ML20138N961

Forwards Addl Corrections to Errors Found in NRC 851010 Transmittal of Proof & Review Tech Specs & as Suppl to Util 851030 & 1107 Ltrs

ML20138N961
Person / Time
Site:	Catawba
Issue date:	12/17/1985
From:	Tucker H DUKE POWER CO.
To:	Harold Denton, Youngblood B Office of Nuclear Reactor Regulation
References
NUDOCS 8512240225
Download: ML20138N961 (21)

Text

{{#Wiki_filter:'Y .. L;. .c. t 3 ,j ~ 'Duxe POWER GOMPAhT P.O. Box 33180 CHARLOTTE, N.C. 28242 HAL B. TUCKER TELEPHONE vsosremarcamT '. (704) 373-4531-pt10Lkaa PSODUCTION L December [17,1985 Mr. Harold.R. Denton,(Director.

Office.of NuclearcReactor-Regulation

.U.-;S._ Nucle'arfRegulatory Commission Washington,-D'. C.l 20555 g . Attention: JMr'.'B. J. Youngblood,. Project Director PWR~ProjectfDirectorat'e No. 4 . R e': Catawba Nuclear-Station,-Units 1 and 2 ^ Docket = Nos..,50--413 and ;50-414 -Proof.and Review-Technical Specifications i< . De'ari Mr.,f Denton : :- In' response to;your-October'10, 1985 letter which transmitted the . Proof ano 9eview' Technical Specifications for Catawba' Units 1 and:2 and'as~a supplement'to my letters _of October 30, 1985 and November o 7,l1985,fattached are-additional' corrections to errors found by our Longoing~ review.- ~ If.youLhavelanyl questions regarding this response please contact

Mr.-Roger W. Ouellette at (704)373-7530.

1.Very.truly yours,-

Hal B.l Tucker

RWO

slb

. Attachment cc: Dr. J.JNelson' Grace, Regional Administrator U.JS.-Nuclear Regulatory' Commission ~ RegionLII-101 Marietta Street, NW, Suite 2900 Atlanta, Georgia-30323 NRCtResident Inspector . Catawba' Nuclear Station g 'PWR = A/RC's TECH SUPPORT AD - J. EMICHT (ltr only) EB (BAL1JutD) l -l EICSB (ROSA) ~ PS8 (CAMMILL) - 8512240225 851217- ] Rss (sERuncER) { g .PDR-ADOCK 05000413l MB (BENAROYA) t P :. _, _, _. PDR . vg -

n + w =, ll go; Neck. Yene. Aom - 7742. ~ 62.) 6ND '3' o

3

-{ ~ l ~ iJ TABLE 3.3 f S' g ~ FIRE DETECTION INSTRIMENTS - il .g ^ ~ FIRE g MINIMUN INSTRt#ENTS OPERA 8LE* ~ ZONE-DESCRIPTION ~ LOCATION Swat FLNE HEAT-FUNCTION **- q o m .g [ w. 1 R.H.R. Pump 18 GG-53 E1.522 + 0-1; O. 1 'A i g 2' 'R.H.R. Pump.IA FF-53 E1.522.+ 0-1 0 1 A. ~ i 3 3 Cont.' Spray' Pump 18? 'GG-54 - E1.522 + 0 3- -0 3 A f 4 Cont. Spray Pump 1A GG-55 E1.522 + 0~ '2 0 '2 A = i 5 R.H.R. Pump 28 GG-61 :E1.522 + 0 .1 0' 1. A 1 6 '- .R.H.R. Pump 2A FF E1.522,+ 0 1 0. -1 A-7 Cont. Spray Pump 28, . GG E1.522 + 0 3 0 3-A 8 Cont. Spray Pump 2A GG-59. E1.522 + 0 .2 0 2 A 9 Aux. F. W. Pumps B8-51 -E1.543 + 0 14-0 12(6). A(8)- 10 Mech. Pene. Room JJ-52 E1.543 + 0 3 0 1 A w 11 Corridor / Cables .NN-51 E1.543 + 0 6 0. 6 A j ) 12 Recip. Chg. Pump 'JJ-53 E1.543 + 0 1 0-1 A i 13 Safety Inj Pacup 18 HH-53' E1.543 + 0 1-0 1 A w 4' 14 Safety Inj Pump 1A 'GG-53 E1.543 + 0 1 0 1 A j 15 Cent. Chg. Pump 18 .JJ-54 E1.543 + 0 2 0-2 A i i 16 Cent. Chg. Pump 1A JJ E1.543 + 0 2. 0-2 A 17 Aisles / Cables KK-56 E1.543.+ 0 18 '0 18 A 18 Aisles / Cables EE-55 E1.543 + 0 6 0' 6 A' 19 AFW Pumps (Unit 2) 88-63 E1.543 + 0 14 0 12(6).A(8) 1 Aisles / Cables NN-6,1 E1.543 + 0 6 0 6 4M 22 Recip. Chg. Pump JJ-60 E1.543 + 0 l' 0 1 A 23 Safety Inj. Pump 28 W-60 E1.543 + 0-1 0 1 -A j 24. Safety Inj. Pump 2A GG-60 E1.543 + 0 1 0 1 A .se 25 -Cent. Chg. Pump 28 . JJ-59. E1.543 + 0 2 0 2 A-26 Cent.:Chg. Pump 2A JJ-58 E1.543 + 0 2 0 2. A i j 27 Aisle,s/ Cables KK E1.543 + 0. 20-0 20 A g. 28 Afsles/ Cables EE-58 E1.543-+ 0. 6 0. 6 A i C 29 SW Gear Equip. Room AA-50 E1.560 + 0 7-0 0 A 8 M 30 Elect. Pene. Room CC-50 E1.560 + 0 8 0-0 A o-31' Corridor / Cables 'EE-53: E1.560 + fi 5-0 5 A i 32 Corridor / Cables KK-52 E1.560 + 0 8 0 8 A. j Q 33 Corridor / Cables NN-54 E1.560_+ 0 -10 0-10 A m i i j

