ML24214A019
| ML24214A019 | |
| Person / Time | |
|---|---|
| Site: | 07109235 |
| Issue date: | 07/31/2024 |
| From: | NAC International |
| To: | Office of Nuclear Material Safety and Safeguards |
| Shared Package | |
| ML24214A017 | List: |
| References | |
| ED20240097 | |
| Download: ML24214A019 (1) | |
Text
July 2024 Docket No. 71-9235 Revision 24A NAC-STC NAC Storage Transport Cask SAFETY ANALYSIS REPORT Shielding Integrity Test Changes Non-Proprietary Version Volumes 1 and 2 Changed Pages Atlanta Corporate Headquarters : 2 Sun Court, Suite 220, Peachtree Corners, Georgia 30092 USA Phone 770-447-1144, www.nacintl.com to ED20240097 Page 1 of 3 List of SAR Changes NAC-STC SAR, Revision 24A July 2024 to ED20240097 Page 2 of 3 List of Changes, NAC-STC SAR, Revision 24A Chapter 1
No Changes Chapter 2
No Changes Chapter 3
Pages 3.4-31 thru 3.4-33, modified text throughout Section 3.4.4.1 where indicated.
Page 3.5-7, modified text in Section 3.5.4.1 where indicated.
Page 3.8.6-1, modified and replaced most of the text in Section 3.8.6 where indicated.
Chapter 4
Page 4-i, modified Table of Contents to reflect changes within the chapter where indicated.
Page 4.1.2, modified and replaced text in Section 4.1.2 where indicated.
Page 4.1-3, modified text in Section 4.1.3.1.1 where indicated.
Pages 4.2-1 thru 4.2-2, modified text in Section 4.2 where indicated.
Page 4.2-3, modified text near the end of Section 4.2.1 where indicated.
Page 4.2-4, modified text in Section 4.2.2.1 where indicated.
Page 4.2-5, text flow changes.
Page 4.2-6, modified text in Section 4.2.3 where indicated.
Page 4.2-7, modified text in the first paragraph of Section 4.2.3.2 where indicated.
Pages 4.2-8 thru 4.2-9, text flow changes.
Page 4.2-10, modified text near the top of the page where indicated.
Pages 4.2-11 thru 4.2-12, text flow changes.
Page 4.2-13, modified Tables 4.2.1 and 4.2-2 and inserted new Note 1 and renumbered Notes 2 and 3 following Table 4.2.2 where indicated.
Page 4.2-14, Modified Tables 4.2-3 and 4.2-4, modified Note 1 and added Note 2 to Table 4.2-3 where indicated.
Page 4.2-15, modified Table 4.2-5 and added new Note following Tables 4.2-6 and 4.2-7 where indicated.
Page 4.2-16, modified page 1 of Table 4.2-8 where indicated.
Page 4.2-18, modified page 3 of Table 4.2-8 and added new Note where indicated.
Page 4.3-1, modified text in Section 4.3 where indicated.
Page 4.3-2, modified text in Section 4.3.2 where indicated.
Page 4.3-4, updated Tables 4.3-1 thru 4.3-3, added new Note 1 to Table 4.3-1 and renumbered Note 2 where indicated.
to ED20240097 Page 3 of 3 Chapter 5 Page 5-iv, modified Table of Contents to reflect new Section 5.12 where indicated.
Pages 5-v thru 5-xxi, text flow changes.
Pages 5.12-1 thru 5.12-3, added new Section 5.12.
Chapter 6 No Changes Chapter 7 Page 7-ii, modified the List of Tables to reflect changes within the chapter where indicated.
Page 7-2, modified last paragraph of Section 7.0 where indicated.
Pages 7.1-10 thru 7.1-12, deleted and modified text from Item 22 thru Item 24c, including deletion of the footnote at the bottom of the page, in Section 7.1.3.1 where indicated.
Pages 7.1-13 thru 7.1-17, text flow changes.
Page 7.4-2, modified the first paragraph of Section 7.4.1 where indicated.
Pages 7.4-3 thru 7.4-4, text flow changes.
Page 7.4-5, modified text in the first two paragraphs on the page in Section 7.4.3 where indicated.
Page 7.4-6, modified text at the end of the first paragraph in Section 7.4.4 where indicated.
Pages 7.4-7 thru 7.4-8, text flow changes.
Pages 7.4-9 thru 7.4-10, modified text in the last two pages of Table 7.4-1 where indicated.
Chapter 8 Page 8.1-12, deleted text near the end of the third paragraph of Section 8.1.5.1 where indicated.
Pages 8.1-13 thru 8.1-15, replaced text in the last paragraph of Section 8.1.5.1 and made extensive modifications to the text in Section 8.1.5.2, including deleting sections 8.1.5.3 and 8.1.5.4, where indicated.
Chapter 9 No Changes of ED20240097 Page 1 of 2 List of Calculations NAC-STC SAR, Revision 24A July 2024 of ED20240097 Page 2 of 2 List of Calculations Supporting Revision 24A
- 1. 423-5010, Revision 0
- 2. 423-5011, Revision 0 Calculation withheld in its entirety per 10 CFR 2.390.
to ED20240097 Page 1 of 1 SAR Changed Pages and LOEP NAC-STC SAR, Revision 24A July 2024
July 2024 Docket No. 71-9235 Revision 24A NAC-STC NAC Storage Transport Cask SAFETY ANALYSIS REPORT Non-Proprietary Version Volumes 1 and 2 Changed Pages Atlanta Corporate Headquarters : 2 Sun Court, Suite 220, Peachtree Corners, Georgia 30092 USA Phone 770-447-1144, www.nacintl.com
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A List of Effective Pages 1 of 6 Chapter 1 Page 1-i thru 1-viii........................ Revision 21 Pages 1-1 thru 1-12....................... Revision 21 Pages 1.1-1 thru 1.1-8................... Revision 21 Pages 1.2-1 thru 1.2-49................. Revision 21 Page 1.3-1..................................... Revision 21 Pages 1.4-1 thru 1.4-24................. Revision 21 100 drawings in Sections 1.3.2 and 1.4.3.2 Chapter 2 Pages 2-i thru 2-lxviii................... Revision 21 Page 2-1........................................ Revision 21 Pages 2.1.1-1 thru 2.1.1-5............. Revision 21 Pages 2.1.2-1 thru 2.1.2-5............. Revision 21 Pages 2.1.3-1 thru 2.1.3-15........... Revision 21 Pages 2.2-1 thru 2.2-8................... Revision 21 Pages 2.3.1-1 thru 2.3.1-2............. Revision 21 Page 2.3.2-1 thru 2.3.2-5............... Revision 21 Pages 2.3.3-1 thru 2.3.3-2............. Revision 21 Pages 2.3.4-1 thru 2.3.4-3............. Revision 21 Pages 2.3.5-1 thru 2.3.5-2............. Revision 21 Pages 2.3.6-1 thru 2.3.6-5............. Revision 21 Page 2.3.7-1.................................. Revision 21 Page 2.3.8-1.................................. Revision 21 Page 2.4-1..................................... Revision 21 Page 2.4.1-1.................................. Revision 21 Page 2.4.2-1.................................. Revision 21 Page 2.4.3-1.................................. Revision 21 Pages 2.4.4-1 thru 2.4.4-10........... Revision 21 Page 2.4.5-1.................................. Revision 21 Page 2.4.6-1.................................. Revision 21 Pages 2.5.1-1 thru 2.5.1-38........... Revision 21 Pages 2.5.2-1 thru 2.5.2-29........... Revision 21 Pages 2.6-1 thru 2.6-2................... Revision 21 Pages 2.6.1-1 thru 2.6.1-7............. Revision 21 Pages 2.6.2-1 thru 2.6.2-8............. Revision 21 Page 2.6.3-1.................................. Revision 21 Page 2.6.4-1.................................. Revision 21 Pages 2.6.5-1 thru 2.6.5-2............. Revision 21 Page 2.6.6-1.................................. Revision 21 Page 2.6.7-1.................................. Revision 21 Pages 2.6.7.1-1 thru 2.6.7.1-17..... Revision 21 Pages 2.6.7.2-1 thru 2.6.7.2-19..... Revision 21 Pages 2.6.7.3-1 thru 2.6.7.3-11..... Revision 21 Pages 2.6.7.4-1 thru 2.6.7.4-59..... Revision 21 Pages 2.6.7.5-1 thru 2.6.7.5-13..... Revision 21 Pages 2.6.7.6-1 thru 2.6.7.6-13..... Revision 21 Pages 2.6.7.7-1 thru 2.6.7.7-5....... Revision 21 Pages 2.6.7.8-1 thru 2.6.7.8-4....... Revision 21 Page 2.6.8-1.................................. Revision 21 Page 2.6.9-1.................................. Revision 21 Page 2.6.10-1................................ Revision 21 Pages 2.6.10.1-1 thru 2.6.10.1-2... Revision 21 Pages 2.6.10.2-1 thru 2.6.10.2-4... Revision 21 Pages 2.6.10.3-1 thru 2.6.10.3-7... Revision 21 Pages 2.6.11-1 thru 2.6.11-2......... Revision 21 Pages 2.6.11.1-1 thru 2.6.11.1-4... Revision 21 Pages 2.6.11.2-1 thru 2.6.11.2-11. Revision 21 Page 2.6.11.3-1............................. Revision 21 Pages 2.6.12-1 thru 2.6.12-5......... Revision 21 Page 2.6.12.1-1............................. Revision 21 Pages 2.6.12.2-1 thru 2.6.12.2-5... Revision 21 Pages 2.6.12.3-1 thru 2.6.12.3-7... Revision 21 Pages 2.6.12.4-1 thru 2.6.12.4-3... Revision 21 Pages 2.6.12.5-1 thru 2.6.12.5-3... Revision 21 Page 2.6.12.6-1 thru 2.6.12.6-2..... Revision 21 Pages 2.6.12.7-1 thru 2.6.12.7-22. Revision 21 Pages 2.6.12.8-1 thru 2.6.12.8-2... Revision 21 Pages 2.6.12.9-1 thru 2.6.12.9-11. Revision 21 Page 2.6.12.10-1........................... Revision 21 Page 2.6.12.11-1........................... Revision 21
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A List of Effective Pages (continued) 2 of 6 Page 2.6.12.12-1............................ Revision 21 Page 2.6.12.13-1............................ Revision 21 Pages 2.6.12.13-2 thru 2.6.12.13-4............................... Revision 21 Pages 2.6.13-1 thru 2.6.13-3.......... Revision 21 Pages 2.6.13.1-1 thru 2.6.13.1-2.... Revision 21 Pages 2.6.13.2-1 thru 2.6.13.2-7.... Revision 21 Pages 2.6.13.3-1 thru 2.6.13.3-4.... Revision 21 Pages 2.6.13.4-1 thru 2.6.13.4-5.... Revision 21 Pages 2.6.13.5-1 thru 2.6.13.5-2.... Revision 21 Pages 2.6.13.6-1 thru 2.6.13.6-2.... Revision 21 Pages 2.6.13.7-1 thru 2.6.13.7-2.... Revision 21 Page 2.6.13.8-1.............................. Revision 21 Page 2.6.13.9-1.............................. Revision 21 Page 2.6.13.10-1............................ Revision 21 Pages 2.6.13.11-1 thru 2.6.13.11-3............................... Revision 21 Pages 2.6.13.12-1 thru 2.6.13.12-2............................... Revision 21 Pages 2.6.14-1 thru 2.6.14-8.......... Revision 21 Pages 2.6.14.1-1 thru 2.6.14.1-2.... Revision 21 Pages 2.6.14.2-1 thru 2.6.14.2-16.. Revision 21 Pages 2.6.14.3-1 thru 2.6.14.3-3.... Revision 21 Pages 2.6.14.4-1 thru 2.6.14.4-4.... Revision 21 Pages 2.6.14.5-1 thru 2.6.14.5-3.... Revision 21 Page 2.6.14.6-1.............................. Revision 21 Pages 2.6.14.7-1 thru 2.6.14.7-14.. Revision 21 Pages 2.6.14.8-1 thru 2.6.14.8-6.... Revision 21 Page 2.6.14.9-1.............................. Revision 21 Page 2.6.14.10-1............................ Revision 21 Pages 2.6.14.11-1 thru 2.6.14.11-5............................... Revision 21 Pages 2.6.14.12-1 thru 2.6.14.12-5............................... Revision 21 Page 2.6.15-1................................. Revision 21 Pages 2.6.15.1-1 thru 2.6.15.1-2.... Revision 21 Pages 2.6.15.2-1 thru 2.6.15.2-7.... Revision 21 Pages 2.6.15.3-1 thru 2.6.15.3-4.... Revision 21 Pages 2.6.15.4-1 thru 2.6.15.4-4.... Revision 21 Page 2.6.15.5-1.............................. Revision 21 Pages 2.6.15.6-1 thru 2.6.15.6-3.... Revision 21 Page 2.6.15.7-1.............................. Revision 21 Page 2.6.15.8-1.............................. Revision 21 Page 2.6.15.9-1.............................. Revision 21 Page 2.6.15.10-1............................ Revision 21 Pages 2.6.15.11-1 thru 2.6.15.11-3............................... Revision 21 Pages 2.6.15.12-1 thru 2.6.15.12-2............................... Revision 21 Pages 2.6.16-1 thru 2.6.16-6.......... Revision 21 Pages 2.6.16.1-1 thru 2.6.16.1-2.... Revision 21 Pages 2.6.16.2-1 thru 2.6.16.2-11.. Revision 21 Pages 2.6.16.3-1 thru 2.6.16.3-3.... Revision 21 Pages 2.6.16.4-1 thru 2.6.16.4-3.... Revision 21 Pages 2.6.16.5-1 thru 2.6.15.5-3.... Revision 21 Pages 2.6.16.6-1 thru 2.6.16.6-3.... Revision 21 Pages 2.6.16.7-1 thru 2.6.16.7-12.. Revision 21 Pages 2.6.16.8-1 thru 2.6.16.8-7.... Revision 21 Page 2.6.16.9-1.............................. Revision 21 Page 2.6.16.10-1............................ Revision 21 Pages 2.6.16.11-1 thru 2.6.16.11-4............................... Revision 21 Pages 2.6.16.12-1 thru 2.6.16.12-2............................... Revision 21 Pages 2.6.16.13-1 thru 2.6.16.13-2............................... Revision 21 Page 2.6.16.14-1............................ Revision 21 Pages 2.6.17-1 thru 2.6.17-13........ Revision 21 Pages 2.6.18-1 thru 2.6.18-6.......... Revision 21 Pages 2.6.19-1 thru 2.6.19-23........ Revision 21 Pages 2.6.20-1 thru 2.6.20-20........ Revision 21 Pages 2.6.21-1 thru 2.6.21.-2......... Revision 21 Pages 2.7-1 thru 2.7-2.................... Revision 21 Page 2.7.1-1 thru 2.7.1-2............... Revision 21
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A List of Effective Pages (continued) 3 of 6 Pages 2.7.1.1-1 thru 2.7.1.1-15..... Revision 21 Pages 2.7.1.2-1 thru 2.7.1.2-15..... Revision 21 Pages 2.7.1.3-1 thru 2.7.1.3-9....... Revision 21 Pages 2.7.1.4-1 thru 2.7.1.4-11..... Revision 21 Pages 2.7.1.5-1 thru 2.7.1.5-3....... Revision 21 Pages 2.7.1.6-1 thru 2.7.1.6-16..... Revision 21 Page 2.7.2-1.................................. Revision 21 Pages 2.7.2.1-1 thru 2.7.2.1-5....... Revision 21 Pages 2.7.2.2-1 thru 2.7.2.2-9....... Revision 21 Pages 2.7.2.3-1 thru 2.7.2.3-6....... Revision 21 Pages 2.7.2.4-1 thru 2.7.2.4-7....... Revision 21 Page 2.7.2.5-1............................... Revision 21 Page 2.7.2.6-1............................... Revision 21 Page 2.7.3.1-1............................... Revision 21 Pages 2.7.3.2-1 thru 2.7.3.2-5....... Revision 21 Pages 2.7.3.3-1 thru 2.7.3.3-3....... Revision 21 Pages 2.7.3.4-1 thru 2.7.3.4-2....... Revision 21 Page 2.7.3.5-1............................... Revision 21 Page 2.7.3.6-1............................... Revision 21 Page 2.7.4-1.................................. Revision 21 Page 2.7.5-1.................................. Revision 21 Page 2.7.6-1.................................. Revision 21 Pages 2.7.7-1 thru 2.7.7-4............. Revision 21 Pages 2.7.8-1 thru 2.7.8-4............. Revision 21 Pages 2.7.8.1-1 thru 2.7.8.1-43..... Revision 21 Pages 2.7.8.2-1 thru 2.7.8.2-2....... Revision 21 Pages 2.7.8.3-1 thru 2.7.8.3-6....... Revision 21 Pages 2.7.8.3-7 thru 2.7.8.3-13..... Revision 21 Pages 2.7.8.4-1 thru 2.7.8.4-10..... Revision 21 Page 2.7.8.5-1............................... Revision 21 Pages 2.7.9-1 thru 2.7.9-40........... Revision 21 Pages 2.7.10-1 thru 2.7.10-12....... Revision 21 Pages 2.7.11-1 thru 2.7.11-16....... Revision 21 Pages 2.7.12-1 thru 2.7.12-10....... Revision 21 Pages 2.7.13-1 thru 2.7.13-4......... Revision 21 Pages 2.7.13.1-1 thru 2.7.13.1-18. Revision 21 Pages 2.7.13.2-1 thru 2.7.13.2-2... Revision 21 Pages 2.7.13.3-1 thru 2.7.13.3-4... Revision 21 Pages 2.7.13.4-1 thru 2.7.13.4-8... Revision 21 Pages 2.7.13.5-1 thru 2.7.13.5-2... Revision 21 Pages 2.7.14-1 thru 2.7.14-13....... Revision 21 Pages 2.7.15-1 thru 2.7.15-16....... Revision 21 Page 2.8-1..................................... Revision 21 Pages 2.9-1 thru 2.9-11................. Revision 21 Pages 2.10.1-1 thru 2.10.1-4......... Revision 21 Pages 2.10.2-1 thru 2.10.2-93....... Revision 21 Pages 2.10.3-1 thru 2.10.3-7......... Revision 21 Pages 2.10.4-1 thru 2.10.4-288..... Revision 21 Pages 2.10.5-1 thru 2.10.5-22....... Revision 21 Pages 2.10.6-1 thru 2.10.6.-36...... Revision 21 13 drawings in Sections 2.10.6.6 and 2.10.6.7 Pages 2.10.6-37 thru 2.10.6-88..... Revision 21 Pages 2.10.7-1 thru 2.10.7-26....... Revision 21 Pages 2.10.8-1 thru 2.10.8-24....... Revision 21 Pages 2.10.9-1 thru 2.10.9-11....... Revision 21 Pages 2.10.10-1 thru 2.10.10-11... Revision 21 Pages 2.10.11-1 thru 2.10.11-8..... Revision 21 Pages 2.10.12-1 thru 2.10.12-31... Revision 21 4 drawings in Section 2.10.12 Pages 2.11.1-1 thru 2.11.1-2......... Revision 21 Pages 2.11.2-1 thru 2.11.2-2......... Revision 21 Page 2.11.3-1................................ Revision 21 Page 2.11.4-1................................ Revision 21 Page 2.11.5-1................................ Revision 21 Pages 2.11.6-1 thru 2.11.6-6......... Revision 21 Page 2.11.6.12-1 thru 2.11.6.12-62............................ Revision 21 Pages 2.11.6.13-1 thru 2.11.6.13-35............................ Revision 21
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A List of Effective Pages (continued) 4 of 6 Pages 2.11.6.14-1 thru 2.11.6.14-10............................. Revision 21 Page 2.11.6.15-1............................ Revision 21 Pages 2.11.7-1 thru 2.11.7-8.......... Revision 21 Pages 2.11.7.8-1 thru 2.11.7.8-34.. Revision 21 Pages 2.11.7.9-1 thru 2.11.7.9-14.. Revision 21 Pages 2.11.7.10-1 thru 2.11.7.10-5............................... Revision 21 Page 2.11.8-1................................. Revision 21 Pages 2.11.9-1 thru 2.11.9-10........ Revision 21 Page 2.12.1-1................................. Revision 21 Pages 2.12.2-1 thru 2.12.2-2.......... Revision 21 Page 2.12.3-1................................. Revision 21 Page 2.12.4-1................................. Revision 21 Page 2.12.5-1................................. Revision 21 Page 2.12.6-1 thru 2.12.6-29......... Revision 21 Page 2.13.1-1................................. Revision 21 Page 2.13.2-1 thru 2.13.2-2........... Revision 21 Page 2.13.3-1................................. Revision 21 Page 2.13.4-1................................. Revision 21 Page 2.13.5-1................................. Revision 21 Pages 2.13.6-1 thru 2.13.6-60........ Revision 21 Chapter 3 Pages 3-i thru Pages 3-viii............. Revision 21 Pages 3.1-1 thru 3.1-12.................. Revision 21 Pages 3.2-1 thru 3.2-14.................. Revision 21 Pages 3.3-1 thru 3.3-6.................... Revision 21 Pages 3.4-1 thru 3.4-30.................. Revision 21 Pages 3.4-31 thru 3.4-33............. Revision 24A Pages 3.4-34 thru 3.4-86................. Revison 21 Pages 3.5-1 thru 3.5-6.................... Revision 21 Pages 3.5-7................................. Revision 24A Pages 3.5-8 thru 3.5-16.................. Revision 21 Page 3.6-1...................................... Revision 21 Pages 3.6.1-1 thru 3.6.1-4.............. Revision 21 Pages 3.6.2-1 thru 3.6.2-3.............. Revision 21 Pages 3.6.3-1 thru 3.6.3-3.............. Revision 21 Pages 3.6.4-1 thru 3.6.4-24............ Revision 21 Pages 3.6.5-1 thru 3.6.5-3.............. Revision 21 Page 3.7-1...................................... Revision 21 Pages 3.7.1-1 thru 3.7.1-3.............. Revision 21 Pages 3.7.2-1 thru 3.7.2-2.............. Revision 21 Pages 3.7.3-1 thru 3.7.3-2.............. Revision 21 Pages 3.7.4-1 thru 3.7.4-9.............. Revision 21 Page 3.7.5-1 thru 3.7.5-2............... Revision 21 Page 3.8-1...................................... Revision 21 Pages 3.8.1-1 thru 3.8.1-4.............. Revision 21 Pages 3.8.2-1 thru 3.8.2-3.............. Revision 21 Pages 3.8.3-1 thru 3.8.3-3.............. Revision 21 Pages 3.8.4-1 thru 3.8.4-15............ Revision 21 Pages 3.8.5-1 thru 3.8.5-2.............. Revision 21 Page 3.8.6-1................................ Revision 24A Chapter 4 Pages 4-i..................................... Revision 24A Page 4-ii thru 4-iii.......................... Revision 21 Page 4.1-1...................................... Revision 21 Pages 4.1-2 thru 4.1-3................. Revision 24A Pages 4.1-4 thru 4.1-10.................. Revision 21 Pages 4.2-1 thru 4.2-16............... Revision 24A Page 4.2-17..................................... Revison 21 Page 4.2-18................................. Revision 24A Pages 4.3-1 thru 4.3-2................. Revision 24A Page 4.3-3...................................... Reivsion 21 Page 4.3-4................................... Revision 24A Page 4.4-1...................................... Revision 21 Pages 4.5-1 thru 4.5-38.................. Revision 21 Pages 4.6-1 thru 4.6-2.................... Revision 21 Pages 4.7-1 thru 4.7-3.................... Revision 21
