Text

Provides Addl Comments & Suggestions Re Proof & Review Tech Specs Distributed in Co Thomas .Various Sections of Tech Specs,W/Summary Explanation of Changes,Encl
	ML20090G624
Person / Time
Site:	Byron
Issue date:	07/16/1984
From:	Tramm T; COMMONWEALTH EDISON CO.
To:	Harold Denton; Office of Nuclear Reactor Regulation
References
	8934N, NUDOCS 8407250093
	Download: ML20090G624 (81)
	v • d • e

_

m yN Commonwealth Edison

) One First National Plaza, Chicago, Illinois

[; O j - Addr:ss Reply to: Post Othee Box 767

\ ,/ . Chicago. Ilknois 60690 July 16, 1984

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

Subject:

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

Thomas.

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

-to H. R. Denton.

(c): A- 'l 2, 1984 letter from T. R. Tramm t . R. Denton.

(d): April 9, 1984 letter from T. R. Tramm to H. R. Denton.

(e): May 2, 1984 letter from L. O. DelGeorge to H. R. Denton.

(f): June 20, 1984 letter from T. R. Tramm to H. R. Denton.

Dear Mr. Denton:

This is to provide additional comments and suggestions regarding

~ the proof and review version of the Byron 1 Technical Specifications that

- cas distributed in reference (a). NRC review of the specific ' changes proposed here is necessary before the Technical Specifications can be finalized.

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

A number of similar changes were submitted in references (b) through ( f) . We understand that the NRC will review each of these proposed changes and . inform Commonwealth Edison of their acceptability.

8407250093 840716 PDR ADOCK 05000 gg gl

(H. R.'Denton July 16, 1984

~

Please direct any questions you may have regarding this matter 4

.to.this office.

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

Very truly yours, 7/2[e T. R. Tramm

, Nuclear Licensing Administrator y"

.- 1m

- cc: ! Byron Resident Inspector a

8934N-W

ATTACHMENT A ,

(Bases Section 2.0)

1) Section 2.1.2 (pg B2-2) Reactor Coolant System Pressure Change "3107 psig" to "3110 psig" in the last paragraph.

This section stated that the entire RCS is hydrotested at 3107 psig which is 125% of the system's design pressure. 125% of design pressure is 1.25 times 2500 psia = 3125 psia -15 psig - 3110 psig.

2) Section 2.2.1 (pg B2-6) Reactor Trip System Instrumentation Setpoints The second to last line reads ". . 10% of full equivalent." This should be changed to read ". . 10% of full powe; equivalent."

This change is for clarity.

I P

a e yy BASES REACTOR CCRE (Continued) .

insalance is not within the tolerance, the axial power imoalance effect on the

.,.' . Overtamperature AT trips will reduce the Setpoints to provice protection

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consistant with core Safety Lleits.

2. L 2 REACTOR C00t. ANT SYSTEM P9E55URE The restrictien of this Safety Limit protects the integrity of the Reactor Coolant System (RC3) from overmosurization and theremy prevents the 1' -'

release of radionuclides contained in the reactor coolant from reaching tae containment atmosphere.

The reactor vessel, pressurizer, and the RC3 piping, valves, and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which

.. permits a maximum tran?ient pressure of 11CE (2735 psig) of design pressure.

O The safety 1.isit of 2735 psig is therefore consistant with the design criteria and associated Code requirements.

3110 The entire RCS is hydrotested at Maig,1255 of design pressure, to demonstrate integrity prior to initial operation..

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T PEOF & REE7 COPY LIMITING 5AFETY SYSTEM SETTING 3 BASES O' N "*" 47 .

. The Overpower AT Reactor trip provides assurance of fuel integMty (e.g.. ne fuel pellet molting and less than 15 cladding strein) under all possible overpower conditions, limits the requind range for overtasserature af tMp. and provides a backup to the Hign Neutron Flux tMp. The Setpoint is autoestically varied with: (1) coolant weature to cornet for tammerature induced changes in density and heat capacity of wetar, and (2) rate of change of tamperature for dynamic compensation for piping delays from the core to tae leep tassersture detectors, to ensure that the alloweale heat generation rata (W/ft) is not exceeded. The Overpower AT trip provides protection to sitigate the consequences of vaMous size steam breaks as reported in 'aCAP-9225, " Reactor Core Response to Lucassive Secondary Steam Releases.*

PressuM zer 8vessu n In each of the pressure channels, there are two indeoendant histacles, each with its own trip setting to provide for a High and Low Pressure tMp thus limiting the pressure range in which reactor operation is permittad. The Law setseint tH p protects against low pressure which could lead to ONS by

} ,tripeing the reactor in the event of a loss of reector coolant pressure.

On decnasing power the Law 5etpoint tMO is autceatically blocked by P-7 (a power level of approximately 105 of RATED THUMAL PChu with turnine insulse channer pressure at aooreximately 105 of full power equivalent); and on increasing power, automatically reinstated by P-7.

The High Satsoint trip functions in conjunction with the pressuMzer relief and safety valves to protect the Reactor Coolant Systas against systas overpressure.

P*essuM:er Water t.evel The Pnssurizar High Water Level tMp is provided to prevent water relief through the pressuMzer safety valves. On decnasing power the PnssuMzer High Water lavel tMO is autoestically blocxed by P-7 (a power level of approxiestely 105 of RATED THUMAL PChu with a turnine isoulse chamoer pressun at amorexiestely ICE of full equivalent)1 and on incnasing power, automatically reinstated by P-7. f PowCR SYRCN - UNIT 1 8 2-4

ATTACHMENT B (Section 3/4.0)

1) Section 4.0, Surveillance Requirements (pg 3/4 0-2)

Add the following as new item 4.0.6.

"4.0.6 Performing a surveillance that results in equipment being inoperable due to requirements of the surveillance (e.g. abnormal valve alignments) is not considered a violation of the Limiting Condition for operation and does not need to be repotted as a violation provided the equipment is made operable at the completion of the surveillance."

This is being added to clarify when Byron Station will be in violation of an LEO.

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PR00F & REVIEW COPY APPLICA8ILITY SURVEILLANCE REQUIREMefTS 4.0.1 Surveillanca Requirements shall be set during the OPERATIONAL MCOES or other conditions specified for individual Limiting Conditions for ll- Operation unisss otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillanca Requirement shall be performed within the specified time interval with:

a. A anximum alloweble extension not to exceed 25% of the surveillance

. faterval, but

b. The combined time intarval for any three consecutive surveillanca intervals shall not exceed 3.25 times the specified surveillance interval.

4.0.3 Failure to perform a Surveillance Requirement within the soecified 0 time interval shall constituta a failure to meet the OPERA 81LITY requirements

. for .a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. Surveillanca Requirements do not have to be performed on inoperaale equipment.

s 4.0.4 Entry into an CPERATIONAL MODE or other specified condition shall not be

. ? anda unless the Surveillance Requirement (s) associated with the Limiting

- Condition for Operation have been performed within the stated surveillanca intatval or as otherwise specified.

4.0.5 Surveillanca Requirements for inservica inspection and tasting of ASME Code Class 1, 2, and 3 components shall be applicante as follows:

O a. Inservice inspection of ASME Code Class 1, 2, and 3 casoonents and.

inservice tasting of ASME Code Class 1, 2, and 3 puses and valves shall be performed in accordanca with Section XI of the ASME Sciler and Pressure Vessel Code and applicaele Addenda as required by 10 CFR Part 50, Section 50.55a(g), except wnere specific written relief has been granted by the Commission pursuant to 10 CFR Part 50,

/ Section 50.55a(g)(6)(1);

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i ATTACHMENT C (Section 3/4.1)

circled items noted in this attachment have been previously submitted.

Section 3/4.1.1.5 (pg 3/4.1-6x) Boron Dilution

1) Section 3/4.1.2.1 Botation Systems (pg. 3/4 1-7)

Delete "350*F" and insert "or equal to 330*F" in the note that has been previously submitted at the bottom of the page.

The reason for this change _is to permit the plant to do heatups and cooldowns without violating Tech Specs. This change will make it permissible to restore Charging and Safety Injection pumps to operability i between 330*F and 350*F. This revision is possible because a review of the Byron cold-overpressure transient analysis reveals that above 330*F the Cold overptessure Mi':igation System should prevent Appendix G limits

' from being exceeded even for a worst case spurious SI actuation. The relieving capacity of a single pressurizer PORV at 1050 psig. the Cold 1 -overpressure Mitigation System valve-opening pressute setpoint corresponding to 330*F, is adequate for dealing with the charging and

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Safety Injection pump flows resulting from the SI actuation.

I Also, it should be noted that, with primacy temperatures of 330*F or ,

greater, the Appendix G overpressure limits are calculated to be above the pressurizer safety valve setting of 2485 psig. Hence, in addition to the 90EV's, the Safety Valves should also prevent a spucious SI actuation i transient at temperatures above 330*F from exceeding the limits.

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2) Section 3/4.1.2.3 Charging Pump-Snutdown (pg. 3/4 1-9)

- Insert, "whenever the temperatute of one or more of the BCS cold legs is less than oc equal to 330*F' before the first sentence in Section 4.1.2.3.2.

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- Delete the asterisk following the word "one" in Section 3.1.2.3. Add !

the asterisk behind Mode 4 under

applicability".

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- Delete "350*F" and insect "or equal to 330*F" in the note that was previously submitted at the bottom of the page.

This is for the same reason as #2 above.

3) Section 3.1.3.3 (pg 3/4 1-18) i In the 140 delete the words " shutdown oc". This change is necessary l

because the digital rod position indication system does not indicate the actual position of the shutdown mode between 18 steps and 210 steps

. withdrawn. Therefore, it is not possible to verify that the DRPI agrees l:

with the demand position indicator within 12 steps.

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'J - 3/4.L 2 80AATTCN SYSTBS Ft.0W PATH - SHUTCCWN Limi tmi CCNOITION POR OPERATTQM 1

"C 3.LL1 As a sinimas, one of the following beren infection flow paths sh'all be OPERASLE and capaele of bein6 peered fne an OMRA8LE amargency power ,

souns:

b. a. A ffsw patte fren'the Beric Acid Storage System via a boric acid tnesfer pump and a charpng peaup to the Aeactar Coolant System if the Boric Acid Storage 3/stam in Specification 3.LL54. is CPERA8LE, w .

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b. The flow path from the rethling water storage tank via a cursing

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OPERASLE emergency power source, suspend all operations involv OnRAna ., .a.a.,. .f b.1n,ing C AE ALTERATIONS or positive reactivity changes.

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4 4.LLI At least one of the ateve required flow paths shall be demonstrated OPERASLE:

a. Atileast once per 7 days by verifying that the tamperature of the

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i host, traced portion of the flow path is greater than or equal to GB*F when a flow path from the Boric Acid Storage System is used, l

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' b. At least once per 31 days by verifying that each valve (annual, p r 5- J.ed, or automatic) in tne flow path taat is not locked,

' sealed, or staenrise secured in position, is in its escroct position.

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LIMITING CCMOITION FOR OPBATTON ei

3. L 2.5 trging pump in the boron injection flow path required by Specificat .L2.1 shall be OPERA 8LE and capatie of being powered from an CPERA8LE emergency power source.

4 APM.ICA8It.ITY: M00E5 4, 5. and 8.

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, With no charging pump OPGA8LE or capable of being powered from an CPGA8LE emergency power sourca, suspend all operations involving CORE ALTU ATIONS or .

nesitive reactivity changes.

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$URVIILLANCE REQUIREMENTS

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4.L2.3.1 The above required charging pump shall be demonstrated CPGASLE by .. ;.- - -

verifying, on recirculation flow, that a differential pressure across the

, pump of greater than or equal to 2418 paid is developed when tastad pursuant to l Specification 4.0.5.

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4. L 2. 3. All charging pumps, excluding the above required OPGA8LE pues, ,

! shall be demonstrated inoperaale at least once per 31 days, except when the t

I reactor vessel head is removed, by_ verifying that the notar circuit breakers are secured the open position.]WA1._..m g 7,,._ .. .t % %g ,

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POSITION INDICATICN SYST D - SHUTOCWN LIMITING CONDITION FOR OPERATION 3.1.3.3 One digital rod position indicator (excluding bank demand position indication) shall be OPERA 8LE and capable of datarsining the control rod s position within a u steps for each fr^r r *"c'ontrol rod not fully inserted.

AP9tICA8ILITY: MODES 34 , 44 and SM. _._ ,

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With less than the above required position indicator (s) OPERA 8LE, fenediately open the Reactor Trip System breakers.

j $URVIILLANCE REQUIREPENTS

'4.1. J rara af the .above required digital rod position indicator (s) shall dotarsined OPERA 8LE by verifying r.n.s G. 41gital rd ;;;;iti ' tine-agrees with the demand position indicator within u steps when exercised over the full-range of rod travel at least once per la sentas. %t Q.*g,'(a(

hh "Indicabn %sba does rd indic.ak the. Scha( pos&rt eg. 84. $ddo Md4 Ocef( IS Abp% Ahd 260 Ab'.ps WMdfawr),

Nita tae Aeac."oi" Trip Systas breakers in the closed position.

$ee Special Test Exception 3.10.5.

8YRON - UNIT 1 3/41 'J

ATTACHMENT D (Section 3/4.3)

Circled items noted in this attachment have been previously submittc<i.

