Text

Forwards Progosed Changes to Final Draft Tech Specs. Justification for Changes Provided W/Each marked-up Section
	ML20098F822
Person / Time
Site:	Byron
Issue date:	09/19/1984
From:	Tramm T; COMMONWEALTH EDISON CO.
To:	Harold Denton; Office of Nuclear Reactor Regulation
References
	9205N, NUDOCS 8410030423
	Download: ML20098F822 (63)
	v • d • e

_.

-/

h Commonwealth Edison f

) one First Nation 11 Pitza Chictgo Ittinois g C Address Reply to: Post Othee Box 767 Nj] Chicago, Illinois 60690 September 19, 1984

-Mr., Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S.-Nuclear Regulatory Commission Washington,'DC 20555 Sub ject:

Byron Generating Station Units 1 and 2 Technical Specifications NRC Docket Nos. 50-454 and 50-455 Reference (a):

August 27, 1984 letter from B. H. Youngblood to D. L. Farrar.

Dear Mr. Denton:

This is to provide proposed changes to the final draft version of the Byron 1 Technical Specifications that was distributed in reference (a).

NRC review of the ~ specific changes proposed here is necessary before the Technical Specifications can be finalized.

Attachments 1 through 19 to this letter contain marked-up pages of various sections of the Technical Specifications.

The justification for the changes is provided in each attachment.

We understand that the NRC will review'each of these proposed changes and inform Commonwealth Edison of their acceptability.

.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, I

.AAAW T. R.-Tramm Nuclear Licensing Administrator Im cc:

Byron Resident Inspector 8410030423 840919 m

gDRADOCK 05000454 V

(j j.

9205N PDR

(

F ATTACMENT 1 Byron Station proposes to modify Surveillance Requirement 4.7.7.d.3, on the Non-Accessible Area Exhaust Filter Plenum Ventilation System (page 3/4 7-19) as indicated on the attached copy.

Justificatio_n The deletion of " greater than or equal to 1/4 in, water gauge" is requested because sections 6.5.1.1.2b, 9.4.5.1.la (items 1 and 7) and 9.4.5.1.214 of the i

FSAR as well as section 9.4.3 of the February 1982 SER states that these rooms are to be maintained at a negative pressure. No specific value for negative l

pressure is provided in these references. However, Table 3.11-2 of the FSAR l

indicates that the pressure in water gauge should be maintained in a range typically between -0.25 to 0.0.

Cuttent wording in the Technical Specificaticns is more testrictive than what is assumed in the PSAR.

Byton Station in their Pre-Operational Test 2.84.11, Auxiliary Building t

Ventilation Test, will verify the BOCS Equipment Rooms to be at a negative pressure of at least 1/8 in.

(1/4 in i 1/8 in) water gauge relative to the adjacent clean areas in the normal operating mode and in the accident orde (LOCA coincident with LOEP) with only the booster fans operating. The equipment rooms include:

1A and IB Safety Injection Pump Rooms 1A and IB Centrifugal Charging Pump Rooms 1A and IB Containment Spray Pump Rooms IA and IB Residual Heat Removal (RHR) Pump Rooms i

1A and IB RHR Heat Exchanger Rooms

(

It is also requested that the BCCS Equipment Rooms negative pressure be j

verified relative to the adjacent clean areas rather than the outside l

atmosphere. Sections 9.4.5.1.la3 and 9.4.5.1.2a6 of the FSAR as well as section 9.4.3 of the SER state that the system controls radioactivity by supplying air from the clean areas to areas with greater potential for contamination. Maintaining a negative pressure in the BCCS Equipment Rooms '

ensures that air flows from the adjacent clean areas into the potentially contaminated equipment rooms to control the spread of contamination. The '

system design does not include installed instruments 4 1ch compete the pressure differential between the BOCS Equipment Rooms md the outside s

l atmosphere. To make this measurement, temporary tubing would have to be installed from the BOCS Equipment Rooms to the outside atmosphere and therefore is not a recommended solution. The current system design allows the 80CS Equipment Room pressure to be compared relative to the adjacent clean areas and this is why the change has been requested.

s

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i l

AUG 2 e 1984 FIN 8l D H H

~

PLANT SYSTEMS

(

, SURVEILLANCE REQUIREMENTS (Continued) 3)

Verifying a system flow rate of 66,'900 cfm 10% through the exhaust filters plenum during operation when tested in accord-ance with ANSI N510-1975.

I c.

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

- within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52 Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, when the average of the methyl iodide penetration for the three samples is less than 7.1%

l d.

At least once per 18 months by:

l 1)

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks of less than 6.0 inches Water Gauge while operating the exhaust filters plenum at a flow I

rate of 66,900 cfm 1105, and 2)

Verifying that the exaust filters plenum starts on manual initiation or Safety Injection test signal.

l 4

l i (( -

3)

Verifying that the system maintains the ECCS equipment rooms at a negative pressure M ;r::t:r O... ;r ;,.;l te 1/t t.

t;,"-

? ::r;: relative to the =t:": e r x p.;r; during system operation.

odjacee denn orees e.

After each complete or partial replacement of a HEPA filter bank, by verifying that the exhaust filters plenum satisfies the in place penetration testing acceptance criteria of less than 1.05 in accord-

+(

ance with ANSI N510-1975 for a 00P test aerosol while operating at a flow rate of 66,900 cfm i 105; and f.

After each complete or partial replacement of a charcoal adsorber bank, by verifying that the exhaust filters plenum satisfies the i.

in place penetration testing acceptance criteria of less than 1.05 l(

remove greater than in accordance with ANSI N510-1975 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 66,900 cfm ! 105.

t

.l A

BYRON - UNIT 1 3/4 7-19

ATTAQ0ert 2 Byron Station propoces to modify Surveillance Requirement 4.3.4.2, on Turbine Overspeed Protection (page 3/4 3-75) as indicated on the attached copy.

Justification h is change is requested based on an operation and Maintenance Memo 041 issued by Westinghouse.

Following a Westinghouse teview of testing frequency and performance data fton turbine and component incidents records and a 1982 survey of utilities operating Westinghouse nuclear turbines, it was concluded that there was no significant difference in failure cates between valves tested weekly and those tested monthly. Westinghouse also noted that a monthly versus a weekly valve testing frequency may be beneficial because it reduces the time a plant is operating in a " transient state". Westinghouse recommended that the throttle, governot, interceptor and teheat stop valves be tested monthly.

With the requested change, Surveillar.:.e Requirements 4.3.4.2 a and c would be performed on the same frequency and therefore were combined. It is also requested that the words "During turbine operation" be added since if the unit were shutdown for any period of time it would be unnecessary to verify every month that the Turbine Overspeed Protection System was OPERAEi.E.

(0548M)

INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION f

LIMITING CONDITION FOR OPERATION M'OM d

3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

l With one throttle valve or one governor valve per high pressure turbine a.

1 steam line inopersble and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoperable.

restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , l

or close at least one valve in the affected steam line(s) or isolate l

the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With the above required Turbine Overspeed Protection System otherwise

)

inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months isolate the turbine from the steam supply.

j r

SURVEJLLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.

4.3.4.2 t(

The above required Turbine Overspeed Protection System shall be demonstrated OPERABLE:

pu,;q +oc W a. operuHow

.31 d;,.ee obserwhow of %e rnovement-of %4

a. A a#t least once per ?' days by,yt!'n; cach ef the fwlle !3 valves bel.w through at laest one complete cycle from the running position:

1)

Four high pressure t'urbine throttle valves, 7

2)

Four high pressure turbine governor valves, 3)

Six turbine reheat stop valves, 4)

Six turbine reheat intercept valves, and by cycling

,b.

Witin 7 days prior to entering MODE 3 from MODE 4,4each of the 12 extraction steam nonreturn check vale s :h:!' be cyc!:d from the L

closed position.

t At lest cara p='

11 daye hy dir;;t sw.civauvis vi thesev::ntOf C OOCh Of the n'www i;l'/05 th"0"7 000 009 3 c i.c sywic f e ei. th

"'Jaa#ng *-

w Duri$ v%e opfebiti:n," -

e::

i C # AaAt least once per 31 days by direct observation, ve*4y freedom of

, S-movement of^the 12 extraction steam nonreturn check valve weight arms.

eech of i

d,e:

At least once per 18 months by performance of, CHANNEL CALIBRATION on the Turbine Overspeed Protection Systems,ind

e. A At least once per 40 months by disassembling at least one of each of the valves given in Specifications 4.3.4.2a. and b. above, and per-t-

forming a' visual and surface inspection of valve seats, disks and l

stems and verifying no unaccept'able. flaws or corrosion.

k BYRON - UNIT 1 3/4 3-75 L

ATTACHMENT 3 Byron Station is requesting a change to Technical Specification 3.9.4,

" Containment Building Penetrations" as shown on the attached copy.

This change is requested by Byron Station for the following reasons:

1.

The removal of the equipment hatch during refueling outages will reduce occupational exposure by eliair.ating the wait time for passage j

l through the personnel hatch and thereby augment ALARA.

2.

