ML20335A495
| ML20335A495 | |
| Person / Time | |
|---|---|
| Site: | 07109302 |
| Issue date: | 11/30/2020 |
| From: | Shaw D Orano TN Americas, TN Americas LLC |
| To: | Document Control Desk, Office of Nuclear Material Safety and Safeguards |
| Shared Package | |
| ML20335A494 | List: |
| References | |
| E-57283 | |
| Download: ML20335A495 (90) | |
Text
Orano TN 7160 Riverwood Drive Suite 200 Columbia, MD 21046 USA Tel: 410-910-6900 Fax: 434-260-8480 Enclosures transmitted herein contain SUNSI. When separated from enclosures, this transmittal document is decontrolled.
November 30, 2020 E-57283 U. S. Nuclear Regulatory Commission Attn: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852
Subject:
Application for Revision 11 to Certificate of Compliance No.
9302 for the Model No. NUHOMS-MP197 Packaging, Docket No. 71-9302
References:
[1] Revision 10 to Certificate of Compliance No. 9302 for the Model No. NUHOMS-MP197 Packaging
[2] NUHOMS-MP197 Transportation Package Safety Analysis Report, Revision 20, September 2019 In accordance with 10 CFR 71.31(b), TN Americas LLC submits an application for revision to the Certificate of Compliance (CoC) No. 9302 for the Model No. NUHOMS-MP197 Packaging [1].
A revision to the CoC is needed to add optional specifications to the packaging and radioactive waste canister(s) (RWC) design. These changes provide assurance that an RWC currently in storage or an RWC that will be loaded and placed in storage can be transported in the MP197HB. The MP197HB is currently in use to ship low-level radioactive waste (LLW) in RWC to disposal facilities, and will continue to be used primarily for this purpose.
Changes to the Safety Analysis Report (SAR) [2] are summarized in, and preliminary changed pages provided as Revision 21A in. A consolidated SAR Revision 21 will be submitted upon completion of the NRC review.
E-57283 Document Control Desk Page 2 of 2 Should the NRC staff have any questions or require additional information regarding this submittal, please contact Mr. Peter Vescovi by telephone at (336) 420-8325, or by e-mail at peter.vescovi@orano.goup.
Sincerely, Don Shaw Licensing Manager TN Americas LLC cc:
Pierre Saverot, Senior Project Manager, U.S. Nuclear Regulatory Commission Eric Pernice, Senior Project Manager, Orano TN Americas
Enclosures:
- 1. Summary of Changes to MP197HB SAR Revision 21
- 2. NUHOMS-MP197HB SAR, Revision 21A, Changed Pages and Drawings, Proprietary Version
- 3. NUHOMS-MP197HB SAR, Revision 21A, Changed Pages and Drawings, Public Version
- 4. Affidavit Pursuant to 10 CFR 2.390 Don Shaw Digitally signed by Don Shaw Date: 2020.11.30 06:01:06
-05'00' to E-57283 Summary of Proposed Changes (Safety Analysis Report, MP197HB, Revision 21A) 1 of 12 This enclosure provides a summary of the proposed changes to guide the review of the affected sections of the safety analysis report (SAR). The substantive changes to the SAR are described in the following summary and more detailed description and justification of the changes is provided in the table that follows.
1.0 Summary of changes Contents Specification Contents for the radioactive waste canister (RWC) have been modified to add allowance for the presence of organic material and residual water with the operating conditions added to require that the effect of gas generation from radiolysis is evaluated prior to shipping, including consideration that flammable gases may be generated. New activity limits have been added to allow use of an RWC with a 0.5-inch minimum shell thickness.
Engineering Drawings Drawing NUHRWC-71-1001 has been revised to add new RWC lid closure detail, and to provide the option to have a 0.5-inch minimum RWC shell thickness. Drawings MP197HB 1002, MP197HB-71-1003, and MP197HB-71-1005 for the MP197HB cask have been revised to include options for the ram access cover plate material.
Shielding Evaluation An additional shielding evaluation has been performed to determine an activity limit for an alternate RWC with a minimum 0.5-inch shell thickness.
Operating Instructions, and Acceptance Tests and Maintenance Instructions Several changes have been made to include consideration of residual water that may remain in the RWC after drying, contents limits for an alternate RWC with reduced shell thickness, instructions for preparing empty package for shipment to allow flexibility in the package classification, additional instructions for external radiation measurements and controls during shipment, and leak test criteria.
2.0 Detailed description and justification of changes Each item in the table below corresponds to a description and justification of changes that affect one or more sections in the SAR. Design changes made to the package design are subject to design control measures commensurate with the original design as previously approved. A TN Americas Quality Assurance program implementing procedure requires the review of design changes for impact on the SAR. A record of this review is noted for each item as Licensing Review (LR) 719302-XXXX or Transportation Licensing Change (TLC) 719302-XXXX.
to E-57283 Summary of Proposed Changes (Safety Analysis Report, MP197HB, Revision 21A) 2 of 12 Item Chapter/Appendix/
Section Description and Justification Chapter A.1 General Information
1.1 Section
A.1.4.10.1 LR 719302-0044
==
Description:==
Update the revision level of the drawings.
Justification:
This is an administrative change to reflect the drawing revisions required to incorporate the change.
1.2a Drawings:
MP197HB-71-1002 MP197HB-71-1003 MP197HB-71-1005 LR 719302-0044
==
Description:==
Split Item 33 (ram access cover plate) into two items: 33A (Type 1 ram cover plate, material SA-203 Gr. E only), and 33B (Type 2 ram cover plate, material SA-203 Gr. E or SA-240 Type 304 / SA-182 Type F304).
Justification:
This change is needed to restrict the material of the Type 1 ram closure plate to carbon steel only per the results of updated Calculation MP197HB-0204, which shows that the higher bolt torques required to compress the metallic seal of the Type 1 ram cover plate are not compatible with stainless steel mechanical properties, and follows the design drawing update.
1.2b Drawing:
MP197HB-71-1002 LR 719302-0044
==
Description:==
Clarifications to the nomenclature of Items 36 and 37 Justification:
This change is only to clarify when Items 36 and 37 are meant to be used.
1.3a Drawing:
NUHRWC-71-1001 Sheet 1 and Sheet 2 TLC 719302-0005
==
Description:==
- 1) Add a metallic option for the seal (Item 6).
- 2) Update Note 11 to allow for a penetration through the outer top cover plate (Item 4) to remain open during transportation.
- 3) Add a new closure detail for the alternate RWC-B configuration
- 4) Add a note for alternative weld configuration for the bolting ring.
to E-57283 Summary of Proposed Changes (Safety Analysis Report, MP197HB, Revision 21A) 3 of 12 Item Chapter/Appendix/
Section Description and Justification 1.3a (cont.)
Justification:
The bottom configuration of the alternate RWC-B is in alignment with the current SAR drawing Alternate Section A-A. The quality classification of RWC components are the same or greater than the classification on the SAR drawings. For the multiple plate shield plug, separate calculations were performed for the RWC-B, as shown in Calculation RWC197-0203 R2. Section 7.1.2 concludes that if the bolts for a mutli-plated shield plug failed for a Service Level D case, there would be no loss of function for the shield plug. Since the RWC is vented, only bulk water removal can be performed. During transportation, the MP197HB serves as the containment boundary, where the MP197HB is filled with helium to assist with heat removal and provide a non-reactive environment during the transportation of the RWC. The Alternate RWC-B uses a 61BTH dry shielded canister (DSC) shell for the design, which is the spacer within Table A.7-1 of the SAR.
- 1) Justification of seal material: The use of a metallic seal is an improvement to the existing SAR drawing. Compared to the use of rubber material, the metallic seal is able to last significantly longer in service when properly installed. The use of the metallic seals used for the Nine Mile Point RWC is analyzed in Calculation 35208-0200. The project-specific metallic seals have a pressure rating of 14 psi and temperature rating of 125 °F, which is in alignment for the design conditions of the RWC per the design criteria document 35208-DCD-001.
- 2) Justification for Note 11 update: In an accident event during transportation of the RWC within MP197HB, if all of the bolts failed, the radioactive materials would remain contained within the RWC. The depth of the centering part of the shield plug is much larger than the RWC/cask cavity gap of 0.5, specified in Table A.7-1 and, therefore, ensures that the RWC lid cannot slide by more than 0.5 if the lid bolts fail and the radial gap between the shield plug and the shell is less than 0.5".
- 3) Justification for new closure configuration: The new bolted alternate RWC-B configuration, like the existing RWC-B design, does not have a welded plate above the bolts within the SAR and is, instead, more similar to the RWC-DD configuration. The axial gap between the top cover plate and shield plug is a maximum of 0.41. Additionally, a note was added for an alternate configuration of the bolting ring to shell weld for fabrication.
- 4) Justification for the alternate weld: The previously evaluated 3/8 weld on the top and bottom of the bolting ring within the shell may, alternatively, have a circumferential complete joint penetration weld. This weld is substantially stronger than the previous weld and provides an easier fabrication alternative. Additional details have been added to the SAR drawing for further clarification.
to E-57283 Summary of Proposed Changes (Safety Analysis Report, MP197HB, Revision 21A) 4 of 12 Item Chapter/Appendix/
Section Description and Justification 1.3b Drawing:
NUHRWC-71-1001 Sheet 1 and Sheet 2 TLC 719302-0010
==
Description:==
The RWC-WA is a variant of the RWC-W that is previously approved and included in CoC No. 9302.
- 1) Add the following to Note 8 :
As an alternative to AWS D1.1 or D1.6 requirements, it is acceptable for the RWC-WA/liner to be welded to ASME Section IX requirements and inspected to ASME Section V requirements.
2)Add the following to Note 4:
A combination of multiple plates may be used for the top closure.
- 1) These welding alternatives are acceptable for DSCs.
- 2) Total thickness of steel is the criteria for the shielding material.
- 3) These material alternatives are acceptable for DSCs.
1.4a Sections:
A 1.4.9A.1 A 1.4.9A.1.4 Table:
A.1.4.9A-1 TLC 719302-0005
==
Description:==
Section A 1.4.9.1, which adds the Alternate RWC-B to the types of design configurations.
- 1) Add A 1.4.9A.1.4 to summarize the Alternate RWC-B Design
- 2) Add column for dimensions of Alternate RWC-B in Table A.1.4.9A-1 Justification:
See Item 1.3a.
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Section Description and Justification 1.4b Sections:
A.1.4.9A.1 A.1.4.9A.1.1 A.1.4.9A.1.2 Tables:
A.1.4.9A-1 A.1.4.9A-2 TLC 719302-0010
==
Description:==
A.1.4.9A.1 Radioactive Waste Canister Description
- 1) Change description of welded configuration to:
Welded Design (RWC-W)
- 2) Add the following to RWC acceptable welding requirements:
As an alternative to AWS D1.1 or D1.6 requirements, it is acceptable for the RWC-W to be welded to ASME Section IX requirements and inspected to ASME Section V requirements.
A.1.4.9A.1.1 Welded Top Shield Plug Design (RWC-W)
- 3) Delete the shell and liner thicknesses from the text.
- 4) Change Four lifting lugs to Lifting attachments.
- 5) Add The combined thickness of the RWC-W cylindrical shell and liner is 1.75 inches.
- 6) A.1.4.9A.1.2 Bolted Top Shield Plug Design (RWC-B)6) Change The RWC-B cylindrical shell is 1.75 inch thick to provide the same shielding as the RWC-W 1.25 inch thick shell used with the 0.50 inch thick inner liner to The combined thickness of the RWC-B cylindrical shell and liner is 1.75 inches.
Table A.1.4.9A-1 Nominal Dimensions of the RWC:
- 7) Change RWC-W Canister Length to 186.50 or 196 Change RWC-W Cavity Length and Cavity Diameter to (min)
Change footnote (2) to The shell thickness for the RWC-W is 1.75 inches, which includes the RWC-W inner liner thickness.
Add footnote (3) Dimensions may vary depending on the shell and cover plate thicknesses.
Table A.1.4.9A-2 Nominal Dimensions of the RWC Inner Liner
- 8) Change shell thickness to 0.50 (min).
Add footnote (1) Dimensions may vary depending on the shell and cover plate thicknesses.
Justification:
- 1) RWC-WA is a three-part top closure configuration with the inner top cover plate and outer top cover plates welded. The top shield plug is not a welded lid. The RWC-WA closure configuration conforms to the specification for the RWC-W closure.
- 2) These welding alternatives are acceptable for Dry Shielded Canisters.
- 3) SAR Drawing NUHRWC-71-1001 allows for a combination of multiple plates to be used to achieve the overall axial and radial thicknesses required for shielding. These may include internal liners, baskets, sleeves or top and/or bottom shield plugs located inside the RWC.
to E-57283 Summary of Proposed Changes (Safety Analysis Report, MP197HB, Revision 21A) 6 of 12 Item Chapter/Appendix/
Section Description and Justification 1.4b (cont.)
- 4) Allows flexibility in the design of lifting attachments to be lifting lugs or other alternative where lifting lugs cannot be used.
- 5) See Justification 3.
- 6) See Justification 3.
- 7) The RWC-DD allows for longer length, which can also apply to the RWC-WA. SAR Drawing NUHRWC-71-1001 allows for a combination of multiple plates to be used to achieve the overall axial and radial thicknesses required for shielding. These may include internal liners, baskets, sleeves or top and/or bottom shield plugs located inside the RWC. The shell cavity length will vary depending on the shell length and liner shielding thicknesses.
- 8) See Justification 3.
1.5 Page:
A.1.4.9A-3 Section:
A.1.4.9A.2, RWC Contents TLC 719302-0011
==
Description:==
Revise to allow loading of organic material with provisions for evaluation.
Justification:
The RWC Contents description is modified to be consistent with NRC guidance on organics to say that gas generation from organics or biological growth need to be considered and evaluated for each shipment.
The package user will be responsible for characterizing the contents and evaluating the gas generation from thermal decomposition, radiolysis, or biologic growth.
1.6 Pages
A.1.4.10-i A.1.4.10-2 through A.1.4.10-7 TLC 719302-0015
==
Description:==
Update Appendix A.1.10 Table of Contents and headers for drawing lists.
Justification:
Editorial
1.7 Pages
A.1.4.9A-1, A.1.4.9A-2a, A.1.4.9A-3 A.1.4.9A-4 A.1.4.9A-5 TLC 719302-0018
==
Description:==
Add description of Alternate RWC with 0.5 shell thickness.
Justification:
Calculation is performed to provide the basis for loading radioactive waste content in the MP197HB Unit 01 transport cask. The calculation determined maximum activity in Ci of Cobalt-60 or equivalent and the equivalent activity limits as function of gamma energy for isotopes other than Cobalt-60. The analysis is performed assuming a minimum steel shell thickness of 0.5 and same top and bottom lids as those of RWCs already licensed in MP197HB.
to E-57283 Summary of Proposed Changes (Safety Analysis Report, MP197HB, Revision 21A) 7 of 12 Item Chapter/Appendix/
Section Description and Justification Chapter A.2 Structural Evaluation 2.1a Page:
A.2.13.2-15, Section:
A.2.13.2.9 LR 719302-0044
==
Description:==
Mention the fact that the ram closure plate may be made of either SA-203 Gr. E (carbon steel) for Type 1, or SA-203 Gr. E (carbon steel) or SA-240 Type 304 / SA-182 Type F304(stainless steel)for Type 2.
Justification:
The option for ram closure plate to be made out of stainless steel was previously not mentioned in the SAR.
2.1b Section:
A.2.13.2.9 Tables:
A.2.13.2-1, A.2.13.2-9, A.2.13.2-9a, A.2.13.2-10, A.2.13.2-11 LR 719302-0044
==
Description:==
Add an analysis of the stainless steel ram closure plate bolts Justification:
The current SAR analysis only considers carbon steel for the ram cover plate material. However, this cover plate may also be made of stainless steel in the case of the Type 2 ram cover plate, which has inferior mechanical characteristics to carbon steel and, therefore, must be evaluated to show its adequacy, and also requires a lower bolt torque. The case of a Type 2 ram cover plate made out of carbon steel is covered by the analysis of the Type 1 carbon steel cover plate.
2.1c Pages:
A.2.13.2-i A.2.13.2-1 LR 719302-0044
==
Description:==
Editorial changes to align titles and text with the Appendices summary in Chapter A.2 and the content of Section A.2.13.2.
Justification:
This is an administrative change to ensure consistency with previous Section A.2.13.2.
2.1d Section:
A.2.13.7 Table:
A.2.13.14-1 TLC 719302-0005
==
Description:==
Added Alternate RWC-B to Table A.2.13.14-1.
Justification:
See Item 1.3a.
to E-57283 Summary of Proposed Changes (Safety Analysis Report, MP197HB, Revision 21A) 8 of 12 Item Chapter/Appendix/
Section Description and Justification 2.2 Page:
A.2.13.13-2 Table:
A.2.13.13-1 TLC 719302-0007
==
Description:==
Change ASME Code Alternatives for ASME NB-3122.1 for the NUHOMS MP197HB Cask Containment Boundary to the requirements imposed during fabrication and periodic maintenance to the weld overlays/plate cladding.
Justification:
The ASME NB-3122.1 code requirement does not allow structural strength to be attributed to weld overlays and cladding. However, the structural evaluation for the MP197HB described in Chapter A.2 does take credit for the thickness of these weld overlays and cladding. Therefore, the alternative to this code requirement, as stated in SAR Revision 19 as supplemented, requires inspections during fabrication and periodic inspections during the service life of the transportation cask that ensure the thickness and lamination to the base metal of the weld overlays and cladding. However, such an annual inspection is cost-prohibitive, and a detailed structural evaluation showed that its periodicity could be safely decreased.