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A List of Effective Pages (continued) 5 of 6 Chapter 5 Pages 5-i thru 5-iii......................... Revision 21 Pages 5-iv thru 5-xxi.................... Revison 24A Pages 5-1 thru 5-4......................... Revision 21 Pages 5.1-1 thru 5.1-30................. Revision 21 Pages 5.2-1 thru 5.2-40................. Revision 21 Pages 5.3-1 thru 5.3-33................. Revision 21 Pages 5.4-1 thru 5.4-42................. Revision 21 Pages 5.5-1 thru 5.5-61................. Revision 21 Page 5.6-1..................................... Revision 21 Pages 5.6.1-1 thru 5.6.1-9............. Revision 21 Pages 5.6.2-1 thru 5.6.2-20........... Revision 21 Pages 5.6.3-1 thru 5.6.3-13........... Revision 21 Pages 5.6.4-1 thru 5.6.4-34........... Revision 21 Page 5.6.5-1.................................. Revision 21 Pages 5.6.6-1 thru 5.6.6-57........... Revision 21 Page 5.7-1..................................... Revision 21 Pages 5.7.1-1 thru 5.7.1-5............. Revision 21 Pages 5.7.2-1 thru 5.7.2-5............. Revision 21 Pages 5.7.3-1 thru 5.7.3-10........... Revision 21 Pages 5.7.4-1 thru 5.7.4-14........... Revision 21 Page 5.7.5-1.................................. Revision 21 Pages 5.7.6-1 thru 5.7.6-22........... Revision 21 Page 5.8-1..................................... Revision 21 Pages 5.8.1-1 thru 5.8.1-9............. Revision 21 Page 5.8.2-1 thru 5.8.2-8............... Revision 21 Pages 5.8.3-1 thru 5.8.3-6............. Revision 21 Pages 5.8.4-1 thru 5.8.4-3............. Revision 21 Pages 5.8.5-1 thru 5.8.5-3............. Revision 21 Pages 5.8.6-1 thru 5.8.6-5............. Revision 21 Pages 5.8.7-1 thru 5.8.7-10........... Revision 21 Pages 5.8.8-1 thru 5.8.8-6............. Revision 21 Pages 5.8.9-1 thru 5.8.9-26........... Revision 21 Pages 5.9-1 thru 5.9-5................... Revision 21 Pages 5.10-1 thru 5.10-6............... Revision 21 Pages 5.11-1 thru 5.11-3............... Revision 21 Pages 5.12-1 thru 5.12-3............ Revision 24A Chapter 6 Pages 6-i thru 6-ix......................... Revision 21 Pages 6.1-1 thru 6.1-6................... Revision 21 Pages 6.2-1 thru 6.2-11................. Revision 21 Pages 6.3-1 thru 6.3-10................. Revision 21 Pages 6.4-1 thru 6.4-2................... Revision 21 Page 6.4.1-1.................................. Revision 21 Pages 6.4.2-1 thru 6.4.2-11........... Revision 21 Pages 6.4.3-1 thru 6.4.3-29........... Revision 21 Pages 6.4.4-1 thru 6.4.4-30........... Revision 21 Pages 6.5-1 thru 6.5-2................... Revision 21 Pages 6.5.1-1 thru 6.5.1-21........... Revision 21 Pages 6.5.2-1 thru 6.5.2-20........... Revision 21 Pages 6.6-1 thru 6.6-2................... Revision 21 Pages 6.7-1 thru 6.7-333............... Revision 21 Page 6.8-1..................................... Revision 21 Pages 6.8.1-1 thru 6.8.1-6............. Revision 21 Pages 6.8.2-1 thru 6.8.2-2............. Revision 21 Pages 6.8.3-1 thru 6.8.3-20........... Revision 21 Pages 6.8.4-1 thru 6.8.4-34........... Revision 21 Pages 6.8.5-1 thru 6.8.5-34........... Revision 21 Page 6.8.6-1.................................. Revision 21 Pages 6.8.7-1 thru 6.8.7-27........... Revision 21 Page 6.9-1..................................... Revision 21 Page 6.9.1-1.................................. Revision 21 Page 6.9.2-1.................................. Revision 21 Chapter 7 Page 7-i......................................... Revision 21 Page 7-ii..................................... Revision 24A Page 7-1........................................ Revision 21 Page 7-2..................................... Revision 24A Page 7-3........................................ Revision 21
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A List of Effective Pages (continued) 6 of 6 Pages 7.1-1 thru 7.1-9.................... Revision 21 Pages 7.1-10 thru 7.1-17............. Revision 24A Pages 7.2-1 thru 7.2-5.................... Revision 21 Pages 7.3-1 thru 7.3-11.................. Revision 21 Page 7.4-1...................................... Revision 21 Pages 7.4-2 thru 7.4-10............... Revision 24A Page 7.5-1...................................... Revision 21 Pages 7.6-1 thru 7.6-6.................... Revision 21 Chapter 8 Page 8-i thru 8-ii............................ Revision 21 Page 8-1......................................... Revision 21 Pages 8.1-1 thru 8.1-11.................. Revision 21 Pages 8.1-12 thru 8.1-15............. Revision 24A Pages 8.1-16 thru 8.1-39................ Revision 21 Pages 8.2-1 thru 8.2-7.................... Revision 21 Page 8.3-1...................................... Revision 21 Pages 8.4-1 thru 8.4-13.................. Revision 21 Chapter 9 Page 9-i.......................................... Revision 21 Pages 9-1 thru 9-13........................ Revision 21
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 3.4-31 operating pressure (MNOP) is obtained from 26 Westinghouse 15 x 15 (Standard) PWR fuel assemblies, using a maximum burnup of 45,000 MWD/MT. The standard burnup, up to 45 GWd/MTU, configuration bounds the pressure of the high burnup (HBU) configuration under NCT conditions (HBU analysis is included in Section 3.8). Calculation of the NAC-STC cavity maximum operating pressure utilizes the gas volume of the cavity, the temperature of the cavity gases and the volume of gases released by the fuel to the cavity. The characteristics of the Westinghouse 15 x 15 fuel assembly pertinent to this analysis are given in Table 3.4-8. The cask cavity dimensions are:
Cask Cavity Inner Diameter 71.0 in Cask Cavity Length 165.0 in Cask Cavity Volume 653,267 in3 The volumes for the directly loaded fuel basket components (Vb) are given in Table 3.4-9.
Vb = 77,448 in3 The volume of the fuel assemblies (Vf) is:
Vf = (Vrods + Vtubes + Vendfittings + Vgrids) (26)
= (4,358.6 in3 + 71.4 in3 + 133.1 in3 + 85.8 in3 ) (26)
= 120,871 in3 The free gas volume in the cask cavity (Vgv) is:
Vgv = Vc -Vb - Vr
= 454,947 in3
= 7,455 liters The fuel rod total free gas volume (Vfg) is:
Vfg = (26)(1.3)(204)
= 6,895 in3 = 113 liters The gaseous fission product inventory can be determined from the ORIGEN-S fission product inventory and the Ideal Gas Law. Regulatory Guide 1.25 states that, of the gaseous fission product inventory in the fuel, 10 percent of all noble gases except Kr, 30 percent of the available Kr, and 10 percent of the 127I and 129I should be considered for release. Conservatively, a 30 percent release rate has been assumed for all of the fission product gases. The fission gas inventories are taken directly from SAS2H output for a burnup of 45,000 MWD/MTU, with an initial enrichment of 3.5 wt % 235U and 40 years cool time. The total number of releasable
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 3.4-32 moles, adjusted by the 30% release fraction, is 8.96 moles/assembly. Isotopes considered include 3H, 4He and all the isotopes of krypton, xenon and iodine.
The Ideal Gas Law can then be used to determine the volume of gas at room temperature and atmospheric pressure.
PV = nRT where:
n
= number of moles of gas R
= gas constant = 0.0821 atm mol K
T
= temperature in K = 293.15 P
= pressure = 1 atm V
= volume of gas and:
nrodbackfill = (3.10E-2 moles/rod)(204)(26)
= 164 moles nfissiongas = (8.96 moles)(26)
233 moles nSTCbackfill
15 293 0821
.0 455
,7 1
= 310 moles nTotal = nrodbackfill + nfissiongas + nSTCbackfill
= 707 moles The bulk average temperature is calculated for both helium and air cavity gas using the results from the 180-degree section three-dimensional cask finite element heat transfer model. The temperature of the cavity gas is defined as being equal to the local metal temperature. The temperature of the gas within the fuel tube is defined as the value calculated for the individual assembly cladding. The average temperature of the gas is then obtained by integrating the gas temperature in both the radial and axial directions. Following this method, the actual calculated value for the bulk average temperature with a helium-filled cavity is 401°F and with an air-filled cavity is 411°F. Based on a conservative bulk average gas temperature of 450°F, the maximum operating pressure within the cask cavity assuming 100 percent fuel rod failure is:
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 3.4-33 P =
7070.0821 7,568
= 3.87 atm = 56.9 psia = 42.2 psig The operating pressure for three percent fuel rod failure is calculated based on the heatup of helium in the cask cavity together with the assumed release of fission gases from the failed rods.
The calculated molar inventory of the released fission gases (nrf) with three percent fuel rod failure is:
nrf
= nfissiongas x 0.03 nfissiongas = Total moles of released fission gas = 233 moles nrf
= 7 moles The calculated molar inventory of the released rod backfill gases with 3 % fuel rod failure is:
nrr
= nrodbackfill x 0.03 nrodbackfill = Total moles of released rod backfill gas = 164 moles nrf
= 5 moles The total moles available for release with 3 % fuel rod failure is:
ntotal = nrf + nrr + nSTCbackfill ntotal
= 322 moles The total helium released from the assumed failed rods is:
Vrh = Vfg x 0.03 Vfg = Fuel free gas volume = 6,895 in3 Vrh = 207 in3 = 4 liters The total gas volume for normal conditions of transport is then:
VNC = Total gas volume = Vgv + Vrh
= 455,154 in3 = 7,459 liters Based on a conservative bulk average gas temperature of 450F, the operating pressure within the cask cavity assuming three percent fuel rod failure is:
P2 =
3220.0821 7,459
= 1.79 atm = 26.3 psia = 11.6 psig
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 3.5-7 3.5.4 Maximum Internal Pressure 3.5.4.1 Maximum Internal Pressure Due to Directly Loaded Fuel From Section 3.4.4, it is known that the maximum pressure in the cask cavity is 56.9 psia, for a cavity gas temperature of 232°C. The maximum fuel rod cladding temperature during the fire transient is 402°C, as listed in Table 3.5-1. Thus, the maximum hypothetical accident pressure can be calculated based on the ratio of these temperatures as follows:
P2 = P T T
1 2
1 P2 = 56.9 675 505 P2 = 76.1 psia 3.5.4.2 Maximum Internal Pressure Due to Yankee Class Canistered Fuel The maximum internal pressure for the canistered fuel configuration is calculated for the canister and for the NAC-STC cavity. The calculated maximum post fire accident temperature of the helium gas is 616F determined by the method discussed in Section 3.5.1.1.1 (442F for normal condition + 174F). A temperature of 650F is conservatively used to calculate the maximum pressure in the NAC-STC.
The internal pressure is a function of rod-fill, fission and backfill gases. The design basis fuel assembly for the internal pressure calculation is the Combustion Engineering Type A assembly.
This assembly has the highest rod back-fill pressure (315 psig) and received the highest burnup (36,000 MWD/MTU). There are three different gases contributing to the canister internal pressure and four gases contributing to the cavity internal pressure. The canister gases are the fuel rod back-fill and fission gases, and the canister backfill gas. The cavity gases are these plus the cavity backfill gas. All of the gases, except the fission gases, are assumed to be helium. The total pressure for each volume is found by calculating the molar quantity of each gas and summing those directly.
The number of moles of the backfill gases are calculated using the Ideal Gas Law, PV = NRT.
Backfill gases for the canister and cavity are assumed to be initially at 1 atmosphere. The quantity of fission gas is derived using the SAS2H fraction of gas atoms of 0.3125 atoms of gas
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 3.8.6-1 3.8.6 Maximum Pressure During Normal and Hypothetical Accident Conditions (HAC) of Transport The pressure calculation documented in Sections 3.4.4 and 3.5.4 is repeated in this section for the high burnup/high heat load payload discussed in Section 3.8.
Parameters that differ between low burnup (LBU) fuel and high burnup (HBU) fuel are summarized below.
Parameter LBU HBU Burnup (MWd/MTU) 45,000 60,000 Maximum # of Assemblies 26 20
- of Heat Transfer Shunts 0
6 Fission Gas Inventory (moles/assy) 29.87 39.46 Fission Gas Release, NCT/HAC 30%/30%
(NUREG-2216) 15%/35%
Fuel Assembly Volume (inch3) 19,185 14,758 Heat Transfer Shunt Volume (inch3) 0 47,014 Cask Cavity Free Volume (liters) 7,455 7,214 The gas temperatures for LBU fuel bound those for HBU fuel and are used in the pressure evaluation.
Using the parameters above and the methodology outlined in Sections 3.4.4 and 3.5.4, pressure results are as follows:
The maximum normal operating pressure assuming 100 percent fuel rod failure is 3.09 atm
= 45.5 psia = 30.8 psig.
The normal operating pressure assuming three percent fuel rod failure is 1.76 atm = 25.9 psia = 11.2 psig.
The maximum hypothetical accident pressure is 5.33 atm = 78.3 psia = 63.6 psig.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4-i Table of Contents 4.0 CONTAINMENT........................................................................................................ 4.1-1 4.1 Containment Boundary................................................................................................. 4.1-1 4.1.1 Containment Vessel.......................................................................................... 4.1-2 4.1.2 Containment Penetrations................................................................................. 4.1-2 4.1.3 Seals and Welds................................................................................................ 4.1-2 4.1.4 Closure.............................................................................................................. 4.1-4 4.2 Containment Requirements for Normal Conditions of Transport................................ 4.2-1 4.2.1 Containment of Radioactive Material............................................................... 4.2-3 4.2.2 Pressurization of Containment Vessel.............................................................. 4.2-3 4.2.3 Containment Criterion for Normal Conditions of Transport............................ 4.2-5 4.3 Containment Requirements for Hypothetical Accident Conditions............................. 4.3-1 4.3.1 Fission Gas Products......................................................................................... 4.3-1 4.3.2 Containment of Radioactive Material............................................................... 4.3-2 4.3.3 Calculation of Allowable Leak Rate for Directly Loaded Fuel With Viton O-rings........................................................................................... 4.3-2 4.3.4 Containment Criterion for Accident Conditions............................................... 4.3-3 4.4 Special Requirements.................................................................................................... 4.4-1 4.5 Appendix....................................................................................................................... 4.5-1 4.5.1 Metallic O-rings................................................................................................ 4.5-1 4.5.2 Blended Polytetrafluoroethylene (PTFE) O-rings.......................................... 4.5-14 4.5.3 Expansion Foam.............................................................................................. 4.5-23 4.5.4 Fiberfrax Ceramic Fiber Paper........................................................................ 4.5-26 4.5.5 Viton O-rings.................................................................................................. 4.5-32 4.5.6 Sample SAS2H Input File............................................................................... 4.5-37
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.1-2 4.1.1 Containment Vessel The primary containment vessel for the NAC-STC consists of a 71.0-inch inside diameter, 1.5-inch thick inner shell, two 1.5-inch to 2.0-inch thick transition sections, a 6.2-inch thick bottom inner forging, and a 7.85-inch thick top forging. The containment vessel components, except for the transition sections, are fabricated from ASME Boiler and Pressure Vessel Code, Type 304 stainless steel nuclear pressure vessel material. The two transition sections are ASME Boiler and Pressure Vessel Code, Type XM-19 stainless steel nuclear pressure vessel material.
The weld examination requirements for the cask body are defined in Table 4.1-2 and are shown on the drawings in Section 1.3.2.
4.1.2 Containment Penetrations The physical penetrations in the NAC-STC containment vessel are the inner lid and the vent and drain ports in the inner lid. The penetrations are designed to ensure sealing of the containment boundary and to ensure that the leakage from the boundary does not exceed:
1 x 10-7 ref cm3/sec using metallic seals, or Using nonmetallic O-rings:
o 1 x 10-7 ref cm3/sec (leak tight criterion), or o 9.4 x 10-5 ref cm3/sec (non-leak tight LBU fuel contents), or o 2.6 x 10-6 ref cm3/sec (non-leak tight HBU fuel contents), or o Preshipment leakage test at a minimum sensitivity of 1 x 10-3 ref cm3/sec.
The quick-disconnect fittings installed in the vent and drain openings and in the interseal test port in the inner lid are not considered to be part of the containment boundary.
4.1.3 Seals and Welds 4.1.3.1 Seals The O-rings of the inner lid, the vent port coverplate and the drain port coverplate are the seals that provide primary containment, as described in Section 4.1 and as shown in Table 4.1-1.
Section 4.5 contains the specifications that describe the PTFE O-rings of the interlid and pressure port covers, and the metallic or nonmetallic O-rings used in the containment boundary and outer lid. Also included in Section 4.5 are the manufacturers technical data bulletins for the expansion foam and the Fiberfrax Ceramic Fiber Paper. Leak testing of the cask is performed prior to acceptance from the manufacturer. Leak testing is also performed following fuel loading
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.1-3 for either immediate transport or for transport following a storage period. Technical information for Viton O-rings is provided in Section 4.5.5.
4.1.3.1.1 Containment System Fabrication Verification Upon completion of fabrication, a Containment System Fabrication Leakage Rate Test shall be performed on the cask containment boundary as described in Section 8.1.3. These leakage tests verify that the leakage rate of containment components does not exceed the maximum allowable leakage rate of 1 x 10-7 ref cm3/sec (leaktight testing applies to individual seals) for metallic O-rings or the total (cumulative) leakage rate of containment components does not exceed 9.4 x 10-5 ref cm3/sec or 2.6 x 10-6 ref cm3/sec for Viton O-rings with LBU or HBU fuel contents, respectively. Viton O-rings may also be tested to leaktight conditions for HBU PWR spent fuel assembly contents by confirming the maximum allowable leakage rate of 1 x 10-7 ref cm3/sec (leaktight testing applies to individual seals). The allowable leakage rate test shall conform to the O-ring design, since the inner and outer lids, and the vent and drain port coverplates must be fabricated using the O-ring groove appropriate to the O-ring design. Metallic O-rings and nonmetallic O-rings cannot be used interchangeably.
4.1.3.1.2 Containment System Verification The containment system preshipment leakage rate test shall be performed on the NAC-STC package containment boundary seals and components prior to each shipment for packages assembled with metallic O-rings in accordance with the leak test acceptance criteria defined in the procedures in Chapter 8. For cask transport immediately after loading, the preshipment leakage rate test for NAC-STC packages assembled with Viton O-ring containment boundary seals shall be performed in accordance with the procedures and acceptance criteria described in Section 7.4.1. For cask shipments following storage, the verification leakage rate test shall be performed in accordance with the procedures and acceptance criteria described in Section 7.4.2.
Whenever a containment component is replaced, the containment component shall be leak tested following replacement using the maintenance leakage rate test (Section 8.2.2.2). This test will verify that the replacement component has been properly fabricated and installed, and that the leakage rate meets acceptance criteria. For packaging using Viton O-rings, the periodic leakage rate test shall be performed annually, and the maintenance leakage rate test is performed if a Viton O-ring is replaced during loading operations in accordance with the acceptance criteria in Section 8.2.2.2.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-1 4.2 Containment Requirements for Normal Conditions of Transport The NAC-STC has been designed to safely transport spent fuel assemblies in either of two configurations. The spent fuel assemblies may be sealed in a transportable storage canister (canistered), or loaded directly into a fuel basket installed in the cask cavity.
In the canistered configuration, the NAC-STC can transport Yankee Class or Connecticut Yankee spent fuel and GTCC waste. The NAC-STC in this configuration is designed and tested to leaktight conditions as defined by ANSI N14.5-1997 and, therefore, meets the requirements of 10 CFR 71.51 for containment of radioactive materials.
For directly loaded fuel, a reference 1717 fuel assembly is used to establish the source term for the containment analysis. For LBU fuel, a burnup of 45,000 MWd/MTU, an enrichment of 2.3 wt % 235U, and a cool time of 5 years is evaluated. For HBU fuel, a burnup of 60,000 MWd/MTU, an enrichment of 3.5 wt % 235U, and a cool time of 5 years is evaluated. The reference fuel assembly is also used in the shielding analysis and is described in Section 5.1.2.