It will reduce the outage time by eliminating unnecessary movement of the equipment t)atch and provide ceady and convenient access to containment.

3.

It will reduce maintenance on the equipment hatch by reducing the frequency of removal and installation.

Justification A fuel handling accident in the Fuel Handling Building has been addressed in l

Section 15.7 of PSAR, and additional information presented in response to Duestion 311.13. Therefore, the only potential accident affected by the proposed change to the Technical Specification would be a fuel handling accident in containment.

This accident was addressed in response to question 311.2, where it was shown that the containment radiation monitocing system would cause the containswnt purge isolation valves to close before any significant radiation would be released from containment.

Since the fuel handling accident in containment assumes the same scenario as a fuel handling accident in the Fuel Handling Building, the same radioactive release would occur in containment as was assumed for a fuel handling accident in the Fuel Handling Building. However, with the equipment hatch removed, this volume of radioactive gas would now be diluted by the volume of air in both containment ~and the Fuel Handling Building, thus reducing the concentration below that assumed for a fuel handling accident in the Fuel Handling Building. Thus the consequences of a fuel handling accident in the Fuel Handling Building envelope the consequences of a fuel handling accident in containment with the equipment hatch cemoved.

No analysis has been done to confirm that a release of radioactive gas in containment would activate the radiation monitoring system in the Fuel Handling Building, and thereby route the released activity through the Fuel l

Hardling Building emergency exhaust system. Therefore, when the equipment hatch is removed, and fuel is being tundled in containment, the Fuel Handling Building emergency exhaust system must be operating.

l (0548M) l

~. -.

ATTAQGENT 3 (Continued)

In summary, the proposed change will not involve an increase in the probability or consequences of accidents previously considered, ard does not reduce any previously considered safety margin. Therefore, thete is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner. The proposed change will, however, allow Byror. Station to reduce occupational exposure and reduce the time required during a refueling outage.

(0548M)

)

REFUELING OPERATIONS C

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS g gg g I

LIMITING CONDITION FOR OPERATION l

I 3.9.4 The containment building penetrations shall be in the following status:

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

one time and the equipment hatch shall be in place and held by a minimum of four bolts, or (I

c W.

A minimum of one door in the personnel emergency exit hatch is l

closed, and f

o F.

Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

{D 1)

Closed by an isolation valve, blind flange, or manual valve, or 2)

Capable of being closed by an OPERABLE automatic containme'nt purge isolation valve.

APPLICABILITY:

During 64HHi=Atf5 rat!Citiw movement of i-W fuel within D^

the containment.

l ACTION:

With the requirements of the above specification not satisfied, immediately l

suspend all operations involving COREM*SMMeNs ts movement of 2.

.J

D fuel in the containment building.

l SURVEILLANCE RE0VIREMENTS

D 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 OPERABLE automatic containment purge isolation valve within l

100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of and at least once per 7 days during CORE g

ALTERATIONS or movement of irradiated fuel in the containment building by:

a.

Verifyirg the penetrations are in their closed / isolated condition, or b.

Testing the containment purge isolation valves per the applicable g

portions of Specification 4.6.3.2.

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ATTAatqENT 4 Byron Station proposes to modify Technical Specification 4.6.4.1, " Hydrogen Monitors" as indicated on the attached copy.

Justificatio_n In surveillance requirement 4.6.4.1 a) and b), the term " nominal" is ambiguous and does not indicate the variation about the one and four volume percent H2 that is acceptable when purchasing gas samples. In addition, the cuccent surveillance proceduce used at Byron Station to calibcate these hydrogen monitors requires five gas samples, with hydrogen contents which vary from zero percent H2 to greater than 20 percent H, balance N. This 2

2 proceduce provides a more accurate calibration than can be obtained with two gas samples which only vacy in hydrogen content from one to four volume percent. The proposed change to the Technical Specification will allow the more accurate proceduce to be used, while allowing some flexibility in the purchase of gas samples.

(0548M)

O

'AUB 2 8 24 M ha b?

?

CONTAINMENT SYSTEMS

[ifi kre _.

CLT 5

3/4.6.4 COMBUSTIBLE GAS CONTROL O

HYOROGEN MONITORS LIMITING CONDITION FOR OPERATION O

3.6.4.1 Two independent containment hydrogen monitors shall be OPERABLE.*

APPLICABILITY:

MODES 1 and 2.

ACTION:

O a.

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

b.

With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at leas. HOT STANOBY 3

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK and a check that the monitor is in standby mode at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and at least once per 92 days 6 by performing a CHANNEL O

CALIBRATION using e d : ;:: -

-t

{ -i ;-

e a.

Nomidal on volum perce hydro en, ba nce n' roge, and i

b.

inal our vo uma p. cent h drogen, balance' nitr gen.

O "The monitors must be in standby mode to meet the requirement in NUREG-0737, Item II.F.1.6.

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O 4u p - p & W & O~t^@

A & JL,n + r~~tfr%~ M O,

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BYRON - UNIT 1 3/4 6-23 O

ATTAQ9amT 5 Byron Station proposes to change Table 3.7-6 of Technical Specification 3.7.12, " Area Temperature Moiiltoring" by deleting certain areas previously iccluded.

Justification The APPLICABILITY statement of this specification requires temperature monitoring only.on components and systems which are required to be OPERABLE.

Byron Station has corducted a review of all areas previously identified in Table 3.7-6 and deleted those areas which do not contain components or systems required to be OPERABLE as defined in Technical Specifications.

4

-(0548M)-

o AUS 2 8 984 o

FmatDRAFT 3/4.7.12 AREA TEMPERATURE MONITORING O

LIMITING CONDITION FOR OPERATION

j,;ni ts ow r b e ex a eeci*el

.n 3.7.12 The temperature of each area shown in Table 3.7-6 shall 5: ::'-t ir.ed-v m4*hin tha'1#=#+- #-'#2*-d in Tahla 1 7-Liri~ n '5 TcOfs) o~ ~fy %s(y'e i

"Nr-O* * " 3 Y en er<.

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APPLICABILITY:

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

O ACTION:

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

Table 3.7-6 for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the

,O temperature in the affected area (s) exceeded the limit (s) and n l

analysis to demonstrate the continued OPERABILITY of the affected equipment.

b.

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

SURVEILLANCE REQUIREMENTS O

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

lO Q

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TABLE 3.7-6 AREA TEMPERATURE MOWITORING AREA TEMP. *F 1.

Misc. Electric Equipment and Battery Rooms 108 2.

ESF Switchgear Ras 108 3.

Division 12 Cable Spreading Rs 108 4.

Upper and Lower Cable Spreading Ras 90

5. ' Diesel-Generator Ras 132 6.

Diesel Oil Storage Rooms 132 7.

Aux. Building Vent Exhaust Filter Cubicle 122 8.

Centrifugal Chargirg Pump Room 122 9.

Containment Spray Pump Rooms 130

10. RHR Pump Rooms 130

11. Safety Injection Pump Room 130 (0548M)

ATTAQ5 TENT 6 Byron Station proposes to modify item 6b on Table 3.3-1 Reactor Trip System Instrumentation (pg 3/4 3-2), Table Notations (pg 3/4 3-5) and Action Statement 5 (pg 3/4 3-6), as indicated on the attached copies.

i Justification This change is requested to accurately reflect the operation of the Boron Dilution Protection System. The change includes an action statement if two channels of the source range neutron flux monitors are not operable in Modes 3, 4 and 5 and allows for the block of the Boron Dilution Protection System when rods are being withdraun.

In Mode 3, Byron procedure BGP 100-2 allous the operator to manually block both trains of the Boron Dilution Protection System (BDPS) prior to withdrawing the shutdoun banks. Current wording in the Technical Specifications does not allow the BDPS to be blocked which could result in actuation of the system when a Reactor Startup is commenced. Therefore, double asterisks have been added to Modes 2, 3, 4 and 5 to allow for the block of the Boton Dilution Protection System when rods are being withdraun. Block l

switches are provided on the main control board to perform this function. By design, this action disables only the automatic switchover of the charging I

pump suction from the VCT to the RMST on a flux doubling signal. The Reactot Trip and all alarms generated from the source range channels are neither bypassed nor blocked.

In addition, there is currently no action statement.for loss of both Scurce Range Channels. In Mode 5 the provisions of 140 3.0.3 do not apply.

Therefore, it is proposed that Action Statement 5 on page 3/4 3-6 be modified to have an item a. which addresses the situation of the number of operable channels one less than the minimum channels opetable requirement and item b I

was added to address the situation of the number of operable channels two less I

than the minimum channels operable requirement. In the event of the loss of both source range channels which resultr in the loss of both trains of the BDPS, Action 5b ensures the Reactor trir breakers are opened as soon as possible, the valves ir.he path of the boron dilution are secured closed and the Shutdown Margin is verified.

l l

i (0548M)

L e,, n,.n..y,

TABLE 3.3-1 REACTOR TRIP SYSTEM INSlRUMENTATION_

5 MINIMUM g

ac CHANNELS CHANNELS APPLICABLE ACil0N TOTAL NO.