The weld overlays and cladding are integrally bonded on one side with the base layer of carbon steel and meet the requirements of SA-264 (clad plate) and SA-578 (weld overlays). Per the ASME SA-264 requirements, the clad plate was load-tested and heat-treated and showed full compliance with the requirements. Weld overlayed materials and weld joints were subject to 100% ultrasonic testing (UT) examination per ASME SA-578 with the acceptance standard of ASME NB-5331, with the exception of the bottom and cask body flange forging, which the analytical evaluation shows meet the stress criteria, even without attributing any strength to their weld overlays, and of port sealing surfaces, which are non-structural surfaces.
The clad plate was tested to 45 ksi shear bond strength, whereas SA-264
§7.2.1 (Addenda 2006) only requires 20 ksi. It was also tested by UT with 100% coverage (S1) with acceptance criteria of Quality Level Class 1 according to SA-264 Section 13. In SA-264, the UT coverage is at the discretion of the manufacturer and the default acceptance criteria are from quality level Class 5. That is, the clad plate under this code alternative is required to have the highest extent of UT coverage and the strictest acceptance criteria from SA-264. There is no ASME code requirement for UT exam of weld overlays.
Although the weld overlays (with the exceptions mentioned above) and cladding were 100% inspected during fabrication and showed no delamination, a starting delamination of 1 diameter corresponding to the Quality Level Class 1 UT acceptance criteria is assumed for the fracture mechanics analysis under fatigue loading. All the transportation loads that are cyclic in nature are considered in the evaluation.
The results of the analytical evaluation demonstrate that the toughness of the materials is sufficient to prevent any fracture instability or significant crack propagation of the delamination of the weld overlays and cladding for the service lifetime (300 one-way trips).
to E-57283 Summary of Proposed Changes (Safety Analysis Report, MP197HB, Revision 21A) 9 of 12 Item Chapter/Appendix/
Section Description and Justification Chapter A.5 Shielding Evaluation
5.1 Pages
A.5-1b A.5-3 A.5-6a A.5-80w TLC 719302-0018
==
Description:==
Add shielding analysis for RWC content loaded in 0.5 shell thickness.
Justification:
See Item 1.7.
Chapter A.6 Criticality Evaluation 6.1 Page:
A.6.5.14-28l
==
Description:==
Replace with correct table.
Justification:
Editorial Chapter A.7 Operating Instructions
7.1 Table
A.7-1 Appendix:
A.7.7.10 TLC 719302-0005 Description of Change:
Add Alternate RWC-B to Table A.7-1.
Add guidance for the loading procedures for the Alternate RWC-B in Appendix A.7.7.10.
Justification:
See Item 1.3a.
7.2a Table:
A.7-1, DSC, Fuel, and Basket Spacer Nominal Heights for Each Type of DSC/RWC TLC 719302-0010
==
Description:==
Change footnote (5) to Height of spacer for RWC-W is 11.75" or 2.25" for the 186.5" or 196" canister length, respectively. The height of spacer for the RWC-B is 11.75" and height of spacer for RWC-DD is 2.25".
Justification:
The RWC-WA will be either the length of the RWC-W or the RWC-DD requiring the associated spacer length.
7.3a Page:
A.7-18 Table:
A.7-2b TLC 719302-0011
==
Description:==
Add items (d), (e), (f) to Table A.7-2b Applicable Content Specification for RWC Justification:
Items (d) and (f) were previously added to Section A.1.4.9A.2, RWC Contents, but were not added to Table A.7-2b.
Item (e) adds the revised allowable contents for organic material as added to Section A.1.4.9A.2, RWC Contents, in change discussed above.
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Section Description and Justification 7.3b Page:
A.7.7.10-1 TLC 719302-0011
==
Description:==
Delete step A.7.7.10.2.3, which is to disengage top shield plug from the lifting yoke and position yoke clear of the task.
Justification:
This level of detail is not needed and the lift yoke may or may not be used to lift the shield plug.
7.4 Pages
A.7-5c, A.7-14a, A.7.7.10-2, A.7.7.10-3 TLC 719302-0013
==
Description:==
Add steps to Chapter A.7-Operating Instructions to require an evaluation of residual water that may remain in the cask cavity or RWC after preparing for transport.
Justification:
Residual water may remain in RWC not dried to 10 mbar. The amount of residual water may result in exceeding the MNOP or design pressure for the RWC or cask during fire HAC. Both vapor pressure of the water and gas generation from radiolysis need to be considered. In addition, radiolysis could result in exceeding the lower flammability limit for hydrogen.
7.5 Pages
A.7-5 A.7-12 TLC 719302-0015
==
Description:==
- 1) Change Table A.6-17 to Table A.6.5.8-7.
- 2) Delete this sentence referring to flow chart of assembly verification in Figure A.7.2.
Justification:
- 1) Editorial issue was created in Revision 12 (Feb 2012) when the table number was changed in Section 6, but the reference here was not updated.
- 2) Editorial issue was created in Rev 19 when Figure A.7-2 was removed from the SAR but this reference was not removed.
to E-57283 Summary of Proposed Changes (Safety Analysis Report, MP197HB, Revision 21A) 11 of 12 Item Chapter/Appendix/
Section Description and Justification 7.6 Page:
A.7-8 Section:
A.7.1.4.1 TLC 719302-0015
==
Description:==
Revise Section A.7.1.4.1 step 14 of the SAR to require supplemental dose rate measurements in the normally occupied spaces. The following wording shall be added at the end of step 14:
This final radiation survey shall include a measurement of the dose rate in the normally occupied spaces (i.e. truck cabin). If the measured dose rate in the normally occupied spaces exceeds 2 mrem/hr, the location of the package shall be changed or supplementary shielding added as necessary to reduce the dose to an acceptable level. Supplementary shielding may be added to the conveyance (e.g. attached to the sides of the trailer or truck cab) to reduce the external radiation levels, but shall not be attached to the package without prior NRC approval. Alternatively, the carrier may implement the radiation dosimetry requirements of 10 CFR 71.47(b)(4) and 49 CFR 173.441(b)(4) to satisfy the requirements of 10 CFR 20.1502.
Justification:
Added to ensure compliance with 10 CFR 71.47(b) and DOT Hazardous Material Regulations for exclusive use shipments.
7.7 Page:
A.7-11, -14 TLC 719302-0019
==
Description:==
SAR Section 7.2, Preparation of Empty Package for Transport, refers to 49 CFR 173.427 in the DOT Hazardous Material Regulation for shipping Low Specific Activity (LSA) material. Prior to a change that added rules for shipping LSA, HMR 173.427 was the rule for shipping empty radioactive material packaging. In the current 49 CFR, 173.427 is for shipping LSA and 173.428 is the rule for empty packaging. The reference to a specific HMR rule for shipping empty packaging is replaced with a general instruction to use the appropriated HMR rules.
Justification:
Residual internal contamination may prevent shipping the empty packaging as allowed by 49 CFR 173.428. The package may be shipped as Type A, Type B, or Surface Contaminated Object (SCO) depending on the level of residual contamination. Not specifying the type of package allows operational flexibility in preparing the empty package for transport.
7.8 Pages
A.7-1a, A.7-18, A.7-18b1 TLC 719302-0018
==
Description:==
Add shielding analysis for RWC content loaded in 0.5 shell thickness.
Justification:
See Item 1.7.
to E-57283 Summary of Proposed Changes (Safety Analysis Report, MP197HB, Revision 21A) 12 of 12 Item Chapter/Appendix/
Section Description and Justification 7.9 Page:
A.7.7.10-2 TLC 719302-0020
==
Description:==
Change the backfill pressure range in instruction step 18 from 2.5 +/- 1.0 psig to 0 to 3.5 psig. This will keep the upper limit the same, but will reduce the lower pressure limit.
Justification:
Thermal properties of backfilled gases (helium or air) used for design basis thermal evaluation of the MP197HB with the RWC are based on its properties at atmospheric pressure (0 psig) as listed in MP197 SAR Section A.3.2.1, Item 15 (helium) and Item 16 (air). The backfilled pressure range between 0 to 3.5 psig does not reduce any thermal conductivity of the backfilled gases (helium or air) in the RWC.
Therefore, the proposed change in the backfill pressure range has no impact on the design basis thermal evaluation in the SAR.
Chapter A.8 Acceptance Tests and Maintenance Instructions
8.1 Sections
A.8.1.10 A.8.2.3.1a TLC 719302-0008
==
Description:==
SAR Table A.2.13.13-1 includes an alternative to the NB-3122.1 code requirement that no structural strength shall be attributed to cladding. Add the requirement in the code alternative for a periodic UT inspection of the cladding to Chapter A.8 of the MP197HB SAR (in new Section A.8.2.3.1a). Also, add the fabrication requirement to Section A.8.1 of the MP197HB SAR (in new Section A.8.1.10).
Justification:
The code alternative in Table A.2.13.13-1 requires specific testing during fabrication as well as a periodic UT inspection of the cask cladding. These requirements are added to the cask fabrication and maintenance requirements specified in Chapter A.8 to ensure that the code alternative specified the Chapter A.2 - Structural Evaluation is implemented in the periodic inspections. Details of the inspection plan are not included in Chapter A.8 to allow flexibility in implementing the inspections in more detailed procedures based on sampling plan for any both weld-overlay and bonded plate cladding.
8.2 Page:
A.8-15 TLC 719302-0015
==
Description:==
Change Each component individually to The sum of all components.
Justification:
ANSI N14.5 requires the summation of all individual leakage rates when the criterion is not leak-tight.
to E-57283 SAR Changed Pages (Proprietary)
Withheld Pursuant to 10 CFR 2.390 to E-57283 NUHOMS-MP197HB SAR, Revision 21A, Changed Pages and Drawings (Public Version)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.9A-1 Appendix A.1.4.9A Radioactive Waste Canister NOTE: References in this appendix are shown as [1], [2], etc., and refer to the reference list in Section A.1.4.9A.3.
A.1.4.9A.1 Radioactive Waste Canister Description The radioactive waste canister (RWC) is designed to contain dry irradiated and/or contaminated non-fuel-bearing solid materials (described further in paragraph A.1.4.9A.2). Under normal transport conditions, the canister rests on four canister rails, attached to the inside surface of an inner sleeve of the NUHOMS-MP197HB transport cask. The RWC is designed to transport its payload dry and in an air or inert gas environment. When a wet-load procedure (i.e., in-pool) is followed for cask loading, the RWC and transport cask cavities are drained and dried in order to ensure that free liquids do not remain in the package during transport. The heat generated by the contents of the RWC is transferred through the transport cask to the environment by conduction, convection and radiation. No forced cooling is required.
Each RWC assembly consists of a cylindrical shell, top shield plug, outer top cover plate, bottom shield plug, and outer bottom cover plate. As shown in Table A.1.4.9A-1, the RWC system consists of five design configurations:
- Welded Design (RWC-W)
- Bolted Design (RWC-B)
- Alternate Bolted Top Shield Plug Design (Alternate RWC-B)
- Dismantling and Decommissioning Design (RWC-DD)
- Alternate RWC design with minimum 0.5-inch steel shell thickness Table A.1.4.9A-1 provides the overall dimensions for each RWC configuration. The details of each configuration are included in the drawings contained in Section A.1.4.10.11 of Appendix A.1.4.10.
The RWC assembly is constructed of steel materials with welded or bolted configurations that provide for handling the contents and biological shielding. The RWC assembly provides a minimum steel thickness of 1.75 inches in the radial direction. The RWC assembly provides a minimum steel thickness of 5.75 inches below the payload and a minimum steel thickness of 7.00 inches above the payload in the axial directions. Local thin spots of the RWC resulting in a reduced thickness is acceptable provided it meets the allowable stress limits listed in Appendix A.2.13.7. An alternate RWC design with minimum 0.5-inch steel shell thickness and similar minimum steel thicknesses in axial directions may be used with the MP197HB Unit 01.
Material properties are listed in Chapter A.2, Table A.2-4. All internal structural components and payloads are the same or similar alloys of stainless steel or carbon steel. These materials are not subject to chemical or galvanic interaction. No hydrogen gas generation is induced by chemical, galvanic, thermal, or radiolytic reactions.
All RWC welding procedures, welders, and welding are performed in accordance with the requirements of AWS D1.1 [1] and AWS D1.6 [2]. All inspections are performed in accordance with AWS D1.1 [1] and AWS D1.6 [2]. As an alternative to AWS D1.1 or D1.6 requirements, it is acceptable for the RWC-W to be welded to ASME Section IX requirements and inspected to ASME Section V requirements.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.9A-2 A.1.4.9A.1.1 Welded Top Shield Plug Design (RWC-W)
The RWC-W shell assembly is a steel welded configuration that provides confinement of radioactive materials, encapsulates the contents in an air or inert atmosphere, and provides biological shielding. The RWC-W shell has redundant seal welds that join the shell and the top and bottom cover plate assemblies to seal the canister. The bottom end assembly welds are made during fabrication of the RWC-W shell. The top end closure welds are made after content loading. Both top plug penetrations (siphon and vent ports) are sealed after the RWC-W drying and backfilling operations are complete.
An inner liner assembly that is used with the RWC-W is a steel welded cylinder with a bottom plate. The bottom plate is designed with drain holes on the bottom to allow liquid from the inner liner to drain into the bottom of the RWC for dewatering. Lifting attachments are provided on the inner liner for lifting the inner liner either empty or loaded. The lifting attachments are designed, fabricated and tested to the requirements of ANSI N14.6 [3]. The inner liner is manufactured with a keyway for alignment in the outer RWC-W canister.
The combined thickness of the RWC-W cylindrical shell and liner is 1.75 inches.
A.1.4.9A.1.2 Bolted Top Shield Plug Design (RWC-B)
The RWC-B shell and bottom are the same as the RWC-W, except the RWC-B provides an option for a bolted top shield plug and does not utilize an inner liner. The bolted top shield plug allows for multiple loadings. Both the top shield plug and outer lid are seal welded after final loading. The combined thickness of the RWC-B cylindrical shell and liner is 1.75 inches.
A.1.4.9A.1.3 Dismantling and Decommissioning Design (RWC-DD)
The RWC-DD is a variant of the RWC design configurations that are approved for use with the MP197HB. The RWC-DD canister nominal diameter is the same as the approved RWC configurations, but longer in length. The RWC-DD is intended to be a reusable canister for transport of secondary liners containing waste or single use for disposal of low level waste in shallow earth disposal sites. The RWC-DD configuration is not intended for extended dry storage like the RWC -W and RWC-B. The RWC-DD shield plug and outer lid are an integral bolted construction without seal welding an outer lid. This bolted outer lid configuration allows reuse of the canister for loading, transport, and unloading for disposal of contents resulting from decommissioning activities. Threaded inserts are allowed at the bolt locations for the outer lid of the RWC-DD.
A.1.4.9A.1.4 Alternate Bolted Top Shield Plug Design (Alternate RWC-B)
The Alternate RWC-B is a variant of the RWC-B design configuration for use with the MP197HB. The Alternate RWC-B canister nominal diameter is the same as the approved RWC configurations and the length falls within the 197-inch max limit. The RWC-B is intended to be used for disposal of low level waste; i.e., irradiated control rod blades and stellite bearings. The RWC-B configuration is intended to be stored within a licensed storage system such as an HSM on-site until the RWC is ready to be transported off-site. The alternate RWC-B consists of a welded vessel that maintains the structural configuration of a package containing solid radioactive wastes (specifically irradiated reactor hardware), encapsulates the contents and
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.9A-2a provides biological shielding. The canister consists of a stainless steel outer shell and inner liner. The top and bottom shield plugs provide shielding at the canister ends. The alternate RWC-B has a bolted top cover plate that seals the canister and can be filtered and vented.
A.1.4.9A.1.5 Alternate RWC with 0.5-inch Steel Thickness An alternate RWC design may be used with the MP197HB Unit 01. The alternate RWC is intended to be used for storage of disposal of low level waste in a licensed storage system such as a HSM on-site until the RWC is ready to be transported off-site. The alternate RWC is similar to a DSC with a minimum of 0.5-inch steel thickness, the minimum steel thicknesses in axial directions are similar to those of RWC-W, RWC-B, RWC-DD and Alternate RWC-B.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.9A-3 A.1.4.9A.2 RWC Contents The NUHOMS-MP197HB packaging is designed to transport a payload of up to 56.0 tons of dry irradiated and/or contaminated non-fuel bearing solid materials in the RWC. The safety analysis of the cask takes no credit for the containment provided by the RWC.
The quantity of radioactive material is limited to a maximum of 90,000 Ci of cobalt-60 or equivalent, except for MP197HB Unit 01 where the limit is reduced to 70,000 Ci of cobalt-60 or equivalent. Equivalent activity limits as a function of gamma energy for isotopes other than Co-60 are shown in Table A.7-2c for the 90,000 Ci limit and Table A.7-2d for the 70,000 Ci limit. The quantity of radioactive material is limited to 9,000 Ci of cobalt-60 or equivalent in the alternate RWC design with minimum 0.5-inch steel shell thickness in MP197HB Unit 01.
