Text

Cycle 2 Reload Rept.
	ML19319A780
Person / Time
Site:	Oconee
Issue date:	09/30/1974
From:	; BABCOCK & WILCOX CO.
To:	;
References
	BAW-1409, NUDOCS 7912120579
	Download: ML19319A780 (31)
	v • d • e

,

D*Whexy FNqt g BAW-1409 Y)

~

4 September 1974

~ ,

e sepea y

i

g m

p ,

~%

go -?df OCONEE 1, CYCLE 2

- Reload Report -

~

11 J Snee ?

.D Y0i}/ 'h

' b "

Babcock & Wilcox 9988

- '"*'""*F

a.W-1 ?.0 9 Se p te=b er 19 74 OCONEE 1. CYCLE 2

- Reload Report -

BABCOCK & WILCOX Power Generation Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 2:*50 5 Babcock s.Wilcox

. . ~ . . . . .

ny E y r-

+

4W.T.l.*;T.

S l' age

~. L l'iTRtWC110 ; A!.3 R"!ARY . . . . . . . , . . .. . . . . .. 1-1 Li i.. ..r. s - .in.

..S. i t....

2.I. IHI r.tdL&6* t 1 *n . . . . . . . . . . . .. .. . . . . .. 2-I

..J. Descripti.m . . . . . . . . . . . ... . . . . . .. 2-1

..I. L.. ore l,iev sics . . . . . . . . . . . ..... . . . ..

.'.'.. 6.rs i .- i.' . A .t . t. - . 1%c l . . . . . .. . . . . .. 2- 1

-L. l'iRFiiR :ANCE: X.ALYSIS . . . . '. . . . . . .. . . . . . . .. 1-1

' s. t . Therrait-Ind raulie Analysin . . . . .. ... . . . .. 5-1 1.1.1. IA si::t Linitati*nn . . . . . . . . . . . . .. 1-1 l.l.J. P. w r 8 pike ::.idel . . . . . . . .. . . . . .. 3-1 1.1. l . fr.SR Ai.ilysis . . . . . . ... . . . . . .. 1-2 l .1. '. . Fi.c i 1.cpe r.it ie res . . . . . .. . . . . . . .. 1-2 1.1. > . sit =: .o ry . . . . . . . . . .. ... . . . .. 1-3 1.1. !;utle.ir An.nlysis . . . . . . . . . . . . . . . . . .. 1- 1

l. l. Safety An.41ysis . . . . . . . . . .. .. . . . . .. 1 '.

5-I.

~

4.1.1. General Satrty Analysi, . . . .. . . . . ..

3.1.2. IsCA Atalysi.s . . . . . . .. .. . . . . .. 1 '.

l . '. . !! eels.m i. .s i An.a l ys i m . . . . . . . . . .. . . . . .. 1-3

- .. C in. lia slons . . . . . . . . . . . . . . . . .. . . . . .. '. - !

3. F.eferences . . . . . . . . . . . . . . . . . ... . . . .. 5-1 1.ist of Tables Tahle J-1. -Cvete 2 Finel Det. . . . . . . . . . . . .. . . . . ... 2 '.

2-J. tic.ince 1 4:.s le . 11.v s ! .

l'.ir eme t e rs . .. . . . . . ... 1 *>

c :. 8:rit . ..va *: .r c . e e , ia l .it ii.n , is ... 1. . .;< / . . ... J - e.

5- l . o;,c rat ine c. nai t i.-i . . . . . . . . . . . .. . . . ... -6

. i g, _ Babcock s. Wilcox

- 4 q k -

~wv+ -, ,wp w , r, w ws 4 y y y 4 m- g- e <-w----a y v ~

7 List of F igu,ryp '

l'. l g e riL are l.. r. ri 1.o .: t a.: Di.e.:ra . .............. 2-7

.' - ! . .. / *

' Ovile - -

.-2. '

Cent ro l koJ ' I.... a t i..u . ....... . . . -

.'- 1 04C, Cycle 2 Iwo-liicon..lonal Relative P wer 5

D:-t r Ibist i. n ' roli P.+er. i.e;utIibr1. . Xenon.

.ormal Rod Positions (Croups: 7 and h Inserted) . . . . .. 2 't l-7 1-1.- M.axir:um Cap Si/c' '.'s Axial Posit ion . . . . . . ... . ..

i-J . Po or 'ipike Factor Vs Axial. Position . . . . . . . . . .. 3- %

1-9 c a'ety-Limits ..............

1- 3. Core Protection sip. rat ion.a t Powen abalante Envelope . .. . . . . . . .. 3-10 t *. .

b 3. Cent rol Red Grcup Withdr.swal Lirnit . s er Four-Pu .p

.............. 1-11 4 0;eratten . . . . . . . . . .

1-6 Control Rod Group Withdrawal Limits for Feur-Pur.p Operation . . . . . .............. 3-12

- 7. Control Rod Group Withdrawal Limits for

.Thr.e- and Two-Pump Operatlon. .............. 3-13 Altowable LOC \ kWitt Limits . .............. 1-l '.

- l- 8 i

__gy _ Babcock s.Wilcox l l

m-

1. I:.7R0-)CCTio5 AND SD0t\RY i h i .4 repert describes t:1e basis to justify operation of the Ocence

%cic ar Stasien, tat t 1. Cymie 2 at the rated core power of 2368 .Vit.

f r.e Luevd herein are the ressired analyses including fuel densification et f ect s .!nd reviaed ECCS (FAC) criteria supporting the necessary revi-

.ston to the Technical Specification. ass elated with cycle 2 oper.s-tion.

