Text

Forwards Listed Tech Specs Accepted by Reviewers as Result of 840918-20 Meetings.Addl Tech Specs Submitted for Review as Listed.Expeditious Closure of Remaining Open Items Requested to Support Tech Spec Certification
	ML20098F526
Person / Time
Site:	Byron
Issue date:	09/20/1984
From:	Tramm T; COMMONWEALTH EDISON CO.
To:	Harold Denton; Office of Nuclear Reactor Regulation
References
	NUDOCS 8410030090
	Download: ML20098F526 (51)
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) One Feret National Plais. Chicago. tilenois d

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~) Address Reply u Post Offic] Bon 767 q

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r September 20, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission i

Washington, D.C.

20005

SUBJECT:

Byron Generating Scation Units 1 and 2. Technical I

Specifications, NRC Docket Nos. 50-454 and 50-455 Dear Mr. Denton l

Per discussions with Mr. Calvin Moon and numerous NRC reviewers, listed below and attached are those pages which Commonwealth Edison understands have been accepted by the principal reviewers as a result of our meeting in Bethesda, Maryland Sept. 18,1984 thru Sept. 20,1984.

PAGE B 2-9 3/4 7-41 3/4 10-4 6-9 3/4 3-38 3/4 7-42 5-4 6-10 3/4 4-2 3/4 8-30 5-5 6 3/4 5-1 3/4 8-31 6-2 6-13 3/4 5-2 3/4 8-32 6-7 6-24 3/4 6-23 3/4 8-33 6-8 Also included with this letter are the following pages which are being formally submitted for your review. These page changes were discussed with the individual reviewers during the meetings and commonwealth Edison feels that mutual concurrence was achieved PACE 3/4 2-14 3/4 3-19 3/4 6-6 B 3/4 2-5 3/4 3-2 3/4 3-22 3/4 6-11 B 3/4 4-1 3/4 3-5 3/4 3-23 3/4 6-12 8 3/4 5-1 3/4 3-6 3/4 4-2 3/4 9-4 B 3/4 6-1 3/4 3-18 3/4 4-20 3/4 9-14 8 3/4 9-1 j

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The following open items remain, as of 9/20/84, as part of the continued Commonwealth Edison /NRC Tech Spec review. The item number identified below corresponds to the Attachment number of the handwritten list provided to Calvin Moon by Commonwealth Edison.

Attachment,#

[$N Sub.iect Status I

4.7.7.d.3

-1/4" pressure Conference call scheduled

]

for 9/24/84 2

4.3.4.2 Turbine Overspeed NRC awaiting formal letterfromCECo(9/24/84) 8 Table 4.11-2 Gaseous Waste Conference call scheduled Analyses for 9/24/84 17 4.7.1.2.3 Diesel Oil Samplina NRC reviewing 4.7.5.3 4.7.10.1.3 4.8.1.1.2 25 3.7.6 Ventilation NRC/ CECO telcon 3.7.7 set for 9/24/84 3.9.12 27 RSB Quastions NRC to determine impact on Tech Specs 29 3.7.5 Ultimate Heat Sink a.) NRC/ Ceco litg.

set for 9/24/84 b.)NRCtodetermine how contingency plan affects Tech Specs 30 Misc.

Omissions, typo's NRC to provide pages from Final Draft 31 Table 3.4-1 RHR Suction Valves Ceco reviewing NRC changes 34 3/4.3-2 Undervoltage NRC reviewing Surveillance Int.

Connonwealth is proceeding with the Tech Spec Certification process based on the August 28 Final Draf t, the 71 pages of corrections, and the attached marked up page changes, connonwealth requests an expeditious NRC closure of the remaining open items to support our Tech Spec Certification.

}'

T. R. Trana Nuclear Licensing Administrator cc: Byron Resident Inspector Senior Tech Spec Coor.

Calvin floon O4

LIMITING SAFETY SYSTEM SETTINGS 9 e/d Ep29T BASES v

Reactor Trip System Interlocks The Reactor Trip System Interlocks perform the following functions:

P-6 On increasing power, P-6 allows the manual block of the Source Range Reactor trip (i.e., prevents premature block of Source Range trip),

provides A backup block for Source Range Neutron Flux doubling, and de energizes the high voltage to the etectors. On decreasing power, M

Source Range Level trips,ar_e automat ally reactivated and high voltage restor y w -

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FL. M 3ciee

,a

%ck On increasing (power +, F-/ au wmu.icaliy enables Reactor trips on low gt P-7 flow in more than one reactor coolant loop, more than one reactor coolant pump breaker open, reactor coolant pump bus undervoltage and underfrequency, Turbine trip, pressurizer low pressure and pressurizer high level.

On decreasing power, the above listed trips are automatically blocked.

P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops.

On decreasing pcwer, the P-8 automatically blocks the single loop low flow trip.

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range Reactor trip and the Low Setpoint Power Range Reactor trip; andautomaticallyblockstheSourceRangeReactortripand,de-energize [

the Source Range high voltage power. On decreasing power, the Inter-mediate Range Reactor trip and the Low Setpoint Power Ranga Reactor trip are automatically reactivated.

Provides input to P-7.\\

l p.s, vu P-13 Provides input to P-7.

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TA8tE 4.3-2 (Continued)

?g ENGINEEPED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION 2

5URVEILLANCE REQUIREMENIS g

IRIP q

ANALOG ACTUAllNG MODES CHANNE L DEVICE MA5fLR SLAVE FOR 1A11C11 y

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACIUAT!DN RLtAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST 1E51 LOGIC IEST 1EST IEST 15 REQUIRED

7. Automatic Opening (Coaticued)

Automatic Actuation a.

Logic and Actuation Relays N.A.

N.A.

M.A.

M(1)

Q 1, 2, 3, 4 b.

SWST Level-tow-Low F5 R

ist. n M.A.

y pt. A,

M.A.

N.A.

1, 2, 3, 4 Coincident With Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements A 8. Loss of Power 2

4.

ESF Bus Undervoltage N.A.

R N.A.

N.A.

M.A.

1, 2, 3, 4 b.

Grid Degraded Voltage M.A.

R N.A.

M.A.

1, 2, 3, 4 9.

Engineered Safety feature Actuation System Interlocks a.

Pressurizer Pressure, M.A.

R M

.N.A.

N.A.

1, 2, 3 P-Il b.

Reactor Trip, P-4 M.A.

N.A.

R N.A.

N.A.

1, 2, 3 Low-Low T,,g, P-12 N.A.

R M

M.A.

N.A.

1, 2, 3 c.

d.

Steam Generator Water 5

R M

N.A.

M(1)

Q 1, 2, 3 Level, P-14 (High-High) eM 8

IABLE NOIAIION k[b D (1)

Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

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9/nh1 REACTOR COOLANT SYSTEM HOT STANDBY l

LIMITING CONDITION FOR OPERATION 3.4.1.2 At least three of the reactor coolant loops listed below shall be OPERABLE and at least two of these reactor coolant loops shall be in operation:*

a.

Reactor Coolant Loop A and its associated steam generator and reactor coolant pump, b.

Reactor Coolant Loop 8 and its associated steam generator and reactor coolant pump, 4

c.

Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and d.

Reactor Coolant Loop D and its associated steam generator and reactor coolant pu:np.

APPLICABILITY:

MODE 3.**

ACTION:

With le'ss than the above required reactor coolant loops OPERABLE, a.

restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT 3HUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With only one reactor coolant loop in operation, restore at least two loops to operation within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or open reactor trip breakers within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

[

c.

With no reactor coolant loo,: in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and irreediately initiate corrective action to return the required reactor coolant loop to operation.

SURVEILLANCE REQUIREMENTS i

4. 0.1. 2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once par 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 The required steam generators shall be detensined OPERABLE by verifying secondary side narrow range water level to be greater than or equal to 41% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.2.3 At least two reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All Reactor Coolant pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2)' core outlet temperature is maintained at least 10*F below saturation temperature, alfit Sam Me 5hk 3.1 o.4-0 8YRON - UNIT 1 3/4 4-2

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3/4.5.1 ACCUMULATORS

v v "v e 9/n N LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with; a.

The isolation valve open, 38 feb b.

A contained barated water level of between.34% and 66%,

X c.

A boron concentration of between 1900 and 2100 ppm, and d.

, A nitrogen cover-pressure of between 617 and 662 psig.

APPLICABILITY:

MODES 1, 2, and 3*.

ACTION:

a.