A ~Coce; docs l day 4K4Y nl. S'

S:

'o: e A } fl TABLE 3.3-11 (Continued)- .t 9 s . FIRE DETECTION INSTRWENTS. j r FIRE F g NINIMUM INSTRW ENTS OPERABLE 8 'd ZONE: DESCRIPTION LOCATION: 5HimE .FLAE C FUNCTION ** i I L34 Aisles / Cables JJ E1.560 + 0 14 0 14'~ A- [ A-m 35 Motor Control Centers GG-56 E1.560.+ 0 2 0 2 3 36 Cable Tray Access- .FF-56 E1.568 + 0-2 0 2. A 37 Equip. Batteries DD-55 E1.554 + 0; 5 0: 4 A w { '38 Equip. Batteries CC-55 E1.554 + 0 5 0 4 AJ l 39 Battery Room CC-56 E1.554 + 0 17 0 0. ~A 41 SW Gear Equi AA-64 E1.560 + 0 7. 0. O A . Elect. Pene.p. Room 4 I 42 Room CC-65 E1.560 +'0. 8 0 0 -A 43 Corridor / Cables FF E1.560'+ 0- -5 0-5' A J l: 4 45 Aisles / Cables NN-60 E1.560 + 0 13' 0 M A" 46 Alsles/ Cables-HH-59 E1.560 + 0 II 0 33-A w j- ) 47-Motor Control Center GG-58 E1.560 + 0 2 0 .2 A 48 Cable Tray Access FF-58 E1.560 + 0-2- 0 2 A o j. 4 49 Equip.1 8atteries ,00-E0 E1.560 + 0 5 0 4 A F 50 Equip. Batteries CC-60' E1.560.+ 0. 5 0 4 A 51 Battery Room- 'CC E1.560 + 0-17 0 '0 'A 53 SW Gear Equip. Room AA-49 E1.577-+ 0 7 0 0. A 54 Aisles / Cables CC-50 E1.577 + 0 10 0 0 A } 55 Alsles/ Cables NN-52 E1.577 + 0 9 0 9 A. 1 56 Aisles / Cables PP-55 E1.577 + 0 '13 0 13 A: I 57 Aisles / Cables LL-55 E1.577 + 0-11 0 11' A k 58 Aisles / Cables NH-55 E1.577 + O 21 0-21 'A 3 i 59 Motor Control Center EE-54 E1.577 + 0 2 -- 0 2 A. E 60 Cable Room . CC-56 E1.574 + 0 18 0 15 4 Mi 62 SW Gear Equip. Room AA-64 E1.577 + 0 '7 0 0 A E 63 Elect. Pene. Room CC-64 E1.577 + 0 10 0 O A. g. 64 Aisles / Cables PP-62: E1.577 + 0 9 0 9-A u i 65 Alsles/ Cables PP-59 E1.577'+ 0 16 .0 16 A. T l 66 Afsles/ Cables LL-59 E1.577 + 0 11 O !11-A ] o 67 Alsles/ Cables HH-53 E1.577 + 0 21 0 21 A i S 68 Motor Control Center 'FF-50 E1.577 + 0 2 0 2 A 69 Cable Room CC-59 E1.577 + 0 18 0 15 A N 1 v = -+-

4 N =. . TABLE 3.3-11 (Continued): f-g . FIRE DETECTION'INSTRWENTS b ~ FIRE c j' .{ ZONE. DESCRIPTION.. -MINIMUM INSTRUMENTS OPERA 8LE*- LOCATION 5mmE FLAME _ HEAT FUNCTION **. -m i. 71' Elect Pene. Room CC-51. 'E1.594 + 0 0 A 72 Control Roon. CC-56' -E1.594 +:0' 7 ES 0 6 A y 8 173 Vent. Equip. Room FF E1.594 + 0 0-0 A i 74 Aisles / Cables LL-56' E1.594 +-0 .25; 0 25 A u i 76 ~Alsles/ Cables PP-54 E1.594 + 0 15 0 15 A 79 Elect..Pene. Room 88-63. E1.594'+ 0- '11s 0 0 A 80' Control Room 88-59. E1.594 + 0 22 0-6.

A 81.

