W2-0-56NUREG-0361
PLUTONIUM AIR TRANSPORTABLE PACKAGE
MODEL PAT-1
Safety Analysis Report
Manuscript Completed: February 1978Date Published: June 1978
Office of Nuclear Material Safety and SafeguardsU. S. Nuclear Regulatoi y Commission
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4511(107pvy-M
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r U.S, KPARTMIBIT OF WMEIClhatimal Technical Informatim Sorvie
PB-282 356
Plutonium Air
Transportable Package
Model, PAT- INuclear Regulatory Commission, Washington, 0 C
(
Jun 78
00
£ %JJ~I
RE PODUrC.ED eftNATIONAL TECHNICALINFORMATION SERVICE
U. S. DEPAITWrPT OF COUMERC[
NRC om 335LT 1. REPORT NUMBER (Assipnedby DDC)1N-77 U.m .. uS. NUCLEAR REGULATORY COMMISSION
.BIBLIOGRAPHIC DATA SHEET NUREG-0361
4.TITLE AND SUBTITLE (Add vo.'um.No., itoppopriam) 2. tLeave blank)
Safety Analysis Report for the Plutonium AirTransportable Package, Model PAT-I N
7..AUTHORIS) 5. DATE REPORT COMPLE 1ED
MONTH YEAR
Febriary 19789. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS Ilnclude Z,p Code) DATE REPORT ISSUED
U.S. Nuclear Regulatory Commission MONTH YEAR
hime ,197R
Office of Nuclear Materials Safety and Safeguards 6. (Leawe blank)
Washington, D.C. 20555 8. kLeave blank)
12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include Zap Code)10. PROJECT/TASKAVORK UNIT NO
Same as 9 11. CONTRACT NO.
13. TYPE OF REPORT i PERIOD COVERED lInclusive dares)
Safety Analysis Report
f .\
15. SUPPLEMENTARY NOTES 14. (Leave blank)
Ai ABSTRACT P200 words or hess)
The document is a Safety ,•nallysis Report for the Plutonium Air TransportablePackage, Model PAT-l, which was developed by Sandia Laboratories under contractto the Nuclear Regulatory Commission (NRC). The document describes the engineeringtests and evaluations that the NRC staff used as basis to determine that thepackage design meets the requirements specified in the NRC "Qualification Criteriato Certify a Package foi- Air Transport of Plutonium," (NUREG-0360). By virtue ofits ability to meet the NRC Qualification Criteria, the package design is capableof safely withstanding severe aircraft accidents. The document also includesengineering drawings and specifications for the package.
17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS
ib., IDENTIFIERS/OPEN-ENDOE TERMS
18. AVAILABIL.ITY STATEMENT
Unlimited1q. SECURITY CLASS ,This report) 121. NO OF PAGE
I
20, SECURITY CLASS (Th'is page) 22 PRICE
'C FORM 335 17-77 .
ABSTRACT
The document is a Safety Analysis Report f-.# the Plutonium Air Transportable
Package, Model FAT-i, which was developed by Sandia Laboritories under contract
to the Nuclear Regulatory Commission (NRC). The document describes the
engineering tests and evaluations that the NRC staff used as a basi! to determine
that the package design meets the requiremnts specified in the NRC "Qualifi-
cation Criteria to Certify a Package for Air Transport of Plutonium" (NUREG-0360).
By virtue of its ability to meet the NRC Qualification Criteria, the package
design is capable of safely withstanting severe aircraft accidents. The document
ilso includes engineering drawings ana spec ficaticns for the package.
iii
CONTENTS
Page
ABSTRREC ........................................... ........................... iix
FIGURES ........................................................................ ix
TABLES ....................................................................... xiii
PREFACE .................................................................... xv
1.0 GE;IERAL INFORMATION ...................................................... 1-1
1.2 Package Description ................................................. 1-1
1.2.1 General ...................................................... 1-11.2.2 Packaging .................................................... 1-11.2.3 Allowable Contents ........................................... 1-8
REFERENCES ............................................................. 1-9
2.0 STRUCIURAL EVALIVATION .................................................... 2-1
2.1 Discussion .......................................................... 2-12.2 Weights and Center of Gravity ....................................... 2-12.3 Mechanical P-operties of Materials .................................. 2-12.4 General Standards for All Packages (10 CFR 071.31) .................. 2-2
2.4.1 Chemical and Galvanic Reactions .............................. 2-22.4.2 Positive Closure ............................................. 2-22.4.3 Lifting and Tiedown Devices .................................. 2-2
2.5 Standards for Type B and Large Quantity Packaging (10 CFR §71.32)... 2-2
2.5.1 Lead Resistance .............................................. 2-22.5.2 E. ternal Pressure ............................................. 2-4
2.6 Normal Conditions of Transport (10 CFR §71.35) ...................... 2-4
2.6.1 General ...................................................... 2-42.6.2 Heat ......................................................... 2-42.6.3 Cold ......................................................... 2-52.6.4 Pressure ..................................................... 2-52.6.5 Vibration ............... .................................... 2-52.6.6 Water Spray .................................................. 2-62.6.7 Free Drop ........ ........................................... 2-62.6.8 Corner Drop .................................................. 2-6i.6.9 Penetration .................................................. 2-62.6.10 Compression ................................................. 2-6
2.7 Hypothetical Accident Condition (IC CFR 571.42) ..................... 2-6
2.7.1 General ...................................................... 2-62.7.2 30-Foot Free Drop ............................................ 2-92.7.3 Puncture ...................................... ............. 2-92.7.4 Fire Test .......................... ....................... .2-92.7.5 Water Immersion .............................................. 2-152.7.6 Post-Test Suimmary of Package Damage ......................... 2-15
2.8 NRC Qualification Criteria Requirements ............................. 2-15
V
CONTENTS (tmwt.)
2.8.1 Discussion .................................................. 2-152.8.2 Sequential Tests .............................. ............. 2-172.8.3 Hydrostatic Test .................... 2-'92.8.4 Terminal Free-Fall Velocity ............................... 2-202.8.5 Other Requirements of the NRC Qualification C iAeria ......... 2-20
REFERENCES ................................................ ............. 2-48
APPENDIX 2-A Maximum Weight uf Contents ........... ..................... 2-Al
APPENDIX 2-B Test Facility Description ................................ 2-41
3.0 THERMAL EYALUATION ...................................................... 3-1
3.1 Discussion ............................................. ........... 3-13.2 Summary of Thermal Properties Materias ............................. 3-13.3 Technical Specifications ............................................ 3-I3.4 Therual Evaluation for Normal Conditions of Transport ............... 3-4
3.4.1 Thermal Models ............................................... 3-43.4.2 Maximum Temperatures .......... .............................. 3-Il3.4.3 Minimum Temperatures ........................................ 3-153.4.4 Maximum Internal Pressure ................................... 3-153.4.5 Maximum Thermat Stress ....................................... 3-153.4.6 Evaluation of Package Perforwnice Under Normal
Conditions of Transport .................................... 3-15
3.5 10 CFR 71 - Thermal Accident Evaluation ............................ 3-16
3.5.1 Thermal Models .............................................. 3-163.5.2 Package Conditions and Enviroinent .......................... 3-173.5.3 Package Tevperatures ......................................... 3-173.5.4 Maximum Interral Pressure.................................... 3-213.5.5 Maximum Thermal Stresses .................................... 3-213.5.6 Evaluation of Package Perforvwnce for 10 CFR 71
Accident Conditions ........................................ 3-21
3.6 NRC Qualification Criteria - Thermal Accident Evaluation ........... 3-21
3.6.1 Thermal Models ............. .............................. 3-213.6.2 Package Co'idit,-.. and Envirvmwent .......................... 3-233.6.3 Package Temperatures ......................................... 3-233.6.4 Maximum Internal Pressure .................................... 3-233.6.5 Maximum Thermal Stresses ..................................... 3-233.6.6 Evaluation of Package Perforwmce for NRC Qualification
ThemaIl ýond tions ......................................... 3-24
REFERENCES ................................................................ 3-25
APPENDIX 3-A Test to Establish Thermal resistance Values ofPAT-' Components .......................................... IA-
4.0 CONTAINMENT .............................................................. 4-1
4.1 ContairwrFent Boundary ................................................ 4-14.2 Norwrzl Conditions of Transport - 10.7R !71.35 ..................... 4-2
4.2.1 TB-I Containment Vessel Leak.:- htness ........................ 4-2
vi
CONTENTS (Cont.)
Page
4.2.2 PC-i Product Can Integrity ................................... 4-24.2.3 Pressurization of Containment Vessel Product Can ............. 4-2
4.3 Hypothetical Accident Conditions (10 CFR §71.36) .................... 4-2
4.3.1 TB-1 Containment Vessel Leaktightness ........................ 4-24.3.2 PC-1 Product Can Integrity ................................... 4-34.3.3 Pressurization of Containment Vessel and Product Can ......... 4-3
4.4 NRC Qualification Cr'teria .......................................... 4-3
4.4.1 Release of Radioactive Contents .............................. 4-34.4.2 Pressurization of Containment Vessel ......................... 4-7
REFERENCES ............................................................... 4-8
5.0 SHIELDING EVALUATION ..................................................... 5-i
5.1 Discussion and Results .............................................. 5-15.2 Calculational Method ................................................ 5-I5.3 Source Specification .................................... ............ 5-2
5.3.1 Gamr a Source ................................................. 5-25.3.2 Neutron Source .................... .......................... 5-3
5.' Model Specification ................................................. 5-5
5.4 1 Description of Radial and Axial Shielding Configuration ...... 5-55.4.2 Shield Reoionai Densities .................................... 5-5
RErERENCES ............................................................... 5-12
6.0 CRITICALITY EVALUATION ................................................... 6-1
6.1 Discussion and Results .............................................. 6-16.2 Calculational Method ................................................ 6-26.3 Contents ............................................................ 6-26.4 Model Specification ................................................. 6-3
b.4.J Normal Co.ditions (10 CFR :71.35 and 10 CFR 071.38) .......... 6-36.4.2 Accident Londitiois - NKZ Qualification Criteria ............. 6-76.4.3 Single Package (IJ CFR ý71.33) ............................... 6-10
6.5 Validation of Calculatic-' Method .................................. 6-10
REFERENCES .............................................................. 6-1i
7.0 OPERAI1NG PR2,EDURES ..................................................... 7-1
7.1 Loading the PAT-I Packac for Transport ............................. 7-1
7.1.1 Loading PC-i Pro:ct Can with Plutonium Oxide .............. 7-17.1 Loadinq PC-i PrzJcct Can into TB-i Containment Vessel ........ 7-17.1.3 Loading TB-i Conrtainment Vessel into AQ-1 Overpack ........... 7-3
7.? Procedures for Unloadinc the Package ................................ 7-3
vii
CONTENTS (Cont.)
Page
8.0 ACCEPTANCE TESTS AND MAINTENANCE PMOGRAH ................................. 8-1
8.1 Acceptance Tests to be Performed Prior to First Use of EachPackage.......................................................81
8.1.1 Fabrication Inspections ...................................... 8-18.1.2 Structural, Pressure, and Leak Rate Tests of the TB-i
Containment Vessel ....................... ................. 8-1
8.2 Tests to be Conducted Prior to Each Shipment ........................ 8-1
8.2.1 Visual Inspection ............................................ 8-18.2.2 PC-I Product Can ............................................. 8-28.2.3 TB-i (.ontainment Vessel ...................................... 8-2
8.3 Peric-dic Test and Maintenance ....................................... 8-2
8.3.1 TB-i Containment Vessel ...................................... 8-28.3.2 Replacement of Gaskets on Containment Vessel ................. 8-2
9.0 SPECIFICATIONS ANP DRAWINGS .............................................. 9-1
9.1 Quality Assurance ................................................... 9-19.2 Fabrication Requirements ............................................ 9-1
9.2.1 General Requirements ......................................... 9-19.2.2 Documents .................................................... 9-19.2.3 Standards of Manufacture ........... ......................... S-29.2.4 Quality Assurance Provisions ................................. 9-4
9.3 Final Acceptance Testing............... .......................... 9-4
9.3.1 Visual :lspection ............................................ 9-49.3.2 Structural, Pressure, and Leak Test Rates of the TB-l
Containment Vessel ......................................... 9-59.3.3 Function and Fit ............................................. 9-5
PAT-1028 TESTING, LEAK RATE, MASS SPECTROETER, PAT PACKAGE ............. 9-6
PAT-1029 MATERIAL SPECIFICATION, REDWOOD FOR PAT PACKAGE ................ 9-11
PAT-1030 WELDING, CORROSION RESISTANT STEEL, PAT PACKAGE ............... 9-13
ENGINEERING DRAWI.nGS ......................................... ........ . . 9-16
ACKNOWLEDGMENTS ............................................................... 9-68
viii
FIGURES
Figurýe Page
1.1 Exterior of PAT-i Package ......................................... 1-2
1.2 Plutonium Air-Transportable Package (PAT-i) ....................... 1-3
1.3 PAT-I Outer Drum Assembly ......................................... 1-4
1.4 Assembled TB-I Containment Vessel .................................. 1-6
1.5 Cutaway Drawing of TB-i Containment Vessel ........................ 1-6
1.6 Component Parts of TB-! Containment Vessel and PC-i Product Can .... 1-7
2.1 Model for Calculating Load Resistance ............................. 2-3
2.2 PAT-! Package Following Heat Test ................................. 2-7
2.3 Lateral Vibration Test of the PAT-i Package ....................... 2-7
2.4 Longitudinal Vibration Test of the PAT-I Package .................. 2-7
2.5 Water Spray Test .................................. ............... 2-7
2.6 PAT-I Package Following Four-Foot Drop Test ....................... 2-8
2.7 PAi-l Package Following Fenetration Test .......................... 2-8
2.8 Compression Test - Longitudinal Axis .............................. 2-8
2.9 Compression Test - Lateral Axis ................................... 2-8
2.10 PAT-I Package Following 30-Foot Drop Test ........................ 2-10
2.11 Set Up for Puncture Test ......................................... 2-10
2.12 Water Immersion lest ............................................. 2-11
2.13 Package Section Following 10 CFR Part 71 Appendix B Te-ts ......... 2- 2
2.14 Redwood Char Following 10 CFR Aopendix B Tests ................... 2-13
2.15 Discssembled TB-I Containment Vessel Following 10 CFR Part 71Appendix B Tests ................................................. 2-14
2.16 PAT-I Package Followvnu 442-FPS Top End Impact .................. 2-22
2.17 Radiograph of PAT-I Packaie Following 442-FPS Top End Impact ..... 2-22
2.18 PAT-i Dimensions Following 442-FPS Top End Impac. ................ 2.23
2.19 PAT-I Package Following 451-FPS Top Corner Impact ................ 2-24
2.20 Radiograph of PAT-I Parcage Followina 451-FPS Top Corner Impact... 2-2?
2.21 PAT-I Dimensions Following 451-FPS Top Corner Impact ............. 2-25
2.22 PAT-I Package Following 445-FPS Side Impact ...................... 2-26
ix
1
FIGURES (Cont'd)
Fl igre Page
2.23 Radiograph of PAT-1 Package Following 445-FPS Side Impact .......... 2-26
2.14 PAT-i Dimensions Following 445-FPS Side Impact .................... 2-27
2.?5 PAT-I Package Following 443-FP5 Bottom Corner lIact .............. 2-28
2.26 Radiograph of PAT-i Package Following 443-FPS Bottom ConerImpact ............................................................. 2-28
PAM-I Dimensions Following 43-FPS Bottom Corner Impa". ........... 2-29
2.2s PAT-I Package Following 466-FPS Bottom End Impact ................. 2-30
2.2? Radiograph of PAT-I Package Following 466-FPS Bottow. Ed Impact .... 2-30
2.30 PAT-I Dimensions Foliowing 466-FPS Bottom End Impact .............. 2-29
2.31 Crush Test (Package Impact-Tested on End) ........................ 2-31
2.32 Crush Test (Package Impact-Tested on Side) ........................ 2-31
2.33 Crush Test (Package Impact-Tested on Corner ....................... 2-31
2.3-1 Fincture Test (Package Impact-Tested on Top Corner) .... : .......... 2-32
2.35 Puncture Test (Package Impact-Tested on Side) ..................... 2-32
2.3f Puncture Test (Package Impact-Tested on Bottom Cor-ner ............. 2-32
2.37 'uncture Test (Package Impact-Tested on Bottom End ................ 2-32
2.2 Puncture Test (Package Impact-Tested on Top End) .................. 2-33
2.3,- Slash Test Damage ................................................. 2-34
2.C -lose Up of Slash Test Penetration ................................. 2-34
2.41 Arrangement of Package for Fire Test ............................... 2-35
2.z1 Fire Test in Progress ............................................. 2-36
2.t3 PAT-I Package Being Lifted Fro- Immersion Test Pool ................ 2-37
2.-'-' Appearance of PAT-I Packages Following Testing ..................... 2-37
2.1: Post-Tess Disassembly of a----A -I- Package .......................... 2-38
2.L4 Package Impact-TesteO on Tcp End ................................... 2-39
2.17 Redwood Inside Load-Spreader Tube of PP-l (Package Ioact-Testedon Top End) ....................................... ................ 2-40
2.• TE-I Containment Vessel Within Load Spreader lubp (Pa:kage Impact-
Tested on Top End) ............................... ................. 2-40
2.1S TT-l Containment Vessel (Packace Impact-Tested on Top End) ......... 2-41
2.3C Disassembled PAT-l (Packige Top-Corner Impact-Tested ............... 2-42
2.1 Disassembled PAT-I (Package Side Impact-Tested' .................... 2-42
2-:-- Disassembled PAT-l (Package Bottom Corner hi,:act-Test-) ........... 2-43
'C
FIGURES (%zRnt'd)
Figure page
2.53 Charred Redwood and TB-1 Vessel Inside Load Spreader (PackageBottom-Corner Impact-Tested) ................................. 2-43
2.54 Disassembled PAT-i (Package Bottom-Corvw Imrpct-Tested) ...... 2-43
2.55 TB-i Containment Vessel (Package Bottoff-Corner Impact-Tested) 2-43
2.56 Disassembled PAT-I (Package Bottom-End Npact-Tested) ......... 2-44
2.57 Hydrostatic Test Chamber with TB-I Contzinment Vessel ......... 2-45
2.58 PAT-1 Package Following 433-FPS Side I]ract at Approximatcly-40'F ....................................................... 2-46
2.59 Radiograph of PAT-i Package Following -T3-i ." Side Impact atApproximately -40'F ........................................... 2-46
2.60 PAT-i Package Following 424-FPS Side Iczact at Approximately200°F ......................................................... 2-47
2.61 Radiograph of PAT-I Package Following C4-FPS Side Impact at
Approximately 200OF ................. ......................... 2-47
2-B.1 Rocket Pulldown Facility ...................................... 2-B2
2-B.2 Rocket Pulldown Facility ...................................... 2-B3
2-B.3 Towline Attachment to a PAT-i Package ........................ 2-B4
2-6.4 Static Test Machine ............................................ 2-25
2-B.5 Set Up fo.- Puncture Test ..................................... 2-B6
2-2.6 Conihal Probe Used to Conduct PuncturE Test ................... 2-B6
2-8.7 Set Up for Slash Test ........................................ 2-B7
2-B.8 Fire Teb Facility ........................................... 2-B9
3.1 Charrinr Data from Reference 3 .............................. 3-3
3.1a Char Velocity for Various Materials .......................... 3-3
3.lb Char Depth vs. Time for Charring Mate'als .................... 3-3
3 -lc-- Char Depth vs. Density ........................................ 3-3
3.2 '.chematic of Nurerical Model Used to lsess Solar Heating ..... 3-5
3.3 Schematic of Internal Heat-Flow Path .......................... 3-6
3.4 Results of Thernal Tests (Solid Linei and Comparison witsNumerical Cdlculation (Symbols) ............................... 3-10
3.5 Daily Thermal CyclE of PAT-I Subjecte: to Solar Radiatior ..... 3-12
3.6 Temperature Distribution in Redwood Clerpack Just BeforeSundown (Most Severe Case) ................................... 3-23
3.7 Temperature of PAT Redwod-Maximum N'-Tial Environment ......... 3-14
3.8 10 CFR Part 71 Apperdix B Fire Test .......................... 3-18
FI(J•6-5 (Cont'd)
Figure Pag
3.9 Depth of Retdwnod Char Followi.-" 1i CFR Part 71 Appendix 2Fire Test ......................................................... 3-19
.3 Post-Flre View of Redwood ......................................... 3-2
.1 Cross S.ction of TB-I Seals ............... ..................... 44-i
4." Z Maximum lb-l Te mperature and Pressure Profile During KRCQualificatin Criteria Fire Test .................................. 4-
6.1 TB-I Geometry Used in KENO for Tw.-Density Cortents ConsideredNormal and Accident Conditions ................................... 6--1
6.2 Model of Repeating Undanmaged Ari Used in KE.iO................... 6--Z
6.3 KENO MOdel for the 280 Damaged Ca.e (X Axis O•ly Shown) .............-
6.4 Modcl of Repeating Damaged Array Lsed in KENO with GeneralizedG nometry .......................................................... 6-_:
7.1 Assembled PAT-i .................................... .............. 7--
xii
TABLES
Table Page
2.1 10 CFR Part 71 Appendix B Fire Test Data ..................... 2-15
2.2 Summary of the NRC Qualification Tests, PAT-] Package ........ 2-20
2-A.I Kinetic Enerqy of TB-i Containment Vessel .................... 2-A2
3.1 Summary of PAT-1 Package Temperatures at Normal Conditionsof Transport ................................................. 3-2
3.2 Thermal Properties of Materials ............................. 3-2
3.3 Experimentally Determined Thermal Resistances ................ 3-7
3.4 10 CFR Part 71 Appendix B Fire Test .......................... 3-17
3.5 Averaged Results-NRC Qualification Criteria Fire Tests ....... 3-23
3-A.1 Steady-State Temperatures Attained During Test to EstablishThermal Resistances .......................................... 3-Al
4.1 PAT-1 Package Post-Test Contvi,,,.:ent .......................... 4-1
4.2 Post-Test TB-1 Air Leakage Rates ............................. 4-3
5.1 Calculated Radiation Dose Rates for a PA7-I Package Loadedwith 2.0 Kg PuO2 Having High Americium Content ................ 5-2
5.2 Gamma Source Spectra for 1648 gms PuO2 with High AmericiumConten ...................................................... 5-3
5.3 Spontaneous Fission Neutron Source Strength by Isotope for1648 grns Recycle PuO 2. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4
5.4 Spontaneous Fission Neutron Spectrum for 164F gms PuO2 withHigh Americium Content ...................................... 5-4
5.5 ,-Neutron Source Strength by Isotope for 164F gms PuO with2High Americium Content ....................................... 5-5
5.6 ,-n Nutron Source Spectrum for 1648 gms PuOo with HighAre'rricium Content ........................... ................ 5-6
5.7 Total Neutron Source Summary for 1648 gms PuO, with HionAmericium Content ...................................... ..... 5-7
5.8 Undamaced Package Spherical Geometry Configuration ........... 5-8
5.9 Damaged Package Spherical Geometry Conficuration ............. 5-8
5.10 Isotopic Densities for 'uO2 with High Americium Content ...... 5-9
5.11 Regional Material Densities for Calculational Model .......... 5-11
6.1 KENO Calculations Establishing PAT-i Package as FissileClass I for 2.0 KG PuO2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1
6.2 Atom P nmber Densities ........................................ 6-5
6.3 KENO Keff with Generalized Geomrtry for 238. rrushed PAT-IPackages (Fioure E.4) ........................................ 6-7 (
.Xiii
TABLES (Cont'd)
Table
6.4 KENO Keff Results for the Criticals ......................... 6-10
i.1 Formulation of Rubber-Modified Epoxy Sealant ................. 7-1
8.1 Repiacement Schedule for Copper Sea1 and 0-Ring .............. 8-3
9.1 Limits of Imperfections in Acceptable Welb- .................. 9-15
xiv
PREFACE
The package described in this Safety Analysis Report was designed for air trans-
port of plutonium. The package was developed and tested by Sandia Laboratories
under contract to the Nuclear Regulatory Commission (NRC). The purpose of this
document is to desc-ibe the engineering tests and analyses that the NRC staff
used as a basis to dete'-mine that the package design meets the requirements
specified ir, the NRC "Qvalification Criteria to Certify a Packagc for AirIransoort of Plutonium" (NUREG-0360). By virtue of its ability to meet the NRC
Oialification Criteria, the package design is capatle of safely withstanding
severe aircraft accidents.
Prior to publication of this document, the NRC Qualification Criteria and the
Model PAT-i package design received an independent technical review and endorse-
P~ent by both the NRC Advisory Conrittee on Reactor Safeguards and the Aeronautics
and Space Engineering Board of the National Academy of Sciences.
xv
1.0 GENERAL INFORMATION
1.1 Introduction
The Plutonium Air Transportable Package, Model PAT-i, is designed for shipment of plutonium by
air. The package design was physically tested to demonstrate that it meets the criteria speci-
fied in NLIREG-0360, "Qualification Criteria to Certify a Package for Air Transport of Plutonium'
(Ref. 1 ) and the requirements of 10 CFR Part 71 (Ref. 2). The package has been assessed for
'ransport of up to 2.0 kg of PuO2 having a maximum decay heat of 25 watts. The package quali-
fit- as Fissile Class I.
1.2 Package Description
1.2.1 General
The PAT-i package has a gross weight of approximately 500-lbs. The exterior sn-pe of the
package is a right-circular cylinder, 24-1/2 inches diameter by 42-1/2 inches long
(Figure 1.1).
1.2.2 Packaging
The PAT-l packaging is composed of three basic parts: (1) a stainless steel containmnent vessel
(designated the TB-l), (2) a protective overpack assembly (designated the AQ-l), and (3) a
stainless steel product can (designated the PC-I) within the TB-I containment vessel. The TB-I
serves as the containment. vessel for the purpose of meeting the requirements of 10 CFR Part 71
and the NRC Qualific3tion ,Criteria. The PC-l serves as a separate inner container as requiired
by 10 CVR !71.42.