With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

D.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REOUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1)

Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and 2)

Verifying that each accumulator isolation valve is open.

" Pressurizer pressure above 1000 psig.

BYRON - UNIT 1 3/4 5-1

9l19lfy EMERGENCY CORE COOLING SYSTEMS s

SURVEILLANCE REOUIREMENTS (Continued) b.

At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 70 gallons by verifying the boron concentration of the accumulator solution, At least once par 31 days when the RCS pressure is above 1000 psig c.

by verifying that power to the isolation valve operator is disconnected from the circuit by re :. S; tr.;

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4. 5.1. 2 Each accumulator water level and pressure channel shall be demonstrated OPERABLE at least once per 18 months by the performance of a CHANNEL CALIBRATION.

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BYRON - UNIT 1 3/4 5-2

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Ck CONTAINMENT SYSTEMS n

3/4.6.4 COMBUSTIBLE GAS CONTROL NYDROGEN MONITORS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen monitors shall be OPERABLE.*

APPLICABILITY:

MODES 1 and 2.

ACTION:

a.

With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . -

SURVEILL NCE REQUIREMENTS 4.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance of a CHANNEL CHECX and a check that the monitor is in standby mode at least once per12 hours,anANALOGCHANNELOPERATIONALTESTa(leastonceper31 days,and at least once per 92 days :

!!/.. R;; TE$T a l by performing a CHANNEL CALIBRATION usingg =r'

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Item II.F.1.6.

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BYRON - UNIT 1 3/4 6-23

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PLANT SYSTEMS 3/4.7.12 AREA TEMPERATURE MONIT0ldNG LIMITING CONDITION FOR OPERATION

% ;t J Wp;. M 3.7.12 The temperature of each area shown in Table 3.7-6 shall 5:

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s 3

c'th'- the li:it; i.; dire +-d in 7=h t e 3. '- S. " -

@ APPLICABILITY:mwu th 3 h, c. Ag % h S o*f.

t Whenever the equipment in an affected area is required to be OPERABLE.

ACTION:

g,@ 2 With one or more areas exceeding the temperature limit (s) shown in a.

Table 3.7-6 for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , prepare and s mit to the Commission u

within30 days,pursuanttoSpecification0.7.faSpecialReportthat provides a record of the cumulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment.

b.

With one or more areas exceeding the temperature limit (s) shown in Table 3.7-6 by more than 30 F, prepare and submit a Special Report as required by ACTION a. above, and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either return the area (s) to within the temperature limit (s) or declare the equipment in the affected area (s) inoperable.

SURVEILLANCE REOUIREMENTS 4.7.12 The temperature in each of the areas r.'.own in Table 3.7-6 shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

O e

BYRON - UNIT 1 3/4 7-41

9I61sv TAB 1.E 3.7-6 o

AREA TEMPERATURE MOWITORING AREA TEMP. *F 1.

Misc. Electric Equipment and Battery Rooms 108 2.

ESF Switchgear Ras 108 3.

Division 12 Cable spreading Rm 108 4.

Upper and Lower Cable Spreading Ras 90 5.

Diesel-Generator Rus 132 6.

Diesel 011 Storage Rooms 132 7.

Aux. Building Vent Exhaust Filter Cubicle 122 8.

Centrifugal Charging Pump Room 122 9.

Containment Spray Pump Ro m 130

10. RHR Pump Rooms 130

11. Safety Injection Pump Room 130 sia

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TABLE 3.8-2 9/

FY Y-MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES VALVE NUMBER FUNCTION 00G059 Unit 1 Suct Isol Vlv H Recomb 2

00G060 Unit 1 Discharge Isol Viv H Recombiner 2

00G061 Unit Discharge Xtie for H Recombiner 2

00G062 Unit Xtie on Discharge of H Recumbiner 2

00G063 Unit Suction Xtie for H Recombiner 2

00G064 Unit Suction Xtie for H Recombiners 2

10G065 OB H Analyzer Inlet Isol Viv 2

00G066 OB H Recomb Disch Isol Viv 2

10G057A OA H Recomb Disch. Isol. Valve 2

10G079 H Recomb Disch. Cnmt. Isol. Valve 2

10G080 H Recomb Suct. Cnmt. Isol. Valve 2

10G081 H Recomb Suction Cnmt. Isol. Valve 2

10G082 OA H Recomb Disch Cnmt Isol Viv 2

10G083 0A H R'ecomb Disch Cnmt Isol Viv 2

10G084 OA H Recomb Cnmt Outlet Isol Viv 2

10G085 H Recomb Cnmt Outlet Isoi Vlv 2

1AF006A 1A AF Pp SX Suct Isol V1v 1AF006B IB AF Pp SX Suct Dwst Isol Viv 1AF013A AF Mtr Orv Pmp Disch Hdr Dwst Isol Viv 1AF013B AF Mtr Drv Pmp Dsch Hdr Dwst Isol Viv 1AF013C AF Mtr Dry Pp Disch Hdr Dwst Isol Viv 1AF013D AF Mtr Drv Pp Disch Hdr Dwst Isol Viv 1AF013E AF Ds1 Drv Pm Dsch Hdr Dwst Isol Viv 1AF013F AF.Dsl Drv Pp Dsch Hdr Dwst Isol Viv 1AF013G AF Ds1 Drv Pp Dsch Hdr Dwst Isol Vly,,

IAF013H AF Ds1 Drv Pp Dsch Hdr Dwst Isol Viv, 1AF017A 1A AF Pp SX Suct Upst Isol Vlv 1AF017B 18 AF Pp SX Suct Upst Isol Viv 1CC685 RCP Thermal Barrier Outlet Hd'r Cnmt Isol Viv 1CC9412A CC to RH HX 1A Isol Viv ICC94128 CC to RH HX 18 Isol Viv ICC9413A RCP CC Supply Dwst CNMT Isol ICC9413B RCPs CC Supply Upst CNMT Isol ICC9414 CC Water from RCPs Isol. Valve ICC9415 Unit 1 Serv. Loop Isol V1v ICC9416 CC Wtr from RCPS ! sol. Valve ICC9438 CC Wtr from RC Pumps Thermal Bar Isol. Valve ICC9473A Disch Hdr X-tie Isol Viv

)

ICC9473B Disch Hdr X-tie Isol Viv.

BYRON - UNIT 1 3/4 8-30

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. TABLE 3.8-2 (Continued) p MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES VALVE NUMBER FUNCTION ICS001A 1A CS Pp Suct from RWST 1C5001B 1B CS Pp Suction from RWST 1C5007A CS Pp 1A Disch Line Dwst Isol Viv 1C50078 CS Pp 1B Disch Line Downstream Isol Viv ICS009A 1A Pump Suction from 1A Recirc Sump 1C50098 18 CS Cont Recirc Sump B Suct Isol V1v to C5 1C5019A CS Eductor 1A Suction Conn Isol V1v ICS0198 CS Eductor IB Suction Conn Isol V1v ICV 112B MOV VCT Outlet Upstm Isol VCT Viv.

ICV 112C MOV VCT Outlet 3'r"ejsol VCT Viv 1CV112D MOV RWST to' Chg Pp Suct Hdr

' Dwn5Y.rn 1CV112E MOV RWST to Chg Pp Suct Hdr ICV 8100 MOV RCP Seal Leakoff Hdr Isol ICV 8104 MOV Emerg Boration Vlv ICV 8105 MOV Chrg Pps Disch Hdr Isol V1v ICV 8106 MOV Chrg Pps Disch Hdr Isol V1v ICV 8109 MOV PD Chrg. Pp Miniflow Recirc. V1v ICV 8110 MOV A & B Chg. pp Recirc Downstream Isol ICV 8111 MOV A & B Chg Pp Recirc Upstream Isol 1CV8112 RC Pump Seal Water Return Isol. Valve.

1CV8355A MOV RCP 1A Seal Inj Inlet to containment Isol ICV 83558 MOV RCP 18 Seal Inj Inlet Isol 1CV8355C MOV RCP IC Seal Inj Isol ICV 83550 MOV RCP ID Seal Inj Isol ICV 8804A MOV RHR Sys X-Tie Viv to Chrgng Pump Suction Hdr A.B.

