Text

License Amendment Request 196 Alternative Source Term and Conforming Amendment
	ML092050277
Person / Time
Site:	Turkey Point
Issue date:	06/25/2009
From:	Jefferson W; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2009-133
	Download: ML092050277 (213)
	v • d • e

0 FPL.

POWERING TODAY.

EMPOWERING TOMORROW."

June 25, 2009 L-2009-133 10 CFR 50.90 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001 Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 License Amendment Request 196 Alternative Source Term and Conforming Amendment Pursuant to 10 CFR 50.90, Florida Power and Light Company (FPL) hereby requests an amendment to Facility Operating Licenses DPR-31 and DPR-41 for Turkey Point Units 3 and 4.

FPL proposes to revise the Turkey Point Units 3 and 4 licensing bases to adopt the alternative source term (AST) as allowed in 10 CFR 50.67 Enclosure 1 provides a description of the proposed change and supporting justification including the determination of no significant hazards and environmental considerations. The Attachments to the Enclosure are as follows: Attachment 1 provides the existing TS pages marked up to show the proposed changes. Attachment 2 provides existing TS Bases pages marked up to show the proposed changes. Attachment 3 provides Numerical Applications, Inc., NAI-1396-045, "AST Licensing Technical Report for Turkey Point Units 3 and 4," Revision 1. Attachment 4 provides a summary of the regulatory commitments made in this submittal. Attachment 5 contains a comparison of the inputs and assumptions from the Turkey Point current licensing bases with the inputs and assumptions used for the AST. Attachment 6 contains a Regulatory Information Summary (RIS) 2006-04 compliance matrix. Attachment 7 contains a list of plant modifications required to support the AST.

FPL requests approval and issuance of the proposed amendments by June 30, 2010.

The Turkey Point Plant Nuclear Safety Committee has reviewed the proposed license amendments. In accordance with 10 CFR 50.91(b)(1), a copy of the proposed amendments is being forwarded to the State Designee for the State of Florida.

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1), using the criteria in 10 CFR 50.92(c). FPL has determined that the proposed changes do not involve a

significant hazards consideration.

an FPL Group company

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Page 2 of 3 License Amendment Request 196 Alternative Source Term and Conforming Amendment Should you have any questions regarding this submittal, please contact Mr. Robert J. Tomonto, Licensing Manager, at (305) 246-7327.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on June /4_ _,2009.

Very truly yours, William Jefes Site Vice Presi nt Turkey Point Nuclear Plant Attachments cc: Regional Administrator, Region II, USNRC USNRC Project Manager, Turkey Point Nuclear Plant Senior Resident Inspector, USNRC, Turkey Point Nuclear Plant Mr. W. A. Passetti, Florida Department of Health

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Page 3 of 3 License Amendment Request 196 Alternative Source Term and Conforming Amendment TABLE OF CONTENTS ATTACHMENTS

" Enclosure 1 Alternative Source Term and Conforming Amendment

Attachment 1 Technical Specification Change Markups

Attachment 2 Technical Specification Bases Markups

Attachment 3 Numerical Applications, Inc. NAI- 1396-045, "AST Licensing Technical Report for Turkey Point Units 3 and 4," Revision 1

Attachment 4 Regulatory Commitments

Attachment 5 AST Input / Assumption Comparison Table

Attachment 6 Regulatory Information Summary (RIS) 2006-04, "Experience with Implementation of Alternative Source Terms" Compliance Table

" Attachment 7 Description of Modifications

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 1 of 40 Alternative Source Term and Conforming Amendment LICENSE AMENDMENT REQUEST LAR 196 ALTERNATIVE SOURCE TERM AND CONFORMING AMENDMENT ENCLOSURE 1

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 2 of 40 Alternative Source Term and Conforming Amendment TABLE OF CONTENTS LICENSE AMENDMENT REQUEST ALTERNATIVE SOURCE TERM AND CONFORMING AMENDMENT SECTION TITLE PAGE Cover Sheet 1 Table of Contents 2

1.0 Purpose and Scope

3 2.0 Background Information 3 3.0 Description of Proposed Changes 5 4.0 Basis/Justification of Proposed Changes 9 5.0 List of Commitments 35 6.0 Conclusion 35 7.0 No Significant Hazards Determination 36 8.0 Environmental Consideration 39 9.0 Summary of Results 40 10.0 References 40 Technical Specification Marked Up Pages 17 pages Technical Specification Bases Markups 20 pages Numerical Applications, Inc. NAI-1396-045, "AST Licensing 82 pages Technical Report for Turkey Point Units 3 and 4" Regulatory Commitments 2 pages AST Input / Assumption Comparison Table 39 pages Regulatory Information Summary (RIS) 2006-04, 8 pages "Experience with Implementation of Alternative Source Terms" Compliance Table Description of Modifications 2 pages

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 3 of 40 Alternative Source Term and Conforming Amendment

1.0 Purpose and Scope

Florida Power and Light Company (FPL) proposes to revise the Turkey Point (PTN)

Units 3 and 4 licensing bases to adopt Alternative Source Term (AST) as allowed pursuant to 10 CFR 50.67 (Reference 1). Sections 3.0 and 4.2 and Attachment 1 provide the descriptions and justifications for the Technical Specifications (TS) being revised to support the AST analyses. Attachment 2 provides the associated TS Bases, supplied for information only, to support the NRC review. Implementation of AST will allow an increase in the design basis assumption regarding control room envelope unfiltered inleakage, while maintaining the accident dose consequences within the limits specified by 10 CFR 50.67.

The input values assumed in the AST analyses bound both the current plant operation and a planned extended power uprate (EPU). This LAR is requesting approval of AST for dose consequences and is not asking for approval of a licensed thermal power uprate.

Additionally, a new TS addressing the use of sodium tetraborate decahydrate (NaTB) for containment sump pH control will be added. This TS and its associated Limiting Condition for Operation (LCO), Actions, and Surveillance Requirements provide assurance that the proper amount of NaTB is maintained in the containment to ensure that the containment sump pH is at the value assumed by the AST loss of coolant accident (LOCA) radiological consequence analysis.

2.0 Background Information The PTN Units 3 and 4 current licensing bases for the radiological consequences analyses of accidents other than the fuel handling accident (FHA) are based on source methodologies and assumptions derived from Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites" (Reference 3).

The accidents are discussed in the Chapter 14 of the Turkey Point Updated Final Safety Analysis Report (UFSAR) (Reference 2). The current PTN dose consequences of design basis events are based on the criteria stated in 10 CFR Part 100 and 10 CFR 50, Appendix A General Design Criterion (GDC) 19. The current licensing basis for the FHA radiological consequence analysis is 10 CFR 50.67, Accident Source Term, is being re-analyzed to address the planned EPU conditions.

PTN predates the NRC issuance of the 10 CFR 50 Appendix A, "General Design Criteria (GDC) for Nuclear Power Plants." The PTN GDCs identified in the UFSAR were based on the Atomic Energy Commission (AEC) 1967 proposed GDCs. However, PTN has committed to 10 CFR 50, Appendix A, GDC 19 - Control Room. The AST analyses meet the criteria stated in 10 CFR 50.67 and 10 CFR 50 Appendix A, GDC 19 - Control Room.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 4 of 40 Alternative Source Term and Conforming Amendment 10 CFR 50.67 was issued by the NRC to permit holders of facility operating licenses to voluntarily revise the traditional accident source term used in the design basis accident radiological consequence analyses with AST. 10 CFR 50.67 requires a licensee seeking to use the AST to apply for a license amendment and requires that the application include an evaluation of the consequences of affected design basis accidents. Regulatory guidance for the implementation of the AST is provided in NRC Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" and NRC Regulatory Issue Summary (RIS) 2006-04, "Experience with Implementation of Alternative Source Terms," (References 4 and 5, respectively).

NEI-99-03, "Control Room Habitability Guidance" (Reference 6) indicated that several nuclear power plants demonstrated control room unfiltered inleakage rates in excess of amounts assumed by the radiological consequence analyses during testing. To address this issue, NRC Generic Letter (GL) 2003-01, "Control Room Habitability" (Reference 7) required utilities to assess the most limiting unfiltered inleakage into the control room envelope. The PTN response to GL 2003-01 (Reference 8) confirmed that the PTN control room envelope and ventilation system meet all applicable habitability regulations and are designed, constructed, configured, operated, and maintained in accordance with the current licensing basis, as described in the PTN UFSAR.

The full implementation of the AST methodology is being used to calculate the offsite and control room radiological consequences for PTN Units 3 and 4 for events discussed in RG 1.183. Attachment 3, "NAI- 1396-045 - AST Licensing Technical Report for Turkey Point Units 3 and 4" (Reference 9). Section 2.0 describes the compliance with RG 1.183 Regulatory Positions for each of the accidents listed below.

The following limiting UFSAR Chapter 14 accidents are analyzed:

Loss-of-Coolant Accident (LOCA)

Main Steam Line Break (MSLB)

Steam Generator Tube Rupture (SGTR)

" Locked Rotor

" Rod Cluster Control Assembly (RCCA) Ejection

Fuel Handling Accident (FHA)

Spent Fuel Cask Drop, and

Waste Gas Decay Tank (WGDT) Rupture RG 1.183 does not address the Spent Fuel Cask Drop or the WGDT Rupture Events. For the Spent Fuel Cask Drop event, the FHA dose limits in RG 1.183 have been adopted. For the WGDT Rupture event, the Branch Technical Position 11-5 of the Standard Review Plan (SRP) gives dose acceptance criteria for the exclusion area boundary (EAB) to 100 mrem.

RIS 2006-04 establishes the AST dose limit to a member of the public for this event at 100 mrem TEDE. Since RIS 2006-04 associates this acceptance criteria with the annual limit established in 10 CFR 20, the EAB dose will be evaluated over the duration of the event

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure I License Amendment Request 196 Page 5 of 40 Alternative Source Term and Conforming Amendment rather than for the worst two-hour period. Branch Technical Position 11-5 does not require the dose consequences to be evaluated at the LPZ or the control room. Although no limit is given for the low population zone (LPZ), the LPZ is evaluated against this 0.1 rem TEDE limit. A TEDE limit of 5.0 rem stated in 10 CFR 50.67 is applied to the control room consistent with other AST events. All doses are evaluated for a 30-day period.

RG 1.183, Appendix A, "Assumptions for Evaluating the Radiological Consequences of a LWR Loss-of-Coolant Accident," indicates that the chemical form ofradioiodine released to the containment should be assumed to be 95% cesium iodide, 4.85% elemental iodine, and 0.15% organic iodine, if the containment sump pH is controlled at values of 7.0 or greater, consistent with the recommendation in RG 1.183.

PTN will install a Recirculation Sump pH Control System that will ensure that the pH is controlled at 7.0 or greater prior to the onset of containment spray recirculation. The system will consist often stainless steel wire mesh baskets containing NaTB in the lower regions of containment and is described in Section 4.2.10. The system will ensure that the pH is maintained at 7.0 or greater for the entire recirculation spray cycle.

3.0 Description of Proposed Changes This license amendment request (LAR) revises the PTN Units 3 and 4 licensing basis to fully implement AST, as described in RG 1.183, Section 1.2, through reanalysis of the radiological consequences of the UFSAR Chapters 11 and 14 accidents listed above.

Full implementation of AST for PTN Units 3 and 4 does not include Environmental Qualification (EQ) of safety related equipment inside the containment, as described in Section 4.2 in this LAR. RG 1.183, Positions 1.3 and 4.3 identify various licensing requirements for which compliance may have been demonstrated, in part, by the evaluation of radiological consequences of design basis accidents. Since this amendment is not requesting approval of a thermal power increase, the current source term and the PTN responses to these licensing commitments are not being revised as a result of implementation of AST remain applicable. In addition, Vital Area Access dose and shielding calculations are unchanged with the implementation of AST.

The following changes to the licensing basis are assumed in the supporting analyses:

0 The TEDE acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11 and 10 CFR 50, Appendix A, GDC 19.

New onsite (Control Room) and offsite (EAB and LPZ) atmospheric dispersion factors (X/Q) have been determined using current meteorological data following the guidance in RG 1.194 and 1.145 (Reference 12 and 13, respectively).

Dose conversion factors for inhalation and submersion use Environmental Protection Agency (EPA) Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 6 of 40 Alternative Source Term and Conforming Amendment Factors for Inhalation, Submersion, and Ingestion," and EPA Federal Guidance Report No. 12 (FGR 12) "External Exposure to Radionuclides in Air, Water, and Soil" (References 10 and 11, respectively).

Control room unfiltered inleakage limits were increased until the dose limit was approached for LOCA. This unfiltered leakage value is assumed for all other events.

A more restrictive TS primary coolant specific activity limit for DOSE EQUIVALENT 1-131 (DEI-131) is utilized.

" The primary-to-secondary SG leakage rate assumed in the accident analysis and specified in the accident induced leakage performance criterion of TS 6.8.4.j.b.2, "Steam Generator Program," is reduced from "1 gpm through all steam generators and 500 gpd through any one steam generator" [at accident conditions] to "0.6 gpm through all steam generators and 0.2 gpm through any one steam generator at room temperature conditions." The current primary to secondary operational leakage limit remains unchanged.

A more restrictive TS containment leakage is utilized.

" Containment sump pH control is provided by NaTB stored in ten stainless steel baskets inside containment.

The Containment Emergency Filtration System is no longer credited to maintain the AST dose consequence analyses within regulatory limits.

The Containment Spray System and the proposed Recirculation pH Sump Control System are being credited for post-LOCA iodine removal.

" For the Secondary RCCA Ejection, FlA within the Fuel Building, and Spent Fuel Cask Drop events, credit is taken for the operators to manually initiate the emergency recirculation mode of the Control Room Ventilation System within 30 minutes.

The Control Room Emergency Ventilation System intakes are being relocated to diverse wind sectors for post-accident contaminants.

The methyl iodide penetration criterion is more restrictive than the current TS limit for the control room emergency ventilation filter.

Accordingly, the following changes are proposed to the PTN TS. Justification for these proposed changes is provided in Section 4.2: (Attachment 1 provides the marked-up TS pages for the changes listed below.)

