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Proposed Changes to Tech Specs & Bases to Reflect Changes to Correct Typos & Administrative Errors
	ML20236K506
Person / Time
Site:	Davis Besse
Issue date:	11/02/1987
From:	; TOLEDO EDISON CO.
To:	;
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ML20236K470	List: ML20236K466; ML20236K488; ML20236K506;
References
	1407, NUDOCS 8711090248
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{{#Wiki_filter:INou A A<dt $ 07 i LIMITIitG CCMDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 'I 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AX I A L POW ER IM3 ALAN C E....................... -........ 3/4 2-1 3/4.2.2 NUCLEAR HEAT FLUX HOT. CHANNEL FACTOR - F................................ 3/ 4 2-5 ~ g '3/4.2.3 NUCLEAR ENTHALPY RISE y HOT CHANNEL FACTOR - F,g........................... 3/42-7 3/4.2.4 QUADRANT POWER TILT.................................. 3/42-9 3/4.2.5 DNB PARAMETERS....................................... 3/4 2-13 3/4.3 INSTRUMENTATION 4 i 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............ 3/4 3-1 3/4.3.2 SAFITf SYST&.5 In5TRUM!hTATION S a fe ty Fea tures Ac tua ti on Sys te:n..................... 3/4 3-9 Feedw e.v-Stenm and Rupture Con:rol Sys:12................ 3/4 3-E3 Anticipa tory Reactor Trip Sys tem................... 3/4 3-30a 3/4.3.3 MONITORING INSTRUMEhTATICN Ra di a tien Moni tori ng I ns t r=2n:s ti on................. 3/4 3-31 Ine:re 0*tecters..................................... 3/4 3-33 S e i s=i c Ins tr=enta ti on..............................3/4 3-37 Me te:r:l ogi cal Ins tn:menta ti on.................... -.. 3/4 3-40 . Remote Shutdown Instru:nenta ti on... ;.................. ' ' 3/4 3-4 3 Po s t-Acci d ent Instnnnentati on...................'..... 3/4 3-45 Chl ori ne Detecti on Sys t ers......................'...... 3/4 '3-51 Fi re Dete cti on Instruments tion........... 3/4 3-52 Radioactive Liquid Effluent Ponitoring Instrumentation 3/4'3 ~ 1%dioacMve. Gase.ous Effluent Meni+oring InskumenkHot 1/4 3-@, 3/4.4 REACER COOLANT SYSTEM 3/4.4.1 CDOLAh7 LOOPS AND COOLANT CIRCb.ATION S ta rtup an d Powe r 0pera ti on............................ 3/4 4, Sh ut down an d Ho t ' 5 tan dby... '.......................... 3/44-2 -i 3/4.4.2 SAFETY YALY ES - SHUTDOWN.............................. 314 x 3 a l 3/4.4.3 '5AFITY VALVES AND ELECTROMATIC ' ELIEF YALYE '0PERATING 3/4 4-4L R 8711090248 B71102 DR ADOCK 0500 6 .IV ; Amendment No. 4. 86-o 4 a

V &ust M07 QR r L INDEI' L!f4tTING CON 0!TIONS FOR OPERATION AND SURVE!LLANCE REOUIRD4ENf5 SECTION PAGE 3/4.5 CCNTAINNENT SYSTP.5 i e 3/4.3.1 PRIMARY CONTA!NNENT Co n ta i nme n t I n te g r i ty................................. 3/46-1: Co nta i nmen t L e a ka g e.................................. 3/4 6 Containment Air Locks................................. 3/4'6-6 I n t e rn a l P r es s u r e.................................... 3/4 7 A i r T em p e ra tu r e...................................... 3/4 6-8 Containment lessel Structural Integrity............. 3/4 L 6-9 Conta inment Ventila tion 5ystem....................... 3/4 6-10 3/4.6.2 OEPREISURIZATION AND CCOLING SYSTEMS Conuinment So ray ~ 5ystem............................. - 3/4 6-11 1 Conta inment Cool ing Systas........................... 3/4 6-13 j 3/4.5.3 CONTAINMENT !!OLATION - VALVE 5......................... 2/4,6-14 3/4. 5. : ' COMSU5TIELI GAS CONTRCL h l Hyd ro g en An a l yz e rs................................... 3/4 6-23 D e l e t ed......................... /.................... 3 / 4 6 - 2 4 - Containment Hycregen Oilution 5ystas.................. 3/4 6-25 Hy d re g e n. Pu rg e S y s t em................................ 3/4.6-26 .) 3/4.3.5 C .'P.'?'" 'C."* * : "T '* SHIEt.D, BunLDIPI(q l 1 Emerg ency Ventil a tt en 5ys tz.:......................... 3/4 6-23 Shiel d Bu il di ng Integri ty....................,........ 3/4 6-31 Shiel d Bu il ding Structural Integri ty.................- '3/4'5 32 3 9 OAVIS-3 ESSE. UNIT 1 V! . Anenment Ho. Ja,: 56- ~

ADDR10NAL CHANGES PREVIOUSLY 4NM N7 I i PROPOSED BY LETTER-INDEX 3 Serial No. ' /306 Date 5-20-87 ~ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0UIREMENTS SECTION_ ,PAGE 3/4.7 PLANT SYSTEMS q 3/4.7.1 TURBINE' CYCLE S a f e ty Va 1 ve s........................................., 3/ 4 7-l ' Auxil ia ry Fe edwa te r Sys tem........................... 3/4 7-4 j Condensate Storage Tank.............................. 3/4 7-6' ') Ac t i v i ty............................................. 3 / 4 7 - 7 Main S team Line l's ola ticn Valves..................... 3/4 7,9 l [. c .,.... ~ <.~.... m............................. m - j 3/4.7.2 STEAM GENERATOR PRE 55URE/ TEMPERATURE LIMITATION...... 3/4 7 l i 3/4.7.3 COMPONENT COOLING WATER 5YSTEM........................ 3/4.7-14 3/4.7.4 SERVICE WATER SYSTEM................................. 3/4 7-15 3/4.7.5 ULTIMATE HEAT SINK................................... 3/4 7-16 3/4.7.6 CONTROL POOM EMERGENCY VENTILATION SYSTEM............ 3/4 7-17 '3/4.7.7 !;"TA L "O SNU8BERS.......;........................... 3/4 7-20 l j 3/4.7.8 S EALED SOURCE CONTAMINATION........................... 3/4 7-36 l 3/4.7.9 FIRE SUPPRESSION SYSTEMS Fi re Supp ress ion Wa te r Sys t em......................... 3/4 7-38 Spray and/or Sprinkler Sys tem........................ 3/4 7-42 Fi re Ho s e S t a t i o ns.................................... 3/4 7-44 3/ 4. 7.10 PEN ETRAT ION FI RE B ARRI ERS............................. 3/4 7-47 3/4.8 El'ECTRICAL' POWER SYSTEMS 3/4.8.1 A.C. SOURCES O pe ra ti n g........................................... 3/ 4 8 - 1 Shutdcwn.............................................. 3/4 8-5'. ~ 3/4.8.2 CNSITE POWER DISTRIBUTION SYSTEMS. A.C. Di s tribution - Opera ting........................ 3/4 5-6

A.C. 0 distribution - Shutdown..........................'3/4 6-7 0.C. Di s tribu tion :- Ope ra ting....... 2................ 3/488 1

D.C. Di stribution - Shutdown.......... J....-.......... 3/4 8-10 I -DAVIS-BESSE, UNIT 1 yII . Amendment No.-38

{

] x 1

Ad4 :No7 ADDiliONAL CHANGES FREyl00 SLY g = PROPOSED Bf LETIER Serial fio.J3__85____ Date 5-20 87 !! CE.1 BASES 1 SECTION PAGE 3/4.7 Pl. ANT SYSTEMS - i TU R8 t N E CY C1.E.................................. ..i 3/4.7.1 STEAM GENERATOR PRESSURE /TU!PE?.ATURE LIMITATICN.- 3/4.7.2 3 3/4.7.3 COMPONENT COOLING WATER SYSTEM.................. 8 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM............................ 8.3/4.7-4 I 3/4.7.5 UL T IMATE HEAT S IN K....................... -....... 8 3/ 4 7-4 3/4.7.5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM........,8' 3/4 7-4 3/4.7.7 "rm IC SNUB 8 ERS............................... 3 3/4 7-5 l 3/4.7.8 SEALED SOURCE CONTAMINATION......................., 8 3/4 ' 7-6 ] 3/4.7.9 FIRE SUPPRESS ION SYSTEMS..................'....... 8 3/4 7 3/4.7.10 PENETRATION FIRE 8 ARRI EiW........................ 8 - 3/4 7-6 3/4.8 El. ECTR ICA L POWER SYSTEMS........................... 3/4.9 RE JELING PERAT*0NS - 1 3/4.9.1 BORON CONCENTRATION............................. 8 :3/4 9-1 3/4.9.2 IN S TRUMENTATION.......,........................... S 3/ 4 9-1 3/4.9.3 O E CA Y T I ME....................................... 8 3 3/*-1. 4 CONTA INMENT PENETRAT ICNS........................ 3 3/ 4 ?-1 3/4.9.5 CCMMUN I CAT I ONS................................... 3 - 3 0 AVIS-BESSE, UNIT 1 XII Amendment No.J33 f ...g m- _________.__m_____ a

NJAA& }'/0 pp 5 ' o 4 INDEX ADMINISTRATIVE CONTROLS 1 SECTION PAGE 6.1 RESPONSIBILITY.......................................... 6-1 6.2 ORGANIZATION 0ffsite.............'................................... 6-1 1 Fa c i l i ty S ta f f.......................................... 6-1 4r44 I Facilih staff ovemrne 6.3 FACIL ITY STAFF 0 VAL IFICATIONS........................... - 6-5' '6. 4 TRAINING................................................ 6-5 6.5 REVIEW AND AUDIT w

6. 5.1 STATION REVIEW BOARD Function..............................................

'6-5 i

?. position..........................................

5-6 Alternates............................................ 5-6 6 Meeting Frequency..................................... l .6-6 l Qu o rum................................................ i Responsibilities......................................- 6-6 5-/ 8 ( Authority............................................. 6-3 R ec o r d s............................................... 6.5.2 COMPANY NUCLEAR REVIEW BOARD q u F u nc t i o n.......................................'.......:6-8. 1 ~6-9 C om po s i t i o n........................................... Al t e rna t e s.............................'............... 6-9 C o "n s u l ta n t s........................................... L6-9 g + i - DAVI5-BESSE. UNIT 1 ' XV . Amendment No. p 1 3' I d~

g,4y gp j ( INDEX ci!NTSTRATIVE CONTROL $ !.ECTION \\ PAGE Me e ti ng F re q ue ncy...................................... 6-9 t Quorum................................................ 6-9 Review................................................ 6-10 i i Audits.................*................................ 6-11 Authority............................................. 6-X 12 l Records............................................... 6.12 EVENT 5.6 R EP ORTABL E 464iaAG665. ACTI0N............................ 6-12 [j $. 7 SA F ETY L IM I T V 10LATI ON................................. 6-13 i L.8 PROCEDURES......................................,....... 6-13

f. 9 REPORTING REOUIRESENTS RoUTIN E.

5. o..

m-ue'.....-R.Er0RT3 m., ~......,m..m-........... 6-148- .i

6. 9. 2 SPECIAL REPCRTS.......................................

