ML19318C528
| ML19318C528 | |
| Person / Time | |
|---|---|
| Site: | 07105980 |
| Issue date: | 05/14/1980 |
| From: | Cunningham G GENERAL ELECTRIC CO. |
| To: | Rawl R TRANSPORTATION, DEPT. OF |
| Shared Package | |
| ML19318C529 | List: |
| References | |
| 16492, NUDOCS 8007010595 | |
| Download: ML19318C528 (58) | |
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ENGINEERING GENERAL ELECTRIC COMPANY, P.O. BOX 460, PLEAsANTON, CALIFORNIA 94566 DIVISION
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l May 14,1980 1
U. S. Department of Transportation Office of Hazardous Materials Operations Washington, D.C.
20590 Attention:
Mr. Richard R. Rawl
Reference:
(1)
USNRC Certificate of Compliance No. 5980 (2)
IAEA Lertificate of Competent Authority No. USA /5980/B( )F (3)
Consolidated Application for Renewal of Certificate No. 5980
Dear Mr. Rawl :
The General Electric Company, Vallecitos Nuclear Center (VNC), currently holds I AEA Certificate of Competent Authority No. USA /5980( )F (Ref. 2) for the G.E. Model 600 Shipping Container.
This Certificate is based on the NRC Certificate No. 5980 (Ref.1) and the IAEA 1967 Edition of "The Regula-tions for the Safe Transport of Radioactive Materials".
l As numerous countries have adapted the 1973 Edition of the IAEA regulations j
as their standard, VNC wishes to request that the D0T issue a Certificate of Competent Authority for the Model 600 under the provisions of the 1973 Edition of the IAEA regulations.
t In support of this application, VNC encloses Attachment A to this letter which makes a paragraph-by-paragraph assessment of the Model 600 in reference j
to the 1973 IAEA regulaticns.
Except when the IAEA 1973 requirements differ from those of the NRC, all of the information contained in Attachment A has been previously submitted to the NRC as the bases for issuing Certificate No. 5980.
For your convenience this application is enclosed as Attachment B j
to this letter.
Please note that for purposes of this application VNC proposes to limit the i
contents for foreign shipments to Special Form materials only. Such limita-l tion will provide the necessary assurance that no leakage would occur under
?
normal or accident conditions. All such Special Form materials would, of i
course, receive prior approval and certification'to the 1973 IAEA regulations l
by the D0T.
VNC also proposes to limit the. fissile content 'to 300 grams of Pu or 500 grams of U-235 or a prorated ratio of the two.
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GENER AL @ ELECTRIC f
Mr. Richard R. Rawl May 14,1980 i
The Model 600 has been historically shipped as a Fissile Class III because it has normally been transported as " sole use".
Because the 1973 I AEA regulation requires multilateral approval of such packages, VNC requests a Fissile Class II category.
The previously submitted criticality information t
supports the category shift as it demonstrates that 33 containers are " safe"
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in the worst possible case, f
In summary, VNC then requests a 1973 IAEA certification for the Model 600
! i Shipping Container based on the applications supporting NRC Certificate No.
5980 but limited to Special Form materials meeting the 1973 IAEA requirements l
and not exceeding the fissile limits listed above.
l Sincerely,
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Al!
i G. E. Cunningham Senior Licensing Engineer t
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o ATTACHMENT A G.E. MODEL 600 TRANSPORT CONTAINER 1973 IAEA REGULATORY ASSESSMENT MAY 1980 i
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G.E. MODEL 600 TRANSPORT CONTAINER 1973 IAEA REGULATORY ASSESSMENT The standards for safety for packaging radioactive materials for international transport are contained in the International Atomic Energy (IAEA) Safety Series No. 6, " Regulations for the Safe Transport of Radioactive Materials",
1973 Revised Edition. 40mpliance of the Model 600 package with the pertinent sections is discussed by paragraph number.
Section I - Definitions The Model 600 transport package is designed to fall under the category defined in paragraph 127 and 131 as a Type B (U) package.
Section II - Packaging and Package Design Requirements General Design Requirements 201 The package is equipped with adequate provision for fork type material handling equipment and adequate hardware to secure the package during transport.
202 Not applicable.
(Package weight is in excess of 50 kg) 203' Reference Paragraph 201.
204 The oackage is intended to be lifted either by fork type equipment or by the shackles and ears attached to the side of the jacket.
205 Two rectangular " eyes" for lifting the jacket off the base are provided adjacer*
.0 the tie-down ears.
During transoort, cover plates are bolted over these " eyes" to preclude their use for lifting or tie-down of the complete package. The cover plates are stamped with a warning "NOT FOR TIE-DOWN".
206 The outer surface of the package is designed to avoid the collection and retention of rainwater.
Section II - Packaging and Package Design Requirements (Continuad)
General Design Reouirements - cont'd.
207 The external surface of the package is smooth and painted for easy decontamination.
208 General Electric Standard Operating Procedures (S0P) preclude adding or deleting features of the package without a complete Engineering Analysis and review.
Additional Requirements for Strong Industrial Packages 209 Not applicable.
Additional Requirements for Type A Packages 210 All external dimensions exceed 4 inches (10cm).
211 The outside package is sealed with a wire and lead seal which prevents the protective jacket from being removed from the pallet-base assembly without breaking the seal or wire.
212 The cask is designed so that the external surface is free from protruding features except for the fixtures required for tie-down.
213 The structural properties of the materials used in the packaging components, i.e., steel jacket, steel shell, and lead metal, are suitable for service 0
0 between the temperature extremes of -40 F (-40 C) and 148 F (70 C).
The package is not highly stressed during normal transit and reduction of ductility at low temperatures is not expected to impair serviceability.
214 All weld: are made in accordance with General Electric standard welding oractices including dye penetnant examination and mass spectrometer leak detection tests. These standards are based on ASME and AWS code requirements.
215 Inspections of the Model 600 packages used for Type B and large quantity shipments since 1969 reveals no evidence of vibratory or acceleration-induced damage of significance to transport safety.
Section II - Packaging and Package Design Requirements (Continued)
Additional Requirements for Tyoe A Packages (cont'd.)
216 The containment system of the radioactive material is a special form encapsu-lation.
Prototypes of these encapsulations have been tested in accordance with the requirements of paragraphs 726-737, and have demonstrated retention of their contents. The containment is retained in the cask cavity which is sealed by the gasket and lid and the drain line plug. This seal is used only to retain the special form capsule in the cask cavity.
Denonstration of cask cavity leak rate is not required for special form contents.
217 Competent authority certification of special form material will be considered to be evidence of leak-tightness for the purposes of normal and accident conditions of transport.
218 While not defined as part of the contai ment system for the purposes of this application, it is noted that the cask lid is bolted in place and those bolts are independent of any other part of the package.
219 Ten years of experience with the Model 600 show no problems of material in-compatibility or pressure generation.
220 Not applicable - No liquid contents authorized.
221 The containment system is unaffected by full vacuum conditions as demonstrated in the special form leakage tests.
222 No /alves are used in the design.
223 The cask shielding closure is independent of the containment system, and cannot be opened inadvertently without removing the jacket hold-down bolts and breaking the seal wire.
224 The protective jacket is bolted closed during transport.
If the lug structures, used in the tie-down system, failed in such a manner to allow the package to become detached from the vehicle, the basic protective features of the protective jacket and the enclosed cask would remain unchanged. L' J
Section II - Packaging and Package Design Reauirements (Continued)
Additional Reouirements for Type A Packages (cont'd.)
225 Comments pertinent to normal conditions of transport are addressed in Section VII.
226 Not applicable - no liquid contents authorized.
227 Not applicable - no gaseous contents authorized.
Basic Additional Requirements The Model 600 package meets all Basic Additional Reauirements for Type B (U)
Packages (paragraphs 228 through 233) with pertinent details as follows:
228 The cask meets all additional requirements specified for Type A packages.
229 The demonstrations of compliance of the Model 600 to the requirements in and Section VII, paragraphs 713-721 is based on the test results of the 230 Model 100 and the appropriate applications of modeling theory in determining the major design parameters of the Model 600. The test of the Model 100 demonstrated its capability to withstand the conditions described in Section VII with no degradations of radiation shielding or loss of radioactive contents through leakage. Therefore, similar outcome can be projected for the Model 600.
The thermal analysis conducted for thermal test simulations on the Model 600, showed a maximum temperature gradient through the shielding cask below the melting point of lead (See Attachment 1 ).Special form contents, during the above conditions, would therefore have been subjected to conditions less severe than those required to demonstrate special form qualifications. The special form containment would therefore be retained in the cask cavity and would suffer no reductions in its certified integrity.
231a When solar energy absorption under normal ambient conditions is added to the internal heat generation, the maximum cask cavity temperature of the package will reach about 232 F (111 C).
This cavity temperature was determined by the application of THTD computer code, G.E.-N.E.D. version, a widely accepted computer code. The program assumed a heat load in the cavity of 600 watts.
(See Attachment 1 ).
This temperature is far below the special form test temperature of 800 C which causes no damage to the source.
