Text

Cycle 2 Reload Rept
	ML19308B546
Person / Time
Site:	Oconee
Issue date:	06/30/1977
From:	; BABCOCK & WILCOX CO.
To:	;
References
	BAW-1432, NUDOCS 8001090541
	Download: ML19308B546 (75)
	v • d • e

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

l c? Ode BAW-1432 June 1976 d

OCONEE UdIT 3 CYCLE 2

- Reload Report -

m i

NOTICE -

i THE ATTACHto FILis ARE OFFICI AL RECOMOs OF THE Division OF DocuwtNT CONTROL THEY wavt BitN CH ARGt0 TO YOU FOR A LlwalE D f tWL Pt nico AND l uusi et RituRNeo to int Rf CoRos.s ACilsn sRANCs oss ce Ast oo Nor stND coCouts swARoto our isRoucw ine wait Rtuovat or ANv, r 6ctis: Frow ooCoutNT Fon RtrRoDuctioN wust i

Si RE F E RRE D 10 f it t PE RSONNE L.

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Of ADLINE RETURN DATE i

i RECORDS F ACattT v SH ANCH Babcock &Wilcox 8001000 S 91 P

Sr.- 14 L'

une 14: ~

OCO.NEL L'N! T 3. CYCLF. 2

- Reload Report -

4 B\\BCOCK & WILCOX Power Cencration Grou;.

Nuclear Power Generation Division P. O. Box 1260 Lynchburg. Virginia 24505

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Babcock & Wilcox

l C_O.N.T_E NTS Pace 1.

INTRO *ECTION.

J 1-1 2.

OPEkATING illSTORY 1-1 Gil.ER.5L DESCRIPT!.)N 1-1 4.

Ft'EL SYSTEM DESIGN.

4.1.

Fuel Assembly Mechanical Design 4-1

4. 2.

Fuel Rcd Design 4-1 4.2.1.

ClaJding Collap*c 4-1 4.2.2.

Cladding Stress 4-2

4. 2.1.

Fuel Pellet Irradiation Svelling.

4-2

4. s.

T he r m.n l Den i g n.

4-2

4. 1.1.

Power Spike Madel 4-1 4.1.2.

Furt Teeperature Analysis 4-3 4.4 Material Design 4-1 4.5.

operating Experience.

4*

5.

NUC!. EAR ~tESIGN.

5-1 5.1.

Physics Charac ter f at ics 5-1 5.2.

An.nlytical Input 5-2 5.3.

C hanges in Nuclear Design 5-2 6

litLR't\\L-HYDRAt!LIC DESIGN.

6-1 4.1.

Therma l-Hyd rau l ic Design Calculations 6-1 5.1.1.

Introduction of Mirk B4 Assemblics 6-1 6.1.2.

Increased RC Syst em Flow.

6-1 6.1.3.

EaW-2 DNB Correlation b-2 6.2.

DNHR Analysis 6-2

6. 3.

Pressure-Tempera.ure Limit Evaluation 6-3 6.4 Flux-to-Flow Setpoint Evaluation.

6-1 7.

ACCIDE'.*T AND TRANSIENT. ANALYSIS 1-1 7.1.

Cencrat Safety Analysis 7-1 7.2.

Rad Wit hdrawal Accident s 7-1 7. 1.

Moderator Dilution Accident 1-2 7.4.

Cold Water (Pump Startup) Accident 7-5 7.5.

Lons of Coolant Flow.

1-1

7. b.

Stuck-Out. Stuck-In. or Dropped Control Rod 7-4 7.7 Loss of Elect ric Power...................

1-4 7.3.

Steam Line Failure.

1-5 7.9 Steam Generator Tube Failure.

7-5

. gig.

Babcock s. Wilcox

[N N.5.3.91' lid,}

Page 7.10.

Fuel Handling Accident 7

's 7.11.

Pod Election Acsident 7-5 7.12.

Maximon Hypothetical Acc: dent

.' - S 7.13.

Vaste Cas Tank Rupture.

7-S 7.14.

LOCA Analyhis I-6 8.

l'ROPOSED !ODIFILATIONS TO TECHNICAL SPECIFICATIONS 3-1 9.

STARTL'P PR'OCRAM.

9-1 REFERENCES A-1 List of Tables Table 4-1 Fuel Design Parameters 4-5 4 2.

Fuel Rod Dimensions 4-5 4-3.

Input Summary for Cladding Creep Collapse Calculations 4-6 4-4.

Fuel Thermal Analysis Parameters 4-7 5-1.

Oconee 3. Cycle 1 and 2 Physics Parameters 5-3 5-2.

Shutdown MarAin Calculation f or Oconee 3. Cycle 2....

5-5 b-1.

Cyc le I and 2 Maxinus Design Conditions.

6-5 7-1.

Comparison of Key Parameters for Accident Analysis 7-7 List of Figures, Fi gure 3-1.

Core loading Diagram for Oconee 3. Cycle 2 3-3 3-2.

Enrichment and Burnup Distribution for Oconee 3. Cycle 2 3-4 3-3.

Cont rol Rod Locations for Oconee 3. Cycle 2........

3-5 4-1.

>bximum Cap Size Ve Axial Position - Oconce 3. Cycle 2 Batch 4 4-E 4-2.

Power Spike Factor Vs Axial Position - Ocnnee 3. Cycle 2 Batch 4 4-9 5-1.

BOC (4 EFPD) Cvele 2 Two-Dimensional Relative. Power D!ntribution - Full Power. Fquilibrium Xenon Normal Rod Posittens (Croups 7 and 8 Inserted) 5-6 8-1.

Oconee 3. Cvele 2 - Core Protection Safety Limits 3-2 6-2. Oconce 3. Cycle 2 - Core Protection Safety Limits..

6- )

S-3.

Oconec 3. Cvele 2 - Core Protection Safety Limits 3-4 6-4.

Oconec 3. Cycle 2 - Protective System Maximun Allowable..

B-5 Sctroints H-5 l

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Ck & WCOR

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f 1.pa re Page s-5.

nonce 3. Cycle 2 - Pre t ec t i ve Systes.%x t r.ua All.sible tetpoint.

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6-4.

nronce

3. Cycle 2 - Fe>d Position Limits f or Four-P snp s te r-st ion Fro m 0 tn !!) ( 10) Ef1D s-7

%-1.

skonce 1 C/a le 2 Pod Position Itmits f or Four-harp o,seration Fron !!5 ( 10) to 226 ( 10s EFPD

%-o M-8.

e ncemcc 3. Cycle 2 Fod Position-l.imits f or Four-Pupp i

optr.ition After 226 (*IO) EFPD

-4 h-9.

ocone.- ~). Cycle 2 Had Position Limits for Two-and Three-Pu p oper.st ion From 0 to !!5 (?!O) EFPD.

?-10 M-I fs, oconce-3. CVele 2 -- Ho.1 Pos i t ion I.!mi t s for Two-ard thes e-Pep operat ion From !!'s ( 10) to 226 (*10) EFPD.

-11 n-ll.

Ocunee !. Cvele 2 - Pod Positten Limits for Two-and Three-l*nnp Operatio After 226 (*10) EFPD.

4-12 h-12.

encensc 3. Cycle 2 - Operation.nl Power Imbalance Envelope f o r ope r.s t ion from 0 to 115 (!!O) EFPD M-I l M-1 1.

Oconee 1. Cycle 2 0;.crat ion.a l Power Inh.elance Envelope f or Oper.at ion Fron 115 ('IO) to 226 (*10) E FPl>

6- ! !.

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1. Cyc!c 2 Operat ion-sl Power Imb.slance Envelope for oper.4 tion Altsr 226 (*10) I:ITD X-IS l

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TKODt LTIO'4 AM St&.%"r TSi% relert J ust if ic% t he oper.at ion of the sciend tycle of sk once % sci..er S t.s-tion, l'n i t ' ) Jt the r.sted core power o f 2 %8.'E t.

In.liacJ art-the t r.;u a r e !

.an s ty*es as est l ined in t he l'SNkr dor ueent "Gu i d.a nc e f o r s' t o e.. e.1 f i.. a..

An.n.iment* Ne!.iting. to Refueling.",func 1915.

Vo support evele 2 oper.st ion ut Oconce Unit

's, t h i s repor t cep lov..an.e i ve <.il techn t.lue s.end design bases established in reports tiat s<re p r.-v i o.a.1 v - u 5-esit t eil.and.secept ed by t he 1:SNkC.ind it s predecesso, acc retcren.in..

A brief wurn.iry of cycle 1.ind 2 reecrier p.ir4 meters r e la t c4 to sewer c.sp.abilit y 1+ includril in section 5 of this report. All of the 4.*cadents.in.n l y zed la t he FSAM h.svc been reviewed for cycle 2 operation.

i n t hm.e c a sc* shire sycle 2 ch.or.acter ist ics proved to be conserv.at ive witta r e pe<t to r aio se.in.e l v e nt for cycle 1. no new an.nlysen were per f ormed.

1he Technie.sl Speci f icat ions fusvc heen reviewed, and t he rmx!!! ic.it iens requ a rn it 1or cyclc 2 oper.ation are justificd in thiS report.

Itssed on t he analye.cs per f ormed, which t ake into accetint the postut.ated ettccts of f uel demit ic.at ion.snd the Final Acceptance Criteria f or Energency Lore Cool-ing Syst ems, it h.sw been concluded t h st skonce t' nit 1.'Cyc!c 2. c.in be s.s t e l y oper.it e d.st t he r.at e4 power level ot' 2 %8 'Si t.

1-1 Babcock & Wilcox -

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2.

OPtA\\ TING HIS10MY t.'n i t 3 of the thonee Nuclear Station achieved init i.a t cr it ic.al tt y on Lpt.nber 5.1974. and power escalat ion crane nced on September i t.1974 The IN% p.

cr 1.sel-of 2 568.%'t was rc. ached on December 16, 197..

