ML24284A264
| ML24284A264 | |
| Person / Time | |
|---|---|
| Site: | 07109405 |
| Issue date: | 10/03/2024 |
| From: | QSA Global |
| To: | Office of Nuclear Material Safety and Safeguards |
| References | |
| Download: ML24284A264 (1) | |
Text
Attachment 3: Non-Proprietary Documents List of Affected Pages SAR Revision 0 Cover page and indices pages pages 1-1, 1-2 & 1-8 pages 1-16 thfu 1-20 including Drawing R1100-USER Rev A pages 2-1 and-2-4 pages 2-33 and 2-34 pages 2-425 including appendix 2.12.10 pages 2-431 including appendix 2.12.11 pages 4-1 and 4-2 pages 5-1 thru 5-6 Sections 6 thru 9 of the SAR in total
Revision 0, October 2024 Safety Analysis Report for the Model 1100 Transport Package List of Affected Pages Original Submission.
Safety Analysis Report QSAGLOBAL Model 1100 Type B(U)
Transport Package October 2024 Revision 0
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts Contents October 2024 - Revision 0 Page i SECTION 1 - GENERAL INFORMATION................................................................................................................. 1-1
1.1 INTRODUCTION
.......................................................................................................................................... 1-1 1.2 PACKAGE DESCRIPTION.............................................................................................................................. 1-1 1.2.1 Packaging.......................................................................................................................................... 1-3 1.2.2 Contents............................................................................................................................................. 1-5 1.2.3 Special Requirements for Plutonium.................................................................................................. 1-6
- 1. 2.4 Operational Features.......................................................................................................................... 1-6 1.3 APPENDIX................................................................................................................................................. 1-7 1.3.1 Drawing R1100................................................................................................................................... 1-9 1.3.2 Drawing R1100-USER..................................................................................................................... 1-16 SECTION 2 - STRUCTURAL EVALUATION............................................................................................................ 2-1
2.1 DESCRIPTION
OF STRUCTURAL DESIGN........................................................................................................ 2-1 2.1.1 Discussion.......................................................................................................................................... 2-1
- 2. 1. 2 Design Criteria.................................................................................................................................... 2-1
- 2. 1. 3 Weight and Centers of Gravity........................................................................................................... 2-1
- 2. 1. 4 Identification of Codes and Standards for Package Design............................................................... 2-1 2.2 MATERIALS............................................................................................................................................... 2-2 2.2.1 Material Properties and Specifications............................................................................................... 2-2 2.2.2 Chemical, Galvanic or Other Reactions............................................................................................. 2-3 2.2.3 Effects of Radiation on Materials........................................................................................................ 2-3 2.3 FABRICATION AND EXAMINATION.................................................................................................................. 2-3 2.3.1 Fabrication......................................................................................................................................... 2-3 2.3.2 Examination....................................................................................................................................... 2-4 2.4 GENERAL REQUIREMENTS FOR A LL PACKAGES............................................................................................. 2-4 2.4.1 Minimum Package Size...................................................................................................................... 2-4 2.4.2 Tamper-Indicating Feature................................................................................................................. 2-4 2.4.3 Positive Closure................................................................................................................................. 2-4 2.5 LIFTING AND TIEDOWN STANDARDS FOR A LL PACKAGES................................................................................ 2-5
- 2. 5. 1 Lifting Devices.................................................................................................................................... 2-5 2.5.2 Tie-Down Devices.............................................................................................................................. 2-5 2.6 NORMAL CONDITIONS OF TRANSPORT (NCT)............................................................................................... 2-5
- 2. 6. 1 Heat.................................................................................................................................................... 2-5 2.6.2 Cold.................................................................................................................................................... 2-7 2.6.3 Reduced External Pressure............................................................................................................... 2-7
- 2. 6. 4 Increased External Pressure.............................................................................................................. 2-8 2.6.5 Vibration............................................................................................................................................. 2-8 2.6.6 Water Spray....................................................................................................................................... 2-8
- 2. 6. 7 Free Drop........................................................................................................................................... 2-8
- 2. 6. 8 Corner Drop..................................................................................................................................... 2-13
- 2. 6. 9 Compression or Stacking................................................................................................................. 2-13
- 2. 6. 1 O Penetration.................................................................................................................................. 2-14
- 2. 7 H YPOTHETICAL ACCIDENT CONDITIONS (HAC)........................................................................................... 2-19
- 2. 7. 1 Free Drop......................................................................................................................................... 2-20
- 2. 7.2 Crush................................................................................................................................................ 2-23
- 2. 7. 3 Puncture........................................................................................................................................... 2-23
- 2. 7.4 Thermal............................................................................................................................................ 2-27
- 2. 7.5 Immersion - Fissile Material............................................................................................................. 2-31
- 2. 7.6 Immersion -All Packages................................................................................................................. 2-31
- 2. 7. 7 Deep Water Immersion Test (for Type B Packages Containing More than 105 A2).......................... 2-31
- 2. 7.8 Summary of Damage....................................................................................................................... 2-31 2.8 ACCIDENT CONDITIONS FOR AIR TRANSPORT OF PLUTONIUM OR PACKAGES VV1TH LARGE QUANTITIES OF RADIOACTIVITY...................................................................................................................................................... 2-33 2.9 ACCIDENT CONDITIONS FOR FISSILE MATERIAL PACKAGES FOR AIR TRANSPORT........................................... 2-33
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page ii 2.10 SPECIAL FORM.......................... ****** ***** *............................................................................... ******............. 2-33 2.11 FUEL Roos............................................................................................................................................. 2-33 2.12 APPENDICES........................................................................................................................................... 2-34 2.12.1 Test Plan 237 Rev 1 (4/19/2024)................................................................................................. 2-35 2.12.2 Test Plan 237 Report Rev O minus Appendices B & D (Sep 2024)............................................. 2-84 2.12.3 Test Plan 79 and Report Vultafoam Compression Test minus Attachment B (Oct 1998).......... 2-193 2.12.4 Test Plan 80 Report (minus manufacturing records) (June 1999)............................................. 2-196 2.12.5 Test Plan 240 Rev O Model 1100 Handle Wrench Test (May 2024).......................................... 2-210 2.12.6 Test Plan 240 Report Rev 1 Model 1100 Handle Wrench Test minus App B (August 2024).... 2-220 2.12.7 Technical Report 413 Model 1100 Transport Package Tie-Down Analysis (5/20/2024)............ 2-243 2.12.8 Test Plan 239 Rev O ISO/ANSI Performance Testing (June 2024)........................................... 2-283 2.12.9 Test Plan Report 239 Rev O ISO/ANSI Performance Testing minus App B & C (Sep 2024)..... 2-328 2.12.10 USDOT Special Form Certificate USA/0335/S-96 Rev 14......................................................... 2-425 2.12.11 USDOT Special Form Certificate USA/0502/S-96 Rev 13......................................................... 2-431 SECTION 3 -THERMAL EVALUATION................................................................................................................... 3-1
3.1 DESCRIPTION
OF THERMAL DESIGN............................................................................................................. 3-1
- 3. 1. 1 Design Features................................................................................................................................. 3-1
- 3. 1. 2 Decay Heat of Contents..................................................................................................................... 3-1 3.1.3 Summary Tables of Temperatures..................................................................................................... 3-2 3.1.4 Summary Tables of Maximum Pressures........................................................................................... 3-2 3.2 MATERIAL PROPERTIES AND COMPONENT SPECIFICATIONS............................................................................ 3-2 3.2.1 Material Properties............................................................................................................................. 3-2 3.2.2 Component Specifications.................................................................................................................. 3-4 3.3 GENERAL CONSIDERATIONS........................................................................................................................ 3-4 3.3.1 Evaluation by Analysis....................................................................................................................... 3-4 3.3.2 Evaluation by Test.............................................................................................................................. 3-5 3.4 THERMAL EVALUATION UNDER NCT............................................................................................................ 3-5 3.4.1 Heat and Cold.................................................................................................................................... 3-5 3.4.2 Temperatures Resulting in Maximum Thermal Stresses.................................................................. 3-10 3.4.3 Maximum Normal Operating Pressure............................................................................................. 3-10 3.5 THERMAL EVALUATION UNDER HAC.......................................................................................................... 3-11
- 3. 5. 1 Initial Conditions............................................................................................................................... 3-11 3.5.2 Fire Test Conditions......................................................................................................................... 3-11 3.5.3 Maximum Temperatures and Pressure............................................................................................ 3-12 3.5.4 Temperatures Resulting in Maximum Thermal Stresses.................................................................. 3-12 3.5.5 Fuel/Cladding Temperatures for Spent Nuclear Fuel....................................................................... 3-13 3.5.6 Accident Conditions for Fissile Material Packages for Air Transport................................................ 3-13 3.6 APPENDIX............................................................................................................................................... 3-13 SECTION 4-CONTAINMENT.................................................................................................................................. 4-1
4.1 DESCRIPTION
OF THE C ONTAINMENT SYSTEM............................................................................................... 4-1 4.1.1 4.2 Special Requirements for Damaged Spent Nuclear Fuel................................................................... 4-1 CONTAINMENT UNDER NCT........................................................................................................................ 4-1 4.3 CONTAINMENT UNDER HYPOTHETICAL ACCIDENT CONDITION......................................................................... 4-1 4.4 LEAKAGE RATE TESTS FOR T YPE 8 PACKAGES............................................................................................. 4-1 4.5 APPENDIX................................................................................................................................................. 4-2 SECTION 5 - SHIELDING EVALUATION................................................................................................................. 5-1
5.1 DESCRIPTION