1) Table 3.3-2, item 7, (pg 3/4 3-7)

Item 7. Overtemperatu're AT response time should be changed from "5 6 seconds" to "5 4 seconds". The FSAR analysis assumes a total ovettemperature AT response time of 6 seconds. This 6 seconds represents a pure 2 second delay plus a 4 second log allowance. This change should be made in order to maintain a consistency with the FSAR.

2) Table 3.3-3, item Id, (Pg 3/3 3-14)

Note number "18" should be changed to "19". This maintains a consistency with Tech Spec Section 4.3.1.2, Table 3.3-1 item 9 action statement number on pg 3/4 3-3.

3) Table 3.3-11. (pg 3/4 3-57)/ Inserts A & B/ (pg 3/4 3-58) on Table 3.3-11, pg 3/4 3-57 change the following:

, Instrument 1.ocation Chance From Chance To

1. Zone 76, Elev. 426 9 13 Zone 7, Elev. 414 6 7 Zone 24, Elev. 414 10 16

2. Zone 69, Elev. 451 8 10 Zone 75, Elev. 451 10 16

3. Zone 77, Elev. 426 9 20 Zone 78, Elev. 426 9 19

5. Zone 13, Elev. 383 5 6 i

l - Insert A to pg 3/4 3-57 Zone 45, Elev. 463 8 9 Zone 46. Elev. 463 8 9 Zone 47, Elev. 463 4 5 l 5 Zone 48, Elev. 463 4 l

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ATTAC19 TENT D (CONTINUED)

(Section 3/4.3)

- Insert B to pg 3/4 3-57 Zone 49 Elev. 439 8 23 Zone 50, Elev. 439 8 23

-Zone 51. Elev. 439 10 13 Zone 52, Elev. 439 10 13 Zone 53, Elev. 439 8 9 Zone 54 Elev. 439 8 9 Zone 55. Elev. 439 4 6 Zone 56, Elev. 439 4 6

- Table 3.3-11, pg 3/4 3-58

6. Zone 67. Elev. 451 3 11 9.- Zone 52, RSH 6 8

10. Zone 39, Elev. 401 3 7 Also, the ** Table Notation, add "These switches are not 72D supervised."

These changes are being made because new detectors are being installed in safe.ty related areas to upgrade coverage.

4) Tablo 4.3-9, item 4, (pg 3/4 3-73)

Delete the words "a nominal" and replace with " hydrogen and nitrogen".

Also, delete items 4a and 4b.

As part of the channel calibration for this equipment, Byron Station makes use of the two point method of calibration utilizing a minimum of two standard gas samples each varying in the total percentage of hydrogen and nitrogen. It is not necessary to calibrate this equipment using only the one and four volume percent hydrogen, balance nitrogen gas samples.

5) Table 4.3-9, item 5. (pg 3/4 3-73)

Delete the words "a nominal" and replace with " hydrogen and nitrogen".

Also, delete items 5a and 5b.

As part of the channel calibration for this equipment, Byron Station makes use of the two point method of calibration utilizing a minimum of two star.dard gas samples each varying in the total percentage of oxygen and nitrogen. It is not necessary to calibrate this equipment using oniv the one and four volume percent oxygen, balance nitrogen gas samples.

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%ene 24 Elev 414 Detection 40-Ka

2. Centrol Aeos Zone 64 Elev 451 Setection 3 4- /0 Zane 65 Elev 461 Detection Zane 75 Elev all Ostection -W /G :

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FIMt CE*1CTicN INSTit9tNTS IMS*ttvtNT LCCATION INSTttytNT TY9t* TOTAL NUM*(1t CF INf*ttMtVTS M Flam !moke

6. Station lattery Rose Zone 87 Elev 451 ,

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10. Fuel Starege Detection 4*7 Zone 39 flev 441 3 Zone 3a Elev 428 Detection .. .... .

TAatt NOTATf0N1

A single detectar in a zone mereed "Setection" will alars in the Main Cantret tese.

A single estacur in a zone earted "Segrusten" will initista susoressten and alars in the Main Central Rees.

"These are Cantainment Ventilation tassersture switanes. Upon reset t ofMa Hi-Hi esien avet be annually initiated.M, M

tassegture.s Mrs setectisit instrumenta located within the containment are not recuind to to CPtRA4L1 suring the perfenance of Type A containment teakage reta tests.

i

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R00 h hf l{h (Off

-; TA8LE 4.3-9 (Continued)

TASLE NOTATICMS

At all times.

, _ ,.,. ** During WASTE GAS HOLOUP SYSTEM operation. .

(1) The DIGITAL CHANNEL OPERATICMAL TEST shall also demonstrate that automatic isolation of this pathway and control room alars annunciation occur if any of the following conditions exists:

! 4. Instrument indicatas asasured levels above the Alars/ Trip Setpoint, .

U,

or l b. Circuit failure, or

c. Instrument indicates a downscale failure, or

- d. Instrument controls not set in operata mode.

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

. a. Instrument indicates esasured levels above the Alars Setcoint, or b.

Circuit failure, or

c. Instrument indicates a downscale failure, or

d. Instrument controls not set in operata mode.

'C (3) The initial CHANNEL CALIBRATION shall be perfomed using one or sort of the reference standards cert 1fied by the National Sureau of Standards .

(N85) or using standards that have teen ootained free suppliers that l participata in seasurement assurance activities with NES. These standards shall pemit calibrating the systas over its intended range of energy and

, ~

asasurement range. For sue.tequent CHANNEL CALIBRATICN, sources that have l .;' been related to the initial calibration shall be used.

l (4) The CHAlelEL CALIBRAT'CN shall include the use of standard gas sasoles containing 4 aemina.1.:AQsp4 &p.

l e. e r e ; = xt r,r .;x , :! re: -i tn ;= , - =e "--

t.

.. vo: - ,... .t; ...;..., Lei a. ..'..w...

(5) The CHANNEL CALIBRATION shall include the use of standard gas sasoles containing : w .'oe!.' g a d M .

. c= e 7 c;.r.: a ,,..., .:_a ai =; a, z "--

5. "=r ;;i - p c ; .t .,c,;ea, L.! e.ae .i; .. w....

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BYROM - UNIT 1' 3/4 3-73'-

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ATTACHMENT E (Section 3/4.4)

Circled. items noted in this attachment have been previously submitted.

1) Section 4.4.6.2.1(b), (pg 3/4 4-20)

Delete the words " inventory and" from item b. At Byron Station there is no capability of monitoring the reactor cavity sump inventory. The method of calculating the containment floot drain and reactor cavity sump discharge is attached (See BOS 0.1-1 Revision 1, Page D-ll).

2) Section 4.4.9.1.2, (Pg 3/4 4-30)

Delete the word "and" between Figures 3.4.2.. 3.4.3. Also, following the -

3.4.3 add "and 3.4.4". By adding reference to Figure 3.4-4, this ensures that the figure will be updated periodically to account for the irradiation of the pressure vessel.

3) Section 3/4.4.11 (pg 3/4 4-39)

Add the attached new submittal Limiting Conditions for operation for the Reactor Coolant System Vents.

This is being added as a result of NUREG-0737.

r- , --- ,,---g - , , - - - -, ,r,, .-g.g--- - - - -

,,w-,. ,wy.w.-w, .- , ,., p me, ,r,g.

REACTOR CCOLANT SYSTEM !.I t

SURVEILLANCE RECUIREMENTS 4.4.6.2.1 Reactor Coolant Systern leakages shall be demonstrated to be within each of the above limits by: -

I a. MonitoHng the containment atmosphere gaseous and particulate ~

radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

b. Monitoring the containment floor drain and reactor cavity sume inu. to j .ad- discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; I c. Measurement of the CONTROLLED LEAXAGE to the reactor el TAM seals when the Reactor Coolant System pressure is 2235 m p g at least once per 31 days with the modulating' valve fully The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4; ,

, d. Performance of a Reactor Coolant System watar inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

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

I % 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation valve scecified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to te within its limit:

a. At least once per 18 months, .

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

c. PM or to returning the valve to service following maintenance, repair or replacement work on the valve, and p d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

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

D (3

vJ SYRCN - UNIT 1 3/4 4-20 F

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C, BOS 0.1-1 Revision 0 Date / /

[d CONTAINMENT FLOOR DRAIN AND REACTOR CAVITY SUMP INVENTORY AND DISCHARGE NOTE: Run time meters are located inside OPLO2J j 1). Reactor Cavity Sump Pump 1RF0lP Shift Previcus Readino Present Readinc Run time Flow Discharce_

1 Min Min Min -

71.lgpm gals 2 Min Min Min *ll.lgpm gals 3 Min Min Min *11.lcre cals 2). Containment Floor Drsin Sump Pump 1RF02PA Shift Previous Readinc Present Readinc Run time Ficw Discharce 1 Min Min Min 142.5gpm gals 2 Min Min Min 142.5gpm gals 3 Min Min Min 142.Scom cals 3.) Containment ~ Floor Drain Sump Pump 1RF02PB.

U Shift Previous Readinc Present Readino Run time Flew Discharce 1 Min Min Min 98.5gpm gals 2 Min Min Min 98.5gpm gals 3 Min- Min Min 98.5com cals Previous Reading (Min) - Present Reading (Min) = Run time (Min)

Run time (Min) X flow (gpm) = Discharge (gals) ,

.h W

f st **6 f

th f APPROVED D-ll JUL 181983 B. O. S. R.

l

- - _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ 1

- 1 1

1

"'a" c= 5=* PR00F & REVlEW COPY l

3 3/4.4.9 PRESSURE /TDPERATURE t.IMITS REACTOR C001. ANT SYST S LIMITTNG CONDITION FOR OPERATION

.7 3.4.9.1 The Reactor Coolant System (except the pressuri:er) tamperature and pressure shall be 11mitad in accordance with the limit lines shown on Figures 3.4=1 and 3.4-3 during heatup, cooldown, criticality, and inservica leak and hydrostatic testing with:

. . '- a. A anximm heatup of 100*F in any 1-hour period,

. b. A maximum cooldown of 100*F, in any 1-hour period, and

c. A maximum temperature change of less than or equal to 10*F in any 1-hour period duMng insanica hydrostatic and leak tasting operations

,3 above the heetup and cooldown limit curves.

APPLICA8ILITY: At all times. .

ACTION:

] With any of the amove limits exceeced, restore the temperature and/or pressure to within the limit.within 30 minutas; perfors an engineeMnq evaluation to datarsine the effects of the out-of-limit condition on the structural integMty of the Reactor Coolant System; detacuine that.the Reactor Coolant System remains acceptable for coctinued operation or be in at least NOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T and pressure to less than 200*F and

.. 50d psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l SURVEIL!.ANCE REQUIRD OT3 9 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutas duMnq systam

! hestup, cooldown, and inservice leak and hydrostatic tasting operations.

1 1 4.4.9.1.2 The reactar vessel antaMal irradiation surveillance specimens shall be removed and examined, to datarsine changes in satarial procerties, as required by 10 CFR Part 50, Appendix H, in accordance with the schedule in l' Taale 4.4-5. The results of these examinations shall to used to updata l A Figures 3.4-2 j es#3.4-3%d 3,+.4, I

~

n.

3YRON - UNIT 1 ,

3/4 4-30

.i :, -. . - - -

n . -.m i

REACTOR COOLANT SYSTEM 3 4.N ll REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION E

3.4.11 At least Two REAcrog VE5SEL HEAD vent PATRs,g AcH consisting of two valves in series powered from emergency busesjshall be OPERABLE and closed.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With one of the above REAC.To&YESSEL HEAD vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and'in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3}9 4-39 L

l

b. With BoTH of the above REACTOR. VEME L HEAT >

vent paths inoperable; maintain the inoperable vent paths closed with power removed from the valve actuators I

of all the valves in the inoperable vent paths, and restore at least oNE of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months -

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

SURVEILLANCE REQUIREMENTS i

I 4.4.11 Each REACTOR VESSEL HEAD vent path shall be demonstrated O'PERABLE at least once per 18 months by: ,

1. Verifying all manual isolation valves in each vent path are locked in the open position.

2. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING.

l

! 3. Verifying . flow through the REAc. Top. VE%EL HEAD l

vent paths during venting COLD SHUTDOWN 7

or REFUELING.

ONN4 AT a

3/4 y-4o i _ . _ _ _ _ ._ . .

ATTACHMENT F I (Section 3/4.5)

Circled items noted in this attachment have been previously submitted.

1) Section 3/4.5.3 ECCS Subsystems (pg. 3/4 5-7)

Delete "350*F" and insert "or equal to 330*F" from the note at the bottom of the page.

Section 4.5.3.2 (pg. 3/4 5-8)

Delete "350*F" and insert "or equal to 330*F".

This is consistent with the change which allows the Charging and Safety Injection pumps to be made operable between 330*F and 350*F.

+

l r

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D ,

EMERGENCY CCRE COOLING SYSTEMS ~

I Ml00F & EBV COPY

. 3/4.5.2 ECCS 51J85YSTEMS - T ave < 350*F D- .

