Text

Forwards Addl Comments & Suggestions Re marked-up Proof & Review Version of Tech Specs Distributed by 831216 Memo
	ML20092L150
Person / Time
Site:	Byron
Issue date:	06/20/1984
From:	Tramm T; COMMONWEALTH EDISON CO.
To:	Harold Denton; Office of Nuclear Reactor Regulation
References
	8838N, NUDOCS 8406290259
	Download: ML20092L150 (36)
	v • d • e

T

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f -

'. Commonwealth Edison

) One First NitionIl Plum. Chicigo. Illinois j

i ~J Addr;ss R: ply to: Post Offico Box 767

\ ,/ Chicago, Illinois 60690 June 20, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

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

Thomas.

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

(c): April 2, 1984 letter from T. R. Tramm to H. 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.

Dear Mr. Denton:

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

Attachments A through J 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 (e). We understand that the NRC will review each of these proposed changes and inform Commonwealth Edison of their acceptability.

8406290259 840620 k PDR ADOCK 05000454 A PDR

'1h lii: - -- -_-

I H. R. Denton June 20, 1984 Please direct any questions you may have regarding this matter

~to this office. .

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

Very truly yours,

'ki ASb*~

T. R. Tramm Nuclear Licensing Administrator 1m cc: Byron Resident Inspector 8838N ku

ATTACHMENT A (Bases Section 2.0) i Circled items noted in this attachment have been previously submitted.

~

t Section 2.1.1 (pg. B2-1) Reactor Core 1)

I Delete "and a reference cosine with a peak of 1.55 for axial power shape." !

Delete the previous change which changed 1.55 to 1.49. The number is now !

1.55.

This change is based on WCAP 10315 " Nuclear Design and Core Physics ,

Characteristics of the Byron Unit 1 Nuclear Power Plant Cycle 1". [

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PR00f & REY!EW C0py 2.1 SAFETY LIMITS I r

BASES t

2.1.1 REACTOR CORE l l i The restrictions of this Safety Limit prevent overheating of the fuel and i possible cladding perforetion which would result in the release of fission !

, products to the reactor coolant. Overheating of the fuel cladding is prevented ;

l by restricting fuel operation to within the nucleate boiling regime wnere the .

heat transfer coefficient is large and the cladding surface tesoeratu n is (

slightly above the coolant saturation temperature. ;

Operation above the upper boundary of the nucleate boiling regiee could .

result in excessive cladding temperatures because of the onset of ceparture !

Jros nucleate boiling (DNS) and the resultant sharp reduction in heat transfer

coefficient. DN8 is not a directly seasurable parameter during operation and [

therefore THERMAL PCWER and Reactor Coolant Temperature and Pressure have been i related to ONS through the WR8-1 correlation. The WRS-1 DNS correlation has l been developed to predict the ON8 flux and the location of DNS for axially [

uniform and nonunifom heat flux distributions. The local DNS heat flux ratio t (DNBR) is defined as the ratio of the heat flux that would cause DN8 at a t

! particular core location to the local heat flux, and is indicative of the l l

margin to DNS. -

e, min val of the du ng A ta coeration, orma) er al , tien N add a ic N8R[pa tra nts Klia(tedt f( for/ a " A" '

I i ty)/ cal call a%1.32 fer a, thimble.celf. fThis/v'alue corresponds \to a 95%g pretsanilitv at a ,55% confidence hvel that DNKwithnot4ccuRc and de ghosen as ~pM at mirgin ,g f#r .fl S.c.. inditTuns S (The curve;T of igh 2.1-1 --: ' ' ' sr.o@ temperature loci of points of THERMAL i for which the l

l POW , eactor er ant - e---="= =* h ve l sini-__ 0 - . r;o less than 1.34 for a typical call and L32 for a thinole !

cell, or the average enthalpy at the vessel exit is equal to the enthalpy of V

I saturated liquid. /. .Sf-Thsse curves are t ased on an enthalpy hot channel factor, F , of

-- ' ; c;fe. ;.;;; ;;;i . sitt e i;;;% -f L M %- =? 4 =1 ~~ " x : . An allowance j N

is included for an increase in Fg at reduced power based on the excression: l 1.S8 '

j 1+0.3(1-P)) x Fh .

l Where P .is the fraction of RATED THERMAL PCWER. ;

These limiting heat flux conditions are higher than those calculated for l the range of all control rods fully withdrawn to the saximum allowanle control ,

rod insertion assuming the axial power imoalance is within the limits of the l f g (AI) function of the Overtemperature trip.

When the axial power j l

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i BYRON - UNIT 1 82-1 l l

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ATTACHMENT B (Section 3/4.2) l f

Circled items noted in this attachment have been previously submitted.

Section 3/4.2.1 (pg. 3/4 2-1) Axial Flux Difference.

1)

In Sections 3.2.1.a and 3.2.1.b, delete the number "3000" and replace wih "5000". Also, in Section 3.2.1.b delete the words "+3%, -12%" and I replace with "+3%, -9% for the initial cycle and +3%, -12% each subsequent cycle". I This change is required for initial PWR cores and has been recomended by Westinghouse. Subsequent cycles will revert to 3000 MWD /MTU and +3% to l

-12%.

2) Section 3/4.2.3 (pg. 3/4 2-8) RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor (pg. 3/4 2-8) !

Delete the original LCO and replace with "3.2.3 Indicated Reactor Coolant System (RCS) total flow rate and H shall be maintained as follows:

i

a. RCS Total Flowrate >_ 386,000 GPM

[. b. H < 1.55 [1.0 + 0.3 (1.0-P)]

where: r Measuredvaluesof(Hareobtainedbyusingthemovableincore ;

detectors to obtain a power distribution map., and P= THERMAL POWER RATED THERMAL. POWER j Delete "the combination of".

! Delete "and R" and replace with "and H".

Delete "shown on Figure 3.2-3".

f I

-(pg. 3/4 2-9)

Delete Figure 3.2-3 ;

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I ATTACHMENT B (Continued) '

l (Section 3/4.2)

-(pg. 3/4 2-10)

Item b delete " comparison" and insert " determination". I Deletethewords"thecombinationofR"andinsert"[H".

Item c delete "the combination of R" and insert " H".

Delete " comparison" and insert " determination".

Delete the words "within the region of". ,

t Delete the words " operation shown on Figure 3.2-3." !

Section 4.2.3.2 delete "the combination of". t Delete"R"andreplacewith"(H". f Delete the words "within the region of" and also " operation of Figure i 3.2-3." >

Section 4.2.3.3 delete the words "within the region of", " operation of Figure 3.2-3" and "the most recently obtained value of R obtained per fj Specification 4.2.3.2 is assumed to exist." Following the words "12 V hours when" insert "outside the above limits".

i I

l With the elimination of Rod Bow Penalty, all references to R can be deleted. Theproposedchangesdefine(Hdirectlywithoutusing'aratio.

i Since there is no rod bow penalty, Figure 3.2-3 (pg. 3/4 2-9) could be l simplified and described by a rectangle. The proposed change to the Lco Section 3.2.3 incorporates the requirements of Figure 3.2-3: therefore, the figure we've previously submitted can be deleted.