" TS 1.12, Definition of DOSE EQUIVALENT 1-131 is revised to reference EPA FGR 11, as the source of thyroid dose conversion factors.

TS 1.13, Definition of E-AVERAGE DISINTEGRATION ENERGY is deleted and replaced with the definition of DOSE EQUIVALENT XE-133 that includes a reference to EPA FGR 12 as the source of whole body dose conversion factors.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 7 of 40 Alternative Source Term and Conforming Amendment TS 3.4.8, Reactor Coolant System Specific Activity o LCO 3.4.8 a., RCS specific activity limit for DOSE EQUIVALENT 1-131 is reduced from "less than or equal to 1.0 microcurie per gram" to "less than or equal to 0.25 microcuries per gram."

o LCO 3.4.8 b., RCS specific activity limit is changed from "less than or equal to I 00/P microcuries per gram of gross radioactivity" to "less than or equal to 447.7 microcuries per gram DOSE EQUIVALENT XE-133."

o APPLICABILITY is changed from "MODES 1, 2, 3, 4, and 5" to "MODES 1, 2, 3, and 4."

o ACTION MODES 1, 2, and 3*, a. and b. and ACTION MODES 1, 2, 3, 4, and 5, are deleted and replaced with new ACTIONS a., b., c., d., and e.

a. With the specific activity of the reactor coolant greater than 0.25 microcuries per gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 is less than or equal to 60 microcuries per gram once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. With the specific activity of the reactor coolant greater than 0.25 microcuries per gram DOSE EQUIVALENT 1-131, but less than or equal to 60 microcuries per gram, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 0.25 microcuries per gram limit. Specification 3.0.4 is not applicable.

c. With the specific activity of the reactor coolant greater than 0.25 microcuries per gram DOSE EQUIVALENT 1-131 for greater than or equal to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or greater than 60 microcuries per gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With the specific activity of the reactor coolant greater than 447.7 microcuries per gram DOSE EQUIVALENT XE-133, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT XE-133 to within the 447.7 microcuries per gram limit. Specification 3.0.4 is not applicable.

e. With the specific activity of the reactor coolant greater than 447.7 microcuries per gram DOSE EQUIVALENT XE-133 for greater than or equal to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

o Footnote on page 3/4 4-26 is deleted.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure I License Amendment Request 196 Page 8 of 40 Alternative Source Term and Conforming Amendment o FIGURE 3.4-1, DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1giCi/gram DOSE EQUIVALENT 1-131 - This figure is deleted and replaced with "This page is intentionally left blank."

o TABLE 4.4-4, Item 1 is deleted and replaced with "NOT USED."

o TABLE 4.4-4, Item 3 is changed

" from "Isotopic Analysis for DOSE EQUIVALENT 1-131 Concentration" to "Isotopic Analysis for DOSE EQUIVALENT 1-131,"

" from "1 per 14 days" to "a) 1 per 14 days" and "b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period."

o TABLE 4.4-4, Item 5 is changed

" from "Radiochemical for R Determination" to "Isotopic Analysis for DOSE EQUIVALENT XE-133,"

" from "1 per 6 months*" to "1 per 7 days,"

o TABLE 4,4-4, Item 6, Isotopic Analysis for Iodine Including 1-131, 1-133, and 1-135 is deleted and replaced with "NOT USED."

o TABLE NOTATIONS are deleted and replaced with "THIS PAGE INTENTIONALLY BLANK."

TS 3/4.6.3, Emergency Containment Filtering System is deleted, and replaced with "Not Used."

" TS 3/4.7.5, Control Room Emergency Ventilation System Surveillance Requirement 4.7.5.c 2, the methyl iodide penetration criteria is reduced from "less than 2.5%" to "less than 1.25%."

TS 3.7.9, Gas Decay Tanks LCO is clarified to change the phrase within the parentheses from "(considered as Xe- 133 equivalent)" to "(DOSE EQUIVALENT XE-133)".

TS 6.8.4.h, Containment Leakage Rate Testing Program is changed to reduce the maximum allowable containment leakage rate acceptance criterion from "0.25%

of containment air weight per day" to "0.20% of containment air weight per day."

" TS 6.8.4.i.b.2, SG Program is revised to change the accident-induced performance criterion from "1 gpm total through all steam generators and 500 gallons per day through any one steam generator" [at accident conditions] to "0.60 gpm total through all steam generators and 0.20 gpm through any one steam generator at room temperature conditions."

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 9 of 40 Alternative Source Term and Conforming Amendment TS 6.9.1.2, Annual reports is changed to address the change in RCS specific activity discussed above. This change addresses the change in the LCO associated with the DOSE EQUIVALENT 1-131 being reduced from "1.0 microcurie per gram" to "0.25 microcuries per gram." The footnote "** This tabulation supplements the requirements of §20.2206 of 10 CFR Part 20" at the bottom of the page is deleted. There is no text on the page that requires this footnote.

The following new TS is created:

TS 3/4.6.2.3, Recirculation pH Control System, This is a new TS to control the post-LOCA containment sump solution pH using NaTB. This passive system will consist often stainless steel baskets containing NaTB in the lower regions of the containment.

4.0 Basis/Justification for the Proposed Changes 4.1 AST Compliance to Regulatory Guide 1.183 Attachment 3 Section 2.0 describes the compliance with RG 1.183 Regulatory Positions for each of the accidents being analyzed.

4.1.1 Accident Source Term The full core isotopic inventory for PTN Units 3 and 4 is determined in accordance with RG 1.183. The bounding values for the core thermal power and fuel parameters were based on a planned extended power uprate (EPU) value. The fission product inventory in the core and coolant systems available for release to the containment is based on the maximum full power operation of the core and the proposed licensed values for fuel enrichment and fuel burnup that equal or bound current licensed values. This isotopic source term bounds the licensed thermal power. This LAR is only requesting approval of AST for dose consequences at a power level above the current licensed power limit. This amendment request does not seek approval to operate at a power level beyond the current licensed power.

NAI-1396-045, "AST Licensing Technical Report for Turkey Point Units 3 and 4,"

(Attachment 3) provides the details of the LOCA and non-LOCA accident dose consequence analyses performed according to the guidelines set forth in RG 1.183. The radiation source terms for PTN are calculated as described in Attachment 3, Section 1.7.

The fission product inventory and assembly source term are based on a bounding maximum core power of 2644 MWt with a calorimetric uncertainty of 0.3% or 2652 MWt which exceeds the current licensed rated thermal power of 2300 MWt. For non-LOCA events with fuel failures, a radial peaking factor of 1.65 is applied to conservatively simulate the effect of power level differences across the core that might affect the localized fuel failures for assemblies containing the peak fission product inventory.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 10 of 40 Alternative Source Term and Conforming Amendment The primary source term methodology described in Attachment 3, Section 1.7.2 is based on maximum equilibrium concentrations from operation at 2652 MWt with defects in one percent of the fuel rod cladding. The equilibrium iodine activities were adjusted to achieve the proposed TS 3.4.8 DEI-131 limit of 0.25 microcuries per gram using the proposed TS definition for DEI- 131.

Secondary system coolant activity is limited to a value of less than or equal to 0.10 microcuries per gram DEI-131 in accordance with TS 3.7.1.4. Noble gases entering the secondary system coolant are assumed to be immediately released; thus the noble gas activity concentration in the secondary system coolant is assumed to be 0.0 microcuries per gram. As such, the secondary side iodine activity is 40% of the primary coolant activity (the ratio of 0.1 microcuries per gram to 0.25 microcuries per gram, i.e., the ratio of the two TS limits of the secondary side coolant source term to the primary side coolant source term.)

4.1.2 Dose Calculation The PTN radiological dose calculations were performed using the AST methodology applying TEDE acceptance criteria. Dose calculations follow the guidelines of Regulatory Positions cited in RG 1.183. Conformance with these regulatory positions is described in Attachment 3, Section 2.0 for each of the accidents analyzed. Offsite and control room doses meet the guidelines of RG 1.183 and the requirements of 10 CFR 50.67. RG 1.183 does not address the Spent Fuel Cask Drop and WGDT rupture events. The criteria adopted for these events are described in Section 4.1.5 below.

Analyses consider the radionuclides listed in Table 5 of RG 1.183. The chemical form of the radioiodine released to the containment is assumed to be 95% cesium iodide (CsI),

4.85% elemental iodine and 0.15% organic iodine, including gap releases and fuel pellet releases. With the exception of elemental, organic iodine, and noble gases, fission products are assumed to be in particulate form. The accident specific transport of these iodine species are described in Attachment 3 Section 2.0.

Increased values for control room unfiltered inleakage where determined by increasing the unfiltered inleakage assumption in the LOCA dose calculation until the dose limit was approached. This unfiltered leakage value was assumed for all other events.

4.1.3 Atmospheric Dispersion (X/Q) Factors X/Q factors for various release-receptor combinations were developed to replace previously approved values used for control room and offsite dose consequences. The X/Q factors were determined in accordance with the guidance provided in RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," and RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants" (References 12 and 13, respectively).

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure I License Amendment Request 196 Page 11 of 40 Alternative Source Term and Conforming Amendment The Control Room Emergency Ventilation outside air intakes are being relocated from their current locations to new locations that fall within diverse wind sectors for post-accident contaminants. See Attachment 3, Figure 1.8.1-1. The relocated intakes will be designed to seismic criteria, protected from environmental effects, and will meet the requirements of 10 CFR 50 Appendix A, GDC 19 to satisfy the operability requirements for the control room emergency ventilation system. In the emergency/ recirculation mode of operation, the intakes will have balanced intake flow rates and capable of drawing outside makeup air from both analyzed intake locations. The intakes will be located near the ground level beyond the southeast and northeast corners of the auxiliary building. In accordance with RG 1.194 (Reference 12), the dispersion factors for the most conservative intake are selected and the dual intake dilution credit allows the X/Qs to the most limiting emergency intake to be reduced by a factor of two in the event analyses for all release points in separate wind sectors. Since the control room envelope remains pressurized while in the emergency/recirculation mode, unfiltered inleakage is modeled as makeup air which bypasses the filtration system. Modeling of the control room conservatively addresses these factors, as they apply to the various release locations for each analyzed event.

A number of release-receptor combinations are considered for the onsite control room atmospheric dispersion factors. The different cases are used to determine the limiting release-receptor combination for the analyzed events. New X/Q factors are also developed for offsite receptor locations. The EAB offsite dose evaluation conservatively applies the worst case two-hour X/Q for the entire event to the worst two-hour dose contribution from different time periods. Attachment 3, Section 1.8 identifies the codes used and details the methodology and assumptions used in developing the new X/Q factors.

Meteorological data over a five-year period (2003 through 2007) is used in the development of the new atmospheric dispersion (X/Q) factors used in the analysis. This is described in detail in Section 1.8.3 of Attachment 3. The data is being provided in a separate submittal to facilitate NRC review of the models used in the analysis.

4.1.4 Assumptions and Methodologies The AST analyses performed for PTN Units 3 and 4 use assumptions and models defined in RG 1.183 to provide appropriate and prudent safety margins. Attachment 5 provides a comparison of the current dose consequence analyses inputs/assumptions to the new AST dose consequence analyses for each of the events being analyzed. The current PTN dose consequences analyses, excluding FHA, use the TITAN5 computer code. The current FHA dose consequences analysis uses the RADTRAD-NAI computer code.

Section 2.0 of Attachment 3 describes the specific assumptions for each radiological consequence analysis, the methodology, transport inputs (e.g. ventilation), removal inputs (e.g. reduction in airborne radioactivity due to containment spray), and the radiological consequences. Selected numeric input values use the limiting or bounding input for either the current power or the uprated power to assure a conservative calculated dose.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure I License Amendment Request 196 Page 12 of 40 Alternative Source Term and Conforming Amendment The following computer codes are used in performing the AST analyses:

ARCON96

Microshield

ORIGEN

PAVAN

GOTHIC

RADTRAD-NAI Attachment 3, NAI- 1396-045, Section 1.5 provides a discussion of the function of each computer code. Each of the codes has been used in previously approved implementations of AST, including the St. Lucie Unit 1 LAR and Amendment (References 14 and 15, respectively).

Overfill is not assumed to occur as a result of a steam generator tube rupture event. This design basis assumption has been previously confirmed for Turkey Point for the current licensed power conditions by a supplemental analysis using the LOFTTR2 computer code demonstrating margin to overfill following a postulated steam generator tube rupture event assuming extended operator action times for break flow termination. A similar confirmatory analysis demonstrated margin to overfill for the planned EPU conditions.

These analyses used a computer code (LOFTTR2) and followed a methodology previously approved by the NRC for the SGTR event utilizing modeling assumptions that are consistent with the Turkey Point Licensing basis (Reference 35).

4.1.5 Dose Consequences Results Analysis of the dose consequences for the LOCA, FHA, MSLB, SGTR, Locked Rotor, and RCCA Ejection Events are made using the RG 1.183 guidelines to support increasing the unfiltered control room inleakage. The analyses used assumptions consistent with the proposed changes to the PTN Units 3 and 4 licensing basis. With these changes, the resulting calculated doses meet the acceptance criteria as defined in RG 1.183.

RG 1.183 does not address the Spent Fuel Cask Drop or the WGDT Rupture Events. For the Spent Fuel Cask Drop event, the FHA dose limits in RG 1.183 have been adopted. For the WGDT Rupture event, the Branch Technical Position 11-5 of the SRP gives dose acceptance criteria for the exclusion area boundary (EAB) to 100 mrem. RIS 2006-04 establishes the AST dose limit to a member of the public for this event at 100 mrem TEDE.

