Text

Forwards marked-up Pages of Sections of Proof & Review Tech Specs.Addl Changes Will Be Submitted by 840409
	ML20088A456
Person / Time
Site:	Byron
Issue date:	04/02/1984
From:	Tramm T; COMMONWEALTH EDISON CO.
To:	Harold Denton; Office of Nuclear Reactor Regulation
References
	8376N, NUDOCS 8404120062
	Download: ML20088A456 (200)
	v • d • e

'N Commonweilth Edison O ) One First National Plaza, Chicago. Illinois (O ,

Address Reply to: Post Office Box 767 Chicago tilinois 60690 April 2, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Byron Generating Station Units 1 and 2 Technical Specifications NRC Docket Nos. 50-454 and 50-455-References (a): December 16, 1983 memorandum from Cecil 0.

Thomas.

(b): March 26, 1984 letter from T. R. Tramm to H. R. Denton.

Dear Mr. Denton:

This is to provide additional comments and suggestions regarding the proof and review version of the Byron 1 Technical Specifications that was distributed in reference (s). NRC review of specific changes proposed here is necessary before the Technical Specifications can be finalized.

Attachments A through Z to this letter contain marked-up pages of various sections of the Technical Specifications. A summary explanation of the changes is provided for each attachment. Justifications are provided where appropriate.

A number of similar changes were submitted in reference (b). We understand that the NRC will review each of these proposed changes and inform Commonwealth Edison of their acceptability. A few additional changes will be submitted by April 9, 1984, t

Please direct any questions you may have regarding this matter to this office.

One signed original and fifteen copies of this letter and the attachments are provided for NRC review.

Very truly yours, 8404120062 840402 i i PDR ADOCK 05000454 PDR

, A T. R. Tramm Nuclear Licensing Administrator 1m I

ly cc: Byron Resident Inspector , gh B376N L

ATTACHMENT A (Definition 1.0)

Circled items noted in this attachment have been previously submitted.

I') Definition 1.30 (Page 1-5) Site Boundary The phrase ",as defined by exclusion area," has been added before "shall". Also, reference to Figure 5.1-1; page 5-2 has been added.

"As defined by exclusion area" has been added to define what site boundary they are referring to, and (figure 5.1-1; page 5-2) is the drawing that shows the exclusion area. ,

0 6

(0433M)

PR00F & ilE! COPY DEFINITIONS '

(

PURGE - PURGING 1.24 PURGE or PURGING shall be any controlled process of discharging air or gas

-from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.25 QUADRANT P5WER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever 4 g, eater. u th one excore detector inoperable, the remaining three detectors

!' =_ r computing the average.

RATED THERMAL POWER 1.26 RATED THERMAL POWER shall be a total core heat transfer rate to the reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTA8LE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. >

SHUTDOWN MARGIN 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor bcri or would be subcritical from its present condition assuming al -- O k. st.

d cluster assemblies (shutdown and control) are fully inserteu exceps ror the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY

/ 'gg dgQud h edtsbr, are8, 1.30 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor othemise controlled by the licensee. (ke {ip 5.L 3? M* S' -

SLAVE RELAY TEST 1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

I 1 BYRON - UNIT 1 1-5 I

ATTACHMENT B (Section 2.0)

Circled items noted in this attachment have been previously submitted.

1) Table 2.2-1 (pg. 2-5) Reactor Trip System Instrumentation Trip Setpoints Number 2, a and b; Values for Allowable Value were changed to reflect current PLS values 111.1% and 27.1% respectively.

2) Table 2.2-1 (pg 2-5) Reactor Trip System Instrumentation Trip Setpoint Number 5; Values for Allowable Values has been changed to reflect current PLS value of 30.9%.

3) Table 2.2-1 (pg. 2-5) Reactor Trip System Instrumentation Trip Setpoints Number 6 and 7; Values for Z "10.0 and 5.15" and Sensor Error "0" were changed to reflect current PLS values. Also, note 5 was added under Sensor error column.

4) Table 2.2-1 (pg. 2-5) Reactor Trip System Instrumentation Trip Setpoints Number 9; Values for Total Allowance "S.0", Z "2.21", and Allowable Value "1871" were changed to reflect current PLS values.

5) Table 2.2-1 (pg. 2-6) Reactor Trip System Instrumentation Trip Setpoints Number 15; Value for Z "0", has been changed to concur with Data in Westinghouse Statistical Study for Byron Station.

6) Table 2.2-1 (pg. 2-6) Reactor Trip System Instrumentation Trip Setpoints Number 16; Values for Trip Setpoint "540" and Allowable Value "N/A" were changed and 16a heading was changed to Emergency Trip Header Pressure to reflect what the trip comes off. Also, item 16b heading was changed for clarity.

7) Table 2.2-1 (pg. 2-6) Reactor Trip System Instrumentation Trip Setpoints Loop Design flow - 95,700 gpm has been changed to 94,400 to reflect value from FSAR Table 5.1-1.

8) Table 2.2-1 (pg. 2-7) Reactor Trip System Instrumentation Trip Setpoints Number 19b "or" was inserted to reflect that either can be used as an input for RTS interlocks.

Number 19d "> 7.8%" and 19e "<12.2" changes to Allowable Values have been made to reflect corrections made by Westinghouse in reference to Westinghouse Statistical Setpoint Study for Byron.

(0433M)

ATTACHMENT B (Continued)

(Section 2.0)

9) Table 2.2-1 (pg 2-11) Table Notations

" Note 5: The sensor error for temperature is 1.2 and for pressure is 1.0" has been added for clarity.

(0433M)

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TABLE 2.2-1 y REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E SENSOR Q TOTAL ERROR w FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.

2. Power Range, Neutron Flux 1

a. High Setpoint 7.5 4.56 0 1109% of RTP" 1111. 5 of RTP*

i

b. Low Setpoint 8.3 4.56 0 125% of RTP* 127.5 of RTP*

3. Power Range, Neutron Flux, 1.6 0.5 0 15% of RTP* with <6.3% of RTP* with High Positive Rate a time constant i time constant g2 seconds 12 seconds n 4. Power Range, Neutron Flux, 1.6 0.5 0 <5% of RTP* with <6.3% of RTP* with u, High Negative Rate a time constant i time constant 12 seconds 12 seconds

5. Intermediate Range, 17.0 8.4 0 -<25% of RTP* 48t% of RTP*

Neutron Flux 30.1 %

0

6. Source Range, Neutron Flux 17.0 10 0 .kf9- 110s cps 11.4 x 105 cps

7. Overtemperature AT 8.7 # ee Note 1 S See Note 2

8. Overpower AT 4.3 1.3 S EE HoTE S

1. 2 See Note 3 See Note 4

}

C:3 1971 gleM-psig R

9. Pressurizer Pressure-Low g ir-M= 1. 5 11885 psig

+

10. Pressurizer Pressure-High 3.1 0.71 1.5 12385 psig 12396 psig f

11. Pressurizer Water Level-High 5.0 2.18 1. 5 <92% of instrument <93.8% of instrument R ipan ipan G c3 C:3 "RTP = RATED THERMAL POWER Q

TABLE 2.2-1 (Continued) y REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS e

E SENSOR M TOTAL ERROR s FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWARLE VALUE

12. Reactor Coolant Flow-Low 2.5 1.77 0.6 >90% of loop ->89.2% of loop design flow

ign ' ow*

Ti.1To

13. Steam Generator Water 27.1 7 J. lo' 1.5 >40.8% of narrow >365 of rrow Level Low-Low I gg range instrument an ~ strument span 49.10

14. Undervoltage - Reactor 3.3 0 0 >4099 v s- >4768 volts -

Coolant Pumps -

M 57 0 f.7. 6

15. Underfrequency - Reactor 14.4 .34re. 0 >4HVf_B >&hfr y Coolant Pumps g cn

16. Turbine Trip fre

.$~ 4 0 W

a. Let-rger,e.;;uTdy 0;; PressureHender N.A. N.A. N.A. g psig psig

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b. Turbine Stop Valve N.A. N.A. N.A. ->1% open ->1% open Closure 4 .We. I-12 c::3

17. Safety Injection Input N.A. N.A. N.A. N.A. N.A. c;;3 from ESF

R*

18. Reactor Coolant Pump N.A. N.A. N.A. N.A. N.A. p Breaker Position Trip 2

~~

q tl, '/0 0

^ Loop design flow =_6 gpm 23

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I. .i i TABLE 2.2-1 (Continued) -

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REACTOR TRIP SYSTEM InSTRWENTATION TRIP SETPOINTS i z e

i e SENSOR

TOTAL ERROR U FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWA8LE VALUE

19. Reactor Trip System Interlocks _

v

a. Intermediate Range N.A. N. A. N.A. 11 x 10 18 amp 16 x.10 18 amp Neutron Flux, P-6 b

l s t.

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b. Low Power Reactor Trips L Riock, P-7 );

. c. . :

i 1) P-10 input N.A. N.A. N.A. 110% of RTP* 112.2% of, RTP"-

'} g

et r.

2) P-13 input N.A. N.A. N. A.- <10% RTP* Turbine <12.2% RTP" Turbine ;..

? Tapulse Pressure Tapulse pressure rt N Equivalent , Equivalent ~ ,t

. .. e

'l* c. Power Range Neutron N.A. 'N.A. ~ N.A. ~<30% otATP* -<S0.2% of RTP" E

. Flux, P-8 - ~

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! d. Power Range Neutron N.A. N.A. H.A 110% of RTP" - 2h8%ofRTP" t Flux, P-10 'c [

e. Turbine Impulse Chasher N.A. N.A. N.A. <105 RTP* Turbine RTP* Turbine - 3

) Pressure, P-13 Tapulse Pressure Impulse Pressure 5 i j

. Equivalent Equivalent Q [

, 20. Reactor Trip Breakers N.A. N.A. N.A N.A. N.A. M [

(1 21. Automatic Trip and Interlock N.A. N.A. N.A. N.A. N.A. [

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ATTACHMENT C (Section 3/4.1)

Circled items noted in this attachment have been previously submitted.

1) Section 3.1.1.3 (pg 3/41-4) Reactivity Control Systems - moderator Temperature Coefficient Under Action a. deleted the word "above" from sentence to add simplicity. "In lieu of any other report required by Specification 6.7.1, "has been deleted to be consistent with Tech Specs.

2) Section 3.1.2.1/4.1.2.1 (pg 3/4 1-7) Reactivity Control System - Boration Systems In Applicability, Mode 4 has been added because Section 3.1.2.1 is referring to Mode 4.

A note stating "MA maximum of one centrifugal charging pump shall be operable whenever the temperature of one or more of the RCS Cold leg is less than 350*F" because it is applicable to Mode 4.

3) Section 3.1.2.2 (pg 3/4 1-8) Reactivity Control System - Operating The "x" was deleted from 3.1.2.2 because it is applicable to Mode 4.

4) Section 3.1.2.2 (pg 3/41-8) Reactivity Control System - Operating Mode 4 was deleted because not applicable to this section.

5) Section 3.1.2.3 (page 3/41-9) Charging Pump - Shutdown Add Note

"* A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 350*F."

This note is added for clarification as it has been in other LCO's.

6) Section 3.1.2.4 (Page 3/41-10) Charging Pumps - Operating.

Deletion of

and the footnote is requested because they are applicable to Mode 4.

7) Section 3.1.2.5 (Page 3/41-11) Borated Water Source - Shutdown.

The suggested change to 2652 from 2650 is necessary in order to be consistent with bases.

Also 6.7% has been changed to 7.0% and 8.7% has been changed to 9.0% to make monitoring from panel easier for operators.

(0433M) -

ATTACHMENT C (Section 3/4.1) Continued

8) Section 3.1.2.6 (pg 3/41-12) Borated Water Sources Item 5(b) added from Section 3.5.4 for consistency discussion.

Also, delete "as required by Specification 3.1.2.2" because it is not applicable below 350*F (Mode 4) and these tanks must be operable in Modes 1 through 4.

9) Section 4.1.2.6 (pg. 3/4 1-13) Reactivity Control System - Operating Added "C" which states "At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST vent path temperature to be greater than or equal to 35'F when the outside air temperature is less than 35'F." This was added to complete Surveillance Requirements.

10) Section 3.1.3.1 (pg.-3/4 1-14/15) Reactivity Control System - Movable Control Assemblies Deleted the words " full length" from the page because all rods at Byron are of the same length.

11) Section 3.1.3.1 (pg S/41-14 and 1 15) Reactivity Control Systems Under Actions, 3a. has been switched with 3d. to add clarity and for its Operating Importance.

12) Section 4.1.3.3 (pg. 3/41-18) Reactivity Control System - Shutdown.

Add sentence stating "The Digital Position Indication System does not indicate the actual position of the Shutdown rods, between 18 steps and 210 steps withdrawn" from pg. B 3/41-3 to Section 4.1.3.3 for System Capability Clarifications.

13) Section 3.1.3.4 (pg. 3/41-19) Reactivity Control System - Rod Drop Time Deleted the word " full length from section 3.1.3.4 to indicate all rods are of one length at Byron.

14) Section 3.1.3.4 (pg. 3/41-19) Reactivity Control System - Rod Drop Time Deleted step (a) from Action because Rod Drop time must be verified before reaching Modes 1 & 2. Therefore (a) will never apply.

(0433M)

L _ _ _ _ _ _ _ _ _ _ _ . _ _ _ __ _ . _ _ _

ATTACHMENT C

-(Section 3/4.1) Continued

15) Section 3.1.3.b (pg. 3/4 1-21) Reactivity Control Systems In Section 3.1.3.6 the "s" was deleted from Figures and "and 3.1-2" was also deleted to get rid of all references to three loop Operations In Action a. The word "immediately" was inserted at the beginning of the

.. sentence and "within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months " was deleted. The phrase "Immediately initiate reduction in" was added to the sentence upon the deletion of

" Reduce" and "within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months " to give clarity to proper operations.

"b" was turned into "c", "c" is now "d" and a new "b" was created stating "Immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing g.t. or equal to 7000 ppm boron or equivalent boration rate, until the req'd control bank position is restored." This was added to be consistent with Surveillance 4.1.1.1.1.b pg 3/41-1.

16) Section 4.1.3.b (pg 3/4 1-21) Reactivity Control System.

Deleted the word " Monitor" and placed " Alarm" in its place.

s e

5 (0433M)-

6

-- --. . l

[ -.. -. REACTIVITY CONTROL SYSTEMS

~C MODERATOR TEMPERATURE COEFFICIENT I

, .. .. LIMITING CONDITION FOR OPERATION 4

3.1.1.3 The moderator temperature coefficient (MTC) shall be: ;

l

a. Less positive than 0 ak/k/*F for the all rods withdrawn, beginning j of cycle life (80L), hot zero THERMAL POWER condition, or '

l l, b. Less negative than -4.1 x 10 4 Ak/k/*F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL PCWER condition.

APPLICABILITY: Specification 3.1.1.3a. - MODES 1 and 2* only#.

Specification 3.1.1.3b. - MODES 1, 2, and 3 only#.

ACTION: -

c

a. . MTC more positive than the limit of Specification 3.1.1.3a.

dx? eration in NODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained

-- sufficient to restore the MTC to less positive than 0 ak/k/*F i(' '

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6;

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that j the MTC has been restored to within its limit for the all rods withdrawn condition; and

3. -In-Hett-of er,y ot.tr .epen-required by 5;;;if4 cation-Gr7th X Special Report is prepared and submitted to the Commission pursuant to Specification 6.7.2 within 10 days, describing the value of the sensured MTC, the interim control rod withdrawal

3. limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b. With the MTC more negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

t I

"With K,ff greater than or equal to 1.

See Special Test Exception 3.10.3.

3 BYRON - UNIT 1 3/4 1-4 1 1

1

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REACTIVITY CONTROL SYSTEMS n

"n, 3/4.L 2 BORATICN SYSTEMS FLOW PATH - SHUTDOWN LIMITTNG CONDITION FOR OPERATION !

l C I l

3.L2.1 As a sinimum, one of the following boron injection flow paths sh'all be OPERA 8LE and capable of of 4 powered from an OPERA 8LE emergency power source:

U a. A flow path fron'the Soric Acid Storage System via a boric acid transfer pump and a charging pump to the Reactor Coolant System if the Boric Acid Storage System in Specification 3.L2.5a. is OPERA 8LE, or

b. The flow path from the refueling water storage tank via a charging 0- pump to the Reactor Coolant System if the refueling water storage

~

. ;. _ . . tank in Specification 3.L2.5b. is OPERABLE. _ . . . _ _

q .- . ..-

APPLICA8ILITY: MODES [5and6.

ACTION:

oO_ '

With none of the above flow paths OPERA 8LE or capable of being powered from an OPERA 8LE emergency power source, suspend all operations involving CDP.E ALTERATIONS or positive reactivity changes.

SURVEILLac! REQUIREMENTS

' ?.

4.1.2.1 At least one of the above required f1cw paths shall be demonstrated OPERA 8LE:

1

^

a. Ati least once per 7 days by verifying that the temperature of the heat traced portion of the flow path is greater than or equal to 68'F when a flow path from the Soric Acid Storage System is used, and

b. At least once per 31 days by verifying that each valve (manual, s:

power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

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8YROM - UNIT 1 3/4 1-7

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I t-REACTIVITY CONTROL SYSTEMS ,

beh FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three baron injection flow paths shall be OPERA 8LE:

a. The flow path from the Boric Acid Storage Systas via a boric acid I transfer pump and a charging puep to the Reactor Coolant System, and

b. Two flow paths from the refueling water storage tank via charging pueps to the Reactor Coolant System.

M APPLICABILITY:

MODES 1, 2, 3, end-b

. ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERA 8LE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERA 8LE 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 and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERA 8LE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated l OPERABLE:

i

a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path from the boric acid tanks is greater than or equal to 65'F when it is a required water source;

b. At least once per 31 days by verifying that each valve (manual, power-operated, or. automatic) in the flow path that is not locked, sealed, or otherwisa secured in position, is in its correct position; i c. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal; and i d. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpa to the Reactor

( Coolant System.

"Only bo I ection'f1 path i iredAo} OPERA egs is less than or8LE8ntver equ&

a of on re of col 3 F. .

)b BYRON - UNIT 1 3/4 1-8 h .

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9 V

f REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN gggg '

q LIMITING CONDITION FOR OPERATION

f 3.1.2.3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERA 8LE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY: MODES 4, 5, and 6.

.7, ACTION:

With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

O SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by . r verifying, on recirculation flow, that a differential pressure across the pump of greater than or equal to 2416 psid is developed when testad pursuant to Specification 4.0.5.

-C 4.1.2.3.2 All charging pumps, excluding the above required OPERA 8LE' pump, shall be demonstrated inoper2ble at least once per 31 days, except when the reactor vessel head is removed, by verifying that the actor circuit breakers are secured in the open position.

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REACTIVITY CONTROL SYSTEMS %% g QQ []

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.( CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION i

3.1.2.4 At least two charging pumps shall be OPERA 8LE.

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APPLICA8ILITY: MODES 1, 2, and 3.

ACTION: ,

With only one charging pump OPERA 8LE, restore at least two charging pumps to q. OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STAND 8Y and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at lust two charging pumps to OPEPABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

( 'I SURVEILLANCE REQUIREMENTS 4.1.2.4.* At least two charging pumps shall be demonstrated OPERABLE by 7 verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 2416 psid is developed when tested pursuant to Specification 4.0.5.

O 8YRON - UNIT.1 3/4 1-10 v.

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i l

REACTIVITY CONTROL SYSTEMS h k h h {.]

/N BORATED WATER SOURCE - SHUT 00WN U

LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

A Boric Acid Storage System with:

a.

$0Yo066*3.

, 1) A minimum contained barated water level of M(M50 gallons),

2) A minimum boron concentration of 7000 ppm, and

3) A minimum solution temperature of 65*F.

b. The refueling water storage tank (RWST) with: o

1) A minimum contained borated water level of (38,740 gallons),

I

2) A minimum boron concentratioa of 2000 ppm, and

3) A minimum solution temperature of 35'F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the baron concentration of the water,

2) Verifying the contained borated water volume, and

3) Verifying the boric acid storage tank solution tempe rature when it is the source of barated water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of barated water and the outside air temperature is less than 35'F.

m)

BYRON - UNIT 1 3/4 1-11

'~.',"~~~' *:-~..... . _ , . . . , , _ _ _ _ .y ,,,g., .

. . - . _ , - . - . . - - ~. .-

l

, I l

REACTIVITY CONTROL SYSTEMS u I hh BORATED WATER SOURCES - OPERATING l

l LIMITING CONDITION FOR OPERATION l

l 3.1.2.6 As a minimum, the following barated water source (s) shall be OPERABLE. )

' % ;;;ific: ticr. ?.1 9

= r;; -

a. A Boric Acid Storage System with:

j 1) A minimum contained borated water level of 40% (15,780 gallons),

2) A minimum boron concentration of 7000 ppe, and

3) A minimum solution temperature of 65'F.

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water level of 89% (395,000 gallons),

2) A minimum boron concentration of 2000 ppa,

3) A minimum solution temperature of 35'F, and

4) A maximum solution temperature of 100*F.

5) Heat tra.ced yt**rd of the ssee:aA +/. 419 PAO 6 50 h ONU APPLICA8ILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With the Sc:ic Acid Storage System inoperable and being used as one of the above required borated water sources, restore the system to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or 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 borated to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200*F; restore the Boric Acid Storage System l to OPERA 8LE status within the next 7 days or be in COLD SHUTOCWN I

within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l

b. With the RWST inoperable, restore the tank 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 STANOBY 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 following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3

SYRON - UNIT 1 ,

3/4 1-12

e me = -

e. -

_ g&A y ._ .

-. , , , , . - - - . , , - . , , - - , - . . . - - . - - - , , - - , , , _ , , - , , . - - - - . - _ _ ,----,,.c--,- . , . _ , , , - , - _ ~

l-1

, PROOF & RD1&l t";py REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERA 8LE:

a. At least once per 7 days by:

1) Verifying the boron concentration in the water,

2) Verifying the contained borated water volume of the water source, and

3) Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature 'is either less than 35*F or greater than 100*F.

$. per 24 hours by Af le a S r once Ve r ify i ncy .rhe RWST vent path I f ern per a +ure ro be 9reater + /i a n or c c3 a a l to 35 F ujhen t- h e ou r g ,cte le s.r +h a s, 35 *F.

air r ep,p er a tu re i .s O

I i

b BYROM - UNIT 1 3/4 1-13

r 4x

$%L e .

REACTIVITY CONTROL SYSTEMS P200f & ggggy pa i ,

3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION f.[ 3.1.3.1 Al fu!!-h ;t utdown and control rods shall be OPERABLE and

, V .(positioned ithin 2 s (indicated position) of their group step counter l .. f demand position.

[{M ' MICA 8ILITY: MODES la and 2*.

' / ACTION:

MI

a. With one or no full-h;;[ ds inoperable due to being immovable l

$' bks as a result of exc...ive friction or mechanical interference or known to be untrippable, determine that the SHUT 00WN MARGIN require-l ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in Q

HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

i hjgf [- b.

With more group step than on [rh;;.

counte 1 A go.i inoperable or misaligned from the on by more than 2 12 steps (indicated g, , position), be in HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

E c. With on f.11 k;;.Pr)d trippable but inoperable due to causes oth o thou - N M 4y ACTION a. above, or misaligned from its

.%3 . group step counter demand height by more than 2 12 steps (indicated c MC position), POWER OPERATION may continue provided that within 1 hour:

_j

$h 1. The rod is restored to OPERA 8LE status within the above gg ggnap : 2.

alignment requirements, or The rod is declared inoperable and the remainder of the rods in l Sjy#d- the group with the inoperable rod are aligned to within i

~M 2 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figures 3.1-1 and 3.1-2. The O THERMAL POWER level shall be restricted pursuant to Specification f 3.1.3.6 during subsequent operation, or l ~i ? 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER

~ <[ OPERATION may then continue provided that:

A reevaluation of each accident analysis of Table 3.1-1 is 1

d [1Ph.

47 performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents l_,je remain valid for the duration of operation under these v* conditions; Nhb-b b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 MQi e is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; special Test Exceptions 3.10.2 and 3.10.3.

m h

k0 1 h % - UNIT 1 3/4 1-14

_ .. . . _ - - - .. _ - , _ _ . MMC a

j 00F& Rf/IEW C0psl REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable incore detectors and Fq(Z) and F are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and 8 4)* The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.

SURVEILLANCE REOUIREMENTS 4.1.3.1.1 The position of each fuM-lengtbr)d shall be determined to be within the group demand limit by erifying th'a individual rod positions at least one 1er rs except during time intervals when the rod position deviation =f teR is operable, then verify the group positions at least once per 4 hou ALAM 4.1.3.1.2 Each % il l determined OPERA M e.gt -.m nt of h..e d not at least fully inserted 10 steps in any one in the core direction at shall be least once per 31 days.

I O

n N)

BYRON - UNIT 1 3/4 1-15 O

__.____s

I g REACTIVITY CONTROL SYSTEMS ( MDF& RENEW copy '

POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1. 3. 3 One digital red position indicator (excluding bank demand position indication) shall be OPERA 8LE and capable of determining the control rod position within i 12 steps for each shutdown or control rod not fully inserted.

APPLICABILITY: MODES 3*#, 4"# and 5*#. _

ACTION: _.

With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers.

O SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the above required digital rod position indicator (s) shall be determined OPERABLE by verifying that the digital rod position indicator agrees with the demand position indicator within 12 steps when exercised over the full-range of rod travel at least once per 18 months. The Q,p(,a(

$5iUon Indic8hn %Am does nok indicBle the Baha( posMet o9- ihe. ShMwh t'od t beheeq a &ps and so alfes wiAdvawn, l

"With tne Reactor Trip System breakers in the closed position.

See Special Test Exception 3.10.5. I O

8YRON - UNIT 1 3/4 1-18 l

1

b REACTIVITY CONTROL SYSTEMS P200F & RM Cm R00 OROP TIME _

LIMITING CONDITION FOR OPERATION elde 3.1.3.4 The individ fe!'-h:73 shutdown and control rod drop time from the fully withdrawn ns less than or equal to 2.2 seconds from beginning of decay of s onary gripper. coil voltage to dashpot entry with:

a. T,yg greater than or equal to 550*F, and

b. All reactor coolant pumps operating.

APPLICA8ILITY: MODES 1 and 2. ___

ACTION: __

4. 'd ' rod drop Fany full-length rod _datermined to exceed ~

N abo limi tore the'tod drJo Atee'to within Me_=have limit prior to ing to MODE 1 ort.

Q. E. With the rod drop time within limits but determined with three raector coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 66% of

(_ RATED THERMAL POWER.

SURVEILLANCEREQUIREMENTS 4.1.3.4 The rod drop time of Nil-h;,-bds shall be demonstrated through seasurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and

c. At least once per 18 months.

BYRON - UNIT 1 3/4 1-19 m

PR00F & RDilEW COPY REACTIVITY CONTROL SYSTEMS CONTROL R00 INSERTION LIMITS O

LIMITING CONDITION FOR OPERATICN 3.1.3.6 The control banks ^shall be limited in physical insertion as shown in Figure # 3.1-1, nd 0.12. -

APPLICA8ILITY: MODES 1* and 2"#. _.

ACTIOM: __

With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:

Im41abitJ

a. Restore the control banks to within the limits withi& 2 hm m<, or hedaielL niU3e i

tvMeien in C /. Reduce THERMAL POWER within--2-hour + to less than or equal to that fraction of RAT RMAl,-PCWych is allowed by the bank position using the abov figure p o 3,t g ,

d. Be in at least TWOBY withf3A hours.

k. 'Intnedlahly inRiale ad CCSnue heabh al; reab Nb or epaf to .$oapm of a Schdhn C,cedSining d 3 . or epal b '70% pprn beran or et ioxluf borern gale, o W I

& re(d corArol bant pes 31en is reskred .

SURVEILLANCE REOUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion when the Rodlimits at leastponcrper\s12 Insertion Limi "0niten i hours except noperable, during then verifytime theintervals individual rod positions at least onc per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

hm I "See special Test Exceptions 3.10.2 and 3.10.3.

With K,ff greater W an or equal to 1.

O BYRON - UNIT 1 3/4 1-21

ATTACHMENT D (Section 3/4.2) i f

l 1) Section 3.2.3 (pg. 3/4 2-8) Power Distribution Limits N

i- A period was added af ter R and "since - - - - - - F " was deleted SH.

because it is stated in Bases.

! 2) Table 3.2-1 (pg. 3/4 2-15) DNB Parameters Under Limits for Pressurizer Pressure > 2220 psig has been changed to

- > 2205 psig because it is the indication provided to Operator.

Also, all other pressures in the Technical Specifications are in gage.

1' I

f a

4 Y

(0433M)

I l

_____, POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITINGCONDITiONFOROPERATION 1

1 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow (

rate and R shall be maintained within the region of allowable operation shown on Figure 3.2-3 for four loop operation.

1 Where: -

F"

- * - - ~ ~

N

1.49 [1.0 + 0 (1.0 - P)]

THERMAL POWER , and b* -

P = RATED THERMAL POWER

c. Fh=MeasuredvaluesofFhobtainedbyusingthemovableincore detectors to obtain a power distribution map. The measured N shall be used to calculate R,64 ace-Figure-3r24 values of Fg

--includes-penalties-for-undetected-feedwater = turi-fouliv

%f-0.-3-and -for-seasurement- uncerta i nti es-o f-2.-N-for-fl ow--

. -and-45-fee-incere-measurement-of N g .

APPLICABILITY: MODE 1. - . ..

ACTION: .

With the combination of RCS total flow rate and R outside the region of acceptable operation shown on Figure 3.2-3:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore the combination of RCS total. flow rate and R to within the above limits, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Fluy-High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

O BYRON - UNIT 1 3/4 2-8

-. . .m , , , _ . _. y

__g. -y-__-__. -- . .

O O O 1 -

?

5 g TABLE 3.2-1 z

e DNS PARAMETERS E.

) PARAMETER LIMITS l'our Loops in Three Loops in operation Operation Reactor Coolant System T $ 592*F ave my ,. :d Pressurizer Pressure i 222^ PS "

M

+

7

1 5

3 a

c2 '

ri C=2 1 ::rJ rv1

16 C""2

' Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL O POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

" ^^These values left blank pending NRC approval of three-loop operation.

s

I ATTACHMENT E (Section 3/4.3)

Circled items noted in this attachment have been previously submitted.

1) Table 3.3-1 (pg. 3/4 3-3) Reactor Trip System Instrumentation Number 16; Emergency Tripheader EHC has been deleted from heading and added in Step A in place of Low Auto Stop Oil to indicate what turbine trip indications come off. Train has also been added after MS 3, 2, 2 to give clarification. Step b Stop has been changed to Throttle to correct the valve type.

2) Table 3.3-2 (pg. 3/4 3-8) Reactor Trip System Instrumentation Response Times.

Number 16; Step (a) Low Auto Stop Oil has been deleted and Emergency Trip header has taken its place to indicate what the turbine trip comes off.

Step (b) Stop has been deleted and Throttle added to indicate the correct valve type.

3) Table 4.3-1 (pg. 3/4 3-10) Reactor Trip System Instrumentation Surveillance Requirements number 16; Same explanation as above.

4) Table 4.3-1 (pg. 3/4 3-12) Table Notations Sentence number 12 has been rearranged as shown to add clarification and give a more accurate description.

5) Table 3.3-3 (Page 3/4 3-19) Engineered Safety Features Actuation Instrumentation.

"3/stm. gen." has been changed to "4/stm. gen." because there are 4 channels / gen.

6) Table 3.3-4 (Page 3/4 3-22) Engineered Safety Feature Actuation Instrumentation Change the Allowable Value for 1.c to "15.5 psig".

The changes in table 3.3-4, Pg. 3/4 3-22 were made to concur with the values from the PLS (Precautions, Limitations and Setpoints).

7) Table 3.3-4 (pg. 3/4 3-22) Engineered Safety Features Actuation. System Instrumentation Trip Setpoints Number 1(c), (d), (e) has changed values for Allowable value from 3.5, 1839, 160 (psig) to 5.8, 1823, 617 (psig) respectively. 1(d) has changed values for Trip Setpoint from 1850 to 1829 (psig). Number 2(c) has changed value for Allowable value form 22.0 to 21.0 (psig) to concur with Westinghouse Statistical Setpoint Study.

_1 (0433M)

ATTACHMENT E (Continued)

(Section 3/4.3)

8) Table 3.3-4 (pg. 3/4 3-23) Engineered Safety Features Actuation System Instrumentation Trip Setpoints.

