ML20217M731

Certificate of Compliance 9228,rev 17,for Model 2000 Package

ML20217M731
Person / Time
Site:	07109228
Issue date:	10/21/1999
From:	Brach E NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20217M724	List: ML20217M720 ML20217M731
References
NUDOCS 9910280018
Download: ML20217M731 (12)

Text

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g NRC FORM 618 U.s. NUCLEAR REGULATORY Commission El iN CERTIFICATE OF COMPLIANCE Nl FOR RADIOACTIVE MATERIALS PACKAGES E

g I a CERTinCATE NUMBER

(

b REVlilONNUMBER c PACKAGE IDENTinCATION NUMBER d PAGE NUMBER

c. TOT AL NUMBER PAGES l

1. l 9228 17 USA /9228/B(U)F-85 1

8 l ki f 2 PREAMBLt~

ll I

N Nl a This certificate is issued to cerufy that the packaging and contents desenbed in licm 5 below. meets the apphcable safety standards set fonh in Title 10.

N!

g Code of Federal Regulations. Part 71. " Packaging and Transportation of Radioactive Matenal?

ll 1

b. This certificate does not reheve the consignor from comphance with any requirement of the regulations of the U.S. Department of Transportation or other q

apphcable regulatory agencies includmg the government of any country through or into whach the package will be transponed.

kj 7THIS CERTinCATE IS 155UED ON THE B A515 OF A SATITY AN ALYSIS REPORT OF THE PACKAGE DESIGN OR a IS$UED TO(Ams and AJdres2J b TTTLE AND IDENTinCATION OF REPORT OR APPLICATION:

kI

.g g'

j General Electric Company General Electric Company application g

g Vallecitos Nuclear Center dated May 19,1988, as supplemented.

i P.O. Box 460, Vallecitos Road g

g Pleasanton, CA 94566 g

c. DOCKET NUMBER 71 9228 g

? 4 CONDITIONS g-j 1

This centficate is conditional upon fulrilhng the requirements of 10 CTR Pan 71. as applicable. and the conditions specified below.

g j

?g 5.

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Nl E

E (a)

Packaging I

e (1)

Model No.: 2000 gl g

5 E

3 (2)

Description g

5 g

A steel encased lead shielded shipping cask. The cask is within a double-walled overpack g

9 with toroidal shellimpact limiters at each end. The overall dimensions are approximately (I

131.5 inches in height and 72.0 inches in diameter. The cask is transported in the upright E

S; or horizontai position. The gross weight of the package is approximately 33,550 lbs.

E-Tj E

1 h

The cask is constructed of two concentric 1-inch thick 304 stainless steel cylindrical shells E

p (ASTM A 240) joined at the bottom end to a 6-inch thick 304 stainless steel forging (ASTM N

y A 182). The annulus between the two shells is filled with lead approximately 4 inches thick.

2 The cask is approximately 71.0 inches in height and has an outer diameter of 38.5 inches.

g j

The cask cavity is approximately 26.5 inches in diameter and 54.0 inches deep.

g 5

The cask lid is 304 stainless steel and lead, has a stepped design, and is fully recessed g

pf into the cask top flange. The lid is secured to the cask body by 15,1.25-inch diameter g

9 socket head screws. The cask is sealed by elastomeric O-rings bonded to a thin aluminum g

g s

disc-shaped ring. The cask is equipped with a seal fest port on the side of the cask body, a E

j h

vent port in the cask lid, and a drain port near the bottom of the cask.

El

1. l E

I l

The cask is positioned within an overpack constructed from two 0.5-inch thick concentric E

h 304 stainless steel cylindrical shells (ASTM A 240). The shells are separated radially by E

t jl eight equally spaced tubes and horizontally by two tube sections. A 304 stainless steel f

al toroidal shellimpact limiter is attached to each end of the overpack. The overpack opens gl j'

just above the lower impact limiter for access to the cask. The top of the overpack is joined gl 3

to the base by 15,1-3/8-inch diameter shoulder screws.

