ML20058M133

Obtains Commission Authorization,Pursuant to 10CFR2.764(c), for NMSS to Issue Initial License Authorizing NSP to Possess Spent Fuel from Plant in ISFSI

ML20058M133
Person / Time
Site:	Prairie Island
Issue date:	09/27/1993
From:	Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
References
SECY-93-269, NUDOCS 9310010228
Download: ML20058M133 (166)

Text

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_ _ _m-m, RELEASED TO THE PDR l _Lo,k /93

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\\.~....f September 27, 1993 SECY-93-269 (Notation Vote)

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The Commissioners i

FROM:

James H. Taylor j

Executive Director for Operations 1

SUBJECT:

PROPOSED LICENSE, UNDER 10 CFR PART 72, FOR DRY STORAGE OF SPENT FUEL AT NCRTHERN STATES POWER C0fiPANY'S PRAIRIE ISLAND NUCLEAR GENERATING PLANT PURPOSE:

j 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 authorizing Northern States Power Company (NSP) to possess spent fuel from Prairie Island in an independent spent fuel storage installation (ISFSI).

The license would give NRC approval to the ISFSI on the Prairie Island Nuclear Generating Plant site, using the Transnuclear Inc., TN-40 storage cask, and consisting of up to 48 casks on concrete pads.

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SUMMARY

The staff has completed thorough health and safety, safeguards, and i

environmental reviews of the proposed storage of spent fuel in TN-40 casks on NSP's Prairie Island site. A proposed license (Enclosure 1) has been prepared for signature by the Director, NMSS, or his designee, based on the staff's Safety Evaluation Report (SER) (Enclosure 2). Only_ spent pressurized water reactor (PWR) fuel from Prairie Island is to be stored at the site.

Contact:

JSchneider, NMSS, STSB NOTE:

TO BE MADE PUBLICLY AVAILABLE 504-2692 or WHEN THE FINAL SRM.IS MADE FSturz, NMSS, STSB AVAILABLE 504-2684

'Section 2.764(c), which provides for express Commission authorization of each specific Part 72 license, is the subject of ongoing rulemaking to eliminate the need for express Commission authorization.

Public comments have been submitted and are under consideration, and 0GC anticipates forwarding a draft final rule to the Commission for consideration in the near future.

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The Commissioners The storage activities and location are covered under the existing Price Anderson indemnity agreement for the Prairie Island site. No offsite transportation of spent fuel is involved. Heavy loads involved in cask handling will be within the auxiliary building's crane capacity of 125 tons.

An NSP evaluation has shown that dry storage activities do not represent an unreviewed safety question for reactor operations, and that no 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 48 casks. At its own risk, NSP has nearly completed site construction and initiated the purchase of storage system components from the vendor. However, NSP recently stopped fabrication in light of the need for State legislative approval discussed below.

Although NSP applied to the Minnesota Public Utilities Commission (MPUC) for a certificate of need for 48 casks, the MPUC granted approval for only 17 casks.

This MPUC approval was thereafter appealed, by citizens groups, through the Minnesota State courts.

Based on the Minnesota Supreme Court decision not to hear the case, approval of the ISFSI will be required by the Minnesota State Legislature.

The Legislature is not scheduled to convene until February 1994; however, committee hearing will begin this fall. The use of 17 casks would provide NSP enough additional spent fuel storage capacity to continue operating Prairie Island for over 10 years.

It is the staff's understanding that, if NSP needs more dry storage in the future, it would need to obtain another certificate of need from the State.

DISCUSSION:

On August 31, 1990, Nuclear Regulatory Commission staff received an application from NSP to store spent fuel in dry storage casks to be located at the Prairie Island site. NSP's Safety Analysis Report (SAR) and its Environmental Report are the basis for this design. The staff reviewed all relevant health and safety aspects of the design including criticality, structural, thermal, and shielding considerations under both normal and accident conditions. The staff concluded that the TN-40 cask designed by Transnuclear Inc., could be used to safely store 40 PWR spent fuel assemblies, as proposed by the applicant. Characteristics of the spent fuel allowed to be stored and other operating limits are set forth in the SER, and are included in the technical specifications of the proposed license (Enclosure 1).

On July 28, 1992, NRC issued its environmental assessment (EA) for the proposed ISFSI at Prairie Island (Enclosure 3).

On August 4,1992, the staff, having completed its environmental review of the Prairie Island ISFSI, published a " Finding of No Significant Impact," in the Federal Reaister (57 FR 34319).

In its safety review of the Prairie Island site application, the staff took into account that the application, as supplemented, includes confirmation by the applicant's reactor safety committee that:

(a) no technical specification changes are required, under the Prairie Island Facility Operating Licenses, to

J The Commissioners,

accommodate a 10 CFR Part 72 license for onsite storage; (b) the joint operation of the reactor and of 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 NSP's application, on the basis of the staff's review, to be acceptable. Based on the staff's review of the applicant's t

submissions, the staff has found that there are no remaining unreviewed safety questions, and that all other pertinent regulatory requirements for l

authorization of issuance of the requested license have been satisfied.

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While the fuel assemblies stored in the TN-40 cask have been. evaluated to remain subcritical in compliance with 10 CFR Part 72, the proposed. license-

-r includes an exemption from 672.124(b) which requires the cask design to include a positive means to verify the continued. efficacy of solid. neutron absorbing materials. Specifically, because the neutron absorbing material (boron) is incorporated within metal matrix (Boral) of the cask basket material no credible mechanism has been identified to lose the boral in the i

basket of this cask. Thus, the staff has concluded that there is no public health and safety need to verify the continued. efficacy of the neutron absorbing material. Therefore, given the above circumstances, the staff has determined-that an exemption from the 672.124 requirement is authorized by law, will not endanger life or property or the common defense and security, and is otherwise in the public interest (10 CFR 72.7). Accordingly, an.

exemption is recommended from application of this general design criterion.

In view of the above, the staff has prepared its SER (Enclosure 2), for the Prairie Island ISFSI, 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 Prairie Island site spent fuel storage.

CONCLUSIONS:

8 The NRC staff has found that, based on its enclosed SER and the previously issued EA, there is reasonable assurance that the activities, authorized by a-license, to_ construct and operate the Prairie Island ISFSI, can be conducted without endangering the health and safety of the public, without significant environmental impact, and in compliance with the conditions of the. license, t

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.

3 COORDINATION:

Th'e' Office of the General Counsel has reviewed this' paper and has no legal-objections.

In a June 25, 1993 memorandum to the Commission, the General e

Counsel provided an analysis of the State of Minnesota court decision

- t concerning the proposed Prairie Island ISFSI..

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4 The Commissioners RECOMMENDATION:

That the Commission authorize issuance of a license to NSP, under Part 72, to receive, transfer, and store spent fuel in dry casks in an ISFSI, on the Prairie Island Nuclear Generating Plant site.

d-1 ies M. Tay1 xecutive Director for Operations

Enclosures:

1.

Proposed License

• 2.

Safety Evaluation Report

• 3.

Environmental Assessment

• Commissioners, SECY, 0GC only.

Commissioners' comments or consent should be provided directly to the Office of the Secretary by COB Wednesday, October 13, 1993.

Commission Staff Office comments, if any, should be submitted to the Commissioners NLT Tuesday, October 5, 1993, with an information copy to the Office of the Secretary.

If the paper is of such a nature that it requires additional review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.

DISTRIBUTION:

Commissioners OGC OCAA OIG OPA OCA OPP REGION III EDO ACRS ASLBP SECY

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UNITED STATES l

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NORTHERN STATES POWER COMPANY PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION MATERIALS LICENSE NO. SNM-2506 The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application, filed by the Northern States Power Company (NSP)

(applicant) for a materials license to receive, store, and transfer spent fuel from Prairie Island Nuclear Generating Plant in an independent spent fuel storage installation (ISFSI) located at its Prairie Island Nuclear Generating Plant site, meets the standards and requirements of the Atomic Energy Act of 1954, as amended (Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The Prairie Island ISFSI will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; 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 Prairie Island Nuclear Generating Plant Units 1 and 2; E.

The applicant's proposed ISFSI design complies with the 10 CFR Part 72, Subpart F, with the exception of 10 CFR 72.124(b);

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 Part 72; I.

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 1;

L.

The applicant's proposed decommissioning plan, pursuant to 10 CFR 72.30 provides reasonable assurance that the decontamination and decommissioning of the Prairie Island ISFSI at the end of its_.useful life will provide adequate protection to the health and safety of the publ ic.

M.

The applicant's proposed emergency plan complies with 10 CFR 72.32; N.

The applicant has satisfied the applicable provisions of 10 CFR Part 170; 5

0.

There is reasonable assurance (1) that the activities authorized by the license can be conducted without endangering the health and safety of the public, and (2) that such activities will be conducted in compliance with the regulations of the Commission set forth in 10 CFR Chapter I; and P.

The issuance of this license will not be inimical to the common defense and security nor to the health and safety of_ the public.

Accordingly, based on the foregoing findings, Materials License No. SNM-2506 is hereby issued to Northern States Power Company to read as follows:

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0 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 1, Part.72, and in reliance on statements and representations heretofore i

made by. the licensee (in the licensee's Prairie Island Independent Spent Fuel Storage Installation Technical Specifications and. Safety Analysis Report, submitted by letter dated August 31, 1990, as revised, supplemented, and submitted by letters dated October 29,1990; April 2, June 5, October 9, 31, November 15, December 11, 20, 23, 1991; January 17, February 6, 10, 12, March 2, 5, April 3, 22, 23, July 10, August 12, 13, 14, 1992; a license is hereby issued authorizing the licensee to receive, acquire, and possess the power reactor spent fuel and other radioactive materials associated with spent fuel storage designated-below; to use such materials for the purposes and at the place designated belo,,; and 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 hereaf ter in effect and to any conditions specified herein.

Licensee 1.

Northern States Power Company 3.

License Number: SNM-2506 2.

Address:

1993 414 Nicollet Mall 4.

Expiration Date:

, 2013 Minneapolis, Minnesota 55401 5.

Docket Number:

72-10 i

6.

Byproduct., source, and/

7.

Chemical and/or 8.

Maximum amount or special nuclear physical form that licensee may material possess at any one time under this license clad with A.

715.29 TeV of A.

Spent fuel assemblies A.

As U0,ium or zircon spent fuel from Prairie Island Nuclear Station Units 1 and 2 zirconium alloys assemblies reactor, using natural water for cooling and enriched not greater than 3.85 percent U-235, and associated radio-active materials related 'to receipt, storage, and transfer of the fuel assemblias l

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:

i The licensed material is to be received, possessed, transferred, and stored at the Prairie Island ISFSI located on the Prairie Island ' Nuclear Generating Plant site in Goodhue County, Minnesota.

11.

This site is described in Chapter 2 of the licensee's Technical Specifications and Safety Analysis Report (TSSAR) for the Prairie Island ISFSI.

12. The Technical Specifications _ contained in Appendix A attached hereto are i

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 ISFSI physical security, guard training and qualification, and

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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:

" Prairie Island Nuclear, Generating Plant Independent Spent Fuel Storage Installation Physical Security Plan," with Revision 0, submitted by letter dated March 10, 1992;

" Prairie Island Nuclear Generating Plant Independent Spent Fuel Storage Installation Security Force Training and Qualification Plan" with Revision j

0, submitted by letter dated March 10, 1992; and " Prairie Island Nuclear Generating Plant Independent Spent Fuel Storage Installation Safeguards i

Contingency Plan," with Revision 0, submitted by letter dated March 10, 1992.

14.

The Technical Specifications for Environmental Protection contained in Appendix A attached hereto are incorporated in the license.

Specifications required pursuant to 10 CFR 72.44(d), stating limits on the f'

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. TN-40 cask external surface contamination within. the limits of Technical Specification.3.4.1 ensures that the offsite_ dose will be i

inconsequential.

In addition, there are no normal or off-normal releases or effluents expected from the double-sealed storage casks of the ISFSI.

Specifications required pursuant to 10 CFR 72.44(d)(1), for operating procedures, for control of effluents, and for the maintenance and use of i

equipment in radioactive waste treatment systems, to meet the requirements

-i of 10 CFR 72.104 are not applicable.

There are,. by the design of the sealed storage casks at the ISFSI, no effluent releases.

'Also, cask i

loading and unloading operations and waste treatment will occur at the-1 Prairie Island Nuclear Generating Plant, under the specifications' of its i

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operating licenses.

15. No spent nuclear fuel shall be allowed to be loaded until such time as the following preoperational license conditions are satisfied:

i A.

A training exercise (Dry Run) of all TN-40 cask 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.

Moving cask in and out of spent fuel pool area b.

Loading fuel assembly (using dummy assembly) c.

Cask drying, sealing, and cover gas backfilling operations d.

Moving cask to, and placing it on, the storage pad e.

Returning the cask to the auxiliary building f.

Unloading the cask 9

Decontaminating the cask h.

All dry-run activities shall be done using written procedures I

i.

The activities listed above shall be performed. or modified and performed to show that each activity can be successfully executed before actual fuel loading.

B.

The Prairie Island Nuclear Generating Plant Emergency Plan shall be reviewed and modified, as required, to include the ISFSI.

C.

A training module shall be developed for the Prairie Island Nuclear Generating Plant Training Program, establishing an ISFSI Training and Certification Program that will include the following:

a.

TN-40 Cask Design (overview) b.

ISFSI facility Design (overview) l c.

ISFSI Safety Analysis (overview) 3 d.

Fuel loading and cask handling procedures and off-normal 1

procedures e.

ISFSI License (overview).

D.

The Prairie Island Nuclear Generating Plant Radiation Protection Procedures shall be reviewed and modified, as required, to include the ISFSI.

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E.

The Prairie Island Nuclear Generating Plant Administrative Procedures shall be reviewed and modified, as required, to include the ISFSI.

F.

A procedure shall be developed and implemented for the documentation of_the characterizations performed to select spent fuelLto be stored in the casks.

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 cask cavity for fuel loading and unloading activities.

H.

Written procedures _ shall be implemented to describe actions to be taken during operation and off-normal / 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 Reaulations.

All commitments to the applicable NRC regulatory guides and to engineering and construction codes shall be carried out.

17.

Fuel and cask movement and handling activities that are to be performed in the Prairie Island Nuclear Generating Plant Auxiliary Building will be governed by the requirements of the Prairie Island Nuclear Generating Plant Facility Operating Licenses (DRP-42, and -60) and associated Technical Specifications.

18.

Pursuant to 10 CFR 72.7, the licensee is hereby exempted from the provisions of 10 CFR 72.124(b) with respect to;providing positive means to verify the continued efficacy of the solid neutron absorbing materials.

4 19.

This license is effective as of the date of issuance shown below.

For the U.S. Nuclear Regulatory Commission Date of Issuance:

1993 by Division of Industrial-and Medical Nuclear Safety Washington, DC 20555

Attachment:

Technical Specifications 4

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION l

f APPENDIX "A" TO MATERIALS LICENSE SNM-2506-1 TECHNICAL SPECIFICATIONS ISSUED BY THE UNITED STATES NUCLEAR REGULATORY COMMISSION i

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t CONTENTS SECTION PAGE 1.0 DEFINITIONS 9

ADMINISTRATIVE CONTR0LS.......................

1-1 DESIGN FEATURES...............................

1-1 FUEL ASSEMBLY.................................

1-1 l

FUNCTIONAL AND OPERATING LIMITS...............

1-1 LIMITING CONDITIONS...........................

1-1 LOADING 0PERATIONS............................

1-1 SURVEILLANCE INTERVAL.........................

1-1 SURVEILLANCE REQUIREMENTS.....................

1-2 2.0 FUNCTIONAL AND OPERATING LIMITS 2.1 CASK VACUUM PRESSURE DURING DRYING............

2-1 2.2 CASK HELIUM BACKFILL PRESSURE.................

2-2 2.3 MAXIMUM CASK LIFTING HEIGHT...................

2-3 3/4.0 LIMITING CONDITIONS / SURVEILLANCE RE0VIREMENTS 3/4.1 FUEL TO BE STORED AT ISFSI....................

3/4-1 3/4.2 DISSOLVED BORON CONCENTRATION.................

3/4-3 3/4.3 MAXIMUM HELIUM LEAK RATE......................

3/4-4 3/4.4 MAXIMUM CASK REMOVABLE SURFACE CONTAMINATION...3/4-5 3/4.5 MAXIMUM CASK SURFACE TEr.PERATURE..............

3/4-6 3/4.6 DOSE RATES....................................

3/4-7 3/4.7 PRESSURE M0NITORING...........................

3/4-8 3/4.8 SAFETY STATUS SURVEILLANCE....................

3/4-9 5.0 DESIGN FEATURES......................................

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6.0 ADMINISTRATIVE CONTR0LS..............................

6-1 6.1 GENERAL.......................................

6-1 6.2 ENVIRONMENTAL MONITORING PROGRAM..............

6 6.3 ANNUAL ENVIRONMENTAL REP 0RT...................

6-1 TABLES 3/4-1 TN-40 CASK OPERATING LIMITS...................

3/4-10 3/4-2 SURVEILLANCE REQUIREMENTS

SUMMARY

3/4-11 i

' -j INTRODUCTION These Technical Specifications govern the safety of the receipt, possession, and storage of irradiated. nuclear fuel at the Prairie Island :ndependent Spent Fuel Storage Installation and the transfer of such irradiated nuclear fuel to and from Units I and 2 of the Prairie Island Nuclear Generating Plant and the Prairie Island Independent Spent Fuel Storage 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 TN-40 cask at the Prairie Island Nuclear Generating Plant Auxiliary Building is governed by the existing Prairie Island 10 CFR Part 50 operating licenses (DPR-42 and

-60), technical specifications, and new specific procedures.

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V I.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 Prairie Island 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 LIMITS:

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 nreparation steps before cask transport from the auxiliary building area.

g.

SURVEILLANCE INTERVAL: A surveillance interval is the interval-between a surveillance check, test, or calibration.

Unless specifically stated otherwise, the specific frequency for each surveillance requirement is met if the surveillance is performed within 1.25 times the interval specified in the frequency, as measured from the previous performance.

For frequencies specified as "once," the above interval extension does not apply.

If a required action requires performance of a surveillance, or its completion time requires periodic performance of "once per.

" the above frequency extension applies to the repetitive portion, not to the initial portion of the completion time.

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SURVEILLANCE 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, t

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SECTION 2.0 T i FUNCTIONAL AND OPERATING LIMITS

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2.1 CASK VACUUM PRESSURE DURING DRYING SPECIFICATION:

The cask cavity vacuum pressure during drying shall not exceed 10 mbar after stepped evacuation. The vacuum pressure shall be maintained for not less than 30 minutes.

APPLICABILITY:

Applicable to all casks.

ACTION:

If the required vacuum cannot be obtained.

1.

Check and repair vacuum drying system as necessary.

2.

Check and repair the cask seals as necessary.

If the specification is still not met, remove fuel from the cask.

BASIS:

A stable vacuum pressure of less than 10 mbar indicates that all liquid water has evaporated in the cask cavity, and that the resulting inventory of oxidizing gases in the cask is less than 0.25 percent of the volume.

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2.2 CASK HELIUM BACKFILL PRESSURE SPECIFICATION:

The cask cavity shall be backfilled with helium. The backfill pressure shall be 20 psia (1.4 bar)

I psia (70 mbar).

APPLICABILITY:

Applicable to all casks.

ACTION:

If the required pressure cannot be obtained:

1.

Check and repair the cask seals as necessary.

2.

If the backfill pressure exceeds the criterion, release a sufficient quantity of helium to lower the cask cavity pressure.

If the specification is still not met, remove fuel from the cask.

BASIS:

The thermal analysis performed for the cask assumes the use of helium as a cover gas. Also, the use of an inert gas (helium) ensures long-term maintenance of fuel clad integrity.

The value of 20 psia (1.4 bar) was selected to ensure that the pressure within the cask remains within the pressure design limits.

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i 2.3 MAXIMUM CASK LIFTING HEIGHT SPECIFICATION:

The cask lifting height with a non-single-failure-proof lifting device shall not exceed 46 cm (18 in.).

APPLICABILITY:

This specification applies to handling of a loaded cask outside the auxiliary building.

ACTION:

In the event of a cask drop from a height greater than 18 inches (45 cm), with fuel in the cask, 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 cask shall be removed from service and evaluated for further use or disposed of, as may be appropriate.

BASIS:

The drop analyses performed for cask drop incidents, for a cask loaded with spent fuel, confirm that drops up to 18 inches (45 cm) can be sustained without unacceptable damage to the cask. This limiting condition ensures that the i

handling height limits will not be exceeded at the storage pad nor in transit to and from the spent fuel pool. Design of the cask is to ASME B&PV Code Section III, Division 1, Subsection NB for Class I components, Service Level D requirements.

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SECTION 3/4.0

- i LIMITING CONDITIONS / SURVEILLANCE REQUIREMENTS t

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s 3/4.1 FUEL TO BE STORED AT ISFSI LIMITING CONDITION FOR OPERATION 3.1.1 The spent nuclear fuel to be received and stored in the TN-40 cask.

at the Prairie Island ISFSI shall meet the following requirements:

(1)

Only fuel irradiated at the Prairie Island Nuclear Generating Plant Unit Nos. I and 2 may be used.

(2)

Maximum initial enrichment shall not exceed 3.85 weight percent U-235.

(3)

Maximum assembly average burnup shall not exceed 45,000 megawatt-days per metric ton uranium.

(4)

Fuel shall have cooled a minimum of 10 years after reactor discharge, before storage in the ISFSI.

(5)

Fuel shall be intact unconsolidated fuel.

Partial fuel assemblies, that is, fuel assemblies from which fuel pins are missing must not be loaded unless dummy fuel pins are used to displace an amount of water equal to that displaced by the original pins.

(6)

Fuel assemblies known or suspected to have structural defects or gross cladding failures (other than pinhole leaks) sufficiently severe to adversely affect fuel handling and transfer capability shall not be loaded into the cask for storage.

APPLICABILITY:

This specification is applicable to all spent fuel to be loaded and stored in the TN-40 cask at the Prairie Island ISFSI.

ACTION:

If the requirements of the above specification are not met, do not load the fuel assembly into the TN-40 cask.

SURVEILLANCE RE0VIREMENTS:

4.1.1 Each fuel assembly to be loaded shall have the above specifications independently verified and documented.

4.1.2 Before inserting a spent fuel assembly into a cask and again before closing the cask, the identity of each fuel assembly shall be independently verified and documented.

BASIS:

The design criteria and subsequent safety analyses of the Prairie Island ISFSI and storage casks assumed certain characteristics and limitations for the fuel that is to be stored.

Specification 3/4.1 ensures that the 3/4-1

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4 integrity of the fuel is protected by defining characteristics such as:

the source of the spent fuel, maximum initial enrichment, irradiation history, and i

l minimum post-irradiation cooling time.

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This specification was derived to ensure that the peak fuel rod temperatures, cask surface contact dose rates, reactivity, and fuel mass are below-the design values.

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3/4.2 DISSOLVED BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.2.1 The cask cavity shall be moderated only by water with a boron concentration greater than or equal to 1800 ppm.

APPLICABILITY: Applicable to all loading and unloading of casks.

ACTION:

1.

With the measured boron concentration less than the specification-before the beginning of cask loading and unloading operations,.

suspend all activities involving cask loading and unloading.

2.

With the measured boron concentration less than the specification during cask loading and unloading operations, suspend all loading and unloading operations until the baron concentration is increased to 1800 ppm or greater.

SURVEILLANCE RE0VIREMENTS 4.2.1.1 Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before insertion of the first spent fuel assembly into a cask, verify and document that the dissolved boron concentration in water in the spent fuel pool and introduced into the cask cavity satisfies the limits specified above, in accordance with the requirements of the Prairie Island Nuclear Generating Station Operating Licenses (DPR-42 and -60).

4.2.1.2 Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before flooding the cask cavity for unloading the fuel assemblies, verify and document that the dissolved boron concentration in water in the spent fuel pool and the water to be introduced into the cask cavity satisfies the limits specified above, in accordance with the requirements of the Prairie Island Nuclear Generating Station Operating Licenses (DPR-42 and -60).

If the water introduced into the cask cavity'is not from the spent fuel pool, then the dissolved boron concentration shall be independently determined by chemical analysis (two samples analyzed by two different individuals). All boron concentration measurements shall be documented.

BASIS: This specification ensures that k., is less than 0.95, and therefore, the spent fuel is subtritical during fuel loading and unloading.

3/4-3

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3/4.3-MAXIMUM HELIUM LEAK RATE LIMITING CONDITION FOR OPERATION 3.3.1 The standard helium leak rate for all closure seals shall not exceed 10-5atm-cc/s.

APPLICABILITY:

Applicable to all casks.

ACTION: With the requirements of the above specifications not satisfied, the seals shall be repaired or replaced in accordance with approved procedures and -

re-examined in accordance with these specifications.

SURVEILLANCE RE0VIREMENT:

4.3.1 During cask loading operations, the cask seals shall be tested in equal to 10',ith ANSI N 14.5, to ensure that the seal leakage is less than or accordance w atm-cc/s.

BASIS: The safety analysis of the cask is based.on the seals being tight to maintain a leak rate less than 10',atm-cc/s.

Seal tightness at this leak rate

-t will ensure the helium atmosphere in the storage cask is maintained for the licensed period.

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4 3/4.4 MAXIMUM CASK REMOVABLE SURFACE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.4.1 Removable contgmination on the cask exterior surfaces shall be less 2

than 1,000 dpm/100 cm (p.2 Bq/cm ) from beta and gamma sources, and 20 e

2 dpm/100 cm (0.003 Bq/cm ) from alpha sources.

APPLICABILITY: Applicable to all casks.

ACTION:

If the limit is exceeded, the cask external surfaces shall be.

decontaminated to meet the specification before movement to the ISFSI.

SURVEILLANCE RE0VIREMENT:

4.4.1 Contamination surveys shall be taken on the accessible cask exterior surfaces. The contamination surveys for removable surface contamination shall be conducted after fuel loading and before moving the loaded cask to the ISFSI site.

EASIS: Compliance with this limit ensures that the offsite dose limits in 10 CFR Part 20, 10 CFR Part 50 - Appendix I, 10 CFR Part.72, and 40 CFR 190 are met.

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3/4-5

3/4.5 MAXIMUM CASK SURFACE TEMPERATURE LIMITING CONDITION FOR OPERATION 3.5.1.

The equilibrium cask surface temperature shall not exceed 250' F (121*C).

APEljCABILITY: This temperature limit applies to all casks stored at the ISFSi.

ACTION:

If a cask surface temperature greater than 250' F (121*C) is observed for any cask, then this indicates that the cask is not performing as intended, or that fuel assemblies not meeting Specification 2.1 have been loaded into

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the cask.

If after verification, fuel assemblies meeting Specification 2.1 have been loaded into the cask and the cask surface temperature is greater l

than 250' F (121*C), then the cask shall be unloaded. A written report shall be submitted to the Nuclear Regulatory Commission Region III Office, with a copy to the Director, Office of Nuclear Material Safety and Safeguards, within 30 days of this incident.

SURVEILLANCE RE0VIREMENT:

4.5.1.

Cask surface temperatures shall be measured and recorded at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after completing cask loading and before moving the cask to the ISFSI.

BASIS: This is to ensure that the fuel clad will be at a temperature such that it will be protected against degradation that leads to gross rupture.

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9 3/4.6 DOSE RATES LIMITING CONDITION FOR OPERATIONS 3.6.1 The contact dose rate (neutron + gamma) on the surface of the cask shall not exceed 200 mrem /hr (2.0 mSv/hr) or the equivalent dose rate at 2 yards (2 meters).

APPLICABILITY: This specification is applicable to the accessible top and side surfaces of a loaded cask.

ACTION:

If the measured dose rate exceeds tne limit, correct fuel loading shall be verified.

If correct fuel is loaded, specific analysis must demonstrate compliance with 10 CFR Part 20, and 10 CFR Part 72 radiation

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protection requirements, or appropriate action must be taken to comply with-acceptable limits.

If acceptable limits cannot be achieved, the cask shall not be placed in service at the ISFSI.

SVRVEILLANCE REOUIREMENTS:

4.6.1 Cask surface gamma and neutron dose rates shall be measured before moving a cask to the ISFSI. Measurements shall be taken near the accessible top and side surfaces.

4.6.2 Two (2) thermoluminescent dosimeters (TLDs) shall be placed on the fence at each side of the ISFSI site (8 total) and read quarterly to determine ISFSI radiation levels.

BASIS: The basis for this specification is.the shielding analysis presented in Appendix 7A of the ISFSI Safety Analysis Report (SAR).

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 the casks. Compliance with the cask surface dose rates (as described in ISFSI SAR Table 7A-4) will ensure compliance with the ISFSI site radiation protection requirements (as described in ISFSI SAR Tables 7A-5 and 7A-7, and ISFSI SAR Figure 7A-6).

3/4-7 l

4 3/4.7 PRESSURE MONITORING LIMITING CONDITION FOR OPERATIONS 3.7.1 The alarm board that monitors cask pressure shall be checked daily and tested annually to ensure that the helium atmosphere in the casks is-maintained.

APPLICABILITY: Applicable to all casks.

ACTION:

If monitcring of pressure between the cask double seals indicates loss of pressure and seal leakage, return the cask to the auxiliary building and repair or replace the seals, as necessary, to return the cask to proper operation.

SURVEILLANCE RE0VIREMENTS:

4.7.1 The alarn board, to which pressure monitoring devices are connected, shall be checked daily.

4.7.2 The alarm board shall be tested annually, to ensure proper functioning.

BASIS:

Pressure between the cask seals must be maintained to ensure that the helium atmosphere in the cask is maintained.

Periodic testing of the alarm board ensures proper functioning of the pressure monitoring and alarm system, to provide timely corrective action.

3/4-8 1

3/4.8 SAFETY STATUS SURVEILLANCE LIMITING CONDITION FOR OPERATIONS 3.8.1 The cask shall be free of damage or debris, to maintain' proper functioning of the cask.

APPLICABILITY: Applicable to all casks.

ACTION:

If significant damage, deterioration, or debris accumulation occurs to the cask surfaces such that the safety. functions of the cask are impaired, take appropriate corrective action to return the cask to proper operation.

SVRVEILLANCE RE0VIREMENT:

4.8.1 A visual serveillance of all casks at the ISFSI shall be conducted, on a quarterly basis, to determine that no significant damage nor deterioration of the exterior of the casks has occurred and that no significant accumulation of debris on cask surfaces has occurred.

SASIS:

These surveillance requirements shall ensure cask maintenance.

i 3/4-9

J Table 3/4-1 TN-40 CASK OPERATING LIMITS Operating Limit Maximum Lifting Height with a Non-Redundant Lifting Device 18 inches (45 cm)

Maximum Cask Surface Temperature 250*F (121*C)

Maximum Surface Dose Rate (or equivalent at 2 meters) 200 mrem /hr Maximum Removable Surface Contamination 2

- Beta and Gamma 1000 dpm/100 cm 2

- Alpha 20 dpm/100 cm Maximum Helium Leak Rate 10~5 atm-cc/s Initial Helium Pressure (Cask Cavity) 20 i I psia (1.4 bar i 70 mbar)

Pressure During Cask Drying Test (held for 30 min.)

$ 10 mbar Boron Concentration in Pool & Cask 1 1800 ppm Storage Capacity 1 40 assemblies Fuel Assembly Characteristics

- Initial Enrichment, U-235 s 3.85 wt. %

- Average Burnup 5 45,000 MWD /MTU

- Time after Irradiation 2 10 Years i

i 3/4-10 i

Table 3/4-2 SURVEILLANCE REQUIREMENTS

SUMMARY

Specification Quantity or Item Period 3/4.1 Fuel to be Stored at ISFSI P, C 3/4.2 Dissolved Boron Concentration P

3/4.3 Maximum Helium Leak Rate L

3/4.4 Maximum Removable Surface Contamination L'

3/4.5 Maximum Cask Surface Temperature S

3/4.6 Dose Rates Cask surface or equivalent at 2 yards-(2 meters)

L At the Fence Q

3/4.7 Pressure Monitoring D, A 3/4.8 Safety Status Surveillance Q

Leoend P - Prior to cask loading.

C - Prior to cask closure following loading.

S - At least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after cask loading and prior to moving cask to storage pad.

L - Prior to moving cask to the storage pad.

Q - Quarterly. -- at least once per 92 days.

A - Annually -- at least once per 366 days.

D - Daily.

Note:

Specified time periods or frequencies may be adjusted by 25 percent to accommodate normal test schedules (see Section 1.0 Definitions,

g. " Surveillance Interval.")

3/4-11 i

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,9' SECTION 5.0 DESIGN FEATURES P

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5.0 DESIGN FEATURES The Prairie Island ISFSI design approval was based on use of the TN-40 storage.

cask and review of specific design drawings, some of which have been deemed appropriate for inclusion in the Prairie Island ISFSI Safety Evaluation Report (SER).

Drawings listed in Section 1.2 of the Prairie Island ISFSI SER have been reviewed and approved by NRC.

These drawings may be revised under the provisions of 10 CFR 72.48, as appropriate.

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SECTION 6.0 V

lL ADMINISTRATIVE CONTROLS i -.

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4 6.0 ADMINISTRATIVE CONTROLS 6.1 GENERAL The Prairie Island ISFSI is located on the Prairie Island Nuclear Generating Plant site and will be managed and operated by the Northern States Power-Company staff. The administrative controls shall be in accordance with the requirements of the Prairie Island Nuclear Generating Plant Facility Operating Licenses (DPR-42, and -60) and associated Technical Specifications, as appropriate.

6.2 ENVIRONMENTAL NONITORING PROGRAM The licensee shall include the Prairie Island ISFSI in the environmental monitoring program for the Prairie Island Nuclear Generating Plant. An environmental monitoring program is required pursuant to 10 CFR 72.44(d)(2).

The licensee shall include the ISFSI in the environmental monitoring report for the Prairie Island Nuclear Generating Plant, and a copy shall be sent to the Director, Office of Nuclear Material Safety and Safeguards.

6.3 ANNUAL ENVIRONMENTAL REPORT An annual report, as required by 10 CFR 50.36a(a)(2), which is the Prairie Island Nuclear Generating Plant Annual Radioactive Effluent Release Report, shall include the Prairie Island ISFSI and shall be submitted to the NRC Region III, Office, with a copy to the Director, Office of Nuclear Material Safety and Safeguards, within 60 days after January 1 of each year. This report should specify the quantity of each of the principal radionuclides released to the environment in liquid and in gaseous effluents during the previous year of 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.

The report under this specification is also required pursuant to 10 CFR 72.44(d)(3).

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1 SAFETY EVALUATION REPORT FOR TIIE PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION s

i U.S. Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards July,1993 i

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at Table of Contents Page 1 GENERAL DESCRIPTION 1-1 1.1 In trod uction.............................

1-1 1.2 General Description of the Storage Cask 1-1 1.2.1 Cask Design Characteristics......................

