ML19290C042
| ML19290C042 | |
| Person / Time | |
|---|---|
| Site: | 07109073 |
| Issue date: | 08/29/1979 |
| From: | NUCLEAR PACKAGING, INC. |
| To: | |
| Shared Package | |
| ML19290C040 | List: |
| References | |
| 14102, NUDOCS 8001090072 | |
| Download: ML19290C042 (80) | |
Text
I Instructions for Incorporating Revision 10 Amendments to Model CH-142 Application, Dated August 29, 1979 Insert new page 0-3 Remove old page 0-3 Add new'page 0-3a Insert new page 0-4 Remove old page 0-4 Insert new page 0-6 Remove old page 0 6 Insert new page 1-3 Remove old page 1-3 Insert new page 1-5b Remove old page 1-5b Insert new page 1-7a Remove old page 1-7a Insert new page 1-13 Remove old page 1-13 Add new page 1-13a Insert new page 1-18b Remove old page 1-10b Insert new page 1-20c Remove old page 1-20c Insert new page 1-20e Remove old page 1-20e Add new page 1-20f Add new page 1-20g Add new page 1-20h Add new page 1-20i Add new page 1-20j Add new page 1-20k Add new page 1-201 Add new page 1-20m Add new page 1-22c-1 Add new page 1-22c-2 Add new page 1-22c-3 r
0 253 Add new page 1-22c-4 i
/
Add new page 1-22c-5 Add new page 1-22c-6 Add new page 1-22c-7 Insert new page 1-22d Remove old page 1-22d Insert new page 1-22d-l Remove old page 1-22d-l Insert new page 1-22d-2 Remove old page 1-22d-2 Insert new page 1-22d-3 Remove old page 1-22d-3 Insert new page 1-22d-4 Remove old page 1-22d-4 80C1090077 a
t Add new page 1-22g-1 Insert new page 1-22h Remove old page 1-22h Insert new page 1-22i Remove old page 1-22i Insert new page 1-24b Remove old page 1-24b Insert new page 1-24c Remove old page 1-24c Insert new page 1-24d Remove old page 1-24d Insert new page 1-24e Remove old page 1-24e Insert new page 1-25 Remove old page 1-25 Insert new page 1-25a Remove old page 1-25a Insert new page 1-25g Remove old page 1-25g Add new page 1-52a Add new page 1-52b Add new page 1-52c Insert new page 3-1 Remove old page 3-1 Insert new page 3-2 Remove old page 3-2 Insert new page 3-3 Remove old page 3-3 Add new page 3-4 Add new page 3-5 Add new page 3-6 Add new page 3-7 Add new page 3-9 Add new page 3-9 Add new page 3-10 Add new page 3-11 Insert new page 6-2 Remove old page 6-2 Insert new page 7-2 Remove old page 7-2 Insert new page 3-1 Remove old page 8-1 Insert new page 3-2 Remove old page 3-2 Insert new pace 8-3 Remove old page 8-3 Insert new page 8-4 Remove old page 3-4 Insert new page 8-5 Remove old page S-5 Insert new page 8-6 Remove old page 8-6 Inr.*rt new page 9-7 Remove old page 3-7 Insert new page 3-3 Remove old page 3-8 Insert neu page 8-9 Remove old page 3-9 1729 256 b
Insert new page 3-10 Remove old page 3-10 Add new page 8-11 Add new page 9-12 Add new page 8-13 Add new page 8-14 Insert new page i Remove old page i Insert new page ii Remove old page ii Insert new page lii Remove old page iii Insert new page iv Remove old page iv 1729 257 C
August 29, 1979 Revision 10 Table of Contents Pace 0.0 GENERAL INFORMATION 0-1 0.1 Introduction 0-1 0.2 Package Description 0-2 0.2.1 Packaging 0-2 0.2.1.1 General Description 0-2 0.2.1.2 Materials of Construction, Dimen-sions & Fabricating Methods 0-2 0.2.1.3 Containment Vessel G-4 0.2.1.4 Neutron Absorbers 0-4 0.2.1.5 Package Weight 0-5 0.2.1.6 Receptacles 0-5 0.2.1.7 Drain Fort 0-5 0.2.1.8 Tiedewns 0-5 0.2.1.9 Lifting Devices 0-6 0.2.1.10 Pressure Relief System 0-6 0.2.1.11 Heat Dissipation 0-6 0.2.1.12 Ccolants 0-6 0.2.1.13 Protrusions 0-6 0.2.1.14 Shielding 0-7 0.2.2 Operational Features 0-7 0.2.3 Ccntents of Packaging 0-7 1.0 STRUCTURAL EVALUATION 1-1 1.1 Structural Design 1-1 1.1.1 Discussion 1-1 1.1.2 Design Criteria 1-1 1.2 Weights and Center of Graviev l-2 1.3 Mechanical Properties of Materials 1-3 1.4 General Standards for all Packages 1-5 1.4.1 Chemical and Galvanic Peactions 1-5 1.4.2 Positive Closure 1-5 1729 258 1
August 29, 1979 Revision 10 Pace 1.4.3 Lifting Devices 1-5 1.4.4 Tiedowns 1-7a 1.5 Standards for *ype "B"
and Large Quantity Packaging 1-9 1.5.1 Load Resistance 1-9 1.5.2 External Pressure 1-10 1.6 Normal Conditions of Transport 1-12 1.6.1 Heat 1-12 1.6.2 Cold 1-13 1.6.3 Pressure 1-13a 1.6.4 vibration 1-14 1.6.5 Water Spray 1-14a 1.6.6 Free Drop 1-14a 1.6.7 Corner Drop 1-14a 1.6.8 Penetration 1-14a 1.6.9 Compression 1-15 1.6.10 Conclusion 1-15 1.7 Hypothetical Accident Conditions 1-16 1.7.1 Free Drop 1-17 1.7.1.1 Free Drop Impact Analysis -
End Impact 1-13 1.7.1.2 Free Drop Impact Analysis -
Corner Drop 1-19 1.7.1.3 Free Drop Impact Analysis -
Side Drop 1-24a 1.7.2 Puncture 1-25 1.7.3 Thermal 1-26 1.7.3.1 Summary of Pressures and Temperatures 1-26 1.7.3.2 Thermal Analysis 1-26 1.7.3.3 Pressure Evaluation 1-44 1.7.4 Water Immersion 1-53 1.7.5 Summary of Damage 1-53 1729 259
August 29, 1979 Revision 10 Pace 1.8 Special Form 1-53 1.9 Fuel Rods 1-53 1.10 Appendix l-54 2.0 THERMAL EVALUATION 2-1 3.0 CONTAI:01EN*
3-1 3.1 Containment Boundary 3-1 3.1.1 Containment vessel 3-1 3.1.2 Containment Penetration 3-9 3.1.3 Seals and Welds 3-9 3.1.4 Closure 3-9 3.2 Requirements for Normal Conditions of Transport 3-9 3.2.1 Release of Radioactive Material 3-9 3.2.2 Pressurization of Containment vessel 3-10 3.2.3 Coolant Contamination 3-10 3.2.4 Coolant Loss 3-10 3.3 Containment Requirements for the Hypothetical Accident Conditions 3-10 3.3.1 Fission Gas Products 3-10 3.3.2 Release of Contents 3-11 4.0 SHIELDING EVALCATICN 4-1 4.1 Discussion and Results 4-1
5.0 CRITICALITY EVALUATION
5-1 6.0 CPERATING PRCCEDURES 6-1 6.1 Procedures for Loading the Package 6-1 6.2 Procedures for Unicading the Package 6-2 1729 260 11:
August 29, 1979 Revision 10 Pace 7.O ACCEPTANCE TESTS AND MAINTENANCE PROGRAM 7-1 7.1 Acceptance Tests 7-1 7.2 Maintenance Program 7-1 8.0 QUALITY ASSURANCE 8-1 8.1 Organisation 8-1 8.2 Qu111ty Assurance Program 8-2 8.3 Cesign Control 8-3 8.4 Procurement Document Control 3-4 8.5 Instructions, Procedures and Drawings 8-4 8.6 Document Control 8-5 8.7 Control of Purchased Materials, Parts and Components 8-5 8.8 Identification and Control of Materials, Parts and Components 8-6 8.9 Control of Special Processes 6-6 8.10 Inspection 8-7 8.11 Test Control 8-7 8.12 Control of Measuring and Test Equipment 8-8 8.13 Handling, Shipping and Storage 8-8 8.14 Inspection, Test and Cperating Status 8-9 8.15 Non-Conforming Material, Parts or Components 8-9 8.16 Corrective Action 8-10 8.17 Quality Assurance Records 8-10 3.18 Audits 8-10 1729 261 17
August 29, 1979 Revision 10 filled with a shock-and-thermal-insulating material consisting of rigid polyurethane foam having a density of approximately twenty pounds per cubic foot.
The insulating material is poured into the cavity between the two shells and allowed to expand, completely filling the void.
Here it bonds to the shells creating a unitized construction for the packaging.
Mechanical properties of these materials are further described in Section 1.0 below.
The upper impact limiter is permanently affixed to the shield lid.
The lower impact limiter is an integral part of the shield body and is secured by means of eight 1 3/4 inch diameter ratchet binders (160,000 lbs. each).
A 24 inch diameter secondary lid is located in the center of the primary lid.
It is secured by means of eight 7/8 inch diameter bolts.
Both are f abricated from twc thicknesses of 3 inch thick steel plate.
The shield body censists of an external 1 1/8 inch thick cuter and 1/2 inch thick inner carben steel shell.