6-18 6.10 R EC O R D R ETEN T10N....................................... 6-183 l 6.11 RADIAT*0N PROTECTION 9 90 GRAM........................... - 6-20 5.12 H I G H RAD I A T I O N AR EA.................................... - 6420 8.13 ENVIRONMENTAL QUAL IFICA T10N....................... '6-21 6 14 Process cWTeoc-PRoderAM (PcP ),.............. G' 23-3 15 0FF51TE. Dost CALCULATION M ANUAL (occ M ),< ~ $ - a. 2.,

g. l 4 M AJoR - c H ANC2Es Tb RADicACTIVs L(aut0. 6ASE005

-~ -app Sott o w ASTE TREATMENT SYSTEMS _... 6-13 ) OAV!!-5E!!E. UNIT 1 XVI Amendmen t'.No. 38. 2

A%/Ydk lpp 7[ ' ) '3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREME l' i ,3/4.0 APPLICABILITY LIMITING CONDITION FOR~0PERATION q 3.0.1 Limitin0 Conditions for Operation and ACTION requirements shall be- ~ applicable during the OPERATIONAL MODES or other conditions specified for each specification. 3.0.2 Adherence to the requirements-of the Limiting Condition for ;0pera-tion and/or associated ACTION within the specif'ied 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 Wen a Limiting Condition for Operation is not met, = except as provided in the associated ACTION requirements, action shall be initiated within I hour to place the unit in 4 MODE in which the Specification does not apply to placing it. as applicable, in: 1. At least BOT STAND 5Y vithin 6 hours, 2. At lesse 30T SHUTDOW within the f allowing 6 hours, and 3. At least COLD SBUTDOW vithin the subsequent 24 hours. Were corrective measures are completed that permit operation under the ACTION requirements,. the ACTION may to taken in accordance with the specified time 11mics ss seasured from the time of failure to meet the Liniting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications. 3.0.4 Entry into an OPERATIONAL MODE or other specified applicability I condition shall not be made unless the conditions of the Limiting Con-i 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 ACTION statements. 3.0.3 Wen a systen, subsystem, train, component or device is determined co be inoperable solely because its emergsney power source is inoperable. ( or solely because its normal power source is inoperable, it may be considered oPERABLZ for the purpose of satisfying the requirements of its ~ applicable Limiting Condition for Operation provided 'i corresponding normal or ' emergency power sour,ce is OPERA 3tI(1);its of its redundant system (s), subsystea(s). train (s), compone;nt(s)1and-and.(2) al.1' device (s) are OPIRA3tI, or likewise satisfy: the requirements of this -j specification. Unless. both conditions. (1) and (2) are satisfied, vichin i 2 hours action shall be initiated to place the unit in a MCDE in which the applicable limiting Condition for Operation does not apply by placin J it as applicable in ~ g 1. At least HOT STAND 37 within 6 hours, 2. At least HOT SUUTDOW within the following 6 hours, 'and 4 3. At least COLD SHUTDOW vi"Ehin the subsequent 24 hours. Ihis Specification is not applicable in MODES $ or 6. - 1 s DAVIS-BESSE, UNIT 1 3/4 0-1 Amendment No. 71-i

/lf07 @P i l l REACTIVITY CONTROL SYSTEMS MAKEUP PUMPS - OPERATING LIMITING CONDITION FOR OPERATION l 3.1.2.4 Two makeup pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one makeup pump OPERABLE, restore the inoperable pump to OPERABLE status within 72 hours or. be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1% ak/k at 200*F within' the next 6 hours; restore two pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. j i SURVEILLANCE REQUIREMENTS 4.1.2.4 In addition to the requirements of Specification 4.0.S, [ l at least two makeup pumps shall be demonstrated OPERABLE at least once per 31 days by: i a. Starting (unless already operating) each pump from the control room. b. Verifying, that on recirculation flow, each pump develops a discharge pressure of > 2400 psig. ~ c. Verifying that each pump operates for at least 15 minutes. d. Verifying that each pump is' aligned to receive electrical power from separate OPERABLE ossential busses. With RCS. pressure > 150 psig. DAVIS-BESSE, UNIT 1 3/4 1-10

gM gg[ ] pgy ] t REACTIVITY CONTROL SYSTEMS q ACTION: (Continued) c) Apowerdistributionmgpisobtainedfrom-theincore detectors and F and F are - their limits wi9hin 72 Nours. verified to'be within ]q 3 d) ' Either the. THERMAL POWER. level is. reduced. to I, 60%

l of the ' THERMAL POWER allowable for' the reactor..

'o . coolant _ pump combination within one hour and within i the next 4. hours -the High Flux' Trip Setpoint is 1 reduced.to d 70% of the-. THERMAL P0WER allowable for. the reactor coolant pump-combination,'or e) The remainder of the rods in.the group' with the.. inoperable rod are: aligned to :within + 6.5% of. the inoperable' rod within one hour while maintaining the-rod sequence.1inser. tion and overlap limits ~ of Figures 3.1-2-and 3.1-3; the THERMAL' POWER leveli 1

s. hall be restricted pursuant' to Specification >

j 3.1.3.6 during' subsequent operation, j j 1 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 I positions at least once per 12 hours.except during_ time interval' :when - s the Asymetric Rod Fault Circuitry is inoperable,. then verify.the indi-vidual rod position (s) of the -rod (s). with inoperable I+trt Circuitry at- .l-least once per 4 hours. ' Fau t & < ~ 4.1.3.1.2 ' Each control rod not fully inserted shall L be7deterniined ;to i be OPERABLE by movement of at least 2% in any one direction at leasti once every 31-days, o DAVIS-BESSE,' UNIT _1 -3/4 1-20.. - L w____,_-_-_-_.----_-_.____ .____.___._______t_-_...

Anudt /vo7 @ l0 3/4.2. POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING _ CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shal'1 be maintained within the limits shown on Figures 3.2-la -lb, -1c, and -id and 3.2-24, -2b, -2c and -2d. APPLICABILITY: MODE 1 above 40% of RATED THERMAL POWER.* { ACTION With AXIAL POWER IMBALANCE cxceeding the limits specified above, either: ) a. Restore the AXIAL POWER IMBALANCE to within its limits within 15 minutes, or l b. Within one hour reduce power until imbalance limits are met or to 40% .i of RATED THERMAL POWER or less. j i SURVEILLANCE RE0VIREMENTS j 4.2.1, The AXIAL POWER IMBALANCE shall be determined to be within limits at least once every 12 hours when above 40% of. RATED TliERMAL POWER except-when the AXIAL POWER IMBALANCE al ans is inoperable, then calculate the l AXIAL POWER IMBALANCE at least once per hour. 1 -l l 6 e

SeeSpecialTest/xception3.10.1.

DAVIS-BESSE, UNIT 1 3/4 2-1 b ndac.t No. )),f2,- O,H, M e 80

douaE N07 T g//- POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS d N 4.2.3.1 F shall be determined to be within its limit by using the incoredethUtorstoobtainapowerdistributionmap: Prior to operation above 75 percent of RATED THERMAL' POWER a. af ter each fuel loading, and

l b.

At least once per.31 Effective Full Power Days, The provisions of Specification 4.0.4 are not applicable. c. N 4.2.3.2 The measured F of 4.2.3.1 above, shall be increased by 5% ' formeasurementuncerta}Nty. i .) f s' ] i DAVIS-BESSE, UNIT 1 3/4 2 8

Ilil8 PAGE PR0l!!DEO

WF FORINFORETION ONLY POWER DISTRIBUTION LIMITS j

QUADRANT POWER TILT. j LIMITING CONDITION FOR OPERATION ) of Table 3.2-2. d 3.2.4 THE QUADRANT POWER TILT shall not exceed the Steady State Limit ( 1 APPLICABILITY: MODE 1 above 15% of ~ RATED THERMAL POWER

I ACTION:

j a. With the QUADRANT POWER TILT determined to exceed the Steady State Limit but less than or equal to the Transient Limit of Table 3.2-2. 1. Within 2 hours: a) Either reduce the QUADRANT POWER TILT to within'its l Steady State limit, or b) Reduce THERMAL POWER so as not to exceed THERMAL POWER, including power level cutof.f, allowable for the reactor coolant pump combination less at least 2% for each 1% of QUADRANT POWER TILT in excess.of j the Steady State Limit and within 4 hours, reduce j the High Flux Trip Setpoint.and the Flux-o Flux-Flow q Trip Setpoint at least 2% for each 1% of QUADRANT POWER TILT in excess ~of the Steady Stat'e Limit. 2. Verify that the QUADRANT POWER TILT is within'.its Steady. State Limit within 24 hours after exceeding the Steady. State Limit or reduce THERMAL POWER'to less than 60% of THERMAL' POWER allowable for the reactor coolant. pump combination within the next 2 hours and reduce'the High Flux Trip Setpoint to < 65.5% of THERMAL POWER allowable for the reactor coolant pump combination within the next 4 hours. 3. Identify and correct the cause of the-out of limit con-dition 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 acceptable at 95% or greater RATED THERMAL POWER.

See Special Test Exception 3.10.1.

DAVIS-BESSE, UNIT 1 .3/42-9

b aA N0/: Pg 13 1 POWER DISTRIBUTION LIMITS -j 1 LIMITING CONDITION FOR OPERATION (Continued) ACTlos: (Cen%ued) 3 b. With the QUADRANT POWER TILT determined to exceed the Transient I limit but less than the Maximum Limit of Table 3.2-2, due to misalignment of either a safety, regulating or axial power-3 shaping rod: I 3 1. Reduce THERMAL POWER at least 2'; for each'1% of -indicated I QUADRANT POWER TILT in _ excess of the Steady State Limit within 30 minutes. 2. Verify that the QUADRANT POWER TILT.is within its Transient Limit within 2 hours after exceeding the Transient Limit or reduce THERMAL POWER to less than 60% of THERMAL POWER allowable for the reactor' coolant pump combination within i the next 2 hours and reduce the High Flux Trip Setpoint to < 65.5% of THERMAL POWER allovable for~ the reactor cooTant pump' combination within the next 4 ho~ rs'. u 3. Identify and correct the cause of the' out of limit con- ) dition 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 l Steady State Limit at least' once per-hour for 12 hours or-until verified acceptable at 95% or greater RATED THERMAL _j POWER. c. With the QUADRANT POWER TILT determined to exceed the Transient Limit but less than the Maximum Limit of Table 3.2-2, due to-causes other than the misalignment of either a' safety, regulat-ing or axial power shaping rod: 1. Reduce THERMAL POWER to less than '60% of THERMAL POWER allowable for the reactor coolant' pump combination within 2 hours and reduce the_High Flux Trip Setpoint to < 65.5% ~ of THERMAL POWER allowable for the reactor coolant pump combination within the.next 4 hours. 2. Identify and correct-the cause of the out of limit con-dition prior to increasing THERMAL POWER; subsequent-POWEP. 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. DAVIS-BESSE, UNIT 1 . 3/ 4,2-10 '

- ~ - - - P3 /Y GuADRMT PONOt TILT 4 Ta bl e 3.2-2 g g p' g* L j;e,i s,3 Steady state Transient-Maximum j limit limit limit i Measurement independent 4.92 11.07 20.0: i QUADRANT POWER TILT QUADRANT POWER TILT a's measured by: 1 Synenetrical incore detector 3.37-8.52. 20.0-i system, 0-50 t10 EFP0 e Symmetrical incere detector 3.02-8.52 20.0-system, af ter 50 10 EFP0 i Power range channels 1.96-6.96 20.0 Minimum incere detector system 1.90-4.40 20.0 i a t .] .i l cAvis-c:ssE, unn 1 3/4 2. Anendment flo. 77, 33, (), E, 00 q

i hbAlQ.l~ /Ydy' Pg/5-TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) and the inoperable channel above may be bypassed for. up-to 30 minutes in any 24 hour period when necessary-to test:the trip breaker associated with the logic i of the channel being tested per Specification 4.3.1.1.1, and c. Either, THERMAL POWER 'is restricted to < 75% of. .SATftT* RATED THERMALTM the -High Flux Trip Setpoirit is reduced to < 85.. of AATED THERMAL' POWER within 4 hours.or the-QUADRANT POWER TILT is-monitored at least' once. per 12 hours. ACTION 3 With the number of OPERABLE channels one less 'than the 1 Total Number of Channels ~ STARTUP and POWER OPERATION may l proceed provided both of the following conditions. are - satisfied: a The inoperable channel is placed in the tripped a. _{ condition within one hour, j 1 b. The Minimum Channels OPERABLE requirement is met; however, one additional channel.may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1.1, and the inoperable channel' above may be bypassed for.up.to 30 minutes-in any 24 hour period when neces sary to test' the trip. breaker associated with the logic of the channel being ~ tested per Specification 4.3.1.1.1. i ACTION 4 With the number of channels OPERABLE one less' than re-quired by the Minimum Channels 0PERABLE requirement and with the THERMAL AewW level: l- . POWER. a. < 5% of RATED THERMAL POWER restore the. inoperable channel to 0PERABLE status prior to increasing j THERMAL POWER above 5*, of RATED THERMAL POWER. b. > 5% of RATED THERMAL POWER,-POWER OPERATION may continue.