Figure 1.2 shows the principal elements of the package. A set of engineering drawings ar..4
specifications for the packaging is included in Section °.0.
The AQ-l iverpack (Figure 1.3) consists.of a 65-gallon drum that is fully lined with an inner
drum; both drums are made of 304 stainless steel. The inner drum has a cylindrical center
section, and is bonded (fixed) to the outside drum. The end sections of the inner drum are
separate and have rounded corners. The drum covers have integral shirt extension. wvhic•. upo•
assembY., fit betwtLerv the center and end sections of thE inrer drur. The 12-qa e C-ri; covkr
nla,,i, -,a h ski t,- e> cension that overlaps a recicr. of the o-tside dr,;". T•e'ty-three "Ic-in.
diameter cLIts pass .hrouah these five layers of sheet 77eta (i.e., the C-clamp extensic, the
outside drum, the inner liner central section, th.e cover extension skirt, and tne top or bottor
end sections of the inner liner) and fasten into nut plates secured inside the inner er;c
sections. The C-,ir: at the top end of the pac'aae can be -eriove-: t. rer:c.vinq-tý-e draw ----lt.
The C-rir,.: at thi bc:ton end of the package is welded srjt.
The outer redwood assembly is made of select, kiln-dried redwood to take advantage of redwood's
high specific energy and fire resistant characterist'cs. The asser5.ly is made of three elemer,-s:
1-1
P. , .. r -i
41.
Figure 1.1 Externor View of PAT-1 Package
1 -2
AOl- OVERPACKOUTSIDE .DRUM
OUTE RREDWOOD
ASSEMSLY*
INNERREDWOODASSEMBLY
LOAD SPREADERASSEMBL"
Nolt Wood Grain Orieniation IS-e Table 3-2)
Figure 1.2 Plutorwum Air TyansprIable Pa~age (PAll)
I -3
• C-RING WITHI SKIRT
"1B- I CONTAINMENT. -s,-VESSEL
INNNER REDWOC'"LUG
LOAD SPREADER DISC
INSULATION PAD,
OUTER REDWOODPLUG
Figu,e 1.3 PAT-1i ter Drum Assembly
(
(1) a removable plug with longitudinally oriented grain, (2) a cylindrical annulus with radial
grain orientation (fabricated as a series of wedqes arranged in a ring), and (3) a fixed plug
with longitudinlly oriented grain. The removable plug affords access tI the TB-I vessel. The
redwood annulus is bonded to the inner drum The large fixed plug is bonded to the wooden
annulus and to te inner drum; tre large fixed plug is also bonded to adjacent load spreader
members.
The bonding agent used to join the wooden elements together or to adjacert netal is a polyester
flexibilized epoxy adhesive, which has resilience over a wiie temperature range. When impact
forces cause deformation, this bond causes the wooden elements and their adjacent retal members
to act in unison-
The load spreader assembly is a 24-inch long alurinur- tube, 0.5-inch thick by i2-inc-; outside
diameter. One-inch thick by ll-incn diamete- aluminum discs are located at eacw. end of tVe -ube.
)ne disc is removable to access the TB-I vessel. The other disc is bonded to a-tiacen, members.
The load spreader distributes inertial loads fror the containment vessel ',ver a rela:ivel'y larme
area of shock-absorbing material (reJwood). For side impacts, the tube is the orincipal loaý
spreader. For end impacts, the discs are the principal load spreaders. rot severe corner
impacts, the extended region of the tube buckles or deforms inward, cons:rictinc outward movement
of the discs.
The inner heat conductor tube is made of :cpper. The tube conducts inte-nal decay heat fror the
TB-i containment vessel. The corduction path leads from the contents to the produc: can, tc, the
TB-l containment vessel, to the copper tube, into the fixed disc, and ir.:o the alu::iLJ::! loc-
spreader tub:. Heat is conducted from the Icad spreader tube through tY-, outer redwood to t-e
outer drum assembly, then to an exchange witn ar": t air. The inner conductor tube, the disc,
and the load spreader tube are mechanically cc:.nc.,.ed to insure an effec:ive heat conductive
path.
The TE-l containment vessel, Fi.ures 1.4, 1.5, and 1.6, is comprised of a body, a lid secured b.
tolts, a copper gasket, and an 0-ring. The vessel body and lid ire fabricated from 'ýHI3-SMC
nrecipitation hardened stainless steel. A HP075 temper enhances the ductility of the steel while
preserving its strength from low to high temc,eratures. The lid is hermetically sealed to the
body ty a copper gasket with knife-edge sealing beads on both the body a'd lid. The sealinc
surfaces are arranged to protect the knife-edg, sealing beads during handling. The lid has 4
pilot diameter region which fits closely intu the mating internal diameter of the d-iv. The TF--
containment vessel is also equipped with an 0-ring and groove which serve as a seconjav, sea'.
The twelve 1/2-inch diameter clcsure bolts, Figure ý,.ý, are forced from !-28E stainless steE.
.his r~terial exhibits excellent corrosion resistance and prcvides strer,.-th at high .empera'e
to ma:ntain the TB-l seal. The bolts art silver plated to prevent galliru with the stainle._
steel TB-I.
PQOD .,CT c 4N
TH I fq)DY
Figure 1., Cutaway Drawing of TB-1 Containment VesselFiqure 1.4 Assembled TB-1 Contninment Vessei
•J
Figure 1.8 Component Parts ol TB. 1 Containment Vessel andPC.I Product CRn
A spacer within the TB-i containownt vessel is fabricated from aluminur., honeycomb (see
Figure 1.6). The spacer cushiz:,s the flat end of the PC-] product can under impact loadinas
and serves as a thermal conductor for internal decay heat.
The PC-i product can (Figure 1.6) is fdbricated from 304 stainless steel and is closed by
crimping in a canning machine. After crimping, the can is sealed by welding or silver soldering.
The product can meets the separate inner container requirements specified in 10 CFR §71.42.
1.2.3 Allowable Contents
The authorized contents of the PAT-I package are limited as follows:
Materi.! Type and Form:
(i) Plutonium oxide and its daughter products, in any solid form.
(ii) Mixtures of natural or depleted uranium oxide and plutoniun, oxide and their
daughter products, in any solid form.
Miximum Quantity: Maximum 2.0 kg total material
ý'aximum Internal Decay Heat: 25 watts
Additional Permissible Contents:
(i) Maximum 16 grams of *ater
(ii) Two single-layer polvethlene bags weighing no more than 9 grams. The baes may
be taped closed or heat-sealed.
Fissile Cliss: Fissile Class I
1-8
REFERENCES
. "Qualification Criteria to Certify a Package for Air Transport of Plutonium," U.S. NuclearRegulatory Conmmission Report NUREG-G360, January 1978. Available for purchase fromNational Technical Information Service (NTIS), Springfield, VA 22161.
2. "Packaging of Radioactive Material for Transport and Transportation of RaoioactiveMaterial Under Certain Conditions," Title 10, Code of Federal Regulations, Part 71,Revised AuqAt 19, 1975. Available from the Gov"-rnent Printing Office.
I-9
2.0 STRUCTURAL EVV&UATION
2.1 Discussion
Physical testing of specimen packages was the primary method used to demonstrate that xne Model
PAT-1 package design meets the structural intcgrity requirements of 10 CFR Part 71 and the NRC
"Qual ification Criteria to Certify a Package for Air Transport of Plutonium,* (NUREG-0360).
Five specimen PAT-1 packages were subjected to the se.uential tests prescribed by the qual fica-
tion criteria. These tests ;re described in Section Z.8. In addition, a TB-i containment
vessel was subjected to the 600 psi deep submersion mest. This test is described in
Section 2.8.3.
Section 2.6 describes the response of tl. packA&e design when subjected to the normal
conditions of transport specified in Appe.... P of )( CFR Part 71. The respomse of the PAT-)
package design when subjected to the hypcthetic'dl accident conditions in Appendix B of 10 CFR
Part 71 is described in Section 2.7.
/
2.2 Weights and Center of Gravity
The approximate weights of the three basic components
O,-10verpack
TB-1 Containment Vessel
PC-1 Product Can (including alum. spacer)
Contents (maximum)
Totals
The centreid of the Model PAl-i package is located oT
20 inches from the btttom end.
of the PAT-I package are as follows:
206*
16.8
0.1
2.0
225 Kg.
Lbs.
454*
37
0.3
4.4
496 lbs.
its longitudinal centerline, approximately
2.3 Mec;anrical Properties ot Materials
The following ma:.-ial properties are used ir thne st-ictural analysis in thi5 section.
Stairless Steel 304
Yield Stress fy
-Modulus of Elasticity E
= 30,00 psi
= 29 x 106 psi
The weic-t of the AO-1 overpack can vary, due to naziral weight variation of kiln-dried redwxtd.
-1
2.4 Genera. Standards for All Packaqes (10 CFR §71.31)
2.4.1 Chemical and G31vanic Reactions
There is no potential for a significant chemical, galvanic or other re;_ction to occUr either
between the various PAI-I package components or between the packaqe aod its contents.
Metal-to-metal cot.tact on the exterior of the AQ-i overpack is limited to the 304 stainless
steel drum and cadmium plated steel bolts. The alu.ainuni load spreader tube is joined to the
bottom load spreader disc by cadmium-plated steel spring pins. The borto' disc of tfý lo*-d
spreader is joined to the cadmium-plated copper heat-conductirn tube by, self-tapping cadmium-
plated steel screws, These two joints are coated with i, moisture-resistant epoxy. The inside
of the copper heat conducting tube is lined with epoxy-resin fiterglass cloth to prevent abra-
siu! of the cadmium plating when loading and unloading the TB-] vessel. In the TB-I contain-
ment vessel, metal-to-metal contact is made between PH 13-8Mo martensitic stainless steel,
silver-plated stainless steel bolts and the copper gasket. Within the vessel, contact is made
between the aluminutr honeycomb spacer and the stainless steel PC-) product c3n. The plutonium
contents, which may be double wrapped and taped (or heat-sealed) within polyethylene bags, are
contained inside the PC-] product can.
These interfaces have no significdnt potential for corrosion.
2.4.2 Positive Closure
Positive closure of the cover L' che outer drum of the PAT-] is provided ty a skirted C-clamp
ring and twenty-three 3/8-inch toits. The bottom end of the PAl-i is similarly secured, except
hat the clamp ring ib welded closed. Positive closure of the TB-] containment vesse? is
provided by twelve 1/2-inch diameter bolts. Positive closure of thE PC-l product can is
provided by a crimped closure which is sealed by welding or silver soldering.
2.4.3 Lifting and Tiedown Devices
The PAT-I package design has no lifting or tiedown features which are a structural part of the
package. The package can be handled by slir.g. pallet, or other means.
2.5 Standards for Type B and _e Quantity Packaing (10 CFR s71.32)
2.5.1 Load Resistance
The PAT-] package meets the requi.,ement of 10 CFR E71.32(a), that when treated as a siPply
supported beam and subjected to a uniformly distributed load equal to five times the package
weight, the stresses in the package materials do not exceed yield.
For the purpose of evaluation, the outer drum of the pa-kage is assumed to support the entire
Vo.ad* (see Figure 2.1). The maximum stress given by simple bear theory is then:
Assumption is conservative, since inner assermlies will add additional bending resistanceto the package.
2-2
'COMPONENTS HERE IGNOREDFOR BENDING STIFFNESS
a- PACKAGE CROSS 3ECTION AT NIIDSPAN
TOTAL LOAD = 5W
b. ASSUMED LOADING AND BOUNDARY CODITIONS
Figure 2.1 Model for Calculating Lo3d Resistance
2-3
Si WL I0 ~ m -x ý_V
where W - 500 lbs.
L 42.5 inches
R = 11.27 iirhes
t = 0.059 inches
r = 565 psi
This stress is negligible in comparison to the yield stress of the outer drum material
(Section 2.3).
2.5.2 External Pressure
As discussed in Section 2.8.3, the PAT-I package design is capable of meetinq the 600 )si test
requirement specified in the qualification criteria- (NUREG-0360). Therefore, it will a'so meet
the 25 psig test requirement specified in 10 CFR Part 71.
2.6 Normal Conditions of Transport (10 (FR 071.35ý)
2.6.1 General
A PAT-I package was subjectt to the Normal Conditiions of Transport specified in Appendix A of
10 CFR Part 71. Following these tests, the 1B-I co)ntainment vessel exhibited an air leak-rate
of l10-1 atm cc/sec. A leak-rate of 10-7 atm cc/sec or less represents leak tightness (Ref. I).
The observed test results demonstrate tnat the P*T-, package meets the 10 CFR Part 71 contain-
ment requirements for Norma. Conditions of Transpor-t.
Since the geometric shape and dimensions of the PAT-I package and its TB-i containment vessel
were essentially unchanged after the tests, the foTlowing requirements in 10 CFR Part• 71 for
Normal Conditions of Transport are also satisfied: (1) no substantial reduction in the, effec-
ti.eness of the package, (2) the total effective voluae on which nuclear safety was assessed
was not reduced by more than five percent, (3) the ;package dimensions upor which nuclear safety
was assessed were not reduced by more than five per-cent, and (4) no aperture occurred in the
outer surface of the package large enough to pernit entry of a four-inch cube.
-2_6-72- f4eat
A PAT-I package was subjected to a 215'F environment for 48 hours (Figure 2.2). As indicated
by the thermal analysis, Table 3.1, this test environnent will generate package tempera:ures
that are greater than those which would occur if the package were to be exposed to direct
sunligh, at an ambient temperature of 130'F, in stil air, with 25 watts of internal decay
heat. Water was placed in the vessel to assure tha: the internal pressure during the test
would exceed the maximum normal operating pressure of 34.3 psi (Section 4.2.3).
To asse~s the adequacy of the TB-1 cortainment vessel for prolonged service under high
temperature environments, a TB-1 was progressively cycled between ambient temperature and
2-Z
200°F. 400°F, 600-F, 700'F and 800'F. Each successive peak temperatur,. was heid for approxi-
mately 24 hours. At the conclsion of the tests, helium leakage from the TB-l containment
vessel was not detectable.
The internal decay heat limitation of 25 watts assures that, under the Normal Co:,d.V-ions of (Transport specified in 10 CFR Part 71, thp mean temperature of the redwood will (E exceed
182°F and the peak temperature will not exceed 225"F. No significant redwoud degradatior would
result from prolonged exposure to this range of temperatures (Refs. 2 and 3).
The properties ef the polyester epoxy used as the bonding agent between redwood elements and
';ween the :-edwood and metal surfaces is -jch that its adhesiveness may be slightly enhanced
by exposure to a temperature of 225CF.
2.6.3 Cold
Following the heat test, the P;T-1 package was cold soaked at -40nF for 48 hours. No deorada-
tion of the package was observed.
2.6.4 Pressure
A TB-l containment vessel was ýeated to 255^F for eight hours, with sufficient water inside tle
vessel to assure that the internal pressure would be at least 51 psia (1.5 times the maximum
normal operating pressure of 3A psia). The unit was helium leak testd before being heated,
while hot, and .'fter cool down. The measured leak-rate was less than 10-10 atm cc/sec in each
case. A leal-r-te of 10-7 atir cc/sec or less represents leaktightness (Ref. I). This result
demonstrates that the package design is capable of withstanding 1.5 times the maximum normal
operating pressure. (10 CFR .71.53(b) requires each package to be pressure tested to tnis
value prior to first use.) The leak-rate measurements were conducted at a pressure less than
0.5 atmospheres. This demonst-ates that the package design meets the reduced pressure test
requirement of 10 CFR Part 71.
2.6.5 Vibration
The PAT-I package was subjectec to the following vibration environments (Figures 2.3 and 2.4),
representative of vibration co-ditions in transport (Ref. 4).
0.2 g 21Hz 30-15- Hz*
6 db/octave rol'off 150-2000 HZ
8 hours duratic-- loncitudinal axis
I hours duratie-- vertical axis.
The vibration tests had no effi-ct upor: the package.
*Randor. vibration levels are secified in g:/Hz units (which are analogous to power levels)
versus frequenry. The method 'o translat2 this level to an approximate equivalent sinusnidalvibration is to choose the badwidth of interest and integrate 3nd take the squart root. Thisprovides an rms g-level that corresponds to sinusoidal motion in that bandwidth.
2-5
2.6.6 Water Spray
The PAT-i package was subjected to the water spray test required by Appendix A of 10 CFR 71.
The test, shown in Figure 2.5, used a three-inch diameter hose equipped with a fog nozzle to
drench the package with more than 124 gallons of water per minute. The upper surfaces of the
package were continuously sprayed for more than 30 minutes, followed by spraying onto the side
surfaces. 10 CFR Part 71 does not require the effects of this test to be assessed individually.
However, the free drop test is to be conducted within 1-1/2 to 2-1/2 hours following the spray
test.
2.6.7 Free Drop
The PAT-] package was subected to a four-foot free drop onto an essentially unyielding surfacF
in a side iipact orientation. To provide assurance that the PAT-I has a wide design margin t*
withstand the rigors of normal handling and transport, the package was also dropped four feet
onto its top end, top-corner, bottom end, and bottom corner. Figure 2.6 shows that the effect
of these tests was inconsequential. As described in Section 2.6.1, the PAT-! package design
meets all the acceptance standards specified in 10 CFR Part 71 for Norma.l Conditions of Transport.
2.6.8 Corner Drop
This requirement does not apply to packages weighing more than 210 pounds and is therefore not
applicable to the 500-pound PAT-1 package.
2.6.9 Penetration
The PAT-] package was subjected to the penetration test specified in Appendix A of 10 CFP 71.
This test requires that a 13-pound, 1-1/4 inch diameter, steel cylinder having a hemispherical
end be dropped onto the package from a height of 40 inch:-s. The result was a slight dimple in
the outer dn.ri of the package as shown in Figure 2.7.
2.6.10 Cororession
The PAT-] package was subjected to ti.e compression test specified in Appendix A of 10 CFR 71.
This test requires that a compressive load equal to five times the package weight (5 X 500 lbs.
= 2500 lbs.) be applied uniformly against the top and bottom of the package for 24 hours. A
load of 3115i lbs., a conveniently available mass, was placed on top th'e PAT-l package, which
rested or. a :'assive steel plate, which in turn rested or, a massive corcrete block (Fioure 2.8).
The load wE- maintained for a period exceedino 24 hours. There was no observable effect to the
,cac~aee. >-e package was then placed on its side on a concrete pad, ard a concrete blob•
weighinc .5rD, lbs was leaned on the package, on top cf which a 3150-It concrete bloct, as( .
placed. Tr' total load, amounting to: rore than 5600 lbs. (Figure 2_.`o produced no eff•,t tc
the package.
2.7 HypothEtical Accident Condition - 10 CFR .71.42
2.7.1 General
The ability of the PAT-l package to meet the Hypothetical Accident Condition, specified in
Appendix 6 -- 10 CFR Part 71 :an be inferred from its ability to meet the more strinqent NRC
2-6
Fiqure 2.2 PAT-1 Package Following Heat Test Figure 2.3 Lateral Vibration Test of the PAT-1 Padc•ag
I. .1
Figure 2.4 Longitudinal Vibration Test of the PAT-1 Package Figure 2.5 Water Spray Test
N,
Fictire 2.6 PAT-I Packaai Follnwina Fn i,.Foot Drnp Toot
Figure 2.8 Compression Tesl - Lonqilgrlinal Axis
Flitire 2.7 PAT, I Package Following Penraieuton Test
7;,
Fr
c)pir" 2.9 Crmpression Teti- Lateral Axii
Qwalification Criteria to Certify a Package f;- Air Transport of Plutoniup. However, effective
J3"e 17, 1978, for shipments in excess of 20 curies per package, 10 CFR Part 71 requires certain
fw'ms of plutonium to be packaged within a separate inner container which will not release
material when subjected to the normal and accident test conditions specified in Appendices A and
B ef 10 CFR Part 71.
A WAT-1 package, including a PC-l product can loaded with 130: surrogate material, was subjected
to- the Hypothetical Accident Conditions in Appendix B of 10 CUR Part 71. The UOj surrogate
material was not contained withIn polyethylene bags (i.e., tht m',aterial was in direct contact
with the inner surfaces of the PC-I product can) Also, the PC-1 product can was sealed with
ep'xy rather than by welding or silver soldering.
Pc'.t-test examination of the PC-l ;roduct can indicated that the crimped closure and the erxv
ovf.bond ;iad remained intact. The can itself had several mintr dents (Fiaur, ?.15). Heliur leai
tei-ing indicated a leak rate beyo-d the provisions of Regulatory Guide 7.4, however. no uraniut
su-0gate material was detected to have been released us~ng a wipe L.st a.d fluorimeter assay
te:'-nique on the interior of the Th,-l vessel and the exterior of the PC-i product can. As
disussed in Section 4.3.2, the PC-I product can, within the PAT-I package, will meet the
pr:.,sions of 10 CFR §71.42 when closed by crimpinS and seale, by welding or silver solderinq.
Th- TB-I containment vessel was foird to be leaktic-Olt.
- 30-Foot Free DrOq
A .ZT-1 packaQe was subje-zted to a 30-foot drop onto an essentially unyielding surface in the
Sic., ,rienta~ior-. This o-ientatior was estimated to be tt;e •ost daraging to the PAT-I.
Fi-re Z. 0 illustrates the minor 6-enting which resulted to the_ outer drum of the packaoe.
2.7.3 Fuc'ure
The PAT-i package was theri subjecteW to a free drop of 40 inches onto a six-inch diameter, ten-
inc- long, steel bar mounted on an essentially unyielding horizntal surface (Figure 2.11). The
bar made contect at.mid-length on tne package because this is :he point at which minimum redwocd
is zrovided between the TE-i containment vessel and the outer drurr. Damage to the PAT-i
cor-'.sted of a minor imprint on the outer drum.
2.7-1 Fire Test.
ThE tAT-I package was sub.Jected to i fire test whiu exceeded the severity Qf the test reqc. red
by " CFR Part 71. the facility useiI for this test .= a ten-foot diameter pool of JP-4 aviation
fue" pool surrounded by a _.imney, 16 feet in diameter by ten feet high. Temperatures measured
wit+4n the fire and on the package surface exceeded the 14765F 4alue specified in the regulations
(Ta:;e 2.1). Following the fire, t.e package was permitted to cool naturally.
2-9
Jr
F iou re 2. 10 PT. 1 Pack FoIlow;nL 33 Foe:' D-: tc t-t:
Figure_211 Set Up for Puncture lest
-i n
I
Figiore '. 12 Wateir Immersion Test
(
,N3
N)
F~cpiro 2.1.1 Pachuoom S*,tltni Foltovwtnq 10 CFRk Pwrt 11
CaJ
Figure 2.14 Redwood Char Following 10 CFR ArIpendix B Tests
Figujre 2.15 Dkasgemble~d TB- 1 (Zotantienmrt V~msel Following 10 CF R
Part 71 Alpgi.nli x B t9Stt%
Table 2.1
10 CFR PART 71 APPENDIX B FIRE TEST
Average temperature on AQ-1 drum: Approx. IBOO°F
Flame temperatures in vicinity package: 2200-2300'F
Duration of fire: 52 minutes
Char depth in outer redwood: 3.8 inches
TB-I temperature: Appro). 200'F
2.7.5 Water Immersaon
The package was then submerged under three feet of water within the basin of the -ire test
facility for approximately 24 hours- (Figure 2.12).
2.7.6 Post-lest Summary of Packac. Damage
The sequential impact, puncture, fire, and irner-ion tests produced only minor da'oge to the
outer drum ard its closure rings. The outer redwood annulus had charred to a dep-n of about
three inches (Figure 2.13). while redwood inside the load spreader was essentiallh un-ffected
(FigurL 2.14). As discussed in Section 2J.1, the TB-i containnp.nt vessel was ieiktigi.t
f'llowirg the tests an,.: me. 10 CFR Part 71 containment reoq'irements. Since the tsts did not
significantly effect the geometry of the PAT-1 packaoe, its shielding effectiveness following
the Hypothetical Accident Conditiors in Appendix B of 10 CFR Part 7) would be essentially the
sane as under Normal Conditions of Transport.
2.8 NRC Qualification Criteria Re:-jirements/
2.8.1 Discussion
Ire NRC Qualification Crite'ia to Certify a Packag. for Air Transport of Plutoniuý (NULREC-0360)
specify that the structural integr;ty of the package be deronstrated by physical usting. Themajor requirement is for sequential testinc to the followira conditions:
1. Irrpact at a velocity cf not less than 422 ft/sec at a right angle onto a flat. essentially
unyieldirng surface, in the orientatior. (e.g., side, end, corner) expected to -esult in
traxin'ur. damage at the conclusion of tie test sequence.
2. A static co.m.pressive load of 72,000 pounds applied in the orientation expecte: to result
iF, maximur damage at the conclusion of test sequence. The force on the packeae to be
developed betweer a flat stee- surface and a two-inch wide. str:--ht, solid, *:eel 'ar.Tne length of the bar to be at lepst as long as the diameter nf tne package a_- the
longitudinal axis of the bar i:. be parallEl tj ti; ,,lare of the fiat surface. The load to
be applied to the bar in a marner th-.S- prer.ents any mrer.,,rs or devices used t, support the
bar from contacting tý-e packace.
2-15
3. Packages weighing less than 500 pounds to be placed upon a flat, essentially unyielding,
horizontal surface and subjected to a weight of 500 pounds falline from a heict of ten
feet, striking in the position expected to result in maxium, damag- at the coclusion of
the test sequence. The end of the weight. contacting the packaoe U be a soliý probe nade
of mild steel. The probe to be the shape of the frustum of a righ: circular cone, 12
inches long,,eight inches in diameter at the base, and one inch in diameter a: the end.
The longitudinal axis of the probe shall be perpendicular to the tmrizontai s-rface. For
packdges weighing 500 pounds or more, the base of thie probe to be placed on a flat,
essentially unyielding surface and the package dropped fror a heigt of ten feet onto the
probe, striking in the position expected to result in maxirpx da-ace at the cUnclusion of
the test sequence.
4. The package to he firmly restrained and supported such that its. "ooritudinal axis is
inclined approximately 45 degrees to the horizontal. The area - -he package which made
first contact with the in-pact surface in test 1, above, to be in te lowermost position.