IRC8001A RC Loop 1A Hot Leg Stop Valve

~

1RC80018 RC Loop IB Hot Leg Stop Valve IRC8001C RC Loop IC Hot Leg Stop Valve IRC8001D RC Loop ID Hot Leg Stop Valve IRC8002A RC Loop 1A Cold Leg Stop Valve 1RC80028

. RC. Loop 18 Cold Leg Sto'p Valve IRC8002C RC Loop IC Cold Leg Stop Valve 1RC8002D RC Loop 1D Cold Leg Stop Valve IRC8003A RC Loop 1A Bypass Leg Stop Valve IRC8003B RC Loop IB Bypass Leg Stop Valve IRC8003C RC Loop IC Bypass Leg Stop Valve 1RC80030 RC Loop 10 Bypass Leg Stop Valve l

1RH610 RH PP 1RH01PB Recire, Line Isol.

i 1RH611 RH PP 1RH01PB Recire, Line Isol.

1RH8701A RC Loop 1A to RHR Pump Isol. Valve BYRON - UNIT 1 3/4 8-31

HIRIU4 B +69 TABLE 3.8-2 (Continued)

MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES VALVE NUMBER FUNCTION 1RH8702A RC Loop IC to RHR Pump Isol. Valve 1RH8701B RC Loop 1A to RHR Pump Isol. Valve IRH8702B RC Loop 1C to RHR Pump Isol. Valve 1RH8716A RH HX 1RH02AA Dwnstrm Isol Viv 1RH8/16B RH HX 1RH02AB Dwnstrm Isol Valve 1RY8000A Prz. Relief Isol. Valve 1A 1RY80008 Frz. R41ief Isol. Valve 18 ISI8801A SI Charging Pump Disch Isol Viv 1518801B SI Charging Pump Disch Isol Viv 1SI8802A SI PP 1A Disch Line Dwst Cont Isol V1v 15I8802B SI PP 18 Disch Line Dwst Isol Viv 15I8804B SI Pump 1B Suct X-tie from RHR HX 1SI8806 SI Pumps Upstream Suction Isol ISI8807A SI 'c Chg FP Suction Crosstie Isol Viv 15I8807B SI to Chg PP Suction Crosstie Isol V1v 1SI8808A Accum. lA Disch. Isol. Valve 15I8808B Accum. 1B Disch. Isol. Valve ISI8808C Accum. IC Disch. Isol. Valve 1SI8808D Accum. 10 Disch. Isol. Valve ISI8809A SI RX HX 1A Dsch Line Dwst Isol Viv 1518809B SI RX HX 1B Dsch Line Dwst Isol Viv ISI8811A SI Cnmt Sump A Outlet Isol Viv 1SI8811B SI Cnmt Sump B Outlet Isol Viv ISI8812A SI Rwst to RH Pp 1A Outlet Isol V1v 15I88128 SI Rwst to RH Pp IB Outlet Isol V1v 1518813 SI Pumps 1A-1B Recirc Line Dwst Isol 15I8814 SI Pump 1A Recirc Line Isol Viv 15I8835 SI Pumps X-tie Disch Isol Viv 1518840 SI RHR HX Disch Line Upstrm Cont Pen Is1 Viv ISI8821A SI PP 1A Disch Line X-tie Isol Viv 15I88218 SI Pump 18 Disch Line X-tie Isol Viv 1SI8920 SI Pump 1B Recirc Line Isol V'1v ISI8923A SI PP 1A Suction Isol Viv 15I8923B SI Pump 1B Suct Isol Valve ISI8924 SI Pump 1A Suction X-tie Dwnstrm Isol Viv 1SX016B RCFC B&D Sx Supply MOV ISX016A RCFC A&C SX Supply MOV ISX027A RCFC A&C Return ISX0278 RCFC B&D SX Return MOV fY BYRON - UNIT 1 3/4 8-32

A

9/alr4 H 4t TABLE 3.8-2 (Continued)

MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES VALVE NUMBER FUNCTION OSX007 CC HX Outlet V1v OSX063A SX to Cont Rm Refrig Cdsr OA.

OSX063B SX to Cont Rm Refrig Cdsr OB OSX146 CC Hx "0" return V1v to Unit 1 MDCT OSX147 CC Hx "0" return Viv to Unit 2 MDCT OSX157A SX M/U Pp OA Supply Fill to MDCT 05X1578 SX M/U Pp OB Supply to MDCT OB MOV OSX158A SX M/U -Pp OA Supply Fill to MDCT MOV OSX158B SX M/U Pp 08 Supply to MDCT OB MOV OSX162A MDCT OA Bypass to basin M0V OSX162B MDCT OB Bypass to basin MOV OSX152C MDCT OA Bypass to basin M0V OSX1620 MDCT OB Bypass to basin MOV OSX163A MDCT OA Riser Isol V1v MOV OSX163B MDCT OA Riser Isol V1v MOV OSX163C MDCT OA Riser Isol V1v MOV OSX163D MDCT'OA Riser Isol V1v MOV OSX163E MDCT 08 Riser Isol V1v MOV OSX163F MDCT OB Riser Isol V1v MOV OSX163G MDCT OB Riser Isol V1v MOV OSX163H MDCT OB Riser Isol Viv MOV ISX001A 1A SX Pp Suct V1v M0.

ISX0018 18 SX Pp Suct Viv MOV ISX004 U-1 SX Supply to U-1 CCW HX MOV ISX005 1B SX Pp Supply to O CCW HX MOV 1SX007 CC HX Outlet Viv ISX010 U-1 Trn A return Viv AB ISX011 Trn A Trn B Unit I return X-tie V1v AB ISXO33 1A SX Pp Disch X-tie MOV ISX034 18 SX Pp Disch X-tie MOV 1SX136 Unit 1 Trn B return V1v AB 1SX150A Sx strn drn to waste treatment b1dg MOV i

1SX150B Sx strn drn to TR b1dg MOV f IWoooG A c.uit.t.ED WYr Coils l A k IC. Su.pply 754) VIV TS*l Yk l

t w o o o (= 6 c w it. t.e D W b Coils I g 4 ID Surpl1 t w o o 2.o A cM ii.t G Wtr Coils

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Ret.orn Tsol Yk IW o o 2.o 6 CML ucG Vltr Cod 5 16(.IO ket.orn 15alYk

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( wu.s.e 9 Wa.ttr %nt 13.l Va.lv BYRON - UNIT 1 3/4 8-33

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Scwo.rd W 9/19hy 3[4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION

'.s 3.10.4 The limita ions of Specification 3.4.1.1 may be sus _nded during the performance of star 'p and PHYSICS TESTS provided:

a.

The THERMAL WER does not exceed the P-7 In rlock Setpoint, and b.

The Reactor Tri Setpoints on the OPERABL Intermediate and Power Range channels ar set less than or equa to 25% of RATED THERMAL POWER.

APPLICABILITY:

During operatio below the P-7 nterlock Setpoint.

ACTION:

With the THERMAL POWER greater than t e " 7 Interlcck Setcoint, immediately open the Reactor trip creakers.

SURVEILLANCE RECUIREMENTS

/

\\

4.10.4.1 The THERMAL POWER shall be determined t be less than P-7 Interlock Setpoint at least once per hour during startup an PHYSICS TESTS.

4.10.4.2 Each Interme ate and Power Range channel. and P-7 Interlock shall be subjected to an AN' OG CHANNEL OPERATIONAL TEST wi hin 12 nours prior to initiating startup a d PHYSICS TESTS.

o k-c %

syG s/+. io.+

BYRON - UNIT 1 3/4 10-4

9/s/W SPECXAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS 1

\\#

LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of the fof3owing requirements may be suspended:

a.

Specification 3.4.1.1 - During the performance of startup and PHYSICS TESTS in MODE 1 or 2 provided:

l 1)

The THERMAL POWER does not exceed the P-7 Interlock Setpoint.

and 2)

The Reactor Trip 5etpoints on the OPERABLE Intermediate and Power Range channels are set less tnan or equal to 25% of RATED THERMAL POWER.

b.

Specification 3.4.1.2 - During the performance of hot rod drop time measurements in MODE 3 provided at least three reactor ccolant ico:s as listed in Specification 3.4.1.2 are CPERAELE.

APPLICABILITY:

Ouring operation below the P-7 Interlock Setpoint or performance of not rod drop time measurements.

ACTION:

a.

With the THERMAL POWER greater than tne P-7 Interlock Setpoint during the performance of startup and PHYSICS TESTS. immediately open the Reactor trip breakers.

D.

With less than the above recuired reacter coc! ant 1 cops OPERABLE during performance of hot rod dr00 time measure. tents, immediately place two reactor coolant loops in operation.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Inttriock 5etpoint at least once per hour during startup and PHYSICS TESTS.