OF SHIELDING DESIGN........................................................................................................... 5-1
- 5. 1. 1 Design Features................................................................................................................................. 5-1 5.1.2 Summary Table of Maximum Radiation Levels.................................................................................. 5-1 5.2 SOURCE SPECIFICATION............................................................................................................................. 5-3
- 5. 2. 1 Gamma Source.................................................................................................................................. 5-3 5.2.2 Neutron Source.................................................................................................................................. 5-3 5.3 SHIELDING MODEL..................................................................................................................................... 5-3 5.3.1 Configuration of Source and Shielding............................................................................................... 5-3 5.3.2 Material Properties............................................................................................................................. 5-4 5.4 SHIELDING EVALUATION.............................................................................................................................. 5-4
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page iii 5.4.1 Methods............................................................................................................................................. 5-4 5.4.2 Input and Output Data........................................................................................................................ 5-4 5.4.3 Flux-to-Dose-Rate Conversion........................................................................................................... 5-6 5.4.4 External Radiation Levels................................................................................................................... 5-6 5.5 APPENDIX................................................................................................................................................. 5-6 5.5.1 Shielding Profile Model 1100 SN TP239-18 with Se-75 (9/12/2024).................................................. 5-7 5.5.2 QSA R&D Report 05102 Issue 1 Surface Dose Rate Correction Factors (February 2005)............... 5-11 SECTION 6 - CRITICALITY EVALUATION.............................................................................................................. 6-1 SECTION 7-PACKAGE OPERATIONS.................................................................................................................. 7-1 7.1 PACKAGE LOADING.................................................................................................................................... 7-1
- 7. 1.1 Preparation for Loading...................................................................................................................... 7-1 7.1.2 Loading of Contents........................................................................................................................... 7-3 7.1.2.2 Preparation for Transport........................................................................................................... 7-3 7.2 PACKAGE U NLOADING................................................................................................................................ 7-3 7.2.1 Receipt of Package from Carrier........................................................................................................ 7-3 7.2.2 Removal of Contents.......................................................................................................................... 7-4 7.3 PREPARATION OF EMPTY PACKAGE FOR T RANSPORT.................................................................................... 7-4 7.4 OTHER O PERATIONS.................................................................................................................................. 7-5 7.4.2 Package Transportation By Consignor............................................................................................... 7-5 7.4.3 Emergency Response........................................................................................................................ 7-5 7.5 APPENDIX................................................................................................................................................. 7-6 SECTION 8 -ACCEPTANCE TESTS AND MAINTENANCE PROGRAM................................................................ 8-1 8.1 ACCEPTANCE TEST.................................................................................................................................... 8-1 8.1.1 Visual Inspections and Measurements............................................................................................... 8-1 8.1.2 Weld Examinations............................................................................................................................. 8-1 8.1.3 Structural and Pressure Tests............................................................................................................ 8-1
- 8. 1.4 Leakage Tests.................................................................................................................................... 8-2
- 8. 1. 5 Component and Material Tests.......................................................................................................... 8-2 8.1. 6 Shielding Tests................................................................................................................................... 8-2
- 8. 1. 7 Thermal Tests.................................................................................................................................... 8-2 8.1.8 Miscellaneous Tests........................................................................................................................... 8-2 8.2 MAINTENANCE PROGRAM........................................................................................................................... 8-3 8.2.1 Structural and Pressure Tests............................................................................................................ 8-3 8.2.2 Leakage Tests.................................................................................................................................... 8-3 8.2.3 Component and Material Tests.......................................................................................................... 8-3 8.2.4 Thermal Tests.................................................................................................................................... 8-3 8.2.5 Miscellaneous Tests........................................................................................................................... 8-3 8.3 APPENDIX................................................................................................................................................. 8-4 SECTION 9 - QUALITY ASSURANCE..................................................................................................................... 9-1 9.1 U.S. Q UALITY ASSURANCE PROGRAM REQUIREMENTS.................................................................................. 9-1 9.2 CANADA Q UALITY A SSURANCE PROGRAM REQUIREMENTS............................................................................ 9-1 List of Tables TABLE 1.2.A: MODEL 1100 PACKAGE INFORMATION.......................................................................................... 1-2 TABLE 1.2.8: ISOTOPE INFORMATION PERMITTED IN THE MODEL 1100 PACKAGE............................................... 1-6 TABLE 2.2.A: MECHANICAL PROPERTIES OF MATERIALS IMPORTANT TO PACKAGE INTEGRITY............................ 2-2 TABLE 2.6.A:
SUMMARY
TEMPERATURES NORMAL TRANSPORT........................................................................ 2-6 TABLE 2.7.A:
SUMMARY
TABLE OF TEMPERATURES.......................................................................................... 2-28 TABLE 2.7.8:
SUMMARY
TABLE OF MAXIMUM PRESSURES................................................................................ 2-28 TABLE 2.7.C:
SUMMARY
OF DAMAGES DURING TESTING.................................................................................... 2-32 TABLE 2.10.A: TYPICAL SPECIAL FORM CAPSULE TRANSPORT INFORMATION.................................................... 2-33
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision O Page iv TABLE 3.1.A:
SUMMARY
TABLE OF TEMPERA TURES.......................................................................................... 3-2 TABLE 3.1.B:
SUMMARY
TABLE OF MAXIMUM PRESSURES................................................................................ 3-2 TABLE 3.2.A: MATERIALS OF COMPONENTS IMPORTANT TO SAFETY................................................................... 3-2 TABLE 3.2.B: THERMAL PROPERTIES OF PRINCIPAL TRANSPORT PACKAGE MATERIALS..................................... 3-3 TABLE 3.4.A: INSOLATION DATA....................................................................................................................... 3-7 TABLE 5.1.A: MODEL 1100 SN TP237(C)
SUMMARY
TABLE OF EXTERNAL RADIATION LEVELS EXTRAPOLATED TO CAPACITY OF 150 Cl IR-192 (NON-EXCLUSIVE USE) AND HYPOTHETICAL ACCIDENT TRANSPORT CONDITION TESTING3................................................................................................................................ 5-1 TABLE 5.1.B: MODEL 1100 SN TP237(C)
SUMMARY
TABLE OF EXTERNAL RADIATION LEVELS EXTRAPOLATED TO CAPACITY OF 150 Cl IR-192 (EXCLUSIVE USE)1.................................................................................... 5-2 TABLE 5.1.C: MODEL 1100
SUMMARY
TABLE OF EXTERNAL RADIATION LEVELS EXTRAPOLATED TO CAPACITY OF 150 Cl SE-75 (NON-EXCLUSIVE USE)1*2............................................................................................... 5-3 TABLE 7.1.A: ISOTOPES PERMITTED IN THE MODEL 1100................................................................................... 7-1 List of Figures FIGURE 1.2.A - MODEL 1100 PACKAGE WITH OPTIONAL JACKET....................................................................... 1-1 FIGURE 1.2.B - MODEL 1100 PACKAGE WITHOUT OPTIONAL JACKET................................................................. 1-2 FIGURE 1.2.C - MODEL 1100 TRANSPORT PACKAGE......................................................................................... 1-3 FIGURE 1.3.A - SKETCH OF MODEL 1100 WITHOUT OPTIONAL JACKET PREPARED FOR TRANSPORT................... 1-8 FIGURE 1.3.B-SKETCH OF MODEL 1100 WITH OPTIONAL JACKET PREPARED FOR TRANSPORT.......................... 1-8 FIGURE 2.6.A - MODEL 1100 TEST SPECIMEN CONFIGURATION FOR RECTANGULAR ACCESS HOLE.................... 2-10 FIGURE 2.6.B-MODEL 1100 REDESIGNED CONFIGURATION FOR CIRCULAR ACCESS HOLE............................... 2-10 FIGURE 2.6.C-MODEL 1100, 1.2 M DROP TEST ORIENTATION-END DROP (REAR PLATE)................................ 2-11 FIGURE 2.6.0-TEST UNIT TP239-1C REAR PLATE DAMAGE AFTER 1.2 MEND DROP........................................ 2-11 FIGURE 2.6.E - TEST UNIT TP239-1 C BOTTOM SHELL DAMAGE AFTER 1.2 M END DROP................................... 2-11 FIGURE 2.6.F - MODEL 1100, 1.2 M DROP TEST ORIENTATION -OBLIQUE DROP (SLAPDOWN)........................... 2-12 FIGURE 2.6.G - TEST UNIT TP239-1 A FRONT SHELL IMPACT DAMAGE AFTER 1.2 M SLAP DOWN DROP............. 2-13 FIGURE 2.6.H - TEST UNIT TP239-1 C REAR SHELL IMPACT DAMAGE AFTER 1.2 M SLAP DOWN DROP............... 2-13 FIGURE 2.6.1-MODEL 1100 COMPRESSION TEST ORIENTATION......................................................................... 2-14 FIGURE 2.6.J-SPECIMEN TP239-1C ORIENTATION FOR THE PENETRATION TEST-LOCK................................... 2-14 FIGURE 2.6.K-SPECIMEN TP237(C) ORIENTATIONS FOR THE PENETRATION TEST-LOCK SLIDE....................... 2-15 FIGURE 2.6.L-TEST UNIT TP237(C) LOCK SLIDE MOVEMENT AFTER PENETRATION TEST................................. 2-16 FIGURE 2.6.M-TEST UNIT TP237(C) LOCK SLIDE IMPACT MARK AFTER PENETRATION TEST............................ 2-16 FIGURE 2.6.N - TEST UNIT TP237(C) SELECTOR RING DAMAGE FROM LOCK SLIDE AFTER PENETRATION TEST.. 2-16 FIGURE 2.6.0-SHELL CUTOUT MODIFICATION PER TMI 1194......................................................................... 2-16 FIGURE 2.6.P-TEST UNIT TP237(C) SHELL AND LOCK SLIDE DAMAGE AFTER PENETRATION TEST................... 2-17 FIGURE 2.6.Q-TEST UNIT TP237(C) SELECTOR RING DAMAGE AFTER PENETRATION TEST.............................. 2 -17 FIGURE 2.6.R - SHELL CUT OUT MODIFICATION PER TMI 1200......................................................................... 2-18 FIGURE 2.6.S - TEST UNIT TP239-1 B SHELL DEFORMATION AFTER PENETRATION TEST.................................... 2-18 FIGURE 2.6.T - TEST UNIT TP239-18 SHELL DAMAGE AFTER PENETRATION TEST OUST COVER REMOVED........ 2-18 FIGURE 2.6.U - TEST UNIT TP239-18 END VIEW OF DAMAGE AFTER PENETRATION TEST.................................. 2-18 FIGURE 2.7.A-MODEL 1100, 9 M DROP TEST ORIENTATION - END DROP......................................................... 2-20 FIGURE 2.7.8-TEST UNIT TP239-1C SHELL AND LOCK COVER DAMAGE AFTER 9M DROP TEST VIEW 1............ 2-21 FIGURE 2.7.C-TEST UNIT TP239-1C SHELL AND LOCK COVER DAMAGE AFTER 9M DROP TEST VIEW 2............ 2-21 FIGURE 2.7.D - MODEL 1100, 9 M DROP TEST ORIENTATION-LEFT SIDE DROP................................................ 2-22 FIGURE 2.7.E-TEST UNIT TP239-1A REAR END SHELL DAMAGE AFTER 9M DROP TEST................................... 2-23 FIGURE 2. 7.F - TEST UNIT TP239-1 A FRONT SHELL DAMAGE AFTER 9M DROP TEST......................................... 2-23 FIGURE 2.7.G - MODEL 1100 TP239-1C PUNCTURE DROP ORIENTATION - END DROP...................................... 2-24 FIGURE 2.7.H - MODEL 1100 TP239-1C PUNCTURE DROP REAR PLATE DAMAGE-END DROP......................... 2-25 FIGURE 2.7.1-MODEL 1100 TP239-1A PUNCTURE DROP ORIENTATION-SLAP DOWN...................................... 2-26
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Pagev FIGURE 2.7.J-MODEL 1100 TP239-1A PUNCTURE DROP REAR PLATE DAMAGE-SLAP DOWN....................... 2-26 FIGURE 3.4.A: MODEL OF CYLINDRICAL PACKAGE FOR HEAT ANALYSIS............................................................. 3-5 FIGURE 5.4.A. - SAMPLE SURFACE CORRECTION FACTOR DISTANCE CRITERIA................................................... 3-5
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 1-1 Section 1 - GENERAL INFORMATION 1.1 Introduction The Model 1100 is designed as an industrial radiography exposure device and transport package for Type B quantities of special form radioactive material. It conforms to the Type B(U) criteria for packaging in accordance 10 CFR 71, 49 CFR 173, IAEA Regulations for the Safe Transport of Radioactive Material No. TS-R-1 (2009 Edition) and SSR-6 Editions 2012
& 2018, and Canadian Nuclear Safety Commission (CNSC) PTNS Regulations SOR/2015-145. This submission is formatted in accordance with NUREG-1886 "Joint Canada - United States Guide for Approval of Type B(U) and Fissile Material Transportation Packages" dated March 2009.
