ML20216H529

Certificate of Compliance 6003,rev 19,for M-130.W/approval Record

ML20216H529
Person / Time
Site:	07106003
Issue date:	03/13/1998
From:	Haughney C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20216H515	List: ML20216H509 ML20216H529
References
NUDOCS 9803230064
Download: ML20216H529 (11)

Text

l swwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwww p NRCFORMSit nw CERTIFICATE OF COMPLIANCE u.s. NUCLEAR RE2uLAToRY CoMM2 SON l l w cm n FOR RADIOACTIVE MATERIALS PACKAGES I I l 1.a CERT!!1CATE NUMBER b. REVISION NUMBER c. PACKAGE IDENTIFICATION NUMBER 9 d. PAGE NUMBER e. TOTAL NUMBER PAGES 6003 19 USA /6003/B()F 1 5 I 1(

I I .

2. PREAM3LE I I

a. This certificate is issued to certify that the packaging and contents describei in Item $ below. meets the applicable safety standards set forth in Title 10. I 1 Code of Federal Regulations. Part 71, " Packaging and Transportation of Radioactive Material? E I

b. This certificate does not relieve the consignor from compliance with any requirement of the regulations of the U.S. Department of Transportation or other I

applicable regulatory agencies. tacluding the government of any country through or into which the pacLage will be transponed. I l-I 3. THIS CERTIF1CATE 15 ISSUED ON THE BASIS OF A SAFETY ANALYS15 REPORT OF THE PACKAGE DESIGN OR APPLICATION

a. ISSUED TO(Name and Address) b. TTTLE AND IDENTIFICATION OF REPORT OR APPLICATION: b b

l-

~

3 U.S. Department of Energy Safety Analysis Report for M-130 shipping I

3 Division of Naval Reactors container dated December 30,1968, as Washington, DC 20585 i j supplemented.  ;

1 L

3 c. DOCKET NUMBER 71-6003 g 3 ( CONDITIONS L

3 This certificate is conditional upon fulfilling the requirements of 10 CFR Part 71, as applicable, and the conditions specified below.

L 1 5.

I 1

• L 1 (a) aging i 1

l-1 Model No.: M-130 L I 1 '

1 1 (2) Description I 1

1 3

The Model No. M-130 shipping container is an upright cylinder 84 inches in diameter by 1 1

158 inches overall height. The container walls consist of a finned 1-inch thick outer shell I 3

fabricated from either carbon steel, carbon steel with stainless steel clad, or solid stainless l l 1

steel,10 inches of lead shielding, and a 1-inch thick inner pressure vessel fabricated from I i l

carbon steel clad with stainless steel. The top of the container is covered with a shielded I 3

closure head which is boited to the container and seals the pressure vessel. An access I 3

opening with a bolted shield plug is provided in the closure head for loading and unloading I 3

spent fuel. I 1

1 3

The pressure vessel has an inside diameter of 55 inches l The central region contains a L I

l secondary heat exchanger (not used during shipment) surrounded by 1/2-inch thick carbon 3

steel backup cylinder 29 inches in diameter. The annulus which remains between the I j backup cylinder and the pressure vessel provides a spaca 13-inches wide and 130-inches 3 high for spent fuel. The spent fuel is contained in the annulus by module holders designed j for the particular core to be shipped.

3 The container has extemal penetrations to the pressure vessel for steam and water relief l 3 ,

3 lines and a fill and drain line (which are capped during shipment) and a pressure sensing i

3 line which remains open to a pressure gage during shipment. The container also has g 3 penetrations which do not open to the pressure vessel for secondary heat exchanger lines  ;

i (which are capped during shipment) and a temperature sensing line. i I L 1 The container is supported on its transport vehicle by an "A" frame structure. Gross weight i i of the loaded container without its support structure is approximately 228,000 pounds. L 3 L 2 L 3 L L

9803230064 980313 i PDR ADOCK 07106003 r C PDR g mammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmE

gwwwwwwwwwwwwwwwwwwwwwwww.wwwwwwwwwwwwwwwwwwwww p NR FORM 618A coNomoNs(c,mnand; u.s. NUCLEAR REtuLAToRY CoMMISsK)N D E Page 2 - Certificate No. 6003 - Revision No.19 - Docket No. 71-6003