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3/4.2 PCW G OISTRIBUTTCH t.IMITS a

b $$ Qf{ l 3/4.2.1 AXIAL Ft.UX OIF88R OCE +3 '~

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.e. 3*la ,- t 2.7o wk Oy Y l l

LIMITING C::NCITTON FOR CPERATTCN ,

)

3. 2.1 The indicataa AX1AL F OIFFGDCE (AFO) snall be saintained witnin tne I following target hand (flux ifference units) anout the target flux cifferenca: j

a. 2 5% for core rage ace:mulated turnuo of less taan or equal to l

3000 Mbe/MTU d !

Scoo for care average accumulated turnus of greater than f

b. ~ %, :

V !

-3000 MWO/MTU. /\

Soo iata outside the move required target Sand at greater i The indicated AFO s than or equal to 5( less tnan 90% of RATED THERMAL PCWG proviced one indi- i cated AFD is wi ui Acceptacle Coeration !.f aits of Figure 3.2-1 and tne l c !

cumulative penalty deviation time does not exceed I hour curing the previous i; 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . e j

The indicated AFO say deviata outside the acove required target band at greater i taan 15% aut less tnan 50% of RATED THEW. AL PCWG proviced the c:mulative penalty deviation time does not exceed I hour curing us previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l t.

AP9tICA8ILITY: MODE 1 aoove 15 of RATED THERMAL PCWG". l l ACTICN:

a. With the indicated AFD outside of us aoove required target tand and !

with THERMAL PCWG greatar taan or equal to 9C% of RATED THER.wAL l l

PCWG, within 15 sinutes, eitner: t

1. Restore the indicated AFD to witnin the anove required .

target band limits, or ,

2. Reduce THGMAL PCWG to less than 9C% of RATED THEWAL PCWG. t

3. With the indicated AFD outsics of the aeove required target band for more than 1 nour of cumulative =enalty deviation time curing us previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside ce Accactante Operation Lf: nits of Figure 3.2-1 and wita THERMAL PCWG 1ess taan SC: :ut equal to or

[

greater tnan 50% of RATED THERMAL PCWG, recuca:

1. THGMAL PCWU ta less taan 50% of RATED THGPAL 80WER witnin .

20 sinutas, and .

I

2. The Power Range Neutron F1 Hign Set:oints to less man or g equal to 55% of RATED *HERPAL PCWER witnin ce next 4 nours.

l l "See 5 cec 1al rest Excsotion 3.1D.2. >

9 Sur reillanca tasting of tne Power Range Neutron " lux cnannel say se performeo pursuant.to Scecification 4.3.1.1 steviced ce incicated AFD is saintainec l witnin ce Acceptante Coerstion Limits of Figure 3.2-1. A total of 15 nours l

ooeration may be aulated witn the AFD outside of us aeove esquired target !

3and during tasting witacut penalty caviation.

r SYROM - UNIT 1 3/42-1 ,.

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PRDOF & REVH COPY O _____, POWER OISTRIBUTION LIMITS d

3/4.2.3 RCS FLOW RATE ANO NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR !

LIMITING CONDITION FOR OPERATION i

-0.2.3 ine coactnation of maicar.ea neector coolant systas (sud) tota 6 rw

- en e..4 ; an.ii as amin s i wii.uin v.n= mglon of allowenie operar.ivn I

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l APPLICA8ILITY: MODE L . - . - -

ACTION:

M or F3a -

With '.r :: Tin;;i;; ;$ RCS total flow rate out. side the region of j acceptable operation cr ;; '*!yvii 3. 2 g ,

AO

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

L Restore 2: -h:tha ." RCS total. flow rate to within the acove limits, or ,

1 I

2. Reduca THERMAL POWER to less than 50% of 4ATED THE?NAL PCWER and reduce the Power Range Neutron Flux-Hign Trip Setcoint to l less than or equal to 55% of RATED THE.4 MAL POWER within tae next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4 3/4 2-8 SYRON - UNIT 1 1

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PROCF & RMY CCPY PCWER OISTRIBUTION LIMITS LIMITING CONDITION FCR OPERATION ACTION (Continued) e

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limits, voyify through incore flux sacping and RCS total flow rat's -- ':: '.nat

' Pand RCS total flow rata are restored to N /within the above limits, or reduce THERMAL PCWER to less taan 5% of p44 RATED THERMAL PChER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; and N c.

( Fu Identify and correct the cause of the out-of-liaNt condition prior to increasing THERMAL PChER acove the reduced i'r ERMAL PChER limit required by ACTION a.2. and/or b. above; sucsegg ent PChER OPERATION

f , and indicated RCS say proceed provided that e: - ncti::

total flow rate are demonstrated, through incere flux samoing anc RCS total ' flow rata : :ri e

be ui W a =e r;;isa f accaotaole

. . .i:: On.n ;; "';.. . 2.2 . prior to exceecing the following

^

~

THERMAL PChER levels: Q

1. A nominal 50% of .uTED THERMAL PCWER, i 2. A nominal 75% of RATED THERMAL PChER, and

3. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or ecual to 95% of l [] RATED THERMAL PC*aER.

V - . .

SURVEILLANCE 9EOUIREMENTS 4.2.3.1 The provisions of Soecification 4.0.4 are not acolicaole. N 4.2.3.2 0.; ;: rin;ti;a ehdicated RCS total flow rata and [n 1 be detarsined to be -ie'- C; r;;i;; ;fMccaotaoleW,... .;.. ;n ;f %r; 2.2- V l a. Prior to operation anove 7E% of RATED THERFAL PChER aftar eacn fusi loading, and l

i

(.,,

b. At icut ones per 31 Effective .:ull Power Days.

4.2.3.3 The indicated RCS total flow rata at shaK leastbe verified once ;er 12 to es wiW- On.surs

-1:: M ccaptacle :::- ti:n :f "';.r; 2.2 P :::: r;;d. :, =; .n- ..: e ;f 1, ; a ....w aur hi 'i ;;.1.i 1. 2. 2. 2, ' :t Mc>VE j

....-...4oo_

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4. 2. 3. 4 The RCS total flow rata indicators sna11 be suefectad to a 0.9ANNEL CALIBRATICN at least onca per 13 tontas. .

