ML20140E046

Responds to 841207 Request for Info Re Issuance of Amends to Licenses DPR-31 & DPR-41,permitting Increase in Licensed Storage Capacity to 1,404 Spent Fuel Assemblies.Description of Fuel Reprocessing History Provided

ML20140E046
Person / Time
Site:	Turkey Point
Issue date:	01/02/1985
From:	Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	Fascell D HOUSE OF REP.
Shared Package
ML20140E060	List: ML20140E046 ML20140E068
References
NUDOCS 8501100068
Download: ML20140E046 (4)

Text

{{#Wiki_filter:r .I-1 l Q tw [ g UNITED STATES NUCLEAR REGULATORY COMMISSION j o 7. WASHINGTON, D. C. 20666 JAN 0 21965 % 2 5'() (54,251 The Honorable Dante B. Fascell United States House of Representatives Washington, D.C. 20515

Dear Representative Fascell:

I am pleased to respond to your December 7,1984, letter requesting information relating to the issuance of amendments to the Turkey Point Plant Unit Nos. 3 and 4 Operating Licenses which pemit the increase in the licensed storage capacity from 621 spent fuel assemblies to 1404 spent fuel assemblies for each of the two Turkey Point spent fuel pools. TP.e Conunission issued Amendment No.111 to Facility Operating License No. DPR-31 and Amendment No.105 to Facility Operating License No. DPR-41 l on November 21, 1984, for the Turkey Point Plant Unit Nos. 3 and 4, respectively. An Environmental Assessment related to this action was issued on November 14, 1984. The Notice of Issuance of Environmental Assessment and Finding'of No Significant Impact was published in the Federal Register on November 16,1984 (49 FR 45514). - I have enclosed. copies of the amendment issuance and the Environmental Assessment. The amendment issuance includes the Safety Evaluation and Notice of Issuance and Final No Significant Hazards Consideration. The request for these amendments was individually noticed on June 7,1984 (49 FR 23715) followed by a monthly notice on July 7, 1984-(49 FR 29925). Consnents, request for a hearing and petition for leave to intervene were l initiated on July 9.1984, by the Center for Nuclear Responsibility and Ms. Joette Lorion. The conenents and concerns relevant to these amendments are addressed in the Safety Evaluation. Under NRC regulations, the Commission may issue and make an amendment l insnediately effective, notwithstanding a request for a hearing, in advance l of holding the hearing where, as here, it has detemined that the amendment i involves no significant hazards consideration. Such issuance is also l consistent with Section 132 of the Nuclear Waste Policy Act of 1982 which l requires the Consnission to encourage and expedite the effective use of l available storage at civilian reactor sites. t I will briefly describe fuel reprocessing history, the need for increased storage capacity and alternatives considered in assessing the acceptability.of increasing the storage capacity for spent fuel assemblies at Turkey Point and other nuclear power reactor sites. Currently, spent fuel is not being reprocessed on a commercial basis l in the United States. The Nuclear Fuel Services (NFS) plant at West Valley, New York, was shut down in 1972 for alterations and expansion; in B501100068 850102 PDR ADOCK 05000250 U _. PDR i' l j

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. $ 45 i Mr. Fascell 2 t ' September 1976, NFS infomed the Connission that it was withdrawing from the nuclear fuel reprocessing business. The Allied General Nuclear Services-(AGNS) proposed plant-in Barnwell, South Carolina, is not licensed to operate. On April 17, 1977, President Carter issued a policy statement on comercial reprocessing of spent nuclear fuel which effectively eliminated reprocessing as part of the relatively near term nuclear fuel cycle. The General Electric Company (GE) Morris Operation (formerly Midwest Recovery Plant) in Morris, Illinois, is in a decomissioned condition. Although no plants are licensed for reprocessing fuel, the storage pools at West Valley are not full, but the licensee

• is presently not accepting any additional spent fuel for storage even from those power generating facilities that had contractual arrangements with West Valley.** On May 4, 1982,' the license held by GE for spent fuel storage activities at its

. Morris operation was-renewed for another 20 years; however, GE is comitted to accept only limited quantities of additional spent fuel for storage at this facility from Cooper and San Onofre Unit 1. When originally licensed, the spent fuel pools for each of the Turkey Point Units had the capacity to hold.217 fuel assemblies. This represented the requirement for one refueling of each unit with reserve capacity to receive a full core. At that time it was expected that the spent fuel would be removed from the site shortly after it was discharged to the spent fuel pools. o The Turkey Point Licenses were amended to allow modifying the fuel pool racks to accommodate 621 fuel assemblies which would be adequate to retain the' reserve capacity for full core unloading (157 assemblies) until about 1986. Since this date is earlier than the date a Federal depository is expected to be 'available for spent fuel [1998 - Nuclear Waste Policy Act of 1982, Section 302(a)(5)] the rack modifications were essential to allow continued

• operation beyond 1986..These current amendments allow expanding

. the storage capacity of each unit to acconnodate 1404 assemblies which would. extend.the full core discharge capability for each generating unit to the

year 2005 for Unit 4 and the year 2006 for Unit 3.

I Comercial reprocessing of spent fuel has not developed as had been originally' anticipated. In 1975 the Nuclear Regulatory Comission directed the staff to prepare a Generic Environmental Impact Statement (GEIS) on spent fuel storage. The Comission directed the staff to analyze alternatives for.the handling and storage of. spent light water power reactor fuel with particular emphasis on developing long range policy. The Statement was to consider alternative methods of spent fuel storage as well l as nuclear power plant shutdown, i - A final Generic Environmental Impact Statement on Handling and Storage l of Spent Light Water Power Reactor Fuel (NUREG-0575), Volumes 13 (the FGEIS) was issued by the NRC in August 1979. The finding of the FGEIS is i j

• The current' licensee is New York Energy Research a'nd Development Authority.

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• • In fact, spent fuel is being removed from NFS and returned to various.

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m 3 Mr.' Fascell 4 that the environmental impact costs of interim storage are essentially negligible, regardless of where such spent fuel is stored. A comparison 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 i filled, the cost of replacing nuclear stations before the end of their nomal lifetime makes this alternative uneconomical. 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. One spent fuel' pool storage alternative considered in detail in the FGEIS is the expansion of onsite fuel storage capacity by modification of the existing spent fuel pools. Applications for approximately 108 spent fuel pool capacity increases have been received and over 100 have been approved.. The remaining ones are still undo review. The finding in each case has been that the environmental impact of such increased storage capacity is negligible. However, since there are variations in storage . designs and limitations caused by the spent fuel already stored in some of the pools, the FGEIS recomends that licensing reviews be done on a case-by-case basis to resolve plant-specific concerns. The enclosed Safety Evaluation and Environmental Assessment provide details and resolution of } the plant-specific concerns related to the Turkey Point site. f .Your constituents may be concerned with public exposure resulting from L the increased storage capacity approved by the Comission. The staff has completed an analysis of radiation exposure experience, based on estimated l source tenns and. assessment of public doses resulting from 38 prior spent fuel pool modifications at 37 plants. Estimated -doses to a hypothetical maximally exposed individual at the boundary of a plant site, during such modifications, have fallen within a l range from 0.00004 to 0.1 millirem per year, with an average dose of 0.02 ~ millirem per year. Similarly, estimated total doses to the population within a 50-mile radius of these plants have fallen within a range from 0.0001 to 0.1 person-rem per year, with an average population dose of 0.006 ~ l person-rem per year. Doses at these levels are essentially unmeasurable. l Based on the manner in which the Florida Power and Light will perform l the modifications; Ltheir radiation protection /as low as reasonably achievable (ALARA) program; the radiation protection measures proposed for the modification tasks including radiation, contamination, and airborne radioactivity monitoring; and relevant experience from other operating i reactors that have performed similar spent fuel pool modifications, the staff concluded that adequate radiation protection measures have been taken to assure worker protection and the Turkey Point spent fuel pool modifications can be perfomed in a manner that will ensure that doses to l workers and the general public will be ALARA. l l l l l L ~ c m -.- ~

Mr. Fascell Based on this review of :.istorical data (" Natural Radiation Exposure in the United States," Donald T. Oakley, U.S. Environmental Protection Agency, Office of-Radiation Programs (ORP/SID 72-1, June 1972)) relating to the storage of spent fuel, we concluded that for the spent fuel pool expansions at Turkey Point the additional dose to the total body that might be received by an individual at the. site boundary and by the population within a 50-mile radius, respectively, would be less than or equal to 0.1 millirem and 0.1 person-rem per year, respectively. These doses are very small compared to annual exposure to natural background radiation in the United States which varies from about 70 millirems per year to about 300 millirems per year depending on geographical location. I trust you find this responsive to your request and of assistance in assuring your. constituents that the Commission's decision was based on sound technical judgement by the staff. This decision is consistent with the Commission's policy of ensuring that operating facilities, such as Turkey Point, achieve and maintain adequate levels of protection of public health and safety. If we can be of further assistance, please do not hesitate to contact us. Sincerely, WZ53@ t{nitanq u,4 William J. Dircks Executive Director for Operations

Enclosures:

As stated e 0 e w 4 " _ 2.l

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,k [pagg *o, UNITED STATES p .[ 7. NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 206SS a %,],,, # November 21,1984 Docket Nos. 50-250 and 50-251 Mr. J. W. Williams, Jr., Vice President Nuclear Energy Department Florida Power and Light Company Post Office Box 14000 Juno Beach, Florida 33408

Dear Mr. Williams:

The Cosmission has issued the enclosed Amendment No.111 to Facility Operating License No. DPR-31 and Amendment No. 105 to Facility Operating License No. DPR-41 for the Turkey Point Plant Units Nos. 3 and 4 respectively. The amendments consist of changes to the Technical Specifications in response to your application transmitted by letter dated March 14, 1983, as supplemented. + These amendments allow spent fuel pool storage capacity expansion from 621 to 1404 spaces for each spent fuel pool. The expansion is to be achieved by reracking each spent fuel pool with two discrete regions within each pool. Region I is for storage of new fuel with an enrichment equal to or 'ess than 4.5% U-235. Region II is for storage of irradiate fuel meeting the burnup requirements defined in the Technical Specifications. The request for these amendments was individually noticed on June 7,1984 (49 FR 23715) followed by a monthly notice on July 7, 1984 (49 FR 29925). Conments, request for a hearing and petition for leave to intervene were initiated on July 9, 1984, by the Center for Nuclear Responsibility and Ms. Joette Lorion. The comments and concerns relevant to these amendments are addressed in the enclosed Safety Evaluation. The Safety Evaluation also j includes a final determination of No Significant Hazards Consideration. under NRC regulations, the Connission may issue and make an amendment l immediately effective, notwithstanding a request for a hearing, in advance I of holding the hearing where, as here, it has determined that the amerdment l involves no significant hazards consideration. Such issuance is also consistent with Section 132 of the Nuclear Waste Policy Act of 1982 which requires the Comnission to encourage and expedite the effective use of available storage at civilian reactor sites. Copies of the Safety Evaluation and Notice of Issuance and Final Determination of No Significant Hazards Consideration are enclosed. l The Environmental Assessment related to this action was transmitted to you j on November 14, 1984 The hatice of Issuance of Environmental Assessment ~ MM j O g,. 3

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Y Mr. J. W. Williams -2 November 21, 1984 and Finding of No Significant Impact was published in the Federal Register on November 16, 1984 (49 FR 45514). Sincerely, I ll Daniel G. Doiia d. Jr., Project Manager Operating Reactors Branch #1 Division of Licensing

Enclosures:

1. Amendment No. Illto DPR-31 2. Amendment No. 105to DPR-41 3. Safety Evaluation 4. Notice cc: w/ enclosures See next page g l i j l t i i t l

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J. W. Williams, Jr. . Turkey Point Plants Florida Power and Light Company Units 3 and 4 I cc: Harold F. Reis, Equire Administrator-Newman and Holtziner P.C. Department of Environmental 1615 L Street, N.W. Regulation Washington, DC 10036 Power Plant Siting Section State of Florida 2600 Blair Stone Road Bureau of Intergovernmental Relations Tallahassee, Florida 32301 660 Apalachee Parkway Tallahassee, Florida 33130 James P. O'Reilly Regional Administrator, Region II Nonnan A. Coll, Esquire U.S Nuclear Regulatory Conunission . Steel. Hector and Davis Suite 2900 4000 Southeast Financial 101 Marietta Street Center Atlanta, GA 30303 Miami, F1orida 33131-2398 Martin H. Hodder, Esquire 1131 N.E. 86th Street Mr. Ken N. Harris, Vice President Miami, Florida 33138 Turkey Point Nuclear Plant Florida Power and Light Company Joette Lorion P.O. Box 013100 7269 SW 54 Avenue Miami, Florida 33101 Miami, Florida 33143 Mr. M. R. Stierheim Mr. Chris J. Baker, Plant Manager County Manager of Metropolitan Turkey Point Nuclear Plant Dade County Florida Power and Light Company Miami, F1orida 33130 P.O. Box 013100 Miami, Florida 33101 Resident inspector . Turkey Point Nuclear Generating Station Attorney General U.S. Nuclear Regulatory Commission Department of Legal Affairs Post Office Box 57-1185 The Capitol Miami, Florida 33257-1185 Tallahassee, Florida 32304 Regional Radiation Representative Mr. Ulray Clark, Administrator EPA Region IV Radiological Health Services 345 Courtland Street, N.W. Department of Health and Atlanta, GA 30308 ~ 1323 Winewood Blvd. Rehabilitative Services Mr. Jack Shreve Tallahassee, Florida 32301. Office of the Public Counsel Room 4, Holland Building Tallahassee, Florida 32304 v, n

[* **'* 1 UNITED STATES es NUCLEAR REGULATORY COMMISSION wAsHmGTON. D. C. 20SSS FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-250 TURKEY POINT PLANT UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.111 License No. DPR-31 1. The Nuclear Regulatory Comission (the Comunission) has found that: A. The application for amendment by Florida Power and Light Company (the licensee) dated March 14, 1984 as supplemented complies with the standards and re as amended (the Act)quirements of the Atomic Energy Act of 1954, and the Comunission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comunission; C. . There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be , conducted in compliance with the Comunission's ragulations; D. The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and The issua'ce of this amendment is in accordance with 10 CFR Part E. n 51 of the Consnission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraon 3.8 of Facility Operating License No. OPR-31 is hereby amenced to read as follows: 1 i b Or A - [ n ~,

, (B) Technical Specifications The Technical Specifications contained in Appendix A and B, as revised through knendment No.111, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of the date of issuance and shall be implemented within 60 days of issuance. FOR THE NUQLEAR REGULATORY C0pmISSION hiUq db,ml5 Steven A. , Chief Operating Reactors Branch #1 Division of Licensing

Attachment:

Changes to the Technical Specifications Cate of Issuance: November 21, 1984 e e e l l i l {- 7 y; ' w? '"

e e sh StC 'o, UNITED STATES ' i, NUCLEAR REGULATORY COMMISSION {. ,I wAsemocrow, o. c. 2osas %.S... FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-251 TURKEY PCINT PLANT UNIT NO. 4 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.105 License No. OPR-41 1. The Nuclear Regulatory Commission (the Cosmission) has found that: A. The application for amendment by Florida Power and Light Company (the licensee) dated March 14, 1984 as supplemented complies with the standards and re as amended (the Act)quirements of the Atomic Energy Act of 1954, and the %ission's rules and regulations set forth in 10 CFR Chapter I; 8. The facility will operate in conformity with the application, the provisions of the Act, and the rules and. regulations of the Cosmission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be ,. conducted in compliance with the Cosmission's regulations; D. The issuance of this amendment will not be inimical to the cosmon defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Cosmission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-41 is hereby amended to read as follows: ,p -.. _, _.. _ _ _ _ _.. _ _ _ _ -, -. _ - _. _ ~ _

-2 9 (B) Technical Specifications The Technical Specifications contained in Appendix A and B, as revised through Amendment No.105, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical j Specifications. 3. This license amendment is effective inunediately and shall be implemented within 60 days of issuance. FOR THE NUCLEAR REGULATORY COPetISSION hl0 ' NWm. ' -[ ffg Steven A.'Vdrga, Chief Operating Reactors Branch #1 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: November 21,.1984 A k e 4 g_ '4-(

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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. Ill FACILITY OPERATING LICENSE NO. DPR-31 AMENDMENT NO.105 FACILITY OPERATING LICENSE NO. DPR-41 DOCKET NO. 50-250 AND 50-251 Revise Appendix A as follows: Remove Pages Insert Pages 11 11 iv iv v v 3.12-1 3.12-1 B3.12-1 B3.12-1 1 3.17-1 3.17-1 Table 3.17-1 'Able 3.17-1 83.17-1 B3.17-1 Table 4.1-2 (Sheet 2 of 3) Table 4.1-2(Sheet 2of3) 5.4-1 5.4-1 O d 6 4 s.*L**, . ? *s 4* J-w

l TABLE OF CONTENTS (Continued) EEE!!!n Tji{g Pggg 3.4 Engineering Safety Features 3.4-1 Safety injection and RHR Systems 3.4-1 1 Emergency Containment Cooling Systems 3.4-3 Emergency Containment Filtering System 3.4 4 Component Cooling System 3.4-4a intake Cooling Water System 3.4 -5 Post Accident Containment Vent System 3.4-6 Control Room Ventilation 3.4-6 3.3 Instrumentation 3.5-1 3.6 Chemical and Volume Control 5ystem 3.6-1 3.7 Electrical Systems 3.7-1 3.8 Steam Power Conversion Systems 3.5-1 3.9 Radioactive Materials Release 1.9-1 Liquid Wastes 3.9-1 Gaseous Wastes 1.9-3 Containerized Wastes 3.9-5 3.10 Refueling 3.10-1 3.!! Miscellaneous Radioactive Materials Sources 3.11-1 3.12 Cask Handling 3.12-1 3.13 Snubbers 3.13-1 3.14 Fire Protection Systems 3.14-1 3.13 Overpressure Mitigating System 3.15-1 3.16 Reactor Coolant System Pressure isolation Valves 3.16-1 3.17 Spent Fue! Storage 3.17-1 l 4.0 SURVEILLANCE REQUIREAENTS 4.1 1 4.1 Operational Safety Review 4.1 1 4.2 Reactor Coolant System in Service Inspection 4.2-1 4.3 Reactor Coolant System Integrity 4.3-1 4.4 Containment Tests 4.4 -1 Integrated Leakage Rate Test - Post Operational 4.4-1 Local Penetration Tests 4.4-2 Report of Test Results 4.4-2 Isolation Valves 4.4-3 Residual Heat Removal System - 4.4-3 Tendon Surveillance 4.4 4 End Anchorage Concrete Surveillance 4.4-6 Liner Survelliance 4.4 7 4.3 Safety injection 4.5-1 4.6 Emergency Containment Cooling Systems 4.6-1 4.7 Emergency Containment Filtering and Post Accident Containment Vent Systems 4.7-1 4.8 Emergency Power System Periodic Tests 4.5-1 4.9 Main 5 team isolation Valves 4.9-1 11 Amendment Nos. and mr

i', i TABLE OF CONTENT 5 (Continued) Section Title

_Pgge, B3.5 Bases for Limiting Conditions for Operation, Instru nentation B3.5-1 B3.6 Bases for Limiting Conditions for Operation, Chemical and Volume Control System B3.6-1 53.7 Bases for Limiting Conditions for Operation, Electrica! Systems 33.7-1 53.8 Bases for Limiting Conditions for Operation, Steam and Power Conversion Systems B3.8-1 B3.9 Bases for Limiting Conditions for Operation, Radioactive hterials Release 53.9-1 B3.10 Bases for Limiting Conditions for Operation, Refueling B3.10-1 B3.!!