A control rod inter h.inge v.a s pe r f o rmed.e t 257 effective full-power days ( F.F PD ). The design f uel cycle i scheduled for completion in %ptember 1976 af ter 46M EfPD.

The f i r.* t ry. le involve,J no operating anomatics t hat would adversely at t evt fuel perf orm.ance during t ht-necond cycle.

Operat ton of cycle 2 in schedulni to begin in early November 1976 Tlw Jewign cycle length in 265 EFPD,.and no control rod interchanges are pl.anned.

2-1 Babcock & Wilcox

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3. ' t ENEPA!. DESCetII'T10*.

The Oconee _ l'n t t i reactor core is described in detal! f a chapt er 3 of t he

',n i t 1 11AR. I 'Ibe cycle 2 core consists of 17.' f uc t a sse:.blies, eac n of which I.~.a 15 by 15.irray. ont aining 20rl f uct rods, 16 control ro.2.:u id e tabes, and one incore inst r uraent guide tube.

Th.- f uel rud c ladd inr :. t old-s.;rned Zirca19y-4 wit h.in OD of 0.4 30 inch and a wall t h!ckness nf 0.02%$ inth.

The fuel con-siww of dishe.i-end, cylindrical pelletr. of :ranius dios ade wht. h. ore 0.70d inch in !cagth.and 0.16's5 inch in diameter.

(See T.bles 4-1 and 4-2 t or addi-

- t inna t data. ) A!! fuet..steabiles in cycle 2 a.aintain.: constant nie t n.n l f ue l load leir. ut 463.6 bg of uranium. The undensit led nocalnal active tuel lengths

.irwi t heoret ical densit ies vary wilght ly betweer, hatches.

Specitic v.!ces are given in Tablee 4-1 and 4-2.

Figure 1-1 16 the core loading diagr-am for Oconee 3,aycle 2.

The in it i.i t en-r a c t.=en t s of b.s t c hes JA 2B, and 3 were 2.60, 2.67, and 3.00 wt u r.a n i um-2 3 5 respeitively. Batch 4 1-enriched to 2.53 wt I ur.an iu s-J 15.

All t he b.s t e h I assenblies will be-discharged at the end of cycle 1.

The ba t c h 2A, 2 h..ind I ass.nblics will Se shuffled to new locations. The ba t c h 4 a v..e b l i c s will oc-cupy prim.arily t he periphery of the core and eight loc.a t s on s in it s interior.

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~ Figure 2 is an eight h-cort-map showing the assembly burnup and enrier.nint I

distribution at t he beginning of cycle 2.

Heac t ivit y cont rol is supplied 'by tal f ull-length Ag-in-Cd cont rol rawls and soluble bornn shim.

In adJit ton to the full-lengt h cont rol rods, eight pa r t i.s !-

lengt h axial power sl.sping rods (APMts) are provided tor. addit ional cont rol of

.ix tal power dist ribut ion.

The cycle 2 loc.ations of the M s ont rol ro.f s arsi the

. group designations are indicated in Figure 3-3.

The core locat ions ut the t o t.a !

pat t ern (69 cont rol rods) for cycle 2 are identical to tiene ut the reference cycle indicat ed in Chapt er J of t he FSAR. I However, t he group design.st ion 6 dif f er het' ween cycle 2 and t he reterence cycle to minimite-pow r peaking.

Neit her cont rol rod interchange nor burnable poison rods

.a r e necessary f or cyc le 2..

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%/* : o pressure is 2200 psia, ar.d the core aver.:.e cens a t..c Ibe l 8.* *61 ratt is $.84 W/f at the rated Core power Of 25t,5 ?Mt.

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1 TOTAL 61 33 Babcoca L W:!ces

4 I L'E!. SYSTEM Di' SIGN 4.1.

f ul As cable Pechanical Dest.:n pertt:ent t uel dest n parameters are listed in Table 4-1.

A!! :uel.nssenblics are al etteal in ce.scspt and are nech.inically int err tung...adle.

The ns w tuel a e -. -: 9 a w s incorp*r. ate minor modit tra tons to the.r.d ift t t.v s. prinarit,. to riane s tuel as ratils pressure drop and increa.e lio lddown

=.o rc in.

All other results pre-ent ed ta t he FSAR fuel a.sembly mechanical discu< ton are appli-eabl. to t he relsud t uel assembli s.

4.2.

Fuel Rod re ico Pert ie. -t tuel rs4 da,siensions f or residu.il and new t uel.ar.

l a st ed in Table 4-2.

TV sechan acal. valu.st ion of t he f uel rod I. discussed N los.

4.J.1.

c 1..dd ing,1 lay y t*t. ep c el lapse analwes were per f ormed f or ttiree-ry le ansr.bly power histories.

II.e hatch 4 tuel is more limit ing tlun b. itch 2 and I luel cue t o conserv.at (vel v

.ns-tc.w d valises f or e l. add ing t hickness and initial ovality. Tab!c 4-3 is a sum-n.i r s et tN hatches 2, 1. and 4 t uel rod de.igns.

The anstebly power histories were an.a ls :nt..ind t N most limit ing a sechly f or eyele 2 v.a* determiness.

The proitetid asssebly rewr history t or the most limiting asscebly was used t o deterstne t he most 11eiting collapse tine as described i n E.W-I tMti4 PA. -

The con-e m ttsus in the analytical procedure are suns.trized below.

1.

T5 MT computer code was used to pre dict t he t ime to collapse. CROV con-

-errat avely piedict s collapse ti=es..ns demonstrated in reterence 2.

2.

k e r oi t t is taken for tission g.as release. Therefore, tN net differentia' pr ener es tased in t he an.slysis are conservat ively high.

.3.

The 'ut ch 2 and 3 cladding thickness used was t he I.TI. (loser t olerance limit )

of t N-.n s-buil t. mea su rement s.

The init ial ovalit y of the cl.idding used was t %.*n (upper telcrance limit) of the as-built measurement..

ihese values wre t nen t ren.k statistical s.ampling of the cladding.

F.a t c h 4 va l ues for cla0J tu t hiek: ess.ind ov. slit y were conservativelv asmed.

_t Babcock & Wdcox

4 Batch 2 and 3 cladding te=peratures were calculated using assembly outlet te=peratures.

This results in cladding temperaturas trat are conservative-ly high when combined with the =axi=um axial peak.

The rcs: limit ing assembl' y e as found to have a collapse t ime greater th.n the maxi =us projected lif e of the tuel assembly, as stuwn in Table 4-1.

This analysts was performed using the assumptions on densif'. cation describec referwece 2.

4. 2._2. Cladding Stress Since the fuel in batches 2. 3 and 4 is identical from a cladding stress point of ele.. tlw (alculat ions perf orwd in the Oconec 3 f uel densificat ion report '

are t h - cos t limiting.

4._2. 3.

thel pellet Irrad_tation Swelling The f uel design criteria spec tf y a limit cf 1.0% on cladding circumferential p l.a s t it' strain.

The pellet design is set so t ha t the plastic cladding strain is les-t han 11.it 55. 000 K.'d /st L'.

The conservat isms in t his analysis are l ist ed below.

1.

IFe nasicum specification value for the fuel pellet diameter was used.

2.

The =asieum specification value f or t he fuel pellet density was used.

n.

The ( ladd in g ID used was the lowest permitt ed spet it icat ion tolerar.ce.

4.

!te saximum expected three-cycle local pellet burnup is less t han 55.000 P-4/mtt!.

4,. 1.

! berna 1 Design The core loading for cycle 2 operation is shown in Figure 3-1.

There are 56 fresh e5atch 4) fuel assemblics and 121 once-burned (batches 2 and 1) fuel

.is-semblics. All fuel assemblics of this core are thermally and geomet rically similar.

natch 1 f uel, however, was resintered to increase tne init ial nominal densit y to 45.5% TD.

Although resintering of manufactured f uel result s in variatacns in pellet diameter and density, these variations were rot large enough to reduce the heat rate to which batch 3 can be exposed.

The linear heat rate capabilit y of the f uel in the cycle 2 core meet s or exceeds the de-sign Itnit of 20.15-kk'/ft.

1.inea. heat rate capabilit ies are based on (enter-line fuel melt and were established utilizing the TAFY-l' code with full fuel densktac_stion penalties.

l 4_2 Babcock t iMilcox i

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4. 3.1.

Power _ Sal,k_c Model The pever spik.: model used 1:. ti.c analyses f or t hti.

re;s : t a the same as t mat presented in S.W-lG955 wit h two :.oci fic.ations.

These ns.ai: icat ions h.sve seen applied to the nadel probabt11 ties F and F. and were rmte to reflec t adi tt f ana.

P.

K re act or opes at ing data t h.at support a somewn.it difterent approach and vield le s severe pen.alties Jue La power -pikes.

F was chucm tros 1. 0 t o si. 5, a nd f.

g was change d t ron a (.aussian to a lin* ar dist rib at i n. shici. ret tect s a d.creas-ing t ra rpse:ncy with increasing gap size. Wit h t 'ars.- shanro...n in it la ! ! + 1 density of 90.0 TD. and.m enric.Sent of 3.0 st u ran ie 2 D. t ut m isur cap site versus axial posit ten a:.2 t h. power pike fac t. r u r as axial po.it aen c.it. ulat ed f or t he batch a f uel (Figures 4-1 sere

d 4-L.

t he corresNnding f.ict or. for t he batch 2 and 5 f ut 1 wo.s id be e mal le r. ; r i ar il y becau se of t t.e higher init ic.! densit y of thee fuels.

Thus, applyinc th. -e gap.f re and twer spake f actors to t he cycle core de.ign gives coa 4crvative renetts.

4. _A.-l..._I"F.I..T cpera t u re Anal v s i s A

thermi analysi. of the f uel rain assumed in-re setor tuel densitzt.itaon to %.R TD.

The ba.is f or t he.anais sts is given in references J..ind i. s i t h t t.c rollow-l ing cwlifleationst 1.

The code opt ion for no re.t ructuring of f uel has been used in t his ata l s s i s in accordance with the NRC's interim evaluation et l

Im.

l 2.