LIMITING CONDITION FOR OPERATION 0 3.5.3 As a sinimus, one ECCS subsystas comprised of the following snali be OPERA 8LE:

! a. One OPERA 8d centrifugal charging pump,"

I b. One OPERA 8 2 RHR heat exchanger g

c. One OPERA 8LE RHR pump, and

d. An OPERA 8LE ' flow path capadle of taking suction from the refueling I water storage tank upon being manually realigned and transferring l suction to the containment sump during the recirculation pnase of operation.

b APPLICA8ILITY: MODE 4. _

o ACTION:

M

,4 ,

a. With no ECCS subsystem CPERA8LE uecause of the inopersaility of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERA 8LE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDCWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With no ECCS subsystas OPERA 8LE because of the inoperanility of 0 either the RHR heat exchanger or RHR pump, restore at least one ECCS subsystas to CPERA8LE status or saintain the Reactor Coolant System T,,, less than 350*F by use of altarnata heat removal methods.

c. In the event the ECCS is actuated and injects water into the React.or Coolant Systas, a Special Report shall be prepared and submitted to the Commission within 90 days, pursuant to Specification 6.7.2, 9 ,

describing the circumstances of the actuation and the total accumulated actuation cycles to data. The current value of the usage factor for each affected Safety Injection noz::le shall be provided in this Special Report whenever its value exceeds 0.70.

p -

" A max 1mus of one centrifugal charging pump sna11 be OPERA 8LE never .

tamperature of one or more of the RCS cold legs is less than m m. ,

? ! ^* ^ en. o m m_0_ t, 3 3c o r, x

. O

% 3/4 5-7 BYRON - UNIT 1 0 .

4

- --y-w- -

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6

^

EMERGENCY CORE C001.ING SYSTEMS P00E.&.s 3 m SURVEII.t.ANCE REQUIREME.VTS _

4.5.3.1 The ECCS subsystem shall be demonstrated CPERABLE per the aoplicaele requirements of Specification 4.5.2.

4.5.3.2 All charging pumps and Safety Injection pumps, except the above required CPERA8LE pumps, shall be demonstrated inoperable *by verifying that the motor circuit breakers are secured in the open position at least once per 12 ho- enev ,the tamperature of one or more of the RCS cold legs is less th - nt;! *2'25^**!- an- +. ss o*F. *

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! M y-- - ? -W 4t.L. 9 C.S . To w% A -Pm , Wu SI S8OL A ST Sati A M JLi L 64L eM. Tb. 3- 9q % k. M h MM3 cJa-A A+ SIS 8 36, .sr sso LA %d s:ssot.8 4 h ET SSLI 0! * *.2#u.

AM ST se 2.t B p ww.tha,. 4 a M) 4d e- A ,-t.f4 a,75 A'0" lh ( SI 380t A , s2 stoEB , sr wrotc d SI%etD) M oho JLa. %de l c9.ed .

,f-BYRON - UNIT 1 3/4 5-8

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_,__~__________ _ _

ATTACHMENT G (Section 3/4.6)

1) Section 3.6.1.7.b (pg 3'/4 6-11) and Action Statement b.

Delete.the number "1000" and replace with "2000". Technical Specification 3.8.1.4 (pg 3/4 6-6) requires an extremely small primary containment internal pressure range. The 1000 hour0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months limit for the isolation valve will be exceeded given normal operation of the unit.

. 2) Section 4.6.4.1 (pg 3/4 6-23)

Delete the words "on a Staggered Test Basis". Following the words " sample gas containing" add the words " hydrogen and nitrogen". Also, delete items 4.6.4.la and 4.6.4.1.b.

As part of the*ehannel calibration for this equipment, Byron Station makes use of the two point method of calibration utilizing a minimum of two standard gas samples each varying in the total percentage of hydrogen and nitrogen. It is not necessary to calibrate this equipment using only the one and four volume percent hydrogen, balance nitrogen gas samples.

The staggered test basis definition does not apply here.

L L-

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, , .m---.-y---m -- w- -g- e <.c-m-.-e ------g -- ye-..mwg,-- -,e. m.= e w w w -ww e ww'r rw- w w w w w enn--er me v r- -w- -rv-e+-e-e- e---mme*mui-- - - '

,e*e.'

) CONTAIW eff SYS m s FM0F & HEW COPY Q- CONTAINNENT Vef7!1.AT!CN SYST194 Lim nmi CONDITION FOR CPGATTCH

!O 3. 6. L 7 Eacft containment purge supply and exhaust isotation valves shall be l

OPGASLE and: ,

, s. Eacit 40-inste contairment shuteam. supply and exhaust isotation valve shsIT be closed and fL a L ?*wr temewg.J

. b. The D= inch containment purge sueply and eznaust isalation valve (s)

D may be open for,s.p u

.<,to-- _,,,,_ m n hours during a calendar year.;-- c- " = "

ApptICA8IUTY NOCES 1, 2, 3, and 4.

A ** O Q a. With a 48-f ach containment p:rge supply and/or annaust isolation valve open or not sealed closed, close and/or seal close t3at valve or isolate tae penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otneerise be in at

'n' least NOT STANDEY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTD0tet l- wittrin the following 30 heurt.

r . -

2 0e 0

f. .

D. Witit the 8-inch containment purge upply and/or exhaust f sotation l valve (s) ocen for more than . during a calendar year, close the open 8-inch valve (s) or isolate the ;;.e.kd. ion (s) witnin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otheerise be in at least H07 STANOSY witatt the next 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , and in CDLD SHUTDOMt witain tae following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

j.1~CO c. Wftir a containment purge supply and/or exhaust isolation valve (s) having a sessured feskager rata in escasa of the limits of Specif1-cations 4.8.L7.3 and/or 4.8.L7.4, restors the inoperehle valve (s)

': to OpetAOLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , othenvise be in at least HGT STAlWEY witflin tne next & hours, and in COLD SHUTDOWN within the

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

J O, 2 p

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SY110M - UNIT 1 3/4 6-11

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. 3/4.6.4 CCM80STISLE GAS CONTROL HYOROGEN MONITOR $ ,

LIMITING CCNDITICN FOR OPERATION C

t. 8.4.1 Two independent containment hydrogen sonitors shall be OPERA 8LE.

AP9CCA8ILITY: MODES 1 and f.

With one hydrogen annitor inoperable, restore the inoperable sonitor to OPERA 8LE status within 30 days or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

C-

$URVEILLANCE RECUIREMENTS ___

t

'4. 5. 4.1 Eaca hydrogen sonitor shall be demonstrated OPERA 8LE by the perforsanca of a CHANNEL CHECK at least ones per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an ANALOG CHANNEL OPERATICNAL TEST at least onca per 31 days, and at least once per 92 days = : C'ft~JC i - -

M performing a CHAMEL CALIBRATION using sample gas containing- y o*A

5. ICT 'd ' - pw. .asm nyw w=u , lei. L ^it.7 -- b C

i SYRCN - UNIT 1 3/4 6-23 e emm.

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I ATTACHMENT H (Section 3/4.7)

1) Section 3/4.7.3 (page "AA" (2 of 3) after page 3/4 7-11)

Component Cooling Water System Delete paragraph 4.7.3.1.b from the attached Surve:illance Requirements of

the Tech epecs.

There are no valves in the component Cooling Water System that are automatically actuated on an SI signal.

2) Section 3/4.7.1 Fire Suppression Systems (page 3/4 7-28)

- Insert the word " Supply" in 5 places on this page between the words *

" Fire Suppression Water" and " System".

- Replace LCO 3.7.10.1.b (page 3/4 7-28) with the following:

"An OPERABLE flowpath capable of hking suction from the flume and ;

transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to: 1) the yard hydrant isolation valves (for hydrants near buildings containing safety related equipment),' 2) the last valve ahead of each hose standpipe as required by

. Specification 3.7.10.5, 3) the last valve ahead of the deluge valve (on

! the diesel generator oil storage room foam system and manual containment charcoal filter deluge systems), or 4) flow alarm valves (on sprinkler systems) as required by specification 3.7.10.2."

This change is made to limit the fire suppression system operability to safety related areas.

3) Section 3.7.10.2 (pg 3/4 7-31)

- Change section heading from " Foam Systems" to " Water Systems" i'

- LCO 3.7.10.2 (page 3/4 7-31) shall be modified to read as follows:

! "The Water Systems including the foam systems in the diesel generator fuel oil storage tank room the sprinkler systems in the Auxiliary Building and the containment charcoal filter manual deluge system shall be OPERABLE."

Delete the word " foam" in the applicability statement and replace with "above". Delete the word " foam" in the Action Statement.

This change is made to include the containment manual deluge system and new sprinkler systems.

4) Renumber Surveillance Requirements 4.7.10.2 to "4.7.10.2.1" this is necessary so two new sections can be added.

6

/

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ATTACHMENT H (CONTINUED)

(Section 3/4.7)

5) Delete the existing Surveillance Requirement 4.7.10.2.1.c(1) and insert the following:

1) By performing a system functional test which includes:

a) Verifying a " Fire Trouble" alarm is received when the isolation valve is closed, and b) Verifying a " Fire" alarm is received on actuation of the alarm switch.

These changes are necessary to provide the surveillance requirements for the two systems added: the containment manual deluge and the new sprinkler system.

This change is necessary to delete automatic actuation of the foam system per a NRC concern on spurious actuation of the foam system.

6) Add as new surveillance requirements 4.7.10.2.2 and 4.7.10.2.3 the following:

"4.7.10.2.2 The above required sprinkler system shall be demonstrated OPERABLE:

a) At least once per 12 months:

1) By flowing water out the inspectors test connection and verifying a " Fire" alarm

2) By cycling the isolation valve and verifying the " Fire Trouble" alara

3) By a visual inspection of the sprinkler header to verify integrity and that the head spray pattern is not obstructed.

4) By verifying water flow through the 2" test drain at the riser.

w - - . . - - . - - .

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

4 ATTACHMENT H (CONTINUED)

(Section 3/4.7)

4. 10.2.3 The above required manual deluge system shall be demonstrated OPERABLE.

a) At least once per 18 months

1) By cycling the isolation valve and verifying the " Fire Trouble" alarm

2) By cycling the deluge valve and verifying a " Fire" alarm

3) By a visual inspection of the deluge header to verify integrity and that the spray heads are not obstructed.

This change notes the distinction between the surveillances for the various water systems.

?

3.

7) Section 3.7.10.3 CO2 Systems (pg 3/4 7-32)

- Delete the words " low pressure" in Sections 3.7.10.3, 4.7.10.3.1 and 4.7.10.3.2. This change is for clarity. There is only one CO2 system.

- To action statement a. add the words "(lower cable spreading room)"

between the words " damaged;" and "for other areas".

- Section 4.7.10.3.2 b(1) and the word "both" after actuates, delete the words " manually and" after actuates and add the words " manually, and" at the end of the sentence.

- On insert A. replace 290 with 275.

This change is necessary in order to be consistent with vendor, the manual. Chemetron Operations & Service Manual.

8) Table 3.7-5 Fire Hose Stations (pg. 3/4 7-36)

Insert " LOCATION ELEVATION HOSERACK ANGLE l M-8 South Wall 451 279 0FP638 of battery room M-26 South Wall 451 280 0FP639" of battery room aft)r L-11 and before M-18.

These new hose reels have been added.

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ATTACHMENT H (CONTINUED)

(Section 3/4.7)

9) Section 4.7.11.1 Fire Rated Assemblies (pg. 3/4 7-38)

- Following the words a " visual inspection of" add "a)' Performing".

- Also, renumber 4.7.11.1 a, b, and c to 4.7.11.1 (1), (2) and (3)

- Also, add as a new item "4.7.ll.l.b Performing a functional test of at least 10% of the fire dampers. If any nonconforming dampers are found, an additional 10% will be inspected.

This process will continue until an acceptable sample is fcund."

This addition is being made due to NRO concerns.

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PROOF & REN CDM 3/4.7.10 FTRE SUP9RESSION SYSTD*5 1 SU ppg 4 FIRE SUP*RESSICN WATER,5YSTEM LIMITING CCNCIT'CN PCR CP9ATTCM ..

'C suPPy w 3.7.10.L The Hre Suppressioer Watar,5ystas shall be_0PGA81.Lvita. ~s

o. fue fire segresafest ^ " "t" tasir discastge aligmod ! --- t;h,dam_a nrw saeprtpfan

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APPt ICASILITY: At all times. IM ACTION:

't ' ) -

a. With one ptas and/or one water supply inoperable, restare ce fnopersale equipment ta CPGAat.E status wiuin 7 cays or stevide an alternata bact:m pump or supo!y. The previsiens of Soecifica-l tions 3.0.3 and 3.0.4 are not asolicaele.

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! b. With me Fire Suporession Watar.5ystas 4therwise inoperacie Q estaelish a barJc.o Fire suportssion Watar,5ystas.wiuin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Su9ty SURVEII.!.ANCE RfCUIRE8ENTS SuP99 '

Q g 4.7.10.1. 1 The Fire 5gpression Watar 5ystan x shall be demonstrstad CPGA8LE:

l s. At feast once per 7 days my verifying ue contained water susoly volume,

!*. At least onca per 31 days on a STAGGES TEST 3 ASIS :y starting us electric motor-triven ptmo and coersting it for at least 15 zinutas iQ on recirculation flow, l

1 l c. At least onca per 31 days by verifying cat esca valve (sanual,

! power-coarstad, or aut:aatic) in ce flew pata is in its esrv.

position, SYtCN - UNIT 1 3/4 7-28

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, , , _ PROOF & HB COPY

, uJA tER. M i v u 45 LIMITING CONDITICN FOR OPERATTON uwess. systems w,wene on .