1-1 l.2.2 Operational Features 1-3 1.2.3 Cask Controls...............

1-3 1.3 Identification of Agents and Contractors....

. 1-4 l.4 Cask Arrays...............

1-4 2 PRINCIPAL DESIGN CRITERIA....

2-1 1

2.1 Introduction..................

............. 2-1 2.2 Fuel to be Stored.............

...................21 I

2.3 Quality Standards

.................................2-1 2.4 Protection against Environmental Conditions and Natural Phenomena

.. 2-2 r,

2.4.1 Tornado and Wind Loading..

. '2-2 2.4.2 Flood

.................2-2 2.4.3 S ei s mic................................ 2 -3 2.5 Protection against Fire and Explosions....................

2-3 2.6 Confm' ement Barriers and Systems......................

2-3 2.7 Instrumentation and Control Systems......................

2-41:

2.8 Criteria for Nuclear Criticality Safety.....................

2-4 2.9 Criteria for Radiological Protection......................

2-4 2.10 Criteria for Spent Fuel and Radioactive Waste Storage and Handling..

2-5 2.11 Criteria for Decommissioning..........................

2-6 3 STRUCTURAL EVALUATION..........................

3-1 3.1 Area of Review

..................................3-1 3.2 Acceptance Criteria.................

..............3-1 3.3 Review

....................................... 3-1 3.3.1 Cask Loads................................

3-2 3.3.2 Cask Materials..............................

3-2 3.3.3 Review of Cask Body 3-3 3.3.4 Review of Basket...

3 -4 3.3.5 Review of the Neutron Shield, Trunions, and Other Components 3-4 3.3.6 Fuel Assemblies........

3-5 3.4 Findings and Conclusions.............................

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4 THERMAL EVALUATION...............

............4-1 4.1 Area of Review...................................

4-1 4.2 Acceptance Criteria

. 4-1 4.3 Review....

4-2 4.3.1 Material Properties and Component Limitations...

4-2 4.3.2 Normal Conditions of Storage 4-2 4.3.3 Accident Conditions

......................,..4-3 4.4 Findings and Conclusions.............

4-3 5 SHIELDING EVALUATION 5-1 5.1 Area of Review....

5-1 5.2 Acceptance Criteria

......5-1 5.3 Review

. 5-1 5.3.1 Source Specification 5-2 5.3.2 Model Specification.

...................5-5 5.3.3 Shielding Evaluation

. 5-9 5.4 Findings and Conclusions..

5-12 6 CRITICALITY EVALUATION.........

6-1 6.1 Area of Review......

6-1 6.2 Acceptance Criteria..............

.................6-1 6.3 Review 6-1 6.4 Findings and Conclusions.

6-3 7 CONFINEMENT EVALUATION 7-1 7.1 Area of Review..................................

7-1 7.2 Acceptance Criteria 7-1 7.3 Review

..................... 7-1 7.3.1 Cask Confinement Description and Design Features.......

7-1 7.3.2 Gaseous Activity Inventory within the Cask Cavity.......

7-4 7.3.3 Estimated Annual Does Equivalent from Gaseous Activity Release due to Normal Operations............

7-4 7.3.4 Estimated Annual Dose Equivalent from Gaseous Activity Release due to Postulated Off-normal Events...........

7-.5 7.4 Findings and Conclusions.

7-5 8 OPERATING PROCEDURES..

..........................8-1 8.1 Area of Review...............................

. 8-1 8.2 Acceptance Criteria................................

8-1 8.3 Review

.......................................8-1 8.4 Findings and Conclusions...............

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9 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM...........

9-1 9.1 Acceptance Tests....

......,9-1 9.2 Maintenance Program.....

. 9-1 10 RADIATION PROTECTION....

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10.1 Area of Review..............

10-1 10.2 Acceptance Criteria...

10-1 10.3 Review

...............................10-2 10.3.1 Ensuring That Occupational Radiation Exposures Are ALARA 10.

10.3.2 Radiation Protection Design Features........

10-3 10.3.3 Estimated Occupational Radiation Exposure Assessment...

10-5 10.4 Findings and Conclusions...

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11 ACCIDENT ANALYSIS 11-1 11.1 Area of Review.......

11-1 11.2 Acceptance Criteria 11-1 11.3 Review 11-1 11.4 Findings and Conclusions.

. 11-3 12 DECOMMISSIONING..

12-1 12.1 Area of Review......

12-1 12.2 Acceptance Criteria.....

12-1 12.3 Review

. 12-1 12.3.1 Unloading of the Cask.................

. 12-1 12.3.2 Decommissioning of the Cask Components 12-2 12.3.3 Decommissioning of the Storage Pad..............

12-3 12.3.4 Decommissioning Funding 12-3 12.4 Findings and Conclusions.............................

12-4 13 OPERATING CONTROLS AND LIMITS......................

13-1 13.1 Area of Review...

13-1 13.2 Acceptance Criteria 13-1 13.3 Review 13-1 13.4 Findings and Conclusions...........

13-1 14 QUALITY ASSURANCE...

..., 14-1 15 REFERENCES 15-1 i

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I GENERAL DESCRIPTION 1.1 Introduction This Safety Evaluation Report (SER) documents the staff's review and evaluation of the Technical Spe ifications and Safety Analysis Report (TSSAR) for the Independent Spent Fuel l

Storage Installation (ISFSI) that is planned by Northern States Power Company (NSP) to store fuel assemblies discharged from the Prairie Island Nuclear Generating Plant, Units 1 and 2 l

(Reference 1). Transnuclear designed the TN-40 Spent Fuel Storage Cask specifically for this ISFSI. The TSSAR was prepared primarily by Transnuclear, Inc., with input provided by NSP and Stone and Webster, Inc. (Archited Engineer for the Prairie Island ISFSI), using Regulatory Guide 3.48 (Reference 2) format, as applicable. This SER uses the format of Regulatory Guide 3.61 (Reference 3), with some differences in the section numbering.

The staff's review of the TSSAR addresses the handling, transfer, and storage of spent fuel in a TN-40 dry storage cask for an at-reactor site ISFSI. Such storage in an ISFSI would be licensed under 10 CFR Part 72, Licensing Requirements for the Storage of Spent Fuel in a Independent Spent Fuel Storage Installation (ISFSI)." In the TSSAR, a single dry storage cask design, the TN-40, is presented.

The staff's assessment is based on the application meeting the applicable requirements of 10 CFR Part 72 for independent storage of spent fuel, and of 10 CFR Part 20 for radiation protection.

Decommissioning, to the extent that it is treated in this TSSAR, presumes, as a bounding case, unloading of a TN-40 cask at the reactor site and subsequent decontamination of the cask before its disposition or disposal.

Use or certification of the TN-40 cask, under 10 CFR Part 71, for offsite transport of spent fuel, is not a subject of this safety evaluation.

1.2 General Description of the Storage Cask 1.2.1 Cask Design Characteristics The TN-40 dry storage cask (see Figure 1.3-2 to 1.3-7 in the TSSAR) was developed by Transnuclear, Inc., and is designed for the storage ofirradiated spent fuel assemblies. The TN-40 cask body is a right circular cylinder that is composed of the following components:

confmement vessel with bol:ed closure; basket for fuel assemblies; gamma shield; trunnions; neutron shield; weather cover; pressure monitoring system; and outer shell.

I The confmement vessel is a welded cylindrical stainless steel (SA-203) vessel that is 38-mm (1.50-inches) thick. A flange is welded to the confmement vessel, to accommodate a bolted closure. The closure consists of two pieces welded together. The outer piece is SA-350, Grade LF3 forged steel and the inner piece is SA-203 stainless steel, with a combined thickness of 267 mm (10.50 in.). The closure lid uses a double barrier seal system with two metallic O-rings 1-1

I forming the seal. The annular space between the metallic O-rings is connected to a helium-filled tank placed between the lid and the protective cover. Pressure m the tank is maintained above the pressure in the cask cavity, to prevent either flow of fission gases out, or air into, the cask cavity, which, under normal storage conditions, is filled with helium.

Surrounding the confinement vessel is a forged steel gamma shield (SA-516, Grade 70) with a wall thickness of 222 mm (8.75 in.). The gamma shield on the bottom end is made of the same material and has the same thickness. The bolted closure provides the gamma shielding at the upper end of the cask body.

i Neutron emissions from the stored fuel are attenuated by a borated polyester neutron shield located on the outside of the outer shell. The borated polyester is ll4-mm (4.50-in) thick and is encased in aluminum boxes that are 3-mm (0.12 in.) thick, which are held in place by a shell of 13-mm (0.50-in.) thick SA 516 Gr 55 steel. Neutron emissions from the top of the cask are attenuated by a slab of borated polyester that is 102-mm (4.0-in.) thick, which is encased in SA-516 Gr 55 steel 13-mm (0.50-in.) thick.

The TN-40 spent fuel storage cask has a cylindrical cavity 1829 mm (72.0 in.) in diameter, 4140 mm (163 in.) long, which holds a fuel basket that is designed to accommodate 40 fuel assemblies that have been discharged from the Prairie Island Nuclear Generating Plant, Units I or 2. The i

fuel basket has 40 cavities, each 204-mm (8.05-in) square, to hold the fuel bundles. The fuel cavities are formed by a matrix of aluminum plates, boral plates, and stainless steel tubes. The plates and tubes are attached by a series of stainless steel plugs that are attached to the stainless steel tubes by gas tungsten arc weld (GTAW). The combined wall thickness is 19-mm (0.75-in.)

thick. The basket is guided into the cask body and held in place by aluminum rails that run the axial length of the cask body.

A protective cover, 9.5-mm (0.375-in.) thick, is bolted to the cask body, to provide weather protection for the lid penetrations. In addition, all exposed surfaces have a sprayed metallic coating, Zn/Al, and are painted for weather protection and ease of decontamination. The neutron shield, helium pressurization system, and shield cap are placed on top of the cask after fuel loading and removal from the spent fuel pool.

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The TN-40 cask body has four trunnions that are welded to the gamma shield. Two of these are located near the top of the cylindrical steel forging, spaced 180 degrees apart, and are used for lifting the cask. The remaining two tamnions are 180 degrees apart, and located near the i

bottom of the cask. The lower trunnions are used to support the cask when it is empty and in a horizontal orientation, which is intended to occur only during transport from the fabrication

j facility to the reactor site, or when the cask is being rotated to a vertical position, on receipt at f

the reactor site. The lifting trunnions have an effective diameter of 286 mm (11.25 in) and are hollow, to permit installation of neutron shielding material, to eliminate a path for neutron streaming. Each lifting trunnion is designed to meet the requirements of NUREG-0612 for a t

non-redundant lifting fixture. Two similar rotation inmnions are attached to the lower part of the cask body. The rotation trunnions are not used when the cask contains fuel assemblies, and 1-2 i

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have not been analyzed as a safety component.

The TN-40 cask has three containment penetrations; one cask cavity drain, one cask cavity vent, and one interseal overpressure port. Each of these penetrations is in the lid. The drain and vent ports are covered by bolted flanges with double concentric metallic O-ring seals. The cavity r

drain line penetrates the closure lid and terminates in a relief in the bottom of the cask cavity.

This is used to drain water from the cask cavity after underwater fuel loading. It is also used during the drying and helium back-filling of the cask cavity. The drain valve is of the quick-i disconnect type and was not analyzed as part of the primary confinement system. The cavity vent is identical to the drain line.

The overpressure port penetrates the closure lid to the space between the two 0-ring seals. A bolted flange with a single metallic O-ring seal connects the line from the overpressure tank to the closure lid.

The overall dimensions of the cask are 5131-mm (202-in.) long and 2565-mm (101-in.) in diameter. The cask weighs approximately 109 metric tons (240,690 pounds), when loaded and placed at the ISFSI.

1.2.2 Operational Features The TN-40 cask is designed to store 40 intact fuel assemblies, after irradiation and cooling in the Prairie Island Nuclear Generating Plant, Units 1 or 2. Each fuel assembly is assumed to have a maximum initial enrichment not to exceed 3.85 */, U-235 in uranium.

Funher assumptions limit the fuel to a maximum of 45,000 MWD /MTU burnup, a minimum decay time of 10 years after reactor discharge and a maximum decay heat load of 0.675 kW per assembly for a total of 27 kW for a TN-40 cask.

1 The heat rejection capability of the TN-40 cask maintains the maximum fuel rod clad temperature below 336*C (636*F), based on normal operating conditions with a 27 kW (25.6 Btu /s) decay heat load, 38'C (100*F) ambient air, and full insolence. The fuel assemblies are stored in an inert helium gas atmosphere.

The shielding features of the TN-40 cask are designed to maintain the maximum combined

-i gamma and neutron surface dose rate to less than 1.25 mSv/hr (125 mrem /hr), under normal

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operations conditions.

The criticality control features of the TN-40 cask are designed to maintain the ' neutron f

multiplication factor k-effective (including uncertainties and calculational bias) at less than 0.95, under all conditions.

-i 1.2.3 Cask Contents The type of spent fuel to be stored in the TN-40 cask is light water reactor (LWR) fuel of the i

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pressurized water reactor (PWR) type, which is unique to the Prairie Island Nuclear Generating Plant, Units 1 or 2. The 14 x 14 Optimized Fuel Assembly (OFA) was used as the reference design in this TSSAR. PWR fuel is made of short cylinders (pellets) of high-fired ceramic uranium dioxide (UO ). These pellets are 8.75 mm (0.3444 in)in diameter. A 3658-mm (144-2 in.) long stack of these pellets is loaded and hermetically sealed into a zirconium alloy tube.

Fuel rods are assembled into bundles in a square array, each spaced and supported by grid structures. The assembly has a top and bottom fitting. The overall dimensions of a 14 x 14 OFA assembly are approximately 197.2-mm (7.763-in.) square by 4097-mm (161.3-in.) long.

1.3 Identification of Agents and Contractors Transnuclear Inc., provides design, engineering, and analysis for the TN-40 cask. It was incorporated in the State of New York in 1965 and has offices in Hawthorne, New York, and Aiken, South Carolina. Transnuclear shares are privately held by Transnucleaire, S. A. of Paris, France.

4 The TN-40 cask may be manufactured by one or more qualified organizations. There are no other agents or contractors involved with the TN-40 cask.

Northern States Power Company provides quality assurance for the manufacturing of the TN-40 cask. In addition, NSP will be the owner and operator of the TN-40 casks and will be responsible for all operations of the ISFSI, including cask loading and transport from the fuel building to the ISFSI. NSP also has responsibility for the security of the ISFSI.

Stone and Webster Engineering, Inc., provides the design, engineering, analysis, and quality assurance for the ISFSI, particularly the earthen berm that partially surrounds the ISFSI and the concrete pad that supports the TN-40 casks.

1.4 Cask Arrays The ISFSI may include up to 48 T'

> casks. The casks will be arranged in two groups with I

each consisting of two rows of 12 casks aligned approximately in an east-west direction. The TN-40 cask shall be stored vertically on its bottom plate on a concrete pad. The TSSAR provides analyses of this mode of storage.

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2 PRINCIPAL DESIGN CRITERIA i

2.1 Introduction l

Subpart F of 10 CFR Part 72 sets forth general design criteria for the design, fabrication, construction, testing, and performance of structures, systems, and components important to safety in an ISFSI.

In this chapter, we discuss the applicability of these criteria to the Transnuclear TN-40 dry storage cask and the degree to which the TSSAR is in compliance with these criteria. Section headings in this chapter generally correspond to sub-sections of Subpart F of Part 72.

2.2 Fuel to be Stored

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The TN-40 cask is designed to store in a dry condition irradiated PWR fuel from nuclear power stations. The design basis fuel is LWR fuel of the PWR type, which is unique to the Prairie Island Generating Plant, Units I or 2. The 14 x 14 OFA was used as the reference design in this TSSAR. PWR fuel is made of short cylinders (pellets) of high-fired ceramic uranium dioxide clad with zircaloy and enriched to 3.85 percent "U by weight. This fuel assembly type 2

is considered to be limiting for thermal, radiological and criticality considerations. The design basis fuel is assumed to have been irradiated to an exposure no greater than 45,000 MWD /MTU and cooled for not less than 10 years. Estimates of the radionuclide' activity in the spent fuel assemblies described above were made using the ORIGEN 2 computer code.

2.3 Quality Standards Classification of quality standards for structures, systems, and components important to safety are required by 10 CFR 72.122 (a). Section 5.4 of the TSSAR identifies three categories of safety that are applied to all significant cask components. The categories are identified as

" Safety-Related," " Augmented Quality," and "Non-QA Related." Safety-related implies critical or major impacts on safety, and would be required to be designed to accord with quality standards. Augmented quality implies minimal contribution to mitigation of accidents; however, the component provides a significant function during normal operation. A quality standard provides numerical criteria or acceptable methods or both for the design, fabrication, testing, and performance of these structures, systems, and components important to safety. These j

standards should be selected or developed to provide sufficient confidence, in the capability of the structure, system, or component, to perform the required safety function. Since quality standards are generally embodied in widely accepted codes and standards dealing with design procedures, materials, fabrication techniques, inspection methods, etc., judgments regarding the adequacy of the standards cited in the TSSAR are presented in the sections of this report where the standards are applicable.

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2.4 Protection against Environmental Conditions and Natural Phenomena 10 CFR 72.122(b) requires the licensee to provide protection against environmental conditions and natural phenomena. Section 3.2 of the TSSAR describes the structural and mechanical criteria for tornado and wind loadings, flood potential, tornado missile protection, seismic design, snow and ice loadings, thermal loadings, combined load criteria, and structural design criteria.

In this section, the discussion is limited to the adequacy of the criteria for protecting against 1

environmental conditions and natural phenomena. The technical basis for accepting these criteria is defined by the regulatory requirement to consider the most severe of the natural phenomena reported for the site, with appropriate margins to take into account the limitations of the data.

The regulatory requirement is interpreted to mean that protection against environmental conditions and natural phenomena should be accomplished either by the criteria specified in the TSSAR or for the most severe natural phenomena that may occur within the boundaries of the United States.

2.4.1 Tornado and Wind Loading The TSSAR establishes the following conditions in Section 3.2.1 as the design basis tornado loadings.

A differential pressure of 20.7 kPa (3 psi), developing during a period of 3 seconds.

A wind having a peripheral tangential velocity of 483 km/h (300 mph) and a forward progression of 97 km/hr (60 mph).

The design tornado-driven missile was assumed to be either an airborne 10 cm x 30 cm x 366 cm (4"x12"x12') plank traveling end on at 483 km/h (300 mph), or a 1814 Kg (4000-lb.) automobile traveling at 80 km/h (50 mph) not more than 7.6 m (25 ft.) above the ground level.

These are consistent with the tornado and wind loading previously found acceptable in the Safety Analysis Report for the Prairie Island Nuclear Generating Plant.

2.4.2 Flood Section 3.2.2 of the TSSAR adopts flood criteria, for the ISFSI, that are identical to the criteria for Class 1 (Safe Shutdown) structures of the Prairie Island Nuclear Generating Plant. Section 2.4.1 of the TSSAR describes the flood. The 100-year flood depth of the water is 209.5 m (687.4 feet) above Mean Sea Level (MSL). The upper surface of the concrete pads that support the storage casks is at an elevation of 210.8 m (693.5 feet) above MSL, so the 100-year flood will not result in water reaching the storage casks. However, the probable maximum flood including maximum discharge from Lock and Dam Number 3, is 214.4 m (703.6 feet) above.

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MSL and wave effects would increase the maximum water level to 215.4 m (706.7 feet) above MSL. The cask cavity seals are located 4.553 m (179.25 inches) above the bottom of the cask which is equivalent to 215.9 m (708.4 feet) above MSL. Consequently, the probable maximum flood would partially submerge the casks, but would not result in water reaching the level of the seals, so no overpressure would be applied to the seals.

During the probable maximum Dood, the maximum velocity of the flood water is 1.9 m/s (6.2 fthec.) Section 3.2.2 of the TSSAR demonstrates that flood waters with this velocity will not cause a cask to tip over or slide.

2.4.3 Seismic Section 3.2.3 of the TSSAR adopts seismic criteria, for the ISFSI, that are identical to the criteria for Safe Shutdown Earthquakes (SSEs), at the Prairie Island Nuclear Generating Plant which was evaluated under the criteria of 10 CFR Part 100, Appendix A.

The peak accelerations for the SSE are 0.12 g horizontal and 0.08 g vertical. The response spectra are presented in Figure 2.5-8 of the TSSAR. Section 3.2.3 of the TSSAR presents the calculations that prove that a cask will not slide or tip over during the SSE. The concrete pad is designed -

to withstand an earthquake without breaking.

2.5 Protection against Fire and Explosions Pursuant to 10 CFR 72.122 (c), the licensee is required to provide protection against fires and i

explosions. In Section 2.2 of the TSSAR, an explosion of a barge loaded with 1270 t (1400 l

tons) of TNT, located in the middle of the channel of the Mississippi River, was evaluated and shown to produce an overpressure at the ISFSI that is much less than the design basis overpressure that was established as 170 kPa (25 psig).

)

In addition, an analysis was performed to evaluate the consequences of a fire resulting from the combustion of all the fuel that is carried by the vehicle that is used to tow the transporter from the fuel handling building to the ISFSI. The estimated fire duration was 10 minutes, and the flame conditions were assumed to be 800*C (1475'F), with an emissivity of 0.9, and the surface of the cask assumed to have an emissivity of 0.8. This fire constitutes an upper bound that is unlikely to be exceeded within the Prairie Island site. No temperature criteria would _be exceeded as a result of this fire, due to the large thermal inedia of the storage cask.

2.6 Confinement Barriers and Systems Pursuant to 10 CFR 72.122(h)(1), the licensee must protect the fuel cladding against degradation and gross ruptures. The TSSAR addresses the issue of fuel cladding degradation, in Section i

3.3.7.

The design criterion for the TN-40 cask requires that the maximum fuel cladding temperature criterion be calculated in accordance with PNL-6189 (Reference 4). The approach used by PNL is less stringent, but substantially in accord with the criterion adopted by the staff to ensure that degradation and gross rupture do not occur over the design life of the ISFSI. For 2-3

'l 10-year old PWR fuel, the limiting temperature was determined to be 340*C (644 *F) for 10-year cooled fuel and 335'C (635'F), for 15-year cooled fuel. A confirmatory analysis, to verify the validity of this limiting temperature, was performed by the staff, and the results reported in Section 3.3.5 of this SER. 10 CFR 72.122(h)(3), though specifically referring to ventilation and off-gas systems that could be associated with an ISFSI, is interpreted to apply to cask storage as a requirement to confine airborne radioactive particule.e materials ~during normal and off-normal conditions. Consequently, closures secured by bolts or other fasteners should be designed to limit leakage to levels that do not exceed the regulatory limits of 10 CFR 72.104 and 72.106. The TN-40 design features a single-closure lid incorporating two metallic O-ring seals.

The design criterion for each seal is a leakage rate not exceeding 104 cc/sec of helium. The staff considers the leakage rate to be acceptable for maintaining the cask helium atmosphere for storage periods of at least 20 years.

The design also provides capability to detect seal failure through pressure monitoring. If seal failure should occur, leak tightness can be restored by any of several means. The acceptability of the leak criterion, with respect to leakage of airborne radioactive particulate and gaseous materials, is addressed in Chapter 7 of this SER.

2.7 Instrumentation and Control Systems Pursuant to 10 CFR 72.122(i), the licensee must provide instrumentation and control systems that monitor systems important to safety over anticipated ranges for normal and off-normal operation. Instrumentation and control systems are not necessary for this type of storage system, because ofits passive nature. The TN-40 cask incorporates a pressure-monitoring system that serves as a cask tightness surveillance system. The design criteria and description of this system appear in Section 3.3.2 of the TSSAR. The staff finds this pressure-monitoring system acceptable for detecting off-normal cask leakage and for complying with the 10 CFR 72.122(i) requirement.

2.8 Criteria for Nuclear Criticality Safety 10 CFR 72.124 requires that spent fuel handling, transfer, and storage systems be designed to be maintained suberitical. The margins ofsafety 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. Section 3.3.4 of the TN-40 TSSAR addresses nuclear criticality safety criteria. Criticality analysis and prevention are reviewed in Chapter 6 of this SER.

The TSSAR establishes a maximum effective multiplication factor of 0.95, for all credible.

j configurations and environments, for the prevention of criticality. This factor is widely accepted as a criticality prevention criterion, and the staff concurs with its application to the TN-40 cask.

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2.9 Criteria for Radiological Protection 10 CFR 72.126 requires that the licensee provide adequate (a) protection systems for radiation exposure control, (b) radiological alarm systems, (c) systems for monitoring efnuents and direct radiation, and (d) effluent control systems, in a radiological protection program. Section 3.3.5 of the TSSAR addresses radiological protection. The detailed evaluation for compliance with the regulation is discussed in Chapters 5,7, and 10 of this SER.

The principal design features of the TN-40 cask for exposure control are the inherent shielding capability of the cask and the integrity of the seals at the closure joints. Radiological alarm systems and systems for monitoring effluents and direct radiation are not applicable to the design of the storage cask. Effluent: =e not a normal consequence of the passive dry storage operation; consequently, control systems to provide radiological protection for this condition are not applicable. Only provision (a) above is applicable to the cask, with respect to shielding capability, and the possibility of leakage from seals that may degrade or suffer damage as a result of an accident.

However, it should again be noted, as in Section 2.7 above, that the sealing system of the cask 4

uses a pressure-monitoring device as a leak surveillance system. Leakage past the outer metallic seal will be manifested by a drop in interseal system pressure.

The shielding capability of the cask, for gamma rays, relies primarily on the thickness and attenuation property of the steel cylinder and the steel closure lids that comprise the primary barriers to radiation.

The TSSAR, in Section 10.1.2 (" Criteria"), also establishes the surface dose criterion as.6 mSv/hr (60 mrem /hr). The staff believes that this criterion is acceptable (see Sections 5.2 and 5.4 of this SER). However, in finding these limits acceptable, the staff notes that, for site-specific analyses, consideration must be given to cumulative dose rate, because of reactor operations, and to individual residency time at or near the site boundary. The nearest individual has been conservatively assumed in this evaluation to be present continuously at the site boundary.

2.10 Criteria for Spent Fuel and Radioactive Waste Storage and Handling Pursuant to 10 CFR 72.128, the licensee is required to design the spent fuel storage and waste storage systems to ensure adequate safety under normal and accident conditions. These systems must be designed with: (a) a capability to test and monitor components important to safety, (b) suitable shielding for radiation protection, under normal and accident conditions, (c) confinement structures and systems, (d) a heat-removal capability having testability and reliability consistent with its importance to safety and (e) means to minimize the quantity of radioactive wastes generated.

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This section of the regulations defines the requirements for the spent fuel storage system within the context of the entire ISFSI. The TSSAR presents a summary that addresses only cladding temperawre criteria and nuclear criticality safety in Section 3.3.7. Actually, the entire TSSAR serves to demonstrate compliance with the details of this part of the regulations.

2.11 Criteria for Deconunissioning Pursuant to 10 CFR 72.130, the licensee is required to design the ISFSI for decommissioning.

For dry cask storage, this requirement applies to the cask design itself. Thus, decommissioning provisions should address decontamination of the cask components after removal of the radioactive spent fuel. The quantity of radioactive wastes produced and contamination of equipment should be minimized. The TSSAR addresses this requirement, in Section 3.3.7.2, in detail.

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3. STRUCTURAL EVALUATION 3.1 Area of Review This chapter evaluates the structural response of the TN-40 dry cask storage system to loadings under normal operating conditions, accident conditions, and loads due to environmental conditions and natural phenomena. The TN-40 cask is designed by Transnuclear, Inc., and will be used at the Prairie Island ISFSI which is operated by NSP.

The structural review summarized in this chapter addresses assumed loads and material properties, allowable stress and displacement criterion, and an evaluation of the structural analyses provided in the TSSAR for each of the components and systems important to safety. Safety requirements of the TN-40 cask.

are specified in 10 CFR Part 72.

3.2 Acceptance Criteria The structural integrity of the cask will be deemed adequate if it can be demonstrated that the stresses and displacements induced by the loads noted in Section 3.4.1 of this SER are lower j

than the allowable limits for the cask components important to safety. Allowable stress limits i

are documented in the Prairie Island ISFSI TSSAR, in Tables.4.2-4 through 4.2-6a, and Figure 4B.5-1. The structural design criteria for the basket are discussed in Section 4.2.3.3.3 and Section 4B.5 of the TSSAR. Information about tests performed on the basket panels is included in Section 4C.

10 CFR 72.122 does not provide explicit structural criteria except to satisfy the safety requirements of that section. To meet the safety concerns of confinement, shielding, and suberiticality, the structural components cannot have stresses or displacements that exceed allowable limits. For the cask components, except the basket, the TN-40 cask uses the provi-sions of Section III of the ASME Boiler and Pressure Vessel Code (BPV code).

The TSSAR lists in Tables 4.2-4 through 4.2-6a and Figure 4B.5-1 stress intensity limits for primary service loads and Levels A and D service loads, for the confinement vessel, confinement bolts, non-confinement structure, and basket components. In general, the stress intensity limits are in accordance with the standards established by the ASME BpV Code.

Consequently, they conform to the quality standard requirement of 10 CFR 72.122(a).

3.3 Review The TSSAR was reviewed for compliance with 10 CFR 72.122(a), which refers to quality standards that govern the characterization of materials, the establishment of stress intensity limits, and the design and analysis methods that provide confidence in the capability of the structure, system, or component to perform the required safety function.. The TSSAR was also reviewed for compliance with 10 CFR 72.122(b), which requires that the cask be designed to '

accommodate the effects of postulated accidents. Judgment of compliance is based either on the information provided in the TSSAR or results from confirmatory analyses.

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3.3.1 Cask Loads The TSSAR specifies, in Tables 3.2-3 and 4.2-3, the normal operating internal pressure of 689 kPa (100 psi) since it envelopes all internal pressure effects.

In Section 3.2.5.3.2, the trunnion loads are based on ANSI N14.6, which lequires that lifting

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devices be capable of lifting 6 times and 10 times the cask weight, without exceeding the yield and ultimate strengths of the material.

The design basis loads due to environmental conditions and natural phenomena are summarized in Section 3.2 of the TSSAR.

l 10 CFR 72.122(b)(1) requires that the cask be designed to accommodate the effects of postulated accidents. The review included calculations to establish an appropriate deceleration for a 46-cm (18-inch) drop onto an unyielding surface. Results from these calculations (using both a lumped.

parameter model and a detailed dynamic finite element model) indicate a deceleration of 350 g.

This deceleration value was used for confirmatory analyses performed during the review.

The review has not addressed any cask tip-over due to design changes to the ISFSI concrete pad.

and safety assurance calculations for the transport vehicle. The concrete pad has been classified as Seismic Category 1, per letters NSP to the Nuclear Regulatory Commission, dated December 20, 1991, and February 10,1992 (References 5 and 6). This prevents cask tip-over from seismic-induced pad dislocation. The safety assurance calculations for the transporter are documented in the letter from NSP, to NRC dated March 5,1992 (Reference 7). This analysis concludes that the cask will not tip over because of a failure of the cask transporter, based primarily on the fact that the cask's center of gravity is not lifted high enough to allow the cask to tip over. Any failure of the transporter, will allow the cask to fall to the underlying pavement and return to a vertical orientation.

3.3.2 Cask Materials Tables 4A.2-1 and 4A.2-2 of the TSSAR describe the mechanical properties used for the components of the cask body. Tables 48.1-1 through 4B.1-3 describe the mechanical properties l

of the basket materials. All materials are identified by ASME code designations that are related to ASTM Specifications. The reviewers consider these specifications to be quality standards, i

in accordance with 10 CFR 72.82(a).

l Tables 4A.2-1 and 4A.2-2 list the mechanical properties including the design stress parameter used for the components of the cask body. Tables 4B.1-1 through 4B.1-3 list the mechanical properties of the basket materials.

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3.3.3 Review of Cask Body Section 4A.3.3 of the TSSAR describes the analysis of the cask body for the following cases:

1.

Lid bolt preload corresponding to 172 MPa (25,000 psi) direct stress in the bolt shank.

2.

Internal pressure of 689 kPa (100 psig).

3.

External pressure of 172 kPa (25 psig).

4.

Lifting loads of 3 g vertical up.

5.

Gravity loads of 1 g down.

6.

Worst-temperature distribution in the cask body (off-normal condition).

7.

Bounding loads of 2 g down and 1 g lateral, with the cask standing in a vertical orientation on the pad.

The maximum stress intensities for each load case, shown in Tables 4 A.3.3-1 through 4A.3.3-7, are far below the allowable stress intensity limits. Section 4A.3.5 describes the method used -

for load combinations, and Tables 4A.3.5-1 through 4A.3.5-8 show the maximum stress intensities resulting from load combinations. The maximum combined strest intensities are less than the allowable stress intensity limits for each component.

The TSSAR states in Sections 3.2.1.2.1 and 3.2.1.2.2 that the cask will not tip over as a result of the design basis tornado wind loads and missile impacts. With the change in design of the-concrete pad to Seismic Category 1, the cask will not tip over as a result of dislocation of the pad because of an earthquake.

Sections 3.2.2, 3.2.4, and 3.2.5.2.8 state that cask integrity will be maintained for floodmg conditions, snow and ice loadings, and for lightning strikes.

Section 4A.3.3.2 addresses the subject of tornado missile impact involving two missiles: Missile A is a 1814-kg (4000-lb) automobile traveling at 80 km/h (50 mph) and Missile B is a 90 kg (200 lb) plank traveling at 483 km/h (300 mph). Missile A has a greater effect on cask stability than does Missile B because of its greater momentum. However, the peak impact conditions involve Missile B since it delivers greater force and applies it to a smaller area than does Missile A. For an impact by Missile B, local damage of the neutron shielding might occur, since the outer shell is relatively thin, but neither the massive gamma shielding forging nor the confinement vessel will be punctured. If Missile B were to strike the top of the cask, it could puncture the weather cover and neutron shield, but would not puncture the cask lid.

The TSSAR describes analyses of the cask body, for accident conditions, in Chapter 8. The impact conditions considered in the TSSAR are bottom end drop and tip-over. Section 8.2.8.2.1 indicates that a 50-g deceleration is used for the design basis for the bottom end drop. These decelerations assume impact'on a concrete storage pad and are based on the methodology of EPRI NP-4830. Transnuclear and NSP have insisted ~ that the TSSAR reflect the 50-g H

deceleration for the bottom end drop. Since the methodology of EPRI NP-4830 is not benchmarked, the review did not consider this to be a valid technique and thus focused upon an 3-3

unyielding surface. The predicted impact loads after impact on an unyielding surface were about 350 g, which proved to be acceptable during confirmatory calculations. Since both analyses in the TSSAR and confirmatory calculations are in agreement that the cask meets the acceptance criteria during the bottom end impact, it must be concluded that this cask design will withstand the effects of 46-cm (18-inch) free fall.