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l 9910280018 991021 E

Ell PDR ADOCK 07109228 a

C PDR g

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1. w w w # # w w w w w w c FORM 618A coNDrnoNs <<, um-se u.S. NUCLEAR REGULATORY Commission g

l Pcg3 2 - C:rtificata No. 9228 - Rsvision No.17 - Dock;t No. 71-9228 El 5;I E '

I h

5(a)(2).

Description (Continued) f d

gl Gussets on the top and bottom impact limiters provide tie-down points for the package.

d The cask body is equipped with attachment plates for lifting devices. The cask lifting s

devices are detached during transport.

s, (l

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s, (3)

Drawings E

sj E

s (i)

The packaging is constructed and assembled in accordance with General Electric E

j 5l Company Drawing Nos.129D4946, Rev.10; 105E9520, Rev. 4; and 105E9521, E

5 Rev. 5.

E s

E fl (ii)

Packaging Serial No. 2001 is constructed and assembled in occordance with l

Generai Electric Company Drawing Nos.129D4946, Rev.10; 101E8718, Rev.12; and 101E8719, Rev.12.

y I

g i.

E 9

(iii)

The HFIR fuel basket and liner are constructed and assembled in cecordance with g

i General Electric Company Drawing No.105E9523, Rev. 3.

q E

s (iv)

The multifunctional rack is constructed and assembled in accordance with General E

i 5;

Electric Company Drawing No.105E9555, Rev. 2.

E

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E h

(v)

The barrel rack is constructed and assembled in accordance with General Electric E

s'l Company Drawing No.166D8066, Rev.1.

E E

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(vi)

The material basket is constructed in accordance with General Electric Company y

Drawing No.183C8356, Rev.1. The material basket may be used with the g

9 multifunctional rack and the barrel rack.

g E

3 (vii)

The TSR fuel basket is constructed and assembled in accordance with General q

s Electric Company Drawing No.105E9560, Rev. 2.

E D

E s

(viii)

The MTR fuel basket is constructed and assembled in accordance with General E

f Electric Company Drawing No.105E9557, Rev. 9.

E e

E (b)

Contents h

(1)

Type and form of material U

y E

s E

g (i)

Irradiated fuel roda which may be cut or segmented.

g s'

5 p

(ii)

Byproduct, source, or special nuclear material in solid form.

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E D

(iii)

Irradiated High Flux Isotope Reactor (HFIR) fuel assembly, positioned within the El h

HFIR fuel basket and liner as specified in E(a)(3). The HFIR fuel assembly is E

i fabricated in accordance with Oak Ridge National Laboratory Drawing Nos.

E I

M-11524-OH-101-D, Rev. O, and M-11524-OH-102-D, Rev. O.

E 4

E ;

f h (iv)

Irradiated Tower Shielding Reactor (TSR) fuel elements, positioned within the TSR h

p fuel basket specified in 5(a)(3).

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• m & 2 m 2

• 2 m 10 mjphlyR2 jim'1 & m & 4 NLm m m mf

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$I NR'c FORM 61&A mNDm0N5 amm,h tim. NUCLEAR REGULATORY Commission ci om l

$1 P ge 3 - C:rtificat] No. 9228 - R; vision No.17 - Dock;t No. 71-9228 q

Il

$i 5.(b),1)

Type e:.d form of material (Continued)

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(v)

Irradiated MTR type fuel assemblies, positioned within the MTR fuel basket specified in 5(a)(3).

+I The fuel assemblies may be sectioned only in the non-fuel bearing region of the assembly. The sf fuel assemblies are composed of aluminum clad plates, and are limited as follows:

+,

i

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Fuel material SQ,

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Max. uranium enrichment

[li (w/o U-235) 94.0 94.0 95.0 El il Max. active fuel thickness (in) 0.023 0.020 0.020 il j!

Min. clad thickness (in) 0.014 0.015 0.015 le El Max. U-235 per fuel assembly (g) 355 290 110

• ls Max. U-235 mass per g

j fuel basket cell (g) 710 580 220 Il

?i Max. burnup (GWd/MTU) 568 568 568 fi

[I Min. cool time (days) 120 120 120 I

SI jI Fuel material QSh UAl.