Since RIS 2006-04 associates this acceptance criteria with the annual limit established in 10 CFR 20, the EAB dose will be evaluated over the duration of the event rather than for the worst two-hour period. Branch Technical Position 11-5 does not require the dose consequences to be evaluated at the LPZ or the control room. Although no limit is given to the low population zone (LPZ), the LPZ is evaluated against the 0.1 rem TEDE limit. A TEDE limit of 5.0 rem stated in 10 CFR 50.67 is applied to the control room consistent with other AST events. All doses are evaluated for a 30-day period.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure I License Amendment Request 196 Page 13 of 40 Alternative Source Term and Conforming Amendment Following the Three Mile Island (TMI) accident, additional evaluations were performed to evaluate the control room to assure operators would be adequately protected. In response to NUREG-0737, "Clarification of TMI Action Plan Requirements" Item II.B.2, FPL installed additional shielding to reduce containment shine radiation exposure from the Unit 3 Purge Valve Exhaust to the control room (Reference 31). Based on this response, the NRC concluded that the modifications were acceptable in accordance with the action plan (Reference 32). As such, the direct shine calculated as part of the AST analysis dose also includes a contribution from the Unit 3 Purge Valves.

Attachment 3, Section 1.6.4 discusses the LOCA shine dose. The 30-day direct shine dose to personnel in the control room, considering occupancy is provided in Attachment 3, Table 1.6.3-2. The total LOCA direct shine dose is 0.723 rem, which is a significant contributor to the total control room LOCA dose consequences (4.87 rem) cited in Attachment 3, Table 2.1-6. Direct shine dose is determined from four different sources to the control room operator after a postulated LOCA event. These sources are containment, containment shine through the Unit 3 purge valves, control room recirculation filters, and the external cloud that envelops the control room. All other sources of direct shine dose are considered negligible. Direct shine dose was calculated using the NRC-approved MicroShield 5 computer code. The applicable components of the LOCA shine dose are used as a conservative assessment of the direct shine dose contribution for all other accidents.

Results of the PTN Units 3 and 4 radiological consequence analyses using the AST methodology and the corresponding allowable control room unfiltered inleakage are summarized in Table 1 below. The analyses support a revised design basis maximum allowable control room unfiltered inleakage of 115 cfm, inclusive of 10 cfm allowance for control room ingress and egress.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure I License Amendment Request 196 Page 14 of 40 Alternative Source Term and Conforming Amendment Table 1 Summary of AST Analysis Results Unfiltered Exclusion Control Area Low Case Room Boundary Population Control Inleakage DoseM 1) Zone Dose12) Room Dose(2 )

(cfm) (rem TEDE) (rem TEDE) (rem TEDE)

LOCA 115 5.85 1.58 4.87 MSLB Pre-accident Iodine Spike 115 0.03 0.02 1.53 SGTR Pre-accident Iodine Spike 115 0.82 0.18 2.85 Acceptance Criteria _<

25(') _25(3) < 5(4)

MSLB Concurrent Iodine Spike 115 0.05 0.04 1.57 SGTR Concurrent Iodine Spike 115 0.28 0.07 1.10 Locked Rotor 115 0.58 0.62 1.33 Acceptance Criteria 2.5 _2.5 (3) _ 5 (4)

FHA in Containment 115 0.91 0.20 1.32 FHA in Fuel Handling Building 115 0.91 0.20 3.80 RCCA Ejection - Containment 115 0.88 0.40 2.47 Release RCCA Ejection - Secondary Release 115 0.61 0.57 1.19 (Automatic Control Room Isolation)

RCCA Ejection - Secondary Release 115 0.38 0.36 3.46 (Manual Control Room Isolation)

Spent Fuel Cask Drop* (with overall 115 0.40 0.09 2.05 DF of 200)

Acceptance Criteria 6.3() < 6.3(3) <_5(4)

Waste Gas Decay Tank* 115 0.08 0.02 0.33 Acceptance Criteria < 0. (51 < 0.1i(5) _<5(4)

(1) Worst 2-hour dose applying worst case two hour X/Q (2) Integrated 30-day dose (Refer to Attachment 3, Section 2.1.5)

(3) Accident Dose Criteria from RG 1.183, Table 6 (4) 10 CFR 50.67 (5) SRP BTP 11-5

see appropriate event summary in Reference 9 for bases of acceptance criteria 4.2 Changes to the PTN Technical Specifications The PTN Units 3 and 4 current licensing basis will be revised to fully implement AST, as described in RG 1.183, through reanalysis of the radiological consequences of the UFSAR

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure I License Amendment Request 196 Page 15 of 40 Alternative Source Term and Conforming Amendment Chapters 11 and 14 accidents listed previously. Attachment 3 Section 1.5 describes the Computer Codes used in the AST analyses and Section 1.6 describes the radiological evaluation methodology.

RG 1.183, Regulatory Position 6 states that the NRC Staff is assessing the effects of increased cesium releases on Environmental Qualification (EQ) doses to determine whether licensee action is warranted. NUREG-0933, "Resolution of Generic Safety Issues" (Reference 30) Section 3.0 Item 187 resolved the issues related to the effect of increased cesium releases on EQ doses. In this document, the staff concluded that there was no discemable risk reduction associated adopting the AST for EQ. PTN Units 3 and 4 EQ analyses will continue to be based on TID-14844. Vital Area Access dose and shielding calculations are unchanged with the implementation of AST.

4.2.1 TS 1.12 Definition of Dose Equivalent Iodine Definition of DOSE EQUIVALENT 1-131 is revised to reference EPA FGR 11, as the source of thyroid dose conversion factors.

Current TS 1.12 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microCurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance for Power and Test Reactor Sites" or Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.

Proposed TS 1.12 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using thyroid dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 1-,1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

Justification A change to the definition of DEI- 131 in TS 1.12 is proposed to reference EPA FGR 11 as the source of thyroid dose conversion factors. This change is consistent with the dose conversion factors used in the applicable dose consequence analyses. A precedent for using dose conversion factors from EPA FGR 11 is established in the St. Lucie Unit 1 LAR and Amendment (References 14 and 15, respectively).

4.2.2 TS 1.13 Definition of E-Average Disintegration Energy

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 16 of 40 Alternative Source Term and Conformin2 Amendment The definition of E-AVERAGE DISINTEGRATION ENERGY is deleted and replaced with the definition of DOSE EQUIVALENT XE-133 that includes a reference to EPA FGR 12 as the source of whole body dose conversion factors.

Current TS 1.13 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample isotopes, other than iodines, with half lives greater than 30 minutes, making up at least 95 percent of the total non-iodine activity in the coolant.

Proposed TS DOSE EQUIVALENT XE-133 1.13 DOSE EQUIVALENT XE-133 shall be that concentration of XE-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. Ifa specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors of air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

Justification The new definition for DOSE EQUIVALENT XE-133 is similar to the definition for DEI-131. The determination of DOSE EQUIVALENT XE-133 will be performed in a similar manner to that currently used in determining DEI- 131, except that the calculation of DOSE EQUIVALENT XE-133 is based on the acute dose to the whole body and considers the noble gases Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 which are significant in terms of contribution to whole body dose. Some noble gas isotopes are not included due to low concentration, short half life, or small dose conversion factor.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 17 of 40 Alternative Source Term and Conforming Amendment When P is determined using a design basis approach in which it is assumed that 1.0%

of the power is being generated by fuel rods having cladding defects and it is also assumed that there is no removal of fission gases from the letdown flow, the value of E is dominated by Xe-133. The other nuclides have relatively small contributions.

However, during normal plant operation, there are typically only a small amount of fuel clad defects and the radioactive nuclide inventory can become dominated by tritium and corrosion and/or activation products, resulting in the determination of a value of E that is very different than what would be calculated using the design basis approach.

Because of this difference, the accident dose analyses become disconnected from plant operation and the Limiting Condition for Operation (LCO) becomes essentially meaningless. It also results in a TS limit that can vary during operation as different values for F_are determined.

The current FEdefinition includes radioisotopes that decay by the emission of both gamma and beta radiation. This change will implement a LCO that is consistent with the whole body radiological consequence analyses which are sensitive to the noble gas activity in the primary coolant, but not to other non-gaseous activity currently captured in the E definition. LCO 3.4.8 specifies the limit for primary coolant gross specific activity as 1001E microcuries per gram.

This change incorporating the new definition of DOSE EQUIVALENT XE-133 in TS 1.13 using the EPA FGR 12 as the source of whole body dose conversion factors is acceptable from a radiological dose perspective since it will result in an LCO that more closely relates the non-iodine RCS activity limits to the dose consequence analyses which form their bases. This change is consistent with the dose conversion factors used in the applicable dose consequence analyses.

4.2.3 TS 3/4.4.8 RCS Specific Activity RCS specific activity limit for DEI- 131 is being reduced from "less than or equal to 1.0 microcurie per gram" to "less than or equal to 0.25 microcuries per gram" and is being changed from "less than or equal to 100/E microcuries per gram of gross radioactivity" to "less than or equal to 447.7 microcuries per gram DOSE EQUIVALENT XE-133." The changes to the LCO, ACTIONS, and SURVEILLANCE REQUIREMENTS are described below:

Current TS LCO 3.4.8 The specific activity of the primary coolant shall be limited to:

a. Less than or equal to 1.0 microcurie per gram DOSE EQUIVALENT 1-131, and

b. Less than or equal to 100/B_ microcuries per gram of gross radioactivity.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 18 of 40 Alternative Source Term and Conforming Amendment APPLICABILITY: Modes 1, 2, 3, 4, and 5 ACTION: MODES 1, 2, and 3*:

a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown in Figure 3.4-1, be in at least HOT STANDBY with average reactor coolant temperature less than 500F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and

b. With the specific activity of the reactor coolant greater than 100o microcurie per gram, be in at least HOT STANDBY with average reactor coolant temperature less than 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION: MODES 1, 2, 3 4, and 5:

With the specific activity of the reactor coolant greater than I microcurie per gram DOSE EQUIVALENT 1-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of Item 6.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.

With the average reactor coolant temperature greater than or equal to 500°F.

FIGURE 3.4.-i DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > I pCi/gram DOSE EQUIVALENT 1-131.

TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM Item I - Gross Radioactivity Determination - At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months -

MODES 1, 2, 3, 4 Item 3 - Isotopic Analysis for DOSE EQUIVALENT 1-131 Concentration.-

1 per 14 days - MODE I Item 5 Radiochemical for F Determination - 1 per 6 months* - Mode 1 Item 6 Isotopic Analysis for Iodine Including 1-131, 1-133, and 1-135 Sample And Analysis Frequency -

a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months whenever the specific activity exceeds 1 gtCi/gram DOSE EQUIVALENT 1-131 or 100/E jtCi/gram of gross radioactivity, Modes 1#,

2#, 3#, 4#, 5# and b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period.

Modes 1, 2, 3 TABLE NOTATIONS

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

Until the specific activity of the Reactor Coolant System is restored within its limits.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 19 of 40 Alternative Source Term and Conforming Amendment Proposed TS LCO 3.4.8 The specific activity of the primary coolant shall be limited to:

a. Less than or equal to 0.25 microcuries per gram DOSE EQUIVALENT 1-131, and

b. Less than or equal to 447.7 microcuries per gram DOSE EQUIVALENT XE-133.

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

ACTION:

a. With the specific activity of the reactor coolant greater than 0.25 microcuries per gram DOSE EQUIVALENT 1-13 1, verify DOSE EQUIVALENT 1-131 is less than or equal to 60 microcuries per gram once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. With the specific activity of the reactor coolant greater than 0.25 microcuries per gram DOSE EQUIVALENT 1-131, but less than or equal to 60 microcuries per gram, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 0.25 microcuries per gram limit. Specification 3.0.4 is not applicable.

c. With the specific activity of the reactor coolant greater than 0.25 microcuries per gram DOSE EQUIVALENT 1-131 for greater than or equal to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or greater than 60 microcuries per gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With the specific activity of the reactor coolant greater than 447.7 microcuries per gram DOSE EQUIVALENT XE-133, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT XE-133 to within the 447.7 microcuries per gram limit. Specification 3.0.4 is not applicable.

e. With the specific activity of the reactor coolant greater than 447.7 microcuries per gram DOSE EQUIVALENT XE-I133 for greater than or equal to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Page 3/4 4-27 This page is intentionally left blank.

TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM Item I NOT USED Item 3 Isotopic Analysis for DOSE EQUIVALENT 1-131-

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure I License Amendment Request 196 Page 20 of 40 Alternative Source Term and Conforming Amendment Sample And Analysis Frequency -

a) 1 per 14 days- Mode l and b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period -

Mode 1 Item 5 Isotopic Analysis for DOSE EQUIVALENT XE-133 - 1 per 7 days - Mode I

Item 6 NOT USED Page 3/4 4-29 This page is intentionally left blank.

Justification The proposed limit for DEI-131 is more restrictive than the limit currently in place and consistent with that assumed inthe accident dose consequences analysis. The proposed TS limit for DEI- 131 is approximately 40 times higher than the highest measured DEl-131 from previous operating cycles that contained fuel failures. Thus, reducing the RCS specific activity limit from 1.0 microcurie per gram to 0.25 microcuries per gram ofDEI-131 will not result in an undue burden on plant operation.

TS LCO 3.4.8.b replaces 100/1 with DOSE EQUIVALENT XE-133 less than or equal to 447.7 microcuries per gram. This limit is established based on the RCS activity corresponding to 1% fuel clad defects with sufficient margin to accommodate the exclusion of those isotopes based on low concentration, short half life, or small dose conversion factors and is consistent with that assumed in the accident dose consequences analysis. The primary purpose of the TS 3.4.8 LCO on RCS specific activity and its associated actions is to support the dose analyses for design basis accidents. The whole body dose is primarily dependent on the noble gas activity, not the non-gaseous activity currently captured in the E definition.

TS 3.4.8 Applicability removes MODE 5. It is necessary for the LCO to apply during MODES 1 through 4 to limit the potential radiological consequences of a SGTR or MSLB that may occur during these MODES. In MODE 5 with the RCS loops filled, the SGs are specified as a backup means of decay heat removal via natural circulation.

In this mode, however, due to the reduced temperature of the RCS, the probability of a design basis accident involving the release of significant quantities of RCS inventory is greatly reduced. Therefore, monitoring of RCS specific activity is not required. In MODE 5 with the RCS loops not filled, the SGs are not used for decay heat removal; the RCS and SGs are depressurized and primary to secondary leakage is minimal.