Bases for Limiting Conditions for Operation, Miscellaneous Radioactive hterial Sources B3.11-1 B3.12 Bases for Limiting Conditions for Operation, Cask Hand!!ng B3.12-1 . B3.13 Bases for Limiting Conditions for Operation, Snubbers B3.13-i B3.14 Bases for Fire Protection System B3.14-1 B3.15 Bases for Limiting Conditions of Operation, Overpressure Mitigating System B3.15-1 B3.17 Bases for Limiting Conditions for Operation, B3.17-1 Spent Fuel Storage B4.1 Bases for Operational Safety Review B4.1-1 B4.2 Bases for Reactor Coolant System In-Service Inspection B4.2 1 B4.3 Bases for Reactor Coolant System Integrity B4.3-1 B4.4 Bases for Containment Tests B4.4 1 B4.5 Bases for Safety injection Tests 54.5-1 B4.6 Bases for Emergency Containment Cooling System Tests 54.6-1 B4.7 Bases for Emergency Containment Filtering and Post Accident Containment Venting Systems Tests

• B4.7-1 B4.3 -

Bases for Emergency Power System Periodic Tests 54.8-1 84.9 Bases for Main Steam Isolation Valve Tests B4.9-1 B4.10 Bases for Auxiliary Feedwater System Tests 84.10-1 B4.11 Bases for Reactivity Anomalies B4.11 -1 B4.12 Bases for Environmental Radiation Survey B4.12-1 B4.13 Bases for Fire Protection Systems B4.13-1 B4.14 Bases for Snubbers B4.14-1 B4.15 Bases for Surveillance Requirements, Overpressure Mitigating System B4.15-1 iv Amendment Nos. and

e s LIST OF TABLES Table Title 3.5-1 Instrument Operating Conditions for Reactor Trip 3.5-2 Engineering Safety Features Actuation 3.5-3 Instrument Operating Conditions for Isolation Functions 3.5-4 Engineered Safety Feature Set Points 3.13-1 Safety Related Snubbers 3.14-1 Fire Detection System 3.17-1 Spent Fuel Burnup Requirements for Storage in Region 11 of the Spent Fuel Pit 4.1 -1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.t-2 Minimum Frequencies for Equipment and Sampling Tests 4.2-1 Reactor Coolant System In-Service inspection Schedule 4.12-1 Operational Environmental Radiological Surveillance Program 4.12-2 Operational Environmental Radiological Surveillance Program Types of Analysis 6.2-1 Operating Personnel v Amendment Nos. and .p. ~

3.12 CASK HANDLING Anglicability: Applies to limitations during cask handling. Obiectives To minimize the possibility of an accident during cask hand!!ng operations that would affect the health and safety of the public.

• ---ifWtions:

Ouring cask hand!!ng operations: (t) The spent fuel cask shall not be moved into the spent fuel pit until all the spent fuel in the pit has decayed for a minimum of 1525 hours.** (2) Only a single element cask may be moved into the spent fuel pit. (3) A fuel assembly shall not be removed from the spent fuel pit in a i shipping cask until it has decayed for a minimum of 120 days.' (4) HEAVY 1.OADS shall be prohibited from travel over irradiated fuel assemblies in the spent fuel pool (refer to T.S. 3.10.10). 8

• The Region 10 fuel which was in the Unit 3 reactor during the period of April 19,1981, through Aprl! 24, 1981, may be removed from the Unit 3 spent fuel pit in a shipping cask af ter a minimum decay period of ninety-five (95) days.
  • The soent fuel cask can be moved into the Unit 4 Spent Fuel Pit af ter a minimum decay of 1000 hours until the new two-region high density spent fuel racks are installed.

111 105 3.12-1 Amendment Nos. and

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r 83.12 SA5ES FOR LIMITING CONDITIONS FOR OPERATION, CASK HANDLING Requiring spent fuel decay time to be a minimum of 1525 hours prior to moving a spent fuel cask into the spent fuel pit will keep potential offsite doses well within 10 CFR Part 100 limits should a dropped cask strike the stored fuel assemblies. The restriction to a!!ow only a single element cask to be moved into the spent fuel pit will ensure the maintenance of water inventory in the unlikely event of an uncontrolled cask descent. Use of a single element cask which nominally weighs about twenty-five tons will also increase crane safety margins by about a factor of four. Requiring the spent fuel decay time be at least 120 days prior to moving a fuel assembly outside the fuel storage pit in a shipping cask will ensure that potential offsite doses are a fraction of 10 CFR 100 limits should a dropped cask and ruptured fuel assembly release activity directly to the atmosphere. The restriction on movement of MEAVY LOADS over irradiated fuel assemblies in the spent fuel pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the FSAR. For the ' purpose of this specification, HEAVY LOADS are defined as loads greater than 2000 pounds.(I) (Refer to T.S.1.36 and T.S. 83.10) t i i l

References:

(1) FSAR Table 3.2.3-1 ( I B3.12-1 Amendment Nos. and l L L .n, 3.< 3 -+-m:

j 3.17 SPENT PUEL STORAGE Appucability: Applies to limitations on the storage of spent fuel assemblles. Otnoctive To minimize the possibility of exceeding the reactivity design limits for storage of spent fuel. Soecifications: (1) Fuel asse nblies containing more than 4.1 weight percent of U-235 shall not be placed in the tingle region spent fuel storage racks. After installation of the two-region high density spent fuel racks, the maximum enrichmentloading for fuel assemblies in the spent fuel racks is 4.5 weight percent of U-235. (2) The minimum boron concentration while fuel is stored in the Soent Fuel Pit shall be 1950 ppm. (3)* Storage in Region II of the Spent Fuel Pit shall be further restricted by bumup and enrichment limits specified in Table 3.17-1. (4)* During the re-racking operation only, fuel that does not meet the burnup rwquirements for normal storage in Region !! may be stored in Region 11 in a checkerboard arrangement (i.e., no fuel stored in adjacent spaces). l

• This Technical Specification is applicable only after installation of the new two-region high density spent fuel racks.

i 3.17-1 Amendment Nos. and k -x s +~m. ee w w e4

TABLE 3.17-1 SPENT PUEL BURNUP REOUIREMENTS POR STORAGE IN REGION D OF THE SPENT PUEL PTT W W Burnup w/o GTD/MT 1.5 0 1.75 5.0 2.0 9.0 2.2 12.0 2.4 14.8 2.6 17.6 2.8 20.1 3.0 22.6 3.2 25.0 l 3.4 27.4 3.6 29.6 3.8 31.8 4.0 34.0 4.2 36.1 4.5 39.0 Linear interpolation between two consecutive points will yield conservative results. t i I l l l l l i 111 105 Amendment Nos. and

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l i 53.17 BASES FOR LIMITING CONDITIONS FOR OPERATION, SPENT PUEL STORAGE 1. The spent fuel storage racks provide safe suberitical storage of fuel assemblies by providing sufficient center-to-center spacing or a combination of spacing and Poison tr. assure k,ff s equal to or less than 0.95 for normal operations and i postulated accidents. l 5 F.-

2.

• IThe spent fuel racks are divided into two regions. Region ! racks have a 10.6 inch center-to-center spacing and the Region !! racks have a 9.0 inch center-to-center spacin6. Because of the larger center-to-center spacing and poison (B10) concentration of Region I cells, the only restriction for placement of fuel is that the initial fuel assembly enrichment is equal to or less than 4.5 weight percent of

. U-235. The limiting value of U-235 enrichment is based upon the assumptions in the spent fuel safety analyses and assures that the !!miting criteria for s criticality is not exceeded. Prior to placement in Region !! cell locations, strict controts are employed to evaluate burnup of the spent fuel assembly. Upon d': termination that the fuel assembly meets the burnup requirements of Table 3.17-1, placement in a Region 11 cell is authorized. These positive controls assdre t$e fuel enrichment limits assumed in the safety analyses will not be exceeded. s' d l l l i-l l l. i

• This Technical Specification is app!! cable upon Installation of the new two-region high density spent fuel racks.

83.17-1 Amendment Nos. UI and 105 f ,4 g'N'fr i f j

TABLE 4.12 (Sheet 2 of 3) WINIMUM FREQUENCIES FOR EOUIPMENT AND SAMPLING TE5TS Max. Time Q!agh Proevency Between Tests (Days)

5. Control Rods (cont'd)

Partial movement of Biweekly while 20 full length rods critical

6. Pressurizer Safety Valves Set Point Each refueling NA i

shutdown

7. Main Steam Safety Valves Set Point Each refueling NA shutdown

3. Containment Isolation Trip Fmetioning Each refueling NA shutdown

9. Refueling System Interlocks Fmetioning Prior to each NA refueling

10. Accumulator Baron Concentration At least once per 31 days and within 6 hours after each solution volume increase of >l%

of tank vehime.t !!. Reactor Coolant System Evaluate Daily NA Leakage

12. Diese! Fuel, Supply Fuel inventory Weekly 10

13. Spent Fue! Pit Baron Concentration Monthly 45 l

14. Fire Protection Pump and Operable

%mthly 45 Power Supply

15. Turbine Stop and Control Closure Monthly

• 45 Valves, Reheater Stop and Intercept Valves

16. LP Turbine Rotor Inspector V, MT, PT -

Every 5 years 6 years (w/o rotor disassembly)

17. Spent Fuel Cask Crane Functioning Within 7 days 7 days when Interlocks-crane is being used to maneuver spent fuel cask.

Amendment Nos. and Y** ~*

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5.4 PUEL STORAGE 1. The New and Spent Fuel Pit structures are designed to withstand the anticipated earthquake loadings as Class ! structures. Each Spent Fuel Pit has a stainless swel liner to ensure against leakage. ~ 2. The spent fuel storage racks provide safe suberitlcal storage of fuel assemblies by providing sufficient center-to-center spacing or a combination of spacing and poison to assure Keff, is equal to or less than 0.95 for normal operations and postulated accidents. Fuel assemblies containing more than 4.1 weight percent of U-235 shall not be placed in the single region spent fuel storage racks. After insta!!ation of the two-region high density spent fuel racks, the maximum enrichment loading for fuel assemblies in the spent fuel racks is 4.5 weight percent of U-235. The racks for new fuel storage are designed to store fuel in a safe suberitical array. The fuel is stored vertically in an array with sufficient center-to-center spacing to assure Keff equal to or less than 0.93 for optimum moderation conditions and equal to or less than 0.95 for fully flooded conditions. Fuel containing more than 4.5 weight percent of U-235 shall not be placed in the New Fue! Storage Area. 3. Credit for burnup is taken in determining placement locations for spent fuelin the two-region spent fuel racks.* Strict administrative controls are employed to evaluate the burnup of each spent fuel assembly stored in areas where credit for burnup is taken. The burnup of spent fuel is ascertained by careful analysis of burnup history, prior to placement into the exage locations. Procedures l shall require an independent check of the analysis of suitability for storage. A complete r3 cord of such analysis is kept for the time l period that the soent fuel &ssembly remains in storage onsite. j I i l l l

• During rack installation, it will be necessary to temporarily store Region I fuel in the Region U spent fuel racks. Strict administrative controls will be utilized to maintain a checkerboard storage configuration, i.e., alternate cell occupation, in the ' legion !!

acks. 111 105 5.4-1 Amendment Nos. and . ~ m y ~ m r -:.

[ '*\\ UMTED STATES es NUCLEAR REGULATORY COMMISSION

I W G.? ';6 }

e us cuason. o. c.sosss s,...../ s SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.111 TO FACILITY OPERATING LICENSE NO. DPR-31 AND AMEN 0 MENT NO. 105 TO FACILITY OPERATING LICENSE NO. DPR-41 FLORIDA POWER AND LIGHT COMPANY TURKEY POINT UNIT N05. 3 AND 4 DOCKET N05. 50-250 AND 50-251 l 1.0 Introduction By letter dated March 14, 1984 and supplemented on July 2 and 23, August 14 and 22, September 10 and 28, October 5, 9,18 and 26, and November 16, 1984 Florida Power and Light Company (FP&L) submitted an application to increase the storage capacity of the spent fuel pools (SFPs) for Turkey Point Units 3 and 4. by replacing the existing racks with new storage racks. Amendment 20 to Facility Operating License DPR-31, dated September 24, 1976,- temporarily allowed the storage capacity of the Unit 3 SFP to be increased from 217 to 235 fuel assemblies. Amendment Nos. 23 and 22 for Units 3 and

4. respectively, dated March 17, 1977, increased the SFP storage capacity at each facility to 621 fuel assemblies.

1.1 Discussion These proposed amendments will allow the licensee to expand the SFPs from .the current capacity of 621 fuel assemblies to 1404 fuel assemblies. This expansion will be accomplished by reracking the existing SFPs with neutron l absorbing (poison). spent fuel racks composed of individual cells made of [ stainless steel. The new spent fuel storage racks will be arranged in two discrete regions within each pool. Region I will consist of 286 locations which will nonnall/ be used for storage of spent fuel with an enrichment equal to or less than 4.55 U-235 at it's most reactive point in life. pegion 2 will consist of 1118 locations and will provide storage for spent fuel assemblies meeting required burnup considerations. The existing fuel storage racks (621 locations) have a nominal centerline-to-centerline spacing of 15.7 inches. The new Region 1 racks will have a 10.6 inch centerline-to-centerline spacing and Region 2 will be !~ 9.0 inches centerline-to-centerline spacing. The major components of the fuel rack assemblies are the fuel assembly cell, boraflex (neutron absorbing) material and the wrapper. The wrapper covers the Boraflex f -material and provides venting of the Boraflex to the pool environment. The existing racks have 635 total storage cells; however due to piping and other interferences the Unit 3 racks have 621 usable cells and the Unit 4 . racks have 614 usable cells. In the 1986-1987 time frame, the units will lose their full-core reserve storage capacity (157 assemblies) and in 'm $aau1g uuryq?kTD - t

. 1990-1991 time frame they will no longer have the capacity to store fuel discharged from the operating units. Since these dates are earlier than the date a federal depository should be available for spent fuel (1998),** additional capacity for the storage of spent fuel is needed. - Increasing the SFPs capacity to 1404 cells, as proposed, will allow plant operation with full core reserve in the SFPs to about the year 2005 for Unit 4 and 2006 for Unit 3. These time frames are based on the present FP&L fuel management. The proposed expansion of the SFP storage racks to 1404 cells should t'e adequate until the federal government begins accepting spent fuel from civilian power reactors. 2.0 Evaluation The " Spent Fuel Storage Facility Modiffcation Safety Analysis Report" provided by the licensee on March 14, 1984, in support of this application for amendments was the basis for the NRC staff cvaluation. Supplemental information provided by the licensee is also reflected in the following Safety Evaluation which sumarizes the NRC staff effort. 2.1 Criticality Considerations i Each pool will contain rz,cks that provide 1404 designatea locations for the storage of reactor fuel. The storage racks will be divided between two regions - one containing 286 locations and one containing 1118. The smaller region, having sufficient capacity for approximately 11/2 full cores, will be used for the storage of fresh fuel and feel not suitable for Region 2. The larger region will normally be restricted to fuel having a specified minimum burnup.. The licensee proposed that, during installation of the new racks, stora enrichmerit) ge of high reactivity spent fuel (up to fresh 4.5 percent be permitted in a checkerboard array with every other location empty. Administrative controls will be used to prevent storage in the empty locations. The Region I racks will consist of stainless steel cans of 8.75 inch square - interior dimension and 0.75 inch wall thickness. On the outer surface of each side of the cans Boraflex sheets having a minimum area density of 0.02 grams per square centimeter of 8-10 are held in place by a thin-walled '~ stainless steel wrapper plate. The rack structure maintains these cans on a 10.6 inch center-to-center spacing. The Region 2 rack design consists of stainless steel cans welded together to form a honeycomb type structure. The cans have an interior square dimension of 8.80 inches and are made of stainless steel.

• • Nuclear Waste Policy Act of 1982, Section 302(a)(5)

, ' ** ? 7 J.4.4ye T f *ee 'e

- All four sides of interior cans have Boraflex sheets containing 0.012 grams of B-10 per square centimeter of surface area that are held in place by a stainless steel wrapper which is spot welded to the can. Thr resulting structure maintains the stored fuel assemblies at a center-to-center spacing of 9.0 inches. 2.1.1 Calculation Methods The calculation of the effective multiplication factor, K , for Region 1 makes use of the AMPX system of codes for neutron cross-sI(( ion preparatica and the Monte-Carlo Code KENO-IV for reactivity. This code set has been verified against a set of 27 critical experiments that simulate various features of the rack design. A calculational methc4 bias of zero and uncertainty of 0.013 based on a 95 percent probability at the 95 percent confidence level (95/95) was inferred from these comparisons. The calculation of the criterion for acceptable burnup for storage in Region 2 makes use of the concept of reactivity equivalence. Since the KENO-IV code cannot handle burned fuel assemblies it is necessary to obtain the fresh fuel assembly enrichment which yields the same pool K assembly. Because of the presence of the poison in the RegT$$ as the burned 2 racks, a multigroup transport theory code is more appropriate than diffusion theory for this calculation. The PHOENIX code was used. The calculation proceeds as follows: 1. An end-point of 39.0 GWD/MT burnup for a bundle having an initial enrichment of 4.5 weight percent U-235 is chosen.

2. ' PHOENIX is used to calculate the K. of such an assembly in the rack geometry (including can and Boraflex absorber).

3. The burnup required to produce the same b is calculated for a number of, smaller enrichments. 4. The enrichment required to produce the same b without burnup is obtained (in the present case the value is 1.5 weight percent U-235). i l-5. KENO-IV is used to calculate the rack multiplication factor for ( the 1.5 weight percent enrichment assembly. The advantage of this procedure is that only relative multiplication factors are computed by PHOENIX. The final value of the.ack multiplication factor is obtained from the more powerful KENO-IV code. 2.1.2 Trettment of Uncertainties For the Region 1 analysis the total uncertainty is the statistical combination of the method uncertainty, the uncertainty in the particular KENO calculation, and mechanical uncertainties due to tolerances, spacing, l: etc. The mechanical uncertainties were treated either by making worst case l

L

e 4 assumptions (e.g., using the minimum rather than nominal value af the boron loading) or by perfoming sensitivity studies and obtaining a value of the uncertainty in rack multiplication factor due to uncertainty in dimensions, etc. In the Region 2 analysis the same uncertainties are considered along with others that are unique to the rack design and usage. These include uncertainty due to particle self-shielding in the boron (actually bias), 4 uncertainty in the plutonium reactivity and uncertainty in the re"tivity as a function of burnup. Including both the plutonium and burnup reactivity uncertainties is conservative since the latter includes the former as one of its components. The PHOENIX code was qualified for burnup calculations by comparing calculated isotopic ratios to measurements made in Yankee-Rowe Core 5, and by comparison of equivalent reactivity burnup between PHOENIX and the LEOPARD / TURTLE codes. A set of 81 critical experiments was analyzed to qualify the code for zero i burnup conditions. Conservative uncertainties of 5 percent of the reactivity change due to burnup have been assigned to these parameters. 2.1.3 Results of Analysis Nomal Storage For Region 1, the rack multiplication factor is calculated to be 0.9403, including uncertainties at least at the 95/95 level, when fuel having an enrichment of 4.5 weight percent U-235 is stored therein. Fuel of either the Westi.nghouse 15X15 standard or OFA design may be stored as well as Con 6ustion Engineering 14X14 or 16X16 and Exxon 14X14 designs. Pure water at 1.0 grams per cubic centimeter is assumed. For Region 2, the rack multiplication fa'ctor is 0.9304 for the most reactive irradiated fuel permitted to be stored in the racks, i.e., fuel with the minimum burnup pemitted for each initial enrichment, including at least 95/95 uncertainties. For fresh fuel (4.5 percent enrichment) stored in a checkerboard array in the racks, the effective multiplication factor is 0.8342. Calculation of the remaining uncertainties w:s not deemed necessary . in this case since assuming conservative values for these tems would still result in a final K'Niculations are obtained for pure water at-a density of for the checkerboard configuration well below the required 0.95. All . 1.0 grams per cubic centimeter. Burned fuel of the same designs as allowed in Region 1 may be stored in Region 2. Analyses were perfomed for all allowable fuel types and the proposed curve of burnup versus initial enrichment bounds the results of the calculation. Abnomal Storage Conditions Most abnormal storage conditions will not result in an increase in K of the racks. For example, loss of a cooling system will result in an TNrease in pool temperature but this causes a decrease in the K,ff value. ~ ,,,-n. u. ---w-- = = = ~ - ' * - -' " ~ '

