ML20211H746
| ML20211H746 | |
| Person / Time | |
|---|---|
| Site: | Surry, 07200002  |
| Issue date: | 06/06/1986 |
| From: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| Shared Package | |
| ML20211H753 | List: |
| References | |
| FOIA-86-542 SECY-86-174, NUDOCS 8606260026 | |
| Download: ML20211H746 (83) | |
Text
f ..m e A E #
J""e e,1988 POLICY ISSUE s,cy_,,_1,4 (Notation Vote)
For: The Commissioners Victor Stello, Jr. .
From: Executive Director for Operations
Subject:
PROPOSED LICENSE UNDER 10 CFR PART 72 FOR DRY CASK STORAGE OF SPENT FUEL AT VIRGINIA ELECTRIC AND POWER COMPANY'S (VEPCO) SURRY POWER STATION
Purpose:
To obtain Commission authorization pursuant to 10 CFR Sec-tion 2.764(c) for NMSS to issue an initial license authorizing VEPCO to possess spent fuel from Surry Units 1 and 2 in an independent spent fuel storage installation (ISFSI) on the Surry site using CASTOR V/21 storage casks. The ISFSI would consist of up to 3 concrete pads (one of which has been constructed) holding a total of 84 casks.
Summary: The staff has completed its safety, safecuards and environmental reviews for the storage of spent fuel in dry casks to be located on VEPCO's Surry Power Station site, approximately one-half mile from Surry Units 1 and 2. NMSS staff licensing reviews have been coordinated with NRR, IE and Region II staff. A proposed license (Enclosure A) has been prepared for signature by the Director, HMSS, or his designee, based upon the staff's Safety Evaluation Report (Enclosure B). Only spent PWR fuel from Surry Units 1 and 2 is to be stored ir nodular cast iron casks at the Surry site.
The storage activities and location are covered under the existing Price Anderson indemnity agreement for the Surry site. No offsite transportation of spent fuel is involved. Heavy loads consider-ations (Fuel Building authorized crane capacity is 125 tons) were addressed with respect to reactor dry cask handling by NRR staff in 1983 under amendments to the operating licenses of Surry Power Station's Units 1 and 2. An evaluation by VEPCO has shown that the dry storage activities do not present an unreviewed safety question for reactor operations and that no changes to the technical specifications of the reactor operating licenses are necessary.
Contact:
John P. Roberts ,
42-74205 s 1
hk(bbN '
l i The Commissioners 2 2 -
4 The independent spent fuel storage installation in this instance consists of a fenced area on the site, which is to contain up to 3 concrete pads which will hold a total of 84 storage casks. At its own risk VEPCO has constructed one storage pad (230 feet by 32 feet), which has been inspected by NRC Region II staff. It has purchased 5 nodular cast iron' storage casks of the CASTOR
! V/21 type from General Nuclear Systems, Inc. (GNSI). The cask-t vendor is a partnership involving Chem-Nuclear Systems, Inc., a
, United States firm, and Gesellschaft fur Nuklear Service abH (GNS), a West German firm. Each cask is designed to store 21 PWR
! assemblies, aged 5 years or more. A quality assurance audit of cask fabrication was performed by the Vendor Program Branch, IE,
] on GNS and its subcontractors in West Germany.
Discussion: In October 1982, NMSS staff received an application from. Virginia Electric and Power Company (VEPCO) for spent fuel storage in a dry ,
cask ISFSI to be located at the Surry Power Station site. Since i
VEPC0 had not chosen a specific cask design, its Safety Analysis j Report and Environmental Report were drafted to bound a potential i cask design. VEPCO stated that it would rely on submittal by cask
[ vendors of topical safety analysis reports for their respective
- designs and the safety reviews of these by NMSS staff, prior to
)
selecting one or more cask designs for its application. With the
' submittal'of the'GNSI CASTOR V topical safety analysis report (TSAR) in January 1984, VEPCO notified NRC in March 1984 by letter of its selection of the CASTOR V (subsequently designated CASTOR V/21) l design for its application. Subsequent revision of VEPCO's safety analysis report and updating of its environmental report were
- based on this design, i
In April 1985, the staff completed its environmental review for l the Surry site application. A notice was published in the Federal
- Reaister of a Finding of No Significant Impact (50 FRN 15517, l April 18, 1985) and an Environmental Assessment (Enclosure C) was j issued.
1 Relying on these actions, VEPCO in July 1985, submitted a request i to NRC to commence construction, at its own risk, of one of the 3 concrete pads for its ISFSI prior to issuance of a Part 72
- license. The NRC staff informed VEPC0 that it would not bar such
- construction (SECY-85-190), and the pad was constructed with l inspection performed by NRC Region II staff. VEPCO's action, l assuming subsequent granting of a license, was taken to allow
! VEPC0 to avoid shipment of spent fuel from its Surry site to its l North Anna plant site for storage there. The receipt and storage
- of up to 500 Surry assembifes at North Anna has been authorized l after hearings before an Atomic Safety and Licensing Board
, (L8P-85-34). However, this alternative is not, as VEPCO has made i clear, its preferred alternative, which is to provide onsite dry i cask storage capability at the Surry site for Surry spent fuel on j a schedule that will allow it to avoid losing full core storage -
1 1
--- v g - r r -vw g= .,v-e+
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I The Commissicners .
3 capacity in the Surry reactor storage pool. Loss of full core reserve will occur with the scheduled refueling outage of Surry Unit 2 this fall. VEPC0 has informed the staff that by June 2
1986, it must nake decisions regarding actions to be taken to i maintain full core reserve to avoid later interference with out-age activities. Early issuance of the requested license under '
10 CFR Part 72 would eliminate the need to transship fuel to its
, North Anna Power Station, an option that VEPCO and the local community wish to avoid.
In September 1985, NMSS staff completed its safety review of the .
TSAR for the CASTOR V/21 cask design and issued a letter of approval with an SER. .The staff reviewed criticality, structural, thermal, and shielding aspects of the cask design, including its fuel basket, under normal and accident conditions. The staff I
concluded that the CASTOR V/21 cask could be used to safely store, as proposed by the vendor, 21 PWR assemblies with 5 years or more decay since discharge (1 kW heat output per assembly, and maximum
- burnup of 35,000 MWO/MTU). Characteristics of the spent fuel
! allowed to be stored and other cask operating limits are set fprth in the safety evaluation report (SER) for the cask topical report.
1 These limits are carried forward in the safety evaluation for VEPCO's application and are included in the proposed a
license's (Enclosure A) Technica? Specifications.
Under provisions of the Nuclear Waste Policy Act of 1982, the j
Department of Energy and VEPCO have jointly supported an unlicensed research and development demonstration at the Idaho
- National Engineering Laboratory (INEL) of CASTOR V/21 cask han-i dling and the storage of 21 spent fuel assemblies. The fuel was supplied by VEPCO from its Surry site and shipped in an NRC cer-l tificated transport TN-8L cask to INEL. This demonstration (des-cribed in Enclosure D) covered, in particular, shielding and thermal aspects of the cask design and confirmed computer code
, analyses. It has provided the staff with added confidence that l the cask design description and the analyses evaluated by the
! staff and found acceptable censervatively address the CASTOR V/21 cask and its safe operation, as specified for the Surry Power i
Station site in the proposed license (Enclosure A). Photographs of the CASTOR V/21 are included with Enclosure 0 In October 1985, staff was informed by DOE, VEPCO and GNSI of cracks observed in non-structural welds in the CASTOR V/21 cask
. basket during the INEL demonstration. These were subsequently investigated. It was initially concluded that these cracks resulted from thermal stresses caused by an excessive heat load (beyond design limits specified in the cask topical report) and i dimensional tolerances inadequate to allow for basket expansion.
l While no safety significance was ascribed to this incident by !
' DOE, and it is not clear that such cracks could occur under !
design conditions, nevertheless NRC staff is re-evaluating the -
i
The Ccomissionsrs ,
4 ,
4 design of the cask basket which is constructed of a borated stainless steel not produced in the United States. The investi-gation by the vendor since October 1985, has raised questions regarding the material properties of the borated stainless steel, which have yet to be resolved. The vendor is performing. tests and analyses of the original basket design, which may req'uire modifications.
In the interim the vendor has submitted, and the staff has reviewed ,
and approved, an alternative design of a cask basket that is i fabricated of stainless steel without added boron content and for i which the initial enrichment of the stored fuel is appropriately I limited (12.2 weight percent U-235) to offset the absence of boron as a neutron absorber. VEPCO has informed the staff that it concurs with the inclusion of the alternative basket design in the CASTOR V/21, and that it has over 100 aged spent-fuel _
assemblies with suitably low enrichment for such storage. The proposed license authorizes VEPC0 to use the stainless steel 1 basket in the CASTOR V/21 cask, and upon the staff's determina-tion of the adequacy.of the vendor's design for a borated +
! stainless steel basket, VEPC0 may store higher initial enrich-3 l ment withoutspent further fuel (Tcense I amendment.<3.5 weight percent U-235) using this ba -
i VEPC0 has already purchased 5 CASTOR V/21 casks. One of these is being used, as mentioned above, to store Surry spent fuel at INEL in cooperation with the Department of Energy. Of the other 4 casks, o
the first is expected to arrive at the Surry site in adequate time to allow Surry's trained cask handling operators to familiarize themselves with transfer handling characteristics of CASTOR V/21 casks. .
The staff has completed its safety review of the Surry site appli-cation. The application in the applicant's safety analysis report (SAR) includes confirmation by the applicant's reactor safety committee that no technical specification changes are required '
under the Surry operating license to accommodate a Part 72 license for onsite storage, that the joint operation of the reactor and onsite storage does not affect the safety margins of either one,
~
L and that onsite storage is an independent operation as defined in Part 72. Based on the staff's review of the applicant's submis-sions, the Staff has found that there are no remaining unreviewed safety questions and that all pertinent regulatory requirements ,
for authorization of issuance of the requested license have been satisfied. Accordingly, the staff has issued its SER (Enclosure B) for Surry, which references the CASTOR V/21 SER, making the appro-priate findings. The staff has completed preparation of a pro-posed Part 72 license with technical specifications, including license conditions which satisfy safeguards requirements of Part 73, for Surry site spent fuel storage.
k
l l
The Commission:rs , 5 C nclusion: The staff believes that issuance of a license for dry cask spent fuel storage at VEPCO's Surry Power Station site is an action in accord with public health and safety, that it is an environmentally
, sound action and that it is in the public interest.
Recommendation: That the Commission authorize issuance of a license to VEPCO under 10 CFR Part 72 to receive, transfer and store spent fuel in dry casks, using stainless steel baskets for storage of low initial enrichment spent fuel, on the Surry Power Station site; with authorization to store higher initial enrichment spent fuel to become effective upon NMSS decision that the questions with respect to the use of a borated stainless steel basket design for the CASTOR V/21 cask have been satisfactorily resolved.
(
, a-2;fN. , /'
ctorStello,Jr.L/
. Executive Director for Operations
Enclosures:
A. Proposed License -
B. SER C. EA D. CASTOR V/21 Cask Storage Demonstration Experience l'
Commissioners' comments or consent should be provided directly to the Office of the Secretary by c.o.b. Tuesday, June 24, 1986.
l Commission Staff Office comments, if any, should be submitted to the Commissioners NLT Tuesday, June 17, 1986, with an infor-mation copy to the Office of the Secretary. If the paper is of such a nature that it requires additional time for analytical review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.
[ DISTRIBUTION:
! Commissioners OGC OPE OI OCA OIA
ASLAP SECY
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ENCLOSURE A
=
VIRGINIA ELECTRIC AND POWER COMPANY
-DOCKET NO. 72-2 SURRY INDEPENDENT SPENT FUEL STORAGE INSTALLATION MATERIALS LICENSE NO. SNM-2501 -
The Nuclear Regulatory Commission (the Commission) has found that:
A. The application filed by the Virginia Electric and Power Company (applicant) for a materials license to receive, store and transfer spent fuel from Surry Units 1 and 2 in an independent spent fuel storage installation (ISFSI) located at its Surry Power Station 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;
- 8. The Surry ISFSI will operate in conformity with the application, as amended, the provistons 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 Surry Power Station, Units 1 and 2; E. The applicant's proposed ISFSI design complies with 10 CFR Part 72, Subpart F; F. The applicant is qualified by reason of training and experience to conduct the operation covered by the regulations 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; 1
Enclosure A l
'H. The applicant is financially qualified to engage in the activities in l 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 I; L. The applicant's proposed decommissioning plan and its financing pursuant to 10 CFR $ 72.18 provide reasonable assurance that the decontamination and decommissioning of the Surry ISFSI at the end of its useful life will provide adequate protection to the health and safety of the public; M. The applicant's proposed emergency plan complies with 10 CFR $ 72.19; N. The applicant has satisfied the applicable provisions of 10 CFR Part 170;
- 0. There is reasonable assurance (1) that the activities authorized by the 4 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 I
and security or to the health and safety of the public.
l Accordingly, based on the foregoing findings, Materials License No. SNM-2501 is hereby issued to the Virginia Electric and Power Company to read as follows:
Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganiza-tion 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 hereto-fore made by the licensee (in the licensee's Safety Analysis Report, Surry _
t 2
e .
Power Station, Dry Cask Independent Spent Fuel Storage Instaliation, as sub-mitted by letter dated October 8, 1982, and as revised and submitted by letters dated January 25, 1983; October 25, 1983; June 40, 1985; and February 19,1986),
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 pur-poses and at the place designated below; to deliver or transfer such materials to persons authorized to receive these materials iri accordance with the regula-tions 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 I Act of 1954, as amended, and is subject to all applicable rules, regulations and orders of the Nuclear Regulatory Commission now or hereafter in effect and to any conditions specified herein.
Licensee
- 1. Virginia Electric and Power Company
- 3. License Nurier: SNM-2501
- 2. Address:
- 4. Expiration Date: May __, 2006
- 5. Docket Number: 72-2
- 6. Byproduct, source, and/or 7. Chemical and/or 8. Maximum amount that special nuclear material physical form licensee may possess at any one time under this license A. Spent Fuel assemblies from A. As UO2 clad with A. 811.44 Teu of Surry Unit 1 & 2 reactors zirconium or spent fuel using natural water for zirconium alloys assemblies -
cooling and enriched not greater than 3.5 percent 3
i U-235 and associated radio-active materials related to receipt, storage, and transfer of the fuel assemblies
- 9. Authorized Use:
The material identified in 6.A and 7.A above is authorized for receipt, possession, storage and transfer.
- 10. Authorized Place of Use:
The licensed material is to be received, possessed, transferred, and stored at the Surry ISFSI locate,d on the Surry Power Station site in Surry County, Virginia near Surry, Virginia.
- 11. This site is described in Chapter 2 of the licensee's Safety Analysis Report for the Surry ISFSI*. -
- 12. The Technical Specifications contained in Appendix A attached hereto are incorporated in the license. The licensee shall operate the installation in accordance with the Technical Specifications in Appendix A.
- 13. The Safeguards License Conditions contained in Appendix B attached hereto are hereby incorporated into this License. The licensee shall maintain a physical protection program for the ISFSI in accordance with those condi-tions.
- 14. The Technical Specifications for Environmental Protection contained in Appendix C attached hereto are incorporated in the license. The licensee l shall operate the installation in accordance with the Technical Specifications in Appendix C.
l l
- 4 t
- 15. This license is effective as of the date of issuance shown below.
For the U.S. Nuclear Regulatory Commission .
Date .
of Issuance by Material Safety Washington, DC 20555 "Hereafter referred to in this license as the SAR.
e I
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Appendix A .
VIRGINIA ELECTRIC AND POWER COMPANY SURRY INDEPENDENT SPENT FUEL STORAGE INSTALLATION
~
. TECHNICAL SPECIFICATIONS FOR SAFETY LICENSE SNM-2501 er t
TABLE OF CONTENTS .
P, age
1.0 INTRODUCTION
.................................................... A-1 1.1 Definitions ................................................ A-1 1.2 Preoperational Li cense Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . A-2 -
1.3 General Li cense Condi tions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '. . A-4 2.0 FUNCTIONAL AND OPERATING LIMITS.................................. A-5 2.1 Fuel To Be Stored At ISFSI ................................. A-5 2.2 GNS CASTOR V/21 D ry Storage Cas k . . . . . . . . . . . . . . . . . . . . . . . . . . . A-6 2.3 Dry Storage Cask Internal Cover Gas ........................ A-6 2.4 Dry Storage Cask Surface Contamination . . . . . . . . . . . . . . . . . . . . . A-8 3.0 LIMITING CONDITIONS ............................................. A-9 .
3.1 Limiting Condition - Handling Height ....................... A-9 3.2 Limiting Condition - Pressure Switch . . . . . . . . . . . . . . . . . . . . . . . A-9 4.0 SURVEILLANCE REQUIREMENTS ....................................... A-11 4.1 Cask Interlid Pressure (CASTOR V/21) ....................... A-11 4.2 Dose Rates ................................................. A-11 4.3 Alarm Board .......................................~......... A-13 4.4 Fuel Parameters ...~......................................... A-13 4.5 Cask Contamination ......................................... A-14 4.6 Cask Seal Testing .......................................... A-14 5.0 DESIGN FEATURES ................................................. A-16 5.1 Site ....................................................... A-16 5.2 Cask Design ..................*.............................. A-16 5.3 Storage Pad and Cask Arrangement ........................... A-16 5.4 Total Storage Capacity ..................................... A-16 iii m = wee e - o
O
- 1. 0 INTRODUCTION These Technical Specifications govern the safety of the receipt, possession and storage, of irradiated nuclear fuel at the Surry Dry Cask Independent Spent Fuel Storage Installation and the transfer of such irradiated nuclear fuel to and from the Surry Nuclear Power Station and the Surry Dry Cask Independent Spent Fuel Storage Installation.
1.1 DEFINITIONS The following definitions apply for the purpose of these Technical Specifications.
- a. Administrative Controls: Provisions relating to organization and management procedures, recordkeeping, review and audit, and reporting necessary to assure that the operations involved in the storage of spent fuel at the Surry ISFSI are performed in e. 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. Functional and Operatina 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.
- d. 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 rods held together by end fittings, spacers, and tie rods.
- e. Limitina Conditions: the lowest funtional capabilities or performance levels of equipment required for safe operation of the facility.
O A-1
-.m.y_
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- f. Surveillance Requirements: Surveillance requirements include:
(1) inspection, test and calibration activities to ensure that the necessary integrity of required systems, components und the spent t.
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.
- g. Tonne-(Te): One metric ton, equivalent to 1000 kg or 2204.6 lb. Fuel quantity is expressed in terms of the heavy metal content of the fuel measured in metric tons and written Teu.
- h. Loading Operations: Loading Operations include all cask preparation steps prior to cask transport from the fuel building area.
- 1. 2 PREOPERATIONAL LICENSE CONDITIONS The license issued under Part 72 shall not allow the loading of spent nuclear fuel until such time as the fbilowing preoperational license conditions are satisfied:
- 1. A training exercise (Dry Run) of all cask loading and handling activities shall be neld which shall include but not be limited to:
- 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 reactor.
- f. Unioading the cask assuming fuel cladding failure.
A-2
- g. . Decontaminating the cask.
- h. All cask handling shall be done.using written procedures.
- 2. The Surry Power Station Emergency Plan shall be reviewed and modified as required to include the ISFSI. (Abnormal event notifications will have to be updated for ISFSI events.)
- 3. A training module shall be developed for the Surry Power Station Training Program establishing an ISFSI Training and Certification Program which will include the following:
- a. Cask Design (overview)
- b. ISFSI Facility Design (overview)
I
- c. ISFSI Safety Analysis (overview)
- d. Fuel loading and cask handling procedures and abnormal' procedures
- e. ISFSI License (overview).
l
- 4. The Surry Power Station, health physics procedures shall be reviewed and modified as required to include the ISFSI.
- 5. The Surry Power Station Administrative Procedures shall be reviewed .
and modified as required to include the ISFSI.
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- 6. A procedure shall be developed for the documentation of the character-izations performed to select spent fuel to be stored in the casks.
- 7. Written operating and abnormal / emergency procedures shall be prepared.
A-3
1.3 GENERAL LICENSE CONDITIONS 1.3.1 Quality Assurance Activities at the Surry Dry Cask ISFSI shall be conducted in accordance with the requirements of Appendix B, 10 CFR Part 50, as described in the Virginia Power Topical Report, " Topical Report on Quality Assurance Program . Operating Phase." This progrei ts implemented through the Virginia Power Nuclear Power Station Quality Assurance Manual (NPSQAM).
1.3.2 Fuel and Cask Handling Activities Fuel and cask movement and handling activities which are to be performed in the Surry Power Station Fuel Building and Crane Enclosure Building will be governed by the requirements of the Surry Power Station Facility Operating Licenses (DPR-32 and DPR-37) and associated Technical Specifications.
1.3.3 Administrative Controls .
The Surry Dry Cask ISFSI is located on the Surry Power Station site and will be managed and operated by the Surry Power Station staff. The administrative controis-shall be in accordance with the requirements of the Surry Power Station Facility Operating Licenses (DPR-32 and DPR-37) and associated Technical Specifications.
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- A-4 ;
2.0 FUNCTIONAL AND OPERATING LIMITS 2.1 FUEL TO BE STORED AT ISFSI 2.1.1 Specification -
The spent nuclear fuel to be received and stored at the Surry Dry Cask ISFSI shall meet the following requirements:
(1) Only fuel irradiated at the Surry Power Station Unit Nos.1 and 2 may be used.
(2) Maximum initial enrichment shall not exceed 2.2 weight percent U-235 for fuel stored in the stainless steel basket reviewed and found acceptable. For higher enrichment fuel, up to 3.5 weight percent U-235, staff review and finding of acceptability for an appropriate basket must be completed prior to storage of such fuel in a CASTOR
. V/21 cask. .
(3) Maximum assembly average shall not exceed 35,000 megawatt-days per metric ton uranium.
(4) Maximum heat generation rate shall not exceed 1 kilowatt per fuel assembly.
(5) Fue'l shall have cooled a minimum of 5 years after reactor discharge and prior to storage in the Surry Dry Cask ISFSI.
l (6) Fuel shall be intact unconsolidated fuel.
(7) Prior to insertion of a spent fuel assembly into a cask, the identity of the assembly will be verified by an individual other than the one who previously identified the assembly.
A-5 yr -ter-w+r = = = - - + - --
2.1.2 Basis The design criteria and subsequent safety analysis of the Surry Dry Cask ISFSI assumed certain characteristics and limitations for the fuels that are to be received and stored. Specification 2.1.1 assures that these bases remain valid l by defining the source 'of the spent fuel, maximum initial enrichment, irradia-tion history, maximum thermal heat generation, and minimum post-irradiation I cooling time.
The radiological analyses are based on a radiation spectrum for 3.5 weight percent U-235 fuel at 35,000 MWD /MTU burnup (these analyses envelope the lower enrichment spent fuel, < 2.2 weight percent U-235, case), multiplied by a ,
factor of three. Compliance with the enrichment and burnup limits will ensure that the Dry Storage Casks design criteria are not exceeded.
2.2 GNSI CASTOR V/21 ORY STORAGE CASK 2.2.1 Specification .
The GNSI CASTOR V Dry Storage Casks used to store spent nuclear fuel at the Surry Dry Cask ISFSI shall have the operating limits shown in Table 2-1.
2.2.2 Basis The design criteria and subsequent safety analysis of the GNSI CASTOR V/21
( assumed certain characteristics and operating limits for the use of the casks.
l This specification assures that those design criteria are not exceeded.
l 2.3 ORY STORAGE CASK INTERNAL COVER GAS 2.3.1 Specification l
The dry storage casks shall be backfilled with helium.
A-6
___ . _ _ _ . . __ __ _~~~ _" " " ~ _ _ _ . _ _ _ _ . _ _ _ - _ _ _ . .
. Table 2-1 GNSI CASTOR V/21 OPERATING LIMITS .
Operating Limit Max. Lifting Height with a Non-Redur; dant Lifting Device 5' with impact limiters without impact limiters 15" Dose Rate '
2 m Distance S 10 arem/hr Surface 5 200 arem/h'r (These limits conform to transportation cask -
dose rate limits. Actual dose rates for the loaded CASTOR V/21 will be significantly less.)
Cask Tightness -
(Standard He-Leak Rate) ,
Primary Lid Seal 5 10.s mbar 1/s
' Secondary Lid Seal 5 10 s abar 1/s Max. Specifid Power of One 1.0 kW Fuel Assembly .
Max. Cladding Temperature during All Phases of Operation 370*C Including Loading Helium Pressure Limit (Cask Cavity) 800 t.100 mbar Pressure during Cask Ofying (Cask Cavity) $ 3 abar (holding for 10 min.)
Water Content (Cask Cavity) S 50 gram
.A-7
2.3.2 Basis .
The thermal analysis performed for the dry storage casks assumes the use of helium as a cover gas. In addition, the use of an inert gas (helium) is to ensure long-term maintenance of fuel clad integrity.
2.4 ORY STORAGE CASK SURFACE CONTAMINATION 2.4.1 Specification Removable contamination on the dry storage cask shall not exceed 1000 dis / min /
100 cm2 , from p, y sources and 20 dis / min /100 cm2 from a sources.
