ML17355A429

Forwards Response to NRC Telcon Questions Re License Amend Request Dtd 990727,proposing Amend on one-time Basis to Modify TS 3.8.1.1 & TS 3.4.3 & 3.5.2 to Extend Allowed Outage Time for EDG from 72 H to 7 Days

ML17355A429
Person / Time
Site:	Turkey Point
Issue date:	10/04/1999
From:	Hovey R FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-99-216, NUDOCS 9910120267
Download: ML17355A429 (134)

Text

REGULAR Y INFORMATION DISTRIBUTION~ SYSTEM (RIDS)

ACCESSION NBR:9910120267 DOC.DATE: 99/10/04 NOTARIZED: NO DOCKET FACIL:50-250 Turkey Point Plant, Unit 3, Florida Power and Light C 05000250 AUTH. NAME . AUTHOR AFFILIATION HOVEY,R'iJ. Florida Power &. Light Co.

RECIP.NAME RECIPIENT AFFILIATION Records Management Branch (Document Control Desk)

SUBJECT:

Forwards response to NRC telcon questions re license amend request dtd 990727,proposing amend on one-time basis to modify TS 3.8.1.1 & TS 3.4.3,&. 3.5.2 to extend allowed outage time for EDG from 72 h to 7 days.

DISTRIBUTION CODE: A001D COPIES RECEIVED:LTR I ENCL I SIZE: I TITLE: OR Submittal: General Distribution E

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL CLAYTON,B 1 1 JABBOUR,K 1 1 SC 1 1 INTERNA : FILE CENTER 01 1 1 NRR/DSSA/SPLB 1 1 NRR/DSBA/'SR'KB 1 1 NUDOCS-ABSTRACT 1 1 OGC/RP 1 0 EXTERNAL: NOAC 1 1 NRC PDR 1 1 D

'E j"t iJ~'i 'jl,'iii ijt'IIIJ NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE LISTS DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 TOTAL NUMBER OF. COPIES REQUIRED: LTTR 10 ENCL 9

OCT 04 1999 L-99-216 PPIL 10 CFR 50.36 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington D. C. 20555 Re: Turkey Point Units 3 Docket Nos. 50-250 One Time Only Proposed License Amendment for Unit 3 Cycle 17 Emergency Diesel Generators Allowed Outage Time Extension Res onse to Re uest for Additional Information By letter L-99-162, dated July 27, 1999, Florida Power and Light Company (FPL) requested that Appendix A of Facility Operating License DPR-31 be amended on a one-time basis to modify Technical Specification (TS) 3.8.1.1, and TS 3.4.3 and 3.5.2 (conforming changes) to extend the Allowed Outage Time (AOT) for an inoperable Emergency Diesel Generator (EDG) from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days.

FPL staff participated in a telephone conference call with NRC staff to address questions regarding the above referenced license submittal. The response to these questions is attached.

In accordance with 10 CFR 50.91 (b) (1), a copy of this letter is being forwarded to the State Designee for the State of Florida.

Should there be any questions on this request, please contact us.

R. J. Hovey Vice President Turkey Point Plant SM Attachment cc: Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Florida Department of Health and Rehabilitative Services 99iOi202b7 99%004 PDR AoaCV, 0S0002S0 I P ~ ., PDR an FPL Group company

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L-99-216 Page 2 STATE OF FLORIDA )

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COUNTY OF MIAMI-DADE )

R. J. Hovey being first duly sworn, deposes and says:

That he is Vice President, Turkey Point Plant, of Florida Power and Light Company, the Licensee herein; That he has executed the foregoing document; that the stateinents made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.

R. J. Hovey STATE OF FLORIDA COUNTY OF MIAMI-DADE Subscribed and sworn to before me this day of 1999, by R. J. Hovey is personally known to me.

Quq( A-Name of Notary Public - State of Florida GHERYL A. STEV@SON i':a'- f MY GOMMSSOII CC5S+17 EXPIRES: Joe 19, 2900 4" Bonded 1lw %by Pub5c 100reOa (Frint, type or stamp Commissione arne o otary u ic

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L-99-216 Attachment Page1of5 Attachment During telephone conversations with the NRC, the staff requested additional information regarding the One Time Only Proposed License Amendment for Unit 3 Cycle 17 Emergency Diesel Generators (EDG) Allowed Outage Time (AOT) Extension submitted on July 27, 1999.

The following provides additional information to clarify the above referenced submittal:

Discuss tlie standards or processes by which the FPL Reliability Risk Assessment Group (RRAG) controlled modifications to the Turkey Point Probabilistic Safety Assessment (PSA) models in 1993, 1994, aird 1995. Do the standards or processes include 10 CFR Appendix B reqriirements or other requirements? Also, discuss the processes employed to assure quality ofthe baseline PSA including internal or external peer reviews and any follow up to the review findings.

Since the Nuclear Regulatory Commission (NRC) approved the Individual Plant Examination (IPE) on October 15, 1992, the Turkey Point PSA model was updated in 1993, 1995, and 1997. The first update to the Turkey Point PSA models was performed by Science Applications International Corporation (SAIC) contractors in 1993. For the subsequent updates, FPL adapted SAIC's processes into FPL standards (desktop procedures).

Since the approval of the IPE, the FPL RRAG has maintained the (PSA) models consistent with the current plant configuration such that they are considered "living"PSA models. The PSA models are updated for different reasons, including plant changes and modifications, procedure changes, accrual of new plant data, discovery of modeling errors, advances in PSA technology, and issuance of new industry/ PSA standards. The update process ensures that the applicable changes are implemented and documented in a timely manner to ensure that risk analyses performed in support of plant operation reflect the plant configuration, operating philosophy, and transient and component failure history. The PSA maintenance and update process is described in detail in the FPL RRAG Standard STD-R-002, PSA Update and Maintenance Procedure.

Standard STD-R-002 defines two different types of periodic updates: 1) a data analysis update, and 2) a model update. The data analysis update is performed every five years.

Model updates consist of either single or multiple PSA changes and are performed at a frequency dependent on the estimated impact of the accumulated changes. Guidelines to determine the need for a model update are provided in the standard.

The RRAG is part of the FPL Engineering department with procedures in accordance with the Engineering Department's Quality Instructions. Procedures, risk assessment documentation and associated records are controlled and retained as QA records.

~>99-2'1 6 Attachment Page 2 of 5 The original development of PSA was classified and performed as Quality-Related under the FPL 10 CFR Appendix B quality assurance program. Subsequent data updates and risk assessments were performed using PSA methods and models. The revisions and applications of the PSA models and associated databases continued to be handled as Quality-Related. This includes PSA specific procedures and follows the independent review process for all model changes and applications. Risk assessments are performed by one individual, independently reviewed by another, and approved by the Department Head or designee.

The computer soAware is also controlled and maintained (classified as Quality-Related) under the quality assurance program with procedures in accordance with the Engineering's Quality Instructions, RRAG's standard STD-R-001, PSA Sofbvare Control Procedure, provides guidance for computer sofbvare control and establishes specific requirements for the use of PSA software, the completion of the associated documentation, and directions on processing changes to software and hardware. Furthermore, it documents the RRAG policy on PSA sofbvare safety classification, 10 CFR 50.59 applicability, soAware deficiency resolution, training requirements, verification and validation requirements, control of batch files and macros, and Quality Assurance (QA) controls for PSA processes and outputs.

Standard STD-R-001 provides the policy on QA control of the PSA processes and outputs.

QA requirements for Quality-Related PSA analytical processes and output documents consist of controlling PSA software as required by Standard STD-R-001 and requiring independent review of all aspects of the model development and its Quality Related applications. Model developments and updates are documented in reports and sent to Document Control. Compliance with 10 CFR 50 Appendix B consists of: 1) controlling software used for PSA model development and for applications which are Quality-Related as defined in STD-R-001, and 2) requiring independent reviews of each subtask while developing/revising the PSA model, and of each Quality-Related application thereafter.

The Turkey Point PSA baseline model is an updated version of the original Turkey Point IPE submittal. Prior to the IPE being submitted to the NRC, a peer review was conducted by an outside contractor. Allreview findings were addressed prior to the IPE submittal to NRC. The Turkey Point IPE was submitted to the NRC on June 25, 1991. It was reviewed extensively by the NRC and NRC contractors. It received "Step 1" and "Step 2" reviews.

Following the reviews, the Turkey Point IPE was revised in 1992. FPL received the NRC Safety Evaluation Report (SER) for the Turkey Point IPE on October 15, 1992. The NRC concluded that the process used to develop the Turkey Point PSA was acceptable in meeting the intent of Generic Letter 88-20.

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L-99-216 Attachment Page 3 of 5 There have been no additional external peer reviews because the changes that were implemented are considered not extensive. However, all updates completed since the initial IPE submittal have been reviewed, independently verified, and documented via Engineering calculations and Engineering Evaluations in accordance with the FPL Engineering Department's Quality Instructions and RRAG standards. Turkey Point intends to participate in the Westinghouse Owners Group Industry Certification program to be scheduled in 2001.

2. Discuss the RRAG's software validation and verification for quality assurance. Do the procedures conform to 10 CFR 50 Appendix B requirenients or other requirements?

All programs that process PSA model inputs are verified and validated as needed. The RRAG policy on verification and validation of QA controlled/procured software, as well as the verification and validation for sofbvare and computers when used for Quality-Related applications is described in RRAG Standard STD-R-001.

Sofbvare verification is the process used to ensure the sofbvare meets the software requirement specifications. The PSA software that is procured with a QA option and is developed under a 10 CFR 50 Appendix B QA program does not require further software verification by the RRAG. However, PSA software which is not procured with a QA option can be verified by comparison of results to previously approved software.

Validation of software is performed for different conditions such as: 1) a new installation of sofbvare, 2) any new database or configuration file changes issued by the RRAG, 3) unreasonable results, 4) computer configuration (software, hardware), and 5) use of software for Quality-Related applications for the first time.

Validation requirements for each Quality Related PSA program are documented in a Software Verification/Validation Plan (SVVP) procedure. These requirements include the method of validation, the frequency of validation, the documentation required and the acceptance criteria. A SVVP procedure is submitted for each program. Actual validation benchmark problems can exercise more than one program, but a separate Software Verification/ Validation Report (SVVR) must be submitted for each program. Each SVVP procedure and SVVR is independently reviewed and then approved by the RRAG supervisor. Software validation tests both the soAware and the hardware. Validation tests are also performed following any significant change in the hardware, operating system, or program or ifthe validation period established in the SVVP procedure expires. Sample formats for the SVVP and SVVR are provided in the Engineering Quality Instruction (conforming to the pertinent 10 CFR 50 Appendix B requirements) for computer software control.

L-99-216 Attachment Page4of5

3. In the Tier 2 discussion, it is stated that on line replacement ofthe radiators willnot be scheduled during the South Florida liurricane season and that the 1999 South Florida hurricane season begins on June 1 and ends on November 30. Explain the possible intent to replace the radiators on-line during Novetnber.

There is no intent to replace the EDG radiators during the month of November 1999.

4. Additional information on compensatory actions FPL willbe taking various compensatory actions to minimize the potential for a Loss of Offsite Power event (LOOP) during the 7-day EDG outage. The potential for an external, weather-related LOOP event to occur during the proposed 7-day EDG outage willbe minimized by scheduling the radiator replacement activity outside the South Florida Hurricane season, or when no adverse weather is expected. Therefore, voluntary entry into an LCO action statement will not be scheduled when adverse weather is expected.

The stability of the offsite electrical distribution system willbe considered by notifying in advance the appropriate system personnel for the 7-day EDG outage. Specifically, the Turkey Point Work Controls department notifies the load dispatcher (approximately 6 weeks) in advance for any scheduled outages that increase the risk of system instability.

Additionally, the Turkey Point management communicates to the load dispatcher any scheduled load threatening surveillance that could impact the electrical system (part of the morning phone call with the sites and system load dispatcher). The load dispatcher may at times request Turkey Point to avoid performing any load threatening tasks that would increase the risk of creating system instability during peak load demand periods.

During the EDG radiator replacement outage the potential for LOOP events to occur willbe minimized by a) postponing the performance of any load threatening surveillance tests until after the affected EDG is returned to service and b) administratively controlling personnel access to the Turkey Point switchyard.

Ropes and appropriate signs restrict the access to the sensitive relay area. Posted signs require that the personnel that need to gain access need to contact the System Protection Department or the Nuclear Plant Supervisor. 0-ADM 701, Control of Plant Work Activities, addresses the request for access to different sensitive areas by requiring the individual directly in charge of the job in the relay area to complete the Red Sheet. The Red Sheet is a work evaluation form, 0-ADM-701 Attachment 7, which requires the approval of not only the job supervisor/ Assistant Nuclear Plant Supervisor and the Nuclear Plant Supervisor, but the approval of the Plant General Manager. The Red Sheet evaluation is performed immediately prior to commencing work on sensitive systems (including the relay area) and is valid for a designated period without permitting any substantial break in work.

L-99-216 Attachment Page 5 of 5 FPL will ensure, in accordance with O-ADM-210, On-Line Maintenance/Work Coordination, that the systems, components, and devices that depend on the redundant EDG as a source of onsite power are operable prior to removing the EDG from service. The EDG outage task activities to be performed on-line will follow guidance outlined in 0-ADM-210. This procedure provides guidance for on-line maintenance activities to ensure adequate coordination between the Operations and Maintenance departments. Furthermore, it provides instructions to ensure that the on-line maintenance is conducted in an effective, consistent manner in accordance with the operating licenses, plant procedures, and applicable regulatory requirements.

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Distri45.txt Distribution Sheet Priority: Normal

~'SS C From: Elaine Walker Action Recipients: Copies:

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OGC/RP Not Found OE Not Found NRR/DSSA/SPLB Not Found NRR/DLPM/LPD3 1 Not Found J Segal Not Found

~FI E CENTER 0 1 Not Found KCRS Not Found External Recipients:

NRC PDR Not Found Total Copies: 9 Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 993400399

Subject:

Application for amends to licenses DPR-31 & DPRP1,to modify TS 3/4.6.3,3/4.6.6 & 3/4.7

.5 re laboratory testing of nuclear grade activated charcioal Body:

PDR ADOCK 05000250 P Docket: 05000250, Notes: N/A Docket: 05000251, Notes: N/A Page 1

L-99-239 10 CFR $ 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Re: Turkey Point Units 3 & 4 Docket Nos. 50-250 & 50-251 Proposed License Amendments Laborato Testin of Nuclear Grade Activated Charcoal In accordance with 10 CFR $ 50.90, Florida Power and Light Company (FPL) requests that Appendix A of Facility Operating Licenses DPR-31 and DPR-41 be amended to modify Technical Specification (TS) 3/4.6.3, Emergency Containment Filtering System, TS 3/4.6.6, Post Accident Containment Vent System, and TS 3/4.7.5, Control Room Emergency Ventilation System. The proposed license amendments request that charcoal samples from these filter units be tested in accordance with the American Society for r

Testing and Materials (ASTM) Standard D3803-1989, Standard Test Method for Nuclear-Grade Activated Carbon.