Ven. Equip. Room-FF E1.594 + 0: 22 - 0 .O A 82 Aisles / Cables KK-58 E1.594 + 0. . ~28 0.l 27 A-84 Aisles /CablesJ w 89 Fuel Pool Area #1. NN-58. E1.594 + 0 17 0 17 A -PP-50 E1.605 + 10 19 7l 19 'A 90 Fuel Pool Area (Unit 2) PP-64 E1.605 + 10 19-7. 19 A' w 128 UHI Bldg. HH-44 E1.550 + 0 2 3-2 A .{_ 4 129 Fuel Pool Purge Room 'NN E1.631 +'6 6' O 6 A-i 130 .UHI 81dg. (Unit 2) HH-71 E1.594 + 0 2 3 2 A' j. 131 Heactor Bldg. 0*-45' Be1. E1.565 + 3 4 0 0 A 132 -Reactor Bldg. '45*-90* Be1. E1.565'+ 3 3 0 0 A 1 l 133 Reactor Bldg. 90*-135* 8el. E1.565 + 3-4 0. O A 8 134 Reactor _ Bldg. 135*-180 Be1. E1.565 + 3' 5 0 0 'A 't } 135 Reactor _B1dg. 180*-225*~ Be1. E1.565 + 3 4' 0 0-A- 8* {' 136 Reactor Bldg. 270*-315' 8el. E1.565 + 3-3 0-0 A. 30 i 137 Reactor Bldg. '315*-0* Be1. E1.565 + 3 8 0 0 A 138 Reactor Bldg. 0*-45* 8el. E1.586 + 3 6 0 0 A- ' !!3 139 Reactor Bldg. 45*-90*.

Be1.'E1.586 + 3.

4 0 0' A j 140 Re~ actor 81dg. 90*-135' Be1..E1.565 + 3 3 0 0 A g i 141 Reactor Bldg. 135*-180-. Be1. E1.586 + 3 8 0 0 -A .N 142 Reactor 81dg. 180*-225' 8el. E1.586 + 3; 5 0 0 A 143 Reactor Bldg. 315*-0*

Be1. E1.586 + 3 5-0 0

A E 144 Reactor 81dg. ~ 0*-45'. liel. - E1.593 ' + 2 14 0 0 A' 145' Reactor Bldg. 45*-90* Pe1. E1.593 + 2 - 17 0 0 A 4 i 146 Reactor Bldg. 90-135' .Be1. E1.593 + 2. 11 'O O A 4 147 Reactor Bldg. 135*-180* Del..E1.593 + 2% TO O-0 A

i l j. ~TA8LE 3.3-11 (Continued): k FIRE DETECTION INSTRimENTS' ( } E 20NE [f W~ _ b FIRE'. DESCRIPTION _LOCATIONL [ N Q h @ ( MCTIO $ ' MINIMELINSTRIBENTS OPmm_E*" ,y" 148 Reactor 81dg.' -180*-225* Be1. E1.593 +.2% 2 0-0' A g" 150 i 149- -Reactor Bldg. 315*-0*' Be1.1 1.593 + 2% .7 u o A Reactor 81dg (Unit 2)' 0*-45* Be1. E1.565 + 3-4- 0 0= A 151 Reactor 81dg (Unit 2) 45'-90* Be1. E1.565 + 3- -3 /* . Js o - A 152 Reactor 81dg (Unit 2) 90*-135*~_ 8el. E1.565 +'3= 4 yo go A L 153 . Reactor Bldg (Unit. 2)

135*-180*

8el. E1.565 + 3 5 A 154 . Reactor 81dg (Unit 2) .180*-225' 8el.'E1.565 + 3 3. 0 0 A { 155 Reactor 81dg (Unit 2) '270*-315* Be1. E1.565 + 3_ 4 0 0 A-156 Reactor 81dg (Unit 2)- 315*-0* 8el. E1.565 + 3' 6 0 0 A 157-Reactor Bldg (Unit 2) 0*-45* Be1. E1.586 + 6 6.- 0 'O_ A i-w 158 Reactor Bldg (Unit 2) 45*-90* LBe1. E1.586 + 6-4 0 O A 1 159 Reactor 81dg (Unit 2) 90*-135' Be1. E1.586 + 6 3-0 0' A l w 160 Reactor 81dg (Unit 2). 135*-180*' 8el. E1.586 + 6 8 0 0 A j-0 161 Reactor 81dg (Unit 2) 180*-225* Be1. E1.586 + 6 5. O_ 0 A-162 Reactor Bldg (Unit 2) .315*-0* Be1. E1.586.+ 6 5 0 0 A 163 Reactor 81dg (Unit 2) 0*-45* Be1.,E1.593.+ 2 13 0. ' O.- A. 164 Reactor Bldg (Unit 2) 45*-90*- i 165 Reactor Bldg (Unit 2) 90*-135*- Be1. E1.593 + 2% 17 0 0-A ^ Be1. E1.593 + 2% 13 0 0; -A

D -

1 166 Reactor Bldg (Unit 2) 135*-180* Be1. E1.593.+ 2 10 0 0 A-O o 167 Reactor 81dg (Unit-2)

180*-225*

Be1. E1.593 + 2% 2 O 0 A m-168 Reactor Bldg (Unit 2) 315*-0* 8el. E1.593 + 2% 7 0 0 A 2* I 169 RCP-1A Reactor 81dg. 'E1.593 + 2% 0 0 1 A