4.10.4.2 Each Intermediate and Power Range enannel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating startup and PHYSICS TESTS.

4.10.4.3 At least the above required reactor coolant loops shall be determined OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to initiation of the hot rod drop time measure-ments and at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during the hot rod drop time measurements by verifying correct breaker alignments and indicated power availability.

-s CALLAWAY - UNIT 1 3/4 10-4

C.

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~

OESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4.

Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 4000 grams uranium.

The initial core loading shall have a maximum enrichment of weight percent U-235.

Reload fuel shall be similar in physical design the initial core loading and shall have a maximum enrichment of M' weight g

percent U-235.

9,o h

CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control red assemblies.

The full-length control rod assemblies shall contain a nominal 142 inches of absorber material.

All control rods shall be hafnium, clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 Ttis Reactor Coolant System is designed and shall be maintained:

a.

In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of 2485 psig, and c.

For a temperature of 650 F, except for the pressurizer which is 680 F.

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,257 cubic feet at a nominal T of 588.4 F.

gg 5.5 METEOROLOGICA1. TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

BYRON - UNIT 1 5-4

b DESIGN FEATURES C1/lq)8,'/-

5.6 FUEL STORAGE CRITICALITY 5.6.1Mhe spent fuel storage racks are designed and shall be maintained with:

A k,ff equivalent to less than or equal to 0.95 when flooded with a.

unborated water, which includes a conservative allowance of 3.31%

ak/k for uncertainties as described in Section 9.1 of the FSAR; and b.

A nominal 14 inch center-to-center distance between fuel assemblies placed in the storage racks.

I

5. 6.1. / The k,ff for new fuel for the first core loading, stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent,io_ advertent draining of the pool below elevation 423 feet 2 inches.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than -iW/ fuel assemblies.

lobo

5. 7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

t e

BYRON - UNIT l 5-5 l

@ln h

D[. 8,9 I

d ' C... M.1 8U s ACMINISTRATIVE CONTROLS UNIT STAFF (Continued) f.

Acministrat#ve procedures shall be develoced and imclemented to limit tne working hours of unit staff wne perform safety-related functions; e.g., licensed Senior Operators, licensed Operators, health physics personnel, equipment operators, and key maintenance personnel.

The amount of overtime worked by Unit staff members performing safety related functions sna11 be limited in accordance with the NRC Pelicy Statement on working hours (Generic Letter No. 82-12).

ONs t rE.

Nuce EAR SAFETY Group (ONEG 5"EE' CZINECIM GEUF GZO}~

6.2.3

!"0EPE'OE"7 FUNCTION ONSG 6.2.3.1 The MEs shall function to examine plant operating characteristics, NRC issuances, industry advisories, REPORTABLE EVENTS and other sources of plant design and operating experience information, including plants of similar design, which may indicate areas for improving plant safety.

The ME6 oNSC3 shall make detailed recommendations for revised procedures, equipment modifica-tions, maintenance activities, operations activities or other means of improving plant safety to the Manager of Nuclear Safety, and the Superintendent, Byron Station.

COMPOSITION ON54 6.2.3.2 The !$f6 shall be composed of at least four, dedicated, full-time engineers located on site.

RESPONSIBILITIES t

onsq 6.2.3.3 The ME9 shall be responsible for maintaining surveillance of plant activities to provide independent verification" that these activities are performed correctly and that human errors are reduced as much as practical.

l l

RECORDS l

ON5G 6.2.3.4 Records of activities performed by the HE6 shall be prepared, main-and the Superintendent, Byron}eed.ar yrptpl to the Manager of Nuclear Safe tained,andforwarded@cchCg l

Station.

l 6.2.4 SHIFT TECHNICAL ADVISOR The Station Control Room Engineer (SCRE) may serve as the Shift Technical Advisor (STA) during abnormal operating or accident conditions.

During these conditions the SCRE or other on duty STA shall provide technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering and plant analysis with regard to the safe operation of the unit.

l

Not responsible for sign-off function.

BYRON - UNIT 1 6-2 l

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9 /lyg</

$4AQ f}

ADMINISTRATIVE CCNTROLS

6. 5 REVIEW INVESTIGATION AND AUDIT (Continued) 0FFSITE qong e,,,p gue;,,,,3,pf fy 6.5.1 The Superviso the Offsite Review and Investigative Function shall be appointed by the m c.tiec Vic; ^ esideftt responsible for nuclear activ-ities.

The audit function shall be the responsibility of the Manager of Quality Assurance and shall be independent of operations.

a.

Offsite Review and Investigative Function d

The Supervisor of the Offsite Review and Investigative Function shall:

(1) provide directions for the review and investigative function and appoint a senior participant to provide appropriate direction, (2) select each participant for this function, (3) select a complement of more than one participant who collectively cessess background a'nd qualifications in the subject matter under review to provide comprehensive interdisciplinary review coverage under this function, (4) independently review and approve the findings and recommendations' developed by personnel performing the review' and investigative function, (5) approve and report in a timely manner all findings of non-compliance with NRC requirements to the Station Superintendent, Division Vice President and General Manager -

Nuclear. Stations, Manager of Quality Assurance, and the Vice President -

Nuclear Operations.

During periods when the Supervisor of Offsite Review and Investigative Function is unavailable, he shall designate this responsibility to an establish. J alternate, who satisfies the formal trainina and experience for the Supervisor of the Offsite WE9 eview and Investigat# Function.

The responsibilities of the per-sonnel performing this functi.on are stated below.

The Offsite Review and Investigative Function shall review:

1)

The safety evaluations for:

(1) changes to procedures, equip-ment, or systems as described in the safety analysis report, and (2) tests or experiments completed uncer the provision of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question.

Proposed changes to the Quality Assurance Program description shall be reviewed and approved by the Manager of Quality Assurance; l'

2)

Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in 10 CFR 50.59;

!~

3)

Proposed tests or experiments which involve an unreviewed l-safety question as defined in 10 CFR 50.59; 4)

Proposed changes in Technical Specifications or this Operating License; SYRON - UNIT 1 6-7

thelev w23m

.m x m

T b3 fb.mi.

GalAl b ADMINISTRATIVE CCNTROLS OFFSITE (Continued) 5)

Noncompliance with Codes, regulations, arcers, Tecnnical Saeci-fications, license recuirements, or cf internal procecures, or instructions having nuclear safety significance; 6)

Significant operating abnormalities or deviation from normal and expected performance of plant equipment that affect nuclear safety as referred to it by the Onsite Review and Investigative Function; 7)

All REPORTABLE EVENTS; 8)

All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures, systems, or components; 9)

Review and report findings and recommendations regarding all changes to the Generating Stations Emergency Plan prior to implementation of such change; and

10) Review and report findings and recommendations regarding all items referred by the Technical Staff Supervisor, Station Superintendent, Division Vice President and General Manager -

Nuclear Stations, and Manager of Quality Assurance.

b.

Audit Function The audit function shall be the responsibility of the Manager of Quality Assurance independent of the Production Department.

Such responsibility is delegated to the Director of Quality Assurance for(0peratin and the 5b f? ^...i tz:- t; _:= _. ;;x # Quality Assurance Naintenance)S= ? ' O - r -c^'- 4 General 5 vpuvist Either shall approve the audit agenda and checklists, the findings and the report of each audit.

Audits shall be performed in accord-ance with the Company Quality Assurance Program and Procedures.

Audits shall be performed to assure that safety-related functions are covered within the period designated below:

1)

The conformance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months; i

2)

The adherence to procedure, training, and qualification of the l

station staff at least once per 12 months; i

l 3)

The results of actions taken to correct deficiencies occurring l

in facility equipment, structures, systems, or methods of l

operation that affect nuclear safety at least once per 6 months; 4)

The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months; BYRON - UNIT 1 6-8

9)IQgy mnn

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)

[ l" '; '

1 h

_s s4.

4 ADMINISTRATIVE CONTROLS OF: SITE (Continued) 5)

The Facility Emergency Plan and imolementing ::rocecures at least once per 12 months; 6)

The Facility Security Plan and implementing procecures at least once per 12 months; 7)

Onsite and offsite reviews; 8)

The Facility. Fire Protection programmatic controls including the implementing procedures at least once per 24 months by qualified QA personnel; 9)

The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside indepencent fire protection consultant.

An outside independent fire pro-tection consultant shall be used at least every third year;

10) The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months;

11) The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months;

12) The PROCESS CONTROL PROGRAM and implementing procedures for solification of radioactive wastes at least once per 24 months; and

13) The performance of activities required by the Company Quality Assurance Program for effluent and environmental monitoring at least once per 12 months.