Therefore, the monitoring of RCS specific activity is not required. The change to modify the TS 3.4.8 Applicability to include only MODES 1 through 4 retains the necessary constraints to limit the potential radiological consequences of a SGTR or MSLB that may occur during these MODES and is therefore acceptable from a radiological dose perspective.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 21 of 40 Alternative Source Term and Conforming Amendment Figure 3.4.-1, DEI-131 Reactor Coolant Specific Activity Limit vs. Rated Thermal Power with the Reactor Coolant Specific Activity > I .ici/gram DEI-131, is deleted.

The new limit is 60 microcuries per gram DEI-131 at all power levels. Action a. has been changed to delete the reference to Figure 3.4.-i and to state greater than 60 microcuries per gram DEI- 131. The change from a graph that changes with power level to a specific value is consistent with dose consequence events that are analyzed at full-power and assume a pre-accident spike of 60 microcuries per gram DEI-131. The full power transients that allow a DEI- 131 spike are not changed.

New Actions a. through e. are incorporated to implement the new LCOs for DEI-131 and DOSE EQUIVALENT XE-133 and toreplace Figure 3.4-1 with a specific value for DEI- 131 of 60 microcuries per gram. Actions a. through c. provide the requirements for DEI- 131. Actions d. and e., provide the requirements for DOSE EQUIVALENT XE-133. The required actions and completion times are consistent with the required actions and completion times in the current TS. However, Actions c.

and e. have been revised to include a requirement to be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This requirement will ensure the unit is placed in a Mode where the TS is not Applicable.

Actions b. and d. are revised to state that LCO 3.0.4. is not applicable. This will allow entry into a Mode or other specified condition in the LCO Applicability when LCO 3.4.8 is not being met. The proposed change to Action b. would allow entry into the applicable Modes from MODE 4 (HOT SHUTDOWN) to MODE 1 (POWER OPERATION) while the DEI-131 is greater that the 0.25 microcuries per gram and less than or equal to 60 microcuries per gram and DEI-131 is being restored to within its limit. The change to Action d. would allow entry into the applicable Modes from MODE 4 (HOT SHUTDOWN) to MODE I (POWER OPERATION) while the DOSE EQUIVALENT XE-133 is greater than 447.7 microcuries per gram and the DOSE EQUIVALENT XE-133 is being restored to within its limit. This Mode change is acceptable due to the significant conservatism incorporated into the DEI-131 and DOSE EQUIVALENT XE-133 specific activity limits, the low probability of an event occurring which is limiting due to exceeding the specific activity limits, and the ability to restore transient specific excursions while the plant remains at, or proceeds to power operation.

TS Table 4.4-4, Item 5 is changed from "Radiochemical for F_Determination" to "Isotopic Analysis for DOSE EQUIVALENT XE-133." This surveillance requires a gamma isotopic analysis as a measure of the noble gas activity of the reactor coolant at least once every 7 days. The measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. The surveillance provides an indication of any increase in noble gas specific activity. Trending the results of this surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The surveillance frequency of 7 days considers the low probability of a gross fuel failure during this time. Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and 1-134, it is acceptable to include the minimum

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure I License Amendment Request 196 Page 22 of 40 Alternative Source Term and Conforming Amendment detectable activity for Kr-85 in the surveillance calculation. If a specific noble gas nuclide listed in the definition for DOSE EQUIVALENT XE-133 is not detected, it will be assumed to be present at the minimum detectable activity. The Table Notation ""

is deleted. This is an administrative change because the notation was only applicable to the F Determination which has been replaced with DOSE EQUIVALENT XE-133. TS Table 4.4-4, Item 1 "Gross Radioactivity Determination" is deleted since the measurement of gross radioactivity applied to the E Determination which has been replaced with DOSE EQUIVALENT XE-133. This is also an administrative change to conform with a DOSE EQUIVALENT XE-133 limit.

TS Table 4.4-4 Item 3 is revised to delete the word "concentration." This is an administrative change. The surveillance ensures that the specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The 14 day frequency is adequate to trend changes in the iodine activity level, considering noble gas activity is monitored every 7 days.

The Mode Applicability and frequency are unchanged from the current TS for Item 3 a). The frequency of between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change greater than or equal to 15% rated thermal power within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period is established because the iodine levels peak during this time following fuel failure; samples at other frequencies would provide inaccurate results. The sample frequency remains unchanged. The Mode Applicability is appropriate since there is a low probability of fuel failure in Modes 2 and 3.

TS Table 4.4-4, Item 6 "Isotopic Analysis for Iodine Including 1-131, 1-133, and 1-135" is deleted since this is redundant to Action a. Likewise, the sampling requirement following a 15% rated thermal power change is included in Item 3. The Table Notation

"#" is deleted. This is an administrative change because the notation was only applicable to Item 6 which has been deleted.

The proposed TS changes for Reactor Coolant Specific Activity have been previously approved for Millstone Power Station Units 2 and 3 by Reference 16.

4.2.4 TS 3/4.6.3 Emergency Containment Filtering System The Emergency Containment Filtering (ECF) System TS is deleted and replaced with Not Used.

Justification The ECF System is credited in the current PTN accident analyses for dose consequences. The ECF system is not credited in the AST analyses and no longer required to mitigate any design basis accident. ECF System iodine removal capability is no longer required since the dose analyses are within regulatory limits. As such, the ECF system does not meet the 10 CFR 50.36 criteria to for inclusion in the TS. Credit for the Emergency Containment Filtering System will be removed from the PTN Units

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 23 of 40 Alternative Source Term and Conforming Amendment 3 and 4 licensing basis. The proposed change in this LAR will delete TS 3/4.6.3, ECF System.

The ECF System changes will not have a significant impact on the PTN PRA Level 2 analysis, since the ECFs are not modeled in the PTN PRA. The Level 2 PRA results are focused on calculating Large Early Release Frequency (LERF). LERF is dominated by bypass sequences, such as core damage sequences initiated by interfacing system LOCAs or SGTRs.

The ECF System is a feature in the Severe Accident Mitigation Guidelines (SAMG).

Since the calculated doses are within regulatory limits, removal of this system does not create an adverse impact on the severe accident management program.

4.2.5 TS 3/4.7.5 Control Room Emergency Ventilation System Control Room Emergency Ventilation System Surveillance Requirement 4.7.5.c 2, is changed from "methyl iodide penetration criteria of less than 2.5%" to "methyl iodide penetration criteria of less than 1.25%."

Current TS SR 4.7.5.c.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 Regulatory Guide 1.52, Revision 2, March 1978, and analyzed per ASTM D3803 -

1989 at 30'C and 95% relative humidity, meets the methyl iodide penetration criteria of less than 2.5% or the charcoal be replaced with charcoal that meets or exceeds the stated performance requirement, and Proposed TS SR 4.7.5.c.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 Regulatory Guide 1.52, Revision 2, March 1978, and analyzed per ASTM D3803 -

1989 at 30'C and 95% relative humidity, meets the methyl iodide penetration criteria of less than 1.25% or the charcoal be replaced with charcoal that meets or exceeds the stated performance requirement, and Justification TS 4.7.5.c.2 requires a change from "methyl iodide penetration criteria of less than 2.5%" to "methyl iodide penetration criteria of less than 1.25%." The analyses assume a Control Room Emergency Ventilation Filtration organic removal efficiency of 97.5%.

The TS is being reduced to a methyl iodide penetration criteria of 1.25% to ensure that a minimum safety factor of 2 is maintained as required by GL 99-02, "Laboratory Testing of Nuclear- Grade Activated Charcoal" (Reference 17). PTN submitted a LAR and received TS amendments 205/199 (Reference 18) to address the Control Room

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 - Page 24 of 40 Alternative Source Term and Conforming Amendment Emergency Ventilation Systems. The safety evaluation report for the TS amendments required a minimum safety factor of 2 for iodine penetration. This change will ensure that the requirements of GL 99-02 and TS amendments 205/199 safety factors are maintained. Recent TS surveillance results have shown that methyl iodide penetration is historically less than 1.25%, such that reducing the methyl iodide penetration criteria from less than 2.5% to methyl iodide penetration criteria less than 1.25% will not result in a burden on plant operation.

4.2.6 TS 3.7.9 Gas Decay Tanks Gas Decay Tanks LCO is clarified to change the phrase within the parentheses from

"(considered as Xe- 133 equivalent)" to "(DOSE EQUIVALENT XE- 133)".

Current TS The quantity of radioactivity contained in each gas decay tank shall be limited to less than or equal to 70,000 Curies of noble gases (considered as Xe-133 equivalent).

Proposed TS The quantity of radioactivity contained in each gas decay tank shall be limited to less than or equal to 70,000 Curies of noble gases (DOSE EQUIVALENT XE-133).

Justification The terminology used in LCO 3.7.9 to establish the maximum curie content for noble gases is changed from "considered as Xe-133 equivalent" to "DOSE EQUIVALENT XE-133." This is an administrative change to provide clarity. The maximum curie content limit remains unchanged.

4.2.7 TS 6.8.4h. Containment Leakage Rate Testing Proaram A proposed change to the TS maximum allowable containment leakage limit in TS 6.8.4.h, will reduce the La limit from 0.25% to 0.20% of containment air weight per day.

Current TS The maximum allowable containment leakage rate, La at Pa, shall be 0.25% of containment air weight per day.

Proposed TS The maximum allowable containment leakage rate, La at Pa, shall be 0.20% of containment air weight per day.

Justification The proposed limit is more restrictive than the limit currently in place. The maximum allowable leakage rate La of 0.20% supports the value assumed in the design basis LOCA and the RCCA Ejection event analyses to determine the dose consequences.

Attachment 3 Section 2.1 discusses the radiological consequences of a design basis

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 25 of 40 Alternative Source Term and Conforming Amendment LOCA and Section 2.6 discusses the radiological consequences of the RCCA Ejection event.

The result of the most recent containment Integrated Leakage Rate Tests (ILRT) for PTN Units 3 and 4 were leak rates of 0.08% and 0.11% of containment air weight per day respectively, which is significantly less than the acceptance limit criterion of 0.1875%. The acceptance criterion is defined as 0.75 La, where in this case, La is equal to 0.25% of the containment air weight per day. By reducing the TS La value to 0.20%

of the containment air weight per day, the acceptance limit criterion now becomes 0.15%. Given the new acceptance criterion of 0.15%, the actual tested leak rates of 0.08% and 0.11% are still below the limit. Additionally, the PTN Units 3 and 4 ILRT As-Found Leak Rate data from 1982 to 1992 resulted in leak rates between 0.05% and 0.11%. A review of the historical data from ILRT and.an engineering evaluation validates the data and supports the change in the allowable limit acceptance criterion.

4.2.8 TS 6.8.4.i.b.2, Steam Generator Program The SG Program is revised to reduce the accident-induced primary to secondary leakage performance criterion from one gpm through all SGs and 500 gallons per day (gpd) through one SG at accident conditions to 0.20 gpm through any one SG and 0.60 gpm through all SGs at room temperature conditions (approximately 398 gpd at accident conditions).

Current TS Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm total through all SGs and 500 gallons per day through any one SG.

Proposed TS Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 0.60 gpm total through all SGs and 0.20 gpm through any one SG at room temperature conditions.

Justification The accident induced leakage performance criterion for primary to secondary leakage is being reduced as described above. The proposed limits are lower than the limits currently in place are consistent with the leakage rates assumed in the accident analyses, other than SGTR, in terms of total leakage rate for all SGs and leakage rate for an individual SG. The operational limit on primary to secondary leakage limit in any one SG is 150 gpd at room temperature. Therefore, there is approximately a 2 to 1 margin between the operational and the accident induced primary to secondary leakage limits.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 26 of 40 Alternative Source Term and Conformina Amendment TS 3.4.6.2, RCS Operational Leakage, defines the limit for primary to secondary leakage through any one SG as 150 gpd and is unchanged by the AST analyses. This will have no impact on plant operations. The AST analyses conservatively assume the primary to secondary leakage to be maximized at 0.20 gpm using a room temperature water density.

4.2.9 TS 6.9.1.2 Annual Reports Annual report for specific activity analysis is changed to address the changes in TS 3.4.8 RCS specific activity discussed above. The footnote ** at the bottom of the page is deleted.

Current TS (5) The time duration when the specific activity of the primary coolant exceeded 1.0 microcurie per gram DOSE EQUIVALENT 1-13 1.

- This tabulation supplements the requirements of §20.2206 of 10 CFR Part 20.

Proposed TS (5) The time duration when the specific activity of the primary coolant exceeded 0.25 microcuries per gram DOSE EQUIVALENT 1-13 1.

- The footnote is deleted.

Justification This is an administrative change to ensure that the reporting requirements agree with theapplicable TS LCO and ACTION requirements and to delete an extraneous footnote. There is no text on the page that requires this footnote.

4.2.10 TS 3/4.6.2.3 Recirculation pH Control System This is a new TS to control the pH of the post-LOCA containment sump solution using NaTB. This passive system will consist often (two large and eight small) stainless steel wire mesh baskets containing NaTB located in the lower regions of the containment.

Proposed TS See Attachment 1, Marked Up Technical Specification Pages, Insert - page 3/4 6-15.

Justification RG 1.183 Appendix A Regulatory Position 2 requires the recirculation sump pH be controlled at values of 7 or greater at the onset of containment spray recirculation mode. To achieve a pH of 7 at the onset of recirculation, NaTB basket will be used to neutralize the containment sump water. This passive system will consist often (two large and eight small) stainless steel wire mesh baskets containing NaTB located in the lower regions of the containment.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure I License Amendment Request 196 Page 27 of 40 Alternative Source Term and Conforming Amendment As stated in WCAP-16596-NP (Reference 19), a fully-hydrated form of NaTB is proposed which makes it less likely that large amounts of water will be absorbed from the potentially humid containment conditions during plant operation. If exposed to dry containment conditions, the hydrated NaTB has the potential for some loss of water.

The chemical properties of the buffer do not change as a result of the potential water loss, but the weight does decrease. The use of NaTB for post-LOCA containment sump pH control has been previously approved by the NRC in Indian Point 3 license amendment 224 (Reference 20).