The package to be struck at approximately the center of its verticil prjjectic• by the end
of a structural steel angle section falling from a height of at leist 150 feel. Tne angle
section to be at least six feet in length withequal legs at least five inches long and
one-half inch thick. The angle section to be guided in such a way to fall end-on, without
tumbling. The package to be rotated approximately 90 degrees about its longitudinal axis
and struck by the steel angle section, falling as before.
5. The package to be exposed to luminous flames from a p.ol fi.-e of JF_4 or JP-5 aviation
fuel for at least 60 minutes. The luminous flames to extend an average of at least three
feet and no more than ten feeL beyond the package iPt all horizental directions. The
position and orientation of the package in relation to the fiel to be that which is
expected to result in maximum damage at the conclusion of the test sequence. Ar alternate
method of thermal testing may De substituted for the above fire test provided -. at the
alternate test is not of shorter duration and would not result in a lo.;er heating rate to
the package. At the conciusion of the thermal test, the package shill be allowed to cool
naturally or shall be cooled by water sprinkling, whichever is expKcted to res.lt in
maximum damage at the conclusion of the test sequence.
6. Immersion under at least three feet of ý;ater for at least eight hou-s.
The acceptance standards for the packaye following this series of tests are:
1. Containment - The containrment vessel must not be ruptured ir. its Doost-tested condition and
the package must provide a sufficient degree of containment to restrict accumv½ted loss
of plutonium contents to not more than an A2 quantity* in a period -f one wee.
Ar A. quantity of plutonium is defined in Table VII of the !nternationE' Atomic En-gy AgencyPeoulations for the Safe Transport of Radioactive Materials, Ip Safet: Series No. 6.
2-16
I
2. Shielding - DemonsLration that the exte. ial radiation level would not exceed one Rem per
hour at a distance of three feet from the surface of the package in its post-tested
condition in air.
3. Sub-Criticality - A single package_ and an array of packages shall be demonstrated to be
sub-critical it accordance with 10 CFR Part 71, except that the damaged condition of the
package shall be considered to be that which results from the above qualification tests
rather than the conditions specified in Ap~endix B of 10 CFR Part 71.
The NRC Qualification Criteria also require tha: the package withstand an external water
pressure of at least 600 psi for not less than eight hours, without detectable water leakage
into the containment vessel.
A further requirement of the NRC Qualificatior Criteria for impact testing at terminal free-
fall velocity is not applicable to the PAT-i package becaust the terminal free-fall velocity of
the packag.Ž at sea-level is less than 422 fps (Section 2.8.4),
To demonstrate that the PAT-I package meets the structural integrity requirements of the NRC
Qualification Criteria, five prototypt PAT-i packages were subjected to the ,escribed sequen-
tial tests. In these five tests, the impact orientation of the package wa, varied. Since the
orientation which would vroduce maximurn package damaoe at the conclusion oi the test sequence
could not b>e readily determined by analysis, incact tests were performed on the top end, top
corner, side, bottorm corner and bottom end of tke packages.
Ic. demonstrate that the package desior. mee:s the deep submersion requir-ment prescribed in the (NRC Qualification Criteria, a TB-l containment vessel was subjected to a 600 Dsi pressure in a
hydrostatic test cha;Pber.
Appendix 2.2 briefly describes the test facilities and outlines how the individual ano sequen-
tial tests were performer!. Section 2.-S.2 assesses the results of the sequential tests.
Section 2.8.3 describes the results cf the hydrostatic test and Sectior 2.8.5 describes other
tests and analytical assessments which are requ!red by the 'JRC Qualification Criteria.
2.8.2 Sequential Tests
2.8.2.1 Imoact Test at 422 fps
2.8.2.1.1 Top End Inpact (r,,
A PAT-] package was impact tested orto its top end (designated as the 0 orientatior) at 442 fps.
The impact compressed the original cEaci:aoe fror 42.5 inches to appro.irately 30 inches in
length. No opening o.,curred in the outer 6rurý and nn redwood was exposed (Fiuures 2.16 and
2.18). Post-test radlographs indicated no discer,,able effect to the TE-! containne-t vessel
(Vigure 2.7).
/- 7
2.8.2.1.2 Top Corner (30-) Impa-t
A PAT-i packaoe .a- impact tested onto its top corner (designated the 30' oriertation) at
451 fps wit'. the clemi, r'ng draw bolt oriented ir the downward pr'ition. The :e_•t resulted in
extensive deformation of 'he top end of the pa.kaqe. A srall tear in the driw- co,.tr exposed
some redwood (Figures 2.19 at,, 2.21). LA no wood was lost. Post-test radiographs ndicated rn
discernable effect to the TB-i contain-,er.z vessel (Fiqure 2.20).
2.8.2.1.3 Side Impact (90ý)
A PAT-I package was impact tested ornto its side •designated the 90 orientatior) at 445 fps.
The impact flattened the package to azproximately one-half of its original dia-:eter and exposed
the inner drum liner at thI corners. However, no redwood was visible (Fi~ures 2--2 and 2.24)
Post-test radiographs indicate that t..is impact orientation wa: the most da,",c~nn to the red-
wood, but there was no disccrnable effect to thu T.-i containment vess'.I (Fic,-e §.22).
2.8.2.1.4 Bottom Corner Impact (5.
A PAT-i pacraqe was impact tested or,:% its bottom corner (designated the 150 orientation) at
443 fps with the welded joint in the nottom clamp ring oriented downward. The test resulted in
extensive deformation of the bottom end of the package. A tear in the outside drurl exlosed the
bonding material between the drum ant the drum liner; however, neither the inter Tiner nor the
redwood were eAposed (Fioures 2.25 and 2.27). The post-test radiourahs indicaec no discern-
able effect to the TB-i containment vessel (Figure 2.26).
2.8.2.1.5 Bottom End Impect. (180')
A PAT-l package was impact tested ont its bottom end (designated the 180 orientacion) at
466 fps. The impact reduced the oricinal package length from approximately !3 inches to
approximately 30 inches. There was c-. opening through the drum or drur, 'iner cnd no redwood
was exposed (Figures 2.2E and 2.10). Pust-test radiographs indicated no disce-naie effect tc
the TB-I containment vessel (Figure 2.29).
2.8.2.2 Crush Test
A 70,000-lb crush force vas applied tnrough a two-inch wide steel bcam to the rom-i vulnerable
point on the packages fo]lowing impa-: testin.. The effects of the crush te-. we'-e n-:liqibie
(Figures 2.31, 2.32 and 2.33).
2.2.2.3 Puncture Test
A 500-lb steel spike was dopped ter;-'eet onto the point on the package where -:ie .receding
cr'jsh force had been applied. The proture test general". resulte( in a two-i,'-c• oiameter hcse
t'-ouch the drum and drur liner and ,_rtial penetration into the redwood. F-r thE side impac.-
tested package, penetration was not a_ great because of -edwood compaction po-.iced by the
iP-:act test (Fioures 2.3-., 2.3c'" 2.3-. 2.37 arid 2.3P'.
2-!F
?.8.2?.4 Slash Teq
The slish test pettrated the outer stainless-steel drur and the drum liner to a depth of
approximately four ,-rhes along the trajectory of toe fea',: angle section. The angle se-tiori
did "ot penetrate to the load spreader imbedded within Zrnt c..ter redwood components .f.i1:lre•
?.39 and 2.40).
S. •.2.5 Fire Te%-
Toe 60-minute f ire :t.picaI ly produced a flaIme teMrea tu,'& of appro• iuotu lI ??y00 F at th" h;ierno!
of the PAT-i paclaye, because the outer drums and cuter- redwood aninlus were penetrated b, the
sCash test, the fire caused all redwiod in the PAT-, pa. age to h, charr.." However, th!-
a.uinu'-. load spre.,0 elements did not redcA t he : i id... te ,cra .u-,. :a,-nre ir a,( 1 F
a,-.d the copper heat .ondu t nr tube W'J. inteqral . ",e re , oo, way.. ..'ec t Q athvn.
Arrangem' ent of a packege for the fire tent is shoh,- in 9 igure 2A.4. lira t est in prrv•. P.-
siown :r Figure 2A.4. The resilts of the tests art dis:usseo in Bectic 1:.1., .7 halow..
2.8.2.f Imr~ersio,:
Toe inu:ersinn test wasned away the redwood ash near the two slasy hWle- in eacb PAW-l :acla~e
and caused blacl so:t (suspended in the water) to spreaj ontin the Ti-I containr'ent vessel
(Figures 2.43 arc 2.44).
2.8. .7 Suilmary of Post-Test Oai,.e
Disasser.bly cf the five packages after the tests 'ea2uired use of cutting torches, band sa~s.
and hand tools (Figure 2.45).
Tne disassembled -ackages are shown in Figures 2.4" throuqV 2A.
, heat-responsive lacquer coatinc that had been a..:Iied to the base c4 ,he 75-1 containp:,C-.:
vessel indicated that the vessel temperature had rEacheJ approxir:ateiy lCT<F.
Ts the cackage were disassembled, the TB-I contairrent vessels were surveyed for contara:n.on
froy te ,'raniur si-'rogate materi•1. Using a wipE test and fluorimeter- assay technioue. ,-
urai, surrogate material was detected on the ev.er÷io, of the TE- vesseii.. TOhW T.,
srimaizps the post-test leak rates that were mea;.*ed on each TF-I corta inoent vessel.
3 Hvdrcs tat.t -est
o-T7r.-. con-ainrE-t . ssel was sutjected to an ex.tEr.al prssure ca• e -. hr PIC psi, for
-r,.. • -cer water (Figu•re 2.57). The v'e .=- leak.raie of toe Ye"s"e was l-- tha '
an ccose both .efore, and after, hydrostatic teo.ing. After- 'yic the eterior, the We
a' th, ,essel was iWentical to its pre-test weigh_. :-.., operino. tnere was no indicati:.,. -ha,
Sreen die or water "ad leaked inside the vessel.
Table 2.2
SLuARY OF THE NRC QUALIFICAIION TESTS, PAT-1 PACKAGE
Ma ximumImpact Impact Velocity Fire Duration Leak-Rate
Orientation (fps' Crush Puncture Slash (winutes) (atm. cc!sec
lop (0") 442 x X x 66 4.5 x I0"6
Top Corner (30') 451 X X X 66 4.5 x 10-5
Side (90') 445 X X X 66 1.4 x 1o.6
Bottom Corner(150ý) 443 X x X 63 5.5 x 106
End (IPO0 ) 446 x X X 66 1.9 A 10.6
2.8.4 Terminal Free-Fall Velocity
A computer program at Sandia Laboratories was used to calculate the free-fall velocity of the
PAT-i package at sed-level. The package was represented as a right-circular cylinder and was
conservatively assumed to be aerodynamically unstable (i.e., it would tumble while falling).
The package was considered to fall from an altitude of 35,000 feet. Appropriate coefficients
were used to account for the effects of drag. When tne package had fallen to sea-level, its
velocity was calculated to be 350 ft/sec. Because of the convergent nature of the calculations,
the sea-level velocity of the package would not be affected by releasing the package from ar
altitude hiaher than 35,000 ft. (i.e., 350 ft/sec is the terminal free-fall velocity of the
package at sea-level). Since thc' terminal velocity at sea-level is less than 422 ft/sec, the
NRC Qualification Criteria do not require the package to be subjected to an individual free-
fall impact test.
2.8.5 Other Requirements of the NRC Qualification Criteria
2.8.5.1 Cold Anbient Temperature Tests
An early prototype of the PAT package was cold soaked at -50ýF for Prre than 48 hours, wrapped
with insulating materials, and 2 hours-IS minutes later, was impact-tested on a 40'F day.
The bulk temperature of the package at the time of impact was calculated to be less than -410r.
The package whas impact-tested onto an essentially unyielding target at 433 fps in a side crienta-
t ior. The packaae was slightly less crushed than similar pack3oes inpact-tested under amhient
tem-e-ature conditions '(iqure 2.5k'). Radiographic inspection (Fiqg-e 2.59) indicated slicAtly
less crushing of the redwood between the outer drum and the TB-! corcainment vessel then ir
similar tests at ambient temperature. hr' damage was observed to thý TB-i containment vessel.
,he package was then crushed, punctured, burned, and iirmersed. The nost-test leak rdte of the
TE-I containment vessel was measured to be 2.4 x 10-6 atm cc/sec air. This leak rate is
comparable to those observed in similar tests at ambient temperature (Table 2.2).
This result supports a conclusion that testing at -40VF would have no significant adverse
effect on the P4T-I package.
2-20
2.8.5.2 High Ambient Temperature Tests
An early prototype of the PAT package was. hot soaked at +200OF for more than 48 hours. wra•.•ed
in insulating materials, and quickly rigged for side impact onto the essentially unvieldinz
target. The impact ve',ocity was 424 fps. The bulk temperature of the package at the time rf
the impact test was approximately 200'F.
The package was crushed similar to packages tested at ambient temperature (Figure 2.60).
Radiographic inspection indicated slightly more crushing of the redwood between the outer c'L-u
and the TB-1 containment vessel than in tests conducted at ambient temperature (Fig.re 2.6,.
There was no damage observed to the TB-I containment vessel. The package was then crushed,
punctured, burned and immersed. The post-test leak rate of the TB-1 containment veýsel was
measured to be 7 x 10-8 atn. cc/sec. This leak rate is comparable with those observed in
similar tests at ambient temperature (Table 2.2).
This result supports a conclusion that testiN the package with a full heat load (2• watts in
equilibrium with a 130lF-ambient temperature would have no significant adveise effect on tt'
PAT-l package.
2.8.5.3 Individual Application of Sequential Tests
The NRC Qualificatior, Criteria rtauire that the ability of a packaqe to withstand ".e acci.'.n:
condition test sequence must. no: ,.e adversely affected if one or more tests i!-, thu sequent' ,ere
to be omitted. The purpose of tois requirement is to assure that safety doe,. not ceoend u:,:.
the necessary occurrence o' a prior or subseauent accident condition (e.g., it may, conceiva:-ly
be possible to design a package so that adverse effects frorr. impact could be correc.ed by
melting materia, to for--. a seal during the fire test). (
The PW-I package has no design features whic': would allow any of the tests prescribed in :-,e
sequence to correct damage dune •y another test in the sequence. For the PAM-I desion, dar.qF
produced by the individual tests is cumulative throunhout the sequence and omission of one -,r
more tests in the series would not affect the ability of the package to meet the prescribec
acceptance standards.
2-2,
Figure 2.16 PAT-I Pacdage Fo(~owirvg442-FPS To End ImPscl
Figure 2.17 Radiograph of PAT-1 Paage Following 442-FPSTop End Impact
2-22
I
24" 26"'
Bottom View
(J
1 5"
ITop View
Figure 2.18 PAT.1 Dimansions Fnolnwing 442.FPS Top End Impact
Figure 2.19 PAT- I Package Following 451-FPS Comer Impect
Figure 2-20 Radiograp of PAT-1 Package Foloin9g451-Fr-STop Comet Impaci
2-24
Exr
LnL
LIn
qL
ioC
2-2
5
Figure 2.22 PAT-1 Package Follow"mq 445-FPS Side Impact
Figure 2-23 Radiograph of PAT-1 Pakage Following 445-FPSSide Impact
2--26
42.75"
A
ý-A15
15.5"
r'J
"4
15.5"
I°
29 V5
Bottom View
- 29.5
Top View
Figure 2.24 PAT-i Dimiensions Following 445. FPS Sid@ Impact
jti(~ ~?
Figure Z25 PAT-I Package Following443-FPS Botanm Corner Impact
Figurt 2.26 RadKioraph of PAT-1 Package Foilowin1 443-FPSBoflom Corner Impact
2-28
I
2?.75'
tgl
Figure 30 PAT-1 Dimensions Following 466-FPS Bottom End Impact
Figure 2.27 PAT. I Dimensions Following 443-FPS Bottom Corner Impact
Sr - *-. -.
As
A..
V.
4N'
r
Sq- A *-
M
Figur 228 PAT-1 Package Following 4a-FPS Botion End Impact
Figure 2-29 Radiograph of PAT-1 Pamage Following 466- FPSBottom End impact
2-30
(7
(.
Figure 2.31 Crush Test (Package Impact Tested on End)
Figure 2.32 Crush Test (Package Inmact-Tested on Side)
Figure 2.33 Crush Test (Package Impct-Tested on CornerO
2-31
Figure 2.34 Puncture Test (Package on Top Corner) Figure 2.35 Puncture Teit (Padcks. impwc-Taeud an Sid.I
r iqwp 2.3C) rlwnctimi I ei I Par6 11"n "n"i T-Opt I on f1w to m f7.,, nor ý Fiqiirf07.37 PlInrctiorp T#-%t IP~odaqn Impact Tomfed on. Bntfom Cornwtl
(
(.
Figure 238 Puncture Test (Package Impact-Tested on Top End)
2-33
Cb
0l
C-
rA
~i*~~ri•I •'•
- •
(
L
C'
'CC
C,:
C
K2-34
(
C
C.
Figure 242 Fire Test in Progress
2-36
Figure Z43 PAT-I Package Be-.rg Lifted Fr'Dm lmnwrson Test Pool
ARM
A-
1w.->
Figuss 244 Appearance of PAT-i1 Pack&W Folloowing Testing
(2-37
Ni
Figure 2.45 Pos-Trest Disassembly of a PAT-1 Packag
Figure 24S Package Impar-Tested on Top-End
2-3£-
Figure 2.47 Redwood tinkle of Load-Spader Tube of PAT-1
(Package Impact-Tosted on Top End)
Figure Z48 TB-i Containment Vessel Within Load-Spreader
Tube(Package impact--Te1•_
on Top End)K2-40
NJ
9' 'I
V
t . -- Pb
VIW . . 4
q-.
Figure 2.49 TB-I Containment Vl,. (Pmck~p Impact-Twtdon Top End)
Figure Z50- DismsembledPAT-1 IlPack age Top-ComnerImpact Tonted)
F
Figure 2.51 Diassembled PAT-1 lF-aiage Side- Impact Tested)
2-42
.0W
Figure 2.52 Disassembled PAT.1 (Package Botiom.Corner Impact Tested) Figure 2.53 Charred Redwood and TB,13 Vessel Inside Load Spreaders(Pm*age Bottom-Coener Impact Tested)
FgýVirn 7.54 Dlsansmrhlril PAT-1 lPnr.kngo BnttomCoarnar Impact Tested) Figure 2.55 TB-I Containment Vesal (Package Bottom.Cornerimpar:t Tested)
Figure 2.56 Disassemblea PAT-1 (Package Bottom-End impact Tested)
r-4
Figure 2,58 PAT-1 Package Following 433- FPS Side Impact atApproximately 40: F
Figure 2.59 Radiograph of PAT-1 Package Following 433-FPS SideImpact at Approximately -40 F
2-46
Figure 2.60 PAT-I Pacdage Followiv:q 424-FPS Side Impact aMApproximnately 200-F
Figure Z61 Radiograpn of PAT-1 Package Following 424-FPS SideImpact at Approximately 200 F
2-47
REFERENCES
I. "Leakage Tests on Packages for Shipments of Radioactive Materials," USNRC Regulatory
Guide 7.4, June 1975. Available from the Division of Technical Information and DocumentControl, U.S. Nuclear Regulatory Commission. Washington, D.C. 20555.
2. W. A. Von Riesemann and 1. R. Guess, "The Effects of Temperature on the Energy AbsorbingCharacteristics of Redwood," Sandia Laboratories Report SAND 77-1589, 1977. Availablein source file for USNRC Report NUREG-0361, February 197S.
3. N. J. DeLollis, "Durability of Structural Adhesive Bonds (A Review),* Sandia Laboratories.Available in source file for USNRC Report NUREG-0361, February 1978.
4. "Transportation, Shock and Vibration Tests for PARC," Menorandur from L. 1. Wilson toJ. A. Andersen, Sandia Laboratories, January 26, 1977. Available in source file forUSNRC Report NUREG-0361, February 1978.
2-48
APPENDIX 2-A
MAXIMUM WEIGHI OT CONTENTS
The NRC ual ificat'ior Criteria (NUREG-0360) require a demonstration, or analytical assessment
showing that the physical test results would not be adversely affected to a sionificant extent
by the presence, during the tests, of the actual cortents that .iill It* transported.
During qualification testing of the PAT-I package. X.2 powder was use-1 as a urrouate for the
intended package contents (pluitonium). However, :,h amount (weig!', of pluyoniuri oxide powder
that could be accoirnodated within the PC-i product can is larger tf,!- the amount (weight) of
UO2 powder used as a surrogate during the tests.
The maximum weight of radioactive material authorized for transpor4 -n the PAT-I package is 4.4
pounds (2.0 ka). Since the combined weight of the 'C-I product can, the pol-'thylene bags,
and the honeycomb spacer is aDproxinately 0.3 pound:, the maximur:t•o.al weight within the TB-I
containment vessel will be aoorox.imately 4.7 pounds. The maximur, we-'aht of thp TB-1 containment
vessel, when loaded for shiprent, will be approximaelyy4 1.7 pounds.
As described in Section 2.8, five PAT-I packages we-_ subjected to t'*, sequential tests specified
in the NRC Qualification Criteria. The amount (wei:ht) of 1102 powde- used ir each package is
showri in Table 2-A-1. Also shown. is the velocity a- wrbich. each paclage was impact tested. In
each case, the kinetic energy of the TB-I containnt-t vessel and cor:ents vwas at least 6
higher during the qua'lification tests than it would have been if :hs package had impacted at a
velocity of 422 ft/sec with its maximum. authorized weight of conten:s. This supports a
concusion that the physical :est results would no: have been advers.ely effected to a significant
extert if 2.0 kg of plutoniu'r had been present in t-e packaces durir.: the te-ts.
2-Al
Table 2-A.1
(IN[TIC EN~ERGY OF TB-i (ONTAý."NT VESSEL
Impact.Orienzatioj
Top End
Top Corner
Side
Bottor Corner
Bottom. End
Velocity(f t/sec.
442
451
445
Wt. of LoadedWt. of LIt0 TB-I Ccntairvent_(iLbsI) Vessel (Ibs)
2.97
2.31
2.66
2.37
2.3P
L1a. 3
"9. t.
13.0.:39.7
Kinetic Energy of T[,-?Containment Vessel at
Impact (ft-lbP_
1.??2 (iC) 5
1.25 (IC
.23
1 .21 (inlY
1 .-4 (C
443
466
If impact-testec at the spe:ified test velocity If 422-':'sec with maximur authorized contentweight, the kinetic energy 0f the 41.7 lb (loadcd weiaht. TB-I containment vessol wo ud be1.. 15 (O)Y ft-lbs.
2-A2
I
APPZV•.;IX 2-B "
TEST FACILTY DESCRIPTION
The following facilities were used to perforr the sequential tests: (1) a rocket pulldow'
facility for the high velocity impact tests, (?) a static test wachine for the crush test, Jý
two tower facilities for the puncture and slashing tests. (4) a fire test facility, ald ("' ar
immersion pool.
Rocket Pulldowr, Facility
The rocket pulldown facility consists of a catie suspended betweer two ridges, a carriage mr;cr.
can be .oisted from the ground to the aerial cible, an essentially unyielding target, a mo.'crail
sled track, a rocket propelled sled, and contrAl and instrumnenttion eqjipment (figures ?-•.1
and Z-B.2). To conduct the high-velocity impa:t tests, packages are suspedIed at a heý-jhl of
approximately ISS feet above the essentially Lryieldinq target. Rocket sled towlines are attachef
(Figure 2-B.3) in such a manner that the packa.es impact in the desired orientation. The zrline•
are routed vertically downward fror the packao-es and pass through pulleys that straddle tth
target. The towlines are then routed horizontally to':he rocket sled, where they are firr.T
attached to a rockft sled. The rocket sled ca&- be propelled by a variable number of high-
velocity aerial rocket motors. By adjusting tLe package height above the target and varyi,.;
the number of rockets installed on the sled, tie desired impact velocities can be attained.
The essentially unyielding target is an extensivel. reinforced 526,000-lb mass of concrete.
dpproxinately 2?• feet in diameter by 11.5 feet deep and foundec on prepared earth. The to:
surface of the concrete slab is faced with a ten-foot by ten foot plate of battlesrip arnc-.
three tc five ifches thick. The steel facing ;s welded to the steel reinforcino rre•bers iP the
concrete and grouted in place.
High-spesd photoaetrics were used to iwasure te impact velocity and orientation of the PA7-I
package.
Static Test itacr:ne
The crush test bps accomplished through use of ý static test ;ia:rnire (Figure 2-B.4,. Radic:-aFs
of the package were used to establish that the -rush and pu.nctu'e tests were perfomed wit, ".ic
package orientec in a manner tu cause maximur, t-image.
jro' Towr Faciiities
Two drop. towers and 30C, feet) mere used t, conduct the punc-tre an- siash'm tests. 0:-
towers provide fir controlling the Tocation anc orienta:ion of -:e test probes strio.ing thE
:ackage. The ge-eral arrangement o- the Epunct--e test ;s shown i. Figures 2-2.5 arn -
The gene-al arra'.-;er-ent of the slasing test v shown in Fiqure 2-. 7.
r•3
Figure 2-B.1 Rocket Pulldown Facility
AERIALCABLE
- a.~~.3
Figure 2.6.2 Rocket Pulldown Facility
(
-~ r -.S. ~-r
I-ti~-.\ A
Figure 2- -3 Towline Attachmen: to a PAT-1 Package
K2-B4
. I i I I
* '.:'
I II
N ,
Figure 2-B.4 Static Test Machine
Figure 2-8.5 Set Up forPuncture Test
Figure 2-B.6 Conical Probe Used to Conduct Puncture Test
2-P6
Figure 2-B.7 Set Up for -S%ýash Test
(.
2- 7
Fire Test Facility
Tht fire test facil)iv is basically a steel tub, approximately 10 feet in diameter. The tut is
jet into the ground with minimum freeboard above the surface. JP-4 aviation jet fuel is floated
on water within the tub. A chimney, approximately 16 feet in diameter and 10 feet hiah, is
centered over the tub (Figure 2-B.8). The chimney is suspended on wire ropes and cart be lifted
to ýrovide draft for the fire. A four-foot high fence is situated about three feet away from
the draft opening and surrounds the chirtney to lessen the effects of wind.