Number 3(b); Values for Allowable Value has been changed from 22.0 to 21.0 to concur with Westinghouse Statistical Setpoint Study.

9) Table 3.3-4 (pg. 3/4 3-24) Engineered Safety Features Actuation System Instrumentation Trip Setpoints Number 4(c), (d) has changes for Allowable Value 12.0, 610 has been changed to 9.2 and 617 respectively. Number 4(c) has changed for Trip Setpoint 10.0 psig to 8.2 psig. Number 5(b) has changes for Trip Setpoint and Allowable Value to concur with Westinghouse Statistical setpoint Study.

10) Table 3.3-4 (pg. 3/4 3-25) Engineered Safety Features Actuation System Instrumentation Trip Setpoints Number 6 (c), (d) has changes for Trip Setpoint and Allowable Value to concur with Westinghouse Statistical Setpoint Study.

11) Table 3.3-4 (pg 3/4 3-27) Engineered Safety Features Actuation System Instrumentation Trip Setpoints.

Number 9 (a) has changes for Trip Setpoint 1950 psig has been changed to 1930 psig. to concur with PLS.

12) Table 3.3-9 (pg 3/4 3-50) Remote Shutdown Monitoring Instrumentation Number 4; Under Pressurizer Pressure, Total No. of Channel, the number of Channel was changed from 2 to 1. Because there is only one channel.

13) Table 4.3-6 (pg 3/4 3-51) Remote Shutdown Monitoring Instrumentation Number 2; Next to Channel Check, an asterisk has been added to M to show that it is applicable below P-6.

14 Table 3.3-10 (pg 3/4 3-53) Accident Monitoring Instrumentation.

Number'll has been deleted because it is not needed in reference to letter,

15) Table 4.3-7 (pg 3/4 3-54) Accident Monitoring Instrumentation Surveillance Requirements.

Number 11 has been deleted because it is not necessary in reference to letter.

(0433M)

ATTACHMENT E (Continued)

(Section 3/4.3)

16) Section 4.3.3.9 (pg 3/4 3-60) Radioactive Liquid Effluent Monitoring Instrumentation The following was added after the word Digital and before Channel; ", and Analog". This is added to give page consolidation.

17) Section 3.3.3.9 (pg 3/4 3-60) Delete page which discusses " Digital Channel Operational Check"

18) Table 4.3-9 (pg 3/4 3-63) Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements. The column was added for Analog Channel Operational Test because we are using Analog Channels not Digital.

19) Table 4.3-8 (pg 3/4 3-64) Table Notations.

Number 5 was added to represent the fact that we are working with Analog Instruments.

20) Table 3.3-13 & Table 4.3-13 (Page 3/4 3-66, 67 16 3/4 3-70.71)

" Monitor" has been changed to " sampler" in the following tables:

1. Table 3.3-13; lines Ib, Ic, 4b, and 4c.

2. Table 4.3-9; lines 1b. Ic, Ab, and 4c.

The revision is the result of comments made by the NRC during a meeting concerning process radiation monitor calibrations, noting that several other stations have designated particulate and iodine channels as

" samplers". This designation lessens the calibration requirements placed on particulate and iodine channels, thus making the current Byron calibration procedures acceptable to the NRC.

21) Table 4.3-9 (pg 3/4 3-72) Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements.

Number 5; 1RE-PR090 to 1RE-PR009 to correct type.

(0433M)

(s)I (v )

^

TABLE 3.3-1 (Conticued)

E REACTOR TRIP SYSTEM INSTRUMENTATION g~

I :=

i '

MINIMUM

' TOTAL NO. CHANNELS CHAhNELS APPLICABLE E OF CHANNELS TO TRIP OPERABLE M00ES ACTION q FUNCTIONAL UNIT e

9. Pressurizer Pressure-Low 3 1 6#

4 2 i (Above P-7) 2 3 1, 2 6#

I 10. Pressurizer Pressure-High 4 I

11. Pressurizer Water Level-High 7#

3 2 2 1 (ebove P-7)

U 12. Reactor Coolant flow-Low -

i Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 7#

m a.

any oper- each oper-

) ating loop ating loop l

T

b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop in 1 7#

two oper- each oper-

- below P-8) ating loops ating loop 4/sta. gen. 2/sta. gen. 3/sta. Gen. 1, 2 6# -g

13. Steam Generator Water A in any each Level-Law-Low operating sim. gen.

operating sta. gen.

97 23

14. Undervoltage-Reactor Coolant 6#

4-1/ bus 2 3 1 Pumps (Above P-7) 3

15. Underfrequency-P.eactor Coolant Cr) 4 2 3 1 6#

Pumps (Above P-7) 4-1/ bus 1

C"2 cm

16. Turbine Trip -Eneu ve . ci

-ft1C (Above P-7)

Tel Fh;;d;;'

M r, cmy y w a + 2l Tem 2/1run 7 y a. -is,f. t; Et y OG Pressure 3 horn 1 7

4 1 1

, b. Turbine Step Valve Closure 4

ThmMt

i l

TABLE 3.3-2 (Continued) -

$ REACTOR TRIP SYSTEM INSTRONENTATION RESPONSE TIMES '

z RESPONSE TIME g FUNCTIONAL UNIT 4 Low Reactor Coolant Flow - Low g 12.

i

a. Single Loop (Above P-8) 1 I second

b. Two Loops (Above P-7 and below P-8) i I second

13. Steam Generator Water Level-Low-Low 1 2 seconds

14. Undervoltage-Reactor Coolant P s (h p.7) 1 1.5 seconds Underfrequency-Reactor Coolant P 5 0.6 second 15.

s{ p,3

16. Turbine Trip (h P.- '

y e muge nT ME" N.A.

' e a. LOe Aete Ste; O!! Pressure w b. Turbine Ive Closure N.A.

do rtveHit

17. Safety Inject o nput from ESF N.A.

Reactor Coolant Pump Breaker Position Tri N.A. l

18. (QQ3ev.o 9-7)

N.A. t2

19. Reactor Trip System Interlocks :::-m O

20. Reactor Trip Dreakers N.A. a i

n

21. Automatic Trip and Interlock Lo01c H.A. Da f :;rJ E

n a

"t3 i

I l

I j

c- - -

._ a i

. TABLE 4.3-1 (Continued) .

E REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS .

.- 8 i lj' TRIP ,

ANALOG ACTUATING MODES FOR c

z CHANNEL DEVICE WHICH j U CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE

,' - H FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

13. Steam Generator Water Level- S R H N.A. N.A. 1, 2 I

I"' l" i

14. Underv tagesaeacter Coolant N.A. R N.A. M N.A. 1 Pump (Ab6ve PM j 15. Underfrequenc -Aaarlo N.A. R N.A. M N.A. 1 l Coolant ausp (4.ye. [-

.I 16. Turbine TripfUAl*Ve P 'ld y' Emerpug Tr@ R4

a. tew-Aute-Step-OH Pressure N.A. R M.A. S/U(1, 10) N.A. 1

b. Turbi Step Va ve N.A. R N.A. S/U(1,10) N.A. 1

$ Closur (b W

17. Safety injection Input from N.A. N.A. N.A. R N.A. 1, 2 .

ESF .u zu N.A. N.A. N.A. 1 O

18. Reactor Coolant D'- Sr N.A. R Position Tripf Abe$e f ') ,

M

' po

19. Reactor Trip System Interlocks ' u Intermediate Range r*

a.

Neutron Flux, P-6 N.A. R(4) M N.A. N.A. 2## -.

j

'"U

, b. Low Power Reactor :hll:

Trips Block, P-7 N.A. h(4) M (a) N.A. N.A. 1 l

i c. Power Range Neutron v

-4*

Flux, P-8 N.A. R(4) M (8) N.A. N.A. 1

.i

I g3

(

)

TABLE 4.3-1 (Continued) hh h hh hk -

TABLE NOTATIONS

With the Reactor Trip System breakers closed and the Control Rod Drive System capable of rod withdrawal.

M8elow P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

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

(1) If not performed in previous 7 days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistant with calorimetric power if absoluta difference is greater than 21 The provisions of Speci-fication 4.0.4 are not applicable for entry into MODE 2 or 1.

> (3) Single point comparison of incore to excore axial flux difference above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

> (5) Detector platcau curves shall be obtained, evaluated and compared to manufacturer's data. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provi-O sions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7) Each train chall be tested at least every 62 days on a STAGGERED TEST BASIS.

(8) With power greater than or equal to the interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the inter-lock is in the required state by observing the permissive annunciator window.

(9) Monthly surveillance in MODES 3*, 4*, and 5* shall also include verification

) that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Monthly surveillance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to an increase of twic'e the count rate within a 10-minute period.

(10) Setpoint verification is not applicable.

7 (11) At least once per 18 months and following maintenance or adjustment of the I Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall l include independent verification of the Undervoltage and Shunt trips.

l verify that on a simulated .,

' (12) AtBaron leastOilution once perDoubling 18 months testduring signalshutg;al the .cr CVCS discharge valveswi+1 M 0 6 t: '

) open ed 2close and the-centrifugel-chargitigy.,, ___t4en-*elva: frem tN n!37 wi+1 g '(e13)f open CHANNEL within 30 seconds.

CALIBRATION shall include the RTD bypass loops flow rate.

,a BYRON - UNIT 1 3/4 3-12

p ~

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ll EE MINIMUM El TOTAL NO. CHANNELS CHANNELS APPLICABLE

- FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

9. Engineered Safety Features Actuation System Interlocks

a. Pressurizer Pressure, 3 2 2 1,2,3 20 P-11

b. Reactor Trip, P-4 2 2 2 1,2,3 22

c. Low-Low T,yg, P-12 4 2 3 1,2,3 20

$b d. SteamGeneratorWaterLevel,fA7sta. 2/sta. gen. 2/sta. 1, 2, 3 20 w P-14 (High-liigh) gen. in any gen. In .

J. operating each

. sta, gen. operating stm. gen.

c.

L23 m

De

rs Ei!

n C3

=

v v v 7 v v r- r v v o

o '

. u-d TA8LE 3.3-4 5

E z

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS '

TOTAL SENSOR TRIP ALLOWA8LE E FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (S) SETPOINT VALUE 4 -

1. Safety Injection -

(Reactor Trip, Feedwater Isolation, Start Diesel Generators, Containment Cooling Fans, and Essential

. Service Water) .

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation M.A. N.A. N.A. N.A. N.A.

Logic and Actuation w Relays

+

w c. Containment Pressure- 61 Nigh-1 2. 5 0.71 1.5 1 5.0 psig 5-44 psig

d. Pressurizer Pressure- 1:23 ID3 Low 13.0 10.71 1.5 1 M50 psig 1 16 R psig

K3

~

e. Steam Line Pressure- - ' 617 Low (above P-11) 14.2 10.71 1.5 3 640 psig 3 frle psig* q

2. Containment Spray $2*

3

a. Nanual Initiation N.A. N.Ac N.A. N.A. N.A.

Q

b. Automatic Actuation Logic and Actuation h

Relays N.A. N.A. N.A. N.A. N.A. O

c. Containment Pressure- 21 High-3 5.0 0.71 1.5 1 20.0 psig 1 -22.0 psig

U _V U U v v- v--- v v v y TABLE 3.3-4 (Continued) f ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS SENSOR TRIP ALLOWA8LE TOTAL.

SETPOINT VALUE E FUNCTIONAL UNIT Alt 0WANCE (TA) Z_ ERROR (S)

-s e 3. Containment Isolation i

a. Phase "A" Isolation

1) Manual Initiation N.A. N.A. N.A. N.A. N.A.

2) Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A. N.A. N.A.

3) Safety Injection See Ites 1. above for all Safety Injection Trip Setpoints and Allowable Values.

b. Phase "B" Isolation I N N.A. N.A. N.A. N.A.

?l [ 1) Manual Inttfation N.A.

t N.A. N.A.

i U' 2) Automatic Actuation M.A. N.A. N.A.

Logic and Actuation Relays 21

3) Containment 5-22.0 psig Pressure-High-3 S.0 0.71 1.5 1 20.0 psig

c. Containment Vent Isolation ]

C:3

1) Manual Initiation N.A. N.A. N.A. N.A. N.A. c2 m

2) Automatic Actuation no

' Logic and Actuation m Relays N.A. N.A. N.A. N.A. N.A.

3) Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation Trip Setpoints and Allowable Values.

Q cm C3

U u v v v v v -f v y v os o yo:

TABLE 3.3-4 (Continued)

E g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPolNTS TOTAL SENSOR TRIP ALLOWABLE ,

ALLOWANCE (TA) Z ERROR (S) SETPOINT VALUE E FUNCTIONAL UNIT 4

e 4 Steam Line Isolation

a. Manual InttIation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic and Actuation Relays M.A. N.A. N.A. N.A. N.A.

c. Containment Pressure- 1A 31 High-2 5.0 0.7} 1.5 1 le:6 psig 1-12r8 psig I Gn w d. Ste ne r e- 14.2 10,71 1. 5 -> 640 psig ~> 6te psig*

! 5 Low ( h P.rt) w A e. Steam Line Pressure-

N ve - igh 8.0 0.5 0 < -100 -< -110.0 psi /s**

ist/s 5.

{h e-n)

Turbine Trip and feedwater Isolation _

a. Automatic Actuation _

\

. Logic and Actuation ] Eta Relays N.A. N.A. N.A. N.A. N.A. <g

b. Steam Generator Water St.9 32 S R*

Level-High-High (P-14) 5.0 2.18 1.5 51H!M of 5 43% of :::a2

. narrow range narrow range instrument instrument Q-span span h b

23

7 w w = w '

} s-

, a I .]

i .

l t .

TABLE 3.3-4 (Continued) ,'

l i 5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS l 1

z .

' . }

TOTAL SENSOR TRIP ALLOWABLE E FUNCTIONAL UNIT ALLOWANCE (TA) -

Z ERROR (S) SETPOINT VALUE 4

6. Auxiliary feedwater

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic and Actuation

, Relays N.A. N.A. N.A. N.A. N.A. .

c. Steam Generator Water Level-Low-Low-Start Motor- S'I 33.1 w Driven Pump and 30.0 27.18 1.5 > 41% of > 40E of A Diesel-Driven Pump iiarrow range liarrow range w instruement instrument i 4 span span 4'PO Yoll.3 y'% 17 YON S

d. Undervoltage-RCP Bus- N.A. N.A. N.A. 1=70~ hti i 0% ".C."

Start Motor Driven Pump A w]Gg; h es ug}tage and Diesel-Driven Pump ( M^^ "cits) _(4674-volts-)

e. Safety Injection- m Start Motor- ::*2 j Driven Pump and See Item 1. above for all Safety Injection Trip Setpoints and @

Diesel-Driven Pump Allowable Values. ri Do

acs Ei!

, e

< C3 1

' N l

1 1

~

(-

l d { \ J

(_/ '

l i

TABLE 3.3-4 (Continued) l !

i 5

z ENGINEERED SAFETY FEA10RES ACTUATION SYS1EM INSTRUMENTATION TRIP SETPOINTS ,

SENSOR TRIP ALLOWABLE

' TOTAL Z ERROR (S) SETPOINT VALUE E FUNCTIONAL UNIT ALLOWANCE (TA) 4

8. Loss of Power ESF Bus _Lindervaltage H.A. N.A. N.A. 2870 volts 2870 volts i 143

a. 1.8 1 0.ls time delay hekeM<cd Rek7s;u.h

b. rid-Degraded- 3804 volts t 76 N.A. N.A. N.A. 3804 volts

- VoEage_ 9s time delay 9 1 0.9s time delay el;,\ 5}de PMc

9. Engineered Safety Feature Actuation System Interlocks IS30 N EW

$ a. Pressurizer Pressure, N.A. H.A. N.A. 14958 psig 5-2050 psig w P-11 A N.A. N.A. N.A. N.A.

" b. Reactor Trip, P-4 N.A.

N.A. N.A. H.A. 550*F > 548"F and

c. Low-Low Iavg, P-12 $ SS2*F _

See Iten S.b. above for all Steam Generator Water Level Trip N

d. Steam Generator Water

Setpoints and Allowable Values. c.:3 Level P-14 c.3 (lligh-liigh) ',

1 w

o J

. m C3 T

M

y v v -

v - - - - - y -

O O lO, t

TABLE 3.3-9 '

$ REMOTE SHUTDOWN MONITORING INSTRUMENTATION 8

I : TOTAL NO. MINIMUM READOUT OF CHANNELS c_

$ INSTRUMENT LOCATION CHANNELS OPERABLE

--e '

- 1. Intermediate Range Neutron Flux IPLO6J 2 1

2. Source Range Neutron Flux IPLO6J 2 1

3. Reactor Coolant Temperature -

Wide Range ,

a. Hot Leg IPLO5J 1/ loop 1/ loop

b. Cold Leg IPLO5J 1/ loop 1/ loop

. 4. Pressurizer Pressure IPLO6J 7i 1 4

5. Pressurizer Level IPLO6J 2 1

6. Steam Generator Pressure IPLO4J/IPLO5J 1/sta gen 1/sta gen

7. Steam Generator Level IPLO4J 1/sta gen 1/sta gen

8. RHR Flow Rate LOCAL 2 1

9. RHR Temperature LOCAL 2 1

10. Auxiliary feedwater Flow Rate IPLO4J/IPLD5J 1/sta gen 1/sta gen E

E

~

1

w~

l !

l 1

, TABLE 4.3-6

-c i

E !

REMOTE SHUTDOWN MONITORING INSTRUMENTATICH ;

7 SURVEILLANCE REQUIREMENTS !

E CHANNEL CllANNEL

. 4 INSTRUMENT CHECK CALIBRATION i "

1. Intermediate Range Neutron Flux M N.A.

i

[ 2. Source Range Neutron Flux M N.A.

3. Reactor Coolant Temperature - Wide Range M R

4. Pressurizer Pressure M R

. 5. Pressurizer Level M R

6. Steam Generator Pressure M-e R

j 7. Steam Generator Level M R

'.. Y

. $ 8. RHR Flow Rate M R

9. RHR Temperature M R

10. Auxiliary feedwater Flow Rate M R y

.. 2 C"l3

. c3 M

$ \lhert i

gl,s'foie beks P-L -

SI-3 m

2 ca

!. ~=

C"3

v v - v v v v v v v i

I i TABLE 3.3-10 l 5

$ ACCIDENT MONITORING INSTRUMENTATION l

I l TOTAL MINIMUM E

NO. OF CHANNELS ?

U INSTRUMENT CHANNELS OPERABLE

. .p

1. Containment Pressure 2 1 I 2. 2 1 -

Reactor Coolant Outlet Temperature - TH0T (Wide Range)

3. 2 1 Reactor Coolant Inlet Temperature - TCOLD (Wide Rage)

4. Reactor Coolant Pressure - Wide Range 2 1

5. Pressurizer Water Level . 2 1

6. Steam Line Pressure 2/ steam generator 1/ steam generator

7. Steam Generator Water Level - Narrow Range 1/ steam generator 1/ steam generator

8. Steam Generator Water Level - Wide Range 1/ steam generator 1/ steam generator y

9. Refueling Water Storage Tank Water Level 2 1

[ "

10. Auxiliary Feedater flow Rate 2/ steam generator 1/ steam generator l $

1

@ &xter-Coelant-Systen-Subcoollag-Margin-Monitor 1

12. PORV Position Indicator (0 pen / Closed) 1/ Valve 1/ Valve

13. PORV Block Valve Position Indicator (0 pen / Closed) 1/ Valve 1/ Valve 1/ Valve 1/ Valve

14. Safety Valve Position Indicator (0 pen / Closed) t2 2 1 .. 3

15. Containment Floor Drain Susp Water Level (Narrow Range), (".3

16. Containment Water Level (Wide Range) 2 1 c?>

91 2/ core quadrant

17. In Core Thermocouples 4/ core qJadrant g

18. Containment High Range Area Radiation 2 1 :::: 2

19. Containment Hydrogen Concentration 8 2 I h

20. Neutron Flux (Power Range) 4 2 Q

21. Auxiliary Building Vent Stack - Wide Range Noble Gas 1/ stack 1/ stack n

22. Main Steam Line Radiation 1/sta line 1/sta line h

23. Reactor Vessel Water Level 2 1

m -- -

y -- --- a w a w w w w

[

v!V TABLE 4.3-7 E

g

ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENT CilANNEL CHECK CHANNEL CALIBRATION I

g 1. Containment Pressure M R

2. M R Reactor Coolant Outlet Temperature - TN T (Wide Range)

3. M R Reactor Coolant Inlet Temperature - TCOLD (Wide Range)

4. Raactor Coolant Pressure - Wide Range M R

5. Pressurizer Water Level M R

6. Steam Line Pressure M R

7. Steam Generator Water Level - Narrow Range M R

8. Steam Generator Water Level - Wide Range M R

9. Refueling Water Storage Tank Water Level M R y 10. Auxiliary feedater flow Rate M R

[ -11r R eac to r-Cool an t- Sys t ea -Subcool i ng- Ma rg i n-Mon i tor P

]

<j 12. PORV Position Indicator (0 pen / Closed) 11 R Q

T

13. PORV Block Valve Position Indicator (0 pen / Closed) M R

14. Safety Valve Position Indicator (0 pen / Closed) M R

' :::o a-

15. Containment Floor Drain Sump Water Level (Harrow Range) M R

16. Containment Water Level (Wide Range) M R

17. In Core Thermocouples M R i 18. Containment HIOh Range Area Radiation M R* b

19. Containment Hydrogen Concentration S Q

20. Neutron Flux (Power Range) M R

21. Auxiliary Building Vent Stack - Wide Range Noble Gas M R

22. Main Steam Line Radiation M R

23. Reactor Vessel Water Level M R

CilANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, i for range decades above 10R/h and a one point calibration check of the detector below 10R/h with an

, installed or portable gamma source. -

O p _

INSTRUMENTATION g -

-- RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION '

__ _ LIMITING CONDITION FOR OPERATION

_ 3_

3.3.3.9 The radioactive liquid effluent monitoring instrumentation channels O-shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to "

ensure that the limits of Specification 3.11.1.1 are not exceeded. The Alarm /

Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (00CM).

O APPLICABILITY: At all times. ._

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel Alam/ Trip Setpoint less conservative th' an required by the above B specification, immediately suspend the release of radioactive liquid affluents monitored by the affected channel, or declare the channel inoperable.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERA 8LE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERA 8LE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specifica-tion 6.7.1.7 why this inoperability was not corrected within the time specified.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

r. - _ . - -

I 4.3.3.9 Each radioactive ifquid effluent monitoring instrumentation channel shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and DIGITAL g CHANNEL OPERATIONAL TEST at the fre-quencies shown in Table 4.3-8.

and RMLOG

?

1 C 3/4 3-60 BYRON - UNIT 1

)

l

. . .= . . .

T c _. . . . . r M

'"5 " "'^ * " .

'-rid 0F HEMEW COPY

-. RADI0 ACTIVE LIQUID EFFt.UENT MONITORING INSTRUMENTATICN .

LIMITING CCNDITION FOR CPERATION - - -

. . . ..~..~

n 3.3.3.9 x The radioactive liquid effluent monitoring instrumentation channels

shown inqable 3.3-12 shall be OPERA 8LE with their Alare/ Trip 5etpoints set to ensure that the limits of Specification 3.11.L1 are not exceeded. The Alam/

Trip 5etpof'nts of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL

, (00CM). \

\

APPLICA8ILITY: Atxall times. . . _ .

4 ACTION:

a.

\

With a radioactive liquid effluent monitoring instrumentation channel m Alare/ Trip 5etpoint less conservative than required by the above specification, immediately suspend the release of radioactive ,

liquid affluents monitored by the affected channel, or declare the channel inoperable. \

, b. With less tnan the minimum number of radioactive liquid affluent monitoring instrumentation, channels OPERA 8LE, take the ACTION shown j,

] in Table 3.3-12. Restore the inoperable instrumentation to CPERA8LE status within the time specified in the ACTION, or explain in the next i

Sesiannual Radioactive Effluent Release Report pursuant to Specifica-l tion 6.7.1.7 why this inoperability was not corrected within the time '

specified. j s

\

j r^- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

. j' '

/

/ \

. ,,_, ,,, _._, SURVEILLANCE REGUIREMENTS . . . . _ _ , _

n / \'

4.3.3.9

/

Each radioactive liquid effluent monitoring initrumentation channel shall be demonstrated,0PERA8LE pe ormance of the CHANNEL CHECX, SOURCE CHECX, CHANNEL CALIBRATION and DIGITA CHANNEL OPERATICNAL TEST at the fre-

\

quencies shown in able 4.3-8. N s s

o \

\

,~

N

. o' SYROM - UNIT 1 3/4 3-60 x '

o

o " " " ' '

d b 'O l ' r TABLE 4.3-8 f 8

z RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 1

DIGITALt 0kRlbb CHANNEL cShKilEl.

Q CHANNEL SOURCE CllECK CHANNEL CALIBRATION OPERATIONAL TEST g

INSTRUMENT CilECK r

1. Radioactivity Monitors Providing .

Alara and Automatic Termination of Release Liquid Radwaste Effluent Line (ORE-PR001) D P R(3) Q(1) g,A,

2. Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination -

of Release ,- % y pd ic gjet

$ a l ssential Service Water Line w (1RE-PR002 F 4 rc .a D /H R(3) 4RyNA q(s)

J, b. Euesh\ Seto, w fd pkuRE/c' ' p.1 m oJhtM Of G f4003)) R(3) -Q(4) N.A. q(5 )

" c h. Station Blowdown Ine Rolo) 0 ,/

~'

3. Flow Rate Measurement Device

a. Liquid Radwaste Effluent Line _

(Loop-WX001) D(4) H.A. R Q m I

.a N. A.

b. Station Blowdown Line (Loop-CWO32) D(4) N.A. R Q @

m

, u.a.

oo a

m C3 o

W

'1 l

o TABLE 4.3-8 (Continued)

Pi100F & REYlEW COPY kJ _ _ _ _ ,

<-- m TABLE NOTATIONS DIGl'IE)

(1) b The(lation'i)f this pathway and control room alarm annunciation occu

. iso any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or

b. Circuit failers, er j c. Instrument indicates a downscale failure, or

d. Instrument controls not set in operate mode.

(2) The OIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

b a. Instrument indicates measured levels above the Alarm Setpoint, or

b. Circuit failure, or

c. Instrument indicates a downscale failure, or n

b d. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that

. participate in measurement assurance activities with NBS. 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.

l (4) CHAhNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.

l (51 The fhah C,hamel Op.nund led M alz dwah ed cdd mro ah m&

Mmmdalioq ocors 3 anj d the fd*I") cadihns was*,

3. InsWa trace meawd kveb tbwe h Rhrm sapu , .r.

b. C ;reu't h:lua , or l 0.%%.4 ha:caks s hes\e hNre , a d.TnsWvA conkds not ed I n egrak nede .

g l

g)

BYRON - UNIT 1 3/4 3-64 l u P _ _ , . _ _ _ . . . . . _ . , _ .

1 j TA8tE 3.3-13 I =<

!, 5 RADIGACTIVE GASE005 EFFLUENT NONITORING INSTRUNENTATION MININMM CHANNELS

[

z INSlutstENT OPERABLE APPLICA81LITY ACTION

-4 5

e. 1. Plant Vent Monitoring System - N flif /

a. Noble Gas Activity Monitor-Providing Alare

39

1) High Range (IAE-PR0280) 1 1

8 39

,g 2) Low Raaea (18E-PR02Sl4r)

S 0

40

.I b. Iodine :^mkr

_.. 6 (IAE-PR028C) 1 1

40

c. Particulate (IRE-

% d. Effluent System flow Rate

36 w Measurine Device (L**P-VAtet) 1 alg 1* 1 a 36

! e. Sampler Flow Rate Measur{ng Device

-(Wi-palea) ClFT- rJtl6U

2. Gas us Wast ~----t

~

ysten j Hydroge/a " (GAT- / / ** * ~"

n Analyzer )

15TE =

c3

/b. Oxygen Analyzer (SAT-GWR04 and OAl-GWe003} / 2

41 m Ro .

o E

s

. . 3 1

4

s

-t TABLE 3.3-13 t -< -

I E

=

RADI0 ACTIVE GASE005 EFFLUENT leNIT0AING INSTALSENTATION

!i-

! MINittel CHAlelELS APPLICA81LITY ACTION E IN5ilh84ENT OPERABLE

. ~,

e 2 f. Plant Vent Monitoring System - l/aif 7.yo i a. Moble Gas Activity Monitor-l Providing Alare i 9

39

1) High Range (JRE-PA0280) 1 1

8 39 lti 2) LowRange(p-PR028K)

'

40

'h b. Iodine k-P9028C) 1

!' u

c. Particulate h- 1

40 D d. Etfluent System Flew Rate

36 I w Measuring Device (Loep-vat 49) 1 1 020

36 Sampler Flow Rate Measuring Device 1

!! $ e.

(JFI-PR162)

L l Jf. Gaseous Waste Management Systes

a. Hydrogen Analyzer (OAT-GNIO90) 1

- 38 g i C2 Oxygen Analyzer (OAT-GW004 and tts l . b. ** M 2 41 DAT-GW8003)

R* .

=z 1

- e, C3 N

. -4ll 1

b.

,s

,~ m TAttE 3.3-13 (Continued)

, 5 l 8 RADI0 ACTIVE GASEGUS EFFLUENT IElIIITORING INSTRIMNTATION z .

MINIIERG CHAleIELS l

INSTRIM NT OPERAttE APPLICARILITY ACTION i

E

'G f c f. Gas Decay Tank System

a. Itohle Gas Activity Monitor - Providing Alarm and Automatic Termination of

Release (ORE-PR002A and 2R) 2 35 l

36

[ b. Flow Rate k nitor 1

c. Sampler Flow Rate Monitor *

(OFI-PR159) 1 36 ,

l m . Containment Purge Systes m a. Noble Gas Activity Monitor - ProvidinR

J, Alarm (IRE-PR001A) 1 37

b. Iodine

i (lRE-Pit 001C) 1 40

~

Saanp/tr

c. Particulate atewMor *

(IRE-PROGIS) l' 44

'-3 C2 d System flow Rate Monitor 1

3d i M

E

e. Sampler Flow Rate Monitor (IFI-PR100) 1

J L C*

u

~<

((. Radioactivity k nitors Providing Alarm and Automatic Closure of Surge Tank Vent Camponent Q Cooling Water Line (ORE-PR009 and IRE-PR009) 2

4 ?h f C3 y i I -

' . { (_ C ,

3 m

I i -

h

} 'E , TASLE 4.3-9 1

8

RADICACilWE GASEGUS EFFLMNT leIITOB1mG ltIST8tefIITATiest SURVEILLAllCE MeulaEMNTS

BIGITAL CHAleIEL M00ES FOR nelICH

) .E CalAfeEL SaueCE OtAteIEL OPERATIONAL SURVEILLAleCE Q IS K QUIRED FLeeCTIONAL MIIIT CMCK CMCK CAllt0AT10Il TEST

1. Plant Vent Moniterlag System - f/tsi+ /

a. Iloble Gas Activity Iteatter -

! Providlag Alare

1) Higin Range (18E-P9029) B M R(3) Q(2) 7 .

2) Low Range (IAE-PA02 W 8 M R(3) Q(2)

$ b. Iodiae *

(IAE-P9029C) 3 M B(3) Q(2) u

.', c. ParticulateSaaf ! (IAE-P90 A in.f*rr M B M R(3) Q(2) i * *

d. Effluent system Flaw Rate 3 N.A. R Q Measuring Device (&eep-VA0lS) l O FE

d

e. Sampler Flow Rate temasuring 8 N.A. R Q Device (IFI-PR162) g i Gaseous Wste Management Systee $

l2. agen yyzer ( T y- / pELETE j('4) **

~

R2 My 5 li. A. M

/ /)/ane

/

r' xs

b. OxygesVAnalyz OAT-Ind8003) '

(84

/ B/ II. A. Q(5) M _

b g

Ji!

. j. -

I 5, TABLE 4.3-9 '

I I E

RADIGACTIVE GASEQUS EFFLUENT MDNITORING INSTauMENIATION SURVEILLANCE REQUIREMENTS

OIGITAL C CHANNEL MODES FOR nellCH z ' SURVEILLANCE CHANNEL 50ueCE CHANNEL OPERATIONAL

< ^ ' Q CHECK CAtl8AAT10N TEST 15 RE0u!REO FUNCTIONAL UNIT CNECK r

l fg Plant Vent Monitorine System - (intf ~/~wo l

i

. a. Noble Gas Activity Monitor -

! Providing slare 2 ^

- 1) High Range (ASE-PR0280) O M R(3) Q(2) l 2 s .