Three and one half inches of lead is located between the two for shielding.
Lif ting lugs and tiedowns are a structural part of the package.
The Mark I configuration differs from the above in that the 24 inch diameter secondary lid has been eliminated and replaced by 6 small 19 inch diameter lids.
The Mark II configuraticn differs frca the baseline design in that the package bottem has been made identical to the existing top.
The 24 inch secondary lid is not included in the bottom.
The package will be almost totally symmetrical about the hori:en-tal center line.
l l729 262 i
i i
0-3
August 29, 1979 Revision 10 The icwer impact limiter in the Mark II version is retained in place by ratchet binders which are attached to the upper impact limiters in 8 locations.
This causes the lower and upper impact limiters to be pulled toward each other and against the ends of j
the cask effectively securing them in place.
This arrangement I
allows the removal of both the lower and upper impact limiter for j
top or bottcm loading.
l l
The lower impact limiter in the Mark I version is fully welded i
to the cask outer shell and bottom.
This version can only be I
top loaded.
I i
i Material on the Mark II is A516 and uses 1" plate on the external skin vs. the 1 1/3" A-36 used on the baseline package.
In all other respects the Mark I and II are identical with the baseline configuration.
All Mk II cask welds are designed as full penetration and are made in accordance with weld procedures qualified to ASME,Section IX requirements.
This is verified on all longitudinal welds via full radiographic inspection.
All circumferential welds joining the inner and cuter cask shells are made utilitizing groove configu-rations, that assure full penetration.
Integrity of these welds i
is verified via magnetic particle inspection.
i 1729 263 0-3a
August 29, 1979 Revision 10 0.2.1.3 containment vessel The overpack itself is not intended to be the containment vessel.
It's prime function is to reduce the severity of the hypothetical accident conditions st. that u.. transportation cask can serve as the containment vessel.
As can be seen from the drawing in Appendix 1.10.1, the con-tainment vessel uses two 3" thick steel plates icined tocether to provide a 6" thick steel lid assembly.
A high temperature silicone gasket is employed in the primary and secondary lid interfaces.
The secondary lid also uses a redundant neoprene seal.
To assure seal integrity, an operation and maintenance program is prescribed together with a leakage test en the containment vessel prior to its first use.
(Refer to Section 6.0 and 7.0 belcw)
Waste products are contained within heavy gauge disposable steel liners.
In the case of dewatered icn exchange resins, the liner is pressure tight to 7 psig, thereby providing redundant contain-ment capability (22. 5 psig for Mark II cnly).
t 0.2.1.4 Neutron Absorbers There are no materials used as neutron abscrbers or T.cderators in the Mcdel CH-142 packaging.
1729 264 0-4
August 29, 1979 Revision 10 0.2.1.9 Lifting Devices Lifting devices are a structural part of the package.
From the general arrangement drawing it can be seen that three reinforced lugs are provided in the Mark II configuration and two reinforced lugs are provided in ene Marx I configuration.
0.2.1.10 Pressure Relief System There are no pressure relief valves.
0.2.1.11 Heat Dissipation There are no special devices used for the transfer or dissipation of heat.
The package maximum design capacity is 400 watts.
How-ever, this value may be exceeded if it can be demonstrated that actual equilibrium temperatures with the higher heat load are still within allowable limits.
0.2.1.12 Coolants There are ao coolants involved.
0.2.1.13 Protrusions There are no outer or inner protrusions except the external ratchet binders described in 0.2.1.2, above, and these are located within the envelope protected by the overpack.
1729 265 0-6
August 29, 1979 Revision 10 package is located at the approximate geometric center of gravity.
A reference point for locating the center of gravity is shown on Drawing Y-20-200D.
(See Appendix 1.10.1).
1.3 Mechanical Procerties of Materials The Mcdel OH-142 packaging uses an outer and inner shell fabri-cated of various thicknesses of low carbon hot rolled steel.
Material properties of the steel are as follows:
A-516 A-516 A-36 (Grade 60)
(Grade 70)
Per MIL-HDBK-V ASME ASME 55,000 psi 60,000 psi 70,000 psi F
=
tu 36,000 psi 32,000 psi 38,000 psi F
=
y 35,000 si 36,000 esi 42,000 osi F
=
su brg 90,000 osi l
90,000 osi 90,000 esi F
=
Rigid polyurethane foam fills the cavity between the steel shells of the overpack.
This material will have a density of approximarely 20 pcf and be of a self-extinguishing variety.
Figure 1 represents the stress-strain curve for the NuPac NPI.F6 foam used for this package.
The curve provides both minimum and maximum compressive properties and was derived from twenty samples of varying density and grain direction.
A 95% probability factor was applied to the standard deviation to establish the spread shown.
Foam Specificaticn NP:.F6 defines the detail foaming testing pro-cedure.
It specifies that foam samples will be taken during the actual foaming process and tested to verify that they are within i 10% of the mean curve at 10%, 30%, and 60% strains.
1729 266 1-3
August 29, 1979 Revision 10 2
f
= 6M/bt
= 6 PL/8bt b
where:
P = 96000 lbs.
L = 27.5 in.
b=
7.0 in.
t=
3.0 in.
(6) (96000) (27. 5) / (8) (7) (3)2 f
=
b f
31428 psi
=
b Margin of Safety is:
M.S.
= 36000 psi /31428 psi -1 M.S.
= +.15 This is conservative since the bending stress varies as the square of the plate thickness and only three inchs of the six inch total thickness were used.
Additional conservatism lies in the use of a uniform beam of minimum width.
Therefore, frcm the above it can be concluded that the lug, bolts and lid can react the lifting load.
Lif ting devices on the secondary lids are covered by sheet metal assemblies during transit.
This prevents inadvertent lifting of the cask assemblies by lugs not designed to withstand the load.
1729 267 1-5b
August 29, 1979 Revision 10 Using standard 40 shear out equation:
-d/2cos40]
P=F 2t E.M.
s
= (35,000 psi) (2) (3/S) 1.5 - 1/2 cos 40
= 29,320 lbs.
Margin of Safety:
M.S.
= 29,320 lbs./ ( 3000 lbs. ) (3 g's) /3 lugs -1 M.S.
= + Large The 3/8 thick lugs will easily buckle under end drop impact condition producing no significant effect on the overpack or package.
These lugs will be covered during transit.
1.4.4 Tiedowns Four tiedowns are provided.
The total load carried by each can be calculated as follcws:
10c Longitudinal
/'
i 4
nF sg N
- v. *'
P A
4
\\
98 1
/
Y 1729 268 1-71
August 29, 1979 Revision 10 1.6.2 Cold The materials of construction in this package are identical to those approved and used in numerous existing Type "3"
licensed packages.
All of the following utilize the same materials.
1.
DOT 6400 Super Tiger 2.
COT 6272 Poly Panther 3.
00T 6679 Half Super Tiger 4.
DOT 6553 Paducah Tiger 5.
COT 6744 Poly Tiger 6.
SN-1 Shipping Container N.U.S.
7.
Hittman - HN-300 Cask 3.
NRC No. 9069 Westinghouse MO-1 9.
Cntario Hydro Overpack - CDN C33 U33 Specifically, the Ontario Hydro package has operated continuously at these cold temperatures with no evidence of problems.
Therefore, on the basis of years of actual operating experience it is safe,to conclude that cold will not substantially reduce the ef fectiveness of the package.
To improve the cold weather capability of the CH-142, MK 2,
the A-36 material on the shell and lid will be replaced by A-516.
This material provides improved notch sensitivity and strength.
Foam material has been tested to -40 F and found to exhibit a small increase in compressive strength.
Samples failed in the 1729 269 l-12
~
August 29, 1979 Revision 10 U
same manner at -40 F as they did at recm temperature, indicating i
i that brittle fracture is not apparent at this temperature range.
Compressure Strength i
(esi)
% Strain Room Temo.
-400F Variation l
~
10 960 960 20 1050 1218 1.16 40 1300 1495 1.15 I
50 1600 1792 1.12 1
60 2260 2305 1.02 i
70 4310 4310 1.00 t
t 80 7640 8098 1.06 j
i i
l Frcm the impact analysis shown on Page 1-20j, it can be seen that I
impact accelerations decrease as compressive strength increases.
Therefore, cold conditiens will not produce detrimental loading and testing indicates that brittle fracture characteristics are not cresent
~
1.6.3 Pressure A differential pressure of.5 atmosphere vill be reacted by the lid and its associated ratchet binder tie dcwns.
From Section 1.5.2, it can be seen that the containment vessel can safely react a 25 psig external pressure.
It is, therefore, safe to conclude that a positive margin of safety will exist for shells when subjected to the.5 atmospheric pressure.
l Leads on the binders are calculated as folicws:
(A inb Ip psi) / (No. of Sinders)
P
=
3 (66)'
(14. 7/2) / ( 3 ) (4)
~
=
1729 270
= 3143 lbs/ binder l-13a
August 29. 1974 Revision 10 Since the primary lid will be reacting these loads in direct compression, the binder will not be loaded.
The secondary lid attachments must react these loads.
Therefore, bolt stress can be found from the following:
The secondary lid weight is:
W
= 1776 lbs.
3 Assume the lid must also react the projected area portion of the payload.
(10000 lbs) (24)'/ (66) 2 7
W
=
7 W
= 1322 lbs 7
Total equivalent wt.
r WT = 177 6 lbs + 1322 lbs WT = 3098 lbs Bolt Loads:
P=
( 3098 lbs) (114 g)/3 bolts t
i P=
44146 lbs/ bolt Using 115 ksi 7/3 inch diameter bolts (SAE Grade 5) their strength is 56,000 lbs.