4.2.3.5 The RCS total flow rata sna11 :s detar sinec by precision neat :41anca i seasurement at least onca per 13 months.

l0 -

BYRCN - UNIT 1 3/4 2-10 l

l ATTACHMENT C (Section 3/4.4)

/N Circled items in this attachment have been previously submitted. ,

1) Section 3/4.4.2.1 and 3/4.4.2.2 (pg. 3/4 4-7,8) Safety Valves and Shutdown The starred (*) item of Technical Specification 3.4.2.1 and 3.4.2.2 needs I to be deleted based upon the FSAR response to NRC Q212.116 which is '

attached and describes the approved ASME Section XI method of testing. :

The station will bench test the valves using nitrogen at the ambient temperature. This practice is consistent with the majority of the Nuclear power Industry. Bench testing of relief valves in this manner is 3 consistent with ASME Section XI requirements.

2) Section 3/4.4.10 (pg. 3/4 4-38) Structural Integrity ;

Delete the words "each reactor coolant pump flywheel.... Revision 1, August 1975." Also, insert the following: i "4.4.10.b I

1. In-place ultrasonic volumetric examination of the areas of higher stress concentration at the bore and key ways will be performed each '

40 month period during refueling or maintenance shutdowns coinciding '

with the inservice inspection schedule as required by Section XI of the ASME Code.

2. Visual examination of all exposed surfaces will be performed each 40 !

month period and a surface examination of the bore and keyway surfaces will be performed whenever the flywheels are removed for ;

maintenance purposes, but not more frequently than once each 10 year !

interval."

l The requirements for examination procedures and acceptance criteria as described in the Regulatory Guide 1.14, Revision 1, August 1975 will be followed. .

The proposed change is consistent with our response to Question 251.7 of the FSAR.

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's REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES ,

O g OPERATING !

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LIMITING CONDITION FOR OPERATION O - -

3.4.2.1 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig 2 1%.X APPLICABILITY: MODES 1, 2, and 3. l O -- - -

ACTION: . - .

With one pressurizer Code safety valve inoperable, either restore the l inoperable valve to OPERABLE status within 15 minutes or be in at least HOT

STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following O 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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Ow - SURVEILLANCE REOUIREMENTS 4.4.2.1 No additional requirements other than those required by Specification 4.0.5. '

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cO, BYRON - UNIT 1 3/4 4-7 P

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l i REACTOR COOLANT SYSTEM ;

, ,Jq & Ragg ggpy ,i em . 2 i

. LIMITING CONDITICN FOR OPERATION l I

l 3.4.2.2 A sinimum of one pressurizer Code safety valve shall be CPE.tA8LE wita a lift setting of 2485 psig 2 IL* 5L.

AP9UCA8ILITf: MOES 4 and 5.

M* ,

t With no pressurizer Code safety valve CPERA8LE, immediately susnend all ;

operations involving positive reactivity c. and place an CPERA8LE RNR !

loop into operation in the shutdown coolin MGC. -

pole l

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SURVEILLANCE RECUIXEMENT3 !

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4.4.2.2 No additional requirements other than those required by j

Scecification 4.0.5. !

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SYRCN - UNIT 1 3/44-6 1 c i

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i REACTOR COOLANT SYSTEM P W' J & E H M .

l 3/4.4.10 STRUCTURAL INTEGRITY e

i LIMITING CONDITICM FOR OPERATICM j i

. 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components :

shall be maintained in accordance with specification 4.4.10. ,

i l

APPLICA81LITY: All MODES. ]!

ACTION: ,

l -

[

C a. With the structural integrity of any ASME Code Class 1 casoonent(s) I not conforming to the above requirements, restore the structural !

to within its limit or isolata l l

I integrityffec*.edof the affected casoonent(s) prior to casoonent(s)6..cr inc :::!M --

)

System tancerature =n '. . . ;-^ .. i .. ;ir.ir mntr; " T L

- = . :,7 10T Gaei n .L eae. " - -- M 2.c o

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! b. With the structural integrity of any ASME Code Class 2 casoonent(s) not confaming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolata the affected conoonent(s) prior to increasing the Reactor Coolant i System taseerature above 200*F. j l

c. With the structural integrity of any ASME Coda Class 3 comoonent(s) l not confoming to the above requirements, restore the structural -

integrity of the affected component (s) to within its limit or isolata the affected component (s) free servica. ;

! d. The provisions of Specification 3.0.4 are not applicable. !

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i i SURVEILLANCE RECUIREMENTS I

4.4.10.AIn._add ,,ition to the requirements of Specification 4.0.5. ::: .wa;r.s,

,am,m r "- ;

=3._t n-,..n - e: = y 2.,m

itic" C.t.b Of ? Ohta=y Sie l_ '.4, 9:ri:i; .1, yet 1075.s n.n.to.6 E d A.aute.e M m Q M 6. 5 & l .

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SYROM - UNIT 1 h 423/4 4% 4s d._._Q%gM q C'" & ,

ut m 4.p_e pm 3.ua to pLhnd. , '

- - . . . _ _ - _ _ _ . . . - . -_ s

l , , B/B-FSAR AMENDMENT 28 OCTOBER 1980

! QUESTION 212.116

(

t' "Recent operating experience has indicated that relief

(' valve setpoints may be temperature sensitive. Discuss this effect on Byron /Braidwood relief valves."

l

! RESPONSE l Changes to relief valve setpoints due to temperature variations are understood and have been considered. Temperature changes affect the spring rate of the valve spring, reducing the setpoint as the temperature increases. Normal ambient air temperature variations do not significantly affect the setpoint.

However, when a cold valve relieves hot fluid, the setpoint l can be reduced. This effect has been considered in the design of the valves and fluid systems. l r

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Q212.116-1

t i

ATTACHMENT D !

(Section 3/4.6) L i

r 1). Section 3.6.4.2 (pg. 3/4 6-24) Electric Hydrogen Recombiners.

The words " power setting" and " power" are changed to " temperature !

controller" and " setting" as there is no power setting on the recombiners :

installed at Byron, only temperature controllers.

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-' CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RECOMBINERS Pi!00:& IEYiEW COPY

-O j LIMITING CONDITION FOR OPERATION _

3.6.4.2 Two independent Hydrogen Recombiner Systems shall be OPERABLE.

APPLICABILITY: MODES 1 and 2 _ . . . . _

ACTION:

! With one Hydrogen Recombiner System inoperable, restore the inoperable system

. to OPERABLE status within 30 days or be in at least HOT STANDBY within the O next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

4

'i SURVEILLANCE REQUIREMENTS .

!O 4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE:

a. At least once per 6 months by verifying, during a Recombiner System

. functional test that the minimum heater sheath temperature increases to greater than or equal to 1200*F within 90 minutes. L'pon reaching ,

b tt S; to maximum powee for 2 minutes and '

5 1200*F, verify that increase themete the power r f::eads r greater than of equal to 38 kW, and <

$' -Irekf*'rAbe "+16 seH Y

b. At least once per 18 months by: l i

ll 1) Performing a CHANNEL CALIBRATION of all'recombiner instrumen-j tation and control circuits, ;

O 2) Verifying through a visual examination that there is no evidence l J

4 of abnormal conditions within the recombiners enclosure (i.e.,

it loose wiring or structural connections, deposits of foreign I; materials,etc.),and .