It is possible to postulate events (e.g., a seismic event) which could lead to an increase in pool reactivity. However for such events credit may be taken for the approximately 1950 ppm of boron in the pool water. The reduction in the K value caused by the boron (approximatefy 0.25) more than offsets the rINtivity addition caused by credible accidents. 2.1.4 Sumary of Evaluation The following discussion sumarizes our evaluation of the proposed re-racking of the Turkey Point SFPs. We have reviewed the assumptions made in the perfonnance of the criticality analyses. These include use of the highest pennitted reactivity bundle, pure water moderator at a density of 1.0 gram per cubic centimeter, and an infinite array of assemblies. These are consistent with NRC guidelines and are acceptable. We have reviewed the uncertainties which have been included. For Region 1, these include variation in poison pocket thickness, stainless steel thickness, cell interior dimensions, center-to-center spacing, boron particle self shielding, and cell bowing. Other parameters, such as boron loading, are taken at their most conservative limits. For Region 2, additional uncertainties due to burnup calculations and calculations of plutonium worth are included. For both regions,calculatior:al uncertainties and biases are included. These uncertainties meet our requirements and are acceptable. We have reviewed the verification of the calculation methods. The KENO-IV code is widely used in the industry for the purpose of calculating fuel rack criticality. The set of benchmark critical experiments used to verify the calculatiions method encompasses the enrichment, separation distance and separating material used in the racks. 'The set of experiments used to verify the PHOENIX c do e for the reactivity equivalence calculations is adequate and encompassed the pellet size and enrichment of the fuel proposed for storage in the Turkey Point racks. The uncertainties in the burnup and plutonium worth are verified against Yankee Core 5 isotopics and comparisons with the Westinghouse design LEOPARD / TURTLE code package. We find that adequate verification of the codes used in the t - criticality analyses has been perfonned. The technique of using reactivity equivalencing to define the storage criterion (burnup as a function of initial enrichment) is, in some form,- in widespread use in the industry and is acceptable. For Region 1 racks we have compared the results of the Turkey Point calculation to a generic study and found them to be compatible. Finally the results of the calculation for Region 1 and 2 meet our acceptance criterion of less than or equal to 0.95 including all uncertainties at the 95/95 level. I' '- + y e s 3 p(* /j

l 6-We have reviewed the proposed Technical Specifications 3.17. B3.17, and 5.4 and find that they are consistent with the assumptions in the safety analysis and are acceptable. 2.15. Conclusions Based on our review, which is described above, we find the criticality aspects of the design of the spent fuel racks to be acceptable. We conclude that fresh Westinghouse 15X15 fuel of either the standard or OFA design as well as Combustion Engineering 14X14 or 16X16 and Exxon 14X14 designs may be safely stored in Region 1 so long as enrichment does not exceed 4.5 w/o U-235. We further conclude that any of these fuel types may be stored in Region 2 provided it meets the burnup and enrichment limits specified in Table 3.17-1 of the Turkey Point Units 3 and 4 Technical Specifications. During the installation of the new racks, fuel which does not meet this criterion may be stored in Region 2 provided it is stored in a checkerboard arrangement with every other location vacant. 2.2 Materials The safety function of the SFP and storage rack system is to maintain the spent fuel assemblies in a subcritical array during all credible storage conditions. We have reviewed the compatibility and chemical stability of the materials, except the fuel assemblies, wetted by the pool water. The only new material or components to be added during the proposed l modification are the nuclear absorter strips. The new spent fuel racks to be installed in both regions are constructed entirely of Type 304 stainless steel..except for the nuclear poison material. The existing spent fuel liner is constructed of stainless s storage racks will utilize Boraflex} eel. The high density spent fuel sheets as a neutron absorber. Boraflex has previously been approved as a neutron absorber and is currently being used in several SFP storage facilities. Boraflex consists of boren carbide powder in a rubber-like silicone polymeric matrix. The spent fuel storage rack configuration is composed of individual storage cells interconnected to fonn an integral structure. The major components of the assembly are the fuel assem6ly cells, the Boraflex material, and the stainless steel wrapper a~:end _tha Boraflat. The Boraflex absorber will not be sealed within the storage cell and vent paths for any gas generated during exposure will be available to the pool. The pool contains oxygen-saturated domineralized water containing boric acid. The water chemistry control of the spent fuel pool has been reviewed elsewhere and found to meet NRC recomumendations. 2.2.1 Corrosion and Material Compatibility The pool liner, rack lattice structare and fuel storage tubes are stainless steel which it compatible with the storage pool environment. In this environment of oxygen-saturated borated water, the corrosive deterioratgon of the Type 304 stainless steel should not exceed a depth of 6.00 X 10~ 2 inches in 100 years, which is negligible relative to the initial thickness. Dissimilar metai contact corrosion (galvanic attack) between the stainless steel of the pool liner, rack lattice structure, fuel storage tubes, and the g C YO ' ' ' * *. 4 ^ yy w y, .s w .n-o ,n.

=-. I - Inconel and the Zircaloy in the spent fuel assemblies will not be significant because all of these materials are protected by highly passivating oxide films and are therefore at similar potentials. The Boraflex is composed of non-metallic materials and therefore Will not develop a galvanic potential in contact with the metal components. Boraflex has undergone extensive testing to study the effects of gamma irradiation in various environments, and to verify its structural integrity and suitability as a neutror absorbing material. The evaluation tests have shown that the Boraflex is unaffected by the pool water environment and will not be 3 degraded by corrosion. Tests wp perfomed at the University of Michigan, exposing Boraflex to 1.103 X 10 rads of ganna rcdiation with substantial concurrent neutron flux in borated water. These tests indicate that Boraflex maintains its neutron attenuation capabilities after being subjected to an environment of borated water and gamma irradiation. Irradiation will cause some loss of flexibility, but will not lead to break up of the Boraflex.4 Long tem borated water soak tests at high temperatures were also conducted. The tests show that Boraflex withstands a borated water innersion of 240*F for 260 days without visible distortion or softening. The Boraflex showed no evidence of swelling or loss of ability to maintain a uniform distribution of boron carbide. The space which contains the Boraflex is vented to the pool at each storage tube assembly. This venting will allow gas generated by the chemical degradation of the silicone polymer binder during heating and irradiation to escape, and will prevent bulging or swelling of the inner stainless steel wrapper. The tests have shown that neither irradiation, environment nor Boraflex composition has a discernible effect on the neutron transmission of the Boraflex material. The tests also show that Boraflex does not possess-leachable halogens that might be released into the pool environment in the presence;of elemental boron from the Boraflex. Boron carbide of the grade nomally in the Boraflex will typically contain 0.1 wt.% of soluble baron. The test results have confimed the encapsulation function of the silicone polymer matrix in preventing the leaching of soluble species from the boron carbide. To provide added ass'urance that no unexpected corrosion or degradation of materials will compromise the integrity of the racks, the licensee has connitted to conduct a long tem poison coupon surveillance program, which will be representative of the material used in both the Region 1 and Region 2 locations. There will be four sets of coupons, each containing not less 4 than 24 jacketed poison coupons, each set will be designed to be hung on the outside periphery of Region 1 and Region 2 modules. The initial surveillance of the specimens will be perfomed after approximately five years of exposure to the pool' environments. Subsequent surveillances will be based on the initial results to assure acceptable material perfonnance throughout the life of the plant. Construction materials will conform to the requirements of the ASME Boiler and Pressure Vessel Code Section II-NP. i-t nn:

. 2.2.2 Conclusion From our evaluation as discussed above, we conclude that the corrosion that will occur in the SFP environment should be of little significance during the life of the plant. Components in the SFPs are constructed of alloys which have a low differential galvanic potential between them and have a high resistance to general corrosion, localized corrosion, and galvanic corrosien. Tests under irradiation and at elevated temperatures in borated water indicate that the boraflex material will not undergo sionificant degradation during the expected service life. We further conclude that the environmental compatibility and stability of the materials used in tne expanded SFPs is ad64uate based on the test data cited above and actual service experience in operation reactors. We have reviewed the license 6's surveillance program and conclude that the monitoring of materials in the SFPs will provide reasonable assurance that the Boraflex material will continue to perfom its function for the life of the pools. The materials surveillance program will reveal any instance of deterioration of the Boraflex that might lead to the loss of neutron absorbing power well before significant deterioration will occur. We do not anticipate, however, that such deterioration will occur. We, therefore, conclude that the compatibility of the materials and coolant used in the SFPs is adequate based on tests, data, and actual service experience in operating reactors, and the selection of of appropriate materials and adoption of a surveillance program by the licensee meets the requirements of 10 CFR Part 50, Appendix A, Criterion 6I, having a capability to pemit. appropriate periodic inspection and testing of components and criterion 62, preventing criticality by maintaining structural integrity of the components and boron poison and is, therefore, acceptable. '2.2.3 References - Materials 1. J. S. Anderson', "Boraflex Neutron Shielding Material -- Product l Perfomance Date," Brand Industries, Inc., Report 748-30-1, (August 1979). p' 2. J. R. Weeks. " Corrosion of Materials in Spent Fuel Storage Pools." BNL-NUREG-23021, July 1977. f 3.. J. S. Anderson, " Irradiation Study of Boraflex Neutron Shielding Materials," Brand Industries. Inc., Report 748-10-1, (August 1981). i 4 J. S. Anderson, "A Final Report on the Effects of High Temperature Borated Water Exposure on BISCO Boraflex Neutron Absorbing Materials," Brand Industries, Inc., Report 748-21-1, (August 1978). 2.3 Structural Design l Our evaluation of the structural aspects of the proposed modifications are [ based on a review perfomed by the staff's consultant, Franklin Research l Center (FRC). The FRC Technical Evaluation Report TER-C5506-529, is appended i L i y y e v asy * ,.n. ,~-

J to this Safety Evaluation and provides additional details relating to the structural evaluation. 2.3.1 Description of the Spent Fuel Pools and Racks There are two SFPs at Turkey Point, one for each unit. They are constructed of reinforced concrete whose walls and floors are lined with a 1/4 inch-thick water tight stainless steel liner. The fuel assembly storage area is approximately 41'-4" wide by 25'-4" long. Wall thicknesses are 5'.-6" on three sides and 4'-0" on the fourth side. The floors of the pools are supported directly on foundation soil. The Region I storage racks are composed of individual storage cells made of stainless steel. The cells within a module are interconnected by grid assemblies to forin an integral structure. Each rack module is provided with leveling pads which contact the SFP floor and are remotely adjustable from above throughout the cells at installation. The modules are freestanding and are not anchored to floor nor braced to the pool walls. The fuel rack assembly consists of three major sections whici, are the leveling pad assembly, the lower and upper grid assemblies, and the cell assembly. The Region 2 storage racks consist of stainless steel cells assembled in a checkerboard pattern, producing a honeycomb type structure. The cells are welded to a base support assembly and to one another to forin an integral structure without the use of grids which are used in the Region 1 racks. This design is also provided with leveling pads which contact the SFP floor and are remotely adjustable from above through the cells at installation. The modules are free standing and are not anchored to the floor nor braced to the pool walls. The fuel rack module consists of two major sections which are-the base sup ort assembly and the cell assembly. 2.3.2-Applicable Codes, Standards and Specifications Load combinations and acceptance criteria were compared with those found in the " Staff Position for Review and Acceptance of Spent Fuel Storage and l. Handling Applicatio'ns" dated April 14, 1978 and amended January 18, 1979. The existing concrete pool structure was evaluated for the new loads in accordance with the requirements of the Turkey Point FSAR Section 3.8.4 2.3.3 Loads and Load Combinations. Loads and load combinations for the racks and the pool structure were reviewed and found to be in agreement with the applicable portiens of the staff position and the Turkey Point FSAR as identified in Section 2.3.2 of c-l this SE. Additional details are provided in the Appended TER. 2.3.4 Seismic and Impact Loads Seismic loads for the rack design are based on the oriainal design floor acceleration response spectra calculated for the plant at the licensing 0.159 safe shutdown earthouake (g operating basis earthquake (08E) and stage. This was based on a 0.05SSE). The seismic loads were applied to the i . model in three orthogonal directions. Loads due to a fuel bundle drop I l ~ ~ <-m;

o . accident were considered in a separate analysis. The postulated loads from these events were found to be acceptable. Additional, description and details are provided in the appended TER. 2.3.5 Design Analysis of Procedures

a. Design and Analysis of the Racks The dynamic response and internal stresses and loads are obtained from a seismic analysis which is performed in two phases. The first phase is a time history analysis on a simplified nonlinear finite element model. The second phase is a response spectrum analysis of a detailed linear three dimensional rack assembly finite element model. Two percent damping is.used in the seismic analysis for both the OBE and SSE. Further details on the methodology is discussed in the appended TER.

Calculated stresses for the rack components were found to be within allowable limits. The racks were found to hav2 adequate margins against sliding and tipping. An analysis was conducted to assess the potential effects of a dropped fuel assembly on the racks and results were considered satisfactory. An analysis was conducted to assess the potential effects of a stuck fuel assembly causing an uplift load on the racks and a corresponding downward load on the lifting device as well as a tension load in the fuel assembly. Resulting stresses were found to be within ecceptance limits.

o. Analysis of the Pool Structures The SFPs'are reinforced concrete plate structures supported on compacted limerock fill. The SFP walls are lined with 1/4-in. stainless steel liners.

These existing structures were analyzed for the modified fuel rack loads using a finite element computer program. Original plant response spectra and damping values were used in consideration of the seismic loadings. Design criteria, including loading combinations and allowable stresses, are in compliance with Turkey Point FSAR Appendix SA and the existing SFPs are determined to safely support the loads generated by the new fuel racks. 2.3.6 Conclusions Based on the above and appended TER, the staff concludes that the proposed rack installation will satisfy the requirements of 10 CFR 50, Appendix A (GDC 2, 4, 61 and 62), as applicable to structures. 2.4 Installation of Racks and Load Handling There is spent fuel in both Turkey Point Unit 3 and 4 SFPs. A temporary crane will be used to move the racks into ani out of the SFPs. The movement of the temporary crane will be over the excitsion areas as defined in the licensee's Phase I submittal for NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." However, the licensee has performed a load drop analysis which indicates that the consequences of a postulated load drop or d 9 g Tt

f 11 - temporary construction crane drop would be bounded by the cask drop accident. Furtherinore the licensee has re-evaluated the cask drop accident using the assumption that all of the spent fuel in the pool was damaged and the newest fuel in the pool had been cooled for at least 1525-hours. Technical Specification 3.12 has been revised to require a decay time of 1525 hours for all fuel in the spent fuel pool prior to cask handling operations. This evaluation is conservative in that not all of the fuel would be damaged in a real cask drop accident. The NRC staff's independent evaluation of the cask drop accident in suoport of the existing SFP racks dated March 17, 1977, resulted in conservatively estimated two-hour radiation doses at the exclusion area boundary (EAB) of 24 Rem to the tyroid and less than 1 Rem to the whole body. Our independent evaluation of the cask drop for the proposed SFP reracks resulted in conservatively estimated two-hour radiation doses at the EA8 of 26 Rem.to the thyroid and less than 1 Rem to the whole body. The slight increase to the thyroid is insignificant when comparted to the 10 CFR 100 guidelines for 'the two-hour dose of 300 Rem to the thyroid and the 1 Rem to the whole body, 4 -in both cases,1s.significantly less than the two-hour dose of 25 Rem whole body provided in the 10 CFR 100 guidelines. l Based on the above, the staff concludes that load handling accidents associated with these SFP modifications will not have any adverse consequences as identified in NUREG-0612, are well within the 10 CFR 100 guidelines, and are acceptable. 2.5 Radiological Consecuences of Accident __ Involving Postulated Mechanical Damage to the Spent Fuel This portion of the staff's review was conducted in accordance with the cuidance provided in NUREG-0800 " Standard Review Plan", Section 15.7.4, NUREG-0612, and NUREG-0554 with respect to the accident assumptions. . For evaluation of accidents involving the spent fuel pool, three types of accidents were considered; a cask drop or tip, a construction accident during rack replacement and a fuel assembly drop while handling fuel. As l. 'noted in Section 2.4 of this SE, the effects of a postulated load drop'are i bounded by the cask drop accident. I 2.5.1 Cask Drop /Tip Accidents Proposed. technical specification 3.12 will require a minimum of 1575 hours of decay for all spent fuel stored in either pool prior to cask handling l operations. A conservative estimate of damage to stored spent fuel assemblies would be from impact of a cask which is sufficient to damage 91 f assemblies (in the appropriate strike sector) and result in the release of their concomitant volatile gap activities. In performing our independent accident radiological consequences analysis, we assumed that the fuel has been discharged from the reactor after operation at a steady-state power level of 2300 MW for an extended period of time. The calculated (0-2 hr.) eh offsite accident radioloaical consequences are estimated to be 26 Rem thyroid and less than 0.1 Rem whole body at the Exclusion Area Boundary. These consequences are well within the radiological guideline values i l ..,c. o -+ --,-----,,,ve ,w w n-+.r,---.,-- -, m - - - - - > ---e -, - -,, -, + - ~, - ~ ~,,-------w-4 ~, - - - - - ->wn- ~

( I specified in 10 CFR 100. See Section 2.4 of this SE for additional details. Radiological consecuences at the Low Population Zone Boundary (LPZ) are connensurately less than those at the Exclusion Area Boundary (EAB). 2.5.2 Construction Accidents For purposes of ensuring that a conservative estimate of damage to stored fuel assemblies from impact of an unspecified object in a non-mechanistically defined construction accident is made, sufficient damage to 157 assemblies (a full core offload) to result in the release of their concomitant volatile gap activities was postulated conservatively. The-licensee has indicated in their submittals that the reracking operation will take place no sooner than 2150 hours after shutdown for the last batch of spent fuel.placed in the SFP. This is to compensate for an 8 ft. water level reduction in the spent fuel pool during rack handling operations. The additional cooldown time compensates for a reduction in pool fodine decontamination factor from 100 to 10 during this period, based upon staff analyses used to detemine the Regulatory Guide 1.25 value of 100 for a 23 foot water depth. In perfoming our independent accident radiological consecuence analysis, we assumed that the fuel has been discharged from the reactor after operation at a steady-state power level of 2300 PW for an a extended period of time. The calculated (0-2 hr.) offsite acciddHt radiological consequences are estimated to be 45 Rem thyroid and 0.5 Rem whole body at the EA8. These conseovences are well within the cuidelines of. 10 CFR 100. Radiological consequences at the LPZ are commensurately less than those at the EAB. 2.5.3 Fuel Handling Accident ^ The postulated fuel handling accident is not directly related to the rereackihg application. The fuel handling accident involves the release of ). the equivalent gap activity of one assembl.y recently discharged from the . reactor for the current fuel exposure of 50,000 Mwd /t. L In perfoming our independent radiolocical consequence analysis for the fuel-handling accident, we assumed that the fuel has been discharged from the reactor after operation at a steady-state power level of 2300 MW for an w extended period of time. Thecalculated(0-2hr.)offsiteacciMHt radiological consequences are estimated to be 30 Rem thyroid and 0.1 Rem whole body at the EAB, well within the guidelines of 10 CFR 100 for the two-hour dose of 300 Rem to the thyroid and 25 Rem to the whole body at t'he EAB. Radiologica'l consequences at the LPZ are commensurate 1y less than those at the EAB. 2.5.4 Conclusions The staff concludes that a cask drop /tip or construction accident resulting in damage to either ninety-one 50,000 mwd /t spent fuel assemblies or 157 similar assemblies with at least 1525 hours and 2150 hours of cooldown time, respectively, will result in atmospheric radionuclide releases with consequences which are well within the dose guidelines of 10 CFR 100. 4 "<rrr, s +. + __ y