(Area =.487 in per Mil Edbk V)
Margin of Safety:
1729 271 56000/ 44146
-1 M.S.
=
. 2_7 Secondary Lid Closure Scits M.S.
=
1-lab
August 29, 1979 Revision 10 Overnack Bottom Circle X2+Y2 2
=R C
Z
= -1 /2 c
Payload Cylinder X
+Y
=R P
Payload Bottom Circle X
+Y
=R g Z
=-l /2 g
Void Circle at Payload x2y23 2 1729 272 Z
= -l /2 g
Void Circle at Overpack Exterior 2 + Y~ = R,2 X
3
= -l !
c l-20c
August 29, 1979 Revision 10 Case No. 4 Maximum mechanical properties Central hole filled From the printouts it can be seen that columns 6 and 7 provide the total kinetic energy and absorbed strain energy for incremental crush depths.
When these valuas become identical the package is in equilibrium.
Column 8 provides this ratio for rapid eval-uation.
It is important to note that a large amount of additional energy absorbing capability remains in the package.
From Case No. 1 it can be seen that the package comes to rest after reaching a crush depth of 20.50 inches.
1729 273 1-20e
August 29, 1979 Revision 10 N
N
/
i
/ s\\fr i
\\
i Af\\\\
\\
\\
N s
'N l
N
/'
d
~
6: 40.08 Y
l The available fcam thickness can be calculated as follows:
l
-1
-c = tan (101/120) = 40,03 j
0 1729 274 1-20f
August 29, 1979 Revision 10 The actual available thickness varies slightly from the diagonal thickness by the fellowing amount:
I
-1 o
a = tan (101-76.25)/(2)/18
= 34.51 j
i A.
i 18 i
Y v
"' ' \\
s\\
+
h D
/
e 6
,~ h
[13'+12.375',
= 21.84 (Diagonal Thickness)
%\\
/
n W O ' Cl 9 ] ' P l.
l 4%
i F
I i
x = 21.84 cos (40.08-34.51) x = 21.74 in.
(Actual Available Thickness) i From Page 1-21a, the maximum strain energy is 32,360,741 in-lbs.
at 21.74 in deflection.
l
'4argin of Safety 32,360,741/25,200,000 -1 M.S.
=
- 1. S.
+ 23
=
1729 275
August 29, 1979 Revision 10 s
If the corner drop was to occur in the flattened area of the overpack, some additional deformation would take place.
The deformation would be directly related to the loss of available foam volume.
This volume can be calculated as follows:
l=
4a-
=
i 2b i
A i
r = 50.5 A = r'(a-sin 2a/2)
Uhere:
r = 50.5 in.
a = cos a = 48/50.5 = 18.1 A= 50.52(13.10 -sin ( 2 ) (13.1) /2 )
A= 52.4 in or Vol = 52.4 in /in of affected length The effected =cne will extend down the package by the following:
S = 20.5 in (Per Pg. 1-21a) m i
i
/
L A
L= 5/cos 40.08 L= 20.5/cos 40.08 40 OYk 4= 26.7 in.
1729 276 1-20h
August 24, 1979 Revision 10 Lost Volume is:
3 V=
(2 6. 7 in) (52. 4 in /in) 3 V= l102 in The total volume of fcan used during the cenpression is given :n Page 1-21a, Colunn 3.
V = 22684 in 7
or Less = 1402/22684 =.06 Loss = 6% due to flattened sides Therefore, the absence of foan in the local flat area of the over-pack results in enly a 6% reduction in the available fcan as calculated.
This small reduction is more than offset by the availability of the additional crush depth and asscciated additional foan as calculated above.
A pcsitive Margin cf Safety e:<ists.
An alternate nethod of evaluating the effect of the overpack flats is to conservatively assure that the cceplete everpack is reduced to a diameter equal to the width across the flats.
This apprcach analytically reduces the ancunt of fcan available for energy abscrb-tien and is therefore conservative.
96" dia. -
i i
i N,
l A
l
('
/
/
96' s (
x
/
s s
N_/
i i
s 's
/
T 6
~ x
/
1729 277 I w eia.
=
=
l-20i
August 29, 1979 Revision 10 1
In order to establish the ma::imum deformation for this case, the i
following CYDRCP (corner) case was run.
From the attached output the maximum deformation was found to be 20.16 in. for a 31 foot l
l drop.
The available foam is:
A 7
1 (96-76.25/2)2 + 19'
-cos tan ~96/120-tan-l ( 9 6-7 6. 25 ) / 2/18'l t
=
a t
= 20.22 in (Available) a t
= 20.16 in (Recuired) i Af ter impact a small amount of foam remains.
i This analysis conservatively neglects the presence of energy abscrbine fcan in the region from 96 inch diameter cut to 101 inch diameter.
Therefore, it can be again concluded that a positive l
1 Margin of Safety will exist for impact on to the flat portion of j
l I
the overpack.
Leads or acceleratiens experienced by the package are also plotted l
as a function of crush depth.
Accelerations were fcund to vary i
frc= a low of 72.2 g's to a high of 76.4 for the full range of conditions.
The small spread in these accelerations is explained by the following.
As the compressive strength is reduced the t
crush depth and centact area increase.
Therefore, the small stress times a large area produces numbers ecuivalent to the product of high stresses times smaller impact area.
Thus, normal variations in crush strengths do not significantly affect package leading.
1729 278 1-20j
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42382.
.7 23Y36000.
25662.
.001 100.00 0.00 0.00 0.00 0.00 2.50 131.1 133.
70224.
1.1 23968000.
53813.
.002 100.00 0.00 0.00 0.00 0.00 3.00 171.6 209.
103bS/.
l.6 24000000.
972b0.
.004 100.00 0.00 0.00 0.00 0.00 3.50 21b.3 306.
141672.
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IS8566.
.007 100.00 0.00 0.00 0.00 0.00 4.00 261.Y 425.
183806.
2.Y 24064000.
239Y35.
.010 100.00 0.00 0.00 0.00 0.00 4.b0 311.1 b60.
228129.
3.6 24096000.
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212777.
4.3 24128000.
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.019 100.00 0.00 0.00 0.00 0.00 5.50 416.S 932.
317427.
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6lb746.
.02b 100.00 0.00 0.00 0.00 0.00 6.00 472.4 1154.
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5.7 24192000.
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40/188.
6.4 24224000.
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7.1 24256000.
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7.S0 651.0 1995.
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t
August 29, 1979 Revisinn. In b
/
lb
- on YOo A
N
/
R 1
/
)
/
'\\
/kss
- sA
/
'\\/
\\
3 E3(1-cos45 )
Eb R3(l'cos45 )
$M
=0 3
2E'o(1+Cos45)g}
0 RL Cos 40,pb. 2Eb (1-Cos 45 ) +2Rb WgR Cos 40
=
b 9
1729 ggg
August 29, 1979 Revision 10 Where:
W=
(Payload + Lid + Overpack)
(10000 lbs. + 6900 lbs. + 3000 lbs.)
=
= 19900 lbs.
g = 76.4 g's (Max.)
R 38.125 in.
b R= 1,011,600 lbs.
L = 15.83 in.
i (1,0ll,660) (15. 8 3) Cos 400, (19900) (76. 4) (38.12 5) Cos 40
=
(2) ( 3 8.12 5) Pb(1-cos 45 +1+1+Cos 45 +.5) 44,402,789 = 12,267,878 + 266.785 P b P
= 120,454 lbs. per binder b
Therefore, the maximum binder load will be 120,454 lbs.
Capacity of the high strength 1 3/4 in. binder is 3.60,000 lbs.
The Margin of Safety is:
M.S.
= 160,000/120,454-1 M.S.
= +.33 Closure System 1729 gg3
August 29, 1979 Revision 10 As noted on Page 1-22b, the analysis was conservatively based, on the assumption that a uniform nominal 1000 psi crush force was reacted across the projected area.
This is a simplifying but con-servative assumption that does not take into consideration the strain hardening effect of the foam.
In reality, the compression force will vary from zero of the central edge of the contact surface to over 14,000 psi at the cask corner.
The 1000 psi compression stress acting over the axial projected area was found to produce a maximum force of 1,011,600 lbs.
(Ref.
Page 1-22c)
Any increase in this force aill reduce the actual binder load.
The actual maximum impact force is calculated by the CYDROP program and is presented on Page 1-22a, Column 4 (i.e.,
4,391,323 4,891,323 lbs. ces 40 lbs.) or an axial component of F
=
=
a 3,262,000 lbs.
Therefore, the actual value will be 3,262,000 lbs.
vs. the 1,011,600 lb. force conservatively assumed.
If this load was applied directly to the lid as shown in the analysis, it would show that the lid was in full ccmpression and, therefore, the binder would experience no lead.
It should be noted that F will not be applied at the centroid a
but will be applied at the true center of pressure.
Since the determination of the center of pressure involves integrating discrete i
pressures over unit area, the CYDRCP progran was mcdified to calcu-l late the true center of pressure.
The follcwing output represents the rerun of the acst critical corner drop case.
The center of pressure location was found to be:
C of P = 23.155 in.
