ML19316A906

Proposed Tech Specs,Allowing Use of All U Fuel Types as Reload Fuel

ML19316A906
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	04/15/1977
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML19316A901	List: ML19316A900 ML19316A904 ML19316A906
References
NUDOCS 8005280045
Download: ML19316A906 (44)

Text

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i, O

CONSTJ2ERS PC'4ER CCIG.GY Dockee 50-155 Request for Change to the Teak-d eal Specificatices License D?R-6 For the reasons hereinafter set forth, it is requesced t'..at the Technical Speci-ficaticas centained in Provisional Cperating License DPR-6, Dcchet 50-155, be e5 anged as described in Section I, below.

I.

Changes A.

Add er replace the follow _:gi 1.

Table 5 1 (Page 3ha).

2.

Page hib.

3 Table 1 (Page k3).

k.

Table 2 (Page 43a).

5 Table 8.2 (Page 91).

l l

i i

l 541799 1

Sk?cP5289QL5

1 TA3I2 5.1 (Additier.at I:fc=atics)

See Page 34 1

General Reload G-1U Relead G-3 1

Gecmetry, Fuel Rod Array ExE 11 x E

~

Rod Pitch, Inches 0 577 0.577 CO Reds 109 E3 2

k 0

Ccbalt - 3 earing Corner Rods Gadoliniu:2 - 3 earing UO R ds k

k 2

Inert Spacer Capture Rod (Zr-2) 1 1

Zirealey Reds 3

3 Spacers per Bu= die 3

3 Fuel Red Cladding Material Zr-2 Zr-2

'Jan *ick=ess, Inches 0.03h 0.03h Fuel Reds Outside Red Diameter, 7-+ es 0.kh9 0.hk9 Tuel Stacked Density, Percent Theoretical 91.6 91.5 Actite Fuel Length, Inches Standard Rod 70 70 Till Gas Eelius B 55 Heliu= c9F:

1

~.

541800 l

3u, e

r

~FIEL 3GILZ Smar.C G-3 EZLCAD FCIL

~

(To 3E StJP?r.7m) f l

541801 i

~. =.

• (w l

l

6 e

e TA3LE 1 Reload E-G and Modified E-G Reload Reload F, J-1 4. J-2 Reload G G-1U G-3 F44-= Core 3urnout Ratio at Over;cver 1.58 1.5**

1 5**

1.5**

Transient Minimum Burucut Ra*gio in Event of Loss of Recirculation Pumps From Rated Power 15 1.5 15 15 y=vd m-Eeat Flux at Overpower, Stu/h-ft 500,000 395,000 407,000 392,900 Maximus Steady State Heat nux, Stu/h-ft 110,000 324,000 333,600 302,100 Fa vd = = Average Placar Linear Heat Generation Rate. Steady State, W/ft Stability Criterion:

y=vd um Measured Zerc-to-Peak nux A=plitude, Percent of Average Operating Flux 20 20 20 20 Fa vd -- Steady State ?cver Level, W 240 2ho 240 240 g

y=vdw = Value of Average Core Power censity 2 2ko -W., W/L h6 h6 h6 h6 Nomi=al Reactor Pressure During Steady State ?cver operation, psis 1335 1335 1335 1335 Minimus Recirculation Flev Rate, Lb/h (Except During Pu=p Trip Tests or Natural Circulatioc Tests as Outlined in 6

6 6

6 x 10" Sectica S) 6 x 10 6 x 10 6 x 10 Rate-of-C'.ange-of-scactor Pcver During

?cver Operation:

Centrol red withdrsval during pcver cperation shall be such that the average rate-of-change-of-reactor ;cver is less than 50 W per sinute v'en ;cver is n

less than 120 W, less than 20 Wg per sinute when ;over is between 120 and 200 Ws, and 10 Wg per minute when power is between 200 n.d 240 W.

• Based on correlatica given in "Desiga 3 asis for Critical Eeat Flux Condition in Soiling Water Reactors," by J M Eealcer, J Z Eench, I Janssen and S Lerf, September 1966 (APD 5286 sad APED 5286, Part 2).

• • 3ased on Exxon Nuclear Cor;cration Synthesi:ed Zecch Lerf.

• • • To be determined by linear extrapolation fres Table 2 attached.

I 541802

(

h3 l

l

TA3LE 2 W_*3GR (W/Ft )

Planar Average Exposure 7, E-G, Reicad G i

(t&d/STM)

Modified 7 J-1. J-2 and 3?SDA Reload G-1U Reload G-3 6.k33 6.k91 6.554 0

200 95 94 6.750 6.T-58 214 6.807.

216 6.887 6.388 437 6 973 kk3 6 360 88k 6.978 885 T.033 893 6.929 1,758 6.970 1,769 6 98k 1,773 6.885 3,k94 6 913 3,509 6.983 3,5k5 5,000 99 9.T 6.838 6,939 6.865 6,970 6.978 7,085 10,000 99 97 6.8kT 10,k22 6.882 10,kS2

)

T.019 10,690 6.867 13,938 6.90k 1k 019 T.069 ik,355 15,000 98 9.6 20,000 8.7 8.6 6.905 21,022 6.958 l

21,194 T.171 21,843 25,000 8.4 8.3 6.8k3 27,778 6.903 28,035 T.161 29,08k 6.703 34,013 6.923 35,1k7 6.958 35,322 l

541803 u.

i

9 g

l TA3LZ c.2 Centerselt

)

III UO -

4 2

Inter-y*

2

=ediat e Advanced NFS-DA c~, Minimun Core Su= cut Ratio at Over;cver 1 5*'

1.5*

1.5*

15 Transient Minix.n 3umout Ratio is Event of I4ss of Recirculation Frca Rated ?cver 15 15 1.5 15

~

Maximum Heat Flux at Ove.,,cver, h02,000

_l Stu/h-Ft2 500,000 i

i Maxirm Steady State Eest nux, Stu/h-Ft' 410,000 500,000 500,000 329,000 M=* = Average Planar Linear Heat

~

(Refer tu J

Generation Rate, Steady State, W/?t Table 2,

?sse k3a) l Stability Criterien: Mad - Measured Zerc-to-Peak nux Amplitude, Percent 20 of Average 0;erating nux 20 Maxirm Steady State Fever Level, W 240 240 Nemisal Reacter Pressure Du d:g 1,335 Steady State ?cver Operation, psig 1,335 Mini =un Recirculation Flov Rate, Lb/h (Ixcept During Pu=p Trip Tests or 3atural Circultion Test.1 as Outlined 'a 6

Sec 8) 6 x 10' 6 x 10 Humber of 3undles:

1 3

Pel.let CO2 1

2 Power UO2 Rate-of-Change-of-Reactor Power Du%.

Pever 0;aratica:

Centrol red withdrawal during pcver c;eratica shall be such that the average rate-of-change-of-reactor ;cver is less than 50 W. Per sinute -*.t.es ;cver is less than 120 We, less tian 20 W. por =inute vte?. ;cver is between 120 and 200 W, and 1015(t per =isute' vhed pcver is betveen 200 and 2ko W:.

i

+3ased upon critical heat flux correlation, APID 5286.

• • No longer used in reactor.

l 91 541804 1

1

.~.

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II.

DISCUSSION 1.0 I3TRODUCTION AND SU W A.E The purpose of this proposed change is to anov the use of an an uranius fuel with acceptable ICCS perfor=ance characteristics in the Big Rock Point reactor and to delete the Technical Specifications li=itation en the design bu=up of fuel bundles.

Presently, licensed fuel types are n x n all uraniu=, n x 11 =ixed-czide and 9 x 9 an uraniu=. ne n x n an uranium fuel (denoted G-1U) was used as reload fuel at the last relceding. Se proposed 11 x n an uranic = reload fuel (denoted G-3) is very s'-d'ar to the G-lU fuel asse=blies with three basic differences. For G-3 asse=blies, the four corner cobalt target reds have bec: repisced with fueled reds.

nere have been changes =ade to the bundle enrich =ent distribution which reduces the overan bundle enric$.=ent frc= 3.885 to 3.lk%. Se placenest of the gadoi'ata poisen pins has been altered for better peaking character-(

istics. nese effects have 'been accounted for is the subsequent analyses presented.

