ML20127E375
| ML20127E375 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 10/28/1992 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| References | |
| SECY-92-366, NUDOCS 9211030069 | |
| Download: ML20127E375 (303) | |
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POLICY ISSUE October 28, 1992 SECY-92-366 For: The Commissioners From: James M. Taylor F Executive Director for Operations
Subject:
PROPOSED LICENSE, UNDER 10 CFR PART 72, FOR DRY CONCRETE MODULE STORAGE OF SPENT FUEL AT BALTIM0RE GAS AND ELECTRIC COMPANY'S CALVERT CLIFFS NUCLEAR POWER PLANT SITE
Purpose:
To obtain Commission authorization, pursuant to 10 CFR 2.764(c), for the Office of Nuclear Material Safety and Safeguards (NMSS) to issue an initial license au*,horizing Baltimore Gas and Electric Company (BG&E) to possess spent fuel from Cahert Cliffs in an independent spent fuel storage installation (ISFSI). It would be on the Calvert Cliffs Nuclear Power Plant site and would employ the Pacific Nuclear Fuel Services, Inc. (PNFS, formerly NUTECH) NUHOMS-24P concrete module and steel canister dry storage system.
The ISFSI would consist of up to 120 modules on concrete pads.
Summary: The staff has completed its safety, safeguards, and environmental reviews of the proposed storage of dry canistered spent fuel in concrete modules on BG&E's Calvert Cliffs site. A proposed license (Encl. 1) has been prepared for signature by the Director, NMSS, or his designee, based on the staff's Safety Evaluation Report (SER) (Encl. 2).
Only spent pressurized water reactor (PWR) fuel from
Contact:
FSturz, NMSS, SCDB 504-2684 SECY NOTE: TO BE MADE PUBLICLY AVAILABLE WHEN THE FINAL SRM IS MADE AVA1(ABLE.
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.. b e' The Commissioners Calvert Cliffs is to be stored in stainless steel canisters in horizontal concrete modules at the site. The storage activities and location are covered under the existing Price Anderson indemnity agreement for the Calvert Cliffs site.
No offsite transportation of spent fuel is involved. A canister containing 24 spent fuel assemblies will be transferred to each module in a transfer cask. Heavy loads involved in cask handling will be within the auxiliary building's crane capacity of 125 tons, when authorized upgraf s have been completed. A BG&E evaluation has shown that dry storage activities do not represent an unreviewed safety question for reactor operations, and that no C additional changes to the technical specifications of the reactor operating licenses are necessary. The ISFSI consists, in this instance, of a fenced area outside the reactor protected area, which is to contain up to a total of 120' concrete modules. At its own risk, BG&E has begun construction of 48 storage modules and has initiated the purchase of otner storage system components from vendors.
Discussion: On December 21, 1989, Nuclear Regulatory Commission staff received a BG&E application for spent fuel storage in a dry modular storage system to be located at the Calvert Cliffs site. BG&E has relied on some of the topical report submitted by NUTECH, Inc. (now PNFS), for its NUTECH Horizontal Modular Storage System (NUHOMS) typo 24P, a concrete module and stainless steel canister design, and on the safety review of this design by the staff. BG&E's Safety Analysis Report (SAR) and its Environmental Report were based on this design and referenced the basic NUHOMS topical report design, where appropriate.
In April 1989, the staff completed its safety review of the topical report for the NUHOMS system design and issued a letter of approval with an SER. The staff reviewed criticality, strtctural, thermal, and shielding aspects of the design, under normal and accident conditions. The staff concluded that the NUHOMS system could be used to safely storc 24 PWR spent fuel assemblies (a maximum of 0.66 kW heat output per assembly), as proposed by the vendor.
Characteristics of the spent fuel allowed to be stored and other operating limits are set forth in the SER for the NUHOMS topical report. These limits are carried forward, with appropriate modific dion, in the safety evaluation for BG&E's application, and are included in the technical specifications of the proposed license (Encl. 1).
In January 1990, NRC issued a materials license, under 10 CFR Part 72, to Duke Power Company, for an ISFSI, at its Oconee Nuclear Station, using a similar NUHOMS-24P system.
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l-The Commissioners In March 1991, the staff, having completed its environmental review of the Calvert Cliffs ISFSI, published a " Finding of No Significant Impact," in the Federal Reaister, and issued an environmental assessment (EA) (Encl. 3).
The staff has completed its safety review of the Calvert Cliffs site application. The application, as supplemented, includes confirmation by the applicant's reactor safety committee that: (a) no technical specification changes are required, under the Calvert Cliffs Facility Operating Licenses, to accommodate a Part 72 license for onsite 3 storage; (b) the joint operation of the reactor and onsite -
storage does not affect the safety margins of either one; and (c) onsite storage is an independent operation, as defined in Part 72. The staff has found the applicant's review acceptable. The staff has also found BG&E's application, on the basis of the staff's review, to be acceptable. Based on the staff's review of the applicant's submissions, the staff has found that there are no remaining unreviewed safety questions, and that all other pertinent regulatory requirements for authorization of issuance of the requested license have been sat;sfied. Regarding removal of canistered spent fuel before ISFSI decommissioning, BG&E expects to be able to directly ship canistered fuel offsite and to avoid spent fuel unloading in reactor basins.
Depending on economic and regulatory ce "itions at the time of decommissioning, the dry shielded canister (DSC) could be returned to the spent fuel pooi, cut open, and the spent t fuel assemblies loaded into a certified transportation cask, for shipping to the repository. Also, in response to Commission comment, the Department of Energy has stated its agreement with the Commission on the issue of such offsite transportation -in its February 1989 " Final Version Dry Cask Storage Study" (D0E/RW-0220) and stated its plans to increase its efforts to ensure coordination with utilities.
This proposed license includes an exemption from providing )
instrumentation and controls systems to monitor systems that are important to safety, as required in the general design criteria of the Commission's regulations. Specifically, no instrumentation and control systems are provided or requf red for the DSC and HSM during storage operations because the NUH0MS-24P is a passive design. Instrumentation for pre-storage operations is provided for draining, purging, and drying of the DSC as well as the backfilling of the DSC. In the Oconee and H. B. Robinson ISFSI licenses, which used dry spent fuel storage designs similar to Calvert Cliffs, the staff concluded that because of the passive design of the NUHOMS system, no systems important to safety are required l
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Af b The Commissioners. i for storage' operations and no instruments important to safety are required. The staff's finding, in the safety l evaluation reports (SERs) for these license-applications, as-well as Calvert Cliffs, was that instrumentation and control systems were not required. However, because of questions raised about whether or not 10 CFR 72.122(i) provides for ,
this type of flexibility, the staff has now determined that '
it is not enough to simply state that instrumentation and- -)
controls are not required. :Rather, an exemption from the requirement should also be granted in_ each of the above Part 72 licenses to make it clear that instrumentation and controls are not required to ensure the operation of the systems important to safety. Therefore, an exemption is recommended from application of the general design criterion. In addition, the staff intends-to issue exemptions to the Oconee and H.B.' Robinson Part 72 licenses to reflect this interpretation regarding instrumentation and controls for passive systems that are considered important to safety. At the earliest opportunity, staff will request rulemaking to include this clarification in Part 72.
Accordingly, the staff has prepared its SER (Encl. 2), for the Calvert Cliffs ISFSI, which references the staff's April 1989 SER for the NUHOMS-24P design, making the appropriate findings. The staff has prepared a proposed Part 72 license-with technical specifications, including license conditions that also satisfy safeguards requirements of 10 CFR Part 73, for Calvert Cliffs site spent fuel storage.
Conclusion:
The NRC staff has found that, based upon'its enclosed SER-and the previously issued EA, there is reasonable assurance that the activities, authorized by a license, to construct and operate the Calvert Cliffs ISFSI, can be conducted without endangering the health and safety of the pubite, without significant environmental impact, and in _ compliance with the conditions of the license and the Commission's regulations. The staff also finds that the issuance of this -
license will not be inimical to the common defense and security.
Coordination: The Office of.the General Counsel has reviewed this paper and has no legal objections.
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o u The Commissioners Recommendation: That the Commission authorize issuance of a license to BG&E, under 10 CFR Part 72, to receive, transfer, and store spent fuel in a dry storage modular system ISFSI, on the Calvert Cliffs Nuclear Power Plant site.
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Enclosures:
- 1. Proposed License
- 2. Safety Evaluation Reoort
- 3. Environmental Assessment DISTRIBUTION:
Commissioners OGC CAA IG OPP EDO ACRS ACNW SECY Commissioners comments or consent should be provided directly to SECY by c.o.b. Thursday, November 12, 1992.
Commission staff office comments, if any, should be submitted to the Commissioners NLT November 4, 1992, with an information copy to SECY. If the proer is of such a nature that it requires additional review and crmment, the Commissioners and the Secretariat should be apprised of when comments may be expected.
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BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS INDEPENDENT SPENT FUEL STORAGE INSTALLATION MATERIALS LICENSE NO. SNM-2505 The Nuclear Regulatory Commission (the Commission) has found that:
A. The application, filed by the Baltimore Gas and Electric Company (CP&L) (applicant) for a materials license to receive, store, and transfer spent fuel from Calvert Cliffs Nuclear Power Plant in an independent spent fuel storage installation (ISFSI) _ located at its Calvert Cliffs Nuclear Power Plant site, meets' the standards and requirements of the Atotic Energy Act of 1954, as amended (Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The Calvert Cliffs ISFSI will operate in conformity with the application, as amended, " ' provisions of the Act, and the rules and regulations of the Co sion;
- c. The proposed site complies with the criteria in Subpart E of 10 CFR Part 72; D. The proposed ISFSI will not pose an undue risk to the safe operation of the Calvert Cliffs Nuclear Power Plant Units 1 and 2; E. The applicant's proposed ISFSI design complies with the 10 CFR 72, Subpart F, with the exception of 10 CFR 72.122(i), with respect to providing instrumentation and control systems for the DSC and HSM during storage operations; F. The applicant is qualified by reason of training and experience to conduct the operation covered by the regulation in 10 CFR Part 72; G. The applicant's proposed operating procedures to protect health and to minimize danger to life and property are adequate; H. The applicant is financially qualified to engage in the activities in accordance with the regulations in 10 CFR Par; 72;
- 1. The applicant's proposed quality assurance plan complies with 10 CFR Part 72, Subpart G; J. The applicant's proposed physical protection provisions comply with 10 CFR Part 72, Subpart H; K. The applicant's proposed personnel training program complies with 10 CFR Part 72, Subpart I; L. The applicant's proposed decommissioning plan pursuant to 10 CFR 72.30 provides reasonable assurance that the decontamination and decommissioning of the Calvert Cliffs ISFSI at the end of its useful ENCLOSURE 1
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l life will provide adequate protection to the health and safety of the public.
M. The applicant's proposed emergency plan complies with 10 CFR 572.32; 1 I
N. The applicant has satisfied the applicable provisions of 10 CFR Part i 170; j
- 0. There is reasonable assurance (1) that the activities authorized by ;
the license can be conducted without endangering the health and i safety of the public, and (2) that such activities will be conducted ;
in compliance with the regulatior3 of the Commission set forth in i 10 CFR Chapter I; and P. The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public.
Accordingly, based on the foregoing findings, Materials License No. SNM-2505 is hereby issued to Baltimore Gas and Electric Company to read as follows:
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a o Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974 (Public Law 93-438), and Title 10, Code of Federal Regulations, Chapter I, Part 72, and in reliance on statements and representations heretofore made by the licensee (in the licensee's Independent Spent Fuel Storage Installation Safety Analysis Report, Calvert Cliffs Nuclear Power Plant submitted by letter dated December 21, 1989, as revised, supplemented, and submitted by letters dated April 26, 1990, June 29, 1990, November 1, 1990, December 20, 1990, February 1, 1991, February 12, 1991, September 30, 1991, October 18, 1991, December 19, 1991, Decamber 27, 1991, August 18, 1992, and September 4, 1992), a license is hereby issued
'horizing the licensee to receive, acquire, and possess the power reactor fuel and other radioactive materials associated with spent fuel storage
( s aated below; to use such materials for the purposes and at the place designated below; to deliver or transfer such materials to persons authorized to receive these materials in accordance with the regulations of the applicable parts of 10 CFR Chapter I. This license shall be deemed to contain the conditions specified in Section 183 of the Atomic Energy Act of 1954, as amended, and is subject to all applicable rules, regulations, and orders of the Nuclear Regulatory Commission now or hereafter in effect and to any conditions specified herein.
Licensee
- 1. Baltimore Gas and Electric Company
- 3. License Number: SNM-2505
- 2. Address: 1992 Charles Center 4. Expiration Date:
P.O. Box 1475 8altimore, Maryland 21203 5. Docket Number: 72-8
- 6. Byproduct, source, and/ 7. Chemical and/or 8. Maximum amount or special nuclear physical form that licensee may mai.eri al possess at any one time under this license A. Spent fuel assemblies A. clad with A. 1,111.68 TeV of from Calvert Cliffs Nuclear As zirconU0,ium or spent fuel l
Station Units 1 and 2 zirconium alloys Assemblies reactor using natural water for cooling and enriched not greater thn 4.5 percent U-235 and asse ia.ted radio-active materiais related to receipt, storage, and transfer of the fuel assemblies 1
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- 9. Authorized Use:
The material identified in 6.A and 7.A above is authorized for receipt, -
possession, storage and transfer.
- 10. Authorized Place of Use:
The licensed material is to be received, possessed, transferred, and stored at the Calvert Cliffs ISFSI located on the Calvert Cliffs Nuclear Power Plant site in Calvert County, Maryland.
- 11. This site is describcd in Chapter 2 of the licensee's Safety Analysis Report for the Calvert Cliffs ISFSI.*
- 12. The Technical Specifications contained in Appendix A attached hereto are incorporated in the license. The licensee shall operate the installation.
in accordance with the Technical Specifications in Appendix A.
- 13. The licensee shall fully implement and maintain in effect all provisions of the Independent. Spent Fuel Storage Installation (ISFSI) physical security, guard training and qualification, and safeguards contingency plans previously approved by the Commission and all amendments made pursuant to the authority of 10 CFR 72.56 and 10 CFR 72.44(e) and 72.186.
The plans, which contain safeguards information protected under 10 CFR 73.21, .are entitled: "Calvert Cliffs Nuclear Power Plant Independent -
Spent Fuel Storage Installation Physical Security Plan" with Revision 2, submitted by letter dated September 4, 1992; "Calvert Cliffs Nuclear-Power P1 ant Independent -Spent Fuel Storage Installation Training and -
Qualification Plan" with Revision 1, submitted by letter dated ~
February 1,1991; and "Calvert C1'iff.s Nuclear Power Plant Independent Spent Fuel Storage Installation Safeguards Contingency Plan" with Revision 1, submitted by letter dated February 1,1991.
- 14. The Technical Specifications for Eneiror ental Protection contained in Appendix A attached hereto are incorporated in the license.
Specifications required pursuant to S 72.44(d), stating limits on the release of radioactive materials for compliance with limits of 10 CFR Part 20 and "as low as is reasonably achievable objective" for effluents are not applicable. Dry Shielded Chnister (DSC) external surface ,
contamination within the limits of Technical Specification 3.2.3.1 ensures that the offsite dose will be inconsequential. In addition, there are no-normal or off-normal releases =or effluents expected from the double-sealed storage canisters of the ISFSI.
Specifications requireo pursuant to S 72.44(d)(1) for operating procedures, for control of effluents, and for the maintenance and use of equipment in radioactive waste treatment-systems to meet the. requirements of 9 72.104 are not applicable. -There are, by the design of the sealed storage canisters at the ISFSI, no effluent releases. Also, all Calvert-2 4
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Cliffs site DSC and Transfer Cack (TC) loading and unloading operations and waste treatment therefrom will occur at the Calvert Cliffs Nuclear Power Plant under the specifications of its operating licenses.
- 15. No spent nuclear fuel shall be allowed to be loaded untii such time as the following preoperational license conditions are satisfied:
A. A training exercise (Dry Run) of all DSC, TC and horizontal storage module (HSM) loading and handling activities shall be held which shall include but not be limited to those listed and which need not be performed in the order listed:
- a. Loading DSC in cask.
- b. DSC (lennth may be truncated) drying, welding, and cover gas backfilling operations.
- c. Moving cask to and aligning and docking with HSH on the storage pad.
- d. Insertion of DSC in HSM.
- e. Withdrawal of DSC from HSM.
- f. Returning the cask to the decontamination pit.
- g. Removing the cask lid and cutting open the DSC (length may be truncated) assuming fuel cladding failure.
- h. Removing the DSC from the cask.
- i. All dry run activities shall se done using written procedures.
J. The activities listed above shall be performed or modified and performed to show that each activity can be successfully executed prior to actual fuel loading.
B. The Calvert Cliffs Nuclear Power Plant Emergency Plan shall be reviewed and modified as required to include the ISFSI.
C. A training module shall be developed for the Calvert Cliffs Nuclear Power Plant Training Program establishing an ISFSI Tr aining and Certification Program which will include the following:
an (overview)
- b. ISFSI Facility /k :gn (overview)
- c. ISFSI Safety Analysis (overview) 3
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procedures i
- e. ISFSI License (overview).
D. The Calvert Cliffs Nuclear Power Plant health physics procedures shall be reviewed and modified as required to include the ISFSI.
E. The Calvert Cliffs Nuc1 car Power Plant Administrative Procedures shall be reviewed and modified as required to include the ISFSI.
F. A procedure shall be developed and impicmented for the documentation of the characterizations performed to select spent fuel to be stored in the canisters and modules. Such procedure shall include independent verification of fuel assembly selection by an individual other than the original individual making the selection.
G. A procedure shall be developed and implemented for two independent determinations (two samples analyzed by different individuals) of the boron concentration in the water used to fill the DSC cavity for fuel loading and unloading activities.
H. Written procedures shall be implemented to describe actions to be taken during operation and abnormal / emergency conditions.
- 16. The design, construction, and operation of the ISFSI shall be accomplished in accordance with the U.S. Nuclear Regulatory Commission Regulations specified in Title 10 of the U.S. Code of Federal Regulations. All commitments to the applicable NRC Regulatory Guides and to engineering and construction codes shall be carried out.
- 17. The double closure seal welds at the bottom end of the DSC shall satisfy the Liquid Penetrant Acceptance Standards of ASME B&PV Code Section III, Division 1, Subsection NB-5350 (1983) Addf tionally, these seal welds at the bottom of the DSC shall be leak tested in accordance with ANSI N14.5 (1987). The leakage rate shall be quantified and shall not exceed the leak rate specified in Technical Specification 3.2,2.2 of Appendix A.
(i.e., standard helii- leak rate shall not exceed 10 atm-cc/s).
l 18. Fuel and TC movement and handling activities wH ch are to be performed in the Calvert Cliffs Nuclear Power Plant Auxiliary Building will be I governed by the requiremet. t of the Calvert Cliffs Nuclear Power Plant f acility Operating Licenses (000-53, tnd -69) and a:isociated Technical Specifications.
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- 19. Pursuant to 10 CFR 72.7, the licensee is hereby exempted from the provisions of 10 CFR 72.122(i) with respect to providing instrumentation and control systems for the DSC and HSH during storage operations.
- 20. This license is effective as of the date of issuance shown below.
for the U.S. Nuclear Regulatory Commission Date of Issuance: 1992 by Division of Industrial and Medical Nuclear Safety Washington, DC 20555
- Hereafter referred to in this license as the SAR.
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CALVERT CLIFFS INDEPENDENT SPENT FUEL STORAGE INSTALLATION ;
l APPENDIX "A' TO MATERIALS LICENSE SNM-2505 TECHNICAL SPECIFICATIONS ISSUED BY THE UNITED STATES NUCLEAR REGULATORY COMMISSION
o e TABLE OF CONTENTS i
1.0 DEFINITIONS SECTION fME ADMINISTRATIVE CONTR0LS.................... 1-1 DESIGN FEATURES.............. ............. 1-1 FUEL ASSEhBLY.............................. 1-1 FUNCTIONAL AND OPERATING LlHITS............ 1-1 LIMITING CONDITIONS........................ 1-1 LOADING 0PERATIONS......................... 1-1 SURVEILLANCE INTERVAL...................... 1-1 SURVEILLANCE REQUIREMENTS.................. 1-2 2.0 E.UNCTIONAl AND OPERATING LIMITS SECTION 2.1 FUEL 10 BE STORED AT ISFSI................. 2-1 2.2 DRY 41ELDED CANISTER (DS0)................ 2-2 2.2.1 DSC VACUUM DURING DRYING................... 2-2 2.2.2 DSC HEllUM BACKFILL PRESSURE............... 2-2 2.3 TRANSFER CASK (TC)......................... 2-3 2.4 HORIZONTAL STORAGE MODULE (HSM)............ 2-4 BASES SECTION 2.1 FUEL TO BE STORED AT iSFSI................. B 2-1 2.2 DRY SHIELDED CANISTER (D5C)................ B 2-2 2.2.1 DSC VACUUM PRESSURE DURING DRYING.......... B 2-2 2.2.2 DSC HELIUM BACKFILL PRESSURE............... B 2-2 2.3 TRANSFER CASK (TC)......................... B 2-3 2.4 HORIZONTAL STORAGE MODULE (HSM)............ B 2-4 3/4.0 LIMITING CONDITIONS / SURVEILLANCE RE0VIREMENTS SECTION 3/4.1 FUEL TO BE STORED AT ISFSI....... ......... 3/4 1-1 3/4.2 DRY SHIELDED CANISTER (0SC)................ 3/4 2-1 3/4.2.1 DISSOLVED BORON CONCENTRATION.............. 3/4 2-1 3/4.2.2 DSC CLOSURE WELDS.......................... 3/4 2-2 3/4.2.3 .DSC EXTERIOR SURFACE CONTAMINATION......... 3/4 2-3 3/4.3 TRANSFER CASK-(TC)......................... 3/4 3-1 3/4.3.1 AMBIENT TEMPERATURE........................ 3/4 3-1 3/4.4 HORIZONTAL STORAGE MODULE (HSM)............ 3/4 4-1 3/4.4.1 MAXIMUM AIR TEMPERATURE RISE............... 3/4 4-1 3/4.5 FIRE PR0TECTION............................ 3/4 5-1 1
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1 BASES SECTION ,
3/4.1 FUEL 10 BE STORED AT ISFSI................. B 3/4-1 3/4.2 DRY SHIELDED CANISTER (D50)................ B 3/4-2 3/4.2.1 DISSOLVED BORON CONCENTRATION.............. B 3/4-2 3/4.2.2 DSC CLOSURE WELD 5.......................... B 3/4-2 3/4.2.3 DSC EXTERIOR SURFACE CONTAMINATION......... B 3/4-2 3/4.3 T RAN S F E R C AS K ( TC ) . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4-3 3/4.3.1 AMBIENT TEMPERATURE........................ B 3/4-3 3/4.4 HORIZONTAL STORAGE MODULE.................. B 3/4-4 3/4.4.1 MAXIMUM TEMPERATURE RISE................... B 3/4-4 3/4.5 FIRE PROTECT 10N............................ B 3/4-5 SECTION 5.0 DESIGN FEATURES............................ 5-1 6.0 ADMINISTRATIVE CONTROLS 6.1 GENERAL.................................... 6-1 6.2 ENVIRONMENTAL MONITORING PROGRAM........... 6-1 6.3 ANNUAL ENVIRONMENTAL REPORT................ 6-1 11
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r INTRODUCTION These Technical Specifications govern the safety of the receipt, possession, and storage of irradiated nuclear fuel at the Calvert Cliffs Independent Spent Fuel Storage Installation and the transfer of such irradiated nuclear fuel to and from Units 1 and 2 of the Calvert Cliffs Nuclear Power Plant and the Calvert Cliffs Independent Spent Fuel Sturage Installation. The protection of the environment during the activities described above is also governed under these technical specifications. The loading of spent fuel.into the dry shielded canister (DSC) and transfer cask (TC) at the Calvert Cliffs Nuclear Power Plant Auxiliary Building is governed by the existing Calvert Cliffs 10 CFR Part 50 operating license (DPR-53 and -69), technical specifications, and new specific procedures.
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e O SECT 10ti 1.0 DEFINIT!0 tis 4
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e r 1.0 DEFINITIONS The following definitions apply for the purpose of these Technical Specifications:
- a. ADMINISTRATIVE CONTROLS: Provisions relating to organization operating, emergency, and management procedures, recordkeeping, review and audit, and reporting necessary to ensure that the ,
operations involved in the movement, transfer and storage of spent fuel at the Calvert Cliffs ISFSI are performed in a safe manner.
- b. DESIGN FEATURES: Features of the facility associated with the basic design such as materials of construction, geometric arrangements, dimensions, etc., which, if altered or modified, could have a significant effect on safety,
- c. FUEL ASSEMBLY: The unit of nuclear fuel in the form that is charged or discharged from the core of a light-water reactor (LWR). Normally, will consist of a rectangular arrangement of fuel and non-fuel held together by end fittings, spacers, and guide tubes.
- d. FUNCTIONAL AND OPERATING L1 HITS: Limits on fuel handling and storage conditions necessary to protect the integrity of the stored fuel, to protect employees against occupational exposures, and to guard against the uncontrolled release of radioactive materials.
- e. LIMITING CONDITIONS: The minimum or maximum functional capabilities or performance levels of equipment required for safe operation of the facility.
- f. LOADING OPERATIONS: Loading Operations include all cask preparation steps prior to cask transport from the auxiliary building area.
- g. SVRVEILLANCE INTERVAL: A surveillance interval is the interval between a surveillance check, test or calibration. Unless specifically stated otherwise, each surveillance requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval. '
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- h. SVRVElllANCE RE0VIREMENTS: Surveillance requirements include:
(1) inspection, test, and calibration activities to ensure that the necessary integrity of required systems, components, and the spent fuel in storage is maintained; (ii) confirmation that operation of the installation is within the required functional and operating limits; and (iii) a confirmation that the limiting conditions required for safe storage are met.
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i SECTION 2.0 FUNCTIONAL AND OPERATING LIMITS I l
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i 2.0 FUNCTIONAL AND OPERATING LIMITS 2.1 FVEL TO BE STORED AT ISFSI SPECIFICATION: Any fuel not specifically filling the requirements of Section 3.1 for maximum burnup and post irradiation time may be stored if it meets the minimum cooling time listed in the Calvert Cliffs ISFSI SAR Table 10.3.1 and all the following requirements are met:
Neutron Source Per Assembly s 2.23 x 10s n/sec/ assembly, with spectrum bounded by Table 3.1-4 of the Calvert Cliffs ISFSI SAR Gamma Source Per Assembly s 1.53 x 10" MeV/sec/ assembly with spectrum bounded by that shown in Table 3.1-4 of the Calvert Cliffs ISFSI SAR ;
APPLICABILITY: This specification is applicable to all spent fuel to be stored in the Calvert Cliffs ISFSI.
ACTION: If the requirements of the above specification are not met, do not load the fuel assembly into a DSC for storage.
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2.2 DRY SHIELDED CANISTER (DSC) 2.2.1 DSC VACUUM DURING DRYING SPECIFICATION: The DSC Cavity vacuum pressure during canister drying shall not exceed 3 torr (3 mm Hg) after stepped evacuation, The vacuum pressure shall be maintained for not less than 30 minutes.
APPLICABILITY: Applicable to all DSCs.
ACTION: If the required vacuum cannot be obtained.
- 1. Check and repair vacuum drying system as necessary.
If the specification is still not met, remove fuel from the DSC.
2.2.2 DSC HELIUM BACKFILL PRESSURE SPECIFICATION: The DSC cavity shall be backfilled with helium. The backfill pressure shall be 2.5 psig 2.5 psi.
APPLICABILITY: Applicable to all DSCs.
ACTION: If the required pressure cannot be.obtained:
- 1. Check and repair the vacuum drying system as necessary.
- 3. If the backfill pressure exceeds the criterion, release.a sufficient quantity of helium to lower the DSC cavity pressure.
If the specification is still not met, remove fuel from the DSC. -
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- 6 2.0 FUNCI DNAL AND OPERATING Li_MITS 2.3 TRANSFER CASK (TC)
SPECIFICATION: The transfer cask lifting height with a non-single-failure-proof lifting device shall not exceed 80 inches. j APPLICABillTY: This specification applies to handling of a loaded ISFSI transfer cask outside the Auxillary Building. I ACTION: In the event of a transfer cask drop from a height greater than 15 inches, (0.38 m) with fuel in a DSC and the DSC in the TC, the fuel shall be returned to the spent fuel pool and visually inspected. If the spent fuel meets the requirements for storage in the ISFSI, the fuel may be subsequently transferred to the ISFSI. The DSC shall be removed from service and evaluated-for further use or disposed of, as may be appropriate.
2-3
-. - - . - . - . . - - . _ - _ - _ = _ - - - - - . . .- _ ~.
, e 2.0 FUNCTIONAL AND OPERATING LIMITS 2.4 HORIZONTAL STORAGE MODULE (HSM)
SPECIFICATI0B: The contact dose rate on the surface of the HSH access door shall not exceed 100 mrem /hr (1 mSv/hr). The contact dose rate on the surface '
of the HSM sides shall not exceed 20 mrem /he (0.2 mSv/hr).
APPLICABILITY: This specification is applicable to initially loaded HSMs ACTION: If the.above dose rates are exceeded, take immediate action to determine the cause and bring down the dose rate to an acceptable level. If an acceptable level cannot be achieved, the DSC shall be removed from the HSH and returned to the spent tuel pool.
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BASES FOR SECTION 2.0 FUNCTIONAL AND OPERATING LlHITS
- =. - . - - - - . . . . .- . . . ~ . . . . - --
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i 2.1 LuEt TO BE STORED _ AT ISFSI BASES The design criter , and subsequent safety analysis of the Calvert Cliffs ISFSI assumed certain characteristics and limitations for the fuel assemblies that are to be stored. Specification 2.1 ensures that these bases remain valid by 4 defining the source of the spent fuel, maximum initial enrichment, irradiation history, maximum thermal heat generation, and minimum post-irradiation cooling time.
The radiological analyses are based on a radiation spectrum for 3.4 weight percent U-235 fuel at 42,000 MWD /HTU burnup. Compliance with the enrichment burnup limits and cooling time will ensure that the design criteria are not
-exceeded.
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, o 2.2 DRY SHIELDED CANISTER (DSC)
BASES 2.2.1 DSC Vacuum Pressure Durina Drvina A stable vacuum pressure of less than three torr (3 mm Hg) indicates that all liquid water has evaporated in the DSC cavity, and that the resulting inventory of oxidizing gases in the DSC is less than 0.25% (Vol.%).
2.2.2 DSC Helium Backfill Pressure The thermal analysis performed for the DSC assumes the use of helium as a cover gas. Also, the use of an inert gas (heilum) ensures long-term maintenance of fuel clad in'.egrity.
The value of 2.5 psig was selected to ensure that the pressure within the DSC is within the pressure design limits.
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o o.
l 2.3 TRANSFER CASK (TC)
BASES ,
The drop analyses performed for cask drop incidents for a DSC loaded in a TC confirm that drops up to 80 inches can be sustained without unacceptable damage to the cask and DSC. This limiting condition ensures that the handling i.eight limits will not be exceeded at the storage pad or in transit to and from the spent fuel pool. Design of the OSC is to ASME B&PV Code Section III, Division 1, Subsection NB for Class I components, Service level D requirements. Based on engineering judgment, drops through a height of 15 inches (0.38m) are not judged to be a concern.
B 2-3
4 e 2.4 HORIZONTAL STORAGE MODULE (HSM)
BASES The dose rates stated in this specification were selected to maintain as-low-as-is-reasonably-achievable exposure to the general public and to onsite personnel inspecting and/or maintaining the air vent openings on the HSMs.
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SECT:0N 3/4.0 .
Lfi41 TING CONDITIONS / SURVEILLANCE REQUIREMENTS i
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3/4.1 FUEL TO BE STORED AT ISFSI LIMITIllG CONDITION FOR OPERATION 3.1.1 The spent nuclear fuel to be received and stored at the Calvert Cliffs ISFSI shall meet the following requirements:
(1) Only fuel irradiated at the Calvert Cliffs Units 1 or 2 may be used. (14 x 14 CE type PWR Fuel)
(2) Maximum initial enrichment shall not exceed 4.5 weight percent U-235.
(3) Maximum assembly average burnup shall not exceed 47,000 megawatt-days per metric ton uranium.
(4) Minimum t,urnup shall exceed the minimum specified in SAR Figure 3.3-1.
(5) Maximum heat generation rate shall not exceed 0.66 kilowatt per fuel assembly. *(This requirement is met, if Requirements 2, 3 and 6 are met.)
(6) Fuel shall have cooled a minimum of ten years after reactor discharge prior to storage in the Calvert Cliffs ISFSI, or as specified in ISFSI SAR Table 10.3-1, (7) Maximum assembly mass including control components shall not exceed 1300 lb.(590 kg).
(8) Fuel shall be intact unconsolidated fuel.
(9) Fuel assemblies known or suspected to have structural defects (other then pinhole leaks) sufficiently severe to adversely affect fuel handling and transfer capability shall not be loaded into the DSC for storage.
APPLICABillTY: This specification is applicable to all spent fuel to be stored in Calvert Cliffs ISFSI.
ACTION: If any fuel does not specifically meet the requirements for maximum burnup and post irradiation time (items 3 & 6 above), confirm to see if the requirements of Section 2.1 are satisfied. If any other requirements of the above specification are not satisfied, do not load the fuel assembly into a DSC for storage.
This number may be analytically determined.
3/4 1-1
, e 3/4.1 FUEL TO BE STORED AT ISFSI SURVEILLANCE REQUIREMENTS 4.1.1 Prior to insertion of a spent fuel assembly into a DSC, the identity of the assembly shall be independently verified by two individuals and shall-be documented.
r 4.1.2 Each spent fuel assembly to be loaded into a DSC shall have the parameters listed in Section 3.1 independently verified by two individuals and docuaented.
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$ V 3/4.2 DRY SHIELDED CANISTER (DSC) 3/4.2.1 DISSOLVED BORON CONCENTRATION LJMITING CONDITION FOR OPERATION 3.2.1.1 The DSC cavity shall be moderated only by water with a boron concentration greater than or equal to 1,800 ppm.
APPLICABILITY: Applicable to all DSCs.
ACTION: ,
- 1. With the measured boron concentration less than the specification prior to the beginning of DSC loading and unloading operations, suspend all activities involving DSC loading and unloading. _!
- 2. With the measured boron concentration less than the specification during DSC loading and unloading operations, take action to increase boron concentration while unloading fuel from the DSC.
SURVEllLANCE RE0VIREMENTS 4.2.1.1 Within one hour pricr to insertion of the first spent fuel assembly into a DSC, the dissolved boron concentration in water in the spent fuel pool and introduced into the DSC cavity shall be independently determined by chemical analysis (two samples analyzed by two different individuals). The boron concentration in-the water shall be reconfirmed at intervals not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> until such time as the DSC is removed from the spent fuel pool. All bcron concentration measurement shall be documented.
4.2.1.2 Within one hour prior to flooding the DSC cavity for unloading the fuel assemblies, the dissolved boron concentration in water in the spent fuel pool and introduced into the DSC cavity shall be independently determined.by chemical analysis (two samples analyzed by two different individuals). The boron concentration in the water shall be reconfirmed-at intervals not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> until such time as the fuel has been removed from the DSC.
All boron concentration measurements shall be documented.
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l 3/4.2 DRY SHIELDED CANISTER (DSC) ;
llMITING CONDITION FOR OPERATION 3.2.2.1 The top shield plug closure weld, the siphon and vent port cover welds, and the top cover plate weld shall satisfy the liquid Penetrant Acceptance Standards of ASME B&PV Code Section III, Division 1, Subsection NB-5350 (1983).
3.2.2.2 The standard helium leak rate for the top shield plu and the siphon and vent port cover welds shall not exceed 10'g closure weld, atm-cc/s. .
APPLICABILITY: Applicable to all DSCs.
ACTION: With the requirements of the above specifications not satisfied, the weld shall be repaired in accordance with approved procedures and re-examined-in accordance with these specifications.
SVRVEILLANCE RE0VIREMENTS 4.2.2.1 During DSC loading operations, the top shield plug closure and the siphon and vent port cover welds shall be tested using a helium leak detector to ensure that, for each weld, leak tightness is less than or equal to 10a tm-cc/ s . These welds and the DSC top cover plate weld shall be dye penetrant tested.
i
.1 3/4 2-2
e e 3/4.2.3 DSL EXTERIOR SURFACE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.2.3.1 Rpmovable contamination on the DSC exterior shall be less than 22,000 dpm/100 cm (3j7 Bq/cm2) from beta and gamma sources, and 2,200 dpm/100 cm (0.37 Bq/cm ) from alpha sources.
APPLICABILJJJ: Applicable to all DSCs.
ACTION: With the DSC contamination level in excess of the above specification, the DSC shall not be inserted in the HSM. If already inserted, it shall be removed from the HSM and both the DSC and the HSH shall be decontaminated to meet the specification before the DSC is re-inserted in the HSM.
SURVEILLANCE RE0VIREMENTS 4.2.3.1 Prior to inserting the DSC into the Transfer Cask (TC), the TC interior shall be smeared to ensure that removable contamination levels on the interiorsurfacesoftpecask,excludingthedrainand-ventlines,areless 2
than 22,00p dpm/100 cm (dpm/100 0.37 Bq/cm cm 2(3.7
) frcm Bq/cm alpha ) from sources. beta and gamma sources and and 2,200 4.2.3.2 After fuel loading, but prior to moving the loaded DSC and TC to the HSH, the top of the sealed DSC, the top six inches of the DSC sides, and the TC exterior surfaces shall be smeared }o ensure that 2 removable contamination levelsarelessthan22,000 ppm /100cm sources and 2,200 dpm/100 cm (0.37 Bq/cml) from alpha sources.( .7 Bq/cm ) from beta 4.2.3.3 After TC unloading, the interior surfaces of the cask shall be smeared to ensure that removable contamination levels on the interior surfaces of cm 2the
( .7cask, Bq/cmop)cluding the drain and vent lines, are less than 23,000 dpm/100 Bq/cm{) from alpha sources.from beta and gamma sources and 2,200 dpm/100 cm (0 3/4 2-3
e o 3/4.3 TRANSFER CASK (TC) 3/4.3.1 AMBIENT TEMPERATURE LIMITING CONDITION FOR OPERATION 3.3.1.1 Fuel transfer operations to and from the ISFSI, in the transfer cask, shall not take p. ace when daylight ambient temperatures exceed 103*F(39.4*C).
APPLICABILITY: This specification is applicable to all outdoor spent fuel transfer operations.
ACTION: If the daylight ambient temperature is expected to exceed of the above specification, do not commence a fuel transfer operation from the Auxiliary Building; if fuel transfer operation from the HSH is required because of another specification, provide shading. If the daylight ambient temperature exceeds 100*F during transfer operation, provide shading.
SURVEILLANCE RE0VIREMENT 4.3.1.1 When temperatures are expected to approach 100*F (37.8*C) or more, the outdoor ambient temperature in full sunlight shall be measured and recorded within one-half hour prior to movement of a TC loaded with fuel, to or from the ISFSI. Additionally, the outdoor ambient temperature in full sunlight shall be measured and recorded once per hour when the loaded transfer cask is outside the Auxiliary Building.
3/4 3-1
. = 1 3/4.4 HORIZONTAL STORAGE MODULE (HSM) 3/4.4.1 MAXIMUM AIR TEMPERATURE RISE LlHITING CONDITION FOR OPERATION 3,4.1.1 The air temperature rise from the HSH inlet to the HSH outlets shall not exceed 60*F (33.3'C).
APPLICABILITY: Applicable to all HSMs.
ACTION: If the temperature rise is greater than 60*F, (33.3*C) the air inlet and outlets should be checked for blockage. If any blockage is cleared and the temperature rise is still greater than 60'F, (33.3*C) the DSC amd HSH cavity shall be inspected, using video equipment or other suitable means.
Analysis of the existing conditions shall be performed to confirm that conditions adversely affecting the fuel cladding integrity do not exist.
Subsequent actions to return to acceptable conditions such as, providing temporary forced ventilation and/or retrieval of the DSC and verification that an assembly fuel with no more than 0.66 kW was loaded shall be performed.
SURVIELLANCE REQUIREMENTS 4.4.1.1. The maximum temperature rise from the HSH inlet to ' outlets shall be checked at the time t~ne DSC is stored in the HSM, again 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later, and again after 7 days.
4.4.1.2 The HSH shall be visually inspected to verify that the air inlet and outlets are free from obstructions when there is fuel in the HSM. The visual inspection frequency _shall be every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3/4 4 I
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le e 3/4.5 FIRE PROTECTION LlHITING CONDITION FOR OPERATION 3.5.1 Onlydiefelpoweredvehicleswithfueltankcapacitynotmorethan100 gallons (0.38 m ) shall be permitted within the ISFSI fenced area. When such vehicles are present, portable fire suppression equipment shall be present in the ISFSI fenced area.
3.5.2 During spent fuel transfer operations to and from the Auxiliary Building and the ISFSI fenced area, there shall be no fossil fuel tanker truck movements or fossil fuel transfer operations to or from such trucks on the >
Calvert Cliffs Nuclear Plant Site inside the plant site entrances. Any such tanker trucks shall not be located within 100 meters of the ISFSI fenced area and transfer route during spent fuel transfer operations.
APPLICABILITY: Specification 3.5.1 is applicable whenever there is spent fuel in the ISFSI or during all spent fuel transfer operations between the Auxiliary Building and the ISFSI. Specific 1 tion 3.5.2 is applicable only during spent fuel transfer operations.
ACTION: With the requirements of the above specifications not satisfied, do not remove the loaded Transfer Cask (TC) from the Auxiliary Building or do not remove a DSC from the HSH.
SURVEILLANCE REQUIREMENT 4.5.1 Prior to removal of the loaded TC from the Auxiiiary Building or a DSC from the HSH, a visual inspection of the transfer route shall be made to ensure compliance with the above specifications.
3/4 5-1
.m.
BASES FOR SECTION 3/4.0 LIMI11NG CONDITIONS / ~JRVEILLANCE REQUIREMENTS
.o e 3/4.1 FUEL TO BE STORED AT ISFSI BASES This specification was derived to ensure that the peak fuel rod temperatures, HSM surface contact dose rates, reactivity, and fuel mass are below the design values.
B 3/4-1
e e 5/4.2 -DRY SHIELDED CANISTER (DSC)'
BASES 3/4.2.1 Dissolved Boron Concentratio2 This specification ensures subcriticality during fuel loading and unloading.
3/4.2.2 DSC Closure Welds The safety analysis of leak tightness of the DSC as discussed is based on a weld being leak tight to 10atm-cc/s. This check is done to ensure compliance with this design criterion.
3/4.2.3 DSC Exterior Surface Contamination Compliance with this limit ensures that the offsite dose limits in 10 CFR.
Part 20, 10 CFR Part 50 - Appendix 1, 10 CFR Part 72, and 40 CFR 190 are met.
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. o- e 3/4.3 IRfNSFER CASK (TC)
LAlf1 3/4.3.1.-Ambient Temperature Thermal analysis of the TC with design' basis fuel and solar heat load, -
indicate that at ambient temperatures above the design maximum of 103'F -
(39.4*C), the temperature of the TC neutron shield material could exceed its-design temperature limit.
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3/4.4 HORIZONTAL STORAGE MODULE (HSM)
BASES 3/4.4.1 Maximum Temoerature Rise s The 60*F (33.3*C) temperature rise was selected to limit the hottest rod in the DSC to below 635'F (336*C) at 70*F (21*C) ambient air inlet temperature.
The expected temperature rise is less than 60*F (i.e., 49'F [27.2*C]; see-Table 8.1-2 of the NUHOMS-24P, NUH-002, Rev 2A topical report) and hence, the current design provides adequate margin for this specification. If the temperature rise is within the specifications, then the HSM and DSC are performing as designed and no further temperature measurements are required during normal surveillance.
The visual inspection interval is selected to ensure that no HSM air inlet or outlets are plugged for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and to ensure that blockaae of inlet and outlets due to an accident will be removed in less than 48 i,ours.
Analysis in Chapter 8 of the Calvert Cliffs ISFSI SAR showed that no temperature limits are exceeded if a module is completely plugged for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Therefore, for normal operations, an inspection of the inlet and outlets once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> will ensure that any local obstructions are identified.
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3/4.5 FIRE PROTECTION BASES I
The fire safety analyses assume characteristics of a fire caused by diesel i fuel limited to the amount carried by security and maintenance vehicles and I
the prime mover and crane used for transfer operations. The consequences of a ;
fossil fuel carrying tanker truck induced fire or explosion have not been analyzed for the Transfer Cask (TC) loaded with spent fuel. A gasoline and-diesel fuel refueling depot for vehicles is situated close to the transfer route; and therefore, periodic storage tank replenishment from a tanker truck is anticipated. The objective is to preclude an accident involving fire or explosion near the'TC due to large amounts of fossil fuels.
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a e SECTION 5.0 DESIGN FEATURES p
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A &
5.0 DESIGN FEATURES The Calvert Cliffs ISFSI design approval was based upon review of specific design drawings, some of which have been deemed appropriate for inclusion in the Calvert Cliffs ISFSI Safety Evaluation Report (SER). Drawings listed in Section 1.5 of the Calvert Cliffs ISFSI SER have been reviewed and approved by the NRC. These drawings may be revised under the provisions of 10 CFR 72.48 as appropriate.
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A O 6.0 ADMINISTRATIVE CONTROLS 6.1 FENERAL The Calvert Cliffs ISFSI is located on the Calvert Cliffs Nuclear Power Plant site and will be managed and operated by the Baltimore Gas.and Electric Company staff. The administrative controls shall be in accordance with the requirements of the Calvert Cliffs Nuclear Power Plant Facility Operating Licenses (DPR-53, and -69) and associated Technical Specifications as appropriate.
6.2 ENVIRONMENTAL MONITORING PROGRAM The licensee shall include the Calvert Cliffs ISFSI in the environmental monitoring for Calvert Cliffs Nuclear Power Plant. An environmental monitoring program is required pursuant to 10 CFR 72.44(d)(2).
6.3 ANNUAL ENVIRONMENTAL REPORT The semi-annual radioactive effluent release reports under 10 CFR 50.36(a)(2) license requirements for the Calvert Cliffs Nuclear power Plant shall also specify the quantity, if any, of each of the principal radionuclides released to the enviroment in liquid in gaseous effluents during the ISFSI operation and such other information as may be required by the Commission to estimate maximum potential radiation dose commitment to the public resulting from effluent release. Copies of these reposts shall be submitted to the NRC Region I office and to the Director, Office of Nuclear Material Safety and Safeguards. The report under this specification is required pursuant to 10 CFR 72.44(d)(3).