The results of the 46-cm (18-inch) bottom end drop were obtained by multiplying the results of the 1-g vertical load t,e described in Section 4A.3.3 and Table 4A.3.3-4. The results of the bottom end drop are shown in Figure 8.2-la, with stress results combined for bolt preload and internal pressure of 689 kPa (100 psi) in Figure 8.2-lb. The TSSAR states that the stress intensities are less than the confinement allowable limits and are acceptable.-

The review included confirmatory analyses using a 2-D axisymmetric finite element model subjected to a 350-g load. The maximum stress intensities for the confirmatory analp.s are less than the allowable limits.

3.3.4 Review of Basket Table 4.2-12 in the TSSAR summarizes the stresses in the basket and compares them to the allowable stresses. The highest Level A stress in the stainless steel boxes is less than the allowable limit of 3 S,.

The largest primary plus secondary stress in a weld is also less than the 3 S limit.

A buckling evaluation of the basket, when subjected to the 46-cm (18-inch) end drop (350-g load) was based on the theory of plates, and assumed that the 304 stainless steel supports the full weight of the basket. If the 304 stainless steel supports only its own weight, buckling is unlikely to occur. The critical load for the aluminum panels is higher than the critical load for the stainless steel tubes. According to the NSP letter dated August 12,1992, (Reference 8) the stainless steel can withstand a 160-g impact load, assuming the 304 stainless steel tubes support the full weight of the basket, whereas the aluminum panels can withstand a 250-g impact load, assuming that they support the full weight of the basket.

3.3.5 Review of the Neutron Shield, Trunnions, and Other Components l

The analyses of the neutron shield (outer shell) are provided in Section 4A.7 of the TSSAR.

The stresses acting on the outer shell and closure plates are summarized and compared with the allowable stresses in Table 4.2-15 of the TSSAR. The maximum stress intensity'in the outer shell is less than the allowable stress.

The TSSAR states, in Sections 3.2.5.4.3 and 4.2.3.4.5, that the trunnions were evaluated for vertical lifting reaction applied on the centers of the lifting shoulders for loads of 6 times and 10 times the maximum weight of a fully loaded cask. When the load is equal to 6 times the weight, the maximum tensile stresses are less than the yield strength of the trunnion material, as required by ANSI N 14.6. When the load is equal to 10 times the weight, the stresses are less than the ultimate strength, as required.

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No impact limiter is included in the TN-40 design. The review did not encompass the cask tip-over accident, because of the classification of the concrete pad as a seismic Category 1 structure, and the safety assurance calculations for the transport vehicle.

Section 4. A.4 of the TSSAR describes the lid bolt analyses. Section 4.2.3.4.3 and Table 4.2-11 of the TSSAR summarize the results. The highest design stress and the combined Level A and Level D stresses are well below the allowable values of 5, and 3 S., respectively.

3.3.6 '

Fuel Assemblies The integrity of the fuel rods under dry !torage conditions was evaluated with reference to the damage mechanisms that are likely to be effective. There are several potential mechanisms, for fuel cladding failure, that include fracture as the terminal event of stable or unstable crack propagation, stress corrosion cracking induced by fission products, hydriding, stress rupture because of creep, oxidation, and diffusion-controlled cavity growth. Since the cask is designed to maintain an inert gas (helium) environment for the fuel rods, oxidation is precluded and need not be considered further as a potential damage mechanism. The effects of the remaining damage mechanisnis were assessed based upon a review of the available empirical data and the conclusions of the researchers involved in cladding integrity (see Reference 10).

Three fundamental agents contribute to fuel cladding degradation under dry storage conditions:

stress, temperature, and an aggressive environment. Under normal conditions, the stress in the cladding is caused by internal gas pressure in the fuel rod. The major component of this gas j

is helium, which is introduced into the free fuel to moderate the effects of the external pressure, while in the reactor core. In the course of time, fission products accumulate in the fuel rod cavities. Besides contributing to the internal pressure of the fuel rod, the fission products may also attack the inner surface of the cladding. The effect of temperature manifests itself by accelerating the rate of degradation mechanisms activated by both stress and corrosion.

Stress corrosion cracking (SCC) occurs as a result of synergistic combination of a susceptible material, an aggressive environment, and high stress. The corrosive environment associated with SCC of fuel rods has been attributed to fission products generated during irradiation.

Although, the specific agent has not yet been identified, iodine, cesium, and cadmium are considered the most likely agents. SCC may also be related to pellet cladding interaction (PCI),

but this has only been observed during reactor operation because of, in part, the large external pressure on the fuel rods. The only known cause of cladding failure due to SCC occurred in a reactor during a ramp-up. No other failures from this cause are known to have occurred either -

during pool storage or under dry storage conditions. One explanation may be that the pellet temperatures during dry storage are much lower than those in_a reactor.. Consequently..the accumulation of fresh fission products at the cladding is slowly reduced during dry storage.-

Furthermore, the activation of SCC requires stress levels substantially above those that can reasonably be expected to prevail under dry storage conditions. The possibility exists, however, that cracks may be present that were initiated during reactor operation. Under these conditions, the stresses generated at the crack tips may be large enough to cause crack extension. However,-

should such a crack penetrate the cladding, it is likely that the internal pressure will be relieved 3-5

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-1 and, as a consequence, effectively terminate the progress of the SCC damage mechanism. The staff concludes, therefore, that SCC is not a damage mechanism that can lead to gross rupture of the fuel rod cladding.

Hydrides in zircaloy have been know to cause cracking by embrittling the cladding. Terminal solubilities of hydrogen in zircaloy increase with temperature. If the temperature subsequently decreases, hydrides will precipitate in an orientation determined by the stress level. Normally the hydride precipitates in a circumferential direction and is not a problem even at hydrogen concentrations up to 400 ppm. At hoop stress levels of 90 to 95 MPa, (13 to 13.8 ksi) the hydride will precipitate in a radial direction that can encourage crack penetration. At 400^C (725'F), the hydrogen concentration could be as high as 200 ppm. Brittleness may be induced as the fuel rods decrease in temperature during dry storage. However, the hoop stresses in the cladding are not expected to be high enough to cause a radial orientation of the hydride and consequent crack initiation. It is remotely possible that pre-existing cracks under stress can i

induce the diffusion of hydrogen to the crack tips, where substantially higher concentrations could precipitate hydride in a manner that would encourage crack extension. However, as is the l

case of SCC, crack penetration would result in a loss of fuel rod internal pressure and termination of the damage mechanism.

The staff concludes, therefore, that the delayed hydriding is not a damage mechanism that can lead to gross rupture of the fuel rod cladding.

Creep rupture is a potential failure mode under dry storage conditions. Researchers have I

demonstrated that using a Larson-Miller approach, temperatures from 380 C (716 F) to 400 C (725"F) could be tolerated for creep rupture lives well beyond that required for interim storage of spent fuel. The Larson-Miller approach, however, is somewhat empirical, since it depends on the existence of experimental data, to establish the appropriate parameter. Practicality limits the duration of creep rupture tests, which are usually conducted at stress levels and temperatures far higher than those that prevail under dry storage conditions. The creep damage mechanisms in the high-temperature, high-stress regime are different from those that occur at lower I

temperatures and stresses.

Consequently, predictions based on a Larson-Miller mode are clouded with sufficient uncertainty to warrant a more fundamental approach to cladding degradation under creep conditions.

The staff examined this matter to determine potential mechanisms for significant creep damage under dry storage conditions. The only mechanism for any of the failure modes considered above, that the staff found represented a potential for cladding degradation and gross rupture, was diffusion controlled cavity growth (DCCG), which is most applicable to the conditions of dry storage. Damage is manifested by the nucleation and growth of cavities at the grain boundaries which, in effect, reduces the area of material available to resist loads. The measure of damage is the fraction of the grain boundary area that undergoes decohesion. The reviewers

}

developed a method to determine the level of damage as a function of time (see Reference 9).

A confirmatory thermal analysis revealed that the maximum temperature for the fuel rods could be as high as 340"C (644 F), with a temperature decay curve representing fuel assemblies with 10 years cooling time, after irradiation for three cycles. Assuming a maximum internal gas pressure of 14.8 MPa (2110 psia) at 317 C (603 F) as specified in Section 3.3.7.1 of the 3-6

TSSAR, for the limiting fuel assembly type (Exxon) and computing the pressure at the temperature of 340"C (644*F), the progress of damage, based on the DCCG analysis indicated the area of decohesion after 20 years of storage would be less than 15 percent. Consequently, an initial fuel rod temperature not exceeding the design criterion of 340"C (644*F) is acceptabie for meeting the requirements of 10 CFR 72.122(h).

3.4 Findings and Conclusions The maximum stresses in the confinement vessel specified in the TSSAR are less than the allowable stress under normal storage conditions. The confirmatory analysis supported these values as reasonable while remaining conservative. The stresses that result from the hypothetical impact and fire conditions are less than the allowable stresses as indicated in the TSSAR and supported in the confirmatory analyses during the review.

The tip-over accident has not been included in the review, because NSP elected to design the storage pad as a seismic Category 1 structure and to provide an analysis of the cask transporter that demonstrates that failure of the transporter will not cause the cask to tip over. These arguments preclude the possibility of a cask tip-over accident and apply, at present, only to this site.

The impact loads indicated in the TSSAR for the 46 cm (18-inch) end drop on a concrete pad i

of 50 g is based on the methodology of EPRI NP-4830, which has not been benchmarked for

-l short drops, because of the absence of empirical data. The review has been based on an impact on an unyielding surface as a bounding condition. Since the results piesented in the TSSAR and the results of the confirmatory calculations both indicate no stresses exceed the allowable stresses, it must be concluded that the design is adequate to protect the health and safety of the public and workers in the vicinity of this spent fuel cask.

The confirmatory analysis of the fuel rods indicates that the peak clad temperatures will remain less than the values that are indicated by the review to be required for significant loss of strength becuase of DCCG.

The confirmatory analysis indicates that the cask design is structurally adequate to meet the requirements of 10 CFR 72.122.

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4 TIIERMAL EVALUATION 4.1 Area of Review This section discusses the thermal design and evaluation of the TN-40 spent fuel storage cask.

The significant thermal design criteria and operating characteristics of the dry storage cask are identined and reviewed for accuracy to provide their intended safety functions. The thermal loads that the cask must endure are also reviewed. Thermal specUications of components must be described in the TSSAR and reviewed to ensure that all components can perform their function at the predicted temperatures. These reviews are performed for normal conditions of storage and for hypothetical accident conditions.

4.2 Acceptance Criteria 10 CFR 72.122(b) requires that structures, systems, and components important to safety must be designed to accommodate the effects of, and to be compatible with, site characteristics and environmental conditions associated with normal operation, maintenance, and testing of the ISFSI and to withstand postulated accidents.

10 CFR 72.122(c) requires that structures, systems, and components important to safety must be designed and located so that they can continue to perform their safety functions effectively under credible fire and explosion exposure conditions. Noncombustible and heat-resistant materials must be used, wherever practical, throughout the ISFSI, particularly in locations vital to the control of radioactive materials and to the maintenance of safety control functions.

Explosion and fire detection, alarm, and suppression systems shall be designed and provided with sufficient capacity and capability to minimize the adverse effects of fires and explosions on i

structures, systems. and components important to safety. The design of the ISFSI must include provisions to protect against adverse effects that might result from either the operation or the failure of the fire suppression system.

10 CFR 72.122(h) requires that the spent fuel cladding must be protected, during storage, I

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 with respect to its removal from storage. This may be accomplished by canning of consolidated fuel rods or unconsolidated assemblies or other means, as appropriate.

10 CFR 72.128(a) requires that spent fuel storage, high-level radioactive waste storage, and other systems that might contain or handle radioactive materials associated with spent fuel or high-level radioactive waste, 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.

The thermal design will be deemed acceptable ifit is demonstrated that the maximum cladding temperature does not exceed 340*C (644*F), during normal storage conditions, and 566*C 4-1

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1 (1050*F), during the hypothetical fire accident. These thermal criteria are addressed in Section 3 and ensure acceptable degradation of cladding integrity over the design storage life of 20 years, or during the hypothetical accident.

4.3 Review The review is divided into three principal parts: (1) material properties and component limitations, (2) normal conditions of storage, and (3) hypothetical accident conditions. These will be addressed in the following sections.

4.3.1 Material Properties and Component Limitations Section 3.3.2.2 of the TSSAR describes the thermal model that was used to evaluate the ability of the TN-40 cask to transfer the heat generated by the spent fuel assemblies to the environment.

The thermal properties of the materials are presented in Table 3.3-2. The properties of aluminum in the basket have been reduced to incorporate the effects of the boral plate that is sandwiched between the two aluminum plates. The properties have been reviewed and the staff concurs.

The manufacturer of the metallic O-rings indicates a maximum temperature of 299"C (570*F) for long-term service while maintaining a seal. Similarly, the maximum service temperature of 1

the neutron shield material is 149 C (300*F) without any degradation of structure of the material.

These temperatures are compared with the maximum predicted temperatures within the cask, in the following table, demonstrating a large margin between the predicted and allowable temperatures.

Location Limiting Temp. (*C)

Predicted Temp. ( C) 0-Ring Seals 299 (570*F) 117 (243*F)

Neutron Shield 149 (300*F) 134 (233*F)

Confinement Vessel 151 (304*F) 4.3.2 Normal Conditions of Storage The thermal analysis in the TSSAR was reviewed and one-dimensional confirmatory calculations were performed to ensure that the fuel rod cladding temperature does not exceed that specified in the TSSAR. The steady-state thermal analysis in the TSSAR was performed with the finite element computer program ANSYS, assuming a solar absorptivity of 0.3 for the cask surface.

In this confirmatory analysis, a power peaking factor of 1.1 was assumed in the hottest region of the cask basket, a time-averaged value ofinsolation was used, and a modified Wooton Epstein Correlation was used to calculate the maximum temperature of the fuel. cladding. The 4-2

4 con 6rmatory calculations were in agreement with the results presented in the TSSAR.

4.3.3 Accident Conditions Section 3.3.6 (" Fire and Explosion Protection") of the TSSAR states that no hydrocarbon fuel will be stored at the ISFSI. A transformer will contain about 380 L (100 gallons) of non-flammable dielectric fluid. The pressure wave created by explosions outside the ISFSI will be deflected by the earthen berm that partially surrounds the ISFSI, resulting in less overpressure than a tornado.

The TSSAR includes a calculation to show that an external pressure of 716 kPa (104 psia) is required for the initiation of the yielding of the cask outside surface. It is concluded that the TN-40 cask is structurally adequate to withstand any credible explosive o'.erpressure.

The thermal analysis for an accidental fire was reviewed in the TSSAR to determine if any radioactive release could occur in violation of 10 CFR 72.106. The TSSAR assumed that the cask is exposed to a 800*C (1472 F) engul6ng fire, for 10 minutes. The confirmatory analysis of the fire accident was based upon an analytic model of the cask that assumed infinite thermal conductivity. The short duration of the fire resulted in cask temperature increases of less than 10 C (50"F) which will not cause damage to the fuel rod cladding.

Section 3.3.2.2.2 defines the assumptions employed for the analysis of a buried cask. The results of the analysis indicate that recovery must be effected before five hours elapses, or the neutron shield will begin to outgas because its temperature will exceed 149*C (30&F). If recovery is not effected within about 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />, the seals will reach a temperature of 299'C (570'F). At this temperature, the manufacturer of the seals does not recommend long-term usage, but the seals will probably continue to function for short periods of time (hours). The staff review of these analyses indicates that it is conservative in both the initial conditions and assumptions during the burial period; however, it serves as a lower bound on the recovery times for this type of accident.

4.4 Findings and Conclusions The maximum cladding temperature specified in the TSSAR in Table 3.3-1 is 336 C (547'F) under normal conditions. The confirmatory analysis performed by the reviewers supported this'-

value as reasonable while remaining conservative. Since the modified Wooton-Epstein correlation is a conservative method for estimating the temperature rise across a fuel bundle, the fuel cladding will remain below 340aC (644*F) during storage, thus preventing cladding degradation, which leads to gross ruptures, and, therefore, the requirements of 10 CFR 72.122(h) are satisfied.

The maximum clad temperature during any operations associated with this cask will occur during the drying period, when a vacuum is present in the cask cavity. The assumptions employed in both the thermal analysis in the TSSAR and the confirmatory calculations are consistent with the '

4-3

vacuum drying operation, so there is'no possibility of the fuel rod temperatures exceeding the thermal criterion during any normal operation of this cask. Backfilling the cask with helium will cause a slight reduction in the peak fuel rod temperatures and restore the conservatism in the analyses presented in the TSSAR and the confirmatory calculations, i

The confirmatory analysis for the fire indicated a small increase, in the cask surface temperature, that would not cause the seals or peak fuel rod temperatures to exceed their respective design criterion. During a fire, the neutron shield material can be expected to offgas; however, this will i

not cause any significant degradation in the ability of the cask to perform its function of

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protecting the health and safety of the public. The cask design is structurally adequate to meet the requirements of 10 CFR 72.122(c) and 72.122(h).

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SIIIELDING EVALUATION 5.1 Area of Review This section of the review addresses the shielding design analyses for normal and sccident conditions of storage. The description of the contents and the gamma and neutron source terms used in the analyses are reviewed. The model used in the shielding evaluation is reviewed. This review includes an assessment of the degree to which the model reflects the geometry, dimensions, and material properties of the spent fuel storage cask during. normal storage and accident conditions. The method used to determine the gamma and neutron dose rates at selected points outside the cask for normal and accident conditions of storage is reviewed. This review includes consideration of the following: (a) the assumed spatial source distribution, (b) the computer programs used, (c) validity of the basic ;uput parameters, and (d) flux-to-dose rate conversion factors as a function of energy.

5.2 Acceptance Criteria 10 CFR 72.104(a) requires that during normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area shall not exceed

.25 mSv (25 mrem) to the whole body,.75 mSv (75 mrem) to the thyroid, and.25 mSv (25 mrem) to any other organ as a result of exposure to: (1) planned discharges of radioactive materials (radon and its daughters excepted), to the general environment; (2) direct radiation from ISFSI operations; and (3) any other radiation from uranium fuel cycle operations within the region.

10 CFR 72.106(a) requires that for each ISFSI site, a controlled area must be established.

10 CFR 72.106(b) requires that the minimum distance from the spent fuel storage facilities to the nearest boundary of the controlled area shall be at least 100 meters (328 feet). Our review focuses on the annual dose equivalent, from an array of 48 casks, to any individual who is located beyond the controlled area from direct and scattered radiation, because of normal operations. In the context of this review, exposure of the individual is continuous and the point beyond the controlled area where the individual is located is defined as the nearest site boundary.

From the TSSAR, the nearest site boundary is established as 180 meters (591 feet) due west of the center of the ISFSI; 110 meters (361 feet) due west of the westernmost cask within the ISFSI; the location of the nearest resident is established as 732 meters (2400 feet) northwest of the center of the ISFSI.

Cask shielding is deemed acceptable if it can be shown that the annual dose equivalent, due to normal operations, from an array of 48 casks, to any individual located beyond the controlled-area, is not in excess of the applicable criterion given in Section 5.2 above.

5-1

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5.3 Review The review is divided into three parts: (1) source specincation (gamma and neutron); (2) model specification; and (3) shielding evaluation.

5.3.1 Source Specification 5.3.1.1 Gamma Source Gamma source activities (MBq) and strengths (photons /sec) for the TN-40 cask are addressed in Sections 3.1.1 (" Spent Fuel To Be Stored"), 7.2.1 (" Characterization of Sources"), 7A.2 (Shielding Analyses), and 7A.4 (Direct Radiation E-W) of the TSSAR. NSP also provided supplemental information in Revision 2, dated September 1991, and its response to questions, dated December 23,1991 (References 10 and 11).

Section 3.1.1 (" Spent Fuel To Be Stored") summarizes the design basis fuel parameters, irradiation conditions, cooling time, and gamma source strength (photons /sec) for the active fuel length of a spent fuel assembly as a function of time after discharge from the reactor. The-gamma source strength is determined from an ORIGEN2 calculation, using the Westinghouse 14x14 array OFA design basis fuel described in Tables 3.1-1 (" Fuel Assembly Parameters") and 3.1-2 (" Thermal, Gamma and Neutron Sources for the Design Basis 14x14 Westinghouse OFA Fuel Assembly") of the TSSAR. In this calculation, the average burnup is 45,000 MWD /MTU; the specific power is 37.5 MW/MTU; the initial fuel enrichment is 3.85 percent; there are 380 kg (838 lb) per assembly of heavy metal; and the minimum time after discharge is 10 years. The irradiation time and interim shutdown periods are not specified. Contributions to the gamma source strength, from activation of the structural materials in the active fuel region, are also not addressed.

Section 7.2.1 (" Characterization of Sources") summarizes again the irradiation conditions for the ORIGEN2 calculation. In addition, the irradiation time and interim shutdown periods are addressed. A three-cycle operating history, with interim shutdown of 30 days, is specified.

Contributions to the gamma source strength from activation of the structural materials in the fuel assembly are also addressed. The specific activities (Ci/MTU) for the major fission products present in the active fuel length of a spent fuel assembly, and the major activation products in the structural materials in the active fuel, upper and lower end fitting, and gas plenum regions, as functions of time after discharge from the reactor, are summarized in Tables 7.2-3 (" Fission Product Activities (Ci/MTU), Westinghouse OFA 14x14 3.85 w/o 2"U, 45,000 MWD /MTU, l

10 Year Cooling Time") and 7.2-3A (Activation Activities (Ci/MTU)) of the TSSAR",

respectively. The specific activities (Bq/MTU) for the activation products in the structural materials are determined from an' ORIGEN2 calculation, using the spatially and spectrally adjusted neutron flux generated by ORIGEN2 for the active fuel region. The assembly gamma source strengths (photons /sec/ assembly) for the active fuel, upper and lower end fitting, and gas plenum regions at 10 years after discharge from the reactor are summarized in Table 7.2-2 (Gamma and Neutron Radiation Sources, Westinghouse OFA 14x14 3.85 w/o "U, 45,000 2

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I MWD /MTU,10 Year Cooling Time) of the TSSAR. The assembly gamma source strengths associated with the upper end fittings and lower end fittings are apportioned at 52.4 percent and 47.5 percent of the total assembly gamma source strength, respectively. ORIGEN210-group gamma source spectra for the TN-40 cask active fuel, upper and lower end fitting, and gas plenum regions are summarized in Table 7.2-4 (Primary Gamma Source Spectrum, ORIGEN2 Group Structure, Westinghouse OFA 14x14 3.85 w/o n'U, 45,000 MWD /MTU,10. Year Cooling Time) of the TSSAR. In addition, the ORIGEN2 gamma source spectra for the active.

fuel region is converted to the SCALE 18-group structure via simple apportioning of overlapping groups and a weighting ratio of the ORIGEN2 and SCA.LE library average group energies. The SCALE 18-group structure gamma source spectra is summarized in Table 7.2-6 (" Parameters.

for the SCALE 27N-18G Library, Westinghouse OFA 14x14 3.85 w/o SU, 45,000 MWD /MTU,10 Year Cooling Time") of the TSSAR.

Section 7A.2 (" Shielding Analyses") describes the use of a uniform axial distribution for the active fuel source region and the generation of gamma source spectra as a function of time after discharge from the reactor. The specific period ofinterest for the' gamma source spectra is 10 to 30 years after discharge from the reactor; the time increment is 1 year.

Section 7A.4 (" Direct Radiation E-W") describes the calculation of an equivalent gamma point source for the evaluation of the indirect radiation resulting from air scattering of the direct radiation component. The bases for equivalent gamma point source terms are the source terms in Table 7.2-6 (" Parameters for the SCALE 27N-18G Library, Westinghouse OFA 14x14 3.85 i

w/o *U, 45,000 MWD /MTU,10 Year Cooling Time") of the TSSAR.

The staff gamma source strengths used for the active fuel, upper and lower end fitting, and gas plenum regions in the NSP TN-40 shielding evaluation are the gamma' source strengths and spectra in Table 7.2-4 (" Primary Gamma Source Spectrum, ORIGEN2 Group Structure, Westinghouse OFA 14x14 3.85 w/o "5U, 45,000 MWD /MTU,10 Year Cooling Time") of Revision 2 to the TSSAR, dated September 1991.

The staff has assumed a uniform axial distribution for the active fuel region gamma source strength throughout the shielding evaluation.

5.3.1.2 Neutron Source Neutron source activities (MBq) and strengths (neutrons /sec) for the TN-40 cask are addressed

'i in Sections 3.1.1 (" Spent Fuel to be Stored"), 7.2.1 (" Characterization of Sources"), 7A.2

(" Shielding Analyses"), and 7A.4 (" Direct Radiation E-W") of the TSSAR. NSP has also provided supplemental information in Revision 2, dated September 1991, and its response to questions dated December 23,1991.

Section 3.1.1 (" Spent Fuel To Be Stored") summarizes the design basis fuel parameters, irradiation conditions, cooling time, and neutron source strength (neutror s/sec) for the active fuel length of a spent fuel assembly, as a function of time after discharge from the reactor. The 5-3

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.. l neutron source strength is determined from an ORIGEN2 calculation, using the Westinghouse 14x14 array OFA design basis fuel described in Tables 3.1-1 (" Fuel Assembly Parameters") and 3.1-2 (" Thermal, Gamma and Neutron Sources for the Design Basis 14x14 Westinghouse OFA Fuel Assembly") of the TSSAR. The major input parameters for this calculation, are described in Section 5.3.1.1 (" Gamma Sources") of this SER.

l Section 7.2.1 (" Characterization of Sources") summarizes again the irradiation conditions for the ORIGEN2 calculation. In addition, the irradiation time and interim shutdown periods are addressed. A three-cycle operating history with interim shutdown of 30 days is specified. The specific activities for the major actinides present in the active fuel length of a spent fuel assembly as a function of time after discharge from ti.e reactor are summarized in Table 7.2-3

(" Fission Product Activities (Ci/MTU), Westinghouse OFA 14x14 3.85 w/o "U, 45,000 2

MWD /MTU,10 Year Cooling Time") of the TSSAR. The assembly neutron source strength (neutrons /sec/ assembly) for the active fuel region at 10 years after discharge from the reactor is summarized in Table 7.2-2 (" Gamma and Neutron Radiation Sources, Westinghouse OFA 14x14 3.85 w/o "U,45,000 MWD /MTU,10 Year Cooling Time") of the TSSAR. The neutron 2

source consists primarily of spontaneous fission of 2"Cm and (a.n) reactions from the alpha particles of 2"Cm and the oxygen in the fuel. The neutron source' strength associated with spontaneous fission and (a,n) reactions is apportioned at 93.7 percent and 6.3 percent, respectively. A SCALE 7-group neutron source spectrum for the TN-40 cask active fuel region is summarized in Table 7.2-5 (" Neutron Source Distribution, Westinghouse OFA 14x14 3.85 w/o 2"U,45,000 MWD /MTU,10 Year Cooling Time") of the TSSAR.

F Section 7A.2 (" Shielding Analyses") describes the use of a uniform axial distribution for the active fuel source region and the generation of neutron source spectra as a function of time after discharge from the reactor. The specific period of interest for the neutron source spectra is 10 to 30 years after discharge from the reactor; the time increment is 1 year.

Section 7A.4 (" Direct Radiation E-W") describes the calculation of an equivalent neutron point source for the evaluation ci the indirect radiation resulting from air scattering of the direct radiation component. The basis for equivalent neutron point source term is the source term in Table 7.2-5 (" Neutron Source Distribution, Westinghouse OFA 14x14 3.85 w/o "U,45,000 2

MWD /MTU,10 Year Cooling Time") of the TSSAR.

The staff neutron source strength used for the active fuel region in the NSP TN-40 shielding evaluation is that of Revision 2 to the TSSAR, dated September 1991. The staff neutron source spectrum is independently determined from the information provided in the TSSAR and yields a finer group structure than that employed by NSP. The Bucholz (TSSAR Reference 7) 2"Cm normalized spontaneous fission and (a,n) reaction neutron spectra are used. The "Cm (a,n) 2 reaction normalized neutron spectrum is converted to the group structure (34 groups) of the 2"Cm spontaneous fission normalized neutron spectrum, using simple linear apportioning. A combined normalized neutron spectrum is then created, by adding the two in the proportions of I

the TSSAR neutron source fractionation results: 93.7 percent spontaneous fission and 6.3 percent (a,n) reaction. The staff neutron source spectrum is the product of the combined normalized 5-4 1

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neutron spectrum, the assembly neutron source strength in Table 7.2-2 (" Gamma and Neutron Radiation Sources, Westinghouse OFA 14x14 3.85 w/o 2"U, 45,000 MWD /MTU,10 Year l

i Cooling Time") of Revision 2 to the TSSAR, dated September 1991, and the number of fuel assemblies.

The staff has assumed a uniform axial distribution for the active fuel region neutron source strength throughout the shielding evaluation.

5.3.2 Model Specification The review is divided into two parts: (1) description of the radial and axial shielding configuration and (2) shield regional densities.

The radial and axial shielding configurations for the TN-40 cask are addressed in Sections 2.3.4.2 (" Calculations"), 2.5.1.7 (" Plan and Profile Drawings"), 5.1.2 (" Flow Sheets"), 7.4

(" Estimated Onsite Collective Dose Assessment"), 7A.1 (" Shielding Design Features"), 7A.2

(" Shielding Analyses"),7A.3 (" Direct Radiation N-S"), and 7A.4 (" Direct Radiation E-W") of the TSSAR. NSP has also provided supplemental information in Revision 2, dated September 1991, and its response to questions dated June 5,1991 (Reference 12), and December 23,1991.

1 Section 2.3.4.2 (" Calculations") provides the dose point distance from the ISFSI to the nearest site boundary in each of 16 directional sectors. The distances are measured from the center of the ISFSI to the nearest point on the site boundary, within a 45* sector centered on the compass direction of interest and are summarized in Table 2.3-1 (" Site Boundary Dispersion Factor (X/Q)") of the TSSAR.

Section 2.5.1.7 (" Plan and Profile Drawings") provides drawings of the site grading plan, with sections and details. Included is a layout of the proposed 48-cask array and a cross section of the earthen berm.

Section 5.1.2 (" Flow Sheets") provides the dose point distances associated with various cask-handling operations. The distances are measured from the surface of the cask to the individual and are summarized in Table 5.1-2 (" Anticipated Time and Personnel Requirements for Cask Handling Operations") of the TSSAR.

Section 7.4 (" Estimated Onsite Collective Dose Assessment") provides the dose point distances from the ISFSI to various onsite locations where personnel may be found. The distances are

+

measured from the southeast corner of the eastern row of casks to the onsite locat%n in question and are summarized in Table 7.4-4 (" Dose Rates at Onsite Locations Due to Cask Storage") of -

the TSSAR.

Section 7A.1 (" Shielding Design Features') illustrates and tabulates the shielding dimensions of the TN-40 cask. Included are the overall thicknesses of the steel cask body wall 24.13 cm (9.5 in.) bottom, 26.04 cm (10.25 in.); and lid, 26.67 cm (10.5 in.); the radial neutron shield, 3

5-5 s

9 I1.43 cm (4.5 in.) and steel shell,1.27 cm (.5 in.) the polypropylene drum 10.16 cm (4 in.) and steel shell 1.27 cm (.5 in.); and the steel torispherical protective cover, 0.95 cm (.37 in.).

Aluminum boxes (0.305 mm-thick) (0.012 in.), used to contain the resin, are included in the overall radial neutron shield thickness.

Section 7A.2 (" Shielding Analyses") describes and illustrates the gamma and neutron shielding models; an assumed loading sequence for the west (first) and east (second) pads of the ISFSI; and the model used to determine relative gamma and neutron dose equivalent rate factors as a function of time after discharge from the reactor, normalized to 10 years. The gamma source geometry is a homogenized right circular cylinder with a height less than that of a fuel assembly (400.61 cm) (157.7 in.) and a cross-sectional area equal to that of the cask cavity (26,268 cm)

Axially, it is divided into five regions: upper end fittings and basket (8.84 cm) (3.48 in.);

plenum and basket (18.19 cm) (7.16 in.); active fuel and basket (365.76 cm) (144 in.); void (1.27 cm) (.5 in.); and lower end 6ttings and basket (6.55 cm) (2.58 in.). With the exception of the radial neutron shield, the steel-encased polypropylene drum, the torispherical protective cover, the lid vent and drain port penetrations, and trunnions, all shield configurations

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correspond to design geometries and/or dimensions. The resin and aluminum boxes in the radial neutron shield and the steel encased polypropylene drum are each homogenized into single compositions via their respective volume fractions.

The torispherical protective cover is modeled as a right circular cylindrical shell instead of a dish. The lid vent and drain port penetrations and trunnions are not modeled. The radial neutron and capture gamma source geometry is a homogenized cylindrical one-dimensional model of the active fuel and basket with 2

a cross-sectional area equal to that of the cask cavity (26,268 cm )(4072 in.). Shield regions are cylindrical shells corresponding to actual dimensions. The resin and aluminum boxes in the radial neutron shield are homogenized into a single composition via their respective volume fractions. The axial neutron and capture gamma source geometries are homogenized one-dimensional plane geometry models. For the top end model, the axial regions are: active fuel and basket (182.88 cm) (72 in.); plenum and basket and upper end fittings and basket (27.03 cm)

(10.64 in.); steel lid (26.67 cm) (10.5 in.); polypropylene drum and steel shell (11,43 cm) (4.5 in.), and steel torispherical protective cover (0.95 cm) (.37 in.). All void dimensions are neglected. For the bottom end model, the axial regions are: active fuel and basket (182.88 cm)

(72 in.); lower end fittings and basket (6.55 cm) (2.58 in.), and steel bottom (26.03 cm) (10.25 in.): Again, void dimensions are neglected. The model used to determine the direct radiation at long distances from the cask is spherical. The source region is a sphere of radius such that the volume of the sphere is equivalent to the volume of the cask cavity. Shield regions are spherical shells corresponding to actual dimensions. The west pad (Grst) loading sequence for the ISFSI is assumed to be eight casks to the middle of the pad at the start of the Drst year and two casks each year thereafter, alternating ends. The east pad (second) loading sequence of the ISFSI is assumed to be two casks to the middle of the pad at the start of the 10th year and two casks each-year thereafter, alternating ends. The entire two-pad loading sequence is assumed to require 21 years to complete. The model used to determine relative gamma and neutron dose equivalent rate factors as a function of time after discharge from the reactor, normalized to 10 years, employs the long distance spherical model described above.

5-6 4

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Section 7A.3 (" Direct Radiation N-S") describes the model used to determine the direct radiation at the centerline of the west and east pads in the north-south (N-S) direction from an array of 24 and 48 casks at 9 and 21 years, respectively. The direct radiation contribution from each front-row cask is determined at each dose point location using the long distance spherical model described in Section 7A.2 (" Shielding Analyses") of the TSSAR. The direct radiation contribution from each back-row cask is derived at each dose point from point kernel shielding analyses of a cask array in which the front row of casks is modeled as uranium cylinders during evaluation of each back-row cask contribution.