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Max. uranium enrichment 9

(w/o U-235) 20.0 20.0 5

Max. active fuel thickness (in) 0.020 0.069

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?l Min. clad thickness (in) 0.015 0.015 Il

?i Max. U-235 per fuel assembly (g) 347 150 E

i Max. U-235 mass per jI fuel basket cell (g) 694 300 s

)I Max. burnup (GWd/MTU) 122 122 4j Min. cool time (days) 120 120

?

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al;l Note: The enrichments, masses, and dimensions shall he based on values prior to irradiation.

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I g NRc FORM 618A CONDinoNs #camids U.s. NUCLEAR REGULATORY Commission J G M)

Pcg2 4 - C rtificate No. 9228 - R; vision No.17 - Dock;t No. 71-9228 f

i (vi)

Irradiated TRIGA fuel elements, positioned with the MTR fuel basket specified in 5(a)(3). The fuel

,I material consists of UZrH, in cylindrical elements, with aluminum, steintess steel, or inconel 1

9 cladding. The H to Zr ratio in the fuel ranges from approximately 1.0 to 1.7. Some fuel elements l

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contain graphite reflectors in each end of the fuel element. The fuel elements are limited as l

l e

follows:

AI l

q 9;

Approximate rod 1-1/2 1/2 1-1/2 1-1/2 1/2

'l diameter (in) b Graphite reflectors with or w th or wth wth wthout without without reflectors renectors reflectors 9

reflectors reflectors 9

9 Uranium concentration 8-45 10-45 8.5 min.

8.5 min.

10 min.

I in fuel (w/o U) e

[l C

Max. rod length (in) 30 30 30 30 30

l)

Max. active fuel il length (in) 15 22 15 15 22 19 l

Min. clad thickness (in) 0.02 0.016 0.02 0.02 0.016 e

Al h

Max. uranium enrichment I

h (w/o U 235) 20.0 20.0 70.0 94.0 94.0 i

"I h

Max. active fuel

?j diameter (in) 1.435 0.51 1.435 1.435 0.51 "I

!i Max. U-235 per rod (g) 165 44 140 220 44 Ij (max.15 rods per (max.15 rods per yj basket cell) basket ce!!)

Il S

33 33 h*j (max. 20 rods per (max. 20 rods per g

basket cell) basket cell) 21 5

Max. U-235 mass per fuel basket cell (g) 560 660 560 660 660 e

["l Max. burnup (GWd/MTU) 427 427 427 568 568 il il Min. cool time (days) 120 120 120 120 120 g

Note: The enrichments, masses, and dimensions shall be based on values prior to irradiation.

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P:g) 5 - C;rtificata No. 9228 - R2 vision No.17 - Docktt No. 71-9228 g

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'5.b(2)

. Maximum quantity of material per package gj' Nl lg Not to exceed 5,450 lbs, including fuel baskets, carrier racks, shoring, secondary 4

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containers, and shielding liner.

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(i) For the contents described in 5(b)(1)(i):

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1. l 600 watts decay heat; and

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Fissile contents not to exceed 1175 grams U-235 equivalent maa +ith initial k

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enrichment not to exceed 5 weight percent in the fissile isotope; n.inimum pellet f

g diameter of 0.3 inch, maximum bumup of 45 GWd/MTU, and minimum cooling time gl 9

of 120 days; or gl

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Fissile contents not to exceed 1750 grams U-235 equivalent mass with initial gl gI enrichment not to exceed 5 weight percent in the fissile isotope; minimum pellet

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i diameter of 0.35 inch, maximum burnup of 38 GWd/MTU, and minimum cooling N;

i time of 120 days. Fuel rods must be contained in closed,5 inch schedule 40 pipe, dl nl i

with a maximum of 437.5 grams U-235 equivalent per pipe; or El m

Fissile contents not to exceed 242 grams U-235 equivalent mass with initial I

i enrichment not to exceed 5 weight percent in the fissile isotope; minimum pellet k

y diameter of 0.3 inch, maximum burnup of 52 GWd/MTU, and minimum cooling time f

I of 180 days.