l l

DAVIS-BESSE, UNIT 11 3/4 3-4 _a

. ADDID0NAL CH5NGES PfiL/IOUSE3 "' E NL%G

WGF

.1 5 PROPOSED BY LEITER' ' v ["h i g'-/G Date 3-47-8 7 fj SerialNo. /3/2 r l q E_3[3-l'(Continued) l ACTION $TATEMENTS(Continued) ] l Vith the number of channels OPERABLE one less than required y ACTION 5 by the Minimum Channels' OPERABLE requirement and with the THERMAL POWER levelt q 0 ~ a. -< 10 amps on the Intermediate Range (IR) in-strurr?ntation,. restore the inoperable channel to ] OPERABLE,gatus prior to increasing THERMAL POWER j above 10 mos on the IR' instrumentation. -j i 0 i b. > 10 amps on the IR instrumentation, o pration may continue. f ACTION 6 With the number of channels OPERABLE'c.ne less than rem quired by the Minimum Channels OPERABLE requirement, f verify cotfr,iance with the SHUTDOWN MARGIN requirements i } i of Specification 3.1.1.1 within one hour and ~at least once pdr 12 hours thereafter. /, ACTION 7 With the number of OPERASLE channels one less than-'the Total Num$er of Chbnnels STARTUP and/or POWER OPERATION 'may,prbceed provided all of the following conditions are stMsfied:

a. a Within 1 hour:

3 1. P1 ace the inoperable chan'nel in the tripped conditi.o.n, or 2. Remove power supplied to'.the control rod trip 1 . device associated with the inoperative chan'nel, b. One additional channel may be bypassed for 'up to 2 ] hours for surveillance testing per Specification j - 4.3.1.1.1, and the inoperable channel above.may Lbe / bypasseoforJto.30 minutes-.inany24 hour-period [ 1 , when.necessary:to test the trip breaker associated with the logic of the channel being tes: 51 per { Specification 4.3.1.1.1. The inoperable channel above mt.y net be bypassed to test'the logic of a channel j ^

of the trip system associated with_ the inoperable j

ciennel.. With the number of channels 0PERABLE less than required ACTION 8 by-the Minimum Channels OPERABLE requirement, be in at-least J. HOT STANDBY within 6' hours. J DAVIS-BLSSE, UNIT 1 '3/4 3-5 e ___-______

M/)Nd /VC7 ADDil!C"At CliANGES PREVIOUSLY ~ i h PI,0 POSED BY LETTER 7 0eriat No. ~ /3/2 . 0ated-074 7 TABLE 4.3-1 (Continued) ~. NOTATION _ If not performed in previous 7 days. (1) l Heat balance only, above 15% of RATED THERMALLPOWER. (2) 1 When THERMAL POWER [TP] is above 305 'of RATED THERMAL 1 .(3) POWER [RTP),compareout-of-coremeasuredAXIALPOWER-IMBALANCE [ API]toincoremeasuredAXIALPOWERIMBALANCE [ API;]. RecalTbrateif: h [ API, - AP1 ] y,,3.5% ] 3 AX1AL POWER IMBALANCE and loop flow. indications only. .i (4) urify at least.one decade overlap if not verified in previous ] (5) j 7 days. Each train tested every other. month. (0) Neutron detectors may be excluded frs CHANNEL CALIBRATION. J (7) i (8)'- Flow rate measurement sensors may be excluded from CHANNEL- ~ CALIBRATION. However, each flow measurement' sensor shall be '{ calibrated at least once per 18 months.X l i With any control rod drive. trip brea'ker closed. When Shutdown Bypass is actuated. { C y(ir,3..,yy. 3, q[y;" kr.;; t;;; d"* " 'tre,'^S2 deh yed ............,,, i m. I j DAVIS-BESSE, UNIT 1 3/4 3-8 Amendment No. 43 0

0 0 0 0 0 T 9 9 9999 1 1 1 1 1 C A E 4 4 4 4 6, LBS 6, ~3 3333 3 3 3 3 AE CD 4, 4, O 3 I LM 2 2222 3, 2 2 2 2 PP 2, 2, A 1 1 1 1 i 1 1 1 1 1 1 E M L U B MSA N I TR 3 3 3333 2 2 2 2 2 O NI E NP I I T MUO A T N E M UR P T SI S TR N I T 3 2 2222 1 1 1 1 's N e I UO M T E 3 T S 3 Y S 3 N E O L I B T OS A A NT T U I T LN C AU 4 4 4444 2 2 2 2 2 A T OF S TO E R U TA E F e n v Y w o n i T n o n o s i E o e e L w o t i s F i r r - o i cl t i A t u u wwL

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A E L 6, ' 5, B AS 4, 4, 4, 4, 4, 4, CE I D LO 3, 3, 3, 3, 3, 3, 5, PM P 2, 2, 2, 2' 2, 2, 4, A 1 1 1 1 1 1 3 N O E I M L T USB i A MTA T I I R 2 2 3 2 2 1 2 N NNE E I UP M M O U RTS N I P M SI E TR 3 T T 2 2 2 1 1 1 2 I S N 3 Y UO S T 3 N E O L I B T A A T U T C OS A NT I S LN E AU 2 2 4 2 2 1 2 R T U OF ~- T TO AEF YTEFA S ev t p l nm a eu tl V mS ra n r ai n i y S e t t o s ac L d) S n i r t n y E e% e t e ne a N e0 rs a t og r N F9 os l a C r) p A ( t E S o le en S l s a L s l N t mo C up r n E l I O pEi t Bi eo N r I e t n C ' r n N t e T cd a e I r 1 T ed) A a z A xnl m G e -a G e% e i t U ea u n O c i r h9 Ci H r T ( c i L n t e l S5 u C yr a e nk e-( K y s T A S ai t E u e a sd C a s Arc n C q s e e a's O c e I N L F pe o N e s r i ou L e r U A SSR C E S EB DLB R D P A U U E Ni N Q T iG A OL I E N M a b S a b c I a b T T CC N "L U FT 3 4 5 ^ E<g&guf @se ~ u2 v'i y8.3; + ) a i: t

I&Of}

Ppo TABLE 3.3-3 (Continued)-

1 TABLE NOTATION-i Trip function may be bypassed in this MODE'with RCS pressure below 1800 psig. Bypass shall be automatically removed when RCS pressure exceeds:1800 psig. Trip function may be bypassed in this MODE with RCS pressure below 600 psig. Bypass shall be automatically. removed when RCS' l pressure exceeds ~600 psig. _One must be in SFAS Channels #1 or #3, the other~must be in Channels #2 or #4. This instrumentation must be OPERABLE during core alterations j or movement'of irradiated fuel within the containment to meet -) the requirements of Tech. Spec 3.9.4. -] 1 All functional units may be bypassed for up to one minute when starting each Reactor Coolant Pump or Circulating Water Pump. ] 1 The provisions of Specification 3.0.4 are not applicable. L 1 ACTION STATEMENTS l ACTION 9 - With the number of OPERABLE functional units y e;1esg than the Total Numbe'r 'of _ Units, etTS$[ and/or pE:5 8-$$55 -l [ may proceed provided both of the following. conditions'are j satisfied: } The inoperable functional unit is.placed in the a. tripped condition within one. hour. For functional-j unit 4a the sequencer channel shall~be_placed in the H tripped condition by physical removal of the~ sequencer module. b. The Minimum Units OPERABLE _ requirement is met; however, one additional functional unit may be bypassed for up to 2 hours _ for surveillance testing per Specification 4.3.2.1.1. ACTION 10 - With any component in the Output Logic. inoperable, trip j the. associated components within'one hour or be in at least HOT STANDBY within. the'next 6 hours and in; COLD SHUTDOWN within the following-30: hours. ACTION 11 - With the number of 0PERABLE Units one less than the Total Number of Units, restore the inoperable' functional. unit.to. OPERABLE status within 48 hours or be in at'least' HOT: STANDBY within the'next 6 hours andlin COLD SHUTDOWN within'the following 30.houbs. ACTION 12 - a. With less than the Minimum Units OPERABLE.and reactor. coolant pressure > 438 psig,( both. Decay Heat ' Isolation ~ Valves (DH11 and DH12)'shallLbel verified closed.' ~ s DAVIS-BESSE, UNIT 1_ 3/4 3-12' Amendment No.-28,L37; if,- ^ [ = A 2 L

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== aa==== = a a: m a II Ie = l bJ E==, o <E eG j 9 3 WW d. ta C ) i d he' C 1 P* N .A : .I ~m' { .,J 8 i 2. 1 h,' J ww ++memm e e m e e ~ i 1m I : aC' E2 1 N ' *= 67 W ' 1 4 E < J

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9 3 e m 6# >= i u m C aC. .E.. +' i >=' .g ,== g, g 4 ) E .c. b-Si W = C I -! i ed C. L ad me > h.c. W4 .e e 2 C C. w- .eCi E.C b8 -- e'. j 4= 3 m. .i== e- = C'.C". bO E e a== e-m

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== == - v -) m3 1 E< .EC V Q 4 m S-4 e eJ b 3 O. W AC w 3 C P* 3 C C t y C-C..# C5 C .u g >= m ed C C O C .J o bJ = b od - b# bp C E { C s 1 = b b ea e O' = D, I e.a 3333g ee ed e. s e== a e r l P*e .9 # e O..O e as m

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C 4 p OMb i IF . 6 .a -C u s.o.. m e w .o -3 .v a i : .u um W - e, 2 E. e. e L., e.> . u. c> e, u, = i.. - D e C C C.3 3==, d.O..J O. O.J Ce .J V O a.C e u o n .e .a: et E a' > ., e,.> ~ v an v w .s< aa au.>- au w u. E.. u C,. C u..b.C. v. c..w e .c w n - I., as. 8.,: b >= >= 4 e 8

9. 0.'O-1C'O.*=C - C..O.v - u - 5. s t b..

p m. - = w.e.s e,6. 4. . 3 e- - g .J tJ x. CCCmem a C v-vmum .u e W D' 4 O O O v U 3.a C.C C C C <mC w a. vvv:. m.C 0 W, 6 R e. a v 5 W m s e a

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e 2 5-W = ~oavm. w e a m J _. N DAVIS-BESSE, UNIT 1~ 3/4,1 2) Amendment No. X./s 43 1

WN 4 A NLD I LE 3 I R 5, SEI EVU 2 L DRQ 4, A OUE N l O MSR 1 3 l I ATC .N U L nF A w N oL S LO dE T EI t N N NT uN E NCT hA M ANS sH i E HUE R CFT ~ C g l na I U i Q re E uv R N-O-

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E I sc C T h e ) N LA t r A ER n t d L NB R R ol t L NI ml e AL a i I u E HA 8h V CC MU 1 s R rn S eo t mN pi t r O ea h ( cu t I T nt A L N oc 2 T E O a N NK I t i 3 E NC 5 S T ss M A [- T er l A ad i 4 U ll i R CC-O l a E T N u L S t g B N E ae A I L f T B da e M A es d E T t u T sl l S ea c Y t u n S n i ea N b m l O l I e lh a T v l t s h A l aiy s U a h wa T V s d T C d S 1 A n se1 E o r et3 T e d. h a S i E t t ci r L 2 4 R a a t ce A U l T o le i op N A S s l 5 ws O sse I E L I 'r 4 ac T F E n-n C e n N t z oyo N Y N a o i r U i T A e i ttt F r E t l t ais i l F C u a uua L A T y s c t ce E S I K a s crl N i N C c e f ai N U O e r i ct A L D P c l a H L R e ar C A E p ueT N T O N S nhS e ate h I I a b e mot T C-e L T N S ) ) U 1 2 F 5 ( ( y2 mE S5Y%a". 5m. m

Yy

.g f

hak :Mo 7 + PgDR TABLE ~3.3-11 (Continued) TABLE NOTATION

May be bypassed when steam pressure is below 650 psig.