Report all findings of noncompliance with NRC requirements and M ea D er, recommendations and results of each audit to the Station Superin-tendent,10i ::ter of Nuclear Safety, the Division Vice President and General Manager - Nuclear Stations, Manager of Quality Ass.. ance, the Vice Chairman, and the Vice President - Nuclear Operations.

c*

Authority tvia no Se r-op re P*r+ * %

0 **

.nc h*8 The Manager of Quality Assurance r;;;rt; '

n. e C :

th; b... ;...:f N 077;,;...%..;- ;d[p;ti;;ti c:

a n L r.. b ec4er, Nuclear Safety

".,ctima Either the Manager of Quality Assurance or the Super;i;;r f the Offsite 1:fie 2d fY1omber ch

%d e e, Sofc q I-weesti; t Ma c actMn has the authority to order unit shutdown u

or request any other action which he deems necessary to avoid unsafe plant ecnditions.

l

, BYRON - UNIT 1, 6-9

9/n/ry

. C ':',:i :~ 7,E CONTE LS OFFSITE (Continued) d.

Records 1)

Reviews, audits, and recommendations shall be documeated and distributed as covered in Specification 6.5.la. and 6.5.lb.;

anc 2)

Copies of documentation, reports, and correspondence shall be kept on file at the station.

e.

Procedures Written administ ative procedures shall be prepared and maintained for the offsite reviews and investigative functions described in Specification 6.5.la. and for the audit functions described in Specification 6.5.lb.

Those procecures shall cover the following:

1)

Content and method of submission of presentations to the Supervisor of the Offsite Review and Investigative Function, 2)

Use of committees and consultants, 3)

Review and approval,

~4)

Detailec listing of items to be revie ed, 5)

Method of:

(1) appointing personnel, (2) cerrcrming reviews, investigations, (3) reporting findings and recommendations of reviews and investigations, (4) approving reports, and (5) distributing reports, and 6)

Determining satisfactory completion of action required ba(d on approved findings and recommendations reported by personnel performing the review and investigative function.

f.

Personnel 1)

The persons, including consultants, performing the review and investigative function, in addition to the Suoervisor of the Offsite Review and Investigative Function shall have expertise in one or more of the following disciplines as approcriate for the'sub' ject or subjects being reviewed and investigated:

a) nuclear power plant technology, b) reactor operations, c) utility operations, d) power plant design, e) reactor engineering, f) radiological safety, g) reactor safety analysis, BYRON

_ UNIT 1 6-10

=

9/M ry kkA []&Y

'M 5 2 a m a

v:N:5':

.: : NTRC'.5 1

.0F: SITE (Continued) h)

Instrumentation and Control Engineering graduate or equivalent witn at least 5 years of experience in instrumentation and control oesign and/or operation.

i)

Metallurgy Engineering graduate or equivalent with at least 5 years of experience in the metallurgical field.

3)

The Supervisor of the Offsite Revies and Investigative Function shall have experience and training which satisfy ANSI N18.1-1971 requirements for plant managers.

ONSITE 6.5.2 The Onsite Review and Investigative Function shall be supervised by the Station Superintendent.

a.

Onsite Review and Investigative Function ~

l The Station Superintendent shall:

(1) provide directions for the i

Review and Investigative Function and appoint the Technical Staff Supervisor, or other comparably qualified individual as the senior participant to provide appropriate directions; (2) approve partici-pants for this function; (3) assure that at least two participants who collectively possess background and qualifications in the sub-ject matter under review are selected to provide comprehensive interdisciplinary review coverage under this function; (4) indepen-dently review and approve the findings and recommendations developed by personnel performing the Review and Investigative Function; (5) report all findings of noncompliance with NRC requirements, and provide recommendations to the Division Vice President and General Manager - Nuclear Stations and the Supervisor of the Offsite Review and Investigative Function; and (6) submit to the Offsite i

Review and Investigative Function for concurrence in a timely manner, those items described in Specification 6.5.la which have been approved by the Onsite Review and Investigative Function.

[

b.

Responsibility The responsibilities of the personnel performing this function are:

1)

Review of:

(1) procedures required by Specification 6.8.1 and changes thereto, (2) all programs required by Specification 6.8.4 and changes thereto, and (3) any other proposed procedures or l

cha.nges thereto as determined by the ++etre Superintendent to l.

affect nuclear safety; Stati0h I

2)

Review of all proposed test and experiments that affect nuclear safety 13Lo Uh 10 C FR 60 59)'

i BURON - UNIT 1 6 12

9//9/sy EM MWT d3w ACMINISTRATIVE CONTROLS ONSITE (Continued) 3)

Review of all proposed changes to the Technical Specifications; 4)

Review of all proposed changes or modifications to l a,nt systems or equipment that affect nuclear safety (2o W

foCFRSO 59; i

5)

Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to tne Division Vice President and General Manager - Nuclear Stations and to the Supervisor of the Offsite L A. and Investigative Function; Review 6)

Review of all REPORTABLE EVENTS; 7)

Performance of special reviews and investigations and recorts thereon as requested by the Supervisor of the Offsite Review and Investigative Function; 8)

Review of the Station Security Plan and implementing procedures and submittal of recommended changes to the Division Vice President and General Manager - Nuclear Stations; 9)

Review of the Emergency Plan and station implementing procedures and shall submit recommended changes to the Division Vice President - Nuclear Stations; A a,d General marmSer-

10) Review of Unit operations to detect potential hazards to nuclear safety;

11) Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Division Vice President and General Manager - Nuclear Stations and the Supervisor of the Offsite L G Review and Investi-gative Function; and

12) Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE CALCULATION MANUAL, and the Radwaste Treatment Systems.

c.

Authority The Technical Staff Supervisor is responsible to the Station Superintendent and shall make recommendations in a timely manner in all areas of review, investigation, and quality control phases of plant maintenance, operation, and administrative proceduras relating to facility operations and shall have the authority to request the action necessary to ensure compliance with rules, regulations, and procedures when in his opinion such action is necessary.

The Stat. ion Superintendent shall follow such recommendations or select a course SYRON - UNIT 1 6-13

MR MFT

~

ADMINISTRATIVE CONTROLS ClfWk HIGH RADIATION AREA (Continued) source or-from any surface which the radiation penetrates shall be barricaded and' conspicuously posted as a high radiation area and~ entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).

Individuals qualified in radiation protection procedures (e.g., Rad / Chem Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protection procedures for entry

'into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device which continuously indicates the radiation dose rate in the area; or b.

A radiation monitoring device which continuously integrates the radiatinn dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or An individual qualified in radiation protection procedures with a c.

radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the Radiation Work Permit.

~

6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be provided with locked doors to. prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Foreman on duty and/or health physics supervision.

Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for i

individuals in that area.

In lieu of the stay time specification of the RWP, l

direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

For individual nigh radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large' areas, such as l

l PWR containment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricade conspicuously posted, and a flashing light shall be activated as a warning de ice.

(JWL N"W l

6-24 BYRON - UNIT 1

TABLE 3.2-1 Eg DNB PARAMETERS z

e g

PARAMETER LIMITS 4

avg

- 5^2*f

a' c) W $91.2.

Indicated Reactor Coolant System T g

22^E;!;*{"'a^5';}*J.1/9psiy#

Indicated Pressurizer Pressure e d R

lh b

9 h.

4

=N

" Limit not applicable during either a 7tiERMAL POWER ramp in excess of 5% of RATED TilERMAL POWER per minute or a THERMAL h '

E POWER step in excess of 10% of RATED THERMAL POWER.

c

['y c

L' &, m a yl1 M M

TABLE 3.3-1 E

g REACIOR TRIP SYSTEM INSTRUMENTATION MINIMUM 2

TOTAL NO.

CHANNELS CHANNELS APPLICABLE y

' FUNCTIONAL UNIT OF CHANNELS TO TdIP OPERABLE MODES ACil0N 1.

Manual Reactor Trip 2

1 2

1, 2 1

2 1

2 3*,

4*, 5^

10 2.

Power Range, Neutron Flux a.

High Setpoint 4

2 3

1, 2 2#

b.

Low Setpoint 4

2 3

1###, 2 2#

3.

Power Range, Neutron Flux 4

2 3

1, 2 2#

High Positive Rate 4.

Power Range, Neutron Flux, 4

2 3

1, 2 2#

High Negative Rate w."

5.