1.2 Package Description This transport package is constructed in accordance with drawing R 1100 contained in Appendix 1.3.1. Note that this drawing is provided as a separate attachment to the application cover letter as it is considered proprietary and intended for use only by regulators in the review of the package design's compliance. An additional drawing, R 1100-USER, contained in Appendix 1.3.2 is intended for use by users of the package for Type B(U) shipments. Drawing R 1100-USER is not proprietary and is intended for reference on the Type B certificate.
Table 1.2.A lists the maximum activity capacities for the Model 1100. The Model 1100 package allows for the use of an optional jacket which facilitates its use as a radiography device and transport package. The jacket does not impair the package's ability to meet the Type B requirements as described in this Safety Analysis Report (SAR).
Figure 1.2.A-Model 1100 package with Optional Jacket
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 1-2
/
Figure 1.2.B-Model 1100 package without Optional Jacket The Model 1100 package without the jacket measures approximately 5 inches (127 mm) in diameter by 13 inches (330 mm) long. The package with the jacket measures approximately 13 inches (330 mm) long by 6.6 inches (168 mm) wide by 81/2 inches (226 mm) tall.
The general package information is shown in Table 1.2.A. The maximum weight of the package contents will not exceed 0.04 lbs (18 grams) as special form sources attached to a source assembly.
Table 1.2.A: Model 1100 Package Information Maximum Maximum Maximum Nuclide Form Maximum Content Maximum Weight Weight Capacity1 Weight2 DU Weight without with Jacket Jacket lr-192 Special Form 5.55 TBq Sources3 150 Ci 0.04 lbs 34Ibs 44 lbs 48Ibs Se-75 Special Form 5.55 TBq (18 grams)
(15.4 kg)
(20 kg)
(22 kg)
Sources3 (150 Ci) 1Maximum activity for lr-192 is defined as output Curies as required in ANSI N432 and 10 CFR 34.20 and in line with TS-R-1, SSR-6 and Rulemaking by the USNRC and USDOT published in the Federal Register on 26 January 2004.
2Maximum content weight includes the mass of the radioactive material and the source capsule handling assembly that can be transported in the package.
3Special form is defined in 10 CFR 71, 49 CFR 173 and IAEA TS-R-1 and SSR-6.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 1-8 Figure 1.3.A-Sketch of Model 1100 without optional Jacket Prepared for Transport Figure 1.3.B - Sketch of Model 1100 with optional Jacket Prepared for Transport
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts 1.3.2 Drawing R 1100-USER October 2024 - Revision 0 Page 1-16
QSA Global, Inc.
Burlington, Massachusetts MODEL NUMBER MAX. WEIGHT PACKAGE WITH JACKET Safety Analysis Report for the Model 1100 Transport Package October 2024 - Revision 0 MAXIMUM WEIGHT WITHOUT JACKET (SEE SHEET 2) 6 MAX. WEIGHT SHIELD Page 1-17 MAXIMUM MAXIMUM MAXIMUM NUMBER PACKAGE AND PACKAGE AND OF SOURCES SOURCE CAPACITY SOURCE CAPACITY (lr l92)
(Se75)
D t--~l ~l00~-~--~48~1=b=s. __
4:....:4..:.:lbc:.=s __
_,__.....:<e34::.cl.:,:cbs'-----'----"-----L-l!.::50~C~U~R!.:IE~S_.J__!.l50
~ C~U:!.!:R,!!1Ee!.S_j LABEL 6 1/2 REF.
'2 C
OPTIONAL POLYURETHANE JACKET A
NOTES:
REAR PLATE LOCKING ASSEMBLY SEE SHEET 2 REAR PLATE 1/4-20 X 1 LG TAMPER PROOF SECURITY SCREWS (4X).
TORQUED TO 70 +/-5IN-LBS.
i-------13 REF. ------i
</JS REF. CONTAINER ASSEMBLY 1100 WITH JACKET
- 1.
ALL COMPONENTS MUST BE REPLACED BY QSA SUPPLIED OR APPROVED COMPONENTS.
THIS INCLUDES FASTENERS.
(4X) 10-24 OR 10-32 BINDING BARREL AND SCREWS - 2 PER END FRONT PLATE ASSEMBLY SEESHEET4 SHOULDER SCREW 1/4-20 TAMPER INDICATING SEAL WIRE (SEE SHEET 4) u+nW onmtWtSE iHClflEO:
FRONT PLATE 114-20 X 1 LG TAMPER PROOF SECURITY SCREWS (3X).
TORQUED TO 70 +/-5IN-LBS WELDED CONTAINER ASSEMBLY All DIMENSIONS ARE INCHES, TOLERANCE +/-1/1 6 QSAGLOBAL DESCRIPTIVE DRAWING 40 NORTH AVE. BURLINGfON, MA 01ao:I TITLE MODEL 1100 SIZE DWG. NO.
B R1100-USER SCALE: 1:4 SHEET OF 4 REV A
D C
D C
I\\
QSA Global, Inc.
Burlington, Massachusetts B
Safety Analysis Report for the Model 1100 Transport Package October 2024 - Revision 0
</>5 REF.
14-----
JJREF.----
1100 WITHOUT JACKET Page 1-18 All DIMENSIONS ARE INCHES, TOLERANCE +/-1/16 QSAGLOBAL DESCRIPTIVE DRAWING 40 NORTH AVE. BURLINGTON, MA 01803 TITLE MODEL 1100 SIZE OWG. NO.
B R1100-USER SCALE: 1:2 SHEET 2
OF 4 REV A
3 D
C
D C
A QSA Global, Inc.
Burlington, Massachusetts LOCK SLIDE "OPEN" GREEN INDICATOR Safety Analysis Report for the Model 1100 Transport Package October 2024 - Revision 0 6
CONTAINER END PLATE GREEN INDICATOR PROTECTIVE COVER Page 1-19 SELECTOR RING REAR PLATE LOCKING ASSEMBLY RED INDICATOR SOURCE UNLOCKED LOCK ASSEMBLY 10-32 X 3/4 LG SCREWS LOCK SLIDE "CLOSED" RED INDICATOR SOURCE LOCKED (SOURCE AND CONTROLS NOT SHOWN)
REAR PLATE LOCKING ASSEMBLY UMLUS OUtllwtSf M'fClfltD:
ALL DIMENSIONS ARE INCHES, TOLERANCE +/-1/ 16 QSAGLOBAL DESCRIPTIVE DRAWING 40 N0R fH AV[. BURLINGTON. MA 0180:1 TITLE MODEL 1100 SIZE DWG. NO.
R1100-USER REV B
SCALE: 1:2 SHEET 3 OF 4 A
7 D
C
D C
A QSA Global, Inc.
Burlington, Massachusetts SHIPPING POSITION TAMPER INDICATING SEAL PORT COVERED &
SHIELDED Safety Analysis Report for the Model 1100 Transport Package October 2024 - Revision 0 6
LOCKED POSITION LIFT KNOB & ROTATE 90 "
PORT SHIELDED FRONT PLATE ASSEMBLY Page 1-20 FITTING ROTOR PORT SHIELD CONNECT POSITION INSERT FITTING AT CONNECT &
ROTATE FITTING TO SECURE PORT SHIELDED UNI.US OTHtlWtSE SJ'ECIAEO; SECURITY SCREW (SEE SHEET1)
EXPOSE POSITION ROTATE FITTING TO OPERA TE POSITION PORT OPEN All DIMENSIONS ARE INCHES, TOLERANCE ! 1/16 QSAGLOBAL DESCRIPTIVE DRAWING 40 N0RHi AVE, BURLINGTON. MA 01803 TITLE MODEL 1100 SIZE DWG. NO.
