ML19326B697
| ML19326B697 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 12/21/1977 |
| From: | ARKANSAS POWER & LIGHT CO., BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19326B682 | List: |
| References | |
| NUDOCS 8004170536 | |
| Download: ML19326B697 (160) | |
Text
{{#Wiki_filter:- _ - . . _ es Q INDEX m '[ s DEFINITIONS /^'r (,_/ SECTION PAGE 1.0 DEFINITIONS DEFINED TERMS.............................................. 1-1 THERMAL P0WER.............................................. 1-1 RATED THERMAL P0WER........................................ 1-1 OPERATIONAL M0DE........................................... 1-1 ACTI0N..................................................... 1-1 OPERABLE - OPERABILITY....... ............................. 1-1 s' REPORTABLE 0CCURENCE....................................... 1-2 Ag7ha-rent aWicnieur l j CCMTA'MM:MT INTEGRITY...................................... 1-2 l CHANNEL CALIBRATION........................................ 1-2 CHANNEL CHECK.............................................. 1-2 CHANNEL FUNCTIONAL TEST.................................... 1-3 l CORE ALTERATION............................................ 1-3 l MARGIN................. 1 SHUTDOWN ......................... 1-3 l 1 I DENT I F I ED L EA KAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3
.- UNIDENTIFIED LEAKAGE....................................... 1-4 l PRES SURE BOUNDARY LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 '
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QdADRANT POWER TILT........................................ 1-4 DOSE EQUIVALENT I-131...................................... 1-4 E-AVERAGE DISINTEGRATION ENERGY............................ 1-4 STAGGERED TEST BASIS....................................... 1-5 I FREQUENCY N0TATION......................................... 1-5 AXIAL POWER IMBALANCE...................................... 1-5
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1 f w PHYSICS TESTS.............................................. 1-6 OPERATIONALMODES(TABLE 1.1).............................. 1-7 FREQUENCYNOTATION(TABLE 1.2)............................. 1-8 1 8&W-STS 1 I June 1, 1976 l 6u041"l0 h
t INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS Reactor Core.............................................. 2-1 Reactor Coolant System Pressure........................... 2-1 ! 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Setpoints...................... 2-4 l r J BASES SECTION PAGE
-2.1 SAFETY LIMIT.S.
Re a c to r Co re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2-1 Reactor Coolant System Pressure........................... B 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Setpoints....................... B 2-4 v 9 ' B&W-STS II June 1, 1976
INDEX J ('s LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE
, j 3/4.0 APPLICABILITY........................................... 3/40-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL 1
Shutdown Margin...................................... 3/4 1-1 i Boron Dilution ...................................... 3/4 1-3 Moderator Temperature Coefficient..... .............. 3/4 1-4 Minimum Temperature for Criticality.................. 3/4 1-5 3/4.1.2 B0 RATION SYSTEMS
" ^'" M th: S h u t,de'."- .. . ... ..... ........ 3/4 10 .
Fl ow Pa t h s - Op e ra ti n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-7 u m . . ., on ., _ u ,, m o ,n. l e,r a Ma keu p Pump s - Opera ti ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-10 000 y "02 t 0:0'/;' Pump Shutd=; .. ........ ..... 3/ ' ' ' '
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1 m _ ,,, , 3., eousuum............................ .,,, . - Boric Ac id Pumps - Opera ti ng. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-13
cra ted 'e ter Scur :: Shutd=:n.................... 3/ ^ 1 l a-Bora ted Wa ter Sources - Opera ting. . . . . . . . . . . . . . . . . . . . 3/4 1-16 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height - Safety and Regulating Rod Groups...... 3/4 1-18 Group Height - Axial Power Shaping Rod Group. . . . . . . . . 3/4 1-21 Pos i tion Indica tor Channel s-0perating . . . . . . . . . . . . . . . 3/4 1-21 3c # tic" ?-di:2ter Ch:r- m Shutd::- 3/' '
'd Rod Drop Time........................................ 3/4 1-25 i Sa fe ty Rod In sertion L imi t. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-26 Regulating Rod Insertion Limits...................... 3/4 1-27 Rod Program.......................................... 3/4 1-31 Xenon Reactivity.....................................
3/4 1-34 B&W-STS III Januar.y 1,1977
INDEX i ra s. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
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U SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS - 3/4.2.1 AXIAL POWER IMBALANCE................................ 3/4 2-1 2/M ,2.? *!U";Ef o "E?' cLUY "0T C""NNCL Fi T0" F.. q ...... ............ . . 3/' 2 5 7 /M 9 7 etterat c A o ratTunt DV DTer
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E. 3 g . ... . . ... ., - 3/4.2.4 QUADRANT POWER TILT.................................. 3/4 2-9 3/'.2.5 D N S " * " "i : T : 3 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 2/d 2 !? 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION. . . . . . . . . . . . 3/4 3-1 O 3/4.3.2
) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION.................................... 3/4 3-9 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation................. 3/4 3-24 Incore Detectors..................................... 3/4 3-28 Se i smic In s trumen ta tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-30 "etec c!cgica! !artrumentatica 3/d ? 23 oe ete Shutder- !actre entatica. ? 'd ?-?"
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2/^ ? 29 Ch' e-4 ~ Detectier Syster:. 3/d 2 ^? 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT L00PS................................ 3/4 4-1 3/4.4.2 SAFETY VALVES - SHUTD0WN............................. 3/4 4-3 h 0 3/4.4.3 SAFETY VALVES - 0PERATING............................ 3/4 4-4 ( B&W-STS IV January 1, 1977
b INDEX
- LIMITING CONDITION FOR OPEE TION AND SURVEILLANCE REQUIREMENTS SECTION . PAGE 3/4.4.4 PRESSURIZER.......................................... 3/4 4-5 3/4.4.5 STEAM GENERATORS..................................... 3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Sys tems. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-13 Opera ti onal Lea kage. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-15 3/4.4.7 CHEMISTRY............................................ 3/4 4-17 3/4.4.8 SPECI FIC ACTIVITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-20 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Cool ant Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-24 Pressuri
- er.......................................... 3/4 4-28 3/4.4.10 STRUCTURAL INTEGRITY................................. 3/4 4-29 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 CORE FLOODING TANKS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T avg >-( N5)"F..................... 3/4 5-3 O
3/4.5.3 ECCS SUBSYSTEMS - T <V 7'G;)^F ..................... 3/4 5-6 avg 3/4.5.4 BORATED WATER STORAGE TANK. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-7 1 B&W-STS V January 1,1977 l l
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'ECTION 1.0 DEFINITIONS l
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3 1.G DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications. THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. RATED THERMAL POWER Sbealy Sba be-1.3 RATED THERMAL POWER shall be a to the reactor coolant ofM4SMWt. jjkhl- reactor core heat transfer rate OPERATIONAL MODE l '. 4 An OPERATIONAL MODE shall correspond to any one inclusive combina-tion of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1. ACTION 1.5 ACTION shall be those additional requirements specified as corollary statements to each principle specification and shall be part of the specifications. OPERABLE - OPERABILITY 1.6 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equip-ment, that are required for the system, subsystem, train, component or device to perform its function (s), are also capable of performing their related support function (s). l
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( l B&W-STS 1-1 June 1, 1976 l
p,, DEFINITIONS s O V REPORTABLE OCCURRENCE 1.7 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specifications 6.9.1 and 5.0.1.0.
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-00'""10"6mf4nkEGRITY IN 1.8 Ra,.&c c B4.id.ng/n /g e if y C0::T,",:::'t.':T !"TJ:T7 shall exist when:
- a. All penetrations required to be closed during accident con-ditions are either:
- 1. Capable of being closed by an OPERABLE ::nt:imentrec.cdoc buillm automatic isolation system, or
- 2. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table (.3.6-lj of Specification Wi e (3.6.-}. + l{.
- b. AM- equipment hatches-+/3ce closed and sealed.
l T4e pusona sl andweqestc,y /e:Jc a.r c.
- c. Cchcirlock%.0PERARLEpursuanttoSpecification(3.6.1.3f. l ree.cdoc- bua ldm Q d. The ::nt:frm:nt leak $ge rates are within the limits of v U Specification (3.6.1.2{.
- e. The sealing mechanism associated with each penetration (e.g. welds, bellows or 0-rings) is OPERABLE.
CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with necessary range and accuracy to known values of I.he parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and , alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. CHANNEL CALIBRATION may be performed by any series of sequential, over-lapping or total channel steps such that the entire channel is calibrated. CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of chamel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. B&W-STS 1-2 January 1, 1977
DEFINITIONS O b CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be:
- a. Analog channels - the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPEAABILITY including alarm and/or trip functions.
- b. Bistable channels - the injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions.
CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of anyAcom-h/orconbl ponert within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position. SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a. No change in axial power shaping rod position. I
- b. All control rod assemblies (safety and regulating) are fully '
insertedexceptforthesinglerodassemblyofhighestreactivity worth which is assumed to be fully withdrawn. IDENTIFIED LEAKAGE i 1.14 IDENTIFIED LEAKAGE shall be:
- a. Leakage (:=gt COT".0LL:0 LCA:'a0C) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank. I b.
redr- baaldmq Leakage into the :::t:f nnt atuo' sphere from sources that are both specifically located and known Ofther et te i nter'er:
'"' t'0 e?e tfer of 10:h::: detect *:
cy:t:= r not to be PRESSURE-BOUNDARY LEAKAGE. I i i B&W-STS 1-3 January 1, 1977
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- c. Reactor coolant system leakage through a steam generator to the secondary system.
UNIDENTIFIED Lc,,~ .E 1.15 UNIDENTIFIED LEAXAGE shall be all leakage which is not IDENTIFIED LEAXAGE r COTP.0'.' :: '.:."."" :. PRESSURE BOUNDARY LEAXAGE 1.16 PRESSUREBOUNDARYLEAKAGEshallbeleakage(exceptsteamgenerator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. 49NM9t+E+%L H,E
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QUADRANT POWER TILT 1.18 QUADRANT POWER TILT is defined by the following equation and is v expressed in percent. QUADRANT POWER TILT = Power in any core quadrant 100 (Average power ci' all quadrants ,)) DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (uCi/ gram) which alone would produce the same thyroid dose as the quantity and ' isetcaic mixture of I-131,1-132,1 133,1 134 and 1-135 actually present. The t1yroid dose conversion factors tied for this calculation shall be those listed in Table !!! of T!D-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." f - AVERAGE DISINTEGRATION ENERGY 1.20 T-AVERAGEDISINTEGRATIONENERGYshallbetheaverage(weightedin proportion to the concentration of each radionuclide in the reactor coolant at the time of samplir)g) of the sum of the average beta and gamma energiet B&W-STS 1-4 January 1,1977
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( '. _, DEFINITIONS As i l i per disintegration (in MeV) for isotopes, other than iodines, with half l lives greater than 15 minutes, makirg up at least 95% of the total non-iodine activity in the coolant. 1 STAGGERED TEST BASIS l 1.21 A STAGGERED TEST BASIS shall consist of:
- a. A test schedule for n systems, subsystems, trains or designated components obtained by dividing the specified test interval into n equal subintervals. I
- b. The testing of one system, subsystem, train or designated ;
components at the beginning of each subinterval. FREQUENCY NOTATION 1.22 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2. AXIAL POWER IMBALANCE i i 1.23 AXIAL POWER IMBALANCE shall be the THERMAL POWER in the top half l of the core expressed as a percentage of RATED THERMAL POWER minus the THERMAL POWER in the bottom half of the core expressed as a percentage of RATED THERMAL POWER. I l S"!ELO EL'!LO!':C '"TEC"!TY i . o. n. e.u. . --
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I PHYSICS TESTS 1.27 PHYSICS TESTS shall be those tests performed to measure the funda-mental nuclear characteristics of the reactor core and related instru-mentation and 1) described in Chapter . ) of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission. 13 t,B&W-STS l-6 June 1, 1976
(~ TABLE 1.1 (D OPERATIONAL MODES U 1 REACTIVITY % RATED AVERAGE COOLANT ! MODE CONDITION, K THERMAL POWER
- TEMPERATURE eff 3Fa
- 1. POWER OPERATION -> 0. 99 > 5% > (-3054 *F
- 2. STARTUP > 0.99 < 5% > )*F
- 3. HOT STANDBY < 0.99 0 > *F
- 4. HOT SHUTDOWN < 0.99 0 *F > T,yg > 200*F
- 5. COLD SHUTOOWN < 0.99 0 < 200*F
- 6. REFUELING ** < 0.95 0 < 140*F l
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) Excluding de. cay heat.
Reactor vessel head unbolted or removed and fuel in the vessel. O V . f'\- 0 B&W-STS 1-7 June 1, 1976
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(' _. TABLE 1.2 V FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours. O At least once per 24 hours. W At least once per 7 days. M At least once per 31 days. Q At least once per 92 days. SA At least once per 6 months. R At least once per 18 months. S/U Prior to each reactor startup. ' N.A. Not applicable. 9 v 0 1 'y/ B&W-STS 1-8 June 1, 1976 e
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O T i O I l I l SECTION 2.0 SAFETY LIMITS l AND LIMITING SAFETY SYSTEM SETTINGS V
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I {b '2 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE ' 2.1.1 The combination of the reactor coolant core outlet pressure and outle't temperature shall not exceed the safety limit shown in Figure 2.1-1. APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of reactor coolant core ! outlet pressure and outlet temperature has exceeded the safety limit, be in HOT STANDBY within one hour. REACTOR CORE 2.1.2 The combination of reactor THERMAL POWER and AXIAL POWER IMBALANCE shall not exceed the safety limit shown in Figure 2.1-2 for the various combinations of two, three and four reactor coolant pump operation. I
/N APPLICABILITY: MODE 1.
O ACTION: Whenever the point defined by the combination of Reactor Coolant System flow, AXIAL POWER IMBALANCE and THERMAL POWER has exceeded the appropriate safety limit, be in HOT STANDBY within one Hur. REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The Reactor Coolant System pressure shall not exceed 2750 psig.u>ke,is.
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APPLICABILITY: MODES 1, 2, 3, 4 and 5. ; i ACTION: MODES 1 and 2 - Whenever the Reactor Coolant System pressure has ex-ceeded 2750 psig, be in HOT STANDBY with the Reactor ! Coolant System pressure within its limit within one hour, f 6 , MODES 3, 4 - Whenever the Reactor Coolant System pressure has and 5 exceeded 2750 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. D B&W-STS 2-1 September 1, 1976 4 4
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$ UNACCEPTABLE OPERATION h -
18~00 f 1600 560 5B0 600 620 640 660 Reactor Outlet Temperature. 'F
', Figure 2.11 Reactor Core Safety Limit J
B&W-STS 2-2 June 1, 1976 i . .
THERMAL POWER 1.EVEL. ', UNACCEPTABEE OPERAil0N ( ) .
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CURVE i RE ACIOR COOL ANT FL OW (itym)
) J 374,880 2
280.035 3 184 del Figure 2.1 7 . i Reactor Core Safety Limit u i ryM35is 1
t! SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l 2.2 LIMITING SAFETY SYSTEM SETTINGS _ l REACTOR PROTECTION SYSTEM SETPOINTS 2.2.1 The Reactor Protection System instrumentatio's setpoints sha 1 . be set consistent with the Trip Setpoint values shown in Table 2 APPLICABILITY _: As shown for each channel in Table 3.3-1. i l ACTION _: t int lessf Table conserv- 2.2-1, With a Reactor Protection System instrumentation se po t tement ative than the the 'ialue shown inand theapply Allowable Values columns i declare channel inoperable tored the applicable to ACTION h r O equirement of Specification 3.3.1.1 until the chan Trip Setpoint value. O v . J B
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( \ TABLE 2.2-1 g! REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS f . g FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 1. Manual Reactor Trip Not Applicable Not Applicable
- 2. Nuclear Overpower 5 feF.2]% of' RATED THERMAL POWER 1 ilss.31% of RATED THERMAL POWER with four pumps operating with four pumps operating
< (/o(2)% of RATED THERMAL POWER < 6eE5l% of RATED THERMAL POWER with three pumps operating with three pumps operating 1 (fas,2)% of RATED THERMAL POWER with 5 5sts)% of RATED THERMAL POWER wi-one pump operating in each loop one pump ooerating in each loop ,
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- 3. RCS Outlet Temperature-High i f6 4).*F + 116/9f *F b4 ,
m 4. Nuclear Overpower . . ,. b5. _ r.... *
. . . . . Y Y .. _._ U._.__ ._+_ +__ ?.A___^_.
- a Based on RCS Flow and II) execed the lir,it lirc of . - the lim.i + " 1 AXIAL POWER IMBLANCE Fi;;urc 2 ' 1 **Ms red &nc cf das ci 4 ,mbel ure 2.2 2. i mones re&oo.n isne y*sh$nedsQ (nsex) as dehned gby yg,g,g,ance(s) g l
- 5. RCS Pressure-Low (I) T*(4000)psig* 83 b**O YF'3 22-1 1 {/p g psig &-M i 1sts
- 6. RCS Pressure-High 1 (2955-) psig N < lissal psig * *.
7. d* *' RCS Pressure-Variable Low (1) > ((10.25-) 3't* F - (-7006)) psig$;- Il T ' -fc outT > ((10. .h
,1 T out F - $ k j) psi I E tD 1
31
- - _ - - _ _ _=_-
q _
'k .
g TABLE 2.2-1 (Centinuid) z
& REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS d
FUNCTION UNIT TRIP SETPOINT ALLOWABLE VALUES
, 8. Nuclear Overpower '
)! cf o^TED THEP"at P0"ER --
( ): cf oaTED m EP"'1 P0tER based on Pump Monitors IjI Cit(h three pump:cpercting Gith tb-ee pumps crerati5
< lF4.I,5)% of RATED THERMAL POWER < (55.e)% of RATED THERMAL POWER with one pump operating in each loop with one pump operating in each 10 1 ( 0 )% of RATED THERMAL POWER with 1 ( O )% of RATED THERMAL POWER wi a
two pump operating in nne loop and two pumps operating in one loop an no pump operating in the other loop no pump operating in the other loo 1 ( 6 )% of RATED THERMAL POWER with 1 ( 0 )% of RATED THERMAL POWER wi no pumps operating or only one pump no pumps operating or only one pum operating operating
- 1. 8
- 9. Reactor Containment Vessel <-4-psig (/ t.5 ps/.) 1 (4,0) psig [gg.7ps#4.)
(1) Trip may be manually bypassed when RCS pressure 1 (1720) psig by actuating Shutdown Bypass provided that:
- a. The Nuclear Overpower Trip Setpoint is < 5% of RATED THERMAL POWER b.
The' Shutdown Bypass RCS Pressure - High Trip Setpoint of < (1720) psig is imposed, and
- c. 75 Shutd^"- Byp:ct it rc=cved "Scr oCE orc Curc ' (1300) p;ig.
? + See F q vre z,2--5 ; & # .5ee Rysr-e 2 2-4 o,
t t ( (
Il 120 100 00 - cc 5 2 4
?"
=
e-a W E o ' 40
,V I
s - 20 C 60 40 20 0 +20 W +M AXIAL POWER IMBALANCE % Figure 2.21 Trio Setpoint For Nuclear Overpower Based On RCS Flow and AXIAL POWER IMBALANCE v - h B&W-STS 2-7 January 1, 1977
.m..- . - - . _ , , ..-,,_._m. . . . , . . , - - . . . .
- . . . - . ~ .
( ,, THERWAL PCTER LEYE . ',
- 120 v
UNACCEPTABLE O-OPERAll0N (106 5) p 2 100 +, 4 '- A(C(PIABL! 4 %
/
PuwP OP(Raft 0m #
/j (79 5) gg ACC(PiaBL(
3& 4 PUNP , OP[R11104
-- 60 (52.3)
A C C ( P T A B L E 2.3 4 4 PUMP OPERATION (
-- 40 20 b :
il Il ll Il i I* l 1 m 1 60 40 20 0 20 40 60 P0wer imoalance. 5 - Figure 2.2 2 Allowable Value For Nuclear Overpower Based On RCS Flow and AXIAL POWER IMBALANCE J% Q BW-STS 2-8 January 1,1977 y
,- -,a-, e,- -en - -r--- r wwe-,-w , < - - - -
fO 4 A o L d t R 1-s 4 s d E , r , he.e Tor Ga t/s t Temp re ner e. , *f
- fig a re- 2.2-3 T ip Tetpoir,t for Protect &e fyryen ,
2-9
.(,, ;
.) [ i O
o 2500 - - P = M ' s , i. T = GIS'I t i '
= 2300 - - -= - --- - -
E I
. ACCEPTABLE
{ OPERATION . E m 2100 / - Z P = 11 75 Tout' 8 5103 psig a (4 5 1900 --
-=
'h _, UN ACCEP T A8L E OPERATION ( P = 1800 psig . 1700 --- - 1500 __ 500 580 600 620 640 660 e Reactor Outlet Temperature *F ARKANSAS POWER & LIGHT CO. PROTECTIVE SYSTEM FIG. NO.
,, ARKANSAS NUCLEAR ONE UNIT I ALLOWABLE V4tur 2.2-4 Q
\s r 2-/o
- - -- - 'v,'-m*w,
BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i Os , l 1 l L (
l { l l l 1 NOTE The sumary statements contained in this section provide the bases for the specifications of Section 2.0 and are not considered a part of the.ie technical specifications as provided in 10 CFR 50.36. l e
\
l L l
V* 2.1 ( SAFETY LIMITS A Q BASES 2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding 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 tne cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime would 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 therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through the IB&W-24-Si DNB correlation. The DNB correlation has been developed to predict the DNB hectf flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the /[
- U ratio of tne heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to {l.301 This value corresponds to a 1951 percent probability at a Mpercene 45 confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. The curve presented in Fi a minimum DNBR of I' .22/1.30)isgure 2.1-1 for predicted represents the conditions the maximum at which possible thermal y power (1121% when the reactor coolant flow is l'" ' 3 to45Io- ' O the design flow rate for four operating reactor coolant pumps."'" This '" wnsfE 7 curve is based on the following nuclear power peaking factors with putential fuel densification effects: F"={2.67T; N F3H = 0.78h = 0.50} g The design limit power peaking factors are the most restrictive ! calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawa1, and form the core DNBR design basis. v m lQ B&W-STS B 2-1 June 1,1976
( SAFETY LIMITS v b' BASES The reactor trip envelope appears to approach the safety limit more closely than it actually does because the reactor trip pressures are measured at a location where the indicated pressure is about (4-5) psi less than core outlet pressure, providing a more conservative margin to the l safety limit. The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densificationn.nd foi rod
- 1. bowig2 .:
The (1 ,1.30) DNBR limit produced by a nuclear power peaking factor of F = (2.677 or the combination of the radial peak, axial peak and position of the axial peak that yields na less than a (-1-G2fl.30) DNBR.
- 2. The combination of radial and axial peak that causes central fuel melting at the hot spot.