Bypass shall-be automatically removed when the steam pressure exc.eeds 650~ psig, f The provisions of Specification 3.0.4 are not applicable. ACTION STATEMENTS ACTION 13 - With the number of OPERABLgCgnnels one gF ggg l. t Total Number of Channels,.... _$ and/or',... y.....- may proceed until performance of the next required CHANf1EL FUllCTIONAL TEST provided the inoperable sec- ) tion of the channel is placed in the tripped condition within 1 hour. ACTION 14 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. l 1 I ) DAVIS-BESSE, UNIT 1 3/4~3-27 Amendment.No. 37-i

LO EI NT NCT H H M H R ANS HUE ~ CFT o w e 2 N e O I e T ~ LA m e M5 ER R R R R t ET NB r TN NI t SE AL YM HA = SE CC R LI i OU RQ t TE x. NR O CE ~_ C 1 EN 1 RA UL 3 1L PI L l 4 E E AV NK i. l E - R NC A L L RU MlE S 5 N B ES 'vT I l A T CC AN WO DI ET .m EA FT F N f DE o NM w r AU o eh s R w L t g s L MI o ai o r AS L wl L i EH d - T l ee s e Sl t e er p e v Fu m S r e s u u L - s P s e e s r rr t e o oP n l r t t a e e P a al l n r ra o n M n e e el o o x, _. a n n nL C i

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t m m me t t I n a a af c c N e e e ef a A m s. e U m t t ti e u S S SD R l L r a A t u N s n O n a I I a b c d M T C

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2M B0 7; a a f INSTRU".ENTATION l ANTICIPATORY REACTOR TRIP SYSTEM INSTRUMENTATION. LTMITING CONDITION FOR OPERATION 3.3.2.3 ' The Anticipatory Reactor Trip System instrumentation channels of Table 3.3-g shall be OPERA 5LE.- 17 APPLICABILITY: As shown in Table 3.3-yf 17. ~ ACTION: As shown in Table 3.3-)617 ] 4 1 l J h ) ~ l . I 1 SURVETIIANCE REOUTREMENTS

==

4.3.2.3 The Anticipatory Reactor Trip System shall be.' demonstrated OPERABLE by the performance of'the CHANNEL CHECK,. CHANNEL CALIBRATION.. and CHANNEL FUNCTIONAL TEST for' the modes and'at the frequencies shown in Table 4.3-yf. l-17 o DAVIS-BESSE' UNIT l' 3/4,3-30a Amendment :Ho.73.', e l .I, 1 L.- ----__---__----. _ _ _ - _

-C S ) I EE LLD ( PBO 1 1 1 PAM A N O ITA SE T MLL N UEB E MNA M I NR 3 3 3 U NAE R I iP t T MCO S N R I EW M O E P g. S TS L Y A 3 M S R 3 P L ) E I E P

l E

R NOI ( TR L T NTR 2 2 2 DWE B A T A R l EO i T O C TP T A C RL A A E fM R oRE Y tH R nT O e T cD A rE P eT I pA C P I O S 5 T N L 2f N E o A LFN 4 4 4 ot w AON T A l n O ee t i T C b cr d e ep s s5 a2 pye b vo yb l a p l m ay p h u c cl i t P i i n r o g t o T Bd o a es L me e f ee ol n oFn t tb L i i u ua A b pnb p ac N r iir t OT u rau u pl i II T TMT O ip TN CU rp N TA U F 1 2 3 m1 Yh Eaian

$ 4M M '/40.7 fjj27 J TABLE 3.3-)5' (CONTINUED)- [ 1 -ACTION STATEMENTS' . ACTION 16 - With the. number of chaa' nels OPERABLE one'less than required by the Minimum Channels OPERABLE requirements, restore the inoperable channel. to OPERABI.E ' status :within ~ 72 hours or reduce reactor power to less than.25. percent of RATED THERMAL POWER within the next-6 hours. ACTION 17 - With the number of channels > OPERABLE one lessi t.han. required by the Minimum Channels.' OPERABLE requirements, restore the inoperable channel to OPERABLE status within' 72 hours or be in at least HOT STANDBY within the'next 6~ hours. ACTION 18 - With the number of OPERABLE channels one'less than the Total Number of Channels, STARTUP anp0VER OPERATION l may proceed provided both of the follcwing conditions 'y are satisfied: a) The control rod drive trip' breaker associated with the-inoperable channel is placed in the tripped' condition.within one hour. 1 b) The Minim'um Channels.0PERABLE requirement is met;. however,-one additional control rod drive-trip breaker associated with 'another channel may be trippe'd for up to 2 hours for surveillance; testing-. per Specification 4.3.2.3,.after reclosing the. control rod-drive Ltrip breaker opened _in a) above. l 3/4 3-30c Amendment'No. 73. DAVIS-BESSE: UNIT 1

( eN S l m* iI CI ED I CE I VNR N AI NLU ILQ I 1 I I E SER EV DR OU MS ~ STN E M ER IUQ E L R LA EN E NOT C NIS N ATE H H H I A CT I L CN L U I F EV R US R N E O W I O T P A 76 T L )- N e e e A 1 E l l l M 3 N N b b b R U O a a a E 4 R LI c c c H T ET i i i TR E S NA l l l E L N NR p p p DW B I AB p p p EO A I A A A TP l l T M CL A E A t t t RL T C o o o A S N N N fH Y oR S E t P I nT I e R cD T rE eY R e pA O l R T A L. b 5 C a 2f c o E EK i w R NC l ot NE S S p l n Y AH p ee R iC A bc l O C r T t d e A o ep P N s IC s5 a2 IT pye N bv A o yb l a ( p-l m ay p u 'p c cl i Pi i in T r .r g t o I T dT o a N e L me U e ee ol n 'F n t tb L i i u ua A b nb p ac N r ir t i O u au u pl I T MT O ip T rp C TA NU F 1 2. 3 ma ~ 9$ 91 F>) I $> y

02 NO 8 9 9 9 I 1 1 1 1 T CA A . L N Y O T l I I T I )' ) ) ) T B N A 1 1 1 1 E C ( ( ( ( M I U L R P T P S A N I R GN I O .SE 5 T MLL UEB 1 I H HNA 1 1 1 1 3 0 NR I H HAE i l 3 i P I HCO E H L f B u n A l i I f a f x r f n o D D ea) e B y ir l ney n t t~ l n ion c i il l i Q c LC s L o vNo we I i u s i i i L st t ro e t t ttt S ra noe c n c cua e n i e e aBn m E om V t o u 's a v u l o i r ut e l l iam o I it l T nu f ol D f o drr t C oA e f eu f C aae 5 D d a ai n o A A M s E nm t' E Rl T f yne el s e e t a c m 'gi I i m t D t al tl A i e s et e s w mnt e R vsR act r a o aia l Gd m m ws o u w l l i t rf dib s d F i o u cao aH a a rvt B al R t e H n oou - e oAn ro H o rA aP e n d o d en i t t gs i i h e d gi i i a nt ut t t u u es na b Ri a qi u a q l B ri e r eb R i i odl u T d n I N sii 1 ( L D sti e I E svm w si vR M oor o ono U rre rorf R GPT a. l F a b GMP o a T S N I 1 2 3 - wAo> y$&g+c._ ~ C$' 5i. - -e l 2

THIS' PAGEPROVIDED eWR"" l FORINFORIMTON ONLY TABLE 3.3-15 (Continued) y TABLE NOTATION j 4. v (1) During radioactive releases via this pathway : ACTION 18 Vith the number of channels OPERABLE less than required by', the Minimum Channels OPERABLE requirement,=eff1uent releases _ may be resumed,'provided that. prior to initiating a release:- 1. At least two independent samples are analy:od in ~ 1 accordance with Specification -4.11 1.1.1 - f or - analyses performed with each batch;. 2. At least two independent verifications'of the release ~ I rate calculations are performed; 3. At least two independent verifications of-the discharge .] valving are performed; i l Othemise, suspend release of radioactive affluents 'via this pathway. ACTION 19 with the number.of channels OPERABLE-less than required by the. Minimum Channals OPERA 8LE requirement, effluent. releases via p'i this pathway may continue provided the flow; rcta is' estimated j at least once per A hours during actual releases. Pump curves may be used to estimate flow. ACTICH 20 With the number of channel's CPERABLE -less than required'by the Minimum Channels OPERABLE requirement, 'ef fluent releases _via. this pathway may continue provided.that,.at least once per- ~ 12 hours, grab samples are. collected.and analyzed for gress radioactivity (beta or gams) at a lower limit of-detection no great' - .i to-7 ci/sl. ] er than, p i DAVI!-BESSE, UNIT'l 3/4'3 ' Amendment: No.. 86'. ~ R P .M_--________________

L LA EN NOT ) S NI S 2 T LATE ( N CT O q Q i E CN M U E F R I UQ E N R O LI E ET ) C NA 3 N NR I A LAB R R R L CL L i I A I E C VRU S EK N CC O RE t A. A. I Ui P T OC H N A S T N E 5 M L 1 U EK I ) ) R NC I 4 4 4 N .HC 3 T NE I I I S Al D D D i C I E L G B N A I T R O T x I e o N y n e B O t i n H n i L i v L o T i s i N td t t t E cn n n c U aa e e e L o u u l F im l l l F dr f f o E aa f f C Rl E E D A o I a n e e t U s go t s t n Q a ni s r s w I ait a o a o L Gda w t w d F l d il i E rvo a n a V oos R o R n I rI M o T aP d d i i t C t c e i A esi u t u u D Brt q a q l I I oa i R i l D N st m L L O A E sio w .R M ont o U rou l R GMA a F a b. T S N I 1 2 o$t M- ,s* Y8 R g.- O.

Ilil8 PAGE PRMIDEE M ar 10RINFORKION ONlY TABLE 4.3-15.(Continued) L TABLE NOTATION (1)- Ouring releases via this pathway. (2) The CHANNEL FUNCTIONAL TEST shall also--demonstrate that automatic isolation of this pathway and control room' alarm annunciation occurs if the instrument indicates' measured levels above the alarm / trip setpoint. (3) The initial C!iANNEL CALIBRATION for radioactivity measurement in>trucentation shall to p eformed using-one or more of the reference standards certified by the National Bureau of Standards or using'stancards that have been-obtained from suppliers -that participate in measurement assurance ' activities with NBS. These; standards should permit calibrating the systedt over its intended range of energy and rate capabilities. ^ For subsequent CHANNEL CALIBRATION, sources-that have been related to the 'nitial calibration should be used, at intervals of at least ance per. eighteen months. For high range monitoring instrumentation, where calib'ation with a radioactive source is i r impractical, an electronic calibration may'be substituted for the radiation source' calibration. (4) CHANNEL CHECK shall consist of. verifying indication of flow during periods of release. CHANNEL CHECX shall be mace at least once daily on any day on which continuous, periocic, or batch releases are made. l 'l i i 0 AVIS-BESSE, UNIT 1 ' 3/4 3-61 Amendment flo. BS a __--__--L__:-

q l S_ T __ N EM E R. I U Q ER Q E S C N N A O L LI L ET NA A. / I E NR V AB N R R R R R R R HI l U CL l S AC N O 6 I T 3 A T 4 N L 5 E EK E M NC L U NE M M H M M H H B R Ai l A T l C l T S C N. I_ G_ e N_ s r I_ g u l s R_ e s e e O_ L s v h h T e e c M n t e r L t I N o lo r P i O i u e w M t f s m g S a e s a n N c r e e a t W i u r t R i O d t P S m D n a 'p i T I r m t u L 0l r p t l w e e e t i r n S e m s t a o W k e y u t i E a T S l O S t T e e i M O r t t v r r s E M B n n e o o o 9 E a a L t t P R p l l a a E o o r r r d ir o o e e e o T C C z n n R W e e i r r r r G G l M o o o u o T t t t s m m r N c c c s a a t E a a a e e e n M e e e r t t o 9 U R R R P S S C 9 R T S g N I 1 2 3 4 S 6 7 l Se ~ 'f y C54 ~ u + Y#n g &2n t l!