' L
Section II - Packaging and Package Design Requirements (Continued)
Basic Additional Requirements '(cont'd.)
U 231b With an ambient temperature of 100 F (38 C) in the shade, the surface temperature of the package will not exceed 117 F (47 C) with a 600-watt decay heat load.
(See Attachment 1
).
232 Conditions used as stated in paragraoh 232 and Table III.(I) 233 The thermal protective jacket will not be degraded by the tests specified in paragraphs 709 through 714 and 719, nor will it be rendered ineffective by comon material handling mishaps.
(See Section VII).
Specific Additional Requirements The Model 600 package meets all Specific Additional Requirements for Type B (U)
Packages (paragraphs 234 through 241) as follows:
234 No filter or cooling system is used.
235 &
No venting or pressure relief system is used on the Model 600 package.
236 237 Normal operating 'ressure in the void space of the special form encapsulation (assuming 600 watts and 232 F (111 C) cavity' temperature during transport) is approximately 4.4 psig (30.4 kPa).
The testing required for special form certification demonstrates that the containment is adequate to withstand pressures substantially higher than 1.5 times the sum of this and reduced atmospheric pressure without failure.
238 Maximum cavity temperature during the thermal test of oaragraph 720 is 0
shown to be 486 F (252 C). This results in a 11.5 psi (79.5kPa) pressure rise in the void space of the source.
Again the special form test assures that the source will remain intact.
2 239
.iaximum operating pressure is less than 100 psi (7 kg/cm ), 689 kPa.
' L_
Section II - Packagin'g and Package Design Reouirements (Continued)
Specific Additional Requirements (cont'd.)
240 The maximum surface temperature of the Model 600 package has been estimated at 161 F(72"C)under normal conditions of transport (decay heat plus solar radiationabsorbed).
241
. Not applicable - no liquid authorized.
Section III - Exemptions - No exemotions authorized.
Section IV - Activity Limits The demonstration of leak tightness required to successfully pass the Special Form Certification tests effectively eliminates the need for the individual activity limits.
The -activity limits of the package are therefore set by the practical heat-transfer and radiation shielding capability. The activity limits for the Model 600 can be generically stated as: Any Solid Radionuclide in Certified Special Form con-tainment with a decay heat load up to 600 thermal watts.
Section V - Controls for Transport and Storage in Transit Mixed Packing 501 Mixed packaging is not authorized.
~.
502 The painted exterior surface of the package is designed to be easily decontaminated so the nonfixed radioactive contamination can be kept below the levels specified in Table XI, IAEA Section V.
503 The Model 600 package is categorized after loading in accordance with the surface radiation and transport index measured prior to consignment.
504-509 See paragraph 503, and 510-513.
' 510-513 The package will be labeled as required by IAEA and USD0T regulations.
514 The gross weight of the Model 600 package is marked in lbs. and kilograms on the nameplate.
515 Not applicable to Type B packages. m
Section V - Controls for Transport and Storage:in Transit Mixed Packing (Continued) 516 The package is marked with the identification of the competent authority and Type B (U) designation on the metal nameplate.
517 The outside of the package is marked with the Trefoil symbol on a metal nameplate.
518-The package will be labeled as required by IAEA and USD0T regulations.
519 Section VI - Fissile Materials 601 The Model 600 package, with fissile radioactive contents, has been analyzed for compliance with Section VI as follows:
General Provisions for Nuclear Safety 602 The Model 600 package has been designed to transport a maximum of 500 grams of U-235 or equivalent mass of mixed fissile Special Nuclear Material (SNM).
The equivalency is calculated using the following mass-isotope relationship:
Grams U-235, Grams U-233 + Grams Pu (Fissile) < l.0 500 300 300 The Model 600 package contents, as defined and restricted above, cannot be rearranged to reach a critical assembly under any circumstances of transport.
This analysis is based on the minimum quantity of fissile material required 1
to produce a critical assembly under optimum conditions These optimum conditions include:
(a) Water submersion to achieve optimum water moderation and full water reflection.
(b) Reduction in effectiveness of built-in neutron absorbers or moderators.
(c) Rearrangement into more reactive arrays.
(d)
Increased reactivity due to temperature changes.
' Carter, R.D., etal, " Criticality Handbook" ARH-600
' L
General Provisions for Nuclear Safety (continued) 603 The quantity of fissile :antents described in paragraph 6)2 shall be calculated on the basis of the burn-up.
If the irradiation history is not known, the contents shall be regarded as unirradiated with original fissile content.
Unspecified fissile materials with unknown enrichment, fissile mass, con-centration, etc. will not be considered for shipment in this package.
604 The Model 600 transport container is designed for classification as a fissile Class II package.
Provisions Specific to Fissile Class I Packages 605-Not applicable to fissile Class II packages.
614 Provisions Specific to Fissile Class II Packages 615 The Model 600 package has been designed to preclude damage from normal con-ditions of transport (ref. paragraphs 709-714) which would reduce the volume or increase the reactivity significantly or change the geometry or spacing of fissile contents more than 5%, or permit the entry of a 10-cm cube.
Water inleakage has been assumed in the limitation of fissile contents and allowable number of packages, as mentioned under paragraph 602.
616 The Model 600 package is designed to satisfy the nuclear safety criteria described in paragraphs 617 through 619.
The Individual Package Considered ih Isolation 617 The Model 600 package is,for the purpose of this analysis, assumed to be damaged in such a way as to allow water to fill all void spaces in the package.
618 The Model 600 package with fissile contents described in paragraph 602 will remain sub-critical with a mass basis safety margin in excess of 20% in the most reactive configuration with close full water reflection.
It is anticipated that the Model 600 package will totally contain and shield the radioactive special form fissile contents and that the special form con-tainment will preclude the leaching or dispersement of the radioactive contents. u
The Individual Package Considered in Isolation (Continued) 2 619 The Density Analog Method was used to calculate the allowaole number of similar package designs which could be stacked together and remain sub-critical under conditions in which the fissile material in a cask had come out of the special form containment and came together, and water had been introducs 'nto the cask cavity such that optimum water moderation and full water reflection of the fissile material was achieved.
The results of the analysis are tabulated below for various fissile contents:
Fissile Contents Maximum Number of Packages (assume 100%)
to Remain Sub-Critical 500 gms U-235 21 0 300 gms U-233 51 300 gms Pu 33 As there is no degradation of the cask or loss of the heat / crash shield under accident conditions, the flooding assumed in the analysis represents the worst case accident condition.
Considering the most restrictive case of Pu, an " allowable number" may be conservatively arrived at as follows:
Number Safe 33
=
Allowable Number 33/5 = 6
=
620 This package, with conten+.s as described in paragraph 602, has been previously approved by the U.S. competent authority in the issuance of Certificate No.
USA /5980/B( ) F, Rev. 3.
Paragraph 620, giving examples of package designs not requiring competent authority approval is not applicable.
621-Provisions specific to fis? 'e Class III packages are not applicable to 624 Class II packages.
2See License SNM-960, Section 5.4.4, Amend. No. 5, General Electric Company, Vallecitos Nuclear Center, Pleasanton, Ca., April 18, 1966 or Paxton, H.C., " Density - Analog Techniques", in Proceeding of the Livermore Array i
Symoosium, TID-4500, UC-46, CONF-680909, Livermore, Ca, September 23-25, 1968.
l L
ANALYSIS OF THE MODEL 100 PACKAGE FOR NORMAL AND ACCIDENT TEST CONDITIONS IAEA Section VII. Test and Inspection Procedures Demonstration of Compliance with the Tests 701 Demonstration of compliance with the test requirements of Section VII for the Model 600 was accomplished by performance of tests with a model (Model 100 cask) and the application of modeling theory in determin,ing the major design parameters of the Model 600. Geometric similarity and identical choices of materials were established between the Model 100 and Model 600 designs. The Model 100 has been approved pursuant to the 1973 IAEA Regulations under IAEA Certificate of Competent Authority No. USA /5926/
B(U), March 17,1980.
702 An ambient temperature of 100 F (38 C) was the basis for all test and analysis condition adjustments.
Tests of Packaging 703 One Model 100 specimen package was assembled as a prototype in 1968.
All destructive tests were accomplished on this prototype.
704 The Model 100 assembly was examined before testing, with no detrimental corro-sion and deterioration observed.
Some superficial scratching and rusting of the cask surface was observed, but these were due to normal material handling (since the cask had been in prior service for some 11 years) and would have no effect on the performance of the package. The Model 100 specimen used for the demonstration tests was within design tolerances.
705 The Model 100 specimen was dropped with its cavity full of water.
The cask drain plug impacted the protective jacket base causing a partial fracture of the protruding plug. A minor amount of water leaked out through this plug following the test, but no other leaks were sustained IAEA Section VII. Test and, Inspection Procedures (Continued) 705 Continued during the drop tests and no other evidence of cask deformation, fastener breakage, or other loss of integrity was observed. The drain plug was subsequently replaced by a socket head type which is fully recessed in the cask body. This replacement clearly eliminates *e potential for re-occurrence of this failure of integrity.
For IAEt ification, the con-tainment system is defined as the special form con ainment.