Equivalent activity limits as a function of gamma energy for isotopes other than Co-60 are shown in Table A.7-2e. The radioactive material is typically in the form of neutron activated metals, or metal oxides in solid form. Surface contamination may also be present on the irradiated components. When a wet-load procedure (i.e., in-pool) is followed for cask loading, the cask cavity and RWC are drained and dried to ensure that there are no free liquids in the package during transport.
The RWC shall contain dry irradiated and/or contaminated nonfuel bearing solid materials. The dry irradiated and/or contaminated non-fuel bearing solid materials whose total RWC payload meets concentration requirements as low level radioactive waste (LLRW) per 10 CFR 61.55.
Waste characterization per 10 CFR 61.55 is the basis for demonstrating compliance with activity limits for transportation. The contents will not include liquid wastes, sludge or resins. Waste containing fissile material is acceptable provided the quantity of fissile material is limited such that it can be exempted from being classified as fissile material per 10 CFR 71.15 (e.g., fission chambers for in-core detectors). Waste containing organic material is acceptable provided that gas generation from water and organic materials does not lead to exceeding the internal pressure limits, potential flammability of explosive conditions, including the formation of corrosive constituents from radiolysis, biodegradation, or chemical reaction. Except potential flammability, the total gas generation from water and organic materials also affects maximum normal operating pressure (MNOP).
A.1.4.9A.2.1 Type and Form of Material The NUHOMS-MP197HB packaging is designed for shipment of various types of irradiated and contaminated reactor hardware. The payload will vary from shipment to shipment. Typical composition of the payload consists of the following components either individually or in combinations:
The typical cobalt-60 specific activity ranges for these items are as follows:
- 1. BWR Control Rod Blades 1.3x10 1.1x10-2 Ci/g
- 3. BWR Fuel Channels 7.8x10 2.0x10-4 Ci/g
- 4. BWR Poison Curtains 6.2x10 4.0x10-2 Ci/g
- 5. PWR Burnable Poison Rod Assemblies (BPRAs) 3.8x10 1.3x10-3 Ci/g
- 6. BWR and PWR Reactor Vessel and Internals 2.0x10 1.3x10-2 Ci/g A.1.4.9A.2.2 Decay Heat load The RWC heat load does not exceed 5kw, well below the limit of 26 kW limit for the DSC contents in MP197HB without external cooling fins.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.9A-5 Table A.1.4.9A-1 Nominal Dimensions of the RWC RWC Design Parameters RWC-W RWC-B RWC-DD Alt. RWC-B Alt. RWC Shell Thickness (in.)
1.75(1) 1.75 1.75 1.75(2) 0.5 Canister Length (in.)
186.50 or 196 186.50 196 196 196.3(3)
Outside Diameter (in.)
67.19 67.19 67.25 67.25 67.19(3)
Cavity Length (in.)(3) 167.30 (min) 167.30 183.25 183.25 180.2 Cavity Diameter (in.)(3) 64.69 (min) 63.69 63.75 63.75 66.19 Note:
(1) The shell thickness for the RWC-W is 1.75 inches, which includes the RWC-W inner liner thickness of 0.50 inches.
(2) The shell thickness for the Alt. RWC-B is 1.75 inches, which include the Alt. RWC-B inner liner thickness of 1.25 inches and outer shell thickness of 0.50 inches.
(3) Dimensions may vary depending on the shell and cover plate thicknesses.
Table A.1.4.9A-2 Nominal Dimensions of the RWC Inner Liner RWC-W Inner Liner Design Parameters(1)
Shell Thickness (in.)
0.50 (min)
Outside Length (in.)
166.30 Outside Diameter (in.)
63.69 Cavity Length (in.)
162.11 Cavity Diameter (in.)
62.69 Note:
(1) Dimensions may vary depending on the shell and cover plate thicknesses.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-i Appendix A.1.4.10 Drawings of Transport Packaging and DSCs TABLE OF CONTENTS A.1.4.10.1 NUHOMS-MP197HB Drawings........................................................................ A.1.4.10-1 A.1.4.10.2 NUHOMS 24PT4 DSC Drawings...................................................................... A.1.4.10-2 A.1.4.10.3 NUHOMS 32PT DSC Drawings........................................................................ A.1.4.10-3 A.1.4.10.4 NUHOMS 24PTH DSC Drawings..................................................................... A.1.4.10-4 A.1.4.10.5 NUHOMS 32PTH DSC Drawings..................................................................... A.1.4.10-5 A.1.4.10.6 NUHOMS 32PTH1 DSC Drawings................................................................... A.1.4.10-7 A.1.4.10.7 NUHOMS 37PTH DSC Drawings..................................................................... A.1.4.10-8 A.1.4.10.8 NUHOMS 61BT DSC Drawings........................................................................ A.1.4.10-9 A.1.4.10.9 NUHOMS 61BTH DSC Drawings................................................................... A.1.4.10-10 A.1.4.10.10 NUHOMS 69BTH DSC Drawings................................................................... A.1.4.10-11 A.1.4.10.11 Radioactive Waste Canister Drawing................................................................ A.1.4.10-12
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-1 A.1.4.10.1 NUHOMS-MP197HB Drawings Drawing Number Title MP197HB-71-1001 Rev 5 NUHOMS-MP197HB Packaging Transport Configuration (2 sheets)
MP197HB-71-1002 Rev 10 NUHOMS-MP197HB Packaging Parts List (2 sheets)
MP197HB-71-1003 Rev 4 NUHOMS-MP197HB Packaging General Arrangement (1 sheet)
MP197HB-71-1004 Rev 7 NUHOMS-MP197HB Packaging Cask Body Assembly (1 sheet)
MP197HB-71-1005 Rev 10 NUHOMS-MP197HB Packaging Cask Body Details (3 sheets)
MP197HB-71-1006 Rev 5 NUHOMS-MP197HB Packaging Lid Assembly and Details (1 sheet)
MP197HB-71-1008 Rev 5 NUHOMS-MP197HB Packaging Impact Limiter Assembly (1 sheet)
MP197HB-71-1009 Rev 5 NUHOMS-MP197HB Packaging Impact Limiter Details (1 sheet)
MP197HB-71-1011 Rev 1 NUHOMS-MP197HB Packaging Transport Configuration Outer Sleeve With Fins Option (1 sheet)
MP197HB-71-1014 Rev 3 NUHOMS-MP197HB Packaging Internal Sleeve Design (1 sheet)
Proprietary and Security Related Information for Drawing MP197HB-71-1002, Rev. 10A Withheld Pursuant to 10 CFR 2.390
Proprietary and Security Related Information for Drawing MP197HB-71-1003, Rev. 4A Withheld Pursuant to 10 CFR 2.390
Proprietary and Security Related Information for Drawing MP197HB-71-1005, Rev. 10A Withheld Pursuant to 10 CFR 2.390
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-2 A.1.4.10.2 NUHOMS 24PT4 DSC Drawings Drawing Number Title NUH24PT4-71-1001 Rev 0 NUHOMS 24PT4 Transportable Canister For PWR Fuel Basket Assembly (5 sheets)
NUH24PT4-71-1002 Rev 0 NUHOMS 24PT4 Transportable Canister For PWR Fuel Main Assembly (8 sheets)
NUH24PT4-71-1003 Rev 0 NUHOMS 24PT4 Transportable Canister For PWR Fuel Failed Fuel Can (4 sheets)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-3 A.1.4.10.3 NUHOMS 32PT DSC Drawings Drawing Number Title NUH32PT-71-1000 Rev 0 NUHOMS 32PT Transportable Canister For PWR Fuel Summary Dimensions (1 sheet)
NUH32PT-71-1001 Rev 1 NUHOMS 32PT Transportable Canister For PWR Fuel Main Assembly (5 sheets)
NUH32PT-71-1002 Rev 1 NUHOMS 32PT Transportable Canister For PWR Fuel Shell Assembly (3 sheets)
NUH32PT-71-1003 Rev 1 NUHOMS 32PT Transportable Canister For PWR Fuel A Basket Assembly (16 Poison/16 Compartment Plates) (8 sheets)
NUH32PT-71-1004 Rev 1 NUHOMS 32PT Transportable Canister For PWR Fuel Aluminum Transition Rail - R90 (2 sheets)
NUH32PT-71-1005 Rev 1 NUHOMS 32PT Transportable Canister For PWR Fuel Aluminum Transition Rail - R45 (1 sheet)
NUH32PT-71-1006 Rev 1 NUHOMS 32PT Transportable Canister For PWR Fuel A/B/C/D Basket Assembly (20 Poison/12 Compartment Plates) (6 sheets)
NUH32PT-71-1007 Rev 1 NUHOMS 32PT Transportable Canister For PWR Fuel A/B/C/D Basket Assembly (24 Poison/8 Compartment Plates) (8 sheets)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-4 A.1.4.10.4 NUHOMS 24PTH DSC Drawings Drawing Number Title NUH24PTH-71-1000 Rev 1 NUHOMS 24PTH Transportable Canister For PWR Fuel Main Assembly (5 sheets)
NUH24PTH-71-1001 Rev 1 NUHOMS 24PTH Transportable Canister For PWR Fuel Basket-Shell Assembly (4 sheets)
NUH24PTH-71-1002 Rev 1 NUHOMS 24PTH Transportable Canister For PWR Fuel Shell Assembly (4 sheets)
NUH24PTH-71-1003 Rev 2 NUHOMS 24PTH Transportable Canister For PWR Fuel Basket Assembly (8 sheets)
NUH24PTH-71-1004 Rev 1 NUHOMS 24PTH Transportable Canister For PWR Fuel Transition Rails (4 sheets)
NUH24PTH-71-1008 Rev 1 NUHOMS 24PTHF Transportable Canister For PWR Fuel Failed Fuel Can (2 sheets)
NUH24PTH-71-1009 Rev 1 NUHOMS 24PTHF Transportable Canister For PWR Fuel Basket Assembly (8 sheets)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-5 A.1.4.10.5 NUHOMS 32PTH DSC Drawings Drawing Number Title NUH32PTH-71-1001 Rev 2 NUHOMS32PTH Transportable Canister for PWR Fuel Parts List (1 Sheet)
NUH32PTH-71-1002 Rev 1 NUHOMS32PTH Transportable Canister for PWR Fuel Main Assembly (1 Sheet)
NUH32PTH-71-1003 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Siphon Pipe Details (1 Sheet)
NUH32PTH-71-1004 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Inner Top Cover Details (2 sheets)
NUH32PTH-71-1005 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Outer Top Cover Details (1 Sheet)
NUH32PTH-71-1006 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Shell Assembly (1 Sheet)
NUH32PTH-71-1007 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Shell Bottom Details (1 Sheet)
NUH32PTH-71-1008 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Grapple Ring Details (1 Sheet)
NUH32PTH-71-1009 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Basket Assembly (1 Sheet)
NUH32PTH-71-1010 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Basket Assembly Details (1 Sheet)
NUH32PTH-71-1011 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Basket Assembly Details (1 Sheet)
NUH32PTH-71-1012 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Basket Assembly - Details (1 Sheet)
NUH32PTH-71-1013 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Basket Rail A180 (1 Sheet)
NUH32PTH-71-1014 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Basket Rail A90 (1 Sheet)
NUH32PTH-71-1015 Rev 0 NUHOMS32PTH Transportable Canister for PWR Fuel Damaged Fuel End Caps (1 Sheet)
NUH32PTH Type 1-71-1000 Rev 1 NUHOMS 32PTH Type 1 Transportable Canister For PWR Fuel Main Assembly (4 sheets)
NUH32PTH Type 1-71-1001 Rev 2 NUHOMS 32PTH Type 1 Transportable Canister For PWR Fuel Basket Shell Assembly (4 sheets)
NUH32PTH Type 1-71-1002 Rev 1 NUHOMS 32PTH Type 1 Transportable Canister For PWR Fuel Shell Assembly (4 sheets)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-6 Drawing Number Title NUH32PTH Type 1-71-1003 Rev 2 NUHOMS 32PTH Type 1 Transportable Canister For PWR Fuel Basket Assembly (7 sheets)
NUH32PTH Type 1-71-1004 Rev 2 NUHOMS 32PTH Type 1 Transportable Canister For PWR Fuel Transition Rails (4 sheets)
NUH32PTH Type 1-71-1010 Rev 1 NUHOMS 32PTH Type 1 Transportable Canister For PWR Fuel Alternate Top Closure (6 sheets)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-7 A.1.4.10.6 NUHOMS 32PTH1 DSC Drawings Drawing Number Title NUH32PTH1-71-1000 Rev 1 NUHOMS 32PTH1 Transportable Canister For PWR Fuel Main Assembly (4 sheets)
NUH32PTH1-71-1001 Rev 1 NUHOMS 32PTH1 Transportable Canister For PWR Fuel Basket Shell Assembly (5 sheets)
NUH32PTH1-71-1002 Rev 1 NUHOMS 32PTH1 Transportable Canister For PWR Fuel Shell Assembly (4 sheets)
NUH32PTH1-71-1003 Rev 2 NUHOMS 32PTH1 Transportable Canister For PWR Fuel Basket Assembly (8 sheets)
NUH32PTH1-71-1004 Rev 1 NUHOMS 32PTH1 Transportable Canister For PWR Fuel Transition Rails (7 sheets)
NUH32PTH1-71-1010 Rev 1 NUHOMS 32PTH1 Transportable Canister For PWR Fuel Alternate Top Closure (6 sheets)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-8 A.1.4.10.7 NUHOMS 37PTH DSC Drawings Drawing Number Title NUH37PTH-71-1001 Rev 2 NUHOMS 37PTH Transportable Canister For PWR Fuel Main Assembly (4 sheets)
NUH37PTH-71-1002 Rev 3 NUHOMS 37PTH Transportable Canister For PWR Fuel Basket Shell Assembly (5 sheets)
NUH37PTH-71-1003 Rev 3 NUHOMS 37PTH Transportable Canister For PWR Fuel Shell Assembly (4 sheets)
NUH37PTH-71-1004 Rev 3 NUHOMS 37PTH Transportable Canister For PWR Fuel Alternate Top Closure (6 sheets)
NUH37PTH-71-1011 Rev 2 NUHOMS 37PTH Transportable Canister For PWR Fuel Basket Assembly (7 sheets)
NUH37PTH-71-1012 Rev 1 NUHOMS 37PTH Transportable Canister For PWR Fuel Transition Rails (7 sheets)
NUH37PTH-71-1015 Rev 0 NUHOMS 37PTH Transportable Canister For PWR Fuel Damaged Fuel End Caps (1 sheet)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-9 A.1.4.10.8 NUHOMS 61BT DSC Drawings Drawing Number Title NUH61BT-71-1000 Rev 1 NUHOMS 61BT Transportable Canister For BWR Fuel Parts List (1 sheet)
NUH61BT-71-1001 Rev 1 NUHOMS 61BT Transportable Canister For BWR Fuel Basket Assembly (1 sheet)
NUH61BT-71-1002 Rev 0 NUHOMS 61BT Transportable Canister For BWR Fuel Basket Details (1 sheet)
NUH61BT-71-1003 Rev 0 NUHOMS 61BT Transportable Canister For BWR Fuel General Assembly (1 sheet)
NUH61BT-71-1004 Rev 0 NUHOMS 61BT Transportable Canister For BWR Fuel General Assembly (1 sheet)
NUH61BT-71-1005 Rev 0 NUHOMS 61BT Transportable Canister For BWR Fuel Shell Assembly (1 sheet)
NUH61BT-71-1006 Rev 0 NUHOMS 61BT Transportable Canister For BWR Fuel Shell Assembly (1 sheet)
NUH61BT-71-1007 Rev 0 NUHOMS 61BT Transportable Canister For BWR Fuel Canister Details (1 sheet)
NUH61BT-71-1008 Rev 0 NUHOMS 61BT Transportable Canister For BWR Fuel Canister Details (1 sheet)
NUH61BT-71-1009 Rev 0 NUHOMS 61BT Transportable Canister For BWR Fuel Basket Details (1 sheet)
NUH61BT-71-1010 Rev 1 NUHOMS 61BT Transportable Canister For BWR Fuel Additional Basket Details - Damaged Fuel (4 sheets)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-10 A.1.4.10.9 NUHOMS 61BTH DSC Drawings Drawing Number Title NUH61BTH-71-1000 Rev 1 NUHOMS 61BTH Type 1 Transportable Canister For BWR Fuel Main Assembly (5 sheets)
NUH61BTH-71-1100 Rev 2 NUHOMS 61BTH Type 2 Transportable Canister For BWR Fuel Main Assembly (7 sheets)
NUH61BTH-71-1101 Rev 1 NUHOMS 61BTH Type 2 Transportable Canister For BWR Fuel Shell Assembly (2 sheets)
NUH61BTH-71-1102 Rev 2 NUHOMS 61BTH Type 2 Transportable Canister For BWR Fuel Basket Assembly (8 sheets)
NUH61BTH-71-1103 Rev 1 NUHOMS 61BTH Type 2 Transportable Canister For BWR Fuel Transition Rails (2 sheets)
NUH61BTH-71-1104 Rev 1 NUHOMS 61BTH Type 2 Transportable Canister For BWR Fuel Damaged Fuel End Caps (1 sheet)
NUH61BTH-71-1105 Rev 1 NUHOMS 61BTHF Type 2 Transportable Canister For BWR Fuel Failed Fuel Can (2 sheets)
NUH61BTH-71-1106 Rev 2 NUHOMS 61BTH Type 2 Transportable Canister For BWR Fuel Top Grid Assembly Alternate 3 (2 sheets)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-11 A.1.4.10.10 NUHOMS 69BTH DSC Drawings Drawing Number Title NUH69BTH-71-1001 Rev 3 NUHOMS 69BTH Transportable Canister For BWR Fuel Main Assembly (4 sheets)
NUH69BTH-71-1002 Rev 3 NUHOMS 69BTH Transportable Canister For BWR Fuel Basket -
Shell Assembly (4 sheets)
NUH69BTH-71-1003 Rev 3 NUHOMS 69BTH Transportable Canister For BWR Fuel Shell Assembly (4 sheets)
NUH69BTH-71-1004 Rev 6 NUHOMS 69BTH Transportable Canister For BWR Fuel Alternate Top Closure (7 sheets)
NUH69BTH-71-1011 Rev 3 NUHOMS 69BTH Transportable Canister For BWR Fuel Basket Assembly (5 sheets)
NUH69BTH-71-1012 Rev 4 NUHOMS 69BTH Transportable Canister For BWR Fuel Transition Rail Assembly And Details (6 sheets)
NUH69BTH-71-1013 Rev 4 NUHOMS 69BTH Transportable Canister For BWR Fuel Holddown Ring Assembly (2 sheets)
NUH69BTH-71-1014 Rev 2 NUHOMS 69BTH Transportable Canister For BWR Fuel Damaged Fuel Modification (1 sheet)
NUH69BTH-71-1015 Rev 2 NUHOMS 69BTH Transportable Canister For BWR Fuel Damaged Fuel End Caps (1 sheet)
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.1.4.10-12 A.1.4.10.11 Radioactive Waste Canister Drawing Drawing Number Title NUHRWC-71-1001 Rev 6 NUHOMS System Radioactive Waste Canister (2 sheets)