6-1
A o SECTION 6.0 ADMINISTRATIVE CONTROLS
6- .-
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SAFETY EVALUATION REPORT FOR THE BALTIMORE GAS AND ELECTRIC COMPANY'S SAFETY ANALYSIS REPORT FOR AN INDEPENDENT SPENT FUEL STORAGE INSTALLATION AT CALVERT CLIFFS
.i September 1992 Enclosure 2
l-A :O TABLE OF CONTENTS-Section Pace
1.0 INTRODUCTION
........................... 1-1 1.1 Background ......................... 1-1 1.2 Discussion of Review Scope and Objectives . . . . . . . . . . 1-2 1.3 Context .......................... 1-3 2.4 Approach .......................... 1-5 1.5 General Description of.NUH0MS-24P System at Calvert Cliffs . . . . . . . . ......... . ... 1-7 1.5.1 Horizontal Storage Module . ............ 1-8 1.5.2 Dry Shielded Canister ............... 1-9 1.5.3 Transfer Cask .......... . . . . . . . 1-10 1.5.4 Fuel Transfer Equipment . . . . . . . . . . . . . 1-12 1.5.5 Fuel Handling & Storage Operations . . . . . . . . 1-16 2.0 SAFETY EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 System Design Criteria and Suitability at Calvert Cliffs ....................... 2-1 2.1.1 Spent Fuel Elements to Be Stored at Calvert Cliffs ................. 2-1 2.1.1.1 Criteria for Nuclear Criticality Safety ....... .... 2-2 2.1.1.2 Criteria for Radiological Protection ................ 2-3 2.1.2 Criteria for Spent Fuel Handling and Transportation ................. 2-3 2.1.3 Design Criteria for Confinement Barriers and Systems ................ 2-6 2.1.4 Criteria for Fuel Stability ............ ~2-7 2.1.5 Protection Against' Environmental Conditions and Natural Phenomena .......... 2-7 2.1.5.1 Normal Operating Conditions .... .. 2-8 2.1.5.2 Off-Normal Operating-Conditions . . . . . . . . . . . . . . . . 2-9 2.1.5.3 Accident Conditions ........... 2-9 2.1.6 Protection Against Fire and Explosion ....... 2-9 2.1.7 Instrumentation and Control Systems . . . . . . . 2-10 2.1.8 Criteria for Decommissioning ..... ... . . 2-12 2.1.9 Findings and Conclusions ........ . . . . 2-14 11
& o TABLE OF CONTENTS (continued)
Section Pace 2.2 -NUHOMS-24P System Design at Calvert Cliffs . . . . . . . . 2-16 2.2.1 Acceptability of NUH0MS-24P at the Calvert Cliffs Nuclear Power Plant . . . . . . . . .-2-16 2.2.1.1 Natural Phenomena .. ..... . . . . 2-19 2.1.1.2 Site Work . . . . . . . . . . . . . . . 2-22 2.2.2 Criticality . Evaluation . . . . . . . . . . . . . . 2-22 2.2.2.1 Description of Review ... . . . . . . 2-23 2.2.2.2 Discussion of Results ..... . . . . 2-24 2.2.2.3 Findings and Conclusions . . . . . . . . 2-25 2.2.3 Structures and Equipment Important to Safety . . . . . . . . . . . . . . . . . . . . 2-26 2.2.3.1 Horizontal Storage Module . . . . . . . 2-31 2.2.3.2 Dry Shielded Canister and Internals . . . . . . . . . . . . . . . 2-36 2.2.3.3 Transfer Cask . . . . . . . . . . . . . 2-50 2.2.4 Evaluation of Handling and Transfer Equipment Not important to Safety ...... . . 2-57 2.2.4.1 TC Lifting Yoke System . . . . . . . . 2-58 2.2.4.2 Transfer Components ........ . . 2-58 2.2.4.3 Vacuum Drying' System . . . . . . . . . . 2-59 2.2.4.4 Automatic Welding Systems . . . . . . . 2-59 2.2.5 Fuel Stability . . . . . . . . . . . . . . . . . 2-60 2.2.5.1 Description of Review .. . . . . . . . 2-60 2.2.5.2 Discussion of-Results. ... . . . . . . 2-60 2.2.5.3 Findings and Conclusions ...... . . 2-62 2.2.6 Thermal Evaluation . . . . . . . . . . . . . . . . . 2 2.2.6.1 Description of Review ..... . . . . 2-62 2.2.6.1.1 Applicable Parts of 10 CFR Part 72 . . . . . . . . . . . 2-62~
2.2.6.1.2 Review Procedure . . . . . . 2-63 2.2.6.2 Discussion of Results ...... . . . 2-65 2.2.6.3 Findings and Conclusions ..... . . . 2-67 iii
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i 9 J ,
i TABLE OF CONTENTS (continued)
Section Pace 2.2.7 Shielding Evaluation . . . . . . . . . . . . . . 2-68 2.2.7.1 Description of Review (Source Specification and Analyses) . . 2-68 2.2.7.2 Discussion and Results . . . . . . . . . 2-69 2.2.7.3 Findings and Conclusions . . . . . . . . 2-71 2.2.8 Radiological Protection Evaluation . . . . . . . . 2-71 2.2.8.1 Description of Review . . . . . . . . . 2-71 2.2.8.2 Discussioa of Results . . . . . . . . . 2-74 2.2.8.3 Findings and Conclusions . . . . . . . . 2-76 2.2.9 Infrastructure . . . . . . . . . . . . . . . . . . 2-76 2.2.9.1 Organization . . . . . . . . . . . . . . 2-77 2.2.9.2 Training . . . . . . . . . . . . . . . . 2-78 2.2.9.3 Procedures . . . . . . . . . . . . . . . 2-78 2.2.10 Surveillance and Monitoring . . . . . . . . . . . . 2-79 2.2.10.1 Description of Review . . . . . . . . . 2-79 2.2.10.2 Applicable Parts of 10 CFR Part 72 . . . 2-79 2.2.10.3 Review Procedure .......... . 2-80 i 2.2.10.4 Design Description . . . . . . . . . . 2-80 2.2.10.5 Discussion of Results . . . . . . . . . 2-81 2.2.10.6 Findings and Conclusions . . . . . . . . 2-82 3.0 CONDUCT OF OPERATIONS . . . . . . . . . . . . . . . . . ...... 3-1 3.1 Procedures ......................... 3-2 3.1.1 Administrative Procedures ............. 3-2 3.1.2 Health Physics Procedures ............ 3-2 3.1.3 Maintenance Procedures ............... 3-3 3.1.4 Operating Procedures ................ 3-5 3.1.5 Test Procedures .................. 3-6 3.1.6 Pre-Operational Test Procedures . . . . . . . . . . 3-7 3.2 Records . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-7 3.3 Training and Certification ...,............. 3-8 3.4 Physical Protection . . . . . . . . . . . . . . . . . . . . . 3-9 3.5 Emergency Planning ..................... 3-9 4.0 QUALITY ASSURANCE . . . . . . . . . . . . . . . . . . . . . . . . 4-1 5.0 OPERATING CONTROLS AND LIMITS . . . . . . . . . . . . . . . . . . . 5-1 iv i
. e TABLE OF CONTENTS (continued)
Section Paae 6.0 DECOMMISSIONING . . . . . . . ........... . . . . . 6-1
7.0 CONCLUSION
S . . . ................ . . . . . . . . 7-1
8.0 REFERENCES
...................... . . . . . 8-1 Appendix A - Chronology of Principal Actions . . . . . . . . . . . . . . A-1 um V
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LIST OF FIGURES fiquy Pace 1.1 Horizontal Storage Module Concept Illustration . . ... . . . . . 1-18 1.2 Dry Shielded Canister Concept Illustration . .. . .. . . . . . 1-19 1.3 Transfer Cask Concept Illustration . . .. .. . ... . . . . . . 1-20 1.4 Transfer Cask with Dry Shielded Canister . . . . . . . . . . . . 1-21 1.5 System for Tranferring DSC Between TC and HSH . . .. . . . . . . 1-22 1.6 Hydraulic Ram System Concept Illustration .. . . .. . . .. . . 1-23 l
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o o LIST OF TABLES Table Pane 1.1 Incorporation of Docketed Material by Reference . . . . . . . . . . . . . . . . . . . . . . . 1-24 1.2 Principal Differences Between NUHOMS-24P Systems as Defined in the TR and the CC SAR . . . . . . . . . . . . . 1-34 2.1.5-1 Evaluation of Design Criteria For Normal Operating Conditions . . . . . . . . . . . . . . . . . . . 2-83 2.1.5-2 Evaluation of Design Criteria For Off-Normal Operating Conoitions . . . . . . . . . . . . . . . . . . . 2-86 2.1.5-3 Evaluation of Design Criteria For Accident Conditions . . . . . . . . . . . . . . . . . . . . . . . . 2-88 2.2.3-1 Load Combinations Used For HSM Reinforced Concrete .. . 2-92 2.2.3-2 Load Combinations Used For DSC Support Assembly . . . . . . 2-94 2.2.3-3 Summary of HSM Design Analysis ........ . . . . . . 2-96 2.2.3-4 Summary of DSC Support Assembly Design Analysis . . . . . . 2-97 2.2.3-5 HSM Load Combinations Results . . . . . . . . . . . . . . . 2-98 2.2.3-6 DSC Support Assembly Rail Load Combination Results (a) . . . . . . . . . . . . . . . . . . . . . . . . 2-100 2.2.3-7 DSC Support Assembly Transverse Member Load Combination Results . . . . . . . . . . . . . . . . . . . . 2-101 2.2.3-8 DSC Stress Analysis Results For Normal Loads Service Level A . . . . . . . . . . . . . ........ 2-102 2.2.3-9 DSC Stress Analysis Results For Off-Normal Loads Se rv i ce Level B . . . . . . . . . . . . . . . . . . . . . . 2-103 2.2.3-10 DSC Load Combinations For Normal and Off-Normal Operating Conditions - Service levels A and B . . . . . . . 2-104 2.2.3-11 DSC Stress Analysis Results For Accident Conditions - Service Level C ......... ...... 2-105 2.2.3-12 DSC Load Combinations For Accident Service Level C' Cases . . . . . . . . . . . . . . . . . . . 2-106 2.2.3-13 DSC Drop Accident Loads Service Level D . . . . . . . . . . . . . . . . . . . . . . 2-107 viii
LIST OF TABLES (continued)
Table Pace 2.2.3-14 DSC Enveloping Load Combination Results For Accident Loads - Service Level D ........... 2-108 2.2.3-15 Transfer Cask Stress Analysis For Normal Loads - Service Levels A and 8 Allowables . . . , . . . . . 2-109 2.2.3-16 Transfer Cask Load Combinations For Normal Operating Conditions - Service Levels A and B . . . . . . . 2-110 2.2.3-17 Transfer Cask Stress - Analysis Results For Accident Loads - Service Level C Allowables . . . . . . . . 2-111 2.2.3-1B Transfer Cask Drop Accident Loads Service Level D Allowables ................ 2-112-2.2.3-19 Transfer Cask Stress Results For Tornado Driven Missile Impact . .................. 2-113 2.2.3-20 Transfer Cask Load Combinations For Accident Conditions - Service Level D . . . . . . . . . . . 2-114-2.2.3-21 Summary of Stress Analyses For Upper Lifting Trunnions and Lower Resting, Weld Regions and Cask Shell . 2-115 2.2.4-1 Summary of Design Criteria and Parameters of Handling and Transfer Equipment Not important to Safety ...... 2-116 2.2.7-1 Comparison of Licensee Calculated Dose Rates to Audit Results . . . . . . . . . . . . . . . . . . . . . . . . . . 2-118 2.2.7-2 Comparison of Applicant and Audit Review Calculation Results for Total Dose Rate At A Distance From The Planned HSM Array . . . . . . . . . . . . ......... 2-119 ix
o e GLOSSARY BG&E Baltimore Gas and Electric Company BRC Below Regulatory Concern CFR Code of Federal Regulations DBE Design Basis Earthquake DBT Design Basis Tornado DSC Dry Shielded Canister ER Calvert Cliffs ISFSI Environmental Report, Reference 2 ERP Emergency Response Plan FSAR Final Safety Analysis Report (for Calvert Cliffs unless otherwise stated)
HRS Hydraulic Ram System HSM Horizontal Storage Module IFA Irriadiated Fuel Assembliee ISFSI Independent Spent Fuel Storage Installation LNG Liquified Natural Gas LC0 Limiting Condition of Operation NRC U.S. Nuclear Regulatory Commission NUTECH NUTECH Engineers, Inc. (See PNFSI)
Oconee Duke Power Company Oconee Nuclear Power Plant PNFSI Pacific Nuclear Fuel Services, Inc. (formerly NUTECH)
RC Reinforced Concrete RG Nft Regulatory Guide SA, Safety Analysis Report SER NRC Safety Evaluation Report SPS Skid Positioning System TC Transfer Cask TR Topical Report, Reference 3, unless otherwise stated TSAR Topical Safety Analysis Report (also, Topical Report) x
1.0 INTRODUCTION
1.1 Background
Baltimore Gas and Electric Company (BG&E) operates the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, on the west shore of the Chesapeake Bay in Calvert j County, Maryland, in the vicinity of Prince Frederick, Maryland. Calvert i Cliffs has been operating since May 8, 1975, and has discharged more than 1350 spent fuel assemblies. These are currently stored in the fuel pool. However, 1 additional onsite storage facilities are required to permit continued -l operation beyond 1995. BG&E evaluated alternatives to the use of a dry storage facility, in the design of a dry storage facility, and to the siting of an independent spent fuel storage installation (ISFSI). Results of the evaluations led BG&E to apply for a license to construct and operate an ISFSI at Calvert Cliffs. The Safety Analysis Report ("the SAR," Reference 1) describes the proposed facility. The Environmental Report (ER, Reference 2) describes analyses supporting the choice of sites and the suitability of the proposed facility.
The proposed ISFSI is based on a system that was the subject of a topical report (TR, Reference 3) and corresponding NRC Safety Evaluation Report (SER, Reference 4). The generic NUTECH Engineers, Inc. NUH0MS-24P system proposed for use has also been used by Duke Power Company for an ISFSI at the Oconee Nuclear Station. This was proposed by SAR (Reference 5). The evaluation was reported by NRC SER (Reference 6).
The NUHOMS-24P system is designed to have 24 pressurized water reactor irradiated fuel assemblies (IFAs) stored in dry shielded canisters (DSC) kept in reinforced concrete (RC) horizontal storage modules (HSM). The system-is very similar to the NUHOMS-7P system, except that the 7P provides for the storage of 7 pressurized water reactor IFAs per DSC, kept in HSMs. The NUHOMS-7P has been the subject of a TR (Reference 7) and corresponding SER (Reference 8). The NUHOMS-7P has been instal' led and is in use at the Carolina Power and Light Company (CP&L) H.B. Robinson 2 (HBR2) Steam Electric Plant.
The HBR2 installation was the subject of a SAR (Reference 9) and corresponding 1-1 t
l
. e SER (Reference 10). CP&L has also proposed another NUHOMS-7P installation, at the Brunswick Steam Electric Plant (Reference 11).
1.2 Discussion of Review Scope and Objectives BG&E has submitted an application to the NRC for a license to construct and operate an onsite ISFSI at the Calvert Cliffs site (Reference 12). This facility would consist mainly of an approximately 3% acre fenced area containing up to 120 NUTECH Horizontal Modular Storage system modules of the 24P type, holding 24 pressurized water reactor IFAs per module. This would provide capacity for storage of up to 2880 IFAs.
The NRC staff has reviewed this license application, evaluating the safety of the proposed operation and the qualifications of the applicant to protect the public health and safety and to meet the requirements of the reoulations for independent spent fuel storage installations (10 CFR Part 72, Reference 13).
In its review of this application, the NRC staff did not evaluate all aspects of the generic NUHOMS-24P system (Reference 3). The staff had previously evaluated this system and approved it for use subject to specific operating controls and limits (Reference 4). This review also did not reevaluate existing licensed reactor or pool storage facilities or activities at Calvert Cliffs.
The review focused on the safety of the proposed on-site dry storage operations, the transportation of dry shielded canisters (DSC) . rom the fuel building to and from the storage area, the insertion of the DSC into the horizontal storage module (HSM) and its subsequent removal for decommissioning, and the applicant's management system for carrying out these onerations in a safe manner.
The objectives of this SER are to document the NRC staff's 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 i requirements of 10 CFR Part 72.
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l.3 Context This Safety Evaluation Report provides NRC staff analyses and recommendations on the SAR (Reference 1) submitted by BG&E in conjunction with an application for a specific license to construct and operate an ISFSI on the Calvert Cliffs site (Reference 12).
The SAR was submitted in accordance with the requirements of 10 CFR Part 72 (Reference 13). This SER is based on review for compliance with 10 C -
Part 72. Changes, clat ifications, and additional information submitted to the NRC subsequent to the SAR during the review process (as listed at Reference 1) are considered to have the full effect and to express the same commitment as the SAR itself.
The BG&E Calvert Cliffs SAR incorporates some material from the Topical Report (TR) on the generic NUH0HS-24P system prepared by NUTECH Engineers, Inc.
(Reference 3) by reference. Reference to the TR is considered to include changes, clarifications, and additional information to the TR submitted to the NRC. A separate SER was prepared by the NRC staff on the NUTECH TR (Reference 4).
The SAR incorporates and refers to information contained in the BG&E Final Safety Analysis Report (FSAR) for Calvert Cliffs (Reference 14). Since the proposed ISFSI is to be on the Calvert Cliffs site, much of the site environment and surrounding area information is identical.
Table 1.1 identifies the references to docketed material made in the SAR and the limits of material incorporated. Information incorporated by references (as provided by 10 CFR 72.18) or included by subsequent submittal (Reference 1) is considered identically with information contained in the SAR. Where such information is already the subject of NRC approval, as by approval of the reconrnendations 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 with any qualifiers included in the referenced document and/or the corresponding SER.
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Some of the technical information on the proposed ISFSI in the BG&E SAR is included in References 1, 3, and 14. In accordance with the intent of separate submittal and approval of a TR, this SER does not repeat the contents of Reference 1, but incorporates portions of the SAR by reference where appropriate.
Use of the proposed ISFSI includes operations and provision and use of special equipment within the existing Calvert Cliffs Auxiliary Building structure housing the fuel pool. These include operations and equipment related to safety. The fuel handling operations require limited amendments to BG&E's existing technical specifications under 10 CFR Part 50 (Reference 15) in order to operate the ISFSI. This SER does not constitute the formal safety evaluation review for the safety of operations and equipment within the fuel storage facility of the Auxiliary Building. This SER does, however, examine the suitability of the transfer cask and DSC for mutual compatibility, and satisfaction of 10 CFR Part 72 requirements.
The SAR and the TR follow the outline for ISFSI SAR recommended by Regulatory Guide 3.48 (Reference 16). Portions of the TR are incorporated by specific references within the SAR, which are considered to adequately comply with the referencing requirements. The SAR has been reviewed and this SER has been prepared considering that only those portions of the TR are included in the SAR as are specifically referenced. Where the CAR contradicts or effectively repeats sections of the TR but with modifications, the SAR text is considered to supersede the TR text, and corresponding comments of the SER (Reference 4) on the TR (Reference 3) are considered to be moot. The BG&E Calvert Cliffs updated Final Safety Analysis Report (FSAR) (Reference 14) is also incorporated by specific references.
The NUHOMS-24P system has been installed and is in use by Duke Power Company at the Oconee Nuclear Power Station. Although the SAR ' Reference 5) and SER (Reference 6) for the Oconee installation are not referenced in the Calvert Cliffs SAR, the staff considered the results of the Oconee evaluation effort in the review of the Calvert Cliffs SAR. The Oconee and Calvert Cliffs ISFSI ;
were both designed by the same engineering team (NUTECH Engineers, Inc. and 1-4
.q ~ o' the Duke Engineering and Services, Inc., subsidiary of Duke Power Company).
The NUHOMS-24P designs used for Oconee and Calvert Cliffs are not identical to the design of the original TR. The principal differences are identified in Table 1.2. The staff considered the evaluation of the Oconee design in areas wher it and the Calvert Cliffs designs are the same but are modifications of the TR design.
Staff evaluation (Reference 8) of the TR (Reference 7) and reviews (Reference
- 10) of existing installations at H.B. Robinson (Reference 9) and that proposed for the Brunswick (Reference 11) Steam Electric Plants, which are similar NUHOMS-7P systems, were considered in the preparation of this SAR. In addition, the results of heat transfer and shielding performance testing and evaluation of the H.B. Robinson NUHOMS-7P ISFSI, reported in Reference 17, have been considered. The H.B. Robinson ISFSI testing may not be directly applicable to the Calvert Cliffs NUHOMS-24P system because the 7P system has only 7 IFAs per DSC. However because the thermal performance and heat shielding calculations of both generic designs were performed using similar methodologies, the test results provided empirical evidence that the methodologies are conservative (more severe temperatures were predicted and used for the design than would actually occur).
1.4 Approach The NRC staff used different approaches for evaluating the safety of the j storage operation and the applicant's qualifications for operating an ISFSI.
The staff approach for evaluating the safety of the storage system involved a i
six-step analytical technique, including:
- 1. Reviewing the site based on the description in the BG&E Calvert Cliffs ISFSI SAR (Reference 1) in order to determine all the credible natural and man-induced phenomena which the ISFSI system would have to withstand. The NRC staff determined whether the applicant identified all these phenomena in its specification for the site-specific design.
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- 2. Determining the design basis severity of-the site phenomena. If-previous-NRC determinations of phenomenon severity were available, they were used; if not, independent determinations were made on the basis of existing-data for the Calvert Cliffs sitel
- 3. Comparing the identification and severity of phenomena'at the Calvert Cliffs site with those for which the generic NUHOMS-24P-storage system had previously been approved. On-the basis of this comparison, the staff determined the acceptability of the NUH0HS-24P system environmental hazard analyses for-the Calvert Cliffs site.
- 4. Reviewing the proposed operations of transporting the transfer cask containing the loaded DSC from the Calvert Cliffs Auxiliary Building to the storage area, inserting the DSC into the HSM, and-retrieving the DSC from the storage areas to the Auxiliary '
Building. The staff assessed the health and the safety _-
implications of these operations and identified areas where license restrictions may be appropriate.
- 5. Reviewing the proposed operations of storage and monitoring and inspecting the storage system, including the on-site and off-site rad lological health and safety implications of these operations,-
and identifying areas where license conditions may be' appropriate.
- 6. Reviewing the nuclear _ criticality safety of the' design and operation of the storage system including specification of the fuel to be stored, and DSC loading.and unloading operational requirements.
In evaluating.the applicant's qualifications, the staff reviewed the applicant's conduct of operations. The review examined BG&E's. procedures, recordkeeping, training, and management systems. The staff also reviewed and evaluated the applicant's quality assurance program and decommissioning plan.-
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-i The staff's assessment of the operation of' storing, monitoring, and inspecting the storage system, including the on-site and off-site radiological health and safety implications of these operations, and identification of areas where license conditions may be appropriate, is presented ;.7 Section 3.0 of this SER.
A verification of the applicant's compliance with'the licensing requirements of 10 CFR Part 72 for independent spent fuel storaga installations is presented in Section 7.0 of this SER.
1.5 General Description of ISFSI-at Calvert Cliffs The following descriptions of the proposed Calvert Cliffs ISFSI are based on the more complete descriptions provided by Reference 1 and are only included for the convenience of readers of the SER. The SER is based on the descriptions provided in the SAR, and not on these summary descriptions. The NUHOMS system components for IFA storage at the ISFSI are the DSC and the HSM.
Additional systems required for the DSC closure and transfer include the transfer cask, the skid and skid positioning system, the trailer, the hydraulic ram system, and the DSC vacuum drying system. More complete descriptions of components are in Reference 1 as follows:
Component SAR Text SAR Figures Drawings Calvert Cliffs Para 2.1, 4.1 1.1-1, 2.1-2 Site ISFSI Site Para 4.1 1.2-1, 2.4-1 84-075-E Rev. 1 HSH Para 1.3.1.2, 4.2.3.1 84-080-E to 84-082-E Rev. I 84-085-E to-84-087-E Rev. 1 84-091-E Rev. 2 84-092-E Rev. 1 84-093-E Rev. 2 84-Oll-E to 84-013-E Rev. 0 1-7
- o Component SAR Text S.'" Figures Drawings DSC Para 1.3.1.1, 4.2.3.2 1.3-1 84-001-E to 84-002-E Rev. 3 84-003-E to 84-007-E Rev. 2 TC Para 1.3.1.3, 4.7.3.3 1.3-2, 4.7-1 84-021-E to 84-030-E Fuel Transfer Para 1.3.1.3-1,3.1.6 4.7-2 Yoke Assembly ;
Equipment Para 4.7.3.4-4.7.3.9 4.7-3, 84-036-E &
4.7-5-4.7-9 84-037-E Rev. 2 Figures 1.1 through 1.5 which are cited below are concept illustrations. The figures do not include proprietary information and may therefore differ in details from the actual design. The figuras may also include material which has been or may be patented.
1.5.1 Horizontal Storage Module The Calvert Cliffs ISFSI will employ HSMs constructed in units of 12, configured in 2x6 arrays. The HSMs will be constructed in place at the ISFSI with pairs of 2x6 arrays placed end to end. Each array _of 12 HSMs will be constructed on a common reinforced concrete foundation slab. The HSM is designed to provide neutron and gamma shielding to achieve a nominal 15 mrem /hr contact dose rate. Nominal contact dose rates at the HSM access door and vents are designed to be less than 100 mrem /hr.
Three-foot-thick exterior walls and roof provide shielding of each HSM array.
The front walls of the HSMs are thickened to three and a half feet at the access opening and vent inlet of each module. Two-foot-thick interior ccmmon walls provide shielding between modules to reduce radiation frca adjacent modules during DSC loading and retrieval. Internal slab and roof caps are provided to shield the ventilation inlet and outlet openings, respectively.
The basic HSM design is shown by Figure 1.1 (extracts of Reference 3),
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The HSH provides structural support for the DSC, protects the DSC against extreme natural hazards such as tornado missiles, and provides radiation shielding. The HSH 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 in the HSH roof slab. A thermal radiation shield is used to reduce the HSH roof temperature to within acceptable limits for all conditions.
The approach slab in front of the HSMs is constructed separately from the HSH foundations. The transfer system is designed to accommodate any credible differential settling between the two slabs. The approach slab and HSM foundation have been designed to minimize differential settlement over the life of the facility.
1.5.2 Dry Shielded Canister The DSC is illustrated in Figures 1.2 and 1.3. A DSC is shown in storage position in Figure 1.1. 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 main component of construction of the DSC is a stainless steel cylindrical confinement vessel.
The internal basket assembly is comprised of 24 guide sleeves suppcrted by spacer disks at intervals corresponding to the fuel assembly spacer grids.
Support rods maintain the 9 spacer disks in location. All canister structural components are fabricated from type 304 stainless steel. Lead gamma shielding
-is used in both the top and bottom end shield plugs.
Criticality safety during wet loading operations is maintained through the geometric separation of the fuel assemblies within the internal basket assembly, the inherent neutron absorption capability of the stainless steel guide sleeves, the proper selection of sufficient' loleted fuel assemblies, 1-9
< 0 l l
and adequate boron concentration in the pool water. However, credit for burn l up is not currently permitted by the NRC staff, l The DSC provides mechanical confinement for the stored fuel assemblies and all 1 radioactive materials for two purposes: to prevent the dispersion of )
particulate or gaseous radionuclides from the fuel, and to mair.cain a barrier i of helium around the fuel in order to mitigate corrosion of the fuel cladding l and prevent oxides from forming in the fuel itself. ;
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 HSH door area dose rates l during storage. The DSC shielding is designed for a maximum contact dass of 100 mrem /hr (flooded capacity).
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. This is accomplished by a combination of surface finisher and dry film lubricant coatings applied to the DSC and the DSC support assembly in the W5n. The transfer operation is illustrated in Figu e 1.4.
1.5.3 Transfer Cask The principal components of the transfer cask (TC) are shown in Figure 1 't (SAR Figure 1.3-2). Figure 1.4 (SAR Figure 4.7-1) shows the TC - 05C.
Figure 1.5 (SAR Figure 4.7-5) shows the TC in position for DSC tr. . fer 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 operat hns.
The structural and outer shells form an ann: Jar pressure vessel. A solid neutron absorbing material is cast between the structural shell and outer shell to provide neutron shielding when the DSC is the TC.
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._ - -_ . - - -. - = . . - -. - _ . - . .
, w .
l The cask bottom end assembly is welded to the cylindrical shell assembly. It includes two closure assemblies for the ram / grapple access penetration. A water tight bolted cover plate, with a core of solid neutron absorbing material, is used for transfer operation within the Auxiliary Building. The bolted ram access penetration 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 by gravity in brackets. At the HSH 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 encapsultting 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 aH as supports during transport to/from the ISFSI. The TC is not designed as a pressure vessel.
The ncutron 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 material 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-condensible gases when heated above 212 8F. 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 95 psig. Pre-set safety relief valves are included to protect the neutron shield cover in the event that its design pressure is exceeded.
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e r 1.5.4 Fuel Transfer Equipment fuel is transferrod in ISFS! operations by means of a transfer cask. Inside the Auxiliary Building, the TC with DSC is transferred from the fuel pool to a position where decontamination, drying, sealing. and installation of the 1C 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 HSH 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 IFA handling. This equipment has been subject to NRC review and approval associated with the Calvert Cliffs FSAR (Reference 14). It is not further addressed in this SER.
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 100 tons. A liftin3 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.
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S .
Lifting slings are used in the Auxiliary Building for placement and removal of the OSC and TC shield plugs and covers. Eyebolts are installed on the items to be lifted to facilitate rigging for lifting.
Equipment used for fuel transfer both inside and outside of the Auxiliary Building is subject to formal NRC evaluation by this SER and as part of the 10 CFR Part 50 license review of updates to the FSAR. This equipment must be determined to be acceptable by both reviews. The fuel transfer equipment subject to both evaluation reviews is identified below:
=
Transfer trailer with integral positioning
. TC support skid 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 cre 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 rigure 1.5. It is pulled by a conventional tractor.
The trailer has eight hydraulic suspensions which carry four pneumatic tires each and are located in four axle lines. There are a total of 32 tires.
Hydraulic suspensions enable coupled steering of all axles around a common point, thus minimizing tire scuffing and the resulting damage to pavement and tires. The suspensions also allow other advantages, such as adjustable deck height, in-site lockout or repair of failed suspensions or tires, and automatic compensation for road surface irregularities. The trailer has all-wheel braking using industrial grade air / spring brakes.
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 1-13
< v 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 operatcr 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 prior to insertion or withdrawal of tne 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. Use of the SPS is illustrated in Figure 1.5 (of this SER).
The hydraulic jacks are designed to support the cask setdown load, and the loads applied to them during the HSH 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 chuck 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 HSH prior to insertion or retrieval of the DSC. The positioning 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.
Anti-friction pads constructed from woven teflon pads and steel are used to reduce the force required to align the cask. Four pads are mounted to the trailer frame. Steel beres or, the skid slide on the teflon surfaces and 1-14
s e protect them from the weather. The travel of the skid is restricted by the stroke of the hydraulic positioning cylinders.
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.
Equipment used for fuel transfer only outside of the Auxiliary Building is subject to evaluation by this SER. This equipment includes slings tu be used with a mobile crane for lifting and positioning the HSH access cover, the TC cover, and the outside TC shield plug covering the rear access port on the bottom of the TC. Eyes are attached to the lifted stems for these operations.
The only equipment used for fuel transfer only outside the Auxiliary Building which is unique to the ISFSI is the hydraulle ram system (HRS).
The HRC provides the motive force for transferring the DSC between the TC and the HSH, The hydraulic ram consists of a double-acting hydraulic cylinder with a capacity of 80,000 lb. In either push or pull mode and stroke of 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 groJnded through the transfer cask. The hydraulic ram system includes a grapple at the l
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 4.7-8).
The HRS includes the following main subcomponents: one single-stage, double-acting, hydraulic cylinder; one grapple assembly; one hydraulic power unit; one ram / cask support frame assembly; one tripod support assembly; hydraulic l-15
,yw , t- - - - p--r m y de - - e
hoses and fittirigs; one hose reel; all necessary appurtenances, pressure limiting devices and controls for the system operation; and, one light duty trailer (for transport and storage of all HRS equipment).
The ram hydraulic cylinder is provided with a support and alignment system which prnvides for the range of vertical and lateral motion necessary for alignment with the DSC, cask, and HSM. The front of the ram hydraulic cylinder is aligned using a ram trunnior, support assembly, and the rear of the ram is aligned using an adjusting tripod assembly. All controls are mounted in one trailer-mounted control panel.
1.5.5 Fuel Handling and Storage Operations The proposed Calvert Cliffs ISFSI is intended to provide a sealed pressure vessel for IFAs in storage. When DSCs with IFAs are in place in the HSMs, the HSM doors are securely fastened by welding. Operations then consist of periodic inspections and recordkeeping. The principal ISFSI operations are therefore those associated with placing the IFAs in storage from their locations in the Auxiliary Building fuel pool and eventually returning them following storage in the HSMs. Contingency operations involve inspection (and any repair) of HSM following " accident" type events, and return of DSCs to the Auxiliary Building for IFA retrieval following any out of tolerance DSC handling accidents.
The operational sequence of placing IFAs into storage is summarily outlined below. The steps and procedures are more fully described in SAR section 5.1.1.
Operations in the Auxiliary Building:
. Fill and seal the DSC-TC annulus with borated water.
- Load IFAs into DSC.
- Place top shield plug on DSC.
1-16 1
5 .
- Place and bolt TC top cover plate.
- Secure TC.
- Remove bottom ram access cover plate from TC and install the inner and outer shield plugs.
Operations at HSM site:
Remove HSM access cover.
- Move transfer trailer and align TC with HSM access port.
Remove TC cover.
- Insert TC into HSC access port and secure.
- Remove outer TC shield plug.
- Withdraw ram, replace outer shield plug.
- Withdraw TC from HSC access port.
Install HSM access and TC covers.
- Withdraw transfer trailer.
- Weld HSH access cover.
DSC and IFA retrieval operations would essentially involve the above steps in reverse order (with cutting replacing welding).
1-17
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CASK TOP SHIELD PLUG COVER PLATE A E / A DSC DRAIN AND DSC SPACER jf0 ~t! : 1I LI H
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\ CASK BOTTCM NEUTBCW SHIELD DSC GRAPPLE RING DSC BOTTOM AND CASK RAM SHIEW PLM PENETRATON Figure 1.4 Transfer Cask with Dry Shielded Canister 1-21
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( =l SECTION VIEW Figure 1.5 System for Transferring DSC between TC and HSM 1-22
O 5 TRANSFER CASK GRAPPLE RAM SUPPORT FRAME HYDRAULC RAM SYSTEM
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Figure 1.6 Hydraulic Ram System Concept Illustration 1
1-23 l
Table 1.1 Incorporation of Docketed Material by Reference Docketed Material (other than sletttals with the SAR. Reference 1) incorporated by reference Staf f Coronents Paragraph CC 1$FS! Envirorrnental Report (ER) Chapters 9.10.11 For need to provide additional on-site storage 1.1 facilities.
(Ref 2)
NOTECH NUH-002 Rev 1A (TR) (Ref 3)
For fuller description of NUHOMS-2CP systens For definition of minismsn factitty design life of So TR. Table 1.2-2 years.
1.2.1-1.2.2 None TR For detailed description of storage system criticality 1.2.3 safety, shielding, structural and decay heat ternoval features.
1.2.4 Mone w
TR. Paragraph 1.2.4 For additional detatis of the safety features of the 1.2.5 NUHOMS-24P systeve.
1.2.6-1.3.1 None TR For detalled description of generic MUHOMS-24P system 1.3.1.1 design.
For stellar design. " differing only in construction 1.3.1.2 TR details."
1.3.1.3-1.4 None TR Incorporated by reference. The statement of Paragraph 1.5 1.5 does not satisfy the 10 CIR 12.18 requirement that the reference be specific. However, specifte references are included elsewhere in the SAR.
2.1.1-2.1.3.3 None For agricultural activltles in the leswdlate area of the 2.1.4 ER ISISI.
2.2-2.2.2 None 9
Table I.I Incorporation of Daciteted platerial by Reference *
(Continued)
Docketed Naterial (other than Paragraph submittels with the SAR. Reference 1) incorporated by reference Staff Ci_. ...ts 2.3.1.1 Reader directed to 3 referenced (2 NOAA Does not constitute a clear or spectfic reference per 10 climatological data and one USAF stsinaary of surface CFR 12.18.
weather observations) but reason for reference or '
material Incorporated not stated.
2.3.1.2-2.3.2.2 None 2.3.3 ER. Table 6.1-IA For joint frequencv. tables of wind direction and speed and atmospheric stability.
2.3.4-2.4.4 None 2.4.5 U.S. Army Coastal Engineering Research Center. For procedures used in the tidal surge analysts for the lechnical Report No. 4
- Shore Protection - Planning open ocean across the Continental Shelf, and Design."
w 4.n 2.4.6-2.4.9 None 2.5 CC FSAR For detailed discussion of the groundwater aquifers.
flow directions, recharge points, and users.
2.5.1-2.1 None 3.I None 3.1.1-3.1.1.I None 3.1.1.2 TR. Paragraph 3.1.1.2 For calculation criteria.
3.I.I.3-3.1.2.3 Mone 3.2 IR. Paragraph 3.2 Tor basis of identification of comonents toportant to safety.
TR Table 3.2-1 For simmary of design loadings for equipment that is tsportant to safety, of applicable codes and standards, and for description of structural and mechanical safety criterta for regnalning design loadings.
3.2.1 1R CC FSAR For design severe wind and tornado loadtngs, for description of site tornado wind and missile loadings (enveloped by design values of the IR)-
iable 1.I Iscorporation of Docketed Material by Reference (Continued)
Docketed Material (other than sutzetttals with the SAR, Refere se 1)
Paragraph incorporated by reference Staf f Corrents 3.2.I.1 None TR, Paragraph 3.2.1.2 For manner of calculating tornado wind and alssile 3.2.1.2 loadings for the Hsie.
TR, Faragraph 3.2.1.3 For details of HSM protection cf DSC from adverse 3.2.1.3 environmental effects.
3.2.1.4-3.2.2 None TR, Paragraph 3.2.3 For basis of aglific ation fae. tor and calculattor= of 3.2.3 domant itSM frequencies.
3.2.4 Mone TR, Paragraph 3.2.5.1 For H5N load coretnations and design criteria.
Y ro 3.2.5.1 For DSC load cruelnations and design criteria.
3.2.5.3 TR, Paragraph 3.2.5.3 3.2.5.4 None TR, Paragraph 3.3.2 Addresses key safety protection systevas.
3.3.1 TR, Paragraph 3.3.2.1 For detailed discussion of the DSC physical containment 3.3.2.1 barriers, and description that there are no credible events leading to breach of DSC contalrunent barrier.
3.3.2.2-3.3.3.2 None TR, Paragraph 3.3.4 Addresses / details of NUHOMS-24P nuclear criticality 3.3.4 safety.
3.3.4.1-3.3.4.2 None For basic reactivity equivalence and crit; calf ty 3.3.4.3 TR methodology.
3.3.4.4, 3.3.4.5, Mone MUREG/CR-0200, Rev 3 For criticality aulysis sequence and 123 GROUP MTH 3.3.4.6 master cross-section library.
O
-~-
Table 1.1 Incorporation of Docketed Material by R=ference (Continued)
- Docketed Material (other than sutzeittals =lth the SAR, Reference 1)
Paragraph incorporated by reference Staff Cc w nts 3.3.5-3.3.6 None 3.3.7.1 TR, Paragraph 3.3.7.1 For discussion showing that all spent fuel handling outside of spent fobi pool done with ITA in DSC.
3.3.7.2-3.4.1 None 3.4.2 10 CTR Part 50. BG&E QA Prmgram For " safety related" requirements in design, s construction, and test. i 3.4.3-3.6 None 4.1.1 IR, Paragraph 8.2 For transport route deston criterta.
i 4.1.2.1-4.2.2.2 None w
,8 4.2.2.3 TR For full description of DSC containment features.
N 4.2.3.1 TR, Paragraph 4.2.3.2. ACI-349 For discussion of array stre criteria. For concrete temperature lim 6ts.
4.2.3.2, 4,3 None 4.3.1 CC FSAR, Paragraph 9.8.2.3 For discussion of Aumillary Butiding weet tletion.
4.3.2-4.3.6 Mone l 4.3.7 CC ISFSI Security Plan For description of security alarm systems.
4.3.8-4.7 None l
4.7.1 CC FSAR, Paragraph 9.7 For descriptions of CC Auniliary Butiding cask hand!Ing crane and spent fuel handling machine.
4.7.2 CC FSAR. Figures 1-5 through I-16 For Auxiliary Butid6ng plans and sections applicable to fuel handling operations.
CC FSAR, "aragraph 9.8 For description of Auxillary Building wenttiation system con f l.. ...t features.
4.7.3.I kone i
t
6 Table 1.1 Incorporation of Docketed Material by Reference (Continued)
Docketed Material (other than sutnittals with the SAR, Reference 1)
Paragraph incorporated by reference Staf f Coments 4.7.3.6 CC F5AR For description of Auxillary Butiding cask handitn3 crane 4.7.3.3-4.7.4.2 Mone 5.1-5.1.3.5 None 5.2.1 CC FSAR For description of Auxiliary Building cask handling crane and spent fuel hand 16ng macntne.
5.2.2-5.6 None _
i l 6.1-6.2 None 6.3 CC F5AR. Paragraph 11.1.2.1 For description of power plant Liquid Idaste Processing d
System.
8 l g 6.4 CC FSAR, Paragraph 11.1.2.3 For description of Solid Waste Processing System.
L i
6.5 None i
7.1.1 NW T.1.2 TR For discussion of 15FSI design considerations that f f ensure that occupational esposures are ALARA. !
TR, Para 1.1.2(2) For discussion of HSM concrete valls and roof
?
TR, Para 7.1.2(3) For d*scussion of DSC lead shield plugs.
1R, Para 7.1.2(5) For discussion of fuel loadtng procedures.
TR, Para 7.1.2(6) For discussion of HSM access opening.
TR, Para 1.1.2(7) For discussion of DSC seal welds.
TR, Para 7.1.2(9) For discussion of HSM enternal shleiding blocks.
TR, Para 7.1.21 0) For discussion of pressure systeve design.
IR, Para 7.1.2(11) For discussion of HSM internal shielding.
IR, Para 7.1.2(12) For discussion of use of procedures. '
4
Table 1.1 Incorporetton of Doc 6eted Material by Reference (Continued)
Docketed Mateetal (other than sutnittals with the SAR. Reference 1) incorporated by reference $ t a f f Conewmt s Paragraph For discussion of use of water in DSC cavity.
1.1.2 (Continued) TR. Para 1.1.2(13)
- 19. Para 1.1.2fl5) For discussion of use of tevrary sh.elding.
7.1.3 None for full description of source wedeling ev thmtology.
7.2.1 TR for discussion of potential airborne releases frce IT A 7.2.2 CC FSAR in the pool.
For detalled discussion of CC 15FSI design features.
7.3.1 TR For ctrnplete description and 111ustration of ISTSI 7.3.2 TR shleiding design. For carelete descriptions of l
shielding methodologtes models.
w s
ORML: CCC-514 Micro. CCC-493. ORNL-5521 For shielding cceputer program packages and cross
@ section data used the analyses.
Grove fngineering: Micro $kysh'ne Manual Version 2; Microshleid tJser's Manual . Verst on 3.
7.3.3 TR For descriptions of 15FSF ventilation system.
7.3.4 None For detatis of modeling techniques used for occupattonal 7.4.1 TR doses.
7.4.2 None 7.5.1 CC FSAR For description of the Radiation Safety Program aMinistrative organtration.
CC FSAR For discussion of Radiation Safety Frogram equipment, 7.5.2 instrtsauantation. and persor_ .e1.
7.5.3 Radiation Safety Manual. ALARA Program For descriptic7_:? Radiation Safety Program Procedures.
7.6.1-7.6.4 None For analytical assssnptions, methodology, and conpater 8.0. 8.1 TR codes used to generate the results in Section 8.
[
Table 1.1 Incrrporation of Docketext staterial by Reference ,
(Continued) !
Docketed Material (other than sutmittals with the SAR. Reference 1) i Paragraph incorporated by reference Staff Carm nts '
8.1.1 TR. Table 8.1-1 For normal operating ioads for twportant to safety twoonents.
TR. Paragraph 8.1.1.2-8.1.1.9 for method of analy' sis.
TR. Table 8.1-2 For material properties.
8 i.1.1 1R. Paragraph 8.1.1.1 For detailed description of loMs applicable to the i normal operation structural analysts. For material ,
densttles.
8.1.1.2 None 71.1.1.3 1R. Paragraph 8.1.1.3 For DSC oasket deed weight and thermal analysts sethodology.
Y c.7 TR Paragraph 8.1.1.3.8 For thermal espaeston algorittes.
8.l.1.4 None 8.1.1.5 (B) TR For enveloping of .possible HsM configurations TR. Paragraph 8.1.1.5 For mei.olology used for ff5M dead and li ve loads.
TR. Paragraph 8.1.1.5 For methodology used for creep ami shrinkege loads.
(D) TR. Paragraph 8.1.1.5 For examination of ef fects of radiation on the cwpresstwe strength and smoules of elasticity of concrete.
(E) TR. Paragraph 8.1.I.5.E For description of uittaute strength method used 8.1.1.6 None 8.l.1.7 TR. Paragraph 8.1.1.7 For applicable heat shleid analysis.
8.1.l.8 None 8.1.1.9 (8) TR, Paragraph 8.1.1.9 For description of TC handling loads and their method of analysts.
(C) TR For stellar method of analysts for thermal loads.
O
Table 1.1 Incorporation of Docketest Material by Reference (Continued) nw. ,n 7
1 S Docketed Material (other than Paragraph sutnittals ulth the SAR. Reference 1)
, incorporated by reference Staf f Coments
+
q IR, Table 8.1-la k For list of of f-nonnal operating loads for the nUHOMS-
_ 24P system.
8.1.2.1 (A) TR, Paragraph 8.1.2.1 For discussion of s'ianisse tolerable DSC misaltg ..t.
(C) TR, Paragraph 8.1.2.1 For discussion of DSC under asstened Jamed and binding l conditions.
I' 8.1.2.2 (A) TR, Paragraph 8.1.1.5 For description of methodology used for HSM off-normal thermal loads.
! (A) TR. Paragraph 8.1.1.4 For description of support assembly slotted holes.
i 8.1.3 TR, Paragraph 8.1.3 !
For description of the HSM, DSC, and TC thermal analysis j models.
- 7 TR, Tables 8.1-5, 8.1-6 For thermalphysical properties of materials of
[
g construction.
1 (A) TR, Paragraph 8.I.l.C For discussion of lifetime an6 tent teuperature.
i 8.1.3.1 TR, Paragraph 8.1.3.1.C t
For description of model used for design basis ambient j air temperatures. '
8.1.3.2 TR, Paragraph 8.1.3.2.A i For description of analytical model used for DSC and IFA heat transfer analyses.
TR Paragraph 8.1.3.1 For description of stainer anetent conditions.
TR, Paragraph 8.1.1 For description of model used for spacer disk thermal analysts.
8.1.3.3 TR, Paragraph 8.1.3 i for defined design basis ambient air temperatures.
TR. Paragraph 8.1.3.3 For description of models used for TC thermal analyses. t 8.2-8.2.2.1 None 8.2.2.2 IR, Paragraph 3.2.1 For methodology used for DBf and D8T missile forces acting on HSM. ,
(A) TR, Paragraph 8.2.2.2 i
. For methodology used in evaluating HSM appited and resisting forces and unents.
s 6
Table 1.1 Incorporation of Dacketed Material by Reference (Continued)
Docketed Material (other than submittals with the SAR. Reference 1)
Paragraph incorporated by referex e Staf f Ctreents 8.2.2.2 (C&tinued) (8) TR, Paragraph 8.2.2.2 for methodology used to evaluate TC DBT ad pressures.
(C) IR. Paragraph 8.2.2.2 For methndology used*
for ef fects of D81 missile loeds on the HSM.
8.2.2.3-8.2.3.1 None 8.2.3.2 TR Paragraph 8.2.3.2 For stellar analytical niethods for evaluating 08E loads.
For discussion of seismic accelerations.
8.2.3.3-8.2.4 None 1
8.2.5.1 TR, Paragraph 8.2 5.1 For discussion shn= lag drop event not credible.
8.2.5.2 TR For discussion of bounding drop deceleration.
Y TR. Paragraph 8.2.5.2 For DSC shell drop, analytical methods. For i.
M methodologies used for DSC c m ponent and TC drop evaluations.
8.2.5.3-8.2.7.1 None 8.- .2 TR. Paragraph 8.2.2.2 for methodology used for thermal analysis of the blocked went case.
For discussion of design basis accident pressure.
For methodology used for HSM thermally induced stresses.
8.2.7.3 None 8.2.8 TR. Paragraph 8.2.8 For description of postulated leak.
8.2 8.1 TR Paragraph 8.2.8.1 for description of basis for excluding rupture of all !
fuel rods frtze DSC leakage case.
8.2.8.2-8.2.10.3 None 8.2.11 BGM Correspondence 3/28/81 For analysts of probability of LNG spill impacting CC.
8.2.12-8.3 Mone 9.1.1-9.1.1.1 CC FSAR. Paragraph 12.1 For description of corporation management organization responsible for CC.
Table 1.1 Incorporatime of Docateted Material by Reference (Continued)
Docketed Material (other than Paragraph submittals alth the SAR. Reference 1) incorporated by reference St a f f C(_ .- . ..t s 9.1.1.2. 9.1.1.3 None i
9.1.1.4 CC FSAR. Paragraph 12.1 1 For description of corporation technical staff.
For description of organtration of CC.
9.l.2.2 CC FSAR. Paragraph 12.1.3 For descriptions of functions, responsiblittles and authorttles of major personnel positions.
9.1.3 CC FSAR. Paragraph 12.1.2 For persennel quallfication requirceents.
9.1.4-9.2.3 None 9.3 CC FSAR. Paragraph 12.2 For description of existing CC training program w
9.3.1.1-9.4.2 None w 9.5 CC FSAR. Paragraph 12.6 w For further details of CC Emergency Response Plan.
9.6 None
! 10.0-10.2.2.1 None 10.2.2.2 TR Paragraph 10.2.2.2 For description of overall technical and operational considerations.
10.2.3 None 10.2.4 TR. Paragraph 10.2.4 For outilne of system design features important to safe operation.
10.2.5-10.3.4 None i
11.1 CC FSAR. Paragraph 1.8 For full description of BG&E QA program.
i, i
,i b
fable 1.2 Principal Difh-ces Betimen IrJDES-24P Systeams as Definerl 'n the TR and the CC SAR Congwnent Area of Difference st90MS T- or sae Materials All materials with exception of Same as Nt#10MS TR s::pport rods in 05C (see below)
HSM DSC loading and unloading Ram in front, port in IC Ram in front, port in TC HSM groups 6 module groups of 3 back-to- 12 module groups of 6 back-to-back module pairs.
back module pairs, ccernon side caninon side and back walIs and back walls Exterior front walls 3' to 3*-6" R.C. 3* thickness to 3*-6" R.C. at access openings and vent i inlets Intermodule walls (side-by-side 2* R.C. side-by-side 2* R.C.
modules) 2* R C. end-to-end Roof slab 3
- R . C. f 10 reba r a t 6" o . c . 3* R.C. #10 rebar at 12" o.c., spanning. each face #8 4
y each way top spanning, at lower at 12" longitudinal, each face 4 face w
j Foundation design 3* R.C. with #9 rebar at 6" o.c. 3* R.C. =4th #10 rebar at 12" o.c. each way, each face each way, each fece Exterior side walls 3* R.C. with #10 rebar at 6" 3* R.C. =tth #10 rebar at 12" each way, each face each way, each f ace 3* R.C.
Design criterton for contact dose 20 mree/hr i 15 mrem /hr rate for HSM ext. surfaces away from door or penetrations
, H94 access door 3" st. plate and 2" solid 1 3/4" st. plate and 10 3/4" cnnerete fill and 1/4" neutron shielding material and steel plate 1/2" steel plate HSM door support frame Continuous L8m6 3/4 on sides and I a-L9142*-0 (2 per side) welded to continuous LFa4 i
bottom with 13-4"n6"m5/8" pl welded to 3/4" n 6" plate on sides and bottom.
with 13-1 1/8" diameter anchor Ea6edment s: Pairs of 1/2" x 6" Nelson studs 9 1/2" -
) bolts 10" o . c . Flange entends 13 1/8" from face.
Air outlet shielding blocks Not designed to withstand Designed to remain in place and withstand all design tornado effects. _
events. including tornado missiles.
e e
9 Table 1.2 Principal Differences Between RAOts-24P Systems as Defined in the it and the CC 5AR (Continuenf)
Area of Dif ference NUHOMS TR CC SAR Conponent HSM DSC support assembly:
(Continued)
Ra;is WT 6 x 115 3/4" x 16" plate on V8 x 40 I Cross Meneer Beams 3-W 10 x 68 2-WS x 48 Cross Member Supoorts ,
Plate 3'-10" x 1*-2" x 7/8' (1/2' p1 4-l* sq. 2-6~ sq) i Inte&ents (ea) 6-2" x 3/4" bar with 3" x 3' x (Ctr) 4-5/8 x 6 9/16" Nelson studs 1/2" ends (End x-been) 4- 3/8 x **
6 t/8' Nelson studs (Acce*s End) 4-7/P' .6" Nelson studs Rall stops Not detailed L 5 x 5 x //8 x 0*-6" t
( Setssic restraint:
Pad plate 3/4" n 1*-6* sq. 3/4" thick u 7"x12" Vertical stop 6' x 1*-6' 3/4" thick n 4' x 12" g Gusset plates 2-1/2" thick. 1-3/4" thick 2-1/2" Shear stops 2-approx 14" long (ea) 1-3/4' x I" x l'0" long
{
4.n Access opening:
Main sleeve 3/4" th 3/4" th Collar disk I/2" th 3/4" th i IC mating sleeve 3/4" th 3/4" th j Guter collar 1/2" th 3/4" th l l
DSC support rail plate 3/4" Enee&ents 16-5/8" x 8 3/16" anchors (=tth 16-5/8" x 6" Nelson studs heads)
Cask restraint assembly Not detailed 4 eneedded 2" x 3*-0~ ASTM A-489 eyebolts with nut Heat shield 11 ga. ASIM A 240 11 ga. ASTM A 240 5t Steel. =tth 2* standoff St Steel, with 2" standoff l
Table 1.2 Principal Ot f ferences Between FJIDMS-; M Systews as Defined in the IR and the CC SAR (Continued)
Caponent Area of 01fference NUHOMS TR CC SAR HSM Intermediate walls 2* R.C.. f3 rebar at 12" each way, each face (stde by (Continued) side and end to end) side by side 2* R.C.
end to end 2* A.C.
Roof 3* R.C. #10 at 6" o.c. 3* R.C. #10 at 12'*o.c. each face lat. #8 at 12' e.f.
each way top. lat. bot. longitudal Foundation 3* R.C. #10 at 6' each way, each 3* R.C. fl0 at 12" each way, each face face Constr. Joint vert. dowels:
Fdn to Int. walls #9 at 12' each face fat. side walls fl0 at 12' each face Y
t.s m l DSC Welds and cover design Shielding covers dif ferent froen NUHOMS-24P et Oconee Design drop height 80" 80' Design mantaman vertical 75 g 75 g decelerati.m.
Design maxinsa hortrontal 75 g 75 g dereferation _
Fuel guide sleeves it ga. SA-240 12 ga. ASTM A-240 No. of spacer disks 8 9 Spacer disk thickness 2" ,
1.5" e
e
I r
i o. !
P l
i: Table 1.2 Principal Differences Between MARBES-24P Syntamm as Befined is the TR and time CC SAR (Contimm:d) ,
?
2 '
Camponent Area of Of f ference i leUH08t5 TR CC SAR
- OSC Support rod 3" disseter 3" diameter I (Continued) 4-Support rod amaterial ,
Lower end caps: (+1/2" lead)
Inner pressure plate 1.00" 2"
, Lead plug Appron. 4.75" r 5" + 2x I/2" 55 Outer cover - 1.25" 0.25" >
Upper end caps: (+1/2" lead) ?
Inner pressure plate 0.5" 1.75" Lead plug Approxlestely 3.5" 5" + 2 x I/2" 55 i 7 Plug assenely top plate 0.5" '
to Outer pressure boundary 1.25" 1.25" N ;
Fuel asses 411es to be stored: !
Type k Initial enric! ament Any PWR CE 14's 14 PWR t i- fossil content i 3.5% 1 4.5%
l' Burnup. IWd/MT 1 33.000 1 45.000 - 4/ 000 1
Decay power, kW/assy 1 0,660 1 0.660 Post Irradiation cooling <1 <!
Neutron source. 15 1 5-h/sec/assy 1 1.43[8 1 1.67[8
- i Gansna source. -
!- photons / rec /assy i 7.76[15 1 5.73[15 nielght-5 688 69 5 660 kg f
Compatibility with TC w/NUHOMS TC w/NUMONS TC i, Siphon and went tube geometry identical witte TR
! i i
Radial gap between DSC and spacer. 0" 0.13"'(nominal) l; disk inside i
i
. . _ - y . . .. -
w . -. . --- - _ , , - - - - - - - - - - - _ -
.Wmim l
Table 1.2 Principal Dif ferences Between IKRDtS-24P Systeuns as Defined in tie IR and the CC SAR (Continued)
Area c Olfference NLTIDMS TR CC SAR Ccmonent C
DSC Material to be stored:
(Contlesed)
Neutron sourca 1.548f8 2.23 x E8 strength (neut/sec)
Gama source 4.62E15 4.21 x EIS strength photons /sec/lFA "(photons /sec/lFAl" Total decay < 0.66 6W i 0.66 kW best power /assy __
, I l
l ; Fuel assenblies:
Design B&W 15 x '15 CE 14 x 14 1.8 w/o i/"
l'quiv. enrictsne'a 1.4 w/o if" Transfer Cask Neutron shell liqu d in annulus with expansion i $olid hydrogenous neutron shleiding. No fluid and no tank ewpansIon tank.
Shielding Design details differ from TR Y
w
- Design criterton for nominal i 200 arem/hr i 400 erem/br contact dose rate on TC ,
fuel Pool Water in TC-DSC annulus Dentnereltred Deminerallied l Operations Water drained when DSC dralred Use of water in TC-DSC annulus Water lef t in annulus da.rtngcic}ureoperations ,
l i
t a
l
- o l
I 2.0 SAFETY EVALUATION 2.1 System Design Criteria and Suitability at Calvert Cliffs Subpart F of 10 CFR Part 72 discusses general design criteria for an ISFSI.
This section of the SER examines all aspects of the design criteria as stated in Subpart F and relates them to design criteria as stated in the BG&E SAR.
Section 72.122(a) of 10 CFR Part 72 outlines the requirements for quality standards. Basically all " structures, systems, and components important to safety must be designed, fabricated, erected, and tested to quality standards commensurate with t' importance to safety of the function to be performed."