D l

170. RCP-18 Reactor Bldg. E1.593 + 2's _ 0 0 1 A E ) 171-RCP-1C Reactor B1dg. E1.593 + 2% 0 0 1 A i 172 RCP-ID Reactor 81dg. E1.593 + 2% 0 0 1 A i 173 RCP 2A 45' 8el. E1.593 + 2% 0 0 1 A g, 174 RCP 28 135' 8el. E1.593 + 2% 0 0 1 .A o cm 175 RCP 2C 225' Be1.-_E1.593 + 2% 0 'O 1 A l Q 176 'RCP 20 315' Be1. E1.593 + 2 0 0 1 A 177 Filter Bed Unit 18 Reactor.81dg.Be1. E1.565 + 3. 2. 0 2-A i a 178 Filter Bed Unit IA Reactor 81dg.8el. E1.565'+ 3-2 0 2 A a 179 Filter Bed Unit 2A Reactor 81dg.Be1. E1.565 + 3' 2-0 ;. 2 .A i 85 ) o

.a 00F& Rn/EWCopy TABLE 4.3-8 (Continued) TABLE NOTATIONS (1)' The ANALOG CHANNEL OPERATIONALLTEST shall also demonstrate that aut a. A zisolation of this pathway and control room alarm annunciation' occur if any of the following conditions. exist: e r Instrument indicates measured levels above the Alarm / Trip Setpoint, a.- .or b.~. ' Circuit failure (Alarm only), or c.- Instrument i icates a downsc' ale failure (Alarm only). (2) The initial C N)$E CALIBRATION shall be performed using one or more of ~ the reference t ards certified by the National Bureau of Standards (NSS) c ""or using standards that have been obtained from suppliers that participate in measur.ement assurance' activities with N85. These standards shall ! permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION,' sources that have-been related to the initial calibration shall be used. (3)- CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on , days on which continuous, periodic, or batch releases are made. L ~ m i e s e-e e 9 ~ OCT 7N 1 CATAWBA - UNITS 1 AND 2 3/4 3-84

PROOF & REyggy(gpy REACTOR C0OLANT SYSTEM 3/4.4.'4~ RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERA 8LE. APPLICA8ILITY:. MODES 1, 2, and 3. ' ACTION: a. With one or more PORV(s) inoperable, because of excessive seat leakage, within 1 hour either restore the PORV(s) to OPERA 8LE status or close the associated block valve (s); otherwise, be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. b. With one PORV inoperable due to causes other than excessive seat leakage, within I hour either restore the PORY to OPERA 8LE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STAN08Y within the next 6 hours and in COLD SHUTDOWN within the fo owing 30 hours. c. With more than one P V M noperable due to causes other than exces-sive seat leakage, w thi hour either rostore the PORV(s) to OPERABLE status or close their sociated block valve (s) and' remove power from the block valve (s) and be in HOT STAN08Y within the next 6 hours and COLD SHUTDOWN within the following 30 hours. 'd. With one or more block valve (s) inoperable, within I hour. (1) restore the block valve (s) to OPERA 8LE status, or close the block valve (s) and remove power from the block valve (s), or close the PORV.and remove power from its associated solenoid valve; and (2) apply the ACTION b. or c..above, as appropriate, for the isolated PORV(s). The provisions of Specification 3.0.4 are not applicable. a. 7-g 1B CATAW8A - UNITS 1 AND 2 3/4 4-10 QCT I . - - -. -... - ~

u. . - - - ~ - - - - - - - -

Lp4(TW$ '. C#4blTlodi. Fo4 OPERATr oa/

I'R00F & g LANT SYSTEMS YCOPy ,P SURVEILLANCE REQUIREMENTS APPLICA8ILITY:.Whenever equipment protected by the Spray / Sprinkler System is _ required to be OPERA 8LE. ~ ACTION: With one or more of the above required Spray and/or Sprinkler Systems a. inoperable, within 1 hour establish a continuous fire watch with

backup fire suppression equipment for those areas in which redundant.

I systems or components could be damaged; for Other areas, establish an hourly fire watch patrol. b. _The provisions of Specifications 3.0.3 and 3.0.4 are not. applicable. 4.7.10.2_ Each of the above required Spray and/or Sprinkler Systems shall be ' demonstrated OPERABLE:. At least once per.31 days by verifying that each vane (sanual, power-a. operated,-or automatic) in the flow path which is acceseible during _ plant operations is in its correct position, 4

b. - - At least' once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel,-

At 1sast once per 18 months-by verifying that each valve (manual, c. L power-operated, or automatic) in the flow path which is inaccessible I during plant operations is in its correct position, and c. At least.once per 18 months: 1) By performing-a system functional test which includes simulated L-automatic actuation of the system, and cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.