T he packages being tested are placed on a stand centeyed over the tub and located about three
feet above the fuel. Flame temperatures were measured with thermocouples at various heights
within the fire. In addition, thermocouples were munted on the surface of the packaaes.
Water and JP-4 fuel were gravity-fed to the burner site from storage tanks.
Water Immersion Facility
The 10-foot diameter oy 6-foot deep tub used for the burr test was also used for the water sub-
mersion test.
2-BE
i
r0,
Figure 2-1.8 Fire Test Faculty
3.0 THERMAL EVALUATI(O
3.1 Discussion
The thermal performance of the PAT-I package is adequate to ensure that its contents are safely
contained under the test conditions specified in 10 CFR Part 71 and in the NRC Qualification
Criteria, NUREG-0360 (Ref. 1). This sectior summarizes the thermal evaluations which complement
the structural evaluation in Section 2.0 and the containmect evaluation in Section 4.0.
The structural evaluation required package temperatures to be calculated under Normal Conditions
of Transport. This information was used to: (1) establisý tne experimenta! test conditions for
the heat test in Section 2.0, (2) assure that differential thermal expansior of package materials
had been properly considered, (3) assure that the long-terrn Performance of the package materials
would not degrade significantly over the container's usefui life, and (4) establish test condi-
tions for the pressure test in Section 2.0.
The containment evaluation required that maximum temperatures and internal pressures be calcu-
lated for the TB-I containment vessel. This information was used to assess the degree of
leaktightness of the five PAT-] Packages tested to the NRC Qualification Criteria.
The results of the thermal evaluation of the package under the Normal Conditions of Ttransportpecified in Appendix A of 10 CFR 11 are shown in Table 3.1.
The maximum normal operating pressure (MNOP) that will occur within the TB-I containment vessel
is 34.3 psia. For the accident tests specified in the NRC Qualification Criteria, the ma~imum
TB-I containnent vessel temperature and pressure are 1080-F and 1110 psia, respectively.
3.2 Summary of Thermal Properties of Materials
The thermal conductivity and specific heat of the redwood will vary with moisture content and
with variations in the assembly of the glued joints. The thermal property values stated in
Table 3.2 are based on experimental data from an actual assembly (Appendix 3.1), and varied
slightly fron the values given in Reference 2.
The charring data shown in Figure 3.1 (Ref. 3) was used t- assess the thernal protection
provided by the redwood. Experimental evidence from the five packages subjected to the perform-
ance tests of the N:C Qualification Criteria indicates t'-at this data is valid for predicting
TB-I temperatures.
3.3 Technical Speifications
Engineering specifications and drawings for the PAT-] package are provided ;n Section 9.0.
3-1
Table 3.1
SUMMARY OF PAT-I PACKAGE TEVI ERATURESNORMAL CONDITIONS OF TRANSPORT*
Location
TB-I Vessel and Seals
Aluminum Load Spreader
Redwood: Mean
Maximum
Stainless Steel Outer Drum:
Maximum
212°F
188°F
182°F
225CF
Minimum
-40' F
-40OF
-40ýF
-40'F
Note:
Mean 1760F -40°F
Peak 225'F -40'F
Under ambient conditions in the absence of solar heating, the outer drum
temperature will exceed ambient by 10F.
*As specified by 10 CFR 71, assuming the maximum 25-watt content decay heat in establishingmaximum temperatures.
Table 3.2
THERMAL PROPERTIES OF MATERIALS
Material
Aluminum Honeycomb
Aluminum
ETP Copper
Stainless Steel
Thermal Conduc-tivity (Ref. 2)(Btu/hr/ft'F)
N.A.
Densit)(Ibm/ft
8.1
169.0
558.0
Specific Heat•Btu/ r,,/,F)
0.21
0.21
0.09
Solar Absorptance/Emittance
90.0
N.A.
N. A.
N.A.220.0
10.0
N.A.
500.0
N.A.
0.11
0. E
0.45/0.2 @ 200°F
N.A.U02
Redwood:
Parallel to GrainPrependicular to
Grain
0.14(l+0.006T( 0 F))0.05(1+-0.006T( 0 F)) 22.0 0.19(1+0.OlT(0F)) ?4 F_.
3-2
0
U
13
4, 103
3. 103
2. 103
00 1000 2000 3000. 4000 50b 0
7;%'1F AFTER RADIANT SOURCE REACHED 1850 F iseo
Figure 3.1.a Char Velocity for Various Materials
C
U
6
4
3
2
5000
EXPOSURE TIME Iwoe
Figure 3.1-b Char Depth vs Time for Charring Materials
C
6
5
4
2
:0 20 20 40 50
DENSITy Iq, fr3I
Figure 3.1.c Char Depth vs Density
Figure 3.1 Charring Data from Reference 3
3-3
3.4 Thermal Evaluation for Normal Conditions of lramsoort
Two models were ised in the thermal evaluation: a sivplified steady-state model and a finite/ "
difference numerical model. The steady-state model assumed a constant solar heating value while
the finite difference numerical model (Figure 3.2) coosidered the daily thermal cyle (i.e.,
130°F ambient temperature with 16-hour exposure to direct sunlight). lo establi&. basic input
data for both models, a thermal test reported in Appendix 3.1 was conducted to deerrn.ine the
thermal resistance of PAT-l components along the heat path lealing froir the TB-l containment
vessel, through the copper heat. conducting tube, aluminum load spreader tube and tiscs, the
redwood, and to the outside stainless steel drum. A schematic diagram of this heit path is
shown in Figure 3.3. The thermal resistance values determined through this experim'ent are shown
in Table 3.3.
3.4.1 Thermal Models
3.4.1.1 Steady-State Model
The outer wall temperature of the PAT-] depends upon tne external environment. A steady-state
heat balance on the stainless steel surface can be written as follows:
Q+qs csA A (s -T ) (1)
where Q is the decay heat, q is the incident solar flux, os is solar absorptivitý, A is the
S S I
area of the container projected normal to the solar flax, A is the total surface area fo-
convective and radiative exchange with the local surroundings, ard 7s and 7 are vie surface
and amoient temperatures.
The surface hea' transfer coefficient h includes botý radiation and convection:
= h + h (2)c r
Dato correlations for natural convection (Ref. 4) give the mean surface Nusselt numer Nu as a
function of Grashoff Gr and Prandtl Pr numbers:
Nu = 0.53 [GrT)Pr(T)]'0 2 5 '3)
where Nu = h D/k, D is a characteristic length (diame-er), and k is the -onductivi-v of air.c
The convection heat transfer coefficient h, is obtainer from Equatiun (3).
hc= 0.53 (k/D)[GrPrIO- 2 5 '4)
The radiative heat transfer coefficient hr is defined as:
h s e (T 2 + T
hr s_ T )(T S + T) 5
3-4
REDWOOD
LUMPED ALUMINUM LOAD SPREADER(ASSUMED ISOTHERMAL)
ASSUMED INSULATED SURFACES-I (SIDE V/EWO.F .PAT 1)
0= 0
FINITE DIFFERENCEMESH SYSTEMTijk T (rj. Oj. 1k)
0=
SUNDOWNHOUR =16es= T
SUNRISEHOUR = 06s = 0
MIDDAYHOUR = 88s = ;/2
Figure 3.2 Schematic of Numerical Model Used to AssessSolar Heating
3-5
5
3
C9-
o.- rr '
~ ~(7.62)
r= 115.241
r3= 1" (27.94)
COPPER CYLINDER: kc 220(0,9u b =THICKNESS INCHES tcm)
6=0.25(0.63 k=CONDUCTIVITY Biu _a-!2 ALUMINUM PLATE: kA = 90 (0.37)
6 = 1" (2.54)3 ALUMINUM TUBE: kA = 90 (0.37)
b = 0.5" (1.27;43 REDWOOD LINER: k'- = 0.31 @ 200 F (0.0013)
STAINLESS WALL: k, = 10 (0.04)
Figure 3.3 Shematic of Internal Heat-Flow Path
3-6
Iable 3.3
EXPERIMENMALL-V *TERMINED THERMAL RESISTANCES
_ocation Thermal Resistance
TB-i Vessel and copper heat tube R0 ,1 0.56'F/watt
Copper heat tube 3nd Al tub2 load snreader, P1,3 0.3b6"F/watt
axial, resistance in copper sleeve, P, O.l 0 F/watt
radial resistance in lo-4er aluinnur. plate, R 0.05°F/watt
axial resistance in aluminum tube, R. 0.19°F/watt
contact resistance Cu/Al and A*!Al jcins,R1, 2 4 R2, 3 0.02 F/watt
Through Redwood Liner (varies sligl:ly ;ilh 7), R, O.5"F/watt
From TB-I to outside surface, RTota' 1.4'F/watt (0.41 JF hr/Btu)
The net radiant heat flux q as given by:
q- T-1 (6)
in which 7 ý 0..2 is the mean surfaze emrissively (and absorptivity) in the long wavelength range
and 0- = .1714 x TO-8 (Btu/Hr/ft2/iR4 ) is tht Steffan-Boltzman constant.
With solar heating, the local external surface terperaturE of the PAT-l may deviate significantly
from the isothermal condition implied in Equation (1). The most severe attitude occurs whcen the
sun imoinges directly on the cylindiicai side of the package. An upper bound on T for theses
circumstances can be obtained by ne-1ectirg heat loss from the ends of the package, and using
the heat transfer areas:
A = 2t. 3r3 (7)
A 2=- •r1 (8)
For ;. and r 3 of 24-inches and li-irches, calculate 11.5 ft2 and 3.7 ft2 for areas Ac and As,
respectively.
By appropriate substitution of equa:ions '21. I4), 1.5), (7), and (8) into (I) , the heat balance
at the package surface is obtai;.ed.
3_7
0 + qs,.sAs {0.53(k/D)[Gr Pr1 0. 2 5 4 ol(T2 + T 2 * + ))(A c)(TS - ) (9)5 5 S S
Then, for Q 0 25 watts (85 Btu/hr) qs = 87 watts/ft 2 (295 Btu/hr/ft 2 ), cs = 0.45, and T_
130'F (5900R), the average surface temperature TS and the TB-1 temperature T TB are obtained:
T 176'F (lO)S
and in terms of the internal heat , and , - total thermal resistance RTotal (Table 3.3).
TT= Q R Total Ts = 211OF (11)
The maximum termerature on týe surface was estimated by considering an elemental surface area
norn-al to the irninging solar flux, for which the heat balance is:
Q/A dA + ý dA : h (T - 1 ) dA (12)
Making the appripriate substitutions into equation (12), and solving for TS, .:hils the 6
maximum surfacf temperature (T)X
(Ts" MAX = 237 0 F (13
However, it is expected that circumferential conduction in the PAT-l would reduce the maximum
surface tempera!ure.
These analytical estimates are based on steady-state heat balance. It is likely that the heat
capacity of th= package will somewhat reduce the temperature rise caused by insoiation, partic-
ularly since tke sun is absent at night. The effect of these transients is addressed in the
next section.
3.4.1.2 Mathenatical Model
The transient temperature response of the PAT-I was calculated using finite difference methods
(Figure 3.2). Since the influence of solar flux is of primary interest, heat flow in the
radial and circumferential directions was-modeled, with temperature assumed uniform in the axial
direction. This is equivalent to removing the ends of the package (at the ends of the aluminum
tube) and replacinc this material with insulated planes. Such a model overpredicts the PAT-I
temperatures because internal heat and absorbed solar radiation are not lost to the surroundinos
from the end sirfaces.
To furth.r sinIify the model, a few assumptions were made concerning the internal metallic
components. Tre TB-I, the copper tube, and the aluminum losd spreader assembly were each
treated as lumzecd masses of uniform temperature. The thermial resistances between these compo-
nents, taken from the test results of Appendix 3.1 which are shown in Table 3.3, have a negli-
gible influence on temperature distribution in the wood.
/3-8
The energy equation applied to the redwood liner is the transient conduction equation:
T (rk, ?T 1 (Rk T (14)ct r•r - "r r' To
in which c, k1 , and k2 were taken as the specific heat and thermal conductivities parallel and
perpendicular to the redwood grain direction given in Tatie 3.2; the other parameters are
illustrated in Figure 3.3. The boundary conditions applied at the outer wall are as follows:
- k h (T - q) 4 os qs F(b) (15)
cos (0 < e s
< - (16)
where s is the angular l::r'n of the sun which varies during the day and r is the angular
coordinate on the surface as noted in Figure 3.2. As the sun passes over the package, is
gradually increased from 0s 0 to Us = during a 16-hour period, with qs set to zero for the
next eight hours. The daily cycle was repeated until a steady periodic behavior was established.
Lecause of the cylindrical shape of the package, the a-ea projected norraa to the insolation
remains the same throughout the 16-hour period of sunliht. This is in contrast to the case of
a flat plate for which the energy absorption varies approximately as a half-sine wave due to
geometric considerations (Ref. 5). Thus, the most severe orientation for the PAT-I package is
the one considered here with the sun moving in the (r,O)-plane.
The finite difference equations derived from (14) and (15) comprise a standard, explicit
schemee. The nodal equations for the outer boundary include the heat capacity and the circum-
ferential conduction in the stainless steel wall as well as the convective and radioactive
exchange with the surroundings. Stability and sufficient accuracy were obtained using nine
radial divisions, 15 angular divisions, and a time step of one-half mi'iute.
Before considering the effects of insolation, a test case was run corresponding to the condi-
tions of the thermal test described in Appendix 3.1. A comparison of experirental and numerical
resul:s is shown in Figure 3.4. The steady-state temperatures are in precise agreement because
the coded expressions for redwood conductivity and resistance between metallic components are
based on the test data. The transient behavior is in reasonably good agreement. The thermo-
couples which measure wall temperature were located on the ends of the PAT-l and are, there-
fore, not in intimate contact with the redwood. The calculated wall temperature refers to the
side wall where the temperature rise due to environmental heating is slowed by conduction into
the redwood. The predicted internal temperatures increase somewhat faster than the experi-
mental data, particularly in the early period. This is partly because the resistance heater
was not in direct contact with the TB-I, thus causing an initial time lag in the experimentally
observed tempeature rise. Also, the mathematical model neglects axial sinking of the internal
neat from the TB-I. These fectors would not have a significant effect on. the calculations
concerned with insolation.
3-9
200
1126. 7240
115.bl220 4
- I .w.) ----
'-30 13. 3)
160 LOWCM ION 1 RMOCOUPIIS NUM[RI CAt CALCULA ION
1 3 2 COPPER TUBE120 ALUMINUM TUBE 0
08. 4,,STAINLESS STEEL WALL_ _._
InlO{ 37. 8)
80(26. 7l
115,610 2 4 6 8 10 12 14 16 18 20 22 24 26 Z8 30 32 34 36 38 40 42 4.1 46 48 50
ELAPSED TIME ItlOURS)I
Figure 3.4 Results of Thermal Test (Solid Lines) and Comparison with Numerical Calculation (Symbols)
The response of tie PAI-1 package was calculated for a solar flux of q. = 295 Btu/hr/ft 2
passing over ,.he cylindricel surface from 0s. = 0 to es = , in a 16-hour period followed by
:-ig5. hours of darkness. The ambient temperature was held steady at 130'F throughout the day
for five conseci:tive days. After the third day the thermal cycle shown in Figure 3.5 was
well established. The representative temperatures increase steadily during insolation, reaching
maxinum values at sunset or shortly after. The maximum TB-i tec-perature, the mean surface
temperature Ts, and the mean redwood temperature w are as follows:
TTB 212'F
T 176cF (17)
T-W < 182'F
These values are essentially the same those obtained by the analytical estimates in equations
(10) and (11) of the previous section, indicating that 16 hours of insolation has almost
brought the package up to quasi-steady state.
Ihe most severe temperature distribution in the redwood is shown in Figu-e 3.6. The peak
temperature occurs at the surface near = where the insolation impinc~s the surface.
(T )MAX < 224VF (18)
This is about 14'F less than the analytical estircate in the prev•ious section because circur-
ferential conduction and transient effects are no* includee. Since the circumferential c.,n-
ductivity of the wood is small, the aluminum load spreader plays a signi~icant role in carrying
heat from the hot side to the cold side of the package. Also, the tem;mpe-ature distribution is
skewed slightly to the left of = because that side has beer subjectef to greater solar flix
in the preceeding hours of the day.
Figure 3.7 shows the amount of the redwood overpack which is cooler r,Ži.:ive to an, given
temperature. Ninety percent of the redwood is cooler than 200:F even at the end of the day.
This is a conservative estimate because the ends of the package, which are significantly
cooler, have not been included in the analysis.
3.4.2 Maximum Temperatures
ihe above analysis and testing of the PAT-i package supports the estima-es for maxinum termpera-
tures under nc-mal conditions of transport as shown in Table 3.1- The c-nservative estimates
that Table 3.1 is based upon are enpha-sized by a review of modelino assumptions:
a. An assum-ed geometric orientation that maximizes absorption of solar flux.
b. An exposure of solar flux for -5 hours (295 Stu/hr/ft2).
c. A high ratio of solar absorptivity to surface emissivity.
(
220(104.4)
200193.3) ALUMINUM
u- 180
,12.2)LLJ
-r
< 160c: (71.1)
2 AVERAGE)Lu
- 140
(60.0) AMBIENT
120M48.9) fSUNRISE SUNSET-- l
0 4 8 12 1. 20 24
E LAPSED TIMES IHRS)
Figure 3-5 Daily Thermal Cycle of PAT-1 Subjected toSolar Radiation
(.3-1:
~4I
220(104.4)
C-,
uj.
I-
200(93.3)
180(82.2)!
160(71.1)
140(60.0)
0 60(0) (1.02)
120 180 240 300(2.04) (3.06) (4,08) (5.10)
0 ANGULAR POSITION DEGREES (RAD.)
i•n Tomlirature Dlttributlon In odwanrd Overpaok.irst Before Sundown (Moit gavitn C'.,r!-
360(ri. 12)
(
!L
'U
9-
7)
'U
220(104.4)
210(98.9)
200(93.3)
190(87.8)
180(82-2)
170(76.7)
160(71.1)
130(65.6)
00 0.2 0.4 L.6 Q.c 1.0
CUMULATIVE FRACTION OF REIDWOOD
Figure 37 Temperature of PAT Raiwood--tRaximum NormalEnviroamern
3-14
d. An ambient temperature of 1300 F.
e. No heat loss from ends of PAT-].
f. No wind to enhance convection.
The thermal stresses and internal pressures in the TB-I are negligible at these temperatures.
Also, the characteristics of the redwood are not significantly affected by a temperature of
200*F, as demonstrated in Reference 6.
3.4.3 Minimum Temperatures
In an external ambient of -40*F with no direct solar flux, the temperatures of the outer wall
and the TB-I are estimated as -301F and +5'F, respectively, for an internal heat load of 25
watts. For a -40°F ambient without an internal heat load the entire package is a uniforin -WF.
The associated thermal stresses and pressures at these temperatures are insignificant.
M.4.4 Maximum Internal Pressure
For a T3-1 containment vessel at a temperature of 215'F with a 25-watt decay heat load, the
internal pressure is 34.3 psia (see Section 4.2.3).
3.4.5 Maximum Thermal Stresses
Temperature differences within the TB-I are not excessive under wormal conditions of transport,
and corresponding stresses are insignificant. Differential thermal expansion within the PAT-i
package is not significant in tenns of thermal contact resistance or secondary stresses whi&'
result.
3.4.6 Evaluation of Package Performance for Normal Conditions of Transport
The evaluation of the ability of the PAT-i package to safely contain its contents under norim1
conditions of transport (Section 2.0) is supported by information in this ;ection.
The TB-i containment vessel is leaktight over the range of its nor-mal operating temperatures.
The copper gasket and elastomeric a-ring are unaffected by this range of temperature, over their
periods of intended use.
[ne PC-i product can would not be affected by the normal conditia performance tests and would
Jot t.E degraded during its single shipment usage.
The metal components of the AQ-l overpack are unaffected by the tenperature range of normal
transport (e.g., -40'F to 225°F for the outer stainless steel drun). Redwood material proper-
ties experience neither short nor long-term degradation from the vean temperature range of
normal transport (-40°F to 182*F).
3-15
*~--0~- -I'' -
3.5 10 CFR 71 - Thermal Accident Evaluation
Physical tests were the primary method used to show .hat the PAT-i package meets the test
requirements specified in Appendix B of 10 CFR Part 71. However, one assessment was required to
establish chat the maximum TB-1 temperature assued in Section 4.0 is reasonable.
3.5.1 Thermal Models
3.5.1.1 PAT-i Package Response in a 10 CFR 71 Fire Environment Analytical Model
The response of the package to the fire test can be calculated utilizing data of charring rates
in wood (Refs. 3, 7).
In Reference 3, an exposed surface of wood was-sheathed in a steel skin similar to that covering
the PAT-I package following the 30-foot drop and puncture tests specified in Appendix B of
10 CFR 71. However, data was generated utilizing an external 1850°F black body radiation source
rather than a 1475°F radiation environment. Figure 3.1 shows that after 30 minutes of fire
exposure, a char depth in the range of 2.1 inches is predicted for the 22-lb/ft3 density redwood
typical of the PAT-l package. After 30 minutes, approximately 2.6 inches of virgin redwood
remains between the char front and the aluminum load spreader tube. The char front is shown to
move at a velocity of about 0.28-ft/hour (see Figure- 3.1-a).
Several tests of the PAT-] and other prototype packages indicate that if the redwood remains in
an integral condition following the fire, char front progression is curtailed (as demonstrated
in Section 3.5.1.2, to follow). Since the PAT-i package is essentially undamaged prior to the
fire test, charring stops shortly after the fire is extinguished. Thi temperature in front of
the char front can be esti.mated by equation (1g).
(T - Ti)/(Tc - Ti) = exp (-vxl/w) (19)
where T is the temperature at a distance x in frort of the progressing char front (0.22 ft), Ti
is the initial internal temperature at- x, v is the char front velocity (0.28 ft/hr), and Tc is
the char front temperature, - 550°F. Based on the high thermal resistance of uncharred redwood,the value chosen for Ti would not be :ignificantly higher-than the temperature ('\ 20O0F -
Table 3.1) of the load spreader. The thermal diffusion coefficient (ow = k/pc) can be calcu-
lated as 0.024 ft /hr, Table 3.2. The temperature of the wood adjacent to the load spreader
tube is calculated as:
I redwood @ load spreader = 227°F (20)
Since the physical condition of the PAT-i package throughout the accident tests assures that
reasonable thermal contact between the TB-i and the aluminum load spreader is maintained, the
TB-i temperature would also be at approximately the same.
3.5.1.2 Test Model - PAT-i Performance During 10 CFR 71 Accident Condition Tests
The therval test specified in Appendix B of 10 CFR 71 requires a heat input to the whole package
not less than would result from radiation exposure of 1475°F for 30 minutes with an emissivity
3-16
coefficient of 0.9 (assuming a surface absorption coefficient of 0.8). The ability of the fire
test facility to produce this environment is discussed in Section 3.6.2.2. This fire test,which lasted S2 minutes, shows that the 10 CFR 71 requirements were. exceeded (see Table 3.4).Figure 3.8 shows that the luminous flame zone extends essentially a,:ross the diameter at the topof the chimney. The internal condition of the packagi and char depth are shown in Figures 3.9and 3.10). The maximum TB-i temperature during the test was estimated at over 200°F by post-test examination of Tempilaque coating painted on the bott-.s and insidc the cover of the vessel.This coating, which degrades at 200'F, indicated a tererature slightly in excess of 2000F.*The 200°F coating inside the TB-1 cover was discolored but still intact, confirming the esti-mated 200'F temperature, the 300°F coating was completely intact. Tempilabels placed on thesurface of the PC-l indicated temperatures of .70OF and 180 0F.
Table 3.4
10 CFR 71 APPENDIX B FIRE TEST
Observed -averae-temperature-on-AQ--l-dr num: ... .. .... 800 0F- .
Observed flame temperatures in vicii,1ty of PAT-i: 2200-2300OFCuration of above temperatures: 52 minutesChar depth in cater redwood: 3.825"TB-i temperature: average 200OF
range (170F to 210°F)Cumulative char rate:* 0.37 ft/hr
*Total char depth observed was divided by 52 minutes; this is a conservative assumotion becausechar would persist beyond the 52 minute duration of the JP-4 fire.
The TB-i t.onperature achieved in the test and the corresponding temperature calcuiited byanalysis (assuring maximum PAT-I package normal operating temperatures, Table 3.11 were essen-tially the same.
3.5.2-Package Conditions and Environment
The minor dents and scratches caused by the 30-foot free drop and puncture tests would have nosignificant effect on the package during the thermal test.
3.5.3 Package Temperatures
The above analysis and testing of the PAT-i package supports a conclusion that approximately2270 F would be the maximum. TB-i temerature during the thermal test specified in Appendix B of10 CFR Part 71.
*The 200°F coating on the exterior of the TB-i vessel was degraded, possibly from the effectsof the water t5mnersion test, but the 300'F coating was cor4pletely intact.
3-17
.~ ~*.:* * ~
Figure 3.8 10 CFR Pert 71 Appefidlx B• Fire Test ,
Fiorwm 3.9 Depth of Redod Char Folowing 10 CFR Part 71 Appwnix BFire Tea
3-1I:
. .. • .... .• r••t
i >•2:::A
:it!f'•
.
Figu re 3.10 Post- Fire View of Redwood •:,
N
3.5.4 Maxiwn Internal Pressure
For a TB-I containment vessel at 2271F, the irnternal pressure is 38.7 psia (Section 4.3.3).
3.5.5 Maximum Thermal Stresses
Temperature differences and the resulting differential expansion and secondary thermal stresses
within the TB-i are not excessive under the tests specified in Appendix B of 10 CFR Part 71.
The performance is verified by test.
3.5.6 Evaluation of Package Performance for the 10 CFR 71 Accident Conditions
Because the maximum temperature predicted fovr the TB-1 (227°F) exceeds the peak normal operating
range by only 15'F, the containment vessel would not be degraded by the thermal environment in
Appendix B of 10 CFR Part 71.