Since the test conditions required for qualification of Special Form encapsulations are significantly more severe than the conditions that this encapsulation would experience in the Model 600 cavity during IAEA accident conditions, no reduc-tion in special form integrity would result and no loss of radioactive material would occur.
706 External photographs were taken of the package during assembly and external and internal photos clearly demonstrate the location and extent of damage to the protective jacket and absence of damage (see 705) to the cask itself.
707 The integrity of the Model 600 package radioactive containment and radiation shielding follow.ing the IAEA specified normal and accident conditions is demonstrated by three separate bases.
First, the tests requireo for Special Form Certification demonstrates that the source integrity is mair.tained under severe conditions (Ref. 705).
Second, the drop test series (Raf. 719) on the Model 100 specimen and their extrapolation to the Model 600 demonstrate that in both cases the cask and protective jacket are not damaged by severe impact conditions and finally, the Transient Heat Transfer Analysis (see (Attachment 1) on the Model 600 demonstrates that no melting of lead would occur as a result of the 30-minute thermal test.
708 The target for the drop test series (Ref. 719) was a concrete pad 20 ft.
by 15 ft. by 6 ft. thick covered with a 10 ft square by 3-inch i; hick carbon steel plate. The total weight of the pad is estimated at 290,000 lbs and it was placed on 3 ft. of 95% engineered fill. The ratio of package weight to target pad was therefore approximately 1:64.
The target for the second drop was a vertical 6-inch diameter steel bar with a 1/4-inch edge radius, bolted vertically to the stcel target pad. L
IAEA Section VII.
Test and Inspection Procedures (Continued)
Normal Transport Test Conditions (Daragraohs 709 through 714) 709-710 (Ref. 711) 711 Water Soray Test. The protective jacket is constructed of welded and painted steel. Water spray will have no detrimental effect on this package.
712 Free Drop Test. The package is demonstrated to withstand a 30 ft.
(9 m) free fall with no impairment of radiation shielding, and no loss of dispersal of radioactive material (Ref. 719); therefore, it meets the requirements of the 1.2-meter (4-foot) free fall.
713 Compression Test. This test was deemed unnecessary as it is less restric-2 tive than the 25 psig (187,000 kg/m ) external pressure required by the USNRC.
714 Penetration Test, This test is less severe than the penetration test of 719b and was, therefore, not conducted on the Model 100 package.
l 715-717 No liquids or gases authorized.
Summary and Concl.
ins The assessments set forth above provide assurance there is no loss or dispersal of the contents during normal transport conditions, and there is no reduction in shielding effectiveness of the package.
Accident Test Conditions (paragraohs 718 through 721) 718 The ptodel 100 specimen package prototype was subject to the drop tests of paragraph 719a and 719b. The thermal test of paragraph 720 was examined analytically for the Model 600 cask. (See Attachment 1 )
719 Mechanical Test.
The Model 100 specimen was dropped 30 feet (9 m) on the drop pad described in paragraph 708.
It was then dropped 40 inches (1 m) on the vertical punch bar.
No evidence was found that the integrity of the package was diminished by these tests, nor that the damage would have any significant effect on the thennal analyses.
By the application of the Accident Test Conditions (paragraphs 718 through 721) continued) 719 Continued
... scale modeling relation established in the design of the Model 600.
Similar structural responses can be reasoned for the Model 600.
720 Thermal Test. The Model 600 package was analyzed for thermal transient effects due to the specified fire conditions using the THTD computer code, GE-NED version, a widely accepted computer code to model the actual package.
The results indicate that the maximum temperature in the lead shielding would be 519 F (270 C). On this basis, no melting of lead would be expected and the shielding integrity of the cask would be assured.
(See Attachment 2).
721 Water Imersion Test.
Prevention of water in leakage is not necessary since the special form contents will not release any radioactivity and the criticality analysis assumed optimum water moderation and reflection.
Water In-Leakage Test for Fissile Material Packages 722 The assessment described in Section VI assumed optimum water moderation and reflection.
Since the allowable number of packages (Ref. par. 619) for both normal and accident conditions far exceeds the practical quantity of casks available for consignment, the package meets this requirement with the maximum proposed fissile loading.
723-These paragraphs describing immersion test conditions is not applicable.
724 (Ref, par. 722).
Summary and Conclusions 725 The effect of accident conditions described in paragraphs 709 through 720 to the Model 100 prototype package and the extrapolation of these effects to the Model 600 package indicate no degradation to the Model 600 package that would adversely affect the criticality safety, or integrity of the "diation shielding or the special form radioacti/e material containment.
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ATTACHMENT 1 THERMAL ANALYSIS MODEL 600 TRANSPORT PACKAGE MAY 1980 l
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THERMAL ANALYSIS
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GE MODEL 600 TRANSPORT PACKAGE A.
PACKAGE AND CONTENTS DESCRIPTION 1.
Package The General Electric Model GE-600 shipping cask, Figure 1, is a stainless steel weldment consisting of inner and outer concentric cylindrical shells each closed at the bottom and connected to a common flange at the top.
The 6" gap between the shells is filled with lead.
The inner shell has a 20.5 inch inside diameter, is 3/8-inch thick, and 46 inches high.
The outer shell is 34 inches in outside diameter, is 3/8-inch thick, and approximately 60 inches high.
The cask uses a lead filled, stainless steel-weldment, plug-type closure.
Lid shielding thickness is 6".
The cask is sealed with a silicone rubber gasket, the closure is retained by six, one-inch diameter steel bolts.
The cask is protected from mechanical a' d thermal assault hy a double-n walled protecti e enclosure.
Each of the close-fitting concentric enclosure shells is 1/2-inch thick carbon steel.
This structure fits over the cask and ' olts to the shipping pallet.
The double-walls pro-z vide thermal protection to the cask under firo-accident colditions.
The enclosure also functions as an ener j absorbing structure to s
preclude mechanical damage to tne cask under accident conditions.
The cask is shipped in the upright position and the total loaded weight is approximately 21,000 pounds.
Heat transfer through the structures and dissipation to the environs by the package is accomplished by a combination of modes.
The content's s
decay heat is transferred to the cask cavity walls by thermal convection and radiation within the cavity, as well as by gamma energy deposition.
Heat flow through the cask structure is by conduction.
It is believed that a small gap may exist at the lead-to-outer shell interface (created 1
by differential contraction) and a combined conduction-radiation process occurs across it.
Heat transfer from the, 1) cask surface to the i -
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FIGURE 1.
GENERAL ELECTRIC MODEL 600 SHIELDED CONTAINER L I
protective enclosure ID, 2) between protective enclosure shells, and
- 3) from the outer surface of the protective enclosure to the environ-ment is by radiation and natural convection.
Conduction, of course, occurs through the protective enclosure structure.
The design of the shipping pallat and protective enclosure allows some vertical air flow through the various annuli by providing entry ports at the base and exit ports at the top.
The amount of heat removed by this convective process is thought to be relatively small.
2.
Contents For purposes of this analysis, the contents are considered to be a collection of irradiated fuel rods each contained in a Special Form capiule.
The number of such capsules could range from one to 37, each held in a basket or fixture within the cask cavity to maintain its position during transit.
The total decay heat load of the collective contents will not exceed 600 watts; no single rod or capsule is thought to exceed 65 watts.
Heat transfer from the contents to the surrounding cask cavity is by a combination of thermal radiation and natural convection, with radiation being predominant.
Air is assumed to be the working fluid in the cask cavity. A certain fraction of the decay heat is deposited in the cask structure through the gamma-heating phenomenon.
Conduction of heat from the internal fuel holder to the cask is thought to be negligible.
B.
SUMMARY
DESCRIPTION OF ANALYTICAL METHODOLOGY 1.
Package The package consisting of the cask, pallet and overpack, was modeled axisymmetrically (two-dimensions) using the Transient Heat Transfer computer program, Versio.1 D (THTD).
This program has been used by General Electric Company extensively over the past 15 years on a variety of thermal problems, including shipping casks.
The ability to accurately model a shipping package such as the 600 was established by benchmarking the THTD Code against the thermal test t
4 1
results of a geometrically similar cask, the GE Model 1500.
An analysis of the 1500 cask THTD model and the thermal test results shows very good agreement.
For the purposes of this evaluation, the Model 600 package THTD model with a 600-watt internal heat load was subjected to both nomal conditions of transport and accident conditions, as defined in IAEA regulations, (Safety Series 6, 1973 Revision).
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2.
Contents The contents of the Model 600 cask were roughly simulated in the THTD model to provide the proper heat load, axial distribution, mass and specific heat, but the model geometry differed from the actual fuel array.
The Special Fonn capsules forming the contents are required by regulations 0
to sustain a temperature of 1475 F for ten minutes, confirmation of the 4
design is by prototype testing. Thus, the requirements for analysis of the contents are limited to those bounding cases which indicate that the capsules, when contained in the Model 600 package, will not exceed the Special Form test parameters.
THTD was not employed for this purpose, rather more simplistic calculations were found to provide sufficient accuracy.
THTD model results constituted the boundary temperatures for the evaluation of the contents.
C.