[

5.(a) Continued

[

h (3) Drawings

[

b L p The packaging is constructed in accordance with General Electric Drawing Nos. 247E209, g g Sheet 1 Rev. R; Sheet 2, Rev. K; Sheet 3, Rev. T; Sheet 4, Rev. U; Sheet 5 of 5, Rev. F, g p and 247E228, Rey, F. g R I p (b) Contents I B L p (1) Type and form of material l B L p frradiated fuel assemblies, activated corrosion products and structural parts containing up i p to 40 gallons of residual contaminated water. The fuel assemblies and structural parts are r b of the following types: r B L b (i) Deleted. L N L Dj (ii) Deleted. I Dj i Dj (iii) Deleted. t D r D (iv) D1G fuel modules of core types 1 or 2. F B L N (v) D1G removable fuel assemblies of core types 1 or 2. I b I  !

N I (vi) Deleted.

D L ,

I I (vii) Deleted.

b r I I (viii) S3G-3/3A fuel module with or without control rods. The core age must be at least 4000 logging-corrected full-power hours.

[

(ix) Deleted. [

-(x) S3G-3/3A irradiated thermocouples and thermocouple cases. g D L p (xi) S8G full size fuel cell with or without control rod. g D L p (xii) S8G partial size fuel cell with or without control rod, t D L p (xiii) Deleted. r b 0 p (xiv) Deleted. L

,l D (xv) D2W fuel cells with control rods. L D L D- L (xvi) NR-1 fuel modules with or without control rods.

b L E E (xvii) Deleted.

D L E I (xviii) . A1W-3 recoverable irradiated fuel modules. Fuel modules that use control rods shall have control rods inserted.

[

L ............................................a

7--wwwwwwwwwwwwwwwwwwwwwwww.ww.wwwww.ww www.w.m g NRC FORM 618A CONornONS (amruu,,4; u.s. NUCLEAR RE!uLAToRY Commission N Page 3 - Certificate No. 6003 - Revision No.19 - Docket No. 714003 I N .

I W 5.(b)(2) Maximum quantity of material per package. I D

I b Deleted.

(i) I b

t I

(ii) Deleted. I B

I b

(iii) 6 fuel assemblies as described in 5(b)(1)(iv) and 4 fuel assemblies as described in I 5(b)(1)(v). I (iv) Deleted. I I

l (v) 10 fuel assemblies as described in 5(b)(1)(viii). ,

p (vi) 9 fuel assemblies as described in 5(b)(1)(viii).

~ ~

(vii) 9 fuel assemblies as d' escribed in 5(b)(1)(viii) and 1 structure as described in f p 5(b)(1)(x). g B

L p

(viii) 4 fuel cells as described in 5(b)(1)(xi) or 2 fuel cells as described in 5(b)(1)(xi) and 2 g p fuel cells as described in 5(b)(1)(:di). t I

p (ix) Deleted. g b

I p (x) Deleted.

Bl' I I i b (xi) 4 fuel cells as described in 5(b)(1)(xv) plus 2 corner fuel cells or 1 RFA fuel cell. r B

L D (xii) 4 fuel modules as described in 5(b)(1)(xvi). E B

j L <

B (xiii) Deleted. I D L D (xiv) For contents described in 5(b)(1)(xviii),6 fuel modules or 8 fuel modules, as I b described in supplement dated March 30,1992. I i

> r <

I Shipments shall be further limited by thermal requirements as follows: '

(3) L b L I

(i) Shipment of contents specified in 5(b)(1)(iv) and 5(b)(1)(v) and limited in 5(b)(2)(iii) L I

shall be made no earlier than 75 days after shutdown and shall have a decay heat I N

load not to exceed 33,500 Btu /hr per shipment. I D r (ii) Deleted. I (iii) Shipment of contents specified in 5(b)(1)(viii), and 5(b)(1)(x) and limited in p 5(b)(2)(v), 5(b)(2)(vi), and 5(b)(2)(vii) shall be made at a time after shutdown, as [

g determined from Bettis Atomic Power Laboratory report WAPD-OP(PP)S-4401 g y dated June 29,1979, and shall have a decay heat load not to exceed 28,620 Btu /hr g g for the shipboard core and 30,000 Btu /hr for the prototype core. g 5'

p t

I (iv) Deleted. t D L

> L D E D E D E

1. -ammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmm g

p' NRCFORM 618A CONDITIONS fromwd) U.s. NUCLEAR REZULAToRY CoMMassioN g b r Page 4 - Certificate No. 6003 - Revision No.19 - Docket No. 71-6003