The calculated gap conductance was reduced by 25! in.ac s ordance i. i t h t he NRC's int erim evaluat ion of TAFY.

Ihsring cycle j oper.3 tion the hichest relative assea.b!v pv.'er level occur. in i

hatch I tuel (see Figures 3-1 a nd 5-1 ). The f uel t o:perat ur e analvsit d oc.2-mented in the Oconee 1 f uel demit ication report ' is ba-cd.in b.it ch J f uel.

1 hts analysis provides a ce.

+rvative est imate f or haten }.and i s a;>p i t t.a b h j

to batch 4 The average caua 4 fuel teciperature at t he no -in.al hea t r.ite L.

estimated to be the same as f or h.a t ch 2. al t hough t he nem ana l l i nc.a r heat rate for batch 4 is slightly higher. These fuel ecperatures and pellet diantters are cocpared in Table 4-4. whic i gives thermal paranctors tor the t uel in the cycle 2 core.

'. 4. Material Desig i

I The batch. 4 t'uel a.senblics are not new in concept, nor do they ut il f re dit-ferent component materials. Therefore, the chemical ce:patibility of all 43 Batn.ock s. Wilcox 1

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f

'possible fuel-cladding-coolant-assembly interactions f or the batch *. fuel

.n-semblics are' identical to those of the present iuel.

J

,4. 5.

Elperating Experiences I

B&W's operating experience has been demonstrated in the operation of six 177-fuel' assembly plants utilizing this fuel assembly design.

i i

i l'

4-4 j._

i 4

i i

l.

i i

k

.f i

+

4 1

3 J

1~

Babcock & Wi cox 8

L

4. 4 n

j.

1-

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

,,-....,,..m......

"Iable 4-1.

Fuel Design Pa r t.m e t ee r s kesidual :uel a.;s.ediv New FA:

Cc=ponent Batch 2 B ?ch 4 Ra t. b _ '.

Fuel asstably type Mark al Mari B1 brk %

Nueber 25/33 60 5e Initial fuel enrich.. wt! i:

2.60/2.67

.l. cs) 2.51 t

Initial fuel densitv. :; TD 94.0(*}

9 ". 0 (' '

v4.o Init ial f ill ;;as pre 6sure, psia (b) tS)

(np l'-s t rh Burnap 4 lu)C. mwd /mt U 16.950!!6.80rl 11.304 0

rle.dding collapse time. EFPil

>29.000 29.000 2.'.tk 9 Design life. EFPil 24.600

24. Nk)

.'O. U n (a)1hese v.alues represented the worst case and are conservative valuss.

(b)1*ropr letary..

Table 4-2.

Fuel Rod Dimensions Residual fuel:

New tuel:

Component batches 2 and 3 batch 4 Fuel rods OD. inch 0.430 ID. inch 0.430 0.377 0.377 Fuel pellets OD. inch

<0.3694IdI 0.1695 Fuel. density. 1 TD 9;.o(al 14.0 l'nJensifled' active fuel length. in.

142.0 ")

I 142.2 Flexible spacers.. type Spring Spring Solid spacers, m.a t er ia l Zr-4 Zr-4 (a)These values represented the worst case and are con <ervatIve values.

Babcockr.WJcom c.3

Table 4-1.

Input Sunnary for Cladding Creep Coll. apse Calculation.

Batches 2 and 3 Catch 4 Pellet 03 (mean speciffed). in.

0.3694 0.3700 Pellet density (mean specified). I TD-

>94.0(*

93.5'D' Densified pellet OD. in.

0.3664 0.3663 Cladding ID (mean specified). in.

0.377 0.377 Cladding ovality. (L'TL). in.

(c)

Cladding thickness (LTI.), in.

(c)

Prepressure (minimum specified), psia (c)

(c)

Post densific tion prepressure (cold), psia (c)

(c)

Reactor system prensure, psia 2200 2200 Stack height (undensifled), in.

142.0

142.6

(" These values represented the worst case and are conservat ive'.

These values conservat ively envelop the design conditions.

' PROPR I ETARY.

l

=

8 4-6 Babcock & Wilcox i

i i

Table 4-4.

Feel Therr.at Analysis Parameters Ratch 2 hatch J Batch 4 Fuel pellet (nominal)

Init ial density. % TD 96.0(4) 95.5(b) 94.0(a) g 4

Initial stack length, in.

141.0,)

141.O(b) 1 2.2,)

43) g Initi.sl dianoter, in.

0.3670 0.3640(b) 0.3695,)

Densif ied stack length, in.

140.65 140.30 140.45 Densified diancter, in.

0.3646 0.3649 0.364o llCF on linear heat rate 1.014 1.014 1.014 I

E Nonima l linear heat rate, kW/ft 5.79 "}

5.50

5.80 *'

Avg fuel temp at nominal LHA. F 1310 1305 1310 I

Id' 1.liR to centerline fuel relt, kW/ft 20.15 20.15'd' 20.tS (a)Specificatton values.

(b)Noninal values after resintering.

(e) Based on dens ified length.

(d) Design limit. The linear heat r.te capability of each fuel. assembly is determined from as-built or resinter data.

4y Babcock & Wsicox

i Flp,ure 4-1.

Nxtz:nm Cap Sire Vs Axial Position - oconce 3. Cycle 2. Ntch 4 (Bssed on dennification f rom 94 to 96.5% TD) 3.0 4

Ex E

2.0 C

5 t

=

3 1.0 I

5 0

8 i

i j

0 20 40 60 8C 100 120 140 4

Asia! Locatfon. inches 4

E X

Rw P

$X R

1

A f

Y g

,a c

r t

u meal C

e.

u s.

~u Av i

I I

s

\\c

., n c.

cc o%

uo ++

c l

e cr o

oa

.o.

.u.

.=

C U

34 ac c

oe 4

n. u e.

G e

.e. 6.

O M

u

< c c

a

-e=

o s.

c A

n

-t e be

.o. w

.M.

o-q a x<

w 4, l

- i s

.2 ed 7~4

~

m c

i

3

~r w u

J

> n o sa A ~*

",8

-e o

u s.s'u

l w

l t

f r

I a

+

~r m_

a i

~

o.

.o.

o.

c.

g n

c

~

Joased avds Janod l

4-9 BabC0Ck & Wilcar

5.

!.TCt. EAR DE513

,5,. l. Phyvsics Characterint in Table 5-1 ccep. ares the core physics parameters of cyclea 1 and 2; the salues for both cycles were generated using PDQO7. Since t he co r e ha s not yet resen-ed an equilibriu:s cycle, dif f erences in core physics para eters are to

>e ex-pected between t he eyeles. The shorter cycle 2 will prod.ce.

r-all r rysle dif f erent ial burnup than that fo cycle 1.

The accumulated. average core burn-up will be higher in cycle 2 than in cycle 1 because of the presence of the once-burned b.stch 2 and 3 f uel.

Figure 5-1 illustrates a representative rel.a-t ive power distribut aon f or the beginning of the second cycle at f ull pswer wit h equilibr aun menon and norr-31 rod posit ions.

The critical boron concentrations for cycla 2 are lower in all c.ame's than for cycle 1.

As ind icat ed in Table 5-2. t!.c cont rol rod wort hs are sufficient to is.s i n t a i n t he required ahutdown nargin. However, due to changes in isotopics and the radial flux dist ribution, the BOC hot, f ull-power cont rol rod worths are generally less t han those f or cycle 1.

The cycle 2 ejected rod wortns are lower than t hose in cycle 1 f or t he same number of regulat ing banks inserted.

It is difficult to ceapare values between cycles or between rod patterns since neither t he rod pat terns f roir. which the CRA is assumed to be ejec ted aor t he isotopic distributions are identical. Calculated ejected rod worths an 2 their adherence to criteria are considered at all times in life.and at all power levels in t he development of the rod insertion limit s presented in section M.

The m.sximu:n stuck rod worths f or cycle 2 are less t han those in cycle 1.

The adequacy of t!w shutdown margin wit h cycle 2 stuck rod worths i s d eenns t r.a t ed in Table 5-2.

The following conservatisms were applied for the shutdown cal-

{

culatlona:

1.

Poison material depletion allowance.

l 2.

10% uncertainty on net red wo r t h.

j 3.

Flux redistribution pen.alty.

5-1 Babcock s. Walcox

Flux redist rihution was accuunted for since t he -hut down.s aa l v s t s wa. c a l. u-1.tted using.: two-d imensional rmdel. I tie shutdown c.sl. ul.st ien.it the end of cycle 2 is analyzed est approximately 226 EFPD.

T his is the latest t i: c (*

10 days) in vore life at w:.ich the t ransient bconk is nearly tully in-crte-J.

Atter 226 ETPD, the transient b.ank will be almost f ully withdrawn, thas inc re..s i ng the available shutdown margin. The reference fuct evele shut down : a r gi n i-pre cnted in the Oconce 1, 2, and 3 FSAR, T.able 3-5.

The cycle 2 power def icits from hot zero power to hot full power are: similar to but slightly higher th.an those f or cycle 1.

Doppler cocificients. raid e r.a t o r coefficients, and xenon worths are similar tor tt.e twe cycles.

The ditfcrenti.it boron worths for cycle 2 are lower than those f or cycle ! due tn depletion ot the fuel and the associated buildup of fission products. The eftestive d e l.s y-ad neutrun fractions f or bot h cycle's show.a decre. awe wit h ~ irnup.

$..?. Analvtical_ iny g The c nle 2 (acore measurceent calculat inn const. ant s u.cJ for comput ing core power distributiong were prepared in the same manner as for the ref er ent e cycle.

's. 1.

Ch.inics in Noelcar Desgn, There have been no relevant changes in core des ign between t he retcrence.ind reload cycles. The name c. tic ulat i.n.31 methods and design intornatten were used t o obt.ain t he important nuclear design parameters.

In addition, no *ignif icant operat ional procedural changes exist from the reference cycle with regard to

.extal or radial power shape cont rol. xenon cont rol, or t ilt control.