For 227-F the maximum internal pressure is X.7 psia, 12 psia less than the internal pressures
generated in the l-I/2_x maximum normal operating pressure test. Testing the PAT-i package to
this pressure is routinely required in the verification tests for containment systeo fabrication
specified in Section 8.0. The elastomneric 0-ring and copper gasket were not degraded under
these performance test conditions.
The PC-i product can, when closed by crimping and sealed by v.,elding or silver soldering, meets
the requirements of 10 CFR 071.42.
The redwood within the AQ-l overpack served its intended purpose. During the f4re test, the
redwood external to the aluminum load spreaders charred to a depth of about 3 inches (consistent
with the analytical assessment described in Section 3.5.1.1). The redwood within the load
spreader tube and discs was unaffected by the- performance tests in Apperdix B of 10 CFR Part 71.
3.6 NRC Qualification Criteria - Thermal Acc-ident Evaluation
Physical tests were also the primary means uised to show that the PAT-i package meets the require-
ments specified in the NPC Qualification Criteria. The purpose of the assessment in this
section is to show that the maximus TB-i temerature assumed in Section 4.4.2 is a reasonable
upper limit which would bound all test resul'ts.
3.b.l Thermal Models
3.6.1.1 PAT-i Package Response tc the Fire Environment of the NRC Qualification Criteria -Analytical Assessment
The PAT-i packages subjected to the fire test specified in the NRC Qualification Criteria had
Leer. damaged by being impacted, cru.shed, purttr•ed, and slashed. Such distortions of package
geometry and alterations in the packaging materials (e.g., compressed redwood,) limit any
theoretical evaluation to conservative estimates.
3-21
The two ripping/slashing tests are especially significant because the resulting penetrations
exposed a significant amount of redwood to the fire. As a result, the wood continued to char
after the fire stopped.
During development testing, the burning and smoldering temperature of the redwood was observed
to occur in two ranges. If the redwood ws contained within an essenti~ly integral stainless
steel outer drum after the high velocity impact test, any continued combustion occurred through
the progression of a char front, whose temperature would be expected to range between 550°F and
620*F (see Ref. 3). If, however, the breakup of the redwood was extensive and the damage to the
outer stainless steel drum allowed "open" glowing combustion of the charcoal', temperatures of
approximately 1100*F were recorded.
Although the PAT-1 packa~e outer stainless steel drum renmins integral following the high
velocity impact test, it sustained two five-square inch penetrations as a result of the ripping/
slashing tests. Therefore, the maximum temperature to which the T1B-1 will be directly exposed
(during the fire test) will be between the temperature of the redwood char front, 550'F to
6200F, and the glowing combustion temperature of charcoal of 1100°F. The TB-1 surface tempe.'a-
ture would not be expected to vary significantly from these values (i.e., the temperature
differential caused by the 25-watt internal heat source would be inconsequential).
3.6.1.2 PAT-1 Package Response to the Fire Environment of the NRC Qualification Criteria -
Test Model (.
Five PAT-l packages were subjected to the tests specified in the NRC ualification Criteria. In
each, the TB-l contained from 1.05 to 1.25 kg of UO12 surrogate contents. To simulate TB-i
containment vessel internal pressures (as described in Section 4.4.2). 19.3 grams of water were
added to the surrogate contents.
The major variable in the sequential tests was the package orientation with respect to the
target at impact. Side, top, top corner, bottom and bottom corner impacts were conducted
followed by the crush, puncture and slashing tests, after which packages were subjected to the
specified fire test. The fire test requires exposure to luminous flames from a pool of JP-4
fuel for at least 60 minutes. The luminous flames are to extend an average of at least three
feet and no nore than ten feet beyond the package in all lhrizontal directions. The ability of
the fire test facility to produce this environment is discussed in Section 3.6.2.2. The time"
temperature records for the three fires used to test the five packages were recorded. The
estimated average flame temperature at package height, estimated average package skin (AQ-l
drum) temperature, duration of the fires, and maximum TB-l temperature during each test are
indicated in Table 3.5.
Thermocouple data plots from the tests are scattered over a fairly wide band (Table 3.5 sihows
maximum package temperatures). The estimates were made by post-test exai.ination of Tempilaque
coatings painted on the bottom and inside of the TB-I vessel covers. Although the Tempilaque
was affected by the adjacent redwood char, and by the mechanical effects fror the impact test, a
rough estimate indicated a response of about 10000 F (538%) for each TBl-l. Also, four of the
five TB-ls tested was dark blue, a color roughly indicative of 1000°F (538%) for PH13-8 Mo
3-22
Z
Table 3.5
AVERAGED RESULTS - NRC QUALIFICATION CRITERIA FIRE TESTS
Flame Temp. @ Pkg.PAT-1 Package Level for Time
I.D. Reported
Top Impact 1850-2100 0F (1)
Duration of Max. Temp.Fire of
Above 1850'F AQ-1 Drum
63 minutes > 2400'F
MinimumTotal Exposure
Time to EngulfingJP-4 Flames
66 minutes
Approx. TB-iTemp.
10O0°F
Top CornerImpact 1850-2100 0 F (2) 58 minutes
58 minutes
2150OF 66 minutes
2200OF 66 minutesSide Impact 1850-2200°F (2)
10000F
1000°F
10000F
1000°F
Bottom CornerImpact
Bnttom Impact
1700-2100-F (2)
1850-2100°F (1)
50 minutes (3) > 2400-F 63 minutes
63 minutes > 2400°F 66 minutes
I. Flame temperature 6" below packages
2. Flame temperature at package median
3. Short time dip below 1850'F during fire; package temperature continued to rise.
stainless steel heated in an oxidizing atmosphere. The ether TB-I was colored
indicative of a temperatb.,e of less than 1000*F (538'C).
mediu. yellow,
3.6.2 Package CondiL.ions and Environment
The damage to the PAT-I packages resulting from the high-velocity impact, crush, puncture, and
slashing tests (described in Section 2.0), significant~y deforms and exposes specific areas of
redwood in the outer overpack. This exposure (1) affords the fire direct access to the redwood,
and (2) permits long-term charring of the redwood following the fire.
3.6.3 Package Tempe-atures
Based on the analysis and test results in Sections 3.6.1.1, and 3.6.1.2, the TB-I is estimated
to have attained a maximum temperature of approximately 1080'F during the thermal test specified
in the NFC Qualification Criteria.
3.6.4 Maximum Internal Pressure
For a TB-I containment vessel at 10800 F, the maximum internal pressure is 1110 psia
(Section 4.4.2.).
3.6.5 Maximum Thermal Stresses
The fire tests support the conclusion that therval stresses in the TB-l do not affect its
containment integrity. Temperatures of the TB-i varied, depending on the progression of the
redwood char front, but were equalized by the aluminum load spreader. The conductivity of the
stainless steel used in the TB-i is such that significant temperature differences would not be
expected within a vessel.
3-23
3.6.6 Evaluation of Package Performance for NRC Qualification Criteria Thermal Conditions (The ability if the PAT-i package to safely cont-in its contents throughout the tests specified
In the NRC Qualification Criteria is demonstrated in Section 4.4.
Both analyses and testing show that maximum temperatures reached by the TB-l contairment vessels
during the fire test range fron' 900'F to llqO1F.
Five tests confirm that the PAT-l pdckage offers sufficirit thermal protection to limit TB-i
temperatures to beluw 11000 F. when subjected to the fire test specified in the NRC Qualification
Criteria. Post test examination of the package indicated that all of the redwood had charred.
The condition of the charred redwood indicated tha' the ree jood external to the aluminum load
spreader was Lubject to glowing combustion (characterized by 1lO0'F), while the redwood internalto the aluminum load speader was subject to smolderina (char:cterized by 550°F to 620 0 F). The
different charring conditions are attributed to available oxygen during conmustion. Further
damage from ripping/tearing of the AQ-l would not sign' '"cantly effect the thermal performance
of the TB-1 since the redwood glowing combustion teerature is limiting and the redwood
internal to the aluwinum load spreader is consumed at the lower (<620 F) terperature. The
naximum 1100°F TB-I temperature is forced by the aluminum load spreader temperature. The
aluminum -.,d spreader has a direct thermal connection through the copper liner. A good safety
margin is afforded by the fact that when the 'up' fire has burned-out, a considerable amount
of vncharred redwood remains around the containment vessel. Following the fuel fire, theremaining redwood undergoes smolderin and glowing combustion. with a correspor.Jing maximum TB-1 (/iessel temperature of approximately 1100°F.
3-24
REFERENCES
I. U.S. I..--ear kegulat-ry Commission, Qualification Criteria to Certify a Package for Airlranýport of Plutonium, NUREG-0360, January 1978. Available for purchase from Nationallechnica; Infornat,on 3ervict (NTIS), Springfield, VA 22!61.
2. F. F. Wangaard, "He;at Transmissivity of Southern ,ne wood, P•voOd, F~berboard, &dPaw tic;ebcard,7 Wood Science, Vol. 2, No. 1, PP 54-60, 1969 Available in public technicallibra'ie!
3. R. E. Brry, T. K. Hill, W'. W. Joseph, and R. K. Clarke, "Accidcnt-Resistant Container:.Materials am' Structures Evaluation,' 5andia Laboratcries Report SAhD74-O010, AugLt 1975.Available for purchase from National Technical Infoimation Service (NTIS), Springfield, VA22161.
4. A. .1. Chapman, Peat Tram.s• r, ttird edition, MacMillan, New York, 1974. .Available inpublic technical libraries.
5. J. A. Duffie and W. A. 6eckman, Solar Energy Thermal Processes. Wiley-Interzcience, NewYork, 1974. Availatle in public technical libraries.
6. W. A. Von Riesemann and T. P. Guess, 'The Effects of Te;perature on the Energy-AbsarbingCharacteristics of Redvood," Sandia Laboratories Report SAND7/-1589, to be publisked.Availabl, i.T suurze filE for USNRC Report NUREG-0361, February 1978.
7. E. I.. Schaffer, "Revicw of Information Related to the Charring Rate of Wood," U.S. ForestService Research Note FMt-0145, U.S. Department of Agriculture, November 1966. AVilablein source file fur USNPC NUREG-0361, February 1978.
3-25
IApPendix 3-A (.
Test to Establish Therml Resistance Values
of PAX-I Coponents
P thermal test was conducted to determine Me thermil resistance values for PAT-1 components.A resistance heater was placed inside the M9-1 of the PAT-] package and thermocouples wereattached at the locations noted in Table 3-A.1.
Table 3-A.1
STEADY-STATE TENPERATURES ATTAINED DURING TEST TO ESTABLISH THERMAL RESISTANCES
Designation Steady-Stat e
TC # Location on Fig. 3.4. Teiperat,,. e
2, 3. 4 Lid of TB 1 2350F
5, 6 Cu Sleeve 2 221F
7, 8 Al Tube 3 212 0F
9 Al Plate (upper) 3 212°F
10, I, 12 Outer SS Iall 4 200°F
The heat was .,un in a temperature-control1e& cim-ber maintained at approximately 20OVF. Theinternal heater was maintained at 25 watts using a variable resistance power supply. Thetransient response of the thermocouples is indicated by the solid lines in figure 3.4. Thesymbols on this plot represent comparative n~umerical calculations which are discussed inSection 3.4.1.2.
The steady-state temperatures listed in Taible 3-A.1 were used in conjucti-n with the knownheat load of 25 watts to calculate thermal resistances along the heat flow path. The steady-state temperature difference between the 1'-l and the outer surface of the FAT-l was written
as follows:
TTB - ls = KQ (1)
where Q is the internal heating rate (25 wtts) and R is the overall thermal resis*ance of theprimary conduction path shown in Figure 3-3. The overall resistance R is the !um of thesepard te resistances along the path. The axial heat flow at the ends was reglected since itcontributes less than 15% to the total comductance.
The temperature d'ifference between the T1-1 amn the copper sleeve ;ias 14'F. Since thermo-couples five and six were located near the top of the copper sleeve, this temperature drop wasattributed to the resistance %.I of the -fiberg91ss protective layer at the slip-fit joint
between the TB-i containment vessel and tihe cadiium-plated copper sleeve.
3-Al
R 0 - TcB .56°Flwatt (2)
The measured temperature difference between the copper sleeve ai•J tkhe aluminum tube was 9'F,
corresponding tc a thermal resistance of
R TAl - TCu 0.36°F/watt (3)(l,3) Q
Since thermocouples seven and eight were located near the Lop of th.• aluminum tube, this
resistance includes the axial resistance in the tLube as well as the axial resistance of the
-,1,,jer sleeve, the radial resistance in the lower aluminum plate, and the contact resistance
at the two joints
R( 1 , 3 ) = I + R2 + R3 + R 1 , 2 + R2,3
The axial resistance in the copper sleeve is
R = 1 0.1 0 F/watt (5}2-ir1 61 kc
where the parameters are defined in Figure 3.3, 3nd the radial resistance in the lower
aluminum plate is
In (r 2 /rd)R2 = 272kA O.05*F/watt (6)
As will be explained below, the axial resistance in the aluminum tube is on the order of
R3 = 0.19°F/watt (7)
This leaves a total contact resistance of approximately
R1,2 + R2.3 = 0.02°F/watt (8)
for the copper/aluminum joint R1,2 an: the aluminum/aluminum joint P2, 3. Since these are
jointed with metallic fasteners, a very small resistance was expected.
The measured temperature difference between the top of the alurninum tube and the outer wall
was
AT3,5 = 12°F ()
Since this temperature drop is related to the thermal resistance of the wood liner, it
provided a means for checking the redwood conductivity. The axial conduction in the aluminum
cylinder was coupled with outward radial conduction along the grain of the wood. Thus, the
3-A2
\
f
fin equation was applicable provided that the convection coefficient h and the outer wall
(forcing) temperature ar'e unifor. The usual convection coefficient h was r.,deflned in terms
of the radial conductance of the wood, as follows
k w (10)r2 In(r 3/ Y
Then, from the textbook solution [4],
f 2 1/2 112 1/
aT(hPk ) 12 ta nh 5- + tanh ;A (11A AT
In which AT is the temperature difference between the lower end of the aluminum tube (Lase
of the fin) and the outer wa. . Equation (11) was rearranged as follows
in(r-, r2) F4 ~ (12)T Uta 15 + tanh
in which
kA r 363 (r 3 /r 2j•-l( ("
Also, the relationship between tT and AT3 , 5 was available from the solution of the fin
equation [4]
AT = AT3,5 cosh I7 'e L (14)
-3J
Then, by combining (12) and (14),
3 n (r 3/r 2 ) inh T + coshhT3i = Q 2l."[ 'w cosh
2-3'w L -33 1 (15)
Since Q, AT3 ,5' and the geometry were all known, the value of kw is calculated from (13) and
(15) as
k = 0.31 Btu/tr/ft/°F (16)
This is in good agreement with published empi-ical exoressions noted in Table 3.2.
(
3-A3
The temperature drop along the aluminum tube was estimated using (14)
aT 3 1 AT - A13 , 5 4.8°F (17)
and the corresponding thermal resistance was
R3 = 4T3 = O.19'F/watt (18)
as given previously in Equation (7).
The thermal resistance of the redwood liner was evaluated as
R = AT3,5 -. 5°F/watt (19)4 Q
but this resistance decreases as the temperature increases due to temperature dependence of
the redwood conductivity.
Contact resistance was neglected in the analytical model (fin approxi.mation) which describes
the heat flaw from the aluminum tube to the outer wall. Nevertheless, the inferred value for
redwood conductivity, k. = 0.31, already exceeds the expected value based on published
correlations. If contact resistance had been included, the inferred value of k would be evenwgreater. This assessment appears to confirm that contact resistance is negligible at the
wood/metal glue joints.
In suLmmary, the overall then•al resistance R between the TB-i and the outer stainless steel
drum wall consists of the three major contributions discussed separately above:
R = RO,1 + R1, 3 + '4 = l.4'F/watt (20)
Although the redwood condkictivity varies with temperature, the total R changes by only 5% with
a temperature change of 50°F. Thus, the steady-state temperature difference between the TB-l
and the outer surface can be taken as
T TB - Ts = QR = 350F (21)
The experimental temperature measurements in the PAT-I were found to be consistent with a
simple analytical model of conduction heat flow. Contact resistance and redwood conductivity
were estimated from these tests results.
3-A4
(4.0 CONTAINMENT
The PAT-1 package meets the containment acceptance standards specified in 10 CFR Part 71 and in
the ERC Qualification Criteria set forth in MUREG-0360 (Ref. 1).
The ability of the TB-i vessel to meet these standards was assessed by helium leak testing with
a mass spectrometer. The assessment indicates that the PAT-1 package provides a greater degree
of cortainment than is required by the acceptance standards. The results are summarized in
Table 4.1.
4.1 Containment Boundary
The TI-i containment vessel (Figure 1.4) provides the primary containment boundary for the
PAT-] package. Within it, the PC-1 product can (Figure 1.5) provides the separate inner
container required by 10 CFR 971.42.
Table 4.1
PAT-i PACKAGE POST-TEST CONTANMENT
Component Test Condition
Noral Conditions oftransport (Appendix Aof 10 CFR 71)
RegulatoryAcceptanceStandard
No Release
Post TestHelium
Leak-Rate(atm-cc/sec)
less than IxlOT10
less than lIxlO-O
less than 4.5xi0-5
ResultsMax. Mass ofPowder Release
0
0
0.17
(
TB-1 Hypothetical Accident No ReleaseConditions (Appendix Bof 10 CFR 71)
NRC QualificationCriteria
A2 /week*
4.2 Normal Conditions of Transport - 10 CFR 971.35
4.2.1 TB-l Containment Vessel Leaktightness
During assembly of the PAT-i package, the TB-I containment vessel was filled wit;h helium at
ambient temperature and pressure and checked with a mass spectrometer for leaktightness. (The
spectrometer is capable of detecting Irakage as low as I0-l1 atm cm3 Isec).**
*For a typical mixture of plutonium oxide powder, an A2 quantity is approximately 2.55 mg.
**Leaktightness is defined in USNRC Regulatory Guide 7.4 (Ref. 2) to be leakage less than10-7 atM-cm3/sec. (
4-1
After completing the normal condition of transport tests, the 1B8- ve'.sel was placed in the mass
spectrometer. A near vacuto was drawn external to the 15-1 vesse; and helium leakage tas
measured. No helium leakage was detected. The presence of helium in the TB-I vessel was
confirmed at disassembly.
These results verify that the TB-i containment vessel remained leaktight throughout the normal
condition of transport tests and meets the regulatory acceptance standards for containment.
4.2.2 PC-] Product Can Integrity
Since the PC-i product can is closed by crimping and sealed by welding or silver soldering, no
materi3l would be released from, the product can if the package were to be subjected to the
Normal Conditions of Transport in Appendix A of 10 CFR Part 71.
4.2.3 Pressurization of Containment Vessel and Product Can
The TS-I vessel and the PC-] product can are designed for a maximum of 2.0 kg of material with a
maxinm moisture content of 16 grams of water. The authorized contents can generate a maximum
decay heat of 25 watts.
Under the heat test specified for normal conditions of transport in Appendix A of 10 CFR
Part 71, the maximum temperature of the TB-l vessel (with 25 watts internal decay heat) would be
215°F. The maximum internal pressure within the vessel would be 34.3 psia.*
4.3 Hypothetical Accident Conditions - 10 CFR §71.36
4.3.1 TB-l Containment Vessel Leaktightness
Following testing of a PAT-I package to the conditions specified in Appendix B of 10 CFR
Part 71, the TB-i vessel w-s subjected to a helium leak test similar to the test described in
Section 4.2.1. No helium leakage was detected.
These results verify that the TB-i containment vessel remained leaktight throughout the hypo-
thetical accident conditions in Appendix B of 10 CFR Part 71 and meets the regulatory acceptance
standards for containment.
4.3.2 PC-l Product Can Integrity
After leak testing, the TE-I vessel was disasefmbled and the PC-l product can was examined
(Section 2.7.1). The criPpd closure and the epoxy overbond had remained intact. The can itself
had several minor dents (Figure 2.15). During the tests, the PC-l product can was loaded with
606 grams of the UO2 surrogate material, 2545 grams of lead shot, and 19.3 grams of water. The
tto polyethylene bags, which normally hold the contents, were not used so that the powder
"'! vapor pressure of water at 215'F (15.6 psia) plus pressure from the original volume ofi:.3ted air (18.7 psia).
4-2
retaining ability of the product can alone could be tested. Helium leak testing indicated a leak
rate beyond the provisions of USNRC Regulatory Guide 7.4; however, no uranium surrogate mterial
was detected to have been released using a wipe test and fluorimeter assay technique on the
interior of the TB-1 vessel and the exterior of the PC-I product can. Since the PC-I product can
maintained its basic structural integrity during the tests, and no uranium surrogate material was
found to have been released with the can sealed by epoxy; the NRC staff concluded that with a
welded or silver soldered seal, the PC-i product can meets the requiresnts of 10 CFR §71.42.
4.3.3 Pressurization of Containment Vessel and Product Can
For the fire test specified in Appendix B of 10 (R Part 71 , the maximum temperature of the TB-l
vessel (with 25 watts internal decay heat) would be approximately 227*F. The maximum internal
pressure within the vessel would be 38.7 psia.*
4.4 NRC Qualification Criteria
4.4.1 Release of Radioactive Contents
4.4.1.1 Summary
A ne'ium leakage measurement was used to verify that the TB-1 vessel meets the acceptance
standards for containment specified in the NRC (oalification Criteria (NUREG-0360). Specifi-
cally, containment was confirmed by measuring the post-test leak rate of helium across the seal
of the TB-1 vessel. The results indicated, through correlation with experiments involving PuO2
powder (Ref. 3), that the bounding magnitude of potential PuO2 leakage from the TB-l vessel would
be less than 0.17 mg in one week.**
4.4.1.2 Gas Leak-Rate - Powder Loss Correlation
During assembly of the five PAT-i packages, the 18-1 containment vessels were charged with helium
at ambient temperature and pressure. Each TB-i vessel was then leak tested. None of the vessels
indicated a detectable leakage of helium (less than 10-10 atm cm 3/sec), After being tested to
the conditions specified in the NRC qualification Cr4iteria, the TB-I vessels were again leak
tested. The results, converted to air leakage, are indicated in Table 4.2.
Table 4.2
POST-TEST TB-1 AIR LEAKAGE RATES
MaximumPackage Impact Air Leakage RateOrientation (atm cm3 /sec)
lop End (00) 4.5 x I0"6
Top Corner (300) 4.5 x 10"5
Side (90*) 1.4 x 10-6
Bottom Corner (150°) 5.5 x 10-6
Bottom End (1800) 1.9 x 10-6
*The vapor pressure of water at 227°F (19.6 psia)plus pressure from the original volume of
heated air (19.1 psia)."This assessment represents a conservative upper limit (i.e., conservatively high differential
pressures were assumed to exist between the environment and the TB-1 vessel, and the vesseland its PuO 2 contents were assumed to be continuously vibrated for one week).
4-3
The potential PuO2 powder release was deiermined using these air leakage measurements. The
evaluation is based upon the following conservative assumptions.
(1) the seal of the TB-i vessel (Figure 4.1) leaks gas through a single straight circular
channel of constant diameter, rather than through the more probable distribution of smaller
holes,
(2) the PuO2 particles are entrained without affecting the fluid flow escaping through the l1•k
(i.e., no settling, blockage or attenuation occu.rs, and particle diameters << hole
diameter),
(3) the diameter ef this ciiatnel can be calculated using the measured post-test air leakage
rates (Table 4.2),
(4) the time-history of the pressure within the TB-l vessel is boinded by the curve shown in
Figure 4.2,
(5) the TB-i -and-its contents are subjected-to continuous agitation following the -tests, and
(6) release of PuO2 powder from the TB-i vessel (through a channel) would be the same as
observed in experimental measurements of PuO2 leakage through an orifice having a diaimeter
approximately the same as the effective diameter (see items 1, 2, and 3, above) of the TB-I
leak channel.
Using the equation below, a diameter, D, of 6.6 ljm was calculated to correspond to a leak-rate of
5 x lO-5 atm-cm3/sec* and a 0.716 cm channel length. The 0.716 cm length corresponds to the
copper seal width. The equation below is based on Poiseuille laminar continuum flow and free
molecular flow. The TB-l internal pressure would typically be about 34.3 psia (Section 4-2.3)
but would increase for a short time during the fire envirornment to 1110 psia** maximum
(Section 4.4.2).
L = 3810 R-- [323 R (Pu2 -F d2 )+ -•(pu pd)]
where:
L = Gas leakaoe, atm-ca 3/sec
D = Effective diameter of leak, cm
a = Effective length of leak, cm
ti = Gas viscosity, centipoises
Pu = Upstream pressure, atm
Pd = Downstream pressure, atm
T = Gas stagnation temperature, *K
M = Gas molecular weighc
The 5 x l" 5 atm cM3/sec valuf bounds all values in Table 4.2.
Above 30 psia pressure differential, gas flow is limited to sonic velocity.
4-4
. . . . .... .. . . .. .. . . . .. i
rII
1 K
COPPER GASKET
\$
DOUBLE-OPPOSEDKNIFE EDGE SEAL
Figure 4.1 Cross Section of TB-i Sods
4-5
I ./-I ..
1100
Z0
U.0
U)
4 aysLongest Fire Duration
6001
-Typical TB.1 Twnpoatwur:Fire Test of Development Unit)
4001
200 - 21 rF
0 I " AA
0 20 40 60 80 100 120 4 DAYS I
TIME FROM BEGINNING OF FIRE EXPOSURE (MIN)
Figure 4.2 Maximum TB-i Temperature and Pressure ProfilewLharing NRC Qualification Criteria Fire Test
I¥EF%•
Potential release from the TB-1 vessel was evaluated using data from experiments which measured
the release of Puv 2 thrwgh 5 zm, 10 on and 20 pm orifices with differential driving pressures of
500 psi and 1000 psi. b these experiments, the anparatus containing the PuO 2 powder and the
orifice was subjected to continuous vibration. The maximum values of PuO2 release that were
measured im the tests, at pressures of 1000 psi and 500 psi and orifice diameters of 10 um and
5 um, were used to calcailate an upper bound of PuO2 release from the TB-i vessel over a one week
period. By interpolatisg the data for 10 us ar.j 5 Igm hole diameters, release rates of 0.225
ug/10 min and 0.029 gng/o mi-n were obtained for 1000 psi and 500 psi pressures. The post-test
pressure within the TB-I vessel is bounded -y four days at 1110 psia, followed by three days at
500 psia (Figure 4.2). The release rate data for 1000 psi was extrapolated to 0.270 Lg/lO min at
1110 psia. By integratian, the maximum release was determined to be 0.17 j.g of PuG2 within a one
week period. This quartity of material is much less than typical A2 quantities of mixed oxides
and is also less than av A2 quantity of any isotope of plutonium.