Intermediate Range, Neutron Flux 2

1 2

1###, 2 3

6.

Source Range, Neutron Flux 2## g k 4

a.

Startup 2

1 2

l b.

Shutdown 2

1 2

3,4,5 5*

l

)

7.

Overtemperature AT 4

2 3

1, 2 6#

7aY

)

~

'2 J

l 8.

Overpower AT 4

2 3

1, 2 6#

y j

9.

Pressurizer Pressure-Low (Above P-7) 4 2

3 1

6#

g=y b

'b ep <;b.n

.3 s %1

?e. 4 oC l

4

h TABLE 3.3-1 (Continued)

TABLE NOTATIONS M -3 Q

With the Reactor Trip System breakers in the closed position and the

~

Control Rod Drive System capable of rod withdrawal.

The provisions of Specification 3.0.4 are not applicable.

1. 8elow the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

M#8elow the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

.g g $mEdh M E f3 MM k

A ACTION STATEMENTS M) #

M*

ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for survelliance testing of other channels per Specification 4.3.1.1; and c.

Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Nautron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.2.

ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

~

BelowtheP-6(IntermediateRangeNeutronFiuxInterlock) a.

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint; and b.

Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

BYRON - UNIT 1 3/4 3-5

FlWo'WERMT

~ABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)'

Y 19 h

. S13 @

ACTION 4 - With the number of CPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving positive reactivity changes, a,

ACTION 5 - With the number of OPERABLE channels pne less than the Minimum Channels OPERABLE requirement restor % the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or within the next hour open the reactor trip breakers, suspend all operat, ions invq)ving positive 6=.TM *AA, reactivity changes, and verify valves ICV 1118, ICV 8428, ICV-8439,

1Cv-8441 and IC,-8435 are closed and secured in position.

F b.

V ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; and b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing of other channels per Specification 4.3.1.1.

ACTION'7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1, provided the other channel i.s OPERABLE.

ACTION 10 - With the number of OPERABLE channels one less t'han the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor trip breakers within the next hour.

ACTION 11

With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

BYRON - UNIT 1 3/4 3-6 w

9/n/ry E o wcr A A v eac,6 al4 3 -ca D-ha Less-than %e.

m.nnmum Chaon e U GPEVA 6 LG reg u. a.r,r,.t Veri fy _ cthe Recch e thy brenh-co are open, susgemiol/

op erch c>u myc.lving pof, hvf. recich've h; c h up.s, ven'fg c ornpl.o ne-e.

6..% ilw._ SH UTD>n' thAAGirV reg v remer ra of Sp ec. h Cch e 5.1.1. l o r 3.l.1. 7-43 cppl ced.alc_ %t-Hus 1 hc g and ver-ify v4lveJ 1C V U1 B.., a dv - EV 2 0_f__3 cv -19 3 )

.. 1C V -8 4 ') I cnet 1 CV - 8's 4 5 ore closed ancL.Se curecI sn f CS' h 0'h WOm i

bou /J.

. mme w -

- W e

e e enum.

O e

em a4e emeh4 e

a

_e e

um 6

m b

Meo e oM 6e eMWm e m

a e

-6 O

h*

g56 e ame ge

ee I

e e-.

y,1A O.

TABLE 3.3-3 (Continued)

Eg ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION z

e c

MlHIMUM 5

TOTAL NO.

CHANNELS CHANNELS APPLICA8tE

[

FUNCTIONAL UNIT Of CHANNELS TO TRIP OPERABtE M00ES ACIION 6.

Auxiliary Feedwater a.

Manual Initiation 2

1 2

1,2,3 22 b.

Automatic Actuation Logic 2

1 2

1,2,3 21 and Actuation Relays c.

Sta. Gen. Water Level-Low-Low

1) Start Motor-Driven Pump 4/sta. gen.

2/sta. gen.

3/sta. gen.

1, 2, 3 19" in any opera-in each w1 ting sta gen.

operating w

~

stm. gen.

s-*

2) Start Diesel-Driven Pump 4/sta. gen.

2/sta. gen.

3/sta. gen.

1, 2, 3 19" in any in each operating operating sim. gen.

stm. gen, m3p cs=

d.

Undervoltage - RCP 4-1/ bus 2

3 1, 2 19' NI9 Bus-Start Motor-P Driven Pump and Diesel-Driven Pump g

e.

Safety injection -

Id Start Motor-Driven Pump See Item 1. above for all Safety injection initiating functions gp 7.)

and Diesel-Driven Pump requirements.

p nsyg f.

Division 11 ESF Bus Undervoltage-g'k Start Motor-Driven Pump (Start as part 2

2 2

1, 2, [

W, L%

of DG sequencing) 25

84 '

TABLE 3.3-3 (Continued)

E ENGINEERED SAFETY FEATURES' ACTUATION SYSTEM INSTRUMENTATION e

c MINIMUM h

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION g

6.

Auxiliary Feedwater (Continued) g.

Auxiliary Feed-water Pump Suction Pressure-Low (Transfer to Essential Service Water) 2 2

2 1,2,3 15*

7.

Automatic Opening of Containment Sump Suction

,g Isolation Valves Y

a.

Automatic Actuation Logic 2

1 2

1,2,3,4 14 and Actuation Relays b.

RWST Level - Low-Low 4

2 3

1,2,3,4 16 Coincident With Safety injection See item 1. above for Safety injection initiating functions and requirements.

T 8.

Loss of Power i

Le A

2.6 2

a.

ESF Bus Undervoltage 2/ Bus 2/ Bus 4/ Bus 1, 2, 3, 4 W

g Ia b.

Grid Degraded Voltage 2/8us 2/ Bus t/ Bus 1, 2, 3, 4 J9" k

A Y

Q i

@lf jf5$3a o$'w

"% H 3,

4

9[19/F'l I M ++ po.y_ 3/4 3 22 02 Aet en as g

oPGtt 6%E M

_ 0_., h AAs M N

k

/

bTne rve J/n Pauseg OPEg.WTioM pm &

& kk.

A n cA *4.' A :

~ % '+ Q - p as.

~ea e A A,J

b. % % u n.

,j S t 9 *. a b 4%

t-

%4AM & %:pLA +.3.2. i.

- ~ -,

-w.

,,,-~-,,,,,--.--,,--,,,~-..,-,,-w

---+.-,,.-w-w--,-,,-

O TABLE 3.3-4.

.h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

TOTAL SENSOR TRIP ALLOWAHLE

[

FUNCTIONAL UNIT ALLOWANCE (TA)

Z ERROR (SEJ SETPOINT VALUE zO 1.

Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Containment Cooling Fans, Control Room Isolation, Phase "A" Isolation, Turbine Trip, Auxiliary Feedwater, Containment Vent Isolation and Essential Service Water) a.

Manual Initiation N.A.

N.A.

'y b.

Automatic Actuation N.A.

N.A.

3 Logic and Actuation Relays c.

Containment Pressure-S7 34 High-1 h

0.71

1. 5 5 v'8.-7 psig

$ S.8 psig l

d.

Pressurizer Pressure-Y memes tow (Above P-11) 16.1 14.41

1. 5 3 1829 psig

? 1823 psig E]O e.

Steam Line Pressure-Low (Above P-11) 21.2 14.81

1. 5 3 640 psig" 1 617 psig*

2.

Containment Spray D

1F

ts a.

Manual Initiation N. A.

N.A.

y g<y b.

Automatic Actuation logic and Actuation a DN Relays N.A.

N.A.

a c.

Containment Pressure-liigh-3

8. 0 0.71 1.5

< 20.0 psig i 21.0 psig

REACTOR COOLANT SYSTEM HOT STANDBY

' ~

~

LIMITING CONDITION FOR OPERATION twc 3.4.1.2 At least thees of the reactor coolant loops listed below shall be OPERABLE and at least two of these reactor coolant loops shall be in operationg when +he Rew Try Syswm preakers are cics* d and of lead coolmW Ic'P shall be ;w opernhen when%c 6 4 der Trip Spgm bre4kg one,cea s e a.

Reactor Coolant Loop A and its associated steam generator and 4rc opm reactor coolant pump, b.

Reactor Coolant Loop B and its associated steam generator and reactor coolant pump, c.

Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and d.

Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.

APPLICABILITY:

H0DE 3.**

- ACTION:

a.

With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

m b.

With only one reactor coolant loop in operationer:nm-ni=

[breakEj 'm +bs clod-+ *"ithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .' i 1; 7 * ' " " " ' ' '

""Om=

open reactor trip breakers

-r Qhe.