)lll NM ~ @c+ n ^ ^ I 7 N Z O I t I lI l i / l A A A A A A A ? i l l t l H H N t t t l ' s s Ai s l 1 l l t t t 1 l t l t l I n t l I l. u I l Ll S i t A i C l t t U. i l i I, t l l ^ i R C E C l lA L. l l E V L i EK 5 t b HC l i l [ H H H H H H H H H M H H H H S t l Ai s i l C I t l 0 u C 1 l l 3 A I I 4 l [ 1 [ H l m t l R n l l o a I A 1 S i 2 l t h i l a S i I G d l a s l u t I i t I 0 t a s n t u s l i e S t N d a II n t w O i H n c o S o e s e c i l u i r u g A t t F d e u t o a n l l e n E r t a r t l i D u a l t d s o m o I s r e S y o s i g 0 i C s e v P I l t l 1 C e p e r u c A r m L e l l l q e ~ P e l t e e e E j I l e e e a s s s n s r v g w s s s s I S i t u, e n d e e e e 2 O n P e e' s L a e V V V r e I t l s R e u s r I S t e r F t t t s t u u I u r e p n n n u a s t s t O P z u y e e e t e u a s e i t r m m $ a f t t e l p p r r a n n t a S r S y t P t o o u a l i i l t u o o s t l a a a t S S O L L s S i t t t S e C i [ e x n n n A f S R 9 li 1 t l G C C r G u o o o E a P I 1 i 1 l S H t P S A C C C S S R S I 1 tR I S O 1 2 3 4 5 6 7 8 9 0 I 2 3 4 l l I 1 I 1 1 1 ,?! y> f

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fy 36 I I INSTRUMENTATION CHLORINE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.3.3.7 The chlorine detection system, with the alarm / trip setpoint adjusted to actuate at a chlorine concentration of 1 5 ppm, shall be OPERABLE with at least two OPERABLE chlorine detectors located in the Reactor Control Room ventilation air intake. APPLICABILITY: 1, 2, 3 and 4 00

ACTION:

a. With one chlorine detector or the chlorine detection system inoperable, within i hour initiate and maintain operation of the control room ventilation system in the recirculation mode of operation; restore,the inoperable detection system or detector to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours and in LOLD SHUTDOWN-within the following 30 hours. i b. The. provisions of Specification 3.0.4 are not. applicable. SURVEILLANCE REQUIREMENTS 4 4.3.3.7 The chlorine detection system shall be demonstrated OPERABLE by i performance of a CHANNEL CHECK at least once per 12 hours, a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months. DAVIS-BESSE, UNIT 1 3/4'3-51 l _- N

{ {}lQ. [NOh INN $$N h bb S FIRE CT N I STRtMENTS i MINIMUM INSTRLIMENTS OPERABLE INSTRUMENT LOCATION HEAT - FLAME SHOKE-1 ) 1. Cont ainment_ I fa. FDZ-RCP 1 Elev. 603' 0-0 1* j lb. FDZ-RCP 2 Elev. 603' 0 0 1* )

c.

FDZ-RCP 3 Elev. 603' O O 1* id. FDZ-RCP 4 Elev. 603' O 0-1*

e.

FDZ-PZR Elev. 603' O O 1 *. 1 ff. FDZ 214 - Core Flooding Tank 1-1 Area Elev. 565' O O 3* i l ig. FDZ 215 - Ctzt. Letdovp Cooler Area ' Elev. 565' 0 0 2* a lh. FDZ 220 - Incore Instrument Trench Area Elev. 565' O O 4*

1. FDZ 317 - Ratch Area - Elev. 565' 0

0 20 * { fj. FDZ 410 - East Passage - Elev. 603'/657' O 0 9* j I 2. Containment Annulus _ J 3

a.

FDZ-A208 Elev. 590' O O 10 !b. FDZ-236H Elev. 774' O O 3 l

c. FDZ-236L Elev. 590' 0

0 9 3. Auxiliary Building FDZ 402 - #1 Electrical Penetration Rm. a. Slav. 603' 0 0 12 b. FDE 405 - Auxiliary Building Storage Rs. Elev. 603' 0 0 1 FDZ 427 - #2 Electrical Penetration Rm. ~ $' c. Elev. 603' 0 0 7 ~ d. FDZ 303 - #3 Mechanical Pentration Rm. Elev. 585' 0 0 12 FDZ 304 - Corridor to Mech. Pent Ras 364 e. Elev. 585' O 0 4 f. FDZ 310 - Passage to BA Mix Tank 8 Elev. 585' 0 0 FDZ 312 - Spent Fuel Fool Pump Rm. g. Elev. 585' O 0 4 ~ h. FDZ 314 - #4 Mech. Pent. Room Elev.~585' 0 ,0 17 1. FDZ 300 - Fuel Handling Area Elev. 585' 0 0 5- ,A

Fire Detectors in high radiation areas which are NOT accessible.

a _A ' A E ^ ~ Mm e NA

QQ "/407 1 TABLE 5.3-N (Can+inued) pp31' Sheet 2.of 4-INSTRLHENT LOCATION 'MINIMLS INSTRLM NTS opEgA3tt-HEAT FLEE ' SMOKE i 3. Auxiliary Building.(Continued) q j.- FDZ 209 - Corridor. to #1 Mech.. Pent. Rm. -{ Elev.iS65'- OL 'O-3. k. FDZ 227 - Boric Acid Evap Passagevoy. 1 Elev. 565' 'O O 6 1 '1. FDZ 208 - #1 Hechanical Penetration Rs.- 1 Elev. 565' 0 -O. '10 1 m. FDZ 231 - Cl e a n A*a s t e Boo s t e r P ump Rs. 'l Elev. 565' 0 0 -1

n. - FDZ 232 - Primary & Deborating Demin Viv '

Rm.' -1Elev. 565 O O =1 o. FDZ 234 - Boric Acid Evaporator Rm.1-2 Elev. 565' 'O O-

1 p.

FDZ 235 Boric Acid Evaporator Rm 1-1 Elev. 565'

0 0

l' q. FDZ 236 - #2 Mechanical-Penetration Rm. l Elev. 565' O 0. 4-r. FDZ 240 - Boric. Acid Addition Tank' Rm. Elev. 565' 0 0 ' 5.. s. FDZ 241 - Passage to B. A. Addition Tk Rm. Elev. 565' O. O 2 c. FDZ 101 - Equipment and Pipe Tunnel-Elev. 545' 0 0 -1' u. FDZ 105 - ECCS Pu=p Room 1-1 Elev. 545 0 'Oi 4' ] v. FDZ 110 - Corridor to Central Area of l Aux Bldg. Elev. 545 ' - 0- 'O. 5 j v. FDZ 113 - De. cay Heat Exchanger Pit .l Elev. 545' 0-0 1 1 x. FDZ 115 - ECCS Pump Room 1-2 Elev. 545' -0 0 2 y. FDZ 124 - Clean ' Waste Receiver Tank' Rm. 1 Elev. 545' 0 O. 4 4 Auxilisrv Buildine Tan Rooms 1 a. FDZ 500 - Radwaste & Fuel Handling Area 1 and Air Supply Area '- Elev.' ' 623 ' b. FDZ-501 - Radwaste Exhaust ' Equipment 'and -. 0 0 '20 ~ Main Station Exhaust Fan. Room Elev. 623 0 0 22-c. FDZ 515 - Purge and Exhaust Equipment Rs. 1 Elev. 623 .O. 0-22' d. FDZ 516 .'Non-rad Air :and Exhaust Equip. Rm.' J - Elev. 623' !0 ,0 5-DAVIS-BESSE, UllIT 1 3/413-54 Amendment No. #,'34 O _s

> duad n'o 7 hN l Sheet 3 of 4 TABLE 3,5-14 @nfinu.ed.) i MINIMLS INSTRLY.:.NTS OPE?.ABLE 'TRUMENT LOCATION _ HEAT F1.A E. ' SMOKE ' 'f 1 5. Control Room Complex j FDZ SOS - Main Control Board Cabinets '3 a. 0 0 Elev. 623' b. FDZ 505 - Control Cabinet Room Elev. 623 0 0 5 .i FDZ 505 - Computer Room - Elev. 623' O 0 ,1 - c. 6. Cable Spreading Room O O '5 FDZ 422A - Elev. 613' i a. 7. A/C Equipment Room 0 0 u FDZ 603 - Elev. 643' a. l 1 8. Diesel Generator Rooms _

- a.

FDZ 318 - Diesel Generator Rm. 1-1 Elev. 585' O O 5 {

- b.

FDZ 319 - Diesel Generator Rs. 1-2 i Elev. 585' O O 4 FDZ 321A - Diesel Generator Day Tank c. Room 1 El&v. 5 0-0 1 ( d. FDZ 320A - Diesel Generator Day Tank ) 0 0 1 Room,1 Elev. 5 4.j 9. Battery Rooms FDZ 428A - Battery Room 3 - Elev. 603' 0 0 2 FDZ 429B - Batteiy Room A - Elev. 603' 0 0 2 e a. b. 10. Cocponent Cooling Water Pumo Rocm 0 0 '9 s,,,, FDZ 328 Elev. 585' a. a 11. Feed Pump Rooms _ l FDZ 237 - Auxiliary Teed Pu=p 1-1. Elev. 565' O 0 3 .a. b. FDZ 238 Auxiliary Feed Pu=p 1-2 Elev.'565' 0 0 '3 ' AVIS-BESSE, UNIT 1 3/4 3-55 Amendment No. #, 34

>derta 4 N07 .i Sect 4 of 4 "T' A BLE,, 35-P4 (Con 4loued) N9 N M N!K'* l':57 '.?"STS C?!U3t! IN57:LMNT LOCAT!CN KT AT TLGE SMCXI 1 12. Switchtear Rooms aq TDZ 324 - CD High Voltage Switchgear Elev. 585' 0 0 3 ) a. b. TDI 325 - A High Voltage Switchgear 0 0 3 Elev. 585' i FDZ 323 - 3 High Voltage Switchgear 0 0 11 j c. Elev. 585' d. FDZ 4 28 - F High Voltage Svir.chgear Elev. 603' 0 'O 12 l . FDZ 429 - E High Voltage Switchgear e. 0 0 6 Elev. 603* 3 13. Int ale S tructure I FDZ 052 - Diesel Fire ? ump Koom s Elev. 576' 0 0 1 ] a. -{ b. FDZ 052-Service Vater Fusp Room 0 3 Elev. 376' O 1 TDZ 053 - Service Water Viv. Room Elev. 565' O O 6' c. } 1 1

The fire detection instruments located within th' Containment are not required-e to be OPE?.A3LE during the per formance of. Type A Contain.2ent Leakage Rate Tes ts.
- These detecto[s automatically actuate fire suppress {on systems.

. 4 DAVIS-BESSE, UNIT 1 3/4 3-56 ' x Q endment.No. / 34 3: 'w. ,TA

hal NO7: 1 97 40 'l 3/4' 4.' REACTOR COOLANT SYSTEM

.,3/4'.4.1.