4.4.2 Pressurization of Containment Vessel
At the peak temperature attained from the fire test specified in the NRC Qualification Criteria
(10800 F), the internal vressure within the TB-I vessel would be less than 1110 psia. This total
includes approximately 772 psia from superheated steam at 10800F, 49 psia fromi heated air within
the vessel, znd 285 psia from the ethylene gas (the assumed decomposition pr~duct from the two
polyethylene bags).
These pressures were based on an original free volume of 1156 cm3 .* The five PAT-] packages
tested to the NRC Qualification Criteria contained from 1048 to 1254 grams of UO , to which 19.3
grams of mater were added. Four of the TB-l vessels within tVise packages appeared to have
reached tenperatures approximating the estimated maximum of 1080'F. The total internal pressure
attained in these tests was calculated to range between 1144 and 1183 psia, which exceeds the
1110 psia design pressure. The total pressure includes 833 to 867 psia from supertwated steam,
49 psia from heated air, and 262 and 267 psia from decomposition of the polyethylene bags. These
pressures were based ona calculated free volume ranging from 1235 to 1261 cim3 . The calculated
pressure in the TB-l vefsel would be slightly less than the actual pressures achieved luring the
tests since the UO02 matEr-ial had a small initial moisture content (no greater than 0.4 w/o or
approximateli 4 gm H20 M This effect was not considered in the internal pressure calculations.
It is concluded from ttE above that the internal pressures which were generated within the
packages during the fire test properly simulated the pressures which would have been experienced
by a PJ.T-1 package loadhd with maxinun PuO2 contents having maximium moisture content.
*The original free volme was calcalated by taking the TB-i internal volume (1460 cm-) and
subtracting (1) the vY!ume of the PK-i product can (25 cm3), (2) the volume of tke aluminumhoneycomb spacer (13 C3), (3) the volume of PuO2 contents assuming maxinum particle densityof 8 gm/cm3 (250 cm3). and (4) the original volume occupied by the H20 (16 on-). K
4-7
REFERENCES
I. U.S. Nuclear Regulatory Coimission, "Qualification Criteria to Certify A Package for AirTransport of Plutonium," NUREG-036C, January 1978. Available for purchase from NationalTechnical Information Service (NTIS), Springfield, VA Z2161.
2. Leakage Tests on Packages for Shipment of Radioective Materials, USNRC Regulatory Guide 7.4,June 1975. Available from the Division of Technical Informationi and Document Control, U.S.Nuclear Regulatory Commission, Washington. DC 20555.
3. L. C. Schwendiman, et al., "Study of Plutonium Oxide Leak RaLes from Shipping Containers,'Battelle, Pacific Northwest Laboratories Report fo. BNWL-2260-4. Available in source filefor USNRC Report NUREG-0361, February 1978.
4-8
5.0 SHIELDING EVALUATION
5.1 Discussion and Results
The PAT-1 package meets the rauiation dose-rate standards prescribed in DOT (49 CFR 6173.393(1))
and ERC (10 CFR 571.36(a)(1)) regulations. The package also Meets the dose-rate standar-";
specified in the "NRC Qualification Criteria to Certify a Package for Air ;ransport of F Monium"
(NUREG 0360). Table 5.1 summarizes the dcse rates calculated for the PAT-i package under normal
and accident conditions, which are well within regulatory standards.
The w. terials to be transported in the PAT-I package will not require extensive shielding aor
dose rates to be within allowable limits. Thus, the difference in dose rate between normal
conditions (undamaged AQ-l overpack) and acridpnt conditions (damaged overpack) for the PAT-l
package will be small. Since these dose rates do not differ greatly, a PAT-I package which
meets the radiation levels for normal conditions would also be within allowable limits fo"
accident con Itions.
The contents of the PAT-1 package are li a a maximum internal decay heat load of 25 vatts.
The results in Table 5.1 are based upon i .ckage being loaded with 2.0 kg of recycle pluto-
nium oxide having a high concentration of ap2ricium. Due to the 25-watt heat load limitation,
the maximum quantity of nigh americium corcentrated material that could be transported in the
PAT-I would actually be limited to approximately 1648 grams. Accordingly. the dose rates shown
in Table 5.1 would be reduced.
5.2 Calculational Method
Radiation dose-rate calculations were made with AMISN (Ref. 1), the one-dimensional discrete
ordinates multig.oup computer program. Karsen and Roach 16-group neutron cross sections
(Ref. 2) and an 11-group gamma cross section library, generated by GAMLEG (Ref. 3), were used
with the program. The first-order Legendre polynomial anisotroplc scattering approximation was
used with fourth-order direction cosines and weighting value quiadiatures. The difference
between calculated results with first-order and third-order Legendre polynominals for scattering
was found to be insignificant for these problems.
The gamma and neutron sourze strengths and spectra used in this evaluation were derived by
assuming a reactor initial fuel charge wtrch consists of 1 .25 w/o U2 35 and 98.75 w/o U2 38 .
Through use of the ORIGEN computer progras (Ref. 4) this fuel was irradiated in a thermal
neutro:i flux of 1014 n/cm 2-sec for one year. The plutonium isotopes were then separated and
radioactive decay was allowed to take place for 13.5 years to permit a high build-up of 95 2 41
This calculation is based on the conservative assumption of a greater-than-normal neutron
source within t he package and is used to demonstrate that if the package meets the radiation
levels for normal conditions of transport (49 CFR §173.393(i)) it will also meet the accident
dose rate requirements in 10 CFR s7l.36(a)(1), regardless of age of contents.
5-1
Table 5.1
CALCULATED RADIATION DOSE RATES FOR A PAT-I PACKAGE hOADED WITH 2.0 KgPuO2 HAVING HIGH AMERICIUM CONTENT
A. Undamaged Package
(i) Dose Rate at Container Surface
Primary gimnas 12.9Secondary gammas 0.3Neutrons 26.5
Total 39.7 mremlhr
kequlatory Limit 200 rmrem/hr
(ii) Dose Rate at Three Feet fron Surface
Primary gammas 0.7Secondary garmmas -
Neutrons 1.2Total T9 mrem/hr
Regulatory Limit 10.0 P•rem/hr
B. Damaged Package-.(-TB-1-Containment Vessel Assumed to be Bare)
(i) Dose Rate at Three Feet from Surface
Primary gammas 1.5Secundary gammas -Neutrons 3.1
Total 4.6 mrem/hr
Regulatory Limit 1000 wtm/hr
aPlutonium oxide daughter product.
5.3 Source Specification
The gamma and neutron source strenvths calculated later in this section are based on a 1648 gms
of recycle PuO 2, which corresponds to the maximum authorized internal decay heat load of 2S
watts. The results reported in Table 5.1 for 2000 grams of this material were extrapolated from
the calculations for 1648 grams of PuO2 described below.
5.3.1 Gamrwa Source
The neutron and gamma radiation source trms were obtained from decay chain calculations with the
ORIGEN computer program. Identical radiation sources were used in the damaged and undamaged
package models.
Calculations inJicate that the radiation scurce strength of the fuel increases as a function of
time, primarily as a result of the decay of 9 4 Pu' 4 1 into " ,241 which has a. If life ofappro):imatel5 24 wit
approximately 13.5 years. The controllirg isotope for alpha decay thermal power is , with
a half-life of approximately 466 years.
Table 5.2 shows the ganyna source spectra for the assumed package contents and the weighted
nergy spectrum values for converting photor. source strength into gamma dose rates (Ref. 5).
5-2
Table 5.2
GAMMA SOURCE SPECTRA FOR 1648 gas PuO2VITH HIGH AmCONTENT*
Group
1
2
3
4
5
6
7
8
9
10
11
Mean Energy(Rev)
3.25
2.75
2.38
1 99
1.55
1.10
0.630
0.300
0.150
0.050
0.005
Upper Emergy(New)
3.50
3.00
2.60
2.20
1.80
1.35
0.900
0.400
0.200
0.100
0.010
Total
Source Strength(Photons/se c
6.40 x 104
1.56 x 107
8.17 x 104
4.46 x 105
1.02 x 106
2.05 x 106
1.83 x 108
5.09 x Inl8
1.28 x 1010
5.69 x 1012
5.70 x 1012
Responsesmrem/hr
(Pho tons /cm-sec-T
4.3F x 10 " 3
4.00 x 10-3
3.71 x 10"3
3.24 x 10-3
2.77 x 10-3
2.30 x 10-3
1 .51 x lo-3
8.30 x 10-4
3.60 x 10-4
3.70 x 10-4
3.70 x 10-4
*25 watts total internal decay beat.
5.3.2 Neutron Source
The neutron source strength for PuO2 is primarily produced by spontaneous fission and by o-n
reactions, the calculated neutron source strength for which is shown in Table 5.3. The isotopic
composition of the PuO2 as a function of decay time was calculated with the ORIGEW decay chain
program.
The calculated spontaneous fission neutron source strength of the various isotopes and other
datd are presented in Table 5.3.
The spontaneous fission neutron spectrum was obtained by distributing the source strength over
a 94Pu239 neutron fission spectrum to give the values shown in Table 5.4.
The a-n neutrons are produced by interaction of the energetic alpha decay particles (from
actinide isotopes) with the lighter nuclides (primarily oxygen) in the PuO2 . The o-n neutron
sources wert calculated from the following relation (Ref. 4):
neutrons = 1.0 X 10 Ea 3.65alpha disintegration
whLre Ea is the alpha particle energy in MeV.
5-3
(
7
l
Table 5.3
SPONTANEOUS FISSION NEUTIRON SOURCE STRENGTHBY ISOTOPE FOR 1648 9ms RECYCLE' Pu()2
N Sr F
Isotope
Pu-236
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
NNo. Atoms
6.61311015
3.63311022
2.22801024
9.178iO123
2.78001023
1.996,1023
(s ec1
6.227x10" 1
5.11100- 19
3.997x10 24
1.801xlO 19
3.997x10.24*
3.232010 19
(n/fit ssion)
2.33
2.89
2.26
3.05
2.18
Total
Sr(n/s ec)
0.
4.33xlO4
2.57x 101
3.73xl 05
3.40
1.41x105
5.57x105
FFraction
0.
7.77x10-2
4.62x10-5
6.70x10-1
6.10x10"6
2.52x10-1
'Value for Pu-24) not available; Pu-239 value used.
**25 watts total internal decay heat.
SPONTANEOUS1648 gms
Table 5.4
FISSION NEUIRN SPECTRUM FORPuO2 WITH HJR' Am CONTENT*
NeutronEnergy
Group1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
Upper Energy(Mev)
10.00
3.012
2.300
1.496
0.9072
0.4979
0.1111
3.183 x 102
9.119 x lO-3
3.355 x 10-3
9.611 x 10-4
2.754 x 10-4
6.144 x 10-5
1.371 x 10-S
3.059 x 10-6
4.140 x 10-7
fforrja Ized PuO 2Pu-239 Fiss. Spont. Fiss.Spectrum (n/sec.)
0.225 1.25 x 105
0.133 7.41 x 104
0.208 1.16 x 105
0.167 9.30 x 104
0.142 7.90 x 104
0.109 6.08 x 104
0.0143 7.97 x 103
0.0017 9.47 x 102
0. 0.
0. 0.
0. 0.
0. 0.
0. 0.
0. 0.
0. 0.
0. 0.
Total 5.57 x 105
*25 watts total internal decay heat.
5-4
The results of these calculations are presented In Table 5.5 for the PuO2 with high Am content.
Table S.5 shows that isotopes g4Pu'M and AM are primarily responsible for the alpha
particles which generate a-n neutrons.
The a-n neutron spectrum was obtained by fitting the neutron source to an experimental PU
and boron a-n neutron source spectrm (Ref. 6), the results for which are given in Table 5.6.
5.4 Nodel Specification
5.4.1 Description of Radial and Axial Shielding Configuration
Dose-rate calculations were made with ANISN, a one-dimensional computer program using spherical
geometry. A spherical geometry approximation is Justified on the basis of source size and
actual geometry in relation to the distances at which allowable dose rates are prescribed.
Spherical geometry is also more accurate with the ANISE program than either cylindrical or plane
geometry. In the spherical model, the thicknesses of the various material regions correspond to
the actual design thicknesses, and the diameter of the source region corresponds to the inside
diameter of the PuO2 container. The spherical geometry configuration for the undamaged package
model is given in Table 5.8; the geometrical configuration of a damaged package model is given in
Table 5.9.
S.4.2 Shield Regional Densities
The material composition and the atomic-nwnber densities of all regions used in the ANISN program
are shown in Table 5.11.
Table 5.5
a-NEUTRON SOURCE STRENGTH BY ISOTOPEFOR 1648 gas PuO2 WITH HIGH Am CONTENT"
N 1 Neuts Neuts
Isotope No. Atoms (sec -)is . Sec Fract.
Pu-236 6.613x101 5 7.711x10 9 5.904x10 8 3.01 --------
Pu-238 3.633x10 22 2.54311010 4.945x10 8 4.97x10 3.78x10"
Pu-239 2.228x10 24 9.0200xO- 1 3 3.933x10"8 7.91x10 4 6.02x10-2
Pu-240 9.178x10 23 3.340x10"1 2 3.550x10- 8 1.09x10 5 8.29x10- 2
Pu-241 2.790x10 23 1. 665x10"9 * 3.261x10- 8 3.641102
Pu-242 1.996x10 2 5.799x10" 14 3.273xlO 8 3.790102 --------
Am-241 2.480x10 2 3 5.074x10&11 4.99?x10" 8 6.2X1O 5 4.78x0"1
P-243 2.685x1020 2.981xlO 12 4.326x40" 8 3.46x101
Total 1.31x10 6
* a 0 8 emissions
**25 watts total internal decay heat.
5-5(
Table 5.6
u-n NEUTRON SOURCE SPECTRUM FOR 16t8 gisPuOZ WITH HIGH Am CONTENT*
NeutronEnergyGroup
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
Total
The total neutron
Upper Energy(Mev),
10.00
3.012
2.300
1.496
0.9072
0.4979
0.1111
3.183 x 10.2
9.119 x 10-3
3.355 x 102 3
9.611 x 10.4
2.754 x 10-4
6.144 x 70-5
1.371 x 10-5
3.059 x 10-6
4.140 x 10*7
Norsalizeda-n Spectrum
0.1047
0.3964
0.2392
0.1526
0.1071
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
o-nNeutrons
n/sec
1.375 x 105
5.205 x 105
3.141 x 105
2.051 x 105
1.406 x 105
0.
0.
0.
0.
0.
0.
0.
0.
O.
0.
1.317 x 106
in Table 5.7.source strength of the PuO2 is summarized
*25 watts total internal decay heat.
5-6
Table .7 .
TOTAL NEUTRON SOURCE SW YARV FOR1648 ps PuO2 WITH HIGH An CONTENT*
NeutronEnergy SF •-n TotalGroup (n/sec) (n/see) (nisec) Fract
1 1.25x10S 1.38x105 2.63x105 0.140
2 7.41x10# ,5.21x10 lO5.500S 0.317
3 1.16xlO5 3.14xlO5 4.30x 10 5 0.229
4 9.330104 2.05x 105 2.98xlO5 0.159
5 7.9OxiO4 l.41x10 5 2.2ox 105 0.117
6 6.08X, 04 0. 6.0ex10 4 3.24x 10 2
7 7.97x103 0. 7.97x10 3 4-25x10-3
8 9 .4 7x1 0e 0. 9.47x10 2 5o5X10-4
9 0. 0. 0.
10 0. 0. 0.
11 0. 0. 0.
12 O. o. O. 0
13 0. 0. 0.
14 0. 0. 0.
15 0. 0. 0.
16 0. 0. 0.
Total 5.57x105 1.32x10 6 1.88x10 6
'25 watts total internal heat.
5-7
Table 5.8
UNDAMAGED PACKAGE SPHERICAL SEOMETRY CONFIGURATION
Zone
12
3
4
5
6
7
8
9
MaterialPNO2
304 SS
PH13-81o SS
Air Gap
Redwood
6061 Al
Redwood
304 SS
Air
Thickness(cm)5.351
3.3100 x 10-2
1.397
2.032
5.105
1.270
12.62
0.3048
91.44
Outer Radius
5.3515.384
6.781
8.81313.92
15.18
27.81
28.11
119.6
Table 5.9
DAMAGED PACKAGE* SPHERICAL GEOMETRY CONFIGURATION
Zone
1
2
3
4
Material
PuO 2
304 SSPH13-8Mo SSAir
Thickness(c.)
5.351
3.300 x 10-2
1.397
91.44
Outer Radius
5.341
5.384
6.781
98.22
*TB-.1 contatirb;---t vessel assumed to be bare.
The volume of the PuO2 region in the spherical geometry model is 646.5 cm3 , and the total
density (including radioactive decay products) of the 1751-W'am source material (1648 grams of
PuO2 ) is, therefore, 2.708 gn/cm3 . These weights are consistent with the 25-watt internal decay
heat limitation.
The inside volume of the actual PuO2 contairnent vessel is approximately 1080 cm3 . This is 1.67
times the volume of the spherical model. Therefore, the total PuO2 material density would be
1.62 gm/cm3 for the actual container. The PuO 2 region composition is given in Table 5.10.
The gamma and neutron radiation source terms were calculated on the basis of the above composi-
tion. The PuO2 region for the gamma dose rate problem was represented in the computer program
by the following elemental densities:
Element
Pu
0
Atomsbarn-cm
6.060xlO-3
1.133x10-2
5-8
III
IC,
Table 5.10
ISOTOPIC ENSITIES FOR Pd2 WITH HIGH Am COMTENT
AtomsIsotope
Pu-236
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Am-241
Am-243
U-234
U-235
U-236
Np-237
0
Weig9t
2.588 x K -6
1.434 x U1I
8.e32 x 32
3.653 x U,
1.115 x 1 2
8.009 x X1
9.910 x U1
1.082 x V-1
1.557
3.337 x U-1
4.985 x U-1
1.172
1.944 x U'2
Atoms__barn-cm
1.023 x 10"11
S.620 x 10.5
3.446 x 10'3
1.420 x l0-3
4.316 x 10 4
3.087 x 10-4
7.836 x 10-4
4.154 x 10"7
6.205 x 10-6
1.325 x 19-6
1.907 x 10"6
3.010 x 10-6
1.133 x 10-2
The PuO2 region for
follows:
the neutron dse rate problem was represented in the computer proram as
E l ement
Pu-239
Pu-240
U-238
U-235
0
Atomsbaro-cm
3.8V x
1.4M5 x
7.6l9 x
1 .325 x
1.133 x
10-310- 3
10 4
10 610-2
Appropriate gamma and neutron radiation source te'ms were calculated on the basis of the above
material composition. However, in order to accoodate the computer program cross-swtion
library, the PuO2 region for the Sa dose-rate problem was actually represented by the
following elemental densities:
E l ement
Pu
0
Atirsbarn-L.,
4.01 x 10"3
8.4M x 10-3
The PuO2 region for the corresponding neutron dam rate problem was represented as follows for
the computer program:
5-5 J!I!
If
Isotope Atoms -or
_[lement
Pu-239 2.N90 x 10-3Pu-240 1.052 x 10-3U-238 9.220 x 10-4U-235 9.820 x 10-70 8.486 x 10"3
The raterial composition of other regions in the model, surrounding the PuO2 region, is given inTable 5.11. The damaged package model consists of only the first four regions iteoized in Table5.9.
5-10
Table 5.11
REGIONAL MATERIAL lIENSIG EIS FOR CALCULATIONAL MODEL
(Element
Region 1 - PuO2
See Table 5.10
Region 2 - 304 SS (IMnsity 7.92 om/cm3 )
Atom•--•-l-n
CSiCrNiFe
3.184 x1.700 x1.745 x7.728 x6.021 x
Region 3 - PH13-WIo SS (Density 7.76 gm/cn3)
04103-10"2103
10-210 -310-310"2
CrNiA]Fe
1.1476.3761.9076.358
x
xx
Region 4 - Air (Density 1.0 x 10-3 gm/cm3)
N0
3.290 x 10-.5it.860 x 10 -6
Region 5 - Redwood (Density 0.359 /cm 3 )
A literature survey provided very little information on the composi-tion of redwood; however, since redwood is very similar to spruce, itscomposition was substituted for that of redwood. Chemical analysisof a redwood sample at Sandia Laboratories later verified theappropriateness of this substitution.
ReQion 5 - Redwood (apDroximated by spruce)
(,
CHN
CaFeSi
8.93 x 10-31.338 x 10 -22.781 x 10 -35.949 x 10"-37.914 x 10-61 .291 x 10 62.007 x 10-7
Region 6 - 6061 Al (Density 2.70 g/cma3)
SiFeCuCrTiA)
3.4792.0416.9177.8275.0986.080
xx
x
10-10-41-05
10 -51010 2
Region 7 - Redwood
Same as Region 5
Region 8 - 304 SS
Same as Region 2
Region 9 - Air
Same as Region 45-11
REFERENCES
1. ANISI-W, RSIC Computer Code Lollection, Oak Ridge National Laboratory Report CCC-255.Available from Radiation Shielding Information Center, Oak Ridge, TN.
2. G. E. Hansen and W. N. Roach, Nos Alamos.Group Averaged Cross-Sections,* Los AlamosScientific Laboratory Repoqrt LX15-Ml, September 1963. Available for purchase fromNational Technical Information Service (NTIS), Springfield. VA 22161.
3. J. H. Reaken and K. 6. Adams, 'M Improved Capability for Solution of a Photon TransportProblem by the Method of Discrete Ordinates," Sandia Laboratories Report SCRR-69739,December 1969. Available for pirchase from National Technical Information Service (NTIS),Springfield, VA 22161.
4. ORIGEN, RSIC Computer Code Collection, Oak Ridge National Laboratory Report CCC-210.Available from Radiation Sbieleig Information Center, Oak Ridge, TI.
5. D. E. Bartime, J. R. Knight. J. V. ftce III and R. W. Roussin, "Production and Testing ofthe DNA Few Group Cross Sectiom Library," Oak Ridge National Laboratory ReportORNL-TN-4840, October 1975. Available for purchase from National Technical InformationSevice (KTIS)-.-So~ifielid,7-VA216I7-
6. Tyufakor, N. D_ et a]., "3 1vestigation of the Spectral Characteristics of NeutronSources, Based on P1-238, Cm-2i14 an Cf-253," Radiation Technolog, Issue 5,AEC-tr-7314, 1975. Available fir purchase from National Technical InformationService (NTIS). Springfield. VA 22161.
7. E. Hagglumd, The Co sttm of Wood, Swedish Forest Products Research Laboratory,Stockholm, Sweden. J cadew-a" Press, Inc., New York, 1951. Available for purchasefrom publisher or for viewing in puAlic technical libraries.
5-12
6.0 CRITIA.ITY EVALUATION
6.. IBscussion and Results
The A-i package meets the subcriticality nluirement; of 10 CFR Part 71 and the criteria set
forth *a the NRC mQualification Criteria to Certify a Package for Air Transport of Plutonium"
(NUR100360). A single PAT-1 package and an array ol PAT-i packages are subcritical under
normal conditions of transport ad the accidt conditions specified in the NRC Qualification
Criteria. Since the PAT-1 packae meets the subcriticality requirements for the NRC qualifica-
tion tests, it would also meet the subcriticality requirements for the iccident tests in
Appendix B of 10 CFR Part 71. The PAT-] package design meets Fissile Class I requirements when
fully loaded with its maximum authorized contts of 2.0 kg of PuO2 and its associated daughter
prodocts in any solid form, or mwxtures of mturnl or depleted UO2 with PuO2 . The fissile
contests may be at any density up to their miimum theoretical value, containing the equivalent
of 16 grams of water. Two single-layer polpethelene bags my be used to packaqe the contents.
Results of the supporting criticality calculations are given in Table 6.1.
Table 6.1
KENO CALCULATIONS ESTABLISHING PAT-] PACKAGEAS FISSILE CLISS I FOR 2.0 KG PuO
CalculatedCondition Requirements-Subcritical keff*
NORMAL Infinite number of undamaged packages. 0.402 + 0.005
NRC QUALIFI- 2?B** damaged packages, maximum reac- 0.390 + 0.005TION CRITERIA tive array; water reflecton on all sides
SINGLE Single TB-1 vessel with water leakage 0.584 + 0.006PACKAGE homgeneously uixed with 2.0 kg PuO2 ;
water reflected.
*15,00o neutron histories; Hanse-Aoach 16-group neutron cross sections.**250 damaged packages required to be subcritical pursuant to 10 CFR 71.
Parametric calculations indicate that any degree of interspersed moderation between PAT-1
packages loaded with PuO2 or between isolated TB-I containment vessels will increase neutron
absorption in the TB-i steel and thus reduce reactivity. Hence, void among an infinite array of
touching PAT-] packages for normal conditiors, and void among an array of 250 damaged PAT-1
packages represent the most reactive enviroment for conditions of transport.
Simplifications in the calculations which result in overestimating the keff are as follows:
a. Dse of Pu-239 as the only plutonium isope in the PuO2 .
(6-1
b. Use of redwood only to surround the loaded TB-1 for normal conditions; other package
materials are neglected.
c. Use of maximum theoretical density of 11.46 gps/cm3 for PuO2 for normal and accident
conditions.
d. Use of only the clustered TB-1 containment vessels are considered in the 280 damaged PAT-i
array for accident conditions; other package materials are ietglected.
e. Choice of local clustering of TB-i containment vessels considered the damage to the packages
to he greater than the damage produced by the qualification tests.
