ML23156A374
| ML23156A374 | |
| Person / Time | |
|---|---|
| Issue date: | 06/02/1994 |
| From: | Thomas Taylor NRC/EDO |
| To: | |
| References | |
| PR-072, 59FR28496 | |
| Download: ML23156A374 (1) | |
Text
{{#Wiki_filter:ADAMS Template: SECY-067 DOCUMENT DATE: 06/02/1994 TITLE: PR-072 - 59FR28496 - LIST OF APPROVED SPENT FUEL STORAGE CASKS ADDITION : CASE
REFERENCE:
PR-072 59FR28496 KEYWORD: RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete
STATUS OP RULEMAKING PROPOSED RULE: PR-072 OPEN ITEM (Y/N) N RULE NAME: LIST OF APPROVED SPENT FUEL STORAGE CASKS: ADDITION PROPOSED RULE FED REG CITE: 59FR28496 PROPOSED RULE PUBLICATION DATE: 06/02/94 NUMBER OF COMMENTS: 27 ORIGINAL DATE FOR COMMENTS: 08/16/94 EXTENSION DATE: 09/30/94 FINAL RULE FED. REG. CITE: 59FR65898 FINAL RULE PUBLICATION DATE: 12/22/94 NOTES ON EXTENDED BY FED. REG. NOTICE PUBLISHED ON 8/29/94 BY 59 FR 44381. TATUS FINAL RULE SIGNED BY EDO. FILE LOCATED ON 16-G. F RULE TO FIND THE STAFF CONTACT OR VIEW THE RULEMAKING HISTORY PRESS PAGE DOWN KEY HISTORY OF THE RULE PART AFFECTED: PR-072 RULE TITLE: LIST OF APPROVED SPENT FUEL STORAGE CASKS: ADDITION ROPOSED RULE PROPOSED RULE DATE PROPOSED RULE SECY PAPER: SRM DATE: I I SIGNED BY SECRETARY: 05/12/94 FINAL RULE FINAL RULE DATE FINAL RULE SECY PAPER: SRM DATE: I I SIGNED BY SECRETARY: 12/15/94 STAFF CONTACTS ON THE RULE CONTACT1: G. E. GUNDERSEN MAIL STOP: T-9F33 PHONE: 415-6195 CONTACT2: K. C. LEU MAIL STOP: PHONE: 504-2685
DOCKET NO. PR-072 (59FR28496) In the Matter of LIST OF APPROVED SPENT FUEL STORAGE CASKS : ADDITION DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 07/18/94 07/ 13/94 COMMENT OF OHIO CITIZENS FOR RESPONSIBLE ENERGY (SUSAN L. HIATT, DIRECTOR) ( 1) 07/28/94 07/21/94 COMMENT OF N. J. DEPT. OF ENVIRONMENTAL PROTECTION (DR. JILL LIPOTI) ( 2) 08/01/94 07/27/94 COMMENT OF JOHN KIELY ( 3) 08/04/94 08/01/94 COMMENT OF DORIS G. KELLER ( 4) 08/05/94 07/ 15/94 COMMENT OF JOHN TRAPP ( 5) 08/09/94 08/01/94 COMMENT OF VISTULA MANAGEMENT COMPANY (WILLIAM HIRT, PRESIDENT) ( 6) 08/11/94 08/09/94 COMMENT OF SIERRA CLUB (CONNIE KLINE) ( 7) 08/12/94 08/11/94 REQUEST FROMCONNIE KLINE SUBMITTED TO CHARLES J. HAUGHNEY, NRC FOR A EXTENSION OF THE COMMENT PERIOD TO SEPTEMBER 30, 1994 08/12/94 08/09/94 COMMENT OF FAWN SHILLINGLAW ( 8) 08/15/94 08/07/94 COMMENT OF ALICE H. HIRT ( 9) 08/15/94 08/12/94 COMMENT OF CENTERIOR ENERGY (JOHN P. STETZ, V. P. ) ( 10) 08/16/94 08/15/94 COMMENT OF STATE OF OHIO AND OHIO CITIZENS (LEE FISHER, ATTORNEY GENERAL) ( 11) 08/16/94 08/15/94 COMMENT OF UTILITY RADIOLOGICAL SAFETY BD . OF OHIO (EDITH A. BINFORD, SECRETARY) ( 12) 08/16/94 08/15/94 COMMENT OF CITIZENS' UTILITY BOARD (DENNIS DUMS, RESEARCH DIRECTOR) ( 13) 08/16/94 08/15/94 COMMENT OF DUKE POWER COMPANY (GARY R. WALDEN) ( 14)
DOCKET NO. PR-072 (59FR28496) DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 08/17/94 08/16/94 COMMENT OF OHIO DEPARTMENT OF HEALTH (ROBERT E. OWEN) ( 15) 08/17/94 08/16/94 COMMENT OF U.S. ENVIRONMENTAL PROTECTION AGENCY (SHIRLEY MITCHELL) ( 16) 08/17/94 08/10/94 COMMENT OF COALITION FOR SAFE ENERGY (TERRY JONATHAN LODGE) ( 17) 08/17/94 08/14/94 COMMENT OF ROBB. SMITH ( 19) 08/18/94 08/16/94 COMMENT OF VECTRA TECHNOLOGIES, INC. (MOSES TAYLOR l JAMES W. AXLINE) ( 20) 08/18/94 08/17/94 COMMENT OF NUCLEAR ENERGY INSTITUTE (JOHN F. SCHMITT) ( 21) 08/18/94 08/10/94 REQUEST FOR EXTENSION OF COMMENT PERIOD SUBMITTED TO THE SECRETARY BY CONNIE KLINE REPRESENTING THE SIERRA CLUB (PREVIOUSLY FAXED ON 8/11/94) 08/22/94 08/15/94 COMMENT OF COMMONWEALTH EDISON (WILLIAM F. NAUGHTON) ( 18) 08/22/94 08/13/94 COMMENT OF CHARLENE JOHNSTON ( 22) 08/22/94 08/16/94 COMMENT OF COALITION FOR A NUCLEAR FREE GREAT LAKES (MICHAEL J. KEEGAN, CHAIRPERSON) ( 23) 08/29/94 08/19/94 LETTER FROM ERIC BECKJORD, DIRECTOR, RES TO CONNIE KLINE, SIERRA CLUB ADVISING THAT THE COMMENT PERIOD FOR THE RULE IS EXTENDED TO SEPT. 30, 1994 08/29/94 08/22/94 FEDERAL REGISTER NOTICE EXTENDING THE COMMENT PERIOD FOR THE PROPOSED RULE TO SEPTEMBER 30, 1994 09/28/94 08/16/94 COMMENT OF FAWN SHILLINGLAW ( 24) 09/28/94 09/09/94 LTR FM BERNERO, NMSS, NRC TO FAWN SHILLINGLAW ACKNOWLEDGING LTRS OF JUNE 27, JULY 15, AND AUGUST 16, 1994 09/29/94 09/25/94 COMMENT OF DINI SCHUT ( 25) 09/30/94 09/29/94 COMMENT OF OYSTER CREEK NUCLEAR WATCH (WILLIAM DECAMP JR.) ( 26) 09/30/94 09/23/94 COMMENT OF CITY OF SYLVANIA (MARGARETT. RAUCH) ( 27) 09/30/94 09/30/94 SUPPLEMENTAL COMMENTS OF TERRY J. LODGE AND TOLEDO COALITION FOR SAFE ENERGY
DOCKET NO. PR-072 {59FR28496) DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 11/29/94 11/23/94 LITTER FROM ROBERT E. OWEN REPRESENTING THE OHIO DEPARTMENT OF HEALTH AND REQUESTING WITHDRAWAL OF COMMENT NO . 15. 12/15/94 12/15/94 FINAL RULE PUBLISHED ON 12/22/94 AT 59FR65898. 12/16/94 12/16/94 MEMORANDUM FROM C. J. HAUGHNEY TOE. L. JULIAN TRANSMITTING A DECEMBER 1994 SAFETY EVALUATION REPORT AND CERTIFICATE OF COMPLIANCE NO. 1004
DOCKET NUMBEH PROPOSED RULE PR *7:2.._ UNITED STATES ( 'q FR 2. f-- Lf q ' J NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 16, 1994 MEMORANDUM TO: Emile L. Julian, Chief OOCV :°: T .:n Docketing and Services Branch us1,,1o._; Office of the Secretary FROM: Charles J. Haughney, Chief *94 0[ 16 P 3 :4 9 Storage and Transport Systems Branch Division of Industrial and Medical Nuclear Safety on.~-_
- DOCK£ !.
SUBJECT:
DOCUMENTS TO BE INCLUDED AS PART OF RULEMAKING 10 CFR PART 72 (59 FR 28496) I am enclosing a copy of the staff's "Safety Evaluation Report of VECTRA Technologies, Inc., a.k.a. Pacific Nuclear Fuel Services, Inc., Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel," dated December 1994, and a copy of the unsigned Certificate of Compliance No. 1004 for the Standardized NUHOMS-24P and
- NUHOMS-52B. These documents should be made available at NRC's public and local document rooms as part of the 10 CFR Part 72 rulemaking (59 FR 28496).
My staff is also forwarding these documents to Tyrone Greene for docketing under Docket 72-1004. The Final Rule "10 CFR Part 72, List of Approved Spent Fuel Storage Casks: Addition" has been forwarded to the Office of the Federal Register for publication. The effective date of the Rule and Certificate of Compliance No. 1004 will be 30 days from the date of publication of the Federal Register notice. ~ * """.__ Charles J. Haugd, Chie Storage and Transport Sys Division of Industrial and Medical Nuclear Safety
Enclosures:
- l. Safety Evaluation Report dated December 1994
- 2. Unsigned Certificate of Compl iance No. 1004 59 FR 28496 Docket No. 72-1004
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20e5e-0001 December 16, 1994 MEMORANDUM TO: Emile L. Julian, Chief Docketing and Services Branch Office of the Secretary FROM: Charles J. Haughney, Chief Storage and Transport Systems Branch Division of Industrial and Medical Nuclear Safety
SUBJECT:
DOCUMENTS TO BE INCLUDED AS PART OF RULEMAKING 10 CFR PART 72 (59 FR 28496} I am enclosing a copy of the staff's "Safety Evaluation Report of VECTRA Technologies, Inc., a.k.a. Pacific Nuclear Fuel Services, Inc., Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel," dated December 1994, and a copy of the unsigned Certificate of Compliance No. 1004 for the Standardized NUHOMS-24P and NUHOMS-52B. These documents should be made available at NRC's public and local document rooms as part of the 10 CFR Part 72 rulemaking {59 FR 28496). My staff is also forwarding these documents to Tyrone Greene for docketing under Docket 72-1004. The Final Rule "10 CFR Part 72, List of Approved Spent Fuel Storage Casks: Addition" has been fo-rwarded to the Office of the Federal Register for publication. The effective date of the Rule and Certificate of Compliance No. 1004 will be 30 days from the date of publication of the Federal Register notice. ~ 11 Charles Haugd: J. Ch,,.ie--~-- Storage and Transport Sys Division of Industrial and Medical Nuclear Safety
Enclosures:
- 1. Safety Evaluation Report dated December 1994
- 2. Unsigned Certificate of Compliance No. 1004 59 FR 28496 Docket No. 72-1004
December 16, 1994 MEMORANDUM TO: Emile L. Julian, Chief Docketing and Services Branch Office of the Secretary FROM: Charles J. Haughney, Chief Storage and Transport Systems Branch Division of Industrial and Medical Nuclear Safety
SUBJECT:
DOCUMENTS TO BE INCLUDED AS PART OF RULEMAKING 10 CFR PART 72 (59 FR 28496) I am enclosing a copy of the staff's "Safety Evaluation Report of VECTRA Technologies, Inc., a.k.a. Pacific Nuclear Fuel Services, Inc., Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel," dated December 1994, and a copy of the unsigned Certificate of Compliance No. 1004 for the Standardized NUHOMS-24P and NUHOMS-528. These documents should be made avail~ble at NRC's public and local document rooms as part of the 10 CFR Part 72 rulemaking (59 FR 28496). My staff is also forwarding these documents to Tyrone Greene for docketing under Docket 72-1004. The Final Rule 11 10 CFR Part 72, List of Approved Spent Fuel Storage Casks: Additionn has been forwarded to the Office of the Federal Register for publication. The effective date of the Rule and Certificate of Compliance No. 1004 will be 30 days from the date of publication of the Federal Register notice. ORIGINAL SIGNED BY Charles J. Haughney, Chief Storage and Transport Systems Branch Division of Industrial and Medical Nuclear Safety
Enclosures:
- 1. Safety Evaluation Report dated December 1994
- 2. Unsigned Certificate of Compliance No. 1004 59 FR 28496 Docket No. 72-lu04 Distribution:
PUBLIC NRC File Center IMIF R/F STSB R/F IMNS R/F GGunderson TGreene DOC NAME: OFC NAME DATE /94 94 C*COPY N=NO CO ICIAL REC
SAFETY EVALUATION REPORT OF VECTRA TECHNOLOGIES, INC. a.k.a. PACIFIC NUCLEAR FUEL SERVICES, INC. SAFETY ANALYSIS REPORT FOR THE STANDARDIZED NUHOMS HORIZONTAL MODULAR STORAGE SYSTEM FOR IRRADIATED NUCLEAR FUEL U.S. NUCLEAR REGµLA TORY COMMISSION OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS December 1994
TABLE OF CONTENTS Section
1.0 INTRODUCTION
, GENERAL DF.SCRIPTION . . . . . . . . . . . . . . . . . 1-1 1.1 In.ttc>clucti.on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1
- 1. 2 Context . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1
- 1. 3 General Discussion of Reference Materials and Role of Inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.3.1 General Description of Standardized NUHOMS System . . . . 1-4 1.3.2 Horizontal Storage Module . . . . . . . . . . . . . . . . . . . . . 1-4 1.3.3 Dry Shielded Canister . . . . . . . . . . . . . . . . . . . . . . . . 1-6
- 1. 3 .4 T.ransfer- Cas.k . . * * . . * * * . . * . * . . . . . . . . * . . * * . . 1-7 1.3.5 Fuel Transfer Equipment . . . . . . . . . . . . . . . . . . . . . . . 1-8 2.0 PRINCIPAL DP.SIGN CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 Inttoo.ucti.on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.2 Fuel to be Stored . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1
- 2. 3 Quality Standards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.4 Protection Against Environmental Conditions and Natural Phenomena . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.4.1 Normal Operating Conditions . . . . . . . . . . . . . . . . . . . . 2-3 2.4.2 Off-Normal Operating Conditions . . . . . . . . . . . . . . . . . 2-4 2.4.3 Accident Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . 2-5 2.4.4 l.oad Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . 2-5 2.5 Protection Against Fire and Explosion . . . . . . . . . . . . . . . . . . . . 2-6 2.6 Confinement Barriers and Systems . . . . . . . . . . . . . . . . . . . . . . 2-7
- 2. 7 Instrumentation and Control Systems . . . . . . . . . . . . . . . . . . . . . 2-9
- 2. 8 Nuclear Criticality Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-10 2.9 Radiological Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-11 2.10 Spent Fuel and Radioactive Waste Storage and Handling . . . . . . . . . 2-12 2.11 Decommissioning/Decontamination . . . . . . . . . . . . . . . . . . . . . . 2-13 2.12 Criteria for Fuel Stability . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-14 2.13 Findings and Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-14 3.0 STRUCTURAL EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3 .1 Horizontal Storage Module . . . . . . . . . . . . . . . . . . . . . . . . . . 3-6 3 .1.1 Design Description of HSM . . . . . . . . . . . . . . . . . . . . . 3-6 3 .1. 2 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-7 3.1.2.1 Normal and Off-Normal Operations . . . . . . . . . 3-7 3.1.2.2 Accident Analysis . . . . . . . . . . . . . . . . . . . . 3-8 3.1.3 Discussion and Conclusions . . . . . . . . . . . . . . . . . . . . . 3-10 i
TABLE OF CONTENTS (Continued) Section ~ 3.2 Dry Shielded Canister * . . . . . . . . . * . * * . . . . . * . . . *. . . . . . 3-11 3.2.1 Design Description of Dry Shielded Canister and Internals . . . . . . * . . . * . . . . . . . . . . . . . . . . . . . 3-11 3.2.2 Design Evaluation for DSC . . . . . . . . . . . . . . . . . . . . . 3-12 3.2.2.1 DSC Normal Operation Conditions . . . . . . . . . . 3-12 3.2.2.2 DSC Off-Normal Events . . . . . . .. .. .. .... 3-15 3.2.2.3 DSC Accident Conditions . . . . . . . . . . . . . . . . 3-16 3.2.2.4 DSC Fatigue Evaluation . . . . . . . . . . . . . . . . 3-23 3.2.2.5 DSC Corrision . . . . . . . . . . . . . . .. ...... 3-23 3.2.3 Discussion and Conclusions for DSC . . . . . . . . . . . . . . . 3-24 3 .3 Tran.sfer Cask * * * * * * * . * * * * * . . . . * * * * . . . * * * * * * * . * . 3-24
- 3. 3 .1 Desi.gr. Description of Transfer Cask . . . . . . . . . . . . . . . 3-24 3.3.2 Design Evaluation of the Transfer Cask . * * * * * . . . * . . .
- 3-25 3.3.2.1 TC Nonnal Operating Conditions . . . . . . . . . . . 3-25 3.3.2.2 TC Accident Conditions . . . . . . . . . . . . . . . . 3-28 3.3.2.3 TC Fatigue Evaluation . . . . . . . . . . . . . . . . . 3-30 3.3.2.4 TC Trunnion Loads and Stresses . . . . . . . . . . . 3-30 4.0 THERMAL EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4 .1 Revi.ew Prt>c:ed.llre * . . * * * * * * * * * * * * . * * . * * * * * . . . . * * *
- 4-2 4.1.1 Design. Description . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2
- 4. 1. 2 Acceptalloo Criter:ia . * . . . . * . . . . . . . . . . . . . . . . . . . 4-3 4 .1. 3 Revi.ew Metllcx:I * * * * * . . * * . * . * * * * * . . . * * * * * . * . 4-3 4.1.4 Key Design Information and Assumptions . . . . . . . . . . . . 4-4 4.2 Horizontal Storage Module (HSM) * . * . . . . . . . . . . . . . . . . * . . 4-5 4.2.1 Design. Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5
- 4. 2.1.1 Normal Operation . . . . . . . . . . . . . . . . . . . . 4-5 4.2.1.2 Off-Normal Conditions . . . . . . . . . . . . . . . . . 4-5 4.2.1.3 Accident Conditions . . . . . . . . . . . . . . . . . . . 4-5
- 4. 2. 2 Discussion and Conclusions . . . . . . . . . . . . . . . . . . . . . 4-6 4.3 DSC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 4-6 4.3.1 Design Evaluation . . . . . . * . . . . . . . . . . . . . . . . . . . . 4-6 4.3.1.1 Normal Operating Conditions . . . . . . . . . . . . . 4-6 4.3.1.2 Off-Normal Conditions . . . . . . . . . . . . . . . . . 4-7 4.3.1.3 Accident Conditions . . . . . . . . . . . . . . . . . . . 4-7 4.3.2 Discussion and Conclusion . . . . . . . * . . . . . . . . . . . . . . 4-7 4.4 TC . . . . . . . * . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-8 4.4.1 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-8 4.4.1.1 Normal Operating Conditions . . . . . . . . . . . . . 4-8 4.4.1. 2 Off-Normal Operating Conditions . . . . . . . . . . . 4-8 ii
TABLE OF CONTENTS (Continued) Section ~ 4.4.1.3 Accident Conditions . . . . . . . . . . . . . . . . . . . 4-9 4.4.2 Discussion and Conclusions . . . . . . . . . . . . . . . . . . . . . 4-9 5.0 CONFINEMENT BARRIERS AND SYSTEMS EVALUATION . . . . . . . S-1 5.1 Description of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1
-5.2 Design Evaluation .. -. . . . . . -....-. . . . . . . . -...- . . . . . . . . . 5-1 5.3 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 6.0 SHIRT.DING EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1 :Design Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2 Design Evaluation (Source Specification and Analysis) . . . . . . . . . . 6-1 6.3 Discussion and Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 7.0 NUCLEAR CRITICALITY SAFErY EVALUATION . . . . . . . . . . . . . 7-1 7 .1 Design Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.1.1 Standardized NUHOMS-24P Design . . . . . . . . . . . . . . . . 7-1 7.1.2 Standardized NUHOMS-52B Design . . . . . . . . . . . . . . . . 7-1 7.2 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.2.1 Standardized NUHOMS-24P Design . . . . . . . . . . . . . . . . 7-1 7.2.2 Standardized NUHOMS-52B Design . . . . . . . . . . . . . . . . 7-2 7.3 Conclusions .. -. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-3 7.3.1 Standardized NUHOMS-24P Design . . . . . . . . . . . . . . . . 7-3 7.3.2 Standardized NUHOMS-52B Design . . . . . . . . . . . . . . . . 7-5 8.0 RADIOLOGICAL PROTECTION EVALUATION . . . . . . . . . . . . . . . . 8-1 8.1 Design Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 8.2 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-2 8.2.1 On-Site Radiological Protection . . . . . . . . . . . . . . . . . . . 8-2 8.2.2 Off-Site Radiological Protection . . . . . . . . . . . . . . . . . . 8-4 8.2.2.1 Normal Operations . . . . . . . . . . . . . . . . . . . . 8-4 8.2.2.2 Off-Normal Operations . . . . . . . . . . . . . . . . . 8-5 8.3 Discussion and Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . 8-5 8.3.1 On-Site Radiological Protection . . . . . . . . . . . . . . . . . . . 8-5 8.3.2 Off-Site Radiological Protection . . . . . . . . . . . . . . . . . . 8-6 9.0 DECOMMISSIONING/DECONTAMINATION EVALUATION . . . . . . . 9-1 9 .1 Design Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.2 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1
- 9. 3 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-2 iii
TABLE OF CONTENTS (Continued) Section ~ 10.0 QUALITY ASSURANCE . * . . * . . . . . . * . * . . . . . . * . * . . . . . . . . 10-1 11.0 OPERATIONS, MAINTENANCE, TESTING, AND RECORDS * . . . . . . 11-1 11.1 Oj;>erat::ions . * * * * . . . * . * * * * * . * . * * * * * * . . . . . * . . **.
- 11-1 11.2 Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 11-1 11.3 Testing. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . 11-2 11.4 Records * . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 11-3 12.0 CONDffiONS FOR SYSTEM USE . . * . . . . . . . . . . . . . . . . . . . . . . 12-1 12.1 General Requirements and Conditions . . . . . . . * . . . . . . . . . . . . 12-1 12.1.1 Regulatory Requirements for a General License . . . . . . . . 12-1 12.1.2 Operating Procedures . . . . * . * * . . . . . . . . . . . . . . . . 12-2 12.1.3 Q-uality Assuran.ce . . . . . . . . . . . . . . . . . . . . . . . . . . 12-3 12.1.4 Heavy Loads Requirements *.*.*...*..........* 12-3 12.1.5 Training Module . . . . . * . . . . . . * . . . * . . . . . . . . .
- 12-3 12.1.6 Pre-Operational Testing and Training Exercise . . . . . . . . . 12-4 12.1.7 Special Requirements for First System in Place . . . . . . . . 12-4
- 12. 1. 8 Surveillance Requirements Applicability . . . . . . . . . . . . 12-5 12.2 Technical Specifications, Functional and Operating Limits . . . . . . . 12-5 12.2.1 Fuel Specification . . . . . . * * . . . . . . . . . . . . . . . . . . 12-5 12.2.2 DSC Vacuum Pressure During Drying * . . . . . . . . . . . . 12-11 12.2.3 DSC Helium Backfill Pressure . . . . . . . . . . . . . . . . . . 12-12 12.2.4 DSC Helium Leak Rate of Inner Seal Weld . . . . . . . . . . 12-13 12.2.5 DSC Dye Penetrant Test of Closure Welds * . . . . . * . . .
- 12-14 12.2.6 DSC Top End Dose Rates . * * . . * . . . . . . . . . . . . . . . 12-15 12.2. 7 HSM Dose Rates . . . . . . . . . * . . . . . . . . . . . . . . . . 12-16 12.2.8 HSM Maximum Air Exit Temperature . . . . . . . . . . . . . 12-17 12.2.9 Transfer Cask Alignment with HSM . . . . . . . . . . . . . . . 12-18 12.2.10 DSC Handling Height Outside the Spent Fuel Pool Building * . . * . * * . . * . . . * . . . . . . . . * . . 12-19 12.2.11 Transfer Cask. Dose Rates .*****.*.**.........* 12-20 12.2.12 Maximum DSC Removable Surface Contamination . . . . .
- 12-21 12.2.13 TC/DSC Lifting Heights as a Function of Low Temperature and Location . . . * * * . . . . . . . . . . . . 12-22 12.2.14 TC/DSC Transfer OJ;>erat::ions at High Ambient Temperatures . . . . . . . . . . . . . . . . . . . . . . . 12-24 12.2.15 Boron Concentration in the DSC Cavity Water (24-P Design Only) . . . . . . . . . . . . . . . . . . . . . . . . . 12-25 iv
TABLE OF CONTENTS (Continued) Section 12.2.16 Provision of TC Seismic Restraint Inside the Spent Fuel Pool Building as a Function of Horizontal Acceleration and Loaded Cask Weight . . . . . . . 12-26 12.3 Surveillance and Monitoring . * . * . . . . * . . . . * . . . . .. . ... . 12-27 12.3.1 Visual Inspection of HSM Air Inlets and Outlets (Front Wall and Roof Birdscreen) . . . . . . . . . . . . . . . . 12-27 12.3.2 HSM Thennal Performance . . . . * . . . . . . . . . .. . . .. 12-28 13.0 REFERENCFS 13-1 LIST OF FIGURES Fi~ure f.w 1.1 NUHO~ Horizontal Storage Module Arrangement . . . * . . . . .. . . . . 1-11 1.2 NUHOMS* Dry Shielded Canister Assembly Components . * . . . . . . . . . 1-12 1.3 NUHOMS'-' On-Site Transfer Cask . . . . . . . . . . * . . . * . . . . . .. . . . 1-13 1.4 Composite View of NUHOMS'-9 Transfer Cask and DSC with Spent PWR Fuel * . . . . . . . . . . . . . . . . * . * . . . . . . . . . . . . . . . . 1-14 1.5 Composite View of NUHO~ Transfer Cask and DSC with Spent BWR Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. . . 1-15 1.6 NUHOMS9 System Components, Structures, and Transfer - 12.1 Equipment Elevation View . . . . . . * . . . . . . . . . . . . . . . . . . . . *. . PWR Fuel Criticality Acceptance Curve . . . . . . . . . . . . . . . . . . . * . . LIST OF TABLES 1-16 12-9 Thbk 2.1 Design Criteria Sources Cited in the SAR . . . . . . . . . . . . . . . . .
- 2-16 2.2 Evaluation of Design Criteria for Normal Operating Conditions . . . . . 2-20 2.3 Evaluation of Design Criteria for Off-Normal Opera.tin,g Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-23 2.4 Evaluation of Design Criteria for Accident Conditions . * . . . . . . . . . 2-25 2.5 Load Combinations Used for HSM Reinforced Concrete . . . . . . . . . . 2-29 2.6 Load Combinations Used for DSC Support Assembly . . . . . . . . . . . 2-31 V
LIST OF TABLES (Continued)
~
3.1.2-1 HSM Load Combination Results . . . . . . . . . . . . . . . . . . . . . . . . 3-32 3.1.2-2 DSC Support Assembly Load Combination Results . . . . . . . . . . . . . 3-j3 3.2.2-1 DSC Stress Analysis Results for Normal Loads . . . . . . . . . . . . . . . 3-35 3.2.2-2 DSC Stress Analysis Results for Off-Normal Loads . * . . . . . . . . * . . 3-36 3.2.2-3 DSC Load Combinations for Normal and Off-Normal Operating Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-37 3.2.2-4 DSC Stress Analysis Results for Accident Conditions . . . * . . . . * . . . 3-38 3.2.2-5 DSC Load Combinations for Accident . . . . . . . . . . . . . . . . . . . . . 3-39 3.2.2-6 DSC Drop and Internal Pressure Accident Loads . . . . . . . . . . * . . . 3-41 3.2.2-7 DSC Enveloping Load Combination Results for Accident Loads . . . . . 3-42 3.3.2-1 Transfer Cask Stress Analysis Results for Normal Loads . . . . . . . . . . 3-43 3.3.2-2 Transfer Cask Load Combinations for Normal Operating Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-44 3.3.2-3 Transfer Cask Stress Analysis Results for Accident Loads . . . . '. . . . . 3-45 3.3.2-4 Transfer Cask Drop Accident Loads . . . . . . . . . . . . . . . . . . . . . . 3-46 3.3.2-5 Transfer Cask Results for Tornado Driven Missile Impact . . . . . . . . . 3-47 3.3.2-6 Transfer Cask Load Combinations for Accident Conditions . . . . . . . . 3-48 3.3.2-7 Summary of Stress Analyses for Upper Lifting Trunnions and Lower Resting Trunnions, Weld Regions and Cask Shell . . . . . . . 3-49 4.2 Summary of Component Temperatures as a Function of Ambient Temperatures . . . . . . . . . . . * . . . . . . . . . . . . . . . . . . 4-10 12-la PWR Fuel Specifications of Fuel to be Stored in the Standardi7.ed NUHOMS-24P DSC . . . . . . . * . . . . . . . . . . . . . .
- 12-8 12-lb BWR Fuel Specifications of Fuel. to be Stored in the Standardu.ed NUHOMS-52B DSC . . . * . . . . . . . . . * . . . . . . . . . 12-10 12.3.1 Summary of Surveillance and Monitoring Requirements . . . . . . . . . . 12-29 vi
1.0 INTRODUCTION
, GENERAL DESCRIPTION 1.1 Introduction VECTRA Technologies, Inc. (VECTRA), formerly Pacific Nuclear Fuel Services, Inc. (PNFS) has submitted a Safety Analysis Report (SAR) and supplementary docketed material (Reference 1) to support issuance of a Certificate of Compliance under 10 CFR Part 72, Subpart L (Reference 2). The application does not request NRC approval for installation at any specific site. The subject of the SAR is the *standardized NUHOMS horizontal modular cask" storage system. For the purposes of this review, the system will be referred to as the
- standardized NUHOMS system" to distinguish it from two previous versions of horizontal storage systems designed by NUTECH Engineers. The standardized NUHOMS system is different in many ways from the previous designs. Consequently the NRC staff reviewed all _*..;atures of the standardized NUHOMS.
10 CFR 72.238 provides that a Certificate of Compliance for a cask model will be issued by NRC on a finding of compliance with 10 CFR 72.236(a) through (i). In addition, 10 CFR 72.234(a) through (f) contain conditions of approval of the spent fuel cask design, including requirements for compliance with 10 CFR 72.236, quality assurance requirements according . to Subpart G of 10 CFR Part 72, and other administrative requirements for which the vendor is responsible. The review focused on the specific requirements for spent fuel storage casks contained in 10 CFR 72.236(a) through (m). These requirements cover fuel specifications, design criteria and administrative aspects. As noted, issuing the Certificate of Compliance will be based on an NRC finding that the requirements in 10 CFR 72.236(a) through (i) are met. Additionally, the staff will address whether the requirement of 10 CFR 72.236(m) has been considered. Issuance of the certificate is also subject to compliance with conditions specified in 10 CFR 72.236(j) through (1) for which the vendor is responsible. The objectives of this Safety Evaluation Report (SER) are to document the NRC staffs review and evaluation of the Safety Analysis Report (SAR) (Reference 1), and to clearly state the compliance (or noncompliance) of the license application to the applicable requirements of 10 CFR Part 72, Subpart L. 1.2 Context This SER provides NRC staff analyses and conditions on the SAR submitted by PNFS in conjunction with an application for certification of the standardized NUHOMS system described in the SAR. 1-1
The SAR was submitted in accordance with the requirements of 10 CPR Part 72, Subpart L. This SER is based on review for compliance with 10 CPR Part 72. Changes, clarifications, and additional information submitted to the NRC subsequent to the SAR during the review process (as liste.d at Reference 1) are considered to have the full effect and to express the same type of commitments as if they were included in the SAR itself. The SAR presumes that the standardized NUHOMS system will be used on the site of a nuclear power reactor licensed by the NRC under 10 CPR Part 50, and that fuel loading and unloading will occur within a fuel pool of the licensed facility. The SER does not include identification of additional requirements should fuel loading and unloading not be within the fuel pool of a facility licensed under 10 CPR Part 50. Information incorporated by reference or included by subsequent submittal (Reference 1) is considered as if it were information set out in the SAR itself. Where such information is already the subject of NRC staff approval, as by approval of a SER (e.g., References 3 and 4), that approval is considered to extend to the document incorporating the information by reference, to the extent of such incorporation, and subject to any qualifiers included in the referenced document and/or the corresponding SER. Use of the proposed standardized NUHOMS system will include operations and use of equipment related to safety within a fuel pool of the facilities licensed under 10 CPR Part 50. Fuel handling operations for an independent spent fuel storage installation (ISFSI) may require amendments to existing license technical specifications for the facility licensed under 10 CFR Part 50. This SER does not constitute the formal safety evaluation review for the safety of operations and equipment within fuel pool facilities. This SER does, however, examine the suitability of the transfer cask and DSC for mutual compatibility and for satisfaction of 10 CPR Part 72, Subpart L requirements. The proposed standardized NUHOMS system uses designs for its components which have evolved from designs in use or under construction as ISFSis at existing facilities. The approval of these ISFSis have involved NRC SERs for topical reports and license application SARs prepared in compliance with 10 CPR Part 72 and Regulatory Guide 3.48 (Reference 5). These documents have provided a context to the review which assisted in determining suitable criteria and design acceptability. 1.3 General Discussion of Reference Materials and Role of Inspection This SER refers in several places to fabrication specifications and engineering drawings for major components of the standardized NUHOMS system. The following paragraphs provide a general explanation for these references; they indicate the referenced specifications and drawings were not a basis for the staff's safety approval of a particular design topic in the SER. Rather, the staff reviewed aspects of the specifications and drawings to verify they accurately incorporated information that was part of the staff's basis for approving the cask design. 1-2
In basic terms, the cask vendor's design commitment, contained in codes, standards, and design criteria, is identified in the SER and serves as a design input for the vendor,s design calculations. The vendor's calculations both demonstrate compliance with design inputs and produce design details, e.g., reinforcing steel sizing, shield lid thickness, and many other results called design outputs. Much of the design output is contained in the vendor's engineering drawings and fabrication specifications. These drawings and fabrication specifications provide the vendor's constructor and component fabricator with detailed instructions for constructing the standardized NUHOMS system and its components. These drawings and specifications are not approved by the NRC as a part of the staff's review of the vendor's standardized NUHOMS system. As reflected in the SER, from the vendor's entire set of design information, the staff's design approval mainly relies on the vendor's criteria and design commitments and certain calculations or parts of calculations. The staff generally uses these portions of the vendor's design information to conduct an independent review and analysis to determine whether there is reasonable assurance that the vendor's design will perform its intended safety function. Another aspect of the staff's activities reflected in the SER, separate from and related to its safety review, is the delineation of the requirements for NRC inspections. For instance, the staff may prepare inspection procedures for the regional or headquarters vendor inspection staffs to conduct certain types of inspections of spent fuel storage cask vendors, fabricators, and constructors. These inspection procedures may specify, in addition to the information in the procedures and the design commitments contained in the SER, that inspectors should use information in the vendor's fabri<;ation specifications, engineering drawings, procurement documents, and material certifications to perform their field inspections. Where the inspection procedures refer to the vendor's drawings and specifications, the staff has typically reviewed selected aspects of the vendor's drawings and fabrication specifications to verify that the results contained in the vendor's design calculations have been accurately transposed into the drawings and specifications. By so doing, the staff provides added assurance that the inspectors will have accurate documentation to inspect the adequacy of construction. It is important to note that this NRC inspection does not constitute an additional NRC review of the standardiz:ed NUHOMS system design or a further NRC safety determination of the adequacy of the standardiz:ed NUHOMS system design. Rather, inspection activities address the adequacy of component construction, fabrication, and quality assurance (QA). Therefore, as previously noted, while the staff did not rely upon the fabrication specifications or drawings in approving the design, the SER will reflect the staff's check of portions of these documents to verify they contain accurate design output information to be used by the fabricator and checked by NRC inspectors. A further NRC check on the validity of the design output information is through QA requirements that review, approve, and link the individual QA programs of utility, vendor, fabricator, and constructor. Among other things, these QA programs ensure the contrQl of changes to drawings and specifications for accuracy and ensure proper engineering review. 1-3
10 CPR 72.234(a) through (f) which contain the conditions of approval for spent fuel cask design, require compliance with the specific design criteria of 72.236 and the quality assurance requirements in subpart O and identify other administrative requirements for which the system vendor is responsible. 10 CPR 72.236(a) through (m) contain the specific requirements for spent fuel storage cask approval, including spent fuel specification, design criteria, and administrative requirements. As noted, the Certificate of Compliance is issued by the NRC on a finding that the requirements of 72.236(a) through (i) are met, and after the staff determines that the requirement of 72.236(m) has been considered. The issuance is also subject to the conditions specified. in 72.236(j), (k), and (l) for which the vendor is responsible. 1.3.1 General Description of Standardized NUHOMS System The following descriptions of the standardized NUHOMS system are based on the more complete descriptions provided by Reference 1 and are only included here for the convenience of readers of the SER. The SER is based on the descriptions provided in the SAR. The standardized NUHOMS system components for irradiated fuel assemblies (IFA) storage at an ISFSI are the Dry Storage Canister (DSC) and the Horizontal Storage Module (RSM). Additional systems required for the DSC closure and transfer include the transfer cask (TC), the skid and skid positioning system, the trailer, the hydraulic ram system, and the DSC vacuum drying system. 1.3.2 Horizontal Storage Module The standardized NUHOMS system uses HSMs assembled from standardized units, as illustrated in Figure 1. 1. These are: *
- Base Unit Assembly, consists of the monolithically poured reinforced concrete (RC) base unit of floor and four walls, with DSC access opening, inlet and outlet ventilation openings, and embedments for attachment of restraints, the DSC support structure roof slab, heat shields, spacers and shield walls. The base unit side walls are 0.46 m (l'-6"), the front wall is 0.76 m (2'-6"), and the rear wall and floor are 0.30 m (1 ') thick.
- DSC Support Structure, a structural steel frame with rails installed within the base unit to provide for sliding the DSC in and out of the HSM, supports the DSC within the HSM, and resists and transfers forces associated with a jammed DSC or a design basis earthquake.
- Roof Slab Assembly, a rectangular 0.91 m (3 foot) thick RC slab which is bolted to the base unit to complete the shielded enclosure for DSC storage. It includes embedments for attachment to the base unit for positioning, for lifting, and for 1-4
attachment of screens between adjacent modules and between modules and external separate shield walls.
- Second Shield Wall, a rectangular 0.61 m (2 foot) thick RC slab installed vertically at the outer side of HSM at the ends of rows of HSMs. The end shield walls are installed with channel spacers, shielded bolt assemblies attaching them to the RSM, and screens across the gaps between the walls.
- Single Module Re.ar Shield Wall, used when HSMs are placed in single rows (the alternative placement is with two rows back-to-back). The rear shield wall is a rectangular 0.46 m (l '-6") thick RC slab installed vertically against a base unit rear wall without an intervening space. The rear shield walls are installed with shielded bolt assemblies.
- Shielded Door, composed of a 5.1 cm (2") thick steel plate and 14.9 cm (5-7/8")
of RC, which closes the DSC access opening and provides radiation shielding and resistance to natural phenomena.
- Basemat, cast-in-place RC foundation on which the HSMs rest. The HSMs are not connected to the ba.semat and are held in place against any horizontal forces by friction. Thickness of the basemat is to be determined by site foundation analysis.
- Approach Slabs, a cast-in-place RC slab providing for access and support of the DSC transport and transfer systems. This slab is structurally connected to the Basemat. Thickness of the approach slab is to be determined by site foundation analysis.
The RSM protects the DSC from the potentially adverse effects of natural phenomena, such as earthquake, tornado, tornado missiles, flood, and temperature. The modular RSM system is considered acceptable for layout variations from a single RSM to unrestricted numbers of RSMs in single or back-to-back rows,* without additional shielding, as approved in this SER, if criteria of the SAR are also met. The RSM dissipates decay heat from the spent fuel by a combination of radiation, conduction, and convection. Natural convection air flow enters at the bottom of the HSM, circulates around the DSC, and exits through the flow channels between the HSM roof slab and side walls. A thermal radiation shield is used to reduce the HSM concrete temperatures to within acceptable limits for all conditions. 1-5
1.3.3 Dry Shielded Canister The DSC is illustrated in Figure 1.2. A DSC is shown in its storage position in Figure 1.1. The principal component subassemblies of the DSC are the she.11 with integral bottom cover plate and shield plug and rain/grapple ring, top shield plug, top cover plate, and basket assembly. The main component of construction of the DSC is a type 304 stainless steel cylindrical confinement vessel. The internal basket assembly for the PWR fuel is comprised of 24 guide sleeves supported by 8 spacer discs at intervals corresponding to the fuel assembly spacer grids. Support rods maintain the spacer disks in location. The internal basket assembly for the channelized BWR fuel i.~ similar to the PWR except that 52 guide sleeves are used for the BWR application, supported by 9 spacer discs. Borated stainless steel poison plates are used for all BWR baskets. Steel shielding is used in both the top and bottom end shield plugs. Criticality safety during wet loading operations for the PWR fuel is maintained through the geometric separation of the fuel assemblies within the internal basket assembly, the inherent neutron absorption capability of the steel guide sleeves, the proper selection of sufficiently depleted fuel assemblies, and adequate boron concentration in the pool water. For BWR fuel assemblies, criticality safety during wet loading operations is maintained by similar means. except that borated stainless steel plates are used in the guide sleeve assemblies and borated water is not required. Credit for bumup is not currently permitted by the NRC staff. The DSC provides mechanical confinement for the stored fuel assemblies and all radioactive mate.rials for two purposes: to prevent the dispersion of particulate or gaseous radionuclides from the fuel, and to maintain a barrier of helium around the fuel in order to mitigate corrosion of the fuel cladding and prevent further oxidation of the fuel. The DSC provides radiological shielding in both axial directions. The top shield plug serves to protect operating personnel during the DSC drying and sealing operations. The bottom shielding reduces the HSM door area dose rates during storage. The DSC shielding is designed for a maximum contact dose of 2 m.Sv/hr (200 mrem/hr) before draining the DSC cavity. The DSC is designed to slide from the transfer cask into the HSM and back without undue galling, scratching, gouging, or other damage to the sliding surfaces. This is accomplished by a combination of surface finishes and dry film lubricant coatings applied to the DSC and the DSC support assembly in the HSM. The transfer operation is illustrated in Figure 1.6. 1-6
1.3.4 Transfer Cask The principal components of the transfer cask (TC) are shown in Figure 1.3 (SAR Figure 1.3-2b). Figures 1.4 and 1.5 (SAR Figures 4.2-9 and 4.2-9a) show the TC with DSC. Figure 1.6 (SAR Figure 1.1-2) shows the TC in position for DSC transfer to the HSM. The transfer cask is a cylindrical vessel with a bottom end closure assembly and a bolted top cover plate. The cask's cylindrical walls are formed from three concentric steel shells with lead poured between the inner liner and the structural shell to provide gamma shielding during DSC transfer operations. The structural and outer shells form an annular pressure vessel. A solid neutron absorbing material is cast between the structural shell and outer shell to provide neutron shielding when the DSC is in the TC. The cask bottom end assembly is welded to the cylindrical shell assembly. It includes two closure assemblies for the ram/grapple access penetration. A watertight bolted top cover plate, with a core of solid neutron absorbing material, is used for transfer operation within the Auxiliary Building (or Spent Fuel Storage Building in some plants). The bolted ram access penetration bottom cover plate assembly is replaced, after the TC is horizontal on the transport trailer and while still in the Auxiliary Building, by a two-piece neutron shield plug assembly for transfer operations from/to the Auxiliary Building to/from the HSM. The inner plug of this assembly is bolted to the TC. The outer plug is held in brackets by gravity. At the HSM site, the outer plug of the assembly is removed to provide access for the ram/grapple to push/pull the DSC into/from the HSM. The top plate cover is bolted to the top flange of the cask during transport from/to the Auxiliary Building to/from the ISFSI. The top cover plate assembly consists of a thick structural plate with a thin shell encapsulating solid neutron shielding material. Two upper lifting trunnions are located near the top of the cask for downending/uprighting and lifting of the cask in the Auxiliary Building. Two lower trunnions, located near the base of the cask, serve as the axis of rotation during downending/uprighting operations and as supports during transport to/from the ISFSI. The TC is not designed as a pressure vessel. The neutron shield material is BISCO Products NS-3. NS-3 is a shop castable, fire resistant material with a high hydrogen content which is designed for nuclear applications. The materlaJ is used in the cask outer annulus, top and bottom covers, and temporary shield plug. It produces water vapor and a small quantity of non-condenslole gases when heated above 100°C (212°F). The off-gassing produces an internal pressure which increases with temperature. As the temperature is reduced, the off-gas products are reabsorbed into the matrix, and the pressure returns to atmospheric. The annular neutron shield containment is designed for an internal pressure of 655 kPag (95 psig). Pre-set safety relief valves are included to protect the neutron shield cover in the event that its design pressure is exceeded. 1-7
1.3.5 Fuel Transfer Equipment With the exception of the TC, fuel transfer and auxiliary equipment necessary for ISFSI operations are not included as a part of the standardized NUHOMS system to be reviewed for a Certificate of Compliance under 10 CFR 72, Subpart L. However this equipment will be described for general information only. Fuel is transferred in ISFSI operations by means of the TC. Inside the fuel pool facility, the TC with loaded DSC is transferred from the fuel pool to a position where decontamination, drying, sealing, and installation of the TC cover take place. The TC and DSC are then transferred to the transfer trailer, still within the Auxiliary Building. The TC with DSC is moved to position for coupling with the HSM access opening by the transfer trailer, with final positioning by movement of the TC support shield over the trailer. The DSC is transferred from the TC to the HSM by use of the ram acting through the ram access opening of the TC. - Equipment used to physically grip, lift, inspect, and position the IFAs in the fuel pool is the same as that already in place and in use for Auxiliary Building IPA handling. There is special equipment involved with fuel transfer within the Auxiliary Building unique to the ISFSI application. Of this, only the TC lifting yoke is used exclusively within the Auxiliary Building and is thereby subject to evaluation as part of the 10 CFR Part 50 license review of updates to the FSAR. The lifting yoke is a special lifting device which provides the means for performing all cask handling operations within the plant's Auxiliary Building. It is designed to support a loaded transfer cask weight up to 90.7 t (100 tons). A lifting pin connects the Auxiliary Building cask handling crane hook and the lifting yoke. The lifting yoke is a passive, open hook design with two parallel lifting beams fabricated from thick, high-strength carbon steel plate material with a decontaminable coating. It is designed to be compatible with the Auxiliary Building crane hook and load block. The lifting yoke engages the outer shoulder of the transfer cask lifting trunnions. To facilitate shipment and maintenance, all yoke subcomponent structural connections are bolted. Lifting slings are used in the Auxiliary Building for placement and removal of the DSC and TC shield plugs and covers. Eyebolts are installed on the items to be lifted to facilitate rigging for lifting. The transfer trailer is used to transport the transfer cask skid and the loaded transfer cask from the Auxiliary Building to the ISFSI. The transfer trailer is an industrial heavy-haul trailer with pneumatic tires, hydraulic suspension and steering, and brakes on all wheels. Four hydraulic jacks are incorporated into the transfer trailer design to provide vertical elevation adjustment for alignment of the cask at the HSM. The transfer trailer is shown in Figure 1.6. It is pulled by a conventional tractor. 1-8
The trailer is pulled using a drawbar steering unit. The steering unit includes hydraulic master cylinders to provide motive force for the slave steering cylinders in the trailer. The trailer may also be steered manually using a remote steering control located on a pendant. This feature allows precise control as the trailer is backed up to the HSM. The pendant allows the operator the freedom to observe the trailer from the side and also reduces the operational exposure by increasing operator distance from the DSC and reducing operator time. The trailer incorporates a skid positioning system which holds the TC support skid. The functions of the skid positioning systems (SPS) are to hold the TC support skid stationary (with respect to the transport trailer) during cask loading and transport, and to provide alignment between the transfer cask and the HSM before insertion or withdrawal of the DSC. It is composed of tie down or travel lock brackets, bolts, three hydraulically powered horizontal positioning modules, four hydraulic lifting jacks, and a remotely located hydraulic supply and control skid. The hydraulic jacks are designed to support the cask setdown load and the loads applied to them during the HSM loading and unloading. Their purpose is to provide a solid support for the trailer frame and skid. Three measures are taken to avoid accidental lowering of the trailer payload: the hydraulic pump will be de-energized after the skid has been aligned (the jacks are also hydraulically locked out during operation of the horizontal cylinders); there are mechanical locking collars on the cylinders; and pilot-operated check valves are located on each jack assembly to prevent fluid loss in the event of a broken hydraulic line. Three positioning modules provide the motive force to horizontally align the skid and cask with the HSM before insertion or retrieval of the DSC. The positionin~ module controls are manually operated and hydraulically powered. The system is designed to provide the capability to align the cask to within the specified alignment tolerance. The hydraulic power supply and controls for the SPS are located on a skid which is normally stored on the hydraulic ram utility trailer. Directional metering valves are used to allow precise control of cylinder motions. The SPS is manually operated and has three operational modes: simultaneous actuation of the four vertical jacks or any pair of jacks, actuation of any single vertical jack, or actuation of any one of the three horizontal actuators. Simultaneous operation of the vertical jacks and the horizontal actuators is not possible. Fourteen small hydraulic quick-connect lines provide power to the seven SPS hydraulic cylinders. The hydraulic ram system provides the motive force for transferring the DSC between the TC and the HSM. The hydraulic ram consists of a double-acting hydraulic cylinder with a capacity of 36,290 kg (80,000 lb.) in either push or pull mode and stroke of 6.4 m (21 feet). The ram will be supported during operation by a frame assembly attached to the bottom of the transfer cask and a tripod assembly resting on the concrete slab. The operational loads of the hydraulic ram are grounded through the transfer cask. The hydraulic ram system 1-9
includes a grapple at the end of the piston which is used to engage a grapple ring on the DSC for retrieval operations. Figure 1.6 shows main components of the hydraulic ram system (SAR Figure 1. 1-2). 1-10
ENO MODULE SHIELD WALL HORIZONTAL STORAGE
~E SHEWED ACCESS DOOR CANISTER AXIAL RETAINER -
........' AIR INLET BASEMAT DRY SHIELDED CANISTER CASK DOCKING - COllAR Figure I. I NUHOMS Horizontal Storage Module Arrangement
SHIELD PLUG OUTER COVER PLATE INNER COVER PLATE
\ TOP END
)
DRAIN ANO FILL PORT I N DSC SHELL IIOJJOM
~
- SUPPORT ROD
- GUIDE SLEEVE" RAM GRAPPLE *-
RING BASKET ASSEMBLY I OUTER COVER SIPHON TUBE PLATE
- FIXED NEUTRON ABSORBE HS SHIELD PLUG FOR CHANNELU:O awn flJEI INNER COVER PLATE Figure 1-2 NUHOMS Dry Shielded Canister Assemhly Components
TOP COVER PLATE ASSEMBLY CASKCAVlTY AUXILIARY PORT
.. I' I
l I !1 I: I I,. I i ti I! i
!I LiI I'
; . , 1*
LEAD :I I I SHIELD I I I
! I I1 1 11\
.. ,I I,
SOLID NEUTRON SHIELD ! 11 I OUTER JACKET il 1/1, I :1
; ;~ I
- RAMACCESS PENETRATION
- ii I!
\ ~ Il.
. 'j I!I I I I iJI III LOWER SUPPORT TRUNNION I;i CASK CAVITY DRAIN PORT RAM ACCESS PENETRATION BOTTOMEND COVER PLATE ASSEMBLY
~,
Figure 1.3 NUHOMS* On-Site Transfer Cask 1-13
DSC INNER TOP COVER PLATE CASK TOP NEUTRON SHIELD PLATE \
\ \
CASKNEU~~ FUEL ASSc~.IBL v SHIELD UPPER LIFTING -TRUNNION DSC GUIDE SLEEVE DSC SPACER DISK 01./TEAJACKET CASK STRUCnJRAL SHELL LOWER SUPPORT TRUNNION DSC SHELL CASK BOTTOM OSC BOTTOM SHIELD PLUG ENO Pl.ATE OSCOUTER CASK BOTTOM NEUTRON SHIELD BOTTOM COVER Pl.ATE CASK RAM PENETRATION COVER PlATE OSC INNER BOTTOM COVER Pt.ATE OSC GRAPPLE RING Figure 1.4 Composite View of NUHOMS* Transfer Cask and DSC with Spent PWR Fuel 1-14
CASK TOP NEUTRON SHIELD - .\
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CASK TOP \ COVER PLATE CASK EXTENSION --. COLLAR ASSEMBLY
.* -- FUEL ASS EMBLY CASK N5UTRON DSC SUPPORT ROO SHIELD UPPER LIFTING TRUNNION DSC NEUTRON ABSORBERS DSC SPACER DISK OUTER JACKET CASK INNER LINER CASK STRUCTURAL SHELL LOWER SUPPORT TRUNNION DSC SHELL DSC BOTTOM SHIELD PLUG CASK BOTTOM ENO PLATE DSC INNER BOTTOM COVER PLA iE CASK BOTTOM NEUTRON SHIELD DSC OUTER BOTTOM COVER PLATE CASK RAM PENETRATION COVER PLATE DSC GRAPPLE RING Figure 1.5 Composite View of NUHOMS* Transfer Cask and DSC with Spent, BWR Fuel 1-15
RAM TRUNNION - SUPPORT RAM PENETRATION TEMPORARY SHIELD PLUG RAM HYDRAULIC HORIZONTAL CVLt'DER STORAGE MODULE ADJUSTABLE RAM TRIPOD ... ::::::::: ,.f.."'.:_ . SUPPORT I °' APPROACH SLAB
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-0SC(STOAED POSITION)
\ lSFSI BASEMAT TRANSPORT TRAILER GRAPPLE ASSEMBLY - TRANSFER CASK
- CASK SUPPORT SKID AND POSITIONING SYSTEM Figure 1.6 NUHOMS Sy~tem Components, Structures, and Transfer E4uipment Elevation View
2.0 PRINCIPAL DESIGN CRITERIA 2.1 Introduction 10 CPR 72.236(b) requires that design bases and design criteria must be provided for structures, systems and components imJX)rtant to safety. The criteria for the design, fabrication, construction, testing and performance of components imJX)rtant to safety are set forth in the general requirements of 10 CPR 72.236(a) through (i). In addition, this SER addresses the staff's consideration of design criteria in 10 CPR Part 72, Subpart F, *oenera1 Design Criteria For Independent Spent Fuel Storage Facilities (ISFSI)." The following subsections discuss the design criteria applied by the NRC staff to the standardized NUHOMS system and the degree to which the design as described in the SAR is in compliance with these criteria. The. subsection headings generally correspond to criteria in 10 CFR 72, Subpart F and 10 CFR 72.236. Section 3.0 of the SAR contains the design criteria proposed by the standardized NUHOMS system vendor. It also identifies sources for these design criteria. The sources and their acceptability are summarized in Table 2.1 of this report. As shown in the table, the identified sources were determined to be acceptable. 2.2 Fuel to be Stored 10 CPR 72.236(a) requires that a specification for the spent fuel to be stored in the cask be provided, including type of spent fuel (i.e., BWR, PWR, both), maximum allowable enrichment of the fuel before irradiation, bum-up, minimum acceptable cooling time of the spent fuel before storage in the cask, maximum heat designed to be dis.~pated, maximum spent fuel loading limit, condition of the spent fuel, and inerting atmosphere requirements. The fuel specified to be stored in the standardized NUHOMS system is intact (not consolidated) PWR or BWR, with physical characteristics presented in Table 12-la and 12-lb of the SER. The related characteristics and parameters of the fuel to be stored are determined by assumptions used by the vendor to analyze and evaluate the capabilities of the system, such as criticality safety, shielding, heat removal, confinement, and the limitations imposed by these analyses in meeting acceptance criteria. The fuel specification, accepted by the staff, is presented in Section 12.2.1 of this report. 2.3 Quality Standards The quality standards considered by the staff for the spent fuel storage system are in 10 CFR 72.122(a) and in 10 CFR 72.234(b). 10 CPR 'n.122(a) provides that structures, systems and components important to safety must be designed, fabricated, erected, and tested to quality standards commensurate with the importance to safety of the function to be performed. 2-1
10 CPR 72.234(b) requires that the design, fabrication, testing and maintenance of spent fuel storage casks must be conducted under a quality assurance program that meets the requirements of Subpart G of 10 CFR 72. Quality standards dealing with the design, materials, fabrication techniques, inspection methods, etc., are cited by the vendor in the sections of the SAR where the standards are applicable. Judgments regarding the adequacy of these standards are also presented in the corresponding sections of this report. The Quality Assurance program proposed by the vendor for the design, fabrication and construction of the standardized NUHOMS system is presented in Section 11.0 of the SAR. The staff's evaluation of the vendor's Quality Assurance program is discussed in Section 10.0 of this report. 2.4 Protection Against Environmental Conditions and Natural Phenomena 10 CFR 72.236(b) requires design bases and design criteria for structures, systems and components important to safety. 10 CFR 72.122(b) provides that structures, systems, and components important to safety must be designed to accommodate the effects of, and to be compatible with, site characteristics and environmental conditions associated with normal operation, maintenance, and testing of the ISFSI and to withstand postulated accidents. It also provides that structures, systems, and components important to safety must be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, lightning, hurricanes, floods, tsunami, and seiches, without impairing their capability to perform safety functions. - The standardized NUHOMS system is intended to withstand environmental conditions and natural phenomena that may occur under "normal," "off-normal," and* "accident" circumstances. Normal operations, off-normal operations, and accidents are described in NRC Regulatory Guide 3. 62 (Reference 72). In Regulatory Guide 3. 62, normal operations of an ISFSI consists of the set of events that are expected to occur regularly or frequently. Off-normal operations consist of the set of events that, although not occurring regularly, can be expected to occur with moderated frequency or on the order of once during a calendar year. Design criteria call for normal and off-normal conditions to both satisfy allowable limits for routine operations. Accidents are any credible incident that could result in a potential radiation dose of 25 mrem or more beyond the controlled area and situations wherein direct radiation or radioactive materials may be released in such quantities as to endanger personnel within the controlled area. Accidents include two sets of events. The first consists of that set of infrequent events that could reasonable be expected to occur during the lifetime of the ISFSI. The second is concerned with natural phenomena or low probability events that are postulated because their consequences may result in the maximum potential impact on the immediate environs. Their consideration establishes a conservative design basis for certain systems with important to safety. 2-2
.)
The principal design criteria used for structural design are listed in Tables 2.2 (normal), 2.3 (off-normal), and 2.4 (accident). Column (5) of Table 2.2, 2.3, and 2.4 includes comments on applicability of the criteria for use of the standardized NUHOMS system at possible sites. Some criteria are acceptable and are not affected by the actual site. Some criteria vary by location, but the values used in the SAR are sufficient to envelope the values that may be appropriate for credible sites in the continental United States. Other criteria should be verified as bounding the proper values for the specific installation location. The SAR does not present a HSM foundation design for certification. The foundation design shown in the drawing is considered a nominal design for illustration. This foundation design is probably adequate or conservative for many potential sites given appropriate site preparation. A foundation analysis should be performed for verification of the adequacy of the nominal design or to provide the basis for a new design specific to the installation. The foundation is not relied upon to provide safety functions. There are no structural connections or means to transfer shear between the HSM base unit module and the foundation slab. However, the user must evaluate the foundation in accordance with 10 CFR 72.212(b)2 and (b)(3) to ensure that, in an unlikely event, no gross failures would occur that would cause the DSC to jam during transfer operation, or cause the standardized NUHOMS system to be in an unanalysed situation, and would prevent removal of a DSC from a HSM. Evaluation of an ISFSI design is accomplished by evaluating the stated criteria and the actual design as separate review stages. Criteria may be acceptable, but if the actual design does not meet the criteria the system may not be acceptable. Similarly, some proposed criteria may not be acceptable, however, because of consexvatism in the actual design, it may be determined to satisfy more stringent, yet acceptable, criteria. 2.4.1 Normal Operating Conditions The staff considers that the design criteria as stated and referenced in Table 2.2 are acceptable for certification with the following exception: in principle, the DSC should be considered a live load rather than treated as a dead load as in the SAR. The weight of the DSC is precisely known; however, any additional loads associated with its transfer are treated as live loads. Based on staff review of the design, factors of safety, and impact of treatment as a dead load, the usage in the SAR is accepted. It has been determined that the factor of safety, if the DSC were treated as a live load, would still be acceptable for the actual design - Section 8.1.1.1 of the SAR states that the long-term average yearly ambient temperature is assumed to be 21 °C (70°F). This temperature bounds most, but not all, reactor sites in the Continental United States. Reactor sites which exceed this temperature are Palo Verde, Turkey Point, and St. Lucie. 2-3
The SAR states that the design basis operating temperatures are -40°C to 52°C (-40°F to 125°P). While this temperature range is acceptable for storage, it is not acceptable for on-site transfer or lifting and handling of the DSC. Paragraph 10.3.15 of the SAR states that the minimum ambient temperature for transfer of the loaded DSC inside the TC manufactured from ferritic steel shall be -17.8°C (0°F)., Furthermore, if the ambient temperature exceeds 37.8°C (100°P), a solar shield shall be provided to protect the solid neutron shield material contained in an annular volume of the TC. No lifting above 203 cm (80 inches) of the loaded DSC is permissible below -28.9°C (-20°F) inside the spent fuel pool building. If the lift height exceeds 203 cm (80 inches), then the minimum temperature is restricted to -17.8°C (0°F) for lifting inside the spent fuel pool hui)ding. The appropriate criteria for impact testing of ferritic steels for the DSC shell or basket is NURBG/CR-1815 (Reference 7). Similarly, the SAR states that the design basis operating temperatures are -40°C to 52°C (-40°F to 125°P) for the TC. This temperatures range is acceptable for handling the empty TC; however, for lift heights of 203 cm (80 inches) or lower the minimum limiting temperature for handling a TC with a loaded DSC shall be restricted to -28.9°C (-20°F). This limiting minimum temperature shall apply inside the spent fuel pool building. For lift heights above 203 cm (80 inches) inside the spent fuel pool building, the minimum temperature is restricted to -17.8°C (0°F). For all transfer operations outside the spent fuel pool building, the maximum height is limited to 203 cm (80 inches) and the minimum temperature is limited to -17.8°C (0°F). The appropriate criteria for impact testing of ferritic steels for the TC is ANSI N14.6 paragraph 4.2.6 (Reference 8). 2.4.2 Off-Normal Operating Conditions Table 2.3 lists summary design criteria used for off-normal operating conditions. The staff considers that the design criteria stated and referenced in Table 2.3 are acceptable and appropriate with the following exceptions:
- The jammed condition loading for the DSC support assembly is properly listed as an off-normal condition; however, it is actually used in the load combinations as though it were an "accident* loading. The NRC requires normal and off-normal loads to be evaluated similarly in load combination expressions without use of the increases in stresses permitted for *accident* type loads.
- The DSC support assembly does not have criteria identified for off-normal temperature rise. This is considered not acceptable, however the actual design is determined to satisfy criteria which should have been listed.
- Identical minimum service temperature restrictions apply to the transfer of the loaded DSC and the TC with a loaded DSC, as described in 2.4.1 above.
2-4
2.4.3 Accident Conditions Table 2.4 lists summary design criteria used for accident conditions. These conditions include extreme natural phenomena, accidental drops and impacts, fire, and explosions. The staff considers that the design criteria as stated are acceptable with the exception that the jammed loading condition was treated as an accident (discussed above). Where other criteria as determined by the staff are considered more appropriate, the criteria are stated in the table and are determined to be conservative and thereby acceptable. The following accident design criteria should be verified as equal to or exceeding appropriate parameters for the actual installation site.
- Flood parameters, especially the 4.6 m per second (15 foot per second) maximum velocity.
- Seismic maximum horizontal and vertical ground accelerations.
- Maximum ambient temperature.
- Potential for fire or explosion, Section 2.5, below.
- Requirements for lightning protection.
- Extreme low temperature, see Section 3.0, below.
- Maximum lift height of loaded DSC to 203 cm (80 inches) outside the spent fuel pool.
It is recognized that some other environmental condition limits used in the SAR and SER may not envelope all points within the continental United States. These are not included in the above list of criteria to be verified due to their negligible impact on the design and due to the otherwise unsuitability of the exceptional locations for ISFSI. 2.4.4 Load Combinations Load combinations presented in the SAR for use in verification of design are presented for the HSM in Table 2.5 and for the DSC Support Assembly in Table 2.6. Staff comments are included in the tables on the acceptability and use of the load combinations. As annotated in the tables, the load combinations used and omitted are considered acceptable with the exception of that used for the *off-normal* case of a jammed DSC loading the DSC support assembly (fable 2.5). The combinations used (numbers 15 and 16) have factored strengths (1.7 for strength or stress in other than shear, 1.4 for shear) which are not appropriate for off-normal loads. In addition to specifying load combinations to be used for the design of the HSM, the SAR also specified design load combinations for the DSC and the TC. In both cases, parts of the ~ME B&PV Code Section ill are used (Reference 9). These definitions of normal, off-2-5
normal, and accident operations are discussed. Tables 8.1-1 and 8.1-la in the SAR define types of loads for all components for normal and off-normal conditions respectively. Normal loads for the DSC shell include deadweight, internal pressure, thermal loads, and normal handling loads. The DSC basket is not subjected to internal pressure loads. Normal loads for the TC include deadweight, thermal, normal handling and live loads. These load combinations correspond to Service Level A in the ASME Code. Off-normal loads for the DSC shell include deadweight, internal pressure, off-normal temperature loads and off-normal handling loads. The DSC internals are subjected to deadweight and off-normal thermal loads. The TC is subjected to combined loads including deadweight, off-normal thermal and off-normal handling loads. These load combinations correspond to Service Level B in the ASME ~ - - Table 8.2-1 of the SAR defines the various load combinations for the accident loads, or Service Levels C and D of the ASME Code. The effects of loads are considered separately. For example, a design basis earthquake at the ISFSI site (.26g), followed by the design basis flood 17m (50 ft) submersion would not be additive. The accidents considered for the DSC include: earthquake, flood, accidental drop, blockage of HSM inlet and outlet vents, and accidental internal pressure. The accidents considered for the TC include: tornado wind and tornado wind driven mismles, earthquake, accidental drop, and loss of cask neutron shield. In evaluating whether the cask design bases envelope site parameters under 72.212(b)(3), the licensee must consider all conditions that are a consequence of the occurrence of the accident conditions under evaluation. 2.5 Protection Against Fire and Explosion
- 10 CFR 72.172(c) contains criteria for fire protection which the staff has considered .
Section 3.3.6 of the SAR addresses. the credibility of ISFSI initiated fires and explosions. As noted in Table 2.4, the SAR states that design criteria for fire or explosion are "enveloped by other design events.* This SER evaluation recognw.es that the probability of a fire or explosion affecting standardized NUHOMS system nuclear safety varies with potential installation sites, procedures and equipment used for transfer actions, and possible accidents at or in the vicinity of the system (e.g., aircraft and vehicle crashes, railroad, truck, or pipeline fires and explosions). The SAR did not identify specific criteria for fire and explosion. It stated that such events were bounded by other criteria. Externally initiated explosions are considered in the SAR to be bounded by design basis tornado generated missiles. The DSC can withstand the external pressure of a flood of a head of water equal to 15.2 m (50 feet). For certification of the design, appropriate basic criteria could be based on limits associated with nuclear safety such as: 2-6
- Maintenance of acceptable radiation shielding to keep exterior surface dose rates within acceptable limits.
- Maintenance of physical protection of the DSC from other events.
- Limiting the maximum temperature reached by cladding to the acceptable limit.
- Umiting stresses and deformations of the DSC shell due to temperature and/or loads to ensure that rupture of that confinement barrier is not risked.
- Limiting stresses and deformation of the interior spaces, support, and positioning elements with the DSC due to temperature and/or accelerations to ensure acceptable spacing and retrieval of IFAs.
The load limits, expressed in load combinations involving other design loads, should provide adequate criteria to satisfy the above. However, the user must not assume that the temperatures, accelerations, missile impacts, and other loads examined for the certification cannot be exceeded by credible fires and explosions regardless of site location and other circumstances. 10 CFR 72.212(b)(2) requires written evaluations to establish that certificate conditions are met with respect to fire and explosion because a potential exists for all sites in the use of internal combustion engine-powered transport trailer. Accordingly, verification that loadings resulting from potential fires and explosions do not exceed those used in the SAR for other events and conditions, is required for installation of the standardiz.ed NUHOMS systems in accordance with 10 CFR 72.212(b)(2). 2.6 Confinement Barriers and Systems The staff considered 10 CFR 72.122(h) which provides that confinement barriers and systems shall: "(l) protect the spent fuel cladding against degradation that leads to gross ruptures or the fuel must be otherwise confined such that degradation of the fuel during storage will not pose operational safety problems*; (2) must provide ventilation and off-gas systems *where necessary to ensure the confinement of airborne radioactive particulate materials during normal or off-normal conditions*; (3) *must have the capability for continuous monitoring in a manner such that the licensee will be able to determine when corrective action needs to be taken"; (4) "must be packaged in a manner that allows handling and retrievability without the release of radioactive materials to the environment or radiation exposures in excess of 10 CFR Part 20 limits." The staff has reviewed the features of the DSC design which provide confinement of radioactive material and, specifically, protection of the spent fuel assemblies. The protection of the spent fuel assemblies depends on two conditions: (1) appropriate fuel clad temperature is not exceeded, and (2) the inert helium atmosphere does not leak out. The first condition is met by thermal hydraulic analyses. See SER Section 4.0. The maintenance of the helium 2-7
atmosphere is discussed below. The review was directed at two aspects of the design: the integrity of the DSC and the allowable leak rate. Confinement is ensured by a combination of inspection techniques, including radiographic inspection, dye penetrant testing, and helium leak testing. The SAR takes the position that the inert helium atmosphere in the DSC will not leak out and that the fuel cladding temperature will be held below levels at which damage could occur. The staff determined that this position is acceptable as a criterion. The staff accepts that the helium atmosphere will be maintained during storage. This is based on the specified acceptance leak rate for the primary seal weld of s 0.01 kPa-cc/sec (104 atm-cc/sec), as well as on the integrity of the DSC. The confinement integrity is ensured by the use of stainless steel, thus precluding corrosion of the DSC, and also by the design criteria which include accident cases such as a drop. 10 CFR 72.236(e) requires that the cask must be designed to provide redundant sealing of confinement systems. Although not expressed as a design criterion, the standardized NUHOMS system employs redundant sealing as discussed in Section 5.0 of the report. 10 CFR 236(j) requires that the cask must be inspected to ascertain that there are no cracks, pinholes, uncontrolled voids, or other defects that could significantly reduce its confinement effectiveness. The quality standards under which the DSC is fabricated and welded provide the assurance of confinement integrity. In addition, the DSC is pressurized and leak tested after all confinement welds have been completed, in accordance with procedures described in SAR Section 10.3.4. The criteria for continuous monitoring are issues which have also been evaluated by the NRC staff. To date, under the general license, NRC has accepted continuous pressure monitoring of the inert helium atmosphere as an indicator of acceptable performance of mechanical closure seals for dry spent fuel storage casks. The NRC does not consider such continuous monitoring for the standardized NUHOMS system doQble weld seals for the DSC to be necessary because: (1) there are no known long-term degradation mechanisms which would cause the seals to fail within the design life of the DSC; (2) the posajbility of corrosion has been provided for in the design because the canister is stainless steel; (3) the creep mechanism is not plausible, because the internal storage pressure is approximately atmospheric, (4) cyclic loading has been considered, and it is below the threshold which the ASME B&PV Code Section ill has established; (5) the internal atmosphere in the DSC cavity is inert helium gas. Therefore, an individual continuous monitoring device for each DSC is not necessary. However, the NRC considers that other forms of monitoring storage confinement systems including periodic surveillance, inspection and survey requirements, and application of preexisting radiological environmental monitoring programs of 10 CFR Part 50 licensees during the period of use of the DSC canisters with seal weld closures can adequately satisfy the criteria in 10 CFR 72.122(h)(4). 2-8
- 2. 7 Instrumentation and Control Systems The staff considered 10 CPR 72.126 which provides the provision of (1) protection systems for radiation exposure control; (2) radiological alarm systems; (3) systems for monitoring effluents and direct radiation; and (4) systems to control the release of radioactive materials in effluents. The SAR takes the position (Paragraph 3.3.3.2) that, because of the passive nature of the standardized NUHOMS system, no safety related instrumentation is necessary.
Since the DSC was conservatively designed to perform its confinement function during all worst-case conditions, as has been shown by analysis, there is no need to monitor the internal cavity of the DSC for temperature or pressure during normal operations. The staff also considered 10 CPR 72.122(i) which provides that an ISFSI should have the capability to test and monitor components important to safety. The user of the standardized NUHOMS system will, as provided in Chapter 12, be required to verify by a temperature measurement, the system thermal performance on a daily basis to identify conditions which threaten to approach design temperature criteria. The user will also be reqffi *'.!d to conduct a daily visual surveillance of the air inlets and outlets as provided in Chapter 12. Therefore, the criteria in of 10 CPR 72.122(i) are satisfied. While the DSC and HSM are considered components important to safety that comprise the standardized NUHOMS system design, they are not considered operating systems in the same sense as spent fuel pool cooling water systems or ventilation systems which may require other instrumentation and control systems to ensure proper functioning. Heri<tC, due to this passive design, temperature monitoring and surveillance activities are appropriate and sufficient for this design. They ensure adequate protection of the public health and safety and meet the criteria in 10 CPR 72.122(i). Given the passive nature and inherent safety, there is no technical reason to require other instrumentation and control Jystems for monitoring the standardized NUHOMS system during storage operations. Non-safety related instrumentation that would be used within a fuel pool facility during loading, unloading, and decontamination is considered by the SAR (Paragraph 3.3.3.1) as being covered by the user's 10 CPR Part 50 liceruie. Instruments used in fuel pool facilities that would be used with DSC loading and unloading operations, and for other operations, include instruments me.asuring the boron content of the spent fuel pool water and the surface contamination and/or dose rates of the DSC and TC. Additional instrumentation that may be used in fuel pool facilities that may not already be used in current operations would provide: helium leak detection of the DSC welds, helium pressure in the DSC, and vacuum measurement of the DSC. These instruments may also be used for weld inspection. Formal NRC evaluation of instrument use within fuel pool facilities is in conjunction with 10 CPR Part 50 review of an updated FSAR and associated documentation. 2-9
Instrumentation used outside of the fuel pool facility and specifically associated with the ISFSI operations would be as follows:
- Prime mover instruments. The principal concern is that prime mover instruments be operational, support any velocity restrictions and reduce the probability of vehicle malfunction or fire.
- Measurement of the HSM surface dose rates.
- Measurement of air temperature rise through the HSM following loading.
- Measurement of hydraulic pressure for the ram with pressure gauges.
- Alignment of cask and ram with HSM using optical survey equipment.
Use of the instrumL~ts is summarily described in Sections 5 and 10 of the SAR (by statements and by inclusions SAR paragraph 10.3.5.2, 10.3.5.6, and 10.3.4.1 by reference). The staff considers the descriptions of instrumentation usage and commitment to preparation of operating procedures, which should include use of the instruments, to be satisfactory for certification.
- 2. 8 Nuclear Criticality Safety To address nuclear criticality safety, the staff considered the provisions of 10 CFR 72.124 and 10 CFR 72.236(c). 10 CFR 72.124 provides that the system should be designed to be maintained subcritical and to ensure that, before a nuclear criticality accident is possible, at least two unlikely, independent, and concurrent or sequential changes have occurred in the conditions essential to nuclear criticality safety. The design of the system must include margins of safety for the nuclear criticality parameters that are commensurate with the uncertainties in the data and methods used in calculations and demonstrate safety for the handling, packaging, transfer and storage conditions and in the nature of the immediate environment under accident conditions. The design must also be based on favorable geometry, permanently fixed neutron absorbing materials, or both. 10 CFR 72.236(c) requires that the cask must be designed and fabricated so that the spent fuel is maintained in a subcritical condition under credible conditions.
The proposed criticality safety criteria for the standardized NUHOMS system are discussed in Section 3.0 of the SAR. The standardi7.ed NUHOMS system is designed to maintain nuclear criticality safety under normal handling and storage conditions, off-normal handling, and hypothetical accident conditions. According to the SAR, the principal criticality design criteria is that ~' which includes error contingencies and calculational and modeling biases, remain below 0.95 during both normal operation and accident conditions. The design basis 2-10
accident is defined as the inadvertent misloading of the DSC with unirradiated fuel of the maximum allowable enrichment. The NRC staff considers that the proposed criteria satisfy 10 CFR 72.124 and 72.236(c) with the following conditions/observations: For the Standardized NUHOMS-24P Design
- 1. The vendor is required to include an additional constraint of limiting the initial enrichment equivalent of stored PWR fuel assemblies to 1.45 wt.% U-235 so that the optimal moderated array reactivity is less than 0.95 ( including bias and uncertainties). The initial enrichment equivalent of an irradiated fuel assembly is the U-235 enrichment of unirradiated fuel assemblies which would give the same reactivity as the irradiated fuel array. Although not included as a criterion, this constraint is-included in the proposed specification for the fuel to be stored. (See Table 12-la.)
For the Standardized NUHOMS-52B Design
- 2. The vendor is required to include a constraint of limiting the initial fuel enrichment of stored BWR fuel assemblies to 4.0 wt.% U-235.
- 3. The vendor is required to ensure a minimum fixed absorber plate boron content of 0.75 wt.% boron in the fabrication of the DSC. (See Table 12-lb.)
The staff's evaluation of the nuclear criticality safety for the standardiud NUHOMS system is included in Section 7. 0 of this report. 2.9 Radiological Protection With re.,pect to on-site protection, Section 20.1201(a) of 10 CFR Part 20 states that the licensee shall control the occupational dose to individual adults to the dose limits specified in 1201(a)(l) and 1201(a)(2). Also, section 20.1101 of 10 CFR Part 20 states that each licensee shall develop, document, and implement a radiation protection program and that the licensee shall use, to the extent pmctical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable (A.LARA). Section 72.126 provides for the provision of: .(1) protection systems for radiation exposure control; (2) radiological alarm systems; (3) systems for monitoring effluents and direct radiation; and (4) systems to control the release of radioactive materials in effluents. 2-11
Guidance for ALARA considerations is also provided in NRC Regulatory Guides 8.8 and 8.10 (References 10 and 11). For off-site radiological protection, the staff considered the requirements contained in 10 CFR 72.104(a) for normal operations and anticipated occurrences, and 10 CFR 72.106(b) for design basis accidents. In addition, the staff considered the dose limitations in 10 CFR Part 20 including the requirement that doses to members of the public must be as low as is reasonably achievable. The radiological protection design features of the standardized NUHOMS system are described in Chapters 3 and 7 of the SAR and are evaluated in Section 8.0 of this SER. These features consist of: (1) radiation shielding provided by the transfer cask, DSC, and HSM; (2) radioactive material containment within the DSC; (3) prevention of external surface contamination; and (4) site access control Access to the site of the standardized NUHOMS system array, which is a site-specific wue not specifically addressed in the SAR, would be restricted to compl:,* with 10 CFR 72.106 controlled area requirements. Based on analyses presented in the SAR (discussed in Section 8.0 of this SER), the staff concludes that the standardized NUHOMS system, if properly sited, meets the design criteria for on-site and off-site radiological protection, including the incorporation of ALARA principles. 2.10 Spent Fuel and Radioactive Waste Storage and Handling The staff considered 10 CFR 72.128(a), which provides that the spent fuel and radioactive waste storage systems should be designed to ensure adequate safety under normal and accident conditions. These systems must be designed with (1) a capability to test and monitor components important to safety; (2) suitable shielding for radiation protection under normal and accident conditions; (3) confinement structures and systems; (4) a heat-removal capability having testability and reliability consistent with its importance to safety, and (5) means to minimize the quantity of radioactive waste generated. Section 72.128(b) further states that radioactive waste treatment facilities should be provided for the packing of site-generated low-level wastes in a form suitable for storage on-site awaiting transfer to disposal sites. Criteria covering items (1) through (4) above have been addressed throughout the preceding sections in this SER in the preceding sections of this Chapter. The SAR does not specifically address the issue of minimization of radioactive waste generation. Solid wastes will likely be limited to small amounts of sampling or decontamination materials such as rags or swabs, while liquid wastes will consist mainly of small amounts of liquid resulting from decontamination activities. Contaminated water from the spent fuel pool and potentially contaminated air and helium from the DSC, which are generated during cask loading operations, will be treated using plant-specific systems and procedures. No radioactive 2-12
wastes requiring treatment are generated during the storage period during either normal operating or accident conditions. The staff agrees that the design of the standardized NUHOMS system provides for minimal generation of radioactive wastes, and that any wastes that are generated would be easily accommodated by existing plant-specific treatment or storage facilities. 2.11 Decommissioning/Decontamination Under 10 CFR 72.236(1), considerations for decommissioning and decontamination must be included in the design of an ISFSI. In this regard the staff has considered 10 CFR 72.130 that provisions should be incorporated to: (1) decontaminate structures and equipment; (2) minimize the quantity of waste and contaminated equipment; and (3) facilitate removal of radioactive waste and contaminated materials at the time of decommissioning. 10 CFR 72.30 defines the need for a decommissioning plan which includes financing. Such a plan, however, is not considered applicable to this review. The cost of decommissioning the ISFSI must be considered in the overall cost of decommissioning the reactor site. To facilitate decommissioning of the HSM, the design should be such that: (1) There is no credible chain of events which would result in widespread contamination outside of the DSC; and (2) Contamination of the external surfaces of the DSC must be maintained below applicable surface contamination limits. The SAR uses the following smearable (non-fixed) surface removable contamination limits as a limiting condition for operation: Beta-gamma emitters: 36.5 Bq/100 cm2 (2200 dpm/100 cm2) Alpha emitters: 3.65 Bq/100 cm2 (220 dpm/100 cm2) Decommissioning considerations are descn'bed in Sections 3.5 and 9.6 of the SAR and are evaluated in Section 9. 0 of this report. The staff acknowledges that decommissioning considerations are sometimes in conflict with other requirements. The reinforced structure of the HSM, for example, will require considerable effort to demolish. Although it is not likely that significant contamination can spread beyond the DSC, demolition of the HSM may generate slightly contaminat.ed dust. However, the staff concurs that primary concern in such cases rests with operational safety considerations, and ease of decommissioning is a secondary consideration. In this regard, the staff concludes that adequate attention has been paid to decommissioning in the design of the standardized NUHOMS system. 2-13
2.12 Criteria for Fuel Stability The staff considered the general design criteria set forth in Section 72.122(h) on
- Confinement Barriers and Systems.* Paragraph (1) of this section provides that *spent fuel cladding must be protected during storage against degradation that leads to gross rupture*
and "that degradation of the fuel during storage will not pose operational safety problems with respect to its removal from storage.* This aspect of the standardized NUHOMS system design is discussed in Section 2.6 of this SER. Paragraphs (2) and (3) in Section 72.122(h) relate to underwater storage of fuel and to ventilation and off-gas systems, respectively, and are therefore not considered in this review. Paragraphs (4) and (5) deal with monitoring and handling and retrievability operations, respectively, and are addressed in Sections 2.6, 2.7 and 2.10 of this document 2.13 Findings and Conclusions Tables 2.2, 2.3, and 2.4 summarize the principal design criteria for the standardized NUHOMS system components important to safety. Criteria identified in the SAR for design of the standardized NUHOMS system are acceptable with the exceptions noted below. These findings and conclusions apply to criteria and not the actual design (see "Conclusions/Discussion* paragraph at the end of each Section). Exceptions related to the HSM and its integral DSC Support Assembly and their resolution are summarized below:
- There are criteria used which may not be acceptable for all potential sites in the continental United States for: earthquake maximum ground accelerations, lightning, flood, and maximum ambient temperature. Uses of the standardized NUHOMS system at individual sites requires verification that the appropriate site parameters and that these parameters are within the acceptable design criteria.
- Fire and explosion loads are assumed to be within maximums for other included loadings. Use of the standardized NUHOMS system at a site requires examinations of potential causes and magnitudes of fires and explosions, and verification that site parameters are bounded by appropriate design criteria evaluated in this SER.
- The loads associated with a jammed DSC are acceptable; however, evaluation of those loads in a load combination expression unintended for *accidents* is not acceptable. Although the usage of the criteria is not acceptable, the staff has determined that the actual design, evaluated with the acceptable load combination expression, is acceptable.
2-14
The SAR includes a nominal design for the HSM foundation. The foundation only has nuclear safety implications in the event of gross failure, since it is structurally independent of (although loaded by) the supported HSM. Suitability of the HSM foundation design of the SAR should be verified for the actual site by a foundation analysis, or an alternative foundation design should be used for the site. The user must perform written evaluations before use to establish that cask storage pads have been designed in accordance with 10 CFR 72.212(b)(2) and (b)(3) to ensure that no gross failures occurred that could cause the standarized NUHOMS system to be in an unanalyzed situation. 2-15
TABLE 2.1 Design Criteria Sources Cited in the SAR [*OS-Docketed submittals which modify and/or extend the SAR presentation]' SAR NRC Reference Source Use Comments 3.2.5.1 ANSI/ANS 57.9-1984 Load combinations for RSM Design Acceptable 3.2.5.2 ASME B&PV Code Subsection NB 118ed for stress analysis and Acceptable (1983) Section m, Div. 1, allowable stresses for DSC shell and lids. Subsection NB and NF for Subsection NF Wied for stress analysis and Class 1 Components and allowablo stresses for DSC basket. Supports 3.2.5.2 ASME B&PV Code (1983) TC stress analysis and allowable stresses Acceptable Section m, Div. 1, excluding the lifting/tilting trunnions. Subsection NC for Class 2 Components 3.2.5.3 ANSI N14.6-1986 Allowable stresses for 1iftin& trunnions inside Acceptable N fuel builmD1. I ...... Table 3.2-1 ACl-318-83 Constroction criteria for concrete HSM. Acceptable O"I Table 3.2-1 AISC Code for Structural Steel DSC Support .A.s8embly Dosien- Acceptable for design mesees, but not load combinations. Table 3.2-1 ASME B&PV Code (1983) Allowable stre8se8 for lifting and support Acceptable Section ill, Subsection NC trunnions on--site transfer fO£ TC. 3.3.4.1. 1.A. ORNLJNUREG/CSD-2 *sCALE-3* Code Wied for Criticality Analysis Acceptable 3.3.4.1.2.A. STUDSVIK/NR-81/3 *cASMO-Z- Code for Fuel Bumup Acceptable 3.3.4.1.2.A. DPC-NE-1002A Duke Power Co. Reload Methodology Acceptable 3.3.4.1.2.A. RF-78/6293 STUDSVIK CASMO Bmcbmm: Acceptable
\
3.3.4.1.2.A. STUDSVIK/NR-81/61 CASMO Benchmark Acceptable 3.6 NUREG/CR-2397 Fuel Assembly Thermal Parameters Acceptable 3.6 ORNL/TM-7431 Fuel Assembly Thermal Parameters Acceptable 3.6 ANSI/ANS-5.1-1979 Fuel Assembly Thermal Parameters Acceptable
,/
Table 2.1 Design Criteria Sources Cited in the SAR (Continued) [*DS-Docketed submittals which modify and/or extend the SAR presentation] SAR NRC Reference Source Use Comments 3.6 A.O. Little, Inc., *Tt,ch. Supt Fuel .Aseemhly Thermal Parameteni Acceptable for Rad Stds. Hi-Lvl Rad Waste Mgt" 3.1.1.3, Thi 3.1-4& NUREG\CR-2397 Development of radiological clwacteristics Acceptable Tbl 3.1-4b NUREG\CR-0200 using ORIOEN DOE\RW-0184 3.1.2.1 Reg. "Guide 1.60 Seismic Design Response Spectra Acceptable 3.1.2.1 Ree- Guide 1.61 Seismic Design Dampin& Values Acceptable 3.1.2.1 ANSUANS-57.9-1984 Operational Handling l.oads Acceptable 3.1.2.1 ANSUANS-51 .9-1984 Accidental Drop Loads Acceptable N 3.1.2.1 ANSUANS-57.9-1984 Thormal and Dead Loads Acceptable L... ........ 3.1.2.1 Ree- Guide 1.76 Tornado Wmd Loads .Acceptable 3.1.2.1 NUREG-0800 Impact Force Criteria, Tornado Missiles, Acceptable Rocommmded Empirical Formula Ueo 3.2 ANSUANS-57.9-1984 Extreme F.nviromnental and Natural Phenomena Acceptable 3.2.1.2 ANSI ASS.1-1982 Tornado Wmd MPH to Pie8IIU.l'e Conversion Accoptable 3.2.1.2 Bechtel BC-TOP-9-A Method for Determining Impact Force for Not an accepted BOUrCe, but the Design of Local Reinforcing results of PNFS calculatiODB are acceptable. 3.2.3 10 CFR 72 Seismic Criteria and Basis for Criteria Acceptable 10 CFR 100, Appendix A Reg. Guide 1.60 Reg. Guido 1.61 I 3.2.4 ANSI ASS.1-1982 Snow and Ice Loads Acceptable
Table 2.1 Design Criteria Sources Cited in the SAR (Continued) [*OS-Docketed submittals which modify and/or extend the SAR presentation] SAR NRC Reference Source Use Comt1W1ts 3.2.S.1 ACI-349-1985 Reinforced Concrete Design Acceptable 3.3.4.1.2A .SAND 86-0151 Major neutron absorbers Reference has not been used in NRC review. The NRC accepts credit for boron in pool watec. The use of bumup credit for storage casks is not approved. 3.3.4.1.3A ANSI/ANS-57 .2-1983 Criticality Criteria Acceptable 3.3.4.1.3A ANSI/ANS-S.17-1984 Credit for Bumup Refetmce not used in NRC review. NRC accepts credit for boron in pool wattt. Use of bum.up credit for storqe casts N is not approved. I co 3.3.4.1.3A PNL-2438 Sources of Negative Reactivity Refcrmce not uacd in NRC review. NRC accepts credit for boron in pool water. Uec of bumup credit for storage casks is not approved. 3.3.4.1.3C EPRI - NP-196 Critical Experiment Reoclnnarlcs Reference not used in NRC review. 3.3.4.1.4A ANSI/ANS 8.17-1984 Double Contingency Principle Reference not used in NRC review. 3.3.4.2.2 ORNL CCC-548 "KENOSA-PC" Monte Carlo Code Acceptable 3.3.4.2.3A NUREG/CR-1784 Criticality Experime.nts Refermcc not used in NRC review. 3.3.4.2.3C NUREG/CR-0073 Criticality Separation Reference not used in NRC review.