COOLANT LOOPS AND COOLANT CIRCULATION t ! STARTUP AN'O' POWER OPERATION LIMITING CON 0! TION FOR OPERATION 3.4.1;1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation. APPLICA8ILITY: MODES 1 and 2*. ACTION: With one reactor coolant pump not in operation STARTUP.a_nd POWER OPERA-a. TION may be initiated and may proceed provided THERMAL POWER is-re-f stricted to less than 79.7* 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 wi.th' three reactor cool-: ant pumps operating-1. High Flux 2. Flux-aFlux-Fl ow u SURVEILLANCE RE0VIREMENTS 9 4.4.1.1/j,Th'e above required reactor coolant loops shall be verified to be-l in operation and circulating reactor coolant at least once per 12 hours. j 4.4.1/.The reactor protective instrumentation' channels'specified in the l 14 a applicable ACTION statement above shall be verified to.have had their' trip setpoints changed to the values specified in Specification 2.2.1' for the ao-plicable number of reactor coolant pumps operating either: Within 4 hours after switching to a different pump cochinat'an if the a. switch is made while operating,'or Prior to reactor criticality if the switch is made while shutdown. b. 1

See Special Test Excepti.on 3.10.3.

hnendment 1!o 4. 33. 33,'#3, Y OAVIS-8 ESSE, UNIT 1 3/44-1 80-G e

^ 7 ' 3/4.4 FI. ACTOR C00txtT ST57EM [OR IN. {0W]'0N O pb c o SHITTDOVN AND HOT STMDBT j l LIMITINC CONDITION FOR OPERATION - } 3.4.1.2 a. At least two 'of the coolant loops _ listed below shall be'- d OPERABLZ: } q 1. Reactor ' Coolant'Imop 1 and its associated steam' l generator, '1 -1 2.. Reactor Coolant Imop 2 and its associated etsam 1 generator, 3. Decay Heat Removal 14op 1,* i 1 3 4. , Decay HeatJRe=cval I4op'2.* R b. At-least one.of the above coolant loops ihall be faL operation.** Not rare than one decay heat remo' val pump usay be operated. c. vich the sole suction path through DE-11' and 08-12 unless the control power has been removed from the. DE-11 and DE-12 valve operator, or manual valves DE-21 and DE-23 are - opened. [ d. The previsions of specifications 3.0.3' and '3.0.4 are not - i applicable. U l APPLICABILITY: MODES 3, 4 and 5 ACTION: s. With less than the above required coolant loops CPIRA312, I immediately initiate corrective 'setion to return the required coolant loops to OPDABLE status asisoon as possible, or be 'in COLD SHUTDOWN within '20 hours. l b. With none of the above required coolant loops in ' operation, suspend all operations involving a reduction in boron concentrationoftheReactorCoolantSystemandimmediately initiata corrective action to' return the' required coolant loop to operation. i h

The normal or. emergency power ~ source may be inoperable. in MODE 5.

- This, loop may not be selected in: MODE 3 unless the, primary side. temperature and . pressure are-. ithin the decay heat removal system's design conditions [ w

- The deca provided (y heat removal pumps may be de'-energized for:up'to 1 hours 1) no the' reactor coolant system boron concentration, and-(2) core outlet' temperature is-maintained at least 10*F below saturation temperature.

9 DAVIS-BESSE UNIT.1 '3/4'4-2J AmendmentNo.N,'A',188, 88, 92

,.>duual 1907 } 29 4W [ 3/4.4 REACTOR COOLANT SYSTEM fM V SURVEILLANCE REQUIREMENTS 4.4.1.2.1 The required decay heat removal loop (s) shall be.. determined OPERABLE per. Specification 4.0.5. 4.4.1.2.2 The required steam generator (s) shall be determined OPERABLE - by verifying secondary side level. to be' greater _ than or equal to (a) 18 inches. above the lower tubef sheet once per = 12 hours if an associated. reactor coolant pump is operating; or, (b) 35 inches above the lower tube sheet once per 12 hours if no reactor ~ coolant pumps.are operating. 4.4.1.2.3 At least one coolant loop shall'be verified to' be in operation and circulating reactor coolant at. least once per 12 hourr. OV k ^ 1 \\ i s >l I s I i -w i . DAVIS-BESSE UNIT 1 3/4 4-2a Amendsent ne, 30 1 _=__:____.___o

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a............__ ..s. a... i.. _.......i......... 3..s.... ........s.. . ;.... s. w ...n. Reactor N 1 ant System Pressure - Pressurizer Level Limits for inoperabia Decay Heat k ova? System Relief Valve in MODE 5 Figure -b - 3,4 ,33 l Davis-Besse Unit 1 3/4 4-4b Amendment No. 57 9 0 m_m_______m.___

>dnua.L eM7 FS 45 l i (.. REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 1 l 3.4.4 The pressurizer shall be OPERABLE with: J a. A steam bubble', b. A water level between 45 and 305 inches. l I APPLICABILITY: MODES 1 and 2. ACTION: With the pressurizer inoperable, be in at least H0T STANDBY with the control rod drive trip breakers open within 6 hours.- 1 o l f: 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 I-DAVIS-BESSE,. UNIT 1l '3/44-5 --- w

.>&u al /yo 7 h REACTOR COOLANT SYSTEM OPERATIONAL LEAXAGE LIMITING CONDIT10fl F0,R OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to: No PRESSURE BOUNDARY LEAKAGE, a. b. 1 GPM UNIDENTIFIED LEAKAGE, j 1 GPk total primary-to-secondary leakage through steam generators, c. d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, q e. 10 GPM CONTROLLED LEAKAGE, and ) f. 5 GPM leakage from any Reactor Coolant System Pressure Isolation ] Valve as specified in Table 3.4-2. ) APPLICABILITY: MODES 1, 2, 3 and 4 o I ACTION: With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STAND 8Y ( a. within 6 hours and in COLD SHUTDOWN within th'e following 30 hours. l b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAXAGE, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours except as pemitted by paragraph c below, In the event that integrity of any pressgigo atg9.maygalve specified in Table 3.4-2 cannot be demonstrated, p_ l' c. continue, provided that at least two valves in each high pressure line having a non-functional valve are in and remain in, the mode corresponding to the isolated condition.(a) f The provisions of Sections 3.0.4 and 4.0.4 are not, applicable for entn d. into MODES 3 and 4 for the purpose of testing the isolation valves in Table 3.4-2. { (a)Motor operated valves shall be placed in the closed position and power j supolies deenergized. 3/4 4-15 Order dtd. 4/20/81 DAVIS-8 ESSE, UNIT 1

duval N07 F%'17 j REACTOR COOLANT SYSTEM i l SURVEILLANCE REQUIREMENTS j i Reactor Coolant System leakages shall be demonstrated to be within

4. 4. 6. 2.1 each of the above limits by:

Monitoring the containment atmosphere particulate radioactivity a. monitor at least once per 12 hours. b. Monitoring the containment sump inventory and discharge at least a once per 12 hours. frorn i Measurement of the CONTROLLED LEAXAGE-to the reactor coolant pump seals j c. to the makeup system when the Reactor Coolant System pressure is 2185 f,20 psig at least once per 31 days, d. Perfonnance of a Reactor Coolant System water inventory balance at least once per 72 hours during steady state operation. 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Yalve specified in Table 3.4-2 shall be individually demonstrated OPERABLE by verifying' leakage testing (or the equivalent) to be within its limit prior to entering ' MODE 2: After each refueling outage, a. b. Whenever the plant nas been in COLD SHUTDOWN for 72 hours, or more, and if leakage testing has not been performed in the previous 9 months, and Prior to returning the valve to service following maintenance, repair c. or replacement work on the valve. 4.4.6.2.3 Whenever integrity of a pressure isolation valve listed in Table 3.4 2 cannot be demonstrated, the integrity of theLremaining pressure isolation valve or the integrity of the remaining pressure isolation valve in series with the motor-operated containment isolation valve in each high pressure line having a leaking valve shall be determined and recorded daily. In. addition, the position of the closed motor-operated containment isolation valve i located in the high pressure piping shall be recorded daily. ( i ) o DAVIS-BESSE, UNIT 1 3/4 4-16 pg, Andndment Mo. 54 4 -J m.

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j i

. i 1 e i ) l i I 6 } s 1 i i 6 i 20 30 40 .50 60 - 70. 80 90 ' '100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER.with the Primary Coolant Specific ' Activity >1.0gCi/ gram OmCwMk,a 1131 1 I DoSC EAulVAlmi DAVIS-BESSE, UNIT 1 3/4 4-23 ^ __._1_.____.__2____

sdAJual NC7 i REACTOR COOLANT SYSTEM g 4./ q REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.11 The following reactor coolant system vent paths shall be oper-able: a. Reactor Coolant System Loop I with vent path through valves RC 4608A and RC 4608B. b. Reactor Coolant System Loop 2 with' vent path through valves RC 4610A and RC 46108. c. Pressurizer; with vent path through EITHER valves RC11 and RGa- (PORV) OR valves RC 239A and RC 200. [ Rc2A APPLICABILITY: Modes 1, 2 and 3 ACTION: a. With one of the above vent paths inoperable, restore 'the inoperable vent path to.0PERABLE status within 30 days, or, be in HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 30 hours. b. With two of the above vent paths inoperable, restore at least one of the inoperable vent paths to OPERABLE status within 72 hours or be in HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 30 hours, c. With three of the above vent paths inoperable, restore at least two of the inoperable vent paths to OPERABLE status within 72 hours or be in HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 30 hours. d. The provisions of specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.4.11 Each reactor coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by: 1. Verifying all manual isolation _ valves in each vent path are locked in the open position, and 2. Cycling each valve in-the vent path through at least one complete cycle of full travel from the control room ~during COLD SHUTDOWN or REFUELING,-and 3. Verifying flow through the. reactor coolant vent syst'em vent'- paths during COLD SHUTDOWN or REFUELING.. DAVIS-BESSE, UNIT 1 3/4 4-32 Amendment No. 85-s .i J

FOR WF0 Mil 0N DiU 1 EMERGENCY CORE COOLING SYSTENS -l ECCS SUBSYSTEMS - T.... > 280*F 1 LIMITING CONDITION FOR OPERATION l 3.5.2.Two independent ECCS subsystems shall be OPERABLE with each-subsystem comprised of: a. One OPERABLE high pressure injection (HPI) pump, j b. One OPERABLE low pressure injection (LPI) pump, c. One OPERABLE decay heat cooler, and d. An OPERABLE flow path capable of.taking suction from the borated water storage tank (BWST) on a safety injection signal-and manually transferring suction to the containment sump'during - the recirculation phase of operation. ') l Q APPLICABILITY: MODES 1, 2 and 3. J ACTION: i 1 a. With one ECCS subsystem inoperable, restore the inoperable j subsystem to OPERABLE status within 72 hours or be in HOT 4 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 Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: 1 At least once per 31 days by_ verifying that each. valve (manual,. a.. power operated or automatic) in the flow path that is not: locked, sealed or othentise secured in position. -is in its correct position. 4 0 AVIS-BESSE, UNIT 1 3/4 5-3 Amendment No. 36 ' e 9

'! SURVEILLANCE: REQOlREMENTS [ Continued) / j b. At:1 east once per 18 months, or prior. to operation af ter ECCS-piping has been drained by verifying that the ECCS piping is full of water by venting the ECCS-pump casings and discharge piping high points. N . By a visual inspection which verifies that no loose debris c. ~ (rags, trash, clothing, etc.) is present in the. containment which could be transported to the containment emergency sump and cause restriction of the pump suction during LOCA conditions.: This visual : inspection shall be performe' l 1. For all accessible areas of the contairm !nt prior to establishing C0!/TAlHMENT INTEGRITY and j j i i 2. Of the areas affected within containraent'at the completion of each containment entry when CONTAINMENT INTEGRITY is c <.tabli s hed..- d. At letst once per 18 months t,y: j j 1. Verifying that the iritetlocks: a) Close DH-11 end DH-12 and deenergire the pressurizer heaters, if either DH-11 or DH-12 is open and a simult.ted reactor coolant system kressure which'is greater than the trip setpoint-(<438'psig) is applied. The interlock to close OH-ll, and/or DH-12 is not' required if-the valve is closed. and 480 V' AC ' power is disconnected from.its motor ope'rators, b) Prevent the opening' of DH-11 and DH-12 when a simulated. or a>ctual reactor cocl6nt system pressure which is greater than the trip setpoint-(<438-psig) is applied. I 2. a) A visual inspection of the containment emergency sump 1 i! which verifies that the subsystem suction inlets are not I restricted by debris and that the sump components.(trash; racks, screens, etc.) show no' evidence of structural distress or corrosion. b) Verifying that on a Borated Water Storage Tank '(BWST) s Low-Low Level interlock trip, the BWST Outlet Valve HV-DH7A (HV-DH78) automatically close.in <75 seconds.. a {ter the operator manually pushes the coiitrol switch to .d open the'Contiir. ant Craergency Semp Valvc HV-DH9A (HV-DH98)' ~ which should bc verified to open in,<75 seconds. 3. Verifying a total leak rate < 20 galle' ns' per h'uur for the ~ LPI system at: a) Normal operating pressure or hydrostatic test pressure of > 150 nsig for those parts of the system downstream of the. pump suction isolation valve, and. b)' > 45 psig for the piping from the containment emergency sump isolation' valve to the pump 4 suction isolation -valve.- g. [ LDAVIS-BESSE, UNIT:1; 3/4 ' 5',4 : TAmebdment,NoO,; 25, fN 77s l