Proprietary and Security Related Information for Drawing NUHRWC-71-1001, Rev. 6A Withheld Pursuant to 10 CFR 2.390
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-i Appendix A.2.13.2 MP197HB Cask Lid Bolt/Ram Access Closure Plate Bolt Analyses TABLE OF CONTENTS A.2.13.2.1 Purpose........................................................................................................ A.2.13.2-1 A.2.13.2.2 Lid Bolt Load Calculations.......................................................................... A.2.13.2-2 A.2.13.2.2.1 Bolt Preload....................................................................................... A.2.13.2-2 A.2.13.2.2.2 Gasket Seating Load.......................................................................... A.2.13.2-3 A.2.13.2.2.3 Internal Pressure Load...................................................................... A.2.13.2-3 A.2.13.2.2.4 Temperature Load.............................................................................. A.2.13.2-4 A.2.13.2.2.5 Impact Load....................................................................................... A.2.13.2-4 A.2.13.2.2.6 Puncture Load.................................................................................... A.2.13.2-5 A.2.13.2.2.7 External Pressure Load of 290 psig................................................... A.2.13.2-5 A.2.13.2.3 Lid Bolt Load Combinations........................................................................ A.2.13.2-6 A.2.13.2.3.1 Additional Prying Bolt Force............................................................. A.2.13.2-6 A.2.13.2.3.2 Bending Moment Bolt Force.............................................................. A.2.13.2-7 A.2.13.2.4 Bolt Stress Calculations............................................................................... A.2.13.2-8 A.2.13.2.4.1 Average Tensile Stress....................................................................... A.2.13.2-8 A.2.13.2.4.2 Bending Stress.................................................................................... A.2.13.2-9 A.2.13.2.4.3 Shear Stress........................................................................................ A.2.13.2-9 A.2.13.2.4.4 Maximum Combined Stress Intensity............................................... A.2.13.2-10 A.2.13.2.4.5 Stress Ratios..................................................................................... A.2.13.2-10 A.2.13.2.4.6 Bearing Stress Under Bolt Head...................................................... A.2.13.2-11 A.2.13.2.5 Results........................................................................................................ A.2.13.2-11 A.2.13.2.6 Lid Bolt Fatigue Analysis........................................................................... A.2.13.2-11 A.2.13.2.6.1 Vibration/Shock................................................................................ A.2.13.2-12 A.2.13.2.6.2 Damage Factor Calculation............................................................ A.2.13.2-12 A.2.13.2.7 Lid Seal Contact Evaluation...................................................................... A.2.13.2-12 A.2.13.2.7.1 Assumptions...................................................................................... A.2.13.2-13 A.2.13.2.7.2 Analysis............................................................................................ A.2.13.2-13 A.2.13.2.7.3 Results.............................................................................................. A.2.13.2-13 A.2.13.2.8 Minimum Engagement Length for Bolt and Flange.................................. A.2.13.2-13 A.2.13.2.9 Ram Closure Plate Bolts Analysis............................................................. A.2.13.2-15 A.2.13.2.9.1 Type 1 Ram Closure Plate Bolt Load Calculations......................... A.2.13.2-16 A.2.13.2.9.1a Type 2 Ram Closure Plate Bolt Load Calculations......................... A.2.13.2-20 A.2.13.2.9.2 Type 1 Ram Closure Plate Bolts Load Combinations.................... A.2.13.2-20a A.2.13.2.9.2a Type 2 Ram Closure Plate Bolts Load Combinations...................... A.2.13.2-22 A.2.13.2.9.3 Type 1 Ram Closure Plate Bolt Stress Calculations...................... A.2.13.2-22b A.2.13.2.9.3a Type 2 Ram Closure Plate Bolt Stress Calculations........................ A.2.13.2-24 A.2.13.2.9.4 Type 1 Ram Closure Plate Bearing Stress under Bolt Head.......... A.2.13.2-24b A.2.13.2.9.4a Type 2 Ram Closure Plate Bearing Stress under Bolt Head.......... A.2.13.2-24b A.2.13.2.9.5 Results............................................................................................. A.2.13.2-24b
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-ii A.2.13.2.9.6 Minimum Engagement Length for Bolt and Flange........................ A.2.13.2-24c A.2.13.2.10 Conclusions................................................................................................ A.2.13.2-26 A.2.13.2.11 References.................................................................................................. A.2.13.2-27 LIST OF TABLES Table A.2.13.2-1 Design Parameters for Bolts Analysis................................................ A.2.13.2-28 Table A.2.13.2-2 Bolt Data............................................................................................. A.2.13.2-29 Table A.2.13.2-3 Allowable Stresses in Closure Bolts for Normal Conditions.............. A.2.13.2-30 Table A.2.13.2-4 Allowable Stresses in Closure Bolts for Accident Conditions............ A.2.13.2-31 Table A.2.13.2-5 Lid Bolts Individual Summary............................................................. A.2.13.2-32 Table A.2.13.2-6 Lid Bolts Normal and Accident Load Combinations.......................... A.2.13.2-33 Table A.2.13.2-7 Lid and Ram Closure Plate Bolt Stresses........................................... A.2.13.2-34 Table A.2.13.2-8 Damage Factor Calculation............................................................... A.2.13.2-35 Table A.2.13.2-9 Type 1 Ram Closure Plate Bolts Individual Summary........................ A.2.13.2-36 Table A.2.13.2-9a Type 2 Ram Closure Plate Bolts Individual Summary.................... A.2.13.2-36a Table A.2.13.2-10 Type 1 Ram Closure Plate Bolts Normal and Accident Load Combinations.................................................................................. A.2.13.2-37 Table A.2.13.2-11 Type 2 Ram Closure Plate Bolts Normal and Accident Load Combinations................................................................................ A.2.13.2-37a LIST OF FIGURES Figure A.2.13.2-1 MP197HB Transport Cask (CG Over Corner Lid Drop-Hot) Lid Seal Decompression as a Function of Circumferential Location... A.2.13.2-38 Figure A.2.13.2-2 Deformation Plot Near Lid-Flange Interface................................ A.2.13.2-39
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-1 Appendix A.2.13.2 MP197HB Cask Lid Bolt/Ram Access Closure Plate Bolt Analyses NOTE: References in this appendix are shown as [1], [2], etc. and refer to the reference list in Section A.2.13.2.11.
A.2.13.2.1 Purpose This appendix analyzes the ability of the cask closure bolts to maintain a leak-tight seal under events defined by normal conditions of transport (NCT) and the hypothetical accident conditions (HAC). Also evaluated in this section are the stresses in the bolt threads and in the internal threads, and the lid bolt fatigue. The stress analysis is performed in accordance with NUREG/CR-6007 [1].
Appendix A.1.4.10 contains reference drawings for the lid and ram access cover plate bolts.
The closure lid has a diameter of 77.18 in. and consists of a 4.50 in. thick plate with a 3.94 in.
thick outer flange. The lid is bolted directly to the end of the containment vessel flange by 48 high-strength alloy steel 1.5 in. diameter bolts on a 74.81 in. diameter bolt circle. Close fitting alignment pins ensure that the lid is centered in the vessel. The bolts material is SA-540 Gr. B23 Cl. 1 or Gr B24 Cl. 1, which has a yield strength of 139.1 ksi and a tensile strength of 165.0 ksi at 350 °F.
The bolt material (SA-540 Gr. B23 Cl. 1 or Gr B24 Cl. 1) used to fabricate the lid bolt will have a minimum tensile strength of 175 ksi as specified in the SAR drawing (Dwg No. MP197HB 1002, sheet 2 of 2, rev. 2, note 16). Fine thread series will be used (11/2-12 UNF, Dwg No.
MP197HB-71-1002, sheet 1 of 2, rev. 2, item 21), which has a tensile stress area of 1.58 in2.
The lid bolt analysis presented in this appendix is in accordance with NUREG/CR-6007 and conservatively uses a 11/2-6 UNC (coarse thread) lid bolt with a tensile strength of 165 ksi. The lid bolt evaluation due to delayed impact is presented in Appendix A.2.13.14.
The following ways to minimize bolt forces and bolt failures for shipping casks are taken directly from [1], page xiii. All of the following design methods are employed in the NUHOMS-MP197HB closure system:
Protect closure lid from direct impact to minimize bolt forces generated by free drops (use impact limiters).
Use materials with similar thermal properties for the closure bolts, the lid, and the cask wall to minimize the bolt forces generated by a fire accident.
Apply sufficiently large bolt preload to minimize fatigue and loosening of the bolts by vibration.
Lubricate bolt threads to reduce required preload torque and to increase the predictability of the achieved preload.
Use closure lid design which minimizes the prying actions of applied loads.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-15 Therefore:
(
)
2 in 750
.2 3500
.1 3772
.1 57735
.0 6
2 1
3500
.1 09
.1 6
1416
.3
=
x
+
x x
x x
x
=
s A
(
)
2 in 616
.3 4075
.1 4703
.1 57735
.0 6
2 1
4703
.1 09
.1 6
1416
.3
=
x
+
x x
x x
x
=
n A
So:
79
.1 0.
70 616
.3 0.
165 750
.2
=
x x
=
J Therefore, the minimum required engagement length Q = J x Le = 1.79 x 1.09 = 1.96 in.
The actual minimum engagement length is equal to:
5.00 (bolt length) - 2.24 (thickness of the closure lid under the screw head) - 0.17 (washer thickness) = 2.59 in > 2.25 in (length of lid bolt insert) > 1.96 in.
The above calculation bounds the minimum required engagement length if inserts are used because Su of inserts is higher than the Su for the lid, thus lowering the J value.
A.2.13.2.9 Ram Closure Plate Bolts Analysis This section analyzes the ability of the ram closure plate bolts to maintain a leak-tight seal under events defined by Normal Conditions of Transport (NCT) and the Hypothetical Accident Conditions (HAC). Also evaluated in this section are the stresses in the bolt threads and in the internal threads. The stress analysis is performed in accordance with NUREG/CR-6007 [1].
The ram closure plate has a diameter of 28.88 in. and consists of a 5.00 in. thick plate with a 2.50 in. thick outer flange. The ram closure plate is bolted directly to the bottom of the containment vessel by 12 high-strength alloy steel 1 in. diameter bolts on a 27.00 in. diameter bolt circle.
The bolts material is SA-540 Gr. B23 Cl. 1 or Gr B24 Cl. 1, which has a yield strength of 137.9 ksi and a tensile strength of 165.0 ksi at 400°F. The ram closure plate is made of SA-203 Gr. E (type 1), or either SA-203 Gr. E, SA-240 Type 304, or SA-182 Type F304 (type 2). SA-240 Type 304 or SA-182 Type F304 is conservatively considered for the type 2 ram closure plate evaluation.
The following evaluations are made in this section:
- Bolt preload
- Gasket seating load
- Internal pressure loads
- Temperature load
- Impact load
- Puncture load
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-16
- External pressure loads
- Load combinations for normal and accident conditions
- Bolt stresses and allowable stresses
- Thread engagement length evaluation
- Bearing stress The design parameters for the ram closure plate analysis taken from [1] are summarized in Table A.2.13.2-1. The ram closure plate bolt data and material allowables are presented in Tables A.2.13.2-2 to A.2.13.2-4. A temperature of 400°F (type 1) and 225°F (type 2) is used in the bolts region during NCT and HAC based on results of thermal analyses.
The following load cases are considered in the analysis:
- 1. Preload + Temperature Load (NCT);
- 2. Gasket Seating Load + Internal Pressure
- 3. Internal Pressure + 30 Foot Corner Drop (HAC);
- 4. Internal Pressure + Puncture Load (HAC).
A.2.13.2.9.1 Type 1 Ram Closure Plate Bolt Load Calculations Bolt Preload The method of analysis is described in Table 4.1 of [1].
A bolt torque range of 225 to 250 ft.lb is required to ensure leak tightness against normal and accident loadings.
Bolt preload for the minimum torque is:
lb/bolt 000 20 0.1 135
.0 12 225 D
K Q
F b
a
=
x x
=
x
=
Bolt preload for the maximum torque is:
lb/bolt 222 22 0.1 135
.0 12 250 D
K Q
F b
a
=
x x
=
x
=
Residual torsional moment for the minimum torque is:
in.lb/bolt 350
,1 12 225 5.0 Q
5.0 Mtr
=
x x
=
x
=
Residual torsional moment for the maximum torque is:
in.lb/bolt 500
,1 12 250 5.0 Q
5.0 Mtr
=
x x
=
x
=
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-20
(
)
(
)
lb/in 958
,1 4
290 0
00 27 4
=
x
=
x
=
lo li lb f
P P
D F
The fixed edge cover plate moment is:
(
)
(
)
in.lb/in 607
,6 32 00 27 290 0
32 2
2
=
x
=
x
=
lb lo li f
D P
P M
The ram closure plate shoulder takes the shear force, so the shear bolt force per bolt is Fs = 0.
The loads calculated in this section are summarized in Table A.2.13.2-9.
A.2.13.2.9.1a Type 2 Ram Closure Plate Bolt Load Calculations Bolt Preload The method of analysis is described in Table 4.1 of [1].
A bolt torque range of 100 to 125 ft.lb is required to ensure leaktightness against normal and accident loadings. A nut factor range of 0.192 to 0.193 is assumed for Neolube No.1 lubricant for this evaluation.
Bolt preload for the minimum torque is:
lb/bolt 218
,6 0.1 193
.0 12 100
=
x x
=
x
=
b a
D K
Q F
Bolt preload for the maximum torque is:
lb/bolt 7,813 0.1 192
.0 12 125
=
x x
=
x
=
b a
D K
Q F
Residual torsional moment for the minimum torque is:
in.lb/bolt 600 12 100 5
0 5
0
=
x x
=
x
=
Q Mtr Residual torsional moment for the maximum torque is:
in.lb/bolt 750 12 125 5
0 5
0
=
x x
=
x
=
Q Mtr Residual tensile bolt force:
Far = Fa = 7,813 lb/bolt.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-20a Gasket Seating Load An elastomer o-ring is used; therefore, the gasket seating load is negligible, Fs = 0 lb/bolt.
Internal Pressure Load See Section A.2.13.2.9.1.
Temperature Load The analysis is described in Table 4.4 of [1].
The cover plate bolt material is SA-540 Gr. B23 Cl. 1 or Gr B24 Cl. 1, the cask bottom is made of SA-350 Gr. LF3, and the Type 2 ram closure plate is made of SA-240 Type 304. The Type 2 ram closure plate is used for RWC contents that are limited to 5 kW; therefore, the temperature of the materials at the bolt location is assumed to be 225 °F for this evaluation.
The coefficient of thermal expansion of the bolts at 225 °F is 6.8 x 10-6 in/in/°F, and the coefficient of thermal expansion of the ram closure plate is 9.0 x 10-6 in/in/°F.
It is assumed that the initial temperature of the cask body is at room temperature (70 °F) and final temperature as 225 °F, which is conservatively high.
The axial force per bolt Fa due to the thermal expansion difference between the RACP and bolt is:
(
)
b b
l l
b b
a T
T E
D F
a a
x x
x
=
2 25
.0
(
)
)
70 225
(
10 8.6 0.9 10 27 1
25
.0 6
6 2
x x
x x
x x
x
=
Fa = 7,232 lb/bolt.
Impact Load See Section A.2.13.2.9.1.
Puncture Load The puncture load is not considered because of the protection provided by the impact limiters.
External Pressure Load of 290 psig See Section A.2.13.2.9.1.
The loads calculated in this section are summarized in Table A.2.13.2-9.
A.2.13.2.9.2 Type 1 Ram Closure Plate Bolts Load Combinations A summary of normal and accident condition load combinations is presented in Table A.2.13.2-
- 10. The method used for the following combination is taken from [1], Table 4.9.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-20b Additional Prying Bolt Force The analysis is described in Table 2.1 of [1].