R1100-USER REV B
SCALE: 1:1.5 SHEET 4 OF 4 A 3
D C
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-1 Section 2 - STRUCTURAL EVALUATION This section identifies and describes the principal structural engineering design of the packaging, components, and systems important to safety and compliance with the performance requirements of 10 CFR Part 71, TS-R-1 and SSR-6.
2.1 Description of Structural Design 2.1.1 Discussion The Model 1100 transport package is described in Section 1.2, "Package Description."
2.1.2 Design Criteria The Model 1100 transport package is designed to comply with the requirements for Type B(U) packaging as prescribed by 10 CFR 71, 49 CFR 173, IAEA TS-R-1 (2009 Edition),
IAEA SSR-6 (2012 & 2018) and CNSC PTNS SOR/2015-145. All design criteria are evaluated by a straightforward application of the appropriate section of these requirements.
In general, the design was based on the Type A and Type B(U) container requirements of 49 CFR, 10 CFR 71, CNSC PTNS and IAEA regulations as identified in Section 1.1. In addition to the transport design criteria, the Model 1100 transport package is designed to meet the performance requirements for industrial radiography exposure devices as specified in 10 CFR 34, Subpart C - Equipment.
The primary design criteria for the Model 1100 transport package ensure radiation safety.
This prevents release of radioactive material, and keeps external package dose levels within specified limits during normal conditions of transport, during normal use, and in the event of a hypothetical accident condition (HAC) defined in the references above.
2.1.3 Weight and Centers of Gravity The transport package weight maximum is 48 lbs (22 kg). The center of gravity of the Model 1100 transport package is approximately 2.5 inches (64 mm) above the bottom of the package without the jacket and 3.0 inches (76 mm) above the bottom of the package with the jacket.
2.1.4 Identification of Codes and Standards for Package Design See Section 2.1.2 relating to design criteria of the package. Any applicable, specific codes or standards related to the finished assemblies for this transport package are specified on the drawings contained in Section 1.3. All component fabrication (including assembly) is controlled under the QSA Global, Inc. Quality Assurance Plan approved by the USN RC and ISO.
All welding under this plan adheres to the standards referenced on the drawings in Section 1.3. All hardware meets the standards referenced on the drawings in Section 1.3. All external fabrication deemed critical to safety is either verified to equivalent in-house standards or dedicated as appropriate for use prior to release as part of this transport package.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-4 for compliance to the drawings provided in Section 1.3 prior to initial use as part of a Model 1100 transport package.
2.3.2 Examination Section 8 describes the acceptance testing and routine maintenance requirements for package components used on the Model 1100 package.
QSA Global, Inc. also reviews and evaluates the impact of regulatory changes to pre-existing product to determine whether any modifications are necessary to maintain compliance with regulatory requirements. When applicable, these reviews also address changes in the state of the product design during periods of storage.
2.4 Gene
~ments o Pac a es 2.4.1 Minimum Package Size The Model 1100 transport package is cylindrically shaped, with the smallest overall dimensions of 5 inches (127 mm) in diameter when transported without the jacket. All other external dimensions of the package with or without the jacket exceed 5 inches (127 mm).
Therefore, it exceeds the minimum package size requirements specified in the referenced regulations.
2.4.2 Tamper-Indicating Feature The front port of the Model 1100 package is designed to require a special tool (i.e., guide tube fitting) to be placed in the front port and rotated before the tungsten shielding in the front plate can be moved away from the end of the source tube. This prevents any inadvertent or unintentional opening of the package during transport. A provision for a tamper indicator seal wire around the knob of the front plate assembly is provided. This seal wire is not readily breakable, therefore if it is broken during transport, it serves as evidence of possible unauthorized access to the contents. Use of this feature meets the tamper indicator requirements for this package. Alternate means of tamper-indicating seals (e.g., seal labels, etc.) can be used on the front or rear plate assemblies so long as their breakage would serve as evidence of possible unauthorized access to the contents.
2.4.3 Positive Closure This package does not involve complex containment systems for source securement. The sources for this package are all special form, welded capsules. The source assembly is held securely in the packages by components of the rear plate assembly. These components prevent the source wire from moving from the secured position in the package when in the locked position.
When the Model 1100 package is prepared for transport, the sleeve and the lock slide maintain the source assembly locked in the secured position preventing source movement.
A cover/cap over the source wire connector prevents access to the source assembly during transport. These features maintain positive closure of the transport package and containment of the radioactive material during transport.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-33 2.8 Accident Conditions for Air Transport of Plutonium or Packages with Lar e Quantities of Radioactivity Not applicable. This package is not used for transport of plutonium or normal form radioactive material. This package is also not used for transport of special form material in quantities~ 3,000 A1.
2.9 Accident Conditions for Fissile Material Packages for Air Transport Not Applicable. This package is not used for transport of Type B quantities of fissile material.
2.10 Special Form The Model 1100 transport package is designed for use with special form source capsules attached to a flexible source assembly. All special form sources transported in the Model 1100 will meet a minimum ANSI/HPS N43.6 and ISO 2919 Classification of 3 for pressure testing.
Typical special form sources transported in this container, and their associated source assemblies, are shown in Table 2.1 0.A.
Table 2.1 O.A: T ical S ecial Form Ca sule Transport Information Special Form Radionuclide Special Form Typical Source Assembly Ca sule Reference Identification 875 Ca sule lr-192 USA/0335/S-96 A424-9 X540/1 Se-75 USA/0502/S-96 A424-25W Based on performance testing, any source capsule that has been tested and achieved special form classification from a Competent Authority, and has achieved an ANSl/1SO1 Pressure Classification rating of 3 can be safely transported in the Model 1100 package so long as the source assembly properly locates the source capsule within the package radiation shielding. Therefore, any compatible source capsule/source assembly meeting these criteria is acceptable for transport in the Model 1100 package without requirement of amendment to the Type B(U) certification.
Details of encapsulation, as well as chemical and physical form of the radioactive material, will comply with specifications approved under U.S. Department of Transportation (or equivalent Competent Authority) special form certifications. Examples of typical special form certifications, including the current approved capsules referenced in Table 2.1 0.A are included in Section 2.12.
2.11 Fuel Rods Not applicable. This package is not used for transport of fuel rods.
1 ANSI/HPS 43.6-2007 (R2013), ISO 2919:2012(E) (or later editions with equivalent classification requirements).
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-34 2.12 Appendices 2.12.1 2.12.2 2.12.3 2.12.4 2.12.5 2.12.6 2.12.7 2.12.8 2.12.9 2.12.10 2.12.11 Test Plan 237 Rev 1 (4/19/2024)
Test Plan 237 Report Rev O minus Appendices 8 & D (Sep 2024)
Test Plan 79 and Report Vultafoam Compression Test minus Attachment 8 (Oct 1998)
Test Plan 80 Report (minus manufacturing records} (June 1999)
Test Plan 240 Rev O Model 1100 Handle Wrench Test (May 2024)
Test Plan 240 Report Rev. O Model 1100 Handle Wrench Test minus App 8 (August 2024)
Technical Report 413 Model 1100 Transport Package Tie-Down Analysis (5/20/2024)
Test Plan 239 Rev O ISO/ANSI Performance Testing (June 2024)
Test Plan Report 239 Rev O minus Appendices 8 & C (Sep 2024)
USDOT Special Form Certificate USA/0335/S-96 Rev 14 USDOT Special Form Certificate USA/0502/S-96 Rev 13
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-425 2.12.10 USDOT Special Form Certificate USA/0335/S-96 Rev 14
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-426 0
East Building, PHH-23 1200 New Jersey Ave, SE Washington, D.C. 20590 U.S. Department IAEA CERTIFICATE OF ca,,pETENT AUTHORITY SPECIAL FO' RADIOACTIVE MATERIALS of Transportation Pipeline and Hazardous Materials Safety Administration CERTIFICATE USA/0335/S-96, REVISION 14 This certifies that the sources described have been demonstrated t o meet the regulatory requirements for special form radioactive material as prescribed in the regulations of the International Atomic Energy Agency1 and the United States of America 2 for the transport o f radioactive material.
- 1. Source Identification -
QSA Global, Inc. Model 875 Series.
2. Source Description -
Cylindrical single or double encapsulations with the outer capsule made of Type 304L stainless steel and tungsten inert gas or laser welded.
Approximate outer dimensions are 6. 35 mm (0. 25 in. ) in diameter and either 19. 05 mm (0. 75 in. ) or 24. 2 mm (0. 954 in. ) in length.
Inner capsules, when present, are made of stainless steel or titanium.
Constructi on of t he outer capsule shall be in accordance with attached QSA Global,
Inc.
Drawing No. R875 OUTER, Rev. E.
Construction of any inner capsule shall be in accordance with attached QSA Global, Inc. Drawing No.
R875 INNER, Rev. C, or QSA Global, Inc. Drawing No. R87527 - 40, Rev.
A.
3. Radioactive Contents -
No more than either : 14.8 TBg (400 Ci) of Iridium-192 as a solid metal ; 8. 14 TBg (220 Ci) of Cobalt - 60 as a solid metal ;
5. 56 TBq (150 Ci) of Selenium-75 in the form of a physically inert and stable metal - selenide compound; 1. 11 TBg
( 30 Ci) of Cesium-137 as encapsulated CsC12 ;
1. 85 TBg (50 Ci) of Thulium-170 as Tm203 ; or 7.4 TBg (200 Ci) of Ytterbium-169 as Yb203.
Only the activity of Ir-192 in special form may be determined from a measurement of the rate of decay or a measurement of the radiation level at a prescribed distance from the source.
1 "Regulations for the Safe Transport of Radioactive Material, 2012 Edition,
No. SSR-6" published by the International Atomic Energy Agency ( IAEA),
Vienna, Austria.
2 Title 49, Code of Federal Regulations, Parts 100- 199, United States of America.
Safety Analysis Report for the Model 11 00 Transport Package QSA Global, Inc.
Burlington, Massachusetts
(-2-)
October 2024 - Revision 0 Page 2-427 CERTIFICATE USA/0335/S-96, REVISION 14
- 4. Management System Activities Records of Management System activities required by Paragraph 306 of the IAEA regulations shall be maintained and made available to the authorized officials for at least three years after the last shipment authorized by this certificate.