The directly loaded fuel assemblies are the only payloads having the option of employing the nonmetallic Viton O-ring seals which are not required to be tested to leaktight conditions and are, therefore, the only payloads addressed in the containment evaluation. A minimum 5-year cool time is employed in this analysis. Note that shorter cool times are allowed for loadings of HBU fuel. These loadings have significantly less inventory such that the loading is bounding.
As shown in Chapter 5, Table 5.2-1, the reference 17x17 fuel assembly contains a slightly lower
(<1% difference) fuel mass than the 15x15 reference assembly, but significantly higher mass/source than the 14x14 and 16x16 reference assemblies. Combined with the highest in-core power per assembly (see Table 5.2-2), maximum source terms are calculated for the 17x17 reference assembly.
A minimum enrichment of 2.3 wt % was chosen as a reasonable lower enrichment band for the 45,000 MWd/MTU fuel. A minimum enrichment of 3.5 wt % was chosen as a reasonable lower enrichment band for the 60,000 MWd/MTU fuel. These initial enrichments are significantly lower than typical fuel discharged at these burnup levels.
NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-2 The power history, including power per assembly and downtime at the end of the cycle, for each reference assembly evaluated is included in SAR Table 5.2-2. For the 17x17 reference fuel assembly type, a power level of 18.55 MW per assembly and 60 days of downtime between cycles is applied. The power level is based on a review of power plant thermal output divided by the number of assemblies in the core (dependent on fuel type selected). The downtime between cycles is an estimate based on overall cycle length, including ramp-up and down in power. A conservative three-power cycle history is applied to generate the 60,000 MWd/MTU burnup.
This power history produces a set of three approximately 18-month-long cycles producing a combined time of 54 months between BOL (beginning of life) and discharge. A 54-month fuel life is considered bounding for generating a 60,000 MWd/MTU burnup fuel assembly. For 45,000 MWd/MTU fuel, three cycles are also used.
The complete SAS2H model, including power level and cool time, for the generation of containment source terms is included in Section 4.5.6.
For direct loading for immediate transport or transport after interim storage using metallic O-rings in the containment boundary, the containment boundary preshipment leakage rate test is performed is tested to a leaktight condition as defined in ANSI N14.5-1997. For direct loading for immediate transport using reusable Viton O-rings, the containment boundary preshipment leakage rate test is performed to confirm no detected leakage of each containment seal when tested to a sensitivity of 1 x 10-3 ref cm3/sec for standard burnup and HBU fuel ( 55 GWd/MTU assembly average). If Viton O-rings are replaced due to wear, damage or leak test failure, the replacement seals shall be tested to the Maintenance Leakage Rate Test acceptance criteria of either:
leaktight condition of 1.0 10-7 ref cm3/sec for high burnup fuel, or
2.6 10-6 ref cm3/sec, or 4.0 10-6 cm3/sec (helium) for high burnup fuel, or
9.4 10-5 ref cm3/sec, or 1.1 10-4 cm3/sec (helium) for standard burnup fuel
( 45 GWd/MTU assembly average).
The structural integrity of the cask containment during normal conditions of transport is demonstrated in Section 2.6.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-3 4.2.1 Containment of Radioactive Material The NAC-STC uses one of three O-ring configurations based on the loading condition. For directly loading of fuel for transport without interim storage, the inner and outer O-rings of the inner lid and vent and drain port coverplates may be either metallic or Viton in the following potential arrangements: inner and outer Viton O-rings; inner and outer metallic seals; and inner metallic seal and outer Viton O-ring. For direct fuel loading for storage, both the inner and outer O-rings must be metallic. For loading of canistered fuel or GTCC waste, or HLW overpacks for transport, the inner and outer O-rings must also be metallic. For configurations using metallic O-rings, the containment boundary is designed and tested to leaktight conditions as defined by ANSI N14.5-1997. For direct fuel loading for transport of standard burnup fuel without interim storage using Viton O-rings, the allowable leak rate is calculated using the methodology of NUREG/CR-6487. For high burnup (HBU) fuel the Viton O-ring or metallic inner seals may be tested to leaktight conditions, otherwise the allowable leak rate is calculated using the methodology of NUREG/CR-6487. Consequently, the cask meets the requirements of 10 CFR 71.51 for directly loaded and for canistered fuel or GTCC waste, and HLW overpacks.
4.2.2 Pressurization of Containment Vessel The maximum normal operating pressure in the cask during normal transport conditions is conservatively based on 100% failure of the fuel rods, using the methodology presented in Section 3.4.4. The cask cavity under normal transport conditions is backfilled to one atmosphere with 99.9% pure helium gas. To determine the limiting temperature conditions, it has been assumed that the helium gas could possibly be replaced by air. Therefore, the normal operating pressure is determined for both gas conditions. From Section 3.4.4, the free gas volume, fuel fill gas volume, and fuel fission gas volumes for the two spent fuel configurations are presented below. The GTCC waste and HLW do not release any gas. The Reconfigured Fuel Assemblies and Damaged Fuel Cans contain failed fuel. The initial charge gas and any significant fission product gases have already been released from the Reconfigured Fuel Assemblies and from fuel in the Damaged Fuel Cans.
Regulatory Guide 1.25 suggests that 10% of the tritium and 30% of the krypton-85 should be assumed to be released from each failed fuel rod. It is conservatively assumed that 30% of both tritium and krypton-85 escape each failed fuel rod. Other radiologically important gaseous nuclides are present only in negligible amounts after the minimum cooling period for the design basis directly loaded and canistered fuels. The postulated release of other radionuclides,
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-4 including volatiles, fines, particulates and crud, does not contribute to an increase in internal pressure.
4.2.2.1 Containment Pressurization Due to Directly Loaded Fuel An increase in pressure within the containment boundary results from an increase in the cask cavity temperature and the postulated failure of 100 percent of the fuel rods in normal conditions of transport (MNOP).
The pressure with air in the cask cavity, based on a conservative bulk air temperature of 450°F (Section 3.4.4), is 3.87 atm (56.9 psia = 42.2 psig). For standard burnup fuel (45 GWd/MTU assembly average) based on a bulk average gas temperature of 401ºF when helium is the cover gas, the pressure in the cask cavity is:
P2= (3.87)(478/505) = 3.66 atm = 53.8 psia = 39.1 psig This is less than the containment boundary design pressure of 75 psig.
High burnup fuel as content was determined in Section 3.8.6 to produce a slightly higher average gas temperature under normal conditions of 438ºF. This is still bounded by the 450ºF temperature used in the Section 3.4.4 pressure evaluations. Normal and hypothetical accident condition pressures are addressed in Section 3.8.6 for high burnup fuel.
To address flammability concerns within the context of PWR fuel transport, the hydrogen (in particular tritium) content was extracted from a high burnup (60 GWd/MTU assembly average)
SAS2H output. Tritium quantity generated was determined to be larger for higher burnup and lower enrichment. A conservative 1.7 wt%, 60 GWd/MTU, was applied and yielded 0.0429 grams of tritium per assembly at reactor discharge (lower quantity at transport due to decay to helium). For 26 assemblies (conservative) tritium quantity is thereby < 1.2 grams (0.4 mole). At standard conditions, 1 mole of gas occupies ~22.4 liters. As the cavity volume of the system is documented as greater than 7000 liters (see Table 4.2-4 and Section 3.8.6), no flammability hazard exists (volume fraction <0.1%).
4.2.2.2 Canister and Cask Pressurization Due to Yankee Class Fuel The maximum normal operating pressure (MNOP) during transport conditions in the transportable storage canister is calculated in Section 3.4.4, and found to be 3.23 atmospheres, or
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-5 32.8 psig. This pressure is conservatively calculated at 450F, compared to the calculated maximum normal conditions of transport bulk gas temperature of 442F, and conservatively assumes the rupture of 100% of the fuel rods. The MNOP is below the design pressure of 55 psig. The GTCC waste, Damaged Fuel can contents and Reconfigured Fuel Assemblies classified as failed do not release gases to the canister cavity due to rupture of fuel rods.
Consequently, there is no increase in canister internal pressure due to these contents.
Since the canister does not fail in any of the evaluated normal transport or accident conditions, this pressure increase occurs within the canister. There is no pressure increase in the cask cavity except that due to the increase in cavity temperature. As the cask cavity is backfilled to 0 psig, a hypothetical canister failure would result in a containment vessel pressure lower than the canister pressure. Hypothetical canister failure would, therefore, result in significantly lower system pressure than the containment boundary design pressure.
4.2.2.3 Canister and Cask Pressurization Due to Connecticut Yankee Fuel The MNOP during transport conditions in the transportable storage canister is calculated in Section 3.4.4 and found to be 3.9 atmospheres, or 42.3 psig. This pressure is conservatively calculated at 450F, compared to the calculated maximum normal conditions of transport bulk gas temperature of 402F, and conservatively assumes the rupture of 100% of the fuel rods. The MNOP is below the design pressure of 55 psig. As described above, the GTCC waste, Damaged Fuel Can contents and Reconfigured Fuel Assemblies do not release gases to the canister cavity due to failures. Consequently, there is no increase in canister internal pressure due to these contents.
Since the canister does not fail in any of the evaluated normal transport or accident conditions, this pressure increase occurs within the canister. There is no pressure increase in the cask cavity except that due to the increase in cavity temperature. As the cask cavity is backfilled to 0 psig, a hypothetical canister failure would result in a containment vessel pressure lower than the canister pressure. Hypothetical canister failure would, therefore, result in significantly lower system pressure than the containment boundary design pressure.
4.2.3 Containment Criterion for Normal Conditions of Transport The NAC-STC is designed and fabrication leakage rate tested to meet the containment criteria of 10 CFR 71.51. The 10 CFR 71 limit for the release of radioactive material under normal
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-6 conditions of transport is 10-6 A2 per hour. The containment criteria are met for the inner metallic seals (outer seals provided by metallic seal or Viton O-rings) configuration by testing the NAC-STC to leaktight conditions, as defined by ANSI N14.5-1997, of 110-7 ref cm3/sec.
This corresponds to a helium leakage rate of 210-7 cm3/sec and a test sensitivity of 110-7 cm3/sec (helium).
The containment criteria are met for the inner Viton O-ring configuration by testing the NAC-STC containment boundary to either a leaktight condition (110-7 cm3/sec), which is equivalent to 2.010-7 cm3/sec (helium) at a test sensitivity of 1.010-7 cm3/sec (helium), which may be applied for the directly loaded shipment of HBU PWR fuel assemblies, or to the cumulative directly loaded leakage rates in Table 4.2-3 for LBU or HBU fuel, or to no detected leakage at a test sensitivity of 1 x 10-3 ref cm3/sec (air) for directly loaded standard and HBU PWR fuel assemblies. The calculation of the allowable fabrication and maintenance leakage rate for the Viton O-ring configuration for standard PWR fuel assemblies is provided in Section 4.2.3.2. The allowable fabrication and maintenance leakage rate test acceptance criteria for all containment boundary conditions (A, B1, B2 and B3 from Table 4.1-1) are detailed in Chapter 8.
4.2.3.1 Permissible Release Rate for the NAC-STC with Metallic O-rings Metallic O-rings are required to be used in the containment boundary when the cask is directly loaded with PWR spent fuel for long-term storage with subsequent transport, and when the cask is loaded with a transportable storage canister or HLW overpack. Metallic O-rings may also be used for the containment boundary seals when the cask is directly loaded with standard (e.g.,
burnup of 45 GWd/MTU) or HBU (e.g., burnup of > 45 GWd/MTU) PWR fuel assemblies (e.g., burnup of 45 GWd/MTU) for transport without interim storage. For the metallic O-ring configurations, either inner and outer metallic seals or inner metallic and outer Viton O-ring seals, the containment boundary closures are tested to leaktight conditions as defined by ANSI N14.5-1997 and, therefore, meets the requirements of 10 CFR 71.51 for containment of the radioactive contents.
Since the cask containment boundary is tested to demonstrate a leaktight condition, an allowable release rate, based on gases, fines, volatiles and particulates that are available for release from the directly loaded PWR spent fuel, spent fuel or GTCC waste in the transportable storage canister, or HLW overpack is not calculated.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-7 4.2.3.2 Permissible Release Rate for the NAC-STC with Viton O-rings Viton O-rings may be used as the containment boundary seals when the cask is directly loaded with PWR spent fuel for transport without interim storage. For the Viton O-ring configurations, the containment fabrication and maintenance leakage rates are summarized in Table 4.2-3. A leakage rate of 1.010-7 ref cm3/sec (e.g., leaktight) may be applied for high burnup (HBU) fuel assemblies with burnup > 45 GWd/MTU. As described in this section, these leak rates meet the requirements of 10 CFR 71.51 for containment of directly loaded PWR spent fuel.
The 10 CFR 71.51 limit for the release of radioactive material under normal conditions of transport is 10-6 A2/hr. In this analysis, A2 for a mixed gas is determined by using the method described in 10 CFR 71, Appendix A. The release fractions for the various radionuclides transported in the NAC-STC are obtained from NUREG/CR-6487 and summarized in Table 4.2-1. The curie content for gases, volatiles, fines and particulates for the directly loaded 5-year cooled PWR reference fuel assembly is provided in Tables 4.2-6 through 4.2-8.
In addition to the radionuclides produced by the fuel material, fuel assemblies develop a coating of impurities deposited by cooling water during power generation. This coating is known as crud. Crud contains mostly non-radioactive elements but also contains a significant amount of 60Co.
NUREG/CR-6487 lists the maximum 60Co concentrations on spent fuel assemblies to be 140 Ci/cm2 for PWR assemblies at initial discharge. The surface area of the reference 1717 PWR assembly is calculated to be 3.54105 cm2, based on the assembly characteristics provided in Table 5.2-2.
The maximum permissible leak rate from the cask under normal conditions of transport is determined from the 10 CFR 71 limit of 10-6 A2/hr.
1 10 2
N N
N 1
6 2
N N
N sec 10 x
2.78 x
A C
L R
or hr 10 x
1 x
A C
L R
where:
LN
= Volumetric gas leakage rate [cm3/s]
= Curies per unit volume (termed activity density) of the radioactive material that passes through the leak path [Ci/cm3]
RN
= Release rate for normal transport conditions [Ci/sec]
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-8 Activity Density of Radioactive Material (CN)
The total inventories of fission product gases, volatiles and fines are shown in Table 4.2-6 through Table 4.2-8. These inventories are calculated by using the source terms produced by the SAS2H sequence, the release fractions and the postulated crud (60Co). The 60Co content is decayed 5 years from discharge.
Fines FissionGas Volaties Crud n
C C
C C
C
V
)
S (N
N S
f V
M f
C AR R
A C
C T
C Crud
where:
Ccrud
= Activity density inside containment vessel resulting from crud spallation [Ci/cm3]
=
Total crud activity inventory [Ci]
fc
=
Crud spallation factor V
=
Free volume inside containment vessel [cm3]
SC
=
Crud surface activity [Ci/cm2]
NR
=
Number of fuel rods per assembly NA
=
Number of assemblies SAR
=
Surface area per rod [cm2]
and, V
f N
N A
W f
C B
A R
R R
F fines where:
Cfine
=
Activity concentration inside containment vessel resulting from fines released from cladding breaches [Ci/cm3]
fF
=
Fraction of fuel rod mass released as fines resulting from cladding breach fB
=
Fraction of fuel rods that develop cladding breach WR
=
Mass of the fuel in fuel rod [g]
NR
=
Number of fuel rods per assembly NA
=
Number of assemblies AR
=
Specific activity of fines emitted from cladding breach in fuel rod [Ci/g]
V
=
Containment vessel void volume [cm3]
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-9
- and, V
)
f A
f (A
W f
N N
C C
C G
G V
V R
B A
R gas vol vg
where:
Cvg
=
Releasable activity concentration inside the containment vessel resulting from gases and volatiles released from cladding breaches [Ci/cm3]
Cvol
=
Releasable activity concentration inside the containment vessel resulting from volatiles released from cladding breaches [Ci/cm3]
Cgas
=
Releasable activity concentration inside the containment vessel resulting from gases released from cladding breaches [Ci/cm3]
=
Mass of the fuel in a fuel rod [g]
NR
=
Number fuel rods per assembly NA
=
Number of assemblies fB
=
Fraction of rods that develop cladding breaches AV
=
Specific activity of volatiles in fuel rod [Ci/g]
fV
=
Fraction of volatiles in fuel rod released if rod develops cladding breach AG
=
Specific activity of gas in fuel rod [Ci/g]
fG
=
Fraction of gas that would escape from fuel rod that develops cladding breach V
=
Void volume inside containment vessel [cm3]
Activity Values for Radionuclides A2 values used in this analysis (based on 10 CFR 71 Appendix A) are listed in Tables 4.2-6 through 4.2-8 for all radionuclides produced by the SAS2H analysis (plus 60Co). The mixture A2 value is shown in Table 4.2-2. For those isotopes for which no specific A2 values are given in 10 CFR 71 Appendix A, the generic values listed in Table A.2 of Appendix A are applied. A2 values for mixed isotopes are calculated from the following:
A i
2 =
1 F
A i
2
where:
F = S S
i i
n and
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-10
[Ci]
activity group Total
=
S
[Ci]
i isotope of Activity
=
S mixture entire the respect to with i
isotope of Fraction
=
F n
i i
Mixture A2 values are determined for gas, volatile, fine and crud mixtures and are then combined for a total cask mixture A2 value. Table 4.2-2 provides the source term and A2 values per group for directly loaded PWR fuel release rate calculations.
Maximum Allowable Leak Rate for Viton O-rings On the basis of the methodology described, the maximum allowable volumetric leak rate for PWR fuel directly loaded for immediate transport in normal conditions of transport is summarized in Table 4.2-3.
Correlation of Allowable Leak Rates to Air Standard The volumetric gas leak rate, L, is independent of cask pressure and temperature. The maximum allowable release must be correlated with air standard leak rates, which depend on gas temperatures, pressures, and leakage path length and diameter. This correlation requires calculation of the capillary opening diameter through which the flow occurs. Depending on pressure and condition of the flow, a combination of continuum and molecular flow occurs.
Continuum flow and molecular flow equations are obtained from NUREG/CR-6487, Section 2, which are adjusted to upstream flow rate in accordance with NUREG/CR-6487 and ANSI N14.5-1997. The continuum volumetric flow rate of the gas (cm3/sec), Lc, is given by:
u a
d u
c u
a d
u 4
6 c
P P
)
P (P
F P
P
)
P (P
a D
2.48x10 L
where:
Fc
=
Coefficient for continuum flow [cm3/atm-s]
D
=
Capillary diameter [cm]
a
=
Capillary length [cm]
=
Fluid viscosity [cP]
Pu
=
Upstream pressure [atm] - pressure inside containment Pd
=
Downstream pressure [atm] - pressure outside containment Pa
=
Average pressure (Pu+Pd)/2 [atm]
and, the molecular volumetric flow rate of the gas (cm3/sec), Lm, is given by:
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-11 u
a d
u m
u a
d u
a 3
3 m
P P
)
P (P
F P
P
)
P (P
aP M
T D
3.81x10 L
where:
Lm
=
Volumetric flow rate of gas at Pa [cm3/sec]
Fm
=
Coefficient for molecular flow [cm3/atm-s]
D
=
Capillary diameter [cm]
T
=
Gas temperature [K]
M
=
Gas molecular weight [g/mole]
Pa
=
Average pressure (Pu+Pd)/2 [atm]
Pu
=
Upstream pressure [atm]
Pd
=
Downstream pressure [atm].
a
=
Capillary diameter [cm]
For this analysis, the gas temperature used for molecular flow analysis is identical to the upstream temperature. Pressure and temperature at normal operating conditions are summarized in Table 4.2-4. Based on the pressure, temperature and allowable leakage rate (LN), the capillary diameter of the leak is determined. The calculated capillary diameter is then used to determine the air standard leak rate and helium test leak rate. Air standard condition leak rates are determined for air leaking from 1 atmosphere to 0.01 atmosphere at a temperature of 298K. The test gas is helium leaking from 1 atmosphere (0 psig) to a vacuum. Table 4.2-3 provides the reference and test leak rates. The minimum sensitivity for these tests is one-half the air reference leak rate. Key containment analysis parameters are summarized in Table 4.2-5.
This analysis is conservative, since a higher upstream pressure, which could result from a higher average gas temperature based on decay heat, results in a higher allowable leak rate assuming that the leak path length and the leak path diameter (calculated based on the reference air condition) are held constant. Since the test condition pressure cannot be less than 1 atmosphere and the average gas temperature does not have a first order effect on calculated leak rate, the helium test condition is conservative with respect to the allowable reference leak rate.
4.2.3.3 Permissible Release Rate for Canistered Fuel and GTCC Waste, and HLW Overpacks The transportable storage canister welded closure is leak tested at final assembly to leaktight conditions, 1 x 10-7 ref cm3/ sec, as defined by ANSI N14.5-1997. To meet this requirement, the allowable leak rate is 2 x 10-7 cm3/sec (helium). The leak test sensitivity applied in testing the canister at the time it is closed is 1 x 10-7 cm3/sec (helium), or less. Consequently, the canister
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-12 provides adequate containment for the spent fuel or GTCC waste. For the Yankee-MPC configuration, the allowable leak rate for the canister is specified as 8 x 10-8 cm3/sec (helium),
with a corresponding test sensitivity of 4 x 10-8 cm3/sec (helium). This specified test condition is conservative with respect to the leaktight condition specified by ANSI N14.5-1997.
HLW overpacks contain up to five (5) canisters of glassified HLW, and the closure lid weld is not leakage tested prior to storage. The glassified HLW welded canister contents will not generate releasable radioactive material, and therefore, no confinement credit is required or applied to HLW overpack. The containment of the HLW overpack contents is provided by the containment condition B1 of Table 4.1-1, which provides a leaktight containment boundary provided inner and outer with metallic seals.
Correlation to Air Standard Conditions The air standard leak rate is 1 x 10-7 ref cm3/ sec, the leak tight condition as defined by Section 2.1 of ANSI N14.5-1997. Leak testing of the NAC-STC cask and the transportable storage canister is performed using helium gas. The NAC-STC cask leak test is performed using an allowable leak rate of 2 x 10-7 cm3/sec (helium) with a detection sensitivity of 1 x 10-7 cm3/sec (helium).