OPERABLE MODES _

OF CHANNEL _5 T_0 TRIP 1

g FUNCTIONAL UNIT 2

1, 2 10 1

3*,

4*, ha q

2 2

1 Manual Reactor Trip 2

1.

28 Power Range, Neutron Flux 3

1, 2 2#

2.

2 l###, 2 4

3 2

High Setpoint 4

2#

a.

Low Setpoint 3

1, 2 b.

2 4

Power Range, Neutron Flux 2#

3.

High Positive Rate 3

1, 2 2

4 Power Range, Neutron Flux, w

3 4.

)

High Negative Rate 2

l###, 2 1

2 m

Intermediate Range, Neutron Flux 4

S.

2## *',** $"

2 5

3".4 m

Source Range, Neutron Flux 2

2 1

6.

1 i

Startup 2

6#

b.

Shutdown 3

1, 2 M

a.

2 4

6#

({D Overtemperature AT 3

1, 2 7.

2 i

4 8.

Overpower AT 6#

pyy l

3 1

Pressurizer Pressure-Low 4

'M 2

9.

(Above P-7)

&,% 1 3

i m q i

c.

O*

i l

E

TABLE 3.3-1 (Continued)

TABLE NOTATIONS Ad6 t. 3 *$[A "With the Reactor Trip System breakers in the closed position and tne Control Rod Drive System capable of red withdrawal.

The provisions of Specification 3.0.4 are not applicable.

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

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

- The Baron Dilutica %Kch'on Spem may be. blocked when rods are beirw) withde.

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 beurs.

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:

The inoperable channel is placed in the tripped condition a.

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance 'cesting 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.2.

ACTION 3 - With the number of :hannels OPERABLE ane less than the Minimum Channels OPERA 81 eaquirement and with the THERMAL POWER level:

'a.

Below the P-6 (Intermediate Range Neutron Fiux Interlock)

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

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

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

BYRON - UNIT 1 3/4 3-5

,,-,.-.-.-,---n-...,-.,--.,

TABLE 3.3-1 (Continuec)

ACTION STATEMENTS (Continued)

AUG c.8 W64 ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirem'ent suspend all operations involving positive reactivity changes.

a.One ACTION With the number of OPERABLE channels:em less than the Minimum Channels OPERABLE requirement restore,the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or within the next hour open the l

reactor trip breakers, suspend all operations involving positive reactivity changes, and verify valves ICVil19, 1CV6428, ICV-8439, 1CV-8441 and 1C-8435 are closed and secured in position.

b..s ** Insert A A "

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

(

a.

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

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

ACTIOE7-WiththenumberofOPERABLEchannelsonelessthantheTotal Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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

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

ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERA 8LE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open.the Reactor trip breakers within the next hour.

ACTION 11 - With the number of OPERABLE cnannels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

BYRON - UNIT 1 3/4 3-6

r n sert- "A /4 2

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ATTacteerr 7 Byron Station proposes to change item 7b on Table 4.3-2 (page 3/4 3-38) of

. Technical Specification 3.3.2 as shown on the attached copy.

Justification There is no provision in the current design for testing the automatic opening of the contairent Sump suction isolation valves on a RWST level low-low signal coincident with a safety Injection Signal by means of an ACTUATION LOGIC TEST. The testing of.this function should be performed undet the ANR14G CHAkhiEL OPERATIONRL TEST. This testing would vecify the operability of the trip function and the proper setpoint. This change makes Table 4.3-2 consistent with NUREG-0452 Rev. 4 Standard Technical Specifications.

(0548M) i

4 TABLE 4.3-2'(Continued)

Eg ENGINEERED SAFEIY FEATURES AC10ATION SYSTEM INSTRUMENTATION-2 SURVEILLANCE REQUIRER Ni$_

~

g IRIP.

q ANALOG ACTUAllNG M(M)L S g.

CHANNEL DEVICE MASIER SIAVL 10R WillCil

. CHANNEL CalANNEL OPERATIONAL OPERATIONAL AClUATION HELAY RELAY SURVElttANCE FUNCTIONAL UNIT.-

CHECK CALIBRATION TEST TES1 LOGIC TESI TESI IEST

,lS REQUIRED

7. Automatic Opening (Continued) a.

Automatic Actuation Logic and Actuation Relays N.A.

N.A.

M(1)

Q 1, 2, 3, 4 b.

RWST Level-tow-Low N.A.

R

.Ek M N.' A.

~ N. A.

N.A.

1, 2, 3, 4 Coincident With Safety injection See Item 1. above for all Safety injection Surveillance Requirements m

D 8. Loss of Power d.

ESf Bus Undervoltage N.A.

R N.A.

N.A.

1, 2, 3, 4 b.

Grid Degraded Voltage N.A.

R N.A.

N.A.

1, 2, 3, 4 9.

Engineered Safety l-eature Actuation System Interlocks a.

Pressurizer Pressure, N.A.

R H

.N.A.

N.A.

1, 2, 3 P-11 b.

Reactor Irip, P-4 N.A.

N.A.

R N.A.

N.A.

1, 2, 3 Low-Low 1,yg, P-12 N.A.

R M

N.A.

I, 2, 3 c.

d.

Steam Generator Water S

R H

N.A.

M(1)

N(1)

Q 1, 2, 3 I*

,k"*

Level, P-14 l

(lligh-liigh) c.3%

IABLE NOIATION E

l (1)

Each train shall be tested at least every 62 days on a STAGGERED 1ESI BASIS.

8D

? %M i

.h

ATTAcimENT 8 Byron Station proposes to change Table 4.11-2, Radioactive Gaseous Waste Sampling and Analysis Program (page 3/4 11-10) es shown on the attached copy.

Justification The requirement for monthly analysis of tritium in the Containment Purge should be deleted from item 2 of Table 4.11-2 because any containment purge effluent released from the plant exits via the Auxiliary Building vent stack where sampling and analysis of tritium will be provided in accordance with iten 3 of Table 4.11-2.

Since at Byron Station containment purge is not directly connected with the outside atmosphere it would be redundant to sample tritium in the containment purge anu' then in the Auxiliary Building vent stack.. In addition, the containment purge monitor at Byron Station is not designed to detect tritium.

The requirement of continuous sampling of containment Purge should also be deleted from item 4 of Table 4.11-2 because continuous sampling is performed by the Auxiliary Building vent stack monitors which includes contributions from containment purge.

(

(0548M)

e lA8LE 4.11-2 Eg RAD 10ACllVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM z

e c

HINIMUM LOWER 1IMIT Of

(

SAMPLING ANALYSIS TYPE OF DETECIl0H (LLD)IIs GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (fi/mi) y 1.

Waste Gas Decay Each Tank Each Tank Principal Gamma Emitters (2) lx10~4 Tank Grab Sample I3)

I3)

I2) lx10

2.

Containment Purge Each PURGE Each PURGE Principal Gamma Emitters Gra5 Sample

-6 o

-H j: 3 ixiG I4)Ib)

-4 3.

Auxiliary Bldg M

Principal Gamma Emitters (2) lx10 y

Vent Stack Grab Sample M

-6 (Units 1 and 2) 11-3 lx10 4.

All Release lypes Continuous (6) g(7) l-131 lx10' 2 y

g a s l i s ted i n 4-8--

-10 J-and 3. above.

Charcoal Saingle I-133 lx10 Continuous (6) g(7)

Principal Ganima Emitters lx10 Particulate Sample Continuous (6)

M Gross Alpha Ix10'II Composite

.,, x, Particulate Sample

y*

Continuous (6)

-11 Q

Sr-t19, Sr-90 lx10 g

Composite y

Particulate Sample Continuous Noble Gas Monitor Noble Gases Gross Beta on-lx10 (3

haaraa 4

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i ATTACHMENT 9 Byron Station proposes to modify paragraph 6.2.2.c (page 6-1) as indicated on the attached copy.

Justification It is proposed that the unit organization include two Radiation Cheelstry

' Technicians on site when fuel is in the reactor. This change is consistent with what is currently required in the Byron General Site Emergency Plan (GSEP). Therefore, for consistency we request that this change be made.

l (0548M)

5 M9ALDW.FT W6 2 8 saa

-1.C O ISTRATIVE CCNTROLS C1 RESPONSIBILITY 6.1.1 The Superintendent, Byron Station, snall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility curing his absence.

6.1. 2 The Shift Engineer (or during his absence from the control room, a designated individual) shall be responsible for the control room command function.

A management directive to this effect, signed by the Division Vice President and General Manager-Nuclear Stations shall be reissued to all station personnel on an annual basis.

6.2 -ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown in Figure 6.2-1.

UNIT STAFF S.2.2 The unit organization shall be as snown in Figure 6.2-2 and:

a.

Each on duty shift shall be ccmposed of at least the minimum shift

, crew composition shown in Table 6.2-1; and b.

At least one licensed Operator shall be in the control room when fuel is in the reactor.

In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room; he

Radiation Chemistry Technician',* qualified in radiation protection c.

procedures, shall be on site when fuel is in the reactor; d.

ALL CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation; e.

A site Fire Brigade of at least five members

shall be maintained onsite at all times.