In haly 1973 34,coci

e. Wilco.x fIIed non proprietary topia al re-

m r t MAW- Ann (Rev. I1

" Fuel Dvnsi! 1. at ion Report." which desc ribes t he methods used in analyzinA f uel densification ef fect s.1 and BAW-1ISS (Rev. 1) . Oconce 1 " Fuel Dec.si2 icat ion Report."- BN4-1 MS ut i!ized the r e t h. ds described in BAW-lug 3 5 and supported the eperat ion of t he cycle 1 of Oconee I at the rated core power of 2568 .T4t. An additional re-port. BAW-lug 79. " Operational Paraseters for E6W Rodded Plants." filed in October 1971. set forth the core operating parameters for B&'4 rodded pl..ats and outlined the anal . 31s used to ceterr:ine plant operating re.trictivas owing to postulated effest. et fuel densification.

The calcular ional =ethods and procedures used in the Arkansas and S"ID "Classi fication and Selec t ive 1.oadin.:" let ter report s (Dece b ?

19 71) were used to deternine the as-built fuel linear heat rate capa-bilit les and as a guideline for fuel place .ent. -

This report employs the analytical techniques and design bases e ! .sia l i hed an the report. = sata 1ca . S. ve to .upport e ve 1 s- 2 operat ion of oconee I at 2568 .MWt.

A br ief su::r..ary of cycle 1 and 2 reactor parameters that are re-lated to power capability (si.ilar to those this report.

in the FSAR) is included in In those cases where cycle 2 characteristics proved to be conservative with respect to those analyzed for cycle 1 operat ion, no new analysis was conducted.

In several instances, credit for new techno- 1 logy and operating experience, where applicable, have been employed to provide increased design nar,:in.

1-1 Babcock s, Wilcox l

i i

l

Based on the analyses perfor:ned, which take into account the pos-tulated-effects of fuel densification and the Final Acceptance Criteria, it' has been concluded that Oconee 1, cycle 2 can be safely operated in the proposed manner at the rated core power level of 2568 .Wt.

449g - 4 1-2 Babcock & Wilcox s_ .; -

, . ~-y. -

w ,-

- .-c..,,. , y-g- ~ -- me-- -u y we--w-

2. CORE DESIG:e 2.1. Introduction Oconee 1.nchieved initi nl criticality un .*.pril 19 197 3. .ind power gener.ition commenced on >L2y 4,197 3. The 100 power level of 2568 W t was reached on :;ovember 8 1973. Control rod interchanges were perfor=ed at 92 and 196 ef fective full-power days (EFPD). The design fuel cycle of 310 EFPD is scheduled for co apletion in the latter part of October 1974.

Oper.ition ut cycle 2 1.i.a.cheduled to began in early Decer.ber 197;.

The design cycle length is 290 ETPD with one control rod interchange to be perforced at approxinstely 50 EFPD.

2.2. Dese.ription The Oconee 1 reactor core is described in detail in section 3 of the Oconee !;uc le.a r s t.it i on . Fin ! Safety Anah-is Report.

The cycle 2 core consists of 177 Fuct assemblies, each of which is a 15 by 15 arr.ny cont.nining 208 fuel rods. 16 control rod guide tubes.

and one incore instrument guide tube. The fuel rods have an undensified no. sinal active length of 144 inches. The cladding is cold-worked Zire.n-loy-4 with an OD of 0.430 inch and a wall thickness of 0.0265 inch. The fuel consists of dished end, cylindrical pellets of uranium dioxide which are 0.700 inch in length and 0.370 ineh in dia::wter (see Tabic 2-1 for additional data).

Figure 2-1 is the core loading diagran for oconee 1. cycle 2. The initial enrichr.ents of batches 2 and 3 were 2.10 and 2.15 wt % EMU re-

.spectively.- tiatches 4A .ind 4B are enriched to 2.60 and 3.20 wt : 2M'. t respectively. All of the batch 1 assemblies and 20 of the batch 2 as-seul.lles will be discharged .it the end of cycle 1. The res.sining batch 2 and 3 assenh11es will be shuf fled to new locat ions. The ba t c h 4 A .n s-semblies . ire inserted in the center core loc.ition and on the major axes.

The batch 43 assemblies will occupy the periphery and one location on e.a.h of the najor axe.e and diagen.ile..

.,-1 Babcock a.Wilcox

w

'F.cactivity sentrol is supplied by 61 tull-l N th Ag-in-Cd control rods and olubic baran shin. In addit ion to tne :ia;1 'ength cont rol rods. cient ; ar t :al lengt n i cat rol rels are provi;ej t ar additional con-trol of axial ;,ower distribution. Ihe loc.a t ionr. of the 69 control rods a r *: Indicated on F la: 2re 2-2.

~De systen pre sure is 2200 ;ni s. .ind the nn.f. nsit ied nominal heat r.i t e 1:. 5. W V4/:- at the rated core powe r o f 2 W T4t . These values are tne ssmt- as Ic. .ycle 1 o;. erat Lon.

2. 3. Core Physics Table 2-2 co: pares the' core physics parameters or rycles I and 2.

The values f or both cycles were gener.sted using P!;q07. The cycle 1 values .stso reflect applicat ion of data f ron t he Oconee 1 Startup Report. !

.s i ni s- the core oas nat yet re.as h ed .c1 equilibriun .

.le. .Ilt f erences in core physics pararete.rs are to be expected between t he cycles.

The shorter cycle 2 will produce. a smaller cvele differential h.arnup tsun th.st f or t he cycl. 1. although the lower batch '. loading (ut t ) w!!! result in a slightly higher burnup per EFPD of operation C."Jd/nt C-EFPD) . !!.e accumulated .sverage core burn sp will be higher in tne cycle 2 tion f or ev(le 1 becau.e at the presence or the once-aurned hatch 2 and I fuel. Figure 2-1 ill. strate.. i representative relative pawer oistrib.ition f or the be.: inning at the second sycle at full power with equilibrius xenon an.1 nornal rod posit ions.