A 3) Verifying the integrity of all heater electrical circuits by

}O perfoming a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than or equd} to 10,000 ohms.

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[1 BYRON - UNIT 1 3/4 6-24

1 IIIa .

r ATTACHMENT E ,

(Section 3/4.7) i P

Circled items noted in this attachment have been previously submitted.

1) Section 3/4.7.10.l(a) (pg. 3/4 7-28) Fire Suppression Water System.

-Delete the words "each with a capacity of 2500 gpm". !

t Section 3/4.7.10.l(b) !

-Delete the words " Deluge or". I Section 4.7.10.1.1 F(1) (pg. 3/4 7-29) j Delete " pump develops at least 2500 gpm at a system head of 347 feet +

(150 psig)" i Insert " fire suppression discharge performance with a capacity of l 3750110% gpm at 98110% psig. a The proposed change incorporates results of an NRC audit requesting l Technical Specifciations to be updated to reflect NFPA requirements. The j 150% rated capacity is mentioned in NFPA-20-Centrifugal Fire Pumps (See attachment). I r

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O PR00F & EiG E 3/4.7.10 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSICN WATO SYSTEM LIMITING CONDITION FOR OPG ATION lC 3.7.10.1 The Fire Sgpressioer Watar System shall be OPGA8LE wita. ,

r.e rire suppre.aio. , - + - w; e r- _ Y~ta

, wi !

their discharge aligned to the fire suppression header, and .

.Q b. An CPGASLE flow path capable of taking suction from the flime ana ,

transferring the water through distribution piping with OPGA8LE !

sectionalizing control or isolation valves to the yard hydrsnt cure valves, the last velve ahead of the water flow alars device on enca x sprinkler or hose stando - , -" the last valve ahead of the deluge l required to be OPSA8LE per C'.,

- valve on each ". 6 er Sy Specifications 3.7.10.2 and 7.10.

APPLICA8ILITY: At all times. IM f ACTION:

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a. With one pa p and/or one wetar supply inoperable, restore the inoperable equipment to CPGA8LE status within 7 days or provida an alternata beckup pump or supply. The provisions of Specifica-i

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tions 3.0.3 and 3.0.4 are not applicable.

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b. With the Fire Suppression Water Syftan 4the: vise inoperable establish a backup Fire Suppression Water Systes.witain 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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. L SURVEILLANCE REQUIREMENTS i

Q 4.7.10.1. 1 The Fire Suppression Water System shall be demonstrated OPGA8LE:

a. At least once per 7 days by verifying the contained water supply l volume,

b. At least onca per 31 days on a STAGGGED TEST SASI$ by starting the electric estar-driven pues and operating it for at least 15 sinutes Q on recirculation flow, [

c. At least once per 31 days by verifying that each valve (sanual, power-operated, or automatic) in the flow patn is in its correct position, 1 C, J SYROM - UNIT 1 3/4 7-28

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i O Pt. ANT SYSTEMS omma con l l

k SURVEILLANCE REQUIREMENTS (Continued) l At least once per 6 aanths by performance of a systas ring header !

d.

l flush, t

e. At least once per 12 months by cycling each testable valve in the ,

c- flow path through at least one complete cycle of full travel, ;

f. At least once per 18 months by performing a system functional test l which twiudas simulated automatic actuation of the systas throughout ;

i its operating sequence, and: . !

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C 1) Verifyi"E,tha,t eac,h ; m s v hT5[di (' ( E((O 5fE'/*Ip I O.h 95 tlO 7 p d g. ' "'"-- "

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2) Cycling each valve in the flow path that is not tastable during ,

plant operation through at least one complete cycle of full l travel, and C.

3) Verifying that each fire suppression pump starts (sequentially) i to maintain the fire suppression water systes pressure greater [

than or equal to 125 psig.

g. At least once per 3 years by performing a flow test of the system in accordance with Chantar 8, Section 16 of the Fire Protection Hanceoon, >

( 4 _ . _ _ . lith Edition, published by the National Fire Protection Association. .

4.7.10.1.2 The fire pump diesel engine shall be demonstrated OPERA 8LE: f

a. At least once per 31 days by verifying: l l

C 1) The fuel storage tank contains at least 325 gallons of fuel, and l i

2) The diesel starts from ambient conditions and operatas for at !

least 30 minutas on recirculation flow.

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b. At least once per 92 days by verifying that a sample of diesel fuel from the fuel oil day tank, ontained in accordance with ASTM-0270-1975, l is within the acceptable limits specified in Table 1 of ASTM 0975-1977 i when checked for viscosity, watar, and sediment; and l

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

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service.

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O SYRON - UNIT 1 3/4 7-29 l t

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6 l I j l 20 86 (INTRIFt GAL FIRE PUMPS i Buried iron' A 1. 4 Water sources containing salt or other materials '

against corrd deleterious to the fire protection systems should be asoided. equisalent staj i

A-2-3.1 A centrifugal fire pump should be selected in the rance nt l oneration trom 90 percent to 150 percent nf its rated capacity. The i performame of the pump when apphed at capacities oser 140 per. l cent of rated capacity may be adsersely affected bs the suction condi- )

[ tions Application of the pump at capacities less than 90 percent of the rated capacity is not recommended l The selection and application of the fire pump should not be con-fused with pump operating conditions. With proper suction condi.

d l 5ew.

tions. the pump can operate at any point on its characteristic curse from shutoff to 150 percent of its rated capacity.

A 2-7 Some locations or installations may not require a pump house. When a pump room or pump house is reqaired. it should be ,

of ample sire and located to permit short and properly arranged pip- !

ing T he sut tion piping should receise first consideration. The pump !

i house should preferably be a detached buildmg of noncombustible

( construction. A one. story pump room with a combustible roof. either detached or well cut off from an adjoining one. story building. is ac-I 3 ceptable if sprinklered. When a detached building is not feasible. the ,

pump room should be so located and constructed as to protect the pump umt and controls from falhng floors or machinery. and frcm ,

fire that might drise awas the pump operator or damage the pump unit or controls. Access to the pump room should be prosided from outside the building. Where the use of brick or reinforced concrete is

not feasible. metallath and plaster is recommended for the construc- ,!

j tion of the pump room. The pump room or pump house should not i F

Nmt l be used for storage purposes. Vertical shaft turbine. type pumps may require a remosable panel in the pump house roof to permit the pump to be remosed for inspection or repair.

gC Jockey P.

A 2-7.1 Impairment. A fire pump which is inoperative for any reason at any time constitutes an impairment to the fire protection system. It should be returned to service without delav. '

Future A.2 A-2 7.6 Pump rooms and pump houses should be dry and free of condensate. Some heat may be required to accomplish this.