SUMMARY
OF RESULTS AND CONCLUSIONS 1.
Normal Conditions of Transport - Package IAEA regulations require the thermal evaluation of the package under both full-sun and shaded conditions.
In each case, the ambient air is at a temperature of 100 F with no movement other than free convection from the package.
The following table, Table I, shows steady-state temperatures at various U
locations in the Model 600 shipping package under conditions of 100 F (33 C) ambient still-air, with and without maximum insolation and maximum inte.nal heatload.(600 watts). As can be seen, the solar radiation contributes significantly to the package temperatures.
It is estimated that the a
4_
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TABLE 1 NORMAL TRANSPORT TEMPERATURES With Solar Radiation Shade Location Temoerature, OF Temoerature, OF Cask Cavity Maximum 196 151 Average 194 143 Cask Closure Seal 188 132 Cask Outside Surface Maximum, Side 194 146
~
Average, Side 191 139 Maximum, Top 190 112 Average, Top 189 111 Protective Enclosure Inner Shell Maximum, Side 181 129 Average, Side 177 125 Maximum, Top 190 122 Average, Top 186 121 Protective Enclosure Outer Shell Adjusted Top 161 Maximum, Top
- 196 113 Average, Top
- 188 112 Maximum, Side
- 177 122 Average, Side
- 170 117 Maximum Lead Temperature 197 151 U
Ambient Air Temperature 100 (38 C) 100 (38 C)
- These are THTD model temperatures which have not been corrected for shadowing and sun-angle. _
u-
solar radiation introduces approximately 4 KW to the package. This is thought to be conservative since the THTD model has such insolation on all surfaces, except the bottom; no sun-angle shading or exposure time were considered.
To account for the shadowing of the top surface of the protective enclosure by its irregular configuration (see Figure 1), the average temperatures of that surface under both full-sun and shaded conditions were computed.
. For the full-sun condition, this temperature is 187.8 F, while for fully-shaded condition the temperature is 111.2 F.
The actual temperature of the top surface considering the partial shadowing may be estimated by summing the area-fractions of both full-sun and shade surface temperatures. This is reasonable due to the good thermal conductivity of the carbon steel enclosure.
This propor-tioning of temperature is as follows:
Total top area, A = 13.6 ft.2 T
Mininum shaded area, A3 = 4.8 ft.2 Maximum exposed area, AEX = 8.8 ft.2 Top (111.2 F) + 1,8.8 (187.8 F) 0 161 F T
=
=
The above estimate is thought to be conservative since actuai areas rather than projected areas were used.
The enclosure side temperatures as calculated by THTD are also overstated but no adjustment will U
be performed since they are less than the 180 F limit.
This is the result of a high sun angle which minimizes the shadows but still shines on the largest surface area. _
2.
Accident,- Thermal Conditions - Package The following table, Table 2, shows maximum temperatures at various locations and the time of the occurrence of these temperatures, in 0
minutes following the start of the 30 minute 1475 F (800 C) fire.
TABLE 2 ACCIDENT CONDITION TEMPERATURES Minutes After Location Peak Temperature, OF Start of Fire Cask Cavity 486 101 Cask Closure Seal 538 55 Lead Temperature 519 61 Cask Outside Surface 598 43 Protective Jacket Inner Shell 1335 30 Outer Shell 1398 30 3.
Normal / Accident Conditions - Contents As described above, the contents have been evaluated by bounding calculations to determine if there are normal or accident conditions which could lead to a Special Form capsule temperature in excess of 1475 F.
The contents were assumed to be held in a new internal structure with a 36 capsule capacity in two concentric rows.
A maximum 65-watt heat load per pin was also assumed The results of this analysis are as follows:
cavity wall = 486 F (Max. Fire conditions)
T Table 2 712 F T
=
pin
~
_7_
t_
65 watts x 1.5 peaking factor O
=
pin 97.5 watts
=
4.
Conclusions The results of the THTD analyses of the GE Model 600 Cask under the normal and accident condition, as prescribed in Federal and International regulations, show that the package is capable of meeting all thermal requirements.
The protective enclosure produces sufficient thermal insolation of the cask during the accident conditions to prevent melting of any lead shielding. Under shaded normal conditions, jacket surface temperature 0
0 is less than the 122 F (50 C) limit.
Additionally, the 180 F (82 C) limit is not exceeded when solar radiation is considered.
Even under package peak thermal conditions, the temperature of the worst-case Special Form capsule remains substantially below its qualifying test temperature. Thus, the contents are thermally safe under all regulatory conditions.
- L
k.'
ATTACHMENT B COMPILATION OF APPLICATIONS IN SUPPORT OF CERT 1 CATE OF COMPLIANCE N0. 5980 AND CERTIFICATE OF COMPETENT AUTHORITY NO. USA /5980/B( )F MAY 1980
DEP ARTMENT OF TRANSPORTATION
/
{
RESEARCH ANO SPECI AL PROGR AMs ADMINis7 RATION
,J WAsmNc CN : C.
20=s0 g%
car.:t c' IAEA CERTINCATE OF COMPETINT ALTHORITY THE A W"w ST R ATc P ngags :
Tvee B Fissile Radioactive Material Package Design Certificate Number USA /5980/3( )F (Revision 3)
This establishes that the packaging design described herein, when loaded with the authorized radioactive contents, has beet. certified by the National Co=petent Authority of the United States as meeting the regula-tory requirements for Type B packaging for fissile radioactive materials I
as prescribed in IAIA Ragulgtions and 49 CFR 89173.393a,173.395(c)(2) and 173.395(c)(3) cf the USA Regulations for the transport of radioactive = steri-1s.
I.
Package Identification - CE-600.
II.
packagine Description - Packaging authorized by this certificate consists of a lead-shielded steel cylinder 34" in diameter by 60" high further enclosed in a double-valled protective jacket of 1/2" steel plates and weighing abcut 18,500 pounds.
III.
Authorized Radioactive Contents - The authorized contents consist of radioactive esterials, n.c.s., and fissile radioactive materials as further limited in U.S. Nuclear Regulatory Co= mission Certificate No. 5930 (Appendix A).
Fissile shipments are authorized as Fissile Class III with no more than two packages per single vehicle or stovage area.
IV.
General Conditions -
a.
Each user of this certificate must have in his possession a copy of this certificate.
b.
Each user of this certificate, other than General Electric Company, Pleasanton, California, shall register his identity in writing to the Office of Hazardous Mattrials Regulation, U.S. Department of Transportation, Washington, D.C. 20590.
c.
This certificate does not relieve any consignor or carrier fro: compliance with any requirecent of the Govern =ent of any country through or into which the package is to be transported.
i I
l 1
Certificate Number USA /5950/E( )?, Revisi:n 3 Fage 2 V.
Markin: and Labelin: - The package cus: bear the =arking USA /5950/E( )7 as well as the o:her =arking and labels prescribed by the USA Regula: ions.
VI.
Erpiration Date - This certificate, unless renewed, expires on December 31, 1984.
This certificate is issued in accordance with the requiremen:s cf the IAEA and USA Regula: ions and in response to the January 15, 1980, pe:ition by General Electric Company, Pleasanton, California, and in consideration of the associated information provided in U.S. Nuclear Regulatory Co==ission Certificate of Compliance 5950 (Appendix A).
Cer:ified by:
b 0$ld R. R. Rawl y'
Designated U.S. Co=peten: Authority for the International Transportation of Radioac:ive Materials Office of Hazardous Materials Regulation Materials Transporta: ion Bureau U.S. Department of Transportation 1" Safety Series No. 6, Regula: ions for the Safe Transport of Radioactive Materials, 1967 Edition" published by the International Atomic Energy Agency (IAEA), Vienna, Austria.
Title 49, Code of Federal Regula: ions, Parts 100-199, USA.
Original issued in response to :he March 12, 1975, petition by the General Electric Co pany, Pleasanton, California.
Revision 1 issued to incorporate Revision 1 of USNRC Certificate No. 5980 and to ex:end expiration date in response to the January 17, 1978, petition by tne General Electric Company, Pleasanton, California.
Revisie.t 2 issued to incorpora:e Revision 3 of USNRC Certificate No. 59S0 and to axtend expiration date in response to the October 10, 1979, petition by the General Electric Company, Pleasanton, California.
Revision 3 issued to incorporate Revision 4 of USNRC Certificate of Compliance Nc. 5980 and :o extend expiration date.
e Form NRC.618 U.S. NUCLEAR REGULATORY CoMMISSloN n2 31 CERTIFICATE OF COMPLIANCE
,g e p 7, For Radioactive Materiets Packages 1.tz) Certificate Number 1.(b) Revision No.
1.(c) Package Identifiestion No.
1.(d) Pages No. 1.(e) Total No. Pages 5980 4
USA /5980/Bf IF 1
3
- 2. PRE AMBLE 2.(a)
This certificate is issued to satisfy Sections 173.393a,173.394,173.395, and 173.396 of the Department of Transportation Haaerdous Materials Regulations (49 CFR 170139 and 14 CFR 103) and Sections 146-19-10s and 146-19-100 of the Department of Transportation Dangerous Cargoes Regulations (46 CFR 146-149), as amendeo.