2). By a visual inspection of each Sprinkler System starting at the 1

L_ system isolation. valve to verify the system's integrity; and 3) By a visual inspection of each nozzle's spray area to verify the j spray pattern is not obstructed. i o i 001 't maa L -CATAWBA - UNITS 1 AND 2 3/4 7-29

.~ -..---.n. . a ~ '. t %bFE #{gg, ~ TABLE 3.7-3 (Continued) FIRE HOSE STATIONS i LOCATIdN' 1 ELEVATION HOSE RACK # 59,'00 574+0 160, AA' 574+0 ~ iRF480-1RF481 49, 88-CC 577+0 1RF490 -45, 88 -577+0 1RF491 55, DO .574+0 1RF492 54, AA 574+0 1RF493

63,~AA-577+0' 1RF993 51, AA 577+0 1RF998 62,~ ' NN 594+0 1RF205 57, MM t 594+0 1RF222 63,'JJ-594+0-1RF231-

57, HH 594+0 1RF245

-57, EE 594+0 1RF253 51, JJ 594+0 1RF259 53, NN 594+0 1RF275 64, 88 594+0 1RF984 50, 88 594+0 IRF985 51, JJ 605+10 -1RF265 63,-JJl 605+10 IRF233 ' 64 M 631+6 1RF483 ' 50-51, M., 631+6 1RF495-L 2.. .ve rPo'ols. ~ 4$ ^ TT-UU' -605+10-1RF208 p 605+10 1RF276 605+10 1RF483 605+10 1RF822 gs '3. -Nuclear Service Water Purp Structure East Section 600+0 1RF939 . West Section 600+0 1RF940 4 O L L CATAWBA - UNITS l'AN0 2 3/4 7-34 =. -... =.. -. ~ -. - . - -. ~.. -.. -... -.. -.

f PLANT SYSTEMS. Efhpyr 3/4.7.13 ' STAND 8Y SHUTDOWN SYSTEM LIMITING CONDITION FOR OPERATION' 3.7.13.The' Standby Shutdown System (SSS) shall be OPERA 8LE. APPLICA8RIT i M00ESb7and-3.s &gh3 AC a. With.the Standby Shutdown System inoperable, restore the inopera51e equipment to OPERA 8LE status within 7 days or be in at least HOT. STAND 8Y within the next~6 hours ar.d in at least HOT SHUTDOWN w the following 6 hours. "b. With the total leakage from UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE and reactor, coolant = pump ~ seal: leakage greater than 26 gpa, declare the Standby Makeup Pump inoperable and take. ACTION a., above. c'. - - The provisions ~ of Specifications 3.0.3 and '3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS I -4.7.13.1 ;The Standby' Shutdown System diesel generator'shall be demonstrated OPERABLE: t ?At least once per 31: days by verifying: ~ a.'

1)~ The fuel level in the fuel storage tank is greafer_ than or equal to 67 inches, and

- 2) ~ The diesel starts from ambient conditions and' operates for at-least 30 minutes at' greater than or equal to 700 kW.~ b. At.least once per 92 days by verifying that a sample of diesel fuel from the fuel, storage tank, obtained in accordance with ASTM-0270-1975, is within the acceptable limits specified in Table 1 of ASTM-0975-1977 when checked for viscosity and water and sediment; and s At least once per 18 months, during shutcown, by subjecting the diesel c. 'to an inspection in accordance with procedures prepared in conjunction with.its. manufacturer's recommendations for the class of service. 4.7.13.2J The Standby Shutdown System diesel starting 24-volt battery bank and charger shall'be. demonstrated OPERABLE: n = At least once per 7 days by verifying that: .a. ~ ~ 1) The electrolyte leve1 ~ of each battery is above the plates; and 2) The'overall battery voltage-is greater than or equal to 24 volts. I CATAWBA - UNITS 1 AND 2 3/4 7-39 OCT 7 1985 i . s _--_..-__,,....-___.-_,,._,..,m, ._,,._,_-,..m_,

i REVlEW COPY ~ TABLE-~3.8-la (Continued) ~ UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTI . DEVICE NUMBER &' LOCATION . SYSTEM POWERED 2.- 600 V_AC MCC (Continued) IMXR-F01B~ . Primary Bkr Incore Instrument Room Ventila-Backup Fuse tion Unit 18 Fan Motor .IMXR-F028 Primary Skr Control Rod Drive Vent Backup Fuse Fan Motor ID IMXR-F03A-Primary Skr Lower Containment Ventilation- -Backup Fuse Unit ID Fan Motor .IMER-F04C ' Primary Bkr~ Upper Containment Ventilation' . Backup Fuse Unit 1D Fan Motor IMXY-F02A Primary Bkr NC Puap 1A 011 Lift Pump Motor 1 Backup Fuse 'IMXY-F028-Primary Bkr NC Pump ID 011 Lift Pump Motor 1 Backup Fuse D IMXY-F03A { ~ l'BackupFuse-Primary Bkr. u. Reactor Coolant' Drain Tank Pump Motor IA .IMXY-F030: Primary Bkr-Ice Condenser Refrigeration Backup Fuse Floor Cool Pump Motor IA-IMXY-F05A Primary Bkr Lighting Transformer Backup Fuse ILR8 IMXY-F058 Primary Skr Lighting Transformer - Backup Fuse ILR11 iMXY-E 02.C Primary & hc.b Ba.; ling 6erdufa/omed B***F Fuse LUeflinj fja.cidne 8ec9 ade 14dfLof85 / (# CATAWBA - UNITS 1 AND 2 3/4 8-35 L.--....-...