)

nis submittal centains infer =ation cencerning fiel syste= desig=, nuclear desig., ther=al hydraulic design and accident and transient analysis as recem= ended by the " Guidance for Fre;csed License A=end=ents Rela.ing to Ref'a1* 5."

Ivery feasible atte=ct to present the infer =ation requested by the guide has bee: =ade.

In genersi, *he =sjer difficulty 1: previdi:g this data was is fer=ulating a suitable " reference cycle" as defined in the guide. For Sections k and o, the reference cycle used was Cycle 14, specif-f icany, the Reicad G-lU 01e1. For Section 7, there was ne si:gle suitable reference cycle available. Sun i the latest analysis found acceptable by the Com=ission was used as the reference cycle for each specific accident or transient. ~.: =acy cases this dated back to the I'ESR.

F'-*1'7, for Section 5, since =any of the ;crs=eters required by the guide vere not.rcu-tisely calculated or s+=4tted is previcus relcad licensing sub=ittals, no reference cycle is giver..

It is Ccusu=ers ?cver Cc=;any's intent to utilice Cycle 15 as the reference cycle for Section i for subsequent reload licensing sub=it:als.

54180.5 2

f e

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l 2.0 CPERAT UG F.~diTORY Cycle lk power production began on July 28, 1976 fonoving*a refueling The core leading censisted of 38 - 9 x 9 fuel assemblies and h6 -

outese.

n x n nel assemblies, vith residual fuel assemblies relocated in the core to ;,rovide adequate shutdevn =a.rgin and acceptable cycle power peak-ing. The off-gas release rate stabilized fo noving start-up at approxi-

stely 750 uci/s (corrected for specific grarity). The plant has oper-ated*since that time at-;cver levels ranging fren 2C6 to 216.W,..

~

~

This reduced power level resulted frc= reaching Technical Specifica-1 ticas MA;LEGR -4 ts, pri=arily for the ? fuel. The of*-gas release i

rate for the latter ;crtics of the cycle is ateraging a;;reximately TCO to SCO uCi/s. This is the lovest off-gas release rate of any cycle and is attributed pri=arily to the rer. oval of copper based =a-terials from the pri=a:/ system se tersi cycles ago, and the subsequent discharge of fuel bearing co;;er based c:c.d.

Cycle ik was crig'-="y desigted for an energy pr duction of 61 Gi4D; however, due to the. ex-tended operating period, energy productics is cov expected to exceed slightly this figure with a ;cwer ceastdevn at the end of the cycle.

A sm/ of Cycle 14 starc-up and tests perfor:ed at the beg'- ' g of the cycle is centained in Special Reper-No 2k daced Hove ber 2k, 1976.

541806 3

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3.0 GZ3EFE DESCRIPTION This ccrres-Cycle 15 is designed te produce a target energy of 72 G'a0.

pcuds te a cycle length of a;;rezi=ately 325 days of power c;eratics at 220 &,,. The projected cere leadi g for this c7 ele, including cebsit distributics, is shown in Figre 3-1.

This'icading scheme is subject to

=1ser changes, depending upon the :esults of hel si;;ing cor.! 2ted during the June 1977 cutage, with the goal of -*d g effluent releases as icv as reasc ably achievable. Figre 3-2 details the fuel red arrange =ent, the initial fuel enrich =ent and the gadcliniu= distributics and concentratics for the G-3 hel. The gadcliniu= is designed to turn up in a singla cycle; thus, c=17 the new assemblies <catain sie,nificant a=custs of turnable ;cisc=.

Figure 3-3 is provided to indicate the beg'- '"g of life fuel bursu; dis-tributics for C7 ele 15 The C7cle 15 hel leading patter: has been designed to icecr; crate 180 rotatic a1 sy: etry throughout the ecre. The hel distributics has been devele;ed te ecc;17 vith Technical Specifica icas

'd dtaticus and safet7 analysis criteria. These !*

• ts and criteria include MAPLEGR, ' '-"-

critical heat flux ratics, -=vd-'

heat flux, -avd-~

centrol red verch and =i d-"

shutdev: argin a=c=g others.

l i

541807

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Figure 3-1 CYCLE 15 BIG ROCK POINT

-N-CORE CONFIGURATION A

B C

D E

F

+

so e.

3.>'4444 e'

'#e% s%e% s%

+' \\

e e

4 440044 e'O 4'4 %e% % 4 4 t o # o % o #' #

1. # # o o#o#o%o e

c

'e#

'*4 4 4 4

^

4 4 4's s s #

's % +'# o #

'% ? % %

e %%%

%@A A4 4 NN*kh

/ '%

' g%

e 9

e Fig *,re 3-2 FUEL ROD ARRANGEMENT - BIG ROCK POINT RELOAD G 3

~

"OOeeeeeeeeO e@ee@@e@ee@O 80888 O8888 OOOOOOOOOOO 8088@@O0808 88@O0888@G8 O@OOOOOOOGO 80880888888 08000@e8888 OGOGGGGGGGO g)OOOOOOOOOQ NUM8ER OF RODS DESCRIPTION 3

INERT RODS 12

.l.50 wt% 235U 40 2.52 wt% 235y Q

61 3.82 wt%235U 0

4 3.82 w,*uS u + i.25 wt*Ga2 3 e

.h 12 TIE RODS h

1 INERT SPACER CAPTURE RCD 541809 5

Figuro 3-3 CYCLE 15 BIG ROCK POINT BOL RNUP D15 R BU ION 8

C D

E 1

4 e,'s4444 s 4 4e4 404 4 s44444444 4444e4444

O 400 4 4 See#

/4%444#

O V4444440'

'^4 4 4 See 4 4 4

/ 44444444

/ 4 4e4 See 4

+N?

Os

/ 444 oc

./

/

/

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7 541810

k.0 FUIL SYS':"rC4 DESIGN The G-3 reload fuel design for Cycle 15 has =echanical, ther=al hydraulic l

and neutronic perfer=ance characteristics s4 d'ar to the G-lU reload fuel design for Cycle 14, the reference cycle. 3cth G-1U a=d G-3 fuels empicy an 11 x 11 red =strix vith four inere reds; hcvever, G-lU hel design incer-porated four corner cobalt reds necessitating a slightly higher enrich =ent A ccmplete description of the =echanical, ther=al hydraulic than G-3 hel.

and neutronic characteristics of the G-lU fuel was presented in our letter dated October 13, 1975 k.1 Fuel Design As* discussed previcusly, the ajor fuel design change for Relcad G-3 fuel when ec= pared to Relcad G-lU hel is 'the e~4-ation of the fcur

' enrichment.

cobalt target reds and the reduction of the overall U Table b.1-1 lists the design paraceters for both the G-3 pre;csed fuel and the G-1U reference fuel. Table 4.1-2 delineates the fuel infectory at ECC for Cycle 15 It censiscs of the fuel type,==ber t,

of assemblies,==her of cycles in cere, and initial ECC enrich =ent, 1

density and average burnup. The G-3 hel is designed to te free-staM* g thrcughcut its life in cere, which is censistent with previcus l

G relcad fuels and, like other G hels, G-3 vas initially filled at nr-m*-=' at=cspheric pressure.

i i

i S

541811 c

w 7

7 v 7

-c-*e4 y

e

- g m

y

• w

i TABLE k.1-1 DE. SIGN PARA!EIIRS FOR 3IG RCCK POIr G-3 VS G-lU All Ursnium Desir s Rel'ad G-3 G-1U 1

e Fue1 Asse=bly 11 x 11 11 x 11 Red Array Red Pitch, h 0.577 0.577

'Jater-to-Fuel Vol=e Ratic of Lattice 2.60 2.69 Heat Trs sfer Area', Ft 80.23 77.48 Hu=ber of Spacer Grids 3

3 Reds per Eundle 0

h Cobalt Target 16 T.cv hrich=ent Ursnia 12 Intermediate Enrieb=ent Urania k0 ( heludes k 32 (Includes k (2cepoisen)

Tie Reds)

Tie Reds)

. o :.