Table 2.1 Design Criteria Sources Cited in the SAR (Continued) [*OS-Docketed submittalB which modify and/or extend the SAR presentation] SAR NRC Reference Source Use Commemts 3.3.4.2.3C NURBG/CR-0796 Criticality Experiments R.efmeoce not 1l8ed in NRC review. 3.3.4.2.3C BAW-1484-7 Criticality Experiments B&W Acceptable 3.3.4.2.3C PNL-6838 Reactivity Measurements Acceptable 3.3.4.2.3D ORNL/I'M-10902 Physical Olaract.eristics of GE Fuel Reference not used ~ NRC review. 3.3.7.1.1 PNL-6189 Fuel Cladding Temperature Limits Acceptable 3.3.7.1.1 PNL-4835 Fuel Cladding Temperature Limits Acceptable 3.4.4.1 NURBG/0612 Lifting Devices Criteria Acceptable N I Not cited NURBG/CR-1815 :Brittle Fracture Criteria for Ferritic Steel Acceptable, used in NRC <.O review.
Table 2.2 Evaluation'Design Criteria for Normal Operatin,Conditions [Columns (1) - (5) are extracted from SAR Tables 3.2-1 and 3.2-5] DcaipLoad Doaign NRC Staff' Compoocd Typo R.oferenco Puameten Applicable Codot Cnmnwrt* (1) (2) (3) (4) (S) (6) HSM Dead Load SAR. 8.1.1.5 Dead wolght lnclDding loeded DSC ANSI 57.9-1984 Acceptable, lite lrnrn terial ACI 349-85 and Ahhouah DSC ii actually * "livo ACI 3492.-85 load," DSC weight ia prechely mown. Load Combinatioo SAR. Table 3.2-5 Strqth requircmeu for peoific loed ANSI 57.9-1984 Acceptable. combination DoaliI1 Bui Normal SAR. S.1.1.5 DSC with pent ftJel rejecting 24.0 k.W ANSI 57.9-1984 Acceptable fur mott of Temperature decay heat for 5 yr coo1iDe time. Ambient Cnntinoot*I ns for lhi ayum. air temperature range~* to + 125°F. SAil 8.1.1.1 Averaae yearly ambicm tcmpccature = Thia temperature bounda IDOlt 70op reactor lite,, Normal Hmdlina Load11 SAR.8.1.1.1 Hydmilic nm load: 20,000 lb. ANSI 57.9-1984 .Acceptable, site ttnmJ1w-!al N Mnirmm load: 110 plf I Snow and l0o Loada SAil 3.2.4 ANSI 57.9-1984 Acceplable for all of couinem1 N 0 us. Livo Loada SAil 8.1.1.5 Deaip load: 200 plf ANSI 57.9-1984 Acoeplllble for all of cootJnmt-1 us forthis yatom. ShieldhJa SAil 7.1.2 Averqe oootact doee mo on HSM e.xtorloc ANSI 57 .9-1984 A.ccq,table. surfaco < 400 mrem/hr at 3 feet from HSM umce. H8M Static Load, SAR3.4.3 To be dcaljllCCI fur iDdividual itc balled on 10 CFR. Acceptable fur Certificate. Foundation itc foundation analym for tatic 1oadt 72.212(b)(2)(ll) Dry Dead Load, SAR. 8.1.1.2 Wclght of loedcd DSC: 65,000 lb. oomiml, ANSI 57.9-1984 Acceptable. Shielded 80,000 lb. cnvcloping CaniJtcr Dcalgn Buil Intcmal Preume SAR. 8.1.1.2 DSC intemal prcnure 9 .6 pllig ANSI 57.9-1984 Acceptable. Load Structural. Deman SAR. Table 3 .2-6 Service Level A and B ASME B&PV Code A.cceptablc.
- Streu Allowab!CII Sec. Ill, Div. 1, NB, CluaI
Table 2.2 Evaluation of Design Criteria for Normal Operating Conditions (Continued) Deaignl.oad Doman NllC Staff' Compoucsa Type llofenmco Puamct6l:I Applloablo Codoa Commom (1) (2) (3) (4) (S) (6) Dry Shielded Delian Bula Opentuli SAR.U.1.2. DSC decay heat 24.0 kW 11:ir 5yr ooo1ina' ANSI 57.9-1984 Acceptable. Canister Tompemw:o lo&da dmo. Ambicut a i r ~ -40*p lo (Cont'd) 125°F. Uftina imido tho apom fuel pool ASME B&PV Code Allcaptablo with teltrlotiom OD buildina ofloadc:d DSC reltlioted to-20°F Sect. DI, Div. 1, NF- WIO, ambient air tompenturo if lift height ii so 2300 SAR.10.3.15 iDchoa or !cu. If lift hoi,pt ii abovo SO SAil Table 1.1 -2 iDchca, minimum twipCtatun. la relt1ictcd NOREO/CR.-lSIS Acccptablo for impact tcltiug. to 0°F. Outlide r,ped fuel pool buildinB. maximum lift boigbt ia 80 incboa IIDd ntlnia*1111 temperature la o*p, Operatl.ona1 lhndlina SAllU.1.2 Hydraulic ram load: 20,000 lb onvcloping ANSI 57.9--1984 ~- Crlticality SAR.3.3.4 K.a, lea than 0.95 ANSI 57.2-1983 kceptable. DSC Support Load Comblnmom SAil Table 3 .2-SC Allowablo t'iu;tored llre.- b: apccifi.o I08d ANSI57.9-19M A.ocepublo, ahe im:m.aw:ial. Aaembly combmatioo:a. N I DeadLoedl SAR.8.1.1.4 i.o.ded DSC + IOlf'woight ANSI 57.9--1984 Acoeptablo, alto lnmmw:ia[. N Nomlll HandJ.ina SAil S.1.1.4 DSC hactlon load with hydnu& ram ANSI 57.9-1984 ~
- i:mmaterlal.
load: 20,000 Iba. NO!IDlil Tempenturo SAR. Table 3.2-So Factored allowable lltouca b: 'PCl¢i& load ANSI 57.9-1984 Deman oocop(ablo. combimtiooa. Stroa Evalnadoo SAR. Table 3 .2-7 Streu al.lowabloa. AISC Stool Allceptablo. Comtruction Mmmal Tranafer NOJ:lllll1 Opon.tl.ae Condition SAR. Tablo 3 .2-S Scrvico Level A IIDd B ASME B&PV Code Acoeptablo. Cuk: Streu lllowabloa Sec. DI, Div. 1 NC-3200 Structure: DeadLoedl SAil 8.1.1.9 1) Vertical orientati.oo, self weight+ ANSI 57.9-1984 Aoccptablo. loaded DSC + water in cavity: 200,000 lb. Shell, R.i:ap, otc. cnvclopinaJ. b) Horlzootal orientatioo. self wei&ht + loldcd DSC oo 1nmfcr lkid: 200,000 lb. eavclopiog. SDOW IIDd Ice Loeda SAR.3.2.4 Extcmal IUrl'aco tempenn.u:,, of cul: will 10 CFR. 72.122.(b) Acoep<<ablo. prcclade buildup of mow uid iDo load, when in mo: 0 paf
Table 2.2 Evaluation of Design Criteria for Normal Operating Conditions (Continued) Dctlpl.olld Dcllgn NRC&aff
~ Typo ~ Pa:runeten App&abl6CodN Commom (1) (2) (l) (,4) (5) (6)
TrammCut Dealp. Bula Opcnlma SAlt.11.1.U Loadod DSC rojectiD,g 24.0 tW decay beat ANSI 57.9--1984 Acceptable. Noto, Ibo riobmnn (Coot'd) TanperalUIO Low Outlido 1.1.2.2 with Syr CClOlh!g ti.mo. Amblors air handlin& ~ al Ibo Spom Fuol Pool Bulldloa ttsmporatw:o llllWO -40"F to 1.2.S °F with loaded DSC lmido Ibo TC ii I01ar lidold, -40°F to lOO*F w/o 10W ANSIN14.6 0°F (upper trmmiom HO SAR. 10.3.15 llblold. fotrltio). DNlp Bui.I Opcra!q SAR. Table 1.1-2 ~ rnbritmm handl1:os tcmpcntmo of a ASMB B&PV Cod6 Acceptablc. Tb6 ASMB B&:PV Too:pnlnl"I Loads Jmido loaded DSC imldo a TC la -10*F, fix Socd.oo.Ill,Div.1, Codo ii acccptablo with Ibo Spia Fuol. Pool Buildma height of 80 lnoboa or lea. For lift helghtl NC-1300 ~ ltatod in cblan aroator d:um 80 bw.:boa lhe n:dnimun parametel'I (upfHII' trw:llllom - handlh:ia lempo,tatllto it 0°F. Impact tcadlJI tenitl.o). SAit 10.3.15 at -40*p nqulred per SAIL ANSI N14.6 1 4.2.6 Aoceptablo f u r ~ to.mu,. ShioJdiDe SAR. 7.1.2 Avera,,o comet dolo mo u 1ban 100 ANSI 57.9--1934 Aoceptable. mremlhr. TCUppc.r ~llmdllna SAlt.8.1.1.9 a) Uppel' liting trw:mione while in ANSI N14.6-1978 .Aoceptabl6. Trwmlom Auxiliary Bulld.lna: I) Streu mut be lea 1ban yield ( atreu fix 6 1i1ma ca:klcal loadlltum:dolli oomiDa1 ii) Streu IIJlllt be a- d:um ul:timato 11trOa1 fix 10 tlmea orldoal load SAit.App. C b) Upper ~ tiwlmoDII b: omile ASMBB&PVCod6 Acceptable. tl:anlflr. Seo. Ill, NC 111,000 lb.lwmnoo Clall2 94,000 l>J.a-t 29,500 lb.ltrmml,on u:ia1 TC Opcnt1ooal Handling SAil B.1.1.9 Lower mppo.rt trwmi.om weight of loaded ASMB B&PV Cod6 Acoeptable. Lower cut dutm,g downloadiDe and tramit to Seo. ID, NC, Clul 2 Tmnnhml RSM TC Sholl Operatlooa1 llaDdllos SAR.1.1.1.9 Hydrau& ram load duo to ti:kl&n of ANSI57.9-1984 Aoeeptable. extraotiDg loaded DSC: 10,000 lb. envelopina TCBob Normal Operating SAil Tablo 3.2-9 SeIVice Lovell A, B, md C ASME B&PV Codo Acoeptablc. Ava. 1trOa lou lhan 2 s_ Scctioo Ill, NC, MAL 11:ma loa d:um 3 S.. Cius 2, NC-3100
' Xlll-1180
Table 2.3 Evaluation of Design Criteria for Off-Normal Operating Conditions [Columns (1) - (5) are extracted from SAR Tables 3.2-1 and 3.2-5] Doafanl.oed Doalgn NR.C Staff Component Typo Reference Parameten Applicable Codea C.ommeuta (1) (2) (3) (4) (S) (6) HSM otr-Nonml Temperature SAR. 8.1.1.5 -4()°F to + 125*F ambiem tomporaturo ANSI 57.9-1984 Accep1ablc for mott of Couinental lJS for thia .yum. SAR. 8.1.1.1 70°F average yearly ambient Off-Nocmal SAR. 8.1.1.4 Hydraulic ram load of 80,000 lb. ANSI 57.9-1984 Acceptablo, aito immatorlal. (Jammed Cooditioo) Handling Load Combimdoo SAR. Tablo 3 .2-5 Strcnath rcqoircmonll for lpCCific Load ANSI 57 .9-1984 ADceptablo, alto immatcri&l. Combimtions DryShwdcd Off-Nonnal Tompentme SAR. 8.1.1.2 -40°F lo 125*p ambient tempelatwe ANSI 57.9-1984 Acoeptablo for llOngO. Canittcc SAR 8 .1.2.2 UftiIJ8 of loaded DSC ICltrlctcd to -20*F ASMB B&PV Codo, Acceptablo for tramport. SAR 10.3.15 oc moro amblem air tcmporaturc foe B1b of Sec. Ill, NF-2300 N I SAR. Tablo 8.1-2 lea than 80 incboa. Outlido apoot fuel pool NURE<J/Cll-1815 Acceptablo for Impact tottina-N buildlna mn:111111111 n.ft height ii 80 incbN, w and mlnlm1111 temperature ii 0°F. Off-NOillllll PreNure SAR. 8.1.1.1 DSC intemal preaure Ion than 9.6 pq ANSI 57.9-1984 AQoeptablo. SAR. 8.1.1.2 Jammed C.oodition llmdlhia SAR. 8.1.2.1 Hydraulic ram load equal to 80,000 lb. ANSI 57.9-1984 Acceptable. DOlllinal Structural Deaip Off-NODlllll. SAR. Tablo 3.2-6 SemcoLevel.C ASME B&PV Code Acceptable. Conditiom Streu Allowablea Seo. Ill, Div. 1, NB Clul 1 DSC Support Jammed Handling Condition SAR. 8.1.1.4 Hydraulic ram load: 80,000 lb. nomiml ANSI 57.9-1984 Acceptablo, ~ Immaterial.
~ly However, med In loed combination u though an "acoklom" load. Not acceptable application of criteria.
Off-Normal Tempemturoa SAR. Table 3 .2-Sc Faotored allowable l1rcuca for apccHio load ANSI 57.9-1984 Actual doalan acoeptablo. oombinatiom. Load Combination SAR. Table 3.2-5C Factored allowahlo ltreNel foe apccific load ANSI 57.9-1984 Acceptable, llitti immaterial. combination
Table 2.3 Evaluation of Design Criteria for Off-Normal Operating Conditions (Continued) Doqnl..old Design NllC&d' Compoooat Type Re~ Panmeten Applicablo Codel Commom (1) l2) (3) (4) (5) (6) Tramf<<Cut Ol'l'-Nonnal Tompeiature SAllll.1.1.1 100°F ambient tcmponture w/o tollr ANSI S7.9-1984 ADoepalble. llhlcld, 12S *F with IO!ar llhlold Brittle fncture of Imme lleCl SAR. 8.1-2 Lowa: tempenlmo limit ii -40*p for Ute ASMB B&PV Code, ~le. trwml,ooa iDaide apom fi.l.c1 pool bulldl.oe for my lift Sec. Ill, NC-2300 without loaded DSC. Lower temperature ASME B&PV Code, liml1 fi>r loedcd DSC ii -20*F for litbl loN Sec. Ill, NF-2300 than 80 lnchea. ANSI Nl4.6 1 4.1.6 Acceptable h ~ fcalin&. Ja:mmod Condition Handlln,g SAR.1.1.2.1 Hydmtl.ic ram Iced: 80,000 Ib. oomiua1 ANSI 57.9-1984 AQoop1abl6. Structma1 Dollp Off-Normal SAil Table 3.2-8 Se.rvico Levol C ASME B&PV Code Acceptable.
~ ~ .ABc,.,,ab!ca Seo. Ill, Div. 1, NC, N CluaZ I
N ..i:,:,. Bob, Off-Nonml Cooditiom SAil Table 3.2-9 Service Level C ASM.E B&PV Code Aooeptable. Av,. 11r011 lcaa than 2 S.. Sec. Ill, Div. 1, NC, Max. 1Uea lea than 3 S.. Clua 2 NC-3200
Table 2.4 Evaluation of Design Criteria for Accident Conditions [Columns (1) - (5) are from SAR Table 3.6-3 and Paragraph 8.2.6] Doalgn Load Doaip NR.C Staff COlJ\)OllCll1 Type Rofeienco Puameten Applicable Codol Comlll6ID (1) (.'2) (3) (4) (5) (6) HSM DNlp Bull Tornado SAil 3.2.1 Max. voloo1ty 360 mph Rogulatory Oulde Acoop1able for us. 1.76 Max. wind preuure 397 pal ANSI ASS.1-1982 Load Combimtiou SAR. Tablo 3 .2-S Strength requiremeatl for apeol1lc load ANSI 57.9-1984 Acceptable, ~ bmmterial. comblnatiom l>N1p Boil Tomado Milliloa SAil 3.2.1 Max. velocity 126 mph NUltEG-0800 .ADoeptablo for us. Doea not
' Typoa: Seo. 3.5.1.4 include an NUREG-0800 Automobilo 3,967 lb. miailea but tbl.o med are Ibo B in. diam llbell, 276 lb. IDOll critical for 1ho HSM.
1 in. IOli.d llphere Flood SAil 3.2.2 Maximum water height: .so feet 10 CFR. 72.122(b) Aocep1ablo for Certification. Maxirmun velocity: 15 (pl Vodficatioa that dcllp cdleda N bound Bile puameterL I N Seiamc SAlt.3.2.3 Hmimotal ground *rcelonrion 0.251 NR.C R.egulatmy ADceptablo for Certi&ation. (.11 (boch diroctiom) Ouida 1.60 a.ad 1.61 Vedficadoo that deaign cdteda Vcdical ground accelemlon 0.17g bound lito parameten. ADcideat Condldoo SAR. B.2.7.2 DSC wbh apent fuel rejecting 24.0 kW ol ANSI 57.9-1984 Low teaqNM llllO acoeptabt. b Tempe,:&twea docay boat for 5yr cooling lime. Ambicm ronrinentJ. US. Veritictioo of lr tempenture range of -40'F to + 125°11 mnlmum lempeUtwi, for with HSM voDtl blocbd for 5 day or lea. mdlvldaal llitea. Bb;bp crlteda cceptablo wbh pproprilo dally IUIVOillmco. Fire nd Exp!OROOI SAR3.3.6 "&veloped by other doalgn ovouta,
- o.g. 10 CFll 72.122(0) Analym of potcmlal flroa nd oxploliom from any crecfiblo
- doalgn balil tornado IOOlCOI roquirod for each lite.
- dotign ham tomdo and mi.ai1o, Vorifiction that awrtion (Column (4)) bound alto parmeton.
Table 2.4 Evaluation of Design Criteria for Accident Conditions (Continued) DealpLoad Dealp NR.C Staff CompollOll1 Type lloisrence Parameten Appllcablo Codoa Commom (1) (2) (3) (4) (S) (6) HSM IJshtnlog SAll B.2.6 "ll&htnlna protection IYltem .Acceptable for Certificate. NFPA 78. (Continued) mquhemeD!I are lito apoclftc ***
- TJgbtnlng l'rotection Codo ii lo bo mod tor evaluatioo of need and doaisn of llghtnhJi protectioaatdto.
RSM S1atic load Sil.3.4.3 To be deliiJICd for indlvidual lite buod 1c, CF1t. Aooeptablo fur Certi&atc. Foundation on altc fnuud*riou analyia b ltadc 72.21l(b)('2)(ll) Nominal SAll deaip Ill' altemative de-,11 loada. lhould bo verified b individual site. DSC AocidontDrop SAll 8.2.5 Equivalent ltalU' deceler,tloo: 10 CFR. 72.122(b) Acceptable. AdmiDiltrativo coutrok mill bo 75g vertical ,end drop ~ to proveat lifting Ol' tramportina Ibo 751 homoot*l aido drop loadod DSC outido tho lpOlll f'l10l pool 2Sg comer drop with l1p down buildiDa hiper than 80 iuchol. (correapooda to 11D 80 iD0h drop beiaht) Structural ~ durlna drop: 10" Ilea, Guido 1.61 10~ damping valuo oxceedt R..G. 1.61 pdanoo. A n, valuo hu beon evaluated by Ibo lltaff and bu been acoepted. N I N Flood SAil 3.2.2 V*xhrnun water beiabt: 50 feet 10 CFll 72.122(b) Acceptable for Certification. Vedftottlon O'I required for individual ilea. Sehmic SAil 3.2.2 Hodzou J ,-md aooolentioa 0.2Sg Acooptablo. Vertical ground aoockarioo 0.171 SAil 8.2.3.2 Hodroota1 accelention: l.51 N1lC llegultory Vordcalaccelcratioo:1.0, Gwdoa 1.60 and 1.61 3" crlticaJ damping ADcideu1 lramaJ Pre.tte SAll 8.2.7.2 DSC iutemal preaure: 50.3 pq hued 10 CFR. 72.12l(b) ADceptablo. (HSM vcm blocked for 5 daya) Table 8.1....._ on 100~ ftJe1 clad IUptme and fill gu relouo, and ambicm air tomp. = 125°F DSC bell tempemare: 587°F Accidou1 Conditionl SAll Table 3.2-6 SciviceLovelD ASME B&PV Codo Acceptablo. Stren allowablcl Sec. Ill, Div. 1 NB, Claal Fire and Exploaom SAll 3.3.6 "bound by" otbec ovem, O.J, l0*CFll 72.122(0) VorificttJon that aaortion (cobnn (4)) ~ llito puamoton for bo(h tim and ~loliom.
- dolan buil tomado
- doalp bull tornado with milallca
- poatulateddrop
- oxtom1 prcaure duo to 50 feot hoad ofwator -
Table 2.4 Evaluation of Design Criteria for Accident Conditions (Continued) Dolign Load Doaip NR.C Staff' Compooent Typo R.ofmenco Paramoten Applicablo Codca Commom (1) ('2) (3) (4) (5) (6) DSC Support Sehmic SAR. 3.2.3 DSC roaction losdi with Horimntal NR.C R.ee, Ouidoa 1.60 ADcepwblo fix Certiftcadon. Voriftcatlon Aaaombly around accolomion: 0.lSg and 1.61 of critoria. Vertical ground accolcmiou: 0.l 7i Jammod Coodirion (I'roatod u SAil Tablo 8.2-12 Hydranlio ram load of 80,000 Iba. ANSI 57.9-19M Off-normal loadt mould be treatod u live a drop of 1-vy 1oed accidom in loeda (or otbor non-ecoldont catesorY) in load combination) "noaml.. load oombinatiom. Doalgn ls acocptablo. Load Combination SAil Tablo 3.2-So Factocod allowablo atrc-. tor apecific ANSI 57.9-1984 Acceptablo, lilti immatoriaL load combinatiollll. Tramfct-Cuk Doalgn Baaia Tornado SAll..3.2.1 Max. wind velocity: 360 mph NRC RoJ. Ouldo 1.76, .ADceptablo. Max. wind preaure: 397 paf ANSI 58.1-1982 N Do1ip Buis Tornado Mlniloa SAil 3.2.1 Automobile, 3967 lb. NUREG-0800 .ADceptablo. MI.no. eo&ootod bound J 8 In. diametor lholl, 276 lb. Seo. 3.5.1.4 oft'octa of other millilea In NUllEG. N Flood SAil 3.2.2 Flood oot inclodod in deaiiJ1 buii. 10 CFR. 72.122 .Acceptablo for C'-Cdifioarion. Appropriate Cuk mo to bo remicted by adminhtrativo admlniilrBtive comrola muat bo In placo controla. at any slto whoro 1loodiIJi ii a pollibDity. Sobmio SAR. 3.2.3 Horlzoutal JIOUDd accelention 0.lSg NR.C R.og. Ouidel Acoeptablo b Cortification. Verification (both dirocti.om) 1.60 and 1.61 of cdtoria required b individual. aitea. Vortical around acce1endoa 0.17a, 3 % crltioaJ. dampina Aocidont Drop SAR S.2.5 Equivaleut ltatio deceleration: 10 CFR. 72.122(b) Al:ceptablc. Adminbtrativo cootrru D1IUl 15i vertical end drop bo lmpo8cd to proved HftiDg or 75g horlrom11l ddo drop trull{>ortUli TC with loedod DSC outlido 25g come,: drop with alapdown of tho lpOll1 fuol buildlna higher than 80 (correapood to an 80 Inch drop bciaht) inchoa. Structural. damping during drop 10 % 10~ damping cxcced, :a..o. 1.61 JUidao,ce, however, 7~ hu been ovalualcd by taff' and acccptod. Scrvico LevoJ. D ASME B&PV Code Acceptable. Boha, Acoidont Drop SAR Tablo 3.2-9 Streatallowabw I Soo. Ill, Div. 1 NC, Cius 2, NC-3200
Table 2.4 Evaluation of Design Criteria for Accident Conditions (Continued) Delipl.oad Delip. NK.C Staff' Compooeiit Type hference Puameten Applicable Codol Commem (1) (l) (3) (4) (5) (6) Tramtilr Cut. StmotunIDN!p.Accidcut SAil Table 3 .2-1 s<<vioo Level D ASMB B&PV Code Accopcablc. (C'-Oatinwl) Seo. Ill, Dlv. 1 NC, Clual NC-3200 lmmall'relluro !All Table 3 .2-1 Not appl:icablo becauN DSC pn:rridol 10 CFR. 72.122(b) Accopuble. pmNme boundary. IJahtninl Not~ [N.R.C Staff: SbaDld not pemlil damtao to ~ l e , hued on IOpU1llft aft DSC or affect D S C ~ ] aml)'lil of bar.ml wbilo on lrllDlit. Fb:c md Bxplollom. !Alt.3.3.6 "Buvolopod by ollw&qn buil .-vom.* 10 CFR. n.122(0) Vtrlll,cmoo that aWl1ioo (cobmm (,4)) o.,. bound llile puameten for b<ICh fiJ:e md Ollploliom.
- dNlan bub tornado pnomcd milli1c N bd.
I N' CX)
Table 2.5 Load Combinations Used for HSM Reinforced Concrete Load Comb. Load Combination Description Correlation. to Standards NRC Staff C'.ommenhl 1.2 1.4 D + 1.7 L ANSI 57.9, Paragraph 6.17.3.l(a) Acceptable. 3,4 0.75 (1.4 D + l.7L + 1.7 H + 1.7 T ANSI 57.9, Paragraph 6.17.3.l(c) [Note: Uses W1 for W]
+ 1.7 W) .ru:ceptahle. Conservative relative to ACI 349.
Paragraph 9.2. l(S) which is also acceptablc. 5 D+L+H+T+E ANSI 57.9, Paragraph 6.17.3.l(e) Acceptable. 6 D+L+H+T+F ANSI 57.9, Paragraph 6.17.3.l(f) Acceptable. 7 D + L + H + T. ANSI 57.9, Paragraph 6.17.3.l(g) Acceptable. WHERE; N I D= Dead Weight *1.0S ANSI 57.9, Paragraph 6.17.1.1 Acceptable. N !.O L= Live Load (varied 1::let\1veen 0-100~ for worat ANSI 57.9, Paragraph 6.17.1.1 Acceptable. case) H= Latem1 Soil Pressure Loads (H taken as - 0) ANSI 57.9, Paragraph 6.17.1.1 Acceptable. W= Tornado W'md Loads NRC Reg. Guide 1.76 and ANSI Acceptable. AS8.1 T IIC Normal Cooditioo. Thermal Load ANSI 57.9, Paragraph 6.17.1.1 Acceptable. T. = Off-Normal or Accident Thermal Loads ANSI 57.9, Paragraph 6.17.1.3 Acceptable. E= Eartbquak.e Load ANSI 57.9, Paragraph 6.17.1.2 Acceptable. F= Flood Load ANSI 57.9, Paragraph 6.17.1.3 Acceptable. A= Accident (e.g.' drop accident) None 2-28
Table 2.5 Load Combinations Used for HSM Reinforced Concrete (Continued) Load Combinaticm. Description Correlation to Standards NRC Staff Comments Omitted Load Coml!i111tism!! Qf ANSI S1,9 1.4 D + 1.7 L + 1.7 H (L.C. #2) ANSI 57.9, Paragraph 6.17.3.l(b) Omission acceptable [with H=O same as LC..
#1].
0.75(1.4 D + 1.7 L + 1.7 H + 1.7 T) (L.C #4) ANSI 57.9, Paragraph 6.17.3.l(d) Omission acceptable [with H-O encompuaed by LC. #3] D+L+H+T+A ANSI 57.9, Paragraph 6.17.3.l(f) Omission would not be acceptable except that tornado missile loadings are acceptably analyz.ed, DSC Support Structure (Stroctural Steel), and that potential consequencee of accidental drop See Table 2-6 of HSM access door is not considered a nuclear safety situaticm. N I w 0
Table 2.6 Load Combinations Used for DSC Support Assembly Load Combination Description Correlation to Standards NRC Staff Comments Equation 1 S > DL + HLf ANSI 57.9, Paragraph 6.17.3.2.l(a) Acceptable. Equation 2 1.5S > DL + HU + T ANSI 57.9, Paragraph 6.17.3.2.l(d) Acceptable. Equation 3 1.6S > DL + HLf + T + E ANSI 57.9, Paragraph 6.17.3.2.l(e) Acceptable. [H = O] Equation 4 1.7S > DL + Ta ANSI 57.9, Paragraph 6.17.3.2.l(g) Acceptable. [H, L = O] Equation. 5 1.7S > HIJ ANSI 57.9, Paragraph 6.17.3.2.l(f) Acceptable. [L, H, T = 0, and dead load of support assembly is negligible] N
~
I w ...... DL= Dead Load Support Auembly including DSC wei&Jrt HLf= Normal Handlin& (tnmsfec) Loads due to friction T= Normal Thermal Load E- Seismic Load Ta= Accident Thermal Load HlJ = Off-Normal Haudling Loads due to a jammed DC including weight of'DSC (accident condition) H= Lateral Earth Pressure - 0 W= Wmd or tornado mi.Miles = 0 (Support assembly is shielded by HSM) L cs Live load not applicable when HSM is closed Omitted Load Combinatiops of ANSI 57.9 S>D+L+H ANSI 57.9, Paragraph 6.17.3.2.l(b) Omission acceptable [H = O] 1.33S > D + L + H + W ANSI 57.9, Paragraph 6.17.3.2.l(c) Omission acceptable [H, W = O]
3.0 STRUCTURAL EVALUATION Introduction This section evaluates the structural designs of the HSM, DSC, and TC. The designs are evaluated against design criteria as presented in the SAR, or otherwise determined to be acceptable (discussed in Section 2 .1). Although 10 CFR Part 72 is the basis for review, it does not specify the criteria that must be used. The staff summary and conclusions are therefore presented in terms of: (1) criteria suitability and any restricting conditions that might apply, and (2) whether or not the standardized NUHOMS system design satisfies the criteria and any restricting conditions. The structural and mechanical systems of the standardized NUHOMS system important to safety are the TC, the DSC, and the HSM including the DSC Support Assembly. Loading conditions for the individual components in the system result from all phases of normal operating conditions, exposure to natural phenomena, and accident conditions. The NRC staff evaluated all analyses for all components submitted in or with the SAR (See Reference 1). All calculations were reviewed by the NRC staff. The review included spot checks, parallel calculations, and validations of sources or expressions used. Assumed loads, material properties, and ASME, ACI, AISC, or ANSI code allowable stress limits were checked. Am,licable Parts of 10 CFR Part 72 The SAR was submitted as part of the application for a Certificate of Compliance under 10 CFR Part 72, Subpart L. Applicable design requirements are therefore stated in 10 CFR 72.236. This SER evaluation also used 10 CFR Part 72, Subpart F, for review of design bases and criteria. The guidance of Regulatory Guide 3.48 has been used for review of the comprehensiveness of the material presented in the SAR and supplementing and modifying docketed documentation. The review was performed in stages. The stages addressed: the sources of requirements and the criteria stated as constituting the basis for the design (SER Section 2.0), the structural evaluation of the actual design against the stated and other appropriate criteria (Section 3.0), and other evaluations (Sections 3 through 13). Materials The materials used for fabrication of HSM (and DSC Support Assembly), DSC, and TC are identified in the corresponding fabrication specifications and/or drawings submitted in supplement to the SAR. The mechanical properties of the materials used for the design and the sources of those properties are shown in SAR Table 8 .1-2. 3-1
The sources identified in SAR Table 8.1-2 for properties of steel are the ASME Boiler and Pressure Vessel Code, Section m-1 (Reference 9), Appendices, Code Case N-171-14 ASTM, and Handbook of Concrete Engineering by Fintel (Reference 12). The ASME Code is an acceptable standard and is in compliance with the quality standards in 10 CFR :Part 72, Subpart F. The source identified in SAR Table 8.1-2 for the mechanical properties of concrete and reinforcing steel is the Handbook of Concrete Enweering (Reference 12), a document that is not considered to meet the quality standards of 10 CFR 72.122. However, the staff has compared the data in Table 8.1-2 with ASTM specifications for steel and the pertinent American Concrete Institute specifications for concrete which do meet Subpart F standards. The staff concurs with the data in SAR Table 8.1-2. The source identified in SAR Table 8.1-2 for the structural properties of lead (Reference 13) is not considered a recognized standard that is consistent with the quality standards of 10 CFR 72.122(a). However, the material strength properties for lead were used conservatively. The staff concludes that the way the data were used meets l'°'e intent of the quality standards of 10 CFR 72.122(a) for material properties. The supplemental material provided with the SAR includes supporting design calculation packages, construction drawings, and fabrication specifications. The SER review is based on supporting design calculation packages, and summary data included in Chapters 4 and 8 of the SAR. The construction drawings and fabrication specifications were used to verify that there is a one-terone correspondence of dimensional and material property data between the drawings and the calculation packages. 10 CFR 72.3 defines structures, systems, and components important to safety which have features that: "(l) maintain the conditions required to store spent fuel er high-level radioactive waste, (2) prevent damage to the spent fuel or the high-level radioactive waste container during handling and storage, or (3) provide reasonable assurance that spent fuel or high-level radioactive waste can be received, handled, packaged, stored, and retrieved without undue risk to the health and safety of the public. " The HSM is considered as important to safety because it provides radiation shielding and protects DSCs from damage (features 1 and 2). The DSC is important to safety since it forms the secondary confinement boundary and prevents and controls criticality (feature 1). The TC is important to safety since it provides radiation shielding during transport and prevents radioactive releases (features 1, 2, and 3). The DSC and TC are also "safety related" equipment in conjunction with their use in fuel pool facilities, per 10 CFR Part 50. Evaluation of Ferritic Steels Against Brittle Fracture The standardized NUHOMS system uses ferritic steels in portions of the DSC and the TC. Because ferritic steels are subject to brittle fracture at low temperatures when movement of the component may involve an impact, the use at low temperature must be evaluated. The two components are subject to slightly different criteria due to the following reasons. 3-2
Brittle Fracture Considerations for the DSC In the case of the DSC, the brittle fracture question has two aspects. The first aspect concerns maintenance of the confinement boundary during an impact (drop accident) at low operating temperature. Because the DSC confinement boundary is manufactured entirely of SA 240 Type 304 steel, brittle fracture is not an issue. However, because the basket materials are manufactured entirely of ferritic steels, the concern is maintenance of favorable basket geometry required to ensure subcriticality. The NRC staff considers this factor to be equally important to maintenance of confinement Hence, the staff accepts NUREG/CR-1815 (Reference 7) as appropriate for brittle fracture test methods for the DSC. As described in the SAR, the basket components are designed according to the ASME B&PV Code, Section ID, Subsection NF for component supports. The basket materials shall, according to the SAR, be impact tested in accordance with the requirements of NF-2300 at - 28.9°C (-20°F). 1-iowever, the NRC staff notes that this requirement is not equivalent to NUREG/CR-1815, and therefore the staff imposes limiting conditions of operation on the use of the DSC as follows.
- 1. No lifts or handling of the DSC at any height are permissible at basket temperatures below -28.9°C (-20°F) inside the spent fuel pool building.
- 2. The maximum lift height of the DSC shall be 203 cm (80 inches) if the basket temperature is below -17.8°C (0°F) but higher than -28.9°C (-20°F) inside the spent fuel pool building.
- 3. No lift height restriction is imposed if the basket temperature is higher than
-17.8°C (0°F) inside the spent fuel pool building.
- 4. The maximum lift height and handling height for all transfer operations outside the spent fuel pool building shall be 203 cm (80 inches) rulil the basket temperature may not be lower than -17.8°C (0°F).
Brittle Fracture Considerations for the TC In the case of the TC, which serves only as a lifting and transfer device, and not as a confinement structure, the brittle fracture question only deals with the possibility of dropping the TC/DSC, and consequences of the DSC or DSC basket brittle fracture. The staff accepts ANSI Nl4.6 and NUREG-0612 (References 8 and 14) as appropriate for brittle fracture test methods for the TC. As described in the SAR, for all operations except lifting, the TC is designed and tested in accordance with the ASME B&PV Code, Section ill, Subsection NC for Class 2 Components. For critical lifts, ANSI Nl4.6 has been used. However, paragraph 4.2.6 in 3-3
ANSI N14.6, which specifies impact testing, was not used by PNFS for the trunnions or the shell, which are ferritic steel. The SAR specifies that impact testing of ferritic steels is required in accordance with ASME requirements of Table NC-2332.1-1, and that tests shall be made at -40°C (-40°F). The guidance in ANSI N14.t> for impact testing ferritic steels is more conservative than the ASME code, i.e., the nil ductility transition temperature (NDT) shall be 4.4 °C (40°F) lower than the lowest service temperature. The impact test procedure used by PNFS will, in fact, never determine the NDT. Therefore, in order to maintain the 4.4 °C (40°F) margin, use of the loaded TC will be limited to a minimum temperature of
-17.8°C (0°F) outside the spent fuel pool building.
In previous NRC SERs, on-site ferritic transfer casks have had limiting conditions of operation with regard to lift height and temperature (References 15 and 16). The staff imposes limiting conditions of operation on the use of the TCffiSC as follows.
- 1. No lifts or handling of the TCffiSC at any height are permissible at D~C basket temperatures below -28.9°C (-20°F) inside the spent fuel pool building.
(The DSC basket is limiting.)
- 2. The maximum lift height of the TCffiSC shall be 203 cm (80 inches) if the basket temperature is below -17.8°C (0°F) but higher than -28.9°C (-20°F) inside the spent fuel pool building. (The DSC basket is limiting.)
- 3. No lift height restriction is imposed on the TCffiSC if the basket temperature is higher than -17.8°C (0°F) inside the spent fuel pool building.
- 4. The maximum lift height and handling height for all transfer operations outside the spent fuel pool building shall be 203 cm (80 inches) IDlQ the basket temperature may not be lower than -17.8°C (0°F).
It should be noted that the DSC is designed to maintain the confinement boundary for drop heights of 203 cm (80 inches) or less. Thus, even if the TC trunnion were to fail due to brittle fracture, the DSC would not release any radioactive material. The only situation which might involve lift heights above 203 cm (80 inches) would be inside the spent fuel pool building, where the -17.8°C (0°F) minimum temperature shall apply and handling of the DSC is controlled by 10 CFR Part 50 requirements. Discussion of Concrete Constituents and Temperature Suitability The SAR indicates that HSM concrete temperatures might exceed ACI 349 (Reference 17) limits, i.e., the 65.6°C (150°F) limit for bulk concrete, the 93.3°C (200°F) limit for local areas for normal operation or any long term period, and 177°C (350°F) for accident or other short term period. The above limits are imposed by ACI 349 for concrete in the absence of tests to evaluate the reduction in strength and to show that the concrete will not deteriorate with or without load (ACI 349, Section A.4). 3-4
The NRC staff accepts the ACI 349 criteria and, based on separate research and analysis, also accepts the following as alternative criteria in lieu of the ACI 349 temperature requirements for ISFSis only:
- 1. If concrete temperatures of general or local areas do not exceed 93.3°C (200°P) in normal or off-normal conditions/occurrences, no tests or reduction of concrete strength are required.
- 2. If concrete temperatures of general or local areas exceed 93.3°C (200°P) but would not exceed 149°C (300°P), no tests or reduction of concrete strength are required if Type II cement is used and aggregates are selected which are acceptable for concrete in this temperature range. The staff has accepted the following criteria for aggregates (fine and coarse) which are considered suitable:
- a. Satisfy ASTM C33 requirements and other requirements as referenced in ACI 349 for aggregates.
- b. Have demonstrated a coefficient of thermal expansion (tangent in temperature range of 21 °C to 37.8°C (70°F to 100°F)) no greater than lxl0-5cm/cm°C (6xlo-6 in/in/ 0 F) or be one of the following minerals:
limestone, dolomite, marble, basalt, granite, gabbro or rhyolite. The above criteria in lieu of the ACI 349 requirements (for ISFSI only) do not extend above 149°C (300°P) for normal or off-normal temperatures for general or local areas and do not modify the ACI requirements for accident situations. For an ISFSI, use of any Portland cement concrete, where normal or off-normal temperatures of general or local areas may exceed 149°C (300°P), or where "accident" temperatures may exceed l77°C (350°P), require tests on the exact concrete mix (cement type, additives, water-cement ratio, aggregates, proportions) which is to be used. The tests are to acceptably demonstrate the level of strength reduction which needs to be applied, and to show that the increased temperatures do not cause deterioration of the concrete either with or without load. The NRC staff considered an exception to the second criteria above for the requirements for fine aggregates only. It should be noted that the HSM roof temperature is calculated to be 121 °c (250°P) on a 52°C (125°P) ambient day, for off-normal conditions, and therefore does not qualify for the following exception. This exception should not be construed as general acceptance for ISFSI usage for any normal temperatures exceeding 93.3°C (200°P) or any off-normal temperatures exceeding 107°C (225°P).
- 1. Fine aggregates composed of quartz sand, s!Uldstone sands, or any sands of the following minerals: limestone, dolomite, marble, basalt, granite, or rhyolite; or any mixture of these may be used without further documentation as to the coefficient of thermal expansion.
3-5
- 2. Fine aggregates must satisfy requirements of ASTM C33 and ACI 349, and of the documents incorporated in those by reference.
Desi1m Descriptions A description of the standardized NUHOMS system is included in Section 1 of this SER. More detailed descriptions are given in this section, where appropriate, to provide the context of the evaluation. The formal description is given by the SAR and subsequent docketed documentation provided (Reference 1). The SER is based on the fonnal description in the SAR and not on the descriptions as summarized or extracted in the SER. 3.1 Horizontal Storage Module 3.1.1 Design Description of HSM A general description of the HSM is included at Section 1.5.1 of this SER. Each HSM is essentially a monolithic reinforced concrete structure with a separate, tolted-on roof slab. The wall and roof thicknesses are dictated by radiation shielding considerations. The reinforcing steel must satisfy requirements for minimum steel as well as the strength requirements for all load combinations. Embedments must provide for attachment of the roof slab, DSC support assembly, door, TC, shield walls, and screens covering gaps between HSMs and between HSMs and shield walls. The front wall of the HSM contains a round port for DSC access which is closed by a round, shielded steel and concrete door welded in place when the DSC is in place. The roof and the front wall of the individual HSM are of sufficient strength to resist tornado missiles. The HSM is unique in the way in which the modules can be configured. They may be located singly, in single rows, or in back-to-back configurations. Shielding requirements for adjacent modules are provided by the adjacent module itse1f. For the end modules, the 0.46 m (1 ft.-6 in.) wall thickness is not sufficient to provide the required shielding alone, and an additional 0.(5() m (2 ft.) thick end module shield wall is attached to the side of the ( HSM. If the modules are located in a back-to-back configuration, the rear walls are protected by the abutting HSM. If the modules are located singly or in single rows, the 0.3 m (1 ft.) thick rear wall is not sufficient to provide the required shielding, and an additional 0.46 m (1 ft.-6 in.) thick rear shield wall is attached to the rear of the HSM. As the passive air cooling uses vents at the sides of the base unit at floor and roof levels, a 15 cm (6-in.) gap is left between adjacent HSMs and end shield walls. The shield walls have been designed to the same standards as the HSM and have been analyzed for the loadings of dead weight, live load, thermal loads, and accident loads of tornado winds/missiles, earthquakes, and floods. The resulting stresses for the end module shield wall are summarized in Table 3.1.2-1 and shown to be acceptable. Significant effects resulting from the tornado missile load are discussed in Section 3.1.2.2.A. The rear shield walls, which abut the HSM, were analyzed and shown in Table 3.1.2-1 to have lower 3-6
stresses than the end module shield walls. The tornado missile load was calculated to be within the allowable limits even while using very conservative analytical assumptions. Located within and attached to the concrete structure, the DSC support structure is a welded steel assembly which supports and restrains the DSC. It is designed to satisfy the structural loads of dead weight, seismic forces, thermally induced loads, and handling loads. 3 .1. 2 Design Evaluation The SAR was reviewed in conjunction with the calculation package NUH 004.0200 (Reference 18). The computer runs which were made to simulate the load conditions for the controlling PWR or BWR DSC design were also included in the review. 9 3.1.2.1 Normal and Off-Normal Operations A. Dead Weight and Live Load Analysis Tables 8.1-3a and 8.l-3b of the SAR provide the dead weights of both 24 PWR Spent Fuel Assemblies and 52 BWR Spent Fuel Assemblies respectively. The vendor has chosen to use a design weight of approximately 36,290 kg (80,000 lbs.), somewhat higher than the total dry DSC loaded weight of the heaviest assembly, for the analyses of the RSM and DSC support assemblies. Because the weight of the DSC is known, the vendor has chosen to treat it as a dead load. The weight of the concrete HSM is included as dead load. The weight of the steel DSC support structure is trivial. The vendor has also chosen to increase the dead load by five percent for all load combinations. The dead and live loads were applied to the finite element model depicted in SAR Figure 8.1-lOa. B. Concrete Creep and Shrinkage Analysis The vendor has cho:ien, to neglect concrete creep and shrinkage effects based on the summary analysis that thermal expansive forces would mitigate rather than aggravate the creep and shrinkage forces. This is acceptable for the HSM design as a conservative simplification. The HSM design satisfies minimum steel requirements of ACI 349-85 (Reference 17), which are partly based on creep and shrinkage considerations and which are more restrictive than the requirements for shrinkage and temperature reinforcement of ACI-318 (Reference 19). C. Thermal Loads The results of thermal analyses performed by the vendor are given in SAR Table 8.1-9b and in Figure 8.1-3a. They are derived from calculations documented in NUH004.0416 (Reference 20). For the normal operations case the thermal gradients, calculated with a long time ambient air temperature of 37.8°C (100°F), were applied to the finite element model depicted in the figure on NUH 004.0200, Rev. 5, page 10b. The thermal loads are the greatest inputs to the normal load combinations and result in the lowest margin of safety for 3-7
both the concrete HSM and the steel DSC SupJX)rt structure. In accordance with ACI 349-85 Appendix A (Reference 17), the ratio of cracked section modulus to gross section modulus is applied to the stresses obtained from the thennal analyses. The NRC staff accepts this approach. D. Radiation Effects on HSM Concrete The vendor calculated the neutron and gamma energy flux deposited in the concrete and determined these levels to have negligible effect on the concrete properties. The NRC staff accepts this determination. E. HSM Design Analysis The vendor analyzed the HSM with its DSC support structure using the ANSYS finite element analysis computer program (Reference 21) and documented the results in reference 18 [NUH004.0200]. The staff reviewed these computations included in the original SAR and in supplemental and modifying docketed material submitted subsequently and considered as part of the SAR. The final design analysis calculations were determined to be acceptable. The analysis resulted in no load cases where the margin of safety for any structural component was less than 0.1. Margin of safety is defined as the allowable load divided by the calculated load minus 1. 3.1.2.2 Accident Analysis A. Tornado Winds/Tornado Missiles Tornado forces used in the SAR treat the tornado forces as normal or off-normal wind loads. This is considered very conservative. ANSI 57. 9 (Reference 22) does not identify tornado loads. Such loads may be considered as "accident" loads and are so treated in ACI-349 (Reference 17) load combination expressions. The vendor chose to use the most severe tornado wind loadings specified by NUREG-0800 and NRC Regulatory Guide 1.76 (References 23 and 24) as the design basis for the standardized NUHOMS design. An analysis was also performed to determine whether the HSM would overturn or slide due to the tornado wind. To demonstrate the adequacy of the HSM design for tornado missiles, a bounding analysis of the end and rear modules in an array was performed. The end module shield walls were evaluated for the direct impact of a 1,799 kg (3,967 lb.) automobile having a 1.86 m 2 (20 sq. ft.) frontal area and traveling at 56.3 rn/sec (184.8 ft./sec). Upon impact, the three spacer plates at the top of the shield wall collapse, and the shield wall is expected to form a yield line along the length of the shield wall at mid-height. No damage will occur to the HSM; however, the damaged end module shield wall will require replacement. The rear shield walls which abut the HSM were also evaluated for the same impact load; no damage is expected. Both end and rear walls have been designed to meet the impulsive and impactive 3-8
requirements of ACI 349-85. The HSM was shown to meet the minimum acceptable barrier thickness requirements for local damage against tornado generated missiles as specified in NUREG-0800. B. Earthquake The standardiz.ed NUHOMS HSM was analyzed for a peak horizontal ground acceleration of 0.25 g and a vertical acceleration of 0.17 g in accordance with NRC Regulatory Guide 1. 60 (Reference 25) and a 7% damping coefficient in accordance with NRC Regulatory Guide 1.61 (Reference 26). These ground accelerations are in agreement with 10 CFR 72.102(a)(2) for sites which are underlaid by rock east of the Rocky Mountain Front except in areas of known seismic activity. Frequency analyses and response spectrum analyses were performed. The modal responses were combined in accordance with Regulatory Guide 1. 92 (Reference 27) and the directional responses were then combined by the square root of the sum of the squares method. The vendor determined that the HSM would neither slide nor overturn due to the seismic input. The NRC finds this approach acceptable and concurs with the findings. C. Flood The HSM was analyzed for a 15.2 m (50 ft.) static head of water and a maximum flow velocity of 4.6 m (15 ft/sec). For this condition the vendor showed that the maximum flood induced moment is considerably less than the ultimate moment capacity of the HSM. Further calculations showed that the HSM would neither slide nor overturn under the design load condition specified. Based on the docketed material, the NRC staff finds the results acceptable. D. Lightning Lightning protection system requirements are site specific and depend upon the frequency of
- occurrences of lightning storms in the proposed location and the degree of protection offered by other grounded structures in the vicinity. NFPA 78 Lightning Protection Code (Reference 28) is to be used for evaluation of need and design of lightning protection at the site.
E. Blockage of Air Inlet and Outlet Openings The vendor defined the design basis accident thermal event as one in which the inlet and outlet vents are blocked for 5 days with an extreme ambient temperature of 52°C (125°F) and maximum solar heat load. The HSM was analyzed for this condition referred to in NUH004.0200 as Accident Thermal (f.). The results of iliermal analyses performed by the vendor are given in SAR Table 8.1-9b. They are derived from calculations documented in NUH004.0418 and NUH004.0419 (References 29 and 30). For the accident case the thermal gradients were applied to the finite element model depicted in NUH004.0200, Rev. 5, 3-9
page 10b. The thermal loads are the greatest inputs to the accident load combination and result in the lowest margin of safety for both the concrete HSM and the steel DSC support structure. In accordance with ACI 349-85 Appendix A, the ratio of cracked section modulus to gross section modulus is applied to the stresses obtained from the thermal analyses. The NRC staff accepts this approach. F. Load Combinations The HSM is designed and evaluated for satisfaction of load combination criteria, as identified in Table 2.5, derived from SAR Table 3.2-5. These load combinations are as stated in Regulatory Guide 3. f,O {Reference 31) and ANSI 57.9 (Reference 22, paragraph 6.17 .3 .1) which are incorporated into the Regulatory Guide by reference. The load combinations incorporating tornado forces used in the SAR treat the tornado forces as normal or off-normal wind loads. This is considered very conservative. Load combinations identified in the SAR for the DSC support structure are shown in Table 2.6, derived from SAR Tables 3.2-5c an1 8.2-11. These load cc,mbinations are acceptable with the exception that the off-nonnal jammed DSC handµng load was treated as an *accident* load rather than in an expression for normal (and off-normal) loads. The actual design was checked by the staff by separate calculation. It was determined that if the loads were used in the acceptable expression, the factor of safety would still be acceptable. Therefore the design is considered acceptable despite inappropriate use of the load combination. 3.1.3 Discussion and Conclusions The maximum loads on the five major concrete structural components of the HSM (floor slab; side, front, and rear walls; and roof slab) are listed in SAR Tables 8.1-10 and 8.2-3 and in supplemental and modifying docketed material. These data were checked by the staff and found to be acceptable. Allowable loads for bending and shear are included in Table 3.1.2-1. These are derived from the structural design and analysis package, NUH004.0200 (Reference 18), which is part of the docketed material and which has been verified by the NRC staff. The staff review included independent development of load combinations acceptable to the NRC. The forces, computed by the vendor and the staff, as well as the resulting margins of safety computed by the staff for the concrete components of the HSM are included in Table 3.1.2-1 and found to be acceptable. Table 3.1.2-2 presents the results of examination of the DSC support structure stresses and load combinations. The submitted data are extracted or derived from the structural design and analysis package which is part of the docketed material (Reference 1). Table 3.1.2-2 shows analyses for the load combination for the support rails, cross beams, support columns, 3-10
and lateral tie beams of the DSC support assembly. The allowable stresses shown in the tables are those developed by the NRC staff based on the submitted calculations. The calculated maximum combined load stresses shown in the table are below the allowable stresses. The axial and bending stresses, divided by their respective allowable stresses, are further combined in order to obtain an interaction margin of safety. This combination, per the AISC Specification for Structural Steel, June 1, 1989, Paragraph Hl (Reference 32) must have a value not greater than 1.0. Review of Table 3.1.2-2 shows that the selection of the steel sections used for DSC columns, cross beams, rails, and tie beams was found to be acceptable. The rail-transverse member interconnection assembly, web stiffeners installed in the W8x35 members, and other miscellaneous HSM steel were checked and determined to be satisfactory. This included door and supports, collars, brackets, TC restraint assembly, heat - shield, seismic restraints, and end stops. The overall result of the review of the HSM and DSC support assembly structural design criteria, load combination, and final design is that the HSM and DSC support, as represented in the current docketed material (Reference 1), are considered to be structurally acceptable and meet the requirements of 10 CFR Part 72.
- 3. 2 Dry Shielded Canister 3.2.1 Design Description of Dry Shielded Canister and Internals There are two DSCs for the standardized PWR and BWR NUHOMS systems. The DSC is the secondary confinement barrier for the spent fuel. The primary confinement barrier is considered to be the fuel cladding. Each DSC will accommodate 24 PWR irradiated spent fuel assemblies or 52 BWR irradiated spent fuel assemblies. The DSC fits inside the transfer cask for handling and *transfer operations, and is moved out of the TC and into the HSM with the hydraulic ram.
The main structural parts of both versions of the DSC consist of the following stainless steel items: a 1.6 cm (5/8-inch) thick shell, a thick outer bottom cover, a thick outer top cover plate, a thin inner top plate and a thin inner bottom plate. The BWR DSC has a total of nine 3.8 cm (1.5-inch) thick spacer discs made from SA-516 ferritic steel, whereas the PWR DSC has eight 5.1 cm (2-inch) thick spacer discs. Each DSC has four 7.6 cm (3-inch) diameter spacer support rods. The PWR DSC has twenty-four square fuel guide sleeves. The BWR DSC has slots cut in the spacer discs to accept borated stainless steel poison plates. Square shaped holes accommodate fifty-two BWR assemblies. In addition to the above structural items, there are two steel shield plates, and numerous small items associated with a grapple, vent and siphon system, and lifting lugs. 3-11
The SAR was reviewed in conjunction with the calculation package NUH 004.0202 (Reference 33) plus all of the computer runs which were made to simulate all the load conditions for both PWR and BWR DSC designs. With one exception, the DSCs are designed as pressure veuels in accordance with the ASME B&PV Code Division 1 Section m Subsection NB-3000-1985 (Reference 9). Material qualifications arc in accordance with Subsection NB-2000. Fabrication and inspection are to be done in accordance with Subsections NB-4000 and NB-5000, respectively. Proof pressure tests are to be carried out according to NB-(,()()(), The exception is the weld design and inspection at the top and bottom of the DSC. The double seal welds at the top and bottom of the DSC do not comply with all the requirements for the ASME B&PV Code, Section m, Subsection NB. The inspection procedures outlined in the SAR do not comply with the code; however, the NRC staff has determined that an exception to Code requirements for volumetric weld inspection is permissible due to the following reasons:
- 1. The closure to the confinement boundary is a double-weld design, i.e., two weld joints provide confinement.
- 2. The gauge pressure (for normal operation) inside the DSC is on the order 1 psig. Therefore, pressure stresses are very low.
- 3. The test method of ensuring a gas tight seal for the inner top seal weld is helium leak detection which is very sensitive. Also dye penetrant testing will be performed at two levels including the weld root pass and cover pass on the outer seal weld to ensure no weld surface imperfections. The test method of ensuring a gas tight seal for the bottom welds consists of a helium leak test by the fabricator for the inner seal weld in accordance with ASTM FA-99, in addition to two levels of dye penetrant testing for this weld. For the outer seal weld a multi-level dye penetrant test is specified.
3.2.2 Design Evaluation for DSC 3.2.2.1 DSC Normal Operating Conditions The dry shielded canister was analyml for. (1) dead weight loads, (2) design basis operating temperature loads, (3) internal pressure loads and (4) normal handling loads. Table 3.2.2-1 of this SER summarizes all the stress analysis results for normal operating conditions. The summary table shows stresses for each DSC component for each load condition analy7.ed by PNFS and the corresponding stress as verified by the NRC staff. Each stress intensity value was compared to the allowable stress for the particular material at the stated temperature as defined by the ASME Code for Service Levels A and B conditions. All calculated stresses are below allowable levels. 3-12
A. Dead Weight Loads for DSC The dead load analysis for the DSC is presented in Section 8.1.1.1.A of the SAR. Both vertical and horizontal orientations of the DSC were considered. For the horizontal orientation the DSC inside the HSM, resting on the support rails, as well as the DSC inside the TC were modeled. The weights are shown in Tables 8.l-3a and 3b of the SAR, and stresses are shown in Table 8.1-7a and 7b of the SAR. The NRC staff reviewed these stress levels and reports them in Table 3.2.2-1 of this SER. Basically, all stresses are lower than the ASME B&PV Code allowable stresses by a substantial margin. B. Design Basis Internal Pressure The design basis normal internal pressure for the DSC is 47.6 kPag (6.9 psig), however, the analyzed pressure is 69 kPag (10 psig). This provides some conservatism in the analysis. Tables 8.1-4a and 8.1-4b of the SAR show five cases for operating and accident pressures. The ANSYS (Reference 21) finite element code was used to model the internal pressure load for the top and bottom portions of the DSC. PNFS used 345 kPag (50 psig) for the internal pressure and then multiplied the stress results by a factor corresponding to the particular load case per SAR Table 8.1-4a and 4b. Additionally these cases bound the internal pressure load of 55.2 kPa (8 psi) which exists during the helium leak test of the bottom inner seal weld during fabrication. Thus, there are two seal welds for the pressure boundary at the top and the bottom of the DSC; i.e., the weld for the outer top cover plate, and an inner weld applied to the inner top plate. The outer top cover plate is the primary structural component, and the weld at that joint is much more substantial than the weld at the inner cover plate. The pressure stresses in the weld of the top inner and outer cover plates were evaluated for normal and accident cases and found to be below the allowable limits. The same type of analysis used to evaluate the top portion was used to evaluate the bottom position of the DSC. Shell stresses were evaluated for the remainder of the DSC by an ANSYS model. The computer model used 345 kPag (50 psig) as an internal pressure load. Because of a linear response of stress to the internal pressure load, the normal, off-normal and accident pressure cases co ...1d be evaluated simply by using factors of 0:2 and 1.01, respectively. All stress intensities were evaluated and were below allowable levels for pressure stress. Tables 8.1-7, 8.1-7a, 8.1-7b and 8.1-7c of the SAR report the DSC pressure stresses for normal pressure of 69 kPag (10 psig). The staff reviewed these pressure stresses and concurs with them. The results of the SER are shown in Table 3.2.2-1 of this SER. C. Design Basis Operating Temperature PNFS has provided for axial thermal expansion of the basket assembly and the inner surfaces of the top and bottom end plates. Thus, no thermal stresses are induced due to restriction of expansion of internal parts. Similarly, PNFS has sized the spacer disc smaller than the 3-13
inside diameter of the DSC shell to preclude induced thermal stresses. PNFS performed four different finite element analyses to detennine thermal stresses for differential expansion of the shell, the spacer disc, and the sheWend cover interface. The axial thermal gradient as well as the circumferential thermal gradient for the shell were modeled using thermal input from separate temperature evaluation (NUH004.0407 Rev. 0, Reference 34). These analyses were perfonned at all ambient conditions ranging from -40°C to 52°C (-40°F to 125°F), with and without solar loads, both horizontal and vertical orientation, and with and without air gaps. The parametric study was perfonned for both the PWR and BWR designs. The maximum temperature calculated determines the material allowable stresses, and the NRC determined that 260°C (500°F) oounds all cases for the DSC shell, disc, ends, and rods. Tables 8.1-13, 8.l-13a, in the SAR report the results of the DSC temperature distribution. The thermal stresses are always defined as "secondary stresses" by the ASME B&PV Code. This means that higher allowable stresses are permitted amt only Service Level A (for normal operations) and Service Level B (for off-normal operations) need be considered. For normal operations at an ambient temperature of -40°C (-40°F), the maximum primary plus secondary stress for all thermal cases considered is 86,STI kPa (12.6 ksi) for the DSC shell. The allowable stress is 386,810 kPa (56.1 ksi). The BWR spacer disc has a thermal stress of 265,460 kPa (38.5 ksi), and, the allowable stress for the disc material is 448,860 kPa (65.1 ksi), so this is acceptable. The staff has reviewed all the documentation provided with the SAR and concurs that thermal stresses for the DSC for normal operations meet ASME B&PV Code requirements. They are shown in Table 3.2.2-1 of the SER. D. DSC Handling Stress The DSC handling load cases were divided into three groups, each requiring different analytical techniques. The design basis handling load is 50% of the DSC dry weight applied axially as it would be during normal ope. . ations when the loading ram is used to insert or extract the DSC from the HSM. The 50% factor is based on actual data obtained during the operation of a similar design at the ISFSI for the Oconee nuclear plant. Other P.'Jnnal cases are dead loads applied +/- I g vertically, +/- I g horizontally, and +/- 1 g axially, and +/- 1/2 g acting in all three orthogonal directions simultaneously. These could occur during transfer in the TC. The off-normal case is a jammed condition occurring inside the TC or HSM. All stresses in all components were evaluated and found to be below the ASME B&PV Code allowable. In addition to the confinement boundary, the grapple and lifting lugs were analyzed for the design basis loads, lx>th normal and off-normal. Both components were evaluated against ASME allowables and found to be satisfactory. The resulting stresses are much lower than allowable stresses, as shown in Table 3.2.2-1 of the SER. 3-14
3.2.2.2 DSC Off-Normal Events Three off-normal events were evaluated by PNFS for the DSC. They were off-normal pressure, jammed DSC during transfer and off-normal temperature. The off-normal temperature of -40°C (-40°F) ambient and the jammed DSC bound the range of loads. A. Jammed DSC During Transfer The basis for the postulated off-normal event, involving jamming of the DSC during transfer into the HSM, is the axial misalignment of the DSC. Should this occur, the hydraulic ram could exert an axial force equal to the weight of the loaded dry DSC, before a relief valve would prevent further load. A detailed finite element model including the actual load path through the grapple ring was performed to estimate this loading. The weight chosen by PNFS was 36,290 kg (80,000 pounds), a figure which exceeds the actual dry loaded weight, thereby affording additional conservatism. The bending stress in the bottom cover plate of the DSC is the highest stress anywhere in the DSC and is smaller than *the allowable. Also, the bending stress in the DSC shell is well below the allowable stress. These results are shown in Table 3.2.2-2 of this report. B. DSC Off-Normal Thermal/Pressure Analysis The off-normal temperature range was taken as -40°C to 52°C (-40°F to 125°F) for the DSC inside the HSM (and inside the TC). The off-normal ambient temperature of -40°C (-40°F) is the basis for the high thermal gradient for the spacer disc, and the top and bottom comers of the shell. These high thermal gradients result in high thermal stresses which are shown to be lower than the allowable stresses for secondary stress. C. DSC Off-Normal Pressure The design basis off-normal internal pressure acting in the DSC is 38.6 kPag (5.6 psig), however the value used in the analysis is 69 kPag (10.0 psig). Both inner and outer DSC pressure boundaries were analyzed for the off-normal pressure case. Because the applied value of 69 kPag (10 psig) is the same for the off-normal and the normal, the stress results shown in Table 3.2.2-2 are the same as shown in Table 3.2.2-1. D. DSC Load combination for Normal and Off-Normal Conditions Table 3.2-5a of the PNFS SAR outlines the different load combinations considered for normal and off-normal conditions and accident These conditions correspond to Service Levels A, B, C, and D of the ASME B&PV Code. Altogether, Table 3.2-5a of the SAR shows 17 combinations for all service levels. However, due to the fact that PNFS combined several combinations because normal and off-normal pressure cases are actually identical, and all thermal cases are bounded by one temperature providing the highest thermal gradient, 3-15
only nine unique combinations are shown in Tables 3.2.2-3, -5, and -7 of this SER. The staff summarized the combinations as Jescribed and finds that all stresses are below the allowables for Service Levels A and B. These are given in Table 3.2.2-3 of the SER. PNFS references are Tables 8.2-9a and 9b. 3.2.2.3 DSC Accident Conditions Section 8.2 of the SAR defines the accident conditions associated with the standardized NUHOMS system. The accident conditions which were examined for the DSC are: (1) earthquake, (2) flood, (3) accident pressure, (4) accident thermal, and (5) accidental drop of the TC with DSC inside. Of these accidents, the drop case is by far the most severe. The SAR classifies the thermal accidents, the pressure accident, and the drop accidents as Service level D conditions, and the remaining accidents including seismic and flood as Service Level C conditions. The NRC ~taff concurs with this classification. A consequence of classifying the thermal accidents as Service Level C or D is that the ASME B&PV Code does not require any stress analysis because of the ASME definition of thermal stresses as "secondaryn stresses or "self-relieving" stresses. The only required consideration of the accident thermal cases was in a reduction of material properties at the higher temperature, which was properly accounted for. A. DSC Seismic Analysis The standardized NUHOMS system is designed to withstartd seismic events which have a maximum horizontal ground acceleration of 0.25 g and a maximum vertical component of 0.17 g. These ground acceleration values are in agreement with 10 CFR 72.102(a)(2) for sites which are underlaid by rock east of the Rocky Mountain Front, except in areas of known seismic activity. NRC Regulatory Guide 1.60 (Reference 25) was used to determine dynamic load amplification factors for the horizontal and vertical directions. NRC Regulatory Guide 1.61 (Reference 26) was used to estimate the critical damping value for the DSC and the RSM. The DSC was conservatively correlated with large diameter piping and therefore has a damping value of 3 %. The DSC was evaluated for two distinct modes of vibration to establish fundamental frequencies, which in tum was used with Figures 1 and 2 in Regulatory Guide 1. 60 to estimate the amplification. The shell cross-sectional ovating mode turned out to be the only mode of interest since it is 13.8 Hz. The beam bending mode is too high to cause a dynamic amplification factor at 62.8 Hz. The resulting spectral accelerations for the DSC shell ovating mode are 1.0 g and 0.68 g for horizontal and vertical directions, respectively. PNFS applied a factor of 1.5 to these accelerations to account for a multi-mode excitation. The applicant used the results of 75 g vertical drop analyses factored by (1.5 x 0.68/75) to obtain stresses for the DSC. For the horizontal orientation, PNSF used the results of the horizontal drop analysis factored by (l.5xlx2/75) to obtain the DSC stresses. DSC shell stresses 3-16
obtained from vertical and horizontal analyses are summed absolutely. These are recorded in Table 3.2.2-4. The DSC was also evaluated for roll-out of the support rails. Horizontal and vertical accelerations of 0.37 g and 0.17 g were applied to the center of the DSC. The resulting factor of safety against roll-out was 1.23 according to an NRC staff evaluation. This corresponds to 1.30 as calculated by PNFS (Reference 1). B. DSC Flood Condition The design basis flood is specified in the SAR as 15.2 m (50 feet) of water with a maximum flow velocity of 4.6 mis (15 feet per second). The flood condition is postulated to occur only when the DSC is housed inside the RSM. The consequences of the water flow will not affect the DSC inside the HSM, and the consequences to the HSM are reported in another section of the SER. Therefore, the DSC is only affected by the static head. The DSC shell and outer cover plates and inner cover plates were modeled with a finite element analysis. PNFS modeled both inner and outer cover plates coupling the nodes of both plates to allow transmission of forces perpendicular to the plates. A more conservative approach would have been to assume no inner plates. However, the resulting stresses due to the 149.6 kPa (21. 7 psi) external pressure are so small ( - 6,895 kPa (1 ksi)) that even if this conservative approach had been used, the resulting stresses would still be lower than the allowables. See Table 3.2.2-4. C. DSC Accident Pressure The bounding DSC internal accident pressure is 379.7 kPag (50.3 psig) according to Section 8.2.9 of the SAR. The maximum ambient temperature of 51°C (125°F) is assumed. This accident is postulated for a DSC inside the HSM, which has all inlet and outlet vents blocked, i.e., the adiabatic heat-up case. Assumptions are that the cladding of all fuel rods failed and that 100% of the fill gas and 30% of the fission gas are released ir..,ide the DSC. Under these conditions, the internal pressure could reach 379.7 kPag (50.3 p~;g). Table 3.2.2-4 of this SER shows the stress results of this case. All stress intensities are lower than the allowables. The heat-up time period which is postulated by PNFS is five days. At that time the fuel cladding temperature is still below the cladding limit of 570°C (1058°F) for accident conditions. The adiabatic accident case bounds all thermal accidents and shows the need for daily inspection of air inlets and outlets. A more complete discussion of thermal performance may be found in Section 4 of the SER. It should be noted that PNFS stated that 204.4 °C (400°F) is the appropriate temperature to select the allowable stresses for the materials in the DSC (NUH004.0202 p.161). Table 8.2-9e of the SAR stated that 260°C (500°F) was the correct temperature. However, Table 8.1-3-17
13 of the SAR indicates that the DSC shell reaches a maximum temperature of 303.9°C (579 °F) for this accident; therefore, the NRC staff used lower material allowable stresses for this case. D. DSC Load Combination (Thermal Accident) for Service Level C Accident Conditions Tables 8.2-9c and -9d of the SAR show the results of two load combinations. These are the enveloping load combinations defined in Table 3.2-5a of the SAR. Table 3.2.2-5 in the SER shows the results of the three unique load combinations of Cl, C2, and C7. There is a slight discrepancy between the allowable stress as reported in the SAR and as reported in this SER. The discrepancy arises because the maximum DSC temperature, as reported in the SAR is 304.4 °C (580°F) for the accident pressure case; whereas, the allowables used in the SAR
- were based on 2(i()°C (500°F). The NRC staff used the allowables associated with 304.4 °C (580°F). The staff has recorded the results in Table 3.2.2-5 of the SER. The conclusion is that the design for the DSC is adequate.
E. Accidental Drop of TC with DSC Because the cask drop accidents postulated in the SAR cause the highest stresses in the both the DSC and the transfer cask, it is appropriate to discuss the basis for selecting some of the parameters and assumptions for this case. All drop situations that were postulated in the SAR involve dropping the TC, with the DSC inside, at a maximum height of 203 cm (80 inches). The NRC staff considers these assumptions reasonable, because the loaded DSC will always be in the TC or inside the RSM whenever it is outside of the spent fuel pool builcling. The centerline of the RSM is located at 259 cm (102 inches) above the base pad; and therefore, the maximum drop height would be about 173 cm (68 inches) for the DSC, should it fall off of the transport trailer during loading or during transport between the spent fuel pool building and the ISFSI site. Thus, 203 cm (80 inch) drop is conservative. Discussion of PNFS Desi~n Methodolo~y One of the major cornerstones of the PNFS justification for the deceleration levels associated with the postulated cask drop accident is a research report published by EPRI (Reference 35). This work has attempted to correlate average deceleration values acting on the cask as a function of several parameters, including drop orientation, drop height, and concrete target hardness. The latter is a non-dimensional variable which includes the following parameters: concrete elastic modulus, concrete ultimate strength, soil elastic modulus, soil ultimate strength, steel reinforcement ratio and footprint of cask. The cask itself is considered to be infinitely rigid, so that from an absorbed energy standpoint, all kinetic energy would be absorbed by the target. The resulting cask deceleration values would represent an upper bound compared to an assumption which pennitted the cask to absorb any elastic or plastic energy as a result of the impact. The EPRI report NP-4830 (Reference 35) was supplemented by a second report NP-7551 (Reference 36) which correlated a small sample of 3-18
existing experimental evidence of cask drops to the analytical presentation made in the NP-4830 report. The magnitude of the deceleration for each drop case was selected as the design criteria in Section 3 of the SAR as 75 g for either vertical or horizontal drop orientations and 25 g for the comer drop. The SAR values are based on an EPRI report (References 35 and 36). The target chosen for this scenario is a 91 cm (36-inch) thick under-reinforced concrete slab. PNFS argues that drop accidents which might occur while the DSC, inside the TC, is enroute to the HSM, would be less severe than a drop accident of the DSC/TC on the reinforced concrete pad/apron adjacent to the HSM. Their argument is based on the fact that the road which would be used as the route between the spent fuel pool building and the HSM location would typically be 30 cm (12 inches) or less of concrete or asphalt on a compacted gravel bed. The "target" or impact surface would thus be significantly "softerw than the loading/unloading approach slab near the HSM. The thickness of this slab is not specified by PNFS, but would be no-thicker than 91 cm (36 inches) and would be designed in accordance with ACI-318-83 (Reference 19). Because references 35 and 36 do not document the deceleration time history, it was necessary to establish damping coefficients and the representative time histories for the three orientations, in order to predict appropriate dynamic load factors (DLF). The SAR provided additional material in Appendix C that included references to drop test data for a 81.6 t (90-ton) rail cask (Reference 37). The time histories from this reference were used to determine the DLFs for the different drop orientations. Based on the documentation provided and the references cited, NRC staff concludes that the DLFs for the vertical, horizontal, and comer drops are 1.50, 1.75, and 1.25, respectively. These factors, when multiplied by the unfactored deceleration levels obtained from reference 36, produced values of 73.5 g, 66.5 g, and 25.0 g for the three drop orientations, which compare favorably with the deceleration values of 75 g, 75 g, and 25 g selected by PNFS in their design criteria. The staff determined that a damping value of 7 % is conservative. This was based on sources in the open literature as well as the information provided by PNFS. Discussion of NRC Staff Evaluation of Accidental Drop The vendor's use of the EPRI report methodology (References 35 and 36) to determine design deceleration loads is not currently endorsed by the NRC. Therefore, the staff independently calculated elastic and plastic strains associated with the absorption of the kinetic energy resulting from dropping a fully loaded DSC through 203 cm (80 inches). The height of 203 cm (80 inches) is conservative because the DSC is not raised more than 173 cm (68 inches) during all transfer operations outside the spent fuel pool building. The total strain rating on the DSC shell was calculated to be 1.28% compared to the minimum 40% elongation (strain) required for the material by the ASME B&PV Code, 3-19
Section IT, Part A. Thus, the strain due to a vertical drop is a very small fraction of the total strain capacity. The general membrane stress in the DSC shell W'.iS calculated from the strains and the moduli of elasticity and found to be equal to 142,730 kPa (20.7 ksi). This value is well below the ASME Code allowable of 289,590 kPa (42 ksi) for Service Level D conditions. Therefore, the factor of safety, as determined by the energy method is slightly in excess of 2 for the general membrane stress. Based on these independent calculations, the NRC staff confirmed that the design of the DSC will provide ample margin of safety during a drop accident. In addition to the independent analysis described above, NRC staff evaluated all the design calculations submitted by PNFS and reported the results in Table 3.2.2-6. These design calculations were performed using an NRC-approved finite element computer program and are described in the next section of this SER. The staff concluded that a drop of the loaded DSC from a height greater than 38 cm (15 inches) may cause damage to the DSC and the stored fuel. Because the ASME Code, Section ill for Service Level D permits plastic deformation, portions of the DSC shell and basket may sustain damage, without compromising the confinement boundary or geometry of the spent fuel array. However, such potential damage is cause for limiting conditions of operation and surveillance.
- a. The loaded DSC/TC shall not be handled at a height greater than 203 cm (80 inches) outside the spent fuel pool building.
- b. In the event of a drop of a loaded DSC/TC from a height greater than 38 cm (15 inches) (a) fuel in the DSC shall be returned to the reactor spent fuel pool; (b) the DSC shall be removed from service and evaluated for further use; and (c) the TC shall be inspected for damage. The affected fuel may be subsequently transferred to dry storage if it meets the requirements for storage.
The DSC may be returned to service or disposed of depending on the results of the evaluation. The TC may also be returned to service if the sustained damage is repairable. These conditions are reflected in the conditions for system use, Section 12.2.10. F. Discussion of Finite Element Models for Cask Drop Reference 33 is a calculation package which presents all the structural analyses for the DSC. Together with 33 separate ANSYS computer runs, the calculation substantiates the design of the PWR and BWR DSC. NRC staff evaluated the entire package. In all cases, the SAR uses the ANSYS (Reference 21) finite element code to model the DSC and TC cask components. PNFS ran nineteen models for all drop cases. These cases included both PWR 3-20
and BWR DSCs, in vertical and horizontal orientations. E.ach part of each DSC modeled. For the vertical drop, an axisymmetric load and an axisymmetric geometry were modeled, using an equivalent 75 g static load. For the horizor.tal drop case, an axisymmetric structure with non-axisymmetric loading was modeled. The asymmetrical loading was approximated with a Fourier series technique in conjunction with an ANSYS element type designed to facilitate the use of the Fourier (harmonic) series. PNFS did not model the comer drop because they stated that the 75 g vertical and 75 g horizontal drop orientations are bounding for the 25 g comer case. PNFS mcxleled thirteen cases for the horizontal drop orientation and seven cases for the vertical drop orientation. The shell, top cover plate and inner top plate were modeled using axisymetric geometry for top end drops. The shell, bottom cover plate and inner bottom plate were also modeled using axisymetric geometry for bottom end drops. The PWR loaded basket mass is greater than the BWR basket mass, therefore, the loads and the stresses are greater for the PWR DSC than for the BWR DSC. The shell, top and bottom cover plates, and top and bottom inner plates were modeled for 3 horizontal orientations, i.e., 0°, 18.5°; and 90° azimuth oriented upward. These orientations correspond with possible drops at 0° and 90° azimuth for the DSC inside the TC falling off the transfer trailer, and 18.5° for a TC/DSC falling directly onto one of the cask rails on the support trailer. Because the spacer discs for the PWR and BWR DSCs are completely different parts, it was necessary for PNFS to analyze both types of discs and both variations of plate thickness. For the 75 g vertical drop cases, the loads acting on the PWR spacer discs include the support rods, guide sleeves and oversleeves. For the 75 g vertical drop cases, the loads acting on the BWR spacer discs include the support rods, poison plates and poison plate support bars. An elastic plastic analysis using classical bilinear material properties with a conservative tangent modulus of 5% was used by PNFS for the top spacer disc for the 52-B DSC. The stresses were evaluated according to the ASME B&PV Code requirements and found to be acceptable. The calculation package reported the results of each .,omputer run and summarized the results by listing the highest stresses in a particular comr')nent for the given drop orientation. E.ach type of spacer disc was also modeled separately for two side drop orientations. The two orientations used for the 75 g 24-P basket were 0° and 90° azimuth upward. Elastic-plastic analysis was used to account for local yielding. Three azimuth orientations were used to model the 52-B spacer disc in the horizontal drop case. Elastic-plastic material properties were specified to predict stresses and displacements more accurately than using simple elastic properties. Again, all of the stresses which were calculated for each of the separate runs were reported by PNFS and evaluated by the NRC staff. Summary tables showing only the maximum stresses were provided in the calculation package. 3-21
Buckling analyses were performed on the spacer disc and support rods for vertical drop orientations, and for the 52-B poison plates in the side orientation. There is a factor of safety of 2.2 for the 52-B spacer disc out-of-plane, 1.8 for the 24-P spacer disc out-of-plane, 8.36 for the 24-P support rod (vertical drop), and 1.37 for the vertical drop case for the 52-B poison plates. The support rods were analyzed for compliance with stress levels below allowable stress levels for the side drop and vertical end drop orientations. The top end drop resulted in a much higher stress level than the horizontal drop; however, the stress is below the allowable stress level. Table 3.2.2-6 of the SER presents a summary of all the results of the 20 ANSYS computer analyses which were outlined above. Both vertical and horizontal results are given. The table shows that all calculated stress levels are below the ASME B&PV Code allowables for Service Level D. The table lists results that PNFS reported in the calculation package and the results obtained by NRC staff evaluation of the calculation package including all of the computer runs. The reader of Table 3.2.2-6 will note some differences between results of PNFS and NRC staff. These differences are attributed primarily to location of stresses in the actual models. NRC staff has consistently been more conservative than PNFS, and still the stress levels are lower than allowable levels. The weld stresses for the critical secondary confinement joints between the top outer cover plate and the shell and the bottom outer cover plate and the shell were also evaluated by NRC staff. A joint efficiency factor of 60% per ASME B&PV Code ND-4245a(3) was used for a Class C, Type 3 weld, which is non-volumetrically examined. Table 3.2.2-6 shows the results of the individual load cases. All of the calculated stresses for the welds are below the allowable stress levels. G. DSC Load Combination for Service Level D Accident Conditions The SAR uses Service Level D for accident case allowable stresses. While NRC s:aff concurs with this decision, it must be coupled with the operating controls and limits as proposed in Section 10 of the SAR. Following a cask drop of 38 cm (15 inches) or greater, the DSC must be retrieved, and the DSC and the internals must be inspected for damage. NRC staff sets this operational control because it is in keeping with the high allowable stress of the Service Level D, i.e., permanent deformations of the DSC confinement boundary and the DSC internals are permitted under Service Level D conditions. Additionally, given the predicted failure of the weld between the guide sleeve and spacer disc at a deceleration below the 75 g level predicted, there is justification for inspection of the 24-P DSC and internals following any cask drop of 38 cm (15 inches) or greater. Note that for the 52-B DSC basket, the poison plates are designed to remain in place during the postulated drop accidents. Therefore, there is no possibility of an unanalyzed geometry in the poison plates with regard to subcriticality. 3-22
In both of the load combination cases D2 and D4, the term Pa, or accident pressure, is used. The maximum temperature associated with the maximum internal pressure of the DSC shell is given as 304 °C (579°F) for the adiabatic heatt1p inside the HSM. However, this temperature is not reached when the DSC is inside the TC, which is the only scenario considered for drop. The maximum DSC shell temperature for the DSC inside the TC appears to be 231 °C (447°F) from Figure 8. l-3b. This temperature is higher than 211 °C (411 °F) given in Table 8.1-13 for the DSC inside the TC. Therefore for the load combination cases of D2 and D4 which include accident pressure, the material properties for a DSC temperature of 260°C (500°F) is used to be conservative. Also conservative is the fact that the maximum pressure stress is used in cases D2 and D4; however, this pressure occurs under adiabatic HSM heatup conditions which are not possible with the DCS inside the TC. Table 3.2.2-7 summarizes the results of the bounding two load combinations, D2 and D4. As noted in the table, the stress intensities are conservatively combined irrespective of location in the DSC, unless otherwise noted. For the case of the DSC shell, this conservative procedure was not used. However, the ASME B&PV Code requires that the stress intensities at any point for all load combinations shall be lower than an allowable stress; i.e., it is not required to combine stresses irrespective of location. As may be seen from Table 3.2.2-7, these conditions are met. 3.2.2.4 DSC Fatigue Evaluation Section NB-3222.4a of Section III of the ASME B&PV Code (Reference 9) requires that components be qualified for cyclic operation under Service Level A limits unless the specified service loadings of the components meet all six conditions defined by NB-3222.4d. Although it is superficially clear that the DSC is inherently not subjected to high cycles of pressure, temperature, temperature difference, or mechanical loads, the Topical Report (fR) (Reference 38) previously evaluated each of the six conditions defined by the ASME B&PV Code in the submittal of the TR. NRC staff evaluated the previous analysis and concurs with the finding that the service loading of the DSC meets all conditions (Reference 39). Therefore a separate analysis is not required for cyclic service. NRC staff does not find any basis for finding that the service loading will deviate from the conditions assumed in the SAR, consequently no fatigue analysis is required. 3.2.2.5 DSC Corrosion The suitability of stainless steel casks for containment of spent fuel was reported in "Laboratory Experiments Designed to Provide Limits on the Radionuclide Source Term for the Nevade Nuclear Waste Storage Investigateion's Project" by V. M. Overby and R. D. McCright. (SAND 85-0380, Reference 71). Stainless steel performs well in oxidizing conditions. Average oxidation rates for stainless steel are: 0.15 micrometer/yr submersed in water at temperatures in the 50°C to 100°c; 0.16 micrometer/yr in 100°c saturated steam, and 0.001 to 0.08 micrometers/yr in 150°C unsaturated steam. The oxidation environment of 3-23
the DSC in the standardized NUHOMS system is expected to be less than the above conditions. It is therefore concluded th1t even under worst oxidation condition, the corrosion depth of the DSC for a 50 year design life is insgnigicant and will not affect the DSC from performing its intended safety functions. 3.2.3 Discussion and Conclusions for DSC Tables 3.2.2-1 through 3.2.2-7 have summarized the structural evaluation of the DSC for Service Levels A, B, C and D. All of the results show that the DSC complies with all of the requirements for ASME B&PV Code, Section ill, Subsections NB and NF (Shell and basket respectively). This evaluation includes many conservatisms as discussed in Section 3.2.2.
- 3. 3 Transfer Cask
- 3. 3 .1 Design Description of Transfer Cask The TC is used to house the DSC inside of the spent fuel pool building and during transport operations between the spent fuel pool building and the HSM. It is designed to provide radiological shielding during all operations when the DSC has spent fuel in it. It is also designed to provide protection to the DSC against potential natural and operational hazards during transport of the DSC to the HSM.