The Fire Brigade shall not include the Shift Engineer, and the two other members of the minimum shift crew necessary for safe shutdown of the unit and,any personnel required for other essential functions during a fire emergency; and "The Radiation Chemistry Technician and Fire Brigade compcsition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence provided immediate action is taken to fill the requi ed positions.

BYRON - UNIT 1 6-1

ATTaceuprr 10 Pyton Station proposes to change Technical Specification 5.3.1 (page 5-4) as shown on the attached copy.

Justification The initial core loading contains some fuel assemblies with enrichments greater than 3.10 but less than 3.20 weight percent U235, The actual enrichment of future reload cotes has not been determined. The upper limit on the enrichment of reload cores is set at 4.0 weight percent U235 by the value used in the criticality analysis for new fuel stocage (PSAR Section 9.1).

't

]

\\

(0548M)

%15 DRMT DESIGN FEATURES (D

o 5.3 REACTOR CORE

'AUS 2 4 ou FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4.

Each fuel rod shall have a O

" "I"'I

'C*i #"'I I'"9th

144 I"Ch'* '"d C "t'I" * ***I""* * **I "'I9ht of 1619 grams uranium.

The initial core loading shall have a maximum enrichment of.3:40 weight percent U-235.

Reload fuel shall be similar in physical design

./q do the initial core loading and shall have a maximum enrichment of 3::$$ weight

6" 5. zo percent U-235.

4. o ON R OD ASSEMBU ES O

5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies.

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

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

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

a.

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

For a pressure of 2485 psig, and O

c.

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

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is g'

12,257 cubic feet at a nominal T,yg of 588.4*F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The seteorological tower shall be located as shown on Figure 5.1-1.

g 10 m BYRON - UNIT 1 5-4 r

ATTACHumT 11 i

Byron Station proposes to modify Table 3.8-2, Motoc-Operated Valves Thermal Overload Protection Devices (pages 3/4 8-30 to 3/4 8-33) as indicated on the attached copies.

Justification Based on discussions with the Nac, it is our understanding that Table 3.8-2 of the Technical Specifications should contain all safety related motoc-operated valves with thermal overload protection devices. As such, Table 3.8-2 has been revised to include all safety related motor-operated valves with thermal overload protection devices.

(0548M)

AU6 2 9 884 TABLE 3.8-2 MOTOR-0PERATED VALVES THERMAL OVERL0A0 PROTECTION DEVICES VALVE NUMBER FUNCTION RC8001A RC Loop 1A Hot leg Stop Valve C80018 RC Loop 1B Hot Leg Stop Valve IR 001C RC Loop 1C Hot Leg Stop Valve k

1RC 010 RC Loop 10 Hot Leg Stop Valve 10G081 H2 Recemo Suction Cnmt. Isol. Valve ICC9438 CC wtr from RC Pumos Thermal Bar Is Valve ICC9416 CC Wtr from RCPS Isol. Valve 10G057A H2 Recomo Cnmt. Isol. Valve Ois "H"

1RC8003A RC Loop 1A Bypass Leg Stop Val' 1RC80030 RC Loop 10 Bypass Leg Stop V ve 1RH8701A C Loop 1A to RHR Pump I

1. Valve

(

1RH8702A R Loop IC to RHR Pump sol. Valve ISI8808A Acc. lA Disch. Iso. Valve ISIS 8080 Accum. 10 Disch.

ol. Valve IRY8000A Pzt. Re 'ef Iso. Valve 1A 1W0056B Chilled W er nmt. Isol. Valve IRC8002A RC Loop 1A d Leg Stop Valve IRC8002B RC Loop Col eg Stop Valve IRC8002C RC Loo 1C Cold L Stop Valve IRC80020 RC L p 10 Cold le top Valve IRC80038 R

cop 13 Bypass Leg top Valve IRC8003C Loop 1C Bypass Leg 5 Valve IRY80008 Pzr. Relief Valve 18 10G080 H2 Recomb Suct. Cnmt. Isol.

Ive 1 WOOS 6A Chilled Water Cnmt. Isol. Valve 10G079 H2 Recomb. Ofsch. Cnat. Isol. Va e ICV 8112 RC Pump Seal Water Return Isol. Va e IRH870 RC Loop 1A to RHR Pump Isol. Valve IRH87 RC Loop 1C to RHR Pump Isol. Valve ISI 088 Accum. 18 Disch. Isol. Valve 1 8808C Accum. IC Disch. Isol. Valve CC9414 CC Water from React. Chg. Pumos Isol. Valve 00G059 Unit 1 Suct Isol Viv H Recomb 2

00G060 Unit 1 Discharge Isol V1v H2 Recombiner 00G061 Unit Ofscharge Xtie for H2 Recombiner 00G062 Unit Xtie on Discharge of H2 Recombiner 00G063 Unit Suction Xtie for H Recombiner 2

00G064 Unit Suction Xtie for H Recombiners 2

00G065 08 H Analyzer Inlet Isol Viv 2

1 AU6 2 e ss4 TABLE 3.8-2 (Continued)

MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION OEVICES VALVE NUMBER FUNCTION 00G066 08 H Recomb Disch Isol Viv 2

10G057A CAH Recomb MIsol. Valve (Disch Y 2

10G079 H2 Recomb Disch. Cnmt. Isol. Valve 10G080 H Recomb Suct. Cnmt. Isol. Valve 2

10G081 H2 Recomb Suction Cnmt. Isol. Valve 10G082 OA H Reconb Disch Cnmt Isol Viv 2

10G083 OA H Recomb Disch Cnmt Isol Viv 2

10G084 OA H2 Recomb Cnmt Outlet Isol V1v 10G085 H2 Recomb Cnmt Outlet Isol Viv Asuf A 1AF013A AF Mtr Orv Pmp Disch Hdr Dwst Isol Viv 1AF0138 AF Mtr Orv Pmp Osch Hdr Owst Isol Viv 1AF013C AF Mtr Orv Pp Disen Hdr Owst Isol Viv IAF0130 AF Mtr Orv Pp Disch Hdr Owst Isol Viv 1AF013E AF Ds1 Orv Pm Osch Hde Owst Isol Viv

( '

1AF013F AF Ds1 Dry Pp Osch Hdr Owst Isol Viv 1AF013G AF Os1 Orv Pp Osch Hdr Dwst Isol Viv 1AF013H AF Os1 Drv Pp Osch Hdr Owst Isol Viv ICCS85 RCP Thermal Barrier Outlet Hdr Cnmt Isol V1v h ed " 1CC9413A RCP CC Supply Dwst CNMT Isol ICC94138 RCPs CC Supply Upst CNMT Isol

" " J CC9414 CC Water from."aeet.. Chg. N. - Isol. Valve 40 1CC9416 CC Wtr from RCPS Isol. Valve S cps R

M"+ E QCC9438 CC Wtr from RC Pumps Thermal Bar Isol. Valve ICS001A 1A CS Pp Suct from RWST 404

1C50018 la CS Pp Suction from RWST 304' t ICS007A 5 7 Pp 1A Disch Line Dwst Isol Viv ICS0078 CS Pp 18 Disch Line Downstream Isol Viv 1CS009A 1A Pump Suction from 1A Recire Sump 1CS0098 18 CS Cont Recirc Sump B Suct Isol Viv to CS 1CS019A CS Eductor IA Suction Conn Isol Viv 1C50198 CS Eductor 18 Suction Conn Isol Viv 1CV1120 MOV RWST to Chg Pp Suct Hdr ICV 112E MOV RWST to Chg Pp Suct Hdr 1CV8100 gyg 4 MOV RCP Seal Leakoff Hdr Isol ICV 8105 MOV Chrg Pps Disch Hdr Isol Vlv ICV 8106 MOV Chrg Pps Disch Hdr Isol Viv ICV 8109 MOV P0 Chrg. Pp Miniflow Recirc. Viv 1CV8110 MOV A & B Chg. pp Recirc Downstream Isol ICV 8111 MOV A & B Chg Pp Recire Upstream Isol ICV 8112 RC Pump Seal Water Return Isol. Valve

AUS 2 g m TABLE 3.8-2 (Continued)

MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES VALVE NUM8ER FUNCTION 1CV8355A MOV RCP 1A Seal Inj Inlet to containment Isol ICV 83558 MOV RCP 1B Seal Inj Inlet Isol 1CV8355C MOV RCP 1C Seal Inj Isol 1CV83550 MOV RCP 10 Seal Inj Isol 1CV8804A MOV RHR Sys X-Tie Viv to Chrgng Pump Suction Hdr A.B.

1RC8001A RC Loop 1A Hot Leg Stop Valve IRC80018 RC Loop 18 Hot Leg Stop Valve IRC8001C RC Loop 1C Hot Leg Stop Valve IRC80010 RC Loop 10 Hot Leg Stop Valve IRC8002A RC Loop 1A Cold Leg Stop Valve IRC80028 RC Loop 18 Cold Leg Stop Valve

-1RC8002C RC Loop 1C Cold Leg Stop Valve IRC80020 RC Loop 10 Lold Leg Stop Valve

-(

1RC8003A RC Loop 1A Bypass Leg Stop Valve IRC80038 RC Loop 18 Bypass Leg Stop Valve IRC8003C RC Loop IC Bypass Leg Stop Valve IRC80030 RC Loop 10 Sypass Leg Stop Valve IRH610 RH PP 1RH01PB Recire, Line Isol.