2.(b)
The packaging and contents described in item 5 below, meets the safety standards set forth in Subpart C of Title to, Code of Federal Regulations, Part 71," Packaging of Radioactive Materials for Transport and Transportation of Radioactive Material Under Certain Conditions."
2.(c)
This certificate does not relieve the consignor from compliance with any requirement of the regu'letions of the U.s. Department of Transportation or other applicable regulatory agencies, including the government of any country through or into which the package will be transported.
- 3. This certificate is issued on the basis of a safety analysis report of the package design or application-3.(a)
Prepered by (Name and address):
3.(b)
Title and identification cf report or application:
Gsneral Electric Company General Electric Company application dated P. O. Box 460 October 10, 1979 Pleasanton, California 94566 71-5980 3,,,,
4 CoNDIT7NS This certificate is conditional upon the fulfilling of the requirements of Subpart D of 10 CFR 71, as applicable, and the conditions specified in item $ below.
- 5. Description of Packaging and Authorized Contents, Model Number, Fissile Class, Other Conditions, and
References:
(a)
Packaging (1) Model No.:
GE-600 (2)
Description A steel-encased lead shielded shipping cask.
The basic cask body is a cylinder 34 inches in diameter by 60 inches high formed by two concentric steel shells whose annular region is filled with 6 inches of lead.
The cavity is 20-1/2 inches ID by 46 inches high, 3/8-inch thick stainless steel cylinder. A recessed plug-type cask lid, consisting of a steel weldment filled with lead, is secured to the cask body by six,1-inch diameter steel bolts.
A silicone rubber gasket provides the seal. A protective jacket consisting of a double-walled structure of 1/2-inch thick carbon steel plates is placed over the cask and bolted to a steel pallet by eight, 2-inch diameter steel bolts.
The cask has one 1/2-inch diameter drain line from the cavity to the outer shell.
The drain line is closed with a plug which has a melting temperature of 200*F..
The cask is shipped in the upright position.
The total weight of the package is approximately 18,500 pounds when loaded.
t
Page 2 - Certificate No. 5980 - Revision No. 4 - Docket No. 71-5980 5.
(a)
Packaging (continued)
(3) Drawings The packaging is constructed in accordance with the following General Electric Co. Drawings Nos. :
)
212E247, Rev. 5 10603898, Rev. 3 144F650, Rev. 1 693C293, Rev. 3 161F470, Rev. 2 106D3892, Rev. 3 12904685, Rev. 0 129D4684, Rev. 0 195F162, Rev. 0 (b) Contents (1) Type and form of material (i) Byproduct material and irradiated special nuclear material in solid or solid oxide form, but specifically not loose powders.
Contents are to be clad, encapsulated or contained in a metal encasement of such material as to withstand the combined effects of the internal heat load and the 1475'F fire with the closure pre-tested for leak tightness.
Also, byproduct material and irradiated special nuclear material in special form.
(ii) Neutron sources in special form.
(2) Maximum quantity of material per package Plutonium in excess of twenty (20) curies per package must be in the form of metal, metal alloy or reactor fuel elements, and (1)
For the contents described in 5.(b)(1)(i) the maximum decay heat not to exceed 600 watts and the fissile content not to exceed 500 grams of U-235, 300 grams U-233, 300 grams Pu, or a prorated quantity of each such that the sum of the ratios does not exceed unity; or no't to exceed 1200 grams fissile provided:
(1) the' fissile material is contained 'a schedule waste liners constructed of 5-inch schedule l
40 pipe with a maximum inside length of 39-5/16 inches, (2) no more than four such liners are shipped at one time, (3) each liner contains no more than 300 grams fissile, and (4) the cask is provided with a positioning lattice to maintain separation between the liners.
j Page 3 - Certificate No. 5980 - Revision No. 4 - Docket No. 71-5980 5.
(b)
Contents (continued)
(2) Maximum quantity of material per package (continued)
(ii) For the contents described in 5(b)(1)(ii) the maximum decay heat not to exceed 50 watts and not more than 500 grams U-235 equivalent mass.
The external dose rate not to exceed 1000 mrem /hr at 3 feet from the surface of a package without internal shiciding.
l (c)
Fissile Class III Maximum number of packages per shipment Two (2) 6.
The U-235 equivalent mass shall be determined by the following method:
U-235 equivalent mass equals U-235 mass plus 1.66 times U-233 mass plus 1.66 times Pu mass.
7.
Except for the neutron sources in special form (10 CFR 571.4(o)), the package contents shall be dry and the fissile material unmoderated (H to X atomic ratio less than 2).
8.
Prior to each shipment, the packaging primary lid closure seal shall be inspected.
The primary lid closure seal shall be replaced with a new seal if inspection shows any defects or every twelve (12) months, whichever occurs first.
9.
The package authorized by this certificate is hereby approved for use under the general license provisions of 10 CFR 571.12(b).
10.
Expiration date:
December 31, 1984.
REFERENCES General Electric Company's application dated October 10, 1979.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION hCharlesE.MacDonald, Chief
.)
Transportation Certification Branch Division of Fuel Cycle and Material Safety DEC 311979 De
i e
v u,j, e.t :. u m L [9 E L E u, t ti l v.
f d
NUCLEAR ENERGY
- r. -
r ENGINEERING l
GENERAL ELECTRIC COMPANY, P.O. BOX 460 PLEAsANTON. CALIFORNI A 94566 DIVISION t
October 10, 1979 Mr. Charles E. MacDonald, Chief Transportation Branch Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Ref:
Certificate of Como11ance No. 5980
Dear Mr. MacDonald:
The General Electric Co., Vallecitos Nuclear Center (VNC), requests that Certificate of Compliance No. 5980 for the G.E. Model 600 shipping con-tainer be renewed.
In support of,this request we are enclosing a consolidation of our original application.with subsequent applications'for amendment into a single package. This consolidation' includes the changes contained in our appli-cation of October 8, 1979. The submittal of May 21, 1974, a criticality analysis demonstrating the safety of the approved 1200 gram fissile loading is enclosed as Attachment A.
As the certificate expires on October 31, 1979, and as the container is in frequent use, VNC requests that a temporary extension be granted to permit the use of the container while the application for renewal is evaluated.
Enclosed is a check for $150.00 for the renewal fee as required by 10CFR170.31.
If your staff has any questions concerning this application for renewal, please contact me at 415-862-2211, Ex. 4330.
Sincerely,
- 8. [. [
G. E. Cunningham Sr. Licensing Engineer
/11 cc:
R. R. Rawl U.S. Dept. of Transportation Office of Hazardous Materials Operations Washington, D.C. 20590 l
GENERAL ELECTRIC SHIELDED CONTAINER - MODEL 600 1.0 Package Description - Packaging (a)
General:
All containers of this model, for purposes of constructing additional containers of this model, will have dime,nsions of plus or minus 5% of the container dimensions spec-ified in this application, and all lifting and/or tiedown devices for additional con-tainers of this model if different from the lifting and/or tiedown devices described in this application will satisfy the requirements of 11CFR71.31(c)(d).
Thi.s cantainer 'is detailed in G.E. drawings 161F470, 693C293, 144F650, 212E247, 106D3892, 129D4684, 129D4685, 195F162, and 106D3898 attached.
Shape:
An upright circular cylinder shielded cask and an upright circular cylinder protective jacket with attached square base.
Size:
Shielded cask is 34 inches in diameter by 59-7/8 inches high. The protective jacket is.72-7/8 inches high by 63-1/2 inches across the box section.
The base is 63-1/2 inches square.
Construction:
The cask is a lead-filled carbon and stainless steel weldment. The protective jacket is a double walled structure of 1/2 inch carbon steel plate and surrounds the cask during transport. The square base is 1/2 inch carbon steel with four I-beams attached.
1.0 Packags Description - Packaging (continued)
(a)
(continued)
Weight:
The cask weighs 15,000 pounds. The protective jacket and base weigh 3,000 pourids.
The weight of the additional plastic shielding for neutron sources is approximately 750 pounds. The lead liners weigh 500 pounds and 4810 pounds respectively for a maximum total weight of approximately 23,300 pounds.
(b)
Cask Body Outer Shell:
3/8 inch thick steel plate, 59-7/8 inches high by 34 inches diameter with a 3/8 inch bottom plate and a 3/8 inch top flange.
Cavity:
3/8 inch stainless steel wall and bottom plate. 20-1/2 inches' inner diameter by 46 inches deep.
Shielding Thickness:
6 inches of leak on sides, 6 inches of lead beneath cavity.
In the wet shipment case'the cavity will be filled with water with a 5 inch air space.
Penetration:
One 1/2 inch outer diameter by 0.063 wall stainless steel tube gravity drain line from the center of the cavity bottom to the side
~
of the outer shell near the cas t bottom with a 1/2 - 14 NPT pipe plug.
Genert.1 Electric may, at its discretion, permanently close and seal the drain line for this container model with no interference to other structural properties of the cask.
Filters:
None.