61 ( heludes 8 Eigh E= rich =ent Urania 61 (Includes 8 ~ Tie Reds)

Tie Reds )

Pcisen (Urania-Gadclinia) h h

Spacer Capture (3cnfeeled) 1 1

hart Red (3cnteeled) 3 3

1m 121 Fuel Red Dia=strsi Pellet-to-Clad Gap, S 0.0095 0.0095 Ctcrall Fuel Ecd Langth, h 78.501 78.501 Actite Fuel Length, b 70.00 70.00 Ple:n:= 7clu=e i.ength 3.901 3.901 Eeliu=

Heli =

Fill Gas I

9 1

541812 t

i 1.

l

)

l TA31E k.1-1 (Centd) i Relcad G-3 G-1U Fuel Red Weights UO Reds (Total Carsmic Grams) 12k8

~

' 1248 2

Poison Rods (Total Ceramic Grs=s) 12h2 12142 s

Fuel Pellet

!'.aterial Sintered UO Sistere UO 2

2 Diameter, b 0.3715 0 3715 length, In 0.300 0.300

ensity, 5 Theoretical (TD = 10.96 g=/c=3) 93.5 93.5 hitial Erich=ent Lov hrich=ent Ecds (Wt5 U-235) 1.50 2.30 Inte.~ediate hrichant Ecds (Wt5 U-235) 2.52 3.20 Eigh Erihnt Ecds (Wt5 U-235) 3.82 E.60 UO -1.20 Wtn Gd 0 P*i'*'

23 (W!5U-235) 3.82 k.60 krerage for Bundle 3.1k 3.88 Dishing Both Eds 3cth hds Dish Volu=e, 5 of Undished Pellet Volume 2

2

<k

<k

• ESC of W. ties, ;;n Pei'sen ?ellet Material 1.20 Wt5 1.20 Wt5 0

Gd 0 0

go3 2

23 2

Diameter, b 0.3715 0 3715

(,

Length, h 0.300 0.300

• !3C = Equitalent 3eren Centent 541813 10 g

TA3LE k.1-1 (Conti)

Reload G-3 G-1U Poison Pellet (Ccatd)

Density, 5 of Theoretical

~

(TD = 10 91 g=/c=3) 93.5 93.5 3cth hds 3cth Inds Dishing Dish Volu=e, % of Undished Pellet Volu=e 2

2 DC of I= purities, pp=

<7

<7 Cladding Material Ziresicy-2 Zirealcy-2, Cold Worked and Cold Worked and Stress Relieved Stress Relieved Cutside Dia=eter, In (After Itching) 0.kh9 0.hh9 hside Dia=eter, E O.3810 0.3810 1

t Hominal Wall ?ick=ess, b (After Itching) 0.034 0.03k M* *-- Wall Thick =ess, In (After 0.032 Itchi:6) 0.032 DC, Tot.d, Ecluding S;:urities

< ko

< E0 hsula.or Fellet 2 3) 0 Al e m (A1 0

~

Material 23 Dia=eter, I 0.365 C.365 0.200 0.200 IAngth, b IBC, Total, Including gties, ?;=

< 76

< 76 e

541814

4 TA3LE k.1-2 FUEL ~MriTCRY TA3LE Cycles ITo of Initisi Initial Fuel krerage 3CC in Assen-E:.ich=ent (v/o)

Stacked Density Sur=up Core blies Fuel ?ree U

Pu

($ neeret'eal)

(%'D/ST) 5 6

?

3 52 0

94 16,6k1 5

2 F-Medified 3 51 0

94 16,703 5

2 G

3.c8 0.90 91.5*

22,851 4

12 F-Medified 3 51 0

94 13,711 3

18 G

3.08 0.90

' 91.5*

13,828 2

8 G

3 08 0 90 91.5*

6,518 2

14 G-1U 3 88 0

91.6*

5,863 1

6 G-1U 3.88 0

91.6*

^

0 1

16 G-3 3 14 0

91.5" 0

d

'Fallets 25 Dished 541815 12

j h.2 Mechanical Design The mechanical design of the Reicad G-3 fuel is ccesistens with the This reference G-1U hel with the exceptics of the upper tie plate.

plate was =cdified slightly in the Reload G-3 desis: to provide stan-dard fuel red location holes i= each cer:er 'of the plate replacing 1

the leching slots u.111:ed by the cobalt target rods. 37 letters dated, June 16, 1972 and Octcher 13, 1975, mechanical desis: analyses for Relcad G and Relcad G-1U fueli vere submitted; these analyses are applicable for Reload G-3 Ne1. Table 4.2-1 describes the G-3 hel assechly cc=;cuents, their pur;cse and cc=;csitics.

f 541816 13

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l TABLE k.2-1 DISCRIPTION o? ':T?E G-3 PUIL ASSE!BLY CCMPOIE3Ts Item Purcose Material / Rationale Upper Tie Plate and Maintains fuel red array.

Cast SS, Grade CF-3 Eardle Provides lifting fixture.

- Strength

- Corrosien Resistance Cc=pression Springs Accez::medates differential Incenel I-750 fael red lengths and sup-

- Corrosien Resistance e,rts upper tie plate.

- Strength at C;erating Cen-

~

ditions.

- Springs leaded high encugh

~

to si"4 *:e fretting and icv encugh act to cause excessive red beving.

Fuel Rod End Cap Prevides high q *'*ty seal TIG - Finet Eead

'ields of fael reds.

- Ircenent penetratics.

- Extremely icv peresit7

- Eigh strength integrity.

Plenus Spring Xaistm*-s ce=;act fael Incenel X-730 '41 e coh=in during >=ad' d E

- Jithstand autcclave treat =ent.

and ship;ir.g.

- Maia*ain spring lead during reactor c;eration.

Plenun Cha=her Conects fission gases.

- Assures /that gas pressure Provides space for axial vill not overstress ci ded g.

expansion of fcel.

Claddd g Contains fissica gases and Iirealcy-2 keeps water fres contacting

- lfi-d-4:e neutren abscr; tics.

fuel.'

- C1sedd g is auteclaved for pref *i # g for corresien i

resistance and to provide a.

corrosien-; reef test.

Penet Claddi=g Provides clearance betJeen

- Designed to -=* *ce fuel l

Gap feel and cladding.

red fissile centent and to d-d*d :e penet-clad i=ter-I actics frc= sve"

• g expected at high bcr up.

541817 tu

e i

TABLE h.2-1 (Cc=td)

Ita=

Purecse Material /Ratienale Insulator ?en et Reduces penet-cc=;enet Al 023 interface te=;eratue.

. - Maintains ta=perature belev those causing excessive stress levels asi below those of cen-cern with =etal-fael reaction.

- Controls hydride precei;tation.

At=csphere Heat transfer =ediu= be-Eeliu=

tseen pe net and clad.

- Gecd heat transfer character-istics.

- Prevides a= easy and reliable lesk detectic: c=1tering

=eans.

Spacers Maintains correct red-to-Zircalcy E Trs=e, Incesel 718 red spacing.

Springs

- Corrosien -d d-d eed.

- Mechanical stability.

- Spring leads en cladding =st be sufficient to d d 'ce

' lateral and rotational =cre-

=ent of fuel red but =st act cause excessive claddi:g or spring stress.

j

- Spacar =st not cause e ces-sive ccclant flow resistance.

i Inert Reds Displaces the highest peak Zircalcy #. Cladding Ici Caps

?*"**

clad temperature reds unde-LCCA conditicus and provide

- Corrosics resistance.

a radiatic sink.

- Lov abscr; tic cross section.

l 3cttru Tie Plate Ma1=tains fuel red array Cast SS,' Grade C7-3 and distributes ecolant

- Strength,

)

te fael reds.

- Cerrosien resistance.