1RH611 RH PP 1RH01PB Recire, Line Isol.

1RH8701A RC Loop 1A to RHR Pump Isol. Valve IPH8702A RC Loop 1C to RHR oump Isol. Valve 1RH87018 RC Loop 1A to RHR Pump Isol. Valve 1RH87028 RC Loop 1C to RHR Pump Isol. Valve IRH8716A RH HX 1RH02AA Ownstrm Isol Viv t

-lRH86168 RH HX.1RH02AB Ownstrm Isol Valve j

1RY8000A Prz. Relief gI Valve 1A 1RY80008 Prz. Relief (Valv M W~

1SI8801A SI Charging Pump Disch Isol Viv 15I88018 SI Charging Pump Disch Isol Viv l

ISI8802A SI PP 1A Disch Line Owst Cont Isol Vlv 15I88028 SI PP 18 Disch Line Owst Isol Viv ISI8804B SI Pump 18 Suct X-tie from RHR HX 1SI8806 S1 Pumps Upstream Suction Isol ISI8807A SI to Chg PP Suction Crosstie Isol Viv ISI8807B SI to Chg PP Suction Crosstie Isol Viv ISI8808A Accum. lA Disch. Isol. Valve 15I88088 Accum. 18 Disch. Isol. Valve 1SI8808C Accum. 1C Disch. Isol. Valve 15I88080 Accum. 1D Disch. Isol. Valve BYRON - UNIT 1 3/4 8-32

AUS 2 8 $84

&' f'f

J' TABLE 3.8-2 (Continue MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES VALVE NUMBER FUNCTION ISI8809A SI RX HX 1A Osch Line Owst Isol Viv 1SI8809B SI RX HX 18 Osch Line Owst Isol Viv ISI8811A SI Cnmt Sump A Outlet Isol Viv 15I88118 SI Cnmt Sump B Outlet Isol Viv A

1SI8812A SI Rwst to RH Pp IT0utlet Isol Viv 15I8812B SI Rwst to RH Pp 18 Outlet Isol Viv 15I8813 SI Pumps 1A-18 Recirc Line Dwst Isol ISI8814 SI Pump 1A Recire Line'Isol Viv ISI8835 SI Pumps X-tie Disch Isol V1v 15I8840 SI RHR % Disch Line Upstrm Cont Pen Is1 Viv 2

1178821A SI PP 1A Disch Line X-tie Isol Viv j 8821B SI Pump 18 Disch Line X-tie Isol Viv 8920 SI Pump 1B Recirc Line Isol Viv 2

8923A SI PP 1A Suction Iscl Viv i

8923B SI Pump 18 Suct Isol Valve 1

18924 SI Pump 1A Suction X-tie Ownstra Isol Viv 1SX016B RCFC B&D Sx Supply MOV 1SX016A RCFC A&C SX Supply MOV 1SX027A RCFC A&B' Return c.

ISX0278 F.FC B&D SX Return MOV 6A Chilled Wtr Coils 1A & 1C Supply Isol '

1W000.

Chilled Wtr Coils 18 & 10 Sue Viv 1 WOO 20A Chilled Wtr Coils 1A &

eturn Isol Viv 1W0020B tr Coi

& 10 Return Isol Viv 1 WOO 23A Chiller il Cooler Return Viv l

1 WOO 238 C

r 19001CB 01 er Return Viv

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B Chilled Water Cnmt. Isol. Valve Inser bH

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ATTAOWEDIT 12 i

Byron Station proposes to chance Technical Specificatimt 5.6.3 (page 5-5) as shown on the attached copy.

Justification The spent fuel storage pool is designed to store 1050 spent fuel assemblies and 10 failed fuel assemblies for a total of 1060 fuel assemblies. This

~ ~

change is consistent with the PSAR.

4 4

(0548M)

.. -. ~.. -... -... - - - -. -

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=ww FlNA!. DRAFT O

DESIGN FEATURES 5.6 FUEL STORAGE O

gtg 3 g

CRITICALITY 5.6.1 / The spent fuel storage racks are designed and shall be maintained with:

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

unborated water, which includes a conservative allowance of 3.31%

Ak/k for uncertainties as described in Section 9.1 of the FSAR; l

and lO b.

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

l I

5.6.1.f The k,ff for new fuel for the first core load,ing, stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

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

CAPACITY 5.6.3 The spent fuel storage poo.1 is designed and shall be maintained with a storage capacity limited to no more than 20$@ fuel assemblies.

/06o C

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

~O e

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O BYRON - UNIT 1 5-5

-O

3 ATTActeuprf 13 Byron Station proposes to modify paragraph 6.12.2 (page 6-24) of the Technical Specification as indicated on the attached copy.

Justification Byron Station will completely barricade and conspicuously post all entrances

]

1 into individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas, such as i

in containment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area. These areas will be temporarily established as required and Byron Station does not J

have a system for installing flashing lights as a warning device. There is an engineering concern with providing a suitable power supply in such an adverse environment and with the installation of the flashing lights in containment.

In addition, the requirements of the Code of Federal Regulations are met for entry into a high radiation area and there are no additional requirements for the situation described above. It is felt that barricading and posting of the areas would provide adequate control and it is not necessary to provide an additional warning device. Therefore, it is requested that the requirement for a flashing light be deleted.

l l

(0548M) t

CT Y A0 h'

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

Individuals qualified in radiation protection procedures (e.g., Rad / Chem Technician) or

-personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

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

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

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

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency

(

specified in the Radiation Work Permit.

l l

6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to perscnnel with radiation levels greater than 1000 mR/h at 45 cm (18 in.) from i

[

the radiation source or from any surface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Foreman on duty and/or health physics supervision.

Doors shall remain locked except during periods of i

access by personnel under an approved RWP which shall specify the dose rate l

1evels 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, direct or remote (such as closed circuit TV cameras) continuous surveillance

~

i 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 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 car; be reasonably constgr cted around the individual area, that individual area shall be barricaded 4 conspicuously posted, and ; fh2Sg 'P 2-et,e;; u. W,;.e4.... err.h; 2; ice u i

i BYRON - UNIT 1 6-2.4 l

ATTAC15EENT 14 Byron Station proposes to modify pages 6-2, 6-6, 6-7, 6-8, 6-9, 6-10, 6-12 and 6-13 as indicated in the attached copies.

Justification In section 6.2.3 on page 6-2 the name of the Independent Safety Engineering Group (IS8G) has been changed to the Onsite Nuclear Safety Group (OMSG). This change is as a result of a General company order issued by the Chairman and President of Commonwealth Edison Company.

It is also requested that records of activity by the ONSG be forwarded

" quarterly" rather taan "each calendar month". Byron Station intends to document weekly reports and these will be compiled and issued quarterly.

It is felt that issuing quactacly repotts will be more efficient than issuing monthly reports. Quartecly reports are consistent with the program currently established by the office of Nucleat Safety.

In section 6.4.1 on page 6-6 ISBG is changed to the Office of Nuclear Safety.

Within the Commonwealth Edison otganization the responsibility for collecting and documenting operational experience feedback (OPEX) resides with the headquatters Office of Nuclear Safety.

Since Commormealth Edison operates several nuclear units, the headquarters office of Nuclear Safety performs several functions such as the OPEK function to aupnent the on-site groups. Those expeciences determined by the onsite group as needing feedback are focwarded to the group at the headquacters office for formal issuance as an OPER.

In section 6.5.1 on page 6-7 the Supervisor of the offsite Review and Investigative Function is appointed by the Manager of Nucleat Safety and r.ot by the Executive Vice President. This change is requested to reflect Commonusalth Edisons organizational responsibilities.

The changes are made on page 6-9 to reflect the current organizational and functional responsibility at Commonwealth Edison.

The changen; proposed on pages 6-8, 6-10, 6-12 and 6-13 are to cocrect typographical ectors or to maintain corsistency with the functional titles used throughout Section 6.

Several of the changes proposed ace also to maintain consistency with terminology between othec Commonwealth Edison operating units and the Byron Station.

1 (0548M)

ma n h.

- ';..'M I

A 4. 4 M1 g

ACHINISTRATIVE CONTROLS UNIT STAFF (Continued)

Schedo)<ci f.

Acministra*.ve proceduris shall be developed and imolemented to limit the working hours cf unit staff wno perform safety-related functions; e.g., licensea Senior Operators, licensed Operators, health physics personnel, equipment operators, and key maintenance

~

personnel.

'sched* led The amount of/ overtime worked by Unit staff members performing safety-related functions shall be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12).

ONstrE.

Nuca EAA 5AFErry GRopp (ONEG S**ETYCN0!NCECINGGRG0F(ICCO'}--

t 6.2.3

!"0EPf"05"7 FUNCTION l

SNSG 6.2.3.1 The +9E6 shall function to examine plant operating characteristics, l

NRC issuances, industry advisories, REPORTA8LE EVENTS and other sources of plant design and operating experience information, including plants of i

similar design,.which may indicate areas for improving plant safety.