Lifting Devices:
Two diametrically opposed ears welded to sides of cask, covered by protective jacket during transport. Two additional ear mounting hole i
patterns are provided for redundant ears which are shipped separately.
1.0 Package Description - Packaging (continued (b)
Cask Body (continued)
Primary Coolant:
l Air (c)
Cask Lid Shape:
A right conical ' cylinder attached to flat plate Size:
Top plate is 34 inches diameter by 3/4 inch thick.
Bottom plate is 24 inches diameter
- by 3/8 inch thick.
The conical cylinder is 26-1/4 inches diameter at top by 6-3/8 inches high by 24 inches diameter at bottom.
Construction:
Lead-filled steel c'.ad cylinder welded to circular steel plates.
Closure:
Six - 1 inch UNC-2A steeT bolts equally spaced 60* apart on a 30 inch diameter bolt circl.
Closure Seal:
A minimum 3/16 inch thick flat silicone rubber or equivalent gasket between body and lid.
None.
Snield Expansion Void:
None.
Weight:
l 1450 pounds.
Lifting Device:
Single steel loop,1 inch diameter steel rod located in center of lid top. Covered by pro-tective jacket during transport.
4
1.0 Package Description - Packaging (continued)
(d)
Liner - Inner. Liner:
Shape:
Basically a right circular cylinder with a hollow center.
Size:
7-1/2 inch outer diameter by 3-3/8 inch inner diameter by 36 inches high.
Construction:
1/8 inch thick top and bottom circular plates welded to 1/8 inch thick lead-filled stainless steel clad cylinders.
Weight:
l 500 pounds Attachments:
3/8 inch diameter stainless steel curved rod 7 inches long is attached to the top plate.
Liner - Outer Liner (Body)
Shape:
Basically a right circular cylinder with a hollow center.
Size:
20 inch outer diameter by 7-7/8 inch inner diameter by 43 inches high.
Construction:
3/8 inch thick top and bottom circular plates welded to 3/8 inch thick lead filled stainless steel clad cylinders.
Weight:
l 4530 pounds (without lid).
Attachments:
Two - 3/4 - 10VNC-2B x 2 inch deep holes 180 apart in top plate to accept a 3/4 - 100NC-2A eyebolt.
Liner - Outer Liner (Lid) i Shape:
Two right_ circular cylinders of different i
diameters attached to flat plates.
1.0 Package Description - Packaging (continued)
(d)
Liner - Inner Liner (continued)
Liner - Outer Liner (Lid):(continued)
Size:
All plate is 3/8 inch thick. Top plate is 17 inches diameter while the bottom plate is 14 inches diameter.
Both cylinders are 2 inches high.
Construction:
Lead filled steel clad cylinders welded to i
circular steel plates.
Weight:
l 280 pounds.
Liner - Poly-ethylene:
Shape:
A right circular cylinder with a hollow center cavity.
Size:
20-inch outer diameter by 38-3/4 inche's high;-
the inner cavity is 2-1/4 inches in diameter by 22-1/4 inches deep; the liner is equipped with an 8-inch diameter by 8-1/2 inch stepped plug.
Construction:
The liner is of 6061 aluminum with welded construction. The cylinder walls are 1/8-inch thick. The top and bottom plates on the lid and the bottom of the cylinder are 1/4-inch thick. The top plate of the cylinder is 1/2 inch thick.
Shielding is provided by 2%
borated polyethylene filler.
Weight:
l Approximately 750 pounds.
Attachments:
Two 3-inch by 3-inch by 1/2-inch ears with l-1/2 inch diameter connection holes for lifting device attachments.
I l
1.0 Package Description - Packaging (continued)
(e)
Protective Jacket 'qdy_
Shape; Basically a right circular cylinder with open bottom and with a protruding box section diametrically across top and vertically down sides.
Size:
72-7/8 inches high by 63-1/2 inches wide across the box section.
Outer cylindrical diameter is 40-1/2 inches.
Inner diameter is 36 inches.
A 5-1/2 inch wide by 1/2 inch thick steel flange is welded to the outer wall of the open bottom.
Construction:
Carbon steel throughout.
Double walled con-struction.
The walls are 1/2' inch thick.
One and a quarter inch air gap between cask shell and inner jacket wall and a one and a half inch gap between inner and outer jacket walls, throughout.
Six 12 inch high by 1/2 inch thick gussets are welded to the outer cylin-drical wall and flange.
Including the two box sections, the gussets are spa.ced 45* apart.
Attachment:
Eight inch bolts connect the protective jacket body, through the flange to the pallet.
Jacket Lifting Two rectangular 1 inch thick steel double Device:
loops located on top of the box section at the corners. The loops are respectively 7 inches long by 3 inches high by 6 inches wide and 9 inches long by 3 inches by 6 inches. These loops are used only for lifting the jacket, not the complete assembly.
1.0 Package Description - Packaging (continued)
(e)
Protective Jacket Body (continued)
Assembly Lifting and Two diametrically opposed 3 inch thick steel Tiedown Devices:
ears welded to sides of box section, each ear has a 1-1/2 inch hole.to accept cable clevis or cable; or an optional rectangular tie-down yolk fabricated 3/16-inch wall rec-tangular milTed steel tubing. This yolk also has 3-inch thick shackle ears and surrounds the protruding box section on top of the cylindrical jacket.
Slots along periphery of the protective jacket at the bottom, slots in box section under lifting loops. Allows natural air circulation for cooling.
(f) protective Jacket Base Shape:
, Hollow cylindrical weldment.with square bottom plate.
Eight I-beams are welded to the bottom pTate.
Size:
Bottom plate is 63-1/2 inches square and 1/2 inch thick.
The cylindrical collar is 36 inches in outer diameter by 3 inches high.
The I-beams are 3 and 4 inches hi.gh by 63-1/2 and 62-3/8 inches long.
Construction:
The cylindrical collar houses two sets of
'l 1-1/2 inch by 1-1/2 inch by 1/8 inch steel energy absorbing angles separated by a 1/2 inch thick carbon steel mid-plate. The cask rests on this assembly. The collar is welded M the 1/2 inch thick carbon steel base plate.
Four I-beams are welded in parallel to the
~
base plate.
Atcachment:
Eight 2-inch diameter nuts are welded to the i
bottom of the base for jacket attachment.
l
e.
2.0 Package Description - Contents (a)
General Radioactive material as the metal or metal oxide, but specifically not loose pc<ders.
Neutron sources and other radioactive materials in special fonn.
(b)
Form Clad, encapsulated or contained in a metal encasement of such material as to withstand the combined effects of the internal heat load and the 1475'F fire with the closure generically pre-tested for leak tightness.
(c)
Fissile Content Not to exceed 500 grams of U-235, 300 grams U-233, 300 grams Pu, or a prorated quantity of each such that the sum of the ratios does not " exceed unity; or not to exceed 1200 grams fissile provided:
(1) the fissile material is contained in standard waste liners con-structed of 5-inch schedule 40 pipe with a maximum i'nsid' length of 39-5/16'inchei, e
(2) no more than four such liners are shipped at one time, (3) each liner contains no more than 300 grams fissile, and (4) the cask is provided with a positioning lattice to main-tain separation between.the liners.
(d)
Radioactivity That quantity of any radioactive material which does not generate spontaneously more than 600 thermal watts by radioactive decay and which meets the requirements of 49CFR173.393.
Shipments of neutron sources will be limited to 50 thermal watts.
(e)
Heat Total maximum internally generated heat load not to exceed 600 thermal w'tts.
(50 thermal a
watts for neutron sources).
-9 2.0 Package Description - Contents (e)
Heat (continued)
Although equilibrium temperature recordings were not taken for this package loaded to 600 watts themal, General Electric will analyze by test or other assessment each con-tainer heat loading with and without liners, as the specific case dictates, prior to ship-ment to verify that the requirements of 10CFR71.35 will be satisfied.
Reference is made to the GE - Model 100 Application, Section 5.1, Exhibit B, for a method of internal heat load analysis and heat dissipation.
3.0 Package Evaluation (a)
General There are no components of the packaging or its contents which are subject to chemical or galvanic reaction; there is no coolant except for air.
The protective jacket is bolted closed during transport.
A lock wire and seal of a type that must be broken if the package is opened is affixed to the cask closure.
If that portion of the pro-tective jacket which is used in the tie-down system or that portion which constitutes the principal lifting device failed in a manner i
to allow the protecti've jacket to separate from the tiedown and/or lifting devices, the basic protective features of the protective jacket and the enclosed cask would be retained.
The package (contents, cask, and protective jacket) regarded as a simple beam supported at its ends along its major axis, is capable of withstanding static load, nomal to and distributed along its entire length equal to five times its fully loaded weight, without generating stress in any material of the packaginj 3.0 Package Evaluation (continued)
(a)
General (continued) in excess of its yi ld strength. The pack-aging is adequate to retain all contents when subjected to an external pressure of 25 pounds per square inch gauge.
Reference is made to the GE-Model 100 Application, Section 5.1, Exhibit C, for a method of determining static loads.
The calculative methods employed in the. design of the protective jacket are based on strain rate studies and calculations and on a literature search of the effects on materials under im-pact conditions.