The main structural parts of the TC consist of the following items: a 3.8 cm (1.5-inch) thick shell, top and bottom machined rings which join the shell to a 5 .1 cm (2-inch) thick bottom cover plate, and a 7. 6 cm (3-inch) thick top cover plate. Some of these items are stainless steel and other are ferritic steel. For lifting and transporting purposes, two ferritic steel upper trunnions are welded to the structural shell. For tilting and transport purposes only, two stainless lower trunnions are welded above the centerline of the structural shell. The shell its.:!lf may be either stainless steel or ferritic steel, depending on fabricator's option. The top cover plate is fixed to the top structural ring with sixteen 1. 75-8 UNC bolts. The payload of the TC is 40,826 kg (90,000 pounds) and the total gross weight with fuel and water but no top lid is 89,566 kg (197,500 pounds) enveloping 84,186 kg (185,600 pounds) with fuel, top lid but no water. These values are for the 52-B DSC assembly, which are slightly larger than the TC for the 24-P design. However, the load used for the analyses is 90,700 kg (200,000 pounds), thus providing a small conservatism. The TC is classified as "important to safety" and has been designed to meet several criteria depending on the function. The primary function of transporting the DSC inside the TC is covered by the ASME B&PV Code, Section ill, Subsection NC for Class 2 components. Load combinations have been extracted primarily from the ASME B&PV Code. The lifting and tilting trunnions have been designed to meet ANSI N14.6-1978. Table 3.2-1 of the SAR provides a complete summary of the design criteria. Material qualifications are in 3-24
accordance with Subsection NC-2000. Fabrication and inspection are to be done in accordance with Subsection NC-4000 and NC-5000, respectively. The review of the structural integrity of the TC is presented according to function, i.e., either transfer function, or lifting/and tilting function. The ASME Code governs for transfer, whereas ANSI Nl4.6 governs for the lift and tilt trunnions. The version of the TC which is specific for the PWR fuel has previously been evaluated by the NRC staff (Reference 16). However because the standardized NUHOMS system design heat load for 5-year-old PWR fuel is 24 kW instead of 15.8 kW, PNFS performed new thermal stress analysis which is evaluated in this SER. Additionally, because the BWR fuel is typically longer than the PWR fuel, PNFS provides a collar which is to be bolted on the top of the PWR TC. The stainless steel collar is designed in accordance with the ASME B&PV Code Section III, Subsection NC. Material Considerations This SER previously discussed the possible brittle fracture of ferritic steels. Paragraph 4.2.6 of ANSI N14.6 establishes low temperature criteria which are acceptable to the NRC staff for the TC design. This criteria was not used by PNFS. Instead the vendor used a test temperature of -40°C (-40°F) and an impact test procedure according to the AS:ME B&PV Code, Section III, Subsection NC. As a consequence of not using ANSI N14.6, the use of the TC shall be restricted. Inside the spent fuel pool building, for DSC/TC lift heights above . 203 cm (80 inches), the minimum temperature shall be -17.8°C (0°F) or higher. For DSC/TC lift heights of 203 cm (80 inches) or lower, inside the spent fuel pool building, the minimum temperature shall be -28.9°C (-20°F) which corresponds to the minimum DSC basket use temperature. For all use outside the spent fuel pool building, the minimum temperature shall be -17.8°C (0°F). This corresponds with ANSI N14.6 paragraph 4.2.6, as well as the SAR paragraph 10.3.15. (Note all TCs have ferritic steel trunnions which are part of the lifting device.) 3.3.2 Design Evaluation of the Transfer Cask 3.3.2.1 TC Normal Operating Conditions The calculation packages concerning the stress analysis for the TC are contained in several PNFS submittals because the Standardized TC derives from a design that was first evaluated by the NRC staff in the NUHOMS 24P Topical Report. (References 3, 4 and 40). There have been some additional analyses performed in conjunction with the TC for a site-specific application (Reference 41, calculation package BGE 001. 0202 Revision 4), and the current submittal has two calculation packages NUH 004.0205 and NUH 004.0206 (References 42 and 43) which deal with the BWR collar and a new thermal stress analysis. All of these packages have been evaluated, and the summary tables in this SER re11ect portions from all above analyses. It should also be noted that, compared with previously reviewed submittals, 3-25
portions of drawings for the Standardized TC differ dimensionally and with respect to material options. NRC staff has taken the various combinations into consideration in this SER. The TC was designed for: (1) dead weight loads, (2) design basis thermal loads, and (3) handling and transfer loads. Table 3.3.2-1 of this SER summarizes all the stress analysis results for normal operating conditions. The summary table shows stresses for each TC component for the three load conditions analyzed by PNFS and the corresponding stress as verified by NRC staff. Each stress intensity was compared to the allowable stress for the particular material at the operating temperature as required by the ASME Code for Service Levels A and B conditions. For TC parts which allow more than one material specification, the lower allowables are listed in Table 3.3.2-1 and all subsequent SER tables. A. Deadweight Loads for the TC The TC is evaluated for two dead weight loads, e.g., a fully loaded cask hanging vertically from its two lifting trunnions and a fully loaded cask supported horizor.tally from its trunnions at top and bottom ends of the TC on_the transport skid. Review of calculation package NUH 004.0206 (Reference 43) indicates that the dead weight loads are trivial when compared to the stress allowables. The results are broken out by orientation, i.e., vertical, horizontal or corner in Table B-1 of NUH 004.0206. All the calculations were made using ANSYS runs for the three orientations (75 g vertical, 75 g horizontal and 25 g corner drops) and then factoring the drop accelerations to obtain 1 g for dead weight. B. Thermal loads for the TC Calculation Package NUH 004.0206 presents the thermal stress analysis associated with the TC. The normal temperature range is considered to be 17.8°C to 37.8°C (0°F to 100°F} and t.ie off-normal excursions go to -40°C (-40°F} and 52°C (125°F) for the 5-year-old PWR and BWR fuel assemblies. The TC has been analyzed for the combined effects of the worst case radial, axial, and circumferential thermal gradients. Tables 1, 2 and 3 in the calculation package show that the 5-year-old PWR fuel on a 21.1 °C (70°F} ambient day causes the highest gradient for radial and axial surfaces, whereas the 10-year-old PWR and BWR fuel cause the highest circumferential gradient for 21.1 °C (70°F} ambient day. For ambient temperatures above 37.8°C (l00°F), use of a solar shade for any operations involving the TC is required, because the 37.8°C (l00°F} ambient temperature with solar load is outside of the design envelope for the neutron shielding of the TC. The effects of dissimilar materials has been accounted for in the analyses by modeling the material properties of all four structural and non-structural (shielding) materials. 3-26
Table 5 of the calculation package summarized the results of three ANSYS runs, the top end axial, bottom end axial, and circumferential thermal stresses. PNFS conservatively summed the thermal stress intensities for axial and circumferential orientation for the shell regardless of actual location in the shell. The staff notes that Table 8.1-lOa in the SAR does not reflect the results of the latest thermal analyses as provided in the calculation package NUH 004.0206 (Reference 43). However, the results as given in the SER in the Table 3.3.2-1 and 3.3.2-2 do incorporate the appropriate material. Also the tables in the SER incorporate different material allowables to account for the least strong material which could be used according to the PNFS drawings. C. Operational handling loads for TC As described in the dead weight load section above, there are two normal operation handling cases for the TC: vertically supported by the crane, and horizontally supported by the skid. The former is governed by ANSI Nl4.6 rules (Reference 8) and the latter is governed by the ASME B&PV Code (Reference 9). The ANSI code is concerned with critical lift loads and consequently only addresses the lifting trunnion design and the TC shell in the vicinity of the lifting trunnion. The evaluation of this aspect of normal handling is discussed in a subsequent section of 3. 3. 2 of this SER. The actual transportation and transfer handling cases which are considered are 1 g vertical, 1 g horizontal, 1 g axial, and +/- 1/2 g applied simultaneously in all three directions. Table 3.3.2-1 of this SER summarizes the results of stress analysis for the TC shell and top and bottom rings and cover plates. All results for the normal handling case are satisfactory for Service Level A. Loads acting on the upper and lower trunnions are discussed in a subsequent portion of this SER. D. TC Load Combinations for Normal and Off-normal Conditions Table 3.2-5b of the SAR defines the different load combination for normal and off-normal events. These conditions correspond to Service Levels A and B of the ASME B&PV Code. There are five Level A conditions and two Level B conditions. Table 3.3.2-2 of this SER has combined the conditions as follows. Load combination 1 combines the worst case of load cases Al, A2, A3, A4 and Bl, and load case 2 combines the worst case of Al, A2, A3, A5, and B2. Note that thermal stresses are the same for all cases, i.e., 21.1 °C (70°F) ambient day for 5-year-old PWR fuel for radial and axial gradients, and 10-year-old PWR fuel for circumferential gradients. Cases Al, A2, and A3 are all exceptionally low. Case A4 and Bl correspond to TC transport outside the fuel pool building, and A5 and B2 correspond to transfer of the DSC into/out of the HSM. Calculation Package NUH 004.0205 (Reference 42)describes the structural evaluation of the BWR cask collar. PNFS evaluated transfer loads as well as accidental drop loads for the collar, bolts and welds. Of these loads, only the transfer loads relate to Service Levels A and B. In order to estimate the dead load and thermal load, PNFS argued that loads would 3-27
be similar to those of the TC without the collar in the vicinity of the top structural ring. The NRC accepts this position based on similar geometry. In all cases, the allowable stresses were evaluated for a material temperature of 204 °C (400°F), a conservative value. As shown in Table 3.2.2-2 all the stresses are lower than the allowables. 3.3.2.2 TC Accident Conditions Section 8.2 of the SAR defines the accident conditions that affect the transfer cask. These conditions are: (1) earthquake, (2) accidental drop of the TC with the DSC inside, and (3) tornado wind loads. Tornado generated missiles, although not discussed in the SAR, was the subject of an NRC staff concern. It is addressed by PNFS and evaluated in this SER. A. TC Seismic Condition The design basis earthquake for the standardized NUHOMS system is 0.25 g peak horizontal ground acceleration and 0.17 g peak vertical ground acceleration. The SAR evaluates the effects of a seismic event on a loaded DSC inside the TC for two conditions. The first case postulated was for the TC in a vertical orientation in the decontamination area during closure of the DSC. For this case, the SAR shows that the loaded TC would not overturn during an earthquake, provided the loaded TC weighs 453.5 kg (190 kips) and experiences a horiwntal acceleration not greater than 0.40 g. N.B. This 0.4 g horizontal acceleration is not ground acceleration, which is limited to 0.25 g; rather, it is the peak acceleration at some elevation above ground level, and could result from a ground acceleration of 0.25 g multiplied by an amplification factor. The second case postulated in the SAR is for a seismic event occurring during the normal transport of the TC loaded on the trailer. The SAR stated that this case is enveloped by the handling case of +/-0.5 g acting in the vertical, axial, and transverse directions simultaneously. In Section 8.2.3 of the SAR the statement is made that the seismic stress inte11S1ties are to be taken as the normal transport stress intensities, because the accelerations for seismic are the same as those assumed for transport. NRC staff has included these stresses in Table 3.3.2-3. The individual stress intensities as well as the three load combination stress intensities are below ASME B&PV Code allowables. B. Design Basis Tornado Wind Loads Acting on TC The SAR shows that if the height to the top of the cask is 371 cm (146 inches), and the half wheel base of the transport vehicle is 168 cm (66 inches), there is safety factor of 1.5 against overturning when the TC is subjected to Design Basis Tornado (DBT) winds. Shell stresses were also evaluated and found to be 26,201 kPa (3.8 ksi), well below the 232,362 kPa (33.7 ksi) allowable for Service Level C. NRC staff concurs with the results for the DBT winds, provided the site-specific equipment, i.e., the trailer and the skid, correspond dimensionally with the example in the SAR. 3-28
C. Cask Drop Accident This SER presents a detailed discussion of the cask drop accidents postulated by the SAR. This includes the basis for the selection of the parameters and the assumptions used for the ANSYS finite element models. All drop scenarios assume that the DSC is inside the TC. Thus all previous discussions about the drop apply equally to the DSC and the TC. The ANSYS models predict that the stresses will exceed the yield stress for all major structural TC components except the top cover. The results in the columns entitled "NRC" in Table 3.3.2-4 are somewhat higher than those of the PNFS columns. This is due to the NRC staff conservatively selecting locations in the TC which may be more localized than locations selected by PNFS. However it is important to note that, in spite of this conservative process, the NRC staff results are still lower than allowable stress intensities. As discussed in the structural analysis of the DSC, any drop height higher than 38 cm (15 inches) shall require the retrieval and inspection of the DSC and its internals, in keeping with the guidelines of the ASME B&PV Code when using Service Level D allowables. Because the TC is also designed to ASME B&PV Code requirements, it will be necessary to inspect the TC as well, should it be subjected to a drop height higher than 38 cm (15 inches). Results are given in Table 3.3.2-4 of this SER. In all cases, the stresses are below the code allowables. D. Tornado Generated Missiles In docketed responses to NRC staff's questions for the NUHOMS-24P TR, PNFS presented results of an accident condition, namely design basis tornado (DBT) generated missiles. The two missiles considered are those suggested in NUREG-0800 (Reference 23), a 1,677 kg (3,697 pound) automobile, and a 125 kg (276 pound) 20 cm (eight-inch) diameter shell. TC stability, penetration resistance, and shell and end plate streises were calculated and shown to be below the allowable stresses for Service Level D stresses. Although the SAR for the standardized NUHOMS system did not include specific reference to these loads, the NRC staff has included them. The NRC staff believes that there is no need to recr ~culate stresses for this accident case because identical shell, bottom plate material and thickness were used, and the identical tornado missiles were postulated for both versions of the TC. The top plate material for the standardized NUHOMS system is a higher strength material as noted in Table 3.3.2-5. The results from the TR are shown for completeness in Table 3.3.2-5. E. TC Load Combination for Service Level D Accident Conditions Table B-2 of the design calculation NUH 004.0206 (Reference 43) summarizes the load combination for the three accident cases postulated in the SAR. Three drop cases were considered (1) vertical drop, (2) corner drop, and (3) horizontal drop. In each drop case the dead weight loads were combined with the drop loads. Table 3.3.2-6 of this SER shows the results and the material allowables at 204 °C (400°F) for the materials specified in the drawings. These allowables are somewhat lower than given in the SAR, but they represent 3-29
the values for the specified materials for 204°C (400°F). In all cases the actual stress intensities are lower than the allowables. Thus the TC meets the ASME B&PV Code for Service Level D conditions. 3.3.2.3 TC Fatigue Evaluation Section C.4.2 of the SAR for the standardized NUHOMS system presents an evaluation of the loading cycles of the TC to show that the six criteria associated with NC-3219.2 of the ASME B&PV Code are met. NRC staff evaluated Section C.4.2 and concurs that all six criteria are met. 3.3.2.4 TC Trunnion Loads and Stresses The relevant design criteria for lifting a "critical load," i.e., the spent fuel loaded in the DSC inside the TC while in the fuel building, are covered by ANSI N14.6, 1987 (Reference 8) and NUREG-0612 (Reference 14). Critical loads, defined by N14.6, are loads "whose uncontrolled movement or release could adversely affect any safety-related system or could result in potential off-site exposures comparable to the guideline exposures outlined in 10 CPR Part 100." In the case of the transfer cask, the cask lifting trunnions shall be considered as special lifting devices for the DSC. Because its design does not provide a dual-load path, the design criteria require that load bearing members shall be designed with a safety factor of two times the normal stress design factor for handling the critical load. Thus, the load bearing members must be sized so that yield stresses are no more than one-sixth minimum tensile yield strength of the material or no more than one-tenth the minimum ultimate tensile strength of the material. An additional allowance for crane hoist motion loads is recommended by NUREG-0612. Although Reference 14 does not quantify the magnitude of this dynamic load, ANSI NOG-1-1983 (Reference 44) does specify 15%, which was used in the SAR. Therefore the allowance for dynamic loads is appropriate. Because the titling trunnions are used in tilting and horiwntal transfer and transfer modes instead of lifting, the lower tilting trunnions are designed to ASME ill Class 2 criteria. Table 3.3.2-7 summarizes the results for the lifting trunnion assemblies, weld regions, and cask shell. This table presents summary results for the lifting and supporting trunnions that are designed in accordance with: (1) ANSI N14.6 for critical lift loads, and (2) ASME for horiwntal support loads. The local stresses in the TC at the intersection of the trunnion sleeve and the shell stiffener insert are calculated by using finite element analyses which appear in reference 18. The summary Table 3.3.2-7 shows that all stresses are less than the allowable stresses for both the ANSI N14.6 critical lift load conditions and the ASME B&PV Code for on-site transfer. Table 3.3.2-7 also shows the results for the lower tilting trunnion assembly. Comparisons between the PNFS values and NRC staff values are given. All stresses are below the ASME ill allowable stress intensities for Class 2 components. This table has included the consequences of using various material options which are noted in the drawings for the 3-30
standardized TC. For instance, the shell and trunnion sleeve materials have two options from which the fabricator may choose. 3-31
Table 3.1.2-1 HSM WAD COMBINATION RESULTS Section: Floor Slab Side Wall Front Wall Rear Wall Force: Shoar Trana. Loog. Shear Tram. Lona. Shear Trana. Long. Shear Trana. Loog. Moment Moment Moment Moment Momcm Moment Moment Moment Load Comb. 1 0.2 2.9 2.3 1.6 11.4 8.5 1.1 10.7 13.4 1.9 6.4 5.9 Load Comb. 3 7.5 154.0 117.0 20.7 231.0 185.0 6.4 494.0 293.0 12.3 118.0 127.0 Load Comb. 5 5.3 111.0 73.9 11.2 167.0 140.0 6.8 435.0 283.0 3.1 102.0 86.7 Load Comb. 6 5.1 110.0 67.0 13.7 152.0 127.0 3.0 308.0 193.0 4.4 105.0 78.7 Allowabl,c 14.6 217.0 229.0 23.9 724.0 762.0 40.4 910.0 910.0 15.3 477.0 313.0 MOS 0.9 0.4 1.0 0.2 2.1 3.1 4.9 0.8 2.1 0.2 3.0 1.5 Load Comb. 7 3.8 146.0 155.0 8.4 340.0 340.0 5.3 691.0 574.0 1.5 366.0 142.0 Allowabl,c 13.5 185.0 195.0 22.7 650.0 618.0 38.3 774.0 774.0 14.5 470.0 267.0 MOS 2.6 0.3 0.3 1.7 0.9 0.8 6.2 0.1 0.3 8.7 0.3 0.9 w Section: Roof Slab End Shield Wall Roar Shield Wall I w Force: Shear Tram. Long. Trana. Long. Tram. Long. N Moment Moment Moment Moment Moment Moment Load Comb. 1 12.8 392.0 399.0 0 0 0 0 Load Comb. 3 27.1 755.0 928.0 114 60.4 172 172 Load Comb. 5 13.0 390.0 533.0 23 12.2 32 32 Load Comb. 6 15.3 445.0 596.0 92.5 49 167 167 Allowable 49.8 2,151.0 2,087.0 516 1,593 423 1,178 MOS 0.8 1.8 1.2 3.5 25.4 1.5 5.8 Load Comb. 7 20.0 618.0 1,035.0 0 0 0 0 Allowable 47.2 1,780.0 1,830.0 MOS 1.4 1.9 0.8 n.a. n.a. n.a. n.a. ahear in unit5 of kipa/ft.; tramvcnc and longitudinal momcntl in units of in.-kipa/ft. MOS .. Margin of Safety ., (allow/calc)-1 n.a. not applicable Load Comb 1 .., 1.4 DL + 1.7 lL Load Comb 3 = 0.75 (1.4 DL + 1.7 lL + 1.7 T + 1.7 W) Load Comb 5 = DL + lL + T + E Load Comb 6
- DL + lL + T + F Load Comb 7 "" DL + lL + Ta
Table 3.1.2-2 DSC Support Assembly Load Combination Results Actual and Allowable S ~ Values COMPONENT AISC Load Combinationw fa Fa fbx Fbx fby Fby Sect 1.6(1,) fv fv/Fv Co~ COLUMN Equation 1 2.43 16.63 1.65 20.99 0.38 20.99 .24 .10 .01 A Equation 2 3.21 16.63 3.51 20.99 2.43 20.99 .33 .29 .02 A Equation 3 5.88 16.63 6.04 20.99 3.75 20.99 .54 .38, .02 A Equation 4 4.62 14.07 9.94 15.96 3.22 15.96 .72 .52 .03 A Equation 5 4.43 18.66 1.13 23.76 3.68 23.76 .27 .19 .01 A caoss BEAM w Equation 1 0.88 18.36 2.77 19.08 0.08 19.08 .20 1.95 .15 A I w w Equation 2 1.19 18.36 1.35 19.08 1.2 19.08 .13 5.33 .30 A Equation 3 2.02 18.36 6.63 19.08 1.55 19.08 .34 4.28 .24 A Equation 4 1.69 15.4 6.08 15.96 3.09 15.96 .40 3.36 .23 A Equation 5 0.47 20.74 4.31 21.6 11.47 21.6 .44 3.89 .19 A RAIL Equation 1 0.57 15.98 4.63 20.99 1.04 23.85 .30 1.07 .08 A Equation 2 0.81 15.98 3.15 20.99 1.87 23.85 .19 3.25 18 A Equation 3 Equation 4 1.48 0.51 15.98 13.57
-- 10.98 9.09 20.99 17.56 2.94 3.38 23.85 19.95
.46
.43 2.15 2.17
.12
.15 A
A Equation 5 3.93 17.89 12.2 23.76 8.38 27 .61 2.1 .10 A
Table 3.1.2-2 DSC Support Assembly Load Combination Results (Continued) Actual and Allowable Stress Values COMPONENT AISC ' Load Combination(a) fa Fa fbx Fbx fby Fby Sect 1.6~ fv fv/Fv Comment#l 11BBEAM Equation 1 0.07 18.15 2.53 20.99 0.62 20.99 .15 .25 I
.02 A Equation 2 1.27 18.15 4.47 20.99 5.32 20.99 .36 1.08 .06 A Equation 3 4.92 18.15 13.63 20.99 11.11 20.99 .91 6.09 .34 A Equation 4 3.52 15.24 12.58 27.56 8.04 17.56 .83 1.22 .08 A Equation 5 1.23 20.S 0.82 23.76 3.47 23.76 .14 .09 .01 A (a) Refer to Table 2.6 for definition of Equations 1 through 5
~ (b) Combined axial and bending stresses as a fraction of the allowable stress per AISC 8th ed., Section 1.6.1 ~ (c) A - Allowable, U., Unallowable
Table 3.2.2-1 DSC S ~ Analysis Results For Nonnal Loam Service Level A Stress (ksi) DSC Strcal 10.0pmg lOO"F Nomw AJ.toonblo*
~ Typo Dead Weight Int. ~ Thermal Handlini LevclA&B ftil:§1 ~ ~ l:il?& ~ ~ ~l:mQ DSC Shell PrlMemb 1.1 0.6P 0.5 0.5 N/A N/A 3.3 3.3 17.5 Mcmb + Bend 4.7 6.2 4.5 5.4 NIA N/A 9.6 9.6 26.3 Prl + Sccond 11.0 18.4 7.7 7.8 12.6 17.3 15.0 lS.0 52.S Outcr PrlMcmb 0.1 0.1 0.4 0.9 NIA NIA 0.3 0.3 17.S Top Memb + Bend 0.1 0.1 4.2 5.6 N/A NIA 0.3 0.3 26.3 c - Plato Prl + Second 0.1 0.1 13.8 14.0 0.2 0.2 0.3 0.3 52.5 Imier PrlMemb 0.1 0.1 0.4 0.7 NIA N/A 0.3 0.3 17.S Top Memb + Bend 0.1 0.1 2.5 2.9 NIA NIA 0.3 0.3 26.3 Plato Prl + Second 0.1 0.1 13.1 13.2 0.2 3.2 0.3 0.3 52.S Outer PrlMemb 0.1 0.1 0.2 1.2 NIA N/A 2.0 2.3 17.5 Bottom Memb + Bend 0.1 0.1 3.3 4.2 N/A NIA 10.3 10.3 26.3 Covet Plato Prl + Seoood 0.1 0.1 4.4 4.4 0.3 0.6 13.4 13.4 52.S w 24-P PrlMcmb 0.9 0.1 NIA NIA NIA N/A 2.7 2.7 20.S I
w Spac<< Memb + Bend 1.S 1.6 NIA NIA 4..S 4.S 30.8 <.n Dile Prl + Secood 2.2 2.2 26.5 32.7 6.6 6.6 61.5 S2-B PrlMemb 1.0 1.0 N/A NIA NIA NIA l.S 3.0 20.5 Spac<< Momb + Bend 1.7 L8 NIA NIA 3.1 5.1 30.8 Due Prl + Second 3.1 3.1 38.5 42.9 S.6 9.3 61.S InDiill' PrlMcmb 0.1 0.1 0.2 0.3 NIA NIA 0.3 0.3 17.5 Bottom Memb + Bend 0.1 0.1 0.5 0.5 NIA NIA 0.3 0.3 26.3 Plato Prl + Second 0.2 0.2 o.s 0.5 2.3 1.6 0.7 0.6 52.S Support PrlMemb 0.4 0.4 NIA NIA NIA NIA 0.4 0.4 19.3 R.odt Memb + Bend 0.8 0.76 NIA NIA 0.9 0.9 29.0 Prl + Secocd 0.8 0,76 -0 -o 0.9 0.9 S7.9
- A.llowab~ 111.U1 for Service Levels A and B Prlmm:y Membrane Sm* 17.S bi Sm= 20.5 bi Sm c: 19.3 bi Prlmm:y Mcmb + Bend 1.5 X Sm* 26.3 1.5 X Sm
- 30.8 LS xSm"' 29.0 Prlmary+Seoooda.ry 3 x Sm"" 52.S 3 x Sm 61.5 3 X Sm* 51.9 Shell, Diec and end SA 240 Type 304 SA516 SA36 platet for SOO°F
** Allowablo 1tre11 for 300°F for 1p11Cer ditc 3.0Sm = 60.0
Table 3.2.2-2 DSC Stress Analysis Results For Off-Normal Loads Service Level B Stress (ksi) Imetna1 DSC Stnm ~ 1bemial otr-Nonnal : CompoDent Typo 10.0 ptlg 100°F Hmdlina Allowable*
~ ~ ~ tIBQ mm! rm&
DSC Sholl PriMcmb 0.5 0.5 NIA NIA 0.6 3.2 17.5 Memb + Bond 4.5 5.4 N/A NIA 9.6 13.2 26.3 Pri + Second 7.7 7.8 12.6 17.3 30. 16.3 52.5 Outer PriMomb 0.4 0.9 NIA NIA 0.1 0.1 17.5 Top Mcmb + Bond 4.2 5.6 NIA ' N/A 0.1 0.1 26.3 Cover Pd+ Socood 13.8 14.0 0.2 0.2 0.6 0.6 52.5 Plato Inner PriMemb 0.4 0.7 NIA NIA 0. 0. 17.5 Top Momb + Bond Z.5 Z.9 NIA NIA 0. o. 26.3 w Plato Pri + Second 13.1 13.2 0.2 3.2 0.8 0.8 52.S I w Oater PriMomb 0.2 1.2 NIA NIA 4.0
- 4.7 17.5 O"\
Bottom Momb + Bond 3.3 4.2 NIA N/A 20.6 20.6 I 26.3 Covet Pri +Sccood 4.4 4.4 0.3 0.6 26.8 26.8 52.5 Plato l+P PriMcmb N/A N/A N/A NIA 0 - 20.5 Spacer Momb + Bond NIA NIA 0 - 30.8 Dile Pri + Socood 26.5 32.7 0 - ' 61.S 52-B Primary Mcmb N/A N/A NIA NIA 0 - 20.5 Spacer Momb + Bond N/A N/A 0 - 30.8 Dile Pri + Socond 38.5 42.9 0 - 61.5 Inner PriMomb 0.2 0.3 NIA N/A 0. 0. I 17..S Bottom Momb + Bend 0.5 0.5 NIA NIA 0.3 0.3 26.3 Plato Pri + Second 0.5 0.5 2.3 7.6 1.3 1.3 52.S Support PriMomb NIA NIA NIA NIA 0 - 19.3 Rod, Momb + Bond NIA NIA 0 - 29.0 Pri + Secoo.d -0 -0 0 - 57.9
- Allowable 11rcq ii taken for Service Level B for SA 240 Typo 304, SA 516 and SA 36, material at 500°F.
Table 3.2.2-3 DSC Load Combinations For Nonna! and OffwNormal Operating Conditions Service Levels A and B Stress (ks!) DSC Strc8I Cuc Cuc1 Caw Allowable* Component Typo A2 A3/A4 B2 I..eve.lA&B
~ ;tIB& ~ HE& ~ ~
DSC Shell PriMcmb 1.0 1.1 4.3 4.4 1.6 4.4 17.5 Momb + Bend 9.2 11.6 18.8 21.2 18.8 24.8 26.3 Prl + Second 31.3 43.5 46.3 51.1' S0.5 52.4' 52.S Outer PrlManb o.s LO 0.8 1.3 0.6 1.1 ! I 17.5 Top Mcmb + Bend 4.3 S.7 4.6 6.0 *U 5.8 I' 26.3 Covcc Plate Prl + Second 14.4 14.3 14.4 14.6 14.7 14.9 52.S lnnerTop PrlMcmb 0.5 0.8 0.8 1.1 0.5 0.8 17.S Covcc Plate Memh + Bend 2.6 3.0 2.9 3.3 2.6 3.0 ' 26.3 Prl + Second 13.4 16.S 13.8 16.8 14.2 17.3 52.5 Out<< PrlMcmb 0.3 1.3 2.3 3.6 4.3 6.0 17.S Bottom Memb + Bend 3.4 4.3 13.7 14.6 24.0 24.9 26.3 CovccPlate Prl + Second 4.8 S.l 18.2 18.S 31.6 31.9 52.S w Inna: PriMcmb 0.3 0.4 0.6 0.7 0.3 0.4 17.5 I w Bottom Mani,+ Bend 0.6 0.6 0.9 0.9 0.9 0.9 26.3 -.....i Plate Prl + Second 2.9 8.3 3.6 8.9 4.2 9.6 52.5 24-P PriMcmb 0.9 0.9 3.6 3.6 0.9 0.9 20.S Spac<< Mcmb + Bend 1.5 1.6 6.0 6.1 1.5 1.6 30.8 Di.le Prl + Second 28.7 34.9 35.3 41.5 28.7 34.9 61.5 Support PriMcmb 0.4 0.4 0.8 0.8 0.4 0.4 : 19.3 Rods Manb + Bend 0.8 0.8 1.7 1.7 0.8 0.8 29.0 Prl + Second 0.8 0.8 1.7 1.7 0.8 0.8 57.9 52-B PrlManb 1.0 1.0 2.5 4.0 1.0 1.0 20.S Spacer Manb + Bend 1.7 1.8 4.8 6.9 1.7 1.8 'I 30.8 Disc Pri + Second 41.6 46.0 47.2 SS.3 41.6 46.0 61.5 Strca intcmitie& coll.SC£Vltivdy combined irrclpective of location unlcu otherwise noted.
*Allowable 11trc:111 ia takm for Service Lovcls A and B for SA 240 Typo 304 Material at SOO'F.
- 1. Load cues A3 and A4 arc combined Into one caso bc:causc the lltrCSIICI for the normal and off-normal prcuurc cum a.re the aunc.
- 2. Load cases Bl, B2, B3, and B4 arc combined into one cue becawic the preaurc for normal and off-normal conditions arc the 111UDC and all theqnnl loads a.re bounded by the lOO"F case.
- 3. The lll4Ximum mess intcnaity In the DSC Bhdl for thia load combination wu found to be near the bottom of the ahell, howcvet tho max S.I. of 18.4 Jcai for the deed weight case is in the shell near a spacer di.le. Consequently a lower DW l!trcn value near the bottom of the shell WllB U8Cd for the combined lltfflll. I
Table 3.2.2-4 DSC S ~ Analysis Results For Accident Conditions Service Level C 1 Stress (k:si) Accideat.2 DSC Stress Pre&sure Flood Accident Compow::nt Type Seismic (S0.3 pslg) (21.7 psi) Handling Allowable*
~ HR£ ~ ~ ~ ~ ~ lIB& ~ ~
DSC Sbc:U Pri Mcmb 1.8 1.7 2.7 2.7 1.2 1.1 0.6 3.2 21. 19.9 Memb + Bend 18.2 14.3 22.6 16.8 1.2 1.1 9.6 18.6 29.1 27.6 Outer PrlMemb 0.4 0.5 1.9 4.4 0.1 18.0 0.1 0.1 21. 19.9 Top Mcmb + Bend 0.4 0.5 20.9 26.9 0.2 18.7 0.1 0.1 29.1 27.6 Cover Plate Weld - - - - - - - 0.5 - - lnlll:c PriMemb 0.4 0.5 2.1 3.5 0.2 0 0. - 21. 19.9 Top Memb + Bend 0.4 0.5 12.5 14.2 0.2 0 0. - 29.l 27.6 Plate w Outer PriMcmb 0.4 o.s 0.9 6.2 0.1 9.6 4.0 4.7 21. 19.9 w I Bottom Mcmb + Bend 0.4 o.s 16.5 20.8 0.4 9.6 20.6 23.6 29.1 27.6 co Cover Plate Weld - - - - - - - 12.4 - - 24-P PriMcmb 3.2 2.9 N/A NIA 0 0 0 - 30.7 28.6 SpllCCI" Mcmb + Bend S.2 5.3 N/A NIA 0 0 0 - 36.9 34.3 Disc 52B PriMcmb 4.0 3.8 N/A NIA 0 0 0 - 30.7 28.6 Spacer Memb + Bend 5.6 6.7 NIA NIA 0 0 0 - 36.9 34.3 Disc Inner Pri Mcmb 0.4 0.5 1.0 1.0 0.1 0 0 - 21. 19.9 Bottom Memb + Bend 0.4 0.5 2.5 2.5 0.4 0 0.3 0.3 29.1 27.6 Plato Support Pri Mcmb 0 0 NIA N/A 0 0 0 - 23.2 21.6 Rods Memb + Bend 0.2 0.2 N/A N/A 0 0 0 - 34.7 32.4
- Allowable stress for Service Level C @ 500°F for all ..:ases except accident pressure.
P. larger of 1.2 Sm or Sy
- 21.0 for SA 240 Type 304 PL+ P11
- smaller of 1.8 Sm or 1.5 Sy = 29.1 for SA 240 Type 304
- 1. No secondary streu needs to be evaluated according to ASME Code for Scrvlcc Level C. This includes thcrm8l u well as secondary bending streaea for p1'C881.11'C CMC8.
- 2. Accident pressure applied to inner cover plates, alllO allowable stressc8 for this condition should be based on 580"F.
Table 3.2.2-5 DSC Load Combinations for Accident Service Level C Cases1 Stress (ksi) DSC Streu Cuc2 CastJI Cuc" Allowable' Componeot Type Cl C2 C3/C4/C5/C6/C:,
~ Im& ~ ID?& ~ ?m&. ~ ~
DSC Shell PriMemb 4.4 s.o 4.4 2.2 3.8 6.5 21. 19.9 Memb + Bcnd 24.9 23.2' Zl.1 12.7 22.0 25.87 29.1 27.6 Outer Pr! Memb 2.4 5.0 2.1 19.0 2.1 4.6 21. 19.9 Top Memb + Bend 21.4 27.5 21.2 24.4 21.l 27.1 29.1 27.6 Cove.- Plate I Inner PriMemb 2.6 4.1 2.4 0.8 2.2 3.6 21. i 19.9 Top Mcmb + Bend 13.0 14.8 12.8 3.0 12.6 14.3 29.1 27.6 P1ato w w I Outer Pri Memb o.s 6.8 0.2 10.9 4.1 11.0 21. 19.9 I.O Bottom Mcmb + Bcod 0.9 21.4 0.9 13.9 21.l 24.31 29.1 27.6 Cover Plate I 24-P PriMcmb 4.1 3.8 0.9 0.9 0.9 0.9 30.7 28.6 Spacer Memb + Bend 6.7 6.9 l.S 1.6 I.S 1.6 36.9 34.3 Dile 52-B Pri Mcmb 5.0 4.8 1.0 1.0 1.0 1.0 30.7 28.6
~ Mcmb + Bend 7.3 8.5 1.7 1.8 1.7 1.8 36.9 34.3 Di.le Inner PriMemb 1.5 1.6 1.5 0.4 1.1 1.1 21. 19.9 Bottom Mcmb + Bend 3.0 6.1 3.0 o.s 2.9 2.9 29.1 27.6 P1ato I
Support PriMcmb 0.4 0.4 0.4 0.4 0.4 0.4 23.2 21.6 Rods Memb + Bend 1.0 1.0 0.8 0.8 0.8 0.8 34.7 : 32.4 Strea Intcnsltiell coDJ1CrVativdy combined m:cspcctivo of location unlcas otbc:rwisc noted.
- 1. Secondary ltrcaes are not required for Scrvico Lovcl C.
- 2. Scimtlc lltreNOI are considered "mccha.nical low* and mtllt be combined with dead weight and accident prcaurc for Cl.
Table 3.2.2-5 DSC Load Combinations for Accident Service Level C Cases (Con~ued)
- 3. Cue C2 is dead weight, normal prcaurc and flooding.
-4. Bccau1e thcnnal ltrcuel need not bo evaluated for Scrvicc Level C, cuca C3 through C6 arc bounded by Cl and consilt of deed wdght, accident pressure, and accldcnt handling. :
S. Allowablca arc hued on a maximum DSC temperature of S00°F, cxcc:pt for the accidait prcaure condition for C3-C7 which bu a maximum DSC tcmpenturc of S79°F.
- 6. The maximum ltrea intem1ty in the DSC lhell for tJ\il load combination wu found to bo near a spacer dilC, where the prcllR.lrC ,streu ii only 2. 7 bi. Thus the 26.9 bi strea doc to prcaurc shown in Table 3.2.2-4 ii not mod bocaUlo it occun noar tho end of tho lhcll.
- 7. The maximum ltrCa inteody in tho DSC lhdl for this load condition ls at tho bottom end of tho lhol1 and ii duo primarily to tho accldcnt loading cue. Tho dead weight 1trc11 "" 6.2 kl.I, and tho accidont prcslW'O Btrc81 - 1.0 at node S1S.
- 8. Tho maximum ltrca intcmity in tho outer bottom plato for thit load condition is noar tho grapple connocdon and ii duo primarily to tho accident hendling cue.
Tho accident proauro at thia location is 0.6 bi. w I +:>, C>
Table 3.2.2-6 DSC Drop and Internal Pressure Accident Loads Service Level D Stress (ksi) DSC StrCII Accident' Component Type Vcrtic:al (75 g) Horirontal2 (75g) Prcuurc (50.3 psig) Allowablc1
~ rm& ~ NE& ~ liB& csoo*r ~
DSC Shell Pri Mcmb 10.2 12.7 26.1 32.9 2.7 2.7 42.0 ! 39.8 Mcmb + Bend 27.1 27.1 40.4 S0.7 22.6 26.9 63.0 59.8 Inner Pri Mcmb 6.2 10.2 10.3 10.3 2.1 3.5 42.0 39.8 Top Mcmb + Bend 10.2 10.2 10.3 11.3 12.5 14.2 63.0 59.8 Plate Outer Pri Mcmb 1.7 3.2 10.3 10.3 1.9, 4.4 42.0 39.8 Top Memb + Bend 3.2 3.8 10.3 11.3 20.9 26.9 63.0 59.8 Cover Plate w Outer PrlMemb 3.9 3.9 10.3 10.3 0,9 6.2 42.0 39.8 I = Bottom Cover Mcmb + Bend 4.5 4.6 10.3 11.3 16.5 20.8 63.0 59.8 Plate ' Inner PrlMemb 3.3 3.6 10.3 10.3 1.0 1.0 42.0 ' 39.8 Bottom Memb + Bend 13.5 17.9 10.3 11.3 2.5 2.5 63.0 ; 59.8 Plate I 52-B (bound) PrlMemb 32.5 32.5 48. 48. NIA NIA 49.0 49.0 Spacer Memb + Bead 60.5 60.S 65. 67.5 NIA NIA 70.0 I 70.0 Dile Support Primary 32.4 33.2 0.3 3.0 NIA NIA 40.6 40.6 Rods (24-P Memb + Bend 57.3 57.27 7.2 7.0 58.0 58.0 bound) Top End Primary - . - - - - - 25.2 I I 23.9 Struct. Weld* Pri + Bend 4.7 6.6 - 11.3 19.7 25.5 37.8 35.9 Bottom End Primary - - - - - - 25.2 23.9 Struct. Wdd* Pri + Bend 6.3 7.5 - 11.3 10.0 12.6 37.8 35.9
- 1. Allowablea ta1ccn at wont cut temperature, i.e., for Case Dl, T-S00°F llhell temperature, CJtCCPl accident pressure.
- 2. These columns for ltrcl&cB for shell, top coven and bottom cover arc talce.n from NUH004.0202 and the SAR for the standardiz.cd NUHOMS.
- Efficiency f&ctor for Clau C, Type 3 non-volumetric irupectcd ,;vdds = 0.6.
- 3. Accident prcsaurc load applied to outer cover platcl!, also allowable lltrcH for tbia condition should be baaed on 580°F.
Table 3.2.2-7 DSC Enveloping Load Combination Results for Accident Loads Service Level D Stress (ksi) DSC Stmll CucD2 CascD4 AllowablCS Component Type DW + To + Pa + FD DW + To + Pa + Lo
~ !:IB& ~ ~ ~
DSC Shell Pri Mcmb 29.3 36.2 3.8 6.5 -42.0 Mcmb + Bend 47.8 59.e 36.9 -41.6 63.0 Outer PriMcmb 12.3 14.8 2.1 4.6 42.0 Top Mcmb + Bend 31.3 38.3 21.1 27.1 63.0 Cover Plate I Inner PriMcmb 12.5 13.9 2.2 3.6 -42.0 Top Mcmb + Bend 22.9 25.6 12.6 14.3 63.0 Flaw Outer PrlMcmb 11.3 16.6 4.1 10.9 42.0 w Bottom Mcmb + Bend 19.5 32.2 10.1 4-4.5 63.0 I .p. Cover I N Plate inner Primary 11.4 11.4 1.1 1.1 42.0 Bottom Mcmb + Bcnd 12.9 13.9 2.9 2.9 63.0 Plate 52-B (bound) PrlMcmb 49.0 49.0 1.0 1.0 49.0 Spacer Mcmb + Bend 6£,.7 69.3 1.7 1.8 70.0 Disc Support Primary 32.8 33.6 0.4 0.4 -40.6 Rods Mcmb + Bend 58.0 58.0 0.8 0.8 58.0 ' I Top End** Primary - - - - 25.2 I Struct. Wcld Pri + Bend 30.4 36.9 24.9 26.1 37.8 Bottom End** Primary - - - - 25.2 Struct. Wcld Pri + Bend 19.7 2-4.0 23.8 25.1 37.8 Streu intcnaiticl colllCl'V8tivcly combined irrclpcctive of location unleas othcrwiBo noted.
- Allowablcs arc based on a maximum DSC temperature of 500°F.
** Efficiency factor for C1au C, Typo 3 non-volumcaic lnspcctcd wolds = 0.6.
- 1. Tho maxim.um. ltrC88 intensity in tho DSC llholl for thi1 load combination was found in the lhdl near tho IUppOrt rail for the 18.5* drop cue. The accident prosmrc l5trcu at this location is 2. 7 ks!.
Table 3.3.2-1 Transfer Cask Stress Analysis Results for Normal Loads Service Levels A and B Allowables Stress (ksi) I Cuk Streu Dead Thermal** Norm.all Componcut Type Weight HAN"lliog Allowable*
~ ~ ~ ~ ~ !IB&
Cuk.Sbcll PriMcmb 0.7 0.7 NIA NIA o.s 0.5 18.7 Memb + Bend 0.9 0.8 NIA NIA 4.1 4.1 28.1 Prl + Second 0.9 0.8 20.3 46.2 (top) 42.6 4.1 S6.1 19.9 (bottom) Top PriMcmb 0.2 0.9 NIA NIA - - 21.7 Cover Memb + Bend 0.6 1.0 NIA NIA 6.3 6.3 32.6 Plate Prl + Second 0.6 1.0 11.7 11.7 6.3 6.3 65.1 Bottom PrlMemb 0.2 0.8 NIA NIA - - 18.7 Cover Mcmb + Bend 1.3 0.8 NIA NIA 14.2 14.2 28.1 w I Plate Pri + Second 1.4 1.4 5.3 19.4 14.2 14.2 56.1 ~ w Top PrlMemb 0.2 0.8 NIA NIA - 1.5 20.3 Ring Mcmb + Bend 0.2 0.8 NIA NIA - 1.5 30.5 Prl + Second 0.5 0.8 8.4 23.6 - 6.4 61.0 Bottom PrlMcmb 0.4 0.4 NIA NIA - 4.6 20.3 Ring Mcmb + Bend 0.4 0.6 NIA NIA - 4.6 30.S Pri + Second 0.6 0.6 17.4 22.8 - 26.9 61.0 Transfer Calk Collar PriMcmb 0.2 0.8 NIA NIA 0 1.5 20.3 for BWRDSC Mcmb + Bend 0.2 0.8 NIA NIA 0 1.5 30.S Pri + Second o.s 0.8 11.8 23.6 0 6.4 61.0
- Allowable. taken 111 400°F
** Tbeanal lltrcucll arc comiden:d secondary streaea only
- 1. The PNFSI ihcll strca reported for handling is located at the trunnloo, whcrcu the NRC lltrca la located In the middle of the lheU where bending would be maximum.
Table 3.3.2-2 Transfer Cask Load Combinations for Normal Operating Conditions Service Levels A and B Stress (ksi) Calk Streu Load Comb 1 Load Comb 2 Component Type Al-A4, Bl Al-A3, A5, B2 Allowable*
~ ~ ~ ~
Ca.akShcll PriMcmb 1.2 0.7 0.6 1.2 18.7 Memb + Bend 1.4 0,8 4.2 4.9 28.1 Pri + Second 63.41 47.1 62.7 1 51.1 56.1 Top PriMcmb 0.2 0.9 0.2 0.9 21.7 Cov<< Meinb +Bend 0.6 0.9 6.9 7.3 32.6 Piatc Pri + Second 18.6 19.0 11.9 12.9 65.1 Bottom PriMemb 0.2 0.8 0.2 0.8 18.7 Cover Memb + Bend 9.9 1.3 8.8 lS.0 28.1 Plate Pri + Second 14.4 21.0 13.2 35.0 56.1 Top Pri Memb 0.2 0.2 0.2 2.3 20.3 Ring Memb + Bend 0.2 0.2 0.2 2.3 30.S Pri + Second 8.6 24.6 8.6 30.8 61.0 Bottom Pri Mcmb 0.4 0.4 0.4 5.0 20.3 Ring Mcmb + Bend 0.4 0.4 0.4 5.2 30.S Pri + Second 18.0 23.6 17.6 50.3 61.0 Transfer Cask: PriMemb - 2.3 - - 20.3 Collar for BWR. Memb + Bend - 2.3 - - 30.S DSC Pri + Second - 30.8 - - 61.0
- Allowablca taken at 400°F
- 1. This streiis ia reported by PNFSI in Table B-2 of NUH 004.0206. In that table an allowable st:rcu for primary plua sccondacy 8trCIII wu taken 111 70 bi. However, the NRC staff has taken lower allowable strenes hued on worat case me.tcriAJ.I which may be uacd according to drawing NUH-03-8001. The NRC staff has comequently summed streuea at a point, rath<< tban the more consocvatlve approach of fillmming ltrelsC8 rogardlca of location.
Table 3.3.2-3 Transfer Cask S ~ Analysis Results for Accident Loads Service Level C** Allowables Stress (ksi) Calk Strca DBT Load Component Type Handling Seiamlc Wmd Combination Al.lowablea"' C2***
~ l:m& ~ ~ fHE'§I ~
Cuk Shell PriManb o.s o.s o.s o.s - 1.7 1.7 22.4 Mcmb + Bend 4.1 4.1 4.1 4.1 3.8 S.4 9.0 33.7 Top PriMcmb - - - - - 0.2 - 26.0 Cover Pl.ate Mcmb + Bend 6.3 3.2 6.3 3.2 o.s 13.2 7.4 39.1 Bottom PriManb - - - - - 0.1 0.8 22.4 Cover Plate Mcmb + Bend 14.2 14.4 14.2 14.4 o.s 28.6 29.2 33.7 Top Ring PriMcmb - 1.S - 1.S - 0.1 3.8 24.4 Manb + Bend - 1.S - 1.5 - 0.1 3.8 36.5 Bottom Ring PriMcmb - 4.6 - 4.6 - 0.1 9.6 24.4 Mcmb + Bend - 4.6 - 4.6 - 0.1 9.6 36.5 Transfcc Cuk PriMemb - 1..S - 1.5 - - 3.8 24.4 Collar for Manb &Bend - 1.5 - 1.5 - - 3.8 36.S BWRDSC
- Allowablea tAk:ai at 400°F
** No 1CCOndary atrcaca need to bo C'Nl.!uated according to the ASME Code for Sa:vicc l..cvd C.
- The C2 loed combination includCII dcadwcight, seill:nic, and ha.ndling loads.
Table 3.3.2-4 Transfer Cask Drop Accident Loads Service Level D Allowables Stress (ksi) I Cuk. Strca Vertical Vcrtical Horizontal Comer Comet Componcm Type Top Drop Bottom Drop Drop with DW Top Bottom Allowablea*
~ 1'.:IB& PNFSI ~ ~ ~ ~ NF& ~ l:IB&
CukSbell PriMcmb 9.6 30.1 8.7 8.7 3.8 21.9 8.3 7.6 4.6 8.5 44.9 Memb + Bead 10.2 33.6 8.7 8.7 15.5 22.7 13.9 7.6 13.9 11.3 64.,4 Top PrlMcmb 25.2 24.2 - - 12.2 17.3 2.1 7.5 - - -48.7 Ring Memb + Bend 25.2 46.4 - - 12.2 22.6 2.9 12.6 - - 73.1 Top PriMemb 24.2 24.2 - - 5.8 7.7 2.7 11.7 - - 49.0 3* Cover Memb + Bead 2-4.2 24.2 3.7 3.7 5.8 8.0 14.1 1-4.1 - - 70.0 Bottom Pri Mcmb - - 22.9 22.9 5.8 6.4 - - - 33.1 44.9 w 2" Cover Mcmb + Bead 14.4 14.4 22.9 22.9 5.8 11.6 - - 33.1 33.l 6-4.-4 I .po O'I Bottom PrlMcmb - - 14.0 26.7 12.2 12.2 - - 9.7 I 10.7 48.7 Ring Mcmb + Bead - - 14.0 26.7 12.2 25.9 - - 9.7 33.9 73.1 TrannerCaik PriMcmb 13.0 13.0 - - 12.2 17.3 - - - - 48.7 Collar for BWR Mcmb + Bend 25.2 46.4 - - 12.2 22.6 - - - - 73.1 DSC Bolts for Ave. Ton1ion - - - - - - 27.1 'J!J. 7 - - 77.0 Top Cover ' Bollll for Ave. Tcnlion 0 0 0 0 - - - - 74.3 74.3 153.0 Collar Shear 0 0 0 0 56.9 56.9 - - 39.6 39.6 64.3
- Allowable1 talcen at 400°F
Table 3.3.2-S Transfer Cask S ~ Results for Tornado Driven Missile Impact s~ (kg) Stress Massive Pen. Resist Allowable* , Cask Type Missile Missile Component Cask Pri Memb 6.4 4.9 44.9 I Shell Pri + Bend 20.5 30.3 64.4 Top Pri Memb 0 0 49.0 l Cover Pri + Bend 19.7 13.2 70.0 ; Bottom Pri Memb 0 0 44.9 w Cover Pri + Bend 17.5 22.2 64.4 I .+:>,
- Allowable stresses based on Service Level D Allowables at 400°P
Table 3.3.2-6 Transfer Cask Load Combinations for Accident Conditions Service Level D Stress (ksi) Ca.de Streu Case Case C8.llo Component Type D1 (Vert) D2 (Comer) D3 (Hori.z) Allowabl~ rm! ~ ~ l:IB& ~ NB& Cade Shell Pri Memb 9.7 30.8 4.7 9.2 3.9 22.6 44.9 Mcmb + Bend 10.3 34.4 14.3 12.1 15.6 23.5 64.4 Top PriMcmb 25.4 25.0 2.2 8.3 12.3 18.l 48.7 Ring Mcmb + Bend 25.4 47.2 3.0 13.4 12.3 23.4 73.1 Top PriMcmb 24.4 25.1 2.9 12.6 5.9 8.6 <19.0 Cover Memb + Bend 24.4 25.2 14.7 15.1 6.0 9.0 70.0 w I .j::,, Bottom PriMemb 14.1 27.1 10.1 11.1 12.3 12.6 48.7 (X) Ring Memb + Bend 14.1 27.3 10.1 34.5 12.3 26.5 73.1 Bottom PriMcmb 23.1 23.7 0 33.9 5.9 7.2 44.9 Cover Mcmb + Bend 23.1 23.7 34.4 33.9 6.0 12.4 64.4
- Service Lcvd D Allowablea at 400°F
Table 3.3.2-7 Summary of Stress Analyses for Upper lifting Tnmnions and Lower Resting Trunnions, Weld Regions and Cask Shell Critical Handling Loads On-Site Transportation Loadl (per ANSI Nl4.6) (per ASME ill Cua 2) Component Location Strculntenlhy(lcm) Allowable (bl) Strea lntauity (bl) Allowable (Ian")*
~
Upper Trunnion (lift pin) A-A S.9 13.1 NIA NIA (IUppOrt pin) B-B 10.0 13.1 NIA NIA Upp<< Tnmnion C-C 6.3 9.0 S.9 45.0 Sloc,vc, (ferritic materiaJ) Sbdl at Sleeve s.o S.4 27.7 4S.O
~ P!!D2 Weld l 6.9 9.0 1 6.4 4S.O Sleeve/Trunnion 2 7.7 9.0 2 8.3 45.0 i
(Upper Trunnion) - w ~ f1a2 I ..i::,. Weld 1 s.o 9.0 1 6.0 '
' 45.0 I.O SICC"t"C/lnacrt 2 s.s 7.0 2 S.4 7.0 (Upp<< Tnmnion) 3 4.4 S.4 3 4.2 32.6 l..o'wc£ Trunnion NIA NIA 4.1 28.1 l..o'wc£ Trunnion (304 material) NIA NIA 5.6 I 28.1 Sloc,vc, Shdl at Sleeve 26.9 28.1 11!!!2 Weld NIA NIA 1 S.6 28.1 Sleeve/Trunnion 2 7.3 28.1 (Lowa:-Trunnion) 3 S.1 28.1
~
Wdd I.IA NIA Sleeve/Cask: (Lowcc Trunnion) 2 6.0 28.1
- Material allowable. arc taktin at 400°F lltreuel and materials have been comcrvativcl.y combined 10 that worst cue for material options ia recorded.
4.0 THERMAL EVALUATION Introduction This section evaluates the thermal hydraulic aspects of the designs for the HSM, DSC, and TC. The designs are evaluated against design criteria as presented in the SAR or otherwise determined to be acceptable. Below .is _a description .of the..thermal hydraulic review which .. was made followed by the actual evaluation. Description of Review Two similar standarized NUHOMS system designs have previously been reviewed, the 7-P design (Reference 45) and the 24-P design (Reference 40). Safety evaluation reports have been issued for both of these designs (Reference 46 and 39, respectively). The standardized NUHOMS design, which is the subject of this SER, differs from the 24-P thermal design in several respects. The standardized NUHOMS system design has increased maximum heat load capacity (1 kilowatt per PWR assembly) compared to the 24-P design (0.66 kilowatt per PWR assembly). This increased heat load capacity was achieved by redesign of the air flow passages (larger flow area, but less height difference from the bottom of the DSC to the air outlet), increased fuel temperature limits for normal operation, and more realistic thermal calculations. The design has also been extended to include the capability to store 52 BWR spent fuel assemblies, referred to as the 52-B design. In light of the two previous reviews of similar designs, this review focused on the differences from the previously approved designs. The review addressed the capability of the standardized NUHOMS system design to maintain fuel cladding temperatures and concrete temperatures within the acceptance criteria during normal, off-normal, and accident conditions, and also on the correctness of thermal gradients determined for use in the structural analysis. Limiting temperature gradients used for structural and confinement integrity evaluations have been reviewed and found to be determined in an acceptable manner. In view of the differences in thermal hydraulic characteristics between the standardized NUHOMS system design and the previously approved designs, the staff considered it prudent to require a thermal performance verification for the first standardized NUHOMS system to be used. The discussion for the thermal performance verification is included in Section 12.1. 7. 4-1
Awlicable Parts of 10 CFR Part 72 10 CFR 72.236(f) requires the cask design to have adequate heat removal capacity without active cooling systems. The staff considered Subpart F criteria as well. Section 72:_.122(h) provides that the fuel cladding should be protected against degradation and gross rupture.
. . Section 72.122(b) ~ta~ that structures,__syste,ms, and comP'.(ments imPQrtr!flt to safety_ ~hould .
be designed to accommodate the effects of, and be compatible with, site characteristics and environmental conditions associated with normal operation, maintenance, and testing; and to withstand postulated accidents. Section 72.126(a) provides that radioactive waste storage and handling systems should be designed with a heat-removal capability having testability and reliability consistent with their importance to safety. Also, Section 72.122(f) states that systems and components that are important to safety should be designed to permit inspection,
- maintenance, and testing .
4.1 Review Procedure The material reviewed consisted of the SAR, including several sets of responses to staff comments and concerns, submitted by the applicant A number of design calculations provided by the applicant were also utilized in the review. This material was evaluated to establish compliance with the applicable requirements of 10 CFR Part 72. The review addressed the adequacy of natural convection cooling to maintain fuel cladding and HSM concrete temperatures within acceptable limits during normal, off-normal, and accident
' conditions. Thermal hydraulic aspects of the transfer cask design were also evaluated.
4 .1.1 Design Description The standardized NUHOMS system provides for the horizontal storage of irradiated fuel in a dry shielded canister which is placed in a concrete horizontal storage module. Decay heat is removed from the fuel by conduction and radiation within the DSC and by convection and radiation from the surface of the DSC. Natural circulation flow of air through the HSM and conduction of heat through concrete provide the mechanisms of heat removal from the HSM. Spent fuel assemblies are loaded into the DSC while it is inside a transfer cask in the fuel pool at the reactor site. The transfer cask containing the loaded DSC is removed from the pool, dried, purged, backfilled with helium, and sealed. The DSC is then placed in a transfer cask and moved to the HSM. The DSC is pushed into the HSM by a horizontal hydraulic ram. The dry, shielded canister is constructed from stainless steel plate with an outside diameter of 107.8 cm (67.25 inches), a wall thickness of 1.6 cm (0.625 inches), and a length of 473 cm (186.25 inches). Within the DSC, there is a basket consisting of either 24 square cells in the PWR design or 52 cells for the BWR design. An intact spent fuel assembly is loaded into each cell yielding a capacity of either 24 PWR or 52 BWR assemblies per DSC. Spacer disks are used for structural support. The DSC has double seal welds at each end and rests on two steel rails when placed in the HSM. 4-2
The HSM is constructed from reinforced concrete, carbon steel, and stainless steel. Passageways for air flow through the HSM are designed to minimize the escape of radiation from the HSM but at the same time to permit adequate cooling air flow. Decay heat from the spent fuel assemblies within the canister is removed from the DSC by natural draft convection and radiation. Air enters along the bottom of each side of the HSM, flows around the canister, and exits through flow channels along the top sides of the module. Heat is also radiated from the DSC to J4e inner &¢~ o{ the HS_M walls wh~, again, _µatu;ral _ _ convection air flow removes the heat. Some heat is also removed by conduction through the concrete. The standardized NUHOMS system utilizes a transfer cask, transporter, skid, and horizontal hydraulic ram. The transporter, skid, and horizontal hydraulic ram are not affected by the thermal analysis. During transport and vacuum drying of the fuel in the DSC, heat is " removed by conduction through the transfer cask. 4 .1. 2 Acceptance Criteria Peak fuel cladding temperature for normal operation, calculated according to the methodology of PNL-6189, Reference 47, is the acceptance criterion relative to the fuel. The staff has reviewed this methodology and found it to be acceptable. Resulting peak fuel cladding initial storage temperature limits are 384 °c (724 °F) for PWR fuel and 421 °c (790°F) for BWR fuel based on the long term average ambient temperature not exceeding 21 °C (70°F). For accident events, the staff has accepted a peak fuel cladding temperature limit of 570°C (1058°F) based on Reference 48. Meeting these criteria for storage with an inert cover gas ensures that the criteria in of 10 CFR 72.122(h) are satisfied. In Table 3.2-1 of the SAR, the applicant cites ACI-349-85 and ACI-349R-85 as the applicable criteria for concrete design. These criteria are acceptable to the staff with the exception that calculated concrete temperatures for both normal operation and accident conditions could exceed those of the ACI criteria. Use of a concrete mix and aggregate specification for higher temperatures is therefore required in lieu of the ACI 349 criteria. (See the materials discussion in Section 3.0 of t.hb SER.) This review of the thermal analysis also addresses the correctness of the calculated maximum temperatures and of temperature gradients used for input to structural evaluations. The manner of calculating maximum temperatures and thermal gradients for the structural analyses has been found to be acceptable. 4.1.3 Review Method The thermal analysis was reviewed for completeness, applicability of the methods used, adequacy of the key assumptions, and correct application of the methods. Thermal analysis was performed primarily with the HEATING-6 (Reference 49) computer program. HEATING-6 is a part of the Oak Ridge National Laboratory SCALE package and is an 4-3
industry standard code for thermal analysis. Representative input and output was reviewed to establish that the code use was appropriate and that the results were reasonable. Independent calculations were performed to check other portions of the analysis which did not use the HEATING-6 code. This includes the natural convection cooling calculation which determines the magnitude of the air flow through the HSM. Since the heat flux through the DSC surface is significantly increased for the standardized NUHOMS system design compared to the previous 24-P design (Reference 40), the ability to remove heat by air cooling is particularly important. An independent determination of the form losses. and friction pressure drop, together with a balancing of the buoyancy and flow loss, confirmed the adequacy of t!te analysis. Review effort was directed toward establishing the validity of the analyses and their applicability to the design. The analyses were reviewed for completeness, validity of input, reasonableness of results, and applicability of results to support conclusions regarding the design. In general, independent analyses were not performed. However, in some cases energy balances and simplified calculations were performed as a check.
- 4.1.4 Key Design Information and Assumptions The key assumptions made in the thermal analysis are listed below.
- 1. The total heat generation rate for each fuel assembly is less than or equal to one kilowatt for each PWR assembly and equal to or less than 0.37 kilowatts for each BWR assembly. All heat is assumed to be generated in the fuel region.
- 2. Each dry storage canister contains 24 intact PWR assemblies or 52 intact BWR fuel assemblies.
- 3. A factor of 1.08 to account for uneven heat generation along the length of the fuel was assumed for thermal analysis inside of the DSC.
- 4. Design long term average ambient temperature of the external environment is taken as 21 °C (70°F) with normal solar heat load. Limiting normal conditions of -17.8°C (0°F) and 37.8°C (l00°F) ambient are also considered.
- 5. Off-normal temperatures of -40°C (-40°F) and 52°C (125°F) ambient temperature are considered. The 52°C (125°F) case assumed maximum solar heat load for the HSM, but use of solar shades and hence no solar heat load for the transfer cask is permissible above 37.8°C (100°F).
- 6. Accident condition is assumed to be total blockage of all inlets and outlets for five days with either -40°C (-40°F) or 52°C (l25°F) and maximum solar heat flux ambient conditions.
- 7. A helium atmosphere is assumed to be maintained within the DSC over the entire storage life of the standardized NUHOMS system.
4-4
t, 4.2 Horizontal Storage Module (HSM) 4.2.1 Design Evaluation The SAR was reviewed in conjunction with three calculation packages, References.SO, 51 and 52, and responses to several rounds of staff questions. -- 4-.2. L 1 Norma1 Operation A total of three cases were considered for normal operating conditions based on the temperature of the air at the inlet of the module. These are: (1) entering air at -17.8°C (0°F) representing "minimum normal conditions,* (2) entering air at 21 °C (70°F) representing "normal conditions,* and (3) entering air at 37.8°C (l00°F) representing "maximum normal conditions." The method of calculating concrete temperatures is acceptable. Concrete temperatures on the inside surface of the HSM reach 100°c (212°F) for the 24-P design, and 89.4°C (193°F) for the 52-B design, when the ambient temperature is 37.8°C (l00°F). As long as the air temperature at the outlet remains within 37.8°C (100°F) of the ambient (for a heat load of 24 Kw), and maximum long term ambient temperature limits are not exceeded, the conservative design calculations show that the HSM concrete temperature limits and fuel cladding temperature limits will not be exceeded. The applicant determined that the HSM wall temperature gradients for the 37.8°C (l00°F) ambient temperature case are bounding among the normal and off-normal cases. These thermal gradients are either best estimate or conservative and are suitable for use in the structural loading analysis. 4.2.1.2 Off-Normal Conditions The off normal conditions considered were an inlet temperature of -40°C (-40°F) representing extreme winter minimum and 52°C (125°F) representing extreme summer maximum. Solar heat flux of 1,397 kJ/hr-m2 (123 Btu/hr-ff) was included for the extreme summer case. The concrete temperature on the inside surface of the HSM reaches a maximum of 121 °C (250°F) for the 24-P design and 110°c (230°F) for the 52-B design for the extreme condition of 52°C (125°F) ambient temperature. 4.2.1. 3 Accident Conditions The total blockage of all air inlets and exits was analyzed as the accident case (Reference 50). Adiabatic heatup of the various components was assumed, with the HSM providing the slowest heatup rate. Adiabatic heating starting at the 52°C (125°F) inlet temperature condition is the limiting case for maximum concrete and fuel cladding temperatures. A heatup period of five days was assumed. The resulting concrete temperatures 249°C (480°F) on the floor and 231 °c (448°F) on the roof are significantly above the acceptance criteria of l 77°C (350°F) for accident conditions. 4-5
4.2.2 Discussion and Conclusions Since concrete temperatures may exceed 93.3°C (200°F) during limiting normal conditions, a concrete mix and aggregate specification must be included for the elevated temperatures expected. These are considered acceptable as an alternative to satisfaction of the testing requirements of ACI 349, Section A.4.3. Satisfaction of the limiting condition for operation of a 37.~°C (10Q 0 F) maximum aj.r tempe~re rise on exit from the HSM gives a reasonable_ degree of assurance that adequate cooling is achieved. Based on the plot of HSM inside roof temperature response shown on page 25 of Reference 50, the HSM concrete temperature may exceed l77°C (350°F) sometime after 40 hours of flow blockage. Concrete temperature over 177°C (350°F) in accidents (without the presence of water or steam) is not acceptable due to reduction in strength and durability. In view of the facts that a PNFS proposed 4-day inspection frequency of the air inlets and outlets result in: (1) exceeding the ACI 349 and NRC staff recommendations for maximum concrete temperature limits, and (2) approaching the fuel clad temperature limits recommended by PNL-6189 (Reference 47), the NRC staff requires a daily inspection for the air inlets and outlets. The applicant used thermal load inputs for the HSM stress analysis from the 37.8°C (100°F) ambient temperature case. Thermal gradients are best estimate or conservative and are suitable for use in the structural loading analysis. See Table 4.2 for a summary of some component temperatures as a function of ambient temperatures. 4.3 DSC 4.3.1 Design Evaluation The SAR was reviewed in conjunction with five calculation packages, References 53 through 57, and responses to several rounds of staff questions. 4.3.1.1 Normal Operating Conditions Fuel cladding temperature limits based upon the methodology of PNL-6189 (Reference 47) have been proposed by the applicant. These limits are acceptable to the staff. The licensee must demonstrate that all fuel to be stored meets the criteria of PNL-6189, which are the accepted limits. The applicant has provided analyses demonstrating that these limits can be satisfied for normal and off-normal conditions provided that the fuel meets the acceptance criteria for storage. The normal operating condition at 21 °C (70°F) ambient air inlet temperature and the high temperature limiting normal case at 37.8°C (l00°F) ambient air inlet temperature 'Yere 4-6
analyzed for the DSC and internals. HEA.TING-6 input and output for the 21 °c (70°F) case and the corresponding HSM run were reviewed. No errors were detected. Trends and magnitude of the resulting temperature distributions are reasonable. The HEATING-6 computer program is an industry standard code which is widley used for nuclear power plant thermal design analyses and has been in use for about twenty years. Application Qf the code for thermal design analysis of the standarized NUHOMS spent fuel storage system was perform~ in a con~tiye manner wh~_inpu_t parameters were chosen so that __ _ ~ _ conservatively high fuel cladding and HSM concrete temperatures were calculated. For the 21 °C (70°F) ambient case, maximum cladding temperatures of 366°C (691 °F) for PWR fuel and 417°C (782°F) for BWR fuel are below the acceptance criteria of 384°C (724°F) for PWR fuel and 421 °C (790°F) for BWR fuel. For the limiting normal case of 37.8°C (100°F) ambient, the cladding temperatures are 371 °C (699°F) for PWR and 420°C (788°F) for BWR fuel. These cladding temperatures are also below the acceptance criteria of 384 °C (724 °F) and 421 °C (790°F), respectively, for PWR and BWR fuel. The temperature distribution within a spacer disk was determined from HEATING-6 calculations for the 37.8°C (l00°F) ambient temperature case. Results of calculations with both helium and steel in the space between the guide sleeves and the DSC shell were used to determine the temperature distribution. The method used is appropriate for determining a temperature distribution for use in structural loading evaluations. 4.3.1.2 Off-Normal Conditions The off-normal condition considered is the 52°C (125°F) ambient inlet air temperature. HEATING-6 calculations were performed which yielded a maximum cladding temperature of 423°C (793°F) for BWR fuel and 374°C (705°F) for PWR fuel compared to the acceptance criterion of 570°C (1058°F). 4.3.1.3 Accident Conditions The applicant has analyzed the complete blockage of all air inlets and outlets for a 5-day period. This adiabatic heatup is addressed in References 50, 51 and 52. The fuel temperatures were calculated for this 5-day heatup period. A steady-state temperature distribution was assumed within the DSC, since its heatup rate is faster than that of the HSM. The resulting temperature distribution is acceptable for use in determining thermal loads. At the end of this time, BWR fuel has reached a temperature of 495°C (923°F), while PWR fuel has reached 447°C (836°F). These temperatures are below the acceptance criteria of 570°C (1058°F) for accident conditions. 4.3.2 Discussion and Conclusion For normal operating temperatures the maximum fuel cladding temperature is below the acceptance criteria for both the 52-B and the 24-P designs. Therefore the fuel cladding is expected to be protected against degradation leading to gross rupture during long-term storage and the proposed maximum heat loads are acceptable. 4-7
Maximum temperatures for both PWR and BWR fuel remain below the acceptance criteria of 570°C (1058°F) off-normal conditions and for accident conditions following five days of adiabatic heatup. With a daily inspection frequency, as required by the NRC, the concrete temperature of the HSM does not exceed the l77°C (350°F) accident limit. 4.4 TC The SAR was reviewed in conjunction with two calculation packages, References 58 and 59. 4.4.1 Design Evaluation During loading, evacuation, and transport to the RSM, the DSC is located within a transfer cask. In this case the steady-state temperature distribution through the cask was determined by modeling the cask as a series of cylindrical annular regions to determine the radial distribution and as a series of heat slabs to determine the axial distribution. Both vertical and horizontal orientations of the transfer cask were considered. 4.4.1.1 Normal Operating Conditions Ambient temperatures of -17.8°C (0°F), 21 °C (70°F) and 37.8°C (l00°F) were considered for normal operation. The surface temperature at the top and bottom of a horizontal DSC was determined for thermal stress evaluation. Axial temperature distribution was also determined for each of the three ambient temperatures. Maximum DSC surface temperature of 113°C (235°F) occurred at the top of the horizontal cask for the 37.8°C (l00°F) case with the 24-P design heat load. 4.4.1.2 Off-Normal Operating Conditions Extreme ambient conditions of -40°C (-40°F) and 52°C (l25°F) were considered as off-normal events. Axial and radial temperature distributions were determined using the same methods as for normal operating conditions. A solar shade will be used whenever temperatures exceed 37.8°C (100°F). Therefore the 37.8°C (l00°F) case becomes the limiting case for thermal gradient determination, yielding a 16.1 °C (61 °F) through wall temperature gradient. Vacuum drying of the DSC before backfill with helium will result in increased fuel temperatures relative to normal conditions due to the decreased heat transfer within the DSC. Methods simi1ar to that used for the normal operation case were used to determine the temperature gradient through the transfer cask wall, except that in this case the cask was oriented vertically. With an internal vacuum in the DSC, the maximum fuel cladding temperature was calculated to be 531 °C (988°F) for BWR fuel and 487°C (909°F) for PWR fuel, both below the short term or accident temperature limit of 570°C (1058°F). These fuel clad temperatures for PWR and BWR fuels are calculated as steady state temperatures. It should be noted that the actual time required for vacuum drying of the DSC is small 4-8
compared to the time necessary for the fuel cladding temperature to reach the calculated maximum value. 4.4.1.3 Accident Conditions While references 58 and 59 consider the accident condition of loss of the neutron shield material, these results were not used in the structural loading evaluation since complete loss of the solid neutron shield material is not postulated to occur. Instead thermal loads from the 37.8°C (100°F) normal operation case were used. 4.4.2 Discussions and Conclusions The limiting thermal condition of 37.8°C (l00°F) with solar heat load was used to determine the thermal loading for all cases. Provided that a solar shade is used whenever the ambient temperature exceeds 37~8°C (100°F), the use of the 37.8° (l00°F) case to determine the thermal loads is acceptable. Since the DSC will be in the transfer cask for relatively short periods compared to the storage lifetime, use of the short term accident temperature limit for maximum fuel temperature is acceptable. The maximum fuel temperature is below this limit for all of the cases considered. 4-9
Table 4.2 Summary of Component Temperatures as a Function of Ambient Temperatures Temperatures ( 0 F) DSC Inside HSM DSC Inside TC Ambient Cladding Cladding Limit Concrete Concrete Limit Cladding Cladding Limit 70 691 (PWR) 724 (PWR) 175 (PWR) 200 (normal steady 782 (BWR) 790 (BWR) 158 (BWR) state operation) 100 699 (PWR) 724 (PWR) 212 (PWR) 2001 (DSC Vacuum) 1058 (off-normal, 788 (BWR) 790 (BWR) 193 (BWR) 909 (PWR) but not 988 (BWR) accident) 125 705 (PWR) 1058 250 (PWR) 350 (steady state) 793 (BWR) 230 (BWR) 125 836 (PWR) 1058 480 (PWR) 35()2 (5 day 923 (BWR) 480 (BWR) adiabatic (350@ 40 heatup) hrs.)
- 1. If concrete temperatures of general or local areas .exceed 200°F but would not exceed 300°F, no tests or reduction of concrete strength are required if Type II cement is used and aggregates are selected which are acceptable for con~rete in this temperature range.
- 2. Use of any Portland cement concrete where "accident" temperatures may exceed 350°F requires subrnis~ion of tests on the exact concrete mix (cement type, additives, water-cement ratio, aggregates, proportions) which is to; be used. The
' tests are to acceptably demonstrate the level of strength reduction which needs to be applied, and to sho\v that the increased temperatures do not cause deterioration of the concrete either with or without load. ~
I 4-10
5.0 CONFINEMENT BARRIERS AND SYSTEMS EVALUATION
- 5. l Description of Review The primary confinement boundaries for fission products which are contained in tlle spent fuel are the intact fuel cladding (no known or suspected gross cladding breeches) and the DSC, which is a welded steel cylinder that is vacuum dried and backfilled_ with helium._ Th~ _
HSM is designed to provide shielding, structural support, ventilation, and weather protection for the DSC, but is not part of the confinement boundary. Similarly, the TC is designed to provide shielding during handling and transfer operations, but is not part of the confinement boundary. The main parts of the secondary confinement boundaries for both versions of the DSC are a shell, outer bottom and top cover plates, top and bottom shield plugs, and inner top and bottom cover plates. The basket assembly is not part of the confinement boundary. The only penetrations required in the DSC are in the siphon and vent block, which is a part of the top shield plug. Two penetrations (with quick disconnect fittings) for vacuum drying and backfilling with helium are located in this block. No credit for confining the helium atmosphere is taken by the disconnect fittings. Two cover plates that mate with the siphon block are seal welded over the penetrations after the drying and helium back.filling operations have been completed. No components are required to penetrate the DSC after helium backfilling is completed and the structural lid is welded in place. No penetrations are used during spent fuel storage. Tables 8.l-4a and 8.1-4b of the SAR report the design basis internal helium pressure for the PWR and BWR versions of the DSC. The PWR canister has marginally higher internal pressure but in all cases is very low. For normal operations with intact fuel cladding on the design basis average ambient temperature day 21 °C (70°F), the internal helium pressure is 34.5 kPag (5.0 psig). For a 37.8°C (100°F) ambient temperature day, the internal helium pressure is only 47.6 kPag (6.9 psig). The accident case considers a 52°C (125°F) day with the HSM vents blocked and 100 percent cladding failure with the release of all of the fuel rod fill gas and 30 percent of the fission gas geHerated in PWR assemblies irradiated to 40,000 MWD/MTU. The DSC is designed to meet the requirements of ASME Code, Section ID, Subsection NB, and constructed in accordance with the ASME Code, Section m, Article NB-4000. 5.2 Design Evaluation The staff considered Paragraph (1) of 10 CFR 72.122(h) as pertinent to storage of spent fuel in DSC .. It requires that "spent fuel cladding must be protected during storage against degradation that leads to gross ruptures" and "that degradation of the fuel during storage will not pose operational safety problems with respect to its removal from storage." Also, 5-1
10 CFR 72.236(e) requires that the cask must be designed to provide redundant sealing of confinement systems.
- The confinement barriers and systems design are considered acceptable if it is demonstrated that: (1) there is a high likelihood that the DSC internal helium atmosphere will remain intact; (2) there is no operable corrosion mechanism that will lead to failure of the DSC to
_provide confinement; (3)-there is no long-term cladding degradation mechanism_ in a helium atmosphere which could cause significant degradation or gross ruptures; and (4) there is insufficient time for cladding or fuel degradation during cask dry-out or off-normal behavior that could pose operational problems with respect to the removal of fuel from storage. An NRC review of the NUHOMS design was made and documented in Section 5.0 of the Safety Evaluation Report for NUHOMS-24P, Revision 1, dated April 1989 (Reference 39). The NRC staff revie.w was directed at two aspects of the design: (1) the mechanical integrity of the DSC, and (2) the long-term behavior of the cladding in an inert environment. The review was also directed at the impact of cask dry-out and off-normal behavior on fuel removal. The present review was directed at examining the review made in Section 5. 0 of the SER of NUHOMS-24P to ensure that the results of that review also apply to the Standardized NUHOMS-24P and 52-B DSCs. Because all of the parts of the confinement boundary are fabricated from stainless steel, the DSC is adequately protected from corrosion mechanisms. The staff reviewed DSC integrity from the point of view of weld quality and inspections, adequacy of leak check methods on welds, other leakage paths, and long-term helium migration. Reviewers also checked the calculated stresses in the DSC under normal, off-normal, and accident conditions in order to verify that they are in the acceptable range. Cyclic fatigue of the DSC was also reviewed. The staff evaluated cladding degradation by reviewing the pertinent technical literature in order to identify known and postulated mechanisms of gross failure of fuel in an inert atmosphere. Based on the literature search, calculations were performed for postulated failures by the mechanism of diffusion controlled cavity growth using a conservative set of assumptions. This was the only failure mechanism considered likely under the DSC storage conditions. The staff also evaluated the possible long-term creep or sag of the fuel cladding under the storage conditions since creep could affect removal of the fuel from storage. The effects of oxidation during the fuel dry-out period were also considered. In its analysis of the cavity growth mechanism, the NRC staff determined that the area of decohesion at the end of a 20-year storage life is less than 4 percent, not high enough to cause any concern. The NRC staff found that creep or sag of the fuel cladding might equal 0.05 cm (0.020 inch), much less than the clearance available for removal of the rods. For postulated fuel oxidation of defective fuel rods during cask dry-out or off-normal behavior, cladding strain was determined to be much less than 1 percent so that fuel defect e~tension or 5-2
fuel powdering is not anticipated. For all these areas of potential fuel degradation, the NRC staff calculations for the B&W fuel gave such conservative results, that they can equally well be applied to the fuels which meet the fuel specification cited in this SER. The NRC staff concludes that the standardized NUHOMS system design provides sufficient means to ensure that the fuel cladding is adequately protected against degradation that leads to gross ruptures. The staff verifi_ed ~ the design of the DSC _provid~ redundant sealing.
- 5. 3 Conclusions The staff concludes that the standardized DSC design conforms to the relevant criteria in 10 CPR 72.122(h), and 10 CPR 72.236(e) regarding redundant sealing of confinement systems. Confinement is ensured by a combination of inspection techniques, including radiographic inspection, helium leak testing, and dye penetrant testing. The confinement capability of the empty DSC shell without either bottom or top plate assemblies is ensured by radiographic inspection of the longitudinal full penetration weld and the girth weld. Helium leak testing is performed to ensure adequate sealing of the inner bottom cover to the DSC cylindrical shell. The confinement capability of the loaded DSC is ensured by helium leak testing after welding and dye penetrant testing of the inner top cover plate and vent port block to the DSC shell. All partial penetration welds are multiple pass welds subjected to dye penetrant testing. The inner seal welds are also helium leak tested. The outer seal welds are dye penetrant tested.
A number of tests and specifications relevant to confinement integrity were proposed by the vendor and determined to be appropriate. These include: DSC vacuum pressure during drying (SAR Section 10.3.2) DSC helium back.fill pressur~ (SAR Section 10.3.3) DSC maximum permissible leak rate of inner seal weld (SAR Section 10.3.4) DSC dye penetrant test of closure welds (SAR Section 10.3.5). These are included as conditions for system use in Section 12.2 of this report. The staff also requires the following condition for the use of the system:
- 1. If fuel needs to be removed from the DSC, either at the end of service life or for inspection after an accident, precautions must be taken against the potential for the presence of oxidized fuel and to prevent radiological exposure to personnel during this operation. This can be achieved with this design by the use of the penetration valves which permit a determination of the atmosphere within the DSC before the removal of the shield plug. If the atmosphere 5-3
within the DSC is helium, then operations can proceed normally with fuel removal either via the transfer cask or in the pool. However, if air is present within the DSC, then appropriate filters should be in place to preclude uncontrolled release of airborne radioactive particulate from the DSC via the penetration valves. This will protect both personnel and the operations area from potential contamination. For the accident case, personnel protection is required as ~ppropri~te. The above is included as a condition for system use in Section 12.1.2 of this report. Because the DSC confinement barrier material is stainless steel, adequate provision for corrosion protection is part of the DSC design. Additionally, the fluence of the neutron flux for a 20-year period of storing the spent fuel is eight orders of magnitude less than the fluence-encountered within an operating reactor. For this reason embrittlement due to neutron flux is not considered to be a concern. A discussion of neutron embrittlement follows. The fuel cladding which has been placed within the DSC has been subject to extensive neutron irradiation while it was present in the reactor core producing power. A representative order of magnitude total neutron flux within the core of an operating nuclear power plant is approximately 1 E+ 12 neutrons/sq. cm.-sec. Core residence time at power for nuclear fuel prior to its placement in the NUHOMS ISFSI is about 28.8 months (80% power operation over a 36-month time period). This results in a total fuel cladding irradiation neutron fluence of 7.5 E19 neutrons/sq. cm. In comparison, the bounding total neutron emission per fuel assembly delineated in the NUHOMS fuel specification is 2.23 E8 neutrons/second. This source, distributed over the entire surface of the decay heat producing section of the fuel rods in a fuel assembly, results in an average neutron flux of 8.9 E+2 neutrons/sq. cm.-sec. Over the 20-year life of the ISFSI, this additional neutron fluence to the cladding would be 5.6 El 1 neutrons/sq. cm. which is ei~ht orders of m~nitude lower than the fluence already accumulated during power operation. The neutron fluence to the fuel cladding represents an insignificant addition to the fluence absorbed by the cladding during power operation. The long-term ISFSI storage would, therefore, not be expected to create any neutron irradiation induced damage to fuel cladding beyond that already caused by irradiation of this fuel while residing in the reactor core during power operation. The staff considered three potential mechanisms for the deterioration of the integrity of fuel rods. The first was potential failure of the cladding by the diffusion controlled cavity growth mechanism. The staff determined that the area of decohesion was less than 4 percent, not high enough to cause any concern. The second mechanism examined was creep of the fuel cladding. It was found to be a maximum of 0.05 cm (0.020 inches), much less than the clearance available for removal of the rods. The third mechanism examined was oxidation of the fuel during the dry-out period. Cladding strain was determined to be much less than 1 ~rcent for postulated fuel oxidation of defective fuel rods. The staff concludes that the 5-4
DSC design has provided sufficient me.ans to assure that the fuel cladding is adequately protected against degradation that leads to gross rupture. 5-5
6.0 SHIELDING EVALUATION 6.1 Design Description The radiation shielding for the stored fuel assemblies .is provided by a variety of shielding materials and operational procedures. The RSM has thick concrete walls and roof as well as ~ ~~YY- shiel9ed <!oor ai:id shielded air outlet vents to reduce radiation.- The DSC has thick shield plugs on both ends to reduce the dose to plant workers. The TC has shielding incorporated in both ends as well as the entire shell. From the operational side of the design, placement of demineralized water in the annulus between the TC and DSC provides shielding as well as reducing contamination of the DSC exterior as well as the TC interior surfaces. The use of water in the DSC cavity during placement of the DSC inner seal weld reduces direct and scattered radiation exposure. Temporary shielding is used during DSC draining, drying, inerting and closure operations. 6.2 Design Evaluation (Source Specification and Analysis) The neutron and gamma ray radiation source terms were calculated for design basis fuels (for the PWR B&W 15x15 fuel and for the BWR GE 7x7 fuel). For both fuels a maximum initial enrichment of 4.0 wt% U-235 is assumed and a post-irradiation cooling time equivalent to five years is assumed. The PWR fuel is assumed to be subjected to an average fuel burnup of 40,000 MWD/MTU; the BWR fuel is assumed to be subjected to an average fuel burnup of 35,000 MWD/MTU. Neutron source are based on spontaneous fission contributions from six nuclides (predominantly Cm-242, Cm-244, and Cm-246 isotopes) and (a, n) reactions due almost entirely to eight alpha emitters (predominantly Pu-238, Cm-242, and Cm-244). The fission spectrum used in shielding calculations is a 1,ueighted combination of the principal contributors. The total neutron source strength for PWR fuel is 2.23E8 neutrons per second per assembly (or 5.35E9 neutrons per second for 24 PWR fuel assemblies). The total neutron source strength for BWR fuel is 1.01E8 neutrons per second per fuel assembly (or 5.25E9 neutrons per second for 52 BWR fuel assemblies). For the BWR fuel the neutron and gamma source strengths and gamma energy spectrum and decay heat were calculated using the ORIGEN 2 computer code (References 60, 61). ORIGEN 2 is a widely used and validated code which has been utilized and approved for previous ISFSI radiation source term calculations. The heavy metal weight and the weights of other materials of the fuel assembly are chosen to bound those for BWR fuel assemblies to give the highest possible values for neutron and gamma source strengths and decay heat. Gamma radiation sources include 70 principal fission product nuclides within the spent fuel and several activation products and actinide elements present in the spent fuel and fuel assemblies. The gamma energy spectrum includes contributions for each source isotope as 6-1
calculated by ORIGEN 2 calculations. The total gamma source strength for BWR fuel is 4.86E15 Mev/s/MTHM (or 1.37E17 pI,otons/second for 52 BWR fuel assemblies). For the PWR fuel neutron source data and neutron spectrum were taken from previous ORIGEN 2 calculations; the results of which bound data from the Office of Civilian Radioactive Waste Management (OCRWM) database (Reference 62). Gamma ray sources were d~tennined using _the OCRWM database with the gamma spectrum _determined -using -the - Microshield computer program (Reference 63). The spectrum results were segmented into the 18 energy group structure used in the shielding calculations with the results normalized to preserve the total gamma power calculated in the OCRWM database. The total gamma source for PWR fuel is 5.81E15 Mev/s/MTHM (or 1.79E17 photons/second for 24 PWR fuel assemblies). The neutron-energy spectrum used for the shielding analysis is given in Table 7.2-la of the SAR for PWR fuel and in Table 7.2-lb of the SAR for BWR fuel. The gamma energy spectrum used for shielding analysis is given in Table 7.2-2a of the SAR for PWR fuel and in Table 7.2-2b of the SAR for BWR fuel. The shielding analysis used the same suite of computer codes as those used in the NUHOMS-24P Topical Report. These computer codes are: ORIGEN-2, ANISN (Reference 64), QAD-CGGP (Reference 65), MICROSKYSIIlNE, and MICROSHIELD. Collectively these codes were used to calculate both the gamma and neutron direct and scattered dose rates and, as previously discussed, the radiation source terms for spent fuel assemblies. All of the these computer codes have been used and benchmarked throughout the nuclear industry. - 6.3 Discussion and Conclusions The staff's review of the standardized NUHOMS system shielding calculations included a combination of reviewing the files provided by the applicant and performing independent check calculations on the files. There were no independent audit calculations of the dose rates since the applicant has demonstrated its proficiency in the application of these same methods and the validation and verification of these computer codes for previous ISFSI applications. The check calculations of the shielding analysis files did not reveal any arithmetic and/or other numerical errors or indicate any changes to be made in the calculations. The calculated dose rates appear to be comparable to previously calculated dose rates. This conclusion also applies to the applicant's calculation of direct and air-scattered dose rates in and around the HSM. Dose rates at locations of interest were calculated for 5-year cooled PWR fuel and presented in Figure 7.3-2 of the SAR. The SAR states that "Consistent with the relative design basis source strengths, the shielding analysis results for the NUHOMS-24P envelop those of the NUHOMS-52B systems." The calculation packages submitted by the applicant presented sufficient information to support this assumption. 6-2
The applicant has performed an extensive number of shielding dose rate analyses for the standardized NUHOMS system design using a neutron and gamma ray source term for 4.0 wt% PWR fuel irradiated to 40,000 MWD/MTHM and cooled for a period of five years. This source term has been shown by the applicant to bound that calculated for 4.0 wt% BWR fuel irradiated to 35,000 MWD/MTHM and cooled for a period of five years. For both the PWR and BWR fuels the source terms were calculated to be conservative relative to the fuel types of possible concern. The PWR neutron source data and source spectra were_ calculated --using-the ORIGEN-2 co-mputer code and tlie results were found to bound data from the OCRWM database. Gamma ray sources were taken from the OCRWM database with the gamma spectrum determined using the MICROSHIELD computer program. The computer codes and methods used to delineate the neutron and gamma sources for the standardized NUHOMS system design are acceptable. Using the ANISN, QAD-CGGP, MICROSKYSHINE, and MICROSIIlELD computer codes, the applicant calculated both direct and air-scattered dose rates in and around the HSM. These codes and methods have been previously used by this applicant and were reviewed and approved for the NUHOMS-24P Topical Report. The calculated dose rates presented for the standardized NUHOMS design appear to be consistent and conservative relative to previously presented results. Independent calculations of the applicant's shielding analysis files did not reveal any arithmetic and/or other numerical errors or indicate any changes to be made in the calculations. Based on a detailed review of the inputs, methods, computer codes, assumptions and dose rate results, including calculational checks of the shielding analysis files, the shielding design analysis for the standardized NUHOMS system has been found to be acceptable and should be sufficient to meet the requirements of 10 CFR 72.104 and 10 CFR 72.106 as required by 10 CFR 72.236(d). In accordance with 10 CFR 72.212(b)(2)', users must perform written evaluations to establi_¥I that these requirements have been met. 6-3
7.0 NUCLEAR CRITICALITY SAFETY EVALUATION From the standpoint of criticality safety, the standardized NUHOMS system consists of two separate designs; one for the storage of 24 irradiated PWR fuel assemblies which is referred to as the standardized NUHOMS-24P design; and the one for the storage of 52 irradiated BWR fuel assemblies which is referred to as the standardized NUHOMS-52B design. Therefo~ the cri_ticality _safety_ ey-;µuatiQn of the two designs will_ be discussed separately. 7 .1 Design Description 7.1.1 Standardized NUHOMS-24P Design Criticality safety, according to the vendor, is ensured by the inherent geometry and material characteristics of the standardized NUHOMS-24P system and by establishing specific criteria for acceptance of irradiated fuel assemblies for storage. There is not a specific design feature such as fixed neutron poisons intended to provide assurance of nuclear criticality safety. The system is designed to provide nuclear criticality safety during both wet loading and unloading operations. 7.1.2 Standardized NUHOMS-52B Design Criticality safety, according to the vendor, is ensured through a combination of geometrical and neutronic isolation of fuel assemblies. Fixed neutron absorbers in the form of borated stainless steel plates are used to control the reactivity of the assembly of stored BWR fuel assemblies such that criticality safety is assured under optimum moderation conditions for all initial fuel enrichments equal to or less than 4.0 wt.% U-235.
