ML20155K373
| ML20155K373 | |
| Person / Time | |
|---|---|
| Site: | 07109205 |
| Issue date: | 05/31/1986 |
| From: | TELEDYNE ENERGY SYSTEMS |
| To: | |
| Shared Package | |
| ML20155K363 | List: |
| References | |
| TES-3205, TES-3205-R02, TES-3205-R2, NUDOCS 8605280111 | |
| Download: ML20155K373 (20) | |
Text
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RADIATION, STRUCTUIUE, AND i
THERMAL EVALUATIQ4 (Application for Type B(U) Package Approval)
TES-3205
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September 1985 t
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Revision 1 February 1986 4
j Revision 2 May 1986
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110 W. Timonium Road Timonium, Maryland 21093 i
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1 TABLE OF CONTENTS t
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I Section Description a Pm 1
GENERAL INFORMATION 1-1 1.1 Introduction 1-1 l,
- 1. 2 Package Description 1 -1 2
STRUCTURAL EVALUATION 2-1 1
2.1 Structural Design 2-1 2.2 Weights and Center of Gravity 2-2 2.3 Mechanical Properties of Materials 2-2 l!
2.4 General Standards for All Packages 2-3 2.5 Standards for Type B and Large Quantity Packaging 2-12 2.6 Normal Conditions of Transport 2-13 j
2.7 Hypothetical Accident Conditions 2-17 2.8 Special Form 2-35 1
2.9 Fuel Rods 2-47 2.10 Appendix 2-48 3
THERMAL EVALUATION 3-1 J
3.1 Discussion 3-1 1
3.2 Summary of Thermal Properties of Materials 3-2 3.3 Technical Specifications of Components 3-8 3.4 Thermal Evaluation for Normal Conditions of Transport 3-9 3.5 Hypothitical Accident Thermal Evaluation 3-21 3.6 Appendix 3-31 4
CONTAINMENT 4-1 f
4.1 Containment Boundary 4-1 l
4.2 Requirements for Normal Conditions of Transport
,.4-2 j
4.3 Containment Requirements for the Hypothetical 1
Accident Conditions
~4-2 i
4.4 Appendix 4-3 l
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5 SHIELDING EVALUATION 5-1 j
5.1 Discussion and Results 5-1 5.2 Source Specification 5-la i
5.3 Model Specification 5-3 t
5.4 Shielding Evaluation 5-6 l R2
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- 5. 5 Appendices 5-9 5/86 6
CRITICALITY EVALUATION 6-1 1
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R2 5/86 DETAILS, SENTINEL SS 5/86 SHEET 2
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4 In Figure 1.2 the RM is shown installed and preloaded inside the inner container.
The generator is centered within the inner container by three
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long fins which can be adjusted to contact the inside of the container wall.
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The prescribed preload is obtained by compressing a silicone rubber pad 2/86 between the generator and the inner container's lid.
A welded, structural member occupies the space between the silicone rubber pad and the generator and transmits the load from the container lid to the RM.
The RM houses the fuel, which is doubly encapsulated in a fuel capsule assembly, and which in turn is contained within a shield assembly.
The shield assembly is preloaded within the generator and is largely surrounded by thermal insulation.
The shield assembly provides the required radiation attenuation following the sequential series of hypothetical accident conditions detailed in 10 CFR 71.73.
The package, including the generator and its components are discussed further in the following paragraphs.
Component weight breakdown is shown in Table 1.1.
1.2.1.1 Shipping Cask Outer Container.
The outer container is comprised of a welded body assembly and a welded lid assembly.
The two assemblies and their detailed components are shown in Figures 1.3a through 1.3e.
The body assembly is welded to e pair of I-beam members to provide access for handling by a forklift truck.
The lid assembly is designed with R1 structural ribs to provide lifting and tie-down points. When assembled, the structural ribs on the lid align with ribs welded to the body and tie into 2/86 the I-beam members.
The outer container surrounds the inner container with approximately 0.14 inch radial clearance.
Type 304L stainless steel was m
selected as the outer container material because of its combination of strength, elongation and weldability.