~ ^ PROOF & REVIEW C TABLE 3.8-la (Continued) UNIT 1 CONTAIM ENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE .0EVICE NUMBER & LOCATION-SYSTEM POWERED ' 5.. '.120 VAC Pane 1 boards (Continued)- IKPN Primary Bkr Backup Fuse- . NC Pump Motor ID Space Heater 1KPN-7-1 Primary Bkr' Lower Containment Vent Unit Backup Fuse-18 Fan Motor Space Heater 1KPN-8-l' Primary Bkr Lower Containment Vent Unit . Backup Fuse-ID Fan Motor Space Heater IKPN Primary Bkr.. Misc Control Power Backup Fuse ' for 1ATC 24 I. T U Ideldib} Cirem*fs A EQCB oool Lower Coob. nment Petawy Skr - AA pg,,, g, g ;f %kg % AB Q 'i 4. W kO**L ye, Con /ao'nmenf y f riq a q % - M D c W e / / /nl deWe[ S Q Skr - AB 1M CATAWBA - UNITS 1 AND 2 3/4 8-43 - - -=. =. - -.

PROOF & ggYlDYC TABLE 3.8-1b (Continued)

UNIT-2 CONTAllmENT PENETRATION CONDUCTOR OVERCURRENT PROTEC

, DEVICE NUMBER & LOCATION SYSTEM POWERED L2. 600 VAC MCC (Continued) - ~ 2EMXS-F05A-Primary Skr S/G 2A Blowdown Inside cont -Backup Fuse ~ Isol Viv 2BB56A 2EMXS-F05B Primary Bkr LS/G 2C Blowdown-Inside Cont (,,,,,, . Backup Fuse Isol V1v 2BB60A

2EMXS-F05C Primary Bkr:

Pzr Liquid Sample Line Inside

. Backup Fuse-Cont Isol Viv 2NM3A 2EMXS'-F06A-

Primary Skr Pzr Steam Sample Line Inside Backup Fuse Cont Isol V1v 2NM6A 1

2EMXS-F068. Primary Skr. NC Hot' Leg A Sample Line Backup Fuse Inside Cont Isol V1v 2NH22A 12EMXS-F06C ~ Primary Bkr NC Hot tsg C Sample Line . Backup Fuse Inside Cont Isol Viv 2NM25A 2MXM-F01A Primary Bkr Reactor Coo an Pump Motor Backup Fuse Drain Tank'P Motor 2MXM-F02A Primary Skr NC P 11 Lift Backup Fuse._ Pump Mo 1 2MXM-F028 . Primary Bkr NC Pump 2C 011 Lift Backup Fuse-Pump Motor 1~ 2MXM-F03A Primary Skr Ice Condenser Power Backup. Fuse Transformer ICT2A 2MXM-F03B~ ~ Primary Bkr Ice Condenser Air Handling Unit r 4 Backup Fuse-286 Fan Motor A & B 1# ocT . CATAWBA - UNITS 1 AND 2 3/4 8-51 =._,_.-,...u.u_._.__....___..,_..--.-_.___.___-.~.-_._..______._-__._._.

r ROOF & REVIEW COPY E ~-TABLE 4.11-l'(Continued) TABLE NOTATIONS ^ (1)The LLD is' defined, for purposes of these specifications, as the smallest . concentration of radioactive material in a sample that will yield a not coun', above system. background, that'will be detected with 95 probability c with only M probability of falsely concluding that a blank observation -represents a "real" signal. . For. a particular measurement system, which may include radiochemical separation: .-4 . 4.66 s ' b LLD =: G E V: 2.22 x Iff* Y exp (-Aat) .Where: -LLD =~ the "a priori" lower limit'of detection (microcurie per unit nass or volume),

sg = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate-(counts per

~. , minute), E =:the counting efficiency-(counts per disintegration),- V = the sample size (units of mass or volume), 2.22 x 10e = the number.of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable,- g p-r oactive, decay constant for the particular radionuclide A '(M)', and - C s.ee-8 t = the lapsed time between-tpe midpoint of sample collection and of counti (g Typical values of E, V,-Y j d At should be used in the calculation. It'should be recognized that the LLD is defined as an a priori (before the fact)-limit representing the. capability of a measurement system and' not as an a posteriori (after the fact)' limit for a particular measurement. 2(2)A batch release is the d<scharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling. n ~ 1 ' CATAWBA l-UNITS 1 AND 2^ 3/4 11-3 'l M 00T v w v w - ~ tr -ww-*=e wwww w te - *e we gee.+, e-ww---m--w ,--e eae-wi',--,.,,wer* n w s w,,e - w t -w, - w = e-o r ew-vm e > y, -,,m--g,wv-y--w-w-