2.4.2 Basis Compliance with this limit ensures that the decontamination requirements of 49 CFR 173.443, as discussed in the Dry Cask ISFSI SAR Section 6.3.3, will:be l met over the lifetime of the cask in storage. .
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3.0 LIMITING CONDITIONS 3.1 LIMITING CONDITION - HANDLING HEIGHT i
- 3.1.1 Specification This specification applies to handling of a cask being used for spent fuel storage outside of the Fuel Building and Crane Enclosure Building,
- a. The CASTOR V/ 21 dry storage cask shall not be handled at a height of greater than 15 inches without an impact limiter.
! b. With the impact limiter the CASTOR V/21 dry storage cask shall not be handled at a height greater than 5 feet.
3.1.2 Basis
( The drop analyses performed for the CASTOR V/21 dry storage cask requires that an impact limiter be used "for~ postulated carx drop incidents on the Surry ISFSI storage pad for drops greater than 15 inches up to 62 inches without sustaining unacceptable damage to the storage cask and fuel basket. This limiting condi-l tion ensures.that the handling height limits will not be exceeded at the
! storage pad or in transit to and from the reactor.
3.2 LIMITING CONDITIONS - PRESSURE SWITCH 3.2.1 Specification The pressure switch used to monitor the leak tightness of the CASTOR V/21 dry storage cask shall have the performance characteristics shown in Table 3-1..
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Table 3-1 CASTOR V/21 ORY CASK PRESSURE SWITCH CHARACTERISTICS l
- 1. Physical specifications Switching pressure, working space 4 bar Switching pressure, reference space 3.5 bar Response sensitivity 0.5~ abar Switching precision 110 mbar Leak rate 10 8 mbar 1/s
- 2. Electrical specifications Maximum allowable switch contact voltage 10 V Maximum allowable switch contact current 10 mA
- 3. Environmental requirements Maximum allowable service temperature 150 *C Temperature coefficient 10 2 abar/*C o
O m
A-10 .
l 4.0 SURVEILLANCE REQUIREMENTS -
Requirements for surveillance of various radiation levels, cask internal pressure, contamination levels, cask seal leak rates, and fuel related param-
! eters are contained in this section. These requirements are summarized in ~
~
Table 4-1 from details contained in Section 4.1 through 4.6. Specified time intervals may be adjusted plus or minus 25 percent to accommodate normal test schedules.
4.1 CASK INTERLIO PRESSURE (CASTOR V/21) 4 4.1.1 Specification -
The cask inter 11d pressure shall be monitored by use of a pressure switch having the characteristics described in Table 3-1. The switching pressure shall be factory set at 4 bar foi the interlid space, and a functional test shall be performed during cask preparation.
~
4.1.2 Basis -
This specification requires the interlid space to be maintained to detect any 4
possible leakage of either cask seal.
4.2 OOSE RATES i
4.2.1 Specification The following dose rate measurements shall be made for the Surry Dry Cask ISFSI:
- a. Cask Surface Gamma and Neutron Dose Rates: After completion of cask loading, gamma and neutron measurements shall be taken on the outside surface or within 2 meters of the cask surface. These dose rates shall be less than the surface dose rates stated in 4.2.2 or their equivalent at a distance of up to 2 meters from the cask surface.
~
i A-11 -
Table 4-1 SURVEILLANCE' REQUIREMENTS
SUMMARY
Section Quantity or Item Period 4.1.1 Pressure Switch Parameters P&L 4.2.1 Dose Rates (Cask surface or up to 2 meters L from cask surface)
Oose Rates (Fence) Q 4.3.1 Alarm Board A 4.4.1 Fuel Parameters P 4.5.1 Cask Contamination L 4.6.1 Cask Seal Testing . L
~
P - Prior to cask loading L - During loading operations Q - Quarterly A - Annually l1 l
A-12 i
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. - .. ~
- b. Dry Cask ISFSI Boundary: TLDs shall be placed o'n the fence of the Dry Cask ISFSI site fence and shall be read on a quarterly basis.
There shall be 2 TLDs on each side of the ISFSI site (8 total).
4.2.2 Basis ,
i The dose rates must be within the dose rates described in Section 7.3.2.1 of the Surry Dry Cask ISFSI SAR as these dose rates were used in the safety environmental analysis performed for this installation (24 mres/hr neutron and 67 ares /hr gamma side surface; 74 mrem /hr neutron and 2 mrem /hr gamma lid surface). These surface measurements will ensure compliance with the dose rate limits at the Dry Cask' Installation fence which will be measured with the TLDs. .
4.3 ALARM BOARD 4.3.1 Specification The alarm board to which all of the pressure switches are' connected shall be functionally tested annually to ensure proper operation of the board.
4.3.2 Basis The alarm board must be checked periodically for general maintenance purposes to ensure that all components of the alarm board are working properly.
- 4.4 FUEL PARAMETERS 4.4.1 Specification i
Prior to cask loading, the fuel selected to be loaded shall have been reviewed to ensure that it is within the parameters of Specification 2.1.1.
- a. Initial enrichment 5 2.2 percent U-235 by weight (Upon staff review and acceptance of an appropriate basket design, the value may be raised to 3.5 percent U-235 as originally proposed).
4 m
A-13
- b. , burnup 5 35,000 MWD /MTU ,
- c. heat generation 5 1.0 kW/ fuel, assembly
- d. fuel cooling period 2 5 years This information shall be documented for each fuel assembly to be loaded in the fuel storage cask.
4.4.2 Basis .
See Basis 2.1.2. ,
4.5 CASK CONTAMINATION 4.5.1 Specification After cask loading and prior to. moving the cask tg the storage pad, the cask shall be swiped to ensure that removable surface contamination levels are less than 1000 dis / min /100 cm 2 , from p, y emitting sources and 20 dis / min /100 cm 2 from a emitting sources.
4.5.2 Basis This surveillance requirement will ensure compliance with the decontamination
- j. requirements of 49 CFR 173.443 during storage in the Surry Dry Cask ISFSI.
4.6 CASK SEAL TESTING i'
4.6.1 Sp'ecification During cask loading operations, each cask seal shall be tested using a helium leak detector to ensure that the seal leak tightness is less than 10.s abar 1/s.
1 A-14
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4.6.2 Basis ,
The safety analysis of leak tightness of the cask as dit. us:cd in the topical report is based on the seals being leak tight to 10 s mbar 1/s. This check is
, . done to ensure compliance with this design criteria. -
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A-15
5.0 DESIGN FEATURES 5.1 SITE .
5.1.1 Specification The Surry Dry Cask ISFSI is located on the Surry Power Station site as shown in Figure 2.2-3B of the SAR.
5.2 CASK DESIGN -
- 5. 2.1 Specification .
The casks used in the Surry Dry Cask ISFSI shall be GNSI CASTOR V/21 casks.
5.3 STORAGE PA0 5.3.1 Specification .
The ISFSI cask storage pads are reinforced concrete pads nominally 32 feet x 230 feet x 3 feet thick with a 20-foot ramp on each end for vehicle access.
Each pad is . designed to hold 28 casks arranged in two rows, nominally 16 feet apart center to center and in each row spaced nominally 16 feet apart center to center. The total facility will have three storage pads if required. Design criteria of the storage pads are contained in Section 3 of the Surry Dry Cask l
5.4 TOTAL STORAGE CAPACITY 5.4.1 Specification The total storage capacity of the Surry Dry Cask ISFSI is 811.44 TeU.
A-16
. , _ _ _ _ _ ,n - ____ m____, , _ , - - , _ . . - - _ - - . , _ _ . _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
e l- .
il Appendix 5 -
VIRGINIA ELECTRIC AND POWER COMPANY SURRY INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFEGUARDS LICENSE CONDITIONS-
~
, LICENSE NO. SAM 2501 I
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i i
1.0 INTRODUCTION
1 These license conditions supplement the requirements of Subpart H, 10 CFR Part 72 to prescribe specific measures for the physical protection of the Surry Independent Spent Fuel Storage Installation (ISFSI).
1.1 PHYSICAL PROTECTION REQUIREMENTS FOR SPENT NUCLEAR FUEL IN ORY STORAGE 1.1. A The licensee shall establish and maintain a physical protection program in accordance with the provisions of his physical security and safeguards contingency plans, published under the title,
" Security Program, Dry Cask Independent Spent Fuel Storage Instal- ,
lation, Surry Power Station," dated July 29, 1983, as revised August 1, 1984 and March 15, 1985; and as it may be further revised under the provisions of 10 CFR 72.33(e) and 72.84.
1.1.B The licensee's physical protection program shall be supported by a i security organization, with personnel trained and qualified in accordance with the' provisions of the plan published under tne title,
! " Nuclear Security Personnel Training and Qualification Program, Surry Power Station Units 1 and 2, and Dry Cask Independent Spent Fuel Storage Installation," dated August 18, 1979, as revised July 14, 1980, September 15, 1980, August 15, 1981, September 17, 1982, February 10, 1983, August 10, 1983, February 10, 1984 and February 15, 1985. Portions of this plan which impact on the Independent Spent
. Fuel Storage Installation may be changed or revised under the condi-tions as provided for revision of the physical protection plan, as contained in 10 CFR 72.33(e) and 72.84. A copy of this plan shall be filed along with the physical protection program identified in Condi-tion 1.1.A above.
B-1 L_ ____ _
O O
s APPENDIX C VIRGINIA ELECTRIC AND POWER COMPANY SURRY INDEPENDENT SPENT FUEL STORAGE TECHNICAL SPECIFICATIONS FOR ENVIRONMENTAL PROTECTION LICENSE SNM-2501 -
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1.0 INTRODUCTION
1
{ These technical specifications govern the protection of the environment during the receipt, possession, storage, and transfer of spent fuel at the Surry ISFSI.
1.1 RADI0 ACTIVE MATERIAL RELEASES 1.1.1 Specification (pursuant to S 72.33(d))
Not applicable.
1.1.2 Basis Specifications are required pursuant to S 72.33(d), stating limits on the release of radioactive materials for compliance with limits of 10 CFR Part 20 and the "as low as reasonably achievable objectives" for effluents. However, there are no normal or off-normal releases or effluents expected from the double-sealed storage casks of the ISFSI.
1.1 EFFLUENT
CONTROL AND WASTE TREATMENT 1.2.1 Specification (pursuant to S 72.33(d)(1))
Not applicable.
1.2.2. Basis Specifications are required pursuant to S 72.33(d)(1) for operating procedures L for control of effluents and for the maintenance and use of equipment in radio-
! active waste treatment systems to meet the requirements of 9 72.67. However, there are, by the design of the sealed storage casks at the ISFSI, no effluent releases, and all Surry site cask loading and unloading operations and waste l treatment therefrom will occur at the Surry Power Station under the specifica-l tions of its operating licenses.
l 1
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- 1. 3 ENVIRONMENTAL MONITORING PROGRAM 1.3.1 Specification The licensee shall include the Surry ISFSI in the environmental monitoring for the Surry Power Station.
1.3.2 Basis An environmental monitoring program is required pursuant to S 72.33(d)(2).
1.4 ANNUAL ENVIRONMENTAL REPORT 1.4.1 Specification An annual report, which is the Surry Power Station Radiological Environmental Annual Operating Report, will be submitted to the NRC Regiori III office with a copy to the Director, Office of Nuclear Material Safety and Safeguards, within 60 days after January 1 of each year, specifying the quantity of each l of the principal radionuclides released to the environment in liquid and in gaseous effluents during the previous 12 months 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.
1.4.2 Basis l
The report of Specification 1.4.1 is required pursuant to 10 CFR S 72.33(d)(3).
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ENCLOSURE 8
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SAFETY EVALUATION REPORT OF SURRY i
DRY CASK INDEPENDENT SPENT FUEL STORAGE INSTALLATION l
Docket No. 72-2 Virginia Electric and Power Company e
\.
l' May 1986 '
e Enclosure B
- . . _ _ _ _ _ . _ , _ - - _ _ _ _ _ _ _ . - _ _ - _ . . __. .___._._.o__
ABSTRACT A Safety Evaluation Report (SER) has been prepared by the Nuclear Regulatory Commission's Office of Nuclear Material Safety and Safeguards for the Virginia Electric and Power Company (VEPCO) application for a materials license pursuant to 10 CFR Part 72, Docket No. 72-2. The license application is for the receipt and storage of spent fuel for a period of twenty (20) years in an independent spent fuel storage installation (ISFSI). Granting the license would authorize VP to receive, possess, store and transfer spent nuclear fuel from Surry Units 1 and 2 at its Surry Power Station (SPS) Site near Surry, Virginia. This SER summarizes the results of the staff's radiological safety review and evaluation of the proposed licensing action.
m 9
iii
TABLE-OF CONTENTS Section Page ABSTRACT .............................................................. iii
1.0 INTRODUCTION
..................................................... 1-1 1
1.1 Background .................................................. 1-1 l 1.2 Application ............,.................................... 1-1 l 1.3 Approach .................................................... 1-3 ,
1.4 Summary of Findings and Conclusion .......................... 1-4 l 2.0 SAFETY EVALUATION ................................................ 2-1 4
2.1 Cask Acceptability at the Surry Site ........................ 2-1 2.1.1 Natural Phenomena .................................... 2-4 l 2.1.2 Man Induced Accidents ................................ 2-9 2.1.3 Summary of Findings about CASTOR V/21 Compatibility with the Surry Site .................................. 2-10 2.2 Transportation of the CASTOR V/21 Cask ...................... 2-11 2.3 Sto rage o f the CASTOR V/21 Cas k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-15 2.4 Findings and Conclusions .................................... 2-17 3.0 CONDUCT OF OPERATIONS ......................:..................... 3-1 3.1 Procedures .................................................. 3-1 3.2 Records ..................................................... 3-3 3.3 Training and Certification .................................. 3-4 3.4 Physical Protection ......................................... 3-5 3.5 Emergency Planning .......................................... 3-5 4.0 QUALITY ASSURANCE ................................................ 4-1 5.0 OPERATING CONTROLS AND LIMITS .................................... 5-1 6.0 DECOMMISSIONING .................................................. 6-1
7.0 CONCLUSION
S ...................................................... 7-1 l
8.0 REFERENCES
....................................................... 8-1 APPENDIX A - Chronology of Principal Actions .......................... A-1 APPENDIX B - Abbreviations ............................................ B-1 I
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7ABLE OF CONTENTS (Continued)
LIST OF TABLES Table P. age 2.1 Approved Limits for the GNSI CASTOR V/21 Cask ............... 2-13 LIST OF FIGURES Figure P_ age 2.1 CASTOR V/21 Cask ............................................ 2-2
- 2. 2 Surry Site Plan ............................................. 2-3 2.3 Storage Cask Transportation. Route ........................... 2-12 e
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i
1.0 INTRODUCTION
1.1 Background
Virginia Electric and Power Company (VEPCO) operates two nuclear power reactors near Surry, Virginia. After more than 12 years of operation, the two reactors have nearly exhausted the capacity of the existing fuel storage pool.
With the scheduled refueling outage for Surry Unit 2 this fall, the Surry plant will be unable to discharge a full core of fuel from either reactor. Anticipa-ting the need for additional storage capacity, VEPCO selected the alternative of dry spent fuel storage casks at the Surry site and applied for license authority under 10 CFR Part 72. As another option to maintain full core dis-charge capability at Surry, VEPCO applied and has received 1.icense authority for receipt and storage of up to 500 Surry spent fuel assemblies at its North Anna Nuclear Station. However, VEPCO has stated that it prefers the onsite, dry-cask storage alternative rather than transshipment to provide for 1-nterim spent fuel storage pending development of a national repository or storage l facility.
- 1. 2 Application VEPC0 has submitted an application to NRC for a license to receive, store and transfer spent nuclear fuel from VEPCO's Surry Units 1 and 2 in an onsite, dry-cask, independent spent fuel storage installation (ISFSI) at its Surry site, to be constructed and operated as set forth therein, (Reference 1). This faci-lity would consist mainly of a 16-acre fenced area for containing storage casks positioned on concrete pads and would hold about 811 metric tons of spent fuel in 84 CASTOR V/21 storage casks.
NRC reviewed this license application, evaluating the safety of the pro-posedoperatiogandthequalificationsoftheapplicanttoprotectpublichealth and safety, and to meet the requirements of the regulations for independent spent fuel storage installations (10 CFR Part 72). In its review of this appli-
~
cation, NRC staff did not evaluate the General Nuclear Systems, Inc. (GNSI) 1-1
I s
CASTOR V/21 cask itself because the staff had previously evaluated this cask i
and approved it for use under specific conditions (References 2 and 3). This review also did not reevaluate existing licen'ed s reactor or pool storage i facilities or activities. The review focused on the safety of onsite dry stor-i age operations, the transportation of storage casks from the reactor fuel build-ing to and from the storage area, and the applicant's management system for carrying out these operations in a safe manner.
The staff in its topical review for the CASTOR V/21 cask, referenced above, i has raised questions regarding the use of borated stainless steel for the fuel
- basket design considered for the CASTOR V/21 cask. These questions were gener-ated by the detection of weld area cracks observed in a cask basket in October 1985 during an unlicensed Department of Energy spent fuel handling and storage demonstration at the Idaho National Engineering Laboratory. This occur-red just after issuance of the staff SER for the cask design (Reference 2) and involved the storage of some spent fuel (8 assemblies out of 21 had decayed for I
about 2 years) with considerably less than 5 years decay time specified for the cask. .
By letter, dated October 14, 1985, the staff requested from GNSI an analysis and evaluation of the cause(s) of the basket cracks and an assessment of the need for design changes.
4 Since October 1985, the staff has received further information from GNSI I
and met with GNSI staff regarding the reported basket cracks. Questions regarding the basket design and materials remain to be fully resolved. While i continuing its tests and analyses of the original basket design, GNSI submitted l by letter dated February 28, 1986, an alternative design for a cask fuel basket which does not employ borated stainless steel as a basket material.
The staff has reviewed this alternative design of a stainless steel basket
- for use in the cask. fhis additional safety evaluation has been completed and-l is included with a letter of approval, which the staff has issued (Reference 3).
The only additional restriction set upon the spent fuel to be stored for the stainless steel' basket design is that the initial enrichment of the fuel to be stored not exceed 2.2 weight percent U-335. All other fuel parameters remain .
those for the 3.5 weight percent U-235 case originally considered, i.e., burnup,
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l etc. Thus,'no other analyses for the cask itself have been performed again,-
that is, no other derating was necessary. l Assuming the uncertainties raised by the Idaho occurrence are satisfac-torily resolved with respect to the borated stainless steel basket design, the staff, after reviewing any experimental data and analyses submitted'and design modifications proposed by the cask vendor, will also issue a letter of approval and safety evaluation to allow storage of higher initial enrichment fuel, up to 3.5 weight percent U-235 in the cask.
1.3 Approach NRC used different approaches for evaluating the safety of the storage
, operation and the applicant's management system. Evaluating the safety of the storage operation involved an analytical approach consisting of five steps.
The staff:
- (1) Reviewed the site based on the description in the Safety Analysis Report (Reference 4) and the site visits in order to determine all the natural and man-induced phenomena that could occur at the site and which might impact a storage cask. The staff determined whether the applicant identified all these phenomena in their storage cask specification.
(2) Determined the design basis severity of these site phenomena. If previous NRC determinations'of severity were available, they were used; if not, independent determinations were made on the basis of existing data.
(3) Compared the identification and severity of phenomena at the site with those for which the cask has been approved. On the basis of this comparison, the staff determined the acceptability of the Castor V/21 cask for the Surry site.
(4) Reviewed the operation of transporting the loaded Castor V/21 cask from the fuel building to the storage area and back again. The .
1-3
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h l-staff assessed the health and safety implications of this operation and identified any appropriate license restrictions.
(5)i Reviewed the operation of storing and monitoring casks, assessed the health and safety implications of this operation, and identified any appropriate license conditions.
In-evaluating the applicant's qualifications, the staff reviewed the appli-cant's approach.to conducting its operations. The review examined VEPCO's procedures, recordkeeping, training, and management systems relative to the proposed dry storage activities.. The staff also reviewed and_ evaluated the applicant's quality assurance program and decommissioning plan.
A verification of the applicant's and application's compliance with the licensing requirements for independent spent fuel storage installations is pre-sented in Chapter 7.
1.4 Summary of Findinas and Conclusions Based on the review and evaluation of both the applicant and the proposed ISFSI operation, the staff concludes:
(1) The Castor V/21 cask will withstand the natural and man-induced phenomena that can occur at the Surry site.
(2) The spent nuclear fuel which VEPC0 proposes to store at the ISFSI complies with the use restrictions for the Castor V/21 cask.
(3) The ISFSI can be operated safely if the applicant complies with the series of restrictions identified in this document.
(4) VEPCO, because of its experience in operating four nuclear reactors and its experience with cask handling and because of its preparation for this ISFSI operation, is qualified to conduct these ISFSI operations.
1-4
a .
(5) The radiological doses associated with normal operation and worst-case potential accidents are not significant.
(6) The, applicant has performed an' appropriate evaluation that concludes that the activities associated with dry cask spent fuel handling and storage do not present an unreviewed safety question with respect to reactor operations and that no changes to the technical specifications of the reactor operating licenses are necessary. Crane capacity for i handling casks at the. storage pool, as approved under the reactor operating licenses, is adequate for the ' loaded CASTOR V/21 storage cask.
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1-5
2.0 SAFETY EVALUATION This safety evaluation reviews the site characteristics, equipment, and
- operations associated with the proposed ISFSI. The evaluation addresses the i
acceptability of the GNSI-CASTOR V/21 cask at the Surry-site, the safety of onsite cask-transportation, and the safety of onsite storage. The review and-evaluation also identifies recommended technical specifications for the license.
The GNSI CASTOR V/21 cask is a large, monolithic, nodular cast iron cask with two bolted lids. The cask, shown in Figure 2.1, is cylindrical with a i
diameter of about 7 feet 9 inches and a height of 16 feet. Side wall thickness is about 17 inches and bottom thickness is about 14 inches. The primary lid is about 11 inches thick and the secondary lid is about 3-1/2 inches thick. The i
cask weights about 102 tons when empty and about 117 tons when full. The NRC
, staff has reviewed and evaluated this cask and approved it for use as a dry spent fuel storage cask subject to certain restrictions concerning fuel char-acteristics, service environment and handling (References 2 and 3).
2.1 Cask Acceptability at the Surry Site The Surry site is located in southeast Virginia in Surry County on a j peninsula in the James River (Reference 5). The towns of Surry, Newport News, and Williamsburg are within 10 miles of the site. Figure 2.2 shows the general
! location of the site and the location of the proposed storage area to the east of the reactor area.
The site is considered to have a moderate, humid climate. Occasional 4
thunderstorms, tornados, hurricanes and certain man-induced phenomena occur j and have been considered in evaluating the acceptability of the cask at the site. This section reviews these various phenomena and their effects on the cask.
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O 2.1.1 Natural Phenomena There is a range of site-specific natural phenomena which the ISFSI.must accommodate. These phenomena are reviewed and evaluated in this section.
Natural phenomena of a meteorological, geological and hydrological nature are addressed.
Ambient Temperature Ambient air temperature is an important factor in cask cooling and fuel cladding protection. A review of the 85 year temperature history for the area shows the temperature extremes for the region:
Area Record High( F) Record Low ( F)
Richmond 107 -17
. Norfolk 105 2 Over this time period, temperatures approaching these records have occurred about three times for the low at each site, three times for the high at Norfolk, and six times for the high at Richmond.
VEPCO has established a cask design specification that the cask operate in the temperature range of -20'F to 115'F. The staff reviewed this range and considers it to be acceptable in view of the site historical data.
The GNSI CASTOR V/21 has been reviewed and approved by the NRC staff for operating in the range of -40'F to 129'F (Reference 2). Therefore, the cask is acceptable for this site.
Insolation Solar radiation is a phenomenon which must be considered in heat transfer analyses for the casks. VEPCO has proposed a maximum solar insolation value of 800 g-cal /cm2 for a 12-hour day. The NRC staff has recommended (Regulatory Guide 7.8) that shipping cask heat transfer analyses be conducted on the _
2-4
assumption of a maximum 12-hour insolation of 800 g-cal /cm 2 regardless of the site. Therefore, the VEPC0 maximum value is considered acceptable.
Since GNSI V/21 cask used the same solar insolation value (800 g-cal /cm2. day) in the design which was reviewed and approved by the NRC staff (Reference 2),
the staff concludes here that the CASTOR V/21 cask can accommodate Surry insolation.
Lightning Lightning is a phenomenon which occurs in the area of the Surry site.
Local data suggest that 40 to 50 lightning storms occur per year.
VEPCO has not identified any maximum lightning surge for the site, although information provided by VEPCO suggests that surges as high as 275 kA could be expected.
GNSI did not establish any design basis lightning strik'e in their CASTOR V/21 topical report, but they did claim that no damage would occur because the cask would act as a Faraday cage and thereby protect the fuel from any lightning. The NRC staff, in its review of the CASTOR V/21 topical report, found that the cask would indeed act as a Faraday cage.
The NRC staff concluded that the storage cask will withstand lightning at the Surry site.
High Winds Wind speeds in the general region have been recorded since the late 1800s.
The record winds for the area's major monitoring stations at Richmond and Norfolk are:
Maximum Observed Period of ,
Area Wind Speed (mph) Observation l s Norfolk 80 1873-1984 .