Le crit ical boron conevnt rat ion. for eyele 2 . ire lower in all cases tiun for the cycle 1. *he control ruit worth, for hot full power (due to changes in radial t lux distribut ion and isotoples) are so cwhat leas (nan those for cvele 1 (at liOL) . although they are sufficient to

u intain - the req's t red . hot.fown marg in. as indicated an !able 2-3. The stuck rod and elected rod worths.are lightly nigher t!un ' for cycle 1.

W adverse sat ty suplicat ions are .issociated with these higher worths since the prevacu. .a r c t y ana l v s i.s .is s u .ed an e l ec t ed rod wor t h we ll in excess of the values . . !s ulat ed n or c: c le .' ; the adequacy of the shut-down nargin with s yc le .' stui. k rod wo r t hs is de: on .t rated in fable 2-3.

The cycle 2 power deficits iron hat' zero power t o hot full power are higher t!un those for cycle 1 due to a nore negat ive noderator co-cf f ic let.t in cycle J.

The p.n.er . txippler coef f ic lent - in t>oth eyeles are nearly equal'.

_2 Babcock s. Wilcox

Ihe dif f erent s.s! boron wart h:s and total xenon worths for cycle 2 are lower ttun tor cycle I due to deplet ion ot~ the fuel and the . associated buildup of t'ission prodnets.

.' . 4 . Core leadins n.s t ch . t oel The batet 4 tuel assenalies will be loa fed as shown in Figure 2-1.

As-built data have been used to ca=ure eighth-core symmet ry in 'U loading.

Also, fuel assemblies with higher 2PU loaditas will be placed in locations of low power density to minimize power peaking.

As specia led' in the Nuclear Analysis scetion of this report. a 20.15 kW/ft fuel melt limit has beun employed in calculatlag the reactor protection system (RPS) setpaints. U Is value is the same as that used in the cycle 1 analysis. Based ora as-built data, all batch 4 assemblics meet or exceed this ersterion except for three assemblies that have been assigned a maximum linear power rating of 20.02 kW/ft. These assemblies will be placed in core locations that will have maximum heat rates less t han 15. 3 kW/f t in cycle 2 ind less than 19.8 kW/f t in cycle 3 to main-tain adequate fuel melt margins. These values have been calculated conservatively with respect to .he calculational method used in the Arkans.ss and SMUD "Classif ic.st ion and Selective Loading Reports"*.

In addition, assembly ID61 will be placed in core location D-14 in conjunction with B&W's cont inuing program to evaluate fuel perfor-mance.

Contained in one f uel rod of assembly ID61 are three ceramic

. spacers whit'h sistslate f uel densificat ion gaps. The proposal to insert this speci.1 as.cmbly into ocunce Unit I had been described in a letter (6/18/74) to Angelo Giambusso. USAEC.

2-3 Babcocic a Wilcox I.

fable 2-1. J cle 2 Fuel Dat.

hatch

, 2 1 _ _4 A 4R Enriche.cnt, wt2 Nil 2.10 .' .1 5 . no 1,20 Loinal ge<m. dentsity, t f l.cor 93.5 93.5 95.5 95.5

.%. of assemblies 36 80 i 56 Burnup .it BoC 2. M'.!d/ cit u 11,n95 7,100 o o 2-4 Babcock s. Wilcox

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

p - 3 c

r b l e.~~ J .' . ' i k e nco_. t C -- . c_l e_' - 2 t hv N s ' P_.ir.e. t e r.s--

k .'.

4;

(.. )t l . - '

a-

[ c E t.. :..nat h. ~u ! *) E .4-a :e S er n sp .M t t. .o a *

7. .- t u,. . ..r u burr..p - 0 4:. ". : * - t r r, s .

r. . e t ..: . cre 1.+.s-i nu . n t~ 9? - * '.*

(. r . t 1.ui t ' N.r'. . la. w t- eis r u . .t ...-i 1.. ,

' .JP - b.enk s 7 .a nd M 17.vtfr4 r

!. lie !..'s

_rFP = hanks 7 .. .J M t ract r ed : .'JJa ..

8 r . t s c.a t t= r. n - 1.uc. pn

%? P .e l l r.*t e c.a t 21 .>

-if - bank M ill.V kJ. c';u i l h e 71 .

I r:. r.t r..! rtel vor t hs - itt P. P.8 4.. *as A rt s 1-7 gb ink =, ll.'>' ws. I'#. N .!..!

E.nk 6 8 . . .* 1.;

S-eni ~ 1 l.15 1..-

. '. .:ik 9 (11.si wd) 9.>.. i t . "> ',

j.

. t'. e.f reil rod war t s.* - hFr. e tw . *kik rL enim I-/' i 1. J ) 1 i. : >

1 P mb 7 l . '* : 5.1.

E. se- M ( 17. i' wJ p rs. ~. 1 .a..*

4.

i ter t ed r od wor t h - l*/.P. .* k lis 1.C a. v. <...

P.T. ') . J ~, 1. J t Stuca r od w. r t ie t*2 P . ' I

k / >

bC J. % 0 M 1. i*. ;.*r*

f F.+er .leficit. l(IP t o !!FP. * *k /k

!> C

- 1. l . -1.<*

1.L4

-1.'n -1.'-

P e r Doppler socff - Bos;. _

1-liJ-'( *?k/k-7. piewer )

.10 J ?. p.w r t o Xc e -o.ve L , ~ % power fu Xe)

- 9. 4 .