2) Low Range (ASE-PR0200 0 M R(3) Q(2) u Sanrkt 1 *

b. Iodine dese+eer (JRE-P9028C) O M R(3) Q(2) 4 )

.', c. Particulate f[ 8 d ( k - 0 M R(3) Q(2) i . o ^ '

D d. Effluent System flow Rate O N.A. R Q Measuring Device (4eer4Nett) l OFE-V4010 ^

e. Sampler Flow Rate Measuring 0 N.A. R Q

-13 G.wice (# 1-PR162) :=2 C2

2 c-3/ Gaseous Waste Management System

^^ p2

a. Hydrogen Analyzer (0AT-GW8000) 0 N.A. Q(4) M

. 33

' rm

b. Oxygen Analyzer (0AT-GWo04 and ^^

OAl-GW8003) 0 N.A. Q(5) M _

h

' N i l 1

i

~'

t U " '

O d , ,

i i

l 5 T_ARLE 4.3-9

. 8

. RADIDACTIVE GASEGUS 5 FLUENT IGNITORING INSTRtRENTATION SURVEILLANCE REQUIRE 8ENTS DIGITAL

. E CHAlelEL te0ES FOR tRtICH q SURVEILLANCE

' CHAIRGEL SOURCE CHAIRIFL OPERATIONAL FUNCTIONAL UIIIT CIECK CNECK CAL 1RRATICII TEST IS REQUIRED Y/. Gas Decay Tank System

a. Noble Gas' Activity Monitor - P P R(3) Q(1)

Providing Alars and Automatic Termination of Release (ORE-PR002A and 28)

b. Flow Rate Monitor P N.A. R Q

_ (0F1T-GnJood Sampler Flow Rate Monitor 0 N.A.

} c. R Q m (OF1-PR159)

[f Containment Purge Systes

.I a. Noble Gas Activity Stonitor -

Providing Alars (IRE-PR00lf) O P R(3) Q(2) l

b. Iodine P P R(3) N.A.

C (IRE-PR001C) d

c. ParticulateSe* n;"frler P P R(3) N.A.

98 '

a:s (IRE-PR001F) o N.A.

q-i

d. Systen Flow Rate Monitor 0 R Q e.

(IFT- VG 0 93)

Sampler Flow Rate Monitor D N.A. R Q h

Q (IF1-PR100) 2E

)

\ /

y- .

m--- y - - p s ,

5 TABLE 4.3-9

E 2

4 l RADI0 ACTIVE GASEOUS EFFLUENT NONITORING INSTRUNENTATION SURVEILLANCE REQUIRENENTS g DIGITAL q CHANNEL N0 DES FOR WIICH g CHANNEL SOURCE CHANNEL DPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CHECK CALIBRATION TEST IS REQUIRE 0 3

5. Radioactivity Monitors Providing Alarm and Automatic Closure of Surge Tank Vent Component Cooling Water Line (ORE-PR009 and IRE- M ) D M R(3) Q(1)

FROH

+

w a

n

. e i

e

t. .

a:s C3 C::3 m

90

o

i Ei!

n C'3 i *U W

i

ATTACHMENT F (Section 3/4.4)

Circled items noted in this attachment have been previously submitted.

1) Section 3.4.3 (pg 3/4 4-9) Reactor Coolant System. '

In Action 3.4.3.a, delete the word " backup" from " backup pressurizer heaters" and " backup heaters."

The work backup is deleted from ACTION to be consistent with what they are referring to in 3.4.3.

2) Section 4.4.4.3 (pg 3/4 4-10) Reactor Coolant System - Relief Valves Surveillance 4.4.4.3 has been deleted. There is no emergency power supply for the PORV's and the block valves. The backup force to close the valves in an emergency is a supply of nitrogen filled accumulators.

3) Section 4.4.b.1 (pg 3/4 4-18) Reactor Coolant System.

Surveillance Requirement 4.4.6.1.a replaces " ANALOG" with " DIGITAL".

Byron Station will use Digital Channels in the Operational test because this (Digital Channels) happens to be the available instrument.

'4) Section 4.4.6.1 (pg 3/4 4-18) Reactor Coolant System Leakage.

A new surveillance has been added to verify the containment floor drain collection sump is filled. The commitment for the surveillance is addressed in an NRC inspection report response.

5) Section 3.4.6.2 (pg 3/4 4-19) Reactor Coolant System - Operatio ial Leakage Action; Step b, "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months " has been changed to "24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months " to allow time to do corrective actions without causing plant transient.

6) Section 4.4.6.2.1 (pg 3/4 4-20) Reactor Coolant System 4.4.6.2.1.c has a change in tolerance of 115 to 120 to be consistent with values previously given.

7) Section 3.4.8, Table 4.4-4 (pg 3/4 4-25, 4-26, 4-28) Reactor Coolant System.

Delete the phrase "of gross radioactivity" for the following:

Section 3.4.8.b Action 3.4.8.d (two places)

Item 4.a for table 4.4-4 This was deleted because it is an inappropriate wording of the sentence which implies you are measuring " gross radioactivity" in grams.

(0433M)

ATTACHMENT F (Continued)

(Section 3/4.4)

8) Section 3.4.93 (pg 3/4 4-35) Reactor Coolant System.

Alteration of sentence in step a; "a lift setpoint not to exceed the limits of" is inserted af ter "with" and before " Figure 3.4-4". The words

" nominal on" were deleted to avoid having to follow curve exactly.

(0433M)

REACTOR COOLANT SYSTEM l 3/4.4.3 PRESSURIZER PR00F".+ RE' lie'/I COPY LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with at least two groups of pressurizer heaters each having a capacity of at least 150 kW and a water level of less tr an or equal to 92% (1755. cubic feet).

APPLICABILITY: MODES 1, 2, and 3.

1 ACTION:

a. With one group ofabeetup pressurizer heaters inoperable, restore at least two groups of bookup heaters 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 and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor trip breakers open within 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 .

{

4 SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water level 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 .

4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit current at least once per 92 days.

4.4.3.3 The cross-tie for the pressurizer heaters to the ESF power supply shall be demonstrated OPERABLE at least once per 18 months by energizing the heaters.

BYRON - UNIT 1 3/4 4-9

O REACTOR COOLANT SYSTEM N00F & H3 E9y 3/4.4.4 RELIEF VALVES O

_. LIMITING CONDITION FOR OPERATION O 3.4.4 All power-operated relief valves (PORVs) and their associated bicch valves shall be OPERA 8LE.

APPLICA8ILITY: MODES 1, 2, and 3. .

ACTION:

O' a. With one or more PORV(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore

., the PORV(s) to OPERA 8LE status or close the associated block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STAN08Y 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 following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

O b. With one or more block valve (s) inoperable, within i nour either 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 its control power; 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 SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

O c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS g

4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERA 8LE at least once per 18 months by:

! a. Performance of a CHANNEL CALIBRATION, and g b. Operating the valve through one complete cycle of full travel.

l 4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle nf full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION a. of Specification 3.4.4.

4:4 4r3-The-emergency wr supply-for-the-PORVs-and-blockwalves-sheH-be

,fesonstrated -OPERA 8tE-et -l eas t -ence- per-la - mon ths-by s l - e. Menually-transferring motive anduntrol power-from the normal-to

--the emergency power-supply,-and-tr. Operating-the valves-through-a-complete-cycle-of-full-travel, e

BYRON - UNIT 1 ,

3/4 4-10 w - -

---w -p-- - , - - - , - - m--- --

_nm--w.. , - -----.-p y 4, q- -_,-,--,w-,m,e-,w- mq_ww- ---ew-g o,-y,,vm , p mem p---n-,-,--mm 4g---~

m.

REACTOR COOLANT SYSTEM 0 0 & i U T iV C 0 W ,

_3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE !

_ LEAKAGE DETECTION SYSTEMS j

~

LIMITING CONDITION FOR OPERATION l l

f 3.4.6.1 The following Reactor Coolant Syste:n Leakage Detection Systems shall '

be OPERABLE:

a. The Containment Atmosphere Particulate and Gaseous Radioactivity Monitoring System,

b. The Containment Floor Drain and Reactor Cavity Flow Monitoring System, and

c. The containment air pressure instrumentation and reactor containment fan cooler outlets and inlets Dewcell and dry bulb temperature instrumentation.

APPLICABILITY: H0 DES 1, 2, 3, and 4.

ACTION:

Q rv P With only two of the above required Leakage Detection Systems OPERABLE, 1I I operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed for gaseous and particulate i

radioacitivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required Gassous or '

Particulate Radioactivity Monitoring System is operable; otherwise, be in at l g least HOT STAN08Y 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 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . >

._ SURVEILLANCE REQUIREMENTS i

4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by: . . _

Containment Atmosphere Gaseous and Particulate Monitoring DN

a.

System performance of CHANNEL CHECK, CHANNEL CALIBRATION, and ANAL CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3,

b. Containment Floor Drain and Reactor Cavity Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months, and J,[. Containment air pressure and reactor containment fan cooler outlet and inlet temperatures-performance of CHANNEL CHECX, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL TEST, at least once per M l' 18 months.

the d lf.sc ,<L.sf.>w..f, l l C. VJ .:s < *he h w . R .'{ n s w .g3 /44** .ht yt b ~

slU)'lZ l f 'p, 'jff/tt,,, n.,,lll, 4-18 /

RYRON - UNIT 1 C g4 cAA, g/Gru Asa+.~~ a.- u d. fnC L. 4J'iuteJ

> A - m / ,G " M )y 3"*"7^h" 2.',* ~ g< &l. g%

O E

REACTOR COOLANT SYSTEM hk h k]@ gy OPERATIONAL LEAKAGE 3

LIMITING CONDITION FOR OPERATION

9 3.4.6.2 Reactor Coolant System leakage shall be limited to

a. No PRESSURE 80VNOARY LEAKAGE,

b. 1 gpa UNIDENTIFIED LEAKAGE, g c. 1 gpa total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator,

d. 10 gpa IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 40 gym CONTROLLED LEAKAGE at a Reactor Coolant System pressure of O 2235 20 psig, and

f. 1 gpa leakage at a Reactor Coolant System pressure of 2235 2 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.

O APPt.!CA8ILITY: MODES 1, 2, 3, and 4.

.cV ,

ACTION:

a. With any PRESSURE SOUNCARY LEAKAGE, be in at least HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

O b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE 80VNOARY LEAKAGE and leakage from ReactorCoolantSystemPressurg,8oursorbeinatleastHOTSTAN08YIsolation rate to within limits within + Valv within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00hN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

c. With any Reactor Coolant System Pressure Isolatioit Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD O SHuTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

BYRON - UNIT 1 3/4 4-19

D f3 REACTOR COOLANT SYSTEM

(' )

D SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above Ifmits by:

I a. Monitoring the containment atmosphere gaseous and particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

b. Monitoring the containment floor drain and reactor cavity sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; D c. Measurement of the CONTROLLED LEAKAGE to the reactor coolgot pump seals when the Reactor Coolant System pressure is 2235 2-K psig at least once per il days with the modulating valve fully open. The provisions of' Specification 4.0.4 are not applicable for entry into MODE 3 or 4; D d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

) 4.4.6.2.2 Each Reactor Coolant System Pressure Isolat. ion Valve scocified in D Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months, .

b. Prior to entering MODE 2 whenever the plant has been in COLD D SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months,

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve, and

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

)

sc 1

w 8YRON - UNIT 1 3/4 4-20 3

. = + - a m === ess en . e e am . 9 . _ _ - _ - a poossys

i I

REACTOR COOLANT SYSTEM

< k. .. GEW COPY 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 1 microCuria per gram DOSE EQUIVALENT I-131,

- and

b. Less than or equal to 100/E microCuries per gram :f ;rn. . ..,~,. 6f.-

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1, 2 and 3*:

a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 but within the allowable limit'(below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided that the cumula-tive operating time under these circumstances does not exceed 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months in any consecutive 12-month period. The provisions of

( Specification 3.0.4 are not applicqble;

b. With the total cumulative operating time at a reactor coolant specific activity greater than 1 microcurie per gram DOSE EQUIVALENT I-131 exceeding 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months in any consecutive 6-month period, prepare and submit a Special Report to the Commission within 30 days, pursuant to Specification 6.7.2, indicating the number of hours above this limit. The provisions of Specification 3.0.4 are not applicable;

c. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T"#9 less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and

d. With the specific activity of the reactor coolant greater than 100/E microcuries per gram ef ge : .i m unity, be in at least HOT STANDBY with T,yg less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With T avg greater than or equal to 500 F.

BYRON - UNIT 1 3/4 4-25

l c

l l

(, "'^"" ^^"'""'" !

. PROOF & REVIEW COPY LIMITING CONDITION FOR OPERATION ACTION (Continued)

M00E5 1, 2, 3, 4, and 5:

With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gras

- ' ;c: :: .. "Mkni, perfom the sampling and analysis requirements of

.Itani 4a of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its 1,imits.

L For this ACTION statament, prepare and submit a Special Report to the Casumis-stoner pursuant to Specification 6.7.2 within 30 days with a copy to the Direc-tor, Nuclear Reactor Regulation. Attention: Chief, Core Performance Branch, and Chief, Accident Evaluation Branch, U.S. Nuclear Regulatory Commission, Washington, D.C., 20555. This report shall contain the results of the speci-fic aethity analyses together with the following information:

'l .

1. Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, Sample

2. Results of the last isotopic analysis for radiofodi perfo rior to exceeding the limit, while limit was exceeded and ftme

.',' radiciodine activity was reduced to less than the 1 neluding for

'(.

each isotopic analysis, the data and time of sampling and the radiciodine concentrations,

3. Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, O 4. History of degassing operations, if any, starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the -

first sample in which the limit was exceeded, and

5. The time duration when the specific activity of the primary coolant exceeded 1 microcurie per gran 005,E EQUIVALENT I-131.

ll l

SURVEILLANCE REQUIREMENTS j 4.4.8 The specific activity of the reactor coolant shall be datamined to be within the limits by performance of the sampling and analysis' program of Table 4.4-4.

BYRON - UNIT 1 3/4 4-26

- -:-- ... : .- . _ _ _ _ --____.____ TX__ __ _ _ Li-[

m ,

TABLE 4.4-4 h REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM MODES IN WHICH SAMPLE h

H TYPE OF MEASUREMENT AND ANALYSIS SAMPLE AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED

1. Gross Radioactivity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1, 2, 3, 4 Determination **

2. Isotopic Analysis for DOSE EQUIVA- Once per 14 days 1 LENT I-131 Concentration

3. Radiochemical for E Determination *** Once per 6 months

1

4. Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 1#, 2#, 3#, 4#, 5#

Including I-131, I-133, and I-135 whenever the specific activity exceeds 1 pCi/ gram DOSE

, EQUIVALENT I-131 or 100/E pCi/ gram

) _a ------n....:..ss.,

i and m 1,2,3 b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWIR change '

exceeding 15%

of the RATED THERMAL c:3 r

POWER within a 1-hour CQ-period.

a3

\ -

a t

-=<

t

r 1

i

) l l

REACTOR COOLANT SYSTEM

) OVERPRESSURE PROTECTION SYSTEMS PRCOF & it!V,5 COPY LIMITING CONDITION FOR OPERATION ,

3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:

g gg Qpm e to ed th Msd

a. Two power-operated relief valves (PORVs) with nowinei 3 i.gvini.5 which

!, vary with RCS temperature as shown on Figure 3.4-4, or I b. The Reactor Coolant System (RCS) depressurized with an RCS vent of

greater Mo'e equal s to 2 square inches.

APPLICABILITY: MODE who he temperature of any RCS cold leg is less than or equal to 340*F y 5 a MODE 6 with the reactor vessel head on.

) ACTION: 1

a. With one PORV inoperable, either restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the RCS through at least a 2 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

)q V b. With both PORVs inoperable, depressurize and vent the RCS through at least a 2 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

t c. In the event either the PORVs or the RCS vent (s) are used to mitigate I

an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.7.2 within 30 days. The report shall describe the circumstances initiating the

) transient, the effect of the PORVs or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.

d. The provisions of Specification 3.0.4 are not applicable.

)

1 D

BYRON - UNIT 1 3/4 4-35 h

l

ATTACHMENT G (Section 3/4.5)

1) ' Section 3.5.3 (pg 3/4 5-7) Emergency Core Cooling System Delete the phrase "or equal to" from the Note, .it comes after "than" and before "350*F". This is done to make note describe Mode 4.

2) Section 4.5.3.2 (pg 3/4 5-8) Emergency Core Cooling System Delete the phrase "less than or equal to" from Section 4.5.3.2. It comes af ter "is" and before "350*F". This is done to make section be in description of Mode 4.

3) Section 3.5.4 (page 3/4 5-9) Refueling Water Storage Tank.

Delete this page. Information included in Section 3.1.2.6.

i-a 1

4 (0433M)

)

B S EMERGENCY CORE COOLING SYSTEMS

-- 3/4.5.2 ECCS SUBSYSTEMS - Tava < 350*F PR00F & REYH COPY D -- - _ _ _ _ _ _ _ _ _ _ _ _

LIMITING CONDITION FOR OPERATION D 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE: ;

a. One OPERABLE centrifugal charging pump,*

b. One OPERABLE RHR heat exchanger, D

c. One OPERABLE RHR pump, and

d. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of g operation.

APPLICABILITY: MODE 4. _

ACTION:

a. With no ECCS subsystem OPERABLE because of the inoperability of I4 either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With no ECCS subsystem OPERABLE because of the inoperability of p~ either the RHR heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERA 8LE status or maintain the Reactor Coolant System T,yg less than 350*F by use of alternate heat removal methods,

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to

} ,

the Commission within 90 days, pursuant to Specification 6.7.2, describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than vi awai Lv 350*F.

[ %

L)

BYRON - UNIT 1 3/4 5-7

I l

l EMERGENCY CORE COOLING SYSTEMS f%{ggg

> :. SURVEILLANCE REQUIREMENTS - ---- ~- - .

4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2.

> 4.5.3.2 All charging pumps and Safety Injection pumps, except the above required OPERABLE pumps, shall be demonstrated inoperable by verifying that the motor circuit breakers are secured in the open position at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the temperature of one or more of the RCS cold legs is hss t.'.= :r ;;gi te 350*F.

)

O .

1 l

l l

! BYRON - UNIT 1 3/4 5-8

~ ~

p EMERGENCY CORE C ING SYSTEMS t -

PriDDT & HEW COPY 3M. 5. 4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION ,

3.5.4 The fueling water storage tank (RWST) and the heat traced portions of the assoc tad flow paths shall be OPERA 8LE with:

c 5 -/

a. A min contained borated water level of $8-N (395,000 gal [ons),

b. A mini baron concentration of 2000 ppe, -

c. A minimum ter temperature of 35'F, and

d. A maximum wat temperature of 100*F.

APPLICA8ILITY: MODES 1, 3, and 4.

ACTION:

With the RWST inoperable, restore the tank to OPE BLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be ir, at least HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and ip' COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I SURVEILLANCE REQUIREMENTS

/.

4.5.4 The RWST shall be demonstrated OPERA 8LEt

\

I a. At least on per 7 days by: -

1) Ver ing the contained borated wat volume in the tank, and

2) erifying the baron concentration of the water. -

\

b. A least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when e outside air temperature is either less thh 35'F or greater than 100*F.

. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWS vent path i

temperature to be greater than or equal to 35'F w en the outside air temperature is less than 35'F.

8YROM - UNIT 1 3/4 5-9 i

l

ATTACHMENT H (Section 3/4.6)

Circled items noted in this attachment have been previously submitted.

1) Section 4.6.1.5 (pg 3/4 6-7) Containment Systems.

Changed numbering System to Reflect the subject in Question.

Changed Description Title to put in main control vernacular.

2) Section 3.5 1 ; (pg 3/4 6-11) Containment Systems Deleted "provided. . . time" from 3.6.1.7b because Byron only has one line.

3) Section 4.7.1.7.3/4.6.1.7.4 (pg 3/4 6-12) Containment System.

Replace " Operable" from 4.6.1.7.3 and insert "to maintain integrity".

Replace " Operable" from 4.6.1.7.4 and insert "to maintain integrity".

4) Section 4.6.2.3 (pg 3/4 6-15) Containment System Deleted af ter " verifying" and before " flow" and phrase "a cooling water" and its place put "an essential service water" to describe the proper system description.

b) Table 3.6-1 (Pages 3/4 6-18 to 22) Containment Isolation Valves For clarification, "s" has been changed to "sec".

Section 4.6.1.7.4 (pg 3/4 6-12) Containment System The measured leakage rate of 0.05La is more appropriate for the 8-inch valve instead of 0.012a. We cannot determine.the basis for 0.01La.

l

_ _ ~ ._ _ _ . _ _ _ ._ . _ _ . - . __

. C.

Os e umBV COPY

, CONTAI!WENT SYSTEMS V.. . . . ..

O'. AIR TEMPERATURE

LIMITING CONDITION FOR OPERATION

~

dOJ 3. 6.1. 5 Primary containment average air tamperature shall not exceed 122*F.

APPt.ICABILITY: MODES 1, 2, 3, and 4.

a E .

I Q'; With the containment average air tamperature greater than 122*F, reduce the average air temperature to within the Ifnit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be in at least HOT STAN08Y 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 following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l}O ii a.

SURVEILL" REQUIREMENTS 4.6.1.5 The primary containment average air temperatureI[ tall be the .

,~.

arithmetical average of the tamperatures of the running fans at the

'- following locations and shall be determined at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

Location l&. RCFC Wit A .L... Air Riar,* %hn. Alatr L

! (y IS. RCFC 2:1t 5 tt ; Air-Riset'Dq Ba mg_ ._y l 16. RCFC tit C dr-- Ai- R14erran e 'D9'B e W rw p.- A w l& RCFC Unit 2 S*"-- Air Ate ~Dq p w7 t.}

.o .

'S 8YRON - UNIT 1 3/4 6-7 l

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

?

CONTAINNENT SYS G S P 00F & HEW COPY

-Q CONTAIMENT VENTILATICA SYSTEM LIMITING CONDITION FOR OPERATION O. 3. 6.1. 7 Each containment purge supply and exhaust isolation valves shall be QPERABLE and:

, s. Each 40-inch containment shutdown. purge supply and exhaust isolation valve shall be closed and sealed closed?and Vwu nrr.yy

- b. The 8-inch containment purge _ly and exhaust isolation valve (s)

. e may be open for up to 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months during a calendar year.pnif:f = v t<- 4. .- .* m e4 c APPLICA8ILITY: MODES 1, 2, 3, and 4.

ACTION:

S a. With a 48-inch containment purge supply and/or exhaust isolation valve open or not sealed closed, close and/or seal close that valve or 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 STAN08Y within the next 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and in COLD SHUTDOW g ) within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

h b. With the 8-inch containment purge supply and/or exhaust isolation valve (s) open for more than 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months during a calendar year, close the open 8-inch valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , othe mise be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in i

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

M c. With a containment purge supply and/or exhaust isolation valve (s) having a asesured leakage rate in excess of the limits of Specifi-

'- cations 4.6.L7.3 and/or 4.6.1.7.4, restore the inoperable valve (s) h to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , othe mise be in at least HOT li stale 8Y 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 following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

" Oi h

Jo:

lI H ,

1 i4 ,

SYRON - UNIT 1 3/4 6-11 l -

w .e- 3-- p- --w-.-w-g.- y

.wo----. ----.-i g y. , -.g. c yy, iy,,.---p-w--,,.yy.-.-%,-...-r,e-.w-wer.-,.. e,-w --, wag,es,---e-ie - . . .%y.m.a

m , . .

CONTAINMENT SYST1[NS

) ~

P SURVEILLANCE REQUIRENENTS _

- /m 4.6.1.7.1 Each 38-inch c,pntainmen and exhaust isolation valve (s) l

' shall be verifi ' sealed 4}o' sed ar[tclosed purg6attupplys least once per 31 days.

Q Pour retro

4. 6.1.7. 2 The clamulative time that alt 9-inch containment purge supply and/or

~

exhaust isolation valves have been open during a calendar year shall be p

deteristned at least once per 7 days. 4 k- 4. 6.1. 7. 3 At least once per 6 months on a STAGGERED TEST BASIS, the inboard 6 and outboard valves with resilient asterial seals in each sealed closed 4 containment purge supply and exhaust penetration shall be demonstrated 6PERA8tf"- x by verifying that the measured leakage rata is less than 0.05 L, when pressurized to at least P ,, 43.6 psig.

n.

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

exhaust isolation valve with resilient natorial seals shall be demonstrated verifying that the sensured leakage rata is less than 0<01 L, when pressurized to at least P,, 43.6 psig, o,of C. , Tb

%./ m A e 84Tn 8M I NTf.6 # L l

Q C

l J

-) SYRON - UNIT 1 3/4 6-12

.. ..- a .. . - . . - . -. . . . . . - .

I

(

CONTAINMENT SYSTEMS h hh gg J

-., CONTAINMENT COOLING SYSTEM

LIMITING CONDITIONS FOR OPERATION 3.6.2.3 Two electrically independent systems of containment cooling fans shall be OPERA 8LE with one fan to each systas.

i G' '

APPLICA8ILITY: MODES 1, 2, 3, and 4.

a. With one system of the above required containment cooling fans

.n inoperable and both Containment Spray Systans OPERA 8LE, restore the

'O inoperable system of cooling fans to OPERA 8LE status within 7 days or be in at least HUT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COG SHUTD0hm within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With two systams of the above required containment cooling fans inoperable and both Containment Spray Systems OPERA 8LE,' restore at

',/~ least one systes of cooling fans to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTD0tM within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore both above required systems of cooling fans to OPERA 8LE status within 7 days of initial loss or 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 SHUTD0tM within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

O*.v Y c. WithonesystemoftheaboverequiredcontainmenU.coIingfans inoperable and one Containment Spray System inoperable, restore the inoperable Spray System to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTD0tm within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore the inoperable system of )

containment cooling fans to GPERA8LE status within 7 days of initial

' g,' - loss or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in l' COLD SHUTD0tm within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ,

le .

SURVEILL=*E REQUIREMENTS 4.6.2.3 Each systen of containment cooling fans shall be demonstrated OPERA 8tE:

a. At least once per 31 days by: -

1) Starting each fan systan in slow speed from the control room, and verifying that each fan system operates for at least z -

15 minutes, and 'E mW.

Verifying.e-eee M eq ..w qe4Tw. ;;^ . sEyflow rata of greater than or equal to

2) -

2660 gym to each cooler.

b. . At least once per 18 sonths by verifying that each fan system starts automatically on a Safety Injection test signal.

O.

SYRON - UNIT 1 3/4 6-15 l.

l l

\

.-_-' Q ,'

l . . . .

1.- , . ~ . . - . - . - - _ . , . _ _ _ _ , - , - - - _ - . . _ _ _ _ _ . _ _ , . _ . . _ - - . _ , _ _ _ . ~ . - - . - _ . - . _ _ _ . _ _ _ . _ _ - _ __

TABLE 3.6-1 PRDOF & REVIEW COPY CONTAINMENT ISOLATION VALVES

(

ISOLATION TYPE OF VALVE NO. FUNCTION TIME ( h f 4 C OPERATOR

1. Phase "A" Isolation .

ICV 8100 Chemical and Volume Control 10 Motor ICV 8112 . Chemical and Volume Control 10 Motor ICV 8152 Chemical and Volume Control 10 Air Operator with solenoid accessory 1CV8160 Chemical and Volume Control 10 Air Operator with solenoid accessory 1 WOOS 6A Chilled Water 50 Motor 1 WOOS 6B Chilled Water 50 Motor 1W0020A Chilled Water 50 Motor 1W0006A Chilled Water 50 Motor 1W00208 Chilled Water 50 Motor IW0006B Chilled Water 50 Motor 1CC9437B Component Cooling 10 Air Operator with solenoid accessory 1CC9437A Component Cooling 10 Air Operator with solenoid accessory IFP010 Fire Protection 12 Air Operator with solenoid accessory IFP011 Fire Protection 12 Air Operator with solenoid accessory

' IIA 065 Instrument Air 15 Air Operator with solenoid accessory IIA 066 Instrument Air 15 Air Operator with solenoid accessory 10G079 Off gas 60 Motor 10G080 . Off gas 40 Motor .

IOG081 . Off gas 60 Motor 10G057A Off gas 60 Motor 10G082 Off gas 60 Motor 10G083 Off gas 60 Motor 10G084 Off gas 60 Motor 10G085 Off gas 60 Motor IPR 001A Process Radiation 4.5 Air Operator with solenoid accessory IPR 0018 Process Radiation 4.5 Air Operator with solenoid accessory IPR 066 Process Radiation 5 Air Operator with solenoid accessory BYRON - UNIT 1 3/4 6-18

t TABLE 3.6-1 (Continued) h sk gf []

( . . _ . . ,

CONTAINMENT ISOLATION VALVES ISOLATION TYPE OF VALVE NO. FUNCTION TIME 6 OPERATOR

1. Phase "A" Isolation (Continued) $##

1PS228A Process Sampling Solenoid -

1PS229A Process Sampling it 0-1& N/J*#

NA Solenoid 1PS230A Process Sampling 0-22N A # Solenoid IPS228B Process Sampling 11TN 4 # Solenoid 1P5229B Process Sampling 4-le NM Solenoid 1PS2308 Process Sampling 0-12*Al# Solenoid 1PS9354A Process Sampling 10 Air Operator with solenoid accessory 1P593548 Process Sampling 10 Air Operator with solenoid accessory 1PS9355A Process Sampling 10 Air Operator with solenoid accessory 1P59355B Process Sampling 10 Air Operator with solenoid accessory 1PS9356A Process Sampling 10 Air Operator with solenoid accessory 1PS93568 Process Sampling 10 Air Operator with i solenoid accessory 1PS9357A Process Sampling 10 Air Operator with solenoid accessory 1PS93578 Process Sampling 10 Air Operator with solenoid accessory 1RE9157 Reactor and Containment 10 Air Operator with Drains to Radiowaste solenoid accessory 1RE9159A Reactor and Containment 10 Air Operator with Drains to Radiowaste solenoid accessory 1RE91598 Reactor and Containment 10 Air Operator with Drains to Radiowaste solenoid accessory 1RE9160A Reactor and Containment 10 Air Operator with Drains to Radiowaste solenoid accessory 1RE91608 Reactor and Containment 10 Air Operator with Orains to Radiowaste solenoid accessory 1RE1003 Reactor and Containment 10 Air Operator with Orains to Radiowaste solenoid accessory 1RE9170 Reactor and Containment 10 Air Operator with Orsins to Radiowaste solenoid accessory 1RY8025 Reactor Coolant Pressurizer 10 Air Operator with solenoid accessory 1RY8026 Reactor Coolant Pressurizer 10 Air Operator with solenoid accessory 1RY8033 Reactor Coolant Pressurizer 10' Air Operator with solenoid accessory 1RY8028 Reactor Coolant Pressurizer 10 Air Ooerator with solenoid accessory g B NtN - UNIT 1 3/4 6-19

~

Y Pr0Per V9ive oper a tion will be deixon streted by V6r1/vina +k, yafn 49 pof.g g tp_ __jy_ _Y G g ul ted P osi Ho n.