Although the methodology developed in [1] applies to full cover plates that extend to the entire diameter of the cask, it can also be applied to the ram closure plate analyzed in this calculation, even though its bolt diameter does not extend all the way to the cask walls. The contents of the cask provide the bottom part of the cask with a rigid seating surface, rigid enough to consider that the ram closure plate is bolted on the walls of a thick cylindrical shell, the inner diameter of which is the diameter of the ram closure plate opening through the cask bottom.
Since the prying forces applied in load case 3 act inward, normal to the cask lid, an additional prying bolt force, Fap, is generated (Table 2.1 of [1]). The additional force generated for the outward loadings is considered negligible since its main source - the contents of the cask -
cannot be in contact with the ram closure plate.
Fap for an inward force is calculated in the following way:
+
x
x
x x
x
=
2 1
2 f
1 lb li f
b lb ap C
C
)
P B
(
C
)
F B
(
C D
D M
2 N
D F
p Where C1 = 1, and:
(
)
(
)
x x
+
x x
x
x
=
lb lf lf li lo ul l
l b
b b
b lb li D
t E
D D
N t
E E
D N
L D
D C
3 3
2 2
2 1
3 8
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-22 5
4 6
4 10 46
.1 64 0.1 00 27 10 2.
26 25
.1 12 64 x
=
x
=
=
b lb b
b b
b D
D E
L N
K lb/bolt
(
) (
)
(
) (
)
7 2
2 2
3 6
lb 2
lo lb 2
ul 2
ul 3
l l
l 10 02
.3 00 27 88 28 00 27 3.0 1
3.0 1
3 00
.5 10 2.
26 D
D D
N 1
N 1
3 t
E K
x
=
+
x x
=
+
x
=
lb/bolt In the case of the internal pressure load, Mf = 683 in.lb.
Therefore:
0 in.lb/bolt 23 683 10 02
.3 10 46
.1 10 46
.1 12 00 27 M
7 5
5 bb
=
x x
+
x x
x x
=
The maximum bending bolt moment is considered equal to 0.
A.2.13.2.9.2a Type 2 Ram Closure Plate Bolts Load Combinations A summary of normal and accident condition load combinations is presented in Table A.2.13.2-
- 10. The method used for the following combination is taken from [1], Table 4.9.
Additional Prying Bolt Force The analysis is described in Table 2.1 of [1].
Although the methodology developed in [1] applies to full cover plates that extend to the entire diameter of the cask, it can also be applied to the ram closure plate analyzed in this calculation, even though its bolt diameter does not extend all the way to the cask walls. The contents of the cask provide the bottom part of the cask with a rigid seating surface, rigid enough to consider that the ram closure plate is bolted on the walls of a thick cylindrical shell, the inner diameter of which is the diameter of the ram closure plate opening through the cask bottom.
Since the prying forces applied in load case 3 act inward, normal to the cask lid, an additional prying bolt force, Fap, is generated (Table 2.1 of [1]). The additional force generated for the outward loadings is considered negligible since its main source - the contents of the cask -
cannot be in contact with the ram closure plate.
Fap for an inward force is calculated in the following way:
+
x
x
x x
x
=
2 1
2 f
1 lb li f
b lb ap C
C
)
P B
(
C
)
F B
(
C D
D M
2 N
D F
p
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-22a Where C1 = 1, and:
(
)
(
)
x x
+
x x
x
x
=
lb lf lf li lo ul l
l b
b b
b lb li D
t E
D D
N t
E E
D N
L D
D C
3 3
2 2
2 1
3 8
(
)
(
)
x
+
x x
x x
x
x
=
00 27 50
.2 00 22 88 28 3.0 1
0.5 10 2.
26 0.1 12 10 2.
26 25
.1 00 27 00 22 3
8 C
3 3
6 2
6 2
2 C2 = 2.03 B is the non-prying tensile bolt force, and P is the bolt preload; both are derived from the load applied to each bolt in normal operating conditions (bolt preload + temperature load + internal pressure load). Conservatively, the fixed-edge closure lid moment is considered for the case of the external pressure load, for which Mf = -6,607 in.lb/in./bolt.
lb/in 128
,2 00 27 12 15,045
=
x x
=
x x
=
lb b
a D
N F
P For all inward loadings such as the external pressure load, Ff is supported by the cask wall and thus has no effect on both the non-prying and prying bolt forces, B and R. Therefore, Ff = 0.
B = Ff if Ff > P and B = P otherwise. Since Ff = 0 < P, B = P = 2,128 lb/in.
Furthermore, the internal pressure load is not included because it decreases the magnitude of the applied prying moment, which is less conservative.
Therefore:
(
)
+
x
x
x x
x
=
03
.2 1
)
128
,2 128
,2
(
03
.2
)
0 128
,2
(
1 00 27 00 22 607
,6 2
12 00 27
a F
Fap = 1200 lb/bolt.
Since this bolt load is less than the load generated by the minimum bolt preload, the prying force generated by the external pressure load is not critical with respect to bolt stress, and will not result in loss of the ram closure plate closure seal.
Bending Moment Bolt Force The analysis is described in Table 2.2 of [1].
The maximum bending bolt moment Mbb generated by the applied load is evaluated as follows:
f l
b b
b lb bb M
K K
K N
D M
x
+
x
=
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-22b The coefficients Kb and Kl are based on geometry and material properties and are defined in Table 2.2 of [1]. By substituting the values given above:
5 4
6 4
10 46
.1 64 0.1 00 27 10 2.
26 25
.1 12 64 x
=
x
=
=
b lb b
b b
b D
D E
L N
K lb/bolt
(
) (
)
(
) (
)
7 2
2 2
3 6
lb 2
lo lb 2
ul 2
ul 3
l l
l 10 02
.3 00 27 88 28 00 27 3.0 1
3.0 1
3 00
.5 10 2.
26 D
D D
N 1
N 1
3 t
E K
x
=
+
x x
=
+
x
=
lb/bolt In the case of the internal pressure load, Mf = 683 in.lb.
Therefore:
0 in.lb/bolt 23 683 10 02
.3 10 46
.1 10 46
.1 12 00 27 M
7 5
5 bb
=
x x
+
x x
x x
=
The maximum bending bolt moment is considered equal to 0.
A.2.13.2.9.3 Type 1 Ram Closure Plate Bolt Stress Calculations The analysis is described in Table 5.1 of [1].
Average Tensile Stress The bolt preload is calculated to withstand the worst case load combination and to maintain a clamping (compressive) force on the closure joint, under both normal and accident conditions.
Based upon the load combination results (see Table A.2.13.2-10), it is shown that a positive (compressive) load is maintained on the clamped joint for all load combinations.
The maximum non-prying tensile force for both normal and accident conditions is Fa = 22,222 lbs, from load case 1.B (maximum torque preload + temperature load).
The average tensile stress caused by the tensile bolt force Fa is:
2 2732
.1 ba a
ba D
F S
x
=
Dba is the bolt diameter for tensile stress calculation: Dba = Db - 0.9743 x p, where p is the pitch of the bolt. According to Table 1, p. 1714 of [3], p = 0.125 in. Therefore, Dba = 0.8782 in. This value is conservatively used for all bolt stress calculations.
psi 685 36 8782
.0 222 22 2732
.1 S
2 ba
=
x
=
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-22c Bending Stress The bending stress caused by the bending bolt moment Mbb is null since the bending bolt moment is equal to 0.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-24 For accident conditions:
318
.0 500 115 685 36 F
S R
HAC tb ba t
=
=
=
163
.0 300 69 279 11 F
S R
HAC vb bt s
=
=
=
1 128
.0 163
.0 318
.0 R
R 2
2 2
s 2
t
=
+
=
+
A.2.13.2.9.3a Type 2 Ram Closure Plate Bolt Stress Calculations The analysis is described in Table 5.1 of [1].
Average Tensile Stress The bolt preload is calculated to withstand the worst case load combination and to maintain a clamping (compressive) force on the closure joint, under both normal and accident conditions.
Based upon the load combination results (see Table A.2.13.2-10), it is shown that a positive (compressive) load is maintained on the clamped joint for all load combinations.
The maximum non-prying tensile force for both normal and accident conditions is Fa = 15,045 lbs, from load case 1.B (maximum torque preload + temperature load).
The average tensile stress caused by the tensile bolt force Fa is:
2 2732
.1 ba a
ba D
F S
x
=
Dba is the bolt diameter for tensile stress calculation: Dba = Db - 0.9743 x p, where p is the pitch of the bolt. According to Table 1, p. 1714 of [3], p = 0.125 in. Therefore, Dba = 0.8782 in. This value is conservatively used for all bolt stress calculations.
psi 837 24 8782
.0 045 15 2732
.1
,2 =
x
=
ba S
Bending Stress The bending stress caused by the bending bolt moment Mbb is null since the bending bolt moment is equal to 0.
Sbb = 0
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-24a Shear Stress For both normal and accident conditions, the average shear stress caused by shear bolt force Fs is:
Sbs = 0.
The maximum shear stress caused by the torsional moment Mt is:
3 093
.5 ba t
bt D
M S
x
=
For both normal and accident conditions:
psi 640
,5 8782
.0 750 093
.5 3 =
x
=
bt S
Maximum Combined Stress Intensity The maximum combined stress intensity is calculated as follows:
(
)
(
)
2 2
4 bt bs bb ba bi S
S S
S S
+
x
+
+
=
For normal conditions, it combines tension, shear, bending, and residual torsion:
(
)
(
)
psi 278 27 640
,5 0
4 0
837 24 2
2
=
+
x
+
+
=
bi S
Stress Ratios In order to meet the stress ratio requirement, the following relationship must hold for both normal and accident conditions:
Rt 2 + Rs 2 < 1 Where Rt is the ratio of average tensile stress to allowable average tensile stress Ftb and Rs is the ratio of average shear stress to allowable average shear stress Fvb.
For Normal conditions:
260
.0 400 95 837 24
=
=
=
NCT tb ba t
F S
R 098
.0 200 57 640
,5
=
=
=
NCT vb bt s
F S
R 1
078
.0 098
.0 260
.0 2
2 2
2
=
+
=
+
s t
R R
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-24b For Accident conditions:
215
.0 500 115 837 24
=
=
=
HAC tb ba t
F S
R 081
.0 300 69 640
,5
=
=
=
HAC vb bt s
F S
R 1
053
.0 081
.0 215
.0 2
2 2
2
=
+
=
+
s t
R R
A.2.13.2.9.4 Type 1 Ram Closure Plate Bearing Stress under Bolt Head The maximum axial force is 22,222 lb for both normal and accident conditions. A washer of outer diameter 1.50 is used. The diameter of the bolt hole is 1.16 in.
The bearing area is 0.25 x x (1.502 - 1.162) = 0.71 in2.
Therefore, the bearing stress is:
psi 300 31 71
.0 222 22
=
The allowable normal condition bearing stress on the ram closure plate is taken to be the yield stress of the ram closure plate material at 400F, 34.2 ksi.
A.2.13.2.9.4a Type 2 Ram Closure Plate Bearing Stress under Bolt Head The maximum axial force is 15,045 lb for both normal and accident conditions. A washer of outer diameter 1.5 in is assumed for the stainless steel RACP. The diameter of bolt hole is 1.16 in.
The bearing area is: 0.25 x x (1.52 - 1.162) = 0.71 in2 Therefore, the bearing stress is:
psi 190 21 71
.0 045 15
=
The allowable normal condition bearing stress on the ram closure plate is taken to be the yield stress of the closure plate material (stainless steel) at 225F, 24,400 psi.
A.2.13.2.9.5 Results A summary of the ram closure plate bolt stresses calculated above is listed in Table A.2.13.2-7.
The calculated bolt stresses are all less than the specified allowable stresses.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-24c A.2.13.2.9.6 Minimum Engagement Length for Bolt and Flange The minimum engagement length Le for the bolt and flange is ([3], page 1490):
(
)
x x
+
x x
x
=
max min max 57735 2
1 1416
.3 2
n s
n t
e K
E n
K A
L At is the tensile-stress area of the screw and is given by the following formula2:
2 min 16238
.0 2
x
=
n E
A s
t
2 Formula valid if the ultimate tensile strength of the screw is over 100,000 psi
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-26 So:
77
.1 0.
70 569
.1 0.
165 178
.1
=
x x
=
J Therefore, the minimum required engagement length Q = J x Le = 1.77 x 0.73 = 1.29 in.
The actual minimum engagement length is equal to:
4.00 (bolt length) - 1.25 (thickness of the cover plate under the screw head) - 0.177 (washer thickness) = 2.57 in > 1.50 in (ram closure plate bolts inserts) > 1.29 in.
The above calculation bounds the minimum required engagement length if inserts are used because Su of inserts is higher than the Su for the lid, thus lowering the J value.
A.2.13.2.10 Conclusions A lid bolt torque range of 950 to 1,040 ft.lb is required.
A ram closure plate bolt torque range of 225 to 250 ft.lb is required (type 1) or 100-125 ft.lb (type 2).
For the required preloads:
- 1. Bolt stresses meet the acceptance criteria of NUREG/CR-6007 "Stress Analysis of Closure Bolts for Shipping Casks."
- 2. A positive (compressive) load is maintained during all load combinations, except, for the lid bolts, for the accident condition impact plus pressure load case. A more detailed analysis is performed to evaluate the closure of the lid during this event and shows that there is no decompression in the seal during this event, and therefore no leak of the contents during the worst case loading condition.
- 3. The bolt and flange thread engagement lengths are acceptable.
The MP197HB cask lid bolts will not fail due to fatigue for 250 round trip shipments.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-28 Table A.2.13.2-1 Design Parameters for Bolts Analysis Parameter Lid Ram Closure Plate Type 1 Type 2 b
Thermal coefficient of expansion of the bolts (in/in/°F) 7.0 x 10-6 7.1 x 10-6 6.5 x 10-6 c, l Thermal coefficients of expansion of the cask, closure lid and ram closure plate (in/in/°F) 7.0 x 10-6 7.1 x 10-6 9.0 x 10-6 ai Maximum rigid-body impact acceleration of the cask for Hypothetical Accident Conditions - 30 ft C. G. over corner drop (g) 40 xi Impact angle between the cask axis and target surface for Hypothetical Accident Conditions - 30 ft C. G. over corner drop 60.3° Db Nominal diameter of closure bolt (in) 1.50 1.00 Dlb Bolt circle diameter (in) 74.81 27.00 Dlg Outer seal diameter (in) 72.737 25.258 Dli Inner edge diameter (in) 70.44 22.00 Dlo Outer edge diameter (in) 77.18 28.88 Dpb Puncture bar diameter (in) 6.0 DLF Dynamic Load Factor 1.1 Eb Young's modulus of bolt material (ksi) 26.45 x 103 26.2 x 103 27.0 x 103 Ec, El Youngs modulus of cask flange, cask bottom, closure lid and ram closure plate material (ksi) 26.45 x 103 26.2 x 103 27.4 x 103 K
Nut factor for empirical relation between applied torque and achieved preload 0.135 Lb Bolt length between top and bottom surfaces of closure plate at bolt circle (in) 2.41 1.25 Nb Total number of closure bolts 48 12 Nul Poissons ratio of closure plates 0.3 Pli Pressure inside the cask (psig) 30 Plo Pressure outside the cask (psig) 290 Q
Applied preload bolt torque (ft.lb) 950-1,040 225-250 100-125 Sub Ultimate strength of bolt material (ksi) 165.0 165.0 Sul Ultimate strength of closure plates material (ksi) 70.0 69.8 Syb Yield strength of bolt material (ksi) 139.1 137.9 143.1 Syl Yield strength of closure plates material (ksi) 32.6 34.2 24.4 tc Thickness of cask wall (in) 7.0 tl Thickness of closure plates at center (in) 4.5 5.00 tlf Thickness of closure plates flange (in) 3.94 2.50 Wc weight of contents (lbs) 118,500 N/A Wl weight of plate (lbs) 6,100 530 Note:
Material properties for Lid at 350°F. Material properties for type 1 (carbon steel) Ram closure plate at 400°F. Material properties for type 2 ram closure plate at 225°F conservatively for type 304 stainless steel.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-36 Table A.2.13.2-9 Type 1 Ram Closure Plate Bolts Individual Summary Load Case Applied Load Non-Prying Tensile Force Fa (lb/bolt)
Torsional Moment Mt (in.lb/bolt)
Prying Force Ff (lb/in)
Prying Moment Mf (in.lb/in)
Preload (P)
Residual torque Minimum 20,000 1,350 0
0 Maximum 22,222 1,500 0
0 Gasket (G)
Seating load 18,654 0
0 0
Internal Pressure (Pi) 30 psig internal 1,218 0
203 683 Thermal (T) 400°F 0
0 0
0 Impact (I) 30 ft accident conditions drop 2,262 0
320 1,080 Puncture (Pu)
Drop on six inch diameter rod 0
0 0
0 External pressure (Pe) 290 psig external 0
0
-1,958
-6,607
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-36a Table A.2.13.2-9a Type 2 Ram Closure Plate Bolts Individual Summary Load Case Applied Load Non-Prying Tensile Force Fa (lb/bolt)
Torsional Moment Mt (in.lb/bolt)
Prying Force Ff (lb/in)
Prying Moment Mf (in.lb/in)
Preload (P)
Residual torque Minimum 6,218 600 0
0 Maximum 7,813 750 0
0 Gasket (G)
Seating load 0
0 0
0 Internal Pressure (Pi) 30 psig internal 1,218 0
203 683 Thermal (T) 400°F 7,232 0
0 0
Impact (I) 30 ft accident conditions drop 2,262 0
320 1,080 Puncture (Pu)
Drop on six inch diameter rod 0
0 0
0 External pressure (Pe) 290 psig external 0
0
-1,958
-6,607
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-37 Table A.2.13.2-10 Type 1 Ram Closure Plate Bolts Normal and Accident Load Combinations Load Case Combination Description Non-Prying Tensile Force Fa (lb/bolt)
Torsional Moment Mt (in.lb/bolt)
Prying Force Ff (lb/in)
Prying Moment Mf (in.lb/in)
- 1. NCT (P+T)
Preload +
Temperature A. Min. Torque 20,000 1,350 0
0 B. Max. Torque 22.222 1,500 0
0
- 2. NCT (Pi+G)
Internal Pressure + Seating Load 19,872 203 683
- 3. HAC (Pi+I)
Internal Pressure + Impact 3,480 0
523 1,763
- 4. HAC (Pe)
External Pressure 0
0
-1,958
-6,607
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.2-37a Table A.2.13.2-11 Type 2 Ram Closure Plate Bolts Normal and Accident Load Combinations Load Case Combination Description Non-Prying Tensile Force Fa (lb/bolt)
Torsional Moment Mt (in.lb/bolt)
Prying Force Ff (lb/in)
Prying Moment Mf (in.lb/in)
- 1. NCT (P+T)
Preload +
Temperature A. Min. Torque 13,450 600 0
0 B. Max. Torque 15,045 750 0
0
- 2. NCT (Pi+G)
Internal Pressure + Seating Load 1,218 0
203 683
- 3. HAC (Pi+I)
Internal Pressure + Impact 3,480 0
523 1,763
- 4. HAC (Pe)
External Pressure 0
0
-1,958
-6,607
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.7-3a Group 5 The top and bottom end assembly dimensions for the Group 5 RWCs are given in the following table.