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(ii) For the contents described in 5(b)(1)(ii):

gl s

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lp 2000 watts decay heat Fissile contents not to exceed 500 grams U-235 equivalent ll, I

p mass. Carrier racks specified in 5(a)(3)(lv) or 5(a)(3)(v) must be used for conte its ll Is exceeding 600 watts decay heat per package, d

M N,

s (iii)

For the contents described in 5(b)(1)(iii):

N I

I One HFIR fuel assembly. The fuel assembly is composed of one inner fuel k

$1 element, with up to 2628 grams U-235, and one outer fuel element, with up to 6872 Ul

[l grams U-235. The mar.imum uranium enrichment is 93.2 weight percent U-235.

f The maximum burnup per assembly is 2300 mwd, the minimum cool time is two gl j

years. Decay heat not to exceed 600 watts per package.

gl a

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(iv)

For the contents described in 5(b)(1)(iv):

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A maximum of 4393 grams U-235 per package. The maximum dranium enrichment E{

bl is 94.0 weight percent U-235. Decay heat not to exceed 35 watts per package.

Nl D

The TSR fuel element', must be positioned and limited within the TSR fuel basket as Nl B

follows:

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5 I

Lower fuel basket section - Up to 4 upper or lower fuel elements, or a combination f si of upper and lower fuel elements, for a total U-235 ql jl mass of 1412 grams.

gl k'

R k m m m m m m m m m m mm m m m m m m m m m m m n

f w r W F w Ist w w is w is< is< w Tsi w w w

• m w w w w w w w w m W m

• w isr W Ts In< w is< + is W w

• a e is< is I

l g NRC FORM 618A cuNumoss awwuwds U.s. NUCLEAR REGULATORY CoMMissaoN J 43 96) l Pcg] 6 - C:rtificat] No. 9228 - R; vision No.17 - Docket No. 71-9228 l

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5

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lp 5.(b)(2)

Maxirnum quantity of material per package (Continued) s!

Middle fuel basket section - Up to 4 fuel cover (lune) plates, for a total U-235 mass

1. l of 304 grams.

s;j Upper fuel basket section - Up to 6 annular fue! elements plus one cylindrical fuel y

element, for a total U-235 mass of 2677 grams.

a p

(v)

For the contents described in 5(b)(1)(v):

-l h

Weight of contents, including fuel elements, spacers, shoring, and hardware, not to j

exceed 42.8 lbs per fuel basket cell.

a Ij!

Decay heat not to exceed any of the following: 1500 watts per package,120 watts p

per cell,35 watts per cellin the upper half of the fuel basket,85 watts per cellin the y

lower half of the fuel basket,765 watts in the lower half of the fuel basket (i.e., the B

lower half of all 21 cells combined).

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fl Failed fuel elements are permitted provided the damage is limited to cladding 9

defects due to corrosion, nicks, and scratches. Failed fuel elements must be

$l structurally and geometrically intact.

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(vi)

For the contents described in 5(b)(1)(vi):

l ll Weight of contents, including fuel elements, spacers, shoring, and hardware, not to y

exceed 42.8 lbs per fuel basket cell.

B h

For stainless steel and inconel clad fuel, decay heat not to exceed any of the h

following: 1500 watts per package,120 watts per cell,35 watts per cellin the upper 9

half of the fuel basket,85 watts per cellin the lower half of the fuel basket,765 s

watts in the lower half of the fuel basket (i.e., the lower half of all 21 cells combined).

j s'

i For aluminum clad fuel, decay heat not to exceed either of the following: 630 watts y

per package,30 watts per cell.