TABLE 3.6-1(Continuedl

( CONTAINMENT ISOLATION VALVES l

ISOLATION TYPE OF VALVE NO. FUNCTION TIME (f)Me C OPERATOR

1. Phase "A" Isolation (Continued)

ISI8880 Safety Injection 10 Air Operator with solenoid accessory 15I8964 Safety Injection 10 Air Operator with solenoid accessory ISI8871 Safety Injection 10 Air Operator with solenoid accessory 15I8888 Safety Injection 10 Air Operator with solenoid accessory 15A032 Service Air 4.5 Air Operator with solenoid accessory 1SA033 Service Air 4.5 Air Operator with solenoid accessory ISD002C Steam Generator Blowdowr. 7.5 Air Operator with solenoid accessory ISD005B Steam Generator Blowdown 3 Air Operator with solenoid accessory 1S00020 Steam Generator Blowdown 7.5 Air Operator with

( solenoid accessory ISD002A Steam Generaton Blowdown 7.5 Air Operator with solenoid accessory ISD005A Steam Generator "'owdown 3 Air Operator with solenoid accessory 15D0028 Steam Generator Blowdown 7.5 Air Operator with solenoid accessory ISD002E Steam Generator Blowdown 7.5 Air Operator with solenoid accessory ISD005C Steam Generator Blowdown 3 Air Operator with solenoid accessory IS0002F Steam Generator Blowdown 7.,5 Air Operator with solenoid accessory ISD002G Steam Generator Blowdown 7.5 Air Operator with solenoid accessory l

150005D Steam Generator Blowdown 3 Air Operator with e solenoid accessory l ISD002H Steam Generator Blowdown 7.5 Air Operator with solenoid accessory l

1RF026 Waste Disposal 15 Air Operator with solenoid accessory 1RF027 Waste Disposal 15 Air Operator with l solenoid accessory l \

BYRON - UNIT 1 3/4 6-20 i

i l

l TABLE 3.6-1 (Continued)

(

CONTAINMENT ISOLATION VALVES TYPE OF ISOLATILN VALVE NO. FUNCTION TIME (a) OPERATOR I

2. Feedwater Isolation 1FWOO9A S/G Feedwater 5 Hydraulic Operator 1FWOO98 S/G Feedwater 5 Hydraulic Operator 1FWOO9C S/G Feedwater 5 Hydraulic Operator 1FWOO90 S/G Feedwater 5 Hydraulic Operator 1FWO35A , S/G Feedwater 6 Air Operator with solenoid accessory 1FWO35B S/G Feedwater 6 Air Operator with solenpid accessory IFWO35C S/G Feedwater 6 Air Operator with solenoid accessory IFWO35D S/G Feedwater 6 Air Operator with solenoid accessory 1FWO39A S/G Feedwater 6 Air Operator with solenoid accessory 1FWO398 S/G Feedwater 6 Air Operator with solenoid accessory IFWO39C S/G Feedwater 6 Air Operator with solenoid accessory

(, 1FWO390 S/G Feedwater 6 Air Operator with solenoid accessory 1FWO43A S/G Feedwater ,

6 Air Operator with solenoid accessory 1FWO43B S/G Feedwater 6 Air Operator with solenoid accessory 1FWO43C S/G Feedwater 6 Air Operator with i

solenoid accessory IFWO430 S/G Feedwater 6 Air Operator with solenoid accessory 9

BYRON - UNIT 1 3/4 6-21

l TABLE 3.6-1 (Continued) PRODF & HM COPY I CONTAINMENT ISOLATION VALVES i

ISOLAT N TYPE OF VALVE NO. FUNCTION TIME (1(0f4 C OPERATOR

3. Containment Ventilation Isolation IVQ005A Containment Purge 5 Air Operator with solenoid accessory IVQOO58 Containment Purge 5 Air Operator with solenoid accessory IVQ005C Containment Purge 5 Ai* Operator with solenoid accessory IVQ003 Containment Purge 5 Air Operator with solenoid accessory IVQ002A Containment Purge 5 Hydraulic Operator IVQ0028 Containment Purge 5 Hydraulic Operator IVQ004A Containment Purge 5 Air Operator with solenoid accessory IVQOO4B Containment Purge 5 Air Operator with solenoid accessory IVQ001A Containment Purge 5 Hydraulic Operator IVQ001B Containment Purge 5 Hydraulic Operator

4. Phase "B"/ Components Isolation ICC9414 Component Cooling 10 Motor 1CC9416 Component Cooling 10 Motor ICC685 Component Cooling 10 Motor

. ICC9438 Component Cooling 10 Motor ICC9413A Component Cooling 10 Motor

- ICC9413B Component Cooling 10 Motor

5. Safety Inject. ion / Main Steam Isolation IM5001D Main Steam 5 Hydraulic IM5101D Main Steam 10 Air Operator with

. Solenoid Accessory 1M5001B Main Steam' 5 Hydraulic IMS101B Main Steam 10 Air Operator with Solenoid Accessory 1MS001A Main Steam 5 Hydraulic IMS101A Main Steam 10 Air Operator with Solenoid Accessory IMS001C Main Steam 5 Hydraulic IMS101C Main Steam 10 Air Operator with Solenoid Accessory BYRON - UNIT 1 3/4 6-22 f

ATTACHMENT I (Section 3/4.7)

Circled items noted in this attachment have been previously submitted.

1) Section 4.7.1.2.1 (pg 3/4 7-5) Plant Systems Delete 4.7.1.2.1.a.3, from Section 4.7.1.2.1 because (a.2) verifies correct position of system. Number (a.3.) does not address any valves not already addressed in (a.2). There is no auto control referred to in (a.3).

2) Section 4.7.1.2.1 (pg 3/4 7-5) Plant System Inserted in [b.2(d)], between " Bus" and "Undervoltage" the number 141, to add a more complete description of the Bus desired.

3) Section 3.7.3 (pg 3/4 7-11) Plant Systems The Component Cooling Water System loops do not lend themselves to the present LCO. A pump or heat exchanger may service any loop. The specification must be expanded to account for the components.

4) Section 3.7.6/4.7.6 (pg 3/4 7-15) Plant System Changed all phrases " Control Room Emergency Air Cleanup Systems" to

" Control Room Ventilation Cleanup Systems" to give proper identification of System. In Section 4.7.6.b the word " duct" was inserted af ter "the" and before " heaters" to give a thorough description of equipment.

5) Section 4.7.6.c (pg 3/4 7-16) Plant System In Section 4.7.6.c.1) The word " Pressurization" was deleted from the Section and in its place, af ter the word "The" and before " System" the word " Makeup", was inserted to give proper description to system being used.

In Section 4.76.6.c.2) The Criteria 0.2% was changed to 0.175% to Comply with number from Regulatory Guide 1.52.

In Section 4.7.6.c.3) The word " Pressurization" was deleted from the Section and in its place, after the word "The" ard before " System" the word " Makeup", was inserted to five proper description to system being used.

6) Section 4.7.6.d (Page 3/4 7-16) Control Room Emergency Air Cleanup System Change to 0.175% from 0.2% is necessary.

Table 3 of Regulatory Guide 1.52 " Laboratory Tests for Activated Carbon" has a value of .175% for methyl iodine penetration.

(0433M)

ATTACHMENT I (Continued)

(Section 3/4.7)

7) Section 4.7.6.e (pg 3/4 7-16) Plant System In Section 4.7.6.e.1) The word " Pressurization" was deleted from the section and in its place, after the word "The" and before " System" the word " Makeup", was inserted to give proper description to system being used.

In Section 4.7.6.e.2) The phrase " Smoke Density High" was deleted and in its place "ESF actuation signal" after the word "a" and before the word "or" to give proper name to the System described in the Section.

8) In Section 4.7.6.e.2) the word " recirculation" was deleted and in its place is inserted the word " makeup" to give proper name description to the mode of operation.

9) Section 4.7.6.e.4 (pg 3/4 7-17) Plant System Operation In Section 4.7.6.4 the word "that" located after " Verifying" and before "the" was deleted to give correct grammatical format. The "s" was deleted from heater to express the use of a singular heater unit. The word

" dissipate" was deleted and in its place the words " power consumption is" to give description of what is actually occurring in process. The words "in accordance with ANSI ,N510-1975." ,

After number 4 a number 5 stating " Verify heaters perform in accordance with N510-1975." was added to ensure that heaters are in accordance with N510-1975.

10) Section 4.7.6.f (pg 3/4 7-17) Plant System In Section 4.7.6.f the words "and bypass leakage" was deleted because bypass leakage testing will not be used at Byron.

The word " Pressurization" was deleted from 4.7.6.f and in its place, after the word "the" and before the word " System", the word " Makeup" was inserted to give correct name description to the System.

11) Section 3.7.10.1 (pg 3/4 7-28) Plant System.

Changed the word " Spray" located af ter "or" and before " System" to " Foam" to give system proper description.

12) Table 3.7.5 (pg 3/4 7-35) Fire Hose Stations.

The corrections made in Table 3.7-5 were to add clarification to where equipment are located and to give proper name of equipment used at Byron Station.

(0433M)

ATTACHMENT I (Continued)

(Section 3/4.7)

13) Table 3.7-5 (pg 3/4 7-35) Fire Hose Stations.

The corrections made in Table 3.7-5 were to add-clarification to where

equipment are located and to give proper name of equipment used at Byron

, Station.

14) Table 3.7-5 (pg 3/4 7-37) Fire Hose Stations. ;

i

~

The corrections made in Table 3.7-5 were to add clarification to where equipment are located and to give proper name of equipment used at Byron

, Station.

w-I (0433M)

e

} ' PUUf SYSTEMS PR00F & EEW CDPY SURVEILLANCE REQUIREMENTS (Continued)

2) Verifying by flow or position check that each valve (manual .

power-operated, or automatic) in the flow path that is not .

locked, sealed, or othe mise secured in position is in its O correct position; and C

[3) Vertfri fu 1y open

$c val @in tSe flow /pathAs try the i ting - ig'-- the Ajuxili,a'ry Feedwated Syst4m is 4

( y)::d/" < * **a*7= ** Pa ?. arfRAip ==

0

? b. At least once per 18 months during shutdown by:

l 1) Verifying that each automatic valve in the flow path activates to its correct position upon receipt of an Auxiliary Feedwater Actuation test signal, and ._

O Verifying that the motor-driven pump and the direct-driven C 2) diesel pump start automatically upon receipt of each of the following test signals:

l c O, y b) 5 - morwt.,t-i - fem-s=

ic%

1 i

generator, or .

c) Undervoltage on Reactor Coolant Pump 6.9 kV Buses (2/4), or 1H -

d) ESF Bus s undervoltage (notor-driven pump only). y .

c'C_

,~~

-.. - 4,7.1.2.2 An auxiliary feeeeter flow path to each steen generator sPall be - -

~~

demonstrated OPERABLE following each COLD SHUTDOWN.of greater-thart 30-days ~

prior to entering ISOE 2 by verifying normal flow to each steen generator.

i 4.7.1.2.3 The auxiliary fesowater pump diesel shall be demonstrated OPERA 8LE:

CO a. At least once per 31 days by verifying the fuel level in its day tank;

b. At least once per 92 days by verifying that a sample of diesel fuel from its day 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, water, and sediment; and i

-O '

c. At least once per 18 months, during shutdown, by subjecting the diesel to an inspection in accordance with its manufacturer's recommendations for this class of service.

l ..

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, PLANT SYSTEMS .

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/ COMPONENT COOLING WATER SYSTEM g4.7.3

/ ITING CONDITION FOR TION / &M (d'MW /WE- MM jf

{ 3.7.3 At least two component cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4. ![

, g:

With only one component cooling water loop OPERABLE, restore at least two [':

loops 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 STAND 8Y 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 following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

f SURVEILLANCE REQUIREMENTS 6 0

4.7.3 At least two component cooling water loops shall be demonstrated

'W* '

OPERABLE: _

i At least once per 31 days by verifying that each valve (manual, f- g. t

a. ,

power-operated, or automatic) servicing safety-related equipment

! that is not locked, sealed, or otherwise secured in position is in j E , i its correct position; and ;

i, i .:

'f %

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s b. At least once per 18 months during shutdown, by verifying that: i

1) Each automatic valve servicing safety-related equipment actuates

' to its correct position on a Safety Injection test signal, and

2) Each Component Cooling Water System pump starts autoestically f x

on a Safety Injection test signal. ;

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" " ve 3/4.7.6 CONTROL ROON ':"~ C= A!"-CMSTEM PROOF & EE# COPY LDetrini CONDITION FOR OPERATION f 2:%t:m ,

3.7.6 Two independent Control Room "c:. ai Air Cid5ystems shall *be

'O OPERASLE.

AMLICMILITY: ALL MODES M .

~W MODES 1, 2, 3 and 4:

% MD- % y

- , A;e Cleenap Systas inoperable, restore a With one Control Room -. -

the inoperable system to OPERABLE statue within 7 days or be in at least HOT STAfESY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTD0km within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . .

MODES 5 and 6:

qs/NANm y *

. a. With one Contre) Room ~ ,cc; ^ P CI:c a System inoperable.

restore the inoperable system to OPERABLE statur within 7 days or -

initista and asistain ion of the remaining QPERASLE Control Ross ~ ystas in the recirculation mode.

h"E C1 N 3.0 %

j O." ). b. With both Control Room Emergency Air C1 Systans inoperable, or

with the OPERASLE Control Ross ^1r Ct c. , ystas, required to be in the recirculation mode by ACTION a. not capable of being powered by an OPERASLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes, j',j SURVEILLM R8NIREMENTS

%gKo,t'm 4.7.6 Each control Room ----. rai Ati-C1-.dystemshallbedemonstrated x

0 GPERASLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air tamperature is less than or equal to 90*F;

b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal

"- adserters and verifying that the system operates for at least 10 continuous hours with the heaters operating;

,,) 8YRON - UNIT 1 3/4 7-15 2 - . - -

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PR00F & REY!EW COPY :

Pt.Nff SYSTEMS

_ SURVEILLANCE REQUIREMENTS (Continued) n 7

c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any venttlation zone --

communicating with the system by:

O, 1) Verifying that the cleanup system satisfies the in place C penetration and bypass leakage testing acceptance cHtaria of less then 0.0EK and uses the test procedure guidance in Regulatory Positions C.5.a. C.5.c. and C.5.d of Regulatory and the systam . flow _ rate _is _

Guide 6000 cfts 1.52, Revision

+.15.for the 2 March, z:d- 1978, *f- 83ystan and 31,~000 .i e I N G M. Eg" ~ ~ ~

o f hw atfesF8 W 7 _ _ _ --

- C - <

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtiined in accordance with Regulatory Position C 6.h of Regulatory Guide 1.52 Revision 2 March 1978, asets the laboratory testing critaria of Regulatory Position C.6.a of Regulatory Guide 1.52, C

O Revision 2, March 1978, for a methyl iodine penetration of less then E D; and (-7 04 O v

) o,R57o CWu2 l

3)

Verifying action-System a sy(stem flow rata of 6000 efin + 10K for the'17...... ; -

and-51i0002ees415-for-thNeeulation-Sy~stasi) ~ ~~

i

~

when tasted in accordanca witK ANSI N610-1975.~~ ~

j cOi

_. d. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorter operation by verifying

within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory t

. Position C.8.h of Regulatory Guide 1.52. Revision 2, March 1978, ,

asets the laboratory testing critaria of Regulatory Position C.6.a of Regulatory Guide 1.52. Revision 2. March 1978, for a methyl

LU iodine penetration of less theno.era,7. r15

._ .-- JA _ At least once per 18 months by. __ . ._., .__ _ _ _ _ _ _ . . _ . . _ , _ .

i

, 1) Verifying.that the pressure drop across the combined HEPA filters and charcoal adsoreer banks is less than 6.0 inches c idster Gauge while opersping_the_ system 8000 cfin +.15 for the/< . z;--- '--^i stam 4md 51iOe9 Q4he-RetircElaiIes-SystemNAKQat EQ a flow

-mgw4 Verifying that on a "t L. _it, Mi,. or High Radiation-Control 8

2)

Room, Outside Air Intake, or Turnine 8uilding Intake test signal,

! the system automatically switches into ab L s1.;.6. sode of I

l OD. operation with flow through the HEPA filters and cha.

adsorter banks, .

l ,

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O. ,uMr - NNN a i, O . SURVEILLANCE REQUIREMENTS (Continued) c.

3) Verifying that the system amintains the control room at a posi-tive nominal pressure of greater than or equal to 1/8 inch Water Gauger relative to ambient pressure in areas served by the system, o

d P v.

O 4) Verifying dthe heaterbowere,cm4somPraw s M '# 27.2 + 2.7 ts kW :' Q w d 4e-eseerdance-wttit-ANSI-N510-M x

5) r each coglete or partial replacement of a HEPA filter bank, by

f. Afte% h 46h ond. AMsr NSLo- t%5.

verifying that the cleanup system satisfies the in place penetration x f 7,; -- 1 ting acceptance criteria of less than 0.055 in 0 accordance'with ANSI 1810-1975 for a 00P test aerosol while operating

'O the system at__a_ flow rate of 6000 eft + 105 for the presserfu'A.-

'"""#p Systaq'siit-51;400-eftglat for'thm-emet 19:alation~5pstas;= ant

g. After each cogleta or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place

'n penetration and bypass leakage testing acceptance critaria cf less C. than 0.055 in accordance with ANSI NS10-1975 for a halogenated hydro-k carhonrefrigeranttestgaswhileoperatingthesystasatallow]fe rate of 6000 cfm + 105 for the Systan & 51;000-c '

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PR00F & RUIN 2 3/4.7.10 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESS!ON WATER SYSTEM LIMITING CONDITION FOR OPERATTON g

,C 3.7.10.1 The Fire suppression Water System shall be OPERA 8LE with:

a. Two fire suppresafon pumps, each Uth a capacity of 2500 gym, with their discharge aliped to the fire suppression header, and

!:h b. An OPERA 8LE flow path capable of taking suction from the f1ume and U transferring the water through distribution piping with OPERA 8LE sectionalizing control or isolation valves to the yard hydrant curh valves, the last valve ahead of the water flow alam device on each sprinkler or hose standpipe, and the last valve ahead of the deluge x

- valve on each Deluge or M 5ystas required to be OPERA 8LE per O's Specifications 3.7.10.2 and 317.10.5.

APPLICA8ILITY: At all times. 0* M M:

a. With one pump and/or one water supply inoperable, restore the inoperable equipment to OPERA 8LE status within 7 days or provide an alternata backup pump or supply. The provisions of Specifica-l- tions 3.0.3 and 3.0.4 are not applicable.

b. With the Fire Suppression Water System 4the mise inoperabie

Q

establish a backup Fire Suppression Water Systas.within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS b 4.7.10.1. 1 The Fire Suppression Water System shall be demonstrated OPERA 8LE:

a. At least onca per 7 days by verifying the contained water supply volume,

b. At least once per 31 days on a STAGGERED TEST SASI$ by starting the 0,

electric entor-driven pump and operating it for at least 15 minutes on recirculation flow,

c. At least once per 31 days by verifying that each valve (eanual, power-operated, or automatic) in the flow path is in its correct position, C.

8YROM - UNIT 1 3/4 7-28 N ***=*.e- * *- m- - +mm.- -

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W Y 5 TABLE 3.7-5

! FIRE HDSE STATIONS LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE

=

Q Aux. Roof 11 - 1 4

" L-10: South wall of safety valve penthouse 481 1 0FP331 L-26: North wall of , safety valve penthouse 481 2 0FP338 Aux. Bldg. N S-18: By dumb waiter 480 233 0FP458 S-15: By U-1 prefilters C.Rt%r Sl airsT 471 176 0FP329 S-21: By U-2 prefilters (,titar Shr$) 471 177 0FP334 ,

Q-17: Wall by elevator in upper cable room 469 2448' 0FP469 Q-19: Wall by stairs in upper cable room 469 252r 0FP477 cukstat elsk t' L-ll: bouthweetco -;;;; m.4,;e 7:::UCSRd467 240 0FP465

- L-14: By the sout r of room *EE W CSR c,1 467 241

OFP466 Y N-13: By the nort corner of cr ;;Clutus A-s467 242

OFP467 lN gg Q-13: In the nort corner of room LEE 94tShtH467 243

OFP468 w g uess,c : West of the elevator in upper cable room 467 245

OFP470 N-18: North wall of 7::: 1550 uc5% c-1 467 246 ** OFP471 outq,.de. N-18: South 11 of rr: 2CC 1 UCM c-1 467 247 +' 0FP472 L-25:* Nort corner of room SEE-tuC% A-2, 467 248 0FP4 73

g thec L-22: " Nort M-23i" South corner of room 266-4ucSR c.-2, corner of room 2EE-t uc% p,-2 467 467 249 +

250

OFP414 OFP475 M C:3 in ut._ P-20: West wall of room EEE-.3 uts0 c-2 467 251 + OFP476 CM Q-2Ti' Southwest corner of room 2EE-2 uc.sE B-2, 467 253 P OFP478 p S-21: By U-2 pr:f!!tersw(;lkes (u-2 side.1 464 232 0FP457 - a-S-15: By "-! p-af11**rs n (;llers t u-t s; del 464 234 0FP459 q'~

S-21: By U-2-NEPA

pr:f!!!er room % 411 hrs (u-2 s41456 231 0FP456 y S-15: By S-1-;;CPA . i.e.f f!ter --- NA CNescu a sW1456 235 QFP460 4 .

L-10: By Control room refrig. units L-12: By blowdown after filters p, F 387 387 106 107 0FP385 0FP384 Q.g N-18: By '.u .-F::i :'. i.. J iv.. 7 ; 1A 387 108 , OFP383 -4 N-23: By FeaQLR_shutdma p r 7 } gg g ggp 387} OfP376 Q-15: By 480V NCC 132X3 381 113 0FP382 V-18: By 54 :: H-- letdown heat exchanger 387 114 0FP379 . A Fire Rose hh d do d Sgpttj Ehe. primag meus of Are- hypres.s*m

! wiw w '"W 95 W ' "

g5 35- p' i p, l
W .

U-TA8LE 3.7-5 (Continued) co y FIRE HOSE STATIONS 2

LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE C

Aux. Bldg. (Continued) g ,11 t.i Ky swikk ed- 3 y M -P-It: S y c e..t i e l r .. m i ihrisr w - 455 20 0FP324 L-11: Sy ccel br, =i h ^ f= h V C. MR M OA *f 455 22 0FP332 LCSR c-L 238

OFP463 M-18:_ North wall ;) 1 :, .:s;- 444 p.ad Mll ~ L-7: ' S=theas t cerner of ' -- --- ^ ' - -- -- 1/SE h'1443 207 4 0FP327 By aac* drer ie : : 1Z1 S400 career e5 9:n M443 208
OFP327 In de P-10: Sy-west .le,.s te rc:: !?h.A*4 = .o "

443 209

OFP325 M-13: South wall of re= 123. LCsn c-1 443 210
OFP326 P-16 --P-13L: West Wall of r.~ 124 t,cq, p, g .g 443 211 + OFP328 S-21: By cabinet 2RY01EC W e c. ten ared 431 229 0FP455 5-24: By U-2 cont. shield wall Lelm. vert. %reA) 431 230 0FP456 S-12: By U-l cont. shield wall tela. pen. grea) 431 237 0FP462 m

                                             )                                                       P_11 . o,M?..M
                                                                                                              .      .- d. ...M.. 9%. . f.9..M. . .._ . ._ _ _
                                                                                                                                        .                                             430            52        0FP313 Q-19: By U-2 haraa iaiar*!^a a -ac McT vnbe itsle 430                                           54        0FP342 p

w P-24: By radwaste evaporator 430 55 0FP343

58 0FP319 V-17: By d' -l 1,u en inje-tie.. p r. ed be h d 430 0FP320 V-17: By W4 door to decon/ change area 430 61 eh L-11: By U waste oil tank room M 405 90 0FP315 P-18: By elevator 405 91 0FP318 P-23: By B-2 spent resin pumps 405 92 0FP349 i Q-11: By laundry tanks hMm gjl5 93 0FP314 "O g og

                                                             *g ,p                                   S-21: _Sp U-2 p!?e t .. I t..iivy pean
                                                                                                                                 .                                                    OS             94        0FP348        @

V-21: -1)f U-2 hydrogen recombiner 405 95 0FP345 co

                                                                          # *g                                                                                                                       %         OFP316        m V- Ab:; I)p U-l hydrogen recombiner {erira! pen:!                                405 O* % S-15: ,1)f U-1 pipe trece! tre!!e" be==                                                  405            97        0FP317        90 W

hudvosen recombirte.r- 368 130 0FP373 N-ll: By the recycle holdup tanks M-14: By the, a of U-1) 368 131 0FP374 P-14: By panel L 4JB _ 368 132 0FP369 k@c; L-20: By %(QTiTjof U-2 A 368 133 0FP355 .= P-21: By the blowdown condenser VW AltuPsa PS 368 134 0FP356 c3 L-25: By the ' - -- ' - - - - - " a - *-" 368 135 0FP361 c3 ) N-25: By r^;;e::eratice .:::te#c.htmic'at By panel IPL86J drain tankt 368 368 136 138 0FP357 0FP362 5-18: Q-ll: By Aux. Bldg. floor drain tanks 368 139 0FP368 U-15: By positive-dtsplech-charging-pump 368 140 0FP372 u-1 Sprag 60 lank

    %                   w                     any                     es-          ge           **   -

W W p app + O TA8tE 3.7-5 (Continued) E B FIRE HOSE STATIONS x LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE i C 2 Aux. Bldg. (Continued)

                 -4 P-11: By recycle evaporator feed pumps                                     350                151      OfP381 e

M-13: By 480V-MEG-t33N6 U-1 S W ra, 350 152 0FP370 N-23: By gas decay tanks 350 154 0FP352 , Q-19: By "B" Aux. 81dg. Equip. drain tank 350 155 0FP365 Q-17: By "A" Aux. 81dg. Equip. drain tank 350 156 0FP371 Q-13: By collection sump pumps 350 157 0FP380 S-18: Between moderating heat exchangers 350 158 0FP354 , V-18: 8etween(tht11er units ' B Fs 350 161 0FP353 350 163 0FP367 W-15: By C::t. ge, ; ; A- pmp in M-13: By leak detection sump 334 165 , 0FP448 P-18: By elevator pit I 334 166 . OFP449 w

                 )     Fuel Hand. Bldg.

Z-15: South of decon. area 430 170 0FP389 T 171 0FP386 1:j X-21: North of spent fuel pool 430 , Z-15: By 480V MCC 134X6 405 172 0FP388 ! AA-19: "y s p -* f =1 p!' .i;. ==:._ ,, rs 405 173 0FP387 Cont. #1 OMe FC. %g %m R-17: By reactor head assembly area 430 62 IFP163 R-2: Dy accumulator tank 1 C. 430 63 IFP154 3 C3 R-7: Bydialcli~ Eguipmed 430 64 IFP160 - Q l R-12: By charcoal filter IA 430 65 IFP157 R "-'--- r-*a-a ricaec B3 %S shirg 403 98 1FP164 g2e ' R-2: By 's.ce.e lh: '..;; gtyc ic. 403 99 IFP155 ' m R-7: By pressurizer (misde miW did{} 403 100 IFP161 a R-12: By panel IPL69J 403 101 IFP158 l m R-12: By p:::! 1." S^.J- t RT 381 143 IFP159 ::gg i R-17: By RC f :: 1" . IA- suunghtr5 381 144 1FP162 n - R-2: By 7:--! l'" ZJ - ECFC 1C. 381 145 IFP156 i : es R-7: By panel IPL52J 381 146 1FP165 U ]

ATTACHMENT J (Section 3/4.8) Circled items noted in this attachment have been previously submitted.

1) , Section 3.8.1.1 (pg 3/4 8-1) Electrical Power System.

In sections 3.8.1.1.a and 3.8.1.2.a, change "the Onsite Class 1E Distribution System with:" to "each Onsite Class 1E 4KV BUS with:" In sections 3.8.1.1.a.1 and 3.3.1.2.a.1, change the word " circuit" to "line". This clarifies the beginning and end points of the two required circuits.

       -     In section 3.8.1.1.a.2 and 3.8.1.2.a.2, the sentence should read "Either of the two transformers forming a system auxiliary transformer bank capable of supplying the buses which are normally supplied by both transformers forming the system auxiliary transformer bank." This clarifies that either transformer is an SAT bank and can be used to supply either Clas 1E 4KV bus.

Items a(1) and (2) are being replaced to provide clarification of the two required offsite power circuits. Also provides clarification that either transformer of a System Auxiliary Transformer bank is adequate.

2) Section 4.8.1.2.e.4.b (pg 3/4 8-4) Electrical Power System Add "of the loads" following "Af ter energization" on e.4.b. This is done for clarification.

3) Section 3.8.1.2 (pg 3/4 8-9) A.C. Sources Items a (1) and (2) are being replaced to provide clarification of the required offsite power source. Clarifies that either SAT bank transformer is adequate.

4) Action 3.8.2.1.a (pg 3/4 3-10) A.C. Sources Add "and/or Charger" after the word " bank" so that the Action refer to a

       " battery and/or charger bank. . .". This is added for clarification of the Action.

(0433M)

ATTACHMENT J (Continued) (Section 3/4.8) . 5) Section 4.8.2.1.2 (pg 3/4 8-11) Electrical Power System Add "M" af ter the word " ohm: for Surveillance Requirea.onts 4.8.2.1.2.b.2 and 4. 8. 2.1. 2. c . 3. Add the following note at the bottom of the page as follows:

"Obtained by subtracting the normal resistance of the connecting bus bar from the measured cell-to-cell and terminal connection resistance." At

~ Byron, that batteries are made up of 4 rows of battery cells. While the resistance limit can be met on cell-to-cell measurements of adjacent cells, it cannot be met on cell-to-cell measurements between cells in different rows. This is due to the intrinsic resistance of the connecting bus bars, which must be much longer between rows of cells that between adjacent cells. Because the concern here is that the terminal connections might be poor, subtracting the normal resistance of the bus bar itself from the measured value of cell-to-cell resistance will still provide the resistance of the terminal connection. Surveillance Requirement 4.8.2.1.2.b.3 should read "3) The electrolyte temperature of all connected cells is above 60*F". This new statement is for clarification of what need to be checked.

6) Section 3.8.3.1 (pg 3/4 8-14) Electrical Power System Section 3.8.3.1.c should read as follows:

                  '"c.      120-Volt AC Instrument Busses consisting of:

1. Instrument Bus 111 energized from its associated inverter connected to DC Bus 111.; 2. Instrument Bus 112 energized from its associated inverter connected to DC Bus 112,; 3. Instrument Bus 113 energized from its associated inverter connected to DC Bus 111,; 3. Instrument Bus 114 energized from its associated inverter
connected to DC Bus 112.

This places the AC Instrument Busses in a form similar to that of parts a ard b.

            -      Change " vital" to " Instrument" in both places in Action 3.8.3.1.b. to reflect station terminology.
            -      Change Action 3.8.3.1.c to read:

(0433M)

_~ ATTACHMENT J (Continued) (Section 3/4.8)

            "c. With a maximum of an A.C. inverter inoperable or not connected to its D.C. power supply, its associated A.C. Instrument bus may be powered from its regulating transformer power supply for up to 72 hrs; otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours."

Byron's design provides a self-regulating power supply to th A.C. inst. busses in addition to the AC inverter. This self-regulating power supply is energized from an ESF 480-volt bus. The change to Action C Will allow us to operate a maximum of one AC INST BUS from the self-regulating power supply for up to 72 hours. This will allow more time to repair a faulty inverter. Note that if the self-regulating power supply were to be lost, Action b. Will still apply.

      -     Delete NoteN because Byron does not need to disconnect the inverters from their DC bus in order to place an equalizing charge on the batteriel; therefore this note is not required.

7) Surveillance Requirement 4.8.4.1.a.1 (pg 3/4 8-17)

Chane "7kV" to "6.9kV" to agree with plant equipment and other points of the Technical Specification Basis.

8) Surveillance Requirements 4.8.4.1.a.2 (pg 3/4 8-18)

Chane the phrase ".. 10% of each type of 480 volt circuit breaker." to "10% of each type of 480-Volt /125-Volt DC circuit breaker." This change is made so that the new sentence agrees with table 3.8-1, 4 (0433M)

                                     . -       ..- __- _ ._. _- -                . - ~ .

F n,

                  "i 3/4.8 ELECTRICAL POWER SYSTEMS

((}((g g {llCOPY 3/4.8.1 A.C. SOURCES

                    .           OPERATING
               ?'             _

LIMITING CONDITION FOR OPERATION i

3. 8.1.1 As a minimum, the following A.C. electrical power sources shall be

                    }          OPERABLE:

physically independent circuits between the offsite network and the Onsite Class 1E Distribution System with:

1) Each system auxiliary transformer energized from an independent transmission circuit, and

2) One of the two transformers forming a system auxiliary trans-former bank.

REflif C C Wi T*// "8 L b. Two separate and independent diesel generators, each with:

1) A separate day tank containir.g a minimum volume of 450 gallons of fuel, L

2) A separate Fuel Oil Storage System containing a minimum volume 7 ,

of 42,000 gallons of fuel, and

 ;                                            3)     A separate fuel transfer pump.

APPLICABILITY: MODES 1, 2, 3, and 4. 1 ACTION: one one-

a. With either en offsite circuit orAdiesel generator of the above required A.C. electrical power sources inoperable, demonstrate the i OPERABILITY of the remaining A.C. sources by performing Specification 4.8.1.1.la or Specifications 4.8.1.1.2a.4) and 6) within 1 hour and at

                            .                 least once per 8 hours thereafter; restore at least two offsite circuits

i. and two diesel generators to OPERABLE status within 72 hours or be in at l least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within

( the following 30 hours.

b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the i OPERABILITY of the remaining A.C. sources by performing Specifications s . 4.8.1.1.la and 4.8.1.1.2a.4) within 1 hour and at least once per og 8 hours thereafter; restore at least one of the inoperable sources A to OPERABLE status within 12 hours or be in at least HOT STANDBY N within the next 6 hours and in COLD SHUTDOWN within the following

        ,                                     30 hours. Restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours from the time of
      +

initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. BYRON - UNIT 1 3/4 8-1 Q

        +                                 .-                 ..- .-       --       -        .. -   -              -

l

l y

/?