Richmond 68 1898-1984 2-5 0
4 .
4 .
! VEPCO has estimated the maximum straightline wind at the site to be 105 mph, with an instantaneous gust as high as 137 mph. The NRC (AEC) staff previously reviewed and approved this extreme wind selection (Reference 6).
d i In addition to straightline winds, winds associated with tornados which have both translatiocal and rotational velocity components were rev'fewed.
l VEPCC has selected a maximum wind speed 300 mph rotational and 60 mph transla-j tional. These values are the same as the NRC (AEC) approved values for Class I structures at the reactor (Reference 6).
- NRC staff reviewed the performance of the CASTOR V/21 cask when subjected 4
to a tornado with 290 nph rotational and 70 mph translational wind speeds (Reference 2) and concluded that no failure could occur. Because the combined
, wind velocity is the same in both cases (360 mph), the staff concludes that the cask can accommodate the worst-case Surry site wind forces. Hurricane wind
{ velocities seldom reach half the velocity of the postulated tornado winds and are therefore enveloped by the worst-case site tornado wind forces.
Tornado Pressure Drop 1
- NRC (AEC) staff previously reviewed and approved a site maximum tornado
) pressure drop of 3 psi in three seconds for Surry Units 3 and 4 (Reference 6).
j This drop is more severe than has been recorded anywhere in the United States.
Since GNSI topical report did not identify a design basis tornado pressure
! drop, the staff review of this report could not address this phenomena. The staff in this review noted that the outer CASTOR V/21 lid contains a pressure of 102 psi, and that the previous staff review of the cask found that both the lid and lid bolt stress intensities are well below maximum allowable values.
On this basis, the staff finds that the 3 psi pressure drop associated with a
! tornado will not stress either lid or lid bolts beyond the allowable limits.
j The staff concludes that the CASTOR V/21 can withstand Surry design basis l tornado pressure drop.
i l
e 2-6 l
. . - - , - . - . - . - , _ - ~ _ _ _ . - _ _ _ . . . , _ _ . - _ _ - - . . . - _ _ __ _ _.. _ - __. _ .. - ____ _ _ ..
Tornado Missiles NRC regulations do not require that an ISFSI be protected from tornado missiles (10 CFR Section 72.72(b)(2)). However, NRC staff believes from the con-text of Section 72.72(b)(2) that the rule was in this instance primarily con-corned with massive structures, such as water pool basins, and not with struc-tures such as dry storage casks. Although VEPCO has not specified any tornado generated missile criteria as part of their cask acceptance criteria, the staff nonetheless considered tornado missiles in this review.
Current NRC staff regulatory position for maximum design basis tornado generated missiles for the Surry site can be found by using NUREG 0800, Sec-tion 3.5.1.4 (Missiles Generated by Natural Phenomena) in conjunction with Regulatory Guide 1.76 (Design Basis Tornado for Nuclear Power Plants). Based on these sources, the design basis tornado generated missiles for the Surry site are: (1) a 1800 kg automobile, (2) a 125 kg, 8 inch armor piercing artillery shell, and (3) a one inch solid steel sphere. The horizontal velocity for all these missiles is 126 mph.
The GNSI CASTOR V/21 cask was designed to withstand these missiles. The NRC staff review and evaluation of the cask concluded that the cask could with-stand the missiles (Reference 2). The automobile would not tip the cask and the artillery shell would not penetrate the lid.
The NRC staff concludes that the CASTOR V/21 can withstand maximum credible Surry tornado generated missiles.
Earthquake The Surry site is an area considered to have a limited potential for earth-quakes. No earthquake within the last 200 years has been severe enough to cause structural damage in the site area.
VEPCO has reviewed the local earthquake history through 1984 and estimated that a .07g earthquake occurs about every 500 years. It has established this frequency as the design basis for the site storage pad which is the working or ,
2-7
support surface for the casks during storage. VEPC0 stated that the cask, when stored on the pad, would not tip during a .07g earthquake.
VEPCO also analyzed the stability of the underlying soils during a .07g earthquake. The analysis utilized information obtained from soil samples of the proposed ISFSI storage area. The analysis showed that liquifaction'would not occur during or following a .07g earthquake.
The NRC staff reviewed the data and methods used to develop the estimate for the 500 year earthquake and concluded that the selection of the .07g earth-quake for design basis was acceptable. The staff evaluated the GNSI CASTOR V/21 cask and determined that it would not tip with an earthquake of .25g in the horizontal direction. The staff also determined that the cask would maintain its containment inte;rity if a tip-over accident should occur (Reference 2).
Therefore, the staff concluded that the CASTOR V/21 cask will withstand the site design basis earthquake.
The NRC staff also reviewed and evaluated the liquifaction analysis per-formed by VEPCO. The staff was satisfied with the way field data was utilized in the analysis as well as the analytical methods. The staff therefore agrees that soil liquifaction will not occur at the Surry site following a .07g earthquake.
Floods l
l The Surry site, located along the James River, experiences some degree of
- flooding. The river has a normal elevation of zero feet above mean sea level (msl). The largest recorded flood in the area was associated with Hurricane l Agnes (June, 1972) and caused a flood level of 2 ft msi; i.e., a 2 foot increase in water level at the site.
l VEPC0 has calculated a maximum possible flood level of 28.2 ft ms1 in their Updated Final Safety Analysis Report for Surry Power Station (Reference 5).
This is the same value that had been reviewed and accepted by the AEC regulatory staff (NRC predecessor) in their evaluation of site flooding for the Surry l
l 2-8
- - - , - - _~,--,- ,..,_ _ _.-,- -. - --
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Units 3 and 4, which were never built (Reference 6). The proposed location for the cask storage area is at an elevation of 35 ft as1. Therefore, the staff concludes that flooding is not a credible phe'nomenon which the cask must withstand.
4
- 2.1.2 Man-Induced Accidents l
{ There are several human activities at or near the site which could pose hazards for the ISFSI. These potential hazards include turbine missiles; explosion or fires from the local storage, processing, and transportation of '
combustible materials; and from the operation of aircraft in the area, i
Turbine Missiles Material failure in an onsite turbine was considered. The main steam turbines associated with the reactors and the 16 anc 20 megawatt gas peaking turbines at the electrical substation could generate turbine missiles. NRC staff, following the position in NUREG-0800 and Regulatory Guide 1.115, has I
concIuded that missiles from these turbines do not pose a threat to the cask j storage area because of the location and orientation of the turbines.
Explosion VEPCO identified several sources of possible explosions including industrial plants, overland transportation routes, water transportation routes, pipeline
- routes, and an onsite fuel tank. VEPCO also postulated accidents and their pro-l bable consequences, indicating that in all cases the ISFSI would be subject to overpressures of less than 1 psi. The most extreme explosion considered was the rupture of the Commonwealth Natural Gas Corporation pipeline located 400 yds. south of the ISFSI.
The NRC staff previously reviewed and approved this design basis peak l explosive overpressure when it was presented in the Updated Final Safety Analysis l Report for the Surry Power Station. The staff continues to find this design basis overpressure acceptable for the Surry site ISFSI as well.
I 2-9 '
l
The NRC staff, in its previous review of the GNSI CASTOR V/21, concluded that it could withstand the effects of a gas explosion and that it could withstand the maximum site explosion overpressure. The staff also concludes that the cask would not tip over if subjected to this overpressure.
Fires .
VEPCO identified the design basis fire as one involving a 320,000 gallon tank of No. 2 fuel oil which is stored 3770 feet west north-west of the ISFSI.
VEPCO stated that the only effect would be an increase of the air temperature' at the ISFSI by 8*F.
i The NRC staff reviewed information in the ISFSI Safety Analysis Report and i made independent calculations to verify estimated consequences. These calcula-4 tions determined that the fuel oil fire was an acceptable design basis fire, that the only impact would be minimal thermal radiation, and that the local air temperature would increase by less than 10*F.
The NRC staff evaluated a much more severe fire accident for the CASTOR V/21 cask and concluded that the cask could withstand the fire (Reference 2). From this analysis, the staff concludes that the cask can withstand the maximum Surry fire.
! Aircraft
- A previous NRC (AEC) staff review performed for Surry Power Station Units
, Nos. 3 and 4 concluded that local flying activities did not pose a hazard to
! the safe operation of the nuclear facility.
j On the basis of this previous review, the staff concluded that the flying i
i activities do not need to be addressed as a hazard for the ISFSI.
2.1.3 Summary of Findings about CASTOR V/21 Compatibility with the Surry Site l The NRC staff has reviewed the Surry site and identified those natural and
! man-made phenomena which the CASTOR V/21 cask must accommodate. The staff ,
i 2-10 l
l l
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established required levels of performance relative to these phenomena, compared them to the performance capability of the CASTOR V/21 cask and found that the CASTOR V/21 cask complied with all the identified performance factors.
2.2 Transportation of the CASTOR V/21 Cask The operation of moving the storage cask from the fuel handling building to the storage area and possibly back again (see Figure 2.3) must be conducted
- in a manner which does not jeopardize the integrity and performance of the cask or the reactor equipment. The NRC staff reviewed this proposed ISFSI operation in order to determine if there were any safety issues and to identify any safety measures which should be implemented.
The NRC staff previously reviewed the storage cask and determined that it provides adequate protection for the operating personnel, public, and fuel
- material if certain fuel characteristics are met, if the cask is prepared in accordance with certain limiting conditions, and if the cask is hand 19d accord-ing to certain restrictions. These conditions on fuel, cask, and handling are identified in Table 2.1 ISFSI operations will be conducted in accordance with the limits. The remaining potential safety problems involve transportation of heavy loads from
- the fuel handling building to the storage site. Any such problems will be -
smaller than the ones previously managed for the Surry steam generator replace-ment effort (Reference 7).
First, in moving from the fuel building to the ISFSI storage area, the cask will be transported up a slight grade in the vicinity of the containment structure for Unit 2 (see Figure 2.3). If the haul vehicle were to experience
- both a transmission failure and a brake failure or the trailer coupling were broken while moving upgrade, the vehicle could roll back toward the operating unit. A lag or lead vehicle, such as a bulldozer, behind or in front of the transporter will therefore be used when the casks are hauled up or down this grade. This will preclude the potential for a collision by a transporter against safety-related structures of the reactor in the event of a transport vehicle malfunction. .
2-11
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EXISTING LOW-LEVEL RADWASTE STORAGE FACILITY FIGURE 2.3. Storage Cask Transportation Route I,
i 2-12
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i
- l' Table 2.1 APPROVED LIMITS FOR THE GNSI CASTOR V/21 CASK F
Fuel Limits for Use in Cask (1) Pressurized Water Reactor (PWR) fuel; 14 x 14, 15 x 15, or 17 x 17 assemblies with Zircaloy cladding.
(2) Twenty-one or less fuel assemblies per cask.
(3) Initial fuel enrichment not greater than 2.2 percent Unas by weight (upon staff review and acceptance of an appropriate basket design the value may be raised to 3.5 percent Unas originally proposed).
' (4) Fuel Assembly Burnup shall not exceed 35,000 MWD /MTU at not more than 35 MW/MTU specific power.
(5)
- Spent fuel assembifes known or suspected either to have gross cladding defects or to have structural defects sufficiently severe to adversely affect fuel handling and transfer capability shall not l be loaded into the cask for storage. Partial assemblies, that is, i
assemblies from which fuel pins are missing must not be stored unless dummy fuel pins are used to displace an amount of water equal to that displaced by the original pins.
Filled Cask Requirements
- (6) Fill gas of helium.
l
'(7) Max, Surface Dose Rate 200 mrem /h (or equivalent)
! (8) Cask Tightness
! (Standard He-Leak Rate)
Primary Lid Seal 10.s mbar 1/s Secondary Ltd Sea) 10.s mbar 1/s i
1 (9) Helium Pressure Limit 7 bar
} (Inter 11d Cavity)
(10) Helium Pressure Limit (Cask Cavity) 800 1 100 mbar (11) Partial Pressure of Air (Cask Cavity) 3 mbar (holding for 10 min)
(12) Maximum Water Content (Cask Cavity) 50 gram (13) Removable Surface Contamination 1000 dpm/100 cm2 l Limits j beta gamma 20 dps/100 cm2 alpha -
i 4
J 2-13 4
O
~
Table 2.1 (Continued)
APPROVED LIMITS FOR THE GNSI CASTOR V/21 CASK Filled Cask Requirements (Continued)
(14) Physical specifications Switching pressure, working space 4 bar Switching pressure, reference space 3.5 bar Response sensitivity 0.5 mbar Switching precision 110 mbar Leak rate 10 S mbar-1-s 1 l Cask Handling Restrictions (15) Maxium Lift Heights Without Impact Limiter 15 inches With Impact Limiter 5 feet P
i 1
e 2-14
. The second issue is the hauling of heavy loads over the emergency diesel generator fuel oil line. .It is possible that pressure from the weight of a loaded transporter could crush the line and thereby render the generator inoperable. However, the NRC staff does not expect the line to fail due to the passage of loaded casks because of a previous VEPCO analysis pe.rformed in connection with the movement of a steam generator, and because the actual i steam generator hauling resulted in no problems. However, operability of the line will be verified after each haul since the line is important to safety.
The Environmental Assessment (EA) for this proposed activity identified the occupational doses associated with cask transportation (Reference 8). In
- this evaluation, the staff concluded that the occupational radiological con-i sequences will be less than 4.1 man-rem per year. The EA also shows that the i
radiological consequences of a worst-case cask handling accident would be very small at the site boundary, about 1 mren whole body exposure.
i In summary, the NRC staff reviewed the cask transiport operation and has j co'ncluded that the proposed operation is acceptable if the fuel and cask meet the restrictions of Table 2.1, if the cask is handled according to the restric-i tions of Table 2.1, if a lag or lead vehicle is used during cask movements up or down the grade of the transport route near the reactor building, and if diesel line operability is checked after each haul.
2.3 Storace of the CASTOR V/21 Cask After the cask has been transported to the storage area, it will be stored along with other casks for periods of time ranging up to 20 years or more.
Long-term safety is provided by keeping the casks in their proper storage condi-tion as approved by the NRC staff in their review of the CASTOR V/21 cask (Reference 2). In reviewing VEPCO's proposed storage operation, the NRC staff focused on the storage pad, the procedures for monitoring and main-taining casks in the approved storage condition, and the program for monitoring and reporting on dry fuel storage operations.
2-15
The storage pad which supports the casks is not defined as being a component which is important to safety. It is a storage surface for the casks and is a massive reinforced concrete pad (3 ft' thick, 32 ft wide, and 230 ft i
long) designed to support up to 28 casks each weighing 117 tons and placed at a nominal 16 ft center-to-center intervals. The NRC staff evaluated the VEPCO slab design to determine if it would withstand the .07g design basis' earthquake l and the loads which would be associated with high winds, thermal stresses which ,
result from the heated casks, and some extreme cases of soil settlement. The
- staff concluded that the slab would withstand a .07g earthquake, that the ther-mal stresses from any expected temperature gradients in the concrete would be small, and that the slab would remain intact even in the unexpected case of extreme soil settlement (15 ft of cantilevered pad loaded with casks).
The staff previously reviewed and approved the CASTOR V/21 casks for use as dry fuel storage packages if the loaded fuel meets certain conditions and helium atmosphere is maintained within the cask. These restrictions were pre-
- sented in Table 2.1. The GNSI cask has a seal-monitoring system which consists of a set of pressure activated switches that would be activated if a seal fails.
This seal-monitoring system will be used and maintained by VEPC0 for . :
I its dry storage casks. The pressure switches will be connected to an alarm board at the ISFSI fence which will activate a trouble light readily visible during routine patrols if any of the pressure switches are opened or any of the wiring fails. There are two redundant alarm lights on separate circuits which l are lighted when the switch is opened. Back-up power is supplied by a diesel I
generator in the event of a normal electrical power failure. .If the trouble light goes on and it cannot be determined that the cause was other than the seals, the cask will be returned to the fuel building within 7 days. The staff has also determined that the alarm circuits will be checked at least annually and that any failed components will be repaired or replaced within 15 days.
VEPCO will monitor the radiation levels associated with dry cask storage of spent fuel. The primary devices for measuring the radiation levels will be 8 thermoluminescent dosimeters (TLDs) placed on the fence surrounding the cask storage area (2 TLDs in the middle on each side). The readings from these TLDs will be used to help determine the offsite doses associated with dry cask fuel storage.
2-16
1 The NRC staff estimated the onsite and offsite doses associated with the l ISFSI in the EA for this proposed , operation (Reference 8). In this evaluation, the staff estimated the annual occupation dose due to storage operations to be less than 13.2 man-res/ year. This value includes exposure of people at the ISFSI, at the low-level waste storage facility, and at the Surry Power Station.
, The. staff estimated the dose from the casks.to the nearest resident by to be less than 6 x 10.s aren/ year. It also estimated that the cumulative site-
'l
. originated dose to this same nearest resident to be about 3 mrem / year, which is !
below the limit of 25 mrem / year specified in 10 CFR 72.67.
l The staff also estimated the doses associated with a maximum credible accident to be about 1.35 mrem whole body at the. site boundary and about 7.7 x 10 2 area whole body for the nearest resident. These doses are below values for which protective actions are required. On this basis, the staff concludes
'that the site boundary at the ISFSI-controlled area as proposed by VEPCO, meets the requirements of the regulations (10 CFR Section 72.68). The NRC staff also concludes that use of the 10-mile reactor emergency planning zone envelopes the requirements of the regulations (10 CFR Section 72.69) for the ISFSI.
Reviewing the entire cask storage operation, the staff concludes that main-taining the casks in proper condition will protect public health and safety
) during the stcrage operation.
2.4 Findinas and Conclusions On the basis of its review of the Surry site, the CASTOR V/21 cask as approved by the staff, and the operations planned by VEPCO, the staff concludes:
(1) All of the natural and man-induced phenomena that one can expect to occur at the site have been identified and design basis events established.
(2) The GNSI CASTOR V/21 cask is compatible with the Surry site.
2-17
(3) ISFSI transfer and storage operations, if conducted under limiting conditions and specifications, will protect public health and safety.
(4) The 10 CFR Part 72 ISFSI siting requirements are met.
i 2-18
O 3.0 CONDUCT OF OPERATIONS VEPCO is a regulated utility that has assembled an organization and man-agement system for operating the Surry Power Station, Units 1 and 2. The NRC staff has previously approved this organization and management system for VEPC0's reactor operations. VEPCO proposes to use this same organization and management system for the ISFSI, with modifications as necessary to accommodata the needs of the ISFSI. The effect will be to integrate the ISFSI into site operations.
The NRC staff reviewed this management system, evaluating the adaptability of the system to the ISFSI requirements and determining what modifications had to be made to ensure safe operation of the ISFSI. Components of the management system which were reviewed are: procedures (administrative, health physics, maintenance, operating, test, and pre-operational test), records, training and certification, physical protection, and emergency planning. Each of these was reviewed, and the staff determined the kinds of changes needed to meet the requirements of the ISFSI regulations. The review and the necessary changes are discussed in this chapter.
3.1 Procedures Administrative Procedures The current administrative procedures present the operating philosophy for j the site, establish management policies, and provide rules and instructions to all site personnel. In order to operate the ISFSI safely, VEPCO pro-poses to modify these procedures so that they specifically include ISFSI opera-tions. The NRC staff has determined that this approach is acceptable and that the modifications should clearly state that while the ISFSI operations are licensed separately from reactor operations, they are to be considered an inte-gral part of plant operations and subject to the same requirements.
i l
3-1 I
I
Health Physics Procedures The current health physics procedures implement the radiation control program at the Surry Power Station. These procedures address VEPCO's assessment and management of a variety of health physics situations throughout the plant.
The NRC staff reviewed this program of procedures for reactor operations, found them adequate, and has approved them.
In order to accommodate the needs of the ISFSI with the existing procedure system, the NRC staff fi1nds that existing procedures need to be modified or new' procedures prepared to (1) identify the location and processing of storage area perimeter TLDs and (2) address the radiological support requirements for personnel transporting the storage casks or entering the cask storage area. The applicant has committed to such actions.
l Maintenance Procedures Current maintenance procedures address the requirements and methods for doing both preventive and cor'rective maintenance throughout the Surry Power Station. In order to meet the requirements of the ISFSI, the NRC staff finds that existing procedures will have to be modified or new procedures prepared to address the maintenance of pressure switches on the casks and the maintenance of the alarm panel, circuit, and lights. The applicant has committed to the implementation of such procedures.
f Operatina Procedures The Surry Power Station currently has detailed procedures for all opera-tions, as required for Part 50 licenses. Under the Part 50 license for the fuel handling building, detailed procedures will be prepared for loading, fill-ing, sealing, and decontaminating of the Castor V/21 cask. In order to accom-modate the requirements of the ISFSI, existing procedures will have to be modified or new procedures prepared in order to include the transportation of the cask from the fuel handling building to the cask storage area, the position-ing of the casks upon the pad, and the monitoring of casks within the storage 3-2
[
area. These revised or new procedures should conform to the limits contained in the ISFSI license technical specifications. The applicant has committed to the preparation of the. required procedures.
l Test Procedures l
l l VEPCO's Surry Power Station currently has test procedures for items or components which must be tested or inspected regularly. In order to accommodate i the needs of the ISFSI, test procedures must be revised or additional procedures prepared in order to address the test requirements for the alarm panel board.
The applicant has committed to this action.
?
Preoperational Test Procedures - i VEPCO performs a dry run for new operations by using pre-operational test routines to check out both procedures and equipment. For the ISFSI operations, the dry run must use both equipment and procedures involved in (1) moving the loaded cask from the fuel handling building to the storage pad, (2) positioning the cask on the storage pad and connecting monitoring equipment, and (3) return-ing the cask from the storage pad to the fuel handling building.
After reviewing the overall VEPC0 system of procedures for managing site operations, the NRC staff concludes that the individual procedures, when modified to incorporate the findings of the NRC"s'taff, will meet the requirements of ISFSI regulations and will adequately protect public health and safety and minimize the danger to life and property. The applicant has committed to the described pre-operational tests of procedures and equipment.
3.2 Records In order to meet the technical specifications of their operating license for the reactor, VEPCO maintains detailed records on the operation and mainten-ance of the station and personnel activities such as duty assignments, qualifi-cations, training, and radiation exposure. The NRC staff has reviewed the 3-3
. . _ . . . - . . -- . _ _ . . _ _ _ . _ - _ __ - . - = . .
.recordkeeping requirements for reactor operation as listed in Surry Power Station Technical Specification 6.5. The NRC staff has concluded that extending the current Surry Power Station recordkeeping system to include the ISFSI and maintaining a detailed record of fuel characteristics and location within the ISFSI, as proposed by the applicant, is acceptable and will meet the record-keeping requirements for the ISFSI (10 CFR 72 Part Subpart D).
3.3 Trainina and Certification VEPC0 currently has a training program which has been approved by NRC staff for reactor operations. Included in this certification program is the requirement that the operator be physically qualified for the activity he .
or she is being trained to perform. The NRC staff considers the ISFSI opera-tions to require training similar to that required for operators who currently ,
handle casks and fuel in the Surry Power Plant. The NRC staff concludes that
- personnel who are either conducting or have direct visual supervision of the cask handling operations should have the same training as required for operators within the plant who hand 1'e casks or fuel. In addition, VEPCQ should have ISFSI training modules which address:
- a. Cask Design;
- b. ISFSI Design, Layout, and Function;
- c. ISFSI Safety Analysis; I
- d. ISFSI Technical Specifications; and
- e. ISFSI Procedures.
The NRC staff finds that the existing training and certification program for cask and fuel handling operators, when modified to include the ISFSI training module, as committed to by VEPCO (in a September 9, 1985 letter to Richard E.
Cunningham, Director, Division of Fuel Cycle and Material Safety, NMSS) will be satisfactory for complying with the ISFSI training and certification require-ments (10 CFR Part 72 Subpart I).
3-4
3.4 Physical Protection VEPCO has an NRC-approved physical protection program for the Surry Power Station. They have 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 ISFSI (10 CFR Part 72 Subpart H) will be met with the incorporation of the amendment into the site physical protection plan.
3.5 Emergency Planning -
VEPCO has a site specific Surry Emergency Plan which was developed for reactor operations. This plan, which meets the requirements 10 CFR Part 50, Section 50.47 and Appendix E, describes how emergencies will be addressed at the site. VEPCO proposes to modify this plan to meet the potential emergencies at the ISFSI by (1) requiring notification of NRC in the event of cask seal leakage or cask drop and (2) requiring an alert notification if there is a loss of cask containment barriers. The NRC staff has reviewed this proposal and concluded that there should also be-an alert notification in the event of an accidental criticality associated with the ISFSI according to the require-t ments of 10 CFR Section 72.52. The applicant has committed to this modification of the emergency plan.
In summary, the NRC staff finds that a Surry Emergency Plan, when modified to include the changes proposed by VEPCO and the staff findings, will meet the regulatory requirements for ISFSI emergency plans (10 CFR Section 72.19).
3-5
I 4.0 QUALITY ASSURANCE The regulations for independent spent fuel storage installations (10 CFR Section 72.80) require that VEPCO establish a Quality Assurance (QA) program based on the program required of power plants (10 CFR Part 50, Appendix B).