3.7.10.2 TheJoam Systans in tae diesel generator fuel. storage tant rece[-t)u. -2 MO fA- .

^

y Yoo AP91.ICA8I.ITY: Whenever equipment protacted by the ystem is required to De OPERA 8 J.

ACTION:

a. With one or more of the above required Mystans inoperaale, .

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months estaalish a continuous fire watan wita backup fire C suppression equipment for those areas in which redundant systans or components could be damaged; for other areas, establish an hourly fire weten patrol.

! b. The provisions of Specifications 3.0.3 and 3.0.4 are not aggiicable.

C SURVEILLANCE REQUIREMENTS 4.7.10.2.1 MTae aeove required Foam Systa[shall be demonstrated

( OPERA 8LE:

a. At least once per 31 days, by verifying that each valve (sanual, l

power-operated, or aut=matic) in the flew pata is in its cor oct

. position,

b. At least once per 12 months, by cycling each testante valve in the flow path througn at least one completa cycle of full travel,

c. At least once per la sonths:

S

1) By performing a systan functional test waien includes::f=?rd

.ausamat h tf = cf - r;; =, xi -

4) Verifying t' t t.t r tf f:?;;; '- e: '* ;.; ;;O. *-

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By a visual inscoction of the dry pipe * ; :-t

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PR00F & REM CDPY PUNT SYST BS _, , , , .
t CD SYSTB5 LIMITING CONDITICN FOR OPERATION 3.7.10.3 The following '.n ' ;..=[CO Systems 2 sna11 he OPE.UBLE:
a. Ofesel generatogrooms and day tank rooms,
b. Lower cable spreading roce,
c. Auxiliary feedwater diesel room and cay tank rece, and w A .. _+' 0 g -
uxru.
. d. Otesel-driveg SW sake-up ousps and day tank rooms.
APPt.ICA8f .ITY: Whenever equipment protacted by the CCs systaes is required to be OPERA 8.E: -
ACTION:
4. With one or more of the above required CDs systass inoperanie, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous fire watch with backup fire suppression equipeant for those areas in whica redundant systaes or components could be damag ; for other areas, estan11sh an nourly fire watch patrol.
( g ,, ,)
b. The previsions of Specifications 3.0.3 and 3.0.4 are not applicante.
SURVE!LLANCY RfCUIRDENTS i
4.7.10.3.1 Eaca of the amove required 'a " unYC s Systans snall be demonstrated CPE.U8LE at least once per 31 days by verifying that esca valve (aanual, power operated or automatic) in the flow peta is in its correc-position.
4.7.10.3.2 Eaca of the aeove rsauired '_. ' x r M o, systems sna11 he demonst " OTG4d: gg m.g ,
a. At least once oer 7 days by verifying t e CD storage tanz level tot i
{ 290 be greater psia, andthan i 75% (7.5 tons) and pressure to be greater than I
b. At least once per la months by verifying:
( 1) The systas, incluainggsociated ventilation systas fire camcors, actuates tr us"; r_ automatically upon receist of a simulatac l actuati signal, and .fec e ._. t h , % A, I
Batk 2)- Flow free esca no=le during a " Puff Test.
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SYRON - UNIT 1 3/4 7-32
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C ox sbope bk level b be. preater A r} r y. O.n.e),
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3K 33 IABLE 3.7-5 (Continised)
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FIBE E SE STATIONS .
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8 80 Call EIEVAll0N HDSE BACK BEEL ANGtE VA M T Assu. Idg. (Cea inesed) gg g,g gy 3,;gA3 ,,,,
ofP324 I 1 P-it: "y : -J. ) . i I, rid ' * * * - 455 20 Ofr332 l J4er-hs;ET L-II: SV C *! -si ^.:'h
5 3a 1C 14sc lut M d 455 22 N-18:' frr"
_Noeth all-0-1 444 238
Of?463
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0- ag ...e a--- e, ---- ogwe3n c ,,g iggs r 443 e 2D8 se 7 gg, 209 5 0FP325
!!! h b t a .. ..a 443
{ P-10: "y u::t d::. te 7:
210 # DIP 326 j M-13: South wall of 7::: !!3. %csn c-t 443 P-l"* 4 4 3: West wall of ;:1 124 g,cm g.g 443 211
OfP328 l l OfP455 5-21: By cabinet 2RV01EC (;tle c. ten. 3 red 431 229 l [ 5-24: By U-2 ont. shield wall Leim. ven tred 431 230 OfP456 t
' 5-12: 431 237 ' OfP462 U By U-l cong. 4 sigteld wall (sta. pen, sna)
l*W_1. !.'.49't
.m reapethtag-room 430 52 OfP313
+ P-II: -
54 ofP342
a 4-19

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By U- toa-,e !en p. r= Mc.t vduc alsk. 430 55 OfP343 l w P-24: By raenaste evaporator 430
cad A.r6* 430 58 ofP319 V-17