Q. Ea e), C/as.,.26 YKV bas cappa Ue o f bei,,, poweyaJ fr.- :

1) E:Me,- &, ,,, f,,.,,,, < ,nJ ;,, y a , , , , , ,; J , , ,,, t ,

norma / Sy < h,,. Ari/;,,, % .,,, % ,, ja,,4 g 3

2) Elfie> l<a ,,< fa ,i>, < ,- ina ht,,, up 64, spy.<;& a,,; /r Sy< h>, t?~ ilia <, L ,, , fa m . s a A t v i , u , u ,,; +

ceou 6ie hr.,/c<a) w ir2

                                       $      hh        $            Y h     h NS h       h ,% $   kfY     Y C<.m a-                       ;,,s y . L.f    tra <m;<<ia. li,, a.

E f { ELECTRICAL POWER SYSTEMS l SURVEILLANCE REQUIREMENTS (Continued) l

4) An impurity level of less than 2 og of insolubles per 100 al when tested in accordance with ASTM-02274-70, analysis shall be completed within 7 days after obtaining the sample but any be performed after the addition of new fuel oil; and

5) The other properties specified in Table 1 of ASTM-0975-1977 and Regulatory Guide 1.137, Revision 1, October 1979, Posi-tion 2.a. , when tested in accordance with ASTM-0975-1977, ,

analysis shall be completed within 14 days after obtaining l

   ..                                              the sagle but may be performed after the addition of new                                                              ,
       '                                           fuel oil.

e. At least once per 18 months, during shutdown, by:

1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's

( recommendations for this class of standby service,

2) Verifying the generator i il reject a load of greater than or equal to 1034 W '% 74 ile maintaining voltage at 4160 1 420 volts and 1 6n 1 4.5 Hz,

   <       (                                 3)    Verifying the diesel generator capability to reject a load of 4
           \                                       5500 W without tripping. The generator voltage shall not exceed 4784 volta during and following the load rejection,

4) Simulating a loss of ESF bus voltage by itself, and:

s a) Verifying de-energization of the ESF busses and load a shedding from the ESF busses, and b) Verifying the diesel starts on the auto-start signal, energizes the ESF busses with pomanently connected loads within 10 seconds, energizes the auto-connected safe shutdown loads through the load sequencing timer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After . o.f the / sed! energizatiorb, the steady-state voltage and frequency of the E5F busses shall be maintained at 4160 + 420 volts ~ and 60 + 4.5 Hz during this test. I o 8YRON - UNIT 1 3/4 8-4

k ELECTRICAL POWER SYSTEMS ( [* A.C. SOURCES ,

 ,.                                 SHUTDOWN                                                                      _

c LIMITING CONDITION FOR OPERATION 6 3. 8.1. 2 As a minimum, the following A.C. electrical power sources shall be g OPERA 8LE: _ ~ . _ _ i One circuit between the offsite transmission network and the Onsite Class 1E Distribution System with: [ [ T

1) A system auxiliary transformer energized by one circuit of the j' 4 offsite transmission network, and,

_ 2) One of the two transformers forming system auxiliary transformer SEPG1cc W int B

                 <v
                                                                                                               ~

b. One diesel generator with:

1) A day tank containing a minimum volume of 450 gallons of fuel, g

4 2) A fuel storage system containing a minimum volume of 42,000 gallons of fuel, and b 3) A fuel transfer pump. APPLICA8ILITY: MODES 5 and 6. N ACTION: f With less than the above minimism required A.C. electrical power sources OPERA 8LE, immediately suspend all operations involving CORE ALTERATIONS, positive reactiv-9 ity changes, movement of irradiated fuel, or crane operation with loads over the spent fuel pool, and within 8 hours, depressurize and vent the Reactor Coolant . System through at least a (2) square inch vent. In addition, when in MODE 5 p with the reactor coolant loops not filled, or in M00E 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiata corrective jl . action to restore the required sources to OPERA 8LE status as soon as possible. SURVEILLANCE REQUIREMENTS y h 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated i 4 OPERA 8LE by the performance of each of the requirements of Specifications l 1 4.8.1.1.1, 4.8.L1.2 (except for Specification 4.8.1.1.2a.5)), and 4.8.1.1.3. r j

          .             3

I g BYROM - UNIT 1 3/4 8-9 1

' - - + + . ,,..-.m-._,_, ,y. , . .,, , _ _

1 8 t R. One Class dE YKV ba capa ble o f be;,,9 power ed from :

1) E; fla, 6 rantfe,m er in a ja;,,, up LL o c.cs e;a /aj u niLa c,,An Avilia e, knefora er han k ciz .

a)Elffav 6 vane form er me ki., up 6 L cppost/. u n; /< Sp < km A'uxi/tary % < form er han4 ( via the u,,;F cross- 6;e b recs /e err ) w a rol 74e Syr f.,,, Aux;/;e ry %,,e form er lan /c supplir.y 6he y

'l Kv hus          e nergi ead from       an  ofl<ile dra ~ <~irhu ed Bus. lll. emet; en/. from its                                                    655oc;6ied  inveties.

_. _. .. c m ed,) to Dc But lli . 2.3xsttwwed Bus._ ll1 exerg:c,[S.s its asroc:ded inuette&
. _ . _ .ceweded } DC Dus lu . . . _ _3. ~f.wkwed Bus .113 ederg;ced how ilt Assoc'.a}cd _invettet
- . , . - - - co m ech ol 40 .DC- Dus. \ \ \. .. . - ._ _ .. 4.3rbked bs lW ekep;e<d $s its Assoc;4}<d lauertek con c),J Jo Dc But 11 2 .
/ flachwest @
c. .. ktl 3 say; muted N A.c.hverlev inoperutte .F s.t <ew <de/
. . .4* ib .
D.C ysweV Sugly, its sssoc:ded A.c. bs>wea \>u.s usy . le. f*w<&cJ ft*H . ifs rejultliry }vasbFuev powek .svgly ? F ug.
.. . .. W N NV51 65kebwise be. in si eui SOT STAND 8t V:Y;n Ike 3ed 6. kauv5. p d.in Cbt.0 5H4706W M w:K:w M f,j(.w:6g . .. -. Wkours s-+- Jdp. _,
P.3@ww e,-- .$ .M g s gb #p- m.-d y, I "7m -" ,w, e4 -'
,.ii9 . p .,
_ . _ . _ _ . . . _ _ . . _ . _ . . . - . . _ ~ . _ . _ . . . . t
!-MEWCOPY
{ ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES _ _ _ _ CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES _ LIMITING CONDITION FOR OPERATION 3.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be OPERA 8LE. APPLICABILITY: MODES 1, 2, 3, and 4. ___ ACTION: , With one or more of the above required containment penetration conductor overcurrent protective device (s) inoperable:
a. Restore the protective device (s) to OPERABLE status or de-energize the circuit (s) by tripping the associated circuit breaker or racking out or removing the inoperable circuit breaker within 72 hours, declare the affected system or component inoperable, and verify the circuit breaker to be tripped or the inoperable circuit breaker
racked out, or removed, at least once per 7 days thereafter; the
{ provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their circuit breakers tripped, their
~-
inoperable circuit breakers racked out, or removed, or
b. Be in at least HOT STAN08Y within the next 6 hours and in COLD SHUTOOWN within the following 30 bcurs.
SURVEILLANCE REQUIREMENTS 4.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be demonstrated OPERABLE:
a. At least once per 18 months:
69
1) By verifying that the )/ kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of the circuit breakers, and performing the following:
a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed and as specified in Table 3.8-1, and BYRON - UNIT 1 3/4 8-17
ELECTRICAL POWER SYSTEMS PR00F & HEW COPY SURVEILLANCE REQUIREMENTS (Continued) c) For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested. W& DC
2) By selecting and functionally testing a epresentative sample i of at least 10% of ecch type of 480-vol circuit breaker. !
Circuit breakers selected for functional testing shall be ; selected on a rotating basis. The functional test shall 1 consist of injecting a current input at the specified Setpoint to each selected circuit breaker and verifying that each circuit breaker functions as designed and the response time is less than or equal to the specified value. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be func-tionally tested until no more failures are found or all circuit breakers of that type have been functionally tested; and
3) By selecting and functionally testing a representative sample of each type of fuse on a rotating basis. Each representative sample of fuses shall include at least 10% of all fuses of that type. The functional test shall consist of a nondestructive resistance measurement test which demonstrates that the fuse meets its manufacturer's design criteria. Fuses found inoper-able during these functional tests shall be replaced with OPERABLE fuses prior to resuming operation. For each fuse found inoperable during these functional tests, an additional representative sample of at least 10% of all fuses of that type shall be functionally tested until no more failures are found or all fuses of that type have been functionally tested.

b. At least once per 60 months by subjecting each 7 kV circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.
BYRON - UNIT 1 3/4 8-18
O o RAOI0 ACTIVE EFFI,UORS PRODF & REVIEW COPY
* ~
n,f 3/4.11.2 GASEQUS EFFLUENTS 005 h LIMITING CONDITION FOR OPERATION Q . 3.11.2.1 The dose h due to radioactive antarials released in gaseous a effluents from the site to areas at and beyond the SITE BOUNGARY (see
.: Figure 5.1-1) shall be Ifmited to the following:
,f0 a. For noble gases: Less then or equal to 500 arems/yr to the whole body and less than or equal to 3000 aress/yr to the skin, and ' b. For Iodine-131 and 133, for tritium, and for all radionuclides in particulata form with half-lives greater than 8 days: Less than or equal to 1500 arems/yr to any organ.
.C.O APPLICAOILITY: At all times.
M: m
/
A. With the dese$cete(s) exceeding the above limits, immediately restore the l Cy.) . j release rata'tolithin the above Ifmit(s). b, W gro, 3,o, b d 3.o,q. % M 9:~oA,V 0 4 39% SURVEILL*** REQUIREMENTS 3 4.11.2.1.1 The desehdue to noble gases in gaseous effluents shall be determined to be within'the above limits in accordance with the methodology
, and parameters in the 00CM.
j.@ 4.11.2.1.1 The dose due to Iodine-131 and 133, tritium, and all radio-nuclidee in particulate form with half-lives greater than 8 days in gaseous effluents shall be detamined to W within the above limits in accordance with the methodology and parameters in the 00CM by obtaining representative samples and performing. analyses in accordance with the sampling and analysis program specified in Table 4.11-2.
?
l-0
BYRON - UNIT 1 3/4 11-9
. .. w . _ . .,9 . . . . . . . . . . . . . . . ..m -- . -. ~~ '^ -. .Z7.-'Z:J_'Z Z~ _, _ _ ._L.. .
__.____.___1._1_._._~____Z_.1__ _ . , _ _
ATTACHMENT K (Section 3/4.9)
1) Section 3.9.1 (pg 3/4 9-1) Refueling Operations.
Deleted "M" and Note at bottom of page because note doesn't say anything
'that Mode 6 doesn't say. " Filled portion" has bee, changed to "unisolated portions" for clarification.
2) Section 4.9.1.3 (pg 3/4 9-1) Refueling Operations.
In section 4.9.1.3 valves ICV 8455,1CV8464A, IPWO46, and 2PWO46 have been changed to ICV 8430,1CV8441, ICV 8435, and 1CV8439 respectively. This was done to be consistent with Byron FSAR Amendment 41 Page 15.4-24.
3) Section 3.9.4 (pg 3/4 9-4) Refueling Operation In section 3.9.4.c.2 "Be" was deleted and " Capable" was made to be the first word of sentence.
Section 3.9.4.a was clarified to distinguish between the personnel hatch and equipment batch.
4) Section 3.9.9/4.9.9 (pg 3/4 9-10) Refueling Operations.
Changed all phrases stating "Containmes.t Ventilation System" to
" Containment Purge Isolation System" to reflect the way it is referred to in bases and title page.
Sect ion 3.9.12 (pg 3/4 9-13) Refueling Operations Section 3.9.12 has been rewritten and an asterisk added to clearly identify the conditions for exhaust ventilation operation. In Action: a) the words " discharging through" were deleted and in its place the words "taking suction from" were inserted because the system operates that way. (0433M)
m o , o k ' 2 i PR00F& REflEW copy
9 3/4.9 REFUEU NG OPERATIONS 3/4.9.100 ROM CONCDmtATION LDtITING COMOITION FOR OPERATION
^ r. unhohUd I 3.9.1 The heron concentration of all MHed portions of the Reactor Coolant System and the refueling canal shall be asintained unifore and sufficient to ensure that the more restrictive of the following reactivity conditions is i est,either: h a. A K,ff of 0.98 or less, or ,
b. A boron concentration of greater than or equal to 2000 ppe.
APPUCMIUTY: MODE [ , 4.s ' i M: With the requirements of the above specification not satisfied, imediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continus beration at greater then or equal to 30 ps l of a salution containing greater than or equal to 7000 pas baron or its equiv-g- alent until K g is reduced to less then or aquel to 0.98 or the boron consentretien is restored to greater than or equal to 2000 ppe, whichever is ' the more restrictive. I y SURVIILL*M R8MIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to: !. a. Removing or unbelting the reacter vessel head, and
b. Withdressi of any ne1* L.,,- control rod in excess of 3 feet from its fully inserted position within the reactor vessel.
4.9.1.2 The beren concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours. C tcv845o, Icv 844: ,Tc.y s435 e41<.ve+39
4. 9.1. 3 Valves -nGWO405;-98YG4644 7 .CT^^. 2 225 shall be verified closed i and secured in position by anchanical stops or by removal of air or electrical power at least once per 31 days.
n "me reacter snall no enintained in M00E 6 whenever fuel is in the reactor U- vessel with the vessel head closure bolts less than fully tensioned or with the head removaC 6j l j j SYNON - UNIT 1 3/4 9-1 G- . . . - - . - -- - - . . .
REFUELING OPERATIONS P,%ya gg gg t 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
a. The personnel hatch closed and held in place by a minimum of four (cNacc polts or removed, yk

b. A minimum of one door in the personnel emergency exit hatch is closed, and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1) Closed by an isolation valve, blind flange, or manual valve, or

2) dgapableofbeingclosedbyanOPERABLEautomaticcontainment 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 or movement of irradiated fuel in the containment building. ,
.s SURVEILLANCE REQUIREMENTS 4.9.4 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 OPERA 8LE automatic containment purge isolation valve within 100 hours 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:
a. Verifying the penetrations are in their closed / isolated condition, or

b. Testing the containment purge isolation valves per the applicable portions of Specification 4.6.3.2.
m .- m N a - w mia,a o u p u 4 GWJhDdCrW Aerd kL.M M& BYR0K - UNIT I 3/4 9-4
n 1%
" REFUELING OPERATIONS i y., -..
3/4.9.9 CONTAINMENT PRD0F& RFliEW CDPy Q. - LIMITING CONDITION FOR OPERATION 90R6E Tsowrio 4 3.9.9 The Containment %.;11.tM,ystem shall be OPERA 8LE. - k ?PPLICA8ILITY: During CORE ALTERATIONS or movement of irradiated fuel within
~. t containment.
ACTIL4: (bhE TSo'9Tto4
a. With the Containment L .;1 h M d 5ystem inoperable, close each of I-0 . the purge valves providing direct access from the containocnt atmosphere to the outside atmosphere.
~
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

r. .
SURVEILLANCE REQUIREMENTS
' A>6 E Isot.RTsoeJ 'C 4.9.9 The Containment m. iiniorf5ystem shall be demonstrated OPERA 8LE ~'
within 100 hours prior to the start of and at lesst once per 7 days during CORE ALTERATIONS by verifying that containment purge isolation occurs on l . .- ..n n_huar Initiation and on a ESF test signal from each of ths.~ containment -- - - - . - - - ' l radiation monitoring instrumentation channels. e. m
.d 'N BYRON - UNIT 1 3/4 F 10 d
l* ... .. . , ..
. . .. s;= - . .* 7 =;. .,9_ , p ..w .. . , , ,
E00F & REVIEW COPY { REFUELING OPERATIONS 3/4.9.H FUEL HANDLING BUILDING EXHAUST VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9. u Two independent Fuel Handling Suilding Exhaust Ventilation Systems shall be OPERA 8LE. APPLICABILITY: Whenever irradiated fuel is in the storage pooler da.g Aish((, ;
.Am A. o rea.sel .< L it s pece f b /44 s hw Am u ,.,+, .,q,q = 4 L&Ln ,c.....,
w .,.. a u w L c..d.%ed%
a. With one Fuel Handling ding Exhaust Ventilation System inoperable, fuel movement within the torage pool or craner operation with loads over the storage pool may proceed provided the OPERA 8LE Fuel Handling Suilding Exhaust Ventilation System is capable of being powered from an OPERA 8LE emergency power source and is in operation and C w 4through at least one train of HEPA filters and charcoal adsorters.
*Ls eu h Ge* c . 4.a .I Y ,<
b. With no Fuel Handling Suilding Exhaust ventilation System OP LE, suspend all operations involving movement of fuel within the storage pool or crane operation with. loads over the storage pool until at least one Fuel Handling Building Exhaust ventilation System is restored to OPERABLE status.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. . .

2. . ; .'. . .*.:. _ . , __. . - .
$URVEILLANCE REQUIREMENTS I '
4.9.u The above required Fuel Handling Suilding Exhaust Ventilation Systems shall be demonstrated OPERA 8LE:
a. At least once per 31 days on a STAGGCRED TEST BA515 by initiating,
, from the control room, flow through the HEPA filters and charcoal ; adsofters and verifying that the system operatas for at least L 15 minutes;
b. At least once per la months, or (1) after any structural maintenance ,
on the NEPA filter or charcoal adsorter housings, or (2) following ' t painting, fire, or chemical release in any ventilation zone ' caemunicating with the system, by: 4 wLes 4 .p:p d kd.L o re nel .< uti c l.gl I,7 leu Re. b , l l h.th ,, As~ e y n n ,el k&L n op k s G.at utL.3 8YRON - UNIT 1 3/4 9-13
- . . . . . - . . __ . - - _ v_ a_
ATTACHMENT L (Section 3/4.10) Circled items noted in this attachment have previously been submitted.
1) Section 3.10.2 (pg 3/4 10-2) Special Test Exceptions In section 3.10.2 and Action 3.10.2; add "[Fq (z)]" af ter "3.2.2" and
"[RCS Flow Rate]" after "3.2.3".
In Action 3.10.2 add "(Rod Height)" after "3.1.3.1", "(RIL)" af ter "3.1.3.5" and "3.1.3.6", "( AI)" af ter "3.2.1" and "QPTR" af ter "3.2.4". In 4.10.2.2.a & b, add "(Fxy)" after 4.2.2.2, "[Fq (z)]" after 4.2.2.3 and "[RCS Flow Rate]" af ter 4.2.3.2. The changes were made to be more informative in relation to what the specification number is referring to.
2) Section 3.10.4 (pg 3/4 10-4) Special Test Exceptions Add " Reactor Thermal power does not exceed 10% of rated thermal power." to 3.10.4.a.
- Add "For testing under no flow conditions. (10% thermal power)" to APPLICABILITY: in section 3.10.4 Add "(10% thermal)" to sections 4.10.4.1 and 4.10.4.2.
To be more informative about subject matter and to add clarification.
3) Section 3.10.5 (pg 3/4 10-5) Special Test Exceptions Delete " full-length" in section 3.10.5. All rods are the same length at Byron Station.
I, . Pil00F & REVH COPY ( SPECIAL TEST E)(CEPTIONS . 3/4.10.2 GROUP HEIGHT. INSERTION. AND POWER OISTRIBUTION LIMITS l LIMITING CON 0! TION FOR OPEPATION 3.10.2 The group height, insertion, and pou r distribution Ifmits of Specifications 3.1.3.1, 3.1.3.5, 3.1:3.6. 3.2.1, and 3.2.4 may be suspended during the performance of PNYSICS TESTS provided:
a. The THE W L POWEA is maintained less than or equal to 855 of RATED M
L ** IR* *** )
'U b. rne limits of specifications 3.2.2kg and 3.2.3r7 are{intainedRCSflowR as and 4steinined at the frequencies spe:1fied in Specification 4.10.2.2, below. AprLICASILIT_Y: ,'MOCE 1., {fg (t}}
^
t g .,- , , RCSThefa+e) Witn'any of 'the Ifmits of Specifiestiono 3.2. 3.2.3 ing exceeded while the requirements of Specifications and 3.2.4(4PTR) , are swi, either-3.1.3.) ( N e d #e334f) " (3.1.3.p RIL) 3.1.3 (#h7 (d J. 3.2. ( a. ' Reduce THEWL POWEA sufficient ta satisfy the ACTION requirements
- . - of Specifications 3.2. 3. 2. er .

b. Se in fit wi In 8 hours.- *#"}
SURVE1LLNeCE REQUIRD O T5 s __ l
. . . _ . . _ . _ . ,s ., .~ . . _ _ . _ _ . . ~"--- ~ ' ~~~ ~'~ ~ E16. 2.~ 1 The THEML POWEA shall be oe+sreined to be less than or equal 13--"
855 of RATED THERMAL POWER at lost ex4 per hour during PHYSICS TESTS. 4.10.2.2 The requirement.1 of.the 641'ow Ifsted se x fications'shall be performed at least once per 12 hours during MfYSICS TESTS: . [ f 8 (s)
~
a. Specif tentions 4.2.2.2 end 4.2.t.s& , and
~
g
b. '
Spec'1fication 4.2.3.2([ w e.xcms : BYRON - UNIT 1 . .3/4 10-2
( SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance ord Ti,^7JT :-QPHYSICS TESTS provided:
a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and 8eecter T)p ertnet fewer doet het 62994 /* % */ rated thermer power.

b. The Reactor Trip Setpoints on the OPERA 8LE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.
APPLICA8ILITY: During operation below the P-7 Interlock Setpointy for f e s ting L4nder no flou conofHion t. (fc % +), gym f ppyg,) ACTION: With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the Reactor trip breakers. SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall t:e datermined to be less than P-7 Interlock Setpoint at least once per hour duringE"""JP @ PHYSICS TESTS. (/#$ -f*/;gfg) 4.10.4.2 Each Intermediate and Power Range channel, and P-7 Interlock shall be subjected ta an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiatingQi^7Jr :-jPHYSICS TESTS. (/pg 1Xpen./) i SYRON - UNIT 1 3/4 10-4
4 SPECIAL TEST EXCEPTIONS d I 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of SpecificationL3.1.3.3 may be suspended during the performance of individual t" :- 7 shutdown and control rod drop time measurements provided;
a. Only one shutdown or control bank is withdrawn from the fully inserted position at a time, and

b. The digital rod position indicator is OPERABLE during the withdrawal of the rods.*
APPLICABILITY: NODES 3, 4, and 5 during performance of rod drop time measurements. -
ACTION: With the Position Indication System inoperable 'or with more than one bank,of rods withdrawn, immediately open the Reactor _ trip breakers. I SURVEILLANCE REQUIREMENTS'
'7' I
4.10.5 The above required Position I ication Systems shall be determined to be OPERABLE within 24 hours prior to 4he start and at least once per 24 hours thereafter during rod drop t me measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree:
a. Within 12 steps when the rods are stationary, and

b. Within 24 steps during rod motion.
i "This requirement is not applicable during the initial calibration of the Position Indication System provided: (1) K,ff is maintained less than or equal to 0.95, and (2) only one shutdown o'r control rod bank is withdrawn from the fully inserted position at one time. BYRON - UNIT 1 3/4 10-5
. - - . . . . , . . ~, . . - - - - -- , - , ,- , - - - . , - . - , r , , . , - - , . .- ._-
. ATTACHMENT M (Section 3/4.11)
Circled items noted in this attachment have been previously submitted.
1) Section 3.11.1.1 (pg 3/4 11-1) Radioactive Effluents A second Action Statement has been added stating Specifications 3.0.3 and 3.0.4 are not applicable. The cause of the radioactive material released to unrestricted areas should limit unit operation.

2) Section 3.11.2.1 (pg 3/4 11-9) Radioactive Effluents A second Action Statement has been added stating Specifications 3.0.3 and 3.0.4 are not applicable. The cause of the_ radioactive material released to unrestricted areas should limit unit operation.
(0433M)
l l > Pn00F & REVIEW COPY 3/4. n RADI0 ACTIVE EFFLUENTS 3/4.H.1 LIQUID EFFLUENTS CONCENTRATION Q; LIMITING CONDITION FOR OPERATION
- 3.11.L1 The concentration of radioactive antarial released in liquid effluents to (MtESTRICTED AltEA5 (see Figure 5.1-1) shall be Ifmited to the concentrations q". . specified in 10 CFit Part 20, Appendix B, Table II, Colums 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be Ifmited to 2 x 10
sicrocurie/a1 total activity.
APPLICA8ILITY: At all times. C ACTION:
a. With the concentration of radioactive antarial released in liquid effluents to UNRESTRICTED AltEA5 exceeding the above limits, immediately restore the con-s centration to within the above limits.
G . b, Th 4 gelg;ygamp 3,o,3 %d 3,o,q m 4 mh_ A, x SURVEILLANCE REQUIREMENTS 4.1L L L 1 Radioactive liquid westas shall be sampled and analyzed according
.? to the sampling and analysis program of Table 4.u-L
m. .
4.1L L L 2 The results of the radioactivity analysis shall be used in accordance with the methodology and parameters in the 00CM to assure that the concentrations at the point- af release are maintained within the limits of Specification 3.1L L L C v-L s BYRON - UNIT 1 3/4 E-1 y . . - - . .
a b
- D RAOI0 ACTIVE EFFLUENTS PRDOF & REVIEW CDPY * ~
n.f-3/4. u.2 GASEOUS EFFLUENTS 00Sh LIMITING CONDITION FOR OPERATION e._ 1 3.u.2.1 The dose h due to radioactive materials released in gaseous L effluents from the s1ta to areas at and beyond the SITE SOUNDARY (see Figure 5.1-1) shall be limitad to the following:
,m
a. For noble gases: Less than or equal to 500 arems/yr to the whole body and less than or equal to 3000 arems/yr to the skin, and

b. For Iodine-131 and 133, for tritia, and for all radionuclides in particulate fom with half-lives greater than 8 days: Less than
, or equal to 1500 arems/yr to any organ. 'O APPt.ICA8ILITY: At all times.
ACTION:
/
q A. With the dosescetofsj exceeding the above limits, immediately restore the
C, _j . release rats to iifthin the above limit (s).
b, Ikg. p ,l ,, Q 4 9 ,l4 ; , h 3 ,o ,3 _ f 3 ,o ,q. m g c,og,y p SURVEILLANCE REQUIREMENTS 1
4. u.2.1.1 The dose hdue to noble gases in gaseous effluents shall be detamined to be with n the above limits in accordance with the methodology and parameters in the 00CM.
G 4. u.2.1.2 The dose due to Iodine-131 and 133, tritium, and all radio-nuclides in particulate fom with half-lives greater than 8 days in gaseous effluents shall be determined to lte within the above limits in accordance with the methodology and parameters in the ODCM by obtaining rep nsentative samples and perfoming. analyses in accordance with the sampling and analysis program specified in Table 4.u-2. G
^ .3 J .
BYRON - UNIT 1 3/4 u-9 N - . . - - . .-s.. . - -
,,. ; ,. ,9,,,, y j . . . . . . - - . . ~ . . - - - . . .~.~ . -
i
. , ___ . .= .. - - _ _ _ _ - ._- . _ . . .
ATTACHMENT N (Section 3/4.12) 10 Table.4.12-1 (Page 3/4 12-11) Table Notations For clarification, "(s-1)" has been changed to "(sec-1)" and "(s)" has been changed to "(sec)". l
/.
t-f. (: [. r h i. (0433M) L i
P.10DJ & REYEW COPY I TABLE 4.12-1 (Continued) TABLE 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 nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6.7.1.6. (2) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13. ' (3) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a. sample that will yeild 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 particuler measurement system, which may include radiochemical separation: 4.66 s b E V 2.22 Y exp (-Aat) ( Where: LLD = the "a priori" lower limit of detection (picoCuries per unit g mass or volume), s = 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 = the number of disintegrations per minute per picocurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide, I (g),and at = the elapsed time between sample collection, or end of the sample collection period, and time of counting ( gg Typical values of E, V, Y, and at should be used in the calculation. BYRON - UN,IT 1 3/4 12-11
ATTACHMENT 0 (Section Bases)
1) Bases 3/4.1.2 (pg B 3/4 1-3) Borat'.on Systems Add attachment "BB" which was taken from previous Refueling Water Storage Tank bases section which has been deleted.

2) Bases 3/4.5.4 (pg B 3/4 5-2) Refueling Water Storage Tank This is being deleted and information added into the Boration System Bases on pg B 3/4 1-3.