The EGEG-CM45G

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

COMPOSITION ONS4 6.2.3.2 The P999 shall be composed of at least four, dedicateo, full-time engineers located on site.

RESPONSIBILITIES ossq 6.2.3.3 The F9fe shall be responsible for maintaining surveillance of plant activities to provide independent verification" that these activities are performed correctly and that human errors are reduced as much as practical.

RECORDS OHSG 6.2.3.4 Records of activities oseformed by the 16E6 snail be prepared, main-tained, and forwarded :: ' ::I_r. ;r 7.ec.;7 to the Manager of Nuclear Safety, T-qbuerterly and the Superintendent, Byron Station.

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

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

Not responsible for sign-off function.

BYRON - UNIT 1 6-2

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' ADMINISTRATIVE CONTROLS 6.2.4 SHIFT TECHNICAL ADVISOR (Continued)

To assure capability for performance of all STA functions:

(1) The shift foreman (SRO) shall participate in the SCRE shift relief turnover.

(2) During the shift, the shift engineer and the shift foreman (SRO) shall be made aware of any significant changes in plant status in a timely manner by the SCRE.

(3) During the shift, the shif t engineer and the shift foreman (SRO) shall remain abreast of the current plant status.

The shift foreman (SRO) shall return to the control room two or-three times per shif t, where practicable, to confer with the SCRE regarding plant status. Where not practicable to return to the control room, the rhift foreman (SRO) shall periodically check with the SCRE for a olant status update.

The shift foreman (SRO) shall not abandon duties original to reactor operation, unless specifically ordered by the shift engineer.

6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or evceed the minimum qualifi-cations of ANSI N18.1-1971.

The Rad / Chem Supervisor or Lead Health Physicist, who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, for a Radiation Protection Manager.

The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the suppie-mental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees.

6.4 TRAINING

~

6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Production Training Department and shall meet or exceed the requirements and recommendations of Section 5 of ANSI /

ANS 3.1-1978 and Appendix A of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience identified by the ESE6. offke of Nuc/w Jefebj.

6.5 REVIEW INVESTIGATION AND AUDIT The Review and Investigative Function and the Audit Function of a:tivities affecting quality during facility operations shall be constituted and have the responsibilities and authorities outlined below.

BYRON - UNIT 1 6-6

- ~ -. - -

h.i G Q Q ( '[

ADMINISTRATIVE CONTROLS

.7 1

l 6.5 REVIEW INVESTIGAT!ON AND AUDIT (Continued)

OFFSITE Plana 3 w of Nucle.ne.5nfehj 6.5.1: The Superviso the Offsite Review and Investigative Function shall i

be appointed by the m.c tf = Vice Preside..t responsible for nuclear activ-ities.

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

a.

Offsite Review and Investigatise Function The Supervisor of the Offsite Review and Investigative Function shall:

(1) provide directions for the review and investigative function and appoint a senior participant to provide appropriate direction, (2) select each participant for this function, (3) select a complement of more than one participant who collectively possess background a'nd qualifications in the subject matter under review to provide comprehensive interdisciplinary review coverage under

.this function, (4) independently review and approve the findings and recommendations developed by personnet performing the review and investigative function, (5) approve and report in a timely manner all findings of non-compliance with NRC requirements to the Station Superintendent, Division Vice President and General Manager -

4 Nuclear Stations, Manager of Quality Assurance, and the Vice Prosident'-

Nuclear Operations. 'Ouring periods when the Supervisor of.0ffsite Review and Investigative Function is unwailable, he shall designate this responsibility to an established alternate, who satisfies the formal trainina and exoerience for the Supervisor of the Offsite 8# - - Review and Investigatop Function.

The responsibilities of the per-

+

sonnel performing this function are stated below. The Offsite Review and Investigative Function shall review:

1)

The safety evaluations for:

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

Proposed changes to the Quality Assurance Program description shall be reviewed and approved by the Manager of Quality Assurance; 2)

Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in 10 CFR 50.59; 3)

Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59; s

4)

Proposed changes in Technical Specifications or this Operating License; SYRON - i; NIT 1 6-7 g

M 23 @

N,3[l. UdNfi i

2 '; " "

Th1b ACMINISTRf~IVE CONTROLS OFFSITE (Continued) 5)

Noncompliance with Codes, regulations, orcers, Tecnnical Soeci-fications, license requirements, or of internal procecures, or instructions having nuclear safety significance; 6)

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

All REPORTABLE EVENTS; 8)

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

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

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

Nuclear Stations, and Manager of Quality Assurance.

b.

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

Such responsibility is delegated to the Director of Quality Assurance for(0perating ard the Si;f f "... A:: n d= _.u Assurance for}faintenance),q==?'y - - r' g ef Qualityk General S uper i'?=

Either shall approve the audit agenda and checklists, the findings and the report of each audit.

Audits shall be performed in accord-ance with the Company Q'Jality Assurance Frogram and Procedures.

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

1)

The conformance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months; 2)

The adherence to procedure, training, and qualification of the station staff at least once per 12 months; 3)

The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems, or methods of operation that affect nuclear safety at least once per 6 months; 4)

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

M$dM f

t b

b6 64, a.

ADMINISTRATIVE CCNTROLS OFFSITE (Continued) 5)

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

The Facility Security Plan and implementing procedures at

.least once per 12 months; 7)

Onsite and offsite reviews; 8)

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

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

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

10) The Radiological Environmental Monitoring Program and the results i

thereof at least once per 12 months; l

l

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

12) The PROCESS CONTROL PROGRAM and implementing procedures for l

solification of radioactive wastes at least once per 24 months; and

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

Report all findings of noncompliance with NRC requirements and l

Mona0ee recommendations and results of each audit to the Station Superin-tendent,10ir::ter of Nuclear Safety, the Division lice President and General Manager - Nuclear Stations, Manager of Quality Assurance, the Vice Chairman, and the Vice President - Nuclear Operations.

M ana y r-of The Manager of Quality Assurancebp;d 'tk f rep *'+ 4* +h* C ***

e D*8N'"E rt; p;;,., ice e.........

nd tt;- 4,,;;7 a~'-

r;;;in L tL D; sir;;t;r, Nuclear Safety [Either the Manager of

,7

i;;;$; 57,;;;g Quality Assurance or the Su cr"f;;r of the Offsite 3: fie eg meager-oh

.5cf(] or re;que;st any other action whict he deems necessary to avoid Irae ti tiaa Cuncticn has the authority to order unit shutdown Noc.g e gr unsafe plant conditions, i

l BYRON - UNIT 1 6-9

"} r.1 mQ~

A '

UnAt'"!

m 2g m ACMD,*5AU'.E CONTROLS OFF5ITE (Continued) d.

Records 1)

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

and 2)

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

e.

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

Those procedures shall cover the following:

1)

Content and _ method of submission of presentatf ore to the Supervisor of the Offsite Review and Investigative Function, 2)

Use of committees and conseltants, 3)

Review and approval,

~4)

Detailed listing of items to be reviewed, 5)

Method of:

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

Determiningsatisfactorycompletionofactionrequiredba(don approved findings and recommendations reported by personnel performing the review and investigative function.

f.

Personnel 1)

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

a) nuclear power plant technology, b) reactor operations, c) utility operations, d) power plant design, e) reactor engineering, f) radiological safety, g) reactor safety analysis, BYRON - UNIT 1 6-10

) @ 2 A SEta 6

10Y!NIS' m :!E CCNT70LS OFFSITE (Continued) h).

Instrumentation and Control Engineering graduate or equivalent with at least 5 years of experience in instrumentation and control cesign 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 6.5.2 The Onsite Review t.nd Investigative Function shall be supervised by the Station Superintendent.

a.

Onsite Review and Investigative Function

'The' Station Superintendent shall:

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

. Review and Investigative Function for concurrence in a timely manner,.

those items described in Specification 6.5.la which have been approved by the Onsite Review and Investigative Function.

b.

Responsibility The responsibilities of the personnel performing this function,are:

1)

Review of:

(1) procedures required by Specification 6.8.1 and changes thereto, (2) all programs required by Specification 6.8.4 and changes thereto, and (3) any other proposed procedures or changes thereto as determined by the Mmt Superintendent to affect nuclear safety; Statich 2)

Review of all proposed tests and experiments that affect

' nuclear safety; BYRON - UNIT 1 6-12

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

d'323 W Ff' ACMINISTRATIVE CONTROLS ONSITE (Continued) 3)

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

Review of all proposed changes or modifications to plant systems or equipment tnat affect nuclea? safety; 5)

Investigation of all violaticns of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Division Vice President and General Manager - Nuclear Stations and to the Supervisor of the Offsite Miem and Investigative Function; Review l

6)

Review of all REPORTABLE EVENTS; 7)

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

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

Review of the Emergency Plan and station implementing procedures and shall submit recommended changes to the Division Vice President - Nuclear Statiors; i

A ced General mana3er-

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

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

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

c.