Consignors in the United States exporting shipments under this certificate shall satisfy t he requirements of Subpar t H of 10 CFR 71.
- 5. Expiration Date -
This certificate expires on January 31, 2028.
Previous editions which have not reached their expiration date may continue to be used.
This certificate is issued in accordance with paragraph(s) 804 of the IAEA Regulations and Section 173. 476 of Title 49 of the Code of Federal Regulations, in response to the December 2, 2022 petition by QSA Global, Inc., Burlington, MA, and in consideration of other information on file in this Office.
Certified By:
William Schoonover Associate Administrator for Hazardous Materials Safety Revision 14 - Issued to extend the expiration date.
January 13, 2023 (DATE )
Safety Analysis Report for the Model 11 00 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-428
.02 TUNGSTEN INERT GM:.
OR LASER WELDED f
¢B
¢.250
_l_
D __i__j--.094
A----------,~
SECTION A-A NOTES:
- 1.
INTERNAL VOID TO BE 0.010 ml OR GREATER.
- 2.
MATERIAL: 304L STAINLESS STEEL
- 3.
INNER CAVITY Dltv'!ENSIONS MAY VARY. METALLIC SPACERS SPRINGS AND G UARDS, 'M-JICH SECURE AND/OR LOC ATE THE RADIOACTIVE MATERIAL 'MTHIN THE CAPSULE, MAY BE USED, AND SHALL HA VE A MELTING POINT ABOVE 800°C.
- 4.
MNIMUM WALL THICKNESS TO BE 0.02 INCHES.
CAPSULE NO.
A 08 C
D 87501
.954
.190
.150
.522 88702
.750
.1 90
.118
.522 3788 l.tlUSSOTHERWSE 5P£CIFIEO*
All DIMENSIONS NIE INCHES TOLERANCES:
FRACTIONS
- 1/8 x.x
- 0.12 x.xx
- 0.06 x.xxx
- 0.020 QSA GLOBAL_
DESCRIPTIVE DRAWING
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-429 TUNGSTEN INERT GAS OR LASER WELDED
¢.205 NOTES:
- 1. MATERIAL: 304L STAINLESS STEEL.
.045
.090
.051
- 2. INTERNAL VOID VOLUME TO BE 0.010 ml OR GREATER.
- 3. INNER CAVITY DIMENSIONS MAY VARY. METALLIC SPACERS, SPRINGS AND GUARDS WHICH SECURE AND/OR LOCATE THE RADIOACTIVE MATERIAL WITHIN THE CAPSULE MAY BE USED.
- 4. MINIMUM WALL THICKNESS TO BE 0.019.
ERF II:
1739 I.NL£SS 01liCRWISE SPCancD ca<<>ISIONS " K>tCS
'lllWtANCES; nucTIONS 1/4 1/0 x.x :t 0.12 x.xx :t: o.oe x.xxx :t 0.020
¢.191 DESCRIPTIVE DRAWING TITLE 875 SERIES INNER CAPSULE SIZE owe. NO.
R875 INNER REV A
SCALE:
NONE SHEET OF C
- I
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts ERF #
Ql.16 1739 t'o
.19 Ql.20 i
October 2024 - Revision 0 Page 2-430 TUNGSTEN INERT GAS OR LASER WELDED NOTES:
- 1. MATERIAL: 316L STAINLESS STEEL OR EQUIVALENT, OPTIONAL MATERIAL: COMMERCIALLY PURE TITANIUM, GRADE 4.
- 2. INNER CAVITY DIMENSIONS MAY VARY. METALLIC SPACERS, SPRINGS AND GAURDS WHICH SECURE AND/OR LOCATE THE RADIOACTIVE MATERIAL WITHIN THE CAPSULE MAY BE USED.
- 3. MINIMUM WALL THICKNESS TO BE 0.009.
UNUSS 01H[RWIS£ SP£C1fEJ DIMO<SIONSl<IH<MS TOU1WICCS, fRACTlOHS :t: 1/8
)(.)( Z 0.1:Z x.xx :t 0.06 x.xxx :t 0.020 DESCRIPTIVE DRAWING TITL£X540N CAPSULE ASSEMBLY SIZE DWG. NO.
R8 52 7-40 REV A
- SCALE, NONE SHEET 1
OF A
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-431 2.12.11 USDOT Special Form Certificate USA/0502/S-96 Rev 13
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-432 0
East Building, PHH-23 1200 New Jersey Ave, SE Washington, D.C. 20590 U.S. Department IAEA CERTIFICATE OF COMPETENT AUTHORITY SPECIAL FORM RADIOACTIVE MATERIALS of Transportation Pipeline and Hazardous Materials Safety Administration CERTIFICATE USA/0502/S-96, REVISION 13 This certifies that the sources described have been demonstrated to meet the regulatory requirements for special form radioactive material as prescribed in the regulations of the International Atomic Energy Agency1 and the United States of America 2 for t he transport of radioactive material.
- 1. Source Identification QSA Global,
Inc.
Model Nos.
X54 (Manufactured before January 1,
1998),
X540 (Manufactured on or after February 17,
1981),
and X540/1 (Manufactured on or after September 27, 2000).
2. Source Description -
Tungsten inert gas or laser seal wel ded cylindrical single or double encapsulations.
The outer encapsulation is made of titanium or stainless steel and the inner encapsulation, if used, is made o f t i tani um, stainless steel, or aluminum.
Appr oximate exterior dimensions are 5. 15 mm (0. 2 in. )
maxi mum diameter and 15. 15 mm (0. 6 in. ) i n length (Model X54) ; and 5. 16 mm (0. 2 i n. ) in diameter and 7. 65 mm (0. 3 in. ) in length (Models X540 and X540/l).
Construct i on shall be i n accordance wi t h attached Amersham Drawing No. A10639, Issue C (Model X54) or QSA Global Inc. Drawing No. R87527, Rev. H (Models X540 and X540/1).
1 "Regulations for the Safe Transport of Radioactive Material, 2012 Edition, No. SSR-6" published by the International Atomic Energy Agency (IAEA),
Vienna, Aust ria.
' Title 49, Code of Federal Regulations, Parts 100- 199, United States of America.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts
(* 2 *)
October 2024 - Revision 0 Page 2-433 CERTIFICATE USA/0502/S-96, REVISION 13 3. Radioactive Contents - *No more than 17. 0 TBq (459. 5 Ci) of Cobal t-60, in the form of a metal, in the Model X54.
No more than either :
20. 0 TBq (540. 5 Ci) of Cobalt-60 in the form of metal ;
17. 0 TBq (459. 5 Ci) of Iridiurn-192 in the form of metal; or 5. 56 TBq (150. 3 Ci) of Seleniurn-75 in the form of physically inert and stable metal-selenide compound, in the Models X540 and X540/1.
Only the activi t y of Ir-192 in special form may be determined from a measurement of the rate of decay or a measurement of the radiation level at a prescribed distance from the source.
4. Management System Activities Records of Management System acti vities requi red by Paragraph 306 of the IAEA r egulations shall be maintained and made available to the authorized officials for at least three years after the last shi pment authorized by this certificate.
Consignors in the United States exporting shipments under this certificate shall satisfy the requirements of Subpart H of 10 CFR 71.
- 5. Expiration Date -
This certificate expires on January 31, 2028.
Previous editions which have not reached their expiration date may continue to be used.
This certificate is issued in accordance with paragraph(s) 804 of the IAEA Regulations and Section 173. 476 of Title 49 of the Code of Federal Regulations, in response to the December 2, 2022 petition by QSA Global, Inc., Burlington, MA, and in consideration of other information on file in this Office.
Certified By :
William Schoonover Associate Administrator for Hazardous Materials Safety Revision 13 - Issued to extend the expiration date.
January 13, 2023 (DATE)
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 2-434 DRG A10639 NO.
FOR ENGRAVING DETAIL¢ SEE DRAWING A62615 I\\CTUAL SIZE TOUJWICES W.mMI.
...... ~-
APPROVAL
............,~
Item 1
2
¢ 3
USED ON Deacrtptfon Moterfcll BODY STAIN.Sil.
PLUG STAIN.stl.
ACTIVE MATERIAL
¢ 15.15 MAX.
~~ A3
~~ A10639 Drawing No.
No.off A10636 A10638 rTEM.1 SHT l OF SHTS l
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts
, 7.65 MAX I/)
4.08 4.tl"3" NOTES:
5.00 MAX Lili:
- 1. MATERIAL: SEE TABLE ERF #
October 2024 - Revision 0 Page 2-435 X540,540/1 LID SHANK MODEL MATERIAL X540 31 6L STAINLESS STEEL
- 1. INTERNAL VOID TO BE 0.010 ml OR GREATER.
- 2. MATERIAL: SEE TABLE J. INNER CAVITY DIMENSIONS MAY VARY. METALUC SPACERS, SPRINGS AND GUARDS WHICH SECURE ANO/ OR LOCATE THE RADIOACTIVE MATERIAL OR INNER SOURCE CAPSULE WITHIN THE CAPSULE MAY BE USED.
- 4. MINIMUM WALL THICKNESS TO BE 0.22.
- 5. DIMENSIONS ARE IN MIWMETERS UNLESS OTHERWISE SPECtflED OIMENSMJNS IN INCHES T0l.ERAHCES,
~~ \\ 1'~2 x,xx +/- 0.06 XJOCX :I: 0.020 DESCRIPTIVE DRAWING TITLE X540 CAPSULE SERIES SIZE DWG. NO.
R87527 A
SCALE:
NONE SHEET Of REV H
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts Section 4 - CONTAINMENT 4.1 Description of the Containment System October 2024 - Revision 0 Page 4-1 The primary containment system for the package is the welded radioactive source capsule.
Source capsules transported in the Model 1100 package qualify as special form radioactive material under 49 CFR 173 and IAEA TS-R-1 and SSR-6. The special form source capsule is attached to a flexible handling wire assembly. The source assembly is maintained within the shielded configuration of the package by means of the rear plate lock assembly after the source assembly is retracted into the package's shield source tube.