Subpart G of 10 CFR Part 72 spells out 18 topics of a quality assurance program which must be considered. These topics are in most cases identical to the topics defined in Appendix B of 10 CFR Part 50. Section 4 of this SER evaluates the suitability of the quality assurance program as proposed by BG&E 1
"r the ISFSI.
w : tion .J 3 of 10 CFR Part 72 defines the phrase " structures, systems, and nponents important to safety" to mean those features of an ISFSI whose anction is: "(1) to maintain the conditions required to store spent fuel or _
high-level radioactive waste safely, (2) to prevent damage to t% spent fuel or the high-level radioactive waste container during handling and storage or.
(3) to 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."
2.1.1 Spent Fuel Elements to be Stored at Calvert Cliffs The generic NUHOMS-24P system is designed for dry, horizontal storage of irradiated PWR fuel from nuclear power stations. The site-specific criticality safety analysis performed for_ the Calvert Cliffs ISFSI determined g that the Calvert Cliffs fuel was less reactive than the design basis fuel for the generic NUHOMS-24P system. This result has been taken into account in the
O t site specific Safety Analysis Report resulting in somewhat different operating controls and limits from the generic system.
The fuel to be stored in the Calvert Cliffs ISFSI is restricted to irradiated Calvert Clif fs fuel assemblies with a maximum initial enrichment 4.5 wt %
235 U. The Calvert Cliffs fuel assemblies designed by Combustion Engineering consist of a 14x14 array of zircaloy clad UO2 fuel rods which are referred to as CE 14x14 fuel. Irradiated fuel assemblies acceptable for storage are limited to those assemblies with sufficient burnup to reduce the fissile 235 content to an equivalent initial enrichment of 1.8 wt. % 0. For an initial _
235 enrichment of 4.5 wt.% U, this corresponds to a burnup of approximately 39 GWD/MTV at a specific power of 32.2 MW/MTV.
Criticality and thermal considerations respectively require a post-irradiation cooling period for a fuel assembly for the minimum of: (1) 5 years; or (2) the time required to limit the total decay heat power to 0.66 kW.
2.1.1.1 Criteria for Nuclear Criticality Safety Section 72.124 of 10 CFR Part 72 requires that spent fuel handling, transfer, and storage systems be designed to be maintained subcritical. The margins of safety should be commensurate with the uncertainties in the handling, transfer, and storage conditions; in the data and methods used in the -
calculations; and in the immediate environment under accident conditions.
Section 72.124 also requires that the design be based on either favorable
-geometry or permanently fixed neutron absorbing materials. For accident conditions under monitored loading and unloading operations in the spent fuel pool, the staff.has considered soluble neutron absorbers in the canister fill water as an acceptable method to maintain a subcritical configuration.
Section 3.3.4 of the proprietary supplement to the SAR addresses the nuclear criticality safety criteria. The review of the safety analysis is given in Section 2.2.2 of this SER.
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.c- .
The SER for the generic NUHOMS-24P TR design establishes a maximum effective reactivity of 0.95 for all credible configurations and environments under unmonitored storage conditions, and a maximum effective reactivity of 0.98 for all credible configurations and environments during monitored operations with a 50-hour drain down limit. Monitored operations are considered to be " wet" operations, and unmonitored operations are considered to be " dry" operations.
This SER establishes the criteria for the accident case to be k,,, 50.95 with no drain down limit.
2.1.1.2 Criteria for Radiological Protection Section 72.24 of 10 CFR Part 72 requires the licensee to provide the means for controlling and limiting occupational radiation exposures within the limits given in 10 CFR Part 20 (Reference 23), and for meeting the objective of maintaining exposures as low as is reasonably achievable (ALARA).
Section 72.126(a) of 10 CFR Part 72 requires 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.
Section 20.101(a) of 10 CFR Part 20 states that any individual in a restricted area shall not receive in any period of one calendar quarter from radioactive material and other sources of radiation a total occupational. dose in excess of 1.25 rems to the whole body. Section 20.101(b) states that, under certain conditions, the quarterly dose limit to the whole body is 3 rems in any calendar quarter.
2.1.2 Criteria for Spent fuel Handling and Transportation Section 72.128 requires that spent fuel systems that "contain or handle radioactive materials asenciated with spent fuel or high-level radioactive-waste must be designe to ensure adequate safety under normal and accident conditions." They must be designed with: "(1) a capability to test and monitor components important to safety, (2) suitable shielding for radioacti.,
protection under normal and accident conditions, (3) confinement structures 2-3
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 wastes generated."
Fuel Handlino 00erations Inside the Auxiliary Buildino Section 4.7 of the SAR discusses the fuel handling operations and the general technical requirements and criteria for the necessary equipment. The fuel handling equipment, which is used inside the spent fuel pool building, consists of the cask handling crane, the spent fuel handling machine and the TC and lifting yoke. All of this equipment and the associated procedures for use of the equipment is governed by 10 CFR Part 50 and applicable technical specifications for Calvert Cliffs.
The single-failure proof crane which is rated for a critical load of 125 tons has recently been installed at Calvert Cliffs specifically to meet 10 CFR Part 50 requirements for moving spent fuel over the spent- fuel pool. Because the critical load to be moved is 100 tons, the new crane has appropriate capacity. The crane will meet the design requirements of CMAA Specification 70 for 1983 (Reference 18).
Prior to ISFSI operations, the utility intends to obtain a new spent fuel handling machine capable of accessing all spent fuel pool locations. This new design will meet all 10 CFR Part 50 criteria, and be more flexible in accessing fuel than the existing spent fuel handling machine.
The criteria for the tranefer cask are discussed in Section 2.1.5 of this SER and Section 3.2.5.3 of the NUTECH TR for the NUH0M';-24P design. Basically, the TC is designed to meet the criteria established by the ASME B&PV Code Section III, Subsection NC for all functions except lifting (Reference 19).
For lifting, the trunnions are designed to meet the criteria for lifting heavy loads per ANSI N14.6 (Reference 20). ANSI N14.6 includes reference to ferritic materials, which must be taken into consideration as discussed in Section 5 of this SER. Further discussion on this topic is includad in the evaluation of the TC in Sections 2.2.3.3 and 5.0 of this SER.
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j ,- .
The criteria for the lifting yoke are discussed in a calculation package entitled BGE001.0209 (Reference 1). The yoke is designed to meet the requirements as specified by ANSI N14.6 for single load path lifting devices.
This is satisfactory.
Fuel Handlino Operations Outside the Auxiliary Buildino The four major components necessary for transporting the DSC to the HSM outside the spent fuel pool building (auxiliary building) are the TC, the TC trailer and skid, the skid positioning system and the hydraulic ram system.
The design criteria for the TC are discussed in Section 2.1.5 of the SER. All of the remaining equipment are classified as not important to safety, primarily because no accidents involving the use of the TC trailer, skid or ram were postulated which would cause a nuclear criticality accident, a nuclear material release or an inability to recover safely. Various accidents involving a loaded TC are evaluated in Section 2.2.3.
The design criteria for the TC trailer are summed up by the SAR as " commercial grade commonly used to haul very heavy loads." Section 4.7.4.2 of the SAR references the AISC, " Manual of Steel Construction" (Reference 21) and the AWS Dl.1, " Structural Welding Code-Steel" (Reference 22) as the relevant design codes. Since the TC is never lifted higher than 80 inches after it is secured to the trailer, the credible drop from the trailer has been analyzed and addressed in Sections 2.1.5 and 2.2.3 of this SER. This is satisfactory.
The design criteria for the skid positioning system are referenced in Section 4.7.4.2 of the SAR. They are the AISC, " Manual of Steel Construction," and the AWS DI.1, " Structural Welding Code-Steel." This is satisfactory since this equipment is not important to safety.
The design criteria for the hydraulic ram, which is used to push and pull the DSC into/out of the HSM, consists of the AISC, " Manual of Steel Construction,"
the AWS D1.1, " Structural Welding Code-Steel," and unspecified National Fluid Power Association Standards.
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M The ram force is limited to a maximum of 80,000 pounds. A force equal to or less than 80,000 pounds has been used for design for the most serious jamming accident. The design criteria for the hydraulic ram are considered acceptable.
2.1.3 Design Criteria for Confinement Barriers and Systems Section 72.122(h) requires that cviflnament barriers and systems: "(1) 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" (Reference 23).
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 review was directed at two aspects of the design: =the integrity of the DSC and the allowable leak rate. As a result of this review, the staff concludes that the DV design conforms to applicable parts of 10 CFR 72.122(h). Confinement is ensured by a combination of inspection techniques, including radiographic inspection, dye penetrant testing, and helium and soap bubble leak testing.
The NVHOMS-24P TR 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 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 5; 10 atm-cc/sec, as well as on the integrity of the DSC. The confinement integrity is ensured by the 2-6
c .
l use of stainless steel, thus precluding corrosion of the DSC, and also by the I design criteria which include accident cases such as a drop, The staff also analyzed the impact of long-term storage on the behavior of spent fuel, using a diffusion controlled cavity growth (DCCG) mechanism as the basis for this calculation since it appears that-this damage mechanism:is the only one applicable to these storage conditions. Under the influence of stress and temperature, this damage mechanism progresses by the nucleation and growth of cavities along grain boundaries. The results of these concerns are addressed in Section 2.2.5 of the SER.
The an1 .yses are predicated on the knowledge and control of the character of the spent fuel loaded into the DSC, particularly the quantity, specific power, and age of the fuel assemblies, and the heat dissipation properties of the system. The thermal evaluation is addressed in Section 2.2.6.
2.1.4 Criteria for Fuel Stability The general design criteria for an independent spent fuel storage installation are given in Subpart F of 10 CFR Part 72. Section 72.122(h) covers
" Confinement Barriers and Systems." Paragraph (1) of this section is pertinent to storage of spent fuel as proposed by Baltimore Gas and Electric Company. It requires 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." Paragraphs (2) and (3) in this section relate to underwater storage of fuel and to ventilation and off-gas systems, respectively, and are not applicable to this review. Paragraphs (4) and (5) l deal with monitoring and' handling and retrievability operations, respectively, and are addressed-elsewhere in this document.
2.15 Protection Against Environmental Conditions and Natural Phenomena The Calvert Cliffs ISFSI is designed to withstand environmental conditions and
- natural phenomena that may occur under " normal," "off-normal," and " accident" l
l 2-7 l
l
circumstances. Design criteria require normal and off-normal conditions (such as would probably occur at sometime during the operational life of the installation) to both satisfy allowables and limits for routine normal operations. Extreme environmental conditions which could possibly occur, such as tornadoes, earthquakes, and floods, are treated as accidents. Higher allowable stresses and some permanent deformation may be permitted for
" accidents." Some phenomena, such as temperature extremes, may have both normal /off-normal and accident sets of values, and corresponding stress or response design criteria.
2.1.5.1 Normal Operating Conditions Table 2.1.5-1 lists summary design criteria used for the Calvert Cliffs ISFSI for normal operating conditions. The first five columns are a repetition of Table 3.6-1 of the SAR (modified in accordance with Reference 1 submittals).
The final column presents the staff assessment of the appropriateness and acceptability of the criteria.
Evaluation of a prepared 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 tt,a criteria the system may not be acceptable. Similarly, criteria may not be acceptable, but the actual design may be determined to satisfy criteria which are acceptable, in which case the ISFSI design would be acceptable. Conservatism has often produced designs which satisfy more stringent acceptable criteria although the stated criteria were not acceptable.
The staff considers that the design criteria as stated and referenced in Table 2.1.5-1 are acceptable. Treatment of the DSC as a dead load instead of a live load is not considered to be within accepted practice, however, the staff considers that it may be accepted in this case due to the precision to which the DSC weight is known and the use of additional handling loads for the transfer situation.
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2.1.5.2 Off-Normal Operating Conditions Table 2.1.5-2 lists summary design criteria used for the Calvert Cliffs ISFSI for off-normal operating conditions. The first five columns are a repetition of Table 3.6-2 of the SAR (modified in accordance with Reference I submit- .
tal s) . The final column presents the staff assessment of the acceptability of the criteria. The staff considers that the design criteria stated and referenced in Table 2.1.5-2 are acceptable and appropriate.
2.1.5.3 Accident Conditions Table 2.1.5-3 lists summary design criteria used for the Calvert Cliffs ISFSI for accident conditions. These conditioris include extreme natural phenomena, accidental drops and impacts, fire, and explosions. The first five columns are a repetition of-Table 3.6-3 of the SAR (modified in accordance with Reference I submittals). The final column presents the staff assessment of the acceptability of the criteria.
The staff considers that the design criteria as stated are acceptable. Where other criteria are considered more appropriate, the criteria stated have been determined to be conservative and thereby acceptable. Design criteria have not been stated for lightning strike of the TC with DSC in transit or at the HSM.
Separate staff review of these conditions has indicated that the design is acceptable for the hazard, although design criteria have not been stated.
2.1.6 Protection Against Fire and Explosion.-
As noted in Table 2.1.5-3, the SAR includes only limited criteria related to fire and explosion. The sole fire situation for which criterion is provided is a forest fire of one hour duration 130 feet from a HSM. With regard to an explosion due to an LNG -spill affecting the ISFSI, the licensee has committed to providing the results of a deterministic consequence analysis currently underway no later than 60 days prior to the start of the nearby Cove Point LNG facility.
This analysis is being done under the provisions of the 10 CFR Part 50 Reactor Operating licenses. ,
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. o Damage limits or other response criteria are not stated in the SAR for fire or explosions. The staff considers that appropriate criteria are that radioactive material containment, radiation shielding, and IFA retrievability should not be affected by fire or explosion; and that IFA cladding temperatures should not exceed the 335 C limit (SAR Paragraph 10.3.IG). The staff used these as criteria for a separate evaluation of potential fires.
Regulatory Guide 3.48 (Reference 16) includes the statement that: " Portable fire suppression equipment will be provided within the fenced boundary whenever a motor vehicle is stationed on the ISFSI site. In addition, signs will be posted stating " DIESEL POWERED VEHICLES ONLY WITHIN THE ISFSI SITE,"
to minimize the fire hazard.
The staff considers that the Calvert Cliffs ISFSI provides adequate protection for fire hazards, based on the designs of the HSM, TC, and DSC, submitted analyses, the stated precautions, and separate staff comments.
The consequence of an explosion from the liquified natural gas plant or pipeline 3-1/2 miles from Calvert Cliffs is being assessed and the results will be submitted by BG&E when they become available. No other potential explosion sources were identified. The staff considers that an explosion due to a vehicle fire might occur. However, the system's inherent resistance to blast pressures, and the other measures indicated are such that the threat tn nuclear safety is negligible.
As a result of the above, the staff considers that the proposed ISFSI design and procedures are acceptable for fire and explosion protection.
2.1.7 Instrumentation and Control Systems The SAR states (Paragraph 3.3.3.2) that "no important-to-safety instrumenta-tion is required for operation of the facility." Instruments are used, however, in conjunction with the operations associated with the fuel pool to HSM transfer sequences.
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.. _ ~ .
l- . .
Instruments used in the Auxiliary Building which would be used with the ISFSI, as for other operations include: instruments measuring the boron content of the spent fuel pool water and the surface contamination on the DSC and TC, and instruments measuring the surface dose rates for the DSC and TC.
Additional instrumentation to be used in the Auxiliary Building which may not already be used in current operations would provide: helium leak detection of the OSC welds, helium pressure in the DSC, and vacuum measurement of the OSC.
These instruments may also be used for weld inspection. Use of the above instruments for ISFSI operations within the Auxiliary Building appears appropriate but is not formally evaluated by the SER. Formal NRC evaluation is in conjunction with 10 CFR Part 50 review of updated FSAR and associated documentation.
Instrumentation used outside of the Auxiliary Building and specifically associated with the ISFSI would be as follows:
Prime mover instruments. The principal concern is that prime mover instruments be operational, support velocity limits, and reduce -the probability of vehicle malfunction or fire.
Measurement of HSM surface dose rates.
Measurement of air temperature rise through each HSM following loading.
Use of the instruments is summarily described in Sections 5 and 10 of the SAR (by statements and by inclusions TR paragraph 10.3.5.2, 10.3.5.6, and 10.3.4.1 by reference).
Controls and evaluations of field measurements are not described; however, these should be included in operational procedures prior to initial use ~of the--
ISFSI for transfer and storage of IFA.
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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.
2.1.8 Criteria for Decommissioning The SAR (Paragraph 3.5) notes that it is anticipated that the DSC will be transported intact with the contained fuel to a federal repository when such a facility is operational, and thereby would never be decommissioned and disposed of as radioactive waste. The SAR also notes that should DSCs be emptied and need to be disposed of as contaminated waste, their decontamination and disposal would be as for other materials and equipment used at Calvert Cliffs.
The BG&E Decommissioning Plan (Reference Ic) includes an analysis of the
" Residual Radioactivity and Activation of the HSM's." This analysis was based upon the assumption that a portion of the HSM building materials will become activated due the presence of neutron radiation emanating from the DSC.
Should the HSM building materials become activated, then a portion of the HSM would have to be disposed of as radioactive waste. The decommissioning plan calls for the removal of up to 6 inches of the inside surface of the HSM to remove the activated material. This portion would be processed as low level radioactive waste, while the remainder of the HSH is disposed of as clean trash. A discussion of the techniques for concrete removal was included.
The assumptions concerning amount of material activated appear to be conservative, and the techniques required to accomplish the decontamination were found to be acceptable.
Action on the BG&E application is not to imply NRC approval for off-site shipment of spent fuel assemblies in DSCs. This is a possibility and may be desirable, if the provisions of 10 CFR Part 71 and other federal (and, where appropriate, state) requirements are met, but the alternative of disposal of the DSCs as radioactive waste (or as containers for the disposal of low-level waste) should also be planned. The BG&E comments that all of the components 2-12
l of the NUHOMS system are of materials similar to those used for other Calvert Cliffs facilities, which may also require decommissioning. This is acceptable as a broad statement.
There is no evidence in the SAR that the DSC design criteria include any criteria which are to facilitate decommissioning. This would be unacceptable except that the staff considers that criteria to facilitate decommissioning must necessarily have lower priority to those relating to the functional HSH safety requirements. Design criteria to facilitate decommissioning directly conflict with structural requirements t' at provide for confinement under all postulated conditions. In other areas, the selection of materials, suited to decontamination and whose surfaces will not significantly degrade, aids decommissioning decontamination. Satisfaction of " ready retrievability" (after confinement is ensured) also supports decommissioning.
The staff considers that although criteria to facilitate decommissioning are not included for the DSC design, the design criteria used provide for decommissioning to the extent practical in view of the higher priority design requirements.
Design criteria specifically to facilitate decommissioning are not provided for the transfer cask. Functional requirements include surfaces permitting ready decontamination, but otherwise practically preclude use of a system that might be more conveniently fragmented. Disposal of the TC as radioactive waste (or as a low-level waste container) is a potential requirement.
The staff considers that the TC design criteria are acceptable regarding decommissioning in view of the competing higher priority functional safety requirements.
The staff considers that the design criteria which could facilitate decommissioning of the HSH may also require design practices not required for other structures with greater probability of eventually constituting radioactive waste. The use of reinforced concrete in itself does not facilitate decommissioning. The use of concrete design practices acceptable 2-13
e to the NRC (e.g., ACI 349) (Reference 24) results in generally monolithic structures whose dest.uction will involve formation of dust and large volumes of rubble, and will generally require extensive demolition effort carried out by site personnel.
The staff considers that the design criteria used for the HSM is consistent with that permitted by the NRC for prior structures which also were required to have decommissioning plans and to be designed to facilitate decommis-sioning. The HSM criteria are therefore considered to be acceptable for decommissioning. _
Other elements of the proposed Calvert Cliffs ISFSI are used exclusively in the Auxiliary Building and therefore would be decommissioned in accordance with 10 CFR Part 50, or are not expected to become radioactive and would therefore not be the subjects of decommissioning.
2.1.9 Findings and Conclusions On the basis of the analysis presented in the SAR, the supplementary information presented in response to questions, and independent confirmatory analysis, within the operating controls and limits specified in this SER, it is concluded that the NUH0MS-24P system is designed to be in compliance with all sections of 10 CFR Part 72. -
The staff has summarized its findings in two forms in this SER. Tables 2.1.2-1, 2.2.2-2, 2.1.2-3, 2.1.2-4, 2.1.5-1, 2.1.5-2, 2.1.5-3, and 2.1.6 present a summary of all design criteria for the civil ano mechanical structures designated as important to safety as well as those not designated as important to safety. Chapter 7 of the SER evaluates the Calvert Cliffs ISFSI with respect to every section in 10 CFR Part 72 and presents staff conclusions on acceptability.
The criteria for system design for protection against environmental conditions and natural phenomena have been examined and compared with the FSAR and prior NRC action on the Calvert Cliffs site,10 CFR Part 72, applicable NRC 2-14
Regulatory Guides and NUREGs, national consensus codes, and the-functional objectives of the system. The criteria pr esented in the SAR are acceptable and the staff recommends their approval, but also finds them to be incomplete.
There are omissions in the~1oad combinations, nonetheless, the design is acceptable, as noted in Table 2.1.5-1.
As noted in Table 2.1.5-3, the SAR does not develop criteria for design or suppression of fire or explosion in the NUHOMS-24P installation. Instead, the system design is oriented on not having combustible fuel present. The system has high inherent resistance to the effects of fire or explosion from internal or external sources, due to type of construction and other design criteria parameters (principally tornado pressure and missiles). The staff finds the effective results of this approach to be acceptable and recommends approval of the SAR presentation on fire criteria. With regard to explosion criteria, the possibility of the Cove Point LNG Plant or pipeline becoming operational by 1993 has been rarcrted by the licensee. A postulated leak of LNG could result in : oRpission from dispersed cloud of natural gas over the ISFSI Site.
Similar to its requirements for the reactor operator licenses, BG&E has committed to provide a hazards consequence analysis of such a postulated leak 60 days prior to Cove Point opr. ration.
The staff concludes that Cove Point currently does not pose an explosion risk to the ISFSI Site. However, in accordance with BG&E's commitments, at least 60 days before Cove Point becomes operational, the licensee shall provide a hazards analysis for review by the staff. With the aforementioned stipulation, the staff concludes that there is no explosion hazard at the ISFSI site.
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 review was directed at two aspects of the design: the integrity of the DSC and the allowable leak rate.
The staff concludes that the DSC design and the DSC leak test procedure will ensure the integrity of the DSC over its storage lifetime. .The staff found the design acceptable for potential oxidation and degradation of the cladding.
The detailed evaluations are found in Section 2.2.5.
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ISFSI transfer and storage operations, if conducted in accordance with the limiting conditions and specifications in the SAR as well as the single addition as mentioned below, will protect public health and safety.
Siting of the ISFSI is in accordance with the off-site dose requirements of 10 CFR Part 72.104(a).
The NRC staff review of the design criteria for the NUHOMS-24P mechanical equipment necessary for transportation, handling, and storage of spent fuel is examined in Sections 2.1.2 and 2.1.5 of the SER. Tables 2.1.2-1, 2.1.2-2, ,
2.1.2-3, 2.1.2-4, 2.1.5-3 summarize all the criteria for normal and accident conditions. With few exceptions, the NRC staff review accepts the proposed design criteria.
Specific cases to which the NRC staff takes exception are:
BG&E has not adequately considered the implication of the use of ferritic steel used as trunnions for the TC. ANSI 14.6 paragraph 4.2.4 is specific about the testing requirements as well as restrictions for use at low temperatures. This concern is discussed in Sections 2.1.5 and 2.2.3. For the purposes of the 10 CFR Part 72 license, the only restriction that the staff imposes on the use at low temperature is that any lift must be restricted to material temperatures higher than the nil ductility transition (NDT) tempera-
- ture plus 40 F.
2.2 NUHOMS-24P System Design at Calvert Cliffs 2.2.1 Acceptability of NUHOMS-24P at the Calvert Cliffs Nuclear Power Plant Site location The proposed ISFSI facility is to be located on the Calvert Cliffs site near Lusby, in Calvert County, MD, at latitude 38'-25'-39.7" N and longitude 76 -
26'-45" W. The site is approximately 45 miles from 9ashington, DC and 60 miles from Baltimore, MD. The Calvert Cliffs site covers 2300 acres. The 2-16 l
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s a controlled area is totally within the BG&E owned property boundary and has a minimum radius of 1189 meters. The proposed ISFSI will be centrally located-within the BG&E_ controlled area.
The ISFSI will be within two parallel security fences, enclosing an overall-area of 226 by 666 feet. There are no towers or stacks at the site whose collapse could have an effect on the ISFSI.
Nearby Industrial and Military Facilities The only nearby military installation is the Patuxent Naval Air Test Center, located 10 miles south of the ISFSI. The only industrial facility within 10 miles of the ISFSI site is the Cove Point Liquid Natural- Gas terminal and pipeline, located about 3 miles south of the ISFSI.
Transoortation There are two general aviation airports within 10 miles of the ISFSI. The private Chesapeake Ranch Airport is 6 miles to the southeast and has a single 2500 foot runway. Representative aircraft using this airport have weights of 5300 pounds and below. The information provided by the SAR (paragraph 2.2.1.2) indicates that the approach to the field passes over or near the site. The St. Mary's County Airport is 10 miles southwest of the ISFSI. This airport is expected to handle aircraft with a gross take off weight up to 12,500 pounds.
An underground 36 inch natural gas pipeline passes approximately 1 3/4 miles to the southwest of the ISFSI. The pipeline runs from a liquified natural gas tanker terminal near Cove Point approximately 31/2 miles southeast of the ISFSI on the Chesapeake Bay.
A rail line passes more than 9 miles to the southwest of the ISFSI. A state highway (combined MD state routes 2 and 4) is adjacent the Calvert Cliffs site and is the closest off-site road to the ISFSI. This road comes within approximately 5000 feet of the ISFSI.
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The Calvert Cliffs site is on the western shore of the Chesapeake Bay. This serves as the channel for ships to the major port of Baltimore. This involves general cargo and tanker traffic and includes tanker barges.
The principal threats to the ISFSI from transportation would be a fuel-air explosion due to 2 rupture of the natural gas pipeline, explosion of cargo on the state highway, crash of a light aircraft on the approach to the Chesaceake Ranch Airport, the crash of a naval military aircraft, or forest fire ignited by any of these. .
Population The 1980 census showed a population of 3358 within 5 miles and an additional 14,508 living between 5 and 10 miles from the ISFSI site. The projected population for 2010 is approximately 10,000 within 5 miles and an additional 31,607 living between 5 and 10 miles from the site. Neither the 1980 or 2010 data show any persons living within 1 mile of the ISFSI site.
Fire and Exolosion Protection The ISFSI is sited near a pine forest. Depending on the extent of clearing for the site the trees are assumed to be as close as 130 feet from HSM. There are no combustible materials to be used for IS'SI components. The principal fuel to be within the ISFSI perimeter will be in security and maintenance vehicles, and the prime mover and crane used for DSC transfer and placement operations. These vehicles also constitute the only significant explosion threat within the ISFSI perimeter.
t The sources of other possible explosions (the pipeline, vehicles on the state
! highway, or ships in the Chesapeake Bay) are considered to present a lower i blast pressure threat than the design pressure requirements for tornadoes.
The blast pressure due to a potential explosion associated with the Cove Point LNG facility will be part of the analysis performed by BG&E, which is to be submitted 60 days prior to the startup of the Cove Point facility.
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The amount of fuel potentially present in vehicles on the site or spilled by the crash of a light airplane is not considered to constitute a threat of HSM or spent fuel assemblias temperature rise that would exceed the situations addressed in the off-normal and accident temperature analyses. A forest fire would constitute a significant threat and is discussed at paragraph 2.2.1.1.
2.2.1.1 Natural Phenomena The normal and off-normal site conditions and extreme environments (treated as
" accidents") are considered to be adequately addressed and included in the Environmental Report and SAR (Sections 2.3, 2.4, and 2.6). These have been used for design criteria, as expressed in the SAR, Section 3.2. The conditions and environments of the principal concern for the ISFSI evaluations are summarized in the paragraphs below.
Ambient Temperature The extreme temperature range recorded at the Patuxent River Naval Air Station is -3 to 103 F. A range of -7' to 105 F has been recorded at Baltimore-Washington International Airport. The analysis used -3 to 103 F as the normal condition temperature range, An extreme temperature of -3*F was used for the " accident" situation. The staff notes that the 103 F ambient temperature is limiting for the TC solid neutron shield material. The staff -
considers that the temperatures and ranges used for the analyses are acceptable, provided BG&E incorporates an additional limiting condition of operation which restricts TC movement when daylight ambient temperature exceeds 103'F. (See Section 5 of this SER.)
Precioitation. Liahtnino. Snow. and Ice The SAR includes precipitation data for Patuxent Naval Air Station, and Baltimore-Washington International and Washington National Airports. The Patuxent NAS data are considered most pertinent. The 24-hour maximum rainfall is 5.88 inches. The 24-hour maximum rainfall recorded at Patuxent NAS is 2-19 l
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11.7 inches.
Regional extremes are 7.82 inches of rain and 22.8 inches of snow, recorded at Baltimore-Washington Airport.
Thunderstorm records for Patuxent NAS show an average of 54+ per year with a average duration of 1 bour and 20 minutes.
The area _is subject to high winds and heavy precipitation from hurricanes.
The area is threatened by an average of one hurricane or tropical storm per year, and experiences such a storm an average of every 10 years.
The topography of the ISFSI site cause it to not be susceptible for flooding.
The staff considers the identification of precipitation extremes and ranges and the use made thereof to be acceptable.
The possibility of lightning strikes on the HSM is assumed.
The HSM is therefore required to meet the Lightning Protection Code (NFPA-78, . The 1987) potential by for lightning strike of the TC with DSC has been previously reviewed the staff.
The TC and DSC are considered to acceptably protect the spent fuel assemblies without further provisions for lightning.
The Calvert Cliffs FSAR provides a 30 psf design basis snow loadA 200 psf roof live load was used for the HSH to envelop all potential live loadings including snow and ice. ,
No snow or ice loadings were used to the TC due to the nature of use, curve surface, and internal heat.
A DSC loaded with spent fuel assemblies is never exposed to direct snow or ice loadings. The staff concurs in the snow and ice loadings and total live load used for the HSM, and in not using such loadings for the TC.
Extreme Winds The wind forces due to hurricanes or other off-normal storms are relatively minor compared to the tornado forces, even when considering the higher saf factors required for structural resistance to normal and off-normal loads than for accidental tornado loading.
The region of the Calvert Cliffs plant is subject to tornadoes. The tornado design parameters used are the most severe for the continental United States .
These provide for 360 mph maximum wind speed, 290 mph maximum rotational 2-20
e e speed, 70 mph maximum translational speed _(with a 150-foot radius to maximum),
maximum pressure drop of 30 psi occurring at a maximum rate of. 2.0' psi per second, and minimum translational speed of 5 mph (to establish a maximum transit time). The values meet the most severe NRC-tornado requirements (Reference 25) and envelope the parameters for the site included in the FSAR.
The NRC staff considers that the ISFSI design criteria for wind forces to be acceptable.
Tornado Missiles Potential tornado generated missiles used for design criteria for the ISFSI are the most severe of those included in NUREG-0800 Section 3.5.1.4 (Reference 26). These envelope missiles that might be generated by the Calvert Cliffs site tornado identified in the FSAR. The staff considers that th" tornado missiles parameters used for design are acceptable.
Flood The Calvert Cliffs ISFSI site is not susceptible to flooding due to streams or runoff. The site is at an elevation of 114 feet. Maximum wave runup from the Chesapeake from all causes is to an elevation of 28 feet. This is the result of maximum waves with a maximum storm surge and normal high tides. Lower elevations would result from tidal surges, and the site was determined to not be susceptible to tsunamis. The site is therefore not considered to subject to flooding from any sources. The staff concurs with this assessment.
Earthauake The NUHOMS-24P system was designed for a maximum horizontal ground acceleration component of 0.25g and vertical component of 0.17 g. The seismic spectra were based on those of Regulatory Guide 1.60 (Reference 27). These-j values exceed the earthquake parameters (0.15 horizontal and 0.10g vertical) for the Calvert Cliffs site presented in the FSAR.
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Damping factors of 3% for the DSC and 7% for-the reinforced concrete-HSM were used. These are derived from Regulatory Guide 1.61 (Reference 28). Use of these resulted in design equivalent static forces of 1.0g horizontal and 0.689 vertical for the DSC and 0.329 horizontal and 0.21g vertical for the HSM. The l staff considers the seismic criteria used for the design and accident analyses to be acceptable.
2.2.1.2 Site Work The foundation analysis to provide for HSM foundation design was based on 12 borings in the ISFSI vicinity (and experience gained in design and construction of the Calvert Cliffs plant). Analysis determined that liquefaction of the soils is not a threat.
The HSMs are to be placed on a site which is a combination of cut and fill.
The HSM will be a minimum of 60 feet from any cut or fill slope. There is considered to be no threat to the HSM from slope failure. The HSH will be founded on compacted in-situ soil or soil fill from adjacent areas of the site.
2.2.2 Criticality Evaluation The criticality safety analysis presented in the SAR was reviewed to determine if the Calvert Cliffs ISFSI is designed to be subcritical and to prevent a nuclear criticality accident. The SAR establishes that the Calvert Cliffs site specific NUHOMS-24P system is designed to be maintained subcritical under both normal and accident conditions. Although the Calvert Cliffs CE 14x14 fuel was determined to be less reactive than the generic design basis fuel, the license applicant electad to perform a site-specific ~ criticality safety analysis. The methodology used in the site-specific cr_iticality safety analysis was similar to that used in the TR with some differences as discussed below.
Criticality safety is assured in the Calvert-Cliffs ISFSI by a combination of .;
design features for the site specific NUHOMS-24P design and by administrative 2-22 l
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e a controls for fuel assembly identification, verification, and handling as described in the SAR. An essential criticality. safety control consideration in the design is the exclusion of the possibility of introducing'unborated water in to the DSC cavity during dry storage operations. This is accomplished by a proprietary design of the DSC and its sealing operation.
Administrative controls restrict fuel assemblies for storage to initial enrichment and burnup criteria which provide an additional ~ criticality safety margin.
2.2.2.1 Description of Review The criticality analysis presented in the SAR uses an identical methodology to that described in the TR with the following differences:
- 1. The design parameters for the criticality safety analysis are based on CE 14x14 fuel with a maximum initial enrichment of 4.5 wt.% "SU.
- 2. The ROCS computer code was used to generate the irradiated fuel actinide inventories over burnup and enrichment ranges of interest.
- 3. The Criticality Safety Analysis Sequence No. 4 (CSAS4) included in the SCALE-3 nuclear criticality safety code system was used in calculating reactivities. The CSAS4 includes the Kr10-Va critical.ity code.
- 4. The analysis of the equivalent initial "50 enrichment included credit for stable fission product absorbers.
The criticality analysis in the SAR was reviewed to ensure that the Calvert Cliffs ISFSI is designed to be subcritical at all times during the operations and storage. Since the site-specific criticality analysis utilized substantially the ~same methodology as used in the TP. for the generic NUHOMS-24P system, independent verification calculations were not performed. The criticality analysis, the verification method presented in the SAR, and the response to questions were reviewed.
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Key factors and assumptions in the criticality safety analysis were:
Assumotions:
- 2. No possibility exists for the DSC to be flooded with unborated water or borated water with less that 1800 ppm boron.
- 3. No accident can occur which could alter the mechanical configuration of the stored array of irradiated fuel assemblies.
Factors:
- 1. The maximum initial fuel enrichment of Calvert Cliffs fuel assemblies is 4.5 wt.% '"U.
- 3. Only irradiated fuel assemblies with an initial enrichment equivalent 5, 1.8 wt.% 2nU are acceptable for storage in the Calvert Cliffs ISFSI.
2.2.2.2 Discussion of Results The criticality safety analysis presented in the SAR was performed using the Criticality Safety Analysis Sequence No. 4 (CSAS4) included in the SCALE-3 nuclear criticality safety system. The criticality safety analysis determined that the maximum reactivity for a Calvert Cliffs NUHOMS-24P DSC fully loaded-with 24 unirradiated assemblies of maximum equivalent initial enrichment of 1.8 wt.% 2"U is 0.93144 at optimum moderator density conditions in pure unborated water including all applicable uncertainties and biases. Since flooding of the DSC with unborated water during unmonitored dry storage operations is considered highly unlikely, this worst case reactivity satisfies the acceptance criteria of k,,,fn.95 under these conditions.
A criticality safety analysis of irradiated fuel assemblies was used to determine the burnup required as a function of initial enrichment to give an equivalent initial enrichment of 1.8 wt.% 2n U.
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The results of this analysis are given in Table 10.3-2 and Figure 3.3-1 of the SAR. Only fuel assemblies satisfying this burnup criteria are acceptable'for-storage in the Calvert Cliffs ISFSI.
The accidental misloading of 2' maxi.num enrichment (4.5 wt.% 235 U) fuel assemblies in a DSC with water containing 1800 ppm boron at optimum moderator density and including all geometrical and mechanical uncertainties and biases '
was determined to be 0.94481.
Additional discussion on the criteria and design acceptability for the ._
accidental misloading condition is required because BG&E has request i that there be no time limit associated with drain-down of the borated water after DSC is removed from the spent fuel pool and moved into the decon pit.
The NRC staff has discussed the acceptance of the genet ic NUHOMS-24P design in reference 4 and of a similar license application in reference 6. For the accidental misloading of 24 unirradiated fuel assemblies and optimum moderator i density at Oconee, k,,,50.98. However, the staff required that a limiting condition of operation include a 50 hour5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> DSC drain-down time limit, in order to preclude the possibility of realizing boiling and therefore optimal moderator density.
The maximum k,,, for the Calvert Cliffs DSC with 24 unirradiated fuel 235 assemblies (enrichment of 4.5 wt.% 0) and water containing 1800 ppm boron ,
at optimum moderation is 0.94481, including the effects of calculational, geometrical, and mechanical uncertainty. This is less than the k,,, criteria of 0.95 that-is applied to situations without extra administrative controls of a drain-down time limit. Because the Calvert Cliffs system has a k,,, of less than 0.95 for this accident situation, no drain-down time limits are required. '
2.2.2.3 Findings and Conclusions l
Based on the review of the site-specific-nuclear criticality safety analysis of the Calvert Cliffs ISFSI, the staff concludes that criticality safety is assured under all credible normal and accident conditions.
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l 2.2.3 Structures and Equipment important to Safety I
introduction Section 2 of this SER discusses all aspects of the structural and mechanical safety evaluation of the Calvert Cliffs ISFSI system. As discussed in Section 2.1, the design criteria are an integral part of the structural and mechanical evaluation because 10 CFR 72.120 through 72.130 does not explicitly-state what criteria must be used. The staff summary and conclusions are therefore presented in terms of: (1) criteria suitability and any restricting conditions as they might apply to an applicant, and also (2) the degree to which the NUHOMS-2AP ISFSI design satisfies the criteria and any restricting conditions.
The structural and nachanical components comprising the NUHOMS-24P ISFSI include the transfer cask (TC), the dry shielded canister (DSC), and the horizontal storage module (HSM). These components have been defined by BG&E as important to safety in Table 3.4-1 of the SAR. Loading conditions for she 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 which BG&E submitted in the SAR. All calculations in all appendices, non-proprietary as well as proprietary, were reviewed by the NRC stai Descriotion of Review Section 2.2.3 evaluates the structural response of the proposed ISFSI system to loads due to normal operating conditions, off-normal conditions, accident conditions, environmental conditions, and natural phenomena. The review procedure discusses the assumed loads, the material properties, and the ASME, ACI, AISC, or ANSI code allowable stress limits. The review provides an evaluation of the design analyses supplied in or with the SAR (Reference 1) l for each of the components and systems important to safety. This SER review is for the site-specific design.
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. - . - - -_m _m___ m
Aeolicable Parts of 10 CFR Part 72 10 CFR Part 72 states the basic regulations establishing the requirements, procedures,-and criteria for the issuance of licenses to possess power reactor spent fuel and other radioactive materials ssociated with spent fuel storage in an independent spent fuel storage installation. The BG&E SAR presents a design for an ISFSI for approval at the Calvert Cliffs site.
The parts of 10 CFR Part 72 which are applicable to the review of-the NUHOMS-24P ISFSI described in this section are: 72.122(a) whit. deals with quality standards; 72.122(b) which requires that structures important to safety be protected against environmental conditions and natural phenomena, as well as appropriate combinations of effects including accident conditions; 72.122(c) which requires protection against fire and explosions; 72.122(f) which requires design to permit inspection, maintenance and testing; 72.122(g) which requires design for emergencies; 72.122(h) which requires protection of the fuel cladding against degradation and gross rupture; and 72.122(1) which requires ready retrievability of the spent fuel.
Review Procedure The SAR was reviewed for compliance with the applicable parts of 10 CFR Part 72 as outlined above. 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 2.1), the structural-evaluation of the actual design against the stated and other appropriate criteria (SER Section 2.2.3), and other evaluations (SER Sections 3 through 7).
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o 4 Desian Descriptions i A description of the proposed Caivert Cliffs ISFSI is included in Section 1 of this SER. More detalied descriptior; 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 by BG&E, The SER is based on the formal description in the SAR and not on the descriptions as summarized or extracted in the SER.
Materials ^
The mechanical properties of all materials used in the fabrication of components important to safety are listed, identified, and/or included by '
reference in the SAR, Paragraph 8.1.1. This paragraphs references the TR for-the NUHOMS-24. The sources identified in this paragraph for properties of steel are the ASME Boiler and Pressure Vessel Code, Section III-1, Appendices, Code Case N-171-14 ASTH, and Sandbook of Concrete Enoinetrj.ng n by Fintel ,
(Reference 29). The ASME Code is an acceptable standard and is in compliance with the quality requirements of 10 CFR Part 72, Subpart F. -The source r identified in TR Table 8.1-2 for the mechanical properties of concrete and i reinforcing steel is the Handbook of Concrete Enaineerino (Reference 29), a document that is not considered to constitute a standard meeting the quality requirements of 10 CFR 72.122. However, the sieff has compared the data in Table 8.1-2 of the TR with ASTM specifications for steel a -d the pertinent American Concrete Institute specifications for concrete. ce. staff concurs with the data in Tabl; G.1-2 of the TR. i The sources ide..tified in TR Table 8.1-2 for the structural properties of lead are not recognized standards that are consistent with the quality requirements of 10 CFR 72.122(a). However, the material strength properties for lead shown
- in tha TR were used in a conservative way that would not invalidate the analysis. The staff concludes that the way the data were used. meets the intent of the quality-requirements of-10 CFR 72.122(a) for material.
properties.
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The Calvert Cliffs SAR presents many construction drawings and specifications.
The staff review is based on these drawings and all the supporting design calculation packages which were submitted by BG&E for the Calvert Cliffs site.
In cases where actual construction drawings were not submitted, there was adequate documentation to successfully complete the SAR review.
4 There are three basic structural and mechanical components identified as important to safety which are reviewed in this chapter. They are: (1) the horizontal storage module: (2) the dry storage canister; and (3) the_ transfer cask. BG&E cites 10 CFR 72.4 as a source of definitions for "Important to Safety." In fx? 10 CFR 72.3 defines structures, systems, and components important to safety which have features that: " (1) maintain the conditions.
required to store spent fuel or 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 hign-level radioactive waste can be received, handled, packaged, stored, and retrieved without undue risk to the health and safety of the public."
Furthermore, BG&E cites 10 CFR 50.49 which discusses " Safety Related" equipment which " ensures (1) the integrity of the reactor coolant pressure ,
boundary, (2) the capability to shutdown the reactor and maintain it in a safe shutdown condition, and (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to 10 CFR Part 100." Broad interpretation of 10 CFR 72.3 results in a list of items and components important to safety-as given below:-
- 1. The structural concrete of the HSH is "important to safety" (provides radiation shielding and protects OSCs from dam:ge) ,
[ definitions 1and2]. ,
- 2. The DSC is "important to safety" (provides secondary confinement boundary). [ definition 1). It could also be considered " safety related" (by providing criticality control it could prevent an accident comparable to 10 CFR Part 100) [defi .ition 3].
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-.r . -- -----,-.wvrvnv--- .m- .*er <-v,v+ --w- w--ym-e-
3.
The TC is "important to safety" (prevents and controls criticality provides radiation shielding during transport, and prevents radioactive releases) (definitions 1, 2, and 3]. The TC is also
" safety related" because it is the primary load path for the fuel during transfer operations. The drop of a loaded TC in the auxiliary building would have to be analyzed under the requirements of the 10 CFR Part 50 reactor operating licenses.
The above items are defined in Table 3.4-1 of the SAR.
Evaluation of Ferritic Steels Aaainst Brittle Fracture Of the three components classified as important to safety, only the TC design calls for the use of ferritic steel. The upper trunnion specifies the use of SA-564 Grade 630 ferritic steel as well as SA-533 Grade B, Class 2 ferritic steel. There h oe been two previous ISFSI cases which discussed .pecial considerations necessary for the use of ferritic steels at low temperatures where the possibility of brittle fracture exists (References 30 and 31). The NRC has published several documents which provide guidance for protecting against brittle fracture in ferritic steel shipping containers. These documents include: NUREG/CR-1815, Draft Regulatory Guide DG-7001, NUREG/CR-3019, Regulatory Guide 7.8, and Regulatory Guide 7.6 (References 32-36). The NRC staff recognizes that the above references were developed specifically for 10 CFR Part 71 applications. The staff considers that these guidelines are conservative and suitable for 10 CFR Part 72 requirements for secondary confinement boundaries primarily because they offer the most comprehensive guidelines for evaluating the design against the possibility of brittle fractures due to impact at low temperature.
In the NUTECH design, the TC is not the secondary confinement vessel. For the TC, the staff considers that the guidance given in ANSI N14.6 paragraph 4.2.6 is suitable (Reference 20). Indeed, NUTECH cites ANSI N14.6 as the design code for the lifting portions of the TC. Since NUTECH did not design according to paragraph 4.2.6 of ANSI N14.6, there are two post,1ble limiting cases which need to be considered.
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The first case involves the use of the TC inside the spent fuel pool building, where NUREG 0612 and ANSI N14.6 requirements apply. TC lifting should be restricted to TC service temperatures (inside the spent fuel pool building) equal or greater than the non-ductility transition (NDT) temperature plus 40'F or more. The limiting temperature is taken from ANSI N14.6. If the NDT temperature is not determined by test, then the temperature limit should ha the lowest test temperature at which ductility was proved by virtue o' ine Charpy impact test plus 40'F. For example, if the lowest test temperature which proved ductile behavior were -30'F, then the minimum service temperature of the trunnion shall be +10*F. _
The second case involves the use of the TC outside the spent fuel pool area.
For this case, the lowest temperature is -3 F because the SER for BuaE's ISFSI at Calvert Cliffs cites -3*F as the lowest temperature possible at the site.
An argument can be made that because the DSC inside the TC is never lifted, transported or handled at a height greater than 80 inches outside the spent fuel pool building, that no drop at 80 inches or less will result in a rupture of the DSC cor'inement boundary. Thus, even if the TC trunnion were to fail due to brittle .racture, the drop height is inside the design envelope.1 2.2,3.1 Horizontal Storage Module A general description of the HSM is included at Section 1.5 ' of this SER.
Table 1.2 compares the HSM design as proposed for use in ths 'J vert Cliffs ISFSI with the design reviewed in the NRC evaluation of the TR (References 3 and 4).
There are significant structural differences between the HSM designs of the TR and SAR. These are such as to cause the Calvert Cliffs HSM to be essentially
'In either of the above two cases, it is important that the minimum ambient operating temperatures for handling operations be established.
Appropriate techniques for determining minimums are shown in Section 5, and should be referenced in all applicable opera'.i g procedures.
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a new design requiring full evaluation for this SER rather than permitting reference to the SER for the TR. l The HSH are essentially monolithic reinforced concrete structures. The wall and roof thickness are dictated by radiation shielding considerations. The reinforcing steel is designed to satisfy structural loads of the dead weight, tornado, and seismic forces, thermally induced stresses, and the relatively trivial other live loadings. The steel must also satisfy requirements for '
minimum steel.
2 The HSM is designed and evaluated for satisfaction of load combination criteria, as identified in Table 2.2.3-1, derived from SAR Table 8.2-11.
These load combinations are as stated in Regulatory Guide 3.60 (Reference 37) and ANSI 57.9 (Reference 38), which is incorporated into the Regulatory Guide, !
(paragraph 6.17.3.1). The load combinations incorporating tornado forces used -
in the SAR treat the tornado forces as normal or off-nsrmal wind loads. This ,
is considered very conservative. ANSI 57.9 does not identify tornado loads.
Such loads may be considered as " accident" loads, and are so treated in ACl-349 (Reference 24) combinations of load expressions (ACI 349, paragraph 9.2.1 (5)). The net effect of the load combination expression used in the BG&E SAR is to calculate a safety factor approximately 80% of what the staff would compute by treating the tornado loads as " accident" loads.
The HSH loads combinations shown in the SAR are considered to be acceptable, except that tornado missile forces are not included. These forces are of the nature of other " accident" forces and could therefore be treated by substituting the missile impact forces (with appropriate dynamic analysis) for the E, or earthquake, in load combinations 5, and 6 (per Table 2.2.3-1 of the -
SER).
Tornado missile impacts were discretely analyzed for the HSH, using methodology accepted for the TR. The SAR methodology did not include the impact forces in load combination expressions.
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Load combinations identified in the SAR for the DSC support structure are shown in Table 2.2.3-1. These load combinations are acceptable with the exception that, per ANSI 57.9 paragraph 6.17.3.2.1, in no case is shear strength to exceed 1.4 5 (strength of section based on the allowable design method). Load combination 13 (per designation in Table 2.2.3-2 of the SER) is therefore unacceptable for shear. Although this criterion is unacceptable for shear, the design can be acceptable if the 1.4 S maximum for shear (allowed for accident situations only,1.0 S is the maximum for normal and off-normal conditions) is satisfied by the actual design. This was checked by the staff review of the design, and found to be acceptable. -_
The HSH and DSC support as: e f 'v aalyses were reviewed. These reviews are summarized in Tables 2. NM : 1 sf the SER. As indicated, the calculations and the load comv. w , .s (snown in Table 2.2.3-1 and 2.2.3-2) were det::rmined to be acceptable. The review addressed all calculations included in the original SAR and ir supplemental and modifying docketed material submitted subsequently and considered as part of the SAR (Reference 1). The final design analysis calculations were determined to be acceptable. Remaining calculational errors were examined and determined not -.
to impact the safety of the design.
The maximum forces on the four major structural components of the HSM (floor slab, inner and end walls, and roof slab) are listed in SAR Tables 8.1-1 and 8.2-2 for normal, of f-normal, and accident loadings. These data were checked by the staff and are shown to be satisfactory on Table 2.2.3-5 of the SER.
Tornado missile loadings in the SAR are treated the same as in the TR and are acceptable. This is despite the missile forces not being included in the load combinations. If these were included in load combinations, as by substituting for the seismic forces in load combinations 5 and 6, some negative margins of safety (MOS) would be calculated. These could occur for bending in the exterior end wall and the roof slab upon impact. This would be if the method of combination were as used in the SAR. The SAR is very conservative in that in combining forces, all forces are assumed to be positive and additive regardless of point and direction of occurrence in the structural component.
2-33
. e The staff does not consider that this method of load combination is necessary for the monolithic HSM since: (1) multiple concurrent missile strikes need not be assumed, and (2) the analysis of resistance capability does not include the capability of adjacent members to assume load on any initiation of yield in a single wall or roof panel. As a result, the treatment of the tornado missile forces is considered to be acceptable.
As calculated by the staff, marginal (negative but rounded to zero) and negative margins of safety may occur under seismic loadings for the floor slab and inner wall. The inner wall Margin of Safety of -0.1 would be unacceptable without further analysis. Staff analysis indicates that, for this situation-and design, the slightly negative Margin of Safety for the inner wall would not result in a nuclear safety hazard. This is based on the monolithic nature of the HSM and the reserve capacities (positive Margin of Safety) of the adjacent external walls (and of the abutting interior walls, at 90', which would not be concurrently loaded to the same force levels). As a result of the above considerations, the HSH structural design is considered to be acceptable.
The design analysis approach to the OSC support structure was reviewed and determined to be acceptable. The staff considers that the distribution of the DSC weight as a uniform load on the longitudinal rails (W8 x 40 sections) to be unrealistic but conservative. The stiffness of the DSC as a beam is considered to be greater than that of the rails, with the results that the OSC dead load would essentially be imposed on the rails at the points.where they
-are supported by the end bracket or transverse members. The staff independently checked the capability of the unstiffened web to transfer these loads. This was found to be satisfactory. The full design analysis of the DSC support structure was checked, including all miscellaneous steel and connections. These were determined to be satisfactory.