Section 7A.4 (" Direct Radiation E-W") describes the model used to determine: (1) the' direct radiation at the centerline of the west and east pads in the east-west (E-W) direction from an array of 24 and 48 casks at 9 and 21 years, respectively; (2) the indirect radiation resulting from i

air scattering of the direct radiation component; and (3) the attenuation of the earthen berm as a function of thickness. The direct radiation contribution from each end row cask is determined

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at each dose point location using the long distance spherical model described in Section 7A.2

(" Shielding Analyses") of the TSSAR. The direct radiation coniribution from each subsequent cask in each row is derived at each dose point from point kernel shielding analyses of a cask array in which the preceding casks in each row are modeled as uranium cylinders during

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evaluation of the subsequent cask in each row's contribution. The model used to determine the indirect radiation resulting from air scattering of the direct radiation component from each cask j

at each dose point location employs an equivalent point source representation of each cask. The equivalent points sources are determined using the long-distance spherical model described in Section 7A.2 (" Shielding Analyses") of the TSSAR. The model used to determine the attenuation of the earthen berm as a function of thickness employs the long distance spherical model described in Section 7A.2 (" Shielding Analyses") of the TSSAR. It is surrounded by various thicknesses of earth with the dose point distance set at 100 meters (109 yds).

i NSP dose point locations for cask surface,1-meter, and 2-meter distances are located on the axial centerline and at the midplane of the active fuel region.

The staff radial and axial shielding configurations used in the TN-40 cask shielding evaluation are independently determined. The source geometry is a homogenized right circular cylinder with a height equal to that of the fuel assembly (405.66 cm) (160 in.) depicted in Figure 3.1-4 i

("OFA Fuel Assembly Dimensional Data") of the TSSAR and a cross-sectional area equal to that

.l 2

2 of the cask cavity (26,268 cm ) (4072 in.). Axially, it is divided into six regions: upper end fittings, basket, and rails (8.84 cm) (3.5 in.); void and end plugs (3.99 cm) (1.57 in.); plenum springs, fuel cladding, guide tubes, basket, and rails (18.19 cm) (7.16 in); active fuel, fuel cladding, guide tubes, grid spacers and springs, basket, and rails (365.76 cm) (144 in); void and end plugs (2.34 cm) (.92 in.); and lower end fittings, basket, and rails (6.55 cm) (2.58 in.).

With the exception of the radial neutron shield and trunnions, all shield configurations correspond to design geometries and/or dimensions. The resin and aluminum boxes in the radial neutron shield are homogenized into a single composition via their respective volume fractions.

The trunnions are not modeled.

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The staff radial and axial shielding confbcation used in the TN-40 cask shielding evaluation as a function of distance is limited. The radial and axial shielding configuration is as described above, with the exception that the cask is surrounded by air to a height of 3 meters (3 yds) above the cask torispherical protective cover and a radius of ? meters (3 yds) from the radial neutron shield shell.

Material densities are addrcssed in Sections 3.3.4.1 (" Control Methods for Prevention of Criticality"), 7A.1 (" Shielding Design Features"), 7A.2 (" Shielding Analyses"), and 7A.4

(" Direct Radiation E-W") of the TSSAR. Material quantities and elemental composition

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(constituerts and/or weight percent) are addressed in Sections 4.6 (Decommissioning Plan) and 7.2.1 (Characterization of Sources). Atom number densities (atoms / barn-cm) are addressed in Section 7A.2 (" Shielding Analyses").

NSP has also provided Supplemental information Revision 2, dated September 1991, and in its response to questions dated December 23,1991.

Section 3.3.4.1 (" Control Methods for Prevention of Criticality") provides the material densities and major elemental constituents of the fuel, fuel rods, basket, cask body, and lid materials in Table 3.3-3 ("Matenal Composition for KENO Model (TN-40 Cask)") of the TSSAR.

i Section 4.6 (" Decommissioning Plan") provides the material mass and major elemental

'j composition (weight percent) of the basket and supports, cask body, lid, radial neutron shield and shell, and torispherical protective cover in Table 4.6-1 (" Data for TN-40 Activation Analysis") of the TSSAR.

d Section 7.2.1 ("Chametei.ation of Sources") provides the structural material mass of the active fuel, gas plenum, and upper and lower end fittings regions in Table 7.2-1 (" Material Distribution in Westinghouse 14x14 OFA Fuel Assembly") of the TSSAR.

Section 7A.1 (" Shielding Design Features") provides the material densities of the cask component ; in Table 7A-1 ("TN-40 Cask Shield Materials").

Section 7A.2 (" Shielding Analyses") provides the material densities and atom number densities (atoms / barn-cm) used in the TN-40 shielding analyses. Material densities used in the gamma source shielding analyses are summarized in Table 7A-2 (" Materials Input for QAD Model (TN-40 Cask)") for the homogenized upper end fitting and basket, plenum and basket, fuel and basket, and lower end fitting and basket source regions; the cask body; the torispherical protective cover; the radial neutron shield; the homogenized radial neutron shield; and the homogenized steel-encapsulated polypropylene drum. Material densities tabulated are only for the major elements: Al, Fe, Zr, and/or U in the source regions; and C, O, Al, and/or Fe in the cask. Atom number densities used in the neutron and capture gamma source shielding analyses are summarized in Table 7A-3 (" Material Input for XSDRNPM (TN-40 Cask)") for the I

homogenized upper end fitting, plenum, and basket, fuel and basket, and lower end fitting and basket source regions; the cask body; the torispherical protective cover; the radial neutron e

shield; the homogenized radial neutron shield; and the homogenized steel encapsulated polypropylene drum. Atom number densities tabulated are only for the major elements O, Al,-

{

l 5-8 9

I I

e L

t

E Fe, Zr, and/or 2nU, 2"U in the source regions, and H, B, C, O. Al, and/or Fe in the cask. To estimate the reduced shielding effectiveness of the basket in the axial direction, NSP has reduced tne basket material component in the homogenized upper end fitting and basket, plenum and basket, and lower end fitting and basket source regions by 75 percent.

Section 7A.4 (" Direct Radiation E-W") provides the material densities of the soil in the earthen berm.

The staff shield regional densities used in the TN-40 shielding evaluations are independently determined from the information provided and most closely resemble the shield regional densities of Revision 2 to the TSSAR. Current differences are associated with a staff effort to represent the various regional densities as accurately as possible. Major, minor, and trace element constituents of all materials have been included wherever known. The upper end fittings, basket, and rails source region contains the upper end fittings and appropriate portions of the fuel basket and rails. The void and end plugs regions are void. The plenum springs, fuel cladding, guide tubes, basket, and rails source region contains the plenum springs and appropriate portions of the fuel cladding, guide tubes, fuel basket, and rails. The active fuel, fuel cladding, guide tubes, grid spacers and springs, basket, and rails source region contains the active fuel, grid spacers, and grid spacer springs and appropriate portions of the fuel cladding, guide tubes, fuel basket, j

and rails. The lower end fittings, basket, and rails source region contains the lower end fittings and appropriate portions of the fuel basket and rails. Staff predicted regional densities are greater l

than those of NSP in all source regions and the radial neutron shield. Differences are greatest for the plenum springs, fuel cladding, guide tubes, basket, and rails source region and least for the lower end fittings, basket, and rails source region. The reasons for the differences are numerous and diverse. First, the staff evaluation of the Al content in each source region includes the Al in the aluminum basket supports; the NSP evaluation does not. Second, the staff evaluation of the Zr content in each source region includes the Zr in the guide tubes and grid spacers; the NSP evaluation does not. Third, the staff evaluation of the Fe content in each source region is based on the appropriate elemental composition of Fe in steel and includes the Fe in the stainless steel basket supports; the NSP evaluation assumes all steel to be Fe and does not include the stainless steel basket supports. Fourth, the staff evaluation of Fe in the plenum springs, fuel cladding, guide tubes, basket, and rails source region includes the Fe in the plenum springs; the NSP evaluation does not.

5.3.3 Shielding Evaluation The shielding evaluation and calculational results are addressed in Sections 3.3.5.2 (" Shielding"),

7.4

(" Estimated Onsite Collective Dose Assessment"), 7.5 ("Offsite Collective Dose Assessment"), 7A.2 (" Shielding Analyses"), 7A.3 (" Direct Radiation N-S"), 7A.4 (" Direct Radiation E-W") and 7A.5 (" Dose Rate Around the ISFSI"), and 7A.6 (" Experimental Results")

of the TSSAR. NSP has also provided supplemental information in Revision 2, dated September.

1991, and its response to questions dated June 5,1991, and December 23,1991.

5-9

t Section 3.3.5.2 (" Shielding") addresses the compliance of the TN-40 casks with respect to the radiation limits of 10 CFR Part 72. Reference is made to Section 7 (" Radiation Protection") of the TSSAR, for specific dose estimates.

Section 7.4 (" Estimated Onsite Collective Dose Assessment") provides the estimated dose equivalent rates at dose point locations associated with single cask handling operations, annual surveillance and maintenance activities, and various onsite locations in Tables 7.4-1 (" Design Basis Occupational Exposures for Cask Loading, Transport, and Emplacement ("One Time Exposure)"), 7.4-2 (" Design Basis ISFSI Maintenance Operations, Annual Exposures"), and 7.4-4 (" Dose Rates at Onsite Locations Due to Cask Storage"), respectively, of the TSSAR. In addition to direct radiation, indirect radiation resulting from air scattering of the direct radiation component is included where appropriate.

Section 7.5 ("Offsite Collective Dose Assessment") provides the estimated dose equivalent rate at the dose point location of the nearest resident to the ISFSI. Direct radiation and indirect radiation resulting from air scattering of the direct radiation component are included.

Section 7A.2 (" Shielding Analyses") summarizes the calculational method of the NSP shielding evaluation; tabulates the direct radiation gamma, neutron, and total dose equivalent rates at various locations about the TN-40 cask; illustrates the direct radiation gamma and neutron dose equn alent rates as a function of distance; and illustrates the direct radiation relative gamma and neutron dose equivalent rate factors as a function of time after discharge from the reactor, normalized to 10 years. Primary direct radiation gamma dose equivalent rates are determined with the point kernel code QAD-CGGP. Direct radiation neutron and capture gamma dose equivalent rates are determined with the one-dimensional transport code XSDRNPM and XSDOSE and include suberitical neutron multiplication. Flux-to-dose rate conversion factors j

used are those of ANSI /ANS-6.1.1. NSP calculated direct radiation gamma, neutron, and total dose equivalent rates at various locations about the TN-40 cask on the surface and 1-meter and 2-meter distances from the surface are summarized in Table 7A-4 ("TN-40 Dose Rates at Short Distances") of the TSSAR. Direct radiation gamma and neutron dose equivalent rates as a function of distance are calculated with the long distance spherical model and XSDRNPM and are presented in Figure 7A-6 (" Dose Rates at Long Distances (mrem /hr)") of the TSS AR. Direct radiation relative gamma and neutron dose equivalent rate factors as a function of time, after discharge from the reactor, normalized to 10 years, are also calculated with the long-distance spherical model and XSDRNPM and are presented in Figure 7A-7 (" Relative Dose Rate Factor (Normalized to 10 Years)") of the TSSAR.

i Section 7A.3 (" Direct Radiation N-S") summarizes NSP's calculational method for determining the direct radiation at the centerline of the west and east pads in the north-south (N-S) direction from an array of 24 and 48 casks at 9 and 21 years, respectively. The direct radiation contribution from each front-row cask is determined at each dose point location, using the long-distance spherical model and XSDRNPM. Decay factors are applied to correct each cask for the appropriate storage time. The direct radiation contribution from each back-row cask at each dose point is calculated from the product of the direct-radiation contribution from each front-row cask 5-10

and a back-row to front-row contribution fraction derived from point kernel shielding analyses of the direct-radiation relative dose equivalent rates at each dose point from each front row cask and each back-row cask with the front-row of casks, modeled as uranium cylinders. The dose equivalent rates from each cask at each dose point in the point kernel shielding analyses are totaled and the back-row to front row contribution fraction calculated. Direct radiation relative dose equivalent rates at specified distances from the front and back-rows, and the back-row to front-row contribution fractions are summarized for the west pad at 9 years (24 casks), wer.t pad at 21 years (48 casks), and east pad at 21 years (48 casks), in Tables 7A-5 (Relative Dose from Front and Back Row of TN-40 Casks (" West Pad at 9 Years)", 7A-5a (" East Pad (21 Years)

Relative Cask Dose Rates at Specified Distances"), and 7A-5b (" West Pad (21 Years) Relative Cask Dose Rates at Specified Distances") of the TSSAR, respectively.

Section 7A.4 (" Direct Radiation E-W") summarizes the calculational method of NSP for determining: (1) the direct radiation at the centerline of the west and east pads in the east-west (E-W) direction from an array of 24 and 48 casks at 9 and 21 years, respectively; (2) the indirect radiation resulting from air scattering of the direct radiation component; and (3) the attenuation of the earthen berm as a function of thickness. The direct radiation contribution from each end-row cask is determined at each dose point location, using the long-distance spherical model and XSDRNPM. The direct radiation contribution from each subsequent cask in each row is the product of the direct radiation contribution from each end-row cask and a subsequent cask to-preceding-casks contribution fraction. This fraction is derived at each dose point from point kernel shielding analyses of the direct radiation relative dose equivalent rates at each dose point from each end row cask and each subsequent cask in each row with all preceding casks in the row modeled as uranium cylinders. NSP evaluations of the direct radiation contribution from each sub',equent cask in each row is described as insignificant. Direct radiation gamma and neuwn dose equivalent rates at specified distances are summarized at 9 years (24 casks) and 21 years (48 casks) in Table 7A-6 (" Total Direct Dose Rate (TN-40 Casks) (mrem /hr)") of the TSSAR. The indirect radiation resulting from air scattering of the direct radiation component is determined at each dose point location using SKYSHINE II. Equival:nt gamma and neutron point sources are employed with decay factors applied to correct each for the appropriate storage time. Indirect radiation gamma and neutron dose equivalent rates at specified distances are summarized at 9 years (24 casks) and 21 years (48 casks) in Table 7A-7 (" Total Skyshine Dose Rate (TN-40 Casks) (mrem /hr)") of the TSSAR. Attenuation of the earthen berm as a function of thickness is determined using the long distance spherical model and XSDRNPM.

Direct radiation gamma and neutron dose equivalent rates and attenuation factors are summarized as a function of earthen berm thickness in Table 7A-8 (" Attenuation Factor for Earth Berm")

4 of the TSSAR.

I Section 7A.5 (" Dose Rate Around the ISFSI") describes and illustrates the calculational method of NSP for determining the annual total dose equivalent rate resulting from ISFSI operations.

Direct and indirect radiation gamma and neutron dose equivalent rates are simply added and then multiplied by the appropriate number of hours in 1 year. Annual total dose equivalent rates for an ISFSI surrounded by a 1.8 m (6-ft) thick earthen berm as a function of distance in east-west l

and north-south directions are summarized in Figures 7A-10 ("ISFSI Dose Rate (N-S Direction 5-11

-l mrem /Yr)") and 7A-11 ("ISFSI Dose Rate (E-W Direction, mrem /Yr)") of the TSSAR, respectively.

Section 7A.6 (" Experimental Results") provides the INEL experimental measurement results for -

the TN-24P, a prototype cask. Gamma and neutron dose equivalent rates are summarized'for dose point locations on the cask surface and at I meter (1 yard) from the cask surface on the axial centerline and at the midplane of the active fuel region. NSP corrections to the experimental measurement results to account for differences in fuel parameters and material thicknesses between the TN-40 and TN-24P, are also provided.

}

The staff evaluation of the TN-40 shielding is performed with the MicroShield and COG -

(Reference 13) codes. MicroShield is a microcomputer adaptation of ISOSHLD, a point kernel code for the evaluation of gamma dose equivalent rates. Buildup factors employed in this use of MicroShield are the geometric progression fitting function representation of appropriate cask materials at the dose point of interest. COG uses the Monte Carlo method to transport both neutrons and gamma rays and was used for the shielding evaluation. With MicroShield, the cylindrical source with cylindrical shields was used to evaluate the dose rate on the sides of the cask; and the cylindrical source with slab shields geometries was used to evaluate the dose rate on the ends. The gamma dose equivalent rates at dose point locations on the axial centerline of the top and bottom surfaces of the cask and at the radial midplane of each source region on the side surface of the cask were determined. With COG, a three-dimensional fm' ite cylinder analysis was used. The average gamma and neutron dose equivalent rates at the top, side, and -

bottom surfaces of each cask were determined. Areas used in the determination of these average dose equivalent rates are at the cask surface and above, about, and below the cask cavity. COG was also used in the evaluation of the TN-40 cask shielding as a function of distance from the-cask. Here the average gamma and neutron dose equivalent rates at distances of 1, 2, and 3 meterr (yds) above the cask torispherical protective cover and from the radial neutron shield shell were determined. Areas used in this evaluation of the average dose equivalent rates are based on the projection of the cask cavity to the distance of interest. Dose rate conversion factors used in all COG calculations are those of ANSI /ANS-6.1.1.

5.4 Findings and Conclusions Staff calculations show:

(1) NSP gamma dose equivalent rates to be higher than staff calculations at all common dose point locations; (2) NSP neutron dose equivalent rates to be lower than staff calculations at all common dose point locations, except the cask bottom and 2 meters (2 yds) above the cask torispherical protective cover and from the radial neutron shield; and (3) NSP total dose equivalent rates to be higher than staff calculations at all common dose point locations except 2 meters (2 yds) above the cask torispherical protective cover.

NSP calculated gamma dose equivalent rates at the top, side, and bottom surfaces of a TN-40 cask are.234 mSv/hr (23.4 mrem /hr),.451 mSv/hr (45.1 mrem /hr), and.798 mSv/hr (79.8 mrem /hr), respectively; staff calculated average gamma dose equivalent rates for these surfaces are metric 0.09 mSv/hr (9.0 mrem /hr)i29 percent,.309 mSv/hr (30.9 mrem /hr)il3 percent, l

^

5-12

= _.

~

and 0.59 mSv/hr (52.9 mrem /hr)il4 percent, respectively.

NSP calculated gamma dose equivalent rates at 2 meters (2 yds) from the top and side surfaces of a TN-40 cask are.059 mSv/hr (5.9 mrem /hr) and.168 mSv/hr (16.8 mrem /hr), respectively; staff calculated average gamma dose equivalent rates for these surfaces are.043 mSv/hr (4.3 mrem /hr) 42 percent and.084 mSv/hr (8.4 mrem /hr)il6 percent, respectively. Differences between staff and NSP average gamma dose equivalent rates are due in some measure to the choice of codes: three-dimensional point kernel (QAD-CGGP) versus three-dimensional Monte Carlo (COG). Source region density variations are also a factor, t

NSP calculated neutron dose equivalent rates at the top, side, and bottom surfaces of a TN-40 cask are.026 mSv/hr (2.6 mrem /hr);.124 mSv/hr (12.4 mrem /hr); and 11.95 mSv/hr (1195.0 mrem /hr), respectively. Staff calculated average 1.eutron dose equivalent rates for these surfaces are.064 mSv/hr (6.4 mrem /hr) 143 percent,.173 mSv/hr (17.3 mrem /hr), ill percent and 5.203 mSv/hr (520.3 mrem /hr) i6 percent, respectively. NSP calculated neutron dose equivalent rates at 2 meters (2 yds) from the top and side surfaces of a TN-40 cask are.005 mSv/hr (0.5 mrem /hr) and.029 mSv/hr (2.9 mrem /hr), respectively; staff calculated average l

neutron dose equivalent rates for these surfaces are.03 mSv/hr (3.0 mrem /hr) 38 percent and

.037 mSv/hr (3.7 mrem /hr) i10 percent, respectively. Differences between staff and NSP

?

average neutron dose equivalent rates are due in some measure to the choice of codes: one-dimensional transport (XSDRNPM) versus three-dimensional Monte Carlo (COG). Source region spectrum group structure variations may also be a factor. Either sets of doses are acceptable i

for safe operations, NSP calculated total dose equivalent rates at the top, side, and bottom surfaces of a TN-40 cask are.26 mSv/hr (26.0 mrem /hr),.575 mSv/hr (57.5 mrem /hr), and 12.748 mSv/hr (1274.8 I

mrem /hr), respectively. Staff calculated average total dose equivalent rates for these surfaces are.154 mSv/hr (15.4 mrem /hr)i25 percent,.482 mSv/hr (48.2 mrem /hr) 10 percent; and 5.732 mSv/hr (573.2 mrem /hr) 16 percent, respectively. NSP calculated that total dose equivalent rates at 2 meters (2 yds) from the top and side surfaces of a TN-40 cask are.064 mSv/hr (6.4 mrem /hr) cnd.197 mSv/hr (19.7 mrem /hr), respectively. Staff calculated average total dose equivalent rates for these surfaces are.073 mSv/hr (7.3 mrem /hr)i29 percent and

.121 mSv/hr (12.1 mrerr/hr)ill percent, respectively.

Annual dose equivalent at 180 meters (591 feet) due west from the center of the ISFSI from an array of 48 casks to any individual, conservatively assumed to be continuously present, is calculated by staff to be less than.485 mSv (48.5 mrem). This is less than the 1 mSv (100 mrem) allowed under 10 CFR Part 20.

Annual dose equivalent at 732 meters (2400 feet) northwest from the center of the ISFSI from an array of 48 casks to any individual, conservatively assumed to be continuously present, is calculated by staff to be less than.00042 mSv (0.042 mrem). This is greater than the.00028 mSv (0.028 mrem) calculated by NSP; however, both are substantially less than the.25 mSv (25 mrem) allowed under 10 CFR 72.104(a).

t 5-13

)

6 CRITICALITY EVALUATION 6.1 Area of Review The review of the criticality safety evaluation is to confirm that the design features of the TN-40 spent fuel storage cask that ensure suberiticality are based on conservative and acceptable analytical and/or test methods. This chapter addresses the requirement to ensure that margins of safety for the effective multiplication factor are commensurate with the uncertainties in the storage conditions, data employed in the calculations, and analytical methods. This requires review of the analytical models used to represent the dry storage cask, the material properties in the models, and the convergence criteria. The benchmarking of the computational tools is also reviewed for accuracy and applicability of the selected critical experiments.

6.2 Acceptance Criteria 10 CFR 72.124(a), (" Design for criticality safety"), requires that spent fuel handling, packaging, transfer, and storage systems must be designed to be maintained suberitical and to ensure that, before a nuclear criticality accident is possible, at least two unlikely, independent, and concurrent or sequential changes have occurred in the conditions essential to nuclear criticality safety. The design of handling, packaging, transfer, and storage systems must include margins of safety, for the nuclear criticality parameters, that are commensurate with the uncertainties in the data and methods used in calculations and must demonstrate safety for the handling, packaging, transfer, and storage conditions, and in the nature of the immediate environment under accident conditions.

Subparagraph (c) (" Criticality monitoring"), requires that a criticality monitoring system shall be maintained, in each area where special nuclear material is handled, used, or stored that will energize clearly audible alarm signals if accidental criticality occurs. Underwater monitoring is not required when special nuclear material is handled or stored beneath water shielding.

Monitoring of dry storage areas where special nuclear material is packaged in its stored configuration under a license issued under this subpart is not required.

The requirements of these sections of 10 CFR Part 72 can be satisfied ifit is demonstrated that the effective multiplication factor is less than 0.95 (including all uncertainties) for all credible configurations of the design basis fuel assemblies in the TN-40 storage cask. The double contingency requirement will be satisfied by assuming that the fuel assemblies loaded into the cask are unitradiated; thus, maximizing the multiplication factor.

6.3 Review The design features that are intended to maintain suberiticality are described in section 3.3.4.1

(" Control Methods for Prevention of Criticality") of the TSSAR. Three features of this design identified as significant to the maintenance of subcriticality are: (1) boron is an integral part of the basket; (2) boron is dissolved in the spent fuel pool water to a concentration of at least 1800 6-1

.i

-}

ppm, whenever a spent fuel storage cask is being loaded; and (3) the weather cover, seals, and seal pressurization system prevent water from entering the cask confinement vessel during-storage.

The criticality analysis described in the TSSAR was performed with AMPX cross section processing codes that included BONAMI and NITAWL, and the Monte Carlo criticality code KENO V.a combined with the 27 group cross-section set based on the Evaluated Nuclear. File, version) B-IV data. These codes and cross sections are on the SCALE system available from the Reactor Shielding Information Center, Oak Ridge,-Tennessee. The effective multiplication of the cask, at the conclusion of the loading process, while the cask is immersed in the spent fuel pool with its lid removed, was 0.8988i0.0033, as described in section 3.3.4.1 (" Control Methods for Prevention of Criticality"), of the TSSAR.

Section 3.3.4.3 (" Verification Analyses-Benchmarking") of the TSSAR describes the benchmarking process, and the results of calculations of five critical experiments at PNL are presented in Table 3.3-7 (" KENO V.a Benchmark Results"). The average calculated k,g is less than unity, so a bias of 1.2 percent should be applied to all calculated effective multiplication factors (Reference 14). When the bias is included, the value of k,y becomes 0.9162.

Section 3.3.4.1 (" Control Methods for Prevention of Criticality") of the TSSAR also addresses the variation of the effective multiplication factor as the water level in the cask cavity is lowered during draining, with the results presented in Table 3.3-4 ("TN-40 Reactivity during Draining").

The variation of the effective multiplication factor for various water densities in the entire cask cavity is presented in Table 3.3-5 ("TN-40 Reactivity Versus Water Density"). In both cases, the maximum value of the effective multiplication factor is 0.9011 0.0036. This value, when corrected for the bias, is less than 0.95, so subcriticality is claimed.

Table 3.3-5 ("TN-40 Reactivity Versus Water Density") of the TSSAR presents the variation of k,y with the density of pure water in the cask cavity. When the water density reaches unity, the cask cavity is filled with pure water and the value of k,y is reported as 0.9306 +.0036 which i

results in a k,y that is less than 0.95 after inclusion of the bias. However, the TSSAR does not seek relief from the requirement for 1800 ppm of boron in the spent fuel pool water whenever a spent fuel cask is being loaded, so this calculation is not considered as limiting the analysis.

Section 3.3.4.2 (" Error Contingency Criteria") of the TSSAR addresses the need for additional conservatism in the evaluation of the effective multiplication factor by noting that adequate conservatism is incorporated into the geometric modeling, so no additional contingency is required to accommodate errors.

The fuel assembly analyzed in the TSSAR is the Westinghouse 17 x 17 OFA (Reference 15).

This fuel assembly design is stated in the TSSAR to yield higher k-effective values with respect to similar analyses using a 14 x 14 or a 15 x 15 PWR design.

A series of confirmatory calculations were performed to provide assurance that the results 6-2

presented in the TSSAR appropriately characterize the maintenance of suberiticality of the reference fuel assemblies in the TN-40 spent fuel cask.

The criticality review of the fuel basket configuration was performed with the SCALE system, using the KENO-V.a code combined with the 123GROUPGMTH cross-section set in Reference

16. Criticality calculations using these computational tools were performed on a personal computer. The fuel assembly analyzed in the SER was the Westinghouse 14 x 14 OFA. The fuel rods, basket, and cask were modeled discretely in three dimensions.

Both off-centered fuel loading conditions and water draining conditions in the cask were l

considered in the review. These issues are discussed below. In one calculation, the off-centered loading had the fuel bundles placed as closely as possible near the corner of each hole in the basket nearest the center of the cask. In another calculation, the off-centered loading had the fuel bundles placed as closely as possible near the corner of each hole in the basket nearest the outside of the cask. The k-effective result for the to-the-center loading was higher (but within I sigma) than the centered loading; the to-the-outside loading was 2 percent absolute lower than the centered loading. In another calculation, draining of the cask was studied at 2.5-cm (1-inch) intervals. K-effective results were calculated for 366 cm (144 inches), 363 cm (143 inches), 361 cm (142 inches), etc., of active fuel covered with borated water between 366 cm (144 inches) and 213 cm (84 inches). K-effective values were calculated at 15-cm (6-inch) intervals below 213 cm (84 inches) until the fuel was uncovered. The peak k-effective value was found when about 305 cm (120 inches) of fuel was covered with borated water, and this k-effective value was within 2 sigmas of the k-effective value for the cask fully flooded with borated water.

Confirmatory calculations with the cask cavity completely filled with partial density water were also performed (water densities between 0.001 and 1 grams water /cc air-water mixture). No k-effective values higher than the effective multiplication factor for normal storage conditions were found.

l The calculation method and cross-section values used in the criticality review were verified oy comparison with critical experiment data for assemblies similar to those for which the cask was designed. Twelve critical experiments were analyzed using the same computer program, cross-l sections, and personal computer. These experiments included low enriched (< 5 wh) uranium oxide fuel rod arrays in water, in borated water, and in borated water with fuel assemblies spaced between borated aluminum plates (Reference 17 and 18). The concentrations of boron in the water and in the aluminum plates were smaller than those in the TSSAR. From these results, the bias of the computational tools used in the criticality review was determined.

6.4 Findings and Conclusions There is no credible way to lose the boral in the basket in this cask; therefore, there is no need to verify the continued efficacy of the neutron absorbing material, as required by 10 CFR l

72.124(b).

l The value of k-effective that is reported in the TSSAR for normal storage conditions is 0.916, 6-3

based upon a two-dimensional model of the cask and basket with the design basis fuel assemblies. This is an upper limit value at 95 percent confidence and includes a bias developed from the evaluation of five critical experiments that had fuel rod arrays and materials that are comparable to the materials of construction of the TN-40 cask. The change in the effective multipli:.ation factor was examined in the TSSAR for partially filled casks and for casks filled with a mist (reduced density pure water), and the effective multiplication factor did not increase above the value for normal storage conditions. The largest k-effective value found in the confirmatory axialysis was 0.937 for 305 cm (120 inches) of fuel covered with borated water.

This is an upper limit value, at 95 percent confidence, with biases applied. The confirmatory analysis models of the TN-40 cask bound the actual fuel and basket configurations and materials.

The calculated results for these models show the largest effective multiplication factor to be less than the design c 'terion of 0.95. On the basis of the TSSAR evaluation and the confirmatory i -

analysis, the staft concludes that the fuel assemblies stored in TN-40 cask will be maintained suberitical, in compliance with the applicable sections of 10 CFR Part 72.

?

1 l

l 6-4 i

I

7 CONFINEMENT EVALUATION 7.1 Area of Review i

Cask confinement is deemed acceptable if it can be shown that the dose equivalent to any

.1 individual from gaseous activity release due to normal operations and postulated off-normal events is not in excess of the applicable criteria given in Section 7.2.

7.2 Acceptance Criteria 10 CFR 72.24(1) requires, in part, a description of the equipment to be installed to maintain control over radioactive materials in gaseous effluents produced during normal operations and expected operational occurrences. The description must include an estimate of each _of the principal radionuclides expected to be released annually to the environment in gaseous effluent produced during normal ISFSI operations.

10 CFR 72.104(a) requires that during normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area shall not exceed

.25 mSv (25 mrem) to the whole body,.75 mSv (75 mrem) to the thyroid and.25 mSv (25 mrem) to any other organ, as a result of exposure to: (1) planned discharges of radioactive materials, radon and its daughters excepted, to the general environment; (2) direct radiation from ISFSI operations; and (3) any other radiation from uranium fuel cycle operations within the region.

The review focuses on the dose equivalent from a single cask to any individual from gaseous activity release because of normal operations and postulated off-normal events. In the context of this review, occupational radiation exposure is not addressed, and off-normal events are anticipated occurrences. Dose equivalent because of normal operations and postulated off-normal events is based on exposure of an individual at the minimum site boundary distance for a continuous period of 1 year. The minimum site boundary distance used is the minimum site boundary assumed in the TSS.AR (110 meters) (120 yds).

7.3 Review The review is divided into four main parts: (1) cask confinement description and design features, (2) gaseous activity inventory within the cask cavity; (3) estimated annual dose equivalent from 3

gaseous activity release because of normal operations; and (4) estimated annual dose equivalent

~

from gaseous activity release because of postulated off-normal events.

7.3.1 Cask Confinement Description and Design Features 1

The TN-40 cask confinement barriers and design features are described in Sections 1.3 (" General Storage System Description"), 3.3.2.1 (" Confinement Barriers and Systems"), 3.3.2.2 (" Heat 7-1

(

Transfer Design"),3.3.3.2 (" Instrumentation"),3.3.5.3 (" Radiological Alarm Systems"),4.4.5.2

(" Alarm System"), 5.1.1 (" Narrative Description"), and 5.1.3.2 (" Instrumentation") of the TSS AR. NSP also provided supplemental information in Revision 2, dated September 1991, and its response to questions dated December 23,1991.

Sections 1.3 (" General Storage System Description"), 3.3.2.1 (" Confinement Barriers and Systems"), 3.3.3 (" Protection by Equipment and Instrumentation Selection"), and 3.3.5.3

(" Radiological Alarm Systems") describe the cask confinement barriers.

The cask body confinement consists of an SA-203 steel flange forging that is welded to a right circular inner shell of SA-203 steel, which, in turn, is welded to a SA-203 steel bottom closure that has a 434.3 cm (171.0-in.) overall length, a 231.1-cm (91.0-in.) cross-sectional diameter at the flange j

forging, a 190.5-cm (75.0-in.) cross-sectional diameter at the side wall, a 3.8-cm (1.5-in.) thick l

side wall, and provides a cavity that is 182.88-cm (72.0-in.)in diameter and 414.02-cm (163.0-in.) long. The cask confinement lid consists of a flanged SA-350, Grade LF3 steel lid that has a 231.1-cm (91.0-in.) cross-sectional diameter and is 11.43-cm (4.5-in.) thick. It is secured to the upper end of the cask body confinement with 48 SA-320, Grade L43 bolts. There are two penetrations through the cask lid confinement: one for draining and one for venting of the cask cavity. Separate lids made 21.29-cm (8.38-in.) diameter and 2.54-cm (1.0-in.) thick SA-240, Type 304 steel are provided to cover the drain and vent connections. Each lid is secured with eight SA-193, Grade B-8 bolts. The cask lid confinement and penetrations use a double barrier seal system, with two metallic O-rings forming each seal. To preclude leakage of air into the cask, the cask cavity is pressurized above atmospheric pressure with helium. The interspace between the metallic O-rings in the cask lid confinement and lid penetration seals is monitored and connected to a helium-filled tank that sits atop an SA-516, Grade 55 steel encased polypropylene drum that is bolted to the cask lid confinement. The tank pressure is maintained above the cask cavity pressure, to ensure that any seal leakage would be into rather than out of the cavity. A 0.97-cm (0.38-in.) thick SA-516, Grade 55 steel flanged torispherical protective cover provides weather protection for the cask lid confinement penetrations. It is sealed with a viton 0-ring and secured to the cask body confinement with 48 SA-193, Grade B-8 bolts.