1729 284 i
1-?2c '
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=
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CYDROP(CORNER)
HUCLEAR PACKAGING PR0PRIEIARY 09.40.56 79/06/27 PAGE 2
MODEL OH-142 DVERPACK - SOFT F0AN (CASE 1)
++ CRUSil PLANE ++
- + INPACT +***
- ++t*
ENERGY +++***
DISTRIBUTION OF STRAIN RATIOS BY CRUSH PERCENT OF CONTACT AREA DEPIH AREA VOLUME FORCE ACCEL.
KINEiiC STRAIN RATIO LE.70 GT.70 01.80 GT.90 GT.95 (IH)
(IN2)
(IN3)
(LBS)
(G)
(IN-LB)
(IN-LB)
(SE/KE)
LE.80 LE.90 LE.95 l.00 33.6 8.
7506.
.I 23872000.
1876.
.000 100.00 0.00 0.00 0.00 0.00
.927 1.553, CENTER OF PRESSURE (IH)
CRUSil PLANE ORIGIN SETBACK (IN)
=
=
1.50 61.5 32.
20773.
.3 23904000.
8946.
.000 100.00 0.00 0.00 0.00 0.00 2.329, CENTER OF PRESSURE (IN) =
l.392 CRUSH PLANE ORIGIN SETBACK (IN)
=
2.00 94.4 71.
41730.
.7 23936000.
24572.
.001 100.00 0.00 0.00 0.00 0.00 1.851 3.106, CENTER OF PRESSURE (IN)
CRUSH PLANE ORIGIN SETBACK (IH)
=
=
2.50 131.4 128.
69382.
1.1 23968000.
52350.
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CRUSH PLANE ORIGIN SEIBACK (IH)
=
=
3.00 172.0 203.
102550.
1.6 24000000.
95333.
.004 100.00 0.00 0.00 0.00 0.00 2.742 4.658, CENTER OF PRESSURE (IN)
CRUSH PLANE ORIGIN SETBACK (IN)
=
=
3.50 215.9 300.
140413.
2.2 24032000.
156073.
.006 100.00 0.00 0.00 0.00 0.00 3.177 5.435, CENIER OF PRESSURE (IN)
CRUSH PLANE ORIGIN SETBACK (IN)
=
=
3 4.00 262.7 420.
182146.
2.8 24064000.
236713.
.010 100.00 0.00 0.00 0.00 0.00 6.211, CENTER OF PRESSURE (IN) 3.604 CRUSil PLANE ORIGIN SETBACK (IN)
=
=
i 4.50 312.1 564.
227167.
3.5 24096000.
33*S41.
.014 100.00 0.00 0.00 0.00 0.00 6.988, CENTER OF PRESSURE (IN) 4.026 CRUSH PLANE ORIGIN SETBACK (IN)
=
=
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275161.
4.3 24128000.
464623.
.019 100.00 0.00 0.00 0.00 0.00 7.764, CENIER OF PRESSURE (IN) 4.445 CRUSH PLANE ORIGIN SEIBACK (IN)
=
=
5.50 418.3 928.
325761.
5.1 24160000.
614853.
.025 100.00 0.00 0.00 0.00 0.00 8.541, CENTER OF PRESSURE (IN) 4.870 CRUSH PLANE ORIGIN SEIBACK (IN)
=
=
6.00 474.6 1152.
377288.
5.9 24192000.
790615.
.033 100.00 0.00 0.00 0.00 0.00jff 9.317, CENTER OF PRESSURE (IN) 5.322
< ul CRUSH PLANE ORIGIN SETBACK (IN)
=
=
6.50 532.9 1403.
429702.
6.7 24224000.
992363.
.041 100.00 0.00 0.00 0.00 0.00$~$
5.799 b "
10.093, CEN1ER OF PRESSURE (IH)
CRUSH PLANE ORIGIN SETBACK (IN)
=
=
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482641.
7.5 24256000.
1220448.
.050 100.00 0.00 0.00 0.00 0.00 M h) 6.304 raI
d CRilSH PL ANE ORIGIN SETBACK (IN) 10.870, CENTER OF PRESSURE (IH)
=
=
I\\3 7.50 654.8 1997.
536434.
8.4 24288000.
1475217.
.061 100.00 0.00 0.00 0.00 0.00 )
C
'#3 11.646, CENTER OF PRESSURE (IN) 6.830 CRUSH PLANE ORIGIN SETBACK (IN)
=
=
b.00 718.2 2340.
590769.
9.2 24320000.
1757010.
.072 100.00 0.00 0.00 0.00 0.00 pg CX)
CRUSil PLANE URIGIN SElBACK (IN) 12.423, CENTER OF PRESSURE (IH) 7.372
=
=
Cys 8.50 783.1 2715.
647180.
10.1 24352000.
2066506.
.085 100.00 0.00 0.00 0.00 0.00 13.199, CENTER OF PRESSURE (IN) 7.911 CRUSH PLANE ORIGIN SEIBACK (IH)
=
=
9.00 849.4 3124.
704837.
11.0 24384000.
2404510.
.099 100.00 0.00 0.00 0.00 0.00 CRUSH PLANE ORIGIN SETBACK (IN) 13.975, CENTER OF PRESSURE (IN) 8.454
=
=
9.50 917.1 3565.
763659.
11.9 24416000.
2771634.
.114 100.00 0.00 0.00 0.00 0.00 CRilSH PLANE ORIGIN SEIBACK (IN) 14.752, CENTER OF PRESSURE (IN) 9.004
=
=
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826029.
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,4
,6
,7
,8
,9
,1
,2
,4
,6
,8
,0
,3
,6
,9
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,8
,4
,0
,7
,6
,6
,7 81 5 1 1 1 7 1 4 1 01 7 23 292 62229353 23 83 4 4 1 4 75 4 6 0 667 3 89 96 2
0 8
5 3
1 8
6 3
1 9
6 4
2 9
7 5
2 0
8 5
3 0
8 5
3 0
8 6
4 9
7 5
2 0
8 6
3 9
7 5
2 0
8 6
3 1
1 5
6 7
7 8
9 0
0 1
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4 4
5 6
7 7
8 9
0 1
1 2
3 2
2 2
2 2
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2 2
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3 3
3 1
1 1
1 1
1 7
8 1
4 2
6 2
2 0
8 5
9 1
0 9
9 7
0 6
6 8
0 7
=
1
8 = 2 = 5 = 9 = 7 = 0 = 7 = 8 = 6 = 0 = 9 = 4 = 4 = 7 = 9 = 0 = 2 = 8
1
= 6
6 = 9
0 3
3 3
2 0
8 1
0 3
6 4
8 8
3 1
9 5
5 5
9 8
4
) 9
)
6 ) 0 )
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5 5
6 6
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l I
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l R
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k R
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l i
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- 0. E 1 E 2. E 3. E 2. E 0. E 5. E 8. E 6. E 1
- 0. E 5.E 4 E 7.E 3
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- 5. f E
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N NA6 A7A9 A2A6 A1 A 6A2A9 A7A5A 3 A2 A1 A1 A1 A 1 A 1 A2 A 2A3 A3 A 4 A 5 L 2L 9 L 7 L 4 L 2 L 9L 7L 4 L 2 L 0L 8 L 6 L 4 L2 L 0 L 8L 6 L 4 L 2 L 0 L 8 L 6 I lP0P 1 P 1 P2P 3P4 P4 P 5P6 P7P 8P8 P9P 0P1 P2 P 2P 3 P4 P 5P 6 P 6P 7 l'
i l
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l 2
2 t 2 l 2 2
l 2
l l
i l
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1 l
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l l
l 2 l 2 1
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1 1
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i H
i H
i i
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i S
S S
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S U
U U
U U
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m R
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l C
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f 0
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, a))
_ sst
)XDd N
_ - rs PC' 80}A
August 29, 1979 r.evision 10 l
j
/
/
N' 4 C ',e j
//
3.57
/
II.4
/
//l
/
k
/
N j
f
/
N
/
/
/ 'y M,g,w/\\23 'Q '
R =/4, s el,a a s
'N wl.2.:.
a. N NTER 3= Wii3'5E
- =JEH PLPilE
/
CR16It I SET-kRCK If we were to assume that one of the eight binders was destroyed prior to the 31 foot (9.45 m) drop and that the lid was infinitely rigid so that a triangular load distribution took place, the following maximum binder load can be calculated.
I 1729 288 1-22c-7
August 29, 1979
. Revision 10 O
N'x' n
O
/
x N
\\
y s
s O
F~
5 9.125 h
/"6 " ' C C ll.l8
/
[,w ea 6
- -.. ~
/
/
/
+ ll.15 -"
- 33.l23
- 5E,03 I
i Di
=0 A
[
2(65.03)P+2(38.125) P/(65.08)+2(11.13)2P/65.08 i
+3,262,000(11.4) = (19,900 lbs ) (76. 4) ( 3 8.12 5) cos 40 173. 67 P + 3 7,18 6,8 00 = 4 4,40 2,78 9 P = 40,387 lbs/ binder 1729 289 l-22d
August 29, 1979 Revision 10 Therefore, if one of the eight binders are lost prior to the i
f 31 foot drop, impact loads will be a maximum of 40,387 lbs. per binder.
The Margin of Safety, based on the new binder load, will be:
M.S.
= 160,000 lbs/40,387 lbs -1 M.S.
= +2.95 The analysis presented on Page 1-22-b is conservative.
Actual maximum binder or associated bolt loads will be 40,387 lbs.
The 1 5/8 inch diameter binder retain pins are loaded in double shear and have the following capacity:
P = 2F A s
Where:
F
=.6Fu j
g i
(. 6) (105,000 psi)
=
i
= 63,000 psi A = 7 (1. 625) 274
= 2.07 in~
2 P=
(2) (63,000 psi) (2.07 in )
P= 261,316 lbs.