The proposed license amendments are submitted in response to Generic Letter (GL) 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, which requires that ASTM D3803-1989 be used for testing both new and used charcoal in engineered safety feature (ESF) applications. A description of the amendments request is provided in Attachment 1. FPL has determined that the proposed license amendments do not involve a significant hazards consideration pursuant to 10 CFR $ 50;92. -.The no significant hazards determination in support of the proposed TS changes is provided in Attachm'ent 2.

Attachment 3 provides the proposed revised TS pages.

The next laboratory surveillance test for Engineered Safety Feature (ESF) charcoal filters at Turkey Point is required to be performed in March of 2000. Assuming the proposed amendments are approved I or specific enforcement discretion is granted prior to that time, FPL will conduct the charcoal surveillance tests in accordance with ASTM D3803-1989. Any replacement charcoal will also meet the 1989 ASTM standard. FPL is therefore requesting the approval of these amendments by February 14, 2000, to suppo'rt this schedule.

~l GL 99-02 states that the Staff will exercise enforcement discretion for licensees in Group 2 to eliminate unnecessary testing of charcoal samples to both ASTM D3803-1989 and the current TS testing protocol during the period of the time between issuance of the GL and approval of the TS amendment. According to the terms of GL 99-02, Turkey Point is a" Group 2'plant. In the event that the Staff does not approve the proposed license amendments by February 14, 2000, FPL hereby requests the Staff to issue a notice of enforcement discretion that excuses FPL from performing charcoal testing using the current TS testing protocol and that permits FPL to test charcoal samples using the ASTM D3803-1989 standard in accordance with the acceptance criteria presented in this submittal.

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L-99-239 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendments Laborato Testin of Nuclear Grade Activated Charcoal In accordance with 10 CFR $ 50.91(b), a copy of the proposed license amendment is being forwarded to the State Designee for the State of Florida.

The proposed license amendments have been reviewed by the Turkey Point Plant Nuclear Safety Committee and the FPL Company Nuclear Review Board.

Should there be any questions, please contact us.

Very truly yours, R. J. Hovey Vice President Turkey Point Plant SM/MG cc: Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant I'lorida Department of Health and Rehabilitative Services

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L-99-239 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendments Laborato Testin ofNuclear Grade Activated Charcoal STATE OF FLORIDA )

) ss.

COUNTY OF MIAMI-DADE )

R. J. Hovey being first duly sworn, deposes and says:

That he is Vice President, Turkey Point Plant, of Florida Power and Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.

R. J. I.Iovey Subscribed and sworn to before me this at g.Z day of II+K 1999.

Name of Notary Public (Type or Print) GHERYL A. ~@

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F. o tN tNMMese~cc~'(pNEs; JQAO 19 20EI R. J. Hovey is personally known to me.

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L-99-239 Page 1 of 9 ATTACHMENT1 DESCRIPTION OF AMFNDMKNTSRE UFST 1.0 Background and Purpose Florida Power and Light Company (FPL) requests that Appendix A of Facility Operating Licenses DPR-31 and DPR-41 be amended to modify Technical Specification (TS) 3/4.6.3, Emergency Containment Filtering System, TS 3/4.6.6, Post Accident Containment Vent System, and TS 3/4.7.5, Control Room Emergency Ventilation System in response to Generic Letter (GL) 99-02. GL 99-02, "Laboratory Testing of Nuclear-Grade Activated Charcoal," dated June 3, 1999, requests that licensees of operating power reactors amend their TS to reference either the American Society for Testing and Materials (ASTM)

Standard D3803-1989, "Standard Test Method for Nuclear-Grade Activated Carbon," or propose an alternate test protocol.

Periodic laboratory analysis of activated charcoal used in the engineered safety features (ESF) ventilation systems of nuclear power plants is required to verify its ability to remove radioiodine from air during normal operation and during postulated accident conditions. Difficulty in achieving accurate and consistent test results has been a long-standing issue with the NRC and the nuclear industry due to the sensitivity of the adsorption mechanism to variations in the process conditions. Interlaboratory comparisons conducted since the early 1980's have demonstrated that the results for these analyses can vary significantly between the various testing laboratories. This disparity has raised concerns regarding the adequacy of these analyses and the specifications used to interpret their results. The NRC staff considers the ASTM D3803-1989 to be the most accurate and most realistic protocol for testing charcoal in ESF ventilation systems because it offers the greatest assurance of accurately and consistently determining the capability of the charcoal. The staff considers that the ASTM D3803-1989 standard provides a consistent and reproducible test method for evaluating the adequacy of charcoal.

The purpose of the proposed license amendments is to adopt ASTM D3803-1989 as the protocol for conducting laboratory tests on both new and used charcoal in the emergency containment, post accident containment vent, and control room emergency ventilation system filtering units which are the affected filter units at Turkey Point.

L-99-239 Page 2 of 9 2.0 System Description The filter units affected by the proposed TS changes include the emergency containment filters, control room emergency ventilation filters, and the post accident containment ventilation filter.

Emer enc Containment Filters ECFs Each reactor at Turkey Point is provided with three ECF units located inside containment. Each unit contains a demister bank, a high efficiency particulate air (HEPA) filter bank, a charcoal filter bank and a fan. The charcoal filter bank in each ECF is comprised of 112 standard Type II tray-type adsorber cells having a nominal face velocity of 40 feet per minute (fpm) and a gas residence time of 0.25 seconds when operated at the design volumetric air flow rate of 333 cubic feet per minute (cfm). The filter units are designed to draw air from the lower levels of containment during an accident and discharge it to the upper regions of the containment building. They were installed to reduce the iodine concentration in the containment atmosphere following a maximum hypothetical accident (MEIA) such that the offsite dose at the site boundary would not exceed 10 CFR 100 guidelines.

The air filtering capacity used to satisfy the design basis is determined from the following conditions:

a) Postulated iodine release to the containment is calculated with the ORIGEN2 code using TID 14844 release fractions at a power level of 2346 MWbased on the equilibrium fission product inventory from a 24 GWD/MTU, two region, equilibrium cycle.

b) Twenty-five percent of the total core iodine inventory is available for leakage from the containment. This assumes 50% of the total core iodine is released to containment and 50% of this activity immediately plates out on the containment walls.

c) The containment leak-rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is 0.25% per day and 0.125% per day thereafter.

d) The iodine in the containment atmosphere is assumed to be comprised of 4% methyl iodide, 91%

elemental iodine and 5% particulate iodine.

Operation of two ECFs for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is credited in the offsite and control room dose analyses associated with the large break Loss of Coolant Accident (LOCA). A removal efficiency of 90% is assumed for elemental iodine. The removal efficiency for methyl iodide is assumed to be 30%.

Operation of the ECFs is also credited in the offsite dose calculation associated with a control rod ejection accident.

'L L-99-239 Page 3 of 9 Control Room Emer enc Ventilation Filters The control room HVAC charcoal filters are located in the common emergency air intake duct. They are placed into service upon detection of high radioactivity in the normal control room HVAC air intake path.

The high radioactivity signal causes isolation dampers in the normal intake duct to close and isolation dampers in the emergency air intake duct to open. An air supply fan draws a limited quantity of outside air through the charcoal filters along with air recirculated from the control room to maintain positive pressure in the control room envelope. The charcoal filter bank is comprised of 3 Type II tray-type adsorber cells to accommodate the 1000 cfm control room HVAC design flow.

Operation of the control room emergency ventilation system is credited in the dose analysis associated with the large break LOCA. A removal efficiency of 95% is assumed for both elemental iodine and methyl iodide in the dose analysis.

Post Accident Containment Ventilation AC Filter Turkey Point uses a common post accident containment vent system to facilitate controlled venting of either reactor containment building through HEPA and charcoal filters to the waste gas tanks and to the atmosphere during post-accident conditions. The system provides the primary means of controlling containment hydrogen concentration during accidents and is placed in service when the containment hydrogen concentration reaches 3.0 volume percent. Service air is used to establish a low containment pressure under these conditions and enables a controlled flow rate to be maintained through the vent and vent filters. The design flow rate for the PACV system is 55 cfm.

The PACV system uses a standard 12" x 12" x 5 7/s" charcoal filter in a bag-in/bag-out type housing. The filter is a Type IV charcoal adsorber bank containing 8 1-inch thick charcoal beds arranged in a V-Bank configuration. The filter has a nominal face velocity of 14 fpm and a gas residence time of 0.35 seconds at the 55 cfm PACV design flow rate.

3.0 Current Technical Specification Requirements TS 3.6.3 requires that three emergency containment filtering units (ECFs) be operable in Modes 1, 2, 3, and 4. If one of the required ECFs become inoperable, it must be returned to operable status within 7 days or the plant must be brought to hot standby conditions within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to cold shutdown conditions within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Various surveillance requirements are listed in Section 4.6.3 of the TS to demonstrate filter unit operability. Surveillance requirement 4.6.3b.2 specifies the charcoal testing that must be performed to demonstrate operability. Testing is required at least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following operational exposure of filters to effluents from painting, fire, or chemical release or (3) aAer every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation. The test requires verifying within 31 days aAer removal, that a laboratory analysis of a representative carbon sample obtained in accordance with applicable portions of Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and performed in accordance with ANSI N-510-1975, meets the acceptance criteria of greater than 99.9% removal of elemental iodine; and that any charcoal failing to meet this criteria be replaced with charcoal that meets or exceeds the criteria of position C.6.a of Regulatory Guide 1.52, Revision 2.

L-99-239 Page4of9 TS 3/4.7.5 requires that the control room emergency ventilation system be operable in all plant operating modes. If the system becomes inoperable in Modes 1, 2, 3, or 4, all movement of fuel in the spent fuel pool must be suspended and the system must be restored to operable status within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. Ifthe system can not be restored to operable status within the 84-hour limit, the plant must be brought to hot standby conditions within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to cold shutdown conditions within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Ifthe action applies to both units simultaneously, the units must be brought to hot standby conditions within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to cold shutdown conditions within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

If the control room emergency ventilation system becomes inoperable in Modes 5 or 6, all operations involving core alteration, movement of fuel in the spent fuel pool, or positive reactivity changes, must be suspended. This action applies to both units simultaneously.

TS 3/4.7.5 describes the various surveillance tests that must be performed to demonstrate operability of the control room emergency ventilation system. Surveillance requirement 4.7.5c specifies the charcoal testing that must be performed to demonstrate operability. Testing is required at least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following operational exposure of filters to effluents from painting, fire, or chemical release or (3) after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation. TS surveillance requirement 4.7.5c.2 requires verifying within 31 days afler removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and performed in accordance with ANSI N-510-1975, meets the acceptance criteria for methyl iodide removal efficiency of greater than or equal to 99% or the charcoal be replaced with charcoal that meets or exceeds the criteria of position C.6.a of Regulatory Guide 1.52, Revision 2.

TS 3/4.6.6 requires that the post accident containment vent (PACV) system be operable in Modes 1 and 2.

Ifthe PACV system becomes inoperable, it must be returned to operable status within 7 days or the plant must be brought to hot standby conditions within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. TS surveillance requirement 4.6.6b specifies the charcoal testing that must be performed to demonstrate operability. Testing is required at least once per 18 months or (1) afler any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following operational exposure of filters to effluents from painting, fire, or chemical release or (3) after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation or (4) after replacement of a filter. TS surveillance requirement 4.6.6b.2 requires verifying within 31 days after removal, that a laboratory analysis of a representative carbon sample performed in accordance with ANSI N-510-1975, meets the methyl iodide removal criteria of greater than or equal to 90% and that any charcoal failing to meet the criteria be replaced with charcoal that meets or exceeds the criteria of position C.6.a of Regulatory Guide 1.52, Revision 2.

L-99-239 Page 5 of 9 4.0 Design Basis Requirements and Safety Analysis Iinpact The ECFs, control room emergency ventilation filter, and PACV filter were included as engineered safety features at Turkey Point to mitigate the consequences of postulated accidents by removing radioactive material from the containment and control room atmospheres. The charcoal filters were specifically installed to remove radioactive iodine and methyl iodide from these locations and maintain post-accident doses within regulatory limits.

The design basis of the ECFs is to provide sufficient iodine removal capability from the containment atmosphere during radiological accidents to maintain offsite doses within 10 CFR 100 limits and control room doses within limits specified in Criterion 19, "Control Room," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR 50. The large break Loss of Coolant Accident (LOCA) is the most limiting design basis event for the ECFs. The amount of iodine released to the containment during a design basis LOCA is based on the assumptions provided in Atomic Energy Commission Technical Information Document TID-14844. The capacity of the system is such that the released iodine can be adsorbed by two of the three ECF Units. The operation of two ECFs is credited in the offsite and control room dose analyses associated with:

a) Large break Loss of Coolant Accident, and b) Control Rod Ejection Accident.

A removal efficiency of 90% is assumed for elemental iodine in these analyses. The removal efficiency for methyl iodide is assumed to be 30%. These removal efficiencies are based on the guidance that is provided in Table 2 of Regulatory Guide 1.52 for 2-inch thick charcoal beds designed to operate inside containment.

The design basis of the control room emergency ventilation system is to mitigate the consequences of an accident by ensuring that the control room will remain habitable during and following all credible accident conditions. General Design Criterion 19, "Control Room," contains the dose limits that must be met by the system during radiological accidents. Operation of the control room emergency ventilation system is credited in the dose analysis associated with a large break LOCA. A removal efficiency of 95%

is assumed for both elemental iodine and methyl iodide in the analysis.

The PACV system is not specifically modeled in any of the plant safety analyses. A methyl iodide removal efficiency of 90%, however, is referenced in the TS for surveillance testing purposes. The requirement was added to the TS in the early 1980's and was derived from the Westinghouse standard TS that were in place in the mid-1970's.

The adoption of ASTM D3803-1989 for laboratory analysis of the above charcoal filters does not impact the design bases of the ESF systems, alter post-accident source terms, or modify the removal efficiencies credited in the dose calculations. Although the Turkey Point accident analyses credit both elemental iodine and methyl iodide retention in the ESF filtration systems, the ASTM standard only provides a measurement of the charcoal's ability to retain methyl iodide. Testing charcoal solely for methyl iodide retention, however, is considered to provide a valid measure of the charcoal's ability to remove radioiodine in any chemical form from the attendant plant gas stream. Supplemental testing for elemental iodine retention is

L-99-239 Attachment l Page 6 of 9 not considered necessary to verify the charcoal's ability to fulfillits design basis function. This position is bolstered by the NRC contention that elemental iodine released to the containment atmosphere will be aggressively removed through the use of the containment spray system such that the only form of iodine anticipated to require treatment by the ESP charcoal filters is methyl iodide. Additionally, an elemental iodine test protocol that provides reliable and reproducible results, and provides the ability to adequately discriminate between good and bad charcoal, has not been endorsed by the NRC.

Based on the above, the proposed changes in test method and acceptance criteria do not impact the plant safety analyses.