• +w--

eg y-w v +-*w---m

. ~ PR00F4 REVM06 RADICACTI E EFFLUENTS CHEMICAL TREATMENT PON05 . LIMITING CONDITION FOR OPERATION 3.11.1.5 treatment pond sha}1'be limited by the following expression:T A. ~ 264 I T, j 1 < 1.0 t C 3 .g ~ excluding tritium and dissolved or entrained noble gases, Where:- A = pond inventory limit for single radionuclide "j", in Curies; y ~C) = 10 CFR Part 20, Appendix B Table II, Column 2, concentration for single radionuclide "j", microcuries/al; V _=. design volume of liquid and slurry in the pond, in gallons; and ~ 264 = conversion unit microcuries/ Curie per milliliter / gallon. APPLICABILITY: At all times. -ACTION: With the quantity of radioactive material in any of the above listed .a. ponds exceeding the above-limit, immediately suspend all additions - cf radioactive material to the pond and initiate corrective action to reduce the contents to within the limit. <b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.1.5 -The quantity of radioactive material contained in each batch of resin / water slurry to be transferred to the chemical treatment ponds shall be determined to be within the above limit by analyzing a repres n ive sample of the batch to be transferred to the chemical treatment po s, s 11 be limited by,the expression: I < 0.006 -j C 3 Where: c) = radioactive resin / water' slurry concentration for radionuclide "j" entering the UNRESTRICTED AREA chemical treatment ponds, in microcuries/ milliliter; and C) = 10 CFR Part 20, Appendix 8.-Table II, Column 2, concentration for single radionuclide "j", in microcuries/ milliliter.

f s0# q#

CATAWBA - UNITS 1 AND 2 3/4 11-8 t _ _ _.... _ _,. _ _ _ ~. -... ~. - _..,. _ -., _. _... - _ - ... _. _ _ - _. _, _. _,. _,. - ~,

50f & REVIEW COPY ~ TABLE 4.11-2 (Continued) TABLE NOTATIONS (1) The'LLD is ' defined, for purposes of these specifications, as the smalle'st concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. 'For a particular measurement system, which may include radiochemical separation: 4.66 s 2. LLD = b + E V .2.22 x 108 Y exp (-Aat) Where: LLD = the "a priori" lower li,mit of detection (microcurie per unit mass or volume), e = the standard deviation of the background counting rate or of b the counting rate of a blank ' sample as appropriate. (counts per minute), 'E =~the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22 x 10s = the number of' disintegrations per minute per microcurie, ~ 4 Y = the fractional radiochemical yield, when applicable, 6,

tive decay constant for the particular radionuclide I-(s4),and ksec-'

t = the eJa sed ti 4M' e midpoint of sample collection and _timedfcounti (L). s sec. Typical values of E V, Y, At should be used in the calculation. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. / .g is 85 g is *. CATAWBA'- UNITS 1 AN. 2 3/4 11-11 g 'l

.s- / PROOF & REVIEW COPY 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CON 0! TION FOR OPERATION 3.12.1 The Radiological Environmental Monitoring Program shall be conducted as specified in Table 3.12-1. APPLICABILITY: At all times. ACTION: With the Radiological Environmental Monitoring Program not being

a. -

conducted as specified in Table 3.12-1,' prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.6 a description of the reasons for not conducting the program as requ, ired and the plans for preventing a recurrence. b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of-Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission 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 effluents so that the potential annual 4 dose

• to a MEMBER OF-THE PUBLIC is less than the calendar year limits of Specification 3.11.1. 2, 3.11. 2. 2, or 3.11. 2. 3.

When more than one of the radionuclides in Table -2 are detected in the sampling medium, this report shal -be s tted if: a concentration ( ) - con ntration (2) + ...) 1.0 reporting level (1)($) re rting level (2) ~ .When radionuclides ot h t an those in Table 3.12-2 are detected and L are the result of plant effluents, this report shall be submitted if the potential annual dose

• to A MC ~,ER OF THE PUBLIC from all radio-nuclides is equal to or greater than the calendar year limits of Specification 3.11.1.2, 3.11.2.2, or 3.11.2.3.

This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating' Report required by Specification 6.9.1.6.' m h "The methodology and parameters used to estimate the potential annual dose to a MEMER OF THE PU8LIC shall be ' indicated in this report. CATAWBA - UNIT *, 1 AND 2 3/4 12-1 Jyt 181985 g64 L 06 pt..

PROOF & REVIEW COPY RADIOLOGICAL ENVIRONMENTAL NONITORING LIMITING CONDITION FOR OPERATION ACTION -(Continued) I' g Withafik'orfreshleayvegeteldefs c. s unavailable from one or.

more of the sample loc tions rgered by Table 3.12-1, identify-specific locations for obtatnTng replacement samples and add them within 30 days to the Radiological Environmental Nonitoring Program given-in the 00CM.

unavailable may then be deleted from the monitoring pr Effluent Release Reto Specification 6.14, submit-in the next Semiannua Pursuant a revised figure (s) port documentation for a change in the 00CM including and table for the 00CM reflecting the new loca-tion (s) with supporting information identifying the cause of the location (s) for obtaining samples. unavailability of. samples a v d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS '4.12.1 The radiological environmental monitoring samples shall be collected- ~ . pursuant to Table 3.12-1 from the specific locations given

Table 3.12-1 and the detection capabilities required by Table 4.12-1.