. Additionally, the staff concludes that a fuel handling accident resulting in damage to a recently discharge 50,000 mwd /t spent fuel assembly will result in atmospheric radionuclide releases which are well within the dose guidelines of 10 CFR 100. t 2.6 Occupational Radiation Exposure The occupational exposure for the licensee's plan for the removal and ' disposal for the high density racks, and installation of the higher density racks is approximately 59 person-rems. This estimate is based on the . licensee's detailed breakdown of occupational exposure for each phase of the i modification. The licensee considered the number.of individuals perfoming a specific job, their occupancy time while perfoming this job, and the 1 I average dose rate in the area where the job is being perfomed. The spent fuel assemblies-themselves contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel. One potential source of radiation is radioactive activation or corrosion products called crud. Crud may be released to the pool water because of fuel moverients during the proposed SFP modifications. This could increase radiation levels in the. vicinity of the pools. During refuelings, when the spent fuel is first moved into the fuel pool, the addition of crud to the pool water from the fuel assembly and from the introduction of primary coolant to the pool water is greatest. However, the licensee does not 1 expect to have significant releases of crud to the pool water during. modification of the pool. Another source of radioactivity in the SFP water is fission products. The fission products are released through minute defects in.the fuel cladding and are significantly reduced when removed from ( the reactor vessel and are no longer being irradiated. The purification system for the pool, which has kept radiation levels in the vicinity of the rol to' low levels,-includes filters and domineralizers to remove crud and l'~. redionuclides. The purification systems will be operating during the modification of the pools. FPL's operating experiences has shown that the storage of additional fuel due to reracking will not contribute to the amount.of crud released to the pool. If crud deposits should become a significant contributor to pool doses, measures will be taken to reduce such doses to ALARA. The licensee has presented two alternative plans for removal and disposal of the old racks. These are (1) to decontaminate and dispose of as radioactive [ . waste for burial or (2) decontan;inate and dispose of as nonradioactive waste in accordance with existing Turkey Point health physics procedures. The old racks will be rinsed by hydrolasing to remove any loose contamination. This i operation will be performed. underwater to minimize airborne radioactivity levels. -In any event, the disposal methodology will follow ALARA guidelines l for each of the alternatives. I~ Divers will not be used during the reracking operation and no underwater l work will be necessary except some simple manipulations which can be l perfomed from above the surface of the pool using special tools. If divers are needed, detailed procedures will be developed and submitted to the staff for review. l ,7 zy ~ -~r v

14 - The licensee has taken measures to ensure that personnel expnsures during the SFP modifications are ALARA. These meesures are described in the licensee's radiation protection program which assures compliance with established procedures to maintain doses ALApA. FPL's radiation protection program was reviewed prior to the last rerack and was detemined adeouate and acceptable by the staff. Based on the manner in which the licensee will perfom their modifications, their radiation protection program, including area and airborne radioactivity monitoring, and relevant experience from other operating i reactors that have perfomed similar SPF modifications, the staff concludes that the licensee's SFP modifications can be performed in a manner that will ensure as low as is reasonably achievable (ALARA) exposures to workers. We have estimated the increment in onsite occupational dose during nomal operations after the pool modifications resulting from the proposed increase in storage fuel assemblies. This estimate is based on infomation supplied by the licensee for occupancy times and for dose rates in the spent fuel aret from radionuclide concentrations in the SFP water. The spent fuel assemblies themselves contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel. Based on present and pro.iected operation, we estimate that the proposed modification should add less than one percent to the total annual occupational exposure of 870 person rem / year / unit (for the years 1970-1982). - 2.6 Conclusion The basis of our acceptance of Turkey Point's occupational dose control programs is that doses to personnel will be maintained within the limits of 10 CFR 20 " Standards for Protection Against Radiation", and as low as is reasonably achievable. Based on present and projected operations in the SFP area, we estimate that the proposed modifications should add less than one . percent to the total annual occupational radiation exposure at both units. The small increase in radiation exposure should not affect the licensee's ability to maintain. individual occupational doses to as low as is reasonably' l achievable levels and within the limits of 10 CFR 20. Thus, we conclude [- that storing additional fuel in the two pools will not result in any significant increase in doses received by workers. , 2.7 Spent Fuel Pool Cooling and Makeup Systems l Each SFP cooling loop consists of a pump, heat exchanger, filter, domineralizer, piping, and associated valves and instrumentation. The pump draws water from the SFP pit, circulates it through the heat exchanger, and returns it to the pit. Component Cooling Water cools the heat exchanger. ' Redundancy of this equipment is not required because of the large heat capacity of the pit and its corresponding slow heat-up rate. Nonetheless, a 100-percent-capacity spare pump which is permanently piped into the SFP i - cooling system has been installed. This pump is capable of operating in l-place of the originally installed pump, but not in parallel with the originally installed pump. Also, alternate connections are provided for [ - connecting a temporary pump to the spent fuel pit loop. l E- . n.,, s --+r e ,~ .-,-,,,,y n. sc-.--m.-gen.~-,3, ,--m.--- -e-

i The existing cooling systems for the SFPs are not safety grade and there are no connections to the shutdown cooling system or other safety related - cooling systems. Therefore in accordance with the Standard Review Plan Section 9.1.3, we assumed that all pool cooling would be lost Tollowing a safe shutdown earthquake. Assuming the loss of cooling, boiling would occur after 7.6 hours for.the normal heat load condition and after 1.6 hours for the maximum heat load condition for the new racks. This would result in a boil off rate of 37.0 and 72.0 gpm, respectively. The licensee has connitted to upgrade the SFP cooling systems such that they will remain functional after a safe shutdown earthquake. The SFPs will be analyzed and modified, i, as necessary, to assure that the cooling function is not lost as the result of the seismic event. The design, procurement, and construction associated with this upgrade will be completed by the end of the second refueling outage after issuance of approval for the re-racking of the SFPs. The structural considerations of the thennel loads imposed by a pool water - temperature of 212*F on the steel liners and the concrete have been reviewed by the Structural Engineering Branch. The resulting tensil stress is 38 ksi versus the allowable value of 36 ksi. However, realizing the self-relieving nature of the themal stresses and further acknowledging that the section in general remains elastic, pool function and structural integrity are maintained. See Section 3.4.3 of the appended TER for further details..The radiological effects have been reviewed by the Accident Evaluation Branch. An independent accident evaluation of the radiological consequences of SFP boiling was performed. The offsite radiological consequences were found to be a small fraction of the 10 CFR 100 guidelines, provided that sufficient make up water capacity is available. The proposed rerack will result in no significant change in the time to boiling under the presently authorized storage. Until the upgrade is complete the amount of fuel that will be stored will be less than the capacity of the existing racks. Multiple alternate means of makeup water are' available until seismically upgraded. Temporary connections can be provided from the fire water system or from the primary water storage. tank.. Additionally, there are two firehouses nearby such that, should a safe shutdown earthquake occur before the upgraded cooling system is operational, fire engines could be available in less than an hour and provide makeup water to the pools. Thus, adequate time is available to provide the - necessary makeup water. 2.7.1 Support Systems - The'SFP cooling system heat exchangers are cooled by the component cooling water systems. The component cooling water system heat exchangers are cooled by the service water systems. The licensee proposed no modifications to these two systems as part of this spent fuel pool expansion project. These systems were a reviewed as to their adequacy to remove the additional heat load and were found to be capable of removing the additional heat. a' 2.7.2 Decay Heat Loads The licensea's calculated spent fuel discharge heat load to the pools, which was determined in accordance with the Branch Technical Position ASB 9-2, e s. ..n, m-,,,..c-_ ,,.w .we.., ,_,w_, ,m.._p..-,en,.,3,.- -y

1-i 16 - " Residual Decay Energy for Light Water Reactors for Long Term Cooling", and the Standard Review Plan Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System", indicates that the expected maximum normal heat load following the last refueling will' be 17.9 MTU/Hr. This heat load will result in a maximum bulk pool temperature of less than 143'F. This normal pool . temperature (143'F) is higher than the acceptance criteria of 140*F as defined in the Standard Review Plan, however, it is acceptable because the heat load calculations considered each reload to consist of one half of a core instead of the actual reloads being thirds of a core. Had the calculations been performed using the third core reloads, the pool temperature would have been less than the 140'F. The expected maximum . abnormal heat load fcilowing a full core discharge is 35.0 MTU/Hr. This F abnormal heat load results in a maximum bulk pool temperature of less than 183'F which is below boiling (212*F) and within the acceptance criteria identified above. 2.7.3 Conclusions Based on the above, we have concluded that the preposed overall SFP modifications are acceptable with respect to the storage rack capacities, .the SFP cooling system capabilities, support system capabilities, the heat loads and pool water temperatures. 2.8 Radioactive Waste Treatments The Turkey Point plant contains radioactive waste treatment systems designed to collect and process the gaseous, liquid, and solid wastes that might contain radioactive material. The radioactive waste treatment systems were evaluated in the Safety Evaluation dated March 1972, in support of the issuance-of the Operating Licenses. There will be no change in the conclusions given regarding the evaluation of these systems because of the proposed spent fuel pool rerack. 2.8.1 Conclusion Our evaluation of the radiological-considerations supports the conclusion that the proposed installation of new spent fuel storage racks at Turkey Point Unit Nos. 3 and 4, is acceptable based on the fact that previous-conclusions relating to the radioactive waste treatment systems, as found in the Turkey Point Unit Nos. 3 and 4 Safety Evaluation, are unchanged by the installation of new spent fuel storage racks. 3.0 Significant Hazards Consideration Comunents The request for these amendments was individually noticed on June 7, 1984 (49 FR 23715) followed by a monthly notice on July 7, 1984 (49 FR 29925). Conenents, request for a hearing and petition for leave to intervene were filed on July 9,1984, by the Center for Nuclear Responsibility and Ms. Joette Lorion. We have addressed the relevant consents in the text of this 4 Safety Evaluation. The petitioners contend: l + ___..m.m., ..,__._m_ ,,_.,,m,..

~

- "A.1 The Comission has traditionally held, in a series of case law that expansion of the spent fuel facility constitutes a significant safety hazards consideration."

Under the Comission's regulations in 10 CFR 50.92, an initial determination that the proposed arendments involve no significant hazards consideration was made based on a determination that on the operation of the facilities in accordance with the proposed amendments would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of-safety. Section 4.0 of this Safety Evaluation contains the Final No Significant Hazards Consideration i - Determination based on our evaluation and the fact that the reracking i technology in this instance, has been well developed and utilized (over 100 ) similar applications'have been approved) and the K,ff of the SFPs will be maintained less than or equal to 0.95. "A.2 Acceptance criteria for criticality will not be met and thus, FPL will not be 'able to ensure that the fuel storage facility will always be subcritical by a safe margin in both normal operating and accident conditions." . This contention is addressed in Section 2.1 (Criticality Considerations) of this SE. The criterion for the neutron multiplication factor (X storage of spent fuel is less than or equal to 0.95 including alt f) for f uncertainties at the 95/95 probability confidence lavel. As noted in Section 2.1, this criterion is met for all normal and abnormal conditions for storage of the spent fuel in.the proposed configuration at the Turkey Point facilities. "A.3 The recitation and notice in 48 (sic) Federal Register Notice 23715, Vol. 49, No. Ill, June 7,1984, that the established acceptance criteria for criticality in the spent fuel pool shall be kept at or below K 0 untrue as evidence.by 48 (sic) Federal Register Notice 25360, Nume.95 is 49, No. 120, June 20, 1984." This contention is incorrect. As noted above in response to contention A.2, i the K for the SFPs is maintained equal to or less than 0.95 including allubrtaintiesatthe95/95probabilityconfidencelevel. The June 20, 1984 Federal Register Notice (49 FR 25360) was related to a separate action addressing the existing new fuel (unirradiated) storage racks which are not affected by these proposed amendments. "A.4 In light of the fact that the utility, FPL, wants to operate the facility with a K proposed undertakTk of 0.98 (FR 25360), as above referenced, places the in the Significant Safety Hazards Category, and there can be no issuance of a license amendment to expand the spent fuel facility without a public hearing required by the Atomic Energy Act of 1954." - In support of contentions A.1 - A.4 the petitioners note the position taken by the Comission in Policy Issue SECY-83-337, STUDY ON SIGNIFICANT SAFETY HAZARDS, August 15 1983: n:., m- -~ .u ~ ~. - -, - - -, -, - - - - - - - - ~ - - - - - - - - - - - - - - -

.1 18 - "A X of greater than 0.95 may be justifiable for a particular applTbtionbutitwouldgobeyondthepresentacceptedstaffcriteria and would potentially be a significant hazards consideration." page 5-6. This contention is factually incorrent. As indicated in responses to contentions A.1 through A.4, the SFPs for Turkey Pont Units 3 and 4 utilize current and accepted technology and the K,ff will be maintained less than or i_ equal to 0.95. 4.0 Final No Significant Hazards Consideration The standards used to arrive at a proposed determination that a request for amendments involves no significant hazards consideration are included in the Commission's regulations,10 CFR 50.92, which state that the operation of the facilities in accordance with the proposed amendments would not (1) involve a significant increase in the probability or consequencs of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The proposed SFP expansion amendments are very similar to the initial SFP expansions, identified in Section 1 of this SE, in which many of the same issues were raised and resolved when the initial expansions were approved. Each specific aspect of this request was reviewed in detail and was very much a repeat of the initial expansion review. The knowledge and experience gained by the NRC staff in reviewing over 100 similar requests was also 4-utilized. The current expansion request does not use any new or unproven technology in either the construction process or in the analytical-techniqu.es necessary to support the expansion request. The same postulated accidents were looked at again and the same precautions have been proposed by the licensee during the installations. In addition, the neutron multiplication factor (K -lessthan0.95 including'Nc)ertainties.of the pools will be maintained equal to or Accordingly, the staff has detemined that the request for amendments to expand (reracking to allow closer spacing) does not significantly increase the probability or consequences of accidents previously evaluated; does not create new accidents not previously evaluated; and does not result in any significant reduction in the margins of safety 'with respect to criticality, t cooling or structural considerations. The following evaluation in relation to the three standards demonstrates that the proposed amendments in support of the SFP expansions do not involve a significant hazards consideration. First Standard - Involve a significant increase in the probability or consequences of an accident previously evaluated. The following potential accident scenarios have been identified: 1. A spent fuel assembly drop in the spent fuel pool. p s*$'", .--,.,--m- ,,c -,ya-nn.- --,.p ---,,,-y- .,-p

19 - 2. Loss of spent fuel pool cooling system flow. 3. A seismic event. 4 A spent fuel cask drop. 5. A construction accident. The probability of any of the first four accidents is not affected by the . racks themselves; thus reracking cannot increase the probability of these accidents. As for the construction accident, FPL does not intend to carry any rack directly over the stored spent fuel assemblies. All work in the spent fuel pool area will be controlled and perfonned in strict accordance with specific written procedures. Details on the precautions and requirements related to the installation and load handling during the SFP expansion activities and the licensees compliance to the recuirements of NUREG-0612 " Control of Heavy Loads at Nuclear Power Plants" are provided in i our SE dated August 29, 1984 Accordingly, the proposed expansion does not significantly increase the probability of an accident previously evaluated. ,e The consequences of (1) a spent fuel assembly drop in the SFP and (4) a spent fuel cask drop and (5) a construction accident are discussed in detail in Sections 2.4 and 2.5 of this SE. i As noted in Section 2.4 of this SE, a load drop analysis was perfonned and indicates that the effects or consequences of.a postulated load or temporary construction crane drop are bounded by the cask drop analysis. The consequences of the cask drop accident analysis results in a slight increase from the previous analysis for the existino racks in the estimated two-hour i radiation doses at the EA8 of 2 Rem to_the thyroid with no change to the . estimated doses to the whole body. The estimates resulting from our current analysis of 26 Rem to the thyroid and 1 Rem to the whole body are significantly less.than the two-hour dose of 300 Rem to the thyroid and 25 Rem to the whole body at the EAB provided in 10 CFR 100 guidelines. The postulated fuel handling accident is not directly related to SFP expansion request as stated in Section 2.5.2 of this SE. The results O our analysis assuming fuel exposure of 50,000 Mwd /t and steady-state power level ofr 2300 W results in 30 Rem th:vroid and 0.1 Rem whole body at the EAB, h ' well withii the 10 CFR Part 100 guidelines identified above. There will be no significant increase in the consequences in that the fuel handlinn accident is not directly related to the SFPs storage capacity but is dependent on the release of the equivalent gap activity of a sinole assembly - recently removed from the reactor. Section 2.3.4, and 2.3.5 and the Appended TER of this SE indicate that the postulated loads from a seismic event will not result in failures to the racks or pool structures, thus their integrity will be maintained. Neither the staff nor the license could identify any new means of losing cooling water. Therefore, since the integrity of the racks and SFP will be maintained there will be no significant change in the consequence of a 1-

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4 - , p y. --a - - - - - - -ar---,-r m-.v.-a e ve=w.-- w-- -w,- ,w-<-wrw ht'=rw---yr-m4v**----- * -- - * - - -------- -g-y w w- + -, - -, - wr ww-vem----

20 - seismic event as the result of this amendment than previously evaluated seismic events. As stated in Section 2.7 of this SE, the proposed rerack will" result in no significant change in time to boiling under the presently authorized storage. The existing SFP cooling systems are not seismic Category 1, however, the licensee has connitted to upgrade the systems to assure functional capability. Adequate time is available to provide the necessary makeup water from either on-site sources or fire engines from a nearby fire house. Thus,-the time 1 available and alternate means of providing makeup water to the SFP result in I no significant increase in the consequences of loss of flow from that previously evaluated. Therefore, based on the above, the probability or consequences of previously analyzed accidents will not be significantly increased as the result of the proposed SFP expansions. Second Standard - Create the possibility of a new or different kind of accident from any accident previously evaluated. The propt'ed SFP expansions have been evaluated in accordance with the guidance of the NRC position paper entitled, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", appropriate NRC - Regulatory Guides, appropriate NRC Standard Review Plans, and appropriate Industry Codes and Standards as identified in this SE. In addition..several previous NRC SEs for SFP expansions similar to this proposal have been reviewed. Neither the licensee nor the NRC staff could identify a credible mechanism for breaching the structural integrity of the SFPs which could result in loss of cooling water such that cooling flow could not be maintained or any other accidents not previously evaluated that might result from these amendments. I .As a result of this SE and these reviews, the proposed 5FP expansions do not, in any way, create the possibility of a new or different kind of accident from any accident previously evaluated for the Turkey Point SFPs. Third Standard - Involve a significant reduction in a margin of safety. l The NRC staf F safety evaluation review process has ' established that the (- issue of margin of safety, when applied to a SFP modification, will need to address the following areas: 1 1. Nuclear criticality considerations. 2. Themal-Hydraulic considerations. r 3. Material, Structural and Mechanical Considerations. The established acceptance criteria used to assess the adequacy of SFP l facilities assure maintenance of the necessary margins of safety. The i staff's SE addresses the three areas identified above. The current request is very similar to the first request for expansion in that it raises the l . m.~ = - ? .-.u- .m-,

21 - same issues that were raised and resolved in the first request. Whereas each aspect of this request was of course reviewed in detail, the review process and scope was very much a repeat of the first expansion. In both reviews, the established criteria have been met. With the criteria met, the necessary and intended safety margins are maintained and there is no significant. reduction in margin. The criterion used in addressing nuclear criticality considerations for the storage of spent fuel is that the neutron multiplication factor (K less than or equal to 0.95 including all uncertainties at the 95/9$ ) is ff probability confidence level. As noted.in Section 2.1 of this SE, the criterion is met for all normal and abnormal conditions for the storage of spent fuel in the proposed ) configuration. The proposed amendments, therefore, do not significantly reduce a margin of safety for criticality. The criteria used in addressing themal-hydraulic considerations for the storage of spent-fuel are the methodologies and assumptions identified in Branch Technical Position ASB 9.2 and the SRP Section 9.1.3 to assure the temperatures for the SFP do not exceed 140*F under normal conditions during reloads and not exceed 212'F (boiling) during abnormal conditions following a full core discharge. As noted 2.7 of this SE, the criteria are met for the normal third of a core reload and for the abnormal full core discharge conditions for bulk-pool temperatures. The proposed amendments, therefore, do not significantly. . reduce the margin of safety for spent fuel cooling. The criteria used in addressing material, structural and mechanical considera'tions are that the compatibility and chemical stability of the materials wetted by the SFP water be demonstrated and no significant corrosion occur. The structural and mechanical design of the SFP and storage racks maintain the fuel assemblies in a safe configuration through all environmental and abnormal loadings using the codes, standards and specifications identlfied in Section 2.3.2 of the SE. As noted in Section 2.2 of this SE, the corrosion that will occur in the SFP environment will be of a little significance for the life of the ' plant and the environmental compatibility and stability of the materials used is adequate based on test data and actual service experience in operating reactors. As c noted in Section 2.3 of this SE and the Appended TER, the structural and mechanical design of the SFPs and storage racks can withstand the environmental and abnormal loading and the SFP structure can sustain the higher density floor loadings with adequate margin. The proposed. amendments, therefore, do not significantly reduce the margin of safety with regard to materials, structural, and mechanical integrity. As the result of this SE and these reviews, the proposed SFP expansions do not result in a significant reduction in a margin of safety with respect to criticality, cooling or structural considerations. 1 4 i e y, C