Section 3.3.2.1 (" Confinement Barriers and Systems") describes the cask lid confinement and torispherical protective cover confinement barriers, details the cavity pressure variations associated with fuel clad failures, and defines and illustrates the monitoring system pressure in terms of the helium test leakage rate. With the exception of specifications for the seal leak rate and the operating pressures of the cask cavity and the monitoring system overpressure tank, the cask lid confinement and torispherical protective cover confinement barrier descriptions reflect the information presented in Section 1.3 (" General Storage System Description") of the TSSAR, 7

Cask seals are stated to be capable of limiting leak rates to less than 1.000x10 atm-cc/sec of helium. The cask cavity maximum operating pressure is initially set at 2.23 bar (2.2 atm).

However, after 20 years of storage, the cask cavity operating pressure will decrease because of a decrease in the average gas temperature of the cavity. The initial operating pressure of the cask monitoring system overpressure tank is set at 5.57 bar (5.5 atm). Any decrease in the pressure of the cask monitoring system will be signaled by a pressure transmitter mounted at the side of the cask. Cavity pressure variations are summarized for two assumed fuel clad failure 7-2

~

r conditions: 10 percent and 100 percent of the cask fuel rods. For an off-normal cavity gas temperature of 226 *C (439 *F), the helium and radioactive gases and vapors released from the 10 percent and 100 percent failed fuel rods into the cask cavity can cause the absolute pressure in the cask. to increase to 2.49 bar (36.13 psia) and 4.82 bar (70.0 psia), respectively.

Expressions for the time-dependent monitoring system pressure in terms of the helium test leakage rate are derived for choked and unchoked flow conditions. The basis for the choked and unchoked flow condition derivations is ANSI N14.5. The determination of the maximum allowable test leakage rate is illustrated in Figure 3.3-2 ("TN-40 Cask Pressure Monitoring System Test Leakage Rate"). Monitoring system gas and cavity gas pressures as a function of time for various test leak rates are illustrated in Figure 3.3-3 ("TN-40 Cask Pressure Monitoring System - System Pressure").

Section 3.3.2.2 (" Heat Transfer Design") sets a maximum temperature criterion of 299 *C (570

• F) for the double metallic O-rings (Helicoflex seals) in the cask lid confinement.

Section 3.3.3.2 (" Instrumentation") addresses the availability and design capabilities of the cask pressure monitoring instrumentation.

Section 3.3.5.3 (" Radiological Alarm Systems") addresses the availability of the cask pressure monitoring instrumentation and specifies that the ISFSI Operating Procedures will contain the procedures to be followed if an alarm associated with a cask pressure monitor is activated.

Section 4.4.5.2 (" Alarm System") addresses the cask pressure monitoring alarm system. The pressure monitoring devices will provide an analog input signal to actuate alarm indication at a monitoring panel outside the ISFSI gate.

Section 5.1.1 (" Narrative Description") describes the confinement design features associated with loading the cask. Loading of the spe' : fuel is to take place in spent fuel storage pool No.1 of.

the reactor building. Once the fuel is loaded and the lid is in place, the cask is lifted to the pool surface and the lid bolts installed. The internal cask cavity is drained by displacing the water with air from the plant compressed air line. The, cask is then returned to the auxiliary building rail bay. The cavity is dried using a vacuum drying system. Once the cavity is dry, it is backfilled with dry helium gas to a pressure of approximately 2.03 bar (2 atm) and the lid seals are leak-checked. The top neutron shield drum is then installed. Once it is secured. the i

overpressure system is pressurized with helium to a pressure of approximately 5.57 bar (5.5 i

atm) and the interspace seals are leak-checked. Proper functioning of the pressure monitoring -

system is ensured before the torispherical protective cover is installed and the cask is transferred to the ISFSI.

Section 5.1.3.2 (" Instrumentation") addresses the use of pressure transducers to monitor the space between the metallic O-rings, to provide an indication of seal failure before any release can occur.

The use of two identical transducers to ensure a functional system through redundancy is also specified.

7-3 h

R 7.3.2 Caseous Activity Inventory within the Cask Cavity Gaseous activity (MBq) calculations and inventories expected to be found within the cask cavity are addressed in Sections 3.1.1 (" Spent Fuel to be Stored"), 7.2.1 (" Characterization of Sources"), and 7.2.2 (" Airborne Radioactive Material Sources") of the TSSAR. NSP has als9 provided supplemental information in Revision 2, dated September 1991.

Section 3.1.1 (" Spent Fuel to be Stored") summarizes the design basis fuel parameters, irradiation conditions, and cooling time for a spent fuel assembly, Gaseous activity is determined for a spent fuel assembly from an ORIGEN2 calculation, using the Westinghouse 14x14 array OFA design basis fuel described in Tables 3.1-1 (" Fuel Assembly Parameters") and 3.1-2 (" Thermal, Gamma and Neutron Sources for the Design Basis 14x14 Westinghouse OFA Fuel Assembly") of the TSSAR. In this calculation, the average burnup is 45,000 MWD /MTU, the specific power is 37.5 MW/MTU, the initial fuel enrichment is 3.85 percent, there are 380 kg (838 lb.) of heavy metal per assembly, and the minimum ti.ae after discharge is 10 years.

The irradiation time and interim shutdown periods are not specified.

Section 7.2.1 (" Characterization of Sources") summarizes again the irradiation conditions for the ORIGEN2 calculation. In addition, the irradiation time and interim shutdown periods are addressed. A three-cycle operating history with interim shutdown of 30 days is specified.

Section 7.2.2 (" Airborne Radioactive Material Sources") examines gaseous activity inventory (MBq) (Ci) within the cask cavity as a function of time after discharge from the reactor, and selected fuel rod gap gas release fractions. Specific nuclides identified for release into the cask cavity are 'H, "Kr, *Cs, and "7Cs. Total cask cavity inventories (MBq) (Ci) are summarized as a function of time after discharge from the reactor in Table 7.2-7 (" Fission Gas and Volatile Nuclides Inventory (Curies /40 Assemblies), Westinghouse OFA 14x14 3.85 w/o "U, 45,000 MWD /MTU,10 Year Cooling Time") of the TSSAR. Cited fuel rod gap release fractions of 30 percent for Kr and 10 percent for all other noble gases are those obtained from Regulatory Guide 1.25.

7.3.3 Estimated Annual Dose Equivalent from Gaseous Activity Release because of Nonnal Operations NSP provides no estimated annual dose equivalent from gaseous activity release because of normal operations. Sections 3.3.5.3 (" Radiological Alarm Systems") and 7.3.3 "(Area Radiation and Airborne Radioactivity Monitoring Instrumentation") of the TSSAR state that there are no credible events that could result in releases of radioactive products.

The staff did not evaluate the annual dose equivalent from gaseous activity release because of normal operations, because the region between the 0-rings is pressurized to prevent any leakage of the cask contents.

7-4

7.3.4 Estimated Annual Dose Equivalent from Gaseous. Activity Release due to Postulated Off-normal Events The review is divided into two main parts: (1) postulated off-normal events, and (2) annual dose equivalent from postulated off-normal events.

Section 8.1.1 (" Event") identifies one event that may result in off-normal operations: loss of electrical power to the ISFSI. Postulated causes for the loss of electrical power to the ISFSI are found in Section 8.1.1.1 (" Postulated Case of the Event"). The means of detecting the loss of electrical power to the ISFSI is discussed in Section 8.1.1.2 (" Detection of Event"). Analysis of the effects and consequences for the loss of electrical power to the ISFSI appears in Section 8.1.1.3 (" Analysis of Effects and Consequences"). Proposed corrective actions for the loss of electrical power to the ISFSI appear in Section 8.1.1.4 (" Corrective Actions").

NSP provides no annual dose equivalent from postulated off-normal events. Sections 8.1.1.3

(" Analysis of Effects and Consequences") and 8.1.2 (" Radiological Impact from Off-Normal Operations") of the TSSAR state that there is no radiological release or impact associated with the off-normal event of loss of electrical power to the ISFSI.

The staff did not evaluate the annual dose equivalent from postulaul off-normal events, because of the double 0-ring system, which will prevent any leakage of the cask contents.

7.4 Findings and Conclusions The cask lid and interspace seals must be leak-tested, for a period of 30 minutes, to a sensitivity of 1.000x10-5 atm-cc/sec.

Gaseous activity release is not considered credible under normal operations and postulated off-normal events.

Cask confinement is in compliance with appropriate guidance and/or regulations.

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8 OPERATING PROCEDURES 8.1 Area of Review Operating procedures for the TN-40 cask are described in Chapter 5 and Section 7.1 of the

[

TSSAR. This SER review is limited to the procedures as presented by NSP in this TSSAR.

8.2 Acceptance Criteria 10 CFR 72.24(h) requires the applicant to submit "...a plan for the conduct of operations including the planned managerial and controls system, and the applicant's organization, and program for training of personnel...." Although this provision applies primarily to the ISFSI, the operations involved in loading, transporting, and storing of the spent fuel are closely -

associated with the design of the cask, to the extent that design features are incorporated to facilitate the conduct of these operations. Spent feel loading and cask handling procedures in the auxiliary building are governed by the 10 CFR Part 50 license. Consequently, the review of the operating procedures is limited to the specific operations of handling the cask from the time it leaves the auxiliary building until it is placed on the storage pad. Managerial and administrative controls would only be relevant if the cask design were such that only administrative controls could ensure that the spent fuel could be safely handled and stored under conditions that would not pose a hazard to operating personnel or the public.

10 CFR Part 20 covers the standards, for protection against radiation, that must be met during the operation of an ISFSI.

Regulatory Guides 8.8 and 8.10 (References 19 and 20) provide guidance to ensure that occupational radiation exposures will be "as low as is reasonably achievable" (ALARA).

8.3 Review TSSAR Section 5.1.1 provides a general description of the operational procedures for loading the cask and preparing it for storage. More detailed procedures describing the receipt and loading of the TN-40 cask at the ISFSI are described in flowsheet form in Fig. 5.1-1.

Inspections and tests are described as part of the preparation for loading.

TSSAR Section 7.1 describes the general procedures to be followed to meet the requirements of 10 CFR Part 20, Regulatory Guide 8.8, and Regulatory Guide 8.10.

These general procedures for radiation protection and meeting ALARA criteria for occupational exposure apply to the cask loading procedure.

NSP has an NRC-approved physical protection program for the Prairie Island Nuclear Generating Plant. It has also developed an amendment to this program, to accommodate the needs of the ISFSI. The NRC staff reviewed and approved this amendment, concluding that the physical protection requirements for the ISFSI (10 CFR Part 72, Subpart H) will be met with 8-1

the incorporation of the amendment into the site physical protection plan.

NSP currently has a training program that has been approved, by NRC staff, for reactor operations. Additional sections will be added to this program, as needed, to include information pertinent to the ISFSI. All personnel performing cask and fuel handling functions will be given additional training in specific areas, as required by the radiation protection program in effect at Prairie Island. The NSP ISFSI training should address the following:

a. TN-40 cask design;

b. ISFSI Design;

c. ISFSI Safety Analysis;

d. ISFSI Technical Specifications; and

e. ISFSI Normal and Off-normal Procedures.

The NRC staff finds that the existing training program, when modified to include the ISFSI, will comply with the requirements of 10 CFR Part 72, Subpart I.

The TN-40 Cask TSSAR addresses the cask receipt, loading, and some onsite transportation procedures at the ISFSI. Procedures for unloading the cask are not covered in the TSSAR, even though this operating procedure is inseparable from decommissioning. This review only covers the inspections, tests, and special preparations of the cask for loading spent fuel. Section 5.1.1 of the TSSAR addresses the loading procedures, whereas Section 7.1 of the TSSAR addresses the issue of ensuring that the occupational radiation doses are ALARA.

(

Detailed procedures for loading, draining and drying the cask, creating an inert environment for the spent fuel, ensuring the effectiveness of the seals at the bolted closure joints, transporting the loaded cask to the storage pad, and ensuring that occupational radiation exposures are maintained ALARA should be developed before NSP loads fuel into the first cask..This has been made a license condition.

8.4 Findings and Conclusions r

The operational procedures for loading the TN-40 cask are in compliance with the appropriate guidance and/or regulations. These procedures must be incorporated into the operational procedures for the reactor. The written procedures for onsite transportation, unloading, and preoperational testing must be added to the operational procedures for the ISFSI.

8-2

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O 9

ACCEPTANCE TESTS AND MAINTENANCE PROGRAM 9.1 Acceptance Tests Section 9.2.3 of the TSSAR addresses the subject of acceptance tests. Specific requirements for -

testing to demonstrate satisfactory fabrication of the cask include insertion of a dummy fuel assembly into the basket while the cask is submerged in the spent fuel pool, to ensure freedom from obstructions. In addition, leak-testing of the cask shall be completed after loading and before transporting the cask to the ISFSI. The leak-testing requirements are reviewed in Section 7 of this SER.

9.2 Maintenance Program Maintenance is addressed briefly in Section 5.1.3.3 of the TSSAR. The position is taken that the TN-40 scent fuel storage cask satisfies its functions of shielding, confm' ement, and criticality control in a totally passive manner and, as such, no maintenance should be required. The possibility of having to repair local areas of the corrosion-inhibiting coating (paint and/or Zn/AL coating) is acknowledged, but no description of techniques is provided. In addition, the description of maintenance activities acknowledges the potential need-for replacement or recalibration of the pressure-monitoring instrumentation.

This position is acceptable to the staff, provided that it is recognized that any maintenance that involves the seals or requires reinsertion of a cask into the spent fuel pool would require procedures to be developed and reviewed before their execution. This approach is justified because it is impossible to anticipate maintenance needs other than cleaning of the outer surface of the casks.

u R

9-1 i

V

s 10 RADIATION PROTECTION 10.1 Area of Review Radiation protection is deemed acceptable if it can be shown that the occupational radiation exposures are ALARA and are in compliance with appropriate _ guidance and/or regulations.

Also, it must be shown that the occupational radiation exposures from transport, storage, and repair activities are not in excess of 10 CFR Part 20 limits.

10.2 Acceptance Criteria J

10 CFR 20.101(a) requires, in part, that no licensee shall possess, use, or transfer licensed material in such a manner as to cause any individual in a restricted area to receive, in any one calendar quarter, from radioactive material and other sources of radiation, a total occupational dose equivalent in excess of 12.5 mSv (1.25 rem) to the whole body.

10 CFR 20.103(a)(1) requires, in part, that no licensee shall possess, use, or transfer licensed material in such a manner as to permit any individual in a restricted area to inhale a quantity of

)

radioactive material in any period of one calendar quarter greater than the quantity that would result from inhalation for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week for 13 weeks at uniform concentrations of radioactive material in air specified in Appendix B, Table I, Column 1 of this part.

10 CFR 72.24(e) requires the licensee to provide the means for controlling and limiting occupational radiation exposures within the limits given in 10 CFR Part 20 and for meeting the objective of maintaining exposures ALARA.

1 10 CFR 72.126(a) requires, in part, that radiation protection systems must be provided for all areas and operations where onsite personnel may be exposed to radiation or airborne radioactive materials.

Guidance is also provided in Regulatory Guide 8.8, and Regulatory Guide 8.10.

One area of this review focuses on those site-specific policy, design, and operational considerations associated with occupational radiation exposures ALARA. Another focuses ors the estimated quarterly occupational radiation exposure to any individual from radiation exposure and gaseous activity release. In the context of this review, quarterly occupational radiation exposures include: (1) cask handling operations, (2) cask surveillance and maintenance operations, and (3) exposure of an individual at the nearest boundary of the restricted area for a period of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> / week for 13 weeks (one calendar quarter). The nearest boundary of the restricted area is defined as that location on the ISFSI nuisance fence boundary where the greatest quarterly occupational radiation exposure occurs.

Radiation protection is deemed acceptable if it can be shown that the site-specific considerations for occupational radiation exposures ALARA are in compliance with appropriate guidance and/or 10-1

a regulations, and that the dose equivalent to any individual from quarterly occupational radiation exposure is not in excess of the applicable limits given in Section 10.1 above.

NSP will need to comply with the new 10 CFR Part 20 when it becomes effective, January 1, 1994. Under the new regulations ALARA is a requirement.

10.3 Review The review is divided into three main parts: (1) ensuring that occupational radiation exposures are ALARA, (2) radiation protection design features, and (3) estimated occupational radiation exposure assessment.

10.3.1 Ensuring That Occupational Radiation Exposures Are ALARA Site-specific policy, design, and operational considerations are addressed in Sections 7.1.1

(" Policy Considerations and Organization"), 7.1.2 (" Design Considerations"), and 7.1.3

(" Operational Considerations"), respectively, of the TSSAR.

Section 7.1.1 (" Policy Considerations and Organization") describes the policy and procedures that ensure that ALARA occupational exposures and contamination levels are achieved. Included among the policy and procedures are:

- The radiological protection program will be implemented at the ISFSI in accordance with the requirements of 10 CFR 72.126 and will be based on policies in existence at the Prairie Island Nuclear Generating Plant as described in Section 12.3 of the Prairie Island Updated Safety Analysis Report;

- The ALARA program follows the general guidelines of Regulatory Guides 1.8 (Reference 21), 8.8, and 8.10, and 10 CFR Part 20;

- Plant and design personnel are trained and updated on ALARA practices and dose-reduction techniques;

- Design and implementation of systems and equipment are reviewed to ensure ALARA criteria are met on all new and modification projects;

- The primary goal of the radiation protection and ALARA programs is to minimize exposure to radiation such that the total individual and collective exposures to personnel, in all phases of design, construction, operation, and maintenance, are kept ALARA;

- Trained personnel adequate to develop and conduct all necessary radiation protection and ALARA programs are provided;

- Training programs in the basics of radiation protection and exposure control are 10-2

provided to all facility personnel whose duties require working in radiation areas;

- The administrative organization is responsible for and has appropriate authority for ensuring that the basic objectives of the radiation protection program are achieved, and for maintaining occupational exposures as far below the specified limits as reasonably achievable;

- The plant manager is responsible for the protection of all persons and for compliance with NRC regulations and license conditions;

- The general superintendent plant engineering and radiation protection is responsible for the radiation protection program, including the program for handling and monitoring radioactive material, which is designed to ensure compliance with applicable regulations, technical specifications, and regulatory guides; 1

- The superintendent radiation protection is responsible for radiation safety; c hieving the goals of the radiation protection program and fulfilling the responsibilities for radiation protection, radiological monitoring, survey; and to ensure that personnel exposure control work is performed on a continuing basis for station operations and maintenance, including the ISFSI.

Section 7.1.2 (" Design Considerations") describes how the design considerations are ALARA._

Included among the design considerations are: the fuel will be stored dry, inside sealed, heavily-shielded casks; the casks will not be opened nor will fuel be removed from the casks while at the ISFSI; the exterior of the casks will be decontaminated before leaving the rail bay of the Prairie Island Nuclear Generating Plant Auxiliary Building; the casks will contain no active components that require periodic maintenance or surveillance; the ISFSI will be constructed before ISFSI operation; an instrumentation or annunciation panel monitoring cask pressure will be located outside of the ISFSI protected area; the location of the ISFSI will be of sufficient distance from frequently occupied areas of the Prairie Island Nuclear Generating Plant that the increased dose to personnel will not be significant; and, the objectives of Regulatory Guide 8.8, with regard to access control, radiation shielding, process instrumentation and controls, and decontamination are satisfied.

Section 7.1.3 (" Operational Considerations") describes how the operational considerations are ALARA. Included among the operational considerations are: the ALARA procedures for the ISFSI will be the same as those used in the radiation protection program for the Prairie Island Nuclear Generating Plant; storage of spent fuel in storage casks is expected to involve lower.

exposures than other altemative methods or designs for onsite storage; operational requirements for surveillance are incorporated into the design considerations in that the casks are stored with adequate spacing to allow ease of site surveillance; instramentation and annunciation will be available outside the ISFSI protected area to minimize surveillance time; operational-considerations are incorporated into the radiation protection design features in that the casks are heavily shielded to minimize occupational exposure; and since the ISFSI contains no systems that 10-3

process liquids or gases or contain, collect, store, or transport radioactive liquids or solids other than the stored spent fuel, it meets ALARA requirements.

l The staff evaluated the site-specific information that NSP provided, comparing it with the j

guidance and/or regulations cited in Section 10.1 of the SER.

10.3.2 Radiation Protection Design Features Installation design features are addressed in Sections 1.2 (" General Description of Iecation"),

1.3 (" General Storage System Description"), 2.1.1 (" Site Location"), 2.3.4 (" Diffusion Estimates"), 2.5.1.7 (" Plan and Profile Drawings"), 3.1.1 (" Spent Fuel to be Stored"),

3.2.5.2.10 (" Buried Cask"), 3.2.5.3.3 (" Internal Pressure"), 3.3.2 (" Protection by Multiple Confinement Barriers and Systems"), 3.3.3 (" Protection by Equipment and Instrumentation Selection"), 3.3.5 (" Radiological Protection"), 3.3.7.1 (" Spent Fuel Handling and Storage"),

3.3.7.2 (" Radioactive Waste Treatment"), 3.4 (" Summary of Storage Cask Design Criteria"),

4.1.1 ("Ixcation and Layout"), 4.2.1 (" Structures"), 4.2.2 (" Storage Site Layout"), 4.2.3

(" Storage Cask Description"),4.4.2 (" Decontamination Systems"),4.4.3 (" Storage Cask Repair and Maintenance"). 4.4.5 ("Other Systems"),4.6 (" Decommissioning Plan"),5.1.1 (" Narrative l

Description"), 5.1.2 (" Flow Sheets"), 5.1.3.2 (" Instrumentation"), 5.1.3.3 (" Maintenance i

Techniques"), 5.2 (" Control Room and Control Areas"), 6.1 (" Design"), 7.1.2 (" Design l

Considerations"), 7.1.3 (" Operational Considerations"), 7.2 (" Radiation Sources"), 7.3

(" Radiation Protection Design Features"),7.4 (" Estimated Onsite Collective Dose Assessment"),

7.6 (" Radiation Protection Program"),7.7 (" Radiological Environmental Monitoring Program"),

7A.1 (" Shielding Design Features"), 7A.2 (" Shielding Analyses"), 7A.3 (" Direct Radiation N-S"), 7A.4 (" Direct Radiation E-W"), 7A.5 (" Dose Rate Around the ISFSI"), and 8.1.2

(" Radiological Impact from Off-Normal Operations").

NSP also provided supplemental information in TSSAR Revisions 1 (Reference 22) and 2, dated April 1991 and September 1991, respectively, and its response to questions dated December 23,1991.

Sections 1.2 (" General Description of Iecation"),1.3 (" General Storage System Description"),

2.1.1 (" Site Location"), 2.5.1.7 (" Plan and Profile Drawings"), 4.1 (" Location and l2yout"),

4.2.1 (" Structures"),4.2.2 (" Storage Site Layout"),7.1.2 (" Design Considerations"), and 7.1.3

(" Operational Considerations") provide information specific to the ISFSI design features.

Included are descriptions of the proposed site location, the cask transporter access road, the ISFSI controlled area, the earthen berm, and the general arrangement of the ISFSl; drawings of the site grading plan with section and details; and dimensions and design criteria for the ISFSI concrete pads; and how the design and operational considerations of the ISFSI are ALAPA.-

Sections 1.3 (" General Storage System Description"), 3.1.1 (" Spent Fuel to be Stored"),

3.2.5.2.10 (" Buried Cask"), 3.2.5.3.3 (" Internal Pressure"), 3.3.2 (" Protection by Multiple Confinement Barriers and Systems"), 3.3.3 (" Protection by Equipment and Instrumentation Selection"), 3.3.7.1 (" Spent Fuel Handling and Storage"), 3.4 (" Summary of Storage Cask Design Criteria"), 4.2.3 (" Storage Cask Description"), 4.4.5 ("Other Systems"), 5.1.3.2

(" Instrumentation"),7.1.2 (" Design Considerations"),7.1.3 (" Operational Considerations"),7.2 10-4

(" Radiation Sources"), 7A.1 (" Shielding Design Features"), provide information basic to the principal design features of the cask. Included are descriptions of the dimensions, materials, and design characteristics of the TN-40 cask components: basket assembly, confinement vessel, confinement barriers and seals, gamma shield, neutron shield, outer shell, torispherical protective cover, pressure monitoring system, and trunnions; the design basis fuel assemblies and associated radiological source term characteristics; and how the design and operational considerations of the cask are ALARA.

Sections 2.3.4 (" Diffusion Estimates"), 3.3.5 (" Radiological Protection"), 7.3 (" Radiation Protection Design Features"),5.2 (" Control Room and Control Areas"),7.4 (" Estimated Onsite Collective Dose Assessment'), 7.6 (" Radiation Protection Program"), 7.7 (" Radiological Environmental Monitoring Program"), 7A.2 (" Shielding Analyses"), 7A.3 (" Direct Radiation N-S"), 7A.4 (" Direct Radiation E-W"), 7A.5 (" Dose Rate around the ISFSI"), and 8.1.2

("RadiologicalImpact from Off-Normal Operations") provide information specific to radiological protection. Included are short-term dispersion factors (X/Q) at the site boundary and at varying

]

downwind distances; descriptions of the ISFSI access control; the location of the alarm panel for the cask pressure monitoring system and the procedures to be followed if an alarm is actuated; the occupational collective dose equivalent (man-rem) associated with ISFSI operation to station j

personnel and with various cask loading, transport, emplacement, surveillance, and maintenance j

activities; area radiation and airborne radioactivity monitoring instrumentation requirements; the Prairie Island Nuclear Generating Plant management position responsible for radiation protection activities at the ISFSI and the bases for the radiation protection practices and procedures; the use

.i of (TLDs) for environmental monitoring of the ISFSI; the radiological source terms, models, materials, analyses and results for the total dose equivalent rates at various locations at and near the surface of a single cask and at the centerlines of the west and east pads in the north-south I

(N-S) and east-west (E-W) directions from ISFSI operations; an assumed cask loading sequence for the ISFSI; and the radiological impact from off-normal events.

j j

Sections 3.3.7.2 (" Radioactive Waste Treatment"),4.4.2 (" Decontamination Systems"), and 6.1

(" Design") provide information specific to waste generation and treatment. Included are descriptions of radioactive waste generation and treatment associated with the ISFSI and the cask loading and decontamination operations that occur within the Prairie Island Nuclear Generating Plant Auxiliary Building.

Sections 4.4.3 (" Storage Cask Repair and Maintenance") and 5.1.3.3 (" Maintenance Techniques") provided information specific to cask repair and maintenance. Included are the identification of typical maintenance tasks such as occasional replacement or calibration of' monitoring instrumentation and recoating of some casks with corrosion-inhibiting coatings.

Section 4.6 (" Decommissioning Plan") provides information specific to decommissioning of the casks and storage pads. Included are descriptions of the decontamination of the interior and surface of the cask; the neutron activation of the cask body, spent fuel basket, and concrete pad materials over the storage period; and the disposal of the cask bodies and spent fuel baskets.

10-5 i

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Sections 5.1.1 (" Narrative Description") and 5.1.2 (" Flow Sheets") provide information specific to handling operations in the cask loading, decontamination, and storage areas. Included are descriptions of the sequence of operations to be performed and the number of personnel, time,

+

and distance from the cask associated with the various cask loading, decontamination, and storage operations.

10.3.3 Estima'ed Occupational Radiation Exposure Assessment f

This review is divided into three main parts: (1) estimated quarterly dose equivalent from cask handling operations, (2) estimated quarterly dose equivalent from cask surveillance and maintenance operations, and (3) estimated quarterly dose equivalent at the nearest boundary of the restricted area.

10.3.3.1 Estimated Quarterly Dose Equivalent from Cask Handling Operations In the TSSAR, NSP does not specifically address the quarterly occupational dose equivalent from cask-handling operations.

The occupational dose equivalent from TN-40 cask handling operations is addressed in Sections 4.4.2 (" Decontamination Systems"), 5.1.1 (" Narrative Description"), 5.1.2 (" Flow Sheets"), 7.4

(" Estimated On-Site Collective Dose Assessment"), and 7A.2 (" Shielding Analyses") of the TSSAR. NSP has also provided supplemental information in Revision 2, dated September 1991, and its response to questions dated December 23,1991.

Section 4.4.2 (" Decontamination Systems") describes decontamination methods that will be used to remove surface contamination, from the casks, resulting from their submersion in the spent fuel pool and specifies the location of the decontamination area as the rail bay of the Prairie Island Nuclear Generating Plant Auxiliary Building.

Section 5.1.1 (" Narrative Description") describes the handling operations in the cask loading and cask storage areas. The cask is moved from the rail bay of the auxiliary building to the surface of the fuel storage pool, for loading. The cask confinement lid is removed and the cask is lowered into the spent fuel pool. Fuel assemblies are loaded into the cask by an operator standing on the movable bridge over the pool using a long handled tool suspended from the spent fuel pool bridge crane hoist. Once the fuel is loaded and the lid in place, the cask is lifted to the pool surface and the lid bolts installed. The internal cask cavity is drained by displacing the t

water with air or with a suitable drain pump. The cask is then returned to the auxiliary building rail bay. The cask is decontaminated in the cask decontamination area. The cavity is dried using a vacuum drying system. Once the cavity is dry, it is backfilled with dry helium and the lid seals are leak checked. The top neutron shield is then installed. Once it is secured, the overpressure system is pressurized with helium and the interspace seals are leak-checked. Before transfer from the auxiliary building to the ISFSI, the cask is monitored for contamination, temperature, radiation dose equivalent rates, and proper functioning of the pressure-monitoring system. The torispherical protective cover is installed and the pressure transducer wires fed through the 10-6 1

i l

external fitting. The cask is then towed to the ISFSI site with a transport vehicle. The cask is set in its. storage position, connected to the cask pressure monitoring system, and a functional check of the monitoring system performed.

Section 5.1.2 (" Flow Sheets") provides the sequence of operations to be performed in the cask loading, decontamination, and storage areas and tabulates for each operation the number of persons and time required and the average distance between the individual and the cask.

Section 7.4 (" Estimated On-Site Collective Dose Assessment") provides the estimated direct radiation dose equivalent rates and occupational collective dose equivalents at beginning oflife (10 years since discharge from the reactor) associated with loading, transport, and emplacement of the storage casks in Table 7.4-1 (" Design Basis Occupational Exposures of Cask leading, Transport, and Emplacement (One-Time Exposure)") of the TSSAR. Direct radiation dose equivalent rates at 1 meter (1 yd) are assumed for all cask-handling operations except cask tramsfer, where individuals are assumed to be at least 2 meters (2 yds) away from the cask.

Section 7A.2 (" Shielding Analyses") provides the direct radiation gamma, neutron, and total dose equivalent rates at various locations on the surface and at 1, 2, and 3-meter (yd) distances from the surface of a single cask in Table 7A-4 ("TN-40 Dose Rates at Short Distances") of j

the TSSAR.

l The staff evaluation of the quarterly occupational dose equivalent from cask handling operations consists of the quarterly occupational dose equivalent from a single cask to any individual from direct radiation during cask loading, decontamination, and storage operations. Specific operations considered are those grouped under placement in pool; loading process; removal from pool; transfer to decontamination area; processing of cask; helium leak test; decontamination; installation neutron shield, pressurize, and test; preparation for transport; transfer of cask to ISFSI; and final cask emplacement. The processing of one cask is assumed during the calendar quarter. Gaseous activity release is not considered credible under normal storage conditions and thus is not evaluated.

The staff evaluation of the quarterly occupational dose equivalent from direct radiation during cask-handling operations is based on the times, distances, and dose equivalent rate determinations of the TSSAR or staff. Times and distances required to perform the various operations are provided in Table 5.1-2 (" Anticipated Time and Personnel Requirements for Cask-Handling Operations") of the TSS AR. Gamma and neutron dose equivalent rates associated with specific distances or operations are provided in Tables 7A-4 ("TN-40 Dose Rates at Short Distances") and 7.4-1 (" Design Basis Occupational Exposures for Cask Loading, Transport, and Emplacement (One Time Exposure)") of the TSSAR. From Section 5.4 (" Findings and Conclusions") of this SFR, NSP gamma dose equivalent rates are conservative; neutron dose equivalent rates are not always conservative. To ensure conservatism in the neutron dose equivaient, staff calculated neutron dose equivalent rates are substituted for the Table 7A-4 entries, where appropriate. To add additional conservatism, beginning of life (10 years since discharge from the reactor) is assumed for all operations.

10-7 i

j

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i t

10.3.3.2 Estirnated Quarterly Dose Equivalent from Cask Surveillance and Maintenance Operations NSP does not specifically address the quarterly occupational dose equivalent from cask surveillance and maintenance operations, in the TSSAR.

l The occupational dose equivalent from TN-40 cask surveillance and maintenance operations is addressed in Sections 4.4.3 (" Storage Cask Repair and Maintenance"), 5.1.2 (" Flow Sheets"),

5.1.3.3 (" Maintenance Techniques"),7.4 (" Estimated On-site Collective Dose Assessment"), and 7A.2 (' Shielding Analyses") of the TSSAR. NSP also provided supplemental information in Revision 2, dated September 1991, and its response to questions dated December 23,1991.

Section 4.4.3 (" Storage Cask Repair and Maintenance") identifies TSSAR Section 5.1.3.3 i

(Maintenance Techniques) as the location of information on cask maintenance.

Section 5.1.2 (" Flow Sheets") identifies typical surveillance and maintenance operations and tabulates for each operation the-number of persons and time required and the average distance between the individual and the cask.

Section 5.1.3.3 (" Maintenance Techniques") identifies typical maintenance tasks such as occasional replacement or calibration of monitoring instrumentation and recoating of some casks with corrosion-inhibiting coatings.

Section 7.4 (" Estimated On-site Collective Dose Assessment") provides the estimated dose equivalent rates and occupational collective dose equivalents at beginning oflife (10 years since discharge from the reactor) associated with annual surveillance and maintenance activities in Table 7.4-2 (" Design Basis ISFSI Maintenance Operations, Annual Exposures") of the TSSAR.

Visual surveillance is based on a walkdown of each of the two pads at a distance no closer than 2 meters 2 yds to the casks; the average dose equivalent rate is assumed to be 4 times the calculated dose equivalent rate at 2 meters (2 yds). Instrumentation operability tests and calibration assume the worker to be located at the monitoring panel at the perimeter fence entrance. For instrument or cask surface defect repairs, the worker is assumed to be positioned between two rows of casks with the average dose equivalent rate conservatively assumed to be 6 times the calculated dose equivalent rate at 2 meters (2 yds). For the major maintenance activity (lid seals replacement), the dose equivalent rate is that at the surface of the torispherical protective cover.