L-99-239 Page 7 of 9 5.0 Technical Specification Change Request The following changes to TS Surveillance Requirements 4.6.3b.2, 4.7.5c.2, and 4.6.6b.2 are requested for Turkey Point Units 3 and 4. Text deletions are shown in strikeout. Proposed text additions are shown in bohl:

a) TS3/4.6.3 Emer enc ContainmentFilterin S stem:

Revise the SURVEILLANCEREQUIREMENT 4.6.3b.2 to read as follows:

"Verifying within 31 days afler removal, that a laboratory analysis of a representative carbon sample obtained in accordance with applicable portions of Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and performed in accordance with ANNA-$40-D3803-1989 at 30 'C and 95% relative humidity, meets the methyl iodide penetratioll criteria of less than 35%; and that any charcoal failing to meet this criteria be replaced with 6

stated performance requirement; and" Justification:

The requested change updates the surveillance requirement to reflect the charcoal test standard imposed by GL 99-02. Since the new test standard is based on methyl iodide penetration rather than elemental iodine removal efficiency, a conforming change is made to reflect the appropriate test agent. This includes a change in the test acceptance criteria due to the change in the parameter used to measure filter effectiveness. The existing TS measured the charcoal filter decontamination efficiency, which is a measure, in percent, of the ability of an adsorbent to remove a specific contaminant gas from an air, or gas stream under specified conditions. The proposed TS provides acceptance criteria in terms of penetration. Filter penetration represents the amount of leakage through or around, an adsorber when tested with a challenge agent of known characteristics under known conditions. Filter penetration is expressed as a percentage of the initial challenge agent concentration. The following mathematical formula for determining the appropriate penetration acceptance criteria is provided in Enclosure 2 of the GL.

[100% Methyl Iodide Efficiency in Plant Safety Analysis]

Allowable Penetration =

Safety Factor The GL enclosure notes that the staff will accept a safety factor of greater than or equal to 2 when ASTM D3803-1989 is used with 30 'C (86 'F) and 95% relative humidity (or 70% relative humidity with humidity control). Given that a methyl iodide removal efficiency of 30% was assumed for the ECFs in the LOCA and control room dose analyses, an allowable methyl iodide penetration of less than 35% has been established for the surveillance test.

L-99-239 Page 8 of 9 The ASTM standard does not include provisions for measuring the charcoal removal efficiency for elemental iodine. Consequently, any previous commitments relative to elemental iodine testing are superseded by the adoption of ASTM D3803-1989.

b) TS3/4.7.5 Control RoomEmer enc Ventilation S stem:

Revise SURVEILLANCEREQUIREMENT 4.7.5c.2 to read as follows:

"Verifying, within 31 days afler removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and analyzed per 'teria-fer-methyl-iedine reraevaWSAeieney-ef-greater-er-equal-te-9@4 ASTM D3803-1989 at 30 'C and 95% relative humidity, meets the methyl iodide penetration criteria of 1ess than 2,5% or the charcoal be replaced with charcoal that meets or exceeds the stated performance requirement eateRa-ef gd Justification:

The requested change updates the surveillance requirement to reflect the charcoal test standard imposed by GL 99-02. Since the new test standard is based on methyl iodide penetration rather than methyl iodide removal, a conforming change is made to reflect the appropriate test acceptance criteria. A maximum allowable penetration of 2.5% is established for the control room emergency filters using the equation referenced in part a) above, and a methyl iodide removal efficiency of 95%

as assumed in the safety analysis. Performing the charcoal test at a relative humidity of 95% will bound all moisture conditions expected in the filter inlet air stream.

c) TS 3/4.6.6 Post Accident Containment Vent S stem:

Revise SURVEILLANCEREQUIREMENT 4.6.6b.2 to read as follows:

"Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample performed in accordance with e-methyl-leone-reined IPN

• humidity, meets the methyl iodide penetration criteria of less than 10% and that any charcoal failing to meet the criteria be replaced with charcoal that meets or exceeds the stated Justification:

The requested change updates the surveillance requirement to reflect the charcoal test standard imposed by GL 99-02. Since the new test standard is based on methyl iodide penetration rather than methyl iodide removal, a conforming change is made to reflect the appropriate test acceptance criteria. It should be noted that the PACV system is not modeled in any of the plant accident analyses so a specific methyl iodide removal efficiency is not rigorously documented for the charcoal filter bank. In the absence of a specific analysis value, the existing TS removal efficiency is converted to "percent penetration" and used to establish the maximum allowable penetration acceptance criteria.

L-99-239 Page 9 of 9 6.0 Conclusion The proposed revision to the TS references the new test standard, and the appropriate acceptance criteria for maximum allowable methyl iodide penetration that must be met to satisfy the surveillance requirement. The penetration acceptance criteria proposed for the emergency containment filters (ECFs) and the control room emergency ventilation filter are based on the methyl iodide removal efficiencies assumed in the plant safety analysis with a safety factor of 2. A methyl iodide penetration acceptance criterion is not currently included in the ECF TS so the test requirement represents a new license commitment. Methyl iodide testing, however, is included as part of the control room charcoal filter surveillance test. The proposed revision reduces the safety factor from its current value of 5 down to a value of 2 to coincide with a reduction in the inherent inaccuracies associated with laboratory test standards.

The post accident containment vent (PACV) filter acceptance criteria for maximum allowable methyl iodide penetration included in this license amendments request is derived directly from the removal efficiency for methyl iodide that is published in the current plant TS, without a change in specification safety factor.

Testing representative samples of charcoal used in the Emergency Containment Filters, Post Accident Containment Vent, and Control Room Emergency Ventilation systems in accordance with ASTM D3803-1989 provides the most accurate and reproducible test method available for monitoring the degradation of charcoal over time. The extensive industry experience and the requested action cited in GL 99-02 provide the basis for incorporating ASTM D3803-1989 into Turkey Point's TS.

L-99-239 Page 1 of 2 ATTACHMENT2 NO SIGNIFICANTHAZARDS COiNSIDKRATIONDKTKKVlINATION The Nuclear Regulatory Commission has provided standards for determining whether a significant safety hazards consideration exists (10 CFR $ 50.92(c)). A proposed amendment to an operating license for a facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment 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. Each standard is discussed below for the proposed amendments.

Operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated, The probability of occurrence of an accident previously evaluated for Turkey Point is not altered by the proposed TS changes because no physical modifications are being made to the plant.

The proposed change requires that new and used charcoal in the plant engineered safety feature (ESF) ventilation systems be tested in accordance with ASTM D3803-1989, at a temperature of 30 'C and a relative humidity of 95%. The use of a new or different test standard to satisfy the charcoal surveillance test requirement does not change the radiological consequences of any previously evaluated accident. The adoption of the ASTM standard will, however, require that future charcoal samples from the emergency containment filters be tested for methyl iodide removal rather than elemental iodine removal as permitted by previous test protocols. The revised test method will provide a more uniform test program for the ESF filters, and will not adversely affect the filters affinity for elemental iodine removal. The adoption of the ASTM standard for laboratory analysis of the ESF charcoal does not impact the design bases of the ESF systems, alter post-accident source terms, or modify the removal efficiencies credited in the facility dose calculations.

The ASTM standard is very stringent and has been shown to provide a more reliable measure of the ability of charcoal to fulfillits intended design function, i.e., to remove radioiodine in any chemical form from the attendant plant gas stream, than previous test protocols. Consequently, the adoption of the ASTM standard for laboratory analysis of the ESF charcoal will ensure that Turkey Point is operated in a manner consistent with the licensing basis of the facility as it relates to the protection of the public and the control room operators during radiological accidents.

Based on the above, it is concluded that the proposed amendment does not involve a significant increase in the probability or consequences of any accident previously evaluated.

L-99-239 Page 2 of 2 (2) Operation of the facility in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed change does not create a new or different type of accident for Turkey Point because no physical plant changes are being made, and no compensatory measures are imposed that would create a new failure scenario. The proposed change only imposes a more stringent surveillance requirement for both new and used charcoal in the plant ESF ventilation systems.

Since no new failure modes are associated with the proposed changes, the activity does not create the possibility of a new or different kind of accident from any previously evaluated.

(3) Operation of the facility in accordance with the proposed amendments would not involve a significant reduction in a margin of safety.

The proposed license amendment adopts a more stringent standard for performing laboratory surveillance tests on both new and used charcoal in the ESF ventilation systems. Given the increased accuracy of the proposed test standard, the amendment also supports the adoption of revised acceptance criteria having a lower safety factor to the plant safety analysis limits. The composite change does not impact the design bases of the ESF systems, alter post-accident source terms, or modify the removal efficiencies credited in the facility dose calculations The margin of safety associated with operation of the ESF ventilation systems is established by the facility dose calculations and the acceptance criteria for system performance defined in 10 CFR 100 and Criterion 19 of Appendix A to 10 CFR 50. The proposed amendments will not change this acceptance criteria nor the calculated dose limits used to establish the current plant-licensing basis.

Sllnllnai'y Based on the above discussion, FPL has determined that the proposed amendments do not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (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 and therefore the proposed changes do not involve a significant safety hazards consideration as defined in 10 CFR 50.92.

L-99-239 ATTACKVIENT3 PROPOSED TECHNICALSPECIFICATION PAGES 3/4 6-15 3/4 6-20 3/4 7-17

i CONTAINMENT SYSTEMS 3/4.6.3 EMERGENCY CONTAINMENT FILTERING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.3 Three emergency containment filtering units shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one emergency containment filtering unit inoperable, restore the inoperable filter to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.6.3 Each emergency containment filtering unit shall be demonstrated OPERABLE:

a0 At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes;

b. At least once per 18 months or .(I) after any structural maintenance on the HEPA filter or charcoal'adsorber housings, or (2) following operational exposure of filters to effluents from painting, fire, or chemical release or (3) after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation by:

1) Performance of a visual inspection for foreign material and gasket deterioration, and verifying that the filtering unit satisfies the in-place penetration and bypass leakage testing

'cceptance criteria of greater than or equal to 99K removal of DOP and halogenated hydrocarbons at the system flow rate of 37,500 cfm tlOX;

~

2) Verifying within 31 days after removal, that a laboratory ana1y-sis of a representative carbon sample obtained in accordance with applicable portions of Regulatory Position C.6.b of Regula-f i to Guide 1.5 , Revision 2 ~Narch 5 and erformed in fl, y:

cooed ce w ccepCaa t at~tlC Kf t c arcoa fal ng o meet ss cntena e re ace with charco tl f t ui-d .,

that meets or e c e . an the pic ~~a.oc,& ~~K'mega

3) y ffyf t y t tt f ttryftyyy drop across the HEPA and charcoal filters of less than 6 inches water gauge during system operation when tested in accordance with ANSI N510-1975; TURKEY POINT - UNITS 3 5 4 3/4 6-15 AMENDMENT NOS. &9 AND R4

I 1

/

CONTAINMENT SYSTEMS 3/4. 6.6 POST ACCIDENT CONTAINMENT VENT SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6 A Post Accident Containment Vent System shall be OPERABLE.

APPLICABILITY: MODES 1 and 2 ~

ACTION:

With the Post Accident Containment Vent System inoperable, restore the Post Accident Containment Vent System to OPERABLE status within 7 days or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.6. 6 The Post Accident Containment Vent System shall be demonstrated OPERABLE:

a. At least once per 31 days by demonstrating system flow path operability via a system walkdown to verify that each accessible manual valve is in its correct position.

b. At least once per 18 months or (1) after any structural maintenance of the HEPA filter or charcoal adsorber housings, or (2) following operational exposure of filters to effluents from painting, fire, or chemical release in any ventilation zone communicating with the system, or (3) after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation or (4) after

~ 5 replacement of a filter by:

A visual inspection of the system for foreign materials a'nd gasket deterioration and verifying that the filter system satisfies the penetration and bypass leakage testing acceptance criteria of less than 1X for DOP and halogenated hydrocarbon tests conducted at a design flow rate of 55 cfm +10K;

2) Verifying, within 31 days after removal, that a laboratory analysis of a re resentative carbon sam le erformed in accordance th QQ uA-4o-90K and at any c arcoa ~a> ing to .

meet the criteria be re laced~ charcoal that meets o

442,.

p

~~K+~ 'X +ITIO~C{'-C)N.ITCWIQQ 4

TURKEY POINT - UNITS 3 Ec 4 3/4 6-20 AMENDMENT NOS.~37 AND ~2

R

~Q tD

1) Verifying that the air cleanup system satisfies the in-place pene-tration and bypass leakage testing acceptance criteria of greater than or equal to 99K DOP and halogenated hydrocarbon removal at a system flow ra te of 1000 c fm +10K.

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Re ulatory Guide 1.52, Revision 2, March 1978. and an zed , eels th ~et~%d+rre K E 4'-M s.-emevaMef or the l be I re laced with charcoal that meets or exceeds th

.6-.a.o-f C Reg~~ry-Gu& , and pErgc v-ma.n c~

e u.ilrevnenp

3) Verifying by a visual inspection the absence of foreign materia s an gasket deterioration.

At least once per 12 months by verifying that the pressure drop across the Add combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 1000 cfm +10%;

e. At least once per 18 months by verifying that on a Containment Phase "A" Isolation test signal the system automatically switches into the recirculation mode of operation.

TURKEY POINT - UNITS 3 & 4 3/4 7-17 ,AMENDMENT NOSPf1 AND%85

0 / e DKC 8 ISA

NQV.'3 0 1999 L-99-176 10 CFR $ 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendments Soluble Boron Credit for S ent Fuel Pool and Fresh Fuel Rack Criticalit Anal ses In accordance with 10 CFR $ 50.90, Florida Power and Light Company (FPL) requests that Appendix A of Facility Operating Licenses DPR-31 and DPR-41 be amended to modify Technical Specifications (TS) Table 3.9-1 and 5.6.1.

These proposed changes increase the subcritical margin in the Spent Fuel Pool (SFP) in order to accommodate degradation of the Boraflex panels in the fuel storage racks by permitting credit for soluble Boron. The generic methodology for crediting soluble boron in spent fuel rack criticality analysis, Westinghouse Spent Fuel Rack Criticality Analysis methodology WCAP-14416-NP-A, Revision 1, was approved by the Nuclear Regulatory Commission on October 25, 1996. The Turkey Point Units 3 and 4 specific Criticality Analyses for Fresh and Spent Fuel storage racks and the SFP Dilution Analysis are submitted herein to update the licensing bases which support the proposed TS changes.

A description of the amendments request is provided in Attachment 1. FPL has determined that the proposed license amendments do not involve a significant hazards consideration pursuant to 10 CFR $ 50.92. The no significant hazards determination in support of the proposed TS changes is provided in Attachment 2. Attachment 3 provides the proposed revised TS pages.

Attachments 4 and 5 provide the Criticality Analyses for Spent Fuel Storage for Turkey Point Units 3 and 4. Attachment 6 provides the Turkey Point Units 3 and 4 SFP Dilution Analysis.

Attachment 7 provides the Fresh Fuel Storage Criticality Analysis. Attachment 8 provides the Turkey Point Units 3 and 4 SFP monthly silica concentration data.

In accordance with 10 CFR $ 50.91(b), a copy of the proposed license amendment is being forwarded to the State Designee for the State of Florida.