[

I 5.(b)(3) Continued (v) Shipment of contents specified in 5(b)(1)(xi) or 5(b)(1)(xii), as limited by 5(b)(2)(vii),

g l g shall have a fully loaded container heat load not to exceed 15,400 Btu /hr per g g shipment. 4 g

D L p (vi) Deleted. g b r l

p (vii) Deleted. t B L l l p (viii) Shipment of contents specified in 5(b)(1)(xv) and limited in 5(b)(2)(xi) shall have a I

' l p heat load not to exceed 19,100 Btu /hr and shall be made no earlier than 420 days r i p after shutdown. I '

i b -

l p (ix) Shipment of contents specified in 5(b)(1)(xvi) and limited in 5(b)(2)(xii) shall have a

, L heat load not to exceed 6,000 Btu /hr and shall be made no earlier than 50 days D l-B after shutdown. I b r B (x) Deleted. L l

D l l B (xi) Shipment of contents specified in 5(b)(1)(xviii) and limited in 5(b)(2)(xiv) shall have L B a heat load not to exceed 43,800 BTU /hr and shall be made no earlier than 400 1 5- days or 175 days for A1W-3E and A1W-3J fuel, after shutdown. L l B r l 5 (c) Transport Index for Criticality Control r B L 1 5 Minimum transport index to be shown on L f

U label for nuclear criticality control: L Except for the contents described in 5(b)(1)(iv) (Core 2), 100 I l 5(b)(1)(v) (Core 2) and 5(b)(1)(viii) and limited in 5(b)(2)(iii) and 5(b)(2)(v) g

]

I l For the contents described in 5(b)(1)(viii) and limited in 5(b)(2)(v) 25 g ,

I For the contents described in '5(b)(1)(iv) (Core 2) and 0

h p 5(b)(1)(v) (Core 2) and limited in 5(b)(2)(iii) g

1. L p 6. Deleted. g b r p 7. For shipments involving the contents specified in 5(b)(1)(viii) or 5(b)(1)(x), the thermocouples and g b thermocouple cases if included or the vacant module holder shall be located in the mid-position of r p either cage and module holder assembly. L b L B 8. Shipments shall be made in the dry condition, except for residual water as limited in 5(b)(1). L p;. E Dl 9. Container number three (M-130-3) has been modified by adding two 4-inch thick by 8-inch wide L W steel pistes welded between fins 25 and 50 and between fins 110 and 135 at approximately 14.75 E D inches from the bottom of the container. The cooling fins in this localized area are removed to L permit attachment of the plate directly to the outer shell of the container. L D} L D{

p E

o 8.....m m===ma==mm==== . mamam ====mm=======mm2

jruwwwwwwwwwwwwwwwwwwwwwwwwwwwwwwww&manwwwwwwwww '

I gc, FORM Si4A CONDmONS (ami,a; u.s. NUCLEAR MEfiuLAToRY ColnessON r i 1 Page 5 - Certificate No. 6003 - Revision No.19 - Docket No. 71-6003 l l l 10. ' Container number four (M-130-4) has been modified by adding a 2-inch thick by 4-inch wide steel

, plate welded between fins 32 and 49 at approximately 18.4 inches from the bottom of the l

, container. The cooling fins in this localized area are removed to permit attachment of the plate g j directly to the outer shell of the container. ,

f 11. Containers M-130-3, M-130-4, M-1304, and M-130-7 may be used for the contents specified in f

, 5(b)(1)(viii) and 5(b)(1)(x) only. Containers M-130-10 and M-130-15 may be used for the contents g i specified in 5(b)(1)(viii),5(b)(1)(x), and 5(b)(1)(xviii) only. ,

I I i 12. Container M-130-11 may be used for the contents specified in 5(b)(1)(xvi) only. ,

3 1 3 13. Deleted. L 3 l 1 14. Expiration date: September 30,2002. .