Section 7A.2 (" Shielding Analyses") provides the direct radiation gamma, neutron, and total dose equivalent rates at various locations on and near the surface of a single cask; an assumed loading sequence for the ISFSI; illustrates the direct radiation gamma and neutron dose equivalent rates as a function of distance; and illustrates the direct radiation relative gamma and neutron dose equivalent rate factors as a function of time after discharge from the reactor, normalized to 10 years. Direct radiation gamma, neutron, and total dose equivalent rates at various locations on the surface and at I and 2 meter distances from the surface of a single cask 10-8

are summarized in Table 7A-4 ("TN-40 Dose Rates at Short Distances") of the TSSAR. The west pad (first) loading sequence for the ISFSIis assumed to be eight casks to the middle of the pad at the start of the first year and two casks each year thereafter, alternating ends. The east pad (second) loading sequence for the ISFSIis assumed to be two casks to the middle of the pad at the start of the tenth year and two casks each year thereafter, altemating ends. The entire two i

pad loading sequence is assumed to require 21 years to complete. Direct radiation gamma and neutron dose equivalent rates as a function of distance are illustrated in Figure 7A-6 (Dose Rates at Long Distances (Mrem /Hr)). Direct radiation relative gamma and neutron dose equivalent rate factors as a function' of time after discharge from the reactor, normalized to 10 years are illustrated in Figure 7A-7 (" Relative Dose Rate Factor (Normalized to 10 Years)") of the TSSAR.

The staff evaluation of the quarterly occupational dose equivalent from cask surveillance and maintenance operations, consists of the quarterly occupational dose equivalent from a single cask to any individual from direct radiation during surveillance and. maintenance operations. Specific operations considered are those grouped under visual surveillance of casks; instrumentation operability tests, calibration, and repairs; surface defects repair; and major maintenance. One I

visual survey; one instrumentation operability test, calibration, and repair; one surface defects repair; and one major maintenance are assumed during the calendar quarter. Gaseous activity release is not considered credible under normal storage conditions and thus is not evaluated.

The staff evaluation of the quarterly occupational dose equivalent from direct radiation during cask surveillance and maintenance operations is based on the times, distances, and dose equivalent rate determinations of the TSSAR or staff. Times and distances required to perform the various operations are taken from Table 5.1-2 (Anticipated Time and Personnel requirements for Cask Handling Operations) of the TSSAR. Gamma and neutron dose equivalent rates associated with specific distances or operations are provided in Tables 7A-4 ("TN-40 Dose Rates at Short Distances") and 7.4-1 (" Design Basis Occupational Exposures for Cask Loading, Transport, and Emplacement (One Time Exposure)") of the TSSAR. From Section 5.4

(" Findings and Conclusions") of this SER, NSP gamma dose equivalent rates are conservative; neutron dose equivalent rates are not always. To ensure conservatism in the neutron dose equivalent, staff calculated neutron dose equivalent rates are substituted for the Table 7A-4 entries, where appropriate. To add additional conservatism, beginning of life (10 years since discharge from the reactor) is assumed for all operations.

10.3.3.3 Estimated Quarterly Dose Equivalent at the Nearest Boundary of the Restricted Area Sections 7.4 (" Estimated On-site Collective Dose Assessment"), 7A.2 (" Shielding Analyses"),

7A.3 (" Direct Radiation N-S"), and 7A.4 (" Direct Radiation E-W"), supplemented by NSP Revision 2, dated September 1991, and its response to questions dated December 23, 1991, provide dose equivalent rates from direct and indirect radiation, during normal operations, from which a quarterly occupational dose equivalent may be estimated.

10-9

e Section 7.4 (" Estimated On-site Collective Dose Assessment") provides the estimated dose equivalent rates and occupational collective dose equivalents associated with ISFSI operations to station personnel in Tables 7.4-4 (" Dose Rates at Oasite Locations Due to Cask Storage") and 7.4-6 (" Annual Collective Exposure Estimates to Onsite Personnel") of the TSSAR.

Section 7A.2 (" Shielding Analyses") provides an assumed loading sequence for the ISFSI; illustrates the direct radiation gamma and neutron dose equivalent rates as a function of distance; and illustrates the direct radiation relative gamma and neutron dose equivalent rate factors as a function of time after discharge from the reactor, normalized to 10 years. The west pad (first) loading sequence for the ISFSIis assumed to be eight casks to the middle of the pad at the start of the first year and two casks each year thereafter, alternating ends. The east pad (second) loading sequence for the ISFSIis assumed to be two casks to the middle of the pad at the start of the tenth year and two casks each year thereafter, alternating ends. The entire two pad loading 1

sequence is assumed to require 21 years to complete. Direct radiation gamma and neutron dose equivalent rates as a function of distance are illustrated in Figure 7A-6 (" Dose Rates at I.ong Distances (Mrem /Hr)"). Direct radiation relative gamma and neutron dose equivalent rate factors as a function of time after discharge from the reactor, normalized to 10 years are il!ustrated in Figure 7A-7 (Relative Dose Rate Factor (Normalized to 10 Years)) of the TSSAR.

Section 7A.3 (" Direct Radiation N-S") provides the dose equivalent rates from direct radiation as a function of distance along the centerline of the west and east pads in the north-south (N-S) direction from an array of 24 and 48 casks at 9 and 21 years, respectively in Table 7A-6 (" Total Direct Dose Rate (TN-40 Cask) (Mrem /Hr)") of the TSSAR. Attenuation of the direct radiation by the earthen berm is not included.

Section 7A.4 (" Direct Radiation E-W") provides the dose equivalent rates from direct and indirect radiation and the attenuation of the direct radiation by the earthen berm. Dose equivalent rates from direct radiation, as a function of distance along the centerline of the west and east pads in the east-west (E-W) direction, from an array of 24 and 48 casks at 9 and 21 years, respectively, are found in Table 7A-6 (" Total Direct Dose Rate (TN-40 Cask) (Mrem /Hr)) of the TSSAR. Dose equivalent rates, from indirect radiation as a function of distance from an array of 24 and 48 casks at 9 and 21 years, respectively are found in Table 7A-7 (Total Skyshine Dose Rate (TN-40 Cask) (mrem /hr)) of the TSSAR. Attenuation of the direct radiation by the earthen berm as a function of thickness is summarized in Table 7A-8 (" Attenuation of the Earth Berm") of the TSSAR. However, it is not included in the dose equivalent rate calculations.

The staff evaluation of the quarterly occupational dose equivalent at the nearest boundary of the restricted area consists of the quarterly occupational dose equivalent from an array of 48 casks at 21 years to any individual at the nearest boundary of the restricted area from direct and indirect radiation. The nearest boundary of the restricted area is defined as that location on the ISFSI nuisance fence boundary where the greatest quarterly occupational radiation exposure occurs._ In the north-south (N-S) and east-west (E-W) directions, the minimum distances between the nuisance fence and the closest cask surface are 38.0 meters (124.8 ft) and 22.8 meters (74.8 ft.), respectively. Gaseous activity release is not considered credible under normal storage 10-10

conditions and thus is not evaluated.

The staff evaluation of the quarterly occupational dose equivalent from direct and indirect radiation at the nearest boundary of the restricted area is computed from the dose equivalent rates in the TSSAR. Quarterly occupational dose equivalents, from direct gamma and neutron radiation, are derived from the product of the dose equivalent rates in TSSAR Table 7A-6

(" Total Direct Dose Rate (TN-40 Cask) (mrem /hr)") and staff estimated distance-dependent correction factors. Quarterly occupational dose equivalents from indirect gamma and neutron radiation are derived from the product of the dose equivalent rates in TSSAR Table 7A-7 (" Total Skyshine Dose Rate (TN-40 Casks) (mrem /hr)") and staff estimated distance-dependent correction factors. From Section 5.4 (" Findings and Conclusions") of this SER, NSP gamma dose equivalent rates are conservative; neutron dose equivalent rates are not always conservative.

To ensure conservatism in the neutron dose equivalent, Table 7A-6 and 7A-7 entries are corrected by the ratio of the staff and NSP calculated neutron dose equivalent rates at 3 meters (3 yds) from the radial neutron shield shell and 2 meters (2 yds) above the cask torispherical protective cover, respectively. Assuming an exposure time of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> / week for 13 weeks (one calendar quarter), dose equivalents are computed and compared to determine the location on the ISFSI nuisance fence boundary where the greatest quarterly occupational radiation exposure occurs.

10.4 Findings and Conclusions l

Site-specific policy, design, and operational considerations are in compliance with appropriate guidance and/or regulations.

The quarterly occupational dose equivalent to any individual from direct radiation is estimated i

to be less than 7.5 mSv (0.750 rem) to the whole body and to the thyroid from cask-handling operations. Gaseous activity release is not considered credible from cask-handling operations.

The quarterly occupational dose equivalent to any individual from direct radiation is estimated to be less than 12.4 mSv (1.24 rem) to the whole body and to the thymd from cask surveillance and maintenance operations. Gaseous activity release is not considered credible from cask surveillance and maintenance operations.

1 The quarterly occupational dose equivalent, to any individual at the nearest boundary of the restricted area (E-W nuisance fence), from direct and indirect radiation, from an array of 48 casks, at 21 years, is estimated to be less than 8.73 mSv (0.873 rem) to the whole body and to the thyroid. Gaseous activity release is not considered credible under normal storage conditions.

The occupational dose equivalent to any individual from direct radiation is in compliance with the 10 CFR 20.101(a) and 20.103(a)(1) limits. Gaseous activity release is not considered credible under normal conditions of storage.

Radiation protection is in compliance with appropriate guidance and/or regulations.

10-11

11.

ACCIDENT ANALYSIS 11.1 Area of Review Cask safety in the event of postulated off-normal and accident events is deemed acceptable if it can be shown that the dose from a single cask to any individual from direct radiation and activity release is not in excess of the applicable values given below.

11.2 Acceptance Criteria 10 CFR 72.24(m) requires, in part, an analysis of the potential dose equivalent or committed dose equivalent to an individual outside the controlled area from accidents or natural phenomena events that result in the release of radioactive material to the environment or direct radiation l

from the ISFSI.

10 CFR 72.106(b) requires that any individual located on or near the nearest boundary of the controlled area shall not receive a dose greater than 50 mSv (5 rem) to the whole body or any

{

organ from any design basis accident. The minimum distance required to the nearest boundary of the controlled area is 100 meters (328 feet).

Our review focuses on the dose equivalent from a single cask to any individual from direct and indirect radiation and gaseous activity release after postulated accident events. Both the minimum distance required to the nearest boundary of the controlled area (100 meters) (110 yds) and the nearest site boundary distance assumed in the TSSAR (110 meters) (120 yds) due west from the nearest cask) are used for the evaluation.

Cask safety is deemed acceptable ifit can be shown that the dose equivalent from a single cask to any individual from postulated accident events is not in excess of the applicable limits.

]

11.3 Review

-I Section 8.2 (" Accidents") of the TSSAR describes the types of accident events, the causes of the accident events, the analyses of the accident events, and the accident dose calculations. NSP has also provided supplemental information in Revisions 1 and 2, dated April 1991, and September 1991, respectively, and its response to questions dated December 23,1991. Accidents are of two types: infrequent events that could reasonably be expected to occur during the lifetime of the ISFSI and events that are postulated because their consequences may result in the maximum potential impact on the immediate environs. Among the accident events are earthquake, extreme wind, flood, explosion, fire, inadvertent loading of a newly discharged fuel assembly, cask seal leakage, hypothetical cask drop and tipping accidents, and loss of confinement barrier. Those accident events postulated to have radiological consequences are: hypothetical cask drop and tipping accidents and loss of confinement barrier. The radiological impact of the hypothetical cask drop and tipping accidents is an increase in dose equivalent associated with the assumed loss of the radial neutron shield and shell. Gamma, neutron, and total dose equivalent rates at 11-1

t 2 meters (2 yds) from the side of a single cask without the radial neutron shield and shell are summarized in Table 7A-4 ("TN-40 Dose Rates at Short Distances") of the TSSAR. Dose equivalent rates for the loss of the radial neutron shield and shell accident at distances other than 2 meters (2 yds) are not provided. The radiological impact of the loss of confinement barrier is limited to an increase in dose equivalent associated with the assumed instantaneous release of the cask cavity gas inventory. Total cask cavity activity (MBq) (Ci) inventories a e summarized for 'H, "Kr, *Cs, and mCs in Table 7.2-7 (" Fission Gas and Volatile Nuclides Inventory (Curies /40 Assemblies), Westinghouse OFA 14x14 3.85 w/o:"U,45,000 MWD /MTU,10 Year Cooling Time") of the TSSAR. Only release of the cask cavity "Kr is considered. Release of the cask cavity H is stated to be covered by the "Kr evaluation and the cesium nelides are stated to remain in the fuel assemblies in particulate form. NSP takes no credit for "Kr decay or for personnel protection. Dose equivalent at the site boundary is estimated using the short-term dispersion factors (X/Q) summarized in Table 2.3-1 (" Site Boundary Dispersion Factor (X/Q)") of the TSSAR and the methodology and equations of Regulatory Guide 1.3 (Reference

23) and is summarized in Table 8.2-1 (" Radiological Consequences - Loss of Confinement Barrier") of the TSSAR.

The staff evaluation of the dose equivalent from postulated accident events considers both the loss of the radial neutron shield and shell and instantaneous release of the cask cavity gas inventory. Dose equivalent from the loss of the radial neutron shield and shell and instantaneous release of the cask cavity gas inventory are computed for an individual at 100 meters (328 feet) and 180 meters (591 feet) from a single cask.

The staff evaluation of the dose equivalent associated with the loss of the radial neutron shield f

and shell is computed from the direct and indirect radiation dose equivalent ratu Jeterminations of the TSSAR or staff. Staff direct gamma and neutron dose equivalents are derived from the product of the dose equivalent rates in Figure 7A-6 (" Dose Rates at Long Distances (mrem /hr)")

and staff estimated dose-independent and distance-independent correction factors. Staff indirect gamma and neutron dose equivalents are assumed to be one twenty-fourth of the product of the dose equivalent rates from an array of 24 casks at 9 years in Table 7A-7 (" Total Skyshine Dose Rate (TN-40 Casks) (mrem /hr)") of the TSSAR and the aforementioned staff estimated dose-independent correction factor. Appropriate gamma and neutron dose equivalent rates at the nearest site boundary (110 meters) (120 yds) are predicted through exponential interpolation of the Table 7A-7 data. Dose-independent correction factors are computed for gamma and neutron radiation from the ratio of the direct radiation radial dose equivalent rates at 2 meters (2 yds) without and with the radial neutron shield and shell. For conservatism, direct radiation dose i

equivalent rates required for these calculations are those of the staff for the gamma radiation and those of Table 7A-4 ("TN-40 Dose Rates at Short Distances") of the TSSAR, for the neutron radiation. The nearest boundary to the controlled area (100 meters) (110 yds) is conservatively

[

assumed to be due east. As such, there is no attenuation of the direct gamma and neutron dose radiation by the earthen berm. With the nearest site boundary (110 meters) (120 yds) established i

as due west, attenuation of the direct gamma and neutron radiation by the earthen berm is included. The attenuation factors assumed are those for a metric equivalent 1.8 meter (6-ft) thick berm in Table 7A-8 (" Attenuation Factor for Earth Berm") of the TSSAR. In the case of the 11-2 l

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1 indirect gamma and neutron radiation, no corrections for earth berm attenuation are required.

To assure additional conservatism, it is assumed that no corrective actions are taken for a period of 1 year, even though action would be taken quicker.

The staff evaluation of the dose equivalent associated with instantaneous release of the cask e

cavity inventory represents upper bound consequences for all available 'H and 35Kr gaseous activity. In computing the dose equivalent because of 3H Pr.J "Kr gaseous activity release, the staff has assumed the following: (1) the initial cask activity inventory in Table 7.2-7 (" Fission

- Gas and Volatile Nuclides Inventory (Curies /40 Asser0blies), Westinghouse OFA 14x14 3.85 w/o 235U,45,000 MWD /MTU,10 Year Cooling Time") of the TSSAR; (2) 100 percent cladding tube failure; (3) the release fractions of Regulatory Guide 1.25 (Reference 24) (0.10 for 'H and 0.30 for 85 85 Kr); (4) a shielding factor of 1.0 for Kr; (5) the occupational inhalation rate of Regulatory Guide 1.25 for the onsite individual (3.47x10 meter /sec)(.00ll ft/s) (6) the d

population weighted inhalation rate of Regulatory Guide 1.109 (Reference 25) for the off-site individual; (7) the whole body and thyroid inhalation dose equivalent and total body dose equivalent factors of Regulatory Guide 1.109; and (8) F-stability atmospheric diffusion with a wind speed of 1 meter /sec (3.3 ft/sec) with plume meander.

11.4 Findings and Conclusions The dose equivalent from a single cask, to any individual, from direct and indirect radiation, after postulated accident events, is less than 14.5m Sv (1.45 rem) to the whole body and to the thyroid at 100 meters (110 yds) due east (nearest boundary of the controlled area) and less than

.16 mSv (0.016 rem) to the whole body and to the thyroid at 110 meters (120 yds) due west (nearest site boundary).

4 The dose equivalent from a single cask, to any individual from gaseous activity release after postulated accident events, is less than 5.18m Sv (0.518 rem) to the whole body and to the thyroid at 100 meters (110 yds) due east (nearest boundary of the controlled area) and less than 4.3 mSv (0.430 rem) to the whole body and to the thyroid at 110 meters (120 yds) due west (nearest site boundary).

Dose equivalent consequences, from a single cask, to any individual, from direct and indirect radiation and gaseous activity release after postulated accident events, are less than the 50 mSv (5 rem) limit established in 10 CFR 72.106(b).

i 11-3

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• 1 12.

DECOAGilSSIONING 12.1 Area of Review Cask decommissioning is deemed acceptable if it can be shown that the applicable regulations l

have been followed, as appropriate. In addition, where limits can be applied, these limits have not been exceeded.

12.2 Acceptance Criteria 5

10 CFR 72.30 provides requirements for a site-specific decommissioning plan, including financing. Among the items to be addressed under this part are the decontamination of the site and facilities, the disposal of residual radioactive materials after all spent fuel has been removed, the decommissioning funding plan, and the cost estimate for decommissioning.

10 CFR 72.130 provides requirements for decommissioning and states, in part, that the ISFSI shall be designed for decommissioning. Among the items to be addressed under this part are the provisions to facilitate decontamination of equipment, the provisions to minimize the quantity of radioactive wastes and contaminated equipment, and the provisions to facilitate the removal of radioactive wastes and materials at the time of permanent decommissioning.

49 CFR 173.421,173.423, and 173.435 provide information on the radionuclide activities that may be transported as limited quantity materials.

10 CFR 30.14 and 30.70 address radionuclide concentrations that are exempt from licensing requirements.

10 CFR 30.18 and 30.71 address radionuclide quantities that are exempt from licensing requirements, i

10 CFR 61.55 and 61.56 address the radionuclide concentrations for Class A wastes and the

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characteristics of such waste.

i 12.3 Review

-l The review is divided into four main parts: (1) unloading of the cask, (2) decommissioning of the cask components, (3) decommissioning of the storage pad, and (4) decommissioning funding.

12.3.1 Unloading of the Cask t

The possibility for cask unloading is mentioned in Section 4.6 (" Decommissioning Plan") of the i

TSSAR. NSP has provided supplemental information, on the wet unloading (in the spent fuel pool' procedure, in a response to questions, dated July 10,1992 (Reference 26).

i 12-1 i

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Assuming a normal spent fuel pool transfer, the unloading sequence is essentially a reverse of the loading sequence. The cask. is returned from the ISFSI to the auxiliary building rail bay via the transporter. The weather cover is unbolted and removed. The overpressure system is then removed and the cavity gas sampled through the vent port. Once the cask is moved into the fuel pool area, the cavity is depressurized and the cask lowered into the spent fuel pool. With the cask lid at the pool surface, fill and drain lines are connected to the lid drain and vent ports.

t Borated water is slowly added to fill the cask and gradually cool the fuel in the cask. When the cask is full, the fill and drain lines are removed. The cask is then lowered to the pool bottom, where the lid is removed, making the fuel accessible for transfer.

Decontamination of the cask is addressed in Section 4.4.2 of the TSSAR. The text of this section states that " Standard decontamination methods will be used to remove surface contamination from the casks." It is further stated that " Manual methods will be employed, i

using water detergents and wiping cloths." Decontamination of the cask is addressed in Section 4.6 of the TSSAR. The text of this section mentions surface decontamination using chemical etching and/or electropolishing. These methods are acceptable to the staff, provided that detailed procedures are developed before decontamination of a cask.

12.3.2 Decommissioning of the Cask Components Decommissioning of the neutron activated fuel basket, cask body and lid, neutron shield, and

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neutron shield shell and protective cover is addressed in Section 4.6 (" Decommissioning Plan *;

of the TSSAR. Supplemental information is also provided by NSP in responses to questions dated June 5,1991, and December 23, 1991.

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Neutron fluxes obtained from the XSDRNPM shielding calculations are used by TN in conjunction with ORIGEN2, to calculate the activation products at 30 days subsequent to l

unloading. Only those nuclides with activity greater than.37 MBq (10r Ci) and those nuclides s

listed in 10 CFR 61.55 are reported in the TSSAR. Table 4.6-2 ("Results of ORIGEN2 Activation Calculations") summarizes the activities for Cr, "Mn, "Fe, "Fe, "Co, "Co, "Ni, l

and "Ni in the fuel basket; C, SCr, "Mn, "Fe, "Fe, "Co, "Co, "Ni, and SNi in the cask body I

and lid; 'H, C, and "Zn in the neutron shield; and C and "Fe in the neutron shield shell and protective cover. Some of these nuclides emit no gamma rays ('H, C, and "Ni). Others ("Fe and "Ni) emit gamma rays with energies ofless than 8 kev.

l Materials quantities used by TN in the activation calculations are representative of those in the l

cask itself. Weights for the fuel basket, cask body and lid, neutron shield, and neutron shield shell and protective cover are summarized in Table 4.6-1 (" Data for TN-40 Activation Analysis") of the TSSAR and are 5676 kg (12,513 lb), 69,465 kg (153,144 lb),5766 kg (12,712 lb), and 4131 kg (9107 lb), respectively; volumes for the fuel basket, cask body and lid, neutron l

shield, and neutron shield shell and protective cover are summarized in Table.4.6-3 l

(" Comparison of TN-40 Activity with Class A Waste Limits") of the TSSAR and are 1.43 m' L

(50.5 ft'), 9.27 m' (328 ft'), 3.37 m' (119 ft'), and 0.53 m' (18.7 ft'), respectively.

I i

12-2

t In evaluating the activation products, the staff has assumed the activities, weights, and volumes presented in TSSAR Tables 4.6-2,4.6-1, and 4.6-3, respectively. Thirty days after the removal of the fuel assemblies, the following inventories remain in the principal cask components.

INVENTORY (MBq)

Fuel Cask Body Outer Shell and Isotope Basket and Lid Neutron Shield Protective Cover

'H 7.84x104 "C

7.47x 10-'

l.89x10-8 1.29x 10*

5'Cr 141 23.5 "Mn 17.1 309 55Fe 187 3360

.581 "Fe 3.46 62.5 "Co 21.8 22.2 "Co

.308

.307 "Ni

.117

.I19

Ni 13.5 13.8 65Zn

.414 12.3.3 Deconunissioning of the Storage Pad Decommissioning of the neutron activated storage pad and underlying media is addressed in NSP responses to questions dated June 5,1991, and December 23,1991. Calculations of the neutron activation of the storage pad and underlying media are not performed. Neutron fluxes at the bottom of the cask are estimated to be approximately an order of magnitude less than that used in the cask body activation calculations. With the cask activation product activities as a basis, NSP infers that activation of the storage pad and underlying media is less than that of the cask and, therefore, below the allowable limits for Class A waste. SpeciGc procedures for disposal of the storage pad remnants and underlying media are not provided.

12.3.4 Decommissioning Funding Decommissioning funding is addresse< in Section 10.1 (" Financial Qualification") of the NSP

" Application for a License to Construct and Operate a Dry Cask ISFSI." Decommissioning.

costs are currently estimated to be $3,100,000 and will be added to those for Prairie Island Unit l

1. Financial assurance for ISFSI decommissioning costs is provided by the external sinking fund, with monthly deposits, established for plant decommissioning. Beginning in 1993, the site-specific cost estimate will be revised on a 3-year basis by performing a new analysis. The current costs will be adjusted by inflation, to determine the final cost, before the recovery pattern is determined for the external sinking fund.

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12.4 Findings and Conclusions Cask ~ unloading information is consistent with the regulatory requirements'.

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Activation product concentrations associated with decommissioning of the cask components are such that the cask components contain license-exempt concentrations of 5H, "C, s'Cr, "Mn, 55Fe, "Fe,5"Co, "Co, and "Zn. Furthermore, the activities or concentrations are such that the cask components may be classed as limited quantity materials for offsite transportation and may be disposed of as Class A waste.

Decommissioning funding information is consistent with the regulatory requirements.

The cask design is consistent with the requirements of 10 CFR' 72.130 that an ISFSI be' designed I

for decommissioning. Furthermore, the actions involved in cask disposal and decommissioning funding are consistent with the requirements of 10 CFR 72.30 as feasible elements of a site-specific decommissioning plan.

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13 OPERATING CONTROLS AND LIMITS 13.1 Area of Review Consistent with 10 CFR 72.44(c), the operating controls and limits established in Chapter 10 of the TSSAR will be reviewed to determine if they cover, for the cask, all required safety limits, limiting conditions for operation surveillance requirements, and design features.

i 13.2 Acceptance Criteria Each license issued under 10 CFR Part 72 shall include license conditions pursuant to 10 CFR 3

72.44. In addition to these conditions each application for a license shall include proposed technical specifications pursuant to 10 CFR 72.26 and consistent with 10 CFR 72.44(c). The final approved technical specifications will be made part of the license.

13.3 Review Section 3.3.5.1 of the TSSAR notes that the ISFSI does not require continuous presence of operators or maintenance personnel. Additionally, radiation monitors are'not required. The pressure monitors are considered to be non-safety grade instruments that require replacement only when defective. Operating controls and limits that may serve as a basis for licensing i

conditions are derived from the analyses and evaluation included in the TSSAR.

13.4 Findings and Conclusions The staff reviewed the specific operating limits summarized in Chapter 10 of the TSSAR. The limits established for these parameters reflect the design criteria which the safety analyses were based and are acceptable. With regard to the fuel characteristics limits described in Section 10.1.2.4 of the TSSAR, the maximum initial enrichment is limited to 3.85 percent and the fuel assembly bumup shall not exceed 45,000 MWD /MTU. The Westinghouse 17x17 OFA was used as the design basis fuel in the TSSAR; therefore storage of other assemblies bounded by this analysis is acceptable.

A maximum handling height of 46 cm (18 inches) without impact limiters should be included as an operating limit. Horizontal storage is not permitted.

Upon completion of loading of the fuel assemblies, the seals should be leak-tested and demonstrate a leakage rate less than 1 x 105 atm-cc/sec, for a period of at least 30 minutes.

. The total decay heat for the cask must be limited to 27 kW and the decay heat for each fuel assembly must be limited to 0.675 kW.

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14 QUALITY ASSURANCE Chapter 6 of 'he NSP ISFSI License Application refers to Chapter 11 of the Safety Analysis -

Report (SAR) for the details of the Quality Assurance Program applicable to the Prairie Island ISFSI. In a letter dated April 22,1992, (Reference 27) NSP submitted an update of Section 4.5,

" Classification of Structures, Systems, and Components," and Chapter 11, " Quality Assurance,"

i of the Prairie Island ISFSI SAR and a commitment to incorporate them into the SAR as part of a future revision.

For the construction phase (design, fabrication, construction, start-up testing), the operational phase (operation, maintenance, modification), and the decommissioning phase of the Prairie Island ISFSI, NSP has committed to apply its NRC-approved 10CFR50 Appendix B QA program i

as described in Appendix IB of the Prairie Island updated Final Safety Analysis Report to the items defined in the ISFSI SAR as safety-related. These items are the containment vessel,~ the' i

basket assembly, the trunnions, and the concrete storage pads.

For the hardware associated with the lid gaskets and vessel penetrations, the shielding, and the security system, NSP has committed to apply its augmented quality program. These items, although important to safety, are not safety-related. In accordance with the acceptance criteria given in the Fuel Cycle Safety Branch Technical Position of June 20,1986, (Reference 28) 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. That is, when properly implemented, the NSP QA program for the Prairie Island ISFSI should meet the requirements of Subpart G of 10 CFR Part 72.

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15 REFERENCES 1.

Northern States Power Co, " Prairie Island Independent Spent Fuel Storage Installation, Technical Specifications and Safety Analysis Report," August 31, 1990.

2.

U.S.

Nuclear Regulatory Commission, Regulatory Guide 3.48,

" Standard Format and Content for'the Safety Analysis Report for an Independent Spent Fuel Storage Installation' (Dry Storage)," October 1981.

3.

U.S.

Nuclear Regulatory Commission, Regulatory Guide 3.61,

" Standard Format and Content for a Topical' Safety Analysis Report for A Dry Spent Fuel Storage Cask," February 1989.

4.

I.S.,

Levy, et al.,

" Recommended Temperature Limits for Dry

~

Storage of Spent Light Water Reactor Zircaloy-Clad Fuel Rods in Inert Gas," PNL 6189, May 1987.

5.

Thomas M.

Parker, Northern States Power Co.,

letter to U.S.

Nuclear Regulatory Commission, December 20, 1991.

6.

Thomas M.

Parker, Northern States Power Co.,

letter to U.S.

Nuclear Regulatory Commission, February 10, 1992.

l 7.

Thomas M.

Parker, Northern States Power Co.,

letter to U.S.

Nuclear Regulatory Commisssion, March 5, 1992.

8.

Thomas M.

Parker, Northern States Power Co.,

letter to U.S.

. i Nuclear Regulatory Commission, August 12, 1992.

9.

M. W. Schwartz and M. C. Witte, " Spent Fuel Cladding Integrity During. Dry Storage," UCID 21181, Lawrence Livermore National Laboratory, Livermore, California, September 1987.

10.

Northern States Power Co.,

" Prairie Island Independent Spent Fuel Storage Installation Technical Specifications and Safety Analysis Report," Rev.

2, September 1991.

11.

Thomas M.

Parker, Northern States Power Co.,

letter to U.S.

i Nuclear Regulatory Commission, December 23, 1991.

12.

Thomas M.

Parker, Northern States Power.Co., letter.to U'S.

Nuclear Regulatory. Commission, June 5,

1991, 13.

T.

Wilcox and E.

Lent,

" COG:

A Particle-Transport Code Designed to Solve the Boltzmann Equation for Deep-Penetration (Shielding) Problems - Volume I, Users Manual," UCRL.- Draft

Copy, Lawrence Livermore National Laboratory,- Livermore, California (October 1986).

14.

W. R. Lloyd, " Determination and Application of Bias Values in the Criticality Evaluation of Storage Cask Designs," UCID-21830, Lawrence Livermore National Laboratory, January 1990.

v e

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15.

U.S.

Department of Energy, OCRWM, " Characteristics of Spent Fuel, High Lasvel Waste, and Other Radioactive Wastes Which May Require Long-Term Isolation, Appendix 2A. Physical Description of LWR Fuel Assemblies," DOE /RW-0184, Vol. 3 of 6,

December l

1987.

16.

"U.S.

Nuclear Regulatory Commission, " SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation,"

NUREG/CR-0200, Revision 4

i (ORNL/NUREG/CSD-2/R4),

Vols.

I, II, and III, April 1991.

Available from Radiation Shielding Information Center as CCC-l 545.6.

l 17.

S.

R.,

Bierman, et al.,

" Critical Separation between

~

Subcritical Clusters of 4.29 Wt% 235U Enriched UO Rods in 2

Water with Fixed Neutron Poisons," NUREG/CR-0073, May 1978.

18.

S.

R.,

Bierman, et al.,

" Critical Separation between Subcritical Clusters of 2.35 Wt % 235U Enriched UO Rods in 2

Water with Fixed Neutron Poisons," PNL-2438, October 1977.

19.

U.S.

Nuclear Regulatory Commission, Regulatory Guide 8.8, "Information Relevant to Ensuring that occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is j

Reasonably Achievable," June 19'78.

3 20.

U.S.

Nuclear Regulatory Commission, Regulatory Guide 8.10,

" Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable, May 1977.

21.

U.S.

Nuclear Regulatory Commission, Regulatory Guide 1.8,

" Qualification and Training of Personnel for Nuclear Power Plants, April 1987.

22.

Northern States Power Co.,

" Prairie Island Independent Spent Fuel Storage Installation Technical Specifications and Safety Analysis Report," Rev.

1, April 1991.

23.

U.S.

Nuclear Regulatory Commission, Regulatory Guide 1.3,

" Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized l

Water Reactors," June 1974.

24.

U.S.

Nuclear Regulatory Commission, Regulatory Guide 1.25,

" Assumptions Used for Evaluating.the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," March 1972.

25.

U.S.

Nuclear Regulatory Commission, Regulatory Guide 1.109,

" Calculation of Releases of Radioactiave Materials In Gaseous and Liquid Ef fluents in Routine Releases from the Light-Water-Cooled Reactors, July 1977.

I

26.

Thomas M.

Parker, Northern States Power Co.,

letter to U.S.

Nuclear Regulatory Commission, July 10, 1992.

27.

Thomas M.

Parker, Northern States Power Co.,

letter to U.S.

Nuclear Regulatory Commission, April 22, 1992.

28.

U.S. Nuclear Regulatory Commission, " Branch Technical Position on Quality Assurance Programs for Independent Spent Fuel Storage Installations," June 20, 1986.

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I U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS ENVIIiONMENTAL ASSESSMENT RELATED TO CONSTRUCTION AND OPERATION OF THE PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION DOCKET NO. 72-10 NORTHERN STATES POWER COMPANY lR - j_

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I' TABLE OF CONTENTS

1.0 INTRODUCTION

I

1.1 DESCRIPTION

OF THE PROPOSED ACTION I

1.2 BACKGROUND

INFORMATION 2

1.3 PREVIOUS ENVIRONMENTAL ASSESSMENTS AND SUPPORTING DOCUMENTS 6

2.0 NEED FOR PROPOSED ACTION 6

3.0 ALTERNATIVES 7

4.0 EXISTING ENVIRONMENT........................

11 4.1 SITE LOCATION, LAND USE AND TERRESTRIAL ECOLOGY 12 4.2 WATER USE AND AQUATIC RESOURCES 14 4.3 SOCIOECONOMIC, HISTORICAL, ARCHE 0 LOGICAL AND CULTURAL RESOURCES

. 15 4.4 DEMOGRAPHY...........................

16 4.5 METEOROLOGY 16 4.6 GEOLOGY, SEISMOLOGY AND SOILS

.................16

5.0 DESCRIPTION

OF THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT ISFSI 17 5.1 GENERAL DESCRIPTION

......................17 5.2 ISFSI DESIGN..........................

18 5.3 ISFSI OPERATIONS........................

22 5.4 MONITORING PROGRAM

.......................27 6.0 ENVIRONMENTAL IMPACTS OF THE PROPOSED ACTION

............27 6.1 CONSTRUCTION IMPACTS

......................27 6.1.1 LAND USE AND TERRESTRIAL RESOURCES............

