel, penalties .are a function of the number of VANTAGE 5 fuel assemblies in' the core, as per Reference 6, which Sas been approved by the NRC. The same methodology. is used to calculate VANTAGE SH transition core penalties. The DNBR per.alty is less than 12.5% " l The transition core penalty is covere? by the margin maintained between the ! design and safety limit DNBR. This _ margin is addressed in the proposed i changes to Section 2.1.1 of the Millstone Unit No. 3 Technical Specifications, i Reactor Core Limit Bases. Maintenance of adequate DNBR margin to cover DNBR t penalties is confirmed on a cycle-specific basis ~ during _the reload safety- l evaluation process. 1 Condition 7 1 Plant-specific analysis should be performed to show that the DNBR limit will not be violated with the higher valve of FAH. i i
Response
The core DNB methodology as applied to Millstone Unit 3 with VANTAGE 5H fuel i is presented in Section 4 of the PSE. The PSE contains Millstone Unit No. 3- 1 specific analyses which support the use of an FAH of 1.70 (with appropriate !' treatment of uncertainties) during the transition period with a full core of VAhiAGE SH fuel. All safety criteria are met with an FAH of 1.70 at 100 per-cent rated power (3411 MWt reactor power). , [qndition 8
]
The plant-specific safety analysis for the steam system piping failure event should be performed with the assumption of loss of power if that is the most - conservative case. Response i 1 A specific _ safety evaluation for Major Secondary' Steam System Pipe _ Rupture was ! performed in support of the Millstone Unit No. 3 transition to VANTAGE SH i fuel. This evaluation is described in Section 5.1.3.5 of the PSE. . 1
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U.S. Nuclear Regulatory Commission B13627/ Attachment 3/Page 4 November 1, 1990 Condition 9-With regard to the RCS pump shaft seizure accident,'the' fuel failure-criterion should be 'the- 95/95 :DNBR limit. TheEmechanistic' method mentioned; in WCAP-10444- is not acceptable.
Response
The mechanistic method was not'used with regard to the RCS pump shaft seizure (locked rotor) accident addressed in' Section: 5.1.5.3 Lof- the PSE. Any irods which violated the 95/95 DNBR limit are assumed to fail.. Condition 10 If a positive MTC is intended for VANTAGE 5, the same positive MTC consistent with the plant Technical-Specifications should be used in the plant-specific safety analysis.
Response
In conjunction with the transition to VANTAGE SH fuel,:a positive MTCeis used for Millstone Unit No. 3 as described:in Section 1 of the PSE.- 'The supporting-- i Millstone Unit No. 3 safety analyses have utilizediMTC; assumptions- which ~ are 1 consistent with or conservative with respect to the current' Technical- Specifi - ' cations. Specific discussion of MTC- assumptions within -the_ VANTAGE 5H safety-analysis is presented in Secti.a 5.1.1.2..of the :PSE. All- non-LOCA accidents reanalyzed demonstrate that the appropriate safety criteria'are met.- Condition 11 The LOCA analysis performed for the reference plant with higher Fn of:2.55 has shown thu the PCT limit of 2200'F is' violated during a ;transttional mixed- , core. Plant-specific LOCA analysis imust be done to; show: that- with the. appro- - priate value of F , the 2200*F criterion can' be met:during use: of 'a transi-tional mixed core.0 1 Re:oonm, In accordance. Wn -Condition 11 of : the - VANTAGE :5 L NRC . SER, - Millstone. Unit ho. 3-speci% LOCA- analysis- for cores fueltd with VANTAGE SH and Standard assemblies were performed with consideration of transitional. core effects. The large--- and small-break J.t CA analyses are: presented in Section 5.2.1Eand-5.2.2 of the PSE. As descr' bed therein, the ECCS acceptance criterionL of' j' 2200*F is-; met for , Millstone Unit No.:3 with a LOCALF of-2 operation and 3.00 for three-loop ~ operation.' .The wors9-casepeak
.601for cladfear temper-loop ;
ature. is 2133*F for a four-loep 'LOCA .and 1874*F for the three-loop case. A L! conservative transition . core penalty :of 50'Ff has ~ been' appliedLto the- PCs ofo it the VANTAGE SH assembly, i o , 4 ]
U.S. Nuclear Regulatory Commission - B13627/ Attachment 3/Page 5 November 1, 1990 Condition 12 Our SER on Westinghouse's extended burnup' topical report:WCAP-10125 is not=yet complete; the approval 'of E the VANTAGE 5 -clesignz for operation to extended -
.i burnup levels is contingent on NRC- approval of- WCAP-10125. However, VANTAGE 5 i
fuel may be used. to. those burnups to which . Westinghouse fuel .is -presently j operating. Our review of. the Westinghouse extended burnup . topical report has ? 3 not ' identified any safety -issues with operation to the burnup. value given.-in H the extended.burnup report. Resoonse i WCAP-In125 has been approved (see Reference 1)., The extended burnup-methodol-ogy w.itained in: this topical-has beenlapplied and is considered in4Section- O of .ie PSE. Condition 13- ' Recently, a vibration problem has been re,'orted .inr a Frenchi reaC -J having i
.14-foot fuel assemblies: vibration: below the~ fuel assemblies in tir lewer- ~
position of the reactoi iessel is damaging the movab% incore' instrumentation probe. thimble. The staff is currently evaluating .the implications; of this "j problem to other cores.having 14-foot-long bundle assembDes. "AnyLlimitations. i to the.14-foot core design 1 resulting from = the-? staff n eoluation- must be ' addressed in plant-specific evaluations.. 1 Refo m te Millstone Unit No. 3 has foot-lorrg fuel ' assembly bundles and therefore the above condition is not applicable. However, NNEC0 has= established .a program - to monitor incore thimble tube -degradationt through -use of eddy-current test-ing. NNECO has implemented all NRC requireditesting:(IE: Bulletin 88-09) of. the thimble tubes. i 1 i i
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L L -l L- : l U.S. Nuclear' Regulatory Commission ! B13627/ Attachment 3/Page 6 ! l~ November 1, 1990 .i l. References l L 1-l 1. Davidson, S. L., (Ed.) - et- al., " Extended Burnup -~ Evaluation - of- l
' Westinghouse Fuel," WCAP-10125-P-A (Proprietary),- December 1985. -
- 2. Skaritka, J., '"Onerational _ Experience with Westinghouse Core,"' WCAP-8183, .l Revision.17, August 1989. ;
i
- 3. Westinghouse letter, E. P. Raheb Jr.,i to C. O. Thomas ~(NRC),;" Response.to ;
Request Number l_ for Additional Information on . WCAP-10444E entitled ' VANTAGE 5 Fuel Assembly" . (Proprietary), : NS-NRC-85-3014, dated' Harch 1,' ' 1985. ,
- 4. " Westinghouse:- Improved Thermai :DesignrProcedure LInstrumentIUncertaintyL ;
Methodology," WCAP-11656, December 1987. '
- 5. Letter from NEU to NRC, Transmittal of WCAP-12621, " Westinghouse'. Revised i Thermal Design Procedure. InstrumentL Uncertainty Methodology for Northeast Utilities,' Millstone Unit 3 Nuclear. Power Station,"EAugust:1990.
t
- 6. P. Schueren and Mr. K. R. McAtee, " Extension of Methodology for Calculat- f ing Transition Core:DNBR Penalties," WCAP-11837-F-A,cJanuary '1990..
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