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-13 Table 4.2-1 Release Fractions: Normal and Accident Conditions LBU (NUREG-2216)
HBU (NUREG-2224)
Radionuclide Origin Fraction:
Normal Conditions Fraction:
Accident Conditions Fraction:
Normal Conditions Fraction:
Accident Conditions Volatiles releasable 2.00E-04 2.00E-04 3.00E-05 3.00E-05 Fission gas releasable 0.3 0.3 0.15 0.35 Rod mass released 3.00E-05 3.00E-05 3.00E-05 3.00E-03 Crud spallation factor 0.15 1.0 0.15 1.0 Fraction of fuel that fails 0.03 1.0 0.03 1.0 Table 4.2-2 Allowable Release Rate Source and A2 Values for Directly Loaded PWR Fuel:
Normal Conditions Reference 1717 PWR Fuel1,2 Crud Gas Volatiles Fines Total Total Activity per Assembly (TBq)
N/C3 1.33E+02 4.92E+03 8.15E+03 1.32E+04 Releasable Activity per Cask (TBq) 2.85E+00 2.79E+01 8.85E-02 1.47E+01 4.55E+01 Cask Volumetric Activity (TBq /cm3) 3.96E-07 3.88E-06 1.23E-08 2.04E-06 6.33E-06 A2 Value (TBq) 0.40 10.52 0.41 0.02 11.35 Fraction of Activity 0.063 0.613 0.002 0.322 1.000 Fraction of Activity / A2 (1/ TBq) 0.1565 0.0583 0.0047 16.0229 16.2425 Mixture A2 Value (TBq) 0.06
- 1. Shown for HBU fuel.
- 2. Based on 3% rod failure.
- 3. Not explicitly calculated.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-14 Table 4.2-3 Leak Rate and Leak Test Sensitivity: Normal Conditions Leak Rate (cm3/sec)
Contents O-rings Vol. Activity (TBq/cm3)
Allowable (L)
Air Reference (LR)
Test Sensitivity LBU Viton 4.8E-06 8.8E-05 9.4E-051 4.7E-051 HBU Viton 6.3E-06 2.7E-06 2.6E-062 1.3E-062
- 1. The corresponding helium test leak rates and leak test sensitivities for directly loaded LBU fuel are 1.110-4 cm3/sec and 5.710-5 cm3/sec, respectively, at standard conditions.
- 2. The corresponding helium test leak rates and leak test sensitivities for directly loaded LBU fuel are 4.010-6 cm3/sec and 2.010-6 cm3/sec, respectively, at standard conditions.
Table 4.2-4 Cask Free Volumes and Pressures: Normal and Accident Conditions Contents LBU, 26 Assemblies HBU, 20 Assemblies Cask Operating Condition Normal Accident1 Normal Accident1 Free Gas Volume (liters)2 7440 7540 7200 7280 Pressure (atm)2 1.79 5.18 1.76 5.33 Average Gas Temperature (K) 505.0 675.0 505.0 675.0
- 1. The accident condition for this analysis is 100% rod failure in combination with a fire accident that raises the cask temperature. This hypothetical dual-failure accident conservatively maximizes both available releasable material and cask pressure.
- 2. Bounding values were chosen for free volume (minimum) and pressure (maximum). This conservatively minimizes free volume and capillary diameter.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-15 Table 4.2-5 Containment Parameters for Nonmetallic O-rings in Normal Conditions of Transport Contents Containment Free Volume (cm3)
Capillary Length (cm)
Capillary Diameter (cm)
Upstream Pressure (atm)
Gas Temperature (K)
LBU 7.44E+6 0.597 8.6E-4 1.79 505 HBU 7.20E+6 1.76 Note: Based on 3% of the fuel rods failing in normal transport conditions.
Table 4.2-6 1717 Reference Fuel SAS2H Output and Group A2 Values - Gases Activity/Assembly Fraction of Source Isotope A2 Value Isotope Fraction/A2 Isotope (TBq)
(TBq)
(1/TBq)
Group A2 H3 1.12E+01 6.24E-02 4E+01 1.559E-03 I129 1.01E-03 5.64E-06 1E+60 5.638E-66 KR85 1.68E+02 9.38E-01 1E+01 9.376E-02 Total 1.79E+02 9.53E-02 10.491 Note: Shown for HBU fuel. For LBU fuel, the combined gas activity and group A2 are 1.33E+02 TBq and 10.518, respectively.
Table 4.2-7 1717 Reference Fuel SAS2H Output and Group A2 Values - Volatiles Activity/Assembly Fraction of Source Isotope A2 Value Isotope Fraction/A2 Isotope (TBq)
(TBq)
(1/TBq)
Group A2 CS134 1.23E+03 1.89E-01 7E-01 2.695E-01 CS135 1.24E-02 1.90E-06 1E+00 1.903E-06 CS137 2.95E+03 4.52E-01 6E-01 7.538E-01 RU106 5.59E+02 8.58E-02 2E-01 4.290E-01 SR90 1.78E+03 2.73E-01 3E-01 9.110E-01 Total 6.51E+03 2.36E+00 0.423 Note: Shown for HBU fuel. For LBU fuel, the combined volatile activity and group A2 are 4.92E+03 TBq and 0.414, respectively.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-16 Table 4.2-8 1717 Reference Fuel SAS2H Output and Group A2 Values - Fines Activity/Assembly Fraction of Source Isotope A2 Value Isotope Fraction/A2 Isotope (TBq)
(TBq)
(1/TBq)
Group A2 AG108 4.40E-05 4.41E-09 2E-02 2.20E-07 AG108M 5.03E-04 5.04E-08 7E-01 7.20E-08 AG109M 4.51E-06 4.52E-10 2E-02 2.26E-08 AG110 1.78E-02 1.78E-06 2E-02 8.89E-05 AG110M 1.31E+00 1.31E-04 4E-01 3.27E-04 AM241 2.58E+01 2.58E-03 1E-03 2.58E+00 AM242 2.47E-01 2.47E-05 2E-02 1.24E-03 AM242M 2.48E-01 2.48E-05 1E-03 2.48E-02 AM243 1.55E+00 1.55E-04 1E-03 1.55E-01 BA137M 2.78E+03 2.78E-01 2E-02 1.39E+01 BI212 1.85E-03 1.86E-07 6E-01 3.09E-07 BK249 2.18E-04 2.18E-08 3E-01 7.26E-08 C 14 3.63E-06 3.63E-10 3E+00 1.21E-10 CD109 4.51E-06 4.52E-10 2E+00 2.26E-10 CD113M 1.02E+00 1.02E-04 5E-01 2.04E-04 CE144 2.61E+02 2.61E-02 2E-01 1.31E-01 CF249 3.12E-05 3.12E-09 8E-04 3.90E-06 CF250 1.25E-04 1.25E-08 2E-03 6.26E-06 CF251 1.55E-06 1.55E-10 7E-04 2.21E-07 CF252 1.92E-04 1.92E-08 3E-03 6.41E-06 CM242 1.01E+00 1.01E-04 1E-02 1.01E-02 CM243 1.18E+00 1.19E-04 1E-03 1.19E-01 CM244 2.87E+02 2.88E-02 2E-03 1.44E+01 CM245 3.52E-02 3.52E-06 9E-04 3.91E-03 CM246 2.19E-02 2.20E-06 9E-04 2.44E-03 CM248 1.46E-06 1.46E-10 3E-04 4.87E-07 EU152 8.73E-02 8.74E-06 1E+00 8.74E-06 EU154 3.02E+02 3.02E-02 6E-01 5.03E-02 EU155 1.44E+02 1.44E-02 3E+00 4.79E-03 GD153 4.11E-03 4.11E-07 9E+00 4.57E-08 HO166M 6.22E-05 6.22E-09 5E-01 1.24E-08 NB 93M 9.07E-03 9.08E-07 3E+01 3.03E-08 NB 94 2.98E-06 2.99E-10 7E-01 4.27E-10 NB 95 1.59E-04 1.59E-08 1E+00 1.59E-08 NP235 7.03E-05 7.04E-09 4E+01 1.76E-10 NP236 2.08E-06 2.08E-10 2E-02 1.04E-08 NP237 1.04E-02 1.04E-06 2E-03 5.22E-04 NP238 1.11E-03 1.12E-07 2E-02 5.58E-06 NP239 1.55E+00 1.55E-04 4E-01 3.87E-04
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.2-18 Table 4.2-8 1717 Reference Fuel SAS2H Output and Group A2 Values - Fines (Continued)
Activity/Assembly Fraction of Source Isotope A2 Value Isotope Fraction/A2 Isotope (TBq)
(TBq)
(1/TBq)
Group A2 TH231 8.58E-05 8.59E-09 2E-02 4.30E-07 TH234 5.29E-03 5.30E-07 3E-01 1.77E-06 TL208 6.66E-04 6.67E-08 2E-02 3.33E-06 TM171 1.90E-05 1.90E-09 4E+01 4.75E-11 U232 2.70E-03 2.70E-07 1E-03 2.70E-04 U233 3.63E-07 3.64E-11 6E-03 6.06E-09 U234 2.56E-03 2.56E-07 6E-03 4.27E-05 U235 8.58E-05 8.59E-09 1E+60 8.59E-69 U236 5.37E-03 5.37E-07 6E-03 8.95E-05 U237 6.03E-02 6.04E-06 2E-02 3.02E-04 U238 5.29E-03 5.30E-07 1E+60 5.30E-67 Y 90 1.78E+03 1.78E-01 3E-01 5.94E-01 Y 91 6.62E-06 6.63E-10 6E-01 1.11E-09 ZR 93 3.33E-02 3.33E-06 1E+60 3.33E-66 ZR 95 6.99E-05 7.00E-09 8E-01 8.75E-09 Total 9.99E+03 5.64E+01 0.018 Note: Shown for HBU fuel. For LBU fuel, the combined fines activity and group A2 are 8.15E+03 TBq and 0.020, respectively.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.3-1 4.3 Containment Requirements for Hypothetical Accident Conditions The NAC-STC has been designed to safely transport 26 design basis directly loaded PWR fuel assemblies, up to 20 HBU PWR fuel assemblies, or canistered spent fuel or GTCC waste in either the Yankee-MPC or Connecticut Yankee-MPC configurations, and HLW overpacks. The structural integrity of the cask containment during hypothetical accident conditions is demonstrated in Section 2.7. Therefore, the cask containment is maintained under hypothetical accident conditions. As described in Section 2.7.11, the transportable storage canister does not fail in any of the evaluated transport accident conditions defined in 10 CFR 71.73.
Consequently, its leaktight condition is maintained in the hypothetical accident conditions. As described in Section 4.1, metallic O-rings are required to be used for the direct loading of fuel for long-term storage and subsequent transport and for the transport of transportable storage canisters and HLW overpacks. Either metallic or Viton O-rings may be used for directly loaded standard burnup PWR fuel assemblies ( 45 GWd/MTU) for transport without interim storage or for directly loaded HBU PWR fuel assemblies (> 45 GWd/MTU) for transport without interim storage.
For direct loading for transport without interim storage using Viton O-rings or metallic seals, the containment boundary requirement under hypothetical accident conditions is met by ensuring that the leak rate limits in Table 4.3-2 are not exceeded. High burnup PWR fuel may employ a leaktight limit of 110-7 refcm3/sec. Calculations related to the non-leaktight limit is provided in Section 4.3.3.
For directly loaded fuel, assuming a simultaneous occurrence of a fire accident and a 100% rod failure, and on the basis of a bulk average gas temperature of 675K resulting from air in the cavity, the pressure within the cask cavity is calculated to be 5.18 and 5.33 atm for LBU and HBU fuel, respectively. The hypothetical presence of air in the cask provides an upper bound on the gas temperature. This pressure represents the maximum possible cask internal pressure.
While fuel could fail within the transportable storage canister, no release from the canister occurs.
4.3.1 Fission Gas Products The calculated amounts of fission gases contained by the design basis directly loaded and canistered PWR fuel assemblies for both normal and hypothetical accident conditions are reported in Section 4.2.2. The accident conditions assume a 100% fuel rod failure with 30% of the available tritium and 30% of the available krypton-85 being released to the cask cavity or to
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.3-2 the canister. These gases contribute to an increase in the cask cavity pressure due to the postulated failure of the directly loaded, intact/undamaged fuel and to an increase in the canister pressure due to the postulated failure of the canistered fuel.
Other released radionuclides, including crud, volatiles, fines and particulates, are not assumed to contribute to an increase in internal pressure of either transport configuration. The GTCC waste does not contain any gaseous products and does not have a failure mode in the hypothetical accident conditions. Consequently, there is no increase in pressure due to the GTCC contents.
The release of material from the postulated failure of the intact/undamaged fuel assemblies bounds the possible release of material from the Reconfigured Fuel Assemblies and Damaged Fuel Cans since the allowable contents of these components is less than or equal to that of an intact fuel assembly.
4.3.2 Containment of Radioactive Material For directly loaded fuel intended for transport without interim storage using metallic O-rings, the containment boundary is tested to a leaktight condition as defined in ANSI N14.5-1997. As shown in Section 2.7 for the NAC-STC cask and in Section 2.7.11 for the transportable storage canister, the containment boundary of the cask and canister do not fail during the hypothetical accident events. Consequently, leaktight containment is maintained by both the cask and the canister in the hypothetical accident events. For the Viton O-ring configuration for direct loading, the containment criteria may be either leaktight (for high burnup fuel) or the allowable leak rate in the hypothetical accident condition is as shown in Section 4.3.3. The radionuclide activities for the reference PWR fuel assembly are provided in Section 4.2.3.
4.3.3 Calculation of Allowable Leak Rate for Non Leaktight Directly Loaded Fuel with Viton O-rings The allowable leak rates under hypothetical accident conditions for the non leaktight configuration are calculated by using the method described in Section 4.2.1.1 for normal conditions of transport. The total inventories of fission product gases, volatiles, fines and crud are calculated by using the source terms generated by SAS2H, using the release fractions. Using the A2 values from 10 CFR 71, Appendix A (Table 4.3-1), the mixture A2 values are determined for gas, volatile, fine and crud mixtures. Finally, the maximum allowable release rates are calculated by using the hypothetical accident conditions allowable release limit:
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 4.3-4 Table 4.3-1 Allowable Release Rate Source and A2 Values for Directly Loaded PWR Fuel: Accident Conditions Using Nonmetallic O-rings 1717 Reference1 Crud Gas Volatiles Fines Total Total Activity per Assembly (TBq)
N/C2 1.33E+02 4.92E+03 8.15E+03 1.32E+04 Releasable Activity per Cask (TBq) 1.90E+01 9.31E+02 2.95E+00 4.89E+02 1.44E+03 Cask Volumetric Activity (TBq/cm3) 2.61E-06 1.28E-04 4.05E-07 6.72E-05 1.98E-04 A2 Value (TBq) 0.40 10.52 0.41 0.02 11.35 Fraction of Activity 0.013 0.646 0.002 0.339 1.000 Fraction of Activity / A2 (1/TBq) 0.0330 0.0614 0.0049 16.8679 16.9672 Mixture A2 Value (TBq) 0.06
- 1. Shown for HBU fuel.
- 2. Not explicitly calculated.
Table 4.3-2 Standard Leak Rate for the Accident Condition Contents O-rings Vol. Activity (TBq/cm3)
Leak Rate (cm3/sec)
Allowable (L)
Air Reference (LR)
LBU Fuel Viton 1.5E-04 2.3E-02 8.6E-03 HBU Fuel Viton 2.0E-04 4.9E-04 1.7E-04 Table 4.3-3 Containment Parameters for Non-Metallic O-rings in the Accident Condition Contents Containment Free Volume (cm3)
Capillary Length (cm)
Capillary Diameter (cm)
Upstream Pressure (atm)
Gas Temperature (K)
LBU Fuel 7.54E+6 0.597 2.6E-3 5.72 675 HBU Fuel 7.28E+6 5.33 Note: 100 % of the fuel rods are postulated to fail in the accident condition.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-iv Table of Contents (continued) 5.12 Evaluation for Uncertainty in NS-4-FR Hydorgen Measurement................................... 5.12-1 5.12.1 High Burnup Configuration.................................................................................. 5.12-1 5.12.2 Low Burnup Configuration.................................................................................. 5.12-2 5.12.3 Canistered Configuration...................................................................................... 5.12-3
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-v List of Figures Figure 5.1-1 Detector Locations for Yankee Class Canistered Fuel and GTCC Waste...... 5.1-10 Figure 5.1-2 Maximum Dose Rate Locations for the Three-Dimensional Directly Loaded Fuel Analysis in Normal Conditions.................................................. 5.1-11 Figure 5.1-3 Design Basis Yankee Class Combustion Engineering Fuel Assembly.......... 5.1-12 Figure 5.1-4 Yankee GTCC Waste Container..................................................................... 5.1-13 Figure 5.1-5 Connecticut Yankee Design Basis Fuel Assembly Source Regions and Elevations........................................................................................................ 5.1-14 Figure 5.1-6 Location of Maximum Dose Rates for CY-MPC Fuel and GTCC Waste in Normal Conditions of Transport................................................................. 5.1-15 Figure 5.1-7 Location of Maximum Dose Rates for CY-MPC Fuel and GTCC Waste in Accident Conditions................................................................................... 5.1-16 Figure 5.2-1 Directly Loaded Fuel Design Basis Burnup Profile....................................... 5.2-14 Figure 5.2-2 Directly Loaded Fuel Neutron and Gamma Source Profiles.......................... 5.2-15 Figure 5.2-3 Design Basis Yankee Class Fuel Burnup Profile............................................ 5.2-16 Figure 5.2-4 Yankee GTCC Waste Container Gamma Source Profile Based on Dose Rate Measurements......................................................................................... 5.2-17 Figure 5.2-5 Connecticut Yankee Design Basis Fuel Neutron and Gamma Burnup Profiles............................................................................................................ 5.2-18 Figure 5.3-1 Three-Dimensional MCBEND Model for Directly Loaded Fuel - Normal Conditions - Axial Detail............................................................................... 5.3-11 Figure 5.3-2 Three-Dimensional MCBEND Model for Directly Loaded Fuel - Accident Conditions - Axial Detail............................................................................... 5.3-12 Figure 5.3-3 Three-Dimensional MCBEND Model for Directly Loaded Fuel - Radial Detail............................................................................................................... 5.3-13 Figure 5.3-4 One-Dimensional Radial Shielding Model with Canistered Yankee Class Fuel........................................................................................................ 5.3-14 Figure 5.3-5 One-Dimensional Axial Shielding Model with Canistered Yankee Class Fuel........................................................................................................ 5.3-15 Figure 5.3-6 One-Dimensional Top Axial Model with Canistered Yankee Class Fuel...... 5.3-16 Figure 5.3-7 One-Dimensional Radial Shielding Model with Canistered Yankee GTCC Waste................................................................................................... 5.3-17 Figure 5.3-8 One-Dimensional Bottom Axial Model with Canistered Yankee GTCC Waste................................................................................................... 5.3-18
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-vi List of Figures (continued)
Figure 5.3-9 One-Dimensional Top Axial Model with Canistered Yankee GTCC Waste................................................................................................... 5.3-19 Figure 5.3-10 CY-MPC Three-Dimensional Canister Model Detail..................................... 5.3-20 Figure 5.3-11 Three-Dimensional NAC-STC Model for CY-MPC Analysis....................... 5.3-21 Figure 5.3-12 Three-Dimensional Model of CY-MPC GTCC Waste Basket....................... 5.3-22 Figure 5.4-1 Radial Dose Rate Profiles for Directly Loaded Fuel in Normal Conditions of Transport.................................................................................... 5.4-8 Figure 5.4-2 Radial Dose Rate Profile by Source Type at 2 meters from the Railcar for Directly Loaded Fuel in Normal Conditions of Transport................................ 5.4-9 Figure 5.4-3 Azimuthal Radial Surface Dose Rate Profile by Source Type at Rotation Trunnion Elevation for Directly Loaded Fuel in Normal Conditions of Transport..................................................................................................... 5.4-10 Figure 5.4-4 Azimuthal Radial Surface Dose Rate Profile by Source Type over Heat Fin Axial Extent for Directly Loaded Fuel in Normal Conditions of Transport......................................................................................................... 5.4-11 Figure 5.4-5 Radial Dose Rate Profile by Source Type at 1 meter for Directly Loaded Fuel in the Accident Condition....................................................................... 5.4-12 Figure 5.4-6 Azimuthal Radial Dose Rate Profile at 1 meter for Directly Loaded Fuel in the Accident Condition............................................................................... 5.4-13 Figure 5.4-7 Graphical Comparison of Normal Conditions Radial 2m+Railcar Dose Rate Profile for DRM and Direct Solution - 1414 Assembly at 40,000 MWd/MTU, 3.7 wt % 235U, 7 Years Cool Time............................................ 5.4-14 Figure 5.4-8 Graphical Comparison of Accident Conditions Radial 1m Dose Rate Profile for DRM and Direct Solution - 1515 Assembly at 40,000 MWd/MTU, 3.7 wt % 235U, 7 Years Cool Time............................................ 5.4-15 Figure 5.4-9 NAC-STC Radial Dose Rate Profile - Normal Conditions - Design Basis Connecticut Yankee Stainless Steel Clad Fuel............................................... 5.4-16 Figure 5.4-10 NAC-STC Radial Dose Rate Profile - Normal Conditions - Design Basis Connecticut Yankee Zircaloy Clad Fuel......................................................... 5.4-17
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-vii List of Figures (continued)
Figure 5.4-11 NAC-STC CY-MPC Azimuthal Heat Fin Dose Rate Variations - Normal Conditions - Design Basis Stainless Steel Clad Fuel..................................... 5.4-18 Figure 5.4-12 NAC-STC CY-MPC Azimuthal Heat Fin Dose Rate Variations - Normal Conditions - Design Basis Zircaloy Clad Fuel............................................... 5.4-19 Figure 5.4-13 NAC-STC CY-MPC Radial Dose Rate Profile - Accident Conditions -
Design Basis Stainless Steel Clad Fuel........................................................... 5.4-20 Figure 5.4-14 NAC-STC CY-MPC Radial Dose Rate Profile - Accident Conditions -
Design Basis Zircaloy Clad Fuel.................................................................... 5.4-21 Figure 5.4-15 NAC-STC CY-MPC Radial Dose Rate Profile - Normal Conditions -
GTCC Waste................................................................................................... 5.4-22 Figure 5.4-16 NAC-STC CY-MPC Radial Dose Rate Profile - Accident Conditions -
GTCC Waste................................................................................................... 5.4-23 Figure 5.5-1 SAS2H Input File for Directly Loaded 1414 Fuel at 40,000 MWd/MTU and 2.3 wt % 235U.............................................................................................. 5.5-2 Figure 5.5-2 MCBEND Input File for Directly Loaded 1414 Fuel Gamma Response from Energy Group 7 - Normal Conditions..................................................... 5.5-4 Figure 5.5-3 MCBEND Input File for Directly Loaded 1414 Fuel Neutron Response from Energy Group 2 - Normal Conditions................................................... 5.5-26 Figure 5.5-4 MCBEND Input File for Directly Loaded 1414 Fuel Gamma Response from Energy Group 7 - Accident Conditions................................................. 5.5-44 Figure 5.6.1-1 MPC-LACBWR Fuel Basket Loading Pattern......5.6.1-5 Figure 5.6.1-2 Location of STC-LACBWR Maximum Dose Rates for Normal Conditions of Transport.5.6.1-6 Figure 5.6.1-3 Location of STC-LACBWR Maximum Dose Rates for Hypothetical Accident Conditions..5.6.1-7 Figure 5.6.2-1 STC-LACBWR Fuel Assembly Source Regions and Elevations.................. 5.6.2-5 Figure 5.6.2-2 STC-LACBWR Fuel Bounding Axial Burnup Profile in Active Fuel Region............................................................................................................ 5.6.2-6 Figure 5.6.2-3 STC-LACBWR Fuel Axial Neutron and Gamma Source Profiles in Active Fuel Region.................................................................................................... 5.6.2-7