OfP320
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g V-17: By u $,:
By 46- door. to ;. decon/
hject'e.. change , area cla 430 61 L-ll: By 4>-t wasta all tank room M 405 90 ' 0FP315 405 91 OfP318 P-18: By 91 evator OfP349 P-23: By spent resin pumps 405 92 93 OIP314 "90 gg .( -11: By laundry tanks bg&.3m M[g5 i
g ,g. - 1: _$9 U-2 a';- th; t. ;i.y .e d US 94 95 OfP348 ofP345 co
~ V-21:
U-2Iiydrogenrecombiner 405 *W h ""I,g ' T:15T' $ U-1 hydrocen recombiner re *-a! pr- ' 405 96 .0lP316 405 97 ofP317 po St*
Fs 5-15: , $ U-l p4p:
rrrl 4ee!!e" %sa_.
W old N s 368 -
130 OfP373 j
I N-II: By the rec ci H-li: Oy the a o -
368 '
131 132 ofP374 OIP369 3 h
g-P 14: By panel P 4J8 _
368 1
l ? s t.;-20: By % of U-2 A N 84le44 useps 360 368 133 OfP355 OfP356
! 3 P-21: By the h ow own casulenser f 134 o
--4ter te- ' 368 135 OlP361 c:3 .
l-25: By the sacyeh := ,rreur l
368 136 OlP357 ~j'0 j N-25: By regenevatl-- -' gdrain ahemic1tl tankt 368 138 OfP362 5-18: By panel IP186J i Q-It: By Aux. Sldg. floor drain tanks 368 139 OlP368 -
U-15: By positive ,ilsplacement-charging-pump 360 140 ofP372
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3/4.7.n FTRE RATED ASSEMLIIS -
LIMmM8 CCMCITICM PCR OPGATION 3.7.n All fire rated assemeltes (walls, floer/cailings, caale tr f enclosures and other fire barriers) separating safety-related fire areas or sacarating persions of redundant systems tapertant ta safe shutdown wiuin a firw area and all sealing devicas in fire rated assemalf penetrations (fire doors, fire
. windows, fire daspers, caele, piping and ventilation auct penetration seats) snell be QPGASLI.
AP'UCASIUTY: At all tiens.
snesa.
With one or more of the above required fire rated assemeltes and/or
' sealing devices inopersale, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months eitaer estan11sh a contin- ~
unus fire watch on at least one side of the affected assensly, or verify the OPGA8IUTf of fire detectars on at leest one side of the inoperaale assembly and establish an hourly fire weten patrol,
b. The provisions of Specifications 3.0.3 and 3.0.4 are not aoolicaale.
SURvtit. LANCE RECUIRE?ntfrs 4.7.11.1 At least onca per la sentas the aseve required fire barrier penetra-tions and penett.tfon sealing devicas shall be verified CPGASLE by3 -':---iq S
) y g a visual inspection of:
y. The exposed surfaces of each fire rated assemoly, 2 #. Each fire window / fire dasper and associated hardware,
~
j 3g. ' At least 105 of each type of sealed penetration. If apparent enanges in appearance or annereal degredations are found, a visual inspection
of an additional 105 of each type of sealed penetration sna11 De
' ande. This inspection process shall continue until a 105 sasole j
- with no apparent enanges in appearance or annergal degradation is found. Samples shall be selected such tnat eacn penetration seal will be inssected every 15 years # a 4 4.7.11.2 Each the above requ fire doors shall be verified CPGA8LE by 5 inspecting th _
^^
_..- ~M : releasa and closing secnanism and latenes 4 l at least onca e 5 sentas, and erifying:
a. The CP G A8I UTY of the Fire Deer supervision Systas for each electrically supervised fire door by perforeing a TRIP ACTUATING OtVICE OPG ATIONAL TEST at least ones per 31 days,
b. That each locked closed fire door is closed at leasti onca per
7 days, and
, c. That each unlocked fire door without electrical supervision is closed at least ones per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
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s ATTACHMENT I (Section 3/4.8)
1) Section 3.8.1.1 Action Statement C.2. (pg 3/4 8-2)
Insert between the words " pump" and "and" the words "is operable". Also, delete the words " Division 21, diesel generator is operable" and replace with "2A diesel generator is capable of providing power to Bus 141." (See Below)
2) Section 3.8.1.3 (pg 3/4 8-X)
The attached section represents a new tech spec submittal for the 2A diesel generator. This tro and Surveillance represents the Byron Station action statement and surveillance requirements for the 2A diesel generator. The above change to pg 3/4 8-2 maintains a consistency with this addition.
b
=. . : . .. .. ....:.... .: . . .. .. . . . . .
. . .._ . . - . .- - _.. . . . ....__ .. . . . ..._..~.
ELECTRICAL POW H SYST ES h -
LIMITING CONDITION FOR CPGAT!CN .
1 ACTION (Continued)
c. With one diesel generator inoperable in addition to ACTION a. or b.
g@k above, verify that: (gg L All required systmas, subsystaes, trains, coeconents and devices ! T that depend on the resnining CPERA8 M diesel generator as a Q6
. source of emergency power are also OPERA 8LE, and e$
i i
2. When in MODE 1, 2, or 3, the diesel-driven auxiliary f tar eqg ~
numreand the Pi:;i= " tir:? m a S 1"i"'"'_, f the inoperable diesel generator is the emergency power supply for M the motor-driven auxiliary feedwater pump.
l If these conditions are not satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be in at least NOT STAN08Y within tha next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD ShUTCCWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
. d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPEA8ILITY of two diesel generators by performing Specification 4.8.LI.2a.4) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, unless the diesel generators are already operating; restore at least one of"the inoperable offsite sources to CPGABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STAN08Y within the next t hours. Wit!r only one offsita sourca restored, restore at least
.. , two offsita circuits to CPGASLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months frem time of 1 initial lots 0:- be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
e. With two of the above required diesel generators inoperable, demonstrata the OPE?A8ILITf of two offsita A.C. circuits by perfone-ing Specification 4.8.LLla. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least onca per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereaftar; restore at least one of the inoperable diesel
' generators to CPEA8LE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLJ SHUTDCWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore at least two diesel generators to OPEA8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months acd in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE REQUIREMENTS - 4.8. L L 1 Each of the above required independent circuits between the offsita transmission network and the Onsite Class IE Distribution Systas shall be:
a. Determined OPGA8LE at ler.st once per 7 days by veHfying correct
j. breaker alignments, indicated power availanility, and
! b. Demonstrated CPGA8LE st least once per IS sonths curing shutdown by l' transferHng sanually unit power supply free the nonsal circuit to the alternata circuit.
4.8. L L 2 Each diesel g:nerator shall be demonstrated 0FE8.A8LE:
j.
BYROM - UNIT 1 3/4 8-2 m.
69-m- -meg.
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3.4.8 ELECTRICAL POWER SYSTEMS 3/4.6.1 A. C. SOURCES OPERATINC LIMITING CONDITION FOR OPERATION 3.8.1.3 The 2A Diesel Generator shall be capable of being manually started and crosstied to Bus 141.f APPLICABILITY: Modes 1, 2, and 3 ETION: With the 2A Diesel Generator incapable of being manually started and crosstied to Bus 141, restore the diesel generator to capable status within 7 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE REQUIREMENTS 4.8.1.1.4 The 2A diesel generator shall be demonstrated capable of providing power to ia - 1 .m n m.
a. At least once per day by:
1. Verifying the day tank level is greater than 450 gallons.
2. Verifying DC control power is available to the 2A Diesel Generator local control panel (2PLO7J).
3. Verifying that at least one starting air receiver is at greater then 175 psig.
! 4. ing Verify's pne; 'the Essential
'--d Service to supply cooling Water System is requirements.
l l - b. At least once per 31 days by: ,
1. Verifying the diesel generator starts manually and operates with a load of greater than or equal to 5500 KW for one half hour.
c. At least once per 18 months.by:
1. Verifying the diesel generator een be crosstied to Bus 141.
4g gu,d f loCFR.5o.%b.)h-)d-c /nd %
du-Q ha. pl. f W k Y U 5-BYRON - UNIT 1 3748-K '
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(0464M) l r
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ATTACHMENT J (Section Bases 3/4) 1)- Section 3/4.1.2 (pg. B 3/4 1-2) Botation Systems Change "200*F" to "350*F" in the Second paragraph.
~
(pg B 3/4 1-3)
Change "200*F" to "350*F" in the Eirst paragraph.
This section specifies that at least two boton injection flow paths shall be operable in modes 1, 2, and 3. However, the bases for this section explained the requirement as being applicable above 200*F.
Change "350*F" to "330*F" in the second paragraph.
'2)' Section 3/4.4.1 (pg. B 3/4 4-1) Reactor Coolant Loops and Coolant Circulation Change "I hout" to "6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months " in the first paragraph.
The action. statement indicated that with less than the above required teactor coolant loops in operation. ba in at least HOT STANDBY within 6 hrs. This was not consistent with the bases provided for this Technical Specification which required placing the plant in HOT STANDBY within I hout with less than the required reactor coolant loops in operation.
3)' section 3/4.4.2 Safety valves (pg B 3/4 4-2) i' In the second paragraph following the words " loss-of-load assuming no l
coactor trip" delete the words "until the first Reactor Trip System Trip l- Setpoint-is reached (i.e., no credit is taken for a direct Reactor Trip on l the loss-of-load)".
The sizinq of.the Pressurizer safety valves, as stated in the BYRON overpressure Protection Report, is based upon a loss of load event j assuming no teactor trip or power-operated relief valves.
l The 1.oss of 1.oad Analysis in FSAR Section 15.2 is the limiting transient with respect to overpressure. This analysis verifies that the safety valve capacity (alteady sized) is sufficient to maintain reactor coolant system pressure below 110% of design pressure (2750 psia). In this analysis, no credit is taken for a reactor trip from the second teactor trip setpoint signal which is either ovettemperature AT or high pressurizer pressure.
l-
+
ATTACHMENT J (CONTINUED)
(Section Bases 3/4)
4) Section 3/4.4.9 Pressure / Temperature Limits (pg B 3/4 4-15)
- Change "HPSI" to " centrifugal charging". This is for clarification.
The following should be added at the end of the last paragraph on pg B 3/4 4-14.
"These two scenarios are analyzed to determine the resulting overshoots assuming a single PORV actuation with a stroke time of 2.0 seconds from full closed to full open. Figure 3.4-4 is based upon this analysis and represents the maximum allowable PORV variable setpoint such that, for the two overpressurization transients noted, the resulting pressure overshoots will not exceed the nominal 10EFPY Appendix G reactor vessel NOT limits."
This is being added in response to an NRC question concerning the PORV Setpoint curve.
5) Section 3/4.5.3 (pg B 3/4 5-2) ECCS Subsystems Change "350*F" to "330*F" in the first paragraph.
6) Section 3/4.6.1.5 Air Temperature (pg B 3/4 6-2)
Delete the word " listed" between the words "made at all" and " locations" and replace with "of the listed running fan". Also, delete the word "priot" and cnange " determining" to "deteemine". This change is made Cor consistency to the 140 Surveillance Requirements.
7)~ Section 3/4.8 (pg B 3/4 8-1)
Insert the following as the last paragraph: "The reporting requirement of 10CFR50.36 do not apply during the first eight houts in the Action Statement."
This requirement is necessary to complete the Pre-Operational Test Program on Unit 2. The Unit 2 Auxiliary Power Test will require breaker interlocks to be checked and this will affect the Unit 2 Diesel Generator
-output breaker. Diesel Generator retesting may also be required. This exception to the normal reporting method will only remain in affect until-fuel load on Unit 2.
4
. . . . . . ..___;. u ... _ . . . . . . . - . .
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r g3 RDCTT/ITY CONTROL SYSTBS h g kh hk BASES M00 GATCA TD8PERf URE C"OEFFICI D T (Continued) ,
The east negative MTC value equivalent to the most positive moderatse .
density coefficient (MDC), was obtained by incnnentally correcting the MCC used in the FSAR analyses to nominal operating conditions. These corrgetions
' . involved suttracting the incremental caange in the NOC associatad with a care candition of all rods inserted (most gesitive MC) to an all rods withdrawn conditiert and, a conversion for the rate of change of moderator dansf ty wita
, g. temperature at RATED THERMAL POWS conditions. TMs value of the MCC was then transformed into the lieiting MTC value -4.1 x 10
Ak/k/*F. The MTC value of -3.2 x 10
Ak/k/*F rapresents a conservative value (with corrections for burnup and soluble boron) at a care condition of 300 peu equilibrius boren concentration and is odtained by asking tasse carrections to the limiting MTC value of -4.1 x 10
Ak/k/*F.
'2 The Surveillance Recuirements for seasurement of the MTC at the beginning and near the end of the fuel cycle are adequata to confira that the MTC remains l
witMn its Itaits since this coefficient changes slowly due principally to tae l reduction in RCS boron concantration associated with fuel burnus.
3/4.1.1.4 MINIMun TypgATugg pCR CRITTMITY ,
TMs specification ensures that the reactor will not be made critical with the Reactor Coolant Systam average tamperature less than 550*F. TMs limitation is required to ensure: (1) the sederator tammerature coefficient is witMn it analyzed temperature range, (2) the trip instrumentation is within
!"S its namel operating range, (3) the pressurizer is capable of being in an l~ OPGA8LE status with a steam bunale, (4) the reactor vessel is amove its sinimum RTg tamperature, and (5) the plant is amove the csoldown staan dumo permissive, P-12. .
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!D 3/4.1.2 80AATTOM SYSTees .
The Baron Infection System ensures that negative reactivity control is ave 11aale during eacn mode of facility operation. The enaconents recuired ta perform tMs function include: (1) borated watar sourcas, (2) charging pumps, (3) separata flow patas, (4) boric acid transfer pumos, and (5) an emergency
.7 power supply free CPGA8LE diesel generators. 3 5cF" '
With the RCS average tancerature above da sinimum of two boren injection flow patas are required ta ensure single functional cacanility in the event an assumed failure renders one of the ficw patas inecerante. The boration cacao 111ty of either flow pata is sufficient to provide a SHUTUChN MARGD free expec.ad coerating canditions of 1..I5 ak/k after xenon cecay and cooldown to 200*F. The saximus expec ad : oration cananility requirement
r. oc=urs at ECL free full power equilibriue xenon concitions and recuires 15,780 gallons of 7000-som scrated watar free the toric acid 'starsge tanks or 70,450 gallons of 2cco-Spa borated water free the refueling watar storage tank.
SYROM - UNIT 1 83/41-2
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RDCTT/ITY CONTROL SYSTUS Pi100F & HE'# COPY l BASIS 90RATTON 'YST B S (Continued) .