The intent was to design a protective jacket that would not only satisfy the requirements of the U.S.. Nuclear Regulatory Commission and the Department of Transportation proscribing the procedures and standards of packaging and shipping and the requirements governing such packaging and shipping but would protect the shielded cask from significant deformation in the event of an accident.
In the event that the package was involved in an accident, a new protective jacket could be readily supplied and the shipment continued with minimal time delay.
The effectiveness of the strain rate calcula-tions and engineering intuitiveness in the design and construction of protective jackets was demonstrated with the General Electric Shielded Container - Model 100 (Ref.: Section 5.1.3 of the Model 100 Application). The protective jacket design for the General Electric Shielded Container - Model 600 will be scaled from the design of the Model 100 in accordance with the cask weight and dimensions,
11_
3.0 Package Evaluation (continued)
(a)
General (continued) maintaining static load safety factors greater than or equal to unity, and in l
accordance with the intent to protect the shielded cask from any defomation in the
~
event of an accident.
1 (b)
Normal Transport Conditions Thermal:
Packaging components, i.e., steel shells and lead are unaffected by temperature extremes of -40*F and 130*F.
Package contents, at least singly-encapsulated or contained in spec.ification 2R containers, but not limited to special form, will not be affected by these temperature extremes.
Pressure:
The package will withstand an external pressure of 0.5 times standard atmospheric pressure.
The cask was pressure tested under water to 4
11 PSIG with no detectable leakage.
j Vibration:
Inspection of the Model 600 casks used since 1961 reveals.no evidence of damage of signifi-l-
cance to transport safety.
Water Spray and Since the container is constructed of metal, Free Drop:
there is no damage to containment resulting from dropping the container throu~gh the stan-dard drop heights after being subjected to water spray.
Penetration:
There is no effect on containment or overall spacing from dropping a thirteen pound by 1-1/4 inch diameter bar from four feet onto the most vulnerable exposed surface of the packaging.
r 3.0 Package Evaluation (continued)
(b)
Normal Transoort Conditions (continued)
Compression:
The loaded container is capable of with-standing a compressive load equal to five times its weight with no change in spacing.
Summary and
Conclusions:
The tests or assessments set forth above provide assurance that the product contents are contained in the Shielded Container -
Model 600 during transport and there is no reduction in effectiveness of the package.
(c)
Hypothetical Accident Conditions General:
The effectiveness of the strain rate calcu-lations and engineering intuitiveness' in the design and construction of protective jackets was demonstrated with the GE Shielded Container - Model 100 (Ref.:
l Section 5.1.3 of the Model 100 Application).
~
Extrapolations of the Model 100 data were used in the design and construction of the GE Model 600 protective jacket. The in-
~creased weight and dimensions of the Model 600 container over the Model 100 con-tainer necessitated a protective jacket wall of 1/2 inch steel compared to a 1/4 inch wall for the Model 100.
Drop Test:
The design and construction of'the GE Model 600 protective jacket was based on an extrapo-lation of the proven data generated during the design and construction of the GE Model 100 and on the results of cask drop experi-ments by C.B. Clifford(1)(2) and H. G. Clarke, Jr.(3)
The laws of similitude were used in an analytical evaluation (3)(4) to determine the protective jacket wall thickness
3.0 Package Evaluation (continued)
(c)
Hypothetical Accident Conditions (continued)
Drop Test: (continued) that would withstand the test conditions of 49CFR173.398(c) and 10CFR71.36 without breaching the integrity of the Model 600 cask. The evaluation, described in G$ -
Model 1000 Application, Section 5.9, Exhibit A, indicated a protective jacket wall thickness of 1/2 inch.
(1)
C.B. Clifford, The Design, Fabrication and Testing of a Quar'.er Scale of the Demonstration Uranium Fuel Element Shipping Cask, KY-546 (June 10,1968).
(2)
C.B. Cl-ifford, Demonstration Fuel Element Shipping Cask from Laminated Uranium Metal-Testing Program, Proceedings of the Second InternationaT Symposium on Packaging and Transportation of Radioactive Materials, Oct. 14-18, 1968, pp. 521-556.
(3)
H.G. Clarke, Jr., Some Studies of Structural Response of Casks to Impact, Proceedings of the Second International Symposium of Packaging and Trans-portation of Radioactive Materials, Oct. 14-18, 1968, pp. 373-398.
(4)
J.K. Vennard,*Elementari Fluid Mechanics, Wiley and Sons, New York,1962, pp. 256-259.
Drop Test: (continued)
The intent of the design for the GE M'odel 600 is, during accident conditions, to susta'in damage to the packaging not greater than the damage sustained by the GE Model 100 during its accident condition tests (ref.: Section 5.1.3(c) of the Model 100 Application).
It is expected that damage not exceeding that suffered by the GE Model 100 will result if the GE Model 600 is subjected to the 30 foot drop test.
Puncture Test:
The intent of the design for the GE Model 600 is to sustain less or equal damage to the packaging during accident conditions than the deformation suffered by the GE Model 100.
It is expected that deformation
3.0 Package Evaluation (continued)
(c)
Hypothetical Accident Conditions (continued)
PunctureTest(cont.)
not greater than that sustained by the GE Model 100 will be received by the GE Model 600 in the event that the package is sub-jected to the puncture test.
Thermal Test:
Because of the various combinations of lead liners and thermal loads available with this cuntainer, the THTD fire transient was not run.
However, reference is made to the shielded container Models 100, 700, and 1500 which demonstrate the effectiveness of the double walled steel jacket as a fire as well as a, crash shiel.d.
General Electric will analy'ze by test or other assessment each container heat load with and without liners as the individual case dictates, to verify that the loaded container will withstand the 30 minute 1475'F fire without significant lead melting in the cask.
Water Immersion:
Since optimum moderation of product material is assumed in evaluations of criticality safety under accident condition the water immersion test was not necessary.
The cask 4
was pressure tested under water to 11 PSIG with no detectable leakage.
Summary and The accident tests or assessments described
==
Conclusions:==
above demonstrated that the package is adequate to retain the product contents and that there is no change in spacing.
Therefore, it is concluded that the General Electric Shielded Container - Model 600 is adequate as packaging for the contents specified in 2.0 of this application.
A 4.0 Procedural Controls Vallecitos Site Safety Standards have been established and implemented to assure that shipmento leaving the Vallecitos Nuclear Center (VNC) comply with the certificates issued for the various shipping container models utilized by the VNC in the normal conduct of its business.
Routine audits are performed to assure compliance with these licenses and permits.
Each cask is inspected, leaktested, and radiographed prior to first use to ascertain that there are no cracks, pinholes, uncontrolled voids or other defects which co0ld significantly reduce the effectiveness of the packaging.
After appropriate U.S. Nuclear Regulatory Commission approval, each package will be identified with a welded on steel plate in accordance with the labeling requirements of 10CFR71 and any other information as required by the Department of Transportation.
5.0 Fissile Class - Class III An analysis has indicated that not greater than the following amount of
~
fissile material may be shipped in any single container:
Grams U-235, Grams U-233, Grams Pu(Fissile) < l.0 500 300 300 The container shall be provided with a rigid metal liner so that the fissile material is confined to a cylindrical geometry with an inner diameter not exceeding 7.0 inches (nominal).
The Density Analog Method as described in SNM License Application for VNC, Docket 70-754, Section 5.4.4, dated April 18, 1966, was used for calcu-lations.
Although this method is normally.used to calculate the number of units for transport under Class II, it was used in this case to demon-strate that one shipment of two casks would be sub' critical.
No credit was taken in the calculations for Pu-240 or other poisons present. The cask cavity was filled with water, and'the fuel was homo-genized with the water in the volurne of the 7.0 inch liner.
This water filling was done to represent the accident case and to allow for wet loading of the casks. The full results of the calculations are shown in the table below:
~~
i
i.
- 5. 0 Fissile Class - Class III (continu;d)
Fissile Material Quantity Safe Number Pu-239 0.3 Kg 33 U-233 0.3 Kg 51 U-235 0.5 Kg 210 In all cases, regardless of fissile mixtures involved, the loadings will be assumed to be exclusively Pu. The contents will be shipped dry.
6.0 Modes of Transportation All modes with the exception of passenger aircraft are requested.
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,7
,y
d ATTACHMENT A CRITICALITY ANALYSIS e
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""'f NUCLEAR G E lv. ~, A L w E L E C. riniC en ENERGY GENERAL ELECTRIC COMPANY, VALLECITOS NUCLEAR CENTER. VALLECITOS ROAD DIVISION PLEASANTON, CALIFORNIA 94566, Phone (415) 862 2211 May 21, 1974 i
l i
Mr. C. E. MacDonald, Chief Transportation Branch Direc tora te of Licensing Regulation U.S. Atomic Energy Commission Washington, D.C. 20545 Ref:
- 1) License SNM-960 Docket 70-754
- 2) Amendment 71-31 to Shy 960, 4/18/69
- 3) Amendment 71-57 to shy 960,11/19/73
Dear Mi. MacDonald:
Cencral Electric has shipped large quantitics of byproduct noterials, and limited quantitics of fissile materials in the Model 600 Shipping Cask under Amendment 71-31 (4/18/69) to License ShH-960 without incident for several years. General Electric now petitions the Atomic Energy Coc=h.sion
~
for an amendment to SNM-960 which will increase the allowable loading of fissile ma terial.