(c)

Transport Index for Criticality Control i

i Minimum transport index to be shown on il label for nuclear criticality control:

EI El For the contents described in 5(b)(1)(i),

b 5(b)(1)(ii), and 5(b)(1)(iii); and limited 4

in 5(b)(2)(i),5(b)(2)(ii), and 5(b)(2)(iii):

100 D

For the contents described in 5(b)(1)(iv),

5(b)(1)(v), and 5(b)(1)(vi); and limited in j

l a

5(b)(2)(iv),5(b)(2)(v), and 5(b)(2)(vi):

0.0 l

51 M

6.

Plutonium in excess of twenty curies per package must be in the form of metal, metal alloy or y

reactor fuel elements.

i

$mmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmm+mmmma'

r Ni E F W W Xi W 37 W W Wi W W Ei W IEi W W BI W Wi W W W W W W W W W W W M7 W W W W E E W W W W W W E p gRC FORM 618A CONDrnoNsimmmwm U.s. NUCLEAR RE*uLAToRY COMM!ssioN g

P g] 7 - Certificat] Ns. 9228 - R:Ivision No.17 - Dock;t No. 71-9228 El i

I:

Nj; 7.

The U-235 equivalent mass is determined by U-235 mass plus 1.66 times U-233 mass plus 1.66 E

f,i f,

. times Pu mass.

I g

8.

Bolt torque:

4' pl

' The cask lid bolts must be torqued to 690 ft-Ibs (lubricated).

g

1. l 5

The bolts used to secure the top of the overpack to the overpack base must be torqued to 100 ft-p B

lbs (dry).

R N

I, D.

(a)

For any package containing organic or inorganic substances which could radiolytically I'

W generate combustible gases, determination must be made by tests and measurements or I.

I by analysis of a representative package such that the following criteria are met over a I

period of time that is twice the expected shipment time:

I (i)

The hydrogen generated must be limited to a molar quantity that would be no more f

f gl than 5% by volume (or equivalent limits for other inflammable gases) of the g

g secondary container gas void if present at STP (i.e., no more than 0.063 g-moles /ft' g.

pl at 14.7 psia and 70*F); or g

hj I

9 (ii) -

The secondary container and cask cavity must be inerted with a diluent to essure l

9 that oxygen must be limited to '5% by volume in those portions of the package which I

B could have hydrogen greater than 5%.

I t

sj I

sj For any package delivered to a carrier for transport, the secondary container must be I

Ej prepared for shipment in the same manner in which determination for gas generation is I

?l made. Shipment period begins when the package is prepared (sealed) and must be I

h completed within twice the expected shipment time.

l a

g j

(b)

For any package containing materials with a radioactivity concentration not eteeding that g

g for low specific activity material, and shipped within 10 days of preparation, or within 10 l

g p

days after venting of drums or other secondary containers, the determination in (a) above g

I W

need not be made, and the time restriction in (a) above does not apply.

l 5

i B;

10.

Prior to each shipment (except for contents meeting the requirements of special form radioactive I

B material), the package must be leak tested to 1 x 108 std cm*/sec. Prior to first use, after the third I j I

use, and at least once within the 12-month period prior to each subsequent use, the package must 1

I be leak tested to 1 x 104 std cm'/sec.

I l B

B l l

11.

The cask must be vacuum dried prior to shipment if contents are loaded under water, or if water is l'

g introduced into the cask cavity. During shipments for which vacuum drying is performed, the cask g

cavity must be filled with helium.

5 I

h 12.

In abition to the requirements of Subpart G of 10 CFR Part 71:

i I

l B

B (a)

Prior to each shipment the cask seal must be inspected. The seat must be replaced with a B:

W new sealif inspection shows any defects or every 12 months, whichever occurs first; and i

B t

I (b)

Each package must meet the Acceptance Tests and Maintenance Program of Chapter 8 of I

the application, except that inspections in Section 8.2 of the application must be performed I

1 at least once within the 12-month period prior to each use; and j

f 9

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s< w 3 E w 10 Tsi + w w ist w is< T ist w w w w ini w w isi Tsi w w w w w is Is, is, is, is, is< Tai isi is, ini i in, w w 1s< a e is f NRC FORM 618A CONDmoNs ace =rmue u.s. NUCLEAR RE2ulAToRY CoMMilSloN gl

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?t Pag) 8 - C;rtificate No. 9228 - R; vision No.17 - Dock;t No. 71-9228 Nl 5

E' l

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12.