3) Bases 3/4.7.6 (pg 8 3/4 7-4) Plant System.
Changed all phrases " Control Room Emeng>ncy Air Cleanup" to " Control Room Ventilation Cleanup" to give proper identification of syctem.
4) Bases 3/4.9.9 (pg B 3/4 9-2) Refueling Operations The word " ventilation" was deleted and the words " Purge Isolation" was added to give proper description of System. ~
(0433M)
\ ,,,c m ,cq,,,g,sy,,,,, Pi!0DF & lHEW COPY 8ASES 80 RATION SYSTEMS (Continued)
With the ACS temperature below 200*F, one Borer. Injection System is e acceptable without single failure consideration on tne basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Baron Injection System baramma inoperable. The limitation for a enximum of one contrifugal charging pump to be l OPERA 8LE and the Surveillance Requirement ta verify all charging psamps except the required CPERABLE pump to be inoperable below 350*F provides assurance tdat a mess addition pressure transient can be relieved by the operation of a single PORV. - The baron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIM of 1% Ak/k after xenon decay and cooldown from 200*F to 140*F. This condition requires either 2,652 gallons of 7000 ppe borated water from the boric acid storage tanks or 11,840 gallons of 2000-ppe barated water from the refueling water storage tank (RWST). ~C 0010-The contained wetar volume limits include allomence for water not available igu ! because of discharge line location and other physical characteristics. The limits on contained water volume and boron concentration of the RWST
, also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH bano minimizes the evolution of , iodine and einimizes the effect of chloride and caustic stress corrosion on anchanical systems and components.
5 i The OPERA 8ILITY of one Soron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6. s 3/4.1.3 MOVA8Lf CONTROL A550eLIES The specifications of this section ensure that: (1) acceptable power . distribution limits are meintained, (2) the sinimum SHUTDOWN MARGIN is emin-tained, and (3) the potential effects of rod misalignment on associated-accident analyses are limited. OPERA 8ILITY of the control rod position l indicators is required to determine control rod positions and thereby ensure l compliance with the control rod alignment and insertion limits. The Digital ! Position Indication System does not indicate the actual position of the shutdown rods between 18 staps and 210 steps withdrawn. The ACTION statements which permit limitad variations from the basic requirements are accompanied by additional restrictions which ensure that the l original design criteria are set. Misalignment of a rod requires seasurement '. of peaking factors arm a restriction in THERMAL POWER. These restrictions j provide assurance of fuel rod integrity during continued operation. In addition,
. these s'afety analyses affected by a sisaligned rod are reevaluated to confirm that the results remain valid during future operation.
SYRON - UNIT 1 8 3/4 1-3 s 3.-
@g M W% % S sl4 l-5 The OPERA 8ILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for infection by the ECCS in the event of a LOCA. The limits on RWST sinimum volume and baron concentration ensure tNat: (1) sufficient water is available within containment to pomit recirculaticn cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following afxing of the RWST and the RCS water voltaes with all control rods inserted except for the most reactive control assembly. These assumptions are consistant with the LOCA analyses. , ,.g, . . = - * * * ' " * *
,,r- -
-- - - - , , , ~- ,, - - - - - e,,,., p._n __,,y e,py _,y- r --_w r,- , - -- -,,,- e -- -,,- --- g. - , - ,y
l EMERGENCY CCRE COOLING SYSTEMS , , h hh hf J BASES ECCS SUB5YSTEMS (Continued) The limitation for a maxieum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pimps and Safety Injection pumps except the required CPERA8LE charging pump to be inoper-able below 350*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. The Surveillance Requirements provided to ensure OPERA 8ILITY of each 3. component ensures that at a minimum, the assumptions used in the safety analyses
>- are met and that sutsystem CPERA8ILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary 'to: (1) prevent total pump flow free exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with
the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptacle level of total ECC5 flow to all injection points equal to or above that assumed in the ECC5-LOCA analyses. The Surveillance Requirements for leakage testing of ECC5 check valves ensures that a failure of one valve will not cause an intarsystem LOCA.
. M .5.4 REFUELING WATER STORAGE TANK k( part of the The ILITY of the refuel hiter rat'ad t a sufficient supply of-sta water is a%1e for inj ion ECCS ensures by the ECCS in event of a LOCA. TJe its on RW inimum 'vetiume arid oron s , ratiot ens 'at: (1) suff,icient wa i vailable within p inment to per9e irculation ing f3oW to the core ' (2) the reactorAfiT(remain suberitica n the cold co tion followi sixing $f the RW5TAffd the RCS water /
volumes with a contr914cds apt for thn@st,rt' active controh / assembly. These tions are istent with the QlM analyses. The atained wate o limit inc es an' allowance for water not e use of tank rge line locati er physi .cfiaracteristics. N f' The imits contained s voljsse and baron c entration af'the RWST also ensure value of between 8, Land 11.0 for the so ion recirculated within co inment after a LOCA. is band mini zes the lution of \ iodi of chlori tic stress co fon on
minimizes % e off 3
ical systans and nonts. BYROM - UNIT 1 8 3/4 5-2
~ ~ = = . . . . - . - . . J-
i i- 8 PLANT SYST9tS
)
8ASES ULTIMATE HEAT SINK (Continued) 1 The limitations on minimum water level and maximum temperature are based on providing a 30-day cooling water supply to safety related equipment without exceeding its design basis temperature and is consistant with the recommend-ations of Regulatory Guide 1.27, " Ultimata Heat Sink for Nuclear Plants," March 1974.
. M 3/4.7.6 CONTROL ROOM =0"C" "I" C'i" .[ SYSTEM vh The OPERA 8ILITY of the Control Room " . c..,.i i,. Ch x_ &,.ystemS ensures that: (1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by
).- thi.s system, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and itEPA filters. The OPERA 81LITY of this system in conjunction with control room design provisions is based on limiting the radiation exposLre to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Critarion 19 of Appendix A, 10 CFR Part 50. ANSI M510-1975 will be used as a procedural guide for surveillance testing. 3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FILTER PLENUM VENTILATION SYSTEM
The OPERASILITY of the Non-Accessible Area Exhaust Filter Plenum Ventilation System ensures that radioactive materials leaking from the ECCS equipment within the pump rooms following a LOCA are filtered prior to reaching the environment.
The operation of this system and the resultant effect on offsita dosage calcu-lations was assumed in the safety analyses. ANSI M510-1975 will be used as a procedural guide for surveillance testing. ) 3/4.7.8 SNU88ERS All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is main-tained during and following a seismic or other event initiating dynamic loads.
. Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on any safety-related l
system. Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of ) the 2-kip,10-kip, and 100-kip capacity manufactured by Company "A" are of the BYRON - UNIT 1 8 3/4 7-4
D l I l REFUELING OPERATIONS
' PROOF & UN COM D 8ASES l 3/4.9.6 REFUELING MACHINE The OPERA 8ILITY requirements for the refueling machine and auxiliary I hoist ensure that: (1) manipulator cranes will be used for movement of drive rods and fuel assemblies, (2) eacn crane has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
g. 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool areas ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in a O.' - single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistant with the activity release assumed in the safety analyses. 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION b - The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140*F as required during the REFUELING N00E, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a baron dilution incident and prevent boron stratification. The requirement to have two RHR loops OPERA 8LE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RNR loop will not result in a complete loss of RHR capability. With the reactor vessel head removed and at least 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, ) in the event of a failure of the operating RHR loop, adequata time is provided to initiate emergency procedures to cool the core. 3/4.9.9 CONTAIPMENT P.g.1...L..SYSTEM .Lf".p y The OPERABILITY of this system ensures that the containment purge r penetrations will be automatically isolated upon detection of high radiation L levels within the containment. The OPERA 8ILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment. 4 0 BYRON - UNIT 1 8 3/4 9-2 ) .
=.i-..-w-- n-------ww w -y,re-ww----w,.,,ww w w w- w ee- ,-.m-vee +w-ew -ww---t.--mm--v----we--+ee=grwme.---r we =.a e*wwen-, -e-*- - + + ---v w- --'
ATTACHMENT P (Section 5.0)
1) Section 5.6.2 (pg 5-5) Fuel Storage Deleted "This is-1 foot . . . . 423 feet 10 inches." on section 5.6.2.
The. normal pool water level will be between 423' 6" (low alarm) and 424' 1 1/2" (high alarm). We cannot specify exact normal level at this time. (0433M)
's PROOF & REMEW COPY DESIGN FEATURES ( 5.6 FUEL STORAGE CRITICALITY
5. 6.1.1 The spent fuel storage racks are designed and shall be maintained
.- with: ,
a. A k,ff equivalent to less than or equal to 0.95 when flooded with 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 contar-to-center distance between fuel assemblies placed in the storage racks.

5. 6.1. 2 The k,ff for new fuel for the first core loading stored dry in the
, r, spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assused.
ORAINAGE . 5.6.2 The spent fuel storage pool is designed and shall be maintained to ( prevent inadvertent draining of the pool below elevation 423 feet 2 inches. i;at 0:
' - 1 .^m :.474. 5:'- '" :: . _ _1 , 21 CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a O storage capacity limited to no more than 1050 fuel assemblies.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be asintained within the cyclic or transient limits of Table 5.7-1. Ia BYRON - UNIT 1 5-5
ATTACHNENT Q (Index)
1) (pg IX)
Section 3/4.5.4 has been deleted from text.
2) (pg X)
Section 3/4.7.6 has been changed from " Control Room Emergency Air Cleanup System" to " Control Room Ventilation System" to give proper.name to the system.
3) (pg XII)
Section 3/4.9.9 has been changed from " Containment Ventilation System" to
" Containment Purge Isolation System" to give proper name to the system.
4) (pg XVI)
Section 3/4.5.4 has been deleted from text. (0433M)
O-
& L S;N .j~
O LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-1 MAXIMim ALLOWA8LE POWG RANGE NEUTRON , Q. FLUX HIGH SETPOINT WITH IN0PERA8LE STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION........................................... 3/4 7-2 TABLE 3.7-2 (8 LANK)............................................... 3/4 7-2 TABLE 3.7-3 STEAM LINE SAFETY VALVES PER L00P. . . . . . . . . . . . . . . . . . . . . 3/4 7-3
g Auxiliary Feedwater Systas............................... 3/4 7-4 Condensata Stora0e Tank. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-6 Speci fi c Acti vi ty. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY
" 5AfftE AND ANALYSIS PR0 GRAM......................... 3/4 7-8 Main Steam Line Isolation Va1ves. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-9 3/4.7.2 ST U M GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-10 3/4.7.3 C2 90NENT COOLING WAT B 5YSTEM........................... 3/4 7-11 s.
C- -
) 3/4.7.4 ESSENTIAL SERVICE WATER SY5 TEM........................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT 5 INK....................................... 3/4 7-13 ve nanm n 3/4.7.6 CONTROL ROOM hi AIR CU:.%ySYSTEM................ 3/4 7-15 3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FIL1Pt PLENUM
!C VENTILATION 5YSTEM....................................... 3/4 7-18 3/4.7.8 SMU88ERS................................................. 3/4 7-20 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNU88ER FUNCTIONAL TEST........... 3/4 7-25 3/4.7.9 SEALED SOURCE C0NTANINATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-26 a-
,J es s
8YRON - UNIT 1 X
-sge . oo J . * * * * ** , ,y f .- , =%'** ,, - y .p g 4
,n .v e
T INDEX PRDOF & REY 3 0071
C r-LIMITING CONDITIONS FOR OPERATION t'4 SURVEILLANCE REQUIREMENTS SLCHQ!f . PAGE
;g Motor-Operated Valves Thermal-0verload Protection 0 and 8ypass 0evices..................................... 3/4 8-34 TABLE 3.8-2 NOTOR OPERATED VALVES THERMAL OVERL0A0 PROTECTION.AND/OR SYPA55 0EVICES.................... 3/4 8-35 C 3/4.9 REFUELING OPERATTONS 3/4.9.1 80RON CONCENTRATION...................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION.......................................... 3/4 9-2 n 3/4.9.3 DECAY TIME............................................... 3/4 9-3 c 3/4.9.4 CONTAD00ENT SUILDING PENETRATIONS. . . . . . . . . . . . . . . . . . . . . . . . 3/49-4 3/4.9.5 C0MUNICATIONS........................................... 3/4 9-5 3/4.9.6 REFUELING MACNINE........................................ 3/4 9-6
.<, s. 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY............... 3/4 9-7 CJ ._ 3/4.9.8 , RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION , High Water Lave 1......................................... 3/4 9-8 LawWaterLge,1,.d*&*****o*4***************************** 3/# '~I '^ 3/4.9.9 CONTADSIENT OT6LATIG SY5 TEM........................... 3/4 9-10
'O 3/4.9.10 WATER LEVEL - REACTOR VE55EL. < . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-11 3/4. 9.11 WATER LEVEL - STORAGE P00L. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-12 3/4.9.12 FUEL HANDLING BUILIDNG EXHAUST VENTILATION SYSTEM........ 3/4 9-13 l
l ! O.,, i i I { O
}.. '
8YRON - UNIT 1 XII
~.
Ye
. ..= . . . , , .y.m - =- - . , , , , , , , - , c. u ,, 4 . y
_ - . . ~ _ . _ . . . . . - - - - .
l. . .
1,,,x PRC7 & RM COPY LIMITING CONOITIONS FOR OPERATICN ANO SURVEILLANCE REQUIREMENTS SECTION
PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATOR $............................................. 3/4 5-1 I
3/4.5.2 ECCS SUS $YSTEMS - T avg ,3350* F. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-3 3/4.5.3 ECCS SU8MSTEMS - T avg
< 350*F........................... 3/4 5-7 n ,. . ,- .=~..-_e-_-m...m..... = . , ,
ug g_3o n 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAIMENT Containment Integrity.................................... 3/4 6-1 Containment Laakage...................................... 3/4 6-2 Containment Air Locks.................................... 3/4 6-4 Internal Pressure........................................ 3/4 6-6 Air Tamparature.......................................... 3/4 6-7 Containment Vessel Structural Integrity.................. . 3/4 6-S , Containment Ventilation System........................... 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray Systas................................. 3/4 6-13 Spray Addi ti ve Systas. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-14 Containment Cooling Systam............................... 3/4 6-15 3/4.6.3 CONTAINMENT ISQ LATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-16 TA8LE 3.6-1 CONTAINMENT IsouTION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . ' 3/4 6-18 3/4.6.4 CCM8USTIBLE GAS CONTROL Hydrogen Monitors........................................ 3/4 6-23 El ectri c Hydrogen Recom0iners. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-24 3/4.7 PUNT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves............................................ 3/4 7-1 SYRON - UNIT 1 IX
. ..:---=-.---:.--_ . . , - - - - ---- _mW . ---, r . - - - - . - . - . - , . - . - . . , - - _ - - - . - , . _ . . , - _ - - . , . , . . . - - - - - - - . - . , _ - . _ . - - . - . . - . - - - - - - - - - - - - . - . _ - . . . - - -
i i taca PR00F & RST# COPY 8ASES SECTION PAGE 3/4.4.5 STEAM GDERAT0RS.......................................... 8 3/4 4-3
. 3/4.4.6 REACTOR COOLANT SYSTEM LEAXAGE............................ 8 3/4 4-4 3/4.4.7 CHEMISTRY................................................. 8 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... 8 3/4 4-5 3/4.4.9 PRESSURE /TEW ERATURE LIMITS............................... 8 3/4 4-7 7 TA8LE 8 3/4.4-1 REACTOR VESSEL TOUGHNESS.......................... 8 3/4 4-11 FIGURE 8 3/4.4-1 FAST NEUTRON FLUENCE (E)1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE........................ 8 3/4 4-12 .
FIGURE 8 3/4.4-2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF RT NOT
, FOR REACTOR VESSEL STEELS EXPOSED TO IRRACIATION AT 550*F........................... 8 3/4 4-13 3/4.4.10 STRUCTURAL INTEGRITY..................................... 8 3/4 4-15 3/4.5 !"EMENCY CORE C00 LING SYSTEMS 4
3/4.5.1 ACCUMULATORS.............................................. 83/45-1 . 3/4.5.2 and 3/4.5.3 ECCS SU85YSTEMS............................... 8 3/4 5-1 l =..,.
-wm~
, w u .. - . . - .............................. . .. .. o P 3/4.6 CCNTAIMMENT SYSTEMS 3/4.6.1 PRIMARY C0NTAIPMENT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 6-1 3/4.6.2 DEPRES$URIZATION AND COOLING SYSTEMS...................... 8 3/4 6-3 l 3/4.8.3 CONTAINMENT ISOLATION VALVES.............................. 8 3/4 6-4 3/4.6.4 CONSUSTISLE GAS CONTR0L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 6-4 1 i I l i i i l t BYRON - UNIT 1 XVI k
ATTACHMENT R (Index)
1) Index XI, limiting Conditions for Operation and Surveillance Requirement.s under Section; 3/4.7.12 Area Temperature Monitoring... 3/4 7-39 has been deleted to be consistent with the detection of pg 3/4 7-39, 3/4 7-40 and the Bases 3/4.7.12, also " Table 3.7-6.. 3.4 7-40" has been deleted for same.
W
~
(0438M)
1 INDEX hhh . 0 1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE ! 3/4.7.10 ' FIRE SUPPRESSION SYSTEMS Fire Suppression Water System............................ 3/4 7-28 Foam Systems............................................. 3/4 7-31 CO 2 Systems.............................................. 3/4 7-32 Halon Systems............................................ 3/4 7-33 Fire Hose Stations....................................... 3/4 7-34 TABLE 3.7-5 FIRE HOSE STATIONS.................................... 3/4 7-35 3/4.7.11 FI R E RATED AS S EMB LI ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-38 1/1 7 19 ASEA
- - - - PPun iwn .n en.a,viiwn un n ew-- - --
n n e .~n6evnAau..r............ gjg 9.. p .e , ,_e anc. . --- --._ ..-......,u, 3 j, , .m anoks w.. w r.n wn iR7FCKM4UML NunA6wnAnu. .......................... - wi w . wu 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES t Operating................................................ 3/4 8-1 TA8LE 4.8-1 DIESEL GENERATOR TEST SCHE 0ULE........................ 3/4 8-8 Shutdown................................................. 3/4 8-9 3/4.8.2 0.C. SOURCES 0perating................................................ 3/4 8-10 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-12 Shutdown................................................. 3/4 8- 13 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating................................................ 3/4 8-14 n Shutdow'................................................. 3/4 8-16 I 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective 0evices..................................... 3/4 8-17 TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES...................... 3/4 8- 19 k BYRON - UNTT 1 XI
ATTACHMENT S (Section 3/4.3) Circled items noted in this attachment have been previously submitted.
1) Table 4.3-1 (page 3/4 3-9) Reactor Trip System Instrumentation Surveillance Requirements Channel Calibration requirement #5 for Power Range High Setpoint and Intermediate Range is deleted because there is no Plateau Curve for these detectors, only fcr Source Range Detectors.

2) Table 4.3-1 (page 3/4 3-11) Reactor Trip System Instrumentation Surveillance Requirement Functional Unit 20 Reactor Trip breaker Trip Actuating Device Operational Test "M(7,11)" has been changed to "R(11)" because "11" is in conflict with "7" as it requires the Trip Actuating Device Operational Test (Independent verification of the Shunt and Undervoltage Trip) each refueling outage. The SER (attached) also requires the surveillance each refueling outage.
A system modification to allow easy testing of the Shunt and Undervoltage trips is scheduled for completion the first refueling outage.
3) Table 4.3-1 (page 3/4 3-12) Table Notations The sentence "For the Intermediate Range and Power Range Neutron Flux Channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1" has been deleted because it no longer applies to Power Range and Intermediate Range.

4) Table 3.3-3 (Page 3/4 3-18) Engineered Safety Features Actuation System Instrumentation The suggested change to "2" from "3" is necessary in order to be consistent with auxiliary Feedwater System at the station. IPSL-AF051 is suction transfer over for pump 1AF01PA, and 1PSL-AF055 is suction transfer over for pump 1AP01PB.

5) Table 3.3-3 (Page 3/4 3-19) Engineered Safety Features Actuation System Instrumentation We have 4 channels per steam generator. Therefore "3" has been changed to "4" for Steam Generator Water Level P-14 (High-High) channels.

6) Table 3.3-4 (Page 3/4 3-24) Engineered Safety Features Actuation System Instrumentation Nuclear Lead-Lag (NLL) card in the 7300 series process protection system.
The PLS calls out a setpoint of -100 psig on the bistables with a cimo constant of 50 seconds on the. Therefore, "/s" has been deleted from 4.e Trip Setpoint and Allowable Value to reflect the PLS data. Also, Trip Setpoint for 5.6 Steam Generator Water Level - High-High (P-14) has been changed to 82.4 from 82 to reflect the PLS data. (0438M)
ATTACHNENT B (Continued) (Section 3/4.3)
7) Table 3.3-4 (Page 3/4 3-25) Engineered Safety Features Actuation System Instrumentation Trip Setpoints The suggested change to "40.8" and "36" from "41" and "40" is necessary in order to be consistent with PLS data.

8) Table 3.3-5 (Page 3/4 3-29) Engineered Safety Features Actuation System Instrumentation Trip Setpoints Under Containment Pressure - High - 1 a) 12(2) hps been changed to 12(51 and Under Pressurizer Pressure-Low a) 12(21 has been changed to 12(5). This is to provide Consistency with the FSAR.

9) Table 3.3-5 (Page 3/4 3-30) Engineered Safety Features Actuation System Instrumentation Trip Setpoints 57 has been changed to 55. #5b., 75(1)/65(2) has been changed to
#5a.I 65(I /75(2) to be consistent with FSAR 6.2-5 #8, 7 has been changed to to be consistent with the FSAR #9, 25(2 has been changed to 250g2)tocorrectatypographicalerror.
10) Table 3.3-5 (Page 3/4 3-31) Engineered Safety Features Actuation System Instrumentation Trip Setpoints
#13, 7 has been changed to 5 to be consistent with FSAR Table 6.2-58
11) Table 3.3-5 (Page 3-32) Table Notation
#6 has been deleted because it is not referred to in Tables 3.3-5.
12) Tables 3.3-6 and 4.3-3 (page 3/4 3-39, 41)
Delete Functional Unit number 3, " Criticality Radiation Level (ORE-AR037/38)" and associated line "1; 2; *; i 15m R/h; 28" on table 3.3-6. Delete Functional Unit number 3, " Criticality - High Radiation Level (ORE-AR037/38)" and associated line "S; R; M; *" on table 4.3-3. Renumber "4" to "3", "5" to "4", and "6" to "5" in table 3. 3-6 and 4.3-3 to indicate proper sequencing after the deletion of the old number 3. Although the criticality radiation monitor is required in 10 CFR 70.24, Reg. Guide 8.12 Section C.1 allows for an exemption from 10 CFR 70.24 since Commonwealth Edison has determined a potential for criticality cannot exist. The FSAR shows that K epp in the new fuel vault, spent fuel pool and core during fuel movements will be less than the maximum Kerg allowed by ANSI 18.2-1973 and ANSI 210-1976. (0438M) 4
ATTACHMENT B (Continued) (Section 3/4.3)
13) Surveillance Requirement 4.3.3.3.1 (page 3/4 3-43)
Replace the existing description with attachment "CC".
14) Surveillance Requirement 4.3.3.2 (page 3/4 3-43)
The first sentence of Surveillance Requiremen 4.3.3.2 should be rewritten to read: "Upon actuation of the seismic monitoring instruments, the equipment listed in Table 3.3-7 shall be restored to OPERABLE status within 24 hours following the seismic event.
15) Table 3.3-7 (page 3/4 3-44)
Delete all of item 3; Triaxial Seismic Trigger (PC Board). Renumber item 4, Response-Spectrum Analyzer, to number "3" and renumber item 5, Triaxial Acceleration Sensors to number "4".
16) Table 4.3-4 (page 3/4 3-45)
Delete all of item 3; Triaxial Seismic Trigger (PC Board). Renumber item 4, Response - Spectrum Analyzer, to number "3" and replace the "N.A." under " CHANNEL CHECK" with "Q". Change the title of column " CHANNEL CALIBRATION" to " DIGITAL CHANNEL OPERATIONAL TEST". Renumber item 5, Triaxial Acceleration Sensors to number "4" and:
1) Replace "SA" for 4a-f with "W",

2) Replace "N. A." for 4a-f with "SA",

3) Replace "Q" for 4a-f with "N.A."
Replace "Q" for items 1.a and 1.b with "R". Replace "N. A." under " ANALOG CHANNEL OPERATIONAL TEST" for 2.a, b and c with "R". The original STS system writeup did not adequately describe the as installed system at Byron. Changes in the surveillance requirements were made to conform with vendor manual recommendations. The proposed change to FSAR Section 3.7.4 is attached for information.
13) Table 3.3-9 (page 3/4 3-50) Remote Shutdown Monitoring Instrumentation Pressurizer pressure Total No. of channels has been changed to "1" from "2" because there is only one channel or loop (1P-0455) feeding to the Remote Shutdown Panel.
(0438M)
o _ _. _ .c U.. O- . I
?
i l l TABLE 4.3-1 i 3 5 g REACTOR TRIP SYSTEM INSTRt31ENTATION SURVEILLANCE REQUIRE 8ENTS . i . TRIP I
. E ANALOG ACTu4 TING IRIDESfFOR I Q CHANNEL DEVICE
! . 1811C4 i
. e CHAIRGEL CHAlelEL OPERATIONAL OPERATIONAL ACTUATICII SURVI;ILLAllCE j j FINICTIOllAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

1. Manual Reactor Trip N.A. N.A. M.A. R N.A. 1, 2, 3* , 48, 58

2. Power Range, lleutron Flux

a. High Satpoint S D(2,4), M M.A. N.A. 1, 2

M(3, 4),
6 Q(4 R(,4, 4), 4
b. Low Setpolnt S R(4)
)
M N.A. N.A. 1###, 2 p w j i 3. Power Range, Meutron Flux, M.A. R('4) M N.A. N.A. 1, 2 y High Positive Rate l j lI * , j ;i 4. Power Range. Neutrcn Flux, N.A. R(4) M N.A. N.A. 1, 2 ; ) High Negative Rate . l l
5. Intermediate Range, lleutron Flux S R(4,/) S/U(1),M N.A. N.A. 1###, 2 g
;. 6. Source Ranee, Neutron Flux S R(4,5,12) S/U(1),M(9) N.A. N.A. 2##, 3, 4, 5 1 ! 7. Overtemperature AT S R(13) M N.A. N.A. 1, 2 --
I 1 8. Overpower AT S R M N.A. N.A. 1, 2 @ e l ' i C3
9. Pre rizer Pres e-Low S R M N.A. N.A. 1 M (Oh.c 9 e ) Oo
; 10. Pres or Pr e-High 5 R N N.A. N.A. 1, 2 m
11. P ssurizar Wate Level-High S R. M N.A. N.A. 1
(,0hos P- d
] ]
12. Re r Coolant ow-Low S R N N.A. N.A. 1 Q 4 i ca I .
C. 3 4 i
I 1 I I TABLE 4.3-1 (Continued) ; E I g REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS .
TRIP I !
i g ANALOG ACTUATING MDDES FOR q CHANNEL DEVICE WlICH g CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED q' 19. Reactor Trip System Interlocks (Continued)
d. Low Setpoint Power Range
', Neutron Flux, P-10 N.A. R(4) M (8) N.A. N.A. 1, 2
e. Turbine Impulse Chamber

Pressure, P-13 N.A. R M (a) N.A. N.A. 1, 2

20. Reactor Trip Breaker M.A. N.A. N.A. R / (/, 11) N.A. 1, 2, 3*, 4*, 5*
, 21. Automatic Trip and Interlock N.A. N.A. N.A. N.A. M (7) 1, 2, 38, 4*, 5*
5;> Logic
~M i
3 o c2
-v1 i a= =
2 L s E h = Ji! l
SEPs % gg - - The analog process channel testing is performed'by intioducing_ dummy input signals into the instrumentation channels and observing the tripping of the appropriate output bistables. The power range nuclear channels are tested by superimposing a test signal on the actual detector signal. To test the logic matrices of the solid-state logic protection system, pulse test signals are used in all possible trip and nontrip logic combinations. The test pulses are of short duration, and the trip logic is not maintained sufficiently long to permit opening of the reactor trip breakers. During logic testing of one train, the other train can initiate any required protective action. To test the reactor trip breakers, bypass breakers are provided. After a bypass breaker is closed, the associated reactor trip breaker can be tripped with a signal from the corresponding logic train. In addition to providing inputs to the solid-state logic protection system, analog signals of the protection channels are used for nonprotective functions, such as control, remote indication, and computer monitoring. To protect from faults in the nonsafety circuits affecting the protection system, isolation amplifiers are used. The isulation amplifiers are classified as part of the protection system. 7.2.2 Specific Findings The concerns arising from the staff review of the reactor trip system and their reso'ution follow. 7.2.2.1 Testing the Reactor Trip Breakers and Manual Trip Switches The reactor trip breakers are provided with undervoltage and shunt trip coils. Interrupting power to the undervoltage coil or energizing the shunt coil will trip the breaker. The undervoltage coils receive trip signals from both the solid-stste logic protection system and the manual trip switches (including the manual reactor trip switches and the safety injection switches). The shunt trip coils receive trip signals from the manual trip switches only. This pro-vides diversity and enhances the separation between the automatic and manual reactor trip systems. Testing of the undervoltage coil operation is carried.out with a trip signal from the-solid-state logic protection system. Testing of the manual reactor trip channel does not allow independent verification of the operability of the shunt coil and the undervoltage coil because the operation of a manual trip switch results in a simultaneous trip action by both coils The staff will_ include in the station's Technical Specifications a requirc~ ment to periodically and independently verify the operability of the undervoltage and snunt trip
}functionsatleastonceeachrefuelingoutage.
7.2.2.2 Protection System Sensors and Cabling in Nonseismic Structures Protection system trip circuit inputs that are loc 4ted in the nonseismic tur-bine building are (1) turbinc.stop valve closure limit switches, (2) turbine auto stop oil pressure' switches, and (3) turbine impulse pressure transducers.
t items 1 and 2 above provide inputs to the reactor trip on turbine trip circuit; Item 3 provides inputs to the P-7 interlock. The reactor trip on turbine trip
~
y
' ' ' ~
Byron SER ' 7-5 y , -- -m a f[*- -- ~w'u ~ w
O 7 - - . . . g ~ TABLE 4.3-1 (Continued) k[$ QQ g TABLE NOTATIONS "With System thecapable Reactor TMpwithdrawal. of rod System breakers closed and the Control Rod OMve NBelow P-6 (Intermediate Range Neutron Flux I'nterlock) Setpoint. O fMBelow P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint. (1) If not performed in previous 7 days. ' (2) Cogarison of calorimetMc to excore power indication above 15E of RATED THEMIAL POWER. Adjust excore channel gains consistent with calorimetric power if absoluta difference is greater than 25. fication 4.0.4 are not applicable for entry into MODE 2 or 1.The provisions of Speci-h (3) Single point comparison of incore to excore axial flux difference above 155 of RATED THERMAL POWER. Recalibrate if the absoluta dir'ference greater than or equal to 35. The provisions of Specification 4.0.4 are
' not applicable for entry into MODE 2 or 1.
(4) Neutron detectors any be excluded from CHANNEL CALIBRATION. O (5) Oetector plateau curves shall be obtained, evaluated and compared to manufacturer's data. Er t.'.a Lii .Miete "x
-My= 2 _1;!(0 p. ..isi... of 5pwiiim. ion M " ce 8 r-;=
e.o. Care not (6) Incore - Excore Caffbration, above 755 of RATED THERMAL POWER. The provi-sions of Specification 4.0.4 are not applicable for entry into N00E 2 or 1. (7) Each train shall be tasted at least every 62 days on a STAJGERED TEST SASIS. (8)
%Hth power greater than or equal to the interlock Setpoint the required
ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the inter-lock is in the required stata by observing the pemissive annunciator window.
). (9) Monthly surveillance in N00ES 3*, 4*, and 58 shall also include veriffestion that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the p missive annunciator window. Monthly surveillancs shs11 include verification of the Baron Oilution Alam Setpoint 10 less of than minuta or equal to an increase of twic~e the count rate within a period. (10) Setpoint verification is not applicable. (11) At least once per 18 months and following asintenance or adjustment of the Reactor trip breakers, the TRIP AC111ATING DEVICE OPERATIONAL TEST shall include independent veM fication af +ha ti. .el e--
^ ; ,.. . 6 rips.
(> (12) At least once per 18 months during shutdown v Soron Oilution Doubling test signal O . cc . CVCSthat on a simulaj;ed 04W
//A MNclose and 'f; sct.-iC; C . ing pay . .... .. Mvalve'wt@ ^ . e. . . 42.0%d E LO~ 444A.
open within 30 seconds. 6. (13) CHAl#IEL CALIBRATION shall include the RYD bypass loops flow rate. b SYRON - LMIT 1 3/4 3-12
) .._. - q; ~;- -- - ------ - - - . L; _ . _----_ . :: - '~-~
i 8 i a
=
TABLE 3.3-3 (Continued) f l
~ '
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION - , E MININUM CHANNELS APPLICABLE M TOTAL NO. CHANNELS M00ES ACTION e FUNCTIONAL UNIT OF CHANNELS To TRIP OPERABLE
6. Auxiliary Feedwater (Continued)

g. Auxiliary Feed-water Pump Suction Pressure-Low

(Transfer to Essential Service Water) 42. 2 2 1,2,3 15*
1
7. Automatic Opening of w Containment Sump Suction
) Isolation Valves w . Automatic Actuation Logic 1 2 1, 2, 3, 4 21 g a.
and Actuation Relays 2 , >l ..
b. RWST Level - Low-Low 4 2 3 1,2,3,4 ,.
16 I Coincident With ! Safety Injection See item 1. above for Safety Injection initiating functions and requirements. ] C3 C
8. Loss of Power

a. ESF Bus Undervoltage -2/ Bus 2/ Bus 1/ Bus 1, 2, 3, 4 198 C#

a;z I (Electromechanical j Relaying) Q

b. Grid Degraded Voltage (Solid State Relaying) 2/ Bus 2/ Bus 1/ Bus 1, 2, 3, 4 19" k -
T 1 {
,1 'l
~
O O O O O O O O O y 0 0 O jO, I l TABLE 3.3-3 (Continued) l
-< f y ENGINEERED SAFETY FEATURES AC10ATION SYSTEM INSTRUMENTATION l ! E MINIMUM !
1 4 TOTAL NO. CHANNELS CHANNELS APPLICABLE e FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTIOM i
; 9. Engineered Safety Features 4 Actuation Syst.es Interlocks l *
a. Pressurizer Pressure, 3 2 2 1,2,3 20 l
P-11
b. Reactor Trip, P-4 2 ,
2 2 1,2,3 22 l
c. 4 2 3 1,2,3 20 l Low-Low T,,g, P-12 4 w l D d. SteamGeneratorWaterLevel.fi/sta. 2/sta. gen. 2/sta. 1, 2, 3 20 na P-14 (High-High) gen. in any gen. In e'. operating each
sts. gen. operating sim. gen.
e t es es i Q*
mo i
1 , c3 C7 i i I
s I
, m - . -f t
i TABLE 3.3-4 (Continued) E B ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWABLE , i l ALLOWANCE (TA) Z ERROR (S) SETPOINT VALUE E FUNCTIONAL UNIT - 4 e 4. Steam Line Isolation N.A. N.A. H.A. N.A.
a. Manua) Initiation N.A.

b. Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A. N. N.A.

c. Containment Pressure- 6.2. 9.J.
High-2 5.0 0.71 1. 5 <M sig i lhe ps g (on Ste ne r - 14.2 10.71 1.5 > 640 psig 610 p g a sa d. ~ i Low ( p p_,,) sa - A e. Steam Line Pressure- < -110.0 psi'/t**
ve ote- 8.0 0.5 0 < -100 ~
(S h p-9) Psik
5. Turbine Trip and Feedwater Isolation

a. Automatic Actuation
. Logic and Actuation @Ef2 Relays M.A. N.A. N.A. N.A. N.A. ty /
k. Steam Generator Mater es.4 y).. l R*
Level-High-High (P-14) 5.0 2.18 1.5 1 42% $ 83% o ::a:s range range instrument instrument Q-span span h a 23
!., L) O !O i ' 1 . l TABLE 3.3-4 (Continued) .' B
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
i TOTAL SENSOR TRIP ALLOWABLE 3 E FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (S) SETPOINT VALUE q
e 6. Auxiliary feedwater I
a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic and Actuation
; Relays N.A. N.A. N.A. N.A. N.A. .
i :
c. Steam Generator Water Level-Low-Low-Start Motor- NI JS w Driven Pump and 30.0 27.18 1. 5 ->.41% of > 40% of D Diesel-Driven Pump ange iiarrow range w instruement instrument
! A hy$&n I
en,e mn
; d. Undervoltage-RCP Bus-Start Motor Driven Pump N.A. N.A. N.A. i5"-70 ""."A -i 0 % sii % -
I val s ac- L .elte r, i and Diesel-Driven Pump esaa --i t c y_ d if 71 Jei
c D- -
i i
e. Safety Injection- ~
Start Motor-Driven Pump and E i { See Item 1. above for all Safety Injection Trip Setpoints and @ s 5 Diesel-Driven Pump Allowable Values. vs l l Do a 1
!. E 3 4 o l
i l
. m m ., l l
_ , . , _ _ , . . _ TABLE 3.3-5 PMDF & HEW COPY ! l ENGINEERED SAFETY FEATURES RESPONSE TIMES . . . . _ _ . INITIATING SIGNAL AND FUNCTION _ RESPONSE TIME IN SECONDS _ _.._
1. Manual Initiation _

a. Safety Injection (ECCS) N. A.

b. Containment Spray N.A.

c. Phase "A" Isola' tion N.A.

d. Phase "B" Isolation N.A.

e. Containment Vent Isolation N. A.

f. Steam Line Isolation N. A.

f. Feedwater Isolation N.A.

h. Auxiliary Feedwater N.A.

i. Essential Service Water M.A.
j Containment Cooling Fans N. A.
k. Start Diesel Generator N.A.