Tl.]l V -llll0 elJfjk"{ 9JNDfgi b(Qi$ MUUDIl0NAL CHANGES FHEVIOUSLV - ~~ tiitL i Wth b vo!J .. PROPOSED BY LETTER 53J . y Serial No._(Ek5 . Date 7-31-8to q I nIIEIUR ni tUIU UnlJ TABLE 3.6-2 CONTAINMENT ISOLATION VALVES (Continued) PENETRATION VALVE ' ISOLATION "I NUMBER NUMBER FUNCTION TIME .] i (seconds),. 67-CV5090-Hydrogen Dilution System. Supply 60-68A SS235A Pressurizer Quench Tank Sample. 30 68A SS235B Pressurizer Quench Tank Sample' 30 68B LV501CB Containment. Air Sample 15 ] ESB' CV50lla ontainment Air Sample 15 j '60 -- ] 69 .CV5065 . Hydrogen Dilution System Supply. ~' 71B CV5010A Containment Air Sample. 15-718 CV5011A Containment. Air Sample-15. 71C CV1544 Core Flood Tank N2. Fill 10' ] 738 CV5010C Containment Air Sample' 15 73B CV50 llc Containment Air Sample 15 748 CV50100 Containment' Air Sample 15-74B CV50llD Containment Air Sample ' 15 ) B. CONTAINMENT PURGE AND EXHAUST ISOLATIO,N 33 #f CV5005 Containment Vessel' Purge inlet Line. 10 33 pd CV5006 Containment Vessel Purge Inlet Line~ 10 34 /! CV5007 Containment Vessel Purge Outlet Line' 10-34 ## CVS008-Containment Ves'sel Purge Outlet Line 10' C. OTHER 5 # SW1365 Containment Air Cooling' Units SW Inlet Line N/A L 6 # SW1368 Containment Air Cooling Units 'SW l Inlet Line, N/A 7 4 SW1367 Containment ' Air Cooling. Units SW Inlet Line 'N/A 9 f SW1356 Containment Air Cooling Units SW I Outlet Line .. N/ A.- j l l Amendment: No'. ( 81.7 5 DAVIS-BESSE, UNIT 1 3/4 6 _

A g,907 ADDili0NAL CHANGES PREVIOUSLY PROPOSED BY LETTER 53 l SedalNa_lRiz5 ' Date 7-3/-% TABLE 3.6-2 ] a CONTAINMENT ISOLATION VALVES (Continued) I j PENETRATION YALYE ISOLATION NUMBER NUMBER FUNCTION _ TIME ) (seconas), j 10 f. SW1358 ' Containment Air Cocling Units SW N/A W Line l Ou tle t 11 't SW1357 Containment Air Cooling Units SW Outlet Line N/A 17 CY343 Containment Yessel Leak Test Inlet l Line 17 Flange Co'ntainment Yessil Leak Test inlet Line (Inside Containment) N/A I 19 i HPS/ High Pressure Injection Line N/A 20 f HPS6 High Pressure' Injection Line 4 N/A f 22 i HP49 High Pressure Injection Line. N/A 23 i SF1 Fuel Transfer Tube N/A 23 Fiange Fue1 Trans!ar %be fi/A ' 24 # ST2 Fuel Transfer Tube N/A 24 Flange Fuel Transfer Tube N/A

25 CS33 Containment Spray Line N/A
25 CS17 Containment Spray Line N/A 25 SA536 Containment Spray Line N/A 25 SA532 Containment Spray Line N/A
25 CS36 Containment Spray Line N/A
26 CS18 Containment Spray Line N/A 25 SA535 Containment Spray Line N/A 26 SA533 Containment Spray Line N/A 27 i DHlA Low Pressure. Injection Line N/A 27 i DH76 Low Pressure Injection Line N/A 13 i DH1B Low Pressure Injection Line~

N/A 28 i DH77 Low Pressure t-i;;; tin Line 'N/A Injection l 4 DAVIS-BESSE, UNIT 1 3/4 6-20 hnendment'No. 3

f..

'37;

du<a1 /do 7. g 5'I CONTAINMENT SYSTENS SURVEILLANCE REQUIREMENTS (Continued) 2. Verifying that the system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.S.a. C.5.c and C.5.d of Regula-1 tory Guide 1.52, Revision'1. July 1976, and the. system- ) flow rate is 8,000 cfm +10%. .j ( 3. Verifying within 31 days after removal that-a laboratory I analysis of a representative carbon sample obtained in / ? accordance with Regulatory Position C.6.b of Regulatory d Guide 1.52 Revision 1 July 1976,Lmeets the laboratory testing criteria of Regulatory Position C.6.a of Regula-j tory Guide 1.52. Revision 1 July 1976.* a j i 4. Verifying a system flow rate of 8,000 cfm +10% during system operation when tested in accordance with ANSI H510-1975. After every 720 hours of charcoal adsorber operation by verify-l c. ing within 31 days after removal that.a laboratory, analysis of i 1 a representative carbon sample obtained in accordance with l Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 1, July 1976, meets the laboratory testing criteria of Regula - tory Position C 6.a of Regulatory Guide 1.52. Revision 1 July 1976.* I d. At leart once per 18 months by: 1. Verifying that-the pressure drop across the combined HEPA filters and charcoal adsorber banks isi 6 inches Water-Gauge while operating the system at a flow rate of 8,000 cfm + 10%. ~ 2. Verifying that the system. starts automatically an any. containment isolation test = signal. e l 3. Verifying that the filter cooling bypass-valves can-be manually opened. **--- l

Representative samples. of used activated carbon from the.EYS shall pass the laboratory test given in Table 3 for an activated carbon bed depth of 2 inches (i.e...the two 2 inch filter beds in series shall be tested per Test 5.b in Table 2 at a relative humidity of 70% for a methyl iodide penetration of less than 15). The pre-and postdoading. sweep medium temperature shall-be (d'C for Test' 5.b of

.i . Table 2, Regulatory Guide 1.52, Revision 1 Juiy 1976. '^::dg y r-'" - - ;7; J

,, L..f.

00 Jmgd d' . ~. DAVIS-BESSE, tait 1 3/4 6-29 ' Amendment No. 43 a

M'O2 /NOT ] p9. ss s _( j i PLANT SYSTE_ljs 3/4.7.2 STEAM GENERATOR' PRESSURE / TEMPERATURE LIMITATION i LIMITING CONDITION FOR OPERATION-3.7.2.1 The temperature of the secondary coolant in' the steam generators shall be >:110 F when the pressure of the secondary coolant in the. j steam generator is > 237 psig. APPLICABILITY: At all times.- ACTION:' With the requirements of the above specification not satisfied: Reduce the steam generator pressure. to < 237-psig' within L30 a. minutes, and-b. Perform an engineering evaluation to determine the~effect of-overpressurization on the structural integrity of' the steam generator.. Determine that the steam generator remains acceptable for continued operation prior to increasing its pressure-above 237 psig. SURVEILLANCE REQUIREMENTS coelant 4.7.2.1 The temperature of.the secondary g,pahnff in each steam generator l shall be determined-to be >'110*F at least once per hour when s'econdary, 4 pressure.in the steam generator is > 237 psig and T,yg:.is < 200*F., i l l 1 1 (; ~ 3/4 7-13 DAVIS-BESSE, UNIT 1 A r

& lh M56 1 REFUELING OPERATIONS l STORAGE P0OL VENTILATION LIMITING CONDITION FOR OPERATION l a I 3.9.12 Two independent emergency ventilation systems servicing the { storage pool area shall be OPERABLE. I APPLICABILITY: Whenever irradiated fuel is in the storage pool. ACTION: ) i With one emergency ventilation system servicing the storage pool a. area inoperable, fuel movement within the storage pool or. crane l operation with loads over. the. storage pool may proceed provided the j OPERABLE emergency ventilation system servicing the-storage pool-area is in operation and discharging through at.least one train of j HEPA filters and charcoal adsorbers. b. With no emergency ventilation system servicing the storage pool' l area OPERABLE, suspend all operations involving movement of fuel I within the storage pool or crane-operation with loads over the storage pool until at least one system is restored to OPERABLE status. The provisions of Specificattuns 3.0.3 and 3.0.4 are not ' applicable. c. l SURVEILLANCE REQUIREMENTS 4.9.12,1 The above required emergency ventilation system servicing the storage pool area shall be demonstrated OPERABLE per the applicable-Surveillance Requirements of 4.6.5.1, and at least once per 18 months by - verifying that.the emergency ventilation system servicing the storage pool area maintains the storage pool area at a negative-pressure of- > 1/8 inches W Mr M p r system operation. key.ela tive to the outside atmosphere 'during { ge i 4.9.12.2 The normal storage pool ventilation system shall be demonstrated L OPERABLE 'at least once per 18 months by verifying that the system fans stop L automatically and that dampers automatically divert flow into the emergency ventilation system on a fuel storage area high radiation test signal. l DAVIS-BESSE, UNIT 1 '3/4 9-12 1 1 ______5

M Nd7 j Q57 SPECIAL TEST EXCEPTION REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specification 3.4.1 may be suspended during the performance of s.tet@ and PHYSICS TESTS provided: { $ TART @ a. The-THERMAL POWER does not exceed Si of RATED THERMAL POWER, and b. The reactor trip setpoints on the OPERABLE High Flux. channels are set < 25" of RATED THERMAL POWER. APPLICABILITY: MODE 2. ACTION: r" of RATED THERMAL POWER, imme-With the THERMAL POWER greater than diately open the control rod drive trip' breakers. 1 i SURVEILLANCE REQUIREMENTS 3 4.10.3.1 The THERMAL POWER shall be determined to'be < 5% of RATED-' THERMAL POWER at least once per hour during :t:c;d aad PHYSICS TESTS. \\ STARTtAP 4.10.3.2 Each High Flux Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating'startup or PHYSICS. I TESTS. i 4 0 AVIS-BESSE, UNIT-1 3/4 10-3 i c--_-----. __.,_...---_.---.I.)

M /N7 P)5F 3 s RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE ( l LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas ~ at and beyond the " SITE BOUNDARY (see - Figure 3.11-2) shall be limited to the following: i a. For noble gases: Less than or equal to 500 mrems/ year to the' total i body and less than or equal to 3000 mrems/ year to the skin.. and 1 b. For iodine-131, for tritium, and for all radionuclides in particulate fom with half lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ. APPLICABILITY: At all timos. ACTION: fode With the dose RAff(s) exceeding the above limits 'Without delay l a. restore the release rate to within the above limit (s). b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. I SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance witn the methodology 1 and parameters in the 00CM. 4.11.2.1.2 The dose rate due to iodine-131, tritium, and all radionuclides in-particulate fem with half-lives greater than.8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology-and parameters in the 00CM by obtaining representative samples and performing ~ analyses in accordance with the sampling and analysis program specified.in Table a.11-2. 1 OAVIS-BESSE, UNIT 1 3/4.11 9 , Amendment No. 85 -]

AwsCNo7 rg s ' 3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING. 3/4.12.1 MONITORING PROGRAM g LIMITING CONDITIONS FOR OPERATIGHS 3.12.1 The radiological environmental monitoring program shall be conducted 1 as specified in Table 3.12-1. APPLICABILITY: At all times. ACTION: a. With the radiological environmental monitoring program not being ' conducted as specified in Table 3.12-1, in lieu of a Licensee j Event Report,' prepare and submit to the Connission, in the Annual ( Radiological Environmental Operating Report required by Specifica-1 tion 6.9.1.lyffa description of the reasons for not conducting the '} ) program as. required.and the plans for preventing a recurrence. j b. With the level of radioactivity as the result of plant effluents in an enviraraental sampling medium at a specified location. 1 exceeding the reporting levels of Table 3.12-2.when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Coneission within 30 days, pursuant to Specification' 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive affluents so that the potential annual dose to A MEMBER OF THE PUBLIC is less than the calendtr year limits.of Specification

3. u.1. 2, 3. u. 2. 2, and 3. u. 2. 3.