~~ o>mw c>LSD an FPL Group company

0 Pl 0

L-99-176 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendments Soluble Boron Credit for S ent Fuel Pool and Fresh Fuel Rack Criticalit Anal ses The proposed license amendments have been reviewed by the Turkey Point Plant Nuclear Safety Committee and the FPL Company Nuclear Review Board.

FPL requests the review and approval of the proposed amendments by June 2000.

Should there be any questions, please contact us.

Very truly yours, R. J. Hovey Vice President Turkey Point Plant SM CC: Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant Florida Department of Health and Rehabilitative Services

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L-99-176 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendments Soluble Boron Credit for S ent Fuel Pool and Fresh Fuel Rack Criticalit Anal ses STATE OF FLORIDA )

) ss.

COUNTY OF MIAMI-DADE )

R. J. Hovey being first duly sworn, deposes and says:

That he is Vice President, Turkey Point Plant, of Florida Power and Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.

R. J. Hovey Subscribed and sworn to before me this day of ~ 1999.

Name of Notary Public (Type or Print) alCfltt

+laoseu~ I+pttg, Aq gal%<

/~iq'~> ilYCOM%SSOHf CCSH017 I) gpl+8: PN10 13'

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R. J. Hovey is personally known to me.

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Distri16.txt 0 Priority: Normal J

Distribution Sheet From: Elaine Walker Action Recipients: Copies:

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RidsNrrDssaSrxb OK RidsNrrDssaSplb OK Rids Manager OK RidsAcrsAcnwMailCenter OK OGC/RP Not Found NRR/DSSA/SRXB Not Found t

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'CRS Not Found External Recipients:

NRC PDR Not Found NOAC Not Found Total Copies: 14 item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 993410153

Subject:

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendme nts Soluble Boron Credit for Spent Fuel Pool and Fresh Fuel Rack Criticality Analyses Body:

PDR ADOCK 05000250 P Docket: 05000250, Notes: N/A Page 1

k 0

Distri16.txt Docket: 05000251, Notes: NIA Page 2

h 4 ~, er

Distri68.txt Distribution Sheet Priority: Normal From: Stefanie Fountain Action Recipients: Copies:

NRR/DLPM/LPD2-2 1 Paper Copy K Jabbour Paper Copy B Clayton Paper Copy Internal Recipients:

RidsRgn2MailCenter OK RidsOgcRp OK RidsNrrWpcMail 0 OK RidsNrrDssaSrxb 0 OK Rids Manager OK RidsAcrsAcnwMailCenter 0 OK OGC/RP Paper Copy NRR/DSSA/SRXB ILE CEN ER ACRS 01~ Paper Copy Paper Copy Paper Copy External Recipients:

NOAC Paper Copy Tota I Copies:

Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003693888:1

Subject:

Proposed License Amendments Soluble Boron Credit for Spent Fuel Pool and Fresh Fu el Rack Criticality Analyses Response to Request for Additional Information Body:

ADAMS DISTRIBUTION NOTIFICATION.

Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML003693888.

Page 1 p)e

Distri68.txt A001 - OR Submittal: General Distribution Docket: 05000250 Docket: 05000255 Page 2

NR 0 8 2000 L-2000-054 10 CFR 50.36 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington D. C. 20555 Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendments Soluble Boron Credit for Spent Fuel Pool and Fresh Fuel Rack Criticality Analyses Res onse to Re uest for Additional Information By letter L-99-176, dated November 30, 1999, Florida Power and Light Company (FPL) requested that Appendix A of Facility Operating Licenses DPR-31 and DPR-41 be amended to modify Technical Specification (TS) 3.9-1 and 5.6.1. By letter dated January 31, 2000, the NRC staff requested additional information regarding the above referenced FPL submittal.

The response to the request for additional information is provided in Attachment 1. FPL has a typographical error in Attachment 5 of L-99-176. Attachment 2 of this letter 'dentified provides the corrected report and supercedes Attachment 5 of L-99-176. FPL has determined that the additional information provided herein does not change the conclusions reached in the original no significant hazards consideration provided in FPL letter L-99-176. Attachment 3 provides the environmental consideration statement.

In accordance with 10 CFR 50.91 (b) (1), a copy of this letter is being forwarded to the State Designee for the State of Florida.

Should there be any questions on this request, please contact us.

Very truly yours, R. J. Hovey Vice President Turkey Point Plant SM Attachments cc: Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Florida Department of Health an FPL Group company

y v

f

L-2000-054 Page 2 STATE OF FLORIDA )

) ss.

COUNTY OF MIAMI-DADE )

R. J. Hovey being first duly sworn, deposes and says:

That he is Vice President, Turkey Point Plant, of Florida Power and Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.

R. J. Hovey STATE OF FLORIDA COUNTY OF MIAMI-DADE Subscribed and sworn to before me this RN day of 2000, by R. J. Hovey is personally known to me.

Name of N lorida (Print, type or stamp Commissioned Name of Notary Public)

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L-2000-054 Attachment 1 Page 1 of 3 Attachment 1 The NRC staff requested additional information regarding FPL letter L-99-176, submitted on November 30, 1999, Proposed License Amendments Soluble Boron Credit for Spent Fuel Pool and Fresh Fuel Rack Criticality Analyses. The following discussion provides the response.

~Re uest 2 The NRC staff safety evaluation report contained in @CAP-14416-NP-A presents the required technical specifications for use with tlie approved soluble boron credit methodology. The Fuel if Storage Criticality specifications in the Desigir Features Section for both k-effless than 1.0 fiilly if flooded with unborated water and for k-effless than or equal to 0.95 fiillyflooded with borated water require reference to WCAP-14416-P for a description of the uncertainties included.

Therefore, proposed Technical Specifications 5.6.1.1.a and 5.6.1.1.b should include tlie phrase "whicli includes a conservative allowance for uncertainties as described in 8'CAP-14416-P.

~Res onse 1 FPL agrees with the above recommendation and wording of proposed Technical Specifications 5.6.1.1.a and 5.6.1.1.b to include the phrase, "which includes a conservative allowance for uncertainties as described in WCAP-14416-P.'* The inclusion of this wording does not change the conclusions reached in the original no significant hazards consideration provided in FPL letter L-99-176.

~Re uest2 Please describe the administrative procedures used to select tlie appropriate asseinblies for storage in the burnup-dependent racks in Region 2.

~Res onse2 Spent fuel assemblies assignment in Region II are specified by the Reload Engineering Design Modification Package which is reviewed and approved by the plant's safety review board prior to the offload of the irradiated fuel assemblies from the core. The basis for these assignments is documented in an engineering calculation in accordance with Nuclear Engineering Department Standard STD-F-009 Revision 3, "Irradiated Fuel Storage Assignments." The requirements in this standard are in compliance with Technical Specification 3/4.9.14 regarding the storage of irradiated fuel.

L-2000-054 Attachment 1 Page2of3 At Turkey Point, the movement of fuel assemblies is controlled by Administrative Procedure 0-ADM-556, Fuel Assembly and Insert Shuffle. Guidelines in this procedure along with the designation of assemblies which satisfy the requirements for storage in Region II, are used to proceduralize the movement of each individual assembly by an assembly identification number and an alpha numeric storage location via Fuel Handling Data Sheets. The Fuel Handling Data Sheets are used by operating personnel to coordinate and track the movement of each assembly to assure that it is stored in its proper location. Control of this evolution is via headphone communication between the Control Room and the fuel handling personnel. Once in the pool, an insert shuffle is done and a camera inspection of the assemblies that are going back into the core is performed. This inspection ensures that the assemblies going back in the core have the right insert and are located in the proper storage rack. The Fuel Handling Data Sheets become Quality Assurance records.

~Re uest3 describes the criticality analysis performed with a reduced B-10 loading in the degraded Boraflex. The assuInptions in the analysis include the following:

Region 1: 0.009 glcm absorber B-10 loading and 0.0351 inch thickness Region 2: 0.006 glcm absorber B-10 loading and 0.051 inch thickness The analysis based on these assumptions results in a E~gless than 1.0 with no soluble boron. Please provide your plan to verify that the Boraflex panels have not degraded beyond the assunied thicknesses.

~Res ense 3 Contingent upon approval of the proposed license amendments, FPL plans to perform a test, in 2001, to verify the analysis assumptions for Boraflex degradation.

Currently, FPL has an on-going in-service Boraflex verification program, which consists of measuring the gap formation, gap distribution, and gap size. The program accomplishes these goals through the performance of blackness testing on a &equency of one test every five years in either Spent Fuel Pool.

Upon approval of the proposed license amendments, FPL would commit to perform a test that validates our assumption on the thickness of the Boraflex every five years beginning in the year 2001. FPL would upgrade the blackness testing with a test which willnot only measure the number of gaps and gap size but also validate our assumptions on the thickness of the Boraflex.

L-2000-054 Attachment 1 Page 3 of3 Substituting the blackness testing with an upgraded test, as well as changing the test date &om the year 2000 to 2001, would change FPL's previous commitment as documented in L-95-041, dated September 5, 1995. Upon approval of the proposed license amendments, FPL will notify the NRC by separate correspondence, of the change in commitment.

L-2000-054 Attachment 2 Attachment 2 The value of 0.0006 g/cm that is quoted on page 2 of 4 of FPL letter L-99-176, Attachment 5 (Westinghouse letter 999FP-G-012, Rev 1) is a typo and should read 0.006 g/cm . Westinghouse has corrected the typographical error and the attached report (Westinghouse letter 999FP-G-0102, Rev 2) supercedes Attachment 5 of L-99-176.

999FP-G-0102, Rev. 2 CAB-99-367, Rev. 2 WeStinghauSe EleCtriC COmpany Commercial Nuclear Fuel Division Box 355 Pittsburgh Pennsylvania 15230-0355 January 5, 2000 Mr. Jimmie L. Perryman ENG-JB Room D 4466 Florida Power 8 Light Company P. O. Box 14000 Juno Beach, Florida 33408

Reference:

1) 99FP-G-0067, dated June 15, 1999

2) 99FP-G-0071, dated July 6, 1999

Dear Mr. Perryman:

FLORIDA POWER 8 LIGHT COMPANY TURKEY POINT UNITS 3 8 4 Criticalit Anal sis with Reduced B" Loadin in the De radedBoraflex forRe ions1and2S entFuelStora e, Revision2 Attached are the results for the completed criticality analysis with the reduced Bto loading in the degraded boraflex for Turkey Point Units 3 and 4 Regions 1 and 2 spent fuel storage (no soluble boron). The methodology and assumptions used in the analysis are the same as in References 1 and 2, except that the absorber B'oading and its thickness are reduced to 0.009 g/cm'and 0.0351 inch for Region 1 and 0.006 g/cm'nd 0.051 inch (remain unchanged) for Region 2. For Region 1, the reduction of both the Bto loading and the corresponding thickness is slightly more limiting than the reduction of the Bto loading only.

For Region 2, the reduction of the Bto loading only is slightly more limiting than the reduction of the Bte loading and the corresponding thickness. The final 95/95 Keff is shown in the attached Table 1 and Table 2 for spent fuel rack Region 1 and Region 2, respectively. Since both Keffs are still less than 1.0, the Turkey Point Units 3 and 4 spent fuel racks will remain subcritical when all cells are loaded 15x15 fresh fuel assemblies with nominal enrichments no greater than 4.50 w/o U~'with natural uranium axial blankets in Region 1, and with nominal enrichments no greater than 1.60 w/o in Region 2. This meets the design basis for no soluble boron water in the pool.

This transmittal has been revised to correct the Region 2 absorber Bio loading to 0.006 g/cm'.

Please contact M. F. Muenks or me, if you have any questions or concerns about this criticality analysis.

Very truly yours, kN~~

David E. McKinnon Project Engineer Commercial Nuclear Fuel Division CC: B. Tomonto TP Site J. Garcia Juno Beach C. A. Villard Juno Beach J. R. Dwight Columbia M. F. Muenks Energy Center

/cad Attachment

99FP-G-0102, Rev. 2 CAB-99-367, Rev. 2 Criticality Analysis With a Reduced B>0 Loading in the Degraded Boraflex for Turkey Point Units 3 4 4 Region 1 and Region 2 Spent Fuel All Cell Storage (No Soluble Boron)

January, 2000 S. Srinilta (ND)

Core Analysis B Date: l 5 Z Dog Verified:

J. Seeker (ND)

Core Analy is C Date:

l of4

99FP-G-0102, Rev. 2 CAB-99-367, Rev. 2 Criticality Analysis With a Reduced B>0 Loading in the Degraded Boraflex for Turkey Point Units 3 & 4 Region 1 and Region 2 Spent Fuel All Cell Storage (No Soluble Boron)

A criticality analysis was performed with a reduced 810 loading in the degraded boraflex for Turkey Point Units 3 & 4 Region 1 and Region 2 spent fuel all cell storage (No Soluble Boron). The methodology and assumptions used in the analysis are the same as in Reference I except that the absorber B10 loading and its thickness are reduced to 0.009 g/cm2 and 0.0351 inch for Region 1 and 0.006 g/cm2 and 0.051 inch (remain unchanged) for Region 2. For Region 1, the reduction of both the B10 loading and the corresponding thickness is slightly more limiting than the reduction of the 810 loading only. For Region 2, the reduction of the 810 loading only is slightly more limiting than the reduction of the B10 loading and the corresponding thickness. The final 95/95 Keff is shown in the attached Table 1 and Table 2 for spent fuel rack Region 1 and Region 2, respectively. Since both Keff's are still less than 1.0, the Turkey Point Units 3 and 4 spent fuel racks will remain subcritical when all cells are loaded 15x15 fresh fuel assemblies with nominal enrichments no greater than 4.50 w/o U235 with natural uranium axial blankets in Region 1, and with nominal enrichments no greater than 1.60 w/o in Region 2. This meets the design basis for no soluble boron water in the pool.