-1.0: -t as

, s 751 power see Xe) . - 1.01 -1. .

-40! pcwer t o Xc t ' -1.03 -I...

P. we r Doppler 'ioe f f - Dit*.

I T '( t?.k / k

powe r )

-95* power (e<tui! Xe) -1.15 -1.1.

- Ede r.at o r . .w t t - HFP. 1G**-(.*?a/m *F 3 .. '

' 80C io Xe. 1000 ppe)- -G.7* -0.12 HC (equii Xe. t 7 ppm) -J.15 -2.21 Ikiren worth - NFP. ~ pps/* *.h/k

'boC (1000 ppm) 0.97 4.a4 4 .

-- DC (17 pre) - 0. 's ! O.a2 Ar Xenon wortl -- HFP. 3/k/k W (4 d.sys) 2. f> 4 2.77 IDC (equtilbrium) - 2.69 2.74 4 ,

, 2-5 Babcock & Wilcox '

m j %

j -

4 .

, , _ . ,r. , . . - -..'-- - .- , . , . . 7 -"'""*T" "

1 t-t.

s ble 2- 5. Shutdown N rgin 0.11(ul.it.on -

Oconee I, Cye:c 2 ROC ? *.P I:00 I*

,l . Available Ra d *.'o r_t_h Total rod vorth, HTPa* 10.30 1I.40 W rth redactions b'iwen ruterial burrup 10.1% 10.n; ilFP to il2P H.$) 4. c44 etuck rod worth (HZP) 2.55 1.4%

!;c t 5.94 e.ys 107 uncert.sinty 0.no _ogq

.i. To t .41 .sva l l .a s le rod worth 5.33 ** . 2 -

f. Sequi red Rad _'m' art h Poser def ici'.. HTP to HZP 1.3; 1. + s Inserted rod warth. HZP 1.05 1.s9 Flux redistribstion 0. t.O 1.90

a. Total rec.uired worth 2.79 4.e4 Shutdown Margin (la - 24) 2.59 1.40

For shutdown sargin calculations this is defined as *265 FPD. the tiee at which the transient control roci (7) begins to ove out of the core.

*i!FP denot es Hot . Full Power il2P denotes Hot. Zero Power 2-6 Babcock & Wilcox b

f LaI __ __

-t

~.

Tigure 2-1. 4 ;;c 2 - Core I. Jir.c 01.Kra-2 ) s -

+ a 3 *: s' : n a ss I

f 9

, i n

I t _.

a----.- -

as at 43 si sp I

t g

3' I

- at si at i 89 85 89 !

:1 *I :) !

I' C -. .. ! as 3 1 : 86 2 : 2 3 at 1

- . . P. it . g i; .;I .o e; . e6 . ;- .:a . . .

a

. 3 -- -, og ) 3 : } : ) : ) : 5 1 as s~*

g

. 14 .

g.t! .g it .a *: t. : Pg

.e 6 .

a$ g3 sa

! g .

as 3 of 1 i' da 3 3 . at : } at

. - - - . - - g. 2. e.12 ' '.t: 8 3 *I ri ea sJ t .

43 de 2 3 3 3 1 2 3 ) } } : at og j

.a . I ) - , L.th .4 te a6 8 *t a.! p.S e.1 g2 .i e3 !

! G. -. at 3 2 2' ) ) : 3 : ) 3 ; } as "

,st3,t.itlt.13 a.t e ea e it 03 s.t e2 * .3 . .) eJ

! , _,_43 3 a6 3 sa } 3 sa ) ) as i at } 4 3 '* ,

, n .15 m to e.* I *) :.3 : t . ..? -'

t _,43 3 2 2 } i : ) 2 ' 3 ' a : 3 ' og ~

t-5 t- .G2 s.g i.

(.t3 5 it t i) .s.ta i is a ti :.tk s3 g.3 43 og 2 1 1- 1 3 ) 3 3 3 1 : as og 6- . .. .13.f.th .s.84 .: it 3 11 a.)

3.$ . s.2 s.: s* -) . ..

i J

og 3 2 og a 3 da } } ag . : . } . at 1 12 3 't eI g? g6 34 :.3

- - C.83 . . . . . . . . ,

' s _i as . 3 3 . I 1 .: 1 2 3 2 3 3 as i . D ta. 0 t2 :.te a-1: *t.

. g a . C-6 . & 5 . D.S : a 22 ,

3.3 . . ,

3 .. og '3 4 : og 2 ? as 6 e4

. t3 .$ . as2 .t4 l

-

g 12 0 12 : $ . .t *:

e . . .

og og as 1. 3 og og

**. a3 :.t . . ..

4 .

I gafta y

. 'lTjP8Ett:

L. .

35 C:si L::a* ;g l

i e

g e

. */

5 2

. r w 3 - e,y.. w ,e--. , ,-w p- .r. we r y , . 9 e ,. - - e w eg 3, - y w ww w- y ,_ m p-

A r

Fig.ure 2-2. Cy:1v .' . Qr.*ral Fcd *.ccati.,as a

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

. . ~ ....

F I

+

.G;4 ..} 4

' L1 G;l f f f , i Get gg7

~

! Ge4 8*

,el Gs4 1

. Got e

, ,Ge3 ; Gat hi Get G:3 l . .

jgge ' G;,5 Go! k2 D '* L' n-.

- - . . - . _ . . ~ _

Ge8 l' lEst .

l Gol.il

G:4 g;3 ;.4 G 6' 4

, l' .t . -. ..,.........hl . . . .. _ . ..

583 l* Go! , IGol Gal 42 3?

iGil .