4.1.1 Special Requirements for Damaged Spent Nuclear Fuel Not applicable. This package is not used for the transport of spent nuclear fuel.
4.2 Containment Under NCT As demonstrated in the Test Plan Reports, supported by assessments when applicable (Sections 2.12 & 3.6), and based on information presented in Sections 2 and 3, performance of the NCT testing and assessed NCT criteria will not cause a breach of the source capsules contained in the package. The NCT criteria listed in 10 CFR 71. 71 will result in no loss of transport package containment as prescribed in 10 CFR 71.51 (a)(1 ).
Since the source capsules are the primary containment of the radioactive contents and no release from the source capsules will occur under test conditions, the Model 1100 transport package is determined to meet the requirements of this section.
4.3 Containment Under Hypothetical Accident Condition The HAC outlined in 10 CFR 71. 73 will result in no loss of transport package containment.
This conclusion is based on information presented in Section 2.7 and Section 3.5 which show that the transport package meets the containment requirements of 10 CFR 71.51 (a)(2).
4.4 Leakage Rate Tests or ype B Packages The primary containment for the radioactive material in the Model 1100 Transport package is the radioactive source capsule. All source capsules authorized for Type B transport in the Model 1100 are certified as special form radioactive material under 10 CFR Part 71, 49 CFR Part 173 and IAEA TS-R-1 and SSR-6. After manufacture, and again once every six months thereafter prior to transport, the source capsule is leak tested in accordance with ISO 9978:2020(E) (or more recent editions) to ensure that containment of the source does not allow release of more than 185 Bq (0.005 µCi) of radioactive material. These fabrication and periodic tests ensure that contamination release from the package does not exceed the regulatory limits.
Reference:
ISO 9978:2020(E) - Radiation Protection - Sealed Radioactive Sources -
Leakage Test Methods.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts 4.5 Appendix Not applicable.
October 2024 - Revision 0 Page 4-2
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts Section 5 - SHIELDING EVALUATION 5.1 Description of Shielding Design 5.1.1 Design Features October 2024 - Revision 0 Page 5-1 The principal shielding in the Model 1100 transport package is the DU shield assembly. The shielding is cast as one piece and is essentially enclosed by stainless steel. Dimensional information for the shield is contained in the drawings in Appendix 1.3.
5.1.2 Summary Table of Maximum Radiation Levels Tables 5.1.A and 5.1.B include worst case radiation profile data obtained from the Model 1100 packages that were tested to the NCT and HAC of Transport under Test Plan 237 (see Section 2.1 2). Radiation levels for the Model 1100 package containing Se-75 are included based on a radiation profile of the Model 1100 with Se-75 (see Section 5.5.1 ). In all cases, the Model 11 00 transport package will meet the NCT and HAC transport regulatory limits specified in 10 CFR 71.47 and 71.51.
Table 5.1.A: Model 1100 sn TP237(C) Summary Table of External Radiation Levels Extrapolated to Capacity of 150 Ci lr-192 (Non-Exclusive Use) and Hypothetical Accident Transport Condition Testing3 Normal Conditions of Package Surface mSv/h 1 Meter from Package Surface mSv/h Transport (mrem/hr)2 (mrem/h)2 Radiation Top Side Bottom Top Side Bottom Gamma 1.56 1.68 1.52 0.01 (1.0) 0.011 (1.1) 0.011 (1.1)
(156)
(168)
(152)
Neutron NA NA NA NA NA NA Total 1.56 1.68 1.52 0.01 (1.0) 0.011 (1.1) 0.011(1.1)
(156)
(168)
(152) 10 CFR 71.47(a), §524 2 (200) 2 (200) 2 (200) 0.1 (10)1 0.1 (10) 1 0.1 (10) 1
& 525 of TS-R-1 or
§526 & 527 of SSR-6 Limits Hypothetical Accident Conditions Gamma 0.01 (1.0) 0.011 (1.1) 0.011 (1.1)
Neutron NA NA NA Total 0.01 (1.0) 0.011 (1.1) 0.011 (1.1) 10 CFR 71.51 (a)(2), §657(b)(ii) of TS-R-1 or §659(b)(i) of 10 (1000) 10 (1000) 10 (1000)
SSR-6 Limits 1Transport Index may not exceed 10.
2Based on the cylindrical geometry of this package, the "top" is considered to be the front as profiled, the "bottom" is considered to be the back as profiled and all other surfaces are considered part of the "sides" of the package.
3Survey results after testing were obtained from the Model 1100 without the optional jacket. This produced dose rates which would be higher than the Model 1100 if it had the optional jacket attached.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 5-2 Table 5.1.B: Model 1100 sn TP237(C) Summary Table of External Radiation Levels Extrapolated to Capacity of 150 Ci lr-192 (Exclusive Use)1 Package (or Freight Container) Surface 2 Meters from Outer Vehicle Surface mSv/h (mrem/h mSv/h (mrem/h' Normal Conditions Top Side Bottom Top Side Bottom of Transport Gamma 1.56 (156) 1.68 (168) 1.52 152)
<0.01 (1.0)
<0.011 (1.1)
<0.011 (1.1)
Neutron NA NA NA NA NA NA Total 1.56 (156) 1.68 (168) 1.52 152)
<0.01 (1.0)
<0.011 (1.1)
<0.011 (1.1) 10 CFR 71.47(b},
10 (1000)2 10 (1000)2 10 (1000)2 0.1 (10) 0.1 (10) 0.1 (10)
§569(a) & (c) of TS-R-1 or §573(a) & (c) of SSR-6 Limits Vehicle Surface mSv/h {mrem/h)
Occupied Position mSv/h {mrem/hr)
Gamma
< 0.001
< 0.011
< 0.011 s 0.02 (2)3 (1.0)
(1.1)
(1.1)
Neutron NA NA NA NA Total
< 0.001
< 0.011
< 0.011 s 0.02 (2)3 (1.0)
(1.1)
(1.1) 10 CFR 71.47(b) or 2 (200) 2 (200) 2 (200) 0.02 (2)
§569(b) of TS-R-1 or
§573(b) of SSR-6 Limits Hypothetical Accident Conditions 1 Meter from Package Surface mSv/h
{mrem/hr)
Gamma 0.01 (1.0) 0.011(1.1) 0.011 (1.1)
Neutron NA NA NA Total 0.01 (1.0) 0.011 (1.1) 0.011 (1.1) 10 CFR 71.51 (a)(2), §657(b)(ii) of TS-R-1 or §659(b)(i) of 10 (1000) 10 (1000) 10 (1000)
SSR-6 Limits 1For packages transported by roadway, railway and sea.
2For packages in closed vehicles, otherwise, 2 (200).
3Confirmed at time of vehicle loading prior to shipment.
Table 5.1.C includes radiation profile data used to demonstrate that the Model 1100 package will meet the external radiation level requirements for non-exclusive use transport when loaded to capacity for Se-75. Based on comparisons of radiation profiles for lr-192 after hypothetical accident testing and relative photon energy outputs for lr-192 and Se-75, it is assessed that radiation levels from Se-75 will be essentially unchanged after undergoing the hypothetical accident condition testing.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 5-3 Table 5.1.C: Model 1100 Summary Table of External Radiation Levels Extrapolated to Capacity of 150 Ci Se-75 (Non-Exclusive Use)1*2 Package Surface mSv/h (mrem/h) 1 Meter from Package Surface mSv/h (mrem/h)
Normal Conditions of Top Side Bottom Top Side Bottom Transport Gamma 0.069 (6.9) 0.073 (73) 0.073 (73) 0.002 (0.2) 0.002 (0.2) 0.002 (0.2)
Neutron NA NA NA NA NA NA Total 0.069 (6.9) 0.073 (73) 0.073 (73) 0.002 (0.2) 0.002 (0.2) 0.002 (0.2) 10 CFR 71.47(a), §524 2 (200) 2 (200) 2 (200) 0.1 (10) 0.1 (10) 0.1 (10)
& 525 of TS-R-1 or §526
& 527 of SSR-6 Limits Hypothetical Accident Conditions3 Gamma
~0.01 (1.0)
~0.01 (1.0)
~0.01 (1.0)
Neutron NA NA NA Total
~0.01 (1.0)
~0.01 (1.0)
~0.01 (1.0) 10 CFR 71.51 (a)(2), §657(b)(ii) of TS-R-1 or §659(b)(i) of SSR-6 10 (1000) 10 (1000)
Limits 1Transport Index may not exceed 10.
2Table results above are based on profile results for Model 1100 serial number TP239-1 B with Se-75 source A426-2 serial number 69202M extrapolated from to a maximum capacity of 5.56 TBq (150 Ci) of Se-75.
10 (1000) 3Based on comparisons of radiation profiles for lr-192 after hypothetical accident testing, it is assessed that radiation levels from Se-75 will be essentially unchanged after undergoing the hypothetical accident condition testing. Based on comparison Se-75 results to lr-192, Exclusive Use results for Se-75 will be no worse than those calculated for NCT.
5.2 Source Specification 5.2.1 Gamma Source The gamma sources allowed for transport in the Model 1100 transport package are specified in Sections 1.2.2 and 2.1 0.
5.2.2 Neutron Source Not applicable. The Model 1100 transport package is not used for the transportation of neutron emitting sources.
5.3 Shieldin odel 5.3.1 Configuration of Source and Shielding A shielding model was not used as the primary justification for this package when containing lr-192 or Se-75. Shielding justification was based on direct measurement.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts 5.3.2 Material Properties October 2024 - Revision 0 Page 5-4 Not applicable. The primary shielding for the Model 1100 package is DU contained within a welded enclosure. DU will not experience any degradation in effectiveness, or radiation exposure, based on temperature extremes in this package. A shielding model was not used as the primary justification for this package. Shielding justification was based on direct measurement using lr-192 and assessment for Se-75. Microshield calculations, supported by a comparison of relative photon shielding effectiveness between lr-192 and Se-75, was only used for capacity justification for Se-75.
5.4 Shielding Evaluation 5.4.1 Methods Shielding justification was based on direct measurement and assessment. All packages are profiled using lr-192 prior to final acceptance and shipment (see Section 8.1.6). This profile accounts for the maximum package capacity and detector geometry. Any package not meeting the required dose rates is rejected.