Group 5 RWC Top and Bottom End Assembly Dimensions (in.)
Component RWC-W RWC-B RWC-DD Outer Top Cover Plate 2.00 2.00 2.00 Top Shield Plug 5.00 5.00 5.00 Bottom Shield Plug 3.75 3.75 3.75 Outer Bottom Cover Plate 2.00 2.00 2.00 Note:
These configuration bounds other variations of the RWCs.
The RWC is analyzed using a three-dimensional (3D) 180° half-symmetric finite element. The bottom assembly of the 3D finite element model conservatively includes only the outer bottom cover plate. A single enveloping finite element (FE) model encompassing all three design options is modeled for all structural analyses.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.13-2 Table A.2.13.13-1 ASME Code Alternatives for the NUHOMS-MP197HB Cask Containment Boundary Reference ASME Code Section/Article Code Requirement Exception, Justification & Compensatory Measures NCA All Not compliant with NCA [1].
NB-1100 Requirements for Code Stamping of Components.
The NUHOMS-MP197HB cask containment boundary is designed &
fabricated in accordance with the ASME Code,Section III, Subsection NB [1] to the maximum extent practical. However, Code Stamping is not required. As Code Stamping is not required, the fabricator is not required to hold an ASME N or NPT stamp, or to be ASME Certified.
NB-1131 The design specification shall define the boundary of a component to which other components are attached.
A code design specification is not prepared for the NUHOMS-MP197HB cask. A TN design criteria is prepared in accordance with TNs QA program.
NB-2130 NB-4121 Material must be supplied by ASME approved material suppliers.
Material Certification by Certificate Holder.
All materials designated as ASME on the SAR drawings are certified to meet all ASME Code criteria but is not eligible for certification or Code Stamping if a non-ASME fabricator is used. As the fabricator is not required to be ASME certified, material certification to NB-2130 is not possible. Material traceability & certification are maintained in accordance with TNs NRC approved QA program.
NB-7000 Overpressure Protection.
No overpressure protection is provided for the NUHOMS-MP197HB cask. The function of the NUHOMS-MP197HB cask is to contain radioactive materials under normal, off-normal, and hypothetical accident conditions postulated to occur during transportation. The NUHOMS-MP197HB cask is designed to withstand the maximum internal pressure considering 100% fuel rod failure at maximum accident temperature. The NUHOMS-MP197HB cask is pressure tested in accordance with the requirements of 10CFR71 [2] and TNs approved QA program.
NB-8000 Requirements for nameplates, stamping &
reports per NCA-8000.
The NUHOMS-MP197HB cask nameplates provide the information required by 10CFR71 and 49CFR173 [8] as appropriate. Code stamping is not required for the NUHOMS-MP197HB cask. QA Data packages are prepared in accordance with the requirements of 10CFR71 and TNs approved QA program.
NB-3122.1 No structural strength shall be attributed to cladding The thickness of the weld overlay/cladding is included in the analytical models described in Chapter A.2 to calculate the behavior of the MP197HB cask for NCT and HAC loading conditions. In addition to the requirements of ASME Sections II, III, and IX, the following requirements are imposed to assure a continuous bond between the base metal and cladding:
- 1. Integrally clad plate shall be subject to straight beam UT examination per SA-264 Section 13 (2007) with 100% coverage (S1) and Class 1 bond quality level.
- 2. The shear strength requirement of integrally clad plate per SA-264 (2004 edition through 2006 addenda) §8.1.3 shall be increased from 20 ksi to 45 ksi.
UT examination per ASME SA-578 with the acceptance standard of ASME NB-5331. The bottom and cask body flange forgings, items 5 and 12 on drawing MP197HB-71-1002, and port sealing surfaces are excepted.
UT examination of the cladding and overlay will be performed according to a sampling plan after every 300 one-way shipments. Any debonded area exceeding 1 inch in its longest dimension will be evaluated according to the certificate holder's corrective action program. This examination will also be required in the event of a cask drop or similar off-normal impact, prior to returning the cask to service.
NB-5221 Volumetric (RT) inspection of CAT. B weld required Fabrication sequence for inner and outer shells makes RT of 2 closure welds very difficult. May need to use UT or multilayer MT/PT to provide volumetric examination.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.2.13.14-7 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.5-1b
[
]
MP197HB Unit 01 is the same as the MP197HB with the exception of localized reduced lead thickness on the side of the cask body. The shielding performance of MP197HB Unit 01 is not expected to be affected for the authorized spent fuel content as shown in Section A.5.5.5.1.
70,000 Ci of Co-60 or equivalent is demonstrated to be acceptable for a uniform lead thickness of 2.77 inches instead of 3.00 inches nominal thickness. 9,000 Ci of Co-60 or equivalent is demonstrated to be acceptable for the MP1967HB Unit 01 loaded with an alternate RWC design with a minimum 0.5-in steel shell thickness.
A.5.1 Description of the Shielding Design The MP197HB cask is designed to transport one of several NUHOMS DSCs loaded with spent fuel assemblies or dry irradiated and/or contaminated non-fuel bearing solid materials in a radioactive waste canister (RWC) in accordance with the requirements of the 10 CFR 71. The authorized contents acceptable for transport are described in Chapter A.1, Section A.1.2.3, including appendices A.1.4.1 through A.1.4.9A. A complete list of the NUHOMS DSCs authorized for transport is provided in Chapter A.1, Section A.1.2.3.1. Chapter A.1, Section A.1.2.3.2 (also in Appendix A.1.4.9A) provides a description of the irradiated and/or contaminated non-fuel bearing solid materials authorized for transport in the RWC as well as its respective physical dimensions.
Radiological sources used for the calculation of the dose rates presented in this chapter are determined through ranking using the response function methodology to develop the fuel qualification tables (FQT). Response function results are compared with direct MCNP analysis using a discrete MP197HB transportation package model as described in Section A.5.4.1.2.3.
By definition of the FQTs, the minimum cooling times are determined so that the maximum NCT dose rates for intact fuel at 2 m from the side of the vehicle are 8.2 mrem/hr. For fuel in the peripheral basket locations, additional cooling time is needed for some burnup, enrichment, and cooling time (BECT) combinations due to fuel reconfiguration, as defined using the methodology in Section A.5.4.1.3.3. Further discussion of the fuel qualification methodology is contained in Section A.5.4.1.3 and FQT results are discussed in Section A.5.5.2.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.5-3 Hypothetical accident condition (HAC) dose rates are calculated 1 m from the surface of the cask body. No credit is taken for the neutron shield or impact limiters. Both intact and reconfigured fuel are considered. The maximum radiation dose rates for HAC are shown in Table A.5-2
[
] and Table A.5-2a [
]
Dose rates for RWC are provided in Table A.5-34. This table contains both NCT and HAC dose rate results. Compared to spent fuel, RWC dose rates are low.
A.5.2 Source Specification There are five principal sources of radiation associated with transport of spent nuclear fuel that are of concern for radiation protection.
- 1. Primary gamma radiation from spent fuel.
- 2. Primary neutron radiation from spent fuel (both alpha-n reactions and spontaneous fission).
- 3. Gamma radiation from activated fuel structural materials and fuel inserts.
- 4. Capture gamma radiation produced by attenuation of neutrons by shielding material of the cask.
- 5. Neutrons produced by sub-critical multiplication in the fuel.
There are three source configurations used in the evaluation of the shielding performance of the MP197HB transportation package. These configurations are selected because of their respective bounding parameters on all authorized contents. The bounding configurations are as follows:
90,000 Ci of Co-60 or equivalent in the RWC, (equivalent activities for sources not entirely Co-60 are discussed in Section A.5.2.1.5 and shown in Table A.5-61), except for MP197HB Unit 01 where the limit is 70,000 Ci of Co-60 or equivalent due to a reduction on lead thickness for the as-built condition (equivalent activities for sources not entirely Co-60 are discussed in Section A.5.2.1.5 and shown in Table A.5-62). 9,000 Ci of Co-60 or equivalent of radwaste material (equivalent activities for sources not entirely Co-60 shown in Table A.5-65) in alternate RWC with a minimum 0.5-inch steel shell thickness and MP197HB Unit 01.
69 GE-2,3 7x7 Type G2A BWR spent fuel assemblies in the 69BTH DSC, and 37 B&W 15x15 Mark B-10 PWR spent fuel assemblies in the 37PTH DSC.
For the spent fuel assemblies listed, design basis sources which encompass the allowable burnup and enrichment combinations for the authorized contents are developed in the following subsections. The spent fuel assembly types are selected as bounding mainly because their respective initial uranium loading bounds all others.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.5-6b The limits on payload gamma activity as function of gamma energy are developed using the methodology described above including 18-energy group response function at 2 m from the package side in NCT for MP197HB Unit 01 with uniform 2.77 inches lead thickness throughout side of the cask body. Activity limits for gamma energy emissions other than Co-60 are determined as the equivalent activities per gamma energy to an activity of 70,000 Ci of Co-60 resulting in a dose rate limit of 8.71 mrem/hr, Table A.5-62.
The difference in dose rates estimated at 2m using the response function and explicitly evaluated by scaling the results for 90,000 Ci are due to difference in the tally option used in the MCNP calculations. The response function are developed with one angular bin, circumferential average tally, while dose rate are computed with 71 angular bins.
This paragraph documents the analysis for the MP197HB Unit 01 loaded with an alternate RWC design with a minimum 0.5-inch steel shell thickness. The quantity of radioactive material is determined assuming a smeared material identical to that of carbon steel at a density of 1.0 g/cc in the MCNP model. The radioactive waste content is distributed within the inner diameter of the canister modeled as a carbon steel cylinder with a 67.19 outer diameter and a 0.5 shell thickness. The top and bottom of the canister are modeled conservatively with the RWC top and bottom material and thicknesses. The height of the homogenized radioactive waste is modeled as 100 cm, the analysis includes three configurations for the waste: bottom, cask mid-height, and top. Table A.5-64 shows the dose rate at 2 m from the side for MP197HB Unit 01 loaded with 9000 Ci of Co-60 for the three configurations previously mentioned. Table A.5-65 shows the dose rates for MP197HB Unit 01 loaded with an alternate RWC design with a minimum 0.5-inch steel shell thickness. The equivalent activities for sources that are not entirely Co-60 are shown in Table A.5-66 based on a activity of 9,000 Ci of Co-60 resulting in a 2-m dose rate limit of 8.91 mrem/hr.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.5-80w Table A.5-63 Reduced Lead Evaluation - 69BTH DSC Component of Dose Rate Reduced Lead Thickness - 3 mm deep, 30 wide, full length grooves in lead NCT Dose rate at 2m radial distance from Side of Impact Limiters, mrem/hr 1 m from package surface Gamma 0.228 +/- 0.001 3.790 +/- 0.018 Neutron 6.588 +/- 0.114 848.717 +/- 3.870 (n,g) 1.809 +/- 0.012 2.430 +/- 0.037 Total (1) 8.432 +/- 0.101 854.866 +/- 3.870 (1) Spatial locations of maximums of components of the total dose rate are generally different.
Because of this, the maximum of the total dose rate is generally not equal to the sum of maximums of the components Table A.5-64 MP197HB Unit 01 Loaded with Alternate RWC Design with 0.5-in Steel Shell Radial Distance 2 m from Side of ILs Normal Conditions of Transport (NCT) for 9000 Ci Co-60 Dose Rate, mrem/hr Relative Error %
Waste - Top 8.83 2.05 Waste - Cask mid-height 8.00 1.52 Waste - Bottom 9.46 2.09 Table A.5-65 Summary of Maximum Dose Rates of the Cask Containing the Radioactive Waste Canister -
24PT4 Radial Distance from Side of ILs or Body (1), m Normal Conditions of Transport (NCT)
Hypothetical Accident Condition Dose
- Rate, mrem/hr Relative Error Dose
- Rate, mrem/hr Relative Error Shield Shell 146.05 1.41 158.33 1.57 Package Side Perimeter 68.03 1.45 185.32 1.34 0
53.84 1.60 111.39 1.26 1
18.25 1.62 44.13 1.54 2
9.46 2.09 22.34 2.06 (1) See Table A.5-34 for top and botom dose rates.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.5-80x Table A.5-66 Equivalent Activity Limits as a Function of Energy - MP197HB Unit 01 - 24PT4 Energy (MeV)
Response
(mrem/h//s)
Relative Uncertainty Activity (/s)
Dose Rate (mrem/h)
Equivalent Activity (/s) (2) 0.60 5.24E-17 0.0014 1.70E+17 0.80 2.00E-16 0.132 4.46E+16 1.00 1.84E-15 0.0517 4.85E+15 1.1732 7.54E-15 0.0278 3.33E14 (1) 2.51 1.18E+15 1.3325 1.92E-14 0.0228 3.33E14 (1) 6.39 4.64E+14 1.50 4.17E-14 0.0151 2.14E+14 1.75 9.26E-14 0.0116 9.61E+13 2.00 1.69E-13 0.0097 5.26E+13 2.50 3.83E-13 0.0078 2.33E+13 3.00 6.51E-13 0.0072 1.37E+13 3.50 9.27E-13 0.0067 9.60E+12 4.00 1.18E-12 0.0067 7.56E+12 4.50 1.41E-12 0.0066 6.33E+12 5.00 1.56E-12 0.0066 5.70E+12 6.00 1.83E-12 0.0066 4.86E+12 8.00 2.10E-12 0.0068 4.24E+12 10.00 2.30E-12 0.0069 3.87E+12 Total Dose Rate:
8.91 (1) Activity limit corresponding to 9000 Ci of Co-60 resulting in 8.91 mrem/hr (using response functions at 1.1732 MeV and 1.3325 MeV)
(2) Equivalent activity limits per energy for contents other than Co-60 determined using 8.91 mrem/hr and response functions at energy groups shown in the first column. Equivalent Activity (i) = 8.91 /
Response function at Energy (i)
MP197 Transportation Packaging Safety Analysis Report Rev. $, 11
NUH09.0101 A.6.5.14-l Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-1a form of the content and their maximum quantity to be loaded in any of the nine DSCs are specified in Table A.7-2a. Type and form of the content and their maximum quantity to be loaded in an RWC are specified in Table A.7-2b, Table A.7-2c, Table A.7-2d, or Table A.7-2e.
Equivalent Activity limits by gamma energy shown in Table A.7-2c or Table A.7-2d or Table A.7-2e may not be interpolated in energy. The proper procedure for gamma emitter is to round source energies up to the next higher energy level in Table A.7-2c or Table A.7-2d or Table A.7-2e. Procedures are provided in this section for (1) transport of the cask/DSC/RWC directly from the plant spent fuel pool and (2) transport of a DSC/RWC which was previously stored in a NUHOMS horizontal storage module (HSM). Section A.7.7 contains an appendix for each DSC model detailing its loading procedures. Table A.7-3 lists these appendices.
A.7.1.1 NUHOMS-MP197HB Cask Preparation for Loading Procedures for preparing the cask for use after receipt at the loading site are provided in this section and are applicable for shipment of DSCs loaded with fuel or of RWCs loaded with dry irradiated and/or contaminated non-fuel bearing solid materials.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-5 This section summarizes the steps for transferring a previously loaded DSC under a 10 CFR Part 72 license from the HSM or AHSM (generally referred here as HSM) to the MP197HB cask for transportation. Depending on the most recent use of the cask, several of the initial steps listed below may not be necessary.