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

The Station Superintendent shall follow such recommendations or select a course l

BYRON - UNIT 1 6-13

ATTACHmWT 15 Byron Station proposes to modify surveillance toquirement 4.5.1.lc as shown on the attached copy.

Justification Vocifying that the breaket is open acconsplishes the same objective without requiring the operator to remove fuses.

l l

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(0548M)

AU6 2e q EMERGENCY CORE CCOLING SYSTEMS a

SURVEILLANCE REQUIREMENTS (Continued) b.

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

by verifying that power to the isolation valve operator is disconnected from the circuit by ra w ic W -caZ - > fr.

jete r r yju s-rHe es mae of".

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

4 W

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l l

BYRON - UNIT 1 3/4 5-2

-. ' ~.

i-l l

l l

l ATTACMENT 16 Byron Station proposes to change itea 6f of Technical Specification Table 3.3-3, " Engineered Safety Features Actuation System Instrumentation", as indicated on the attached copy.

Justification The current version of the Specification, which identifies ACTION statement 18, requires a plant shutdown of the numbec of OPERABLE channels is one less j

than the MINIIEM CHAbtlE13 OPERABLE, and the inopete.ble channel cannot be restored to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . With the proposed change, which identified ACTION statement 19, an inoperable channel may be placed in the tripped condition and STARTUP and/or POWER OPERATION may continue. Since there are only two channels available, wher one is placed in the trip condition, the plant protection is conservative since a trip signal from the remaining channel will meet the two channels to trip requirement. The proposed changed provides flexability for plant operation without reducing plant protection.

i I

(0548M) i

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION l

c MINIMUM E

TOTAL NO.

CHANNELS CHANNELS APPLICABLE j

j FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACil0N 6.

Auxiliary Feedwater a.

Manual Initiation 2

1 2

1,2,3 22 i

b.

Automatic Actuation Logic 2

1 2

1,2,3 21 and Actuation Relays c.

Sta. Gen. Water Level-2 Low-Low j

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

2/sta. gen.

3/sta. gen.

1, 2, 3 19^

g in any opera-in each y

Ling sta gen. ' operating sta. gen.

e-.

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

2/sta. gen.

3/r,ta. gen.

1, 2, 3 19^

in any in each operating operating 3

stm. gen.

up artm em d.

Undervoltage - RCP 4-1/ bus 2

3 1, 2 19^

'[f.12 j

Bus-Start Motor-W d

Driven Pump and Diesel-Driven Pump gJ e.

Safety Injection -

"N) w Start Motor-Driven Pump See Item 1. above for all Safety injection initiating functions y 7.J j

and Diesel-Driven Pump requirements.

g uyg f

f.

Division 11 ESF Sus i

Undervoltage-

'3*

j Start Motor-Driven i

Pump (Start as part 2

2

(_2 1, 2, 3 j

of DG sequencing) _

j

/9 f

ATTACl9ENT 17 Byron Station proposes to modify Technical Specification Sections 4.7.1.2.3(b)

(page 3/4 7-5), 4.7.5.3(b) (page 3/4 7-14), 4.7.10.1.2(b) (page 3/4 7-30) and 4.8.1.1.2(d) and (e) (pages 3/4 8-3 and 3/4 8-4) as shown on the attached copies.

Justification The proposed changes to the diesel oil sampling requirements as shown on the attached sheets are in accordance with previous discussions and transmittals to the NBC.

J e

1 1

e (0548M) l

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"'S4 3 LAW' sis ~E"5

,__a SURVEILLANCE RECCIREMENTS (Continued) vdue_(rmM,fewe~ :

gg 2)

. Verifying by flow or position cneck that eachp---- ~-"" ^-

a haVin the flow path that is not locked, sealed, or otnerwise secured in position is in its correct position; and r3) erifyi that e h autAic e in,tne flow-path is in th ~

full open p tion Wreneve e Audliary Feedwa Sy is abovejld%RA p

in toma cont or w THE L

b.

At least once per 18 months during shutdown by:

1)

Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Auxiliary Feedwater Actuation test signal, and 2)

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

a)

ESF; or b)

Steam Generator Water Level Low-Low from one steam generator, or c)

Undervoltage on Reactor Coolant Pump 6.9 kV Buses (2/4), or d)

ESF Bus 141 Undervoltage (motor-driven pump only).

4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERA 8LE following each COLD SHUT 00WN of greater than 30 days prior to entering MODE 2 by verifying normal flow to each steam generator.

4.7.1.2.3 The auxiliary feedwater pump diesel shall be demonstrated OPERABLE:

a.

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

r 04ot i-1981 b.

At least once per 92 days by verifying that a sample of diesel fuel g

from its day tank, obtained in accordance with ASTM-0270 107? is within the acceptable limits specified in Table 1 of ASTM-0975-1977 when checked for'viscositj, water, and sediment; and c.

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

BYRON - UNIT 1 3/4 7-5

F AUG.'. -

PLANT SYSTEMS

,i SURVEILLANCE REQUIREMENTS (Continued) 4.7.5.3 The essential service water make up pump shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that:

1)

The fuel storage tank level is at least 16%,

2)

The diesel starts from ambient conditions on a simulated low basin level test signal and operates for at least 30 minutes, and 3)

Each valve (manual, power-operated, or automatic) in the flow path is in its correct position.

OMwn a

b.

At least once per 92 days by verifying that a sample of diesel fuel D4oS1-19e s

from the fuel storage tank, obtained in accordance with ASTM-0270-10757-is within the acceptable limits specific in Table 1 of ASTM-0975-1977 when checked for viscosity, water, and sediment; and At least once per 18 months by subjecting the diesel to an inspection c.

in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service and by cycling each testable valve in the flow path through at least one complete cycle of full travel.

e BYRON - UNIT 1 3/4 7-14

~

Ai'C Z ; ($34 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

N b.

At least once per 92 days by verifying that a sample of diesel fuel s

from the fuel oil day tank, obtained in accoroance with ASTM-027117s, is within the acceptable limits specified in Table 1 of ASTM 0975-1977 when checked for viscosity, water, and sediment; and titoS7-196(

At least once per 18 months, during shutdown, by subjecting the c.

diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

4.7.10.1.3 The fire pump diesel starting 24 volt battery bank shall be demonstrated OPERABLE:

At least once per 7 days by verifying that:

a.

1)

The electrolyte level of each battery is above the plates, and 2)

The overall battery voltage is greater than or equal to 24 volts.

b.

At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery, and At least once per 18 months, by verifying that:

c.

1)

The batteries, cell plates, and battery racks show no visual indication of pnysical damage or abnormal deterioration, and 2)

The battery-to-battery and terminal connections are clean, tight, free of corrosion, and coated with anticorrosion material.

BYRON - UNIT 1 3/4 7.30

m AUG 2 9 y ELECTRICAL POWER SYSTE.v5 5URVEILLANCE RE0VIREMENTS (Continued)

In accordance with the frequency specified in Table 4.8-1 on a a.

STAGGERED TEST BASIS by:

1)

Verifying the fuel level in the day tank, 2)

Verifying the fuel level in the fuel storage tank, 3)

Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank, 4)

Verifying the diesel starts from ambient condition and accelerates to at least 600 rpm in less than or equal to 10 seconds.* The generator voltage and frequency shall be 4160 + 420 volts and 60 + 1.2 Hz within 10 seconds

after the start signal.

The diesel generator shall be started for this test by using one of the following signals:

a)

Manual, or b)

Simulated loss of ESF bus voltage by itself, or c)

Simuiated loss of ESF bus voltage in conjunction with an ESF actuation test signal, or d)

An ESF actuation test signal by itself.

5)

Verifying the generator is synchronized, loaded to great r tnan y

or equal to 5500 kW in iess tnan or equal to 60 seconcs. operates with a load greater than or equal to 5500 kW for at least 60 minutes, and 6)

Verifying the diesel generator is aligned to provide standby power to the associated ESF busses.

b.

At least once per 31 days and af ter each ooeration of the diesel where the period of operation was greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by checking for and removing accumulated water f rom the day tanks; l

c.

At least once per 92 days by checking for and removing accumulated water from the fuci oil storace tanks:

d.

e pe 92 day nd from w fuel oil rior addi-]

tion t the s rage ta s by veri ing tnat a ample obtained in acc dance

'th ASTM-270-1975 eets the fo owing min 'rirum requi re-m ts in cordanc with the ests specif'.d in ASTM 975-1977:

1)

A ater and sediment ntent of I s tnan or.q,ual to 0.

g lume pe ent; 47 2)

A kinert tic visc ity of 40 C of greater han or e al to

.3

(

centis ckes, but less than r equal to 4.1 centis okes;

The diesel generator start (10 sec) from ambient conditions shall be performed l

at least once per 184 days in these surveillance tests.

All other engine starts l

for the purpose of t:11s surveillance testing may be preceded by an engine pre-lube period and/or other warmup procedures recommenced by the manufacturer so that mechanical stress and wear on the diesel engine is diinimized.

l BYRON - UNIT 1 3/4 8-3

e i

To: Cht. MOOh hM

O s stets T*e pg 319 B-5 l

4.8.1.1.2 d.