Margin of Safety M.S.
= 261,316/120,454-1 M.S.
= +1.17 1
Threfore, it can be concluded that the retaining P n has i
adequate capability of reacting the imposed loads.
Preloads are not an additive factor.
Preloaded joints are relieved by the applied load and only cause an increase in fastener f
load when the applied load exceeds the preload.
I 1729 290 l-22d-l
ugus e
Revision 10 Mark I Overpack Retainer Pins For the Mark I configuration, the overpack is secured to the cask by means of 8 1/2" diameter ball lock pins acting in double shear.
These pins pass through the ratchet binder lugs on the lid and the guide channels on the overpack.
The reaction force required to retain the overpack is directly proportional to Pb' Therefore P
=P (wt of overpack/w) o b
(120,454 lbs. ) (3,000 lbs. ) / (19,900 lbs. )
=
= 18,158 lbs. per pin The pins act in double shear through two thicknesses of 3/16 inch plate.
Shear out capacity can be calculated as follows:
P
=F 2t E.M. - d/2 COS 40 su su (35,000(2)(.375)(1.5 -.25 COS 40 )
=
34,350 lbs.
Shear Out
=
The Carr Lane Model Number CL-8-BLP ball-lock pin has a rated double shear capacity of 32,300 lbs.
Margin of Safety 3 2,9 00 lbs/18,15 8 lbs.
-1 M.S.
=
M.S.
=
. 31 Therefore, it can be concluded that the overpack retainer pins are T. ore than adequate for securing the everpack to the cask for the Mark I configuration.
1729 291 1-22d-2
August 29, 1979 Revision 10 Mark I Stud Substitution for Binders As an option to ratchet binders, the Mark I configuration may use 8-1 3/8 NC Grade 5 studs.
From above it was shown that the overpack for the Mark I is secured by means of retainer pins.
Therefore, the remaining load that must be carried by the bolts can be calculated as follows:
(16900)(76.4)(38.125) cos 40
= 12267878'266.7Pb(71.25/76.25)
P 95391 (76.25/71.25) b P
= 102,085 lbs/ bolt b
The capacity of a 1 3/8" NC Grade 5 stud is:
=F A
s Tu net Where:
F
= 105,000 psi (Min.)
Tu o
A
= 1.15 in' net (105,000)(1.15)
P
=
s P
= 120,750 lbs./ stud s
1-22d-3
August 29, 1979 Revision 10 Margin of Safety M.S.
= 120,750/ 102,085-1 M.S.
= +.18 Each stud is threaded into the top closure ring and high strength doubler.
Total thread engagement includes.75 inches for the closure ring and 1.75 inches into the doubler.
Recommended thread engagement is that equal to the thickness of a heat treated nut of the same tensile strength as the stud.
Minimum thickness for a 1 3/8 NC Heavy Hex Nut is 1.378 in. (Max.),
per Machinery Handbook.
Since the doubler is manufactured from a material of greater strength than the stud, the following conservative margin of safety c.n be calculated.
M.S.
= 1.75 in./1.398 in.
-1 M.S.
= +.27 1729 293 1-22d-4
August 29, 1979 Revision 10 Bearing stresses in the lugs can be calculated as follows:
brg "
!^
f Where:
P = 120,454 lbs. per lug A=
(1.50 in. thick) (1. 625 in. dia. )
f
= 120,454/ (1. 50) (1. 625) brg 49417 psi f
=
brg Allowable bearing stress per Page 1-3 f
= 90,000 psi brg Margin of Safeby M.S. = 90000/49417-1 M.S.
= +.92 1729 294 1-22c-1
August 29 1979 Revision 10 The lug capability in net area is:
P
=F u
tu Where:
F
110,000 psi (ASTM A-517) tu A
(3.75 - 1.625)(1.5)
= 3.19 (115,000 psi) (3.19 in~)
P
=
P
= 366,850 lbs. (Net Area) t 34 / \\
Lug to lid attachment MN b
3;[
/
i l
i i
I l
t l
Weld Shearing:
f l
P
= F A weld s
s Where:
F
= 35000 psi 3
A=
(1. 50 in) (3. 0 in) = 4.5 in~
(3 5000 psi) (4. 5 in')
P=
P= 157,500 lbs
!729 2hb 1-22h
August 29, 1979 Revision 10 Margin of Safety 157,500/120,454-1 M.S.
=
M.1.
= +.31 Since the binder load is reacted eccentrically an inward radial load will be produced.
This load is reacted by the lower cir-cumferential skirt of the overpack.
The load produces a com-pressive load on this skirt or ring.
No additional load will be imposed on the upper lug weld due to this eccentricity condition.
5 I
/8 "
't
<p -- -e-g' l
k/AJ6 Pf = RL
[
h R
= Pf/L lI E
(160000 lbs) (2.5 in)/ (13 in)
=
= 22222 lbs.
Where:
P = Maximum binder capacity If these were totaled at each binder the ring would be loaded as follows:
N 1
1729 296 J
\\NN i
1-22i
August 29, 1979 Revisinn in This conservatively assures 4.00 in, of foam across the maximum damage area.
Again it must be noted that this is extremely conservative since the overhanging portion of the overpack will absorb a large amount of energy thereby greatly reducing the total deformation.
Using the strain calculated above a very conservative acceleration can be calculated by assuming that the foam across the surface has all been stressed to a 67% strain.
From the sketch below it can be seen that only that material directly beneath the cask experiences these strains and as you progress outward they rapidly decrease.
Using this conservative assumption a strain of 67% will produce a maximum stress of 3600 psi (Ref. Figure 1).
The effective area is calculated as:
%//
I I
'7 o
l 0 % E7'E4/W y_
/
.1 s(
1, s
i
~ ~ ___ _ -
I i
4 7 2 : TEA /U(MA :c)
A = 2RL sin (s/2)
A=
(2) (50.5) (44) (sin 33.5 )
A = 2452 in' If we apply the maximum stess of 3600 psi across the whole area che acceleration would be:
1729 297 1-24b
August 29, 1979 Revision 10 2
a= (2452 in ) (36000)/64000 a = 137.9 g's If the same integration analysis was repeated, using the full length of the overpack, an average compressive stress of 1300 psi would produce an impact angle of 58 (1-Cos 29 ) 101/2
=
= 6.33 in This would produce a maximum strain of:
Strain = 6.33 in/12.375 in
= 51%
From Figure 1 the stress at a 51% strain is 1900 psi The effective area is:
A= 2 RL Sin 6/2 o
A=
(2) (50. 5) (80) Sin 29" A = 3917 in a=
(3917 in ) (1900 psi)/64000 lbs a = 116 g's i729 298 1-24c
"ugust 29, 1979 Retrision 10 s
Therefore, maximum acceleration and deformations are experienced when only the projected area of the cask is reacted by the overpack.
As noted earlier the secondary lid bolts do not react these shear loads due to the stepped lid design.
The following analysis is provided to demonstrate what margin it safely would exist if they were required to react these loads.
Load per bolt is:
P=
(1776 lbs Lid) (138 g's)/8 bolts P=
30636 lbs (shear)
Bolt capacity is:
R=
(115,000 psi) (.487 in ) (60%)
R = 33603 lbs Margin of Safety M.S.
= R/P -1 M.S. = 33603/ 30636 -1 M.S.
=
+.~10
==
Conclusion:==
Frcm the above it can be concluded that under the most conservative conditions the package will maintain more than 4 in of foam in the compressed area.
Impact loads will not produce detrimental ef fects on closure system since all loads are carried in direct 1729 299 1-4e
August 29, 1979 Revision 10,
compression across the deeply stepped joints.
Therefore, the side drop of 31 faet will not produce detrimental effects to the package.
Foam material has been tested to -40 C and found to exhibit a small increase in compressive strength ( < 16%).
Samples failed in the same manner at -40 C as they did at room temperature, indicating that brittle fracture is not apparent at this temperature range.
Frcm the SAR it can be seen that impact accelerations decrease as compressive strength increases.
Therefore, cold conditions will not produce detrimental loading and testing indicates that brittle fracture characteristics are not present.
1729 300 1-24e
August 29, 1979 Revision 10 1.7.2 Puncture A 40 inch drop onto a 6 inch diameter pin can occur in three separate regions of the package, i.e.,
overpack area, ends and side walls between overpack.
Since the overpack is backed by side wall or end type construction any impact in this region would be less severe.
Using ORNL-NSIC-68 for the side wall evaluation, the puncture energy can be calculated:
(W/S)
.71 t
=
Where:
St=
55000 psi (A-36) 70000 psi (A-516, Grade 70) Optional Matl.
S
=
2 64000 lb.
l N
=
.71 (64000/55000) t
=
1 a
1.11 (A-36) t
=
y
.94 (A-516, Grade 70) t
=
2 Margin of Safety:
1.125/1.11-1 M.S.
=
+.02 (A-361 (1 1/4" pr ?)
=
M.S.
= 1.00/.94-1 M.S.
= +.06 (A-516)
(la ptT) 1729 301 1-25
August 29, 1979 Revision 10 The ends of the package are constructed from two thicknesses of 3 inch steel plates for a total of 6 inches.
From the above, it l
was shown that the 1 1/8" (A-36) or 1" (A-516) steel sides backed by 3 " of lead were shown to be adequate.
For the sake of complete-ness, the following analysis is presented to substantiate the adequacy of the end plates.