(pamy-w c.

With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loopSto operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side narrow range water level to be greater than or equal to 41% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

%e. repired 4.4.1.2.3

= ' ::

t=: reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"All Reactor Coolant pumps may be cieenergized for up to I hour provided:

(1) no operations are permitted that would cause dilution of the Reactor Coolant System' boron concentration, and (2) ' core outlet temperature is maintained at least 10'F below saturation temperature.

4 - M h b e = E ' D M M 3.1 o.af-

Gba rs

&d 9/1s/99 MN N/A'M/fL u,,,, _S4 mu( c REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a.

The Containment Atmosphere Particulate --" c-- =: Radioactivity Monitoring System, b.

The Containment Floor Orain and Reactor Cavity Flow Monitoring System, and c.

The --

'"i----t 24- ;s=

_ c:- _:_ :-== ^

m--

^

_4 4a:

' a.- _ _ _ _^'_^ m

.,n.

__ l *,.

^- i k a---

-A w

-- J -~='2 AN e[da'seous RobNchk($ k5 rig 5 5*-

1 APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a. or c.

in opera We.

on. With - 'y ' m of the above required Leakage Detection Systems ofter **:;E, operation may continue for up to 30 days provided grab samples o' the containment atmosphere are obtained and analyzed for gaseous and particulate radioacitivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the "g"

following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

T%wt g SURVEILLANCE REOUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

a.

Containment Atmosphere Gaseous and Particulate Monitoring System-performance of CHANNEL CHECK, CNANNEL CALIBRATION, and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, b.

Containment Floor Orain and Reactor Cavity Flow Monitoring System-performance of CHANNEL CALIBRATION at least once per 18 months, and c.

Verify the oil separator portion of the containment floor drain collection sump has been filled to the level of the overflow to the containment floor drain unidentified leakage collection weir box once per 18 months, following refueling, and prior to initial startup.

r--+-<----t_- + e e = : =-1

:- t-

-:- n c i-e n ; 7.c ;;;'.. wikt

- f '

'-t ^_:: ;- :tr;; ;--f - :- - :' FEEL-CALIDh wN t :::'.

- -- h =^n ds.

8YRON - UNIT 1 3/4 4-20

G s t a.T' 8

n PP.6G 3l'+ Y~2*

9ll&

b.

W;% b. of the. above reguired Leakage. DehNm Spkms in opereby, be. in of.leed Hor sTAussy, te+hin -the. nex+- t. he>ves and in COLD sH urtoWN u;&iw the follevig 30 boor.s.

of the abe reguired Leak %<. Dekc.Hw 5 5 km3 3

c.

Lli+b

o. and c.a inoperable.:

0 Reshwe. en'*r o.or c. of +he above reguired L+akap %A,w 3 skms tu Ol'Er2A61 shk.s tJ%)n 72. hova and 3

O Ob%in and analyze. o 3rato sayle. o' N confainmeo a+mosyhort. for ga.seous and parkcul<k. recticachW) ab 14akt onc<. pev-24 hours and

3) Perform a Reachr-CoolanF Syskm uais-invenfoy belance. aF leasF onc< per~

8 h our.s Ofberwae., be in a+ leas F Hor STMDE1 (4% h nevf g, A,,r3 and in CoLo suurvouw uk the fe dewing 30 hou<s.

e.

ie e

e-

=

-+p e.

p

-w-i-m,-

e e

r e

g

.,p ye.-

.m--en h

e v

4em 8N M -

me a

w m.w + e e p.

ewe a m4

- -+

e W

e 4

e e

e a

e4 em ebe, m ene q,M

-emen.GM-ein w

a g

e g

e og

-h*h-m

-e q

s eep

-p--.-eh--9-6

W 6 w e

w see 4

e d

46 pM,peamu++

-m-*gy

--g u

W W

w M-t e

e e

O*-

4-4 W

M -8DP'e & &v6--

@e--emme.--he*W enemu 4.

e+

6 *.

4e

- M

-*-ene4---

4

-W 24 M

e4

+@M'*D-*=*

4 eahawew-e m.W e.N@

- Wm hee-m.y w*W.eur 4

4h*w r-

,,,i.mp,

,e 6

ee#

4e same -

a ma MM

+0

-w*

e4 - met =+

5@

u umW*

O -

-w6 h

e 4

4 9

ge 6

M e*W Dm-

-aWw-*eee ee e m eei +e$ =

h "W4 e

4 W w'e b-e e

+M 4-e e

e e

9

=y S

4

.g e

_W.e___.

_.e e

e y _ __ _

e 6

h@

Seele-6 p

.e,agie p

3 6

l f

y, n-n CONTAINMENT SYSTEMS b

INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between

-0.1 and M psig.

+l. 0 APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN with.in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4

+

-__-.___c..-._..-

9f/fhY 2A sc 2 o enA EN UR ILATION SYSTEM LIP.ITING CONDITION FOR OPERATION

3. 6.1. 7 Each containment purge supply and exhaust isolation valves shall be OPERABLE and:

Each 48-inch containment shutdown purge supply and exhaust isolation a.

valve shall be closed and power removed, and The B-inch containment purge supply and exhaust isolation valve (s) b.

may be open,'-

=p ^- 100T '-

-5 c# 3 -

' - -: - ; + 5+ provided no more than lines open at one timegfor Puf9e3 und /or-h+ing o No are regsral f or safeh reMed purposes such 45 ;

APPLICABILITY:

MODES 1, 2, 3, and 4.

g S e c E n ar *,,A,,

ACTION:

N Witr a 48-inch containment purge Fu;;1y anc/or exhaust isolation a.

valve open anc/cr powerec, close anc remove power to isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following req 30m 01her -Hwn th45e. 8fAW io Sp<lhjc<h b.

With the 8-inch ce tainrrent pu ge supply'and/or exhaust isolation valve (s) open for ::

1^^~

L g -

g ::, close the open 8-inch valve (s) or isolate the penetration (s) witnin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , othe* wise be in at least HDT STANDSY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the follo ing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , With a containment purge supply and/or exhaust isolation valve (s) c.

having a measurec leakage rate in excess of the limits of Specifi-cations 4. 6.1. 7. 3 and/or 4. 6.1. 7. 4, restore the inoperable valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , otherwise be in at least HOT STANDEY within the next 6 hcurs, and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

e e

_. - _ _.. _.. _. _ _ _ _. _ _ -. _ _ _ _ _ _ =.

- n. 7 9.-

' T MS trLT" A ~ ~ Ts. 9A66 ~ 3/4 4,- H g

i

l) (naintaining containmeW purt. Wi%ih +h4. limb of Spc c o n cdhon. 3 A.1

't r

j j.

2) rstdoany A contwnmenF ahmosp'nuc o,rbene. rodsoacha t

O c c ep tz b)<.

le ve l fw pprsonntI

[

.c,r Josesay maknol fo on I.

So f eh;.

l-4.-

i i

e 1

k.

6 2-.

h

I i

t L

i l

4 l-4-

4.

l-ko t

h^!;

i 3

]-

i

[

t-1

(

I I:

?

i II i ^

i

+

l i

O I

f l

h CONTAINMENT SYSTEMS A.1m

.> _ L.' N Atbe oa - --

- v,f y,u w as

/9f SURVEILLANCE REQUIREMENTS 4.6.1.7.1 Each a8-inch containment purge supply and exhaust isolation valve (s) shall be verified closed and power removed at least once per 31 days.

4.6.1.7.2 fThe cu ative t' that al

-inch c ainment p e suoply g or/

exn pasola n valve ave been ren duri a calenda ear snal M e I

l K ermine

__ least ce per 7 y s.)

4.6.1.7.3 At least once per 6 months on a STAGGERED TEST BASIS, the inboard ar.d outboard valves with resilient material seals in each closed 48-inch containment purge supply and exhaust penetration shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.05 L, when pressurizec to at least P,, 43.6 psig.

4.6.1.7.4 At least once per 3 months, each 8-inch containment purge supply and

exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.01 L, when pressurized to at least P,, 43.6 psig.

Documenkhcw shdj be reviwed every i e menes de cmHrm in a cc ord am-e.

m purging cmJ vcmW3 wa.s pc-feemed v,4 spswwhm 3.6.i.9 BYRON - UNIT 1 3/4 6-12

sinin F' R DRAFT REFUELING OPERATIONS l

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS g gg g LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

The personnel hatch should have a minimum of one door closed at any a.

one time and the equipment hatch shall be in place and held by a minimum of four bolts, or 4he. e$viPrnent hatch rernoved pursucan t- %

s ave,l tame. Require m ent-H. 9. 4. 2..

b.