Reference:

1) 99FP-6-0071 Criticality for Spent Fuel Storage for Turkey Point Units 3 &4 (Degraded Boraflex) 2 of4

~ ~

Table 1. Region 1- No Soluble Boron Base Keno Reference Reactivity 0.97155 Calculation and Methodology Biases Range Methodology (Benchmark) Bias 0.00770 Pool Temperature Bias 50 F to 185 F 0.00077 Boron Particles in Boraflex KQQZH Total Bias 0.01231 Tolerances and Uncertainties Parameter Reactivity Variation Variation Fuel Enrichment +0.05/-P P5 0.00191 Fuel Density +2/-2 % 0.00250 Fuel Pellet Dishing 1.187 % 0.00145 Rack Cell Inner Dimension +0.05/-0.025 inch 0.00153 Rack Cell Pitch +0.12/-0.12 inch 0.01022 Rack Wall Thickness +0.007/-0.007 inch 0.00024 Wrapper Plate Thickness +0.002/-0,002 inch 0.00000 Poison Panel Thickness +0.007/-0.007 inch 0.00973 Poison Cavity Thickness +0.010/-0.010 inch 0.00004 Poison Panel Width +0.075/-0.075 inch 0.00047 Asymmetric Assembly Position 0.00534 Calculation Uncertainty 0.00129 Benchnmk Bias Uncertainty Total Uncertainty (convoluted) 0.01590 Final K,fr on 95/95 Basis 0.99976 30f4

Table 2. Region 2- No Soluble Boron Base Keno Reference Reactivity 0.97383 Calculation and Methodology Biases Range Methodology (Benchmark) Bias 0.00770 Pool Temperature Bias 50 F to 185 F 0.00103 Boron Particles in Borafiex 599~

Total Bias 0.01323 Tolerances and Uncertainties Parameter Reactivity Variation Variation Fuel Enrichment . +0.05/-0.05 % 0.00972 Fuel Density +2/-2 % 0.00254 Fuel Pellet Dishing 1 187% 0.00116 Rack Cell Inner Dimension +0.025%).025 inch 0.00000 Rack Cell Pitch +0.07/-0.03 inch 0.00116 Rack Wall Thickness +0.007/-0.007 inch 0.00000 Wrapper Plate Thickness +0.002/-0.002 inch 0.00000 Poison Panel Thickness +0.007/-0.007 inch 0.00582 Poison Cavity Thickness +0.010/-0.010 inch 0.00000 Poison Panel Width +0.075/-0.075 inch 0.00026 Asymmetric Assembly Position 0.00000 Calculation Uncertainty 0.00041 Benchmark Bias Uncertainty Total Uncertainty (convoluted) 0.01213 Final K,~ on 95/95 Basis 0.99919 4of4

L-2000-054 Attachment 3 Environmental Consideration 10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not:

(i) involve a significant hazards consideration, (ii) result in a significant change in the types or significant increase in the amounts of any effluents that my be released offsite, and result in a significant increase in individual or cumulative occupational radiation exposure.

The proposed license amendments change the subcritical margin in the Spent Fuel Pool in order to accommodate degradation of the Boraflex panels in the fuel storage racks by permitting credit for soluble Boron. The proposed amendments do not expand the capacity of the Turkey Point Spent Fuel Pools. As described in UFSAR Section 5.2.4, each spent fuel pool rack has a maximum capacity of 1404 cells available for use, with no blanks inserted. The amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and no significant increase in individual or cumulative occupational radiation exposure. FPL has concluded that the proposed amendments involve no significant hazards consideration and meet the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). FPL has determined pursuant to 10 CFR 51.22(b), that an environmental impact statement or environmental assessment need not be prepared in connection with issuance of the amendments.

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Laboratory Testing of Nuclear Grade Activated Charcoal, Additional Information Body:

ADAMS DISTRIBUTION NOTIFICATION.

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Page 1

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A001 - OR Submittal: General Distribution Docket: 05000250 Docket: 05000251 Page 2

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. MAR 09?ooo L-2000-068 10 CFR $ 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Re: Turkey Point Units 3 4 4 Docket Nos. 50-250 & 50-251 Proposed License Amendments Laboratory Testing of Nuclear Grade Activated Charcoal Additional Information In accordance with 10 CFR $ 50.90, Florida Power and Light Company (FPL) requested in letter L 239, dated November 23, 1999, that Appendix A of Facility Operating Licenses DPR-31 and DPR-41 be amended to modify Tcchnical Spcciflication (TS) 3/4.6.3, Emergency Containment Filtering System, TS 3/4.6.6, Post Accident Containmcnt Vent System, and TS 3/4.7.5, Control Room Emergency Ventilation System. Thc proposed license amendments were submitted in response to Generic Letter (GL) 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, which requires that ASTM D3803-1989 be used for testing both ncw and used charcoal in cngincercd safety feature (ESF) applications.

As a result of conversations with your staff FPL is plcascd to provide the following additional information regarding the Control Room Emergency Ventilation System (CREVS). The face velocity of thc CREVS charcoal filters is not an overt design parameter. Rather, the design flow rate for the CREVS filters is a volumetric flow rate of 1000 cubic feet per minute.

FPL reviewed the Turkey Point UFSAR and available correspondence on control room habitability to determine ifthe CREVS charcoal filter face velocity was previously transmitted to the NRC as part of an earlier submittal. No source documents were found that would indicate that the CREVS charcoal filter face velocity was previously docketed. As a result, FPL has prepared the attached tables. These tables summarize the information previously provided to the NRC in our responses to Generic Letter 99-02. The tables also include the requested information on CREVS face velocity.

The following parameters substantiate the 40 fpm CREVS face velocity specified in the attached table:

CREVS Filter Volumetric Flow: 1000 cfm Number of CREVS Charcoal Cells: 3 Number of Beds in Each Cell: 2 Charcoal Bed Surface Area: 643 in (26.5 in. x 24.25 in.)

Dividing thc filter volumetric flow rate by the number of CREVS charcoal cells gives a volumetric flow mtc of approximately 334 cfm per cell. Dividing this cell volumetric flow rate by the total charcoal bed surface area for fiow in each cell gives the charcoal filter face velocity. Since each charcoal bcd has 643 in of surface area for flow, and each cell has a parallel arrangement of two charcoal beds, the total surface area for fiow is 1286 in'r 8.9 ft'pcr cell. Dividing thc cell volumetric flow rate of 334 cfm by this total surface area for flow gives a face velocity, i.c., linear velocity, of approximately 37.5 fpm. This value is rounded up to 40 fpm to account for a worst case combination of dimensional toleranccs, and the slight reduction in surface flow area caused by thc charcoal bcd framing members.

an FPL Group company

L-2000-068 page 2 of 3 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendments Laboratory Testing of Nuclear Grade Activated Charcoal Additional Information The above parameters were taken from Section 4.7.5c.l of the plant technical specifications, and Revision 1 of drawings 5610-M-38-16 and 5610-M-38-19.

Should there be any questions, please contact us.

Very truly yours, R. J. Hovey Vice President Turkey Point Plant attachment cc: Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant Florida Department of Health and Rehabilitative Services

L-2000-068 page 3 of3 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendments Laboratory Testing of Nuclear Grade Activated Charcoal Additional Information STATE OF FLORIDA )

) ss.

'OUNTY OF MIAMI-DADE )

R. J. Hovey being first duly sworn, deposes and says:

That he is Vice President, Turkey Point Plant, of Florida Power and Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.

R. J. Hovey Subscribed and sworn to before me this day of 2000.

Name of Notary Public (Type or Print)

"y r ",, CHERYL A. STEVENSON IP~ +;- I N(ccMMssoN cc SM17

><'mmeey~uda ~

R. J. Hovey is personally known to me.

)

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L-2000-068 Page 1 of 1 TABLE 1 - CURRENT TS REQUIREMENTS System Description Current TS Requirements Bed Credited Face Test Test Face TS Safety Test Test System Thickness Velocity Penetration Temp~

Section Efficiency'methyl Factor Standard Velocity'ft/min)

(inches) iodide) (ff/min) (methyl iodide) ('C) RH'5%

3/4.6.3 ECFS 30% 40 N/A4 N/A ANSI N510-1975 130 40 3/4.7.5 CREVS 95% 4Q s 1% ANSI N510-1975 25 70% 40 3/4.6.6 PACVS N/As 14 F10 N/A ANSI N510-1975 25 70% 40

'redited as used in the safety analyses

~

Not a current technical specification requirement s

Methyl iodide removal by the PACVS is not credited in the plant dose analyses 4

Methyl iodide penetration in the ECFS is not tested. Current technical specification only requires elemental iodine testing TABLE 2 - PROPOSED TS REQUIREMENTS System Description Proposed TS Requirements Bed Credited Face Test Test Face TS System Thickness Velocity Penetration

$ f Test Temp Test Velocity" Section Efficiency'methyl Standard RH (inches) iodide) (A/min) (methyl iodide) ('C) (ft/min) 3/4.6.3 ECFS 30% 40 35% ASTM D3803-1989 30 95% 40 3/4.7.5 CREVS 95% 40 < 2.5% ASTM D3803-1989 30 95% 40 3/4.6.6 PACVS N/As 14 < 1Q% N/A ASTM D3803-1989 30 95% 40

'redited as used in the safety analyses

~

Not a proposed technical specification requirement s

Methyl iodide removal by the PACVS is not credited in the plant dose analyses

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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003689828

Subject:

Turkey Point Plant - RAI - RE: Potential Risk of the Proposed Civil and Government Aire raft Operations at Homestead Air Force Base (TAC Nos. MA6249 and MA6250)

Body:

ADAMS DISTRIBUTION NOTIFICATION.

Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML003689828.

DF01 - Direct Flow Distribution: 50 Docket (PDR Avail)

Docket: 05000250 Docket: 05000251 Page 1'

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March 8, 2000 Mr. Thomas F. Plunkett Presiddot - NLIclear Division Florida Power and Light Company P.O. Box 14000 Juno Beach; Florida 33408-0420 REQUEST FOR ADDITIONALINFORMATION REGARDING THE POTENTIAL I'UBJECT:

RISK OF THE PROPOSED CIVILAND GOVERNMENT AIRCRAFT

,OPERATIONS AT HOMESTEAD AIR FORCE BASE ON THE TURKEY POINT

'PLANT (TAC NOS. MA6249 AND MA6250) 1

Dear Mr. Plunkett:

I By letter dated November 17, 1999; Florida Power and Light Company's (FPL's) responded to the U.S. Nuclear Regulatory Commission (NRC) staff request regarding the t

above subject. The NRC staff has reviewed FPL's submittal and has determined that

'I 1 additional information is needed by the staff before it can complete its review. The enclosed request for additional information (RAI) has been discussed with Olga Hanek of your staff. A target date for your response has been agreed upon to be 45 days from your receipt of this RAI. Should a situation occur that prevents you from meeting the target date, please contact me at (301) 415-1496..

/RA/ ~

fdPjf~jQpgg:

Kahtan N. Jabbour, Senior Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-250 and 50-251

'\

Enclosure:

Request for Additional Information cc w/encl: See next page Distribution:

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'UBLIC LWcrt OGC RCorreia PDII-2 R/F ESullivan BClayton FAkstulewicz KJabbour MRubin DOCUMENT NAME: G:>PDii-2<Turkey)RAIA6249 *See previous concurrence INDICATE IN BOX: "C"-"COPY W/0 ATfACHMENT/ENCLOSURE, "E"=COPY W/ATT/ENCL, "N"=NO COPY OFFICE PM:PDII-2 E LA:PDII-2 SC:SPSB SC:PDII-2 NAME KJabbour BCla on MRubin RCorreia&

DATE 03/ /00 03/'4 /00 03/ 02 /00 03/i /00 OFFICIAL RECORD COPY

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lf UNITED STATES NUCLEAR REGULATORY COMMISSION i>i

'~Pgf'O WASHINGTON, D.C. 20555-0001 March 8, 2000 IcwrrS Mr. Thomas F. Plunkett President - Nuclear Division Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420

SUBJECT:

REQUEST FOR ADDITIONALINFORMATION REGARDING THE POTENTIAL RISK OF THE PROPOSED CIVILAND GOVERNMENT AIRCRAFT OPERATIONS AT HOMESTEAD AIR FORCE BASE ON THE TURKEY POINT PLANT (TAC NOS. MA6249 AND MA6250)

Dear Mr. Plunkett:

By letter dated November 17, 1999, Florida Power and Light Company's (FPL's)

I responded to the U. S. Nuclear Regulatory Commission (NRC) staff request regarding the above subject. The NRC staff has reviewed FPL's submittal and has determined that additional information is needed by the staff before it can complete its review. The enclosed request for additional information (RAI) has been discussed with Olga Hanek of your staff. A

\

target date for your response has been agreed upon to be 45 days from your receipt of this RAI. Should a situation occur that prevents you from. meeting the target date, please contact me at (301) 415-1496.

Sincerely, led <

Kahtan N. Jabbour, Senior Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-250 and 50-251

Enclosure:

Request for Additional Information cc w/encl: See next page

Mr. T. F,. Plurtkett TURKEY POINT PLANT Florida Power and Light Company CC:

M. S. Ross, Attorney Attorney General Florida Power & Light Company Department of Legal Affairs P.O. Box 14000 The Capitol Juno Beach, FL 33408-0420 Tallahassee, Florida 32304 Mr. Robert J. Hovey, Site Plant Manager Vice President Turkey Point Nuclear Plant Turkey Point Nuclear Plant Florida Power and Light Company Florida Power and Light Company 9760 SW. 344th Street 9760 SW. 344th Street Florida City, FL 33035 Florida City, FL 33035 Mr. Steve Franzone County Manager Licensing Manager Miami-Dade County Turkey Point Nuclear Plant 111 NW 1 Street, 29th Floor 9760 SW. 344th Street Miami, Florida 33128 Florida City, FL 83035 Senior Resident Inspector Mr. John Gianfrancesco Turkey Point Nuclear Plant Manager, Administrative Support U.S. Nuclear Regulatory Commission and Special Projects 9762 SW. 344~ Street P.O. Box 14000 Florida City, Florida 33035 Juno Beach, FL 33408-0420 Mr. William A. Passetti, Chief Mr. J.A. Stall Department of Health Vice President - Nuclear Engineering Bureau of Radiation Control Florida Power & Light Company 2020 Capital Circle, SE, Bin ¹C21 P.O. Box 14000 Tallahassee, Florida 32399-1741 Juno Beach, FL 33408-0420 Mr. Joe Myers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-21 00

RE VEST FOR ADDITIONALINFORIVIATION REGARDING THE POTENTIAL RISK OF THE PROPOSED CIVILAND GOVERNMENT AIRCRAFT OPERATIONS AT HOMESTEAD AIR FORCE BASE ON TURKEY POINT UNITS 3 AND 4 FLORIDA POWER AND LIGHT COMPANY DOCKET NOS. 50-250 AND 50-251

1. The attachment to the FPL June 15, 1998 letter response (L-98-152) on aircraft hazards presents the equation f=. N*P*A"F I.

as part of the Department of Energy methodology for assessing the risk of aircraft crashes to nuclear power plants. The definition of P is given as "in flight crash rate per mile...." In addition, F is defined as "crash probability density over area A," without any mention of units.

If F is dimensionless, then the units of f work-out to be (Flight operation/year)'(crashes/mile)'(sq. miles)*(probability density).

This has the units of Flight operations-crashes-miles/year which is incompatible with the quantity f, whose units are crashes/year.

The same equation is also presented in FPL's attachment to June 24, 1994 letter response (L-94-157) on IPEEE results for aircraft. However, some of the definitions appear to be different. Specifically, on page 27, P is defined as "probability of an aircraft crash per operation." With this definition the units for the equation are (Flight operations/year)*(crashes/flight operations)*(sq. miles)*(probability density).

This works-out to have the units Crashes-sq.miles/year which again is inappropriate for a crash frequency. It appears in this case that if the crash probability density had the units of (1/sq. mile) then the overall crash frequency would have the units of crashes/year.

ENCLOSURE

r Please provide a clarification of the units that were used in both analyses with respect to the crash probability and the crash probability density.

2. With respect to the aircraft risk analyses performed for Turkey Point Units 3 & 4, please indicate how the presence of the adjacent fossil unit chimneys was taken into account when calculating the effective target area used in estimating the on-site crash frequency. Indicate the relative effect of the chimneys on the total calculated effective target area.