The limit is (/9.4-M-7.) kw/ft. Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking. v The specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively. The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in BASES Figure 2.1. The curves of BASES Figure 2.1 represent the conditions at which a minimum DNBR of (1.22/1.30J is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 122f4)%, whichever condition is more restrictive. Using a local quality limit of {22/M-)% at the point of minimum DNBR as a basis for curve 3 of BASES Figure 2.1 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR. . The DNBR as calculated by the (B&W-2/W-G-) DNB correlation continually increases from point of minimum DNBR, so that the exit DNBR is always higher. Extr:p:!:tfer of th: correl: tier beyend it pe5!!:hed qu:lity
-r:r;c cf (2??/!S)" it jutti' fed e- the b !: ef exper % rt:1 tta.
fD
\j
( B&W-STS B 2-2 September 1, 1976
(l
- p. SAFETY LIMITS V
BASES For each curve of BASES Figure 2.1, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than (-h4F1.30f or a local quality at the point of minimum DNBR less than (22/-4)% for that particular reactor coolant pump situation. The (4,43fl.Jul DNBR curve for four pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the left of the four pump curve will be above and to the left of the other curves. 2.1.3 REACTOR COOLANT SYSTEM PRESSURE l The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from re&ching the cont & = nt atmosphere. rear +ec k. %3 The reactor pressure vessel and pressurizer are designed to Section 9 III of the ASME Boiler and Pressure Vessel Code which permits a maximum transient pressure of 110%, 2750 psig, of design pressure. The Reactor (v Coolant System piping, valves and fittings, are designed tar ANSI B 31.7,
/9f,8 Edition, which permits a maximum transient pressure of 110%, 2750 psig, of component design pressure. The Safety Limit of 2750 psig is therefore consistent with the design criteria and associated code requirements.
. Ja.s The entire Reactor Coolant System-i+ hydrotested at 3125 psig,125%
of design pressure, to demonstrate integrity prior to initial operation. t B&W-STS B 2-3 September 1, 1976
2.2 Lill! TING SAFETY SYSTEli SETTINGS V BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETFOINTS The Reactor Protection System Instrumentation Trip Setpoint specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safMy limits. Operation with a trip setpoint less conservative than its Trip Setpoint but within its specified Allowable Value is accept-able on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses. The Shutdown Bypass provides for bypusing certain functions of the . Reactor Protection System in order to perm 1? control rod drive tests, ! zero power PHYSICS TESTS and certain startup and shutdown procedures. ' The purpose of the Shutdown Bypass RCS Pressure-High trip is to prevent normal operation with Shutdown Bypass activated. This high pressure trip setpoint is lower than the normal low pressure trip , setpoint so that the reactor must be tripped before the bypass I is initiated. The Nuclear Overpower Trip Setpoint of < 5.0% prevents l ) g any significant reactor power from being produced. SuTficient '
/ natural circulation would be available to remove 5.0% of RATED THERMAL l POWER if none of.the reactor coolant pumps were operating. -
l Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic Reactor Protection System instrumentation channels and provides manual i reactor trip capability. Nuclear Overpower A Nuclear Overpower trip at high power level (neutron flux) provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. During normal station operatign, reactor trip is initiated when the reactor power level reaches (105.51% of rated power. Due to calibration l and instrument errors, the maximum actual power at which a trip would be actuatedcouldbe1112{%,whichwasusedinthesafetyanalysis. s more conser-va.%s +Aan Hie valae l l B&W-STS . B 2-4 June 1, 1976 l ,
I (. *
,- ' LUllTU:G SAFETY 'iVSTEM .%TTINGS (v l' BASE 5 i
RCS Outlet Temperature - High 411 The RCS Outlet Temperature High trip < HM-)*F prevents the rear. tor outlet temperature from exceeding the design limits and acts as a backup trip for all power excursion transients, f!uclear Overpower Based on RCS Flow and AXIAL PCWER IMBALANCE The power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has been established to accomodate flow decreasing transients from high power where protection is not provided by the Nuclear Overpower Based on Pump Monitors channels. The power level trip setpoint produced by the power-to-flow ratio provides both high power level and. low flow protection in.the event the reactor power level increases or the reactor coolant flow rate decreases. The power level setpoint produced by the power-to-flow ratio provides overpower DNB protection for all modes of pump operation. For every ficw rate there is a maximum permissible power level, and for every power level ( there is a minimum pennissible low flow rate. Typical power level and low A flow rate combinations for the pump situations of Table 2.2-1 are as follows:
/06.5
- 1. Tripwouldoccur)whenfourreactorcoolantpumpsareoperating '
if power is (-lFJ4)% and reactor flow rate is 100%, or flow rate is (M-4)% and power level is 100%. J3 9
- 2. hen three reactor coolant pumps are operating Trip would is if power occurgw)%
(W and reactor flow rate is (74.7}1, or flow rate is (fer+)% and power is (75)%. l
- 70. F
- 3. Trip would occur when one reactor coolant pump is operat.'ng in each loop (total of two pumps operating) if the power is (M,4)% 52.3 and powerreactor flow rate is (#-e.2+h0-)% or flow rate is (Ro45-+)%
level is (49.0)%.
- 411QlSes For safetyFcalculations the maximum calibration and instrumentation errors for the power level were used.
l g V ( , B&W-STS B 2-5 June 1, 1976
LIMITIt.'G SAFETY SYSTEM SETTIPGS ~ BASES The AXII.L P0'c'IR !.MBALANCE boundaries are established in order to pre"ent. reactor thermal limits from being exceeded. These thermal limits are citMr power peaking kw/ft limits or DNBR limits. The AXIAL , POWER IHBALAr!C2 nduces the pcwer level trip produced by the flux-to- l flow rr. tin such that the boundaries of Figure 2.2-1 are produced. The flux-to-ficw ratio reduces the power level trip and associated reactor pcwer-reactor pc.cer-imbalance boundarics by (1.:0)% for a 1% flow reduction. F04S 1 RCS Pre.ssure - Lew, High and Variable low The High and Low trips are provided to limit the pressure range in which reactor operation is permitted. During a sicw reactivity insertion startup accident from low pcwer or a slow reactivity insertion from high pcwar, the RCS Pressure-High setpoint is reached before the Nuclear Overpower Trip Setpoint. The y gg3
, trip setpoint for RCS Pressure-High,122:5 Fpsig, has been established to maintain the systein pressure below the safety limit, {2750} psig, for (L.)
V any design transient. The RCS Pressure-High trip is backed up by the v pressurizer code safety valves for RCS over pressure protection, and is
~ therefore set lower than the set pressure for these valves, {25001 psig.
The RCS Pressure-High trip also backs up the Nuclear Overpower trip. SM9 1801 s/.75 The RCS Pressur Low, -{l000) psig, and RCS Pressure-Variable Low, ((10.20' T *F- DC L 3 psig, Trip Setpoints have been established to l maintain t@d DNB ratio greater than or equal to (-1,4W.l.307 for those design accidents that result in a pressure reduction. It also prevents reactor operation at pressures below the valid range of DNB correlation limits, protecting against DNB. Due to the calibration and instrumentation errors, the safety analysis used a RCS Pressure-Variable Low Trip Setpoint of [(15.25) //. 75 l T out I PSI9 ' Jtd3 Nuclear Ove'rpower Based on Pump Monitors In conjunction with the power / imbalance / flow trips the Nuclear Overpower' Based On Pump Monitors trip prevents the minimum core DNBR from decreasing below 4lw4#1.30f by tripping the rec: tor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power level for the number of pumps in operation. V B&W-STS B 2-6 September 1,1976
7 Y
- S diii g,ik
("_ LIMITING SAF_ETY SYSTEM SETTINGS . I J BASES _ { Bu n.:pixa Reactor Cor.tcirc. cat Vc :cl Pressure - High t gasso ma *8 The Reactor Centcir.r.cnt Vc;;ci Pressure-High Trip Setpoint < M psig, I _ m provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of-coolant accident, even in the absence of a RCS Pressure -Low trip. ; i emeus t R IEEER 6 p ::
- 1. -
W r b J%C hb 1 I 1 i B&W-STS B 2-7 June 1, 1976 i
- h M&.-l .. . .jy ~ * *. * ~"T.*
.. .. . y..1 :n-
w . O g V 1600 - .. .
- - - . -I -- -.
l i
; i !
s 2400 -- .. . ..., -._
! I i
d ; l . 1- 2200 1. ! f
/ :
1:
- i ' ) 1. i . I i l -y
{ l
! . i l l'
I l l 2000 - j ,
- c. I i ,
e
- I es 1 (. '
j i
/
- eno .. :
l
./ -
j . 1 I I i 4
! i 1600 - .-L - ' . - - - - - .
550 'b0
; t' t - U.; r,4 6 6GO
- s r . e , 0,.;. : e ni. ir - -r Cu"VE GPM #0*[R P;'WP S Ol'i 34 ' I N:; *
- T PI 0 F i I W ' I -
1 .114.880 (100.i- , til 'u R F !:9BR tlu.fi
? ?80 035 t 74 l'i B"i 7 IHRif F!!WPS ( P%R 'W!!'
1 IS4 441 (49 l e 59 0 0'.I Ptiui' r. ! ai;a . t e' :!'A ilt i Mii)
"'06 % OF tt sit,N stas BASES Figure 2.1 i
O V < B&W-STS B .:-d - June 1, 1976 w
('_ O( SECTIONS 3.0 AND 4.0 LIMITIt'G CONDITICNS FOR OPERATION AND SURVEILLANCE RECUIREMENTS l I
l 3/4 LIlllTIt!G CONDITIONS FOR OPERATI0fl AND SURVEILLANCE REQUIREMENTS { . 3/4.0 APPLICABILITY l LUlfTit!G CC"OlTIO!! FCR OPERATION 3.0.1 Limi.ing Conditions for Operation and ACTION requirements shall be applicable during the OPERATIONAL MODES or other conditions spacified for I each specification. 3.0.2 Adharence to the requireinents of the Limiting Condition for Opera-tion and/or associated ACTION within the specified time interval shall constitute compliance with the specification. In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval, completion of the ACTION statement is not required. 3.0.3 In the event a Limiting Condition for Operation and/or associated ACTION requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the facility shall be placed in at least HOT STAT!DBY within 1 hour and in COLD SHUTDOWN within the following 30 hours unless corrective measures are completed that permit operation under the permissible ACTION statements for the speci-fied time intarval as measured from initial discovery. Exceptions to these requiremants shall be stated in the individual specifications. 3.0.4 Entry into an OPERATIONAL MODE or other specified applicability condition shall not be made unless the conditions of the Limiting Con-dition for Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with l ACTION statements. SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be applicable during the OPERA-TIONAL MODES or other conditions specified for individual Limiting l Conditions for Operation unless otherwise stated in an individual Sur-veillance Requirement. 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:
- a. A maximum allowable extension not to exceed 25% of the surveil-lance interval, and
- b. -A total maximum combined interval time for any 3 consecutive tests not to exceed 3.25 times the specified surveillance interval.
A. B&W-STS 3/4 0-1 October 1, 1975
(T APPLICABILITY " SURVEILLANCE REOUIREMENTS (Continued) 4.0.3 Performance of a Surveillance Requirement within the specified time interval shall constitute compliance with OPERABILITY requirements for a Limiting Condition for Operation and associated Action statements unless otherwise required by the specification. Surveillance Requirements do not have to de performed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL MODE or other specified applicability conditions shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be applicab!: as follows: E
- 1. c- t'. '5 pericd 'rcr i::uan : cf the T ilitj Opd eting Licen - +c t' e :te-t c' #ecility cc creia' operetter. 4".:cr-
"4-e eh211 te.*47 cf ^.5"E Ccd Cl :: 1, 2 :nd 3 p aps and iel m
- b. per<~ - 4 <- , ::rd:n : .ith n;ti;n n ;; the A =
' hi!cr _ :d " c::u c '!:::cl Ccde ( + ) Editien, :nd ^.dd:nd:
t" cu? ( * ), except '^^r^ :peci'i '-ftten relief h:: b Cr
- ;;r:nted t; the Cc--f::icn v ;
- b. r e- the tiu pe-icd #c11c"4"; ste-t c# #ec414ty e t-cie' l
""a*eticr hservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be perfonned in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g),
except Commission where pursuantspecificto 10 written CFR 50, relief Sectionhas been granted 50.55a(g)(6)(i . by)the Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements. The provisions of Specification 4.0.2 are m applicable to surveillance intervais associated with inservice inspection and testing activities required by Section XI of the above ASME Boiler and Pressure Vessel Code and applicable Addenda. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed l to supersede the requirements of any Technical Specification.
%.mr-i ri, ema_ ra n 4,,
. . a n a a _., a , .. ., u ,.___,,,_2 ____ , .__. .. m
-- - - ., p . .w. . . . ., . . w w . w na vai
..o... . . . , , , .
B&W-STS 3/4 0-2 January 1, 1977
( 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL S!!UTD0WN MARGIN LIMITING COMDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > 1% ak/k. APPLICABILITY: fl0 DES 1, 2*, 3, 4 and 57 # I ACTION: With the SHUTDOWN MARGIN < 1% ak/k, imediately initiate and continue boration at > 25, gpm of (37ed ppm baron or its equivalent, until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be detennined to be > :% ak/k:
- a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable. If the inoperable control rod is imovable or untrippable, the above required SHUTDOWN MARGIN shall be increased.by an amount at least equal to the withdrawn worth of the imovable or untrippable control rod (s).
- b. #
' When in MODES 1 or 2 , at least once per 12 hours, by verifying that regulating rod grou I Specification 13.1.3.6I.ps withdrawal is within the limits of
- c. ##
When in MODE 2 , within 4 hours prior to achieving reactor criti- I cality by verifying that the predicted critical control rod position is within the limits of Specification {3.1.3.6$. ;
- d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading by~ consideration of the factors of e. below, with the regulating of Specification rod.g)roups at the maximum insertion limit
{3.1.3.6 ! With Keff >_l.0. With K,ff < l.0. See Special Test Exception 3.10.4. l
~
) B&W-STS 3/4 4-1 June 1,1976 l
+ *-
" ~ s n r ~ , s c.-
+o be 6fly,4 h.p,+ ,c a.. 6 -ud
([' PEACTIVITY CONTROL SYSTEMS -- SURVEILLANCE REQUIREf1ENTS (Continued) e. When in 140 DES 3, 4 or 5, at least once per 24 hours by consideration of the following factors: d /dt>f
- 1. Reactor coolant system boron concentration.
- 2. i Control rod position. I
- 3. Reactor coolant system average temperature.
- 4. i Fuel burnup based on gross thermal energy generation. i
- 5. Xenon concentration.
- 6. Samarium concentration. i 4.1.1.1.2 The overall core reactivity halance shall be compared to predicted values to demonstrate agreement within + 1% ak/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification J4.1.1.1.1.e}, above.
The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading. t B&W-STS 3/4 1-2 January 1, 1977 v
. 1 2
S b
~
REACTIVITY C0!! TROL SYSTEMS 80R0H OILUTION 7-] U LIllITING C0fl0lTION FOR OPERATION j l 1500 3.1.1.2 The flo rate of reactor coolant through the Reactor C 0at k ye) l
'c Sy:t=shallbe(>s2300)OpmwheneverareductioninReactorCoolant System boron concentration is being made. q APPLICABILITY- " ^
4f5, an/6 f ACTION: !
/Soo 6.sse,/ '
With the flow rate of reactor coolant through the Reactor C;- h t l Sy t = < s gpm, immediately suspend all operations involving a i reduction in boron concentration of the Reactor Coolant System. I f f SURVEILLANCE REQUIREMENTS 15cc V colant through the Reactor CeS5e}t 4.1.1.2 The flow rate of reacto : : r. Sy-':r shall be determined to be > _
) gpm within one hour prior to i the start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration'by either: l 4" ^"""'+4^"
- . '!0r# f;'h; it '"::t Cr :Ecter 00hnt "'_""" 4? !
-ee I
- b. Verifying that at least one DHR pump is in operation and supplying ~> (M 00) gpm through the Reactor ........ .,,- . I 15co Vess ej
'" d'C9 7 ' b ta mcde o{ p.lQ l
I i
$ 'l j B&W-STS 3/4 1-3 June 1, 1976 !
t n
O (~ REACTIVITY CONTROL SYSTEMS N0DERATOR TEMPERATURE COEFFICIENT l.!!ilTIf!G C0!!DITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:
- a. Less ositivethan{0.5$x10-4 Ak/k/*F whenever THERMAL POWER is < 5}%ofRATEDTHERMALPOWER. I
- b. LIff$$ $th:r(n.0} 'O'4 'k/':/"" whenever THERMAL POWER isEfESj%ofRATEDTHERMALPOWER. I
'm cpt 0:: th:r ( 0. 2) x 10-4 /' / P at Pf.:0 CP",'
" 0'.1E P .
ADPLICABILITY: MODES 1 and 2*# , ACTION: With the moderator temperature coefficient outside any of the above limits, be in at least HOT STANDBY within 6 hours. v SURVEILLANCE REQUIREMENTS 4.1.1.3.1 The MTC shall be determined to be within its limits by con-firmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits. 4.1.1.3.2 The MTC shall be determined at the following frequencies and THERMAL POWER Conditions during each fuel cycle.
- a. Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.
- 5. at nr " """L n0 HEP., "ith " ' ty: aft:r re::hin; ; ??.TED
'T!!EP."".L PC','EP, :;uilbriu: beren concentr:ti:n Of (200)- pp.. *With Keff >_ l.0. #See Special Test Exception 3.10.2.
9 B&W-STS 3/4 1-4 January 1, 1977
_ _..-._--_._.___.m l o l REACTIVITY CONTROL SYSTEMS
.(s-MINIMUM TEMPERATURE FOR CRITICALITY i
O LIMITIllG CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant System !=:: aver 9:.e. be>15251*F. 100; temperature (Tavg) s aH ; APPLICABILITY: MODES 1and2*f l' ACTION: avera.3e., '
<(525l'F, restore Withtoa within T Reactor Coolant its limit System within 15 minutes4eeportemperature be H (T}X9)OT STANDBY w i thE9next 15 minutes.
SURVEILLANCE REQUIREMENTS ' 4.1.1.4 r) The RCS temperature (T,yg). shall be determined to be > {5253 F:
- a. Within 15 minutes prior to achieving reactor criticality. l
- b. At least once per 30 minutes when the reactor is critical and '
'.l the Reactor Coolant System T avg is less than (69&)*F.
g :, 1 I With K,ff 1 1.0 . Dec Spec, / 7,~e3 / ( Ex'ceph,9 g,m 3 , t i 6I B&W-STS 3/41-5 January 1,1977 I
Q . d.~~g Q b,Q f Q __, .h. M . D.l u . , ;
~Yu.
o,,,,,,,,m, _m.. .m m mwv...... u v. .muu o,-. a cn..,a 3/0.1.2 00"AT!07; C'f TC;;"' 6 /d-FL^'J PAT :S - :l:UT"A , i LI".!T!!!C C0"0!TIO!? IO" 0?:",ATIOi *
- 3. 2.1 be At least one of the following boron injection flow paths sh RABLE:
I a. A flow path from the concentrated boric acid storage stem used ia a boric acid pump and a makeup or decay heat r al (DHR) p to the Reactor Coolant System, if only the bo c acid sth ge system in Specification (3.1.2.8a) is OP ABLE, or w b. A flow tn from the borated water storage ta via a makeup l or DHR p to the Reactor Coolant System if only the borated i water stor e tank in Specification (3.1. b) is OPERABLE. APPLICABILITY: MODES 5 d 6. gag J ACTION:
'.'! With none of the above flow pat OPERABLE suspend all operations M involving CORE ALTERATIONS or pos one injection path is restored to ive r ctivity changes until at least E status.
SURVEILLANCE REQUIREMENTS f FM 4.1.2.1 At least one of the strated OPERABLE: ove required ow paths shall be demon- l W1 a
- a. At least once r 7 days by verifying th the pipe tempera-
' ture of the at traced portion of the flo path is > (105)*F i
when system is a flo path from the concentrated boric cid storage sed. i 1 ,
- b. At le t once per 31 days by verifying that each lve (manual, ~
powe operated or automatic) in the flow path that s not lo ed, sealed or otherwise secured in position is in ts rrect position. P
/ u
; B&W-STS 3/4 1-6 January 1, 1977 3
f - .
- ///;;" : TJ r . 9..e f, .'l'i.;. 3 y :-
p I;iLO';! P.tilis - OPE.'.b C s U l'
!, . . i ~ s i_ : ._'s_ __. J
- H i %.4 _ 1.1 . .,.'.i?. . r 101 . .
O".l.2.2 J c.d .d the 7011]Mi.ig boron injection flow paths ,.ui! he
. lEETddl.E:
i vlddow & k
- a. .A /1 4 - gi yrw ;-h? - -- -
horic acid -- 9l v" t !: me ac 1 pumo and :nakev: or decay heat rea5 val (*.:HP. ) p t] th W.ter Coolant Sys ::m I
- l=
! ':. A f';w path 9.m '?c bora ted vata" sto-age tsnk via em'<euo or W pump to the Reactor Coolant System.
i. p 2PL:CA3P.!TY: i10CES 's, 2, 3 and 4.
' l' AC-'IO.'! :
A e iPtl 44 k h cn th? ficv . ath from the n _ : ;. boric acid : . . .. i
- - ~
inoparable, restore the inoperable ficu path to 0FER.'4Cf. sta us within 72 hours or be i>1 at least HOT STANO3Y - , -
.q , y.
.,...a.c.,. .v ..
[d \ ;l
'l within the next hours; restore the flow path to OPERABLE rtatas within the .1 ext 7 days or be in COLD SHUTDOWN within the next 30 hours,
!! f 2, i f
- l b. 5'ith th3 flow path from the borated water storage tank in-
;i .T?rable, restoro the flow path to OPERABLE status within one
!! oiar ar be in at least HOT STANDBY within the next 5 hours and -
jj in C0;.0 SHUTDOWN within the following 30 hours. I
,O a SUWIELLANCE REQUIREMENTS t
I
'li j ' 4.1.2.2 Each of the above required flow paths shall be demonstrated 0??RA8LE:
- a. At least once per 7 days by verifying that the pipe tempera-ture
" " '"* of"the -d heat traced portion of the flow p(ath from the boric acid cter:r: ty:t= is >~ 105YF.
add,& k g i ,4 al .