This extensive program is to be applied to those ISFSI activities that are important to safety throughout the life of the facility. Finally, the licensee must maintain records of design, fabrication, erection, testing,' maintenance, and u,tilization of the safety-related components.
VEPCO is currently licensed under 10 CFR Part 50 to operate nuclear power facilities, and a QA program meeting the requirements of 10 CFR Part 50, Appendix B, is already in place. The governing docubent for this program is the VEPC0 Topical Report VEP-1-4A, " Topical Report on Quality Assurance Program
- Operating Phase," which the NRC has reviewed and approved. This program is -
implemented through the VEPC0 Nuclear Power Station Quality Assurance Manual (NPSQAM), which addresses each of the 18 criteria in 10 CFR Part 50, Appendix B.
As indicated in previous chapters, tha casks are the only components with a safety function. For this reasons, it is important that cask activit-ies be controlled by the QA program.
The staff finds that the requirements for QA (10 CFR Section 72.80) will be met when the existing VEPCO QA program is applied, as has been committed to i-by the applicant, to the procurement and handling of the GNSI-CASTOR V/21 cask.
The staff also finds that VEPCO is applying quality standards in a manner which is commensurate with safety requirements as required by the ISFSI regulations (10 CFR Section 72.72(a)).
4 5
4-1
l 5.0 OPERATING CONTROLS AND LIMITS 1
4 .
Each license issued under 10 CFR Part 72 shall include license conditions pursuant to Section 72.33. In addition to the conditions set forth in Sec- ,
tion 72.33(b), each application for a license under 10 CFR Part 72 shall l include proposed technical specifications pursuant to Section 72.16 and consis-
, l tent with Sections 72.33(c) and 72.33(d). The finally approved technical speci-fications will be made part of the operating license. l The technical specifications of a license define certain features, char-acteristics and conditions governing operation of an' installation. Technical specifications cannot be changed without approval of the NRC. Consistent with 10 CFR Sections 72.33(c) and 72.33(d), the technical specifications will cover safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, administrative controls, effluent control, and environmental monitoring.
The licensee originally proposed technical specifications in its applica-tion dated October 8, 1982. Subsequent meetings between NRC and VEPCO staff resulted in revised technical specifications which were submitted February 19, 1986. These proposed technical specifications were derived from limits asso-l ciated with the use of the CASTOR V/21 cask and from the results of NRC staff's f safety and safeguards reviews of the proposed ISFSI and its operation and from the results of the NRC staff's environmental review, as expressed in the staff's Environmental Assessment (Reference 8). The technical specifications, which meet the requirements of 10 CFR Section 72.33, have been found acceptable by the staff.
l e
, 5-1
6.0 DECOMMISSIONING One of the requirements for independent fuel storage installations (10 CFR
,Section 72.18) stipulates that the applicant prepare for decommissioning by submitting a descriptive plan outlining not only the processes but also the financial plan. The descriptive plan is to include enough specific information r
on proposed actions for, decontamination and disposal of residual radioactive materials in order to provide a reasonable assurance that the public health and safety will be protected. The plan should identify and discuss the design
~
features of the ISFSI that will facilitate decommissioning. An outline of the financial arrangements for decommissioning is also needed to ensure that the process will be carried out.
VEPCO identified steps to be taken for decommissioning their casks by referencing the GNSI Topical Report, and committed to decomm,issioning the ISFSI casks according to GNSI specifications. The NRC staff previously reviewed this topical report (Reference 2) and found that it met the requirement of considering decommissioning in the design of the cask. VEPCO also presented a plan for decommissioning the slab. Because no radioactive releases are expected during either normal operation or accidents, Virginia Power is not expecting pad decontamination to be necessary. However, radiation surveys w,ill be made to determine.if any decontamination is necessary.
VEPCO presently owns and operates four nuclear power generating units.
They anticipate that decommissioning costs of the Surry ISFSI will be only a small fraction of the cost of decommissioning the Surry Power Station and therefore will not be an issue.
The staff concludes that VEPCO has considered decommissioning in their design of the ISFSI, meeting the requirements of 10 CFR Section 72.76. The staff also concludes that VEPC0 has a satisfactory decommissioning plan, that financing decommissioning will be possible, and that the requirements for decommissioning (10 CFR Section 72.18) will be met. , ,
6-1
7.0 CONCLUSION
S Introduction The regulations for independent spent fuel storage installations (10 CFR Part 72) call for the Commission to issue a license after determining that the application meets (1) the standards and requirements of the Atomic Energy Act; (2) the regulations of the Commission; and (3) the fourteen points identified
, in 10 CFR Section 72.31(a).
The staff's conclusions on these fourteen points are based largely on the safety evaluation report (SER) for the Surry ISFSI, the topical safety evalua-
. tion report (TSER) for the General Nuclear System, Inc. (GNSI) CASTOR V/21 cask, as supplemented by an additional safety evaluation performed by the staff, and the environmental assessment (EA) for the Surry ISFSI.
This chapter presents in summary fashion the staff's conclusions on the fourteen points of 10 CFR Section 72.31(a). References to sections in backup documents (SER, TSER, and EA) which support the findings are included. This serves as a point-by point check of VEPCO and their application against licensing requirements stated in 10 CFR Section 72.31(a).
Findinos In accordance with 10 CFR Section 72.31(a)(1), the staff finds that the l applicant's proposed ISFSI design complies with the general design criteria contained in Subpart F of 10 CFR Part 72 because:
l (1) The CASTOR V/21 cask, which has been identified as the only component important to safety, has been designed and will be fabricated, tested, and handled under NRC approved quality assurance programs. (See SER Section 4.0 and TSER Section 14.0.)
i 7-1
l (2) The CASTOR V/21 cask has been designed to function under both normal and severe environmental conditions at the Surry site. Both natural phenomena and man-induced conditions have been considered. The ISFSI is not considered to pose any threat to the local aquifer. (See SER Section 2.1.) .
(3) The Surry reactors and the ISFSI do not share structures, systems and components ,important to safety. (See SER Section 2.0.)
(4) The cumulative radiological effect of normal operation of the ISFSI and the Surry reactors are less than 25 mrem /yr to the nearest indi-vidual. No ISFSI accidents were identified which could have an effect offsite requiring protective actions. (See SER Section 2.3 and EA Section 6.2.)
(5) The CASTOR V/21 cask is designed to be inspected and tested and is essentially maintenance free. (See TSER Section 9.0.)
(6) The fuel cladding is protected against degradation and gross rupture by (1) maintaining an inert helium atmosphere in the cask, and (2) maintaining the fuel clad at temperatures that provide reasonable assurance that its integrity will be maintained throughout the period of_ storage. (See TSER Section 2.6.)
(7) The CASTOR V/21 cask provides the necessary confinement of radio-active particulate material during normal and accident conditions.
(See TSER Sections 7.0 and 11.0 and EA Section 6.2.)
(8) The CASTOR V/21 cask has a pressure monitoring switch which would signal if a cask seal failed. (See SER Section 2.3 and TSER Sec-tion 2.7.)
(9) Provisions have been made to provide timely emergency power to the central security alarm station to permit continued safe storage.
(See SER Section 3.4.)
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. ~
(10) The CASTOR V/21 cask is designed to maintain the spent fuel subcriti-cal in all normal and accident conditions through the use of cask fuel baskets designed to maintain stored fuel in subcritical configur-ations. For spent fuel which had a initial enrichment of 1 2.2 weight percent U-235, a stainless steel basket design has been reviewed by NRC staff and found to meet the requirements of 10 CFR Section 72.73.
(See Reference 3.) The basket design proposed for higher initial enrichment fuel (1 3.5 weight percent U-235) remains under review and may not be used until NRC staff review is complete and a final design
, is accepted.
(11) The CASTOR V/21 cask provides acceptable shielding of the stored spent nuclear fuel. The applicant, by use of the existing SPS health physics program, will provide for cask decontamination, access control to the ISFSI, and radiation surveys that ensure radiation protection l
and exposure control at the ISFSI. (See SER Section 3.1 and TSER Section 10.0.)
.(12) The' CASTOR V/21 cask, by virtue of the low leak rate required to main-tain the helium atmosphere, allows essentially no effluents to escape from the ISFSI. Direct radiation of the ISFSI is measured by TLDs and radiation surveys. (See SER Section 2.3, SER Table 2.1 and TSER i Section 2.9.)
(13) Analyses show that radiation exposure limits specified in 10 CFR Section 72.67 and Section 72.68 are met for normal operations and accident conditions. (See EA Section 6.2 TSER Sections 5.4, 7.4 and 7.5, and SER Section 2.3.)
(14) Only minimal amounts of low level radioactive wastes might be generated at the ISFSI during maintenance operations. These will be managed by existing Surry low level radioactive waste management systems.
The CASTOR V/21 cask, which contains the spent fuel, is designed to ensure adequate safety under normal and accident conditions. (See SER Section 2.3 and TSER.)
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. (15) The ISFSI and the CASTOR V/21 cask are designed for decommissioning.
(See SER Section 5.0 and TSER Section 12.0.)
In accordance with 10 CFR Section 72.31(a)(2), the NRC staff finds that site characteristics and. external natural and man-induced events have been investigated and assessed, that acceptable design basis events and conditions have been determined, and that the CASTOR V/21 cask design envelopes parameters associated with these site characteristics and design basis events to provide adequate protection. The NRC staff therefore finds that the proposed ISFSI site complies with Subpart E of 10 CFR Part 72 because:
(1) Based on regional characteristics, natural phenomena have been appro-priately assessed and design basis phenomena identified and evaluated.
(See SER Section 2.1.1.) -
(2) Man-made facilities and activities in the region have been examined, potential man-induced events that affect ISFSI design have been identified, and design basis man-induced events have been appropriately evaluated. (See SER Section 2.1.2.)
(3) Construction, operation, and decommissioning of the ISFSI have been found to cause no significant impact to the region surrounding the ISFSI. (See EA Section 9.0.)
i (4) The CASTOR V/21 cask is designed for peak horizontal accelerations I
i of 0.25g (which exceeds the design earthquake for the Surry ISFSI site) commensurate with the requirements of 10 CFR Section 72.66(b).
(See TSER Sections 2.4 and 3.4 and SER Sections 2.1.1 and 2.3.)
(5) The annual dose (from direct radiation due to ISFSI operations) i equivalent to the nearest real individual (nearest resident) located beyond the controlled area boundary is only a small fraction of the 25 arem, whole body dose criterion. This fraction, added to the dose due to.Surry reactor operations is still within the criteria provided 7-4
-e- e- -.,e-,,_r-.-,--,.--,--..,,_w,,n,..-.,-,w.g,,wy, re..--.-.,m,,,ww,,.y,,,,....-w,,a-.,e.- - - . , - , - - - - - , . - - . . _m _. . __ _____m______
the reactors operate within the numerical guides specified in Appen-dix I to 10 CFR Part 50. (See SER.Settion 2.3, TSER Saction 5.4 and EA Section 6.2.)
(6) The controlled area boundary is greater than 100 meters from the ISFSI and coincides with the SPS site boundary. The dose to an individual located on or beyond this boundary due to ISFSI accidents would be much less than the 5 rem criteria. (See SER Section 2.3, TSER Sec-tions 7.5, 3.3, and 11.4 and EA Section 6.2.)
(7) The ISFSI emergency planning zone (EPZ), for consistency, is the same EPZ as that designated for the SPS. However, there is reasonable assurance that for ISFSI accidents no protective actions beyond the controlled area boundary would be required. (See SER Section 2.3 and EA Section 6.2.)
(8) Because only SPS spent fuel is to be stored in the ISFSI located on the SPS site, there is no environmental impact due to spent fuel being transported into the area. (See SER Appendix A Technical Specification 2.1.)
In accordance with 10 CFR Sections 72.31(a)(3) and 72.72(d), the NRC staff finds that operation of the ISFSI on the SPS site will not pose undue risk to ,
the safe operation of the SPS reactors because:
(1) The ISFSI location is physically separated from the reactor buildings by approximately 3190 feet. (See SER Section 2.0.)
(2) The ISFSI and the SPS reactors do not share any structures, systems or components that are important to safety. (See SER Section 2.0.)
(3) The ISFSI will be operated as an integral part of the overall SPS operations. ISFSI operations can be scheduled and conducted so they do not conflict with reactor operations. (See SER Section 3.0.)
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.___.._...__.-_-.___,_..--_.._-___,.,.._.,,_,m_
-(4) Additional assurance against damage to structures important to safety is provided by using a lag or lead vehicle behind or in front of the transporter when a cask is hauled up or down a grade, and by testing {
diesel fuel oil lines after heavy loads have been hauled over them.
(See SER Section 2.2.). -
-In accordance with 10 CFR Section 72.31(a)(4), the NRC staff finds the applicant, by virtue of possessing 10 CFR Part 50 operating licenses (Nos. DPR-32 and DPR-37), is. qualified by reasons of training and experience to conduct ISFSI operations covered by the regulations in 10 CFR Part 72. (See SER Section 1.4.) !
In accordance with 10 CFR Section 72.3(m), the NRC staff finds the Surry site independant spent fuel storage installation meets the definition of inde-pendence as described in 10 CFR Part 72 since it is not physically connected to the reactor facility and its sharing of utilities and services does not,
"(i) increase the probability or consequences of an accident or malfunction of components structures or. systems that are important to safety or (ii) reduce the margin of safety as defined in the basis for any technical specifications i
of either facility." The applicant has performed an appropriate evaluation that concludes that the activities associated with the ISFSI 'do not represent an unreviewed safety question for reactor operations and that no changes to the technical specifications of the reactor operating licenses are required.
In accordance with 10 CFR Section 72.33(d), the NRC staff finds that the i
applicant will be in compliance with the limits of Part 20 of this chapter and the "as low as is reasonably achievable objectives" for effluents (see Sec-tion 6.2.1 of Reference 8); that the applicant has proposed adequate operating j procedures for control of affluents and equipment maintained and used in radio-I active waste treatment systems to meet the requirements of Section 72.67 of this part (see Section 6.2 of Reference 8); that the applicant has proposed an adequate environmental monitoring system (see Section 5.5 of Reference 8) to
(
ensure compliance with the technical specifications for effluents; and that the l applicant will provide an annual report for an estimate of maximum potential I
radiation dose commitment to the public from effluent releases (see Sec-l tion 5.5.2 of Reference 8).
l l
l l 7-6 I
I . -
i l
In accordance with 10 CFR Section 72.31(a)(5), the NRC staff finds that the applicant has adequate operating procedures to protect health and to mini- I mize danger to life or property. (See SER Section 3.1.) I I
In accordance with 10 CFR Section 72.31(a)(6) and Regulatory Guide 3.50, the staff finds that VEPCO is a regulated utility, and as such is financially qualified to operate an ISFSI on the Surry site in accordance with 10 CFR Part 72.
In accordance with 10 CFR Section 72.31(a)(7), the NRC staff finds that the applicant has a quality assurance plan that complies with Subpart G of 10 CFR Part 72. (See SER Section 4.0.)
In accordance with 10 CFR Section 72.31(a)(8), the NRC staff finds that the applicant's physical protectico plan complies with Subpart H of 10 CFR Part 72.
(See SER Section 3.4.) .
In accordance with 10 CFR Section 72.31(a)(9), the NRC staff finds that the applicant has an existing personnel training program for employees at the SPS, that when amended to include appropriate training modules covering ISFSI operations and cask handling procedures, will comply with Subpart I to 10 CFR Part 72. (See SER Section 3.3.)
In accordance with 10 CFR Section 72.31(a)(10), the NRC staff finds that the i applicant's decommissioning plan and its financing provide reasonable assurance that at the end of its life the Surry ISFSI can be decontaminated and decom-missioned and that public health and safety will be adequately protected. (See SER Section 5.0.)
In accordance with 10 CFR Section 72.31(a)(11), the NRC staff finds that the applicant's emergency plan for the SPS, when amended to include ISFSI emergencies, l will comply with 10 CFR Section 72.19. (See SER Section 3.5.)
l In accordance with 10 CFR Section 72.31(a)(12), the NRC staff finds that the applicant has satisfied the applicable provisions of 10 CFR Part 170.
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_ ._ _ -_ , . . _ . _ . - . .. _ _ _ _ 1 . _ _ _ _ - - - _ _ _ _ _ _ _ __. _ __ - -.
I In accordance with 10 CFR Section 72.31(a)(13), the NRC staff finds that, based on its " Safety Evaluation Report Related to the Topical Safety Analysis Report for CASTOR V/21 Dry Spent Fuel Storage Cask Submitted by General Nuclear Systems, Inc.," as supplemented, its " Environmental Assessment Related to the Construction and Operation of the Surry Dry Cask Independent Spent Fuel Storage Installation," and this safety evaluation, there is reasonable assurance that the l activities authorized by a license to receive, store and transfer spent reactor fuel from Surry Units 1 and 2 at the Surry Dry Cask Independent Spent Fuel Stor- ,
age Installation can be conducted without endangering the health and safety of i the public and that these activities will be conducted in compliance with the conditions of the license, and the Commission's Regulations (10 CFR Part 72, 10 CFR Part 20, 10 CER Part 50, and 10 CFR Part 73).
In accordance with 10 CFR Section 72.31(a)(14), the NRC staff finds that issuance of a license to receive, store and transfer spent reactor fuel from Surry Onits 1 and 2 at the Surry Dry Cask Independent Spent Fuel Storage Installation will not be inimical,to the common defense and security because:
(1) The activities will be conducted within the jurisdiction of the United States. (See SER Section 2.1.) ,
(2) The directors and principal officers of the applicant are citizens of'the United States. (See SER Section 3.0.)
(3) The applicant is not owned, dominated or controlled by an alien foreign corporation or foreign goverr. ment.
(4) The licensee may only receive and store fuel discharged solely from the Surry Power Station nuclear reactors. (See SER Appendix B, TS Section 2.1.)
(5) The fuel was used for civilian purposes and there will be no diver-sions for military surposes. (See License Nos. OPR-32 and 37.)
7-8
8.0 REFERENCES
- 1. Letter from R. H. Leasburg, Vice President, Nuclear Operations, to R. E. Cunningham,.0ffice of Nuclear. Material Safety and Safeguards, NRC, dated October 8, 1982, with License Application for Surry Power Station, Dry, Cask Independent Spent Fuel Storage Installation. Available at the NRC Public Document Room, 1717 H Street NW., Washington, DC 20555 under Docket No. 72-2.
- 2. Letter from Leland C. Rouse, Chief, Advanced Fuel and Spent Fuel Licensing Branch, NMSS, NRC, to Victor J. Barnhart, Vice President General Nuclear Systems, Inc., dated September 30, 1985, with Safety Evaluation Report Related to the Topical Safety Analysis Report for the Castor V/21 Dry.
Spent Fuel Storace Cask Submitted by General Nuc1' ear Systems, Inc.. Avail-
, able at the NRC Public Document Room, 1717 H Street NW., Washington, DC 20555 under Docket No. 72-2.
- 3. Letter from Leland C. Rouse, Chief, Advanced Fuel and Spent Fuel Licensing Branch, NMSS, NRC, to Victor J. Barnhart, Vice President, General Nuclear Systems, Inc., dated April 30, 1986. Available at the NRC Public Document Room, 1717 H Street NW., Washington, DC 20555 under Docket No. 72-2.
- 4. Safety Analysis Report, Surry Power Station, Dry Cask Independent Spent Fuel Storace Installation, VEPCO, February 19, 1986. Available at the NRC Public Document Room, 1717 H Street NW., Washington, DC 20555 under Docket No. 72-2.
- 5. Updated Final Safety Analysis Report, Surry Power Station Units 1 and 2, Virginia Electric and Power Co., June 1985. Available at the NRC Public Document Room, 1717 H Street NW., Washington, DC 20555 under Docket Nos. 50-280 and 50-281.
i 8-1 4
- 6. Safety Evaluation Report of the Surry Power Station Units 3 and 4. U.S.
Atomic Energy Commission, May 23, 1974. .Available at the NRC Public Docu-ment Room, 1717 H Street NW., Washington, DC 20555 under Docket Nos. 50-280 and 50-281.
- 7. Safety Evaluation by the Office of Nuclear Reactor Regulation, December 15, 1
1978. Available at the NRC Public Document Room, 1717 H Street NW., .
Washington, DC 20555 under Docket Nos. 50-280 and 50-281.
o
- 8. Letter from Leland C. Rouse, Chief, Advanced Fuel and Spent Fuel Licensing Branch, NMSS, NRC, to W. L. Stewart, Vice President, Nuclear Operations, Virginia Electric and Power Company, dated April 12, 1985, with Environ- -
mental Assessment Related to the Construction and Operation of the Surry Dry Cask Independent Spent Fuel Storace Installation, U.S. Nuclear Regulatory Commission, April 1985. Available at the NRC Public Document Room, 1717 H Street NW., Washington, DC 20555 under, Docket No. 72-2.
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.________ i-._____ _ _ _ _ _ . _ _ _ . _ . _ _ . _ . . _ _ -_. . _ _ . . _ . _ . . _ _ _
APPENDIX A CHRONOLOGY OF PRINCIPAL ACTIONS October 8,1982 Letter to R. E. Cunningham from R. H. Leasburg; In accordance with Part 72, VEPCO applies for a license to construct and operate a Dry Cask Independent Spent Fuel Storage Installation to be located on the site of the Surry Power Station in Surry County, Virginia.
Enclosures of license application, Safety Analysis -
, Report, and Environmental, Report.
December 3, 1982 Letter to R. H. Leasburg from.L. C. Rouse; NRC acknow-ledges receipt of application and the NRC staff is
, evaluating the appitcation for completeness and accuracy..
NRC yequests VEPCO to serve the Chie'f-Executive of
~
Surry County.
December 9, 1982 Federal Register notice published on Notice of Consideration.
December 14, 1982 Letter to R. E..Cunningham from W. L. Stewart; VEPCO notifies NRC of reorganization of t,ht nuclear. opera-tions department.
- December 17, 1982 Federal Register notice published, correcting December 9 Federal Register notice.
January 3, 1983 Letter to R. E. Cunningham from W. L. Stewart; VEPC0 asks that NRC remove and discard the Physical Security
. Plan from license application.
January 25, 1983 Letter to R. E. Cunningham from W. L. Stewart; VEPCO provides-a supplement to the SURRY ISFSI SAR. _ .
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March 16, 1983 Letter to W. L. Stewart from L. C; Rouse; NRC requests additional information be included on supplemental material (administrative identification for filing ease)-
and encloses NRC comments on Combustion Engineering Topical Report.
April 8, 1983 Letter to W. L. Stewart from L. C. Rouse; NRC comments ,
of Fuel Facility Safeguards Licensing Branch suggests it may be desirable for VEPCO to complete 1~y resubmit the
. required safeguard plans.
July 29, 1983 ,
Letter to R. E. Cunningham from W. L. Stewart; VEPCO ,
~
resubmits Security Plan.
August 19, 1983 Letter to R. E. Cunningham from W. L. Stewart; VEPC0 provides NRC with Nuclear Security Personnel Training and Qualification Program for Surry Dry Cask ISFSI.
September 9,1983 Letter. to W. L. Stewart from L. C. Rouse; NRC transmits questions on application for the Surry ISFSI.
October 25, 1983 Letter to R. E. Cunningham from W. L. Stewart; VEPC0 provides supplement to Dry Cask Independent Spent Fuel
( Storage Installation License Application.
March 2, 1984 Letter to R. E. Cunningham from W. L. Stewart; VEPC0 responds to NRC's request for additional information of September 9, ic33.
March 13, 1984 Letter to R. E. Cunningham from W. L. Stewart; VEPC0 makes selection of cask for the ISFSI and requests NRC give the vendor top priority in completing review of topical report.
- A-2.
r i
_ March 30, 1984 Letter to W. L.' Stewart from W. Brown; NRC discusses physical, protection requirements for dry cask storage of spent fuel. .
l May 21, 1984 Letter to R. E. Cunningham from W. L. Stewart; VEPCO j discusses a meeting on May 23, 1984 addressing tempera-tures for dry cask storage.
June 20, 1984 Letter to R. E. Cunningham from W. L. Stewart; VEPC0
. provides additional information in response to NRC's request for additional information of September 9, 1983.
, June 25, 1984 Letter to R. E. Cunningham from W. L. Stewart; VEPCO
~
provides additional information in response to NRC's request for additional information of September 9,1983.
September 21, 1984 Letter to R. E. Cunningham.from W. L. Stewart; VEPCO provides additional information in r,tsponse to NRC's request for additional information of' September 9 - 1983.
October 1, 1984 Letter to W. L. Stewart from L. C. Rouse; NRC requests j
more detailed information regarding soil liquefaction analysis.
October 24, 1984 Letter to-R. E. Cunningham from W. L. Stewart; VEPCO provides additional information and response to NRC's request for additional information of September 9, 1983.
November 14, 1984 Letter to W. L. Stewart from L. C. Rouse; NRC requests more detailed information about radiological dose calculations.
November 30, 1984 Letter to R. E. Cunningham from W. L. Stewart; VEPC0 provides additional information in response to NRC's
, request for additional information of September 9, 1983.
1 o'
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_ r w December 4, 1984 Letter to R. E. Cunningham'from W. L. Stewart; VEPC0 -
provides , additional information in response to NRC's ,
request for additional information of' September 9, 1983.