- 7. 2 Design Evaluation 7.2.1 Standardized NUHOMS-24P Design In addressing nuclear criticality safety for the st.:.11darclized NUHOMS-24P design, the staff has applied criteria in 10 CPR 72.124 and 10 CPR 72.236(c). 10 CPR 72. lL--+ provides that the system should be designed to be maintained subcritical and to ensure that, before a nuclear criticality accident is possible, at least two unlikely, independent, and concurrent or sequential changes occur. It states that the design of the system must include margins of safety for the nuclear criticality parameters that are commensurate with the uncertainties in the data and methods used in calculations. It calls for the design to demonstrate safety for the handling, packaging, transfer, and storage conditions, and in the nature of the immediate environment under accident conditions. It states that the design should be based on favorable geometry, permanently fixed neutron absorbing materials, or both. 10 CFR 72.236(c) requires that the cask must be designed and fabricated so that the spent fuel is maintained in a subcritical condition under credible conditions.
7-1
The design criteria propoSed by the vendor in the SAR are that~' remains below 0.95 during both normal operation and accident conditions. These design criteria were determined by the staff to be acceptable. The criticality evaluation of the standardized NUHOMS system design was presented in SAR Section 3.3.3. The vendor performed criticality calculations to determine the most reactive fuel type and to show criticality safety of the s t a n ~ NUHOMS-24:P system._ -- According to the vendor,-tlie B&W f5xl5 fuel-is the most reactive PWR fuel assembly and was selected as the design basis for the standardized NUHOMS-24P design. The criticality safety analysis for the standardized NUHOMS-24P system presented in the SAR was performed using a calculational methodology consisting of several standard computer programs: The CASM0-2 (Reference 66) computer program was used to calculate the irradiated fuel actinide number density data as a function of burnup. The SAS2 (Reference 67) sequence in the SCALE-3 (Reference 60) criticality safety analysis code system was used to calculate the reactivity of the array of stored irradiated fuel assemblies. The SCALE-3 code system used in the analysis presented in the SAR was a mainframe version of the programs and included ORIGEN-S to perform fuel bumup, depletion, and decay calculations, and KENO-IV (Reference 68) code for criticality calculations. The cross sections used in the criticality safety analysis were the 123 group library from the SCALE system. The KENO-IV code and the calculational methodology utilized to calculate~ was benchmarked against 40 critical experiments as presented in Section 3.3 of the SAR. The criticality analysis presented in the SAR and supplementary response to requests for additicnal information were reviewed by the staff. Some independent and confirmatory calculations were also performed to verify important sensitivities in the criticality analysis. 7.2.2 Standardized NUHOMS-52B Design For the standardizeo NUHOMS-52B design, the staff used the nuclear criticality safety criteria in 10 CFR 72.124 and 10 CFR 72.236(c). 10 CFR 72.124 provides that the system should .be designed to be maintained subcritical and to ensure that, before a nuclear criticality accident is possible, at least two unlikely, independent, and concurrent or sequential changes occur. It states that the design of the system must include margins of safety for the nuclear criticality parameters that are commensurate with the uncertainties in the data and methods used in calculations. It provides that the design should demonstrate safety for the handling, packaging, transfer, and storage conditions, and in the nature of the immediate environment under accident conditions. It also provides that the design should be based on favorable geometry, permanently fixed neutron absorbing materials, or both. 10 CFR 72.236(c) requires that the cask must be designed and fabricated so that the spent fuel is maintained in a subcritical condition under credible conditions. 7-2
In order to address the criteria of 10 CFR 72.124(b), the staff has considered the following. The staff used the bounding neutron flux of the 52B spent fuel and calculated the reaction rate from thermal neutron absorption in boron and then evaluated this rate for a 20-year storage life. The result of this calculation is a maximum depletion of boron of approximately 0.04 % for 20 years, which is small compared to the design tolerance of the absorber material and can therefore be considered insignificant. Aside from this boron depletion mechanism due to thermal neutron absorption in boron, the staff ,tias not p<>stuhµed other mechanisms which could reduce-the efficacy of the fixed neutron absorber. The design criteria proposed by the vendor in the SAR are that kur, remains below 0.95 during both normal operation and accident conditions for optimum moderation density. These design criteria were determined by the staff to be acceptable. The criticality evaluation of the standardized NUHOMS-52B design was presented in SAR Section 3.3.3. The vendor performed criticality calculations to determine the most reactive fuel type and to show criticality safety of the standardized NUHOMS-24P design. According to the vendor, the GE-2 7x7 fuel is the most reactive BWR fuel assembly and was selected as the design basis fuel for the standardized NUHOMS-52B. The criticality safety analysis for the standardized NUHOMS-52B design presented in the SAR was performed with a calculational methodology using the microcomputer application KENO-Va (Reference 69) and the Hansen-Roach 16 group cross section working library. A small computer program, designated PN-HEf, was developed by the vendor to automate the computation of resonance parameters necessary for mixing the Hansen-Roach cross sections. The KENO-Va/PN-HET code system was benchmarked against 40 critical experiments as presented in a separate computer code QA verification document, KENO5A-QA (Reference 70). The criticality analysis presented in the SAR and supplementary response to requests for additional information were reviewed by the staff. Some independent and confirmatory calculations were also performed to verify important sensitivities in the criticality analysis.
- 7. 3 Conclusions 7.3.1 Standardized NUHOMS-24P Design On the basis of the analysis presented in the SAR, the supplementary analysis presented in response to questions, and confirmatory calculations performed by the staff, it was determined that the standardized NUHOMS-24P design and proposed operating procedures are adequate to maintain the system in a subcritical configuration and to prevent a nuclear criticality ~dent, and therefore satisfy 10 CFR 72.124 and 10 CFR 72.236(c), subject to the key factors assumed by the vendor in the analysis. Specifically:
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- 1. Criticality safety calculations presented in the SAR and independent confirmatory calculations performed by the staff show that criticality safety is ensured for a maximum U-235 initial enrichment equivalent to 1.45 wt.%
which was determined for the design basis B&W 15x15 fuel assemblies.
- 2. The criticality safety analysis of the misloading of unirradiated fuel assemblies presented in the SAR-and independent confirmatory calculations performed by the staff show that the array reactivity can be maintained subcritical Orur<0.95) in this accident situation by filling the DSC with borated water before wet loading or unloading. The minimum level of boration required as determined by the staff analysis, based on 4.0 wt.% enrichment of unirradiated B&W 15xl5 fuel assemblies was determined to be 2,000 ppm. The analysis presented in the SAR determined the minimum level of boration to be 1,810 ppm. In Jj~ of resolution of the difference between the staff and SAR analysis of the required minimum level of boration, the more conservative value of the staff analysis is taken. The SAR assumed that the B&W 15x15 fuel presents a bounding case.
The key factors and assumptions used by the vendor in the criticality safety analysis are as follows:
- 1. The maximum initial fuel enrichment evaluated for irradiated fuel assemblies is 4.0 wt.% U-235.
- 2. The DSC is filled with borated water during fuel loading and unloading ope~tions. The required boron concentration is determined for maximum fuel enrichment.
- 3. Only irradiated fuel assemblies with an initial enrichment equivalent < 1.45 wt.% U-235 will be loaded into the DSC. The criticality acceptability curve of minimum burnup versus enrichment is shown in Figure 3.3-3 of the SAR.
- 4. Fuel assemblies are no more reactive than the design basis 15 x 15 rod array.
- 5. Accidents resulting in altered mechanical configuration of the array of fuel assemblies are not credible.
- 6. Accidents during dry storage that result in the flooding of the DSC with unborated water are not credible.
Key factors 1, 2, 3, and 4 are reflected in the fuel specification discussed in Section 12.2.1. Previous evaluation of vendor topical reports and site-specific applications involving casks with large number of assemblies (e.g., 24) have addressed the potential for criticality when 7-4
water is added to the cask or canister before fuel removal. Because NRC staff position does not yet allow for bumup credit, the past analyses have assumed a full load of fresh fuel and considered the case for optimum moderation. These analyses have been the limiting cases for nuclear criticality safety. Minimum boron concentration in the DSC cavity water, during wet loading and unloading operations, is discussed as a condition for system use in Section 12.2.15. 7.3.2- Standardized NUHOMS-521fDesign The conclusions of the analysis of the standardized NUHOMS-52B design are more straightforward since the standardized NUHOMS-52B system is designed to provide assurance of nuclear criticality safety under optimum moderation conditions for loading of unirradiated fuel assemblies of a maximum enrichment of 4.0 wt.% U-235. This simplification is due to the use of fixed neutron absorbers in the design. The initial SAR also requested certification of a low-enrichment design which contained no neutron absorber plates but limited the initial fuel enrichment. Independent criticality safety calculations performed by the staff did not confirm that criticality safety was ensured in this low enrichment design. The vendor has withdrawn the low-enrichment design from further consideration. On the basis of the analysis presented in the SAR and subsequent revisions, and independent confirmatory calculations performed by the staff, it was determined that the standardized NUHOMS-52B system design and proposed operating procedures are adequate to maintain the system in a subcritical configuration and to prevent a nuclear criticality accident, and therefore satisfy 10 CPR 72.124 and 10 CFR 72.236(c), subject to the key factors assumed by the vendor in the analysis. Specifically:
- 1. Criticality safety calculations presented in the SAR and independent confirmatory calculations performed by the staff show that criticality safety is ensured for a maximum initial U-235 fuel enrichment of 4.0 wt.% which was determined for the design basis GE-2 7x7 fuel assembly.
- 2. The criticality safety analysis assumes a minimum boron density of 0. 75 wt%
boron in the borated stainless steel absorber plates. The key factors and assumptions used by the vendor in the criticality safety analysis are as follows:
- 1. The maximum initial fuel enrichment of fuel assemblies stored in the standardized NUHOMS-52B system is 4.0 wt.% U-235.
- 2. The boron loading in the neutron absorber plates is a minimum of 0. 75 wt.%.
7-5
- 4. Accidents resulting in an alte,.ed mechanical configuration of the array of fuel assemblies are not credible.
Key factors 1, 2, and 3 are reflected in the fuel specification discussed in Section 12.2.1. 7-6
8.0 RADIOLOGICAL PROTECTION EVALUATION 8.1 Design Description The main radiation protection features of the standardized NUHOMS system design are described in Sections 7.1.2 and 7.3 of the SAR and include: (1) radiation shielding; (2) __ radioacti_ve material confinement; (3)_ prevention_ of external surface contamination; and (4) site access control. Shielding includes many features designed to reduce direc~ and scattered radiation exposure, including:
- 1. Thick concrete walls and roof on the HSM which limit the dose rate to site workers and the off-site population;
- 2. A thick shield plug on each end of the DSC to reduce the dose to workers performing drying and sealing operations, and during transfer of the DSC in the transfer cask and storage in the HSM;
- 3. Use of a shielded transfer cask for DSC handling and transfer operations which limits the dose rate to ISFSI and plant workers;
- 4. Filling of the DSC cavity and the DSC-transfer cask annulus with water during DSC closure operations to reduce direct and scattered radiation exposure ; and
- 5. Use of temporary shielding during DSC draining, drying, inerting and closure operations as necessary to further reduce direct and scattered radiation dose rates.
The confinement features of the standardized NUHOMS system control the release of gaseous or particulate radionuclides and are described in Section 3.3.2. These features include:
- 1. The cladding of the stored fuel assemblies, which provides the first level of confinement;
- 2. The DSC confinement p~ure boundary which provides the second level of confinement. The DSC confinement boundary includes: the DSC shell, the inner seal weld primary closure of the PSC, the DSC shielded end plugs, the outer seal weld secondary closure of the DSC, and the DSC cover plates.
The DSC has been designed as a weld-sealed containment pressure vessel with no mechanical or electrical penetrations. All the DSC pressure boundary welds are inspected according to the appropriate articles of the ASME B&PV Code, Section III, Division 1, Subsection NB 8-1
(Reference 9). These criteria ensure that the weld metal is 33 sound 33 the parent metal. As pointed out in the description of the DSC in Section 3.2.1, the double seal welds at the top and bottom of the DSC do not comply with the ASME Code. Consequently, the weld inspection requirements are also not strictly in accordance with Section NB-5000 of the Code. The staff has accepted alternative inspection and test requirements in lieu of the Code. Contamination of the DSC exterior and transfer cask interior surfaces is controlled by placing demineralized water in the transfer cask and DSC during loading operations, then sealing the DSC/cask annulus. In addition, surface contamination limits for the DSC have been established, and are discussed in Section 9.2. Access to the site of the NUHOMS ISPSI would be restricted by a periphery fence to comply with 10 CPR 72.106(b) controlled area requirements. The details of the access control features will vary from site to site, but must meet the requirements of 72.106(b) for the access to the controlled area. In addition to the controlled area restrictions, access to the spent fuel is restricted by an HSM access door, which is welded in place. This door weighs approximately 2. 7 t (3 tons) and would require heavy equipment for removal. 8.2 Design Evaluation This section evaluates the radiation protection features of the standardized NUHOMS design separately with regard to (1) on-site occupational exposures under normal loading and storage conditions, and (2). off-site exposures under normal storage conditions and in the event of accidents. 8.2.1 On-Site Radiological Protection Regulatory requirements for on-site radiological protection are contained in 10 CPR 20.1101 and 20.1201-1208 which require the licensee to provide the means for controlling and limiting occupational radiation exposures within the limits given in 10 CPR Part 20 and for meeting the objective of maintaining exposures 33 low 33 is reasonably achievable (ALARA). Section 20.1201(a) of 10 CPR Part 20 states that the licensee shall control the occupational dose to individual adults to the dose limits specified in 120l(a)(l) and 1201(a)(2). Also, section 20.1101 of 10 CPR Part 20 states that each licensee shall develop, document, and implement a radiation protection program and that the licensee shall use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low 33 is reasonably achievable (ALARA). Section 72.126(a) provides that radiation protection systems shall be provided for all areas and operations where on-site personnel may be exposed to radiation or airborne radioactive materials. 8-2
Guidance for ALARA considerations is also provided in NRC Regulatory Guides 8.8 and 8.10 (References 10 and 11). Radiation protection for on-site personnel is considered acceptable if it can be shown that the non-site-specific considerations (1) will maintain occupational radiation exposures at levels which are ALARA, (2) are in compliance with appropriate guidance and/or regulations, and (3) will ensure that the dose from associated activities to any individual does not exceed the limits of 10 CPR Part 20. The calculational methods used ih the estimation of on-site doses are described in detail in the SAR. These methods focused on the use of the ANISN, QAD-CGGP,
- MICROSKYSHINE, and MICROSHTBTD (References 64, 65, and 63, respectively) radiation transport codes, as well as manual calculations, to calculate exposure rates around the DSC in a transfer cask and an HSM. Dose rate maps in the g~eral vicinity of a 2x 10 array and two lxlO arrays containing 10-year-old fuel were constructed.
- The calculational methods and results presented in the SAR and associated calculation' packages were reviewed for completeness, correctness, and internal consistency.* In addition, confirmatory calculations were performed for the gamma-ray dose rates at various locations around the DSC, TC, and HSM. ,
Radiation doses to on-site workers were not calculated in the SAR. Rather, a summary of the operational procedures which lead to occupational exposures are presented, as are the number of personnel required, the estimated time for completion, and the average source-to-subject distance. This information can be used with dose map results to assess the individual and collective on~site doses. The SAR notes that experience with an operating standardized NUHOMS system has shown that the collective occupational dose associated with placing a canister of spent fuel into dry storage is less than 0.014 person-Sv (1.4 person-rem), and that the application of effective procedures by experienced ISFSI personnel can reduce the collective dose to below 0.01 person-Sv (1 person-rem) per canister. A detailed assessment of operator doses and the possible provision of management or administrative controls to meet ALARA ~riteria is the responsibility of the user in accordance with its 10 CPR Part 50 licensee's radiation protection program and 10 CPR Part 20. Other workers at the nuclear power plant site will also be exposed to direct and air-scattered (skyshine) radiation during the transfer and storage phases of ISFSI operation. Examples of activities involving such exposure are surveillance of the HSMs, and site operations which are not associated with spent fuel storage but which are performed in the general vicinity of the storage area. Major factors influencing the magnitude of the exposures are the occupancy times and spatial distribution of workers, and the intensity of the radiation field. An assessment of the expected on-site doses incurred by site personnel not directly involved 8-3
in ISFSI operations is the responsibility of the user in accordance with its 10 CFR Part 50 licensee's radiation protection program and 10 CFR Part 20. 8.2.2 Off-Site Radiological Protection 8.2.2.1 Normal Operations Regulatory requirements for off-site radiological protection in 10 CFR Part 20 require that dose to members of the public should be kept within the limits of 20.1301 and should be ALARA. Section 72.104(a) of 10 CFR Part 72 requires that during normal operations and anticipated occurrences, the annual dose equivalent to any real individual located beyond the controlled area shall not exceed 0.25 mSv (25 mrem) to the whole body, 0.75 mSv (75 mrem) to the thyroid, and 0.25 mSv (25 mrem) to any other organ as a result of exposure to (1) planned discharges of radioactive materials (except for radon and its daughter products) to the general environment, (2) direct radiation from ISFSI operations, and (3) any other radiation from uranium fuel cycle operations within the region. Off-site radiological protection features of the standardized NUHOMS system are considered acceptable if it can be shown that design and operational considerations, which are not site-specific, result in off-site dose consequences in compliance with the applicable sections of 10 CFR Parts 20 and 72, and that these doses to off-site individuals are ALARA. The two principal design features which limit off-site exposures during normal operations are the confinement features of the double-seal welded DSC, and the radiation shielding of the DSC and the HSM. During transfer operations, shielding in the radial direction is provided by the transfer cask. The confinement features of the DSC control the release of gaseous or particulate radionuclides. There are no liquid effluents from the ISFSI. During normal operations, the only pathway of exposure to the off-site population is direct and scattered radiation from the stored fuel. The review for off-site radiological protection mainly involved a detailed evaluation of the methods applied and the results obtained in the applicable SAR sections, supplemented by additional information (including detailed calculation packages) provided by the applicant on these methods and results. For the case of off-site doses from direct and scattered (or "skyshine") radiation, an evaluation was performed on the application of MICROSKYSHINE, MICROSHIELD, and manual calculation methods, which were used to calculate gamma-ray and neutron dose equivalent rates at various locations in and around the HSM and to construct a dose-versus-distance curve. The dose rates predicted by this curve for various off-site distances was used to assess the general level of compliance with the dose rate criteria of 10 CFR 72.104(a) and for 10 CFR Part 20. The dose to an off-site individual residing at some distance from a filled standardized NUHOMS system array will vary depending on a number of factors, including fuel type, size, and geometry of the array, and the directional orientation of the receptor with respect to 8-4
the array. A conservative estimation of the distance required to reduce the full-time occupancy dose rate from a filled ISFSI array to 0.25 mSv/yr (25 mrem/yr) is approximately 300 meters. Normal operation of a standardized NUHOMS HSM would comply with the dose rate criteria of 10 CFR 72.104(a), provided site-related factors allow for a sufficient distance to the controlled area boundary. As required by 10 CFR 72.212(b)(2)(iii}, this must be evaluated by the user before storing fuel in a standardized NUHOMS system ISFSI. 8.2.2.2 Off-Normal Operations Section 72.106(b) requires that any individual located on or near the closest boundary of the controlled area (at least 100 m) shall not receive a dose greater than 0.05 Sv (5 rem} to the whole body or any organ from any design basis accident. Off-normal events and postulated accidents that could result in the loss of shielding or the release of radionuclides are analyzed in Sections 8.1 and 8.2 of the SAR. In particular, an accident resulting in an instantaneous release of 30 percent of fission gas inventory (mainly Kr-85) is assessed in Section 8.2.8. The SAR rei;x>rts that this accident results in a dose at 300 meters from the ISFSI site of 0.53 mSv (0.053 rem} to the whole body and 0.067 Sv (6.7 rem) to the skin. These results were confirmed by independent calculations. The dose to the whole body is well within the 0.05, Sv (5 rem) limit prescribed by 10 CFR 72.106(b). The calculated skin dose exceeds this limit by a small amount, although the conservative, generic nature of the assessment warrants that the DSC leakage event be further assessed for site-specific applications. It should also be noted that, as indicated in the SAR, no credible conditions have been identified which could breach the canister body or fail the double-seal welds at each end of the DSC. Thus, these dose results are only presented to bound the consequences that could conceivably result, and to evaluate compliance with the 10 CPR 72.106(b} requirement. Other accidents are assessed in Section 8.2 (e.g., floods, tomados, earthquakes, accidental cask drop, blockage of air inlets and outlets, etc.), but the SAR concludes that none of these other accidents represent credible sources of off-site dose consequences.
- 8. 3 Discussion and Conclusions 8.3.1 On-site Radiological Protection The shielding, confinement, and handling design features of the standardized NUHOMS design conform to the on-site radiological protection requirements of 10 CFR Part 20 and are considered acceptable for the set of conditions assumed in this review. Dose rates calculated by the vendor for different locations around the standardized NUHOMS system design are significantly higher than those determined for previous NUHOMS designs. This is specifically reflected in the dose rate limits delineated in Operating Limit 12.2. 7 of this SER.
8-5
Although independent review analyses and more exact dose calculation methods may result in lower dose rates, the relative dose rates for this design are still expected to be higher than comparably calculated dose rates for earlier NUHOMS designs. These relatively higher dose rates are not consistent with the objective of maintaining occupational exposures ALARA. Site-specific applications with this design should provide detailed procedures and plans to meet ALARA guidelines and 10 CFR Part 20 requirements with respect to the operation and . !!}aintenM.~ of thi~ sta!].~ NPIJQMS_systewJSFSI d~ign .. As.~~.~e>v~ *- . ____ . _ details of access control, surveillance, and other operational aspects affecting on-site exposure must be in compliance with existing licensee's radiation protection program. 8.3.2 Off-site Radiological Protection The shielding and confinement design features of the standardized NUHOMS system design conform to the off-site radiological protection requirements of 10 CFR Part 72 and 10 CFR Part 20 and are considered acceptable for the set of conditions assumed in this review. The use of high-integrity double-seal welds on the DSC ensures that, during normal operation, there are no effluents from the standardized NUHOMS system. Off-site dose is, therefore, due strictly to direct and scattered radiation, the intensity of which is a function of distance and direction from the array. Site-specific factors such as the number of HSMs in the storage array, the distance and direction of the nearest boundary of the controlled area, the contribution of reactor plant effluents to the off-site dose, and resultant collective off-site dose must be considered in the compliance evaluation for a proposed standardized NUHOMS system at a specific site. This evaluation must be performed by each user to assure compliance with 10 CFR 72.212 and 10 CFR 20.1207. This requirement is contained in the conditions for system use in Section 12.2.18 of this SER. 8-6
9.0 DECOMMISSIONING/DECONTAMINATION EVALUATION 9 .1 Design Description The standardiz.ed NUHOMS system design recognizes the need for decommissioning at the end of its useful life. External contamination of the DSC is limited by its containment features_and through_the contamination controLproced.ures11~ duriJig PSC fuel l~g. _In_ particular, contamination levels on the external surface of the DSC are minimized by the use of uncontaminated water in the DSC and cask/DSC annulus during fuel pool loading operations. This prevents contaminated fuel pool water from contacting the DSC exterior. Also, there is no credible chain of events which would result in widespread contamination outside of the DSC. The SAR also states that the DSC is designed to interface with a transportation system planned to transport canistered intact fuel assemblies (i.e., filled DSC's) to either a monitored retrievable storage facility (MRS) or a geologic repository. Until the transportation system is available, the DSCs are not approved at this time for transportation. If the fuel must be removed from the DSC at the reactor site before shipment, the DSC will likely require decontamination of the internal surfaces and disposal as low-level radioactive waste. Once the DSC's have been removed, only small amounts of residual contamination are expected to remain in the HSM passages, thereby facilitating easy decommissioning. 9.2 Design Evaluation Section 72.130 of 10 CFR Part 72 provides criteria for decommissioning. It provides that considerations for decommissioning should be included in the design of an ISFSI and that provisions be incorporated to (1) decontaminate structures and equipment; (2) minimize the quantity of waste and contaminated equipment; and (3) facilitate removal of radioactive waste and contaminated materials at the time of decommissioning. Although 10 CFR 7'),.130 does not provide specific criteria for acceptance, the ISFSI must be designed for decommissioning. Therefore, the standardiz.ed NUHOMS system design has been reviewed against good nuclear engineering practices which include (1) means to control the spread of contamination and (2) a design which facilitates decontamination. Section 72.30 of 10 CFR Part 72 defines the need for a decommissioning plan which includes financing. Such a plan, however, is not considered applicable to this review. The cost of decommissioning the ISFSI must be considered in the overall cost of decommissioning the reactor site. 10 CFR 72.236(i). requires that the cask be designed to facilitate decontamination to the extent practicable. The standardiz.ed NUHOMS design places heavy reliance on the prevention of contamination on the outer surface of the DSC. If these levels are kept low, very little contamination will exist on the inner surfaces of the HSMs, and ease of decommissioning will be facilitated. Section 10.3.14 of the SAR specifies a limiting condition for operation (LCO) for smearable 9-1
(non-fixed) surface contamination levels on the outer surface of the DSC. This specification states that smearable contamination levels shall be less than 36.5 Bq/100 cm2 (2200 dprn/100 cm2) (10-5 µCi/cm2) for beta-gamma emitters and 3.65 Bq/100 cm2 (220 dprn/100 cm2) (10-o
µCi/cm2) for alpha-emitting radionuclides. This specification corresponds to surface removable contamination limits in 10 CPR 71.87(i)(l).
-The surveillance requirement for this LCQ-is-to determine the contamination levels-of.the __ _ DSC by taJcing surface contamination surveys of the upper one foot of the DSC exterior while the DSC is in the transfer cask before making the first closure weld. This survey can be used as a representative sample of the DSC body. If the specified limits are exceeded, the annular space between the DSC and transfer cask will be flushed with demineralized water until the contamination levels are within these limits. By minimizing DSC contamination, the potential contamination of the internal surfaces of the HSM is kept to a minimum. The design of the standardized NUHOMS system is based on the intended eventual disposal of each DSC following fuel removal. However, it is also possible that the DSC shell/basket assembly could be reused. Such an alternative would be dependent on economic and regulatory conditions at the time of fuel removal. At this time, it is not known whether demolition and removal of the HSM can be performed by conventional methods. Uncertainty exists with respect to (1) the specific levels of contamination that might exist on the inner surfaces of the HSM and (2) contamination level criteria which will govern whether the HSMs can be disposed of as low-level radioactive waste or as ordinary rubble. The staff also notes that decommissioning of the DSC's, transfer cask, and other equipment are matters which will be properly addressed iJi site-specific decommissioning plans. 9.3 Conclusions The staff concludes that adequate attention has !:>een paid to decommissioning in the design of the standardized NUHOMS system considering the current state of knowledge. The staff also acknowledges that decommissioning considerations are sometimes in conflict with other requirements. The reinforced structure of the HSM, for example, will require considerable effort to demolish. Although it is not likely that significant contamination can spread beyond the DSC, demolition of the HSM may generate slightly contaminated dust. However, the staff concurs that primary concern in such cases rests with operational safety considerations, and ease of decommissioning is a secondary consideration. A specification is proposed by the vendor for maximum DSC exterior surface contamination in SAR Section 10.3.14. The primary reason for requiring a clean exterior surface of the DSC is to reduce the total amount of activity as a source of potential contamination for the TC and HSM interior surfaces. The SER includes this condition for system use in Section 12.2.12 of this report. 9-2
10.0 QUALITY ASSURANCE Chapter 11.0, 1 "Quality Assurance," of Revision 2 of the Pacific Nuclear Fuel Services Group Certification Safety Analysis Report (SAR) for a general license in accordance with Subpart K of 10 CFR Part 72 describes the PNFS quality assurance program. The PNFS quality assurance program is applied to structures, systems, and components of the NUHOMS independent spent fuel storage system -important -to safety. Chapter 11. 0 addresses each of the 18 quality assurance criteria of 10 CFR Part 72, Subpart G, "Quality Assurance," and it includes the commitment that PNFS will implement the quality assurance program controls described in Revision 1 of the VECTRA Quality Assurance Manual dated July 22, 1994. This manual has been reviewed and accepted by the NRC. Chapter 11.0 of the SAR describes the graded quality assurance program that is applied by PNFS to the structures, systems, and components of the NUHOMS spent fuel storage system based on that structure, system, or component's importance to safety. Chapter 11 defines three quality categories (or levels of quality/quality assurance) _for items important to safety, and there are some items that are not important to- safety. Chapter 11 of the SAR describes the differences between the quality assurance program for each category. It also lists the quality category of each structure, system, and component of the NUHOMS spent fuel storage system. The staff has reviewed PNFS's quality assurance program description given and referenced in Chapter 11. 0 of the SAR. 2 The staff finds that the PNFS commitments meet the requirements of Subpart G of 10 CFR Part 72 and are, therefore, acceptable for the issuance of a general Certificate of Compliance in accordance with Subpart L of 10 CFR Part 72. SAR pages 11.1-1 through 11. 3-5 identified as NUH-003, Revision 2, November 5, 1993. 2 The acceptance criteria for quality assurance for independent spent fuel storage installations, based on Subpart G of 10 CFR Part 72, is given in the Fuel Cycle Safety Branch (Currently the Storage and Transport Systems Branch) Technical Position of June 20, 1986. 10-1
11.0 OPERATIONS, MAINTENANCE, TESTING, AND RECORDS 11.1 Operations 10 CFR 72.234(t) requires as a condition of approval of the Certificate of Compliance that: "the cask vendor [PNFS] shall ensure that written procedures and appropriate tests are established befgre_ ~ of Qt~~- A _copy_of these procedures and tests must be provided to each cask user.* Regulatory Guide 3.48, Section 9 (Reference 5) describes-the -- information to be incorporated in operating procedures for loading, unloading, and preparation of the cask. For the Certificate of Compliance for the standardized NUHOMS system, the term "cask" in 10 CFR Part TI, Subpart L, and in Regulatory Guide 3.48 is applied to the full NUHOMS System. Procedures described in the SAR were reviewed and evaluated as part of the staff preparation of this SER. Procedures for loading the DSC are in SAR paragraphs 5.1.1. 2 through 5 .1.1. 6 and are summarized graphically in SAR Figure 5.1-1. These include descriptions of recommended procedures for loading, use of fluids to fill cavities, removal of moisture, sealing, on-site management, and placing and sealing in storage positions. Procedures for unloading the DSC are in SAR paragraphs 5.1.1.8 and 5.1.1.9 and are graphically summarized in SAR Figure 5.1-2. These include descriptions, tests, special preparations for unloading, unsealing, removal of DSC and on-site transfer, opening, removal of IFAs and cask decontamination. Procedures for preparation of the TC and DSC for use &re in SAR paragraph 5 .1.1.1 and are part of the graphical summariz.ation in SAR Figure 5 .1-1. These include descriptions of inspections, tests, and special preparations of the TC and DSC necessary to ensure that they are properly loaded, closed, decontaminated, and transferred. Staff review of the procedures included in the SAR indicates that they are acceptable and are in full compliance with the requirements of 10 CFR TI.234(t) for written procedures and with the guidance of Regulatory Guide 3.48, Section 9, for descriptions of operating procedures. These descriptions provide adequate bases for users to develop more detailed written procedures to follow during cask operations. 11.2 Maintenance 10 CFR TI.234 (a) requires as conditions of approval of the Certificate of Compliance that maintenance must comply with the requirements of 10 CFR 72.236 which require that the "cask must be designed to store spent fuel safely for a minimum of 20 years and permit maintenance as required.* Regulatory Guide 3.48, Section 9.4, provides guidance on description of the maintenance program. The staff based evaluation of the SAR descriptions of maintenance on the 10 CFR Part TI, Subpart L, requirements and Regulatory Guide 3.48 guidance. 11-1
The SAR describes maintenance for the standardized NUHOMS system in paragraph 5.1.3.5 which states that, as the system is totally pruwve, it does not require maintenance. To ensure that the ventilation airflow is not interrupted, the RSM is to be periodically inspected to ensure that no debris is in the airflow inlet or outflow openings. SAR Section 5.1.1. 7 describes these monitoring operations. The TC is expected to be maintained and pi:epared in accordance with the procedure for each DSC IFA loading, transfer, loading into the HSMcycle, and for.the unloading-process. A - single TC may be used at a site. SAR Section 4.5 describes recommended procedures for inspection, maintenance, and repair of the TC. Staff review of the provisions for and descriptions of maintenance included in the SAR indicates that they are acceptable and are in full compliance with the requirements of 10 CFR Part n, Subpart L, and the guidance of Regulatory Guide 3.48. 11.3 Testing 10 CFR 'n..234(a) requires as conditions of approval of the Certification of Compliance that testing must comply with the requirements of 10 CFR 'n.236. 10 CFR 72.234(t) requires that the cask vendor ensure appropriate tests are established before use of the "casks II and that a copy of the tests must be provided to each user. 10 CFR 72.236(j) requires that the
- cask must be inspected to ascertain that there are no cracks, pinholes, uncontrolled voids, and other defects that could significantly reduce its confinement effectiveness. 11 10 CFR
'n.236(1) requires that "the cask and its systems important to safety must be evaluated by appropriate test or by other means acceptable to the Commission, to demonstrate that they will reasonably maintain confinement of radioactive material under normal, off-normal, and credible accident conditions.* Regulatory Guide 3.48, Section 9, provides guidance on describing tests in the SAR. Descriptions of and requirements for testing in conjunction with fabrication of the standardized NUHOMS system components are included in the SAR. Descriptions of tests and inspections associated with preparation for loading and loading operations are included in SAR Sections 5.1.1.1 through 5.1.1.6. Inspections in conjunction with downloading operations are described in SAR Section 5.1. l.9. Instruments to be used dunng loading operations and their functions are listed in SAR Table 5 .1-1. No instruments or control systems are used during the storage cycle due to the passive nature of the standardized NUHOMS system (SAR Section 5.4). Recommended pre-operational testing is described in SAR Section 9.2. This includes the test program description and discussion. Recommended testing is also included in SAR Section 10, Operating Controls and Limits. These include:
- DSC pressure during drying and backfill (SAR Sections 10.3.2 and 10.3.3).
- Tests of DSC inner seal and closure welds (SAR Sections 10.3.4 and 10.3.5).
11-2
- RSM dose rates with DSC in storage (SAR Section 10.3.7).
- HSM temperature rise with DSC in place ( SAR Section 10.3.8).
- TC dose rates (SAR Section 10.3.12).
- DSC surface contamination (SAR Section 10.3.14).
- Ambient temperatures before TC use for DSC transport (SAR Section 10.3.15).
_Staff review of the pI'9visiog_s for and descr_ip_tion~ of testing included in the SAR indicates that they are acceptable and are in full compliance with the requirements-of 10 CPR Part -72~ Subpart L, and the guidance of Regulatory Guide 3.48. 11.4 Records 10 CFR 72.234(d) requires that the cask vendor ensure a record is established and maintained for each cask fabricated under the NRC Certificate of Compliance and describe the information to be included on the record. The SAR does not explicitly identify the information record specified in 10 CFR 72.234(d). The only statement is that, "The ISFSI records should be maintained by the licensee in accordance with the requirements in 10 CFR Part 72 and with the existing plant records retention practices." As the 10 CFR 72.234(d) requirement is placed on the vendor, regardless of the location where the records are maintained, the staff considers that the required assurance is not met in the SAR. The required record does not require any data from the "cask" user other than name and address. The remaining data relates to cask fabrication and the Certificate of Compliance. Specifically, the data required to be recorded and maintained for each "cask" by the vendor VECTRA are [10 CFR 72.234(d)(2)]:
"(i) The NRC Certificate of Compliance number; (ii) The cask model number; [The model number should be marked on each HSM, DSC and TC.]
(iii) The cask identification number; [A unique identification number should be marked on each HSM, DSC and TC.] (iv) Dat~ fabrication was started; (v) Date fabrication was completed; (vi) Certification that the cask was designed, fabricated, tested, and repaired in accordance with a quality assurance program accepted by NRC; (vii) Certification that inspections required by paragraph 72.2360) were performed and found satisfactory; and (viii) The name and address of the cask user." The data marked on the DSC and TC are among those required by 10 CFR 72.236(k) which requires that the data be conspicuously and durably marked and also include the empty weight. The staff considers that the HSM and its included DSC support assembly are important to safety; therefore, maintaining a record and marking the individual HSM would be consistent with the intent of Subpart L. 11-3
12.0 CONDffiONS FOR SYSTEM USE This section presents the conditions which a JX,ltential user (general licensee) of the standardized NUHOMS system must comply with, in order to use the system under a general license that is issued according to the provisions of 10 CPR 72.210 and 10 CFR 72.212 .
.These @nditions hav~ eiAJ.er_been proposed by the system vendor, imposed by the NRC staff as a result of the review of the SAR, or are part of the regulatory. requirements expiesseo fa-10 CFR 72.212.
12.1 General Requirements and Conditions 12.1.1 Regulatory Requirements for a General License Subpart K of 10 CFR Part 72 contains conditions for using the general license to store spent fuel at an independent spent fuel storage installation at power reactor sites authorized to possess and operate nuclear power reactors under 10 CPR Part 50. Technical regulatory requirements for the licensee (user of the standardized NUHOMS system) are contained in 10 CFR 72.212(b). 10 CFR 72.212(b)(2) requires that the licensee perform written evaluations, before use, that establish that: (1) conditions set forth in the Certificate of Compliance have been met; (2) cask storage pads and areas have been designed to adequately support the static load of the stored casks; and (3) the requirements of 10 CFR 72.104 "Criteria for radioactive materials in effluent and direct radiation from an ISFSI or MRS," have been met. 10 CFR 72.212(b)(3) requires that the licensee review the SAR and the associated SER, before use of the general license, to determine whether or not the reactor site parameters (including earthquake intensity and tornado missiles), are encompassed by the cask design bases - considered in these reports. 10 CFR 72.212(b)(4) provides that, as a holder of a Part 50 license, the user, before use of the general 10 CFR Part 72 license, must determine whether activities related to storage of spent fuel involve any unreviewed safety issues, or changes in technical specifications as provided under 10 CFR 50.59. Under 10 CFR 72.212(b)(5), the general license holder shall also protect the spent fuel against design basis threats and radiological sabotage pursuant to 10 CPR 73.55. Other general license requirements dealing with review of reactor emergency plans, quality assurance program, training, and radiation protection program must also be satisfied pursuant to 10 CFR 72.212(b)(6). 10 CFR 72.212(b)(7), (8), (9) and (10) describe record and procedural requirements for the general license holder. Without limiting the requirement identified above, site-specific parameters and analyses, identified in the SER, that will need verification by the system user, are as a minimum, . as follows: 12-1
\J
- 1. The temperature of 21 °C (70°F) as the maximum average yearly temperature with solar incidence. The average daily ambient temperature shall be 37.8°C (l00°F) or less (Reference SER Section 2.4.1).
- 2. The temperature extremes of 52°C (125°F) with incident solar radiation and
-40°C (-40°F) with no solar incidence (Reference SER Section 2.4.1) for stt?rage of the DSC inside t:!1:e HSM~ _
- 3. The horizontal and vertical seismic acceleration levels of 0.25g and 0.17g, respectively (Reference SER Table 2-4).
- 4. The analyzed flood condition of 4.6 mis (15 fps) water velocity and a height of 15.2 m (50 feet) of water (full submergence of the loaded HSM DSC)
(Reference SER Table 2-4).
- 5. The potential for fire and explosion should be addressed, based on site-specific considerations (See SER Table 2-4 and related SER discussion).
- 6. The HSM foundation design criteria are not included in the SAR. Therefore, the nominal SAR design or an alternative should be verified for individual sites in accordance with 10 CPR 72.212(b)(2)(ii). Also, in accordance with 10 CFR 72.212(b)(3), the foundation design should be evaluated against actual site parameters to determine whether its failure would cause the Standardized NUHOMS systems to exceed the design basis accident conditions.
- 7. The potential for lightning damage to any electrical system associated with the standardized NUHOMS system (e.g., thermal performance monitoring) should be addressed, based on site-specific considerations (See SER Table 2.4 and related SER discussion).
- 8. Any other site parameters or consideration that could decrease the effectiveness of csk systems important to safety.
In accordance with 10 CFR 72.212(b), a record of the written evaluations must be retained by the licensee until spent fuel is no longer stored under the general license issued under 10 CPR 72.210. 12.1.2 Operating Procedures Written operating procedures shall be prepared for cask handling, loading, movement, surveillance, and maintenance. The operating procedures suggested generically in the SAR were considered appropriate as discussed in Section 11. 0 of the SER and should provide the basis for the user's written operating procedure. The following additional procedure requested by NRC staff in Section 11.1 should be part of the user operating procedures: 12-2
If fuel needs to be removed from the DSC, either at the end of service life or for inspection after an accident, precautions must be taken against the potential for the presence of damaged or oxidized fuel and to prevent radiological exposure to personnel during this operation. This can be achieved with this design by the use of the purge and fill valves which permit a determination of the atmosphere within the DSC before the removal of the inner top cover plate and shield plugs, prior to filling the DSC cavity with borated water (see SAR paragraph 5.1.1.9). If the atmosphere within the tisc is helium, then operations should proceed- nonruilly with fuel removal -- either via the transfer cask or in the pool. However, if air is present within the DSC, then appropriate filters should be in place to preclude the uncontrolled release of any potential airborne radioactive particulate from the DSC via the purge-fill valves. This will protect both personnel and the operations area from potential contamination. For the accident case, personnel protection in the form of respirators or supplied air should be considered in accordance with the licensee's Radiation Protection Program. 12.1.3 Quality Assurance Activities at the ISFSI shall be conducted in accordance with a Commission-approved quality assurance program which satisfies the applicable requirements of 10 CFR Part 50, Appendix B and which is established, maintained, and executed with regard to the ISFSI. 12.1.4 Heavy Loads Requirements Lifts of the DSC in the TC must be made within the existing heavy loads requirements and procedures of the licensed nuclear power plant. The TC design has been reviewed under 10 CFR Part 72 and found to meet NUREG-0612 (Reference 14) and ANSI N14.6 (Reference 8). However, an additional safety review (under 10 CFR 50.59) is required to show operational compliance with NUREG-0612 and/or existing plant-specific heavy loads requirements. 12.1.5 Training Module A training module shall be developed for the existing licensee's training program establishing an ISFSI training and certification program. This module shall include the following:
- 1. Standardized NUHOMS System Design (overview);
- 2. ISFSI Facility Design (overview);
- 3. Certificate of Compliance conditions (overview);
- 4. Fuel Loading, Transfer Cask Handling, DSC Transfer Procedures; and
- 5. Off-Normal Event Procedures.
12-3
12.1. 6 Pre-Operational Testing and Training Exercise A dry run of the DSC loading, TC handling and DSC insertion into the HSM shall be held. This dry run shall include, but not be limited to, the following:
- 1. Functional testing of the TC with lifting yokes to ensure that the TC can be
_-~~ly_transported_over th~ ~tire route reqajred f()r fuel loading, washdown pit and trailer loading. * - * --- -- * *
- 2. DSC loading into the TC to verify fit and TC/DSC annulus seal.
- 3. Testing of TC on transport trailer and transported to ISFSI along a predetermined route and aligned with an HSM.
- 4. Testing of transfer trailer alignment and docking equipment. Testing of hydraulic ram to insert a DSC loaded with test weights into an HSM and then retrieve it.
- 5. Loading a mock-up fuel assembly into the DSC.
- 6. DSC sealing, vacuum drying, and cover gas backfilling operations (using a mock-up DSC).
- 7. Opening a DSC (using a mock-up DSC).
- 8. Returning the DSC and TC to the spent fuel pool.
12.1. 7 Special Requirements for First System in Place The heat transfer characteristics of the cask system will be confirmed by temperature measurements of the first DSC placed in service. The first DSC shall be loaded with 24 fuel assemblies, constituting a source of approximately 24 kW. The DSC shall be loaded into the HSM and the thermal performance will be assessed by measuring the air inlet and outlet temperatures for normal airflow. Details for obtaining the m~urements are provided in Section 12.2.8, under *surveillance." A letter report summarizing the results of the measurements shall be submitted to the NRC for evaluation and assessment of the heat removal characteristics of the thermal design within 30 days of placing the DSC in service, in accordance with 10 CFR 72.4.
- Should the first user of the system not have fuel capable of producing a 24 kW heat load, or be limited to a lesser heat load, as in the case of BWR fuel, the user may use a lesser load for the process, provided that a calculation of the temperature difference between the inlet and outlet temperatures is performed, using the same methodology and inputs documented in 12-4
the SAR, with lesser load as the only exception. The calculation and the measured temperature data shall be reported to the NRC in accordance with 10 CPR 72.4. The calculation and comparison need not be reported to the NRC for DSCs that are subsequently loaded with lesser loads than the initial case. However, for the first or any other user, the process needs to be performed and reported for any higher heat sources, up to 24 kW for PWR fuel and 19 kW for BWR fuel, which is the maximum allowed under the Certificate of Compliance. The NRC will also accept ~e µ~- of_ artificial thermal loads other than spent - - fuel~ fo satisfy the above requirement. 12.1.8 Surveillance Requirements Applicability
\
The specified frequency for each Surveillance Requirement is met if the surveillance is performed within 1.25 times the interval specified in the frequency, as measured from the previous performance. For frequencies specified as "once,* the above interval extension does not apply. If a required action requires performance of a surveillance or its completion time requires period performance of "once per *.. ,* the above frequency extension applies to the repetitive portion, but not to the initial portion of the completion time. Exceptions to these requirements are stated in the individual specifications. 12.2 Technical Specifications, Functional and Operating Limits 12.2.1 Fuel Specification Limit/Specification: The characteristics of the spent fuel which is allowed to be stored in the standardized NUHOMS system are limited by those included in Tables 12- la and 12- lb. Applicability: The specification is applicable to all fuel to be stored in the standardized NUHOMS system. Objective: The specification is prepared to ensure that the peak fuel rod temperatures, maximum surface doses, and nucle.ar criticality effective neutron multiplication factor are below the design values. Furthermore, the fuel weight and type ensures that structural conditions in the SAR bound those of the actual fuel being stored. Action: Each spent fuel assembly to be loaded into a DSC shall have the parameters listed in Tables 12-la and 12-lb verified and documented. Fuel not meeting this specification shall not be stored in the standardized NUHOMS system. 12-5
Surveillance: Immediately, before insertion of a spent fuel assembly into an DSC, the identity of each fuel assembly shall be independently verified and documented. Bases: The specification is based on consideration of the design basis _parameters_ included in, Jh~- SAR and limitations imposed as a result of the staff review. Such parameters stem from the-type of fuel analyzed~ structural limitations, criteria for criticality safety, criteria for heat removal, and criteria for radiological protection. The standardized NUHOMS system is designed for dry, horizontal storage of irradiated light water reactor (LWR) fuel. The principal design parameters of the fuel to be stored can accommodate standard PWR fuel designs manufactured by Babcock and Wtlcox, Combustion Engineering, and Westinghouse, and standard BWR fuel manufactured by General Electric and is limited for use to these standard designs. The analyses presented in the SAR are based on non-consolidated, zircaloy-clad fuel with no known or suspected gross cladding breaches (See Tables 12-la and lb.) The physical parameters that define the mechanical and structural design of the RSM and the DSC are the fuel assembly dimensions and weight. The calculated stresses given in this SER are based on the physical parameters given in Table 12-la,b and represent the upper bound. The design basis for nuclear criticality safety is based on the standard Babcock & Wilcox 15x15/208 pin fuel assemblies with initial enrichments up to 4.0 wt.% U-235, and General Electric 7x7 fuel assemblies with initial enrichments up to 4.0 wt.% U-235, for the standardized NUHOMS-24P and NUHOMS-52B designs, respectively. The HSM is designed to permit storage of irradiated fuel such that the irradiated fuel reactivity is less than or equal to 1.45 wt.% U-235 equivalent unirradiated fuel for the NUHOMS-24P design, and less than or equal to 4.0 wt.% U-235 initial enrichment fuel for the NUHOMS-52B design. The thermal design criterion of the fuel to be stored is that the maximum heat generation rate per assembly be such that the fuel cladding temperature is maintained within established limits during normal and off-normal conditions. Fuel cladding temperature limits were established by the applicant based on methodology in PNL-6189 and PNL-4835 (References 47, 48). Based on this methodology, the staff has accepted that a maximum heat generation rate of 1 kW per assembly is a bounding value for the PWR fuel to be stored, and that 0.37 kW per assembly is a bounding value for the BWR fuel to be stored. 12-6
The radiological design criterion is that the gamma and neutron source strength of the irradiated fuel assemblies must be bounded by values of the neutron and gamma ray source strengths used by the vendor in the shielding analysis. The design basis source strengths were derived from a bumup analysis for (1) PWR fuel with 4.0 weight percent U-235 initial enrichment, irradiated to a maximum of 40,000 MWD/MTU, and a post irradiation tune of five years; and (2) BWR fuel with 4.0 weight percent U-235 initial enrichment, , irradiated to a maximum of 35,000 MWD/MTU, and a post irradiation time of 5 years. 12-7
Table 12-la PWR Fuel Specificatiom of Fuel to be Stored in the Standardized NUHOMS-24P DSCCl> Title or Parameter Specifications Fuel Only intact, unconsolidated PWR fuel assemblies with the following requirements Physical Parameters Assembly Length See SAR Chapter 3 Nominal Cross-Sectional Envelope See SAR Chapter 3 Maximum Assembly Weight See SAR Chapter J('l) No. of Assemblies per DSC s 24 intact assemblies Fuel Cladding Zircaloy-clad fuel with no known or suspected gross cladding breaches Thermal Characteristics Decay Heat Power per Fuel Assembly S 1.0 kW (this value is maximum for any given assembly, and may not be averaged for all 24 assemblies) Radiological Characteristics Bumup ~40,000 MWD/MTU Post Irradiation Time ~5 years Maximum Initial Enrichment S4.0 wt. % U-235 Maximum Initial Uranium Content ~472 kg/assembly Maximum Initial Equivalent ~ 1.45 wt. % U-235(3) Enrichment Neutron Source Per Assembly ~2.23E8 n/sec with spectrum bounded by that in Chapter 7 of SAR Gamma Source Per Assembly ~7.45E15 photon/sec with spectrum bounded by that in Chapter 7 of SAR (1) The limiting fuel specifications listed above must be met by every individual fuel assembly to be stored in the standardized NUHOMS-24P system. Any deviation constitutes an Unanalyzed Condition and Violation of the Certificate of Compliance. (2) Design valid for fuel weights up to 762.8 kg (1,682 lb). (3) Determined by the PWR fuel criticality acceptance curve shown in Figure 12.1. 12-8
50 . . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - UNACCEPTABLE UNACCEPTABLE 0 ................- - - , . - - ~ -......- - , - -....-~....--...-------,.-.....,i----.,.....--, 1.60 2.00 2.40 2.10 3.20 3.eo 4 00 INmAL ENRICHMENT (w.10 235 U) Figure l 2 .1 PWR Fuel Criticalily Accepcancc Cwve 12-9
Table 12-lb BWR Fuel Specif"ications of Fuel to be Stored in the Standardized NUHOMS-52B nsc<l) Title or Parameter Specifications Fuel Only intact, unconsolidated BWR fuel assemblies with the following requirements Physical Parameters Assembly Length See SAR Chapter 3 Nominal Cross-Sectional Envelope See SAR Chapter 3 Maximum Assembly Weight See SAR Chapter 3 (w/fuel channels) No. of Assemblies per DSC S 52 intact channeled assemblies Fuel Cladding Zircaloy-clad fuel with no known or suspected gross cladding breaches Thermal Characteristics Decay Heat Power per Fuel Assembly s0.37 kW (this value is maximum for any given assembly, and may not be averaged for all 52 assemblies) Radiological Characteristics Bumup S35,000 MWD/MTU Post Irradiation Time ~5 years Maximum Initial Enrichment S4.0 wt. % U-235 (DSC with 0.75% borated neutron absorber plates) Maximum Initial Uranium Content s 198 kg/assembly Neutron Source Per Assembly s l.01E8 n/sec with spectrum bounded by that in Chapter 7 of SAR Gamma Source Per Assembly S2.63E15 photon/sec with spectrum bounded by that in Chapter 7 of SAR (1) The limiting fuel specifications listed above must be met by every individual fuel assembly to be stored in the standardized NUHOMS-52B system. Any deviation constitutes an Unanalyzed Condition and Violation of the Certificate of Compliance. 12-10
12.2.2 DSC Vacuum Pressure During Drying Limit/Specification: Vacuum Pressure: ~ 0.4 k:Pa (3 mm Hg) Time at Pressure: c?: 30 minutes following stepped evacuation Number of Pump-Downs: 2 Applicability: This is applicable to all DSCs. Objective: To ensure a minimum water content. Action: If the required vacuum pressure cannot be obtained:
- 1. Con:finn that the vacuum drying system is properly installed.
- 2. Check and repair, or replace, the vacuum pump.
- 3. Check and repair the system as necessary.
- 4. Check and repair the se.a1 weld between the inner top cover plate and the DSC shell.
Surveillance: No maintenance or tests are required during normal storage. Surveillance of the vacuum gauge is required during the vacuum drying operation. Bases: A stable vacuum pressure of 0.4 kPa ( ~ 3 mm Hg) further ensures that all liquid water has evaporated in the DSC cavity, and that the resulting inventory of oxidizing gases in the DSC is well below the 0.25 volume%. 12-11
12.2.3 DSC Helium Backfill Pressure Limit/Specifications: Helium 17.25 kPag (2.5 psig) +/- 17.25 kPag (2.5 psig) backfill pressure (stable for 30 minutes after filling). Applicability: This specification is applicable to all DSCs. Objective: To ensure that: (1) the atmosphere surrounding the irradiated fuel is a non-oxidizing inert gas; (2) the atmosphere is favorable for the transfer of decay heat. Action: If the required pressure cannot be obtained:
- 1. Confirm that the vacuum drying system and helium source are properly installed.
- 2. Check and repair or replace the pressure gauge.
- 3. Check and repair or replace the vacuum drying system.
- 4. Check and repair or replace the helium source.
- 5. Check and repair the seal weld on DSC top shield plug.
If pressure exceeds the criterion, release a sufficient quantity of helium to lower the DSC cavity pressure. Surveillance: No maintenance or tests are required during the normal storage. Surveillance of the pressure gauge is required during the helium backfilling operation. Bases: The value of 17.25 kPag (2.5 psig) was selected to ensure that the pressure within the DSC is within the design limits during any expected normal and off-normal operating conditions. 12-12
12.2.4 DSC Helium Leak Rate of Inner Seal Weld Limit/Specification:
~ 1.0x10-2 k:Pa
- cm3/s (1.0 x 104 atm
- cubic centimeters per second) (atm
- cm3/s) at the highest DSC limiting pressure.
Applicability: This specification is applicable to the inner top cover plate seal weld of all DSCs. Objective: 1. To limit the total -raofoactive -gases normally released by each ___ ___ _ canister to negligible levels. Should fission gases escape the fuel cladding, they will remain confined by the DSC confinement boundary.
- 2. To retain helium cover gases within the DSC and prevent oxygen from entering the DSC. The helium improves the heat dissipation characteristics of the DSC and prevents any oxidation of fuel cladding.
Action: If the leak rate test of the inner seal weld exceeds l.Ox10-2 kPa
- cm3/s (1.0xlo-4 atm
- cm3/s):
- 1. Check and repair the DSC drain and fill port fittings for leaks.
- 2. Check and repair the inner seal weld.
- 3. Check and repair the inner top cover plate for any surface indications resulting in leakage.
Surveillance: After the welding operation has been completed, perform a leak test with a helium leak detection device. Bases: If the DSC leaked at the maximum acceptable rate of 1.0xl0-2 kPa
- cm3/s (1.0xlo-4 atm
- cm3/s) for a period of 20 years, about 63,100 cc of helium would escape from the DSC. This is about 1 % of the 6.3 x 1Q6 cm3 of helium initially introduced in the DSC.
This amount of leakage would have a negligible effect on the inert environment of the DSC cavity.
Reference:
American National Standards Institute, ANSI Nl4.5-1987, "For Radioactive Materials-Leakage Tests on Packages for Shipment" (Appendix B3). 12-13
12.2.5 DSC Dye Penetrant Test of Closure Welds Limit/Specification: All DSC closure welds except those subjected to full volumetric inspection shall be dye penetrant tested in accordance with the requirements of the ASME Boiler and Pressure Vessel Code Section ill, Division 1, Article NB-5000 (Reference 8.3 of SAR). 'The liquid penetrant test acceptance standards shall be those described in Subsection NB-5350 of the Code. Applicability: This is applicable to all DSCs. 'The welds include inner and outer top and bottom covers, and vent and syphon port covers. Objective: To ensure that the DSC is adequately sealed in a redundant manner and leak tight. Action: If the liquid penetrant test indicates that the weld is unacceptable:
- 1. The weld shall be repaired in accordance with approved ASMB procedures.
- 2. 'The new weld shall be re-examined in accordance with this specification.
Surveillance: During DSC closure operations. No additional surveillance is required for this operation. Bases: Article NB--5000 Examination, ASME Boiler and Pressure Vessel Code, Section m, Division 1, Sub-Section NB (Reference 8.3 of SAR). 12-14
12.2.6 DSC Top End Dose Rates Limit/Specification: Dose rates at the following locations shall be limited to levels which are less than or equal to:
- a. 2 mSv/hr (200 mrem/hr) at top shield plug surface at centerline with water in cavity.
- b. 4 mSv/hr (400 mrein!hr) aftop-oover-plate surface atcenterline without water in cavity.
Applicability: This specification is applicable to all DSCs. Objective: The dose rate is limited to this value to ensure that the DSC has not been inadvertently loaded with fuel not meeting the specifications in Section 12.2.1 of the SER and to maintain dose rates as low as is reasonably achievable during DSC closure operations. Action: a. If specified dose rates are exceeded, the following actions should be taken:
- 1. Confirm that the spent fuel assemblies placed in DSC conform to the fuel specifications of Section 12. 2.1.
- 2. Visually inspect placement of top shield plug. Re-install or adjust position of top shield plug.
- 3. Install additional temporary shielding.
- b. Submit a letter report to the NRC within 30 days summarizing the action taken and the results of the surveillance, investigation and findings. The report must be submitted using instructions in 10 CFR 72.4 with a copy sent to the administrator of the appropriate NRC regional office.
Surveillance: Dose rates shall be measured before seal welding the inner top cover plate to the DSC shell and welding the outer top cover plate to the DSC shell. Basis: The basis for this limit is the shielding analysis presented in Section 7.0 of the SAR. 12-15
12.2. 7 HSM Dose Rates Limit/Specification: Dose rates at the following locations shall be limited to levels which are less than or equal to:
- a. 4 mSv/hr (400 mrem/hr) at 1 m (3 feet) from the HSM surface.
- b. Outside of HSM door on centerline of DSC 1 mSv/hr (100 mrem/hr).
- c. End Shield wall exterior 0.2 mSv/hr (20 mrem/hr).
Applicability: This specification is applicable to all HSMs which corttain--a loaded DSC. Objective: The dose rate is limited to this value to ensure that the cask (DSC) has not been inadvertently loaded with fuel not meeting the specifications in Section 12.2.1 of the SER and to maintain dose rates as low as is reasonably achievable at locations on the HSMs where surveillance is performed, and to reduce off-site exposures during storage. Action: a. If specified dose rates are exceeded, the following actions should be taken:
- 1. Ensure that the DSC is properly positioned on the support rails ..
- 2. Ensure proper installation of the HSM door.
- 3. Ensure that the required module spacing is maintained.
- 4. Confirm that the spent fuel assemblies contained in the DSC conform to the specifications of Section 12.2.1.
- 5. Install temporary or permanent shielding to mitigate the dose to acceptable levels in accordance with 10 CPR Part 20, 10 CPR 72.104(a), and AI.ARA.
- b. Submit a letter report to the NRC within 30 days summarizing the action taken and the results of the surveillance, investagion and findings. The report must be submitted using instructions in 10 CPR n.4 with a copy sent to the administrator of the appropriate NRC regional office.
Surveillance: The HSM and ISFSI shall be checked to verify that this specificantion has been met after the DSC is placed into storage and the HSM door is closed. Basis: The basis for this limit is the shielding analysis presented in Section 7. 0 of the SAR. The specified dose rates provide as-low-as-is-reasonably-achievable on-site and off-site doses in accordance with 10 CPR Part 20 and 10 CPR 72.104(a). 12-16
12.2.8 HSM Maximum Air Exit Temperature Limit/Specification: Following initial DSC transfer to the HSM or the occurrence of accident conditions, the equilibrium air temperature difference between ambient temperature and the vent outlet temperature shall not exceed 37.8°C (l00°F) for ~5 year cooled fuel, when fully loaded with 24 kW heat. Applicability: This specification is applicable to all HSMs stored in the ISFSI. If a DSC is placed in the HSM with a heat load less than 24 kW, the
- limiting differenctfbetween outlet and ambiennemperatures shall be determined by a calculation performed by the user using the same methodology and inputs documents in the SAR and SER.
Objective: The objective of this limit is to ensure that the temperature of the fuel cladding and the HSM concrete do not exceed the temperatures calculated in Section 8 of the SAR. That section shows that if the air outlet temperature difference is less than or equal to 37.8°C (100°F) (with a thermal heat load of 24 kW), the fuel cladding and concrete will be below the respective temperature limits for nonnal long-term operation. Action: If the temperature rise is greater than that specified, then the air inlets and exits should be checked for blockage. If the blockage is cleared and the temperature is still greater than that specified, the DSC and HSM cavity may be inspected using video equipment or other suitable means. If environmental factors can be ruled out as the cause of excessive temperatures, then the fuel bundles are producing heat at a rate higher than the upper limit specified in Section 3 of the SAR and will require additional measurements and analysis to assess the actual performance of the system. If excessive temperatures cause the system to perform in an unacceptable manner and/or the temperatures cannot be controlled to acceptable limits, then the cask shall be unloaded. Surveillance: The te~perature rise shall be measured and recorded cuuiy following DSC insertion until equilibrium temperature is reached, 24 hours after insertion, and again on a daily basis after insertion into the HSM or following the occurrence of accident conditions. If the temperature rise is within the specifications or the calculated value for a heat load less than 24 kW, then the HSM and DSC are performing as designed to meet this specification and no further maximum air exit temperature measurements are required. Air temperatures must be measured in such a manner as to obtain representative values of inlet and outlet air temperatures. Basis: The specified temperature rise is selected to ensure the fuel clad and concrete temperatures are maintained at or below acceptable long-term storage limits. 12-17
12.2.9 Transfer Cask Alignment with RSM Limit/Specification: The cask must be aligned with respect to the HSM so that the longitudinal centerline of the DSC in the transfer cask is within
+/-0.3 cm (+/-118 inch) of its true position when the cask is docked with the HSM front access opening.
Applicability: Tpis _specification is applicable during the insertion and retrieval of all DSCs. - - - Objective: To ensure smooth transfer of the DSC from the transfer cask to HSM and back. Action: If the alignment tolerance is exceeded, the following actions should* be taken:
- a. Confirm that the transfer system is properly configured.
- b. Check and repair the alignment equipment.
- c. Confirm the locations of the alignment targets on the transfer cask and HSM.
Surveillance: Before initiating DSC insertion or retrieval, confirm the alignment. Observe the transfer system during DSC insertion or retrieval to ensure that motion or excessive vibration does not occur. Basis: The basis for the true position alignment tolerance is the clearance between the DSC shell, the transfer cask cavity, the RSM access opening, and the DSC support rails inside the HSM. 12-18
12.2.10 DSC Handling Height Outside the Spent Fuel Pool Building Limit/Specification: 1. The loaded TC/DSC shall not be handled at a height greater than 203 cm (80 inches) outside the spent fuel pool building.
- 2. In the event of a drop of a: loaded TC/DSC from a height greater than 38 cm (15 inches) (a) fuel in the DSC shall be returned to the reactor spent fuel pool; (b) the DSC shall be removed from service and evaluated for further use; and (c) the TC shall be inspected for damage and evaluated for further use.
Applicability: The specification applies to handling the TC, loaded with the DSC, on route to, and at, the storage pad. Objective: 1. To preclude a loaded TC/DSC drop from a height greater than 203 cm (80 inches).
- 2. To maintain spent fuel integrity, according to the spent fuel specification for storage, continued confinement integrity, and DSC functional capability, after a tip-over or drop of a loaded DSC from a height greater than 38 cm (15 inches).
Surveillance: In the event of a loaded TC/DSC drop accident, the system will be returned to the reactor fuel handling building, where, after the fuel has been returned to the spent fuel pool, the DSC and TC will be inspected and evaluated for future use. Basis: The NRC evaluation of the TC/DSC drop analysis concurred that drops up to 203 cm (80 inches), of the DSC inside the TC, can be sustained without breaching the confinement boundary, preventing removal of spent fuel assemblies, or causing a criticality accident. This specification ensures that handling height limits will not be exceeded in transit to, or at the storage pad. Acceptable damage may occur to the TC, DSC, and the fuel stored in the DSC, for drops of height greater than 38 cm (15 inches). The specification requiring inspection of the DSC and fuel following a drop of 38 cm (15 inches) or greater ensures that the spent fuel will continue to meet the requirements for storage, the DSC will continue to provide confinement, and the TC will continue to provide its design functions of DSC transfer and shielding. 12-19
12.2.11 Transfer Cask Dose Rates Limit/Specification: Dose rates from the transfer cask shall be limited to levels which are less than or equal to:
- a. 2 mSv/hr (200 mrem/hr) at 1 m (3 feet) with water in the DSC cavity.
- b. 5 mSv/hr (500 mrem/hr) at 1 m (3 feet)" without -water in the DSC cavity.
Applicability: 'Ibis specification is applicable to the transfer cask containing a loaded DSC. Objective: The dose rate is limited to this value to ensure that the DSC has not been inadvertently loaded with fuel not meeting the specifications in Section 12.2.1 of the SER and to maintain dose rates as low as reasonably achievable during DSC transfer operations. Action: If specified dose rates are exceeded, place temporary shielding around affected areas of transfer cask and review the plant records of the fuel assemblies which have been placed in DSC to ensure they conform to the fuel specifications of Section 12.2.1. Submit a letter report to the NRC within 30 days summarizing the action taken and the results of the surveillance, investigation and findings. The report must be submitted using instructions in 10 CFR 72.4 with a copy sent to the administrator of the appropriate NRC regional office. Surveillance: The dose rates should be measured as soon as possible after the transfer cask is removed from the spent fuel pool. Basis: The basis for this limit is the shielding analysis presented in Section 7.0 of the SAR. 12-20
12.2.12 Maximum DSC Removable Surface Contamination Limit/Specification: 36.5 Bq/100 cm2 {2,200 dpm/100 cm2) for beta-gamma sources 3.65 Bq/100 cm2 (220 dpm/100 cm2) for alpha sources. Applicability: This specification is applicable to all DSCs. - Objective: To ensure_ that ~I~ of non-fixed contamination above accepted limits does not occur. ---- Action: If the required limits are not met:
- a. Flush the DSC/transfer cask annulus with demineralized water and repeat surface contamination surveys of the DSC upper surface.
- b. If contamination of the DSC cannot be reduced to an acceptable level by this means, direct surface cleaning techniques shall be used following removal of the fuel assemblies from the DSC and removal of the DSC from the transfer cask.
- c. Check and replace the DSC/transfer cask annulus seal to ensure proper installation and repeat canister loading process.
Surveillance: Following placement of each loaded DSC/transfer cask into the cask decontamination area, fuel pool water above the top shield plug shall be removed and the top region of the DSC and cask shall be decontaminated. A contamination survey of the upper 0.3 m (1 foot) of the DSC and cask shall be taken. In addition, contamination surveys shall be taken on the inside surfaces of the TC after the DSC has been transferred into the HSM. If the above surface contamination limit is exceeded, the TC shall be decontaminated. Basis: This non-fixed contamination level is consistent with the requirements of 10 CPR 71.87(i){l) and 49 CPR 173.443, which regulate the use of spent fuel shipping containers. Consequently, these contamination levels are considered acceptable for exposure to the general environment. This level will also ensure that contamination levels of the inner surfaces of the HSM and potential releases of radioactive material to the environment are minimized. 12-21
L 12.2.13 TC/DSC Lifting Heights as a Function of Low Temperature and Location Limit/Specification: 1. No lifts or handfug of the TC/DSC at any height are pe.rmism>le at DSC basket temperatures below -28.9°C (-20°F) inside spent fuel pool building.
- 2. The maximum lift height of the TC/DSC shall be 203 cm (80 inches) if the basket temperature is below -17.8°C (0°F) but higher--than-=-28.9°_C_(-70°F) inside the spent fuel pool building.
- 3. No lift height restriction is im.{X)sed on the TC/DSC if the basket temperatUre is higher than -17.8°C (0°F) inside the spent fuel pool building.
- 4. The maximum lift height and handling height for all transfer operations outside the spent fuel pool building shall be 203 cm (80 inches) aru1 the basket temperature may not be lower than -
17.8°C (0°F). Applicability: These. temperature and height limits apply to lifting and transfer of all loaded TC/DSCs inside and outside the spent fuel .{X)Ol building. 10 CPR Part 72 applies outside the spent fuel pool building and 10 CPR Part 50 applies inside the spent fuel pool building. Objective: The low temperature and height limits are im.{X)sed to ensure that brittle fracture of the ferritic steels, used in the TC trunnions and shell and in the DSC basket, does not occur during transfer operations. Action: Confirm the basket temperature before transfer of the TC. If no calculation or measurement of this value is available, then the ambient temperature may conservatively be used. Surveillance: The ambient temperature shall be measured before transfer of the TC/DSC. Bases: The basis for the low temperature and height limits is ANSI Nl4.6-1986 paragraph 4.2.6 (Reference 8) which requires at least 4.4 °C (40°F) higher service temperature than nil ductility transition (ND'I) temperature for the TC. In the case of the standardized TC, the test temperature is -40°C (-40°F); therefore, although the NDT temperature is not determined, the material will have the required 4.4° (40°F) margin if the ambient temperature is -17.8°C (0°F) or higher. This assumes the material service temperature is equal to the ambient temperature.
The basis for the low temperature limit for the DSC is NUREG/CR-1815. The basis for the handling height limits is the NRC evaluation of the s+ructural integrity of the DSC to drop heights of 203 cm (80 inches) and less. 12-23
12.2.14 TC/DSC Transfer Operations at High Ambient Temperatures Limit/Specification: 1. The ambient tempe.tature for transfer operations of a loaded TC/DSC shall not be greater that 37.8°C (100°F) (when cask is exposed to direct insolation).
- 2. For transfer operations when ambient temperatures exceed 37.8°C (100°F) up to 52°C (125°F), a solar shield shall be used to provide J)!Otecti~~ ~t_direct ~lar rac!!a~on.: .
Applicability: This ambient temperature limit applies to all transfer operations of loaded TC/DSCs outside the spent fuel pool building, the spent fuel pool building. Objective: The hj.gh temperature limit 37.8°C (l00°F) is imposed to ensure that:
- 1. The fuel cladding temperature limit is not exceeded.
- 2. The solid neutron shield material temperature limit is not exceeded, and
- 3. The corresponding TC cavity pressure limit is not exceeded.
Action: Confirm what the ambient temperature is and provide appropriate solar shade if ambient temperature is expected to exceed 37.8°C (100°F). Surveillance: The ambient temperature shall be measured before transfer of the TC/DSC. Bases: The basis for the high temperature limit is PNL-6189 for fuel clad limit, the manufacturer's specification for neutron shield and the design basis pressure of the TC internal cavity pressure. 12-24
12.2.15 Boron Concentration in the DSC Cavity Water (24-P Design Only) Limit/Specification: The DSC cavity shall be filled only with water having a boron concentration equal to, or greater than 2,000 ppm. Applicability: This limit applies only to the standardized NUHOMS-24P design. No boration in the cavity water is required for the standardized
- NUHOMS-52B system_ since that system uses fixed absorber plates.
Objective: To ensure a subcritical configuration is maintained in the case of accidental loading of the DSC with unirradiated fuel. Action: If the boron concentration is below the required weight percentage concentration (gm boron/1(}6 gm water), add boron and re-sample, and test the concentration until the boron concentration is shown to be greater than that required. Surveillance: Written procedures shall be used to independently determine (two samples analyzed by different individuals) the boron concentration in the water used to fill the DSC cavity.
- 1. Within 4 hours before insertion of the first fuel assembly into the DSC, the dissolved boron concentration in water in the spent fuel pool, and in the water that will be introduced in the DSC cavity, shall be independently determined (two samples chemically analyzed by two individuals).
- 2. Within 4 hours before flooding the DSC cavity for unloading the fuel assemblies, the dissolved boron concentration in water in the spent pool, and in the water that will be introduced into the DSC cavity, shall be independently determined (two samples analyzed chemically by two individuals).
- 3. The dissolved boron concentration in the water shall be reconfirmed at intervals not to exceed 48 hours until such time as the DSC is removed from the spent fuel pool or the fuel has been removed from the DSC.
Bases: The required boron concentration is based on the criticality analysis for an accidental misloading of the DSC with unburned fuel, maximum enrichment, and optimum mcxleration conditions. 12-25
12.2.16 Provision of TC Seismic Restraint Inside the Spent Fuel Pool Building as a Function of Horizontal Acceleration and Loaded Cask Weight Llmit/Specification: Seismic restraints shall be provided to prevent overturning of a loaded TC during a seismic event if a certificate holder determines that the horizontal acceleration is 0.40 g or greater and the fully loaded TC weight is less than 86,260 kg (190 kips). The determination of
- -horizontal acceleratiqn_acting at the CG of the loaded TC must be based on a peak horizontal ground acceleration at -the site, but sliall not-exceed 0.25 g.
Applicability: This condition applies- to all TCs which are subject to horizontal accelerations of 0.40 g or greater. Objective: To prevent overturning of a loaded TC inside the spent fuel pool building. Action: Determine what the horizontal acceleration is for the TC and determine if the cask weight is less than 86,260 kg (190 kips). Surveillance: Determine need for TC restraint before any operations inside the spent fuel pool building. Bases: Calculation of overturning and restoring moments. 12-26
12.3 Surveillance and Monitoring Paragraph 10.2.3 of the SAR outlines a single surveillance requirement proposed by PNFS. However, as discussed below, there are many items subject to monitoring. The single item subject to surveillance is the HSM air inlet and outlet passages. They shall be inspected once every 4 days to ensure that they are clear of obstructions. The SER notes that this proposed surveillance frequency could result in exceeding the HSM concrete temperature limit of 177°C (350°F) for accident conditions of blocked inlets or outlets. The concrete temperature for this adiabatic heat-up will exceed 1T/°C (350°F) in approximately 40 hours. Furthermore, the maximum fuel clad temperature will be exceeded in a 5-day period. Although the vendor-proposed 4-day inspection frequency will prevent exceeding the fuel cladding temperature, the HSM would need to be removed from service if inlets or outlets are found to be substantially blocked, and it cannot be established that the blockage is less than 40 hours. As a result of this situation, the NRC staff is requiring the following surveillance frequency for the HSM. 12.3.1 Visual Inspection of HSM Air Inlets and Outlets (Front Wall and Roof Birdscreen) Limit/Surveillance: A visual surveillance of the exterior of the air inlets and outlets shall be conducted daily. In addition, a close-up inspection shall be performed to ensure that no materials accumulate between the modules to block the air flow. Objective: To ensure that HSM air inlets and outlets are not blocked for more than 24 hours to prevent exceeding the allowable HSM concrete temperature or the fuel cladding temperature. Applicability: This specification is applicable to all HSMs loaded with a DSC loaded with spent fuel. Action: If the surveillance shows blockage of air vents (inlets or outlets), they shall be cleared. If the screen is damaged, it shall be replaced. Basis: The concrete temperature could exceed lT/°C (350°F) in the accident circumstances of complete blockage of all vents if the period exceeds approximately 40 hours. Concrete temperatures over 177°C (350°F) in accidents (without the presence of water or steam) can have uncertain impact on concrete strength and durability. A conservative analysis (adiabatic heat case) of complete blockage of all air inlets or outlets indicates that the concrete can reach the accident temperature limit of l 77°C (350°F) in a time period of approximately 40 hours. 12-27
12.3.2 HSM Thermal Performance Surveillance: Verify a temperature measurement of the thermal performance, for each HSM, on a daily basis. The temperature measurement could be any parameter such as (1) a direct measurement of the HSM temperatures, (2) a direct measurement of the DSC temperatures, (3) a comparison of the inlet and outlet temperature difference to predicted temperature differences for _each indivi4ual H§M, or (4) other means that would identify and allow for the correction of off-normal tliernial conditions that could lead to exceeding the concrete and fuel clad temperature criteria. If air temperatures are measured, they must be measured in such a manner as to obtain representative values of inlet and outlet air temperatures. Also due to the proximity of adjacent HSM modules, care must be exercised to ensure that measured air temperatures reflect only the thermal performance of an individual module, and not the combined performance of adjacent modules. Action: If the temperature measurement shows a significant unexplained difference, so as to indicate the approach of materials to the concrete or fuel clad temperature criteria, take appropriate action to determine the cause and return the canister to normal operation. If the measurement or other evidence suggests that the concrete accident temperature criteria l77°C (350°F) has been exceeded for more than 24 hours, the HSM must be removed from service unless the licensee can provide test results in accordance with ACI-349, appendix A.4.3, demonstrating that the structural strength of the HSM has an adequate margin of safety. Basis: The temperature measurement should be of sufficient scope to provide the licensee with a positive means to identify conditions which threaten to approach temperature criteria for proper HSM operation and allow for the correction of off-nonnal thermal conditions that could lend to exceeding the concrete and fuel clad temperature criteria. 12-28
Table 12.3.1 Summary of Surveillance and Monitoring Requirements Surveillance or Monitoring Period Reference Section
- 1. Fuel Specification PL 12.2.1
- 2. - DSC Vacuum Pressure During Drying L - -
12.2.2
- 3. DSC Helium Backfill Pressure L 12.2.3
- 4. DSC Helium Leak Rate of Inner Seal L 12.2.4 Weld
- 5. DSC Dye Penetrant Test of Closure L 12.2.5 Welds
- 6. DSC Top End Dose Rates L 12.2.6
- 7. HSM Dose Rates L 12.2.7
- 8. HSM Maximum Air Exit Temperature 24 hrs 12.2.8
- 9. TC Alignment with HSM s 12.2.9
- 10. DSC Handling Height Outside Spent AN 12.2.10 Fuel Pool Building
- 11. Transfer Cask Dose Rates L 12.2.11
- 12. Maximum DSC Surface Contamination L 12.2.12
- 13. TC/DSC Lifting Heights as a Function L 12.2.13 of Low Temperature and Location Legend PL Prior to loading L During loading and prior to movement to HSM pad 24 hrs Time following DSC insertion into HSM S Prior to movement of DSC to or from HSM AN As necessary D Daily (24 hour frequency) 12-29
Table 12.3.1 Summary of Surveillance and Monitoring Requirements (Continued) Surveillance or Monitoring Period Reference Section
- 14. TC/DSC Transfer Operations at ffigh L 12.2.14 Ambient Temperatures
- 15. Boron Concentration in DSC Cavity PL 12.2.15 Water (24-P Design Only)
- 16. Provision of TC Seismic Restraint PL 12.2.16 Inside the Spent Fuel Pool Building as a Function of Horizontal Acceleration and Loaded Cask Weight
- 17. Visual Inspection of HSM Air Inlets and D 12.3.1 Outlets
- 18. HSM Thermal Performance D 12.3.2 Legend PL Prior to loading L During loading and prior to movement to HSM pad 24 hrs Time following DSC insertion into HSM S Prior to movement of DSC to or from RSM AN As necessary D Daily (24 hour frequency) 12-30
13.0 REFERENCES
- 1. Pacific Nuclear Fuel Services, Llc., *safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel," Docket No. 72-1004:
- a. NUH-003, Rev. 0, December 20, 1990.
- b. NUH-003, Rev. 1, September 25, 1991.
- c. NUH-003, Rev. 2, November 5, 1993.
[Note: Where different submittals are at variance in addressing the same subject, the most recent submittal governs.]
- 2. Office of the Federal Register, Code of Federal Regulations, Title 10 - Energy, Chapter I - Nuclear Regulatory Commission (10 CFR), Revised as of January 1, 1993. Following parts referenced:
- a. 10 CFR Part 72, *Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste."
- b. 10 CPR Part 20, 11 Standards for Protection Against Radiation."
- c. 10 CPR Part 50, "Domestic Licensing of Production and Utilization Facilities. 11
- d. 10 CFR Part 71, "Packaging and Transportation of Radioactive Material."
- 3. U.S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the TR for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel Submitted by Nutech Engineers, Inc.," NUH-002, Rev. lA, April 1989.
- 4. U.S. Nuclear Regulatory Commission, "Safety Evaluation Report for a Design Change to the Transfer Cask for the Duke Power Company's ISFSI,
- February 1990.
- 5. U.S. Nuclear Regulatory Commission, *standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage Installation or Monitored Retrievable Storage Installation (Dry Storage), 11 Regulatory Guide 3.48, August 1989.
- 6. U.S. Nuclear Regulatory Commission, "Quality Assurance Program Requirements (Design and Construction)," Regulatory Guide 1.28, February 1979.
13-1
- 7. U.S. Nuclear Regulatory Commission, "Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers up to Four Inches Thick," NUREG/CR-1815, August B81.
- 8. American National Standards Institute, ANSI N14.6-1986, *special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More," 1987.
- 9. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Section m, Division l ,_ 1983 Edition with_ Winter 1985 Addenda. _ __ _ _ ___ _
- 10. U.S. Nuclear Regulatory Commission, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As is Reasonably Achievable,* Regulatory Guide 8.8, Rev. 3, June 1978.
- 11. U.S. Nuclear Regulatory Commission, *Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable," Regulatory Guide 8.10, May 19'n.
- 12. Fintel, M., Handbook of Concrete Enmeering, Van Nostrand Reinhold Co.,
New York, New York (1985).
- 13. Bolz, R.E., and G.L. Tuve, CRC Handbook of Tables for Applied Science, 2nd Edition, Chemical Rubber Co., 1973.
- 14. U.S. Nuclear Regulatory Commission, "Control of Heavy Loads at Nuclear Power Plants," NUREG-0612, July 1980.
- 15. U.S. Nuclear Regulatory Commission, "Safety Evaluation Report for Pacific Sierra Nuclear TR of a Ventilated Storage Cask System for Irradiated Fuel," March 1991.
- 16. U.S. Nuclear Regulatory Commission, "Safety Evaluation Report for the Baltimore Gas and Electric Company's Safety Analysis Report for an Independent Spent Field Storage Installation at Calvert Cliffs," November 1992.
- 17. American Concrete Institute, "Code Requirements for Nuclear Safety Related Concrete Structures and Commentary," ACI 349-85, American Concrete Institute, Detroit, Michigan, 1985.
- 18. PNFS, "Standard NUHOMS Prefabricated Module - HSM and DSC Support Structure Analysis and Design,* NUH004.0200/Rl, R2, R3, R4, RS.
- 19. American Concrete Institute, "Building Code Requirements for Reinforced Concrete, 11 ACI 318-83, 1983.
13-2
- 20. PNFS, "Thermal Analysis of Standardized NUHOMS During Blocked Inlet and Outlet Openings in the HSM Sidewalls for 1 kW Per Fuel Assembly Decay Heat,"
NUH004.0416/R0, December 12, 1991.
- 21. Swanson Analysis Systems, Inc., ANSYS En&ineerin~ Ana]ysis Systems User's Manual, Version 4.4, Volumes 1 and 2, Pittsburgh, Pennsylvania.
- 22. American National Standards Institute, ANSI 57.9-1984 *Design Criteria for an
_]ndepenc;Iept_Spent fuel S~e _l!lst:all@iQ!l. (Dry s~~-1YIJe)."
- 23. U.S. Nuclear Regulatory Commission, *standard Review Plan," NUREG-0800, Rev. 2.
- 24. U.S. Nuclear Regulatory Commission, "Design Basis Tornado for Nuclear Power Plants,* Regulatory Guide 1. 76, April 1974.
- 25. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.60, "Design Response Spectra for Seismic Design of Nuclear Power Plants, Rev. 1," December 1973.
- 26. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.61, "Damping Values*for Seismic Design of Nuclear Power Plants,* October 1973.
- 27. U.S. Nuclear Regulatory Commission, "Combining Model Responses and Spatial Components in Seismic Response Analysis,* Regulatory Guide 1.92, February 1976.
- 28. National Fire Protection Association Codes, *Lightning Protection Code,* NFPA 78.
- 29. PNFS, "Standardized NUHOMS HSM Air Flow Calculation," NU1{004.0418/R0, 4lt December 22, 1992.
- 30. PNFS, "Standardized NUHOMS HSM Heat Transfer Analysis," NUH004.0419/R0, December 22, 1992.
- 31. U.S. Nuclear Regulatory Commission, "Design of an Independent Spent Fuel Storage Installation (Dry Storage),* Regulatory Guide 3.60, March 1987 (which incorporates ANSI/ANS 57.9-1984 (Reference 23).
- 32. American Institute of Steel Construction (AISC), "Specifications for Structural Steel Buildings,* 1989, contained in the AISC Manual of Steel Construction, Nmth Edition, 1989.
- 33. PNFS,
- 10 CPR 71. NUHOMS-24P and -52B DSC Structural Analysis,"
NUH004.0202/R2, January 4, 1993. 13-3
- 34. PNFS, "Cask Axial and Radial Thennal Analysis for Standardized NUHOMS-24P Design with 5-year Old Fuel (1 kW per Fuel Assembly)," NUH004.0407/Rev. 0, September 10, 1991.
- 35. Electric Power Research Institute, *Tue Effects of Target Hardness on the Structural Design of Concrete Storage Pads for Spent Fuel Casks," NP-4830, October 1986.
- 36. Electric Power Research Institute, "Structural Design of Concrete Storage Pads for Spent Fuel Casks," NP-7551-, August 1991.
- 37. Warrant, M. and J. Joseph, "Test Data Report for Quarter Scale NUPAC 125-B Cask Model," Report No. OEND-INF-091, Sandia National Laboratories, February 1988.
- 38. NUTECH Engineers Inc., NUH-002, "Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUHOMS-24P,
- Rev. lA, July 1989, and additional docketed supporting and modifying submittals.
- 39. U.S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel Topical Report NUHOMS-24P, submitted by NUTECH Engineers, Inc.," April 1989.
- 40. Pacific Nuclear Fuel Services, Inc., TR Amendment 2 for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUHOMS-24P (NUH-002, Rev. 2).
- 41. NUTECH Engineers, Inc., "Transfer Cask Structural Analysis," Cale. No.
- BGE 001.0202, Rev. 4, 1990.
- 42. PNFS, "Standardized NUHOMS BWR Cask Collar Structural Evaluation,"
NUH004.0205/R0.
- 43. PNFS, "Standard NUHOMS Transfer Cask Thermal Stress Analysis,"
NUH004.0206/R0.
- 44. American National Standards Institute/ASME N00-1-1983, "Rules for Construction of Overhead and Gantry Cranes," 1983.
- 45. NUTECH Engineers, Inc., "Topical Report NUH-001 for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUHOM-7P," Rev. lA.
- 46. U.S. Nuclear Regulatory Commismon, "Safety Evaluation Report for NUTECH Horizontal Modular System for Irradiated Fuel Topical Report," March 28, 1986.
13-4
- 47. Levy, I.S., et al., "Recommended Temperature Limits for Dry Storage of Spent Light Water Reactor Zircaloy-Clad Fuel Rods in Inert Gas," Pacific Northwest Laboratory Report, PNL-6189, May 1987.
- 48. Johnson, A.B., Jr., and E.R. Gilbert, "Technical Basis for Storage of Zircaloy-Clad Spent Fuel in Inert Gases," PNL-4835, September 1983.
- 49. Elrod, D.C., et al., "HEATING-6: A Multidimensional Heat Conductor Analysis With the Finite Difference Formulati.QIJ," NUREG/CR-200, Vol. 2, Sec. FlO, ORNLJNUREG/CSD-2N2, October 1981.
- 50. Pacific Nuclear Corporation NUH004.0416, "Thennal Analysis of Standardized NUHOMS During Blocked II1let and Outlet Openings in the HSM Sidewalls for 1 kW Per Fuel Assembly Decay Heat," Rev. 0, December 12, 1991.
- 51. Pacific Nuclear Corporation, NUH004.0418, Rev. 0, "Standardized NUHOMS HSM Air Flow Calculation," December 22, 1992.
- 52. Pacific Nuclear Corporation, NUH004.0419, Rev. 0, "Standard NUHOMS RSM Heat Transfer Analysis," December 22, 1992.
- 53. Pacific Nuclear Corporation NUH004.0412, *NUHOMS-24P DSC Thermal Analysis for 1 kW Fuel,U Rev. 1, December 22, 1992.
- 54. Pacific Nuclear Corporation NUH004.0414, "NUHOMS-24B DSC Thermal Analysis,1' Rev. 1, December 22, 1992.