If the optional jacket is used, it will further reduce surface dose rates on the Model 1100 package. The final acceptance profile is performed on each Model 1100 without the optional jacket attached. As such, the use of the jacket will have no detrimental impact on dose rates.
5.4.2 Input and Output Data Radiation measurements included in this section were adjusted to the maximum activity capacity for the package (e.g., activity correction factor), and the surface measurements were also adjusted to correct for off-set of the survey meter probe from the true surface of the package.
Activity correction factors (CFA) were obtained by using the following relationship:
Mo.J..1.mwnPackageAeti ityCapaci y (Ac)
CFA =
Actual Pr ofile Activaty(A11)
For example, if Ap = 35Ci and Ac = 150Ci, then 150Ci C~ = 135Ci =
- Therefore, in the example above, all original surface and 1 meter profile measurements would be multiplied by a factor of 1.1 for a package profiled using 135 Ci and a package capacity of 150 Ci.
Radiation measurements at the surface of the container are also adjusted to compensate for the offset of the survey meter probe from the true surface of the package. Surface correction factors (SCF) are obtained based on information contained in Section 5.5.2. Evaluation of the SCF in this document revealed that direct application of the inverse square law
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 5-5 introduces an error when the material of the shield contains a heavy element such as tungsten, uranium or lead. When heavy shields are involved, there is a build-up of Compton-scattered photons and X-rays which causes scattered radiation to emanate from everywhere within the shield and not just from the source in the center. Under these circumstances, the inverse square law relationship between dose rate and distance overestimates the actual dose rate on the surface of the device.
Experimental measurement using TLDs demonstrated that the SCF for devices using heavy element shielding varies more accurately as follows:
- d.
SCF =
d 3 where d1 and d3 are det ermined as shown in Fi9ure 5.4.A 1
For example, if d 1 = 9 inches and d3 = 10 inches, then SCF=
( Oinches)
=1.05 (9inches)
Therefore, in the example shown, all original surface profile measurements located along the sides of the package shown in Figure 5.4.A would also be multiplied by a factor to account for surface correction of the detector to the package surface. Different SC F's would be calculated for any dimension of the container where the minimum distance from the center of the activity to the center of the radiation probe is different.
d1 =
distance from activity center to surface of container.
d2 =
distance from activity center to surface of container plus radius of the survey meter probe.
d3 =
distance from activity center to back of the probe.
Figure 5.4.A. - Sample Surface Correction Factor Distance Criteria
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 5-6 The radiation profile data showed no increase in radiation dose after testing beyond normal measurement variations. All test specimens met the regulatory requirements.
5.4.3 Flux-to-Dose-Rate Conversion Not applicable. Flux rates were not used to convert to dose rates in any shielding evaluations.
5.4.4 External Radiation Levels Radiation surveys for the Model 1100 showed maximum surface and 1 meter radiation levels from the transport package within regulatory limits. Radiation surveys of Model 1100 transport package test specimens after undergoing normal and accident condition transport testing were compliant with all applicable the regulatory limits.
5.5 Appendix 5.5.1 Shielding Profile Model 1100 SN TP239-1 B with Se-75 (9/12/2024) 5.5.2 QSA R&D Report 05/02 Issue 1 Surface Dose Rate Correction Factors (February 2005)
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 6-1 Section 6 - CRITICALITY EVALUATION All parts of this section are not applicable. The Model 1100 transport package is not used for shipment of Type 8 quantities of fissile material.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts Section 7 - Package Operations October 2024 - Revision 0 Page 7-1 Operation of the Model 1100 transport package must be in accordance with the operating instructions supplied with the transport package, per 10 CFR 71.87 and 71.89. The supplied operating instructions will include the applicable requirements from Sections 7 & 8 of this SAR.
7.1 Package Loading 7.1.1 Preparation for Loading The Model 1100 transport package must be loaded and closed in accordance with procedures that, at a minimum, include the requirements specified in this section. Shipment of Type B quantities of radioactive material are authorized for sources specified in Section 7.1.1.1. Maintenance and inspection of this package is in accordance with the requirements specified in Section 7.1.1.2.
- 7. 1.1.1 Authorized Package Contents The Model 1100 transport package is designed for use with a special form source capsules as approved under a U.S. Department of Transportation (or other Competent Authority) special form certification. The special form source capsule will be part of a source assembly designed for use and transport within the Model 1100 package. The approved isotopes and maximum package activity limits are shown in Table 7.1.A. Details of encapsulation, as well as the chemical and physical form of the radioactive material, will comply with specifications approved under U.S.
Department of Transportation or other Competent Authority special form certification.
Table 7.1.A: Isotopes Permitted in the Model 1100 Maximum Maximum Nuclide Form Maximum Maximum Weight Weight Capacity1 DU Weight without with Jacket3 Jacket3 lr-192 Special Form 5.55 TBq Sources2 (150 Ci) 34Ibs 44Ibs 48Ibs Se-75 Special Form 5.55 TBq (15.4 kg)
(20 kg)
(22 kg)
Sources2 (150 Ci) 1Maximum activity for lr-192 is defined as output Curies as required in ANSI N432 and 10 CFR 34.20 and in line with IAEA TS-R-1, SSR-6 and Rulemaking by the USNRC and USDOT published in the Federal Register on 26 January 2004.
2Special form is defined in 10 CFR 71, 49 CFR 173 and IAEA TS-R-1 and SSR-6.
3Maximum package weight includes the mass of the radioactive material and the source capsule handling assembly that can be transported in the package.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 7-2
- 7. 1.1. 2 Packaging Maintenance and Inspection Prior to Loading
- a. Ensure all markings are legible.
- b.
Inspect the container for signs of significant degradation. Ensure all welds are intact, the container is free of heavy rust or other cracks/damage to the housing which breaches the container. If there is any evidence of significant damage including any bent or cracked welds contact QSA Global, Inc. prior to shipping.
- c.
Assure all bolts and fasteners (hardware) required for assembly of the package, and as specified on the drawings referenced on the Type B transport certificate, are fit for use. Without removing the hardware by disassembly from the package, examine the visible external surfaces of the bolts/fasteners for any signs of fatigue cracking.
Note: A detailed visual examination of the bolt/fastener thread condition is performed after removal from the exposure device as part of the Annual Maintenance inspections required for radiography devices under 10 CFR 34.31 or equivalent Agreement State regulations.
The bolts/fasteners must be replaced if they are no longer fit for use (e.g., threads stripped, unable to fully thread, signs of cracking, etc.).
Ensure any replacement hardware meets all applicable specifications listed on the drawings referenced on the Type B transport certificate.
- d. Ensure the lock assembly plate is properly secured to the container weldment and that the selector ring lock is functional and secures the selector ring in the "LOCK" position with the dust cover installed in place. Ensure the plunger lock assembly is functional and when engaged prevents rotation of the selector ring from the "LOCK" position.
- e. Ensure the front port knob can be properly secured over the outlet hole.
- f.
If the container fails any of the inspections in steps 7.1.1.2.a-e, remove the container from use until it can be brought into compliance with the Type B certificate.
NOTE:
All components, including fasteners, used on the Model 1100 transport package must be replaced by QSA Global, Inc.
supplied or approved components. Contact QSA Global, Inc.
if replacement components are needed to make a compliant Type B shipment.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 7-3 7.1.2 Loading of Contents NOTE:
These loading operations apply to "dry" loading only. The Model 1100 package is NOT approved for wet loading.
7.1.2.1 General Pre-transportation Requirements
- a. The contents are authorized for use in the package.
- b. The package condition has been inspected in accordance with Section 7.1.1.2.
- c.
Ensure that the source is secured into place in the storage position after loading the Model 1100 package in accordance with the applicable licensing provisions for the user's facility related to radioactive material handling.
7.1.2.2 Preparation for Transport NOTE:
- a. Ensure that all conditions of the certificate of compliance are met and that the package is assembled in accordance with the drawing(s) referenced on the certificate of compliance.
- b.
Ensure removable contamination on the outside surface of the package does not exceed the limit specified in 49 CFR 173.443.
- c. Survey all exterior surfaces of the package to ensure that the radiation level does not exceed 2 mSv/hr (200 mR/hr) at the surface. Measure the radiation level at one meter from all exterior surfaces to assure that the radiation level is less than 0.1 mSv/hr (10 mR/hr).
- d.
Ensure a seal wire(s) is properly installed over the front port knob to serve as a tamper indicating seal during transport.
- e. Ship the container according to the requirements for transporting radioactive material as established in 10 CFR 71.5 and 49 CFR 171-178.
The US Department of Transportation, in 49 CFR 173.22(c), requires each shipper of Type B quantities of radioactive material to provide prior notification to the consignee of the dates of shipment and expected arrival.
7.2 Package Unloading 7.2.1 Receipt of Package from Carrier 7.2.1.1 The consignee of a transport package of radioactive material must make arrangements to receive the transport package when it is delivered. If the transport package is to be picked up at the carrier's terminal, 1 O CFR
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 7-4 20.1906 requires that this be done expeditiously upon notification of its arrival.
7.2.1.2 Upon receipt of a transport package of radioactive material:
- a. Survey the transport package with a survey meter as soon as possible, preferably at the time of pick-up and no more than three hours after it was received during normal working hours. Radiation levels should not exceed 2 mSv/hr (200 mR/hr) at the surface of the transport package, nor 0.1 mSv/hr (1 O mR/hr) at a distance of 1 meter from the surface.
Record the actual radiation levels on the receiving report.
If the radiation levels exceed these limits, secure the container in a Restricted area, and notify the appropriate personnel in accordance with 10 CFR 20 or applicable Agreement State regulations.
- b. Inspect the package for physical damage or leaking. If the package is damaged or leaking, or it is suspected that the package may have leaked or been damaged, restrict access to the package. As soon as possible, contact the Radiation Safety Officer to perform a full assessment of the package condition and take necessary follow-up actions.
- c. Record the radioisotope, activity, model number, and serial number of the source and the transport package model number and serial number.