The op-erational limits (Technical Specif ic.st ions changes) for the reload c ycle.are shown in section 8.

i i

I 37 Babcock & Wilcox

Tabl.

~2-1.

Ocon e ) cycle 1 and 2 i t.ysics Parx sters C.clej Cycle 2 Cycle length.-EFPD t64 265 Cycle burnup..'Gd/mtt!

14.646 8.293 Average cure burnup - U)C. M'.41/raU 14.646 18.190

~

Initial core loading atU e2.1 82.1 Criti al - boron - fu)C. pp s (no Xe)

II/.P d#..all rods out 1.559 1.251 liZP. groupa 7.and 8 insert ed 1.410 1.103 ilfr.,troups 7.and 3 inserted 1.312 931 Critical boron - EDC Prm (eq Xc)

It?.P. all rods out

'. 2 9 325 if f P. group H (37.52 ud. eq Xct 134 29 Cont rol rew! wortha - IlFP("I, IWC, %?.k/k Croup 6

1. 5 ".

1.14 t.roup 7 0.99

0. '* 7 t.roup 4 (37.5% wd) 0.44 0.54 Cont rol real worths - I!FP EOC. %Ak/k Croup-7

1. 17 1.24 Group M (17.5: wd) 0.26 0.49 Man ejected rod worth - H2P. %?.k/k DC g

0.4%(#

0.66,"'

IAM:

0.72 0.FO' M.sx s t uc k ro.1 worth - HZP. !?.k/k

!> 4:

4.10

2. 10 IAM:

2. 6',

2.1x Pos.cr delicit. HZP to itFP, !!.k/ k roc

!.10 1.65 uk:

2.10

2. 10 Doppler coeti - fuc 10" (!k/k/*F)

,100 power (0 xe)

-1.51

-1.54 Doppler coet f - FAX. 10" (l.k/k/*F) 1001 p.mer (eq Xe)

-1.67

-1.54 Mtwierator roefI - IlFP.

10-'* ( *.k/ k /

F)

RM (o Xc. 1000 ripe. groups 7. 8 ins)

-0.23

- 1. 0+a FtM: (eg Xe. 17 ppm. group 8 in s)

-2.70

-2.39 Niron wort h - IlFP. ppa /%f.k/k IWC (1000 ppn) 99 107 HM: (17 ppa) 100 101 Xenon worth - HFP %?.k/k r.ak (4 days) 2.71 2.64 IW (equilihrtom) 2.65 2.66 5-3 Babcock & Wilcou i

e Table 5-1.

=(Cont'd)

' f:7C l 1

fycle 2 f

Effertive delayed neutron frktion - HFP B0C EOC-0.00690 0.00535 0.00514 0.00520 (a)HZP:

hot zero power; HFF: Hot full power.

(b) Ejected rod value for gew ps 5.6

7. and 8 inserted.

(c) Ejected rod value for groups 6, 7, and 8 inserted.

f l

i l

i 5-4 Babcock a witcom

Table 5-2.

$Wt down _M ergin Calquia t ion t or 0 o.we 3.jyfic j FC C_ Of 3.)

(

BOC. % *k/k tgg,i 1 a b 1c_ kod__Wo r t n Total rod vorth. HZP!>1 9.73 9.73 Warth reduction due to bu -

s, ral

-0.19

-0.30 poison material Max 1 ua atuck rod. H2P

-2.10

-2.IM net worth 7.29 7.25 s.cw 10% uncertainty

-0.71

-0.71 Total available worth 6.56 6.52 H,aters t red Hawl Wor t h Power deticit. hFP to HZP 1.65 2.30 M.ix.311ow.1 hic inserted rod wrth 1.22

1. 'M Flux redistribution O.40 1.tv)

Total required worth 3.27

'. 8 5 Sheitdown Margin Tot.it av.sil. wrt h - total rey. worth J.29

1. 6 *.

fiot et Required shutdown margin is 1.00

k/k.

(.4)For s.hu'down margin calculations, t

thig is def ined as approxir-stely

.! 26 EFPI). the latent time in core life at w:.ich the tr.instent

b. ink 13 iw.irly full-in.

(1.)II? P: hot zero power; itFP: hot full power.

5_s Bahrr.cka W;Icnv

Fl ;ure 5-1.

E. * : (4 EFPD). Cycle 2 Two-Dimensional Relat i.e Pever i

b i s t r iteo: 1..n -- Full Power. Equilibriun Xe.en. N.r 21 Rc.d Position- (Groups 7 and 6 Ins.rted) 8 9

10 11 12 13 14 15 i

H 1.30 1.19 1.15 1.11 1.18 0.83 0.78 0.63 7

K 1.19 1.33 1.30 1.18 1.22 0.55 0.70 0.62 8

L 1.15 1.30 1.12 1.09 1.12 0.96 1.00 0.55 M

1.11 1.18 1.09 1.34 1.31 1.23 0.97 8

N 1.18 1.22 1.12 1.31 1.12-1.14 0.70 7

0 0.83 0.55 0.96 1.28 1.14 0.77 P

0.78 0.70 1.00 0.97 0.70 R

O.63 0.62 0.55 Inserted Rod Creup No.

1.33 Relative Power Density 5-6 Babcock & Wilcox

8 h

. ti.

IttEM*.AL tlYDRAl'LIC DESICS Therm.al hydr.aulie tW agn Cilculations e.

! 5-- t w real-hydr.aulie de stca calculations for support of cycle 2 operatton i

ut t l a cd t he analyt ic.il : vet hwl dot u:::ent ed in references I and 3.

l'aese cal-sist a n. were ::ade to.a eown ter the introduction of the Nri B4.nssemblics

' %: n it, to consider the anni:aum actual re.*ctor coolant (RC) system flow ratr a. 9easured during tir-t cys le operat ion..and t o incorpo r.a t e the BLW-2 (rF..rrelation in pl.aee of 1 !.e previously used W-1 correl.atlon.

+. I.1.

Iptro,dyctton,ot, N rA re, A-we,: S,1,iy_s A cissu wed in hec t ion 4.1. t he Nrk tl4 assembly dif f ers f rom t he N r k 3 3 rr6,.itilv in t he design er t *>e nd t t t t ing.

This produce. a slight ly a-aller

1.% restitance for t he E. a-enblies.

Thus, introducing B4 asne:nblica into t'e sete causes a slight c hunce in the core f low dist ribut ion.

To obtain the

..c;e 2 sore slow distributton, the thermal-hydraulic model ut tl azed t he ac-t t.a ; % 54 I.*! B1 sont1_,uration with B1 assceblie. In the Lottest core l oc.a-tsoc.

6. l.'....!.n,c,ragsed RC,f v;< tyn,11Ow Re. actor coolant flow dat.: ebtainnt during cycle I eperat ion verit ied th.at the

-s -t e-i t low v.as great er t h.an - t he des ign i Iow.

The ceasured value was 1101 of the Je-agn tlow.

For the esele 2 therm.81-hydraulic design analysis, the system tla

-a. conservatively chosen to be 107.6% of design. The incorporation of tht-i > creased t low in the thernal-hydraulic calculations was accompani d b e

ya sorre-ponding increase in the core inlet t e:npe ra t u re frem 554 to 555.9F.

Such i

an ' t: c r e.a ne is necessary since the integrated control system chaintains an aver-age core t ropera t u re.

The increases in RC flow and inlet temper ature are-dun ge-in calculat tonal parameters only and do not represent c hanges in opera-t ien or t he plant.

t>- 1 Babcock & Wilcox

6.1. 1.

B&'.*-2 DNB Correlat ion The 85*='-2 DSB correlat lon, a realist ic pred act ion of tne burne ut phenocena.'+ '

haas t ven reviewed and approve *1 f or use with the.ars' B fuel anembly design.

In - the.epp:1 cat i6n of this correlat tan to t he Oconec 3, cycle 2 cure, two modi-l fications, which have also been app! icd to the Ocence 1 cycle 1. and Ocence 2 cycle 2 cores.- have. bcen inst ituted t i

l 1.

The liniting design DMBR of 1.30 was used.

This corresponds to. 95:

prob. ability at a 95 confidence level that DNB will not oc:ur.

2.

The pressure range applicable to t he correl.ition has been otendad downwar d f rca 2000 to 1750 psia.

j Both of t hese modit f eat ions have been approved by the NxC.*

The use of this c orre!.st ion in conjunction with increased system flow for the cycle 2 annivsis indicates t ha t the margin to DNB is greater for evele 2 than had been predicted for tiest-rore operation, as shown by the d.ita in Table 6-1.

6.2.

DNBR An.stvsts in addit ion to the it ema discu%ed.above, t he DNER arsalysis f or cyr te J oper.4-tion considered maximum design condit ions.' as-built l uct as cmalv gron.*t ry,

and hot operat ing corsillion i.

Thii. analysis resulted in the hot ch.innel (batih 1 1.sel) einimus DNhR of 1.98.at 1127. power for undensitied fuel.

The DNnet cal-culations f or undenstiled f uel are based on a 144-snch act ive length.

The met heJ used t o determine the effects et dens &tication was the.ame as t h at described in reference 5.

Ih> wever, t he size of t he densif icat ion-induced gar and the nagnitude of the resultant power. pike were deternined on the basis on t he power spike model described in sect ion 4.3.1.

The values used f or th. se two para eters were 3.10 inches and 1.087 respect ively, which are conservative.

The shortened (densified) stack length used in the analysis w.as 141.12 inches.

Alt hough this is lenger t han the den sified stack length of t he t.iten 3 f uel i

(140.10 inches) t he gap size.ind power spike magnitude were large enough to l

give conservat ive t esultn. With this basis, the densification effect results in a 5.91% reduct ion in the minimum DNBR, as shown in Table 6-1.

( T.s h ic e>- 1 compares cycle 1 and 2 thermal-hydraulic design data.)

An analysis was performed with the COBRA 111-C code to deter =ine the efic(t of a fuel rod bowing into the hot channel and reducing its flow are.a.