3. The on-site crash frequency was estimated using parameters that are dependent on aircraft type and flight phase. Specifically, this applies to the parameters N, P, A, and F in the equation f=N*P*A*F.

That is, the equation is really of the form f=gQ N) P(A; F)

I j where i is the ith type of aircraft and j is the jth flight phase. Please provide a sample of representative values (e.g., for a commercial air carrier and a large rhilitary aircraft) that were used in the analyses for each of these parameters. Please indicate the source of the information used to evaluate each parameter.

4. According to the draft SEIS for the proposed disposal of some of the former Homestead Air Force Base, bird strikes can cause aircraft mishaps. Hence, some portion of the overall crash rate for a given aircraft and flight phase may be attributable to bird strikes. To what extent has the possibility of bird strikes been incorporated in the aircraft risk assessment for Turkey Point Units 3 & 4? If the Turkey Point aircraft risk analyses are based on nationally averaged aircraft crash rates, please indicate how representative these rates are of the projected Homestead air operations with respect to the bird strike contribution?

5. The draft SEIS (pp. 2.2-9 to 2.2-11), in discussing the projected air traffic for the proposed Homestead airport conversion, indicates that more than 80% of the traffic is estimated to be in connection with flights from Latin America, the Caribbean, or other international locations. The aircraft crash rates presented in NUREG-0800, SRP 3.5.1.6, are based on data for U.S.

Carriers, General Aviation, and military aviation. Hence, the data may not be representative of the air traffic mix being projected for the Homestead airport.

For example, in an item presented by the National Center for Policy Analysis, reference is made to an 80-page report of the Commercial Aviation Safety Strategy Team in which the U.S.

accident rate from 1987 to 1996 is described to be on the average of 0.5 major accidents per million departures, compared to 0.7 for Western Europe, 4.8 for Eastern Europe and the old Soviet Union, 5.7 for Latin America and 13 for Africa. This suggests that the accident rate could be significantly affected by the mix of air traffic that is being projected. Indicate if this has been taken into account in the FPL aircraft analyses to-date and if not, to what extent would this affect the previously estimated aircraft risks for Turkey Point Units 3 & 4.

• (http://www.ncpa.org/pd/regulat/pdreg/regfeb98e.html)

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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003678652

Subject:

Turkey Point Units 3 and 4 RAI on Soluble Boron Credit for Spent Fue l Pool and Fresh Fuel Rack Criticality Analyses Body:

ADAMS DISTRIBUTION NOTIFICATION.

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~~s'oa N l4'4 oos~7~~

January 31, 2000 ~p/gh W A'~-+//

Mr. Thomas F. Plunkett President - Nuclear Division Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420

SUBJECT:

TURKEY POINT UNITS 3 AND 4 - REQUEST OF ADDITIONALINFORMATION REGARDING SOLUBLE BORON CREDIT FOR SPENT FUEL POOL AND FRESH FUEL RACK CRITICALITYANALYSES (TAC NOS. MA7262 AND MA7263)

Dear Mr. Plunkett:

By letter dated November 30, 1999, Florida Power and Light Company's (FPL's) proposed technical specification changes for Turkey Point Units 3 and 4. The proposed changes would permit taking credit for the soluble boron in the spent fuel pool and fresh fuel rack criticality analyses in order to accommodate degradation of the boraflex panels in the fuel storage racks.

The NRC staff has reviewed FPL's submittal and has determined that additional information is needed by the staff before it can complete its review. The enclosed request for additional information (RAI) has been discussed with S. Mihalakea of your staff. A target date for your response has been agreed upon to be 30 days from your receipt of this RAI. Should a situation occur that prevents you from meeting the target date, please contact me at (301) 415-1496.

Sincerely,

/&W R. Hernan for:

Kahtan N. Jabbour, Senior Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-250 and 50-251

Enclosure:

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pe REGNI Wp 0

UNITED STATES 0

O NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Jaauary 31, 2000 0'w*w+

Mr. Thomas F. Plunkett President - Nuclear Division Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420

SUBJECT:

TURKEY POINT UNITS 3 AND 4 - REQUEST OF ADDITIONALINFORMATION REGARDING SOLUBLE BORON CREDIT FOR SPENT FUEL POOL AND FRESH FUEL RACK CRITICALITYANALYSES (TAC NOS. MA7262 AND MA7263)

Dear. Mr. Plunkett:

By letter dated November 30, 1999, Florida Power and Light Company's (FPL's) proposed technical specification changes for Turkey Point Units 3 and 4. The proposed changes would permit taking credit for the soluble boron in the spent fuel pool and fresh fuel rack criticality analyses in order to accommodate degradation of the boraflex panels in the fuel storage racks.

The NRC staff has reviewed FPL's submittal and has determined that additional information is needed by the staff before it can complete its review. The enclosed request for additional information (RAI) has been discussed with S. Mihalakea of your staff. A target date for your response has been agreed upon to be 30 days from your receipt of this RAI. Should a situation occur that prevents you from meeting the target date, please contact me at (301) 415-1496.

Sincerely, l~l Kahtan N. Jabbour, Senior Project Manag, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-250 and 50-251

Enclosure:

Request for Additional Information cc w/encl: See next page

'v Mr. T. F. Plunkett TURKEY POINT PLANT Florida Power and Light Company CC:

M. S. Ross, Attorney Attorney General Florida Power 8 Light Company Department of Legal Affairs P.O. Box 14000 The Capitol Juno Beach, FL 33408-0420 Tallahassee, Florida 32304 Mr. Robert J. Hovey, Site Plant Manager Vice President Turkey Point Nuclear Plant Turkey Point Nuclear Plant Florida Power and Light Company Florida Power.and Light Company 9760 SW. 344th Street 9760 SW. 344th Street Florida City, FL 33035 Florida City, FL 33035 Mr. Steve Franzone County Manager Licensing Manager Miami-Dade County Turkey Point Nuclear Plant 111 NW 1 Street, 29th Floor 9760 SW. 344th Street Miami, Florida 33128 Florida City, FL 33035 Senior Resident Inspector Mr. John Gianfrancesco Turkey Point Nuclear Plant'.S.

Manager, Administrative Support Nuclear Regulatory Commission and Special Projects 9762 SW. 344~ Street P.O. Box 14000 Florida City, Florida 33035 Juno Beach, FL 33408-0420 Mr. William A. Passetti, Chief Mr. Rajiv S. Kundalkar Department of Health Vice President - Nuclear Engineering Bureau of Radiation Control Florida Power & Light Company 2020 Capital Circle, SE, Bin ¹C21 P.O. Box 14000 Tallahassee, Florida 32399-1741 Juno Beach, FL 33408-0420 Mr. Joe Myers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100

REQUEST FOR ADDITIONALINFORMATION RELATED TO THE AMENDMENTOF THE TECHNICAL SPECIFICATIONS FOR SOLUBLE BORON CREDIT FOR SPENT FUEL POOL AND FRESH FUEL RACK CRITICALITYANALYSES TURKEY POINT UNITS 3 AND 4 FLORIDA POWER AND LIGHT COMPANY DOCKET NOS. 50-250 AND 50-251 The NRC staff safety evaluation report contained in WCAP-1441'6-NP-A presents the required technical specifications for use with the approved soluble boron credit methodology. The Fuel Storage Criticality specifications in the Design Features Section for both k-eff less than 1.0 if fully flooded with unborated water and for k-eff less than or equal to 0.95 if fully flooded with borated water require reference to WCAP-14416-P for a description of the uncertainties included. Therefore, proposed technical specifications 5.6.1.1.a and 5.6.1.1.b should include the phrase "which includes a conservative allowance for uncertainties as described in WCAP-14416-P."

Please describe the administrative procedures used to select the appropriate assemblies for storage in the burnup-dependent racks in Region 2.

Attachment 5 describes the criticality analysis performed with a reduced B-10 loading in

'he degraded boraflex. The assumptions in the analysis include the following:

Region 1: 0.009 g/cm'borber B-10 loading and 0.0351 inch thickness Region 2: 0.006 g/cm'bsorber B-10 loading and 0.051 inch thickness The analysis based on these assumptions results in a K, less than 1.0 with no soluble boron. Please provide your plan to verify that the boraflex panels have not degraded beyond the assumed thicknesses.

ENCLOSURE

0 ,v 0

Distri- l.txt Distribution'Sheet Priority: Normal From: E-RIDS3 Action Recipients: Copies:

RidsNrrPMKJabbour 0 OK Internal Recipients:

RidsRgn2MailCenter OK RidsManager 0 OK RidslroMailCenter 0 OK RidsAcrsAcnwMailCenter 0 OK IRO Paper Copy FIL'ECENTER 01 Paper Copy External Recipients:

NRC PDR Paper Copy NOAC Paper Copy IMS Paper Con~.

Total Copies:

ATTENTION DCD Item: ADAMS Package Library: ML ADAMS"MQNTAD01 PLE<ASE CRE<ATE 4 PROVIDE COPIES ID: 011130369 CD TO ALL RECIPIE<NTS OI<'TTACHED

Subject:

Letter forwarding Turkey Point Units 3 and 4 Updated Final Safety Analysis Report Revisi on 17 Document Date: 04/16/2001 Body:

ADAMS DISTRIBUTION NOTIFICATION.

A053 - OR Submittal: Updated FSAR (50.71) and Amendments

Title:

Letter forwarding Turkey Point Units 3 and 4 Updated Final Safety Analysis Report R evision 17 Page 1

Docket Number: 05000250 Docket Number: 05000251 Document Date: 04/16/2001 Author Name: Hovey R J Author Affiliation: FPL Group Company Addressee Name:

Addressee Affiliation:NRC/Document Control Desk Document Type: Letter Availability: Publicly Available Document Sensitivity: Non-Sensitive Comment: Electronic Drawing files contained on CD-ROM were not added to ADAMS.

A copy of the CD-ROM was provided to the NRC Public Document Room and File Cent er.

Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML011130369.

Page 2

APR l6 2001 L-2001-086 10 CFR 50.4 10 CFR 50.71 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 U dated Final Safet Anal sis Re ort Revision17 Florida Power and Light Company has completed Revision 17 of the Turkey Point Units 3 and 4 Final Safety Analysis Report (FSAR)..

The enclosed information accurately reflects plant changes made since the previous submittal. This revision incorporates changes completed between April 9, 1999 and October 23, 2000. Miscellaneous user comments resolved during this time period have also been incorporated.

A single CD-ROM of this document is being submitted in lieu of hard copies in accordance with guidance provided by RIS 2001-05, "Guidance on Submitting Documents to the NRC by Electronic Information Exchange or on CD-ROM," and NRC letter to Turkey Point dated March 28, 2001, "Florida Power & Light Co., Turkey Point Plant, Request for Exception to 10 CFR 50.4, Written Communications," from Brenda J. Shelton. This CD-ROM submittal of the complete FSAR will make obsolete all previous hard copies of the document. It is requested that you destroy or return to us these obsolete copies.

If you have any questions, please contact Steve Franzone at 305-246-6228.

Very truly yours,

. Hovey Vice President Turkey Point Plant DRL Attachment cc: Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point gD ggc an FPL Group company p ~g ] I QOZ~ Z I

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Distri20.txt Distribution S eet Priority: Normal From: Andy Hoy Action Recipients: Copies:

>/re P K Jabbour 1 Paper Copy Internal Recipients:

RGN2. Paper Copy IRO Paper Copy FILE CENTER 0 Paper Copy ACRS Paper Copy External Recipients:

'RC PDR 1 NOAC 1 IHS Total Copies: 8 DCD has 11 copies of enc losure Item: ADAMS Document Library:. ML ADAMS HQNTAD01 ID: 003677802

Subject:

Revision 17 to Updated Final Safety Analysis Report. for Turkey Point U nits 3 and 4.

Body:

ADAMS DISTRIBUTION NOTIFICATION.

Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching f or Accession Number ML003677802.

A053 OR Submittal: Updated FSAR (50.71) and Amendments Docket: 05000250 ~

Page 1

U

Distri20.txt Docket: 05000251 Page 2

l 4 y ~ ~

Non-contro1led Document Re>>ion Transmittal Inted: 1/1 9/00 t FSAR Name: U.S.N.R.C. US Loc: US MAIL DOCUMENT CONTROL DESK WASHINGTON, D.C. 20555

'ote: 11 COPIES The holder is responsible for verifying that this revision is complete and for destroying superseded copies.

Title FINAL SAFETY ANALYSIS REPORT To Be Removed: To Be Added:

Revision Revision No. of Pa es LIST OF EFFECTIVE PAGES 16 17 8 CHAPTER 4, TABLE 4.1-3~ 16 17-001 1 CHAPTER 5, TEXT PAGE 5.1.3-16 + 13 17-004 1 CHAPTER 5, APPENDIX 5A, PAGE 5A-11 16 17-004 1 CHAPTER 9, TEXT PAGE 9.5-17 V 13 17-003 1 CHAPTER 9, APPENDIX 9.6A, PAGE 96A-63~ 15 17-007 1 CHAPTER 11, CONTENTS PAGE 11-Iii 13 17-008 1 CHAPTER 11, TEXT PAGE 11.4-1 (NEW)17-008 1 CHAPTER 12 (NEW CHAPTER, CONTENTS, TEXT)17-008 10 APPROVED INTERIM REVISIONS TO THE FSAR ARE ISSUED AND WILL BE MAINTAINEDON YELLOW PAPER (PLEASE TAKE CAUTION WHEN FILING)

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~ ~ 14 3.2.3 4 4 3.2.3-28 6 3.2. 15 6 3.2.3-29 1 3.2 -16 6 3.2.3-30 1 3.2.3-31 16 CHAPTER 4 3.2.3-32 0 3.2.3-33 0 Contents 3.2.3-34 0 3.2 '-35 4 4-i 13 3.2.3-36 0 4-A 16 3' '

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PAGE PAGE REV CHAPTER 4 (cont'd) Table Table 4.2-1(shl)g 1 4.2-1(sh2) 16 4.1-4 16 4.2-2 13 4.1-5 16, 4.2-3 8 4.1-6 0 4.2-4 0 4.1-7 0 4.1-8 14 ~ai ere 4.1-9 15 4.2-1 13 Text 4.2-2(Part 1) 0 4.2-2(Part 2) 0 4.2-1 4.2-3 0 4.2-2 0 4.2-4 13 4.2-3 16 4.2-5 0 4.2-3a 6 4.2-6 0 4.2-4 0 4.2-7(shl) 0 4.2-5 1 4.2-7(sh2) 0 4.2-6 0 4.2-7(sh3) 0 4.2-7 13 4.2-7(sh4) 0 4.2-7a 13 4.2-8 3 4.2-8 ~

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PAGE PAGE REV CHAPTER Text contents 5.1.3-16 13 5-i 5.1.3-17 0 16 5.1.3-18 16 5-i 1'-l 13 5.1.3-19 16 11 16 5.1.3-20 5-iv 16 5.1.3-21 16 5-iva 10 5.1.3-22 0

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~ 16 5.1.3-34 0 5.1.1-7 16 5.1.3-35 0 5.1.3-36 0 5.1.2-1 13 5.1.3-37 16 5.1.2-2 16 5.1.3-38 13 5.1.2-3 10 5.1.2-4 0 Tabl e 5.1.2-5 16 5.1.2-6 16 5.1.3-1 5.1.2-7 11 Text 5.1.3-1 13 5.1.3-1a 13 5.1.4-1 16 5.1.3-2 0 5.1.4-1a 13 5.1.3-3 13 5.1.4-1b 13 5.1.3-4 0 5.1.4-2 0 5.1.3-5 13 5.1.4-3 13 5.1.3-6 0 5.1.4-4 0 5.1.3-7 0 5.1.4-5 13 5.1.3-8 0 5.1.4-6 0 5.1.3-9 0 5.1.4-7 13 5.1.3-10 13 5.1.4-8 0 5.1.3-11 0 5.1.4-9 0 5.1.3-12 0 5.1.4-10 0 5.1.3-13 0 5.1.4-11 5.1.3-14 0 0 5.1.3-15 0 5.1.4-12 0 5.1.4-13 0 Rev. 16 10/99

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~cHApTER (cont d) Text Tab1e 5.1.8-1 0 5.1.8-2 0 5.1.4-1(shl) 0 5.1.8-3 16 5.1. 4-1(sh2) 0 5.1.4-1(Sh3) 0 5.1.9-1 5.1.4-1(Sh4) 13 5.1.9-2 5.1. 4-1(Sh5) 13 5.1.9-3 5.1. 4-1(sh6) 0 5.1.9-3a 5.1.9-4 Tex't

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5. 1-1(sh3) 7 5.1.6-1 0 5. 1-2 16 5.1.6-2 0 5. 1-3 9 5.1.6-3 0 5.1-4(shl) 0 5.1.6-4 16 5. 1-4(sh2) 0 5.1.6-5 0 5. 1-5 0 5.1.6-6 0 5. 1-6 0 5.1.6-7 0 S. 1-7 0 5.1 6-8

~ 0 '5. 1-8 0 5.1.6-9 0 5. 1-'9(Shl) 0 5.1.6-10 0 5. 1-9(sh2) 0 5.1.6-11 0 5.1-9/Note 1 13.