....'..l..... . . . .. . . bI . ..

ed4-g,gl gag ; Ep8 Gap Gsa G:t Gal _ ,

Gal

- . . . . . . Gef .I

. G:4 . . . . . . .-

~ 888 ' , G,3 ; ' Gal G;8 U2 ~;'

l:

- - .i.

! Sel

. . . .. . . . . . nt . . .

G;6 + ' G;l Gel D3 U' N$

e

+.=. . . . .

S.i.1

> G;e ; GA G:2 Gs! kt La I l Got 4h G:9 41

} Go) _

-. . _ ..._. . . . .. ... a I

Gol Li gja

~ 5*8 G;F I r,

~ _ -.. . .

hl *- G.I *!

]' ,

8 l <,L t 7 .L 4. .

Bittet letteGmanGE 2 attle initoGuang Gent 31::sf oftia a'i';me%CE Nis*isanGE I D 8 2 8 4 3 6 9 4 ti 9 5 0 l' E 4 4 7 1 '2 f 8 e f*'at it it 2-6 t.'

e. .

Figure 2-3.

BOC, Cycle 2 Twc-Dicensional Relatise Power Distributica - Full Power, Equilibrium Xecer.,

Nor=al Rod Pesitiens (Groups 7 and 6 Inserted)

.88 .93 1.05 1.19 .83 1.05 .96 l 1.16

.93 .96 1.00

)

.94 .75 .63 .93 i 1.14 1

1.05 I.00 1.02 .99 .85 .84 1.40 -

1.04 1

1.19 .94 .99 1.47 1.00 1.02 1.30 l

.83 .75 X

.85 1.00 1.03 .94 .93

\'

l.05 .63 .84 1.02 .94 .95

.96 .93 1.40 1.30 .93 1.16 1.11 1.04 2-9

1 PhKFort\L:r .*i!..\;.* SIS 3.2. Ine r il-H'. d r.:ul i c .*ina i s s is 3.1.1. Decien i.1.itations D,!;)3 - ;vsign linit..t ions l'or evele 2 oper.ation were 1.55 t?"J-2) nininun D:;3R at reference design conditions with vent valves c ;0 sed and 1.3d (B.W-2) sinirun D!'BR on the P-T envelope with one vent

.alve open."

rent ral fuel Melt -- 1.imitat iens on the lir. ear heat rate

.ere established utili: ting fisil s uel densification penalties. This re-

. .;; t s in a nininum linear heat rate capah!!!ty of 20.15 i'4/f t for unre-

.tricted. loading.

Pre **ure-Tenyerature (P-l) Envelope - Modifications to the P-! envelope were ree;uired t o =s et t he design D!;BR li tititions for the 5.Ti

carrelat ion under t he. a wuned worst-case densified conditions.

4.1.2. Po.cr Srike Madel

!he po s r .;ilke -odel ut i l i.ti d in t his analy is is identi-1; to that presented ;n it/W- N.i- ex opt ' r two noditications. The

~

0.:i t ic.a t ions have heen applied to F,...t:d Fy . These prebabilities h r.e 5cen changed to ret leet .. !. lit ional dat.: f ro . operat ing reactors t . ...t sur po r t a so.e.n..t .:i f t t rent ..ppro.ich and yield less severe penal-tiva due to power spikes. F g .aw chan.ed tron 1.0 to 0.i. F., was di engeo f ren . i.iias-san cist rib 4* lon to a linear dist ribut:on, which reticets a decre.asin4 t requency wit h increasing gap size.

The maxi un g.sp si:e .erseas . axial posit ten is sho.n in rinre 1-1. and the power npike tactor versu . axi.al position is shewn in Ti.:ure 3-d. These t'igures also how the init ial and t inal theoreti-eal densitles (TDI TDF) used in t he calculat ions.

4-1 Babcock & Wilcox k

r*"-

's 3.1. 3. DNBR Analysis The DNBk analysis utilized the SLW-2 CI:F correlation and the actual operating primary system flow. The B&W-2 CIIT correlation is a more realistic prediction of the " burnout" phenomena and has been re-viewed and approved for use with the Mark-B fuel assembly design (B&W-100006 ). The actual primary system flow (107.6' of design flow) has been verified as noted in reference 5. Utilization of the B&W-2 CilF correlation and the actual primary system flow provide a more accurate prediction of DNBR margin in the core.

In addition to the maximum design conditions considered in the FSAR, the ef fects of fuel densification on DNBR were taken into account. As input to the DNBR assalysis for the batch 4 fuel, the mini-mum lot average density and the densified as-built stack height were used. Using this input and the corresponding power spike, the most lim-iting DNBR conditions were calculated.

The axial flux shape that gave the maximum DN3R change from the eriginal design value was an outlet peak with a core offset of

+11.R'.3 - The spike magnitude and the maximum cap size used in the analysis are 1.07 and 1.v5 inches, respectively. These two ettetts results

2. -i.4 and -1.O' cnances in nini:.um hot channel DNHR . sad pc.Ain.: n.a r-gin, respectively.

In addit ion, all fuel assembly water channel spacings were measured and each assembly was evaluated as to its DNBR capability. Any penalties incurred were accounted for in the thermal, nuclear and safe-ty analyses.

3.1.4 Fuel Temperatures The : Inimum capability with respect to linear heat rate of the fuel is 20.15 kW/ft. The basis for the analysis utilized is given in SAW-100iil and BAW-100444 with the following additional modifications:

1. In the equiaxed zone, 37. porosity is assumed but is not used in the calculation. That is, the input value for fuel density is used and therefore no credit is taken in the esiculation for increased thermal conductivity of U0,s for the higher fuel density.