The results demonst rat e t hat rod how of the magnitude predicted is.adequatelv tempensated 6-2 Babcock & Wilcox

for by the tiew.ere.e redus t ion t ac tor.

Rod bow.seav tros the hot channel was al so.an_al y d.

In tt.t= annlysis tr.e etteet ot.: peser sp de v.s 4JJed to tne hat red in t he a r e.i et the : ta.inun is hK.

This an.stv-ts also 1.

an s t r.a t ed t ha t tim.mtr.nt o.once 1. tvele 2 M.b K re

41 t s cenw rvatively a....n t ter t he vt-tect'. or tact rod bowing.

6. 1.

Pr. su r.- T. gra t u re !. ic a t f.v.a ! 4.a t t on Pressur.-terperature (P-T) l im i t s f or. ve l e 2 ope r.at ion we r e b.a scJ en t i.e in-t re.ii.e d Mt:

vstem flew (section 6.1.23, the limiting DNisa ot 1. 30 as deter sncJ by t he

.%'.-/ DNB correl.itson (sW t to. *.1. 3 6, and t he d..n-i f s c a t ion pe nal t s J.-

fined in -cition 6.2.

P-T linats tor crele 1 operat ien were he,sd on one vent valve ta tlin,c epen, which reduce-tk cml. ant fish tor heat re mv41 by

.C.

An NRC stait.!a cision ~ relie ved 1% t ram having to inc li.d e., vent valve flev penalty i n

.a t e t y an.a l)

e:..

This dec ts ton v.as ttise l on an ev.a lisat t..n a t oper.st-Ing d.s t.: t ron BW plant e.

in roninnet son wit h a v. -it v.il s e survet !!.ince pregr.ap.

Thereture, t he P-T limit-analynis for evele 2 op sati..n di.! not incl.sJe t he open vent v.eIse pen.al t y.

S. '..

Flun-t o-Flow Setfoint Fv slu it ion Ibc met hed et evalu.at ing t he Ilux-to-fle. *ctpoint f or cycle 2 oper.ition de-v aat ed f r.c previous setpoint an.slyses in t hat the power ver-us tine was in pis t direct ly t o t he t ransient DNBR ca lcul.nt ion.

During an RC pump coastdown. the tine to rvac t er t r ip, and hence t he power t r. ins tent, is dif f erent 2 o r coa c h

~c t -

point.

Therefore, the t echnique used as to. lect a eset point.o i deternins-t t.e power t r. ins tent ter Lt.

This la acces;'I nshed by f irst det ern in in g the flow f r.ac ion annori.et ed wit h t he f low-to-t lov act roint.ind the indicat ed power level (1021 of design).

The flow traction is t hen rel.it ed t o t he.sppropriste punp co.sutdown curve '

pump for Oconee 3) to obtain t he t iric.at whs h alw trip

.ignal occurs.

delay t ros t r ip signal t o st.n r t of control rod :wtion (1.5 is then aJded to determine t he t ime when the power start s on to d. crease.as a result of the r eac t o r ac r.im.

This power transtent.and the pump co.astdown ( t l.v transient) are input to the RADAR

code, and a tr.ansient DNBR is e.n lcu t.a t ed l

based on an assumed init101 peser level of 108:1 f ull power.

This procedure la c ompleted f or a second..and it necessary.

.a third setpoint.

The s.mallest DNBR occurring during e.ach transient. analysis is then plotted

.sgainst the corresponding netpoint.

Frem t his plot is determined she nininum g

Babcock & Wdcom

setpoint necessary to maintain a minimus DNBR greater than 1.30 throughout the transient.

In addition to the procedure just described, the Oconee 3 flux-to-flev setpoint evaluation also included the ineressed RC system flow (section 6.1.2), the liniting DNbk of 1.30 as de:crnined by the BW-2 DNS correlation (section 6.1. 3).

and the f uel densificat ion penalty defined in section 6.2.

The results indi-cate that a flux-co-flow setpoint of 1.10 would provide adequate protection for cycle 2 operation. This v:.lue. however, represents the thermal-hydraulic 11mit t t he actual netpoint specified. is lower to conservatively account for instrument errer and drif t. flow noise. and'the like.

i I

6-4 Babcock & Wilcox

iable 6-1.

Ovele 1 ind 2_ Maximum C.% f en Conditjf_ns Cvele 1 Cwcle 2 Design poser level, Wt 2%8 25bs Systen,,re-sure, ps i.:

2200 2200 kr.u tia c.>lant iIow, I design 100.0 107.b Ve.4el irlet/ outlet coolant temperature, 100%

pownr. F 554.0/603.s 555.9/602.2 Ref design ralial-local power pe.aa' ing tactor 1./b 1.73 kel design. axial flux su pe

1. 5 cosine

1. 5 cosine f(.it channel f.actorp.

Fnth.alpy rise I.011 1.011 elea t flux 1.014 1.014 Flow.ere.:

0.98 0.93 Active fuel length, in.

(T.ible 4-4)

( T.thl e

,- 4 )

Avg hr.it ;luu. 1000 power, Btu /h-tt-175,640 175,640 M.a u he.it ilux. 200% power, htu/h-it' 468,95u 468.95)

Cilf correl.it ion

's.'- l R W'- 2 Minimum DShk (* Power)

%> dennific.atinn pen.alticw 1.55 (1141) 1.98 (1122)

Full den =ific.stion pen.altieu 1.4R (1141) 1.86 (1123)

IdI Hasnt on densifleJ length of batch 3 f uel and hot t uel rod di.ameter.

(b) Baned on.sverage heat flux with reference peaktna.

6-5 Babcock & Wilcox l

7.

ACCIDELT AND TRANSIEhT ANAI.YSIS 7._1.

General Safety Analysis Each F5ARI accident analysis has tiern exaninni wit h respect to thanges in cycle 2 parameters to determine the eff ect s of the i.ych 2 r e t e.s d and t o cr:,ur e that therm.it performance is not degraded during hypothetical t ransient s.

The core thermal parameters used in the FSAR accident.inalysis were design op-erating values based on calculated values plus uncertainties.

Cycle ! v a l.e*

(FSAR v.alues! of core thermal parameters.are coep.ared wit h t hose used in t he cycle 2. analysis in Table 6-1.

The.e parameters are t o:nm to ill of t he-act i-dent analyses presented herein. For each accident of the FSAR

.a discu-wien.ind the key p.ar.wteters are provided. A comparison of t he key p.arameters (wer Table 7-1) i roes the FSAR and the present cycle 2 is provided wit h t he accident J a r.-

cus*lon to show that the init ial condit ions ot' the t r. ens ten t ire bounded by the FSAR anasymis.

The eftects of fuel densification on the FSAR accident results have been eval-uated and are reportal in BAW-1399.'

Since cycle 2 reload fuel assemblies con-tain fuel rods stese theoretical density is higher th.an those considered in reference 3. the conclusions derived in that reference are still valid.

Calcul.ational techniques.ind methods for cycle 2 analyses rema in con. t s t er.t wit h t tane used for the FSAR. Additional DNBR margin is shown ter cycle 2 be-cause t he B&W-2 CitF correlat ion was used in;te.id of t he W-3.

No new dose calculations were performed for the reload report.

The done con-siderations in the FSAR were based on m4ximum peaking and burnup for.all core cyclest therefore, the dose considerations are independent of the rele.aJ batch.

f

_7_. 2.

Rod Wit hdrawal Accident s l

This acetJent is defined as an uncontrolled reactivity addition to the core due to withdrawal of control rods during startup condit ions or f rom rated pcwer 7-1 Babcock s, VVilcox i

conditions.

bot h typew of inc ident s were analyzed in tl.c FSAR.

The 1 port. int para:neter s during a rod wit hdrawal accident are Doppler coe:!!c tent. nede r.s t a r tarperatur. coeftir ent, and the rate at which reactivity is.saard to t r.e core.

Only bl.h-presesre.ind high-t lux t r tFs.are account ed for in the F /.it an.s ! v m s..

whic h ignores t ult iple.il a rm.. interierks, and t rips that norsilly preclude 8his type of incident.

For posit ive react ivity addit ions indicative of the c event *, t he most severe result s occur f or 801, condit ion..

The FSAR v.alue of the key parameters for hol. condit ions were -1.17 - 10

(.*.k/ k /

F ) for the Dop-

~

pler eoetstetent. 0.5 10 * ' k / k i nr t he mode ra t o r t e:rpera t ur a cecificient

.and t od t,roup wor t h* up t o a il t r.c l ud i n g. 10

.~.k/ k rod bank wc r t h.

Coeparable cycle 2 p.irame t r ic va lue a are - 1. % = 10 ' (ak/k/*F) for the IA ;>pler coef f i-

~

Irf* (*k/k/*F) f or t he modt rator tsoperature coef f icient, and.i cient. -1.06 C3aximuni rod b.ank s or t h ot 9.N

1. k / k.

Theref ore, cyc!c 2 parameters are tusinded by de ign values a..amed for the FSAR nalysis.

Thum, f or t he rod i.it tuiraw.al trea, lent. t he con.ca.uefices will be no more *cvere t han t hose presented in the MAR.

For t he rod w it hd tawal t ran ra t ed power, t he t rann ient consequence *

.sre alno le os severe t han t iu.e pt senttd in t he densif icat ion report. I F. l.

%derator 1,ilut ton Ar. i. tent l'.o r on in t he f orm ut bo r i s' acid is utilized to control excess reac t ivit y.

The boron sontent of tio-reattor cool. ant is periodic.nlly reduced to compens.ite for f uct hornup and t ransient xenon ef f ect s wit h dilut ion water supplied by t he makeup anJ pur i t ic.it ten.vstm.

The avJerator dilut ion t ran%ientis cone.idered are the pumping of wat er wit h zero boroe concent rat ion f ro:n t he rLakeup t ank to t he RCS under condit ions at ful1-pewer operatlon, hot shutdown, and refueling.

