ML19338D800
| ML19338D800 | |
| Person / Time | |
|---|---|
| Site: | 07109001 |
| Issue date: | 08/27/1980 |
| From: | Macdonald C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML19338D796 | List: |
| References | |
| NUDOCS 8009230761 | |
| Download: ML19338D800 (6) | |
Text
- - _ - _
Form NRC.618 -
U.s. NUCLEAR.F;EGULA. TORY CoMMisSloN CERTIFICATE OF COMPLIANCE
-10 c R 71 For Radioactive Materials Packsges e
1.la) Certificate Number 1.(b) Revision No.
1.(c) Package Identification No.
1.!c) Pages No. 1.(e) Total No. Pages
-9001 10 IKA/4001/R( V 1
7
- 2. PREAMBLE 2.(a)
This certificate is issued to satisfy sections 173.393a.173.494.173.39s. and 173.396 of the Department of Transportatior Mazardous Materials Regulations (49 CFR 170-189 and 14 CFR 103) and sections 146-19-10s and 146-19-100 of the Department of Transportation oangerous Cargoes Regulations (46 CFR 146-149), as amended.
2.lb)
The packaging and contents described in item 5 below, meets the safety standards set forth in Subpart C of Title 10. Code of
, Federal Regulations. Part 71. " Packaging of Radioactive Meterials for Transport and Transportation of Radioactive Material Under Certain Conditions-
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This certificate does not relieve the consigno-from compliance with any requirement of the regulations of the U.s. Department of Teansportation or other appilcable regulatory agencies, including the government of any country through or into which the package
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will be transported.
- 3. 7his cartificate is issued on the basis of a safety analysis report of the package design or application-3.(a)
Prepared by (Name and address):
3.(b) ' Title and identification of report or' application:
. General Electric Company General Electric Company application dated 175 Curtner Avenue October 8, 1979, as supplemented..
San Jose, CA 95125 3.le cocket No. 71-9001
' 4 CoNDITtoNs This cartificate is conditional upon the fulfilling of the requirements of subpart D of 10 CFR 71, as applicable, and the conditions specified in item s below.
- s. Description of Packaging and Authorized Contents. Model Number, Fissile Class, other Conditions, and References-(a), Packaging s
(1) M3 del No.:
IF-300 (2) Description A stainless steel encased,'depl.eted uranium shielded cask.
The cask is cylindrical in shape, 64 inches in diameter and a maximum of 210 inches long with maximum cavity dimensions of 37-1/2 inches in diameter by 180-1/4 inches long.
Shielding is provided by 4 inches-of depleted uranium, 2-1/8 inches of stainless steel and a minimum of 4-1/2 inches of water.
-Two closure heads are provided for the shipment of BWR and PWR fuel assemblies.<
The~ heads are 304 stainless steel forgings and end plates which encase the 3-inch thick depleted uranium shielding.
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The closure heads are secured to the cask body by means of 32, 1-3/4 inch studs and nuts.
The cask is sealed with a metallic ring gasket.
Thecavityispenetratedbyaventlineitthetokandadrainline'atthe bottom.
These. lines.are sealed by bellows seal stainless steel globe valves and valved quick-disconnect couplings.
The vent.line is also equipped with a 375'psig relief valve.
All valves are housed in protected boxes on the cask exterior.
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'Page~2.
Certificate No. 9001 --Revision.No. 10 - Docket No. 71-9001 5.
-(a) Packaging (continued)
(2) Description.(continued)
Neutron shielding is provided by a liquid-filled, thin-walled, corrugated containment on the cask exterior..This cylindrical structure is-separated into two longitudinal compartments, each equipped with two expansion tanks, fill and relief valves.
The fill line from each compartment is terminated by a stainless-steel globe valve in a protected box (separate from cavity boxes) on the cask exterior.
The vent line from each compartment goes to an expansion tank which is provided with a pressure relief valve set at 200 psig.
The cask has two types of fuel baskets which can be interchanged to accommodat@
various fuels.
The PWR basket holds 7 assemblies, the BWR basket hold 18 assemblies.
The BWR fuel basket may be provided with supplementary shielding (depleted' uranium) near the cask closure.
The cask is shipped horizontally Qith -the bottom supported in a tipping cradle between two~ pedestals and the upper end resting in a' semi-circular saddle; the upper end is pinned to the saddle.
The cask supports are welded to the framing of a~37-1/2-foot long by 8-foot wide structural steel' skid. The ' skid also holds the cask cooling system which consists of two diesel engines driving two blowers which discharge into common ducting.
Four ducts run the length of the cask and direct cooling air to the corrugated 1
surface.
Operation of the auxiliary cooling system is not a requirement of this package approval.
The entire cask and cooling system is covered by a retractable-aluminum enclosure.
Access to'the enclosure is via locked panels in the side and a locked door in one end. Although the Model No. IF-300 cask can be transported-for short distances on the highway, its principal mode of transportation is by railroad.