1-2 Babcock s. Wilcox .

w

2. The option in the code for no restructuring of fuel has been used in the analysis presented here in accordance wit' AEC'r. inter-in evaluation of TATY.-

3. The calculated gap conductance is " educed by 25% by the code, also in accordance with AEC's interin evalus ton of s AFY. 9 All fuel lots were inspected for average and LTL density and diameter values. Each lot was then evaluated as to its limiting linear heat rate in accordance with ref erence 4 As a result, three assemblies will be selectively loaded as described in section 2.4.

3.1. 5. Su= mary This analysis assumes that densification and associated phenomena vill affect the hot channel, which has the most li=iting ther-nal-hydraulic characteristics in the core. In addition, the power spike is assumed to be located at the hot channel position that minimizes the DNBR. The resultant 5.4% F.iBR loss, or 3.0; reduction in power peaking nargin, will be compensated by changes in the Technical Specifications, so that the plant can functinn at rated power without violating the ini-tial design criteria for DNBR and/or fuel melting.

Table 3-1 compares thermal-hydraulic operating conditiens for cycles 2 and 1.

l 3.2. Nuetear Analysis The RPS power / imbalance limits (DNBR and centerline fuel melt pro-tection) and the operational limits (administrative LOCA kW/f t controls) have been established for cycle 2 operation according to the cethods and procedures described'in EAW.10079 3 l and EAU-12s8 . Following is a sucsary of cycle 2 design parameters utilized in the analysis:

Parameter _ ___

Cycle 2 value Fuel melt limit, kW/ft l 20.15 DNS peaking ma gin penalty due to densification, % -3.0 overpower, % of 2568 MWt 112 Densified nominal heat rate at 100* power, kW/ft 5.80 Power spike factor Figure 3-1 l Nuclear power peaking uncertainty 1.075 1

LOCA lisit, kW/ft Figure 3-8 3-3 Babcock & Wilcox 1

I l

The power peaks resulting f rom cycle 2 operation were examined.

The plant can operate at rated power without exceeding DNBR, fuel =elt, and ECCS criteria by adhering to the limits specified in Figures 3-3 through 3-7.

3.3. Safety Analysis 3.3.1. General Safety Analysis The significant effects of fuel densification were identi-fled, and the effects on the safety analysis were reported in BAW-13H32 ,

This detailed analysis showed that no safety margins that previously ex-isted were Jeopardized by the postulated effects of densification.

It was further established that the limiting transients were the rod ejection accident and the loss of coolant flow. Table 2-2 shews that the ejected rod worth for cycle 2 (0.32%) will be much less than the rod worth used in RAW-1388(0.50%). In addition, the moderator and Doppler coefficients of reactivity are more favorable than those used in the previous analysis. Therefore, it can be concluded that the rod ejection accident will result in conditions no more severe than pre-viously reported. The loss-of-coolant-flow t/pe accidents (LOCAs) will be less severe than previously reported since the initial DNBR will be higher.

As shown in Table 3-1, the initial DNBR at the o'vrpower of 114% of rated power for cycle 2, batch 4, is much higher usia.g the measured flow of 107.6 and t he BAW-2 correlation. 5'E Thus, the tran- .

sient results for cycle 2 fuel will be less severe than or equal to the results reported previously.

The peaking values are consistent with the discussion presented in section 3 of BAW-13Mg2, 3.3.2. LOCA Analysis A generic LOCA analysis for B&W 177-fuel assembly nuclear steam systems with lowered steam generators has been performed using the Final Acceptance Criteria ECCS Evaluation Model and is reported in EAW-100917 . That analysis is generic in nature since the limiting values of key parameters for all plants in this category were used. Thus, the anal-ysis provides conservative results for operation of Oconee 1.

34 Babcock a.Wilcox

Portions of the LOCA analyst, nave been repeated for Oconee 1 using specific parameters associated wit the Oconee 1 plant and cycle 2 fuel. The purpose of this re-analysis v2s to reduce se:e of the over-conservatisms iron the generic analysis, emly the 2 and 4-foot eleva-tions relative to the botton of the core were reexamined.

At the 2-foot elevation, an explicit evaluat ion of the LOCA limiting peak linear heat rate to maintain the peak cladding tem-perature below 2200F for f uel batches 2, 3. and 4 (which constitute cy-cle 2 fuel) was perforced. The results shew that the licits are 16.0, 16.0, and 17.0 kW/ft for batches 2, 3, and 4 respectively.

At the 4-foot elevation, fuel batches 2 and 3 were eval-uated for a peak linear heat rate of 17. 5 kV/ f t t the results show that the peak cladding teeparature is less than 2200F. Eatch 4 was not ex-anined at 4 feet since the stored energy in th.* fuel for this hatch is less than that for batches 2 and 3 and therefore would result in a lower peak cladding teeperature than for hatches 2 and 3 at the sa=e linear heat rate.

Figure 3.3-8 shows bounding curves for allowable LOCA peak linear heat rates for cycle 2 fuel, 1.c.,

fuel batches 2 and 3 and the reload batch, 4

3. 4. Mechanical Anal _vsis Cladding strain due to fuel pellet irradiation growth was calcu-lated according to the procedures out!!ned in S&W topical report, EAW-100531 . Results for the batch 2 and 3 assemblics remaining Jn the core for the cycle 2 have been reported in BAW-11ssi. These results indi-cated that cladding strain af ter three cycles was less than the 1% strain limit.

Analyses of the batch 4 fuel through three cycles of operation resulted in a cladding strain of 0.52%, which is well within the 1%

strain limit. Input to the batch 4 analysis included applicable as-built data on cladding and fuel pellet dimensions and a peak pellet burn-up of 42,181 mwd /mtt'. Conservatisms in this analysis included a minimum pellet-to-cladding gap, a maximum pellet density greater than 96.5%, and pellet thermal expansion equal to that at the fuel-melt power limit.