6.2 Calculational Method
All criticality analyses on the PAT-1 package were performed with the 3-D KENO Monte Carlo
coqputer program (Ref. 1), together with the Hansen-Roach 16-group neutron cross-section set
(Ref. 2). The KENO program was especially developed for reactivity estimates for arrays of
units. The cross-sL:tion set has successfully calculated fast and epithermal uranium and
plutonium homogeneous systems, as well as solutions containing these isotopes. The two unmode-
rated PuO2 criticals-discussed in -Section- 65 show-that-k 'efs- are- calculated slightly higher-
than unity when this cross-section set is used. Use of the Hansen-Roach set is therefore
conservative for the PAT-1 package analyses.
Both ordinary and generalized geometry options were employed in KENO to describe explicity in
3-D the various model configurations for normal and accident conditions of transport. Details
are discussed in Section 6.4 on Model Specifications.
To effect an infinite array in KENO in both ordinary and generalized geometry, a specular
boundary cendition is applied on the six faces of the confining planes parallel to the X, Y, and
Z coordiF.ate system basic to [ENO. These planes form a CUBOID, within which is tr. repeating
lattice. For the single-package analysis, this boundary condition is replaced by the no-neutron
return, flux fall-off condition.
6.3 Contents
The PAT-] contents are described inSecticm 6.1. The contents may have a moisture content
equivalent to 16 grams of water. The intt--spersed hydrogen from the plutonium moisture content
is a neutron poison for both bare and reflected systems characterized by a H/Pu of about 0.25.
Additionally, the polyethelene bags slightly increase neutron absorption in the 13-1 steel. For
these rea-mns, the moisture content and polyethelene bags may be neglected in the criticality
modeling.
All plutonium in the PNO2 was assumed to be Pu-239. Fast fission cross sections for Pu-240 and
Pu-242 are smaller than those for Pu-239 ever all neutron energies and fall off rapidly below
1 Nev. Pu-241 fission cross sections are also smaller than Pu-239 above 0.5 Hev, but are signi-
ficantly larger than PI-239 at energies below 0.5 Rev. However, because PI-241 will exist only
in the presence of considerably more Pu-240, Pu-242, or both, the presence of Pu-239 as the
single-fissionable isotope represents the most reactive contents and is a major conservative
assumption in the calculated keff'S. 6-2
Two specific densities for the 2.0 kilograms of PuO .th -csaienvsslWertassumed. The first. taken as 1.608 gus/ca represents comlete filling of the 2.0 kgs P..2
vithin the 1T-1 volume (1243.7 cm3). The second was taken as 11.46 gms/cm ,te theoretical
density of PuO2 . The space reining within the 1T-1 rosel was considered a void. Both normal
and accident conditions were analyzed using each density separately. The reactivitZ of this
density range shows that all densities of Pu0 2 are subcritical for the PAT-1 package. The
results are given in Section 6.4.1.
Since any amount of Pu23902 is more reactive neutronically than a Coresponding amownt of
1123502, the authorized PuO2-Unat02 mixture presents less of a criticality hazard than the
analyzed 2.0 kgs Pu2390 2 contents.
6.4 Model Specification
This section summarize: the geometric modeling r! the rflo cases for the normal and accident
condition criticality analyses of the rAI-; package.
6.4.1 Normal Conditions - (10 CFR 171.35 and 10 UCP s71.38)
Normal conditions were analyzed using the ordinary geometry avail-ble in the KENO program
mploying three concentric cylinders, i.e., the contents region, the 11-1 steel regin, and the
redwood region.
For the cuse in which 2.0 kg of dry PuO2 fills the entire volume o- the TB-i (1.608 9ms/cm3) the
contents region of the TB-i was described by a cylinder with a radilus of 5.36 oms and a height
of 13.78 oms (Figure 6.1(A)). When 2.0 kg of dry NO2 was assumed at a density of 11.46 gms/cm3 3
(theoretical crystalline density) and occupies a corresponding volue (174.5 cm3), the PuO2
region was taken as a cylinder with a radius of 5.36 cms and a height of 1.93 ons. An empty
space (VOID), modeled as a cylinder with a radius 5.36 cas and a height of 11.85 c.s, sits above
the concentrated Pu0 2 cylinder completing the description of the imoer contents (Figure 6.1(0)).
The steel walls of the TB-I were taken as a 1.43-cm thick cylindrical annular regioa with the
same thickness for top and bottom walls for both of the above contemts.
The redeood region of the PAT-1 was taken as a cylinder with a radius of 29.53 cms and a height
104.0 cms. The heights of both cylinders were specified in a manner that places the center of
the contents region at the geometric center of the redwood region.
The nuclides from the Hansen-Roach neutron cross section set and the atom numDer denities
averaged over appropriate regions in the model are given in Table 6.2.
A CUBOID (a rectangular parallelepiped - 6 planes parallel to X, Y. and Z axes basic to the KENO
geometry) was placed tightly encasing the redwood cylinder. On each of the six plames."a
specular (MIRROR) boundary was placed (i.e., 100% return of all neutrons), thus giving an
infinite array of PAT-1 packages in contact with one ancther (Figure 6.2).
6-3
ir
x
"'.'.:.:.'-.'- ........
138
CMS
Pub!
FOR BOTHTB-i CYLINDERSIR: 5.38 CMSSTEEL THICKNESSOF 1.43 CMS
0%
(A)Pu0 2 DENSITY OF 1.608 gm/cm
3 (B)PuO 2 DENSITY OF 11.48 gm/cm 3
Figure 6.1 TU.1 Geometry Used In KENO For Two Defnsity ContentsOenCsamainl Normul and Aaeidont Oodltlor8s
Tabl1 i6.2
ATOM NUMBER DENSITIES(In Units of 10• Atoms/cm3 of Region)
FOR ARRAYS (brml & Deaged) N,
REGION ORMATERMIL NUCLIDE
DENSITY PO 2 (gmslan 3 )1.608 11.46
Coutents Pu-239 0.0035751 0.025473*in TB-1 0 0.0071 02 0.050945*
Steel Cr 0.01745TB-1 N" 0.007728
Fe 0.060210
Re&ood 0 0.05455C 0.008209H 0.01230Fe 0.001437A) 0.001738
Water H 0.06688
Reflector 0 0.03344
*Aveged over 174.5 cm, of volume corresponding to a 2.0 k9Pw~z at density 11.46 qms/an 3 in TB-i.
FOR SINGLE PACKAGE
Cotentsin TB-1
Pu-239H0
0.00357510.0574970.035898
I,
Steel Cr 0.01745TI-i Ni 0.007728
Fe 0.060210
IWter H C.06688Reflector 0 0.03344
l ,-6-5
I
F 9.53 C, S
F(0 /
I TB-ISTEEL CYLINDERIR: 5&36 CMSIH: 13.78 CMSTHICKNESS: 1.43 CMS
REDWOOD
CUBOI D-PLANESPECULARBOUNDARY-CONDITION ONALL SIX PLANES
104.0 CMS
Figure 6.2 Modd of Plepmeing Lkbdinnaged Array Used In K ENO
6-6f
The KEW0 keff's for these infinite arays of PAT-1 packages under norml conditions of transport
were calculated as 0.238 + 0.003 for the 1.608 gjm/cr case and 0.402 * 0.05 for the 11.46 gns/cm3
case respectively.
6.4.2 Accident Conditions - NRC Qualification Criteria
The KENO configuration chosen for damaged conditions (see Section 2.8.2.7) urs an array of 280
crushed packages modeled as 140 hemicylinder pairs with their respective TB-1 containment
vessels touching (Figure 6.3). The array was represented as ten PAT-] packages (five pairs of
crushed henicylinders) along the X axis, seven rows of such crushed cylinders along the Y axis,
and four tiers high along the Z axis. This modeling used two MIXED BOXES; the first was the
TB-i with its contents, the second represented what would ordinarily be the redwood and the
tntra-PAT-1 regions. Calculations were performed considering the density of the PuO2 contents to
be 1.608 gm/cc and 11.46 gm/cc, with the redwood region and intra-package space taken as void.
These calculations were repeated with the void replaced by a five percent density water medium.
All four calculations used a one-foot full density water reflector surrounding the array. For
the number of neutron histories used no distinction was discernable between the keff' s for the
void and five-percent water density. The keff for the 11.46 gm/cm3 PuO2 case was calculated as
0.390 + 0.005 versus a keff of 0.239 + 0.003 for the 1.608 gm/cm3 PuO2 case. A total of 15,000
neutron histories were used in all calculations.
In addition to the calculations described above, a second KENO configuration of 288 crushed
packages (Figure 6.4) was analyzed using the GENERALIZED geometry option. A basic cell of six
damaged packages was repeated three times along the X axis, four times along the Y axis, and
four tiers high along the Z axis. In this 3-D modeling, quadric surfaces and planes were
used to describe the six TB-1 vessels in the six hemicylinders; with the two central TB-1/
vessels touching, surrounded by the four other TB-1 vessels. The array vas considered to be
reflected on all sides by one foot of water at full density. The PICTURE computer program
(Ref. 5) was used to verify that the geometry was modeled correctly. The density of the PuO23contents was taken as 11.46 gm/cm . Calculations were performed considering t0e composition
of the redwood and the intra-package space to be void and to be water at 5%, 10%. and 15% of its
normal density. The k 's that resulted from these calculations are listed in Table 6.3 andeffshow that the void case is the most reactive configuration. A total of 15,000 neutron histories
were used in-aT-l-calculations.
Table 6.3KENO Keff* WITH GENlERALIZED GEOMETRY FOR 288 CRUSHED
PAT-] PACKAGES (FIGURE 6.4)
COMPOSITION OF REDWOOD ANDINTRA-PACKAGE SPACES (PER-CENT OF NORMAL-DENSITY WATER) KENO Keff
15% 0.376 + 0.003
1 0(. 0.374 + 0.003
5% 0.383 + 0.004
0% (void) 0.387 + 0.004
/
*Hansen-Roach cross sections
6-7
IFl
z
30
x
P- SEVEN RCWSY ALONG Y AXIS
FIgiur 6.3 KENO Model For the 280 Dam•aed Case(X AxIs Only Shown)
TB-1
CRUSHED PAT-i
Figure 6.4 Model of RePosting Demaged Array Uud in KENOWth Gumuf zed Goametry
-- ,--.
6.4.3 Sinole Package - (10 CF1R 71.33)
Ordinary KENO geometry was used for a single-package case in which 2,000 grams of Pu-239. inter-
mixed with water, fill the TS-1 region. The ratio of hydrogen to Pu-239 under these conditions
was 16.1, the maximum possible for a TI-l vessel with 2,000 grams Pu-239. The TB-I vessel
was surrounded on all sides by a one-foot normal density water reflector. The KENO keff was
calculated to be 0.584 + 0.006 for 15.000 neutron histories.
6.5 Validation of Calculattonal Method
In order to ascertain the relative accuracy of the calculated k effs for the PAT-1 package.
three criticals were calculated using exactly the same methods and cross sections as in the
PAT-1 analyses. A brief description of the criticals with the result.:.g keff' s are given in
Table 6.4. Since the keff's for the wrooderated PuO2 systems are calculated higher than unity,
it can be reasonably concluded that the keff's for the PAT-) package were calculated
conservatively.
Table 6.4
KENO Keff RESULTS FOR THREE CRITICALS
Details ofNo. Critical
Critical Mass(kg-Pu) keff Ref.
JEZEBEL; fast bare 16.25 1.005 + 0.008 3Pu-239 metal sphere.Crit Dims: Radius of6.31 cms.
2 PuO compacts; unmoderated 114 1.027 + 0.007 4bar system; PUAJ2 -6 5.79gm/c5 3; 18.35 w/o Pu-740.Crit Dims: 30.78 cmsx 30.78 cms x 20.90 cas.
3 Same as Number 2. 38 1.034 + 0.007 4but reflected by % 6"plexiglas on all sides.Crit Dims: 25.65 cmsx 25.65 cms x 10.03 cas.
6-10
!
REFERENCES
1. L. N. Petrie and N. F. Cross, "IENO IV, An Improved Monte Carlo Criticality Program,Oak Ridge National Laboratory Report ORJL-4938, November 1975. Available for purchasefrom National Technical Information Service (NTIS), Springfield, VA 22161.
2. G. E. Hansen and W. H. Roach, "Six and Sixteen Group Cross Sections for Fast andIntermediate Critical Assemblies,' Los Alamos Scientific Laboratory Report LAMS-2543,December 1964. Available for purchase from National Technical Information Service(NTIS), Springfield, VA 22161.
3. G. A. Jarvis, et al., "Two Plutonium Metal Critical Assemblies,"' Nuclear Science andEngineering, Vol. 8, 525-531, 1960. Available for purchase from the publisher or forviewing in public technical libraries.
4. S. R. Bierman and E. D. Clayton, "Critical Experiments with Unmoderated PlutoniumOxide," Nucle'r Technulogy, Vol. 11, June 1971. Available for purchase from thepublisher or for viewing in public technical libraries.
5. D. C. Irving and G. W. Morrison, "PICTURE: An Aid In Debugging Geom Inpst Data,'Oak Ridge National Laboratory Report ORIL-TM-2a9Z. Available for purchase fromNational Technical Information Service (NTIS), Springfield, VA 22161.
(6-11
7.0 OPERATING PROCEMIRES
7.1 Loading the PAT-1 Package for Transport
A cutaway illustration of a PAT-1 package loaded and assembled for shipment is shown in
Figure 7.1. Prior to loading the PAT-i package, verify that the requirements specified in
Section 8.0 have been met.
7.1 .1 Loading PC-1 Product Can with Plutonium Oxide
1. The PC-1 product can must be loaded in a controlled environment which provides adequate
exhaust filtration and control of surface cleanliness.
2. Visually inspect the PC-i product can for cleanliness and freedom from defects.
3. Load the contents into-the PC-i- product-can. Verifythat-the -eight of the radioactive
contents does not exceed 2.0 kg. The contents may be contained within no more than two
single-layer polyethlene bags individually taped or beat-sealed closed. Verify that the
decay heat load cf the contents does not exceed 25 watts and that no more than 16 grams of
moisture (or equivalent) plus nine grams of polyethlene bags are present.
4. Roll crimp the lid to the body of the PC-i with a standard canning tool or machine. Clean
and inspect the exterior of the PC-I product can by standard wipe tests. Seal the product
can using an appropriate welding or silver soldering procedure.
5. Verify that the sealed PC-i product can meets the requirements specified in Section 8.2.2.
7.1.2 Loading PC-I Product Can Into TB-I Containment Vessel
1. Visually inspect the components of the iB-I vessel and top spacer for defects, damage,
cleanliness, and for correct part.
2. Place loaded PC-I product can, followed by top spacer, into the TB-I vessel.
3. Install an unused copper gasket (see Table 8.1) into the seal groove at the top of the TB-I
body.
4. Coat the elast~pmeric O-ring (see Table 8.1) with silicone "ease and install in the groove
on the TB-i lid.
5. Insert the TB-I lid into the TB-1 body, being careful not to danage the O-ring.
6. Install the 12 socket head bolts (0.500-20), fingertight, throuig the TB-'. lid into the
body.
7-1
COVER
COVERLINER
INSULATIONPADS (2)
OVERPACK.(AG-I)
CAPLUG
CLAMPRING
CAP SCREWS
DISC
,SPACER
CONTAINMENTVESSEL(TB-i)
CONTENTS
PRODUCT CAN(PC-i)
Figure 7.1 Assembled PAT-i
7(7-2
7. Tighten the 12 lid bolts, in two steps, in the fo11•ing sequence: 1-7, 4-10. 2-8, 11-5.
3-9, 12-6. Tighten all bolts to 50 foot-pounds in the first step and to 75 foot-pounds in
-the second step.
8. Attach the lifting sling to the TB-1 lid, using the three cap screws (0.250-28UNF) and
washers.
9. Verify that the assembled TB-1 containment vessel meets the requirements specified in
Section 8.2.3.
7.1.3 Loading TB-1 Containment Vessel into AQ-1 Overpad
1. Visually inspect the AQ-l components for defects, damage, cleanliness, an,' for correct part.
2. Using the lifting sling, place the TB-I contairue vessel into the AQ-1 Overpack; make sure
the TB-i is fully seated.
3. Place the inner wood plug (smaller) on top of the TB-1 vessel, aligning the notches in the
plug to clear-the-lifting sling screws in the To-!.
4. Insert in order the aluminum disc, the outer wood plug; and the inner insulation pad
(larger).
5. Insert in order the cover line, the outer insulation pad (smaller), the cover, and the clamp
ring. During Installation, align the index marks on the cover liner, the cover and the
clamp ring with the index mark on the AQ-1 overpak.
6. Install the 23 hex-head screws, fingertight, throwgh the assembled clamp ring, cover, and
cover liner.
7. Tighten the 23 cap screws to 15 foot-pounds.
8. Install the four-inch long hex head screw in the clamp ring and tighten to 50 foot-pounds.
9. Install the locking nut on the four-inch long screw-and-tighten to 30 foot-pounds.
10. A security wire and seal may be installed throagN the holes provided in the clamp ring lugs.
7.2 Procedures for Unloading the Package
1. Remove the PAT-] from the skid, if present.
2. Remove the security wire and seal, if present.
3. Remove the 23 hexagon head screws at the top of the package.
4. Remove the clamp ring bolt and lock nut.
7-3
IL. h em o tvw dnim cover.
Dr- Ren * insaltion pad.
E... Revmae m tnintme cover.
ML Remmve da oute- redwor ?I ug.
VM_. Remwvet l2iic-h diamter aluminum load spreader plate.
1M_ Remind,, inrmm redm.od plug.
IM2. Rmve * TB--1 -montainment vessel.
MIS. RevmmeleI l ng sling from the TB-i vessel.
dJT1ON: Prenme rzwlief action may accompany removal of the TB-1 vessel lid in the followimg
1RM. Reinw-t 12Z •rket head screws which secure the lid to the TS-1 vessel.
T-M. Remmuwh- liUdi -ffrm the TB-1 vessel. This can be done by re-installing the three screws
(ued,-the_- kiftinq sling) and turning them approximately evenly until the lid releases.
T . Rm ow aTlmiiium honeycomb top spacer.
717. Rom•xw PCI ;product can with contents.
VDTE: striing or preparing an empty TB-1 vessel for return shipment, care should be
tm awoid damaging the knife-edges (that engage the copper gasket) on the lid and
a of t vessel.
7-4 (1
8.0 ACCEPTANCE TESTS AND PAINTENANCE PROGRAI
8.1 kceptance Tests To Be Performed Prior to First Use of Each Package
8.1.1, Fabrication InspectionsBefore first use of each pacuging, the Inspections and determinations specified in Section 9.0
shall be completed.
8.1.2 Structural, Pressure, and Leak Rate Tests of the TB-i Containment Vessel
Before its first use. each TB-1 containment vessel must be determined to be leaktight* when
asseabled as for shipment. but without the elastomeric O-ring seal and subjected to an internal
pressure 50 percent higher than maximum rnrmal operating pressure (1.5 x 34.3 psia 51.4 psia).
8.2 Tests To Be Conducted Prior to Each Shipment
8.2.1 Visual Inspection
Note: The following steps are not necessarily in sequential order. To accomplish a complete
visual inspection, the TB-I vessel and AQ-1 overpack would have to be disassebled in
accordance with Section 7.2.
8.2.1.1 AQ-1 Overpack
a. Cteck that there are no indentations or damage points deepe. than one-half inch in the
visible interior redwood assemblies.
b. Check that there are no large breaks in the fiberglass covering of the copper heat conductor
tube.
c. Check that there are no indentations or damage points deeper than one-half inch in the outer
steel drum and covers.
d. Check that the twenty-three bolt head covers are in place around the bottom of the drum.
e. Check that the twenty-three 3/8-' -h diameter bolts are in place around the top end of the
drum (when the AQ-1 overpack is ass ibled).
f. Check that the 5/8-inch diameter, by four-inch long, clamp ring bolt and lock nut are in
place at the top of the drum (when the AQ-1 overpack is assembled).
g. Check that the vent plug covers are in place on the drum bottom and on the drum lid.
10-7 atm-cc.sec. per USNRC Regulatory Guide 7.4 which references ANSI N14.5.
8-1
- -.----.-.--.- .-.-.-.-....,--i-.-,,---,.------.-- -
S.2. .2 TB-i Containment Vesselp. :
Cloc. t k that the copper gasket Is in place on the 78-1 lid, and
b. Check that the elastomeric 0-ring is in place on the TB-i lid
m. lightly lubricated with the specified silicone grease.
that I
and V
It is free of any- gouge
hat it isfree of gouges
I1 body :(the knife edgec. Check for visible damage to the small circular knife edge in the 13-
on the lid will normally be hidden by the copper gasket).
d. Check for the presence of the 12 socket head bolts which secure t closure to the 1B-1 and
three bolts and washers for the nylon lifting sling (when the TB-l vessel is assembled).
8.2.1.3 Rejection
Mechanical damage as described in Section 8.2.1.1 or 8.2.1.2 shall be cause for rejection.
Damage to the knife-edge will be cause to perform a leakage test as described in Section 8.1.2 to
determine that the TB-1 vessel is acceptable.
8.2.2 PC-I Product Can
Prior to placing the loaded PC-i product can into the TB-1 containment vessel, a leakage test
shall be completed on the PC-i. This test sim11 indicate leakage less than 10-3 atm cm 3 'sec
from the assembled PC-1. (An acceptable test would be one in which the PC-1 is placed in a
closely fitting chamber, a vacuum is rapidly drawn, and any subsequent rise in chamber pressure
is correlated with leak rate.)
8.2.3 TB-1 Containment Vessel
As part of preparation for each shipment, the TB-1 containment vessel will be assembled in
accordance with the checklist described in Section 7.1.2 ard a leakage test shall be completed
on the TB-i together with its radioactive contents. This test shall indicate leakage less than
10-3 atm cm3/sec from the assembled TB-I. (An acceptable test -ould be one in which the TB-1
is placed in a closely fitting chamber, a vacuum is rapidly drawn, and any subsequent rise in
chamber pressure is correlated with leak rate.)
8.3 Periodic Test and Maintenance
8.3.1 T8-1 Containment Vessel
After every third use of a TB-1 containment vessel to transport radioactive materials, or if the
leak test described in Section 8.1.2 has not been performed on that particular TB-l vessel
within the preceding 12-month period, the TB-l shall be determined to meet the leak test require-
ment described in Section 8.1.2.
8.3.2 Replacement cf Gaskets on Containment Vessel
The copper gasket and the elastomeric O-ring for the 111-1 containment vessel shall be replaced
in accordance with the schedule shown in Table 8.1.
8-2
(
Table 8.1
REPLACEMENT SCHEDULE FOR COPPER SEAl AND 0-RING
TB- I Replacement Schedule
Copper Seal Prior to final closure for each shipmentof loaded TB-1
Elastomeric O-ring Prior to each 4th shipment of radioactivematerial, or annually, whichever occurs,more fre~ently.
f.- 3
I
9.0 SPECIFICATIONS AND DRAWINGS
9.1 Quality Assurance
The program to design, develop, and test the Plutonium Air Transportable Package, Model fAT-l,
and to demonstrate that the package design meets the NRC Qualification Criteria (NUREG-0360),
was conducted by Sandia Laboratories vnder contract to the Nuclear Regulatcry Commissiom.
Model PAT-1 packages must be procured, fabricated, inspected, accepted, operated, maintained,
and repaired in accordance with a Quality Assurance Plan which meets the requirements of 10 CFR
Part 71, Appendix E, and approved by the NRC.
9.2 Fabrication Requiremmets
9.2.1 General Requirements
The following estdb'ishes the general requirements for fabrication and inspection of the
Plutonium Air Transportable Package, Model PAT-1. All itenis are to be in accordance with the
drawings, standards, procedures, and material specifications to the extent specified herein.
9.2.2 Documents
The following documents form a part of this specification to the extent specified herein.
Document Issue Title
PAT-1028 Testing, Leak Rate, Mass Spectrometer, PAT Package
PAT-1029 Material Specification, Redwood for PAT Package
PAT-1030 Welding, Corrosion Resistant Steel, PAT Package
1001 A PAT-1 Assembly
1002 A Overpack, AQ
1003 A Container, Sibassembly
1004 A Drum
1005 A Cover, Modified
1006 A Ring, Clamp, Modified
1007 A Cover, Liner
1008 A Cylinder, Wood
1009 A Cylinder, Wocd
1010 A Plug, Wood, Fixed
1011 A Cylinder, Wood
1012 A Plug, Wood, Fixed
1013 A Plug, Wood, Removable
1014 A Plug, Wood, Removable
1015 A Spacer, Top
9-1 (
Document
1016
1017
1018
1019
1020
1021
1022
1023
1024
1025
1026
1027
Issue
AAAA
A
A
A
AB
A
A
A
Title
Load Spreader Assemblytootaineent VesselBolt, Socket, Head Special .500-20
Gasket, Copper
Lid. TBForging, TB Lid
Body, TB
Forging, TB BodyCan Assembly
Cylinder, Liner
Pad, Insulation
Pad, Insulation
9.2.3 Standards of Manufacture
9.2.3.1 Quality Control
Unl1ss otherwise specified, the requirements of 10 CFR Port 71, Appendix E, shall apply.
9.2.3.2 Welding Method
Welding and fabrication shall be in accordance with PAT-1030.
9.2.3.3 Glued Assemblies
a. ii gluing wood pieces together (wood to wood), polyvinyl acetate resin emulsion per
1*94-A-180 shall be used. As an alternate, glue per 9.2.3.3.b may be used, unless other-
wise specified.
b. In gluing wood to metal, or metal or metal parts, the bond material shall be a mixture of
the following by weight.
100 parts Uniroyal B635 resin
15 parts Benzoflex 988 plasticizer
9 parts Uniroyal 3080 hardener
The material shall be processed as follows: Upon receipt of the five gallon cans of
Uniroyal B635, heat for 12 hours at 130 + 10°F and then shake for 30 minutes in a paint
shaker. This is necessary because the material will freeze at close to room temperature.
Before use, the material shall be a clear, high viscosity liquid. If it is cloudy, lumpy,
er resembles an "icy slush," repeat the heating and shaking process.