Specifically, General Electric requests that the fissile leading of the Model 600 Shipping Cask be increased to 1200 grams provided:
(1) the fissile material is contained in standard waste liners constructed of five-inch schedt21e 40 pipe with a maximua inside icngth of 39-5/16 inches ;
(2) no more than four such liners are shipped at one time; (3) each liner contains no more than 300 grams fissile; and (4) the cask is provided wi th a.ponitioning la ttice such tha t the geometry shown in Figure 1 in maintained.
Tbi purpose of the positioning lattice is to improve the criticality clwarac-tersstics of the cask.
The waste liners are cicsed with either a bronze or brass screw top with a t-inch "O" ring gasket.
The gasket material may be either buna-N rubber or neoprene.
These waste liners are exactly the same as those used in the Model 1600 container.
The cas'k'will be shipped as Fiss'ile Class I7".
O
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....s.
~f Mr. C. E. MacDonald May 21, 19 74 O
t General Electric requests tha t a very timely review be made of this application in order to expedite shipment of irradiated AEC-owned fissile material to the AEC for disposal. We belie've tha t this request is reasonable in the light of your review of our submittal for the same fissile loading for the Model 1600 container (Re f. 3). The two containers are extremely similar with, from the viewpoint of criticality safety, the significant difference being the smaller cavity of the Model 600. Drawings of the two casks are enclosed as Attach-ment a to this submittal.
The analysis was perforned using the computer codes ANISN ), a discrete ordinates one-dimensional transport code, and KEN 0(2), a Monte Carlo code.
ANISN was used to analyze the normal shipping design geometry while KENO was used for the accident case.
From this analysis we conclude that the 600 series cask is critically safe for the shipment of four standard waste liners each containing 300 gm fissile (Pu239 235) for a total cask limit of 1200 grams fissile'.
This limit is or U safe with no restriction as to fissile type or composition.
It is not intended !
to ship U233 in this container in excess of the presently permitted limits.
The results of the criticality calculations are as follows:
- 1. ' Design Geomett;y k, gg 0.869
=
2.
Accident (Close Proximity - Flooded) = k 0.945 +.029
=
g The error associated with the accident case result was computed at 3:7 The infinite multiplication of this cask (i.e., an infinite array of such casks) i I
was calculated for the design geometry to be i
k, 0.974
=
lhe design geometry was analyzed using the same method as was used to analyze
{
the 1600 series cask. This geometry is shown on Figure 1.
The dimensions and ma terial regions for this figure are given in Tables 1 and 2 The problem wa s s ol ved in two pa r taL using the ANISN code and Los Alamos 16-group cross-section sets (3, 4, 5) with Pi sca ttering.
The firs t part consis ted of defin tng an infinite cylinder cell with a " white" boundary condition on the outer dianeter(
for the individual fuel liner.
This calculation resulted in a 16-group cell weighted macroscopic cross-section set to be used in subsequent calculations.
The outer dimension for the cell calculation was determined so as to preserve the atos densities of the materials within the cavity n'ormalized to the cavity volume.
Each cell consisted of the fissile material / moderator combination I
contained within each liner, the' waste liner itself, and the void surrounding each liner.
Each liner shall be limited to 300 gas of fissile material which will result in a fissile density of 0.023 gm/cm p
=
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Mr. C. E. NacDonald May 21, 1974 Table 1:
600 Series Cask Dimensions -
Radius R-em R
43.18 g
R 42.2275 2
R 26.9875 3
R 26.035 4
R S
15.00 R
03 6
4 Table 2: Material Regions
. l Region Ma terial Identifica tion 1
's taintes s outer Liner 2
Lead Shield
- i i
j 3
S tainless Inner Liner
~4 Void Void 5
Aluminum Waste Liner i
6 Pu & H O Fissile Wasta 2
4
~
This density was used to determine the atom density in the li~ner for the
!I infinite cylinder.
We volume fraction for.the water was determined by I
assuming a theoretical plutonium density of 11.46 gm/cm3 (the most likely
}
density of the fissile waste material to be shipped).
The atom densities l
for this problem are shown in Table 3.
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l
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Mr. C. E. MacDonald 4-May 21, 1974 1
o j
Table 3: Atom Densities N
i (a tom /b-em)
Ma terial Isotooes Atom Densities Fissile Plutonium-239 5.796 x 10' Hydrogen 6.673 x 10" 0xygen 3.337 x 10' Liner Aluminum 6.023 x 10" S tainleca S teel Tr %
6.01 x 10" Chromium' 1.72 x 10' Nickel 8.81 x 10"
-2 Shield Lead 3.31 x 10 With the homogenized' cell-weighted set,of cross-sections obtained for the liner in the first calculation, the entire cask was analyzad with ANISN.
The cell-weighted cross-sections were used 'to desc' ribe the cask cavity and the recainder of the cask was described discretely by catorial region.
It might be noted here that the resultant k,ff for this configuration is higher than the corresponding results obtained for the 1600 series cask -
600 Series Cask:
k,fg 0.869
=
1600 Series Cask:
k,gg 0.720
=
This is the result of the smaller cavity in the 600 cask causing the liners to be closer together than in the 1600 cask.
The accident case where the liners are assumed to be somehow arranged together in the center of the cavity and flooded, was first analyzed in a manner similar to the approach used for the design geometry. The results of this ca*.culation were:
k,ff 0.981
=
d Although this number is less than 1.00, the confidence of suberiticality of the system is reduced because of the necessity in making assumptions deviating from the~ true description of the system to facilitate analysis.
The assumption that could have the most effect on this number is the assumption of an infinite cylinder along the Z-axis.
In order to account for axial leakage, it is h
necessary to go to a two-or three-dimensional code.
For this problem, KENO, a Monte Carlo code, was selected. This code permits three-dimensional descriptios of systems for analysis.* The cask was described in KENO as shown inl Figure 2.
t i
4 e
Mr. C. E. NacDonald 5-May 21, 1974 The liner,s were described as box types in the setup with the " core" region then described as a 2 x 2. array of such boxes.
The cask itself was then input as reflector regions.
The'WilSN problem previously described was re-run to obtain adjoint fluxes with which neutron weighting factors for the reficctor regions were determined.
The number densities used for this problect are given in Table 3.
The 1culations were then performed using a Knight-modified (6)
Hansen and Roach (
set of cross-sections obtained from Oak Ridge National Laboratory.
The results of this calculation are given on Figure 3.
Af ter 4320 histories the average was 0.945 with the maximum of 0.974 and the minimum of 0.915 at the 997. confidence level.
Fig. 2:
600 Series, Cask - Three-Dimensional Representation 1
7.065 cm Liner: R
=
979,
g g
y R,
7.7203 cm
=
I 99.854 H
=
i, 4 Sis-yg-J.
- /s -
H, 106.998
=
H,p Cask R-in R-cm H-in H-cm I
c.vity 10.25 26.035 46 116.84 Ps i
i I
Liner 10.625 26.9875 46.75 118.745 l l
1
,6 Lead 16.625 42.2275 58.75 149.225 l
l 1
I Liner 17.0 43.18 59.875 152.083 I
l I
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=
a 4
s
I Mr. C. E. MacDonald
, May 21, 1974 N
Fissile Box Geometry r---
I i
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R l
g' i
l il i
21.84 cm l
R
=
i I
l 1
I R' = 18.6403 g
g g
l I
Il ll f
I g
t e
L____2_____a General Electric believes that the above analysis clearly demonstratas the safety of the proposed fissile load for the Model 600 Shipping Cask.
GD:'
Attachmen A c this submittal contains the revised pages to our basic eppli-cation for the Model' 600 (Appendix D to License SNM-960).
Thank you for your timely consideration of this application.
-Sincerely, fs G. E. Cunningham Adminiseracor - Licensing SW Att.
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I Re ference s 1.
Engle, W. W., "A Users Manual far ANISN", K-1693, Union Carbide, Oak Ridge, Tenn., March 30, 196/
2.
Whitesides, G. E., and Cross, N. F., " KENO - A Multigroup Monte Carlo Criticality Program", CTC-5, oak xidge Computing Technology Center (1969).
3.
Hansen, b. E., and Roach, W. H., "Six and Sixteen Group Cross-Sections for Fast and Intermediate critical Assemblies", LAMS-2543, Los Alamos Seidntic Lab, Los Alamos, New Mexico, November,1961.
4.
Connolly, L. D., et al, "Los Alamos Group Averaged Cross-Sections",
LAMS-2941 Los Alamos Scientific Lab., Los Alamos, New Mexico. July, 1963.
O 5.
Personal Communication, Smith, D. R. to Walker, E. E., Los Alamos Scientific Lab, Los Alamos, New Mexico, February 26, 1971.
6.
Whitasides, G. E., KENO Cross-Sectinn Library,
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