Figure 1.3 defines the dimensions, weld criteria, inspection criteria and finish requirements.
This container forms the external surface of the package.
1.2.1.2 Shipping Cask Inner Container.
The inner container (Figures 1.4a and 1.4b) is in the form of a cylindrical body with a plug-type lid.
The lid is equipped with three ring bolts, each with a rated capacity of 25,000 pounds, for use with a special sling for handling by overhead crane.
The rings can be used to lift either the lid or the complete container.
The inner container body is machined from a solid steel forged billet i R2 meeting ASWA-181 specifications.
The lid is machined from ASM-A-36 hot i 5/86 rolled steel plate.
The bumper pad and ring segments are fabricated from 6061-T651 aluminum alloy and are attached to the container lid to minimize the clearance between the inner and outer containers with the ring bolts installed in the lid.
The container lid is bolted to the body by twelve 3/4 inch square head bolts recessed in a cavity designed such that special tools are required for installation or removal to negate inadvertent opening of the container.
I Figures 1.4a and 1.4b defines the container dimensions and materials.
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As stated above, the post hypothetical accident configuration is assumed to be the RTG shield assembly containing the fuel assembly.
An (7
additional post accident configuration was analyzed - one which postulates R2
, damage to the RTG shield to the extent of separation of the shield plug from 5/86 the shield body.
The fuel capsule could, then, be in a position such that radiation streaming paths would exist through a gap between the plug and the shield body.
Other conditions are such that the capsule remains within the intact shipping container in a position at or very near to its normal shipping position.
Radiation analysis for this accident configuration conservatively assumed a bare capsule (devoid of the RTG depleted uranium shield) within the intact shipping cask inner container.
For this configuration, the maximum dose rate at one meter from the outer surface of the inner container is 260 mrem /hr.
Exterior surface dose rates of the inner container do not exceed 2.72 Rem /hr. Exterior dose rates are, then, well below the maximum allowable of one Rem /hr at one meter from the resultant configuration for the hypothetical accident sequence.
Details of the postulated accident and the radiation analysis are provided in Appendix 5.5.2.
5.2 Source Specification 5.2.1 Gama Source The radioactive source consists of up to 31,400 Ci of Sr-90 in the SrF 2
fuel form.
Sr-90 and its relatively short lived daughter product Y-90 are considered, for all practical purposes, pure beta emitters.
External radiation consists of Bremsstrahlung radiation (hereafter referred to herein as gamma radiation) which is produced by the betas emitted in the decay process.
Energy dependent gama source distributions are derived using the theory and computational technique of Evans (Ref. 5.1) for external Bremsstrahlung.
Source strength distributions derived for the SrTiO I"*1 3
form, using this method, were verified by comparison of conputed dose rates with measured values beyond varying thicknesses of lead shielding (Ref. 5.2).
o TES-3205 5-la
rate at one meter from the surface is less than one rem / hour.
Hence, the unit complies with the specific requirement of 71.51 (a) (2). The computed i /7 dose' rate of 710 mrem / hour at one meter (Table 5.1) should be a conservative Q
, estimate of the actual.
In other directions conputed dose rates are considerably lower due to the thicker uranium shielding.
5.5 Appendix R2 5.5.1 References (Chapter 5) 5.1 R. D. Evans, "The Atomic Nucleus," McGraw Hill, 1955.
5.2 " Shielding Kilocurie Amounts of Strontium-90,"
A.
M.
- Spamer, MND-P2529, The Martin Co., March 1961.
5.3 NSRDS-NBS 29, " Photon Cross Sections, Attenuation Coefficients, and Energy Absorption Coefficients from 10 kev to 100 GeV,"
J. H. Hubbel, U.S. Dept. of Consnerce, Nat. Bureau of Standards, August 1%9.
5.4 Herbert Goldstein, " Fundamental Aspects of Reactor Shielding,"
Addison Wesley, 1959.
5.5.2 R'IG Shield Damage Accident An accident situation is postulated in which some damage to the Du R2 g
RTG shield occurs such that the shield plug separate from the shield body.