The key p.arametere la thin.i i.a l y s i s are the initial boron concentration, boron react ivit y wort h, and ciodera t o r t.-xp. rat ure coef f icient for power casen.

For ;esit ive react ivit y addit ions of thi s type, t he most severe results occur f or 1W1. cond i t n on e. The FSAR v.ilues of the key parameter!s f or BOL condit ions werc !!.00 ppm f or t he init i.nl boron concentration. 75 ppm /12 (t.k/h) boron re-

~'

u t i v i t y wor t h.and

+0. W. - 10 ak/k/*F for the moderator temper.sture coctfi-

[

cie nt.

Compar.ahle cycle 2 values are 931 pra for the init ial boron concent r.it ioa. M2 p;well: (*k/k) boron reactivity worth and -1.06 - 10" (?.k/ k ) /

F f or t he mode ra-t or t emper.at ure coctficient.

The FSAR shows that the core and RCS are adequ. ate-ly prot ect ed during t his event. Sufficient time f or operator act ion to l

7-2 Babcock s. Wilcox

terminate t his tran.wient is also shown in t he FSAR, even wit h caximum dilut ion ani einica.= shutdc.m r argin.

The predisted cycle.' parar.trie v. slues of in-portance to t he c.oderator dilut ion t ransient.a re baunded hv the FSAR desagn

v. slues: t ru s, the analysis in the FSAR in valid.

7.1 Cold ' etat er Ji* trap St art up) Acc ident There are no chec k or isolat ion valves in the reactor s oolant piping; there-fore, tt' clamaic cold w. ster accident is not rossible. Hawever, when t he-re-in operated with one or more pumps not running, and then thesc.are turned actor on, t he increased f low r. ate will cause the average core t empe r.s t ure t o dec rea r.e.

I f t he =vderator t a1sper.ature coef f ic ient is negative, t hen reas t avi t y will be added to tN core.and a power riac will occur.

Protecttve ants rlocks and procedures prevent s t.a r t ing idle pumps if the re ar-tor Iover is above 222.

How=:ver, t hese restric t ions were ignored, and two-pump.t.artup t rom 50% power was analyzed as the most severe t r.insient.

To n.axintre react ivity addit ion, the FSAR analysis assumed the monit nec.at ive moderator t ruperature coef ficient of -3.0 - 10 (?.k/k)/*F and tim l e.s s t neg.i-tive Doppler cceif icient of -1.30 = 10".*.k/L.

T hs-corres. pond ing most neg.st Ive moder.itor tesperature coeffleient and le.ast neg.atIve Doppler cocificient pre-d ic t ed f or c yc l e 2 a re -2. %

10" and - 1. 54 - 10 " (?.k/k)/*F respecttvely.

Since t he cred'.cted cycle 2 moder.stor temperature coef t icient is less neg.atave and t he Doppler coef f iclent is more negat ive t han t he v.aiucs used in t be-FSAR, the transicat result s would be less severa cl6an those report esi in the FSAR.

7. 5.

Lo.s of Coolant Flow The re.ac tor coolant flow rate decreases if one or cure of the re. actor coolant pump 6 f.n i t.

A pumping f ailure can be caused by mechanic.al f ailure or loss of el ctric.al power.

Wit h four independent pumps available, a mechanical

illure la one pusp; sill not affect the operation of others. Wit h t he reactor.a t ;iowe r,

tieeffectdf loss of cool. ant flow is a rapid increase in coolant t er.:pe r.a t u r e due t o t he reduction of heat removal capabilit y.

This incre.ase could renult in DNB if corrective action were not taken imacdi.itely.

The key parameters for tour-pump coastdown or a locked-rotor incident are the flow rate, tlow coa s t down c h.a rac t e r t at ic s, Dopple r coef ficient, moderator temperature coef f t-cient, and hot clunnel DSB peaking factors. The most on s e r v.a t iv e init a.sl condit ions were assumed for the densif is-it ion report 3: FSAR values of flow 7-3 Babcock a. Welcox

.ind. o.is t d..a. -1.17

.f ' ' ;',>/ r %ppler

a. :

ii icnt. *...i If*O'

'T rodir. ster t.v.vratart. e.

i.:ent. vi t t> d. n. : 4e1 :'.e! r. v.

pne s~

pe.s k-Ing.

Ihe r.

Iis sneued t b it t 6.e MSP re nin.d ab,ve :.6 8%-5) *as t h.

.or-pu -p c mt d. c... a n ! t i..- t ue !

.1.i t.a i n.: t e,irr.it sr.

r.

.iined b. Icv tritsrL linitw ti r....

I.ustd-rctnr tr.c.

4.-et.

The pre dic t.4 ;> ara'a t t it s.s t :4c, tor e l e.!.o re - 1. %

10' t.k / k J / r A ; ; ;'. r 10"' ('k/k)/*r -d er.stor te=per.ature co.Itictent..

.L'

.... i

1. i. n t. -1.86 peaking I.ac t u r.

.v. shewn in Iahle b-1.

Since the ; s ed i. t ed c yi I c.' v.i !.a - are bounded by t ' c c u sed in the densliicat in refert, tnc :esulta o! that ana1y-in represent t he-mi.t severe censequences rom a loss-of-flow incident.

1,.A.,, $t escyd *2. t Sturb-In or j ofped Cnntrol R_od 2

u if.a control rod were dropped into the core while it was operating, a r.apid de c rease in cutron power would occur.secompanied by a decre.a.c in the core

, eve ra ga a in> l. int temperature. The power dist rihist ion night be distorted due to a new control rod p.at tern, under which condition..a return to full power night Ic.id t o localized power densit ics and heat flumeu in exce.s ut denign l i m i t.e-tsons.

The key p.iraseters f or t hlm t r. ins t ent.are moderator temper.iture coefficient.

droppe.1 rod wr t h, and loc.sl peaking factors. The FSAR analv.is w. ins based on

u. f.4 and u. !M *.k/k red wor t hs wit h a moderat or t emperat ure coef ficient of 10' ' (th/k)/*F.

Fo r cyc l e 2. t he m.ix irrum wor t h rod a t power lu O. ?8C

- 1. 0

' k/k.ind a rWerator temper.iture coef t'icient of -2. 39 = 10 ' (f.k / k ) /

P.

Since t he p red ic t ed r od wo r t h i s less posit ive and t he nderator ts eperat ure coet t i-cient is more po.ittve, the consequences of this t r.nns i ent are les6 severe t h.ac. t he re.ul t 6 presented in the FSAR.

7. 7 l.o s o r I:!cetric Power two types of p wcr lossen were considered in the FSAR: (1) a lo w-of-lo.ed condition s.susid by sep.tration of the unit f ro:s t he t ransni ns ton sy. item and (2).i hypothetical condition resulting in a complete loss of.311 a.ystem and l

l unit power except t h.a t f rom t he unit batteries.

The FSAR analvasi:4 ev.slu.sted the loss of load with and wit hout turbine runb.stk.

When there is no runback, a re. actor t rip occurs on high reactor coolant p r e-.-

sure or t emperat ure.

This case results in a non-limiting a cident.

~he largest 'of t ait e done occurs f or t he second case.

1.c.,

loss et all electrical g

7,,

Babcock 4. Wilcox

power except unit batteries, and assuming eperation with failed f uel and steam generator tube l ea kage.

These results are i nd ependent et core laadang; there-fore, the results of the FSAR are a;placaole for any reload.

1 8.

S t rag,L,l_n c Failure A steam line f ailure in definet as a rupeste of any of the steam Itne, trum the.stoaa generators. Upon tat t iat ion of tre rupture, both stean generators stcrt to blow down. causing a v dden decrease in t ne primary syst em te=pera-ture, pressure, and pressurizer level. The tenperature reduction leads to positive reactivity insertion, and the reactor trips on high flux cr low 3C pressure.

T he FS AR ha s identif ted a double-ended rupture of the steam line between the stean generator and steam stop valve as the worst-rase situation at end-of-life conditions.

The key parameter for the core response is the modcrator temperature coef fi-c i ent.. wh i c h wa s a s sumed in t he FSAA t o be -3.0 = 10 ' (?.s /k ) /

F.

~

The cycle 2 predicted value of moderator temperature coe'ficient is -2.39 a 10 ' (f.k/k)/

F.

This value is bounded hy those used i.. the FSAR analysis; hence, the re-mults in the FSAM represent the worst situetion.

79 S t e.ca Genera t o r Tube Fa i lu re 3

A rupture or leak in a steam generator tube allows reactor toolant and associ-ated activity to pass to the necendary system.

The FSAR analysin is based on complete severance of a steam gecerator tuSe.

The prinary concern for this incident in the potential radiological release, whie.h is ir.dependen t os core loading. Ifence, the FSAR results are applicable to this rtioad.

7.10 Fuel _Itandlin LAccident The mechanical da ur accident is considered the maximum potential source of activity release during fuel haraling activities.

The primary concern is

~

radiological releases that are imdependent of core loading: therefore. the FSAH results are applicable to all reloads.

i.,1,1.

Rod Fjectior. Accident For reactivi'y to be added to the core more rapidly than by uneentrolled rod wit hdrawal, physical f ailure of a pressure barrier component in the control rod drive assembly must Such a tailare could cause a pressure dif f eren-occur.

tial to act on a control rod assembly and rapidly eject the assembly fren the 75 Gabcock & VVilcox

core region.

Th!i inciden, Qresents the most rapid reactivit y insertion that can be re.asonably postulated. The values used in the TSAR.ind densification rea.r t at BOL conditions. -1.17 - 10 ~ U.k/k)/'F Doppler coef ficient. +0.5 -

10 ' Uk/k)/*F anderator ter perat'.re coef ficient. and an ej ec t ed rod wort h of 0.65% !.k/k represent the t.ixl: um possiht e t r.ans i en t. The corresponding cycle 2 p.o r a me t r i c v.a l ue s o f - 1. 54 - 10" (ak/k)/*F Doppler. -1.06 10 (ak/k)/*F moderator t enperature coef f icient (both more negative than those used ist refer-ence 4) and a maximum predit ted ejected rod worth of O.19% ak/k ensure that I

the ren.It s will he less severe tiun t!= se presented in the FSAR and the den-sification report l.