With the RC5 tamperature below Mone Beren Injection Systas is ..
accsotaale without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTBATION$ and positive reactivity enanges in the event the single Soron
Injection System becomes inoperaale.
The limitation for a anaism of one centrifugst enarging ptme to be CPGASLE and the Surveillance Requirement to verify all anarging pumos except the required CPGA8LE pump to be inoperable below provides assurance tdat a mass addition pressure transient can be reliev by tae operation of a 5 single P08V. .
M o*p The boven capanility required below 200*F is sufficient to provida a SWTDOWN MARGIN of 1% .ik/k after xenon decay and coeldown from 200*F to 140*F. This condition requires eitner 2.552 gallons of 70CC pon horatad water from the beric acid storage tanas or 11.8a0 gallons of 2000-ppe horated water from the refueling water storage tank (15d5T).
CC218.
Ther contained water volume limits include allowance for wter not ava11 ant o g8 because of discharge line location and at.%e pnysical charactaristics.
The limits. on contained water votes and baron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within contaitument after a LOCA. This pH band sinimizas the evolution of
, iodine and sinimizes the effect of chlorida and caustic stress corrosion on anchanical systans and components.
The CPGA8ILITY of one Baron Injection Systas during REFUELING ensures that this systam is ava11aale for reactivity control while in MODE 6.
i- 3/4.1.3 MOVA8LE C0fm 0L A55 BOLI!5 4
' The specifications of this section ensure that: (1) acceptante power .
distribution limits are asintained, (2) the sinisus SWTDOWN MulGIN is amin-tained, and (3) the potential effects of rod sisalignment on associated accident analyses are 11mitad. QPGASILITY of the control red position i inetcators is required to determine control rod positions and taersey ensure compliance with the control red alignment and insertion limits. The 31gital Position Indication Systas does not indicate the actual position of the snutcown rods betmoen la staps and 23 staps witadrawn.
The ACTION stataments whics permit limitad variations from the basic requirements are ac=capanied by additional restMetions vnfen ensure that the
- original design cMteria are set. Itisalignment of a red reouires seasurement of peaking factors arts a restMetion in THUMAL POWG. These restM etions
-- . provida assurance of fuel rod intagMty duMng continued ocerstion. In aceition,
, these safety analyses affected by a sisaligned red are reevaluated 'a confir s taat the resulta reonin valid during future operation.
EYRON - UNIT 1 8 3/4 1-3
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I 3/4.4 REACTOR COOLANT SYSTEM {].} g g g g .
BASES 3/4.4.1 REACTOR CCOLANT LOOPS AND COOLANT CIRCULATION _
The plant i designed to perate with all reactor coolant loops in operation and maintain DN8 dn I 2 d ring all normal operations and anticipated transients. In DES 1a[w with one reactor coolant loop not in operation this specification _g ... that the plant be in at least HOT STAN08Y within g, 5 hout$. &g rHf AAG.icASLf 6 rarsty ANnwsa Div8A In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERA 8LE.
In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay. heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERA 8LE. .
In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.
,I The operation of one reactor coolant pump (RCP) or one RHR pump provices s adequate flow to ensure m.xing, prevent stratification and produce gradual reactivity changes during boren concentration reductions in the Reactor Coolant System. The reactivity change rate associated with baron reduction will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a reactor coolant pump with one or more RCS cold legs less than or equal to 380*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant Systas, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of .
Appendix G by rest ~ricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.
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l REACTOP COOLANT SYSTEM 8ASES 3/4.4.2 SAFET'f VALVES i .
The pressurizar Code safety valves operata to prevent the RCS free being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated staas at the valve 5etaoint. The
! relief capacity of a single safety valve is adequata to relieve any overpressure l condition which could occur duMng shutdown. In the event that no safety
. valves are OPERA 8LE, an operating RHR loop, connected to the RCS, provides -
overpressure relief capability and will prevent RCS overpressurization. In addition, the Ovenressure Protection System provides a diverse seans of protection against RCS overpressurization at low tamperatures.
During operation, all pressurizer Code safety valves must be OPERA 8LS to prevent the RCS from being pressurized above its Safety Limit of 2725 psig.
~~
. The connined relief capacity of all of these valves is greater than the saximum surge rate resulting from a complete loss-of-load assuming no Reactor tMp
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operation of the power-operated relief valves or staan dung valves.
1 Demonstration of the safety valves' lift settings will occur only duMng shutdown and will be perfomed in accordance with the provisions of Section XI of the ASME Soiler and Pressure Code.
3/4.4.3 P9E550RI7E_R The limit on the seximum water volume in the pressurizer assures that the parameter is saintained within the normal staacy-stata envelope of operation assumed in the SAR. The limit is consistant with the initial SAR assumptions.
The 12-hour peHodic surveillance is sufficient to ensure that tne parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a staas bubble is formed and thus the RCS is not a hydraulically solid systas. The requirement that a ainimum nunner of pressuMzer heatars be OPERA 8L2 enhances the casatilit/ of the plant to control.Reector Coolant Systas pressure and estaolish natural cirtulation.
3/4.4.4 RELIES VALVES The power-operated relief valves (PORVs) and staas subole function to relieve RCS pressure duM ng all design transients un to and including tne design stac load decrease with staan dump. Cperation of the PORVs sinimizes the undesiracle ocening of the spring-loaded pressurizer Code safety valves.
Each PORY has a remotaly ocerated Diock valve to provice a positive shutoff capanility should a relief valve become inopersale.
3 3/4 4-2 SYROM - UNIT 1
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REACTOR C00t. ANT SYST1!M j FR00F & REyg cgpy SASES PRE 55URE/TEMPERAT1JRE LIMITS (Continued) l Although the pressurizer operates in tamperature ranges aeove thost for which there is reason for concern of nonductile failure, operating limits ,
are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.
The OPERASILITY of two PORVs or an RCS vent opening of at least 2 square inches ensures that the RCS will be protected free pressure transients which
" could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 380*F. Either PCRV has adequate relieving capability to protect the RCS free overpressurization when the -
transient is limitad to eitner: (1) the start of an idle RCP with the secon-dary water temperature of the steam generator less than or equal to 50*F aeove the RCS cold leg tamperatures, or (2) the start of a MPtI pimo and its injection into a water solid RC3. ,__ h ,
/4.4.10 STRUCT1JRAL INTEGRITY 7* 7 The inservice inspection and testing programs for ASME Code Class 1, 2,
- and 3 components ensure that the structural integrity and operational readiness t.
of these components will be maintained at an acceptacle level througnout the life of the plant. These progress are in accordance with Section XI of the
. ASME Soiler and Pressure Vessel Code and applicaale Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by t.%
Consission pursuant to 10 CFR 50.55a(g)(6)(1).
Components of the Reactor Coolant System werv. designed to provide accass
) to pemit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code,1974 Edition and Addenda through Summer 1975.
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SYRON - UNIT 1 B 3/4 4-15
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ECC3_ SUS 5YSTPS (Continued) l The limitation for a seximum of one centrifugal charging pump to be j OPDABLE and the Surveillanca Requirement to verify all caarging pumos and..
Safety Injection pumos except the required CPGASLE charging pump to se inoper-S. able below provides assurance that a mass addition pressure transient can be relieved operation of a single PGRV.
H o*F The surveillance Requirements provided to ensure OPGA8ILITY of eacs congenent ensures that at a sinism, the assations used in the safety analyses are est and that subsystes CPGA8ILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECC3 flows will be amintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping systas to each
injection point is necessary*ta: (1) prevent total pues flow free exceeding Mmous carutitions when the systan is in its sinima resistance configuration, -
(2) provide the proper flow solit between injection points in accordance with the assumptions used in the ECC5-LOCA analyses, and (3) provide an ac:sotaole level of total ECCS flow to all injection points equal to or amove that assumed in the ICCS-LOCA analyses. The Surve111anca Requirements for leakage tasting of ECCS checx valves ensures that a failure of one valve will not cause an .
intersystes LOCA.
%f4. 5. 4 REFUst "eG VATU STCRAGE TAME The ILITf of the refueli tar sta tank ( . part of the t a sufficient suoply a water is le for inj fon ECCS ensures nimm 've.lume cron by the ECCS in event of a LOCA. ts on ent weiar i a11aale witain inment to e
-^-- & .sion ens Nat: (1) su A
s
.I pe irculation ing to the core ' (2) the reactor tae , remain water 1on followi zixi f the h57 suscritica a the eslo .ive contro volumes with a cent r*=d - ~ t for tab aost istant with the analyses.
assemely. These tions are allonence not The ined wata limit inc line locati e other pnysica aractari stics. ,
e of tant h
e Th also ensure ts contained vel alue of between 8 and 11.0 for tae and baron v - se- tion of Me h 5T ion recircutatad Mftar a LOCA. band aini es the witain eff is of calori -
tic stress cc lution ionof o \n f odine, sinimites' l
ical systmas and nents.
I 3YRCN - UNIT 1 5 3/4 5-Z
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i A-CONTAIM8ENT SY N 8ASE3 3/4.6.1.5 AIR TDPERATURE oE
& f Fp,J The limitations on containment average air erature ensun that tae overall containment average air temperature does a t exc the initial
( tassersture condition assumed in the accident anal is fa a staas line break accident. Measurements shall be ande at al listed ocations, =netner-
! by fixed or portable instruments, .po6ee to detarsintap tae average air temperature. " .
l 3/4.6.1.6 CONTAINMENT VESSEL. STRUCTURAL INTTGRITY
-- This Itzitation ensures that the structural integrity of the enntainment f
will be asintained comparante to the original design standards for tae life of the facility, Structural integrity is required to ensure taat tae containment will withstand the maximum pressure of 43.9 psig in the event of a estd leg double-ended break ac=1 dent. The seasurement of cantainment tanden lift-off force, the tensile tasts of the tendon wires or strands, the visual examination l of tandons, anchorages and exposed intarior and extarior surfaces of tae containment, and the Type A leakage test are sufficient to demonstnte tais capability.
Repos*d The Surveillanca Requirements for demonstrating tae containse '-
gyg stmetural integrity are 1 latory l
Guide 1.35, "Inservica Sm) Memo 44anceutad eillanca withTendons the rec:amencations in Prestressed Concre of i Containment Structures ^; . _ 7 and osed Regulatory Guide 1.35.1,
"Satarmining Prestress n gress for Inspect 1 of Prestressed Concreta j Cantainments," April 1 .
jg The required Special ortsfromanyendneeringevaluationofcontainment annomalties shall include a %agtae tendon enndition, the condition of the concreta (especially at tanden ancaerages), the inspection precedure, the tolerances on cracking, the results of the engineering evaluation and tae corrective actions takan.
3/4.6.1.7 CONTAIMsENT VENTI 1.ATTCM SYSTD The de-inen containment purge supply and exaaust isolation valves are required to be sealed closed during plant operations sinca case valves have not been demonstrated casante of closing during a '.CCA or staan ifne break 4ccident. Maintaining tassa valves sealed closed curing plant oceration ensures that exces;ive quantities of radioactive meterial wiii not be released via tae Containment Purge Systas. To provide assurance that tne at-inen c:ntainment i
valves cannot to inadvertently coened, the valves are sealed closed in ac:crt-
! anca wita Standard Review Plan 6.2.4 wnien includes secnanical devices .o seal or lock the valve closed, or prevents power free being sucolied ts the valve operator.
BYROM - UNIT 1 8 3/4 6-2
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O 3/a.3 ELIC-"t!".At. 8CaG 5'/5 P5 2ASE3 3/a.9.1. 3/a.3.2 Aho 3/a.3.3 A.C. ICURC ~S. 3. C. ! CURT *15. ANO MS- ? =c=G M5TU307 CM The CPGA8IL'Tf of =e A." and 3.C =cwer scurtas and associatad dstMbution systans curing :nerrtion ansures tnat sufficient ;cwer will ta ava11aale to sucoly tan safae/--statad ecui:mont neu1 red f:r: (1) == :afe snutdown of us facility, and (2) ue sitigaden ana c:nt 31 of ac=1 dent =end-
. .tions vitain t3e facilit/. De sinisua stecified indecencent and mcuncant A.C. and 0.C. sower scurtas and dstMbuden systans satis?f =e ncuirssents of General Casign Critarion 17 of Ascendx A ts 10 CM Part 50. .
The A"CTICM mou1 resents scocified for tne levels of degrscation of =e power scurtas pesvice restMetion .ocn c:ndnued facili?/ ::eradon ==usensurate vita =a level of coqracation. The CPGA8IL7f sf me :cwer scurens an consistant wic tas initial sndidon assumoticns of us safar/ analyses and are tased untn saintaining at least :ne -scuncant sat of :nsita A.C. anc 3.C.
ower sourcas and associar-ad dstMtution systans CPGASLE curing ac=1 cent canoidens coincicant wits an assumed loss-af-stdsits ;cwor ano single failurg O "
of ne ot..ar nsita A.C. scures. Ths A.C. and 0.C. scures allcwente :vt-sf-sorties tians are tased on Regulatsef Lice 1.93, *Ava11acilt;f cf ElectMeal
__ . Power 54urcas,' Cecascer 1974 'when one diesel generztse is inecencle, =ers ,
is an acdticnal AC7*CM recuiriment ta veMP/ cat all neuimd systans, suosystans, trains, =moanents anc cavicas, cat cacend en =e enaining CPGA81.1 desel genersue as a scurts of emer;ency ;cwer, are also CPEA81.5, and taat one desel- Hven aux 111 arf feecwatar su=m is CPGA8LE. This escuire-sent is intanded tm :ravice assuranca cat a loss-of-offsita ;cwor event will not result in a camalete loss of safety func.1on of Mdcal systans cuMnq l tne period one of tae diesel generstars is inecersale. The tar-s veriff as used
' ~
fn mis contart means ta acainistratively =ecx by examining icgs or cuer information to catarsine if tartain c=moonents are out-of-service for saintenanen or ocer reasons. It does not seen ta : erd:n == Surveillanca Requirements needed ta eclenstrata ce CPGASIL'Tf of =e ::moonent.