L 4

3 4 r

i REFERENCES I 3 L i Safety analysis report for M-130 shipping container, MAO-E8-703 dated December 30,1968. I 1 L 1 Supplements Naval Reactors (NR) letters A#2256 dated February 24, and G#1931 dated March 3,1969; I 3 General Electric Company (GE) letter ONP-74520-526 dated April 3,1972; NR letter G#3207 dated I 1 April 27,1972; GE letter ONP-74520-528 dated April 28,1972; NR letter G#3250 dated June 6,1972; GE I 3 letters ONP-74570435 dated October 25, ONP-74570-654 dated December 4, and ONP-14570466 f 3

dated December 12,1972; ONP-74570482 dated January 12, ONP-74570-698 dated January 31, f 3

ONP-74570-687 dated February 6, ONP-7439045 dated March 26, and DLGN45570454 dated f 3 L September 24,1973; and DLGN45570-901 dated January 10,1974; NR letter G#4061 dated 3

January 29,1974; GE letters DLGN-85570-924 dated February 15, DLGN-85570 923 deM March 6, and I 3

DLGN-85570 969 dated May 24,1974; NR letter G#4991 dated November 25,1975; GE letters 1 3

ONP-74340-JTT-73 dated December 17,1975; CGN-85570-1145 dated September 9, CGN-85570-1146 I I I dated September 10, and CGN-85570-1148 dated September 14,1976; Bettis Atomic Power Laboratory 3 I letters WAPD-R(K)-1378 dated August 30,1976, and WAPD-OP(PP)S 4401 dated June 29,1979; NR l letters G#6197 dated July 13,1979, G#7022 dated July 14, WAPD-LP-(CES)SE-170 dated July 1981;

, and WAPD-LD-(CES)SE-181 dated September 1981; WAPD-LP(CES)SE-96 dated February 1982' l G#7136 dated March 17,1982; Gk7160 dated May 18,1982; G#7582 dated September 7,1983; r

, G#C87-5692 dated September 2, and G#C87-5689 dated September 23,1987; G#C87-8008 dated ,

3 January 19, G#C88-5931 dated May 12, and G#C88-5961 dated July 25,1988; G#C89-2825 dated g i 3

March 29, and G#C89-2863 dated August 11,1989; G#C92-03392 dated March 30, and G#92-03729 ,

, dated October 20,1992; G#C93-10935 dated October 8,1993; G#96-03344 dated March 6, and g 3 G#96-03610 dated December 9,1996; G#97-03543 dated July 10, and G#C97-03685 dated g :

December 19,1997.

3 r !

3 L 3 FOR THE U.S. NUCLEAR REGULATORY COMMISSION t

'/

3 3;

3 Charles J. Haug y, Acting D' Spent Fuel Project Office or{' t F

Office of Nuclear Material Safety L 3 and Safeguards 0 3 L 3 Date: March /f ,1998 0 C

1' 3 C L.............................................a

y  %&

. p UNITED STATES g

n &

i NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2000641001 y*****j APPROVAL RECORD Model No. M-130 Certificate of Compliance No. 6003 Revision No.19 By application dated December 19,1997, the U.S. Department of Energy, Division of Naval Reactors (NR), submitted a revised criticality analysis to support certrfication of shipments of D1G-2 fuel in the M-130 shipping container with a Transport index (TI) of zero. As stipulated in 10 CFR Part 71, this entails demonstrating subcriticality for infinite arrays of packages under normal and hypothetical-accident conditions of transport.

Cnbcality Evaluation The staff has completed its review of the revised criticality analysis. The staffs review consisted primarily of an assessment of the applicant's described analysis methodology and did not include confirmatory computations.

Table 1 details the staffs assessment of selected aspects of the applicant's analysis methodology. The overall methodology was found to be complete and adequately conservative, both in its use of pessimistic or bounding assumptions within explicit computational models and in its allowances for uncertainties, potential calculational biases, and various effects not explicitlyincluded in the models.

Using this conservative methodology, the analysis showed k-effective to be less than 0.95 for infinite arrays of normal packages and accident-damaged packages. Therefore, the staff concurs with the applicant's conclusion that the M-130 container loaded with D1G-2 spent fuel l meets the criticality safety requirements of 10 CFR Part 71 for shi;)ments with a Tl of zero. I l

Table 1. Assessment of the Navy M-130 D1G-2 Criticality Analysis Methodology Analysis Methods and Feature Approximations Staff Comments Primary RACER code: Quasi- RACER's geometry modeling capabilities and computational continuous-energy Monte physical rigor are very close to state-of-the-art.

tool Carlo method with exact in presentations pursuant to RAls on other 3D geometry modeling Naval packages, NR has described to the staff capabilities. Uses how the application of this code to M-130 and hyperfine cross section other containers, with various loadings similar to data derived from D1G-2, has been validated against a large set ENDF/B and special NR of experimental critical benchmarks based on evaluations. fresh Naval fuel, including some with neutron spectral characteristics and material arrangements roughly similar to those in moderated transport packages.