Predictions of cladding collapse into an axial gap were made usir.g the CROV computer code described in the B&W topical report pal *-10084. I "

The analysis assu=ed that densification could occur in two ways at the 3-5 Babcock s.Wilcox i

beginning of life or over a 2*.00-hour pernod. ~i he second assueption pro-duced higher cladcing temperatures over the 2400-hour period and the most conservative resulti. Other conservatisms used in the analysis included minimus cladding wall thickness. naximun initial ovality, a decrease in the prepressure level at BOL no fission gas release, and severe burnup and radial peak histories over each of the cycles. The predicted values for time to collapse as a rer. ult of this analysis fot batches 2 and 3 were reported in the letter report, "Oconee 1 Cladding Coll.ipse " August 1974 The most conservative results (densification occurring in 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months ) indicated a time to collapse of 25,650 hours0.00752 days 0.181 hours 0.00107 weeks 2.47325e-4 months as opposed to a three-cycle residence time of 21,500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months . The results for batch 4 predict that collapse will occur at 29.990 EFPH of operation. This corresponds with a projected burnup of 37,800 %*d/ctU, whereas the expected three-cycle burnup and exposure time for batch 4 is 27,751 % d/mtU and 21,500 EFPH.

Table 3-1. Operating Condit tons

_ Cycle I Cycle 2 Power level, L*t 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % FSAR design flow 100.0 107.6 Ref design radial-local power peaking factor 1.78 1.78 Ref design axial flux shape 1.5 cosine 1.5 cosine CliF correlation W-3 BLW-2 Minimum DNBR (max design conditions, 1142 power, no densification effects 1.55 2.05 Mechanical hot channel factor on enthalpy rise (F q) 1.011 1.011 Mechanical hot cf.annel factor on local

surface heat flux (F)q 1.014 1.014 Densification penalty, %

DNBR margin -4.5 -5.4 Power peaking margin -2.6 -3.0 1 Power spike at outlet 12.5 7.1 Linear heat rate to central fuel melting (Class I fuel, AEC restrictions), kW/ft 20.15 20.15 1

3-6 Babcock s.Wilcox 1

I

Figure 3-1. Y.a x i=u=2 Gap Size Vs Axial Fcsition 2.0 y TOF = 96.5%

101 = 94.5', .

, _ g

% i E # !

5 / . ,

I S

E /

0.0 ~

0 20 40 60 80 100 120 14S Arial Position Along Active Lengtn, inene ,

3-7

6 rigure 3-2. N.er sp n e ract er vs /u:..; rositter.

1.08 .i T0F = 95.5*,

i 1.07 - T01 = 94.5*, ;

i l.06 !

I I ;

E g 1.05 -

e A

1 I

1.04 :__.._.______._____..

3 .

' I m

8 1.03 -

1.02 _p_ ._

l __.

I 1.01 -

1.00 0 20 40 60 80 100 120 140 Axial Position Along Active Lengtn. incnes 3-e

. F4.;ure 3-?. .r.. :. :: u - i i:e . m:-::3

' e: 7 .2 .

.es L: e.

20

\

)

--100

\

e f

'\ k l

[

EC 63 (3AND4) - -

\

40 20

-40 -20 -

0 +20 +40 Reactor Poser imoalance. :

CURVC REACTOR COOL ANT FLO8 (LB,HR) 1 131.3 106 2 98.1 106 3' 64.4 106 4 60.1 10 6 3-9

Y

~-s it ;-0. ..r*i... . . . . ,, , . , _

i PGeer. t 256E us:

r 110 f

1

-20l4 .

i4.1

.- lCU 1*2 50 4

a,

-u 4

30. 7 5

]

9*

- aw i

I J'

1 i a

_ t; O ,, !

- 50

-31.2.52 28.1. 52 -

1 !

.. 40

I

-- . _1- >_ ,

t -

30 - 20 -10 0 10 20 30 Co r e m::a l a nc e . '.-

3 3-;D

-c

-a y '- . - - - . 4 p g ,y *mp-- ,.- , -- --4p-r-.w--v-- -w-gg

-rrr sw + q -y- -

ge- y*-w-- y.w.,e a -u--

l

~ i . t. r s '2 - 2 ^ r t r .; l - - *

'. i' J;e.2:.2o j

l l

. ,,res ,g ., ;e .-r

. ,., ,its;. . c.. . . ,

>: .as.

.' **e satrir . . .3 s i .i ! s n e -

-e

e r a t i e r..

-~

173 2l3 1C3 ..

tg

$g '. Li 8:e*' L****

C a t ,- .

Seitra:*e2 negien 83 -

2);

122

m. *,-

~

i. .

60 -

M a

,- Fef ess cie 37 p.

. s l *.. erst *g

,,;.t-1 0

'4 C

1 20 -

e t

0 *- - --

L. - - - - - - - - - - - - - . . - - - - - - - - - - - - - - '

C 50 100 150 200 210 300 p:1 tates s;tastaen ,,

5,. os g .. < 100

. --. ----.2 0 25 to 75 1CO. . . . - .G a .*

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

C

. ,25 ,

50 15 1CC b6 Ca5 3 ..

L

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

A M

f~ 4 Tigure~3-(s. Centre

r nii Crea; .itidra. ; * :- :<-t r i n.r-i . .; c ; . r a- :-

i 3.

' f 4: .Irces a s

e ;e r:eauge w :* ' . .. ": se s :->;..; ;

i m.n .

t

~2. T*e e :r:rsess ...:35 se, ,, ,;,,:: p. 73g.g.