. Based on the foregoing, and the fact that the reracking technology in this instance has been well developed and demonstrated (100 similar applications have been approved), the Commission has concluded that the standards of 10 CFR 50.92 are satisfied. Therefore the Commission has made r final determination that the proposed amendment does not involve a significant hazards consideration. 5.0 Environmental Considerations A separate Environmental Assessment has been prepared pursuant to-10 CFR Part 51. The Notice of Issuance of Environmental Assessment and Finding of No Significant Impact was published in the Federal Register on November 16, 1984 (49 FR 45514). 6.0 Conclusion We have concluded based on the considerations discussed above, that: (1) these amendments will not (a) significantly increase the probability or consequences of accidents previously evaluated, (b) create the possibility of a new or different accident from any previously evaluated or (c) significantly reduce a margin of safety and, therefore, the amendments do not involve significant hazards considerations; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public. Dated: Movember 21, 1984 PrincipIl Contributors: DJ Mcdonald, Project Manager M. Wohl, Accident Evaluation Branch J. ',ee, Meteorology and Effluent Treatment Branch l J. Mins, Radiological Assessment Branch E. Branagan, Radiological Assessment Branch t S. Kim, Structural and Geotechnical Engineering Branch B. Turov11n, Chemical Engineering Branch i J. Ridgley, Auxiliary System Branch L. Kopp, Core Performance Branch. - R. Samworth, Environmental and Hydrologic Engineering Branch i' l-I l f se'"*. ~ ,-.-,a r ---w----.,,-,r- .a

ATTACINEh"I I TECHNICAL EVALUATION REPORT 1 i EVALUATION OF SPENT FUEL RACKS STRUCTURAL ANALYSIS FLORIDA POWER AND LIGHT COMPANY TURKEY POINT UNITS 3 AND 4 NRC DOCKET NO. 50-250, 50-251 FRC PnOJECT Casos NRC TAC NO. 54480, 54481 Pnc ASSIGNMENT 26 NAC CONTRACT NO. NRCe4t.t3D FMC TASK 529 Preparedby Franklin Reseereft Center 20th and Race Streets Philadelphie.PA 19103 FRC Group Leader: R. C. Herrick Propered for Nucieer Reguistory Conwneesson Lead NRC Ergpneer: S. 3. Kim weerungton D.C. 20555 October 25, 1984 This report wea prepared as an account of were sponsored by an agency of the United States 3 Govemment. Neither the Uruted States Govemment not any agency thereof, or any of their employoos. makes any warranty, empressed or impoled, or assumes hay legal teethlity or i reopenenbility for any third party *e use, or me results of such use. of any information. appa-russe, product or procese diecioned in this report, or represents that its use try such thire party would not intnnge prevetely owned rights. Propered by: Reviewed by: Approved by: AC4m&! _ U' ed* 7 E Principal Author Project Meneger Department Director (icting) Dete W-W- /W Date /8Md7 Date: l0/l N Y PRANKUN RESEARCH CINTER DIVISION OF ARVtN/ CAL 5 pan 20th and Race Seeses. Phda.. Ps. 19103 (215) 448 1000 4'&4 1 M /dii * ,. +r *-W" &

A _S 5 T_ TER-C5506-529 ~( 1 CONTENTS A ?& ?. s ssection Title y,, { "j -c Q-1 INTRODUCTION 1 y l.1 Purpose of the Review. 1 Y 1.2 Generic sackground. 1 i -i 2 ACCEPTANCE CRITERIA. 3 E M 2.1 Applicable Criteria 3 w r 212' Principal Acceptance Criteria. 4 3 TECENICAL REVIEW 6 Tr, 3.1 Mathematical Modeling and Seianic Asalysis of i spent Fuel Ract Modules 6 ~ y7 3.2 Evaluation of the Simplified Two-Dimensional M Nonlinear Model 10 y d' e 3.2.1 Description of the Model 10 y \\i 't3.2.2 Aaaumptions Used in the Analysis 12 "j w e N A 3.2.3 Bydrodynamic Coupling Setween fluid T and Rack Structure. 12 f Ig 3.2.4 Seismic Loading 13 2 W 3.2.5 integrationTimestep 14 A A 3.2.6 Rack Displacements 14 g 3.3 Evaluation of the Detailed Three-Dimensional Linear Model 16 3.3.1 Description of the Model 16 3.3.2 Assumptions Used in the Analysis 16 h 2" 3.3.3 Lead Correction Factor. 13 n iii 4 K 'T 5 s _A g N_ b W

w E E m m. I 9 TER-C5506-529 m =- T CONTENTS (Cont.) --( N pection Title Page h) 3.3.4 hodule Asseecly Lift-Off Analysis 18 b 3.3.5 Stress Results. 18 L f k 3.4 Review of Spent Fuel Pool Structural Analysis. 19 htt-3.4.1 Spent Fuel Pool Structural Analysis 19 h 3.4.2 Analysis Procedure. 19 l-l' 3.4.3 Summary of Results. 25 1 3 3.5 Fuel Assemely Drop Accident Analysis 26 4 CONCLUSIONS. 27 28 E 5 ItEFERENCES. r s t i ~ t, a ? E 1. 6 k ) ?,L ..s E. iv t I \\ = s 4 -L (;,' i} f = l .5 y;. a ( K

e e i TER-C5506-529 FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NBC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC. The following staff of the Franklin Research Center contributed to the technical preparation of this report: R. Clyde Herrick, Vincent K. Luk, and Balar S. Chillon (consultant). e 9 / G e V e i .syrw~i

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TER-C5506-529 1. INTRODUCTION 1.1 PURPOSE OF TEE REVIEW This technical evaluation report (TER) c) vers an independent review of the Florida Power & Light Company's licensing report (1) on high-density spent tuel racks for Turkey Point. Units 3 and 4 with respect to the evaluation of the spent fuel racks' structural analyses, the fuel racks' design, and the pool's structural analysis. The objective of this review was to determine the structural adequacy of the Licensee's high-density spent fuel racks and spent fuel pool. 1.2 n menTC BACIGROUND Many licensees have entered into a program of introducing modified fuel l racks to their spent fuel pools that will accept higher density loadings of spent fuel in order to provide additional storage capacity. However, before the higher density racks any be um A, the licensees are required to submit 1 l rigorous analysis or experimental data verifying that the structural design of the fuel rack is adequate and that the spent -fuel pool structure can accomandate the increased loads. The' analysis is complicated by the fact that the fuel racks are fully ismersed in the spent fuel pool. - During a seismic event, the water in the ' pool, as well as the rack structure, will be set in action resulting in fluid-, structure interaction. The hydrodynamic coupling between the fuel assemblies and the rack cells, as well as between adjacent racks, plays a significant role in affecting the dynamic behavior of the racks. In addition, the racks ~ are free-standing. Since the racks are not anchored to the pool floor or the pool walls, the action of the racks during a seismic event is governed by the static / dynamic friction between the rack's mounting feet and the pool floor, me by the hydrodynamic coupling to adjacrnt racks and the pool walls. e Accordingly, this report covers the review and evaluation of analyses submitted for Turkey Point Units 3 and 4 by the ?,1censee, wherein the structural analysis of the spent fuel racks under seismic loadings is of primary concern due to the nonlinearity of gap elements and static / dynamic. g g [ s

1 TER-C5506-529 friction, as well as fluid-structure interaction. In addition to,the evaluation of the dynamic structural analysis for seismic loadings, the design of the spent fuel racks and the' analysis of the spent fuel pool structure under the increased fuel load are reviewed. + r l s' l l l l I-l !l' 2-l c -. - ~ -.... . " ^ *

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. -.. ~ - t t TER-C5506-529 .i. 2. ACCEPTANCE CRITERIA 2.1 APPLICABLE CRITERIA 1 The criteria and guidelines used to determine the adequacy of the high-density spent fuel racks and pool structures are provided in the following documents: o OT Position for Review and Acceptance of Spent Fuel Storage and Bandling Applications, U.S. Nuclear Regulatory Commission, January 18, 1979 (2) o Standard Review Plan, NUREG-0800, U.S. Nuclear Regulatory Commission Section 3.7, Seismic Design Section 3.8.4, Other Category I Structures Appendix D to Section 3.8.4, Technical Position on Spent Fuel Pool Racks Section 9.1, Fuel Storage and Nandling o ASE Boiler and Pressure Vessel Code, American' Society of Mechanical 7 Enginoers Section III, subsection NF, Component Supports Subsection NB, Typical Design Rules o Begulatory Guides, U.S. Nuclear Regulatory Commission 1.29 - Seismic Design Classification 1.60 - Design Response Spectra for Seismic Design of Nuclear Power l Plants I j. 1.61 - Damping values for Seismic Design of Nuclear Power Plants 1.92 - Combining Medal Reeponses and Spatial componants in Seismic Response Analysis l 1.124 - Design Limits and Imading Combinations for Class 1 Linear-Type Component Types I o Otkr Industry Codes and Standards i American National Standards Institute, N210-76 American Society of Civil Engineers, suggested Specification for p Structures of Aluminum Alloys 6061-T6 and 6067-T6. [. L [ l I t .~ ; : 7,w. . _,... -. ~. _,... -. _.. ~.... -..

TER-C5506-529 [ 2.2 PRINCIPAL ACCEPTANCE CRITERIA The principal acceptance criteria for the evaluation of the spent fuel racks' structural analysis for Turkey Point Units 3 and 4 are set forth by the NRC's OT Position for Review and Acceptance of Sysnt Fuel Storage and Bandling Applications (M Position Paper) [2]. Section IV of the document describes the nochenical, material, and structural considerations for the fuel racks and their analysis. i The main safety function of the spent fuel pool and the fuel rects, as stated in that document, is "to maintain the spent fuel assemblies in a safe configuration through all environmental and' abnormal loadings, such as earth-quake, and impact due to spent fuel cask drop, drop of a spent fuel assembly, or drop of any other heavy object during routine spent fuel handling." Specific applicable codes and standards are defined as follows:- " Construction asterials should conform to Section III, Subsection NF of the ASM* Code. All asterials should be selected tc be compatible with the fuel pool environment to minimize corrosion and galvanic effects. Design, fabrication, and installation of spent fuel racks of stainless g steel materials any be performed based upon the AISC** specification or Subsection NF requirements of Section III of the AS M S&PV Code for Class 3 component supports. Once a code is chosen its provisions must be followed in entirety. When the AISC specification procedures are adopted, the yield stress values for staialeas steel base metal any be y = C obtained frem the Section III of the ASM B&PV Code, and the design stresses defined in the AI3C specifications as percentages of the yield - stress any be used. Permissible stresses for stainless steel welds used in accordance with the AISC Code may be obtained free Table NF-3292.1-1 of AS M Section III Code." s Criteria for seismic and impact loads are provided by Section IV-3 of the W Position Paper, which requires the fc110 wings o Seismic escitation along three orthogonal directions should be imposed simultaneously. !~ l-

• American Society of Mechanical Engineers Boiler and PrcsAre Vessel Codes, I,atest Edition.
• t American Institute of Steel Construction, Latest Edition.

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l TER-C5506-529 o The peak response from each direction should be combined by the square root of the sum of the squares. If response spectra are available for vertical and horizontal directions only, the same horizontal response spectra any be applied along the other horizontal direction. o Increased damping of fuel racks due to submergence in the spent fuel pool is not acceptable without applicable test data and/or detailed analytical results. o Local impact of a fuel assembly within a spent fuel rack cell should be considered. Temperature gradients and mechanical load combinations are to be considered in accordance with Section IV-4 of the OT Position Paper. The structural acceptance criteria are provided by Section IV-6 of the of Position Paper. For sliding, tilting, and rack impact during seismic events,

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Section IV-4 of the OT Position Paper provides the following: "For impact loading the ductility ratios utilised to absorb kinetic energy in the tensile, flemural, compressive, and shearing modes should be quantified. When considering the effects of seismic loads, factors of safety against gross sliding and overturning of racks and rack modules under all probable service conditions shall be in accordance with the Section 3.8.5.II-! of the Standard Review Plan. This position on factors of safety against sliding and tilting need not be met provided any one of the,.following conditions is mots (a) it can be shown by detailed nonlinear dynamic analyses that the amplitudes of sliding action are miniasi, and tapact between adjacent rack modules or between a rack andule and the pool walls is prevented provided that the factors of safety against tilting are within the values permitted by Section 3.8.5.II.5 of the Standard Review Plan (b) it can he shown that any sliding and tilting action will be contained within suitable geometric constraints such as thermal clearances, and that any impact due to the clearances is incorporated." .5-

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l z,' TER-C5506-529 3. TECHNICAL REVIEW 3.1 STEEMATICAL MODELING AND SEISMIC ANALYSIS OF SPENT FUEL RACK MODULES The submerged spent fuel rack modules exhibit highly nonlinear structural behavior under seismic escitation. The sources of nonlinearity can generally be categorized by the followings a. The impact between fuel cell and fuel assembly: The fuel assemely standing inside a fuel cell will impact its four inside walls repeatedly under earthquake loadings. These impacts are nonlinear in nature as.d when compounded with the hydrodynamic coupling effect will significantly affect the dynaste responses of the modules in seismic events. b. Friction between module base and pool liner: The modules are free-standing on the pool liner, i.e., they are neither anchored to the pool liner not attached to the pool well. Consequently, the andules are held in place by virtue of the frictional forces between the module base and pool liner. These frictional forces act together with the hydrodynamic coupling forces to both escite and restrain the module during seismic events. All modules at Turkey Point Units 3 and 4 have nearly square cross sections across the axes of fuel cells (1). Modules of this design geometry generally' behave in three dimensional fashion under earthquake loadings. Bence, the modules will exhibit three dimensional nonlinear structural behavior in seismic events, and all seismic analyses of modules should therefore focus on characterizing this behavior. There are tw types of modules at Turkey Point Units 3 and 4 [1]. The undules in Region I have a center-teter storage cell spacing of 10.6 in. They are reserved for temporary core off-loeding, temporary storage of new fuel, and storage of spent fuel above specified levels of reactivity. The modules in Region II, with 9.0-in center-to-center spacing, are used to store irradiated fuel below specific reactivity levels. The designs of modules in Regions I and II are shown in Figures 1 and 2, respectively. The Licensee conducted the seismic analysis of modules in two parts. The first part was a time history analysis of a simplified two-dimensional nonlinear finite element model of an individual fuel cell shown in Figure 3.,

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e TER-C5506-529 3 3. .s. 4 s A 2 3 e 3 4 1 J 3 J e 2 a a 1 a 1 3 a 1 3 3 A 1 J 8 1 j J J J J J 3 4 4 4 4 e j 3 2 s 1 i ] A J e i 2 a / 4 s 1 A 3 ./ 4 4 3 a a 7# i 1 4 i 1 / i N 3 2 - / / 4 s s / I 'l I l 0

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\\ 6 " i= W / w y-p Ah A m. d M.,/ b= / / g-.. c g; c m: W -a, ( D h Y A d 1. 'b d l I' ,g y /- r / d l f* lJf'f [. w / / A u %s% x> n / g StIFPotr FAD Y 7$ a/ 3 ////// /////// / //// / // RIGION I REGION II Figure 3. Two-Dimensional Nonlinear Model .g. 1,cy,

j 1 TER-C5506-529 The second part was a response spectrum analysis of a detailed t,hree- ] dimensional linear finite element model of a rack assembly shown in Figure 4. Both modules consisted of two models to reflect the two different designs of modules in Regions I and II. Structural damping of 24 was used in the seismic analysis for both the operating basis earthquake (UBE) and the safe shutdown earthquake (SSE). In a previous review of similar spent fuel racks, the following issue 4 concerning the modeling technique used in the analysis was discussed (3]: The simplified two-dimensional model does not fully simulate the more complicated three-dimensional structure behavior exhibited by the modules. The two-dimensional model essentially uncouples the two mutually perpendiculst horizontal motions which are nonlinearly interrelated under seismic loadings. Thus, an approach using two models (nonlinear, two-dimensional and linear, three-dimensional model) any have difficulty in resolving peak stresses. The description and evaluation of the two andels are addressed in detail in Sections 3.2 and 3.3. The displacement and stress results are discussed in appropriate subsections. l 3.2 IVAIDATION OF TEE SIMPLIFIED TWD DDWISICIIAL NONLIhEAR MODEL 3.2.1 Descriotion of the Model [. The simplified two-dimensional model was developed to simulate the major ( structural characteristics of an individual fuel cell within a submerged rack l assembly. Two versions of this model are shown in Figure 3 to reflect two "different module designs 1i Regions I and II. The model was developed in accordance, with the NBCAN (Westinghouse Electric Computer Analysis) code., A time history analysis of the andel was performed by the Licensee with the simultaneous application of a vertical and a horizontal component of seismic loads. Nonlinear gap elements were used in the model to represent the possible impact between the fuel cell and the fuel assembly, as well as the friction between the module base and the pool liner. The hydrodynamic coupling effect between fuel cell and fuel assembly, as well as between fuel cell and rigid wall, is simulated by appropriate coupling springs. A damping l i l l l ..r~ ---,---.,,...n.,-_,

TER-C5506-529 /YW 4 '7s - ,-- _.} f6 ~~ s s ~ l s s i i:- I e ,g N s g I l J. n. s I 3 , s n~ 'y, i s + -lE l l.im jp t 9/ 7-n [.) 4-y s r r. r s (- s s u ,s .,,,,y s----- s. s y. s r..,f -~1 "i s y N eV V j[ 1 ] Jr I v-s y s s l s y . s--- y. s-% s s y/ s y, N J ~ r x, r .t==._ /r a s-=... o s a d M g ,y jy '1 J ; s ,s s- % M ~~ ,,y, , s/ "g1,, e -~ ' l,'-; s , y.AP % /s .m . v w-- ,s m..V 3,,. -- a .,v e Tg . ~a j, = r=- ~ -' g .s -4 .~ ,, yl A ~ 3,; ;, -~ s, s duE j, _: 1-j lp -= as / v- ,,,...~s-- ~~ w. s ',\\--- M.. T isi. V __q -u ~. al s e. . y. m m I I ~ u f s e p "'==* y mmma r i i - l i l l REGION I RICION II l l l Figure 4. Three-Dimensional Linear Model i l 6 I l l