The OPGASII.'.Tf of the sinimum scocified A.C. and 0.C. power smurtas and associatad dstMbution systans curing snucewn and -sfueling ensuns mat:
(1) tae facility can :s saintained in =a snuucun or -ifveling c ndsten for extenced due ; ericas, and (2) sufficient inst: .:montatien and ::nt:-:1 :.scacility is ava11aale for eenitoring and saintaining tas unit status.
- The Survei11anca tecuiriments for canonstrating =e OPEASIL*?f of ce desel generturs are in ac srunca vita me asc=amencadons of iequiaurf Guices 1.9, *!alecdon of Of esel enersur !at C.anacity f:r Stancey 8 ewer Sucolies,' . war c 10,1371.1.108, *hriccic Testing of 01esel Generter units used as Onsita Elec-Mc pewer Sys. ass at Muc!aar ? wer 81 ants,* tevis'en 1.
August 1377, and 1.137, %si-011 Systans for Stancay 3tessi Genertum,'
Revision 1, Ocucer 1979.
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ATTACHMENT K (Section 6.0)
1) Table 6.2-1 (pg 6-4)
- Replace the present table with Number of Individuals Required to Position Fill Position Modes 1, 2, 3, & 4 Modes 5 & 6 SE 1 1 SF 1 None M 2 1 A0 2 1
, _SCRE- 1 None or, whenever a SCRE (SRO/STA) is not included in the shift crew composition, the minimum shift crew shall be as follows:
Number of Individuals Required to Position Fill Position Modes 1, 2, 3, & 4 Modes 5 & 6 SE 1 1 SF 1 None RO 2 1 A0 2 1 .
STA 1 None SE - Shift Supervisor (Shift Engineer) with a Senior Operator License on Unit 1 SF - Shift Foreman with a Senior operator license on Unit 1 Ro - Individual with a Reactor Operator license on Unit 1 AO - Auxiliary Operator SCRE - Station Control Room Engineer with a Senior Reactor Operators
License on Unit 1.
STA - Shift Technical Advisor.
I
ATTACHMENT K (Section 6.0)
Delete "(other than the Shift Technical Advisor)" which appears twice in the second paragraph.
This change is consistent with the LsSalle Technical Specification minimum shift crew composition.
2) Section 6.2.4 (pg. 6-5)
Delete "ONSG" as last word of paragraph and replace with " Office of Nuclear Safety" as per Operating Experience Feedback BAp 1260-1.
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S TAst! 6.2-1
.. MINIMUM SHIFT CREW CCMPOSITICM POSITION NUM0ER off INDIVIDUAL 5 REQUIRED TO FIU. POSITI,C
[$ MODES 1, 2, 3 & 4 MODES S g S3 L ~1 SM 1 km
" 20
2 1 M 4 1 5 3 STA 1 Mone Shift Supe er (Shift Eng N ) with a Senior Operator E-11 Unit 1 I fdual with a Senior Operator lies on Unit 1 20 - ndividual with an Operator license on Unit Ad Auxiliary Operator
-
Shift Technic 21 W a--
The Shift Crew Coseeition any be one less than the minisim requirements of Table 6.2-1 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 acesamodate .
unexpected esence of err-duty shift crew saaners provided immediate action is taken to restore tas Shift Crew Comossition to within tne einimum requirement.s
-- p/ -
of Tante 6.2-1. This provision does not posit any shift crew position to be unmanned upon shift change due to an oncoming snitt creman being lats or assent.
During any absence of the Shift Superviser from the control room while the
Unit is in MODE 1, 2, 3 or 4, an imHvidual (Cr r. . ;.. :t.i?. %.i...; ;""
M.;-.71eith a valid Senior Operator license shall be designated to assume V, the control room command function. During any assenes of the Shift Supervisor free the control room while the Unit is in M00f 5 or 6, an individual with a valid Operatar If canse -(? r er N Sh'"_ 'r-'r? ?3;t:Pshallbe designated to assume the control room command function.
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PR00F & HEW COM
^
> ADPINISTRATIVE CONTR01.5 ORGANIZATION (Continued)
e. The Onsite Nuclose Safety Group (CNSG) shall function to examine unit operating charactaristics, NRC issuances, indust 7 advisories, t.icensee Event Reports and other sourcas of plant design and o'oerating amperience information, including plants of similar design, which any r indicata areas for improving unit safety. The ON5G shall be composed of at least four, dedicated, full-time engineers of multi-disciplines located on sita and shall be augmented on a part-time basis by personnel free other parts of the Commonwealth Edison Cossany organiza-tion to provida expertise not represented in the group. The CMSG shall be responsible for maintaining surveillance of Unit activities to provide independent verificationa that these activities are per-l9 "
formed correctly and that himman errors are reduced as such as practical. The ON5G shall make detailed recommendations for revised i
procedures, equipment modifications, maintenanca activities, opera-
tiens activities or other means of tuproving unit safety to the Director, Nuclear Safety. .
f. The Station Control Rose Engineer ** (SCRE serves as the lead control -
' room Itcansed Senior Operator during ncesal operations and as the ,
Shift Technical Advisor (STA) during abnomal operating and accident conditions. In the event of abnormal operating or accident conditions, ~
the SCRE will relinquish his job as contr:. room licensed Senior Operator to a Shift Supervisor and will assume the role of STA. Then he shall provide technical support to the shift in the areas of 9 thermal hydraulics, reactor engineering and plant analysis with regard 3 '
to the safe operation of the unit. .
6. 2. 3 Qualifications of the station annagement and operating staff shall aset sinimum acceptable levels as oescribed in ANSI N18.1, *5 election and Training of Nuclear Power Plant Personnel," dated March 8, 1371. The Rad /Chen Super-viser or I. sed Health Physicist shall aset the requirements of Radation Protection lC Manager of Regulatory Guida 1.8, September, 19T5. The Itcansad Operators and Senior Operators shall also amet or exceed the sinimum qualifications of the supplemental respiirements specified in Sections A and C of Enclosure 1 of. the March 28, 1980 NRC letter to all licensees.
6. 2. 4 The content and quality of retraining and replacement training programs for the unit staff shall be enintained under the direction of the PMouction Training Geoartment. The Station Training Department, under the functional control of the Production Training Department, shall have responsibilities for training activities at the station. These training programs shall meet or exceed the requirements and recammendations 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 1sttar ta all Itcansees, and shall include familiarkation with relevant industry operational amperienca identified by the 4NSS:C ogo..sy.n SQ
" Met responsible for sign-off feature. ,
I
'"'The position of SCRE can be filled by a STA provided a licensed Senior 1 Operator is in the control room.
I syn 0N - uNrr i e-s
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. . . . . . . .. '?~ E- ~-w -
i . - -_-_- - _ _ _ . _ - - - -
9 BAP 1260-1 Revision 1 OPERATING EXPERIENCE FEEDBACK A. STATEMENT OF APPLICABILITY The p.arpose of this procedure is to outline the method to provide for effective and efficient feedback of operating experience to plant personnel.
This procedure has the objective to assure that operating information pertinent to plant safety originating both within and outside the utility organisation is continually supplied to operators and other cognizant personnel, and is incorporated into training and re-training programs, and evaluated for possible modifications or plant procedures.
B. REFERENCES
1. 'NUREG-0737, iten I.C.5.
2. CECO. Operating Experience Review Program
-C. MAIN BODY 1; operating experience assessment and dissemination is the responsibility of the Administrative and Support Services Assistant Superintendent. If applicable, the Training Supervisor is responsible for incorporation of this information, once obtained, into appropriate training and re-training programs. The Operating Assistant Superintendent is responsible for ensuring dissemination of information to Operating Department personnel.
2. Operating experience information is received at the Station from many sources; among which are:
a. Operating Experience Assessment Report Transmittals (OPEX).
b. Deviation Reports.
c. Off-Site Reviews.
d. NRC Bul'.ecins, Circulars, Information Notices, and correspondence.
e. LER Monthly Reports,
f. y, Technical bulletins and vendor advisories.
3. Information pertaiains to the above items is handled as follows:
a. OPELare received e at the Station from Office of Nucle (Safety. The e include NRC Power Reactor Events, ::CC/"f.'.C J 51.mifium t Operating Experience Reports, NSAC Annual Screening and Evaluation'of License Event Reports, Note pad items and DVR's from Ceco stations. Upon receipt, the Administrative and Support Services Assistant Superintendent reviews the report APPROVED and routes it to the Licensing Coordinator who logs the report,
di"**** "'i'1"8 *"Y "S"i"d "8ponse , and obtains seacion OCT 0 71983 review approval. The original copy is routed to the Training Supervisor for dissemination t) the SCRE and License personnel B. O. S. R. indicated on the routing cover sheet. The Training Supervisor is responsible for maintaining a record of the routing completions.
-1
_ amme-es e4- -=
y9-Tw' - ---ry&-------yw& gr e 1-t y--w-M-e,--w*+ pr v g 7 9y -um e--y.--- . - -eww -.-n-e---r-s.y-_
ATTACHMENT L (Tavg SetPoint Change)
1) Table 2.2-1 (pg 2-9)
In NOTE 1, change "587.7" to "588.4" for the Tl setpoint.
2) Table 2.2-1 (pg 2-11)
In NOTE 3, change "587.7" to "588.4" for the Til setpoint.
3) Section 5.4.2 (pg 5-4)
Change "587.7" to "588.4" The above changes to Tavg are being made to reflect the PLS setEcints.
I
i J I, .I u a ,g
_ g. ,
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taste 2.2-1 (Centlaued) ,
TAttf alDTAIIetBS (Centlaued) i ilelE 1: (Centlaued) i E i 4 se = Tlas constaat uttilmed la the seasured T*
las cesponsater, av a e s, 588.4 e e
', T* 1 tet:fT (lienlaal I,, at SATES TilflBIAL reWER), ,
i N. = a.e8134, a
P = pressuriser pressure, psig, p' = 2235 psig (liesla41 SCS aparattag pressure), .
I 5 = Laplace transform operator, s 8, :j f i and f (AI) is a function of the ladicated difference between top and bettes detectors of the r power raeGe mentree les chambers; with galas to be selected based en measured lastrennent respeas. during plant STAR 1HP tests such that:
]
I g he tusen -425 and -65, f (AI) = 8. uhere q g and g are percent j t (I) for q g RATES IllEgi4As. ;esf2 la the top and betten halves of the care respectively, and qg
g is i-total illfA144L POWER la percent of RAlf8 IlifRMAL POWES; (II) for each percent that the magnitude of qg g exceeds -425, the AT Trip setpelat 7, shall he automatically reduced by 2.865 et its value at RATES TitEAHAL POWES; and $
13 l (Ill) for each percent that the magnitiede of gg g exceeds -65, the AT Trip 5etpalat LQ shall he autenatically reduced by 1.865 of lts value at RATES TitERHAL POWER. p
cs i
hamsel's maxlaus Irly Satpalat shall not exceed its computed Trip intpoint by more than b j
x.
$00fE 2:
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g g
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IAttf f.3-1 (Centlanned) l i ,
Y = f 1181A118185 (Centlaume) 4 ilsif 33 (Castlemed) -
. E = 4.eetM/*f for T > I" and Me = 4 for I i I*,
4 Es i e i
-'. T = As deflood la Ilote 1
' i
. I" = Indicated T , at egEggelEAft4L POWER (Cctlbratten tasperature for AI l instrismentatlen.148M*f).
5 = As defleed la liste 3. and - ,
i l
is(AI) = 0 for all AI. i
)
1 Iltif 48 he chamael' 6p 5etpalat shall met exceed its campested Trip setpelat by more than
,,x .ms .1 - e.
2.(
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i l Nove 5 *. Tise seetsost. en.coc. Fog. Ts,neesi.oroM is 1 2. d 'Folt. 92.65S0E8 1510 1
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PRDOF & REMEW CDPY
( OESIGN F.ATURES .
5. 3 4EACTCR CORE l FUEL AS3sM61.35 ' . . .
j IS9 'l-
l 5.3.1 The ce 11 ntain 193 fuel assemolies with each fuel assemely containing fue rods clad with I!rcatoy-4. Eaca fuel rod shall have a -
l*
i nominal . ive fuel *tength of 144 inches and contain a saximum total weigns x.
of uranius. The initial core loading shall have a maximum enrichment of 3.10 weignt portant U-235. Reload fuel snell be similar in anysical design '
to the initial care loaaing and snall have a seafaus enrichment of 3.50 weignf. ~
percent U-235. . ,
.C.I, .
7 -
CONF 401. 400 A53988L , , - -
ne l
5.3.2 The ce ; .@ y I red ;M8' & ,p
, shall contaig 53 # 1-!a w . = c 5"t-; _. control rod assemolies snail contai,n a nominal
' ~ 4? : f assemolies. .
,,ci?+$ 2; i
3.42 incnes of sorser antarial. All control roca shall be ha , clad wita stainless steel .
( ,," -]
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5* $ R A WW ?h _
-4,
. - : .y} }9 y y y'
' "Y , <.fV J 'i ccIGN nt35 tar 20'itwERATURE ,
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5.4.1 The Reactor Coolant system is designed and snail be asintained: .,. j.g' {a, :.s.y l,4') w n. ..e., e ,n; g m u.6
a. In accordance with the cada requirsments scocified in Section 5.2 of i
the FSAA. wita allowanca for normal degradation pursuant ts the ".' $ ?!i9j j
applicaele Surveillanca Requirements, ; . .. "
';ry, 7 Ty i
.. a h-
3. Per pressure;.of 2485 psig, and
c. For a tamperature of 650*1r, escapt for the pressuri::sr wnfcn is . ,., . g. (..h .[y'4[ ( g saaer. m. . . s, m, t d. a.
.o
' ... '., .QQ - Q: :, og:%jt.. 3
$QK , .
. ~^ y.ra.- -
q
5. 4.2 This tant water and staas values of the Reactor Coolant Systas is L
.2 12,257 cusi.: feet at a nominal Tavg of 6SPr4*F. "l~
5BB.Y -
-9 1
, ' "r# .
5.5 wtTICRCt.CGICAt. TCWER !.CCATICM
5. 5.1 The seteorciogical tawer shall be located as snown on Ftgure 5.1-1.
3YRCM - UMI 1 5-4 7
- ~. ...%
- - - - - - - ,+w,w., w-,,,--- -.y .ew m- _w.w.,--,,.--.-.-.-,-,.v.n-, ..r,-,..--..-- ,.,--,.r-
I l
ATTACHMENT M (Radiation Monitoring Instrumentation)
1) Tabic 3.3-6 (pg 3/4 3-39)
Change the alarm / trip setpoint for the Fuel Building - Fuel Handling unit from "12 mR/h" to "12.5mR/h".
2) Table 3.3-6 (pg 3/4 3-39)
Change the alarm / trip setpoint for the Containment - Fuel Handling unit from 12mR/h" to "$100mR/h".
3) Table 3.3-6 (pg 3/4 3-39)
Change the alarm / trip setpoint for the Main Control Room - Outside Air Intake from "12mR/h" to " particulate 1.0 E-9pci/cc '
gas 9.5 E-7pci/cc iodine 1.12 E-9pci/ce".
The above setpoints have been calculated according to ANSI /ANS-8.3 - 1979, Appendix B.
-w _ _. p,, i $ ---- -.,,yyy .---t---- F-+T---rm - - - - -
.e- - - *ev-Wh ^--e r-----*y-------e t-7 T- PM N D-=----te--*--- F*?
- n.A;-...o--1...c- - m. . . . . - . . . - -
.I TA8tr 3.3-6 4
j l RADIATION M3NITORIIIG INSTRtN NTATION FOR PLANT OPERATIONS i E MINIIRM 4
CHAIRIELS CHAleIELS APPLICARLE ALAM/ TRIP i e FUNCTI0lW. UNIT TO TRIP / ALARM JPERARLE N00ES SETP0fMT ACTIC11
-1. Fuel Rullding- '
Fuel Mendline 4*G 2 ** 30 (GRE-A2055/56) 1 1/mR/h -
2. Containment-Fuel 10 0
' , ' Handling (IRE-AR011/12) 1 2 # -
1/all/h 26
3. Criticality-Radiation Level .
2
28
, g (0RE-AA037/38) 1 1 15 mR/h 4 * ,
l 4. Gaseous Radleactivity-
.. RCS Leakage Detection -
(IAE-PRollA) N.A. I 1, 2, 3, 4 N.A. 29
. 5. Particulate Radioactivity-RCS Leakage Detection -g
! (IRE-PRO 113) N.A. I 1, 2, 3, 4 N.A. 29 g,j 8
6. Main Control Rece-Outside M Air Intake (0RE-PR031/32 gp and ORE-PR033/34) 1 2 per All 1 f-elVir 27 ,
intake t ;'*
S pqofic4l41c. l.oE-1.a4[CC.
34s 1 SE-7AQ/cc. Q .
odinc. f.