7,.,l_2. __ fia x i vmsa Hypo t_he t i ca lJ,c c i d e n :

There is no postulated r.cchanism whereby this accident can occur since it would require a multitude of f.a l lu re in the engineered n.ifeguards.

The hypothetical accident las b cd ;:olely on a gross release of radioact ivity to the reacts.r building. The consequentta of this accident.a r e independent of core loading; hence. t he result s repor ted in t he FSAR are.applicabic f or all relo.eds.

7.11.

k'aste Gas Tank Ruytury The wa.ite g.is t. ink v.ar..nssumed to contain the gaseous activity evolved from deg.assin,t all of the reactor coolant following operation with 1*. defective fuel.

Rupture of the tank would re ult in the release of it s radioactive contents to t he pl. int ventilattun system.snd to the atmosphere through the unit vent.

The consequences ui thia incident.are independent of core loading; therefore, the resulte, reported in the FSAR.are applicable to.any relo.ad.

7.14 10CA Analvsis A generic h0CA. analysis los been pertormed for the BW 17 7-FA. lowered-loop hSS using the Final Acceptance Cr iteria ECCS evaluation model. I' The analysis is generic since the limiting v.:!ues of k / parameters for all plants in this c.itegory were used.

Fu r t hertair e.

the combination of average fuel temperatiare I

a. a function of linear he.at r.ite and the lif etime pin pressure dat.: esed in this 10CA limit s analys ts is coinservative compared to those c.stculate d for this r e t o.a d.

Thus, the analysis.ind the LOCA limits reported in referenecII provide conservat ive result s f or the operat ton oi Uconee 3. Cycle 2 fuel.

j The fol!owing t.abulation shows t he bounding values f or allowable LOCA pc.ak i

linear heat rates for Oes.ee

1. Cycle 2 fuel.

7-6 Babcock s.Wilcox

Allowable pear. linear Core elevation d t hc.a t rate, k4/ft 2

15.5 4

16.6 6

18.0 8

17.0 10 16.0 Table 7-1.

Cotaparisor. of Key Parazeters for Accident Analysis FSAR denai!1.-!

Predtctcd Parameter va li.e cyf_le_2 value thL Doppler coef f.10 ' (f.k/k)/*F

- 1.17 * 3

-1.54 D)L Doppler coef f.10'k (f.k/ k ) /

F

-1.13

-1.54 Bot moder.itor.oeff. 10 (t.k/k)/'F

+ 0. i(b)

, g,,,,

ICL moderator coef f. 10 ' (t.k/k) /'F

- 3. 0

-2. W All rod b.ank worth (ilZP), I t.k/k 10.0 9.s init i.it borun cone (liFP). pra 1400 911 Poron reactivity worth (70F). pps/12

'k/k 75 a2 M.tx ejceted rod worth (ilFP). I t.k/k O.65 O. 19 Drepped rod worth (ifFP). : /.k/k O.46 0.29 IdI(-1.2 10" ?.k/k/F) was used for stea a line f ailure an l 10" ? k/k/F) was uwd f or cold water analynis.

a ysis;

(- 1.1 (b)(+0.94

~

= 10 ' "n/k/F) was used f er the coderator dilut ion accident.

i I.

l i

i 1

y_7 Babcock & Wacox

8 PR0 POSED MODIFICATIONS lu TECHNICAL SPLCIFICATIONS The Technic.al Specifications have been reviwd fer cycle 2 operation. Changes the results of the following:

were 1.

L't.ing t he B4W-2 CHF correlat ion rather th.an W-J as 6.1.

fiscussed in section 2.

L' sing a 95/95 confidence Icvel rather th.an 99/95, as discussed in sect ion 6.1. 3.

3.

L's i ng 107.67, of design f low rather t h.an Ir47, as di.scussed in.ect ion 6.1.

4 t*skng the Fir.a t Acceptance Cr it er !.a I.OCA Acalys t.4 in-rest rict ing peaks during operat ion, as discus

sed in ect ion 7.14 5.

Revising the.assumpt ions on which t he f lux-f lew RpS r.et point is b.ased.

This setroint now accounts for signal noise and 1Iow :::casure ment error on the b. asis of data accumulated f rom operating !!f.W re.actora.

6 An an.alysis incorporat ne, the effects of t uel rod bev on core-parameters.

The rod ivow evaluation w.as performed using the methods and proced arcs de-scribed in reterence 12 wit h the power spake calculated so that on t least 95* of the fuel rods will not exceed a givet power spike factor at a 9 57.

confidence level.)

7.

The penalty on core coolant flow d'ic to an asmumed cren ver.t valve h.+n been elininated b.ased on a vent valve 3ury.-allance progra:: perform.ed during each refueling shutdown.

Base 1 on t he Tecnnic.s! Specif icatiossa derived t rom the an.aly ses pleasert ed in thi6 rscort, the Final Accept..nce Crit t r ia Et'CS limit s will not be exceeded, nor will the th rm.il destyn criteria be violatec.

Figures 8-1 through 8-14 tilustrate revisions to previous Technical Specificat ion tafet y li:::its.

i r

l 1

g.t Babcock % Wilcox

i i

Fi gure 8-1.

Oc:mee 3. Cycle 2 - Core Protection saf tv Limits Of@)-

P t

1400

.u ac.

i 2200 t

ir t

b i

2000 4

J-d-

i 4

InM 1600 f

e 500

.580 600 620 no e60 Reactor Outlet Tempen at ure. F 4

e f

I Babcock & Wilcox.

l s.2 t.

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

Fi g re B-2.

Oconee 3. Cyc!c 2 -- Core r' rote ct ion Sa f er s 1.iaits

- 120 l-_.? l, I!?)

(1^ 1.

112)

Acceptabic kl 4-Pt.m

(-40 1*F1)

- - 100 0;e r.a r ion 41, 1M)

(

.'J, M6.4)

<*o.a.

86.46

-- gn i

(-4fl. 74.4)

Acceptable (4 ). 74.4) 3 & 4 Pump operation

(.2 L sR.9)

- - 6n (30.8 58.9)

CD

(-40. 45.4s-Acceptable

'(41, 46.9) 2 3 6 4 Punp Operation 40

- 20 r

f f

-bo

-O

-20

)

20 4 (,

en Reactor Power Irbalance.

Curve

\\

L actor Coolant Flow (1b/h) l 1

t 141.3 106 2

t 105.6 = 10*'

)

3 69.3 = 106 8-3 l

I I

Figure 6-3.

neonee 1. Cyc l e 2 - Co re P ro t e c t ion Sa f e t y I.a mit s 2600 2400

~

DD O

c:

5 y

2 2

2200 -

I t

.~,

.Iy 2000 -

c I.C I N)0 -

16 Cal r

t n

SN) 580 600 620 640 6*30 Re.netor that tet Temperature. F React >r coolant flov Cu rve (15s/h)

Power Pumps operating (type of liat t )

1 141.3 10 (1002) 112:

Four pumps (DNER limit) l 2

105.6 IG' t 74. 7%)

86.4%

Three puesps (DNBR limit) 3 69.3 = 10' (49.0'.)

58.9%

One pump in each loop (quality j

limit) g4 Babcock & Wilcox

Figure 8-4 Oconee 3. Cycle 2 - Protect ive System htximm Allowable Setpoints 2400 P = 2355 psig T = 619 F 2100 2200 w

7c.

i 5

Acceptable z

2I00 Operatton c.

e

.2c

,C L'

2

.~

2000 v

N c,

5 Ac l*na r rep t.nb l e*

g e-Operation 1900 e

4 I

P9ff 1800 (587.5) l 1

9 t

540 560 550 600 620 640 i

Reactor Outlet Temperature. F a_s Babcock 4. Wilcor

l l

Figure 8-5.

Oconce 3, Cycle 2 - Protective System P.eximum Allowable Setpoints Power level. *

- -120

(-11, 107)

(18._107)

+

Four Pu:np y Setpoint 3.o "

- - 100

(-28. 91) +.,

( 0. 40)

Three Pump Setpoints no 11, 19.9) (18, 79.9,

(-28, 65.9)i

>(10. 62.9)

Two Pump

~

Setpointn (18, 52.4,

(-28, 38.4)<

~~

>( 10 15.4) i 3

'o c

c

-6 E

\\

c m

m e

i t

O

-60

-40

.' O O

.' O 40 on Power imbalance. *,

For two pumps in one loop, the flux-flow setpoint must be 0.961.

a_t Babcock a. Wilcox I

Figure S-6.

Oconee 3. Cycle 2 - Rod Position Limits for Tour-Pump Operation From 0 to 115 (*10) EFPD gy) 170. 1020 C202.5. 102 170. 91

( '202.5. 91 """' l'"'I 161. 85 C"'"II go Festricted

O Region Res t ri ct ed f

70 Region 2

?25.8. 65 100, 64 5

y) 129. h2 O

e 9) e, u

~

Pernissible

)

.-3 Opersting 30 -

Region 20 -

10 -

0. 0 0:

i n

e i

i 0 20 40 60 60 100 120 140 160 180 200 220 240 260 260 300 Rod Index.

withdrawn 0

25 50 75 100 0

25 50 75 100 E-8 I

I i

i i

f Croup 5 e

a Croup 7 0

25 50 75 100 i

i f

f I

Croup 6 l

8-7 Babcock a. Wilcox

Figure 8-7.

Oconee 1. Cvele 2 - Red Posi t ion Limi t s fo r Four-Pun 9per.it ion Froci 115 (- 10) EFPD to 226 (?!O)

Ell *D 115.102 170, 102 100- operation in this 9'

C,209.4 102 Region is Not 90.