~ f TZR-C5506-529 value of 254 waa used to represent the impact damping of the fael, assembly [41. This impact damping value was determined from a test consisting of the fuel assembly in air impacting on a grid surface [51 3.2.2 Assumptions Used in the Analysy The following assumptions were'used in the seismic analysis of the models a. A structural desping value of 24 was used for both CBE and SSE events. b. The fluid damping was conservatively neglected. c. Only a constant value of friction coefficient was considered in each seismic analysis. The coefficient of friction remained unchanged whether the module was stationary or in action. Analysis was per-formed for static friction coefficients of u a 0.2 and 0.8. These two caseg would envelop the values of intermediate friction coefficients. d. The initial status of the gap between fuel cells and fuel assembly is immsterial because all fuel cells would move in phase soon after an earthquake occurred. Adjacent modules would also move in phase in seismic events. e. The sicshing movement of the water is in the upper elevations of the spent fuel pool sbove the top of the modules. Therefore, no sloshing loads are imposed on the module structure. . The assumption in Iten d any be valid when adjacent modules are fully loaded, but the out-of-phase response will most likely occur when some modules are either partially loaded or empty. 3.2.3 Bydrodynamic Coupiine 9etween Fluid and Cell Structure The hydrodynamic couplir.g effect between adjacent modules and between the fuel cell and fuel assembly plays a significant sole in affecting the dynamic responses of the module in seismic events. As stated in Section 3.2.2, the modules were assumed to move in phase. This assumption led to consideration of the action of an individual cell surrounded on all four sides by rigid boundaries which are separated from the cell by equivalent gaps as an equiva-lent representation of the entire cact assembly. The hydrodynamic coupling 6

e' e TER-C5506-529 mass between the raca module and the pool wall, as shown in Figure 3, was calculated by evaluating the effects of the gap between the modu'les and the pool wall using the method outlined in the paper by Frits [6]. The technique of potential flow and kinetic energy was used in assessing the hydrodynamic cousling mass between the fuel cell and the fuel assembly. This mass, which depade on the size of fuel assembly and the inside dimen-sions of the fuel cell, was calculated by equating the kinetic energy of the hydrodynamic coupling assa to that of the fluid flowing around the fuel assembly within the fuel cell. The concept of this method was discussed in a paper by De Santo (7]. Fritz's (6] method for hydrodynamic coupling is widely used and provides an estimate of the asas of fluid participating in the vibration of immersed -mass-elastic systems. Fritz's method has been validated by excellent agree-ment with experimental results (61 when employed within the c:mditions upon which it was based, that of vibratory displacements which are very small com-pared to the dimensicas of the fluid cavity. Application of Fritz's method for the evaluation of hydrodynamic coupling effects between tact andules and a pool wall has been considered by this review to serie only as an approxima-tion of,the actual hydrodynamic cocpling forces. This is because the gecoetry of a fuel rack module in its clearsace space, is considerably different than that upon which Fritz's method was developed and esperimentally verified. Thus, the limitations of Fritz's (4] modeling technique for hydrodynamic. coupling of rack modules adjacent to other rack modules or a pool wall reinforce the position of this review that the Licensee's fuel rack dynamic andel be considered conservative only 'for dynamic displacements that are small relative to the available displacement clearance. 3.2.4 Seismic Loading The andel was subject to a simultaneous application of a vertical and a horisontal component of seismic loads. The horisontal seismic loads are identical in the north-south and the east-west directions, but there are two different sets of hydrodynamic coupling masses in these two horizontal 1 l ,rp m

e a TER-C5506-529 directions. Conservative results were obtained by the Licensee by conducting one time history analysis in the horizontal direction having the acre severe hydrodynamic coupling mass. 3.2.5 Intearation Time steo The Licensee performed a time step study in an effort to find the correct integration time step to yield a converged solution (5]. It was found that the convergence of solution occurred at a time step of 0.001 see for modules in Region I and 0.005 see for modules in Region II (4]. These time steps are ~4 much greater than the 2.0x10 see reported by Gilmore of westinghouse in a i similar analysis (S]. The Licensee explained that the wide range of time steps that yield convergence any be responsible for these differing values. 3.2.6 Raet Displacements The Licensee claimed that the displacement of the module would be the same as that of the individual cell found in this model because of the in-phase action assumption used in this analysis. The Licensee found that the anximum combined seismic and thermal module displacements are 0.256 inch in Region I and 0.214 inch in Region II (5]. Both results are smaller than the nominal spacing of 1.11 inch between adjacent modules, and consequently, no collision will occur between adjacent modules. While this result may not be conservative because the twc>4imensional andel used in this analysis uncouples the two horizontal responses under seismic loadings, it does indicate that the displacements are relatively small. The detailed esck displacements are tabulated in Table 1 which is taken from the Licensee's response (5) to questions during the review. The acaents and shear forces generated from this andel were used to calculate the load correction factors. The load results from the detailed model were then multiplied by these factors to yield the stress results in the structural analysis of the module, as discussed in Section 3.3 of this report. A detailed review of this method was given in Reference 3.,

TER-C5506-529 Table 1. computed mack oisplacements l i l REClON I l REGION li SSE Seismic + Maximum Normal Thermal sac. 5eismic + Normal Thermal Max. Sliding Distance,as.2 (N. Linear Results) As in .0001 0.007 Max. Structural Deff.,a=.8 (N-Linear Results) g in .124 0.084 Total Displacement One Rock 4 = As + f A in .1241 0.093 SSRS Combined Disolacement 2 Rocks with only A max in .175 0.127 I sliding g,=g2.g 2 Max. Normal Thermal Displacement 3 in .088 0.087 7 Max. Combined Thermal & Seismie Displacements 3 in .256 0.214 d= 1 + 4 men 7 Rock to Rack Ce GAP in I.!I 1.1l l l REGION I l REGION !! SSE Seismic Slidino + Max Accident Thermal s5E seismie sijoing + Thermal Accident - Max. Sliding Distance, A =.2 As in .0001 0.007 j Max. Accident Thermal Displacement in .175 0.190 7 l Combined Thermal & Seismic Sliding 3 in .1751 0.197 A = A.' + Sr Rock to Rock Gap GAP in 1.11-1.11 ,s 2, i I i NOTE: TM RACK TO WAU. CAPS ARE LARGER THAN THE RACK TO RACK CAPS. l, r , _ _. 2 ),[,1. J.- _ __, --____.____.,_._.m.__ _,__m__,_..._

d TF.R-C5506-529 Because load correction factors based on base sceent and base shear force l were employed by the Licensee to introduce the dynamic response from the nonlinear two-dimensional dynamic displacement analysis model to the linear i-three-dimensional stress analysis, the Licensee provided a comparison of the vertical mounting pad forces in the linear and nonlinear models. Figure 5, which is taken from the Licensee response (5), shows that the summation of vertical forces in the two analysis models is reasonably close and is considered to be satisfactory. 3.3 IVALUATIcet OF TEE DETAILED TEREI-DIM NSICMAL LINEAR MODEL 3.3.1 Description of the Model i-A model was developed to simulate the major structural characteristics of the entire module submerged in the fuel pool. Two versions of the model are shown in Figure 4 to represent two different module designs in Regions I and II. The MCAN code was used to develop these two models. Three-dimensional j beam elements were used to construct the models. According to Reference 5, the seismic analysis was done on the 10x11 ? module in Region I and the 10:14 module in Region II. The model of the module in RegioE I has two fine meshes of elements, one on the top and the other on the bottom of the model to represent the top and the bottoe grip assembly of 'the module, respectively. There are eight horizontal meshes of elements in the model of the wodule in Region II to simulate the eight skip weld locations along the length of cells. A response spectrum analysis of the three-dimensional models was performed. The three components of the seismic loads were applied to the models, one component at a time. 3.3.2 Assumotions Used in the Analysis All the assumptions ezcept the initial status of the gap between fuel cell and fuel assembly used in the analysis of the two-dimensional sodel are applicable here. A few additional assumptions used in this analyeis are described below .) =~ e

Trm-c5506-529 NON LisR MODEL PAD LOADS REGION I lost i NS + DW EW + 0W 73700 72000 73700 54300 54300 54300 42000 32300 Lineer Non Lineer Linear Non Linear Total NS + DW I47400 149600 Total EW + DW I14000 117000 Total DW 108600 112800 Total DW 86600 88000 Ro*io (N5+0W)/ 1.34 1.33 Ratio (EW+0W)/ 1.32 1.33 DW DW REGION 1110x14 N5 + DW EW + DW 68200 87200 .M300 51800 62000 62000 %300 62000 Linsor Non Lineer Lineer. Non Linear Total NS + DW 155400 145300 Total ED + DW 192600 181600 Total D W l13800 101900 Total D W 124000 ll4500 Ratio (N5+0W)/DW l.37 f.43 Ratio (EW+0W)/DW l.55 1.59 Figure 5. Comparison of Mounting Pad Loads for the leonlinear an'd Linear Itack Analysis Modules - - -. -

I. TER-C5506-529 a. A composite distributive mass density was used in the analysis to embody the masses of the fuel cell, the fuel assembly, the poison material, and the hydrodynamic coupling mass, b. No impact between the fuel cell and the fuel assembly was considered. c. The module base was stationary with respect to the pool liner at all times. 3.3.3 Load Correction Factor since the detailed model did not account for the nonlinear effect of a fuel ass e ly impacting a fuel cell and the support pad movements, the internal loads and stresses for the module assembly obtained from this model were modified by load correction factors. The calculation was focused on the bending soments and shear forces obtained at the base plate of this detailed model. The bending soment load correction factor was defined as the ratio of the bending scaent obtained at the base of the simplified model to the average bending soment Serived at the base of the detailed model. Similar definition was used for the shear force load correction factor. The maximum loads from this detailed model were multiplied by these load correction factors and were used in the structural analysis to obtain the streses within the module assembly 'Further discussion is provided in Section 3.4. .3.3.4 Module Assembiv Lift-Off Analysis The modules hav$ng the largest difference between the two horisontal ~ dimensions were chosen to study the possibility of lift-off. The Ox11 module in Region I and the 9x13 andule in Region II were subject to investigation for this purpose. Both modules were found not to lift off the pool liner in seismic events (5). 3.3.5 Stress Results The annimum responses of the detailed model from the seismic components in tnree directions were combined by the SRSS model in the structural analysis. Ctresses from these responses and from dead weight are shown in Tables 2 ard 3 for Region I racks and Region !! racks, respectively. Tables,

e .e TER-C5506-527 Table 2. ' Stresses, Region I Racks RECION I RACXS

SUMMARY

OF DESIGN sir s N Ate MINIMUM MARCINS OF SAFETY Normel & Vaset Canditiore Desipi Allowshie Margin Strese Stress or f.0 Summert Pad Assembly 4.4 hosert Ped Sheer 2009 23150* 10.52 Amiel end Bending 5701 23150* 3.06 Bering 4230 23150* 4.47 IJ Suseert Ped Serow Sieur - 3675 7260 IJ2 IJ Sussert Plate Sheer 2!52 7260 3.30 WoM Sheer 15672 21000* .44 2.0 Con Assentir 2.4 cess to Bettem Crid WeW - wow Sheer 15860 23150' 46 2.2 Cell to To Grid Weld Wold Sheer 15840 23150* .46 2J Cell Amiel and Bandne JI4 1.0** .74 2.4 Cell to wresser weld WeM Steer 4517 f260 f.05 3.0 Grid Ammemhiv Al to Grie See Meneer Steer 2055 7260 3.51 AsielseW Bandng 1657 13390 7.37 3.2 To crid Meseers WoM Siser 13544 21000 J5 3.3 To Grid Outer Member Asiel and Sanding 1707 13B70 7.14 - Sher 146 9260 62.51 14 Sottom Crid Serveture '5heer 3349 7260 1.77 Amiel and Bending 12057 13890 .15 3J Setten Grid Memeers Weeds Wold Steer 15702 21000 .34 l 3.6 Bottom Grid Base Plate [ WeM i WeM Sheer 15741 21000 .32 1.0 Crid Assembly - CentM A7 mortem Grid outer Member Amiei and Sanding 19350 13870 . 15 Steer 763 7260 11.06 l 3.8 Seen Pisee Stiffener to Base Plate WeW l WeM Sheer 13530 21000 .56 i 7hermal Plus 08E Stress is Limiting r i-Allowable Per Appendix XVil-2215 Eo. (24) l i r l - j,'. ,,s +- ,-+---L- -+v--- - - ----- ' *'- -*~ - ***'"" -- ' " - - - - ' - - - - ^ ' - - - - - - - - - - - - - - -

e e TER-C5506-529 Table 3. stresses, Region II Racks REGION 2 RACKS

SUMMARY

OF DESIGN STRESW8: ABC MINIMUM MARG NS OF SAFETY Normal & Unset Conditions Desigi Allowable Margin Stress Stress of M (osi) Safety 1.1 Support Pod Shear 3S04 23150+ S.61 Axial and Bending 10288 23150* 1.25 Bearing 7631 23I50+ 2.03 1.2 Sgport Pad Screw ~t Shear 6974 9260 .33 I.3 Support Plate Shear 4403 9260 1.10 Weld Shear 16SS6 21000* .34 2.0 Cell Assembly 2.1 Cell i Axiol and Bendity .899 1.0 .' 2.2 Cell to Base Plate Wold Weld Shear 15482 21000 .36 2.3 Cell to Cell Weld Wold Sheer 18389 23150' .26 2.4 Cell Seem Weld 6 l Weld Shear 1751' 2194" .25 2.5 Cell to wrapper Weld Weld Shear 10299 18S20** .80 i Thermal Plus OBE Stress is Limiting SSE Stress is Limiting t Allowable per Appendix XVil-221S Eg (24) i l tt Design Load and Allowable Load in Lbs is Shown f t ,m. -_.,,v._,.,,,,_.w1_ ,.r, .,%,w__,.,_ ,..__,__,.___.,_m.w,,m_.,_m .,_,.,mm

e~ TER-C3506-529 2 and 3 were provided by the Licensee (5] and the support plate weld shear ~ stress and allowable stresses were subsequently changed as discussed below. Tables 2 and 3 provide the final data which were found to be acceptable during 4 the review. For Tables 2 and 3, the allowable shear stress in the weld of Item 1.3, support Plate, was changed to 21,000 poi to be in accordance with the allowable weld stress of Table NF-3292.1-1 of the Assa code.* For Table 3, the weld shear stress for Item 1.3 was changed to 16,556 psi, recognizing that i the' support plate compressive load is carried in metal-to-metal contact and is not dependent upon the weld. t 3.4 REVIEW CF SPENT FUIL 700L STRUCTURAL ANALYSIS t 3.4.1 Scent Fuel Pool Structural Analysis The spent fuel pool is a reinforced concrete plate structure supported or I compacted limerock fill. The spent fuel pool walls are lined with 1/4-sn [ stainless steel liner. The Licensee presented an analysis to demonstrate the structural integrity of the spent fuel pool for the postulated 1ruding conditions for the new high density racks. ( l 3.4.2 Analysis procedure The Licensee used the finite element method for the analysis of the spent fuel pool. The structure was modeled with three-dimensional solid elements and the AN3Ys computer code. By approsiasting syumetry along the long (north-south) diretion of the pool, only half of the pool was modeled. The boundary conditions on the plan of symmetry were adjusted to represent symmetric and non-sysmetric loading conditions. The liner plate was not considered to provide structural resistance in the pool analysis. The soil l . medium was represented by vertical compression spring elements. The thermal l l sffects were obtained by imposing a uniform thermal gradient across solid i

elements, i

• American society of Mechanical Engineers, Boiler and Pressure vessel Code, l

Section III, Division 1, subsection NF, 1980 Edition.. p } '.

• e o

TER-C5506-529 The following critical loading combinations were considered. 1. Y = 1.25 (D+P+L) with and without T 2. Y = 1.25 (D+P+L) with and without W 3. Y = 1.25 (D+P+L+E) with and without T 4 Y = 1.0 (D+P+L+E') with and without T where Y = required yield strength of the structure D = weight of the structure plus permanent loads P = hydrostatic pressure of pool water L = weight of loaded fuel racks in pool E = design earthquake load, 0.05g horizontally, 2/3 (0.059) vertically E'= maximum earthquake load. 0.15g horizontally, 2/3 (0.159) vertically T = thermal load (inside face of walla 180*F, exposed face 30'F, and bottom face of slab 50*F) W = wind load. As a result of this analysis, thn Licensee stated the following: 1. Seismic analysis for the new racks showed that these racks do not uplift during the seismic event and, therefore, no additional amplification factors for impact were considered. 2. The analysis showed that the seismic loading created a more severe l effect than the combined effect of tornado, wind, and l ,deptessurisation. t 3. The resulting stresses in the elements caused by mechanical loads i were evaluated by computing the capacities of individual sections and congiaring the capacities to the actual normal forces and soments. 4. For the combinations of mechanical and thermal loads, the sections were analysed following the approach shown in ' commentary to ACI 349-A-40.* 5. A separate analysis was conducted to determine the effects of thermal, hydrostatic, and hydrodynamic loads on the functionality of the liner. The analysis showed that there was no loes of function. ~ The results of the structural analysis were prised in the Licensee's Table A (5), reproduced here as Tables 4-a and 4-b. I i l l f i r, ,n. A. .. - ~ -, ~,. k,-------,----

=- ,.~s-. 9 Table 4-a. Spent Fuel Pool.Loed Combinations and Stresses MECHAMCAL LOAD 5 MECHAMCAI. & TERMAL 1.25ID eP eL) 1.25 (D + P + L) e E 1.25 ID e P

• L) + E e T (1)

(1) (2) (3) N M Ay, Mm/M N M M Mm/M Rober Stress g m Lacesien M/ft) K-ft/ft K-ft/ft E/ft) K-ft/ft K-ft/ft Stress Base Met 18.1 7.4 23 2.95 13.2 16.7 27 1.4 is = 12.8 ksl(5) 2.81 East Well 9.6 -22 -52 2.36 25.0 -29.3 -43 1.47 fv = 142 psl(6) 1.04 f4) 4 (Cenell (fv = 82 ps0 1.80(4) (fv = 142 psi) 1.04 (4) Y East Well 312 122 568 4.66 64.6 163 430 3.0 is = 3i.1 hel 1.03 (Pee 0 Ps = -9.6 kal ) North Well 19.8 -906 -123 f.27 13.1 -140 -151 1.08 is = 27.l ksi Ps -2.65 hel 1.33 Seuch WeII 18.9 -38.5 -192 4.99 23.0 -74.1 -182 2.39 is = 35.3 kal(7) 1.02 Ps = I.4 ksi l Miette Well 28.5 12.6 209 9.44 2.4 32.5 118 6.7 is = 9.6 ksi 3.75 i l Ps = 9.0 ksi N = Applied norneel lerce en section is = Stress in tenslen steel h l M = Applied moment en sectlen Ps = Stress in compression sleet A M a Menineen elastic moment fy = Cencrete sheer stress m u (negelive sipi indiceses congwessive seress) E S 0; i l

i' TER-C5506-529 f Table 4-b. Notes for Table 4-a t gg) Momimum elastic mornent for o section with normal force N imposed on it. g Based on a crocked analysis per the methodology discussed in Reference 2, reinfore;ng steel stress is obtoined directly. i (3) Due to the self relieving noture of thermal loods on reinforced concrete, the rotto of monimum moment copocity to actual moment cannot be uniquely determined. As e olternative, the ratio of dFy to computed reinforcing steel stress is ,dii Since structural integrity is maintoined beyond the alloweble stress for thermal loading, the octual safety factor is poster then the ratio reported. (4) Where shoor stresses control, the ratio providsid is that of allowable sheer stress (conservatively toisen as 148 psi) divided by fy. (5) This strees represents the monimum strees found in the top layer of reinforcing steel in the thinner conter section of the base mot. The top steel in this, cree is importet for teensfer of the tensile foods imposed by the laterol water pressure from the pool. The-bottom steel in the center portion of the base mot of the pool is used primarily for crock control. Since the base mot rests directly on competent fill motoriol, stresses in this bottom (secondary) steel resulting from thermal loads P Jve no adverse effect on the ability of the pool to transfer lood. Therefore, the stress in the bottom steel is.not included in Tele A. l (6) As shown in Figure 6, this section oerurs in the 3 foot wide by 18 inch thick section of the east well between the two canal wells. Becausa of the short ~ soon of this section, ed the large rotle of section thicimess to span length, the section does not resist loods in the fashim of a shallow booms shoor stresses control the section copecity. Since.Nor stirrupc are provided, the allowable shoor stress in the concrete encoeu.148 psi. The reinforcing steel, on the outside foce of this section is used only for crack control and is not needed to resist mechanical loods. Therefore, the flemurol stresses in this reinforcing steel are not included in Table A. ~ (7) This represents on overage stress (total force on the total section) over the top 10 feet of the outside fore horiaantal reinforcing steel. The result indicates that the section in general remoins below the minimum specified yield stress. However, o maimum stress of 38 lesi has been coleviated for the reinfereing steel in the top element of the well. Realizing the self-relieving nature of the thermal stresses and further orleowledging that the section in general remains elastle, pool function and structural integrity are meintoened. Additionally. is accordance with the Turket Point Uedated FSAR. Assendix SA. Section II. Limited vieldina is allowable f,8h p stovided the deflection is checked to ensure that the eJfected Class I svetens sad eesiseent are not stressed bevond their e.,lowables. No Class I systems or eeviseent are attached to this section of wall. l f t l r .__.d,- i.-

s I TER-C5506-529 2 3.4.3 susumary of Results The results of the analysis listed in Table 4-a show that the stress levels under critical loading combinations remain within the. specified allowable values, but with one exception. The review showed that: 1. The average bearing stress under the pool slab is below the allowable pressure of 10 kaf for the compacted limerock fill. 2. The maximum tensile stress in steel is shown to be 35.3 kai compared to the allowable value, Fy = 36.? kai. 3. The sheer stress in concrete controls the design in the 18-in-thick section of the east well between the two canals. The ratio of the allowable sheer stress to the maximum shear stress is shown to be 1.04. The exception to stresses within the allowable values concerns the tensile stress in the steel of the south well, which, in accordance with note i 7 of Tables 4-a and 4-b, was computed to be a maximum of 38 ksi.' For uO, in Table 4-a and for comparison to the allowable value? the Licensee averaged the maximum stresses in the steel over the upper 10 ft of well to yield an average of 35.3 kai which was compared to.he allowable value of 36 kei. Where this procedure may be questioned, the Licensee also cited Appendix SA, Section II of Turkey Point's updated FSAR which states that limited yielding is allowable und'or certain accident conditions. This was reviewed and considered to be acceptable. In addition, the Licensee's response [10] to USNBC Question No. 8 regarding the effects of 212*F water in the spent fuel pool concludes that l i stresses for the thermal load reasin within the original design allowables. For simultaneous occurrences of seismic and therasi conditions, the Licensee reported [10) that localized steel stresses were slightly higher than the allowable stress of 36 kai, and justified their angnitudes by the FSAR statement cited in the paragraph above that would permit local thermal stress yielding under certain sceident conditions. l l Af ter considering this review, evaluation showed that the 212*F pool water temperature resulted from a cooling systes pipe break during a seismic . l l l l m ,y, ,,,,-.-w --- -- a + y-pe---- w-- eev---wegr- - - * ' e. v-w-- - - -* * - - - - - - - - - - - ^ " " - - ^ - - - - - -

e e TER-C5506-529 event. Thus, considering the hours it would take to raise the pool water temperature to 212*F and increase the thermal gradient in the pool structure, the short duration seismic event would have been long past so that the j g structural considerations would remain to be those of thermal and deadweight only. The Licensee's response to usunc Question No. 8 (10) indicates that analysis showed this to be 38 kai versus the allowable value of 34 kai and was justified by statements in the FSAR as discussed above. This review concludes that the spent fuel structure is acceptable for the higher density loading. 3.5 FUEL ASSDSLY DROF ACCIDENT ANALYSIS with respect to accidental dropping of a fuel assembly, the Licensee provided the following: "In the unlikely event of dropping a fuel assembly, accidental deformation of the rack will not cause the criticality acceptance criterion to be violated. For the analysis of a dropped fuel assembly, three accident conditions are postulated. The first accident condition conservatively assumes that the weight of a fuel assemely, control rod assembly and handling