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-viii List of Figures (continued)
Figure 5.6.2-4 STC-LACBWR Axial Moderator Density Study Neutron Source Comparison................................................................................................. 5.6.2-8 Figure 5.6.2-5 STC-LACBWR Axial Moderator Density Study Gamma Source Comparison................................................................................................. 5.6.2-9 Figure 5.6.2-6 STC-LACBWR Axial Moderator Density Study Hardware Source Comparison............................................................................................... 5.6.2-10 Figure 5.6.3-1 STC-LACBWR Three-Dimensional Canister/Basket Model Detail.......... 5.6.3-6 Figure 5.6.3-2 STC-LACBWR Three-Dimensional Cask Model - Axial Detail............... 5.6.3-7 Figure 5.6.3-3 STC-LACBWR Three-Dimensional Cask Model - Radial Detail............. 5.6.3-8 Figure 5.6.3-4 STC-LACBWR Detector Grid Locations.................................................. 5.6.3-9 Figure 5.6.4-1 STC-LACBWR Normal Condition Radial Dose Rate Profiles -
Undamaged Fuel......................................................................................... 5.6.4-4 Figure 5.6.4-2 STC-LACBWR Normal Condition Radial Surface Dose Rate Profile by Source Type - Undamaged Fuel................................................................. 5.6.4-5 Figure 5.6.4-3 Removed..................................................................................................... 5.6.4-6 Figure 5.6.4-4 STC-LACBWR Normal Condition Azimuthal Dose Rate Profile and Detector Map for Trunnions - Undamaged Fuel........................................ 5.6.4-7 Figure 5.6.4-5 STC-LACBWR Normal Condition Radial 2m Dose Rate Profile by Source Type - Undamaged Fuel................................................................. 5.6.4-8 Figure 5.6.4-6 STC-LACBWR Normal Condition Top Axial Surface Dose Rate Profile by Source Type - Undamaged Fuel............................................................ 5.6.4-9 Figure 5.6.4-7 STC-LACBWR Normal Condition Bottom Axial Surface Dose Rate Profile by Source Type - Undamaged Fuel.............................................. 5.6.4-10 Figure 5.6.4-8 STC-LACBWR Accident Condition Radial Dose Rate Profiles -
Undamaged Fuel....................................................................................... 5.6.4-11 Figure 5.6.4-9 STC-LACBWR Accident Condition Radial Surface Dose Rate Profile by Source Type - Undamaged Fuel.......................................................... 5.6.4-12 Figure 5.6.4-10 STC-LACBWR Accident Condition Azimuthal Dose Rate Profile for Lead Slump - Undamaged Fuel................................................................ 5.6.4-13 Figure 5.6.4-11 STC-LACBWR Accident Condition Azimuthal Dose Rate Profile for Trunnions - Undamaged Fuel................................................................... 5.6.4-14 Figure 5.6.4-12 STC-LACBWR Accident Condition Radial 1m Dose Rate Profile by Source Type - Undamaged Fuel............................................................... 5.6.4-15
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-ix List of Figures (continued)
Figure 5.6.4-13 STC-LACBWR Accident Condition Top Axial Surface Dose Rate Profile by Source Type - Undamaged Fuel........................................... 5.6.4-16 Figure 5.6.4-14 STC-LACBWR Accident Condition Bottom Axial Surface Dose Rate Profile by Source Type - Undamaged Fuel........................................... 5.6.4-17 Figure 5.6.4-15 STC-LACBWR Normal Condition Dose Rate Profile Comparison at Cask Radial Surface - Active Fuel Damaged Fuel................................ 5.6.4-18 Figure 5.6.4-16 STC-LACBWR Normal Condition Dose Rate Profile Comparison at Cask Top Axial Surface - Active Fuel Damaged Fuel.......................... 5.6.4-19 Figure 5.6.4-17 STC-LACBWR Normal Condition Dose Rate Profile Comparison at Cask Bottom Axial Surface - Active Fuel Damaged Fuel.................... 5.6.4-20 Figure 5.6.4-18 STC-LACBWR Accident Condition Dose Rate Profile Comparison at Cask Radial Surface - Active Fuel Damaged Fuel................................ 5.6.4-20 Figure 5.6.4-19 STC-LACBWR Accident Condition Dose Rate Profile Comparison at Cask Top Axial Surface - Active Fuel Damaged Fuel.......................... 5.6.4-21 Figure 5.6.4-20 STC-LACBWR Accident Condition Dose Rate Profile Comparison at Cask Bottom Axial Surface - Active Fuel Damaged Fuel.................... 5.6.4-22 Figure 5.6.4-21 STC-LACBWR Normal Condition Dose Rate Profile at Cask Radial Surface - Non-Fuel Hardware Damaged Fuel....................................... 5.6.4-23 Figure 5.6.4-22 STC-LACBWR Normal Condition Dose Rate Profile at 2m from Cask Radial Surface - Non-Fuel Hardware Damaged Fuel........................... 5.6.4-24 Figure 5.6.4-23 STC-LACBWR Normal Condition Dose Rate Profile at Cask Top Axial Surface - Non-Fuel Hardware Damaged Fuel............................. 5.6.4-25 Figure 5.6.4-24 STC-LACBWR Normal Condition Dose Rate Profile at Cask Bottom Axial Surface - Non-Fuel Hardware Damaged Fuel............................. 5.6.4-26 Figure 5.6.4-25 STC-LACBWR Accident Condition Dose Rate Profile at Cask Radial Surface - Non-Fuel Hardware Damaged Fuel....................................... 5.6.4-27 Figure 5.6.4-26 STC-LACBWR Accident Condition Dose Rate Profile at 1m from Cask Radial Surface - Non-Fuel Hardware Damaged Fuel.................. 5.6.4-28 Figure 5.6.4-27 STC-LACBWR Accident Condition Dose Rate Profile at Cask Top Axial Surface - Non-Fuel Hardware Damaged Fuel............................. 5.6.4-29 Figure 5.6.4-28 STC-LACBWR Accident Condition Dose Rate Profile at Cask Bottom Axial Surface - Non-Fuel Hardware Damaged Fuel............................. 5.6.4-30
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-x List of Figures (continued)
Figure 5.6.4-29 STC-LACBWR Normal Condition Radial Dose Rates-Fresh Fuel versus Spent Fuel Isotopics.................................................................... 5.6.4-31 Figure 5.6.4-30 STC-LACBWR Accident Condition Radial Dose Rates-Fresh Fuel versus Spent Fuel Isotopics.................................................................... 5.6.4-32 Figure 5.6.6-1 STC-LACBWR SAS2H Input File for Allis Chalmers Fuel................... 5.6.6-2 Figure 5.6.6-2 STC-LACBWR SAS2H Input File for Exxon Nuclear Company Fuel.. 5.6.6-4 Figure 5.6.6-3 STC-LACBWR MCNP Input File for Normal Conditions Radial Biasing - Fuel Gamma Source - Undamaged Fuel................................. 5.6.6-6 Figure 5.6.6-4 STC-LACBWR MCNP Input File for Accident Conditions Radial Biasing - Fuel Neutron Source - Undamaged Fuel............................... 5.6.6-20 Figure 5.6.6-5 STC-LACBWR MCNP Input File for Normal Conditions Top Axial Biasing - Damaged Upper End Fitting Gamma Source........................ 5.6.6-34 Figure 5.6.6-6 STC-LACBWR MCNP Input File for Accident Conditions Bottom Axial Biasing - Damaged Lower End Fitting Neutron Source............. 5.6.6-46 Figure 5.7.1-1 Location of STC-WVDP Maximum Dose Rates for Normal Conditions of Transport........................................................................... 5.7.1-3 Figure 5.7.1-2 Location of STC-WVDP Maximum Dose Rates for Hypothetical Accident Conditions................................................................................. 5.7.1-4 Figure 5.7.2-1 HLW Canister Sketch.............................................................................. 5.7.2-2 Figure 5.7.3 1 WVDP HLW Overpack/Basket Model Detail......................................... 5.7.3-5 Figure 5.7.3-2 STC-WVDP Three-Dimensional Cask Model - Axial Detail................. 5.7.3-6 Figure 5.7.3-3 STC-WVDP Three-Dimensional Cask Model - Radial Detail............... 5.7.3-7 Figure 5.7.3-4 STC-WVDP Tally Grid Locations........................................................... 5.7.3-8 Figure 5.7.4-1 STC-WVDP Normal Condition Radial Dose Rate Profiles.................... 5.7.4-3 Figure 5.7.4-2 STC-WVDP Normal Condition Radial Surface Dose Rate Profile by Source Type............................................................................................. 5.7.4-4 Figure 5.7.4-3 STC-WVDP Normal Condition Azimuthal Dose Rate Profile and Tally Map for HLW Midplane................................................................. 5.7.4-5 Figure 5.7.4-4 STC-WVDP Normal Condition Top Axial Surface Dose Rate Profile by Source Type........................................................................................ 5.7.4-6
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xi List of Figures (continued)
Figure 5.7.4-5 STC-WVDP Normal Condition Bottom Axial Surface Dose Rate Profile by Source Type............................................................................ 5.7.4-7 Figure 5.7.4-6 STC-WVDP Accident Condition Radial Dose Rate Profiles.................. 5.7.4-8 Figure 5.7.4-7 STC-WVDP Accident Condition Radial Surface 1 Meter Dose Rate Profile by Source Type............................................................................ 5.7.4-9 Figure 5.7.4-8 STC-WVDP Accident Condition 1 Meter Azimuthal Dose Rate Profile for Lead Slump...................................................................................... 5.7.4-10 Figure 5.7.4-9 STC-WVDP Accident Condition Top Axial Dose Rate Profiles.......... 5.7.4-11 Figure 5.7.4-10 STC-WVDP Accident Condition Bottom Axial Dose Rate Profiles..... 5.7.4-12 Figure 5.7.6-1 STC-WVDP ORIGENs Input File for Maximum Source HLW............. 5.7.6-2 Figure 5.7.6-2 STC-WVDP MCNP Input File for Normal Conditions Radial Biasing
-Gamma Source - Maximum Source...................................................... 5.7.6-4 Figure 5.7.6-3 Accident Condition MCNP Input.......................................................... 5.7.6-14 Figure 5.8.1-1 Location of STC-HBU Maximum Dose Rates for Normal Conditions... 5.8.1-4 Figure 5.8.1-2 Location of STC-HBU Maximum Dose Rates for Hypothetical Accident Conditions................................................................................. 5.8.1-5 Figure 5.8.2-1 STC-HBU Design Basis Burnup Profile.................................................. 5.8.2-5 Figure 5.8.2-2 STC-HBU Fuel Neutron and Gamma Source Profiles............................ 5.8.2-6 Figure 5.8.3-1 XZ VISED Slice for the STC with HBU for Normal Conditions............ 5.8.3-4 Figure 5.8.3-2 XY VISED Slice for the STC with HBU for Normal Conditions........... 5.8.3-4 Figure 5.8.3-3 XZ VISED Slice for the STC with HBU for Accident Conditions......... 5.8.3-5 Figure 5.8.3-4 XY VISED Slice for the STC with HBU for Accident Conditions......... 5.8.3-5 Figure 5.8.4-1 STC-HBU Normal Condition Radial Dose Rate Profiles for DRM and Direct Solutions....................................................................................... 5.8.4-3 Figure 5.8.6-1 Loading Radial Surface Dose Rate [mrem/hr] Mesh -
Nrm.......................................................................................................... 5.8.6-2 Figure 5.8.6-2 Loading Radial Surface Dose Rate Mesh FSDs - Nrm..... 5.8.6-2 NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xii List of Figures (continued)
Figure 5.8.6-3 Loading Side 2m + Vehicle Dose Rate [mrem/hr] Mesh
- Nrm....................................................................................................... 5.8.6-3 Figure 5.8.6-4 Loading Side 2m + Vehicle Dose Rate Mesh FSDs -
Nrm.......................................................................................................... 5.8.6-3 Figure 5.8.6-5 Surface Dose Rate Profile by Source Type for Loading -
Nrm.......................................................................................................... 5.8.6-4 Figure 5.8.6-6 1m Dose Rate Profile by Source Type for Loading -
Acc........................................................................................................... 5.8.6-4 Figure 5.8.7-1 HBU Flat Bed Configuration Loading Radial Surface Dose Rate [mrem/hr] Mesh - Nrm - 0.4 g/kg Cobalt Content................ 5.8.7-2 Figure 5.8.7-2 HBU Flat Bed Configuration Loading Radial Surface Dose Rate Mesh FSDs - Nrm - 0.4 g/kg Cobalt Content....................... 5.8.7-2 Figure 5.8.7-3 HBU Flat Bed Configuration Loading Side 2m + Vehicle Dose Rate [mrem/hr] Mesh - Nrm - 0.4 g/kg Cobalt Content................ 5.8.7-3 Figure 5.8.7-4 HBU Flat Bed Configuration Loading Side 2m + Vehicle Dose Rate Mesh FSDs-Nrm - 0.4 g/kg Cobalt Content.........................5.8.7.3 Figure 5.8.7-5 HBU Flat Bed Configuration Surface Dose Rate Profile by Source Type for Loading - Nrm - 0.4 g/kg Cobalt Content.......... 5.8.7-4 Figure 5.8.7-6 HBU Flat Bed Configuration 1m Dose Rate Profile by Source Type for Loading - Acc - 0.4 g/kg Cobalt Content.................... 5.8.7-4 Figure 5.8.8-1 HBU Flat Bed with Shield Ring Configuration Loading Radial Surface Dose Rate [mrem/hr] Mesh - Nrm.................................. 5.8.8-2 Figure 5.8.8-2 HBU Flat Bed with Shield Ring Configuration Loading Radial Surface Dose Rate Mesh FSDs - Nrm......................................... 5.8.8-2 Figure 5.8.8-3 HBU Flat Bed with Shield Ring Configuration Loading Side 2m + Vehicle Dose Rate [mrem/hr] Mesh - Nrm........................... 5.8.8-3 Figure 5.8.8-4 HBU Flat Bed with Shield Ring Configuration Loading Side 2m + Vehicle Dose Rate Mesh FSDs-Nrm.................................... 5.8.8-3 NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xiii List of Figures (continued)
Figure 5.8.9-1 STC-HBU MCNP Input File for Normal Conditions Radial Biasing - Fuel Gamma Response from Energy Group 7............. 5.8.9-2 Figure 5.8.9-2 STC-HBU MCNP Input File for Accident Conditions Radial Biasing - Fuel Neutron Response from Energy Group 2........... 5.8.9-14 Figure 5.8.9-3 STC-HBU SAS2H Light Element Input File for 49 GWd/MTU Burnup, 2.7 wt. % Initial Enrichment, and Cool Time Range of 18 to 40 Years................................................................................................. 5.8.9-25 Figure 5.9-1 LBU Flat Bed with Shield Ring Configuration Radial Surface Dose Rate [mrem/hr] Mesh - Nrm....................................................................... 5.9-2 Figure 5.9-2 LBU Flat Bed with Shield Ring Configuration Radial Surface Dose Rate Mesh FSDs - Nrm.............................................................................. 5.9-2 Figure 5.9-3 LBU Flat Bed with Shield Ring Configuration Side 2m + Vehicle Dose Rate [mrem/hr] Mesh - Nrm............................................................. 5.9-3 Figure 5.9-4 LBU Flat Bed with Shield Ring Configuration Side 2m + Vehicle Dose Rate Mesh FSDs - Nrm..................................................................... 5.9-3 Figure 5.10-1 XZ VISED Slice for the STC with Reduced Lead................................... 5.10-2 Figure 5.10-2 XY VISED Slice for the STC Augmented Top Weldment...................... 5-10-2 Figure 5.10-3 XZ VISED Slice for the STC Accident Condition Lead Slump with Reduced Lead............................................................................................ 5.10-3 Figure 5.11-1 XZ VISED Slice LBU Flat Bed with Non-Truncated Impact Limiters.... 5.11-2 NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xiv List of Tables Table 5.1-1 Type, Form, Quantity and Potential Sources of the Fuel Used for Design Basis Directly Loaded and Canistered Fuel.................................................... 5.1-17 Table 5.1-2 Design Basis Canistered Fuel - Physical Parameters...................................... 5.1-19 Table 5.1-3 Nuclear Parameters of the Canistered Fuels and GTCC Waste............................................................................................ 5.1-20 Table 5.1-4 Directly Loaded Fuel Maximum Dose Rates for Normal Conditions of Transport..................................................................................................... 5.1-21 Table 5.1-5 Directly Loaded Fuel Maximum Dose Rates for Hypothetical Accident Conditions....................................................................................................... 5.1-22 Table 5.1-6 Combined Top, Radial Midplane, and Bottom Canistered Yankee Class Fuel Dose Rates for Normal Conditions of Transport.................................... 5.1-23 Table 5.1-7 Combined Top, Radial Midplane, and Bottom Canistered Yankee Class Fuel Dose Rates for Hypothetical Accident Conditions................................. 5.1-24 Table 5.1-8 Canistered Yankee GTCC Waste Dose Rates for Normal Conditions of Transport......................................................................................................... 5.1-25 Table 5.1-9 Canistered Yankee GTCC Waste Dose Rates for Hypothetical Accident Conditions........................................................................................ 5.1-26 Table 5.1-10 Connecticut Yankee Stainless Steel Clad Fuel Maximum Dose Rates for Normal Conditions of Transport..................................................................... 5.1-27 Table 5.1-11 Connecticut Yankee Zircaloy Clad Fuel Maximum Dose Rates for Normal Conditions of Transport..................................................................... 5.1-28 Table 5.1-12 Connecticut Yankee Stainless Steel Clad Fuel Maximum Dose Rates for Hypothetical Accident Conditions............................................................ 5.1-29 Table 5.1-13 Connecticut Yankee Zircaloy Clad Fuel Maximum Dose Rates for Hypothetical Accident Conditions.................................................................. 5.1-29 Table 5.1-14 Connecticut Yankee GTCC Waste Maximum Dose Rates for Normal Conditions of Transport.................................................................................. 5.1-30 Table 5.1-15 Connecticut Yankee GTCC Waste Maximum Dose Rates for Hypothetical Accident Conditions.................................................................. 5.1-30 Table 5.2-1 Directly Loaded Three-Dimensional PWR Reference Fuel Assembly Descriptions.................................................................................................... 5.2-19
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xv List of Tables (continued)
Table 5.2-2 PWR Fuel Reactor Operating Conditions for Directly Loaded Fuel.............. 5.2-20 Table 5.2-3 PWR Cycle Length Calculation for Directly Loaded Fuel Source Terms...... 5.2-21 Table 5.2-4 Directly Loaded PWR Fuel Assembly Hardware Mass and Activation Scale Factors by Source Region.................................................... 5.2-22 Table 5.2-5 Directly Loaded Fuel Axial Gamma and Neutron Source Profiles................ 5.2-23 Table 5.2-6 Directly Loaded 1414 Fuel Assembly Spectra at 40,000 MWd/MTU, 2.3 wt % 235U, 10 Years Cool Time in MCBEND Group Format.................. 5.2-24 Table 5.2-7 Design Basis Yankee Class Fuel Input Parameters for SAS2H..................... 5.2-25 Table 5.2-8 Design Basis Yankee Class Fuel Neutron Source Spectra at 36,000 MWD/MTU and 8 Years Cooling.................................................................. 5.2-26 Table 5.2-9 Design Basis Yankee Class Fuel Gamma Source Spectra at 36,000 MWD/MTU and 8 Years Cooling.................................................................. 5.2-27 Table 5.2-10 Design Basis Yankee Class Fuel Hardware and GTCC Waste Gamma Spectra............................................................................................................. 5.2-28 Table 5.2-11 Connecticut Yankee Design Basis Fuel Reactor Operating Conditions......... 5.2-29 Table 5.2-12 Connecticut Yankee Design Basis Stainless Steel Clad Fuel Source Term... 5.2-30 Table 5.2-13 Connecticut Yankee Design Basis Zircaloy Clad Fuel Source Term............. 5.2-31 Table 5.2-14 Connecticut Yankee Design Basis Fuel Assembly Hardware Mass and Mass Scale Factors by Source Region............................................................ 5.2-32 Table 5.2-15 Connecticut Yankee Reactor Operational Cycle History............................... 5.2-33 Table 5.2-16 Connecticut Yankee Design Basis Non-Fuel Assembly Hardware Source Spectra............................................................................................................. 5.2-34 Table 5.2-17 Connecticut Yankee Design Basis Non-Fuel Hardware Masses.................... 5.2-35 Table 5.2-18 CY-MPC Axial Gamma and Neutron Source Profiles - Design Basis Stainless Steel and Zircaloy Clad Fuels.......................................................... 5.2-36 Table 5.2-19 Connecticut Yankee GTCC Waste Source Term at 10 Years Cool Time...... 5.2-37 Table 5.2-20 Isotopic Constituents of the Connecticut Yankee GTCC Waste at 10 Years Cool Time............................................................................................. 5.2-38 Table 5.2-21 MCBEND Standard 28 Group Neutron Boundaries....................................... 5.2-39 Table 5.2-22 MCBEND Standard 22 Group Gamma Boundaries....................................... 5.2-40 Table 5.3-1 Directly Loaded Fuel Region Homogenization.............................................. 5.3-23 Table 5.3-2 Directly Loaded Fuel Homogenized Elemental Densities.............................. 5.3-24
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xvi List of Tables (continued)