) , B&W-STS 3/4 1-7 January 1, 1977
E i 5 l V { CTIVITY C0_NTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) ' got orpsb/e. af aa /cmahc aWY
- b. At least once per 31 days by verifying tha ac.hoevoonyaisce each valve , an"21.guore/postbon y
"^':Or Optr.ted er tut;;; tic) in the flow path that is not ;
l locked, sealed, or otherwise secured in position, is in its i t ccrr :t position. , re.p. red
- c. p At least once per 18 months, during shutdown, by verifying that each auter- valve in the flow path actuates to its
**t'5
- =rt '
position)td : injectfen test signal. on a 5:ren a.edu]Q ES ~ 4 I v I 4 I I s 88W-STS 3/4 1-8 June 1, 1976
(. REACTIVITY CONTROL SYSTEMS MAKEUP PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two makeup pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4 . ACTION: With only one makeup pump OPERABLE, restore at least two makeup pumps to OPERABLE
- status within 72 hours or be in at least "0T ST""0SY cr.d h roted SH'JTDOL'" M?.9C!" egiv:1:nt t: '" 3.k/k :t 200*c H th1- the ::t S how Mr;rresture at least two makeup pumps to OPERABLE status within the nextJ7daysorbeinCOLDSHUTDOWNwithinthenext30 hours.
5Y E'N Tn c e a ,a , / h ,,, j.j, , Q g v SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two makeup pumps shall be demonstrated OPERABLE b-- ue-4#y4ag t% t ea ecirculatter '1ces, :05 p=p develep: : dit h:rg; p"eesure 1 act; when tested pursuant to Specification 4.0.5. With RCS pressure > ( ) psig. Boo -
\
I k B&W-STS 3/4 1-10 January 1, 1977
(
-. [) e. tt, 6.. : 0) .}L'fl Jy a tir 3f t'.C.t ,
REACTIVITf CONTROL SYSTEMS DECAY HEAT REMOVA PUMP - SHUTDOWN LIMITING CONDITION R OPERATION 3.1.2.5 At least one dec y heat removal HR pump in the boron injection flow path required by Spec fication (3.1 .1))or (3.1.2.2) shall be OPERABLE and capable cf bei g powered f m an OPERABLE emergency bus. APPLICABILITY: MODES 4*, 5* d 6. ACTION: With no DHR pump OPERABLE, suspen \all operations involving CORE ALTERATIONS or positive reactivity changes u ti at least one DHR pump is restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 4.1.2.5 At least the bove required decay hea removal pump shall be demonstrated Ol'ERABL by verifying, that on rec [culation flow, the pump develops a discharg pressure of > psig Shen tested pursuant to Specification 4.0. .
\
\
\'I e
- RCS Pressure < ( ) psig. -
i I l I l l s B&W-STS 3/4 1-11 January 1, 1977
c, 1 -o c . REACTIVITY CONTROL SYSTEMS v BORIC ACIO PUMP - SHUTDOWN LIMITING CONDITION FOR OP RATION 3.1.2.6 At least one boric acid pump shall be OPERABLE and pable of being powered from an OPERAB E emergency bus if only the f1 path through the boric acid pump Specification (3.1.2.la) is OPERABLE. APPLICABILITY: MODES 5 and 6. ACTION: With no boric acid pump OPERABLE as required to complete the flow path l of Specification (3.1.2.la), sus nd all operatidns involving CORE ALTERATIONS or positive reactivit changes untti at least one boric acid pump is restored to OPERABLE statu . i SURVEILLANCE REQUIREMENTS / 1
/
4.1.2.6 At least the ab ve required boric acid pump shall be demonstrated OPERABLE by verifying, at on recirculati flow, the pump develops a discharge pressure of _ psig when tes ed pursuant to Specification 4.0.5. O l k) B&W-STS 3/4 1-12 January 1,1977
i m REACTIVITY CONTROL SYSTEMS BORIC ACID PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.7 At least one boric acid pum i required by Specification (3.1.2.2a}pshall in the boron injection be OPERABLE flow path sad capd!c Of . haia; p~cOr^d #r^ r OPEPf.BLE :: rg y but if the flow path through i the boric acid pump in Specification 3'1.2.2alisOPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ! ACTION: With no boric acid pump OPERABLE, restore at least one boric acid pump l to OPERABLE status within 72 hours or be in at least HOT STANDBY-and-Scr:ted to u SML'T00', '".".0!'i :quhalaat to 1% ok/k et 200"i within the next hours; restore at least one boric acid pump to OPERABLE status withi{nthenext7daysorbeinCOLDSHUTDOWNwithinthenext30 hours. 1 IL t e [ SURVEILLANCE REQUIREMENTS
& \ed pably Hw : l
'}
4.1.2.7 At least the bove required boric acid pump shall be demonstrated OPERABLE by verif ect cuhtf r '!=., th: j. y q=;; pru=ying, ;e_t'st e- g;is hu tu ta p.......t ; :.r;0,,d:;;: h;; ; I . T% p ,, wo f, , k,4 / mu, g w g, ; ,co, j
% 4 om hd Co < ls m . ds . '
[,Wl$om o n Ye cs(ca fa % hfbn , bkt pump veldp3 t 4 d'5h e. pre 3w re o f(g,)t 7 u7 , o - s4 . B&W-STS 3/4 1-13 January 1, 1977 e i
D-e Q ' Q. t
' \
REACTIVITY CONTROL SYSTEMS
- BORATED WATER S CES - SHUTDOWN LIMITING CONDITION R OPERATION s ,
3.1.2.8 As a minimum, ne of the following borated water so es shall be OPERABLE:
- a. A concentrated oric acid storage system and as ociated heat tracing with:
- 1. A minimum con ined borated water volyme of ,
gallons. I
/
- 2. Between and ppm of boron. I
- 3. A minimum solution temperature o (105)*F. I
- b. The borated water storage tank (BW )with:
- 1. A minimum contained be ated Gater volume of (25,000) gallons.
O 2. A minimum boron concentra ion of (1800) ppm. ' p I v
- 3. A minimum solution tempera re of (35) F.
APPLICABILITY: MODES 5 and 6. ACTION: With no borated water sources PERABLE, suspend 11 operations involving CORE ALTERATION or positive eactivity changes u il at least one borated water source is restored to PERABLE status. SURVEILLANCE REQUIREMEN s 4.1.2.8 The above r quired borated water source shall demonstrated ' OPERABLE:
- a. At lea once per 7 days by:
- 1. erifying the boron concentration of the wate . I
- 2. Verifying the contained borated water volume o the tank. I 8&W-STS 3/4 1-14 January 1,1977
\
M l I REACTIVITY CONTROL SYSTEMS ' SURVEILLANCE REQUIREMENTS (Continued) ___ h
- 3. ng the concentrated boric aci rage system N solution ature when it i e source of borated water. t
- b. At least once hours by verifyin BWST temperature when i e source of borated water and the empera ture is < 35,F. ide) g
;..ryg)
. . . iw u
. O!1 l
W pumi D e i , W
, . m 1
B&W-STS 3/4 1-15 June 1, 1976 I w.-
. . . . . , . ..... m.z g :&.--
( REACTIVITY CONTROL SYSTEMS s BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.9 Each of The fo' lowing borated water sour'ce/ shall be OPEP.ABLE: a N.h o r %h
- a. The cencontr:ted boric acid stora; syst~" and associated heat tracing with:
- 1. M'
^ -i-i kw Me. e : -t egwd elt.n =r:ted l- water 't. .u :lge / of47;:1).:=. is e. es I
C P 2 7c 0 pp,9l Or. cr1.
- 2. S:t.:=r =d pp: cf b rcr,. l
/c
- 3. A mini, mum solution temperature of (4%)*F. chow. S.
Gry:. r a it... h. .. . f emp .woru.r-s ,
-b. The Ser:ted w:ter :ter:ge t:r (B"ST) dth:
3 a __ .., _; t__ m_; . _ _ . . _ , . _ _ _,<_m..___ __; w w i ww a s uww wve uwww ru u w w s v v s uriiw w6 ww b rvw w s s u'au m,m 11 a n e- . . . . . . . l 1
- 2. ": t .;= r. =d ppr cf her^" I v -4'ai l
?. 8 u celut*:r ter.per:tur: ef (25) e v APPLICABILITY: MODES 1, 2, 3 and 4. i ACTION:
rdddain husk
- a. With the =n=ntr;t d boric acid :t =; cyst ^r inoperable, l restore the storaga system to OPERABLE status within 72 hours !
or be in at least HOT STANDBY rrd 5: = t^d t : !"'.'TC^" "." "C U' I
- -d;;1=t t 1% ah/h :t 200"" within the next-6-IfoW /?- '
restore the =n=ntr:::d boric acid ten;: ry *- to OPERABLE status within the next 7 days or b in COLD SHUTDOWN within the next 30 hours. <te, hofi 4,1k
- b. "ith ==e+a-a +ha m_m the m_ m5m r -nted
. . - r:ter :terage
_m... .me,_
, ts-' _iaaa- ebte. . m ,___.
w u s s rs V3 e ,k.Vuin o 6.1 ___ < . . . _ IO .ww TT4 bat 8su Vl aw s tVu l ass ub awudW um,. u M Tu, n b , J BEE b u J t _ . u..w. __ a.. e. u. . ,., u ,. _ - . -a 4.3.cni n cunTnnuu s.-, ,..-v. v. w..... .. .. .
... .u,_ .t. , _ , , _ . ,__ en u_..__
. ... . . . . ... ..v...., .. . . . . . .
B&W-STS 3/4 1-16 January 1, 1977
1 r-REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS T/w. 4.1.2.9 Ge + borated water source shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
'. '.'e. i fy ir.g the here.. cer.cer.t. otie. ir. ;;;h set:r ear;c. I
/e.va. /.s
- 2. Verifying the :--t:i d rited water '/ ! :/:e M
- * ~- ' ~ ~ " l-h e. b ce o <. <Lc. , d c, d t'. / ic , ha ,m . I
- 3. ad/s.ho,, juosk Verifying the nr.:;r.tcuted boric acid eter:; ':y t:r solution temperature.
_t .,_________._.,.t_..__ t. 1,. . : _ _
... . . . ...- ,... m .~..
. u.
. - -. .- .ouer
(::tt'de) 2 - te ;-m+"-- ter th 4 4e - "*F. Hr. bo <^cn nom.endmL c.,_ c f /-4, b o r,c_ cd C,ocldo
%re, Sr..o[y s,, pl2 ork er Ge, a c) iniske up by/ 11c tnare v H reg o" .1Hy &1 ene.:. pa.c dq,. .
~
I v B&W-STS I . 3/4 1-17 January 1, 1977 I l l I
REACTIVITY CONTROL SYSTEMS - m. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT - SAFETY AND REGULATING R0D GROUPS LIMITING CONDITION FOR OPERATIONS 3.1.3.1 positionedAllwithin control (safety
+ 0. and regulating))
L indicated position ofrods their shall groupbeaverage OPERABLE and height. f.osn e/ms (e.5*4 APPLICABILITY: MODES 1* and 2*. ACTION: 11cre Man cne. I n NW d ' N l*** C *1 P S'C" '
- a. With ana ne ma-a control rod / inoperableddue to being innovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within one hour and be in at least HOT STANDBY within 6 hours, m oh e r
- b. With more than one control rod inoperab or misaligned from O .
its grou average height by more than + (dndicated position , be in at least HOT STANDBY within 6 hours.' k S N "
- c. With one control rod inoperable due to causes other than addressed in ACTION a, above, or misaligned from its group average height by more than + (indicated position), POWER OPERATION may continue provi one hour either:
0:ssheshatwithin}6.G*/n
- 1. The control rod is restored to OPERABLE status within the above alignment requirements, or
- 2. The control rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.
POWER OPERATION may then continue provided that:
- ) " rc =h:ti:n ;f :: h :::ident :::ly:': Of T:ble 3.1 1 i: ; rf;n. d .;t th'- 5 d:y:; thi: r:: :':: tier th:!'
-:-"- th t th pr ;i; :13 :::ly::d rc;:lt: Of th:::
- ! dent: --i t v:' id fa- th- de-e+4aa as aa=*=+4aa ceder th::: :: d!ti:n:.
b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours. i l
*5ee Special Test _ Exceptions 3.10.1 and 3.10.2.
lB&W-STS 3/4 1-18 January 1,1977
_ . _ . . .- - - ~ ~ - I REACTIVITY CONTROL SYSTEMS ACTION: (Continued) c) A m- peacr di;tributi;n :e3p i: Obt:ined
'r:r the 'n:Cr:
m _ _ > - _a ,__ ,__4, w .. u m4+w4.
'" ~ ~~
!C77?':4;T. m'49.677 4atC Z ~
' ~ '
4 i d) Either the THERMAL POWER level is reduced to < $60T% l of the THERMAL POWER allowable for the reactor
-d 'ith'-
coolantpumpcombinationwithinonehougipSc'petnt-th; c:at ? h;.c; thc S cice. 0.c.pcacc .c i: ccduccd te ; (70)% ef the Tll:R".AL IO" R oliveoLic
';r the r;;;t ? :.01:nt p=p cc 'i tir , or e) The remainder of the rods in the group with the f 7.0'o ok d inoperable rod are aligned to within +Hr.+t ofthe ( (;. 5 %
inoperable rod within one hour while iiiaintaining the Indoe fe) rod sequence, insertion and overlap limits of Figures 3.1-1, 3.1-2, 3.1-3, 3.1-4 and 3.1-5; the Pn,fa,g i/ THERMAL POWER level shall be restricted pursuant to Specification 13.1.3.7} during subsequent operation. I SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each control rod shall be determined to be within the group average height limit by verifying the individual rod positior.s
+b at least once per 12 hours 4w: pt Sr'n; ti : ' n ter":1: ';n .
e a c etric oducr$tg7 e $ 4 g7373g g, the ,,g74fy egg 73 $ m 4 ge7 4 cd petit *cn: t 'e t ence pc- A ke m 4.1. 3.1. 2 Each control rod not fully inserted shall be determined to be OPERABLE by movement of at least (+}% in any one direction' at least once every 31 days.
- f. 5 %[2,0,nch e, )
p . G - ll B&W-STS 3/4 1-19 January 1, 1977
/
4
@,, E ' ti. 0
( 4 j' TABLE 3.1-1 ACCIDENT ANA YSES REQUIRING REEVALUATION i, THE EVENT OF AN INOPERABLE CONTROL R0D/ Control Rod A sembly Insertion Ch racteristics Control Rod Ass bly Misalignm t loss Of Reactor Co lant From mall Ruptured Pipes Or From Crack In Lar e Pipes Which Actuates The Emergen Cor Cooling System Single Control Rod Ass ly Withdrawal At Full Power Major Reactor Coola Syst m Pipe Ruptures (Loss Of Coolant Acciden Major Secondary ystem Pipe R ture I Rupture of a ntrol Rod Drive chanism Housing
) (Control Rod Assembly Ejection) "
/
&W-STS v 3/4 1-20 January 1, 1977
s [p REACTIVITY CONTROL SYSTEMS GROUP HEIGHT - AXIAL POWER SHAPING R00 GROUP LIMITING CONDITION FOR OPERATION , 3.1.3.2 All axial power shaping rods (APSR) shall be OPERABLE, unless fully withdrawn, and shall be positioned within + fr 33 (indicated position) of their group average height. }. 0 ,4 3 6 . 5 04, APPLICABILITY: MODES 1* and 2*. ACTION: 9.0 :11 cl, e.:. , With a maximum of one APSkinoperable or misaligned from its group 1 average height by more than +4r-!E (dndicated position), operation may I continue provided that withiii 2 hours: s,5 % l 1
- a. The APSR group is positioned such that the misaligned rod is restored to within limits for the group average height, or darm 6A ce
- b. It is determined that th imbalance limits of Specification 3.2.1 are satisfied and[ movement of the APSRslgroup es "p:
g L vented while the rod remains inoperable or misaligned.
'd e fe,a,,,se l f/,,,{ /f , ,,,,j,,j o, u /,n ,/5 e( s,an,M, con SLI we ssh s fie d SURVEILLANCE REQUIREMENTS 4.1.3.2.1 The position of each APSR r.e4 shall be determined to be within the group average height limit by verifying the individual rod positions t--
.e- n . . - _. - at least
- e... onc,e
.. n.sper 12 hours.e u..as.. s cep+ +M~;
s.....,u,. tMc. .*-tervm
+u.. . u. . su. a.n
, s ). 1 . *L2 _ L 1 m .. _ _
,A
..-7......... .....-_.....ro.-
4.1.3.2.2 Unless all APSR are fully withdrawn, each APSR shall be determined to be OPERABLE by moving the individual rod at least 44-)% at least once every 31 days. /. J.,0 ,oc/,c)
~/'c. & y, k.
*See Special Test Exceptions 3.10.1 and 3.10.2.
B&W-STS 3/4 1-21 January 1, 1977 l
('
*~s REACTIVITY CONTROL SYSTEMS w
POSITION INDICATOR CHANNELS - OPERATING LIMITING CONDITION FOR OPERATION Ee.c lu 3.1.3.3 *4&& safety, regulating and axial ggwer shaping coptrol rod absela[c
- d c'.? itch position indicat!O _h r !: !I_ pult cttppi f position' r claf l/g, indicat!?'Eh:rn:1: shall be OPERABLE.:nd :p:ble of deter ' 'n- th:
- ntr:1 c;d p;;iti n: it- * (2)".. -
APPLICABIL7TY: - MODES 1 and 2. ACTION: A embolros udk bolh the absolaf t. scn
- a. "it' 2 r' Of en: erd re t ' position indicat:r ch: n:1 cuid fhe i
r"alafiv e p control red ;rcup or one pul:: Stoppi9 position indicatec-c':r 21 per :^ntr:1 r:d ;r;;; inoperable either: 'CG
"*-^'""""a' .-..~.--m", T a'
._ 1 _.- ^ ^ " " . 'e m-~ s 'm o , ".. v '. .m-
. '.".m"""..'m'
.. ". ^m 'm" " . , . ' . ' . . -
- S!: 'er the a:::ter cc:T:nt m,.._,___
vuw awun n _ _ _ _ _ ,. e_ ";:p :::t,,in:ti:n :nd redu:: t': __,_. ._ ,s. _, .<_ -,,,,o., vv4. yv vv w . a & .y s u w yv s .. w kw ' va wa ss a a i n. se. p [)' _DAUr0 s 11 a._.s k i n Em. 6kn .nseen. , a n T", g &i wn j .a,m n enmh4ns&4an s s mJsE4.-o'E;,,;;', _r '"' - ~ ' ' --' ' ~ ~ ~ r -- r - - -- - t - D e.+e. ,n.n <. +I,e c
.f,
, m 9
- 2. ,,Oper: tier : 3 ;r.y.l a nu; k:p- be.
. .s
.mm.
pr.mbl<. % van 6 on3 +l at '- 1 ss ,- g3 y- a) Thepositionof$1 control rod with the inoperable u position indica % s verified within 8 hours by {. 2 actuating its 0%, 25%, 50%, 75% or,100% position 3 a, reference indicatee. I ren E *s .
- F G b) Thecontrolrodggggp(s)containingtheinoperable position indicat_.
E -y- _"?-- ' is subsequently maintained us at the 0%, 25%, 50%, 75% or,100% withdrawn position 3 3 and verified at this position at least once per 12 ve e
-n hours thereafter. l
' +
';s -
c) ^ :--t':- *: th'- th: -it: -f Sp;;ifi :ti .. g s,. D A_ uf+w mm._ +w_o n_ e- ,o_ n o_ i. e_ o e_+ mm4-m mm_,4+4_- 4 . 2 4. . _- _.-. . - ' . . . . - . . . ' . per:ble , Oper: tion ir. "'0E01 :r.d 2 sej : ..i;no. Tv, 4 up 6v
, T ' : v. . . p.e :d:d':11 ;f th: rc:d n: itch p::iti:r indi::t:r ch: mel: :r: 0 """a"L".
v(-3 B&W-STS 3/4 1-22 January 1,1977 l
(~. REACTIVITY CONTROL SYSTEMS (%% POSITION INDICATOR CHANNELS (Continued) SURVEILLANCE REQUIREMENTS
....... e. -. u. _-..
.; . ..a..-u.
... ,.. e. +. m ., m,.,4 ,. ,, mme_4+4mm 4
,u,.,+m..
,.w,mmoi es 3, tc dct;caincd to t: 0" ".^."Lt by scrifyin; that the pm^ steppi .; petitie-4"Ain ter char- m ':nd the reed c'czitch petitic" 4"dicater 92"" & 2;r00
..a
. 6.4- m,.,
g
~ .. . 3. ., .. . . - -
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( Do!cfe.: ( REACTIVITY CONTROL SYSTEMS ' ( POSITION INDICATOR CHANNELS - SHUTDOWN ' LIMITING CONDITI FOR OPERATION
/
3.1.3.4 At least one ed switch rod position indi tor channel shall be OPERABLE for each sa ty, regulating or axial p er shaping rod not fully inserted. APPLICABILITY: MODES 3*, 4* nd 5*. ACTION: With less than the above required ositio indicator channel (s) OPERABLE, immediately open the Control Rod Dr ve T p Breakers. SURVEILLANCE REQUIREMENTS
/ \
4.1.3.4 Each of the above re ired rod positio indicator channel (s) shall . be determined to be OPERABLE y performance of a HANNEL FUNCTIONAL TEST at least once each 18 month . l
- With the Control d Drive Trip Breakers in the closed sition.
t a 88W-STS 3/4 1-24 January 1, 1977 l
o ( O( Q REACTIVITY CONTROL SYSTEMS ROD OROP TIME LIMITING CONDITION FOR OPERATION i< 25 % wl/s</cw,s {t o 4 ! 'in3o.4ek 3.1.3.5 The individual safety and regulating rod drop time from the I fully withdrawn position shall be < (I.44 seconds from powe- interruption rea-<h; control rod drive breakers T: ( } *ncertier ( " ; citic") with:n.// at the r~ cco/qn f ps'sa/M atx.Tt l n <3 c s- gg. /. so 5;c,sd:, g,,. n o 2- I I avg - -
#kte c,c,3d,6c,u, E ^
- t r :: !:nt pum;; 0;;r;tir.g.
APPLICABILITY: MODES 1 and 2. ACT*0N:
- a. With the drop time of any safety or regulating rod determined ,
to exceed the above limit, restore the rod drop time to within l the above limit prior to proceeding to MODE 1 and 2.
- 5. "ith th; r;d dr;; timc: withir. 1*-it: but d;t; m'n;d ith i m
)' t': d
- te :::1:nt pur; 0;cr: ting, 0; r: tier ;; pr;;;;d pr;;ided that EP"al o0ugo 4: .ogtggetog te ion +w3 m.
equ:1 t the ' Eou^L onugo e s i mm. wi g <m. +w . .-+-- -mmi,-+
'" ; -^ '4 std:r ; riti ; it the t h: Of r;d dr:; the measurement.
SURVEILLANCE REQUIREMENTS 4.1.3.5 The rod drop time of safety and regulating rods shall be demon- I strated through measurement prior to reactor criticality:
- a. For all rods following each removal of the reactor vessel head. I
- b. For specifically affected individual rods following any main-tenance on or modification to the control rod drive system which could affect the drop time of those specific rods. ;
- c. At least once every 18 months.
l D
- lO B&W-STS 3/4 1-25 January 1, 1977 s ~
\
p REACTIVITY CONTROL SYSTEMS SAFETY R0D INSERTION LIMIT LIMITING CONDITION FOR OPERATION
/
3.1.3.6 All safety rods shall be fully withdrawn. 1 Aie da, APPLICABILITY:fl*and2*#. ACTION: With a msximum of one safety rod not fully withdrawn, except for sur-veillance hour either: testing pursuant to Specification f4.1.3.1.2), within one
- a. Fully withdraw the rod, or l
- b. Declare the rod to be inoperable and apply Specification 13.1.3.11 bg v
SURVEILLANCE REQUIREMENTS 4.1.3.6 Each safety rod shall be determined to be fully withdrawn: I a. Within 15 minutes prior to withdrawal of any regulating rod during an approach to reactor criticality.