5.1.6-12 9 5.1-9(sh3) 0 5.1-9(sh4) 0 Tabl e 5.1-10(shl) 0 5.1-10(Sh2) 0 5.1.6-1 5.1-10(sh3) 0 5.1.6-2 (Deleted) 5. 1-10/Note 1 13

5. 1-10(sh4) 0 Text 5. 1-11 0

5. 1-12(Shl) 0 5.1.7-1 16 5. 1-12 (Sh2) 0 5.1.7-2 0 5. 1-13 0 5.1.7-3 0 5.1-14/Note 1 13 5.1.7-4 0 5.1-14 0 5.1.7-5 11 5.1-15 6 5.1.7-6 16 5.1-16 0 5.1 7-7

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~CHAPTER (cont ') TEXT Ficiure 5.4.2-1 16 5.1-18 0 5.4. 3-1 5.1-19(sh1) 0

5. 1-19(sh2) 0 Tabl e

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5. 1-19(shS) 0 5.1-20 0 Tekt 5.1-21 13 Text 5.5.1-1 5.2.1-1 5.2.3-1 16 Text 5.2.3-2 0 Sa-1 14 5 '.4-1 5a-2 Sa-3 14 16 5.3.1-1 5A-4 16 SA-5 16 5.3.2-1 10 Sa-6 15 SA-7 14 5.3.3-1 10 5a-8 15 SA-9 14 5.3.4-1 10 SA-10 16 5.3.4-2 10 SA-11 16 5.3.4-3 10 Sa-12 15 5.3.4-4 10 SA-13 15 5.3.4-5 10 SA-14 14 5.3.4-6 10 Sa-15 16 5.3.4-7 10 Sa-16 14 5.3.4-8 10 Sa-17 16 5.3.4-9 10 5.3.4-10 10 Tabl e 5.3.4-11 10 5.3.4-12 10 SA-1 0 5.3.4-13 10 5A-2(Deleted) 14 SA-3 0 5.F 1-1 Rev. 16 10/99

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~arar>vEa (cunt'd) Ficiure Text 9.6-1 13 9.6-2 13 9.5-3 16 9 '-3 13 9.5-4 3 9.6-4 (Deleted) 12 9.5-5 16 9.6-5 13 9.5-6 15 9 '-7 F 6-6 13 4 9.6-7 13 9.5-7a 16 9. 6-,8 13 9.5-8 2 9.6-9 13 9.5-9 1 9.6-10 13 9.5-10 7 9.6-11 13 9.5-10a 15 9.6-12 9.5-11 15 13 9 '-12 9.6-13 13 9.5-12a 16 16 9 '-14 13 9.5-13 15 9.6-15 13 9.5-14 16 9.6-16 13 9.5-15 16 9.6-17 13 9.5-16 13 9.6-18 13 9.5-17 13 Text Table 9.7-1 0

9. 5-1 16 9.7-2 0 9.5-2 16 9.7-3 16

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~CHAPTER (cont'd) Table Text 9. 11-1(sh1) 16 9.11-1(Sh2) 16 9.9-1 13 9.11-2 8 9.9-2 12 9.9-3 12 Ficiure 9.9-4 12 9.11-1 9.9-5 16 14 9.11-2 13 9.11-3 13 ciciure 9.11-4 13 9.11-5 13 9.9-1 13 9.11-6 13 9.9-2 13 9.11-7 13 9.9-3 sh 1 13 9.11-8 13 9.9-3 sh 2 13 9.11-9 13 9 '-4 13 9.11-10 9.11-11 13 9.9-5 13 13 Text Text

9. 10-1 11 9.12-1 12 9.10-2 9.12-2 14 16 9.12-3 15 9.10-3 ~

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PAG REV PAGE REV Tex't Text 9.6A-96 8 9.6A-97 8 9.6A-SS 16 9.6a-98 15 9.6A-56 15 9.6a-99 16 9 'A-57 16 9.6A-100 8 9.6A-58 15 9.6A-101 10 9 'A-59 15 9.6A-102 16 9.6A-60 8 9.6A-103 16 9.6a-61 16 9.6a-104 .16 9.6A-62 16 9.6A-104a 15 9.6A-63 15 9.6A-105 16 9.6a-64 8 9.6a-106 16 9.6A-65 8 9.6A-107 16 9 'a-66 8 9. 6A-108 16 9.6a-67 15 9. 6A-109 16 9.6A-68 16 9. 6A-110 16

9. 6A-'69 8 9. 6A-111 16 9.6A-70 8 9. 6A-112 10 9.6A-71 8 9. 6A-113 8 9.6a-72 16 9. 6A-114 16 9.6a-73 16 9.6A;115 16 9.6a-74 8 9. 6A-116 16 9.6a-75 8 9.6a-117 16 9.6A-76 16 9.6A-118 16 9.6a-77 8 9.6A-119 8 9.6A-78 8 9.6A-120 16 9.6A-79 8 9. 6A-121 16 9.6a-80 8 9.6A-122 16 9.6a-81 8 9.6A-123 16 9.6A-82 8 9.6A-124 16 9 'a-83 10 9. 6A-125 8 9.6A-84 16 9. 6A-126 8 9.6a-85 16 9. 6A-127 8 9 'a-86 16 9. 6A-128 8 9.6A-87 16 9.6A-129 16 9.6A-88 16 9.6A-130 16 9.6A-89 16 9.6A-131 16 9.6a-90 16 9.6A-132 16 9.6A-91 8 9.6A-133 8 9.6A-92 13 9.6A-134 16 9.6A-92a 16 9. 6A-135 16 9.6A-93 16 9. 6A-136 16 9 'A-94 13 9.6A-137 16 9 'A-95 16 9. 6A-138 8 Rev. 16 10/99

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(* ') Text

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9. 6A-165 16 9.6a-207 16 9.6a-166 16 9.6A-208 8 9.6a-167 16 9.6a-209 10 9.6A-168 16 9.6A-210 16 9.6A-169 16 9.6A-211 8

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9. 6A-177 16 9.6A-219 16 9.6A-177a 16 9.6A-220 16 9.6a-177b 16 Rev. 16 10/99

UPDATED 0

FINAL SAFETY ANALYSIS REPORT TURKEY POINT UNITS 3 & 4 LIST OF EFFECTIVE PAGES VOLUME 4 (Continued)

PAGE REV PAGE REV CHAPTER 10 (Cont'd) Text Ficiure 10.3-1 10.3-2 10.2-52 13 10.3-3 10 '-53 13 10.2-54 13 10.4-1 16 10.2-55 13 10.2-56 13 10.2-57 13 10.2-58 13 10.2-59 13 10.2-60 13 10.2-61 13 10.2-62 13 10.2-63 13 10.2-64 13 10.2-65 13 Rev. 16 10/99

UPDATED FINAL SAFETY ANALYSIS REPORT TURKEY POINT UNITS 3 & 4 LIST OF EFFECTIVE PAGES VOLUME 5 PAGE REV PAGE REV CHAPTER 11 Tabl e Contents 11.1-6 11-1 16 ure 11-ii 15

~Fi 11-111 11-i v 11-v

'3

, 13 1101-1 11.1-2 13 13 13 11.1-3 11-vi 13 11.1-4 9

13 Text 11.1-5 0 11.1-6 (Deleted) 12

~

11 1-1 16 11.1-7 13 11.1-2 16 11.1-8 13 11.1-2a 16 11.1-9 13 11 '-3 11.1-4 16 11.1-10 11.1-11 13 13 16 11.1-5 16 11.1-12 13 11.1-6 0 11.1-13 13 11 1 7

~ 16 11.1-14 13 11.1-8 15 11.1-15 13 11.1-9 16 11.1-16 13 11.1-10 16 11.1-17 13 11.1-11 16 11.1-18 13 11.1-12 16 11.1-19 13 11e 1 13 16 11.1-20 13

11. 1-14 16 11.1-21 13

11. 1-15 16 11.1-22 13 11 '-16 16 11.1-17 16 Text 11.1-18 16 11.1-19 16 11.2-1 16 11.1-20 16 11.2-2 11 '-21 16 11 '-3 16 16 11.1-22 16 11.2-4 0 11.2-5 16 Table 11.2-6 6 11e2-7 0 11 e 1 1 11.2-8 16 11e1-2 0 11.2-9 0 16 11.2-10 0 11.1-3(sh1) 16 11.2-11 11

11. 1-3 (Sh2) 16 11.2-12 11.1-4 16 11.1-5 16 0

11 '-13 16 11.2-13a 16 Rev. 16 10/99

UPDATED FINAL SAFETY ANALYSIS REPORT TURKEY POINT UNITS 3 8c 4 LIST OF EFFECTIVE PAGES VOLUME 5 (Continued)

PAGE PAGE REV CHAPTER 11 (cont'd) Text Text 11.3-1 16 11 \ 3 2 13 11.2-14 16 11.3-2a 16 11.2-15 16 11 ~ 3 3 16 11.2-16 16 11.3-4 16 11.2-17 15 11.3-5 0 11.2-18 13 11.2-18a 15 Tab1e 11.2-19 10 11.2-20 15 11 ~ 3 1 13 11.2-21 10 11.2-22 16 CHAPTER 12 11.2-23 16 11.2-24 15 Text 11.2-24a 16 11.2-25 12 12-1 11 '-26 12 11 '-27 0 CHAPTER 13 11.2-28 11 11.2-29 0 Text Table 13-1 11.2-1 0 CHAPTER 14 11.2-2 14 11 ~ 2 3 14 Contents 11.2-4 14 11.2-5 0 14-i 14 11.2-6 0 14-A 15 11\27 11.2-7a (Del eted) 13 14-iii 14 13 14-iv 14 11.2-8 4 14-v 14 11.2-9 13 14-vi 16 11.2-10 0 14-vl i 14 11.2-11 0 14-viii 14 14-ix 16 Ficiure 14-x 16 14-xi 15

11. 2-1 13 14-xi 14 11.2-2 13 i'4-xiii 14 11.2-3 13 14-xiv 14 11.2-4 0 14-xv 15 11.2-5 0 14-xvi 14 Rev. 16 10/99

UPDATED FINAL SAFETY ANALYSIS REPORT TURKEY POINT UNITS 3 & 4 LIST OF EFFECTIVE PAGES VOLUME 5 (Continued)

PAGE PAGE REV CHAPTER 14 (cont'd) ~Tex Contents 14.1.2-1 14 14-xvii 14 14.1.2-2 14 14-xyiii 16 14.1.2-3 14 14-x1 x 16 14.1.2-4 14 14-xx 15 14.1.2-5 14 14-xx1 15 Tabl e Text 14.1. 2-1 14 14-1 14 14-2 14 Ficiure 14-3 14 14-4 14 14.1.2-1 14 14-5 14 14.1.2-2 14 14-6 14 14.1.2-3 14 Ficiure 14.1.2-4 14 14.1.2-5 14 14-1 14.1.2-6 14 14.1.2-7 14 Text 14.1.2-8 14 14.1-1 Text 14.1-2 14.1.3-1 (Deleted) 13 14.1.1-1 1 14.1.1-2 14 14.1. 4-1 15 14.1.1-3 14 14.1.4-2 15 14.1.1-4 13 14.1.4-3 14.1.1-5 14 14 14.1.1-6 15 Tab1e sable 14.1.4-1 13 14 1.1-1 F 14 14.1.4-2 13 ciciure Ficiure 14.1. 1-1 14 14.1.4-1 14 14.1.1-2 14 14.1.4-2 14.1.1-3 14 14 14.1.1-4 14 Rev. 16 10/99

TABLE 4.1-2a CHEMICAL ANALYSES IN WEIGHT PERCENT REACTOR VESSEL SURVEILLANCE MATERIAL Intermediate Lower Element shell Shell

~uni z Unit 4 Unit 3 Unit 4 C 0.20 0.22 0.20 0.21 Mn 0. 64 0.67 0.61 0.67 P 0. 010 0.010 0.010 0.011 S 0. 010 0.009 0.008 0.009 Si 0.26 0.20 0.20 0.23 Ni 0.70 0.71 0.67 0.70 Cr 0.40 0.33 0.38 0.31 V 0.02 0.002 0.02 0 '01 Mo 0.62 0.56 0.58 0.56 Co 0.011 0.017 0.015 0.015 Cu 0.058 0.054 0.079 0.056

~ Zr Tl

• 0.001 0.010

• 0.001 0.005 0.008

%0.001

~0.001 0.008

• 0.001 0.004 0.008

• 0.001 sb A0.001 *0.001 Zn 0.001 "0.001 0.001 *0.001 As "0.005 0.004 "0 005 F 0 005 F

B "0.003 "0.003 *0.003 *0.003 Al 0.005 0.008 0.005 0.008 N> 0 '03 0.001 0.003 0.002 Nb 0.002 0.001 W *0.001 ~0.001 Pb *0.001 0.001 Ta 0.003 0.002

" Not detected. The number indicates the minimum limit of detection.