In order to demonstrate adequacy of the cylindrical flask over-pack to withstand impact on a 6" diameter cylindrical pin, an "ANSYS" finite element analysis of the circular end plates has been performed.
The analysis approach considered both large deflection behavior of the circular plate and bi-linear characteristics of the mild steel material.
The mathematical model is shown in Figure 1.
This model consists of 52 nodes and 36 iso-parametric quadrilaterals (STIF42) repre-senting an axi-synemetric plate of three inches thickness and 74h" in diameter.
The plate model was-loaded by applying a series of prescribed displacements to node 2, corresponding to the contact perimeter of the 6" diameter puncture pin.
Ar the outer diameter of the plares, node 52, the forces induced by this pre-scribed displacemenr were reacted in an axial fashion.
No radial constraints were imposed upon the model except at the axis of symmetry.
I729 302 1-25a
August 29, 1979 2evision 10 Puncture impact on a binder assembly could result in loss of effectiveness of that binder.
Since the puncture test 'ollows the 30 foot impact, no significant loads are reacted by the binders.
The heavy 6 inch thick lid is stiffened by the over-pack and can be safely retained by the remaining seven (7) binders.
The 7000 pound lid is more than adequate to compress the silicone gasket and seat the lid on its steel stops.
Preload supplied by the binders only serves to secure the lid against the stop and retain it during i:apact.
A torque rate will be applied adjacent to the binders.
This will assure adequate preload.
The pre-scribed torque of 80 f t-lbs will produce a preload of 3000 lbs. in each binder.
This load will have no impact on the structural integrity of the package.
MARK I Lid Puncture Analysis The ends of the package are constructed of two thicknesses of three (3) inch steel plates for a total thickness of six (6) inches.
The top pair of end plates contain six access holes spaced equally about a circumferential line with a 22" radius.
The holes in the outer plate have a diameter of 20 1/2" whereas the holes in the corresponding inner plate have a diameter of 19".
The lower plate is reinforced between holes with a vertical web comprised of a 2" x 4" rectangular steel bar.
1729 303 1-25c
August 29, 1979 Revision 10 In order to evaluate the effect of the reduced insulation associated with a damaged or post-dropped overpack, the following analysis is presented:
Under corner impacts, Cases 1-4, the damage predictions are l'
as follows:
Ref.
Deformation Crush Volume 3
Case Page (in)
(in )
1 1-21a 20.5 22684 2
1-21c 19.5 20161 3
1-21e 19.5 20161 4
1-21g 18.5 17799 The nost severe of these damage predictions will be used to l
scale thermal effects.
i Original Foam 1.
Resistance (P. 1-36)
R
= 859.23 o
2.
Volume
=j (101'- 5 5') (18 ) + (101~-7 6. 2 5~ ) ( 2 2 )
~
Vg i
= 177249 in" Camaged Foam Volume 3
1772J9 - 22684 = 154565 in V
=
D Scale Resistance Change as:
D D
55 R
O
[g KA i729 304 tx: t
= v /a v
=
0 a
o 3
C O^
V 0"
O l-522
August 29, 1979 Revision 10 Thus:
VD R
--2 V
D KA g_ =
= 7_D o
_2 0
^
The damaged foam resistance is:
V (859.28x10-3) 154565 D
~
-=
D O
V 177249 0
= 749.31x10~
This is equivalent to:
!2
-wvwvww-I[I 1 I
I 2f 3
R 2
l7 D"
l W
W
+1 E
R l?s O
A
-wwwww-Mhere R
= an additional parallel resistance corresponding to the 3
damage caused by the 30 ft. drop.
DO R
=
A R -R O
D (749.31)(959.2S)
-3
'x 10
= a'9549
=
359.23-749.31 The heat that ficws across this damage resistor at t=
30 minutes I
is:
I 2 24)
T 9 -24
- R 2
A (1470.1-175.8)
= 221.1 Stu/hr.
=
o.8549 i
In 30 minutes, the total heat flow is conservatively estimated at:
02-24 = 110.5 Stu.
1729 305 1-975
August 29, 1979 Revision 10 Assuming all this heat goes into the lid:
C i=8, 18 C
=
LID g,
U
= 841 Btu / F (Ref. Page 1-42)
The change in lid temperature is thus:
aTD" 8 {
= 0.13 F Thus, the predicted percent increase in seal temperature (Node 24) is:
.13 x100 = 0.074%
%AT
=
24 175.8 Therefore, it can be concluded that the loss of insulation associated with damaged overpack will have no significant effect on seal or l
package temperature.
l 1729 306 1-52c
August 29, 1979 Revision 10 3.0 CCNTAINMENT This chapter identifies the package containment for the normal conditions of transport and the hypothetical accident conditions.
3.1 Containment Soundary 3.1.1 Containment vessel The containment vessel claimed for the Mcdel OH-142 package is the shielded transporta:icn cask as described in Section 0.2 and the general arrangement drawing in Appendix 1.10.1.
I The folicwing is applicable exclusively to the CH-142, MK II.
e 1
The CH-142, MK II will be used to carry a heavy gauge steel liner with a capacity of approximately 120 cubic feet of ion exchange l
resin.
The liner is a cylindrical container with.1340 inch thick walls and.1975 inch thick heads.
Each liner is pressure tested to 22.5 psig.
I I.
During transport the liner will contain a maximum of 120 cubic feet of resin originating from the purification system.
The size of the resin beads ranges from 0.4 - 0.6 nm in diameter.
Type "S" overpacks are designed to provide containment of the radio-active material for normal and transpcrtation accident conditiens.
The fesien of the CH-142 did not take credit for the liner inside the overpack.
Thus, there is no containment credit for the liner.
1729 307 3-1
August 29, 1979 Revision 10 The liner's structural integrity is such that it would most likely retain its contents in case of the hypothetical accident condition.
The SAR considers only the cask as containment, which is a conser-vative position.
It has been demonstrated by calculations that the sealing arrange-ment is unaffected by accident conditions.
For this reason, it is impossible that ion exchance resin, because of the size of the beads, could escape.
Gaseous radioactivity could be present in the package and is considered to be the critical item in contain-ment calculations.
Gaseous activity can arise from decay products of radionuclides carried in the ion exchange columns.
The only radionuclide that decays to a gaseous radioactive material in the ion exchange resin is iodine which decays to the noble gas xenon.
The iodine itself will be in the ionic form, therefore fixed on the ion exchange resin.
There are several isotopes of iodine.
However, the only isotopes of relevance to this evaluation are I-131 and I-133.
The other isotopes are either in very small quantities compared to I-131 and :-133 or they decay to products that are either stable or would not be in the vapour phase.
1729 308 3-2
August 29, 1979 Revision 10 From the list of radionuclides shown it can be seen that the l
l maximum anticipated I-131 content will be 1500 Ci.
It was conser-I vatively assumed that all other iodine is I-133.
For the calcu-lations, it is assumed that when the liner is shipped the above i
I maximum will be jointly present.
There would be no xenon present initially as a result of the dedeutoration process.
The maximum possible xenon activities have been calculated using the following assumptions-i I
i a)
All gaseous fission products originally present in the i
I spent IX resin are removed during the dedeuteration l
process which takes place en site prior to shipment.
i b)
Only the formation of xenon, from the decay of radioacti"a iodine, was considered.
c)
Only two isotopes, vi: I-131 and I-l'L,, were felt to ke i
of significance in this study.
The others were ignored either because they decay to stable isotopes of xenon (eg., I-130, I-132 and I-134) or are expected to be present in negligible amounts (eg., I-lJS).
d)
The xenon isotopes considered were Xe-131', Xe-133" and Xe-133.
e)
The processes of build-up and decay of each of these isotopes is such that their cencentration in the resin 1729 309 3-3
August 29, 1979 Revision 10 reaches a maximum value several days after dedeuteration.
I As is shcwn below, Xe-131" achieves its maximum value after 13.98 days, Xe-133" after 1.96 days and Xe-133 l
after 2.72 days.
Although physically impossible, the l
i i
maximum value of each isotope was assumed to exist j
concurrently in the resin and this was used in the safety assessment of the overpack.
I The maximum activities are calculated to be:
i i
2.7 Ci of Xe-131 (m) after 13.98 days
~
48.3 Ci of Xe-133 (m) after 1.96 days 177.0 Ci of Xe-133 after 2.72 days i
r i
The A, value, when considering several radionuclides, must reflect the mixture.
The following equation is used to determine an A value taking into acccunt Xe-131(m), Xe-133
[
2 and Xe-133(m).
i
- Ci l
A (mixture) =_
2 (Ci/A i) 2 Where:
Ci
- activity of each radionuclide (Ci/A )i - activity of each radionuclide divided
{
2 by its A V"1"*
f 2
i 1729 310 3-4
August 29, 1979 Revision 10 The maximum activities calculated above for the radionuclides of concern are:
2.7 Ci of Xe-131 (m) 48.3 Ci of Xe-133 (m) 177.0 Ci of Xe-133 The individual A values are 100 Ci for Xe-131 (m) and 1000 Ci 2
for Xe-133.
The A value for Xe-133 (m) is based on a comparison 2
with Xe-133.
The effective gamma energy for Xe-133 (m) is 0.023 MeV while the effective gamma energy for Xe-133 is 0.0296 MeV.