- b. At least once per 12 hours. I i
*See Special Test Exception 3.10.1 and 3.10.2.
#With K,ff >_ l.0.
ke B&W-STS 3/4 1-26 January 1, 1977 I E
? %
REACTIVITY CONTROL SYSTEMS O V REGULATING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.7 The regulating rod groups shall be limited in physical inser-tion as shewn on Figures 13.1-13, {3.1-27,13.1-37, {3.1-4} .aad- $3.1-6) a.n d 3. / -6 with a rod group overlap of 25 + 5% between sequential with' drawn groups (5 and 6, and 6 and 7) (5 x d S/7). APPLICABILITY: MODES 1* and 2*#. ACTION: With the regulating rod groups inserted beyond the above insertion limits, or with any group sequence or overlap outside the specified limits, except for surveillance testing pursuant to Specification {4.1.3.1.2f, either:
- a. Restore the regulating groups to within the limits within-e-p hours, or
- b. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figures within-e-hours, or
- c. Be in at least HOT STANDBY withn 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.7 The position of each regulating group shall be determined to be I t within the insertion, sequence and overlap limits at least once every l 12 hours except when:
- . The c;u': tim; d 'r::rtier 3 4 4 t 2':- !: i noper:blc, tha
- ri fy the group: to 5: .itS4- the 4.ccrtier '* 't: :tlcnt ence pe A 'cu c.
- b. The control rod drive sequence alarm is inoperable, then verify the groups to be within the sequence and overlap limits ct least once per 4 hours.
*See Special Test-Exceptions 3.10.1 and 3.10.2
#With Keff " 1.0.
f3
~) B&W-STS 3/4 1-27 January 1, 1977 l
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(
% REACTIVITY CONTROL SYSTEMS R0D PROGRAM i l
i LIMITING CONDITION FOR OPERATION ! l 3.1.3.8 Each control rod (safety, regulating and APSR) shall be pro- I gramed to operate in the core position and rod group specified in Figure (2.' C,' m (3.1-7). l APPLICABILITY: MODES 1* and 2*. ACTION: , With any control rod not programed to operate as specified above, be in HOT STANDBY within 4 hour. 3 l SURVEILLANCE REQUIREMENTS I 4.1.3.8 , I
- a. Each control rod shall be demonstrated to be programed to !
operate in the specified core pcsition and rod group by: 1 l
- 1. Selection and actuation from the control room and verifi-cation of movement of the proper rod as indicated by both the absolute and relative position indicators:
a .: For all control rods, after the control rod drive i patches are locked subsequent to test, reprograming l or maintenance within the panels. ' b) For specifically affected individual rods, following maintenance, test, reconnection or modification of power or instrumentation cables from the control rod drive control system to the control rod drive.
- 2. Verir 9 that each cable that has been disconnected has been properly matched and reconnected to the specified control rod drive.
- b. At least once each 7 days, verify that the control rod drive patch panels are 19cked.
O b *See Special Test Exceptions 3.10.1 and 3.10.2. l B&W-STS January 1,1977 3/4 1-33 1
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- ~ :-
REACTIVITY CONTROL SYSTEMS XENON REACTIVITY LIMITING CONDITION FOR OPERATION 3.1.3.9 THERMAL POWER shall not be increased above the power level cutoff I specified in Figures 3.1-1, 3.1-2, and 3.1-3 unless # the following conditione is satisfied:
- a. Xenon reactivity is within 10 percent of the equilibrium value for RATED THERMAL POWER and it approaching stability,-eg=
S. '"EP"*L " CUES h:: 5 :n ..ithin : . ng; cf ( ) t: ( ) pc. :nt i e' "^TED :EP".aL POUCP f r : ;;ried ;;;;; ding 2 h:07; in th;
-::luble pri:Cr : .-tr:1 mcde, en !;di ; ::::r 'r:: :t:rt _p:.
APPLICABILITY: MODE 1. ACTION: f-'S With the requirements of the above specification not satisfied, educe (j THERMAL POWER to less than or equal to the power level cutoff within 15 minutes. v SURVEILLANCE REQUIREMENTS 4.1.3.9 Xenon reactivity shall be determined to be within 10f. of the l equilibrium __ . _u.,, value for RATED THERMAL POWER and to be approaching stability u- 2-.--_,--2 .u.. .u vurouni onuro u., u--- : , u .--- -,
-[ tA { ') 5 5f l'T:0 i 55.5L E5WC:[' . 5I.hi,-'3',priortoincreasIng' THERMAL POWER above the power level cutoff.
s O V B&W-STS 3/4 1-36 January 1, 1977 '
~
~.
3/4.2 POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on Figures 3.2-1, 3.2-2 and 3.2-3. I APPLICABILITY: MODE 1 above 40% of RATED THERMAL POWER.* ACTION: With AXIAL POWER IMBALANCE exceeding the limits specified above, either:
- a. Restore the AXIAL POWER IMEALANCE to within its limits within 15 m' = t::, or A h co s
- b. -30 'r :t 1:20t 90T ST*"0S Within 2 Scr:-
Redu.ee yea.cfav puse uful Axtr PW EManLAud
$tmrh A.rs me.f.
SURVEILLANCE REQUIREMENTS 4.2.1 The AXIAL POWER IMBALANCE shall be detennined to be within limits at least once every 12 hours when above 40% of RATED THERMAL POWER.eneept i :r = "'!"L
^via'_ on"Eo"Oh'E". !M",g Iuontau,g 7 2aLa"C: = niter 7g. i; i=per: tic, tha =1=1;t: th; See Special Test Exception 3.10.1.
Ol-L B&W-STS 3/4 2-1 January 1,1977 l
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(w POWER DISTRIBUTION LIMITS tt y NUCLEAR HEAT P(.UX HOT CHANNEL FACTOR - F a T 4.5 g , w g4 u 3 EeQtJe.SksMbsoh On h I-hss v. Cc 'ered, b y ' LIMITING CONDITI FOR OPERATION #m hnlGxe 1i mi bs . 3.2.2 Fqshall be 1 ited by the following relationships: I ) F q1 p f P > 0.5 F q12( ) for 1 0.5 [ where P = RAT H OWER and P 1 1.0. APPLICABILITY: MODE 1. l ACTION: With Fq exceeding its limit: a. Reduce THERMAL POWER at leas 1% fo each 1% F exceeds the limit within 15 minutes and similar r uce the Nuc9 ear Overpower Trip Setpoint and Nuclear Overpower ed on RCS Flow and AXIAL POWER i IMBALANCE Trip Setpoint within hours.
- b. Demonstrate through in-core m ppi that F is within its limit within 24 hours after exceed ng th limit 9rreduceTHERMAL POWER to less than 5% of R ED THE L POWER within the next 2 hours.
- c. Identify and correct th cause of the o t of limit condition prior to increasing TH IMAL POWER above he reduced limit re-quired by a or b, ab e; subsequent POWE OPERATION may proceed provided that Fnis emonstrated through 1 -core mapping to be within its limit av a nominal 50% of RATED ERMAL POWER prior l to exceeding this JHERMAL POWER, at a nomina 75% of RATED THERMAL POWER pr'or to exceeding this THERMAL OWER and within 24 hours after ttaining 95% or greater RATED ERMAL. POWER.
SURVEILLANCE REQUIREME S s 4.2.2.1 Fn shall b determined to be within its limit by using he incore detectors to obta a power distribution map: B&W-STS 3/4 2-5 January ,1977 l
( POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) hemme
- a. Pr o initial operation above 75 percent of R THERMAL POWER a each fuel loading. l
- b. At least once per M ective F ower Days.
- c. The provisions of Spe ation .4 are not applicable.
4.2.2.2 The me ed F of 4.2.2.1 above, shall be increa account fo (3)% to nufacturibg tolerances and further increased by k.hy acco or m4asurement uncertainty.
,g t
homme t' w dh " C wy g ,. - b i i t t l
Am .
1 e' B&W-STS 3/4 2-6 January 1,1977 l e - - -
. - . ...w
- -- yy n.
- . . . _==---.:_:------=----------.;------------------:=.
i: - POWER DISTRPOUTION LIMITS g gec Pc
, NUCLEARENTHAqYRISEHOTCHANNELFACTOR-FAH 3 _
d \ $p.V e [ eus [.q Pe ta. W it ec k 4 fW thetb Fh w w >e,x d {y
'. LIMITING CONDITI FOR OPERATION 88h ba {4 nu // m ik
,{ 3.2.3 F H shall be imited by the following relationship:
F H 1( ) [1 + 0.2(1-P)] T ERMAL POWER where P = RATED THERMAL POWER i and P 1 1.0. l APPLICABILITY: MODE 1.
$, ACTION: , With F exceeding its limit: -
H O a. Reduce THERMAL POWER at t 1% for each 1% that F" exceeds the
' limit within 15 minutes a similarly reduce the NubYear Overpower . Trip Setpoint and Nucle erpower based on RCS Flow and AXIAL l i POWER IMBALANCE Trip S poi t within 4 hours. ; b. Demonstrate through n-core m ping that F N is within its limit within 24 hours af er exceedin the limit h reduce THERMAL POWER to less th 5% of RATED HERMAL POWER within the next 2 hours.
- c. Identi#y cnd orrect the cause of he out of limit condition prior to i reasing THERMAL POWER a ove the reduced limit required aorb,aboge;subseque POWER OPERATION may proceed ovided that F is demonst ted through in-core mappin tobewithinit$Hlimit at a n inal 50% of RATED THERM POWER prior to exceeding this ERMAL POWER, at a nomi al 75% of RATED THERMAL POWER prio to exceeding this TH AL POWER and within 24 hours after ttaining 95% or i eater RATED THERMAL POWER.
k O
'w) .
i B&W-STS 3/4 2-7 January 1, 1977 l
. . - . . - - .n -.- -- . . .
m
\
'r0WER DISTRIBUTION LIMITS e' .
s l I ( SURVEILLANCE REQUIREMENTS ' in-. 4.2.3.1 all be determined to be within its limit by us the incoredet8ctor obtain a power distribution map: guun
- a. Prior to opera above 75 percent ATED THERMAL POWER after each fuel loa I
- b. At least once per 31 ctive 1 Power Days.
- c. The provisi of Specification 4.0.4 are applicable.
N summa 4.2.3. e measured F N of 4.2.3.1 above, shall be increased by f asurementuncertaNty. l I l I W A i l l l Ml i l puuss i I I i as l l B&W-STS 3/4 2-8 January 1, 1977 l l _ f
.% -p E.
- L. =
- h ~
_. m - - - _ , _ = - - - -- - - - - - - -------------- . . -- _-. k. POWER DISTRIBUTION LIMITS QUADRANT POWER TILT LD3 TING CONDITION FOR OPERATION 3.2.4 THE QUADRANT POWER TILT shall not exceed the Steady State Limit of Table 3.2-1. - APPLICABILITY: MODE 1 above 15% of RATED THERMAL POWER.* ACTION:
- a. With the QUADRANT POWER TILT determined to exceed the Steady State Limit but less than or equal to the .Tene4ed. Limit of Table 3.2-1: Magemum
- 1. Within 2 hours:
a) Either reduce the QUADRANT POWER TILT to within its Steady State Limit, or fo belous fhe peeper levtl cdaNene b) Reduce THERMAL POWER 0 so as not to exceed THERMAL l
, POWER.
- c'e" ; ; r: r ':=' =t:"f allowable for 4 the reactor coolant pump combination less at least 2% '
for each 1% of QUADRANT POWER TILT in excess of the Steady State Limit and withinf4 hours, reouce the "r' :r Ox rpr.cr T '; *-tpe'-t r ; 'M Nuclear Mand
.0verpower Based on RCS Flow and AXIAL POWER IMSALANCE Trip Setpoint at least 2% for each 1% of QUADRANT POWER TILT in excess of the Steady State Limit. l CS .L.. .k- Af f AnD Auf
-. U...- .a f t. g. . .... . . . . . 3..._. Ont.len TTI 1"_._. .._. 4e m4.k.4. 4.e c...Ju e . . .. i 4 4.----_...m464. o n. u..... . . . _ .- .. ., -,... . .. 24.. .u. e . . 2..
e.... i4 :. __ .__;..__ vuenu . onuen ... . ...
.. .._. _ _,___ _ . . . , . . . .en.. _, ,_ ,
_ ruenumi -. nnuco
.c_ ____m__
___, _. mu.___
^"* * - - t i ^ - " ' " d -
n..... _ v.a. e_ th: :::t 2 aeh:;- _- _>_. ._ n ... r ,d.. nd;;; . . - - th; it.;'.u r
., ...r..... .. - . . . . , . .. ......~ ,v. n
. .- .. . ... . .. ..w.. ... ....
.u.___.,u_..__ - - - - - . M.
- 2. ::t; rd 0:77;.;t th; = =:__,__ _ >__
,........m.......,
,.u
___>__ 0'
. nnuen.
".: 00t_..e_ ^'_'.init - ;di
. _ _ . aaue-
. __,_ . . . , _ m,_ --..,--... . . - - . . >
or ,..on . .. vu, v. . . . . . . . .. n .........u.. .. 1 "aaa+a= "aa'==* " ' - - - 1 M ' :'* .y prx =d ;=;'d;d t';t l
.L. nuannaur v.~. .
anuen vn_v. 4.
.. _.. .. ... 4 ,4.2 m4.u.,- 4., e . . 2., 1 e.... i4_4. .. i...... ... .._ .
-___ -. . . _ _ .._.. _______<_.._,__s.ot___ __ m, r.. . . . . . . . . . . . . . . ..
n--tet a . ...+.u, . . . . . sev .._ ,.
..- .. _ ._.m-. .m.,-- .. . .
. . . - - . . . . -~.--
- n. m rv n.
, *See Special Test Exception 3.10.1.
i B&W-ST:: 3/4 2-9 January 1,1977 l l ___
j i~ -
~.::s .* .. . .: . .- . .. ~.>NNa M AIL u
_ POWER DISTRIBUTION LIMITS sLIMITING CONDITION FOR OPERATION (Continued) ient With the QUADRANT POWER TILT determined to exceed the Tra l Limit but less than the Maximum Limit of Table 3.2-1, d to i misalignment of either a safety, regulating or axial p er ! aping rod: I l. Reduce THERMAL POWER at least 2% for each la' of indicated ADRANT POWER TILT in excess of the Stead State Limit w in 30 minutes.
- 2. Verif that the QUADRANT POWER TILT i within its Transient l ggg Limit w hin 2 hours after exceedin the Transient Limit i or reduce HERMAL POWER to less th 60% of THERMAL POWER l allowable the reactor coolan pump combination within the next 2 h rs and reduce th uclear Overpower Trip !
Setpoint to -
.5% of THERMA POWER allowable for the reactor coolant mp combin g
O 3. Identify and correc ion ~within the next 4 hours. tion prior to increas the THERMALause of the out of limit condi-POWER; subsequent POWER j OPERATION above 60% 4 ERMAL POWER allowable for the
"""' reactor coolant pum comb ation may proceed provided that the QUADRANT POWE TILT is erified within its Steady State Limit at 1 ast once pe hour for 12 hours or until verified accep ble at 95% or eater RATED THERMAL ,
POWER. g c. With the QUADR T POWER TILT determined Limit but le exceed the Transient 7y than the Maximum Limit of T le 3.2-1, due to
~
causes oth than the misalignment of either safety, regulat-f ' ing or a 1 power shaping rod:
' l.
duce THERMAL POWER to less than 60% of THE L POWER llowable for the reactor coolant pump combina on within 2 hours and reduce the Nuclear Overpower Trip Se oint g to < 65.5% of THERMAL POWER allowable for the reac r cooTant pump combination witnin the next 4 hours.
- 2. Identify and correct the cause of the out of limit con 5
""! tion prior to increasing THERMAL POWER; subsequent POWER
. OPERATION above 60% of THERMAL POWER allowable for the reactor coolant pump combination may proceed provided that the QUADRANT POWER TILT is verified within its Steady State Limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER. #
gl B&W-STS 3/4 2-10 January 1,1977 I
4p ( k POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) d. With the QUADRANT POWER TILT determined to exceed the Maximum Limit of Table 3.2-1, reduce THERMAL POWER to < 15% of RATED
~~
THERMAL PCWER within 2 hours. SURVEILLANCE REQUIREMENTS 4.2.4 The QUADRANT POWER TILT shall be determined to be within the limits at least once every 7 days during operation above 15% of RATED THERMAL POWER. eycept "her the QUADo^"7 '0MEP 7!LT erite- it 4 epe 251 , ther th: heums, QUADPa"7 "0ME' 7!LT :h:!' 5: :21: elated :t le::t :: ; r !? (~')\
%w l
o o 88W-STS 3/4 2-11 January 1,1977
9 4
? .
TABLE 3.2-1 v QUADRANT POWER TILT LIMITS STEADY STATE -T" ":::"T MAXIMUM LIMIT L:"!' LIMIT s ....._-... ,.a.a .a..+ o QUADRAATP6dERTil.T ~ (+:96) -(11.^7; (M
/%f f A n n A LIT Yve v __
n n,,em_., s t= f's
.... ~
u_.. . . . .. a w . . ci - -+=4, 1 fnenwa inmene&ne C u e + nra ,, n,3 to e33 ren ni Daum e. Osnan (*hannale l,o nAi
/7 nA4 s....,
/en ni o, . , _ _ _ - _ n_ __.__ e.,_._ ra e,s
. . . . . . . . . . . . .... ,. . . - - . . . _ , - - - . . _r,- oa os t, ,n ni
, r.., -- ,
s ( , v I i l i I i l i I i 1 1 e i
). .
L B&W-STS 3/4 2-12 January 1,1977 1
r 6 ION =
, POWER DISTRI TION LIMITS LIMITING CONDITIwN FOR OPERATION 3.2.5 The followi DNB related parameters shall be main ined within the limits shown on Table 3.2-2:
- a. Reactor Coo ant Hot Leg Temperature.
- b. Reactor Cool t Pressure.
- c. Reactor Coolan Flow Rate.
APPLICABILITY: MODE 1. ACTION: With any of the above parame rs exceedi t's limit, restore the para-meter to within its limit wit in 2 hour or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within e next 4 hours. SURVEILLANCE REQUIREF.CNTS z , 4.2.5.1 Each of the para
\
ters of Ta le 3.2-1 shall be verified to be within their limits at 1 st once per 2 hours. 4.2.5.2 The Reactor oolant System toca flow rate shall be determined to be within its 11 it by measurement at east once per 18 months. B&W-STS 3/4 2-13 January 1,1977
n '.
.A I 'h E Z_ E W E TABLE 3.2-2 j &
4 d DNB MARGIN j t LIMITS
~
i.. I Four Reactor Three Reactor One Reactor I.f;- f Coolant Pumps Coolant Pumps Coolant Pump Operating
.1 Parameter Operating Operating in Each Loop React oolant Hot Leg Temperat T *F <
_.(605.2) -< (605.2)III < .2) i
,y <
{,,f Reactor Coolant Pres psig(2) > (2062.7) > (2058.9)II) > (2092.5) w 2 Reactor Coolant Flow Rate, gpm > (369,600)
> (276, > (181,843)
Y
.i } ,
- F. -
- c
,'4 UI Applicable to the 100 h 2 Reactor Coolant Pumps Operating.
l' c ( Limit not icable during either a THERMAL POWER ramp increase in excess of 5% of k $ RATED L POWER per minute or a THERMAL POWER step incr. ease of greater than 10%
,E o TED THERMAL POWER.
w 2 I 1, a
~
El N k yy 33 . , . .
W i i(% , 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CON)lTION FOR OPERATION 3.3.1.1 As a minimum, the Reactor Protection System instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE."i*h oce " cc l TIMES :: :h wr. in T;bic :. -2. l APPLICABILITY: As shown in Table 3.3-1. l ACTION: As shown in Table 3.3-1. I SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each Reactor Protection System instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations di. ing the MODES and at the frequencies shown in Table 4.3-1. 4.3.1.1.2 The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.
" 3.1.1.2 The ROCT0" "".0TECT!0" SYSTEM RES"0"SE TIME ^f 22:F ':::t0r trip function :h ll be dcren:trat:d to be "ith*- it: "-i t a t '-? : t ^a"^
nar 1Q mnnthe Emek +ne+ ehm11 i ncl e e An 2+ 1ame+ n n_ a ehmanal nar feme+4an
- h th:t :11 ch rn:h :r: :::t:d :t 1 ::t cr. : cv: j " ti..~ . 10 re.th:
+:r: '
it the total au-ber of redund:nt ch:rach in : :p: ift; rc::ter tria 'uactica 2: eke'a ia t'a "Tet:1 ug, ef cg:7 ega celg__ ey 7357 2 . 0 '. . l l B&W-STS 3/4 3-1 , June 1, 1976
TABLE 3.3-1 g REACTOR PROTECTION SYSTEM INSTRUMENTATION z A d TOTAL NO. CHANNELS MINIMUM CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
- 1. Manual Reactor Trip 2 / 1 3- / 1, 2 and
- 1
- 2. Nuclear Overpower 4 2 2 2 1, 2 24
- 3. RCS Outlet Temperature--High 4 2 3.-g 1, 2 2-3#
- 4. Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE 4 2(a)(b) 2, 1, 2 2# l S. RCS Pressure--Low 4 2(a) J 1, 2 2 -3#
, 6. RCS Pressure--High 4 2 -a-2 1, 2 2 -3#
2 7. Y:rf:bk La RCS Pressure V5 Yempe,.4, 2(a) +2 1, 2 2. -Of Y
" 8. Nuclear Overpower Based on Pump Monitor 4 2(a)(b) -3 2 1, 2 2-3# l
- 9. Reactor O..@1..t. Pressure--High 4 2 4-2 1, 2 E-9#
- 10. Intermediate Range, Neutron Flux and Rate 2 0 s%_ / 1, 2 and
- 3 4-
- 11. Source Range, Neutron Flux and Rate A. Startup 2 0 0- / 2## and
- 4 +- ?
, B. Shutdown 2 0 3, 4 and 5 1 54-l
- 12. Control Rod Drive Trip Breakers 2 per tr!;
c I Z _..,.go. u .y 2y 1, 2 and
- 64#
_ . . . +_ .. ,.- ,. _
- 13. Reactor Trip Module 2p 1
- 2. par hye =r + trig 2 ;y;pe-t c,T.t H p 1, 2e 24# .
trip ;y-t . @ 14. Shutdown Bypass RCS Pressure-High 4 2 9-2 2**, 3**, 5 6# u 4**, 5** 4 (. ( (
.m .