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The gross weight of the cask is approximately 140,000 pounds.
The skid and other external components weigh approximately 35,000 pounds.
(3) Drawing The Model No. IF-300 shipping cask is described by the following General Electric Company Drawing No.:
159C5238-Sheets 1 thru 2, Rev. 2; Sheet 3, Rev. 3; Sheets 4 thru 5, Rev. 4; Sheet 6, Rev. 5; Sheet 7, Rev. 3; Sheet 8, Rev. 3 or 4; Sheet 9, Rev. 3; Sheet 10, Rev. 5; cnd Sheet 11, Rev. 2.
(b) Contents - water as primary coolant (1) Type and form of material Irradiated PWR or BWR uranium-oxide fuel assemblies.
The specific power of each fuel assembly shall not exceed 40 kw/kgU and the burnup of each fuel assembly shall not exceed 35,000 MWD /MTU.
The minimum cooling time of each assembly shall-be no.less than 120 days.
Prior to irradiation, the PWR and BWR fuel assemblies have the following dimensions and specifications:
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.. 10l D:cket No.. 71-9001
- Page4 3;-1Certificats Ns. 9001 l-Revision N2.
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!(b)L Contents-. water as' primary coolant ontinued)
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.PWR-BWR Fuel form
. Clad UO ' pellets Clad UO2 pellets.
2
. Cladding material
-Zr or SS-Zr orISS.
Maximum initial:U~
- content / assembly,.kg
'46; 198~
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' Maximum.initialU-235-enrichment, w/o.'
- 4. 0 3.5 2'
' Maximum bundle cross section,jinches-8.75-5.75 Fuel pin array 14x14/15x15 7x7 Fue11 diameter, inch 0.380-0.460 0.500-0.600 J
- Fuel pin. pitch-range,
- inch.
0.502-0.582-0.647-0.809 e
Maximum active ' fuel length,-inches.
145 146 The assemblies may be shipped with or without burnable poison: rods or.
j; control-rods.
(2) Maximum quantity of material per package Maximum decay heat per package not to exceed 210,000 Btu /hr.
Maximum-37,500/ Btu /hr/PWR assembly. Maximum 14,600/ Btu /hr/BWR assembly.
Seven (7) PWR fuel assemblies, or' eighteen (18) BWR fuel assemblies.
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-Above assemblies:to be contained in their respective fuel baskets are shown
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in GE Drawing No. 159C5238-Sheet'6, Rev. 5.
p (c) J Contents'- air;as primary coolant f(1)f Type and ' form of materia 1
' Irradiated.PWR and:BWR uranium' oxide fuel assemb1'ies.. The specific power
-ofieach fuel ^ assembly shall-'not exceed-40 Kw/KgU and the burnup of each
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fuel assembly ~shallinot exceed 35;000: MWD /MTU. The minimum cooling time of
!each assembly.shal_1-be'no less than^120 days..' Prior-to irradiation, the
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'BWR and PWRtfuelL. assemblies shall have'the following' dimensions'and
? specification's:
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3 A..
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- e Page 4 - Certificate No. 9001 - Revision No.10 - Docket No. 9001 5.
(c) _ Contents air as primary coolant (continued)
PWR BWR Fuel form Clad 002 pellets Clad UO2 pellets Cladding material Zr or SS Zr or SS Maximum initial U content / assembly, kg 465 198 Maximum initial U-235 enrichment, w/o 4.0
- 3. 5
' Maximum bundle cross
.section, inches-8.75 5.75 Fuel pin array.
14x14/15x15
~7x7 Fuel diameter, inch 0.380-0.460 0.500-0.600 Fuel pin pitch range, inch 0.502-0.582 0.647-0.809 Maximum active fuel length, inches 145 146 The assemblies may be shipped with or without burnable poison rods or control rods.
(2) Maximum quantity of material 'per package Itaximum decay heat per package not to exceed 40,000 Btu /hr.
Maximum 5,725 Btu /hr/PWR assembly.
Maximum ~2,225 Btu /hr/BWR assembly.
Seven (7) PWR fuel assemblies, or eighteen (18) BWR fuel assemblies.
Above assemblies to be contained in their respective fuel baskets as shown in GE Drawing No. 159C5238-Sheet 6, Rev. 5.
(d) Unloaded' package '- contents and maximum quantity of material Greater than a Type A quantity of residual radioactive material consisting of mixed-fission and activation products adhering to interior cavity and fuel basket sut.ces.
(e) Fissile Class I
6.
(a) The end of life (after irradiation) calculated fuel pin pressure shall not exceed 1,800' psia, at 900*F for contents 5(b).
(b)::The' end of life total calculated residual gas that could become available from the' fuel pins shall not exceed'0.23 lb moles for content 5(c) and-individual
-calculated fuel pin pressure shall not; exceed 2,500 psia, at 900'F.
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Page.5 - Certificate No. 9001'- Revision No. 10 - Docket No. 71-9001 a
7.