- 55. Pacific Nuclear Corporation DUK003.0203, "Dry Storage Cladding Temperature Limits for the 24P NUHOMS System Using the CSFSM Model Presented in PNL-6189," Rev. 0, July 11, 1991.
- 56. Pacific Nuclear Corporation NUH004.0410, anry Storage Cladding Temperature Limits for the Standardized NUHOMS-52B Using the CSFM Model Presented in PNL-6189," Rev. 0, August 28, 1991.
- 57. Pacific Nuclear Corporation NUH004.0507, "NUHOMS-24P Radiological and Thennal Source Term Calculation," Rev. 0, August 14, 1992.
- 58. Pacific Nuclear Corporation NUH004.0407, "Cask Axial and Radial Thermal Analysis for Standardized NUHOMS-24P Design with 5-year Old Fuel (1 kW per Fuel Assembly)," Rev. 0, September 10, 1991.
- 59. Pacific Nuclear Corporation NUH004.0406, "Cask Axial and Radial Thermal Analysis for Standardized NUHOMS-52B Design," Rev. 0, September 10, 1991.
13-5
- 60. Oak Ridge National Laboratory, SCALE-3: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation, NUREG/CR-0200, Rev. 3, December 1984.
- 61. Ryman, J.C., O.W. Herman, C.C. Webster, C.V. Parks, "Fuel Inventory and Afterheat Power Studies of Uranium-Fuel Pressurized Water Reactor Fuel Assemblies Using the SAS2 and ORIGEN-S Modules of SCALE with an ENDF/-B-V Updated Cross-Section Llbrary,
- NUREG/CR-2397 (ORNL-CSD-90), U.S. Nuclear Regulatory_ Co~on _and Oak Ridge National Labora~cy, 1980.
- 62. Office of Civilian Radioactive Waste Management,* Characteristics of Spent Fuel, High-Level Waste, and Other Radioactive Wastes Which May Require Long-Term Isolation," DOFJRW-0184, December 1987.
- 63. Grove Engineering, Inc., "Microshield User's Manual, A Program for Analyzing Gamma Radiation Shielding,* Version 2.0, 1985.
- 64. Oak Ridge National Laboratory, "ANISN/PC - Multi.group One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering," CCC-514 MICRO, Oak Ridge National Laboratory, 1977.
- 65. Oak Ridge National Laboratory, "QAD-CGGP, A Combinational Geometer Version of QAD-P5A, A Point Kernal Using the GP Buildup Factor," CCC-493, Oak Ridge National Laboratory, 1986.
- 66. Ede.mus, Malte, et al., "CASM0 A Fuel Assembly Burnup Program,*
STUDSVIK/NR-81/3, March 1981.
- 67. SAS2 SCALE-3, Op. Cit
- 68. Petrie, L.M. and N.F. Cross, "KENO IV: An Improved Monte Carlo Criticality Program." ORNL-4938, November 1975.
- 69. Oak Ridge National Laboratory, *KENO5A-PC, Monte Carlo Criticality Program with Supergrouping," CCC-548, June 1990.
- 70. PNFS, "QA Category 2 Computer Code Verification Document KENO5A, PNFSI Version 1.2.0," Rev. 0.
- 71. Proceedings of the Workshop on Source Term for Radionuclide Migration from High Level Waste or Spent Nuclear Fuel - Page 175, Albuquerque, NM, November 13-15, 1984, SAND 85-0380, Hunter and Muir, Editors.
13-6
- 72. U.S. Nuclear Regulatory Commission, *standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks," Regulatory Guide 3.62, February 1989.
13-7
Certificate of Compliance FOR DRY SPENT FUEL STORAGE CASKS 10 CFR PART 72
- 1. a. CERTIFICATE tlJMBER: 1004
- b. REVISION NUMBER: 0
- c. PACICAGE IDENTIF TI
- d. PAGE NUMBER: 1
- e. TOTAL nuni.~
- 2. is certificate is issued to certify that the cask a
~:taa in item 5 below, aeet the applicable safety standard Titl , Code of Federal Regulations, Part 72,*Licensing Requir
~~~e Storage of Spent Nuclear Fuel and High-~.,,,,.,. dio
- 3. /-.,1s CERTIFICAT
,_,,,JSk design, No
- a. ~ED BY (Nue ~~;2 VECr_echnol k.a Je:
Pa~~Nuclea ,~nc. 6203 San Ignac he San CA 9 1 N r FueP 2-1004
- 4. This certificate ing the reaents
~!rJ!li*~-rt 72, as applica ' pecif
- Effective Date:
Expiration Date: Charles J. Haughney, Chief Storage and Transport Systems Branch Division of Industrial and Medical Nuclear Safety, NMSS
Certificate of Compliance FOR DRY SPENT FUEL STORAGE CASKS 10 CFR PART 72 2. Services, lnco izontal Modul 5.
- b. Description The Standardized NUHOMS System and its analyses and operations are described in the SAR (Docket 72-1004) identified previously. The Nuclear Regulatory Commission has reviewed the SAR in the Safety Evaluation Report identified previously.
The system which is being certified is described in Sections 1, 3, 4, 5, 6, 7 and 8 of the SAR and in the NRC's SER accompanying the SAR. (The system drawings, which reflect this description, are contained in Appendix E of the SAR.) The Standardized NUHOMS System is a horizontal canister system composed of a steel dry shielded canister (DSC), a reinforced concrete horizontal storage module (HSM), and a transfer cask (TC). The welded DSC provides confinement and criticality control for the storage and transfer of irradiated fuel. The concrete module provides radiation shielding while allowing cooling of the DSC and fuel by natural convection during storage. The TC is used for transferring the DSC from/to the Spent Fuel Pool Building to/from the HSM. The principal component subassemblies of the DSC are the shell with integral bottom cover plate and shield plug and ram/grapple ring, top shield plug, top cover plate, and basket assembly. The shell length is fuel-specific. The internal basket assembly is composed of guide sleeves, support rods, and spacer disks. This assembly is designed to hold 24 PWR fuel assemblies or 52 BWR assemblies. It aids in the insertion of the fuel assemblies, enhances subcriticality during loading operations, and provides structural support during a hypothetical drop accident. The DSC is designed to slide from the transfer cask into the HSH and back without undue galling, scratching, gouging, or other damage to the sliding surfaces. The HSM is a reinforced concrete unit with penetrations located at the top and bottom of the side walls for air flow. The penetrations are protected from debris intrusions by wire mesh screens during storage operation. The DSC Support Structure, a structural steel frame with rails, is installed within the HSM module to provide for sliding the DSC in and out of the HSM and to support the DSC within the HSM. The TC is designed and fabricated as a lifting device to meet NUREG-0612 and ANSI Nl4.6 requirements. It is used for transfer operations within the Spent Fuel Pool Building and for transfer operations to/from the HSM. The TC is a cylindrical vessel with a bottom end closure assembly and a bolted top cover plate. Two upper lifting trunnions are located near the top of the cask for downending/uprighting and lifting of the cask in the Spent Fuel Pool Building. The lower trunnions, located near the base of the cask, serve as the axis of rotation during downending/uprighting operations and as supports during transport to/from the Independent Spent Fuel Storage Installation (ISFSI). With the exception of the TC, fuel transfer and auxiliary equipment necessary for ISFSI operations are not included as a part of the Standardized NUHOMS System to be reviewed for a Certificate of Compliance under 10 CFR Part 72, Subpart L. Such equipment may include, but is not limited to, special lifting devices, the transfer trailer, and the skid positioning system. 2
- c. Drawings The drawings for the dry irradiated fuel storage canister system are contained in Appendix E of the SAR.
- d. Basic Components The basic components of the Standardized NU.HOt4S__ Sy_§:tem J:hat are_ important to__
safety are- the-DSC, HSM,--and-TC~ *These components are described in Section 4.2 of the SAR.
- 6. Fabrication activities shall be conducted in accordance with a quality assurance program as described in Section 11.0 of the SAR.
- 7. Notification of fabrication schedules shall be made in accordance with the requirements of 10 CFR 72.232(c).
- 8. Standardized NUHOMS Systems, which are authorized by this certificate, are hereby approved for general use by holders of 10 CFR Part 50 licenses for nuclear reactors at reactor sites under the general license issued pursuant to 10 CFR 72.210, subject to the conditions specified by 10 CFR 72.212 and the attached Conditions for System Use.
- 9. Changes, tests, and experi11ents The holder of this certificate of compliance may:
(1) Make changes in the cask design described in the Safety Analysis Report, (2) Make changes in the procedures described in the Safety Analysis Report, or I * (3) Conduct tests or experiments not described in the Safety Analysts Report, without prior C011111ission approval, unless the proposed change, test or experiment involves a change in the Certificate of Compliance, an unreviewed safety question, a significant increase in occupational exposure or a significant unreviewed environmental impact. A proposed change, test, or experiment shall be deemed to involve an unreviewed safety question: {l) If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased; or (2) If a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created; or (3) If the margin of safety as defined in the basis for any technical specification or limit is reduced. The holder of this certificate of compliance shall maintain records of changes in the cask design and of changes in procedures if the changes consititute changes in the cask design or procedures described in the Safety Analysis Report. The 3
holder of this certificate of compliance shall also maintain records of tests and experiments it conducts that are not described in the Safety Analysis Report. These records must include a written safety evaluation that provides the bases for the determination that the change, test or experiment does not involve an unreviewed safety question. The records of changes in the cask design and of changes in procedures and records of tests or experiments must be maintained until the C0111t1ission terminates the certificate. The holder of this certificate of compliance shall annually furnish to the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Co11111ission, Washington, DC 20555, a report containing a brief description of changes, tests, and experiments made under this provision, including a sununary of the safety evaluation of each. Any such report submitted by a holder of this certificate of compliance will be made a part of the public record pertaining to this certificate. The holder of this certificate of compliance who desires: (1) To make changes in the cask design or the procedures as described in the Safety Analysis Report, or to conduct tests or experiments not described in the Safety Analysis Report, that involve an unreviewed safety question, a significant increase in occupational exposure, or a significant unreviewed environmental impact, or (2) To change the Certificate of Compliance shall submit an application for amendment of the certificate.
- 10. Effective Date:
Expiration Date: FOR THE NUCLEAR REGULATORY COMMISSION Charles J. Haughney; Chief Storage and Transport Systems Branch Division of Industrial and Medical Nuclear Safety, NMSS 4
ATIACHMBNTA CONDIDONS FOR SYSTEM USE CERTIFICATE OF COMPLIANCE 72-1004
TABLE OF CONTENTS Section fm
- 1. 0 IN'l'R.ODUCTION . . . . . . . . . . . . . * . . . . . . . . . . . . . . . . . . . . . A-1
- 1. 1 General Requirements and Conditions . . . . . * . . . . . . . . . . . . . . A-1
- 1. 1. 1 Regulatory Requirements for a General License . . . . . . . A-1 1.1.2 ~ PrcK:ec:llll'C,S * * * * * * * * * * * * * * * * * . * * * ** A-2 1.1. 3 Quality Assurance . . . * . * . . . . . . . . . . . . . . . . . .. A-3 1.1.4 Heavy Loads Requirements . . . . . . . . . . . . . . . * . . . . A-3
- 1.1.5 Training Module * * * . * . . . . . . * . . . . . -. . . .- . . . *. A-3 1.1. 6 Pre-Operational Testing and Training Exercise . . . . . * .. A-4 1.1. 7 Special Requirements for First System in Place . . . . . . . . A-4
- 1. 1.8 Surveillance Requirements Applicability . . . . . * . . . . *. A-5 1.2 Technical Specifications, Functional and Operating Limits . . . . . . . . A-5
- 1. 2.1 Fuel Specification * * * . . . . . * . * * . . . . * . . * . . . . . A-5 1.2.2 DSC Vacuum Pressure During Drying . . * * . . . . . . . .. A-11 1.2.3 DSC Helium Backfill Pressure . . . . . . . . . . . . . . . . . . A-12 1.2.4 DSC Helium Leak Rate of Inner Seal Weld . . . . . . . . . . A-13 1.2.5 DSC Dye Pe.netrant Test of Closure Welds . . . . . . . . . . A-14 1.2.6 DSC Top End Dose Rates . . . . . . . . . . . . . . * . . . . . A-15 1.2. 7 HSM Dose Rates . . . * . . . . . . . . * . . . . . . . . . ... . A-16 1.2.8 HSM Maximum Air Exit Temperature . . . . . . . . . . . . . A-17 1.2.9 Transfer Cask Alignment with HSM . . . . . * . . . . . . ** A-18 1.2.10 DSC Handling Height Outside the Spent Fuel Pool Building . . . . . . . . . . . . . . . . . . . . . . . . . A-19 1.2.11 Transfer Cask Dose Rates . . . . . . . . . . . . . . . . . .... A-20 1.2.12 Maximum DSC Removable Surface Contamination . . . . . A-21 1.2.13 TC/DSC Lifting Heights 88 a Function of
- 1.2.14 1.2.15 Low Temperature and Location . . * . . . . . . . . . . TC/DSC Transfer Operations at High Ambient Temperatures * * * . * * . * * . . * . * . . * . . Boron Concentration in the DSC Cavity Water (24-P Design Only) . . . * * * * . . . . . . * . . * * . . A-22 A-23 A-24 1.2.16 Provision of TC Seismic Restraint Inside the Spent Fuel Pool Building 88 a Function of Horizontal Acceleration and Loaded Cask Weight . . . . . . A-25 1.3 Surveillance and Monitoring . * * * . * . . * * * . . . . * . . * . . . ...* A-26 1.3.1 VJ.SU&l Inspection of HSM Air Inlets and Outlets (Front Wall and Roof Birdscreen) . . . * * . . . . * . * . . . . A-26 1.3.2 HSM 'Iberma1 Performance . . . . . . . . . . . . . . . .... A-27 l
LIST OF FIGURES Figure ~ 1.1 PWR Fuel Criticality Acceptance Curve A-9 UST OF TABLES IBhlc 1-la PWR Fuel Specifications of Fuel to be Stored in the Standardized NUHOMS-24P DSC . * * . * . . . * . . . * . . . . . . . . . . A-8 1-lb BWR Fuel Specifications of Fuel to be Stored in the Standardized NUHOMS-S2B DSC . * . . . * * . . * . . * . . . . . . . . . . A-10 1.3.1 Summary of Surveillance and Monitoring Requirements . . . . . . . . . . A-28 ii
LO INTRODUCTION This section presents the conditions which a potentia1 user (general licensee) of the standardized NUHOMS system must comply with, in order to use the system under the genera1 license in accordance with the provisions of 10 CFR 72.210 and 10 CFR 72.212. These conditions have either been proposed by the system vendor, imposed by the NRC staff as a result of the review of the SAR, or are part of the regulatory requirements expressed in 10 CFR 72.212. 1.1 General Requirements and Conditions 1.1.1 R.egulatmy Requirements for a General license Subpart K of 10 CFR Part 72 contains conditions for using the general license to store spent fuel at an independent spent fuel storage installation at power reactor sites authorized to possess and operate nuclear power reactors under 10 CFR Part 50. Technical regulatory requirements for the licensee (user of the standardized NUHOMS system) are contained in 10 CFR 72.212(b). Under 10 CFR 72.212(b)(2) requirements, the licensee must perform written. evaluations, before use, that establish that: (1) conditions set forth in the Certificate of Compliance have been met; (2) cask storage pads and areas have beal designed to adequately support the static load of the stored casks; and (3) the requirements of 10 CFR 72.104 "Criteria for radioactive materials in effluent and direct radiation from an ISFSI or MRS,* have been met. In addition, 10 CFR 72.212(b)(3) requires that the licensee review the SAR and the associated SER, before use of the general license, to determine whether or not the reactor site parameters (including earthquake intensity and tornado missUes), are encompassed by the cask design bases considered in these reports. The requirements of 10 CFR 72.212(b)(4) provide that, as a holder of a Part 50 license, the user, before use of the general license under Part 72, must determine whether activities related to storage of spent fuel involve any unreviewed safety issues, or changes in teclmlcal specifications as provided under 10 CPR 50.59. Under 10 CFR 72.212(b)(5), the general license holder shall also protect the spent fuel against design basis threats and radiological sabotage pursuant to 10 CFR 73.55. Other general license requirements dealing with review of reactor emergency plans, quality assurance program, training, and :radiation protection program must also be satisfied pursuant to 10 CFR 72.212(b)(6). Records and procedural requirements for the geneial license holder are described in 10 CPR 72.212(b)(7), (8), (9) and (10). Wrthout limiting the requirements identified above, site-specific parameters and analyses, identified in the SER, that will need verification by the system user, are as a minimum, as follows: A-1
- 1. The temperature of 70°F as the maximum ave.rage yearly temperature with solar incidence. The average daily ambient tempeiature shall be 100°F or~
(Reference SER Section 2.4.1).
- 2. The temperature extremes of 125°F with incident solar radiation and -40°F with no solar incidence (Reference SER Section 2.4.1) for storage of the DSC inside the HSM.
- 3. The horizontal and vertical seismic acceleration levels of 0.25g and 0.17g, respectively (Reference SER Table 2-4).
- 4. The analyzed flood condition of 15 fps water velocity and a height of 50 feet of water (full submergence of the loaded HSM DSC) (Reference SER Table 2-4).
- 5. The potential for fire and explosion should be addressed, based on site-specific considerations (See SER Table 2-4 and related SER discussion).
- 6. The HSM foundation design criteria are not included in the SAR. Therefore, the nominal SAR design QI' an alternative should be verified for individual sites in accordance with 10 CFR 72.212(b)(2)(tl). Also, in accordance with 10 CFR 72.212(b)(3), the foundation design should be evaluated against actual site parameters to determine whether its failure would cause the standardized NUHOMS system to exceed the design basis accident conditions.
- 7. The potential for lightning damage to any electrical system associated with the standardiz.ed NUHOMS system (e.g., thermal perlormance monitoring) should be addressed, based on site-specific considerations (See SER Table 2. 4 and related SER discussion).
- 8. Any other site parameters or consideration that could decrease the effectiveness of cask systems important to safety.
In accordance with 10 CFR 72.212(b)(2), a record of the written evaluations must be retained by the licensee until spent fuel is no longer stored under the general license issued under 10 CFR 72.210. 1.1.2 Operating Procedures Written operating procedures shall be prepared for cask handling, loading, movement, surveillance, and maintenance. The operating procedures suggested generically in the SAR were considered appropriate as discussed in Section 11.0 of the SER and should provide the basis for the user's written operating procedure. The following additional procedure requested by NRC staff in Section 11.1 should be part of the user operating procedures:
If fuel needs to be removed from the DSC, either at the end of service life or for inspection after an accident, precautions must be taken against the potential for the presence of damaged or oxidized fuel and to prevent radiological exposure to personne1 during this operation. This can be achieved with this design by the use of the purge and fill valves which permit a determination of the atmosphere within the DSC before the removal of the inner top cover plate and shield plugs, prior to filling the DSC cavity with borated water (see SAR paragraph 5.1.1.9). If the atmosphere within the DSC is helium, then operations should proceed normally with fuel removal either via the transfer cask or in the pool However, if air is present within the DSC, then appropriate filters should be in place to preclude the uncontrolled release of any potential airborne radioactive particulate from the DSC via the purge-fill valves. This will protect both personne1 and the operations area from potential contamination. For the accident case, personnel protection in the form of respirators or supplied air should be considered in accordance with the licensee's Radiation Protection Program. 1.1. 3 Quality Assurance Activities at the ISFSI shall be conducted in accordance with a Commission-approved quality assurance program whicb satisfies the applicable requirements of 10 CFR Part 50, Appendix B, and which is established, maintained, and executed with regard to the ISFSI. 1.1.4 Heavy Loads Requirements Lifts of the DSC in the TC must be made within the existing heavy loads requirements and procedures of the licensed nuclear power plant The TC design has been reviewed under 10 CFR Part 7l and found to meet NUREG-0612 and ANSI N14.6. (Reference 8). However, an additional safety review (under 10 CFR 50.59) is required to show operational compliance with NUREG-0612 and/or existing plant-specific heavy loads requirements. 1.1.5 Training Module A training module shall be developed for the existing licensee's training program establishing an ISFSI training and certification program. This module shall include the following:
- 1. Standardized NUHOMS Design (overview);
- 2. ISFSI Facility Design (overview);
- 3. Certificate of Compliance conditions (overview);
- 4. Fuel 1-0ading, Transfer Cask Handling, DSC Transfer Procedures; and
- 5. Off-Normal Event Procedures.
A-3
1.1.6 Pre-Operational Testing and Training Exercise A dry run of the DSC loading, TC handling and DSC insertion into the HSM shall be held. This dry run shall include, but not be limited to, the following:
- 1. Functional testing of the TC with lifting yokes to ensure that the TC can be safely transported over the entire route required for fuel loading, washdown pit and trailer loading.
- 2. DSC loading into the TC to verify fit and TC/DSC annulus seal.
- 3. Testing of TC on transport trailer and transported to ISFSI along a predetermined route and aUgned with an HSM.
- 4. Testing of transfer trailer aHgnment and docking equipment. Testing of hydraulic ram to insert a DSC loaded with test weights into an HSM and then retrieve it.
- 5. Loading a mock-up fuel assembly into the DSC.
- 6. DSC ~Jing, vacuum drying, and cover gas backfiltjng operations (using a mock-up DSC).
- 7. Opening a DSC (using a mock-up DSC).
- 8. Returning the DSC and TC to the spent fuel pool.
1.1.7 Special Requirements for First System in Place The heat transfer characteristics of the cask system will be recorded by temperature measurements of the first DSC placed in service. The first DSC shall be loaded with assemblies, constituting a source of approximately 24 kW. The DSC shall be loaded into the HSM, and the thermal performance will be assessed by measuring the air inlet and outlet temperatures for normal airflow. Details for obtaining the measurements are provided in Section 1.2.8, under *Surveillance.* A letter report summarizing the results of the measurements shall be submitted to the NRC for evaluation and assesmnent of the heat removal characteristics of the cask in place within 30 days of placing the DSC in service, in accordance with 10 CPR 72.4. Should the first user of the system not have fuel capable of producing a 24 kW heat load, or be limited to a lesser heat load, as in the case of BWR fuel, the user may use a lesser load for the process, provided that a ca1culation of the temperature difference between the inlet and outlet temperatures is performed, using the same methodology and inputs documented in A-4
the SAR, with lesser load as the only exception. The calculation and the measured temperature data shall be reported to the NRC in accordance with 10 CFR 72.4. The calculation and comparison need not be reported to the NRC for DSCs that are subsequently loaded with lesser loads than the initial case. Eowever, for the first or any other user, the process needs to be perlormed and reported for any higher heat sources, up to 24 kW for PWR fuel and 19 kW for BWR fuel, which is the maximum allowed under the Certificate of Compliance. The NRC will also accept the use of artificial thermal loads other than spent fuel, to satisfy the above requirement. 1.1. 8 Surveillance Requirements Applicability The specified frequency for each Surveillance Requirement is met if the surveillance is performed within 1.25 times the interval specified in the frequency, as measured from the previous performance. For frequencies specified as "once,* the above interval extension does not apply. If a required action requires performance of a surveillance or its completion time requires period performance of *once per*.* ,* the above frequency extension applies to the repetitive portion, but not to the initial JX)rtion of the completion time. Exceptions to these requirements are stated in the individual specifications. 1.2 Technical Specifications, Functional and Operating Limits 1.2.1 Fuel Specification Limit/Specification: The characteristics of the spent fuel which is allowed to be stored in the standardized NUHOMS system are limited by those included in Tables 1-la and 1-lb. Applicability: The specification is applicable to all fuel to be stored in the standardized NUHOMS system. Objective: The specification is prepared to ensure that the peak fuel rod temperatures, maximum surface doses, and nuclear criticality effective neutron multiplication factor are below the design values. Furthermore, the fuel weight and type ensures that structural conditions in the SAR bound those of the actual fuel being stored. Action: Each spent fuel assembly to be loaded into a DSC shall have the parameters listed in Tables 1-la and 1-lb verified and documented. Fuel not meeting this specification shall not be stored in the standardized NUHOMS system. A-5
Surveillance: Immediately, before insertion of a spent fuel assembly into an DSC, the identity of each fuel assembly shall be independently verified and documented. Bases: The specification is based on consideration of the design basis parameters included in the SAR and limitations imposed as a result of the staff review. Such parameters -stem from the type of fuel anal}17,ed, structural limitations, criteria for criticality safety, criteria for heat removal, and criteria for radiological protection. The standardized NUHOMS system is designed for dry, horizontal storage of irradiated light water reactor (LWR) fuel The principal design parameters of the fuel to be stored can accommodate standard PWR fuel designs manufactured by Babcock and Wtlcox, Combustion F.ngineering, and Westinghouse, and standard BWR fuel manufactured by General Electric and is limited for use to these standard designs. The analyses presented in the SAR are based on non-consolidated, zircaloy-clad fuel with no known or suspected gross breaches. (See Tables 12-la and lb.) The physical parameters that define the mechanical and structural design of the HSM and the DSC are the fuel assembly dimensions and weight. The calculated stresses given in the staff's SER are based on the physical parameters given in Tables 1-la and 1-lb and represent the upper bound. The design basis for nuclear criticality safety is based on the standard Babcock & Wtlcox 15x15/208 pin fuel assemblies with initial enrichments up to 4.0 wt.% U-235, and General Electric 7x7 fuel as.,emblies with initial enrichments up to 4.0 wt.% U-235, for the standardiml NUHOMS-24P and NUHOMS-52B designs, respectively. The HSM is designed to permit storage of irradiated fuel such that the irradiated fuel reactivity is less than or equal to 1.45 wt.% U-235 equivalent unirradiated fuel for the NUHOMS-24P design, and less than or equal to 4.0 wt. 9£ U-235 initial enrichment fuel for the NUHOMS-52B design. The thermal design criterion of the fuel to be stored is that the maximum heat generation rate per assembly be such that the fuel cladding temperature is maintained within established limits during normal and off-normal conditions. Fuel cladding temperature limits were established by the applicant based on methodology in PNL-6189 and PNL-4835 (References 1, 2). Base.d on this methodology, the staff has accepted that a maximum heat generation rate of 1 kW per assembly is a bounding value for the PWR fuel to be stored, and that A-6
0.37 kW per assembly is a bounding value for the BWR fuel to be stored. The radiological design criterion is that the gamma and neutron source strength of the irradiated fuel assemblies must be bounded by values of the neutron and gamma ray source strengths used by the vendor in the shielding analysis. The design basis source strengths were derived from a bumup analysis for (1) PWR fuel with 4.0 weight percent U-235 initial enrichment, irradiated to a maximum of 40,000 MWD/MTU, and a post irradiation time of five years; and (2) BWR fuel with 4.0 weight percent U-235 initial enrichment, irradiated to a maximum of 35,000 MWD/MTU, and a post irradiation time of 5 years. A-7
Table 1-la PWR Fuel Speclflcations of Fuel to be Stored in the Standardized NUHOMS-24P nsc<1> Title or Parameter Specifications Fuel Only intact, unconsolidated PWR fuel assemblies with the following requirements Physical Parameters Assembly Length See SAR Chapter 3 Nominal Cross-Sectional Envelope See SAR Chapter 3 Maximum Assembly Weight See SAR Chapter 3(2) No. of Assemblies per DSC :s; 24 intact assemblies Fuel Cladding Zircaloy-clad fuel with no known or suspected gross cladding breaches Thenna1 Characteristics Decay Heat Power per Fuel Assembly :s; 1.0 kW (this value is maximum for any given assembly, and may not be averaged for all 24 assemblies) Radiological Characteristics Bumup :S:40,000 MWD/MTU Post Irradiation Time ~Syem Maximum Initial Enrichment S4.0 wt. % U-235 Maximum Initial Uranium Content :S472kg/~ly Maximum Initial F.quivalent s 1.45 wt. % U-235(3) Enrichment Neutron Source Per Assembly s2.23E8 n/sec with spectrum bounded by that in Cbaptex 1 of SAR Gamma Source Per Assembly S7.4SE15 photon/sec with spectrum bounded by that in Cbaptex 1 of SAR (1) The limiting fuel specifications listed above must be met by every individual fuel assembly to be stored in the standardized NUHOMS-24P system. Any deviation constitutes an Unanalyzed Condition and Vwlation of the Certificate of Compliance. (2) Design valid for fuel weights up to 762.8 kg (1,682 lb). (3) Determined by the PWR fuel crlticality acceptance curve sllown tn Mgme t .1. A-8
so----------------------------- UNACCEPTABLE 35 30 I- 25 20 j 15 10 UNACCEPTABLE 5 - 0 1.80 2.00 2.40 2..10 INT1ALENRICI-MENT('tW0235 U) 3.20 3.!0 4 00 Figum 12 .1 PWR Fuel Critiolity Accc,pWICO Curve 12-9
Table 1-lb BWR Fuel Specfflcatlons of Fuel to be Stored In the Standardized NUHOMS-52B DSc<1> Title or Parameter Specifications Fuel Only intact, unconsolidated BWR fuel assemblies with the following requirements Physical Parameters Assembly Length See SAR Chapter 3 Nominal Cross-Sectional F.nvelope See SAR Chapter 3 Maximum Assembly Weight See SAR Chapter 3 (w/fuel channels) No. of Assemblies per DSC s 52 intact channeled assemblies Fuel Cladding Zircaloy-clad fuel with no known or suspected gross cladding breaches Thermal Characteristics Decay Heat Power per Fuel Assembly s0.37 kW (this value is maximum for any given assembly, and may not be averaged for all 52 assemblies) Radiological Characteristics Bumup S35,000 MWD/MTU Post Irradiation Time ~s years Maximum Initial F.nrichment S4.0 wt. % U-235 (DSC with 0.75% borated neutron absorber plates) Maximum Initial Uranium Content s; 198 kg/assembly Neutron Source Per Assembly s; 1.01E8 n/sec with spectrum bounded by that in Chapter 7 of SAR Gamma Source Per Assembly ::.2.63B1S photon/sec with spectrum bounded by that in Chapter 7 of SAR (1) The limiting fuel specifications listed above must be met by every individual fuel assembly to be stored in the standardized NUBOMS-52B system. Any deviation constitutes an Unanalyzed Condition and Violation of the Certificate of Compliance. A-10
1.2.2 DSC Vacuum Pressure During Drying Limit/Specification: Vacuum Pressure: S:3 mm Hg Time at Pressure: ~ 30 minutes following stepped evacuation Number of Pump-Downs: 2 Applicability: This is applicable to all DSCs. Objective: To ensure a minimum water content. Action: If the required vacuum pressure cannot be obtained:
- 1. Confirm that the vacuum drying system is properly installed.
- 2. Check and repair, or replace, the vacuum pump.
- 3. Check and repair the system as necessary.
- 4. Check and repair the seal weld between the inner top cover plate and the DSC shell.
Surveillance: No maintenance or tests are required during normal storage. Surveillance of the vacuum gauge is required during the vacuum drying operation. Bases: A stable vacuum pressure of :S 3 mm Hg further ensures that all liquid water has evaporated in the DSC cavity, and that the resulting inventory of oxidizing gases in the DSC is well below the 0.25 volume!>. A-11
1.2.3 DSC Helium Backfill Pres.1Ul'e Limit/Specifications: Helium 2.5 psig +/- 2.5 psig backfill pressure (stable for 30 minutes after filling). Applicability: This speci.tication is applicable to all DSCs. Objective: To ensure that: (1) the atmosphere surrounding the irradiatoo fuel is a non-oxidizing inert gas; ('2) the atmosphere is favorable for the transfer of decay heat. Action: If the required pressure cannot be obtained:
- 1. Confirm that the vacuum drying system and helium source are properly installed.
- 2. Check and repair or replace the pressure gauge.
- 3. Check and repair or replace the vacuum drying system.
- 4. Check and repair or replace the helium source.
- 5. Check and repair the seal weld on DSC top shield plug.
If pressure exceeds the criterion, release a sufficient quantity of helium to lower the DSC cavity pressure. Surveillance: No mainterumce or tests are required during the normal storage. Surveillance of the pressure gauge is required during the helium backfi)ljng opention. Bases: The value of 2.5 psig was selected to ensure that the pressure within the DSC is within the delign limits during any expected normal and off-normal operating conditions. A-12
1.2.4 DSC Helium Leak Rate of Inner Seal Weld Llmit/Specification: s; 1.0 x 104 atm
- cubic centimeters per second (atm
- cm3/s) at the highest DSC limiting pressure.
Applicability: This specification is applicable to the inner top cover plate seal weld of all DSCs. Objective: 1. To limit the total radioactive gases normally released by each canister to negligible Jevels. Should fission gases escape the fuel cladding, they will remain confined by the DSC confinement boundary.
- 2. To retain helium cover gases within the DSC and prevent oxygen from entering the DSC. The helium improves the heat dissipation characteristics of the DSC and prevents any oxidation of fuel cladding.
Action: If the leak rate test of the inner seal weld exceeds 1.0x104 (atm
- cm3/s):
- 1. Check and repair the DSC drain and fill port fittings for lea.ks.
- 2. Check and repair the inner seal weld.
- 3. Check and repair the inner top cover plate for any surface indications resulting in leakage.
Surveillance: After the welding operation has been completed, perform a leak test with a helium leak detection device. Bases: If the DSC leaked at the maximum acceptable rate of 1.0x104 atm
- cm3/s for a period of 20 years, about 63,100 cc of helium would escape from the DSC. This is about 1 % of the 6.3 x 1()6 cm3 of helium initially introduced in the DSC. This amount of leakage would have a negligible effect on the inert environment of the DSC cavity. (
Reference:
American National Standards Institute, ANSI N14.5-1987, "For Radioactive ~-Leakage Tests on Packages for Shipment," Appendix B3). A-13
1.2.5 DSC Dye Penetrant Test of Closure Welds Limit/Specification: All DSC closure welds except those subjected to full volumetric inspection shall be dye penetrant tested in accordance with the requirements of the ASME Boiler and Pressure Vessel Code Section m, Division 1, Article NB-5000 (Reference 8.3 of SAR). The liquid penetrant test acceptance standards shall be those described in Subsection NB-5350 of the Code. Applicability: This is applicable to all DSCs. The welds include inner and outer top and bottom covers, and vent and syphon port covers. Objective: To ensure that the DSC is adequately sealed in a redundant manner and leak tight. . Action: If the liquid penetrant test indicates that the weld is unacceptable:
- 1. The weld shall be repaired in accordance with approved ASME procedures.
- 2. The new weld shall be re-examined in accordance with this specification.
Surveillance: During DSC closure operations. No additional surveillance is required for this operation. Bases: Article NB-5000 Examination, ASME Boiler and Pressure Vessel Code, Section m, Division 1, Sub-Section NB (Reference 8.3 of SAR). A-14
1.2.6 DSC Top End Dose Rates Limit/Specification: Dose rates at the following locations shall be limited to levels which are less than or equal to:
- a. 200 mrem/hr at top shield plug surface at centerline with water in cavity.
- b. 400 mrem/hr at top cover plate surface at centerline without water in cavity.
Applicability: This specification is applicab1e to all DSCs. Objective: The dose rate is limited to this value to ensure that the DSC has not been inadvertently loaded with fuel not meeting the specifications in Section 1.2.1 and to maintain dose rates as low as reasonably achievable during DSC closure operations. Action: a. If specified dose rates are exceeded, the following actions should be taken.:
- 1. Confirm that the spent fuel assemblies placed in DSC conform to the fuel specifications of Section 1.2.1
- 2. Visually inspect placement of top shield plug. Re-install or adjust position of top shield plug if it is not properly seated.
- 3. Install additional temporary shielding.
- b. Submit a letter report to the NRC within 30 days summarizing the action taken and the results of the surveillance, investigation and finding". The report must be submitted using instructions in 10 CFR 71..4 with a copy sent to the administrator of the appropriate NRC regional office.
, Surveillance: Dose rates shall be measured before seal welding the inner top cover plate to the DSC shell and welding the outer top cover plate to the DSC shell. Basis: The basis for this limit is the shielding analysis presented in Section 7.0 of the SAR. A-15
1.2. 7 HSM Dose Rates limit/Specification: Dose rates at the following locations shall be limited to levels which are le88 than or equal to:
- a. 400 mrem/hr at 3 feet from the HSM surface.
- b. Outside of HSM door on center line of DSC 100 mrem/hr.
- c. End shield wail exterior 20 mrem/hr.
Applic.ability: This specification is applicable to all HSMs which contain a loaded DSC. Objective: The dose rate is limited to this value to ensure that the cask (DSC) has not been inadvertently loaded with fuel not meeting the specifications in Sectioo 1.2.1 and to maintain dose rates as-low-as-is-reasonably achievable (Al.ARA) at locations on the HSMs where surveillance is performed, and to reduce off-site exposures during storage. Action: a. If specified dose rates are exceeded, the following actions should be taken:
- 1. Ensure that the DSC is properly positioned on the support rails.
- 2. Ensure proper installation of the HSM door.
- 3. Ensure that the required module spacing is maintained.
- 4. Confirm that the spent fuel assemblies contained in the DSC conform to the specifications of Section 1.2.1.
- 5. Install temporary or pemument shielding to mitigate the dose to acceptable levels in accordance with 10 CFR Part 20, 10 CFR 72.104(a), and Al.ARA.
- b. Submit a letter report to the NRC within 30 days summarizing the action taken and the results of the survcillance, investigation and findings. 1be report must be submitted using instructions in 10 CFR 72.4 with a copy sent to the administrator of the appropriate NRC regional office.
Surveillance: The HSM and ISFSI shall be checked to verify that this specification has been met after the DSC is placed into storage and the HSM door is closed. Basis: The basis for this limit is the shielding analysis presented in Section 7.0 of the SAR. The specified dose rates provide as-low-a.s-is-reasonably-achievable on-site and off-site doses in accordance with 10 CFR Part 20 and 10 CFR 72.104(a). A-16
1.2.8 RSM Maximum Air Exit Temperature Limit/Specification: Following initial DSC transfer to the HSM or the occurrence of accident conditions, the equihlnium air temperature difference between ambient temperature and the vent outlet temperature shall not exceed 100°F for :i:!!S year cooled fuel, when fully loaded with 24 kW heat. Applicability: This specification is applicable to all HSMs stored in the ISFSI. If a DSC is placed in the HSM with a heat load less than 24 kW, the limiting difference between outlet and ambient temperatures shall be determined by a calculation performed by the user using the same methodology and inputs documents in the SAR and SER. Objective: The objective of this limit is to ensure that the temperatures of the fuel cladding and the RSM concrete do not exceed the temperatures calculated in Section 8 of the SAR. That section shows that if the air outlet temperature difference is less than or equal to 100°F (with a thermal heat load of 24 kW), the fuel cladding and concrete will be below the respective temperature limits for normal long-term operation. Action: If the temperature rise is greater than that specified, then the air inlets and exits should be checked for blockage. If the blockage is cleared and the temperature is still greater than that specified, the DSC and HSM cavity may be inspected using video equipment or other suitable me.ans. If environmental factors can be ruled out 88 the cause of excessive temperatures, then the fuel bundles are producing heat at a rate higher than the upper limit specified in Section 3 of the SAR and will require additiooal measurements and analysis to asse&1 the actual penormance of the system. If excessive temperatures cause the system to perform in an unacceptable manner and/or the temperatures cannot be controlled to acceptable limits, then the cask shall be unloaded. Surveillance: The temperature rise shall be measured and recorded daily following DSC insertion until equilibrium temperature is reached, 24 hours after insertion, and again on a daily basis after insertion into the HSM or following the occurre.nce of accident conditions. If the temperature rise is within the specifications or the calculated value for a heat load less than 24 kW, then the HSM and DSC are performing 88 designed to meet this specification and no further maximum air exit temperature measurements are required. Air temperatures must be measured in such a manner 88 to obtain representative values of inlet and outlet air temperatures. Basis: The specified temperature rise is selected to ensure the fuel clad and concrete temperatures are maintained at or below acceptable long-term storage limits. A-17
1.2.9 Transfer Cask Alignment with HSM Limit/Specification: The cask must be aligned with respect to the HSM so that the longitudinal centezline of the DSC in the transfer cask is within
+/- 1/8 inch of its true position when the cask is docked with the HSM front access opening.
Applicability: This specification is applicable during the insertion and retrieval of all DSCs. Objective: To ensure smooth transfer of the DSC from the transfer cask to HSM and back. Action: If the alignment tolerance is exceeded, the following actions should be taken:
- a. Confirm that the transfer system is properly configured.
- b. Check and repair the aJignment equipment.
- c. Confirm the locations of the alignment targets on the transfer cask and HSM.
Surveillance: Before initiating DSC insertion or retrieval, confirm the alignment. Observe the transfer system during DSC insertion or retrieval to ensure that motion or excessive vibration does not occur. Basis: The basis for the true position alignment tolerance is the clearance between the DSC shell, the transfer cask cavity, the HSM access opening, and the DSC support rails inside the HSM. A-Ii
1.2.10 DSC HandUng Height Outside the Spent Fuel Pool Building Limit/Specification: 1. The loaded TC/DSC shall not be handled at a height greater than 80 inches outside the spent fuel pool building.
- 2. In the event of a drop of a loaded TC/DSC from a height greater than 15 inches: (a) fuel in the DSC shall be returned to the reactor spent fuel pool; (b) the DSC shall be removed from service and evaluated for further use; and (c) the TC shall be inspected for damage and evaluated for further use.
Applicability: The specification applies to handling the TC, loaded with the DSC, on route to, and at, the storage pad. Objective: 1. To preclude a loaded TC/DSC drop from a height greater than 80 inches.
- 2. To maintain spent fuel integrity, according to the spent fuel specification for storage, continued confinement integrity, and DSC functional capability, after a tip-over or drop of a loaded DSC from a height greater than 15 inches.
Surveillance: In the event of a loaded TC/DSC drop accident, the system will be returned to the reactor fuel handling building, where, after the fuel has been returned to the spent fuel pool, the DSC and TC will be inspected and evaluated for future use. Basis: The NRC evaluation of the TC/DSC drop analysis concurred that drops up to 80 inches, of the DSC inside the TC, can be sustained without breaching the confinement boundary, preventing removal of spent fuel assemblies, or causing a criticality accident. This specification ensures that handling height limits will not be exceeded in transit to, or at the storage pad. Acceptable damage may occur to the TC, DSC, and the fuel stored in the DSC, for drops of height greater than 15 inches. The specification requiring inspection of the DSC and fuel following a drop of 15 inches or greater ensures that the spent fuel will continue to meet the requirements for storage, the DSC will continue to provide confinement, and the TC will continue to provide its design functions of DSC transfer and shielding, A-19
1.2.11 Transfer Cask Dose Rates Limit/Specification: Dose rates from the transfer cask shall be limited to levels which are less than or equal to:
- a. 200 mrem/hr at 3 feet with water in the DSC cavity.
- b. 500 mrem/hr at 3 feet without water in the DSC cavity.
Applicability: This specification is applicable to the transfer cask containing a loaded DSC. Objective: The dose rate is limited to this value to ensure that the DSC has not been inadvertently loaded with fuel not meeting the specifications in Section 1.2.1 and to maintain dose rates as-low-as-is-reasonably achievable during DSC transfer operations. Action: If specified dose rates are exceeded, place temporary shieldjng around affected areas of transfer cask and review the plant records of the fuel assemblies which have been placed in DSC to ensure they conform to the fuel specifications of Section 1.2.1. Submit a letter report to the NRC within 30 days summarizing the action taken and the results of the surveillance, investigation and findings. The report must be submitted using instructions in 10 CPR 72.4 with a copy sent to the administrator of the appropriate NRC regional office. Surveillance: The dose rates should be measured as soon as possible afu-x the transfer cask is removed from the spent fuel pool Basis: The basis for this limit is the shielding analysis presented in Section 7.0 of the SAR. A-20
1.2.12 Maximum DSC Removable Surface CoPramioation limit/Specification: 2,200 dpm/100 cm2 for beta-gamma sources 220 dpm/100 cm2 for alpha sources. Applicability: This specification is applicable to all DSCs. Objective: To ensure that release of non-fixed conramioation above accepted limits does not occur. Action: If the required limits are not met:
- a. Flush the DSC/transfer cask annulus with demineralized water and repeat surface conramination surveys of the DSC upper surface.
- b. If contamination of the DSC cannot be reduced to an acceptable level by this means, direct surface cleaning techniques shall be used following removal of the fuel assemblies from the DSC and removal of the DSC from the transfer cask.
- c. Check and replace the DSC/transfer cask annulus seal to ensure proper installation and repeat canister loading process.
Surveillance: Following placement of each loaded DSC/transfer cask into the cask decontamination area, fuel pool water above the top shield plug shall be removed and the top region of the DSC and cask shall be deconraminated. A conramination survey of the upper 1 foot of the DSC and cask shall be taken. In addition, contamination. surveys shall be taken on the inside surfaces of the TC after the DSC has been transferred into the HSM. If the above surface conrairnination limit is exceeded, the TC shall be deconramioated, Basis: This non-fixed conrarnioation level is consistent with the requirements of 10 CPR 71.87(i)(l) and 49 CPR 173.443, which regulate the use of spent fuel shipping containers. Consequently, these contamination levels are considered acceptable for exposure to the general environment. This level will also en.sure that contamination levels of the innex surfaces of the HSM and potential releases of radioactive material to the environment are rninirnfaed. A-21
1.2.13 TC/DSC Lifting Heights as a Function of Low Tempemture and Location limit/Speciflcatioo: 1. No lifts or bancf:Ung of the TC/DSC at ~y height are permissible at DSC basket temperatures below -20°F inside the spent fuel pool bulleting.
- 2. The maximum lift height of the TC/DSC shall be 80 inches if the basket temperature is below 0°F but higher than -20°F inside the spent fuel pool bunding.
- 3. No lift height restriction is imposed on the TC/DSC if the basket temperature is bigher than 0°F inside the spent fuel pool building.
- 4. The maximum lift height and haodUng height for all transfer opentions outside the spent fuel pool building shall be 80 inches awl the basket temperature may not be lower than 0°F.
Applicability: These temperature and height limits apply to lifting and transfer of all loaded TC/DSCs inside and outside the spent fuel pool building. The requirements of 10 CFR Part 72 apply outside the spent fuel building. The requirements of 10 CFR Part 50 apply inside the spent fuel pool building. Objective: The low temperature and height limits are imposed to ensure that brittle fracture of the ferritic steels, used in the TC trunnions and shell and in the DSC basket, does not occur during transfer opentions. Action: Confirm the basket temperature before transfer of the TC. If calculation or measurement of this value is available, then the ambient temperature may conservatively be used. Surveillance: The ambient temperature shall be measured before transfer of the TC/DSC. Bases: The basis for the low temperature and height limits is ANSI Nl4.6-1986 paragraph 4.2.6 which requires at least 40°F higher service temperature than nil ductility transition (NOT) temperature for the TC. In the cue of the standardized TC, the test temperature is - 40°F; therefore, although the NDT temperature is not determined, the mate.rial will have the required 40°F margin if the ambient temperature is 0°F or higher. This assumes the material service temperature is equal to the ambient temperature. The basis for the low temperature limit for the DSC is NUREG/CR-1815. The basis for the handling height limits is the NRC evaluation of the structural integrity of the DSC to drop heights of 80 inches and less. A-22
1.2.14 TC/DSC Transfer Operations at High Ambient Temperatures limit/Specification: 1. The ambient temperature for transfer operations of a loaded TC/DSC shall not be greater that 100°F (when cask is exposed to direct insolation).
- 2. For transfer operations when ambient temperatures exceed 100°F up to 125 °F, a solar shield shall be used to provide protection against direct solar radiation.
Applicability: This ambient temperature limit applies to all transfer operations of loaded TC/DSCs outside the spent fuel pool building. Objective: The high temperature limit (100°F) is imposed to ensure that
- 1. The fuel cladding temperature limit is not exceeded,
- 2. The solid neutron shield material temperature limit is not exceeded, and
- 3. The corresponding TC cavity pressure limit is not exceeded.
Action: Confirm what the ambient temperature is and provide appropriate solar shade if ambient temperature is expected to exceed 100°F. Surveillance: The ambient temperature shall be measured before transfer of the TC/DSC. Bases: The basis for the high temperature limit is PNL-6189 (Reference 1) for the fuel clad limit, the manufacturer's specification for neutron shield, and the design basis pressure of the TC internal cavity pressure. A-23
1.2.15 Boron Concentration in the DSC Cavity Water (24-P Design Only) Limit/Specification: -
'Ibe DSC cavity shall be filled only with water having a boron concentration equal to, or greater than 2,000 ppm.
Applicability: This limit applies only to the standardized NUHOMS-24P design. No boration in the cavity water is required for the standardiud NUHOMS-52B system since that system uses fixed absorber plates. Objective: To ensure a subcritical configuration is maintained in the case of accidental loading of the DSC with unirradiated fuel. Action: If the boron concentration is below the required weight percentage concentration (gm boron/106 gm water), add boron and re-sample, and test the concentration until the boron concentration is shown to be greater than that required. Surveillance: Written procedures shall be used to independently determine (two samples analyud by different individuals) the boron concentration in the water used to fill the DSC cavity.
- 1. Within 4 hours before insertion of the first fuel assembly into the DSC, the dissolved boron concentration in water in the spent fuel pool, and in the water that will be introduced in the DSC cavity, shall be independently determined (two samples chemically analyud by two individuals).
- 2. Within 4 hours before flooding the DSC cavity for unloading the fuel assemblies, the dissolved boron concentration in water in the spent pool, and in the water that will be introduced into the DSC cavity, shall be independently determined (two samples anal}'7.ed chemically by two individuals).
- 3. The dissolved boron concentration in the water shall be reconfirmed at intervals not to exceed 48 hours until such time as the DSC is removed from the spent fuel pool or the fuel has been removed from the DSC.
Bases: The required boron concentration is based on the criticality analysis for an accidental misloading of the DSC with unburned fuel, maximum enrichment, and optimum moderation conditions. A*24
1.2.16 Provision of TC Seismic Restraint Inside the Spent Fuel Pool Building as a Function of Horizontal Acceleration and Loaded Cask Weight Limit/Specl:fication: Seismic restraints shall be provided to prevent overturning of a loaded TC during a seismic event if a certificate holder determines that the horimntal acceleration is 0.40 g or greater and the fully loaded TC weight is less than 190 kips. The determination of horimntal acceleration acting at the center of gravity (CG) of the loaded TC must be based on a peak horimntal ground acceleration at the site, but shall not exceed 0.25 g. Applicability: This condition applies to all TCs which are subject to horimntal accelerations of 0.40 g or greater. Objective: To prevent overturning of a loaded TC inside the spent fuel pool building. Action: Determine what the horizontal acceleration is for the TC and determine if the cask weight is less than 190 kips. Surveillance: Determine need for TC restraint before any operations inside the spent fuel pool building. Bases: Calculation of overturning and restoring moments. A-25
1.3 Surveillance and Monitoring The NRC staff is requiring the following surveillance frequency for the HSM. 1.3.1 Vuual Inspection of HSM Air Inlets and Outlets (Front Wall and Roof Birdscreen) limit/Surveillance: A visual surveillance of the exterior of the air inlets and outlets shall be conducted daily. In addition, a close-up inspection shall be performed to ensure that no materials accumulate between the modules to block the air flow. Objective: To ensure that HSM air inlets and outlets are not blocked for more than 40 hours to prevent exceeding the allowable HSM concrete temperature or the fuel cladding temperature. Applicability: This specification is applicable to all HSMs loaded with a DSC loaded with spent fuel. Action: If the surveillance shows blockage of air vents (inlets or outlets), they shall be cleared. If the screen is damaged, it shall be replaced. Basis: The concrete temperature could exceed 350°F in the accident circumstances of complete blockage of all vents if the period exceeds approximately 40 hours. Concrete tanperatures over 350°F in accidents (without the presence of water or steam) can have uncertain impact on concrete strength and durability. A conservati.ve analysis (adiabatic heat case) of complete blockage of all air inlets or outlets indicates that the concrete can reach the accident temperature limit of 350°F in a time period of approximately 40 hours. A-26
1.3.2 HSM Thermal Performance SurveiUance: Verify a temperature measurement of the thermal performance, for each HSM, on a daily basis. The temperature measurement could be any parameter such as (1) a direct measurement of the HSM temperatures, (2) a direct measurement of the DSC temperatures, (3) a comparison of the inlet and outlet temperature difference to predicted temperature differences for e.ach individual HSM, or (4) other means that would identify and allow for the correction of off-normal thermal conditions that could lead to exceeding the concrete and fuel clad temperature criteria. If air temperatures are measured, they must be measured in such a manner as to obtain representative values of inlet and outlet air temperatures. Also due to the proximity of adjacent HSM modules, care must be exercised to ensure that measured air temperatures reflect only the thermal perfonnance of an individual module, and not the combined perfonnance of adjacent modules. Action: If the temperature measurement shows a significant unexplained difference, so as to indicate the approach of materials to the concrete or fuel clad temperature criteria, take appropriate action to determine the cause and return the canister to normal operation. If the measurement or other evidence suggests that the concrete accident temperature criteria (350°F) has been exceeded for more than 24 hours, the HSM must be removed from service unless the licensee can provide test results in accordance with ACI-349, appendix A.4.3, demonstrating that the structural strength of the HSM has an adequate margin of safety. Basis: The temperature measurement should be of sufficient scope to provide the licensee with a positive means to identify conditions which threaten to approach temperature critezia for proper HSM operation and allow for the correction of off-normal thermal conditions that could lend to exceeding the concrete and fuel clad temperature criteria. A-27
Table 1.3.1 Summary of Surveillance and Monitoring Requirements Surveillance or Monitoring Period Reference Section
- 1. Fuel Specification PL 1.2.1
- 2. DSC Vacuum Pressure During Drying L 1.2.2
- 3. DSC Helium Backfill Pressure L 1.2.3
- 4. DSC Helium Leak Rate of Inner Seal L 1.2.4 Weld
- s. DSC Dye Penetrant Test of Closure L 1.2.S Welds
- 6. DSC Top End Dose Rates L 1.2.6
- 7. HSM Dose Rates L 1.2.7
- 8. HSM Maximum Air Exit Temperature 24 hrs 1.2.8
- 9. TC Alignment with HSM s 1.2.9
- 10. DSC Handling Height Outside Spent AN 1.2.10 Fuel Pool Building
- 11. Transfer Cask Dose Rates L 1.2.11
- 12. Maximum DSC Surface Contamination L 1.2.12
- 13. TC/DSC lifting Helgb.ts u a Function L 1.2.13 of Low Temperature and Location Legend PL Prior to loading L During loading and prior to movement to HSM pad 24 hrs Time following DSC insertion into HSM S Prior to movement of DSC to or from HSM AN As necessary D Daily (24 hour frequency)
- A-28
Table 1.3.1 Summary of Survelllance and Monltorin& Requirements (Continued) Surveillance or Monitoring Period Reference Section
- 14. TC/DSC Transfer Operations at High L 1.2.14 Ambient Temperatures
- 15. Boron Concentration in DSC Cavity PL 1.2.15 Water (24-P Design Only)
- 16. Provision of TC Seismic Restraint Inside PL 1.2.16 the Spent Fuel Pool Building as a Function of Horizontal Acceleration and Loaded Cask WeJght
- 17. VlSU81 Inspection of HSM Air Inlets and D 1.3.1 Outlets
- 18. HSM Thermal Performance D 1.3.2 wend PL Prior to loading L During loading and prior to movement to HSM pad 24 hrs Time following DSC insertion into HSM S Prior to movement of DSC to or from HSM AN As necessary D Daily (24 hour frequency)
A-29
References
- 1. Levy, I.S., et al., *Recommended Temperature Limits for Dry Storage of Spent Light Water Reactor Zircaloy-C1ad Fuel Rods in Inert Gaa,
- Pacific Northwest Laboratory Report, PNL-6189, May 1987.
- 2. Johnson, A.B., Jr., and E.R. Gilbert, *Technical Basis for Storage of Zircaloy-Clad Spent Fuel in Inert Gases,* PNL-4835, September 1983.
A-30
T DOCKET NUMBER PR 72 PROPOSED RULE-=-.:.:._..;.__ DOCK ETED USN~C (§ Cj F R2~ °It) [7590-01-Pl
*94 0EC rs P2 :34 NUCL£Ak REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AF02 List of Approved Spent Fuel Storage Casks: Addition AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
SUMMARY
- The Nuclear Regulatory Commission (NRC) is amending its regulations to add the Standardized NUHOMS Horizontal Modular System to the List of Approved Spent Fuel Storage Casks. This amendment allows the holders of power reactor operating licenses to store spent fuel in this approved cask under a general license.
EFFECTIVE DATE: (30 days from date of publication in the Federal Register) ADDRESSES: Copies of the environmental assessment and finding of no significant impact, as well as, the public comments received on the proposed rule are available for inspection and/or copying for a fee at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from the individuals listed under the next heading below.
FOR FURTHER INFORMATION CONTACT: Mr. Gordon E. Gundersen, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Conunission, Washington, DC 20555, telephone (301) 415-6195, or Dr. Edward Y. S. Shum, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Cormnission, Washington, DC 20~55, telephone (301) 415-7903. SUPPLEMENTARY INFORMATION:
Background
Section 218{a) of the Nuclear Waste Policy Act of 1982 {NWPA) includes the following directive: "The Secretary [of the Department of Energy (DOE)] shall establish a demonstration program in cooperation with the private sector, for the dry storage of spent nuclear fuel at civilian nuclear reactor power sites, with the objective of establishing one or more technologies that the [Nuclear Regulatory] Co11111ission may, by rule, approve for use at the sites of civilian nuclear power reactors without, to the maximum extent practicable, the need for additional site-specific approvals by the Commission." After subsequent DOE technical evaluations and based on a full review of all available data, the Co11111ission approved dry storage of spent nuclear fuel in a final rule published in the Federal Register on July 18, 1990 (55 FR 29181). The final rule established a new Subpart K within 10 CFR Part 72 entitled
- General License for Storage of Spent Fuel at Power Reactor Sites."
Irradiated reactor fuel has been handled under dry conditions since the mid-1940s when irradiated fuel examinations began in hot cells. Light-water reactor fuel has been examined dry in hot cells, since approximately 1960. 2
Irradiated reactor fuel has been stored continuously at hot cells under dry conditions for approximately two decades. The NRC's experience with storage of spent fuel in dry casks is extensive as described in the proposed rule to establish 10 CFR Part 72, Subpart K (May 5, 1989; 54 FR 19379). Further, the United States has extensive experience in the licensing and safe operation of independent spent fuel storage installations (ISFSis). By the end of 1994, six site-specific licenses for dry cask storage will have been issued: Virginia Power's Surry Station, issued July 2, 1986; Carolina Power and Light's {CP&l) HB Robinson Station, isJued August 13, 1986; Duke Power's Oconee Station, issued January 29, 1990; Public Service of Colorado's Fort St. Vrain facility, issued November 4, 1991; Baltimore Gas and Electric's {BG&E} Calvert Cliffs Station, issued November 25, 1992; and Northern States Power's (NSP) Prairie Island Nuclear Generating Plant, issued October 19, 1993. All except NSP have colllllenced operation and loaded fuel. At the end of 1994, dry storage spent fuel inventories of these utilities are as follows: 500 assemblies at Virginia Power, 60 assemblies at CP&L, 530 assemblies at Duke Power, 1480 fuel elements at Public Service of Colorado, and 190 assemblies at BG&E. NSP plans to begin storing fuel soon. In May 1993, Consumers Power's Palisades Station commenced operation and loaded fuel under the provisions of the general license in 10 CFR Part 72, Subpart K. At the end of 1994, approximately 168 assemblies are stored at Palisades. As a result of the growing use of dry storage technology, the NRC has gained over 35 staff years of experience in the review and licensing of dry spent fuel storage systems. In addition, the NRC draws upon the knowledge and experience of outside scientists and engineers recognized as experts within their respective fields in the performance of the independent safety analysis 3
of the system and component designs submitted by applicants for dry cask licenses or certification. Reviews of numerous applications seeking site-specific licenses, certificates of compliance, or approvals of topical reports, have been conducted over the past eight years. More recently, the NRC published a notice of proposed rulemaking in the Federal Register on June 2, 1994 (59 FR 28496), which proposed to amend 10 CFR 72.214 to include one additional spent fuel storage cask (i.e., the VECTRA Technologies, Inc., Standardized NUHOMS Horizontal Modular Storage System} on the list of approved - spent fue, storage casks that power reactor licensees may use under the provisions of a general license issued by NRC in accordance with 10 CFR Part 72, Subpart K. The Standardized NUHOMS consists of two systems: (I} The NUHOMS-24P holds 24 specified pressurized-water reactor spent fuel assemblies and (2} The NUHOMS-52B holds 52 specified boiling-water reactor spent fuel assemblies. Subsequent to the expiration of the 75-day public conrnent period on August 16, 1994, NRC received a request, dated August 11, 1994, for a 6-week extension of the conunent period from Connie Kline of the Sierra Club on behalf of 12 citizen groups. The extension request asserted that several proprietary documents related to this rulemaking were not available to the public for approximately 2 weeks at the beginning of the co11111ent period. The NRC granted the request on August 29, 1994 (59 FR 44381} by extending the public comment period to September 30, 1994. VECTRA Technologies, Inc. (formerly Pacific Nuclear Fuel Services, Inc.} submitted to the NRC a Safety Analysis Report (SAR) entitled "Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel,N NUH-003, Revision 2, dated November 1993. 4
Subsequently, VECTRA Technologies, Inc. provided additional information to the NRC related to the SAR. In March 1994, the NRC issued a draft Safety Evaluation Report (SER) entitled "Safety Evaluation Report of Pacific Nuclear Fuel Services, Inc. Safety Analysis Report for the Standardized NUHOHS Horizontal Storage System for Irradiated Nuclear Fuel* approving the SAR. The NRC issued a draft Certificate of Compliance by letter to Mr. Robert D. Quinn from Mr. Frederick C. Sturz dated April 28, 1994. These documents are part of the docket and record that support the proposed rule published in the Federal Register on June 2, 1994 (59 FR 28496). The objective of 10 CFR Part 72 is to protect the public health and safety by providing for the safe confinement of the stored fuel and preventing the degradation of the fuel cladding. The review criteria used by the NRC for review and approval of dry cask storage under 10 CFR Part 72 consider the following factors: siting, design, quality assurance, emergency planning, training, and physical protection of the fuel. Phenomena such as earthquakes, high winds, tornados, tornado driven missiles, lightning, and floods are included in the review of a specific system, either for a certificate of compliance or a site-specific license. In addition, applicants must demonstrate to NRC's satisfaction that their proposed dry cask system will resist man-made events such as explosions, fire, and drop or tipover accidents. 1 Based on further NRC review and analysis of public connents, both the SER and Certificate of Compliance for the Standardized NUHOMS were modified. Section M contains a description of changes to the SER and Certificate of 1 The design bases for these events and accidents are contained within 10 CFR Part 72. 5
Compliance in response to public cormients. The NRC finds that the Standardized NUHOMS, as designed and when fabricated and used in accordance with the conditions specified in its Certificate of Compliance, meets the requirements of 10 CFR Part 72. Thus, use of the Standardized NUHOMS, as approved by the NRC, will provide adequate protection of the public health and safety and the environment. With this final rule, the NRC is approving the use of the Standardized NUHOMS under the general license in 10 CFR Part 72, Subpart K, by holders of power reactor operating licenses under 10 CFR - Part 50. Simultaneously, the NRC is issuing a final Certificate of Compliance to be effective on (30 days from date of publication in the Federal Register). A copy of the Certificate of Compliance is available for public inspection and/or copying for a fee at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC. Public Responses In response to the proposed addition of the Standardized NUHOMS, 239 coments in 27 letters with one supplement were received from individuals, public interest groups, an environmental group, an association, industry representatives, a city, states, and one Federal agency. One commenter withdrew his co11111ents. Many of these letters contained similar coments that have been grouped together and addressed as a single issue. All colllll8nts have been grouped into 15 broad issues designated A though 0. A summary of the coments and an NRC analysis and response to those comments is included for each broad issue. The NRC has identified and responded to 89 separate issues that include the significant points raised by each commenter. 6
A number of comments were related to the disposal of high-level waste and the use of dry cask storage technology in general, rather than to the acceptability of this particular cask. Examples of these coments include:
- The Federal Government's failure to resolve questions about the permanent storage of nuclear waste leaves both the plant and public with limited options: additional storage in pools, additional storage in dry casks, or plant shutdown. The Federal Government has an obligation to resolve the issue of permanent or interim storage. It would be difficult to overstate the need for dispatch in doing so, as hundreds of American communities will eventually face this problem.
- It is not fair to the public of Ohio to link Toledo Edison Company's attempts to continue.the safe storage of its nuclear fuel with insistence by others that the NRC shut down Davis-Besse and every other nuclear plant in the country.
- Only dry storage casks that are compatible with future DOE interim or permanent storage operation, including transportation, should be approved for use under the general license and listed in 10 CFR 72.214.
These coments deal with broad policy and program issues relating to the storage and disposal of high-level radioactive waste, including the DOE's repository program and as such are beyond the scope of this rule. However, there is a summary of relevant information on many of these broad issues in Group G. Many conrnents were directed at the Standardized NUHOMS-24P with only a few co11111ents being specific to the Standardized NUHOMS-52B. Many comenters discussed topics that were not the subject of this rulemaking and thus were not specifically addressed by the NRC staff as a part 7
of this final rule action. These convnents express opposition to the use of dry cask storage and included the following suggestions and topics: (1) Nuclear plants generating radioactive waste should be shut down. (2) The production of radioactive waste should be stopped when the existing spent fuel pool (and off-load-reactor capacity) is full. (3) A formal hearing should be required at each site using dry storage casks. (4) The Davis-Besse plant should be shut down. (5) The use of nuclear power should be stopped and existing sites cleaned up. (6) Palisades experienced problems in using the VSC-24 cask. (7) Alternative forms of power should be used. Finally, many co11111enters expressed concern over the ability of dry cask storage designs, presumably including the Standardized NUHOMS, to store spent fuel safely. The following responses r to these comments reflect a small but important portion of the NRC's review of health, safety, and environmental aspects of the Standardized NUHOMS to ensure that the cask is designed to provide protection of the public health and safety and environment under both normal conditions and severe, unlikely but credible, accident conditions. Dry cask storage systems are massive devices, designed and analyzed to provide shielding from direct exposure to radiation, to confine the spent fuel in a safe storage condition, and to prevent releases of radiation to the environment. They are designed to perform these tasks by relying on passive heat removal and confinement systems without moving parts and with minimal reliance on human intervention to safely fulfill their function for the term of storage. The NRC staff has concluded that the methods of analysis are 8
conservative and assure that the design has appropriate margins of safety under both normal and accident conditions. Analysis of Public Comment A. A number of conunenters raised issues relating to cask handling and the ability of the cask to withstand drop and tipover accidents. A.I. Comment. Several commenters wanted the transfer cask containing the Dry Storage Canister (DSC) to be analyzed for the maximum possible drop, regardless of whether that drop can occur inside or outside the spent fuel building. One conmenter alleged that a drop of the transfer cask into the spent fuel pool would damage fuel assemblies in the pool. Another conmenter was concerned about jamming the transfer cask in the spent fuel pool. What would happen to the cask if jammed fuel could not be extricated? Would the entire 40 ton transfer cask be left in the fuel pool? Response. Use of the Standardized NUHOMS inside the fuel handling building would be conducted in accordance with the 10 CFR Part 50 reactor operating license. These cask handling operations, including loading, retrieval, and training, must be evaluated by the general licensee as required by 10 CFR 72.212(b)(4) to ensure that procedures are clear and can be conducted safely. Load handling activities and possible load drop events with structural and radiological consequences related to transfer cask drops inside the spent fuel building are subject to the provisions of 10 CFR 50.59. Thus, the licensee must determine whether the activities involve any unreviewed facility safety question or any change in facility technical specifications. The transfer cask and DSC designs were evaluated by the NRC against the 9
criteria for controlling heavy loads that are found in NRC's NUREG-0612, 2 "Control of Heavy Loads at Nuclear Power Plants,* and American National Standards Institute (ANSI) N14.6, "Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More." The lifting yoke associated with the transfer cask is a special purpose device designed to ANSI N14.6 criteria to ensure that the yoke can safely lift the wet transfer cask containing the DSC out of the spent fuel pool and can safely lift the dry transfer cask and DSC to the transport trailer. Pursuant to 10 CFR 50.59, for those reactor plants with a shipping cask drop analysis, the licensee must verify that the shipping cask drop analysis adequately describes the consequences of a postulated transfer cask drop and that no unreviewed safety question exists. For those reactor plants that may lack a shipping cask drop analysis, each licensee must perform a transfer cask drop analysis pursuant to 10 CFR 50.59. Specific requirements for lifting the transfer cask are contained in the Certificate of Compliance and SER. However, movement of the transfer cask in the spent fuel pool building must, as required by 10 CFR 72.212(b)(4), be evaluated by the licensee pursuant to 10 CFR 50.59. The possibility of jamming a transfer cask while in the spent fuel pool is one of many issues to be evaluated under 10 CFR 50.59. A.2. Comment. One commenter asked why the transfer cask with the DSC can be lifted to 80 inches outside the spent fuel pool building when it has to 2 Copies of NUREG-0612 and NUREG/CR 1815 may be purchased from the Superintendent of Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington, DC 20402-9328. Copies are also available from the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161. A copy is also available for inspection and copying for a fee in the NRC Public Document Room, 2120 L Street, NW (Lower Level), Washington, DC 20555-0001. 10
be unloaded and inspected for damage if it drops from above 15 inches. Why not limit the height to 15 inches? Response. The transfer cask with the DSC rides on the transport trailer at a height of greater than 15 inches and therefore was analyzed for a drop from that height (80 inches). A drop from a height between 15 and 80 inches does not pose a public health and safety hazard. However, to ensure safety the NRC requires the DSC to be unloaded and inspected for damage. A.3. Coment. One comenter asked about the tipover analysis or drop analysis result. Re~ponse. The tipover, end drops, and horizontal drop analyses form part of the structural design basis for the Standardized NUHOMS design. The designer, VECTRA, described these drop and tipover analyses in SAR, Section 8.2.5. The NRC's evaluation of the vendor's analyses is described in SER, Section 3.2.2.3E. The NRC found the results of these analyses to be satisfactory, because the calculated stresses were all within the allowable criteria of the American Society of Mechanical Engineers (ASHE) Code. A.4. Col1'Ullent. Several comenters, citing Section 1.1.1 of the draft Certificate of Compliance, requested that the postulated cask drop accident in the plant fuel handling area be included in the list of parameters and analyses that will need verification by the system user (for the 10 CFR 50.59 safety evaluation). Response. As stated in Section 1.1.1 of the draft Certificate of Compliance, a holder of a 10 CFR Part 50 license before use of the general license under 10 CFR Part 72, must determine whether activities related to storage of spent fuel involve any unreviewed facility safety issues or changes in facility technical specifications as provided under 10 CFR 50.59. Fuel 11
handling including the possible drop of a spent fuel cask is among the activities that are required to be verified. Fuel handling operations, including spent fuels and fresh fuels, are routine within the nuclear power plant and are subject to NRC regulation under 10 CFR Part 50. A holder of a 10 CFR Part 50 license is required to establish operating procedures for spent fuel handling and to provide emergency planning to address a potential cask drop accident in the reactor's fuel handling area (Certificate of Compliance, Section 1.1.4). Therefore the NRC considers it clear that the spent fuel operation in the nuclear power plant should be evaluated to verify that the possible drop of a spent fuel cask does not raise an unreviewed safety issue or require a facility technical specification change appropriately regulated under 10 CFR Part 50. A.5. Co111T1ent. One coll11lenter stated that there is no place to unload the spent fuel in the event of a canister breach. There is no indication that the canister, the canister lifting mechanism, or the transport mechanism to move the canister into the cask, are nuclear grade equipment or have been designed to prevent a single failure from breaching the canister and circumventing the protection provided by the sole barrier provided by the -canister wall itself. Response. According to 10 CFR 72.122(1), storage systems must be designed to allow ready retrieval of the spent fuel in storage. A general licensee using an NRC-approved cask must maintain the capability to unload a cask. Typically, this will be done by maintaining the capability to unload a cask in the reactor fuel pool. Other options are under consideration that would permit unloading a cask outside the reactor pool. 12
With respect to canister equipment and design, the DSC or canister is designed to the ASHE Boiler and Pressure Vessel Code (BPVC), Section Ill, Subsection NB. The DSC provides a containment boundary for the radioactive material and the cladding of the fuel rods provides confinement of fuel pellets. Only intact fuel assemblies (rods) with no known cladding defects greater than pin holes and hairline cracks are permitted to be stored. This approach assures the structural integrity of the fuel to confine the fuel pellets and its retrivability. ~n the unlikely event of a breach that required the canister to be unloaded, the canister can be returned to the reactor spent fuel pool. Therefore, it is incorrect to assert that there is no place to unload a canister. The Horizontal Storage Module (HSM) is designed to American Concrete Institute (ACI) 349, which is the required code for nuclear structures made of reinforced concrete. The transfer cask is designed according to the ASME BPVC, Section Ill, Subsection NC; ANSI-Nl4.6 for heavy loads; ANSI-50.9 for load combinations; and NUREG/CR 1815 for impact testing. Because the cask itself is required to meet such exacting standards of construction, the transport mechanism and the trailer that move the canister into the HSM are not considered to be important to safety. Therefore, the design that meets industry standards is sufficient. B. A number of convnenters raised issues relating to releases of radioactivity from surface contamination and leakage from the casks under normal and accident conditions. 13
B.l. Cognent. One commenter pointed out that the Certificate of Compliance Surveillance Requirement 1.2.12 does not have a section stating the action that is to be taken when the contamination level in the transfer cask exceeds limits after the DSC has been transferred to the concrete HSM. Response. The Certificate of Compliance Surveillance Requirement in Section 1.2.12 has been modified to clarify that decontamination of the transfer cask is required if the surface contamination limit is exceeded. B.2. Conment. One commenter, who was concerned with the seismic events at the Davis-Besse Nuclear Power Station, stated that a displacement pulse of 60 cm, as observed in the Lander's quake in the Mojave Desert northeast of Los Angeles, would completely destroy the HSM and allow a substantial release of radioactivity from the fuel within. Response. The potential for a seismic event is not the same at every reactor site in the United States. For Davis-Besse, the maximum ground displacement has been calculated to be 3.33 inches (8.46 cm), corresponding to a 0.15g maximum ground acceleration. This is substantially less than the displacement observed in the Lander's quake and appears to be well within the 9 design of the Standardized NUHOMS. Each general licensee using the Standardized NUHOMS, including Davis-Besse is required to document their evaluations to determine that the reactor site parameters, including seismic events, envelope the cask design basis, as specified in its SAR and SER. 8.3. Comment. One convnenter, citing a Wisconsin Public Service Commission draft environmental impact statement (EIS) for Point Beach, asked for an explanation of why NUHOMS and metal casks have a greater potential to spread contamination than the Pacific Sierra Nuclear Associates ventilated storage cask (VSC) system, VSC-24 cask. 14
Response. The specific rationale that forms the basis of the statement in the Wisconsin Public Service Commission's draft EIS for Point Beach was not documented. The decontamination requirements for the two designs are comparable. The VSC-24 DSC is loaded into the ventilated concrete cask (VCC) forming the VSC. The VSC is then transported from inside the reactor auxiliary building to the storage pad. During moving and storage of the VCC, the exterior surface remains clean because it has not been exposed to contamination in the spent fuel pool. The NUHOMS DSC is moved in the transfer - cask from the reactor building to the horizontal storage module in the field. Because the transfer cask has been in the spent fuel pool, it may have small amounts of external contamination that have the potential to spread during transit. However, any potential contamination of this type could not be significant. The NRC requires that the limits for surface contamination, workers' dose, and environmental dose must all be met for the operation of the ISFSI, including during any transfer operations. Each 10 CFR Part 50 licensee must have a radiation protection program to monitor operations to ensure that surface contamination and worker and public exposure to radiation are below acceptable levels and as low as is reasonably achievable (ALARA). Past operation of the NUHOHS shows that the doses are well below all NRC limits. 8.4. Co11111ent. One co11111enter "is concerned that heat generated by fission product decay may provide the driving force, the presence of free moisture in water-logged fuel may, in a non-mechanistic way, provide a transport mechanism for fission product release and the ambient air circulating through the cask concrete structure may provide (an unmonitored) pathway to the biosphere.* One comenter remained concerned about the possibility of insufficient drying of the fuel before placement in the DSC. 15
Another commenter, citing the Battelle Pacific Northwest Laboratory Report PNL-5987 on the removal of moisture from degraded fuel during vacuum drying, contends that the mechanism for free moisture and radionuclide release that pertain in nonnal or upset conditions, such as conditions caused by sabotage, have not been simulated adequately. Response. The DSC is a closed vessel. There is no path available for release of fission products from inside the DSC to the atmosphere. During nonnal operation, the circulating air, as it passes through the HSM and around the outside of the DSC to remove the heat, never comes in contact with fission products and therefore, could not remove these products from the cavity of the DSC. Moreover, design basis accidents under upset conditions were postulated and analyzed in the SAR and SER. These analyses show that the heat generated from fission product decay is not capable of breaching the DSC and could not provide the driving force for a release of radioactivity. Further, it 1s not expected that any significant amount of moisture will remain in the fuel after it is loaded into the DSC. The fuel is dried after it has been loaded into the DSC and the topcover plate seal welded to the DSC shell. The Certificate of Compliance requires two pump-downs to a vacuum pressure of less than 3mm Hg each with a holding time of greater than 30 minutes. A stable vacuum pressure of less than 3mm Hg further assures that all liquid water has evaporated in the DSC cavity. The safeguards issue of radiological sabotage of storage casks has been reviewed previously and assessed in the 1989 proposed rule to add Subparts K and L to 10 CFR Part 72 (54 FR 19379). The NRC has detennined that the Standardized NUHOMS is sufficiently robust such that the effects of a successful attack would have low health consequences and are similar to the 16
results presented in the 1989 proposed rule. (see also response to comment N.l) C. A number of comments were received that focused on monitoring, surveillance, and inspection activities associated with dry cask storage of spent fuel, particularly as they relate to the Standardized NUHOMS. C.l. Comment. One conrnenter stated that there are neither active nor passive systems in place to mitigate barrier breaches, nor are there active radiation monitors that would indicate a breach has occurred. There are no monitored drains and sumps nor are there retention basins. The co11111enter stated that the cask is insufficient to be relied upon for the health and safety of Ohioans. Response. The Certificate of Compliance (Section 1.3) for the Standardized NUHOMS includes surveillance and monitoring requirements that are more than sufficient to detect cask degradation in time to ensure that adequate corrective actions can and will be taken. In addition, radiation monitoring and environmental monitoring programs would detect any radiation leak in excess of NRC limits from an NRC-approved cask. In some instances, the NRC has required continuous monitoring where it is needed to detennine when corrective action needs to be taken. Under a general license, to date, the NRC has accepted continuous pressure monitoring of the inert helium atmosphere as an indicator of acceptable performance of mechanical closure seals for dry spent fuel storage casks. However, the NRC does not consider continuous monitoring for the Standardized NUHOMS double-weld seals to be necessary because: 17
(1) There are no known long-term degradation mechanisms which would cause the seal to fail within the design life of the DSC; and (2) The possibility of corrosion has been included in the design (see SER Section 3.2.2.5). These conditions ensure that the internal helium atmosphere will remain stable. Therefore, an individual continuous monitoring device for each HSM is not necessary. However, the NRC considers that other forms of monitoring, including periodic surveillance, inspection and survey requirements, and application of preexisting radiological environmental monitoring programs of 10 CFR Part 50 during the use of the canisters with seal weld cl:sures can adequately satisfy NRC requirements. With respect to the use of instrumentation and control systems to monitor systems that are important to safety, the user of the Standardized NUHOMS will, as provided in Chapter 14 of the SER and in Section 1.3.2 of the Certificate of Compliance, be required to verify, the cask thermal performance on a daily basis by a temperature measurement, to identify conditions that threaten to approach cask design temperature criteria. The cask user will also be required tG conduct a daily visual surveillance of the cask air inlets and outlets as required by Chapter 12 of the SER and Section 1.3.1 of the Certificate of Compliance. While the HSH and DSC are considered components important to safety, they are not considered operating systems in the same sense as spent fuel pool cooling water systems or ventilation systems that may require other instrumentation and control systems to ensure proper functioning. Due to this passive design, temperature monitoring and surveillance activities are 18
appropriate and sufficient to assure adequate protection of the public health and safety for this design. Because the Standardized NUHOMS DSC is welded closed and has been decontaminated before being placed in a HSH, there is no routine radioactive liquid generation that would require a retention basin or sump. Water entering the storage area has no mechanism of becoming contaminated because the DSC is enclosed within the HSM and is expected to be dried by the heat generated during storage. C.2. Conanent. One conmenter exnressed concern over the possible external corrosion of the stainless steel DSC because of exposure to water over decades. Another col'llllenter expressed concern about corrosion of stainless steel under conditions of indefinite duration, stating that while stainless steel corrodes less rapidly than carbon steel, even the plumbing fixture industry is finding unexpected stainless steel pitting and corrosion under conditions far less intense than those in a DSC. Another col'llllenter stated that the system is not designed for remote inspection of the DSC for corrosion while it is in the HSM and that the only way to inspect the DSC is to return ft to the spent fuel pool. Periodic inspection of the DSC is needed to preclude or identify gradual canister deterioration by unknown mechanisms. Another coamenter inquired about a checking system for the NUHOMS in the future. How will corrosion be evaluated on the canister (DSC) and the support rails inside the HSM? Is it possible for them to accumulate moisture and corrode together over possibly many years of storage? What check is required on the possibility that the canister couldn't be removed at the end of cask life? 19 .
Response. The DSC is enclosed within the HSM and 1s not exposed to external water. Laboratory experiments have indicated a general corrosion rate of less than 0.00001 inches per year for similar stainless steels. The NRC believes these experiments more accurately bound DSC corrosion than experiences 1n unrelated industries. For the SO-year design life of the DSC, the expected corrosion would therefore not result in exceeding a corrosion depth of 0.0005 inches. This will not affect the DSC perfonning its intended safety functions. Because of the low corrosion rates expected for stainless 4t steel, p~riodic inspections for deterioration of the DSC are not considered necessary. Therefore, inspections are not required. The support rails for the DSC have an extremely hard-alloy steel applied to the sliding surface, are ground to a smooth finish, and are coated with a dry film lubricant to prevent corrosion and to reduce the coefficient of friction. Furthermore, the environment inside the HSM is protected from rain and it is kept dry by the heat load from the DSC. Therefore, it is highly unlikely that corrosion between the stainless steel and the hard alloy steel surface of the support rail will occur to any significant extent. These conclusions and analyses regarding the very small likelihood of corrosion indicate that there is reasonable assurance that the DSC can be removed from the HSM when required. C.3. Co111110nt. One co111nenter questioned whether the screens between the casks, which are essential to cooling, will remain clear of debris and how they can be cleaned if they become partially clogged. Another corrrnenter was concerned about how the roof screen was inspected, stating that it seems likely that insects, animals, and birds will be attracted to the warm air coming from the outlet vents. Several conmenters remained concerned about vent blockage that can completely cover and block screening and vents 20
particularly from insects such as paper wasps, that build huge nests, and swanns of midges that are co11DJ1on to the Great Lakes. How are the screens attached? Response. As stated in the Certificate of Compliance, a licensee using the Standardized NUHOMS must conduct a daily visual surveillance of the exterior of air inlets and outlets (front wall and roof bird screen). In addition, the licensee must perform a daily close-up inspection to ensure that no material accumulates between the modules to block the air flow. If the surveillance shows blockage of air vents, the licensee is requ~red to clear the vent blockage by following procedures developed by each user of the Standardized NUHOMS. If the screen is damaged, the licensee must replace the screen. The required daily surveillance and temperature measurements should readily detect blockage of the vents or screens by insects, animals, or birds in a timely manner, leading to the removal of the obstruction before damage occurs from high temperatures. The bird screen is made of stainless steel wire cloth tack-welded to stainless steel strips, which are attached to the 4t HSM with stainless steel wedge anchors. C.4. Co11DJ1ent. One connenter expressed concern about the presence of burrowing and other nuisance animals that have posed problems at other waste sites. Response. Burrowing and other nuisance animals are not expected to pose problems for the Standardized NUHOMS. Because of the robust system design, animals will not be able to get to the radioactive material or cause damage such that water could cause movement of the radioactive material. Burrowing under the concrete pad would not cause damage to safety-related components. Further, large-scale burrowing would likely be detected by the 21
daily surveillance or other activities related to the operation of the storage area. C.5. Comment. One co11111enter wanted additional radiation monitoring because of the calculated higher dose rates over previous NUHOMS designs. The connenter stated that these higher dose rates are not consistent with the objective of maintaining occupational exposures ALARA, and that site-specific applications should provide detailed procedures and plans to meet AlARA guidelines and 10 CFR Part 20 requirements with respect to operation and maintenance. Response. No additional radiation monitoring has been specifically identified or required for the Standardized NUHOMS. However, 10 CFR Parts 20, 50, and 72 require that licensees comply with ALARA. In addition, 10 CFR 72.212(b)(6) requires each licensee to review its radiation protection program to determine that their effectiveness is not decreased by use of the Standardized NUHOMS. Further, 10 CFR 72.212 (b}(9) requires each licensee to conduct storage activities in accordance with appropriate written procedures. If the results of these licensee activities indicate that additional procedures are required then the licensee is required to implement the procedures. C.6. Comment. One commenter was concerned about the optical survey equipment used to align the transfer cask with the HSH before transfer. What checks are made on this optical equipment and what regulations apply? Response. The optical equipment used to align the transfer cask with the HSM is optional and is an operational convenience. However, the licensee must meet Technical Specifications 1.2.9 in the Certificate of 22
Compliance. Therefore, only appropriate calibrations or checks to assure compliance with this technical specification are appropriate. C.7. Co11111ent. One commenter wants to know who evaluates the insertion or retrieval of the DSC for excessive vibration and what is the result of excessive vibration. Would this allow crud to be released? Response. The NRC Certificate of Compliance, Section 1.2.9 provides that the cask user observe the transfer system during DSC insertion or retrieval to ensure that motion or excessive vibration does not occur. It also prescribes certain follow-up actions to be taken by the cask user in the event that alignment tolerances are exceeded and excessive vibration occurs. It is possible that excessive vibration could dislodge crud. However, the crud would be contained within the DSC and would not be released to the atmosphere because the DSC is a sealed vessel. Any opening of the DSC will be under controlled conditions that should safely contain the crud and prevent its release to the environment. C.8. Comment. Several coninenters wanted the NRC to set definite methods for the required surveillance and monitoring of NUHOMS, including the daily temperature measurements, so that data are uniform and standardized for future reference on different modules at different reactor locations. Response. The NRC Certificate of Compliance for the Standardized NUHOMS has required temperature measurements. However, the licensee or vendor has latitude in determining how the performance-based temperature requirements will be met. The NRC is not convinced that the possible benefits of a uniform, but prescriptive, surveillance and monitoring system or technique would outweigh the costs of curtailing the freedom of cask users to design an implementation scheme suited to their individual needs. The collection of 23
unifonn data for possible future use, but without a specific regulatory need could lead to additional exposure to workers, or adversely affect safety without any offsetting benefit. C.9. Comment. One coR111enter asked about the design life of this NUHOMS module and on how this is documented. Will the canister be removed from the concrete module at a specific time and be opened? Response. The design life of the Standardized NUHOMS is 50 years as described in the SAR. The Certificate of Compliance has a 20-year approval period that can be renewed by NRC for another 20 years following a safety reevaluation. It is expected, that at the end of operation, the canister will be removed from the concrete module and will be opened in the spent fuel pool facility or an adequate dry environment alternative. The fuel will be transferred to an NRC-approved shipping cask for off-site transportation and ultimate disposal by the DOE. C.10. Comment. One commenter believed it prudent to monitor temperature and air flow to ensure that temperature excursions are not experienced. Response. NRC believes the required temperature measurement stated in Specification 1.3.2 of the Certificate of Compliance, plus the daily visual inspection of HSM air inlets and outlets, are adequate to ensure that temperature excursions exceeding the design basis are not experienced and to determine when corrective action needs to be taken to maintain safe storage conditions. Therefore, air flow measurements are not required to assure safety. 24
- D. A number of commenters raised technical issues related to the thermal analysis of the Standardized NUHOMS and thermal performance of the system under normal, off-normal, and accident conditions.