..._.m. __._
9 TABLE 3.3-1 (Continued) TABLE NOTATION l I
*With the control rod drive trip breakers in the closed position and the control rod drive system capable of rod withdrawal.
**When Shutdown Bypass is actuated.
#The provisions of Specification 3.0.4 are not applicable.
HHigh voltage to detector may be de-energized above 10-10 amps on both ! Intennediate Range channels. l (a) Trip may be manually bypassed when RCS pressure 111720{ psig by actuating Shutdown Bypass provided that: (1) The Nuclear Overpower Trip Setpoint is 1 % of5 RATED THERMAL POWER. I I (2) The Shutdown Bypass RCS Pressure--High Trip Setpoint of 111720) psig is imposed. , (3) TheShutdownBypassisremovedwhenRCSpressure>{l800{psig. (b) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3. ACTION STATEMENTS ACTION 1 - With. the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within Mours or~ (2 be in at least HOT STANDBY within the next 6 hours and/or open the control rod drive trip breakers. ACTION 2 - A With the number of OPERABLE channels one less than the Total Number of Channels STARTUF and/or POWER OPERATION may proceed provided -l' :' th: '-il;wir.;; ;;aditiea; .. .
^;;h'hi. uso c<nJ.hur Hw.hi oroepe<uble an e,n e. /1 a , . c)w,,ne t is p laa d in ne by ,wnal
- 2. 'h S:p;n hl: cher.;;l b ph ;;d h the t.Ipped
- dit'.: Githin ::: h;;r.
5. TS: "*-i _- Chr--'s OPE".8"LE --';_u.___, x -tt.i t _ _ . . < _ _ _ . . _ _ ___ . m .,___i _.. i C ' ' "'"'i'. ' ..""' -'Cl.;"'i '"'""'14 i'i. '
)
In ; i _ : _r.,
... ...., r e _. n , . c.
. . .~
. . i. _
. .. .' ..s.
. .T. . . i. ' ' ' - ~ ~ ~ -
6 #016chana
's nuba of opana e%d egua/ 4 He l l l*s Orciche .s in eru p ad n r o pagg7,ng
"* /> r a videcf o,,,e ,,,y,,z/a, 3,, po,ve z
, ,e , , gg B&W-STS 3/4 3-3 '
January 1.1977 l _. _ _ _ . y_
om een L U s us + h ,t ; p!!:?. J , h " " ' Fe w L *M c& ar . . i.w, h b., (ccn4) U"' ' n I
- f % ~ ' *** W l a ,. ope d um.),b en u.,, , /-/v v._
*'ai r /lc.>,- .
C' TABLE 3.3-1 (Continued) v ACTION STATEMENTS (Continued) , s n. A_ 6ka_ 1 skmo- msu ha ko ( n.ar-na u. skin e k s .- --
- - - - - - - - -s --
-._ .J c .. . . _ -A _ - -en'__=__.'t__
_ _ _ _ . ' . MA t_- s r=***w 'we wy wv su manuwsa an a s sy s. r uvus N n
.60 06wd3Aniy~ bw bcob Osc i,r a p us cakur
+ka k s c e n_ e_ 4. m_ + n.. .A_
. .. u 4 + h + k.a.
- l. a. n, d. , a. t. . . . e. .k.. - . . ,.1 . 4. =g 6--6.~J
~~a 7 ,. . c.,7,,.
- 4. #v.4 - s. +. 4.A a. . .i.m.1 11 I
- e. . c. 4 6. .L. - ., T. .u.rf. u~A t n. nt,ff ~. n, 2. , A.. .. )
.( l
..__.&.=.2.
. . a. m ".P.,. r. \ W i i
n -8 18 1_ . E . n #. D M T. =en., 'Pu.ws
. . rn u.A t as iAn.f nnt . vna f r".n u s EU be e s
_ if - . a.ww.ww. vvw.
. n, nunw - -- Tr4nr Cn+nn4n+ ---r--
4e waA ..aA
+n s / O t* I # n#
s - s -- -- t _ D A T.--en T. u.r.n. .u.A e.n. vnou . nt_f e n . o. . , A L. 2. n A L,.v-. . .. -.
~m 6km '
i 1 ... Af t M n D.A s1T nn.t.f --4 6 .,+ nn en T. Y. T. J. .,
,s.--
..vw n . 6. w. . w .J 1.. . . . ...w.
" ^ ?* M b^ ' ' *f . ..
APTTnM 1 U 4 + k. . ..+ b. -
.. ......L
. . .. .. r nn. v en,n n ., n t e
_L____3. ___ a ___ it _ to w o . . . . .w i a vus acaa wous s wuc Tn+s1 M umha_ , n#. eks-..n1.e_ c_ T a n T. t_t o..s. n. A
. o n.u. . _rD. n. DrD A T. T. nM..
. ... . . ..s
.,o i
_nenen-A n=no4AaA kn+h as +k. j r- ----r-- - --- - - - - - - - - - ra11 n..,4 n n, r- an..
- - - - - - - - --- 4 + 4 nn e - -a-A -.-- i enf4e44aA i 1
l
.a Tha 4 $nnn es hla_ c h s_ n n a_ l 4e n1sen_A in tho trinn,ad raaditia- uith4- ene her s
k Tk. u 4 < =. . . ,_-.. e k. . . , ._
.. .. . . , l. e. n.. o.. .en.
.,. . a .o n e.q w . . . ..w_ ... .?"Y >-
. Ema + . k ru.,m e . a w
, . _ . . . . _ . , a. n. .a. s A_ A,. 4. +. 4. a n e l. ,.. _.k....e. n
. .n_ a l m t ,u ka kunseenA V f___ __ i_ #1 L_ _
- s. ?
sr----- v d.n w,_ __ou..
. . . ,, r um... 399.uuso scou.us vu.
ena-4#4.es+4an A 11 1 1 _, .. -__... _snA +ha 4 n n na_ ps h l. a_ rhmnnal l 2 : : :" he bya=ce=d
#a- "a +a 'n =4a"+=e ia =au '> d
- k n o_ r--
._ . . no 4aA uhan nanoceswu
+n +act +hn +w4n k wam L n w
-- ---- s -- --- - --- --
r - - - - - - 2 C_ C_ n c_ 4 s + a_N_ u f. +.b.
. . . . +ba_ lan4A ..,.. .
.8
+. .b..a
. _. A b. 9 m n a l k_ _4. M.n,
+,e+nA
,-... . . , e . __4.,4.,.. 4a.
..1 i i l
ACTION + - With the number of channels OPERA 8LE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL Power level: M
- a. <-54 of RATED THERMAL POWER restore the inoperable Thannel to OPERABLE status prior to increasing THERMAL POWER above-64 of RATED THERMAL PCWER.
la N
- b. > 5% of RATED THERMAL POWER, POWER OPERATION may continue.
B&W-STS 3/4 3-4 January 1, 1977 i
~.
TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) g ACTIO;l With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
- a. 1 10-10 amps on the Intermediate Range (IR) in-strumentation, restore the inoperable channel to OPERABLE _jgatus pri.or to increasing THERMAL POWER above 10 amps on the IR instrumentation.
S. > 10-10 amps on the IR instrumentation, operation may continue. 5 ACTION - With the number of channels OPERABLE one less than re-quired by the Minimum Channels OPERABLE req;irement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 within one hour and at least once per 12 hours thereafter, arevnu st. ..._L-- m, uvi s. *. .L....i . . . . . , . .. n..n.e
.. n a..n e.L..,..__.1..,
r m_m _.. 1....
,, +ks- +km T. n. +._s 1 n , . . .kn . n. f. Ph s n. n.a_ l. e_ CTADTItD and/nr DnurD ADFDaTTON
".Ij ""9C^ed pro'. O . O IhO O. CY ".h COI . OSS I O
__.4ect e 2 2 U4+h4n 1 kn . . . . _1 Dlmen t h.n_ i nn,nnem b_ l a channal in the te4nnad ennA4f4nn nr 9 Damnun namn. . _ e n __,,. n n1 4. ,_. A_ + m_+ L. . . . .6.
,_ _ .. . .. . 1 . ..> s. . 2.e-d:;';; :::::i:t-> _2" -
'------*' 7 '7-- '
- b. On: 2ddition:1 char 90! may 5- byper:ed 'er "; te ?
haure #cr ?"r"ei'12nce +ertin; per cacci'icatie-4.2.'.'.1, : d th: ' :;:r:51: ch:rn:1 :t: : ::; 5: I ta'.:::::d f:r t: '^ -inct:: 'r. ;33 2' h ur ;;ri:d t:: :::;;;;;, t; t :t th: t m4.u .u.
. ... . . . . i.__: .,. .u t. __,ri; t,r;:k:r :::::i:ted-t_
_Cna,4#4,. 4a. A S 1 1 Tk_
.mmu nnt ha
..1 .
tn
. . . . 4 ....=, _ ._ . _. L 1 ._._ _L._
..._ ._1 . __ .L_._ I n s + k. .
_ kunnecaA.. ____ __ tact._+ba . . , .lan4e. n# n ekmanni
- 4. ,o,.__ ,,,-,4_,+.; m4 6 +u.
.,6....i 7 ,. .. -_-_. __ . . . . . . . _ 4_.-_..wi.
7 __.. t a.rv,n.o, o. u- t...
.t.
. t. .... ._...-. ..._L__ , ono,
- u. __i.. . . . . . . . no. c . . ._..e_ i_... .+u._
.... .. . 4
.. _2 u . . . u. u s _ 4_. ._ . _ . . . . , _ . . - _
_, .... . ._, r_ u. .. __ _. _, , n. o
. . . . . . .. c_ o n o _
, e ..,.. ,4 .. k. 4. .+
i_. . un, ...unnu ..,it,_ , u..__
.__.. . . . . ......~.............__...
( we r) B&W-STS 3/4 3-5 January 1,1977 l
.s
}r..yont.' f *. ]_p I-l1.. n .~ .< I- l 'I' ' ' ' ' ' " 'E " " ' ' "
(- ><. es,, I l. .. v v .h I h -
. : . c d ' < ~~
'rI ' > ' t l ' ' "" "
) ,, I-h e tw e/v.rI r,'l 'lro.'n e I' n<t<^ ,r- [ T< $
ii H,. ai./ n ,ep<.4 .l,h ,l'Is-
,..,y /ie./
4 e. -l. } . v.' d ,d <, i w g o.c h a n s eFA a r;p,,niy,t ov:<r >' /, . i G . b. ) } <sp d r. av.e. nhoil 1.u \
-)r!.l<,e ..n .r.n It, r e.n r../.' ?
\
c, o n ., i I . <. fr. I f,,,,1. .,1; d. he ira,,, . '( t, . .. i f th,cl < <$ t..t u) * !I l,'- rp < r 's.-
+ <l sr . se d' llt ~
4,,p c)<. jour. t t' o't ~t s < < . .- \ l
, Iv> l\ h .. W. h.,) ,e ,ll . , e ue. h + va-r # lo llt:> n o s.no D'c ,s.'och.:r s . I f +l t'w '3 I
t !- *'& l tit >r/(>rhN c so d /he. rw 1n n a o.,9 I r p 'i' s!. * '. < 1.' ~ \ n a V % v e,l- r,<l < a o l lt t.s e fltet (M.7 kk- k,.9, pe o^te 'S, bh u. .
< rnc ]ol' ~ IvtiI h *. . p I,> e .'l tn IIs R,,
f n*'t do.l ,a n 1, ;//Irat spit .
<.>- 4 0 9
.ycli/ /si a. I ,G,. :,- dir e., s .
s O i t i i l l I l N 4 3- S n. _ - __ _, _ _ _ . _,-. _ _ ._ _ ___ __ __ ___ .a
m , e, - TABLE 3.3-2 to REACTOR PROTECTION SYSTEM INSTRUMENTATION RESPONSE TIMES S CTIONAL UNIT RESPON MES m
- 1. 1 Reactor Trip Applicable
- 2. Nuclear 0 wer* seconds 1
- 3. RCS Outlet Temper re--High 1 seconds
, 4. Nuclear Overpower Based o CS Flow and ; AXIAL POWER IMBALANCE
- 1 seconds
.j 5. RCS Pressure--Low seconds 1
g 6. RCS Pressure--High 1 seconds y Variable Low RCS Pressy
- 7. seconds i
- 8. Nuclear Overpower ed on Pump Monitor
- seconds
- 9. Reactor Con ent Pressure--High s nds 1
I
& *Neutr detectors ars exempt from response timo testing. Response time of the neutron g sig portion of the channel shall be measured from detector output or input of first el'act ic c nent in channel.
01 k LZLU
~ ~
- 1. =
LlLE ~ TTZ1
E TABLE 4.3-1 Y g REACTOR PROTECTIM SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED
- 1. Manual Reactor Trip N.A. U(1) N.A.
- 2. Nuclear Overpower S 1, 2 g, Qp'
- 3. RCS Outlet Temperature--High S R H 1, 2
- 4. Nuclear 6verpower Based on RCS
"(2,g.r.d Q(7,8) M F1ow and AXIAL POWER IMBALANCE S(4) 1, 2 R 5. RCS Pressure--Low S R H 1, 2
+
w 6. RCS Pressure--High S R H 1, 2 0 7. MHeh L= RCS Pressure VS.72>tpyg R H 1, 2
- 8. Nuclear Overpower Based on Pump Monitor S R H 1, 2
- 9. Reactor S- 2 " Pressure--High S R M 1, 2
- 10. Intermediate Range, Neutron M/I.
Flux and Rate S @ M And S/U(1)(5) 1, 2 and *
- 11. Source Range, Neutron Flux ////
p and Rate S -Rf7-J- 58 W S/U(1)(5) 2, 3, 4 and 5 l { 12. Control Rod Drive Trip Breaker N.A. N.A. M and S/U(1) 1, 2 and
- Q 13. Reactor Trip Module N.A. d-&A-- M 1, 2, and *
. 14. Shutdown Bypass RCS
- Pressure-High S R M 2 * * , 3 * * , 4 * * , -G**- l 0
~
('. TABLE 4.3-1 (Continued) NOTATION h* Wkeat at S e r v w. s With any control rod drive trip breaker closeo. When Shutdown Bypass is actuated. , (1) - If not performed in previous 7 days. ! (2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference > 12}, percent. _ I (3) - Compare incore to out-of-core measured AXIAL POWER IMBALANCE I above 15% of RATED THERMAL POWER. Recalibrate if absolute difference >T27 percent. (4) - AXIAL POWER IMBALANCE and loop flow indications only. l k.h ba k'e.ut M rntHi- i. hunn d. c (5) d,v,er,,v, ifyHre at least $d decade overlapf ;;ri'ifd 'r p=i:=h. l 3 0 t (6) - d yt. 4 1 s. s o vec.a. ru n s M s4 ra m u f a4.,o,s, Each train tested
,y vther month. "
(7) - Neutron detectors may be excluded fr 'm CHANNEL CALIBRATION. 1 (8) - Flow rate measurement sensors may be excluded from CHANNEL I CALIBRATION. However, each flow measurement sensor shall be calibrated at least once per 18 months. l 1 l N . v B&W-STS 3/4 3-8 January 1,1977 l
m INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumen-tation channels shown in Table 3.3-3 shall be OPERABLE with their trip l setpoints set consistent with the ' lues shown in the Trip Setpoint column of Table 3.3-4.:nd ith PE' E 7?"ES :: de - " 791: ?.? 5. I APPLICABILITY: As shown in Table 3.3-3. ACTION: s
- a. With an ESFAS instrumer.tation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable l and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setooint Value. i
- b. With an ESFAS instrumentation channel inoperable. take the l action shown in Table 3.3-3. I SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABl.c by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CH#NEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-2.
)
4.3.2.1.2 The logic for the bypasses shall be demonstrated OnERABLE 1 during the at power CHANNEL FUNCTIONAL TEST of channels affected by ! bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATIn:1 testing of each channel affected by bypass operation.
' 3. . .0 The CNCI":D.E0 S?fET FEAT"P,ES PESP^"SE T!",C cf ;;;h :: TAC fun;tien sh;;l be d .an;tr;t;d t; be "ithi" +ha 14=4+ =*
!! nnth;. 1==s+ ca-= ra-
;;h t,.;t :h;il M:hd: :t letzt ene ch:rn:1 per fun:th:
- ' th:t :!' 'r n:' :r: t::t:d :t 1:2:t One : :ry " ti.;; 10 ;; nth;
.i:r: " h th; t;t;l n,.
i:r Of reintnt ch:rn:h " : :;;;ifk CSIAC l l
.f4;c-ti:n : :h;wr 5 th; "T;t;l ";. Of Ch:rn:h" C:h= cf T;tk 2.0 -0. l B&W-STS 3/4 3-9 June 1,1976
TABLE 3.3-3 g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION F un MINIMUM , TOTAL N0. CHANNELS CHANNELS APPLIC* isle ' FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
- 1. SAFETY INJECTION s
- a. High Pressure Injection R.cdc puM T 1) Manual Initiation c- t + - t Pressure-2 1 2 1,2,3,4 12 l High 4 (3)* 2 3. (2f 1,2,3 9# M
- 3) RCS Pressure-Low --4-( 3 ) 2 4-f22 1, 2, 3* 94 fe&) A-
- 4) Automatic Actuation Logic 2 1 2 1,2,3,4 10 i R* b. Low Pressure Injection ca h 1) Manual Initiation 2 1 2 1,2,3,4 12 l
- 2) Containment Pressure-High {3) 2 1, 2, 3 12{ 9# 40-}+-
- 3) RCS Pressure-- ,
Low-Low {3] 2 4-121 1, 2, 3** 9# -f9hr-
- 4) Automatic Actuation logic 2 2 1,2,3,4 f' d2.B..ld. g' I!T SPRAY C0!!T.".I..-
1 10 l ' c c a. 5 Manual Initiation p2 /. 2 1,2,3,4 12 I
- s. re.- a *. M.. se
- b. Co# iit NrEdnt he'ss)ure-9 '
y- _ High-High 3 2 #1 1,2,3 0 7 I
- c. Automatic Act ation Logic -@--- 4- 4- 1- M T 4e I
k w B id) %*nt 2 1 2 I , 2, 3, 4 10 l I. m.___, .. _ dO*mul U N 2 / g r
f, 4, h, ff ID
M i f l (r0 (O l l l l l TABLE 3.3-3 (Continued) g *
.F ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION
! w ! -4 m MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE N0 DES ACTION I'<4r 7% 14.ng
- 3. MTAmmT ISOLATION 4#:p coes.ixt sys rEN a.N Crt&- r_k e Bo nt O g olation ar.J c a l.,g .g d m 3 11 Manual Initiation 2 1 2 1,2,3,4 12 l !
w P e.tei.. ..L Pressure-
) h f kY'~~THigh o
+f3) 2 4{21 1,2,3 9#-f9hP- I i
w 3) Automatic Actuation Logic 2 1 2 1,2,3,4 10 4
,L. #6 . 6 . -- --.....a. n..__ J
. . . , . . . . r_ _o L. . . . *...
J e al m + 4. .
.s s u._.., ,_ ,__ o
. . . . ..2..s. . ..-. h u
u, a, N. , C""t & ~.t ***ic'r^- ' us,u
. - 2 t, v, , , ins
, n ,
s
-3) C nt:'.. nt sh, .,6, ... -
_u a r,\ n , in, , - ~ . ,,n n.=.=.A-- S
. 21. ww.
=
. .-. L3_
- s. a.T . . a. s.J..
- u. n. 3. .
. T g ed g h J gLg .g L, J, 7 ..N
&_A.._ L
- _
C 'l " * * " ' ' ' * ' * "*6""*'V 3 y
."Iande 9 1 o , e 6 a i a.- = =
d ,--- T b ., h, 11. m
- _A e ,m.m
, . ..nt.aT.
. A ...
T a mmas=ame. e.A.n u.
. n ee N a Manism1 Yn4t4m+4an 9 1 o a M i a = i
*- s, ', J. 7 i. <- n I ,
-. . . . . . . . . . . TT3ttree '
N4nk .. A [S\ n **' - - n ,~. ~ ~ ' '"# ' #" 5 I* '* #' \#0' ~
.. ...__--s2. n_x..x> _
... n w . w w . ,v..
J naic ? 2 ' , 2, 3, 4 i^ l i i
/ \
TABLE 3.3-3 (Continued) l ENGINEERED SAFETY FEATURE ACTU.\ TION SYSTEMS INSTRUMENTATION
, M un l
' MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1sut Scaa<.
- 5. -MAIM-STEAM 150LAT40N- Z / 2 lj2, .3 12.
/Ats%'or,5WTArroD AND Orvint. S)'Lfdb1h@c) l .. m,.. 1 :,,:t:.t:0, . : 1 : 1, 2, 2 : 12 I i b. ainment Pressure-High 4 (3) 2 3 (2) 1, 2, 3*** 9# (9)# I
- c. RCS Pressure-Lo 4 (3) 2 2 (2) 1, ***
- 9# (9)h l
- d. Automatic Actuation Logic '
1 2 1, 2, 3, 4 10 l
$ 5. CONTAINMENT EMERGENCY w SUMP SUCTION b a. Manual Initiation 2 1 2 1,2,3,4 12 I
- b. Borated Water Stora Tank-Low 4 (3) 2 3 (2) , 3 9# (9)i l
- c. Au c Actuation a ogic 2 1 2 1,2,3,4 10 l g _
~
v M ( ( (
m ( -
. TABLE 3.3-3 (Continued) l TABLE NOTATION
- Trip function may be bypassed in this MODE with RCS pressure below 87F8'1000)psig Bypass shall be automatically removed when RCS pressure exceeds g g psig.
**Trt;s i..- func.ti:n ::y be by;;;;;d in thi- ,-_,,-
. __- - ,, <_ __m_
MODE -ith 9CS pr:::;r: bel -
\ "T VV g ya3y, wypuJa a gge 3 g Wb u w bVillu b 8 b u l UJ 5 w.erv u h u IW .J
_ __ 3_ renn\ .1 pu sh 8 8 VI WJJWB E
--- ... s..., r.. , .
***'-fp fun: tier : 3 he by;::::d '- thi: "0"E "ith cte2- "ea"*2te" pr ;;;r: b lew (725) p;is. By;;; Chall be aut?"St4cil'y "?""" d 6:r Ot::: ;;ncr t:r pr ;;r: cx:::d; (725} ?:fg.
#The provisions of Specification 3.0.4 are not applicable.
ACTION STATEMENTS _nevenu n ut.u'.u. ..._u__ -,nocono,e ,w,..mi, +w,.
~~" ' m.m to,e +wo vi:1,'".'_._d:_-'.'., Ei,T..Ti,
. . . . . . . . . . . . . . . . _ ' _; ,;,_ in;~;,;__- nyn;;,a nynvia;a b:th t f th: f:lle.:' ; ::nditi r.: 07: ;;ti;fi;d.