Tables 2.2.3-6 and 2.2.3-7 of the SER present the results of examination of the DSC support assembly stresses and load combinations. The submitted data are extracted or derived from Calculation No. 4123-04-2002.02-0001 (Refer-ence 1). This calculation and the submitted Drawing 4123-06-2002.02-0002, 2-34 i
l Sheet 2 of 3, show the support assembly rails as W8 x 40 sections and the transverse members as WB x 48 sections. This SER assumes these are the sections to be used, in contrast to the WT6 x 115 rail and W10 x 68 crossbeam identified in SAR Table 8.2-12. The assumed sections have lower but acceptable margins of safety than those in the SAR Table 8.2-12.
t Tables 2.2.3-6 and 2.2.3-7 show analyses for the three load combination !
situations shown on Table 2.2.3-2. The allowable stresses shown in the tables are those developed by the NRC staff based on the submitted calculation and the constraints of ANSI 57.9, Paragraph 6.17.3.2.1, for the load combination and shear. Although the SAR only used a factor of 1.5 S for load combination 13 (per Table 2.2.3-2), ANSI 57.9 allows a factor of 1.6 S, which is reflected under load combination 13 in the Tables. ANSI 57.9 also limits the factor for shear to 1.4 S regardless of the load combination, which is also reflected under load combination 13. The calculated maximum combined load stresses shown in the table are below the allowable stresses. The axial and bending stresses, divided by their respective allowables stresses, are further combined in order to obtain an interaction margin of safety. This combination, per the AISC Specification for Structural Steel Buildings, June 1, 1989, Paragraph HI (Reference 21) must have a value not greater than 1.0.
The stresser, on the W8 x 40 rail (Table 2.2.3-6) are based on an assumption that the DSC dead weight is uniformly distributed. As discussed above, the staff considers that, due to the relative stiffness of the DSC and the rails, the DSC will load the rail over the points at which the rail itself is supported. Web crippling calculations were examined and it was determined that the rails are satisfactory for this I2ading. Review of Table 2.2.3-6 shows that the rails are acceptable for all of the load combinations.
The stresses in the W8 x 48 transverse members are shown and evaluated in Table 2.2.3-7. No lateral forces on the members are shown.
Slotted holes in the rail support assembly are assumed to relieve any significant lateral forces which could be developed due to temperature change of the rails. Other forces (e.g., friction during DSC insertion) in the axis of the rails is 2-35
_ ~ , ._. , _ . ,
e s assumed to be transmitted into the HSH front wall by the rail support, and not into the transverse members. Review of Table 2.2.3-7 shows that the selectio of the steel section used for the transverse members is considered to be satisfactory.
Calculations for the rail-transverse member interconnection assembly, and for the assemblies supporting the ends of the transverse member and other miscellaneous HSH steel were checked and determined to be satisfactory.This includes door and supports, collars, brackets, TC restraint assembly, heat shield, seismic restraints, and end stops.
The design of the HSH roof vent covers was checked.
The covers are intended to be able to survive tornado wind and other forces. The staff considers that a cover might be damaged by a tornado missile but that such an event would not constitute a nuclear safety concern. The designs of the precast plenum cover and supporting baffle and attachments to the HSM were checked and determined to be satisfactory.
The overall result of the review of the HSM structural design criteria, load combination, and final design is that the HSH as represented in the current docketed material (Reference 1) is considered to be structurally acceptable.
2.2.3.2 Dry Shielded Canister and Internals The DSC for the Calvert Cliffs ISFSI is the secondary confinement barrier for the Calvert Cliffs spent fuel. The primary confinement barrier is considered to be the fuel cladding.
Each DSC will accommodate 24 irradiated Calvert Cliffs spent fuel assemblies. The DSC fits inside the transfer cask and is inserted into the HSH with the hydraulic ram.
The main structural parts of the DSC consist of the following stainless steel items:
! a 5/8-inch thick shell, a 1-3/4-inch thick bottom cover, a 1-1/4-inch thick top cover plate, a 1-1/2-inch thick inner top plate, nine 1-1/2-inch thick spacer discs, four 3-inch diameter spacer support rods, and twenty-four 0.105-inch thick square fuel guide sleeves.
i In addition to the above 2-36
,-,,,r.-w--,-*v" ' " ' ' ~
o .
structural items, there are two lead shield plates, and numerous small items associated with a grapple, vent and siphon system, lifting lugs, and shielding casing.
With one exception, the DSC is designed as a pressure vessel in accordance with the ASME B&pV Code Division 1 Section !!! Subsection NB-3000-1983 (Reference 19). Material qualifications are in accordance with subsection NB-2000. Fabrication and inspection are to be d:ne in accordance with Subsection NB-4000 and NB-5000, respectively. Proof pressure tests are to be carried out according to NB-6000. The exception is the weld design and inspection at the top and bottom of the DSC.
The double seal welds it the top and bottom of the DSC do not comply with the ASME Code, Section ill, Subsection NB. Neither the inspection procedures outlined by BG&E in the SAR nor the inspection procedures in the TR comply with the code. The NRC staff has determined that an exception to code requirements 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 of 1 psig. Therefore, pressure stresses are very low. ~
- 3. The test method of assuring a gas tight seal for the top welds is helium leak detection which is very sensitive. Also dye penetrant testing will be performed at two levels including the weld root pass on both seal welds to assure no weld surface imperfections. The test method of assuring a gas tight seal for the bottom welds is soap bubble film per ANSI N14.5 (Reference 39).
2-37
l DSC Normal Ooeratina Conditions The dry shielded canister was analyzed for: (1) dead weight loads, (2) design basis operating temperature loads, (3) internal pressure loads and (4) normal '
handling loads. Table 2.2.3-8 of this SER summarizes all the stress analysis results for normal cperating conditions. The summary table shows stresses for ,
each DSC component for each load condition analyzed by NUTECH 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. '
Dfad Weicht loads for DSC The dead load analysis for the DSC is presented in Section 8.1.1.1.A of the SAR. Both beam bending and shell bending were considered. The Calvert Cliffs version of the NUH0MS-24P DSC is 2880 pounds heavier than the version previously reviewed in the TR. Consequently, BG&E recalculated the stresses '
associated with the dead weight. The weights are shown in Table 8.1-1 of the SAR, and stresses are shown in Table 8.1-3 of the SAR. The NRC staff reviewed these stress levels and reports them in Table 2.2.3-8 of this SER. Basically, all stresses are lower than the ASME Code allowable stresses by a substantial margin.
Desion Basis Internal Pressure Table 8.1-2 of the SAR shows five cases for operating and accident pressures.
'The-ANSYS (Reference 40)_ finite element code was used to model the internal pressure load for the top and bottom portions of the DSC. BG&E used 1 psig for the internal pressure and then multiplied the stress results by a factor corresponding to the particular load case per TR Table 8.1-2.
The DSC has two seal welds for the pressure boundary at the top of the DSC, i.e., the weld for the outer top cover plate and an inner weld applied to the lead cover plug. The outer top cover plate is the primary structural 2-38
a e I
component and the weld at that joint is much more substantial than the weld at the inner lead cover plate. The pressure stresses in the weld of the inner and outer cover plates were evaluated for normal and accident cases and found ,
to be below the allowable limits. Additional shell stresses were evaluated i for the remainder of the DSC by an ANSYS model. The computer model used 1 psig as an internal pressure load. Thus normal, off-normal and accident pressure cases could be evaluated simply by using factors of 10 and 50, respectively. All stress intensities were evaluated and were below allowable levels for pressure stress.
Table 8.1-3 of the SAR reports the OSC pressure stresses for normal pressure -
of 10 psig. The staff reviewed these pressure stresses and concurs with them. '
The results of the SER are shown in Table 2.2.3-8 of this SER.
Desian Basis Operatina Temperature NUTECH 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, NUTECH has sized the spacer disc smaller than the inside diameter of the OSC shell to preclude induced thermal stresses. However, NUTECH did perform four different finite element analyses to determine thermal stresses for differen-tial expansion of the shell, the . spacer disc, and the shell/end cover interface. These an& lyses were performed at one ambient condition of -3'F ambient temperature. Table 8.1-13 in the SAR reports the results of the DSC temperature distribution.
The thermal stresses are always defined as "secordary stresses" by the ASME Code. This means that higher allowable stresses are permitted and 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 _-3*F, the maximum primary plus secondary stress for all thermal cases considered is 25.9 ksi for the OSC shell. The allowable stress is 56.1 ksi. The spacer disc has a thermal 2-39
.,n+, - c,,, -g, y- c -- 4gm.--y-- g-g--my-g y- , +-t m iy y 9
e b stress of 52.4 ksi, and, the allowable stress for the disc material is 60.0 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 code requirements. They are shown in Table 2.2.3-8 of the SER.
DSC Handlina Stress The OSC handling load cases were divided into three groups, each requiring different analysis techniques. The design basis handling load is 25% of the DSC dry weight applied axially as it would be during normal operations when the loading ram is used to insert or extract the DSC from the HSM. Other normal cases are dead loads applied i 1 g vertically, i 1 g horizontally, and i 1 g axially, and i 1/2 g acting in all three directions simultaneously.
These could occur during transportation 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 Code allowable. In addition to the confinement boundary, the grapple and lifting lugs were analyzed for the design basis loads, both normal and off-normal. Both components were evaluated against ASME Codes allowables-and found to be satisfactory. The resulting stresses are much lower than allowable stresses, as shown in Table 2.2.3-8.
DSC Internal Basket Analyses Section 8.1.1.3 of-the SAR discusses the stress analysis conside..tions of the basket components, i.e., the spacer disc, the 24 guide sleeves and the four 3-inch diameter support rods. The spacer disc was analyzed using n finite element program for the 75 g drop case. For the normal dead weight, the-stress levels were divided by 75. The results given in Table 8.1-3 of the SAR show stress values lower the ASME Code allowables.
Thermal stresses for the spacer disc were also analyzed and found to be less than allowable stresses. Stresses due to handling were judged to be the same as dead weight stresses, The support rods and guide sleeves were not analyzed 2-40 f
J
o o for normal operating conditions,-but only for the drop case which bounds the design. All stresses for the individual cases are reported in TE f e 2.2.3-8 of this SER.
DSC Off-Normal Events Three off-normal events were evaluated by NUTECH for the DSC. They were off-normal pressure, jammed DSC during transfer and off-normal temperature. The off-normal temperature of -3*F ambient and the jammed DSC bound the range of loads.
Jammed DSC Durina 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 80,000 pounds, before a relief valve would prevent further load. The revised calculation package BG&E 001.0203 reports the maximum off-normal jammed canister load as 80,000 pounds which corresponds to the SAR load of 80,000 pounds. For a static coefficient of friction of 1.23, the 80,000 pound. jam load provides a margin for the actual loaded weight of the Calvert Cliffs DSC (65,000 pounds). The bending stress in the bottom cover plate of the DSC is smaller. than the allowable. Also, the bending stress in the DSC sh' ell is well below the allowable stress. These results are shown in Table 2.2.3-9 of this report.
Bindina of DSC Durina Transfer "
A variation of the jammed case involves an angular misalignment of the'DSC with respect to the HSM. This condition also results in stresses lower than the allowables.
2-41 4
.--n-. - . , . , , . - - - - - - - - , , , , - . --n - . .,- . -- ~-- - - - .- - -
~ _ - . _ - _ - _ _ _ . - - . -- ._- - - - = . - - - - - - .
o o Dsc Off-Normal Thermal / Pressure Analysis The off-normal temperature range was taken as -3*F to 103 F for the DSC inside the HSM. The off-normal thermal analysis is the basis for the higher thermal stresses for the spacer disc, and the cause of higher internal pressures causing higher shell and end plate stresses. The table in the SAR which reports these stresses (Table 8.1-4) does not, in fact, show higher than normal condition thermal or pressure stresses for any component except the DSC shell for the thermal event. Pressure stresses are shown to be constant and thermal stresses for the spacer disc are also shown to be constant. (Compare SAR Tables 8.1-3 and 8.1-4.) In fact, the only difference in the SAR tables is in the off-normal handling case where higher stresses are recorded.
DSC Load combination for Normal and Off-Normal Conditions Table 3.2-Sa of the NUHOMS-24P TR outlines the different load combinations considered for normal and off-normal conditions. These conditions correspond to Service Levels A, B, C, and D of the ASME Code. Altogether, Table 3.2-Sa of the TR shows 17 combinations for all service levels. However, due to the fact that NUTECH did not present data in their TR for the off-normal thermal cases and off-normal pressure cases, the NRC staff combined load combinations A3 and A4, as well as 82 and 83 for the purposes of presenting the results shown in Table 3.2.3-10 of this SER. The staff summarized the combinations as described and finds that all stresses are below the allowables for Service Levels A and B.
DSC Accident Conditions Section 8.2 of the SAR defines the accident conditions associated with the NUHOMS system at Calvert Cliffs. The accident conditions which were examined for the DSC are: (1) earthquake, (2) accident pressure, (3) accident thermal, and (4) accidental drop of the transfer cask with the DSC inside. Of these 2-42
- ,. .. .- - . - . - - - ..--...z- .- . - - -- -- - . - _ -
\
accidents, the drop case is by far the most severe'.
The SAR classifies the thermal accidents, the pressure accident, and the drop accider.ts as Serv level D conditions and the remaining accidents as Service Level C cond .
The NRC staff concurs with this classification.
A consequence of classifying the thermal accidents as Service Level C o that the ASME Code does not require any stress analysis because of the A definition of thermal stresses as " secondary " stresses or "self-relievin stresses.
The only required consideration of the accident thermal cases was in a reduction of material properties at the higher temperature.
DSC Seismic Analysis An equivalent static load method was used to simulate seismic . The loading vertical and horizontal design loads were taken as 1.0 g and 1.5 g respectively.
Because vertical drop analyses were performed at 75 g, the results were reduced to 1 g for the seismic event.
Horizontal loads were developed by using a model of the OSC supported by one rail only insid HSM.
The allowable Codestresses values. in the shell were evaluated and found to be b rails. The DSC was also evaluated for roll-out of the support Horizontal and vertical accelerations of 0.41 g and 0.26 g were applied to the center of the OSC.
The resulting factor of safety against roll-out was 1.04 according to an NRC staff evaluation.
This corresponds to 1.05 as calculated by NUTECH (Reference 1).
The stress intensities for the 1 g vertical case were calculated by fa the dead load analysis results by 1.0.
The stress intensities for the 1.5 g horizontal case were calculated by assuming that the DSC is supporte single T-section rail inside the HSM.
Lateral bending stresses were conservatively summed absolutely with vertical bending stresses to obtain a combined stress of 17.8 ksi.
2 considered a credible hazard to the DSC during handlin 2-43
e d.
DSC Flood Condition Section 2.5.2.2 of the Calvert Cliffs FSAR (Reference 14) makes the statement that the site is not subject to flooding. The SAR for the ISFSI also states that the Calvert Cliffs site is not subject to flooding, therefore no flood conditions were postulated. ,
DSC Accident Pressure The bounding DSC internal accident pressure is 50.0 psig according to Section 8.2.9 of the SAR. The maximum ambient temperature of 103'F was assumed.
Assumptions were that all cladding failed and that 100% of the fill gas and
. 30% of the fission gas were released inside the DSC. Under these conditions, j the internal pressure could reach 50.0 psig. Table 2.2.3-11 of this SER shows )
the stress results of this case. All stress intensities are lower than the allowables.
It should be noted that NUTECH used 400*F as the appropriate temperature to select the allowable stresses for the materials in the DSC (SAR Table 8.2-9).
However, Table 8.1-13 of the TR indicates that the DSC shell reaches a maximum temperature of 460*F for this accident; therefore, the NRC staff used lower material allowable stresses for this case and all accident load combinations that have this load case as a part of the load combination.
DSC load Combination for Service level C Accident Conditions Table 8.2-9 of the SAR shows the results of one-load combination. This is the enveloping load combination defined in Table 3.2-Sa of the TR. The load combination in the TR are misleading because cases C4 and C5 are actually the same, as are cases C6 and C7. The only difference in both of these sets of cases is thermal stresses, which NUTECH did not evaluate in the TR. As may be seen from Table 2.2.3-12, all stresses are below allowable levels.
Table 2.2.3-12 in the SER indicates that the TR oversimplifies the load 4
combinations. Not all Service Level C load combination are enveloped by 2-44
a .
condition C1, as reported in the SAR. Revision 4 of the calculation package BGE001.0203 shows three load combinations, which lists combination C2 as enveloping. Also, 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 460'F, whereas, the allowables were based on 400*F. The staff has recorded the results in Table 2.2.3-12 of the SER. The conclusion is that the design for the DSC is adequate for the Calvert Cliffs site.
Accidental Dron of OSC Inside TC The DSC design for Calvert Cliffs Nuclear Power Plant has 24 CE 14x14 PWR spent fuel assemblies which are 9700 pounds lighter than the TR design; however, the dead weight of the empty DSC for Calvert Cliffs is almost 3000 pounds heavier due primarily to increased shielding in the end plugs. For the-DSC shell NOTECH performed two new analyses for the site-specific Calvert Cliffs design for the vertical drop; however, reference I states that all other drop cases the TR design bound the stresses for Calvert Cliffs. The NRC staff concurs that for the horizontal drop case this is true for the DSC shell. However, the end cover /shell juncture could be subject to different stresses because of different. spacer disc spacing and higher weights of the end shield assemblies.
A similar situation occurs for the case of the corner drop. Although the SAR analysis of a 30* corner -drop case for the transfer cask which includes upoer and lower portions of- the DSC, the results of the DSC stresses in the TR and in the SAR for the Calvert Cliffs site are not reported. Instead both Reference 1 and 3 state that the stresses for the vertical (75g) and horizontal (759) bound the 300 corner drop. See Table 3.7 of Reference 4.
For the Calvert Cliffs SER, the DSC stress results of the 30 0 corner drop are again reported for completeness.
For the spacer disc, both vertical and horizontal drop cases were calculated for the Calvert Cliffs version of the NUHOMS-24P.
I 2-45
..,-,m-~w
~
=w a - y +rw--,- v-- - - , - n-v., w-r,m.,vn-- -e-- ,n wm.
Discussion of Cask Droo 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. The basis for all drop situations that were postulated are discussed in the TR. They all involve dropping the TC, with the OSC inside, at a maximum height of 80 inches. The NRC staff considers these assumptions reasonable, because the loaded DSC will always be in the TC or inside the HSH whenever it is outside of the building which houses the spent fuel pool. The _
centerline of the HSH is located at 102 inches above the base pad and therefore the maximum drop height would be about 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, 80 inches is conservative.
Five different drop orientations were considered: (1) a horizontal drop, (2 and 3) a vertical end drop onto the top or bottom of the TC, and (4 and 5) a corner drop at an angle of 300 onto the top or bottom corner of the TC. The drop height was 80 inches for all orientations.
The magnitude of the deceleration for each case was defined in Section 3 of the TR as 75 g for either vertical or horizontal drop orientations and 25 g -
for the corner drop. The TR based these values on an EPRI report (Reference 41), which describes a method of predicting maximum decelerations of casks as a function of drop height, target hardness (i.e., hardness of concrete pad),
and cask orientation.
Because Reference 41 does not document the deceleration time history, it was necessary for the NRC staff to establish what the representative time histories and damping coefficients for the three orientations would be, in order to predict appropriate dynamic load factors (DLF), .The TR provided additional material which included references to drop test data for a 90-ton rail cask (Reference 42). The time histories from this reference were used to determine the DLFs for the different drop orientations.
2-46
a .
The NRC staff concludes that the DLFs for the vertical, horizontal, and corner drops are 1.50, 1.75, and 1.25, respectively. These factors, when multiplied by the unfactored deceleration levels obtained from reference 41, produced values of 73.5 g, 66.5 g, and 25.0 g for the three drop orientations. These values compare favorably with the deceleration values of 75 g, 75 g, and 25 g selected by NUTECH 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 NUTECH. Based on the above review, the NRC staff finds that the deceleration levels used are appropriate for the drop cases considered.
Discussion of Finite Element Models for Cask Droo in all cases, the SAR uses the ANSYS (Reference 40) finite element code to model the drop cases for the DSC and TC cask components. For the vertical drop, an axisymmetric load and an axisymmetric geometry were modeled, using an equivalent 75 g static load. For the horizontal and corner drop cases, 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.
Additional analyses for the horizontal drop orientation for the DSC confinement boundary were not performed. The NRC staff agrees that the stress analysis which was performed for the NUHOMS-24P TR for the horizontal drop case envelopes the Calvert Cliffs version of the DSC because the fuel loads are less by about 9700 pounds and there is one more spacer disc to distribute the fuel loads to the shell. Therefore, the horizontal drop stresses as calculated in Table 3.7 of the TR are conservative for this license application. These are repeated in Table 2.2.3-13 of the SER.
The Calvert Cliffs version o.' the spacer disc was analyzed for the horizontal '
drop case by utilizing a bilinear representation of material properties together with a STIF42 element.'The output stresses for this particular element are not well suited to compliance with the ASME code, which requires 2-47
primary membrane as well as membrane plus bending stress intensities, because the output combines both types of stresses at the surface. As a result, the
" stress type" as report' in the Tables of the SARs (References 1 and 3) do not accurately represent .e actual stress type for the STIF42 element.
Element type STIF63 affords this possibility.
The spacer disc was evaluated for vertical end drops in two orientations, bottom and top drop. The vertical top end drop was performed using a b:
linear elastic-plastic model with the guide sleeves providing the major portion of the load to the bottom spacer disc. The maximum G loading which the guide sleeves could exert without failure is 35 Gs, at which point the plug welds on the guide sleeve tabs will fail, allowing the guide sleeves to fall through the spacer disc square holes. The SAR uses 75 Gs for the vertical bottom end drop spacer disc analysis. STIF43 elements are used in the vertical top end analysis and STIF63 elements are used in the vertical bottom and analysis. Both types of elements have six degrees of freedom per node, and are therefore well suited to predict the required primary membrane plus primary bending stresses as required by the ASME Code. The results of these analyses are shown in Table 2.2.3-13 of this SER.
Two 75 g vertical drop analyses were performed for the DSC confinement components, i.e., the shell, bottom cover and two top covers. The SAR uses the STIF42 element. The model incorporates multiple elements through the thickness in local areas of interest, namely at the junction of the shell and the cover and at weld locations. Thus primary bending stres:;es could be estimated and combined with primarily membrane stresses.
The 25 g corner drop case was not reported in the SAR, however a DSC inside a TC for a 300 drop is analyzed. See Figures Bl/B4, B2/S4, B3/B4, and B4/84 of TC calculation package BGE 001.0202 for drawings of the models. The type of elements used to model the cask and DSC components were STIF25 with three translational and no rotational degrees of freedom per node, Only one element j was used to model the thickness of the DSC shell or DSC top and bottom covers.
The staff notes a similar shortcoming for the STIF25 element as for the STIF42 2-48 l
element for the purposes of complying with the ASME code. The NRC staff reports the results of these analyses in Table 2.2.3-13 for completeness.
Examination of Table 2.2.3-13 shows that, with the exception of the inner top cover, the vertical and horizontal drop cases cause the highest stress intensities for the confinement boundary of the DSC. It may also be seen that all stress intensities are less than the ASME code allowable stresses.
The four support rods running the length of the basket were also checked for stress as well as critical buckling during a 75 g vertical drop. For the drop accident, the SAR analysis postulates the load for each rod to be one quarter of the sum of: the dead weight of nine spacer discs, the weight of the guide sleeves, and the self weight of one rod. The primary axial stress is 27.6 ksi compared with an allowable of 28.0 ksi. Also, the critical buckling load was found to be 184 ksi, well above the actual lead. Based on the above evaluation, the NRC staff concurs that the support rod design is satisfactory.
DSC toad Combination for Service level D Accident Conditions Table 5.4 of the design calculation BGE001.0203 Revision 4 summarized the enveloping load combination stress results for the DSC accident conditions.
Table 3.2.5a of the TR defined the load cases for each load combination. The stress intensities in the DSC at various critical locations were evaluated by combining the dead load, accident pressure load, and the worst drop orientation load. Table 2.2.3-14 of this SER uses material allowables for Service Level D for the worst thermal condition reported in the TR. These allowables are somewhat lower than the TR used; however, it may be seen that even with these lower allowable stresses, the DSC components meet the ASME Code requirements.
The SAR uses Service i.evel D for accident case allowahle stresses. While the NRC staff concurs with this decision, it must be coupled with the operating controls and limits as proposed in Section 10 of the TR. Following a cask drop of 15 inches or greater, the DSC must be retrieved, and the DSC and the internals must be inspected for damage, The NRC staff set; this operational
(
2-49 l
l
a e 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 of 35 g (below the 75 g level predicted), there is justification for inspection of the DSC and internals following any cask drop of 15 inches or greater.
DSC Faticue Evt.luation Section N8-3222.4a of Section III of the ASME Code (Reference 19) requires that components be qualified for cyclic operation under Service Level A limits unless the specified service loadings of the :mponents; 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 TR (Reference 3) previously evaluated each of the six conditions defined by the ASME Code in the submittal of the TR. The NRC staff evaluated the previous analysis and concurs with the finding that the servic6 loading of the DSC meets all conditior.s. Therefore a separate analysis is not required for cyclic service. The staff does not find any basis for finding that the service loading at Calvert Cliffs Nuclear Power Plant will deviate from the conditions assumed in the SER for the TR, consequently no fatigue analysis is required.
2.2.3.3 Transfer Cask The transfer cask (TC) is used to house the DSC inside of the spent fuel pool and during transport operations between the spent fuel pool 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 tho HSM.
The main structural parts of the TC consist of the following stainless steel items a 1.5-inch thick shell, top and bottom machined rings which join-the 2-50
0 shell to a 2-inch thick bottom cover plate and a 3-inch thick top cover plate.
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 top cover plate is fixed to the top structural ring with sixteen 1.75-8 UNC bolts.
The payload of the TC is 90,000 pounds and the total gross weight with fuel and water but no top lid is 200,000 pounds enveloping and 193,000 pounds with fuel, top lid but no water.
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 111, Subsection NC for Class 2 components. The lifting and tilting trunnions have been designed to meet ANSI N14.6-1978. Load combinations have been extracted primarily from the ASME B&PV Code. Table 3.2-1 of the TR provides a complete summary of the design criteria. Material qualifications are in 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 transportation function, or lif ting /and tilting -
function. The ASME Code gcv rns for transportation, whereas ANSI N14.6 governs for the lift and tilt trunnions.
Materia' innsiderations 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. To recap briefly, the use of the TC inside the fuel pool building should be restricted to a temperature of 40 F or more higher than the NOT temperature of the trunnion material. In the event that the NDT temperature is not determined, the limiting temperature shall be 40 F higher than the lowest Charpy test 2-51
a .
I temperature at which ductile behavior is proved. Outside the fuel pool building, the lowest temperature is limited to -3 F, as cited in the Calvert 0
TC Normal Ooerittina Conditions The TC is evaltated for two dead weight loads, e.g., a fully loaded cask hanging vertically from its two lifting trunnions and a fully loaded cask supported horizontally from its trunnions at top and bottom ends of the TC on the transport , kid. Handling loads are calculated. Thermal stresses are evaluated for Service Level A and B conditions.
Deadweicht loads for the TC Review of BG&E calculation package BGE001.0202 Revision 2 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 4.1.1.1 of the aforementioned reference. 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.
Thermal loads for the TC Section 4.1.2 of the BG&E calculation package BCE001.0202 describes the th rmal analysis performed to verify that the thermal stresses in the TC are -
below allowable : stresses for Service Levels A and B. These service levels are the only ones required to evaluate accerding to the ASME Code for Class 2 components (Reference 19). Table 4.1.2 of the calculation package defines the temperature gradsents at which thermal load cases were evaluated, i.e., an 8
ambient temperature of ~70 F for normal conditions and an ambient temperature of -3"F and 103"F for off-normal conditions.
The NRC staff reviewed the computer analy;e> for the thermal case and confirmed that only one run was made for the -3': case. The position of the 2-52
0 SAR is that the -3 F ambient condition provides the worst thermal gradient and therefore the worst stress condition for all possible conditions. The NRC staff noted that the thermal stresses reported in SAR and calculation packages show identical thermal stresses for normal and off-normal. The staff checked the computer output and recorded the stresses as shown in the summary Table i 2.2.3-15 of this SER. In all cases, the staff confirmed that the thermal stresses are below allowables.
Operational handlina 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 horizantally supported by the skid. The former is governed by ANSI N14.6 rules (Reference 20) and the latter is governed by the ASME Code.
The ANSI code is concerned with critical loads and consequently only addresses the lifting trunnion design and the TC shell in the vicinity of the lifting trunnion.
The actual transfer handling cases which are considered are I g vertical,1 g horizontal, I g axial and 1/2 g applied simultaneously in all three directions.
Table 2.2.3-15 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.
TC Load Combinations for Normal and Off-normal Conditions Table 3.3. of the calculation package BGE001.0202-defines the different load combination for normal and off-normal events. These conditions correspond to Service Levels A and B of the ASME Code, There are five Level A conditions and two Level B conditions. The SAR does not present data for all the cases, so Table 2.2.3-16 of this SER has combined the conditions as follows. All five Level A conditions and both Level 8 conditions are lumped together as one single load combination.
(Tables 6.1 (p. 6.2) and 6.2 (p. 6.3) of the referenced BG&E calculation package show identical results for all seven load cases). In all cases, the allowable stresses were evaluated for a material 2-53
, m -e-- - - , ,- , ~. ,-,,s,- - , - a s .--n.- - - , e --~, m -:w-- . . , - - , . . , . - - - -- -s
e .
temperature of 400'F, a conservative value. As shown in Table 2.2.3-16, all the stresses are lower than the allowables.
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 OSC inside, and (3) tornado wind loads.
TC Seismic Condition 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 ea thquake, provided the loaded TC weighs 200 kips and experiences a horizontal acceleration not greater than 0.45 g. Since the maximum acceleration of the Calvert Cliffs cask in the washdown pit is 0.30 g (zero period acceleration), the cask should not overturn. The NRC staff calculated a safety factor of 1,06 against overturning.
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. On page 4.30 of the design ,
calculation BGE001.0202, the statement is made that the seismic stress intensities are to be taken as the transfer handling stress intensities on page 4.28. The NRC staff has included these stresses in Table 3.2.3-17. The individual stress intensities as well as the three load combination stress intensities are below ASME Code allowables.
Desian Basis Tornado Wind loads Actina on TC The SAR shows that if the height to the top of the ci-' is 130 inches, and the track of the transport vehicle is 118 inches, there is safety factor of 2.0 5
2-54
against overturning when the TC is subjected to Design Basis Tornado (OBT) winds. Shell stresses were also evaluated and found to be 3.6 ksi, well below the 23.7 ksi allowable for Service Level C. The 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.
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 ensure 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 ring and the shell. As discussed in the structural analysis of the DSC, any drop height higher than 15 inches shall require the retrieval and inspection of the DSC and its internals, in keeping with the guidelines of the ASME Code when using Service Level D allowables. Because the TC is also designed to ASME Code requirements, it will be necessary to inspect the TC as well, should it ba subjected to a drop height higher than 15 inches. Results are given in Table 2.2.3-18 of this SER. In all cases, the stresses are below the code allowables.
Tornado Generated Missiles In docketed responses to NRC staff's questions for the NUH0MS-24P TR, NUTECH presented results of an accident condition, namely design basis tornado (DBT) generated missiles. Tb .20 missiles considered are those suggested in NUREG-0800 (Reference 26), s 3697 pound automobile, and a 276 pound eight-inch
~
diameter snell. TC stabiiity, penetration resistance, and shell and end plate stresses were calculated and shown to be below the allowable stresses for Service Level D stresses. For the Calvert Cliffs site, similar analyses are
, not performed based on the argument that the overall dimensions of the Calvert l
l 2-55 I.
.. . . ~ ..
j l
Cliffs TC and the TC in the TR are very simila- The NRC staff concurs that j there .. no need to recalculate stresses for this ar.cident case because identical shell, top plate and bottom plate material and thickness were used, and the identical tornado missiles were postulated. The results.from the TR are shown for completeness in Table 2.2.3-19.
TC Load Combi.1ation for Service leve' D Accident Conditions Table 3.3 of the design calculation BGE001.0202 summarizes the load combination for the four 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. A fourth case was for design basis tornado missiles. Table 2.2.3-20 of this SER shows the results and the material allowables at 4000 F for the materials specified in the drawings. These allowables are somewhac lower than given in the SAR, but they represent the values for the specified materials for 400 F. In all cases the actual stress intensities are lower than the allowi>1es. Thus the TC meets the ASME Code for Service Level D conditions.
TC Fatique Evaluation Section C.4.2 of the TR for the NVH0MS-24P presents an evaluation of the loading cycles of the TC to show that the six criteria associated with NC-3219.2 of the ASHE Code are met. The NRC staff evaluated Section C 4.2 and concurs that all six criteria are met.
TC Trunnion loads and Stresses lhe relevant design criteria for lifting a " critical load," i.e., the spent fuel laaded in the DSC insice the TC while in the fuel building are covered by ANSI N14.6,1987 (Reference 20) and NUREG-0612 (Reference 43). Critical loads, defied 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 CFR Part 100 " In the case of the transfer cask,- the cask lifting and tilting 2-56
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 tha normal stress design factor for handling _ the critical load. Thus, the load bearing members must be sized so that yield stresses are no me 6 than one-sixth minimum tensile yield strength of the <.aterial or r . w than one-tenth the minimum ultimate tensile strength cf the materiC An additional allowance for crane hoist motion loads is recommended by NUREG-0612. Although Reference 43 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 assumption is appropriate.
Table 2.2.3-21 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 horizontal 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. The summary Table 2.2.3-21 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 transportation.
2.2.4 Evaluation of Handling and Transfer Equipment Not Important to Safety in order to support the operation of the NUHOMS-24P system at Calvert Cliffs, BG&E will supply many systems and components which interface between the existing equipment and facilities at Calvert Cliffs and the three items considered important to safety i.e., the DSC, HSH and TC. The items considered important to safety have been reviewed for compliance with 10 CFR Part 72 in Section 2.2.3 of this SER. The equipment listed in Table 2.2.4-1 is site-specific and is described in references 3 and 5.
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o
- l 2.2.4.1 TC Lifting Yoke System BG&E calculation package BGE001.0209 demonstrates the suitability of the three components required to attach the spent fuel nool crane to the lifting lugs of the TC. The lifting hook, lifting beam and hook box are only used inside the spent fuel pool building and are controlled by 10 CFR Part 50 regulations.
ANSI N14.6 was used as the design criteria. This design code requires a safety factor of 6 compared to yield stress and 10 compared to ultimate stress for combined shear stress or maximum tensile stress, because the yoke system is a single load path syc..m rather than dual load path.
A finite element analysis of the hook profile and was performed and reported.
The SAR states that the local stresses exceed the safety factor for ultimate strength. However, paragraph 4.2.1.2 of ANSI N14.6 is cited in the SAR as permitting slight local yielding in th.e material provided the designer reports and justifies exceeding the stress limit. The justification provided by NUTECH is that the area affected is small, about 10%' of the hook width, and the calculated stress is 15.5 ksi compared with an allowable of 10.5 ksi.
Although the allowable is greatly exceeded, the actual yield is 90.0 ksi.
Thus the safety factor on ultimate is not met locally, but the actual yield is not exceeded. The NRC staff finds that justification is adequate'with respect to paragraph 4.2.1.2 of ANSI N14.6.
The lifting beams were also analyzed in the SAR and found to comply with the safety factors of ANSI N14.6. No local yielding would be caused by the 200,000 pound load.
2.2.4.2 Transfer Components BG&E did not provide drawings or calculations for any of the following equipment necessary for transfer operations: transfer trailer, TC positioning skid, TC r ,iort skid, hydraulic ram and grapple, cask restraints, or optical alignment systems. BG&E did provide general design criteria and/or performance specifications. These are summarized in Table 2.2.4-1 of this SER. The staff notes that none of this equipment is important to safety, 2-58
Q. s therefore the SER review consisted of comparing design parameters of the equipment with the actual conditions which will exist at the Calvert Cliffs site. As recorded in the last column of Table 2.2.4-1, the transfer components are satisfactory.
2.2.4.3 Vacuum Drying System Section 10.3.2.1 of the SAR specifies that the vacuum drying system shall have the capability of reducing the pressure inside the DSC to s 3 torr and the time at pressure shall be 30 minutes fLilowing the stepped evacuation. This result is an oxidizing gas inventory of less than 0.25 volume %. As discussed in Section 2.2.5 of this SER, this level of oxidizing gas, when mixed with 837 120gm of helium does not provide a long-term cladding degradation mechanism for the spent fuel. Although BG&E has not submitted any specifications other than the 1 3 torr for 30 minutes, this should be satisfactory for removing oxidants from the DSC cavity. A second evacuation and second introduction of helium provide added assuranca that the oxygen is removed.
2.2.4.4 Automatic Welding Systems BG&E has not submitted any information on the automatic welding system which they intend to use for closing the top two cover plates for the DSC. However, there are two tests which must be applied to the welds prior to placing the DSC in an HSM for storage. The primary seal weld of the inner cover plate is to be helium leak checked to confirm a rate lower than 10"' atm-cc/sec. The second test which is to be applied to all seal welds for the inside and outside cover plates is a dye penetrant test in accordance with the ASME B&PV Code Section III, Subsection NB-5350. Thus, due to the testing that will be required of the welds, the NRC staff concludes that the unspecified automatic welding system will necessarily have to function well enough to pass the inspection tests.
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2.2.5 Fuel Stability 2.2.5.1 Description of Review The general design criteria for an independent spent fuel storage installation are given in Subpart F of 10 CFR 72. Section 72.122(h) covers " Confinement Barriers and Systems." Paragraph (1) of this section is pertinent to storage of spent fuel as proposed by Baltimore Gas and Electric Company. It requires 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." Other paragraphs in this Section deal with (2) underwater storage of fuel or (3) ventilation and off-gas systems and are n>t applicable to this review. Paragraphs (4) and (5) deal with monitoring and handling and retrievability operations, respectively, and are addressed elsewhere in this SER.
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.
The NRC staff review 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 Calvert Cliffs installation.
2.2.5.2 Discussion of Results In its review of the structural analysis of the DSC, the NRC staff finds the design acceptable. Confinement of radioactive material is ensured by a radiographic-inspection of the longitudinal and girth full penetration welds and the soap bubble and penetrant test of the bottom circumferential weld, and helium leak testing and dye penetrant testing of the two seal welds for the top lead plug and top plate, respectively.
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I The Topical Report for the NUHOMS-24P states that the design of the DSC and the TC are based on the B&W 15x15 fuel assembly parameters. Heat transfer, safety characteristics and radiologic neutron and gamma source terms were calculated assuming fuel with an initial enrichment of 4% and a burnup of 40,000 MWD /HTHM at a specific power of 37.5 MW/MTHM with a post discharge cooling time of ten years. The total heat rejected to the DSC and the HSM is conservatively taken to be less than or equal to 0.66 kilowatts per fuel assembly.
For Calvert Cliffs, the design is based on a loading of 24 CE 14x14 fuel assemblies per canister. The neutron and gamma source terms were calculated for a range of initial enrichments varying from 2.05% to 4.5% to determine the cooling times required to limit the decay heat to 0.66 kilowatts per fuel assembly. At this decay heat rate the largest neutron and gamma sources were found to be 3.4% initial enrichment and for a burnup of 42,000 MWD /MTHM, cooled for eight years. As for heat transfer and safety characteristics, calculations were carried out for fuel having 4.5% initial enrichment with a burnup of 45,000 MWD /MTHM. For cooling time of ten years, this fuel is below the 0.66 kilowatts per fuel assembly.
The cladding material for the B&W fuel and CE fuel is identical but the cladding thickness is slightly larger for the CE fuel. In this review of potential mechanisms for the degradation of the integrity of fuel rods, the NRC staff considered (1) the potential failure due to the diffusion-controlled gravity growth mechanism, (2) creep or sag of the fuel cladding, and (3) oxidation of the fuel during the dry-out period.
In its analysis of the cavity growth mechanism, the NRC staff determined that the area of decohesion at the end of a twenty-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 0.020 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 extension or fuel powdering is not anticipated. For all these areas of 2-61 1
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 CE fuel in the Calvert Cliffs design. The NRC staff concludes that the NUHOMS design provides sufficient means to ensure that the fuel cladding is protected against degradation, and that this conclusion also applies to Calvert Cliffs.
2.2.5.3 Findings and Conclusions The major components of the fuel design and of the NUTECH design proposed for Calvert Cliffs are within those previously analyzed for NUH0MS-24P. The previous key assumptions relating the diffusion rate of helium through the DSC, the properties of zircalloy cladding, and the temperature limits for long term and short-term protection of the cladding are within the limits used for NUHOMS-24P. Thus, the staff's conclusions relative to the generic NUHOMS-24P are:
(1) The design conforms to the applicable parts of 10 CFR 72.122(h) and; (2) The design has provided sufficient means to assure that the fuel cladding is protected against degradation are equally applicable to the Calvert Cliffs implementation of this design.
2.2.6 Thermal Evaluation p 2.2.6.1 Description of Review 2.2.6.1.1 Applicable Parts of 10 CFR Part 72 Section 72.122(h) of 10 CFR Part 72 requires that the fuel be confined such that degradation of the fuel during storage will not pose operational safety problems with respect to its removal from storage. Section 72.122(b) requires that structures, systems, and components important to safety 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.
2-62 m___-__._6.._-.im..a_m.__
e .
Section 72.126(a) of 10 CFR Part 72 states that spent fuel storage and handling systems must be designed to ensure adequate safety under normal and accident conditions. These systems must be designed with a heat removal capability having testability and reliability consistent with its importance to safety.
2.2.6.1.2 Review Procedure The Calvert Cliffs NUHOMS-24P ISFSI system is similar.to that in the NUHOMS-24P topical report and the Oconee NUHOMS-24P system with some design differences which are outlined elsewhere in this report. The only major difference which affects thermal performance is the use of a solid neutron shield in the TC for Calvert Cliffs whereas the NUHOMS-24P and original Oconee designs use ethylene glycol solutions. As in these previous NUHOMS-24P systems, the Calvert Cliffs ISFSI provides for the horizontal storage of irradiated fuel in a dry shielded stainless steel canister (DSC) which is placed inside a reinforced concrete horizontal storage module (HSM)- Decay.
heat is removed from the fuel by conduction and radiation within the DSC, which is voided of air and filled with helium gas. Heat is transferred from the surface of the DSC by convection and radiation. Finally, natural circulation of air through the HSM coupled with conduction through concrete psovide the mechanisms by which heat is removed from the HSM.
-For the purposes of thermal performance, the important components of the ISFSI-are the DSC, TC which is used to transport the DSC loaded with 24 spent 14 x 14 PWR fuel assemblies from the spent fuel pool to the ISFSI site, and the HSM into which the DSC is placed. While conduction and radiation are the predominate heat transfer mechanisms for the DSC and TC, the HSM also relies on natural convection to air. Air enters each HSH through two inlets at the bottom, flows around the DSC, and then exits through two outlets at the top of the HSM, The HSH design also includes conduction and radiation heat transfer mechanisms.
l l
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The applicant uses the same methods as in the NUHOMS-24P topical _ report for calculating temperature limits for the safe dry storage of spent-fuel which resulted in a maximum temperature limit of 335'C for normal operating conditions and 570*C for abnormal and accident conditions. This 570 C limit is based on the following fuel characteristics, which are specific to Calvert Cliffs 14 x 14 PWR fuel assemblies:
1.
Maximum burnup and associated maximum decay time of either: 50,000 MWD /MTV and 12 years; 45,000 MWD /MTV and 14 years; or 42,000 MWD /MTU and 15 years; 2.
Maximum initial fill gas pressure less than 435 psia; and 3.
Maximum individual fuel assembly decay heat of no more than 660 Watts.
The 335 C temperature limit is conservative when compared to the 340 C '
temperature limit in-the NUHOMS-24P topical report. The 570 C accident temperature limit is also conservative when compared to measured inert gas (e.g., helium) environment fuel rod failure temperatures of from 765* to 800'C .
The ISFSI thermal analysis was reviewed for completeness, applicability of the methods used, adequacy of the key assumptions, and correct application of the methods.
The principal analytical tool used by BG&E's contractor is the HEATING-6 computer code which was developed by Oak Ridgt National Laboratory and was previously used for the NUHOMS-24P topical report. HEATING-6 is the latest in a series of HEATING generalized heat transfer computer codes, which were first developed over 30 years ago. HEATING-6 is widely used in the nuclear and other industries and has been extensively verified.
HEATING-6 input and output were checked to establish that code use was appropriate and the results were reasonable. Other calculations were performed to check other manual methods used by BG&E's contractor in-calculating natural convection cooling of the HSM.
Since BG&E used identical methods to the NUHOMS-24P topical report, which has previously been reviewed j
and approved by the staff, the level of review for the thermal. analysis was 2-64 1
4
focused on those aspects of the Calvert Cliffs NUHOMS-24P design which differ from the NUHOMS-24P topical report.
2.2.6.2 Discussion Of Results The analytical methods used for the Calvert Cliffs NUHOMS-24P ISFSI thermal analysis are essentially identical to those reported in the NUH0MS-24P topical report. These methods have been reviewed and approved by the staff. The thermal analysis encompasses normal, off-normal, and postulated accident L
conditions for the TC, DSC, and HSM. The key results are the absolute values and gradients of temperatures in the spent fuel cladding and each material medium that exist between the fuel and the environment when the fuel is inside both the DSC and either the TC or the HSM.
i With one exception for the forest fire analysis which is unique to the Calvert Cliffs site, the remainder of all thermal analyses used the same combination of HEATING-6 models of the HSM and OSC-TC coupled with other manual methods for calculating HSM natural circulation cooling and Calvert Cliffs specific CE 14 x 14 PWR fuel effective thermal conductivity as in the NUHOMS-24P topical report. The aforementioned models and methods were used interactively s in the same manner as the NUHOMS-24P topical report.
A review of assumptions and model inputs for the thermal analysis revealed no -
> errors and confirmed that the selected parameter values were reasonable and I
suitably conservative. In addition, all trends and the magnitude of the results appeared to be reasonable.
For normal operating conditions, three initial conditions are assumed at -3 F, 70 F, and 103'F representing minimum, normal and maximum conditions at the HSM. The last two cases also include a solar heat load. The TC is analyzed for both the -3 F and 103 F scenario with a solar heat load imposed in the 103 F case. For all the aforementioned normal operating condition cases, the calculated concrete and fuel cladding temperatures are within their limits.
The maximum cladding temperature of 322 C at 70 F and 326 C at 103 F are bounded by their respective limits of 335 F and 570 C. Separate structural 2-65
_ - - - _ - - . , - , - - - - - - - - _ - _ - - _ - - - - - _ - - - - - - - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - -- - ~ - - ~~~
analyses which_ are discussed in further detail in Section 2.2.3 of this report conclude that all thermal induced stresses are within acceptable limits.
It should be noted, however, that, for the 103 F case with the TC, the average NS-3 solid neutron shielding material temperature of 278'F is only two degrees less than the maximum value (i.e., 280*F) at which the manufacturer performed tests on enclosed NS-3 material. These tests were designed to measure the hydrogen release due to offgassing and the resultant pressure buildup and final hydrogen content.
These factors are important because, for shielding,
[
the hydrogen content is a major contributor to reducing the neutron dose rate and the shield region internal gas pressure affects the setpoint and flow capacity of the neutron shield region safety relief valve. Any changes to the TC/DSC design, spent fuel decay heat load, or higher ambient temperature / solar heat flux could result in the neutron shield temperature exceeding its tested limit.
Based on this limiting thermal condition-, which is specific to the Calvert Cliffs TC neutron shield design, an additional limiting condition of operation (LCO) based on ambient temperature is being required for TC movement A and is discussed in greater detail in Section 5 of this report.
i Off-normal conditions considered for the thermal analyses included air temperatures of both -3 or 103 F.
Both concrete and fuel cladding off-normal condition calculated temperatures are well below their associated limiting
, values.
i Three accident scenarios were used in analyzing the thermal response of the j Calvert Cliffs ISFSI:
all HSH inlets and outlets blocked for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> with 103 F ambient air and maximum solar heat load; DSC in the transfer cask with-an internal vacuum rather than Helium; and a forest fire at the closest '
location (130 feet) to the-ISFSI.
For the.HSH inlet and outlet blocked accident, the peak fuel cladding f
temperature would be 389 C which is significantly less than the acceptance criteria of 570 C. The corresponding maximum concrete temperature is 391 F which is also acceptable for an accident case. The accident scenario with a vacuum inside the DSC results in a peak fuel cladding temperature of 393 C
- 2-66
which is again well below the limit of 570 C. A review of the calculations supporting these results found no errors.
Due to the fact that the Calvert Cliffs ISFSI site is located adjacent to a forest on its southern perimeter and to a relatively dense growth of younger trees on its western perimeter with the potential of becoming a forest during the license period, the applicant includes a forest fire analysis which calculates the thermal effects of a postulated forest fire on the ISFSI. A review of this at O ysis revealed a number of non-conservative errors in the methodology and ,cputs. Subsequently, the applicant submitted a revised analysis of the forest fire accident using acceptably conservative inputs and methodology. This revised forest fire analysis utilizes the FIRES-T3 computer code which is widely used and accepted in the fire hazards analysis industry.
A more detailed and thorough, but still conservative, analysis of the forest fire is included in the revised calculation. This calculation was reviewed and found to be acceptable. The peak calculated HSM surface concrete temperature was calculated to be approximately 152 F which is significantly less than any temperature which could compromise the structural or shielding function of the HSM.
2.2.6.3 Findings and Conclusions The Calvert Cliffs NUHOMS-24P ISFSI thermal analysis was reviewed to ensure -
that it complied with the appropriate sections of 10 CFR Part 72. This review included inputs, assumptions, methodology, and results. Emphasis was placed on features of the Calvert Cliffs design which differ from the NUHOMS-24P topical report and which impact thermal performance of this system. The review process revealed that a high ambient temperature and solar heat flux represent a limiting condition for the transfer cask solid neutron shield, thus resulting in an added LC0 on ambient daylight temperature during planned TC movement (see Section 5 of this report).
This review concludes that the Calvert Cliffs NUHOMS-24P ISFSI thermal design as it is described in the SAR and further substantiated by the applicants 2-67 1
responses to additional information is in compliance with all appropriate sections of 10 CFR Part 72.
2.2.7 Shielding Evaluation 2.2.7.1 Description of Review (Source Specification and Analyses)
The neutron and gamma ray radiation source terms were calculated for CE 14 x 14 PWR fuel assemblies in use or projected to be used at Calvert Cliffs with an enrichment of up to 4.5 wt.% 235 U and an assembly burnup of up to 47,000 MWD /MTU. The ORIGEN-2 computer code was used to calculate neutron and gamma ray source terms as a function of initial fuel assembly 235 0 enrichment, burnup, and decay time. ORIGEN-2 is a widely used and validated code which has been utilized and approved for previous ISFSI radiation source term calculations. For all possible combinations of initial enrichment, burnup and decay time, it was found that a 3.4 wt.% 23s U fuel assembly with a burnup of 42,000 MWD /MTU and a cooldown time of 8 years resulted in the most limiting neutron and gamma radiation source term. This source term, used for all shielding dose rate calculations for the ISFSI, consists of: 4.27 E+15 photons /second gamma and 2.23 E+8 neutrons /second for each fuel assembly with an energy spectrum distribution which is delineated in Table 3.1-4 of the SAR.
It should be noted that the fact that a 3.4 w/o enrichment 42,000 MWD /MTU burnup assembly with a decay time of 8 years was used for the aforementioned bounding radiation source does not imply that higher enrichment or burnup fuel was not accounted for in the ISFSI design, but only that this particular combination yielded the highest shieldina source term. Other sections of this document address such parameters as cladding thermal limits, decay heat and criticality which are also functions of these fuel parameters.
The shielding analysis of the Calvert Cliffs ISFSI utilized the same suite of computer codes as those used in the NUHOMS-24P Topical report. These computer codes are: ORIGEN-2, ANISN, QAD-CGGP, SKYSHINE-II, and MICR0 SHIELD.
Collectively, these codes were used tc calculate both the gamma and neutron direct and scattered dose rates and, as previously discussed, the radiation 2-68
source term for spent fuel assemblies. All of these computer codes have been extensively used and benchmarked throughout the nuclear industry.
2.2.7.2 Discussion of Results The staff's review of the Calvert Cliffs ISFSI shielding calculations included a combination of reviewing calculation files provided by the applicant and performing independent audit analyses. The licensee's contractor has demonstrated its proficiency in the application of these same methods and the validation and verification of these computer codes for previous ISFSI applications.
A review of some of the shielding calculation files revealed several arithmetic and/or other numerical errors. However, these mistakes were found to either be conservative (i.e., correction of the mistake would reduce the calculated dose rates) or increase dose rates by an insignificant (i.e., less than 3%) amount.
Review audit calculations of the direct shielding dose rates were performed using the LINEDOSE computer code and manual methods. LINED 0SE converts a.
cylindrical gamma source into an equivalent line source and then calculates the dose rate through shielding. Manual methods used in neutron dose calculations were based on an isotropic distribution of the ISFSI design neutron source term, neutron removal cross sections, and dose attenuation data from appropriate shielding information documents. It should.be noted that-these audit calculations are intended as a check of BG&E's submitted shielding design calculations and not as an independent reanalysis. The nature of these audit calculation methods is designed to confirm BG&E's doses by confirming-order of magnitude and should generally underpredict the dose rates calculated by the BG&E's contractor.