After withdrawing B635 material from the can, the air space above the material should be
thoroughly flushed with dry nitrogen, and the can tightly sealed. Care should be taker. to
prevert moisture in the air from reacting with the material and forming a crystalline layer
on the surface. However, if a layer forms, it can be removed and the material used.
9-2
Is mixing, first blend the B635 and the IMmzoflex M3 plastlci•er together, not add the3090 hardener and mix thoroughly. For small batches of 100 to 200 grams, two minutes of
vigo•uS hand mixing with a spatula will be satisfactory. For larger batches, mix
therwghly to insure uniformity. To extend the pot life of the adhesive, mixing should be
done as rapidly as possible.
c. Glue lines, resulting from the dimensions of finished mating parts, shall be 0.001 to
0.1J0 inch.
9.2.3.4 Protection
All parts and asseblies shall be adequately protected from accumulation of foreign matter,
corrosion, physical damage or deterioration. This requirement shall apply to all manufacturing
operatioas from receipt of raw material to completion of a finished product, to product held in
any storage area, and to product prepared for shipment.
a. Protective measures used during processing, fabrication, and packaging must not only guard
against obvious damage and deterioration but also against the creation of latent conditions
that may later cause unsatisfactory performance, accelerating deterioration, or malfunction.
b. Raw material and parts at all levels of production shall be kept adequately segregated and
identified at all times. Parts shall be transported in a manner which will assure adequate
protection from damage.
c. All items such as raw material, parts, subassemblies, assemblies, etc., not in ltimediate
use, shall be adequately packaged, identified, and stored./
9.2.3.5 Cleanup of Parts and Assemblies
All finished parts and subassemblies shall be adequately cleaned before final assembly. Final
assembly and necessary subassembly shall be performed in an envirounent appropriate to the type
of product. All parts and assemblies shall be thoroughly cleaned to remove foreign ani manu-
facturing waste material such as:
Superfluous hardware, wire, and insulation clippings.
Chips, filings, abrasives, machining lubricants.
Soldering, brazing, and welding fluxes, solder droppings, weld spatter, slag, and welding rodends.
Drippings of lubricants, adhesives, and sealing compounds.
Paint droppings, splatter, and overspray.
Residues from liquid baths used in plating and chemical treatments.
Temporary tags and packaging.
9.2.3.6 Part Number Marking
a. All finished parts, assemblies, and subassemblies shall be permanently marked with the
applicable part number, using ink marking with covercoat. The vertical height cf characters
shall be not less than 0.12 inch.
9-3 (
b. Part numbers shall be located in any readily observable location which does not affect
function.
9.2.3.7 Standards
The following documents form a part of this specification to the extent specified herein:
MSI Y 14.5, 1973 Dimensioning and Tolerancing
MSI B1.1, 1974 Unified Inch Screw Threads (UN and UNR Thread Form)
9.2.4 Quality Assurance Provisions
9.2.4.1 The supplier responsible for the manufacture of the PAT Package must establish and
maintain a Quality Assurance Program which meets the requirements of 10 CFR Part 71, Appendix E.
9.2.4.2 Material Certification
Certification from material suppliers verifying that all materials used in the fabrication of
the PAT-i Package are in accordance with applicable drawings and specifications-
9.2.4.3 Drawing Compliance
The PAT-] assembly and component parts shall be inspected either by open set-up or gaging
methods, to assure that the dimensional requirements specified on the applicable drawings are
met.
9.3 Final Acceptance Testing
The following tests and inspections shall be performed on all units. Tests may be performed in
any sequence, as appropriate.
9.3.) '"sual Inspection
The following steps are not necesarily in sequential order.
9.3.1.1 AQI- Overpack
a. Check that there are no indentations or damage points in the visible interior redwood
assemblies.
b. Check that there are no breaks in the fiberglass covering of the innermost tubular member
(copper heat conductor tube).
C. Check that there are no indentations or damage points in the outer steel drum and covers.
d. Check that the 23 bolt head covers are inplace around the bottom of the drum.
e. Check that the twenty-three 3/8-inch diameter bolts are in place around the top end of the
drum (when AQ-I is assembled).
9-4
Check that the $-4ch diaeter, by four-Inch 10ngs clawring bolt ad lock nut are Uplace at the top of the drum (when AQ.-. is assbled).
g. Check that the vmnt plug covers are In place on the drum bottom and on the drum lid.
9.3.1.2 78-1 Containment Vessel
a. Check that the copper gasket Is in place ma the TB-1 lid and that It Is free of any gouge or
irregularity.
b. Check that the elastomeric O-ring is in place an the TB-1 lid and that it is free of gouges
and lightly lubricated with the s"Icone grease specified on the drawings, or equivalent.
c. Check for visible damage to the stall circular knife edge in the TB-1 body (the knife-edge
in the lid will normally be hidden by the copper gasket).
d. Check for the presence of the 12 socket head bolts which secure the closure to the TB-j and
three bolts and washers for the nylon lifting sling (when TB-1 is assembled).
9.3.1.3 Rejection
Nechanical damage as described In .Sections 9.3.1.1 and 9.3.1.2 shall be cause for rejection.
Damage to the knife-c 'jes will be cause to perform a leakage test as described in Section 9.3.2
to determine that the TB-i vessel is acceptable.
9.3.2 Structural, Pressure, and Leak Rate Tests of the TB-1 Containment Vessel
Each TB-1 contairnent vessel, before its first use, must be determined to be leaktight when
assembled as for shipment but without the elastomeric 0-ring seal and subjected to an internal
pressure 50 percent higher than maxim. normal operating pressure (1.5 x 34.3 psia = 51.4 psia).
Leak rate shall not exceed 1 x 10-7 atm-cc/sec (air) as determined in accordance with PAT-1028.
9.3.3 Function and Fit
The PAT-1 shall be visually and
fit into the assembly properly.
mechanically
as sped.fied
inspected to assure that
in the drawings.
all of the component parts
9-5
J(
I . . . .. .. . ..
PAT-1028 TESTING, LEAK RATE, MASS SPECTROMETER, PAT PACKAGE
1.0 GENERAL
1.1 Scope
This specification defines the requirements for the quantitative measurement of the leak rate of
a sealed component of the Plutonium Air Transportable Package, nodel PAT-1. A mass spectro-
i.eter type leak detector is used.
1.2 Product Description
The item to be leak tested per this specification is the Containment Vessel, TB, drawing number
1017, which is a major component of the PAT Package, drawing number 1001. The detail parts of
the Containment Vessel, TB, are:
TB Body
TB Lid
Copper Gasket
O-Ring
Bolt, Socket Head, Special, 0.500-20UNF (12 required)
1.3 Definitions
1.3.1 Background
Test system background is included in the leak rate results. This my be spurious output of the
leak detector expressed in suitable terms, due to the response to all gases other than the
actual leakage of tracer gas from the product being tested and/or the known leak. The background
may be inherent in the detector or extraneous, and includes absorbed tracer gas.
1.3.2 Units
1.3.2.1 Pressure Units
a. Millimeter of Mercury (mmHg). A unit of pressure corresponding to a column of mercury
exactly I millimeter high at O°C-under standard gravity acceleration of 980.665 cm/sec2.
b. Micron of Mercury (uHg). A unit of pressure eqaal to 1/1000 of the millimeter of mercury
pressure unit.
c. Torr. A unit of pressure equal to 1/760 of a stand3rd atmosphere; differs by only one
part in 7 million from a millimeter of mercury. Torr is the preferred pressure unit for
low pressure (vacuum) measurement.
1.3.2-2 Leak Rate Units
(cubic centimeter per second, standard temperature and pressure). A flow rate of
gas in terms of cubic centimeters per second in whicr the gas volume is reduced to stancdard
temperature and pressure. 1.315 cc/sec. = 1 Torr-liter/sec.
9-6
1.3.3 SIP (Standard Pressure and Temerature)
Defined asl 0C and 760 Torr.
1.3.4 K Factor
A factor used to convert from a tracer gas leak rate obtained under the specified test conditions
to an equivalent air leak rate at those specific test conditions. For parposes of this
specification:
K ol ecular weight of helium- Molecular weight of air
K - 0.372
1.3.5 Sealed Product
A product which is capable of maintaining, or of being sealed by special fixtures to maintain, an
internal pressure or vacuum.
1.3.6 Tracer Gas
A gas that is used to measure the leak rate of the product being tested.
1.3.7 Leak Rate
The quantity 4i gas flowing in unit time into, or out of, the product under test, reduced to
units of volume at standard twiperature and pressure.
/
1.3.7.1 Tracer Gas Leak Rate
The leak rate tests results, calculated without application of a K factor, from the leak detector
readings.
1.3.7.2 Total Gas Leak Rate
An estimate of the product leak rate, obtained by multiplying the tracer gas leak rate by the
specified K factor.
1.3.7.3 Maximum Permissible Leak Rate
The xlmimum total gas leak rate limit allowable for product acceptance.
1.3.8 Known Leak
A calibrated device from which tracer gas is .Nnitted at a krown rate.
2.0 DOCW.NTS
The following documents, of the exact issue shown, forn a pa.-t of this specification to the
extent specified herein.
Specifications:
Federal:
BB-H-1168b Helium, Technical9-7
3.0 REQUIREMENTS
3.1 Equipmenmt Capability
3.1.1 Leak Detector System
The system shall consist of a mass spectrometer type leak detector together with a suitable test
chamber that will completly enclose the product, auxiliary pressurization equipment, and instru-
mentation necessary for performance of the test under the specified conditions.
3.1.2 Location of Known Leak
The known leak shall be cormected directly to the vacuum system as closely as possible to that
point at which the test item will be connected.
3.1.3 Vacuum Gages
Gages used to ascertain that the pmvssure in the test chamber or fixture is not greater than the.specified maximum sh3ll be suitably located to read the pressure of the test chamber or fixture.
If leak detectors having gages capable of reading pressures indicated below are available, it is
permissibleto use these gages prov:ide:
a. At indicated test pressure of S x 10-5 Torr (mHg) or lower is used for helium.
b. An essentially short direct comection is maintained between the leak detector and the test
chamber or fixture.
Under the present state-of-the-art, calibration of vacuum gages below one micron of mercury is
not quaranteed by Primary Standards and, therefore, is not required.
3.1.4 Attenuation Setting on Leak Detector
The leak detector shall be operated on the most sensitive readable scale.
3.2 Gases
The tracer and fill gas used shall be heliun per BB-4-1168b, Type 1, Grade A.
3.3 Calibration of Known Leak
The known leaks used shall be calibrated by the Primary Standards Laboratory specified in the
contract or purchase order. Calibration shall be performed prior to initial use and at
intervals thereafter in accordance with poli'ies established by the Primary Standards Laboratory.
3.4 Procedure (Leak Rate Measurema~t)
3.4.1 General
a. Readability of the leak detectar output meter shall be checked per 3.1.3.
b. Pressures at which a test is te be performed shall not exceed 5 x 10-4 Torr (nn1g) helium.
9-8
3.4.2Produ t lek Tst
3.4.2.1 The, 1T3 to be leak tested will be recetved disassembled, -and Itfill be necessary to
assemble the unit as the leak test progresses, as follows.,
a. Invert the TB Body and pl ce it on the edge of a clean.work surface, withapproximately one- I,
fourth the interior diameter of the Body extending out from the work surface.
b. Insert the nozzle of the helium tracer gas line into the interior of the TB Body, and flood
the inteiror with helium for a period of at least 20 seconds.
c. Remove the nozzle and, %ith the TB still in the Inverted position, place the TB Body in
position over the TB Lid (the copper seal should be used for this test, the elastomeric 0-
ring should not be in place at this time). While holding the Lid firmly against the Body,
invert both end place them in the up-right position on the work surface.
d. Align the holes in the Lid with the threaded holes in the Body (orientation optional) and
install the 12 socket heat bolts onto the unit. Using a suitable strap-type holding device
and torque wrench, torque the bolts to 50 +5 foot-pounds.
e. Place the assembled TB in the chamber of the mass spectrometer leak tester. Readings shall
be takem frru the leak detector output meter for each test, as follows:
R = Background
R2 = Backgrond & Known Leak
R3 = Product Leak & Background
3.4.2.2 Leak Rate Calculation
Readings R1 1 R2 , and R3 , as obtained in e. above, permit calculation of the total gas leak rate
as follows:
(C)(K)(R 3 - R1)L3 R2 R OL =
R2 -RI
Where:
L = Total gas Leak rate
C = Calibrated value of known leak
X = See 1.3.4
3.4.2.3 Maximum Permissible Leak Rate
The leak rate for the TB shall not exceed 1 x 10-7 atm-cc/sec (air).
9-9
3.4.3 Test Records
The foloiwing info'ination shall be recorded by the testing laboratory. Copies of test records
shall be distributed as specified in the contract or purchase order.
a. Product part number and serial number.
b. -Purchase order or contract number.
c. Malke and model number of leak detector used.
d. Value of known leaks used.
e. Test pressures (internal and external).
d. Tracer gas and concentration used.
g. Fill gas used.
h. Leak detector output readings per 3.4.2.1.e.
i. Calculated total gas leak rate.
j. Date of test.
9-10
PAT.1029 MATERIAL SPECIFICATION. REDWOOD FOR PAT PACKAGE
I.I Scope i . . - :..
This specification defines the requirements for the redood material used in fabrication of the
Plutonium Air Transportable Package, Model PAT-1.
1.2 Definitions
Definitions of terms used in this specification may be found in the document listed in 2.1.
below.
2.0 DOCJMWNTS MD EQUIPMENT
The following documents and equipment form a part of this specification to the extent specified
herein. Okless otherwise specified the latest issues shall be used. In the event of conflict
between documets referenced herein and the contents of this specification, the contents of this
specification sball be a superseding requirement.
2.1 Docmuents
"Standard Specification for Grades of California Redwood Lumber" published by Redwood Inspection (Service, San Francisco, California.
2.2 Equipment
Moisture Register, Model L, made by the Moisture Register Co., Alhambra. California, or
equivalent.
3.0 REQUIRDIJTS
3.1 Material Description
a. The ruate"ial to be used shall be clear, kiln dried redwood, free of defects, as defimed in
Para.- 104 of the "Standard Specifications" document. Knots are not pernissibie.
b. Asseimblies shall be fabricated of one inch or greater nominal thickness.
c. Moisture content shall not exceed 12%. The material shall be protected, as necessary, to
assure this condition.
d. Finger-joined and edge-glued lumber is acceptable.
1. Parts shall be assembled in accordance with the applicable drawings, using polyvinyl
acetate resin emulsion per MMM-A-180, or Resorcinal adhesive per MNM-A-181.
(99-11\
I - - - -
II
2. Glue l ines shall be 0.030 inch maxima..
3.2 Grain Defects
Burls and birdseyes of less than 0.375 inch diameter are acceptable. providing there are notmore than six such defects in a six inch diameter, in a mmximum of 5% of the lumber ina
subassemly.
4.0 QUALMT ASSUJOCE PROVISIONS
4.1 Visual Inspection
All material shall be visually inspected to assure that it meets the requirements of Section 3
of this specification.
4.2 Moisture Content
The material shall be inspected, after final machining and just before application of sealant,
to assure that the moisture content specified in 3.1.c is mot exceeded. Inspection shall be
conducted as follow:
a. Zero the Moisture Register in accordance with the manufacturer's instruction.
b. Twenty percent of the material in each subassembly shall be checked for moisture content,
using the Moisture Register per the manufacturer's instructions. The Register readings
shall not exceed 24.(5.0 PREPARATION FOR DELIVERY
The lumber to be used shall be packaged by the subcontractor so that it will meet the require-
ments of this specification after processing by the fabricator of the PAT. Vapor barriers shall
be used in packaging to maintain the minimum moisture content requirement specified.
-12
I
f PAT-1030 WELDING, CORROSION RESIS(
KANT STEEL. PAT PACKAGE
i) 1.0 GENERAL
1.1 Scope
This specification defines the fabrication and inspection requirements for the welding of
corrosion resistant steel parts used on the Plutonium Air Transportable Package, Model PAT-1.
1.2 Definitions
1.2.1 Welding Terms and Definitions
Welding terms and definitions used in this specification shall be in accordance with AMS A2.0,
except for the following:
Porosity. Voids in the weld metal of approximately spherical shape.
1.2.2 Welding Symbols
Welding symbols. used on the product drawings shall be in accordance with AMS A2.0.
2.0 DOCUMENTS
The following documents form a part ofthis specification to the extent stated herein. Unless
otherwise specified use latest issues.
.
AWh A 2.0-58
AMS A 3.0-51
MIL-1-6866B
Welding Symbols
Welding Terms and Definitions
Inspection, Penetrant Method of
3.0 REQUIREMENTS
3.1 Welding Process
Welding shall be done by any of the arc welding processes, using manual, semiautomatic, or
automatic techniques.
3.2 Weld Preparation
Loose scale, slag, rust, grease, oil, and other foreign matter shall be removed from surfaces
to be welded. Beveling and weld preparation may be done by flux-oxygen cutting, provided crack-
ing does not occur in the metal and provided at least 0.12b inch of metal is removed from all cut
edges by mechanical means, grinding, etc.
9-13
3.3 Weld Defects
Imperfections that exceed the limits specified in Table 1 shall be considered defects and are
--nacceptable, except as specified below.
,.4 Repair of Defects
Repair of defects is permissible if the required weldmnent, the repair weld itself, and the
adjacent parent metal mect the requirements of the original weldment. A repaired welment
shall be reinspected in the same manner as the original weldnent.
4.0 QUALITY ASSURANCE PROVISIONS
4.1 General
4.1.1 Responsibility for Inspection
The Supplier performing the welding shall be responsible for ti.e performance of all tests and
inspections specified herein.
4.1.2 Inspection Records
The Supplier shall maintain records of all inspections performed-per 4.2. Copies shall be
distributed as specified in the contract or purchase order.
A.l.3 Inspection Sequence
Weldments may be inspected at any time after the weld preparation and cleaning requirements of
'ection 3 have been met.
4.1-4 Rejected Units
Parts that fail to meet any of the requirepents of this specification shall be rejected. Repair
as defined in 3.4 is pernissible.
4.2 Product Inspection and Testing
The following inspections shall be performed on the weldments on parts of the PAT Package on
which welding is performed.
4.2.1 100%' Visual Inspection
Parts shall be visually inspected 100% for the imperfections defined in Table 1. No magnifica-
tion is required.
9-14
Table 1
LIMITS fF IMWERFECTIONS IN ACCEPTABLE WELDS
IM!ERFECTION
Cracks in weld bead
Cracks in parent metal
Crater cracks
Incomplete fusion andinadequate jointpenetration
Porosity (Internal)
Inclusions (Internal)
Undercut
Overlap
Concavity
LIMIT
Unacceptable
Unacceptable
Unacceptable
The aggregate length of the imperfections shall notexceed 1-1/ZT in a weld length of 6T. and the lengthof any individual imperfection shall not exceed 1/2T.If the weld length is less thar. 6T the aggregate lengthof the imperfections shall not exceed 1/4 the weldlength, and the length of any individual imperfectionshall not exceed 1/12 the weld length. (See Note)
Not applicable
Not applicable
Unacceptable
Unacceptabl E
Unacceptable in bitt welds. In fillet welds, actualthroat shall be not less than the theoretical throatfor specified weld size. (
Convexity of butt weldson either side
Size of fillet welds
Note: (T) is the specified weld size.
Weld SizeUp to 0.125 inch0.125 to 0.500 inch0.500 inch and larger
Specified weld size (length
Maximum Rein-forcement Height0.050 inch251 of weld size0.125 inch
of legs) +50%. -0.
C9-15
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1> T-YPICAL 5-rACK L%' MINIML*A 2x8,mAXi~.4m2 x 10 STO(X.. MAC AlIPE ALL ZURFAC-E.6.
6. COAT EXPOSED SURFACES PER PAT-I,'5AR
I1~1
(
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4
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R tATVhIAL: C.DR-K, 9-A rAT.1
3 BOND2
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0>~ TTPICAL srAcA UP wiMIJud 9 'a x to STOCK. I4MN"J ALL SvaAFCr S.
a.k4&T~AL- CLA.EA VVXNIM K-NcDRC PER PA7..o1q
SbOND PERPT *I.SAM $BCI-0- 41.1.6.1
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TYPICAL STACK-UP MINIMUM 2X4(STC.MACHI1NE Al. SURFACES.
2. AATERiAL: CL-EAR REDWOOD, KILDDRIED PER PA'ATr-019
3. OMD ýPER PAT- 1, SAR 5trCTIC'J 9A.&S~ us~fi
4. 5TAGGER ALL- GLUE JOINTS.
2 I
a 7 6 5 4S 7 S 5 I
U--> C-OAT IND)CATED SUP!PAT1-iAR SECTIONJ I.-L
Z fvTF-xrLGL CLEARDvaED PER PAXT-i0tý
a BOND PrRP7-1, 5ARUSILKA mmm~-A-IeSO
4. ST-e.BGEFP AL" GL:
7-11=C A-L. 5TACKSTrOCIL. MACHINIE Al
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8 7 6 5 47 6 5 4
I4 I
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)AT INDICATED SURFAM5. PER
A.I
A-MFMAL CLEAR )PEIDvOM KLNJ:ZED PER vAT-iolV
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YPIOCAl- STACK LP2. MNV%MLA'A 2,s &PTOCiL. MACRIFLE ALL12WEFAEPMS.
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ISIZE D I IDWG NO joiiISCALE 112 ISHET7IO
A 3 2 I
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NOTES
U2Z- TYPICAL 5TACK UP. MINIMUM 2Xr STOCK.MACHINE ALL SURFACES.
a- MATERiAL: CLEAR F tECNV^VOD, KILNDRIED PER PA.- IO?1r-
3. BOND PER PAT-I,SAR 5FCTIOIJ C...3.5 USIMIJMM M-A-180.
SGRA I N __ _
//,-ORIENTATION
t I
2 .9 4 E N
NOTES
rACK mP MINIMUM 2KG ro Tcx(..ý,LL suF Vý'CES.
cLEAR DV~VDOD, KILNR F=AZ. - 102 9
IPAT-I, SAR 5raCTiOK 9.Z.3.5 USIIJG
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SPLUG, WOOD, Fo IXEDjSZE c I DWG NO 1012
ISCALE 112 I SHEET IOF I
aD 7 I 6 5 4
1. COAT EXPOSED .5Ji*
P- -r'A C-bMM- -COtsA
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2. MATERIAL : CLEAR REDWOOD,DRIED PER PA"T-Io09
KILN
3. BOND PERPAr-I,SAR SECTOtN 9).2.3.3ustimG MMM -A - 10
R-TYVI:CAL STACK UP. MINIMuMSTOCK. MACHINE ALL SURFACES.
2,8
ý->PART KiJmPlrýE MARKIýJG FPER. PAT-I 5AR '.-CTiov.a 4
C.77 11, 17 0 z §'6.11 B-
END GRAIN
KOTES
SURFACES PER PAT-I1
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I SAR 5ECTION 9.2. 3. 3
K UP. MINIMUM4IE ALL SURFACES.
RKI'IJG PER PAT- ,5A.R 151EGTIQK 9
2-43"d.40
K
(
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SHOWN EQUALLY SPACED
VVlTNI• .o0
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L ALvNA,. HONEYCOMB 7 5056 ALLOY, .OO7ZFOIL.• 1/8"CELL, 8. I L/CU FT, )800
.L.. I-IHE.CEL., OR APPROVEDEQUI VALEkST.
[ZMETAL BOND FILM ADHESIVE FM?123-5(BLOOMINGDALE DEPARTMENT OF AMERICANCYANAMID CO., HAYRE de_ G&RACE,MD.) ORAPPROVED EQUIVALENT.
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ISHEET I OF Ii m . ,
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kL ~AiG.b-4T STITC141GIit SHALL. COI,1OR M To
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8. CHAMFER PLUS INCOMPLETE THREEADSNOT TO EXCEED 2 PITCHES.
9. DIMENSIONS IN INCHES.
1. MATERIAL: A-2-66 PER AMS 5731.
2. HEAT TREAT: 160 KSI BOLT TENSILE STRENGTH,ýc 36 MIN.
3. FINISH: SILVER PLATE PER AMS 2410.
4. CONCENTRICITY: HEAD DIAMETER TO BE CON-CENTRIC TO THE THREAD PITC(J-i DJA, WITHIN.010 T.IR. SOCKET TO BE CONCENTRIC WITHHEAD DIA. WITHIN .01_T.I.R.
5. SURFACE TEXTURE: PER ANSI 846.). UNLESSOTHERWISE SPECIFIED, THE SURFACE TEXTURESHALL NOT EXCEED 125 MICROINCHES,
G. MECHANICAL PROPERTIES: RATED ULTIMATETENSILE 5TRENGTH IS 30,000 LBS. DOUBLESHEAR MEQUIREMENTS ARE NOT APPLICABLE,TENSILE TESTING SHALL BE PERFORMED USINGA TENSILE BAR E4GAGED AS FULLY AS POSS-IBLE. FATIGUE TESTING IS WAIVED.
.2+7 MEASURED AT BOTTOMOF HEX SOCKET FACET
..50(
EDGE!
I~%5AI visctiPT10*4 akqY6
(NMS 5731.
)LT TENSILE STRENGTH,
R AMS 2410.
AMETER TO BE CON-) PITCH DIA. WITHINE CONCENTRIC WITHr.I.R.
ANSI B46.I. UNLESSHE SURFACE TEXTUREIICROINCHES,
: RATED ULTIMATE0000 LBS. DOUBLE.E NOT APPLICABLE.ýE PERFORMED USIN3) AS FULLY AS POSS-IS WAIVED.
X ,45^-APPRO,.
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BOLT, SOCKET, HEADSPECIAL I500 -20
I SIZ C IDWG NO 10 18ISCALE 112 I SHEET I OFI
NOTES
I. PART WiUIER MAR KWIIG, pR PATr- I:,AR ZC-CTIOb. 9
2. MATERIAL: COPPER, HALF HARD 110 ETPPER QQ-C-576.
3. 6y ALL OVER.
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ACKNOWLEDGEMENTS
Mr. John A. Andersen, project engineer for the PAT program
at Sandia Laboratories, worked tirelessly and contributed
greatly to the successful design and testing of the package.
9-71