2. Containment Pressure-Nich-1 ~
.. 6 - - _ _ ._
i'
a. Safety Injection (ECCS) i 27(1)/12

1) Reactor Trip i2

2) Feedwater Isolation i 7(3)

3) Phase "A" Isolation i 17(2)/27(1)

4) Containment Vent Isolation 1 25(1)/10(2)

5) ,
Auxiliary Feedwater i 60
.6) Essential Service Water < 32(2)f47(1)
7) Containment Cooling Fans 55(1)/40(2)

8) Start Diesel Generator i 10
'3. Pressurizer Pressure-Low - - . - -
g-
a. Safety Injection (ECCS) 1 27(1)/12N

1) Reactor Trip i2 l 2) Feedwater Isolation < 7(3)

3) Phase "A" Isolation < 17(2)/27(1)

4) Containment Vent Isolation 25(1)/10(2)

5) Auxiliary Feedwater 1 60

6) Essential Service Water < 47(1)/32(2)
O 7) Containment Cooiin Fans isSc )/40(2)
8) Start Diesel Generator _< 10 l BYRON - UNIT 1 3/4 3-29 h .-. .
, m m .. , y - - m -
- mw.,, . - - . . . _ . , , [ -- '-PRDCF& RED copy ^ ~
_.,_,~~~... TA8LE 3.3-5 (Continued) g_ ENGINEERED SAFETY FEATURES RESPONSE TIMES _.
._ INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS ,_
l 4. Steam Line Pressure-Low _ _ , _ _ _ ,
a. Safety Infection (ECCS) 1 22(4)/12(5)

1) Reactor Trip i2

2) Feedtsatar Isolation < 7(3) 3') Phase "A" Isolation -
< 17(2)/27(1)
4) Containment Vent Isolation 25(1)/10(2) ,

5) Auxiliary Feedwater 1 60

6) Essential fervice Water < 47(1)/32(2)

7) Containment Cooling Fans 55(1)/40(2)

8) Start Diesel Generator i 10 l

b. Steam Line Isolation i5 l

5. Containment Pressure-High-3

a. Containment Spray i M1)/45(2)

b. Phase "B" Isolation iM1)/45(2)

6. Steam Generator Water Level-High-High l.6 %
l
a. Turbine Trip < 2.5 O) 'l

b. Feedwater Isolation

7. Steam Generator Water Level-Low-Low _ _ _ _ _ _ _ _ _ _.

a. Motor-Oriven Auxiliary Feedwater Pump i 60 .

b. Diesel-Oriven Auxiliary Feedwater Pumps 1 60

8. Containment Pressure-High-2 Steam Line Isolation 14C3 RWST Level-Low-Low Coincident with 9.
' Safety Injection Automatic Opening of Containment i g 2)/255(1)
Sump Suction Isolation Valves A50 O SYRON - UNIT 1 3/4 3-30
---r.-__-
_ +.. _ _ - - i l Pi!DOF & RNEW COPY
-- - - . . TABLE 3.3-5 (Continued) __
ENGINEERED SAFETY FEATURES RESPONSE TIMES ,_ _ _ l _,,_,., __ .. .,,,_ INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS ,,
10. Undervoltaae RCP Sus . _ _ _ ___

a. Motor-Oriven Auxiliary Feedwater Pump i 60

b. Diesel-Driven Auxiliary Feedwater Pump i 11

11. Division 1 ESF Bus Undervoltage Motor-Oriven Auxiliary Feedwater Pump i 60

12. Loss of Power

a. ESF Bus Undervoltage i 10

b. Grid Degraded Voltage i 10

13. Steam Line Pressure - Necative ~ ~ ~~
Rate-Hian ]. Steam Line Isolation 17 b
14. Phase "A" Isolation _ _ , _ _ _ ___
Containment Vent Isolation i5
15. Auxiliary Feedwater Pumo Suction . _
Pressure-Low-Low _ _ _ _ , _ l Automatic Switchover to ESW N.A. i-t v 8YRON - UNIT 1 3/4 3-31 0
mw s= , g_. - =. .a ._
.- .a. .
n__
%...,_, i TABLE 3.3-5 (Continued) J_ kkyf Qg TA8LE NOTATIONS (1) Diesel generator starting and sequence loading delays included. . (2) Diesel generator starting and sequence loading delay not included.
Offsite power available. (3) Air operated valves. (4) Diesel generator starting and sequence loading delay included. RHR Pumps ,nlLt included. _._ (5) Diesel generator starting and sequence loading delays not included. Offsite power available. RHR pumps M included.
-(6) fr2rnet thy: .c i,,cid. ,
v I e e BYRON - UNIT 1 3/4 3-32 0 - _ - - _ _ - .. _ - . _ _ .. _ _ - . . - - -_- - -
t . j t' I I l TABLE 3.3-6 i 5 l' j j RADIATION NDNITORING INSTRtMENTATION FOR PLANT OPERAT100lS l E MINileM 4 CHANNELS CHANNELS APPLICABLE ALARM / TRIP FUNCTIONAL LAIIT TO TRIP / ALARM OPERA 8LE N00ES SETPOINT ACTION e I l
1. Fuel Building-l Fuel Handling **
2 1 2 mR/h 30 - (ORE-AR055/56) 1 t 2. Containment-Fuel Hand 11ag (IRE-AR0ll/12) 1 2 # 1 2 mR/h 26 1 , ,._ ,. ,__ m .- 1
> y n a ss sa-m u
l w 5x caseous Radioactivity-l
.I, RCS Leakage Detection
N.A. I 1, 2. 3, 4 N.A. 29 i (IRE-PR011A) i Yh. Particulate Radioactivity-RCS Leekage Detect!sn o j i (IRE-PRO 118) N.A. I 1.2.3,4 N.A. 29 g,]
! SE Main Control Room-Outside M ! Air Is.take (0RE-PR031/32 g3 3 and ORE-PR033/34) 1 2 per All 1 2 mR/h 27 , intake r i 8
.i a
i i l
. .. i t #
i l i TABLE 4.3-3 E E RADIATION MONITORING INSTRUMENTATION FOR PLANT *
OPERATIONS SURVEILLANCE REQUIREMENTS e
E DIGITAL
*1 CHANNEL r
CHANNEL CHANNEL OPERATIONAL MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATION TEST SURVEILLANCE IS REQUIRED
1. Fuel Building fuel **
[landling(ORE-AROSS/56) S R M l 2. Containment Fuel Handling S R M #
(ORE-AR0ll/12) 2 C7'.t'.c '.'t, " bi
^
h,'555 a u E
w i A 3A Gaseous Radioactivity-
^ H RCS Leakage Detection S R H 1, 2, 3, 4 (IRE-PRollA) afJr. Particulate Radioactivity- '. RCS Leakage Detection H 1, 2, 3, 4 y (IRE-PR0118) S R in C) IE. Main Control Room- '1 Outside Air intake - flu (ORE-PR031/32 and ,; . ORE-PR033/34) S R H , All Y...m ' I *With fuel in the fuel storage areas or fuel building.
**With irradiated fuel in the fuel storage pool.
Q
#Must satisfy Specification 3.9.9 requirements. Q 1
i'
INSTRUMENTATION
%O][ ]
{-~~~~'SEISMICINSTRUMENTATION _ _ , _ , __. _ _ _ ___ LIMITING CONDITION FOR OPERATION
p. - _ . ..
} 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERA 8LE. APPLICA8ILITY: At all times. __ ACTION: 9 a. idith one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.7.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERA 8LE status.
) b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. . . SURVEILLANCE REQUIREMENTS ) / ; 4.3.3.3.1 Each of the above required seismic monitoring instruments shall bh -r demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, CHANNEL , CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies .."..
shown in Table 4.3-4. 4.3.3. M ;;h of the C ;; n ;;i. ,t. sei ;ic .r.r. iter ?
-4.. ing e seiseic ::::t g;;ter- ther, er equel-ts 0.00 [i--instruments-ect hall be restored eteto OPERA 8LE status within 24 hours aNRATIO: i;;rfe. d ithin '
10 dey N o11owing the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground
) motion. A Special Report shall be prepared and submitted to the Commission - pursuant to Specification 6.7.2 within 14 days describing the magnitude, frequency spectrum and resultant effect upon facility features important to safety. _
padua1dA NM )
, % p A(mot ATaam. 3.3-7 'O BYRON - UNIT 1 3/4 3-43 ) . . , . . . , , , _ . . . . .. * - + - ~ ~ - -
l TM "cc." g\' _. N.3.3.3.! M k. Q p-
. . s..
ak a n e tam % s w e g.*u -+g e..
. s . a* ns. e u % % 93 %s-O % & W w ,fe %. % - Q & ut th- W A 4 pr~a .a. p w a w* e . **e++Me s + n.
t..Q* M e &is4%u% g g&ASch.
& & em, Ah. W - & u&
nwamany u s& y.
a. ai .ua _ r a su , wg % ,% .
0 Y@, 3 A L. h h 4 .* b. A-
*)4 %g % ~&.*k. @ a.W W ^&D &" A WAb 'A M 4 sk a - apa p " e .
PPSOF & REYH CDPt TA8LE 3.3-7 SEISMIC MONITORIAG INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRtMENTS ANO SENSOR LOCATIONS RANGE OPERA 8LE
1. Time - History Accelerographs

a. 12M-451' , Aux. Elect. Ihr, OPA02J N.A. 1

b. 7A-703', Byron River Screen House M.A. 1

2. Triaxial Peak Acceleregraphs

a. Cont./ Reactor Eq. Accumulators -2 g to +2 g i O -2 g to +2 g 1

b. Cont./ Reactor piping
. c. Aux. 81dg./ Cat. I piping -2 g to +2 g 1
3. Trie.;;ial-Setssic_Teiggae-(M-Soard)- c _
4
-e.-Aint:-Elect.-Am;-0PAcid -0,42-g *- 4.02 e i o
-..y 0
-tr Sy7.'T River 5.. _.. r'k-- = -G 01-1-te 4 ,42-g 1-4 3 f. Response-spectrum Analyzer 12M 451' . Aux Elect Ra, CPA02J None 1
[f. Triaxial Acceleration Sensors
a. Cont./10W - 377' -2 g to +2 g i

b. Cont /10W - 502' -2 g to +2 g I

c. Cont./10X - 426' -2 g to +2 g 1

d. Free Field /27 + 0.7N, 47 + 71E -2 g to +2 g 1

e. Aux. 81dg./18N - 426' -2 g to +2 g i

f. Byron River Screen House -2 g to +2 g 1

c. .
Nith reactor control room indication.
#Is a component of Time-History Accelerograph.
3/4 3-44 ( SYROM UNIT 1
=_.. .- . = .
l (-- PR00F & RMW mpy ' ( TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS , ,_ _ Ol'
'DtGrrAL Tp^$$nt ANALOG res r CHANNEL CHANNEL -GHANNELC OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATICMt TEST _ _ , _
1. Time History Accelerographs

a. 12M-451', Aux Elect Ra, OPA02J SA R dR

b. 7A-703', RSH SA R gR

2. Triaxial Peak Accelerographs

a. Cont./ Reactor Eq. Accumulators N.A. R G/P.

b. Cont./ Reactor piping N.A. R N.A. R

c. Aux. Bldg./ Cat. I piping N.A. R gR
&--TM ax i e l-S ei smi c-Tri gge F( PC- Bo a rd ) -- N.-A. N. A. N. A - '
3 /. Response-Spectrum Analyzer 12M-451; Aux Elect Ra, OPA02J M,AM N.A. Q Yf.TriaxialAccelerationSensors / ,-c /
a. Cont./10W - 377' i.SA-m/@ N.A. SA ,Q H.A-

b. Cont./10W - 502' SA Q .A. SA Q tt.A-

c. Cont./10X - 426' Sq Q N.A. sA q N.A-

d. Free Field /27 + 07N, 47 + 71E 5 Q N.A. SA p ti A -

e. Aux. Bldg./18N -426' SA 9 N.A. 5A Q ti.A-1' . Byron River Screen House SA q Q SA 1 N.A-BYROM - UNIT 1 3/4 3-45
C/B-FSAR AMENDMEfff 43 SEPTEMBER 1983 h {~$ A f. M p . Lt pcL p g value is determined by testing programs such as was done for the reactor coolant loop (Reference 1). 3.7.4 Seismic Instrumentation 3.7.4.1 Instrumentation fgr Earthquakes Seismic instrumentation is necessary to determine promptly the seismic response of nuclear power plant features important to permit comparison of such response with that used as the design basis. The seismic instrumentation for Byron /Braidwood utilizes two types of sensor-recorders with a playback capability available in the control room area. The location and function of these seismic devices were selected to provide adequately for the determination of seismic event loads into the structures via . x computerized analysis prog rams.5% WMe= i-N dung
% . h e. Location g Ac & andg acDescription h e n 4 Rat A . h . % @ A b w W D L' fT-F j '
of Instrumentation 3.7.4.2 3.7.4.2.1 Playback Unit g 4] A primary playback unit with strip chart recorder, ndicator lights, and playback system is provided in the cont ol area. x The lights indicate whether the system is tig;m and whether the operating basis or safe shutdown maximum accelerations are exceeded in any one of the three orthogonal directions in the basement of the containment structure. These directions coincide with the major axes of the analytical model used in the seismic analysis of the plant structure. 3.7.4.2.2 Time-History Accelerograph a..mw-ThL h %d'T+*t'h y M"h% ff our triaxial ter:7 each of which measures 9 # the absolute acceleration as a function of time in three orthogonal directions: these directions coincide with the major axes of the analytical model of the structure. The se -- ' -----" ^ M are placed at the following locations:
1. in the free field at site coordinates 41+00E, 27+00N and 39+00E, 41+00S for Byron and Braidwood Stations, respectively.

2. on the containment building foundation slab at elevation 377 feet and azimuth 145 degrees, i
1.7-30 l i i L
j/r u/c-ream Anznunswr 03 SEPTEMBER 1983
3. on the containment shell wall at elevation
{ 502 feet an$ azimuth 145 degrees, and
4. on the containment refueling floor at gusaAf E- elevation 426 feet.
'3.7.4.2.3 Peak Accelerographs x A triaxial peak ich measures the absolute pea!:
acceleration in three orthogonal directions coinciding with the major axes of the analytical model is provided at each of the following locations: I
a. on the accumulator tank located at elevation 426 '
feet in the containment building;
b. on the safety injection piping st elevation 421 feet in the containment building;

c. on the essential service return piping at elevation 346 feet in the auxiliary building.
3.7.4.2.4 Response Spectrum Analyzer
---v On; GranLLaliEW6 lei 4KiB1 Lc5hGnG6 SheG Is ud. e661 7sus 15 h6Gv~$dC$ ^ $h3u ch6ns for calculating the peak acceleration vs. f requency measured at the two locations listed below:
a. on the base slab of the containment building, elevation 377 feet. This location serves the dual purpose of monitoring the base slab response and the support motion of reactor equipment recorded from the accelerometer through the time-history accelerograph. This recorder is also equipped to provide signals in the control room area in the event preset acceleration limits are j exceeded (see Subsection 3.7.6.3) .

b. on the floor , elevation 426 f eet, in the counting room in the auxiliary building (recorded directly from the accelerometer). __
one time-history accelerograph is provided at the foundation level of the river screen house (Byron only) . Using the playback 3 the response spectrum can be determined for this location. j These locations are chosen to allow meaningful correlation between the recorded accelerations and those calculated using the analytical model of the structure. i 3.7-31
l
&/g-YhAM / AbENDMENT 25 MARCI' 1990
3.7.4.3 Control Room Operator Notification ,
l The centrally located seismic indicating and recording equip-ment near the main control room is the source of operator information concerning the acknowledgement of an earthquake. An acceleration of 0.02g in any direction activates the seis-mic switch which turns on the seismic monitors and lights up X the seismic alarm lights at the c:ntr:1 t:ti;.."vanel. ssasmec. wns.n m<n Using the spectrum analyzer, an operator can observe the triaxial spectral analysis from the containment building slab monitor and the auxiliary building 426 feet. elevation
>( monitor. (The latte r = ==-' --a== * = *- is direct-connected to x the spectrum analyzer, while th ormer ace =1=ra== is
, connected to the analyzer th ugh the time-histor ccelero-graph.) 5 0/50#- 66N38' i The operator can also use the playback feature of the system to obtain time history and spectral analysis of any of the x
,4 Y[os.u.ww.1=m_3..s#*11sted ED SEN50f S in Subsection 3. 7.2. 4. 2.4 4trft
Observed values which exceed the OBE acceleration threshold values stored in the spectrum analyzer are marked during the printout, and an alarm light is energized. Further analysis is needed to authenticate structural. loads and to
, evaluate observations via the structural response-seismic
' model. An observation which exceeds the SSE acceleration threshold is validated in a similar manner with the structural response-seismic model. When evaluated accelerations exceed SSE threshold values, the reactor is shut down. The alarm lights and the recorder data are available simultaneously
with the seismic event from the Containment Building slab monitor and the Auxiliary Building 426-foot elevation monitor.
Data from all other locations is available af ter the seismic 4 event, using the playback feature. l 3.7.4.4 Comparison of Measured and Predicted Responses The computer program which evaluates the time-history data computes the maximum response accelerations at various
points of the model. The observed response spectra can be compared with the response spectra. Agreement between the
! observed response spectra and the computed response spectra i from the time-history inputs demonstrates the adequacy of the analytical model. The magnitude of actual forces at various structural locations can then be ' compared to design values to authenticate the capability of the plant to continue operation without undue risk to the health and safety of the [ public. ( 3.7-32 i t
I i 4-. _9
, p - - y -- -( { 3 ? t j j g ,
a $ l I l i l l I l 4 . . .
._ i i 4 . ,
y T4 ; i i , i ; ;, ; ;;4 , ;,. t ,. ,
. J_4 _ , . ! ;.!3h.g,+.2.'I Ta q M.. # p ,% 'h s.
w
.. - .I g .tw i hI . b M =ki b N O q.er % d ! Mh4 ' e %- dA M'di@d A &@' ' u to . .@3,Na s.c. Dt sps oJ AAA 2 -g '
_ . o. ,h yaa.e % n M 4g.x h . u m m . %' cec.s @ dA A e 4 g b d .. A hmAh a,m_
- H n.M c % A.Al A s. h.
3- A Ema -
. - i ,-, W:ochsu* A g . e ad 4;tL2 M l e - - )_= . . ;. .; i 4.- .. f . i_.._. ,} 6 .
b.'1% mi% o%' 'h A h ~1 == b eA _h tar L+.k -_ML4 &_ eQ. w & a.!w k M y-Ga. 9 %g - 7 Am . 1
, _ .TM b' 1 j .
i ' I * e i
! l ! ! 6 i i ! . 4 _ .' . . .; _
j_g . ,_ # .. . p . _ . . . . . .
. .NM } N h .ib. t hil .
m Q s 2h,1- g.. - A.%A A mm u o u-g . W A : A 2su
&%A% % A m.s A-%i e +&qm.mm* M m&r -
A. c , . .
% - % %e gm. u m t M; & Q k k & t .,. m L k & & - % .
h - A c a e A s s e h .. . . f --* j(
+ , + . . * ..
7 r i e-1 . i j t 4 7-., . h . 1 ,, i '
! i ~
s . (,3 . l .
y Q v v W W W W w w _ k t TABLE 3.3-9 5 REMOTE SHUTDOWN MONITORING INSTRUMENTATION 1 5
TOTAL NO. MINIMUM READOUT OF CHANNELS c.
LOCATION CHANNELS OPERABLE
-4 INSTRUMENT , ,
H 1. Intermediate Range Neutron Flux 1PLO6J 2 1
2. Source Range Neutron Flux IPLO6J 2 1

3. Reactor Coolant Temperature -
Wide Range ,
a. Hot Leg IPLO5J 1/ loop 1/ loop

b. Cold Leg IPLO5J 1/ loop 1/ loop
. 4. Pressurizer Pressure IPLO6J Il 1 4 2 1
5. Pressurizer Level IPLO6J
6. Steam Generator Pressure 1PLO4J/IPLD5J 1/sta gen 1/sta gen h

7. Steam Generator Level IPLO4J 1/sta gen 1/sta gen

8. RHR Flow Rate LOCAL 2 1 i

9. RHR Temperature LOCAL 2 1 c=3

10. Auxiliary Feedwater Flow Rate IPLO4J/IPLO5J 1/sta gen 1/sta gen B
a i
ATTACHMENT T (Section 3/4.4) Table 3.4-1 (page 3/4 4-21) Reactor Coolant System Pressure Isolation Valves Deletion of valve number "1SI8900A, B, C, D,1SI8815" and function "CHG/SI Check Valve, CHG/SI Backup Check Valve" is requested because, RCS leakage can not propagate through these check valves for two reasons. First, the motor operated valves,1SI8801A and B are normally closed and are only open during a safety injection. Second, the Chemical and Volume Control System is always operating in the normal charging mode which is at a higher pressure than the RCS. (0438M)
PROOF & REWW COPY 7 _ REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCT'.ON 1:I=0CA,3# -CHG/SI-Gheck-Valve
-ifirIO615 -CHG/firI-Backup-Check Valve 1SI8948A,B,C,0 Accumulator Check Valve ISI8956A,8,C,0 Accumulator Backup Check Valve ISI8818A,B,C,0 RHR Cold Leg Check Valve ISI8819A,B,C,0 SI Cold Leg Check Valve ISI8949A,B,C,0 SI Hot Leg Check Valve 1SI8905A,B,C,0 SI Hot Leg Backup Check Valve ISI8841A,8 RHR Hot Leg Check Valve BYRON - UNIT 1 3/4 4-21 ., ----.g.-.- .. -~+ -- - _ _ . , , _ . . , . . _ . , . . , - , - . - , - - - - - , . , - - - - - - - , , - - - . - - - - , , - - , , - - - , - - . - - - - .
ATTACHMENT U (Section 3/4.7)
1) Section 3.7.12 (pg 3/4 7-39) Plant System - Area Temperature Monitoring Page 3/4 7-39 has been deleted because Byron Station has no equipment installed.

2) Table 3.7-6 (pg 3/4 7-40) Plant System - Area Temperature Monitoring Page 3/4 7-40 has been deleted because Byron Station has no equipment installed. .
Safety-related equipment at Byron Station will not be subjected to temperatures in excess of their environmental qualification test temperatures . A program has been undertaken at Commonwealth Edison to ensure that installed equipment at the plant site will be replaced prior to its.end-of-life condition as defined by the results of equipment qualification test program. Temperatures utilized during equipment qualification testing were conservative with respect to Byron site. (0438M)
-L :_-.. - .s- ~.._;;; -- - - - - - . - . . - -. - - ? l
( = 7 5 = 5a5 PRDOF & REydPY j] 3/4.7.12 AREA TEMPERATURE MONITORING Ii
\
LIMITING CON 0! TION FOR OPERATION l-i4 3.7.12 The temperature of each area shown in Table ). -6 shall be maintained within the limits indicated in Table 3.7-6. / APP .ICABILITY: Whenever the equipment in an af area is required to be
- u.
M

..
i!
a. With one or more areas exceed the temperature limit (s) shown in Table 3.7-6 for more than a urs, prepare and submit to the Commission i
within 30 days, purslaant Specification 6.7.2, a special Report that provides a record of the umulative time and the amount by which the temperature in the aff ' ares (s) exceeded the limit (s) and an 6 analysis to demonstra 'the continued OPERASILITY of the affected equipment. \
. \
b. With one or more exceeding the tamperature limit (s) shown in Tabler3.7-6 by than 30*F1 prepare and submit a special Report as
s 4- required by ON a. above, and within 4 hours either return the area (s)
;? to within the reture limit (s) or declare the equipment in the affected a (s) inoperable.
SURVIILLANCE REQUIREMENTS 4 4.7.12 /tamperature in each of the areas shown in, Table 3.7-6 shall be detemi to be within its limit at least once per hours. be ktb MW naffy t 8YRON - UNIT 1 3/4 7-39
. _ _ . h ._& . M h. . - - #
i I N ---
8 . TABLE 3.7-6 AREA TEMPERATURE MONITORING N g AREA TEMPERAT1JRE LIMITA*F) 1.
\ Misc. Elec. Equip. and Battery Rooms, /
10S ESFx5witchgear Res. , Div.12 CablesSpreading N g 2. Upper and Lower Cable Spreading Rooms 90
3. Diesel-Ge\ nerator Rooms, Diesel 011 132 Storage Rooms

4. Aux. 81dg. Orain Tank Cubicles, 122 Recycle Holdup TankxCubicles, Gas 8 DecayTankCubicle,G4sDecayP/pe Tunnel, Recycle Evap.' pe Tunnel, Floor Drain Tank Cubic 1 K Recycle Holdup Pipe Tunnel, Filte ipe Tunnels, Domineralizer Cubi 1,es, PositiveDisplacementfumpRooms, SpentVent B1dg. FuelExhaust Pit Pump /11ter Cubic 1 Roos, Aux.\ ~- - -
Process Filter Cub'icles, Centrif-
, Pusp Rooms ugal Charging /
5. Containment,5 pray Pump Rooms, 130 g RHR Pump Rooms, Safety Injection Pump Rooms, Pentration Areas El.
346, 364 6.
/
Control Roos
/
7. Radwasta Evaporator cubicles 118 g

8. Boric Acid Tank Cubicles 127 t-deWec Mabnaffe BYRON - UNIT 1 3/4 7-40 8
L _ _ __m
l ATTACHMENT V (Bases 3/4.7.12)
1) Bases 3/4.7.12 (pg 0 3/4 7-8) Plant System - Area Temperature Monitoring Page B 3/4 7-8 has been deleted because Byron Station has no equipment installed. Safety-related equipment at Byron Station will not be subject to temperatures in excess of their environmental qualification test temperatures. A program has been undertaken at Commonwealth Edison to ensure that installed equipment at the plant site will be replaced prior to its end-of-life condition as defined by the results of equipment qualification test program. Temperatures utilized during equipment qualification testing were conservative with respect to Byron site, 1
(0439M)
1 { PLANT SYSTEMS PRODF & HBV COPY BASES 3/4.7.12 AREA TEMPERATURE MONITORING The area temperiture limitations ensure that safety-related equipment will not be subjected to ' temperatures in exces's of their environmental qualification temperatures. NExposure to exenssive temperatures may degrade equipment and can cause a lossN f its OPERASILITY. .
/ /
x\/Xs,
/ s / -x . s 1
I l BYRON - UNIT 1 8 3/4 7-8 g =.-.- . , . , - - - . - , - n - . - - -.~-.-,--,.e.-,y - - - - - - - - .- w. -- y - ---, - -- ---- -- - --- -.--y-----.-*w
e ATTACHMENT W .' (Saction 5.0)
1) Figure 5.1-1 (page 5-2) Ev.clusion Area ard Unrestricted Area For Radioactive Gaseous and Liquid Effluents.
This figure replaced old Figure 5.1-1 which did not have meteorological tower on the map.
=
e (0438M) . l
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, ,' ,. . . '. - Q ... .{ .0:.::;:76.. :. .: :., $ 1 .. ...i!yi?+j$i(U .+:i . ,r , u ,, . 0: , ,. .. : %;4+:!::.:.Mg;0:. ;.::; : :+3 , f 93 -@M.i:!{:!:f,:':.:::
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.: !j[;! , ** - :j,!;;;;.. a egeesQy ::yj f:;$ @$hij ee .M - ..
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: ::j.t . . :0::23M+:e+..........d..:W 4;il!!E.is.pi'
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'8 's ,..., f *
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E BYRON STATION O 1 2 FIGURE 5.1-1 s c ALE IN MILE S EXCLUSION AREA ANC UNRESTRICTED AREA
-uma sus EXCLUSION AREA BOUNDARY ......... RESTRICTED AREA BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS i
5-2 BYRON - UNIT 1
ATTACHMENT X (Section 6.0)
1) Section 6.3.2 (pg 6-11) Onsite.
Specification 6.2 does not specifically require programs and procedure; however; Section 6.0 does outline required procedures and programs and should be the section referenced.
2) Administrative Control (pg 6 24)
These changes were recommended by Commonwealth Edison Technical Services Nuclear Dept. They are consistent with practice for other CECO Nuclear station and are consistent with the changes already made in Section' 3/4-12. (0438M)
PRODI& REVIEW CDP [
, ADMINISTRATIVE CONTROLS OFFSITE (Continued) h) Instrumentation and Control Engineering graduate or equivalent with at least 5 years of experience in instrumentation and control design and/or operation.
1) Metallurgy Engineering graduate or equivalent with at least 5 years of experience in the metallurgical field.