An RWC may be stored in an HSM, AHSM or other allowed overpack on the plant site. When the MP197HB cask is dry loaded with an RWC, operational steps similar to dry loading a DSC from an HSM into the MP197HB cask should be used depending on the storage overpack.
CAUTION:
Before initiating any steps described in this section:
- The licensee shall perform an audit of spent fuel pool records from the time of canister loading for the identification of the loaded fuel assemblies, and
- The licensee shall compare the irradiation parameters of the loaded contents against those shown in Table A.6.5.8-7 to ensure compliance with the isotopic depletion analysis.
[
]
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-5c
[
]
Procedure for Transferring a Loaded DSC or RWC from an Overpack into a NUHOMS-MP197HB Cask
- 1.
Verify that the contents are in compliance with the fuel specification requirements or waste requirements in the Certificate of Compliance (CoC). An independent check of this verification is also required.
- 2.
Verify that the NUHOMS-MP197HB cask has been prepared for loading as described in Section A.7.1.1.
2a.
Verify that the DSC or RWC was loaded in accordance with the applicable procedure in Appendices A.7.7.1 through A.7.7.10 as listed in Table A.7-3.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-8
- 9.
Install the transportation skid tie-down straps.
- 10.
Install the impact limiters on the cask and torque the attachment bolts in accordance with the drawings in Chapter A.1, Appendix A.1.4.10.1.
- 11.
Remove the impact limiter hoist rings and replace them with hex bolts.
- 12.
Install the cask tamperproof seals.
- 13.
Install the transportation skid personnel barrier.
- 14.
Perform a final radiation survey to assure the cask radiation levels do not exceed 49 CFR 173.441 [2] and 10 CFR 71.47 [3] requirements. This final radiation survey shall include a measurement of the dose rate in the normally occupied spaces (i.e. truck cabin). If the measured dose rate in the normally occupied spaces exceeds 2 mrem/hr, the location of the package shall be changed or supplementary shielding added as necessary to reduce the dose to an acceptable level. Supplementary shielding may be added to the conveyance (e.g.
attached to the sides of the trailer or truck cab) to reduce the external radiation levels, but shall not be attached to the package without prior NRC approval. Alternatively, the carrier may implement the radiation dosimetry requirements of 10 CFR 71.47(b)(4) and 49 CFR 173.441(b)(4) to satisfy the requirements of 10 CFR 20.1502.
- 15.
Verify that the temperature on all accessible surfaces is < 185°F.
- 16.
Prepare the final shipping documentation and release the loaded cask for shipment.
A.7.2 NUHOMS-MP197HB Package Unloading Unloading the NUHOMS-MP197HB cask after transport involves removing the cask from the conveyance and removing the DSC from the cask. The cask is designed to allow the DSC to be unloaded from the cask into a NUHOMS staging module, hot cell or other suitable overpack, and provisions exist to allow wet unloading into a fuel pool. RWCs can either be unloaded similarly to the DSCs, or the contents can be vertically unloaded with the RWC remaining in the cask. The necessary procedures for these tasks are essentially the reverse of those described in Section A.7.1.
A.7.2.1 Receipt of Loaded NUHOMS-MP197HB Package from Carrier Procedures for receiving the loaded cask after shipment are described in this section. Procedures for receiving an empty cask are provided in Section A.7.1.1.
- 1.
Verify that the tamperproof seals are intact.
- 2.
Remove the tamperproof seals.
- 3.
Remove the hex bolts from the impact limiters and replace them with the impact limiter hoist rings provided.
- 4.
Remove the impact limiters from the cask.
- 5.
Remove the transportation skid personnel barrier and tie-down straps.
- 6.
Remove the external aluminum fins, if present.
- 7.
Take contamination smears on the outside surfaces of the cask. If necessary, decontaminate the cask.
- 8.
If required for unloading, install the front and rear trunnions and torque the bolts as specified in Drawing MP197HB-71-1002, Chapter A.1, Appendix A.1.4.10.1 following the torquing sequence shown in Figure A.7-1.
- 9.
If the packaging contains high burnup fuel assemblies, perform a Radiation Survey (both neutron and gamma) and a Thermal Survey of the cask loaded with the contents to evaluate the axial radiation and thermal source distributions. These surveys shall be performed on the survey locations identified in Section A.7.1.4, Step 3 using the same quality assurance requirements. It is recommended to use the same type and model of the
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-11 This procedure is for removal of contents from within a RWC-DD at a disposal site. The RWC-DD will remain within the NUHOMS-MP197HB cask at all times and may be reused after unloading. The procedure described below is intended to show the type of operations that will be performed and is not intended to be limiting.
- 1.
Lift the cask and transfer it onto an upending device.
- 2.
Upend the cask and place it on a solid surface.
- 3.
Remove the cask lid.
- 4.
Remove the RWC-DD lid.
- 5.
Install sealing surface protection, as appropriate.
- 6.
Lift the contents out of the RWC-DD and transfer to disposal area.
- 7.
Remove the sealing surface protection, if used.
- 8.
Install the RWC-DD lid.
- 9.
Install the cask lid.
- 10.
Downend the cask.
A.7.3 Preparation of Empty Package for Transport
- 1.
Determine the amount and form of residual internal activity within the interior of the empty packaging.
- 2.
Inspect and securely close the empty packaging.
- 3.
Prepare the empty packaging for shipment using the package requirements specified in the Hazardous Material Regulations (HMR) [2], which are appropriate for the amount and form of the residual activity and contamination.
A.7.4 Other Operations A.7.4.1 Cask Cavity Vacuum Drying and Dryness Verification Test The procedure for drying the cask cavity of moisture and performing a dryness verification test is given in this section. These steps are only required if the cask was wet loaded.
- 1.
Connect a vacuum system to the cask vent port.
- 2.
Connect a drain bottle, or similar, to the cask drain port.
- 3.
Evacuate the cask cavity until the vacuum level is 40 mbar, or less.
- 4.
Isolate the vacuum system and vent the cask cavity to allow residual moisture to condense and flow through the drain port.
- 5.
Repeat steps 3 and 4 several times until no more water escapes the drain port.
- 6.
Close the cask drain port.
- 7.
Evacuate the cask cavity until the vacuum level is 10 mbar, or less.
- 8.
Isolate the vacuum system and measure the cask cavity pressure rise over a period of 10 minutes. The acceptance criterion for this dryness test is a pressure rise no greater than 6 mbar.
- 9.
Repeat steps 7 and 8 as necessary to achieve an acceptable result.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-12 A.7.4.2 Pre-shipment Verification Leakage Testing of the NUHOMS-MP197HB Cask Containment Boundary The procedure for assembly verification leakage testing of the cask containment boundary prior to shipment is given in this section. Assembly verification leakage testing shall conform to the requirements of ANSI N14.5 [1] or ISO -12807 [11]. The order in which the leakage tests of the various seals are performed may vary. If more than one leakage detector is available then more than one seal may be tested at a time. Personnel performing the leakage test shall be specifically trained in leakage testing in accordance with SNT-TC-1A [7]. The acceptance criterion for pre-shipment leakage rate testing shall be either (a) a leakage rate of not more than the reference air leakage rate, or (2) no detected leakage when tested to a sensitivity of at least 10-3 ref-cm3/s.
The following steps present one method of performing the pre-shipment verification leakage testing. Alternate methods and order of testing are acceptable as long as the above criteria is satisfied for the MP197HB containment boundary seals.
- 1.
Remove the port plugs from the lid test port, vent port, drain port, and the bottom test port.
- 2.
Attach a suitable vacuum pump to the cask lid test port.
- 3.
Evacuate the volume between the lid O-rings and perform the pre-shipment leak test in accordance with Section A.8.2.2. If either O-ring was replaced, the maintenance leak test in Section A.8.2.2 shall be performed.
- 4.
After meeting the leak test criteria, disconnect the vacuum pump and either tighten the port bolt, or verify it has been tightened, in accordance with Drawing MP197HB-71-1002 in Chapter A.1.
- 5.
Install the port plug.
- 6.
Repeat steps 2-5 for the bottom test port.
- 7.
Attach a suitable vacuum pump to the vent port.
- 8.
Either tighten the port bolt, of verify it has been tightened, in accordance with Drawing MP197HB-71-1002 in Chapter A.1.
- 9.
Evacuate the volume outside of the closed port bolt seal and perform the pre-shipment leak test in accordance with Section A.8.2.2. If the O-ring was replaced, the maintenance leak test in Section A.8.2.2 shall be performed.
- 10.
After meeting the leak test criteria, disconnect the vacuum pump and install the port plug.
- 11.
Repeat steps 7-10 for the drain port.
This concludes the assembly verification leakage test procedure.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-14 A.7.5 References
- 1.
ANSI N14.5-2014, American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment, American National Standards Institute, Inc., New York, 2014.
- 2.
Title 49, Code of Federal Regulations, Subtitle B, Chapter 1, Parts 171 through 180.
- 3.
Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packaging and Transportation of Radioactive Material.
- 4.
U.S. Nuclear Regulatory Commission, Office of the Nuclear Material Safety and Safeguards, Safety Evaluation of VECTRA Technologies Response to Nuclear Regulatory Commission Bulletin 96-04 for the NUHOMS-24P and NUHOMS-7P.
- 5.
U.S. Nuclear Regulatory Commission Bulletin 96-04, Chemical, Galvanic or Other Reactions in Spent Fuel Storage and Transportation Casks, July 5, 1996.
- 6.
Not Used.
- 7.
SNT-TC-1A, American Society for Nondestructive Testing, Personnel Qualification and Certification in Nondestructive Testing.
- 8.
Updated Final Safety Analysis Report for The Standardized Advanced NUHOMS Horizontal Modular Storage System For Irradiated Nuclear Fuel (CoC 1029) Revision 3.
- 9.
Not used.
- 10.
Not used.
- 11.
ISO-12807, Safety Transport of Radioactive Materials - Leakage Testing on Packages, First Edition, 1996.
- 12.
NUREG-1927, March 2011, Standard Review Plan for Renewal of Spent Fuel Dry Cask Storage System Licenses and Certificates of Compliance, United States Nuclear Regulatory Commission.
- 13.
EPRI Report No. 1013524, "Climatic Corrosion Considerations for Independent Spent Fuel Storage Installations in Marine Environments, Electric Power Research Institute, June 2006.
- 14.
Repairing SCC of type 316 SS vessels Materials Performance, September 2007, NACE International (p. 80).
- 15.
NUREG/CR-7030, Atmospheric Stress Corrosion Cracking Susceptibility of Welded and Unwelded 304, 304L, and 316L Austenitic Stainless Steels Commonly Used for Dry Cask Storage Containers Exposed to Marine Environments, Page 47, Nuclear Regulatory Commission, October 2010.
- 16.
E-33299, Evaluation Procedure to Verify DSC Acceptance for Transport, Transnuclear, August 2012, Revision 1.
- 17.
Catherine Houska Deicing Salt - Recognizing The Corrosion Threat TMR Consulting, Pittsburg, PA http://www.imoa.info/_files/pdf/DeicingSalt.pdf
- 18.
Greg Oberson, Darrel Dunn, Todd Mintz, Xihua He, Roberto Pabalan and Larry miller, US NRC-Sponsored Research on Stress Corrosion Cracking Susceptibility of Dry Storage Canister Materials in Marine Environments - 13344 WM2013 Conference, February 24-28, 2013, Phoenix, Arizona USA, US NRC ADAMS, ML13029A490
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-14a
- 19.
Regulatory Guide, 1.21, Revision 2, June 2009, Measuring, Evaluation, and Reporting Radioactive Material in Liquid Gaseous Effluents and Solid Waste, United States Nuclear Regulatory Commission.
- 20.
NUREG-1617, Standard Review Plan for Transportation Packages for Spent Nuclear Fuel Final Report, March 2000.
- 21.
B. Anderson, M. Sheaffer, L. Fischer, NUREG/CR-6673, Hydrogen Generation in TRU Waste Transportation Packages, 2000.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-17 Table A.7-1 DSC/RWC, Fuel, and Basket Spacer Nominal Heights for Each Type of DSC/RWC (All dimensions are in inches)
Canister Type 61BT 61BTH 69BTH 24PTH 24PT4 32PT 32PTH 32PTH Type 1 32PTH1 37PTH RWC Type 1 Type 2 S
L S-LC S-100 S-125 L-100 L-125 S
M L
S M
DSC bottom spacer height(1) 2.20 2.20 2.20 1.24 11.7 5.7 11.7 2.2 11.7 11.7 5.7 5.7 12.5 5.25 12.5 5.25 N/A 16.25 9.0 (5)
DSC top spacer height (1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
(1)
Fuel spacer height (2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
(2)(4)
N/A Basket spacer height (3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
(3)(4)
N/A (1)
DSC/RWC top and bottom spacers can be combined to one spacer. If one spacer is used, it can be installed either on top or bottom of the DSC/RWC. The height of the spacer is to be determined such that the gap between the cask and DSC/RWC is below 0.5 for normal transport conditions. The specified spacer height may include any axial spacing provided by the internal canister sleeve components.
(2)
Fuel spacer can be installed either on top or bottom of the fuel assembly. The height of the fuel spacer to be determined using the formula specified in Appendix A.2.13.14, Table A.2.13.14-2 such that the gap between the fuel assemblies and the DSC is below 1.5 for normal transport conditions.
(3)
Basket spacer can be installed either on top or bottom of the basket. The height of the basket spacer is to be determined such that the gap between the basket and the DSC is below 0.815 for normal transport conditions.
(4)
Fuel and basket spacers can be combined in one spacer.
(5)
Height of spacer for RWC-W is 11.75 for the 186.5 and 196 canister length, respectively. The height of the spacer for the RWC-B is 11.75 and height of the spacer for RWC-DD and Alternate RWC-B is 2.25".
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-18 Table A.7-2a Applicable Fuel Specification for Various DSCs DSC MODEL Applicable Fuel Specification from Chapter A.1 NUHOMS-24PT4 Tables A.1.4.1-1 and A.1.4.1-2 NUHOMS-32PT Table A.1.4.2-2 NUHOMS-24PTH Table A.1.4.3-2 NUHOMS-32PTH Table A.1.4.4-2 NUHOMS-32PTH1 Table A.1.4.5-2 NUHOMS-37PTH Table A.1.4.6-2 NUHOMS-61BT Table A.1.4.7-2 NUHOMS-61BTH Table A.1.4.8-2 NUHOMS-69BTH Table A.1.4.9-1 Table A.7-2b Applicable Content Specification for RWC Type and Form of Material The NUHOMS-MP197HB packaging is designed for shipment of various types of irradiated and contaminated reactor hardware. The payload will vary from shipment to shipment. Typical composition of the payload consists of the following components either individually or in combinations:
- 1.
- 2.
BWR Local Power Range Monitors (LPRMs)
- 3.
BWR Fuel Channels
- 4.
BWR Poison Curtains
- 5.
PWR Burnable Poison Rod Assemblies (BPRAs)
- 6.
PWR and BWR Reactor Vessel and Internals Decay Heat load 5 kW Loading Components with high specific activity are generally placed near the center of the RWC. For each shipment, the RWC is normally filled to capacity, which prevents shifting of the contents during transport.
If the RWC is not full, appropriate component spacers or shoring is used to prevent significant movement of the contents.
Maximum Quantity of Material per Package (a)
For containment, the quantity of radioactive material is limited to a maximum of 8, 182 A2. The radioactive material is primarily in the form of neutron activated metals, or metal oxides in solid form. Surface contamination may also be present on the irradiated components. When a wet load procedure (i.e., in-pool) is followed for cask loading, the cask cavity and RWC are drained and dried to ensure that there are no free liquids in the package during transport.
(b)
The NUHOMS-MP197HB packaging is designed to transport a payload of up to 56.0 tons of dry irradiated and/or contaminated non-fuel bearing solid materials in the RWC. The center of gravity (CG) of the loaded NUHOMS-MP197HB package is to be 102 +/- 4 inches from the bottom of the cask.
(c)
The quantity of radioactive material is limited to a maximum of 90,000 Ci of cobalt-60 or equivalent, except for MP197HB Unit 01 where the limit is reduce to 70,000 Ci of cobalt-60 or equivalent. Equivalent activity limits as a function of gamma energy for isotopes other than Co-60 are shown in Table A.7-2c for the 90,000 Ci limit and Table A.7-2d for the 70,000 Ci limit. The quantity of radioactive material is limited to 9000 Ci of cobalt-60 or equivalent in the alternate RWC design with minimum 0.5-inch steel shell thickness in MP197HB Unit 01. Equivalent activity limits as function of gamma energy for isotopes other than Co-60 are shown in Table A.7-2e.
(d)
The RWC shall contain dry irradiated and/or contaminated nonfuel bearing solid materials. The dry irradiated and/or contaminated non-fuel bearing solid materials whose total RWC payload meets concentration requirements as low level radioactive waste (LLRW) per 10 CFR 61.55. Waste characterization per 10 CFR 61.55 is the basis for demonstrating compliance with activity limits for transportation.
(e)
The contents will not include liquid wastes, sludge or resins. Waste containing organic material is acceptable provided that gas generation from water and organic materials does not lead to potentially flammable of explosive conditions, including the formation of corrosive constituents from radiolysis, biodegradation or chemical reaction. The package user is responsible for characterizing the contents and evaluating the gas generation from thermal decomposition, radiolysis, or biologic growth.