In addition to the requirements of Subpart G of 10 CFR Part 71: (Continued) l g

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g

/

(c)

' The package must be prepared for shipment and operated in accordance with the

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(

/

Operating Procedures of Chapter 7 of the application.

g I

3, 13.

Appropriate carrier racks or shoring must be provided to minimize movement of contents during

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W accident conditions of transport. A lead liner, as shown in General Electric Company Drawing No.

El s

129D4922, Rev. 2, which was included in the March 29,1989, supplement, raay be used inside the U

s;j ll cask.

E 9

N 1

8 14.

Each batch of ethylene propylene seals must be tested in accordance with Section 8.1.4.2 of the f

j application.

g 15.

Fissile mass limits for reactor fuel are based on fissi!e mass prior to irradiation.

l J

/

16.

For the contents described in 5(b)(1)(v) and 5(b)(1)(vi), the package may be transpor'ed qll

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horizontally. For horizontal transport, the package must be secured to the truck bed with the top q

M end of the package (closure end) facing the front (cab) of tne truck, d

y g.

M 17.

The package authorized by this certificate is hereby approved for use under the general license d

9;;

provisions of 10 CFR 971.12.

Nl 9

G ll 18.

Expiration date: July 31,2000.

U u-REFERENCES s

p General Electric Company application dated May 19,1988.

d g

l Supplements dated: March 29 and August 24,1989; May 30, October 11, and December 12,1990; May q

M 22,1991; June 8, July 27, August 4, and December 9,1993; April 29 and July 28,1994; May 19 and 30, g

M 1995; October 31,1996; May 9 June 3, July 14, August 14, and October 7,1997; May 15, June 2 (3),

W M

June 24, October 6, November 10,1998; and February 3, April 29, May 26 and 27, June 9,12,18,22,25, I

f July 27 and 29, and October 18,1999.

l 8

s N

h FOR THE U.S. NUCLEAR REGULATORY COMMISSION U

l d

i h

4 q

j E. William Brach, Director

(.

M Spent Fuel Project Office d

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Office of Nuclear Material Safety I

)

and Safeguards I

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Date:

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s am m m m m m m m m m mm m m m m m m is1 m m m m m m m m

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4 UNITED STATES y

, *j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2tEW0001

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APPROVAL RECORD Model No. 2000 Package Certificate of Compilance No. 9228 Revision No.17 By application dated October 18,1999, General Electric Company requested an amendment to Certificate of Gompliance No. 9228 for the Model No. 2000 package. The applicant requested that the certificate be amended to authorize an MTR-type fuel assembly with the following parameters:

Fuel material

UAI, Maximum uranium enrichment (w/o U-235) 20.0 Maximum active fuel thickness (in) 0.069 Minimum clad thickness (in) 0.015 Maximum U-235 per fuel assembly (g) 150 Maximum U-235 mass per fuel basket cell (g) 300 Maximum burnup (GWd/MTU) 122 Minimum cool time (days) 120 Except for the criticality evaluation, the new fuel type is bounded by the evaluations for previously approved contents. A new criticality evaluation is necessary because the enrichment / fuel dimension / fuel mass combinatiori of the new content is not enveloped by the criticality analyses for prulously approved contents.

CRITICALITY EVALUATION The applicant performed a criticality analysis for the new MTR fuel. The applicant used the GEMER01V computer code to show that the package, with the new content, meets 10 CFR Part 71 criticality requirements for a Transport index for Nuclear Criticality Control of 0.0. The calculations considered an infinite array of packages with the degree of moderation varied from dry to fully flooded with water. Therefore, the criticality analysis applies to both normal conditions of transport and hypothetical accident conditions. The applicant evaluated the MTR fuel with water reflecting the package containment t,ystem, as required by 10 CFR 71.55(b)(3). The applicant determined that the packcge provides greater reflection than water.