FPL has performed an analysis to determine the amount of NaTB needed to maintain the post-LOCA containment sump pH greater than or equal to 7.0 following recirculation at PTN. The analysis used dissolution data contained in WCAP-16596-NP to calculate surface dissolution rates. The analysis considered the minimum and maximum quantities of boron and borated water, the time-dependent post-LOCA sump temperatures, and the minimum time to reach switchover conditions. In addition, the formation of acid from radiolysis of air and water, radiolysis of chloride bearing electrical cable insulation and jacketing, and spilled reactor coolant were considered.

The analysis determined acceptable basket configurations for obtaining a sump pH of 7.0 at the onset of the containment spray recirculation mode. The results showed that the minimum number of baskets would be two large baskets and eight small baskets.

Using this basket configuration, a parametric study was performed to determine the acceptable NaTB level required to ensure a pH of 7.0 was achieved at the onset of the containment spray recirculation mode. Since, not all of the NaTB will be dissolved at the onset of containment spray recirculation, the final long-term pH was also calculated. Table 2 provides a summary of the quantities of NaTB and resulting pH values at the onset of recirculation and long-term.

The analysis performed parametric studies to determine acceptable basket configurations that ensure a sump pH of 7.0 at the minimum time to initiate containment recirculation spray flow. The results of Case 2, provided in Table 2 below, show the minimum quantity of NaTB required. Case 2 conservatively assumed the following:

0 Minimum NaTB density - This maximizes the required number of baskets and dissolution time.

Maximum refueling water storage tank (RWST), accumulator, and RCS inventory and boric acid concentration - This maximizes the NaTB requirement and minimizes the sump pH.

Maximum chloride bearing electrical cable insulation and jacketing mass - This maximizes the quantity of strong acid generation by radiolysis, thereby, minimizing the sump pH.

Minimum sump level versus time profile - The analysis conservatively adjusted the sump level versus time profile, such that the sump level is minimized at the earliest onset of containment spray recirculation flow, thereby maximizing the

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 28 of 40 Alternative Source Term and Conforming Amendment concentrations of boric acid and strong acids in the sump.

The results of Case 3, provided in Table 2, shows that the maximum 30-day pH remains less than 8.2, which is the long-term maximum pH discussed in the following paragraphs. This case conservatively assumes the following:

Maximum NaTB density;

Minimum boric acid concentrations in the RCS and RWST;

No strong acids considered;

Maximum sump temperature;

Maximum RCS inventory in the sump - This is conservative since the assumed boric acid concentration for this case is 0 ppm;

No accumulator volume is considered;

Minimum RWST level;

All baskets are completely full of NaTB;

One additional small basket completely full of NaTB is installed in containment; The concentration of hydrogen ions is based on the relative level of acids and bases and other ions in the sump solution. For evaluating the post-LOCA sump pH, the following chemical compounds are considered:

Boron/boric acid

NaTB

Hydrochloric acid

Nitric acid Hydrochloric acid generated as a result of the irradiation of chloride bearing cable insulation and jacketing in containment and nitric acid generated by irradiation of water in the sump are determined following the guidelines of NUREG/CR-5950. The quantities of hydrochloric acid and nitric acid are determined based on the 30-day integrated post-LOCA gamma and beta doses in the containment air and sump. The minor contributions from other acidic and basic species are assumed to offset each other and are negligible in comparison to the chemicals above.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 29 of 40 Alternative Source Term and Conforming Amendment TABLE 2 Case Purpose Small Basket Large Basket Initial NaTB Dissolved pH at Long-Term No. Min/Max Height Height Mass Mass prior Recirc pH pH (ft) (ft) (Ibm) to Recirc (Ibm)

I Min 2.5 2.77 14264 4687 7.005 7.396 (min) 2 Min 1.9167 2.1867 11061 4637 7.000 7.241 (min) 3 Max 2.5 2.77 1 17034* 17034 8.105 8.108 (max)

This case conservatively assumed that an additional small basket is installed in containment.

The baskets containing the NaTB will facilitate a passive mechanism for controlling post-accident containment sump fluid pH without operator action. The two large baskets and eight small baskets are designed to contain a combined mass of greater than 11,061 lbm of NaTB, thereby ensuring that the minimum pH of the sump at the onset of spray recirculation flow is 7.0. The above quantity of NaTB will ensure that the post-LOCA containment sump pH is maintained greater than or equal to 7.0 to support the containment radioiodine specification assumptions in the LOCA radiological consequences analysis detailed in Attachment 3. A proposed change to the TS will incorporate a new TS 3/4.6.2.3, Recirculation pH Control System. This specification will ensure operability of the passive post-LOCA pH control system.

Chloride induced stress corrosion cracking (SCC) of austenitic stainless steel components is reduced if the pH of the sump is maintained greater than 7.0. Short term exposure of less than one hour to a pH less than 7.0 prior to onset of the recirculation phase will not result in significant SCC, but long term exposure of several days during the recirculation phase may result in significant damage. NUREG-0800, SRP Section 6.1.1.1 (Branch Technical Position MTEB 6-1) (Reference 21) recommends that the long term containment sump pH should be maintained between 7.0 to 9.5. PTN EQ Documentation Packages establish the minimum (7.0) and maximum (8.5) pH limits assumed for equipment qualification.

Currently at PTN, the post-LOCA sump pH is controlled by manually adding NaTB via the Chemical Volume and Control System (CVCS) in accordance with the Emergency Operating Procedures (EOPs) until the pH is adjusted above 7.2. The pH analysis shows that the pH will be adjusted above 7.0 very early into the event, well before the recommended time to prevent SCC of stainless steal components. As shown by Case 3, the maximum possible pH is 8.1. This case included an extra small basket of NaTB to demonstrate that a maximum pH value would not approach pH levels which could cause SCC. The proposed recirculation sump pH control system, therefore, will sufficiently prevent degradation of stainless steals in the containment building.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 30 of 40 Alternative Source Term and Conforming Amendment GL 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors," (Reference 22) discusses the potential for larger quantities of debris to be generated during a LOCA and the effect of more debris on mitigating systems. The installation of the ten stainless steel baskets containing NaTB will maintain the sump pH at or above 7.0. To account for minimum and maximum values that influence pH and fixed quantities of NaTB, the maximum pH can not be adjusted to 7.2. Case 3 shows the maximum pH will be increased above the current value. The increased pH will result in the generation of additional debris.

Currently, the primary source of chemical debris in containment is from aluminum in the normal containment cooler fins. To offset the effects from the increased pH, the normal containment cooler aluminum fins are being replaced with copper fins which generate less debris. The overall result of increasing the containment sump maximum pH and replacing the aluminum fins with copper fins will maintain the post-LOCA debris generation within the quantities currently reported in the PTN sump strainer design basis. The proposed changes will not increase the potential debris in the containment sump; therefore, PTN will continue to satisfactorily address Generic Safety Issue (GSI) 191, "Assessment of [Effect of] Debris Accumulation on PWR Sump Performance."

NRC Branch Technical Position MTEB 6-1 indicates that: "for a pH above 7.5, consideration should be given to hydrogen generation problems arising from corrosion of aluminum." As discussed above, the quantity of aluminum within containment will be substantially reduced with the installation of copper fins for the normal containment coolers. Their replacement will act to offset any potential increase in aluminum corrosion. However, the NRC granted approval of PTN license amendments 217/211 (Reference 23) deleting the TS for hydrogen monitors and the associated post-accident containment ventilation system. The NRC safety evaluation concluded that hydrogen combustion is a small contributor to containment failure for large/dry containment designs due to the robustness of these containment types and the probability of common sources of random ignition from electrical equipment sparks and small static electric discharges. The basis for approval of these amendments is not impacted by installation of the NaTB baskets. Since hydrogen generation is not considered a post-LOCA issue for the PTN design, the pH values that will result from implementation of this LAR are acceptable.

Appropriate TS LCO, Actions, and Surveillance Requirements are proposed for TS 3/4.6.2.3 "Containment Systems - Recirculation pH Control System." Containment sump pH control systems meet the 10 CFR 50.36 Criterion 3 threshold for inclusion within TS since they are part of a primary success path to mitigate the consequences of a design basis accident. The proposed TS, LCOs, and Actions are consistent with NUREG- 1431, "Standard Technical Specifications - Westinghouse Plants" (Reference 24). NUREG-1431 provide surveillance requirements for an active system. As such, the proposed LCO Applicability, Actions and Completion Times for an inoperable Recirculation pH Control System are consistent with other recently approved Containment Systems - Spray Additive System TS (Reference 26 and 27). The proposed Surveillance Requirements ensure that the total amount of NaTB stored in

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 31 of 40 Alternative Source Term and Conforming Amendment containment is sufficient to ensure that a sump pH of 7.0 will be achieved by the start of sump recirculation and maintain the post-LOCA sump water pH greater than or equal to 7.0 thereafter.

4.3 Control Room Envelope Inleakage Control room tracer gas testing per ASTM E-741-2000, "Standard Test Method for Determining Air Change in a Single Zone by Means of Tracer Gas Dilution" (Reference 25) was performed on the control room envelope for PTN Units 3 and 4. The results demonstrated that the control room envelope is at a positive pressure to all adjacent areas and that the outside airflow is the only source of inleakage to the envelope in the emergency pressurization mode and that this entire inleakage is filtered. Specifically, the test determined, within a 95% confidence level, that the average outside air flow exceeded the average total control room inleakage. Therefore, the control room bypass leakage assumed in the AST analyses conservatively bounds the as-tested inleakage.

4.4 Post-LOCA Iodine Removal As discussed in Attachment 3, Section 2.1.3, reduction of the airborne radioactivity in the containment by natural deposition is credited. A natural deposition removal coefficient for elemental iodine has been calculated and is credited for the sprayed and unsprayed regions.

Natural deposition for the removal of aerosols is credited for unsprayed regions only. No natural deposition for the removal of organic iodine is credited.

Containment spray provides coverage of a portion (approximately 1/3) of the containment volume. Therefore, the PTN containment atmosphere is not considered to be a single, well-mixed volume. The containment is divided into three regions, sprayed and unsprayed above the operating deck and an unsprayed region below the operating deck. The mixing rates between the regions are based on a separate sensitivity study evaluating various combinations of containment fans and sprays to produce the most conservative mixing rates. Containment spray is credited for the removal of elemental and particulate (aerosols) iodine. The elemental iodine spray removal and the particulate iodine spray removal coefficients are dependent on the calculated decontamination factors (DF), based on the maximum airborne elemental.and aerosol concentrations in the containment. Elemental iodine removal capability stops when the DF reaches 200 (approximately 2.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ). The spray aerosol removal rate is reduced by a factor of 10 when the DF reaches 50 (approximately 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ).

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 32 of 40 Alternative Source Term and Conforming Amendment Containment spray flow is automatically initiated post-LOCA following actuation of a containment high pressure signal. EOPs currently direct operators to realign the containment spray pump to take suction from the containment sump via the R14R pump discharge during switchover to cold leg recirculation. Therefore, credit for elemental and aerosol DFs is achievable based on the current plant design.

4.5 Control Room Ventilation System Isolation Redundant radiation monitors located on the control room normal air intake ensure that the emergency ventilation system has redundant means of automatic actuation for all postulated accidents having the potential of yielding doses greater than GDC 19 limits. To achieve this for the cask drop accident, FPL installed redundant safety related normal air intake radiation monitors to automatically initiate control room isolation/pressurization upon detection of high radiation (Reference 33). The addition of this automatic isolation signal ensured that LOCA remained the most severe design basis accident with regard to control room doses.

For the AST analysis, the Locked Rotor and secondary release RCCA Ejection Events exceed the control room detector setpoint within 30 seconds. An additional 30 seconds account for the signal processing, and damper closure times. TS 3.3.2 Engineered Safety Features Actuation System Instrumentation, Functional Unit 9.e ensures the operability of the control room air intake radiation monitor and associated trip setpoint surveillance criteria are maintained.

For events which do not initiate a control room isolation signal, the analyses credit operator action to initiate control room isolation/pressurization within 30 minutes. For the FHA, Spent Fuel Cask Drop, and secondary release RCCA Ejection with less than design basis fuel failure events, no credit was taken for an automatic control room isolation signal. As such, these analyses credit manual operator action to isolate the control room in 30 minutes. This action can be recognized using existing control room indications from primary and secondary system process radiation monitors and achieved from the control room within the specified time. Procedures exist to respond to these events.

Similarly, Palisades has set precedence by assuming their control room would be manually isolated and placed into emergency recirculation mode in 20 minutes for all analyzed events (Reference 34).

4.6 Disposition of Recent AST Submittals RAls A review of NRC Requests for Additional Information (RAIs) received concerning the recent AST submittals was performed. The subjects of the RAIs that could be applicable to PTN Units 3 & 4 have been reviewed and incorporated in this LAR, as appropriate.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 33 of 40 Alternative Source Term and Conforming Amendment 5.0 List of Commitments

1. FPL will relocate the Control Room Ventilation System emergency air intakes prior to implementation of AST. The relocated intakes and associated ductwork will be designed to seismic criteria, protected from environmental effects, and will meet the requirements of 10 CFR 50 Appendix A, GDC 19. The new intakes will be located near the ground level of the southeast and northeast corners of the auxiliary building and will fall within diverse wind sectors for post-accident contaminants. FPL will perform post-modification testing in accordance with the plant design modification procedures to ensure the TS pressurization flow remains adequate to demonstrate the integrity of the relocated intakes.

2. FPL will install ten (two large and eight small) stainless steel wire mesh baskets containing NaTB located in the containment basement to maintain pH during the sump recirculation phase following a Design Basis LOCA.

3. FPL will implement the requirements of the proposed Technical Specification for the Recirculation pH Control System to maintain operability of this system and ensure the suitability of the NaTB. This commitment supersedes the commitment from PTN Letters L-2009-062 and L-2009-063. (References 28 and 29, respectively) related to NaTB sampling.

4. FPL will replace the aluminum fins on the normal containment coolers with copper fins. The copper fins will maintain the post-LOCA debris generation within the quantities currently assumed in the PTN sump strainer design basis.

5. FPL will update the necessary procedures to implement the manual operator action for initiation of the emergency control room ventilation system.