A minimum of one door in the personnel emergency exit hatch is closed, and Each penetration providing direct access from the containment c.

atmosphere to the outside atmosonere shall be either:

1)

Closed by an isclation valve, olind flange, or manual salve, or 2)

Capable of being closed by an OPERABLE automatic containment purge isolation valve.

APPLICABILITY:

During CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied. immediately suspend all operations involving CORE ALTERATIONS cr mcvement of irradiated fuel in the containment building.

SURVEILLANCE REQUIREMENTS 4.9.4.1 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic containment purge isolation valve within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:

Verifying the penetrations are in their closed / isolated condition, a.

or b.

Testing the containment purge isolation valves per the applicable portions of Specification 4.6.3.2.

4.9.4.7 S *

D 3#'~ U BYRON - UNIT 1 3/4 9-4

r N >u.r thkV

4. 9. 4. 2 Verify that-the Fuel Handliny Bv,1 cling Euhaush Filler Plenums maintan. -the Fuel Buildiny of a neyante. pressure of-grealer 4hcm or eyl lo V4 inclh coakr gage. rela h'k. k 4ht.

oubid< atmosphore.

wi% Rhe quiernent hatch remored, C. Prior tb COeE ALTEthTioNs or nwemen \\- of nrraclickd fuel and Af lea.sk once. yw d,akl fueI'7 dayf durin}

CC#E ALTTRADONJ or b.

rhovernen& cd srra Verificchvh cf ne Ahve. pressure will be_ pe,4cemed Wih Spkm5 in -Me. nc(mal E F u FLiM 4 IMoDE.

I

i

by a %.

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 1)

Verifying that the exhaust filter plenum satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1.0%, when using the test procedure guidance in Regulatory Positions C.S.a. C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the flow rate is 21,000 cfm i 10%;

2)

Verifying, within 31 days afer removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6 a of Regulatory Guide 1.52, Revi-sion 2, March 1978, and by snowing a methyl iodine penetration of less tnan 4.3%;

3)

Verifying a flow rate of 21,000 cfm i 10% through the exnaust filter plenus during operation when tested in accordance with ANSI N510-1975.

Af ter every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, c.

within 31 days af ter removal, that a laboratory analysis of a re::re-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, by showing a methyl fodice penetration of less than 4.3%;

d.

At least once per 18 months by:

1)

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the exhaust filter plenum at a flow rate of 21,000 cfm i 10%.

2)

Verifying that on a Safety Injection test signal and a High Radiation test signal, the system automatically starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorber banks; and 3)

Verifying that the exhaust filter plenue maintains the fuel building at a negative pressure of greater than or equal to 1/4 inch Water Gauge relative to the outside atmosphere during operations involog thwe. men + of 6el unNn -ihe Sterage ft.>ol j or-Crane. operab web loads ow the Skra3c. pl, BYRON - UN!T 1 3/4 9-14

POWER DISTRXBUTION LIMITS b'

9/Hk/

f w.w Ae 43-t

r. m BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE NOT CHANNEL FACTOR (Continued)

The 12-hour periodic surveillance of indicated RCS flow is suf ficient to detect flow degradation which could lead to operation outside the acceptable limit.

I 4.2.4 QUADRANT POWER TILT RATIO 3/

The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the. power capability analysis.

Radial power distribution measurements are made during STARTUP testing and per'odically during power operation.

The limit of 1.02, at which corrective actica is required, provides DNB and linear heat generation rate protection with x y plane pc-er tilts.

A.imit of 1.02 was selected to provide an allo-ance for the unce ta'nty associated with the indicated pc-er tilt.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time allowance for operation with a tilt condition greater

/

than 1.02 but less than 1.09 is provided to allow identification and corree-tion of a dropped or misaligned control rod.

In the event such ACTION does not correct the tilt, the margin for uncertainty on F is reinstated by reducing g

the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symnetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.

The two sets of four symmetric thittples is a unique set of eight detector locations.

These locations are C-8,t E-5, E-11, H-3, H-13, L-5, L-11, N-8.

3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to *h5 png-ruided paramsters wlte be en uverass of %yzed transient. The.*c<ltul4Mi Wlue4 maintain a design DNBR throughout each anal s inbuka telm he the. QPEM6LL t.benntII.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

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BYRON - UNIT 1 8 3/4 2-5 1

1

N FlML D2M 3/4.4 REACTOR COOLANT SYSTEM BASES 1

3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the applicable design bases DNBR during all normal operations and anticipated transients.

In MODES 1 and 2 with one reactor cool-ant loop not in operation this specification requires that the plant be in at least HO' SIANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing decay heat m--

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In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a hea*.

removing component, require that at least two RHR loops be OPERABLE.

The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a reactor coolant pump with one or more RCS cold legs less than or equal to 350*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50.

The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.

The requirement co maintain the boron concentration of an isolated loop greater than or equal to the boron concentration of the operating loops ensures that no reactivity addition to the core could occur during startup of an isolated loop.

Verification of the boron concentration in an idle loop prior to opening the stop valves provides a reassurance of the adequacy of the boron concentration in the isolated loop.

Startup of an idle loop will inject cool water from the loop into the core.

The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature is within 20*F of the operating loops.

BYRON - UNIT 1 B 3/4 4-1

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3/4.5 ' EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators.

This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, baron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.

A contained borated water level between and ensures a volume of greater than or equal to 6995 gallons but les han o equal to 7217 gallons.

$l 63 The accumulator power operated isolation valves are considered to be

" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met.

In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures If a closed isolation valve cannot be immediataly opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in c mode where this capability is not required.

The requirement to verify accumulator isolation valves shut with power remo d from the valve operator when the pressurizer is solid ensures the y

acct, ators will not inject water and cause a pressure transient when the Reactor Coolant System is on solid plant pressure control.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.

Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward.

In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period.

With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

BYRON - UNIT 1 8 3/4 5-1

9l/ffjl 3/4.6 CONTAINMENT SYSTEMS i

BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY O

Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.

This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accideni.-

analyses at the peak accicent pressure, P,.

As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, or 0.75 L, as applicable, during performance of the periodic g

test to account for possibic degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50, 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure'and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.

Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 0.1 psig, and (2) the containment peak pressure does not exceed the design pressure of 50 psig during steam line break conditions.

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,g The maximum peak pressure expected to be obtained from a cold leg double-ended break event is 43.6 psig. The limit ofMpsig for initials positive containment pressure will limit the ' total pressure to 49-4Fpsig4

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3/4.9 REFUELING OPERATION'S w.

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BASES 9

3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain suberitical during CORE ALTERATIONS, and (2) a uniform baron concentration is maintained for reactivity control in the water v 1ume having direct access to the reactor vessel.

The Ifmitation on Keff of no greater than 0.95 is sufficient to prevent reactor criticality during refueling operations and includes a 1%.ik/k conservative allowance for uncertainties.

includes a conservative uncertainty allowance of 50 ppm.Similarly, the boro These limitations are consistent with the initial conditions assumed for the boron dilution incid:nt in the safety analyses.

The locking closed of the required valves dilution of the filled portions of the RCS.during refueling operations precludes RCS cf unborated water by closing flow paths from sources of unborated water.This a 3/4.9.2 INSTRUMENTATION Th2 OPERABILITY of the Source Range Neutron Flux Monitors ensures that condition of the core. redundant monitoring capability is available to detect changes in the rea 3/4.9.3 OECAY TIME Tha minimum requirement for reactor suberiticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products This d: cay time is consistent with the assumptions used in the safety analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS Tha requirements on containment building penetration closure and OPERASILITY Cnsura that a release of radioactive material within containment will be restricted from leakage to the environment.

The OPERASILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel olement rupture based upon the lack of containmen pre surization potential whilo in the REFUELING MODE.

Q The Byron Station is designed s h that tf a containment opens into the fuel building through the personnel hatch [, In the event of a fuel drop acc the containment, any gaseous radioactivity escaping from the containment buildin sill be filtered through the Fuel Handling aufiding Exhaust Ventilation System.

I 1/4.9.5 COWJNICATIONS Gaticn personnel can be promptly informed of significant changes in

.acility status or core reactivity conditions during CORE ALTERATIONS.

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/RON - UNIT 1 8 3/4 9-1 m

ML20098F526

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