==ehanism of 3,000 pounds impacts the top and fitting of a stored fuel assembly from a drop height of 3 feet. Calculations will show that the impact energy is absorbed by the dropped fuel assembly, the stored t fuel assembly, the cells and rack base plate assembly. If in the unlikely event that two adjacent cells are crushed together for their fuel length,1:ritically, calculations show that k gg 1,0.95. Under these faulted conditions, credit is taken for dissolved bcron in the water, and the critically acceptance criterion is not violated. The second accident condition is an inclinsd drop on top of the ract. Results will be the same as for the first condition. The third accident assumes that the dropped assembly (3,000 lbs) falls straight through an empty cell.and impacts the rack base plate droe a drop height of 201 inches. The results of this analysis will show that the impact energy is absorbed by the fuel assembly and the rack base plate. Criticality calculations shown that k,gg 10.95 and the critically acceptance criterion is not violated." This statement was found to be acceptacle during the review. n e -w.. .-.----,,m-r--,,..-e----.--w,,. ..--.e-wnr,-+----. ,-.-v--e--.=cm. .3--,

TER-C5506-529 4. CONCLUSIONS Sased upon the review and evaluation, the following conclusions were reached: The lia:/.t.ations of the modeling technique empicyed for hydrodynamic o coupling of fuel assemblies within a fuel rack es11 and of fuel rack modulen to other rack modules and the pool walls indicate that the l modeling technique contributes known accuracy only for the condition in which the displacements are small compared to the available clearance space. As the Licensee's reported displacements are small, an w eptable use of the hydrodynamic coupling was employed. o Computed displacements are small relative to clearance between rack modules or between ract modules and the spent fuel pool walls. Thus, the use of two-dimensional dynamic rack module analysis was satisfactory for displacement. o while the methodology employing two dimensional nonlinear models and linear three-dimensional models correlated by load correcting factors to 1,ntroduce the nonlinear impacting load characteristics to the three-dimensional linear andel we,s not considered to be fully acceptable without further validation as a stress analysis method, a detailed step-by-step review of the stress analysis coupled with additional load tabulations requested and supplied indicates that, with the conservatisas noted to be present, the stress analysis is acceptable. o The spent fuel pool structure has design assgin to sustain the higher density floor loadings. . ~ --w, e-z-,, ,,---.,...y..,---we--. - - - -- - -. - --esg rw w-ww ,,w- -wr

O 6 I i TER-C5506-529 5. REFERENCES

1. Florida Power & Light Company Licensing Report on Turkey Point Units 3 and 4 Spent Fuel Storage Facility Modifications, Safety Analysis Report NRC Docket Nos. 50-250 and 50-251

2. OF Position for Review and Acceptance of Spent Fuel Storage and Randling Applications, U.S. Nuclear Regulatory Commission January 18, 1979

3. Franklin Research Center Technical Evaluation Report, " Evaluation of Spent Fuel Packs Structural Analysis for Duke Power Company. McGuire Nuclear Station Units 1 and 2*

August 10, 1984

4. Florida Power & Lighti Company Response to FRC's Request for Information October 5, 1984 o

5. Florida Power & Light Company Response to FRC's Questions October 1, 1984

6. R. J. Fritz "The Effect of Liquids on the Dynamic Motions of Immersed Solids

• Journal of Engineering for Industry pp. 167-173, February 1972

7. D. F. De Santo "Added Mass and Eydrodynamic Damping of Perferated Plates vibrating in Mater

• I ASM, Journal of Pressure vessel Technology I

vol. 103, p. 173, May 1981

8. C. B. Gilmore

" Seismic Analysis of Freestanding Fuel Racks" Presented at 1982 Orlando Pressure and Piping Conference

9. ASM Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NF, 1980 Edition, Table NF-3292.1-1 i

10. Florida Power & Light Company l

Response to USNRC Question No. 8 regarding the effects of a sustained pool j water temperature of 212*F on the pool and cooling systes l l t. i { l i

7590-01 UNITED STATES NUCLEAR REGULATORY COP 911SSION FLORIDA POWER AND LIGHT COMPANY DOCKET NOS. 50-250 AND 50-251 l NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING j LICENSES AND FINAL DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The U.S. Nuclear Regulatory Comission (the Comission) has issued Amendment No.111 to Facility Operating License Nos. DPR-31, and Amendment No.105 to Facility Operating License No. DPR-41, issued to Florida Power and Light Company (the licensee), which revised Technical Specifications for Operation of the Turkey Point Plant Unit Nos. 3 and 4 (the facilities) located in Dade County, Florida. The amendments are effective as of the date of + issuance and shall be implemented within 60 days of issuance. The amendments pennit the expansion of the spent fuel storage capacity for. Turkey Point Plant Units 3 and 4. This expansion would be accomplished by reracking the existing spent fuel storage pools with neutron absorbing (poison) spent fuel racks composed of individual cells made of stainless steel. Reracking the spent fuel pools would increase the Turkey Point Plant, Units 3 and 4 storage capacities from 621 to 1404 spaces for each of the units. The new fuel storage racks will be arranged in two discrete regions within each pool. Region 1 will consist of 286 locations which will normally be used for core off-loading. Region 2 will consist of 1118 locations and i will provide normal storage for spent fuel assemblies meeting required burnup considerations. The existing fuel storage racks have a nominal center-to-centerline spacing of 13.7 inches. The new Region 1 fuel storage racks will l l have a 10.6 inch centerline-to-centerline spacing and Region 2 will be 9.0 l inch centerline-to-centerline spacing. The major components of the fuel rack l l "V \\ a'** W h w"

7590-01 assemblies are the fuel assembly cell, Boraflex (neutron absorbing) material and the wrapper. The wrapper covers the Boraflex material and provides venting of the Boraflex to the pool environment. The effective multiplication factor (K,ff) of the fuel assembly array is designed to maintain the required subcriticality of K,ff equal to or less than 0.95 for both Regions 1 and 2. The transmittal letter requesting the amendments dated March 14, 1984, includes the requested Technical Specification changes, the licensee's determination on significant hazards considerations and the supporting Spent Fuel Storage Facility Analysis Report. The application for these amendments complies with the standards and i requiremerts of the Atomic Energy Act cf 1954, as amended (the Act), and the Comission's rules and regulations. The Connission has made appropriate findings as required by the Act and the Connission's rules and regulations in 10 CFR Chcpter I, which are set forth in these license amendments. Notice of Consideration of Issuance of Amendments and Proposed No Significant-Nazards Consideration Detennination and Opportunity for Hearing in connection with this action was initially published in the FEDERAL REGISTER (49 FR 23715) and in the monthly publication (49 FR 29925) on July 7, 1974 A request for a hearing was filed on July 9, 1984, by the Center for Nuclear i Responsibility, Inc. and Ms. Joette Lorton. Under its regulations, the Connission may issue and make an amendment innediately effective, notwithstanding the pendency before it of a request for a hearing from persons, in advance of the holding and completion of any required hearing, where it has detennined that no significant hazards l-consideration is involved. i I I i I ~ ~-

7590-01 The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that these amendments involve no significant hazards consideration. The basis for this determination is contained in the Safety Evaluation related to this action. Accordingly, as described above, these amendments have been issued and made innediately effective and any hearing will be held after issuance. A separate Environmental Assessment has been prepared pursuant to 10 CFR Part 51. The Notice of Issuance of Enyironmental Assessment and Finding of No ~ Significant Impact was published in the FEDERAL REJISTER (49 FR 45514) on November 16, 1084 For.further details with respect to the action see (1) the application for the amendments dated March 14, 1984, as and supplemented on July 2 and 23 August 14 and 22 September 10 and 28, October 5, 9,18 ard 26 and November 16,1984,(2) Amendment Nos.111 and 105to Facili;y Operating License Nos. DPR-31 and DPR-41 (3) the Commission's related Safety Evaluation and (4) Environmental Assessment and Notice of Issuance of Environmental Assessment - and Finding of No Significant Impact. All of these items are available for public inspection at the Commission's Public Document Room,1717 H Street, N.W., Washington, D.C., and at the Environmental and Urban Affairs Library, i j Florida International University, Miami, Florida 33199. A copy of items (2),' (3) and (4) may be obtained upon -request addressed to the U.S. Nuclear i i 6 4 - f h

7590-01 Regulatory Comission, Washington, D.C. 20555, Attention: Director, Division of Licensing Dated at Bethesda, Maryland, this 21st day of Nevember -1984. FOR THE NUCLEAR REGULATORY COMISSION David L. Wigginton, Acting Branch Chief Operating Reactors Branch No. 1 Division of Licensing 9 e W-

e s 4ssitcy / UNITED STATES

• ^-

NUCLEAR REGULATORY COMMISSION , rl,f;2 ],5 ~ 3, j WASWNGTCN. D. C. 20555 L/ a f November 14, 1984 Docket Nos. 50-250 and 50-251 Mr. J. W. Williams, Vice President Nuclear Energy Department Florida Power and Light Post Office Box 14000 Juno Beach, Florida 33408

Reference:

Technical Assignment Control Numbers 54480 and 54481

SUBJECT:

ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT - SPENT FUEL POOL EXPANSIONS, TURKEY POINT PLANT, UNITS 3 AND 4 By letter dated March 14, 1984, you requested Technical Specification amendments in support of the proposed spent fuel pool expansions at the Turkey Point Plant site. We have enclosed our Environmental Assessment related to this proposed action. Based on our assessment, we have concluded that there are no significant radiological or non-radiological impacts associated with the proposed spent fuel pool expansions and will have no significant impact on.the quality of the human environment. We have also enclosed a Notice of Issuance of Environmental Assessment and Finding of No Significant Impact. This notice is being forwarded to the Office o.f Federal Register for publication. Sincerely, nky. o 1, /u Treven A. 7trga, Chi Operating Reactors Br h #1 L Division of Licensing

Enclosures:

1. Environmental Assessment 2. Notice cc w/ enclosures: See next page ..7 .E

- J. W. Williams, Jr. _ Turkey Point Plants Flc.rier Power and Light Company Units 3 and 4 cc: Harol'd F. Reis, Eouire Administrator Newman and Holtziner P.C. Department of Environmertel 1615 L Street, N.W. Regulation hashington, DC 10036 Power Plant Siting Section State of Florida 2600 Blair Stone Road Sureau of Intergovernmental Relations Tallahassee, Florica 3E301 660 Apalachee Parkway Tallahassee, Florida 33130 James P. O'Reilly Regional Administrator, Region II Norman A. Coll, Esquire U.S Nuclear Regulatory Commission Steel, Hector and Davis Suite 2900 2000 Southeast Financial 101 Marietta Street Center Atlanta, GA 30303 Miami, Florida 33131-2398 Martin H. Hodder, Esquire 1131 N.E. 86th Street Nr. Ken N. Harris, Vice President Miami, Florida 33138 Turkey Point Nuclear Plant Florida Power and Light Company Joette Lorion P.O. Box 013100 7269 SW 54 Avenue Miami, Florida 33101 Miami, Florida 33143 Mr. M. R. Stierheim Mr. Chris J. Baker, Plant itanager County Manager of Metropolitan Turkey Point Nuclear Plant Dade County Florida Power and Light Compary Miami, Florida. 33130 P.O. Box 013100 Miami, Florida 33101 Resident Inspector Turkey Point Nuclear Generating Station Attorney General U.S. Nuclear Regulatory Commission Department of Legal Affairs Post Office Box 57-1185 The Capitol Miami, Florida 33257-1185 Tallahassee, Florida 32304 Regional Radiation Representative Mr. Ulray Clark, Administrator EPA Pegion IV Radiological Health Services 345 Courtland Street, N.W. Department of Health and Atlanta, GA 30308 Rehabilitative Services 1323 Wineweed Blvd. Mr. Jack Shreve Tallahassee, Florida 32301 Office of the Public Counsel Room 4, Holland Building Tallahassee, Florida 32304 t 's b 4 l 4

j 1 1 i f e Environmental Assessirent 4 By The Office of Nuclear Reactor Regulation Relating to Expansion of Spent Fuel Pools Facility Operating License Nos. OPR-31 and 41 Florida Power and Light Company Turkey Point Plant Units Nos. 3 and 4 Docket Nos. 50-250 and 50-251 e l 4 d' dm O 3 y p nV JT " " _

c. \\ s t, TABLE OF CONTENTS 1.0 INTROCUCTION 1.1 Identification of Proposed Action 1.2 -Neec for Increased Storage Capacity 1.3 Alternatives 1.4 Fuel Reprocessing History 2.0 FACILITY 2.1 Spent Fuel Pools 2.2 Radioactive Waste Treatment Systems i l 3.0 ENVIR0!4 MENTAL IMPACTS OF THE PROPOSED ACTION 3.1 Introduction 3.2 Radiation Exposure 3.2.1 Occupational Exposure 3.2.2 Public Exposure 3.3 Radioactive Material Released to the Atmosphere 3.4 Solid Radioactive Wastes 3.5 Radioactive Released to Receiving Waters i 4.0 NON-RADIOLOGICAL IMPACT 5.0

SUMMARY

5.1 Alternative Use of Resources 5.2 Agencies and Persons Consulted 6.0 BASIS AND CONCLUSION FOR NOT PREPARING j AN ENVIRONMENTAL IMPACT STATEMLNI e, e .y

~ , 4 1.0 IriTRODUCTION 1.1 Idertificaticn of Proposed Action The amendments would pemit the increase in the licensed storage capacity from 621 spent fuel assemblies to 1404 spent fuel assemblies for each of the 1 two Turkey Point spent fuel pools. This would extend the full core discharge capability for each generating unit from the 1990-91 time frace to the year 2005 for Unit 4 and the year 20015 for Unit 3. 1.2 Need For increased Storace Capacity When originally licensed, the SFPs for each of the Turkey Point units had the capacity to hold 217 fuel assen611es. This represented the requirement for one refueling of each unit with reserve capacity to receive a full core. At that time it was expected that the spent fuel would be removed from the site. By letter dated March 17, 1977, NRC approved amendments to the Turkey Point Licenses to allow modifying the fuel pool racks to accormadate 621 fuel assemblies. The current rack configuration will be adequate to retain the reserve capacity for full core unloading until about 1986. Since this date is earlier than the date a federal depository is expected to be available for spent fuel (1998 - Nuclear Waste Policy Act of 1982, Section 302(a)(5)] the proposed rack modifications are essential to allow continued operation beyond that 1986. This current application is to expand the storage capacity of the SFP for each unit to accommodate 1404 assemblies. I The additional SFP capacity is achieved by removing the racks not in the fuel pools and installing new racks which can accomodate a greater number of assemblies by reducing the distance between adjacent assemblies. The net a result is that after 1986 the older spent fuel assemblies ranging in age-out-of-reactor up to 13 years can be lef t in the fuel pool while newly spent fuel, assemblies are added. 4 1.3 Alternatives j Comercial reprocessing of spent fuel has not developed as had been originally antit.ipated. In 1975 the Nuclear Regulatory Comission directed the staff to prepare a Generic Environmental Impact Statement (GEIS, the j Statement) on spent fuel storage. The Comission directed the staff to analyze alternatives for the handling and storage of spent light water power e reactor fuel with particular emphasis on 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 plant shutdown. ~ A final Generic Environmental Impact Statement on Handling and Storage of i Spent Light Water Power Reactor Fuel (NUREG-0575), Volumes 1-3 (the FGEIS) was issued by the NRC in August 1979. The finding of the FGEIS is that the environmental impact costs of interim storage are essentially negligible, regardless of where such spent fuel is stored. A comparison 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 9 a + g

4 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 nomal lifetime makes this aIternative uneconomical. 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. One spent fuel storage alternative censidered in detail in the FGEIS is the expansion of onsite fuel storage capacity by modification of the existirg spent fuel pools. Applications for approximately 108 spent fuel pool capacity increases have been received and over 100 have been approved. The remaining ones are still under review. The finding in each case has been that the environmental impact of such increased storage capacity is degligible. However, since there are variations in storage designs and limitations caused by the spent fuel already stored in some of the pools, the FGEIS reconnends that licensing reviews be done on a case-by-case basis to resolve plant-specific concerns. This Environmental Assessment (EA) addresses only the specific concerns related to the proposed expansion of the Turkey Point SFPs. The environmental impacts associated with the operation of the Turkey Point Plant were evaluated in the NRCs Final Environmental Statement (FES) dated July 1972. 1.4 Fuel Reprocessina History Currently, spent fuel is not being reprocessed on a connercial basis in the United States. The Nuclear Fuel Services (NFS) plant at West Valley, New York, was shut down in 1972 for alterations and expansion; in September 1976 NFS infomed the Connission that it was withdrawing from the nuclear fuel reprocessing business. The Allied General Nuclear Services (AGNS) proposed plant in Barnwell; South Carolina, is not ;icensed to operate. On April 17, 1977, President Carter issued a policy statement on connercial' reprocessing of spent nuclear fuel which effectively eliminated reprocessing as part of the.relatively near term nuclear fuel cycle. The General Electric Company (GE) Morris Operation (forsnerly Midwest Recovery Plant) in Morris, Illinois, is in a deconnissioned condition. Although no plants are licensed for reprocessing fuel, the storage pools at Morris and at West Valley are licensed to store spent fuel'. The storage pool at West Valley is not full, but the licensee

• is presently not accepting any i

additional spent fuel for storage, even from those power generating facilities I that had contractual arrangements with West Valley.** On May 4, 1982, the license held by GE for spent fuel storage activities at its Morris operation l

• The current licensee is New York Energy Research and Development Authority.

l

• • In fact, spent fuel is being removed from NFS and raturned to various utilities.