5/86 t
This situation is in addition to the hypothetical accident sequence previously described and analyzed.
Herein, details of the accident scenario and radiation analysis for the resultant configuration are provided.
An accident environment such as the 9 meter drop of the hypothetical accident sequence could result in high stress levels in the Du stainless steel cladding on inpact.
Failure of the cladding could render the attachment mechanism for the shield plug to the shield body ineffective since the shield plug is bolted through collars which are part of the cladding.
This, combined with some crushing of the load bearing insulation between the shield plug and the bottom load bearing plate could produce some measureable separation between the shield plug and the shield body.
The fuel capsule assembly could, then, move to a position where there is a direct radiation streaming path through the gap.
Since both the RM within the shipping container and the Du shield within the RTG were preloaded, it is adjudged that, given this accident the damaged shield would not be released from the RM housing and the fuel capsule assembly could not be fully released from the shield body.
Nominally, the shield damage would be as described above and the final position of the fuel capsule assembly would be at or very near to its nominal shipping position.
Radiation analysis for this post accitent condition conservatively l
assumes the conplete absence of the Du shield (body and end plug).
TES-3205 5-9
Calculations assume a bare fuel capsule assembly at its nominal shipping R2 position within the shipping cask inner container (outer container not 1
present).
Calculations were performed assuming a loading of 31,400 Ci using 5/86 d
the techniques and data discussed and presented in this chapter.
For this calculation, the buildup factor for iron is appropriate.
Dose rates were computed at various locations on the exterior surface of the inner container and at one meter from the container surface. Maximum dose rates occurred on the axial centerline of the configuration above the container.
This is the expected result since the shield thickness is a minimum (top thickness of 4.0 inches as opposed to side of 5.7 inches a j
bottom of 6.5 inches).
The maximum dose rate at the cask surface is 2.72
)
}
Rem /hr.
The maximum dose rate at one meter from the surface is 260 mrem /hr
- a value well below the allowable of one Rem /hr for the configuration resultant from the hypothetical accident sequence.
4 O
i
)
TES-3205 5-10
.=
The RM assembly, sealed with dual Viton O-rings at the receptacles to head interfaces, the head to housing interface, and theJousing to fuel R2 p
access cover interface, is leak tested to less than 1 x 10 cc He/sec STP.
5/86 The sealed stainless steel " canning" surrounding 7_ the uranium alloy shield components is also leak tested to less than 1 x 10 cc He/sec STP.
- However, containment does not depend on the integrity of these seals.
8.1.4 Component Tests In addition to the tests previously described, the capsule and liner hardware are also subjected to radiography as specified in MIL-STD-271D Section 3.
This is a non-destructive test used to detect flaws, such as voids, cracks or inclusions in the finished fabricated parts.
The shield components are subjected to dimensional and weight checks t R1 verify that the minimum material density has been obtained.
2/86 Welds in the RM head and housing and in the shipping cask outer container are subjected to dye penetrant tests.
Test procedures and acceptance criteria are specified on the appropriate drawing.
8.1.4.1 Valves, Rupture Discs and Fluid Transport Devices.
Not applicable. The Sentinel SS package contains none of these units.
8.1.4.2 Gaskets. The shipping cask outer container is equipped with a gasket between its body and lid. The gasket is not required for containment R1 purposes.
However, to preclude the chance of water intrusion, if the gasket 2/86 fm is more than one year old at the time of a shipment, it will be replaced.
{}
8.1.4.3 Miscellaneous.
There are no components, other than the fuel capsule assembly, shield assembly and shipping cask containers previously discussed, whose failure would impair the effectiveness of the package.
8.1.5 Tests for Shielding Integrity Experience has shown that the dimensional inspections and weight measurements mentioned in 8.1.4 above are sufficient to insure that the shield components have been fabricated properly.
The shield integrity (design) is tested (verified) after fueling by Health Physics personnel at ORNL using appropriate radiation survey instruments.