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.A-2-8.1 The exterior of aboseground steel piping should be kept painted. NOTE 2 A 2 is 2 h A-2-9.1 The exterior of steel suction piping should be kept painted.

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t ATTACHMENT F :

(section 3/4.8) !

i O Circle items noted in this attachment have been previously submitted. I V )

1) Section 3/4.8.3 (pg. 3/4 8-14) Onsite Power Distribution '

f Delete "(both) between redundant busses within the unit (and". l These tie breakers do not exist at Byron Station.

2) Section 3/4.8.4 (pg. 3/4 8-18) Electrical Equipment Protective Devices t 1 ,

Delete the previous request to add "125 VOLT-DC". At this voltage level the cable would fail before damage to the penetration could occur, j 4

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4 ELECTRICAL PCWER SYSTD*S PRODF & REYU COPY 3/4.8.3 CNSTTEPCWERb5MI3UTTCN

~~

' CPTRATING , , , (

, LIMITING C::NOTECN .CCR CPERATICN

3. 8. 3.1 The following electrical bussas snall to energi:ad in ce scocified manner wita tie treakers ocen Quw ..L . -i .~.ni :_:m -ir" ;; .cir ;

-fend between units at the same stationt.

a. Divi iaa "' A C. ESF Susses esnsisting of: ;

kyt - l

1) C /e*^ Sus 141, i

2) 480-volt. Bus 131X, and l

3) 440-Volt Bus 131Z.

B. 01visiaa 4.C. ESF Susses c:nsisting of: i

1) beu-voit

" 3 Bus 142 3) 480-volt aus L'E.

aus :.32x,and 'lue sAS GbAwd(1)

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c. 120-volt A.C. Sus El energi:ed fms its associated inver.ar t:nnectad '

to 0.C. Sus ni," l

d. UD-Volt A.C. Sus 113 energi:ed fr:s its associated invertar c:nnec.ad I to 0.C. Bus C1,a y

e. 120-Voit A.C. Sus n2 energi:ed frem its associated invertar- (

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connected to 0.C. Sus 1*.2,= and 120-volt A.C. Sus E4 energi:ed fre's its associated' invertar c nnectad tm 0.C. Sus n2." /

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A "* AIL *"*'; .402a t, z, 3, and 4 .

ACTICN: r

a. With one of the required divisions of A.C. ESF tusses not fully '

energized, reenergize the division wicin 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or te in at least HOT STAN08Y witMa **= next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in C3L3 SHUT';C'eN wicin the .

l following 30 y 9

h. With one A.C w+ tut tu t energi:ed, reenergi:e ce A.C. vital bus within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HQT STANC8Y wicin -as next j .b u m anu in C 0"G "; i iin 7e rollowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . l I

c. With one A.C. invertar inoperaale or not emnnected to its 0.C.

I power supoly, reenergiIs the A.C. vital bus fme its associated i invertar witain 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or to in at least MOT STANOSY witain ce next 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and in COLO SHUTCChN witain ne following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ] l J :

' ~ Ng5 ALE Witk (4ttAtiaan i h .

'Two invertars say te disc =nnec.ad f es their 0.C. Sus for un *4 24 neurs as necessary, tns.faeno of Ms.m..g ni ecc2Ti:ing carse on cair i associatad t at fag,f.aak revi . (1) 1 tusses are energi:ed, and i

(2) the vita 1 sg ass c taaf.Stas air t*at=f : ant areenegi:sd from eneir i ociated invertars ud c:nnectad to =eir asscciated 0.w. :cs. ^

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SYRCN - UNIT 1 3/4 S-14 s

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ATTACHMENT G (Section 3/4.9)

Circled items noted in this attachment have been previously submitted.

1) Section 3/4.9.6 (pg. 3/4 9-6) Manipulator Crane.

-Item b(1), previously a change was submitted to change the rated capacity of the auxiliary hoist from 3000 pounds to 2500 pounds. The actual number for the auxiliary hoist rated capacity in 2000 pounds per Manipulator Crane Stearns Roger Incart.x>retted vendor manual.

-Delete " Refueling Machine" and insert " Manipulator Crane". This is correct Byron Station nomenclature.

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ELECTRICAL PCVER SYSTTMS SURVEILLANCE REOUIREMENTS (Continuec) c) For each circuit breaker found inoceraale curing these .

functional tests, an additional representative samole of at least 10% of all the circuit breakers of the inocertole type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tastec.

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2) By selecting and functionally tasting a ortsentative samole of at least 1C% of each type of 480-volt circuit breaker.

Circuit breakers selected for functional testing shall be selected on a rotat W h>=4= N #"a % a=! teet re1' consis jecting a current input <t - M :t I " - ' *.

t en selected circuit breaker and verifying tnat eacn circuit reeker functions as designed : c *t: - :::^:: tH: ; '..;

t.1 .- r :n ' - in: :::: c -*2. Circuit breakers found .

nocer unctional testing sna11 be restortc. w .

0 8LE status prior to resuming coeration. For each circuit

' breaker found inoperable during these functional tests, an n additional recrosentative samole of at least 10% of all the

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circuit breakars of the inoperante type snall also be func- .

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tionally tasted until no more failurts are found or all circuit

h. breakers of that type have been functionally tasted; and

3) By selecting and functionally testing a coortsentative samole of eacn type of fuse on a rotating basis. Eacn ripresentative sample of fuses shall incluce at least 10% of all fuses of that l type. The functional test sna11 consist of a noncestructive resistance seasursenent test whicn cemenstrates snat the fuse i

I meets its manuf acturtr's design critaria. Fuses found inocer-able during these functional tasts shall be replaced with CPERA8LE fuses prior to rtsu:ning oceration. , For eacn fuse i found incoeraale during these functional tests, an accitional J

representative sasole of at least 10% of all fuses of tnat type sna11 be functionally tasted until no more failurts art founc cr all fuses of that type nave been functionally tastec.

D. At least once per 60 montas by subjecting eacn 7 W circuit breaker to an inspect.f on and preventive maintanance in accorcance witn o-ocacurts prepared in conjunction with its manuf acturtr's encomencations.

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SYRCN - UNIT 1 3/4 8~18

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IJ' . REFUELING OPERATIONS Pn00F & Ray ggpy :

. _ , _ , 3/4.9. 6 ".",; U.e .e^.;;IZ C (q ini cctog- s i- . _ ,

C.W GC. J '

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i LIMITING CONDITION FOR 0?ERATION _

inanipetor c.rt nt.

l 3.9.6 The refa L,; ;;;.;}= shall be used for movement of' drive rods or fuel i assemolies and shall be OPERA 8LE with: ,

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The r;p&puja10r crone---~ = used formvement of fuel assemolies having-

a. , . r, ted / 2 850 ~ '. (

1) A(QM:r apacity o(4864 poinds, and j

2) An overload cutoff limit less than or equal to 2350 pou.,ds.