- . . : 'ne;;r:b!:':h:rn:1 i: ;1 ::d '- th: tri;;;d ennA4+4nn u4+h4n 9.a agg ,
- k. vu us_ _... ek,..mi,
- . . .._.. ..._.. ._ . _ n_ or _- o n_ _at_ e_ ,o n ,,4 ,m m
+ 4e m:t; S:.:r;;r, ::: :dditt:n:1 ch:rn:1 ::; 5 _
byaerrad # r up t: ? heur: f:r :urve!'l: :: tactinn noe Cnne4f4rm+4n. A ._._.1 4 9 1 ACTION [9{- With the number of OPERABLE Channels one less than the Total Number of Channels operation may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour. l ACTION 10 - With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the next 30 hours; however, one channel may be bypassed for up a to I hour for surveillance testing per Specification 4.3.2.1.1.
^CTIO" 'l 'dith 1 :: th;r th: "'-i u- Chir :!: OPEP^"LE. 0;;r:ti::
;;3 ;ntir.;; pr:;fd:d th: ::nt:iic nt purg ;nd ca.u..t V:!?:: ir :!-t i d c'er^d.
%i. -
B&W-STS 3/4 3-13 January 1,1977
( - ( l p t - V TABLE 3.3-3 (Continued) ACTION STATEMENTS (Continued) ACTION 12 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERA 3LE status within 48 hours or be in at least HOT STANDBY wi+.hin the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
'^CTIO" 12 i
- th th; numbe of 0"E"^.SLE Charn:1:
Tvioi n umuc, ei Ch nnel; sperati;n : ::: ?::: th: -ebe-- 3 pr;;;;d pr:;ided th in:per:b!: char e' 4? p1=-ad in +ha hunacead enn-
-dition and the Ediimum ch;nn:1 0" ""."LE r quirement i:
de en ty3ted..4 s<o 1 h a. . . . nno maa4+4nn21 channoi m=v bu bypa336d IGi ap I: 2h Ur: for SurY i'l!" tect#^0
;;r Speci'ic: tier '.2.?.' '
, i 1
<-~ ,
I l
- \
l l l
/
l l I
\.
B&W-STS 3/4 3-14 January 1,1977
[)
- g. f(G q-TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION TRIP SETPOINTS
] FUNCTIONAL UNIT TRIP SETP0 INT ALLOWABLE VALUES
- 1. SAFETY INJECTION
- a. High Pressure Injection g%d# g*2)'Contat= 1) Manual Initiation nt Pressure-High Not Applicable Not Applicable 1>_((5' psig-13.7ps,-t 184.2) psig ), sfr.7 psiq.
- 3) RCS Pressure-Low MOG) psig i5eo h/ps7] > fft$ psig
- 4) Automatic Actuation Logic Not Applicable Not Applicable
- b. Low Pressure Injection Rent 1/8f/'l Manual Initiation - Not, Applicable Not Applicable
{ %)2 Contai= cat-Pressure-High
- 3) RCS Pressure--Low-Low if5t-ps49>?.7psu[cj.ps9)
> (409) psig lino 1((9.I) psig,), /2;7ps.cc
> 094 psig Y 4) Automatic Actuation Logic Eat Applicable Eot Applicable G D' Toe 64e;c,s
- 2. CONTAIN"CNT SPRAY
- a. Manual Initiation Not Applicable Not Applicable Q&, g,jg,f' .' Contaf =:nt Pressure--High-High 1 (30)- psig M7/sm[3pg) 1((50.4 psig),49f
- c. Automatic Actuation Logic Not Applicable Not Applicable f gl Rmclurf1t,Jdy;
- 3. CONTA!N.".CNT ISOLATION art / Ca>A ra g g.
g a. %u% P-u nt Cont:i= Id.,ydlation Is Manual Initiation Not Applicable Not Applicable g b ' Q") g2) ,Centtiment Pressure-High 1(5)psic-/F.7pt[@3'y) y 3) Automatic Actuation Logic Not Applicable s(t No Applicabl,eW.2) psig .) /K f
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _s_ _ _ _ _ _ _ _ _ _ _ _ _ _______ '____ _ _ _ _ _
f l TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION TRIP SETPOINTS F TIONAL UNIT TRIP'SETPOINT ALLOWABLE VALUES
- b. Co inment Purge and Exhaust Isola n -
1 Manual itiation Not Applicable No pplicable 2)) Containme Pressure-High 5 (5) psig ( ) psig 3 Containment 4)J Automatic Actua 'on Logic iation-High < (25) mr/hr _(
< ) mr/hr Not Applicable Not Applicable
- 4. CONTAINMENT COOLING
$ a. Manual Initiation Not Ap cable Not Applicable w b. Containment Pressure-High 1( psig 1( ) psig g c. Automatic Actuation Logic Applicable Not Applicable
- 5. MAIN STEAM ISOLATION
- a. Manual Initiation Not icable Not Applicable
- b. Containment Pressure-High 1 (5) ps ) psig
- c. 1> ((
RCS Pressure-Low 1 (1600) ps ) psig
- d. Automatic Actuation L c Not Applicable Not Applicable
- 6. CONTAINMENT EMERGEN SUMP SUCTION
- a. Manual In ation flot Applicable t Applicable q b. Borate ater Storage Tank-Low > (3) feet
_ 1 feet
- c. Aut tic Actuation Logic Not Applicable Not Ap icable 7.
g L OF POWER 4.16 Ky Emergency Bus Undervoltage ( + ) volts ( + ) volts ( ( 4 sA ren
^'O
'>~e enac u>sr~c-- ( > oo ,,, ~ svs ,,,3 l'"Te n sycr m zc ass y
hamm
# ~ ,.., r c.
_d TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES R41TIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDJr N w
- 1. M anual a Emergency Core Cooling l
' h Pressure Injection NA low ressure Injection Compon t Cooling Water System NA Service W er System NA w
- b. Containment D ressurization Systems l [
Containment Spra NA puut Containment Cooling NA
- c. Containment Isolation tems l Containmc.1t Isolation Containment Purge and Exht st solation NA, Q
9 NA .
- 2. Containment Pressure-High High Pressure Injection 5( )*/( )** '
Low Pressure Injecti 1( )*/( )** Main Steam Isolati n 3( )*/( )** Containment Iso tion 3( )*/( )** g Emergency Ve ilation 3( )*/( }** }i m' . Containme Purge and Exhaust Isolation 5( ) ( )** puum Contai ent Cooling 3( )*/( )** Comp ent Cooling Water System 1( )*/(
- S vice Water System 1( )*/( )** ,
i - l w . 1 B&W-STS 3/4 3-17 - June 1, 1976
'~ *
. .Gn- A .y M ;. 7 *-- m 3- p r ~ y,-~w w ,_,_ -
6 i 7 -- g- m ' , TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES m ATING SIGNAL AND FUNCTION RESPONSE TIME IN SECON g
- 3. M ainment Pressure--High-High A
Conta nt Spray 1( )* )**
- 4. Reactor Co ant Pressure-Low High Pressure ection _ ( )*/( )**
Main Steam Isolati Component Cooling Water stem 3( } gN'
< ( )*/( )**
Service Water System 5. 1 ( }*/( )** janum Reactor Coolant Pressure--Low-Low Pressure Injection 1 ( )*/( )**
- 6. Contair. ment Radioacti y-High Containment Purg nd Exhaust Isolation
- 1( )**
ilation O 7. Emergency V Borated ter Storage Tank-Low _ ( )** ;-Q ;,. , M Co inment Emergency Sump Suction 5( ) ens I i M i l O B&W-STS e i 3/3 3-18 June 1, 1976 l
- g. ;,.r.g,,., --- .,_ .. m, r _ _ . m _
S
. _ - - .=---
Y 1,, W TABLE 3.3-5(Continuedl y ear' .
% LE NOTATION
- Diesel ge r starting and sequence loading dela cluded.
Response time 11 cludes movement of valv attainment of 6 pump or blower discharg sure.
** Diesel generator startin sequent ing delays not included. p Offsite power ava Response time limi
- e. udes movement of .
valves and nment of pump or blower discharge p e. IImum w O -
! .4' User i
i o B B&W-STS 3/4 3-19 June 1. 1976 l _ a. C -- '
- r'r 9 C.WMCE : . . .
?)
S TABLE 4.3-2 E T ENGIhEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION SURVEILLANCE REQUIREMENTS m w I " CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED
- 1. SAFETY INJECTION
- a. High Pressure Injection
- 1) Manual Initiation N.A. N.A. M(1) 1, 2, 3, 4 R d L,d T 0;.tcirc.;.t Pressure-High S R MM 3 3) RCS Pressure-Low S R M 1,
1, 2, 2, 3 3
- 4) Automatic Actuation Logic 'N.A. N.A. M(2) 1, 2, 3, 4
$ b. Low Pressure Injection Y g 4 # J )' Manual Initiation N.A. N.A. M(1) 1, 2, 3, 4
- 2) C::tcir..xat Pressure-High S 1,2,3
$ gl,y,"d 3) RCS Pressure--Low-Low S R
R Mts-)- H 1, 2, 3
- 4) Automatic Actuation Logic N.A. N.A. M(2) 1, 2, 3, 4 Ra%me Banoms.
- 2. -GONTMNMENT SPRAY
. a. Manual Initiation N.A. N.A. 1,2,3,4 M(1)
Reatbc Pa.1d..
- b. Cent & m t P ssure--
g High-High 5 R 1,2,3 MfB-)- 1 1
- c. Automatic Actuation Logic N.A. N.A. M(2) 1, 2, 3, 4 I M
( ( (
P ~
/
TABLE 4.3-2 (Continued)
! ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION SURVEILLANCE REQUIREMENTS.
b, d CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED REkRDR Ban.o &,
- 3. .GQNTADME4T ISOLATION AND 8&dR6.cmuu-b
- a. %k1..... sa.u.a#ation
.. ant Isol li Manual Initiation N.A. N.A. M(1) 1, 2, 3, 4 E ta j " b 'kj'rq'T ? Centai. : t Pressure-7 High S R M(44- 1, 2, 3
- 3) Automatic Actuation Logic N.A. N.A. M(2) 1, 2, 3, 4
. ntainment Purge and Exhaust Isolation w
k
" 1) Manual tion N.A. N.A. M(1) 1 ,4
- 2) Containment Pre -High 5 R M(3) , 2, 3 l
- 3) Containment Radioact v High R 1,2,3,4
- 4) Automatic Actuation Logic N.A. M(2) 1, 2, 3, 4 1
- 4. CONTAINMENT COOLING ,
c. E a. Manual Ini N.A. N.A. 1, 2, 3, 4 M(1) [ b. Co nt Pressure-High S R M(3) , 3 utomatic Actuation Logic N.A. N.A. M(2) 1, 2, , O
- S. MMN-STEAN -ISekhHON L INE n es A g iMTRu? Man 17AT:Od '
/) n D Cold 1tOL SysTa m VJ k & t, z,n
- . .:nn! Initicti:n N.A. ii. A. ii(i) i,2,a
- b. C::t in int '. 35u, c .i f.
S R l"(3) I,
- c. RCS Pre:M e Ler S
" " 1, 2, y 33 at Antruna tic Actuatinn (9 gig u A. y , 8, , y({} },^}
s i
O O 00 '
, Et.c=rc~
TABLE 4.3-2 (Continued) EERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION SURVEILLANCE REQUIREMEN I CHANNEL MODES IN WHICh
. CHANNEL CHANNEL FUN ' #L'2VElLLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED
- 6. CONTAINMENT EMERGENCY SUMP SUCTION l
- a. Phnual Initiatior. . . M(1) 1, 2, 3, 4
- b. Borated Water Storate Ta S R M 1, 2, 3
- c. Automatic Actuatio ic N.A. N.A. 2) 1, 2, 3, 4
?!
m 7. LOSS OF P S m . 6 Kv Emergency Bus S R M 1, 2, A, Undervoltage N Of a
=
x
? ." '
- g. -
O s
l ( ', r l' TABLE 4.3-2 (Continued) TABLE NOTATION (1) Manual actuation switches shall be tested at least once per 18 . months during shutdown. All other circuitry associated with manual safeguards actuation shall receive a CHANNEL FUNCTIONAL TEST at least once per 31 days. (2) Each train or logic channel shall be tested at least every other
, 31 days.
(;} T;,m CNANN:L r'JNCT:0NAL TEST ch:1' include ex^ cir4"a +ha +"""*"i++a" 13 :pp!;#n; cither !: Lur er Prerre e te +"" ?r""a""i St" C d? # 0# +"" tr:r.;ri tter t
- l B&W-STS 3/4 3-23 June 1. 1976 l
h (' INSTRUMENTATION t/ l 3/4.3.3 MONITORING INSTRUMENTATION ! /9 RADIATION MONITORING INSTRUMENTATION *
, LIMITING CONDITION FOR OPERATION 1 55158 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm / trip setpoints within the I soecified limits. '
w APPLICABILITY: As shown in Table 3.3-6. l . . ACTION: ' unus a. With a radi: the value tioninmonitoring shown Table 3.3-6,channel alarm adjust the / trip setpoint to setpoint within exceeding, I the limit within 4 hours or declare the channel inoperable.
- b. With one or more r.idiation monitoring channels inoperable, take '
the ACTION shown in Table 3.3-6. I 4......
- c. The provisions r f Specifications 3.0.3 and 3.0.4 are not applicable.
e gy I I W pyiiii SURVEILLANCE REQUIREMENTS P 4.3.3.1 Each radiation monitoring instrumentation channel shall be demon-strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRA-TION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3. l ff.)- sq. M lI I B&W-STS 3/4 3-24 June 1, 1976 m l
e q (T - k . kb h.- co TABLE 3.3-6 E RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM / TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION
- 1. AREA MONITORS
- a. Fuel Storage Pool Area
* -I 4
- 1. Criticality Monitor (1) 1 15 mR/hr (10 - 10 ) mR/hr 17 9 ii. Ventilation System Isolation (1) ** 1 (2 x background) (1 - 10 5) cpm 15
- b. Containment u i. Put je & Exhaust 2
u Isolation (1) 6 1 (2 x background) (1 - 10 5) cpm 16 b
- 2. PROCESS MONITORS h
h I
- a. Fuel Starage Pool Area j
- 1. Gaseous Activity -
Ventilation System Isolation (1) **
-< (2 x background) (1 - 10 5) cpm 15
- 11. Particulate Activity -
Ventilation System Isolation (1) ** 1 (2 x background) (1 - 10 5) cpm 15 .2 O b. Containment E i. Gaseous Activity Pr a) Purge & Exhatist
@ Isolation (1) 6 1 (2 x background) (1 - 10 5) cpm 16 R
y b) RCS Leakage Detection (l) 1, 2, 3 & 4 Not applicable (1 - 10 ) cpm 14
- 11. Particulate Activity
- a) Purge & Exhaust Isolation (1) 6 1 (2 x background) (1 - 10 5) cpm 16 b) RCS Leakage Detection (l) 1, 2, 3 & 4 Not applicable (1 - 10 ) cpm 14
*With fuel in the storage pool or building
** With irradiated fuel in the storage pool
~ ~ ~
4
~ I"s f 1 I ~ El ~ l 1 L~~ l F
t 5 (ATEC
~
- ('
% IllME TABLE 3.3-6 (Continued) !
O TABLE NOTATION f w ACTION 14 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE require- p ment, comply with the ACTION requirements of Specification (3.4.6.1). ACTION 15 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply , with the ACTION requirements of Specification (3.9.12). h ACTION 16 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply a with the ACTION requirements of Specification (3.9.9). t ACTION 17 - With the number of channels OPERABLE less than j required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with - portable monitoring instrumentation at least once per en==en
- r. 24 hours, n
V w pumma l l l i I oV E B&W-STS June 1, 1976 3/4 3-26
,. m . 4.. ,,_-. .
_t TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS h CHANNEL H0 DES IN WHICH
& CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE l
g INSTRUMENT CHECK CALIBRATION TEST REQUIRED '
- 1. AREA MONITORS
- a. Fuel Storage Pool Area l'.
- i. Criticality Honitor S R H *
- 11. Ventilation System Isolation S R H **
I b. Containment
- 1. Purge & Exhaust Isolation
.q 2. S R H 6 h PROCESS MONITORS { i' Y a. Fuel Storage Pool Area i, O i. Gaseous Activity - ; Ventilation System Isolation S R H **
- 11. Particulate Activity -
Ventilation System Isolation S R H **
?
- b. Containment
- 1. Gaseous Activity
} E' a) Purge & Exhaust l 5 Isolation S R H -
6 ,
~ b) RCS Leakage ) _ Detection 5 R H 1, 2, 3. u, 4 ,
e 2 11. Particulate Activity a) Purge & Exhaust Isolation S R M 6 b) RCS Leakage Detection S R H 1, 2, 3, & 4
- With fuel in the storage pool or building
** With irradiated fuel in the storage pool
( INSTRUMENTATION INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 As a minimum, the incore detectors shall be OPERABLE as specified below.
- a. For AXIAL POWER IMBALANCE measurements:
- 1. Nine detectors shall be arranged such that there are three detectors in each of three strings and there are three detectors lying in the same axial plane with one plane at the core mid-plane and one plane in each axial core half.
- 2. The axial planes in each core half shall be symetrical about the core mid-plane.
- 3. The detector strings shall not have radial symmetry.
b. For QUADRANT POWER TILT measurements with the Minimum Incore Detector System:
- 1. Two. sets of 4 detectors shall lie in each core half. Each !
set of detectors shall lie in the same axial plane. The " two sets in the same core half may lie in the same axial plane. i
- 2. Detectors in the same plane shall have quarter core radial l j
symetry. em. nnanonur nnuco vn y eN ,.a e ,,, , , , , , _ ___ _ _ . , m e . w . w . hb255h 4- in: ora 02:5 cer: Dck$$k$r'h:5SEh9'b555'[$5}5OfEb:EEte:t:r: 12 dr:nt ch:!' b; 0":"'"LE. APPLICABILITY: of: When the incore detectio'1 / stem is used for surveillance
- a. The AXIAL POWER IMBALANCE.
- b. The QUADRANT POWER TILT.
eN . s e 3g . .. 9 ACTION: With less than the specified minimum incore detector arrangement OPERABLE, "d; nt an ir.c;c; d; tat;r; for th: ;;plistic .T.eaf te.-f ag Nactien . The - provisions ndm pwr of Specifications k $ g o yu3.0.3 and 3.0.4 are not applicable.
, 4 M e p c ,, , , , , y, m $ ,, 4 , ( w ,, , y , ,_ v BkN-S'TfFws"pcom4,m4cn. 3/4 3-28 January 1, 1977 l
(' r5 ) t INSTRUMENTATION s v. SURVEILLANCE REQUIREMENTS 4.3.3.2 The incore detector system shall be demonstrated OPERABLE: to EFKV
- a. By performance of a CHANNEL CHECK within-7 ey; prior to its use for measurement of the AXIAL POWER IMBALANCE or the QUADRANT POWER TILT. ;
- b. At least once per 18 months by performance of a CHANNEL CALIBRATION which does not include the neutron detectors.
1 B&W-STS 3/4 3-29 ocuary 1,1977
R.ssolVi Slw dul Z Is spec & / (> IllSTRUMENTATION v SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 I shall be OPERABLE. APPLICABILITY: At all times. ACTION: l
- a. With one or more seismic monitoring instruments inoptrable for i more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not I applicable, a
SURVEILLANCE REQUIREMENTS 4.3.3.3.1 Each of the above seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-4. ! 4.3.3.3.2 Each of the above seismic monitoring instruments actuated during a seismic event shall be restored .. OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 5 days following the seismic event. Data shall be retrieved from actuated instruments and analyzed to detennine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Comission pursuant to Specification 6.9.2 within 10 days describing the magnitude, frequency spectrum and resultant - effect upon facility features important to safety. O O (' B&W-STS 3/4 3-30 June 1,1976
?5 0 fv't G $b W Rat b /5 Spechtch
( TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUti MEASUREMENT INSTRUMENT INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
- 1. Triaxial Time-History Accelographs
- a. 1
- c. 1
- d. 1
- 2. Triaxial Peak Accelographs
- a. 1
- b. I
- c. 1
- d. 1
- e. 1 G 3. Triaxial Seismic Switches
- a. 1*
1
- b. 1* l l
C. l* l
- d. 1*
- 4. Triaxial Response-Spectrum Recorders
- a. 1*
- b. j )
- c. 1
- d. j
- e. -
i l f. I 1
- With reactor control room indication B&W-STS 3/4 3-31 June 1, 1976 I
TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST
- 1. Triaxial Time-History Accelographs
- a. ,
M* R SA
- b. M* R SA
- c. M* R SA
- d. M* R SA
- 2. Triaxial Peak Accelographs
- a. NA R NA
- b. NA R NA
- c. NA R NA
- d. NA R NA fh y
- e. NA R NA
- 3. Triaxial Seismic-Switches .
- a. M** R SA '
- b. M** R SA '
- c. M** R SA '
- d. M** R SA I
- 4. Triaxial Response-Spectrum Recorders
- a. M** R SA '
- b. NA R SA
- c. NA R SA
- d. NA R SA
- e. NA R SA
- f. NA R SA
,
- Except seismic trigger
** With reactor control room indication v
B&W-STS 3/4 3-32 June 1, 1976 i
Q f. ..
. l l
INSTRUMENTATION ' N METEOROLOGICAL INSTRUMENTATION _ ITING CONDITION FOR OPERATION
\
I
/
3.3.3.4 The meteorological monitoring instrumentation c annels shown in Table 3. 8 shall be OPERABLE. I APPLICABILI : At all times. ACTION: I
- a. With one r more required meteorolog* al monitoring channels I inoperabl for more than 7 days, pr are and submit a Special Report to t Commission pursuant o Specification 6.9.2 within the next 10 dqys outlining the c se of the malfunction and the plans for rest ing the channel s) to OPERABLE status.
- b. The provisions o pecificat' ns 3.0.3 and 3.0.4 are not I applicable.
(3 %) % SURVEILLANCE REQUIRE NTS
/
4.3.3.4 Each oi/' the above meteorological monitoring instrumentation channels shall oe demonsf. rated OPERABLE by the performance of e CHANNEL CHECK and CHANNELCALIBpTIONoperationsatthefrequenciesshowninTable4.3-5. l Q - B&W-STS 3/4 3-33 June 1, 1976
knas {
~
(' EL.E T C~ O( / TABLE 3.3-8 ' t W METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUME MINIMUM y LOCATION OPERA
- 1. WIND SPE
- a. Nominal E v. 1
- b. Nominal Elev. 1 W
.,W)
; :.9 ansa
- 2. WIND DIRECTION
'. 1
- a. Nominal Elev. i 1 I
- b. Nominal Elev.
1 M o l
- 3. AIR TEMPERATURE - D AT
- a. Nominal Ele .
1
- b. Nominal lev.
m M
\ '
%,h A
O' B&W-STS 3/4 3-34 September 1, 1976
~~
t !