December 10, 1984 Letter to R. E. Cunningham from W. L. Stewart; VEPCO provides additional information in response to NRC's ,
request for additional information of September 9, 1983.
Fcbruary 8, 1985 . Letter to R. E. Cunningham from W. L. Stewart; VEPCO provides additional information in response to NRC's request for additional information of November 14, 1984.
Msrch 7, 1985 Letter to W. L. Stewart from L. C. Rouse; NRC requests additional information.
March 15, 1985 Letter to J. G. Davis from W. L. Stewart; Virginia Power is identified as the new corporate name.
April 10, 1985 Letter to R. E. Cunningham from W. L. Stewart; VEPCO requests authorization to initiate construction of the ISFSI pad.
April 12, 1985 Letter to W. L. Stewart from L. C. Rouse; NRC trans-mits a copy of the Federal Register notice of Issuance of the " Environmental Assessment Related to the Con-struction and Operation of the Surry Dry Cask Inde-pendent Spent Fuel Storage Installation" and Finding of No Significant Impact.
April 18, 1985 Federal Register notice published on issuance of Environ-mental Assessment and Finding of No Significant Impact for the Surry ISFSI.
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April 23, 1985 Letter to W. L. Stewart from L. C. Rouse; NRC requests revised drawings and specifications for concrete ISFSI storage pad.-
May 9, 1985 Letter to R. E. Cunningham from W. L. Stewart; VEPC0 provides additional information on response to NRC's request for additional information of April 23, 1985.
May 24, 1985 Letter to the NRC Commissioners from W. J. Dircks; Transmittal of a Policy Issue Commission Paper (Infor-mation), SECY 85-190 CONSTRUCTION OF A CONCRETE PAD AS FOUNDATION FOR DRY SPENT FUEL STORAGE CASKS AT THE SURRY NUCLEAR POWER PLANT SITE.
June 10, 1985 Letter to W. L. Stewart from R. E. Cunningham; NRC responds to April 10, 1985 request for permission to initiate construction of initial concrete pad with its attendant security equipment and states that VEPCO can proceed with pad construction at their own risk.
June 10, 1985 Letter to R. E. Cunningham from W. L. Stewart; VEPCO amends its license application, SAR and environmental report.
June 21, 1985 Letter to R. E. Cunningham from W. L. Stewart; VEPCO provides additional information in response to NRC's request for additional information of March 7,1985.
July 30, 1985 Surry Power Station site visit; discussion of proposed license technical specifications (Ref. memorandum of 8/5/85 from J. P. Roberts to L. C. Rouse).
August 7, 1985 Letter to R. E. Cunningham from W. L. Stewart; drawing supplied by VEPCO to be docketed as additional infor-mation.
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September 9, 1985 Letter to R. E. Cunningham from W. L. Stewart; VEPC0 l provides additional information on training of personnel who would perform fuel handling and cask handling for the Dry Cask ISFSI.
October 15, 1985 Letter to W. L. Stewart from L. C. Rouse; NRC requests reassessment of use of CASTOR V/21 cask at Surry and also requests a report of findings along with any necessary license application changes.
December 27, 1985 Letter to R. E. Cunningham from W. L. Stewart; VEPCO transmits additional information on the CASTOR V/21 basket indications and indicates intention to proceed with the use of the CASTOR V/21 cask.
January 9,1986 Meeting to discuss proposed license technical specifi-cations (Ref. memorandum of 1/14/86 from J. P. Roberts to L. C. Rouse).
February 19, 1986 Letter to R. E. Cunningham from W. L. Stewart; VEPC0 amends the SAR to update Chapter 10, ISFSI Tech Specs.
February 24, 1986 Meeting at which VEPC0 proposes use of an all stainless steel basket in the CASTOR V/21 cask with questions concerning use of a borated stainless steel basket to continue to be addressed until their resolution. (Ref.
memorandum of 2/27/86 from J. P. Roberts to L. C. Rouse).
March 25, 1986 Letter to R. E. Cunningham from W. L. Stewart stating over 100 spent fuel assemblies of an initial enrichment of less than 2.2 weight percent U-235 are in storage at Surry Power Station. These are available for storage in a CASTOR V/21 cask with an all stainless steel basket.
VEPC0 concurs in this design change by its cask vendor.
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APPENDIX 8
~
ABBREVIATIONS l
AEC- U.S. Atomic Energy Commission Stu British thermal unit cfm cubic feet pe,r minute cfs cubic feet per second C1 Curie (3.7 x 1010 disintegrations per second)
CFR Code of Federal Regulations
- cm centimeter DF Decontamination Factor DOT U.S. Department of Transportation dpm disintegrations per minute
- C degree Celsius
- F degree Fahrenheit .
EA Environmental Assessment ft foot g accelaration of earth's gravity (about 980 cm/sec2 (32.2 ft/sect ))
l g-cal gram calorie gal gallons gpm gallons per minute h ' hour ha hectare hp horsepower l hr hour in inch ISFSI Independent Spent Fuel Storage Installation aff effective multiplication factor kg kilogram km kilometer kPa nilo pascal kVA kilovolt ampere i
kW kilowatt 8-1
- ~ _ .
I liter
.lb pound LAW low-activity waste a meter i
~
abar millibar (14.5 psi x10 3) min. minute mi milliliter MM modified Mercalli MPa nega Pascal MPC maximum permissible concentration sph miles per hour arem millirem ms1 mean sea level MTU metric ton uranium (TeU)
MWD megawatt-days NRC U.S. Nuclear Regulatory Commission PMF probable maximum flood PMP probable maximum precipitation .
psi pounds per square tnch -
, psig pounds per square inch (pressure gauge)
PWR pressurized water reactor QA quality assurance rem r,,adiation dose unit (roentgen equivalent, man) s second i SER safety evaluation report l SPS Surry Power Station l
Te tonne Teu tonne of uranium TLD thermoluminescent dosimeter l TSAR Topical Safety Analysis Report TSER Topical Safety Evaluation Report V volt VEPCO Virginia Electric and Power Company W Watt 8-2
..+ - - -w. ,
=- w >y.
"*V k k r.
l o 4
I 3
i e
h h
s i
t l
ENCLOSURE C s
? .
(
t e
l l
\
I U.S. NUCLEAR REGULATORY COMMISSION ,
OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS ENVIRONMENTAL ASSESSMENT RELATED TO THE CONSTRUCTION AND OPERATION OF THE SURRY ORY TATx- NDEFENDENT SPENT FUEL STORAGE INSTALLATION -
00CXET NO. 72-2 VIRGINIA ELECTRIC AND POWER COMPANY APRIL 1985 Enclosure C
ENVIRONMENTAL ASSESSMENT FOR SURRY ORY CASK ISFSI CONTENTS .
1.0 INTR 00VCTION ....................... I
1.1 DESCRIPTION
OF PROPOSED ACTION . . . . . . . . .'. . . I
1.2 BACKGROUND
INFORMATION . . . . . . . . . . . . . . . . 2 1.3 PREVIOUS ENVIRONMENTAL ASSESSMENTS AND SUPPORTING DOCUMENTS . . . . . . . . . . . . . . ... . . . . . . 5 1.4 STATUS OF ENVIRONMENTAL APPRO'fALS . . . . . . . . . . . 5 2.0 NEED FOR PROPOSED ACTION . . . . . . . . . . . . . . . . . . 7 3.0 ALTERNATIVES . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1 SHIP SPENT FUEL TO A PERMANENT FEDERAL REPOSITORY , . . 8 3.2 SHIP SPENT FUEL TO NORTH ANNA . . . . . . . . . . . . . 9 3.3 INCREASE THE STORAGE CAPACITY OF THE EXISTING SPENT FU EL POOL . . . . . . . . . . . . . . . . . . . . . . 10 3.4 CONSTRUCT A NEW INDEPENDENT SPENT FUEL STORAGE POOL AT THE SURRY SITE . . . . . . . . . . . . . . . . . . 10
" 3.5 SHIP SPENT FUE1: TO A REPROCESSING FACILITY ...... 11 3.6 SHIP SPENT FUEL TO A FEDERAL INTERIM STORAGE (FIS)
FACILITY ...................... 11 3.7 IMPROVE FUEL USAGE .................. 11 3.8 OPERATE SURRY POWER STATION AT REDUCED POWER ..... 12 3.9 SHIP SPENT FUEL TO OTHER UTILITY COMPANIES' REACTORS FOR STORAGE . . . . . . . . . . . . . . . . . . . . . 12 3.10 CONSTRUCT AN ISFSI AT A SITE AWAY FROM THE SURRY POWER STATION . . . . . . . . . . . . . . . . . . . . 13 3.11 NO ACTION . . . . . . . . . . . . . . . . . . . . . . . 13 3.12
SUMMARY
OF ALTERNATIVES . . . . . . . . . . . . . . . . 14 i
4.0 ENVIRONMENTAL INTERFACES . . . . . . . . . . . . . . . . . . 15 4.1 SITE LOCATION, LAND USE AND TERRESTRIAL RESOURCES . . . 15 4.2 WATER USE AND AQUATIC RESOURCES . . . . . . . . . . . . 18 4.3 SOCI0 ECONOMICS AND HISTORICAL, ARCHAEOLOGICAL AND CULTURAL RESOURCES ................. 20 4.4 DEMOGRAPHY ...................... 20 4.5 METEOROLOGY AND CLIMATOLOGY . . . . . . . . . . . . . . 21 4.6 GEOLOGY AND SEISMOLOGY ................ 26 11 i
CONTENTS (Continued)
5.0 DESCRIPTION
OF SURRY ORY CASK ISFSI . . . . . . . . . . . . . 28 5.1 ISFSI LOCATION . . . . . . . . . . . . . . . . . . . . .- 28 5.2 SITE PREPARATION . . . . . . . . . . . . . . . . . . . . 28 5.3 STORAGE SYSTEM . . . . . . . . . . . . . . . . . . . . . 28 5.4 ISFSI OPERATIONS . . . . . . . . . . . . . . . . . . . . 35 5.5 MONITORING PROGRAMS . . . ............... 36 6.0 ENVIRONMENTAL IMPACTS OF PROPOSED ACTION .......... 39 6.1 CONSTRUCTION IMPACTS . . . . . . . . . . . . . . . . . . 39 6.2 OPERATIONAL IMPACTS . .. ............... 41 6.2.1 Radiological Imoacts from Routine Ooerations .. 41 6.2.2 Non-Radiolacical Imoacts ............ 52 7.0 SAFEGUARDS FOR SPENT FUEL . . . . . . . . . . . . . . . . . . 54 7.1 onsite MOVEMENT . . . . . . . . . . . . . . . . . . . . 55 7.2 FIXED SITE SAFEGUARDS ... ........ 56 7.3
SUMMARY
...... 58 8.0 OECOMMISSIGNING . . . . . . . . . . . . . . . . . . . . . . . 59 9.0
SUMMARY
AND CONCLUSIONS . . . . . . . . . . . . . . . .,. . . 60 -
9.1
SUMMARY
OF ENVIRONMENTAL IMPACTS . . . . . . . . . . . . 60 9.2 BASIS FOR FINDING OF NO SIGNIFICANT IMPACT . . . . . . . 61
10.0 REFERENCES
......................... 63
- 11. 0 LIST OF P REPARERS . . . . . . . . . . . . . . . . . . . . . . 66 I
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l ENVIRONMENTAL ASSESSMENT
' ' RELATED TO THE PROPOSED CONSTRUCTION AND OPERATION 0F THE SURRY ORY CASK INDEPENDENT SPENT FUEL STORAGE INSTALLATION -
1.0 INTRODUCTION
1.1 DESCRIPTION
OF PROPOSED ACTION' By letter dated October 8,1982, Virginia Electric and Power Company (the Applicant or VEPCO) submitted an application 2 for a license to construct and operate a Ory Cask Independent Spent Fuel Storage Installation (ISFSI) to be
' located on the Surry Power Station site in Surry County, Virginia. The scope of this environmental assessment includes the construction and operation of an ISFSI on the Surry site, including impacts specifically derived from the cask to be used, the CASTOR V/21. The function of the Dry Cask ISFSI is to provide interim storage of spent fuel from Surry Units 1 and 2. Loading and initial preparations of the casks will take place within the Surry Power Station fuel handling building. The casks will be stored on concrete slabs constructed on site.
The Surry ISFSI is designed to operate througn the licensed life of Units 1 and 2, i.e., to the years 2007 and 2008, respectively. However, the duration
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of licenses issued under 10 CFR Part 72 is limited to 20 years; thus, the application is for 20 years from the date of issuance. The appitcant expects '
l to request renewal of the Dry Cask ISFSI license, if necessary, prior to its expiration.
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This Environmental Assessment addresses the potential environmental impacts associated with the proposed construction and operation of the Dry task ISFSI on the Surry Power Station Site.
1.2 BACKGROUND
INFORMATION I
The Surry Power Station Units 1 and 2 (Dockets 50-280 and 50-281) were licensed i to operate in May 1972 and January 1973, respectively. Commercial operation began in December 1972 for Unit 1 and in May 1973 for Unit 2.
Until about 1975, it was planned in gen'eral that spent fuel from nuclear powered reactors would be stored in the spent fuel pool at the reactor site where generated for an interim period. After the interim storage, it was anticipated .
that spent fuel would be transported to a reprocessing plant for recovery and recycling of the fuel. Reactor facilities, incluaing Surry, which were designed j
and constructed prior to 1975 provided less caoacity for spent fuel storage on site than required for life of plant.
3 Commercial reprocessing of spent fuel has not coveloped as had been originally anticipated. In 1975 the Nuclear Regulatory Commission directed the staff to prepare a generic environmental impact statement on spent fuel storage. The Commission directed the staff to analyze alternatives for the handling and storage of spent light water power reactor fuel with particular emphasis on i
developing long range policy. The Statement was to consider alternative methods of spent fuel storage as well as the possible restriction or termination of the
- generation of spent fuel through nuclear power clant shutcown.
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The " Final Generic Environmental Imoact Statement on Handling and Storage of Soent Light Water Power Reactor Fuel" (the FGEIS)2 was issued by the NRC in August 1979. In the FGEIS, consistent with long range policy, the storage of spent fuel is considered to be interim storage to be used until the issue of permanent disposal is resolved and implemented.
Options for interim storage considered in detail in the FGEIS include .
(1) onsite spent fuel pool capacity expansion, (2) spent fuel storage capacity expansion at reprocessing plants, (3) f rrdependent spent fuel storage facilities, (4) transshipment of spent fuel between reactors and (S) reactor shut-downs to terminate or reduce the total amount of spent fuel generated.
Of these options,111 onsite spent fuel pool capacity. expansions through raracking modifications have been reviewed and approved by the NRC since t
issuance of the FGEIS. As discussed in Section 3.0 of this Environmental Assessment, further increase in fuel pool capacity at Surry is not a viable alternative. The option of transshipment of a portion of the generated spent fuel from Surry to the applicant's North Anna Station is being considered in a separate licensing action.
l The applicant is participating in a demonstration program with the Department of Energy (DCE) which involves shipment of spent fuel (up to 4 casks of capacity) to DOE's Idaho National Engineering Laboratory (INEL). The first demonstration cask, which is the CASTOR V/21 model, is at INEL awaiting the -
expected delivery of Surry spent fuel in the latter half of 1985. The second comonstration cask (a Westinghouse Model MC-10) is expected to be caliverso to INEL in 1986, and the third and fourth casks have an uncertain delivery m- -- - - -,,m.-, . - - - -- , .-- y -, , -y ,--- ,--- - - - -. -,-- -- - . -
schedule at this time. Spent fuel storage uncer this program is not considered as an alternative to the proposed action since the program,is a !
demonstration only and involves limited storage of a small amount of spent fuel (about 60 metric tons). Participation in the demonstration does nothing to alter the applicant's need as detailed in Section 2.0.
The FGEIS concluded that the independent spent fuel storage installations represented the major means of away-from-reactor interim storage and that a standard design of an ISFSI to be situated at a reactor site had been submitted to and reviewed by the NRC. The FGEIS supports findings that the storage of LWR spent fuels in water pools, whether at the reactor or away from reactor sites, has an insignificant impact on the environment. While the environmental impacts of the dry storage option were not specifically addressed in the FGEIS, the use of alternative dry passive storage techniques for aged fuel appeared to be equally feasible and environmentally acceptable.8 In the case of both dry passive storage and wet storage, environmental impacts would need to be considered on a site-specific casts. This assessment addresses the site-specific environmental aspects related to the Dry Cask ISFSI at the Surry site.
A ccmparison of the impact-costs of various alternatives reflects the advantage of continued generation of nuclear power versus its replacement by coal-fired power generation. In the bounding case considered in the FGEIS, that of shutting down the reactor when the existing spent fuel storage capacity is filled, the cost of replacing nuclear stations before the end of their normal lifetime makes this alternative uneconomical.
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1.3 PREVIOUS ENVIRONMENTAL ASSESSMENTS AND SUPPORTING 00CUMENTS Several previous assessments have been conducted which are specific to the Surry site. A Final Environmental Statement (FES) was prepared for each of the Surry Units 1 and 2; the Unit 1 FES' was issued in May 1972 and the Unit 2 FES' in June 1972.
An application for a construction permit for two additional units to be located at the Surry site was filed in April 1973. An FES* related to this proposed construction of Surry Units 3 and 4 was issued in May 1974.
Subsequent to the issuance of a construct' ion permit in Feburary 1975, the applicant cancelled plans.to construct the two additional units.
t This environmental assessment is tiered on the previously issued FES's for the Surry reactor units and the FGEIS (NUREG-0575), noted in Section 1.2 above.
Additional environmental information, used in this assessment, is provided in the appiteant's Surry ISFSI Environmental Report (ER)' and Safety Analysis '
l Report (SAR)* and supple-mental responses (references 9 through 17) to the NRC staff's questions 1
',1',8' on the Surry ISFSI ER and SAR.
i 1.4 STATUS OF ENVIRONMENTAL APPROVALS The applicant will require a Conditional Use Permit from the Surry County Board '
i of Supervisors to proceed with construction of the Surry ISFSI. All existing recuirements of Federal (other than NRC recuirements, specific to this acclication),
4
6-state and local permits, licenses or other forms of approval issuec for Surry Units 1 and 2 will encompass operation of the Surry ISFSI since, the installation is located onsite.
We have identified no other environmental approvals required for the proposed action and are unawara of any potential licensing difficulties related to environmental protection matters.
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2.0 NEED FOR pR0p0 SED ACT!ON Surry Units 1 and 2 are Westinghouse pressurized water reactors (PWRs') with 157 fuel assemblies per unit'. The spent fuel pool at Surry has a capacity ~
8 of 1044 fuel assemblies . After off-loading spent fuel in March 1985, space for 159 assemblies, or two more than full core reserve capability will remain in the spent fuel poo181 Although maintenance of full core reserve capacity is not a safety matter, many power plant owners consider maintenance i of this capacity desirable for operational flexibility *. According to the FGEIS, experience has shown that the capacity for fully unloading a reactor has been useful in making modifications and repairs to reactor structural components and for periodic reactor vessel inspections 8 . The spent fuel
, storage capacity of the spent fuel pool at Surry has been expanded; additional c
expansion of the spent fuel pool is not viable as discussed, along with other alternatives, in Section 3.0. The applicant's need is to provide spent fuel storage to avoid shutdowns before the end of the useful life of the Surry Power Station. The Dry Cask ISFSI is one of four methods which could provice j additional interim storage capacity. The applicant's preferred alternative is the proposed Dry Cask ISFSIt'.
Assuming all approvals are obtained and the installation is completed by spring 1986, the movement of spent fuel from the pool to the Surry ISFSI could be accompitshed without extended loss of full core reserve capability. The proposed dry cask ISFSI .is designed to store all of the anticipated spent fuel resulting from Surry Units 1 and 2 coeration in excess of tne present design capacity of the spent fuel pool.
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l-l l 3.0 ALTERNATIVES
- The following sections present alternatives to the proposed action. The alternatives were considered against the need for the proposed action discussed in Section 2.0. For Itkely alternatives, impacts are addressed.
- 1. Ship spent fuel to a permanent federal repository.
- 2. Ship spent fuel to North Anna.
- 3. Increase the storage capacity of the existing spent fuel pool.
- 4. Construct a new independent spent fuel storage pool at the Surry site.
- 5. Ship spent fuel to a reprccessing facility.
- 6. Ship spent fuel to a Federal Interim Storage (FIS) facility.
- 7. Improve fuel usage.
- 8. Operate Surry Power Station at reduced power.
- 9. Ship spent fuel to other utility companies' reactors for storage.
- 10. Construct an ISFSI at a site away from the Surry Power Station.
- 11. No action.
3.1 SHIP SPENT FUEL TO A PERMANENT FEDERAL REPOSITORY This alternative would be VEPCO's preferred alternative. The Department of Energy (00E) is developing a repository under the Nuclear Waste Policy Act of i 1982 (NWPA), but is not likely to be ready to receive spent fuel before 1998.
Therefore this alternative does not meet the near-term storage needs of the applicant.
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l 3.2 SHIP SPENT FUEL TO NORTH ANNA l
l VEPCO, under a separate licensing action, has appli.ed for an amenament to the North Anna Power Station operating license to transship up to 500 spent fuel assemblies from its Surry Power Station to its North Anna Power Station for storage in the spent fuel pool there. This alternative is being reviewed in parallel with the Surry ISFSI application and, if approved, would provide additional storage of spent fuel from the Surry Power Station until the early 1990's.
The environmental impact of VEPCO's application for increasing spent fuel storage at North Anna Units 1 and 2 and the transshipment and receipt of Surry 1 and 2 spent fuel at North Anna has been separately assessed by the NRC. The action was found to have no significant impacts.88 '
Subsequent to the application, VEPCO entered into a agreement with Louisa ,
County, Virginia, to limit the number of assemblies transshipped to 130.88 This amount would only provide VEPCO a few years additional storage, then more capacity would be required.
Presently, the North Anna licensing action is in litigation before an Atomic Safety and Licensing Board (ASLB). Becsuse of this, there exists some uncertainty about the availability and timely implementation of this .
alternative. However, if approved, this alternative would only provide a short-term solution to the storage proolem at the Surry Power Station and therefore does not meet VEPCO's extendea spent fuel storage needs.
4
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'3.3 INCREASE THE STORAGE CAPACITY OF THE EXISTING SPENT FUEL PCOL In order to provide. increased spent fuel storage capability, many utiitties are altering the racks that hold the spent fuel assemblies in the spent fuel storage pools. When using this procedure, the structural framework of the spent fuel storage pools must be able to withstand the additional stresses caused by the increase in weight of the spent fuel to be stored. The app 11-cant has already increased the original capacity of the Surry spent fuel pool and has determined that if any significant additional increase in storage capacity were made to the Surry spent fuel pool the structural design safety
- criteria would be exceeded. The applicant has determined that it cannot store more spent fuel than the present licensed capacity at the Surry Power Station spent fuel storage pool.' .
3.4 CONSTRUCT A NEW INDEPENDENT SPENT FUEL STCRAGE POOL AT THE SURRY SITE l .
Additional storage capacity could be achieved by building a new spent fuel storage pool similar to that existing at the Surry Power Station. The i
NRC has generically assessed the impacts of this alternative and found that "the storage of LWR spent fuels in water pools has an insignificant impact on the environment."8 However, it does not appear that a new storage pool and the equipment for transfer could be designed, licensed and constructed in time to meet VEPCO's immediate need. .
)
3.5 SHIP SPENT FUEL TO A REPROCESSING FACILITY 4
Reprocessing of the Surry Power Station spent fuel is not viable because there is no operating commercial reprocessing facility in the United States, nor is there the prospect for one in the foreseeable future.
3.6 SHIP SPENT FUEL TO A FEDERAL INTERIM STORAGE (FIS) FACILITY Under the Nuclear Waste Policy Act of 1982 (NWPA) the federal government has the responsibility to provide not more than 1900 metric tons capacity for
.the . interim storage of spent fuel. The impacts of storing spent fuel at a FIS facility fall within those already assessed by the NRC in NUREG-0575.8 In passing the NWPA, Congress found.that the owners and operators of nuclear power stations have the primary responsibility for providing interim storage of spent nuclear fuel. In accordance with the NWPA and 10 CFR Part 53, ship-ping spent fuel to a FIS facility is considered a last resort alternative.
Therefore, while VEPCO pursues its own licenseable alternatives that can be reasonably provided in a timely manner, this alternative is not considered pertinent.
- 3.7 IMPROVE FUEL USAGE Under this alternative fuel assemblies would be used in the reactor longer '
thereby reducing the amount of spent fuel generated. The applicant is presently particioating in a COE program to extend the burnup of fuel G
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i assemblies at its North Anna and Surry Power S~tations. While this alternative may reduce the ultimate amount stored and lengthen the time when increments of
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increased storage capacity will be needed, it does not eliminate the need for increased storage capacity and is therefore not viable.
i 3.8 OPERATE SURRY POWER STATION AT REDUCED POWER Operating the Surry reactors at reduced power levels would extend the life of the fuel and thereby reduce the amount of soent fuel generated. This alter-native, like improving fuel usage, merely postpones the time when additional capacity is required and the amount needed. Also, operating at reduced power would not make effective use of available resources, thus causing economic penalties., Therefore, this' alternative, is not considered viable.