When more than one of the radio- ~ nuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if: i reporting level (1)

report.ng level (2) concentration ~ (21 +.. 11.

concentration (11-When radionuclides other than those in Table 3.12-2 are detected and are the result of plant affluents, this report shall be submitted if the potential annual dose to A MEMBER OF THE PUBLIC is aqual to or. greater than the calendar year liatts of Specifications 3.11.1.2,.

3. u.2.2 and 3. u.2.3.

This report is not required if the measured level of radioactivity was not the result of plant (Ifluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental;0perating _ Report. c. With milk or fresh leafy vegetable samples unavailable from one or more of the sampla locations required by Table 3.12-1,_ identify : -locations for obtaining replacement samples and if. practica1' add them to the radiological environmental monitoring program within 30 : days. The locations from wtiich sa:cles.were unavailable may then bei deleted from the monitoring program. !!n lieu of-a Licensee Ivent Report and pursuant to Specification 6.9.1'.11,; identify.the cause. of. 1 the unavailtoility of samoles and icentify the.new location (s) for -. obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also -i_nclude in the report a revised figure (s) and table for the 00C.M ' reflecting the new location (s). DAVIS-BES$E, UNIT 1 1- .3/4 12-1 ^LAmendmentiNo.:86'

MOYo7) b6 60MLOGICAL ENVIRONMENTAL MONITORING I NTION I (. Con +\\med ) .ll 'I With specimens unobtainable due to hazardous conditi6ns, seasonal .d. unavailability, malfunction of. automatic sampling equipment and other legitimata reasons, every effort will be made to complete corrective action prior to the end of the next sampling period. All deviations frca the sampling schedule ~will be documented'in the Annual Radiological' Environmental' Operating Report pursuant. ,to Specification 6/1.10. lL! 9 a The provisions' of > Specifications 3.0.3'and'3.0.4'are not applicable. l e. ) SURVEILLANCE RE0VIREMENTS-i 4.12. 1. 1 The' radiological environmental monitoring samples shall e collectee l . pursuant to Table 3.12-1 from the specific locations given in the table-and-l figure (s) in the 00CM and shall be analyzed pursuant to the requirements of i . Table ~ 3.12-1, and the detection capabilities required by Table 4.12-1. d 4.52.1.2 Cumulative potential' dese contributions: for the current Lealendar - year from radionuclides detected in environmental. samples shall be determined o in accordance with the methodology and parameters in the 00CM. I g i 3j

1

-DAVIS-BESSE, UNIT.1

3/4~12 2

. Ainendment No.86 l l-J 1

MO.7 $6/ J 1 i APOLICA3!LITI 1 BASES i a other specified conditions are satisfied. In this case, this would osan that. for one division the emergency power source must be OPERABLE (as must be tne cocoonents supplied by the emergency power sourca) and all reduncant systems, suosystems, trains, components and devices in the other division must be OPERA 3LE, or likewise satisfy Specification 3.0.5 (i.e., be cacanle of our-i forming their design functions and have an emergency power source OPERA 8LI), In other words, both emergency power sources must be OPERABLS. In other words, j both emergency power sources must be OPERA 8L! and all redundant systems, sub-j systems, trains, comconents and devices in -coth divisions must also be OPERASLI. If these conc!tions are not satisfied, action is required in accordance with. tais specification. In MODES 5 or 6, specification 3.0.5 is not applicable, and thus tne individual ACTION stataments for each applicable Limiting Condition for Operation in y these MODES must be adhered to. \\ s i I G l DAVIS-BESSE, UNIT 1 B'3/4 lb Amendment No. 71 )

c .)uld%AlYO7 hb1 3/4.3' INsnUMENTATION f BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND 5h rIY 5T5 TEM INSTRUMENTATION ~ The OPERABILITY of the RPS, SFAS and SFRCS instrumentation systems ensure that 1) the associated action and/or trip will be initiated when the. parameter monitored by.each channel or combination thereof. exceeds its setpoint, 2) the specified coincidence logic is maintained 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional ) capability is available for RPS, SFAS and.SFRCS purposes from diverse parameters. The OPERABILITY of these systens is required to provide tho' overall-reliability. redundme and diversity assumed available in the facility design for the protection and mitigation of accident and transient con. d i t. ions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses. The surveillance requirements specified for these systems ensure that the overall systee functional capability is maintained comparable to the original design standards. The periodic surveillance tests. performed at the minimum frequencies are sufficient to demonstrate this capability. The measurement of response time at the specified frequencies provides assurance that the RPS, SFAS, and SFRCS action function associated with each channel is ccepleted within the time limit assumed in the i afety analyses. No credit was taken in the anal s ith response times indicated as not applicable. yses for those channel, w I Response time may be demonstrated by 'any. series of sequential, overlapping or total channel test measurements provided that such tests l i i demonstrate the total channel response. time as defined. Sensor response-time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with' certified response times. An SFRCS channel consists of 1) the sensing device (s)', 2) associated ' logic and output relays includin Valves and Turbine Trip)(, and.3) g Isolation of Main Feedwater Non Esse powe.r sources. Safety-grade anticipatory reactor: pump turbines. percent of RATED THERMA)(trip is initiated by a trip-(above 25 POWER) or trip of both main feedwater This anticipatory trip ' rill operate in advance of 'the reactor coolant system high pressure reactor trip to reduce the peake reactor coolant systes operated relief valve. pressure and thus reduce challenges to the power This anticipatory reactor trip system was- ~ installed to satisfy item II.K.2.10 of NUREGe0737 DAVIS-BESSE Unit 1 B'3/4 3-1 Anendment No. 73 - j

p Mo7 QW i CONTAINMENT SYGTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure th'at 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psi and 2) the containment peak pressure does not exceed the design pressure. 4 of 40 psig during LOCA conditions. The maximum peak pressure obtained from a LOCA event.is 37 psig. The limit of 1 psig for initial positive containment pressure.will limit the total pressure to 38 psig which is less than.the design pressure and is consistent with'the safety analyses. p.6.1.5 AIR TEMPERATURE The limitations on containment-average air temperature ensure than the overall containment average air temperature.does not exceed the initial temperature condition assumed in the. accident analysis for a LOCA. 3/4.6.1.6 CONTAINENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the contain-ment steel vessel will be maintained comparable to the original design standards for the life of the facility. ' Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 38 psig in the event of a LOCA. A visual inspection in conjunction with Type A i leakage tests is sufficient to demonstrate this capability. j 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The limitation on use of the Containment Purge a'nd Exhaust System i limits the time this system may be in operation with the reactor coolant system temperature above 200*F. ' This restriction minimizes the time - that a direct open path would exist from the containment atmosphere to the outside atmosphere and consequently reduces the probability that an l 'ident dose would exceed 10 CFR 100 guideline values in the event of ~a, [ LOC.;;#:; coincident with purge.systen operation. The use of this { system is.therefore restricted to non-routine usage not to exceed 90 occurrM hours in any consecutive 365 day period which is equivalent to approximately 11 of the. total possible yearly unit operating time.. 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTENS - 3/4.6.2.1 CONTAINMENT SPRAY SYSTE_M The OPERABILITY of the containment spray system ensures that contain-ment depressurization.and cooling capability will be available in'the event of a'LOCA. The pressure reduction and resultant : lower containment DAVIS-BESSE, UNIT 1 8 3/4 6-2

W WO7l gpy CONTAINMENT SYSTEMS _ BASES J 3/4.6.5.2 SHIELD BUILDING INTEGRITY. SHIELDING BUILDING INTEGRITY ensures that the release of radioactive material from the. containment vessel' will be restricted'to those' leakage-paths and associated leak rates assumed in the safety analysis. The j closure of.the airtight doors and blowout, panels listed in Table 4.6-1 ensure that the. Emergency Ventilation System (EVS): can provide a negative ~ within. the annulus _ between-- 'l. . pressure between 0.25 to. l.5 inches water the shield building and, containment'vesse and'within the interconnecting 'j mechanical _~ penetration rooms after a los of-coolant' accident. xThis restriction, in conjuction with the. ope tion of.the EVS, will limit the site. boundary radiation doses to wi hin the limits fof.10 CFR 100: ~ during accident conditions. l 3/4.6.5.3 SHIELO BUILDING STRUCTURAL INTEGRITY This limitation ensures that' the structural integrity' of the contain-ment shield building will be maintained comparable to the original. design standards for the life of the facility. Structural 11ntegrity is required to provide 1) protection for the steel vessel from external missiles, 2) radiationshieldingintheeventofaLOCA,'and.3)anannulussurrounding the steel vessel that can be maintained'at a negative pressure during accident conditions. -) DAVIS-BESSE, UNIT-1 B 3/4-6-5' i

n. Aris.1 /4'67 g65 PLANT SYSTE:d5 BASES' CoMTAMIN ATt od 3/4.7.8 SEALED SOURCE C"T.""LATION -l, . The limitations on removable contamination for sources requirin leak testing, including alpha emitters. is based.on 10 CFR:70.39(c) g ' limits for plutonium. This limitation will ensure that leakage from oy product, source, and special nuclear material sources will;not exceed allowable intake values. 3/4.7.9 -FIRE SUPPRESSION SYSTEMS The OPERABILITY of the-fire' suppression systems ensures,that. adeouate fire suppression capability is available to confine and~extinquish fires accorrinb ' % = S; in any portion of the facility where safety related equipm nt -l is located. The fire suppression system consists of the water system, spray and/or sprinklers, and fire hose stations. The collective; capability of the fire suppression. systems is_ adequate'to minimize, potential-damage to safety related equipment and is a gajor eler.ent in the facility fire protection program. In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting _ equipment is recuired to be made available in the affected areas until the inoperable equipment is. restored to service. In the event the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides-the major fire suppression caoability of the plant. The requirement-for a twenty-four hour report to the Comrnission provides for promot evalua-tion of the accepta911ity of. the corrective. measures 'to provide adecuate: fire suppression s pability for the continued protection.of the nuclear plant. 3 / 4. 7 '.'10 PENETRATION FIRE BARRIERS The functional integrity of the penetration fire barriers ensures-that fires will.be confined or adequately retarded from spreading to; adjacent portions of the facility. This design feature minimi:esEene-possibility of 6. single fire rapidly involving sever:1 areascof the facility prior to detection and. extinguishment. The per.etration fire barriers are a passive element in the facility! fire protection' program.- and are suoject to periodic inspections 4 During' periods of time when the barriers are not functional, r c:n z q tinuous. fire watch is required to be maintained in'the~ vicinity of the. affected barrier until the barrier is restored.to functional. status. DAVIS-GESSE, UNIT 1-- B3/4.7-6) LAmendmentNo;.9- _ _ __z__ x E

r l} PpS 1 DESIGN FEATURES VOLUME l 5.4.2 The total water and steam volume of the reactor coolant system 'is' 12.110 + 200 cubic feet at a nominal T,yg of 525'F. 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be. located as shown.on Figure 5.1 -1. 5.6' FUEL STORAGE a N CRITICALITY j i 5.6.1.1 ' The spent fuel storage racks are designed and shall be maintained I with: ] a. AK equivalent to less than or. equal to 0.95 when flooded with ubd.ff,0., water, which includes 'a conservative allowance of 1% ontw[awd delta k/k for calculation uncertainty, b. A rectangular array of stainless steel cells spaced 12 31/32 inches on centers iri one direction and 13 3/16 inches on centers in the other direction. Fuel assemblies stored in'the spent fuel pool shall be placed in a stainless steel cell of 0.125 inches nominal thickness or in a failed fuel container.. 5.6.1.2 The new fuel storage racks are designed and shall be maintained with: a. AK equivalent to less than or equal to;0.95 when flooded with unb8fIted water, which includes a conservative allowance of'1% delta k/k for uncertainties as described in Section 9.1' of the FSAR. b. A nominal 21 inch center-to-center distance between fuel assemblies placed in the storage racks.' ORAINAGE 5.6.2 The; spent fuel' storage pool is' designed and shall be maintained to prevent-inadvertent draining of the pool below 9 feet above the, top of.the fuel storage racks. DAVIS-BESSE,. UNIT 1 . 5-5 Amendment No.'19 d .}}

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