' Sh_?." .m.. . .. .- - . ,
?
et' s
.i TABLE 4.3-5 , .y g -
METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
?
$m CHANNEL CHANNEL
, 9 INSTRUME CHECK CALIB ON C :;:
. 1. WIND SPEE ;;r.p a. Nominal };i;. Elev. '
D SA
'I) b. Nominal
;f .
i Elev. SA
- 2. WIND DIRECTION
!?] a a
- a. 9 Nominal 1: Elev. D SA i\
.- w g } b. Nominal
, w Elev. '
SA
- 3. AIR TEMPERATURE - DE T
- 1. Nominal Elev. D SA d b. minal
} c Elev. D SA E.
M M_ _ I
~~
- IX ~1 13a 1 I 1 I
INSTRUMENTATION ElC 'E i 9 REMOTE SHUTDOWN INSTRUMENTATION-e ING CONDITION FOR OPERATION 3.3.3.5 remote shutdown monitoring instrumentation channels hown pupus in Table 3.3 shall be OPERABLE with readouts displayed exter al to I the control ro . APPLICABILITY: M0 1, 2 and 3. ACTION: *
- a. With the number o OPERABLE remote sh down monitoring channels less than required Table 3.3-9, ther restore the inoperable tumuu channel to OPERABLE s tus withi 0 days, or be in HOT SHUTDOWN within the next 12 hours
- b. The provisions of Specific n 3.0.4 are not applicable. l SURVEILLANCE RE REMENTS x
p r y l 4.3.3.5 ach remote shutdown monitoring instrumentation channel sh 1 I be d strated OPERABLE by performance of the CHANNEL CHECK and CHAN " CAL TION operations at the frequencies shown in Table 4.3-6. I i
?
4
. i B&W-STS 3/4 3-36 June 1, 1976 1
'"g* I'.8 * * *
, -=
'w. . . . - . . .m. , ,., 'b. _M !
. TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION E MINIMUM f READOUT MEASUREMENT CHANNELS e s 3 IRSTRUMENT LOCATION RANGE OPERAB W ; 1. Po Range Nuclear Flux 1
- 2. Intermed e Range Nuclear 1
. Flux -
P.
- 3. Source Range Nucle Flux 1 i
- 4. Control Rod Drive Trip B Indication ker OP -CLOSE 1/ trip breaker l}
'd
- 5. Reactor Coolant Temperature - 1 Average -
tp 6. Reactor Coolant Flow Rate 1 S
-Qi-
{
- 7. Pressurizer Pressure 1 h
- 8. Pressurizer Level 1
!. 9. Steam Generator Pressure 1/ steam generator j 10. Steam Generator Lev 1/ steam generator l
,' & 11. Control Rod Po ion insertion limit
- g Limit Switc swi / rod I
- 12. DHR F1 Rate 1 g
M 13. Temperature 1 e
'[ . Auxiliary Feedwater Flow Rate 1
$.i 5
. L.L- .-._Jul dM. ;vl Ey I --it 1. .
1 .Im. . ...L. -
' -~
iBLE 4.3-6 ' t7 , I REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS i
' E.
- m. CHANNEL i
a hSTRUMENT CHECK CHANNEL __y CALIBRATIQr
- 1. wer Range Nuclear Flux M t
Q*
- 2. Inte diate Range Nuclear Flux M I Q* .
' ' '. 3. Source R e Nuclear Flux M Q* I
- 4. Control Rod ve Trip Breaker Indication M N.A. l
- 5. Reactor Coolant erature-Average M R
- 6. Reactor Coolant Flow te M R q I 7. Pressurizer Pressure M R s
- 8. Pressurizer Level M R Y
q 8 9. Steam Generator Level M R
]0) i -
- 10. Steam Generator P ssure M R
.l
- 11. Control Rod P ition Limit Switches R i 12. DHR Flow ate M R C
q 5 13. DHR mperature M R yi - , ., _ i ,
- 14. uxiliary Feedwater Flow Rate M ,
'. . e i+ y N
- Neutron detectors may be excluded from CHANNEL CALIBRATION.
l 21 1 lG
;e lw lbl ~
,) k)
)lW I l 14 14I. ..l . f
.i _ .I Ii
, Y INSTRUMENTATION _, M POST-ACCIDENT INSTRUMENTATION . LIMITING CONDITION FOR OPERATION 3.3. . The post-accident monitoring instrumentation channels shown Table 3. 10 shall be OPERABLE. p APPLICABILITY. MODES 1, 2 and 3.
ACTION:
- a. With the number f OPERABLE post-accide monitoring channels ,-
less than require Table 3.3-10, e er restore the , inoperable channel t PERABLE sta s within 30 days, or be in HOT SHUTDOWN within next hours. p
- b. The provisions of Specific 1 3.0.4 are not applicable, w
SURVEILLANCE REQUIREMEt 4.3.3.6 Each po -accident monitoring instrumentation chann shall be N demonstrated RABLE by performance of the CHANNEL CHECK and C NEL CALIBRATIO perations at the frequencies shown in Table 4.3-10. w
~
l 1 dilmen f B o B&W-STS 3/4 3-39 June 1, 1976 12E . ~" Ohveerh
=
O O O3
-I f TABLE 3.3-10
\'. !Q cn POST-ACCIDENT MONITORING INSTRUMENTATION MINIM *t C ELS
, MNSTRUMENT ERABLE
- 1. er Range Nuclear Flux 2
- 2. Reacto ilding Pressure 2
- 3. Source Range lear Flux 2 N
.4. Reactor Coolant Out Temperature 2 i y 5. Reactor Coolant Total Flow te 2
[ 6. RC Loop Pressure o 2 R[
- 7. Pressurizar Level 2 l
- 8. Steam Generator utlet Pressure 2/ steam generator
- 9. Steam G rator Level 2/ steam generator
{ 10. ted Water Storage Tank Level 2 h 1. Startup Feedwater Flow Rate i 't \ a M I T.
- a t
g ~ ~g.~g 3
~y [
f TABLE 4.3-10 h POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS R CHANNEL CHANNEL ___
.c .
MNSTRUMENT CHECK CALIBRATI W t-e i 1. er Range Nuclear Flux M
- i 2. Reacto uilding Pressure M R
- pd '
N . {Jy 3. Source Range lear Flux Q* ~18 4. Reactor Coolant Out Temperature M R S. Reactor Coolant Total Flow te M R R) R g 6. RC Loop Pressure M R
+
Y l
,' e 7. Pressurizer Level M R
- 8. Steam Generator Out Pressure M R
- 9. Steam Generat Level R
, 10. Borate ater Storage Tank Level M R i . E
- ; % 11. rtup Feedwater Flow Rate M R g 5 $
c ?
- i Neutron detectors may be excluded from CHANNEL CALIBRATION.
is l *t f f[:t
.S?
p.. r-. - - - -- - --
man e lt m I ' INSTRUMENTATION l N CHLORINE DETECTION SYSTEMS EtE 7E' " - s LIMITING CONDITION FOR OPERATION
,/ m 3.3.3. Two independent chlorine detection systems, with their al / trip A setpoin . adjusted to actuate at a chlorine concentration of < pm, '
shall be 0 ABLE. - APPLICABILITY: 2, 3 and 4. l ACTION:
- a. . With less than t chlorine detection ystems OPERABLE, within 1 hour initiate an aintain opera ~ n of the control room W emergency ventilation stem in e recirculation mode of operation; restore the 1 pera e detection system to OPERABLE status within 30 days or b at least HOT STANDBY within the nex; 6 hours and in COLD WN within the following 30 hours,
- b. The provisions of Sp fication 3. 4 are not applicable. ~
SURVEILLANCE REQUIREM S x 4.3.3.7 Each lorine detection system shall be demonstrated 0 performanc ABLE by f a CHANNEL CHECK at least once per 12 hours, a CHAN FUNCTION TEST at least once per 31 days, and a CHANNEL CALIBRATION W least ce per 18 months. 4 *'l s N e o v B &W-STS 3/4 3-42 January 1, 1977
<aab l
1 d
p [ 3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS , l LIMITING CONDITION FOR OPERATION l 3.4.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation. 1 APPLICABILITY: As noted below, but excluding MODE 6.* ! ACTION: MODES 1 and 2:
- a. With one reactor coolant pump not in operation, STARTUP and POWER OPERATION may be initiated and may proceed provided THERMAL POWER is restricted to less than ( )% of RATED THERMAL !
POWER and within 4 hours the setpoints for the following trips ! have been reduced to the values specified in Specification 2.2.1 for operation with three reactor coolant pumps operating: b
- 1. (NuclearOverpower3 I
- 2. 1 Nuclear Overpower based on RCS flow and AXIAL POWER
% IMBALANCEJ. I
- 3. 1 Nuclear Overpower based on pump monitors), i ;
- b. With one reactor coolant pump in each loop not in operation, STARTUP and POWER OPERATION may be initiated and may proceed provided THERMAL POWER is restricted to less than ( )". of j RATED THERMAL POWER and within 4 hours the setpoints for the .
following trips have been reduced to the values specified in Specification 2.2.1 for operation with one reactor coolant pump operating in each loop:
- 1. $NuclearOverpowerf. I
- 2. (Nuclear Overpower based on RCS flow and AXIAL POWER.
IMBALANCE (. I
- 3. { Nuclear Overpower Based on pump monitorsi.
i
- I See Special Test Exception 3.10.3.
l LL I B&W-STS 3/4 4 1 January 1, 1977 l l
( ( REACTOR COOLANT SYSTEM O LIMITING CONDITION FOR OPERATION (Continued)
"^0:0 2, i : d 5;
\
- s Oper: tier may precced previd:d at le::t : : r:::t:r :::1::t 1::p
- tr 4a eparaHea wi n aa associated -eacter c::lant pu p er decay
-hert c c:v:1 p rp.
b, The pr;vi:icn: Of Sp :iffecti:n: 3.0,2 :nd 2.0.t ;r: 300 :ppli;;bi:. l l i l SURVEILLANCEREQUIilEMENTS 4,4,1 The Reactor Protective Instrumentation channels specified in the applicable ACTION statement aboya shall be verified to have had their trip setpoints changed to the values specified in Speciftcation 2.2.1 for the applicable number of reactor coolant pumps operating either: ' a, Within 4 hours after switching to a different pump combination if the switch is made while operating, or b, Prior to reactor criticality tf the switch is made while shutdown. O O v-B&W-STS 3/4 4-2 June 1.1976
REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2500 PSIG + 1%*. I APPLICABILITY: MODES 4 and 5. ACTION: With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE DHR loop into operation in the shutdown cooling mode.
~
SURVEILLANCE REQUIREMENTS l 4.4.2 No additional Surveillance Requirements other than those l required by Specification 4.0.5. l
*The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
B&W-STS 3/4 4-3 January 1, 1977 1 .
(. REACTOR COOLANT SYSTEM " SAFETY VALVES - OPERATING LIMITING CONDITION FOR OPERATION
.& vl; 3.4.3 AM-pressurizer code safety valves shall be OPERABLE with a lift setting of 2500 PSIG + 1%*.
A APPLICABILITY: MODES 1, 2 and 3" ACTION: With one pressurizer code safety valva inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours. , (3 (# SURVEILLANCE REQUIREMENTS " 4.4.3 No additional Surveillance Requirements other than those required by Specification 4.0.5. ,
}
*The lif t setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
g Es, Iry 1o1/v Y
! Ied1 3 wd-lt c -oly en - v&lve I " ' 4 M< OP6Cth3LC /5 Glbwd/
w'Oi t 2 4. hca:c ,m ad ic. h.e , s j, s
,,,o,, ., k, 4o d c ps, a/nu e--
\
l l
\
(G%--g l' B&W-STS 3/4 4-4 January 1, 1977
.n l
(
\
,, - 3 '-- REACTOR COOLANT SYSTEM
~-
PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with:
- a. A steam bubble. !
isd!"W n a
- b. A' water Sjhe,l between (45) and (3o5 ) C'M; f t-l APPLICABILITY: MODES 1, 2 and 3.
ACTION: With the pressurizer inoperable, be in at least HOT SHUTDOWN with the ' control rod drive trip breakers open within 12 hcurs. SURVEILLANCE REQUIREMENTS \ 4.4.4 The pressurizer shall be demonstrated OPERABLE by verifying pressurizer level to be within limits at least once per 12 hours. l I l O. L]- B&W-STS 3/4 4-5 January 1, 1977 I 1 1
\
(. REACTOR COOLANT SYSTEM v STEAM GENERATORS l LIMITING CONDITION FOR OPERATION l l i 3.4.5 Each steam generator shall be OPERABLE.with ; ..;ter 10'/c! bet'?re" ( ) wu s j inum. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With one or more steam generators inoperable due to steam generator tube imperfections, restore the inoperable generator (s) to OPERABLE status prior to increasing T above 200*F.
avg h Uith nno n v- mn -a e+omm nonoratnre innnornhlo rie v o +n +ho w tor 10 /:1 being cut:f de thc'"-f tc , be i :t le::t M0T STffl0S ~
"!thi- 5 h0;r: and in COLD SML'TD0'a' within th; ncxt 20 h;;r .
SURVEILLANCE REQUIREMENTS v 1 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Sracification 4.0.5. 4.4.5.1.. Steam Generator Sample Selection and Inspection - Each steam I generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam \ generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be perfomed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except: a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas. v B&W-STS 3/4 4-6 June 1, 1976
( k i REACTOR COOLANT SYSTEM (T SURVEILLANCE REQUIREMENTS (Continued)
- b. The first sample of tubes selected for each inservice inspection l (subsequent to the-prc= r i:c inspection) of each steam generator I shall include: b w l.a ,s
- 1. All nonplugged tubes that previously had detectable wall penetrationsi>20%I.
- 2. Tubes in those areas where experience has indicated poten- ,
tial problems.
- 3. A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. If any selected tube does not pennit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
- c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
- 1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with im-perfections were previously found. ,
l
- 2. The inspections include those portions of the tubes where l imperfections were previously found. j The results of each sample inspection shall be classified into one of the following three categories: ,
fategory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes. yy B&W-STS 3/4 4-7 January 1, 1977
1 -( , - l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
/
C-3 More than 10% of the total tubes in'spected are degraded tubes or more than 1% of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must ; exhibit significant (> 10%) further wall penetratinns ! to be included in tne above percentage calculations. 4.4.5.3 Inspection Frequencies - The above required inservice inspections I of steam generator tubes shall be performed at the following frequencies: h . h m.
- a. The 'i w* i . -" 4 ~ inspection shall be performed *=da.mb Ms first Efft:ti;; Tull Pr.;;r ";n'h: but ithi- 21 ::1:nt r mentn: refad Of q h, sk 4"4t421 cr ticality. Subsequent inservi i performed at intervals of not less thar/%ce nor inspections more thanshall 24 be calendar months after the previous inspection. If two consecu-tive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation 3 has not continued and no additional degradation has occurred,
^
/ the inspection interval may be extended to a maximum of once per 40 months. ,_,.
- b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals fall into Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspec-tions satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a maximum of once per 40 months.
- c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
- 1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to tube sheet welds) in excess of ;
the limits of Specification 3.4.6.2. I
- 2. A seismic occurrence greater than the Operating Basis Earthquake.
1 V B&W-STS 3/44-8 January 1, 1977 ~.
(m
'( REACTOR r.00LAt4T SYSTEM JRVEILLAflCE REOUIREMEtiTS (Continued)
- 3. A loss-of-coolant accident requiring actuation of the ;
engineered safeguards.
- 4. A main steam line or feedwater line breat.
4.4.5.4 Acceptance Criteria
- a. As used in this Specification:
- 1. Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.
- 2. Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
(m
- 3. Degraded Tube means a tube containing imperfections > 20%
of the nominal wall thickness caused by degradation. _
- 4. % Degradation means the percentage of the tube wall thickness affected or removed by degradation.
- 5. Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing e defect is defective. !
- 6. Plugging Limit means tlie imperfection depth at or beyond which the tube shall be removed frora service because it may become unserviceable prior to the next inspection and .
is equal to 140J%* of the nominal tube wall thickness. 1
- 7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.
- 8. Tube Inspection means an inspection of the steam generator tube from the paint of entry completely to the point of exit.
*Value to be determined in accordance with the recommendations of Regulatory Guide 1.121, August 1976.
pr i V B&W-STS 3/4 4-9 January 1, 1977 I l
. l
( REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l l l
- 9. Preservice Inspection means an inspection of the full I length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
- b. The steam generator shall be determined OPERABLE after completing :
the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2. 4.4.5.5 Reports
- a. Following each inservice inspection of steam generator tubes, the number of tubes plugged enerator g#shall Fo be reported to the Commission "f tF'- 15 in each steam ty;.pt<w, 5p ,. /..u N,s L ,iy,3
- b. The complete results of the steam generator tube inservice inspection shall be inchded ia th; ?nnuci Op;r; ting ";;;rt f:r t'e per hd da 9ich thh 4 :pecths *.e: cc ph*ed This report "
shall include: repe d d p e r.;uj
- 1. Number and extent of tubes inspected. b 9% So /*m (e ./2.'/
- 2. Location and percent of wall-thickness penetration for each indication e' an imperfection.
- 3. Identification of tubes plugged.
- c. Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Comission shallbereportedpursuanttoSpecificationS".1%riorto 6.12 3 resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to detennine.cause of the tube degradation and corrective measures taken to prevent recurrence.
i.t.S.S The :;t;;;. ;;ncretor ;h:11 b; t .;n:tr:ted OPEP^"LE by '! r fy' ; i
;;;;;. ;;ncr;t;r I;=1 t; b; iJ.:.,1i;dt; ;t 1;;;t ;;;; p;r l' h;;r;.
a B&W-STS 3/4 4-10 January 1.1977
rx as Os a: s en
-t tn TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTLD DURING INSERVICE INSPECTION Preservice inspection No Yes
[ No. of Steam Generators per Unit Two e Twh Four g whQwo
};so ^" One First inservice inspection I Second & Subsequent inservice inspections One l e l One 2 Ok .
Table' Notation:
- 1. The inservice inspection may be limJtd to one steam generator on a rotating schedule encomp.. sing 3 N % of the tunes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a hke manner. Note that under some circumstances, the operating conditions an one or more steam generators may be found to be more severe than those in other steam generators. Under such circum stances the sample sequence shall be modified to inspect the most severe conditions.
2.';.-...,........,..........e; ...,,;;;;J d 'n; " ; . :::: . c . . :;; .. ;;:.;n .? ;? ? L; . , :._ .'
."H_ ':t g ._, .. c.; _ .. :d 'e::e.. O m . . . . .a . e. . ; J , . .L.; . ' ;,....
o ,""
^' 0 'ICO !" p T_ * : ': M O 2^'r'~
8"'
- 2. C ' ^ O ' '_' O O !' O' * ^;0 :* 0 2 . ; ^ r ? !^' : ^
- I ~'" * " 'i _ I "; * *' O ; " ' 3"
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. T h0 2^'^'d'"^"^* I':; r 'I^ ^:
a
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=
g TA8LE 4.4-2 di M STEAM GENERATOR TUBE INSPECTION 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N /A NA S Tubes per S. G. ' C-2 Plug defective tubes C-1 None NIA NA and inspect additional Plug defective tubes C-1 None 2S tubes in this S. G. C-2 and inspect additional C-2 Plug defective tutes 4S tubes in this S. G. Perform action for C-3 C-3 result of first sample Perform action for C-3 C-3 result of first N/A N/A sample M a C-3 Inspect all tubes in All other this S. G., plug de. S. G r are None N/A N/A A fective tubes and C-1 inspect 2S tubes in Some S. G.s Perform action for N/A N/A each other S. G. C-2 but no C-2 result of second additional sample Prompt notification S. G. are
- to NRC pursuant C-3
- to specification Additional inspect all tubes in N h*2.3 S. G. is C-3 eacn S. G. and pius
] defective tubes. Prompt notification N/A N/A to NRC pursuant to specification 5' 444= f./2 3
- s -
te S = 3 b Where N is the number of steam generator.. n the unit, and n is the number of steam generators inspectt4 n during an inspection 4 Q> I 4
REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE: rent.4
- a. The cent ;c ::nt b,,,/d^lmosphere a particulate radioactivity monitor-ing system. I b.
reache bn,/).,scl The cc-t :- :nt sump ievel and flow monitoring system. I r ecac.k.c ba . /da ,19
- c. ith , Tfie (cent:t ment :f ::c!cr cendentate c:> 2tc) er : cent 9 m nt & c:P c-1 gaseous radioactivity monitoring system.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: (b With only two bf the above r' quired leakage detection systems OPERABLE, operation maf continue for up to 30 days provided grab samples are obtained and analyzed at least once per 24 hours when the required gaseous and/or particulate radioactivity monitoring system is inoperable; otherwise be i in at least HOT STAN0BY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ; l i SURVEILLANCE REQUIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE l by: W B&W-STS 3/4 4-13 January 1, 1977
im i
\. REACTOR COOLANT SYSTEM p(.
SURVEILLANCE REQUIREMENTS (Continued) Re%/s,' C.,, /d.. .) " " '16 ' O )
- a. Centzi- ..: atmosphereparticulatMmonitoringsystem-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3.
- b. Reoc k..~ su C::tci me.nt 6,,M.,$p level tad 'i=. ::rit:r' ; :y:t:r perfonnance of CHANNEL CALIBRATION at least once per 18 months.
I
- c. 'spairy ;;;r:grict: :ur;;in:::: t ::: egendi,; uper t'e typc cf b akoge detecticr. ;j,tcm. ;ciccted.)
( v l l l t IO V B&W-STS 3/4 4-14 January 1, 1977 l 1 i
i r (% - (*- DEACTOR COOLANT SYSTEM fT V OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE B0UNDARY LEAKAGE. I
- b. 1 GPM UNIDENTIFIED LEAKAGE. I
+Ae.
- c. 1 GPM total primary-to-secondary leakage through V steam generators.2nd 'O.5) CP" t H^";" 2 y ^"^ !+^2- a^"^"'+^"
- d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System. I a f i eDM f*n'tTnnt t t s D a m s. t n .= (*nalsnt Rue +nm
!; K.fPRVAf'r y' m~ 1... : T. W4.;s"n , U ; L ~ - s+. . _
/
,, APPLICABILITY: MODES 1, 2, 3 and 4.
(j /W ACTION:
- a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT s'JANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any Reactor Coolant System leakage greater than any one J.EAKAGE, of the above reduce limits,rate the leakage excluding to withinPRESSURE limits withid BOUNDAR,)/ hours or-4 ., ,+ u .. unm m .inov .4.ue .u. __a e u_._. - ; e n. -
-!"'55:5'ItNE 55 75i'-Uh; 35"bE5.AU$v~[SUIra h of ca>ldca'n and M dr kcns v4 s/dlewn h he d d *emon<.d b, ;
ufe&y emelw&, ort he escA. cas a.sd ,~e.pc ched p.a svad h> , a. SURVEILLANCE REQUIREMENTS fpco fu 4en 6,/2,3 1 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be l within each of the above limits by: r en d2v" bar l a. Monitoring the c^-t9xnt a sphere particulate radioactivity i monitor at least once per 49-hours. I
- b. Monitoring the b Mt E8)p inventory and discharge at least once per4B hours. l B&W-STS 3/4 4-15 January 1, 1977 i
.s i
REACTOR COOLANT SYSTEM v SURVEILLANCE REQUIREMENTS (Continued)
-c .