Table 5.3-3 Directly Loaded Fuel Assembly Activated Hardware Region Homogenization.............................................................................................. 5.3-25 Table 5.3-4 Directly Loaded Fuel Assembly Zircaloy Hardware Region Homogenization.............................................................................................. 5.3-26 Table 5.3-5 Regional Densities for Directly Loaded Cask Structural and Shield Materials.............................................................................................. 5.3-27 Table 5.3-6 Yankee Class Fuel and Yankee GTCC Material Compositions..................... 5.3-28 Table 5.3-7 Connecticut Yankee Stainless Steel Clad Fuel Region Homogenization....... 5.3-30 Table 5.3-8 Connecticut Yankee Zircaloy Clad Fuel Region Homogenization................. 5.3-30 Table 5.3-9 Connecticut Yankee Homogenized Fuel Regional Densities......................... 5.3-31 Table 5.3-10 Connecticut Yankee Stainless Steel Clad Fuel Assembly Hardware Region Homogenization................................................................................. 5.3-31 Table 5.3-11 Connecticut Yankee Zircaloy Clad Fuel Assembly Hardware Region Homogenization................................................................................. 5.3-32 Table 5.3-12 Regional Densities for CY-MPC Structural and Shield Materials................. 5.3-33 Table 5.4-1 ANSI/ANS 6.1.1-1977 Neutron Flux-to-Dose Conversion Factors............... 5.4-24 Table 5.4-2 ANSI/ANS 6.1.1-1977 Gamma Flux-to-Dose Conversion Factors................ 5.4-25 Table 5.4-3 Minimum Cooling Time Evaluation for 1414 Reference Fuel..................... 5.4-26 Table 5.4-4 Radial Dose Rate Loading Table Results for Directly Loaded Fuel in Normal Conditions of Transport................................................................. 5.4-27 Table 5.4-5 Loading Table for Directly Loaded PWR Fuel............................................... 5.4-28 Table 5.4-6 Detector Maximum Dose Rates for Directly Loaded Fuel in Normal Conditions of Transport.................................................................................. 5.4-29 Table 5.4-7 Detector Maximum Dose Rates for Directly Loaded Fuel in Accident Conditions....................................................................................................... 5.4-30 Table 5.4-8 Directly Loaded Radial Detector Description for Normal Conditions of Transport..................................................................................................... 5.4-31 Table 5.4-9 Directly Loaded Radial Detector Description for Accident Conditions of Transport.................................................................................. 5.4-31 Table 5.4-10 CY-MPC Neutron Flux-to-Dose Conversion Factors..................................... 5.4-32 Table 5.4-11 CY-MPC Gamma Flux-to-Dose Conversion Factors.................................... 5.4-33
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xvii List of Tables (continued)
Table 5.4-12 NAC-STC CY-MPC Detector Maximum Dose Rates - Normal Conditions - Design Basis Stainless Steel Clad Fuel..................................... 5.4-34 Table 5.4-13 NAC-STC CY-MPC Detector Maximum Dose Rates - Normal Conditions - Design Basis Zircaloy Clad Fuel............................................... 5.4-35 Table 5.4-14 NAC-STC CY-MPC Detector Maximum Dose Rates - Accident Conditions - Design Basis Stainless Steel Clad Fuel..................................... 5.4-36 Table 5.4-15 NAC-STC CY-MPC Detector Maximum Dose Rates - Accident Conditions - Design Basis Zircaloy Clad Fuel............................................... 5.4-37 Table 5.4-16 NAC-STC CY-MPC Detector Average Dose Rates - Normal Conditions - Design Basis Stainless Steel Clad Fuel..................................... 5.4-38 Table 5.4-17 NAC-STC CY-MPC Detector Average Dose Rates - Normal Conditions - Design Basis Zircaloy Clad Fuel............................................... 5.4-38 Table 5.4-18 NAC-STC CY-MPC Detector Average Dose Rates - Accident Conditions - Design Basis Stainless Steel Clad Fuel..................................... 5.4-39 Table 5.4-19 NAC-STC CY-MPC Detector Average Dose Rates - Accident Conditions - Design Basis Zircaloy Clad Fuel............................................... 5.4-39 Table 5.4-20 NAC-STC CY-MPC Detector Maximum Dose Rates - Normal Conditions - Design Basis GTCC Waste....................................................... 5.4-40 Table 5.4-21 NAC-STC CY-MPC Detector Maximum Dose Rates - Accident Conditions - Design Basis GTCC Waste....................................................... 5.4-40 Table 5.4-22 NAC-STC CY-MPC Detector Average Dose Rates - Normal Conditions - Design Basis GTCC Waste....................................................... 5.4-41 Table 5.4-23 NAC-STC CY-MPC Detector Average Dose Rates - Accident Conditions - Design Basis GTCC Waste....................................................... 5.4-41 Table 5.6.1-1 Summary of STC-LACBWR Normal Condition Maximum Dose Rates - Undamaged Fuel...................................................................... 5.6.1-8 Table 5.6.1-2 Summary of STC-LACBWR Accident Condition Maximum Dose Rates - Undamaged Fuel...................................................................... 5.6.1-8 Table 5.6.1-3 Summary of STC-LACBWR Normal Condition Maximum Dose Rates - Damaged Fuel.......................................................................... 5.6.1-9 Table 5.6.1-4 Summary of STC-LACBWR Accident Condition Maximum Dose Rates - Damaged Fuel.......................................................................... 5.6.1-9
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xviii List of Tables (continued)
Table 5.6.2-1 STC-LACBWR Fuel Characteristics for Shielding Evaluations............... 5.6.2-11 Table 5.6.2-2 STC-LACBWR Fuel Reactor Operating Conditions................................. 5.6.2-12 Table 5.6.2-3 STC-LACBWR Fuel Assembly Neutron Spectra...................................... 5.6.2-13 Table 5.6.2-4 STC-LACBWR Fuel Assembly Gamma Spectra...................................... 5.6.2-14 Table 5.6.2-5 STC-LACBWR Activated Hardware Gamma Spectra.............................. 5.6.2-15 Table 5.6.2-6 STC-LACBWR Fuel Assembly Activated Hardware Mass and Mass Scale Factors by Source Region....................................................... 5.6.2-16 Table 5.6.2-7 STC-LACBWR Fuel Assembly Decay Heat............................................. 5.6.2-16 Table 5.6.2-8 STC-LACBWR Axial Gamma and Neutron Source Profiles.................... 5.6.2-17 Table 5.6.2-9 STC-LACBWR Source Term Input for Axial Moderator Density Study.. 5.6.2-17 Table 5.6.2-10 STC-LACBWR Result Comparison for Axial Moderator Density Study. 5.6.2-17 Table 5.6.2-11 LACBWR Normalized Neutron Spectra Comparison for Axial Moderator Density Study........................................................................... 5.6.2-18 Table 5.6.2-12 LACBWR Normalized Fuel Gamma Spectra Comparison for Axial Moderator Density Study........................................................................... 5.6.2-19 Table 5.6.2-13 LACBWR Normalized Hardware Gamma Spectra Comparison for Axial Moderator Density Study................................................................. 5.6.2-20 Table 5.6.3-1 STC-LACBWR Active Fuel Region Homogenization.............................. 5.6.3-10 Table 5.6.3-2 STC-LACBWR Fuel Assembly Hardware Region Homogenization........ 5.6.3-10 Table 5.6.3-3 STC-LACBWR Typical Radial Surface Detector Division....................... 5.6.3-11 Table 5.6.3-4 STC-LACBWR Typical Axial Surface Detector Division........................ 5.6.3-11 Table 5.6.3-5 STC-LACBWR Homogenized Fuel Assembly Regional Densities.......... 5.6.3-12 Table 5.6.3-6 STC-LACBWR Structural and Shield Material Regional Densities.......... 5.6.3-13 Table 5.6.4-1 ANSI Standard Neutron Flux-To-Dose Rate Factors................................. 5.6.4-33 Table 5.6.4-2 ANSI Standard Gamma Flux-To-Dose Rate Factors..................... 5.6.4-34 Table 5.7.1-1 Summary of STC-WVDP Normal Condition Maximum Dose Rates -
Maximum Source......................................................................................... 5.7.1-5 Table 5.7.1-2 Summary of STC-WVDP Accident Condition Maximum Dose Rates -
Maximum Source......................................................................................... 5.7.1-5 Table 5.7.2-1 HLW Canister Characteristics for Shielding Evaluations............................ 5.7.2-3
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xix List of Tables (continued)
Table 5.7.2-2 Bounding Activities for HLW Glass............................................................ 5.7.2-3 Table 5.7.2-3 HLW Canister Neutron Spectra................................................................... 5.7.2-4 Table 5.7.2-4 HLW Canister Gamma Spectra.................................................................... 5.7.2-5 Table 5.7.3-1 STC-WVDP Typical Radial Surface Tally Division................................... 5.7.3-9 Table 5.7.3-2 STC-WVDP Typical Axial Surface Tally Division..................................... 5.7.3-9 Table 5.7.3-3 STC-WVDP HLW Glass Modeled Composition and Density.................... 5.7.3-9 Table 5.7.3-4 STC-WVDP Structural and Shield Material Regional Densities............... 5.7.3-10 Table 5.7.4-1 ANSI Standard Neutron Flux-To-Dose Rate Factors................................. 5.7.4-13 Table 5.7.4-2 ANSI Standard Gamma Flux-To-Dose Rate Factors................................. 5.7.4-14 Table 5.8.1-1 Type, Form, Quantity and Potential Sources of the Fuel Used for Design Basis STC-HBU Contents............................................................................ 5.8.1-6 Table 5.8.1-2 Minimum Cool Time [years] Summary for 14 Assembly Loading of HBU............................................................................................................. 5.8.1-7 Table 5.8.1-3 Minimum Cool Time [years] Summary for 16 Assembly Loading of HBU............................................................................................................. 5.8.1-8 Table 5.8.1-4 Minimum Cool Time [years] Summary for 20 Assembly Loading of HBU............................................................................................................. 5.8.1-9 Table 5.8.2-1 NAC-STC Neutron and Gamma Source Term Axial Profile....................... 5.8.2-7 Table 5.8.2-2 TRITON Neutron Source Term Comparison............................................... 5.8.2-8 Table 5.8.3-1 STC-HBU Cylindrical (RZT) Mesh Radial Detector Description............... 5.8.3-6 Table 5.8.3-2 STC-HBU Rectangular (XYZ) Mesh Radial Detector Description............. 5.8.3-6 Table 5.8.3-3 STC-HBU Cylindrical (RZT) Mesh Top Axial Detector Description......... 5.8.3-6 Table 5.8.3-4 STC-HBU Cylindrical (RZT)
Mesh Bottom Axial Detector Description................................................................................................... 5.8.3-6 Table 5.8.5-1 Example Loading Table for STC-HBU........................................................ 5.8.5-2 Table 5.8.5-2 Loading Source Terms for Maximum Dose Rates -
Normal Conditions....................................................................................... 5.8.5-3 Table 5.8.5-3 Loading Source Terms for Maximum Dose Rates -
Accident Conditions..................................................................................... 5.8.5-3 Table 5.8.5-4 Loading Source Terms for Maximum Dose Rates -
Normal Conditions....................................................................................... 5.8.5-3 NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xx List of Tables (continued)
Table 5.8.5-5 Loading Source Terms for Maximum Dose Rates -
Accident Conditions..................................................................................... 5.8.5-3 Table 5.8.5-6 Loading Source Terms for Maximum Dose Rates -
Normal Conditions....................................................................................... 5.8.5-3 Table 5.8.5-7 Loading Source Terms for Maximum Dose Rates -
Accident Conditions..................................................................................... 5.8.5-3 Table 5.8.6-1 HBU Maximum Dose Rates for Normal Conditions................................... 5.8.6-5 Table 5.8.6-2 HBU Maximum 1m Dose Rates for Accident Conditions........................... 5.8.6-5 Table 5.8.7-1 HBU Flat-Bed Configuration Minimum Cool Times -
Loading......................................................................................................... 5.8.7-5 Table 5.8.7-2 HBU Flat-Bed Configuration Minimum Cool Times -
Loading......................................................................................................... 5.8.7-6 Table 5.8.7-3 HBU Flat-Bed Configuraion Minimum Cool Times -
Conditions.................................................................................................... 5.8.7-7 Table 5.8.7-4 HBU Flat-Bed Configuration Bounding Source Terms - Normal Conditions................................................................................................... 5.8.7-8 Table 5.8.7-5 HBU Flat-Bed Configuration Bounding Source Terms - Accident Conditions.................................................................................................... 5.8.7-8 Table 5.8.7-6 HBU Flat-Bed Configuration Maximum Dose Rates for Normal Conditions -
Loading............................................................. 5.8.7-8 Table 5.8.7-7 HBU Flat-Bed Configuration Maximum 1m Dose Rates for Accident Conditions -
Loading............................................................ 5. 8.7-9 Table 5.8.7-8 HBU Flat-Bed Configuration Maximum Dose Rates for Normal Conditions -
Loading............................................................. 5.8.7-9 Table 5.8.7-9 HBU Flat-Bed Configuration Maximum 1m Dose Rates for Accident Conditions -
Loading............................................................. 5.8.7-9 Table 5.8.7-10 HBU Flat-Bed Configuration Maximum Dose Rates for Normal Conditions -
Loading........................................................... 5.8.7-10 Table 5.8.7-11 HBU Flat-Bed Configuration Maximum 1m Dose Rates for Accident Conditions -
Loading........................................................... 5.8.7-10 NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5-xxi List of Tables (continued)
Table 5.8.8-1 HBU Flat-Bed with Shield Ring Configuration Minimum Cool Tmes -
Loading................................................................................... 5.8.8-4 Table 5.8.8-2 HBU Flat-Bed with Shield Ring Configuration Minimum Cool Times -
Loading................................................................................... 5.8.8-4 Table 5.8.8-3 HBU Flat-Bed with Shield Ring Configuration Minimum Cool Times -
Loading................................................................................... 5.8.8-4 Table 5.8.8-4 HBU Flat Bed with Shield Ring Configuration Bounding Dose Rates -
Loading................................................................................... 5.8.8-5 Table 5.8.8-5 HBU Flat Bed with Shield Ring Configuration Bounding Dose Rates -
Loading................................................................................... 5.8.8-5 Table 5.8.8-6 HBU Flat Bed with Shield Ring Configuration Bounding Dose Rates -
Loading................................................................................... 5.8.8-6 Table 5.9-1 LBU Flat Bed with Shield Ring Configuration Minimum Cool Times.......... 5-9-4 Table 5.9-2 LBU Flat Bed with Shield Ring Configuration Bounding Dose Rages.......... 5.9-5 Table 5.10-1 HBU Flat Bed Change in Cool Time [years] with Augmented Top Weldment and Reduced Lead -
.............................................. 5.10-4 Table 5.10-2 HBU Flat Bed Change in Cool Time [years] with Augmented Top Weldment and Reduced Lead -
.............................................. 5.10-5 Table 5.10-3 HBU Flat Bed Change in Cool Time [years] with Augmented Top Weldment and Reduced Lead -
.............................................. 5.10-6 Table 5.11-1 LBU Flat Bed Change in Cool Time [years] with Augmented Top Weldment and Reduced Lead....................................................................... 5.11-3 NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5.12-1 5.12 Evaluation for Uncertainty in NS-4-FR Hydrogen Measurement This section evaluates the effect on dose rates for reduced hydrogen in the NS-4-FR neutron shield.
The high burnup directly loaded, low burnup directly loaded, and canistered configurations are analyzed.
The baseline hydrogen content and NS-4-FR density for the licensing basis calculations are respectively. Evaluations addressing uncertainty in the hydrogen measurement apply a reduced hydrogen content of using the licensing density of These evaluations were performed for the fuel neutron and n-gamma radial cases only. The fuel neutron and n-gamma dose rates are dependent on hydrogen content in the neutron shield and the reduction in hydrogen density, equivalent to material density times hydrogen weight fraction, must be addressed. Only radial components are re-evaluated as axial dose rates are not near limits and axial neutron shields are small (thin) compared to the radial NS-4-FR thickness. Radial shields will therefore show the maximum effect of hydrogen content changes.
5.12.1 High Burnup Configuration The full licensing range of the cool times included in Section 5.8 were reevaluated using the reduced hydrogen content using dose rate response functions. The method of evaluation is identical to that in Section 5.8. The evaluation demonstrates that at the current licensed cool times, the system meets regulatory dose rates (10 mrem/hr at 2 meter and 1000 mrem/hr on the cask surface).
Slight increases in calculated values are observed but these changes are generally within the statistical uncertainty of the existing evaluation. To allow for such minor variances/uncertainty evaluations, the licensing evaluation limited the 2-meter dose rate to 9.5 mrem/hr. The reduced limit is not a regulatory limit and was established for the specific reason to address unevaluated uncertainties/variances not captured in the licensing evaluation.
Radial, normal condition, maximum calculated dose rates are listed in Table 5.8.6-1 for the 20-assembly configuration. As clearly indicated in the table, the 2-meter dose rate is below the licensing limit and at the NAC limit. The maximum calculated 2-meter dose rate is 9.5 mrem/hr.
The maximum surface dose rate was calculated at 375.9 mrem/hr.
NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5.12-2 Re-evaluating the cool time tables against the reduced hydrogen content produced the following results.
Hydrogen Weight Fraction (%)
Radial 2m Maximum Dose Rate (mrem/hr)
Radial Surface Maximum Dose Rate** (mrem/hr)
Nominal value (see Section 5.3.1.4)
Maximum dose rate occurs at the gap between the top of the neutron shield and the upper impact limiter The reduced hydrogen analysis demonstrates that while dose rates at the 2-meter location show a minor increase, the change is not significant versus the margin to the 10.0 mrem/hr licensing limit (greater than 0.2 mrem/hr of margin remains). The cask surface dose rate change is well within the Monte Carlo uncertainty of the analysis with a negligible change (~1%) and is likely the result the result of Monte Carlo analysis fluctuations as the bounding location is one without the neutron shield.
Based on the rationale above, a minimum hydrogen content of is acceptable and does not significantly affect calculated results, SAR safety conclusions, or conformance with regulatory limits.
The evaluation of the 20-assembly configuration is applicable to the 16-and 14-assembly configurations as all high burnup variations rely on the aluminum heat transfer shunts to shield the 2-meter plane.
5.12.2 Low Burnup Configuration The full licensing range of the cool times included in Table 5.4-5 were reevaluated using the reduced hydrogen content using dose rate response functions. Other than the change in code package from MCBEND to MCNP, the method of evaluation is identical to that in Section 5.4.
The evaluation demonstrates that at the current licensed cool times, the system meets regulatory dose rates (10 mrem/hr at 2 meter and 1000 mrem/hr on the cask surface). Slight increases in calculated values are observed but these changes are generally within the statistical uncertainty of the existing evaluation. To allow for such minor variances/uncertainty evaluations, the licensing evaluation limited the 2-meter dose rate to 9.5 mrem/hr. The reduced limit is not a regulatory limit and was established for the specific reason to address unevaluated uncertainties/variances not captured in the licensing evaluation.
NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 5.12-3 Due to the change in code packages (and cross-section libraries), this evaluation focuses on the change in fuel neutron and n-gamma dose rates. Re-evaluating the cool time tables against the baseline and reduced hydrogen content produced the following results.
Hydrogen Weight Fraction (%)
Radial 2m Maximum Neutron and N-Gamma Dose Rate (mrem/hr)
Radial Surface Maximum Neutron and N-Gamma Dose Rate** (mrem/hr)
Nominal value (see Section 5.3.1.4)
Maximum dose rate occurs at the lower trunnion elevation The reduced hydrogen analysis demonstrates that while dose rates at the 2-meter location show a minor increase, the change is not significant versus the margin to the 10.0 mrem/hr licensing limit (0.3 mrem/hr of margin remains). The cask surface dose rate change is well within the Monte Carlo uncertainty of the analysis with a negligible change (~1%) and is likely the result the result of Monte Carlo analysis fluctuations as the bounding location is one without the neutron shield.
Based on the rationale above, a minimum hydrogen content of is acceptable and does not significantly affect calculated results, SAR safety conclusions, or conformance with regulatory limits.
5.12.3 Canistered Configuration The canistered contents of the NAC-STC, due to the additional shielding of the canister along with significant cool time of decommissioned facilities, have significant margin to regulatory limits compared to directly loaded contents. Applying the bounding dose rate changes from the directly loaded contents to the bounding dose rates of Connecticut Yankee fuel yields only a minor increase in dose rates.
Based on the rationale above, a minimum hydrogen content of is acceptable and does not significantly affect calculated results, SAR safety conclusions, or conformance with regulatory limits.
NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7-ii List of Figures Figure 7.3-1 Cask Cooldown Piping and Controls Schematic......................................... 7.3-11 List of Tables Table 7-1 Torque Table..................................................................................................... 7-3 Table 7.4-1 NAC-STC Containment Boundary Leakage Testing Requirements............. 7.4-7
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7-2 NAC-STC. In the canistered configuration, damaged (failed) fuel will be separately containerized (canned) and sealed in the canister prior to transport.