Spacer Captr e Ecd Maintains correct le:gi-

- Centi =cus clad and tudinal ;csitica cf for=ed Zircalcy sheet

spacers, stock cc=:ectors.

I Tie Red Provides struct=si skelete: Ir-2 clad fuel reds vi$h end cf assembly by securing the fittings fcr attact=ent to I

upper and icver tie plates.

tie plates.

U 541818 1

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h.3 Ther=al Desi n The design basis for the ther=al perfer=ance of Relcad G-3 fuel is identi-cal to that described in our sut=1ttals dated June 16, 1972 and October 13, f

1975 for Reload G and G-1U, respectively, 4.h Chemical Desic The adequacy of =aterial:s selected for the chemical fuel design has been demonstrated through the excellent perfersance of Ixxon Nuclear Fuels to date. Past irradiation tests for assemblies s!*"ar to G-3 have produced no fuel failures or degradatica due to incompatibility with the reactor water chemistry. Results of the post-irradiatica embticus of fuel assembli,es, including those of the G design, are centained in Special Report No 24 dated November 24, 1976.

541819 e.

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5.0 NUCLIAR DISIGN o ;e C-lu and The fresh Mel to be used for Cycle 15 is Ix:xc Nuclear's r

Type G-3 Fourteen bundles of Type G-1U hel are currently used in Cycle 14.

I=portant differences between the reload G-lU and ' reload C-3 fuel bundle designs are the replacement of the four cerner ccealt target rods with lov enrichment fuel reds, a change 1: the gadolinia poison pi locations and changes in the busile enrich =ent distributica which reduce the bu=ile average enrir 5 -t from 3.88% for G-lU fuel to 3.1k% for G-3 hel. The effects of these changes on local peaking factor and fuel bu=C.e reactivity have been accounted for in co=puting the core physics characteristics.

51 Ph'nies Characteristics As discussed earlier, previous reload licensing sub=ittals fer the Big Rock Point Plant did =ct include =any of the physics para =eters requested in the " Guidance for Proposed License A=end=ents Relating to Refueling,"

(

thus these para =eters are ct available for previcus cycles and cen-sequently no reference cycle, =eeting the criteria of " reference cycle" as cefined in the guide, is available. It is the intent of Censu=ers

?cver Ccapacy to calculate the pars =eters required by the guide fer Cycle 15 and to utilice them is subsequent licensing sub=ittals as the reference cycle for ph7 sics parameters. These are included as Table 51-1

'nd Figure 5 1-1.

Table 51-1 includes the full ;cver do;;1er coefficient, delayed eutre fraction, void coefficient and total peaking facters for both the 3eginning of Cycle and end of Cycle Conditiens The - M -

reactivity for in sequence red drop vorth is 2.297, ak/k for 3CC and 1.715 Ak/k for ICC, both cases well belev the Technical Specificatica li=it of 2.5% ak/k.

""'e Cycle 15 core ca be mistained suberitical in the = cst reactive ec -

ditica thrcughout the operating cycle with the =cs reactive red Nily withdrawn and all other reds P'"y inserted.

"'e Technical Specificatics concerning shutdow: =argin is 0 3% ak/k which is sis ificantly less tha:

the 3CC and IOC values for shutdeve =arg1= listed in Table 51-1.

Figure

(

5.1-1 is the Mll shutdevn =argi curve for Cycle 15 541820

,T

The limiting scrsa reactivity curve for Cycle 15 is provided in Fig-ure 5 1-2.

This curve is most limiting at IOC vhen the control red densit/ in the critical rod configuration is the lowest. Figure 5.1-2 is a comparison of Cycle 15 scram reactivity to the bounding curre of Cycle 11. The Cycle 11 cur-te is considereCli=iting since it was uti-lized is the referenge analysis of the red d:c; accident, the accident most sensitive to scram insertion charnetaristics.

Two physics ; ara =eters requested by the guide are cet provided in this

~

submittal. The sederator tempersture coefficients are not ecesidered in any accident or trs=sient analysis, and essentially have no = easing for a boiling water reactor other than in its i=itial heatup prior to ;cver operation. However, these vill be calculated prior to ;cver operation for Cycle 15 The other parameter not provided is the verth of the standby liquid control systa=. The verth of the standby liquid centrol system is being evaluated for this cycle.

TABLZ 5.1-1 Psrsmeter 30C ECC Coppler Ccefficient ( Ak/k/5 Fever)

-7 06x10-I -T.65x10~I y= vd m Radial x Axial Peaking Factor 1.697 1.668 Mavd me Radial x Axial x Local Peaki=g Facter

2. kT9, 2.30 6 Maxim:n Rod Worth (5 ak/k) 2.29 1.7k

.00606

.00588 Delayed 3eutron Fractics 2.21 6.21 Sh'tdown Margin (5 Ak/k)

Void Coefficient (ak/k/ Unit Void)

.1663

.1127 52 Analytical Intut Reactor mver distrib'utions, reactivities, reactivity ecefficients, fuel burcup and margin to thermal limits are calculated with the G20% ce=puter GROK is a three-di=ensional coarse =esh reactor sim:later with program.

ther.a1 hydraulic feedback and is a derivative of the FLARE progrsm.

(D L Delp, et al, "TLARE, A TERII-DIME 2SICNAL 30ILIUG WATIR PJACTCR SDC-o LATOR," GEAP kS98, July 16,196k.) Ihe neutronics pars =eters

'.4, M~ and is 541821 1

local peaking factor, as a function of 1ccal operating state as ec=puted by the fuel designer, are the W ar inputs. Algorith=s have been in-cluded which calculate peak heat flux, !CE'R, MA?LEGR and theoretical flux vire traces for ecmparisc= vith reactor =easurements.

53 Changes in Nuclear Design S.ere are no changes is core design features, calenlational =ethods, data or infor=stica relevant to dete:mising i=portant nuclear design para =sters, other than those"=entiend?. above, for C7cle 15 19 541822 l

=

9 o

o.

FIGURE.5.1-1 SHUTDOWN MARGit. VS EX."OSURE CYCLE 15 7

6 9

<au M.

5 w

E O

4

=3 8

g 3

5 2

l 3

0 0

1 2

3 4

5 6

EXPOSURE (GWD/ T) 20 541823

.I m

n

Figure 5.1-2 BIG ROCK POINT PLANT SCRAM CURVE REFERENCE CYCLE (CYCLE 11)

END OF CYCLE 15 o 1.0 Y

r g 0.9 r

~

E 0.8 I

0.7 wU d 0.6

/

20.5 a

M 0.4

/

/

0.3 f

0 l

u.

0.2

/

O

//

/

1 Z o,;

0 1

0.5 1.0 1.5 2.0 2.5 3.0 m

TIME (SECONDS)

6.0 THI2 MAL EYDRAULIC DESIGN The hydraulic design of the relcad 0-3 assemblies is identical to that of the reload G-1U assemblies. The ther=al ;erfor=ance of reload G-3 fuel differs slightly fren relcad G-lU in that the average red ;cver has changed since the four cobalt corner rods have been replaced with 1

low enrichment UO fuel r ds. This has led to a decrease is average red 2

power, consequently it has also led to a decrease in **m=

cver;cver clad.and fuel t.e=;eratutes and an increase in the over;cwer =isi=u=

4 critical heat flux ratio. A ccmparisen of the ther=al hydraulic pa-rameters for ecaplete cores of relcad G-3 and reload G-lU fuel asse=-

blies is centained is Table 6-1.

Although Cycle 15 vill be predcminantly 11 x 11 reload G-type fuel as-semblies, approxi=ately 25% of the core vill be 9 x 9 : 1 cad F fuel assemblies. Ecvever, the *

• e radial peaking factor for the relcad F fuel for Cycle 15 is expected to be 0.8k.