Table 2.2.7-1 presents a comparison of BG&E's calculated dose rates with those calculated during this review. A total of 8 cases were independently calculated for this review representing a total of 11 different neutron or gamma dose rates. In all but one dose rate (case BGERAD2 neutron dose rate),
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i p _
the audit dose rate was less than BG&E's dose rate, and even in this one case
, the difference between the two calculated dose rates was only 5% which is considered insignificant for dose rate calculations. Based on this comparison, BG&E's calculated direct neutron and gamma dose rates were found to be conservative.
Along with the previously discussed review of direct gamma and neutron doses, an independent analysis of the skyshine dose rates from the completed ISFSI facility was performed during this review. The review audit calculation of direct and skyshine dose rates from the array of HSMs proposed for Calvert Cliffs used a simplified energy balance model computer code developed by the reviewer. The key input to this model is an equivalent point source for each module surface which is derived from BG&E's calculated surface area and dose rates. A photon number weighted average energy was also determined for this 4 method. This audit analysis included air attenuation and some geometrical effects such as the self-shielding of modules which are behind the outermost j row of HSMs. The audit dose rate was calculated using a point kernel method 2
and scatter probabilities were determined from the Klein-Nishina relation.
Although not as accurate as the method used by BG&E's contractor, this review analysis was designed to determine if the results of BG&E's calculation of HSM array dose at a distance are reasonable. A comparison of the two calculations is presented in Table 2.2.7-2.
Although the review calculation dose rates out to 100 meters appear to be significantly higher than BG&E's dose rates, this difference is attributed to geometrical simplificat,ons in the audit calculation such as the use of a point source which would be expected to overestimate dose rates close in to the HSM array. In spite of these and other differences in methodology between the audit and BG&E's calculation, even the doses out to 100 meters are within the same order of magnitude. At 700 and 1000 meters, the audit results are either within 30% or underpredict the dose rate calculated by the licensee.
Based on the aforementioned comparison of calculated dose rates from the HSM between BG&E and the reviewer, the staff concludes that the HSM array dose rates at a distance are acceptable.
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2.2.7.3 Findings and Conclusions BG&E's contractor has performed an extensive number of shielding dose rate analyses for the Calvert Cliffs HSM design using a bounding neutron and gamma L ray source term which was derived from ORIGEN-2 computer code calculations of all possible combinations of Calvert Cliffs fuel initial enrichment, burnup, and decay times. Using the ANISN, QAD-CGGP, SKYSHINE-II, and MICR0 SHIELD computer codes, the contractor calculated both direct and air scattered dose rates in and around the HSM. These codes and methods have been previously
\ used by this contractor and were reviewed and approved for the NUHOMS-24P topical report.
Independent analyses using alternative methods and/or computer codes were performed by the staff as part of the review process. These audit analyses were conducted to check the dose rates reported by BG&E and are not intended to be detailed design calculations. A comparison of the audit dose rates with those presented by BG&E showed close agreement in most cases and when there 4
was not as close an agreement, BG&E's dose rates were greater than the audit results. Although a few arithmetic errors were discovered in BG&E's analysis files, these errors were either insignificant or, if corrected, would have reduced the reported dose rates.
~
Based on a detailed review of the inputs, methods, computer codes, assumptions _
and dose rate results including selected independent review analyses, the Calvert Cliffs ISFSI shielding design has been found to be acceptable and meets the requirements of 10 CFR 72.104 and 10 CFR 72.106.
2.2.8 Radiological Protection Evaluation 2.2.8.1 Description of Review Section 72.24 of 10 CFR Part 72 requires the licensee to provide the means for controlling and limiting occupational radiation exposures within the limits given in 10 CFR Part 20 (Reference 23), and for meeting the objective of maintaining exposures as low as is reasonably achievable (ALARA).
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Section 72.126(a) of 10 CFR Part 72 requires 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.
Section 20.101(a) of 10 CFR Part 20 states that any individual in a restricted area shall not receive in any period of one calendar quarter from radioactive material and other sources of radiation a total occupational dose in excess of 1.25 rems to the whole body. Part 20.101(b) states that, under certain l conditions, the quarterly dose limit to the whole body is 3 rems in any
- calendar quarter.
- . Guidance for ALARA considerations is also provided in NRC Regulatory. Guides 8.8 and 8.10, (Reference 45 and 46).
Thc design considerations which ensure that occupational exposures for the NUH0MS-24P ISFSI are ALARA are discussed in Section 7.1 of the Topical Report for that design. These are listed here along with any differences in the Calvert Cliffs implementation of the NUHOMS-24P generic design which affects the shielding design considerations are listed below. (Differences are denoted by an asterisk [*] before the number.)
- l. The design criterion for the contact dose rate on the HSM exterior surfaces away from the door or penetrations is 15 mrem /hr or less.
This is less than the design for the TR, which cited 20 mrem /hr.
- 2. Thick concrete walls and roof on the HSM to minimize the on-site and-off-site dose contribution from the ISFSI.
- 3. A lead shield plug on each end of the DSC to reduce the dose to workers performing drying and sealing operations, and during transfer and storage of the DSC in the HSM.
- 4. The ' design criterion for the nominal contact dose rate on the transfer cask =is 100 mrem /hr or less. This is less than the design
, for the TR, which was 200 mrem /hr.
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- 5. Fuel loading procedures which follow accepted practice and build on existing experience.
- 6. A recess in the HSM access opening to dock and secure the transfer cask during DSC transfer so as to reduce direct' and scattered radiation exposure.
- 8. The water used to fill the DSC cavity prior to immersion in the spent fuel pool will be borated. The shielding analyses were performed assuming that pure water was used to fill the cavity. The impact on the shielding calculation results is negligible.
- 9. Placing external shielding blocks over HSM air outlets to reduce direct and streaming radiation exposure.
- 10. Use of a passive system design for long-term storage that requires minimal maintenance.
- 11. Use of an internal shielding slab and wall around the HSM air inlet opening-to reduce direct and streaming radiation exposure.
- 12. Use of proven procedures and experience to control contamination during fuel handling and transfer operations.
- 13. Use of water in the DSC cavity during welding of the DSC top shielded end plug seal to reduce direct and scattered radiation exposure.
- 14. The water in the DSC/ cask annular gap will be drained when the water inside the DSC is drained following completion of the top shiel_d plug primary seal weld. Subsequent DSC closure operations will be performed with the DSC cavity and the annular gap dry. The 2-73
i shielding calculations were performed assuming that water would be present in the annular gap when the DSC is flooded, and that the annular gap would be drained when the DSC is drained.
- 15. Use of temporary shielding during DSC draining, drying, inerting and closure operations as necessary to further reduce the direct and scattered dose.
During the design phase of the Calvert Cliffs NUHOMS-24P ISFSI, the NUHOMS-C7P demonstration project was successfully completed at the H. B. Robinson plant.
Comparison between predicted dose rates and those measured during the first fuel load at Robinson confirmed that the ALARA design considerations employed in the NUHOMS ISFSI design are sound and effective. The NUHOMS-24P design incorporates certain improvements in the design and analysis of the radiation shielding as compared to the NUH0MS-07P system (References 3 and 7).
Furthermore, lower exposure design criteria for transfer cask and HSM average surface dose rates have been specified for the Calvert Cliffs ISFSI.
Successful demonstration of the lower design criteria for transfer cask and HSM surface dose rates all ensure that the Calvert Cliffs ISFSI shielding design and occupational radiation exposures will be ALARA.
2.2.8.2 Discussion of Results The calculated methods used in the estimation of on-site and off-site doses were reviewed in detail. The methods and results presented in the SAR and supporting documentation were reviewed for completeness, correctness, and internal consistency. The basis for assumptions and conditions of the dose calculations were reviewed.
The components of storage at the ISFSI are the Dry Shielded Canister (DSC) and the Horizontal Storage Module (HSM). The DSC design is similar to the design of the generic NUH0MS-24P DSC as described in Reference 3. However, revisions have been made to accommodate a slightly shorter fuel assembly design and an additional one-half inch of lead has been added in both shield plugs.
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From a radiological protection point of view, the HSH design is essentially identical to the design presented in Reference 3, differing only in construction details. The HSMs will be constructed in place at the ISFSI with pairs of 2x6 arrays placed end to end. Each array of 12 HMs will be constructed on a common reinforced concrete foundation slab. Three-feet thick end walls provide shielding on the sides of each HSM array. The front walls of the HSMs are thickened to three and a half feet at the access opening and vent inlet of each module. Two-feet thick interior common walls provide shielding betweer "iodules to prevent scatter in adjacent modules during DSC loading and retrieval. The roof slab for the HSH is three feet thick. An internal slab and roof caps are provided to shield the ventilation inlet and outlet openings. The shielding is designed to provide neutron and gamma shielding to achieve a nominal 15 mrem /hr contact dose rate. Nominal contact dose rates at the HSH access door and vents are designed to be less than 100 mrem /hr. The HSMs are independent, passive systems for the dry storage of irradiated fuel assemblies and are designed to ensure that normal operation and credible hazards do not impair their function.
ALARA considerations and radiation protecti've design features of the NUH0MS-24P were reviewed in Section 10.1.3 of the Safety Evaluation Report for the generic design. Changes made in the design for Calvert Cliffs implementation do not change previous evaluations in this area.
Access will be controlled through gates in a double-fence, around the ISFSI area. Normally the gates are locked; guards will be stationed when the gates are opened. Unauthorized-access will be detected by remote sensing devices.
In addition, the HSM steel doors will be tack welded after insertion of a loaded DSC and heavy equipment would be required for an unauthorized entry.
The radiological source terms were calculated using ORIGEN 2 for the range of initial enrichments and burnups of fuel to be stored, assuming cooling times for each assembly corresponding to a heat output of 0.66 kW. The fuel assembly with the largest source term was found to be a 3.4 wt.% initial enrichment, 42,000 MWD /MTU element, cooled for 8 years. All shielding analysis was carried out with this source term.
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0 9 Consideration of the possible release of airborne radioactive material indicated no possibility of significant release during drying and sealing of the DSC or during DSC transfer and storage.
Estimates of the occupational dose received during fuel loading, closure, and transfer of the DSC at the HSM are given in the SAR. The SAR also presents a graph of the enveloring dose rate versus distance from the face of the array of 120 HSMs loaded with design basis fuel. Dose to workers at the power plant due to exposure from the ISFSI is minimal. Also, for the inspection of the HSM air inlet and outlet vents, BG&E should apply its ALARA procedures.
Off-site, the maximum exposed member of the public would recieve a dose of less than 2 mrem /yr from the ISFSI and less than 13.5 mrem /yr from the remaining fuel cycle operations in the vicinity. These totals are less than the regulatory requirement of 25 mrem /yr in 10 CFR 72.104 and 40 CFR 190.
2.2.8.3 Findings and Conclusions The NRC staff has concluded that the radiological protective features of the Calvert Cliffs impicmentation of NUHOMS-24P corform to the on-site and off-site protection requirements of 10 CFR Part 72 and are acceptable for the set of conditions assumed in this review. The ISFSI design and operation procedures, including access control, are also consistent with the objective -
of maintaining occupational exposures as low as reasonably achievable. In addition, the whole body and organ dose to an individual at or beyond the controlled area boundary from the maximum credible accident are well within the 5 rem criteria specified in 10 CFR 76.106.
2.2.9 Infrastructure The BG&E administrative infrastructure associated with system design necessary to operate the BG&E Calvert Cliffs ISFSI is comprised of organization, training, and procedures. Section 10 CFR 72.4(h) requires t!it the SAR provide, "a plan for the conduct of operations, including the planned managerial and administrative controls system, and the applicant's 2-76
organization, and program for training of personnel pursuant to subpart I" .
(training and certification of personnel,10 CFR 72.190, operation requirements, 10 CFR 72.192, operation training and certification program, and 10 CFR 72.194, physical requirements).
2.2.9.1 Organization The BG&E corporate and Calvert Cliffs site organizations are addressed in SAR paragraph 9.1. These are evaluated in Section 3.0 of this SER. Since Calvert Cliffs is an operating licensed nuclear power plant, the in-place organization and the corporate organization supporting nuclear power activities are experienced with operations, procedures, management, and administration of facilities and operations which relate to nuclear safety.
BG&E proposes to provide the ISFSI operating and support organization through the extension of responsibilities within the current corporate and plant organizations. The staff considers that this is most appropriate for the situation, for the nature of the NUHOMS-24P system, and that the organization as described (by reference to the FSAR and description of responsibilities) is acceptable.
The SAR identifies the prime contractor for design, analysis, and component supply as Pacific Nuclear Fuel Services, Inc. (PNFSI, formerly Nutech Engineers, Inc.). Licensing support, geotechnical engineering, and Quality Assurance program revisions are being performed by Duke Engineering and Service, Inc., using Duke Power Company personnel experienced in the similar (not identical) Oconee Nuclear Station NUHOMS-24P ISFSI. Subsurface investigations are being performed by Law Engineering Testing Company.
The staff considers that the indicated use of external contractors and support for design, installation, and preparation of special ISFSI documentation to be appropriate. It is noted that all of the identified -firms have had experience with one or more NUHOMS installations, and that the prior NRC licensing.
actions with which they were involved have led to license approval (References 2-77 l
l 1
1 6 and 10). The staff considers that the indicated supporting firms and their roles are acceptable.
2.2.9.2 Training The SAR describes the proposed training program at paragraph 9.3. This is evaluated in Section 3.3 of this SER. ISFSI training will include general background training on the ISFSI system design and operations and detailed training appropriate to the trainee's specific responsibilities. The ISFSI training is to be implemented by extension of the current Calvert Cliffs training program. The staff considers that the proposed training as described to be appropriate and acceptable.
2.2.9.3 Procedures The SAR describes procedures in Sections 5, 9, and 10. Use and planned modification of existing procedures are summarued in Section 9. Proposed procedures are evaluated in Section 3.1 of the SER. These include procedures relating to administration, health physics, maintenance, operations, testing, pre-operational testing, and records.
The transfer of fuel from the fuel pool to HSH sequence is described in SAR Section 5. This includes a narrative description of the steps which should be a sufficient basis for preparing operational procedures. Those procedures occurring within the auxiliary building would be subject to the requirements of 10 CFR Part 50.
Quality assurance procedures are evaluated in Section 4 of this SER. Other procedures relating to ISFSI safety are to be based on or extensions of procedures used for the current nuclear plant.
! The staff considers the descriptions of procedures and the approach to providing and preparing procedures to be appropriate and satisfactory.
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) -t
J 2.2.10 Surveillr.nce and Monitoring 2.2.10.1 Description of Review The equipment, components, and parameters requiring surveillance and monitoring are discussed in Section 10 of the SAP.: (1) the potential drop height of the DSC, (2) the inlets and outlets of the HSM, (3) DSC helium backfill pressure and primary weld seal leakage rate, (4) DSC closure weld seal integrity, (5) DSC surface contamination, (6) DSC, TC, and HSM surface dose rates, (6) daylight ambient air temperature, (7) presence of fossil fuel tanker trucks on the site property, and (8) selection of acceptable fuel assemblies for insertion into the DSC.
2.2.10.2 Applicable Parts of 10 CFR Part 72 Section 72.122(a) of CFR Part 72 requires that structures, systems, and components important to safety be designed, fabricated, and tested to quality standards commensurate with the importance to safety of the function to be.
performed.
Section 72.122(h)(4) of 10 CFR Part 72 requires that storage systems have the capability of continuous' monitoring in a manner such that the licensee will be able to determine when corrective action needs to be taken to maintain safe storage conditions.
Section 72.'i (c)(1) and (2) of 10 CFR Part 72 states that effluent monitoring syttems must be provided "as appropriate for the handling and storage systern," and systems for measuring direct radiation levels must be provided in and around radioactive material storage areas.
Section 72.172 of 10 CFR Part 72 states that licensees must establish measures to ensure that failures or deficiencies are promptly reported.
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o o 2.2.10.3 Review Procedure The proposed surveillance and monitoring program for the Calvert Cliffs ISFSI 4
was reviewed for compliance with applicable sections of 10 CFR Part 72, for technical soundness and feasibility of implementation. The review procedures consists of review of descriptions and specifications, evaluation of testing, sampling, and radiation monitoring strategies, and discussions with the licensee and vendor representatives.
2.2.10.4 Design Description The Calvert Cliffs design, which is similar to the NUHOMS-24P generic design previously submitted, reviewed, and approved by the staff utilizes the OSC as the primary means of confining any postulated release of fission products or other radioactive material from the spent fuel assemblies which are contained within this structure. The HSM constitutes the structure which provides.
shielding and a natural convection air cooling pathway for decay heat removal from the DSC. The TC provides shielding and additional radioisotope confinement during the transport of the spent fuel loaded DSC from the spent fuel pool to the HSM.
The safety related design envelope for this ISFSI consists of: (1) spent fuel decay heat removal without exceeding fuel, DSC, HSM or TC temperature limits, (2) long-term confinement of the rpent fuel related radioactive source term from the environment, and (3) ALARA doses to the staff involved in all phases of loading the spent fuel into the ISFSI and monitoring its status.
To ensure complia.nce with the aforementioned safety design envelope, the surveillance and monitoring program provides the following major elements:
(1) _ Spent fuel snecifications, which determine acceptable fuel assemblies to be stored in the ISFSI; (2) Assurance of DSC confinement integrity and a non-oxidizing fuel environment by LCOs on drying vacuum pressure, helium backfill pressure, 2-80
e .
primary seal weld heliunleak rate, and closure weld dye penetrant tests; (3) Measurement of and limits on DSC surface contamination as well as DSC, TC, and HSH surface dose rates; (4) Measurement of and limits on the air temperature rise between the spent-fuel-loaded-DSC HSM inlets and outlets; (5) Daily and post-high wind accident visual inspection of the HSH air inlets and outlets; (6) Measurement of and limits on daylight ambient air temperature during loaded TC movement to the ISFSI; (7) Surveillance inspection and restriction of all fossil fuel carrying tanker trucks on the site during loaded TC movement to the ISFSI; (8) Transportation route inspection to ensure that maximum TC drop height is not exceeded; (9) Use of thermoluminescent dosimeters (TLDs) to continuously monitor radiation levels at the controlled area boundary; and (10) Measurements of boron concentration in the spent fuel pool and in fill water used in the DSC for fuel loading and unloading operations.
2.2.10.5 Discussion of Results The proposed system for monitoring and surveillance is based on considerations relating to the ability of the Calvert Cliffs specific NUHOMS-24P ISFSI design to confine fission products and other radionuclides (i.e., activation products) associated with the loading, transporting and placing of spent fuel into the HSMs. In particular, the following considerations are noteworthy:
(1) The review of the DSC design, material, manufacturing, and testing specifications indicates that failure of the closure welds or the DSC body is highly unlikely. No credible mechanisms have been identified by which gross failure of the type that result in significant radionuclide release could occur.
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o .
(2) The Calvert Cliffs design is inherently safe in that a maximum postulated radionuclide release results in doses .t the controlled area Foundary that are a small fraction of both the dose limits specified by 10 CFR 72.104(a) for normal operations and 10 CFR 72.106(b) for a postulated accident.
.ae proposed ISFSI takes advantage of experience obtained in the use of the NUHOMS-24P system, at the Oconee nuclear power station since 1990. Along with seal integrity testing of each DSC prior to loading into its HSM, the applicant has committed to measurement of air temperature rise within each HSM after it has been loaded with a OSC. In addition, the applicant has committed to performing visual inspections of all HSM air inlets and outlets every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and after any high wind speed accidents.
The use of TLDs provides a passive means of continuously monitoring radiation levels at the site area fence. Since these levels are not expected to rise during normal operations, a detectable increase would be indicative of radionuclide release. The NRC staff has determined that a network of TLDs at the ISFSI site area fence represents an adequate means of continuous monitoring for the purpose of ensuring that fuel is being maintained in a safe storage condition.
2.2.10.6 Findings and Conclusions The NRC staff has concluded that the proposed program for surveillance and monitoring for the Calvert Cliffs ISFSI system including the OSC, TC, and HSM meets the requirements of 10 CFR Part 72 and provides an adequate means of ensuring that spent fuel is maintained in a safe storage condition.
l f 2-82 l
l l
e Table 2.1.5-1 Ivaluation of Design Criteria For flormal Operating Conditions (Coluuns (1) - (5) are extracted from SAR Table 3.6-1)
Design load Design NRC Staff Cog onent Type Reference Parameters Appilcable Codes conoment s (1) (2) (3) (4) (5) (6)
HSH Dead Load IR 8.1.E.5 Dead weight including loaded ANSI 57.9-1984 Acceptable. Although DSC DSC ACI 349-85 and ACI is actually a "Ilve load' 349R-85 DSC weight is precisely known.
Load Cmelnation TR Table toad Combination Methodology ANSI 57.9-1984 Acceptable.
3.2-5 Sec. 6.11.1.1 Design Basis Operating SAR 8.1.1 DSC with spent fuel rejecting ANSI $7.9-1984 Acceptable.
Tesnperature 15.8 kW decay heat. Ambient air temerature range -3*F to 103*f.
Norinal Handling toads IR 8.1.1.4 Hydraulle ram load: 20,000 ANSI 57.9-1984 Acceptable.
y Ib. (25% loaded DSC weight) cn W Snow and Ice loads TR 3.2.4 Mantam load: 110 psf ANSI 57.9-1984 Acceptable.
(included in Ilve load)
Live loads TR 8.1.15 Design load: 200 psf ANSI 57.9-1984 Acceptable.
Shielding SAR 1.1.2 Average contact dose rate on ANSI 57.9-1984 Acceptable.
HSM enterior surface < 15 mrem /hr.
Dry Dead Loads SAR Tabl, Weight of loaded DSC: 65,000 ANSI 57.9-1984 Accepteble.
Shleided 8.1-1 lb. nominal. 80,000 lb.
Canister enveloping Design Basis Internal $AR 8.2.7.2 DSC Internal pressure 9.6 ANSI $7.9-1984 Acceptable.
Pressure toad psig
$tructural Design IR Table Service Level A and 8 ASME B&PV Code Acceptable.
3.2-6 Sec. III, Div. 1
, NB, Class !
Design Basis Operating SAR 8.1.1.1 DSC decay heat 15.8 kW. ANSI 57.9-1984 Teg erature Loads Acceptable.
Ambient air temperature -3*F t o 103*F .
Table 2.1.5-1 Evat.setics, of Design Criteria Foe moraal Operating Condittens (Cantinued)
Design toad Design Caponent Type Mct Statf Reference Parame*ers Applicable Codes Co m ts (1) (l) (3) (4I
_ - . . - (5) (6)
Dry Operattona* Handling TR 8.1.1.1 Mydraulle ram load: 20,000 ANSI 51.1-1934 Acceptable.
Shlelded Ib. enveIoping Canister (Con't.) .
OSC Criticality 1R 3.3.4 K less than 0 95 in 1800 AMSI 57.9-1984 Acceptable.
ptwo borated water DSC Dead Loads TR i 1.4 toaded OSC + self waight:
Support A4SI 57.9 1984 Acceptable.
85.000 lb. AISC Code Assen61y '
Operational fiandling TR 8.1.1.4 DSC Reaction load with ANSI ST.9-1984 Acceptable.
bydraulle ram load: 20,900 lb.
Transfer Normal Operating Condition TR Table Service level A and 8 N Cask ASME B&PY Code Acceptable.
3.2-8 Sec. 111. Olv. 1
/c NC-3700 Structure:
shell, Dead loads TR 8.1.1.9 a) Vertical erleatation, self Anst $1.3-1984 Acceptable.
Rings, weight + loaded DSC + =ater j etc. In cavity: 200,000 lb.
enveloping.
b) Horitortal erfentation, self weight + loaded Ost on j
transfer skid: 193,000 lb.
i nominal, 200,000 lb.
enveloping Snow and Ice toads TR 3.2.4 Enternal surface temperature 10 CFR 72.122 Acceptable.
of cask =lli preclude buildup of snow and Ice loads when in use: O psf Design Basis Operating 5AR 8.1.1.1 toaded DSC rejecting 15.8 kW ANSI 57.9-1984 Tewperature loads Dutside Acceptable.
decay hest. An6ient air
, Spent fuel Pool Building terperature range -37 to i 103T.
l
! =
Table 2.1.5-1 Evaluatlan of Destyi Criteria For Norum) eperating Confitions (Continem=ll
{
i Design load Design NRC Staff Cayonent Type Reference Parameters Appilcable Codes Ctmunents
! (1) (2) (3) (4) (5) (6)
I !.
l Transfer Design Basis Operating Not in SAR tower teg erature limit is AN51 N14.6 paragraph This criteria ==rst be l Cask Tegerature toads Inside NOT teap +40*F for use 4.2.6 ae %d SAR.
(cont *d.) Spert Fuel Fool Building inside spent fuel pool l building. -
- Shielding SAR F.1.2 Average contact dose rate ANSI $1.9-1984 Acceptable, j less than 100 arce/hr.
i Transfer Operational Handling TR 8.1.1.9 a) Upper lif ting trunnlans ANSI N14.6-1978 Acceptable, f i
Cask Opper while in Ausillary Bellding: e Trunnlons
- 1. TR 8.1.1.9 l) Stress m et be less 1) Paragraph Il Staff accepts 1' than yleid stress for 4.2.1.2 is justification for
. 6 times critical load cited for e=ceeding factors ,.
! of 115.000 exew tlon to l) of safety on yleid t j' m Ib./ trunnion nominal strength based on i g it) Stress must be less paragraph 4.2.1.2
'm than ultimate stress ANSI N14.6 j for 10 times critical
- icad i
l TR App. C b) Upper lifting trunnions i
for onsite transfer:
I 118,000 lb./trunnlon' ASMC 8&PV Code l 94.000 lb./ shear Sec. III. NC.
29.500 lb./trunnlon antal Class 2 i TC Operational Handling TR 8.1.1.9 tower support trunnions ASNC B&PV Code Acceptable.
Lower weight of loaded cask during Sec. III. NC.
i Trunnlons down1 ceding and transit to Class 2 H5pl ?
i i
, TC Operational Handitog TR 8.1.1.9 Hydravile ram load due to ANSI 57.9-1984 Acceptable Shell friction of entracting loaded DSC: 20.000 lb. enveloping
< TC Normal Operation- TR Table Service levels A. 8. and C ASMC B&PV Code Acceptable. i Bolts 3.2-9 Avg. stress less than 2 5,,, Section Ill. NC.
Max. stress less than 3 5,, Class 2. NC-32000 XIII-II8O '
l
]-
. ~ . . . , - . . , -
i
[
b Table 2.1.5-2 Evaluauw of Deste Celterla for Off-normal Operating Conditions f j [Coliams (1) - (5) are from SAA Table 3.6-2) !
. i
, Design load Design mRC Staff 5 Component Type Reference Parameters Applicable Codes Conuments
' (1) (2) (3) (4) (5) (6) f
(
4 H5M Of f-Normal f ewerature SAR 8.1.1.1 -3*F to 103*F adlent ANSI 51.9-1984 Acceptable. f t<wpera ture
- l i
Jensed Condition Handling TR 8.1.2.1 Hydraalle ran load equal to ANSI 57.9-1984 Acceptable. !
I00% of DSC: 80.000 1b.
naninal l i
Load Camelnation TR Table load Cod ination Methodolegy AN51 57.9-1984 Acceptable.
- 3.2-5 Sec. 6.17.1.1 '
l
! Dry Off-normal temperature SAR 8.1.1.1 -3*F to 103T an6 tent ANSI 57.9-1984 Acceptable.
Shielded [
t emper at ure !
Canister ;
Off-normal pressure SAR 8.2.7.2 DSC Internal pressure less ANSI 57.9-1984 Acceptable.
i 7
00 than 9.6 psig ,
- l Jammed Condition Handling TR 8.1.2.1 Hydraulle ran load equal to ANSI 57.9-1964 Acceptable.
80.000 lb. caelnal [
5tructural Design Of f-Normal TR Table Service Level C ASME 8 p v Code Acceptable. '
Conditons 3.2-6 Sec. Ill. Civ. 1. ,
NB. Class I !
05C Jesued Handling Condition IR 8.1.2.1 Hydraulle ram load: 80.000 ANSI 57.9-1984 Acceptable.
Support Ib. ruzelnal [
! Assen61y '
Load Co dination TR 8.2-11 Load com6ination methodology ANSI 57.9-1984 Acceptable.
i Transfer 1
- Of f-normal temperature SAR 8.1.1.1 103*F ambient tenyerature ANSI 57.9-1984 Acceptable.
j Cask' i
- Brittle fracture of ferritic not in SAA tower teg erature Ilmit is ANSI N14.6 paragraph This criteria must be
- any lift.
, e l4 Janned Condition Handling IR 8.1.2.1 Hydraulle ram load: 80.000 ANSI 51.9-1984 Acceptable.
Ib. nominal '
i !
?
i
.e '
t Table 2.1.5-2 Evoluetten of Deste Criteria for off-morumi operating Canditlems (Cantinued) t
]'
Design Load Desty mRC 51aff
. Component Type Reference Parameters Applicable Codes - Comments
{. ;
(I) (2) (3) (4) (5) (6)
Transfer- Structural Oesty Of f-pormel IR Table Service tevel C A58sE B&PV Code Accg M lte. -
l Cask ronditons 3.2.8 Sec. Ill. Olv. I, (Cont..). NC. Class 7 ...
Bolts. Off-mormel Conditions IR Table Service tevel C A58eE 8&PV Code Accev 4 %
Sec. III.'Olv. I 3.2-9 Avg. stress less than 2 5, ,
Mar. stress less than 3 5, NC. Class 2 [
. NC-3200 l ,
-i 4
. ro
. s gp f j.
i i
! I i
I f
i 5
Y
._. +m n se w - -,-->+w-, , . + e---e --2 - - , r-- n sr s m n e- , n ,ww >--r,- >= --
l i
Table 2 1.5-3 f ealuation of Design Criteria For Aaldent. Cmdttfons (Columns (1) - (5) are trum SAR Table 3.5-3 and Paragrasdi 8.2.6) i I
Design load Design MRC Staff
- Comonent Type Reference Parameters Applicable Co&s C - ts (2) (4) (5) (6)
~
(1) l (3) ,
j H5M Design Basis Tornado TR 3.2.1 Man. velocity 360 mph Regulatory Guide 1.76 Acceptable. Does not AN51 58.I 1962 . refIect "hurr!Iane Mau. =tnd pressure 304 psf oceanline" but this is co mensated for by conservative iT ortacce category.
A Load Con 6tnation IR Table toad Cmbinetton Methodology ARSI $7.9-1984 Acceptable. Use of 081 d
3.2-5 Sec. 6.11.1.1 for W ls very conservative and is acceptable.
! Design Basts Tornado TR 3.2.1.2 Man. velocity 126 aph NURIG-0800 Acceptable. Not all j Missiles Types: Sec. 3.5.1.4 MUREG-0800 alssiles but .
j Autmobile. 3,967 lb. those tesed are the most y 8 in. diam shell. 276 lb. critical for the H5M.
g i in. solid sphere i
Flood SAR 2.4.1.2 Dry site Acceptable.
i SAR 2A.I appropriate Selsele SAR 3.2.3 lbrirontal ground NRC Regulatory Guldes Acceptable.
acceleration 0.15g (both I.60 and 1.61 directions)
Vertical ground acceleration 0.10g 71 critical damping Acel &nt Condition SAR 8.2.7.2 HSM vents (lnlet/ outlet) ANSI 51-9-1984 Blockage criteria Temerature blocked for 48 hrs or less. acceptable with H5M inside '. arface te@: appropriate surseillance.
391*F . Tem erature is a 1 calculation result, not a
- critecton.
Fire SAR 8.2.10 One hour forest fire 130 feet ACI-349 for concrete Acceptable.
fran H5M tey erature 4
j .
i O
.L I.
4 Table 2.1.5-3 [valuatlan of Desty: Criteria For Accident Candittens (Continued)
Design Load Design NRC Stalf i Casvonent Type' Reference Parameters Applicable Codes Comments (1) (2) (3) (4) (5) (61 1 -
[
d
' ll5M [mplosions SAR 8.2.11 Consequence of LNG spill. NUREG-0800 Acceptable. (See 5tt .
(Continued) Plant SER causitment to Section 2.2.3 Paragra;:h 2.1.6) Subject i address 60 days prior to to re-evaluation 60 days opening of lag facility. . prior to LNG facility startup.
Lightning $AR 8.2.6 Preclude damage to ifSM or RfPA 78. Lightning Acceptable.
d contained DCS due to Protection Code lightning-j DSC Accident Drop TR 8.2.5 Equivalent static Acceptable. t deceleration:
75 g vertical end drop -'
75 g horizontal side drop 101 damping value enceeds [
25 g corner drop with slap R.G. I.61 guldance. A T1 "
doun value has been evaluated N (corresponds to an 80 inch by the staf f and has been . l
, - So drop height) accepted.
- Structural dumping during Reg. Guide 1.61 drop: 10 1
, [
4 Flood TR 3.2.2 naulassa water height: 10.CFR 72.122(b) Acceptable. (" Dry $lte' 50 ft. accepted)
Selsmic SAR 8.2.3.2 Hortrontal acceleration: 1.5g NRC Reg. Guides Acceptable. t Vertical acceleration: 1.0g 1.60 and 1.61 31 celtical damping Accident Internal Pressure SAR 8.2.7.2 OSC Internal pressure: 30.9 10 CFR 72.122(b) Acceptable.
l (HSM vents blocked) psig based on 1001 fuel clad i j rupture and fill p s release. t i and ambient air temp. =
!- 103T.
OSC shell temperature: 460*F i'
Accident Conditions TR Table se vice tevel D ASME 86PV Code . Acceptable.
3.2-6 Sec. III. Div. I j- 28. Class I ;
I'i k
- ,i
- 'f i
f.
- , . - ,. ~ _ . ., . ~ . . - . _ ..~.__4 - _ _ _ .
t:
table 2.1.5-3 Evaluatlan of Desle Criterta For Accident Canditlems (Continued) l
!' Design toad Desty j
- RC Staff .
} Cogonent Type Reference Parameters Appitcable Codes Comment s *
(1) (2) (3) (4) 15) (6) t 4
DSC Support Selsele SAR 8.2.3.2 05C reaction loads: DRC Reg. Guides Acceptable. (toads are
- Asse41y teortrontal acceleretten
- 0.6tg 1.60 and 1.61 calculation results)
Vertical acceleration: 0.399
}l . 71 critteal damylng .
[~ load Ceu6tnation IR Table Load combination methodology Amst 57.9-1984 Acceptable.
8.2-11 Sec. 6.17.3.2.1 Transfer Design Basis Tornado TR 3.2.1 Man. wind velocity: 360 mph mRC Reg. Guide Acceptable.
Cask Mau wind pressure: 397 psf 1.76.' Am51 58.1-1982 l Design 8asts Tornado TR 3.2.1 Automablie. 3967 lb. NUREG-0800 Acceptable. Missiles i
I 4
Misslies 8 In. diameter shell. 276 lb. Sec. 3.5.I.4 selected bound etfects of other missiles in NURfG - !
l Flood IR 3.2.2 Cask use to be restricted by 10 CTR 72.122 Acceptable (" Dry Site'
]
7 so administrative controls accepted) ,
C Seismic TR 3.2.3 Hortrontal ground' NRC Reg. Guides Acceptable. Emelope
- acceleration
- 0.25 g (both 1.60 and 1.61 actual site values. (
directions) o j Vertical acceleration: 0.11 g '
31 crtttcal damping '
[
Accident Drop IR 8.2.5 Equivalent static 10 CfR T2.122(b) Acceptable.
! deceleration: !
[ 75 g vertical end drop !
- j. 15 g hortrental side drop 1 j 25 g corner drop with slapdo=m 101 denying exceeds 7..G.
1 (ccrrespreds to an 80 inch I.61 guidance. however. TI
- j. drop height) has been evaluated by Structural denying during drop staff and accepted.
j 10%
i Bolts. Accident Drop TR Table Service 1.evel D ASME B&PV Code Acceptable. !
a- 3.2-9 Sec. til. Olv I I
- WC. Class 2 -
at-3200 4
. p I
T l e
. s 4 y . r* y e -+,.--- = . r , .m +---c-s - v., w r-#,-.t .= -# .en----.c-. - +-
4 .
4 i
et ^
Table 2.1.5-3 Evaluntless of Destyi Criteria For Acetdset casidtttens (tanttamed)
-, I Design toad Design IeRC Staff Caugenent Type Reference Parameters Applicable Codes Comuments
, (1) (2) (3) (4) (5) (6) ;
'< Cask 3.2-8 Sec. III. Olv I (Continued) leC. Class 2 leC-3200 .
Internal Pressure Iso t appilcable because DSC 10 CTR f2.122(b) Accepteble.
provides pressure boundary 4 r
- tightning loot (le4C 5';aff
- Should not Acceptable, based on f'
- Addressed permit demoge to DSC or separate staff analysts of '-
j affect DSC retrievability) hatard while on transit.
1 1
' N e
.. gg3 4
g t f i
I i'
i 1
.I i
b o
- l 4
t i i i- A 5
i i-
Table 2.2.3-1 Load Cm61nstions IJsed foe MSM Reinforced Concrete
[ Derived fras SAR Table 8.2-!!)
Load load Cm6tnation Description Correlation to Standards WRC Staf f Cmunents Em6.
1.2 1.4 0 + 1.7 L ANSI 57.9. Paragraph 6.lF.3.l(s) Acceptable.
3.4 0.75 (1.4 0 + 1.FL + 1.7 H + 1.1 i + 1.7 W) ANSI 57.9. Paragraph 6.lT.3.l(c) (Note: thes V+ for W]
Acceptable. Conseristive relative to ACI 349. Paragraph 9.2.I(5) which is also acceptable.
5.6 0*L+H*i+E ANSI 57.9. Paragraph 6.lF.3.l(e) Acceptable.
7 D+L+H+T. ANSI 57.9.' Paragraph 6.l?.3.1(g) Acceptable.
Where:
0= Dead Weight
- 1.05 ANSI 57.9. Paragraph 6.17.1.1 Acceptable.
N L= Live load (varied between 0-10C% for worst ANSI 57.9. Paragreph 6.17.1.1 Acceptable.
e case) ru H= lateral Scli Pressure Loads (H taken ANSI $7.9, Paragraph 6.11.1.1 Acceptable. Acceptable (encept for tornado as = 0) overturning analysis)
W= Tornado Wind loads Acceptable. (veryconservative]
i= Normal Condition Thermal Lead ANSI 57.9 Paragraph 6.17.1.1 Acceptable.
T. - Off-Normal or Accident Thermal loads ANSI 51.9. Paragraph 6.11.1.3 Acceptable.
E= Earthquake Load ANSI 51.9. Paragraph 6.11.1.2 Acceptable.
A= Accident (e g., drop accident)
=
9
Table 2.2.3-1 toad Custinations Used for HSM Reinforced Concrete (Derived from SAR Table 8.2-11] (Continued) 4 Omitted lead Ccmelnations of A8tS t 57.9 -
t s.4 9 + 1.7 L + l.T H (L.C. #2) ANSI 57.9. Paragraph 6.lF.3.l(b) Outssion acceptable (=lth Ha0 same as L.C. <
- 1) y i
1 0.751%.4 D + 1.7 L + l.T H + .
ANSI $7.9. Paragraph 6.lF.3.l(d) Ontsslog acceptable [=lth H=0 encmssed be .
1.F i) (L.C.84) L.C.#3).
D+L+H+T+A ANSI 57.9. Paragraph 6.lT.3.l(f) Ontssion not acceptable unless tornado ,
! misslie loadings and drop of HSM access door !
3 DSC Support Structure (Structural Steel). See Table are acceptably analyzed. ,
2.2 1.5-2
{
1 i
L i
. .?.
ta t i
a f k
j i t 4- !
I i
e 4
t l I i
Table 2.2.3-2 toad Codinations thd for DSC Leport Assembly (Derived frte SAft Table 8.2-12)
Load Com6ination Correlation to Standards NRC staf f Cawnts
- HL, ANSI 57.9, Paragrap% 6.lF.3.2.l(a) Acceptable.
- 12. S = DW,
- HL, ANSI 57.9, Paragraph 6.lF.3.2.l(a) Acceptable.
- 13. 1.5 5 - DW. + OW. + DBE ANSI 57.9. Paragraph 6.lT.3.2.l(e) [(e) 1.6 5 > Acceptable for spent autels and con 6tned D + L
- H + i + [ except 1.4 5 is manimure for stresses. (Use of I.5 5 in Iteu of 1.6 5 sheer] allowed, is conservative). bot acceptable for shear.
Where.
OW. = Dead Wetght of Support Asser61y Acceptable.
DW = Dead Weight of Cannister Acceptable for steel structures since all j
loads are unfectored (D5C is considered a live load by staff.]
y Ht, = Ilormal Handitng Loads - Friction e (from DSC transfer) Acceptable.
4 y HL, = Off-Normal Handling Loads -
Jammed DSC Acceptable.
DBE = Selsmic toads Acceptable.
Dettted Load Combinations of ANSI 57.9 5*D+L+H ANSI 51.9. Paragraph 6.lF.3.l(b) Oelssion acceptable [H = 0].
1.335 > D + L + H + V ANSI $7.9. Paragraph 6.17.3.l(c) Qalssion acceptable [H & W = 0].
1.55
1.75 > D + L a H + T + A ANSI 51.9. Paragraph 6.11.3.l(f) Ontssion acceptable [H. T & A = 0] .
Outssion acceptable [H & T. = 0].
l.75 > D + L + H + Ta AMSI 51.9. Paragraph 6.ll.3.l(g) s e
o Table 2.2.3-2 toad Cambinettwas lhed for 95C Sgport Asse61y (Derived from SAR Table 8.2-12]
.(Continued)
- [nolanation of Dmitted Loeds H* Lateral Earth Pressure - G. no earth contact =lth Support Assw61y i
Support Assen6ly shleided from ulnd or missile W= Wind (ortornadosortornadomissiles) loads
=0 Analysis shows negilgible temperature 7 and T. - Thennel Loads = 0 stresses. (Alloweble stresses reduced for elevated temperature).
flot subject to drop iceds apart from selsmic loads A= Accident (e.g., heavy drop) 1, d
9 4
-e 40 tit
?
4 i
i I
t a,. , p. . ..,8 - , , - w.,-.,, . . - . ~ .... , , --.. ,,
4 Table 2.2.3-3 Summary of M98 Destyi Analysis j Component Calculation Critical Load teet Staf f Cossments Ne6er/Date Combination hum 6er 1 (fram Table 2.2.3-l)
- Floor Slab 4123-04-2002.01-0001 Shear: 3 (=/o V) (O. V,) Acceptable.
- Mauent
- 3 (=/o Y) (Thermal) Acceptable.
s l Inner Wall 4123-04-2002.01-0001 Shear: 5 (Selsmic) Acceptable.
, NCIEPnt! $ ($e19mlC) ACCeptabIe.
End Wall 4123-04-2002.01-0001 Shear: 5(Sessaric) Acceptable. [ Note: Using V for tornedo causes L.C. 3 to govern)
Acceptable.
j Mmment: 7 (Thermal) i j- Roof Slab 4123-04-2002.01-0001 Shear: 7 (Themal) Acceptable.
Mmment : 7 (Thermal) Acceptable.
l 4 N .
e .I
- 50 m
j' l
-[
i ,
i 4 J
'l e
1.
.i l .
i t
i t
1 O
, , , . 4, --% .- w -
w v e- '
-m_ i *----- - _ --
o-Table 2.2.3-4 Soumery of DSC Support Assembly Design Analysis Component Calculatton Critical Load MtC Staff Ccomments Ikseer/Date Cam 6ination kaser (frrum Table 2.2.3-1)
W10 x 68 4I23-04-2002.02-0001 Shear: 13 (Setssic) . Acceptable.
Cross Beam Antal & Bending: 13 (Setsmic) Acceptable, if ccusined stresses used (A15C Spec. 9th fdttlon. Paragraph fl.
, 92 Wi6 m 115 4123-04-2002,02-0001 Shear: 13,(Setsenc) Acceptable.
Support Rail Antal & Bending: 12 (05C janumedi Acceptable.
i i.
i' y i.
4 e
-m ar
/
Iable 2.2.3-5 HSM toad Cebination Result s Section: Floor Slab Inner Wall End (Enterter) Wall Roof Slab i
Force Component: Shear M ment Shear hwu nt Shear kmt Shear M aent f
l (Units) (Kips /ft) (in-Kips /ft) (Kips /ft) (in-Klps/ft) (Ktps/ft) (in-Ktps/ft) (Kips /ft) (In-Klps/ft)
(Kips = 1000 pounds)
INTERNAL FORCES: (Intries from SAR Tables 8.1-8 and 8 2-2)
Dead Weight (D) 9.0 294 2.0 144 0.1 82 5.1 219 Live load (Inci Creep) 0.8 47 0.2 21 0.3 33.3 1.2 45 Normal Themal (i)' 2.5 1000 0.4 388 0.9 978 2.4 1138 Accident Themel (Ta) 40.8 1338 2.1 522 3.5 1311 42.5 1892 fornado Winds ("W") 3.4 144 1.1 104 4.8 129 4.9 154 Tornado Missiles ( A)8 - - - - 5.5 1882 5.5 1882
?
u) Seismic (E) 19.5 110 9.7 123 6.1 413 10 8 605 co l
Con 6tnations of the above forces per load Con 6tnations of Table 2.2.3-l*
1,2 (w/D x 1.05) 14.6 512 3.2 247 1.6 IIT 9.6 398 l 3,4 (w/0 x 1.05) 18.5 1842 4.4 167 8.4 1416 16.5 1946 5.6 (w/D x 1.05) 32.3 2066 12.4 1283 8.0 1410 19.8 201T l
Allowable for L.C.1-6 188 2047 28 1862 218 204F 188 204F Mintmum M.D.5* 4.8 0 1.3 -0.l* 25 0.4 8.5 0 Combinations of the above forces per load Con 61netton 7 of Table 2.2.3.1-1 7 (w/D x I.05) 51 1694 5.0 694 4.5 1436 49 1516 Allowable for LC 7' ' ITS 1942 26.6 1021 201 1942 178 1942 Allowable for M.0.5. 2.5 0.1 4.3 0.5 45 0.4 2.6 0.3 Staf f Ccmnents: A A A U (A*] A A A A A - Acceptable .
U - Unacceptable
_ . _ = . . . . . .
t, l- Table 2.2.3-5 HSM load Combination Results (Continued) i NOTfS:
I.
Of f-normal thermal is enveloped by accident thermal since L.C. 7 would be used for each (per ANSI $1.9. Paragraph 6.17.I.3).
- 7. Tornado miss;tes are discretely analyred in the design analysis using the methodology used in the TR and previously accepted. Missile loads
! are not included tn the load cabinations.
- 3. Dead loads are increased by 51 per ANSI 57.9. Paragraph 6.11.1.1.
- 4 Margin of Safety (MDS) = Factor of Safety -1. MOS based on the allowable forces for the section and cdttlon should not be negative.
i 5.- Staff considers this merginally unacceptable MOS to be acceptable in considering monolithte nature of the HSM. In which yleid of inner wall j would be resisted by the adjacent structural cwiponents (e.g., abutting Interior walls not concurrently loaded and enterior walls, which would have positive mergins available to assime load). As a result of the above. and not as a general policy, the negative MOS for inner wall is acceptable.-
- 6. Capacities are reduced for the thermal " accident" case for elevated temiperature of concrete and reber ($AR Table 8.1.8).
e
~
e e @
.w I
- t. ;
l' !
r
[
c' a
7 l i I i
.I -
e j.
1
, w -,~ s
+ .-e .,, , , . . . - - . ---, --- , - . - - .
- t 1
Table 2.2.3-6 DSC Support Assed ly Ratt toad C h ination Results (a) m Load Camelnations (SAR Table 8.2-12)
- 12 Off-pormal 13 Ac.cident 11 Nonnel Janned (Seismic)
Stresses / Casements 5* W.(b) + 0% + Ht, (b) 5. W.(b) + DR + Ht, (b) 1.5 $= DW,(b) + W
- Det I
j
^
Autal (ksi) (fa) 1.15 3.42
- 2.27 g ##C Allow (kst) (fa) 14.6 14.6 23.4 8endtng (kst) (fb) 6.08 11.15 14.84 NRC Allow (kst) (fs) 17.56 17.56 28.1 i
. C e lned Antal (fal & Sending (fb) 43 < l.0 .87 < l 0 .63 = 1.0 (AISC 8th, ed., Sect. 1.6) neC Commments Acceptable Acceptable Acceptable N
e Shear (ksi) (fv)' 4.73 T.74 11.54 i
- o t
o WRC Allow (kst) (fv) 10.6 10.6 14.8 I h8C Comments Acteptable Acceptable Acceptable ideb Cripping Acceptable Acceptable Acceptable Deflections (in) .009 .027 .C22 l Notes: a) Thermal esponsion of the sum 6ers are allowed on the slotted bolt holes. Therefore, no thenmal stresses were ronsidered in the calculation. .
, b) ' 9G&E did not consider the weight of the DSC support steel (DWs) and normal handilng was (HL.) in the calculation since the stresses are well below the allowable limits of aulal and bending stresses. The #RC staf f accepted the results.
I i
e
. , , + . . . . , . , - - . . . . , . . . . .
o Table 2.2.3-1 05C Support Assenely Transverse M r Load Combination Results load Ctr61 nations (SAR Table 8.2-12) 12 Off-Normal (a) 13 Accident Stresses /Cors=ents Il Normal (a) Jamed (SelS"tcI
% DW,(c) + 04 + Ht, (c) $= OW,(c) + OR + Hl, (c) 1.5 5
- DV,(c)
- DR + DBE Antal (kst) (fa) .9 .8 1.9 MRC Allow (ksi) (fa) 15 5 15 5 24.8 Bending (kst) (fb ) 8.72 4.06 17.09 NRC Allow (kst) (fb.) 11.56 17.56 28.I Con 6tned Asial (fa) & Bending (fb) .55 < I . 0 .28 < l.0 .68 < l.0 g (AISC 8th, ed., Sect. l.6) '
NDC Co m ts Acceptable Acceptable Acceptable N
e Shear (kst) (fv) (a) 1.67 1.42 3.99 NRC Allow (kst) (f.) (a) 10.6 10.6 15.9 Web Cripping (a) Acceptable by engineering (b) Acceptable by engineering Acceptable ju&rment judywent Deflections (in) .036 .014 .085 Notes: a) Results of the NRC staff calculations. 86&E did not determine the results in Calculation No. 4123-04-2002.02.
b) Thermal stresses are considered " secondary stresses." and are self-Ilmiting. Therefore, they were not considered in the con 6tnation.
c) BG&E did not consider the weight of the D5C support. transverse steel (DWs) and normal handling load (HL,) in the calculation siv re the stresses are well below the allowable lletts. The NRC staf f accepted the results.
Tabte 2.2.3-8 OSC 5 tress Analysis Resulta for Normal tomds Senice level A Stress (kst) 10.0 pstg -3*F Nonnel Allowable
- DSC Stress Handling level A & 8
( C-4+ ent Type Dead Weight int. Pressure thennal NVitCH W3C NUTECH PRC WUTECH h2C NUTECH NEC I 0.3 0.5 N/A N/A 1.0 1.0 18.7 DSC Shell Pri Mem6 0.1 0.2 l
4.9 2.4 2.8 N/A N/A I .0
- 1.0 28.0 I Memb + Bend 4.9 4.9 13.1 13.1 25.9 25.9 1.0 .l . 0 56.1
{ Fri
- Second 4.9 0.3 0.5 N/A N/A I.0 1.0 18.7 Outer Prl Memb 0.1 0.2 0.4 5.5 5.4 N/A N/A 1.0 1.0 28.0 Top Mem6 + Bend 0.2 56.1 l
0.3 0.2 13.1 13.1 11.9 11.9 1.0 1.0 l
Cover Pri + 5econd 91ste N/A N/A N/A 1.0 1.0 18.T Inner Prl Men 6 0.2 0.2 N/A 0.5 N/A N/A N/A N/A 1.0 1.0 28.0 Top Men 6 + Bend 0.2 0.2 N/A N/A 13.9 13.9 1.0 1.0 56.I Cover Pri + 5econd 0.2 Plate N 1.0 18.7 0.1 0.2 0.3 0.4 W/A N/A I.0 b
O Bot taa Pri HsWh Men 6 + Bend 0.2 0.3 0.8 0.8 N/A N/A 3.2 3.5 28.0 Cover Plate 10.1 10.7 3.2 3.5 56.1
" Pri + Second 0.2 0.3 1.0 1.0 Prl Mep6 0.5 0.5 0.0 0.0 N/A N/A 0.5 0.5 18.1 Spacer 2.0 2.0 28.0 Disc Mew 6 + Bend 2.0 2.0 0.0 0.0 N/A N/A 7.5 N/A N/A 52.4 52.4 6.63 1.5 60.0**
Pri + 5econd 6.6
?