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b. The auxiliary hoist used for latening and unlatening drive rods e '

l l having:hafed :$ '

A e4e4eue capacity of pounds, and X l 1)

2) A load indicator which shall be used to prevent lifting loads '

in excess of 1000 pounds. ' ' ~

APPLICA8ILITY: During movement of drive. rods or fuel assenclias witnin the j l reactor vessel. !

1

% ACTION:

With the requirements for cran'e' and/or hoist OPERABILITY not sstisified, suspend l use of any inoceraDie sanipulator crane and/or auxiliary noist from Operations involving the movement of drive rods and fuel assemolies witnin :ne reactor I vessel.

SURVE!LLANCE :tECUIREMENTS

4. 9. 6.1 Eacn manipulator crane used for movement of fuel assehelies witnin the reactor vessel shall be demonstrated CPERABLE witain 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ::rior to the start of such operations by performing a load tast of. at least 9290 EI',3 pounds and demonstrating an autasatic load cutoff wnen tne crane load exceeds 2850 pounds.

4.9.6.2 Each auxiliary hoist ANd associated load indicator useC for movement of drive rods within the reactor vessel shall be demonstrated CPERABLE within 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months sjet9e,to the start of sucn coerations by performing a load test of at leasy4000" pounds.

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SYRON - UNIT 1 -

}/4 94 1

0

. . i ATTACHMENT H (Bases Section 3/4)

[~) Circled items noted in this attachment have been previously submitted. f

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1) Section 3/4.2.2 (pg. B3/4 2-4) Heat Flux Hot Channel Factor, and RCS Flow +

Rate and Nuclear Enthalpy Rise Hot Channel Factor. i Delete the following sentences: "As noted on Figure 3.2-3, RCS Flow I rate....will not be below the design DNBR value." Also, "R as calculated in Specification 3.2.3....is the maximum "as measured" value allowed". ,

2) Section 3/4.2.3 (pg B3/4 2-5).

-Delete the sentence "When RCS flow rate....with the limits of Figure 3.2.3".

-Insert the words "following initial plant startup" between the words f "Therefore" and "a penalty". ,

-Delete the words "is included in Figure 3.2-3" and replace with "as well as a measurement error of 2.1% have been included in the limiting value

of 386,000 gpm for RCS total flow rate. The thermal design flow of 377,600 gpm increased for measurement error and penalized for feedwater i venturi fouling is 385,915 gpm." ,

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3) Section 3/4.7.10 (pg. B3/4 7-7).

p Add the words " Pump capacity is based on NFPA 20 1984 which calls for [

(,,/ 150% flow at 65% discharge pressure." after the second paragraph. .

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PCWER OISTRI3UTION LIMITS h khh '

l BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLCW RATE AND NUCLEAR 5NTHALFY RISE !

HOT CHANNEL FACTOR (Cont 1nuec) !

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c. The control red insertion limits of Specification 3.1.3.5 are I maintained, and I

d. The axial power distribution, excressed in ter-ns of AXIAL FLUX OIFFERENCE, is maintained witnin the limits.

N e g will be maint.iined within its limits proviced Conditions a. througn o--

d. acove are maintained. M n nd n Fig.ce 0.2-3, IC fie- , ;e e..d ~N

- r; : "tr:d:: Of'" :v in.L wn. .nv e. r (;..., . 'e. ;... ..e RC fiv. C as+m 4e m=an+sh1 7 2' ;n. .u.saurvu - la aIsu Iww) ta ec;uT: that thG ;&lC; -

Jete; ::4 R .iii noi . MI:= th: d::f gn := i '.e? The relaxation of F as a function of THERMAL POWER allows enanges in the radial =cwer snace for all permissible red insertion limits. l

! = n::.i.Lau .a $w.siiimian 3. 2.5 wu ...u .c. ~ ige;e 2.2 0, u wiL2 '

R - F' in: ta= Or :;=1 :: 1. r; . T:... ,elw. ;s .ee '

= ;;r':n n:' nnt--

m!pn n; e . inilw.nues param w.o w waa r den 4;R , .. g. , p e rd :!;c 5*--'

. _-_......~..~.,1. R. .uu i u '.= .awivu" 4a!w. . ' e e d .

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hU Fuel red bowing recucas tne value of CNB ratic. C.ecit is availacle to partially offset this reduction. This credit c:mes frem a generic cesign

,,3 ,,

margin wnica totals 9.1% wnen the analysis is perfor red witn tne accreved l l intarim methods. The targin used to partially offset red Ocw :enalties is 9.'.%.

I This margin breaks acwn as felicws: )

f 1) Design Ifmit ONBR (1.5)% i f

2) Grid scacing Ks (2.9)% !

3) Thermal Offfusion Coefficient (1.2)% *

4) DNBR multiolies (1.7)% i

! ; 5) Piten Reduction (1.7)%

The margin used to partially offset red bcw cenalities is (5.9)% *ita tne remain-ing (3.2)% used to trade off against measured ficw nica nay :e as nuen as (2)j ,

lower tnan tnermal design flow plus uncertainties. ThecanaltiesaccliectoFy to account for rod bow as a function of burnuo are censistant with taase des- I cribed in Mr. John F. Stol:'s (NRC) letter to T. M. Ancersen (Westingneusa) ,

cated Acril 5,1979 with the difference ceing cue to the amount of nargin / j eacn unit uses to cartially offset rod bcw cenalities. dj When an qF seasurement is taken, an allcwance for Octa ex er' mental error and manufacturing tolerance must be made. An alicwance of 5% i3 accrocriata for a full core mac taken with the Incore Detector Flux Macoing System, ar.d a 2% allowance is acercoriate for manufacturing tolerance.

.- THIS PAGE OPEN PENDING RECEIPT 0-'

m CN - UNIT :. INFORMp{}ggROM THE APPUCANT j

POWER OISTRIBUTICM LIMITS hh hh g

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he usEs .

d' HEAT FLUX HOT CNANNEL FACTOR, and RCS FLCW RATE ANO NUCLEAR ENTHALPv RISE HOT CHANNEL FACTOR (Continuec)

$ The Radial Peaking Factor, F9(Z), is seasured periodically to provide assurance that the Hot Channel Factor, Fg (Z), remains within its limit. The dt g, ,

'P-h F,y limit for RATED 1HERMAL PCWER (F j k *h Factor Limit Report per Soecification 6.7.L9 was deterwined free expected

) as provided in the Radial Peaking

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power control anneuvers over the full range of burnuo conditions in the core. l

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r: - ; W- t -; ;.-1;;r. wit:; tt;; ii;it ;f pr; . .. .weasurement jo errors of 2.1% for RCS total flow rata and 4% f5r F # have been allowed for

d l

3 N { in detamination of the design CN8R $d$dk value. M N M ,

jg- The se resent error f r Rt3 total ow rata is based ucon perforsing a !