D.l. Convnent. Several coD111enters wanted, in the interest of ALARA principles, the capacity for approximately 24 kW heat removal to be verified by using an artificial heat load. One comenter suggested that the NUHOMS be tested with a full heat load at a testing site such as Idaho National Engineering Laboratory (INEL), and not at each reactor site that may load it with a higher heat generation rate fuel. Another coD111enter cited the ALARA philosophy of loading the oldest fuel first even though design basis fuel is on site. Several connnenters wanted deletion of the requirement (a literal interpretation of draft Certificate of Compliance) to calculate the temperature rise for each HSM loaded with canisters producing less than the design limit of 24 kW for the following reasons: (1) Users are not normally provided the vendor's analytical models for this calculation, (2) The 100 °F rise calculated for the design basis maximum heat load ensures that all safety limits are met for concrete and fuel, (3) Because 24 kW is the limit, virtually all the HSMs will be affected, which places an undue burden on the user to "baseline* the predicted delta-T by calculation considering the inherent safety margins of the system, and (4) Technical Specification 1.3.1 ensures that air flow is not blocked so a false measurement of low temperature rise cannot occur. Response. A licensee is not required by NR~ to load the oldest fuel first but, in the interest of ALARA, it may do so. However, each time 25
hotter fuel is loaded up to the maximum allowed in a DSC, the licensee would need to verify the heat removal performance of the system. For fuel producing less heat than the design limits of the system, the heat removal capacity of the system determined by calculation must be verified by temperature measurements. This process must be repeated each time a DSC is loaded with hotter fuel until the maximum-system designed heat load is reached. When loaded with spent fuel producing 24 kW heat, the system may not have an ambient and vent outlet temperature difference of more than 100 °F for fuel cooled equal to or more than 5 years. This verification process is required to confirm that the as-built system of each licensee is performi~g as designed. A licensee could use an artificial heat source to test an initial cask at a bounding heat load of 24 kW before loading fuel. However, this test would only verify the spent fuel heat removal capacity of the system. It would not verify as-built performance. Experience has shown that adequate verification testing can be performed at the reactor site. Therefore, performing the verification at a testing site like INEL would not provide additional safety margins. D.2. Corrment. Several connnenters pointed out possible conflicting statements about temperature measurements in the surveillance requirements. In discussions about the heat removal capacity test, temperatures are detenuined only during the test period. Daily temperature measurements on each HSM are required to verify thermal performance. Response. These two temperature measurement programs have different objectives. Temperature measurements by licensees to verify the heat capacity calculations need only be done until equilibrium is reached. The daily temperature measurements by licensees are intended to demonstrate 26
continued safe operation within specified limits over the life of the HSM and may not be the same type of measurement done in the initial period to verify heat removal capacity. 0.3. Comment. One co11111enter was concerned about the adequacy of cooling under all atmospheric conditions in the country. The connenter cited conditions such as humidity over 90 percent, temperature over 100 °F, and no wind. Response. Regulatory requirements for general licensee users of dry storage casks are contained in 10 rFR 72.212(b). Each user must verify that the follow;ng conditions are not exceeded at their reactor site for the Standardized NUHOMS: the maximum average yearly temperature with solar incidence is 70 °F; the average daily temperature is 100 °F; and the maximum temperature is 125 °F with incident solar radiation. If the power reactor site high temperature parameters fall within these criteria, the Standardized NUHOMS can be safely used at the site. D.4. Co11111ent. One commenter wants the NRC to establish procedures to measure temperature performance, especially the thermal performance of an individual module dnd not the combined performance of adjacent modules as stated on page A-23 of the draft Certificate of Compliance. Response. As required by the regulations, the licensees are required to develop detailed procedures. NRC in its regulatory oversite role has the opportunity to review the adequacy of the procedures. The requirement cited by the co11111enter is a requirement for the licensee to verify a temperature measurement of the thermal performance for each HSM, not the combined performance of adjacent modules. A cautionary statement is included in the basis of the specification to ensure that licensee measurements of air 27
temperatures reflect only the thermal performance of an individual module and not the combined performance of adjacent modules. D.S. Comment. One coD1J1enter wanted to know how the temperature differences in the roof, side wall, and floor areas are incorporated into the daily temperature measurement. Response. For the first HSH to be emplaced, the user is required to measure the air inlet and air outlet temperature difference of the system at equilibrium. This measurement is to ensure that the heat capacity of the system will not be exceeded and that the concrete temperature criteria will not be exceeded. For the Standardized NUHOMS, this maximum heat capacity is 24 kW. The 24 kW heat load is the design maximum and is the basis for the thermal hydraulic calculations for the cask. The temperature distribution for various parts of the HSM have been calculated (i.e., the roof, walls, and floor) by the cask vendor. Temperature differences causing thermal stresses in the concrete were evaluated and are duly reported in both the SAR and SER. These calculations were reviewed by NRC as a part of the overall process for this design approval. D.6. Comment. One com1enter stated that daily temperature measurements are not necessary to ensure convective air flow, given the requirement to verify that the inlets and outlets are not obstructed. Site-specific NUHOMS require temperature measurements when the DSC is placed into the HSM, 24 hours later, and again at 1 week after loading to ensure adequate thermal performance. Response. The NRC disagrees with this comment. The HSM and DSC are considered components important to safety in the Standardized NUHOMS. Daily temperature measurements of the thermal performance by the licensee are 28
required to provide additional assurance that thermal limits are not exceeded under the general license. This requirement was imposed on the first cask of this type approved by the NRC and listed in 10 CFR 72.214 for use by a general licensee, the VSC-24 cask (58 FR 17967; April 7, 1994) and is now applied to the Standardized NUHOMS. E. A number of co111Denters expressed concern about emergency planning and response contingencies.
- E.1. Corrment. Several co1T111enters expressed concern that in the event of problems and the need to off-load fuel (as in the recent situation at Palisades), a transfer cask may not be available in a timely manner because of inclement weather or because the transfer cask itself has experienced problems or is being used elsewhere. One commenter expressed concern at having to have a transfer cask on site,within 40 hours of vent blockage to prevent concrete damage. If the transfer cask is leased from VECTRA and is not at the licensee's site, who is liable if something happens that would require the use of a transfer cask?
Response. The NRC has analyzed all design basis accidents from the operation of an ISFSI and concluded that there will be no release of radioactive material to the environment. The 40-hour limit on vent blockage is intended to prevent concrete degradation that might occur over a long period of storage. A vent blockage accident would not result in the release of radioactive material because the DSC would not be breached. Therefore, the NRC believes that the potential risk to the public health and safety is extremely small during the time needed to obtain the use of a transfer cask. 29
Thus, there is no requirement that a transfer cask be at an ISFSI site all the time. E.2. Co11111ent. One co11111enter expressed concern that the effects of tornado winds and missiles during movement of the fuel in a transfer cask or in a storage cask on a transporter were not analyzed. Response. Both the vendor's SAR and NRC staff's SER address the effects of tornado winds and missiles during movement of the transfer cask with a loaded canister. These analyses show that, for tornado winds, there is a safety factor of 1.5 against overturning when subjected to Design Basis Tornado winds (a safety factor greater than 1 will generally be adequate for public protection}. The transfer cask stability, tornado missile penetration resistance, and shell and end plate stresses were calculated and shown to be below the allowable stresses for ASHE BPVC Service Level D (accident) stresses. E.3. Comment. One commenter described an October 1972 storm that flooded the entire Davis-Besse plant site, including the (pre-operational) reactor building. There has been subsequent flooding of the site, particularly during spring thaws. Response. Safety analyses by NRC and the cask vendor show the Standardized NUHOMS can withstand floods and will continue to perform acceptably. With regard to the Davis-Besse site, the licensee changed site topography during plant construction. Specifically, the area was built up and some dikes were added. The plant structure's ground floor elevation is 585 feet International Great Lakes Datum (IGLD), which is also the elevation of the pad. The licensing design basis for maximum probable static water level on the site is 583.7 feet IGLD. As noted, the HSM and DSC were 30
evaluated for flood conditions as required by 10 CFR 72.122(b). The HSM can withstand a maximum water velocity of 15 feet per second and a static head of 50 feet of water. The DSC can withstand a static head of 50 feet of water. Any site that intends to use a Standardized NUHOMS design must evaluate the conditions at their site to verify compatibility with the design specifications of the system. F. A number of coD111enters raised issues relating to the design, evaluation, and operation of the Standardized NUHOMS. F.1 Comment. Several comments related to the fuel to be stored in the Standardized NUHOMS. One co1T111enter wanted control components contained in assemblies addressed in the SAR and SER citing DOE acceptance criteria. One commenter questioned how 55,000 MWO/MTU burnup fuel now being used in pressurized water reactors will be handled since the Standardized NUHOMS-24 is rated to handle only 40,000 MWD/MTU burnup fuel. Another co11111enter, citing provisions of current site-specific licensees for other NUHOMS designs, stated that higher burnup should be allowed if the decay heat and radiological source terms are within limits. Another co11111enter asserted that increased fission products from higher enriched fuel may potentially increase embrittlement of the fuel cladding and that this needs to be evaluated in the SER. This commenter further alleged that this would increase the probability of more defective fuel being loaded into dry casks. Response. The vendor designed the cask system for storage of pressurized water or boiling water reactor fuel assemblies meeting certain specifications. By limiting the use of the cask system to assemblies meeting these specifications, the vendor made a decision that may partially restrict 31
the use of the cask. However, the NRC does not require that a cask be universal for all types of fuel or be usable at every reactor site. For example, none of the casks previously listed in 10 CFR 72.214 is usable for boiling water reactor spent fuel. Currently, the 55,000 MWD/MTU burnup fuel and fuel with initial enrichments of greater than 4% will have to remain in the spent fuel pool because dry spent fuel cask designs to store fuel with this higher burnup and initial enrichment or related to DOE acceptance criteria have not yet been reviewed and evaluated by the NRC. F.2 Co0111ent. Several comments were related to criticality safety analysis. One cormnenter questioned the conservatism of using 7.5-year cooled spent fuel when 5-year-cooled fuel is the minimum specified and when older fuel may also be stored in the cask. Another inquired about criticality safety if the original basket geometry were compromised, as might be the case for brittle failure of a spacer disk. In the compromised basket geometry case, the co11111enter also asked about the difference in criticality safety for a helium atmosphere rather than a borated water medium. The commenter, - referring to July 24, 1992, meeting minutes, inquired why all parties agreed not to spend any resources to make these criticality safety calculations. Response. The Standardized NUHOMS nuclear criticality safety analysis is based on the following: (1) Babcock and Wilcox 15 x 15/208 pin fuel assemblies with initial enrichments up to 4.0 wt% of U-235 and (2) General Electric 7 x 7 fuel assemblies with initial enrichments up to 4.0 wt% of U-235, for the Standardized NUHOMS-24P and NUHOMS-52B designs respectively. The age of the fuel that will actually be stored is not relevant in 32
criticality safety analysis because the analysis assumes storage of unirradiated fresh fuel that is more reactive than cooled spent fuel. The Standardized NUHOMS-24P syste~ has administrative controls that limit the irradiated fuel reactivity to less than or equal to 1.45 wt% of U-235 equivalent unirradiated fuel (Certificate of Compliance Section 1.2.1}. The possibility of a criticality accident caused by the brittle failure of the basket should not be a significant concern. No lifting or handling of the DSC outside the spent fuel pool building is permitted if the basket temperature is lower than o*F. If the user does not determine the actual basket temperature, the ambient temperature must be used conservatively. Under these temperature restrictions, the basket materials will not behave in a brittle fashion. Consequently, the basket geometry would not be compromised by brittle failure. As for the criticality safety consideration related to a helium atmosphere versus a borated water medium, the k._ 1 , of the fuel in a helium atmosphere is much less than the k.,, in borated water. Therefore, criticality calculations for the borated water are sufficient because they are more conservative and therefore would bound calculations using a helium atmosphere. F.3 Co11111ent. Two coU1J1enters were concerned with shielding and dose assessments for the Standardized NUHOMS. One co11111enter believed that using IO-year-cooled fuel for the dose assessment was nonconservative when 5-year-cooled fuel is needed to load the DSC to produce 24 kW of heat. Another, referring to an NRC meeting with Pacific Nuclear Fuel Services, Inc. (PNFSI), wanted clarification of an NRC request to delete a clause allowing the utility to perform site-specific shielding calculations. 33
Response. The cask vendor presented dose assessment results in the SAR for both 5- and IO-year-cooled fuel. However, for this rulemaking, NRC used the dose assessment for 5-year-cooled fuel for the shielding analysis radiation source term and for accidental releases of radionuclide material. NRC's use of the 5-year-cooled fuel assessment is conservative ~nd bounding. To ensure safe storage of spent nuclear fuel in NRC-approved casks, the NRC specifies, in Section 1.2.1 of the Certificate of Compliance a number of fuel acceptance parameters. These parameters, which may include burnup, initial enrichment, heat load, cooling time, and radiological source term, define the properties of those assemblies that can be stored in a cask. One such parameter of interest for the Standardized NUHOMS is the radiological source term that forms the basis of the shielding analyses. For this parameter, the vendor proposed an alternative approach. Specifically, for fuel assemblies that fall outside the specified source tenn parameters but satisfy all other parameters, the vendor proposed to allow licensees to do individual cask shielding calculations to show compliance with the design basis dose rates. This could result in more assemblies in a licensee's inventory that wo~ld be eligible for dry storage. In the instance noted in the comment, the NRC did not agree with the vendor proposal. The Certificate of Compliance dose rate specifications provide a ,simple check to ensure that DSCs are not inadvertently loaded with the wrong fuel. The dose rate specifications are based on the shielding analyses provided by the vendor in its SAR. Because of differences in non-fuel components in the ends of some assemblies, dose rates higher than those evaluated by NRC in the SER may occur at the ends of casks than were assumed in the shielding analysis. The Certificate of Compliance specifications allow for this possibility and permit 34
the licensee to store such fuel provided the licensee verifies proper cask fabrication, conformance with all other fuel parameters, and compliance with radiation protection requirements. The site-specific calculations referred to in the comment are not shielding calculations, but rather are the licensee's written evaluations (or dose assessments) required by 10 CFR 72.212(b)(2)(iii) to establish that the radiation criteria for ISFSI in 10 CFR 72.104 have been met. The Certificate of Compliance also requires that the licensee submit a letter report to the NRC su11111arizing its actions in this type of case. F.4 Comment. Several commenter~ were concerned with fuel clad integrity issue~. Particularly, they were concerned with potential problems that may arise because of differences between vertical and horizontal storage. One co11111enter noted that it was essential to inspect the cladding carefully for the minute hairline cracks which would allow the radioactivity inside to escape. Another commenter wanted it made clear that for fuel to be eligible for storage it doesn't need to be specifically inspected nor require special handling or storage provisions within the spent fuel pool. The co11111enter also asserted that pinhole leaks in fuel rod cladding do not constitute gross breaches. The comenter wanted fuel cladding integrity clarified. Another commenter claimed that horizontal storage of fuel rods will lead to cladding deterioration that would challenge the technical specifications of the NUHOMS cask. Another commenter was concerned about the possibility of fuel rod bowing that could result in weighted contact between the fuel cladding/crud and the DSC guide sleeve, with the potential for eventual bonding of the materials over the duration of the storage period. One commenter, noting that some of the fuel in the spent fuel pools could be nearly 20 years old, was concerned that the fuel will not be tested for leaks using specific techniques 35
such as penetrating dyes, eddy current, sipping, or ultrasound before canister loading. A coD1Denter wanted all fuel with known defects and all water-logged fuel retained in the spent fuel pool until the cask integrity under operating conditions is fully demonstrated. Another wanted to know how *grossly breached* fuel will be ultimately handled and shipped off site. Response. In the Standardized NUHOMS, PWR fuel rods are stored in a horizontal orientation and do not normally deflect in the middle of any span so that the rods contact the DSC guide sleeve. However, the possibility exists +~at a bowed rod may come in contact with the guide sleeve. With respect to storage of BWR fuel, the fuel channel that surrounds the fuel bundle (rods) provides a barrier to separate coolant flow paths, to guide the control rod, and to provide rigidity and protection for the fuel bundle during handling. Therefore, the BWR fuel rods inside the channel do not come in contact with the guide sleeves. Even if there were contact with either PWR or BWR fuel rods, the interaction would not present a significant concern because the guide sleeve material is stainless steel, which has a very low rate of corrosion, and the DSC cavity is evacuated and back-filled with inert helium, which further reduces the likelihood of any corrosion or bonding involving the guide sleeve and fuel rods. The Certificate of Compliance requires that the fuel have no known or suspected gross cladding breaches to ensure the structural integrity of the fuel. Known or suspected failed fuel assemblies {rods) and fuel with cladding defects greater than pin holes and hairline cracks are not authorized in the Standardized NUHOMS. Fuel meeting this specification will be safely stored and will remain intact in storage because the dry inert atmosphere and 36
relatively low temperature will prevent deterioration of the cladding. Grossly breached fuel will be handled in site-specific license applications. F.5. Comment. Quite a few comments related to the structural stability of the HSH, particularly its response to earthquakes. Comenters questioned the possibility of vertical storage of the Standardized NUHOMS jnd suggested that it would be very difficult to restrain the HSM if the DSC were in a vertical position. One coirmenter wanted dry storage casks constructed to Building Officials Code Administrators (BOCA) National Building Code (and Ohio - Administrative Code) for structures in use group H-4, high haz1rd use, which includes radioactive materials. Coirmenters questioned whether ground acceleration as used by the NRC in its evaluation could adequately describe all potential earthquakes east of the Rocky Mountain Front and suggested that a ground acceleration of 2.Sg would not be realistic for all sites, despite proximity to fault lines. Another commenter alleged a number of seismic events in the midwest which had some effect in the Ohio area could cause a complete failure of the cask and requested that the NRC insist that the cask, containment structure, and foundation pad be designed to substantially exceed all earthquakes with a potential for 0.60g. One commenter wanted to know if the module had been analyzed for earthquake events at all United States reactor sites, according to Laurand Findmun Seismic Hazard Curves. Other commenters expressed various concerns about the integrity and reaction of the Standardized NUHOMS components under earthquake-conditions and asked the following questions: Could the casks crash against each other as the ground moves beneath them? Could the module shift, crack, or move off the pad? 37
How are the rail support holdings evaluated? Could the DSC be knocked off the rails? and Could the module roof crack and fall on the canister? Response. The Standardized NUHOMS design described in the vendor's applications for approval and the SAR does not address vertical storage. Consequently, NRC neither evaluated nor approved vertical storage for the system. Therefore, it may not be stored vertically. The NRC reviewed the Standardized NUHOMS for compliance with design criteria that are more stringent than those of the BOCA National Building Code (NBC) (see response to Conunent A.5). These more stringent criteria are included in national standards that more closely represent the use of the Standardized NUHOMS. Part 72 specifies a design basis maximum ground acceleration of 0.25g for areas east of the Rocky Mountain Front that are not in areas of known seismic activity. All HSMs and DSCs are designed to withstand a 0.25g earthquake. Any reactor licensee who intends to use the Standardized NUHOMS , must verify that the maximum displacements at the cask's location on the reactor site are within the design criteria for the system. The Standardized NUHOHS is free standing and not dependent on the pad for safety. Failure of the pad caused by seismic events will not cause the Standardized NUHOHS to fail. Therefore, cask safety does not require the pad to be designed to withstand a seismic event. F.6. Conanent. One corrmenter stated that the SAR did not include consideration of the accident events such as: aircraft crashes, turbine missiles, external fires, explosions, and sabotage. 38
Response. Before using the Standardized NUHOMS, the general licensee must evaluate them to ensure the site is encompassed by the design bases of the approved cask. The events listed in the coll'lllent are among the site-specific considerations that must be evaluated. Toe site evaluation for a nuclear plant considers the effects of nearby transportation and military activities. It is incumbent upon the user of the cask to detennine if the SER for the facility encompasses the design basis analysis perfonned for the Standardized NUHOMS or any certified cask. The great majority of the aircraft are single-engine propeller airplanes which typically weigh on the order of 1,500 to 2,000 pounds. The cask's inherent design will withstand tornado missiles and other design loads and also provides protection from the collision forces imposed by these light general aviation aircraft without adverse consequences. NUREG-800, Section 3.5.1.6
*standard Review Plan for Light Water Reactors." contains methods and acceptance criteria for determining if the probability of an accident involving larger aircraft (both Military and civilian) exceeds the acceptable criterion. It is incumbent upon the licensee to determine whether or not the
- reactor site parameters are enveloped by the cask design basis as required by 10 CFR 72.212(b}(3}. These would include an evaluation demonstrating that the requirements of 10 CFR 72.106 have been met. Turbine missile analyses typically snow a very low probability of a turbine missile breaking the turbine casing. The site's turbine missile analyses must be considered as part of the facility's analysis of the suitability of the storage location. External fires are handled by established fire control programs. Explosions are prevented by control of combustibles under the licensee's fire protection program. Sabotage is 39
considered under the criteria for security programs that each licensee must implement. (see also response to comment N.l). F.7. Comment. Several commenters raised issues about the pad and foundation for the Standardized NUHOMS. One conmenter referred to a previous rulemaking that stated that the NUHOMS casks required site-specific approvals because they are constructed in place. Other corrrnenters, concerned with seismic events at the Davis-Besse Nuclear Power Station and soil stability issues similar to cask use at the Palisades Plant, asserted that there was a - necessary relationship of the Standardized NUHOMS cask or module to the pad at a specific site and that evaluation of it could not be based on the reactor site seismic analysis. Each site required singular seismic and soil analysis for dynamic loads and not just static loads. Response. The NUHOMS design referred to in the July 18, 1990, 55 FR 29181, rulemaking includes the site-specific pad as an integral part of the concrete HSM and therefore it is important to safety. The Standardized NUHOMS considered in this rulemaking have the HSMs as free-standing units; that is, they have no structural connections to the pad. The Standardized NUHOMS does not rely on the pad to perform a safety function to protect public health and safety. The vendor analyzed the HSM containing the DSC for peak ground accelerations of 0.25g caused by earthquakes and found that it would neither slide nor overturn. NRC evaluated the Standardized NUHOMS under a wide range of site conditions that could diminish cask safety. Further, under the NRC general license, before using the Standardized NUHOMS a licensee must verify that reactor site parameters are within the envelope of conditions reviewed by NRC for the cask approval. If potential conditions exist at the reactor site (including potential erosion, soil instability, or earthquakes) 40
that could unacceptably diminish cask safety by any credible means, the licensee's analysis must include an evaluation of the potential conditions to verify that impairment of cask safety is highly unlikely. The NRC's regulations do not explicitly require a licensee using a cask under a general license to evaluate the cask storage pad and foundation under such site conditions for erosion or earthquakes. If conditions at the reactor site could unacceptably diminish cask safety by affecting the stability of the supporting foundation so as to put the cask in an unsafe condition, the cask may not be used unless the foundation is appropriately modified or a suitable location at the reactor site is found. Implicitly, therefore, the pad and the underlying foundation materials must be analyzed under site conditions that include erosion, soil instability, and earthquakes, even though the pad has no direct safety function and the cask is designed to retain its integrity even assuming the occurrence of a range of site conditions. The licensee has the responsibility under the general license to evaluate the match between reactor site parameters and the range of site conditions (i.e., the envelope) reviewed by NRC for an approved cask. Typically, the licensee will have a substantial amount of information already assembled in the Final Safety Analysis Report (FSAR) for the nuclear reactor. In addition, the envelope for the approved cask is identified in the NRC SER and Certificate of Compliance and in the cask vendor's SAR for the cask. Of course, the licensee should consider whether the envelope evaluated by NRC adequately encompasses the actual location of the cask at the reactor site. The licensee should also consider whether there are any site conditions associated with the actual cask location that could affect cask design and that were not evaluated in the NRC safety evaluation for the cask. 41
The vendor analyzed the DSC and the HSM for rigid body response {i.e., sliding and overturning) to seismic accelerations. The resultant peak horizontal ground acceleration is 0.37g and the peak vertical acceleration is 0.17g. The margin of safety against sliding is 1.35. Similarly, the design seismic force will. not cause the HSM to tip over because the stabilizing moment of the HSM is greater than the seismic overturning moment. The margin of safety against overturning is 1.26. Thus, no sliding or overturning of the HSM or DSC will occur from the design earthquake. Because the pad is not considered a safety-related item, a specific pad design is not being approved in this rulemaking for the Standardized NUHOMS. F.8. Conment. A few cornmenters had questions pertaining to the operation of and procedures for the Standardized NUHOMS. One commenter inquired whether just one module of the Standardized NUHOMS could be purchased by a utility, or whatever number of modules desired could be procured and easily added like singular casks. One commenter expressed concern about snow removal procedures to prevent blockage of the bottom vents by drifting snow. Another commenter wanted NRC to establish a procedure and criteria for dose - rates discussed on pages A-15 and A-16 in the draft Certificate of Compliance. Several commenters noted that a procedure for opening a storage cask and removing the fuel has not been tried before nor documented in the rulemaking. They were also concerned that unloading of a cask would place workers at higher risk. Response. The NRC Certificate of Compliance does not permit or limit the number of NUHOMS modules that may be purchased by a general license. The NRC does not regulate the commercial arrangements between the cask vendor and 42
the users including any provisions on the number of casks that can be purchased or added to the Standardized NUHOMS. Under the Certificate of Compliance, Section 1.3, the user of the Standardized NUHOMS (general licensee) is required to conduct a visual surveillance of the exterior of air inlets and outlets. If the surveillance shows blockage of air vents, they must be cleaned in accordance with proper procedures. These procedures will minimize the potential impact to the health and safety of workers. The daily temperature measurements indicate proper - thermal performance. The Certificate of Compliance requires each licensee to develop procedures to implement the dose criteria prescribed on pages A-15 and A-16. On page A-15 of the Certificate of Compliance, Section 1.26, the dose rate criteria to be met is equal to or less than: (a) 200 rnrem/hr at the top shield plug surface at centerline with water in the cavity; and (b) 400 mrem/hr at the top cover plate surface at centerline without water in the cavity. On page A-16 of the Certificate of Compliance the dose rate criteria - is less than or equal to: (a) 400 mrem/hr. at 3 feet from the HSM surface; (b) 100 mrem/hr. outside of the HSM door on center line of the DSC; and (c) 20 mrem/hr. at the end shield wall exterior. Each licensee is required to develop its own procedures to implement these criteria. In addition, each licensee must develop operational procedures for the ISFSI for workers' radiation exposure to be ALARA. For the Standardized NUHOMS, removal of spent fuel from the DSC is addressed in Chapter 5 of the SAR and in Chapter 11 of the SER. The process is essentially the reverse of loading operations and would be performed under the reactor license radiation protection program. The Certificate of 43
Compliance requires each user to develop written procedures for these operations and includes precautions to be considered for unloading. ALARA is required to be addressed by 10 CFR Part 20. Specification 1.1.6 of the Certificate of Compliance requires that pre-operational testing and training exercises include the opening of a DSC and returning the DSC and transfer cask to the spent fuel pool. The Certificate of Compliance also requires the training program to include off-normal events. F.9. Conrnent. One coD111enter, citing the May 1993 study prepared for the NRC by the Center for Nuclear Waste Regulatory Analyses of San Antonio, Texas, questioned the relatively higher temperature consequences of dry storage on fuel cladding. The report states that, *the dry environment has the potential of producing such problems as further fuel cladding oxidation, increased cladding stresses, and creep deformation as a result of rod internal pressure *** These possible spent fuel and cladding alteration modes could be quite accelerated under dry storage conditions, since temperatures are much higher than in wet storage.* The commenter does not believe that NRC is fulfilling its obligation in 10 CFR 72.122(h) to see that *spent fuel cladding must be protected during storage against degradation that leads to gross rupture.* Response. The Hay 1993 study addresses the long-tenn geological disposal of high-level waste (spent fuel) and is not directed to the short-term interim storage of spent fuel at nuclear power plants. The report evaluates processes over 10,000 years of repository performance for geological disposal. The conclusions of the report are not applicable for the interim storage period of a 20-year cask certificate during which spent fuels stored in the DSC have to meet the NRC's criteria to ensure that cladding is 44
protected. Under normal operation of the ISFSI, leakage of radionuclides is not expected to occur. The design and the double-seal welding of the DSC covers are checked and tested to provide structural integrity throughout the approved storage period. During nonnal storage conditions, the licensee is required to conduct a radiation monitoring program to ensure protection of workers and the safety of the general public. G. A number of co11111ents were related to broad policy and program issues in connection with the storage and disposal of high-level radioactive waste, including the DOE repository program. Some co11111enters questioned the use of dry cask storage and technology in general. Some commenters stated that only dry storage casks that would be compatible with DOE interim or final repository operations, including transportation, should be approved for use under a general license. G.l. Connnent. *one commenter does not want* any more casks approved until a pennanent Federal repository is opened. The wet fuel pool is a proven technology that has been successful in containing radioactivity. Another commenter stated that dry storage is dangerous. Response. The NRC, in implementing the Nuclear Waste Policy Act of 1982, has an obligation to review dry storage technologi.es and to detennine whether to approve the use of these technologies for the storage of spent fuel if they meet applicable safety requirements. The July 18, 1990; 55 FR 29181, rul emaking found that spent fue.l stored in dry storage casks designed to meet the NRC regulatory requirements can safely contain radioactivity. This rulemaking adds one cask design that meets the safety requirements previously developed to the list of approved casks. The previous responses to comments, 45
as well as the detailed safety and environmental analyses underlying this rulemaking (and described elsewhere in this notice), all reveal that the Standardized NUHOMS will conform to the NRC requirements and that its use should not pose the potential for significant environmental impacts. DOE is required by the Nuclear Waste Policy Act of 1982 to accept spent fuel for ultimate disposal. Moreover, the Comission made a generic determination in its waste Compliance Decisions (September 18, 1990; 55 FR 38474 and August 31, 1994; 49 FR 34658) that safe disposal is technically feasible and will be available within the first quarter of the 21st century. Dry cask storage has significant advantages over wet storage in that the system is passive and requires minimal human intervention. No pumps, filters, or water quality monitoring are needed to maintain the conditions necessary for wet storage. The only monitoring required for the Standardized NUHOMS is daily temperature monitoring and visually checking inlet and outlet vents. G.2. ColllllE!nt. A number of conmenters wanted a full formal trial-type
- public hearing on the use of the NUHOMS cask
- Response. Consistent with the applicable procedure, the NRC does not intend to hold formal trial-type public hearings on the Standardized NUHOMS rule or separate hearings at each reactor site before the use of the dry cask technology approved by the Commission in this rulemaking. Rulemaking procedures, used by the NRC for generic approval of the Standardized NUHOMS, including the underlying NRC staff technical reviews and the opportunity for public input, are more than adequate to obtain public input and assure protection of the public health and safety and the environment. In this rulemaking, the NRC has taken additional steps to elicit and fully consider 46
public co11111ents on the Standardized NUHOMS technology. These steps included NRC participation in public meetings near Davis-Besse and extension of the public co11111ent period by 45 days in response to public requests. This extension provided a total public co11111ent period of almost 4 months. Section 133 of the Nuclear Waste Policy Act of 1982 authorizes the NRC to approve spent fuel storage technologies by rulemaking. When it adopted the generic process in 1990 for the review and approval of dry cask storage technologies, the Conunission stated that *casks **. [are to] be approved~ rulemaking and any safety issues that are connected with the casks are properly addressed in that rulemaking rather than in a hearing procedure" (July 18, 1990; 55 FR 29181). Rulemaking under NRC rules of practice, described in 10 CFR 2.804 and 2.805, provides full opportunity for expression of public views but does not use formal trial-type hearings of the kind requested by conrnenters. In this proceeding, rulemaking clearly provided adequate avenues for members of the public to provide their views regarding NRC's proposed approval of the Standardized NUHOMS, including the opportunity to participate through the submission of statements, information, data, opinions and arguments. In \ this connection, technical evaluations for Standardized NUHOMS and detailed documented findings of compliance with NRC safety, security, and environmental requirements were prepared by the NRC staff for public examination. In November 1993, the NRC staff reviewed the Standardized NUHOMS and approved the design for the purpose of initiating this rulemaking to grant a generic approval of the design. In addition, the NRC staff conducted a second review in response to the public comments on the Standardized NUHOMS in this 47
rulemaking, again finding compliance with NRC requirements as discussed in this document. In addition to reviewing systematically and in depth the technical issues important to protecting public health and safety, and the environment, the NRC has taken extra steps to obtain and fully consider public views on the s,andardized NUHOMS technology and has made every effort to respond to public concerns and questions about th~ Standardized NUHOMS compliance with NRC safety, security, and environmental requirements. The initial public comment period opened on June 2, 1994, and was scheduled to close on August 16, 1994. On August 29, 1994, the public co11111ent period was extended to September 30, 1994. The NRC also participated in an earlier meeting near the Davis-Besse site. Under these circumstances, formal hearings would not appreciably add to NRC's efforts to ensure adequate protection of public health, safety, and the environment and they are unnecessary to NRC's full understanding and consideration of public views on the Standardized NUHOMS. 6.3. ColTIJ)ent. One commenter stated that because there is not now and there may not be a permanent high-level radioactive waste {HLWR) repository for co11111ercial reactor fuel, and since the NUHOMS 24P and 528 casks are non-transportable, any distinction between so called *temporary storage" and "permanent disposal" of this waste is moot. Because of the lack of a pennanent repository or Monitored Retrievable Storage (MRS) in the foreseeable future, a case of a serious spill and the resultant contamination at an environmentally unsuitable site like Davis-Besse where "short and long-term adverse impacts associated with the occupancy and modification of (a) floodplain **. potential release of radioactive material during the lifetime of 48
the ISFSI **. {and location) over an aquifer which is a major water resourcen have been inadequately dealt with. Response. This rulemaking to certify the Standardized NUHOMS is for interim storage of spent fuel in an approved cask for 20 years. It does not authorize or approve the ultimate disposal in a permanent HLRW repository, which is under the responsibility of the DOE. During interim storage, the user (holder of a Part 50 license) must protect the spent fuel against design basis threats, and against environmental conditions and natural phenomena such as tornadoes, tornado missiles, earthquakes, and floods. In regard to flooding, the Certificate of Compliance has a provision (see A-2 of Certificate of Compliance) for flood condition analysis to ensure that there is no release of radioactive material from flooding. G.4. Comment. One commenter stated that projected future uses of land and water within the region are impossible to make given the unknown length of time this waste may remain on site and the options for both cask and reactor license renewal beyond 20 and 40 years, respectively, and the fact that no known man-made structure can last for the length of time that this waste must 4I be isolated from humans and the environment. If an MRS or repository ever become available, this waste may have to be repacked. Each handling of this waste increases the likelihood of an accident, spill, contamination, and worker and public exposures. Response. Projected future land and water use can be made based on the continued safe operation of a reactor and its associated dry cask storage facility. The continued operation of these facilities should have no greater impact on land and water use in the future than they do today. As previously noted, the NRC Waste Confidence decisions concluded there is 49
reasonable assurance that safe disposal of spent fuel by the Federal Government will be available by the year 2025. Therefore, the spent fuel will not remain at a reactor site for the length of time it must be isolated from humans and the environment. It should be noted that the absence of significant environmental impacts from dry cask storage at a reactor site is the conclusion of NRC's environmental assessment for the Standardized NUHOMS and for previously approved dry casks analyzed in earlier rulemakings addressing 10 CFR Part 72, - as well as in the Co111Dission's Waste Confidence decisions in 1984 (August 31, 1984; 49 FR 34658) and 1989 (September 29, 1989; 54 FR 39765). In the 1984 Waste Confidence decision, the Conunission concluded there was reasonable assurance that spent fuel can be safely stored at reactor sites, without significant environmental impacts, for at least 30 years beyond expiration of NRC reactor operating licenses. The 1989 Waste Confidence decision review reaffirmed earlier Conunission conclusions on the absence of significant environmental impacts. G.5. CoD111ent. One corrmenter questioned whether the NUHOMS canister will fit the conceptual design for the DOE multi-purpose canister (MPC). If DOE chooses to use vertical casks (like the VSC) at the MRS, will the NUHOMS inner canister fit into the vertical outer concrete shell in the MPC design? If local reactors choose the VSC-24 or the NUHOMS, will either inner metal canister fit into the overpacks for DOE, or will they have to be opened after storage, returned to the pool, the fuel put in a new canister, and the old one discarded as radioactive waste? Response. The Certificate of Compliance for the Standardized NUHOMS is intended for the interim storage of spent fuels and is not required 50
to confom to, and has not been evaluated by NRC for confomance with, the conceptual design for the DOE MPC. DOE has not yet made final decisions regarding design or deployment of the MPC. Therefore, it is not possible to speculate on conformance of the Standardized NUHOMS to the MPC. 6.6. Conment. One commenter asked what are the criteria for 20-year renewal of this cask design? How will this be checked? If the design is not renewed, what fs the plan? Response. The 1989 proposed rule (May 5, 1989; 54 FR 19379) to add Subparts Kand L to Part 72 indi~ited that the 20-year period represents what the Com:ssion believes to be an appropriate increment for cask design approvals. The application for design reapproval would have to demonstrate the cask's ability to perform the necessary safety functions for the reapproval period. The application would be evaluated by NRC against the Commission's regulatory requirements. If a cask design is not reapproved, the licensee would have to remove casks from service as the 20-year approved storage life expired. This could mean removal of the spent fuel and storing it elsewhere. G.7. Conment. One commenter wanted to discuss the need for an additional cask design, including how it would better meet the need of the interim dry cask storage of high-level waste. Response. Section 218(a) of the Nuclear Waste Policy Act of 1982 (NWPA) provides the following directive: "The Secretary [of DOE] shall establish a demonstration program in cooperation with the private sector, for the dry storage of spent nuclear fuel at civilian nuclear reactor power sites, with the objective of establishing one or more technologies that the [Nuclear Regulatory] Commission may, by rule, approve for use at the sites of civilian 51
nuclear power reactors without, to the maximum extent practicable, the need for additional site-specific approvals by the CoD111ission." After subsequent DOE technical evaluations and based on a full review of all available data, the Comission approved dry storage of spent nuclear fuel in a final rule published in the Federal Register on July 18, 1990 (55 FR 29181). The final rule established a new Subpart K within 10 CFR Part 72, entitled "General* License for Storage of Spent Fuel at Power Reactor Sites." Therefore, there is a need for casks to be approved by NRC to implement the NWPA to meet the demand of the interim dry cask storage of spent fuels in the nuclear power plants. However, the variety of cask designs submitted by vendors for NRC review and approval is mostly dictated by economic reasons that do not involve NRC. H. A number of commenters wanted site-specific analyses done for each use of the Standardized NUHOMS despite the fact that each licensee must determine that the site parameters are enveloped by the cask design specified in the SAR, SER, and Certificate of Compliance. The intent of Subpart K of 10 CFR Part 72 was to grant a general license to licensees of power reactors to use NRC-approved dry storage casks listed in 10 CFR 72.214 without additional licensing review by NRC. H.l. Comment. A number of convnenters wanted site-specific Environmental Impact Statements (EIS). Several commenters stated that an EIS should be required ~n any waste facility that may be pennanent along. the Great Lakes fresh water system. To say that this will have no adverse effect on public health and safety is a prediction most of the public does not accept. The co11111enter believes that the generic ruling to use a dry cask storage 52
design at any reactor site is impossible and should be discarded. By relying on environmental evaluations done in the 1970s before Davis-Besse construction, the NRC was remiss in its responsibility to protect the people of Ohio from hann by its licensee. Another conunenter wants the NRC to prepare, at a minimum, an Environmental Assessment (EA) for each site, including infonnation on sensitive ecosystems, wildlife, demography, meteorology, and geology. The EA should discuss the cask's capability to withstand weather conditions and potential catastrophic events. Response. The potential environmental impacts of utilities using the Standardized NUHOMS (or any of the other spent fuel casks approved by NRC (10 CFR 72.214)) have been fully considered and are documented in a published Environmental Assessment (EA) covering this rulernaking. Further, as described below, the EA indicates that use of the casks would not have significant environmental impacts. Specifically, the EA notes the 30-plus years of experience with dry storage of spent fuel have shown that the previous extensive NRC analyses and findings that the environmental impacts of dry storage are small and succinctly describes the impacts, including the non-radiological impacts of cask fabrication (the impacts associated with the relatively small amounts of steel, concrete, and plastic used in the casks are expected to be insignificant), the radiological impacts of cask operations (the incremental offsite doses are expected to be a small fraction of and well within the 25 mrem/yr limits in NRC regulations), the potential impacts of a possible dry cask accident (the impacts are expected to be no greater than the impacts of an accident involving the spent fuel storage basin), and the potential impacts from possible sabotage (the offsite dose is calculated to be about one rem). All of the NRC analyses collectively yield the singular 53
conclusion that the environmental impacts and risks are expected to be extremely small. NRC EA's for previously approved dry casks also concluded there was an absence of significant environmental impacts from dry cask storage at a reactor site when they were analyzed in earlier rulemakings addressing 10 CFR Part 72 as well as in the Commission's Waste Confidence decisions in 1984 {August 31, 1984; 49 FR 34658) and 198~ {September 29, 1989; 54 FR 39765). In the 1984 Waste Confidence decision, the Convnission concluded there was reasonable assurance spent fuel can be safely stored at reactor sites, without significant environmental impacts for at least 30 years beyond expiration of NRC reactor operating licenses. The 1989 Waste Confidence decision review reaffirmed earlier Commission conclusions on the absence of significant environmental impacts. Given the Commission's specific consideration of environmental impacts of dry storage and the absence of any new information casting doubt on the conclusion that these impacts are expected to be extremely small and not environmentally significant, the NRC is not convinced that meaningful new environmental insights would be gained from either a new site-specific EIS or EA for each site using dry storage methods. The EA covering the proposed rule, as well as the finding of no significant impact {FONS!) prepared and published for this rulemaking, fully comply with the NRC environmental regulations in 10 CFR Part 51. The Commission's environmental regulations in Part 51 implement the National Environmental Policy Act {NEPA) and give proper consideration to the guidelines of the Council of Environmental Quality {CEQ). The EA and FONSI 54
prepared as required by 10 CFR Part 51 confonn to NEPA procedural requirements. Further analyses are not legally required. The regulation 10 CFR Part 72, Subpart K, already authorizes dry cask storage and approves dry casks for use by utilities to store spent fuel at reactor sites. See 10 CFR 72.214 for a listing of infonnation on Cask tertificate Nos. 1000 through 1003, 1005, and 1007. The purpose of this final rule is to add one more cask to the list of casks already approved by NRC. The cask added to the list in§ 72.214 by this final rule complies with all applicable NRC safety requirements. Finally, this final rulemaking applies to the use of this cask by any power reactor within the United States. H.2. Comment. One commenter stated that the January 30, 1994, reply from NRC's Robert Bernero to Mr. Adamkus, EPA, is completely inadequate, as is the March 1994 "Draft Environment Assessment and Finding of No Significant Impact because no consideration is given to the site's unsuitability even for 0 LLRW per NRC's own admission, and new information which could alter the 0 original site evaluation findings" is ignored. Response. This final rule does not provide any site-specific NRC approval or address site-specific parameters that are peculiar to a particular reactor site. The rule only adds one cask design, the Standardized NUHOMS, to the list of approved casks available for use by a power plant licensee in accordance with the conditions of the general license in Part 72. Pursuant to those conditions, each licensee must determine whether or not the reactor site parameters (including earthquake intensity and tornado missiles) are encompassed by the cask design bases considered in the cask SAR and SER. The 55
EA and FONSI for this rule are limited in scope to the Standardized NUHOMS in a generic setting. Unlike interim storage prescribed in 10 CFR Part 72, the in-ground disposal of radioactive material, whether high-level or low-level waste (HLW or LLW), must take into account the geologic, hydrologic, and geochemical characteristics of the site or region to isolate the radioactive waste from the accessible environment. Site criteria for in-ground disposal of radioactive' wastes enable an applicant to choose an appropriate site, one with a combination of favorable conditions that will be a natural barrier to retard or attenuate the migration of any leaked radioactive material over a long period to control releases within acceptable limits. The disposal period for LLW is on the order of 500 years, and for HLW it is greater than 10,000 years. Therefore, site characteristics are investigated and assessed for interim spent fuel storage under Part 72, not to determine their suitability as a barrier to release of radioactive material, but rather to determine the frequency and the severity of external natural and artificial events that e could affect the safety of an ISFSI. Unlikely, but credible, severe events are considered to determine the safety of the storage cask design. H.3. Comment. One commenter stated that the NRC has not approved technologies for the use of spent fuel at the sites of *** without the need for additional site reviews. If that were so, no additional site review would have been necessary at Palisades, nor would an SAR revision or a Certificate of Compliance amendment be called for right after the VSC-24 was certified. Response. The approval and use of dry storage technologies under the provisions of the general license are relatively new. Questions were raised by members of the public about the possible effects of earthquakes and 56
erosion at the Palisades site on the safe storage of spent fuel in the VSC-24 dry casks. As the agency which is responsible for questions about compliance with regulatory requirements, which oversees such matters as the "cop on the beat,w the NRC began an independent assessment to more closely examine the behavior of the pad at Palisades under normal conditions, under the long-term effects of erosion, and under conditions of a postulated earthquake that might cause the sand below or around the pad to move. The results of NRC's assessment were documented in the NRC Final Safety Assessment of Independent Spent Fuel Storage Installation (ISFSI) Support Pad (TAC No. M88875). As is the case at all sites, NRC requires the cask user to determine if the design basis for the storage technology being considered encompasses the site parameters at the location where the fuel is to be stored. The review at Palisades confirmed this to be the case. As experience with use of this new design is gained, modifications to the design described in the SAR are expected and allowed under the provisions of 10 CFR 72.48. H.4. Comment. One commenter wanted the environmental impacts of alternatives, such as: renewable energy sources, conservation of energy, shutting down the nuclear power plants, and wind and solar power evaluated. Response. Energy production is not the subject of this rulemaking and alternative sources of energy are, therefore, not reasonable alternatives requiring evaluation. This rulemaking is limited to the addition of the Standardized NUHOMS to the list of approved casks in 10 CFR 72.214. H.5. Convnent. One commenter stated that the NRC is ignoring the regulatory requirements of a site-specific license as to the feasibility of using the cask or of modifying its design. 57
Response. This rulemaking does not cover site-specific NRC licensees; however, the NRC is not ignoring them. Under NRC regulations, the utility has two options in using dry cask storage of spent fuel: (1) the licensee may apply for a site-specific license from NRC; or (2) the licensee may use an NRC-approved cask under the general license provisions of Subpart K of 10 CFR Part 72. However, not all licensees may be able to use the general license provisions, either because the fuel type they possess is not storable in any cask listed in IO CFR 72.214 or because none of the cask designs envelope the reactor site parameters. The NRC is also not ignoring site-specific license considerations relating to modifying cask desiqns. Quite the contrary, the criteria that apply to modifications of an NRC-approved cask such as the Standardized NUHOMS are the same as the criteria that apply to modifications of site-specific ISFSis. H.6. Co11111ent. Because the populations of several states and provinces, including two-thirds of the population of Quebec, are based along the St. Lawrence Seaway, one commenter wanted an Economic Impact Statement conducted with a cost/benefit analysis citing possible adverse impact on tourism and sport fishing. Response. A regulatory analysis, which considers both benefits and impacts of adding the Standardized NUHOMS to the list of NRC-approved casks under Subpart K of 10 CFR Part 72, was prepared in support of this rulemaking action. It was included as a part of the notice of proposed rulemaking and is also included in this final rulemaking notice. However, this regulatory analysis reflects the limited scope of this rulemaking. Because the rulemaking does not provide any site-specific NRC approvals, NRC did not evaluate site-specific economic impacts. 58
H.7. Comment. One convnenter wanted to restrict the use of the cask to reactor sites that have responded on schedule to NRC Generic Letter 88-20, Supplement 4, *Individual Plant Examination of External Events (IPEEE}." Response. IPEEE response submittals will not address dry cask storage and are not necessary for Standardized NUHOMS use. H.8 Comment. One cof!lllenter stated that NUHOMS must not receive generic approval because site-specific characteristics must be considered. The commenter stated that placing this cask on the shores of Lake Erie is potential ecocide and the cask is not terrorist-proof. Another co1D11enter stated that the potential engineering problems of storing high-level nuclear waste in a variety of climatic and geologic regions of the United states are not considered. Response. A utility's use of the Standardized NUHOMS, for the storage of spent fuel in casks at a reactor site, would not have a significant impact on the environment. This finding is supported by the NRC safety and environmental evaluations for the Standardized NUHOHS, including the applicant's demonstration of compliance of the cask with NRC requirements, as well as by the 1990 rulemaking on dry cask storage and the 1984 and 1989 waste confidence proceedings. Because the Standardized NUHOMS can only be used by a licensee if the site parameters are enveloped by the cask design basis, as specified in the SAR and SER, cask storage of spent fuel near the shore of lake Erie within the specified parameters would not have a significant impact on the environment. 59
I. The following corrments relate to the transportability of dry storage casks to an off-site location. I.l. Comment. One commenter questioned how the cask transport methods used at both on-site and off-site locations are related. Response. In this rulemaking, the NRC reviewed the cask vendor's proposed means for transporting the Standardized NUHOMS canister and transfer cask outside the reactor buildings to the on-site storage pad under the storage requirements of 10 CFR Part 72. This on-site movement occurs within an owner-controlled area where access can be 'limited and where operations would be safely managed by the general licensee. The NRC did not review the Standardized NUHOMS for transport off-site, for example to a DOE MRS or repository. Generally, off-site transport of spent fuel occurs in public places where the shipper has fewer access restrictions and limited control of the surroundings. Off-site spent nuclear fuel shipments must be made in a transportation cask approved by the NRC pursuant to NRC's regulations found in 10 CFR Part 71, npackaging and Transportation of Radioactive Material," and must also comply with pertinent Department of Transportation (DOT) e* regulations. At this time, the NRC is approving the Standardized NUHOMS for storage only. I.2. Co1T111ent. One corrmenter, citing a Wisconsin Public Service Commission EIS for Point Beach, questioned the statement, nThe baskets' heavier weight and larger diameter make the transportability of an intact NUHOMS canister to an MRS site or repository questionable." Response. The NRC has not reviewed the Standardized NUHOMS in this rulemaking for off-site transportation. 60
1.3. ColTll]ent. One connnenter wanted to know the relationship between the Standardized NUHOMS and the NUHOMS MP187 now applying for a Certificate of Compliance. Is the MP187 transportable? Will the canister of all models fit into the transport overpack? Wouldn't a utility be better off waiting for the transportable cask rather than choosing a storage only cask that may have compatibility problems with an MPC system? Response. The MP-187 transportation overpack uses a canister similar to the Standardized NUHOHS. However, it is the subject of a separate NRC review as part of a site-specific licensing application. Both the Standardized NUHOMS and the MP-187 share many coR1110n design features. However, they are separate applications, and the NRC has not been asked by the cask vendor to review whether the Standardized NUHOMS can be'transported in the NUHOMS MP187 transportation overpack. The issue of whether a utility should consider the transportability of dry storage casks is beyond the scope of this rulemaking. 1.4. Comment. One coR1Denter cited a report given at the HLW Conference at Las Vegas, in 1990, *integrated Spent Fuel Storage and Transportation Systems using NUHOMS, 0 by PNFSI (page 671): "While subsequent transfer of an intact DSC from a NUHOMS on-site transfer cask directly to an OCRWM rail/barge 1s feasible, this method of transfer is not preferred since the assemblies would be oriented top down and the DSC bottom shield plug and grapple ring assembly would be orientated top up, thus complicating the canister opening and fuel handling process at the MRS or geologic repository following shipment." Has NRC evaluated this situation? Has it been rectified? 61
Response. Because the cask vendor applied for certification of the Standardized NUHOMS only as a storage cask under 10 CFR Part 72; transportation of this cask is not a subject of this rulemak1ng. Therefore, the NRC review of the standardized NUHOMS did not consider the particular transportation problem described in the comment. J. Several commenters supported the rule stating that it is beneficial to the NRC and licensees, and it is consistent with NRC's direction to avoid unnecessary site-specific licensing reviews. Others disagreed and asked specific questions about NRC's approval and oversight process. J.l. Connnent. One commenter stated that the NRC statement, "The proposed rule will not have adverse effect on public health and safety," cannot be guaranteed and, therefore, even though 1t may be convenient for the nuclear industry and the NRC to avoid site-specific approvals, in this case these are essential for maintaining public safety. Another corrmenter following the same theme questioned how the following determination was made:
"this cask, when used in accordance with the conditions specified in the
. Certificate of Compliance and NRC regulations, will meet the requirements of 10 CFR Part 72; thus, adequate protection of the public health and safety would be ensured." Response. Dry storage casks approved by the NRC for use under the general license are of a robust design that relies on generic cask features to ensure protection of the public health and safety. Additional NRC site-specific approvals are unnecessary. NRC oversight and inspections are sufficient to ensure that general licensees implement NRC conditions on cask use. If specific concerns are raised, the NRC also has the authority to look 62
into them and respond as necessary to protect public health and safety. The NRC has established specific requirements in 10 CFR Part 72 that must be met in order to obt~in a Certificate of Compliance for a cask. The details of the review and the bases for the NRC concluding that the cask meets the requirements of 10 CFR Part 72 are provided 1n the SER. The goal of dry cask storage technology is to store spent fuel safely. That goal, and the effectiveness of the technology, have been demonstrated empirically and experimentally. Different cask designs may require different types of analysis to demonstrate their safety. Therefore, different review methods may be appropriate to reach that conclusion. In each case, the level of review performed is that needed to provide assurance of adequate protection of the public health and safety. J.2. Co11111ent. Several commenters expressed concern over the exemption to 10 CFR 72.234(c) granted to VECTRA to begin transfer cask fabrication (but not use) ~to have the necessary equipment available for use by Davis-Besse Nuclear Power Station (DBNPS) in mid-1995, and thus enable DBNPS to maintain complete full-core off-load capability in its spent fuel pool following the refueling outage scheduled for early 1996." One commenter said that seeking public co11111ent and providing co11111ents is an exercise in futility because cask approval seems to be a fait accompli. Another co11111enter wants no exemptions for fabrication before certification to be allowed, stating that problems have developed when all these exemptions are allowed. Response. The NRC granted VECTRA's request for an exemption to fabricate the transfer cask before issuance of the Certificate of Compliance under its NRC-approved quality assurance program. NRC's exemption decision made a special effort to clarify that fabrication was entirely at VECTRA's 63
financial risk and did not ensure favorable consideration of VECTRA's application. The NRC's finding, based on the SAR for the Standardized NUHOMS and the NRC's SER, concluded that beginning fabrication before the issuance of the Certificate of Compliance would pose no undue risk to public health and safety. Use of the transfer cask is dependent on satisfactory completion of NRC's certification process. The NRC staff carefully considers the public comments received in rulemakings to determine whether changes are needed to the proposed rule. As noted elsewhere in this notice, several public comments received in this and other cask-approval rulemakings have resulted in changes to the SER and the Certificate of Compliance. For this reason, the public comments provide useful inputs to the NRC's safety approval process. J.3. Conment. One commenter wanted a Regulatory Guide outlining the requirements of an SAR for cask certification (CSAR). Requirements for a CSAR have not been clarified. Specific criteria for a TR (TSAR) by a vendor for a generic Certificate of Compliance need to be set. Response. Regulatory Guide 3.61, ftstandard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask, 11 dated February 1989, provides guidance for the preparation of a TSAR. Regulatory Guide 3.62, "Standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks," dated February 1989, provides guidance in preparing an SAR locating an ISFSI at a reactor site. Both Regulatory Guides identify similar information that can be potentially useful to prospective applicants for cask certification. J.4. Comment. One commenter wanted to know why Pacific Nuclear divested itself of any ownership or relationship to the VSC design in January 64
1992. How does this affect proprietary material shared in these two closely related designs? How does it affect their relationship to the DOE MPC system? Response. The key individual involved in the design and development of the VSC-24, who was also involved in the design and development of the NUHOMS design, left Pacific Nuclear and formed a new company, Pacific Sierra Nuclear, for the commercial manufacture and marketing of the VSC-24 storage system. The NRC has experienced no difficulty obtaining the required safety information, including proprietary information or answers to its questions from either firm, either before or after divestiture. The NRC is not aware of any relationship between the vendors. In addition, the NRC fully reviewed the health and safety aspects of each vendor's cask design independently. The NRC did not rely on any assumed relationship between the two vendors. Concerning their relationship to the DOE MPC system, each vendor has to establish its own relationship with DOE. J.5. Co1T111ent. One conrnenter wanted to know how long any model of NUHOMS has been used and if fuel has been taken out and evaluated. Has the 24P or 52B ever been used anywhere and for how long? If not, this is a test of a new cask at a reactor site. Response. The NUHOMS-24P is being used at Duke Power Company, Oconee Nuclear Station, under a site-specific license issued January 29, 1990, and at Baltimore Gas and Electric Company, Calvert Cliffs Nuclear Station, under a site-specific license issued November 25, 1992. Monitoring and surveillance of the system is being performed under the conditions of the site-specific license. However, there has been no need for fuel to be removed for evaluation. 65
The NUHOMS-528 has not been used yet. Pre-operational testing of the first cask system put in place under the general license is to be performed in accordance with Certificate of Compliance, Attachment A, "Conditions for Systems Use." Monitoring and surveillance of the system will be performed under the conditions of the Certificate of Compliance. The first use of the Standardized NUHOMS-52B will not place plant workers, the public, or the environment at risk. Conditions of use for the Standardized NUHOMS-528 ensure adequate safety of the workers, the public, and the environment. The Standardized NUHOMS-528 has been designed and will be fabricated to well established criteria of the ASHE B&PV and A~I codes. It uses construction materials that have well known and documented properties to provide the necessary structural strength and radiation shielding to meet regulatory requirements. While the Standardized NUHOMS-528 is not identical to the NUHOMS-24Pt many parallels in design and function can be drawn to demonstrate that the Standardized NUHOMS-528 will perform as intended. J.6. Col11tlent. One conunenter stated that even though dry cask storage passes all NRC rules and is one of the least expensive methods, it would seem that a different location or more expensive storage method is worth lives, resources, and property. Response. Based on numerous NRC reviews and growing experience with dry cask storage technologies, the NRC has concluded that spent fuel can be safely stored in dry casks without significant risk to the public health and safety. More expensive storage techniques or alternative storage locations would not provide any significant additional public protection. Further, the storage location is a matter of Congressional policy as reflected in Section 218(a) of the Nuclear Waste Policy Act of 1982, which includes the 66
following directive: "The Secretary [of DOE] shall establish a demonstration program in cooperation with the private sector, for the dry storage of spent nuclear fuel at civilian nuclear power reactor sites, with the objective of establishing one or more technologies that the [Nuclear Regulatory] Conmission may, by rule, approve for use at the sites of civilian nuclear power reactors without, to the maximum extent practicable, the need for additional site-specific approvals by the Commission." Section III(a} also finds that the generators of the spent fuel have the primary responsibility to provide for the interim storage of the spent fuel until it is accepted by the DOE. The typP of spent fuel stored in the dry cask storage systems is one factor that allows the cost of the systems to be lower. Because the fuel has cooled a number of years, passive cooling can be used rather than active cooling as is required for fuel just removed from the reactor. Passive cooling reduces the cost by not having active components such as pumps, heat exchanger, water filters, and the maintenance required for these components. J.7. Co11111ent. One commenter opposed licensing any dry cask storage system other than the DOE multi-purpose canister (MPC) because it minimizes handling individual fuel assemblies, standardizes compatibility between storage sites and DOE, and reduces cost. Multiple cask designs lead to less expertise in production, operation, and accident management. Federal regulations need to be amended to mandate only the use of the MPC. Response. The DOE MPC system will not be available for general use until well after 1997. In the meantime, additional storage capacity is needed now at several reactor sites. Once the MPC is available for general use, most utilities might use it. However, given the demonstrated and immediate need of some reactors for an additional storage capacity, and given 67
NRC's responsibility to implement dry cask storage under a general license pursuant to NWPA of 1982, it would not be prudent for NRC now to require use of MPC designs that not even DOE has yet approved. The NRC does not agree that the number of cask designs has a significant effect on the level of expertise available because standard engineering and scientific skills such as mechanical and civil engineers and health safety specialists can be hired as needed. K. Several commenters had concerns about decommissioning issues. K.1. Conrnent. One commenter, citing the draft SER, stated that decon111issioning and decontamination of reactors and reactor sites remain uncertain at best. "At thfs time, it is not known whether demolition and removal of the HSM can be performed by conventional methods *.** The reinforced structure of the HSM, for example, will require considerable effort to demolish." The convnenter continues by indicating that in its typical fashion of putting off until tomorrow what it cannot deal with today, the NRC considers "ease of decorrmissioning (a) secondary consideration." Response. The demolition of the HSM will be more difficult than a typical building because of the large amount of reinforced steel it contains. However, it is technically feasible and represents a likely level of effort similar to that required to demolish a bank vault. Bank vaults are routinely demolished without extraordinary effort. The HSM may become slightly radioactive from being exposed to a neutron radiation field during the spent fuel storage period, which would require some containment during demolition to prevent the spread of contamination. Recognizing this, the NRC considers 68
deco11111issioning a secondary consideration compared to the safety afforded by storage of spent fuel in dry casks. K.2. Conunent. One commenter questioned how, where to, and when the spent fuel and casks will go? How does the decommissioning of NUHOMS affect the reactor deco111111ssioning plan if no repository is sited and the pool must remain open? Another commenter expressed concern that after the operating facility has been decoanissioned, the spent fuel pool may not be available for use in recovery of a breached DSC. Response. The Commission determined in the Waste Confidence decisions that sufficient repository capacity will be available, in the first quarter of the 21st century, to accept spent fuel that is already in storage or that will be generated during the lifetime of the reactor licensed by NRC. In addition, the Convnission determined that spent fuel can be safely stored at reactors until it is disposed. The bases for these determinations are extensively discussed in the Waste Confidence decisions (54 FR 39765; September 28, 1989 and 49 FR 34658; August 31, 1984) and remain applicable today. To operate the dry spent fuel storage area under the provisions of the general license, a license to possess or operate a nuclear power reactor under 10 CFR Part 50 1s required. If the reactors were decomissioned and the license terminated, and if the spent fuel were to remain on site, a specific license issued under 10 CFR 72.40 would be required. At the time of application for a specific license and before the Part 50 license was terminated, the licensee would have to address the subject of how the fuel will be repackaged for shipment to an MRS or repository. {None of the casks now listed in 10 CFR 72.214 are approved for transportation). Deconvnissioning 69
and termination of a Part 50 license for a given reactor site must take into account the proper disposal of any spent fuel. L. A number of positive and negative connents were received about the application of 10 CFR 72.48 or Item 9 of the Certificate of Compliance to general licensees. L.l. Comment. Several conmenters questioned the application of 10 CFR 72.48 to Certificate of Compliance holders for use by a general licensee. Some commenters believe that this regulation is being inappropriately applied to general licensees and cask vendors. These commenters believe that the regulation was intended to apply to site-specific licenses issued under 10 CFR 72.40 only. One commenter cited the parallel application of 10 CFR 50.59 to 10 CFR Part 50 licensees. Any changes to the Certificate of Compliance and the supporting SAR and SER need public input using the rulemaking process. Who would make the decisions in using the terms "unreviewed safety questions,* "significant increase,n and "significant environmental impact*? Other commenters liked this addition, stating that non-safety-significant changes can be made in a timely and cost effective manner. Several coamenters supported the incorporation of item number 9 (in 72.48 type language) in the draft Certificate of Compliance. One cementer wanted similar provisions made for general license holders with recordkeeping requirements applicable to the general license rather than the certificate holder. Changes requiring an amendment to the certificate should be initiated by the certificate holder only. Response. The NRC will not allow changes in the Certificate of Compliance under 10 CFR 72.48. However, the general licensee may make changes 70
in the SAR under 10 CFR 72.48, unless it involves an unrev1ewed safety question, a significant increase in occupational exposure, or a significant unreviewed environmental impact. The general licensee must make the detenninations, in the first instance, that are necessary for application of 10 CFR 72.48. The licensee must also retain its evaluations on its records (which are subject to NRC review). Supporting this application of 10 CFR 72.48 to the general license are the words of 10 CFR 72.48(a)(l) which provides as follows: *The holder of a license issued under this part may: (i) Hake changes in the ISFSI *** described in the Safety Analysis Report, .** {iii) *** without prior CoD1Dission approval, unless the proposed change, test or experiment involves a change in the license conditions incorporated in the license, an unreviewed safety question, a significant increase in occupational exposure, or a significant unreviewed environmental impact." Also supporting the interpretation is 10 CFR 72.210 which provides as follows: MA general license is hereby issued for the storage of spent fuel in an independent spent fuel storage installation at power reactor sites to persons authorized to possess or operate nuclear power reactors under Part 50 of this chapter.* The NRC staff is considering a rulemaking to amend NRC regulations to explicitly state that 10 CFR 72.48 applies to general licensees. L.2. Comment. One coD1Denter stated that the CFR is silent on how a vendor can change a cask SAR and certificate after the final rule. It should be made clear for the vendor that this cask SAR (CSAR) is generic for all United States sites. All seismic, control component, distance, changes in length and weight, changes in transfer devices, etc., need to be clearly defined in the proposed rulemaking for the cask and the CSAR before public 71
comment. Who would be liable if a utility requested the vendor to change a certified cask design? Response. The cask vendor can apply to the NRC for a change to the cask certificate and SAR after the final rule is published in the Federal Register. The vendor must propose the generic revisions to the certificate and SAR and request NRC review of the proposed revision. The NRC will evaluate the proposed revision in an SER, and if appropriate, prepare a draft revised Certificate of Compliance. These documents would then be placed in the NRC Public Document Room and a proposed rule would be published requesting public convnents on the proposed revised Certificate of Compliance. After consideration of public connnents (and assuming an appropriate basis exists), a final rule would be published incorporating the revision in the revised Certificate of Compliance. The SAR (CSAR) is not necessarily generic for all United States operating reactor sites as the co11111ent appears to suggest. The SAR is pertinent for those sites that have parameters that are incorporated by the e cask design bases analyzed in the SAR. From a practical standpoint, it is difficult for a cask vendor to foresee all possible combinations of seismic, control component, distance, changes in length and weight, changes in transfer devices, etc. Revisions are expected when the vendor submits its initial application for approval. The vendor is responsible for the certified cask design. L.3. Conpnent. One connnenter wanted an explanation for not allowing buyer substitution of material for a Certificate of Compliance and that these references should be deleted from fabrication specifications and drawings. Does this mean that no changes in any materials are allowed once the design is 72
certified? If so, explain this in reference to new models of the VSC-24 as far as materials, coatings, etc.? Response. Under 10 CFR Part 72, the licensee is permitted to make changes in the ISFSI as described in the SAR provided the changes do not - - - --- -- -- invol-ve- an unreviewed--safety quest-ion_. The -1 tcensee_ and cask _certificate __ _ holder must have a quality assurance (QA) program that provides control over activities affecting quality of the identified structures, systems, and components to an extent co111Uensurate with the importance to safety and to ensure conformance with the approved design. The NRC does not want buyers (who may not be the licensee or certificate holder) of cask materials to automatically be able to substitute material without the necessary safety evaluations. Rather, the licensee, through the cask certificate holder, has the ultimate responsibility for approving any changes to ensure conformance with the approved design. For structures, systems, and components identified as important to safety, if alternative materials are desired to be used and those specific materials form the basis of the safety evaluation, it would be
- appropriate to identify those materials in the cask application.
Alternatively, the certificate holder may seek an amendment to the SAR and, if necessary, a change to the Certificate of Compliance. For other structures, systems, or components that are needed for the design to be used or are otherwise prudent, but do not perform a safety function and were not relied upon in the basis for design approval, appropriate changes may be permitted provided the licensee and the Certificate of Compliance holder document the appropriate evaluations and use their quality assurance programs to implement the change. New models of the VSC-24 casks are not the subject of this rulemaking. 73
L.4. CoR111ent. One commenter questioned how the draft Environmental Assessment and Finding of No Significant Impact would remain valid if changes to cask design and procedures can be made. Tests or experiments could be conducted under draft Certificate of Compliance Item No. 9 (see also 10 CFR 72.48l leadjng _to the use.of a cask that does not meet the con_ditions specified in the Certificate of Compliance. These changes may adversely impact site-specific public health, safety, and the environment. Response. Given the limiting criteria of 10 CFR 72.48, it is unlikely that any change would materially change the environmental analysis. The licensee's authority under 10 CFR 72.48 does not permit any changes that involve unresolved safety issues, changes to the conditions for cask use in the Certificate of Compliance, significant increase in occupational exposure, or significant environmental impact. In the Environmental Assessment supporting this rulemaking to approve the Standardized NUHOMS, the NRC staff evaluated various types of accidents that could happen to the ISFSI facility. The NRC staff's evaluation encompassed design basis accidents and concluded - that no radioactive material will be released to the environment. The NRC staff also evaluated a worst-case accident and found that the environmental impact is insignificant. Therefore, it is unlikely that the potential impact from changes to cask design or tests or experiments under the control of the licensee would introduce new environmental considerations or impacts that differ from or exceed those as analyzed in the Environmental Assessment. Changes in environmental impacts, as a result of changes to the cask design or procedures, must be evaluated by the licensee. The licensee's evaluations are available for inspection by the NRC. 74
M. A number of technical clarifications and editorial issues were raised. M.l. Conment. One commenter stated that both the SAR and SER on which the Certificate of Compliance 1s based should be dated, as was the case for the VSC-24 Certificate of Compliance. If not, the public will be commenting on-an unfinished document that can be endlessly revised. Response. Both the draft SER and the SAR are dated November 1993. These documents were revised based on public comments. M.2. Comment. One commenter wanted page one of the Certificate of Compliance revised to change the name 11 Pacific Nuclear" to "VECTRA". Response. The Certificate of Compliance has been revised to reflect this. M.3. Convnent. One commenter pointed out a typographical error on page A-19 of the draft Certificate of Compliance. In the Basis paragraph, the sentence starting, "Acceptable damage may occur *.. 11 should read "Unacceptable damage may occur .** " Response. The Certificate of Compliance has been revised to 4t correct this. M.4. ColTll)ent. One co1T1T1enter requested clarification of Technical Specification 1.2.16 on page A-25 of the draft Certificate of Compliance, as to whether the Yearly Average Ambient Temperature is a surveillance requirement or an action statement. It is unclear what action should be taken if either of the two specified limits (Yearly average temperature <70 "For average daily ambient temperature <100 *f) is exceeded. Response. The Yearly Average Ambient Temperature specification is a site-specific parameter that the user must verify in accordance with the 75
requirement of 10 CFR 72.212(b)(3) in order to use the system under the general license. There is no surveillance requirement or further action to be taken. Certificate of Compliance Section 1.1.1, *Regulatory Requirements for General License,* also includes verification of some of the same site-specific temperature parameters and has been amended to include the 100 'for less average daily ambient temperature parameter. Therefore, this specification mentioned in the comment (draft Certificate of Compliance Section 1.2.16) was delete~.
\
M.5. Comment. Apparently in reference to a December 4, 1991, letter from PNFSI that stated *The NUHOMS Certification Safety Analysis Report (CSAR) was ... ,* one comenter believed that the use of the tenn CSAR was a good idea and should have been used by the NRC. The utility SAR should be called SAR as it was and the vendor SAR should be called CSAR just as NUHOMS did in 1990. Also, the acronyms topical report (TR), TSAR, and SAR are being used interchangeably and they need clear definition. This would eliminate confusion on the issue by those involved. Response. The NRC staff generally agrees with the comment. However, the required documents that form the basis of the NRC staff's safety review are clearly identified in the SER and Certificate of Compliance. M.6. Coment. One comenter wanted the term "certificate holder" eliminated because it is ambiguous and misleading. Response. The tenn *certificate holder" has been changed to "holder of a Certificate of Compliance" to be consistent with the regulations. M.7. Comment. One conmenter wanted the draft Certificate of Compliance clarified as to who is responsible for the use of seismic restraints at each 76
reactor site, the vendor or the utility, citing the ambiguous term "certificate holder." Response. The utility is responsible for determining the need for seismic restraints in the spent fuel building based on seismic conditions at the site (Certificate of Compliance, Section 1.2.17). M.8. Colllllent. Several conrnenters stated that the limits on both neutron and gamma emission rates as well as neutron and gamma spectra (Attachment A, Section 1.2.1 of draft Certificate of Compliance) result in excluding some fuel assemblies that would actually produce loder dose rates. The problem for fuel qualification stems from the fact that the neutron dose rate does not decrease as rapidly as the gallllla dose rate during cooling because of the longer lived isotopes. Thus, a high burned fuel assembly excluded on the basis of high neutron source term may remain excluded, even though with extra cooling time the combined neutron/gamma dose rate could be less than the design basis case. Some fuel may not qualify because it exceeds the spectra requirements, even though the energy groups exceeding the limits - may not be significant contributors to the dose rates. Combined neutron/gama dose rates are the real concern; it is recolllllended that the limits on source term be replaced by limits based on dose equivalence. The fuel specification should allow other combinations of fuel enrichment, burnup, and cooling time that would not result in exceeding the fuel cladding temperature or dose rates. Response. The NRC staff agrees that alternative fuel specifications could be beneficial. However, this commenter did not provide a specific alternative, and the NRC staff has not evaluated any other 77
alternative at this time because VECTRA did not include this approach in the SAR. Therefore, no other approach is considered for this rulemaking. M.9. Comment. One co11111enter suggested wording changes to the draft Certificate of Compliance in Attachment A, Section 1.2.6, Action b, as follows: *visually inspect placement of top shield plug. Re-install or - - adjust position of top -shfela plug if -n; 1s not pt*operly seated.* The conmenter also proposed wording changes to Action c of the same section as follows: nrnstall additional temporary shielding or implement other ALARA actions, as appropriate.* Response. The NRC staff agrees with t~e first cormJent and has added the suggested words to the Certificate of Compliance, Section A.1.2.6, Action b. It is not necessary to change Action c because 10 CFR Part 20 ALARA already applies to these activities. H.10. Conment. One conmenter wanted draft Certificate of Compliance, Attachment A, Section 1.2.6, Action d deleted. The user should be pennitted to analyze and document higher dose rates under 10 CFR 72.48, which is available for NRC review. Another commenter wanted the complete Section 1.2.6 of Attachment A to the draft Certificate of Compliance deleted. Given that HSM dose rates are specified, a specification for DSC dose rates is not necessary because only the workers involved in the canister closure operations are affected by them and they are already covered by the reactor radiation protection program. One commenter wanted draft Certificate of Compliance, Attachment A, Section 1.2.11 deleted. Given that HSM dose rates are specified, a specification for transfer cask dose rates is not necessary because only the workers involved are affected, not the general public. The convnenter also stated that if Section 1.2.11 cannot be deleted the action 78
statement should be revised to read as follows: "If specified dose rates are exceeded, place temporary shielding around the affected areas of the transfer cask or implement other ALARA actions, as appropriate. Review the plant records of the fuel assemblies which have been placed in the DSC to ensure they conform to the fuel specifications of Section 1.2.1. The report to the hRC should be deleted with the user being able to analyze and document the higher dose rates under 10 CFR 72.48, which is available for NRC review." Response. The dose rate limits are for design purposes. The dose rate is limited to ensure that the DSC has not inadvertently been loaded with fuel not meeting the vendor/applicant spent fuel specifications. The NRC will require reporting if the specified dose limits are exceeded. For these reasons, the NRC will not grant the above requests. M.11. Comment. One commenter stated that the requirement for a dissolved boron concentration in the DSC of 2000 ppm is in excess of the 1810 ppm site-specific license. The 1810 ppm dissolved boron is sufficient to ensure reactivity below 0.95 K-eff {95/95 tolerance level with uncertainties) assuming 24 fresh fuel assemblies. For the unlikely worst case with water density of 0.2 to 0.7 gm/cc (a condition not achievable for fresh fuel), reactivity remains below 0.98 K-eff. The pool dissolved-boron verification-measurement frequency should be changed from not to exceed 48 hours to once per month to be consistent with 10 CFR Part 50 requirements. Another commenter stated that the NUHOMS-24P canister was designed using burnup credit, the basis for licensing is "credit for soluble boron." Jhe burnup-enrichment curve requirement {Figure 1-1, draft Certificate of Compliance) should be removed until the NRC accepts burnup credit and the pool 79
boron specification (Section 1.2.15, draft Certificate of Compliance} is removed. The NRC has not yet approved the use of burnup credit in criticality analyses for spent fuel storage and transportation casks. The applicant did, however, analyze credit for burnup as an alternative design acceptance basis for the NUHOMS-24P DSC, p*ending *furtner consideraffon of-burnup credit by NRC. As discussed in the SER, the NUHOMS-24P DSC criticality safety is approved based on, among others, the key assumptions of loading with irradiated fuel - assemblies with equivalent enrichment <1.45 wt% U-235, misloading unirradiated fuel with maximum enrichment of 4.0 wt% U-235, and soluble boron in water for wet loading and unloading. The NRC considered the use of the burnup-enrichment curve, Certificate of Compliance, Figure 1-1, as a fuel selection criteria, to be prudent. Its use adds additional unanalyzed conservatism in the criticality safety margin. It is comparable to previous NUHOMS-24P approvals. Its use would also be consistent with the requirement that storage cask designs be, to the extent practicable, compatible with removal of the stored spent fuel from the reactor site, transportation, and ultimate disposition by DOE. Therefore, the NRC disagreed with the comenter's request to allow Standardized NUHOMS-24P users the option of using these burnup-enrichment curve. Response. The conunent appears to refer to the use of a NUHOMS 24P associated with a site-specific license. The ustandardized NUHOMS 24P and 528" are the subject of this general rulemaking and should not be confused with a site license. The SER for this rulemaking is clear about conditions for use, i.e., 2000 ppm boron concentration is required to ensure that the keff remains below 0.95. The SAR for this rulemaking does not request, nor 80
does the SER grant, exemption from the requirement of k..ttt 2 0.95 for all accident conditions, including misloading of 24 unirradiated fuel assemblies and optimum moderation density. The NRC has not yet approved the use of burnup credit in criticality analyses for spent fuel storage and transportation casks. The applicant di~~ however, analyze credit for burnup as an alternative design acceptance basis for the NUHOMS-24P DSC, pending future acceptance of burnup credit by NRC. As discussed in the SER, the NUHOMS-24P DSC criticality safety is approved based on, among other assumptions, the key assumptions of loading with irradiated fuel assemblies with equivalent enrichment <1.45 wt% U-235, misloading unirradiated fuel with maximum enrichment of 4.0 wt% U-235, and soluble boron in water for wet loading and unloading. The NRC still considers the use of the burnup-enrichment curve, Certificate of Compliance Figure 1-1, as a fuel selection criteria, to be prudent. Its use adds additional unanalyzed conservatism in the criticality safety margin. It is comparable to previous NUHOMS-24P approvals. Its use would also be consistent with the requirement that storage cask designs be, to the extent practicable, compatible with - removal of the stored spent fuel from the reactor site, transportation, and ultimate disposition by DOE. Therefore, the NRC disagrees with the co11111enters request to allow Standardized NUHOMS-24P users the option of using the burnup-enrichment curve. H.12. Comment. Several commenters stated that the listing of specific fuel types in the draft Certificate of Compliance is overly restrictive. Allowance should be made for very similar fuel types or a "fuel qualification table" as proposed by the vendor should replace the listing. 81
Response. The NRC agrees that allowance should be made for very similar types of fuel to be stored. The Certificate of Compliance provides this flexibility. The "fuel qualification table* consideration at this time is not subject to this rulemaking. M.13. Comment. One coR111enter citing the first paragraph of page A-27 of the draft Certificate of Compliance states that the postulated adiabatic heatup would result in concrete temperatures being exceeded in approximately 40 hours. As a result, it is appropriate and conservative to perform the visual surveillance to verify no vent blockage on a daily basis to ensure that a blockage existed for less than 40 hours. The last sentence ~n the first paragraph should reflect that the module needs to be removed from service if it cannot be established that the blockage is less than 40 hours, not 24 hours. A 24-hour surveillance interval will adequately verify this. One coD1Denter cited an inconsistency in Section 3 of the draft Certificate of Compliance. Section 3.1 indicates that a module must be removed from service if a vent blockage is in existence for longer than 24 hours. Surveillance - Section 1.3.2 indicates that a module must be removed from service if the concrete accidenL temperature criterion has been exceeded for more than 24 hours. A vent blockage of less than 24 hours would not cause the temperature limit to be exceeded, as explained in Section 1.3 and the objective for the 24-hour frequency required by surveillance 1.3.1. The apparent conflict between Section 1.3 and the action for Surveillance Requirement 1.3.2 should be resolved. It appears that Surveillance Requirement 1.3.2 actions are appropriate. Response. The Certificate of Compliance has been clarified to reflect the comment. 82
M.14. Comment. One conunenter stated that Section 1.2.14 to Attachment A of the draft Certificate of Compliance is unnecessary because the time to transfer the DSC from the transfer cask to the HSM would normally require less than 8 hours. During this time, even with temperatures above 100 °F without the solar shield, any increase in fuel clad temperature and neutron shield temperature would be small and therefore not detrimental. Additionally, the transfer cask is open to the atmosphere and would not pressurize. Response. The vendor, vc:TRA, has proposed this limiting condition of operation in lieu of showing what detrimental effect might occur on the cladding or neutron shield, should the ambient conditions involve temperatures above 100 °F. The NRC concurs with this condition as cited in Attachment A, Section 1.2.14 of the Certificate of Compliance. N. Several commenters raised safeguards/sabotage issues. N.l. Comment. One commenter cited the World Trade Center bombing and - the ease with which a disturbed individual recently breached security and remained undetected at a U.S. reactor. Explosive technology has become very sophisticated in the last 15 years since the NRC and Sandia Laboratories studied the effect of sabotage on shipping casks in the March 1979, NUREG-0459, "Generic Adversary Characteristics Sulllnary Report." Another commenter made reference to an experiment with balloons which failed. Yet another commenter questioned the degree of protection in the spent fuel pool versus dry cask storage. Will the cask be in a vital area? Will safeguards be reviewed as part of the security plan? What is the effect on the security of these casks? 83
Response. The NRC reviewed potential issues related to possible radiological sabotage of storage casks at reactor site ISFSls in the 1990 rulemaking that added Subparts Kand L to 10 CFR Part 72 (55 FR 29181; July 18, 1990). NRC regulations in 10 CFR Part 72 establish physical protection and security requirements for an ISFSI located within the owner controlled area of a licensed power reactor site. Spent fuel in the ISFSI is required by 10 CFR 72.212(b)(5) to be protected against the design basis threat for radiological sabotage using provisions and requirements as - specifi~d in 72.212(b)(5). Each utility licensed to have an ISFSI at its reactor site is required to develop security plans and install a security system that provides high assurance against unauthorized activities that could constitute an unreasonable risk to the public health and safety. The security systems at an ISFSI and its associated reactor are similar in design features to ensure the detection and assessment of unauthorized activities. Alarm annunciations at the ISFSI are monitored by the security alarm stations at the reactor site. Response to intrusion is required. Each ISFSI is periodically - inspected by NRC and annually audited by the licensee to ensure that the security systems are operating within their design limits. The validity of the threat is continually reviewed, with a formal evaluation every six months by the NRC. The NRC is currently conducting a study into the consequences of a vehicle bomb detonated in the vicinity of an ISFSI. Following completion of this study the NRC will make a determination as to whether additional physical protection is warranted. In the interim, the NRC staff believes that the inherent nature of the fuel, along with the degree of protection provided by 84
the approved storage means for spent fuel, provides adequate protection against a vehicle bomb. N.2. Comment. One colllD8nter wanted the emergency plan updated to include initiating events caused by unnatural occurrences, such as sabotage, particularly for this fuel storage option. The commenter believes _that the NRC should determine if upgraded or new security barriers are necessary for the David-Besse site. Response. Under 10 CFR 72.212 requiregients, each general licensee must protect the spent fuel against the design basis threat of radiological sabotage. Also, 10 CFR 72.212 requires each general licensee to review the reactor emergency plan to determine whether its effectiveness is decreased, and if so, to prepare the necessary changes and obtain the necessary approvals. Therefore, the comment is already essentially incorporated into NRC regulations. O. Several commenters had fabrication, quality assurance, and inspection concerns. 0.1. Co11111ent. One co11111enter raised questions about NRC oversight and requirements for proper cask fabrication by licensees. This is based on tests of the faulty welds at the Palisades plant conducted in July 1994 just before the cask was filled, but the test was not reviewed. Response. The ultimate responsibility to ensure proper cask fabrication belongs to the user of the cask. Each Part 50 licensee (general licensee) must have its own quality assurance (QA) program in place to oversee vendor activities. The QA requirements apply to design, purchase, fabrication, handling, inspection, testing, operation, maintenance, repair, 85
modifications of structures, systems and components, and decommissioning that are important to safety. In addition, certified cask vendors have NRC-approved QA programs that control the implementation of these quality activities in a manner appropriate to the safety significance of these activities. In turn, the general licensee reviews, approves, and oversees its vendor's QA programs and activities~ the NRC in~pects both the general
-licensee and the subtiered vendors for compliance with the respective QA program requirements and for the adequacy of the activities performed.
- The faulty welds at Palisades in a loaded cask happened because the radiographs were not read initially. If the radiographs were read in a timely manner, the cask should not have been loaded without corrective action first being taken. NRC oversight and involvement in the process contributed to timely detection of the defective cask weld.
0.2. Comment. One cormnenter wants clarification of the quality assurance program. NRC should have a regulatory guide for vendors with strong criteria for audits and subcontractors, and NRC inspection reports of fabricating facilities need to be put in the PDR. How will a subcontractor of NUHOMS vendor be checked by NRC in the future? If a vendor is going to continuously change subcontractors, the NRC should inspect each cask and carefully inspect the vendor QA manual. Response. Chapters 11 or 13 of Regulatory Guides 3.62 and 3.61, respectively, provide guidance on acceptable quality assurance programs. These chapters state that a QA program meeting the requirements of Appendix B of 10 CFR Part 50 or Subpart G of 10 CFR Part 72 will be accepted by NRC. Both Parts 50 and 72 require an audit program. An NRC Branch Technical Position titled *Quality Assurance Programs for an Independent Spent Fuel 86
Storage Installation (ISFSI) 10 CFR 72,* implements the NRC review of quality assurance programs submitted by applicants. NRC inspection repprts are routinely placed in the PDR except for reports containing sensitive information. Inspection reports of NUHOMS fabrication are available in the PDR. 0.3. Comment~ One commenter wanted to know if any nonconforrnances have been discovered in inspection reports of any fabrication of the NUHOMS canister. If so, what? How was this resolved? How has the QA program for NUHOMS been reviewed? Is there a manual? How will contractors and subcontractors be checked? Response. A notice of nonconformance is documented in NRC Inspection Report No. 721004/93-07 dated August 23, 1993. The NRC staff conducted inspections in three phases at Duke Power Company, its contractor (Pacific Nuclear Fuel Services, Inc.) and subcontractor (Rancor, Inc.), concerning the QA activities with regard to the NUHOMS-24P dry spent fuel storage canisters. The NRC staff found that implementation of Duke Power 4t Company QA Program was satisfactory, in general. However, certain NRC requirements under Subpart G of 10 CFR Part 72 were not met. QA activities cited in the inspection report were documentation of nonconforming materials, parts, or components; quality assurance records; control of purchased material, equipment, and services; control of measuring and test equipment; instructions, procedures, and drawings; licensee inspection; and audits. Nonconformance corrective actions were taken and documented by Duke Power Company. The NRC staff found these corrective actions acceptable and so stated in letters dated January 13, 1994, and April 4, 1994. The corrective 87
actions taken and the implementation of the QA Program are reviewed in periodic inspections by the NRC staff. The latest version of the QA manual is "VECTRA Technologies, Inc., Quality Assurance Manual," Revision l, transmitted July 25, 1994, which reflects the corporation's new name and organization and includes additional changes-to update the manual and clarify QA recordkeeping commitments. The NRC staff found Revision 1 acceptable and so stated in its letter dated August 23, 1994. In its review, the NRC staff compared Revision 1 of the - VECTRA QA Manual with Revision 3, Edition 2, of the PNSI QA manual, which the NRC staff found acceptable by letter dated January 28, 1993. Contractors and subcontractors of cask vendors (or licensees) are subject to periodic QA inspections performed by the NRC staff. 0.4. Comment. One conmenter wanted to know if there is a possible problem, and if there was, how it was resolved, with a material defect in Swagelok tube fittings for NUHOMS? Response. The NRC is not aware of any material defect problem 4t with Swagelok tube fittings on NUHOMS designs. There is no reliance on the Swagelok fittings as part of the confinement boundary for the NUHOMS canister. The fittings are covered by a metal plate that is welded on after the canister is vacuum dried. Therefore, if there is a failure in the fitting it would be the responsibility of the licensee to repair or replace it so that the DSC can be loaded properly, but its failure would not cause a public health and safety concern.
- l Finding of No Significant Environmental Impact: Availability The Convnission has detennined under the National Environmental Policy Act of 1969, as amended, and the Co111Dission's regulations in Subpart A of 10 CFR Part 51, that this rule 1s not a major Federal action significantly affect~~g the quality of -the human environment -and therefore an environmental impact statement 1s not required. This final rule adds an additional cask to the list of approved spent fuel storage casks that power reactor licensees can use to store spent fuel at reactor sites without additional site-specific approvals from the Comission. The environmental assessment and finding of no significant impact on which this detennination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW. (Lower Level),
Washington, DC. Single copies of the environmental asseisment and finding of no significant impact are available from Mr. Gordon E. Gundersen, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission, Washington DC, 20555, telephone (301) 415-6195. Paperwork Reduction Act Statement This final rule does not contain a new or amended infonnation collection requirement subject to the Paperwork Reduction Act of 1980 {44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget approval number 3150-0132. 89
l l Regulatory Analysis The Conmission has prepared a regulatory analysis on this regulation. The analysis examines the costs and benefits of the alternatives considered by the Conunission. Interested persons may examine a copy of the regulatory
.analysis at the NRC -Public Document -Room, 2120 L Street NW. (Lower Level),
Washington, DC. Single copies of the analysis may be obtained from Mr. Gordon E. Gundersen, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Connission, Washington DC, 20555, telephone (301) 415-6195. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Comission certifies that this rule does not have a significant economic impact on a substantial number of small entities. This rule affects only licensees owning and operating nuclear power reactors and cask vendors. The - owners of nuclear power plants do not fall within the scope of the definition of *small entitiesR set forth in Section 601(3) of the Regulatory Flexibility Act, 15 U.S.C. 632, or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121. Backfit Analysis The NRC has detennined that the backfit rules 10 CFR 50.109 and 10 CFR 72.62 do not apply to this final rule. A backf1t analysis is not required for this final rule because this amendment does not involve any 90
) ' provisions that would impose backfits as defined in 10 CFR 50.109{a){l) or 72.62(a). List of Subjects in 10 CFR Part 72 Manpower training programs, Nuclear materials, Occupational safety and health, Reporting and recordkeeping requirements, Security measures, Spent fuel. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended and 5 U.S.C. 552 and 553, the NRC is adopting the following amendments to 10 CFR Part 72. PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE
- 1. The authority citation for Part 72 continues to read as follows:
AUTHORITY: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953, 954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233, 2234, 2236, 2237, 2238, 2282); sec. 274 Pub. L. 86-373, 73 Stat. 688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246, {42 U.S.C. 5841, 5842, 5846); Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123, (42 U.S.C. 5851); sec. 102, Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135, 137, 141, Pub. L. 97-91 \
425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148, Pub. L. 100-203, 101 Stat. 1330-235 (42 u.s.c. 10151, 10152, 10153, 10155, 10157, 10161, 10168). Section 72.44(g) also issued under secs. 142(b) and 148(c), (d), Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b), 10168(c), (d)). Section 72.46 also issued under sec. 189, 68 Stat. 955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42 U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(9)). Subpart J also issued under secs. 2(2), 2(15), 2(19) 117(a), - 14l(h), ~ub. L. 97-425, 96 Stat. 2202, 2203, 2204, 2222, 2244 (42 U.S.C. 10101, 10137(a), 1016l(h). Subparts Kand Lare also issued under sac. 133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252 (42 U.S.C. 10198).