! 1

Table 1. Assessment of the Navy M-130 D1G-2 Criticality Analysis Methodology (cont.)

Analysis Methods and Feature Approximations Staff Comments Contents The model assumes a full Actual loadings will always include D1G-2 fuel modeled loading wherein all full- types that are less reactive than the one l size positions in the modeled. This is because the D1G-2 core has container are loaded with fewer fuel modules of the modeled most-the single most-reactive reactive type than there are full-size loading D1G-2 fuel type. positions in the M-130 container.

Fuel model: Detailed, fully explicit Minor approximations on small details of the fuel l geometry model with nominal geometry are stated to have negligible reactivity dimensions, effect. Dimensional tolerances are treated via material specifications (next item) and a positive delta k-effective allowance (final items in table).

Fuel model: Fresh composition. The potentially positive reactivity effects from materials Effects of selected fuel bumup are accounted for es an additional l manufacturing tolerances delta k-effective allowance as described later in

are bounded by this table. Likewise, a further at;owance is simultaneous density subsequently provided for material tolerance l adjustments that effects not modeled explicitly.

maximize contents of U-l 235 and water and

! rninimize poison content.

Control rod Fully explicit model with Effects of dimensional tolerances are included model: nominal dimensions. in the allowances described later in this table.

geometry Axial shifting relative to Moderate axial shifting was assumed for

, fuelis assumed greater bounding normal conditions and greater shifting than predicted maximum. for bounding accident drop tests.

l l Control rod Nominal fresh Responses to RAls on other Naval packages l model: composition. indicate that 1) rod self-shielding effects make l materials calculated results weakly sensitive to uncer:minties or small changes in rod composition, and 2) effects of control rod depletion are negligible.

Fuel Full-size and smaller fuel in actuality, the fueled regions of the full-size placementin modules are modeled as modules are placed with a substantial axial package, being axially aligned in offset relative to those of the smaller modules.

axial their most reactive positions relative to each other.

I L

2 1

Table 1. Assessment of the Navy M-130 D1G-2 Criticality Analysis Methodology (cont.)

m Analysis Methods and Feature Approximations Staff Comments Fuel Fuel modules are Effects of dimensional tolerances in the module placement in assumed centered in their holder assembly were determined to be package, most reactive onentation negligible and nominal starting dimensions were horizontal within the holder cavites used. Snifting within the holder cavities is not in the accident case, the addressed but is likely bounded for the limiting holder segments are accident case by the assumed shifting of the assumed to shiftinward holder segments.

greater than predicted by structural analysis, infinite array Cylindrical outer surface Curved mirror-reflective BC is nonphysical but model, lateral of containeris modeled probably conservative here. Rationale:

as a mirror-reflective Consider 2D array reactivity with increasing boundary condition (BC). numbers of planar mirror BCs (i.e.,2=line <

4= square < 6= hexagon (max hcp) <

infinity-cylinder). Note: Curved mirror boundaries can cause tracking anomalies and should be used with caution.

Infinite array Plane mirror-reflective Placement of the top BC is clearly conservative.

model, BCs top and bottom; top Consideration of offset stacking, generally not vertical BC is belowlocation of warranted in such cases, would likely be top surface of package, bounded by the conservative placerrent of the top BC.

Intemal Uniform flooding with full- A most-reactive uniform flooding level was moderation, density water up to top of assumed possible following worst-case damage normal active fuel. from the " normal" 1-foot drop.

case 1

3 l

4 Table 1. Assessment of the Navy M-130 D1G-2 Criticality Analysis Methodology (cont.)

Analysis Methods and Feature Approximations Staff Comments intamal Preferential flooding with The assumed preferential flooding configuration moderation, full-density water up to is highly conservative. There appears to be no accident case active fuel bottom in plausible mechanism for producing such a upside-down package. radial distribution of moderator. However, the The fuel modules are rapid climb in reactivity as surrounding water assumed flooded, while density approaches zero (SAR Fig.7) suggests everything surrounding that a selective lowering of intomal water

! the modules, including densities (e.g., near the module periphery) may l the gap between module further increase reactivity. Any such effects are l and holder-cavity wall, is likely bounded by other conservatisms in the assumed dry. The latter presently assumed configuration.

maximizes reactivity by Note: The SAR incorrectly refers to moderation enhancir.g the neutronic within the package as " interspersed" (see next

, coupling between item). Part 71 uses the term " interspersed" only l modules, in the context of arrays. Used as intended, it refers to moderation in the interstitial spaces l between arrayed packages.