28 1

6.1.2 WATER USE AND AQUATIC RESOURCES

.............29 6.1.3 OTHER IMPACTS OF CONSTRUCTION 29 6.1.4 SOCI0 ECONOMIC

......................30 6.1.5 RADIOLOGICAL IMPACTS FROM CONSTRUCTION..........

30 6.2 OPERATIONAL IMPACTS

......................31 6.2.1 RADIOLOGICAL IMPACTS FROM ROUTINE OPERATIONS.......

31

-1. 6.2.1.1 0FFSITE DOS E...................

31 d' 6.2.1.2 COLLECTIVE OCCUPATIONAL DOSE...........

32 14 6.2.2 RADIOLOGICAL IMPACTS OF 0FF-NORMAL EVENTS AND ACCIDENTS 40 6.2.3 NONRADIOLOGICAL IMPACTS

.................47 6.2.3.1 LAND USE AND TERRESTRIAL RESOURCES........

47 6.2.3.2 WATER USE AND AQUATIC RESOURCES

.........47 6.2.3.3 OTHER EFFECTS OF OPERATION............

47 in

M 7.0 SAFEGUARDS FOR SPENT FUEL 48 8.0 DECOP911SSIONING 49-9.0

SUMMARY

AND CONCLUSIONS 50 9.1

SUMMARY

OF ENVIRONMENTAL IMPACTS................

50 9.2 BASIS FOR FINDING OF NO SIGNIFICANT IMPACT...........

51

10. REFERENCES 52 I

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ENVIRONMENTAL ASSESSMENT RELATED TO THE CONSTRUCTION AND OPERATION OF THE PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION

1.1 DESCRIPTION

OF TIIE PROPOSED ACTION By letter dated August 31,1990, Northern States Power Company (NSP) (the Applicant) submitted an application for a Nuclear Regulatory Commission license to construct and operate a dry cask independent spent fuel storage installation (ISFSI) to be located on the site of the Prairie Island Nuclear Generating Plant in Goodhue County, Minnesota. The ISFSI or some other spent fuel storage system is needed to provide interim onsite spent fuel storage.

This Environmental Assessment (EA) addresses the expected environmental impacts associated with the proposed construction and operation of the ISFSI on the Pratrie Island site.

NSP owns approximately 560 acres of land at the Piairie Island Nuclear Generating Plant.

They operate two 1650 MWt nuclear generating units at the site. There are a few areas at the plant site controlled by the U.S. Army Corps of Engineers. The protected area of the ISFSI and the access road connecting the ISFSI and Auxiliary Building will be on land owned by khe site is lo$ated within the city limits of the City of Red Wing, Minnesota, oak. west bank of the Mississippi River. The controlled area for the ISFSI g

corresponds to the exclusion area of the nuclear station. The nearest site boundary (fence) is 360 feet west of the nearest edge of the ISFSI concrete storage pad. The nearest mad (Wakonade Drive East) borders the site boundary to the west of the ISFSI. He edge of the road is 400 feet west of the nearest edge of the pad. Figure 1.1 shows the location of the 1

i proposed ISFSI relative to the other features on the site including the reactor buildings and security fence. Figure 1.2 provides additional detail on the ISFSIlayout.

i The proposed ISFSI is a system designed oy Transnuclear Incorporated of Hawthorne, New York. It is called the TN-40 Dry Cask Storage System and is similar in design to the TN-24, which has already been approved by the NRC. Each cask will hold 40 Prairie Island spent fuel assemblies. The ISFSI will be designed to accommodate a total of 48 TN-40 storage casks stored on two concrete pads, each with two parallel rows of 12 casks. The ISFSI site will be surrounded by an earthen berm to a height of 17 feet above pad grade.

The casks will be loaded with the spent fuel assemblies in the spent fuel pool enclosure of the Auxiliary Building at the plant, decontaminated, lifted by a crane and moved laterally through an access door. They will then be and picked up by the transport, vehicle which will be pulled by the tow vehicle to the concrete pads at the ISFSI site. The casks are self-contained, independent, passive systems, which do not rely on any other systems or components for their operation.

The TN-40 dry storage cask is designed to provide storage of spent fuel for at least 25 years.

The ISFSI will provide adequate capacity to enable Units 1 and 2 at the Prairie Island Nuclear Generating Plant to continue operation until expiration of their licenses in 2013 and 2014, respectively. 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.

1.2 BACKGROUND

INFORMATION g:

~i s The Prairie Islabi Nuclear Generating Plant began commercial operation in December 1973.

Prior to the mid 1970's, the nuclear industry 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

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generated. After an indefinite interim storage period, utilities anticipated that spent fuel l'

would be transported to a reprocessing plant for recovery and recycling of fuel materials.

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Reactor facilities, such as the Prairie Island units, were not designed to provide spent fuel storage capacity for life-of-plant operations.

Because commercial reprocessing did not develop as anticipated, the 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 through nuclear power plant shutdown. A " Final Generic Environmental Impact Statement (FGEIS) on Handling and Storage of Spent lj ht Water Power Reactor Fuel,"

g NUREG-0575, was issued by NRC in August 1979 (Reference 1).

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 derating 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. 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 as feasible as wet storage and eendmamentally acceptable. In the case of both dry passive storage and wet NC storage, envircomental impacts need to be considered on a site-specific basis, g:

The onsite expansion of spent fuel pools has been used by most utilities. NRC has reviewed and approved more than 120 onsite spent fuel pool capacity expansions through reracking modifications since issuance of the FGEIS. At Prairie Island, efforts to maintain sufficient spent fuel storage reserve capacity have included a decrease in the annual spent fuel 5

discharge rate through various core design refinements and two separate spent fuel pool rerackings. Additional expansion of onsite storage capacity will be required by 1994 to ensure uninterrupted operation of the Prairie Island units.

i As required by 10 CFR Part 72 and Part 51, this assessment addresses the site-specific environmental impacts of construction and operation of the dry storage ISFSI at the Prairie Island site.

1.3 PREVIOUS ENVIRONMENTAL ASSESSMENTS AND SUPPORTING DOCUMENTS Several environmental documents have been prepared specific to the Prairie Island site. A Final Environmental Statement (FES) related to the Prairie Island Nuclear Generating Plant was prepared by the U.S. Atomic Energy Commission in May of 1973 (Reference 2). Th$s EA relies on information supplied by NSP in its Environmental Report (ER) (Reference 3) related to the proposed ISFSI for Prairie Island submitted with the application in August 1990 and supplementary information submitted in response to NRC questions in References 4 and 16. In addition, the FGEIS (NUREG-0575) (Reference 1), the applicant's Updated Safety Analysis Report (USAR) (Reference 5) and Technical Specifications and Safety Analysis Report (Reference 6) provided additional information. The Minnesota State EIS (Reference 7) is also referenced.

A 2.0 NEED FOR PROPOSED ACTION

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Discharged assemblies from Prairie Island Nuclear Generating Plant, Units 1 and 2, are currentfpsored onsite in a spent fuel pool. The spent fuel pool provides for long-term storage of 1386 assemblies in high density storage racks (Reference 6). The spent fuel pool will lose capacity for discharge of a full core in 1993. Storage capacity will be exhausted completely in 1994. The capacity of the ISFSI will enable NSP to store an 6

additional 1920 spent fuel assemblies in 48 casks, and will enable Units 1 and 2 to continue opemtion until expiration of their respective Operating Licenses in 2013 and 2014.

3.0 ALTERNATIVFJ NSP evaluated a number of alternatives for the storage of spent nuclear fuel prior to the selection of the dry storage ISFSI. The alternatives did not sufficiently meet the require-ments for storage of spent nuclear fuel generated at the Prairie Island Plant. A brief discussion of these altematives 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 (DOE) is currently working to develop a repository as required under the Nuclear Waste Policy Act of 1982, amended in 1987 (NWPA). DOE is looking at a site at Yucca Mountain, Nevada, to determine if it is a suitable location for a high-level radioactive waste repository. It is not likely that DOE will have a licensed repository ready to receive spent fuel before 2010.

Although DOE 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 does not meet the immediate needs of NSP.

.,.a.,

y-Alternative Dry,Sirage Systems Several alternative dry storage systems other than the Transnuclear TN-40 Dry Cask Storage System exist. The NUTECH Horizontal Modular Storage (NUHOMS) is in place, or planned for installation, at other locations. A vault storage system has been in use in Great Britain, and a vault designed to store United States type spent fuel has received NRC 7

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approval. A concrete cask storage system, similar to NUHOMS except providing vertical storage, is being reviewed by the NRC. In addition, metal dry storage casks have been developed by other companies. While the design may differ, the impacts associated with these alternative dry storage systems are expected to be similar. NSP determined that these alternative dry storage systems did not meet its needs.

Modified Pool Storage Modifications to the existing fuel pit could be made to combine it with spent fuel pool and thereby increase the total pool storage capacity. The modifications are of a sufficient magnitude that the pool would be out of commission until completion. This alternative for expansion would add storage for about 500 more spent fuel assemblies. At current generation rates, this greater storage capacity would be exhausted by about 2002. This does.

not provide a means to store spent fuel for the balance of the plant's licensed operational lifetime. Also, the spent fuel in the pool would need to be stored elsewhere while the modification took place. For these reasons, this alternative was not selected.

Existing Pool Capacity Increase There are several alternatives in this category which involve modifying the existing spent fuel storage pool. Unlike the above alternative, these do not require expansion and the major reconstruction of the fuel pool. Reracking, or changing to racks designed with a more compact array of cells, was last done in 1981. Current generation rack designs are even it might be possible to increase the pool capacity by about 15 percent by more co reracking

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possible way to increase the existing pool capacity is through spent fuel rud con _~ _ % In consolidation, the fuel rods from two spent fuel assembhes are i

1 removed, reconfigured and then placed m a camster, returning to the rack cell formerly

-l occupied by a single spent fuel assembly. 'Ihe use of two-tiered racks is a third method of increasing pool capacity. A second tier of filled storage racks is placed on top of the 1

existing storage racks. None of these alternatives meets NSP's needs. Reracking will not 8

i provide a means to store spent fuel for the balance of the plant's licensed operational lifetime. Spent fuel rod consolidation would not meet life-of-plant needs and would interfere with normal plant operations. The use of two-tiered racks would require considerable support of the fuel pool walls, and there are technical and licensing uncertainties associated with it. Therefore, none of these alternatives alone is adequate to meet Prairie i

Island's storage needs.

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 capacity of the pool is fixed at the time of construction. The NRC has generically assessed this alternative and found that the storage of LWR spent fuel in water pools has an insignificant impact on the environment (Reference 1).

This alternative would require about 5 years to design, obtain state and federal reviews and '

approvals, and construct. It could result in the temporary shutdown of both reactors or a sustained period of operation at reduced power levels, and was, therefore, a less attractive alternative than the proposed action.

Shipment to Existing Storage Facilities This alternative involves shipping Prairie Island spent fuel to Monticello or to Pathfinder, near Sioux Falls, South Dakota, or to a facility at a site owned by another utility.

Monticello's fuel assemblies are smaller than Prairie Island's and the handling tool is different. 'Iheyeh and handling equipment would have to be replaced and modified to store Prairie Is5and fuel. Additionally, the current pool capacity does not provide a means to store spent fuel for the balance of the plant's operational lifetime. Shipping to Pathfinder could not be accomplished without rebuilding the fuel storage system, which was removed 1

during conversion from a nuclear power plant to a fossil fuel plant in 1967. A third possibility entails shipping to a spent fuel storage facility at another site. This alternative requires another utility to agree m this arrangement, and it is unhkely that this will happen.

9

The impacts of storage at other facilities would be similar to the impacts of storage at Prairie Island, but with the additional concem of transportation impacts. Accordingly, these alternatives were not selected.

Shipment to Reprocessing Facility When Prairie Island was constructed, NSP intended to ship the spent fuel to a commercial reprocessing facility in the United States. The reprocessing indust *y did not develop as expected and applications for reprocessing were frozen in 1977 during President Carter's administration because of proliferation concerns. Reprocessing services are available in France and Great Britain. To avail itself of these services, NSP would have to contract with a company to receive and reprocess the spent fuel, manage the plutonium and uranium extracted, and solidify the waste for shipment back to the United States for permanent disposal. This alternative has not been selected by any utility to date because the proced difficulties are thought to be insurmountable. It is not, therefore, considered a viable alternative.

Future Reduction in Rate of Spent Fuel Generation Reducing the rate of spent fuel assemblies generated will defer the date at which space in the spent fuel pool runs out. The use of fuel with a higher burnup is one way to reduce the rate of generation. The combination of fuel and core design currently being used at Prairie Island is achieving the maximum burnup allowed today. This alternative does not, therefore, provide a p the storage problem. Reducing operations and the rate of spent fuel generation at Prairie Island plant until DOE begins to accept spent fuel could only be

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considered a alternative if the forecast time was more certain. While the reduction in rate of spent fuel generation would have less impact on the environment than other storage or shipment alternatives, it does not appear feasible to achieve at Prairie Island. Since large uncertainties remain concerning this whole issue, the potential for Prairie Island to run out of 10

storage space under reduced operation is very real. It is, therefore, not considered a potential alternative.

No Action This alternative would result in NSP filling the existing spent fuel storage capacity at the Prairie Island plant by January 1994, and thereby forcing shutdown of the plant. NSP estimates large baseload facilities cannot be brought into service until the late 1990's.

Peaking plants could be built by 1994, but these cannot generate electricity for baseload use in a cost-effective manner. This alternative was not selected for this reason. The impacts of curtailing the generation of spent fuel by ceasing operation of existing power plants when their spent fuel pools become filled was evaluated by the NRC in the Final Generic Environmental Impact Statement and found to be undesirable.

4.0 EXISTING ENVIRONMFRT The ;eneral environment around the Prairie Island Nuclear Generating Plant is well characterized as a result of studies conducted in support of the construction of the plant, and from additional field investigations at the proposed ISFSI site made in June 1991. This section briefly reviews the environment surrounding the site, with emphasis on those features most likely to be affected by the construction and operation of the ISFSI. 'Ihe assessment of construction and operational impacts is presented in Chapter 6.

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4.1 SITE LOCATION, LAND USE AND TERRESTRIAL ECOLOGY Site Location The proposed ISFSI will be located within the existing plant site for Prairie Island. It is located in the city limits of the city of Red Wing, Minnesota, on the west bank of the Mississippi River. It will be located in Section 5, Til3N, R15W in Goodhue County, at approximately 92 degrees 37.9 minutes west longitude and 44 degrees 37.3 minutes north latitude. The ground surface near the Prairie Island site is fairly level to slightly rolling, ranging in elevation from 675 to 706 feet above mean sea level (1929 adjustment). The surface slopes gradually toward the Mississippi River to the northeast and the Vermillion River on the southwest. Steep bluffs run parallel to this stretch of the Mississippi River and riz to an elevation of over 1000 feet above mean sea level approximately 1.5 miles northeast and southwest of the site.

Land Use Goodhue County, in which the site is located, and the adjacent counties of Dakota and Pierce (in Wisconsin) are predominantly rural. Dairy products and livestock account for most of the farm products with field crops and vegetables accounting for most of the remainder. The region within a 5-mile radius of the site is almost exclusively agricultural. Principal crops include soybeans, corn, oats, hay and some cannery crops at about 4 miles from the plant site. The nearest dairy fann is located more than 3 miles southwest of the plant site. Some beef cattle are raised appuximately 2 miles southwest. Cattle are on pasture from early June to late September;or early October. During the winter, cows are fed on locally produced hay and silage. Behbnd the site boundary, within a 1-mile radius of the plant, there are approximately 30 permanent residences and summer cottages. The closest occupied offsite residence is approximately 2400 feet northwest of the proposed ISFSI site. Located near the site, is the Mdewakanton Sioux reservation. All traffic to and from the plant passes through the reservation (References 3 and 7).

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There are several industrial facilities located within 5 miles of the ISFSI site. No military installations are within 5 miles, and no large natural gas pipelines pass close to the site. The Red Wing airport is located about 7 miles east southeast. High speed railroad traffic occurs on the Soo Line Railroad, the Burlington Northern and the Chicago Northwestern Railroad, all within 5 miles of the site. Truck traffic occurs on Minnesota State Highway 61, which runs within 2.5 miles of the ISFSI site to the south. In addition, several county roads are located within 5 miles. Barge traffic on the Mississippi River occurs in the main channel within one half mile of the ISFSI site.

9 Terrestrial Ecology Terrestrial Ecology studies have been conducted in a 1.5-mile radius around the Prairie Island Nuclear Generating Plant on the Minnesota side of the Mississippi River. These -

t studies identified the quantity and quality of various habitats in the near vicinity. The four major types of habitat are oak openings, lowland forests, prairie or abandoned fields, and sand terrace. The ISFSI site area is approximately 70% wooded and 30% open. Of the wooded area, about 80% of the trees are Siberian elm (Ulmuspumila)..The remaining woody species include American elm (Ulmus americana), white pine (Pinus strobus), box elder (Acer negundo), Cottonwood (Populus deltoides), Aspen (Populus tremuloides), Red.

cedar (Juniperur virginiana), Red oak (Quercus rubra), White spruce (Picea glauca), and Butternut (Juglam cinera). In addition, there are some shrubs and woody vines, such as Staghorn sumac (Rhus typhina), raspberry (Rubus indaeus), prickly ash (Xanthaxylum americanum), grape (Vitis riparia) and poison ivy (Rhus radicans). The three most common

.m A grasses found site are little Bluestem (Andropogon scopariur), Big Bluestem w

(Andropogon~

and Kentucky Bluegrass (Poa pratemis). Since this is an'old farm site, many of were planted. Steps were tram in 1974 to restore prairie vegetation in the vicinity of the plant, through controlled burning. This was continued in 1975 with

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plowing and seeding of several native plant species. The results of the ecology studies and the prairie establishment operation are included in documents contained in the Prairie Island 1

Nuclear Generating Plant Annual Reports (Reference 16).

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t Wildlife ecology studies.were conducted together with the terrestrial ecology studies.

Common species of small mammals, insects, and birds predominate in the vicinity of the proposed ISFSI site. Certain species were targeted for fmther studies. The Great Blue Heron and the Great Egret were studied for four years to determine the effects of power plant operation on them. Direct effects from operation at the plant were determined to be minimal. A Mourning Dove and Grackle study was initiated in 1974. The conclusion was reached that the Mourning Dove population at Prairie Island was stable. Bald Eagle studies were also conducted. The results of all of these studies are included in the Prairie Island Annual Reports.

4.2 WATER USE AND AQUATIC RESOURCES The principal surface waters in the area of the site are the Mississippi River, the Vermillion River, the Cannon River, and Sturgeon Lake. The level: f the Mississippi River and Sturgeon Lake are controlled by Lock and Dam Number 3 which is located approximately one and one half miles downstream from the plant. The Vermillion River enters the main stream of the Mississippi below the dam. There are no withdrawals of river water for supply of city water for at least 300 miles downstream from the site. Some withdrawals of water for irrigation use do occur, with the nearest being 53 miles downstream.

Regionally, the movement of ground water is toward the Mississippi River and its main tributaries. The ground water slopes toward these surface streams, generally at low gradients. The result of boring tests performed in June 1991, show that the ground water generally is found at depths of 16.0 to 20.7 feet below the surface. Due to the permeable nature of the sindy alluvial soils forming Prairie Island, the ground water table responds y

quickly to changes in river stage. There is only minor usage of ground water near the site or immediately downstream. The nearest ground water consumption of magnitude is 6 miles downstream in the town of Red Wing. The water supply comes from four deep wells, which penetrate sandstone aquifers. These wells pump from depths of 400 to 730 feet. Several industries ir the Red Wing area also utilize ground water, principally from the bedrock 14

aquifers. The communities of 12ke City and Wabasha, 25 and 37 miles downstream, also supply their water needs from wells in bedrock. The wells located in and around the Indian reservation pump from depths of 90 to 110 feet.

Dispersion of surface run off/ drainage (effluents) entering the ground water system from the plant would take place principally in the upper poMon of the saturated zone of the river alluvium. Due to the numerous surface waterway.s in the area of the site, most of the surface run off would leave the ground water and mix with surface waters at the borders of Prairie Island. These effluents are from operations in the reactor building, not from the ISFSIitself.

4.3 SOCIOECONOMIC, HISTORICA.L, ARCHEOLOGICAL AND CULTURAL RESOURCES The immediate area surrounding the Prairie Island site is predominately ruial, with the exception of the city of Red Wing. The Prairie Island Indian Reservation is located within '

one mile of the proposed site. There is a large community center nearby, run by the Mdewakanton Sioux.

The area surrounding the P.airie Island Nuclear Generating Plant is one of past Indian and French trader activity. An archaeological survey was conducted in 1967, and nothing significant in the immediate area of the power plant or ISFSI area was found. They did find evidence of an Indian village and burial mounds at the southern boundary of the plant site.

This area, called the Bartron Archaeological Site, has been designated to archaeological

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interests and,was added to the National Registet of Historic Places in February 1971. No other areas of.

, archeological and cultural significance are found within the site soundary. w t.

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4.4 DEMOGRAPHY The population density in the vicinity of the Prairie Island Nuclear Generating Plant is generally low, with the exception of the city of Red Wing. The nearest offsite occupied residence is approximately.45 miles northwest of the proposed ISFSI site. The 1985/1986 population within 2 miles of the site was estimated to be 464 people, and within 10 miles it was estimated to be 23,054 people. An additional 320 people reside within the 10-mile radius on a seasonal basis. The city of Red Wing projected a population of 14,754 people for the year 1990, which accounts for over half of the projected Goodhue County total population for that year. The estimated 1985/1986 permanent population distribution within 50 miles is 2,193,433. This population is capected to increase from 3 to 11 percent per decade, (Reference 3).

t 4.5 METEOROLOGY

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The climate in the region of the Prairie Island Nuclear Generating Plant is basically continental with influence from the general storms which move eastward along the northern I

part of the United States. The geographical location results in frequent changes in weather systems as the polar and tropical air masses alternate. Rainfall averages about 25 inches per year, with 65 percent falling in the months of May through September. Snowfall averages about 44 inches per year. Minnesota lies to the north of the principal tornado belt in the United States. Data collected in 1971-1972 indicate an average wind speed of 6.7 mph with a prevalence of stable vertical stability conditions, (Reference 5).

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4.6 GEOLOGY, SEISMOIDGY AND SOIIE MT :.I The Prairie Island Nuclear Generating Plant is located on a low island terrace associated with the Mississippi River flood plain. 'Ihe Mississippi River flood plain in this area is confined within a valley about 3 miles wide. Rocky bluffs and heavily forested slopes rise abruptly from both sides of the valley to a height of about 300 feet. The overburden materials at the 16 I

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i site are permeable sandy alluvial soils, which were deposited as glacial outwash and as recent river sedimentation. The uppermost bedrock unit at the site is sandstone and is believed to l-be part of the Franconia formation. Underneath the Franconia formation are several hundred feet of lower Cambrian and Precambrian sandstone with minor shale horizons. The dominant structural feature in the area is the Keweenawan Basin which was formed in early Precambrian times. This basin is separated from a smaller basin in the Twin Cities area by the Afton-Hudson anticline. The site is located on the west limb of the Red Wing anticline.

There are several major faults in the Minnesota-Wisconsin region. The principal movements along these faults appears to have been restricted to Precambrian times. The Douglas fault and the Lake Owen fault penetrated Precambrian rocks along the North and South sides of

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the Keweenawan Basin, respectively. A southern portion of the Lake Owen fault, known as the Hastings fault, trends southwest near the city of Hastings, about 13 miles northwest of the site. There is no evidence of recent activity along any of the known fault zones in the Minnesota-Wisconsin region.

l The soils at the site are somewhat frost susceptible, and to avoid any potential problems,-

footings and slabs will be founded below the anticipated frost depth or on fill below the frost' depth. The settlement upon loading the cement slab has been found to be acceptable. In addition, a liquefaction analysis was performed in June 1991, and the subsurface materials have been found to be stable and adequate for the proposed foundation loading.

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5.0 DESCRIPTION

OF THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT ISFSI 5.1 GENERAL DESCRIFFION The proposed ISFSI is designed to safely store spent fuel by confining the fuel and providing shielding from radiation through the incorporation of physical components and a system of 17

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complementary pmcedures to protect both the onsite personnel and the general public from radioactivity in the spent fuel. 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 ISFSIis described in Section 5.4.

5.2 ISFSI DESIGN l

The ISFSI provides for the temporary dry storage ofirradiated fuel assemblies in storage casks. The major physical components of the proposed ISFSI are the spent fuel, the storage casks, the transport vehicle and the concrete pads. Each of these components is discussed-below.

Spent Fuel I

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Spent fuel, because of its radioactive nature, presents a potential hazard to plant personnel, the general public, and the environment. The ISFSI system is designed to safely store spent.

fuel by confining the fuel material and providing bulk shielding from radiation. Fuel to be stored at the ISFSI will originate only from Prairie Island and will hav~e been allowed to cool i

a minimum of 10 years. In addition, only intact unconsolidated fuel must be used. If partial assemblies are to be stored, the missing fuel pins must be replaced by dummy pins.

NSP has identified the spent fuel assemblies to be stored in the ISFSl, Specifically, the spent fuel must comply with the restrictions listed in Table 5.1 before it will be transferred to the j.

ISFSI. 'Iheaslgeggictioon me based on the need to assure that: (1) there is no potential for x

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9) maximum allowable fuel clad temperatures are not exceeded, and (3) z 7'.

dose rates the ISFSI age within the allowable design limits.

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Storage Cask The TN-40 storage cask consists of a basket assembly for support of the fuel assemblies, a containment vessel, the gamma and neutron shields, a weather cover, a pressure monitoring system, and trunnions for lifting and rotating the cask. The casks will be stored in a i

vertical position on a concrete pad. The ISFSI will enable NSP to store 1920 spent fuel assemblies in 48 casks. The storage casks are designed with the objectives of ensuring that fuel criticality is prevented, cask integrity is maintained, and fuel is not damaged so as to preclude its ultimate removal from the cask. A summary of the TN-40 Cask design is 1

included in Table 5.2.

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TABLE il FUEL ASSEMBLY PARAMETERS Maximum weight (w/o control component) 1300 lb.

Assembly dimensions 7.763"x7.7 3"x 61.3" (w/o control component)

Fuel rod array 14 x 14 Number of fuel rods 179 Active fuel length 144" Initial enrichment 3.85 w/o U235 (maximum)

Burnup (maximum) 45 GWD/MTU Cooling time (minimum) 10 years Initial uranium content:

(max.)

400 kg (min.)

360 kgU Fuel Pellet O.D.

0.3444 in.

Fuel Rod O.D.

0.4000 in.

Clad Thickness 0.0203 in.

. Clad Materh1 Zr-4 i $$

{s s Source per Assembly 2.44E+15 photo s/sec

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' Neutron Source per' Assembly 2.19E+8 neutro s/see Decay Heat per Assembly 0.675 kW 20 f

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TABLE 12 DIMENSIONS AND WEIGHT QE TN-40 CASK No. of assemblies / cask 40 Overall length, w/ cover 202.0" Outside diameter 72.0" Loaded weight on storage pad 240,690 lb.

Loaded weight on crane hook 237,533 lb.

I Surface temperature less than 250 degrees Cooling Radiant and convective Maximum Surface Contact Dose Rate:

radial 57.5 mrem /hr top 25.6 mrem /hr bottom 1275.0 mrem /hr Maximum Surface Dose Rate on the pad 200 mrem /hr l

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The transport h moves the loaded storage casks from the Auxiliary Building rail bay to the concrete pads in the ISFSI. The transport vehicle must be designed for a minimum of 100 fully-loaded one way trips over approximately a 25-year period over several different types of ground surfaces. The transporter shall be designed to limit cask lift height to less than 18 inches.

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O Concrete Pad The storage casks will be stored in two parallel rows of 12 casks on each of two 216-foot long x 36-foot wide x 3-foot thick concrete pads. The two slabs will be positioned end to end with 40 feet in between. To improve foundation performance and earthquake safety, 3 feet of soil beneath each slab will be excavated and replaced with compacted structural fill.

The pad elevation will be 693 feet 6 inches above mean sea level (msl) to preclude immer-sion of the cask seals during the probable maximum flood. They will be surrounded by a 17-foot high earthen berm.

5.3 ISFSI OPERATIONS Fuel handling and cask loading operations in the Auxiliary Building will be done in accordance with requirements of the Prairie Island Nuclear Generating Plant 10 CFR Part 50 Operating License: DPR-42 (Unit #1) and DPR-60 (Unit #2). Cask transport and storage at the ISFSI will be subject to requirements of the Prairie Island ISFSI 10 CFR Part 72 License. The major steps associated with the placing of fuel in the Prairie Island ISFSI are presented in Table 5.3.

- 1ABL.S 5 3 ISFSI OPERATIONAL STEPS A. RECEIVING

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1.

Unload' empty cask and separately packaged seals at plant site.

2.

Inspect the following for shipping damage: exterior surfaces, sealing surfaces, trun-nions, seals, accessible interior surfaces and basket assembly, bolts, bolt holes and threads, neutron shield vents.

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3.

Remove' weather shield and install plug in neutron shield vent hole.

i 4.

Remove lid bolts and lid.

t 5.

Install protective plate over cask body sealing area.

6.

Obtain hd and lid seal from storage.

7.

Attach lid seal to lid by means of six retaining screws.

8.

Move to spent fuel pool area.

i B. SPENT FUEL POOL AREA 1.

Lower cask into cask-loading pool.

2.

Imad preselected spent fuel assemblies into the 40 basket compartments.

3.

Verify identity of the fuel assemblies loaded into the cask.

4.

Remove protective plate from cask body flange.

1 5.

Iower lid and place on cask body flange over the two alignment pins.

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Lift to of pool and install lid bolts.

7.

Connect drain line to quick-disconnect coupling in the drain port.

8.

Bolt special adapter, with quick disconnect coupling, to vent port bolt holes.

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9.

Connect plant compressed air line to special adapter quick-disconnect coupling.

10. Pressurize cavity to force water from cavity through drain port to the spent fuel pool.

I1. Disconnect plant compressed air line and drain line from their quick-disconnect couplings.

12. Move cask to the decontamination area.

C. DECONTAMINATION AREA (RAIL BAY) 1.

Decontaminate cask until acceptable surface contamination levels are obtained.

2.

Torque lid bolts using the prescribed procedure.

3.

Remove plug from neutron shield vent and install pressure relief valve.

4.

Connect Vacuum Drying System (VDS) to vent port.

5.

Evacuate cavity to remove remaining moisture using prescribed procedure.

6.

Break vacuum by closing vacuum valve and opening air valve to admit dry air into the cavity.

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7.

Disconnect VDS at vent port and install vent port cover with seal and bolts.

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8.

Connect Vacuum-Backfill System (VBS) to quick-disconnect coupling in the drain port.

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9.

Evacuate cavity to 10 millibar and backfill with dry helium gas,

10. Pressurize cavity to about 2 ATM with helium.

I1. Disconnect VBS at the drain port quick-connect coupling and install drain port cover with seal and bolts.

12. Perform helium leak test of lid seals.

13. Remove over pressure port cover.

14. Install top neutron shield drum.

15. Torque the bolts using prescribed procedure.

16. Pressurize over pressure system with Helium to a pressure of about 5.5 ATM.

17. Perform leak test on over pressure system.

18. Check external surface temperatures using an optical pyrometer.

19. Check surface radiation levels.

20. Install yesective cover with seal and bolts.

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21. Load hisif on transport vehicle.

22. Move cask to Storage Area.

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D. STORAGE ARPA 1.

Unload cask from transport vehicle.

2.

Position cask in preselected location on storage pad.

3.

Check for surface defects.

4.

Connect pressure instrumentation to cask and to monitoring panel.

5.

Check that pressure instrumentation is functioning.

6.

Check surface radiation levels.

The administrative procedures for the ISFSI will be the same as those used for the Prairie Island Nuclear Generating Plant. Any changes to these procedures will be reviewed and approved by the Station Operations Committee and Safety Audit Committee. Before startup and during the lifetime of the ISFSI, the cask monitoring instrumentation, the electrical system, the communications system, and the storage casks will be tested to ensure their proper functioning. The existing training program at the plant will be used to provide and maintain a well qualified work force for safe and efficient operation of the ISFSI. All personnel working in the fuel storage area will receive radiation and safety training and tnose actually performing cask and fuel handling functions will be given additional traimag in specific areas as aquhed by the Radiation Protection program in effect at the Prairie Island n.

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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 also will r ot generate any chemical, sanitary, or solid wastes; or release any radioactive materials in solid, gaseous, or liquid form during normal operations. Similarly, because there are no liquid or gaseous effluents from the ISFSI, special environmental monitoring for these exposure pathways is not necessary. Therefore, a separate environmental measurement program for the ISFSIis not warruted. However, *a help ensure proper operation of the ISFSI system, NSP will incorporate monitoring in the Prairie Island site monitoring program. The site operational surveillance program will also be expanded to include surveillance of the ISFSI.

The current operational meteorological and radiological monitoring programs will be.

continued through the life of Prairie Island Nuclear Generating Plant, Units 1 and 2.

The program is designed to confirm that NSP operations are within regulatory requirements and consistent with the documented As Low As Is Reasonably Achievable (ALARA) program.

The main purpose of this program is to minimize exposure to radiation so that the total exposure to personnel in all phases of design, construction, operation, and maintenance are as low as can be reasonably achieved. As an additional measure of conservatism,.16 i

thermoluminescent dosimeters will be placed at equal intervals along the perimeter fence in the vicinity of the ISFSI. The population dose from the plant will be reported periodically in i

the Station Operating Reports, and these will include the contribution from the ISFSI.

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6.0 ENVIRONMENTAL IMPACTS DE IIIE PROPOSED ACTION l

6.1 CONSTRUCTION IMPACTS 1

The ISFSI site will be developed and managed so as to minimize construction impacts. All construction activities will comply with Federal, State and local regulations for environmental 27

f protection, as well as occupational safety and health. Most of the construction area is covered with prairie grass and weeds or is wooded and will be cleared. Timber resulting from the clearing operation will be collected for appropriate disposal. Portions of the ISFSI site and adjacent areas have been used for the disposal of dredged material taken periodically from the station intake channel.

6.1.1 LAND USE AND TERRESTRIAL RESOURCES Construction of the ISFSI, including the site area, berm and access road, will affect approximately 10 acres of the 560 acre Prairie Island Nuclear Generating Plant site area.

Portions of the ISFSI site and adjacent areas have been used for the disposal of dredged material taken periodically from the station intake channel. The principal tenain alterations to the site area will come from clearing, excavation, grading, and berm construction.

Cleared areas and exposed earth will be seeded, graveled or paved to stabilize and control -

runoff, and to minimize soil erosion. After construction of the concrete slabs is complete, the area immediately surrounding the slabs will be covered with well-compacted crushed rock. The construction will not impact offsite land use.