, r$recision heat balance and ing the results to calibrata tne RCS flow rata :

a w indicators. P antial foul ng of the feedwatar venturi wnica signt not ce r

[ ; detected could ias the re it from the precision heat balance in a nonconsar- (

svative sanner. erefore, a. penalty of 0.1% for undetacted fouling of tne i f: 5?cd f "igre 0.2 0.*- Any fouling wnica signt bias i (dq,gfeedwaterventumthe RCS flow rata sensurement greatar than 0.1% can be detected!

Od$p If detected, action sna11 f j 'j ;

,$and be taken

.e.,

tnnding before various performing plant sucsecuent performance precisionparameters.

heat balance seasurements, either tne effect of fouling shall be quantified and concensatad for in i

Y 0 RCS flow rata sessursment or the venturi shall be cleaned to elisinata }

4 I :ne fouling. t k >

hour periodic suneillance of indicated RCS flow is sufficient to t

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flow degradation which could lead to coeration outside the ac:aotacle 0 Ydetect region eration. = ^ ::..'ig. 3.2 :." - ;

3/a.2.a CUA0 RANT PCWER TILT RATIO j The QUADRANT PCWER TILT RATIO limit assuns that the radial ocwer l l

' distribution satisfies the design values used in tne power caoacility analysis. [

Radial power distribution sessurements are sade curing STARTUP *asting and t periodically during power operation. l t

The limit af LO2, at which cor octive action is requind, provides CNB I and linear heat generation rata protection with x-y plane power tilts. A limit k of LO2 was selected to provide an allowance for the uncertainty associated with the indicated power tilt. j The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time allowanca for oceratien with a tilt condition greater ;

than LO2 but less than LOS is provided to allow identification anc cornc- !

tion of a dropped or sisaligned control rod. In the event suca ACCCN does SYRCN - UNIT 1 3 3/4 2-5 I

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. . I s l PLANT SYSTEMS :

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8ASES !

3/4.7.10 FIRE SUPPRESSION SYSTEMS i

The OPERA 8ILITY of the Fire Suppression Systems ensures that adequate !

,> fire suppression capability is available to confine and extinguish fires occurMng in any portion of the facility where safety-related equipment is located. The Fire Suppression System consists of the water system, spray, and/or sprinklers, CDs, Halon, and fire hose stations. The collective capa-bility of the Fire Suppression Systems is adequate to minimize potential damage ,

to safety-related equipment and is a major element in the facility Fire !

Protection Program. . !

f In the event that portions of the Fire Suppression Systems are inoperable, j alternate backup fire-fighting equipment is required to be made available in ;

the affected areas until the inoperable equipment is restored to service. i When the inoperable fire-fighting equipment is intended for use as a backup !'

anans of fire suppression, a longer period of time is allowed to provide n e alternate means of fire fighting than if the inoperable equipment is the i primary means of fire suppression.

b M m NFP A 2.0 , 63 4 l wb.A ra003 Sm 650% %ba d c da %. l The Surveillance Requirements provide surance that the einimum OPERA 8ILITY [

requirements of the Fire Suppression Systems are set. An allowance is ande t for ensuring a sufficient volume of Halon in the Halon storage tanks by veMfying l either the weignt or the level of the tanks. Level sensurements are made by !

either a U.L or F.M. approved method. J In the event the Fire Suppression Water System ba - inoperable, immediata -

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' corrective measures aust be taken since this system provides the sajor fire

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l l suppression capability of the plant. j 3/4.7.11 FIRE RATED ASSEMBLIES The functional integrity of the fire rated assemolies and barder .!

penetrations ensures that fires will be confined or adequately retardeo from L sorseding to adjacent portions of the facility. These design features minimize r the possibility of a single fire rapidly involving several areas of the facility ;

prior to detection of and the extinguishing of the fire. The fire barrier l

penetrations are a passive element in the facility fire protection program and i are subject to peModic inspections. !

Fire barMer penetrations, including cable penetration barriers, fire t doors and dampers are considered functional when the visually observed condition is the same as the as-designed condition. For those fire barrier peretrations that are not in ther as-designed condition, an evaluation shall be perfonned to show that the modification has not degraded the fire rating of the fire barrier i penetration. j During periods of time when a barrier is not functional, either: (1) a i continuous fire watch is required to be saintained in the vicinity of the e affected barrier, or (2) the fire detectors on at least one side of tne affected barMer must be veMfied OPEP.A8LE and an hourly fire watch patrol estaolished, !

until the barMer is restored to functional status. l l

BYRON - UNIT 1 8 3/4 7-7

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i ATTAC19fENT I [

(Section 5.0) ;

1) Section 5.3.1 (pg. 5-4) Fuel Assemblies.

O)

N Delete the number "1748" and replace with "1594". This new number represents the maximum total gram weight of the uranium in each optimized ,

Fuel Assembly (OFA) rod as reconnended by Westinghouse. ,

2) Section 5.3.2 (pg. 5-4) Control Rod Asemblies.

Delete the words " full-length and no part-length" and " full-length" from this paragraph. Byron Station does not utilize part-length control rods !

and it is therefore unnecessary to reference full-length or part-length l when discussing control rod assemblies. ;

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0,35,13 "EATURES .

5. 3 REACTORCCR$ i FUEL A55088 LIES

}169Y 5.3.1 The ce shall contain 193 fual assemolies with esca fuel assemoly ,

containing 64 fuel rods clad with Iircalcy-4. Eaca fuel rod sna11 have a l x nominal ive fuel length of 144 incnes and contain a saximum total weignt !

of 944 uranitas. The initial core loading shall have a saximum enricament .,

of 3.10 weignt percent U-235. Reload fuel sna11 be sia11ar in anysical ::ssign to the initial core loading and shall have a saximum enricament of 3.50 weigns ,

l i;.. A G U=235. ,

O CONT 40L ROC A55088 LIES , ,

x l l 5.3.2 The core shall contaig53 -M1 -i ,;'. = .a ,,.. ; R.VY. control rod x assasolies. The fu"-1 rp. control rod assemolies sna11 contai.n a nominal !

142 inches of abscreer satarial. All control rods snail be hafnius, clad with !

stainless steel tuning.'

3 i

5.4 REACTOR CCOLANT SYST24 ?

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OESIGN PRE.SSURE AN0' TEdPERATURE l j 1

5.4.1 The Reactor Coolant System is designed and snail be saintained:

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a. In accordanca with the cada requirements specified in Section 5.2 of the FSAA, with allowanca for normal degradation pursuant to the I applicable Surveillanca Requirsments,

b. For a pressure of 2485 psig, and .

,a

c. For a tammerature of 650*F, excapt for the pressurf::ar wnfcn is 680*F. .