I 3.9 SHIP SPENT FUEL TO OTHER UTILITY COMPANIES' REACTORS FOR STORAGE l
l, In 1979, VEPCO explored this alternative with several neighboring utilf ties with unfavorable results'. The NWPA and 10 CFR Part 53 clearly place the responsibility for the interim storage of spent nuclear fuel with each owner or operator of nuclear power plants. Thus, for utility companies faced with their own potential storage capacity limitations, accepting another utility's spent fuel for storage is not very attractive. Therefore, this is not considered a practical or reasonable alternative.
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3.10 CONSTRUCT AN ISFSI AT A SITE AWAY FROM THE SURRY PCWER STATICN 4
The construction of an ISFSI at a location other than at the Surry site could provide additional storage capacity for VEPCO. The ISFSI could be a dry type similar to the one proposed or a pool type similar to alternative four. The only difference between the proposed ISFSI and this alternative would be that -
an ISFSI away from the reactor site would require offsite shipment of spent fuel and construction of a fuel handling facility. This alternative would be more costly thah the proposed action and would have the additional environmental
! impacts associated with offsite transportation of spent fuel. These impacts are generically accounted for in 10 CFR 5 51.52 (Table S-4). However, the NRC
, has generically assessed the impacts for this alternative and found that LWR .
a spent fuel storage in pools has an. insignificant impact on the environment and that dry storage appears to be environmentally acceptable.2 There is some
.i' doubt about the availability of alternative sites and tne timeliness of implementing such an ISFSI (about five years).* Therefore this alternative would not fulfill VEPCO's immediate need for additional storage capacity.
3.11 NO ACTION l
If no action were taken, VEPCO would be forced to shut down operations at its Surry Power Station. This would result in no more production of spent fuel thereby eliminating the need for increased spent fuel storage cacacity. '
The impacts of curtailing the generation of spent fuel by ceasing the operation of existing nuclear power plants when their spent fuel pools become filled was 6
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9 evaluated and found to be undesirable.2 This alternative would be a waste of an available resource, the Surry Power Station itself, and is not considered viable. ,
3.12
SUMMARY
OF ALTERNATIVES In summary, only four of the alternatives could provide a solution to VEPCO's spent fuel storage problem; ship spent fuel to North Anna, construct a new independent spent fuel storage pool at the Surry site, construct an ISFSI at a site away from the Surry Power Station, and ship spent fuel to a FIS f acil t.ty. Transshipping spent fuel for storage at North Anna has been separately assessed by the NRC and found to have no significant impa' cts, but would only
- provide a tusporary solution to VEPCO's need for increased spent fuel storage capacity at the Surry Power Station. Con-struction of an additional spent fuel storage pool at the Surry site or an ISFSI away from the Surry site could provide long term increased storage capacity for VEPCO with insignificant environmental impact. However, they cannot be implemented in a timely manner to meet VEPCO's immediate need for additional capacity. The impacts for the alternative of shipping to a FIS facility would be similar to those for the offsite ISFSI alternative. However, this is only viable as a last resort.
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4.0 ENVIRONMENTAL INTERFACES The environment of the Surry site and region have been described in the previous Surry FES's. The applicant has updated the environmental descrip-tions with information in Chapter 2 of the Surry ISFSI Environmental Report'.
Those environmental features which the staff believes most likely to interface
, with the construction and operation of the Surry ISFSI are summarized in this section. Most of the environmental effects are expected to be limited to the Surry site. For some of the interfaces (e.g., socioeconomics and radiological
, dose to humans), the staff considered the region of interest to extend off-site out to an 80 km (50 mi) radius from the Surry site. The staff's assessments of th'e* potential environmental effects of ISFSI construction and operation are presented in.Section 6...
- 4.1 SITE LOCATION, LAND USE AND TERRESTRIAL RESOURCES The Surry ISFSI is to be located on the Surry Power Station site approximately 1,000 m (3,300 ft) southeast of Units 1 and 2 reactor building within the site boundaries of the station (Figs. 4-1 and 4-2). The ISFSI facility will occupy approximately 6 ha (15 acres) (Reference 7, Section 2.1.1.1).
At present there is an existing low-level waste storage facility consisting of a concrete slab 30.5 m by 30.5 m (100 ft by 100 ft) covered by a Butler -
building and a gravel road leading to it from service , road along the intake
! canal (Figs. 4-2 and 4-3).
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. The remainder of the* ISFSI site consists of an open woods of mixed pines and hardwoods that was last logged in the late 1960's (Reference 24, Section 2.7.7).
A discussion of the wildlife inhabiting the Surry site are provided in Reference 24, Section 2.7.2. ,
A breeding pair of Southern bald eagles, Haliaeetus leucocephalus, is nesting onsite (Reference 7, Section 4.1.6.3). This species is on the U.S. Fish and Wildlife Service's endangered list in Virgin,f a. The nesting site is located approximately 823 m (2700 ft) from the ISFSI site (Referenet 7, page 4.1-5).
The Surry nuclear power site is bounded on the north and approximately 2/3 on its south side by the Hog Island State Wildlife Management area (Fig. 4-2).
The wildlife management area is used primarily for the protection of migratory waterfowl, mostly Canada geese, Anser canadensis, and pintails, Anas acuta, (Reference 6, Section 2.7.1). Peak water-fowl populations approach 25,000 birds (Ibid.).
4.2 WATER USE AND AQUATIC RESOURCES The Surry site is located on Gravel Neck peninsula which is bordered on three sides by the James River (Fig. 4-1). Detailed descriptions of hydrology, water use and aquatic biota in the vicinity of the site are provided in Reference 4, Sections 2.1.3 and 2.7 and Reference 5, Sections 2.5 and 2.7.2. The site grade for the ISFSI concrete slabs will be 35 ft (ms1). Surface drainage is from the ISFSI to the James River toward the north.
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Since issuance of th FES' for Units 3 and 4, the discovery in 1975 of Keoone contamination at the James River has resulted in a total ban or partial closure of fisheries depending on species, river segment or type of fisheries; i.e.,
recreational or commercial (Refereneg,7, Section 2.1.3).
4.3 SOCI0 ECONOMICS AND HISTORICAL, ARCHAEOLOGICAL AND CULTURAL RESOURCES 6
The immediate region surrounding the Surry Power Station is rural. The site, however, is within commuting
- distance of several sizeable metropolitan areas. The closest cities are Williamsburg and Newport News. At further J
i distances are Hampton, Portsmouth, Norfolk, Virginia Beach, Petersburg, Hopewell and Richmond.
- i While the Surry Station is located in a region rich in natural and man-made
)
historical sites there does not appear to be anything of historical interest within the boundaries of the site.
The socioeconomic character of the region and cultural resources have pre-viously been described in other reports by VEPCO',8' and by the AEC' and
! NRC8'.
4.4 DEMOGRAPHY 1
I i Residential population within 5 miles of the Surry site is small, estimatec l by NRC to oe aaproximately 1,360 persons or 17 persons per square mile in 9
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1980. Between 5 and 10 miles the resioential population density increases with the inclusion of Newport News and Williamsburg. NRC estimates that the 1980 residential population within 10 miles of the Surry site was approx-imately 71,300 persons. Within 50 miles of the site the 1980 residential population was approximately 1,684,500 persons. The staff has compared its residential population estimates with the applicant's and finds general consistency. Differences in estimates are attributable to difference in the estimating techniques used. Transient population within 10 miles is relatively large. Peak transient population estimates furnished by the applicant are taken from the " Virginia Radiological Emergency Response Plan" dated June 1983.' Total peak transient population within approximately 10 miles is 63,755 persons. Colonial Williamsburg and Busch Gardens account for 50,400 of this total. Institutional population within approximately 10 miles is, 15,290 persons.
4.5 METEOROLOGY AND CLIMATOLOGY This section summarizes the regional climatology and local meteorology including information on savare weather and atmospheric diffusion conditions.
The effects of heat dissipation from the casks and the potential increase in fogging due to ISFSI operation are addressed in Section 6.2.2. Review of the applicant's assessment of atmospheric diffusion conditions for use in accident analyses is presented in Section 4.5.4. Causes of accidents postulated by the i apolicant include tornadoes which are addressed in Section 4.5.3. Also, see the staff's assessment of accicents in Section 6.2.1.3.
o
4.5.1 Regional Climatology The Surry site is located in a zone of transition between continentai and -
marine climates, exhibiting characteristics of each. The climate is generally moderate, influenced during much of the year by the anticyclonic
- circulation of the Azores-Bermuda high pressure system. Summers are warm and humid, resulting from the dominance of tropical maritime air masses ,
over the area. Winters are generally mild, with continental and maritime air masses alternating over the area. Temperatures are moderated due to the proximity of the Atlantic Ocean. The Appalachian Mountains to the west 1
act as a partial barrier to outbreaks of cold, continental air in the winter, usually delaying the advance of the cold air long enough to moderate the temperatures associated with it. The site is principally ,
affected by storms originating along the southeast coast of the U.S. and tracking northeastward along the coast.
4.5.2 Locaf Meteorolooy Data from Richmond and Norfolk (50 miles NW, and 35 miles SE of the site, respectively) have been used to characterize the local meteorology of the Surry site. Richmond data exhibit more continental characteristics, and Norfolk more maritime. Mean monthly temperatures at Richmond range from 3*C (38'F) in January to 26'C (78'F) in July, while mean monthly temoeratures at Norfolk range from 5'C (41*F) in January to 26'C (78'F) in July. Record minimum temperatures have been -24*C (-12'F) at Richmond I
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(January 1940) and -17'C (2*F) at Norfolk (February 1895). Record maximum temperatures have been.42' (107'F) at Richmond (August 1918) and 41*C (105'F) at Norfolk (August 1918). Temperatures of 31*C (90*F) and above can be expected about 43 days per year at Richmond, and 31 days per year at Norfolk. Temperatures of O'C (32*F) or below can be expected about 92 days per year at Richmond, and only 62 days pe'r year at Norfolk. -
Annual average precipitation in the area is about 1120 mm (44 inches), with the maximum monthly means occurring in June, July and August when about 130 mm (5 inches) can be expected. The maximum monthly precipitatien i recorded at Richmond.was about 480 mm (19 inches) in July 1945. Snowfall varies greatly in the region, with Richmond expecting an annual snowfall of 370 mm (14.7 inches), and Norfolk expecting 190 mm (7.4 inches). The maximum 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> snowfall' reported at Richmond was 550 mm (21.6 inches) in January 1940, while Norfolk has reported 450 mm (17.7 inches) in December 1892.
Wind data from the Surry site for the 45 m (147 ft) level, representing the l
t period 1974-1981, indicates predomine1t wind directions from the southwest and south-southwest, occurring 19.6 percent of the time. North-northeast and northeast winds occurred least frequently, at a frequency of 8.2 percent of the i
time. The prevailing wind direction at Richmond is south, and at Norfolk the prevailing wind direction is southwest. During the time period 1949-1980, mean wind speeds were 3.3 m/s (7.5 mph) and 4.7 m/s (10.5 moh) at Richmond and Norfolk, respectively.
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Data from Richmond and Norfolk indicate averages of 29 and 22 days per year of heavy fog (defined as visibility of h mile or less), respectively.
4.5.3 Severe Weather .
i A variety of severe weather, from ice and snow storms to tornadoes and hurricanes, can affect the site area.
Spring and summer circulation patterns bring a strong flow of warm, moist unstable air into the area, and resulting thunderstorms are not uncommon.
Thunderstorms can be expected on about 37 days per year, being most frequent in July with a monthly average of between 8 and 9 days with thunderstorms. About 4
75 percent of the annual number of thunderstorms days occur during the months
, of May through August. Thunderstorms are least frequent during the months of December, January and February, with monthly averages of less than one-half day ,
with a thunderstorm.
Ouring the period 1954-1981, 27 tornadoes were reported within the one-degree latitude-longituoe square containing the site, giving a mean annual tornado frequency of about 1. The computed probability of a tornado strike at the plant site is 9 x 10 -5 per year.
In the period 1871-1981, 52 tropical storms or hurricane centers passed within 35 km (100 nautical miles) of the site. The maximum " fastest mile" of winds, recorded at Richmond and Norfolk, were 30 m/s (68 mph) and 36 m/s (80 mph), resoectively.
, - - v
f During the period 1955-1967, in the one-degree latitude-longitude square containing the site, there were 7 windstorms of 26 m/s (50 knots) or greater, and 9 reports of hail 19 mm (3/4 inch) or larger.
Ice storms of freezing rain or glaze are not uncommon in the winter, but they are seldom severe enough to do any considerable damage. The most notable glaze storm at Richmond was during January 27-28, 1943, when heavy damage was done to trees and overhead transmission lines of all kinds. No quantitative statistics are available to the staff at this time.
Fifty-one atmospheric stagnation cases of 4 days or more were reported in
-the site region during tne period 1936-2965. The highest monthly frequency of these cases was in October. Three cases of atmospheric stagnation
, lasting 7 days or more were reported for the area in the period 1936-1965.
4.5.4 Atmoseheric Diffusion Conditions for Accidental Releases In & letter 18 to the staff, tne applicant has provided an assessment of atmo-spheric diffusion conditions for use in determining the effects of a loss of confinement barrier accident at une proposed Dry Cask Independent Spent Fuel Storage Facility (ISFSF). To determine relative concentrations (X/Q)
~
for this accident, the direction-dependent atmospheric dispersion model described in Regulatory Guide 1.145, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants" was . ,
, used. Seven years (1976-1982) of onsite meteorological data provided inouts to the evaluation. The input carameters incluced nourly average e , ,. , - . , . , , , - - - , , , , - - - - - - , . - , , - , - - - - - - - -- . - - - - - - ,,-,.,--,-----,.,a..-,,-,-,---...n,-,,,e--a._--
values of wind speed and wind cirection at the 10 m (33 ft) level, and atmoscheric stabiJity determined from temperature cifferences measured between the 10 m (33 ft) and 45.7 m (150 ft) levels of the onsite meteorological tower.
From the array of X/Q values presented by the applicant, the staff selected for use in the accident assessment presented in Section 6.2.1, 6.7 X 10-4 sec/m3 at the controlled area boundary 0.5 km-(0.3 miles) northwest of the ISFSF and 3.8 x 10 -5 sec/m3 at the nearest residence 2.5 km (1.5 miles) south-southwest of the ISFSF. According to the calculations, these values are expected to be exceeded 0.5 percent of the time. These assessments, as made by the appiteant, are reasonable and provide an acceptably low probability of being exceeded.
4.6 GEOLOGY AND SEISMOLOGY i
This section summarizes geological and seismological features in the vicinity of the site. One of the causes of a postulated accident of the Surry ISFSI considered by the applicant is an earthquake. (See the staff's assessment of accidents in Section 6.2.1.3.)
l l
l 4.6.1 Geology l
The ISFSI site is located in the Coastal Plain physiographic province. In Virginia this province has a stair-step character, composed of a series of plains that are successively lower from west to east. In the vicinity of the site the upper 6.1 to 10.7 m (20 to 35 ft) consists of layers of brown and mottled crown sand, silty sand, and organic and inorganic silts and clays.
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In late-Pleistocene time the sea level rose to about +45 feet, accompanied by the deposition of. clayey sands. From the end of this period to the present the sea has been receding and erosion has been occurring. Regional subsidence in the site area has been measured to be 1 to 5 mm (0.04 to 0.2 in) per ; ear.
A site survey conducted in May 1975, indicated that this was not a problem.
In the immediate site area there are no surface features indicative of actual or potential localized subsidence or landslidir.g. There is no history of surface mining or other activities by man, which would cause ground disturbance.
4.6.2. Seismolooy
- No earthquake within the last 200 yea s has been large en'ough to cause structural damage at the' site. There are no known earthquake epicentral locations within 48 km (30 miles) of the site.
l Liquefaction is not likely to occur in local strata for an earthquake having a maximum ground acceleration of less than .07g.
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5.0 DESCRIPTION
OF SURRY ORY CASK ISFSI 5.1 ISFSI LOCATION The Surry ISFSI will consist of ser. led surface storage casks arranged on three concrete slabs which will be constructed onsite (see Figures 4-1, 4-2 and 4-3).
5.2 SITE PREPARATION The 6 ha (15 ac) of approximate land area set aside for the ISFSI may be cleared of vegetation (see Section 6.1.1). The areas to be occupied by the concrete slabs will be excavated one at a time with about 3060 m3 (4000 yd3) of material removed per slab. This spoil material will be stored near the ISFSI excavation site. The excavations w't11 be filled and compacted with suitable material to support the slabs. Each slab will be formed with ready-sixed concrete trucked into the site. The approximate area covered by each slab is
~
684 m2 (7360 f 2). The depth of the slab may be up to 0.68 m (2 ft 3 in);
thus, the expected volume of concrete required for the three slabs is 465 m3 (613 yd3). Temporary construction butidings will be erected on the site and removed at the completion of construction of the ISFSI.
5.3 STORAGE SYSTEM I
'l The Surry ISFSI is designed for the dry storage, in casks, of spent nuclear fuel, originating from the Surry Power Station. The initial cask design selected by the applicant for licensing and use at the Surry ISFSI is the
CASTOR V/21 cask.8* The CASTOR V/21 cask was designed by Gesellschaft fpr Nuklear-Service mbH (GNS) to meet the International Atomic Energy Agency (IAEA) specif tettions for Type 8(U) packaging, corresponding to Nuclear Safety Fissile Class I, " Regulations for the Safe Transport of Radioactive Materials." The cask is a thick wall nodular cast iron cylinder wnich weighs 106 metric tons, fully loaded. It is about 4.9 m (16 ft) high and 2.4 m (8 ft) in diameter (see Figures 5-1 and 5-2). The CASTOR V/21 is designed to safely store 21 PWR spent fuel assemblies by providing confinement, shielding, criticality control and heat removal.
5.3.1 Material to be Stored L
The. material to be stored at the Surry ISFSI is spent nuclear fuel used at the Surry. Power Station. The fuel used during the first years of operation had an initial enrichment not exceeding 3.5 weight percent U-235 and a discharge burnup not exceeding 35,000 MWO/MTU. There are about 900 spent fuel assemblies meeting these criteria stored in the spent fuel pool at Surry and being considered for dry' cask storage. This fuel is the design basis fuel considered for this assessment.
l Although the Surry Power Station has been authorized to operate with fuel of higher initial enrichment and to higher burnups, it is not presently being considered for licensing in the Surry ISFSI. For the purpose of this environmental assessment, the radiological impacts, based on the design basis "r
fuel in CASTOR V/21 casks, have been multiplied by a factor of three to a 2
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4 encompass potential impacts that may result from variations in future cask designs or vendors and from storing spent fuel with higher initial enrienment and burnup.
In order to meet certain cask design criteria, the PWR spent _ fuel to be stored in the CASTOR V/21 cask will'not exceed the following design basis characteristics.
- 1. Initial enrichment of 3.5 weight percer.t U-235.
- 2. Maxim 6m burnup of 35,000 MWD /MTU at a specific power of 35 MW/MTU.
- 3. Peak thermal power of 1 kW/ assembly.
- 4. Out of the reactor not .less than five (5) years. ,
- 5. 9.765 MTU/ cask (.465 MTU/ assembly). .
5.3.2 Cask Design and Safety Features l
The CASTOR V/21 cask design safety report ** is reviewed and a safety evalua-tion issued in addition to a safety evaluation of the use of this cask at the Surry ISFSI before the issuance of the license. The safety evaluations review design adequacy, site parameters, and operations to ensure confinement, shield-ing, criticality control and heat removal of the spent fuel and to ensure safe l
l storage.
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CONFINEMENT Confinement of the spent fuel is achieved by the cask and two stainless steel, lids which are bolted to the cask body. Each lid is sealed shut using multiple metal and elastomer 0-ring seals. The lid system integrity is en.sured by monitoring the pressure between the two-lids. The cask is designed ,
to maintain its structural integrity and stability against external impacts, cask drops and severe environmental loads. Thus, because of the cask struc-tural integrity and tightness, it is a safe confinement barrier against the release of radioactivity and loss of the helium cover gas.
SHIELDING The 379 mm (15 in) thick nodular cast f ran walls of the cask provide gamma shielding. For additional neutron shielding, two concentric rows of axial holes in the cask wall are filled with polyethylene. When the CASTOR V/21 is filled with the design basis fuel, the surface dose rate at the side of the cask wall is about 30 mres/hr (7.8 mrem /hr neutron and 22.3 mrem /hr gamma).
CRITICALITY CONTROL l
The basket (see Figure 5-3) used to hold the spent fuel assemblies inside the cask is made of stainless steel and baronated stainless steel. The boron in the stainless steel provides sufficient neutron absorption to maintain subcritical conoition of the fuel in the cask. The basket is easigned to
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l FIGURE 5-3 CASTOR V-21 FUEL SASKET (Source: Ref. 26, Fig. 1.2-6)
maintain its structural integrity unoer the same accicent conditions as the cask. Thus the scent fuel elements are ensured to remain subcritical under all storage conditions.
i HEAT REMOVAL-
/
I'n order to ensure fuel cladding integrity, the CASTOR V/21. cask is designed to keep cladding temperatures below a maximum specified temperature of 370*C (698'F). Tzp cask is designed for a maxim 0m thermal power load of 21 kW under extreme environmental conditions. The cask has an inert helium gas atmosphere which not only innibits corrosion but acts as a heat transfer medium. The decay heat of the fuel assemblies is conducted through the cask body and
~
transferred to the air surrounding the cask by natural convection and radiation from exterior surfaces. The cask has 73 cooling fins on the outside to assist heat transfer. The passive nature of heat removal is an additional safety feature of the cask. Heat is transferred directly to the air. No cooling water is required. 'The maximum design surface temperature of the cask is 82*C (180*F).2' 5.4 ISFSI OPERATIONS The Surry ISFSI, by the nature of its passive dry cask storage, has simple operations. All cask loading and preparations take place at the Surry Power .
Station spent fuel storage building under the reactor operating licenses.
There, after fuel is placed in the cask, the primary lic is set in place and tigntened. Water is drained from the cask and the cask cavity is vacuum
. - _ - - - - . - - - L. - - . . , , - . - .
i dried. A helium leak test is conducted to ensure tightness of the lid. The cavity is then filled with helium slightly below atmospheric
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pressure (0.8 bar). The second lid is placed on the cask and the seals l tested. The space between the two lids is filled with helium to 7 bar-abs l 1
pressure and the seal monitoring system is then activated. The outside of the cask is decantaminated before the cask is moved out of the Surry fuel and i
decontamination buildings and transferred to the ISFSI. Once the cask is set i-in place on the concrete pad, the seal monitoring system is connected. No maintenance is required other than periodic visual or functional checks of the seal monitoring system. The applicant anticipates that about four casks per year will be pit ed in the ISFSI. Existing opersting plans, practices and procedures (i.e., health physics, environmental monitoring, security, and quality assurance) for the Surry Power Station will be adopted for use at the i Surry ISFSI. !
i 5.5 MONITORING PROGRAMS 5.5.1 Containment Seal Monitorino t-l Preparation of the cask seals is described in Section 5.4. The seal monitoring system is activated while the cask is in the spent fuel storage building.
! When the cask is set in place at the ISFSI site, the seal monitoring system is connected. The system uses a diaphragm-type pressure gauge which monitors -
l the pressure set up between the primary and secondary lids.8' Pressure 1
changes result in opening a contact of the pressure-actuated switch in the main gauge, generating an electrical signal to the control room. The signal
is displayec visually or by an annuciator. A second
- gauge of the same type is used to conitor the reference pressure of 4 bar established in the main gauge. The second gauge actuates a visual and acoustic signal if the reference pressure in the main gauge falls below a preset level of 3.5 bar. '#o maintenance of the casks at the ISFSI is required other than periodic visual or functional checks of the seal monitoring system, 5.5.2 Radiological Monitoring In addition to the existing radiological monitoring program for the Surry Power, Station, the applicant plans to add thermoluminescent dosimeters to be located at appropriate intervals along the ISFSI perimeter fence. The contribution of the ISFSI to the offsite doso will be reported in the station Radiological Environmental Monitoring Annual Reports. The method for deter-mining the ISFSI contribution will be documented in the offsite dose calculation manual.