- Allowable stress for Service Levels A and B Primary Me=4ane 5m = 18.7 ksi Primary M e6 + 5end 1.5 Se = 28.0 Primary + 5econdary 3.0 5m = 55.1 Shell, D'se and end SA 204 Type 304
< plates "lor 400*F
- Allowable stress for 30G*F for spacer disc 3.0 Se = 60.0
(
i _
o t
4 - !
Table 2.2.3-9 DSC Stress Analysis Results For Off-Mormel Leeds 1
Servicar level B
- Stress (ksi) {
1' Internal ;
. DSC Stress Pressure Thermal OfI-Normel l Component' Type 10.0 psig -3*F Hand Allowable
- l fc Ntfi[CH _NE[ MtJIECH N#q WJi[CH NS{
{-
j- DSC Shell Pri Men 6 0.5 0.5 N/A N/A 1.4 1.4
- 18.7
- Memb + 8end 2.4 2.8 N/A N/A 10.0 10.0 28.0 !
l Pri + Second 13.1' 13.1 39.6 25.9 10.0 10.0 56.1 l t
Outer Pri Memb 0.3 0.5 N/A N/A 0.0 0.0 18.7 [
- iop Ibub + 8end 5.5 5.4 N/A N/A 0.0 28.0 t i Cover Prl + Second- 13.1 13.1 11.9 11.9 0.0 0.0 56.1 Plate Inner Pri Men 6 N/A N/A N/A N/A 0.0 0.0 18.7 [
, Top Memb + 8end N/A N/A N/A N/A 0.0 28.0 i Cover Prl + Second N/A N/A 13.9 13.9 0.0 0.0 56.1 !
y Plate 5
Bottae Pri Mest 0.3 0.4 N/A N/A 0.0 0.0 18.7 f L
Cover Mem6 + 8end 0.5 0.8 N/A N/A 6.5 6.5 28.0 l i Plate Pfl + Second 1.0 1.0 10.7 10.7 6.5 6.5 56.1 [
i I Spacer Prl Men 6 0.0 0.0 N/A N/A 0.5 0.5 18.7 [
! Olse Pri + Second N/A N/A 52.4 52.4 5 63 F.5 56.1 i i !
- Allowable stress is taken for Service level 8 for SA 204 Type 304 material at 400T.
I i
P I
f t
y a
i i
i !
J Table 2.2.3-10 DSC toad Ca 6lnations For Normal and Off-normal Operating Conditions Service tevels A arwl 8 4
Stress thst)
I' s OSC Stress Case' Case' Allowable
- Component Type A3 tewel A & 8 81 [
NUTECH NRC muTECH mEC DSC Shell Pri Me=6 1.4 1.7 1.6 1.9 18.7 Mene + 8end 6.7 8.7 17.3 16.2 28.0 ,
Prl + Second 46.4 52.4 53 9 52.4 56.1 l Bottom Prl Mere 1.4 1.6 0.4 1.6 18.7 }
Cover Memb + Bend 4.2 4.6 7.5 7.6 28.0 l
} Plate Prl
- Second 15.1 15.5 18.4 18.5 56.1 ,
j Outer (pressure) Prl Men 6 1.4 1.7 0.4 0.7 18.7 i L Top Memb + Bend 6.7 6.8 5.7 5.8 28.0 ;
l Cover Prl + Second 26.3 26.2 25.3 25.3 56.1 i j, Plate !
y Inner Top Prl Med 1.2 1.2 0.2 1.2 18.7
- Cover Mene + 8end 1.2 1.2 0.2 0.5 28.0 *
$ Plate Pri + Second 15.1 15.1 14.1 14.1 56.1 !
!' 5 pacer Pri Men 6 1.8 1.8 0.5 1.0 18.7 ,
Dise Mene + Bend 14.2 14.2 6.1 4.0 28.0 I Pri + 5econd 56.1 59.3 54.4 59.3 60.0
{
i Support Prl Mene 0.8 0.8 - -
18.7 i Rods Med + 8end - - - -
28.1 i Pri + 5econd - - - -
56.1 l
- Allo =eblo stress is taken for Service Levels A and 8 for SA 204 Type 304 Material at 400*F. -
- 1. Load cases A3 and A4 were cuelned into one case because the stresses for the normal and off-normal pressure cases were not suppled by MUitCH.
- 2. i Load cases 82 and 83 were combined Irto one case because the stresses for the thermal case with the DSC inside the casit or inside the H5M at i l.
T . = 103*F and FD*F were not supplied by NUT (CH. j 1
L 1 e l
t 4' t i
_ . _- - _ , _ .m_. . . __ . __
ll i
Table 2.2.3-11 DSC Stress Analysis Results For Accident ceuiltlers Service Level C' Stress (kst)
DSC Stress Accident Normal Accident Component Type Seismic Pressure 50.0 Handling Handling Allowable *
(psig)
NUTEC*l NGC NUTECH WRC NUTECH NRC NUTECH NRC DSC Shell Prl Mee 0.9 0.8 -
2.5 -
1.0 1.4
- 1.4 21.6 l Memb + 8end 17.8 17.8 -
14.0 -
1.0 10.0 10.0 32.4 .
4 Outer Pri Mee 0.8 0.8 -
2.5 -
I.0 -
0.0 21.6 Top Memb + 8end 0.8 0.8 -
21.0 -
1.0 - 0.0 32.4 i Cover Plate i Inner Prl Mee 0.4 0.4 -
N/A -
1.0 -
0.0 21.6 Top Mee + Bend 0.8 0.8 -
N/A -
1.0 -
0.0 32.4 Cover Plate N
b C
Sottom Pri Mee 0.4 0.4 -
2.0 -
1.0 -
0.0 21.6
, Cover Pl6te Me e + Bend 0.8 0.8 -
4.0 -
3.5 6.5 6.5 32.4 Spacer Prl Mee 2 0.7 -
N/A -
0.5 -
0.5 21.6 Disc Phee + Bend 3.2 3.6 -
N/A -
2.0 -
2.0 32.4
- Allowable stress for Service Level C ~'
P. larger of 1.25 m or Sy - 22.4 Pg=Pg + P, = larger of 1.8 Se or 1.5 Sy = 33.7 I. No secondary stress needs to be evaluated eccording to ASME Code for Service Level C. This includes thermal as well as secondary bending stresses for pressure cases.
4 4
9 6
t
~ _ -- . _ _ _ _ _ .. . . , _
i
=
Table 2.2.3-12 DSC load Cambinations for Accident Service level C Cases' Stress (kst)
DSC Stress Case' Case
- Case
- Case' Component Type Cl C3 C4/05 C6/07 Allowable
- MUTECH NRC NUitCH NRC NUTECH NRC WUltCH NRC l DSC Shell Prl Men 6 2.4 3.5 1.5 3.7 1.5 2.7 1.5 3.9 21.6 i Men 6 + Bend 24.5 25.4* 16.9 19.9 16.9 18.9 16.9
- 27.4 32.4 I i'
Outer Pri Memb 1.9 3.5 1.5 3.7 1.5 2.7 1.5 2.7 21.6 ;
Top Mem6 + 8end 28.4 29.2 27.6 28.4 27.6 27.4 27.6 27.4 32.4
- Cover Plate i Inner Pri Web I.0 0.6 0.2 1.2 0.2 0.2 0.2 0.2 21.6 Top Mene + 8end 1.0 1.3 0.2 1.5 0.2 0.5 0.2 0.5 32.4 j Cover i Plate
' T Y Bottone Prl Mene 2.0 2.6 1.6 3.2 1.6 2.2 1.6 2.2 21.6 5m Cover Plate New6
- Bend 5.2 5.1 4.4 7.8 4.4 4.3 4.4 10.8 32.4 '
Spacer Prl Memb 2.5 1.2 0.5 1.0 0.5 0.5 0.5 1.0 21.6 Dise Mem6 + 8end 5.2 5.6 2.0 4.0 2.0 2.0 2.0 4.0 32.4 e
i Support Prl Mem6 0.8 -
0.4 -
0.4 0.4 21.6 Reds Men 6 + 8end - - - - - - - -
31.0 i
- 1. Secondary stresses are not required for Service Level C.
- 2. Seismic stresses are considered " mechanical loeds" and must be con 61ned with W and accident pressure for Cl.
The location for mentame combined stresses due to seismic, accident pressure and dead weight occurs in the middle (length) of the shell. Thus, t the 25.4 ksi shown here is not obtatned by adding the maalansa stresses from Tables 2.2.3-4 and 2.2.3-4 for Individual load conditions where the ,
' monimum stresses occurred at different locations in the shell. I
- 3. Case C3 is W. accident pressure and normal handling.
- 4. Thermal stresses are secondary and need not be evaluated for Service Level C. Therefore. C4 and C5 are identical cases consisting of accident l pressure and W.
- 5. Because thennel stresses need not be evaluated for Service Level C cases C6 and C7 are identical cases consisting of W, accident pressure, and accident handling.
- 6. Allowables are based on a maxtuman DSC temperature of 460*F. This results in lower allowables than MUTECH reported in SAR.
= 4 h
) .
o Table 2.2.3-13 DSC Drop Accid mt Loads Service Level D Stress (kst)
DSC Stress Allowable' 0%,w.ent Type Vertical (759) Horizontal' (75g) Correr* (259)
NUTECH NPC NUT [CH NRC WRC 8.8 16.4 9.2 IT.6 14.1 43.2 DSC Shell Pri Men 6 9.3 IT.I 12.4 24.1 10.8
- 64.0 Men 6 + Bend 0 16.8 14 16.9 26.5 43.2 inner Pri Mere 13.2 17.6 15.9 18.7 21.4 64.0 Top Pe e + Bend Cover Plate Pri Mep6 0 19.5 9.5 9.5 10.2 43.2 Outer 64.0 Men 6 + Bend 13.1 22.9 14.6 21.0 14.2 Top Cover Plate 9.5 9.5 15.2 43.2 Y Bottan Prl Men 6 Men 6 + Bend 0
14.8 17.1 16.6 14.6 21.0 15.4 64.0 Cover N Plate 25.8 25.8 32.0 31.9 - 43.2 Spacer Pri Mene 43.1 43.1 34.6 34.6 - 64.0 Disc Mene + Bend Primary 27,6 - - - 27.6 28.0 Support Rods Primary 11.9 19.5 9.5 9.5 9.5 21.6*
Top End
)
Struct. Weld (shear) l Primary 11.8 16.4 9.5 9.5 9.5 21.6 j Bottan Ind l Struct. Weld (shear) l l
' ouables taken at worst cast tenperature, t.e.. for Case 01. T=460*F shell tewperature.
1.
- 2. . ese colwes for stresses for shell, top covers and bottom cover are taken fran SER for NUHOMS-24P. since no cew analysis was performed for these elements.
- Ef ficiency factor for non-volumetric inspected elds = 0.5
- 3. This colism is taken frors results obtained in res BPLSB5V and BPL5868.
1 t
Table 2.2.3-14 DSC Emeloping Lead Comelnatim Results fer Accidet Lande 5ervice level O i
Stress (ksil 1 DSC Stress Contrailing Calculated Caponent Type load Allowable
- Ccsubination NUTECM NJC ,
DSC Shell Pri Memb 02 10.7 20.3 43.2 Memb + 8end 29.3 43.6 64.0 i' inner Prl Memb D2 I4.2 26.1 43.2
- Top Men 6 + 8end
' 16.1 27.6 64.0 Cover
! Piste l
Duter Prl Mese D2 11.0 22.2 43.2 Top Mem6 + Bend 42.2 50.3 64.0 Cover
, a Plate
\
"O Bottcme Prl Me=6 02 11.1 19.3 43.2 Cover Memb + 8end 19.2 25.3 64.0
, Plate Spacer Prl Mem6 02 32.5 32.4 43.2 1
Dise Mau6 + 8end 49.2 45.1 64.0 Support Primary D2 28.0 28.0 43.2 '
Rods Mea 6 + 8end 28.0 28.0 64.0 Top End Prisery 02 11.9 19.5 21.6**
l Struct. Weld (shear) !
- 6 i Botton End Primary D2 11.8 16.4 21.6 i Struct. Weld (shear)
Allowables taken at worst cast temperature. I.e., for Case DI. T=460*F shell temperature.
t -
Ef ficiency factor for non-vcitanetric inspected welds = 0.5 i
O 4
l
o Table 2.2.3-15 Transfer Cask Stress Analysis Results for normal Loads Service tevels A and 8 Allomebles Stress (ksi)
Cask Stress Dead Thermal ** Normal Cmponent Type Weight Handling Allowable
- NUT [(H NR( NUIECH NR[ NUl[Cff N,R,(
Cask Shell Pri Men 6 0.3 0.5 N/A N/A 5.0 5.0
- 18.7 Men 6 + Bend 0.7 0.5 N/A N/A 5.0 5.0 28.1 Prl + Second 0.5 -
10.5 13.0 27.7 21.7 56.1 Top Prl Mew 6 0.1 0.9 N/A N/A - -
18.7 Cover Men 6 + Bend 1.0 1.0 N/A N/A 3.2 3.2 28.1 Plate Prl
- Second 1.0 -
4.7 4.7 - -
56.1 Bottae Prl Mer6 0.1 0.8 N/A N/A -
18.7 Cover Mer6 + Bend 0.6 0.8 N/A N/A 16.9 16.9 28.1 Plate Pri + Second 06 -
12.9 46.5 - -
56 i N
i Top Pri Meme 0.6 0.8 m/A N/A 1.5 1.5 18.7 o Ring Men 6 + Bend 0 0.8 N/A N/A I.5 1.5 28.1
- Pri + 5econd 0.2 -
30.8 30.8 6.4 6.4 56.1 Bottcze Pri Mew 6 0.2 0.4 N/A N/A 4.6 4.6 18.2 Ring Mem6 + Bend 0.6 0.6 N/A N/A 4.6 4.6 28.1 Prl
- Second 0.6 0.6 27.2 21.2 26.9 26.9 56.1 Allowables taken at 400T Thermal stresses are considered secondary stresses only
e o Table 2.2.3 18 Transfer Cask Load Combinations for normal Operating Conditions Service Levels A and 8 Stress (kstl Cask Stress Load Cerecaent Type Certinattens Calculated A11ovable NUffCH $
Cask Shell Pet kere Al A5. 81 82 5.3 5.5 18.7 Memo + 8end 5.7 5.5 28.1 Pri + Second 38.7 40.7 56.1 Top Pet kent ALAS 8182 0.1 0.9 18.7 Cover Meme + 8end 4.2 4.2 28.1 Plate Pri
- 5 eend 5.7 5.7 56.1 Bottom Pri Mene Al A5. 81-82 0.1 0.8 18.7 Cover Memb + 6end 17.5 17.7 28.1 Plate Pri + 5econd 13.5 46.5 56.1 Top Pri kenc Al A5, 81-82 1.9 2.3 18.7 Ring Marc + 8end 2.1 2.3 28.1 Pet + 5econd 35.2 37.2 56.1 80ttcn Pri Menc Al-A5. 81-82 4.8 $.C 18.7 Ring Manc + 8end 5,2 5.2 28.1 Pri + 5econd _
,54.7 54.7 $6.1 i
i l Load cattnations for cases Al A5 and 81-82 were defined by NUTECH. However. NOTECH reports all stresses for all ccroonents are the sam for all load cases. Consequently, there is no distinction whatsoever for the vertous load cases. NUTECH only calculated thermal stresses associated with the -3'F terrperature, i
I r
?
I l
l 2-110 l
l 1
- - - - - - - _ --a-----
o-Table 2.2.3-17 Transfer Cask Stress Analysis Results for Accident loads service level C** Allamables Stress (ksil Cask Stress 08I toad Component Type Handlleg Setsale Wind Camelnation Allowables*
gg...
- i. MUT[CH PRC Cask Shell Prl Wue 5.0 '. . i! 0.9 5.8
- 10.5 22.4 Mene + Bend 5.0 5.0 2.9 6.4 10.5 33.7 Top Prl Meu6 - - -
0.1 0.9 22.4 Cever Plate Meu6 + 8end 3.2 3.2 .4 16.6 7.4 33.7 Botton Prl Neu6 - - -
0.1 0.8 22.4 Cover Plate hu6 + 8end- 14.4 14.4 .3 31.8 29.6 33.7 Top Ring Prl Mem6 1.5 1.5 -
1.9 3.8 22.4 Moab + Bead 1.5 - 1.5 -
2.1 3.8 33.7 to Bottom Ring Prl Memb 4.6 4.6 -
4.8 9.6 22.4 L
Men 6 + Send 4.6 4.6 -
5.2 9.8 33.1 N
- Allowables taken at 400*F
- No eecondary stresses need to be evaluated according to the A5MC Code for Service Level C.
- *** lhe C1 load comelnation includes des &elght, seismic, and handling loads.
l t
4 4
Table 2.2.3-18 Transfer Cask Dr g Accident loads Service level D Allowables Stress (kst)
Cask ~
Com onent Stress Verticd Vertical Hortrontal Corner Corner Type Top Drop Bottom Drop Drop with DW Top Botton Allowables*
NUT [CH NRC NUT [CH NRC RUT [CH NRC NUT [CH NRC NUT [CH NRC Cask Shell Pri Me+ 6.5 6.4 4.7 10.1 8.8 9.1 5.9 11.4 3.8 7.3 44.9 Memb + 8end 7.8 7.8 7.8 12.5 - -
11.6 11.6 7.6 7.6 64.4 Top Pri Mes 10.5 14.7 8.6 -
6.9 4.4 13.9 20.0 -
44.9 Ring ;
Merb + 8end 17.0 17.0 8.4 - - -
18.0 20.0 -
64.4 Top Prl Mee 9.1 9.1 5.5 -
3.5 3.5 3.2 23.6 * -
44.9 3" Cover Mese + Bad -
9.1 6.3 ',
23.6 23.8 -
64.4 Bottom Prl Me4 1.1 -
5.7 19.8 3.7 4.1 - -
0.8 18.8 44.d 2" Cover Me,b + 8end 70.5 -
31.4 31.4 - - - -
T5.5 18.8 64.4 Bottom Pri Memb - -
16.0 31.6 8.8 7.4 - -
1.6 10.3 44.9 m Ring Meeb + Bend 13.3 -
22.9 22.9 - - - - ';5.8 8.8 64.4 8
w W
N Bolts for Ave. Tension - - - - - - -
61.1 - -
17.0 Top Cover l
- Allowables taken at 400*F. Sy = 20.7 ksi at 400*F W
i B
o s Table 2.2.3-19 Transfer Cask Stress Results for Torna t Driven Missile tapact Stress Massive Pen. Resist Allowable
- Cass W eneet Type Missile Wisstle Cask Prt Met 6.4 4.9 44.9 shell Pri + Bend 20.5 30.3 64.4 Top Pri kemb 0 0 44.9 Cover Pri + Bend 19.7 13.2 64.4 Bottom Pri Memb 0 0 44.9 Cover Pri + Bend 17.5 22.2 64.4 Allowable stresses based on service Level D allowables at 400*F
- -113
w
~
Table 2.2.3-20 Tr&nsfer Cask toed Carb for Accidnit Conditions Service Level D i i
Stress (ksi) _
l Cask Stress Case Case Case Cayonent Case . Allowable * '
Type 01 (Vert) 32 (Corner) D3 (HoeIz) D4 (D8I) ' kst '
NUffCH NRC NUTECH . MRC NUTECH . NRC NUTECH MRC t
! Cask Shell Pri Memb 6.6 10.6 6.8 11.9 9.0 9.6 6. 7
- 6.9 44.9 Mnse
- Bend 13.0' 12.4 12.1 ~8.8 -
31.0 30.8 64.4 Top Pri Memb - 10.6 15.5 16.4 20.8
, 7.0 5.2 0.6 0.8 44.9 j, Ring' Mei6 + 8end -
17.8 20.8' 20.8 0.8 0.8
+
64.4 4
Top Pri Memb 9.2 9.9 3.4 24.5 3.6 4.4 0.1 .9 44.9 Cover Memb + 8end 5.6 9.9 24.8 24.8 20.F 20.7 64.4 !
s Botton Pri Mes6 . "'.2 32.4 1.1 19,6 8.9 T.8 0.1 0.4 44.9
- i. Rin9 Mes6 + 8end 23.5 16.4 19.6 - -
0.6 0.6 64.4 N Bottse Prl Icsie 5.8 ' 20.6 0.9 10.7 3.8 4.9 0.1 0.8 44.9
.[s-* Cover' Memb + 8end 31.8 32.2- 16.1 9.4 - -
22.8 23
- 64.4 Service level D Allowables at 400*F 1
l [
t-T t .
i 4
M i l
s
th n
Table 2.2.3-21 Susinary of stress Ar.alyses for Upper Lifting Trunnions and Lower Resting Trwintons. Weld Regions aruf Cask Shell Critical Handling loads On-Site Transportation Loads (per AM51 M14.5) (per ASMElli Class 2)
Cceponent location Stress Instensity (ksi) Allowable (kst) Stress Intensity (ksi) Allowable (kst)
Section Upper Trunnion A-A 5.9 13.1 8-8 10.0 tspper Trunnion C-C 6.3 9.0 5.9 45.0 Sleeve 2" Insert Plate 5.0 5.4 27.1 45.0 Lower Trunnion N/A N/A 4.8 28,1 Lower Trunnion N/A N/A 5.6 32.6 Sleeve 26.9 32.6 e*
$ Plane plan, Weld i 6.9 9.0 1 6.4 45.0 Sleeve /Trunnlon 2 7.7 2 8.3 45.0 (Upper Trunnion)
Plane Plane
- Weld 1 5.0 9.0 6.0 1 45.0 Sleeve / Insert 2 5.5 7.0 2 5.4 7.0 (Upper trunnien) 3 4.4 5.4 3 4.2 32.6 Pla ,e Weld N/A I 5.6 28'1 Sleeve / Trunnion 2 7.3 28 l (Lower Trunnion) 3 5.7 28$1 Weld Plane Sleeve / Cask (tower Trunnlon) N/A 2 6.0 32,6
Table 2.2.4-1 54. mary of Design Criteria and Parameters of Handling and Iransfer [quipment tiot Important to safety Conponent Design Design Referenced or Satisfies Site -
Load Parameter Enveloped by Spectfic 9equirements TC Ilfting yoke syst m Critical Lift Dead toad 200.000 lb. BGt001.0209 Yes, actual load ts nminal Design Cale 188.500 lb.
hk ar ,Sy/6
- ANSI N14.6 Beams Design Code Box ar ,Su/10 Controlled by 10CFR Part 50 Transfer Traller Dead Load DSC + Skid + TC - 240.000 SAR Table 3.2-1 Yes l
Ib.
Operating leads on traller longitudinal: 30.000 lb. SAR Table 3.2-1 Yes deck Transverse: 40.000 lb.
Operating loads SAR Table 3.2-1 Yes 7
TC Positioning Skid longitudinal: 30.000 lb.
Trans<erse: 20.000 lb.
$ Vertical: 200.000 lb. 180.000 Extent of Motion Longitudinal: 35 in. SAR Yes l Transverse: +1-5 In. 5AR Yes !
Vertical: 10 In. SAR Yes IC Support Skid Dead toad DSC + TC a 200.000 lb. BGE001.0209 Yes. AISC Steel 180.000 lb. Construction Manual Operating load Positioning: SAR Yes longitudinal: 30.000 lb.
Transverse: 40.000 lb.
Transportation:
0.5g vertical +
0.59 transverse +
0.5g longitudinal Hydraulic flormal Operation Operating force: 20.000 lb. 16.250 lb. Yes Ram i Grapple Speed: 36 in/ min j i
Jamwd Condition Operating force: 80.000 lb. 65.000 lb. Yes Speed: 9 in/ min B
i
A Table 2.2.4-1 L-ry of Design Criteria and Parameters of Handling aruf Iransfer Equipment Ilot Iaportant to Safety Conponent Design Design Referenced or Satisfies Site -
toad Parameter fnveloped by Specifle Requi reinent s Cask Restraints Normel Operation 20.000 lb. 16,2 W 11. Yes Jmseed Condition 80.000 lb. 65.00" Ib. ,
Yes Optical Alignment Not Specified by BG1E NtR10MS-24P 1R Preoperational testing Systems allt verify system works. Hydraulle raes has relief valve set at 80,000- Ib DSC Vacuan System Normal Operation 05C Vacuuu: 1 3.0 torr SAR Yes pressure held for 30 minutes DSC helium backfilling Normal Operation 2.5 pstg i 2.5 psig SAR Yes y system or 837 go + 120 gw w
U DSC top cover automatic Closure of DSC prior to AVS Specification IR for NtROtS-24P Yes welding / cutting machine evacuation, pressure check, hellura backfill
o e Table 2.2.7-1 Caparison of Licensee Calculated Dose Rates to Audit Results Comparison of Licensee Calculated Dese Rates to Audit Results licensee (mrem / hour) Audit Case 10 Gama heutron Gama heutron BGERAD 35.6 33.B 8.2 7.4 BGETOPW C.658 'O.495 BGEBOTNS 1343.7 1307.0 8GEHM5R 0.16 0.09 CONCD00R 21.7 0.9
~
BGERA01 36.69 36.58 9.54 6.5 BGERA02 48.22 856.04 33.4 899.3 BGETOP 1148.425 896.7 2-118
e e Table 2.2.7 2 Comparison of Applicant and Audit Review Calculation Results for Total Dose Rate At A Of stance Frc. 'he Planned H5M Array Front View Total Dose Rate Isrum/ hour) 01574NCE (uf7ERS) LICENSEE AUO!T 10 0.39 0.81 70 0.075 0.20 100 0.043 0.13 700 1.8 E-4 2.3 E-4 1000 2.0 E5 9.1 E-6 End View Total Dose Rate (erum/ hour) 10 0.54 0.66 70 0.14 0.30 100 0.09 C.20 700 3.1 E-4 3.5 E-4 1000 5.2 C-5 1.4 E-5 2-119
- e 3.0 CONDUCT OF OPERATIONS BG&E is a regulated utility with a corporate organization and management system associated with nuclear power that has been previously approved by the NRC. The organization and procedures of the Calvert Cliffs Nuclear Power Plant Department, which has responsibility for operation and maintenance of Calvert Cliffs, have also been previously approved by the NRC. The SAR states that the current responsibilities, procedures, administration, and other aspects of management and operation of the utility and the plant will be extended to include the ISFSI.
The SAR incorporates or refers to appropriate sections of the FSAR (Reference
- 14) (See Table 1.1 of this SER) in addressing the following areas under Corporate Organization: corporate functions, responsibilities, and authorities, in-house organization, and technical staff. The descriptions in the FSAR do not currently refer to the ISFSI.
The staff considers that extension of responsibilities and functions of the current, approved organizations to provide for the ISFSI to be acceptable.
However, modifications of these to the extent of recognizing the ISFSI as a system with additional requirements established by 10 CFR Part 72 is essential. The staff therefore considers the absence of a full presentation in the SAR acceptable only if the statements in the SAR are commitments by BG&E to properly include the ISFSI in the respective portions of the FSAR in its first update, following NRC approval of the Calvert Cliffs ISFSI.
i Bu&E's interrelationships with contractors al,d suppliers, and the in-house organization specifically responsible for ISFSI design, for spent fuel management, and for operation and maintenance of Calvert Cliffs are identified in the SAR and are considered to be appropriate.
l i
The description of the on-site Calvert Cliffs organization, with primary-responsibility for spent fuel storage; personnel functions, responsibilities, and authorities; and personnel qualification requirements are incorporated in the SAR by reference to the FSAR. This is acceptable if the appropriate 3-1 l
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portions of the FSAR are updated to include the ISFSI. This is expected in-the first FSAR update following NRC approval of the Calvert Cliffs ISFSI.
3.1 Procedures 3.1.1 Administrative Procedures The current BG&E administrative procedures for Calvert Cliffs present the operating philosophy for the site, establish management policies, and provide basic rules and instructions to all site personnel. BG&E proposes to apply these procedures to ISFSI operations in order to operate the ISFSI safely.
The NRC staff has determined that this approach, which will integrate the ISFSI operations with the plant operations, is acceptable.
3.1.2 Health Physics Procedures BG&E health physics and ALARA policies and programs are already in place at Calvert Cliffs. These programs are to be extended to apply to ISFSI operations.
The primary goal of the health physics and ALARA programs is to ensure that exposure of personnel to radiation during all phases of design, construction, operation, and maintenance are kept at levels which are as low as is reasonably achievable.
This is achieved by integrating ALARA concepts into the design, construction, and operation of facilities.
The health physics program identifies the positions and responsibilities of participating organizations in conducting these programs. Specific responsibilities of the general office and Calvert Cliffs health physics staffs are contained in the BG&E Health Physics Manual.
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-The ALARA Program follows the general guidelines of NRC Regulatory Guides 1.8, 8.8, 8.10 (References 47, 45, and 46) and complies with the requirements of 10 CFR Part 20. At-Calvert Cliffs, the basic program consists of:
- 1. The BGLE ALARA Manual;
- 2. Continued surveillance of in-plant radiation and contamination conditions, as well as monitoring and control of personnel exposures, by the Calvert Cliffs and General Office health physics staff; and
- 3. An ALARA Committee, consisting of management and representatives from all groups, whose purpose is to assess the effectiveness of the ALARA Program at Calvert Cliffs.
Although the BG&E health physics and ALARA programs are adequate for existing operations, the NRC staff finds that the existing procedures will have to be 1-reexamined, verified as adequate, or modified or extended as necessary in order to accommodate the needs of the ISFSI. In particular, new procedures may be required to:
1.
Expand the on-site and off-site monitoring programs by adding new thermal luminescent dosimetry (TLD) monitoring locations at or around the ISFSI; and
- 2. Address the radiological support requirements for personnel transporting the DSCs, performing new module construction in the vicinity of filled modules, or entering the storage area for other routine tasks.
3.1.3 Maintenance Procedures This evaluation is concerned with maintenance procedures important to continuing safety, and that the maintenance procedures are safe in themselves.
-Maintenance operations include those surveillance, " inspection, test, and-3-3
. e-calibration activities to ensure that the.necessary integrity of-required systems and components.is maintained."-(10 CFR 72.44c(3)(ii))
Maintenance procedures for the ISFSI are addressed in the SAR at paragraphs 5.1.3.5 and 9.4.1, No maintenance procedures are described other than that the HSM will be periodically inspected to ensure that the vent openings are free. The statement is made that " maintenance procedures will-be developed to provide fer periodic maintenance of the transfer equipment." Review of the system and procedures for normal surveillance stated elsewhere (SAR paragraph 10.3.3) results in the following staff conclusions:
- 1. The NUHOMS-24P ISFSI system, as it is to be installed and used at Calvert Cliffs, per the SAR does not require periodic maintenance of the DSC or HSM.
- 2. There are no foreseen normal or off-normal conditions that are
- expected to require repair, application of protective coatings, or other routine maintenance operations for a DSC or HSM when s DSC is in storage within an HSM.
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- 3. Provisions for periodic surveillance of the cooling air inlets and l outlets, and the correction of any blockage of these, as provided j for in the SAR are adequate ard constitute the only significant I maintenance operations for the DSC or HSM when in use.
- 4. Maintenance of the TC, which will be reused, occurs in conjunction with that use, as described in the operation description (SAR Section 5). The description of this is considered adequate.
- 5. Maintenance operations on the HSM prior to use essentially consist of ensuring that the HSM is free of debris, which is included in
( the loading procedure sequence.
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- 6. Maintenance of the transfer and welding equipment used within the Auxiliary Building is properly a subject for the 10.CFR Part 50 review of the updated FSAR or other submitted documentation. This .
includes the DSC welding machine, lifting equipment, transfer trailer, annulus seal, and prime mover, to the extent of safety concern.
- 7. Maintenance is expected to be necessary for the ram, the TC positioner mounted on the transfer. trailer, the associated support skid equipment, and the slings for securing the TC to the.HSM.
Maintenance procedures and schedules for the equipment should be as stated in documentation provided by the supplier. Maintenance of the rigging should be consistent with the requirement for other rigging used in nuclear safety applications and as provided~for in BG&E procedures applicable to rigging used in the Auxiliary Building.
- 8. Maintenance responsibilities that do exist for the ISFSI system are appropriately assigned to the Calvert Cliffs Nuclear Power Plant Department.
The general conclusion of the staff regarding the description of maintenance precedures and plans, and the commitment for future preparation is that they are adequate for the proposed system.
3.1.4 Operating Procedures Operating procedures for the Calvert Cliffs ISFSI system are outlined in the SAR. Those procedures occurring external to the Auxiliary Building and not associated with equipment used solely within-the Auxiliary Building were formally evaluated as part of the preparation of this SER. These outlined procedures have been reviewed and are considered acceptable. Procedures for ISFSI surveillance and for reaction to extraordinary conditions stated-in the SAR have been reviewed and are considered to be satisfactory. The SAR commits BG&E to the preparation of more detailed operating procedures (SAR Paragraph 3-5
a 0 9.4.1) in accordance with existing instructions for such preparation. In addition, the SAR commits BG&E to the preparation of formal Fuel Handling Procedures (FHP) to control preparation of the DSC for transport, including loading, sealing, drying, backfilling, and placement; and transport and placement into the HSM.
The staff considers the description of proposed preparation of procedures to be acceptable. However, the staff further considers that drafts of all operating and testing procedures should be, completed, approved, issued, and in use, not later than the final " dry-run" of the fuel pool to HSM sequence testing.
Operating procedures for operations within the Auxiliary Building (or which relate only to equipment or operations only occurring within the Auxiliary Building) have been reviewed for consistency with the overall system and with prior evaluations of NUHOMS ISFSI TR and SAR. The procedures appear appropriate; however, formal 10 CFR Part 72 NRC review and evaluation are conducted in conjunction with 10 CFR Part 50 review of the updated FSAR or associated separate documentation.
3.1.5 Test Procedures Pre-operational testing is addressed at paragraph 3.1.6, below. The SAR-provides for routine testing associated with operating controls and limits in Section 10 (further evaluated in Section 5 of the SER). The phenomena to be tested are identified; however, specific procedures or instrumentation for the conduct of the testing are generally not included. These test procedures are considered to be part of the operational procedures. The written procedures are to be prepared and used not later than the final " dry-run" of the fuel pool to HSM sequence. The staff considers the subjects and scope of the operational testing to be appropriate and acceptable.
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3.1.6 Pre-Operational Test Procedures The SAR describes a test program associated with initial ISFSI operations (SAR paragraph 9.2.2). This program will use an actual DSC and a mock-up, partial length DSC, the TC, and HSM, and the lifting and handling equipment to verify suitability of the equipment, installation, and the proposed operations prior to use with actual spent fuel. This testing is considered appropriate and acceptable, if the procedures are performed in accordance with the formal written procedures to be used with actual fuel (i.e., also constitute a test and validation of the written procedures).
No pre-operational or initial operation tests are included for heat transfer, temperatures, or air flows of the DSC loaded with a non-radioactive heat source inside the HSM. Such testing was required in the SER for the generic system (Reference 6) for the NUHOMS-24P installation at Oconee. The staff considers that testing at Oconee and the NUHOMS-7P system at H.B. Robinson (Reference 17), have adequately verified the basic heat transfer capabilities of the design.
Further, the temperature rise test to be performed on each HSM following installation of a DSC (SAR Paragraph 10.3.2.7) provides further system validation. The staff concludes that the proposed pre-operational test procedures are appropriate and acceptable. _
3.2 Records The SAR (Reference Ib) identifies existing procedures and describes modifications that will be made to these to meet the record requirements specific to the ISFSI described in 10 CFR Part 72. The staff considers that the records described, if modified to satisfy the appropriate requirements of 10 CFR Part 72, would be appropriate and acceptable. The procedures, records, and responsibilities must be established prior to the receipt of fuel at the ISFSI.
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- a. .o 3.3 Training and Certification The SER is concerned with training and certification of personnel for operation of equipment and controls that have been identified as important to safety, and direct supervisory personnel for such operation. The license applicant must establish a program of training, proficiency testing a' !
certification of such personnel which must be submitted with the license application for approval (10 CFR 72.192).
The proposed training program is described in Section 9.3 of the SAR and the FSAR (Reference 14). The nature of the ISFSI is such that the proposed training, which is essentially an extension of the current Calvcet Cliffs training and qualification program, is considered to be satisfactory. The existing BG&E employee personnel qualification requirements are considered to be appropriate for the " certification" of personnel.
General training is to be provided to operations, maintenance, and health physics personnel on the applicable regulations and standards, and in the engineering principles of cooling, radiological shielding, and structural characteristics of the DSC and HSM. BG&E plans to provide detailed training to its operators in the areas of the DSC preparation and handling; fnel loading; transfer cask preparation and handling; and transfer trailer loading.
The maintenance personnel will receive detailed training in the operation of the aut tic welder for the DSC top end shield plug, the operation of the transfei trailer, alignment of the cask skid with the HSM, and alignment and normal and off-normal operation of the hydraulic ram assembly. Specific training is to be provided on cleaning of the HSM air inlets and outlets. The health physics personnel will receive training in the radiological shielding design, especially of the various ISFSI system shields.
The NRC staff finds that the existing training and -certification program, supplemented to include the planned specific ISFSI training, meets the-ISFSI training and certification requirements (10 CFR Part 72, Subpart I).
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- l 3.4 Physical Protection The ISFSI must be derigned to furnish physical protection from sabotage. The requirements (10 CFR 72.182) are for information'on site layout, design features, design criteria, design bases, materials of construction, and quality assurance sufficient to provide reasonable assurance that the final security system will provide such protection.
The SAR includes security with safety criteria. The resulting system and procedures are considered to fully meet requirements for physical protection of the fuel. This is provided for fuel in storage by the doubly welded seal of the DSC and the tack welds on the HSM access cover, and by inherent security features of heavy weights and limited access, and heavy equipment necessary to move the DSC. Physical protection during transfer operations is provided by the welded-on covers of the DSC and the temporary bolted-on covers of the TC. The staff considers that the physical protection to be provided by the system design and provisions is adequate and acceptable.
BG&E has an NRC-approved physical protection program for the Calvert Cliffs Nuclear Power Plant. They have also developed an amendment to this program to accommodate the needs of the ISFSI. The proposed Physical Security Plan, Safeguards Contingency Plan, and Design for Physical Security and Guard Training Plan are separate from the SAR to withhold them from public disclosure (per SAR Section 9.6)'. NRC staff review of these documents is separate from the SAR review reported in this SER (Reference Docket Nos. 50-137 and 317).
3.5 Emergency Planning "For an ISFSI that is located on the site of a nuclear power reactor licensed for operation by the Commission, the emergency plan required by 10 CFR 50.47 shall be deemed to satisfy the requirements" (for an emergency plan) (10 CFR 72.32). The SAR (Section 9.5) identifies the Emergency Response Plan for Calvert Cliffs in catisfaction of this requirement.
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. i This plan, which meets the requirements of 10 CFR Part 50, Appendix E, and which has been approved by the NRC staff, describes how emergencies will be addressed at the site. The staff considers that the emergency planning is appropriate and acceptable for coverage of the Calvert Cliffs ISFSI.
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4.0 QUALITY ASSURANCE Chapter 6 of the BG&E ISFSI License Application refers to Chapter 11 of the ISFSI Safety Analysis Report (SAR) for the details of the QA program applicable to the Calvert Cliffs ISFSI. In Chapter 11, " Quality Assurance,"
of the Calvert Cliffs ISFSI SAR dated December 19, 1991, BGLE has described the QA program for the life cycle of the ISFSI.
For the construction phase (design, fabrication, construction, startup testing), the operational phase (operation, maintenance, modification), and the decommissioning phase of the Calvert Cliffs ISFSI, BG&E has committed to apply its NRC-approved 10 CFR 50 Appendix B QA program as described in Appendix 1B of the Calvert Cliffs updated Final Safety Analysis Report to the items defined in the ISFSI SAR as safety-related. These items are the transfer cask, the dry shielded canister (basket, spacer disks, support rods, top and bottom end shield plug / support, dry shielded canister body, and end closure plates), and the lifting yoke.
For the horizontal storage module (concrete shielding, dry shielded canister support assembly, and ISFSI foundation), BG&E has committed to apply special ,
QA provisions described in Chapter 11 of the Calvert Cliffs ISFSI SAR during the construction phase. The horizontal storage module, although important to safety, is not safety-related. However, BG&E has committed to apply its NRC- t approved 10 CFR 50 Appendix B QA program to the horizontal storage module after the construction phase. In accordance with the acceptance criteria given in the Fuel Cycle Safety Branch Technical Position of June 20, 1986, regarding QA for ISFSis and the provision in 10 CFR 72.140(b) for a graded QA program for ISFSI items and activities important to safety, this is acceptable to the staff.
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5.0 OPERATING CONTROL AND LIMITS Each license issued under 10 CFR Part 72 shall include license conditions pursuant to 10 CFR 72.44.
In addition to the conditions set forth in 10 CFR-72.44(b), each application for license under 10 CFR Part 72 shall include proposed technical specifications pursuant 10 CFR 72.26 and consistent with 10 CFR 72.44(c). The finally approved technical specifications will be maae part of the operating license.
The technical specifications of a license define certain features, characteristics, and conditions governing operation of an installation.
Technical specifications cannot be changed without approval of the NRC.
Consistent with 10 CFR 72.44(c), the technical specification will cover safety limits, limiting safety system setting, limiting conditions for operation, surveillance requiremants, design features, administrative controls, effluent control, and environmental monitoring.
Sections 10 of the SAR describes operating controls and limiting conditions for operation. Section 10.1 of the SAR discusses the basis for deleting five limits which are included in the TR for the NUHOMS-24P. These areas are evaluated as part of this SER. Each of these items will be discussed below.
In addition, the NRC staff has determined that other limiting conditions of operation are necessary for the NUHOMS-24P to meet.the requirements of 10 CFR 72.44.
Sections 10.3.2.5, 10.3.2.6, and 10.3.4.1 of the SAR for the TR have been deleted from the SAR for Calvert Cliffs. All of these deal with surface dose rates for the DSC, HSH or TC. The staff has evaluated the BG&E basis for these deletions below.
Operating limits 10.3.2.5, Dose Rate at End of DSC Shield Plug, 10.3.2.6, Surface Dose Rate at the HSH while the DSC is in storage, and 10.3.4.1, Maximum Surface Dose Rate on Transfer Cask, all provide for the monitoring of personnel exposure and provide for personnel protection in accordance with ALARA design considerations. Their removal from the Calvert Cliffs ISFSI 5-1
O 4 operating controls and limits was questioned as a part of this review process.
BG&E has agreed to reinstate these three limits in future revisions of the SAR. Subject to later ',er'.?ication, the inclusion of these three operating limits on DSC, HSH and transfer cask dose rates is acceptable.
A fourth operating control, deleted by BG&E in the SAR, concerns the requirements to return the DSC to the fuel pool following a drop of 15 inches or more. BG&E has stated that the controlling drop height should be 80 inches; however, the NRC staff takes exception to this and requires the more conservative drop height of 15 inches. This concern was discussed in a series of NRC staff questions and BG&E responses. The utility agreed in its response of December 20,1990 (Reference 1) to maintain the 15-inch drop limit for return of the DSC to the fuel pool for opening of the DSC and inspection of the fuci assemblies following the drop accident. The basis for this requirement is that the Service level D stress allowables permitted by the ASME B&PV Code include plastic deformation, but require removal from service and inspection following the accident case. Since there is always some room for error in the analysis as well as the actual deceleration levels postulated in the accident, and in the manufacture of the DSC, the conservative requirements of the ASME Code as well as the previous SERs will continue for this BG&E application.
Operating Controls and Limits for Criticality Considerations Topic S_pecification Re feren_ce Fuel Type CE 14x14 Fuel SAR Table 3.3-5 Specifications Burnup Enrichment SAR Table 10.3-2, equivalent Figure 3.3-1
<= 1.8 wt.% "50 U5 Initial s 4.5 wt.% 0 SAR Table 3.3-5 Enrichment DSC Moderator during Loading Boron >-1800 ppm SAR p. 3.3-6 and Unloading concentration Operations 5-2
In addition to the above reinstatements and changes to the BG&E proposed operating controls and limits, the staff has determined that three additional limiting conditions of operation (LCO) neid to be included in Section 10 of the Calvert Cliffs ISFSI SAR. These new LCOs are necessary because of specific design features of the Calvert Cliffs ISFSI Transfer Cask and features of the transportation path between the spent fuel storage pool and the ISFSI site.
The new LCOs should be added to Section 10.3.2 and are delineateo below.
1.4 Maximum Ambient Temoerature Durina loaded Transfer Cask Movement To The ISFSI Site A.
Title:
Maximum Ambient Temperature During 1.oaded Transfer Cask Movement To The ISFSI Site B. Specification: Daylight ambient temperature: $ 103*F.
C. Applicability: All Transfer Cask movement to the ISFSI site while carrying any spent nuclear fuel.
D. Objective: To stay within the thermal limits of the Transfer Cask neutron shielding material.
E. Action: If the measured daylight ambient temperature exceeds 10: F, no spent fuel loaded transfer cask movement outside of the spent fuel pool building will be allowed.
F. Surveillance: Using a calibrated instrument with a known acc; racy, the ambient temperature in full daylight (i.e., not shaded) will be measured no earlier than 30 minutes prior to movement of the loaded transfer cask outside ef the spent fuel pool building. The measured temperature, including an allowance for instrument uncertainty and accuracy, along with its time and date shall be recorded and stored as part of the ISFSI records.
G. Basis: Due to increased neutron shielding for the Calvert Cliffs Transfer Cask as compared to the original 5-3
NUH0MS-24P design, ambient temperatures above 103*F in conjunction with the solar radiation heat load can result in exceeding the shielding material design temperature.
- 2. Loaded Transfer Cask Movement Restriction Durina Onsite Fossil Fuel Tanker Truck Presence A.
Title:
Loaded Transfer Cask Movement Restriction During Onsite Fossil Fuel Tanker Truck Presence B. Specification: When a tanker truck carrying fossil fuel (e.g.,
gasoline, diesel fuel, natural gas, etc.) is situated or in transit within 200 meters of the ISFSI transfer route or ISFSI site, movement of a spent fuel loaded transfer cask outside of the spent fuel pool building is prohibited.
C. Applicability: Applicable to all movement of a spent fuel loaded transfer cask outside of the spent fuel-pool building.
D. Objective: To preclude an accident involving a fire or explosion near the transfer cask which is initiated from a tanker truck containing large quantities of fossil fuels.
E. Action: If a fossil fuel-carrying tanker truck is situated or in transit within 200 meters of the ISFSI transfer route or site, no spent fuel loaded transfer cask movement outside of the spent fuel pool building will be allowed.
F. Surveillance: Prior to and during movement of a spent fuel loaded transfer cask outside of the spent fuel pool building until the OSC has been placed inside the HSH and the shield door has been installed, all fossil fuel-carrying tanker trucks will be prohibited from entering the ISFSI transfer route area. In addition, any such tanker trucks already within this area must first be removed from the area before transfer cask 5-4
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6.0 DECOMMISSIONING Decommissioning is addressed in the SAR (Reference 1) at paragraphs 3.5 and 9.6. A decommissioning plan is included by Reference l'.c. Requirements for decomaissioning planning and for providing for decommissioning in system design are included in 10 CFR 72.24(g) and (q),10 CFR 72.30, and 10 CFR 72.130. The latter effectively summarizes the principal requirements for design:
- The ISFSI must be designed for decommissioning.
. Provisions must be made to facilitate decontamination of structures and equipment.
. Provisions must be made to minimize the quantity of radioactive wastes and contaminated equipment
- Provisions must be made to facilitate the removal of radioactive wastes and contaminated materials at the time the ISFSI is permanently decommissioned.
The BG&E Decommissioning Plan for the Calvert Cliffs ISFSI was prepared subsequent to the SAR. The staff considers the submitted plan (Reference 1.c) to be incorporated with the SAR, thereby satisfying the requirements of 10 CFR 7?.24(g) for inclitsion of a description of the plan.
Paragraph 2.1.8 of the SER addresses the criteria for decommissioning. The staff considers that the criteria and/or the actual design of the elements of the proposed Calvert Cliffs ISFSI are acceptable for decommissioning in view of higher priority nuclear safety requirements for the design, and prior NRC approval of similar designs prepared under the same requirements.
The principal uncertainty associated with the decommissioning is whether neutron-induced radioactivity of the HSH will be at a level that requires its 6-1 I
4 9 disposal as low-level waste or whether the HSM can be broken-up -and disposed of as non-radioactive reinforced concrete rubble. The plan conservatively assumes the former as the basis for the cost estimates. The latter case is highly dependent upon the development of future criteria which could allow the HSH to be treated as non-radioactive. When new criteria are adapted, the cost estimates and funding plan could than be revised accordingly.
The cost estimate and planned trust fund are based on construction of all of the HSMs covered in the license application. Not all of the HSMs are to be built initially.
The staff considers that it is feasible to empirically determine the development of neutron-induced radioactivity in the HSH (as by time lapse activity measurements of a core sample of the reinforced concrete after a period of DSC storage) such that activity levels could be projected for the time of decommissioning. If the projected radioactivity level is below a level determined to be acceptable to the NRC, then a lower decommissioning cost estimate would be acceptable.
The staff concludes that the proposed ISFSI acceptably meets decommissioning requirements, understandably, the decommissioning plan is highly sensitive to uncertainties in disposal criteria and methods. The plan provides for the removal and disposal of activated portions of the DSC and HSM as low-level waste.
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7.0 CONCLUSION
S Introduction Pursuant to 10 CFR Part 72, Subpart C, the Commission is to issue a license after determining that the application meets: (1) the standards and requirements of the Atomic Energy Act; (2) the regulations of the Commission; and (3) specifically, the fourteen points identified in 10 CFR 72.40(a).
The NRC staff conclusions on these fourteen points are based on this safety evaluation report (SER) for the Baltimore Gas and Electric Company ISFSI, the SER for the NUHOMS-24P dry storage system, the Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUHOMS-24P (TR), and the Environmental Assessment (EA) for the ISFSI at the Calvert Cliffs site.
This section presents summarized conclusions based on the fourteen points of 10 CFR 72.40(a). References to sections in backup documents (SERs and EA) which support the findings are included. This section serves as a comparison of the BG&E application against licensing requirements stated in 10 CFR 72.40(a).
Findinas In accordance with 10 CFR 72.40(a)(1), the NRC staff finds that the applicant's proposed ISFSI design complies with the general design-criteria contained in Subpart F of 10 CFR Part 72 because:
- a. Quality Assurance Standards (10 CFR 72.122(a))
BG&E has identified the following components as being important-to safety and controlled by the BG&E QA program: (1) the reinforced concrete HSM and its DSC support structure, (2) the DSC and-its internal basket assembly, and (3) the on-site transfer cask.
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- b. Protection against environmental conditions and natural phenomena (10 CFR 72.122(b))
The BG&E NUHOMS-14P system has been designed to function under both normal and severe environmental conditions at the Calvert Cliffs site. Both natural phenomena and man-induced conditions (i.e., fire and-explosions) have been considered. Appropriate combinations of effects of normal and accident conditions and the effects of natural phenomena have also been considered. (See Section 2.1 and 2.2 of this SER). The ISFSI does not release any-water or radioactive nuclides and is therefore not considered to pose any threat to the local aquifer. (See Section-2.4 of the SAR for the BG&E ISFSI.) The applicant has demonstrated that the acceptance criteria on fuel and concrete temperatures w".1 be met undar extreme temperatures appropriate for the site and changing wind conditions.
- c. Protection against fires and explosions (10 CFR 72.122(c))
The OSC and HSM do not contain any significant flammable or explosive materials. The steel and reinforced concrete materials used in its construction do not support combustion. Although there is no fixed fire suppression system located inside the Calvert Cliffs ISFSI fenced boundary, BG&E will provide portable fire suppression equipment inside the fenced area. See Section.
3.3.6 of the SAR as well as 8.2.10. The effects of a forest fire have been analyzed, and the results do not challenge the. functions of the HSM. Postulated LNG leak induced explosions are not currently possible because the nearby Cove Point facility is not operating. At least sixty days prior to the Cove Point facility becoming operational, BG&E will submit a consequence analysis.
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- d. Sharing structures, systems and components (10 CFR 72.122(d))
The Calvert Cliffs reactors and the ISFSI do not share structures, systems and components important to safety. (See Section 2.2. of this SER.)
i e. Proximity of sites (10 CFR 72.122(e))
The cumulative ri.diological effect of operation of the ISFSI and the Calvert Cliffs reactors is less than 13.5 mrem /yr to the nearest offsite individual. There are not ISFSI-related accidents which could have an effect offsite requiring protective actions.
(See Section 2.2.8 of thi:. 5ER and Section 6.2.1.1 of the EA.)
- f. Testing and maintenance of systems and components (10 CFR 72.122(f))
The NUHOMS-24P system as designed for the Calvert Cliffs site is basically passive and essentially maintenance free. However, Section 9.2 of the SAR describes preoperational tests which BG&E commits to make prior to spent fuel loading and storage. The 4
staff accepts these commitments.
Operational testing specified by the NUTECH-24P TR requires that the maximum air rise temperature from the HSM inlet to outlet, af ter the DSC is initially loaded, shall be limited to 60 F.
(Reference 10.3.2.7 of the SAR.) For normal maintenance, the NRC staff requires that visual inspections will be carried out on a daily basis to ensure that the inlet and outlets of the HSM are unblocked. (Reference 10.3.3.1 of the SAR.)