The user shall verify that the NAC-STC transport cask has the correct O-ring configuration for the intended use. The transport cask may be configured with either metallic O-rings or with non-metallic Viton O-rings in the following combinations: 1) double (i.e., inner and outer seals) metallic seals; 2) double Viton O-rings (i.e., inner and outer seals); and 3) single metallic seal (inner) and Viton O-ring (outer seal). The O-rings may not be used interchangeably, since each O-ring type requires a different lid O-ring groove configuration. Consequently, the inner lid, vent and drain port coverplates and outer lid are machined with a square O-ring groove to accept metallic O-rings or are machined with a truncated triangular (dove-tail) groove to accept Viton O-rings, applicable to the O-ring seal combination.
Double Viton O-rings may be used only when directly loading spent fuel for transport without interim storage. The inner metallic seal and Viton O-ring combination may also be used for directly loaded spent fuel for transport without interim storage. Double metallic O-rings must be used when directly loading spent fuel for an extended period of storage and may be used when directly loading standard PWR spent fuel assemblies having burnups of 45 GWd/MTU for transport without interim storage. Double metallic O-rings must also be used when loading canistered fuel, Greater Than Class C (GTCC) waste, or canistered High Level Waste (HLW) for transport. The double Viton O-ring and inner metallic and outer Viton O-ring seal combinations may be tested to leaktight criteria for the containment boundary for the directly loaded HBU fuel assemblies. The metallic and nonmetallic O-rings have different limits of allowable leakage rates as specified in the procedures and in Table 7.4-1.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.1-10
- 5) Drain approximately 50 gallons from the cask cavity by connecting a helium (99.9% minimum purity) supply to the vent port quick-disconnect (located in the inner lid) and a drain line to the drain port quick disconnect. Purge the water from the cask by pressurizing to 35 to 40 psig. Following removal of approximately 50 gallons, turn the helium supply off and maintain the helium pressure above the cavity water.
- 6) Remove the inner lid interseal test port plug and connect the helium Mass Spectrometer Leak Detector (MSLD) to the interseal test port to verify the new inner lid inner Viton O-ring leakage rate is 2.0 x 10-7 cm3/sec (helium) with a minimum test sensitivity of 1.0 x 10-7 cm3/sec (helium).
- 7) After successful completion of the maintenance leakage rate test of the inner lid O-ring seals, the cask preparation and loading procedures will restart at Step 5 in Section 7.1.2.1.
- 18. Drain the cask cavity by connecting a helium supply to the vent port quick-disconnect and a drain line to the drain port quick-disconnect. Purge the water from the cask by pressurizing to 35 to 40 psig and hold until all water is removed (observed when no water is coming from the drain line). Turn the helium supply off, vent the helium from the cavity and disconnect the helium supply line from the vent port. Then, disconnect the drain line from the drain port quick-disconnect.
- 19. Connect a vacuum drying system to the cask cavity via the vent and drain port quick-disconnects in the inner lid. Evacuate the cask cavity until a pressure of 4 mbar is reached. Continue pumping for a minimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after reaching 4 mbar. Valve off vacuum pump from system and turn vacuum pump off. Using a calibrated vacuum gauge (minimum gauge readability of 2.5 mbar), observe for a cask cavity pressure rise. If a pressure rise (P) of more than 12 mbar in ten minutes is observed, continue pumping until the pressure does not rise more than 12 mbar in ten minutes. Repeat dryness test until cavity dryness has been verified (P < 12 mbar in 10 minutes). Record test results in the cask loading report.
- 20. Without disconnecting the vacuum drying system from the vent and drain port quick disconnects and allowing air to re-enter the cask cavity, turn off and isolate the vacuum pump. Backfill the cask cavity with helium (99.9% minimum purity) through the vent port quick-disconnect to a final helium pressure of 0 psig helium pressure (+1, -0 psi).
- 21. Install the drain and vent port coverplates using new metallic O-rings or inspected Viton O-rings. Torque the bolts to the value indicated in Table 7-1.
- 22. Perform inner lid O-ring leakage testing of a NAC-STC containing PWR spent fuel with metallic or Viton containment seals as follows:
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.1-11 22a. For the inner metallic seal and outer Viton O-ring assembly or the double inner and outer metallic seal assemblies, connect a vacuum pump to the inner lid interseal port via the quick-disconnect in the inner lid. Evacuate the inner lid interseal volume between the inner and outer O-rings until a pressure of 4 mbar is reached to remove water and moisture from the inner lid interseal volume. Disconnect vacuum pump and connect the helium Mass Spectrometer Leak Detector (MSLD) to the inner lid interseal test port and evacuate the volume between the O-rings to <1 mbar. Maintain the vacuum on the interseal for the metallic O-ring assembly region and using the helium leak detector, verify that any detectable leak rate for metallic O-rings is 2 x 10-7 cm3/sec (helium). The test sensitivity shall be 1 x10-7 cm3/sec (helium).
22b. For inner and outer Viton O-rings, connect a vacuum pump to the inner lid interseal port via the quick-disconnect in the inner lid and evacuate the inner lid interseal volume between the inner and outer Viton O-rings until a pressure of 4 mbar is reached. Continue pumping for a minimum of 30 minutes after reaching 4 mbar.
Perform the preshipment leakage rate test to confirm no detected leakage to a test sensitivity of 1 x 10-3 ref cm3/sec by pressurizing the O-ring annulus with helium gas to 15 (+2, -0) psig and isolating for a minimum of 15 minutes. There shall be no loss in pressure during the test period. If test is acceptable vent and disconnect the helium pressure test system from the interseal test port and proceed with cask preparation procedures per Step 23. If test is not acceptable after two attempts, prepare cask for re-immersion in the spent fuel pool for lid removal and Viton O-ring seal replacement.
22c. Following the replacement of inner lid inner Viton O-rings required due to excessive wear of the O-rings or failure of the preshipment leakage rate test, the inner lid maintenance leakage rate test shall be performed. Connect a vacuum pump to the inner lid interseal port via the quick-disconnect in the inner lid and evacuate the inner lid interseal volume between the inner and outer Viton O-rings until a pressure of 4 mbar is reached. Continue pumping for a minimum of 30 minutes after reaching 4 mbar. Disconnect vacuum pump and connect the helium Mass Spectrometer Leak Detector (MSLD) to the inner lid interseal test port and evacuate the volume between the O-rings to <1 mbar. Perform the maintenance leakage rate test to verify the total cumulative leakage rate is as shown in Table 7.4-1.
22d. Upon successful completion of the inner lid O-ring leakage test (e.g., maintenance or preshipment), vent and disconnect the leakage test equipment from the interseal test port and proceed with cask preparation procedures per Step 23.
- 23. Install the test port plug for the inner lid interseal test port using a new metallic or Viton O-ring and torque the plug to the value specified in Table 7-1.
- 24. Perform preshipment of maintenance leakage rate testing of vent and drain port coverplates as follows:
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.1-12 24a. For the inner metallic O-ring assembly, remove vent port coverplate interseal port plug and connect the helium Mass Spectrometer Leak Detector (MSLD) to the vent port interseal test port and evacuate the volume between the O-rings to <1 mbar. Perform the maintenance leakage rate test to verify the leakage rate is 2 x 10-7 cm3/sec (helium). The test sensitivity shall be 1 x10-7 cm3/sec (helium).
24b. For Viton O-ring assembly, perform the preshipment leakage rate test to confirm no detected leakage to a test sensitivity of 1 x 10-3 ref cm3/sec by pressurizing the O-ring annulus to 15 (+2, -0) psig and isolating for a minimum of 15 minutes. There shall be no loss in pressure during the test period.
24c. For new replacement Viton O-rings, use a leak detector connected to the interseal test port to verify the total cumulative leakage rate is as shown in Table 7.4-1.
24d. Upon successful completion of the port coverplate inner O-ring leakage test (e.g.,
maintenance or preshipment), vent and disconnect the leakage test equipment from the interseal test port.
- 25. Repeat Step 24 for drain port coverplate.
- 26. Install the test port plugs for the vent and drain port coverplates using a new metallic or Viton O-ring, as applicable, and torque the plugs to the value specified in Table 7-1.
- 27. Drain residual water from the pressure port, ensuring that the pressure port is clear to also allow water to drain from the interlid region.
- 28. Install the transport pressure port cover on the pressure port. Torque the port cover bolts to the value specified in Table 7-1.
- 29. Perform a functional leak test on the pressure port cover by removing the O-ring test plug and using a test fixture, pressurize the annulus between the pressure port cover O-rings to 15 psig and isolate. During a 10-minute test period, there shall be no loss in pressure during the test period.
- 30. Install the pressure port cover interseal test port plug and O-ring and torque the plug to the value specified in Table 7-1.
- 31. For the metallic outer lid O-ring assembly, remove the O-ring, clean the O-ring seating surface and groove, and install a new metallic O-ring. For Viton O-ring assemblies, inspect the O-ring and replace if damaged.
- 32. Install outer lid and align vent pins.
- 33. Attach the outer lid lifting device to the outer lid and overhead crane. Install the outer lid using the alignment pins to assist in proper seating. Remove the outer lid alignment pins.
Install the outer lid bolts and torque to the value specified in Table 7-1. The bolt torquing sequence is shown on the outer lid.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.1-13
- 34. Attach a supply of air, nitrogen or helium to the interlid port quick-disconnect. Backfill the interlid volume to 15 psig air, nitrogen or helium and hold for 10 minutes. There shall be no pressure loss during the test period. Disconnect air or helium supply.
- 35. Install the interlid port cover using new metallic O-rings. Torque the interlid port cover bolts to the value specified in Table 7-1.
- 36. Remove the test plug from the interlid port cover and, using the O-ring test fixture, pressurize the O-ring annulus to 15 psig with air, nitrogen or helium. Isolate the annulus and hold for 10 minutes. No loss of pressure is permitted during the test period.
- 37. Remove the air, nitrogen or helium supply and vent the annulus pressure. Replace the metallic O-ring on the interlid port cover test plug, install the test plug and torque it to the value specified in Table 7-1.
- 38. If using the optional shield ring assembly, confirm the lower sector is installed. If necessary, install the lower section by attaching a sling to the lower sector and secure to the upper forging using the socket head cap screws. Torque the socket head cap screws to the value prescribed in Table 7-1.
- 39. Perform final external decontamination and perform survey to verify acceptable level of removable contamination to ensure compliance with 49 CFR 173.443. Perform final radiation survey. Record the survey results.
- 40. Perform final visual inspection to verify assembly of the NAC-STC in accordance with the Certificate of Compliance. Verify that the loading documentation has been appropriately completed and signed off.
7.1.3.2 Loading Canistered Fuel, Canistered GTCC Waste, or HLW Overpacks Canistered fuel, canistered GTCC waste, or HLW Overpacks are loaded into the NAC-STC using a transfer cask. This procedure assumes that the canister, or overpack, has been previously loaded, drained, vacuum dried, backfilled with helium and welded closed, as applicable. The canister, or overpack, may have been retrieved from dry storage, or it may have been loaded and sealed immediately prior to loading in the NAC-STC.
Canisters containing spent nuclear fuel that are to be retrieved from storage for off-site transport will be evaluated to ensure that the specific canister stored in the storage overpack, which may have been subject to 10 CFR 72 normal, off-normal, accident and natural phenomena events, retain their ability to satisfy functional and performance requirements of the NAC-STC packaging certified content conditions. Similarly, GTCC Waste canisters and HLW Overpacks will be evaluated to ensure that the specific canister or overpack, which may have been exposed to off-normal, accident and/or natural phenomena events during storage operations prior to
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.1-14 loading for transport, retain their ability to satisfy functional and performance requirements of the NAC-STC packaging certified content conditions.
Canisters containing spent nuclear fuel experiencing only normal or off-normal events during storage, and canistered GTCC waste and HLW Overpacks need only be evaluated for potential corrosion at the welds and any damage caused by removal from the storage cask.
In addition to the evaluation done for normal/off-normal storage, canisters containing spent nuclear fuel that have experienced accident or natural phenomena events must be evaluated for potential degradation of the fuel, basket and neutron absorbers. This evaluation will be performed for each canister as part of the preparation for loading for off-site transport using:
- 1) the annual inspection and surveillance records and off-normal and accident event reports that are maintained by the licensee for each loaded NAC-MPC system in compliance with 10 CFR 72 requirements; and 2) in the case of storage accidents and natural phenomena events, any necessary examinations at the time of transfer to ensure the condition of the canister and contents.
Dry storage systems that have been maintained within an Aging Management Program will include system specific review and assessment of this information record as part of the off-site transport evaluation to ensure the NAC-STC packaging certified content conditions are validated. Maximum assembly average burnup for fuel assemblies retrieved from dry storage for off-site transport is limited to 45,000 MWd/MTU. System loading into the NAC-STC will be observed by operations staff noting any system interferences that occur during canister retrieval from the storage overpack and placement of the canister into the transport overpack. The cause of the interference and potential damage caused by the interference will be determined prior to shipment. Noted interferences will be made part of the canister evaluation record to the extent required to validate NAC-STC packaging content conditions are satisfied when the spent fuel canister is placed within the NAC-STC containment boundary for off-site transport.
This procedure assumes that the sealed canister, or HLW Overpack, conforms to the design basis of the NAC-STC with appropriate spacer configuration and that the canister is already in the transfer cask.
- 1. Attach the transfer cask yoke to the cask handling crane hook.
- 2. Engage the transfer cask yoke to the trunnions of the transfer cask.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.1-15
- 3. Raise the transfer cask over the NAC-STC cask and lower it until it rests on the transfer cask adapter plate. Remove and store the transfer cask lifting yoke. Remove the transfer cask shield door stops.
- 4. Attach the two (2) canister 3-legged lifting sling sets to the hoist rings in the canister lid.
Attach the opposite end of the slings to the crane hook.
Note: Alternative canister lifting systems may be utilized.
- 5. Attach the hydraulic system to the operating cylinders on the transfer cask adapter plate.
- 6. Using the crane, raise the canister just enough ( 1 inch) to take the canister weight off of the transfer cask bottom shield doors.
- 7. Open the transfer cask shield doors.
- 8. Lower the canister or HLW overpack into the NAC-STC cask. Exercise caution to avoid contact with the interior cavity wall.
Note: Prior to loading into the NAC-STC cavity the condition of the spent fuel canister, greater than class C (GTCC) waste canister, or the HLW overpack, and the canister/overpack internals shall be evaluated to verify the canisters/overpacks:
- a.
Meet the design requirements and CoC content conditions of the NAC-STC package;
- b.
Account for the effects of any accident or natural phenomena events that the canisters or overpacks may have been exposed to during storage operations prior to loading in the NAC-STC package, and,
- c.
The vitrified HLW overpack meets the limits in 10 CFR 71.15 for classifying the contents as fissile exempt.
- 9. Disconnect and remove the canister lifting sling from the crane hook and lower it onto the top of the canister.
- 10. Close the transfer cask shield doors and install the door stops.
- 11. Retrieve the transfer cask lifting yoke and engage the transfer cask trunnions. Lift the transfer cask from the transfer cask adapter plate. Store the transfer cask and transfer cask lifting yoke in the designated locations.
- 12. After removal of the lift slings, install the NAC-MPC canister top spacer, as required (for the loading of Yankee-MPC, MPC-LACBWR and HLW Overpacks only).
- 13. Retrieve the cask adapter plate lifting sling and attach it to the transfer cask adapter plate.
- 14. Remove the transfer cask adapter plate and store it in the designated location. Using the appropriate lifting sling, remove the adapter ring and bolts. Install the inner lid alignment pins.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.1-16
- 15. Remove the inner lid O-rings and clean inner lid O-ring groove surfaces. Replace the metallic O-rings on the inner lid, carefully inspecting the new O-rings for damage prior to installation. Secure the O-rings in the groove using the O-ring clips and screws.
- 16. Attach the inner lid lifting slings to an auxiliary crane hook, lift the inner lid and place it on the cask using the inner lid alignment pins to assist in proper lid seating and orientation. Visually verify proper lid position.
- 17. Disconnect the lid lifting device from the crane hook and remove it from the inner lid.
- 18. Install at least 10 inner lid bolts equally spaced on the bolt circle to hand tight. Remove the inner lid alignment pins.
- 19. Install the remaining inner lid bolts and torque all of the bolts to the torque value specified in Table 7-1. The bolt torquing sequence is shown on the inner lid.
- 20. Remove the metallic O-rings in the drain port coverplate, and clean and inspect the O-ring groove. Install new metallic O-rings and install the coverplate. Torque the coverplate bolts to the value specified in Table 7-1.
- 21. Connect the vacuum pump to the cask vent port and evacuate the cask cavity to a stable vacuum pressure of less than, or equal to, 4 mbar (approximately 3 mm of Hg) and backfill the cask cavity with helium (99.9% minimum purity) to 0 psig without allowing air to re-enter the cask. Disconnect the vacuum pump and helium supply from the vent port.
- 22. Remove the metallic O-rings in the vent port coverplate and clean and inspect the O-ring groove. Install new metallic O-rings in the vent port coverplate and install the coverplate.
Torque the coverplate bolts to the value specified in Table 7-1.
- 23. Connect the leak detector to the inner lid interseal test port and evacuate the air between the metallic O-rings until a pressure of <1 mbar is reached. Using the helium leak detector, verify that any detectable leak rate is 2 x 10-7 cm3/sec (helium). The test sensitivity shall be 1 x 10-7 cm3/sec (helium).
Note: See Table 7.4-1 for details on required containment boundary seals leakage test requirements and allowable leakage rates.
- 24. Install the test port plug for the inner lid interseal test port using a new metallic O-ring and torque the plug to the value specified in Table 7-1.
- 25. Connect the leak detector to the vent port coverplate interseal test port. Evacuate the interseal volume until a pressure of <1 mbar is reached. Using the helium leak detector, verify that any detectable leak rate is 2 x 10-7 cm3/sec (helium). The test sensitivity shall be 1 x 10-7 cm3/sec (helium).
- 26. Install the test port plug for the vent port coverplate using a new metallic O-ring and torque the plug to the value specified in Table 7-1.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.1-17
- 27. Repeat Steps 25 and 26 for the drain port coverplate test port.
- 28. Remove the outer lid metallic O-ring. Clean the outer lid O-ring seating surface and groove. Install a new metallic outer lid O-ring. Install the outer lid alignment pins.
- 29. Attach the outer lid lifting device to the outer lid and cask handling crane. Install the outer lid using the alignment pins to assist in proper seating. Remove the outer lid alignment pins. Install the outer lid bolts and torque to the value specified in Table 7-1.
The bolt torquing sequence is shown on the outer lid.
- 30. Attach a supply of air, nitrogen, or helium to the interlid port quick-disconnect and backfill the interlid volume to 15 psig air, nitrogen, or helium and hold for 10 minutes.
No loss of pressure is permitted during the 10-minute test period. Disconnect air, nitrogen, or helium supply.
- 31. Install the transport interlid port cover in the interlid port using new O-rings. Torque the interlid port cover bolts to the value specified in Table 7-1.
- 32. Remove the O-ring test plug from the interlid port cover and, using the O-ring test fixture, pressurize the O-ring annulus to 15 psig with air, nitrogen, or helium. Isolate the annulus and hold for 10 minutes. No loss of pressure is permitted during the test period.
- 33. Vent the annulus pressure, remove the air, nitrogen, or helium supply, replace the metallic O-ring on the interlid port cover test plug and install the test plug. Torque the plug to the value specified in Table 7-1.
- 34. Perform final external decontamination and perform survey to verify acceptable level of removable contamination to ensure compliance with 49 CFR 173.443. Perform final radiation survey. Record the survey results in the cask loading report.
- 35. Perform final visual inspection to verify assembly of the NAC-STC in accordance with the CoC. Verify that the loading procedure and checklist are appropriately completed and signed off.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.4-2 and nonmetallic O-rings have different allowable fabrication, maintenance, periodic, and preshipment leakage rates, as specified in the procedures and summarized in Table 7.4-1.
7.4.1 Fabrication, Maintenance and Periodic Leakage Rate Test Procedures As described in Chapter 4, the NAC-STC primary containment boundary is designed and tested to assure that there is no leakage under any of the normal conditions of transport or accident conditions that exceeds the allowable value determined in accordance with 10 CFR 71.51. This leakage rate is verified prior to transport by the performance of leak tests on the containment boundary sealed with metallic O-rings to ensure that the leakage rate is less than 2 x 10-7 cm3/sec (helium). For NAC-STC intended to transport PWR fuel assemblies sealed with Viton O-rings, the periodic leakage rate test is performed annually or after replacement of a Viton O-ring, and the cumulative leakage rate is less than that shown in Table 7.4-1. Helium leakage test procedures shall be prepared by qualified personnel, and approved by personnel qualified in accordance with the requirements of SNT-TC-1A as a NDT Level III (Leak Testing). Leak tests shall be performed by personnel qualified for helium leakage testing in accordance with the requirements of ANSI/ASNT CP-189-2006, Standard for Qualification and Certification of Nondestructive Testing Personnel.
As described in Section 4.1, the containment boundary is defined differently for transport after long-term storage than for loading for transport without interim storage. As described in this section, leakage tests are performed in accordance with the requirements of ANSI N14.5-1997.
The leakage test requirements and acceptance criteria performed after long-term storage in preparation for transport and performed following cask loading operations for transport without interim storage are described in Sections 7.4.2 and 7.4.3, respectively. The generic procedures used to perform leakage testing are incorporated in the NAC-STC loading procedures in Section 7.2. Detailed helium leakage test procedures describing the equipment and the leak test system used to perform the leakage tests are developed and prepared by qualified personnel and approved by personnel qualified in accordance with the requirements of SNT-TC-1A as a NDT Level III (Leak Testing). As noted in Section 7.1, the GTCC Waste canister or HLW overpack will have been loaded, closed and sealed prior to loading into the NAC-STC. The transportable storage canister for spent fuel is a separate inner container for the transport of damaged fuel.
GTCC Waste canisters and HLW overpacks are not qualified as separate inner containers for confinement/containment purposes.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.4-3 Section 7.4.4 provides the procedural guidance on corrective actions to be taken in the event a leakage test does not meet the acceptance criteria.