3) The Supervisor of the Offsite Review and Investigative Function shall have experience and training which satisfy ANSI N18.1-1971 requirements for plant managers. ,
ONSITE , 6.3.2 The Onsite Review and Investigative Function shall be supervised by the Station Superintendent.
a. Onsite Review and Investigative Function The Station Superintendent shall: (1) provide directions for the f
Review and Investigative Function and apooint the Tecnnical 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-jact 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 Review and Investigative Function for concurrence in a timely manner, those items described in Specification 6.1.G.l.a which have been epproved by the Onsite Review and Investigative Function,
b. Responsibility The responsibilities of the personnel performing this function are:
6.6,l
1) Review of: (1) procedures required by Specification .Grs and 6.6.7 changes thereto, (2) all programs required by Specification Grs-and changes therato, and (3) any other proposed procedures or changes thereto as determined by the Plant Superintendent to affect nuclear safety;
'( 2) Review of all proposed tests and experiments that affect nuclear safety; BYRON - UNI' 1 6-11
-.. . - _ _ _ . _ _ _ . _ . .-m.,.__-
PRDOF & REVH COPY ADMINISTRATIVE CONTROLS { REPORTING REQUIREMENTS (Continued) . ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
6.7.1.6 Routine :.o_...y... -
mv . . w- n i vr - ece..,, Reports covering the operation of the Unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality. ') The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activites for the report period, including a comparison with preoperational studies, with operational controls as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operatior; on the environment. The reports shall also include the results of the Land Use Census required by Specification 3.12.2. The Annual Radiological Environmental Operating Reports shall include t.no results of analysis of all radiological environmental samples and of all
; environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the 00CM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include tne following: a summary description of the Radiological Environmental Monitoring Program; at least two legible maps *" covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program and the corrective actions being taken if the specified program is not being performed as required by Specification 3.12.3; 1 reasons for not conducting the Radiological Environmental Monitoring Program as required by Specification 3.12.1, and discussion of all deviations from the sampling schedule of Table 3.12-1; discussion of environmental sample measure-ments that exceed the reporting levels of Table 3.12-2 but are not the result of plant affluents, pursuant to ACTION b. of Specification 3.12.1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable. F NWEft.T
*A single submittal may be made for a multiple unit station. **0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations, g & c 1,2:th: Ef9 ;:2 Meek -
' - * ** In lieu of submission with the M:rrrr' *: 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. !- BYRON - UNIT 1 6-18
. , . , , , . - - , , , , , , , . , , ,.-,m,.,.,--n,,. -
TM M.E * . _ . - . _ _ _--- }. W _. ..}.m.,, _r.hu..... A. ,,, {,,,,, Q.. ___g_ g m_ 0 The2-. . nm. 1 n. m _ ..-__ &. _. _:'t r 2;.__., 1 ef-; u y. M hall also include an annual summary of hourly
- --- meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind -. speed, wind direction, atmospheric stability, and precipitation (if :::easured), _
or in the form of joint frequency distributions of wird speed, wind oirection, and atmospheric ' stability."* This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the Unit or Station'during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the 5ITE BOUNDARY (Figures 5.1-3 and 5.1-4) during the report period.
' - ' ~
All assumptions used in making these assessments, i.e. , specific activity, exposure time and location, shall be included in these reports. The meteorolo- .. _ gical conditions concurrent 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 00CM.
The : I h b to be submitted 60 days .- - after January 1 of each year shall also include an assessment of radiation doses - .---- to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other _ _ __ nearby uranium fuel cycle sources, including doses from primary effluent path- . _ ways 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 liquid and gaseous affluents are given in Regulatory Guide 1.109, Rev. 1, ~ ~ '
~ ~~ October 1977.
- u_ _ _
l -- - changes made during the reporting period to the PCP : d t: the '"'Wpursuant ^ ~ l to Specifications 6.11 and 6.12 respectively, as well as any major changes to Liquid, Gaseous or Solid Radwaste Treatment Systems, pursuant to Specifica- --- It shall also include a listing of new locations for dose calcula-tion 6.13.
u. _ tions and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.
4e. - . - .' e
...m ..e e, 4-3,,-.. ,. w.< _.,,--.m>-.-w .,.-n. .s%- . , ,<.ww.=.. ----#,.e.1+i.--.v.-- . ..-~.,--w--- -a.e --
I' . . . . - . ._
~ , - .- . -
l
.- - ~ ..- --._. ~ ~ ~ ~ . . - - - - .._. - . _ . _ . _ . -
l l . -. . . - _ . . . - . ... . l-l l.. . , . _ _ _ . - - -,... -. _ - -. . ~, . _ .. _ _ _ _ _ ._ . _ - _ - . .- - - - _ ~ - , ..
PROOF & REVIEW COPY !
,m , ADMINISTRATIVE CONTROLS NN a' REPORTING REQUIREMENTS (Continued)
W hv WA ,%<d"
-EM?""UAL RAO.CACTT.'C C' LUC"I OELEAS 9EP0d* ,
6.7.1.7 Routine
- - < covering the operation of the Unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period O of the first report shall begin with the date of jnitial criticality.
bewd Mg/4. O S g The h ;ionnw.i Rouiva nim Effluent 20.:::Dep' orts shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the Unit as outlined in Regulatory Guide 1.21,
" Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and D Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61),
type of container (e.g. , LSA, Type A, Type B, Large Quantity), and SOLIDIFICA-3 TION agent or absp,rbent (e.g., cement, urea formaldehyde). The Sesiannual Radicactive Effluen Release'ltegort to be submitted 60 d after Januiry 1 of eacVyear shall al,s include an annual /ummary of hourly meteo,rpT'ogical data,,c611ected over,,the prev,f ous year. Jh'is annual suma y may be pi'ther in the, form of an houryby-hour Jisting on magnetic tape of wind Jg) saeed, wind dir,ection, atmospheHe stabpiity, and precipitation (,1f measured), A' ,/fsr in the form of joint frequency distributions of,Avind speed,, wind directiorp and atmosppe'ric stabilityf 7" This s,ame report spall include ,an assessment,,4f the radiation doses due,,to the radjoactive liquid and gaseous effluents yeleasec from ( Unit or Station during the previous , calendar year / This sameA eport shal also include an' assessment /'of the radi,ation doses from radioactive li id and gaseous / effluents to MENBERS OF'THE PUBLIC du'e to their,4ctivities
/
O i ' side the SITE' BOUNDARY (Fig 0res 5.1-3,a'nd 5.1-4) during the report period. / 1 assumptions used in maying these assessments, exposure time'and locatiop, shall be fncluded in these reports'. i.,4. , specifje activity,Themet f gical condjt' ions concurrpnt with the time of relea'se of rad factive materials in gaseous effluents, aJd determined'by sampling dmeasurement[ shall b 'used for detpfmining thy /' gaseous y doses. pathwa/requency The assessme t of g radiat~on doses shaLT be perforrped in accordap'ce with e methodology and ) par '.ters in the 00CM. / / / j
/ *A single submittal may be made for a multiple unit station. The submittal O should combine those sections that are common to all units at the station; however, for units with separate Radwaste Systems, the submittal shall specify
_ the releases of radicactive material from each unit. b **I offubmiMonwtht Semi ual ffueI Releasp party the I enseep tp pti e a'dioactdepsump,ry(ofre(uir ini g'thi
/ mete rologi dat on s te in file fat s 1b Og up requ t. p/rovided[to f the JRtf v
BYRON - UNIT 1 6-19 O
* * * ... .. . . . . - . %. .. - - - = . - - - -**&
1 r
! 200F & HEW COPY I
ADMINISTRATIVE CONTROLS
) REPORTING REQUIREMENTS (Continued) / - f - / ,.e The4femiannuaVRadioactive'Effluen't. Release' Report, t'o be,shbmitted 60 days also incipfe an assessment of radiation desns aftey4anuary the like1 mostPtif eachMEMBEF expossd y,ee'r shall'0F THE P,U8LIC from' reactor releases and 9 arby urani[um fuel, cycle soucces, including dosts' front primaryf fffluent ways and' direct radiation, fo'r the pre (ious ca}endar year to show ance/ confor$ path- / )
with CFR Par 0, "Enyfronmenta Radiatio # Protection Sta'ndards f f NucJiar P Operati . Acceptable met dsforcalcuja).fngthepo'secent butijan fro uid and seousaffluentsa given infReguKatorf Guidd 1.109, ev. Y, October 1977. OM bAUAsy'e*I-
' ^ - ' " - " ' " " ' ' ' - - - ' ' ' - - ' -
orts shall include a list The "
) and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period. . , W The,Sem6nnual,Radioactiye1Effluenge' Tease Reports shpT include ~ any 't X-ch ea'made dur)i tg the reporting p ied to the PCP and to the 00,CM, pursuah p p ecificatp6s6.11and6.12r ectively,-a,s well,at' any major changes to quid, Gassous or 5 9Md Radwas,t Treatmen t tion 6.13/ It shaft also include a listin,t'g Systems /pursuan of new'locatio / o Specifica-for dos'e cal a-o oc c 2 /^ .he s
W - ......... .....____,... .. hN%
..............n.,,,.w. shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specifications 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively.
MONTHLY OPERATING REPORT
6. 7.1. 8 Routine reports of operating statistics and shutdown experience, l
including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, l U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to ) l the Regional Administrator of the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report. RADIAL PEAKING FACTOR LIMIT REPORT ! 6. 7.1. 9 The F,y limits for Rated Thermal Power (F y ) shall be provided to the ) NRC Regional Administrator with a copy to Director of Nuclear Reactor Regulation, Attention: Chief, Core Perfossance Branch, U. S. Nuclear Regulatory Commission, l Washington, 0.'C. 20555 for all core planes containing Bank "0" control rods and ) T all unrodded core planes and tV plot of predicted (F q ,pRel) vs Axial Core Height with the limit envelope at least 50 days prior to each cycle initial ved by the Commission by letter. In addition, criticality unless otherwise m'- O . V l BYRON - UNIT 1 6-20 b
- .. . . . - ::~~ . .. _ . - - --U
) t PROOF & HEW COPY O ADMINISTRATIVE CONTROLS ) HIGH RADIATION AREA (Continued) which the radiation penetrates shall be barricaded and conspicuously posted as !
a high radiation area and entrance thereto shall be controlled by requiring l issuance of a Radiation Work, Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously 3 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 radia-tion 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 radiation dose rate in the area.r.d alm.; ....n a p rn i. ir.tagnt:fA--
t:: i: r;n heJ.^ 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
c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for n providing positive control over the activities within the area and
](/ shall perform periodic radiation surveillance at tne frequen specified P/ 'h; ":d/Chn E;:niser er Lnd "r!+h chuWh in the RWP.
6.10.2 In addition to the requirements of Specification 6.10.1, 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 3 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 axcept during periods of I 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 individuals in that area. In lieu of the stay time specification of the RWP, p 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 high radiation areas accessible to personnel with radiation l levels of greater than 1000 mR/h that are located within large areas, such as 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 barricaded #h!" Eg liG ht '
Mll t:
n":':d es a .cn'n; e/ sk::w::;ic::;l)put:d,nd:
~1QA geAL PO BYRON - UNIT 1 6-23 t - + - . a.--_..,,.i4-,,_ _, ,. _ , n-, 7___., , _ , , , _ , , , . , , _,_,,_._,.,__%,, , _ , , .,,.,,,m_ y.,,_,m.,a,-_,,,_,.__,y-_,_y,,
v PROOF & PHEW COPY ADMINISTRATIVE CONTROLS -- t 6.11 PROCESS CONTROL PROGRAM (PCP) 6.11.1 The PCP shall be approved by the Commission prior to implementation. 6.11.2 Licensee-initiated changes to the PCP: Abnugl N o icd 6hVlr*M>chIAk Offidi,Q R eport
a. Shall be submitted to the Commissio in the # ~ '" " ~
CTTI.. R i.... R., ..; for the period in which the change (s) was made. This submittal shall contain:
1) Sufficiently detailed information to totally support the rationale for the change wthout benefit of additional or supplemental information; a

2) A determination that the change did not reduce the overall conformance of the solidified. waste product to existing criteria for solid wastes; and

3) Documentation of the fact that the change has been reviewed and

( found acceptable by the Onsite Review and Investigative Fanction.

b. Shall become effective upon review and acceptance by the Onsite Review and Investigative Function.
6.12 0FFSITE DOSE CALCULATION MANUAL (00CM)
( 6.12.1 The ODCM shall be approved by the Commission prior to implementation.
6.12.2 Licensee-initiated changes to the ODCM: as cowb\ led rev;s;.as w;h
a. Shall be submitted to the Commission k tt; f r W :21 ".;di;;;the -
E"i....i. " i rre ": :rt 'er tt: " c- - f
.',i:5 -5:n--de dag:e .
Dccukentshen narpolNg affr t h :. it.!e e. ii.i I .ii.Il ; p y c sOe nll c[mt;antain *ira 1 0 d4 3;ri;d eT
1) Sufficiently detailed information to totally support the rationale for the change.lwitnour, oener1t of additional or l supplement.ai inTormation.' Information submitted should consist of a package of those pages of the 00CM to be changed with each
'g --) page numbered, dated and containing the revision number, together with appropriate analyses or evaluations justifying } ~
l the change (s);
2) A determination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations;
and
3) Documentation of the fact that the change has been reviewed and found acceptable by the Onsite Review and Investigative Function.

b. Shall become effective upon review and acceptance by the Onsite Review and Investigative Function.
( . BYRON - UNIT 1 6-24 __ ___ __ _ w
ATTACHMENT Y
1) Bases 3/4.3.3.3 (page B 3/4 3-4)
Replace Section 3/4.3.3.3 with Insert "JJ" The original STS system writeup did not adequately describe the as installed system at Byron. Changes in the bases fully describes the system. _1 (0438M)
l 't INSTRUMENTATION [ ~ BASES -.: : .~.: MOVABLE INCORE DETECTORS (Continued) For the purpose of measuring Fq (Z) or F a full incore flux map is used. Quarter-core flux maps, as defined in WCAP-8M8, June 1976, may be used in ; recalibration of the Excore Neutron Flux Detection System, and full incere flux saps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable. 3/4.3.3.3 SEISMIC INSTRUMENTATION ._
-T OPERABILITY of the seismic ~ instrumentation ensuresyt.-sdfficient "j j "
capability- vailable-toTromptly det'brn(ne the_magnhude of a seismic event
~
(and evafuatepermit requi,r'ad-tE the-responsedthose featurssimiiortant'i.o comparisWof-tOleasurebresponse safety. to-that used in 'theThisl,apability dayfgrr: basis for' the;faciif't'y to determine'i~f71antbshutdown is required pursuant to' Appendix A of'10- CFJL Part 100. The instiumentationciLconsistent with,J.he reTommendationsjf-Rigulatory Guide Id$, "Instrumifi'tationWr^
. Earthquakes," April 1974.
3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that I sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972. 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY oi' the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANOBY of the facility from locations outside of the control room. *This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50. 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During ; and Following an Accident," May 1983 and NUREG 0737, " Clarification of TMI ! Action Plan Requirements," November 1980. 1 ( BYRON - UNIT 1 B 3/4 3-4 l l l
i TaSed, T M " M G 4,4 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient cap:bility is available to promptly determine the magnitude of a seismic event and cvaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. he inthrumenhalien__ Ronsish of bc time-hidory response Apetlewn andyter 3 a playk,ack _. Land b peak _rtcordi 3 actq\ercnteler.:;,, dnd SiA hyluit( acaelevamelers. One . _bime diskni recorde< and one %emor are localed down at 6e Viver Screen Ib The. ved of de egwpment, achAwth e umus 3 is \oeded iw ue Ruu%g Etedu c al 1 m.. $h4 remaining sensors are locaba esbs : uree in canknmet, one in de Rutil:q 1Su;l dig and one at no free bid loch a + on , 97 n r . The _.pok recording acedeemehrs we passivo deurst wIvck he non%hy _ _ on h rest of Ae s3sb and _are toeded on reath egutemed3 reackr pip;n), - . . in A ess Ao unhtummt . on . category 1 prping. _ _%. %dal sceekromeh< is based on he ornoynal 6e.halanced servo-9eaeleromehrs whd generste. a adago Mgn21 p Alab , Tk _. voihge. Signals are tranwnied hhe bme-kishg recorder' in th+ fku;iiary
&ledrkul %cm a digilidand recorded on magnek hp-
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ATTACHMENT Z
.These-changes delete all reference to three loop operation.
This mode of operation is not supported by existing safety analyses. t-4 e 4 L 8376N
1) Index pg (III, X) Safety Limits and Limiting Safety System Settings "During Four Loop Operation" and'"Four Loop In Operation" has been deleted from pg X and III respectively to get rid of all indications that Byron has more than one type of Loop Operations.

2) Section 2.1.1 (Page 2-1) Safety Limits Delete from section 2.1.1 the "s" from " Figures" and "and 2.1-2. . . . .
respectively".
3) Figure 2.1-1 (pg 2-2)
Deleted "Four Loops in Operation". Only one type of Operation.
4) Figure 2.1-2 (Page 2-3)
Delete this page for Figure 2.1-2.
5) Section 2.1.1 (Page B2-1) Reactor Core Delete from paragraph #4, the "s" from " Figures" and "and 2.1-2".

6) Section 3.1.3.4 (Page 3/4 1-19) Rod Drop Time Delete Action 3.1.3.4.b.

7) Section 3.1.3.6 (Page 3/4 1-21) Control Rod Inscrtion Limits Delete from LCO 3.1.3.6 "s" from " Figures" and "and 3.1-2".

8) Figure 3.1-1 (pg 3/4 1-22)
Delete "Four Loop Operation".
9) Figure 3.1-2 (Page 3/4 1-23)
Delete space for Figure 3.1-2.
10) Table 3.2-1 (Page 3/4 2-15) DNB Parameters Delete "Three Loops in Operation". Delete "Four Loops in Operation".
Delete "MM". Delete "MN These values left blank pending NRC approval of three loop operation". (0435M)
3 ATTACHMENT (Continued)
11) Table 3.3-1 (Page 3/4 3-2) Reactor Trip System Instrumentation Delete from 7. Overtemperature ti T "b. Three Loop Operation".
"Four Loop Operation".
Delete from 8. Overpower li T "b. Three Loop Operation". "Four Loop Operation". Delete the astrics "MM" on these lines.
12) Table 3.3-1 (Page 3/4 3-5) Table Notations Delete "MM Values left blank pending NRC approval of three loop operation"

13) Section 3.7.1.1 (Page 3/4 7-1) Safety Valves Delete Action 3.7.1.1.7 The Word "four" Delete Action 3.7.1.1.b.
Change Action 3.7.1.1.c to 3.7.1.1.b.
14) Table 3.7-1 (Page 3/4 7-2)
Delete from Table 3.7-1, the words "During Four Loop Operation", and all of Table 3.7-1. Delete "II These values left blank pending NRC approval of three loop operation".
15) Section 3/4.7.1.1 (Page B 3/4 7-1) Bases - Safety Valves Delete "For three loop operation:", and the word "four" from formula.
SP = (X) - (YlGD_ x (M) X
16) Section 3/4 7.1.1 (Page B 3/4 7-2) Bases-Safety Valves Delete "M" and note with it.
(0435M)
. - =
INDEX t SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS gg lo ulkc. loop god be. de ekd 2.1.1 REACTOR CORE.................... ........................... 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - M"P L^F5 in C^:."ATIO.':. . 2-2 J!C"RC 2.1-2 G h .%)................................... ....... . 2- N
2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... 2-4 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-5 BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 8 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS

2. 2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS............... B 2-3
( BYRON - UNIT 1 III s y
- . .~
I fh b i 1 . LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE l TA8LE 3.7-1 MAXIMUM ALLOWA8LE POWER RANGE NEUTRON g p &c. loo p E 0d FLUX HIGH SETPOINT WITH INOPERA8LE
/ g dejeW.
STEAM LINE SAFETY VALVES.00HNG-feUR-L00P
-6PERAH0ft. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-2 -T'O': 3.7-Z (5vum)............................................... ---3/' '-2M TA8LE 3.7-3 STEAM LINE SAFETY VALVES PER L00P. . . . . . . . . . . . . . . . . . . . . 3/4 7-3 l Auxiliary Feedwater System............................... 3/4 7-4 . Condensata Storage Tank.................................. 3/4 7-6 Specific Activity........................................ 3/4 7-7 TA8LE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM......................... 3/4 7-8 Main Steam Line Isolation Va1ves......................... 3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION. . . . . . . . . . 3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-11 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM........................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK....................................... 3/4 7-13 3/4.7.6 CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM. . . . . . . . . . . . . . . . 3/4 7-15 3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FILTER PLENUM VENTILATION SYSTEM....................................... 3/4 7-18 3/4.7.8 SNUB 8ERS................................................. 3/4 7-20 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUB 8ER FUNCTIONAL TEST. . . . . . . . . . . 3/4 7-25 3/4.7.9 SEALED SOURCE CONTAMINATION.............................. 3/4 7-26 4
SYRON - UNIT 1 X
. Q EMeM w + - - e *- m-,c.-e --mwg-- rg ..w,, ---y yv.-,e--i.mm.,-,y-mmyr 999-y--~w--T -Nr' m' we-" "+Wq"m--'.*'w' ='-- =w C="
PRIDF & RMEW C0FY (P 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figuref 2.1-1,=d 2.12 fer four _:M ::.. luvreperati = ,,m ;::+1 q16 APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by'the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pres-surizer pressure line, be in HOT STANDBY within 1 hour, and comply with the requirements of Specification 6.5.1. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: MODES 1, 2, 3, 4, and 5. ((G ACTION: MODES 1 and 2: l Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be l in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour, and comply with the requirements of Specification 6.5.1. MODES 3, 4 and 5: Whenever the Reactor Coolant System pressure has exceeded 2735 psig, l reduce the Reactor Coolant System pressure to within its limit within l 5 minutes, and comply with the requirements of Specification 6.5.1. . l l I l l l BYRON - UNIT 1 2-1
\
l
~
P!!00F & REVIEW C9PY 670 _ . 660 ' s 650
* ' *.00 PSI A \ \
I 640 2250 PS!A u N E' . I
= i = i > a,0
( I
= *9 2000 PSI A ,
l ! ua 620.l , , , C l
$ 1350 PSIA . ,
4 3 ! I 3 s '> ' 61a l l i : i i i 600 l ' l i i- .
! 1 l '
I 3c, l I I l
l. . ;
' l I l 1
_ l - 530 20
.0 60 30 100 120 0
P O.*.E M '-6) l FIGURE 2.1-1 je[ gem b Sph REACTOR CORE SAFETY LIMIT - """ ' """' ' ""-^~~ Icop Shokd beddM 2-2 SYRON - UNIT 1
~ -- -- - -
l- -- - - . _ . - . _ _ _ _ _ _ _ _ _ __ _ __ __
PRDD & RMEW COPY (
% eay"*A*'% 4 M -
Figu 2 left' blank pending NRC
approval of,ttiree-loop operation. -
f x
/ /
t l !s BYRON - UNIT 1 2-3
-e . _ _ . .
2.1 SAFETY LIMITS ( BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the WRB-1 correlation. The WRB-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB. , The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated transients is limited to 1.34 for a typical cell and 1.32 for a thimble cell. This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
- The curvehf Figure / 2.1-1 end-3.4 3. show 5the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.34 for a typical cell and 1.32 for a thimble cell, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
Thesecurvesarebasedonanenthalpyhotchannelfactor,Fh,of1,55 and a reference cosine with a peak of 1.55 for axial power shape. An allowance isincludedforanincreaseinFhatreducedpowerbasedontheexpression: ( Fh=1.55[1+0.3.(1-P)] Where P is the fraction of RATED THERMAL POWER. These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f1 (AI) function of the Overtemperature trip. When the axial power
\
BYRON - UNIT 1 B 2-1
,-,-y-, -- -,--% --- ,.i-y, , ,~-r%-, - - - ,,w, , - - w ,...,,m-,,.,__,-,--,,,,,,,,,,,-,.y-.#,,,-,.e__- --, , , - - - ,-- --~,
REACTIVITY CONTROL SYSTEMS P?00F & REV13Y CCPY ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length shutdown and control rod drop time from the fully withdrawn position shall be less than or equal to 2.2 seconds from beginning of decay of stationary gripper. coil voltage to dashpot entry with:
a. T,yg greater than or equal to 550*F, and

b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2. ACTION:
a. With the rod drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
. With d drop time within limits _byt determi withtj}rej rgactor c lant pumps hperating d eration proce d provibd WER M restr cte do less thanA equal to 6% of
[ ERMAL RATED THERMnL POWER. b SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:
a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and

c. At least once per 18 months.
O BYRON - UNIT 1 3/4 1-19 l
PR00f & REYlEW COPY REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figurey 3.1-1.=d 0.1 r. , APPLICABILITY: MODES 1* and 2*#. ACTI0K: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:
a. Restore the control banks to within the limits within 2 hours, or

b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above figures, or

c. Be in at least HOT STANDBY within 6 hours.
l ( SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual i rod positions at least once per 4 hours. l "See Special Test Exceptions 3.10.2 and 3.10.3.
#With K,ff greater than or equal to 1.
O BYRON - UNIT 1 3/4 1-21
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RELATIVE POWER (Percem) l i i 1 1 l FIGURE 3.1-1 R00 SANK INSERTION LIMITS VERSUS THERMAL POWER dp P0tfR -6004LO9FRATTnK l BYRON - UNIT 1 3/4 1-22 0
hb b i ICOPY ftp
/
Figure 3.1- efLblank'/pending NRC approval of three-loop - operation N 0 x O BYRON - UNIT 1 3/4 1-23
~( en s U =
8 z TABLE 3.2-1
' DNB PARAMETERS C '[a e PARAMETER iM rrS -LI"IIS
fotw-Laap

irc W
-0;,;r.,uw _ , _ . _ . . _ -
Reactor Coolant System T avg ~ ~ ~ " - i -< 592*F Pressurizer Pressure > 2220 psia * ~ j
; E to U
i i
k ~O
. =
1 C:2 s.: 3 m c-2
U
'! r r-r1 j ~4 1
^ Limit not applicable during either a TilERMAL POWER ramp in excess of 5% of RATED TilERMAL M O
j POWER per minute or a TilERMAL POWER step in excess of 10% of RATED THERMAL POWER.
~we vei= iert sie# n=ew nec agg=i or t.1.4. w, ... .u ;a.
I
TABLE 3.3-1 E
REACTOR TRIP SYSTEM INSTRUMENTATION g .
' MINIMUM CHANNELS APPLICABLE g TOTAL NO. CHANNELS ACTION q FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES e 2 1, 2 1
1. Mar.uol Reactor Trip 2 1 4 2 1 2 3*, 4*, 5* 10

2. Power Range, Neutron Flux High Setpoint 4 2 3 1, 2 2#
a.
b. Low Setpoint 4 2 3 1###, 2 2#
^ P'ower Range, Neutron Flux '4 2 3 1, 2 2# 3. High Positive Rate , i 2 3 1, 2 2# y 4. Power Range,. Neutron Flux l t '4 ) _
High Negative Rate 1 w t "
i E 5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3
6. Source Range, Neutron Flux 2 1 2 2## 4 4

a. Startup

b. Shutdown (Trip Brk Closed) 2 1 . 2 3*, 4*, 5* 10 _
Shutdown (Trip Brk Open) 2 0 1 3, 4, 5 5 1 c. 93
* :: 2 c3
7. Overtemrierature AT t. 3 3 1, 2 6#

e Fuu, -L

p Oper:t ic; 4
^^
2
^^
7 F
A !;.. mm Le p Opcr:tisr.
..:s

8. Overpower AT CU Fou, Luep Operatir, -
4 2 3- 1, 2 6# ::sE
.s. ** -' ** ^^ "
I;,, om L;;p Op;r tis;; A C3 mq
-q I
! - . . i b ._
~ ~ ~ ~ ~ ~ ~
PR00f & HEW COPY TA8LE 3.3-1 (Continued) _ _ _ , TABLE NOTATIONS
*With the Reactor Trip System breakers in the closed position and the Control Rod Drive System capable of rod withdrawal. Jeb.e h't ib.m p.. 5 ; MC .pp d c' th.o, Mes c;e--tien. - #The provisions of Specification 3.0.4 are not applicable.
N8elow the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. N#8elow the P-10 (Low Setooint Power Range Neutron Flux Interlock) Setpoint. ACTION STATEMENTS _ ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERA 8LE requirement, restore the inoperable channel to OPERA 8LE status within 48 hours or be in HOT STAN08Y within the next 6 hours. 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 hour;

b. The Minimum Channels OPERA 8'.E requirement is met; however, the inoperable channel may be bypassed for up to 2 hours for surveillance 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 Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the
, QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2. ACTION 3 - With the number of channels OPERA 8LE one less than the Minimum Channels OPERA 8LE requirement and with tne THERMAL POWER level:
a. Below the P-6 (Intermediate Range Neutron Flux Interlock)
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 OPERA 8LE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER. I
! BYRON - UNIT 1 3/4 3-5 ..,, ,' . ...,.~~-'~.Z-
N I 3/4.7 PLANT SYSTEMS f))}{ { { D} {gpy 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION j l 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-3. APPLICABILITY: MODES 1, 2, and 3. ACTION:
a. With four reactor coolant loops and associated steam' generators in operation and with one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the
next 6 hours and in COLD SHUTDOWN within the following 30 hours.
br Wi reactor coolant loops and associated steam generators in operati + one or more main steam line Co safety valves associated with an o a loop inoperabJer-o eration in MODES 1, 2, and 3 may proceed provid o c it1Hn 4 hours, either the inoperable valve ', ed to OPERA tatus or the Power Range NeutronEurtitg6 Trip Setpoint is reduced pe le 3.7-2; otherwise, be l (at least HOT STANDBY within the next 6 hours a i COLD 4HUTDOWN within the following 30 hours. bf. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5. BYRON - UNIT 1 3/4 7-1
( PR00F & DH COPY TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES-00 RING FCUR LOOP OPERATION MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE . SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER)
, 1 87 2 65 3 43 TABLE 3.7-2 i MAXTMUFAtiOWABLE POWER RANGE NEUTRON FLUX HI ETPOINT WITH INOPERABLE STEAM D NE SAFETY VALVES DURIN ITHREE LOOP OPERATION x
IMUM ALLOWABLE POWER RANGE MAXIMUM NUMBER OF INOPERABLE [ SAFETY VALVES ON ANY NGUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR */ (PERCENT'DFsRATED THERMAL POWER)
\**
1
==
3 O L les=L iwu soiety Valves assali WUPERAtiLt, un ine W r-a @ te d
-94444Nii.wa. /'**Tfiele-vallits-leNWrig NRS-appr$vahlf tfifee loop-operatiga. e
( BYRON UNIT 1 3/4 7-2
t PR005 & R' MEW CDI)Y ( 3/4.7 PLANT SYSTEMS i BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES
. The OPERABILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be limited to within 110% (1320 psia) of its design pressure of 1200 psia during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 102% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 17.958 x 108 lbs/h which is 119% of the total secondary steam flow of 15.135 x 108 lbs/h at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowabic THERMAL POWER restriction in Table 3.7-1. STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced i Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions are derived on the following bases: Te. . ivvp upm . ier.. --
SP = (X) - (Y)(V) x (109).
For three operation: SP = )~(
X 1 l
Where: SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER, V = Maximum number of inoperable safety valves per steam line, U = Maximum number of inoperable safety valves per operating steam line, BYRON - UNIT 1 B 3/4 7-1 l
f PLANT SYSTEMS
PRODF & REVIEW COPY :
,,,,g l SAFETY VALVES (Continued) i 109 = Power Range Neutron Flux-High Trip Setpoint for four loop operation, = Maxt- perce TED THERMAC7 WER permi le by -8 Se cip pending thr C approval o lo operat ree loop o
n. JN ation, val lef Q ]j X = Total relieving capacity of all safety valves per steam line in 1bs/ hour, and ,
i Y = Maximum relieving capacity of any one safety valve in ' lbs/ hour. 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss-of-offsite power. The motor-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 740 gpa at a pressure of 1450 psig to the entrance of the steam generators. The diesel-driven auxiliary feedwater pump is capable j of delivering a total feedwater flow of 740 gpa at a pressure of 1450 psig to : the entrance of the steam generators. This capacity is sufficient to ensure i that adequate feedwater flow is available to remove decay heat and reduce the ( Reactor Coolant Systas temperature to less than 350*F when the RHR System may i be placed into operation. i 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANOBY conditions for 9 hours with steam discharge to the atmosphere concurrent with total loss-of-offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. 3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident 1 gpa reactor to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses. BYRON - UNIT 1 8 3/4 7-2 l l l 1
ATTACHMENT Z These changes delete all reference to three loop operation. This mode.of operation is not supported by existing safety analyses.
'l h&!
8376N' 4
IW %%eL % ~gyzq MAR 2 81984 1 MEETING / TRIP REPORT Docket No.; 40-8027 - Applicant: SequoyahFucisCorporation(SFC) Gore, Oklahoma March 20-21, 1984 - - Date: Place: Oak Ridge National Laboratory (ORNL) Oak Ridge, Tennessee Attendees: HRC - E. Shum, M. Rhodes Oak Ridge National Laboratory - Norm Hinkle John Witherspoon Richard McLean
Purpose:
To discuss and resolve internal staff comments. on the draft Environmental Impact Appr&isal M (EIA) for Sequoyah Fuels Corporation, Sequoyah M Facility, Gore, Oklahoma. g n Summary: On March 21,1984, NRC staff met with ORNL staff to discuss questions and comments raised EW during the internal review of the SPC Draft EIA. O The EIA is being prepared by the ORNL Environmental Review Team as part of the renewal action of SFC's Source Material License SUB-1010. The commcnts will be incorporated into the EIA and another draft will be dubmitted to the Uranium Process Licensing C . Section for staff review. ;1 _N Origi.m1 sigucun 'o ' g E. Y. Shq , , . . _ DISTRIBUTION 4 Q Docket No. 40-8027 E. Y. Shum =- Nt1SS R/F Uranium Process Licensing Section 4 FCUP R/F Uranium Fuel Licensing Branch x VLTharpe DivisioncdffFuel Cycle and - Beveridge/ Cornell 1-23 Material Safety -Eih= W PDR MXRC ]; RHODES W g EYShum M. J. Rhodes Uranium Process Licensing Section g+ MJRhodes Uranium Fuel Licensing Branch j_ r Division of_FuelCycle_and _-- ---
FCUP N FCUP Mb FC g Material Sa ety .hFCUP.g g

MJRhodes/as EYShum VL arpe i l WTCrow .
mi8c m e er j i m e< , urv8a _ _ut _ _ _1 _ g}}

ML20088A456

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