- g. Emergency capability (10 CFR 72.122(g))
Section 9.5 of the SAR covers emergency capability for the Calvert Cliffs Nuclear Power Plant. BG&E states that its emergency program is adequate to manage the consequences of ever.ts which 7-3 ;
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e o might involve the ISFSI. This plan was approved under Section 50.47 and therefore satisfies 10 CFR 72.32. (See Section 3.5 of this SER.)
- h. Confinement barriers and systems (10 CFR 72.122(h)(1))
The fuel cladding is protected against degradation and gross rupture by (1) maintaining an inert helium atmosphere in the DSC, and (2) maintaining the fuel cladding at temperatures that provide reasonable assurance that its integrity will be maintained throughout the period of storage. (See Section 2.2.5 of the SER.)
The NUHOMS-24P system involves a dry storage concer,t, therefore the underwater clause does not apply.
The DSC p-ovides the necessary confinement of airborne radioactive particulate material during normal and accident conditions. (See Sections 5.0 and 12.0 of the SER for NUHOMS-24P TR and EA Section 6.2.)
The surveillance and monitoring program provides a continuous means of monitoring the TC, DSC, and HSH (see Section 2.2.10 of the SER) to determine conditions indicative of when corrective actions must be taken to maintain safe s u rage conditions. The i proposed system for monitoring and surveillance is based on.-
consi' - tions . - 'ng to the ability of the specific design to i
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o o confine the fuel material in connection with the loading, transporting t.nd placing of spent fuel into the HSM.
Tha DSC seal integrity provides assurance that the necessary helium atmosphere will be maintained over the storage lifetime.
(See Section, 2.2.3.2 of the SER and Section 5.3.1 of the SER for the NUHOMS-24P TR.) The DSC design, materials manufacture, and test specifications for seal weld (dye penetrant testing and heliuin leak detection) and the lack of plausible mechanisms, under normal and off-normal conditions of use (such as corrosion, creep or cyclic loadings), which would cause the double seal welds to fall, ensure that the internal 'neliem atmosphere is maintained.
The DSC contains no penetrations for sampling or gauges nor other leakage paths (such as across bolted closure seals that can leak).
Diffusion through the weld and DSC parent metal is not a potential mechanism to permit the escape of helium and ingress of oxygen.
The fuel cladding is protected against degradation that leads to gross rupture (see Section 2.2.5 of the SER) by maintaining an inert helium atmosphere. As shown above, the combination of the sealed DSC design coupled with no known plausible long-term degradation mechanisms ensure that the internal helium atmosphere will remain. Therefore, an individual continuous monitoring device for each DSC is not required.
Section 5.4 of the EA states that the current radiological environmental monitoring program at Calvert Cliffs will also serve as the program for the ISFSI. (See Section 7.6 of the SAR.)
Because the NUHOMS-24P system is passive, no instrumentation important to safety is required for the operation of the faci'lity.
(See 5.4.1 of the SAR.) i 1
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The DSC is designed so that retrieval from the HSH is the reverse process from insertion into the HSM. The top cover plate can be cut off by conventional cutting operations in order to extra:t the ,
IFAs after the DSC is lowered into the spent fuel poul. (See Section 11 of the SER for the TR.)
- 1. Instrumentation and control systems (10 CFR 72.122(1))
However, no instrumentation and control systems to monitor sy, et, important to safety are provided or required during storage. gsee Section 2.7 of the SER for the NUHOMS-24P TR.) The DSC and HSM are considered components important to safety that comprise the NUHOMS ISFS! design. Due to the passive design, they are not considered operating systems in the same sense as spent fuel pool cooling water systems, ventilation systems, or offgas systems, which may require instrumentation and control indicators to ensure proper functioning. Hence, no instrumentation and control systems to monitor and control DSC and HSH functioning have been provided.
An argument can be made that since the DSC and HSM are identified as components that are important to safety and that the requirements of 72.122(i) apply only to " systems important to safety," and that this design requirement therefore does net apply to the DSC and HSM. However, while the DSC and HSH components cor.stitute the NUHOMS design, one can equally conclude that the DSC and HSH could be viewed as a " system" (NUHOMS-24P system) that is "important to safety," and therefore making this design criteria technically applicable. While the NUHOMS-24P systeE is not_specifically called out as a " system important to safety" it certainly is important to providing safe storage'of the spent fuel.
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Whether or not it is viewed as a " system," given the passive design and inherent safety, there is no technical reason to require instrumentation and catrol systems for monitoring the l NUH0MS-24P (DSC and HSM) system during storage operations at the l Calvert Cliffs ISFSI. Therefore, pursuant to 10 CFR 72.7, the Commission has determined to grant an exemption from the ,
requirements of 10 CFR 72.122(i) with respect to the DSC and HSM l for storage operations, and, in support thereof, has further f determined that the grant of such exemption is authorized by law I and will not endanger life or property or the common defense and security and is otherwise in the public interest.
J. Control room or control area (10 CFR 72.122(j))
No control room or control areas are required for the operation of the ISFSI because it is passive. Surveillance of operations will be visual. (See Section 5.2 of SAR.)
- k. Utility or other services (10 CFR 72.122(k))
The "'!:! OMS-24P system, designed for dry passive storage of irradiated fuel, requires no support systems during storage.
Handling and transportation operations in the spent fuel pool area, where lifting equipment is required, are covered under Calvert Cliffs reactor operating licenses. Outside of the spent fuel pool building, transportation systems are required, but-do not pose a concern important to safety, because the 10 and DSC are designed to confine the irradiated fuel assemblies (IFAs) and protect personnel in the event of all postulated ac ' dents. No emergency utility services are required at the Calvert Cliffs ISFSI. (See Section 4.3 of the SAR.)
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- 1. Retrievability (10 CFR 72.122(1))
The NUHOMS-24P system is designed to permit retrievability of the 1 IFAs after storage or following an accident which requires !
inspection of the confinement boundary. (See Section 11 of the '
Independence of installation (10 CFR 72.3)
In accordance with 10 CFR Section 72.3, the NRC staff finds the Calvert Cliffs site ISFSI meets the definition of independence as described in 10 CFR Part 72 since it is not physically connected to the reactor facility and its sharing of utilities and services does not, "(1) increase the probability or consequences of an accident or malfunction of components, structures, or systems that are important to safety; or (2) reduce the margin of safety as defined in the bas.1s for any Technical Specification of either facility." The applicant has performed an appropriate evaluation that concludes that the activities associated with the ISFSI do not represent an unreviewed safety question for reactor operations and that no additional-changes to the Technical Specifications of the reactor operating license are required.
Criteria for nuclear criticality (10 CFR 72.124)
The NUHOMS DSC is designed te provide nuclear criticality safety during wet loading and unloading operations through the use of borated water with a-minimum concentration of 1800 ppm. After fuel loading and DSC drying and sealing, the spent fuel assemblies re not moderated assuring subcriticality during subsequent operations and configurations. (See Section 2.2.2 of this SER.)
The criteria which is applied is consistent with the criteria developed for
! -the generic NUH0MS-24P design (see Reference 4). The maximum effective reactivity is less than 0.95 for wet loading and unloading operations for all credible configurations and environments. The maximum effective reactivity is less than 0.95 for all credible accident cases, including misloading of fresh 7-8
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l fuel in optimum moderator density conditions taking credit for boron, with no restriction on drain-down time during DSC closure operations.
Criteria for radioloaical orotection. exposure control and alarm systems (10 CFR 72.126(a),(b))
The DSC and HSM provide acceptable shielding of the stored spent nuclear fuel.
The applicant, by application and possible expansion of the existing health physics program at Calvert Cliffs, will provide for DSC or TC decontamination, ISFSI access control, and radiation surveys that ensure radiation exposures are controlled at the ISFSI. (See Section 3.1.2 of this SER and Section 10.0 of the NU" CH TR SER.) No separate radiological alarm system is required for the ISFSt.
Analyses show that exposure limits specified in 10 CFR 72.104 and 72.106 are not exceeded for both normal cperations and postulated accident conditions.
(See Section 6.2 of the EA and Section 10.2 of the SER for the TR.)
Criteria for radioloaical protection. effluent monitorina and control (10 CFR 72.126(c),(d))
The DSC, by virtue of its very low leak rate and low surface contamination, allows essentially no radioactive effluents to escape the ISFSI. Direct and air scattered (skyshine) radiation levels from the HSH surfaces are expected to be well within acceptable limits and will be verified by thermoluminescent dosimeter measurements and radiation surveys. (See Section 2.1.2 and 3.1.2 of this SER and Section 6.2 of the EA.)
Criteria for soent fuel. hich-level radioactive waste. and other radioactive waste storaae and handina (10 CFR 72.128)
Only minimal quantities of low-level radioactive waste will be generated as a result of the ISFSI operations. Solid wastes include small size filters and general " housekeeping" materials such as rags, and swabs. Liquid wastes will result from decontamination procedures inside the Auxiliary Building. The 7-9 l
. o contaminated pool water removed from the DSCs will be pumped back to the spent fuel pool. Contaminated air and helium purged from the DSC during evacuation will be pumped to the Auxiliary Building processing system. These wastes will be managed in accordance with existing Calvert Cliffs procedures. (See Section 6.2.1 of the EA.)
Criteria for scent fuel, hiah level radioactive waste. and other radioactive waste storaae and handlina - heat removal capability (10 CFR 72.128(a)(4))
Heat removal for the Calvert Cliffs NUHOMS-24P is by totally passive means. _
The natural draft cooling mechanism has been used in operating spent fuel <
storage systems at H.B. Robinson Nuclear Power Plant and Oconee Nuclear Power Plant and has been demonstrated to be reliable to date.
Criteria for decommissionina (10 CFR 72.130)
The NUHOMS system at Calvert Cliffs is designed for decommissioning. (See Section 6.0 of this SER and Section 11.0 of the SER for the TR.)
Sitina evaluation factors (10 CFR Subpart E)
In accordance with 10 CFR 72.40(a)(2), the NRC staff finds that site characteristics and external natural and man-induced events have been investigated and assessed, that acceptable design basis events and conditions have been determined, and that the design for the NUHOMS system for Calvert Cliffs envelopes parameters associated with these site characteristics and design. The NRC staff therefore finds that the proposed ISFSI site complies with Subpart E of 10 CFR Part 72 because:
General considerations (10 CFR 72.90)
All the general considerations associated with the siting evaluation factors have been evaluated in this SER, or in the EA for Calvert Cliffs, and are considered adequate with regard to the design envelope proposed by BG&E in the SAR.
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1 o e Qesian basis external natural events (10 CFR 72.92)
Based on regional characteristics, natural phenomena have been appropriately assessed and design basis phenomena identified and evaluated. (See Section 2.1, and 2.2.)
Desian basis external man-induced events (10 CFR 72.94)
Man-made facilities and activities in the region have been examined, potential man-induced events that affect ISFSI design have been identified, and design !
basis external mah-induced events have been appropriately evaluated. (See ;
Section 2.1, 2.2 of this SER.)
Sitino limitations (10 CFR 72.96)
Does not apply to this ISFSI application.
Identifyina recions around an ISFSI site (10 CFR 72.98)
The regions around the proposed Calvert Cliffs ISFSI site have been identified and impacts have been evaluated. BG&E has indicated that an LNG facility located at Cove Point (within a 4 mile radius of the Calvert Cliffs site) may be installed. Should this happen, BG&E has committed to providing a ,
consequence analysis of the results of a postulated LNG 1eak no later than 60 days prior to startup of Cove Point.
Definina potential effects of the ISFSI on the reqign (10 CFR 72.100)-
i Construction, operation, and decommissioning of the ISFSI have been found to cause no significant impact to the region surrounding the ISFSI. (See EA Sections 6.0, 8.0 and 9.0.)
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. 9 Geolonical and seismolooical characteristics (10 CFR 72.102)
The design earthquake for the Calvert Cliffs ISFSI NUH0HS-24P system (including TC, HSH, and DSC), which has a peak horizontal acceleration of 0.159 (see Section 3.2.3 of the SAR), has been suitably established and is -
commensurate with the requirements of 10 CFR Section 72.102(f)(2). (See Sections 2.4 and 3.3 of the SER for the TR and Sections 2.1.5 and 2.2.3 of this SER.)
Criteria for radioactive materials in effluents and direct radiation from an ISFSI (10 CFR 72.104)
The annual dose equivalent (from direct and air-scattered radiation originating from ISFSI sources) to the nearest real individual located beyond the controlled area boundary (i.e., the nearest resident) is only a small fraction of the 25 mrem whole body dose equivalent criterion. This fraction, when added to the dose contribution from the Calvert Cliffs reactor operations is still within the 25 mrem criterion, provided the reactor operates within the numerical guides specified in Appendix I to 10 CFR 50. (See Section 2.2 of this SER and EA Section 6.2.)
Controlled area of an ISFSI (10 CFR 72.10b)
The controlled area boundary is greater than 100 meters from the ISFSI. The dose to an individual located at or beyond this boundary due to ISFSI accident is expected to be much less than the 5 rem criterion. (See Section 2.2.8 of this SER and EA Section 6.2).
Soent fuel and biah-level radioactive waste transoortation (10 CFR 72.108)
Because only spent fuel which has been irradiated in Calvert Cliffs reactors is to be stored in the ISFSI located on the Calvert- Cliffs site, there is no l _ environmental impact due to spent fuel being transported into the area. (See l Technical Specification of this SER.)
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o e issuance of license (10 CFR 72.40)
In accordance with 10 CFR Sections 72.40(a)(3) and 72.72(d), the NRC staff finds that operation of the ISFSI on the Calvert Cliffs site will not pose undue risk to the safe operation of the Calvert Cliffs reactors because:
(1) The ISFSI location is physically separated from the reactor ,
buildings by approximately 2,200 feet. (See Figure 1.1-1 of the SAR.)
(2) The ISFSI and the Calvert Cliffs reactors do not share any structures, systems or components that are important to safety.
(See BG&E SAR Section 1.0.)
The ISFSI will be operated as an integral part of the overall (3)
Calvert Cliffs operations. ISFSI operations can be scheduled and conducted so they do not conflict with reactor operations. (See Section 5 of this SER for operating controls and limits and BG&E SAR Sections 3.0, 4.0, 5.0, and 9.0) i in accordance with 10 CFR Section 72.40(a)(4), the NRC staff finds the applicant, by virtue of possessing a 10 CFR Part 50 operating license (No. 50-317 and 50-318), is qualified to conduct ISFSI operations covered by the regulations in 10 CFR Part 72. (See Section 1.5 of this SER.)
In accordance with 10 CFR Section 72.40(a)(5), the NRC staff finds that the applicant has adequate operating procedures to protect health and to minimize danger to life or property. (See Section 3.1 of this SER.)
In accordance with 10 CFR Section 72.40(a)(6) and Regulatory Guide 3.50, the NRC staff finds that BG&E is a regulated utility, and as such is financially qualified to operate an lSFSI on the Calvert Cliffs site in accordance.with 10 CFR Part 72.
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l e o in accordance with 10 CFR Section 72.40(a)(7), the NRC staff finds that the applicant has a quality assurance plan that complies with Subpart G of 10 CFR Part 72. (See Section 4.0 of this SER.)
In accordance with 10 CFR Section 72.40(a)(8), the NRC staff finds that the applicant has a physical protection plan that complies with 10 CFR Part 72 and Section 73.50 of 10 CFR Part 73. (See Section 3.4 of this SER).
In accordance with 10 CFR Section 72.40(a)(9), the NRC staff finds that the applicant has an existing personnel training program for employees at Calvert Cliffs, that when amended to include appropriate training modules covering ISFSI operations and cask handling procedures, will comply with Subpart I of 10 CFR Part 72.
In accordance with 10 CFR Section 72.40(a)(10), the NRC staff finds that the applicant's decommissioning plan provides reasonable assurance that at the end of its life the Calvert Clif fs ISFSI can be aecontaminated and decommissioned and that public health and safety will be adequately protected. (See Section 6.0 of this SER.)
In accordance with 10 CFR Section 72.40(a)(11), the NRC staff finds that the applicant's emergency plan for Calvert Cliffs when amended to include ISFSI emergencies, will comply with 10 CFR Section 72.32. (See Section 3.5 of this SER.)
In accordance with 10 CFR Section 72.40(a)(12), the NRC staff finds that the applicant has paid all appropriate licensing review fees as required by 10 CFR Part 170.
In accordance with 10 CFR Section 72.40(a)(13), the NRC staff finds that, based on its " Safety Evaluation Report related to the Topical Report for the NUTECH Horizontal Modular System for Irradiated Nuclear fuel NUH0MS-24P," its
" Environmental Assessment Related to the Construction and Operation of the Calvert Cliffs Nuclear Power Plant Independent Spent Fuel Storage Installation," and this safety evaluation, there is reasonable assurance that 7-14 l
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the activities authorized by a license to construct and operate the Calvert Cliffs Independent Spent Fuel Storage Installation can be conducted without I endangering the health and safety of the public and that these activities will be conducted in compliance with the conditions of the license, and the Commission's Regulations (10 CFR Part 72, 10 CFR Part 20, 10 CFR Part 50, and 10 CFR Part 73.)
In accordance with 10 CFR Section 72.40(a)(14), the NRC staff finds that issuance of a license to construct and operate the Calvert Cliffs Nuclear Power Plant Independent Spent fuel Storage Installation will not be inimical to the common defense and security because:
(1) The activities will be conducted within the jurisdiction of the United States. (See Section 2.1.1. of this SER.)
(2) The directors and principal officers of the applicant are citizens of the United States. (See Section 3.0 of this SER.)
(3) The applicant is not owned, dominated, or controlled by an alien foreign corporation or foreign government.
(4) The licensee may only receive and store fuel discharged solely from the Calvert Cliffs Nuclear Power Plant. (See the Technical Specifications for this SER.)
On the basis of the analysis presented in the NUTECH SAR for the NUHOMS-24P, BG&E FSAR, EA, SAR, the supplementary information presented in response to questions, and independent confirmatory analysis, and within the operating controls and limits specified in this SER it is concluded that the applicant meets all the applicable requirements of 10 CFR Part 72.
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8.0 REFERENCES
- 1. Baltimore Gas and Electric Company (BG&E), "Calvert Cliffs Independent Spent Fuel Storage Installation Safety Analysis Report," 1989. Docket Numbers 72-8 (50-317/318), and additional docketed supporting and modifying submittals as listed below. [ Note: Docketed additional material, such as design analyses, responses to NRC comments, drawings, and further modifications to these or the SAR are considered as elements of the SAR for the purposes of the SER review. Where multiple designs, criteria, procedures, etc., are included in different documents, that -
contained in the docketed document with the latest date is considered to represent the SAR.
- a. BG&E letter, dated November 1, 1990, with attachments.
- b. BG&E letter, dated December 20, 1990, with attachments,
- c. BG&E letter, dated August 18, 1992, with attachments.
- 2. BG&E, "Calvert Cliffs ISFS1 Environmental Report," 1989.
- 3. NUTECH Engineers Inc., NUH-002, " Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUHOMS-24P," Revision lA, July 1989, and additional docketed supporting
~
and modifying submittals.
- 4. 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 NUH0MS-24P, Submitted by NUTECH Engineers, Inc.," April 1989.
- 5. Duke Power Company, " Independent Spent Fuel Storage Installation, Safety Analysis Report," Oconee Nuclear Station, and additional docketed supporting and modifying submittals.
- 6. NRC, " Safety Evaluation Report for Duke Oconee Independent Spent Fuel Storage Installation," June 1989.
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- 7. NUTECH Engineers, Inc., " Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985, and additional docketed supporting and modifying submittals.
- 8. NRC, " Safety Evaluation Report for NUTECH Horizontal Modular Storage Systems for Irradiated Fuel, (NUHOMS-7P)," March 1986.
- 9. Carolina Power and Light Company, " Independent Spent Fuel Storage Installation SAR," Revision 1,1986, and additional docketed supporting --
and modifying submittals.
- 10. NRC, " Safety Evaluation Report of H.B. Robinson Steam Electric Plant Unit No. 2, Independent Spent Fuel Storage Installation," June 1986.
- 11. Carolina Power and Light Company, " Brunswick Steam Electric Plant independent Spent Fuel Storage Installation Safety Analysis Report,"
August 1989, and additional docketed supporting and modifying submittals.
- 12. BG&E, "Calvert Cliffs Independent Spent Fuel Storage Installation License Application," December 21, 1989, Docket 72-8, 50-317/318. _
- 13. U.S. Nuclear Regulatory Commission, " Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation," Title 10 Code of Federal Regulations, Part 72, Office of the Federal Register, 55 FR 29181, July 18, 1990.
- 14. BG1E, "Calvert Cliffs Nuclear Power Plant Updated Final Safety Analysis Report," Docket No. 50-317/318, 8-2
- 15. U.S. Government, " Domestic Licensing of Production and Utilization i Facilities, Appendix 8 - Quality Assurance Criteria for Nuclear Power >
Plants and Fuel Reprocessing Plants," Title 10 of Code of Federal -
Regulations, Part 50, Office of the Federal Register. 1
- 16. U.S. Nuclear Regulatory Commission, " Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage Installation (Dry Storage)," Regulatory Guide 3.48, October 1981.
- 17. Pacific Northwest Laboratory, " Heat Transfer and Shielding Performances of the H.B. Robinson NUHOMS-7P Spent Fuel Dry Storage Modules," May 1990.
- 18. Crane Manufacturer's Association of America, CHAA Specification 70,
" Specifications for Electric Overhead Traveling Cranes," 1988.
- 19. American Society of Mechanical Engineers, " Boiler and Pressure Vessel Code," Section 111, Subsection NC,'1983,
- 20. American National Standards Institute, ANSI N14.6-1986, "Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More," 1987.
- 21. American Institute for Steel Construction, (AISC), Manual of Steel
[.onstruction, Ninth Edition, 1989.-
1
- 22. American Welding Society (AWS) Dl.1-88, " Structural Welding Code -
Steel," 1988.
- 23. U.S. Government, " Standards for Protection Against Radiation," _ Title 10 of the Code of Federal Regulations, Part 20, Office of the Federal Register.
- 24. American Concrete Institute, ACI-349-85, " Code Requirements for Nuclear Safety Related Structures," 1985.
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- 25. U.S. Nuclear Regulatory Commission. " Design Basis Tornado for Nuclear Power Plants," Regulatory Guide 1.76, April 1974. l
- 26. U.S. Nuclear Regulatory Commission, " Standard Review Plan," NUREG-0800, Revision 2.
l
- 27. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.60, " Design l Response Spectra for Seismic Design of Nuclear Power Plants, Rev 1,"
December 1973.
- 28. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants," October 1973.
- 29. Mark Fintel, Handbook of Concrete Enaineerina, Van Nost, and Reinhold Co., NY, 1974.
- 30. U.S. Nuclear Regulatory Commission, " Safety Evaluation Report of Pacific Sierra Nuclear Topical Report on the Ventilated Storage Cask System for Irradiated Fuel Revision 2," March 1991.
- 31. U.S. Nuclear Regulatory Commission, " Safety Evaluation Report for Public Service Company of Colorado's Safety Analysis Report-for Fort St. Vrain Independent Spent Fuel Storage Installation," October 1991.
- 32. 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 1981.
- 33. U.S. Nuclear Regulatory Commission, " Fracture Toughness Criteria for Ferritic Steel Shipping Cask Containment Vessels With A Haximum Wall Thickness of Four Inches," Draft Regulatory Guide DG-7001, July 1989.
4 34. U.S. Nuclear Regulatory Commission, " Recommended Welding Criteria For Use in the Fabrication of Shipping Containers for Radioactive Materials," NUREG/CR-3019, March 1985.
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l
- 35. U.S. Nuclear Regulatory Commission, " Load Combinations for the Structural Analysis of Shipping Casks," Regulatory Guide 7.8, May 1977.
- 36. U.S. Nuclear Regulatory Commission, " Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels," Regulatory Guide 7.6, ,
Revision 1, March 1978.
- 37. 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 38).
- 38. American National Standards institute, ANSI 57.9-1984 " Design Criteria for an independent Spent Fuel Storage Installation (Dry Storage Type)."
- 39. American National Standards Institute, ANSI N14.5, " Radioactive Material Leakage Test on Packaging for Shipments," 1987.
- 40. Swanson Analysis Systems, Inc., ANSYS Enaineerina Analysis Systems User's Manual, Version 4.4, Volumes 1 and 2, Pittsburgh, PA.
- 41. Electric Power Research Institute, "The Effects of Target Hardness on the Structural Design of Concrete Storage Pads for Spent fuel Casks,"
NP-4830, October 1986.
- 42. Warrant, M. and Joseph, J., " Test Data Report for Quarter Scale NUPAC 125-8 Cask Model," Report No. GEND-INF-091, Sandia National Laboratories, February 1988.
- 43. U.S. Nuclear Regulatory Commission, " Control of Heavy Loads at Nuclear Power Plants," NUREG-0612, July 1980.
- 44. ANSI ASME N0G-1-1983, " Rules for Construction of Overhead and Gantry Cranes," 1983.
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- 45. U.S. Nuclear Regulatory Comission, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As is Reasonably Achievable," Regulatory Guide 8.8, Revision 3, June 1976.
- 46. U.S. Nuclear Regulatory Comission, " Operating Philosophy for !
Maintaining Occupational Radiation Exposures As Low As Is Reasonably J Achievable," Regulatory Guide 8.10, May 1977.
- 47. U.S. Nuclear Regulatory Commission, " Qualification and Training of Personnel for Nuclear Power Plants," Regulatory Guide 1.8. Revision 2, 1987.
i.
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APPENDIX A CHRONOLOGY OF PRINCIPAL ACTIONS FOR DOCKET 72-8 December 21, 1989 Creel, BG&E, submits application to construct and operate ISFSI at Calvert Cliffs including:
(1) ISFSI SAR (proprietary supplement withheld)
(2) ISFSI Environmental Report February 2, 1990 Notice is forwarded from Sjoblom, NRC, to Creel, BG&E, regarding consideration of issuance of material license for storage of spent fuel and opportunity for hearing on December 21, 1989, request to allow storage of spent fuel in dry storage concrete module system at ISFSI.
February 20, 1990 Roberts, NRC, forwards request to Creel, BG&E, for additional information in response to the QA program for the ISFSI submitted with the December 21, 1989, application.
February 21, 1990 Cunningham, NRC, denies the December 21, 1989, request from Creel, BG&E, for withholding information regarding calculation data, results of design ' calculations &
component design details & critical dimensions from public disclosure per 10 CFR 2.790. Information not proprietary and will be placed in PDR.
March 23, 1990 Creel, BG&E, forwards to NRC proprietary Revision 1 to "lSFSI SAR" with amended affidavit. Revision withheld.
April 26, 1990 Creel, BG&E, forwards proprietary calculations and drawings regarding the design of the ISFSI- to the NRC per January 23, 1990 request. Nineteen oversize drawings and calculations withheld.
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l June 6, 1990 Letter from Roberts, NRC, to Creel, BG&E, requesting i proprietary drawings 84-Oll-E, 84-012-E, and 84-013-E, Rev 0, omitted from April 26, 1990, submission.
June 14, 1990 Roberts, NRC, confirms with Creel, BG&E, the planned visit on June 26 and 27, 1990, of F. Sturz to the plant in connection with the environmental review of ISFSI per June 8, 1990 telecon.
June 29, 1990 Creel, CG&E, forwards proprietary drawings BGE-84-Oll/BGE-84-013 and'Rev 3 to calculation BGE001.0202 w/ll oversize drawings to the NRC.
July 23, 1990 Sturz, NRC, forwards comments on SAR for Calvert Cliffs ISFSI to Creel, BG&E.
July 26, 1990 Creel, BG&E, forwards " Decommissioning Plan for Calvert Cliffs ISFSI" to the NRC.
August 3, 1990 Cunningham, NRC, advises Creel, BG&E, ti..t listed documents marked proprietary will be withheld from public disclosure per March 23 request and 10 CFR 2.790(b)(5).
October 16, 1990 Creel, BG&E, advises NRC of comments on SAR and environmental report and requests meeting to discuss pruposed responses.
November 1, 1990 Denton, BG&E, responds to NRC comments on environmental issues regarding BG&E license application for facility ISFSI. Enclosed proprietary withheld.
December 20, 1991 Creel, BG&E, responds to NRC July 23, 1990, comments on SAR regarding application for ISFSI. Enclosed proprietary withheld.
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1 February 12, 1991 Creel, BG&E, responds to NRC February 2, 1990 request for additional information regarding QA program for plant ISFSI. l March 22, 1991 Haughney, NRC, forwards to Creel, BG&E, FR notice of issuance of environmental assessment and finding of no significant impact regarding December 21, 1989, application to construct and operate dry storage ISFSI at Calvert Cliffs.
July 21, 1991 Creel, BG&E, requests the NRC provide schedule for completion of reviews of final issuance of OL for ISFSt.
Utility is proceeding with construction of ISFSI at own risk to meet intended goal of avoiding loss of full core reserve after Spring 1994 refueling outage.
August 2, 1991 Sturz, NRC, responds to Creel's July 21 letter requesting schedule for completing review of ISFSI application.
August 19, 1991 Sturz, NRC, forwards comments to Creel, BG&E, and requests additional information regarding December 20, 1990, response to NRC questions on ISFSI.
September 30, 1991 Creel, BG&E, responds to NRC August 19, 1991, follow up comments on SAR in support of license application for ISFSI. Enclosed proprietary withheld.
October 18, 1991 Creel, BG&E, forwards "Calvert Cliffs ISFSI SAR Chapter 11
'QA'" in response to NRC requests of December 21, 1989, and February 12, 1991, requests for additional information.
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, o November 15, 1991 Creel, BG&E, responds to October 17, 1991, request for additional information regarding the July 2,1991, single-failure-proof upgraded spent fuel cask handling crane TS change request. Response spectrum seismic analysis and stress analysis of main and auxiliary trolley girders enclosed.
November 21, 1991 Sturz, NRC, forwards to Creel, BG&E, request for additional information to enable NRC to review new Chapter 11, "QA" to ISFSI SAR, submitted via October 18, 1991, letter.
December 12, 1991 Sturz, NRC, forwards to Creel, BG&E, a request for additional information to enable NRC to continue review of September 30, 1991, license application for ISFSI covering forest fire analysis and DSC drop analysis.
December 19, 1991 Creel, BG&E, forwards response to NRC November 21, 1991 request for additional information regarding expanded version of QA program (Chapter 11 of SAR) for proposed ISFSI transmitted via October 18, 1991, letter.
December 19, 1991 Sturz, NRC, forwards to Creel, BG&E, request for additional structural analyses for transfer cask drop accident to enable NRC to continue review of September 30, 1991 application for ISFSI.
December 20, 1991 Site visit and meeting with BG&E at Calvert Cliffs.
December 27, 1991 Creel, BG&E, forwards nonproprietary and proprietary responses to NRC requests for additional information on SAR for utility license application for ISFSI. Proprietary enclosures withheld.
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March 3, 1992 Heeting with BG&E to discuss proposed technical specifications.
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July 28, 1992 Sturz, NRC, forwards request to Creel, BG&E, for additional j decomissioning plan information. l August 18, 1992 Creel, BG&E, forwards Revision to ISFSI Decomissioning Plan.
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- f %, UNITED STATES y- g NUCLEAR REGULATORY COMMISSION s l WASHING TON, D. C. 20555
'% , , , , , ,./ March 22, 1991 Docket No. 72-8 (50-317/318)
Baltimore Gas and Electric Company ATTN: George C. Creel Vice President Nuclear Energy Charles Center P. O. Box 1475 Baltimore, MD 21203 Gentlemen:
The Nuclear Regulatory Commission, Office of Nuclear Material Safety _and Safeguards, Division of industrial and Medical Nuclear Safety, has issued an environmental assessment in connection with your application, December 21, _
1989, for at,thority to construct and operate a dry storage independent spent fuel storage in:,tallation to be located on the Calvert Cliffs Nuclear Power Plant site in Calvert County, Maryland. We are enclosing copies of the Notice of Issuance and Finding of No Significant Impact, which has been forwarded to the Office of the Federal Register for publication, and the " Environmental Assessment Related to the Construction and Operation of the Calvert Cliffs Independent Spent Fuel Storage Installation."
Sincerely, fl ,
i ' I f.
Charles'J. Ha ney, Chi f ",
Fuel Cycle Safety Br.anc Division of Industrial anc Medical Nuclear Safety
Enclosures:
- 1. Notice of Issuance and Finding of No Significant Impact
- 2. . Environmental Assessment d
cc: Attached list
' Enclosure 3. - .
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o e Mr. G. C. Creel Calvert Cliffs Nuclear Power Plant Baltimore Gas & Electric Company Unit Nos. I and 2 cc:
Mrs. Mary M. Krug, President Mr. Joseph H. Walter Calvert County Board of Engineering Division Commissioners Public Service Commission of Maryland Prince Frederick, liaryland 20678 American Building 231 E. Baltimore Street Baltimore, Maryland 21202-3486 D. A. Brune, Esq. Ms. Kirsten A. Burger, Esq.
General Counsel Maryland People's Counsel Baltimore Gas and Electric Company American Building, 9th Floor P. O. Box 1475 231 E. Baltimore Street Baltimore, Maryland 21203 Baltimore, Maryland 21202 Mr. Jay E. Silberg, Esq. Ms. Patricia Birr.1e Shaw, Pittuan, Pvtts ar.d Trowbridge Co-Director 2300 N Street. NW Maryland Safe Energy Coalition Washington, DC 20037 P. O. Box 902 Columbia, Maryland- 21044 Mr. G. L. Detter, Director, NRM Calvert Cliffs Nuclear Power Plant MD Rts 2 & 4, P. O. Box 1535 Lusby, Maryland 20657 Resident Inspector c/o U.S. Nuclear Regulatory Commission P. O. Box 437 Lusby, fiarylar.d 20657 Mr. Richard I. McLean Administrator - Radioecology Department of Natural Resources ,
580 Taylor Avenue Tawes State Office Building ,
PPER B3 Annapolis, Maryland 21401 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 1
o ,
7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO 72-8 (50-317/318)
BALTIMORE GAS AND ELECTRIC COMPANY NOTICE OF ISSUANCE OF ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT FOR '4E CALVERT CLIFFS INDEPENDENT SPENT FUEL STORAGE INSTALLATION AT THE CALVERT CLIFFS NUCLEAR POWER PLANT The U.S. Nuclear Regulatory Commission (the Commission) is considering issuance of a materials license under the requirements of Title 10 of the Code of Federal Regulations, Part 72 (10 CFR Part 72), to Baltimore Gas and Electric Company (BG&E or the Applicant), authorizing receipt and storage of spent fuel in an independent spent fuel storage installation (ISFSI) located onsite at its Calvert Cliffs Nuclear Power Plant in Calvert County, Maryland. The Commission's Office of Nuclear Material Safety and Safeguards, Division of Industrial and Medical Nuclear Safety, has completed its environmental review in support of the issuance of a materials license. The " Environmental Assessment (EA) Related to the Construction and Operation of the Calvert Cliffs Independent Spent Fuel Stora0e Installation" has been issued in accordance with 10 CFR Part 51.
Description of the Proposed Action: The proposed licensing action would authorize the Applicant to construct and operate a dry storage ISFSI.
The function of the ISFSI is to provide interim storage for up to 2880 fuel assemblies from Calvert Cliffs Units 1 and 2. Twenty-Four. fuel l
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2-l assemblies are stored in an inert atmosphere inside a stainless steel canister '
which provides confinement, shielding, critica ity control and heat removal.
Spent fuel loading and canister preparation takes place within the Calvert Cliffs reactor buildings. The canister is then transported inside a transfer cask to the onsite ISFSI where the canister is placed inside a concrete horizontal storage module (HSM) which provides additional shielding. Up to a total of 120 storage modules would be required under the requested license.
Need for the Proposed Action: After reracking, the spent fuel storage pools at the Calvert Cliffs have reached their maximum capacity. Design limitations preclude further such expansions. The Calvert Cliffs Units 1 ,
and 2 spent fuei pools are expected to exceed prudent operating reserve capacity in April 1993 and are expected to exceed their ability to hold an entire of floaded reactor core in April 1995. Additional spent fuel is being generated as the two Calvert Cliffs reactors continue to operate, and additional storage capacity will be required in order to recover and maintain a prudent operating reserve of spent fuel storage capability.
The proposed action would provide the additional capacity required to store spent fuel expected to be generated at the Calvert Cliffs Nuclear Power Plant through the year 2016, and maintain prudent operating reserve-storage capacities.
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Environmental Impacts of the Proposed Action: As discussed in the EA, no significant impacts from construction of the ISFSI are anticipated.
Similarly, no significant impacts are expected from ISFSI operations.
The activities will affect only about 6 acres of land area on the-Calvert Cliffs site, resulting in a loss cf less than 1 percent of the biological production onsite. Construction will be phased acrording to storage requirements. Potential impacts from fugitive dust, erosion, and noise levels which are typical for the planned construction activities are minimal, and with good construction practices will be controlled to insignificant levels.
The radiological impacts from liquid and gaseous effluents arising from cask loading and preparation are minimal. They fall within the scope of impacts evaluated for licensed reactor operations and are controlled by the existing Calvert Cliffs reactor technical specifications. The primary radiological exposure pathway associated with ISFSI operation is-direct irradiation of nearby residents and site workers. The highest dose to the nearest resident for any year is about 0.2 mrem, which is well within the 25 mrem /yr requirement of 10 CFR 72.104. The highest collective dose commitment for any year' to the population within 2 miles of the ISFSI is expected to be less than 0.08 person rem. Dose levels beyond this distance are insignificant. The radiological impacts from the postulated worst-case accident at the Calvert Cliffs ISFSI are within the 5 rem criteria for whole body and organ doses to an individual, as l
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. o specified in 10 CFR 72.106(b), and are less than the Environmental Protection Agency Protective Action Guides for individuals exposed to radiation as a result of accidents.
Alternatives to the Proposed Action:
5several alternatives were discussed in the EA, but none sufficiently met the spent fuel storage requirements for the Calvert Cliffs Nuclear Power Plant. The alternative of no action would force a shutdown of power generation at the Calvert Cliffs Nuclear Power Plant. The " Final Generic Environmental Impact Statement (FGEIS) on Handling and Storage of Spent Lfght Water Power Reactor Fuel," NUREC-0575, found that ISFSIs represent a majer means of interim storage ~at a reactor site. While the_ environmental impacts of the dry storage ISFSI option were not specifically addressed in the FGEIS (storage of light-water-cooled-power reactor spent fuel in a water pool was spe .lly addressed), the FGEIS concluded that the use of alternative dry passive storage techniques for aged fuel appeared to be equally feasible and environmentally accepta31e, although environmental impacts need to be considered on a site-specific basis. -
Because the Commission has cor.cluried there are no significant environmental
'mpacts associated with the proposed action, any alternative of equal or greater environmental impacts need not be evaluated.
0
- 5-Alternative Use of Resources: The only resources committed irretrievably and not previously considered in environmental documents relating to the Calvert Cliffs Nuclear Power Plant are the steel, concrete, and c:her construction materials used in the canisters and HSMs for the ISFSI.
Agencies and Persons Contacted: Outside 0; cies that were contacted in connection with the preparation of the . w .samental Assessment include:
the Power Plant and En,' "nmental Review Division, Manyland Department of Natural Resources, and several Calvert County Governmental Departments.
Finding of No Significant Impact: In summary, the ISFSI is located on--
the previously disturbed Calvert Cliffs Nuclear Power Plant site within the confines of the exclusion area and will require only a minor commitment of land resources. The proposed action will cause no release of effluents, and there will be no significant increases in individual and collective radiation doses to either the public or onsite workers. Potential offsite impacts from a postulated worst-case credible accident are 9 small fraction of the regulatory limits of 10 CFR 72.106, and well below+the Environmental Protection Agency's Protective Action Guides. Therefort, the proposed action will at significantly affect the quality of the human environment.
Accordingly, pursuant to the requirements cf 10 CFR 51.31 and 10 CFR 51.32, the Commission has determined that a finding of no significant impact is appropriate and that an environmental impact statement (EIS) need not be prepared for the issuance of a materials license for the Calvert Cliffs ISFSI.
. o The EA for the proposed action, on which this Finding of No Significant Impact is based, relied upon several environmental documents, with independent assessment of data, analyses and results. The following documents were utilized: (1) " Final Environmental Statement Related to Operation of Calvert Cliffs Nuclear Power Plant, dated April 1973; (2)
"Calvert Cliffs Independent Spent Fuel Storage Installation Environmental Report," dated December 1989; (3) Supplemental Information to the "Calvert Cliffs Independent Spent Fuel Storage Installation Environmental Report,"
BG&E, dated November 1, 1990; and (4) " Final Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel,"
NtlREG-0575, Volumes 1-3, August 1979.
The EA and other documents related to this proposed action are available for public inspection and for copying for a fee at the NRC Public Document Room, 2120 L Street, NW., Washington, D.C. 20555, and at the Local Public Document Room at the Calvert County Library, Fourth Street, Prince Frederick, Maryland, 20678.
Dated at Rockville, Maryland, this M day of March, 1991.
FOR THE NUCLEAR REGULATORY COMMISSION Charles J. Ha ghney, C ef a Fuel Cycle Safety Branch Division of Industrial and Medical Nuclear Safety Office of Nuclear Material Safety and Safeguards i
l
^ ,O. _e, U S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR HATERIAL SAFETY AND S/JEGUARDS ENVIRONMENTAL ASSESSMENT RELATED TO CONSTRUCTION AND OPERATION OF THE CALVERT CLIFFS INDEPENDENT SPENT FUEL STORAGE INSTALLATI03 DOCKET NO 72-8 (50-317, -318)
BALTIMORE GAS AND ELECTRIC COMPANY March 1991 5
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T e TABLE OF CONTENTS Page
- 1. 0 INTRODUCTION ............................................... 1 1.1 Description of the Proposed Action .................... 1
- 1. 2 Background Information ................................ 5 1.3 Previous Environmental Assessments and Supporting Documents .................................. 6 2.0 NEED FOR THE PROPOSED ACTION ................. ............, 7
- 3. 0 ALTERNATIVES .................. ............................ 8 4.0 ENVIRONMENTAL INTERFACES ................................... 10 4.1 Site Location, Land Use and Terrestrial Resources ..... 10 4.2 Water Use and Aquatic Resources ....................... 11 .
4.3 Socioeconomics and Historical, Archeological, and Cultural Resources ........................ ....... 11 4.4 Demography ..... ................ ..................... 11 4,5 Meteorology ............... ........................... 12 4.6 Geology and Seismo?ogy ................................ 12
5.0 DESCRIPTION
OF CALVERT CLIFFS NUCLEAR PLANT ISFSI .......... 14 5.1 General Description ............. ..................... -14 5.2 ISFSI Design .......................................... 14 5.3 ISFSI Operation ....................................... 16 5.4 Monitoring Program .. ................................. 18 6.0 ENVIRONMENTAL IMPACTS OF PROPOSED ACTION ...... ............ 20 6.1 Construction Impacts .................................. 20 6.1.1 Land Use and Terrestrial Resources ............. 20 6.1.2 Water Use and Aquatic Resources ................. 20 6.1.3 Other Impacts of Construction................... 20-6.1. 4 Socioeconomics ................................. 21 6.1. 5 Radiological Impacts from Construction ......... .21 6.2 Operational Impacts ................................... 21 6.2.1 Radiological Impacts from Routine Operations ... 21 6.2.1.1 Offsite Dose .......................... 22 6.2.1.2 Collective Occupational Dose .......... 23-6.2.2 Radiological Impacts of Accidents ....... ...... 24 6.2.3 Nonrediological Impacts ........................ 29 6.2.3.1 Land Use and Terrestrial Resources..... 29 6.2.3.2 Water Use and Aquatic Resources . . . . . . . 29' 6.2.3.3 Other Impacts of Operation . . . . . . . . . . . . 30 i
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TABLE OF CONTENTS (Continued)
Pagl,e 7.0 SAFEGUARDS FOR SPENT FUEL .................................. 31 8.0 DECOMMISSIONING ............................................ 32 9.0
SUMMARY
AND CONCLUSIONS .............. ..................... 33 ,
9.1 Sur.na ry o f Env i ronmental Impacts . . . . . . . . . . . . . . . . . . . . . . 33 9.2 Basis for Finding of No Significant Impact ............ 33
10.0 REFERENCES
................................................. 34 -
11.0 LIST OF AGENCIES AND PREPARERS ......................... ... 35 FIGURES 1.1 Calvert Cliffs Site Plan ................................... 2 1.2 ISFSI Layout ............................................... 3 1.3 NUHOMS-24P Horizontal Storage Module ....................... 4 TABLES 5.1 Design Parameters for the Calvert Cliffs ISFSI .............. 15
- 5. 2 Major Operational Steps for Transferring the Fuel from the Spent Fuel Pool to the ISFSI .................... 17 6.1 Collective Occupational Dose to Calvert Cliffs on Workers Directly Involved in ISFSI Activities . . . . . . . . . . . . 25 6.2 Estimate of Collective Dose to Calvert Cliffs Station '
Workers Not Directly Involved in ISFSI Activities ........ 26 6.3 Expected Dose at the Controlled Area Boundary nesulting from a Dry Shielded Canister Leakaje Accident at the Calvert Cliffs Nuclear Power Plant ....................... 27 6.4 Expected Dose at the Nearest Resident Resulting from Dry Shielded Canister Leakage Accident at the Calvert Cliffs Nuclear Power Plant ....................... 28 iii
b o ENVIRONMENTAL ASSESSMENT RELATED TO THE CONSTRUCTION AND OPERATION OF THE CALVERT CLIFFS INDEPENDENT SPENT FUEL STORAGE INSTALLATION
1.0 INTRODUCTION
1.1 DESCRIPTION
OF THE PROPOSED ACTION By letter dated December 21, 1989, Baltimore Gas and Electric Company (the Applicant) submitted an application for a license to construct and operate a dry independent spent fuel storage installation (ISFSI) to be located on the Calvert Cliffs Nuclear Power Plant site in Calvert County, Maryland. The ISFSI or some other spent fuel storage system is needed in order to maintain a prudent operating reserve of spent fuel storage capacity in the two spent fuel pools at the Calvert Cliffs site. This Environmental Assessment (EA) addresses the expected environmental impacts associated with the proposed construction and operation of the ISFSI on the Calvert Cliffs Nuclear Power Plart site.
Baltimore Gas and Electric Company owns and operates two 2700 MWt nuclear generating units t the Calvert Cliffs Nuclear Power Plant. The proposed ISFSI will be located approximately 2300 feet southwest of the power plant abou6 70 feet above the existing plant yard elevation. Figure 1.1 shows the location of the proposed ISFSI relative to the other features on the site including the reactor buildings and security fence. Figure 1.2 provides addi-tional detail on the ISFSI layout.
The proposed ISFSI is a system designed by Pacific huclear Fuel Services, Inc. (formerly Nutech, Inc. ,) of San Jose, California. It is referred to as the Nutech Horizontal Modular Storage System or NUH0MS-24P. The major components of this system are a dry shielded canister (DSC), a transfer cask, and a horizontal storage module (HSM). The DSC is placed inside the transfer cask, filled with 24 assemblies in the spent fuel pool, dewatered, inerted, sealed, decontaminated, and transferred to the storage arua in the shielded transfer cask. Once in the storage area, the DSC is removed from the shielded transfer cask and placed into the HSM which provides bulk shielding and passive, natural convection heat removal. Figure 1.3 illustrates the DSC and HSM of the proposed Calvert Cliffs ISFSI.
The Calvert Cliffs ISFSI is designed to operate for 50 years, well beyond the operating life of the two reactors. Licenses issued for ISFSIs under Part 72 Title 10 of the Code of Federal Regulations (10 CFR Part 72) are for 20 years, but the licensee may seek to renew the license, if necessary, prior to its expiration.
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1.2 BACKGROUND
INFORMATION Both Calvert Cliffs units were granted their construction permits in November 1969. The first unit began commercial operation in May 1975. The second unit began commercial operation in April 1977. Prior to the mid 1970's, the nuclear industry in general and the Calvert Cliffs Nuclear Power Plant in particular, planned to store, for an interim period, spent fuel from nuclear power reactors in a spent fuel pool at the reactor site where it was generated. After an indefinite interim storage period, utilities anticipated that spent fuel would be transported to a reprocessing plant for recovery and recycling of fuel materials. Reactor facilities, such as the Calvert Cliffs units, were not designed to provide spent fuel storage capacity for life-of plant operations.
Because commercial reprocessing did not develop as anticipated, the Nuclear Regulatory Commission (NRC), in 1975, directed the staff to prepare a generic environmental impact statement (EIS) on spent fuel storage. The Commission directed the staff to analyze alternatives for the handling and storage of spent fuel from light water power reactors with particular emphasis on developing long range policy. The staff also considered the consequences of restriction or termination of spent fuel generation throucn nuclear power plant shutdown. A " Final Generic Environmental Impact Statement (FGEIS) on Handling and Storage of Spent Light Water Power Reactor Fuel, NUREG-0575,2 was issued by NRC in August 1979.
In the FGEIS, the storage of spent fuel is considered interim storage until the issue of permanent disposal is resolved and a plan implemented. Interim storage options evaluated in detail and included in the FGEIS are: (1) onsite expansion of spent fuel pool capacity; (2) expansion of spent fuel pool storage capacity at reprocessing plants; (3) use of ISFSIs; (4) transshipment of spent fuel between reactors; and (5) reactor shutdowns or deratings to terminate or reduce the amount of spent fuel generated.
The FGEIS concluded that an ISFSI represents the major means of interim storage at a reactor site once the spent fuel pool capacity has been reached. The FGEIS supports findings that the storage of light water cooled power reactor (LWR) spent fuels in water pools, whether at-the-reactor or away-from-reactor sites, has an insignificant impact on the environment. While the environmental impacts of the dry storage option were not specifically addressed in the FGEIS, the use of alternative dry passive storage techniques for aged fuel appeared to be equally feasible and environmentally acceptable. In the case of both dry passive storage and wet storage, environmental impacts need to be considered on a site-specific basis.
The onsite expansion of spent fuel pools has been used by most utilities. The NRC has reviewed and approved more than 120 onsite spent fuel pool capacity expansions through veracking modifications since issuance of the FGEIS. At the Calvert Cliffs Nuclear Power Plant, both spent fuel pools have been reracked increasing the pool capacity to its maximum. Design limitations preclude further expansion, thus eliminating this as a viable option for meeting increased storage needs.
5
o 4 Baltimore Gas and Electric Company, therefore, proposes to solve the problem of inadequate spent fuel storage capacity at its Calvert Cliffs Station through the construction of an onsite ISFSI. As required by 10 CFR Part 72, this assessment addresses the site-specific environmental impacts of construction and operation of the dry storage ISFSI at the Calvert Cliffs Nuclear Power Plant site.
1.3 PREVIOUS ENVIRONMENTAL ASSESSMENTS AND SUPPORTING DOCUMENTS Several environmental documents have been prepared specific to the Calvert Cliffs Nuclear Power Plant site. A Final Environmental Statement (FES) related to the operation of the Calvert Cliffs Nuclear Power Plant was prepared by the U.S. Atomic Energy Commission in 1973.2 This document relied on information supplied by Baltimore Gas and Electric Company in the its Environmental Report (ER) related to the proposed ISFSI for the Calvert Cliffs Nuclear Power Plant submitted with the application in December 19893 and supplementary information was submitted in response to NRC questions in October 1990.4 This EA is tiered on the 1973 FES, the 1990 ER with supplementary information, and the FGEIS (NUREG-0575). Additional information used in this assessment is provided in the applicant's Final Safety Analysis Report (SAR) for the operation of the Calvert Cliffs Nuclear Power Plant,5 the SAR for the proposed ISFSI6, and the Nutech, Inc., " Topical Report for the Nutech Horizontal Modular Storage System for Irradiated Nuclear Fuel:
NUHOMS-24P."7 6
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2.0 NEED FOR PROPOSED ACTION The two spent fuel pools at the Calvert Cliffs Nuclear Power Plant are shared between Units 1 and 2. The combined capacity of the two pools was originally 400 storage positions, but with the reracking efforts, the capacity of the pools has been increased to 1830 storage positions. Design limitations preclude any further pool storage capacity increases.
Prudent operating practice provides for sufficient spent fuel storage capacity to accommodate a full core off-load (217 assemblies), and allows safe diver access for maintenance during a refueling outage. The prudent operating reserve for the Calvert Cliffs pools will be lost in April 1993. The ability to offload an entire reactor core will be lost in April 1995.
Additional spent fuel is being generated as the units continue to operate, and -
additional storage capacity will be required in order to recover and maintain the prudent operating reserve of spent fuel storage capacity. The proposed action would provide the additional capacity required to store spent fuel expected to be generated at the Calvert Cliffs Nuclear Power Plant through 2016, the end of its currently licensed operating life, if needed.
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o 4 3.0 ALTERNATIVES Baltimore Gas and Electric Company evaluated a number of alternatives for the storage of spent nuclear fuel prior the selection of the dry storage ISFSI.