Rev. 3-7/85

TABLE 4.1-3 PRESSURIZER AND PRESSURIZER RELIEF TANK DESIGN DATA Pressurizer Design/Operating pressure, psig 2485/2235 Hydrostatic Test Pressure (cold), psig 3107 Design/Operating Temperature oF 680/653 water Volume, Full Power, ft3

• 780 Steam Volume, Full Power, ft3 520 Surge Line Nozzle Diameter, in. /Pipe Schedule 14/Sch 140 Shell ID, in. /Minimum Shell Thickness, in. 84/4.1 Mimimum Clad Thickness, in. 0.188 Electric Heaters Capacity, kw (total) 1300 Heatup rate of Pressurizer using Heaters only, oF/hr 55 (approximately)

Power Relief Valves: ¹455C & 456 Number Set Pressure (open), psig i) Normal operation 2335 ii) OMS Actuation during Heatup or Cooldown a) RCS < 285'F 415 +15 b) RCS > 2850F Setpoint increases step-wise to 2335 psig as temperature increases to 750oF (See Table 4.1-1)

Capacity, lb/hr saturated steam/valve 179,000 Safety Valves Number 3 Set Pressure, psig 2485 +1% (as left)

+2%%d/-3X (as found)

Capacity, lb/hr saturated steam/valve 293,330 Pressuri er Rel i ef Tank Design pressure, psig 100 Rupture disc release pressure, psig 100 Design temperature, oF 340 Normal water temperature, oF 120 Total volume, ft3 1300 Rupture disc relief capacity, lb/hr 900,000 60%%d of net internal volume (maximum calculated power)

, Rev 16 10/99

I shows that failure could occur if vertical reinforcing were not provided.

fact the magimum allowable vertical averaged tensile s tress according to Taylor's interaction curve is fa 0.03 ft c therefore fa ~ +150 psi. For this reason, special anchorage zone reinforcing is used in addition to that required by the loading cases.

Such special reinforcing is based on the following considerations:

l. Pull scale load tests of the anchorage on the same concrete mix us'ed in the structure and review of prior uses of the anchorage.

2. The post-tensioning supplier's recommendations of anchorage reinforcing requirements.

3. Review of the final details of the combined reinforcing by the consulting firm of T. Y. Lin, Kulka, Yang and Associate.

For typical detailed Analysis, see Topical Report B-Top-2 dated October 1969, submitted in connection with Docket No. 50-255, a NON-PROPRIETARY report.

'I (b) 'Earthauake or Mind Loadin The stresses in the structure for the earthquake loading conditions exceed the stresses for design tornado or wind. The earthquake analysis is'onducted in the following manner:

The loads on the containment structure caused by earthquake are determined by a dynamic analysis of the structure. The dynamic analysis is made on an idealized structure of lumped

-.,asses and weightless elastic columns acting as springs.

5.1.3-15

The analysis is performed in two stages; the determination of natural frequencies of the structure and its mode shapes, and the response of these modes .to the earthquake by the spectrum response. For the supported equipment, piping, etc. a time history technique is used to develop the floor response spectrum curves, and the supported elements are then analyzed by the response spectrum method as discussed in Appendix SA, Section III.

II I

The natural frequencies and mode shapes are computed using the matrix equation of motion shown below for a lumped mass system. Matrix interation was performed by use of a digital computer program to yield the natural frequencies and mode shapes. The form of the equation is:

(K) ~

( a ) = ii' (M) ~

(a )

K = Matrix of stiffness coefficients including the combined 1,

effects of shear, flexure, rotation and horizontal translation.

M = Matrix of lumped masses a = Matrix of mode shapes m = Angular natural frequency of vibration The results of this computation are the several values of r~and mode shapes an for n .= 1, 2, 3, ---m where m is the number of degrees of freedom (i.e..

lumped masses) assumed in the idealized structure.

To obtain the .loads on the containment structure the response of each mode of vibratio'n to the design earthquake is computed by the response spectrum technique as follows:

5.1.3-16 Rev. 13 10/96

newer structures, wind loads are as required by the edition of the south Florida Building code applicable at the time of design. shape Factors are applied in accordance with Reference SA-4, or as required by the South Florida Building Code applicable at the time of design. No tornado loads are considered.

SA-1.4.2 Turke Point Fossil uni s 1 and 2 Chimne Desi n Re uirements The Fossil unit 1 8 2 chimneys, located directly north of unit 3, do not perform any safety related functions, or directly protect safety related equipment. However, failure of these structures has the potential of adversely affecting safety related systems. Accordingly, these structures have been designed to not fail and cause an adverse interaction with any safety related systems, when subjected to the Class I seismic loads (0.15 g) and wind loads (145 mph hurricane and 225 mph tornado) described in sections SA-1.3.4 and SA-1.3.5 of this appendix.

5A-1.5 Miscellaneous Loads for structures s stems and E ui ment The units are designed for an outdoor temperature range of +30oF to +95oF. No ice or snow loads are considered in the design of the various structures and equipment.

SA-2 ' METHOD OF SEISMIC ANALYSIS SA-2.1 Structures The methods for seismic analysis of the containment and control building structures are described in Section 5.1.3.2.

SA-2.2 Res onse S ectra Response spectra curves for floors at grade and for the containment basemat were developed based on the El Centro, California, earthquake. These curves are shown in Figures 5A-1 for the design basis earthquake event (E), and Figure SA-2 for the maximum earthquake event (E') . The analysis methodology is similar to the technique described in Section S.1.3.2(b). (Reference SA-3)

SA-2.3 Seismic Class I Pl in Anal sis seismic Class I piping systems are typically analyzed as mathematical models consisting of lumped masses connected, by elastic members. The distance from Rev. 16 10/99

the pipe axis to the center of gravity of the valve and operator is considered, with the mass of the valve and operator, for all motor, air, or gear operated valves. When necessary for the integrity of the piping, valve, or operation, the valve structure is externally supported. The stiffness matrix for the pipe is developed to include the effects of torsional, bending, shear and axial deformations as well as change in flexibility due to curved members and internal pressure. Flexibility factors are calculated in accordance with UsAs B31.1. System natur al frequencies and mode shapes for all significant modes of vibration are then determined using equations of motion, and spectral accelerations as determined from the response spectra applied.

The following equations are successively used to determine the response for each mode, maximum displacement for each mode, and the total displacement for each mass point:

p(max) =

M.co'.

(2) V. =cp. Y (max)

(3) V. = gV.

where:

Yn(max) = response of the n<h mode Rn = participation factor for the n<" mode = Mi g

Mi = mass l

$ 'ill mode shape i for n<h mode Sa = spectral acceleration for the n<h mode D = earthquake direction matrix M> = generalized mass matrix for the n<h mode = >Mi )~1n e> = angular frequency of the n<h mode Vill maximum displ acement of mass i for mode n V< = maximum displacement of mass i due to all modes calculated 5A-12 Rev. 15 4/98

9.5.3 SYSTEM EVALUATIQN Underwater transfer of spent fuel provides essential ease and corresponding safety 'in handling operations. Water is an effective, economic and transparent radiation shield and a reliable cooling medium for removal of decay heat.

Basic provisions to .ensure the safety of refueling operations ar':

\

a) Gamma radiation levels in the containment, control room and fuel storage areas are continuously monitored (see Section 11.2.3). These monitors provide an audible alarm at the initiating detector indicating an unsafe condition. Continuous monitoring of reactor neutron flux'provides immediate indication and alarm in the control room of an abnormal core flux level.

b) Containment integrity is maintained when core alterations or movement of irradiated fuel occurs inside the containment.

~ .) Whenever any fuel is being added to the reactor core or is being relocated, a reciprocal curve of source neutron multiplication is recorded to verify the subcriticality of the core.

Incident "Protection Direct communication between'the control 'room and the refueling cavity manipulator crane is required. whenever changes in core geometry which "affect criticality are taking place. This provision allows the control room operator to inform the manipulator crane operator of any impending unsafe conditions detected from the control board indicators during fuel movement.

Malfunction Anal sis An analysis is presented in Section 14 concerning damage to one complete outer row of fuel elements in an assembly, assumed as a conservative limit for evaluating environmental consequences of a fuel handling incident.

9.5-16 Rev. 13 10/96

9.5.4 TEST AND INSPECTION CAPABILITY Upon completion of core loading and installation of the reactor vessel head, certain mechanical and electrical tests can be performed prior to initial criticality. The electrical wiring for the rod drive circuits, the rod position indicators, the reactor trip circuits, the in-core thermocouples and the reactor vessel head water temperature thermocouples can be tested at the time of installation. The- tests can be repeated on th'ese electrical items before initial operation.

9.5.5 REFERENCE

1. Turkey Point Unit 4 Plant Change Hodification (PC/H)05-066, "Turkey Point Unit 4 Cycle 16 Reload," Revision 2, dated Harch 6, 1996.

9.5-17 Rev. 13 10/96

9.5.4 TEST AND INSPECTION CAPABILITY Upon completion of core loading and installation of the reactor vessel head, certain mechanical and electrical tests can be performed prior to initial criticality. The electri.cal wiring for the rod drive circuits, the rod position indicators, the reactor trip circuits, the in-core thermocouples and the reactor vessel head water temperature thermocouples can be tested at the time of installation. The tests can be repeated on these electrical "items before initial operation.

9.5-17

0 2.4 APPENDIX A TO BTP 9.5-1 GUIDELINES (Cont d)

A endix A Guidelines Plant Conformance Alternatives Remarks G.4 Haterials Containin Radioactivit Haterials that collect and contain radio- Haterials containing or collect-activity such as spent ion exchange resins, . ing radioactivity are stored in charcoal filters, and HEPA filters should closed metal containers in areas be stored in closed metal tanks or con- free of ignition sources or tainers that are located in areas free combustibles. Rated fire barriers from ignition sources or combustibles. are provided to preclude exposure These materials should be protected from to fire in adjacent areas.

exposure to fires in adjacent areas as Requirements for control of decay well. Consideration should be given to heat are deve)oped for specific requirements for removal of isotopic storage materials. as required.

decay heat from .entrained radioactive materials.

9.6A-63 Rev. 15 4/98

2.5 . CONFORMANCE TO 10 CFR PART 50 APPENDIX R RE UIREMENTS The information which follows is a lineup of the Turkey Point Units 3 and 4 designs against the requirements of Appendix R to 10 CFR Part 50. Also see the lineup against BTP Appendix A presented in Section 2.4 of this Appendix.

Appendix R requirements are given in the first (left-hand) column of the following tabulations, retaining the numbering, sequence of Appendix R:

Information on various aspects of the Turkey Point Units 3 and 4 Fire Protection Program is given in the second column as necessary to demonstrate conformance to the Appendix R Requirements, or in the third column to describe alternative approaches. The fourth column provides supplemental information as appropriate.

Based on the criteria established in 10 CFR Part 50.48, Turkey Point Units 3 and 4 are required to conform only to Sections III.G, III.J, and,III.O of Appendix R. Additional Sections requiring conformance as a result of prior.NRC review and acceptance of Turkey Point Units 3 and 4 design with respect to BTP APCSB 9.5-1 Appendix A are III.A, III.H, III.I and III.L. All other Sections of Appendix R are not applicable to Turkey Point Units 3 and 4.

9.6A-64 Rev 8 7/90

TABLE OF CONTENTS (Continued)

Section Title Pacae 11.2.3 Radiation Monitoring System 11.2-10 Process Radiation Monitoring System 11.2-11 Containment High Range Radiation Monitors 11.2-13 Containment Air Particulate Monitors 11.2-13

~

Containment Radioactive Gas Monitors 11.2-14 Plant Vent Gas Monitors '1.2-15.

Condenser Air Ejector Monitors 11.2-16 Component Cooling Liquid Monitors 11.2-16 Waste Disposal System Liquid Effluent Monitor 11.2-17 Steam Generator Liquid Sample Monitors 11.2-17

'Hain Steam Line Honitors 11.2-18 Reactor Coolant Letdown Line Activity Monitors 11.2-18a Spent Fuel Pool Vent Monitor - Unit 3 11.2-18a Area Radiation Monitoring System 11.2-19 System Description 11.2-19 The Detector 11.2-20 The Local Indicator 11.2-20 The Channel Indicator 11.2-21 Radiation Monitoring System Cabinet 11.2-21

, Health Physics Program 11.2-22 Facilities and Access Provisions 11.2-22 Personnel Monitoring 11.2-24 Personnel Protective Equipment 11.2-24a Monitoring Instrumentation 11.2-26 11.2.4 Evaluation 11.2-26 11.2.5 Tests and Inspection Capability 11.2-29 11.3 oacti ve Materi al s Safety 11.3-1 11.3.1 Materials Safety Program 11.3-1 11.3.2 Facilities and Equipment 11.3-3 11.3.4'adi Personnel 11.3.3 and Procedures 11.3-4 Required Materials 11.3-5 11-iii .Rev. 13 10/96

LIST OF TABLES Tabi e Title 11.1-1 Waste Disposal System Performance Data (Two Units) 11,1-2 Waste Disposal Components Code Requirements 11.1-3 Component Summary Data 11.1-4 Estimated Liquid Discharge to Waste Disposal 11.1-5 Estimated Liquid Release'y Isotope (Two Units) 11.1-6 Estimated Annual Gaseous Release by Isotope (Two Units) 11.2-1 Radiation Zone Classifications 11.2-2 Primary Shield Neutron Flux and Design Parameters 11.2-3 Secondary Shield Design Parameters 11.2-4 Accident Shield Design Parameters 11.2-5 Refueling Shield Design Parameters 11.2-6 Principal Auxiliary Shielding 11.2-7 Radiation Monitoring System Channel Sensitivities 11.2-7a DELETED 11.2-8 Detecting Medium Conditions 11.2-9 Portable Rad>at>on Survey Instruments 11.2-10 Instantaneous Radiation Sources Released to the Containment Following TID-14844 Accident. Release 11.2-11 Gap Activity Sources Circulating in Residual Heat Removal Loop and Associated Equipment 11.3-1 Byproduct, Source and Special Nuclear Materials; Radioactive Sources Listing 11-iv Rev. 13 10/96

12. CONDUCT OF OPERATIONS Organization and Responsibility This section covered the positions and personnel at the time of initial plant startup and operation. This information can be found in the original docketed FSAR and this is also addressed in the plant operating license (Technical Specifications).

Training This section covered the training program at the time of initial plant startup and operation. This information can be found in the original docketed FSAR and this is also addressed in the Technical Specifications.

Procedures The operating procedures for startup, normal operations, and anticipated emergency operating conditions is addressed in the original docketed FSAR and current requirements indicated in the Technical Specifications. The Emergency Plan in effect for Turkey Point is issued as a separate document.

Records The procedure for maintaining plant operating, maintenance, QA, personnel, training, and instrumentation and control record is addressed in the original docketed FSAR and current requirements indicated in the Technical Specifications.

Administrative Control The necessary administrative procedures are addressed in the original docketed FSAR and current requirements indicated in the Technical Specifications.

Plant Security Plan Turkey Point maintains a Plant Security Plan and is issued as a separate document.

12-1 Rev. l-ll/83