Interspersed Not modeled. SAR states Use of an artificial curved-mirror boundary hydrogenous that mirror-reflective hugging the container surface effectively moderation, boundary on container's eliminates interstices and thus precludes the accident case outer surface is more modeling ofinterspersed moderation. For reactive. arrays of heavily shielded, moderated packages like M-130, it is true that interstitial moderation l

generally lowe s reactivity.

Close Not applicable for 3D The SAR incorrectly refers to "close reflection" l reflection infinite arrays. In discussing what should be called outside array " interspersed moderation," i.e., moderation in I the spaces between packages. "Close full reflection" in Part 71 refers to water just outside the confines of a finite or 2D infinite array.

(Note: Spent fuel casks are generally not among the fissile package types for which finite or 2D-infinite arrays with close reflection by water can be more reactive than a 3D infinite array.

Moreover, Part 71 does not currently provide for evaluating the latter.)

1 4  !

l l

f i

l l

l Table 1. Assessment of the Navy M-130 D1G-2 Criticality Analysis Methodology (cont.) l Analysis Methods and Feature Approximations Staff Comments Allowance for Assumed delta k-effective The SAR notes that bumup-reactivity effects

most-reactive allowance readily bounds are generally less positive in rodded modules, i

bumup conservative core-follow as in M-130, than in unrodded ones. Specific

computations of allowances for fission product decay and l unrodded reactivity operational variability, excluded from the core-l versus bumup. follow analysis, are included separately.

l Allowances Direct benchmark bias includes a conservative allowance for bias l for bias and and uncertainty, plus uncertainty associated with important materials uncertainty extra allowances for present in M-130 and not in the benchmarks.

derived from differences in material benchmarks and configuration between benchmarks and j

package.

l Allowances Conservatively based on These allowances for anticipated cross section l

for pending scoping calculations refinements (a.k.a. present cross section refinements using improved cross deficiencies) appear similar in function to the in RACER section data for important preceding allowances for M-130 materials not l cross section materials in M-130 represented in the benchmark cases. It is data container, noted that both sets of allowances serve to add conservatism and the magnituda of each seem to be conservatively estimated.

Allowance for Statistically combines Tolerance parameters are described in SAR unmodeled maximum effects from Table 8. Based on the nature of the fabrication tolerances in unmodeled tolerances on processes and how each associated tolerance fuel and various parameters. parameter is considered, the statistical control rods combination of effects is reasonable and conservative.

Allowance for Arbitrary value consistent Provides additional conservatism to account for i poter.tially with that used in previous the possible existence of a configuration that is more-reactive package analyses. more reactive than those identified and conditions modeled in the analysis.

5 1

i

CONCLUSION The scope of the stafs review, as detailed in Table 1 above, consisted of an assessment of selected aspects of the applicant's described analysis methodology and did not include i confirmatory computations. The overall methodology was found to be complete and adequately I conservative.

Based on the the stafs review of the revised criticality evaluation, there is reasonable assurance that the proposed use of the M-130 shipping container to transport D1G-2 spent fuel will not endanger public health and safety. The staff concludes that the changes requested by i the applicant will not affect the package's ability to meet the requirements of 10 CFR Part 71. ;

Therefore, in accordance with the application dated December 19,1997, and pursuant to 10 1 CFR Part 71, CoC 6003 for the M-130 package is revised to reflect an assigned criticality Transport Index of zero for core type 2 D1G-2 fuel.

1 All other conditions of COC 6003 shall remain the same. l

. This authoritation is effective upon issuance.

Principal Contributors: D. Carlson S. Whaley l

l FOR THE NUCLEAR REGULATORY COMMISSION Charles J. Haughney, Acting or l Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Date: U;)

6

)