The alternatives did not sufficiently meet the requirements for storage of spent nuclear fuel generated at the Calvert Cliffs Nuclear Power Plant. A brief discussion of these alternatives follows.
Permanent Federal Repository If a permanent Federal repository were available, the preferred alternative would be to ship spent fuel to the repository for disposal. The Department of Energy (00E) is currently working to develop a repository as required under the Nuclear Waste Policy Act (NWPA), but is not likely to have a licensed repository ready to receive spent fuel before 2010. Although the Department -
of Energy recommended that a Monitored Retrievable Storage (MRS) facility be constructed and in operation by 1998, the NWPA prohibits siting an MRS before obtaining a construction permit for the repository. Given the uncertainties of schedules for a repository and MRS, this alternative, therefore, does not meet the near-term storage needs of the Baltimore Gas and Electric Company.
Reracking of the Calvert Cliffs Spent Fuel Pools As discussed in Section 1.2, by reracking the spent fuel pools at the Calvert Cliffs Nuclear Power Plant, the pools have reached their maximum capacity.
Transshipment to other Nuclear Reactor Sites Because of uncertainties in the timing of fuel acceptance for Federal storage and disposal under the Nuclear Waste Policy Act, most utilities are expected to face spent fuel storage shortfalls and are expected to be unwilling to reduce their own storage capacity. Therefore, this option is not considered to be viable.
Full Scale Rod Consolidation Baltimore Gas and Electric Company has been involved in the development of fuel rod consolidation. While disassembling the intact fuel assemblies and reconfiguring the fuel rods into a close packed array in the existing pool could expand storage capacity, it would only extend full core reserve capabilities to the year 2002. However, this alternative appears less attractive than that of the ISFSI because of technology unca.tainties, current consolidation rates, potentially higher worker radiation exposure, and uncertainties about DOE acceptance of non-fuel bearing components that would be generated by consolidation.
Construction of a New Independent Storage Pool Additional storage capacity could be achieved by building a new spent fuel storage pool similar to that existing at the plant site. The NRC has generically assessed the impacts of this alternative and found that "the storage of LWR spent fuels in water pools has an insignificant impact on the environment."1 In contrast with the proposed action, the higher costs for commissioning, operating, maintaining, and decommissioning a new pool storage facility make this alternative less preferable.
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i Other Dry Storage Technologies Other dry spent fuel storage, technologies such as metal casks, concrete casks, and modular dry vault storage, could provide additional storage capacity. These other technologies would be expected to have no significant environmental impact similar to that for the proposed action. However because of uncertainties and cost differentials these other technologies were not considered acceptable to BG&E.
Shipment to a Reprocessing Facility There are no existing commercial reprocessing facilities in the United States, nor is there the prospect for one in the foreseeable future. While there are reprocessing facilities in operation in the United Kingdom, Japan, Germany, and France, in the near term, the political, legal and logistical uncertainties associated with trying to ship spent fuel overseas make this alternative not viable.
No Action {
If no action were taken to provide additional spent fuel storage capacity, the Calvert Cliffs Nuclear Power Plant will be unable to continue operation beyond 1995. This would result in the elimination of 20 years of facility operation, the impacts of curtailing the generation of spent fuel by ceasing operation of existing nuclear power plants when their spent fuel pools become filled was evaluated and found to be undesirable.2
)
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. 4 4.0 ENVIRONMENTAL INTERFACES The general environment around the Calvert Cliffs Nuclear Power Plant is well characterized as a result of the studies conducted in support of construction of the Calvert Cliffs Nuclear Power Plant. This section briefly reviews the environment with emphasis on those envirnnmental features that are most likely to be affected by the construction and operation of the ISFSI. The assessment of construction and operational impacts is presented in Chapter 6.
4.1 SITE LOCATION, LAND USE AND TERRESTRIAL RESOURCES The proposed ISFSI will be located within the existing plant site area for the Calvert Cliffs Nuclear Power Plant. The area is wooded with Virginia and Loblolly Pines. The total storage area associated with the full 120 modules is estimated to be 676 feet by 236 feet. There will be a access 30-feet wide access off an existing road which leads from the developed portion of the site to the proposed ISFSI storage area. The total area developed for the ISFSI will be slightly more than 6 acres within the 2300 acre Calvert Cliffs site.
The Calvert Clif fs site is located in Calvert County, Maryland, on the west bank of the Chesapeake Bay. It is approximately 10.5 miles (17 km) southeast of the town of Prince Frederick, Maryland, at 38.4 degrees north latitude and 76.5 degrees west longitude.
Baltimore Gas and Electric Company owns and controls all property within a exclusion area with a minimum radius of 0.71-mile (from the center of Calvert Cliffs Units 1 & 2). Maryland Routes 2 and 4 run adjacent to the site on the west. The JSFSI will be located approximately 2,300 feet (701 m) southwest of the reactor buildings, placing it approximately 3,900 feet or 0.74 miles (1.2 km) f rom the controlled boundary in the west direction.
The ecology of the Calvert Cliffs site is well documented as a result of surveys performed for the operation of the nuclear power plant.2 There are two species of endangered Tiger Beetles found along the beach area at Calvert Cliffs. Also, another endangered species, the Star Duckweed, is found on BG&E property. The ISFSI is located well outsioe of a 1200 foot protection zone, established by the State of Maryland, for a nest of Bald Eagles located south of Camp Conoy. Because the area proposed for the ISFSI is small (only about six acres), the surrounding area is already disturbed and the ISFSI has a fence surrounding it, there is expected be minimal effect on wildlife.
Within a J0 mile radius of the site, most of the land is predominantly rural and characterized by farmlands and wetlands. The Patuxent Naval Air Station is the only nearby military facility and is located ten miles south of the ISFSI. The are two other small airports in the area. The Chesapeake Ranch Airport is about 6 miles southeast of the ISFSI and the St. Mary's County Airport, located about 10 miles southwest. The only industrial facility within ten miles is the Cove Point Liquid Natural Gas (LNG) terminal and pipeline located 3.5 miles south-southeast of the Calvert Cliffs power plant.
The LNG terminal has been idle since April 1981, but is scheduled to be reopened in 1991.
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4.2 WATER USE AND AQUATIC RESOURCES The ISFSI is located approximately 3,000 feet west of the western shore of the Chesapeake Bay. The Bay is used for recreational fishing, boating and sw % ming. The Bay is also an important fisheries resource and navigational shipping channel. Most potable water used in Calvert County is from subsurface sources and used chiefly for domestic and agricultural purposes.
There are 15 towns and 8 private communities in Calvert County that have public water supplies. Most domestic water supplies in Calvert County come from deep wells. There are about 70 of these within 2 miles of the Calvert Cliffs plant.
4.3 SOCI0 ECONOMICS, AND HISTORICAL, ARCHEOLOGICAL, AND CULTURAL RESOURCES The immediate area surrounding the Calvert Cliffs plant site is rural.
Despite an average 5.2 percent annual population growth in Calvert County between 1970 and 1980 and a projected 75 percent increase between 1980 and 2010, no significant industrialization is expected and the socioeconomic character of the area will remain basically unchanged. The additional work-force required during construction will not be of sufficient size or their stay of sufficient duration to affect the basic socioeconomic characteristirs o* the local area.
The ISFSI site is located within the Calvert Cliffs plant boundary and therefore will not affect any regional historic, archeological, cultural, or scenic resources.
4.4 DEMOGRAPHY The population density in the vicinity of the Calvert Cliffs Nuclear Power Plant is generally low. There are currently no residences 5,ithin 1 mile of the reactor plant. The closest residence is about 0.9 miles west-southwest of the ISFSI or about 1.3 miles southwest of the reactor plant. Additional resiciences are located within 1 to 2 miles from the west-northwest sector and west around to the southeast sector. In the future, the nearest resident might be located as close as 0.75 miles from the ISFSI which is the closest distance to the controlled area boundary.
A little over 300 people are estimated to live within 2 miles of the site and about 5500 people within 5 miles. Due to the recreational uses of the Chesapeake Bay in the area, the summer transient population may increase by about 23 percent. Nearby population centers within 5 miles of the site are as follows: Long beach, Calvert Beach, St. Leonard, Wallville, Mackall, Sellers, Lusby, Appeal, Bertha, and Cove Point.
The Tri-County Council of Southern Maryland has projected about a 75 percent increase in the population in the area between 1980 and 2010. Solomons, St. Leonard, and Longbeach communities are projected to be the major growth areas within 10 miles of the site.
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<. 4 4.5 METEOROLOGY The Calvert Cliffs site lies along the west bank of the Chesapeake Bay. The Bay area generally has mild winters and summers. Most of the weather comes from a westerly direction across the United States. From October to April the winds prevail from the northwest and during the warmer months from the southwest, with average annual wind speeds of 4.7 mph. Precipitation amounts are distributed rather uniformly throughout the year, with typical amounts of about 40 inches annually. Additional climatological data is available in the ER.8 Extremes of weather include maximum wind speeds of 80 miles per hour recorded at the Baltimore-Washington International Airport, Maryland. Extremes in temperature have ranged from -3 to 103 degrees Fahrenheit at Patuxent River Naval Air Station. The heaviest monthly rainfall recorded at the Patuxent River Naval Air Station was 15.5 inches in July 1945. The heaviest monthly snowfall recorded was 32.3 inches in 1979.
The area experiences about 54 thunderstorms per year, and tropical storms may at times affect the area. Hurricane force winds are expected to affect the Calvert Cliffs area about once every 10 years bring heavy rainfall to the region, with amounts of 5 to 6 inches falling within a 24-hour period.
The mean annual frequency of tornadoes in the vicinity of a single latitude-longitude square near the ISFSI is 0.5 per year and the probability of a tornado striking a single point within the area is approximately 3.75x10-4 tornadoes per square mile per year (or a recurrence frequency of about or.e tornado per square mile every 2700 years).5 4.6 GEOLOGY AND SEISMOLOGY The Calvert Cliffs Nuclear Power Plant Site is located within the Coastal Plain Physiographic Province about 50 miles east of the Fall Zone. The regional geology is typical of the Piedmont Province. The buried surface of the basement igneous and metamorphic rock slopes to the southeast at about 50 feet per mile, and at the site, is located about 2500 feet below sea level.
Beneath the Coastal Plain Province overlying the basement are Cretaceous and Tertiary strata consisting of sedimentary deposits of silt, clay, sand, and gravel. The thick sedimentary strata of the Coastal Plain in the vicinity of the site have essentially not been deformed since they were deposited about 135 million years ago. Areas in the vicinity of the site above 70 feet elevation are underlaid by Pleistocue age sediments which consist primarily of silt and sand. Portions of the site below 70 feet elevation are underlaid by relatively impervious Miocene age sediments of the Chesapeake group. Based on test borings at the ISFSI site and laboratory tests no significant potential for liquefaction of soils is anticipated.
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The Calvert Cliffs site is in a region which has experienced infrequent minor earthquake activity, Because the region has been populated for over 300 years', all earthquakes of moderated intensity (Modified Mercalli Scale VI) or greater would probably have been reported during this period. Since the late 18th Century,18 earthquakes with epicenter intensities ranging from Modified Mercalli Scale V to VII were reported within about 100 miles of the site, Most of the reported earthquakes.in the region have occurred in the Piedmont Physiographic Province west of the Fall Zone and were generally related to known faults in the Piedmont rocks. No shock within 50 miles of the site has been large enough to cause significant structural damage. Only one earthquake of intensity V or greater, located 45 miles northeast of the site, has been reported but caused no structural damage, 13
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5.0 DESCRIPTION
OF THE CALVERT CLIFFS NUCLEAR PLANT ISFSI 5.1 GENERAL DESCRIPTION The proposed ISFSI involves physical components and a system of prccedures designed to be used in a complementary fashion to protect onsite personnel and the general public from radioactivity in the spent fuel, and to maintain the integrity of the confinement and shielding barriers which provide this protec-tion. The physical components of the proposed ISFSI are described in Section 5.2, while the operational procedures are described in Section 5.3. The planned monitoring program for the ISFSI is described in Section 5.4.
5.2 ISFSI DESIGN The ISFSI provides f.e the horizontal, dry storage of irradiated fuel -
assemblies in a concrete module. There are six major physical components associated with the proposed ISFSI. These are the spent fuel, the dry shielded canister (DSC), the transfer cask, the transfer trailer, the horizontal st orage module (HSM), and the hydraulic ram. Each of these components is discussed below. Detailed design information is presented in Reference 6. "
Spent Fuel Spent fuel, because of its radioactive nature, presents a potential hazard to plant personnel, the general public, and the environment. The ISFSI systen is designed to safely store spent fuel by confining the fuel material and providing bulk shielding from radiation.
Baltimore Gas and Electric Company has identified the spent fuel assemblies to be stored in the ISFSI. Specifically, the spent fuel must comply with the restrictions listed in Table 5.1 before it will be transferred to the ISFSI.
These restrictions are based on the need to assure that: (1) there is no -
potential for nuclear criticality; (2) maximum allowabic fuel clad temperatures -
are not exceeded, and (3) dose rates outside the HSM are within the allowable design limits.
Dry Shielded Canister .
The DSC provides the primary confinement of the fuel. It consists of a stainless steel cylinder with an internal structure of discs and rods with discrete storage positions for 24 pressurized water reactor (PWR) spent fuel assemblies.
There are shielded end plugs for the DSC which reduce the radiation field at the ends of the cylinders.
Transfer Cask The transfer cask is used to transport the loaded DSC either from the spent fuel pool in the reactor area to the ISFSI, or from the ISFSI to the fuel pool.
The cask has both lead gamma shielding and a water-based solution for neutron shielding. There are removable plates at the two ends so that the DSC can be placed in and removed from the transfer cask. There are also lifting trunnions on the cask so that it can be moved into and out of the fuel building, and lifted onto the transfer trailer.
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Table 5.1. Design Parameters for the Calvert Cliffs ISFSI Category Criterion or Parameter Value Fuel Acceptance Initial Fissile Content 1.8-4.5% 23sU with credit for Criteria burn-up Radiation Source Total Gamma per Assembly 4.27 x 1015 photons /sec Total Neutron per Assembly 2.32 x 10 8 neutron /sec Heat Load per Assembly 0.66 Kw Dry Shielded Capacity per Canister 24 PWR Fuel Assemblies Canister Size Length (typical) 4.42 m (174 in.)
Diameter 1.71 m (67 in.)
Temperature (max. 332 degrees C (630 degrees F) long-term fuel rod clad)
Cooling Natural Convection Design Life 50 Years Material 304 Stainless Steel with Lead End-Shields Internal Helium 2.5 psig i 2.5 psig Horizontal Capacity 1 Dry Shielded Canister per Storage Module Module 4
Size Length 5.8 m (19 ft.)
Height 4.6 m (15 ft.)
Width 2.6 m (8.7 ft.)
Average Surface Radiation 15 mrem /hr-Dose Rate (area weighted average)
Material Reinforced Concrete Design Life 50 years l
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i 4 Transfer Trailer The transfer trailer is used to transport the transfer cask between the fuel building and the ISFSI.
Horizontal Storage Module The HSM is a reinforced concrete shield structure used to store the DSCs. The HSM provides shielding as well as heat removal by natural convection.
Hydraulic Ram The hydraulic ram is used to move the OSC from the transfer cask into the HSM or from the HSM into the transfer cask.
- 5. 3 ISFSI OP.ATIONS The ISFSI will be operated according to procedures which will be incorporated into the existing system of Calvert Cliffs Nuclear Power Plant procedures.
The major steps associated with the placing of fuel in the Calvert Cliffs ISFSI are presented in Table 5.2. As part of these operations, a number of specific actions will be taken to assure protection of operators as well as the general public. The major specific actions are:
Preoperational Testing Prior to any transfer or loading of spent fuel, Baltimore Gas and Electric Company will perform dry runs with the various ISFSI components to ensure the operability of system components and procedures. Any problems identified during these preoperational tests will be resolved through modification of equipment or procedural changes.
Component Quality Assurance Quality assurance procedures will be applied to the acceptance of key ISFSI equipment. The highest level of quality assurance will apply to the DSC and the transfer cask. Lower levels of quality assurance will apply to items which are less critical for the prottetion of operators and the general public, including the HSM and onsite construction activities.
Fuel Selection l Specific procedures along with quality assurance checks will be applied to the l fuel selection process to ensure that only appropriate fuel is selected for
! loading into the ISFSI.
j Contamination Control of the DSC Exterior Surfaces Because external surfaces of the DSC will be directly exposed to the atmosphere as part of the canister ccoling process, Baltimore Gas and Electric Company will l
take steps to keep exterior surfaces as contamination free as possible. This practice will minimize the potential for release of radioactive material to the environment.
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Table 5.2. Major operational steps for tranferring the fuel from the Spent Fuel Pool to the ISFSI (1)
- 1. Receive, inspect and accept the manufactured DSC.
- 2. Position the DSC in the transfer cask, fill the DSC and cask with water, and lower the transfer cask containing the DSC into the spent fuel pool.
- 3. Load the previously selected spent fuel assemblies into the DSC.
- 4. When loading is complete, position the top end shield plug in the DSC.
- 5. Move the loaded DSC/ transfer cask combination from the pool to the decontamination pit.
- 6. Lowe water level in both the DSC and the transfer cask and weld the top end shield plug to the DSC body.
- 7. Purge and dry the DSC, fill with helium, seal the DSC fill and drain ports, weld the DSC top cover plate in place, and deconteminate the upper DSC surface and the transfer cask exterior if necessary.
- 8. Drain the water from the transfer cask and position the transfer cask with the filled and sealed DSC on the transfer trailer.
- 10. Inspect the interior and exterior of the HSM.
- 11. Position the trailer next to the inspected HSH and align the DSC with the HSM opening.
- 12. Transfer the DSC from the transfer cask to the HSM.
- 13. Close the HSM and return the transfer cask and transfer trailer to their storage position. (
(1) Steps for removing the fuel from the DSC are not addressed above, but are considered in References 3 and 8.
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4 Radiation Protection Procedures The operation of the ISFSI will be according to the general radiation protection program which is already in place at the Calvert Cliffs site.
Training The operators for the ISFSI will be trained in the principles and require-ments of the ISFSI.
, Normal and Emergency Procedures Normal and emergency procedures will be established for the operation of the ISFSI and adhered to by all personnel.
- 5. 4 MONITORING PROGRAM An effluent monitoring program is not applicable to the ISFSI, because its operation will not result in any water or other liquid discharges; it will not generate any chemical, sanitary, or solid wastes; and it will not release any radioactive materials in solid, gaseous or liquid form during normal operations.
Similarly, with the lack of liquid or gaseous effluents from the ISFSI, special environmental monitoring for these exposure pathways is not necessary.
Therefore, a separate environmental measurement program for ISFSI is not warranted; however, to help assure proper operation of the ISFSI system, Baltimore Gas and Electric Company will incorporate ISFSI monitoring into the Calvert Cliffs site monitnring program. The site operational surveillance program will also be expanded to include surveillance of the ISFSI.
The Calvert Cliffs Nuclear Power Plant maintains an air, water, and food pathway monitoring program which establishes the basis for evaluation of environmental impacts of facility operation, and is used in the assessment of public and occupational dose from Calvert Cliffs operations. This environmental surveillance program has been conducted continuously at the Calvert Cliffs Nuclear Power Plant since 1969. The program is designed to confirm that Baltimore Gas and Electric Company operations are within regulatory requirements and consistent with the documented As Low As Is Reasonably Achievable (ALARA) program. The main thrust of the health physics and ALARA programs is to minimize exposure to radiation such that the total exposure to personnel in all phases of design, construction, operation and maintenance are kept ALARA. The ISFSI operations are included in the existing ALARA program for the Calvert Cliffs Nuclear Power Plant.
Levels of external radiation exposure from the ISFSI will be estimated by environmental dosimeters strategically placed to confirm that radiation exposures to direct and scattered radiation are as predicted. Changes in ISFSI inventory will be factored into the radiation dosimetry assessment. No measurable increase in radiation levels above normal background is anticipated beyond the Calvert Cliffs controlled area.
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> ' w-An operational surveillance program will be instituted tr monitor the safe operation of the ISFSI. Once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, site personnel will visually inspect all air inlets of each loaded HSM for obstructions and screen damage.
As necessary, removal of obstruction or screen repair will be initiated immediately. The ISFSI will also be included in-routine site patrols by Calvert Cliffs security personnel.
Monitoring program results are published annually. The ongoing monitoring program is describad and results for the most recent 1 year program are contained in Reference 8.
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e L 6.0 ENVIRONMENTAL IMPACTS OF THE PROPOSED ACTION 6.1 CONSTRUCTION IMPACTS The ISFSI site area will be developed and managed so as to minimize construction impacts. All construction activities will comply with Federal, State and local regulations governing safety and health for construction. Work will be monitored by Baltimore Gas and Electric Company personnel.
Construction will be phased according to need. Initial requirements for storage will be cet by construction of two 2 by 6 HSH arrays placed end to end.
Subsequent construction of HSMs will be as required.
6.1.1 Land Use and Terrestrial Resources The 120 storage modules and their access area will occupy about 6 acres (0.02 km2) approximately 2300-feet (701 m) southwest of the reacter containment buildings. The area is totally within the Calvert Cliffs controlled area; thus, no additionel land use impacts will result from construction of the ISFSI. Part of the con".en,-tion area is wooded and will be cleared. The terrain alteration, clearing, t ~ o.ction and grading will result in a loss of biological production of less that oi.. percent of the Calvert Cliffs site area.
Construction of the ISFSI is not expected to have any impact on any known species listed by either the Federal or State government as endangered.
Measures will be taken to ensure that dust created during earthwork will be kept at an acceptable level and existing paved roads remain free of objectionable amounts of earth and rocks. Burning permits will be obtained as needed and requirements for erosion control, such as silt fences, used as necessary.
6.1. 2 Water Use and Aquatic Resources Construction of the ISFSI will not impact local water supplies. Concrete for the slab will arrive on the site ready-mixed. Drinking water for cleaning operations and fugitive dust control (spraying) will be transported to the site by truck. The portable rest rooms provided during construction require no onsite source of water . During clearing, and excavation operations, temporary measures will be utilized to manage storm-water runoff and provide sediment control in accordance with local construction codes. More permanent drainage will be installed as soon as area excavations and backfill allow.
6.1.3- Other Impacts of Construction Air Quality Temporary increases in levels of suspended particulate matter will result from construction activities. In addition, exhaust from construction vehicles will add to levels of hydrocarbons, carbon monoxide and oxides of nitrogen. Measures such as watering of unpaved haul roads will be used to minimize the generation of fugitive dust. In addition, cleared areas and exposed earth will be seeded, graveled, or paved to stabilize and control runoff, and minimize soil erosion.
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-h %r Noise Noise levels'due to construction traffic, grading, and excavation are not expected to be greater than the noise associated with the normal operation of the Calvert Cliffs plant. To protect onsite-personnel, Occupational Safety and Health Administration standards will be followed.
6.1.4 Socioeconomics Construction of the Calvert Cliffs ISFSI is anticipated to be performed by a local contract construction firm and existing BG&E personnel. The size of the construction work force is not expected to exceed 50 persons at any one time.
The Calvert Cliffs work force may be increased by at most two employees associated with ISFSI-related operations. Given the large population growth rate of Calvert County, a small number of i.ew residents due to the ISFSI will have no effect on the community.
6.1.5 Radiological Impacts from Construction Initially, there will be no radiological impacts from construction. However, occupational radiation exposure is expected to result from the construction of additional HSMs, after some are filled. These operations will be conducted under either (1) existing procedures suitably modified and approved for this activity, or (2) procedures to be prepared under the existing Baltimore Gas and Electric Company administrative requirements which meet NRC Quality Assurance (QA) and ALARA requirements. Because radiation fields from filled HSMs are non-uniform, temporary shielding and access controls will be used as necessary to keep occupational exposure to construction workers ALARA. Estimates of construction-related doses are presented in Section 6.2.1.2.
6.2 OPERATIONAL IMPACTS 6.2.1 Radiological Impacts from Routine Operations The primary pathway through which site workers and nearby residents may be exposed as a result af normal Calvert Cliffs ISFSI operations is through external exposure to direct and scattered radiation. Radiological dose estimates were calculated for this pathway using conservative and design basis assumptions: maximum storage module surface dose rates of 21 mrem /hr neutron and 47 mrem /hr gamma; maximum fuel burn-up_ of /* GWD/MTU (gigawatt-days per metric ton of uranium); and post-irradiation tecay period of at least 8 years before dry storage. These assumptions result'in conservative dose estimates; actual doses are expected to be somewhat lower. Because the proposed ISFSI involves only dry storage of spent nuclear fuel in dry, sealed DSCs, there will be essentially no gaseous or liquid effluents associated with normal storage operations. Activities associated with cask loading and decontamination may result in some gaseous and liquid effluents; however, these operations will be conducted under the 10 CFR Part 50 operating license, and radiological impacts from those effluents fall within the scope of impacts from reactor operations which were assessed in the Calvert Cliffs FES.2 21
p
. L 6.2.1.1 Offsite Dose ISFSI operations will result in a small additional dose to n, embers of the public from direct radiation exposure. Section 72.104(a) of 10 CFR Part 72 requires that dose equivalents from normal operations to any real individual located beyond the ISFSI controlled area not exceed 25 mrem /yr to the whole body, 75 mrem /yr to the thyroid, and 25 mrem /yr to any other organ as a result of planned effluent releases, direct radiation from ISFSI operations, and radiation from other uranium fuel cycle operations within the region.
The maximally exposed member of the public is assumed to have continuous occupancy in the nearest residence to the ISFSI which is located 4705 feet from the facility. At that location, the dose rate from 120 HSMs filled to capacity with design basis fuel would be less than 1 mrem /yr from ISFSI, and less than 13.5 mrem /yr from the remaining fuel cycle operations in the vicinity.4 It can be concluded that the radiation exposure due to the Calvert Cliffs ISFSI, combined with all other fuel cycle operations, will not exceed the regulatory requirements of 25 mrem /yr in 10 CFR 72.104 and 40 CFR Part 190.
Appendix I to 10 CFR 50 sets forth design objective dose commitment guides for liquid and gaseous effluents released from nuclear power reactors. For each reactor, the maximum annual dose commitment to an individual in an unrestricted area is 3 mrem /yr due to liquid effluents and 5 mrem /yr due to gaseous effluents. Thus, the maximum design guide dose commitment from effluents due to operations of Calvert Cliffs Units 1, 2 would be 16 mrem /yr. Current dose levels as a result of releases of radioactivity in effluents are less than the design level.
Since no liquid or airborne effluents are postulated to emanate from the ISFSI, the direst and air scattered radiation exposure discussed in the previous sections comprises the total radiation exposure to the public. No estimation of effluent dose equivalents is necessary.
The nearest resident is located 4705 ft. (1.4 km) from the ISFSI center. The maximum expected dose to an <dividual at this location would be about 0.2 mrem /yr. When combined with the dose commitment from reactor operations, the total dose commitment is well within the 25 mrem /yr limit specified in 10 CFR 72.104. In addition, trees and hilly terrain between the ISFSI and these locations provide shielding, such that individuals here would essentially be exposed only to air-scattered radiation from the ISFSI.
By 2010 there are projected to be about 492 people between 1 and 2 miles of the-Calvert Cliffs Station. The collective dose to this population due to Calvert Cliffs ISFSI operations is estimated to be less than about 75 person-mrem /yr.
The collective dose commitment to that same population due to Calvert Cliffs reactor operations without the ISFSI is estimated to be about 101 person-mrem /yr.2 For populations in the region under consideration beyond 2 miles from the ISFSI, direct and air-scattered radiation contribute very little to the collective dase commitment.
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J w 6.2.1.2 Collective Occupational Dose Spent fuel storage at the Calvert Cliffs ISFSI will result in a small increase in the total occupational dose at the Calvert Cliffs site. Occupational radiation exposure for ISFSI operations is expected to result from loading fuel into the DSC, leading the shipping ce,K, moving the shipping cask to the ISF5I, inserting the OSC into the HSM, sealing the HSH, and conducting routine security checks and operational surveillance. Occupational doses to construction workers resulc from exposure to direct and scattered radiation from irradiated fuel in previously filled modules. The license application requests authority to construct and operate a total of 120 HMSs. These modules will be built incrementally in 1 to 5 rows of 24 HMSs per row as needed to match the rec,uirements for additional .torage. Constrcction work performed subsequent to the loading of any hSM with spent fuel may Jesult in work exposures from direct and skyshine radiation in the vicinity of the loaded HSMs. Assuming a maximum -
dose rate of 0.11 mrem /hr and a workforce of 1500 person hours per HSH, a total dose of 0.17 person-rem rar HSH is estimated for occupational exposure resulting 4 from construction activ.cies. Assuming 24 HSMs are constructed and loaded with spent fusi initially, a total of about 4 person-rem is estimated for the occupational exposure associated with constructic. activities of each additional row of storege modules. This exposure would be d eived over the life of the plant. Calvert Clifts Station workers not ' 9ct.y involved in ISFSI operations will be exposed to small increases in the general area radiation level. .
' 6 Engineered features of the storage uodules and application of administrative 6 controls are designed to ensure that all exposures are maintained at levels which are As Low As Is Reasonably Achievable (ALARA). All ISFSI operatior.s will be conducted under either (1) existing procedures suitably modified and approved f or this activity, or (2) procedures to be prepared under the existing Baltimore Gas and Electric Company administrative requirements which meet NRC Quality Assurance (QA) and ALARA requirements. Occupational doses will be controlled to I within the limits of 10 CFR Part 20.
Table 6.1 presents the estimated maximum collective occupational doses from "
annn i operation and construction of the ISFSI, while Table 6.2 estimates the annual collective dose to Calvert Cliffc Station workers not directly involved in ISFSI activities.
A maximum of sixteen DSCs/HSMs are planned to be loaded the first yaar. The upper bound for annual ISFSI occupational exposure is estimated as rollows.
- 1. The maximum occupational dose due to the loading of one DSC/HSH is about 1.5 person-rem.
- 2. The maximum occupational dose during the loading one DSC/HSH due to previcusly loaded HSMs is abov* 0.02 person rem.
- 3. The maximum occupational dose during the loading of one DSC/HSM to the additional support personrel is about 0.02 person-rem.
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.- L Based on thcre results, the maximum occu,qtional expnsure due to the loading of the DSC/HSH is about 1.5 person-rem. This will result in a maximum annual ISFSI occupational dose during USC/HSH f om loading the first 16 CSC/P^Ms of about 24 person-rem.
The dose to Calvert Cliffs Station wo,xers not directly involved in ISFSI activities is expected to be about 16 person-rem /yr. These values constitutes a small, incremental fraction of the total occupational dose commitment at the Calvert Cliffs Nuclear power Plant. During the years 1987, 1988, and 1989, the collective totsi occupational dose at Calvert Cliffs was 413 rem, 291 rcm, and
, 346 person-rem respectively. The 3 year average from reactor operations is 350 person-rem.
Once all 120 mouule; are loaded, the annual occupational collective dose would be less than five percent of the current average occupational collective dose.
6.2.2 Radiological Impacts of Accidents A variety of accident scenarios which may affect the safe operation of the Calvert Cliffs ISFSI have been postulated. These include earthquakes, tornadoes, tornado missiles, lightning, fires, pressurization of the DSC, bloct?ge of air inlets and outlets, cask drre, leakage of the DSC, and loss of air outlet shielding. The canisters and storage modules are designed to withstand the resultant forces from these accidents. However, two of the postulated accidents have possible offsite radiological consequences. These are loss of air outlet shielding and canister leakage. cf these, canister leakage is the bounding case accident. For assessment purposes, an accident is postulated whersin a non-mechanistic simultaneous failure of the DSC and all fuel cladding occurs, re<ulting in the loss of the helium covec gas and 30 percent of the radioactive Kr-85, I-129, and H-3 inventory in the spent fuel for one DSC. Tables 6.3 and 6.4 summarize the radiological impact of this DSC accidcat scenario. The release fraction estimates for particulate radioactivity (i.e., Sr-90, Ru-106, Cs-137, and Cs-134) used in this analysis were based on a worst-case scenario f or cir-cooled transfer casks.8 This reference (Scenario 5), while not directly relatable to a non-mechanistic simultaraous failure of the DSC, is expected to provide a reasonable assumption. Further, this reference clearly indicates that particulate releases contribute an insignificant amount to the radiation dose. The cited scenario considers n11 release mechanisms that are crr dible for air-cooled casks. Once radionaclides have been released from the fuel rods they are postulated to escape the DSC. The radioactivity released to the DSC cavity is based on the design fuel to be stored in the cask (PWR fuel, initial enrichment of 4.0 percent U-235; 42 GWD/MTU hurn-up; 8 years out of tne reactor).
The accident damage is not expected to provide a pathway with a large cross-sectional area from the DSC cavity to the environment;-the most likely release pathway would consist of only a small section of a failed DSC. In addition to the smal' release area, radionuclides can condense, plate out, or be filtered out before escaping the DSC.
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= _ . .- _ . . .
3 %.
Table 6.1 Collective occupational dose to Calvert Cliffs on workers directly ircolved in 15FSI activities Operation Person-rem Person-rem per DSC per year DSC Loading and Cask Decontamination at Rear. tor 1.471 23.536(1)
Transfer of DSC to and Emplacement in HSM Surveillance during loading .0174 .278 Annual Surveiilance .046(2) fotal 23.860 (1) Estimates are based on the following assumed construction and loading schedule: 24 tiSMs are constructed initially, and 16 DSCs are loaded the first year.
(2) This value is derived by cssuming that one inspector performs a daily inspection walking at an averane speed of 3 mph on a path passing directly in front of all 120 air vents of the 15FSI.
25
- Le Table 6.2 Estimate of collective dose to Calvert Cliffs Station workers not directly involved in ISFSI activities Location No. of Dose rate Annual dose employees (mrem /jer) (person-rem /yr)
Office and Training Facility 240 1.2 E-03 0.75 Nuclear Engineering Facility 280 2.2 E-03 1.60 Materials Processing Facility 240 2.2 E-03 1.37 Nuclear Office Facilit, 470 4.7 E-03 5.74 A
Inside the Protected Area 1200 2.2 E-03 6.86 Total 16.32 A
Nuclear Engineering Facility dose rates used as conservative estimate for the plant protected area.
26
~ ~~
)
Table 6.3 Expected dose at the controlled area boundary resulting from a dry shielded canister leakage accident at the Calvert Cliffs Nuclear Power Plant (1)
Whole Body Dose Nuclide OSC Release X/Q Breathing Tot. body Dose at inventory fraction rate inhalatic". '
ry DCF10 3
(uCi) (see/m ) (m3 /sec) (rem /uC1) (rem)
H-3 5.79E+09 3.00E-01 4.00E-04 2.54E-04 1.20E-04 2.54E-02 K r-85 6.61E+10 3.00E-01 4.80E-04 1.00E+00 3.34E-10 (2) 3.18E-03 1-129 3.92E+05 3.00E-01 4.80E-04 2.54E-04 1.80E-01 2.58E-03 Cs-134 1.43E+1) 5.00E-10 4.80E-04 2.54E-04 4.40E-02 3.84E-07 Cs-137 1.07E+12 5.00E-10 4.80E-04 2.54E-04 3.00E-02 1.96E-06 Sr-90 7.38E+11 5.00E-10 4.80E-04 2.54E-04 1.30E+00 5.85E-05 Ru-106 2.09E+10 5.002-10 4.80E-04 2.54E-04 4.70E-01 5.99E-07 Total Dose 0.031 Thyroid Dose Nuclide DSC Release X/Q Breathing Thyroid Dose at inventory f raction rate inhalation boundary DCF10 (uCi) (sec/m3 s) (m3 /sec) (rem /uCi) (rem)
H-3 5.79E+09 3.00E-01 4.80E-04 2.54E-04 1.20E-04 2.54E-02 Kr- 85 5.61E+10 3.00E-01 4.80E-04 1.00E+00 3.34E-10 (2) 3.18E-03 1-129 3.92E+05 3.00E-01 4.80E-04 2.54E-04 8.30E+00 1.19E-01 Cs-134 1.43E+11 5.00E-10 4.80E-04 2.54E-04 7.70E-02 6.71E-07 Cs-137 1.07E+12 5.00E-10 4.80E-04 2.54E-04 4.90E-02 3.19E-06 S r-90 7.38E+11 5.00E-10 4.80E-04 2.54E-04 9.50E-03 4.27E-07 Ru-106 2.09E+10 5.00E-10 4.80E-04 2.54E-04 6.10E-02 7.77E-08 Thyroid Dose 0.14d (1) The distance from the controlled area boundary to the nearest HSM is about 3,900 feet, t
(2) Whole-body submersion DCF in rem-m3 /uCi-sec (Reference 11).
27
i., o-Table 6.4 Expected dose at the nearest residence resulting from dry shielded canister leakage accident at the Calvert Cliffs Nuclear Power Plant (1)
Whole Body Oose Nuclide OSC Release X/Q Breathing Tot. body Dose at inventory fraction rate inhaiation boundary OCF10 (uci) (sec/m3 ) (ms /sec) (rem /uCi) (rem)
H-3 5.79E+09 3.00E-01 3.60E-04 2.54E-04 1.20E-04 1.91E-02 Kr-85 6.61E+10 3.00E-01 3.60E-04 N.A. 3.34E-10 (2) 2.38E-03 1-129 3.52E+05 3.00E-01 3.60E-04 2.54E-04 1.80E-01 1.94E-03 Cs-134 1.43E+11 5.00E-10 3. 60E-04 2.54E-01 4.40E-02 2.88E-07 Cs-137 1.07E+12 5.00E-10 3.60E-04 2.54E-04 3.00E-02 1.47E-06 S r-90 7.38E+11 5.00E-10 3.60E-04 2.54E-04 1.30E+00 4.39E-05 Ru-106 2.09E+10 5.00E-10 3.60E-04 2.54E-04 4.70E-31 4.49E-07 Total Oose 0.023 Thyroid Dose Nuclide OSC Release X/Q Breathing Tot. body Oose at inventory fraction rate inhalation boundary OCF10 (uCi) (sec/m3 ) (m3 /sec) (rem /uC1) (rem)
H-3 5.79E+09 3.00E-01 3.60E-04 2.54E-04 1.20E-04 1.91E-62 Kr-85 6.61E+10 3.00E-01 3.60E-04 N.A. 3.34E-10 (2) 2.38E-03 I-129 3.92E+05 3.00E-01 3.60E-04 2.54E-04 8.30E+00 8.93E-02 Cs-134 1.43E+11 5.00E-10 3.60E-04 2.54E-04 7.70E-02 5.03E-07 Cs-137 1.07E+12 5.00E-10 3.60E-04 2.54E-04 3.30E-02 1.61E-06 Sr-90 7.38E+11 5.00E-10 3.60E-04 2.54E-04 9.50E-03 3.21E-07 Ru-106 2.09E+10 5.00E-10 3. 60E-04 2.54E-04 6.10E-02 5.83E-08 Thyroid Dose 0.111 (1) Nearest residence is approximately. 4705 feet.
(2) Whole-body submersion DCF in rem-ms/uCi-sec (Reference 14).
26
m After the radioactive material escapes the DSC, two factors a e important in determining whether the particles reach the population: the fraction that becomes suspended in air, and the fraction that is respirable (less than 10 microns in diameter). A direction-independent atrmspheric dispersion (X/Q) value was used to calculate a dose as the nearest controlled area boundary (0.74 mi. or 1.2 km), and the nearest residence (0.9 mi. or 1.4 km). The X/Q used is taken from Regulatory Guide 1.4,M and e.sumes Class F stability, 1 m/sec wind speed, and ground-level release.
The upper bound dose at the controlled area boundary due to the postulated accident which releases 30 percent of the tritium, noble gas, and iodine would be about 31 mrem to the whole-body and 148 mrem to the thyroid. The dose at the location of the nearest residence would be about 23 mrem to the whole-body and about 111 mrem to the thyroid. The resultant whole-body dose to an indi-vidual at the controlled area boundary is a small fraction of the 5 rem criteria specified in 10 CFR 72.106(b). These doses are also much less than the Prctective Action Guides (PAGs) established by the Environmental Protection Agency (EPA) for individuals exposed to radi ation as a result of accidents: 1 rem to the whole-body and 5 rem to the most severely affected organ. Thus, the release of ef fluents from the 1 'SI due to accidents, even those with a very low probability of occurrence, will have a negligible impact on the population in the surroundings of the Calvert Cliffs Nuclear Power Plant.
A separate emergency planning zone (EPZ) has not been developed for the ISFSI.
The 10-mile Plume Exposure Pathway EPZ for the Calvert Cliffs Nuclear Power Plant provides a sufficient level of safety for credible accident scenarios related to construction and operation of the ISFSI.
6.2.3 Nonradiological Impacts 6.2.3.1 Land Use and Terrestrial Resources Operation of the ISFSI will not require the use of any land beyond that which was cleared and graded during its construction, and is not expected to adversely impact the terrestrial environment. Heat from the DSCs is not expected to be high enough to affect vegetation growth adjacent to the HSMs. Inhibited access to the ISFSI by the surrounding fence will discourage wildlife species from using the area adjacent to the HSMr. During winter months some birds may roost on the upper surface of the HSMs due to heat from the exit vents. This is not expected to result in adverse impact to individual birds. Wire mesh screens will be placed over the inlet and exit ports of the HSMs to prohibit entry of birds, wind-blown debris, etc.
6.2.3.2 Water Use and Aquatic Resources The Calvert Cliffs ISFSI is a passive, air-cooled system. There is no planned water use or liquid discharge to local surf ace or groundwater supplies associated with operation of the ISFSI. Surface runoff from precipitation will enter Chesapeake Bay under existing drainage routes, but is not expected to result in negative impact to water quality.
The only water required for operation of the ISF . for decontamination of the transfer cask, will be used within the confines of the Calvert Cliffs Station Auxiliary Bufiding and fall within the scope of impacts previously assessed for reactor operations.2 29
G.2.3.3 Other Impacts af Operation Climatolocy During rainy days, precipitation may vaporiz9 upon contact with the surface of the HSMs as a result of the relative higher rperature of the HSM surface or outlet air. Consequently, fog may form above the HSMs. However, a significant increase in the amount of fog extending beyond the plant's exclusion boundary is not expected.
Noise Noise associated with operaticn of the ISFSI will result from transfer of the designated spent fuel from the spent fuel pool facility to the HSMs. The noise associated with this activity is not expected to be distinguishable from other operatio.ial noise at the site or to result in adverse impact to local residents.
1 1
30 '
(
M' Mr
- 7. 0 SAFEGUARDS FOR SPENT FUEL The Commission's requirements fer the protection of an ISFSI are set forth in 10 CFR Part 72, Subparts H and K, which include provisions for security plans, a security organization, response guards, detection aids, response force action, communication capability, and law enforcement agency liaison.
The applicant tas submitted to the NRC a security, contingency and guard training and qualification plan which will commit to protecting the spent fuel against the design basis threat of radiological sabotage. These plans include the following:
- Bartiers to limit unauthorized access to this storage installation,
- Access controls for personnel, vehicles, and packages,
- Search requirements to detect contraband materials,
- Detection and assessment capability for all alarms,
- Site specific training for security force members,
- Pre planned contingency events and security actions,
- Commitments for responding to unresolved alarms.
- Provisions for obtaining support from the local law enforcement agency, and
- Secure transportation of the spent fuel from the reactor site to the ISFSI.
The implementation of these physical security plans will be inspected for effectiveness and operational compliance.
An independent safety review of the horizontal storage module design is being cc-ducted by the NRC. Conservative data are used for safety analysis of the design, including design basis criteria, margins of safety, siting factors, quality assurance and physical protection. The potential for radiological sabotage, theft or diversion of spent fuel from the ISFSI with the intent of
, utilizing the contained special nuclear material (SNM) for nuclear explosives is not considered credible due to (1) the inherent protection afforded by the massive reinforced concrete storage module and the steel storage canister, (2) the unattractive form of the contained SNM, which is not readily separable from the radioactive fission products, and (3) the immediate hazard posed by the high radiation levels of the fuel to persons not provided radiation protection.
Accordingly, the storage of spent fuel at this ISFSI will not constitute an unreasonable risk to the public health and safety from acts of radiological sabotage theft, or diversion of SNM.
31
- a W "
8.0 DECOMMISSIONING All spent fuel assemblies W red in the proposed Calvert Cliffs ISFSI will eventually be shipped to a UOE Monitored Retrievable Storage (MRS) Facility or directly to a Federal Geological Repository for permanent disposal.
Decommissioning of the ISFSI will be performed in conjunction with decommissioning of the Calvert Cliffs Nuclear Power Plant. The costs of decommissioning the ISFSI are expected to represent a small and negligible fraction of the costs of decommissioning the Calvert Cliffs Nuclear Power Plant.
Decommissioning will involve submittal of a decommissioning plan in accordance with 10 CFR 72.30. The only act.ivities expected in decommissioning the Calvert Cliffs ISFSI are the removal of the spent fuel from the site for transfer to a Federal repository, and the decontamination and dismantling of the concrete HSMs. Presently, Baltimore Gas and Electric Company expects to be able to remove the DSCs containing the spent fuel from the HSMs and place them in a transportation cask for shipment to the Federal repository. If the fuel must be removed from the DSCs for transport or disposal, the canister could be decontaminated and disposed of as low-level waste. The HSMs are expected to have minimal contamina+ ion of their internals and air 'assages, which could be easily removed. Subsequent to removal of the DSCs, the reinforced concrete modules could be broken up and removed. No residual contamination is expected to remain on the concrete pads.
Based on a separate NRC staff assessment,13 annual occupational doses associated with unloading spent fuel from an ISFSI, af ter 20 years storage, for subsequent offsite shipment to a Federal MRS or repository are estimated to be small. If the DSC must be returned to the reactor buildings, and the fuel removed f rom the DSC, returned to the spent fuel storage pool, and loaded into a shipping cask, the occupational doses associated with storage cask and fuel handling are expected to be less than one-half of the values shown in Table 6.1. If the DSC is compatible with a certified shipping cask and easily inserted directly into the shipping cask from the HSM, doses to workers are expected to be about one-tenth of the doses shown in Table 6.1.
In accordance with the requirements of 10 CFR 72.30, a decommissioning plan has been submitted by the Applicant. H This document includes a commitment to establish an externally administered, sinking fund as described.in 10 CFR 72.30(c)(5), to fund decommissioning costs. According to the requirements of
- 10 CFR 72.54, the licensee may apply to the NRC for authority to surrender a license voluntarily and to decommission the ISFSI within two years following
! permanent cessation of operations, and in no case later than one year prior to l expiration of the license. The Applicant must receive approval of the final decommissioning plan f rom the NRC prior to the commencement of any decommissioning activities. The NRC will then terminate the license after determining that (1) the decommissioning has been performed in act.' rdance with
! the approved final decommissioning plan and the order authorizing the decommissioning; and (2) the terminal radiation survey and associated documentation demonstrates that the ISFSI and site are suitable for release for unrestricted use.
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-- . -- - - . - . _ _ - , _ - - - = - - - . - - . . _
es .qw i
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- 9. 0
SUMMARY
AND CONCLUSIONS 9.1
SUMMARY
OF ENVIRONMENTAL IMPACTS As discussed in Section 6.1, no significant construction impacts are anticipated.
The activities will af fect only a very small fraction of the land area of the Calvert Cliffs Nuclear Power Plant. With good construction practices, the potentials for fugitive dust, erosion and noise impacts, typical of the planned construction activities, can be controlled to insignificant levels. The only resources committed irretrievably are the steel, concrete, and other construction materials used in the ISFSI storage modules, pads, and canisters.
The primary exposure pathway associated with the ISFSI operction is direct radiation of site workers and nearby residents. As discussed in Section 6.2.1, -
the radiological impacts from liquid and gaseous effluents during normal operation of the ISFSI fall within the scope of impacts from licensed reactor operations, which were assessed in the Calvert Cliffs FES and are controlled by the existing Technical Specification for the reactors. ,
The dose to the nearest resident f rom ISFSI operation is less than 1 mrem /yr, and when added to that of the operations of the two-unit Calvert Cliffs Nuclear Power Plant, is much less than 25 mrem /yr as required by 10 CFR 72.104. The collective dose to residents within 1 to 2 miles of the ISFSI is estimated to be less than .1 person-rem /yr. Occupational dose to site workers during HSM construction (24 person-rem /yr), and during ISFSI operation (24 person-rem /yr),
is a small fraction of the total occupational dose commitment at the Calvert Clif fs Nuclear Power Plant (i.e. , 350 person-rem /yr is the annual average occupational dose over 3 years ending in 1989). Individual doses are controlled to te within the limits established by 10 CFR Part 20.
The upperbound offsite radiological impacts due to accidents at the Calvert Cliffs ISFSI are about 31 mrem to the wholt body and 148 mrem to the thyroid of an individual located at the controlled area boundary, and about 23 mrem whole body and 111 mrem thyroid doses to the noarest resident. These doses are only a small fraction of the criteria specified in 10 CFR 72.106(b) and by the EPA Protective Action Guides. The Emergency Planning Zone (EPZ) for the ISFSI will coincide with that of the Calvert Cliffs Nuclear Power Plant (1.9., a 10-mile Plume Exposure Pathway and 50-mile Ingestion Pathway).
As discussed in Section 6.2.3, no significant nonradiological impacts are expected during operation of the ISFSI. The only environmental interface of the ISFSI is with the air surrounding the storage modules; the only discharge 01 waste to the environment is heat to the air via the passive heat dissipation system. Climatological ef fects which are anticipated in the immediate vicinity of the ISFSI are judged to be insignificant to public health and safety.
9.2 BASIS FOR FINDING OF NO SIGNIFICANT IMPACT We have reviewed the proposed action relative to the requirements set forth in 10 CFR Part 51, and based on this assessment have determined that issuance of a materials license under 10 CFR Part 72 authorizing storage of spent fuel at the Calvert Cliffs ISFSI will not significantly affect the quality of the human environment. Therefore, an environmental impact statement is not warranted, and
, pursuant to 10 CFR Part 51.31, a Finding of No Significant Impact is appropriate.
33
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10.0 REFERENCES
- 1. U.S. Nuclear Regulatory Commission, " Final Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel,"
NUREG-0575, August 1979.
- 2. U.S. Atomic Energy Commission, " Final Environmental Statement Related to Operation of Calvert Cliffs Nuclear Power Plant Units 1, and 2," April 1973.
- 3. Baltimore Gas and Electric Company, "Calvert Cliffs Independent Spent Fuel Storage Installation Environmental Report," December 1989.
- 4. Baltimore Gas and Electric Company, Letter from R. E. Denton to the NRC,
" Response to NRC's Comments on Environmental Issues Regarding BG&E's License Aprlication for Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI)," November 1, 1990.
- 5. Baltimore Gas and Electric Company, "Calvert Cliffs Nuclear Power Plant Final Safety Analysis Report."
- 6. Baltimore Gas and Electric Company, " Independent Spent Fuel Storage Installation Safety Analysis Report Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant," December 1989.
- 7. Nutech, Inc., " Topical Report for the Nutech Horizontal Modular Storage System for Irradiated Nuclear Fuel NUHOMS-24P," San Jose, CA, 1988.
- 8. Baltimore Gas and Electric Company " Radiological Environmental Monitoring Program Annual Report for the Calvert Cliffs Nuclear Power Plant Units 1 and 2, January 1 - December 31, 1990," March 1991.
- 9. Wilmot, Edwin L., " Transportation Accident Scenarios for Commercial Spent Fuel," SAND 80-2124, Sandia National Laboratory Albuquerque, NM, February 1981.
l
- 10. Dunning, Donald E., " Estimate of Internal Dose Equivalent from Inhalation and Ingestion of Selected Radionuclides," WIPP-DOE-176, Evaluation Research Corporation, Oak Ridge, TN, for Westinghouse Electric Corporation.
l
- 11. D. C. Kocher, " Dose Rate Conversion Factors for External Exposure to Photons and Electrons," NUREG/CR-1918, prepared for the Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, TN, August 1981.
- 12. U.S. Nu-lear Regulatory Commission, " Assumptions Used for Evaluating the l Potential Radiological Consequences of a loss of Coolant Accident for l Pressurized Water Reactors," Regulatory Guide 1.4, June 1974.
l 13. U.S. Nuclear Regulatory Commission, Note to File, " Comparative Analysis l Examining Two Storage Systems," October 20, 1988, Docket No. 72-4.
i
- 14. Baltimore Gas and Electric Company, Letter from G. C. Creel to the NRC,
" Transmittal of Decomissioning Plan for Independent Spent Fuel Storage Installation (ISFSI)," July 26, 1990.
l 34
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- m%s 11.0 LIST OF AGENCIES AND PREPARERS Those NRC staff members principally responsible for the preparation of this EA are listed below:
Name Responsibility Frederick C. Sturz Project Manager Michael G. Raddatz Technical Reviewer The following outside agencies were contacted in connection with the preparation of this EA.
Power Plant and Environmental Raview Division, Maryland Department of Natural Resources Calvert County Department of Planning and Zoning Calvert County Department of Emergency Management Calvert County Environmental Planner Calvert County Director of Public Safety 4
t L 35 l.
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