Dose rates at the shipping cask outer container surface and at one meter from the surface are expected to be far less than the allowable limits of 200 mrem / hour and 10 mrem / hour, respectively, for non-exclusive use shipments. See Chapter 5.
8.1.6 Thermal Acceptance Tests No acceptance tests are planned to verify the thermal analysis shown in Chapter 3.
TES-3205 8-2 e
9 8.2 Maintenance Program The following paragraphs discuss the inspections and tests which are performed prior to each subsequent use of the Sentinel SS package.
g 8.2.1 Structural and Pressure Tests No structual or pressure tests are required to ensure the continued performance of the packaging.
8.2.2 Leak Tests
'lhe RTG and the shipping cask inner and outer containers will be
" swiped" to check for the presence of removable radioactive contamination.
This test verifies that containment has not been breached since the previous shipment and is the basis for demonstrating compliance with 10 CFR 71.87(i).
8.2.3 Subsystems Maintenance The maintenance items for the inner and outer containers are the gaskets and the lid fasteners.
By design the R'IG is maintenance-free. The gaskets will be inspected visually prior to each shipment for checks, cracks' R2 crazing, loss of resiliency and other evidence of deterioration or damage.
5/86 The outer container gasket will be replaced if it is more than one year old at the time of shipment in any event.
The inner container gasket will be O
replaced only if it shows evidence of deterioration. Likewise, the tasteners Q which attach the lids to the inner and outer containers will be inspected prior to each shipment for signs of corrosion, structural damage and other forms of deterioration. These will be replaced with new bolts if applicable.
8.2.4 Valves, Rupture Discs, and Gaskets on Containment Vessel The containment [ vessel] was previously defined as the strength member of the fuel capsule assembly.
It contains no valves, rupture discs, or gaskets.
R1 2/86 8.2.5 Shielding The maximum exposure rate at the surface of the package and at one meter from the surface of the package will be determined prior to each shipment.
These measurements can be used to determine whether or not the shielding is still adequate and to document that the requirements of 10 CFR 71.47 have been met.
8.2.6 Thermal No inspections or tests are necessary to check for degradation of thermal performance.
Heat dissipation and hence, component temperatures do not rely on coolant flow and associated circulation systems, special coatings which may degrade, or other extraordinary cooling means. As the radioisotope decays, component. temperatures decrease predictably.
Shipping procedures,
(
) included in an Operation and Maintenance Manual, will require that the v
cooling fins shown on Figure 1.1 be installed.
TES-3205 8-3
8.2.7 Inner / Outer container Maintenance lR2'86
)
I'n addition to the inspections and tests mentioned above, the Operation V and Maintenance Manual will require that:
The bolts which attach the shipping cask inner container's lid to a.
its body and the outer container's lid to its body be installed and torqued to specified values.
g 2/86 b.
The security seal be insta.1 led on the shipping cask outer container.
c.
A visual inspection be conducted to insure that the inner and outer containers' physical condition is unimpaired except for
- dents, marks, scratches and other insignificant surface imperfections which normally occur during transportation.
d.
The shipping container inner cask be visually inspected specifically to determine if there are any cracks in its lid or body. If cracks are found, that inner container will not be used.
If cracks are not found, the RM will be installed and the entire surface of the inner container will be scanned with a radiation survey instrument to check for streaming or points where the exposure rate is substantially greater than the exposure rate at points in the surrounding area.
If no anomalies are found, the container will be used.
If streaming or unusual variations in exposure rate are detected, the suspect area will be reinspected
' (
using a dye penetrant test.
If the dye penetrant test shows a crack, that container will not be used.
Otherwise, the inner container will be deemed satisfacory for use.
1 The shipping cask outer container includes a gasket. Although it e.
is not required for containment purposes, it will be replaced to preclude water instrusion if it is more than one year old at the g
time of a shipment.
5/86 The Operation and Maintenance Manual will also include procedures for loading the RM in the shipping cask inner container, loading the inner container in the outer container, marking and labelling the outer container, R1 and tieing down the package to the transport vehicle.
2/86 I
2 Ov TES-3205 8-4
_~.
.