Allowed 170, 914 4 309,4, 91 Power ievel go-Shutdown 15, 85 C"**II Kirgin -

f-Festrirted IO-Limit Restricted Region ac Region k

129, 62

"- " q 60 225.8, 65 o

50-gg, gn 40-Pe rmi nnible Ope ra t i ng Re gion it;-

'*el 0,

15 1(1-O e

i i __

n i

e s

e O

20 40 60 Ni 100 129 140 160 180 200 220 240 260 260 10 0 Rod Index, % withdrawn 0

2,5 5,0 75 1,00 o

j5 50 75 1,00 Group 5 Group 7 0

25 50 75 100 1

I i

i I

Group 6 l

Rakenebe w:t -

t

i y

figure 6-8.

Oconce 3. Cyc!c 2 - Rod Position Limits fo r Fou r-

~~

Punp operation Af t er 226 (* 10) EF?D Operatton in this 140 102 255 102 gng Region is Not AlIowed 90 -

Power f. eve 1 outoff 240, 91 80 -

Restricted h-Region Shutdown 79 Nrgin T.

I. l a i t 96 67 225.8. 68 s0 54, 50 50 j

!. 0 Permissible 4

Operating 3g Region 20 ~

4 I' 10 -

O. O e

E

'o ed d ig)3 gjo lets Ib ;{xi j$3 j[g do--.', g o

16o

,gg Rod Index. % withdrawn 2,5 50 75 100 0

25 50 75 100 g

I f

f g

g g

Croup 5 a;roup 7 0

25 50 75 trxi a

f i

f g

Group 6 heh S M k OF

Figure 8-9 Oconec 3. Cycle 2 - Rod Fosition Limits for Two-and Three-Pump Operation From 0 to 115 (?10) EFPD Restricted Region 114 102 1 4,102

'.102 Restricted 100 - for 2 and 3 Pump Region for e3 operatton 90

%,5 3 Pump U

Operation c

2

'26.83 300,82 a

o*

c MD

U 129,79 3

a e

g in c.

(

u 60

3 I

w 2

Permissibic Operating Region

.c

]

40 2

10 E

30

~

f 10 c

t I

f a

f f

8 e

f f

s 20 40 60 60 100 120 140 160 1A0 200 220 240 260 280 300 Rod Index, % withdrawn 0

25 50 75 100 0

25 50 75 100 a

s e

a e

i l

i I

f Croup 5 Group 7 0

25 50 75 100 m

i I

I e

Group 6 r.

l

Figure 8-10.

Oconee 3. Cycle 2 - Rod Posit:fon Limits for h and Three-Pussp Operation From 115 (110) to 226 (210)-EFPD Operation in this 115 102 46.102 218 02 Restricted 9

100 Region is Not Region for Allowed

Pump S

99 Operation I

%c' c

80 Shutdowfi e

n, N rgin g 79. 79 226.Sc c

300.82 y

70 - Listt 9 o' r.

u s'

S q

Permissible Operating Region 60 4 e u

y r

U~

3 50 8.50 g i

o[

40 g

- <j' 30 p5 22 N

n,,. '

20 g

15 g

If'-

n.q O

20 40 e>0 80 100 120 140 160 180 200 220 240 260 260 300 stod Index. % withdrawn G

25 50 75 100 0

25 5,0 75 100 i

i i

Group 5 Group 7 0

25 50 75 100 a

i i

e 1

Group 6 l

1 1

8-11 Babcock s. Wilcox

)

i l

Figure 8-11.

Oconee 3. Cycle 2 - Rod Position Limits for heo-and Threc-Pump Operation After 226 (110) ETPD 140.102

'33 c

100

@ ration in this Restricted Region k 102 3

Region is Not for 3 Fump 3

0-Allowed Operation 1

226.87 w

80 Shutdown Margin 20.85

[

t.init 2

k 70-u" 60 Permissible Operating Region U

w 50-54.5 e,

2 f

40.

30.

').

o 30 20.

w 5

R

% entricted for 2 and go,e 2

3 Pump Oper.ation

0. 0,

()

20 40 60 80 100 120 140 160 180 200 220 2(.D 260 280 300 Rod Index. % withdrawn 0

25 50 75 100 0

25 50 75 100 I

I a

f e

a i

I t

i Group 5 Croup 7 0

25 50 75 100 t

I f

f a

Group 6 i

i l

n. a..... - -.

l Figure 8-12.

Oconee 3. Cycle 2 - Operatlonal Power labalance Envelope for Operation From 0 te !!$ 610) FIPD Power. % of 2568 We Restricted Region

- -110

-7.96.102 16.83.102 n

100

-7.78.91 3

I*

--90

-9.67.8 320.12. %

. 80

- -70

-23.64.64(

- -60 i

Permissible Operating

- -50 Region

-40

- -30

- -20

- -10 g

3 3

e f

I i

I O 40

-30

-20

-10 0

10 20 30 40 50

~ Axial Power Imbalance. 2 1

l 3 13 R h bow:s-i

l l

Figure 6-13.

Oconee 3. Cycle 2 - Operational Pa er Imtsalance Envelope for Operation From 115 (:10) to 226 (:10) EFPD Power, % of 2568 Wt Restricteo Region

-16.47.102(;

- 100

) 16.83.102

-15.72.91' U 90 t )16.83.91

-16.94,8 q)I7.00.85

- -80 70

-23.64.64

- - 60 Pernisnible operating 50 Regaon

)

- - 40

- - 10

- - 20

- - 10

~

L f

f I

I i

i f

_g g

-50

-40

-30

-20

-10 0

10 20 30 40 50 Axial Power Imbalance. %

i i

n.u 8

8-a-- 8 * "* ' - - -

Fimre B-14 Oconee 3. Cycle 2 - <perational Power Inhalance E.welope for Operati.~a Af ter 226 (?10) EFPD Power, 2 of 2548.Wt Rentricted Region

-25.5.102 -

16.71,102

-27.86,91 t)

- 90

>23.96.91

-80

-70

-60 Permiusible seperating

,,$n Elegion

- -40

-30 20

- -10 i

e i

f I

t

-50

-40

-10

-20

-10 0 _

e t

_e 10 20 30 40 50 Axial Powc: Imbala:we, I i

l

\\

f!-15 a'--*-"-'

2 9.

STMTUP PROGRAM - PHYSICS TESTING The p'anned startup 'esting associated with core performance is outlined be-low.

These tests verify that core performance is within the assumptions of the s.af orty analysis and provide the necessary data f or continued safe plant operatfon.

P_re-Crit f ea t Test s 1.

Control rod drive trip time Zero Power Teste 1.

Cr it ical teron concentration 2.

Tassperature reactivity coefficient 1.

Control red group worth 4

Ejected rod worth Power fests 1.

C.re puwer distribution verification at approximately 40, 75, and 100% FP, normal control rod group configuration.

2.

Incore/out-of-core detector imbalance correlatton verification at approxi-mately 75 FP.

3.

Power Doppler reactivity coefficient at approminately 100 FP.

6 Temperature reactivity coefficient at approximately lOOZ F1 91 Babcock & Wilcox

REFERENCES I

Ocowe Nuclear Station, l' nits 1, 2, and 3. Fini Salety A5alysit. Neport,

' Docket hs. 5tb269, 50-270, 50-287 A. F. J. Tekcet. H. W. Wilson, and K. E. Yoon Progras to Determine In-Reas ter rett ormawe of B&W Fuels - Cladding Crevp Collapse, B:W-100MPA, Bah (eck & Wilcox, January 1975.

Oconn 3 Fuct Densification Report, HAW-1299, L.bcock & Wilcox, Nov mber 1973.

C. D. Nrgan and H. S. Kao, TAFY - Fuel PJn Tm;nerature and Cas Pressure Anal w is, ILW-1CK%&, Babcock & Wilcox, May 1970.

Fuel ;%csification Report, BAW-10054. Rev 2 (i*retrictary) and B:W-10055, Rgv_1 thn-proprietary), B.abcock & Wilcox. June 973.

Correlat ten of Critical Heat Flux in Bundle Ceoled by Pressurized Water, RAW-lWWM, Babcock & Wilcox, May 1976.

Correlatton et Crit teal lleat Flux in Bundle Ceole.1 by Pressurized Water RAW-lWU6, Babcock & Wilcox, February 1972.

A. Sebn.cncer (NRC) to K. E. Suhrke (B&W), Let ter on review of "MW operat-ing Experience of Reactor Internals Vent Valves.* hva ber 1975.

B&W Oprat ing Esperience of. Reactor Internals Test Yalves. Babcock & Wilcox, August 1475.

IO RADAR - Reactor Thermal and liydraulle Analysis Dr ring Reactor Flow Coast-down, MW-100g, July 1973.

II ECCS Analv41s et B&W a 177-Fuel Assembly, l.oversloop.NSS, BAW-1010,1,,

Rev 2. Kabcock & Wilcox, April 1976.

I *'

W. O. Tarker fl\\ike Power Co.) to B. C. Rusche tXE ), l.et t er, Fe bruary 2 7, 1976.

I t

A_1 Babcock & Wilcox

..@vwh

l t

DOCKET NO. SO - 2. E 9 DATE: [tj y Z. (,, f 97 b NOTE TO wge AND/OR IDCAL PUBLIC DOCL' MENT ROO*iS

the following item submitted with Ictter dated [d 1 2Ll. 19 7 b A

from Od [4 Fo e Com map 4

..is. hein' ' withheld from g

s public disclosure', pending review, in accordance with' Section 2.790.

PROPRIETARY INFORMATION P-.c..+ -

c.., Is *3 TEi:,c~l. %e,1,wlaix R w i3io ns.

1:

~

=.

\\

a if N Y.$

Regulatory File Room 4

O

END MICR0 PHOTOGRAPHERs,s_-

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f%

~ i( \\5:;/ 'd

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MICROFILM SECTION MAVY PUBLICATIONS AND PReNisNC SE RviCE Creact SUILDeNC 15? 2. WA5848NG TON N AV V T Alto usa $wsNGTON O C 20374

..

ML19308B546

Text

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