7.4.2 Leak Testing for Transport After Long-Term Storage This section summarizes the leak test methods used to demonstrate continued containment of PWR spent fuel prior to transport following an extended period of storage in the NAC-STC cask.
The containment boundary for this transport condition is defined as Containment Condition A in Section 4.1 and requires the use of metallic O-rings in the containment boundary. In addition to the steel inner lid and port coverplates, the containment boundary is specified as the outer O-rings of the inner lid and of the vent and drain port coverplates and the O-rings of the test port plugs. As specified in the generic loading procedure, the outer lid must be removed to test the inner lid and the vent and drain port coverplates prior to transport. Note that HBU spent fuel assemblies, canistered spent fuel, GTCC Waste canisters, or HLW Overpacks are not loaded into NAC-STC casks for long-term storage, and therefore, Containment Condition A does not apply for these directly loaded or canistered contents.
To conduct the periodic leakage rate test, the inner seal regions (annulus between the O-rings) of the inner lid and the vent and drain port coverplates are evacuated to less than one millibar (mbar), and backfilled to 0 psig with 99.9% pure helium, and the test port plugs are reinstalled.
The outer lid is reinstalled using a new metallic O-ring. The interlid region (between the inner and outer lids) is evacuated to a vacuum of 1 mbar. After the vacuum condition is reached, a helium leak detector is used to sample the interlid region for helium leakage past the inner lid outer O-ring, the vent and drain port coverplate outer O-rings, and O-ring test port plugs. The allowable leak rate is 2 x 10-7 cm3/sec (helium) with a minimum test sensitivity of 1 x 10-7 cm3/sec (helium). This test method conforms to A5.4 (evacuated envelope) of Appendix A of ANSI N14.5-1997. If helium leakage is detected exceeding the leaktight criteria, corrective action is taken as described in Section 7.4.4.
The outer lid and pressure port are tested using a pressure drop method to confirm the proper installation of the outer lid and pressure port O-rings. The interlid region is pressurized using the interlid port to 15 psig with air or helium, and the pressure is held for 10 minutes. No loss of pressure is permitted during the test period. Following the test, the interlid region pressure is reduced to 0 psig. The interlid port cover is installed and the annulus between the O-rings of the port cover is tested using the same method. This test confirms the proper installation of the
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.4-4 interlid port cover O-rings and conforms to test method A.5.1 (gas pressure drop) of Appendix A of ANSI N14.5-1997.
7Property "ANSI code" (as page type) with input value "ANSI N14.5-1997.</br></br>7" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..4.3 Leak Testing for Transport After Loading without Interim Storage This section summarizes the leakage tests required to demonstrate containment of directly loaded standard PWR and HBU spent fuel without interim storage, and for sealed transportable storage canisters containing spent fuel or GTCC waste, or HLW Overpacks containing separate sealed canisters of glassified high level waste. The containment boundary for these transport conditions is defined as Containment Condition B1, B2 and B3 based on contents and type and material of containment boundary seals as defined in Section 4.1, Table 4.1-1 and Table 7.4-1. In addition to the steel inner lid and port coverplates, the containment boundary is specified as the inner O-rings of the inner lid and of the vent and drain port coverplates. The metallic inner lid O-ring and vent and drain port coverplate O-rings are maintenance leakage rate tested using the evacuated envelope method (test description A5.4 of Appendix A of ANSI N14.5-1997) with a vacuum in the annulus between the O-ringsas the metallic seals are replaced for each loading operation. The Viton inner lid O-ring and vent and drain port coverplate O-rings are preshipment leakage rate tested using a pressure drop method to confirm the proper installation of the inner lid, and vent and drain port O-rings. The interlid region of each component is pressurized using the interseal test port to 15 psig with helium, and the pressure is held for 15 minutes. No loss of pressure is permitted during the test period and the preshipment leakage rate test confirms that there is no detected leakage from any seal to a minimum sensitivity of 1 x 10-3 refcm3/sec. Following the test, the interseal port plugs are reinstalled. This test confirms the proper installation of the inner lid, and vent and drain port cover Viton O-rings and conforms to test method A.5.1 (gas pressure drop) of Appendix A of ANSI N14.5-1997.
The containment boundary O-rings for standard PWR spent fuel assemblies with burnups 45 GWd/MTU and HBU PWR spent fuel assemblies with burnups > 45 GWd/MTU directly loaded for transport without interim storage may be either metallic or Viton. The containment boundary seals for HBU PWR spent fuel assemblies with burnups 45 GWd/MTU and directly loaded for transport are required to be Viton O-rings. The containment boundary seals for HBU PWR spent fuel assemblies with burnups > 45 GWd/MTU and directly loaded for transport may be either Viton O-rings or metallic seals. The containment boundary O-rings for canistered spent fuel assemblies or GTCC waste, or HLW Overpacks are required to be metallic O-rings. Metallic O-ring containment seals are tested with a helium MSLD connected to the interseal test ports of the inner lid, and vent and drain port coverplates to detect helium in the annulus between the O-rings. The
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.4-5 allowable maintenance leakage rate for each metallic O-ring for NAC-STC casks configured with metallic seals defined as the containment boundary is 2 x 10-7 cm3/sec (helium) with a minimum test sensitivity of 1 x 10-7 cm3/sec (helium). For NAC-STC casks loaded with PWR spent fuel assemblies for immediate transport and provided with Viton containment O-rings, the periodic and maintenance leakage rates are a cumulative leakage rate for all containment closure components is as shown in Table 7.4-1.
The preshipment leakage rate test prior to transport of a NAC-STC cask utilizing reusable containment boundary Viton O-rings that have not been replaced is performed to a minimum test sensitivity of 1 x 10-3 refcm3/sec. The higher sensitivity maintenance leakage rate test for the Viton O-rings (as shown in Table 7.4-1) shall be performed during annual periodic testing and for maintenance leakage rate tests after replacement of the Viton O-rings or when the Viton O-rings or other containment components are replaced between annual maintenance periods during operations.
For NAC-STC casks provided with metallic O-rings as the containment boundary, the metallic seals are replaced for each loaded transport, and the maintenance leakage rate test is performed to an acceptance criteria of 2 x 10-7 cm3/sec (helium) with a minimum test sensitivity of 1 x 10-7 cm3/sec (helium).
The series of leakage rate tests described and detailed in Table 7.4-1 confirms that the allowable leakage test rates are satisfied for the types of O-rings used in the containment boundary closure components for Containment Conditions B1, B2 and B3. Section 7.4.4 provides the procedural guidance on corrective actions to be taken in the event a leak test does not meet the acceptance criteria.
Following completion of NAC-STC leakage testing of the containment boundaries, the outer lid and pressure port are tested using a pressure drop method to confirm the installation of the outer lid and pressure port O-rings. The interlid region is pressurized using the interlid port to 15 psig with air and the pressure is held for a minimum of 10 minutes. No loss of pressure is permitted during the test period. Following the test, the interlid region pressure is reduced to 0 psig. The interlid port cover is installed and the annulus between the O-rings of the port cover is tested using the same method. This test confirms the installation of the interlid port cover O-rings.
These components provide an additional barrier against the release of radioactive material but are not part of the containment boundary.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.4-6 7.4.4 Corrective Action If a specific containment component containing an O-ring fails to meet the leakage test acceptance criteria for that component, the component is removed and the O-ring is removed from the groove. The O-ring groove is cleaned and visually inspected to ensure proper cleanliness and surface condition. A new O-ring of the appropriate material, size and identification (i.e., metallic or Viton) is installed. The removed component is reinstalled and the bolts torqued to the appropriate torque value. The component is then retested in accordance with the applicable original test procedure and acceptance criteria. Replacement of a containment boundary Viton O-ring due to failure of the preshipment leakage rate test shall require O-ring replacement and performance of the maintenance leakage rate test to the applicable leakage rate acceptance criteria (see Table 7.4-1).
The replacement of the inner lid O-ring(s) due to a failure of a leakage test, either immediately after loading or after extended storage for the directly loaded spent fuel configurations, will require the NAC-STC cask to be returned to the spent fuel pool for inner lid removal and inner lid O-ring replacement. The cask unloading procedures (Section 7.3.3) will be utilized to prepare the cask for placement in the spent fuel pool, including cask cool down operations, to allow inner lid removal for seal replacement. At cask storage facilities having appropriate dry transfer or hot cell facilities, the inner lid O-ring can be replaced without placement of the cask in a fuel pool for shielding purposes. Prior to removal of the inner lid, a gas sample should be taken at the vent port to verify the condition in the cavity environment. If there are indications that fuel has failed during the storage period, care should be exercised in both flooding the cask and in removing the inner lid (refer to procedures in Section 7.3.2.1, Steps 4.f. and.g.).
For NAC-STC casks being loaded with canister spent fuel, GTCC Waste canisters or HLW Overpacks, the containment boundary seals are metallic seals, and the inner lid metallic seals may be replaced without returning the cask to the pool as the canister or overpack confines and shields the spent fuel, GTCC waste, or HLW.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.4-7 Table 7.4-1 NAC-STC Containment Boundary Leakage Testing Requirements Containment Condition Content Condition Allowable Fabrication, Maintenance and Periodic Leakage Test Rate/Sensitivity Preshipment Leak Test Rate/Sensitivity Fabrication, Maintenance and Periodic Leakage Rate Test Location/Method Preshipment Leakage Test Location/Method A
Using double metallic O-rings Up to 26 directly loaded intact PWR spent fuel assemblies following storage operations per 10 CFR 72 having burnups of 45,000 MWd/MTU Allowable leakage rate is < 2 x 10-7 cm3/sec (helium)
(i.e., leaktight)
Minimum test sensitivity is < 1 x 10-7 cm3/sec (helium)
Allowable leakage rate is
< 2 x 10-7 cm3/sec (helium)
(i.e., leaktight)
Minimum test sensitivity is <
1 x 10-7 cm3/sec (helium)
Leakage tests performed on both inner and outer inner lid and vent and drain port cover O-rings using evacuated envelope methods: envelope provided by outer lid with test performed through the interlid port with helium in interseal regions of inner lid and vent and drain coverplate O-rings for outer seals, or by interseal volume of inner and outer O-ring seals with helium in the cask cavity.
Testing is performed in accordance with ANSI N14.5 requirements prior to re-use of transport cask.
Evacuated envelope method (envelope provided by outer lid with test performed through the interlid port) with helium in interseal regions of inner lid and vent and drain coverplate O-rings.
These series of leakage tests are performed on the NAC-STC containment boundary following directly loaded fuel storage operations.
The outer O-rings are the designated boundary as access to the cask cavity to verify helium backfill conditions is not planned.
Testing is performed in accordance with ANSI N14.5 requirements immediately prior to transport.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.4-8 Table 7.4-1 NAC-STC Containment Boundary Leakage Testing Requirements (Continued)
Containment Condition Content Condition Allowable Fabrication, Maintenance and Periodic Leakage Test Rate/Sensitivity Preshipment Leak Test Rate/Sensitivity Fabrication, Maintenance and Periodic Leakage Rate Test Location/Method Preshipment Leakage Test Location/Method B1 Using double metallic O-rings Up to 26 directly loaded intact/undamaged PWR spent fuel assemblies having burnups of 45,000 MWd/MTU for immediate transport, or canistered Yankee Class or Connecticut Yankee or MPC-LACBWR spent fuel assemblies, Reconfigured Fuel Assemblies, Damaged Fuel Cans or GTCC waste and MPC-WVDP HLW overpack Allowable leakage rate is < 2 x 10-7 cm3/s (helium)
(i.e., leaktight) for each component test\\
Minimum test sensitivity is
< 1 x 10-7 cm3/s (helium)
Allowable leakage rate is < 2 x 10-7 cm3/s (helium)
(i.e., leaktight) for each component test Minimum test sensitivity is
< 1 x 10-7 cm3/s (helium)
Evacuated envelope test of the inner lid inner O-ring, and the vent and drain coverplates inner O-rings with atmospheric helium in the cask cavity.
Test envelope provided by the inner and outer O-rings of the inner lid and vent and drain port coverplates.
Testing is performed in accordance with ANSI N14.5 requirements.
Evacuated envelope test of the inner lid inner O-ring, and the vent and drain coverplates inner O-rings with atmospheric helium in the cask cavity.
Test envelope provided by the inner and outer O-rings of the inner lid and vent and drain port coverplates.
Testing is performed in accordance with ANSI N14.5 requirements.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.4-9 Table 7.4-1 NAC-STC Containment Boundary Leakage Testing Requirements (Continued)
Containment Condition Content Condition Allowable Fabrication, Maintenance and Periodic Leakage Test Rate/Sensitivity Preshipment Leak Test Rate/Sensitivity Fabrication, Maintenance and Periodic Leakage Rate Test Location/Method Preshipment Leakage Test Location/Method B2 Using either double non-metallic (e.g.,
Viton) O-rings or inner metallic seal with outer Viton seal Up to 26 directly loaded intact/undamag ed PWR spent fuel assemblies having burnups of 45,000 MWd/MTU for immediate transport.
Either For double Viton O-ring seals:
Allowable leakage rate is
< 1.1 x10-4 cm3/s (helium) for the total of the three leakage tests Minimum test sensitivity is < 5.7 x 10-5 cm3/s (helium) for double Viton O-ring seals; Or, For inner metallic seals:
Allowable leakage rate is
< 2 x 10-7 cm3/sec (helium) (i.e., leaktight)
Minimum test sensitivity is < 1 x 10-7 cm3/sec (helium)
Either For double Viton O-ring seals:
No detected leakage when tested to a sensitivity of 10-3 ref.cm3/s for double Viton O-ring seals that have not been replaced during loading operations; Or Allowable leakage rate is
< 1.1 x10-4 cm3/s (helium) for the total of the three leakage tests Minimum test sensitivity is <
5.7 x 10-5 cm3/s (helium);
For any replaced inner metallic seals:
2 x 10-7 cm3/sec (helium) at a minimum sensitivity of < 1 x 10-7 cm3/sec (helium)
Evacuated envelope test of the inner lid inner O-ring, and the vent and drain coverplates inner O-rings with atmospheric helium in the cask cavity.
Test envelope provided by the inner and outer O-rings of the inner lid and vent and drain port coverplates.
Testing is performed in accordance with ANSI N14.5 requirements.
Viton O-rings tested with a gas pressure drop test for a minimum of 15 minutes applied to the interseal test ports of the inner lid, and vent and drain port coverplates; Metallic inner seal or replaced Viton O-rings to Maintenance Leakage Test requirements with evacuated envelope.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 7.4-10 Table 7.4-1 NAC-STC Containment Boundary Leakage Testing Requirements (Continued)
Containment Condition Content Condition Allowable Fabrication, Maintenance and Periodic Leakage Test Rate/Sensitivity Preshipment Leak Test Rate/Sensitivity Fabrication, Maintenance and Periodic Leakage Rate Test Location/Method Preshipment Leakage Test Location/Method B3 Using double non-metallic (e.g., Viton)
O-rings or inner metallic seals and outer Viton O-rings Up to 20 directly loaded intact/undamag ed HBU PWR spent fuel assemblies with burnups of 55,000 MWd/MTU for immediate transport.
Either For double Viton O-ring seals:
Allowable leakage rate is
< 4.0 x10-6 cm3/s (helium) for the total of the three leakage tests Minimum test sensitivity is < 2.0 x 10-6 cm3/s (helium) for double Viton O-ring seals; Or, For inner metallic seals:
Allowable leakage rate is 2 x 10-7 cm3/sec (helium) (i.e., leaktight)
Minimum test sensitivity is 1 x 10-7 cm3/sec (helium)
Either:
For double Viton O-ring seals:
No detected leakage when tested to a sensitivity of 10-3 ref. cm3/s for double Viton O-ring seals that have not been replaced during loading operations; Or Allowable leakage rate is
< 4.0 x10-6 cm3/s (helium) for the total of the three leakage tests Minimum test sensitivity is <
2.0 x 10-6 cm3/s (helium);
For any replaced inner metallic seals:
2 x 10-7 cm3/sec (helium) at a minimum sensitivity of < 1 x 10-7 cm3/sec (helium)
Evacuated envelope test of the inner lid inner O-ring, and the vent and drain coverplates inner O-rings with atmospheric helium in the cask cavity.
Test envelope provided by the inner and outer O-rings of the inner lid and vent and drain port coverplates.
Testing is performed in accordance with ANSI N14.5 requirements.
Viton O-rings tested with a gas pressure drop test for a minimum of 15 minutes applied to the interseal test ports of the inner lid, and vent and drain port coverplates; Metallic inner seal or replaced Viton O-rings to Maintenance Leakage Test requirements with evacuated envelope.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 8.1-12 8.1.5 Tests for Shielding Integrity 8.1.5.1 Gamma Shield Test The gamma scan test shall be conducted by continuous scanning or probing over 100 percent of all accessible cask body surfaces, which directly shield regions where lead was poured, using a detector and a 60Co source. Accessible cask surfaces are not only those surfaces that are physically accessible but also cask surfaces where accurate detector readings can occur. The source strength shall be of an intensity sufficient to produce a count rate that equals or exceeds three times the background count rate on the external surfaces of the cask. The count rate shall be maintained for greater than one minute prior to the start of scanning. The detector scan path spacing (cask body exterior surface) will be sufficiently small such that there will be scanning overlap based on the size detector used and the scanning speed will be 4.5 feet per minute or less. The source scan path spacing (cask interior surface) will be on a sized grid pattern that is sufficiently small such that scanning overlap will occur based on the size detector used.
A gamma scan test is not required for the cask inner closure lid, cask outer closure lid, cask inner bottom forging, cask outer bottom forging, or cask outer bottom plate. These components shall be ultrasonic tested in order to demonstrate their soundness as gamma shielding. Ultrasonic testing shall be performed per ASME B&PV NB-2542.1 using the acceptance standards of Section NB-2542.2 for forgings and ASME B&PV NB-2532.1 using the acceptance standards of NB-2532.1(b) for plates.
The acceptance criteria for the cask body shield test shall be that the shield effectiveness of the cask body is equal to or greater than the shield effectiveness of a lead and steel mock-up. The steel thickness of the mockup shall be equivalent to the minimum steel thickness specified on the License Drawings and the lead thickness shall be equivalent to the minimum lead thickness specified in the License Drawings. The shielding mock-up will be produced using the same fabrication techniques as those approved for the cask.
Measured count rates that exceed those established by the test mock-up shall cause the component to be rejected, with exception to the alternate acceptance criteria as specified in Section 8.1.5.1.1 for the Fiberfrax region of the lead. The rejected areas/components shall be evaluated to determine the corrective action to be taken. Any repaired areas shall be retested prior to acceptance.
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 8.1-13 The gamma scan tests ensures that there are no gross voids in the poured lead and that the minimum required lead thickness has been achieved and thus ensures that the radiation dose limits are not exceeded for each cask.
8.1.5.1.1 Alternate Acceptance Criteria (Fiberfrax Region)
In the region of the Fiberfrax insulation shown in Drawing 823-802, Detail G-G, defined as the top 10.18 inches of the cask radial lead shield, an increased count rate equivalent to a reduction in lead thickness of 0.65 inches is acceptable for directly loaded fuel shipments. However, the cask shall be used with the shield plate on the basket top weldment as detailed in drawing 423-872, which compensates for this reduced lead shielding.
8.1.5.2 Neutron Shielding Test The neutron shielding of the NAC-STC is provided by a solid layer of NS-4-FR, which is a hard polymer material. A 5.5-inch layer of NS-4-FR is located in the annulus formed by the outer shell and the 0.236-inch (6 mm) thick neutron shield shell. The neutron shield is divided in sections by the copper/stainless steel fins. A 2-inch thick layer of NS-4-FR is also installed in the cask inner lid and in the cask bottom.
NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 8.1-14 The neutron shielding material is installed into the annulus between the outer shell and the neutron shield shell by pouring it with the cask in an inverted vertical position. Procedures used for installation of the material are validated prior to use by destructive examination of a full scale mock-up of the neutron shield cavity. Qualification of the installation procedure verifies material homogeneous properties and minimizes the potential deleterious voids.
Dimensional inspection of the cavities containing the neutron shielding material shall ensure that the required thickness specified in the License Drawings is incorporated into the cask.
The installation of the neutron shielding material shall be performed in accordance with written, approved, and qualified procedures. The procedures shall ensure that mix ratios and mixing NAC PROPRIETARY INFORMATION REMOVED
NAC-STC SAR July 2024 Docket No. 71-9235 Revision 24A 8.1-15 methods are controlled in order to achieve proper material composition, boron concentration and distribution, and that pours are controlled in order to prevent gaps or unacceptable voids from occurring in the material. Procedures shall be qualified by the use of mock-ups to ensure that the NS-4-FR installation does not result in the creation of unacceptable voids. Wet density data for each mix of installed neutron shield material shall be maintained as part of the quality record documentation package.
The above tests are performed during fabrication of the cask and ensure that the radiation dose limits are not exceeded for each cask.
8.1.6 Thermal Test Prior to acceptance at the factory, a thermal test shall be performed on each fabricated packaging to confirm and verify that the fabricated and assembled cask possesses the heat rejection capabilities predicted by the thermal analyses. The thermal test shall be performed in accordance with approved written procedures.
8.1.6.1 Thermal Test Set-up The thermal test set-up is shown in Figure 8.1-1(a). As depicted, the thermal test shall be performed with the cask positioned horizontally on a test frame. The transport impact limiter or equivalent insulating material shall be installed on each end of the cask to simulate the transport configuration. The cask will be located in a covered building in a still environment. The cask shall be assembled with the basket installed. A thermal test lid with connections for thermocouple leads and electric heater power cables shall be installed in place of the inner lid.
The outer lid will not be installed for the test. The thermal test lid will be provided with an O-ring seal capable of containing the containment cavity helium atmosphere.
Electric heaters shall be installed in each fuel tube. The electric heaters will have an active length of between 120 and 150 inches and be capable of generating a minimum of 22 kilowatts (kw).
The heaters will be supported in the basket so as to not be in contact with the wall of the fuel tube. The power supplied to the heater will be recorded throughout the test duration.