i ~ l 1 i ) \\ 5 e 6 CATAWBA - UNITS 1 AND 2 3/4 12-2 JUI 131!85 OCT 7 1985

,,_4... ii.y- ~ /- -.. d '....+ 4 =. Il t W& REVIEWCOPY DRAFT ~ TA8LE'4.12-1 (continued)' TA8LE NOTATIONS. -(1) This list does not mean that'only these nuclides are to'be considered. Other peaks that are. identifiable, together with those of the above : 1 i nuclides, shall also be analyzed and reported in the' Annua 1' Radiological Environmental Operating-Report pursuant'to Specification 6.9.1.6. t '(2) ' Required detection capaht11 ties for thermoluminescent dosimeters used . for environmental measurements shall:be in accordanc -tions of Regulatory Guide 4.13. I(3) The LLD is defined, for purposes of these specifications, as the. smallest concentrations of radioactive material-in a sample that will yield a net M with only 55 probability of falsely concluding that a blan .; represents-a "real". signal. i For a particular measurement system, which may include radiochemical separation: 4.66 s b 7 E -- - '- V 2.22 Y ' exp(- Aat) g. Where: t LLD = the "a priori" lower. limit of detection (picoCuries per unit 1 mass or volume), =.the standard deviation of the background counting rate or of the sb counting rate of a blank sample-as appropriate (counts per minute). E' = the counting efficiency (counts por disintegration) 'V = the sample size (units of mass--or volume), 2.22 = the number of disintegrations per minute per picncurie, Y. = the fractional' radiochemical. yield, when applicable,. ,'A ~ onctive decay constant for the particular radionuclide b (t>l()s,ec~l L At: = the elap time-between environmental col 1ection g en f 'the s e collection period, and time of counting (#). C-Sec j Typical values of E, V, Y and'At should be used in the calcula) i s n. f CATM4A'--UNITS 1~AHO 2 3/4 12-11 JUL 181985 OCT 7 1985 O; . ~.. -.. - ......_.-.._....._---,.__,_,-.-m.._.-. _ _ ~ -,, -.

e u

INSTRUMENTATION PROOF & REVIEW '8ASES = U,. s -FIRE DETECTION INSTRUMENTATION (Continued') any area. 'As a result the establishment of 'a. fire watch patrol must be .that provide'only early fire warning. 3 initiated at an earlier stage The establishment of frequent fire patrols in the affected. areas is required to provide detection capability until the inoperable' instrumentation is restored to OPERA 8ILITY. ~ 3/4.3.3.9 LOOSE-PART DETECTION SYSTEM w. sufficient capability is available to detect loose metallic pa -Reactor System and avoid or mitigate damage to React The with the rpcommendations-of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981. 3/4.3.3.10 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUME The radioactive Itquid effluent instrumentation is provided to monitor effluents during actual or potential releases of liquid effluents Alare/ Trip Setpoints for these instruments shall be calculated and adjusted in The alarm / trip will occur prior to exceeding the limits of 10 CFR P OPERABILITY and use of this instrumentation is consistent with the requirements The

of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

!J l 3/4.3.3.11 'RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUM ~ f + The radioactive gaseous effluent instrumentation is provided to monitor f and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous affluents. The Alare/ Trip-Setpoints for these instruments shall be calculated and adjusted in l-accordance with the as,thodology and parameters in the 00CM to ensure that the -alare/ trip will occur prior to_ exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially u plosive gas mixtures in the WASTE GAS HOLDUP SYSTEM.- The OPERA 8ILITY and use of this instrumentation is consistent with' the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50 1 The sensitivity of any noble gas activity monito used to show compliance with the' gaseous effluent release requirements o. ~fPa-tion 3.11.2.2 shall be such that concentrations as low as l'x 106p8Mee are measurable. C q 1985 g . CATAWBA -' UNITS 1 AND 2 8 3/4 3-6 9-e' 7 9-4d In we -e ein-w ofe-%w ge gr e te e p-w-+ s e m-owng-e us tv49 7wFTW*" WW

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". Y F& 5EVIEWCOPY ADMINISTRATIVE CONTROLS l SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) l Release Reports shall include the following infwaation for eachMof solid waste (as defined by 10CFR Part 61) shioped offsite during the report period: Total Container volume, hg a. Total Curie quantity fey w b bined by measurement or estima b. Principal radionuclides (, determined by measurement or estimate), c. 5Pecify whe.h ' d. pse. T 4 ype of wasta dewatered spect resin, compacted dry waste, evaporator bot anJ geesassig employed . Tyfe.F c Mer (e.. LsA, TyrA T S leae. panMyh M f. Solidification age or absorben,t 3 ^ 1 L-Z . g., cement - " ^ ;;. n.2 ( ^weq form 4 Ae.hp e.). l -The Semiannual Radioactive Effluent Release Report to be submitted within 60 day.s after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may wind direction, atmospheric stability, and precipitation (if m the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability." This same report shall' include an assessment of the radiation doses due to the radioactive liquid and gaseous affluents released 's from the unit or station during the previous calendar year. shall also include an assessment of the radiation doses from radioactiveThis same rep liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activiti=s inside the SITE BOUNDARY (Figure 5.1-3) during the report period. tions used in making these assessments, i.e., specific activity, exposureAll assump-time and location shall be included in these reports. The meteorological conditions concurr,ent with the time of~ release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (00CM). .The Samiannal Radioactive Effluent Release Report to be submitted within 2 60 days after January 1 of each year shall also include an assessment of radiation doses to'the likely most expcsed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation". Acceptable methods for calculating the dose contribution from liqufd and' gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

• In lieu oi' submission with the Semiannual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

OCT 18 ' CATAWBA - UNITS 1 AND 2 6-18 l -., _ _... _ - - - - - - - - - - - - - -}}