This vill result is a 1225 rated ;cver **-4m m critical heat flux ratio of greater than 2.0 for small and le.rge crifice channel locatices. Therefore, si L'u = critical heat flux ratio units for Eelcad F fuels vill not 1 *-40 core ;cver c;-

ersticus for Cycle 15

%e l 22 541825 I

L

TABLE 6-1 Ther. al Eydraulie Para =eters (Core Contains All of Iach Type Fuel)

Relead 0-3 Reload 0-1U Core Conditions 0

Reference Design Thermal Output, (MW,)/(3tu/h) 2h0/8.191 10 Total Flow Rate, 2 /h 12.3 x IQ 0

Iffective now Rate for Eeat Transfer, Lb/h 9 9 x 10 System Pressure, Nominal in Steam Dome, Psia 1350 Assembly Descrittien Red Diameter, Inches 0.hk9 0.hk9 Red Pitch, Inches O.577 0.577 Number of Active Reds 117 113 Total Fuel Length ;er Assembly, Teet-682 5 659 2 Eest Transfer Area, Ft.

80.23 77.k8 2

2 2

0.163/23.hk 0.163/23.hk new Area, Ft /In Design Pcver Peaking Factors Traction Generated in Fuel, 5 97.0 96.6 Fuel Assembly Power 7 actor 1.k5 1.k5 Local Pesking Factor 1.20 1.20 Axial Peaking Factor 1 51 1 51 Ingineering Eeat nux Factor 1.0k 1.Ok Assembly Thermal Perfe=ance i

Max 1==m Eeating Rate,'44/ft, at 225 Overpover 13.53 14.0 leav%= Eesting Rate, '44/ft, at Rated Pever 11.09 11.kk Average Heat 1=g Rate, 't4/ft k.06 k.20 y=v 4 = Eest nux, 3tu/h-7,t at 22% overpower 392,900 ko7,000 2

leav % m Eest n ux, 3tu/h-Ft at Rated Pcver 322,100 333,600 Average Heat nux, Stu/h-Ft st Rated Peve-

" 7,900 127,100 Tempersture,

'F, at 225 Over;cver 3879 3900 Marm.un CO2

• Max 1=um Clad Temperature, *F, at Overpever 745 754 MC3FR at Overpower Conditions Axial Peak at I/L =.k5 1.68 1.63

. Coolant Subeccling at Core Inlet, 3tu/2 22.8 22.8 Assembly *:Pfdraulic Perferma.nce Average Assembly now kner Crifice Zone, 2 /h 132,900 132,900 Outer Orifice Zone, 2 /h 80,500 80,500 (2k Assemblies en Peripher7 of Cere) 0 Active Core now at Design Pever D/h 9 9 x 10 9 9 x 10 Hot Assembly Flov at 1225 Design Fever (Reference Design Flev) 123,000 123,000 Assembly aP at Average Design Power (Includes Orifice AP) 5 37 Psi 5.37 Psi Ect Assembly F.ngineering Isthal;y Rise Factor 1.10 1.10

• Crui-rree Surface 541826 23

1 l

1 l

I 1

1 l

7.0 TRANSIDT AND ACCIDDT ANAI.1' SIS l

In order to update this section, the NRC Standard Reviev Plans, Regulatory Guide 1.70, and the General Ilectric Standard Safety Analysis Report vere l

thoroughly researched to determine what accidents, trsesients and i'"4 ting design criteria vere necessary for a proper review. The " reference cycle" for accident and transient analysis consists of the latest anslysis run for each. In many cases these date back to the FRSR, but wherever analyses have been run subseque.t to this, they have been used as the reference cycle. The references for this section are ecstained in Subsection 7.3 7.1 Trsesinnt Analysis T.1.1 Significant Reacter Kinetics and Fuel Ther al Eydraulic Desig:

Parsmeters For each reactor trsusient ecesidered in previcus licensing sub=ittals for the Big Rock Pof:t Plant, the reactor kinetics para eters '.tich control the reactor transient response are shown in Table T-1.

Also shown in Table 7-1 are the reference value and the corres;cedit.g Cycle 15 value for each sigr.ificant pars =eter.

3elev is a discussics of the effects that the Cycle 15 vslues are expected to have ce the reactor transient res;cuse.

I=portant to the analysis of the reactor transients is the thermal and hydraulic design of the varicus feel bundles cesprising the reactor core. All Big Rock Point f el bundles up to and including the Relead G-3 feel bundles have been desig=ed to eet the fc11cving constraints.

(Refer to Section 6)

(1)

M4a4-critical heat flux rstic (MCEFR) at design ever;cver (122%) and design peaki=g facters =ust be greater than 1.50.

(2) Maximum fael temperature at lesign over;cver and desig: peaking factors =ust be less than the f el sel:1:s te=; erst =e.

Given

• hat these constraints are set, the the. a1 res; case of each feel type (ie, MCETR, peak fuel te=;ersture, peak clad te=;erstr e) to a given transient vill be as previcusly predicted or better.

~

2k 541827

e 4

TABI.E 7-1 Significant. Henctor Kinet.ica Parametera Nominal Cycle 15 Important Hererence Value Value Event Latest Analysta Kinetica Parameter (s)

• Ious of Ext.ernal Load With Hererence 1 Void Coefficient.

B00:

.208

.1663 EOC:

.10

.1127 8

and Wit.hout. Turbine Bypana Page 13 (Ak/k/ Unit Void)

(Bounda Main Steam I.ine Isolation Valve Closure

-0 and loss of Condensor Doppler coefficient

-5. l 2x10!Ak/k/5 Power

-7.06x10 Vucutus)

• Steam Pressure Hegulator Reference 1 Void Coefficient 110 0

.2 08

.1663 EOC:

.10'

.1127 Failure Heaulting in Page 10 (Ak/k/UnitiVoid)

Heduced St.eam Flow

-0

-E Doppler Coefficient

-5. l:2x10 Ak/k/$ Power

-7 06x10

• Uncont. rolled Hod With-y*

drawal Frosa Suberitical

-N

• Cold St. art.-Up Hererence 1 Doppler Coefficient

-1. is'. 10 Ak/k/*F

.95x10~

Page 6 Maximtun Henct.tvity 3.9% Ak/k 1 77%ss 3

Addition A

p/t*

175 183 H

~5

• Ilot GLart-Up Heference 1 Doppler Coefficient

-1.37x10 Ak/k/*F

.95x10 Page 6 g

Maxistun Heactivity la.2% Ak/k 2.29588 Addition p/t" 175 183 i

Void Coef'ficient 110 0 :

.20

.1663 8

• Uncont. rolled Hod Wit.h-Hererence 1' EOC:

.Y

.1127 Drawal at. Power Page 12 (Ak/k/ Unit Vold)

Dopplet-Coefficient.

-5.le2x10 Ak/k/% Power;

-7 06x10

~

7 e ?

TABI.E 7-1 (Cont.d)

Nominal Important cycle 15 Event Latest Analyala Kinet.ica Parameter (s)

Reference Value Value

• Inactive Hecirculation Heterence 1 Void Coefficient BOC:

.208

.1663 Pump Etart-Up Page 22 EOC:

.1'0 "

.1127 (Ak/k/t! nit Vold) 4 Doppler Coefficient

-5.42xlo ? Ak/k/%.'over

-7 06x10

• Ioss of Hecirculation (This event la reevaluated for each new core loading using the methods described in Appendix B of Reference 2.)

Piumpu I

.n

.A H

00 EC co aHererence 7 8'Heference value fs maximum worth for out-of-sequence rod. Cycle 15 value la maximum worth for in-sequence rod.

Line Pressurizatics Events (Less of Load, T.:.rbine Trip, Main Stet:

7 1.2 Isolatic Valve Cicsure, Stea= Pressure Reguitor Failure) voids

. Pressurizatica events are charatterized by a decrease 1:

The =cre negative void coefficient resulting 1: a power 1: crease.s mi:e the systen pressures a=i severs for BCC conditicas tends toSince the reference cycle SCC void coeffi-reached in these events.

cient is :ssre negative than expected at ar.y ti== during Cycle 15, the reference cycle analysis cesservatively bounds the upccming cycle.