Thus, A for the mixture is:
2 2.7 + 29.3 - 177 A, = 2.7 48._3_ + 177
= 904 Ci e
100 1000 1000 The permissible leak rate can be determined from ANSI-N14.5
" Leakage Tests on Packages for Shipment of Radicactive Materials",
a)
For normal condition of transport:
.,o R,I = A, x 2.73 x 10 R = 904 x 2.73 x 10 ^O 3
R = 2.51 x 10-Ci/sec 3
The free volume within the cask cavity is:
V = 142 ft.
7 120 ft.3 (Resin Volume)
V
=
R 33% of resin volume is void V
=
7 40 ft.3
(. 3 3 ) (12 0)
=
=
1729 311 3-5
August 29, 1979 Revision 10 V
=V
-V
+V Free T
R '
V
= 62 ft.3 or 1.75 m 3 The specific activity of the medium is:
3 3
CN (Xe-131 m) 2.7 Ci/1.75 m 1.54 Ci/m
=
=
3 48.3 Ci/1.75 m' =
27.60 Ci/m C3(::e-133 m)
=
3 Cg(Xe-133
) = 177.0 Ci/1.75 m
= 101.14 Ci/m TOTAL C
= 130.23 Ci/m 3
Permissible leak rate:
4 i
i b
- E !b N
3
~
= 2.51 x 10 Ci/sec/130.23 Ci/m
-3 3
= 1.92 x 10 cm /sec i
i Tesc sensitivity is:
S = L/2 = 10~
cm/sec (Normal) l t
b)
For accident condiciens:
i
_a R
=A x 1.65 x 10 - sec 3
2 R
= 904 x 1.65 x 10 '
~
g
-6 R,
= 1.47 x 10 Ci/sec a
Permissible leak rate is:
-6 3
L,
= 1.49 x 10 Ci/sec/130.23 Ci/m a
-3 3
L
= 1.14 x 10 m /sec 3
-2 3
}[2}^ L = 1.14 x 10 cm /sec 3 l 3-6
August 29, 1979 Revision 10 Test sensitivity is: 3 ~ S& L/2 = 5 x 10 cm /sec (Accident) The CH-142 Cask has been designed to permit a leak test at any i time necessary for the safe transport of radioactive :naterial. i 1729 313 3-7
August 29, 1979 Revision 10 Radionuclides Carried Exclusively in the OH-142 MK II 1 i I Based en analysis of resin samples, it is assumed that the l following maximum activities could be present inside the OH-142 MK II. i t Radionuclide Activity OCuri.e s ) l Arsenic-76 15 Barium-140 30 Ccbalt-60 300 I Ccbalt-53 30 Chromium-51 30 Copper-64 600 Cesium-134 and 137 3000 Iron-59 30 Iodine-131 1500 Icdine (not 131) 1650 Manganese-54 30 Manganese-56 130 l Nicbium-95 75 Sodium-24 15 i 9 Strontium-39 450 Strontium-90 15 i Xenon 1200 Zinc-65 15 I 2irconium-95 75 1729 314 3-3
August 29, 1979 Revision 10 3.1.2 Containment Penetration There are no penetrations into the containment vessel. 3.1.3 Seals and Welds A silicone seal is placed between the primary lid to body inter-face. It is described in Section 0.2.1.3 above. All joints are are welded with full penetration welds. 3.1.4 Closure The closure devices for the lid consist of eight 1 3/4 inch diameter ratchet binders as described in Section 0.2 above. 3.2 Requirements for Normal Conditions of Transport The following is an assessment of the package containment under normal conditions of transport as a result of the analysis per-formed in Chapters 1.0 and 2.0 above. In su==ary, the containment vessel was not affected by these tests. (Refer to Section 1.o above) 3.2.1 Release of Radioactive Material There was no release of radioactive naterial from the centainment vessel. 1729 315 3-9
August 29, 1979 Revision 10 3.2.2 Pressurization of Containment vessel Normal conditions of transport will have no affect on pressurizing the containment vessel. 3.2.3 Coolant Contamination This section is not applicable since there are no coolants involved. 3.2.4 Coolant Loss Not applicable. 3.3 Containment Recuirements for the Hypothetical Accident Conditions The following is an assessment of tk.e packaging centainment under the hypothetical accident conditions as a result of the analysis performed in Chapters 1.0 and 2.0 above. In summary, t le containment vessel was not affected by these tests. (Refer to Section 1.7). 3.3.1 Fission Gas Products Not applicable since there are no fissionable materials involved. 1729 316 1, 3-10
August 29, 1979 Revision 10 3.3.2 Release of Contents The analysis performed in Chapters 1.0 and 2.0 above show that there is no release of radioactive material under the hypothetical accident conditions. 1729 317 3-11
August 29, 1979 Revision 10 standard soap bubble test around secondary lid seals and drain port areas shall be used to detect leaks. Should significant gauge pressure drop be observed, inspact and replace primary lid seals as required. For the Mark II -3 configuration, the acceptance criteria will be 10 atm 3 ~4 em /s. Test sensitivity will be appro:cinately 10 att cm /s. 7) Inspect the package for proper labeling necessary to meet all applicable regulations. 8) Check to see that the primary lid lifting lugs are covered for transit. 9) Install an approved security seal. 10) Using suitable material handling equipment, transfer the package to the transport vehicle. 11) Secure package to the transport vehicle using the approp-riate tie down devices. 6.2 Procedures for Unloading the Package 1) Move the unopened package to the appropriate unloading area. Place it in a suitable unloading attitude. 2) Perform an external inspection of the unopened package. Record any significant or potentially significant observations. 3) Remove the security seal. 4) Repeat steps 1 and 2 in Section 6.1, above, for re-moving the overpack lid. 5) Remove the discosable steel liner. 6-2
August 29, 1979 Revision 10 necessary for the safe and easy operation f the packaging should be given regular inspection and repaired or replaced as necessary. A leak test will be performed on replaced seals or when damaged seals are suspected. The test will be performed in accordance with Section 6.1. As a minimum, sealing gaskets shall be replaced with new gaskets meeting the description in Drawing Y-20-20lD every four (4) months (sooner if visible wear is detected). Ratchet binders nust work frea and easy. Lubricate as required and replace if necessary. Any damaged or lost fasteners will be replaced with equivalent grade and strength. i729 319 7-2
August 29, 1979 Revision 10 3.0 QUALITY ASSURANCE 8.1 Organization Full responsibility for the Quality Assurance (QA) Program adherence to 10CFR71, Appendix E criteria rests with NuPac. Some Quality Program activities are delegated to other organizations, i.e., calibration of measuring equipment, however, the responsibility of the control of quality in the other organizations continues to rest with NuPac. The NuPac Quality Department has sufficient authority and organi-
- stional freedom to identify quality programs, implement corrective action and verify corrective action effectiveness.
Additionally, the Quality Department is independent from other organizations within NuPac and reports directly to the President of NuPac. The Quality Department is headed up by the Quality Manager who is responsible for the development, implementation and administration of the entire NuPac Quality Program. See Figure 2, " Organization Chart, Nuclear Packaging, Inc.". The Quality Manager and all other quality personnel and/cr organi- =ations within, or utilized by NuPac, are fully qualified for their quality responsibilities. Qualification records are maintained in the NuPac Quality Record File. 1729 320 3-1
August 29, 1979 Revision 10 8.2 Quality Assurance Procram NuPac has established and implemented the CA Program described herein. Training and/or evaluation of personnel qualifications are required for all QA functions in accordance with written pro-cedure and are approved by the Quality Manager. The QA Program assures that all quality requirements, engineering specifications, and specific provisions of any package design approval are met. Those characteristics critical to safety are emphasized. Nuclear Packaging, Inc. has developed a quality system to assure traceability and control the quality of all materials and processes utilized in the production of radioactive shielding, casks, containers, and other equipment pertaining to shipping packaging for irradiated fuel, high level waste and plutonium. The Quality Manual delineates requirements and procedures necessary to exercise control over design, documentation, procurement, material, fabrication, inspection, inventory, shipment and quality data retention. NuPac Quality System and implementing Quality Procedures are designed and administered to meet the 13 criteria of 10CFR71, Appendix E. Figure 1 is a matrix delineating the relationship between the 17 NuPac Guality Procedures and the 13 10CFR71, Appendix E criteria. 1729 321 3-2
August 29, 1979 Revision 10 8.3 Design Control NuPac Quality Procedures (QP's) have been developed, approved, and implemented to control design review in such a manner to assure that the following occur: (a) Design activity is planned, controlled and documented. (b) Regulatory and design requirements are correctly trans-lated into specifications, drawings and procedures. (c) Design documents contain quality requirements. (d) D'eviations from quality requirements are controlled. (e) Designs are reviewed to assure adequate design verifi-cation activities, i.e. ; stress, thermal, accident analysis, etc., are performed and that design charac-teristics can be controlled, inspected and tested, and that acceptance criteria are identified. (f) Interface control is established and adequate. (g) Design and specification changes are reviewed and approved by the same organization (s) as the original issue. (h) Design errors and deficiencies aro documented and corrective action to prevent recurrence is taken. (i) Design organization (s) and their responsibilities and authorities are delineated and controlled via written procedure. 1729 322 3-3
August 29, 1979 Revision 10 8.4 Procurement Document Centrol The NuPac QA Program assures that all purchased material, compo-nents, equipment and services adhere to design specifications. Supplier evaluation and selection, objective evidence of supplier quality, assignment of quality requirements to procurement docu-ments and related design documents, and source, in-process and receiving inspection are all administered and controlled in accor-dance with approved NuPac QA procedures. 3.5 Instructions, Procedures and Drawincs Quality planning is developed for all activities requiring quality participation in accordance with approved NuPac QA procedures by qualified Quality Engineers (QE's) and are approved by the Quality Manager. All design documentation, i.e., drawings, specifications, special processes, etc. affecting quality are reviewed by the Quality Department and referenced in quality planning as necessary to assure adherence to package design approvals and the applicable criteria of 10CF371, Appendix E. 1729 323 3-4