5.5.3 Non-radioloaical Monitoring The ongoing onsite meteorological measurements program for the Surry Power Station provides a data base for evaluation of the station and ISFSI operations. The primary meteorological tower measures wind speed, wind direction and horizontal component directional fluctuation at two levels
[i.e. at 10 m (33 ft) and 45.7 m (150 ft)], amoient air and dewpoint tempera-ture at the lower level, differential air temoerature between levels and l
l -
- - - - --- w.-.e ,_,,--.c_, -
-. r - - - , -,-r, --~-- --m,n ,-,mr-- , , - - -e,-- e- - - - -- - - - - - - - - - - - - - - - - - - , , , --
rainfall at the tower base. A backup tcwer monitors wind speed, direction and hori: ental component fluctuation in direction at 9.2 m (30.3 ft) el. No other non-radiological monitoring specific to the ISF5I is required. ~
9 0
e t
e S
O l -
G @ .e6 -
y , -. ,..__ _-- -, -.. , , . . _ , . ...w_, -- -- .w... ,m
O 6.0 ENVIRONMENTAL IMPACTS OF pR0p0 SED ACT!CN 6.1 CCNSTRUCTION IMPACTS 6.1.1 Land Use and Terrestrial Resources , ,
Approximately 6 ha (15 acres) of open. woods and the wildlife habitat this area contains will be destroyed. fhisisasmallareawithnoknownrare, endangered or threatened species of plants or animals. The nesting pair of bald eagles were not disturoad by the construction of the low-level waste
, facilities on the ISFSI site (Reference 7, page 4.1-5). Therefore, the impact of constructing the ISFSI site at the same distance from the nesting site is expected to have a negligible effec't on land use and terrestrial resources, including the nesting bald eagles. *
- 6.1.2 Water Use and Aouatic Resources Construction of the Surry ISFSI is not expected to impact local water users, water quality or aquatic biota. Those activities which could potentially impact surface waters are the clearing and axcavation operations conducted a
prior to pouring the concrete slabs. The applicant indicates that a temporary drainage system may be constructed to collect runoff into temporary settling ponds and that more permanent drainage will be provided as excavation and backfill operations proceed. '
No dewatering during excavation is anticipated. Concrete for the slabs will be ready-mixed so no water use nor wastes from concrete batch ooerations will
- -- , , - - - m--y y- ,-e - . ~ . , - . - _ , - , m.--y-,.- , - . - - . _ , - -. - . , , , , _ . ,,.m.
f result. Existing site facilities for barge off-loacing may be used; no credging or other construction activities en the James River will be required for the ISFSI. -
6.1.3 Other Imoacts of Construction Because of the small size of the construction effort impacts to air quality and ambient noise levels and other potential impacts associated with construction are expected to be well within the bounds of impacts previously considered and found acceptable at the Surry site and unlikely to be discernible from other impacts associated with normal station maintenance activities. With good construction impact control practices (Ref. 7, Section 4.5), the potential.for fugitive dust, erosion and noise impacts
~
typical of the planned constructfon activities can be controlled to minimal level s.
The construction effort for the ISFSI is small relative to initial nuclear station construction. A peak construction force of about 20 peoole is antici-pated (Ref. 7, p. 4.1-3)). Construction of Surry Unit 1 involved a peak work force in excess of 2000 people". A typical scheduled refueling and maintenance outage brings several hundred workers to a nuclear station site. Construction of the ISFSI is also limited in scope, involving primarily clearing and grading, and pouring concrete pads. This type of construction is not unique and is an -
activity of a type which the licensee could do without NRC approval were it not for the purpose herein described, anc the reouirement of a Part 72 license for fuel storage.
- 1 l
6.1.4 Socioeconomics The socioeconomic effects associated with operation of the facility wi.11 be essentially nil as no additional operating work force will be required.'
6.2 OPERATIONAL IMPACTS 6.2.1 Radioloofcal Imoacts from Routine Operations There are three pathways by which workers and members of the public may be 1
exposed as a result of the Surry Independent Spent Fuel storage installation (ISFSI) operation: to direct radiation; to radioactivity released in gaseous effluents; and to radioactivity released in ifquid effluents. Because the proposed ISFSI involves only dry storage of spent nuclear fuel in selected casks, there will be essentially no liquid or gaseous affluents associated with storage activities. Although activities associated with cask loading and decontamination may result in some liquid and gaseous effluents, these operations will be conducted at the Surry Power Station under the 10 CFR Part 50 operating licenses. The radiological impacts from those effluents fall within the scope of impacts from reactor operations which were assessed in the Surry Power Station Unit I and Unit 2 Final Environmental Statements *.
The primary exposure pathway associated with normal Surry ISFSI operations is '
i
! direct f eradiation of nearby residents and site workers. The radiological dose estimates presentsd were calculated using conservative and design basis
,..,_r, . _ _ _ - , . , , , . . _ _ . - - . - . . _ , .
,,. ,, -- .__.,y- - _ _ _
assumptions: Maximum cask surface dose rates of 7.8 mrem neutrons and 22.3
~
mrem /hr gammas, maximum fuel burnuo of 35 GWD/MTU8', fuel out of the reactor at least 5 years before storage, no self-snielding of casks in arrays, and emplacement of four casks per year. The resultant calculated dos,es were then multiplied by a factor of three (3) to provide an upper bound of the radio-logicalimpactsassociatedwiththe.potentialstorageofhigherburnupfueland to accommodate other potential cask designs. . These assumptions result in conservative dose estimates; actual doses are expected to be lower.
6.2.1.1 Offsite Oose Commitments 1
ISFSI operations will also result in additional dose to members of the public -
. from direct radiation exposure. Section 72.67(a) of 10 CFR 72 requires.that, '
from normal operations, dose equivalents to any real individual located beyond the ISFSI controlled area not exceed 25 mrem /yr to the whole body as a result of planned affluents releases, direct radiation and other radiation from uranium fuel cycle operations within the region.
Appendix I to 10 CFR 50 sets forth design objective dose commitment guides for liquid and gaseous effluents released from nuclear power reactors. For each reactor, the maximum annual dose commitment to an individual in an unrestricted i
area is 3 mrom due to liquid effluents and 5 mrem due to gaseous effluents. ;
I Thus, the maximum design guida dose commitment from effluents due to Surry !
I Power Station operations would be 16 mrem /yr. Based on its usage, the Low 1 l
Level Waste Storage Facility (LLWSF) would contribute an additional 4.aE-2 mrem /yr. Actual doses due to release of radioactivity in effluents are less O
m w-- . - - -y- w.-- - - ,,.-3----
than design amount. The estimated radiological doses due to Surry Power Station operations are .36 mrem /yr from gaseous effluents and 2.52 mres/yr from liquid effluents.'
The estimated maximum anneal dose commitment to the nearest real individual (located 2.5 km (1.5 mi) away] due to direct radiation from the casks at the Surry ISFSI is about 6E-5* mrem /yr. This dose is only a very small fraction of the design guide dose commitment and those estimated in the FES for the Surry Power Station operations. When combined with the dose commitment from reactor and low-level waste storage operations, the total dose commitment is well within the 25 mrem /yr limit specified in 10 CFR 72.67 and 40 CFR 190.
. Forty-eight permanent residents are located within 3.2 km (2 mi) of the Surry ISFSI. If all are assumed to be located as close as the nearest resident, then the collective dose commitment would be 3E-6 man-rem /yr due to Surry ISFSI operations. Based on ground-level air concentration dose rates in the Surry Unit 1 FES', the collective dose commitment to this same population within two miles would be about 4.3E-4 man-res/yr due to Surry Power Station operations.
Attenuation of the direct radiation dose rates from the ISFSI beyond two miles contributes little to the collective dose commitment for more distant popula-tions in the region under consideration. Compared to the estimated 92.4 man-rem /yr due to Surry Power Station operations,' the impact of the collecttve dose commitment in the region due to the Surry ISFSI is negligible. '
4
" SE-i = 6x10 ~5 i
6.2.1.2 Collective Occuoational Dose Commitment-Spent fuel storage at the Surry ISFSI will result in a small increase'in the total occupational dose at the Surry Power Station. Engineerett features of the casks and application of administrative controls ensure that all exposures are maintained at levels which are as low as reasonably achievable (ALARA).
t -
! VEPCO has estimated the maximum annual collective occupational dose commitment l
from the operation of the Surry ISFSI. The estimates were based on emplacing four casks per year and the design basis use of the collocated LLWSF. The additional _ exposure to workers at the Surry Power Station assumes the ISFSI is full.
Occupational doses during construction assumes 2060 man-hours to complete one concrete slab on which casks will be placed. Because the slabs are to be constructed as needed, the exposures during construction include contributions from the LLWSF and 28 casks on previously filled slabs.
If all the concrete slabs were initially constructed at one time, a small reduction (18 percent) in the total occupational dose from construction could be realized. However, because the applicant may want to use casks with greater I
i storage capacity, not all the slabs may be needed. Thus, by constructing the l
slabs at one time, the small reduction in occupational exposure must be weighed against higher initial capital outlay for something that eventually may not be
! needed.
l i
i I
. - ~ . . - . _, , . _, . . .._. -- _.
Table 6.2-1 summarizes the maximum collective oc'cupational dose commitments from annual operations and construction.17 The 23 man-rem /yr dose from normal
- operations and an average of 64 man-rem per slab construction constitutes a small fraction of the total occupational dose commitment at th,e Surry Power Station. Actual doses are expected to be less. For example, in 1982 the collective occupational dose at the Surry Power Station was 2119 man-rem, with .
an annual average coll' active occupational dose over ten years, ending with 1982, of 2315 man-rem /yr.2' Individual doses are controlled to be within the limits of 10 CFR Part 20.
6.2.1.3 Environmental Assessment of Accidents
. 'VEPCO, in .its application, postulated accidents of the Surry ISFSI due to a variety of causes: earthquakes, tornadoes, floods, fires, natural gas pipeline explosions, and an accidently dropped cask. All are either not credible for the Surry site or the cask is designed to withstand the resultant forces without losing its mechanical integrity. The only cask components with a potential to malfunction are the lid seals, the pressure gauge or the pressure gauge monitoring system. For assessment purposes, the aop11 cant postulated an accident scenario where a nonmechanistic simultaneous failure of both cask seals and all fuel cladding occurs, resulting in the loss of the helium cover gas and the radioactive noble gas inventory in the spent fuel for one cask. The assessment of accident impacts presented here is based on the inventory contained in the CASTOR V/21 cask which will be initially used by VEPCO.
7 c . . - - . - - - m _ - . . . _ . . . , , . . . . _ .
Table 6.2-1 COLLECTIVE ANNUAL OCCUPATIONAL AND CONSTRUCTION DOSE COMMITMENTS
- Annual Operations Man-Rem / Year Cask loading and decontamination at reactor (1) 5.5
' Transfer of Cask to ISFSI (1) .1 Cask Emplacement (1),(2) 4.0 Surveillance and Maintenance (2),(3) .8 Additional exposure to workers at the Low-Level Waste Storage Facility (2) 8.1 Additional exposure to workers at the -
Surry Power Station 4.3 Total Ffi (1) Four cask per year ..
(2) Assumes the ISFSI and the LLWSF are completely filled (3) Assumes 12 surveys, 2 instrument test /yr,1 instrument /yr and recalibration Construction Man-Rem Slab L 53
,, Slab 2 62
' Slab 3 78 Total T91
- Source: Reference 17 Includes enveloping factor of three l
l
- _ _ _ _ _ _ _ . -,. . . . - . _ _ _ - - - - . . . - . , . . N----. .____-_?_'____
The CASTOR V/21 cask was designed for storage and transportation of irradiated spent fuel assemblies and its design to fulfill the IAEA international specifi-cations for type B(U) packaging. Although storage of spent fuel is the only use evaluated in this report, a hypothetical worst-case accident based on transportation accident scenarios is being evaluated to establish an upper bound accident impact for storage applications. The transportation accident
, scenario is not considered credible for storage situations. It has been chosen merely to determi,ne estimates for release of radionuclides from the spent fuel to the cask cavity and then to the environment rather than arbitrarily assuming a nonnechanistic accident release.
Dht release fractions used in this analysis were based on Reference 28 for scenario 5' (a worst-case for air-cooled casks)~. This scenario considers all -
~
release mechanisms that are credible for air-cooled casks. The mechanism for release of radioactivity considered appropriate for this evaluation was an impact rupture which somehow causes mechanical disruption of the cladding and subsequent depressurization of 10 percent of the fuel rods. The fraction (20 percent) of the spent fuel inventory of noble gases generated in the reactor that are in the fuel pellet gap is released to the cask cavity.
Because of the low temperatures, the remainder of fission products released are assumed to be particulates that are swept out of the rods as they depressuri:e after rupture. The spent fuel inventory fraction that is swept out as particulates is 2E-6.
Once radionuclides have been released from the ' fuel rods they must then find a path out of the cask. The result of accident damage is not expected to provide a pathway with a large cross-sectional area from the cask cavity to the environment. Only a small section of a failed cask seal would,be the most likely release pathway. Before the radionuclides are released to the environ-ment, they must pass many places that are relatively cool and through small '
passages. As,a result, radionuclides can condense, plate out, or be filtered out before escaping the cask. For gas-cooled casks (in this case helium cooled) 60 percent of the noble gases in the cask cavity are assumed to be released and 5 percent of the particulates.
After the radioactive mafiP~fal escapes the. cask, there are two factors important in determining whether. the particles reach peopl'; e tne fraction that i
becomes suspended in air and the fraction that is respirable (less than 10 microns aerodynamic. diameter). Five percent of the particulates were assumed to be scaller than 10 microns and remain as an aerosol.
The radioactivity released to the cask cavity is based on the design fuel to be stored in the cask; PWR fuel, initial enrichment of 3.5 percent U-235, 35,000 MWO/MTU burnup, 5 years out of the reactor. The 0.5 percent ground level 1
direction dependent atmospheric dispersion (X/Q) values were used to calculate
! doses at the nearest controlled area boundary (503m) and at the nearest resident (2414m).15 '
Tables 6.2-2 and 6.2-3 summarize the radiological imoact of a CASTCR V/21 cask accident containing 5 year cooled spent fuel. The upper bound doses (with a l
- o e
w . ,- -- -- , ,v., , - - - , - - ,-ww-----r- , - - - , -
k Table 6.2-2 Radiological Doses at the Controlled Area Boundary from Storage Due to a CASTOR V/21 Accident at the Surry Power Station Whole-body Total Fraction Inhalation
- Cask Released i Inventory (1)*
Breathing Dose Conversion Dose at Controlled Aerosolized + X/Q(3) Rate (4) j Nuclide - (pci) Respirable (2) 3 Factors (5) Area Boundary-(sec/m ) (m8 /sec) (Rem /pCI) (Rem) k H-3 5.44E*9"* IE-2 1 6.73E-4 2.54E-4 1.25E-4 1.16E-3
! Kr-85 7.93E*10 lE-2 6.73E-4 l
1-129 4.05E+5 SE-10 N/A 3.34E-4(6) '"[" 1.78E-4 6.73E-4 2.54E-4 5.0 (thyroid) 1.7E-10(thyroid)
Cs-134 3.45E+11 SE-10 6.73E-4 2.54E-4 4.55E-2 1.34E-6
{ Cs-137 9.77E+11 SE-10 6.73E-4 ' 2.54E-4 3.26E-? -
2.72E-6
) Sr-90 6.89Eill SE-10 6.73E-4 2.54E-4 2.4E-2
- 1.41E-6 i
Ru-106 1.76Etll SE-10 6.73E-4 2.54E-4 6.18E-2 9.30E-7 Il Total Whole Body Dose 1.35E-3
- Footnotes:
1.
~~CNSI CKS10R V/21 Cask Topical SAR, 5-Year Cooled Fuel (Ref. 26).
- 2. SAND 80-2124 (Ref. 28).
- 3. At 503m (Ref. 15).
! 4. Regulatory Guide 1.109.
2 5.
NUREG/CR-0150 Vol. 3. (Ref. 29)
- 6. NUREG/CR-1918 (Ref. 30)
' For example, 5.44E*9 means 5.44X10 8 .
s 4
i' Table 6.2-3 Radiological Doses to the Nearest Resident from Storage Due to a CASTOR V/21 Accident at the Surry Power Station . .
l
) Whole-body i
Total Fraction .
Inhalation Dose to Cask Released .
Breathing Dose Conversion the Nearest'
, inventory (i)* Aerosolized + X/Q(3)
Nuclide -(pCI) Respirable (2) liate(4) Factors (5) Resident (sec/m3 ) (m8 /sec) (Rem /pCl) (Rem) l
- 11- 3 5.44E*9 IE-2 3.84E-5 l
2.54E-4 1.25E-4 6.63E-5 l Kr-85 7.93E*10 1E-2 3.84E-5 **
N/A 3.34E-4(6) 1.02E-5.
1-129 4.05Ei5 SE-10 3.84E-5 i
2.54E-4 5.0 (thyroid) 9.88E-12(thyroid)
Cs-134 4.35E*11 SE-10 3.84E-5 2.54E-4 4.55E-2 9.65E-8 Cs-137 9.77E+11 5E-10 3.84E-5 l '2.54E-4 3.26E-2 1.55E-7 Sr-90 6.89E*11 SE-10 3.84E-5 2.54E-4 2 4E-2 8.06E-8 i
Ru-106 1.76E+11 SE-10 3.84E-5 2.54E-4 6.18E-2 5.30E-8 i
i Total Whole Body Dose 7.69E-5 3
a Footnotes:
l 1.
i GNSI CAS10R V/21 Cask Topical SAR, 5 year cooled fuel, (Ref. 26).
- 2. SAND 80-2124 (Ref. 28).
i 3. At 2414m (Ref.15). *
- 4. Regulatory Guide 1.109. .
- 5. NUREG/CR-0150 Vol. 3 (Ref. 29)
- 6. NUREG/CR-1918 (Ref. 30)
- s r
I
bounding factor of 3) at the controlled area bouncary, due to the postulated l
accident, would be about 4 miem to the whole-cody and thyroid. If all the '
ncble gas (Kr-85) were released, as was assumed by the appitcant, the' dose at the nearest site boundary wculd only be 18 mrem to the whole body. The nearest resident would receive about .24 mrem dose to the whole-body and j
thyroid. The resultant whole-body dose to an individual at the controlled area boundary is a small fraction of the 5 rem criteria specified 'in 10 CFR .
72.68(b). These doses are also much less than the protective action guidelines established by the Environmental Protection Agency (EPA) for individuals exposed to radiation as a result of accidents: I rem to the whole-body and 5 rem to the most severely effected organ. Thus the release of effluents due to accidents at the ISFSI have a negligible impact on the population in the region around the Surry Power Station. .
Another accident associated with ISFSI operations that VEPCO addressed is a fuel assembly dropped in the worst orientation while being loaded into the i
cask. However, cask loading is conducted at the Surry Power Station under the reactor operating licenses. The environmental impact from this type of accident has already been assessed by the staff in the FESS for the Surry Power Station Unit 1 and Unit 2.
l l
1 -
- , - . ,n. , _ . . , - , _ . - - - . . . . , - - - - . - - - - - , , - , - - - ...-,---,
O
- 6.2.2 Non-Radiological Imoacts 6.2.2.1 Land Use and Terrestrial Resources Operation of this facility is not expected to detrimentally impact tn'e terrestrial environment. The only potential terrestrial interaction identi-fied is the heat radiating from the casks. This may limit the ability to maintain A grasi cover close to the. pads and with no vegetative cover, other erosion control measures m&y become necessary. Since erosion adjacent to the pads could interfere with use of the access road, the staff expects that the applicant would correct any erosional problem before erosion became a signifi-cant environmental impact.
During operation, the ISFSI site will present relatively poor' habitat for use by wildlife species.
- Construction of the ISFSI will have reduced the cover types and, thus, the ecological niches on the 6 ha (15 ac) ISFSI site. The reduced cover, plus the inhibited access due to the ISFSI perimeter fence and other human interference due to various operational and maintenance activistes in and around the ISFSI are expected to discourage wildlife use of the area, in general. It can be postulated that some species may demonstrate a preference in winter to the warmer temperatures experienced near the casks. However, birds and other species which might be attracted would be represented by few indi-viduals and no population-level effects are expected to result.
6.2.2.2 Water Use and Aquatic Resources The Surry Dry Cask ISFSI is a passive system cooled by airi there is no planned water use nor liouid releases to surface or groundwater bootes associated with oceration of the ISFSI. Surface runoff from precipitation J
J
.r., ..- . . - , -
l i
, events will be handled by the construction of swales, as necessary, to direct runoff from the ISFSI toward natural drainage patters.
6.2.2.3 Socioeconomics T .
The socioeconomic effects associated with construction of the ISFSI will be extremely small. The project is a small construction project and will involve a peak work force of 20 persons.'
s 6.2.2.4 Cask Heat Dissipation At the request of the staff (Request for Additional Information Q-1.3.8E)28, 1
the applicant ' has performed a conservative analysis of fog enhancement beyond.the site boundary due to precipitation evaporation after impingement on cask surfaces, which may be heated to 127'C (260'F)". The analysis was based on the maximum 24-hour precipitation rate measured at Norfolk, Virginia, and atmospheric dispersion conditions which are typical of rainy periods. The results of the analysis showed that the relative humidity would be increased by a few percent.
The staff has reviewed the analysis and assumptions and concurs in the results.
Since a change in relative humidity of a few percent in an atmosphere that is already near saturation would not appreciably increase fog formation, the staff agrees with the applicant's conclusion that any fog formation, due to evapora-ting water from the heated casks, would be negligible beyond the site boundary.
The temperature of 82*C (180'F) is the maximum cask surface temoerature expected for the CASTOR V/21 cask under extreme environmental conditions (Ref. 26, Section 5). The higner temperature was considered with regarc -
to fog formation.
9
7.0 SAFEGUARDS FOR SPENT FUEL Irradiated (spent) fuel removed from nuclear power reactors is highly .
radioactive and requires heavy shielding for safe handling. Theft or diversion ,
of spent power reactor fuel by subnational adversaries, with the intent of utilizing the contained special nuclear material (SNM) to fabricate nuclear explosives is not considered credible. Radiological sabotage of. spent fuel might be within the capability of potential adversaries, however, and therefore has the potential of a possible hazard to local populations.
There is no history of any confirmed deliberate acts of spent fuel sabotage directed against lice **ed nuclear power facilities within the United States which culminated in a direct or indirect danger to the public health and safety by exposure to radiation. It is likely, however, that there may be people who have the skills necessary to plan and execute an operation against the nuclear industry and that, conceivably, such people could be motivated to conduct such an operation.
This section addresses the possibilities of radiological sabotage of spent fuel while it is being moved between the reactor storage pool and the dry storage facility; and when the fuel is stationary within the dry storage facility.
o
- . . , - - - - .- _ _ _ - , . . . _ _ , . -y
l 7.1- ONSITE M VEMENT '
l The operations of loading and securing the spent fuel within the storage casks will take place within the reactor storage pool. After the casks are-removed from the pool, they will be placed on a dedicated transport vehicle and moval, under armed escort, to.the storage facility. The storage facility is ,
1ccaud less than 0.8 km (0.5 mi) from the reactor, on the same site, which is owned and controlled by the applicant.
The average weight of the loaded storage casks will range between 70 and 100 metric tons. The massive construction, primarily for safety purposes, provides a considerable measure of protection against criminal ~a cts.
The following measures, which are required as part of the appitcant's security-plan, would make a successful attack on spent fuel movements highly unlikely:
(a) Security personnel; 1
(b) Communication with onsite, armed response forces and local law enforcement authorities; (c) Unannounced schedules; and (d) Means/ procedures for halting the movement of the transport vehicle.
1 l
l l
l
l After considering the absence of any information :enfirming an identifiable '
threat; the difficulty of taking possession of and breaching a spent fuel cask, and dispersing the spent fuel; the ~ low significance of the resulting 1
consequences; and the applicable protection measures, the NRC Staff has concluded that the proposed onsite movements of spent fuel would not consti-tute a serious risk to the public health and safety, 1 7.2 FIXED SITE SAFEGUARDS To the extent that acts of sabotage initiate sequences of. events much like those initiated by accidents, the safety features designed into the spent fuel storage casks which will be used at the Surry site for mitigation of consequences of such accidents, also provide protection against potential releases of radiation which could result from sabotage. However,-the
- possibility exists that potential saboteurs may be capable of overcoming the inherent protection and engineered safety features of the casks in an attempt to create a radiological hazard.
Although there is no information available confirming the existence of any identifiable threat to commit acts of sabotage against a domestic spent fuel storage facility, protection against such acts and their possible consequences s
is dictated by prudence, guided by the known capabilities of certain potential adversary groups. For this reason, NRC regulations include reoutrements for the physical protection of scent fuel against sabotage.
e
-E7-The Commission's requirements for protective measures for spent fuel at fixec site facilities are contained in Subpart H of 10 CFR Part 72 and are detailed in 10 CFR Part 73. Principal features include requirements for guards, an armed response force, physical and procedural access controls,, detection aids, communications systems and liaison with local law enforcement agencies.
The applicant has submitted to the Commission a Physical Security Plan which i l
contains commitments to these requirements. This plan must be reviewed by the staff and determined to be satisfactory, as a condition of the appii-cant's license. The implementation of this plan will subsequently be inspected for effectiveness and operational compliance.
The foregoing observations have been analyzed, in light of all of the following considerations:
(a) The absence of any information confirming an identifiable threat to the proposed storage activity; (b) The features of the spent fuel casks and storage site design that provide inherent protection against potential radiological releases; (c) The protection features required by the regulations, as incorporated into the applicant's security plan, which provide detection of intru- '
sion and/or unauthorized activities and a capability for summoning response forces in a timely manner; and e
s - - - -
(d). The limited potential for radiological consequences as reflected in the staff's analysis of various studies of radiological sacotage events involving spent fuel.
As a result of this evaluation, the staff has determined that the sa'botage-related risks to the public health and safety related to the dry cask storage of spent fuel, as may be authorized by the issuance of the subject license, are acceptably small.