4 s '+

. was renewed for another 20 years; however, GE is conritted to accept only limiteo quantities of additional spent fuel for storage at this facility from Cocper and San Onofre Unit 1. 2.0 FACILITY The principal features of spent fuel storage at the Turkey Point Plant, as they relate to this action, are briefly described here as an aid in following the evaluation in subsequent sections of this EA. 2.1 Spent Fuel Pools Spent fuel assemblies are radioactive due to their fresh fission product content when initially removed from the reactor core; also, they have a high themal output. The SFPs are designed for storage of these assemblies to allow for radioactive and themal decay prior to shipment. Space permitting, the assemblies may be stored for longer periods, allowing continued fission proouct decay and thermal cooling. The walls and floor of the spent fuel pit are lined with a 1/4-inch-thick stainless steel liner. Monitoring trenches are provided behind the liner for detecting and collecting any leakage. Any leakage is directed to the waste disposal drainage system, thus preventing uncontrolled leakage of SFP water. Each SFP cooling loop consists of a pump, heat exchanger, filter, demineralizer, piping, and associated valves and instrumentation. The pump draws water from the SFP pit, circulates it through the heat exchanger, and returns it to the pit. Component Cooling Water cools the heat exchanger. Redundancy of this equipment is not required because of the large heat capacity of the pit and its corresponding slow heat-up rate. Nonetheless, a 100-percent-capacity spare pump which is pemanently piped into the SFP cooling system has been installed. This pump is capable of operating in place of the originally installed pump, but not in parallel with the originally installed pump. Also, alternate connections are provided for connecting a temporary ~ ump to the spent fuel pit loop. p 2.2 Radioactive Waste Treatinent Systems The plant contains radioactive waste treatment systems designed to collect and process the gaseous, liquid and solid waste that might contain radioactive material. The radioactive waste treatment systems are evaluated. in the Final Environmental Statement (FES) dated July 1972. There will be.no change in the waste treatment systems described in the FES because of the proposed SFP expansions for Units Nos. 3 and 4. 3.0 ENVIRONMENTAL IMPACTS OF THE PROPOSED ACTION 3.1 Introduction The potential radiological environmental impacts associated with the expansion of the spent fuel storage capacities were evaluated and determined to be environmentally insignificant as addressed below.

i Curing the storage of the spent fuel under water, both volatile and non-volatile radioactive nuclides may be released to the water from the surface of the assefeblies or from defects in the fuel cladding. Post of the material released from the surface of the assemblies consists of activated corrosion products such as Co-58, Co-60, Fe-59 and Mn-54 which are not volatile. The radionuclides that might be released to the water through defects in the cladding, such as Cs-134, Cs-137, Sr-89 and Sr-90 are also predominantly non-volatil?. The primary impact of such non-volatile radioactive nuclides is their contribution to radiation levels to which workers in and near the SFPs would be exposed. The. volatile fission croouct 4 nuclides of most concern that might be released through defects in the fuel ,l cladding are the noble gases (xenon and krypton), tritium and the iodine isotopes. j Experience indicates, however, that there is little radionuclide leakage from spent fuel stored in pools after the fuel has cooled for several menths, i-The predominance of radionuclides in the SFP water appear to be radionuclides becomes mixed with water in the SFP during refueling operations) g (which that were present in the reactor coolant system prior to refuelin or crud l dislodged from the surface of the spent fuel during transfer from the reactor core to the SFP. During and afte'r refueling, the SFP purification system reduces the radioactivity concentration considerably. It is theorized that most failed fuel contains small, pinhole-like perforations in the fuel cladding at reactor i operating conditions of approximately 800'F. A few weeks after refueling, the spent fuel is cooled in the SFP and the fuel clad temperature becomes relatively cool, approximately 180*F. This substantial temperature reduction l should reduce the rate of release of fission products from the fuel pellets l and decrease the gas pressure in the gap between pellets and clad, thereby tending to retain,the fission products within the gap. Iri addition, most of the gaseous fission products have short half-lives and decay to insignificant levels withir) a few months. Based on the operational reports submitted by the licensees and discussions with the operators, there has not been any significant leakage of fission products from spent fuel' stored in the Morris 4 Operation (fonnerly Midwest Recovery Plant) at Morris, Illinois, or at the Nuclear Fuel Services (NFS) storage pool at West Valley, New York. Some spent i fuel asser@ lies which had significant leakage while in operating reactors have been stored in these two pools. After storage in the onsite SFPs, these fuel assemblies were later shipped to either Morris.0peration or NFS for extended storage. Although the fuel exhibited significant leakage at reactor operating I conditions, there was no significant leakage from these fuel assemblies in the l-offsite storage facility. 3.2 Radiation Exposure 3.2.1 Occupational Exposure i taking part in the Turkey Point Unit 3 and 4 spent fuel pool (SFP) y workers k The licensee has estimated that the radiation doses incurred b l modifications will be about 60 person-rems. This represents about a 7'; j increase in the average annual dose from routine occupational radiation i i l

= ~ '.7 exposure at the plant which was about 870 person-rems / year /ur.it over the five-year period 1978-1982 (NUREG-0713, Vol 4, December 1983). i Additionally, we have estimated the increment in onsite oc'cupational cose during normal operations after the pool ~ modifications resulting frcm the proposed increase in stored fuel assemblies. This estimate is bastd on infonnation supplied by the licensee, relevant assumptions for occupancy times i and for dose rates in the spent fuel area from radionuclide concentretiens in the water of the SFPs. The spent fuel assemblies themselves centribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel. Based on present.and projected operations in the SFP area, we estimate that the proposed modification should add less than ene percent of the total annual occupational radiation exposure at both units. The small increase in radiation exposure should not affect the licensee's abil!ty to maintain individual occupational doses to as low as is reasonably achievable (ALARA) levels 'and within the limits of 10 CFR Part 20. Thus, we + conclude that storing additional fuel in the two pools will not result in any significant increase in doses received by workers. i 3.2.2 Public Exposure The staff has completed an analysis of radiation exposure experience, based on estimated source tems and assessment of public doses resulting from 38 prior spent fuel rool modifications at 37 plants. Estimated doses to a hypothetical maximally exposed individual at the boundary of a plant site, during such modifications, have fallen within a range from 0.00004 to 0.1 millirem per year, with an average dose of 0.02 i millirem per year. Similarly, estimated tot &l doses to the population within 1-a 50-mile radius of these plants have fallen within a range frem 0.0001 to 0.. person-rem per year', with an average population dose of 0.006 person-rem per year. Doses at these levels are essentially unmeasurable. Based on the manner in which the licensee will perform the modification; their radiation protection /a.s low as reasonably achievable (ALARA) program; the radiation protection measures proposed for the modification task, including radiation, contamination, and airborne radioactivity monitoring; and i relevant experience from other operating reactors that have perfomed similar i SFP modifications, the staff concludes that adequate radiation protection measures have been taken to assure worker protection, and the Turkey Point SFP modifications can be perfonned in a manner that will ensure that doses to workers and the general public will be ALARA. Based on this review of historical data relating to the storage of spent fuel, we conclude that for the proposed SFP expansions at Turkey Point, the additional dose to the total body that might be received by an individual at the site boundary, and by the population within a 50-mile rad!us, retpectively, would be less than or equal to 0.1 millirem and 0.1 person-rem per year, respectively. These doses are very small compared to annual exposure to natural background radiation in the United States, which varies from about 70 millf rems per year to about 300 millf rems per year decending on geographical location. (

Reference:

" Natural Radiation Exposure in the Unitec

p 8-States," Donald T. Oakley, U.S. Environmental Protection Agency, Office of L Radiation Programs (ORP/SID 72-1, June 1972). 3.3 Radioactive Material Released to the Atmosphere As of February 1984, the Unit No. 3 SFP contained 372. spent fuel i assemblies. The Unit No. 4 SFP contained 313 spent fuel asser+11es and one i new fuel assembly. The current usable storage capacities for tpent fuel assemblies are 621 and 614 for Unit Nos. 3 and 4, respectively. The. proposed amendments will increase the licensed storage capacity to 1404 fuel asserr.blies for each unit. Fifty-two (52) to sixty eight (68) fuel assemblies are expected to be added to the 'SFPs following each refueling. Since space must be reserved to a9connodate a complete reactor core unloading operation (nonnally 157 fuel assemblies), the useful pool capacities are 875 and 934 i fuel assemblies for Unit Nos. 3 and 4. respectively, with the proposed modification. At an input of 52 to 68 spent fuel assemblies per refueling operation (17 months), adequate storage capacity will be available for approximately 20 years. With respect to releases of gaseous materials to the atmosphere, the only j radioactive gas of significance which could be attributable to storing additional. spent fuel assemblies for a longer period of time would be the noble gas radionuclide Krypton-85 (Kr-85). Experience has demonstrated that after spent fuel has decayed 4 to 6 months, there is no longer a significant release of fission products, including Kr-85, from stored spent fuel containing cladding defects. To determine the average annual release of Kr-85, we assumed that all the Xr-85 released from any defective fuel discharged to the SFPs will be released prior to the next refueling. The assumption of prompt release 1:; conservative and nlaximizes the amount of Kr-85 to be released. The enlargec capacities of the pools have negligible effect on calculated average annual quantities, of Kr-85 released to the atmosphere each year. Iodine-131 releases from spent fuel assemblies to the SFP water will not be significantly increased because of the expansion of the fuel storace capacity since the Iodine-131 inventory in the fuel will decay to negligible levels between refuelings. Most of the tritium in the SFP water results from activation of baron and lithium in the primary coolant and this will not be affected by the proposed expanded capacity. A relatively small amount of tritium is added during reactor operation by fissioning of reactor fuel and subsequent diffusion of tritium through the fuel and the Zircaloy cladding. Tritium release from the fuel essentially occurs while the fuel is hot, that is, during operations and, to a limited extent, shortly after shutdown. Thus, expanding SFP capacities will not increase the tritium activity in the SFPs. Storing additional spent fuel assemblies is not expected to increase the bulk water temperature during normal refuelings above the 150*F used in the design analysis. Therefore, it is not expected that there will be any e a 9

= significant change in the annual release of tritium or icdine as a result of the proposed modifications from that previously evaluated.in the FES. 3.4 Solid Radioactive Wastes The concentration of radionuclides in the pool water is contrqlled by the filters and the demineralizer and decay of short-lived isotopes. The activity is highest during refueling operation when reactor ccolant water is introd.ced into the pool and decreases as the pool water is processed through the filters and domineralizer. The increase of radioactivity, if any, due to the propcsed modifications should be minor because of the capability of the cleanup system 4 to continuously remove radioactivity in the SFP water to acceptable levels. The licensee does not expect any significant increase in the amount of solid waste generated from the SFP cleanup systems due to the proposed modifications. While we agree with the licensee's conclusicn, as a conservative estimate we have assumed that the amount of solid radwaste may be increased additionally by two resin beds (120 cubic feet solidified) and four spent filter cartridges (60 cubic feet solidified) per year from both units due to the increased operation of the SFP cleanup systems. The annual average volume of solid wastes shipped offsite for burial from a typical PWR is approximately 20,000 cubic feet. If the storage of additional spent fuel does increase the amount of solid waste from the SFP cleanup systems by about 180 cubic feet per year from both units, the increase in total waste volume shipped from Turkey Point Unit Nos. 3 and 4, would be less than 1% and would not have any significant additional environmental impact. If the present spent fuel racks to be removed from the SFPs because of the proposed modification are contaminated, they may be oisposed of as low level solid waste. We have estimated that approximately 26,000 cubic feet of solid radwaste wil1~ be removed from both units becauw of the proposeo modifications. Averaged over the lifetime of both units, this would increase the total waste volume shipped from the facility by less than 2t. This will not have any significant additional environmental impact. 3.5 Radioactive Material Rel' ased to Receiving Waters e There should not be a significant increase in the liould release of l radionuclides from the plant as a result of the proposed modifications. Since the SFP cooling and cleanup systems operate as closed systems, only water originating from cleanup of SFP. floors and resin sluice water need be considered as potential sources of radioactivity. It is expected that neither the quantity nor activity of the floor cleanup water will change as a result of these modifications. The SFP domineralizer resin removes soluble radioactive materials from the SFP water. These resins are periodically sluiced with water to the spent resin storage tank. The amount of radioactivity on the SFP demineralizer resin may increase slightly due to the additional spent fuel in the pool, but the soluble radioactive material should be retained en the resins. If any radioactive material is transferred from the spent resin to the sluice water, it will be removed by the liquid radwaste system. After processing in the licuid 4 l w ,v.---.r-nw--- -,-,-.+-,=*w , - *,v-,yv -n,.----- - +, - - - - =, - - + - -r---m- -. <-- -+---

10 - radwaste system, the amount of radioactivity released to the environment as a result of the proposed modifications would be negligible. { d.0 NON-RADIOLOGICAL IMPACT The. spent fuel' storage racks that will be removed from the pool will be decontaminated and will be disposed of either as low level radioactive waste or as non-radioactive waste, depending on the effectiveness of decontamination. Because of the small quantity (less than 20 tons), this should pose no significant environmental problem. The new assemblies will be fabricated at a Westinghouse facility at Pensacola, Florida, and moved directly to the fuel pool areas for installation. Installation is not expected to impact terrestrial resources not previously disturbed during original station construction. f The only non-radiological discharge altered by the fuel pool modifications is the waste heat. The contribution of the thirteen year old and older fuel assemblies to the total station heat discharge will be 2 negligible. Heat is removed from the fuel pool by the spent fuel pit cooling system. This is a completely closed system which uses a heat exchanger to transfer the removed heat to the Component Cooling Water System. This system transfers the heat to the station cooling reservoir which also receives the waste heat from the main condensers. The licensee has conservatively estimated that the normal maximum rate of hegt rejection from egch of the two spent fuel pools will increase from 8.8 X 10 Stu/hr to 17.0 10 8tu/hr. This' i is the rate which will occur later in the station life when the pools are again filled to capacity. The total heat load to the plant closed cycle cooling canals will be increased by about 0.3 percent. Because there is no significant environmental impact attributable to the discharge of waste heat from the plant as' indicated in the FES dated July 1972 and the very small increase which will occur as a result of the fuel pool expansions, the staff finds the. impact of the additional heat loao to be negligible. The licensee has not proposed any change in the discharge of chemicals nor changes' to the National Pollutant Discharge Elimination System pennit in conjunction with the fuel pool modifications. No increase in service water usage is proposed. Therefore, we conclude that the Turkey Point Plant spent fuel pool expansion will not result in nonradiological environmental effects significantly greater or different from those alre6dy reviewed and analyzed in I the FES. 5.0

SUMMARY

The Final Generic Environmental Impact State (FGEIS) on Handling and Storage of Spent Light Water Power Reactor Fuel concluded that the environmental impact of interim storage of spent fuel was negligible and the cost of the var.ious alternatives reflects the advantage of continued generation of nuclear power with the accompanying spent fuel storage. Because of the differences in SFP designs the FGEIS recomended licensing SFP expansion on a case-by-case basis. i n -.r-, y,yO.,,,,-.._r-,-- -..,__.--.,--____.-.--.c..----.-..-%.,c.---,

. For Turkey Point Plant, the expansion of the storage capacity of the SFPs will not create any significant additional radiological effects or measurable non-radiological environmental impacts. The additional whole body dose that might be received by an individual at the site boundary is less than 0.1 millirems per year; the estimated dose to the population within a 50-mile radius is estimated to be less than 0.1 person-rems per year. These doses ate small compared to the fluctuations in the annual dose this populatton receives from exposure to background radiation. The occupational radiation dose to workers during the modification of the storage racks is estimated by the licensee to be about 60 person-rems. This is a small fraction of the total person-rems from occupational dose at the plant. The small increase in radiation dose should not affect the licensee's ability to maintain individual occupational dose within the limits of 10 CFR Part 20, and as low as reasonably achievable. 5.1 Alternative Use Of Resources This action does not involve the use of resources not previously considered in connection with the Nuclear Regulatory Comission's Final Environmental Statement dated July 1972 related to these facilities. 5.2 Agencies And Persons Consulted The NRC staff reviewed the licensee's request and did not consult other agencies or persons. F 6.0BASISANDCONCLUSIONSFORNOTPREPARINGANENVIRONMENTALIMPACTSTATEMEN] The staff has reviewed these proposed modifications to the facilities relative to the requirements set forth in 10 CFR Part 51. Based upon the environmental assessment, the staff concluded that there are no significant radiological or non-radiological impacts assochited with the proposed action and that the proposed license amendments will not haver a significant effect on the quality "of the human environment. Therefore, the Comission has determined, pursuant to 10 CFR 51.31, not to prepare an environmental impact statement for the proposed amendments. f l Dated November 14, 1984 Principal Contributors: ( D. Mcdonald, Project Manager R. Samworth, Environmental and Hydrologic Engineering Branch J. Lee, Meteorology and Effluent Treatment Branch J. Minns, Radiological Assessment Branch E. Branagan, Radiological Assessment Branch M. Wohl, Accident Evaluation Branch. . + ,._n

u nn s-7590-01 Uh!TED STATED NUCLEAR REGULATORY COMMISSION FLORIDA POWER AND LIGHT COMPANY DOCKET NOS. 50-250 AND 50-251 NOTICE OF ISSUANCE OF ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT The U.S. Nuclear Regulatory Commission (the Consission) is considering issuance of amendments to Facility Operating License Nos. DPR-31 and DPR 41, issued to Florida Power and Light Company (the licensee), for operation of the Turkey Point Pf ar.t Unit Nos. 3 and 4 located in Dade County, Florida. Identification of Proposed Action: The amendments would consist of changes to the operating licenses and Techaical Specifications (TSs) and would authoriza t.n increase of the storage capacity of both spent fuel pools (SFPs) f.or 621 fuel assemblies to 1404 fuel assemblies with enrichment.s no greater than 4.5 weight percent U-235. The amendments to the TSs are responsive to the licensee's applicatica dated March 14, 1!r84 The NRC staff has prepared an Environmental Assessment ~ of the Propo:ed Action, " Environmental Assessment By the Office of Nuclear Reactor Regulation Relating to the Modification of the Spent Fuel Storage Pools, 0 grating License Nos. DPR-31 and DPR-41, Florida Power and Light Company, Turkey Point Plant Unit Nos. 3 and 4 Docket Nos. 50-251 and 251," t da1!ad I;ovember 14, 1904, k \\ ,f., e hl L

• • h b Y A AO an n s y ="
• V

- v

e o. . 7590-01 Sumarv of Envi~nmental Assessrent: The Final Generic Environmental Impact Statement (FGEIS) on Handling and Storage of Scent Light Water Powey Reactor Fuel (NUREG-0575), Volumes 1-3, concluded that the environmental impact of interim storage of spent fue? was negligible and the cost of the various alternatives reflects the advantage of continued generation of nuclear power with the accompanying spent fuel storage. Because of the differences in SFP designs, the FGEIS recommended licensing SFP expansions on a case-by-case t basis. For Turkey Point Plant Unit Nos. 3 and 4, the expansion of the storage capacity of the SFPs will not create any significant additional radiological effects or non-radiological environmental impacts. 4 The additional whole body dose that might be received by an individual ~ at the site boundary is less than 0.1 millirem per year; the estimated dose to the population yithin a 50-mile radius is estimated to be less than 0.1 person-rem per year. These doses are small compared to the fluctuations in the annual dose this population receives from exposure to background radiation. The estimated radiation doses incurred by workers taking part in the modifications to the SFPs will be about 60 person-rems. This represents about a 7% increase in the average annual dose from routine occupational radiation exposure at the plant which was about 870 person-rems / year / unit over the five year period of 1978-1982. The only non-radiological discharge altered by the modifications to the SFPs is the waste heat. The total load to the station closed cycle cooling \\ canalswillbeincreasedby(about0.3 percent. Thus, there is no significant environmental impact attributable to the discharge waste' heat from thf station due to this very small increase. i,*

a n. ... 7590-01 FINDING OF NO SIGNIFICANT IMPACT The staff has reviewed the proposed modifications to the facilities ~ relative to the requirements set forth in 10 CFR Part 51. Based on this assessment, the staff concluces that tnere are no significant radiological or non-radiological impacts associatec with the proposed action and that the issuance of the proposed ameddments to the licenses will have no significant impact on the quality of the human environment. Therefore, pursuant to 10 CFR 51.31, an environmental impact statement need not be prepared for this action. For further details with respect to this action, see (1) the application for amendments to the Technical Specifications dated March 14, 1984 and supplemented July 23, August 22 and September 16,1984,(2) the'FGEIS on Handling and Storage of Spent Light Water Power Reactor Fuel (NUREG-0575), (3) the Final Environme.ntal Statement for Turkey Point Plant Units 3 and 4 issued July 1972, and (4) the Environmental Assessment dated November 14, 1984 These documents are available for public inspection at the Comission's Public. Document Room 1717 H Street, N.W., Washington D.C. 205!!5 and at the ' Environmental and Urban Affairs Library, Florida International University i Miami, Florida 33199. Dated at Betheser, Maryland, this 14th day of November 1984. FOR THE NUCLEAR REGULATORY C0petISSION ., ~ ~ Dominic V. Vassallo, Acting Assistant Director for Operating Reactors l Division of Licensing e + Y h

• }}