(f)
Waste containing fissile material is acceptable provided the quantity of fissile material is limited such that it can be exempted from being classified as fissile material per 10 CFR 71.15 (e.g., fission chambers for in-core detectors). The package user is responsible for characterizing the contents.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7-18b1 Table A.7-2e Activity Limits for Alternate RWC - MP197HB Unit 01 Energy (MeV)
Equivalent Activity
(/s) 0.6 1.70E+17 0.8 4.46E+16 1
4.85E+15 1.1732 1.18E+15 1.3325 4.64E+14 1.5 2.14E+14 1.75 9.61E+13 2
5.26E+13 2.5 2.33E+13 3
1.37E+13 3.5 9.60E+12 4
7.56E+12 4.5 6.33E+12 5
5.70E+12 6
4.86E+12 8
4.24E+12 10 3.87E+12 For sources not entirely Co-60, the following equation is used:
( )
1
( )
i i
i S E Activity Limit E
where
( )
iS E is the source strength in s
and
( )
i ActivityLimit E is the corresponding maximum emission for the each energy of the radionuclide gamma emission. The energy should be rounded up to the next high energy.
For example:
/ 4.46E+16 1 Where Cs-137 is the source strength of Cs-137 For a content of Co-60 and Cs137 mixture, the allowed activities of Co-60 and Cs-137 is:
/ 4.64E+14 + Cs-137
/ 4.46E+16 1 Where Co-60 is the source strength of Co-60 and Cs-137 is the source strength of Cs-137
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7.7.10-i Appendix A.7.7.10 Radioactive Waste Canister (RWC) Wet Loading Procedures TABLE OF CONTENTS A.7.7.10.1 Wet Loading of the RWC.......................................................................... A.7.7.10-1 A.7.7.10.2 RWC-W/RWC-B Drying and Backfilling (Not Applicable for Alternate RWC-B Design)...................................................................................................... A.7.7.10-1 A.7.7.10.3 RWC-W/RWC-B Sealing Operations (Not Applicable for Alternate RWC-B Design)...................................................................................................... A.7.7.10-2 A.7.7.10.4 RWC-DD Drying and Sealing Operations................................................ A.7.7.10-3 A.7.7.10.5 Alternate RWC-B Water Removal and Closure Operations..................... A.7.7.10-4
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7.7.10-1 Appendix A.7.7.10 Radioactive Waste Canister (RWC) Wet Loading Procedures Note: The steps below apply to all versions of the RWC unless noted otherwise. The term cask, used in these steps refers to either the NUHOMS-MP197HB cask or an acceptable NUHOMS transfer cask. The loading procedure below is also applicable to the Alternate RWC-B design where the top cover plate is only bolted. No welding is required for the closure for the Alternate RWC-B prior to transportation.
A.7.7.10.1 Wet Loading of the RWC The starting condition for the following steps assumes completion of the cask preparation steps in Section A.7.1.2.
- 1.
Lift the cask and position it over the cask loading area of the pool.
- 2.
Lower the cask into the pool.
- 3.
Place the cask in the location of the pool used for the cask loading area.
- 4.
Disengage the lifting yoke from the cask lifting trunnions and move the yoke clear of the cask.
- 5.
Load the RWC cavity (waste may be in a loaded liner). Record contents and location on the cask loading report to the extent practical.
- 6.
Install the liner shield plug (RWC-W).
- 7.
Install the top shield plug (RWC-B and Alternate RWC-B).
- 8.
Install the shield lid (RWC-DD).
- 9.
Inspect the shield plug/lid to verify that it is properly installed. Repeat steps 6 through 8 as necessary.
- 10.
NOT USED.
- 11.
NOT USED.
- 12.
Check the radiation levels at the center of the top shield plug or lid and around the perimeter of the cask.
- 13.
NOT USED.
- 14.
Lift the cask from the pool.
- 15.
Move the cask to the plant designated preparation area.
For Alternate RWC-B, see Section A.7.7.10.5.
A.7.7.10.2 RWC-W/RWC-B Drying and Backfilling (Not Applicable for Alternate RWC-B Design)
- 1.
Check the radiation levels along the perimeter of the cask. The cask exterior surface should be decontaminated as necessary. Temporary shielding may be installed as necessary to minimize personnel exposure.
- 2.
Install the Inner Top Cover Plate (RWC-W) (for multiple plate cover).
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7.7.10-2
- 3.
Decontaminate the exposed surfaces of the RWC cylindrical shell perimeter and remove the annulus seal.
- 4.
Allow water from the annulus to drain out until the water level is approximately twelve inches below the top edge of the RWC shell. Take swipes around the outer exposed surface of the RWC shell and check for smearable contamination as required.
CAUTION: Radiation dose rates are expected to be high at the RWC vent and siphon port locations. Use proper ALARA practices (e.g., use of temporary shielding, appropriate positioning of personnel, etc.) to minimize personnel exposure.
- 5.
Prior to the start of welding operations drain a sufficient amount of water to allow for welding of TSP or ITCP (approximately 100 gallons) from the RWC.
- 6.
NOT USED.
- 7.
Install the automated welding machine onto the top shield plug or Inner Top Cover Plate (for multiple plate cover).
- 8.
Check radiation levels along the surface of the top shield plug. Temporary shielding may be installed as necessary.
- 9.
Take precautions to prevent debris and weld splatter from entering the annulus.
- 10.
Weld the top shield or inner top cover plate (for multiple plate cover) to the RWC shell.
- 11.
Perform required dye penetrant examination of the weld surface(s).
- 12.
NOT USED.
- 13.
Pump out and/or blowdown remaining bulk water from the RWC cavity.
- 14.
Once the water stops flowing from the RWC, close the RWC siphon port and disengage the gas source.
- 15.
Connect the VDS or equivalent drying system to the RWC.
- 16.
Start the VDS and evacuate the RWC cavity to 10 mbar or lower for at least 10 minutes.
If that criterion cannot be achieved, perform an evaluation of the residual water in the canister after closure to verify the hydrogen gas concentration in the RWC will be below the lower flammability limit of 4% by volume and also to verify the cavity gas pressure will not exceed the design basis during the intended period of use. The effects of hydrogen generation due to radiolysis of residual water in the canister can be determined using the methodology developed by the U.S. Department of Energy for TRU wastes [21].
- 17.
Use air or helium to pressurize the RWC to 0.0 to 3.5 psig backfill pressure (stable for 30 minutes).
CAUTION: Radiation dose rates are expected to be high at the vent and siphon port locations. Use proper ALARA practices (e.g., use of temporary shielding, appropriate positioning of personnel, etc.) to minimize personnel exposure.
- 18.
Close the line connected to the siphon port.
A.7.7.10.3 RWC-W/RWC-B Sealing Operations (Not Applicable for Alternate RWC-B Design)
- 1.
Disconnect the VDS from the RWC. Seal weld the prefabricated covers over the vent and/or siphon ports and perform the required dye penetrant weld examination(s).
- 2.
Install the outer top cover plate and the automated welding system onto the RWC.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7.7.10-3
- 3.
Tack weld the outer top cover plate to the RWC shell. Place the outer top cover plate weld root pass.
- 4.
Perform dye penetrant examination of the root pass weld. Weld out the outer top cover plate to the RWC shell and perform the required dye penetrant examination on the weld surface(s).
- 5.
Remove the automated welding machine from the RWC.
- 6.
Drain the water from the cask/RWC annulus.
The cask/RWC is now ready to be prepared for downending as described in Chapter A.7, Section A.7.1.2.2.
A.7.7.10.4 RWC-DD Drying and Sealing Operations
- 1.
Check the radiation levels along the perimeter of the cask. The cask exterior surface should be decontaminated as necessary. Temporary shielding may be installed as necessary to minimize personnel exposure.
- 2.
Disengage lid from the lifting yoke and position the yoke clear of the cask.
- 3.
Install lid bolts.
CAUTION: Radiation dose rates are expected to be high at the RWC-DD drain port location. Use proper ALARA practices (e.g., use of temporary shielding, appropriate positioning of personnel, etc.) to minimize personnel exposure.
- 4.
Decontaminate the exposed surfaces of the RWC-DD cylindrical shell perimeter and remove the annulus seal.
- 5.
Take swipes around the outer exposed surface of the RWC-DD shell and check for smearable contamination as required. The annulus water level may be lowered as necessary to perform this step.
- 6.
Remove bulk water from the RWC-DD cavity.
- 7.
Connect a vacuum system to the RWC-DD.
- 8.
Start the VDS and evacuate the RWC cavity to 10 mbar or lower for at least 10 minutes.
If that criterion cannot be achieved, perform an evaluation of the residual water in the canister after closure to verify the hydrogen gas concentration in the RWC will be below the lower flammability limit of 4% by volume and also to verify the cavity gas pressure will not exceed the design basis during the intended period of use. The effects of hydrogen generation due to radiolysis of residual water in the canister can be determined using the methodology developed by the U.S. Department of Energy for TRU wastes [21].
- 9.
Disconnect the vacuum system and close the RWC-DD port.
- 10.
Drain the water from the cask/RWC-DD annulus.
The cask/RWC is now ready to be prepared for downending as described in Chapter A.7, Section A.7.1.2.2.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.7.7.10-4 A.7.7.10.5 Alternate RWC-B Water Removal and Closure Operations
- 1.
Install O-rings or seal for top cover plate in the groove and bolt the top cover plate to the bolting ring flange to required torque values.
- 2.
Decontaminate the exposed surfaces of the RWC perimeter and remove annulus seal.
- 3.
Allow water from the annulus to drain out until the water level is approximately 12 inches below the top edge of the RWC shell. Take swipes around the outer exposed surface of the RWC shell and check smearable contamination as required.
- 4.
Replace annulus water to within 3 inches of the top of the RWC after smear test.
- 5.
Remove the drain port plug and any temporary shielding on top of the streaming drain hole.
- 6.
Perform bulk water removal by pumping water out of the RWC using a lance. When the pump starts to cavitate and lose suction, bulk water removal is completed.
- 7.
Remove lance and bulk water removal equipment. Use crane if available to limit exposure.
- 8.
Bolt drain port cover manually to required torque values and install drain port O-ring or seal if applicable.
The cask/RWC is now ready to be prepared for downending as described in Chapter A.7, Section A.7.1.2.2.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.8-14 If a thermal conductivity test result is below the specified minimum, at least four additional tests shall be performed on the material from that lot. If the mean value of those tests, including the original test, falls below the specified minimum, the associated lot shall be rejected.
After twenty five tests of a single type of material, with the same aluminum alloy matrix, the same boron content, and the same primary boron phase, e.g., B4C, TiB2, or AlB2, if the mean value of all the test results less two standard deviations meets the specified thermal conductivity, no further testing of that material is required. This exemption may also be applied to the same type of material if the matrix of the material changes to a more thermally conductive alloy (e.g.,
from 6000 to 1000 series aluminum), or if the boron content is reduced without changing the boron phase.
The measured thermal conductivity values shall satisfy the minimum required conductivities as shown in Section A.3.2.1, Table 17 for HLZC #1, #2 and #3, and in Section A.3.2.1, Table 19 for HLZC #4.
In cases where the specified thickness of the neutron absorber may vary, the equations introduced in Section A.3.3.1.5 shall be used to determine the minimum required effective thermal conductivity.
The thermal conductivity test requirement does not apply to aluminum that is paired with the neutron absorber.
A.8.1.10 Weld Overlays and Cladding Tests See alternative to code article NB.3122.1 in Table A.2.13.13-1 for requirements that are imposed to assure a continuous bond between the base metal and the weld overlays and cladding at the time of fabrication.
A.8.2 Maintenance Program A.8.2.1 Structural and Pressure Tests Within 14 months prior to any lift of a NUHOMS-MP197HB transport package, the front trunnions shall be subject to either of the following:
- A test load equal to 300% of the maximum service load per ANSI N14.6 [3], paragraph 7.3.1(a) for single failure proof trunnions or a test load equal to 150% of the maximum service load per ANSI N14.6 [3], paragraph 7.3.1(b) for non-single failure proof trunnions. After sustaining the test load for a period of not less than 10 minutes, accessible critical areas shall be subjected to visual inspection for defects, and all components shall be inspected for permanent deformation.
- Dimensional testing, visual inspection and nondestructive examination of accessible critical areas of the trunnions including the bearing surfaces in accordance with Paragraph 6.3.1 of ANSI N14.6 [3].
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.8-14a A.8.2.2 Leakage Tests The following containment boundary components shall be subject to periodic maintenance, and preshipment leakage testing in accordance with ANSI N14.5 [4] or ISO-12807 [11]:
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.8-15
- Lid
- Ram Access Closure Plate
- Vent Port
- Drain Port Leakage Tests for DSC Shipments Test Frequency Acceptance Criteria Typical Method (ANSI N14.5 TABLE A-1, [4])
Periodic Within 12 months prior to shipment Each component individually 1x10-7 ref cm3/s (He)
A.5.3 A.5.4 Pre-shipment Before each shipment, after the contents are loaded and the package is closed No detected leakage, sensitivity of 10-3 ref cm3/s or better, unless seal is replaced.
A.5.1 A.5.2 A.5.8 A.5.9 Maintenance After maintenance, repair, or replacement of containment components, including inner seals Each component individually 1x10-7 ref cm3/s (He)
A.5.3 A.5.4 Leakage Tests for RWC Shipments Test Frequency Acceptance Criteria Typical Method (ANSI N14.5 TABLE A-1, [4])
Periodic Within 12 months prior to shipment The sum of all components 1.44x10-4 ref cm3/s with a test sensitivity of 7.20x10-5ref cm3/s A.5.1 A.5.2 A.5.8 A.5.9 Pre-shipment Before each shipment, after the contents are loaded and the package is closed No detected leakage, sensitivity of 10-3 ref cm3/s or better, unless seal is replaced.
A.5.1 A.5.2 A.5.8 A.5.9 Maintenance After maintenance, repair, or replacement of containment components, including inner seals The sum of all components 1.44x10-4 ref cm3/s with a test sensitivity of 7.20x10-5 ref cm3/s A.5.1 A.5.2 A.5.8 A.5.9 No leakage tests are required prior to shipment of an empty NUHOMS-MP197HB packaging.
MP197 Transportation Packaging Safety Analysis Report Rev. 21A, 11/20 NUH09.0101 A.8-15a A.8.2.3 Component and Material Tests A.8.2.3.1 Fasteners All threaded fasteners and port plugs shall be inspected whenever removed, and annually, for deformed or stripped threads. Damaged parts shall be evaluated for continued use and replaced as required.
At a minimum, the MP197HB cask lid bolts shall be replaced at least every 250 shipments (round trip) to ensure adequate fatigue strength is maintained.
A.8.2.3.1a Weld Overlay/Cladding A periodic UT examination of the inner shell wall weld overlay/cladding needs to be performed.
See alternative to code article NB.3122.1 in Table A.2.13.13-1.
A.8.2.3.2 Impact Limiters A visual examination of the impact limiters before each shipment will be performed to ensure that the impact limiters have not been degraded between leakage test intervals. If there is no evidence of weld cracking or other damage which could result in water in-leakage, the wood will not be degraded. If there is visual damage, the impact limiter will be removed from service, repaired, if possible, and inspected for degradation of the wood. Impact limiters will be leakage tested once every five years to ensure that water has not entered the impact limiters. If the leakage test indicates that the impact limiters have a leak, a humidity test will be performed to verify that there is no free water in the impact limiters.
AFFIDAVIT PURSUANT TO 10 CFR 2.390 to E-57283 TN Americas LLC St.ate of Maryland County of. oward
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1 Prakash Narayanan, depo and say that I am ChiefTechnical Offi~er of TN Americas LLC duly authorized to execute this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediatel_y ~l~w. I am s~bmitting this affidavit in conformance with the provisions of IO CFR 2.390 of the Comm1ss1on s regulations for withholding this information.
below:
The information for which proprietary treatment is sought is contained in Enclosure 2 and is listed Portions of certain chapters and appendices of the Safety Analysis Report (SAR) for CertificateofCompliance o. 9302 MP197 and MPl97HB, Revision 21A, Docket 71-9302 (Proprietary Version)
This document has been appropriately designated as proprietary.
I have personal knowledge of the criteria and procedures utilized by TN Americas LLC in designating information as a trade secret, privileged or as confidential commercial or financial information.
Pursuant to the pro isions of paragraph (b) (4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.
- 1) The information sought to be withheld from public disclosure involves certain design details associated with the SAR analyses, calculations, and SAR drawings for the MP l 97HB System, which are owned and have been held in confidence by TN Americas LLC.
- 2) The information is of a type customarily held in confidence by TN Americas LLC and not customarily disclosed to the public. TN Americas LLC has a rational basis for determining the types of information customarily held in confidence by it.
- 3) Public disclosure of the information is likely to cause substantial hann to the competitive position of TN Americas LLC because the information consists of descriptions of the design and analysis of a radioactive material transportation system, the application of which provide a competitive economic advantage. The availability of such information to competitors would enable them to modify their product to better compete with TN Americas LLC, take marketing or other actions to improve their product's position or impair the position of TN America LLC' s product, and avoid developing similar data and analyses in support of their processes, methods or apparatus.
Further the deponent sayeth not.
Prakash arayanan Chief Technical Officer, TN Americas LLC efore me this 30th day of November 2020.
Notary Public My Commission Expires J.Q.J_I _ I.&.-3 RO OAJONE.S NOTARY PUBLIC MONTGOMERY COUNTY MARYLAND 2fW SSI
.-vO\\RES OCT. 16, t,GfCOMMI
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