The maximum gross weight of the cavity contents shall not exceed 21,000 pounds.
8.
(a) For the contents described in 5(b) (water coolant) the cavity fill specifications shall include the following:
A 21.0 +'1.0 - 0.5 cu ft cavity air void shall be established for PWR and BWR loadings.
These air voids are established wi.m the bulk water temperature-is at 100 F for both the PWR and BWR' loadings.
If less than the maximum number of fuel ~ assemblies is loaded into the basket, a void
' displacement equivalent to the missing fuel assemblies shall be inserted into the basket.
In addition, the licensee shall take sufficient-time-temperature-pressure data to show that the cavity pressure will not exceed 346 psig during a 130*F day with no auxiliary cooling.
Under freezing conditions, the minimum heat load shall be 36,400 Btu /hr when water is used as the primary cavity coolant.
(b) For the contents described in 5(c) (air coolant) the cavity fill specifications shall include' the following: An air void shall be established such. that not more than 1.0 cu ft of water (corresponding to a bulk water temperature of 70 F) remains in the cavity.
Tha licensee shall take sufficient time-temperature-pressure data to ensure that the cavity pressure will not exceed 45 psig, and that the average cavity wall temperature will not exceed 210*F during the 130 F day with no auxilary cooling.
9.
A determination shall be made for each water coolant shipment (for contents described in 5(b) that the total radioactivity of the primary coolant will not exceed, during the anticipated period of transport, the limits specified in 10 CFR 671.36(a)(2).
This determination shall include monitoring of the coolant and verification of the coolant activity upon arrival of-the package at its destination.
Records of such determinations shall be maintained for a period of-two years after its generation.
- 10..For-the contents described in 5(c) and 5(d), the air coolant is considered part of the package contents.
The radioactivity limits specified in 10 CFR 971.35(a)(4) do not apply.
11.
Prior to each shipment, the licensee shall confirm that the cask is properly sealed by testing as Subsection 11.3.3.1.
12.
The cark contents shail be so limited that under normal conditions of transport, 111 times the neutron dose rate plus 11.3 times the gamma dose rate will not exceed 1,000 mrem /hr at three (3) feet from the external surface of the cask.
' 13.
The neutron shielding tanks shall be filled with water during the months of May through October and approximately a 50/50 volume percent mixture of ethylene glycol and water during the months of October through May if.the total' package decay heat is greater than 183,400 BTU /hr (70% of design basis).
If the total package decay heat
'is:less than 183,400 BTU /hr the ethylene glycol and water mixture may remain in place all year.
l-14.~
In addition the-requirements of Subpar _t D of 10 CFR Part 71, each package prior to first use shall meet all of'the acceptance' tests and criteria specified in Subsections i
6.7.6.2,111.3.1.1, and 11.3.1.7.
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Page 6 - Certificate No.-9001 - Revision No. 10 - Docket No. 71-9001
- 15. The maximum allowable heat load shall be documented for each cask and conspicuously and durably marked.on.the cask.
16.
Each cavity relief valve, typical glove valves, and typical shielding tank (barrel expansion tank) relief valves shall be tested as' stated in Subsections 6.5.3.3, 6.6.1.1, and 6.6.1.2.
In lieu of the requirements of 10 CFR $71.54(h), valve testing and maintenance frequency shall be as stated in Subsections 6.5.3.4, 6.6.2.1, and 6.6.2.2 except during periods of cask inactivity.
During. inactive periods the maintenance and testing frequency may be disregarded provided that the package is brought into full compliance with these requirements prior to the next use of the package.
17.
The cask cavity shall be equipped with a Target Rock 73J pressure relief valve set at a pressure of 375 psig'(450 F).
The valve is shown in Target Rock Corporation Drawing No. 73J-001, Rev. H or J.
.18.
The uranium shielding material shall be separated from all steel surfaces with a mini-mum copper thickness of 4-mils, except that the stud bolts attaching the shield assem-blies to top of the BWR basket shall be coated with a minimum of 1/2-mil of copper.
19.
For casks using air as the primary coolant, the cavity pressure relief valve specified in Item 16 shall be installed and operati,ng during the cooldown prior to unloading.
- 20. No shutoff valve shall be installed between each neutron shield tank and its respective tnermal expansion tank.
- 21. The package authorized by the certificate is hereby approved for use under the general licBnse provisions of 10 CFR 971.12(b).
1 22.
Expiration date:
October 31, 1984.
j REFERENCES General Electric Company consolidated application dated October 8, 1979.
Supplements dated: May 12 and J,uly 21, 1980, Documentation of maximum package heat load as determined by Item 15 above.
Section XI, Quality Assurance and Testing, is deleted from the application.
l FOR THE U.S. NUCLEAR REG'JLATORY COMMISSION i
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[
Charles E..M Donal, Chief Transportation Certification branch Division of Fuel Cycle and Material Safety ~
l Date:
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