- 2. In S 72.214, Certificate of Compliance 1004 is added to read as follows:
§ 72.214. List of approved spent fuel storage casks.
Certificate Number: 1004 SAR Submitted by: VECTRA Technologies, Inc. SAR
Title:
Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, Revision 2 Docket Number: 72-1004 Certification Expiration Date: (20 years after final rule effective date) 92
- l Model Numbers: NUHOMS-24P for Pressurized Water Reactor fuel; NUHOMS-528 for Boiling Water Reactor fuel.
Dated at Rockville, Maryland this ~~ay of
&~ , 1994.
For the Nuclear Regulatory Comission. or, rector for Operations. 93
OHIO DEPARTMENT OF HEALTH 246 N. HIGH STREET GEORGE V. VOINOVICH DOC t( ET ED Post Office Box 118 Governor LJ <'...,. 11 f1-; ('
\..,
Columbus, Ohio 43266-0118 PETER SOMANI. M.D.. Ph .D. Telephone: (614) 466-3543 Director of Health
*94 NOV 29 P1 2 :24 OFFICL cr ~ -, ,c L -* y November 23, 1994 OOCK E. llf'.'"; . :;*; c-*
BHAI* , 1 The Secretary .. U.S. Nuclear Regulatory Commission 6OCKET NUMBER PROPOSED RULE...:..,::____ PR 1 2 _. Washington, DC 20555 Attn: Docketing and Service Branch s C a, f '(?. ;)_yL/ q 6)
Dear Sir or Madam:
This letter is in reference to my letter dated August 16, 1994, which attached our comments to proposed rule in 10 CFR 72 to add the NUHOMS horizontal modular storage system to the list of approved spent fuel storage casks. It has recently been brought to my attention that there were inappropriate comme n ts chal l e nging NRC's sincerity in conducting a thorough review of the integrity of the cask under appropriate conditions. It was further revealed to me that there were also comments within the report that attacked NRC's track record in controlling its licensees and contractors, as well as being "remiss in its responsibility to protect the people of Ohio" by using outdated techniques and obsolete information in the safety assessment of the cask. Given that I had not reviewed this document before this was revealed to me, but only received verbal technical highlights of its content, it was released under my signature without full knowledge of such. I assure you that I do not agree with the derogatory statements made toward the NRC and find it appalling and upsetting that such made its way to you. I have always been of the opinion that the NRC has done a good job overall in protecting the general population from radiation hazards. I sincerely apologize to the NRC, therefore, for any embarrassment this may have caused you and assure you that any further comments provided to you will have my personal review in their entirety, regardless of the signatory. In recognition of the inappropriateness of significant comments made as noted above and assurance of responsible actions on our part, I am hereby withdrawing the referenced comments submitted by us to the NRC in their entirety. I do not wish for such to remain on record as a position of the Ohio Department of Health. Any position offered to the NRC relative to the cask system in question, or any other on-site issues at any of the nuclear power plants, will be provided through the Utility Radiological Safety Board of Ohio of which this agency is a member. I feel a consensus position on such by this board is paramount to assuring all HEA 6413 (Rev. 5/93) An Equal Opportunity Employer/Provider
page 2 concerns within the state are appropriately considered and comments are provided reflecting such. It is my hope that you realize the manner in which this has occurred with the understanding that such comments as referenced above do not reflect the opinion of this bureau or the department, and that such has not adversely impacted our working relationship with the NRC. I will sincerely work toward this improvement utilizing every opportunity and means available to me. I thank you for your consideration of my request, and look forward to hearing from you. I may also be reached by phone at (614) 644-2727. Si ncerely ,
-------- tJ___
Owen, Chief Bureau Radiological Health REO : sp cf: Roland Lickus, NRC Region III Roger Suppes, ODH
Law Office TERRYJONATHANLODGE DOCKETED v Toledo, Ohio 43624 U HRC t 1t; 1 618 N. Michigan Street Suite201
*~" .se..ro 3<J
'94 ~ A1 1 :Ql (419) 255
- 7552 DO K T NUMBER PR -,2 p-: ED RULE
( 5qFR 2 Ho/6)
/
September VIA FAX TO (301) 504-1672 9 AND CERTIFIED MAIL Secretary U.S. Nuclear Regulatory Commission ATTN: Docketing and Service Branch Washington, DC 20555 RE: Supplemental Comments of Terry J. Lodge and Toledo Coalition for Safe Energy on proposed rule to amend NRC regulations to add Standardized NUHOMS Horizontal Modular Storage System to list of Approved Fuel Storage Casks
Dear Secretary:
I am writing to supplement the comments I submitted on August 16, 1994 personally and as general counsel for the Toledo Coalition for Safe Energy, a Toledo, Ohio-based, unincorporated association of persons opposed to the continued commercial generation of electricity via nuclear power. Herewith are our supplemental comments in opposition to the NRC's proposal to add the NUHOMS cask design to the agency's approved list.
- 1. EPA o osition to eneric licensin of casks Lodge and Toledo Coalition for Safe Energy adopt and incorporate by reference, as if rewritten herein, the contents of the August 16, 1994 letter to Samuel Chilk of the NRC from Shirley Mitchell, Chief of the Planning and Assessment Branch, Planning and Management Division, U.S. Environmental Protection Agency (copy attached hereto as Exhibit A).
- 2. OBJECTION to Vectra's intentions of chan in the cask desi n without bein re uired to submit to ublic comment* also OBJECTION with demand that comment period be extended due to failure to make Vectra documents accessible to he public Even as the NRC is considering generic licensing of the NUHOMS system, the Vectra company has signified that it has made
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changes to the shield plug, and that other future changes are contemplated. Se e "Fax Transmission Notes" dated August 1, 1994 and August 2, 1994 from Moses Taylor, Jr. of Vectra to F .C. Sturz, attached hereto as Exhibits Band C. Lodge and Toledo Coalition hereby object to any changes to the proposed cask design without those changes being made subject to the public review and comment process. The proposed changes, which are being tracked outside the comment and rulemaking process, are being made to a nonfinal Certificate of Compliance - - clear evidence that the NRC has abdicated the role of regulator of dry cask storage. The superficial assessments of environmental concerns made by the NRC in conjunction with the C of C must be revised in view of the contemplated changes. Lodge and Toledo Coalition further demand that all documents evidencing the aforementioned design change proposals, and any other contemplated design changes, be incorporated into the draft Certificate of Compliance, and that the revised Certificate be made available for public comment, before to design depicted in the present Vectra C of C become the subject of final action.
- 3. Incorporation by reference of original and supplemental comments submitted b Alice Hirt William Hirt John Keill Wisconsin Citizens' Utilit Board Don't Waste Michi an Waste Ohio and Mary E. Sinclair Lodge and the Toledo Coalition adopt as their own and incorporate by reference, as if rewritten herein, all comments presently or previously submitted in this proceeding by Alice Hirt, William Hirt, John Keill y , Wisconsin Citizens' Utility Board, Don't Waste Michigan, Don't Waste Ohio, and Mary E.
Sinclair. N LODGE t: and as ge eral Toledo Coalition
FROM RV-PMD-PB!l TO 84192464~3 lJNITEO STATES £NVlRONMt;NTAl PRO't~CTION AGENCY REGIONS 77 WEST JACKSON OOIJLEVARO CHICAGO, IL 60604-3590 ME-19J Mr. Samuel ChilJt U.S. Nuclear Regulatory Commission Washtngton, DC 20555 Attrit Docketing and service Branch ~ar Kr. Chilk: we have revieved the proposed rule (S9 FR. 2849&) on the addition to the list o~ approved Gpent fuel storage oasJca. Your agency proposes to add the standardii&d NTJHOMS Horizontal Modular Storage Systq to the list of approved spent tuol storage casks. The casks are proposed to be used for the storage o~ spent, high level radioactive waste on site at nuclea~ power plants once fuel storage pools have reached capacity. We of~er the ~ollmting co1DJ11ents. It is stated in the proposed rule that. *t:hi* cask, when uded in accordance with the eonditlo~* sP'(leified in tha certiricata of compliance and MltC requlationer will aeet the requix-eMnta of 10 CFR Part 72: thu*, adequate protection ot the publJ.Q baalth and safety would~ an11urQCI.* However, it i* not olea.r how this detennination was *ade. The storage ot spent, high-level nuclear fuel in dry casks has the potential to result in significant ad-vea:rse impacts upon the enviroNaent 1~ just one cask is damaged or opened due to sabQtage, degradation of component.a, a
- catastrophic event, hwnan ex-ror, or so11e other reason. Bas.e:d on the limited experience with dry cask storage, and on the potential tor significant i~ct* to occur to bUJIIIµ\ health and the enviral\U\ent if the casks are daqged, thi* issue ahoulcl be further assessed..
It ia stated in the proposed rule that *caakc approved t.bro1.u1b rul.emaking a~e to be suit.able for utut under a range ot environmental conditions sufficiently ~road to enc011.pasa multiple nuclear power plants in the o.s. without the need for fUrtber site-specific approval by nc.* Furtl'1or, your agency oonten<k that it is exempt fro* conducting sit~ specific aa$eBtJJlent.s, in accordanc$ with Section 133 of the Nuclear W~*t* Policy Act of 1982 (NWPA), because genoric cask approval would elhainate the need for sit8-apecific approvals. Hovever, thrOugh the generic l:"Ulemaki.ng process, the potential engi~ring proble.>> of storing high level nuclear waste in a variety of clillatic and geologic regions of th.8 united State* are not considered. Your agency statea in the. propos$4 rule tb~t "tb* alt~rnative to this proposed action is to withhold certification of this new
design and give a site specific licenee to ~ach ~tility that proposed to use th~ oask. '11lis alternative however, would cost the NRC more time and money for each site specific review.~ Your agency ohould assess the safety And environ.ental effects of storing nuclear wastes on ea.c h aite proposing to request a d.ry cask storage license. Becaus* evecy nuclear power plant has site specitie envirorunental considerations, your agency should be required to prep4re, at a a1ni*Wll, an EA tor each plant that applies for a dry caGk storage licen~e. Each EA should provide a detailed discussion about the propo$ed site, including infonuation about sensitive ecosyste*s and typea of wildlife, as well aa information on de=ography, aeteorology, and geology unique to each ite. Each EA should al*o diacu** the casks' capability to withstand v~rioua veather conclitiona os well as
~tential catastrophic events.
It is stated in the proposed rule that Rthe kRC has determined that this rule, if ad.opted, would not be a aajor Federal action significantly affecting the quality ot tbe hu.an environaent, and ther~fore, an environmental illpa~t *tateJDent i* not required *** It would not change safety rQqUireaents and WO\lld not have significant enviroJUD.enql iapacts.M Rowe~*~, specific de~ila on the proposed addi~ional cask'* design va* not provided in the proposed rule. For C.Xallple, infl)ra.ation concerning th* -..terialti proposed tQ bG Wied, th* quant.1..ty ot spent tuel rods each cask ia able to hold, and th* projected veigbt for .acb caaJt on~ it is at capacity, i* not pt-ovide.d. A discuaalon of vhetber or not this cask design hAs been .. used in a de110matration project, or
.whether plans are underway t<> do so, vu *l*o not provided in th*
pt'oposect t'Ule. In addition, the tteed ror an addi~iol\al oaek d**lqn*wa* not
<.<<a:11\inel* that the "t.Etl'lpt)r~ry storage of hig-h-.l*v*l nucleU' vast* in eoncrete ca.-Jts*
iA~l' constitute long teo *torage. Thi* would detm drr cask 3 tox-a9e a signi~i~nt ~vironJDental impa(rt., Which should. warrant
~ite-apeci~ic analyaia ot oach proposed *ite before atorinq i gh level nuclear waste.
ROM ll:2 3SJ !S374
/lJ9 16/1994 14: 40 FROM RIJ-PMD-PBB TO 84192464703 P.04
. Thank you for th~ opportunity to provide comment& on the proposed rule. Yf you have any ~estions on our co~ents, please contact Holly Wirick of ~y staff at (312) 353-6704.
~;.tiT ~y: .~CTt<l VECTRA . 12oi **n IQOle40, '"""' 1C-0
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- SU0JECT: NUtiOMS- S~d1.a,<J S~nt F"'el Sto:&QO TOTAL PAa~S: 19 St!!em C~°'tion Slf!!):.Ma..'.)'als R ~
- Erdot-.d are draft Charq, NM and
- NNi appot',dix fOt the NUHOMS' CSAA ~ict'I Ito being t.nn1,oareo by Vt:CTRA tOt Mure~~ lO ~ CS.AR The d'\ar,ge page, mlta the *d~on of a *MO
,11.e!<j pl.Jg to tne NUHOMs* *i*P dry 11'\ae:oed C,a'\,ltt( (DSC) OUIQfl v.WC'.J'\ may be u&&O In 118:J Of th~
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~ s to tN CSAR Thau Ol"llft chanQt P&gM wit: ~ fu.d to you on 8/11-iW. l wou~ Uu~ ~ l\i9 a ...
writer~ Cllll ..t~ you Ind SAIC ~- ....~to dLSru&l l ~ potential Mu'e ChanQM. l Will ca;I you or. 8/1194 to arr*~ tho c.all. cc JoM Stol(i.y. ~c FAXED
---- _. . --- --.=-~ -- -- -----*-
~ VECT RA i). .. l 620l lM tgnec,tO. Me 100
- S.n JoM, CaWom.a iSt 19
- PHONE (-408) 620-9&00 FAX: (<tOI) 211~
fAI JBANSMIS$0N ffQD 0 URGENT O CONftOENTIM. TIME* 829AM DATE:812194 -*-- TO F. C.Sl&Q FAX NO. 301-415-5369 FROM. MoN1TaylOf'.Jt. CHARGE NO.. SUBJECT. NUH~ S ~ Spent Fuel &orage TOTAL PAGES: 5
~~- .. s~(.;.,. certtrte:M°' Sal!fX Anall!f!. R!e?!! _ _ . .
"')
'I
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\\..J tn accordance With ow recent telephone ailQJIIM)Ot, endOsod a..e dtaft changes to the HSM drawingi ShOWIOg m,soenanoous altemate Jetaels fct attaching HSM oomponenls togethef. These changes art a*~ be*ng constde(ed by VECTRA to, ~ change$ 10 the standardiZed system. l realize that lheSE reduced cop,e, may not be totally legable so I WIii ovtm91t e,cpcess cop.es to you in addition to this fax.
As dlSCU~sed before. I '#tOUkl like 10 Ml 14) 8 conference call with you and SAIC to discussed lheSE
~ .1 Potentaat future changes. ., ...
cc John Stoklev. SAIC FAXEO
- Jr-f._
*94 SEP 30 ,~18 :13 Citq Of Sqlvania Or ;
L ..J - SYLVANIA CITY COUNCIL MARGARETT. RAUCH , CLERK DOCKET NUMBER September 23, 1994 PROPOSED RULE p 72.:.- (_S Cj F {( .2-'ilf--q£) Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 Re: Dry Cask Storage Gentlemen: Sylvania City Council, at its meeting on September 19, 1994, enacted Resolution No. 32-94, calling on the NRC to consider public comments and to require a site specific impact statement prior to issuing any permit for dry cask storage. Enclosed is a certified copy of that resolution pursuant to Section 2 of same. Sincerely, Margaret T. Rauch, CMC/AAE Clerk of Council Enclosure MAR ::1 1995 Acknow1edged by card .....: .........."....---=: 6730 MONROE STREET
- SYLVANIA, OHIO 43560-1948 * (419) 885-8931 FAX (419) 885-8998
l ..:: ,_:...... . ~. r: .*I. j",* 11 ' COM1ilSSION C * ..f::' V CE SECTION. (;r , * .
- rtL ':.' RET ARY OF ThE COMMIS ION Document Statlsb
RESOLUTION NO. 32 -94 CALLING ON THE NUCLEAR REGULATORY COMMISSION TO CONSIDER COMMENTS FROM THE PUBLIC AND TO REQUIRE A SITE SPECIFIC IMPACT STATEMENT AS TO LOCAL CONDITIONS BEFORE ISSUING ANY PERMIT FOR DRY CASK STORAGE; AND DECLARING AN EMERGENCY. WHEREAS, the current licensing practices (called the NRC's "generic" licensing rule) of the NllC ..for storing high level nuclear waste at reactor sites permits dry cask storage of high level nuclear waste at every reactor site in the country with no public hearing or site specific environmental impact statement and without regard to local conditions; and, WHEREAS, the "generic" licensing rule was applied at the Palisades Nuclear Plant in Covert, Michigan, and two of the casks have been loaded; and, WHEREAS, subsequent to the loading of the casks at Palisades plant site it was determined that site specific conditions existed making the critical dunes area unstable and
DOCKET NUMBER Pl 1:2-- PROPOSED RULE..!-!!_:....:-- (5 q FR 2?-19t) DOCKETED US NRC Oyster Creek Nuclear Watch RJBox24.1 *94 SEP 30 P3 :57 Island Heights. NJ 08732 OF FICE OF S~LR'T!"'RY DOC KETI NG & SlRil CE BRANCH September 29, 1994 Attention: Docketing & Setvice Branch U.S. Nuclear Regulatory Commison Washington, D.C. 20555 Re: Comment on proposed generic approval by rule of NUHOMS 52B dry casks.
Dear Sirs:
I am writing on behalf of Oyster Creek Nuclear Watch to ask that the NUHOMS 52B dry casks not be given the generic approval currently proposed under the rulemaking process. We have read of the problem with the dry casks at the Palisades plant in Michigan, where a cask has been found to be substandard, and where there is apparently at present no established procedure for remedying the situation. We believe that it is unacceptable that casks should be approved without a known reliable method for remediating flaws that may later be discovered. To do this would be to jeopardize the health and safety of the people of our area as wen as any others where NUHOMS 52B casks may be used. Oyster Creek Nuclear Watch further objects to this proposed rule because it is our understandng that there are substantial differences between fuel rods and assemblies at different plants, which would make a generic rule inappropriate and unsafe. Sincerely, w ~ ~.i . William deCamp, Jr., Trustee Oyster Creek Nuclear Watch MAR - 1 19951 Acknowledged by card ........................~~
.S. NUCLEF.~ - ~t -.: nY COMMISSION DOCKETIN G ~ :-t HVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postman< Date J/;..qt,'/
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.J DOCKET NUMBER PROPOSED RULE_ . Pl -J__2__ (5"CJF R2<i--Jq~ September 9, Ms. Fawn Shillinglaw SEP 2 8 199 t 1952 Palisades Drive ooc,:1:T:.'lG Pac Appleton, WI 54915 SER '!CE oflAMCH SECY-NRC
Dear Ms. Shillinglaw:
'},-*;---r,- ~~
~ l t ~I Thank you for your letters of June 27, 1994, and August 16, 1994, expr ng your concerns about Part 72 General Licensees' use of 10 CFR 72.48 to make changes to an approved cask as described in the safety analysis report {SAR).
Your comments will be given due consideration as we consider changes to our regulations and Cask SARs and certificates of compliances {CofC). We will endeavor to keep you informed of any notice of proposed rulemaking related to §72.48 and amendments to the VSC-24 CofC when they are issued. The staff is currently conducting its technical analyses of Sierra Nuclear Corporation's (SNC) SAR amendment and evaluating whether changes to the CofC are required. Your concerns about the methods used to move the VSC-24 cask were addressed in a number of previous letters to you by the staff. Nothing new has transpired on t~is issue. We expect this issue to be clarified based on our review of s~c*s ~AR amendment and changes to the CofC. In regard to your support of the Wisconsin Citizen's Utility Board {CUB) letter dated July 15, 1994, the Commission has elected to consider the CUB letter as a request for action under 10 CFR 2.206. I have enclosed a copy of the response to Mr . Dennis Dums. Your Augu st 16, 994 , letter, commented on the inclusion of 10 CFR 72.48 l anguage into the Standardized NUHOMS CofC. Your letter wil l be considered as a comment to the Proposed Rule to add the Standardized NUHOMS Horizontal Modular Storage System to the list of approved casks in 10 CFR 72.214, as published in the Federal Register on June 2, 1994 (59 FR 28496). Si nee ; ~l Y, Qrlgbal : :0_':rd :-., 1 hCiber , , . *. Robert M. Bernero, Director Office of Nuclear Material Safety and Sa fegua :--1 s
Enclosure:
As stated *prior concurrence OFC STSB* STSB STSB OGC STSB DD: IMNS NAME FSturz:jc Mlusardi FBrown WReamer (Haughney BBrach DATE 08/26/94 08/ /94 08/30/94 09/01/94 09/02/94 09/02/94 OFC D: IMNS I DD: IMNS I D:NMSS J~ \ ~ NAME CPaperiello GArlotto RBern~ DATE 9/02/94 9/06/94 9/Y /94 C=COPY E=COVER/ENCLOSURE N=NO COPY OFFICIAL RECORD COPY - G:NMSS0387.FCS
I Distribution: NMSS387 NRC File Center Docket Nos. 72- 1007, 72-1004 PDR/LPDR Docket Nos. (50-266/301), (50-313/368) IMNS Central File CPoland NMSS 9400387 MKnapp i'Begion III, SfSB R/f, LJacobs-Baynard KCLeu WReamer CEstep AHansen TAlexion GKalman Region IV
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D C"ET NUMBER PROPOSED RULE PR ____;7~ [7590-01 p ocKETED ( 51 FR 2~~116) US, R NUCLEAR REGULATORY COMMISSION *94 AUG 29 A9 :S 7 10 CFR Part 72 OF F1CE F SC(.:RETARY OOCK Eri t.G _;_ s::;:;;v 1r:c: RIN 3150-AF02 u A, Ch List of Approved Spent Fuel Casks: Addition AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule: Extension of the comment period.
SUMMARY
- On June 2, 1994 (59 FR 28496), the Nuclear Regulatory Commission (NRC) published for public comment a proposed rule to add the Standardized NUHOMS Horizontal Modular Storage System to the List of Approved Spent Fuel Storage Casks. This amendment would allow the holders of power reactor operating licenses to store spent fuel in this approved cask under a general license. The comment period for this proposed rule expired on August 16, 1994.
On August 11, 1994, the NRC received a request for a 6-week extension of the comment period from Connie Kline of the Sierra Club on behalf of 12 citizen groups. The extension was requested because several proprietary documents releated to this rulemaking were not available to the public for approximately 2 weeks at the beginning of the comment period. The Commission is extending the comment period to September 30, 1994. This should allow the public ample time to review all of the needed information and to provide the NRC with comments on the proposed rulemaking.
DATES: The extended comment period will expire on September 30, 1994. Comments received after this date will be considered if it is practicable to do so but the Commission is able to assure consideration only for comments received before this date. ADDRESSES: Send written comments or suggestions to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Docketing and Service Branch. Copies of comments received may be examined at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. FOR FURTHER INFORMATION CONTACT: Mr. G. E. Gundersen, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6195; or Mr. K. C. Leu, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington DC 20555, telephone (301) 415-7864. Dated at Rockville, Maryland, this ~Tay of ~ ,d--- , 1994. For the Nuclear Regulatory Commission. for Operations. 2
DOCKET N ER 1 ;J- .. PROPOSED RULE_.!...!!.---....:.-- - ( S q fr< tJ>Lf 9~) August 19, 1994 DOCKETED US NRC Ms. Connie Kline Sierra Club *94 AUG 29 I\ 9 :56 38531 Dodds Landing Dr. Willoughby Hills, OH 44094 OF FI CE DF SECRETAR Y
Dear Ms . Kline :
OOC KET l.:G & SlRV IC t BH ANCH SUBJECT : Extension of the Comment Period for 10 CFR 72.214 In accordance with your faxed request to extend the comment period on the subject regulation, the NRC has agreed to extend the date until September 30, 1994, in order to provide more time for members of the public to prepare comments. The notification of the extension will be published in the Federal Register. This letter complies with your request to notify you regarding a decision to extend the comment period. Sincerely, Or1~foal !fi;t,~ 8),
~e.~
~Eric S. Beckjord, Director Office of Nuclear Regulatory Research Distribution:
RDB/Sub/Rdg/Central Docketing &Service Branch {DSB-94-103} JTaylor,EDO STreby, OGC SBahadur EBeckjord RAuluck TSpeis GGundersen/JMate BMorris PDR FCostanzi CHaughney JMilhoan EDO 10361 HThompson JBlaha MFato RES 940244 Document Name: FAX-KLIN OFFICE DRA:RDB DRA:RDB DRA:RDB NAME: JMate RAuluck SBahadur
/jw DATE: / /94
DOCKET NUMBER PR -12 PROPOSED RULE (_5q FR 2i-Lf'1/J
-~-I
- ~
~ USNRC DOCKETED
©@IDDHflH@m U@[r ID [.t)m@D@@(r (F[j'@@ ~rr@@O [Lmli~ .94 AUG 22 An P.O. Box 331 Monroe, Ml 48161
- OF FICE OF~ S~CRE1 ooct,E T\hl; & SER BRANCH Secretary of the Commission August 16, 1994 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Docketing and Service Branch The Coalition for a Nuclear-Free Great Lakes (CNFGL) is submitting th@
following comments in the proposed plans for the storage of spent nuclear fuel at the site of the Davis Besse nuclear power plant (Comments on the 10 CFR part 72 proposed rule). CNFGL is a non-profit coalition of citizens woups from 8 states and 3 provinces concerned with safe energy. The coalition has members in the immediate area who are impacted by the decisions -under consideration. We request Public Hearings be held on the matter of Dry Cast Storage at Davis Besse. We request that an Environmental Impact Statement (EIS) be conducted. We request that the EIS include an Economic Impact Statement with a cost I benefit analysis to be conducted as well. We request non approval of generic rule making pertaining to the dry cask storage at Davis Besse based on the following concerns presented here. There are a multitude of concerns of a specific nature which we have not outlined here. There are major population centers near Oak Harbor, Ohio that will be impacted immediately by an accident or rupture of ck"y storage casks at the Davis Besse site. These population centers include: Detroit, Ml. , Monroe, Ml. , Toledo, OH., Oak Harbor, OH .* Port Clinton, OH., The Resort Islands including Put-N-Bay, Sandusky, OH., Cleveland, OH., Akron, OH. , Findley, OH., Mansfield, OH., Windsor Ontario, Amherstburg Ontario, Leamington Ontario, London Ontario. The placement of the spent fuel from the containment or spent fuel pool at Davis Besse into an unproven and questionable technology warrants comprehensive public review. This material is lethal, and in a democracy the persons at risk must have a voice in the disposition of these lethal materials. We demand Public Hearings aa prescribed by the Atomic Energy Act (AEA). Lake Erie will flush into Lake Ontario and then flush into the St. Lawrence Seaway. This will take decades. When these waters are contaminated with radioactivity with the half lives associated with spent fuel the drinking water for millions of residents will be lost along with their economies. These stakes are too high *to be considered an *acceptable risk*. The hydrology and impact Acknowledged by card ...SEP 2 8 ,_._. _ ..,__ 1994 __
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on the people of the Great Lakes must be examined, explored and take precedence over any financial gains which might be made or saved by private interests. We demand that an Environmental Impact Statement be conducted as presaibed by the National Environmental Policy Act (NEPA). The economy of this region relies heavily on tourism and sports fishing. The land and the waters of the region are at high rist of irreparable damage due to the operation of nuclear power facilities. The deliberate placement of spent nuclear fuel into m-y storage casks is a decision which must have full public input. This decision will impact the public of several states and provinces. The communities along the shores of Lakes Erie and Ontario. and the St. Lawrence Seaway rely heavily on tourism and sports fishing. A decision to place spent nuclear fuel near the shores of Lake Erie could c.-tainly undermine these economies. We demand that a full Economic Impact Analysis be conducted as it pertains to the livelihood of millions of persons in the region. The NRC commissioned a study from the Sandia Labs which was to provide an assessment of a worst case accident at each U.S. nuclear power plant. The 1982 study concluded that there would be billions of dollars (1980 dollars) of damage at Davis Besse, thousands of deaths due to cancer would occur. These figures were developed when the nuclear fuel from Davis Besse was within an *in-depth* containment. Storage of spent nuclear fuel near the shores of Lake Erie would be at considerably higher risk than an *in-depth* containment. The containment of the ck"y storage casts must be sautinized in full view of the public. The economies of several states and provinces encompassing hunm-eds of communities would be placed in jeopardy. Two-thirds of the population from all of Quebec live along the St. Lawrence Seaway. These questions must be considered in the light of day. and the public has every right to be party to these discussions in a demoaacy. For these aforementioned reasons the Coalition for a Nuclear-Free Great Lates demands: public heaings; environmental impact statements; full cost I benefit analysis; certificate of need legislation; and economic impact statements.
, Sincerely.
~l~ Chairperson.
Coalition for a Nucle*-Free Great Lakes
OCCKET NUMBER Pl --, ,, DOCKE1ED 9'1PO~ED RULE_..:..;;~ /- J...
-- August 13, 1994 USNRC
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*94 AUG 22 Al 1 :54 Secretary-U.S. Nuclear Regulatory Commission Attention: Service and Docketing Division OFFn.: c si:-rR~ fr'\HY Washington , D.C. 20555 OOCK EflfG:x ~KvlCE RE: PUBLIC COMMENT - VECTRA DRY CASKS SYSr.049 Secretary:
As a citizen of Northwest Ohio I am vehemently opposed to nuclear power. Radioactive waste is a dangerous substance for which there is no safe, long enduring solution for either disposal or storage. It makes little sense to continue producing radioactive waste as a result of nuclear power given this situation. - Realizing that this is the public comment period for the Vectra Dry Cask Storage System, I have chosen to write to oppose the licensing of this system while the Department of Energy is in the process of developing the multi-purpose cask for the same purpose with greater benefits. DOE has proposed a multi-purpose canister (MPC) system to "provide a standardized system for storage, transportation and disposal of spent nuclear fuel. " DOE by its own informational bulletin I) "stresses the importance of minimizing handling of each individual spent fuel assembly and 2) stresses standardization and compatibility among storage technology used at civilian reactor storage sites and DOE facilities". DOE also cites over all reduction in waste management system costs as an advantage of a single system. If this in fact is DOE's position , why is the Nuclear Regulatory Commission in the business of licensing a variety of dry casks storage systems with variations from the proposed government MPC? Having multiple approved .casks systems and transport casks with separate specs, individual safety systems and separate Quality Assurance programs will surely lead to less expertise in each phase of production, operation and accident management. The NRC needs to stop the generic licensing of further casks systems including Vectra, nuclear power needs to slow down/shut down to allow time to fully develop the MPC, and the Federal Regulations need to be amended to mandate ONLY the use of the MPC. Please take these comments into consideration. Sincerely,
~/}1iMifrrJ Charlene Johnston 3417 Darlington Road Toledo, Ohio 43606
.\ S.NUCLtr.1 * * . . .. ,~A I vH Y COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics A,stmar1( Date £---I r ,;/a,t-/ ~pies Received _ _ __ _ _ __ dd'I Copies Reproducu1 _ _ _ __ peciat Distriout on Jti YJ<;1 1;200,. G UkbA.-er-.; e.,p. , ~h ~
OOCKET NUMBER Pl . . ., I PHOPOSED RULE-=--=._/;...c;;.- - OO CKE TEO (S-1 FR 2i-'19&) NUC LEA R ENE RGY US NRC INS TIT UTE
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*94 AUG 18 p 5 :Q4 John F. Schmitt, CHP DIRECTOR, RADIOLOGICAL PROTECTION, EMERGENCY PREPAREDNESS,&
WASTE REGULATION OFFI CE OF SECRf:. TARY OOCK ETi>JG x. ~'.U,VICE BR ANCH August 17, 1994 Mr. John C. Hoyle Acting Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION: Docketing and Service Branch
SUBJECT:
Comments on the Proposed Rule "List of Approved Spent Fuel Storage Casks: Addition," published by the U.S. Nuclear Regulatory Commission (59 Fed. Reg. 28496)
Dear Mr. Hoyle:
These comments are submitted on behalf of the nuclear power industry by the 1 Nuclear Energy Institute (NEI) , in response to the proposed rule "List of Approved Spent Fuel Storage Casks: Addition," published by the U.S. Nuclear Regulatory Commission (NRC) (59 Fed. Reg. 28496). We endorse the proposed addition of the Standardized NUHOMS Horizontal Modular Storage System to the list of NRC approved casks for Spent Fuel Storage in § 72.214. Based on NRC's review of the use of this cask, it has properly determined that adequate protection of the public health and safety would be ensured. Adding this cask to the "List of Approved of Spent Fuel Storage Casks" in§ 72.214 is appropriate and proper so this cask may be used under a general license. We believe this proposed rule is appropriate and beneficial to the NRC and licensees, as well as consistent with NRC's direction to avoid unnecessary additional NRC site reviews. 1 NEI is responsible for providing a unified nuclear energy industry approach to address and resolve nuclear regulatory issues and related technical matters. NEI's members include every utility licensed to operate a commercial nuclear power plant in the United States, the major nuclear steam supply system vendors, major architect/engineering firms, fuel fabrication facilities, materials licensees and other holders ofNRC licenses, and other individuals and organizations involved in the nuclear energy industry. NEI is the successor organization to the Nuclear Management and Resources Council (NUMARC), Edison Electric lnstitute's (EEi) nuclear activities including the utility waste management and transportation group (UWASTE), the U.S. Council for Energy Awareness (USCEA), and the American Nuclear Energy Council (ANEC). 1776 I STREET NW SUITE 400 WASHINGTON, DC 20006-3708 PHONE 202 739 8000 FAX 202 785.4019 8 19,., ............. Ac,"'nOW1vled q..,~db V card .....0 ...,P...,2......... I
. l~ l..,11.J - ~ .,*...., ;*/ :, ~; , ,..1~ 1'-" ;\J 0OCf\L i ;;,;:., , :.:i t::HV1CE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics stmark Date )-.}A,,,J ~ k t,11-eHL topies Received _ __ ....:,./_ _ __ _
Add'I Copies Reproduced -=- ~- - -- - Special Distribution /2:l: }2~ fJ CJ/2, . c~ u i1 r1.. e ,t S°, B-"" rbu.HA. F- /;;>'JJIJl, /21> /<. n::-./ulduAJ I A-,,,.,,;( T fe"\.--t f-o ' G;t...,>1..,:,{~rS-w'\..
Mr. John C. Hoyle U.S. Nuclear Regulatory Commission ~ August 17, 1994 '::;.. Page2 ~
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NRC is also considering a separate rulemaking ~ l would allow changes regarding the use of spent fuel casks without prior NRC approval after formal, documented review by the responsible licensee that the change will not negatively impact safety. Proposing such a rule change would be appropriate. It would also be consistent with the rationale that motivates the current proposed rule. We encourage the NRC to propose a licensee change review process consistent with§ 72.48. Such provisions should also be incorporated into the Standardized NUHOMS System' s final Certificate of Compliance so that appropriate changes are available for this system until such changes can be generally available. The draft Certificate of Compliance referenced in the proposed rule should be enhanced to be appropriately performance-based when finalized, rather than concentrating on specific fuel characteristics. This will adequately protect public health safety, and avoid unnecessary additional site reviews associated with current site-specific applications of this storage system. This is consistent with the rationale for the current proposed rule. If you have questions on our comments on this subject, please contact me at (202) 739-8108 or Steven Kraft at (202) 739-8116. Sincerely,
~ +d~
John F. Schmitt JFS/SPK:amw
Mr. John C. Hoyle U.S. Nuclear Regulatory Commission August 17, 1994 Page3 be: JFC RWB WHR JFS SPK Chron File - High Level Waste Dry Cask Storage j:\jfs\nrcsp-fl.doc
OOCKET NUMBER PROPOSED RULE PR 7 2. ( Sq FR 2f-3/1t) VECTRA *94 AUG 18 PS :OS August 16, 1994 OF Fi, , DC ~.*. NUH-03-259 Docket No. 72-1004 The Secretary U .S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Docketing and Service Branch
Subject:
Comments on USNRC's Certificate of Compliance for the Standardized NUHOMS Horizontal Modular Storage System
Reference:
Federal Register Notice of June 2, 1994
Dear Secretary:
The enclosed comments are applicable to the Draft Certificate of Compliance, No. 1004 for the subject system. Respectfully submitted, VECTRA Technologies, Inc. W. ~ es W. Axline, P .E . rage Licensing Manager MT/JWA:lds ,, Enclosure VECTRA Technologies, Inc.
- 6203 San Ignacio Ave., Suite 100
- San Jose, CA 95119
- Tel: (408) 629-9800 Fax: (408) 281-6186 Engineering
- Fax: (408) 281-6202 Fuel Services
ooc*. _ j 11\u ~ ..... ... .._. . f! t *~*r, OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Copies Received _ _ ___,__ _ __ Ader! Copies Reproduced _3:-...-- - - SpP.Cial Distributjon ~t ~ /Jf) fl _,Cz ~.p .rJ ~~ 5!ii:!~
Comment No. 1 Topic: Changes, Tests, and Experiments
Reference:
C of C Section 9, page 3-4
Background:
This Section allows the holder of the certificate to make changes without prior Commission approval under certain conditions. Comment: Similar provisions should be made for general license holders with record keeping requirements applicable to the general license rather than the certificate. Changes which would require an amendment to the certificate should be initiated by the certificate holder only. August 16, 1994 1 NUH-03-259
Comment No. 2 Topic: Fuel Specification - Fuel Manufacturers
Reference:
Attachment A, page A-6
Background:
The NUHOMS system is designed to safely store many types of fuel designs. Those designs are typically denoted by the manufacturer and array (i.e. B&W 15x15). Comment: Fuel which is designed by another vendor, such as ANF, that is functionally the equivalent of one of the fuels listed, should not be excluded. The first paragraph in "Bases" on page A-6 should be modified to include equivalent fuel designs for PWR and BWR fuels (for example: standard BWR fuel manufactured by General Electric, or equivalent"). August 16, 1994 2 NUH-03-259
Comment No. 3 Topic: Fuel Specification - Fuel Qualification
Reference:
Attachment A, pages A-5 to A-10
Background:
The NUHOMS fuel qualification is intended to assure that the maximum cladding temperatures, surface doses, and nuclear subcriticality are below design limits. Comment: The fuel specifications, as written in the draft Certificate of Compliance, are overly restrictive and will unnecessarily disqualify many fuel assemblies that would not exceed system safety parameters. The specifications should allow other combinations of fuel enrichment, bumup, and cooling time that would not result in exceeding the fuel cladding temperatures or dose rates (this is the objective of the fuel specification as stated in Section 10.3.1.1 of the SAR). This is consistent with recent licensing practices established for certified concrete systems (the Sierra VSC-24), prior NUHOMS site licenses, and revisions of the SAR which have been available to the public. We would be glad to meet with the Staff if further clarification is needed. August 16, 1994 3 NUH-03-259
Comment No. 4 Topic: Fuel Specification - Fuel Qualification
Reference:
Attachment A, pages A-5 to A-10
Background:
The licensing basis for criticality control in the NUHOMS-24P canister design is soluble boron. Comment: Although the NUHOMS-24P canister was designed using bumup credit, the basis for licensing is "credit for soluble boron". The bumup-enrichment curve requirement (Figure 1-1) should be removed until the time that the NRC accepts bumup credit and the pool boron specification (Section 1.2.15) is removed. Utilities should be allowed, at their option, to use the SAR bumup-enrichment curve when qualifying fuel. This would enhance future licensing of offsite fuel shipment (i.e. Part 71 certified transport). August 16, 1994 4 NUH-03-259
Comment No. 5 Topic: HSM Maximum Air Exit Temperature
Reference:
Attachment A, page A-17
Background:
This Section describes a functional checkout of the thermal performance of a newly loaded HSM. Some of the requirements are unclear and appear to be unnecessarily restrictive as worded in the draft. Comment: A. "Applicability" Section Section 1.2.8, "Applicability", requires the system user to calculate aT for the horizontal storage module vents whenever a DSC with less than 24 kW decay heat is loaded. Since all DSCs will be likely to produce less than 24 kW, this requirement, interpreted literally, would cause the utility to perform thermal calculations for every DSC placed in service. While we understand that the 100°F corresponds to a 24 kW heat load, and that the limit for lower heat loads would be accordingly smaller, we believe that these additional calculations are unnecessary. This is for two reasons:
- The concrete and fuel cladding temperatures have been shown by analysis in the standardized NUHOMS SAR to meet their respective criteria as long as the vent temperature difference does not exceed 100°F, regardless of the DSC decay heat. This is because fuel loads with less heat generation will have accordingly lower fuel cladding, concrete, and exhaust temperatures.
- The extra precision gained by "baselining" the predicted aT is small compared to the calculational effort; especially since periodic visual surveillance (Section 1.3.1) and daily temperature monitoring (Section 1.3.2) are already required.
We therefore ask that 100°F be used as a criterion for all fuel loads, and that no additional aT calculations are required. B. "Surveillance" Section The required surveillance interval is unclear. VECTRA proposes that the first sentence in the surveillance section be replaced with,
The temperature rise shall be measured and recorded immediately following DSC insertion, 24 hours after insertion, and again seven days after insertion into the HSM or following the occurrence of accident conditions. The continued thermal performance of the HSM shall be verified in accordance with Section 1.3.2."
August 16, 1994 5 NUH-03-259
This is consistent with previous NUHOMS site licenses. It also stresses that this test is a functional evaluation for newly loaded HSMs by referencing the requirements on Section 1.3.2 for daily HSM thermal monitoring. August 16, 1994 6 NUH-03-259
D KET NUMBER DOCKETED PF OPf'"ED RULE 71.:. US C CS 1 F (<. 2F-'-lct6)
*94 ~UG 18 PS :OS 38531 Dodds Landing Dr.
Willoughby Hills, OH 44094 August 10, 1994 The Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 ATIN: Docketing and Service Branch RE: June 2, 1994 Federal Register Notice 28496 NUHOMS Cask 24P-52B Mr. Secretary: The twelve citizens groups listed below request that you grant an extension of the comment period on the Nuclear Regulatory Commission's proposition to amend its regulations to add the Standardized NUHOMS Horizontal Modular Storage System to the List of Approved Spent Fuel Storage Casks. We request that you extend the comment period which ends on August 16, 1994 to September 30, 1994. Twenty-eight documents containing proprietary information were withheld from public disclosure for 14 days after the official comment period began. Therefore, the NRC is obligated to extend the official comment period. These documents, which were withheld from public disclosure until June 16, 1994, contain information that forms the basis for the final rule in the rule making to add the Vectra Company's NUHOMS-24P and NUHOMS-52B casks to the "List of Approved Spent Fuel Storage Casks" eligible for use under the general license in Subpart K of 10 CFR Part 72. The requirements of 10CFR 2.790(c) of NRC regulations provide that "information submitted in a rule making proceeding which subsequently forms the basis for the final rule will not be withheld from public disclosure by the Commission... " Moreover, the seriousness and complexities of issues involved with on-site dry cask storage warrant an extension of the comment period. The public needs more time in order to make informed comments on the addition of the Standardized NUHOMS System. On August 19, 1994 the Utility Radiological Safety Board of Ohio is sponsoring a workshop entitled "High Level Radioactive Waste." The workshop is three days after the official comment period ends. Information valuable in making intelligent comments may be gained by the public at this workshop which makes an extension desirable and warranted. Furthermore, serious concerns which demand in-depth study have been raised by the draft Certificate of Compliance for the Standardized NUHOMS cask. The NRC 's attempt
to insert the language and intent of 10 CFR 72.48, which pertains to Specific Licenses, into the Certificate of Compliance for the Standardized NUHOMS is a very serious issue which warrants much thought and research on the part of the general public. In light of the fact that the NRC has seen fit to avoid public hearings and to proceed with on-site dry cask storage without site specific environmental impact statements, it certainly seems prudent for the health and safety of the general public that the NRC allow the public an extension of the comment period. In view of the above considerations, the twelve environmental groups listed below request an extension of the official comment period on the Standardized NUHOMS dry cask system to September 30, 1994. Please inform Alice Hirt or Connie Kline of your decision regarding our request as soon as possible. On behalf of the following groups, c~~~ Connie Kline, Sierra Oub FAX - 216- 946-9012 Alice Hirt, Coalition for Safe Energy-FAX-616-335-8100
- Connie Kline, Sierra Club Paula Ross, Ohio Citizen Action
- Ointon Warne, Consumers League of Ohio Carolyn Monk, Don't Waste Ohio
- David Ellison, Ohio Greens Steve Gannis, Ohio Gtizens Against a Radioactive Environment Keith Haddad, Border Opposed to Nuclaer Dump
- Harvey Wasserman, Greenpeace Shirley Tomasello, Lake Erie Alliance
- David Hughs, Concerned Gtizens of Cuyahoga, Ashtabula, Lake and Geauga Counties Bridgette Mariea, Ohio Enviorunental Council
- Affiliation for identification purpose only Sent via Facsimile and U.S. Mail
08- 14-94 Mr. Ivan Selin u.s.N.R.C. ATT.Docketing and Service Branch Washington, DC 20555 eear Mr. Selin: I am writing in regards to dry storage of nuclear waste at "Davis Bessie" in Toledo Ohio. To my knowledge the plan is to contain the waste material in a dry storage above ground. After contemplating the method that would be used, I feel that this option is dangerous. The exposure of such containment is great to such perils as tor-nados, aircraft crashes, air strikes and possibly earthquakes. I understand that this storage method passes all rules with the N.R.C., however, I also know it is one of the least expensive methods. U.S. citizens and a vital fresh water resource could be at great risk. It would seem to me that a different location
- r a more expensive storage method is worth lives, resources and
~ roperty. Please reconsider a safer solution. Thank SEP 2 8 iSS4--,. Acknowledged by card .......................,__
,lS. NUCLc;..,, . ,i:cuuLATORY COMMISSIO~
DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistic! '°91mar1( Date ti I 5/1'--/ t,opies Rec:eived _ _ _,'/:;,_ _ __ Add'I Copies Aeprocuctd Special Distribution _ ll.;;f;__ >:J,, r?iJJl; . G v-ntl-..~11--S~ ~ .Jk ~
e Commonwealth Edison 1400 Opus Place DOCKET NUMBER Downers Grove, Illinois 6051 5 PROPOSED p 12 RULE_!..!:=------ ( £ C/ f ((.. ~'1'16] DOCKETED US HR C
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August 15, 1994 *94 AUG 22 A11 :LJ 9 OFFI CE CF :JECRETAHY OOCK Ef!r~G &. S[RVICF BRANCH Mr. Samuel J. Chilk, Secretary U.S. Nuclear Regulatory Commission Washington D. C. 20555 Attention: Docketing and Services Branch
Subject:
Proposed Rulemaking to 10 CFR Part 72 List of Approved Spent Fuel Storage Casks: Addition", 59 Fed. Reg. 28496 (June 2, 1994)
Dear Mr. Chilik:
Commonwealth Edison Company (ComEd) appreciates the opportunity to comment on the proposed amendments to amend its regulations to add the Standardized NUHOMS Horizontal Modular Storage System to the List of Approved Spent Fuel Storage Casks in 10 CFR part 72. In general, ComEd is in favor of including the NUHOMS system in the list of approved storage casks. Additional comments are provided in the Attachment to this letter. Sincerely, William F. Naughton, Director Strategic Licensing Policies & Issues SEP 2 8 1994 -, Acknowledged by card ..........................,,,,,,,,
.S. NUC . *_*~....J" v. 1 r vOMMl;SSION DOCKE1 ii,G & SE:RVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics IPostmarl< Date P-./t 6 ...............,_
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Attachment Additional Comments on Proposed Rule to Add the NUHOMS System to the List of Approved Spent Fuel Storage Casks
- 1. ComEd believes that the draft Certificate of Compliance (C of C)
Section 9 related to changes, tests, and experiments is appropriate because it allows design, test, and procedure changes to be made in a manner consistent with existing part 50 licensees under 10 CFR 50.59.
- 2. ComEd agrees that it is appropriate to introduce burnup credit in the fuel criticality acceptance provisions of the draft C of C.
- 3. The language of the draft C of C is overly restrictive in excluding fuel manufactured by vendors other than those specifically listed. As a minimum, page A-6 should specifically allow fuel designs that are very similar to those analyzed. Or even better, a "fuel qualification table",
as being proposed by VECTRA, should replace the listing of fuel types now in the C of C.
Law Office TERRYJONATHANLODGE DOCKETED Toledo, Ohio 43624 USHRC
~~" (419) 255 - 7552 618 N. Michigan Street Suite 201 '94 AUG 17 P3 :53 DOCKET NUMBER PROPOSED RULE PR -, 2 August 18(" F!Oi9~1- ::U'RETARY
( )G/ FR 2<;-'-IU} DOCKET I G & Sc.RV ICE 8"\t-.NCH Secretary U.S. Nuclear Regulatory Commission ATTN: Docketing and Service Branch Washington, DC 20555 RE: Comments of Terry J. Lodge and Toledo Coalition for Safe Energy on proposed rule to amend NRC regulations to add Standardized NUHOMS Horizontal Modular Storage System to list of Approved Spent Fuel Storage Casks
Dear Secretary:
I am writing as a member of the general public and as general counsel for the Toledo Coalition for Safe Energy, a Toledo, Ohio-based, unincorporated association of persons opposed to the continued commercial generation of electricity via nuclear power. Herewith are our comments in opposition to the NRC's proposal to add the NUHOMS cask design to the agency's approved list.
- 1. The NRC is i norin the
- nt of a site-s ecific license as to the the casks or of modifying their design The NRC proposes to make Title 10 of Code of Federal Regulations, part 72 applicable to evaluations of cask design, only an as-needed basis. But the wording of pertinent sections of that title requires a site-by-site technical analysis must be made before NUHOMS can be introduced at any site.
10 CFR §72.24(a) requires [a] description and safety assessment of the site on which the ISFSI or MRS is to be located ... If the proposed ISFSI or MRS is to be located on the site of a nuclear power plant or other licensed facility, the potential interactions between the ISFSI or MRS and such other facility must be evaluated." And §72.90(e)(f) requires: Pursuant to Subpart A of Part 51 of this chapter for each proposed site for an ISFSI ... the potential for SEP 2 8 1994 Ackno\ ledged by card ...............JIIJ.....UMAMl!iut
* ~;-; *:i c. v * .* :)RY 00\1MISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics stmar1( Date £/ti f'l't pies Received _ _ _/_ _ _ __ -.cicr1 Copies Reproduced ____ _ __
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radiological and other environmental impacts on the region must be evaluated with due consideration of the characteristics of the population, including its distribution, and of the regional environs, including its historical and esthetic values. The facility must be sited so as to avoid to the extent possible the long-term and short-term adverse impacts associated with occupancy and modification of floodplains. And §72.98(c)(2)(3) mandates "consideration of present and projected future uses of land and water within the region and any special characteristics that may influence the potential consequences of a release of radioactive material during the operational lifetime of the ISFSI or MRS." And §72.100(b) states: "Each site must be evaluated with respect to the effects on the regional environment ... both usual and unusual regional and site characteristics must be taken into account." And §72.102(a)(1 )(d) dictates: East of the Rocky M. Front, sites will be acceptable if the results from on site foundation and geological investigation, literature review, and regional geological reconnaissance show no unstable geological characteristics, soil stability problems or potential for vibratory ground motion at the site in excess of an appropriate response spectrum anchored at 0.2g ... Site specific investigations and laboratory analyses must show that soil conditions are adequate for the proposed foundation loading. And §72.122(b)(4),(e) enjoins: If the ISFSI or MRS is located over an aquifer which is a major water resource, measures must be taken to preclude the transport of radioactive materials to the environment through this potential pathway ... An ISFSI or MRS located near other nuclear facilities must be designed and operated to ensure that the cumulative effects of their combined operations will not constitute an unreasonable risk to the health and safety of the public. And §72.236(m) requires "to the extent practicable in the design of storage casks, consideration should be given to compatibility with removal of the stored spent fuel from a reactor site, transportation, and ultimate disposition by the Department of Energy." And §72.40(c) counsels: For facilities that have been covered under previous
licensing actions, including issuance of a construction permit under Part 50 of this chapter, a reevaluation of the site is not required except where new information is discovered which could alter the original site evaluation findings. In this case, the site evaluation factors involved will be reevaluated. In his 2/14/94 memo to the NRC's General Counsel, NRC's Haughney states,"This section (§72.48) clearly applies to specific licenses issued individual licenses under Part 72." There is not now and there may never be a permanent HLRW repository for commercial reactor fuel. The NUH0MS 24P and 52B casks are non-transportable. To call them "temporary storage," when in fact they appear to comprise a "permanent disposal" solution, is disingenuous, and that fact alone obligates the NRC to undertake a site-specific inquiry as to whether the cask design is adequate. The NRC clearly attempts to have things both ways: when it is convenient, regulations which clearly have a site-specific applicability are to be invoked, but not in a site-specific manner.
- 2. The ro osed Certificate of Com liance unlawful! authorizes chan es in cask desi n rocedures and testin Paragraph 9 of the proposed Certificate of Compliance purports to authorize its holder to make changes in the cask design described in the SAR, make changes in procedures described in the SAR, or conduct tests or experiments not described in the SAR without prior approval of the NRC so long as certain threshhold parameters are not breached. The entire text of 10 CFR
§72.48 has been incorporated by reference into the Certificate of Compliance (at paragraph 9 of Draft Certificate 1004 for the NUH0MS-24P and 52B casks). Approval of this precept would constitute an unlawful delegation of governmental regulatory authority, which is exclusively reserved to the Commission, to a private entity. It would unlawfully shift critical decisions concerning the significance of proposed changes in design, procedures or testing to the entities which stand to profit most from having such flexibility. Even more crucially, the site-specific applicability of §72.48 is wholly compromised by the unlawful delegation of regulatory authority. The use of §72.48 by a General Licensee under subpart K to modify an SAR or a C of C is not possible. There is no provision in 10 CFR 72 Subparts Kor L to allow users of the General License to modify a vendor's SAR. There also is no provision in Subparts Kor L that allows a vendor of a cask approved under Subpart L to modify its SAR or C of C. Furthermore, vendors of casks approved under Subpart L cannot utilize license provisions
of 10 CFR 72 apply to operators of a spent nuclear storage installation, not to cask vendors. Since the Code is silent on a process to change a generic cask design by changing an SAR or a C of C, the NRC must use a rulemaking procedure which provides for public comment and proprietary release. No changes to a generic SAR or C of C should be allowed without this procedure. Otherwise the whole rulemaking for generic cask acceptance is a sham allowing vendors to rush to get their designs certified "as is" and "as soon as possible" and then permitting utilities to make whatever site-specific changes they want later. This is not the intent of a generic rulemaking.
- 3. The base mat for the HSM has not been analyzed for the conse uences of crackin and breaku The pad on which the proposed HSM reposes is not considered safety-related, but is merely a shallow base mat. The NRC has not apparently addressed in any way the possibility of the pad cracking or shifting -- which could be particularly significant since many such pads will be built on sites where ground water is shallow.
- 4. Comments s ecific to ro osed use of dr cask stora eat the Davis-Besse Nuclear Power Station Davis-Besse Nuclear Power Station, 27 miles east of Toledo, Ohio was built in a marshy wetlands floodplain. A severe Lake Erie storm in October 1972 caused 300 feet of dike to break, which submerged the entire plant site, including the reactor building. People had to be evacuated by air or boat. Fortunately, the plant was pre-operational. There has been serious subsequent flooding of Davis-Besse, particularly during spring thaws, when roads leading to and from the plant are impassable due to water levels.
During September 1986 and June 1987 hearings regarding proposed LLRW sludge disposal on site at Davis-Besse, evidence was developed that Toledo Edison had provided no hydrology study and TE's 1970 geological studies related to construction of Davis-Besse were inadequate and outdated and revealed a limited understanding of soil types, permeability, water flow patterns on site, underground aquifers in the Navarre Marsh area and response to changes in Lake Erie levels or to flooding.
- 5. Criticisms of the use of a transfer cask 10 CFR 72.234(c) states that fabrication of casks under a certificate of compliance must not start prior to receipt of the C of C for the cask model. The NRC has just granted VECTRA an exemption to begin transfer cask fabrication (but no use) "to have the necessary equipment available for use by DBNPS in mid-1995, and thus enable [Davis-Besse] to ma i ntain complete full-
core off-load capability in its spent fuel pool following the refueling outage scheduled for early 1996." One one transfer cask will be shared by several nuclear power plants around the country. In the event of problems and the need to off-load the fuel, a transfer cask may not be available in a timely manner due to inclement weather or because the TC, itself, has experienced problems or is being used elsewhere. Moreover, the crane used for fuel handling in the spent fuel pool building is a single failure device. The C of C and SER discuss drop analyses of 15" up to 80". However, there is no discussion of drop accidents within the spent fuel pool building such as a drop onto the building floor or a drop of the TC into the spent fuel pool, itself, which would surely damage the fuel assemblies in the pool. A drop in either location would be considerably greater than 80 inches. The NRC has recently approved a higher MWD/MTU burnup fuel, 55,000 MWD/MTU for use in PWR's such as Davis-Besse. But the NUHOMS 24 is rated to handle only 40,000 MWD/MTU burnup fuel. There is no reconciliation of this discrepancy.
- 6. Incorporation by reference of comments submitted by Alice Hirt The proponents of these comments incorporate by reference as if rewritten herein all comments submitted in this proceeding by Alice Hirt.
Since ly,
~ otfu~~
ndi ually and as general Counse for Toledo Coalition for Safe Energy
DOCKET NUMBER PROPOSED RULE p 72 (SqfR 2HCJ6) - DOCKETED UNITED STATES ENVIRONMENTAL PROTECTION AGENc#S IR C /1y REGION 5 ((J 77 WEST JACKSON BOULEVARD *94 It CHICAGO, IL 60604-3590 AUG 2:t Al1 :SJ
~ *1 6 1994 OFF1C.t C:F c.;: -~
DOC1 £T/f,~Ft :0"'--'"'""Tiiro: TIONOF : B - - J Mr. Samuel Chilk U.S. Nuclear Regulatory Commission Washington, DC 20555 Attn: Docketing and Service Branch
Dear Mr. Chilk:
We have r eviewed the proposed rule (59 FR 28496) on the addition to the list of approved spent fuel storage casks. Your agency proposes to add the Standardized NUHOMS Horizontal Modular Storage System to the list of approved spent fuel storage casks. The casks are proposed to be used for the storage of spent, high level radioactive waste on site at nuclear power plants once fuel storage pools have reached capacity. We offer t he following comments. It is stated in the proposed rule that "this cask, when used in accordance with the conditions specified in the certificate of compliance and NRC regulations, will meet the requirements of 10 CFR Part 72; thus, adequate protection of the public health and safety would be ensured." However, it is not clear how this determination was made. The storage of spent, high-level nuclear fuel in dry casks has the potential to result in significant adverse impacts upon the environment if just one cask is damaged or opened due to sabotage, degradation of components, a catastrophic event, human error, or some other reason. Based on the limited experience with dry cask storage, and on the potential for significant impacts to occur to human health and the envircnment if the casks are damaged, this issue should be further assessed. It is stated in the proposed rule that "casks approved through rulemaking are to be suitable for use under a range of environmental conditions sufficiently broad to encompass multiple nuclear power plants in the U.S. without the need for further site-specific approval by NR".:. 11 Further, your agency contends that it is exempt from conducting site specific assessments, in accordance with Section 133 of the Nuclear Waste Policy Act of 1982 (NWPA), because generic cask approval would eliminate the need for site-specific approvals. However, through the generic rulemaking process, the potential engineering problems of storing high level nuclear waste in a variety of climatic and geologic regions of the United States are not considered. Your agency states in the proposed rule that "the alte rnative to this proposed action is to withhold certification of this new SEP ~ {ggf d on Recycled Paper 2 Ac1<nowledged by card ................. ..,.,. ,
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U.S. NUCU:A:; i~LuULATORY COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date --~- - - - - Copies Received _ __ _ _ __ Add'I Copies Reproduced _ _ _ _ __ Special Distributon (2_ :;(.0~ RfJll . G ~ - / 'A
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TO: NRC August 9, 1994 OOH PERSPECTIVES AND CONCERNS RELATIVE TO NUHOMS-24P DRY CASK STORAGE SYSTEM FOR SPENT NUCLEAR FUEL BACKGROUND INFORMATION AND DISCUSSION OF ISSUES Following is a discussion of the Ohio Department of Health (ODH) perspectives and concerns on.various issues related to the generic design of dry storage casks, particularly the NUHOMS-24P design that Toledo Edison Company has elected to use at Davis-Besse Nuclear Power Station, near Oak Harbor, Ohio. 1.0 Defense In Depth Versus a Single Barrier Design -- 1.1 It is ODH perspective that the single barrier concept employed in the dry cask storage system violates the NRC general design criteria under which most nuclear plants have been licensed for construction. The NRC general design criteria provides defense in depth to release of radiation. PotentJal releases from the nuclear fuel within the reactor were to 'be contained by several barriers:
- fuel with cladding integrity
- the reactor vessel
- the reactor coolant system
- the containment
- 1. 2 In addition to this defense in depth, there were to be criticality control systems and engineered safety systems designed to keep the core coo1, shut the reactor down, respond to emergency conditions in order to mitigate releases, monitor those releases and contain them on the site, if possible.
1.3 In nuclear power plants today, radioactive fuel is kept in a spent fuel pool, where there are multiple engineered safety systems to prevent and contain releases. The pool is a steel
-lined, reinforced concrete structure in a fuel building adjoining the reactor containment. Discharged fuel is transferred to its storage location in the spent fuel pool from the reactor via an interconnecting water-filled canal and fuel transfer system. Highly borated water provides a means to shield workers from direct radiation, cool the fuel and keep it subcritical. Heat in the pool is removed via a heat exchanger and the water can be . radiologicall-y cleaned to remove contamination from leaking fuel. In addition, the air above the pool is continuously f1ltered to keep it clean.
Spacing between fuel assemblies is maintained by spent fuel racks in which the discharged fuel assemblies are kept fully supported in a vertical position. The spacing is also a
TO: NRC August 9, 1994 ODH PERSPECTIVES AND CONCERNS RELATIVE TO NUHOMS-24P DRY CASK STORAGE SYSTEM FOR SPENT NUCLEAR FUEL measure to prevent criticality. 1.4 Unlike fuel in a reactor or in a spent fuel pool, the NUHOMS-24P dry storage cask and others like it have but a single barrier between the hi,ghly radioactive spent fuel and the biosphere. That sole barrier is a single-walled stainless steel canister with a welded lid. That is the only barrier interposed between the spent fuel assembl'ies, whose cladding may be substantially cracked, water-logged or otherwise qegraded by years of reactor cycling and pool storage, and the ambient air that circulates through vents in the concrete structure that surrounds the canister.
- 1. 5 It is the OOH perspective that the NUHOMS-24P cask single barrier design is insufficient* to rely upon for the he~lth and safety of Obioans, gi~en the complete lack of other detection, safety and mitigation features ..
2.0 Dry Cask Storage Release Pathways 2.1 Release of fission products from the fuel rods whose cladding is already defected requires a breach of the canister, a transport' mechanism for fission products to escape the canister and a pathway to where they may do harm. 2.2 ODH is concerned that the heat generated by fission product decay may provide the driving force, the presence of free moisture in water-logged fuel may, in a norunecha'nistic way, provide a transport mechanism for fission product release and the ambient air circulating through the cask concrete structure may provide (an unmonitored) pathway to the biosphere. 2.3 In the NUHOMS-24P dry storage cask there are neither active nor passive systems in place to mitigate barrier breaches, nor are there active radiation monitors that would indicate a breach has occurred. There are no monitored drains and sumps nor are their retention basins. 3.0 Breached Canister Handling 3.1 A place to unload the highly radioactive spent fuel in the event of a breach of the canister has not been provideq; nor is there any indication that the canister, the canister lifting mechanisms or transport mechani~ms to move the canister into the cask are to be nuclear grade equipment or
TO: NRC August 9, 1994 ODH PERSPECTIVES AND CONCERNS RELATIVE TO NUHOMS-24P DRY CASK STORAGE SYSTEM FOR SPENT IWCLEAR FUEL has been designed to prevent a single failure from breaching the canister, thereby circumventing the protection provided by the sole barrier provided by the canister wall itself. 3.2 If the canister integrity is maintained, it can be unloaded from the concrete structure and returned to the spent fuel pool building for inspection. However, there are no ways designed into the dry cask storage system to remotely inspect the c~sk for corrosion while_ it is in place. 3.3 With the deconunissioning of an operating-facility that has these dry storage casks sitting on a pad outside the fuel building, it is not clear to us that a spent fuel pool will be available to afford its use in recovery from a breached condition. 3.4 Davis-Besse has recently had its license amended to permit it to use more highly enriched fuel ( up to 5 weight percent uranium-235 versus the 3.8 weight percent previously allowed) and achieve higher fuel assembly average discharge burnups (60,000 megawatt days per metric ton). This should have the effect of reducing the number of spent fuel assemblies produced, all other things being equal. However, the ODH is concerned that the increase in fission products available to produce decay heat from the burnup of more highly enriched fuel and potentially increased embri ttlement of the fuel cladding from the extended burnup allowed needs to be evaluated in the NRC safety evaluation, particularly from the heat transfer considerations and the probability of more defected fuel being loaded into dry casks. 1 3; 5 Since the dry cask storage structures are used to store hazardous materials, they should also be evaluated as to the need for compliance with Building Officials & Code Administrators (BOCA) National Building Code (and Ohio Administrative Code) requirements for structures in use group At the August 13, 1993, public meeting of the Utility Radiological Safety Board of Ohio's Working Group held in Columbus, Ohio, the Toledo Edison Company representative, Ted Myers, stated that Davis-Besse staff will make no effort to avoid loading fuel with known cladding defects nor waterlogged fuel into the dry cask storage canisters.
TO: NRC August 9, 1994 ODH PERSPECTIVES AND CONCERNS RELATIVE TO NUHOMS-24P DRY CASK.STORAGE SYSTEM FOR SPENT NUCLEAR FUEL H-4, high haza~d use, which includes radioactive materials. 2 3.6 As a participant on the Utility Radiological Safety Board of Ohio ("URSB), the Ohio Department of Health, Bureau of Radiological Health, received and reviewed the Battelle Pacific Northwest Laboratories Report, PNL-5987, Investigation of Water-Logged Spent Fuel Rods Under Dry Storage Conditions. This report has been cited by the NUHOMS-24P cask vendor 3 as evidence that* spent fuel that has degraded cladding and has taken on water will _not provide sufficient moisture to transport radionuclides nor corrode the weldment sealing the stainless steel canister within the cask. The laboratory investigation simulated the vacuum drying process that follows loading of spent fuel into its steel storage canister. The PNL investigation demonstrated that it is extremely difficult to remove water that tends to accumulate in the fuel rod plenums and absorb onto the ceramic fuel pellets. It concludes that there is insufficient free moisture remaining after drying to cause degradation of the steel canister. While this report's conclusions provide a level of comfort, it is not altogether clear to us that the mechanism for free moisture and radionuclide release that pertain in normal operation or under upset conditions, such as caused by sabotage, have been simulated adequately.
- 3. 7 Per a statement by Toledo Edison Company's representative 4 spent fuel to be loaded into the canisters for storage outside will not be inspected beforehand to eliminate known leakers.
-However, based on the ,concerns discussed above, it is the. ODH perspective that spent fuel with known defects and all water-logged fuel should be retained in the spent' fuel pool and not put in dry storage casks outside, as an additional conservatism, until such time as cask integrity under operating conditions is fully demonstrated.
- 3. 8 Further, it is the ODH perspective that a periodicity for 2
BOCA National Building Code, Article 3; also Ohio Administrative Code (OAC)4101:2-3, Section 306-.5 3 NUTECH Engineers, Inc. (thereafter Pacific Nuclear, now Vectra) Horizontal Modular Storage System 4 Ted Myers statement at the August 13, 1993, public meeting of the URSB Working Grqiup. See also Toledo Blade, August 18, 1993, article by Ann Fisher, p. 3
TO: NRC August 9, 1994 ODH PERSPECTIVES AND CONCERNS RELATIVE TO NUHOMS-24P DRY CASK STORAGE SYSTEM FOR SPENT NUCLEAR FUEL inspection of the integrity of the steel canister should be established to preclude unidentified, gradual canister deterioration by mechanisms that are not yet known. It is our understanding from the NRC that in fact a rigorous inspection program will be made part of the NRC approval requirements. 5 3.9 Their are considerable ramifications of usage of generically approved casks, such as the NUHOMS-24P design, beyond merely the efficacy of the cask itself. The cask does not support itself, nor does it lift itself, nor does it transport itself nor inspect itself to determine it* its vents are blocked. It is ODH perspective that no plant should be allowed to use this cask without meeting strict, codified NRC prerequisites as to the rest of the storage systems structures, supporting equipment and operating processes, and the quality of their design, materials, fabrication, testing, maintenance, inspection and quality control. Further, these- structures, equipment and processes should be of nuclear grade quality because of their importance to safety. Further the NRC should evaluate the adequacy of the entire storage system not just its cask. Allowing the cask to be put into place without attending to these prereq~isites is folly. 4.0 Safeguards/Security Considerations 4.1 At Davis-Besse, as at other nuclear power plants, there is a security plan in effect to keep out intruders 6 and an emergency plan in effect to show how releases would be dealt with respecting the heal th and safety of the surrounding population. Both of these plans are part of the required documentation and need to be regularly updated as part of thee Updated Final Safety Analysis Report submittals. 7
- 4. 2 Further to cask security safeguards, the casks sit ori a reinforced concrete pad within the protected area of the plant;_ however, the degree of protection for the fuel in these 5
Conversation with Frederick Sturz, NRC-Washington, at the URSB Working Group (public) meeting on August 13, 1993. 6 The security plan by its nature is not a publicly available document. 7 The ODH and other state and local ~gencies play an important role in protecting the public during emergencies as part of the emergency plan implementation.
TO: NRC August 9, 1994 OOH PERSPECTIVES AND CONCERNS RELATIVE TO NUHOMS-24P DRY, CASK STORAGE SYSTEM FOR SPENT NUCLEAR FUEL casks versus fuel in the reactor or spent fuel pool, in our view is a safeguards issue that should be addressed by the NRC. Will the casks oe protected as a vital area within the protected area? Will the safeguards be explicitly reviewed as part of the updating of the security plan? We note that the NRC proposes to amend 10 CFR Part 73 to improve barrier design to protect against malevolent intrusion of explosives carried in land vehicles. 8 What is the effect on the security of these casks? 4.3 It is the OOH perspective that the eme~gency plan needs to be updated to include initiating events caused by unnatural occurrences, such as sabotage, particularly for this fuel storage option. Whether specifically for the Davis-Besse site this means upgraded or new security barriers is for the NRC to determine. 5.0 Using the Old Pre-Construction Environmental Evaluation 5.1 Consumers Power's Palisades plant, under the NRC's general licensing provisions (per 10 CFR 72, Subpart K) was allowed to use a cask storage system at its plant without the need for further site specific 'licensing that would require a lengthy NRC review9. It is the ODH' s understanding that this situation is the one under which Davis~Besse seeks to implement its dry cask storage system. 5.2 In utilizing generic licensing of casks, the assumption being employed by *the NRC to avoid a lengthy licensing review for a specific plant, such as Davis-Besse, is that the conditions bounding use of the dry cask storage system will not be worse
, than already analyzed in the current site accident analyses under which its operating license was granted. However, in our view, not being worse than a cqre meltdown is no comfort and sJD,all justification for imposing another risk to the population and environment of this part of Ohio, in order to absolve the federal government of its responsibilities to properly site and license a long-term repository for high level waste.
5.3 By relying on environmental evaluations done in the 1970's, prior to Davis-Besse' s construction, using old evaluation 8 Nuclear News, December 1993, p.75 9 ibid, p. 39
TO: NRC August 9, 1994 ODH PERSPECTIVES AND CONCERNS RELATIVE TO
~OMS-24P DRY CASK STORAGE SYSTEM FOR SPENT NUCLEAR FUEL techniques and now obsolete demographics and land use data, the NRC is remiss in its responsibility to protect the people of Ohio from harm by its licensee. The NRC should be requiring that new environmental impact statements be created for each of these sites, including Davis-Besse, as a prerequisite to the utilization of any of these types of casks at any of the nuclear power plant sites.
- 5. 4 The Davis-Besse site has not been evaluated for suitability as a long term repository for highly radioactive waste. The specific environmental impacts of storing spent nuclear fuel in dry storage casks versus other alternatives, such as spent fuel pool reracking, is needed for each power plant site, regardless of whether this generic cask design is licensed for use. Given the federal government's track record in controlling its licensees or contractors so as to prevent the endangerment of the public at the numerous radiologically contaminated sites in Ohio, and the possibility that the federal government will continue its record of failure in its mission to build a national high-level waste repository, licensing these casks as a temporary measure until a high-level waste repository is ready to take their place is a naivety that the (concerned citizens of Ohio and) ODH do not share nor wish to indulge in with the federal bureaucracy.
II I DOCKET NUMBER PR 12 Duke Power Company PROPOSED RU L E ~ - - - Electric Center
- Q P.O. Box 1006 Charlotte, NC 28201-1006
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Secretary, U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATIN: Docketing and Service Branch
Subject:
Comments on the Proposed Addition of the Standardized NUHOMS Horizontal Modular Storage System to the List of Approved Spent Fuel Storage Casks; Federal Register Notice of June 2, 1994 File no.: OS-230.04 The following comments are applicable to the Draft Certificate of Compliance, No. 1004, for the subject system. Section 9- Changes, tests, and experiments This section is a much needed addition to the certificate since 10CFR72, Subpart K currently includes no provisions for implementing non-safety significant changes. Such changes are permitted for site-specific licenses under 10CFR72.48, which specifies the evaluation criteria for such changes. Without the inclusion of similar provisions for the certificate holders of General License systems, even such minor deviations as altering the sequence of non-critical procedure steps or revising a non-critical weldment for fabricability cannot be accomplished without prior NRC review and approval. This section should be broadened to grant General Licensees similar discretion for implementing non-safety significant changes. Such action would be consistent with licensees' current practices for implementing similar changes for reactor operations under 10CFR50.59, as well as for ISFSI operations under site-specific licenses via 10CFR72.48. Without such action, timely and cost effective changes could not be accomplished. Printed on recycled paper c-:p 2 8 1994 A r:>\wloJ1:,ed uY card ...~:......................,...,
v.- .. . ._.-., -.1li*1 COMMISSIOI\I DOCKi:T!NG & SERVICE SECTION OFFICE OF THE SECRETARY or THE COMMISSION Document Statistics Postmark Date ,9:/J5/'1'1 Copies Aeceived _ _ _-,,lI!..--___ _ hld'I Copies Reproduced -=-- ...--,:,...,,....- - Special Distribution ~ fL=!: = V....:5.J-J..-4,L~- - G U-"'-~.f ICUr,.,;1 Sb W;!Y:-
Attachment A 1.2.1- Fuel Specification Fuel Claddin~ Criteria This section precludes the storage of fuel with known or suspected gross cladding breaches. Previously, in site-specific licenses, the Commission has restricted only fuel with structural defects which could impair handling of the fuel assembly. This more restrictive requirement is not warranted for the following reasons:
- 1. The fuel is confined in a sealed metal canister. Thus, any fuel particles that might exit a fuel cladding breach are contained.
- 2. The dry, inert internal atmosphere of the DSC prevents oxidation of the fuel material in a rod that has sustained a cladding breach.
- 3. The potential for airborne contamination during the fuel retrieval process is addressed by the requirements of Section 1.1.2.
FuelBurnup Bumup is limited to 40,000 MWD/MTU for PWR assemblies and 35,000 MWD/MTU for BWR assemblies. This limit is derived from the design basis fuel assemblies which additionally assume 4.0 weight percent U-235 and 5 years' cooling. As written, this specification would not allow the utility to load fuel with higher bumup, even though such fuel could meet the decay heat requirement while, at the same time, have less significant radiological source terms, if simply cooled for a longer period of time. This section should be amended to allow the user to exceed the stated bumup limits, provided that the decay heat and radiological source t.erms are verified to be within limits. Such a provision is consistent with current site-specific licenses for the NUHOMS system. Neutron and Gamma Source Terms It is recognized that source terms must be limited to ensure that dose limits for radiation workers and the ISFSI offsite contributions are not exceeded. However, the requirement to meet stated limits on both neutron and gamma emission rat.es as well as neutron and gamma spectra results in excluding some fuel assemblies which would actually produce lower dose rates. For example, neutron dose rates are only a fraction of the gamma dose rates- even for the higher neutron dose rates characteristic of higher burned fuel assemblies. The problem for fuel qualification stems from the fact that the neutron dose rate does not decrease as rapidly as the gamma dose rate during cooling because of the longer lived isotopes. Thus, a high burned fuel assembly excluded on the basis of high neutron source term may remain
excluded. Even though with extra cooling time, the combined neutron/gamma dose rate could be less than the design basis case. Additionally, some fuel may not qualify due to exceeding just the spectra requn:ements, even though the energy groups exceeding the limits may not be significant contributors to dose rates. Since combined neutron/gamma dose rates are the real concern, it is recommended that the limits on source term be replaced by limits based on dose rate equivalence. Tables or curves could be used to specify appropriate limiting combinations of burnup, enrichment, and cooling time which produce dose rates equivalent to the design basis fuel. 1.2.6- DSC Top End Dose Rates This specification should be deleted. Given that HSM dose rates are specified, a specification for DSC dose rates is not necessary since only the workers involved in the canister closure operations are affected by them. Historically, dose rates for operations within an operating nuclear facility, have not been limited by Technical Specifications since they do not directly impact public safety. Unlike HSM dose rates which affect exposures at the ISFSI boundary, DSC dose rates primarily affect worker exposure- not the general public. Routine ALARA practices ensure that worker exposure is routinely monitored and controlled. Also, these Technical Specification limits do not provide for "hot spots" that have been observed in actual loading experience. These are localized effects and may be caused by factors such as impurities in the fuel structural materials. These hot spots could exceed the limits, yet have no impact on worker exposure due to their size, orientation, or location. As a minimum, the Action section should be amended as follows: Action b. Visually inspect placement of top shield plug . Re-install or adjust position of top shield plug if it is not properly seated. (This action should not rewiire re-installation or adjustment of the top shield plug. A top shield plug which is not correctly seated is unlikely to result in high centerline dose rates, and adjustment should only be done with great caution. Such action should be taken only if the need is clearly indicated following a thorough examination, and so, should not be mandatory if the need is not clearly indicated.) Action c. Install additional temporary shielding or implement other ALARA actions, as appropriate. (As currently written, this action requires the installation of additional shielding without consideration of the practicality of doing so. The welding machine is normally mounted to the DSC shield plug or top cover plate, or to a shielding device which is similarly mounted. The addition of more shielding material may be precluded by welding machine
operability requirements. Also, adding shielding is not the only means to reduce worker exposures- other compensatory actions such as enhanced monitoring, worker briefings, and decreased stay times could produce the same desired result. Therefore, additional shielding should not be mandatory.) Action d., which requires submittal of a report to the NRC, should be deleted. The user should be permitted to analyze and document the higher dose rates under 10CFR72.48, which is available for NRC review. (Please see comments for Section 9.) 1.2.8- HSM Maximum Air Exit Temperature The applicability section requires the user to calculate the temperature rise for each HSM loaded with canisters producing less than the design limit of 24 kW. This requirement should be eliminated for the following reasons:
- 1. Users are not normally provided the vendor's analytical models for this calculation.
- 2. The 100 degree F. rise calculated for the design basis maximum heat load ensures that all safety limits are met for concrete and fuel.
- 3. Since 24 kW is the limit, virtually all of the HSMs will be affected. This places an undue burden on the user, considering the inherent safety margins of the system.
- 4. Technical Specification 1.3.1 ensures that air flow is not blocked, so a false measurement of low temperature rise cannot occur.
1.2.11- Transfer Cask Dose Rates This Technical Specification should be deleted. Given that HSM dose rates are specified, a specification for Transfer Cask dose rates is not necessary since only the workers involved in the storage operations are affected- not the general public. Also, this specification does not provide for "hot spots" that have been observed in actual loading experience. (Please refer to the comments on Technical Specification 1.2.6.) As a minimum, the Action section should be amended as follows: Action: If specified dose rates are exceeded, place temporary shielding around the affected areas of the transfer cask or implement other ALARA actions, as appropriate. Review the plant records of the fuel assemblies which have been placed in the DSC to ensure they conform to the fuel specifications of Section 12 .1. The requirement for submittal of a report to the NRC should be deleted. The user should be permitted to analyze and document the higher dose rates under 10CFR72.48, which is available for NRC review. (Please see comments for Section 9.)
1.2.14- TC/DSC Transfer Operations at High Ambient Temperatures The requirement to install a solar shield for HSM loading with ambient temperature above 100 degrees F., is unnecessary. Transfer of the DSC from the Transfer Cask into the HSM would normally require less than eight hours. Any increases in fuel clad temperature and neutron shield temperature resulting from solar heating with ambient temperature in excess of 100 degrees F., would be small and short in duration- therefore, not detrimental. Additionally, the Transfer Cask cavity is open to the atmosphere and would not pressurize. 1.2.15- Boron Concentration in the DSC Cavity Water (24-P Design Only) Limit/Specification The requirement for a dissolved boron concentration in the DSC cavity of 2000 ppm is in excess of the 1810 ppm requirement for the site-specific license and should be reduced accordingly. The site-specific requirement for 1810 ppm dissolved boron is sufficient to ensure reactivity is below 0.95 K-eff (95/95 tolerance level with uncertainties) assuming 24 ~ fuel assemblies. Even with the additional unlikely condition of worst case water density- 0.2 to 0.7 gm/cc (a condition not achievable for fresh fuel)- reactivity remains below 0.98 K-eff. Since only irradiated fuel assemblies will be loaded into the DSC, reactivity will actually be well below limits. Surveillance 3, Verification of the dissolved boron concentration in the pool is required every 48 hours while the DSC is in the spent fuel pool. This surveillance frequency is much greater than required for spent fuel pool storage racks with burnup credit. A surveillance frequency of once per month has been found sufficient for spent fuel pool operation under Part 50. This is justified due to the unlikelihood of spent fuel pool deboration. This surveillance requirement for DSC loading should be changed to a frequency of once per month. 1.3.2- HSM Thermal Performance This specification requires that temperature measurements be made daily for each HSM. Daily temperature measurements are not necessary to ensure convective air flow, given the requirement under 1.3.1 to verify that the inlet and outlet vents are not obstructed. The site-specific licenses for the NUHOMS system ensure adequate thermal performance by requiring temperature measurements when the DSC is placed into the HSM, 24 hours later, and again at 1 week after loading.
If any of these comments need clarification, please call me at 704-382-6778. Gary R. Walden, Engineer Nuclear Engineering cc: K.S. Canady R.W. Rasmussen E. D. Price J.R. McLean
16 N. Carroll St., Suite 300 Madison , WI 53703 (608) 251 -3322 CITIZENS' UTILITY BOARD~fs~ER rlo
*94 AUG 16 P4 :46 DOCKET NUMBER R12 August 15, 1994 PROPOSED RULE p (s<JF R 26'-Lf 96)
Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Docketing and Service Branch Re: NRC 10 CFR Part 72 List of Approved Spent Fuel Storage Casks: Addition Standardized NUHOMS Horizontal Modular Storage System NRC Proposed Rule
Dear Sir:
Please find enclosed Wisconsin Citizens' Utility Board's comments on the above proposed rule. Sincerely, ~ Dennis Dums Research Director Enclosure SEP 2 8 1994 Ar., r.Jw edged by card ...........- ...._.....,_.
DOC t<cTIMj & SERVICE SECTION
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Nuclear Regulatory Commission Proposed Rule 10 CFR Part 72 List of Approved Spent Fuel Storage Casks Standardized NUHOMS Horizontal Modular Storage System The Wisconsin Citizens' Utility Board, by Dennis Dums, Research Director, hereby submits comments on the above referenced matter. Comment 1: The proposed rule appears to increase the potential for adver s e e ffe c t on the environment and on public health and safety by allowing the certificate holder (VECTRA) to make changes in the cask design and/or procedures described in the Safety Analysis Report (SAR), and conduct tests or experiments not described in the SAR without prior Commission approval if VECTRA determines that the prop osed change, test or experiment does not involve a change in the Certificate of Compliance (CoC), an unreviewed safety question, a si gn i fi c ant increase in occupation exposure or a significant unreviewed environmental impact. (see Draft Coe, 9. Changes, tests, and experiments). One of the conditions imposed upon VECTRA in making its determination as to whether a proposed change in cask design or procedure, or a particular test or experiment not described in t h e SAR involves an unreviewed safety question is: (3) If the margin of safety as defined in the basis for any technical specification or limit is reduced. This appears to the lay reader to imply that VECTRA can make determinations regarding the impact a proposed change, test or experiment would have on the safety limits established in technic a l specifications that are specific to a Part 50 licensee. Furthermore, VECTRA "shall maintain" records of changes in design or procedure and records of tests and experiments "it" conducts that are not described in the SAR. And, "These records must contain a written safety evaluation that provides the bases for the determination that the change, test or experiment does not involve an unreviewed safety question." It appears that VECTRA acquires the same right to make decisions regarding site specific public health and safety and environmental impacts that holders of a Part 50 license have under 10 CFR 50. 59. How can the Commission assure the public that VECTRA's ( as a self interested cask vendor) use of this power will not adversely effect public health, safety, or the environment?
Comment 2: The Draft Environmental Assessment and Finding of No Significant Impact for the proposed rule is inadequate. The Commission's Draft EA and FONSI is premised upon the use of a cask that meets the conditions specified in the Coe (seep. 4 Draft EA FONS!) which itself is premised upon the Commission's SER of VECTRA's SAR. Yet, Coe 9 provides the opportunity for changes in cask design and procedure, as well as tests and experiments not described in the SAR. How does the Draft EA and FONSI address the possibility that changes in cask design and changes in procedure could be made, and that tests or experiments could be conducted under Coe 9. Changes, tests, a nd experiments that create an unreviewed safety question that goes unidentified by VECTRA or any other entity be it the utility or the NRC, leading to the use of a cask that does not meet the conditions specified in the Coe and which may adversely impact site specific public heal th, safety, and the environment? The requirement that VECTRA submit reports only on an annual basis to the NRC regarding changes, tests, and experiments made under Coe 9 attests to the potentially large number of casks that could be placed into service that may not meet the conditions specified in the Coe.
The Utility Radiological Saf~ 5 d of Ohio
*94 AUG 16 P4 :49 August 15, 1994 The Secretary U.S. Nuclear Regulatory Commission ATTN: Docketing and Service Branch DOCKET NUMBER Pl PROPOSED RULE...:..::.--:---
12 Washington, DC 20555 (SCf FR 2S!Lf'16)
Dear Secretary:
The Utility Radiological Safety Board of Ohio (URSB or Board) is pleased to forward the comments of its Citizens Advisory Council, in the matter of the Commission's proposed rule to add the NUHOMS Horizontal Modular Storage System to the list of approved Spent Fuel Storage Casks. The URSB unanimously endorsed these comments at its quarterly meeting on August 8, 1994 and offers them for the Commission's consideration. The URSB was established by the Ohio General Assembly in July of 1989. The Board's purpose is to:
"... develop a comprehensive policy for the state regarding nuclear power safety. The board's objectives shall be to promote safe, reliable, and economical power; ... and recommend policies and practices that promote safety, performance, emergency preparedness, and public health standards that are designed to meet the state's needs."
Additionally, Ohio Revised Code Chapter 4937 requires the Board to:
"... make recommendations to increase cooperation and coordination among the member agencies toward the promotion of nuclear safety and mitigation of the effects of a nuclear electric facility incident ... and serve as the clearinghouse for recommendations made by independent organizations or private citizens and review and investigate such recommendations."
The URSB consists of six state agencies: the Ohio Departments of Agriculture, Health and Industrial Relations; the Ohio Emergency Management and Environmental Protection Agencies; and the Public Utilities Commission of Ohio. The URSB is advised by the URSB Citizens Advisory Council on Nuclear Power Safety (CAC) on technical and public safety and health issues related to the operation of Davis-Besse, Perry, and Beaver Valley nuclear power plants. The CAC is composed of local government officials, academics, representatives from environmental organizations, scientists, nuclear and health professionals, and
, " SEP 2 8 1994 d *-r ......... "'.....