6.0 Conclusion The use of the AST methodology in accordance with this LAR for the design basis accident dose consequences analyses will support increases in allowable control room unfiltered inleakage. The analyses performed in support of this LAR meet all of the applicable criteria of 10 CFR 50.67, as defined within RG 1.183.

This LAR was developed using the guidance provided in RIS 2006-04. The proposed new TS for Section 3/4.6 "Containment Systems" ensures that the containment sump pH is maintained at or above 7.0 in accordance with the LOCA AST radiological consequence analysis.

7.0 No Significant Hazards Determination The Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazards consideration exists. A proposed amendment to an operating license for a facility involves no significant hazard if operation of the facility in accordance with the

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 34 of 40 Alternative Source Term and Conforming Amendment 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.

Florida Power and Light (FPL) proposes to revise the Turkey Point Units 3 and 4 licensing basis to implement Alternative Source Term (AST), described in Regulatory Guide 1.183 (July 2000). The following limiting design bases accidents are analyzed:

Loss-of-Coolant Accident

Main Steam Line Break

Steam Generator Tube Rupture

Locked Rotor

" Rod Cluster Control Assembly Ejection

Fuel Handling Accident

Spent Fuel Cask Drop, and

Waste Gas Decay Tank Rupture NRC Regulatory Guide (RG) 1.183 does not address the Spent Fuel Cask Drop and the Waste Gas Decay Tank Rupture events. For the Spent Fuel Cask Drop event, the requirements of RG 1.183 for the Fuel Handling Accident have been adopted. For the Waste Gas Decay Tank Rupture event, the limit of 0.1 rem total effective dose equivalent (TEDE) at the exclusion area boundary (EAB) established in RIS 2006-04 has been adopted. Although no limit is given to the low population zone (LPZ), the LPZ is also evaluated against the 0.1 rem TEDE limit. A TEDE limit of 5.0 rem stated in 10 CFR 50.67 is applied to the control room. All doses are evaluated for a 30-day period.

FPL has reviewed this proposed license amendment for FPL's Turkey Point Units 3 and 4 and determined that its adoption would not involve a significant hazards consideration.

The bases for this determination are:

The proposed amendment does not involve a significant hazards consideration for the following reasons:

1. The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

AST calculations have been performed for Turkey Point Units 3 and 4 which demonstrate that the dose consequences remain below limits specified in RG 1.183 and 10 CFR 50.67. For the Spent Fuel Cask Drop and the Waste Gas Decay Tank Rupture Events which are not addressed by RG 1.183, the AST methodology has demonstrated that the dose consequences remain below the limits identified above. The AST calculations are based on the current plant design and operation as modified by the installation of a passive post-LOCA recirculation pH control system, re-location and

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 35 of 40 Alternative Source Term and Conforming Amendment redesign of the control room emergency ventilation intakes, the replacement of the aluminum normal containment cooler fins with copper fins, and for certain events, manual operator actions for initiation of control room emergency ventilation system.

These proposed changes to the plant configuration are not accident precursors for any previously evaluated accidents and support mitigation of the dose consequences of previously evaluated accidents. The proposed modification to the plant configuration will be fully qualified to the appropriate design requirements to assure their required function is available for accident mitigation and to assure the function of other equipment required for accident mitigation are not adversely impacted. The use of the AST changes the regulatory assumptions regarding the analytical treatment of the design basis accidents and has no direct effect on the probability of any accident. The AST has been utilized in the analysis of the limiting design basis accidents listed above.

The results of the analyses, which include the proposed changes to the Technical Specifications (TS), and the installation of the modifications, demonstrate that the dose consequences of these limiting events are all within regulatory limits.

TS 3/4.6.3 Emergency Containment Filtering (ECF) System has been deleted since the dose consequence analyses are within regulatory limits. A new TS is being incorporated to ensure the operability of the Recirculation pH Control System. The remaining TS changes are consistent with, or more restrictive than, the current TS requirements or established precedent. None of the affected systems, components, or programs are related to accident initiators.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes to Turkey Point Units 3 and 4 only affect those systems described above. The proposed Recirculation pH Control System is a passive system that will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed modification to the plant configuration will be fully qualified to the appropriate design requirements to assure their required function is available for accident mitigation and to assure the function of other equipment required for accident mitigation are not adversely impacted. Neither implementation of the AST methodology, establishing more restrictive TS requirements, deleting TS 3/4.6.3, nor installing the modifications described above have the capability to introduce any new failure mechanisms or cause any analyzed accident to progress in a different manner.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 36 of 40 Alternative Source Term and Conformina Amendment

3. The proposed amendment does not involve a significant reduction in the margin of safety.

The proposed implementation of the AST methodology is consistent with NRC Regulatory Guide 1.183. For the Spent Fuel Cask Drop and the Waste Gas Decay Tank Rupture Events which are not addressed by RG 1.183, the AST methodology has demonstrated that the dose consequences remain below the limits identified above.

With the exception of the deletion of TS 3/4.6.3, and the addition of the recirculation pH sump control system, the proposed TS changes are consistent with, or more restrictive than, the current TS requirements or established precedent. The proposed TS requirements and plant modifications will support the AST revisions to the limiting design basis accidents. As such, the current plant margin of safety is preserved.

Conservative methodologies, per the guidance of RG 1.183, have been used in performing the accident analyses. The radiological consequences of these accidents are all within the regulatory acceptance criteria associated with the use of the AST methodology.

The proposed changes continue to ensure that the doses at the exclusion area and low population zone boundaries and in the Control Room are within the corresponding regulatory limits of RG 1.183 and 10 CFR 50.67. The margin of safety for the radiological limits is set at or below the 10 CFR 50.67 limits. An acceptable margin of safety is inherent in these limits.

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

Based on the above discussion, FPL has determined that the proposed change does not involve a significant hazards consideration.

8.0 Environmental Consideration 10 CFR 51.22(c)(9) provides criteria for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment of an operating license for a facility requires no environmental assessment, if the operation of the facility in accordance with the proposed amendment does not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (3) result in a significant increase in individual or cumulative occupational radiation exposure. FPL has reviewed this LAR and determined that the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment. The basis for this determination follows.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 37 of 40 Alternative Source Term and Conforming Amendment Basis This change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:

1. As demonstrated in the 10 CFR 50.92.evaluation, the proposed amendment does not involve a significant hazards consideration.

2. The proposed amendment does not result in a significant change in the types or increase in the amounts of any effluents that may be released offsite. The change does not introduce any new effluents or significantly increase the quantities of existing effluents. As such, the change cannot significantly affect the types or amounts of any effluents that may be released offsite.

3. The proposed amendment does not result in a significant increase in individual or cumulative occupational radiation exposure. The proposed change is mainly analytical. The addition of stainless steel wire mesh baskets inside containment has no impact on the cumulative occupational radiation exposure. The surveillance of the amount of NaTB in the baskets will be done during a refueling outage so as to ensure that no significant dose is received. Additionally, trained operators following procedures will respond to the individual events to mitigate radiation exposures.

Therefore, the proposed amendment has no significant affect on either individual or cumulative occupational radiation exposure.

9.0 Summary of Results Results of the PTN radiological consequence analyses using the AST methodology and the corresponding allowable control room inleakage are summarized on Table 1 in Section 4.1.5.

The total amount of NaTB stored in containment is sufficient to ensure that a sump pH of 7.0 will be achieved by the start of sump recirculation and maintain the post-LOCA sump water pH greater than or equal to 7.0, thereafter. A discussion of the containment sump pH analysis is provided in Section 4.2.10.

10.0 References

1. 10 CFR, Code of Federal Regulations, Title 10 Energy.

2. Turkey Point Units 3 and 4, Updated Final Safety Analysis Report (UJFSAR).

3. TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites, March 23, 1962.

Turkey PointUnits 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 38 of 40 Alternative Source Term and Conforming Amendment

4. USNRC, Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.

5. Regulatory Information Summary (RIS) 2006-04, Experience with Implementation of Alternative Source Terms, May 7, 2006.

6. NEI 99-03, Control Room Habitability Guidance, Nuclear Energy Institute, Revision 0, June 2001 and Revision 1, March 2003.

7. Generic Letter 2003-01, Control Room Habitability, June 12, 2003.

8. J. A. Stall (FPL) Letter (L-2009-299) to USNRC Document Control Desk, Generic Letter 2003-01, Control Room Habitability - 180 Day Response, December 9, 2003.

9. NAI-1396-045, AST Licensing Technical Report for Turkey Point Units 3 and 4, Revision 1, Numerical Applications, Inc., June 18, 2009.

10. EPA Federal Guidance Report No. 11 (FGR 11), Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, 1988.

11. EPA Federal Guidance Report No. 12 (FGR 12), External Exposure to Radionuclides in Air, Water, and Soil, 1993.

12. USNRC, Regulatory Guide (RG) 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, June 2003.

13. Regulatory Guide (RG) 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Revision 1, November 1982.

14. G.L. Johnston (FPL) to NRC Document Control Desk (L-2007-085), License Amendment Request "Alternative Source Term and Conforming Amendment," July 16, 2007.

15. B.L. Mozafari (NRC) to J.A. Stall (FPL), St. Lucie Plant, Unit 1 - Issuance Of

. Amendment Regarding Alternative Source Term, November 26, 2008.

16. C. Sanders (NRC) to D.A. Christian (Virginia Electric and Power Company),

"Millstone Power Station, Units 2 and 3 - Issuance of Amendment Re: Deletion of E Bar Definition and Revision to the Reactor Coolant System Specific Activity (TAC Nos. MD8492 and MD8493)," October 27, 2008.

17. Generic Letter (GL) 99-02, Laboratory Testing of Nuclear Grade Activated Charcoal, June 3, 1999.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 39 of 40 Alternative Source Term and Conforming Amendment

18. Safety Evaluation By The Office Of Nuclear Reactor Regulation Related to Amendment No. 205 to Facility Operating License No. DPR-31 and Amendment No. 199 to Facility Operating License No. DPR-41, March 21, 2000.

19. WCAP-16596-NP, "Evaluation of Alternative Emergency Core Cooling System Buffering Agents," Revision 0, July 2006.

20. J.P. Boska (NRC) To Vice-President, Entergy Nuclear Operations, "Indian Point Nuclear Generating Unit No. 3 - Issuance Of Amendment Re: Changes to Technical Specifications to Replace Sodium Hydroxide Buffer with Sodium Tetraborate,"

June 9, 2008.

21. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 6.1.1, Engineered Safety Features Materials.

22. GL 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During design Basis Accidents at Pressurized Water Reactors, September 13, 2004.

23. License Amendments 217/211, Turkey Point Units 3 and 4 - Issuance of Amendments Regarding Deletion of Technical Specifications for Hydrogen Monitors and Post-Accident Containment Vent System (TAC Nos. MB0334 and MB0335),

December 20, 2001.

24. NUREG-143 1, Standard Technical Specifications Westinghouse Plants, Revision 3.0.

25. ASTM E-741-2000, Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution.

26. J. Pollock (Entergy - Indian Point 3) to U.S. Nuclear Regulatory Commission, NL-08-041, "Proposed Changes to the Technical Specifications Regarding Replacement of the Sodium Hydroxide Buffer with Sodium Tetraborate Decahydrate", February 28, 2008.

27. J. Boska (NRC) to Vice President, Operations (Entergy), "Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Changes to Technical Specifications to Replace Sodium Hydroxide Buffer with Sodium Tetraborate (TAC No. MD8206)," June 9, 2008.

28. L-2009-062, W. Jefferson (FPL) to NRC Document Control Desk, Turkey Point Unit 4 Response to NRC Request for Additional Information Regarding the Reponses to GL 2004 -02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors, TAC No. MC4726, March 19, 2009.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Enclosure 1 License Amendment Request 196 Page 40 of 40 Alternative Source Term and Conforming Amendment

29. L-2009-063, W. Jefferson (FPL) to NRC Document Control Desk, Turkey Point Unit 3 Response to NRC Request for Additional Information Regarding the Reponses to GL 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors, TAC No. MC4725, March 19, 2009.

30. NUREG-0933, Resolution of Generic Safety Issues.

31. R. Uhrig (FPL) to D. Eisenhut (NRC) (L-82-276), "Post-TMI Requirements Supplement 2 to Response to Generic Letter No. 82-05," July 9, 1982.

32. S. Varga (NRC) to R. Uhrig (FPL), "NUREG-0737, Item III.D.3.4, 'Control Room Habitability' Turkey Point Plant Units 3 and 4," November 25, 1983.

33. J. Williams (FPL) to S. Varga (NRC) (L-85-164), "NUREG 0737, Item III.D.3.4 Control Room Habitability," April 17, 1985.

34. M. Chawla (NRC) to M. Balduzzi (Entergy), "Palisades Plant - Issuance of Amendment Re: Alternative Radiological Source Term (TAC No. MD3087),"

September 28, 2007.