Loss of biological production from approximately 10 acres is anticipated. The habitat displaced by the ISFSI consists primarily of trees, shrubs, prairie grasses, and weeds. It is also used by common small mammals, insects and birds. Displacement of resident fauna within the proposed ISFSI is likely to occur due to construction activities which produce noise. Since wildlife egress from the area immediately surrounding the construction site is unrestricted, the construction noise impact is expected to be minimal. The habitat is not w

unique or cri wildlife. The site r.rea is not used for nesting or feeding by bald eagles c.

or migratory,

Disruption of wildlife activities due to construction noise is expected to be minimal (Reference 8). The only resources committed irretrievably are the steel, concrete, and other construction materials in the ISFSI pads and storage casks.

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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 and water for cleaning operations and fugitive dust control 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, a temporary drainage system may be constructed to collect the runoff into temporary settling ponds. More permanent drainage will be installed as soon as area excavations and backfill allow. The drainage system will not alter the natural drainage patterns. As the construction of the ISFSIinvolves no use or degradation of the regional water, its impact on navigation, fish and wildlife resources, water quality, water supply and aesthetics should be negligible.

The ISFSI has been sited to avoid the problems associated with the occupancy and modification of flood plains. It has been designed such that the lowest point of potential leakage into the cask is above the level of the probable maximum flood.

6.1.3 OTIIER IMPACTS OF CONSTRUCTION Noise Construction of the ISFSI will generate noise, but it will be of minimal duration. Due to the distance of the site from the nearest residence, the impact on the surrounding community is considered to be acceptable. By complying with Occupational Safety and Health Administra-tion (OSHA) noise regulations, the impact of noise on the construction workers will be minimal.

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9 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 roads, will be used to minimize the generation of fugitive dust. In addition, cleared areas and exposed carth will be seeded, graveled, or paved to stabilize and control runoff, and minimize soil erosion.

6.1.4 SOCIOECONOMIC Construction of the Prairie Island ISFSI is scheduled to be performed by local construction forces wherever possible. Relocation of construction personnel and their families is therefore not expected. A peak construction force of about 20 workers, including all employees of contractors and their subcontractors, is anticipated. The additional work force-required during construction will not be of sufficient size or their stay of sufficient duration to affect the basic socioeconomic characteristics of the local area.

6.1.5 RADIOLOGICAL IMPACTS FROM CONSTRUCTION All construction activities related to site preparation will have been completed prior to the commencement of fuel transfer. Ambient radiation levels at the construction site do not differ significr.ntly from average background levels in the area. Radiological impacts from construction acti,vities~are considered to be negligible, pg 2

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6.2 OPERATIONAL IhfPACTS 6.2.I RADIOLOGICAL IMPACTS FROM ROUTINE OPERATIONS External exposure to direct and scattered radiation is the primary pathway through which site workers and nearby residents may get a dose commitment from normal operation of the ISFSI. Because the proposed ISFSIinvolves only dry storage of spent nuclear fuel, there will be no ;;aseous or liquid effluent associated with normal operations. Cask loading and decontamination will be conducted within the Prairie Island Nuclear Generating Plant Auxiliary building, and are conducted under the 10 CFR Part 50 operating license. Radio-logical impacts from gaseous and liquid effluent resulting from these operations fall within the scope of impacts of reactor operations and have been previously addressed in the Final Environmental Statement for the Prairie Island Plant (Reference 2).

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6.2.I.1 OFFSITE DOSE ISFSI operations will result in a very small additional dose to members of the public from direct radiation exposure. Section 72.104(a) of 10 CFR Part 72 requires that dose equiva-lents 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. Using conservative assumptions in the ISFSI Safety Analysis Report (Reference 6), the dose to the nearest pumeigre resident from ISFSI operations, in combination with the maximum p,

permissible dose',hom the Prairie Island Nuclear Generating Plant, will not exceed the 25 l

.ps-I mrem per year Emit specified in 10 CFR 72.104.

I In calculating the offsite collective dose, the entire population within a 2 mile radius of the plant was conservatively taken to be at the location of the residence subject to the highest exposur,. The residence of highest exposure was found by determining the dose rate to the l

31 I

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residence nearest to the spent fuel cask storage source. The nearest residence, which is shielded by the 17-foot berm, was determined to be.45 miles northwest of the ISFSI.

The dose rate was calculated assuming that at distances beyond 800 meters, the dose rate falls off inversely with the square of the distance. The dose rate resulting from cask storage to the nearest residence NW of the ISFSI was calculated to be 9.0E-06 mrem / hour.

The dose to the population was obtained by taking the dose rate (in rem /hr) for the NW sector and multiplying it by the total population within a 2-mile radius of the plant to obtain (person-rem)/ hour and then multiplying by 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> / year to obtain (person-rem)/ year.

Because of the conservative assumptions in this dose calculation, the dose to the population residing within 2 miles of the plant, placed at the residence exposed to the highest dose rate, adequately estimates the population dose.

Currently, there are 464 residents within 2 miles of the Prairie Island Nuclear Generating Plant. The total annual population direct dose from cask storage was calculated to be 0.037-person-rem. Based on a dose rate of 9.0E46 mrem / hour, the annual dose to the nearest permanent resident due to ISFSI operations has been conservatively calculated to be 8.0E-02 mrem / year. The maximum annual dose to the nearest resident from the Prairie Island Nuclear Generating Plant has been calculated to be 0.0027 mrem / year due to liquid effluents and 0.334 mrem / year due to gaseous effluents (Reference 6). The total from both the plant and the ISFSI would therefore be less than 1.0 mrem / year. The conservative calculation of dose to residents within 2 miles and the rapid attenuation of neutron and gamma dose rates with distance make the collective doses for the more distant population negligible.

e 6.2.1.2 COLLECTIVE OCCUPATIONAL DOSE Spent fuel storage at the Prairie Island ISFSI will result in a small increase in the total occupational dose at the site. Occupational radiation exposure for ISFSI operations is expected to result during loading, transport and emplacement of the casks, and from surveillance and maintenance activities. One-time exposures which occur during cask 32

loading are expected to result in a radiation exposure of 2.315 person-rem. Annual surveillance, assumed to be done four times per year, will contribute.16 person-rem. Other maintenance operations contribute.604 person-rem to the total collective occupational dose.

The onsite collective dose has been assessed by estimating the number of personnel required to perform specific tasks, the time required to do them, and the estimated radiation levels in the areas in which the tasks are performed. Table 6.1 gives the estimated maximum collective occupational doses from one-time exposures during cask loading, transport and emplacement. These operations will be performed each time one of the casks is filled.

Table 6.2 shows the annual exposures due to ISFSI maintenance operations.

To evaluate the additional dose to station personnel from ISFSI operations, a conservative analysis has been performed. All workers at the Prairie Island Nuclear Generating Plant are considered to be in buildings or in the plant yard. No credit is taken for shieldir.g of personnel by buildings. Dose rates at various site locations were conservatively calculated -

based on the distance from each source using the East-West directed source dose rate versus distance data.

Table 6.3 shows the dose rates at several onsite locations due to cask storage. Distant dependent dose rates at each location are calculated by summing the direct shine and skyshine components. The berms have been shown to essentially climinate the direct radiation dose component (Reference 6). Table 6.4 shows the number of station personnel, and Table 6.5 presents the annual collective exposure estimates to onsite personnel not directly involved iin ISFSI activities.

g;:u..

}%f Table 6.6 sho&%

ws tis collective occupational dose to those workers who are directly involved in ISFSI activities. The surveillance and maintenance activities are conservatively assumed to be performed twice as often during the first year, resulting in a larger dose for these activities during that year. The cumulative dose (in person-rem) is shown over the 21 years it takes to completely fill both pads with 48 casks.

33

4 TABLEfl DESIGN BASIS OCCUPATIONAL DHE TIME EXPOSURES DURING CASK L.OADING. TRANSPORT AND EMPLACEMENT 8 Task Time No. of Dose Rate Dose Required (hr) persons (mrem /hr) (Person-rem)

Placement in pool 2 2

3 5.0 0.03 Imding process 5

5 5.0 0.125 Removal from pool 5

5 30.0 0.75 Transfer to decontamination area 1

3 30.0 0.09 Processing of cask 6.5 2

30.0 0.39 Helium leak test 2

2 30.0 0.12 Decontamination 2

3 30.0 0.18 Install neutron shield, pressurize, test 3

2 30.0 0.18 Preparation for transport 1

3 30.0 0.09 Transfer of cask to ISFSI 1

3 20.0 0.06 Final cask emplacement 2

5 30.0 0.30 TOTAL 2.315

• .1]G..

3;Y 2

\\

1

' Dose rates at 1 meter were utilized for all cases except cask transfer, when individuals will-typically be at least 2 meters away from the cask.

2Steps from Table 5.3.

34

e TABLE M DESIGN BASIS ISFSI MAINTENANCE OPERATIONS ANNUAL EXPOSURES Task Time No.of Dose Rate Annual Dose Required (hr) persons (mrem /hr) (Person-rem)

Visual Surveillance of 1

2 78.8 0.16 Casksi Instrumentation 1

2 1.0

.002 23 Operability Tests Instrumentation 2

2 1.0

.002 Calibration 53 Instrumentation 1

2 118 0.24 Repairs

Surface defect repair.:

1 2

118 0.24 s

Major Maintenance

• 32.5 3

25.6 0.12 TOTAL 0.764 Notes:

1. Assumes 4 yearly surveys,15 minutes each, no closer than 2 meters to cask.

2. Based on two tests per you,30 minutes each.

3. Based on re-calibration of the instruments every 2 years (annualized).

4. Assumes repair of one instrument every year, I hour per repair.

5. Assumes repair of one cask overy year,1 bour per repair.

6. Assumes once in 20 years (total dose of 2.5 person-rems is annualized by dividing total does by 20 years)

7. Assumed to k st imamisaring panel at the perimeter fence entrance.

~

8. Assumed tow between 2 rows of casks.

~

~ yd.y.

1 fi"'h p

.y -

i 1

35

)

TABLE L3 DDSE RATES AI ONSITE LOCATIONS DUE IQ CASK STORAGE

• Location Distance (Ft.)

Dose Rate from cask (mrem /hr)

Administration Building 1517 1.60E-04 Training Building 773 2.43E-03 NPD Building 835 1.89E-03 Construction Warehouse B 733 2.86E-03 Parts Warehouse 1288 3.40E-04 Environmental Lab 1876 5.00E-05 Computer Area 1797 7.00E-05 Outage Trailers 1512 1.60E-04 Security Building 1132 6.10E-04 Power House 2161 2.00E-05 Substation 2098 2.00E-05 Construction Warehouse A 1124 6.20E-04 5yQ r.t 6:l h@ih.

s..

i

• Dose rates were conservatively calculated based on the distances fmm each source using the East-West directed source dose rate versus distance data. Air attenuation was not taken into account in developing these estimates.

36

TABLE 6d NUMBER QE STATION PERSONNEL Location Summer Number location Full Time' Outage Help' 2

1 Administration Building 190 0

0 2

Training Building 55 0

0 3

NPD Building 120 0

0 4

Construction Warehouse B 15 45 0

5 Parts Warehouse 6

0 0

6 Environmental Lab 2

0 2

7 Computer Area 14 0

5 8

Outage Trailers 3

40 0

9 Security Building 117 25 0

10 Power House 92 192 0

11 Substation 2

2 0

12 Constniction Warehouse A 15 45 0

TOTAL 631 349 7

Notes:

1. For full time employees, assume 2500 hours0.0289 days <br />0.694 hours <br />0.00413 weeks <br />9.5125e-4 months <br /> / year.

2. For outage employees, assume 540 hours0.00625 days <br />0.15 hours <br />8.928571e-4 weeks <br />2.0547e-4 months <br /> / year.

3. For summer help, assume 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> / year.

37 j

i

)

TABLE M ANNUAL COI TFCITVE EXPOSURE ESTIMATES TD ONSITE PERSONNEL DlTE ID CASK STORAGE Location Full Time Outage Summer Total No.

Location (person-(person-Help (person-rem) rem)

(person-rem) rem) 1 Administration Building 7.60E-02 0.00E+00 0.00E+00 7.60E-02 2

Training Building 3.34E-01 0.00E+00 0.00E+00 3.34E-01 3

NPD Building 5.67E-01 0.00E+00 0.00E+00 5.67E-01 4

Construction Warehouse B 1.07E-01 6.95E-02 0.00E+00 1.77E-01 5

Parts Warehouse 5.10E-03 0.00E+00 0.00E+00 5.10E-03 6

Environmental Lab 2.50E-04 0.00E+00 4.00E-05 2.90E-04 7

Computer Area 2.45E-03 0.00E+00 1.40E-04 2.59E-03 8

Outage Trailers 1.20E-03 3.48E-03 0.00E+00 4.66E-03 9

Security Building 1.78E-01 8.23E-03 0.00E+00 1.85E-01 10 Power House 4.60E-03 2.07E-03 0.00E+00 6.67E-03 11 Substation 1.00E-04 2.16E-05 0.00E+00 1.22E-02 12 Constmetion Warehouse A 2.33E-02 1.51E-02 0.00E+00 3.84E-02 1

TOTAL 1.30E+00 9.48E-02 1.80E-04 1.40E+00 i

6

+ ~/5,

e.)

its'-

x l

i 38 1

TABLE 6.6 COLLECTIVE OCCUPATIONAL DOSE TO PRAIRIE ISLAND WORKERS DIRECTLY INVOLVED IN ISFSI ACTIVITIES (PERSON-REM)

1. of New Visual General Cumula-Casks Cask Surveil-Mainte-Total at tive Year Loaded Loading' lance 2 2

nance Year End Total 1

8 18.52

.32 1.21 20.05 20.05 2

2 4.63

.16

.604 5.39 25.44 3

2 4.63

.16

.604 5.39 30.84 4

2 4.63

.16

.6(M 5.39 36.23 5

2 4.63

.16

.604 5.39 41.63 6

2 4.63

.16

.604 5.39 47.02 7

2 4.63

.16

.604 5.39 52.41 8

2 4.63

.16

.604 5.39 57.81 9

2 4.63

.16

.604 5.39 63.20 10 2

4.63

.16

.604 5.39 68.60 11 2

4.63

.16

.604 5.39 73.99 12 2

4.63

.16

.604 5.39 79.38 13 2

4.63

.16

.604 5.39 84.78 14 2

4.63

.16

.604 5.39 90.17 15 2

4.63

.16

.604 5.39 95.57 16 2

4.63

.16

.604 5.39 100.96 17 2

4.63

.16

.604 5.39 106.35 18 2.

4.63

.16

.604 5.39 111.75 19 2

4.63

.16

.604 5.39 117.14 20 2

4.63

.16

.604 5.39 122.54 21 2

4.63

.16

.6(M 5.39 127.93 TOTAL 48 111.12 3.52 13.29 127.93

' Occupational exposure of 2.315 person-rem per cask.

' Exposure based on fully loaded pad, all actwities assumed to be performed twice as oAen during first year.

39

i i

6.2.2 RADIOLOGICAL IMPACTS OF OFF-NORMAL EVENTS AND ACCIDENTS A variety of off-normal and accident scenarios which may affect the safe operation of the Prairie Island ISFSI have been postulated by the applicant. These include earthquakes, tornadoes, tornado missiles, lightning, fires, loss of electric power, extreme wind, floods, explosions, inalvertent loading of a newly discharged fuel assembly, cask seal leakage, cask drop and tipping, and loss of confinement barrier.

Of the off-normal operations, only loss of electric power is considered to be applicable to ISFSI operations. A loss of power to the ISFSI may occur as a result of natural phenomena, such as lightning or extreme wind, or as a result of disturbances in the non safety-related portion of the electric power system of the Prairie Island Nuclear Generating Plant. If electric power is lost, the area lighting and receptacles and the cask pressure monitoring instrumentation would be nonfunctional. If the loss of power were localized solely at the ISFSI, detection would occur during periodic surveillance of the site. This event has no safety or radiological consequences because a loss of power will not affect the integrity of the storage casks, jeopardize the safe storage of the fuel, or result in radiological releases.

The consideration of the set of infrequent events which could be expected to occur during the lifetime of the ISFSI (Design Events III and IV), provides a conservative basis for the design of certain systems with confinement features.

The design earthquake (DE) is postulated to occur as a design basis extreme natural phenom-x.

~

enon. The D Ar use in the design of the casks and ISFSI structures is equivalent to the safe shutdow/huake for the Nuclear Generating Plant. Analyses of seismic response

• W.

characteristics'of the casks show that cask leak-tight integrity is not compromised and that no damage will be sustained. The DE is not capable of damaging the cask, and therefore no radioactivity is released.

40

O Extreme winds due to passage of the design tornado are postulated to occur as an extreme natural phenomenon. Tornado loading consists of a differential pressure buildup from normal atmospheric pressure to 3 psi in 3 seconds, a lateral force caused by a funnel of wind-having a peripheral tangential velocity of 300 mph, and a forward progression of 60 mph.

Extreme winds are not capable of overturning these casks nor of damaging their seals. Since no radioactivity is released, no resultant doses will occur. Imcal damage to the neutron shield may be caused by tornado missiles; however, the dose rate at the site boundary, without any shield is less than the allowable dose rate. Corrective actions would be utilized-to keep the dose rate low.

The probable maximum flood has been calculated to reach a level of 703.6 feet above msl with wave action to a maximum level of 706.7 feet. The casks are designed to withstand the forces developed by the probable maximum flood without damage to cask integrity or tipping of the casks. The height of the cask seals will be above the level of the probable maximum flood and associated wave action. No fuel damage or criticality is postulated to occur as a result of flooding, and no resultant doses are projected.

A munitions barge explosion has been postulated to occur at a location approximately 2600 feet from the ISFSI. This occurrence represents the worst-case impact on safe operation of the ISFSI due to a transportation accident. An overestimate of the blast effect is given by assuming that the impacts calculated at the reactor control room would also occur at the ISFSI, which is farther away from mid-channel. A pressure wave of 2.25 pai is estimated to occur at the ISFSI. The cask is designed to withstand a pressure wave of 3 psi. It will not tip as a result the postulated pressure wave, and no cask damage or release of radioactivity is postulated. E The only combustible materials in the ISFSI are in the form of insulation on instrumentation wiring, and paint on the outside surface of the storage casks. In addition, the tow vehicle will contain a small amount of gasoline or diesel fuel. No other combustible or explosive materials are allowed to be stored on the ISFSI slabs. The ISFSI atea will be cleared of

~

'l u

41

trees. The area surrounding the Equipment Storage Building and concrete pad within the perimeter road will be covered with crushed rock. No fires, other than small electrical fires, are considered credible at the ISFSI. The casks are designed to withstand this kind of fire.

No radioactivity is relemi, and therefore no resultant doses would occur.

The possibility of a spent fuel assembly with a heat generation rate greater than 0.675 kW being inadvertently loaded into the cask has been considered. In order to preclude this accident from going undetected, a final verification of the assemblies loaded into the casks and a comparison with fuel management records will be performed to ensure that the loaded assemblics do not exceed any of the specified limits. Due to the multiple administrative controls in selecting fuel assemblies, this accidental loading is not considered credible.

In order to prevent cask seal leakage from occurring, the storage casks feature redundant seals together with an extremely rugged body design. Additional barriers to prevent release of radioactivity include the sintered fuel pellet matrix and the zircaloy cladding which surrounds the fuel pellets. The interseal gaps are pressurized in excess of the cask cavity pressure. Although no credible mechanism that could result in leakage of radioactive products has been found, a complete loss of the cask seal capability has been analyzed and the results found to be negligible.

In an accident where the confinement function is non-mechanistically removed, heat removal and radiation shielding functions operate in the normal passive manner. In the event of broken cask seal barriers or removal of the closure lids, no release occurs. If the cladding in the loaded fuel m'amemblies fails, there is gap activity release. If the fuel pellets themselves fail, the remalidng Kr-85 is released from the fuel matrix. Table 6.7 lists the fission gas and wa volatile nuclides in a cask. Of the nuclides present in a cask containing 40 design basis fuel assemblies, Kr-85 is the only one naturally occurring in a gaseous state and which could escape from the cask in a breach of confinement barrier. No additional credit is taken for decay of Kr-85 during dispersion offsite nor for personnel protection due to shielding -

provided by any structure or system.

42 i

~

In the po***~I acciderit, all of the Kr-85 gas is assumed to be instantaneously released.

The maximally exposed individual is assumed to be located at the site boundary where the least amount of atmospheric dispersion takes place (largest x/Q value). The dose results for this location Are conservative for any individual and may be reported as dose to an individual i

at the nearest boundary. Table 6.8 lists the downwind dispersion factors. These were calculated using the Briggs formula for lateral and vertical plume spread. The equations for atmospheric diffusion found in the SAR are based on Regulatory Guide 1.145 which result in more conservative x/Q values.

Tables 6.9 and 6.10 summarize the expected doses at the site boundary in a containment failure where 30 percent of the K-85 and 10 percent of the H-3 inventory is released. The release fraction estimates for particulate radioactivity (i.e., Cs-134 and Cs-137) used in this analysis were based rn a worst-case scenario for air-cooled transfer casks (Reference 18).

This reference is expected to provide a reasonable assumption of the result of a non-mechanittic failure of the cask seal. Particulate releases clearly contribute a very small amount to the radiation dose.

After the radioactive material escapes the cask, two factors determine whether the particles reach the population: the fraction that becomes suspended in air; and the fraction less than 10 microns in diameter, which is respirable. An atmospheric dispersion value was used to calculate a dose at the nearest controlled area boundary and residence.

The calculated whole body dose at the boundary is 435 mrem and 11 mrem at the nearest residence. 'I1Nka small fraction of the 5000 mrem (5 rem) criteria specified in 10 CFR 72.106(b).

doses are also much less than the Protective Action Guides established by the Environmental Protection. Agency (EPA) for individuals exposed to radiation as a result of accidents: 1000 mrem to the whole body and 5000 mrem to the most severely affected organ.

l

)

43

.,,.-.yv.-.

The release of effluents from the ISFSI due to accidents, even a postulated worst-case accident, will have a negligible impact on the population surrounding the Prairie Island Nuclear Generating Plant.

TABLEfd FISSION GAS AND VOLATILE NUCLIDES INVENTORY (CURIES /40 ASSEMBLIES)

Nuclide 10 year Decay 20 year Decay H-3 6.46E+03 3.68E+03 KR-85 9.67E+04 5.06E+04 Cs-134 1.35E+05 4.68E+03 Cs-137 1.7E+06 1.35E+06

.g.

$l' 4

Westinghouse OFA 14x14,3.85 w/o U 235,45,000 MWD /MTU, 10 year cooling 44 i

TABLE 18 DOWNWIND DISPERSION FACTORS Downwind Section Downwind Distance' x/Q (m)

(sec/m' N

1015 1.95E-04 NNE 905 2.20E-04 NE 650 3.11E-04 ENE 550 4.92E-04 E

550 4.92E-04 ESE 775 2.59E-04 SE 795 2.52E-04 SSE 445 7.27E-04 S

345 1.17E-03 SSW 255 2.08E-03 1

SW 255 2.08E-03 WSW 195 3.49E-03 W

180 4.08E-03 l

W*

110 1.07E-02 WNW 195 3.49E-03 NW 245 2.25E-03 NNW 1055 1.87E l 2 4 -..* All distances and x/Q values are calculated froin the ceofer of the site AiDf#except the 110 meter distance west which is from the edge of the pad

nearest to the site boundary.

+

c:

f 45 1

i

7. J TABLE 63 l

EXPECTED DOSE DE M CONTROLLED AREA BOUNDARY RESULTING FROM A DEX CASK LEAKAGE ACCIDENT M M PRAIRIE ISLAND ISFSI WHOLE D_QSE M BOUNDARY Nuclide Cask Release x/Q Breathing Whole Body Dose at Inventory Fraction (sec/m')

Rate DCF2 Boundary (uCi)

(m'/sec)

(Rem /uCi)

(Rem)

H-3 6.46E+09 1.00E-01 1.07E-02 2.54E-04 1.58E-04

.277-Kr-85 9.67E+ 10 3.00E-01 1.07E-02 1.00E+00 5.10E-10'

.158 Cs-134 1.35E+ 11 5.00E-10 1.07E-02 2.54E-04 9.10E-02 1.67E-05 Cs-137 1.70E+ 12 5.00E-10 1.07E-02 2.54E-04 5.35E-02 1.24E-04 TOTAL

.435 Rem 810 Year cooled fuel 2 Reference 20

$ Whole Body submersion dose conversion factor in rem-m'/uCi-sec (Ref. 20)

TABLE filQ EXPECTED DQSE M M NEAREST RESIDENCE RESULTING FROM A DRY CASK LFAKAGE ACCIDENT M M PRAIRIE ISLAND ISFSI WHOLE BODY DOSE M RESIDENCE Nuclide Cask Release x/Q Breathing Whole Body Dose at Inventory Fraction (sec/m')

Rate DCP Boundary (uCi)'

(m'/sec)

(Rem /uCi)

(Rem)

H-3 (OEE+09 1.00EW1 2.78E-04 2.54E-04 1.58E-04

.0072 i

Kr-85 9 N +10' 3.0(E-01 2.78E-04 1.00E+00 5.10E-10' 4.11E-03 Cs-134 l'.

Eli!

5.00E-10 2.78E-04 2.54E-04 9.10E 02 4.34E-07 Cs-137 1.70E+ 12 5.00E-10 2.78E-04 2.54E-04 5.35E-02 3.21E-06 TOTAL

.011 Rem 10 Year cooled fuel 2 Reference 20

' Whole Body submersion dose conversion factor in rem-m'/uCi-sec (Ref. 20) 46

6.2.3 NONRADIOLOGICAL IMPACTS 6.2.3.1 LAND USE AND TERRESTRIAL RESOURCES s

Gperation of the ISFSI will not require the use of any land beyond that which was cleared and graded during construction, and is not expected to adversely impact the terrestrial environment. Operation of the ISFSI will have a minimal impact on the local wildlife. Birds are not expected to roost directly on the casks due to their high surface temperature. A fence surrounding the concrete storage pads will prevent access by other wildlife.

6.2.3.2 WATER USE AND AQUATIC RESOURCES The operation of the ISFSI requires no active water cooling system; therefore, there will be i no impact on surface or ground water quality or aquatic biology.

6.2.3.3 OTIIER EFFECTS OF OPERATION Noise The only operational noise associated with the proposed action will result from the transfer of spent fuel from the spent fuel pool facility to the dry cask storage facility. Since the noise associated with this operation is not expected to be louder than normal truck traffic, no adverse impacts are expected, f

Climatology g'.

po The surface of the storage casks may approach 240*F. This will cause the air temperature in the immediate vicinity of the casks to be higher than ambient temperature. The affected area is very small and localized. During rainy days, precipitation may vaporize at the cask surface because of these high cask surface temperatures. Any cask-induced fogging episodes 47

T will have the greatest impact at the locations where visibility is important. The county raad to the west of the ISFSI site and the nearest residence,.45 miles NW, were chosen to represent the impacts along the road. Using the EPA Industrial Source Complex Dispersion Model, the two pads were modeled as an area source with a water vapor plume release height of 16 feet. Many conservative assumptions were included in the analysis. The results indicate that the fogging impacts due to the ISFSI casks at the county road and nearest residence would occur.04 percent of all hours during the May-October period and 0.2 percent of all hours during the November-April period.

23 SAFEGUARDS FOR SPENT FUEL he NRC requirements for 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.

On March 10, 1992, the applicant submitted to the NRC a Physical Security Plan which incorporates measures presently in effect for the protection of the Prairie Island Nuclear Generating Plant, and additional safeguards specifically for the spent fuel. These include the following:

Barriers to limit unauthorized access to the ISFSI, e

Access, for personnel, vehicles, and packages, w&!:

a:

Search requirements to detect contraband materials, e

Detection and assessment capability for all alarms, e

Site-specific trainmg for security force members, e

48

O C

Pre-planned contingency events and security actions, Commitments for responding to unresolved alarms, Provisions for obtaining support from the local law enforcement agency, e

Secure transportation of the spent fuel from the reactor site to the ISFSI.

o The implementation of these physical security plans will be inspected for effectiveness and operational compliance.

An independent safety review of the cask design is being conducted by 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 for nuclear explosives is not considered-credible due to the massive size and construction of the cask, the unattractive form of the enclosed radioactive material, and the 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 special nuclear material.

8.0 DECOMMISSIONING

m,.

$k The ease of deqommissioning of the storage casks to be utilized at the Prairie Island ISFSIis

s..

one feature of ^the design concept. A decommissioning plan must be submitted in accordance with 10 CFR 72.30. At the end ofits service lifetime, cask decommissioning could be accomplished by one of the following options:

49

d

1. The loaded storage cask could be shipped to a suitable fuel repository for permanent storage. If licensing requirements at the time allow, the entire cask could be placed inside a shipping container or overpack for shipment.

2. The spent fuel could be removed from the cask and shipped in a certified shipping container to a fuel repository. The cask could then be decontaminated and scrapped.

3. The surface of the ISFSI cask can be decontaminated by chemical etching with hydrochloric or nitric acids, or electropolished to achieve the same results.

h The cask materials will be only slightly activated from the spent fuel, and it is expected that after surface decontamination, the activation products will be negligible and the cask could be scrapped. A detailed evaluation will be performed at the time of decommissioning to i

determine the appropriate mode of disposal.

Due to the leak tight design of the casks, no residual contamination is expected to be left behind on the concrete base pad. The spent fuel pool will remain functional until the ISFSI is decommissioned. This will allow the pool to be utilized to transfer fuel from the storage casks to licensed shipping containers for shipment offsite if this decommissioning option is chosen.

9.0

SUMMARY

AND CONCLUSIONS SUMMIILY OF EhVIRONMENTAL IMPACTS 9.1

% W :/

+v.c As discussed in Section 6.1, no significant construction impacts are anticipated. The activities will affect only a very small fraction of the land area at the Prairie Island Nuclear Generating Plant. With good construction practices, the potentials for fugitive dust, crosion and noise impacts, typical of the planned construction activities, can be controlled to insignificant levels.

50

The primary exposure pathway associated with the ISFSI operation is direct radiation of site workers and nearby residents. As discussed in Section 6.2.1, the radiological impacts from liquid and gaseous effluent during normal operation of the ISFSI fall within the scope of impacts from licensed reactor operations.

The dose to the nearest resident from ISFSI operation is about 8.0E-02 mrem / year, and when added to that of the operations of both reactor units, is much less than 25 mrem / year, as required by 10 CFR 72.104. The collective dose to residents within 1 to 2 miles of the ISFSIis.037 person-rem. Occupational dose to site workers, both directly and indirectly involved in ISFSI activities, is a small fraction of the total occupational dose commitment.

The gamma dose to an individual at the controlled area boundary from a loss of confinement accident has been calculated to be.435 rein, which is well within the 5 rem criteria set forth-in 10 CRF 72.106(b) and less than the EPA Protective Action Guide of 1 rem.

No significant nonradiological impacts are expected during operation of the ISFSI. The heat given off by the casks has been determined to cause an insignificant amount of cask induced fogging. No other effects are anticipated in the immediate vicinity of the ISFSI.

9.2 BASIS FOR FINDING OF NO SIGNIFICANT IMPACT The proposed action has been reviewed relative to the requirements set forth in 10 CFR Part 51, and hased on this assessment, the NRC has determined that issuance of a materials license undery'CFR Part 72 authorizing storage of spent fuel at the Prairie Island ISFSI c-19 h ~ ffect the quality of the environment. Therefore, an environmental will not a

impact at=W.,is not warranted and, pursuant to 10 CFR Part 51.31, a Finding of No Significant Impact is appropriate.

1 51 J

my:

10.

REFERENCES 1.

U.S.- Nuclear Regulatory Commission, " Final Generic Environmental Impact State.

ment 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 the Prairie Island Nuclear Generating Plant," May 1973.

3.

Northern States Power, " Prairie Island Independent Spent Fuel Storage' Installation Environmental Report," Docket Number 72-10, Revision 1, September 1991.

4.

Northern States Power, letter from Thomas Parker, " Responses to NRC Questions j

Regarding the Prairie Island Independent Spent Fuel Storage Installation Technical Specifications and Safety Analysis Report," June 5,1991.

5.

Northern States Power, " Prairie Island Nuclear Generr. ting Plant Updated Safety t

Analysis Report," Revision 8, December 1989.

i 6.

Northern States Power, " Prairie Island Independent Spent Fuel Storage Installation -

Technical Specifications and Safety Analysis Report," Docket Number 72-10, Revision 2, September 1991.

~ 7.

Minnesota Environmental Quality Board, " Final Environmental Impact Statement on the Praiiie Island Independent Spent Fuel Storage Installation," April 12, 1991-

.r 8.

Northern States Power, "Prairic Island, Docket No. 72-10,- Application for a License :

to Construct and Operate a Dry Cask Independent Spent Fuel Storage Installation,"

August 31,1990.

52

W 9.

U.S. Nuclear Rel;ulatory Commission, " Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-level Radioactive Waste," 10 CFR Part 72.

10.

U.S. Nuclear Regulatory CommisQn, " Environmental Protection Regulations for Domestic Licensing and Related R.egulatory Functions," 10 CFR Part 51.

I1.

U.S. Nuclear Regulatory Commission, " Standards for Protection against Radiation,"

10 CFR Part 20.

12.

U.S. Nuclear Regulatory Commission, " Applicability of Existing Regulatory Guides to the Design and Operation of an Independent Spent Fuel Storage Installation,"

Regulatog Guide 3.53, July 1982.

13.

U.S. Nuclear Regulatory Commission, " Preparation of Environmental Reports for Nuclear Power Stations," Regulatory Guide 4.2, Revision 2, July 1976.

14.

American National Standards Institute /American Nuclear Society, " Design Criteria for an Independent Spent Fuel Storage Installation (Dry Storage Type), " ANSI /ANS-57.9,

1984, 15.

U.S. Nuclear Regulatory Commission, " Assumptions Used for Evaluating the Potential Radiological Consequences of a less of Coolant Accident for Boiling Water Reactors," Regulatory ^ Guide 1.3, June 1974.

pg, ~

'y

-(

16.

Prairie l Nuclear Generating Plant Annual Reports.

17.

Northeni States Power, letter from Thomas Parker, " Responses to NRC Questions Regardmg the Prairie Island Independent Spent Fuel Storage Installation, Quality Assurance and Cask Thermal Analysis," February 6,1992.

1 i

53

't 18.

Wilmot, Edwin C., " Transportation Accident Scenarios for Commercial Spent Fuel,"

SAND 80-2124, Sandia National Laboratory, Albuquerque, NM, February 1981.

19.

Environmental Protection Agency, Federal Guidance Report #11, EPA 520,1-884)20.

i l

20.

U.S. NRC " Calculation of Annual Doses to Man from Routine Releases of Reactor 3

Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Reg Guide 1.109, October 1977.

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54 1

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