The fuel elements are low enriched U-Al fuel plates. The plate thickness is 0.069 inches with 10 plates per fuel assembly. The fuel elements and packaging were discretely modeled. The model of the packaging is similar to the previous models, including the shorter basket. Like the previous models, the applicant assumed that one end of the fuel elements protrude from the basket because the MTR fuel elements are stacked two high. The poison plates were modeled _

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using the minimum dimensions (including tolerances) shown in the drawing. The applicant took i

75 percent credit for the B-10 in the poison.

Moderation inside and between the packages was varied from dry to full density water to determine optimum moderation. The applicant also assumed that the structuralintegrity of the fuel assemblies does not remain intact and the water gaps increase and are evenly spaced around the fuel plates. The maximum k-eff occurred for full density water. The maximum k-eff was 0.67345, including bias and uncertainty. Sir..e the maximum k-eff was well below 0.95 and well below the maximum k-eff for the U 0, fuel assembly (0.9078), the applicant did not perform 3

a fuel plate number sensitivity analysis.

The NRC staff performed confirmatory criticality analysis for the new fuel design. Staff used the KENO Va. computer code in the SCALE system and the 238 group ENDF/B-V cross-section set.

The staff performed optimum internal and intetspersed moderation calculations for an infinite array of packages under hypothetical accident conditions.

The staff also assumed that the fuel assemblies did not retain their structural integrity and was allowed to expand to the maximum space allowed by the basket. The outer plates in the staff's models touch the edges of the basket so that the water gaps between the fuel plates are at a maximum. The staff performed criticality calculations to determine the optimum number of fuel f

plates. The number of fuel plates was reduced from the maximum number (10) and the total fuel mass was held constant by varying the amount of fuelin each plate. The staff found that the package was most reactive when the assembly consisted of the maximum number of fuel plates. The staff's maximum k-eff was 0.7751, including uncertainty.

The results of the applicant's analysis and the staff's confirmatory analysis provide reasonable assurance that the new MTR fuel content does not affect the ability of the package to meet the criticality safety requirements of 10 CFR Part 71.

CONCLUSION The Certificate of Compliance is revised to include as an authorized content the MTR-type fuel with the parameters specified above. This change does not affect the ability of the package to meet the requirements of 10 CFR Part 71.

E. William Brach, Director Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Date:

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- USERS OF CERTIFICATE OF COMPLIANCE NO. 9228 MR. CHARLIE C. BOYD, JR.

BABCOCK & WILCOX NUCLEAR ENVIRONMENTAL SERVICES PO BOX 11165 LYNCHBURG,VA 24506 MS. SUSAN KINTNER CHEM-NUCLEAR SYSTEMS, INC.

140 STONERIDGE DRIVE COLUMBIA, SC 29210

- MR. MICHAEL E. WANGLER US DEPARTMENT OF ENERGY i

19901 GERMANTOWN ROAD GERMANTOWN, MD 20874 MR. CHUCK W. BASSETT GENERAL ELECTRIC COMPANY 6705 VALLECITOS ROAD '

SUNOL, CA 94586 -

MS. DEB SCHEBLER 10WA ELECTRIC LIGHT AND POWER COMPANY 3277 DAEC ROAD PALO,IA 52324 MR. THOMAS H. NEWTON, JR.

MASSACHUSETTS INSTITUTE OF TECHNOLOGY 138 ALBANY STREE f l

CAMBRIDGE, MA 02139 MR. ROBERT L. BEILKE NcBRASKA PUBLIC POWER DISTRICT

)

PO BOX 98 l

BROWNVILLE, NE 68321 1

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' MR: STEPHEN SCACE NORTHEAST NUCLEAR ENERGY COMPANY PO BOX 128 i WATERFORD, CT 06385 MR. WALT A. MEYER, JR. _

UNIVERSITY OF MISSOURI COLUMBIA

- ROOM 402 RESEARCH REACTOR FACluTY-I COLUMBIA, MO 65211 i

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