The Utility Radiological Safety Board of Ohio - 180 E. Broad St. - Columbus, Ohio 43266-0573 Department of Agriculture. Department of Health - Department of Industrial Relations - Emergency Management Agency Environmental Protection Allencv
- The Public Utilities Commission
DOCK!: I !Nu o ~tt11/1CE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics stmar1( Date fr: > l I I Lf L-/ CoJjes Received _ _ __ _ _ __ I.defI Copies Reproduced _ L"""'"/_ ___,_ _ Special Distribubon ;z r;q /1!24 G1.,,...,n,d,-,.? lf-- ~ -- . !y; o!~ ~ ~ -
August 15, 1994 'Page citizens residing near the nuclear power plants. The primary objective of the CAC is to represent diverse views and ideas on the safe operation of nuclear power plants and to bring these views and ideas to the URSB. In June of 1996, the spent fuel pool at the Davis-Besse Nuclear Power Station will reach full capacity with full core reserve. The plant has decided to dry store its additional spent fuel using the NUHOMS 24-P horizontal module storage system. The CAC has reviewed the vendor's Safety Analysis Report and the NRC staff's Safety Evaluation Report and extensively discussed many of the features of the standardized design. As a result of this review, the CAC drafted the enclosed observations and recommendations. Thank you for your consideration. Please contact me at (614) 466-4821 should you have any questions regarding these comments. Sincerely, EAB:jc
Comments on the Nuclear Regulatory Commission's Proposed Rule to Amend its Regulations to add the Standardized NUHOMS Horizontal Modular Storage System to the list of Approved Spent Fuel Storage Casks. as prepared by the Utility Radiological Safety Board of Ohio Citizens Advisory Council and Endorsed by the Utility Radiological Safety Board of Ohio August 8, 1994
The Citizens Advisory Council (CAC) to the Utility Radiological Safety Board of Ohio (URSB) is pleased to offer the following comments to the Nuclear Regulatory Commission (NRC) regarding its proposed rule to add the NUHOMS Horizontal Modular Storage System to the list of approved spent nuclear fuel storage casks. The CAC asks the NRC to consider these comments in light of the fact that the spent fuel pool at the Davis-Besse Nuclear Power Station in Oak Harbor, Ohio will be full in June 1996. Toledo Edison Company officials, the operations of the Davis-Besse Station, plan to use the NUHOMS cask system to store spent nuclear fuel at the plant site beginning in the Fall of 1995. DISCUSSION Temperature Monitoring The CAC has reviewed the Safety Analysis Report (SAR) for the NUHOMS cask system and is concerned with the thermal performance of the casks. The NUHOMS casks are fabricated of steel. The shield modules are fabricated of concrete. Concrete contains significant amounts of water bound within the structure. When concrete is elevated to temperatures significantly above the boiling point of water, 212F, that bound water will evaporate. With evaporation, the concrete is degraded and hence loses its structural integrity. Thus, elevated temperatures are detrimental to the safety of these devices. The casks are cooled by natural convection--the heat generated by the cask heats the air which then rises due to density differences between the heated air and the surrounding cooler air. Any blockage of air flow through the small passages at the bottom and top of the storage module will increase the air temperature and subsequently increase the cask temperature and concrete temperature. Although calculations in the SAR show that excessive temperatures will not be reached in the casks, the CAC believes it is prudent to monitor temperature and air flow to ensure that temperature excursions are not experienced. Cask Drop Accidents The CAC reviewed the November 5, 1993, Certification Safety Analysis Report (CSAR), and found that in Revision 2, Section 8.2.5., accidental cask drops were discussed. Also, in Section 8.2.5.1-A, Cask Handling and Transfer Operation it is noted that the NUHOMS Application 'for Certificate of Compliance does not fully address potential cask drops related to handling the cask inside the plant's fuel handling building. Potential cask drops include utilizing the plant's crane for placement of the DSC into the cavity of the transfer cask, lifting the transfer cask/DSC into and out of the plant's spent fuel pool and placement of the transfer 1
cask/DSC onto or off of the transport skid/trailer. In Section 8.2.25 of the CSAR, it states:
"The postulated drop accident scenarios should be consistent with those currently addressed in the plant's 10 CFR 50 licensing basis for handling of a shipping cask. Such postulated accidents are plant specific and should be addressed by the licensee."
Also, Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants identifies, in Table 15-1, spent fuel cask drop accidents as applicable events to analyze; and Table 15-4 identifies parameters and information to be provided. In NUREG-0800, Standard Review Plan Section 15.7.5 addresses acceptance criteria for spent fuel cask drop accidents. Many licensee's Final Safety Analysis reports (FSARs) and subsequent NRC Safety Evaluation reports (SERs) used Regulatory Guide 1.70 and NUREG-0800 to describe and accept spent fuel cask drop accidents. Other licensees completed their analyses and were licensed before these NRC documents applied. A CAC member is aware of one licensee that has deleted the spent fuel cask drop accident analysis from their FSAR. In the Draft Certificate of Compliance for the Standardized NUHOMS System for Irradiated Fuel, Section 1.1.1, Regulatory Requirements for a General License, identifies as a minimum the "site-specific parameters and analysis, identified in the SER, that will need verification by the system user." The recommendations in the Draft Certificate of Compliance do not include the postulated cask drop accident in the plant fuel handling area. -- Security, Fire, Explosion and Sabotage - The SAR explicitly addresses the following postulated accident events:
- Reduced HSM air inlet a~d outlet shielding
- Tornado winds and tornado generated missiles
- Design Basis earthquake
- Accidental transfer cask drop with loss of neutron shield
- Lightning effects
- Debris blockage of HSM air inlet and outlet opening
- Postulated DSC leakage
- Pressurization due to fuel cladding failure within the DSC The analyses show that the dry cask storage facility can accommodate these accidents without presenting undue risk to the public. However, the SAR did not include consideration of aircraft crashes, turbine missiles, external fires and explosions, or sabotage.
2
An aircraft impact into the dry cask storage facility is an obvious challenge that should be addressed. While the probability of such an event would depend on the specific location of the facility, the generic safety analysis should address the capability of the facility to withstand such challenges. The likelihood and potential implications of aircraft crashes are typically considered in nuclear power plant safety analyses. Similarly, turbine missiles have been experienced at power plants and their implication should also be considered. The electric power industry has performed substantial research on turbine missile hazards. Also, the potential for sabotage such as vehicle bomb threats, is an issue of obvious concern to the public. The reference report did not include consideration of sabotage, aircraft crashes, turbine missiles, or external fires and explosions. Such potential hazards should be recognized and their implications to the proposed design discussed. RECOMMENDATIONS Temperature Monitoring
- 1. Accept Section 1.3.1 of the Draft Certificate of Compliance, Visual Inspection of HSM Air Inlets and Outlets (Front Wall and Roof Birdscreen) as written, without modification. In this section the NRC requires daily visual surveillance of the exterior of the air inlets and outlets.
- 2. Accept Section 1.3.2 of the Draft Certificate of Compliance, HSM Thermal Performance, as *written, without modification. In this section the NRC requires that verification of a temperature measurement of the thermal performance for each HSM be done daily.
Cask Drop Accidents
- 1. Verification is required that site specific plant fuel handling cask drop accident analysis exist and is applicable to the NUHOMS DSC. This verification should be a part of a site specific 10 CFR 50.59 safety evaluation or license amendment application. The NRC Certificate of Compliance for the Standardized NUHOMS System should include this analysis in the list of minimum site specific requirements.
- 2. The criteria for acceptance should be consistent and applicable to potential situations during fuel handling operations in the spent fuel pool area.
3
Security. Fire. Explosion and Sabotage
- 1. The generic safety analysis should address the capability of the facility to withstand such challenges as aircraft crashes and turbine missile impacts.
- 2. As a general rule, it is recommended that the cask system be placed in a plant protected area where it is subjected to the vehicle bomb threat rule.
However, issues of security would be better left to the site-specific safety analysis rather than discussed in the generic document. 4
B-lts-l:14 8:04 AM ;on.ro ATTOR.NE~ GE:{ERAI.. DOCKET NUMBER Attomey General PROPOSED RULE p /2 DOCKETED Lea *Flaher USHRC (SCJFR 2~C/6)
*94 AUG 16 P 1 :59 Auguat 15, 1994 OFFIC OF SECRtTAR'1 DOCKET! G &. SfHVICE BRAHCt-1 The Scretary ATTN: Docketing and Service Brenoh U.S. Nudear Regulatory Commission Wahlngton, D.C. 2015156 RE: Propo,ed Rule on Dry Caak Storage of Spent Nuclear fuel, 69 Fed. Reg.
21488 (June 2, 914) Mr. Secretary: Twelve citizen group, repre11nting thoueands of Ohioan, have asked the Nuole*r Regulatory Commie*ion to extend undl September 30, 1994. tha publlc comment period for the NRC's propo1el to amend ragulatlon1 to add the Standardized NUHOMS Horlzontal Modular Storage 8y1t1m to their list of approved spent full storage caw. I ura* the NRC to approve thie reqult. I have attached the letter of August 9, 1994, 1lgned by the Ohio Si*rr* Club. People in Ohio need more time to make Informed comment* on th* ctltloal iaue of storage of pent nuclear fuel in dry casks. The Auguat 9 letter identifies several reaeon* for extending th8 comment period. I look forward to NRC"* favorable con*lder*tion of thla request. SEP 2 S 1994 - Ac nowledgArl card .................................. State Office Tower / 30 East Broad Street / Columbua. Ohio 4321e-3428 I ...
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8-16-94 8: 54 AM ;OHIO .~TTORNEY GENERAL 6146446~35;~ 3/ 4 FRCl"1 : VISTULA !'R-RE1!!NT CO, Pl-DE t-D, : ,1G+3315+910'a ~;. 15 1994 06 :1?PM P0J Sffll Docldl J.anding Dr. Willoupl,y H ii& OH 44094 Auguot 10,109'
'rh, SICNtlry U.S. NudNr Rtaulatory Comnliuion W~t>C20555 ATrN: DDc:ketlng Ind Serrice Drane!h RD: ),me 2, 19H Fed*al Rapt* Notice 28'96 NUHOMS Calk UP-528 Mr.Somm,y.
'fht twelw citlfADI groupe liated Wow requ11t that you grant an atenlion 0£ the CIDIIUlldpmoc:l cm the Nudau Regulato,y Comlnblion's propocitian to am.,d ltl nsuJationl t.> add the Standardiaed NUHOMS Horizontal Modular Storage &y.tam to the Ult ol ApprvYecl Spent Putl Storage Cub. We rcquon that you 9'Cttnd the comment period whie!h tJlda on Aupat J6, f 9N to Septtmber 30, 1994.
~-eight doNlnants mntainlng pmpriltuy information were withheld &on, pubHc:
cffKrJoeun far W.clay* after tho P81tt*l annmantpmiod b.gan. '11\erebo, h NRC ill obligatld ID ..._d the olfidll comment period. Thelo docmMl\11, which were withhold from publk-ctwc.n 1111tiJ June J6, 1~ contain informatian that .IDnM th* 1MN8 lor the ftMl rule fn therumalctng to add tho Vectm Cmnpany'* NUHOMS-26' and NUHOMS-528 cub to the "lJlt of Appzovcd Spoilt Pull Stang* Cub" eJipble for UN undlr the
,.,.,.1 BCINMt in Sut-pal't K of JO OR Put 12. The requireentl ollOCPR 2.'90(e) ~ NRC regulation.I pnwide that infannation am,ittlcl in I rv1e ffllldna pmc:eNm& '#hich IIQt.quently f'onl1I by tho ComnuMian.....
the.,_. r- the &al tulo wm not ho wfflthald hlll ,-He dilclOfW'I Monio,*, the Ndou111* and cmnplexttiel of illu* invol*ed with on*altl dry cuk atorap *mant
- mdallicm of th* mmmet,-iod. Thapabllc nNdt more an. tn order to make inloftucl COIIID\Gllla on tho addition of tho Sta1ldardt&ecl NlJHOMS s,.t.n. On Auguet 19, J9M die Utility Jladiolopal Sl"1y 8olrd of Ohio* lpOIIIOdllg
- worbhop antftftd 1iqh Lewi ladtoadtY* Waste." *J'he worbhop II dll'le dayt after the oflclal camtnent period endl. laformatiOI\ Ylluable 1n llUlk1ftl tntema*t eommcmts may bu plMd by tilt pubic at tlUI WOlbhopwhic:11 makoe an oxtelllltft demablt and wamntN.
Punlllnnan, _.. . c:gn"'"" which clamand In-depth ahldy have bean rat* l,y the drl.A c..rtfflate of Ownpli1nm fmo the St1ftdardiMd NU1 JOMS ca!lk. The NRC '* attempt
8-16-94 8:54 AM ;OHIO ATTORNEY GENERAL 6146446135;# 4/ 4 AU1 : vl STuL.A ~ CO, ~,. 1~ 1994 06:1BPM P04 to iuert tho languqe and intent of 10 CJlR 12.'8, Which pertains to &partftc UcenNt, mtv th* Certificate of Compliance fot the Standarcliztd NUHOMS it av--, aeriou11 iatuo whim wmanta muc:h thought and l'aaMrdl on the part of tlw general pubJSC. bl Bght of the fact that the NJlC hu ,eon fit to avoid public hurinp and ID prc,cood with on-tl dry e11k ltaragt without ai.te IDedftc 111vlronsn111tal impact atatoD.\11111, it cmtainly aoom11 pnnlent fot th* health aralMloty pf th, g<maI pub& that tM NRc dow th publiil an IXtanlion of tb, CX>mll'Mll'lt ptriOd. In ~ew of tho above COll8idsafiol'II. th* twelve environmental grou~ Ji.tad 'below requeet an extlnliOn of the offidAI c:mnmentplriod on tho 8tndardiud NUHOMS dry CIM ayattm to S.ptmnber30, l9H. PIIUI inform Alb J-Jtrt or Connie JCline DI your - dedlton reprdina our nqUllt aa 800I\ u poeaiblt. On behalf of tlle following grwpt, Conm* XUne, Sama Club FAX - 216- 946-9012 AUm Hirt, coalition for Salo BNrgy *PAX-6lt,..S35-,8100
-c.onnle Klme, Sierra Club Paula ROIi, Ohio Ctizen Action
*Cirlton Wame, Consu1nor1 League of Ohio c.c,Jyn ~ Don't Waat. Ohio "David Bllieon, Ohio Cl'COl'\I Steve Gannie, Ohio Oti1.en1 Against e Radioactfvc Environmunt ICmth Haddad_ Border Oppoeecl to Nuclaer Dump fF}{arvey W118e1'1Mt\ Ciroenpeaat Shirley Tmnaaollo, Lake Brie Allianm
-David Hugha, Ccmeernecl Otizenl of CUyahoga, Alhtabul-.. Lake and Ceaga Onmtlea Bridgette Mariea., a.to Bnvionmental Council
- Affiliation for idanWkdcm purpoeo cmly Sent via F*Climlll ancl U.S. Mall
6146446135;# 1/ 4 8-16-94 8:54 AM ;OHIO ATTORNEY GENERAL AHomey General L-* Fl1her FAX COVER SHEET Fax No.: (614) 644-6135 Attention: Mr. Clwlco J. Haught,ey1 Qucf
. Storage and Tnn1pmt Syatema Drench
. TO: Divlaion of Indu1trial and Medical Nuclw_Safoty , NMSS
- - UnitedStatttaNu~legulataryc.ommladcm
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WMhiflaton,O.C.2MS! PAX-30115-9369 . .
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TELEPHONE.NO * .,.1:4.-..;;14::ar;;,_.... lPECIAL INSTRUCTIONS: RB: June 2, 1994 Federal Reglater Notice 28496-NUHOMS C..ak 24P-.52B Deadline Bxtenslon Requeet of the ex>mment period ft>r the Standardized NUHOMS Horizontal Modular Storage Syatem IP YAM PP HAI BIAIIYI 6HY .o, D1C PNIII PBAPMLY, Pf 1AM CQNTACT MMQQ/CALL MG! PlftlPN 41 SOON Al POIIIBLE, MOTi: PLIA81 MAKI ALL CONTACT WITH IINDIIIICALL IACK P!ll80N. IP YOU ARI l!NDING INPOIIMATION, mu, NYlt OONTACT UQEIYIB RfPBI RNQ. State Office Tower I 30An !ut Broad Stre- / Colu,mbu1, Ohio *15-3428
&,1.1111 n-*-~ __ ,_ __
ere CENTERIOR ENERGY 6200 Oak Tree Boulevard Moil Address DOCKETED USNRC Independence OH PO Box 94661 *94 AUG 15 P1 :07 216-447-3100 Cleveland . OH 44101-4661 DOCKET NUMBER Pl
,.i~ . ,- . ,,,er-OFFI CE OF SECR~ TAt1Y Doc E-.* 1,L. ~- ,L, . . .
Bh.1\; 'Cf. PROPOSED RULE-=-==----"-- - - - Docket Number 50-346 ( S Gf f g 2. lr'-IC/6) License Nu~ber NPF-3 Serial Number 2246 August 12, 1994 The Secretary, U.S. Nuclear Regulatory Commission Attention: Docket ing and Service Branch Vashington, D. C. 20555
Subject:
Comments on Proposed Rule to Add the Standardized NUHOHS Horizontal Modular Storage System to the List of Approved Spent Fuel Storage Casks Gentlemen: On June 2, 1994, the Nuclear Regulatory Commission (NRC) published in the Federal Register (59 FR 28496) a proposed rule to add the Standardized NUHOHS Horizontal Modular Storage System to 10 CFR 72.214, List of Approved Spent Fuel Storage Casks. Comments on the proposed rule were solicited by the Federal Register Notice. Toledo Edison welcomes the opportunity to comment on the proposed rule. Toledo Edison commends the NRC for moving forward with the certification of the standardized NUHOHS storage system which is selected for use at the Davis-Besse Nuclear Power Station (DBNPS) to augment on site fuel storage capacity beginning in the summer of 1995. Utilization of the NUHOHS system at the DBNPS will enable Toledo Edison to maintain sufficient capacity in the spent fuel pool to completely off load the reactor core following the refueling outage scheduled in early 1996. Because of this need, Toledo Edison encourages the NRC to complete this rulemaking in an expeditious fashion. Toledo Edison has reviewed the proposed Certificate of Compliance and Draft Safety Evaluation Report for the Standardized NUHOHS system. Toledo Edison's comments are attached. Operating Companies: SEP 2 8 1994 Cleveland Electric Illuminating Toledo Edison Acknowla 'ged by card ..................- ........_..
1/.S. NUCLt ~s~ HH,ULA I vH y COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics 1'ostmar1< Date :z_ / '-I Copies Received _ _ _~/_ _ _ __
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Docket Number 50-346 License Number NPF-3 Serial Number 2246 Page 2 Should you have any questions or require additional information regarding Toledo Edison's comments, please contact Hr. Villiam T. O'Connor, Manager - Regulatory Affairs, at (419) 249-2366. Very truly yours, ~~ Vice President - Nuclear Davis-Besse Nuclear Power Station PVS/eld Attachment cc: L. L. Gundrum, NRC Project Manager J.B. Hartin, Regional Administrator, NRC Region III S. Stasek, DB-1 NRC Senior Resident Inspector USNRC Document Control Desk Utility Radiological Safety Board
Docket Number 50-346 License Number NPF-3 Serial Number 2246 Attachment Page 1 Comments on the Proposed Certificate of Compliance Page A-4 The requirement to load the first canister with a design basis fuel load should be modified to allow more latitude in the selection of fuel. It is likely that Toledo Edison will be the first use of the standardized NUHOMS system. Although, Toledo Edison currently has design basis spent fuel, our philosophy is to perform dry casking of the oldest fuel first, thereby minimizing both heat load and radiation dose. This philosophy is also consistent with our public information campaign which has always stated that the oldest fuel would be loaded first. The requirement for the first user to load design basis fuel seems contrary to the As Low As Reasonably Achievable (ALARA) principle. Page A-6, first paragraph in Bases The reference to Tables 12-la and lb appears to be in error. The reference should be to Tables 1-la and 1-lb. Some clarification of the integrity of spent fuel permitted to be stored is r equested. It should be made clear that f uel need not be specially inspected to be eligible for storage, that pinhole leaks do not constitute gross breeches, and that fuel is eligible for storage provided that it does not require special handling or storage provisions within the spent fuel pool. Page A-8, Assembly Length The specification for assembly length refers to Safety Analysis Report (SAR) Chapter 3. Chapter 3 addresses the length of as built new fuel. Irradiated fuel will be longer than new fuel. This should be clarified. Page A-19, Basis paragraph The sentenc~ in the middle of the paragraph that begins "Acceptable damage may occur *** " should read "Unacceptable damage may occur *.* " Page A-25 The Action for Technical Specification 1.2.16, Yearly Average Ambient Temperature, appears to be a surveillance requirement versus an action statement. It is unclear what action should be taken if either of the two specified limits (Yearly average temperature< 70F or average daily ambient temperature< lOOF) is exceeded.
Docket Number 50-346 License Number NPF-3 Serial Number 2246 Attachment Page 2 Page A-27, first paragraph The first paragraph states that the postulated adiabatic heatup would result in concrete temperatures being exceeded in approximately 40 hours. As a result, it is appropriate and conservative to perform the visual s urveillance to verify no vent blockage on a daily basis to ensure that a blockage has not existed for 40 hours. The last sentence in the first paragraph should reflect that the module needs to be removed from service if it cannot be established that the blockage is less than 40 hours, not 24 hours. A 24 hour surveillance interval will adequately verify this. Page A-28 Section 1.3 indicates that a module must be removed from service if a vent blockage is in existence for greater than 24 hours. Surveillance 1.3.2 indicates that a module must be removed from service if the concrete accident temperature criteria has been exceeded for greater than 24 hours. A vent blockage for less than 24 hours would not cause the temperature limit to be exceeded, as explained in Section 1.3 and the objective for the 24 hour frequency required by Surveillance 1.3.1. The apparent conflict between Section 1.3 and the action for Surveillance Requirement 1,3.2 needs to be resolved. It appears that the Surveillance Requirement 1.3.2 Actions are appropriate.
DOCKET NUMBER PROPOSED RULE _ _ _ __ PR 72 (SC/ FR 2 itCf 6) DOCKETED ~ USNRC i.;.!) August 7, 1994
*94 AUG 15 P1 :06 Secretary OFFICE Or SE RETARY U.S, Nuclear Regulatory Commission DOCKE TING "* sr-RViCE Washington, D.C. 20555 BRANCH ATIN: Docketing and Service Branch Mr. Secretary, I wish to comment on the Nuclear Regulatory Commission's (NRC) proposal to amend it's regulations to add the Standardized NUHOMS Horizontal Modular Storage System to the List of Approved Spent Fuel Storage Casks.
- I strongly object to the NRC's attempt to insert 10 CFR 72.48 into the draft Certificate of Compliance 1004 for the Standardized NUHOMS Systems. The only way that 10 CFR 72.48 may be involved is in a site specific license. The use of 10CFR 72.48 by a General License under Subparts K or L to modify an SAR or a C of C is not possible. Since the Cod e is silent on a process to change a generic cask design after the final rule, the NRC must use a rule making procedure which provides for public comment and proprietary release. If the draft C of C for the standardized NUHOMS is approved containing #9 , then it is obvious that the whole rule making for general cask acceptance is nothing more than the NRC's attempt to avoid a site specific licensing procedure. The inclusion of #9 in the standardized NUHOMS C of C means the cask is not generic. Furthermore, vendors of casks approved of under Part 72 Subpart L cannot utilize 10CFR 72.48 to modify their SARs because vendors of casks are not licensed under lOCRR Part 72.40. The site specific license provisions of Part 72 apply to operators of a spent nuclear fuel storage installation, not to vendors of casks. I also strongly object to the fact that the NRC staff, obviously concerned about the thermal characteristics of the standized NUHOMS system, would recommend that the C of C for the NUHOMS be approved but that after its approval, require a thermal performance verification for the first standardized NUHOMS system to be used(SER 4-1) These NRC concerns about the thermal performance of the standardized NUHOM S were clearly expressed by Mr. Frederick Sturz in a 12/15/1992 letter to Mr. William McConaghy of Pacific Nuclear Fuel Services, "We believe present HSM thermal analysis has removed significant conservatisms to the point where it may be slightly non-conservative, and therefore, little if any significant safety margin remains." These concerns should be totally resolved before the C of C is approved and the standardized NUHOMS is added to the list of approved casks. Approving the C of C before these issues are resolved is totally irresponsible on the part of the NRC I also object to the fact the transfer cask will be leased to the utilities which are using a standardized NUHOMS cask system and that VECTRA will own the transfer cask. The consequence of this situation is that in case of an immediate need to remove the DSC from the HSM , the transfer cask may very well not be on site to remove the irradiated Acknov~ ,edged by card ......................... SEP 2 8 1994 ..---
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., r.'v1tCt: SECTION OFFICE Of rriE SECRETARY OF* 11HE COMMlSSION Document Statistics Poslmark Date l
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fuel. Inclement weather could make the transfer cask unavailabe for some extended period of time. This situation is not acceptable for the health and safety of the public. I also have serious concerns about the horizontal storage of the irradiated fuel rods in the standardized NUHOMS . The Feb. '94 Wisconsin Draft Environmental Impact Statement for the Point Beach Nuclear Power Plant Projects raises serious concerns that the "horizontal storage may cause bowing/ sagging of the rods and there may be a bonding of the fuel cladding/ crud and guide sleeve as they are in weighted contact for the entire storage period." Vectra's response to this concern that "there is no mechanism for "bonding " of the fuel cladding and guide sleeve given the dry inert atmosphere inside the canister ... "(3/ 15/'94 letter to Mr. F. Sturz from Vectra) does not adequately address this most serious concern. Furthermore, in a May 1993 study prepared for the RC by the Center for Nuclear Waste Regulatory Analyses of San Antonio, Texas, serious questions are raised about the consequence s of dry storage on fuel cladding. The report states that, "the dry environment has the potential of producing such problems as further fuel cladding oxidation, increased cladding stresses and creep deformation as a result of rod internal pressure ... These possible spent fuel and cladding alteration modes could be quite accelerated under dry storage conditions, since the temperatures are much higher that in wet storage." I do not believe that given these concerns about potential degradation of the fuel cladding the NRC is fulfilling its obligation to see that "spent fuel cladding must be protected during storage against degradation that leads to gross rupture"(Section 72.1222(h) on "Confinement Barriers and Systems") I also object to the supposition that the ISFSI basemat is not important for safety. For the NRC to state that the "foundation only has nuclear safety implications in the event of gross failure, since it is structurally independent of the supported HSM" (SER 2-14) and therefore the HSM foundation design criteria are not included in the SAR(SER 12-20 , is nothing more than an irresponsible act on the part of the NRC to avoid a site specific licensing procedure at each reactor site where the standardized NUHOMS casks will be used. If the foundation needs to be site specific, then the entire cask system should be site specific and the foundation should absolutely be covered by the SAR. The foundation is obviously very important for safety. I object to the RC's assumption that the standardized NUHOMS cask is designed to withstand any "credible" seismic event that may occur east of the Rocky Mt. front except in areas of known seismic activity. The standardized NUHOMS HSM was analyzed for a peak horizontal ground acceleration of 0.25 g and a vertical acceleration of 0.17 g. The standardized UHOMS cask is being proposed to be used at Davis Besse in northern Ohio. According to a new Educational Leaflet (No. 9) released in 1994 by the Division of Geological Survey, "On the basis of historic seismic activity, it is likely that large earthquakes with epicenters in the state would occur in the western seismic zone or in northeastern Ohio; ... Some researchers have suggested that north-eastern Ohio is capable of a maximum 6.5 magnitude earthquake, whereas western Ohio may be capable of producing an event in the 6.0 to 7.0 magnitude earthquake (maximum MMI of IX). The leaflet goes on to state that" the widely used earthquake hazard/risk map produced in 1969 by the U.S, Geological Survey ... does not incorporate new knowledge of potential earthquake-generating structures in Ohio's basement rocks and a probable greater earthquake hazard in northeastern Ohio." It is irresponsible for the NRC to put these "generic" cask at reactor sites without site specific, up-to-date environmental impact
statements. Furthermore, it is entirely negligent on the part of the NRC to refuse to do such an environmental impact statement in view of the fact that the Federal government has no repository for this high level radioactive waste and consequently these reactor sites will quite possibly become "permanent" high level radioactive waste dumps. End of comments, Alice H. Hirt 2243 Robinwood Toledo, Ohio 43620
. I
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FROM : vISTuLA MANAGEMENT CO. PHONE NO. : 616+335+8100 Aug. 11 1'394 12:25PM P01 fJOCKET NUMBER PR 12- DOCKETED
~RQ OSED RULE~--- U P-'RC
( Sq FR 'J-f-'-/1£)
*94 AUG 12 A9 :31 AttP.ntion: Mr. Charles). Haughney, Chief OFFIC OF SECRETARY DOC tTIH & SE ICr_
Storage and Transport Systems Branch BRANCl lJivision of Industrial and Medical Nuclear Safety, NMSS United Stabl!N Nut'lP.ar Regulatory Commission Washington, D.C. 20555 FAX-301-415*5369 From: Connie Kline -216-946-9012 Fax 216--946-9012 - Alico Hirt - 616-335-3405 Fax - 616-335-8100 RE: June 2, 1994 Federal Register Notice 28496-NUHOMS Cask 24P-52B Deadline Exte1\8ion Request of the comment period for the Sb:mdordizcd NUHOMS Horiz.ontal Modular Storage System Total pages - 3
l~u*..,Lu-r, ,... ~**- __,,, Ji*, Y C,QMMISSION DOCKET ING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date ,.t:l~ J . . £I~ /)rtJwY\, Copies Received _ ___,____ "Z" ____ N f1J') Add'I Copies Reproduced -=-3_ _ __ _ Special Distribution fl:htJ ~/?)2~ - L --:? l l < ~~V\,
FROM vISTuLA MANAGEMENT CO. PHONE NO. 616+335+8100 Aug. 11 1994 12 :26PM P02 38531 Dodds Landing Dr. Willoughby H ills, OH 44094 August JO, )994 The St.acrctnry U.S. Nuclear Regulatoiy Commi"ion Washington, OC ;zu,,:; ATIN: Docketi11g and Service Branch RR: June 11994 Federal Ragisll~ Notice 28496 NUHOMS Cask 2tP-52B Mr. Secretary: The twelve citiz~ns groups listed befow request that y,,u grant an extension (>f the comment period 011 the Nuclear Regulatory Commission's propu~tiou tu ct111CJ1J its rcgulaUons to add the Standardizc..'<1 NUJiOMS Horizontal Modular Storage System tu the List of Approved Spent Fm~I Storage ('.asks. We request thttt you extend the comment puriod which ends on August 16, 1994 to September 30, 1994, Tw~nty-wght documentG conbtiuing prnpri,'!t.ary information were withheld from public di¥Closuro for 14 days after th~ official coI11111e1tl period b<.-gan. Therefore, the NRC is ul,Jigc1t",J lv extend the o£.6cieJ comment period. Thwl, documant&, whid1 were withh11ld from public discl08ure UJltil )Wlc 16, 1994, contain information th.at forms the basis for the final rule in the rule making to ttdd the Vectrd Cumpc111y'i:; NUJIOMS-241' and NUHOMS-5213 casb to lhe "l.wt of Approved Sp,mt Fuel Storage Casks" eligible for use und<.'l' the general Jiceme in Subpart K of 10 CFR l'art n.. The requirements ofJO(:FR 2.190(c) of NRC regulations provide that "information !!!Ubmith.>d in a rule making proceeding which subirequentJy forms the basis for the final rule will not be withhdd from public disclosure by the C.ommiuiun..." Moreover, the: ~ri,>u8lle1t11 wtd complwdtieR of u.wuow involv9d with <.lJMlit':1 dry r.1111k ijtoragc warrant an cxll*nsion of the t-umment ix--riod. The public needs more time in order to make informed comments on the i1c.Jdiliu11 uf the Sbmd11rdi~.ed NUHOMS System. On August 19, 1994 the Utility ludiological Safety Board of Ohio is spoJ\80ring a workshop entitled "High u,,ye} Radioactive Waste." Th~ workshop u; three daya after tht' ufikhal comment period ends. lnforsnation valuable in making intelligent t.'C>mments may be glliued by the public at this w<>rkshop which makci; an extension desirabJe and warranted. Furthermore, serious concerns whicll demand in*depth study }1ave boen nriljtld by the dr.1ft Curtificaite of Compliance fut t hP. ~tandardized NUT IOMS c11t1k. The NRC 's attempt
FROM v lSTuLA MAMAGEMENT CO. PHONE NO. 616+335+8100 Aug. 11 1994 12:26PM P03 tu in~t the Ianguag~ and intent of 10 CFR 72.48, which pertains to Specific License&, into tl1c Ccrt;fielltc of Compliance for tho Standardized NUHOMS fo .i very ocriouo ioouc which warrants much thought and research on the part of the general public. In light of tht~ fact that the NRC hl1~ ~een fi1 to 1tvuid publk h~iUing~ and to pr~d with on-site dry cask storage without site specific environmental impact statements, it certainly i,eems prudent for the health and safety of the general pubhc that the NJ<C aJlow the public an extension of the comment period. In view of the above considerations, the twelve environmental groups listed below n.-qut..at;t an ext~~ion 0£ the official c.*omment period on the Standardized NUHOMS dry cask system to September 30, 1994. Please inform Alice Hirt or Connie Kline of your IJtN.:ji+iun n,gilrdi.ng uu.- ,~qul1fil ab buon itti ,l)V2'tiiU\:. On behalf of the .following groups,
~~--~-~ /-A-.L_
Connie, Klintt, Sitttta dub FAX -216-946-9012 Alice Hirt, Coalition for Safe Energy-FAX-616-335-8100 "Connie Kline, Sierra Club Paula Ross, Ohio Qtizcn Action
*Clinton Warne, Consumers League of Ohio
<.:arolyn Monk, Don't Waste Ohio
~avid Blliso~ Ohio GJWnii; Steve Gannis, Ohio Citizens Against a Radioactive Envh-onment Keith Haddad, Border Opposed to Nudaer lJump
*Harvey Wuaerman,, Gl"ffnpeace Shirley *1*omasel10, Lake Erie Alliance
*David Hughs, Concerned Citizens of Cuyahoga, Ashtabula, Lake and Geauga Counties Bridgette Mariea, Ohio Envionmental Council
*Affiliatiou for identi£i.cati<>n purpcac only Sent via l< acsim1le and U.S. Mail
SIERRACLUB
- Connie Kline, Northeast Ohio Nuclear Committee 38531 Dodds _Landing Dr. USt-'RC DOCKETED ~1
\..!J Willoughby Hills, OH 44094 / t7.IY .
216-946-9012 (Ph. & FAX) 1 f 8/9/94 '94 AUG W P1 :07 COMMENTS ON PROPOSED RULE 11 LIST OF APPROVED SPENT FUEL ST -ij~E[GA~ ~~ *.~ i;;:>D_tTION" 59 Federal Register,.28496)'6/2/94 ' OCt<;c-:* * **.r.*,;*:_
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- 1. The following excerpts from 1o CFR 72 deal direotly with siting issues and indirectly with permanency issues which are the subject of several current lawsuits (Kelley v. Selin filed in May 1993, State of Michigan, et al v U.S. DOE et al, and Northern States Power Co., et al v. U.S. DOE, both of which were filed 6/20/94.) Discussion pertinent to Davis B~sse follows. .
72.24(a) "A description and safety assessment of the site on which'the ISFSI or MRS is'to be located .. .lf the . proposed ISFSI or MRS is to be located on the site of a nuclear power plant or other licensed facility, the potential interactions between the ISFSI or MRS and such other facility must be evaluated." 72.40(c) "For facilities that have been covered under previous licensing actions, including issuance of a -onstruction permit under Part 50 of this chapter, a reevaluation of the site is not required except where new information is discovered which could alter the original site evaluation findings. In t~is case, the site evaluation factors involved will be reevaluated." Subpart E - 72.90(e)(f) "Pursuant to Subpart A of Part 51 of this chapter for each proposed site for an ISFSl ... the potential for radiological and other environmental impacts on the region must be evaluated with due consideration of the characteristics of the population, including its distribution, and of the regional environs, including its historical and esthetic values. The facility must be sited so as to avoid to the extent possible the long-term and short-term adverse impacts associated with occupancy and modification of floodplains." 72.98(c){2){3) "Consideration of present and projected future uses of land and water within the region and any special characteristics that may influence the potential consequences of a release of radioactive material during the operational lifetime of the ISFSI or MRS." 72.1 00(b) "Each site must be evaluated with respect to the effects on the regional environment...both usual and unusual regional and site characteristics must be taken into account." 72.102(a){1 ){d) "East of the Rocky Mt. Front, sites will be acceptable if the results from on site foundation and geological investigation, literature review, and regional geological reconnaissance show no unstable Ageological characteristics, soil stability problems or potential for vibratory ground motion a the site in excess Wot an appropriate response spectrum anchored at 0.2g .... Site specific investigations and laboratory analyses must show that soil conditions are adequate for the proposed foundation loading."
- 72.122(b){4),(e) "If the ISFSI or MRS is located over an aquifer which is a major water resource, measures must be taken to preclude the transport of radioactive materials to the environment through this potential pathway .. .An ISFSI or MRS located near other nuclear facilities must be designed and operated to ensure that the cumulative effects of their combined operations *will not constitute an unreasonable risk to the health and safety of the public."
72.236(m) "To the extent practicable in the design of storage casks, consideration should be given to compatibility with removal of the stored spent fuel from a reactor site, transportation, and ultimate disposition by the Department of Energy." DISCUSSION OF DAVIS BESSE SITE As you know, the Environmental Impact Statement for Davis Besse was done some 20 years ago before the Standard Review Plan was instituted. According to the Army Corps of Engineers, older reactors were licensed under currently outdated environmental guidelines and couldn't be built on their present sites today. At the request of the International Joint Commission (which has called for , l elimination of radioisotopes from the Great Lakes ecosystem), the Army Corps of Engineers made a -~ detailed study of the Great Lakes shoreline which was published in 1993. However, the data for fl>. Ottawa County, Ohio were never published either because the Corps ran out of time or money. Davis Besse was built in a marshy wetlands floodplain . You are undoubtedly aware of the severe SH 2 8 1994 over Acknow'edqed.by card ......................... ,::-;
Page 2 of 6 - Kline Lake Erie storm in October 1972 which caused 300 feet of dike to break, submerging the entire plant site-, ~ncluding the reactor building and forcing people to be evacuated by air or boat; fortunately the plant was pre-operational. There has been serious subsequent flooding of Davis Besse, particularly during spring thaws when roads leading to and from the plant are 1mpassable due to water levels. Given the fact that there is not now and there may netter-. be a permanent HLRW repository for commercial reactor fuel and the fact that the NU HOMS 24P and 52B \ casks are non-transportable, any distinction between so called "temporary storage" and "permanent disposal" of this waste is moot. The Davis Besse site cannot even meet the NRC's bare minimum siting criteria for an above or below-ground 11 low-level 11 radioactive waste disposal facility per 10 CFR 61 which "contains common sense siting requirements (that) the NRC views as minimum ... whether or not engineered enhancements (concrete) are used. The NRC siting requirements are primarily directed at aspects to be avoided : A. Sites should be avoided (with) known natural resources. B. A prospective site must be well-drained and free of flooding or frequent ponding. C. The site should be located far enough above the water table to prevent ground water intrusion. D. Sites and areas where seismic activity and erosion .. .occur... must be avoided. 11 (1) In several documents, the NRG, itself, opposes at-reactor storage of LLRW "beyond 5 years as a significant safety and environmental matter that vvould divert the plant operator from its main task of rea_~tor operation and make it diffic~lt to d~termine if radioacti~e releases were from the r~~~tor or thA fac1hty.11 (2) From 10 CFR 61 , 11 S1tes must not be l~cated in,areas where nearby fac1ht1es ...coulcl-' significantly mask or interfere with the disposal facility's environmental monitoring program." A May 24, 1988 study (attached) entitled "An Evaluation of the Four Licensed and Operating Nuclear Power Plant Sites in Michigan for Co-Location of LLRW Isolation Facility" prepared by Environmental Resources Management for the Michigan LLRW Authority concluded : None of the four nuclear power plants in Michigan are suitable sites for co-location of a LLRW isolation facility (due to) intense geological processes such as mass wasting, erosion, poor drainage .. .the shoreline setting of each of the nuclear power plants does not offer the safety and security of alternative non-shore sites. Wind-driven flooding and seiches will undoubtably play an important role in the integrity and longevity of the site and facility throughout its life. The NRG is also aware of the 12/30/93 letter (attached) from U.S. EPA Region 5 Regional Administrator Valdas Adamkus to the NRC which states: Your agency has assessed dry cask storage systems generically and has also evaluated the environmental impacts of them generically ... We believe the potential for significant adverse impact to either Lake Michigan or the Mississippi River (valuable natural resources providing drinking water and recreational opportunities for many people) is real and was not fully assessed in the generic environmental assessment prepared for the dry cask storage process ...The site specific conditions and the valuable resources of Lake Michigan and the Mississippi River warrant a full and complete evaluation of the impacts and review by other Federal and State agencies as well as the interested public. The 1/30/94 reply to Mr. Adamkus from the NRC 's Robert Bernero is completely inadequate as is the NRC's March 1994 "Draft Environmental Assessment and Finding of No Significant Impact" because no consideration is given to the site's unsuitability even for LLRW per the NRC's own admission, and "new information which could alter the original site evaluation findings" (see below) is ignored. Per recent phone conversations, the U.S. EPA considers this matter unresolved. Below are several finding of fact from court documents presented during the September 1986 and June 1987 hearings regarding LLRW sludge disposal on site at Davis Besse which the State of Ohio ' vigorously opposed. These constitute "new information ... both usual and unusual regional and site characteristics ... which could alter the original site evaluation findings .. .and must be taken into
Page 3 of 6 - Kline account." A. State expert witnesses, Mr. Pavey and Mr. Guy, geologists, and Mr. Voytek, a hydrologist, were astonished that TE had provided no hydrology study and stated that TE's geological studies done in 1970 related to construction of Davis Besse were inadequate and outdated and revealed a limited understanding of soil types, permeability, water flow patterns on sit~, underground aquifers in the Navarre Marsh area and response to changes in Lake Erie levels* or to flooding . Former Attorney General Celebrezze described TE's geology studies as "cursory, flawed , oversimplified, and superficial." (Transcript, p. 49) B. The State of Ohio testified that there had been major technical and equipment advancements in the last decade in both geology and hydrology. The process of deep excavation in the past usually smeared evidence of sand and gravel layers, of cracks, of soil permeability, and of tiny water flow pathways. Bore logs were frequently deceptive where parts of the core were missing. C. The State pointed out the similarity of till , glaciolacustrine, clay and sand patterns of soils for the whole Great Lakes area and especially for Ottawa County with its widespread marsh areas. The State reviewed evidence of early glacial movements in soil patterns and concluded that there was an upper till aquifer which , when saturated , drained into Lake Erie, the Navarre Marsh, and the Toussaint River. D. The State cited indications of drainage pathways - some lateral and then vertical into ground water A and the bedrock lower aquifer. Mr. Pavey insisted that by all indications, the water in the glacial wsediments connected to the bedrock. Using "The Soil Survey of Ottawa County" by Gordon and Huebner, the State supported its findings of cracks, fractures, thin seams, ler.ises, and former tree root flow paths (from early forests) to account _for drainage down to the ground water aquifer from the till above. Even one of TE's own borings (B-125, ATEC Assoc. , Inc. , 1974) documented the presence of sand layers. E. Both the State of Ohio and TE agreed that the limestone-dolomite bedrock was highly permeable and that ground water levels were responsive to weather, seasons, Lake levels, river levels, and marshlands. When high northeast winds raise the Lake Erie water levels at the west end, the groundwater levels also rise. After a storm , the flow of both gradually reverse. TE verified the extent of the ground water system and its permeability from the wide radius affected by its dewatering procedures in the early 1970's. The State observed that ground water was released into Lake Erie through the permeable bedrock that extends into the Lake. The State contended that all of northwestern Ohio depended on the same groundwater bedrock aquifer system which included the entire Ottawa Marsh area. - Due to the lack of a permanent repository or MRS any time in the foreseeable future, the distinction between so-called "temporary storage" and permanent radioactive waste disposal are mere semantics especially in the case of a serious spill and resultant contamination at an environmentally unsuitable site like Davis Besse where "short and long-term adverse impacts associated with occupancy and modification of (a) floodplain ... potential release of radioactive material during the lifetime of the ISFSl ... (and location) over an aquifer which is a major water resource" have been inadequately dealt with. Furthermore, "projected future uses of land .and water within the region" are impossible to make i given the unknown length of time this waste may remain on site and the options for both cask and reactor license renewal beyond 20 and 40 years respectively and the fact that no known man-made structure can last for the length of time that this waste must be isolated from humans and the environment. If an MRS or repository ever become available, this waste may have to be repacked. Each handling of this waste increases the Hkelihood of an accident, spill , contamination, worker and/or public exposure. Decommissioning and decontamination of reactors and reactor sites remains uncertain at best. 9.2 and 9.3 of the Draft SER state, "At this time, it is not known whether demolition and removal of the HSM can be performed by conventional methods ... The reinforced structure of the HSM, for example, will require considerable effort to demolish." Of course, in its typical fashion of putting off until
- tomorrow what it cannot deal with today, the NRC considers "ease of decommissioning (a) secondary over
Page 4 of 6 - Kline consideration."
- 2. You are aware of the controversy regarding whether 10 CFR 72.48 which pertains to Specific Licenses can be used by those issued General Licenses under Subparts K and L. This issue remains unresolved because the NRC General Counsel has not issued a legal interpretation despite a 2/14/94 request to do so from the NRC's Charles Haughney (copy attached}. Because this issue can only be resolved through NRC rulemaking, inclusion of the text of 10 CFR 72.48 as # 9 in Draft Certificate 1004 for the NUHOMS-24P and 52B casks is improper.
In a 10/1 /92 NRC memorandum regarding a 7/24/92 meeting with Pacific Nuclear Fuel Services Inc. (now VECTRA) regarding certification of the NUHOMS cask, the NRC states, 11The only way that 10 CFR 72.48 may be involved is via a site-specific license. 11 In a 1/31 /94 NRC letter, the NRC states, 11 Subparts K and L of 10 CFR Part 72 are silent on cask SAR and certificate changes after the final rule. The NRG staff is currently contemplating rulemaking to clarify these issues. 11 A 6/3/94 memorandum from Mr. Sturz to Mr. Haughney regarding a 5/19/94 meeting between the NRC and SNC stated: Staff indicated that it had written a memo to the Office of General Counsel requesting an interpretation of the applicability of 72.48, that it had not yet A received a reply to the request, that the licensee can make its own interpretation W of the regulations, and that rulemaking may be considered to clarify the regulation. Pacific Sierra Nuclear (SNC) related that the Arkansas Nuclear Plant (ANO) need to load two casks before its next outage presently scheduled for March 1995. The utility wants to use the longer VSC 24 cask currently the subject of the requested amendment 1 to the SAR. In order to meet this schedule, casks are needed by this fall. A 6/2/94 letter from Entergy Operations informs NRC's Robert Bernero that it intends to make modifications to the VSC 24 SAR for use at ANO by applying the provisions of 10 CFR 72.48, that based on its 10 CFR 72.48 evaluation, Entergy had directed SNC to begin fabricating fourteen casks of increased length to accommodate ANO's longer CE 16 x 16 fuel, and that Entergy intends to continue using 10 CFR 72.48 in the future. In his 2/14/94 memo to the NRC's General Counsel, Mr. Haughney states, "This section (72.48) clearly and applies to specific licensees issued individual licenses under Part 72. 11 Yet the 6/3/94 NRC memorandum from Sturz to Haughney seems to give General Licensees the green light to interpre.a 10 CFR 72.48 as they see fit before the General Counsel rules on that part of the Code.
- There is no provision in Subparts K or L of 10 CFR 72 that permit a General Licensee to change a vendor's SAR. Nor do Subparts K or L allow a vendor to modify its SAR or C of C. Cask vendors are not licensed under 10 CFR 72.40. The site specific license provisions of 10 CFR 72 apply to operators of spent fuel storage installations not to cask vendors.
Since the Code is silent on a process to change a generic cask design by changing an SAR or a C of C, the NRC must use a rulemaking procedure which provides for public comment and proprietary release. To issue a general license to a cask vendor so a cask can be used anywhere and then to permit virtually unlimited site specific changes is contradictory and not in keeping with the intent of generic rulemaking. The cask vendors, the NRC, and the utilities can't have it both ways.
- 3. Transfer cask and related issues:
A. 10 CFR 72.234(c) states, "Fabrication of casks under C of C must not start prior to receipt of the C of C for the cask model. 11 The NRC has just granted VECTRA an exemption to begin transfer cask fabrication (but not use) 11to have the necessary equipment available for use by DBNPS in mid-1995, and thus enable DBNPS to maintain complete full-core off-load capability in its spent fuel pool following the refueling outage scheduled for early 1996. 11 This is yet another example of the NRG allowing the vendor to put the cart before the horse, bending NRC rules to facilitate the perpetuation , of the industry. Seeking public comment appears to be nothing more than going through the motions and providing comments is an exercise in futility because cask approval seems to be a fait accompli.
Page 5 of 6 Kline
-The situation is similar to utilities supposedly proceeding at their own risk under limited work authorizations prior to issuance of a reactor construction license. Once the investment was made, a construction license was was a certainty as is a certificate of compliance.
B. It is our understanding that one transfer cask will be shared by several nuclear power plants around the country. We are concerned that in the event of problems and the need to off-load the fuel (as in the recent situation at Palisades) , a transfer cask may not be available in a timely manner due to inclement weather or because the TC, itself, has experienced problems or is being used elsewhere. C. We are concerned tha~ the crane used for fuel handling in the spent fuel pool building is a single failure-proof device. The C of C and SER discuss drop analyses of 1511 up to 80 11
- There is no discussion of drop accidents within the spent fuel pool building such as a drop onto the building floor or a drop of the TC into the spent fuel pool, itself, which would surely damage the fuel assemblies in the pool. Both these drops are considerably greater than 80 inches!
D. We remain concerned about possible jamming of the transfer cask in the spent fuel pool. What would happen to the cask if the jammed fuel could not be extricated? Would the entire 40 ton TC be left in the fuel pool?
- 4. It is our understanding that the test revealing the faulty welds at the Palisades plant was conducted in July just before the cask was filled, but the test was not reviewed. This raises serious questions
- about NRC oversight and requirements for proper cask fabrication by licensees.
- 5. We are concerned about the presence of burrowing and other nuisance animals that have posed problems at other waste sites. It seems likely that insects, animals, and/or birds will be attracted to th e warm air coming from the outlet vents. We rem ain concerned about vent blockage particularly from insects such as paper wasps which build huge nests and swarms of midges common to the Great Lakes which can completely cover and block screening and vents.
- 6. We remain concerned that the fuel will not be tested for leaks using penetrating dyes, eddy current, sipping or ultrasound prior to canister loading despite the fact that some of the rods in the spent fuel pool will be nearly 20 years old . Exactly how will 11 grossly breached 11 fuel be ultimately handled and shipped off site?
- 7. We think additional radiation monitoring should be required, particularly in light of 8.3.1 of the Draft SER which states, "Dose rates calculated by the vendor for different locations around the e standardized NHUHOMS design are significantly higher than those determined for previous NU HOMS designs ... the relative dose rates for this design are still expected to be higher than comparably calculated dose rates for earlier NUHOMS designs. These relatively higher dose rates are not consistent with the objective of maintaining occupational exposures ALARA. Site-specific applications with this design should provide detailed procedures and plans to meet ALARA guidelines and 10 CFR 20 requirements with respect to the operation and maintenance of this standardized NU HOMS ISFSI design."
- 8. We question how the higher 55,000 MWD/MTU burnup fuel now being used in PWR's will be handled since the NUHOMS 24 is rated to handle only 40,000 MWD/MTU burnup fuel.
- 9. We remain concerned about the possibility of insufficient drying of the *fuel before placement in the DSC. We do not feel that the issue of corrosion of stainless steel has been adequately evaluated especially under conditions of indefineite duration. While stainless steel corrodes less rapidly than carbon steel, even the plumbing fixture industry is finding unexpected stainless steel pitting and corrosion under conditions far less intense than those in a DSC.
- 10. The issue of sabatoge does not seem to be adequately addressed in the Draft SER particulary in view of the 1993 bombing of the World Trade Center in New York and the ease with which a disturbed individual recently breached security and remained undetected at a U.S. reactor. Explosive technology has become very sophisticated in the last 15 years since the NRC and Sandia Laboratories studied the effect of sabotage on shipping casks in the March 1979 NUREG-459 - 11 Generic Adversary over
Page 6 of 6 -Kline Characteristics Summary Report". (1) Afton Assoc. Inc., "Explanatory Guide to 10 CFR 61," U.S. NRC, Div. of LLW and Decommissioning, Office of Nuclear Material and Safeguards, Washington, D.C., 1989, pp 5 & 6. _ (2) U.S. NRC Generic Letter 81-38, 1/10/81 ; Generic Letter 85-14, 8/1 /85; NRC Information Notice No. 90-09, 2/5/90; SECY-90-318, 9/12/90; Midwest Compact "Frequently Asked Questions and Answers About LLRW Disposal and the Midwest Compact," St. Paul, MN, Fall 1991, Question 1.8.
- Affiliation for identification. purposes
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FILE: 703-01 FINAL REPORT .l An Evaluation of the Four Licensed and Operating Nuclear Power Plant Sites in Michigan for Co-Location of a Low-Level Radioactive Waste Isolation Facility May24, 1988 PREPARED FOR: MICHIGAN LOW-IEVEL RADIOACTIVE WASTE AUTHORrIY DEPARTMENT OF MANAGEMENT AND BUDGET
. P.O. BOX 30026 HOWSTER BUILDING LANSING, MI 48909 I PREPARED BY:
ENVIRONMENTAL RESOURCES MANAGEMENT
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SUITE 2 ANN ARBOR. MICffiGAN 48105
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SECTION FOUR " Conclusions 1111s study found that none of the four nuclear power plants 1n Michigan are suitable sites for co-location of a low-level radioactive waste isolation facility. Based on the availaole information Usted in Section Five of this report. the nuclear power plant sites and immediately adjacent areas did not meet several key exclusionary criteria. These criterta include those in which areas of intense geologic proce~s such as mass wasting. erosion and the like must be excluded. areas with high values of soil permeability must be excluded. areas exhibiting poor drainage and ponding must be excluded. and areas designated as wetlands must be excluded. - Although some specific detailed information about the sites was unavailable, the information that did exist was enough for a proper initial evaluation as requested by the Authority. The goal of the siting criteria is to select a site with outstanding natural baniers 1n the event of a leak or spill breaching one of the many engineered bamers of the actual facility and disposal process. Relying only on the available information that was reviewed, a low-level radioactive waste disposal facil1ty would not meet the goals of the Siting Criteria Advisory Committee, Act 204 of 1987. the Authority's and the NRC's siting objectives and criteria and the overall goals of the NRC's performance objectives. All of the sites are located near either populated or popular seasonal resort areas of the State and are located adj acent to one of the Great Lakes. These sites do not offer suitable natural protection from an
,e inadvertent spill or undetected leak of the anticipated waste miXture.
Finally, the shoreline setting of each of the nuclear power plants does not offer the safety and security of alternative non-shore sites . Wind-driven flooding and seiches will undoubtably play an important role in the integrity and longevity of the site and facility throughout its life. 4-1
UNITED STATES ENVIRONMENTAL PROTECTION AGENCY REGION 5 77 WEST JACKSON BOULEVARD CHl~O I~ ,~*3590 3 REPLY TO Tt1E ,HT£** r, ME-19J Mr. Jamea M. Taylor Executive Director tor Operations United States Nuclear Regulatory Commission Washington, o.c. 20046-0001
Dear Mr. Taylor:
We are writing in regards to the proposed dry cask storage proposal tor Consumers Power Company's Palisades Nuclear Power Plant near south Haven, Michigan and Northern States Power, Prairie Island Nuclear Power Plant. We have recently become aware o! both o! these proposals. From our understanding, both Consumers Power and Northern States Power have been granted approval by your agency to use a dry cask storage system to store spent nuclear tuel on a concrete pad on each . tacility's property. The license for storage ot these spent tuel rods is for 20 years with the potential ot extending the license for an additional twenty years. The dry cask storage system was authorized by the Nuclear Waste Policy Act of 1982, section 133. Your agency has assessed dry cask storage systems generically and has also evaluated the environmental impacts ot them genericai1y. The Consumers Power Company's dry cask storage site will be adjacent to Lake Michigan which is a valuable resource providing drinking water and recreational opportunities tor many people. Similarly, the Prairie Island Nuclear Power Plant is situated on an island i n the Mississippi R~ver, another valuable natural resource. We believe the potential tor significant adverse impact to either Lake Michigan or the Mississippi River is real and was not fully assessed in the generic environmental assessment prepared for the dry cas~ storage process. Therefore, under our authority under Section 309 of the Clean Air Act, we are requesting to review the environmental documentation that you have used to determine that this action would not have a significant impact upon the human or natural environment. The site specitic conditions and the valuable resources o! Lake Michigan and the Mississippi River warrant a full and complete evaluation ot the impacts and the review ot this analysis by other Federal and State agencies as well as the interested public. EI>O --- 009678 over
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In addition, th* Prairie Island Nuclear Power Plant &hares the island with the Prairie Island Dakota Community. It does not appear that the impact to the Tribe was assessed in the gener i c environmental impact atatemant. You may need to evaluate your agency'* trust responsibilities, to the Indian Tribe, regarding the siting ot the dry cast storage area at the Prairie Island Nuclear Power Plant. We will appreciate your -cooperation in this matter and look forward to a timely reaponse. If you have any questions, please feel free to contact Mr. Robert Springer, Assistant Regional Adminis ator for Planning and Management, at 312/353-2024~ Sine y yo Valdaa V. Ad Regional Adm
for Ruje~~~~~i ~~~ -i~;i-,i~i;*-* Office of the Genertl Counsel .ftB . 1 4 1~ Charles J. H1ughney, Chtef Storage and Transport Systems Branch D1vtston of Industrial ind Medical Nuclear Safety Office of Nuclear HAter1&1 Sifety 1nd Safeguards
SUBJECT:
REQUEST FOR LEGAL INTERPRETATION OF THE APPLICABILITY OF SECTION 12.48 TO GENERAL LICENSEES LICENSED IN SECTION 72. 210 Would your office confirm our understanding of the scope and app11cabi11ty or L. Section 72 .48 of our regulattons? Th1s section clearly applies to spectftc ...,......- licensees tssued ind1v1dual licenses under Part 72. Our question involves whether Section 72.48 miy also 1pply to general licensees as licensed tn Section 72,210, * *** persons authorized to possess or operate nuclear powar reactors under Part SO of this chapter,* Our v1ew 1s that the prov1s1ons of Section 72.48 are avatlable to th1s group of general licensees, as defined 1n Section 72.210. Accordingly. general licensees may mike changes to thetr certified storoge systems, provided such changes do not constitute an unrevtewed safety question as defined 1n Section 72.48. I 1sk that your offtce review our pos1tton and confirm whether or not we are construing this regulation correctly.
- The question is important to ongotng 1ct1vtt1es at the Arkansas Nuclear One s1te and the Potnt Beach Nuclear Statton where the respective licensees need to perform design changes to certified storage systems 1n order to acconnodate thetr particular spent fuel and handling equipment designs. The constder1t1on of these design changes will affect near~tenn. upcoming refueling outages for these stations.
I and ~Y staff are 1v1111ble to discuss this matter w1th_you at any time. We would apprec11te your early 1nd thoughtful cons)der1tton of our request. Charles J, H*'§/,; CHftf i~;1~i Z~Lb6 Stor1ge and Transport Systems Branch Division of Industrial and Medic&l Nuclear Safety 1 ..,00 l 0 Office of HucleAr M1tert1l S1fttY and Safeguards
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(_§ Cf FI?.. :J_f-1-J q 6_) - VISTULA MANAGEMENT COMPANY 1931 Scottwood Avenue Suite 700 P.O. Box 4719 *94 AUG -9 A9 :36 Toledo, Ohio 43620 TEL (419) 242-2300 FAX (419) 246-4703 OF FI Cl CF SEC RE TA 'f OOCKE f I 'G (~ S~ F , !Lt 1 August 1994 BR :- HCh Secretary U.S . Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Docketing and Service Branch
Dear Mr. Secretary:
My name is William Hirt. I wish to comment on the safety analysis report (SAR) dated November 5, 1993, RDQ-93-052, for the standardized NUHOMS system cask for irradiated fuel. Specifically, I wish to draw your attention to 1.3.1.1: "Furthermore, there are no credible accidents which could breach the containment boundary of the DSC as documented by this SAR." During the past fifteen years, my company has substantially rehabilitated over 100 structures in the states of Ohio and Indiana. Most of these structures range in age from 50 to over 100 years. During the rehabilitation process, we often come across evidence of damage apparently due to past seismic activity. This evidence usually takes the form of stress cracks in interior plaster and in exterior brick and masonry materials, movement of foundation plates, and degradation of flooring and roof joist systems. While most of this evidence does not compromise the building' s structural integrity, it does indeed indicate that seismic activity has occurred over a relatively brief period of time from a geological perspective. A brief review of seismic activities in Ohio reveals that the current seismic requirements for the design of the NUHOMS ' cask is clearly inadequate and, furthermore, seismic conditions could and have occurred which could cause a complete failure of the cask and the type of disaster which would then, inevitably, occur.
- On December 16, 1811 and January 23, 1812 and February 7, 1812, the largest earthquakes ever to strike the continental United States occurred at New Madrid, Missouri. These events were felt throughout an area of about 2 million square miles, which included all of Ohio. In Ohio, some chimneys were toppled in the Cincinnati area, which experienced the strongest shaking from these events. Should earthquakes of this intensity be repeated at New Madrid (which many experts are predicting will occur within the next 50 years) they could cause considerable damage throughout western Ohio.
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- June 18, 1875 : This earthquake was felt throughout an area of at least 40,000 square miles in Ohio. Masonry walls were cracked and chimneys were toppled. It is estimated that the Modified Mercalli Intensity (MMI) scale was VII. The MMI scale measures horizontal acceleration during seismic activity. Horizontal acceleration or ground motion can produce tremendous forces that can seriously damage a structure if the duration of the movement lasts for a comparatively long time. During many earthquakes, this destructive motion only lasts for about 10 seconds.
- September 19, 1884: An earthquake covering roughly a 140,000 square mile area, whose epicenter is located in the vicinity of Lima, Ohio, was felt as far away as Washington, D.C. It had an MMI of VI.
- May 17, 1901 : Another quake dislodged bricks and toppled chimneys in Portsmouth and Sciotoville, Ohio. MMI intensities of VI were generated in the epicenter area.
- November 5, 1926: This quake was centered in Migs County. Chimneys were toppled. An MMI of VII was generated.
- September 30, 1930: This quake was centered in the vicinity of Anna in Shelby County, where a chimney was toppled. An MMI of VII has been assigned to this event.
- September 20, 1931 : Again in Shelby County, chimneys were toppled and plaster was cracked. A ceiling collapsed in a public school north of Anna. An MMI of VII has been estimated for this quake.
- March 2 & 9, 193 7: These are two of the most recent quakes, and by far the most damaging to have struck Ohio in recorded history. It is estimated that the March 2nd quake reached VIII on the MMI scale. Both earthquakes were felt throughout a multi-state area and plaster was cracked as far away as Fort Wayne, Indiana. The center of this quake also was along the same fault line in Shelby County. The quake was felt throughout an area of about 150,000 square miles.
- January 31 and July 12, 1986: These are perhaps the two most recent quakes to be felt in the State of Ohio. While little damage was recorded, cracked plaster and broken windows were evident. Both quakes reached VI on the MMI scale.
Secretary, U.S . Nuclear Regulatory Commission I August 1994 Page three The Safety Evaluation Report draft dated May 14, 1994 for the standardized NUHOMS horizontal storage system for irradiated nuclear fuel states on 3-9B, "The standardized NUHOMS HSM was analyzed for a peak horizontal ground acceleration of .25g." Clearly, the entire state of Ohio has experienced earthquakes in the recent past in the range of .15g to .30g, as measured on the MMI scale of VII and VITI. Moreover, less than 60 years ago the state of Ohio experienced an earthquake as noted above which exceeded .25g on the MMI scale. Clearly, therefore, the current design of the NUHOMS HSM does not adequately take into account the very real potential for seismic activity which could seriously compromise the structure of the cask, the cask's foundation and the resultant release of radiation. Radioactive particulates would then follow wind patterns. Persons downwind could then inhale radioactive particulates and gases. Particulates could settle on vegetation and soil, entering the ground water and Lake Erie. Therefore, the cask, containment structure and foundation pad must be designed to substantially exceed all known earthquake potential. On the MMI scale, this should measure X, that is, at least .60g. The NRC's insistence in this matter is critical if the public's health and safety is to be protected. Thank you for this opportunity. Cordially yours, L-Hi]{1 William President WH/mr
Secretary, U.S. Nuclear Regulatory Commission 1 August 1994 Page three The Safety Evaluation Report draft dated May 14, 1994 for the standardized NUHOMS horizontal storage system for irradiated nuclear fuel states on 3-9B, "The standardized NUHOMS HSM was analyzed for a peak horizontal ground acceleration of .25g." Clearly, the entire state of Ohio has experienced earthquakes in the recent past in the range of .15g to .30g, as measured on the MMI scale of VII and VIII. Moreover, less than 60 years ago the state of Ohio experienced an earthquake as noted above which exceeded .25g on the MMI scale. Clearly, therefore, the current design of the NUHOMS HSM does not adequately take into account the very real potential for seismic activity which could seriously compromise the structure of the cask, the cask's foundation and the resultant release of radiation. Radioactive particulates would then follow wind patterns. Persons downwind could then inhale radioactive particulates and gases. Particulates could settle on vegetation and soil, entering the ground water and Lake Erie. Therefore, the cask, containment structure and foundation pad must be designed to substantially exceed all known earthquake potential. On the MMI scale, this should measure X;'--"'1--..!c.he NRC's insistence in this matter is critical if the public's health and safety is to be protecte . Thank you for this opportunity. Cordially yours, William Hirt President WH/mr
DOCKET NUMBER PROPOSED RULE p ~2._ DOCKETED ( ~ q FR 1 y-LJ CJ 6) USNRC John Trapp 713 Roeder Street "94 AUG -5 P4 :22 Monroe, MI 48161 July 15, 1994 OFFI CE OF SECRETARY OOCKETlt~G & SERVICE To whom it may conii.f\#, H This letter is to be considered comment upon the addition of the Standardized NUHOMS Moduler Storage System to the List of Approved Spent Fuel Storage Casks. NUHOMS must not receive generic approval. Sight specific characteristics must be considered for this cask. These casks are not terrorist proof, to place this cask upon - the shores of Lake Erie is potential ecocide. Without a guaranteed federal repository these casks could become permanent. Over time the horizontal storage of fuel rods will lead to cladding decay which will challenge the technical specifications of the NUHOMS cask. No more casks should be approved until a permanent federal repository is opened. The wet fuel pool is a proven technology which has been successful in containing radioactivity. If this cask is to be used, it must only be used in e conjunction with a sight specific Environmental Impact Statement. The people of the Lake Erie Basin demand full public hearings regarding the use of the NUHOMS cask. Very sincerely; John Trapp co-chair Zebra Mussel Alliance SEP 28 1 ~ Acknowledged by card .........- ........."""".,.
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DOCKET NUMBER PR r O OSf D RULE -72 (j) (.§CfF'R2 VISTULA MANAGEMEN1CO DOCKETED 1931 Scottwood Avenue Suite 700 P.O. Box 4719 *94 AUG -1 P4 :03 Toledo, Ohio 43620 TEL (419) 242-2300 FAX (419) 246-4703 27 July 1994 Secretary U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn : Docketing and Service Branch
Dear Mr. Secretary:
My name is John R. Kiely. I am a registered professional engineer in the states of Ohio (Registration #E42326) and California (Registration #E21894). I designed nuclear power plant containment structures for the Bechtel Corporation in the 1970' s, with particular emphasis on the seismic design . I wish to submit a comment regarding the NRC proposed rule under 10CFR Part 72, Sub Part L, for a general license for the horizontal modular storage systems (NUHOMS). I am commenting on the article in the Federal Register dated Thursday, June 2, 1994, submitted by James M. Taylor, executive director for operations regarding the VECTRA Technologies, Inc., design for a Horizontal Storage Module (HSM). I will attempt to show in this comment that the statement "The proposed rule will have no adverse effect on the public health and safety" cannot be guaranteed and, therefore, even though it may be convenient for the nuclear industry and the NRC to avoid site specific approvals, in this case these are essential for maintaining public safety. I have two concerns. The first regards the adequacy of the cooling requirements. The second is the structural adequacy of the HSM during a potential seismic event. I wish to raise concerns regarding the adequacy of the cooling under all atmospheric conditions throughout the country. On a recent day in the Midwest, the humidity stood over 90%, the temperature over 100
- and there was no wind. I am not convinced that a relatively hot load of fuel under prolonged high temperature conditions can be adequately cooled in the cask as it is currently designed. As any engineer can tell you, the structural integrity of concrete can be easily compromised by prolonged high temperatures. I also wish to question whether the screens between the casks, which are essential to the cooling, will remain clear of debris and to ask how can they be cleaned if they do become partially clogged?
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U.S. Nuclear Regulatory Commission 27 July 1994 Page two My primary concern is for the structural integrity of the HSM and by implication, for the embrittled fuel cladding within the Dry Storage Canister (DSC). It would be very difficult to restrain the HSM if the DSC were in a vertical position. However, even with the internal supports, storing embrittled fuel rods in a horizontal position when they have been in a vertical position throughout their lifetime, is not wise. It is a well known fact that changing the loading conditions on an embrittled metal cladding may result in changed stress loading and it is impossible to adequately predict the loading during either a drop accident or a seismic event on this metal cladding. Obviously, the integrity of the entire storage system will be violated if this cladding should crack. When a material has been loaded in one direction, i.e., vertical, within the reactor and is then changed to a horizontal position it is essential to carefully inspect the cladding for the minute hairline cracks which would allow the radioactivity inside to escape. Under the current design, this is not possible. I believe this is a serious problem. On page 3 of the Certificate of Compliance, under item #9, the certificate holder is given permission to make changes in the procedures and to conduct tests, none of which have been described in the safety analysis report. While this may be convenient for the commission and the nuclear plant operator, I believe it is very unwise to give this permission to regulators regardless of their experience, regardless of the local conditions, regardless of the condition of the HSM and the fuel within it. To require an annual report after the event is clearly inadequate. Figure 1. 1 shows the HSM's stored in a horizontal position in close proximity to each other with no restraint. I question the logic of this method of storage in any type of seismic event. Will not the casks crash against each other as the ground moves beneath them and potentially have the structural integrity of each of these vital containment structures compromised? On page 3.9 item #B, the statement is made that under the peak horizontal ground acceleration of 2.Sg's the HSM "would neither slide nor overturn due to the seismic input". I wish to raise several questions about this point. First, I do not believe that a 2.5g ground acceleration can adequately describe all potential earthquakes east of the Rocky Mountain Front. Because an area has not had an earthquake in the recent past is no reason to conclude that it cannot within the lifetime of this cask. It is well known that one of the major earthquakes in the country occurred in the Midwest approximately 190 years ago. There is some discussion that earthquakes of this magnitude could happen every several hundred years. I do not believe that a ground acceleration of 2.5g is realistic for all sites, regardless of proximity to fault lines. However, if you allow these canisters to be used anywhere without further hearings you have made this decision.
U.S. Nuclear Regulatory Commission 27 July 1994 Page three A more serious concern has to do with the data obtained from the June 28, 1992 Landers quake in the Mojave Desert northeast of Los Angeles. This quake showed a significant "displacement pulse" in an area near the fault line. This is raising serious questions about the whole seismic design of all types of structures. However, the HSM design which allows it to sit unanchored on a concrete pad, would be particularly vulnerable to such a "displacement pulse". Clearly a displacement pulse of 60 cm., as observed in the Lander' s quake, would completely destroy the HSM and allow a substantial release of radioactivity from the fuel within. Until an adequate analysis for this new data can be performed I cannot envision how the NRC can prudently allow this cask to be licensed. There are a number of other issues which could be raised; however, I believe this new earthquake information is of such a serious nature that it far outweighs these other issues, and until this one has been resolved I would strongly urge the NRC not to license this canister for storage for nuclear waste. I would appreciate hearing from you about these concerns. JK/mr cc: Alice Hirt William Hirt
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Governor rl~..,.._;~n n..-..-.1-~~n Ra\.Ua.UL.J 4 .1.~ Programs ;-\r ., Commissioner CN 415 Trenton, N .J. 08625--0415 Tel (609) 987-6389 Fax (609) 98Ju6.i39021, 1994 The Secretary of the Commission DOCKET NUMBER PROPOSED RULE~-.:;;..---- PR "/ 2 U.S. Nuclear Regulatory Commission Washington, DC 20555 ( £ '1 Ff< 2-~ L/ 1 b) ATTN: Docketing and Service Branch
Subject:
Draft Certificate of Compliance for the NUHOMS System for Irradiated Fuel
Dear Sir:
The New Jersey Department of Environmental Protection' s Bureau of Nuclear Engineering (BNE) has reviewed the subject draft Certificate of Compli~nce referenced in the proposed rulemaking contained in the June 2, 1994 Federal Register. As a result of our review, we have the following comments:
- 1. The Certificate of Compliance, Section 1.1. 7, establishes special requirements for the first system in place. The user of the first system is required to load the first storage module with fuel assemblies containing a heat load of approximately 24 kW. we note that the NRC will accept the use of an artificial heat load as well. In the interest of AI.ARA principles, we believe the use of an artificial he~t load sh~uld be req1~ired in lieu of fuel assemblies with a 24 kW heat load. Furthermore, loading the oldest non-leaking fuel fir st and for subsequent loads into the modules should be required.
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- 2. Surveillance Requirements (SR) 1.2.8, HSM Maximum Air Exit Temperature, and 1. 3. 2, HSM Thermal Performance, appear to conflict. In the surveillance section of SR 1.2.8, it states that "if the temperature rise is within the specifications or the calculated value for a heat load less than 2 4 kW, then the HSM and the DSC are performing as designed and no further temperature measurements are required." SR 1. 3. 2 requires that a temperature measurement for each HSM be performed on a daily basis.
- 3. Surv.;3illanc.;; R1:;quire:mant 1. 2. 12, Maxin:um DSC Remov.:\ble Surface Contamination, has a surveillance requirement which s tates " contamination s urveys shall be taken on the inside surfaces of the TC after the DSC has been transferred into the HSM. " However, there is no corre sponding requirement in the action section stating what action is to be taken as a result of an unacceptable contamination survey.
If you have any questions, please contact me at (609) 987-6389. Sincerely, JJtr;t J 11 Lipoti, Ph.D. Assistant Director, Radiation Protection Programs cc: Kent w. Tosch, Manager DEP Bureau of Nuclear Engineering
DOCKET NUMBER Pnoroc:Fo RULE PR 11-. DOCKETED DOCKETED US RC (i) July 13, 1994 {_ 5 q r-=-(2 J.J-L;q,tJ USHRC COMMENTS OF OHIO CITIZENS FOR RESPONSIBLE ENJ}RGn i8c P~4~~~ 8 dN4 57 : TION" (59 FED. REG. 28496, JUNE 2, 1994) OFF ICE OF OOCKETI G &
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PROPOSED RULE, "LIST OF APPROVED SPENT FUEL STORAGE CASKS : ADDI-5 '. . V SECRET RY G & SER 1/ICE ANCH The NRC is proposing to add the Standardized Horizontal Modular Storage System to the list of approved spent fuel storage casks in 10 CFR Part 72. OCRE supports this proposed rule. The NUHOMS system has the advantage of using the same dry storage canister for onsite storage, transportation, and ultimate disposal of the spent fuel, thereby greatly reducing fuel handling operations and consequent-ly reducing occupational radiation exposure and the chance for leakage and contamination. OCRE does have the following specific comments regarding the NUHOMS application:
- 1. OCRE is pleased that the NRC is requiring daily measurement of the thermal performance of each Horizontal Storage Module.
Such daily measurements will give early warning of impaired heat transfer which could lead to overheating of the HSM concrete and the fuel rod cladding within the canister. Impaired heat trans-fer could result from blockage of the air vents by wind-blown debris or by animal nesting activity (e.g., paper wasps can block openings with their nests), or from atmospheric conditions lead-ing to stagnation of air flow.
- 2. The criticality analysis (SAR Section 3.3.4.1.3) assumes that irradiated fuel has been cooled 7.5 years following discharge from the reactor, and takes credit for fission product neutron absorbers. Since the minimum cooling time for spent fuel to be stored in the NUHOMS system is 5 years, OCRE assumes that the use of the 7.5 year cooling time in the criticality analysis is done for conservatism; i.e .* as time passes, the quantity of the fission product absorbers will diminish due to radioactive decay.
OCRE would therefore question whether the assumption of the 7.5 year cooling time is conservative, as it is very likely that even older fuel will be stored in the casks. (Compare SAR Section 7.4.2, which states that "given the average age of fuel in U.S. storage pools, and the most probable NUHOMS loading schedules, filled NOHOMS ISFSis should have substantially older fuel than [10 years].")
- 3. SAR Section 7.4, "Estimated On-site Collective Dose Assess-ment," assumes that the spent fuel stored in the NUHOMS system has been cooled for 10 years before placement in the cask. The SAR states that 10 year cooled fuel was chosen "since it is a physical impossibility for a utility to have a facility full of five year fuel." However, it is certainly possible that at least some of the fuel stored in the system will be five year fuel, which will result in higher radiation levels and on-site doses.
In fact, Section 1.1.7 of the Attachment A of the Draft Certifi-1 SEP 2 8 1994 Acknowledged by card ........................." ..*-
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cate of Compliance requires that the first DSC be loaded with fuel assemblies constituting a heat source of approximately 24kW, which means that first DSC will probably use five year fuel. OCRE believes that the assumption of 10 year cooled fuel for the dose assessment is nonconservative.
- 4. The Draft SER (Section 5.2) dismisses corrosion of the DSC with one sentence: "Because all of the parts of the confinement boundary are fabricated from stainless steel, the DSC is ade-quately protected from corrosion mechanisms." The SAR (Section 4.6, "Cathodic Protection") notes that the DSC is filled with helium, which provides an inert atmosphere, and that the DSC temperature is well above ambient air temperature such that there will be no condensation on exterior surfaces. OCRE would note that some stainless steels are subject to corrosion under certain conditions. Situations can be postulated which would result in wetting of the DSC exterior surfaces, such as flooding and rains driven by high winds. The potential for corrosion should be analyzed in a rigorous manner instead of being summarily dis-missed. Since it is likely that the NUHOMS systems will be used to store spent fuel for decades, before a disposal facility is available, the long term potential for corrosion must be evaluat-ed.
- 5. The combination of the transfer cask containing the DSC has only been analyzed for an accidental drop of 80 inches. This is conservative for handling outside the spent fuel building.
However, absolutely no consideration has been given for potential accidental drops of the transfer cask and the DSC inside the spent fuel building, where the potential for accidental drops much greater than 80 inches exists. Indeed, such accidents inside the spent fuel building are more likely because of the lifting of the cask in and out of the fuel pool and onto the transport trailer. The cask must be analyzed for the maximum possible drop, regardless of whether that drop can occur inside or outside a building. Respectfully submitted, Susan L. Hiatt Director, OCRE 8275 Munson Road Mentor, OH 44060-2406 (216) 255-3158 2
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(SC/ Ff< 2'y'-/ 96} NUCLEAR REGULATORY COMMISSION r 10 CFR Part 72 RIN 3150-AF02 List of Approved Spent Fuel Storage Casks: Addition AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule.
SUMMARY
- The Nuclear Regulatory Commission (NRC) is proposing to amend its regulations to add the Standardized NUHOMS Horizontal Modular Storage System to the List of Approved Spent Fuel Storage Casks. This amendment will allow the holders of power reactor operating licenses to store spent fuel in this approved cask under a general license.
DATES: J-/ 16/o/'f Submit comments by (insert date 75 days after date of publication in the Federal Register). Comments received after this date will be considered if it is practical to do so, but the Commission is able to assure consideration only for comments received on or before this date . ADDRESSES: Send comments to: The Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555 . ATTN: Docketing and Service Branch. Deliver comments to: One White Flint North, 11555 Rockville Pike, Rockville, Maryland, between 7:45 a.m. and 4:15 p.m. Federal workdays.
Copies of the comments received and the environmental assessment and l finding of no significant impact can be examined at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of these documents can be obtained from Mr. G. E. Gundersen, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3803. FOR FURTHER INFORMATION CONTACT: Mr. G. E. Gundersen, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3803; or Mr K. C. Leu, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 504-2685. SUPPLEMENTARY INFORMATION:
Background
Section 218(a) of the Nuclear Waste Policy Act of 1982 (NWPA) directs that, "[T]he Secretary [of the Department of Energy (DOE)] shall establish a demonstration program in cooperation with the private sector, for the dry storage of spent nuclear fuel at civilian power reactor sites, with the objective of establishing one or more technologies that the [Nuclear Regulatory] Commission may, by rule, approve for use at the sites of civilian nuclear power reactors without, to the maximum extent practicable, the need for additional site-specific approvals by the NRC." Section 133 of the NWPA states, in part, that "the Commission shall, by rule, establish procedures for 2
the licensing of any technology approved by the Commission under Section 218(a) for use at the site of any civilian nuclear power reactor . " To implement this mandate, the Commission approved dry storage of spent nuclear fuel in NRC-approved casks, publishing a final rule on 10 CFR Part 72 entitled "General License for Storage of Spent Fuel at Power Reactor Sites" (55 FR 29181). This rule also established a new Subpart L within 10 CFR Part 72 entitled "Approval of Spent Fuel Storage Casks," containing procedures and criteria for obtaining NRC approval of dry storage cask designs. The 1990 rulemaking listed four casks in§ 72.214 of Subpart K as approved by the NRC for storage of spent fuel at power reactor sites under general license by persons authorized to possess or operate nuclear power reactors. Since then, two more casks have been listed in§ 72.214, one on April 7, 1993 (58 FR 17948) and another on October 5, 1993 (58 FR 51762). Discussion This proposed rulemaking would add the Standardized NUHOMS Horizontal Modular Storage System to the list of NRC approved casks for spent fuel storage in§ 72.214. Following the procedures specified in§ 72.230 of Subpart L, VECTRA Technologies, Inc. (formerly Pacific Nuclear Fuel Services, Inc. (PNFSI)) 1 submitted an application for NRC approval, together with a "Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel" (SAR), NUH-003, Revision 2, dated November 1993. The NRC evaluated VECTRA's submittal and issued a draft Safety 1 0n January 24, 1994, Pacific Nuclear Systems, Inc., (parent company of PNFSI) changed its name to VECTRA Technologies Inc. 3
Evaluation Report (SER) on VECTRA'S SAR and a draft certificate of compliance for the Standardized NUHOMS Horizontal Modular Storage System. On January 24, 1994, Pacific Nuclear Systems, Inc., (parent company of PNSFI) changed its name to VECTRA Technologies, Inc., after it acquired ABB Impell Corporation. The NRC is proposing to approve VECTRA's Standardized NUHOMS Modular Storage System for Irradiated Nuclear Fuel, for storage of spent fuel under the conditions specified in the draft certificate of compliance. This cask, when used in accordance with the conditions specified in the certificate of compliance and NRC regulations, will meet the requirements of 10 CFR Part 72; thus, adequate protection of the public health and safety would be ensured. This cask is being proposed for listing under§ 72.214, "list of Approved Spent Fuel Storage Casks" to allow holders of power reactor operating licensees to store spent fuel in this cask under a general license. The certificate of compliance would terminate 20 years after the effective date of the final rule listing the cask in § 72.214, unless the cask's certificate of compliance is renewed. The certificate contain*s conditions for use which are similar to those for other NRC-approved casks; however, the certificate of compliance for each cask may differ in some specifics--such as, certificate number, operating procedures, training exercises, spent fuel specification. The draft certificate of compliance for the Standardized NUHOMS cask and the underlying draft SER, are available for inspection and comment at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC. Single copies of the proposed certificate of compliance may be obtained from Mr. K. C. Leu, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 504-2685. 4
Submission of Comments in Electronic Format In addition to the original paper copy, commenters are encouraged to submit a copy of the letter in electronic format on IBM PC-compatible 5.25- or 3.5-inch computer diskette. Data files should be provided in one of the following formats: WordPerfect, IBM Document Content Architecture/Revisable-Form-Text (DCA/RFT), or unformatted ASCII text. Finding of No Significant Environmental Impact: Availability Under the National Environmental Policy Act of 1969, as amended, and the NRC regulations in Subpart A of 10 CFR Part 51, the NRC has determined that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment, and therefore, an environmental impact statement is not required. The rule is mainly administrative in nature. It would not change safety requirements and would not have significant environmental impacts. The proposed rule would add one cask known as the Standardized NUHOMS Modular Storage System to the list of approved spent fuel storage casks that power reactor licensees can use to store spent fuel at reactor sites without additional site-specific approvals by the NRC. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the environmental assessment and finding of no significant impact are available from Mr. G. Gundersen, Office of Nuclear Regulatory Research, 5
U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone (301) 492-3803. Paperwork Reduction Act Statement This proposed rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget, Approval Number 3150-0132 . Regulatory Analysis On July 18, 1990 (55 FR 29181), the Commission issued an amendment to 10 CFR Part 72. The amendment provided for the storage of spent nuclear fuel under a general license. Any nuclear power reactor licensee can use these casks if (1) they notify the NRC in advance, (2) the spent fuel is stored under the conditions specified in the cask's certificate of compliance, and (3) the conditions of the general license are met. In that rulemaking, four spent fuel storage casks were approved for use at reactor sites, and were listed in 10 CFR 72.214. That rulemaking envisioned that storage casks certified in the future could be routinely added to the listing in§ 72 . 214 through rulemaking procedures. Procedures and criteria for obtaining NRC approval of new spent fuel storage cask designs were provided in 10 CFR 72.230. Subsequently, two additional casks were added to the listing in§ 72.214 in 1993. 6
The alternative to this proposed action is to withhold certification of this new design and give a site-specific license to each utility that proposed to use the cask. This alternative however, would cost the NRC more time and money for each site-specific review. In addition, withholding certification would ignore the procedures and criteria currently in pl~ce for the addition of new cask designs. Further, it is in conflict with NWPA direction to the Convnission to approve technologies for the use of spent fuel storage at the sites of civilian nuclear power reactors without, to the extent practicable, the need for additional site reviews. Also, this alternative is anticompetitive in that it would exclude new vendors without cause and would arbitrarily limit the choice of cask designs available to power reactor licensees. Approval of the proposed rulemaking would eliminate the above problems. Further, the .proposed rule will have no adverse effect on the public health and safety. The benefit of this proposed rule to nuclear power reactor licensees is to make available a greater choice of spent fuel storage cask designs which can be used under a general license. However, the newer cask design may have a market advantage over the existing designs in that power reactor licensees may prefer to use the newer casks with improved features. The new cask vendors with casks to be listed in§ 72.214 benefit by having to obtain NRC certificates only once for a design which can then be used by more than one power reactor licensee. Vendors with cask designs already listed may be adversely impacted in that power reactor licensees may choose a newly listed design over an existing one. However, the NRC is required by its regulations and NWPA direction to certify and list approved casks. The NRC also benefits 7
because it will need to certify a cask design only once for use by multiple licensees. Casks approved through rulemaking are to be suitable for use under a range of environmental conditions sufficiently broad to encompass multiple nuclear power plants in the United States without the need for farther site-specific approval by NRC. This proposed rulemaking has no significant identifiable impact or benefit on other Government agencies. Based on the above discussion of the benefits and impacts of the alternatives, the NRC concludes that the requirements of the proposed rule are commensurate with the Commission's responsibilities for public health and safety and the common defense and security. No other available alternative is believed to be as satisfactory, and thus, this action is recommended. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the Commission certifies that this rule will not, if promulgated, have a significant economic impact on a substantial number of small entities. This proposed rule affects only the licensing and operation of nuclear power plants and cask vendors. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121. 8
Backfit Analysis The NRC has determined that the backfit rule (10 CFR 50.109 or 10 CFR 72.62) does not apply to this proposed rule, and thus, a backfit analysis is not required for this proposed rule because this amendment does not involve any provisions which would impose backfits as defined in the backfit rule. List of Subjects In 10 CFR Part 72 Manpower training programs, Nuclear materials, Occupational safety and health, Reporting and recordkeeping requirements, Security measures, Spent fuel. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and 5 U.S.C. 553; the NRC is proposing to adopt the following amendments to 10 CFR Part 72. PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE The authority citation for Part 72 continues to read as follows: AUTHORITY: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 184, 186,187, 189, 68 Stat. 929, 930,932, 933, 934, 935, 948, 953, 954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2071, 2073, 2077, 2092, 9
2093, 2095, 2099, 2111, 2201, 2232, 2233, 2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat. 688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); Pub. L. 95-601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851); sec. 102, Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135, 137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148, Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153, 10155, 10157, 10161, 10168). Section 72.44(9) also issued under secs. 142(b) and 148(c), (d), Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b), 1068(c)(d)). Section 72.46 also issued under sec. 189, 68 Stat. 955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42 U.S.C. 10154). Section 72.96(d) also issued under sec. 145 (g), Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(9)). Subpart J also issued under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-425, 96 Stat. 2202, 2203, 2204, 2222, 2244, (42 U.S.C. 10101, 10137(a), 10161(h)). Subparts Kand Lare also issued under sec. 133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252 (42 U.S.C. 10198). In § 72.214, Certificate of Compliance 1004 is added to read as follows: § 72.214. List of approved spent fuel storage casks. Certificate Number: 1004 SAR Submitted by: VECTRA Technologies, Inc. 10
SAR
Title:
Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, Revision 2 Docket Number: 72-1004 Certification Expiration Date: (20 years after final rule effective date) Model Numbers: NUHOMS-24P for Pressurized Water Reactor fuel NUHOMS-52B for Boiling Water Reactor fuel. d Dated at Rockville, Maryland, this .M._ day of d£ _,, , 1994. For the Nuclear Reg:.eory Commission. or for Operations. 11}}