Inactive Recirculat cs Pu=p Start-up or Cold Water Event i

7 1.3 Like the pressuri:stics events, this event is characterized by a decrease 1: voids and a power increase; therefore, the void coeffi-the i=;crtant kisetics pars =eter fer this event.

cie:t is agai:

Because the reference cycle SCC void coefficie t is =cre negative than expected at any ti=e during Cycle 15, the reference analysis bounds Cycle 15 7 1.k Loss of Recirculatics Pu=ps This eve =t is reenluated for every ecre relcading using the =ethod de Results of this analysis, although scribed i: A;;endix 3 cf Reference 2.

presently incomplete, are expected to shev, as they have sheve for flux ratic never falls previous cores, that the d-4-:= critical heat belev 1.5 and, 1: fact, =cact==1ca117 increase throughet.: the critical J

f Therefore, this event is not li=iting for

ortic of the transient.

Big Rock Point.

7 1.5 Red Withdrawal at Power This event is characterized by increases in core ;cver level and cere The void and dcypler coefficients are the i=portant kinetics voids.

As =cted in Table 7-1 the Cycle 15 parameters for this event.

do;;1er coefficient is =cre negative (ie, =cre e cuservative) tha:

In additics, the Cycle li vas assumed in the reference analysis.

void coefficient is =cre negative than the vers-case (ICC) void I

Therefore, it is cencluded ccetticient for the reference cycle.

that the reference cycle analysis tou=ds Cycle 15 for this event.

27 541880 g

aw

M

-e-

j l

4 i

. T.1.6 Start-Up Ivent he start-up event (or the uncentrolled red withdrsval from sub-critical) was analyzed is Reference 1 for both the cold and het standby initial conditions. The start-up event is charactericed by an extremely rapid increase is nuclear ;over (to approxi=ately 100 ti=es rated ;cver) fc11cved by an equally rapid ;cver reduction due to doppler feedback. D e is;crtant kinetics pars =eters for this even't are the doppler coefficient, the ratio of 3CA/1*,

and the maxi =um reactivity addition due to the withdrawal of a control rod while suheritics.l. As noted is Table T-1 en17 the Core 15 doppier coefficient is significantly noncenservative as De compared to the values assumed in the reference a alysis.

==N reactivity additics is =nch less than assu=ed for the reference analysis, and the 3EIA/f.* ratio is nearly the same as assumed in the reference analysis.

If, however, the event were reanalyzed ass" ' g the Cycle 15 value for the dc;;1er coefficient and assuming the sa=e red verths and 3CA/1* ratios as in the ref-erence esslysis, the ccusequences of this accident vculd still set Assuming a linear relationship between doppler coeffi-be severs.

cient and feel effective ta=;erature rise, the reduced Cere 15 dcPP-1er coefficient would result in fa.1 effective *emperstv.re rises of 1

900*F and 850*? (as compared to 580*? and 590*? is the reference analysis) for the cold and het start-up events, respectively.

- hus, assuming a het spor peaking factor of 3.0, the peak fuel temperatures of 2800*? and 3100*? for the cold and hot start-up events, respectively, vould still be significantly less than the nis is still extremely conservative since fuel melting temperature.

a hot spot peaking factor of 3 0 is significantly greater than vill be allowed during Cycle 15 based en ICCS '*-*tations.

7.2. Accident Analysis I

r l

T.2.1 Loss of Ccolant Accident (1CCA) ne Big Rock Point Loss of Ccolant Accident analysis fer Ex=en Nuclear Ccmpany (I3C) fuel was performed with I"C esiculational sedels which 541831 as

^

4 are consistent with the require =ents of A;;endix I of 10 C?R 50.

The app;c;riate asst =:pticus and results of the ICCS a:alysis for I

Relcad G-3 all-uranium feel vere documented in Reference 3.

nis report was submitted to the Director of Nuclear Reactor Regulation en February 18, 1977 is sup;crt of a. proposed Technical Specifica-tions change dated December 17, 1976 for updating MAPEGR limits for Exxon Relcad G and Reload G-1U fuel. ne sa=e report also in-cluded a' reanaly's'is of T. dss of Ccolant Accident for, Reload G and Reload G-1U fuel. ne '*m4 ting break sice for all three feel types 1

(Reload G-3, Reicad G and Relcad G-1U) was identified to be a 2

0.25 ft small recirculation line break. MAP EGR 1'

'ts as a fanc-tien of burnup vere also provided is this report. m 'dts for all other fael types (General Electric ? and Modified ?) which vill be reloaded into the core for Cycle 15 vill re=ais unchanged frem values approved by the NRC for previcus cycles (Reference 4), with one exception discussed in Sectics 3.0.

• 7 2.2 Red Drop Accident he centrol red drop accident has been previcusly analyzed in Reference 5 Se verst case (hot standby) *.as analyced for both an all-uranium core and a =ixed-czide core. Se i=pertant kinetics parameters for the control red d:cp accident are listed belev along with the values assu=ed in the analysis and the Cycle 15 values.

Assumed Value pars =eter Ursnius Cere Mixed-oxide Cere Core 15 Value Iffective Delayed Neutron Traction

.00591

.c0529

.00588

-5

-I

-I Doppler Coefficient

.916x10 ak/k/*?

.96x10 n/k/*?

.95x10 akik/ ?

M

• e verth of a 2.55 ak/k 2.5% ak/2 2.295 ak/t Single Centr:1 Red na Cycle 15 values for the i=percant kinetics ;arsmeters are very si"d'ar to the values assu=ed for both cores analyced in Reference 5 De Cycle 15 values of de;;1er coefficient asi 3ITA are bounded by the values assu=ed in the two analyses. 3ased cc this ec=; arisen, the reference analysis is censidered conservative for the upcc ing cycle.

l 29 541832 l

l

~

o T.2.3 A:ticipated Trs:sient 'Jithout Sers= (ATJS)

The ec: sequences of the =cs: 11=1 ting ATJS event, the icss of Icad without turbine bypass, were previously evaluated in References 1 and 6.

Sese analyses assu=ed a void coefficient =ch =cre negative

(.20 I > 'c/ unit void) than expected at any time during Cycle 15, 1

~0 and a &. gler coefficient mch less negative (-5.k2x10 ak/k/5) than expected during Cycle 15 Thus, the previous analyses are considered ccuservative for the upccming cycle.

T.3 References 1.

APID kc93, "Trsesiest Analysis, ccesu=ers ?cver Cc=pany 31g acek Point Plant," October 1962, and/or 31g Rock Point Final Eacards Su==ary Report.

2.

GZAP kk96, " Core Perfor=acce and Tra=sient Flev Testing - His Rock ? cia Boiling 'Jater Reactor," July 1965

~

3 I2-UF-76-55, Revisic: 1, "ICCS Analysis for Exxc= Nuclear Coc;any G-3 All Uranius Uo C.cbalt Fuel for Big acek Point (Including Rescalysis of Relcad G and G-1U Designs)," February 1977 k.

"3ig Rcck Point Plant Less-of-Coolant Accident Analysis for General Electric Fuel in Cc for=ance With ICC7R50 A;;endix E," July 11, 1975 (Suhmit.ed as A;;endix A to a Technical Specificatices change request fren Ccesumers ?cver Cc:spany to the NRC dated July.25,1975.)

5 Technicel Specificatices change request fres R 3 Sewell (C? Cc) to J ? 0' Leary (UTAIC) dated June 20, 197k.

6.

3EDZ-21065, " Anticipated Transients 'Jithout Scras Study for 31g acch

?cist ?cver Plant," Cetober 1975 T.

Proposed Technical Specificatic=s change dated January 17, 196k.

3 541833 5.

O G

=

4 PROPCSID !CDIFICATIONS TO TECH 3ICAL SPICI?!CATICUS 8.0 The proposed Technical Specifications are contained under Section I of In genersi, the changes consist of proposing specific this submittal.

parsmeters and drawings for the Reload G-3 fael for the 31g Rock Point prese.;;ed in Sections 1

~

i i

, Technical Specifications with justif cat onThe MAPLEGR limits, through 7 Specifications change posed MAPLEGR lisij:s contained in the Technical 17, 1976. Justification for these limits is con-request dated Decenber h Cc==issics tained in Exxon Report %2-NF-76-55, Revision 1, fc:-arded to t e There is also a sinor conectics to the MAFLEGR for on February IS,1977 3y letter dated July 25, 1975, ve

=odified ? Pael at a burnup of 25,C00.In Amendne t 10, dated June k, 197 proposed a MAPLEGR li=1t of 8.k.