7.2.2 Removal of Contents 7.2.2.1 Unload the package in accordance with the instructions supplied with the package (or QSA Global, Inc.) per 10 CFR 71.89. If instructions are not available, contact QSA Global, Inc. for assistance in obtaining a copy of the applicable instructions.
7.2.2.2 Unloading of the package must also be in accordance with applicable licensing provisions for the user's facility related to radioactive material handling.
7.3 Preparation of Empty Package for Transport In the following instructions, an empty transport package refers to a Model 1100 transport package without a radioactive source contained within the shielded container. To ship an empty transport package:
7.3.1. Unload the container in accordance with Section 7.2.2.
7.3.2 Assure that the levels of removable radioactive contamination on the outside surface of the transport package does not exceed 4 Bq/cm2 (0.0001 µCi/cm2) when averaged over 300 cm 2.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 7-5 7.3.3 Assure that the levels of removable radioactive contamination on the inside surface of the shield container does not exceed 400 Bq/cm 2 (0.01 µCi /cm2) when averaged over 300 cm 2.
7.3.2 When it is confirmed that the Model 1100 transport package is empty, prepare the transport package for shipment and survey to ensure the external surface radiation level does not exceed 5 µSv/h (0.5 mrem/hr).
NOTE:
If the dose rate on the surface equals or exceeds 5 µSv/h (0.5 mrem/hr) based on the DU shielding, then the package, without a radioactive source assembly installed, must be shipped as an excepted package and comply with the requirements of 49 CFR 173.426.
7.4 Other Operations 7.4.2 Package Transportation By Consignor Persons transporting the Model 1100 transport package in their own conveyances should comply with the following:
7.4.2.2 For a conveyance and equipment used regularly for radioactive material transport, check to determine the level of contamination that may be present on these items. This contamination check is suggested if the package shows signs of damage upon receipt or during transport, or if a leak test on the special form source transported in the package or a source tube DU contamination wipe exceeds the allowable limit of 185 Bq (0.005
µCi).
7.4.1.2 If contamination above 4 Bq/cm 2 (0.0001 µCi/cm2), based on wiping an area of 300 cm2, is detected on any part of a conveyance or equipment used regularly for radioactive material transport, or if a radiation level exceeding 5 µSv/h (0.5 mR/hr) is detected on any conveyance or equipment surface, then remove the affected item from use until it can be decontaminated or decayed to meets these limits.
7.4.1.3 Ensure the package is properly blocked and braced prior to transport to prevent movement within the conveyance during transport.
7.4.3 Emergency Response In the event of a transport emergency or accident involving this package, follow the guidance contained in "2024 (or later editions) Emergency Response Guidebook: A guidebook intended for first responders during the initial phase of a transportation incident involving hazardous materials/dangerous goods", or equivalent guidance documentation.
Reference:
"2024 Emergency Response Guidebook: A Guidebook for First Responders During the Initial Phase of a Transportation Incident Involving Hazardous Materials/Dangerous Goods"
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts 7.5 Appen *x Not Applicable.
October 2024 - Revision 0 Page 7-6
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts October 2024 - Revision 0 Page 8-1 Section 8 -ACCEPTANCE TESTS AND MAINTENANCE PROGRAM 8.1 Acceptance Test 8.1.1 Visual Inspections and Measurements Prior to the first use of any Model 1100 for Type B transport, the package is inspected visually for compliance to the following criteria:
8.1.1.1 The transport package was assembled properly to the applicable drawing.
8.1.1.2 The package contains no cracks, pinholes, uncontrolled voids or other defects that could significantly reduce the effectiveness of the package.
8.1.1.3 The transport dose rate requirements for the package are met when loaded to it maximum approved capacity.
8.1.1.4 All fasteners as required by the applicable drawings are properly installed and secured.
8.1.1.5 The relevant labels are attached and contain the required information including the model number, serial number, gross weight and package identification number associated with the Type B(U) package. In addition, the package is also marked in accordance with 10 CFR 20.1904(a), 1 O CFR 34, and 10 CFR 71 or equivalent Agreement State or applicable International regulations.
Visual inspections and measurements will be performed in accordance with a USN RC approved Quality Assurance Program per the requirements of 10 CFR 71.101.
8.1.2 Weld Examinations Weld examinations will be performed in accordance with the applicable drawing requirements and QSA Global, lnc.'s USNRC approved Quality Assurance Program No. 0040.
8.1.3 Structural and Pressure Tests Prior to first use as part of a Model 1100 Transport Package, container structural conformance will be evaluated in accordance with the applicable drawing requirements and QSA Global, lnc.'s USNRC approved Quality Assurance Program No. 0040. The containment system is not designed to require increased or decreased operating pressures to maintain containment during transport, therefore pressure tests of package components prior to first use are not required.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts 8.1.4 Leakage Tests October 2024 - Revision 0 Page 8-2 The source capsules (primary containment) transported in the Model 1100 package are wipe tested for leakage of radioactive contamination upon initial manufacture.
The removable contamination must be less than 185 Bq (0.005 µCi). The source capsules will also be subjected to leak tests compliant to ISO 9978:2020(E) (or more recent editions) at 6 month intervals to ensure continued containment integrity. A source capsule that fails any of these tests is not approved for Type B transport in the Model 1100 package.
8.1.5 Component and Material Tests Component and material compliance is achieved in accordance with the requirements in QSA Global, lnc.'s USNRC approved Quality Assurance Program No. 0040.
8.1.6 Shielding Tests The radiation levels at the surface of the transport package and at 1 meter from the surface are measured upon manufacture. This survey is performed in a low background area and involves a slow scan survey of the entire surface area as well as one meter from the surface of the package. This survey is used to identify any void volumes or shield porosity which could prevent the finished device from complying with the dose limits in 10 CFR 71.47.
The radiation profile survey made of each package surface and at one meter from the package surface, is extrapolated to the maximum capacity of the transport package. These surveys must not exceed 2 mSv/hr (200 mR/hr) at the surface, nor 0.1 mSv/hr (1 o mR/hr) at 1 meter from the surface of the transport package. Failure of this radiation profile survey will cause rejection of the affected Model 1100 package.
Rejected packages which do not comply with the construction requirements on the applicable drawings referenced on the Type B certificate, or that do not comply with the radiation profile requirements will not be distributed as approved Type B(U) packages.
8.1. 7 Thermal Tests Not applicable. The source content of the Model 1100 package has minimal effect on the package surface temperature and therefore no additional testing is necessary to evaluate thermal properties of the packaging.
8.1.8 Miscellaneous Tests Not applicable.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts 8.2 Maintenance Program 8.2.1 Structural and Pressure*Tests October 2024 - Revision 0 Page 8-3 Not applicable. Material certification is obtained for Safety Class A components used in the transport package prior to their initial use. Based on the construction of the design, no additional structural testing during the life of the package is necessary if the container shows no signs of defect when prepared for shipment in accordance with the requirements of Section 7 of the SAR.
The Model 1100 packaging is not designed to require increased or decreased operating pressures to maintain containment during transport, therefore pressure tests of package components prior to individual shipment are not required.
8.2.2 Leakage Tests As described in Section 8.1.4, "Leakage Tests," the radioactive source assembly is leak-tested at manufacture. In addition, the source is leak tested in accordance with that Section at least once every six months thereafter if being transported to ensure that removable contamination is less than 185 Bq (0.005 µCi).
In addition, users of the package are required to perform an annual contamination wipe test of the shield source tube (reference 10 CFR 34.27(e)). A contamination wipe of the source tube is also performed whenever the package is returned to the manufacturer.
8.2.3 Component and Material Tests The transport package is inspected for tightness of fasteners, general condition, and fitness for use prior to each use (see Section 7.1.1 ). Prior to each use, a radiation survey of the transport package is made to assure that the radiation levels do not exceed 2 mSv/hr (200 mR/hr) at the surface, nor 0.1 mSv/hr (1 O mR/hr) at 1 meter from the surface.
8.2.4 Thermal Tests Not applicable. The source contents of the Model 1100 package have no adverse effect on the package surface temperature, and therefore no additional testing is necessary to evaluate thermal properties of the packaging prior to shipment.
8.2.5 Miscellaneous Tests Inspections and tests designed for secondary users of this transport package under the general license provisions of 10 CFR 71.1 ?(b) are provided in Section
- 7.
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts 8.2.5.1 Ageing Mechanism Considerations October 2024 - Revision 0 Page 8-4 Inspections and tests designed for secondary users of this transport package under the general license provisions of 10 CFR 71.1 ?(b) are provided in Section 7. Based on use of other similar QSA Global, Inc. package designs (e.g., Model 880 Series), it is not expected that the Model 1100 package will see large periods of non-use or storage in between container shipments. Even if periods of non-use/storage are experienced on an individual package, inspection as described in Section 7 prior to making a Type B shipment will identify any degradation that might have occurred during storage which could adversely affect the package integrity.
The materials used in the Model 1100 package are not vulnerable to degradation due to irradiation over time, and there will be no chemical/galvanic material interactions between these materials. As such, there will be no degradation of the package internal construction due to storage, empty or loaded, prior to shipment.
Any degradation on externally assessable components will be identified by the inspections and tests specified in Section 7 as these are sufficient to identify package degradation which could adversely impact the package integrity after storage (reference IAEA SSR-6 (2012 & 2018) §503(e) & 613A). Performance of the Section 7 inspections will ensure that the package remains compliant with all applicable Type B(U) package certifications requirements prior to shipment.
Reference:
IAEA Safety Guide No. SSG-26 (Rev 1) Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material (2018 Edition).
8.3 Appendix Not applicable.
l
Safety Analysis Report for the Model 1100 Transport Package QSA Global, Inc.
Burlington, Massachusetts Section 9 - Quality Assurance 9.1 U.S. Quality Assurance Program Requirements October 2024 - Revision 0 Page 9-1 All component fabrication (including assembly) is controlled under the QSA Global, Inc.
Quality Assurance program approved by the USNRC (approval number 0040) and ISO 9001.
9.2 Canada Quality Assurance Program Requirements Not applicable. This package is originally submitted for certification in the United States and complies with the criteria in Section 9.1.