August 29, 1979 Revision 10 9.6 Document Control Policy and procedure for review, approval, release and change control of all controlled, quality related documents are delineated in approved NuPac CA Procedures. Provisions are provided in the QA Procedures for identification of individuals /organisations responsible for review, approval and issuance of documents. Document control responsibilities, facilities and distribution requirements are also addressed. Controlled documents include, but are not limited to: (a) Design specifications; (b) Design and manufacturing drawings; (c) Special process specification and procedures; (d) Procurement documents: (e) QA Procedures and manuals; (f) Cuality Planning for receiving, in-process and source inspection; (g) Source Surveillance and evaluation reports; (h) Test procedures; (i) Audit Reports. 3.7 Control of Purchased Materials, Parts and Components Procurement documents are reviewed for acceptability of suggested suppliers based on the NuPac approved supplier lists. In addition, and as required, supplier surveys are conducted to further assure supplier acceptability. 1729 324 3-5
August 29, 1979 Revision 10 Quality requirements and standard clauses are added to procurement documents to require suppliers to identify material, provide test reports, control special processes, certify equipment and personnel, etc. Quality planning is prepared and approved by the Quality Depart-ment for performance of source and receiving inspections in accor-dance with package design approval requirements and applicable 10CF271 criteria. All described activity is delineated in approved NuPac CA Procedures. 3.9 Identification and Centrol of Materials, Parts and Ccmpenents The identification and control of materials, parts, components and completed and in-process assemblies is administered by the Quality Department in accordance with approved NuPac QA Procedures. These procedures address quality status tags, maintenance of material identification and traceability, part identification, and related documentation. 3.9 Centrol of Special Processes NuPac approved QA Procedures delineate the policies and procedures established to control such special processes as: welding, heat treating, lead pouring, non-destructive examination, etc., in accor-dan:e with applicable codes, standards, specifications, 10CFR71 1729 325 3-6
August 29, 1979 Revision 10 criteria and other requirements. Special processes developed by NuPac suppliers and by NuPac are documented. These documents are controlled as described in Section 3.6. 3.10 Inspection All receiving, source and in-process inspection activities are performed in accordance with approved NuPac QA procedures. All inspection personnel and/or organi:ation qualifications are reviewed and accepted by the Quality Manager prior to inspection activity. The inspection activity is performed in strict accordance with approved quality planning prepared by qualified QE personnel (see also Section 8.5. Mandatory hold points, inspection equipment requirements, accept / reject criteria, personnel requirements, characteristics to inspect, variables / attributes recording instruer. ion, reference documentation and other requirements are included in the inspection planning. 3.11 Test Centrol A test control program as it applies to quality is addressed in approved NuPac QA Procedures and assures, via required planning, that all required testing, such as proof and acceptance tests are identified and performed in accordance with test procedures, design requirements and limitations. Prerequisites, accept / reject criteria, data recording criteria, instrumentaticn calibration, environmental 1729 326 3-7
August 29, 1979 Revision 10 conditions, documentation and evaluation requirements, etc. are delineated in the test procedures and changes to the test procedures are required to be reviewed / approved by the same organization (s) as the original issue. 3.12 Control of Measurinc and Test Ecuipment Administration of the calibration of measuring equipment and instrumentation is performed by the Quality Department in accor-dance with approved NuPac QA Procedures. The calibration system assures that all standard e.easuring instruments (SMI) are calibrated and properly adjusted at specified intervals to maintain accuracy within predetermined limits. Calibration is performed using equip-ment traceable to national standards. All calibrated equipment is statused, identified and controlled in such a manner as to meet the requirements of 10CFR71, Appendix E, Section 3.12. 3.13 Handling, Shippine and Storage NuPac approved QA Procedures require that handling, storage and shipping requirements adherence verification criteria be included in quality planning. These requirements are designed to prevent damage or deterioration of material and equipment. Information pertaining to shelf life, environment, packaging, temperature, cleaning, handling, presevation, etc., is included as required to meet design, NRC package approval and/or U. S. Cepartment of Transportation shipping requirements. 1729 327 3-3
August 29, 1979 Revision 10 Shipping documentation preparation, departure and arrival time and destination data recording is also addressed in the planning when applicable. 8.14 Inspection, Test and Operatina Status The use of inspection status tags, quality inspection stamps and other means to indicate inspection and test status at or for NuPac are delineated in approved NuPac CA Procedures. The clarity of the status indication, prevention of inspection and/cr test step by passing and prohibition of removal or modifi-cation of status indications, except with Cuality Department approval / Material Review disposition is assured via these procedures. See also Section 8.15. 3.15 Ncn-conformine Material, Parts or Ccmponents NuPac approved QA Procedures require that material, components and equipment that do not conform to requirements are centrolled to prevent their inadvertent use. Identification, segregation, discrepancy reporting, dispostion of non-conformances by authorized individuals and reinspection activities are performed and controlled in strict accordance with these precedures. 1729 328 3-9
August 29, 1979 Revision 10 8.16 Corrective Action Failures, malfunctions and deficiencies in material, components, equipment and services are identified and reported to the Quality Manager and the President. The cause of the condition and correc-tive action necessary to prevent recurrence is identified, imple-mented and then followed up to verify corrective action effectiveness. Detail requirements for this activity is delineated in approved NuPac QA Procedure. 3.17 Quality Assurance Records A quality records system is in effect at NuPac and is administered in accordance with approved NuPac QA Procedures. The purpose of the quality record system is to assure that documented evidence pertaining to quality related activities is maintained and avail-able for use by NuPac, its customers, and/or regulatory agencies, as applicable. Quality Records include, but are not limited to, inspection and test records, audit reports, quality personnel qualifications, design reviews, quality related procurement data, supplier evaluation reports, etc. Retention times and record pro-tection requirements are also delineated. 8.18 Audits Quality program audits are performed on a periodic, scheduled basis. Written planning sheets and check lists are utili:ed. 1729 329 3-10
August 29, 1979 Revision 10 Audit results and corrective action activity are reported to management, in writing, and are retained in the quality assurance record files. Current NuPac practice is to audit all quality functions on an annual basis. Details of the NuPac Audit System are delineated in approved NuPac QA Procedures. 1729 330 3-11
August 29, 1979 Revision 10 QUALITY RECUIREMENTS MATRIX 13 CFR 7S NuPac 10CFR50, Appendix 3 NuPac Quality Manual 10CFR71, Appendix I Crganization Quall'y Prcgram i Crganization Chart QP 1 - Quality Control Manual QP 14 - Cuality Assurance Training II. Quality Assurance Program Same As Above l III. Design Control QP 2 - Design Review QP 15 - Engineering Holds QP 17 - Oesign Control IV. Procurement Document Control QP 4 - Procurement Control CP 15 - Engineering Holds 7. Instructions, Procedures and QP 3 - Dccument Centrol Orawings QP 5 - Quality Planning QP 15 - Engineering Holds 7I. Dccument Centrol CP 3 - Occument Control CP 15 - Engineering Holds VII. Control of Purchased Material, QF 4 - Procurament Centrol Equipment and Services CP 12 - Material Control VIII. Identification and Control QP 3 - Cocument Centrol of Materials, Parts and OP 12 - Material Control Compenents 1729 331 3-12
August 29, 1979 2evision 10 IX. Centrol of Special Process QP 4 - Procurement Control OP 5 - Quality Planning CP 6 - Inspection and Verification QP 16 - Special Process Cualifications and Control X. Inspection GP 6 - Inspection and Verificaticn XI. Test Centrol CP 5 - Quality Planning QP 6 - Inspecticn and Verificatica GP 15 - Engineering Ecids XII. Control of Measuring and OP 11 - Calibration Centr:1 Test Equipment XIII. Handling, Storage and OP 12 - Material Centrol Shipping 'cIV. Inspection, Test and CP 6 - Inspection and Verification Cceratinc. Status XV. ';cncenferming Materials, QP 7 - Discrepancy Reporting and Parts, or Ccmponents Centrel XVI. Correction Action OP 3 - Cerrective Action XVII. Cuality Assurance Reccrds CP 1 - Cuality Centrol :!anual CP 9 - Quality Records n 4uw-- v. .-...a ~~ -....-..' = ' 4, ..o.o r c w ww 99 -
- b. s
. h. 1 e .4 ww M. a 1* M wG.9 .sv.... i 729 332 3-13
Au?'Ist 29, 1979 Revision 10
- . X Hl BI T A i
ORGANIZATION CHART l NUCLEAR PACKAGING INCORPORATED PRESIDENI. L. HANSEN i ADMINIST. E NG RG_. Q.A. MGR. PRODUCTICN J. SIMCHUK L.HANSEN J. CLIVADOTI C. SKCRUPA I PROCUREMENT' INSPECTICN RECEIVING SHI? PING J.SIMCHUK 1729 333 i f I t 3-14
s _ss u g % ' a's f i W l , 'l' ' ,s 2 e s g 16,., # 2 89 C ,4,, g e Y N e= a
- {
g, 5Y g h
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- s..
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- ?$
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i;-
- .:=.=. _-.--- ::
{ f ,h .d E df*~. N b d 3 5 I s e is.. [ - t .- / l i r.fa p-j* g, y
- = C n
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