We propose o correct this back value was incorrectly transposed to 8.7 to 8.k.

d One proposed change to the Technical Specificaticas has not been add That change is the deletics, frc= Section 5 2.1(b) and up to this point.

'Gd/T of Cc tained Urasi= for Table 8.2, of the li=itation for the "ld=* ---

This limitation f'.rst appeared is Censu=ers ?cver an Individual 3undle."

d June 1, p;cycsed Technical Specificaticas for the 31g Ecek ?cist Plant date Char-It was contained in Section 5 2.2, "Primeipal Calculated Nucles:

1962.

"is section er.stai ed specific pars =sters releva=t acteristics of the Core."

d void coefficients, to Cycle 1 for the 31g Rock ?cist reacter (eg, =cderator an

)

It dopplers, reactivity balance, average I&d/ Ten of ecstained urasi=, etc i

d with the is apparent that although the other design pars =eters assoc ate original core ec= position vere updated or deleted as necessary to acc for the different reload feels, the li=1tation on

-=%

ccel burnout was maintained intact for each subsequent relcad licensing sub=1ttal.

Consu=ers Pcver Compacy responded to a By letter dated January 20, 1977, gas release frc= fael pellets letter fro = Mr D L Iiemann concerning fissic:

In our response, ve esiculated the relevant pars =eters with high burnup.

for all G series relcad fuels for burnups rseging frc= 30,620 to 38,935 Che resuits This is vell above the desig: burnup of the f els.

!&d/ C.

of this analysis indicated that the fissica gas release =cdel bur =up had 541834 3

=

=-

- - - ~

' ' ' ' " ' * -+

very-little effect on the peak clad ta=;erature predictics (less than 1",

reduction in MA?LEGR ' *ts at end of life) and ccusequently insignificant effect on the 31g Rcek Point 1CCA analysis and, therefore, was of no safety cencern. We further indicated that these results veuld be censistent for the 9 x 9 fuels.

Attached as A;;endix 1 to this report is an analysis of the fission product inventor 7 change with increased burnup for the 31g Rock Point core. This a=alysis was conducted assu=ing a core average burnup of 30,000 MWD /STU, which is also higher than the desig: average burnup of the Relcad G fuel. The results of this analysis indicated that the radiation dose increases due to in-

' creased fissica product inventory were insig=ifiesnt and veuld remais in-significant (less than 15 of total dese) until the fuel burnups reacted approximately 15 x 10 Wd/T.

Further justificatica for deletics of the -

dm-burnup ' * 'tatica exists in the Standard Reviev Plans. Section k.2 states 1: part, "The cladding design should be such as to acccu=cdate the fissicc gas evolved in opera-tien, so that the fuel can reach design burnup withcut exceeding the cladad g structural desig criteria." 37 virtue of the preceding dis -

cussic: and analyses, Censu=ers Power Cc=pa=7 comeludes that it adequately

=eets the Stanciard Reviev Plas criteria and therefore no arbitrar7 burnup

' *-dtatics is necessa:7 Also, the Gen,eral Electric St.andard Tech =ical Specificaticus for Boiling Vater Reactors =akes no =entics of the -"d--

fuel bur =up allevable. Since 31g Rock ?cist is in the precess of eccver-sics to the standard for=at, we feel that censistency veuld also d'etate l

deleting this arbitrary ' *-#tation.

)

i Thus, based en safety analyses perfor=ed ecccerning fuel burnup and en guidance developed frem the Nuclear Regulatory Cc==issics in the fez = cf the Standard Reviev Plans and Standard Technical Specificatices, Censu=ers Power Ccmpany cencludes that the li=1tatica en =axi=== fue'. bur =up is u=-

necessa:7 and shculd be deleted frem the 31g Rock ?cist Te^-4 cal 3;eci-fications.

l b-32 541835 O

e 6

s

,,...,_y 3-

=

I

\\

90 START try pgCoaAx The testing and start-up progrs= pla= ed for the next refueling cutage and subsequent start-up vill include:

(1) Centrol red drive testing, as required by the Technical Specifica-tions, including scra= ti=es.

(2) Core shutdown margin verification with the sost reactive red with-drawn.

(3) Critical contre.~. red patten.

(k) Mecsure=ent of flux shapes during ;cver escalatic: and cet:; arisen to ec=puter predicticcs.

A brief explanation of these tests is contained belev.

Shutdevn Marg 1= 7erificatics:

Core shutdevn rargin is verified at the begd-d I of each cycle and dur-ing the first ecid shutdevn after 35,000 !Gd. generation. The Technical Specificatices require suberiticality to be demonstrated with the = cst reactive red withdrawn frc= the core ~ as well as an i=ediately a'4acer.t or more.

A= analytical dete=1:stics red k=w to contribute.003 %,ff of the highest worth red is =ade, as well as the =u=ber of actches of an.

i=nediately a4 acent red required to ccatribute.e5 ak/%.

35 e.2/% is 1

added to accou=t for a reactivity increase at ta=peratures higher tha:

the ambient ta=;erature at which the test is perfc=ed,(nor-=y 1Cd to 204). Plant precedures call for the individ"=' vithdrawal cf each centrol red in the core, plus at least the su=ber of =ctches specified in the

';hysics analysis cc an aQacent red to verify the analy ical deter '-= tics.

Ccre =enitoring is provided by gas-filled 3cron 10 lined pre;cr:10:a.1 counters and say be supple =ented by portable fission cha=bers positioned above the core when available cou=t rate is icv.

Rod Drive Scrs= Ti=e Testing:

Technical Specificaticus require =ents state the maxi u= cc= trol red drive scrs= ti=e frc= the fully vithdrav position to 905 of insertien shall not exceed 2.5 seconds. This requirement is verified at the begi ing of each cycle by attaching leads frc= a stri; chart recorder traveling at a predeter=ised rate to the position indication for the drive to be tested and to the 25 volt d-c signal frc= the high reacter pressure input en cae 13 l

541836 i

1 l

1

, e i

)

l

[t' of the

.ro safety chancels. The ccatrel red is fully vithdram a=d a high reactor pressure trip is si=ulated by re=cving pcver frc= the reac-tor pressure input to the safety system icgic circu!*ry. Scram time is measured on the strip chart from the ti=e of safety channel trip to full insertion. The test is repeated for all centrol reds using both safety channels.

Critical Red Patter =:

s Control red withdraval sequences and initial critical red patterns are analytically date.~ined at the begirming of each cycle. On ec=;1stien of cere reccustitution and shutdevn =argi verificatice, as initial criticality at a=bient conditices is perfor=ed. Is.egrated centr:1 red vorth curves and shutdown margin are a4usted based en the conditions of actus1 critical red patters. On attaining steady state equilibriu=

cceditions at rsted ;cver, a reactirity balance is perfor=ed cad the ac-1 tual critical red pattern is verified to be within 1.T ak/k of expected.

r Figure 91-1 shows the ecid critical ec= trol red ;stte. at 3CC aci 20C.

?cver Distribution Measurement:

Flux distrihutica =easure=ents are =ade during the escalatics to rated

cver after the beg - ' g of a new cycle by inserting cf ec;;er-citasi

=

d Fredictic s alley wires into the core and coesting the cepper activation.

of this flux vire activation ars =ade with GRCK, a thr'ee-di=ensic:a1 cce group diffusien theory code, usi=g actual operating ccediticas as input.

(See Section 5 2.)

Pcver distrihutica calculaticus are then a4usted based en the flux vire, GRCK calculation ec=parison. In-ccre instru=entation is calibrated to ecnform with flux vire =easure=ents. Ther=al hydraulf analysis based on the flux vire corrected ;over distrihution celculatices are ccupared I

to MAFLEGR, MCEFR and heat flux li=its to insure confc=ance with the Technical Specifica ices.