By sempling new fuel oil in accoccance with ASTM-D4057-prior to addit ion to storage' tanks and:

1)

By verifying in accordance with the tests specified in ASTM-D975-81 pt-ior to addition to the storage tanks that the sample has:

l a)

An API Gravity cf within 0.3 cegrees at 60/60 F,

(-

or e specific gravity of withi'n 0.0016 at E0/60 F, when compar'od to the supoliers ce'tificate, or an ecsolute specific gravity at 62/E0 F of greater-than ov-ecual to 0.83 f

but less than or equal to 0.89, or an ADI g**evity of gre c t er-than ot-crua' to 27 cegreet but 1ers tha* or e.ual to 29 cegtcest b) p k:ncoatic viscestty at 42 F c' greatetr than I

or ecual to 1.9 centistoses, but less snar. cc ecual to 4.1 centistokes, if the gt avity wes; not d et erri i ned by comcarision with the suoplier's certification; c)

A flasn point ecual to or gt eat er than 125 F; and l

d)

A clear and bright accearance with proper color wher. tested in accordance with ASTM-D4176-82.

[

-m__---

2)

By verif g witbin 30 days of obtaining the sample I

that the.,ctt.*ptoperties specified in Table 1 of 4

ASTM-D975-81 are met when tested in accordanca with j

ASTM-D975-81 except that the analysis for sulfur may be performed in accordance with ASTM-D1552-79 or ASTM-D262C-82.

Af leas) once evepf 92)M-ays by,, obtaining a d aim sa pie e.

j Waccorganc[wf4ihA,ST 0 (-85'an [vgrif nyhha tM V Of le -( of/ASPP975*81%re e,ne yA f~ r eker lbsynperffieh i Ta T

d

, n' es ed

'n a cor ne wi

. TM-D9Jd-8,.e t

t w

l

\\

dh an ysi

.f Mu fu.ma tte perfor d i a_ o dhnce ST(1-D2 2-82.

l 2

wi h A TM-1

-7 r

I e.

At least once every 31 days by obtaining a sample of fuel oil in accordance with ASTM-02276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance l

t with ASTM-02276-78. ":t= A,-

ret

AUG 2 8 gg FIc,A,,lD..M,jgFT ELECTRICAL POWER SMTEMS SURVEILLANCE REOUIREMENTS (Continued) f cific g avity,as spec &fied ( the manufacttrer at 60/6 F 3) s of greatep than equ to 0.8J but 1 is tha oregalt p'89 orjtn API _ravit t 60* of gr ter t no to

( /27 degrees bu less an or qual 39 de rees;y equa f.

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 capability to reject a load of greater than or eoual to 1034 kW wnile maintaining voltage at 4160 + 42C volts and frequency at 60 : 4.5.Hz, (transient state), 60 + 1.2 Hz (steady state).

3)

Verifying the diesel generator capability to reject a load of 5500 kW without tripping.

The generator voltage shall not exceed 4784 volts during and following the load rejection, 4)

Simulating a loss of ESF bus voltage by itself, and:

a)

Verifying de-energization of tne ESF busses and load shedding from the ESF busses, and b)

Verifying the diesel starts on tne auto-start signal, energizes the ESF busses with permanently connected leads within 10 seconds, energizes the auto-connected safe shutdown loads through the load secuencing timer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdcwn loads.

After energization, the steady-state voltage and frequency of the ESF busses shall be maintained at 4160 + 420 volts and 60 + 1.2 Hz during this test.

BYRON - UNIT 1 3/4 8-4

1 ATTAQ5cDrr 18 Byron Station proposes to modify Technical Specitication 3.4.1.2 and 3.10.4 as shown on the attached copies:

Justification, The chang 2 to specification 3.4.1.2 is merely a note at the bottom of the page referencing specification 3. M.4.

The change to specification 3.10 4 saintains the requirement of specification 3.4.1.2, that three reactor coolant loops must be OPERABLE in mode 3, but deletes the requirement that two reactor coolant loops must be operating.

This change is ar. extension to mode 3 of the special test exception of specification 3.4.1.1 which allows the reactor coolant pumps to be turned off la modes i and 2 during startup and PHYSICS TESTS. Allowing the reactor coolant pumps to be off in mode 3 during rod drop tests presents no gceater concern than allowing all reactor coolant pumps to be off in modes 1 and 2 during startup and PHYSICS TESTS.

These changes are requested to allow the hot no-flow rod drop time measurements to be performed in a timely mannet without the requirement to start the coactor coolant pumps every hour whkh would result in unnecessary cycling of the ik:P's.

l (0548M) 1

- =-

4 l

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{4%~'f,

}$ ':1 U REACTOR COOLANT SYSTEM

~

3 HOT STANDBY AUS 2 8 $84 LIMITING CONDITION FOR OPERATION l

3.4.1.2 At least three of the reactor coolant loops listed below shall be OPERA 8LE and at least two of these reactor coolant loops shall be in l

operation:"

a.

Reactor Coolant Loop A and its associated steam generator and l

reactor coolant pump, I

b.

Reactor Coolant Loop B and its associated steam generator and j,

reactor coolant pump, l

c.

Reactor Coolant Loop C and its associated steam generator and l

reactor coolant pump, and i'

d.

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

I APPLICABILITY: MODE 3.**

ACTI0ti:

a.

With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be A

in HOT SHUTDOW within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

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

c.

With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concsntration of the Reactor Coolant 4

System and isaediately initiate corrective action to return the required reactor coolant loop to operation.

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

4.4.1.2.2 The required steam generators shall be determined CPERA8LE t'y verifying secondary side narrow range water level to be greater than or equal g

to 41% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

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

A,

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

A+ %. butdT 4 Gqb uo,q BYRON - UNIT 1 3/4 4-2 4

e.

--,--,-,,-.(h,..--,,---..w,,s,

.,..... - -,-,.., - - ee.. -. -,... -,-,--,, -

-,,,-,,wm,ew-w,

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

e.,..v-,,

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f AUS 2 8 tgg 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION tL.CA.c. htions3 d 3.10.4 The limitations o cifi 1.1 may be suspended during the performance of startup an ICS TESTS >IsL(ded.

rev' o,. S p u. h a t. - 3.4.t.t %,:. g w p h og dwm wm. 6 ~ai. \\.,7. progle2-iQ The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and z6)

The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels sre set less than or equal to 25% of RATED THERMAL

.(ER N beat (

mg Q3 j

.,g,4t,c,LJ,\\..os 9,,y u_,q

,,Q, 6 s u.iCA A <el, m.. vs cPeraatc.

u d 6s p

<sb

.c.c..

e APPLICABILITY:

uring operat7on be ow the P-7 Interlock etpoint,,oc puC.,m y

~

oC re n deep M ~Lu,uw\\s k -ka. '1.

ACTION:

With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately ca..

open the Reactor trip breakers.

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SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall b2 determined to be less than P-7 Interlock Setpoint at least once per hour during startup and PHYSICS TESTS.

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

,g initiating startup and PHYSICS TESTS.

M lusikb ukew go:d % J,g ccebd leces SUI htEt-M.co.g. 5 i

cPseA at.s 4 4 4 k.o,s eceo, b wLuem eC (L.ed Jme k esu d

% uS g cu-J 6.dce de pd kh.5b ewco pu 4 t

k codEve M% mon *bs g ocdc.kM

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l BYRON - UNIT 1 3/4 10-4 is u.

?

ATTAcmqENT 19 f

i I

Byron Station proposes to modify the Bases section of Lin'iting Safety System settings an shown on the attached copy.

)

i Justification The proposed changes correct inaccuracies in the P-6 description and provide a more complete description of the P-10 interlock.

I 4

(0548M)

L

LIMITING SAFETY SYSTEM SETTINGS BASES AU6 2 3 964 Reactor Trio System Interlocks The Reactor Trip System Interlocks perform the following functions:

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

provides7 backup block for Sour::e Range Neutron Flux doubling, and oc rnonval block that de energizes the high voltage to theAdetectors.

On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.

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@ doMi$

source RoaSc P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, more than one reactor coolant pump breaker open, reactor coolant pump uus undervoltage and underfrequency, Turbine trip, pressurizer low pressure and pressurizer high level. On decreasing power, the above listed trips are automatically blocked.

P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the single loop low flow trip.

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range Reactor trip and the Low Setpoint Power Range Reactor trip; and autcsatically blocks the Source Range Reactor trip and ie energizes the Source Range high voltage power. On decreasing pow r,dyheInter-sociate Range Reactor trip and the Low Setpoint Power Range Reactor trip are automt.tically reactivated.

?rovides input to P-7.

P-13 Provides input to P-7.

provides on l

Ou toNht bodAug Munthorn orid Sovece. Ron3e.

co hip vcHn3e. fnihe deteetors is cesfored

{

if powee decrease.s belou %< P-G i

serpoi*

i I

l BYRON - UNIT 1 8 2-9 i

i

-

ML20098F822

Text

Dear Mr. Denton:

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