7.3
SUMMARY
The applicant's Security Plan includes measures that provide protection against acts of radiological sabotage of spent fuel while it is being moved between the reactor spent fuel pool and the ISFSI and while at the ISFSI. However, to the extent that acts of sabotage initiate secuences of events much like those initiated by accidents, any resulting offsite consequence would be essentially the same as that assessed in Section 6.2.1.3; i.e., there would be no significant impact.
l D
,vr-- w-,- w ,n--- - - - , - , , . , , . - . - . - . . _ _
4 8.0 CECOMMISSIONING 1
l A proposed decommissioning plan
- was included as a part of the application
. in accordance with 10 CFR 72.18.-* The only activities expected in decom-I missioning the Surry ISFSI are the removal of the spent fuel from the site.and decontaminating the inside surface of the casks. The casks would then be released for re-use or disposal. do residual contamina-4 tion is expected to be left behind on the concrete pads.
The costs of decommissioning the ISFSI are expected to represent a small and negligible fraction of the costs of decommissioning the Surry Power Station Units 1 and 2.
Uncer Section 51.20(b)(10) of 10 CFR Part 51, an environmental impact statement must be prepared in connection with the issuance of a license amendment authorizing decommissioning of an ISFSI. However, the proposed action here is limited to construction and operation. A request for authority to decommission, contemplated by Section 72.38 of 10 CFR Part 72, will come at a later date. New regulations revising the requirements for such applications, as well as the requirements applicable to such authorization, have recently been proposed (50 Fed. Req. 5600 (February 11,'1985)]. Among the proposed regulation changes is the deletion of the requirement in Section 51.20(b)(10) to prepare an e'n 94 onmental impact statement in connection with decommissioning of an ISFSI.
O n-,r- - - - - - . - , . - , , , - -
- 9.0
SUMMARY
AND CONCLUSIONS 9.1
SUMMARY
.0F ENVIRONMENTAL IMPACTS As discussed in Section 6.1, no significant construction impacts are anticipated. The activities will affect only about 2 percent of the land area on the Surry site. With good construction practices, the potentials for 8.
fugitive dust, erosion and noise impacts, typical of the planned construction activities, can be controlled to minimal levels. The applicant is committed to the imolementa-tion of " good construction practices" during ISFSI construction.
The only resource committed irretrievaoly is the concrete used in the three ISFSI storage pads.
i
- As discussed in Section 6.2.1, the radialogical impacts from liquid and
- gaseous effluents during normal operation of the ISFSI fall within the scope of impacts from ifcensed reactor operations which were assessed in the Surry Units 1 and 2 FES's and are controlled by the existing Technical '
Specifications for the Surry units. The primary exposure pathway associated with the ISFSI operation is direct irradiation of site workers and nearby residents. The dose comm.tment to the nearest resident from the ISFSI operation is 6E-5 mres/yr and when added to that of the Surry Power Station operations is less than 25 mres/yr as required by 10 CFR 72.67. The collective dose commitment to residents within two miles of the ISFSI is 3E-6 man-res/yr. Occupational dose of sits a kers during slab construc-tion (64 man-rem per slab) and during ISFSI operation (23 man-rem /yr) is a 6
small fraction of the total occupational cose commitment at the Surry Power Station (i.e., 2315 man-rem /yr as the annual average dose over 10 years ending in 1982). Indivicual doses are controlled to be within the limits established by 10 CFR Part 20.
The radiological impacts due to accidents at the Surry ISFSI are 4 mrem to the whole-body and thyroid of an individual located at the controlled area j boundary and .24 mrem to the nearest resident. These doses are only a small fraction of the criteria specified in 10 CFR 72.68(b) and by the EPA. An Emergency Planning Zone (EPZ) is being considered which would coincide with the ISFSI controlled area and the Surry Power Station site boundaries. If approved, there would be no need for the applicant to have an offsite emergency response
. plan for the ISFSI. -
As discussed in Section 6.2.2, no significant non-radiological impacts are expected during operation. The only environmental interface of the ISFSI is with the air surrounding the casks; the only discharge of waste to the environment is heat to the air via the passive heat dissipation system.
Climatological effects which are anticipated in the immediate vicinity of the ISFSI are judged to be insignificant to public health and safety.
9.2 BASIS FOR FINDING OF NO SIGNIFICANT IMPACT We have reviewed the proposed action relative to the requirements set forth in 10 CFR Part 51 and, based on this assessment, have determined snat issuance of a materials license under 10 CFR Part 72 authorizing storage of spent fuel
--- - - * , - - -- m-- +- - - - - - - - - - - - - - - - - - - - - - -
. j I
i i
at the Surry ISFSI will not significantly affect the cuality of the human environment. Therefore, an environmental impact statement is not warranted and, pursuant to 10 CFR Part 51.31, a Finding of No Significant Impact"(FONSI) is appropriate.
~
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10.0 REFERENCES
Regulatory Commission, submitting Application for Surry Independ'e nt Spent Fuel Storage Installation, Octete. 8, 1982. 1
- 2. U.S. Nuclear Regulatory Commission, " Final Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel," NUREG-0575, August 1979.
- 3. Ibid., p. ES-12.
- 4. U.S. Atomic Energy Commission, " Final Environmental Statement related to Operation of Surry Power Station Unit -1," Docket No. 50-280,
, May 1972.
- 5. U.S. Atomic Energy Commission, " Final Environmental Statement related to Operation of Surry Power Station Unit 2," Docket No. 50-281, June 1972.
- 6. U.S. Atomic Energy Commission, " Final Environmental Statement related to Construction of Surry Power Station Units 3 and 4," Docket Nos.
50-434 and 50-435, May 1974.
- 7. VEPCO, " Environmental Report'- Surry Power Station Dry Cask Independent Spent Fuel Storage Installation," (undated), submitted with Application (see Reference 1).
- 8. VEPCO, " Safety Analysis Report - Surry Power Station Dry Cask Independent t
Spent Fuel Storage Installation," (undated), submitted with Application (see Reference 1). -
- 9. VEPCO, Letter to NRC submitting Responses to NRC Staff's Requests for Additional Information, March 2, 1984.
- 10. Ibid., June 20, 1984.
- 11. Op. cit., June 25, 1984.
- 12. Op. cit., September 21, 1984.
- 13. Op. cit. , October 24, 1984.
i 14. Op. cit., Novemoer 30, 1984.
- 15. Op. cit., December 4, 1984.
l l 16. Oo. cit., December 10, 1984.
l . .
. . . _ . . . . _ - _ _ . __ _ _ _ _ _ . _ _ , _ _ . . _ . . _ _ _ . . _ . , . _ . _ - _ . . _ , . . _ _ _ , . . _ _ _ . _ - - , . _ _ _ _ . . ~ , _ _ _ _ . .
. 17. Virginia Electric and Power Company, Latter to NRC submitting Resoonses to NRC Staff's Requests for Acditional Information, February 8,1985.
- 18. U.S. Nuclear Regulatory Commission, Letter to VEPCO transmitting Requests
, for Additional Information, September 9, 1983.
- 19. Ibid., October 1, 1984. ~
- 20. Op. cit., November 14, 1934.
- 21. NRC Memorandum from J. Roberts to L. Rouse, (
Subject:
Summary of Meeting with VEPCO on January 31,1985), dated February 4,1985.
- 22. NRC, " Finding of No Significant Impact" and " Environmental Assessment related to Increasing the Spent Fuel Storage Capacity and the Storage of Surry Spent Fuel at the North Anna Power Station Units", Occket Nos.
50-338 and 50-339. Enclosure 1 of memorandum from James R. Miller, NRC
' Office of Nuclear Reactor Regulation, to Joseph Rutberg, NRC Office of the Executive Legal Director, dated July 2, 1984.
- 23. Letter from Michael W. Maupin (Hunton & Williams) to Sheldon J. Wolfe, j Atomic Safety and Licensing Board Panel, U.S. Nuclear Regulatory l-Commission, transmitting Settlement Agreement of April 26, 1984 between County of Louisa, Va., and VEPCO, dated May 1, 1984. .
- 24. VEPCO, " Applicant's Environmental Report - Construction Permit Stage -
Surry Power Station Units 3 and 4," Docket Nos. 50-434 and 50-435, April 1973.
- 25. U.S. Nuclear Regulatory Commission, " Socioeconomic Impacts of Nuclear Generating Stations: Surry Case Study," NUREG/CR-2749 Vol. 11, July 1982.
- 26. General Nuclear Systems, Inc., " Topical Safety Analysis Report for the CASTOR V/21 Cask Independent Spent Fuel Storage Installation (Ory Storage)," January 22, 1985.
- 27. U.S. NRC, " Occupational Radiation Exposure at Commercial Nuclear Power Reactors - 1982 Annual Report," NUREG-0713, Vol. 4, Decemoer 1983.
- 28. E.L. Wilsons, " Transportation Accident Scenarios for Commercial Spent Fuel," SANO-80-2124, Sandia National Laboratories, Albuquerque, NM, February 1981.
,e,- - - ,- - - - , un---,_y-+--ww-
- 29. D. E. Dunning, Jr. et al., " Estimates of Internal Dose Equivalent to 22 Target Organs for Radionuclides Occurring in Routine Releases frem Nuclear Fuel-Cycle Facilities, Vol. III", NUREG/CR-0150, Vol. 3, t
prepared for the NRC by Oak Ridge National Laboratory, Oak Ridge, Tennessee, October 1981. -
- 30. D. C. Kocher, " Dose-Rate Conversion Facters for External Exposure to Photons and Electrons", NUREG/CR-1918, prepared for the NRC by Oak Ridge National Laboratory, Oak Ridge, Tennessee, August 1981.
1 I
l l
-6 6 -
11.0 LIST OF PREPARERS Name and Title Resoonsibility C. Billups, Aquatic Scientist Project Leader, Water Use'and Aquatic Resources W. Swick, Secretary- Word Processing, Coordination D. Cleary, Section Leader Socioeconomics, Demography G. LaRoche, Sr. Land Use Analyst Land Use and Terrestrial Resourtes E. Markee, Sr. Meteorologist Meteorology, Climatology i
R. Samworth, Section Leader Air Quality, Noise.
O. Smith, Plant Protection Analyst Safeguards for Spent Fuel F. Sturz, Health Physicist Radiological, Alternatives e
4 m +
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O ENCLOSURE D e
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(
CASTOR-V/21 PWR SPENT FUEL STORAGE CASK PERFORMANCE TESTING AND ANALYSES (EXECUTIVE
SUMMARY
)
J. M. Creer February 1986 a
Work supported by the Virginia Power Company and the U.S. Department of Energy Pacific Northwest Laboratory Richland. Washington 99352 e
G
.- Enclosure D
EXECUTIVE SUM 4ARY 4
This report documents a heat transfer and shielding performance test conducted on a G:sellschaft fur, Nuklear Service (GNS) CASTOR-V/21 pressurized water reactor (PWR) spent fuel storage cask. Performance testing was conducted, under a cooperative cgreement program between the Virginia Power Company and the U. S. Department of Energy (DOE), by the Pacific Northwest Laboratory operated by Battelle Memorial Institute, the Idaho National Engineering Laboratory (INEL) operated by EG&G Idaho,
~
Inc., and Virginia Power. Testing was performed at INEL's Test Area North (TAN)
, cask testing facility and consisted of pretest preparations, performance testing, "
and post-test activities. Protest preparations included conducting cask handling dry runs and characterizing PWR spent fuel assemblies from Virginia Power's Surry Nuclear Power Plant. The performance test matrix included 5 runs consisting of two cask orientations and three backfill environments. Post-test activities included crud collection and video and photographic scans of selected fuel assemblies.
The CASTOR-V/21 PWR spent fuel storage cask consists of a nodular cast iron body.
The ' cast iron / graphite- material exhibits good strength and ductility and .provides I
offective gamma shielding. The overall dimensions of the cask are 4.9 m (16 ft)
Icng and 2.4 m (8 ft) in diameter and tne cask weighs approximately 100 tons loaded with unconsolidated PWR spent fuel. Two concentric rows of polyethylene rods are incorporated in the wall of the cask to provide neutron shielding. The external surface consists of heat transfer fins which are circumferentially oriented around 1 the cask surface. The fuel basket within the cask is configured to hold 21 PWR spent fuel assemblies and is constructed of stainless steel and borated stainless
, steel for criticality control. The Surry spent fuel assemblies used during testing are of a standard Westinghouse 15 X 15 rod design. The cask is closed with two j lids having both rubber and metallic 0-rings to seal the cask cavity from the environ-ment.
l j Dry runs of cask handling and fuel loading were performed prior to Surry' fuel being j leaded in the cask. The objectives of the dry runs were to gain operational experi-cnce and to finalize handling and test procedures. Each dry run was conducted suc-l c0ssfully without encountering unusual problems or requiring significant modifi-cations to the cask or handling equipment.
t t
The Surry PWR spent fuel assemblies were characterized using in-basin ultrasonic cxaminations and video. After testing, the fue.1 assemblies.were videoed and photo-graphed, and smear samples collected. The results of these examinations revealed no indication of any failed fuel before or after the CASTOR-V/21 cask performance test.
Based on pretest ORIGEN2 predictions, fuel assembly decay heat generation rates totaled approximately 28 kW at the start of testing and 27 kW at the end of testing (Table 5-1). Thirteen of the twenty-one fuel assemblies had decay heat rates near 1 kW and the remaining eight assemblies had decay heat rates of approximately 1.8 kW at the start of the month long test. The fuel assemblies were loaded in the cask with the hot assemblies in the outer regions of each quadrant as seen in Figure S-1. Pretest heat transfer predictions using the HYDRA computer program indicated that peak clad temperatures in nitrogen and helium would be below or near 380*C.
The selected fuel loading pattern was predicted to create a relatively flat radial temperature profile across the basket during testing.
Table S-1 weh SURRY PWR SPENT FUEL CHARACTERISTICS Burnup Cooling Time Enrichment Sept. 1985 Pred.
Assembly (GWd/MTU) (Months) (wt%) Decay Heat (kW)
Start End V04, V08. V12. V24 31.1 46 2.91 1.00 0.98 V05 31. 5 46 2.91 1.02 0.99 T03, T07. T08 T09 T11, T12. T13. T16 35.7 46 3.11 1.11 1.09 V11, V13. Vl4, V15 29.8 26 2.91 1.79 1.72
, V01, V09, V25. V27 30.2 26 2.91 1.83 1.75 l
l Total 28.4 27.5 i
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.+* w 7 x u .....
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., , iso.
Figure S-1. CASTOR-V/21 Cask Cross Section Figure S-2 shows the predicted axial decay heat profile assumed for each of the Surry assemblies. Measured axial power profiles for the Surry assemblies were not available as data for predicting axial decay heat profiles of the assemblies used in the CASTOR-V/21 cask performance test. Axial gamma radiation scans previously cbtained on Turkey Point reactor fuel assemblies were used to produce an assembly axial burnup distribution. The Turkey Point and Surry reactors and spent fuel are W:stinghouse PWR designs and essentially the same. ORIGEN2 was used with the assembl~y axial burnup distribution and the Surry operating history to determine the axial decay heat profile shown in Figure S-2. The dips in the decay heat profile are due to grid spacers. Axial decay heat profiles are important because they strongly affect the shape of measured axial fuel temperature profiles, especially in vacuum and in a horizontal orientation where convection heat transfer is minimized.
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o 3 o o.2 o.4 o.6 o.'S 1.o 1.2 1.4 1.6 n.i.tiv. v.iu.
Figure S-2. Measured Gamma and Predicted Decay Heat Axia,1 Profiles The outer surface of the cask was instrumented with 35 thermocouples (TC). 70 gamma dese rate sensors, and 70 neutron dose rate sensors. Sixty TCs contained in ten lances (tubes) were inserted through the cask,1)d into fuel assembly guide tubes or void basket spaces. Of the ten TC lances, eight with six TCs each were inserted into fuel assembly guide tubes, and two with six TCs each were positioned in basket void spaces as shown in Figure S-3.
se l
The cask test' matrix included assessments of performance with a full load of fuel (21 assemblies), vertical and horizontal orientations, and vacuum, nitrogen, and helium backfill environments. The test matrix and corresponding measured peak guide r
tube temperatures and estimated peak clad temperatures are presented in Table S-2.
l
, Pcak clad temperatures were estimated by using calculated guide tube-to-hot rod temperature differences from the HYDRA computer program.
4- -
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. . l
- o. 4 n.wy a.iae Orientation T m Penern Mark -o
+'
+ + 4 t'++2
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+
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+4 9 Trunnion
+ 1 e3 27o*-
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+ +
y-iso- .
Figure S-3. TC Lance Locations Table S-2 -
CASTOR-V/21 Cask Test Matrix and Peak Temperatures Meas. Est.
, Cask Heat Ambient Guide Tube Peak Clad j Run # Loading Orientation Backfill Load. kW Temo, 'C Temp, *C Temo, *C 1 Full Vert. He 28.4 27 347 352 2 Full Vert. N 28.4 24 358 368 3 Full Vert. Vac 28.4 25 414 424 4 Full Horiz. He 28.4 24 360 365 5 Full Horiz. N 28.4 24 395 405 Table S-2 indicates that in a vertical orientation with nitrogen and' helium back-fills, peak clad temperatures were less than the 380*C allowable. This was also th3 case for the horizontal helium run. The vertical vacuum and horizonal nitrogen runs resulted in peak clad temperatures over 380*C, but the temperatures were not G
I
- 5- ,
6 O
cxcessively high (<425'C). None of the peak temperatures occurred in the high decay heat (1.8 kW) outer assemblies (Figure S-1). In general, the cask heat transfer performance was concluded to be exceptionally good because the peak temper-ature in helium, when the cask was dissipating approximately 28 kW. was less than ,
that specified for the cask operating limit of 21 kW in the cask topical report prepared by GNS.
Axial and radial temperature profiles for the five test runs are shown in Figures 5-4 and S-5. Attention should only be given to data points because their corre-sponding lines are provided for clarity and in no way are intended to represent actual profiles. The axial profiles are for the hot center assembly and the radial profiles are for the axial location at which the temperatures peak in a vacuum and in a horizontal orientation. The axial profiles vividly show the effects of canvection in nitrogen and helium In a vertical orientation where peak tempera-
- tures are skewed towards the top of the cask. Both the nitrogen and helium axial profiles show about the same degree of convection. This is surprising because the density and viscosity of helium are not condusive to convection i.e., buoyancy forces in helium are substantia,Ily less than in nitrogen. Note that the peak tcmperature in the vertical helium ,,run is not significantly lower than the peak tcmperature in nitrogen. This is an indication that convection in nitrogen nearly --
makes up for the relatively high thermal conductivity (four times that of nitrogen) cf helium. The indication of significant convection in the CASTOR-V/21 cask implies
~
that the basket is designed to support convection which is obvious from the rela-tively "open" design indicated in Figures S-1 and S-3.
Symmetry over the length of the fuel assemblies in the vertical vacuum and hori-zental axial temperature profiles indicates the absence of convection in these runs. These profiles are similar to the axial gamma and decay heat profiles previ-cusly presented in Figure S-2. The lack of axial convection in vacuum and hori-zental runs is reasonable because significant density gradients cannot develop in a vacuum or an orientation with low axial gravitational forces.
Radial temperature profiles for the five test runs shown in Figure S-5 indicate relatively flat profiles across the basket, and steep gradients from the basket to the cask inner wall. Steep gradients across the basket-to-inner wall gap indi-cate the gap is important to the heat transfer design of the cask. Temperature G 9
l
. 1 e
500 Center Assembly A1 25'C Ambient a <: Surface OVertices vecuum 400 ""$ d O Vertical Nitrogen
) j
~' O Vertical Helium 1 g 1 e Horisontal Nitrogen 300 -
. '. 3 Horizontal Helium g
! s
?
<}
200 -
I j
100 . .;..
1h a h 0
O 100 200 300 400 500
. T. *C Figure S-4 Center Assembly Axial Temperature Profiles 500 ..
400 -
300 -
Zs235 em 25'C AmWnt T. 'C 200 -
O V. tical vecuum O Vertical Nitrogen O Vertical Heleum 100 - e Horisontal Nitrogen a Horisontal Helium I I
! O - A7 A4 A1 a4 87 ,
i I
M?REIEEEEIHERE t t I t i t 120 a0 40 0 40 80 120
- n.46u..
Figure S-5. Radial Temperature Profiles .
l . .
gradients from assemblies A1 and A4 to the gas adjacent to their fuel tubes are seen to be greater in vertical helium and nitrogen runs than vertical vacuum or horizontal runs. This indicates that significant gas flow occured outside assem-bly fuel tubes in a vertical orientation (convection heat transfer is significant) cnd that in vacuum and horizontal runs, axial gas flow was relatively, low (heat transfer by ra,diation and/or conduction was dominant).
D:se rates on the top, side, and bottom of the cask are shown in Figures S-6.
S-7, and S-8. Only data points should be considered because the corresponding lines are provided for clarity and do not represent actual profiles. The radiation source strength was higher for the test fuel than fuel considered in the cask topical report prepared by GNS. A peak gamma dose rate of 43 mrem /hr was measured en the top of the primary lid at 45 degrees (the outer lid was not used during testing). The total dose rate (gamma plus neutrons) was approximately 85 mrem /hr at the center of the primary lid. When the secondary lid (90 mm, 3.5 in. thick) is used on the cask during normal operation, these dose rates should be reduced balow 20 mrem /hr. ,
100 Pnm.cy Ud. 45*
21 Ase.mtesse
',N c o..
o n.
o T.w so -
a b .
j e
II i de I A\
f 80 I
120 n.eu Figure S-6. Gamma and Neutron Dose Rates on Cask Primary Lid ,
8,-. -
- o *
+ .
500 v g 400 - 't j? '
side. 45
- 21 Assemblies
. 4 O Gemme 300 -
E
- a p Og O Neutron
". O Total 200 -i b 0b 0 ()
100 -
V;
. 11 4
0 50 100 150 200 Dose Rete. mR/hr Figure S-7. Gama and Neut~ron D0se Rates on _ Cask Side l ,, ,
2
~ s 4s-
$ 21 Assemelees
. o a..
3 < O Neutrea a ,, ,
Ow c
~ ~
- " N
, , t ,
,0 40 80 120 n
Figure S-8. Ganna and Neutron D0se Rates On. Cask Bottom
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, , -- - ---.---- n. - - , , . , , . , , - , , , , - - . _ , - . . ,n-, ,, ,,-,,e, .--6--,------n- . - -
The total dose rates along most of the cask side are less than 50 mrem /hr (Figure S-7). There are localized peaks in the gamma and neutron dose rates of up to 140 mrem /hr and 21 mrem /hr, respectively. The neutron dose rate peaks are relatively low, but the gamma peaks are substantial. The localized peaks occurred at locations cdjacent to fuel assembly end fittings. However, a minor refinementi involving rcplacement of neutron shielding material with gamma shielding material in locations
- ccrresponding to the peaks should reduce the gamma peaks if desirable.
I Dose rates on the bottom of the cask (Figure S-8) peak at the center (05 mrem /hr t:tal), but are relatively low and uniform oi, the remainder of the surface. These relatively low dose rates are not of concern when the cask is oriented horizontally.
The overall shielding performance of the CASTOR-V/21 cask was good and met the intended design goal of <200 mrem /hr even though the test fuel had a higher source strength than fuel considered in the GNS topical report. With a very minor refine-ment, total dose rates can easily be reduced to less than 75 mrem /hr if lower dase rates are desirable. ,
After' cask testing was completed, selected fuel assemblies were videoed, photo-graphed, and smear samples taken. No unusual anomalies were observed on any of th3 selected assemblies; however, eight indications of cracks were observed in tha CASTOR-V/21 basket. The observed indications had no affect on the ability to rcmove or reinsert fuel assemblies in the basket. Figure S-9 identifies the lo- .
- cations of the crack indications.
(
Tha test basket was designed to have a relatively tight fit to permit the TC lances to pass through the primary lid and into fuel assembly guide tubes. GNS performed a thermal stress analysis of the basket and concluded that the basket expanded and came in contact with the inner wall of the cask (Gap 2). Also, the fuel tubes l ccntaining the outer assemblies adjacent to the diagonals came in contact with the basket barrel (Gap 1). An examination of the measured temperatures, HYDRA i
l temperature predictions, and associated differential thermal expansion between the basket and cask body substantiates the GNS analysis. The cracks in the basket welds did not result in adverse safety implications because the cracked welds S
. i. -
10 -
---,---------,,--,w,-w--,---,w-- , ,,,-,--~- - - - - , ,
~ .
Gap 1 a
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. ~ ., .-
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Figure S-9. ChSTOR-V/21BasketCrackIndications It is highly unlikely that any member could hava cinimal structural requirements. lt robl em.
complctely fracture free from the basket assembly and present a critica i y p i
The cask performance test demonstrated that the CASTOR-V/21 cask It was concluded that the heat transfer per-fact:rily handled and loaded dry. Peak clad temperatures with helium fctmance of the cask was exceptionally good. 380'C with a total and nitrogen backfills in a vertical orientation were less than The shielding performance of the cask met design expec-cask heat load of 28 kW. /hr can be esta-tations (<_200 mrem /hr), and cask surface dose rates of <75 mrem blished with minor refinements. if desired.
. e e- - ---
CASTOR V/21 AFTER FABRICATION AT KRAFTWERK UNION PLANT MULHEIM FEDERAL REPUBLIC OF GERMANY 9
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