,.,, .,g ,

3;er21:3- .

- f . 3:34.c3 ;e ::ee- e.* !.!C . !C *

.:.a i 1

P

__ ~ --

?

4 5

~s the

s. s.

1CQ - /

r \

-/ -/,

3 i,

, e v '

/,

80 Restr.cte- '.

/

'e',-

/

r t

s ' t ae .

+

-- / , .-^- -

a .. -- -

' e e * ~' s

. .I - -

r .

'O -

2

+ *-

s M'

. s%

. o-4 4

3-or <$2 Pi i

. .s4

.=

- / 'e' m i: :e

- .2 og - C. erst. g

_.i:

8eg.:- ;

I 20 _ . 1 t

e

-..t O! 'i , , .

O. 50 100 150 200 25C 300

-E:s case: estraraes E 25 $$ ?S 100 s

h~* '

0 25 -50 75 100

. -. . 1 0 -25 .50 75- 'C: g.g y- _. , , i Gst n 3-l' s

4

' , s -

7 #"

~ P

'l

.; . ,;.--,- , .-. .-. 4 r._ . . , - - , _ , , _ . . . - . - , , , _ . _ . , - , , , . , . . . - , . ~ , . . - _ , - - , ,-..v__-- _ , . . . . , , . . , . . - ,

.#. r

.n, :

N. .

I

'.' F ic.re 34 cc .t r.: . re .: r.*u: v . c. c. - - -

+ : a- .a . . . -i .:: i n:a t, Ron inoes is tne percentage.s.- st t e ertr 'e 2 ' 0' *"f :: " .* : < ' i groups.

s J

j -_-_. . _ _ _ ..

1 e: x 5

.l 2 Pcc se tn:taeas L.

.y x t

~

=

-100 - 15C -"t

Restricten .s

= t, . Re giert . 5,

1. -

e i ee a ,

u 30 ' *

=

E

.-4

= ,

a.-

='

t !

Pe. issis:e fo .-60a_ s p' 0:erating se g, ,

4-a - ,.

I

_E- 4'

=

- = . 40 L

.s o

a.

20 -

'O i i 1 g :

50 100 150 2:0 25C C Roo inces. *. eitnce seaI 0- .2 s

b '

1 f }

..j9 I

l

-* e& *,b ( Y .?

6' e e\es' M's ? e$ *% s b4~

l' " e h - , ; ,3 .,

2+t-18 _ _ _ _ .

i 4 '-

e... : - 54e QL51 .

i

\ . .

15 _ _ _ ., _ _ .. _ . ..

M as:

"m i E

14 L_-- . . .

m M

Z O

G C

. 1 2 ,_. -. . _ . .

4,9 A

i IO .- __ . _ . _ .

0 2 4 5 S :

Dista ce frc ? set. ' '

5 _

_ ddG:,:A-

+

l. . O tM:1.t.*S l o';S ihe an.nlysis deeunented in thi. report uns periormed u%ing ;ienalt les

'. r t uel da r -it irat ion and met hodas and tirocedurer that h.sve been acestted by the USEC. Cycle J eperation at rated power has be n shewn to be cen-si. stent with the thermal design criteria and 14CA kW/f t li=its. An naaly-

.413 of claddins:. creep-collapse performance in accordance with ref erence 3 h.is shewn that no cladding collapse will occur during three cycles of op-eration.

Based upon the Tect.nical Specifications derived from the ar.alyses presented in this repert,. the Final Acceptance crfteria ECCS limits will net be exceeded, and the thermal design criteria vill not be violated.

F J. - I Babcock & Wilcou l 1

i l

.- l i

1

~n,,.~.,.a,,.-.~,,._...,n;a ,, n. . . _ . , ...u., -

_~ -

1 I

5. 'REFERE;CES Fuel Densification Report. BAW-1005 3. Rev. 1. Babcock & Wilcox, Juiv 1973.

Oconec 1 Fuel Densification Report. BA,W-138=, Rev. 1. Babcock & Wilcox, J uly -1** 7 J.

3 Operational Parameters for B&W Rodded Plants. BAW-10079, Babcock & Wilcox, oc t uoer l'17 3.

Ictter Report: " Classification and Select ive Loading of Fuel for Rancho Seco." Babcock & Wilcox. December 1973. (Proprietary).

'- oeonee 1 Startup Report. Duke Power Co.. Toverber 16, 1973.

Correlation of Critical lleat Flux in a Bundle Cooled by Pressurized W.ater. BA,W-10000 Babcock & Wilcox. March 1970. ,.

"Densification Kinet ics and Power Spike Model." Meeting With USAEC.

July 1 197*.; J. F. Ilarrison (B&W) to R. l.obel (USAEC). Telecon.

" Power Spike Factor." July IM. 1974 C. D. Morgan, and H. S. Kao. TAFY Fuel Pin Temperature and C.as Pressure An.nlysis. BN.*-14044. Babcock & Wilcox. Ny 1472.

~*

6 M. l Man, et ..l.. haw's i t C5 s.v.iluat i. a !!. del Rep. rt With ';pesisi.

A plicat ton to I 7 7-Fuel Assembly Plants Wit h Lowered-leop Arrangement.

BA*.'-10091. Babcock & Wilcox. August 1974 '

A. F. J. Eckert, 11. W. Wilron, and K. E. Yoon. Progra= to Determine In-Reactor Performance of B6.* Fuels - s:ladding Creep Collapse.

Br.'-10034. Babcock & Wilcox. May 1974 ,

5-1 Babcock a. Wilcox um

ML19319A780

Text

Navigation menu

Search