L M

3k 541837

,, ~ -.,

-,m--

i Figure 9.11 COLD CRITICAL ROD PATTERNS i

BOC A

B C

D E

F 1

4 6

4 6

J 2

6 0

0 0

0 4

N

'3 4

4 0

0 0

6 i

4 6

0 0

0 0

4

)

5 4

0 0

0 0

6 6

6 4

6 4

BRP CONTROL RCD POSITIONS 1

EOC i

A B

C D

E F

1 23 9

23 9

)

2 9

0 0

0 0

D N

3 23 0

0 0

0 9

4 9

0 0

0 0

23 5

23 0

0 0

0 9

f 6

9 23 9

23 BRP CONTROL ROD POSITIONS

\\_

0 is all in 23 is all cut,,,418 3 8 a

35

o

,e i

III. CC'TCLUSIONS 3ased on the foregoing, Big Rock Point Plant Review Cor:=1ttee has con-cluded that this change does not involve an unreviewed safety questien.

CCNSU!ERS PO'4R CCMPANY By C R 3ilby (Sic ed)

C R Bilby, Vice President Production Is Transmission ~

Sworn and subscribed to before ne this 15th day of April 197T.

(SEAL)

Linda R Thayer (Siced)

Linda a Thayer, Notary Public Jackson County, Michigan My cou:=ission expires July 9,1979 4

e Y

(

1

'6 541839

~

Appendix 1 FISSICU PRODUCT EM_uCRY CHANGE WITE ETCRI.ASED SU?ET - 3IG RCC POINT 3ACKGR0tJND Radiation dose analysis for the "MCA" in the FESR assu=es 1/3 core at 5000 Wd/T,1/3 at 10,000 Wa/T and 1/3 at 15,000 Wi/T. All nuclides which are significant in radiatica dose centribution (Table A-1), with the exception of D-85 and I-129, have achieved steady state equd' <brium at the icvest burnup utili:ed in the FESR. Consequently, only D-85 and I-129 centinue to increase as burnup increases.

CALCtJI.ATIONS The FESR does cet list specific quantities for core inventory but activities for the 105 release case vere dete_ d ed for an earlier analysis (Table A-1) to be 1.k1 x 10 Curies of I-129 (105 of core inventory) and 1.88 x 10 C ries

~

of Kr-85 (30% of core inventery) after two years of full ;cver operatics. Two years were chosen to proMde conservative inventories of Kr-85 and I-129 equal to a full core at 15,000 Wa/T. Activities are calculated for 30,c00 Wd/T by-Equation I:

n" Activity = ( * " *

• 1. ) (?) (1) (240 W ) (GI.) (1-e~

) Iquatica I vhere: Y = Fissien Yield 1 = Decay Ccustant (See )

GR = 0.1 for I-129, 0 3 for D-85 T = k Years Irradistics (1.26 x 10 Sec)

D-85 Activity = 3 53 x 10 O uCi Released to Centa1=ent 3

I-129 Activity = 2.82 x 10 uC1 Released to Centai=ent Centribution to off-site dose frca leakage of the above quantities at -=vd-"-

containment leak rate (?E5R Firres 13 2 and 13 3) of L.3 x 10-9/see is per-fe:=ed as fc11cvs:

D-85: Rad /h = 0.35 Y (I/Q) (Q/Sec) (

• )

Iquatics I~

Y h

where:

I/Q

= Diffe.sion Constant for Ground Level Release at 8k2=, Frcs Her:J.atory 3

Guide 1.3 (k.5 x 10 Sec/m )*

541840 I

s t

NCTI:

• 8.6 x 10 frcs Titre 3.A divided by a vake correctics facter of 1.9 frc= Tirre 2.

1

e Q/See =* Release Rate = (k.3 x 10~9/See) (1.58 x 10 C1) = 8.1 x 10~3 Ci/Se 5 = Average Energy per Disintegratics (0.002 Mev)

Y

~

Frem Iquation II, Kr-85 dose rate equals 6.6 x 10 rad /h to the total body.

In comparison, FESR Section 13 11.4 indicates the ~ M ' = dese rate free the plume (all com;cnents) is 5 x 10-3 rad /h.

Thus', the percentage due to Kr-85 is insignificant at (6.6 x 10 /5 x 10-3) (1C05) = 0.00135.

I-129: Rad /h = (K) -(I/Q) (Q4Sec) (3)

Equation III vhere: I = Dese Conversion Facter per Regulatory Guide 1.109 (5 55 Rem /uci Inhaled) 3 X/Q = Diffusien Cc=stant (k.5 x 10 Sec/s )

~

~9 3

Q/See = Release Rate = (k.3 x 10 /See) (1.k1 x 10 uC1) = 6.1 x 10 uC1/Sec 3 = 3reathing Rate per Regulatory Guide 1.3 (1.25 =3/h)

Frem Equatien III, I-129 thyroid 6:se ec

• t=en free one hour of i:halation

-6

-6 equals 19 x 10 rs=, or 3.8 x 10 res frem two hours of ex;csure. I ec=-

arison, FESR Section 13.14.3 indicates a 2-hour thyroid dose of 2 rs= is ex-

ected frem the total halc6en '* ".ure.

Thus, I-129 is insignificant at (3.8 x 10 /2) (100%) = 0.000195.

CONCLUSION Dese ecstributions fres nuclides affected by increased fuel burnu; are negligible relative to total "MCA" deses. Radiation dese increases due to increased prcduc-tics of I-129 and Kr-85 vill remain insignificant (less thss 1",of total dese) 6 up te core bur =u;s of a;;rox1= ate 1715 x 10.wa/T.

(,

541841 2

e

"'A3LE A-1 Full Core Fuel Fin Gap Radicactivity Released Gac Activity (uct)

Big acek Point Plant Initial Design Calculatices Frcs quanicassee Design Data GI - Meditied for 80/20 Raticed to 31g Rock Isetece 1(Min )

U-235 Pu-238 Fissien Mixture

?cint Pever Levels 1.hu+05 I-129 7 762-14 I-131 5 98E-05 5.62E+12 5 92E+12 I-132 5.-03 8.35I+12 9 02E+12 I-133 5.69E-Ok 1.37I+13 1.33I+13 I-134 1.33I-02 1.45E+13 1.56E+13 I-135 1 73E-03 1.29E+13 1.21I+13 Ie-138 4.882-02

, 1.2 H+12 1.2H+12 Ir-87 9.'0'-03 k.2E+n 5 : ' ' '+'n Ir-88 4.1kE-03 6.k3I+11 7 27I+11 Ir-85m 2.622-02

2. 39I+n 2.66E+n 1.37I+12 3.ThE+11 Ie-135 1.26I-03 i

Ie-133 9.'o'-05 1.36E+12 1.37I+12 Ie-.143 k.33E+01 8.+09 Kr-9k k.16E+01 1.80E+10 D-93 3 22E+01 9 3hE+10 Ie-1k1 2.k2E+01 2.23I+n Kr-92 2.26E+01 3.2kE+n Ir-91 4.8kE+00 5.81E+ n Ie-140 3.06E+00 6.89E+n

~

Ir-90 1.29E+00 3.52E+n Ie-139 1.0hE+00 9.TkE+n Ir-89 2.182-01 8.09E+n Ie-137 1.81E-01 1.20E+12 Ie-135n k.k2I-02 2.12E+n 3.68I+11 Ir-83m 6.18E-03 9.82E+10 Ie-133m 2.13E-ok 3.82E+10 3.kTI+10 Ie-131s k.03I-05 3.62I+09 k.512+09 Ir-85 1.22I-07 1.88I+10 2.02E+10 Note: For I-129 and Ir-85 erd' *trium is cet obtained.

Eence, the activity available is the eq"*' *briu value e.t full pcVer operstice for tvc years s"

ti=es (1 - e-AO) where t is two years.

1 3

541842

.