35. J. Stang (NRC) to R. Powers (Indiana Michigan Power), "Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments (TAC. Nos. MB0739 and MB0740),"

October 24, 2001.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment I License Amendment Request 196 Page lof 17 Alternative Source Term and Conforming Amendment LICENSE AMENDMENT REQUEST LAR 196 ALTERNATIVE SOURCE TERM AND CONFORMING AMENDMENT Technical Specification Change Markups TS Page ix 1-3 Insert Page 1-3a 3/4 4-26 Insert Page 3/4 4-26a 3/4 4-27 3/4 4-28 3/4 4-29 3/4 6-15 (Replaced with new TS for Recirculation pH Control System 3/4 6-16 (Text deleted) 3/4 7-17 3/4 7-25 6-17 6-18a 6-19

L-2009-133 Attachment 1 INDEX Page 2 of 17 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS (ATMOSPHERIC TYPE CONTAINMENT) 3/4.6.1 PRIMARY CONTAINMENT C ontainm ent Integrity .................................................................................... 3/4 6-1 C ontainm ent Leakage .................................................................................. 3/4 6-2 C ontainm ent A ir Locks .................................................................................. 3/4 6-4 Internal P ressure ........................................................................................... 3/4 6-6 Air T e m pe rature ........................................................................................... 3/4 6-7 Containm ent Structural Integrity .................................................................... 3/4 6-8 Containm ent Ventilation System ................................................................... 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containm ent Spray System ........................................................................... 3/4 6-12 Emerqency Containment Cooling System .......................... 3/46-14

"<--Recirculation pH Control System 13/4 6 3/4.6.3 rg,- or--P,,RIv GOrN-r^,NE^NT ,-11 TE-R'NG SvY-T. .......... ........ 7, AQ4r 3 33/46-16 3/4.6.4 CONTAINMENT ISOLATION VALVES ......................................................... 3/4 6-17 3/4.6.5 COMBUSTIBLE GAS CONTROL Hydrogen Monitors ........................................................................................ 3/4 6-19 3/4.6.6 POST ACCIDENT CONTAINMENT VENT SYSTEM .................................... 3/4 6-20 TURKEY POINT - UNITS 3 & 4 ix AMENDMENT NOS. 4-, AND43

L-2009-133 Attachment 1 Page 3 of 17 DEFINITIONS FREQUENCY NOTATION DOSE EQUIVALENT 1-13 1.12 DOSE EQUIV ALENT 1131 shall be that concnRAtRIoen of 1131 (microeGUricigram) which; alonc w9uld Pro-d-uc the camA thyroid doc'1RO1 as the quantity and iatopic of 1,3,1, 1132, 1133,111*"4, andl 1135-2 actually prcscnP-t. The thyroid dosc convc;Fr1sio facto rs uscdF for thiof calculatRIo chp-n" b theoce 10qstcd- in T-Ablc 111 4 of TID) 1111 Ca4culatinf Di~stancc F=actoArs fo-r Powcr.F-. andF- Test RcActor Sitcs" o Table E 7 of NRC Rogulator; Guide 1.109, Rovision !, Octobor 19A77.

E AVERAGE D -IRINTEGRATIO ENERGY Insert B 1.1 ;;

shalpb tho vorago -(Awleghtedin propoWrio to the conccntration of oacR-h ra~d-ionuc-idse in tile roactoFrcoolant at thc timc of sampling) of the sum of thc a1rg bt n gamma ocgo o disintcgratien (McVid) for thc radionuclidcs OR the samplc isetepcs, othcr than icdinac, wthlflives groator than 230 m R -ntee, making uip at least 06 porcont of tho toal no dn ct'iyi tccoat 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GAS DECAY TANK SYSTEM 1.15 A GAS DECAY TANK SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System off gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System (primary-to-secondary leakage).

TURKEY POINT - UNITS 3 & 4 1-3 AMENDMENTNOS. 2-,AND228I

L-2009-133 Attachment 1 Page 4 of 17 INSERT A 1.12 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram).

that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using thyroid dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

INSERT B DOSE EQUIVALENT XE-1 33 1.13 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-1 31 m, Xe-33m, Xe-1 33, Xe-1 35m, Xe-1 35, and Xe-i 38 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-1 33 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

JINSERT Page 1-3al

L-2009-133 Attachment 1 Page 5 of 17 REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 0.25 microcuries 3.4.8 The specific activity of the primaryant shall be limited to:

a. Less than or equal to 1.0 microcuric per gram DOSE EQUIVALENT 1-131, and

b. Less than or equal to microcuries per gram APPLICABILITY: MODES 1,2,4 and-6. 447.7 DOSE EQUIVALENT ACTION: and XE-1 33 MOIDES 1,2 and 3*:

With the spesific arstivity of the rcactor coolant grcater than 1 m:iGrzzuriz pcr gram: DOSE EQUIVALENT 11431 for mA-re th.;n 48 h9urc during onc eontinucuc, timzi interY8I or eXcczding the limit uiRi on

,nFv 3. 1 1,* robto bhe i otRant HtoW STA4DBY with evcnagc,, aet9t ze,, aR,t temperaturz less than 6002F within 6 heurz; and With the ppeifif aativity oreactor fcoolant the gseater tha 4004e n AMmits p ram, be in at

'part HO-T ST.AISNDY wAith ayerage rcaetcr ecclant tempraturc less the n 6002Fwthn6hu.

-,---e programe~t hn1WTbe444 analysis Fgan effntesampling of and Add INSERT from next page

With* the............ ..... ir t..mp

....... at .. ...e than or 9F '--'^a, to 60 2F TURKEY POINT - UNITS 3 & 4 3/4 4-26 AMENDMENT NOS. 4-7 AND 4-a

L-2009-133 Attachment 1 Page 6 of 17 INSERT

a. With the specific activity of the reactor coolant greater than 0.25 microcuries per gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 is less than or equal to 60 microcuries per gram once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. With the specific activity of the reactor coolant greater than 0.25 microcuries per gram DOSE EQUIVALENT 1-131, but less than or equal to 60 microcuries per gram, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 0.25 microcuries per gram limit.

Specification 3.0.4 is not applicable.

c. With the specific activity of the reactor coolant greater than 0.25 microcuries per gram DOSE EQUIVALENT 1-131 for greater than or equal to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or greater than 60 microcuries per gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With the specific activity of the reactor coolant greater than 447.7 microcuries per gram DOSE EQUIVALENT XE-133, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT XE-133 to within the 447.7 microcuries per gram limit. Specification 3.0.4 is not applicable.

e. With the specific activity of the reactor coolant greater than 447.7 microcuries per gram DOSE EQUIVALENT XE-133 for greater than or equal to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

INSERT Page 3/4 4-26a

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% RATED THERMAL POWER FIGURE 3.4.-l DOSE UIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERC T OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC 0 ACT ITY >1 jaCi/gram DOSE EQUIVALENT 1-131.

TURKEY POINT - UNITS 3 & 4 3/4 4-27 AMENDMENT NOS. 137 AND 132

L-2009-133 Attachment 1 Isotopic Analysis for DOSE Page 8 of 17 EQUIVALENT XE-133 I TABLE 4.4-4

-4 C

REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM m

TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE 0 AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED z

Z 1. Gracc, Radiiaacivit' At l4-eatene.e e- 4,-2,34~

C-I [Determination. 72-hBUFe z

cn b) One sample between

2. Tritium Activity 1 per 7 days. 1,2,3,4 Determination 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER

3. Isotopic Analysis for ther 14 days. 1 DOSE EQUIVALENT 1-131 change exceeding 15%

GeRee~teaieR of the RATED

4. Radiochemical Isotopic Monthly THERMAL POWER 1,2,3,4 90 Determination Including within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period.

Gaseous Activity

5. Radiazhemkcal f8r E 1 Detarminateeiz

6. a) Oncz per 4 heUrz, !#,2#f,3#f,4#i, Ied;.e Inellii . I 4131,-

, 4 n*d34 36 a.*1 1 NOT USED speeifie aetllt ecrcezds 1 lC i/gramn DO6E EQUIVALENT m

z 0

m z

--i z b. Gnz cample bet'::zon 1,-2, 0 2Pan d-6 h.A-uWc felloeW igaTHEiRMAL POWER changzxcczing 15%

af th-o R.ATED THERMAL z POWER vithon a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

L-2009-133 Attachment 1 Page 9 of 17 Tal 4. 4 (Gentenwe

Smo'p to hp *bkPnoftpr Rminmum ; ef 2 -F-PD And 20 days ef POWEF!R OPERATION have elaozcd shnee Feactpr was IJCI SUBGREitiai f8 913 felUFS eF ImelgeR Im i i d'i *l faD =in = .d i= =*

ITUnti theZ CPZi~eem aeflitv O the rmcaroto U98oiant bysiem IS Fe6!9F8 Wtnhn its imt I THIS PAGE INTENTIONALLY LEFT BLANK.

TURKEY POINT - UNITS 3 & 4 3/4 4-29 AMENDMENT NOS. 4-&7-AND 4-3

This page replaced by Insert on next page. AhL-2009-133 Agt e c1 0mnPa of 17 CONTAINMENT SYSTEMS 3/4. EMERGENCY CONTAINMENT FILTERING SYSTEM IN DITION FOR OPERATION 3.6.3 Three em ency containment filtering units shall be OPERABLE.

APPLICABILITY: MO S 1, 2, 3, and 4.

ACTION:

7days or be in at least HOT STA lY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and thein inoper CO .* SHUTDOWN With 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ýr t/yscontainmeT filtering iuii one emergency unit inoperable, restore e filter to OPERABLE status within within the following SURVEILLANCE REQUIREMENTS 4.6.3 Each emergency containment filtering unit all b emonstrated OPERABLE:

a. At least once per 31 days on a STA ED TEST BASIS by initiating, from the control room, flow through the HEPA filters and arcoa dsorbers-and verifying that the system operates for at least 15 minutes;

b. At least once per 18 month r (1) after any struc ral maintenance on the HEPA filter or charcoal adsorber housings, or (2 ollowing operational exp ure of filters to effluents from painting, fire, or chemical release or ) after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of sy m operation by:

1) Performa e of a visual inspection for foreign mat 'al and gasket deterioration, and verifyin hat the filtering unit satisfies the in-place p etration and bypass leakage testi acceptance criteria of greater than or equal to 9 06 removal of DOP and ha genated hydrocarbons at the system flow rate of 37, cfm +/-10%;

2) Verifying within 31 days after removal, that a laboratory analys of a representative carbon sample obtained in accordance with applicable portions o:egulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and perfo ed in accordance with ASTM D3803-1989 at 300C and 95% relative humidity, meets th ethyl iodide penetration criteria of less than 35% and that any charcoal failing to me this criteria be I replaced with charcoal that meets or exceeds the stated performance req ment; and I

3) Verifying a system flow rate of 37,500 cfm +/-10% and a pressure drop across t HEPA and charcoal filters of less than 6 inches water gauge during system operation wh tested in accordance with ANSI N51 0-1 975; TURKEY POINT - UNITS 3 & 4 3/4 6-15 AMENDMENT NOS. 205 AND 199

L-2009-133 RAttachment 1 Page 11 of 17 CONTAINMENT SYSTEMS 3/4.6.2.3 RECIRCULATION PH CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.3 The Recirculation pH Control System shall be OPERABLE.

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

With the Recirculation pH Control System inoperable, restore the buffering agent to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.3 The Recirculation pH Control System shall be demonstrated OPERABLE:

a. At least once per 18 months by:

1. Verifying that the buffering agent baskets are in place and intact;

2. Collectively contain > 11061 pounds (227 cubic feet) of sodium tetraborate decahydrate, or equivalent.

L-2009-133 Attachment 1 Page 12 of 17 CONTAINMENT SYSTEMS 3/4.6.3 Not Used SVEILLANCE REQUIREMENTS Continued) c fter maintenance affecting flow distribution, by performance of a visual inspec and an air d itest at a system flow rate of 37,500 cfm +/-10%;

d. At least on er 18 months by:

1) Verifying th e system starts on a Safety Inject' est signal and;

2) Verifying that the filte oling solenoid v s can be opened by operator action and are opened automatically on ss of f* ignal.

e. After each complete or partial repla en a HEPA filter bank, by performance of a visual inspection for foreign material a gasket dete ation and by verifying that the filtering unit satisfies the in-place penet on and bypass leaka esting acceptance criteria of greater than or equal to 99% remov f DOP test aerosol while ope g the system at a flow rate of 37,500 cfm +/-1 0%; and*

f. After eac mplete or partial replacement of a charcoal adsorber ba by performance of a visua spection for foreign material and gasket deterioration and by yeni that the filtering

atisfies the in-place penetration and bypass leakage testing acceptance

nia of greater han or equal to 99% removal of halogenated hydrocarbon while operating the sys at a flow rate of 37,500 cfm +/-1 0%.

TURKEY POINT - UNITS 3 & 4 3/4 6-16 AMENDMENT NOS. 447 AND 43a

L-2009-133 Attachment 1 Page 13 of 17 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

1) Verifying that the air cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of greater than or equal to 99% DOP and halogenated hydrocarbon removal at a system flow rate of 1000 cfm +/-1 0%.

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 Regulatory Guide 1.52, Revision 2, March 1978, and analyzed per ASTM D3803 - 1989 AT 30 0 C and 95% relative humidity, meets the methyl iodide penetration criteria of less than or the charcoal be replaced with charcoal that meets or exceeds the stated perf ance requirement, and 1 f25%

3) Verifying by a visual inspection the absence of foreign materials and gasket deterioration.

d. At least once per 12 months by verifying that the pressure drop across the 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 NOS. 206 AND 9

L-2009-133 Attachment 1 Page 14 of 17 PLANT SYSTEMS GAS DECAY TANKS LIMITING CONDITION FOR OPERATION 3.7.9 The quantity of radioactivity contained in each gas decay tank shall be limited to less than or equal to 70,000 Curies of noble gases (ecnciderzd as v^ 33 -- .. ,alent).

APPLICABILITY: At all times.

""* ý(DOSE EQUIVALENT XE- 133).

ACTION:

a. With the quantity of radioactive material in any gas decay tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Annual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.4.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.9 The quantity of radioactive material contained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank and the Reactor Coolant System total activity exceeds the limit of Specification 3.4.8.

TURKEY POINT - UNITS 3 & 4 3/4 7-25 AMENDMENT NOS. AND 1-82

L-2009-133 Attachment 1 Page 15 of 17 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

9. Limitations on the annual and quarterly doses to a member of the public from iodine-1 31, iodine-1 33, tritium, and all radionclides in particulate form with half lives greater than 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

10. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

g. Deleted

h. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, and as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following deviations or exemptions:

1) Type A tests will be performed either in accordance with Bechtel Topical Report BN-TOP-1, Revision 1, dated November 1, 1972, or the guidelines of Regulatory Guide 1.163.

2) Type A testing frequency in accordance with NEI 94-01, Revision 0, Section 9.2.3, except:

a) For Unit 3, the first Type A test performed after the November 1992 Type A test shall be performed no later than November 2007.

b) For Unit 4, the first Type A test performed after October 1991 shall be performed no later than October 2006.

3) A vacuum test will be performed in lieu of a pressure test for airlock door seals at the required intervals (Amendment Nos. 73 and 77, issued by NRC November 11, 1981).

The peak calculated containment interval pressure for the design basis loss of coolant accident, Pa, is 49.9 psig.

The maximum allowable containment leakage rate, La, at Pa, shall be

of containment air weight per day.