" ::ur m:nt Of th: CC"T"CLLED LEAKAC: to the rc::ter :: !:^t
.ny n se31 7 usan +ho comr+nr enniant syet m nrasenra ie
'???0
, ?O,' p.fg :t lc;;t cac "cr ?! d:y: '"i th the "edu'at4"O I
">1"e-Tull3 ;;n.
- d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours during steady state operation.
( i V B&W-STS 3/4 4-16 January 1,1977
e l - m REACTOR COOLANT SYSTEM i b) CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-1. APPLICABILITY: At all times. I ACTION:
)
MODES 1, 2, 3 and 4.
- a. With any one or more chemistry parameter in excess of its Steady State Limit but within its Transient Lim,it, restore the parameter to within its Steady State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD i SHUTDOWN within the following 30 hours.
(' b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. At all other times: l With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady State Limit for more than 24 hours or in excess of s Transient Limit, reduce the Reactor Coolant System pressure to < psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the system pressure above blW&) psig or prior to proceeding to MODE 4. 6W SURVEILLANCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3. (,a - ( b ! 'O B&W-STS 3/4 4-17 Janua ry, 1, 1977 l
g(s
~
g ,, TABLE 3.4-1 - REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY STATE TRANSIENT PARAMETER LIMIT LIMIT DISSOLVED OXYGEN
- 0 10 ppm i 1.00 ppm 1 l CHLORIDE 1 0.15 ppm 1 1.50 ppm FLUORIDE 1 0.15 ppm 1 1.50 ppm
- Limit not applicable with T avg 1 250*F.
~
v B&W-STS 3/4 4-18 June 1, 1976
( .. TABLE 4.4-3 REACTOR CCA', ANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS SAMPLE AND PARAMETER ANALYSIS FREQUENCY DISSOLVED OXYGEN
- At least once each 72 hours CHLORIDE At least once each 72 hours FLUORIDE At least once each 72 hours
,. Not required with T avg 1 250'F.
B&W-STS 3/4 4-19 June 1,1976
( r' ( REACTOR COOLANT SYSTEM (] %..J SPECIFIC ACTIVITY v LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to: 3.5
- a. 4 1,G.uct/ gram DOSE EQUIVALENT I-131. I
- b. I k uCi/ gram. I APPLICABILITY: MODES 1, 2, 3, 4 and 5.
ACTION: MODES 1, 2 and 3*.
- a. 3. S*
With the specific activity of the primary coolant > 4,4-uCi/ gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown orr Figure 3.4-1, operation may continue for up to 48 hours provided that operation under these circumstances shall not exceed 10% of h the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.
- b. 3.s With the specific activity of the primary coolant > M uCi/ gram DOSE rQUIVALENT I-131 for more than 48 hours during one continuLs time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T . <
1500) F within 6 hours, avg
- c. With the specific activity of the primary coolant > Me/E 71 uCi/ gram, be in at least HOT STANDBY with T within 6 hours. avg < 1500) F MODES 1, 2, 3, 4 and 5:
- a. 36 With the specific activity of the gy,imary_ coolant > +re=
pCi/ gram DOSE EQUIVALENT I-131 or I'WG/E pCi/ gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. A REPORTABLE OCCURRENCE shall - be prepared and submitted to the Comission pursuant to Speci fication C ^. '. . This report shall contain the results of the specific activity analyses together with the following information: - f}' v
*With T avg > (500)*F.
B&W-STS 3/4 4-20 January 1, 1977
f'
'[~ -
REACTOR COOLANT SYSTEM C' ACTION: (Continued)
- 1. Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded. I
{ l
- 2. Fuel burnup by core region. I i
i
- 3. Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded. I i
- 4. History of de-gassing operations, if any, starting 18 hours prior to the first sample in which the limit ,
was exceeded. I
- 5. The time duration when the specific activity of the primary coolant exceeded WCi/ gram DOSE EQUIVALENT I-131. 3.6 SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.
pe \d. - B&W-STS 3/4 4-21 January 1, 1977
f TABLE 4.4-4 g PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE z w h AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND MODES IN WHICH SAMPLE AND ANALYSIS ANALYSIS FREQUENCY AND ANALYSIS REQUIRED
- 1. Gross Activity Detennination At least once each 72 hours 1, 2, 3, 4 '
- 2. Isotopic Analysis for DOSE 1 per 14 days 1 EQUIVALENT I-131 Concentration g 3. Radiochemicil for E Determination 1 per 6 months
- 1 s,
b 4. Isotopic Analysis for Iodine -et- Once per 4 hours, whenever d.2,3,4,5# # # # Including I-131, I-133, and I-135 the specific activity exceeds W pCi/ gram D_0SE EQUIVALENT g I-131 or 400/E pCi/ gram. 72
-F One ci 71: Set: r- ? a-d s 1 9 1 k n, , r c fal l nwi n,,
3 Turouay_
*^"EP change exceedi ; !5 per
-cent of th: "^TED "3IMAk
-P^"E". wits' a tre 'e f ra-iad' E
I
$" Until the specific activity of the primary coolant system is restored within its limits.
.
- Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since the g reactor was last subcritical for 48 hours or longer.
0 ( ( (
gs u .4 E
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- .i.i tiii E0 ' . i ,
LQs D ; ', ';' ,-. l UNACCEPTABLE ll_ 4 Wt' r "-I '
~
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+..-
i
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i si , [150 ~; ~ ~l ! l 5~~ I'l l z
._ ...> .ie ! x i
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i . l 1 i i 3 i (x E 5 i e ii' t l,6 l 6
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, ,. . I i % i i z ! 't ' ACCEPTABLE \ *'i
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11
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8 i i i ll i, o , 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER 1 FIGURE 3.4 'i l DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus I Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 9:9pci/ gram Dose Equivalent 1-131 3.5 O v L, B&W-STS 3/4 4-23 June 1, 1976 l
(s. REACTOR COOLANT SYSTEll 3/4.4.9 PRE 3SURE/TEi1PERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION o.nd 14 3.4.9.1 The Reactor Coolant / System (except the pressurizer) temperature andpressureshallbelimiJte in accordance with the limit lines shown on Figures 3.4-2,eee 3.4-Tduring heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of jl00l'F in any one hour p9riad. I
- b. Amaximumcooldownof{100}*Finanyonehourperiod. I APPLICABILITY: At all timesf ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the i fracture toughness properties of the Reactor Coolant System; determine - that the Reactor Coolant System remains acceptable for continued operation j or T be in at least HOT STANDBY within the next 6 hours and reduce RCS ' and pressure tile 9 following to less than 200 F and 500 psig, respectively, within 30 hours. i SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be dptermined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice ir ' id hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-S. The results of these examina-tions shall be used to update Figures 3.4-2 and. 3.4-3 7 and 3,4-4, N or P lv d ses a.a cedan n w , y ja ( p , g o, ppp
" "* ' #fflY G bode 525*q ,,< H,and B&W-STS 3/4 4-24 January 1,1977
~
u l I I I I I I I I I . C p F 2400 _ ARKANSAS NUCIIAR ONE - UNIT 1 - J aO 52200 - rossi esEss ifwr - E pel *f 0 2 Atr'LICAILE FOR RfATUP RATES
'57 8' r 2000
~
a a ses sas
,, < : o o =,f , -
3 e ses ass ;< so*f lx Asr 1/2 mosa PEstos)
;; 1800 -
o 2sco 2:2 - - n E 3305 244 E F 2588 tes d 1600 -
"? ArrLICAILE FOR C00LBOWS 841E8 SF E 1400 - < '" 'I'" -
g i << .. , I .. ii, .0.. ,E. ..> sa$1200e . Z 1000 - _ f 800 - a 8 U C j $00 - 33, ,,,,,g,,,, ,,,,,,,,,g,,,,,,g,,, ,,,,,,,,,,,, ,,, g,,,, ,,, g, g,, ,,,,, ,, g,, ,,,,, _ o serve (s). The limit serves leclude the pressere differential between the pelat of erstes ha 400 - A "**** *****'***** *"8 * " ** 'b' ****' ***3 'eslea coattelline the limit
~
serve, they laslede an addittenet mercia of eefety for poselble lastremost errer (25 pels 3 and it*f). For coeldews. metes I and 2 en flgere 3.1.2 3 are applicable.
- 200 -
l I I I I I I I I 40 to 120 ISO 200 240 280 320 3c0 . 430 440 laatsates Reactet Coolant Tempefatuts Tc, *f - l N
. AEACTOR COOLANT SYSTEM INSERVICE IfYDR0 STATIC TEST llEA1UP AND COOLD0tfN s'
, LlulTATIONS APPLICA0LE TOR flRST 5.0 EFFECTIVE FULL toter TEARS
. Figufs 3.+:2 4 i
i
- i .
.l i . -
l l Y ', ,l
.jf 2400- I I I I i 1 I I l$
ARKANSAS NUCLEAR ONE - IINIT I , , _ 2200 - - 3 POINT Pet 33 ftWP. _ g 2000 - ,,1 *r
= A ess so 8" 85' _
$D 1800 - 8 Se4 283 C
9 2250 312 o" __ 5 1600 r o 2:3 F set 183 APPLICAttt r0B Et1Ter a17't3 a .s u 323 u < lu'r/u
" 34,g _
Sst s anse g ,,- (< so'r in AnY lla sana reales) 9 . - k' 3 1200
; 1000 -
a CtlTICALITY a Llulf (tron) _ 0 800 -
'
- e
$a 600 -
t ec a The assestante pressere-temperatereteeabloatione are below and tw the right of the llelt 400 A serve (s). The llelt serves laclude the pressere dif ferential between the pelat of erstee k pressere eenserement and the pressera en the remeter vessel reglen centreilles the tielt E cerve, they lacInde se additlenal eersia of esfety for possible les trement errer (25 pel _ g,, ,,,,,,,,,,,,,,,,,,,, 00 -
,,,i.e,y, in, ,,,,g,, ,,,, ,,, g, ,,,, ,,,,,,,, ,,gi, combinettene are to the right of the arttiselity fielt. -
I I I l El 1 I 0 1
~
280 320 360 400 120 160 , 200 240 40 00 thsicated Reactor Coolant Temperature. Te, 'F REACTOR COOLANT STSTEN, NOREAL OPERATION-MEATUP LIMITATIONS. APPLICABLE FOR FIRST 5.0 EFFEcilVE FULL POWER TEARS l ! Flpure 3.4. -
a 48 s, s - % .= (, .* a S g
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( ' O 6 o i0, TABLE 4.4-5 d REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE s
.Ahnl DRAwI)L SPECIMEN CAFA/LE /AnastnoAslEP""_
".E"'".' r.L ! " .
'1. A N[- &~ flas b* u w.thdcawn fa.- Fe F,,y,
- 2. ANr-5 ss>> //,dara fe// car,n3 ) Nuic.le.al Daes.s ~ 6e s.s e 1 w 3. A Nr - A +'s Iladnd 4//w..,3 3 BIcya.la af y Du.1- sas.>e z
? 4. A ^JZ - C & l W *
- E IIc<u' y 7 0 eycle al .
w Da s,s - B e,s c f
- 5. A tH - D /nsech on /ccA hon W 2 barpe d yr,cr i-t
+ " <-We a F %0 >- s e-se 1; w.14 </.ur fa/A ..,3 sztb cy,,/m,
- 6. A tJE - F l'I Y'
** lM b* b n YEb'i'P e ') P r'0 f o ^
e 4- cyde cd Da vi., - 8 esse. f . tJJk je.w i s' full~.o.a3 II tL c.,,,je,
=
E Q , O 4 em
.('I, ,
REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITIO3 FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
- a. A maximum heatup and cooldown of 100*F in any one hour period. l
-b. A maximum spray water temperature differential of ( F. '
l A PLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within limits within 30 minute.;; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce pres-surizer pressure to less than 500 psig, within the following 30 hours. v 1 SURVEILLAMCE REQUIREMENTS 4.4.9.2 The pressurizer temperature shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential s. hall be determined to be within the limit one.e per 12 hours during auxiliary spray operation. s
- r'] '
Q B&W-STS - 3/4 4-28 January 1,1977
l r 4 REACTOR COOLANT SYSTEM - (O 3.4.10 STRUCTURAL INTEGRITY I ASME CODE CLASS 1, 2 and 3 COMPONENTS l l LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1. APPLICABILITY: All MODES. ACTION:
- a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT considerations.
- b. With the structural integrity of any ASME Code Class 2 component (s) j d' not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or 1
isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.
- c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the component (s) to within its limit or isolate the affected component (s) from service.
- d. The provisions of Specification 3.0.4 are not applicatie.
SURVEILLANCE REQUIREMENTS 4.4.10.1 In addition to the requirements of Specification 4.0.5:
- a. The r:::*^r ::chr,t purp flyd.;;;; sh;11 b; h;p;;t;d p;r th; rc: n = t t!::: Of R:;;ht:ry r;;iti;r, C. '.t. ;f r,;;ul;ta,,, --
Ocid 1.li, S:;hur,1, ".;;;;t l';-'C. m)_: B&W-STS 3/4 4-29 January 1, 1977 l
('.( REACTOR COOLANT SYSTEM o SURVEILLANCE REQUIREMENTS (Continued)
- b. Each internals vent valve shall be demonstrated OPERABLE at least once per 18 months during shutdown, by:
- 1. Verifying through visual inspection that the valve body and valve disc exhibit no abnormal degradation. I
- 2. Verifying the valve is not stuck in an open position. I
- 3. Verifying through manual actuation that the valve b:;in:
t: ? r "ith : fer:: equi::!:nt te - 0.M) p;id 2nd is fully open with a force :;tini::t I:(; (0.30) ;;id. o f S 9 w Ibg (appled veekea tfy up,va.nl,)_ 3
)
v i
,e*
B&W-STS 3/4 4-30 January 1,1977 l (
('.
\ , 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) y CORE FLOODING TANKS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system core flooding tank shall be OPERABLE with:
- a. The isolation valve open. I b.
mlakl leve./ of if.c r c,4 Anter' 'n d 5:.;ted water "<"== ha+'"aa / i =d / i 6 feet of borated water. 1 A m ,s,, . beren cenc.en feh ko, c f Z270
- c. ar mer,4,2270) :rd /
, ) ppm of boron. I
- d. A nitrogen cover-pressure of between (575I and (625l psig.
APPLICABILITY: MODES 1, 2 and 3*. ACTION:
/ a. With one core flooding tank inoperable, except as a result of h' a closed isolation valve, restore the inoperable tank to OPERABLE status within one hour or be in HOT SHUTDOWN within thenextghours.
- b. With any core flooding tank inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in "0T STaMDSY ith' :rs S: r :nd 5: ', HOT SHUTDOWN within the next -4 hours.
56 SURVEILLANCE REQUIREMENTS 4.5.1 Each core flooding tank shall be demonstrated OPERABLE:
- a. At least once per 12 hours by:
1. hvol Verifying the contained barated water "^'" and nitrogen cover-pressure in the tanks. 1
-2. '!:r'fying th:t :::h t:rk icehtf er ::h: i: ;;; ,.
*With Reactor Coolant pressure > (800T psig.
f t &y B&W-STS 3/4 5-1 January 1,1977
f' EMERGENCY CORE COOLING SYSTEMS O ( ,/ SURVEILLANCE RE0UIREMENTS (Continued)
- b. At least once per 31 days and within 6 hours of each solution volume increase of > fl% of tank volumel, by verifying the boron concentration of the tank solution. t
- c. At least once per 31 days by verifying that power to the isolation valve operator is disconnected by removal of the breaker from the circuit.
-d. it ic::t :n:: ;;r 19 :: nth: b; v;ri'ying th:t ::F :r 'l : dins vlmmk!a:f 't r4*5::!:tf:r / 1 '/ ^r2 : 2rt^ :ti c y :-d 4: 4 'e ' ':d l t c' sf- : . ;_ _ _ ! 7_ -l7 ,
7, a uwg n- t - gmg. twg ogggtg. cggyggt gy ggg,797 rgg =g r- g-l 3 M p.n n rara 4 7+ nr 3 fes# 7ty infagtign tget} eta-,1 i i
\ l V
e 1 . ' \_^- h) i B&W-STS -- 3/4 5-2 January 1,1977 s . . _ _
(' EMERGENCY CORE COOLING SYSTEMS P ECCS SUBSYSTEMS - T an > J Z )^F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each I subsystem comprised of:
- a. One OPERABLE high pressure injection (HPI) pump. I
- b. OneOPERABLElowpressureinjection(LPI) pump.
l
- c. One OPERABLE decay heat cooler.
annunedi,sa
- d. An OPERABLE flow path capable of taking suction from the Om m Mkg
' borated water storage tank (BWST) on a safety injection signal ('-cal and Oute :tf 11; tr::.:ftr+g cuttf or to th: : n t: f --^ n t 27 ' ' '^
on a borated water storage tank low level signal & d "; the recirculation phase of operation. le e.nte APPLICABILITY: MODES 1, 2 and 3. ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
- b. In the event the ECCS is actuated and injects water into the '
Reactor Coolant System, a Special Report shall be prepared and submitted to the Comission pursuant to SpecificationJ. . 6. /2. 3 within 90 days describing the circumstances of the actuation
~
and the total accumulated actuation cycles to date. ' - J. . 7 ) 1 n Np 1 B&W-STS 3/4 5-3 January 1,1977
[ EMERGENCY CORE L ) LING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At least once per 12 hours by verifying that the followin ves are in the indicated positions with power to the val operators removed:
Valve Number Valve Function VaTve Position .
- a. '
b, c. f
- b. At least once per days by:
- 1. Verif g that the LP injection cross-o loc d, sealed or othemise secured in the valves are throttled sition. "
Verifying)that automatic in the eachflowvalve path(manual, that is notpower operated l'ocked, seale o
, otherwise secured in position, is in its correct position.
A-e. By a visual inspection which verifies that no loose debris j (rags, trash, clothing, etc.) is present in the :: ti- st re& bc/o,scJ f x nt sump and cause whichcouldbetransportedtothecery'LOCAconditions restriction of the pump suction during . This visual inspection shall be performed: C re.acA, 6.M.sg
- 1. For all accessible areas of the N- -
p$1orto establishing 60&ThiNME#T INTEGRITY. I R@cice eatuoix j .-
- 2. Of the areas affected within s,uy g- zt ba,lds at' the completion of each eatW-established. *#*:'t eng'y hi when GOMT*fNMENT- IN'EGRITY is -
f Recgog gg,co,n.3 b 4. _ At least once per 18 months by: 1. Verifying-automatic isolation and intefrloek action of the DHR system from the Reactor C.oolant System when the Reactor Coolant System pressure is > 400 psig. B&W-STS 3/4 5-4 January 1,1977
f%.
,' EMERGENCY CORE COOLING SYSTEMS
_-SURVEILLANCEREQUIREMENTS(Continued}__ 2. A visual inspection of the& buildly which verifies that the subsystem suction not inlets a restricted racks, screensby debris and that the sump components rash distress or cor,rosion.etc.) show no evidence of structural Verifying he4PI system a total at: leak rate < (6) gallonsurper for ho y a) Normal o I ing pressure of 9 parts of the sy d 1 psig for those u. isolation valve eam of the pump suction g b) > psig for the piping from the inment emergency isolation valve. sump isolation valve to the pump on C.c At least once per 18 months, during, by shutdown > 1. Verifying that each automatic valve in test { signal. actuates to its correct&fo, position on a (;the flownp i.j;;th,c 2. Verifying that each HPI and LPI pump test-signal. matica11y upon receipt of a 4::f:t; starts auto-
'nj:;tba test}
J tn wy
- f. By s crif,:ny that each c' r a4
- g. >
f;~. LPr i .dM:t d f M^r', ;rc.;;;. th; ;r. ;, QMWAwLE* r;L :ni pumpf-t ::h;; the [A t ~ tested pursuant to Specification 4.0 .. 5rir;;hti r "5 when
-1. -
lll;h pr;;;;c; inj:: tic, pts; ;
.s
_ ;;i;.-
- 2.
Lo' W :ur; hjecticn p; p ; _ _ p;i;. - C ( t B&W-STS O 3/4 5-5 January 1, 1977 e . . ., p gy +
......._ 4w.-
, :r ,
,, p . . . k y .,_, . , .
4 g. g .n.< 1 1
. . . . _ . . y;
.^
} EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T < )"F yg v
LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
- 2. Cr. : OPEME'.E 'if p ccru e 4-jectie- ("O pu=p - '
8- A- One OPERABLE low pressure injection (LPI) pump. I b ,,, One OPERABLE decay heat cooler. I t--di An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) and transferring suction to the containment emergency sump. APPLICABILITV: MODE 4. ACTION:
- a. With no ECCS subsystem OPERABLE because of the inoperability of !
either the.""! pt ? cr the flow path from the barated water d storage tank, restore at least one ECCS subsystem to OPERABLE v i {' status within en+(> hour or be in COLD SHUTDOWN within the next 40-hours. l 36
- b. With no ECCS subsystem OPERABLE because of the inoperability of either the decay heat cooler or LPI pump, restore at lett one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T less than (Wes-)*F by use of alternate heat removal methods. avg zco
- c. In the event the ECCS is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and submitted to the Commission pursuant to Specification E.0.2 64 3 i within 90 days describing the circumstances of the actuation l
and the total accumulated actuation cycles to date. l SURVEILLANCE REQUIREMENTS
.,. )
- 4. 5. ~5 The ECCS subsystems shall be demonstf1ted OPEPAE per the applicable Surveillance Requirements of {4.5.2'l. '
i f ~
~
B&W-STS 3/4 5-6 January 1, 19/7
h(,
- E_MERGENCY CORE COOLING SYSTEMS O BORATED WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The borated water storage tank (BWST) shall,be OPERABLE with
mduhd AN .ti;d b;r;t d water ;:!:::les e / 37.5 0.' c, feed. {36 ) a. o f ha +""^ " (M 0^0'3>'"d
*** * '** ***1*I
( ) ;;llene. I A c. chf*ff*c"7 .247o 2 208
- b. 2 tucer (1800) d ' l- ppm of boron. I
- c. A minimum water temperature of ( ) F. l APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION: With the vorated water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.5.4 The BWST shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1. Verifying the<infic+W IMI cr m ae_ torates water ;;1 :: in the tank. l
- 2. Verifying the boron concentration of the water.
b. At least once per 24 hours by verifying the water tem'perature when (outside) air temperature < 468F. 4D fh L'"J B&W-STS 3/4 5-7 January 1,1977 } H h .
-}}