a E-b44/cc g, M
e mm -m-o' -m m
- +- +
m - - _ _ - - - _ _ _ _ _ - - - _ _ - _ _
ATTACHMENT N (Setpoint Changes)
1) Table 2.2-1 (pg 2-6)
The Z value for Underfrequency - Reactor Coolant Pumps should be "13.3" not "0".
This change is to be conistent with the Setpoint Methodology.
2) Table 3.3-3 (pg 3/4 3-15)
Item 3c should read
" containment Vent Isolation
1) High Radiation See Table 3.3-6 Item 2 for all High Radiation initiating functions and requirements.
2) Automatic Actuation 2 1 2 1.2.3.4 17 1.ogic and Actuation Relays
3) Manual Phase "A" Isolation See Item 3.a.1 for all manual Phase "A" Isolation initiating functions and requirements.
4) Manual Phase "B" Isolation See Item 3.b.1 for all manual Phase "B" Isolation initiating functions and requirements.
5) Safety Injection
See Item 1 aLove tot all Safety Injection initiating function and requiroments."
The above change is to reflect the signals that cause a containment Vent Isolation at Byron.
3) Table 3.3-4 (pg 3/4 3-22)
Change item Ic Total allowance from 2.5 to 3.0, change item Id Total allowance from 13.0 to 16.1.
Change item Id Z from 10.71 to 14.41.
Change item le Total allowance from 14.2 to 21.2.
Change item le 2 from 10.71 to 14.81.
Change item 2c Total allowance from 5.0 to 8.0.
The above changes are made to be consistent with the Setpoint Methodology.
. . - , - . . . - . . , , , , , , . ~ . . .
L ..
ATTACHMENT N (Coatinued)
(Setpoint Changes)
4) Table 3.3-4 (pg 3/4 3-23)
Ites 3c should read
" Containment Vent Isolation
1) High Radiation See Table 3.3-6 Item 2 for the High Radiation trip setpoint
2) Automatic Actuation N.A. N.A. N.A. N.A. N.A.
Logic and Actuation Relay
3) Manual Phase A See Item 3.a.1 above for all manual Isolation Phase "A" Isolation Trip Setpoints and Allowable Values.
4) Manual Phase B See Item 3.b.1 above for all manual Isolation Phase "B" Isolation Trip Setpoints and Allowable values.
5) Safety Injection See Item 1 above for all Safety Injection Trip Setpoints and Allowable Values". ,
The above change is to reflect the signals that cause a containment Isolation.
5) Table 9.3-4 (pg 3-24)
Change item 4.c Total Allowance from 5.0 to 7.7.
Change item 4.d Total Allowance from 14.2 to 21.2.
Change item 4.d Z from 10.71 to 14.81.
Change itse 4.e Allowable Value from 5-110.0 psi to 1-111.5 psi.
Change item 5.b Total Allowance Ecom 5.0 to 6.0.
Change itse 5.b Z from 2J8 to 4.28.
These changes are necessary to concur with the updated version of the Westinghouse Setpoint Study.
ATTACHMENT N (Continued)
(Setpoint Changes)
6) Table 9.3-4 (pg 3-25)
Change itee 6.c Total Allowance from 30.0 to 27.1.
Change item 6.c 2 from 27.18 to 18.28.
Change item 6.c Alloweble value from 36 to 39.1.
These changes ace necessary to concue with the updated version of the Westinghouse Setpoint Study.
7) Table 4.3-2 (ph 3-34)
Change ites 3.c containment Vent Isolation to read as follows:
"1) High Radiation See Table 4.3-3 Item 2 for all High Radiation Surveillance Regirements.
2) Automatic Actuation N.A. 'N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 1.ogic and Acutation Relays
3) Nanual Phase "A" See Item 3.a.1 above for all manual Isolation Phase "A" Isolation Sueveillance Requirements.
4) Manual Phase "B" See Item 3.b.1 above for all manual Isolation Phase "B" Isolation Surveillance Requirements.
5) Safety Injection See Ites 1 above for all Safety Injection Surveillance Requirements."
The above change is to reflect the signals that cause a containment Vent Isolation at Byton.
1 ,
J f
i IA8tE 2._2-1 (Continued) 5 2
REAC10R TRIP SYSIEM IN51RUMENIAll0N 1 RIP SEIPOINIS e
E . 5fMSOR
' G '
10TAL ERROR Att0WABLf VAllE e fuNCIIONAt UNII Att0WANCE (IA) Z _(R lHIP SEIPSINI
12. Reactor Coolant flow-Low 2.5 1.77 0.6 >90% of lorp >89.2% of loop-ilesign fluw" / esign i =^
! T1.t%
13. Ste.es Generator Water 27.1 M 1.h 140.8% of sierrow range insthiment - .
> 34 4 o f iTow a! Hind struisent g
level Low-Low s jpe span ,
/ t.t o
14. Undervoltage - Reactor 3.3 0 0 ( *000 v
- s- >47f,8 vults -
Coolasit Psamps ,
EAEEli i
0 >
15. Undertsequency - Reactor 14.4 13 * .lara. _ 'jft y Coolaint Pusaps , A i a.
16. Iurliine Trip l Nd 0 f , , ,- y 7t;p Ho< der .5* *f D ;
Mpsig
~
m. tuwd, latc#H Pressure N.A. N.A N.A. >)i(psig ,
i i h. Iurtaine 6to H.A. N.A. N.A. 11% open 11% open flosuie & p.44 Valve -o
c3
^
17. Sa.lety Injection input N.A. N.A. N.A. N,A. H.A. c3 1 1 rom ESF
- po HencLor Coolant Pamap N.A. N.A. N.A. N.A. N.A. -p-
18. l j drealtes Posillon.1 rip q
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Ea q q, s" j Akoop d d Ig~4k IloW 3 % 700 gpm_ h 2d i
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, E ]!
. t* 8' *G i I l ' '
I i! : ,.- :'. I, i -
>l
~
e 1 l
. iAnar i.3-4 4
g (NGINifRf5 1VJ!!_!1Aluttf5 AClua[jeN SYSJIN INSINtW NIATION INIP StlPGINis ,
' AtteWastt l l
' rorAt SENson iNir SflPolNr VAlu[ f E !!M!!aNAlt#8H ," All0WANCEj!g) { (NuoM(9 '
4 -
i. 5.t.tv Inj.ction .
(ae.ctor Trip, fe. *4ter j ,
isolation. Start al.sen l
.- faner48 ors. Coatalnment l l Cooling lens, asul fasantial l Service W4ter) .
N.A. N.A. N.A. N.A. N.A.
a. M4 nuel Initiation f
' i Autamatic Actuallap N.A. N.A.
N.A. N.A. N.A.
b. !
logic and Actuatinal Nulsys q
e 30 6I
c. Containment Pressure- 5-44 psig illutet M a.74 1.5 1 5.0 pelg g ' I l'i .'ll 3:21 1933
d. Fransurizar Pressure- l(a.1 1 le n paig low .13M 14.+t- 1.5 1 M64 psig
~ I
e. Stoaa Line Pressura- 21. 1 . f*4.81 (l's
.)Are .l L M 1. 5 1 640 psig 1 610 pslu* c3
. Iow (above P-II) s1l fl*
2. Containment Spray '
n3
4. Hasiuol Initlettest N.A. N.A N.A. N,A. N.A.
] 8
b. Automatic Actuallest h
toulc and Actuatloss N.A.
NuIays N.A. N.A. H.A. H. A.* . E,$
c. Contelument Pres 6ure- 8.0 2l Illuir3 ett o.71 1.s 5 20.e palu 5 tr.o pstu ,
I S ,
.s 8
.. .., , , , , , - ~ . . , _ . . . . . ,, .. _. . . . . ,
o . o' 'e o * * *
D - .
, JASIE 3.3-4 (roatinued} g '
4 ;
[NGINitStB 1Affly f{glumES ACluAlgo!L5ul!!Ll!LSInt**#faunM inir Stiro Nis
[ .'
I
51N50R 1 RIP At(0WAstt 10lAL . VAtuf 8 E AllDWANCE-(IA} } [Ril0N(5} SflPolM[
(t!!!C[liMiAl INil[
i ,
.". 3. Contalsment Isolatloa these "A" Isolation 4.
N.A. M.A. N.A. N.A.
1) Hanual Inlilallon N.A.
2) A. stomatic Actuation N.A. N.A. N.A.
Lo$.IcandActuation N.A.
j aa ys N.A.
3) Safely injection See Itas 1. Mwe for el Safen Na@a kh h@W M '
Allowable Values.
b. l'hase "A" 16alatlaa N.A. N.A. N.A.
l) Hanual.InltIatton N.A. N.A.
[ N.A.
t a H.A. N.A. N.A.
l l'1 2) Automatic Actuation N.A.
logic and Actuallon kalays
3) Contalement 00 1.5 1 20.0 psig 1-t2.0 psig
.l Pressure-Ill0h-3 M 0.}t
c. Contalsment Vent i#'h 4 r' f %
i liolation
See Table 3 3-G Dem1 #*' + c Nh
~
u W P**P N.A. C3 in Y6Mh b...if.N.
_ . . . . . . . ... u
- . .a- . .
m no
2) Aa.tomatic Actasallon m
' logic and Actuation N.A. N A. **J Relays H.A. N.A. M.A. h am.nI m.na*\
See Itum 3.a.1 above for all Phase "A" Isolation Iry Satpoints Q
3)^ Pliana isolation" A" and Allowhle Values. m,n a 13
i) man,,a ?haar. "Y Sec Item 3.b.t ubeA for all,~ Phon. "B" .rsolehm Tr.e SM'$b M
r w l li m and Allow"l)le Yalut'f-o 5) Safelytnieem See I % .1. h far all Mcy TaFM D N'
and Alloet.a Va.luea *
(
1
e, . . , .
o o o o = , o, -
) - .
, ' i !
t i
' I IAasf 3,3-4 (familauad!
!!!!ilNt(885 sat ((L((A]{lgLf((uA110N SulffLifL51punINialiass inir stIrgj!lIs '
8
' lolAt s(N504 18le AtinWAllt( ,
tenon (s} stiro NI vAlur .
9 tialEljoNAt unit Atlauwt (van I 4 ,
.. 4 54. Lin isolatie, N.A. N.A. N.A. N.A.
s. Nahual laillation N.A.
h. Autusalic Actuation .
i tculc 41ut $ctuetlen N.A. N.A. N.A. ;
Delays M.A. N.A.
9.'I 11 M
c. Contaisissat Pressure- 5 ned pel I 5-at-e psis litair2 -t-e e.fl 1.6 we ss.1 is.gi j
.l . st. vrETJirer - . w e- e i.6 1 se .l. .i 1 sie p /c >
e* ,
, low [(uc e-ii) dE-lit.5 fui Wa n'. a. Steam.llaafratsure- a.6 a 5 -ion
univ. an iPNI a.e
Pal / q
(.6 h e-id
6. Iurbine Irly asus f eessator itulation
a. Automatic f.cLuation ]
ca I
80010 8861 Actuatloa-Relay 6 N.A. N.A. N.A. N.A. N.A. 4] l
'd 2d St.'l 32 3 A*
stum dianurator Water (o.O
b. 1.6 5 erm er 5 aM or m 4 evel-lileirHigli (P-14) .ke- .>.-tts .
I . narrow ras I earrow renu q
.. y..,tru..L 3, Qi.6ir ,, ...
s 3!
o
>. " 9
i O O O. J ., )
) .-
6-I MIf 4.3-2 (C*!tliqued) 4 E MIsitfafs 5Affly giglggf IMIAlls
-l .
E 1Rif ACIDAlledG N00f5 ANAt04 U
CelAltift DEVICE HAilfR SIAVE fot nellCM
e4 AtluAllGil Rfl AV StiAV SINIVfIllANCE ChAllMEL CHANNft. SPIRAllGNAL APfRAIIONAL !s arot_ilm.[g riittg (3tiggyle Igg Itsr _ ggi[E_jtg Jtg ]LH
' f!!!!EllanAs uisit l 3. C tal .t s .l.tl 1
a. Phase "A" Iselatten A- N.A. M.A 1, 2, 3, 4 1
n) Hanuel Initiation N.A. N.A. N.A. 41. A.
, g M(I) H(1) 4 1, 2, 3, 4 M.A. N.A. N.A. M.A.
l 2) Automatic Actuatlea i Oc and Actuallaa -
l , y e. .,. I
'T 3) Safety Injecitoa Sea Iten 1. above for all Safety Injectlan surveillance 8. qui,emaats. .
b. these "S" Iseletion f
! *i ,
1 R N.A. M.A. N.A 1, 2, 3, 4 I) Heaual Initiatlan N.A. M.A. N.A.
N.A. N.A. 83 H(1) M(L) 4 1, 2, 3, 4
.' 2) Automatic Actuallon M.A. N.A.
logic Actuation A*
l l- telay.
== '
N.A. N.A. N.A. N.A. I, 2, 3 I i, 3) containment s a H Q-
' ! Pressure-Niuh-3 E ib
, c. Cantainment Vent Isoletten 7g g 3 h t .for al{ R4.b Surveille(e RetoMemenia.
) ,
gi, II N. ,..b. i.!.'i.'N.,E. ,' ..
. . . .... ,,.... .- y .
.. .o..
.,.,27 4 N.A. N.A. N.A. N.A. H(1) H(1) 4 1, 2, 3, 4
! 2) Automatic Actuation laulc and Actuation M&
(' A-
, , i,...nt .
5,,,vaiii.. . .a.
sa. ii.a 3. .. ai...
,.,. , , .ii n,as. A i ..iat . .
@i..Li.. 4 m_a
' 5ee.'. Chew 5.b.t above for all Phcise"[b"Isolc.hn a surveillance Rego:<cments
{ -
't)MonvalPh*iG or3sketu w.m -
In%li.% . M* c TMw 1 stbNe ih- (41) 0"fEIO IhjMW 50rveillowce ho <c-menth
. . w w w w o w w . w
, b -
3, .
i I
i l l .
l .
IAuls 3.3-4 (Continued) ,
ej INGINithfu SA&&ly [f41Hets Ag uAjjoM Syhi{}t J 5JN15HNjAl10N INIP 5tlPGIN15 1 ,
-18lAt *
$1N50R IRIP Att 0 Wall! L !
8 E IL*0lL*Ll!!!! AllGWANfL Q A} l INRGA Q) StirelN! yAtui .
1 l
e. 6. Auxillary feetkater
4. N nual Initletlaa N.A. N.A. N.A. N.A. N.A. !
l
h. Automatic Actuation logic aiul Actuellen -
Nolays N.A. N.A. N.A. N.A. N.A. .
c. Ste8m Generator W ter gg.)
c level-tow-law- Sc M steri h tor- 29.l 10. W 'M a.t g Driven Pump and .36 2f:mI* 1.5 1 4t% of 1 enafr of e- Diesel-Driven Pump marrow ranck narrow range i y r' astrueg instrument i 5/" 6 pen g c ')y.d p.g .9%v *As
d. Undervolt8uu-ACP Buk-Stest b ier Driven Pump N.A. N.A. N.A. / vin vats)} ubesitage 7aN-SCP a-voltage)
-6W-#6:P j
j and Diesel-Driven Pump (4See-vp) -44674-ve V
a. Safety injection- t/$]9 iby3 3 Start W tor-l firiven Pump and SeeItemt.shoveforallSafetyinjection{rlyintpointsand @
Diesul-Driven Pump Allowable Value6. n no
=_a C
5 s
C].
, s.o
a. ,
.

ML20090G624

Text

Subject:

Dear Mr. Denton:

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