VOLUME

(

5.4.2 The total water and staan volume of the Reactor Coolant Systas is $

12,257 cunic feet at a nominal T,,, of 587.4*F.

5.5 wf7EOROLOGICAL TCWER LCCATION j

5. 5.1 The seteorological :swer shall te located as snown on Figurs 5.1-1. [

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3YRCN - UNIT 1 5-4

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ATTACHMENT J (Section 6.0) i

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

1). Section 6.7.1.9 (pg. 6-20) Radial Peaking Factor Limit Report.

Delete the following words from this section: ,

i

a. "NRC Regional Administrator with a copy to" j

b. " Attention: Chief, Core Performance Branch"

c. " Washington. D.C. 20555" [,

d. "at least 60 days" Also, insert the words "with the reload analysis submitted or technical !

specificaton change" between the words " criticality" and "unless !

otherwise approved". !

2). Section 6.7.1.9 (pg. 6-21) Radial Peaking Factor Limit Report. ;

Remove the words " submittal or an.... Factor Limit Report" and insert *

" reviewed by the commission". Also, delete the words "60 days". :

The above Section 6.7.1.9 changes are consistent with the reporting I method used by the office stations at Commonwealth Edison. l I

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ADMINISTRATIVE CONTROLS

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g REPORTING REQUIREMENTS (Continued)

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,- / Ry i Th emiann Radioact e Efflu 't Relea Reoort,Ao me,5ucai :ac 60 cays i afts wanuary f each y r shall so inci e an assessmoort of rao1'ation cases t e likel most exposed MEMBE OF'THE P LIC fros'reac2sr releases and exhor eerey urytt um fuel cle sourens, inci ing dosesIfroadrisaryfsffluent,pata- / l ways an(direct rg ation, r the p ious calendar year to 5dow conformance / l g 0, "En ronmenta Radiation Protection 5 neards f NugMar with CFR Pa . t i

r Operati

Acc le set. ds for epicu)a$4ng the se cent but eous of uants a given inf Regulle(on Guid 1.109, ev. .,ji n fro P .

quid and Octoder .

& ?!L-&~A GulAsm.d+k The 5 9 ... ' ".3' t'.T. ; N ... .; ".S.-Q'$t-rts tha11 include a list j

, 1

,D and description of unplanned releases from tne site to UNRESTRICTED AREAS of l l

radioactive satorials in gaseous and liquid effluents made during :ne reporting :

1 period. !

~~

, h J j The annual ioacti,WiffluenyA6 Tease RescTts snjM inct3ae any ,, i enan saca duri tne reperting peptd to tae,7C and Jar tne 00CM, cursuarit l' 3 pecificatipos 6.R 6.12 e actively is well,as any u%e onangsis to iquid, Gas us or 5 d Radwa Treatmen Systems /pursuany/ o Scepf' fica- i tion 6. . Lt sna also in ude a lis q of ne /locatic for code cal a- l nd/or env neental nitoring centifiye(by the no use J'ursu !

/ ensus/ '

(tions to ocifica n 3.12.2, l A j .he i

W ^^ M . . .

... . __ . __ _,Nf__EMI .. _. ... _ ... h . ,- % . . . all also incluce s' i g an explanation as to wny tne inoceracility of licuid or gaseous ;

v the following:

l effluent monitoring instrumentation was not corrected witnin :ne time scocified '

in Saecifications 3.3.3.10 or 3.3.3.n, respectively; and description of the events leading to liquid holduo tants or gas storage tanks exceeding ne limits I I

of Specification 3.H.1.4 or 3.H.2.6, resoectively. f O MONTHLY OPERATING REPORT j

6. 7.1. 8 Routine reports of operating statistics and snutccwn excerf ence. l including doc.usentation of all challenges to tne PORVs or safety valves, snail ,

be suositted on a sonthly basis to tne Director, Office of Resource Management, !

l l 20555, witn a cocy to U.S. Nuclear Regulatory Comeission, wasnington, D.C.

tne Regional Administrator of tne NRC Regional Office, no later taan ne 15tn l0 of eacn conta following the calendar senta covered by the report. l f

RAOIAL PEAK *.NG FACTOR t.IMIT REPORT N f 6.7.1.9 The F xy 11eits for Rated Thornal Power (Fxy,) sna11 be scovided to One ,

lO l "" "wi: e -utrat-- sen : =:y :.yOpetorofNuclearReactorRequiation.

Chief, ".m F;-10.xx Pr:n, U. S. Nuclear Regulatory Comeission, l

l '*ttent h .

"r"%.., ^. ;. ;^^:':"i'er all core planes containing Bank '0" control rods and j I

all unrodded core planes and the plot of predicted (F q .Pg ,j) <s Axial Core Heignt with the limit enveloce at h;;t Z c y:$rior to eacn cycleIninitial adaition, O

criticality unless otnerwise approved my :ne Cosumission by lettar.

M N M cec p M n~ M M p pf h ebg . 6-20 j

SYRON - UNIT 1 O .

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' PR00F & REVIEW COPY l ADMINISTRATIVE CONTROLS RACIAL PEAKING FACTOR LIMIT REPCRT (Continued) i in the event that the limit should change requiring a new r:-itt.1 ,

t S #.1 h =' Q M , .

_'-: t.' O 1 ^.. . ., .~.c . Li 1 : "r: Tit shall be sucaitted (+*--

Mr:ior to the date the limit wuld become effective unless otherwise ATP I approved t;y the Cassission by letter. Any information needed to support F l will be by request from the NRC and need not De included in this report. O i'

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! SPECIAL REPORTS 6.7.2 Special reports saall be submitted to the Regional Administrator of the NRC Regional Offica within the time period specified for each report.

2 6.8 RECORD RET 9f7 ION .

l In addition to the applicaele record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the sinieue period indicaied.

6.8.1 The following records sna11 be retained for at least 5 years:

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a. Records and logs of unit operation covering time interval at each power level;

b. Records and logs of principal saintenance activities, inspections, repair and reolacement of principal items of equipment related to nuclear safety;

c. All REPORTA8LE EVENT 5;

d. Records of surveillance activities, inspections, and calibrations y required by these Technical Specifications;

e. Records of changes made to the procedures required by Specification 6.6.1;

f. Records of radioactive snissents;

g. Records of sealed source and fission detector lean tests and results; and l

N. Records of annual pnysical inventorf of all sealed source sacerial of record.

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6.8.2 The following records sna11 oe retained for tne duration of tne Unit Operating License:

a. Records and drawing changes rifiecting unit design nodifications I sade to systems and equipment described in tne Final Safety Analysis l

Report;

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- b. Records of new and irradiated fuel Inventory, fuel transfers and asseemly burnuo histories; SYRON - UNIT 1 6-21

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ML20092L150

Text

Subject:

Dear Mr. Denton:

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