Leakage Rate acceptance criteria are:

1) The As-found containment leakage rate acceptance criterion is < 1.0 La. Prior to increasing primary coolant temperature above 200OF following testing in accordance with this program or restoration from exceeding 1.0 La, the As-left leakage rate acceptance criterion is <0.75 La, for Type A test.

2) The combined leakage rate for all penetrations subject to Type B or Type C testing is as follows:

TURKEY POINT - UNITS 3 & 4 6-17 AMENDMENT NOS. 2-20 AND 2-25

L-2009-133 Attachment 1 Page 16 of 17 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed1f-epn total through all SGs and 50e MIe.AS Rper day through any one SG. 10.60 gm 3 The operational lea age performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant S tem Operational Leakage."

c. Provisions for SG tube repai criteria. Tubes found by inservice inspection to contain flaws with a depth equal to o exceeding 40% of the nominal tube wall thickness shall bge plugged.

The following alternate tube r pair criteria may be applied as an alternative to the 40%

depth based criteria:

1. For Unit 3 during Ref ling Outage 23 and the subsequent operating cycles until the next scheduled in pection, and for Unit 4 during Refueling Outage 23 and the subsequent operating ycles until the next scheduled inspection, flaws found in the portion of the tube elow 17 inches from the top of the hot leg tubesheet do not require plugging.

For Unit 3 during Refu ling Outage 23 and the subsequent operating cycles until the next scheduled ins ection, and for Unit 4 during Refueling Outage 23 and the subsequent operating dles until the next scheduled inspection, all tubes with flaws identified in the p rtion of the tube within the region from the top of the hot leg tubesheet to 17 inc s below the top of the tubesheet shall be plugged.

at room temperature conditions TURKEY POINT - UNITS 3 & 4 6-18a AMENDMENT NOS. 23-AND 22-9

L-2009-133 Attachment 1 Page 17 of 17 ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC pursuant to 10 CFR 50.4.

STARTUP REPORT 6.9.1 .1 A summary report of plant startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.

The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions of characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. Subsequent Startup Reports shall address startup tests that are necessary to demonstrate the acceptability of changes and/or modifications.

Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS*

6.9.1.2 Annual Reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.

Reports required on an annual basis shall include:

The results of specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Fuel burnup by core region; (3) Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) History of degassing operations, if any, starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; and (5) The time duration when the specific activity of the primary coolant exceeded .4- mire'zre per gram DOSE EQUIVALENT 1-131.

i0.25 microcuriesl

A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

- Thic tabulatin cupp19mAnt9 tho roguiromontc of 620.2206 of 10 CFR Part 20.

TURKEY POINT - UNITS 3 & 4 6-19 AMENDMENT NOS. 2a AND-224

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 2 License Amendment Request 196 Page 1 of 20 Alternative Source Term and Conforming Amendment LICENSE AMENDMENT REQUEST LAR 196 ALTERNATIVE SOURCE TERM AND CONFORMING AMENDMENT Technical Specification Bases Markups (Information Only)

TS Bases Page 10 90 55 99 57 Insert Page 99a 60 101 61 120 62 67 68 Insert Page 68a Insert Page 68b 84 89 Insert Page 89a

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0-ADM-536 Technical Specification Bases Control Program 2/18/09 L-2009-133 ENCLOSURE 1 Attachment 2 (Page 3 of 4) Page 2 of 20 INDEX BASES SECTION PAGE 3/4.6 Containment Systems 3/4.6.1 Primary Contain t ................................................................ 83 3/4.6.2 Depress ation and Cooling Systems .......................................................... 87 3/4.6.3 Aie-ee etimn itrigSse ........................... 9 3/4.6.4 Containm ent Isolation Valves ......................................................................... 90 3/4.7 Plant Systems 3/4.7.1 T urbine Cycle ................................................................................................ 91 3/4.7.2 Component Cooling Water System ............................................................... 96 3/4.7.3 Intake Cooling Water System ........................................................................ 96 3/4.7.4 U ltim ate H eat Sink .......................................................................................... 97 3/4.7.5 Control Room Emergency Ventilation System .............................................. 97 3/4.7.6 Snubbers ......................................................................................................... 98 3/4.7.7 Sealed Source Contam ination ......................................................................... 99 3/4.7.8 Explosive G as M ixture ................................................................................... 100 3/4.7.9 G as Decay Tanks ............................................................................................ 100 3/4.8 Electrical Power Systems 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. Sources, D.C. Sources, and Onsite Power Distribution ............................................................. 100 W2003:DPS/ln/nw/In

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0-ADM-536 Technical Specification Bases Control Program 2/18/09 L-2009-133 ATTACHMENT 1 Attachment 2 (Page 44 of 111) Page 3 of 20 TECHNICAL SPECIFICATION BASES 3/4.4.5 (Cont'd)

SG tubing is subject to a variety of degradation mechanisms. SG tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 6.8.4.j, Steam Generator (SG) Program, requires that a program be

,established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.8.4.j, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. The SG performance criteria are described in Specification 6.8.4.j. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1)..

Applicable Safety Analysis tem pe ature on The steam generator tube rupture (SGTR) accident is the limiting desi basis event for SG tubes and avoiding a SGTR is the basis for this Specification. The alysis of a SGTR event assumes a bounding primary-to-secondary leakage rate equal to for each of the two intact SGs plus the leakage rate associated with a double-ended rupture of a single tube in the third (ruptured) SG. The accident analysis for a SGTR assumes the contaminated secondary fluid is released to the atmosphere ,4d safety valves Of atmospheric dump valves. No credit for iodine removal is taken for an/ steam released to he condenser prior to reactor trip and concurrent loss of offsite power.

The analysis for design basis accidents and transients other than a SG"R assume the SG tubes retain their structural integrity (i.e., they are assumed not to rup.ure). In the dose consequence analysis for these events the activity level in the steam discharged to the "atmosphere is based on a primary-to-secondary leakage rate of -l-gpm to al through all SGs and/* ga, pe*or- &)y through any one SG at eeide 'fl ditions,' or is assumed to inc ase to these levels as a result of accident induced nditi s. For accidents that do not nlssumedfuel irolve to be equal to damage, thethe LCO 3.4.8, primary coolantReactor activityCoola levelof D SE EQUIVALENT 1-131 is t S/stem Specific Activity, limits.

or accidents that assume fuel damage, the primary ool it activity is a function of the amount of activity released from the damaged fuel. T d se consequen es of these events are within the limits of GDC 19 (Ref. 2), . , 10 CFf 50.67 (Ref. 7) or the NRC approved licensing basis. 0.60 Steam generator tube integrity satisfies Criterion 2 of 0 CFR 50.36(c)(2 (ii).

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0-ADM-536 Technical Specification Bases Control Program 2/18/09 L-2009-133 ATTACHMENT 1 Attachment 2 (Page 46 of 111) Page 4 of 20 TECHNICAL SPECIFICATION BASES 3/4.4.5 (Cont'd)

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5). 0.20 gpm 06 The accident induced leakage perfo nce criterion ensures that the primary-to-secondary 0.60 leakage caused by a design basis cident, other than a SGTR, is within the accident nalysis assumptions. The acciden nalyses assume that accident leakage does not exceed 4- gpm total through all SGs and 5 0 gallons per day through any one of the three SGs at The accident induced leakage rate includes any primary to secondary Stconditions.

leaka e e nor during the accident."----... to theaccident in addition to primary to secondary leakage induced room temperature The operational leakage performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational leakage is contained in LOC 3.4.6.2 and limits primary-to-secondary leakage through any one SG to 150 gpd at room temperature. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of leakage is due to more than one crack, the cracks are very small, and the above assumption is conservative.

Applicability SG tube integrity is challenged when the pressure differential across the tubes is large.

Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

Reactor Coolant System conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary-to-secondary differential pressure is low, resulting in lower stresses and reduced potential for leakage.

Actions The ACTIONS are modified by a Note clarifying that the ACTIONS may be entered independently for each SG tube. This is acceptable because the ACTIONS provide appropriate compensatory actions for each affected SG tube. Complying with the ACTIONS may allow for continued operation, and sihsquen affected SG tubes are governed by subsequent ACTION entry and application any additional W2003:DPS/In/nw/ln

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0-ADM-536 Technical Specification Bases Control Program 2/18/09 L-2009-133 ATTACHMENT 1 Attachment 2 (Page 49 of 111) Page 5 of 20 TECHNICAL SPECIFICATION BASES 3/4.4.5 (Cont'd)

References

1. NEI 97-06, Steam Generator Program Guidelines

2. 10 CFR 50 Appendix A, GDC 19 3.

4. ASME Boiler and Pressure Vessel Code, Section 111, Subsection NB

5. Draft Regulatory Guide 1.121, Bases for Plugging Degraded PWR Steam Generator Tubes, August 1976

6. EPRI Pressurized Water Reactor Steam Generator Examination Guidelines

7. 10 CFR 50.67, Accident 3/4.4.6 Reactor Coolant System Leakage Source Term 3/4.4.6.1 Leakage Detection Systems The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary to the containment. The containment sump level system is the normal sump level instrumentation. The Post Accident Containment Water Level Monitor - Narrow range instrumentation also functions as a sump level monitoring system. In addition, gross leakage will be detected by changes in makeup water requirements, visual inspection, and audible detection. Leakage to other systems will be detected by activity changes (e.g., within the component cooling system) or water inventory changes (e.g., tank levels).

Background

Components that contain or transport the coolant to or from the reactor core make up the Reactor Coolant System (RCS). Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant Leakage, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational Leakage LCO is to limit system operation in the presence of Leakage from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of leakage.

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0-ADM-536 Technical Specification Bases Control Program 2/18/09 L-2009-133 ATTACHMENT 1 Attachment 2 (Page 50 of 111) Page 6 of 20 TECHNICAL SPECIFICATION BASES 3/4.4.6.2 Operational Leakage 10 CFR 50, Appendix A, GDC (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant leakage. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS leakage varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant leakage into the containment area is necessary. Quickly separating the IDENTIFIED LEAKAGE from the UNIDENTIFIED LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the RCPB from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions. from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

Applicable Safety Analyses The primary-to-secondary leakage safety analysis assumption for individual events varies.

The assumption varies depending on whether the primary-to-secondary leakage from a single steam generator (SG) can adversely affect the dose consequences for the event. In 0.20 gpm at room _w.hh case, the affected SG is assumed to have the maximum allowable leakage (-SW temperature ). Collectively, however, the safety analyses for events resulting in steam discharge to the atmosphere assume that primary-to-secondary leakage from all steam generators (SGs) is .--gpm total and 500 galle per aythrough any one SG aeeident requirement to im'it rima y-to-secondary leakage throug anyone SG to less than or equal For Control Room to 150 gpd at room em erature is significantly less than the conditions assumed in the doses, primary-to- safety analysis. 0.0ROCA Ejection secondary age is a factor in the dose releases o tside containment resulting Sfrom a accident. To a lesser extent, other acci nts or transients involve initially via the secondary steam release to the atmosphere, such as a SG tub rupture (SGTR). The condenser SJAE leakage contaminates the secondary fluid.

and then The UF 3 analysis for SGTR assumes the contaminat secondary fluid is released to the atmosp via the atmospheric dump valves and/o main steam safety valves for a limited period of time. Operator action is taken to isola the affected SG within the time period. Te50glei e-dyp-m~ oseedp d fe

........ "t ....... 0.20 gpm at room temperature conditions W~2003:DPS/ln/nw/ln

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O-ADM-536 Technical Specification Bases Control Program 2/18/09 L-2009-133 ATTACHMENT 1 Attachment 2 (Page 51 of 11) Page 7 of 20

tion TECHNICAL SPECIFICATION BASES 3/4.4.6.2 Cont'd ROCA Ejection cidents for which e radiation dose release path is primary-to-secondary leakage, the

... frtor- ac i t is more limiting for site radiation dose releases. The safety analysis for the e accident assumes that primary-to-secondary leakage from all SGs is 4-gpm total. The dose consequences resulting from the leeked r- accident are well ithln the limits defined in . G. ý2 .!. A e 0 . -mall 060 ~fraetien efthese limits).10FR567 The RCS operational leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Limiting Condition for Operation (LCO) i j.

RCS operational leakage shall be limited to:

a. Pressure Boundary Leakage No PRESSURE BOUNDARY LEAKAGE is allowed, being indicative of material deterioration. Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage. Violation of this LCO could result in continued degradation of the RCPB. Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b. Unidentified Leakage One gallon per minute (gpm) of UNIDENTIFIED LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the leakage is from the pressure boundary.

c. Identified Leakage Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because leakage is from known sources that do not interfere with detection of UNIDENTIFIED LEAKAGE and is well within the capability of the RCS Makeup System. IDENTIFIED LEAKAGE includes leakage to the containment from specifically known and located sources, but does not include PRESSURE BOUNDARY LEAKAGE or controlled reactor coolant pump seal leak-off (a normal function not considered leakage). Violation of this LCO could result in continued degradation of a component or system.

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0-ADM-536 Technical Specification Bases Control Program 2/18/09 L-2009-133 ATTACHMENT 1 Attachment 2 (Page 56 of 11) Page 8 of 20 TECHNICAL SPECIFICATION BASES 3/4.4.7 Chemistry The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits.

The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

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0-ADM-536 TechnicalSpecification Bases Control Program 2/18/09 L-2009-133 ATTACHMENT 1 Attachment 2 (Page 57 of 111) Page 9 of 20 TECHNICAL SPECIFICATION BASES The sample analysis for: determfininig the groess specific activity, and emca exclude the r-adiojodinies because of the low r~eaetor: coolanit limnit of I .. r.ui/ga DOS EQUIALEN I 11, and because, if the limfit is exceeeded, the r-adioiodinie level is tob detenmined ever;y I hour~s. If the gfross specific aetivit,' level and r~adiciodine level in h fefetef eela.t we at thek limint-the eesihtiee

adi*e*ie.,ewould

e appoxiately 1%4. in a r~elease of r-eactor- coolant with a typical mfix~tur-e of r-adioactiviy, th;eatual radioiodinie contr-ibution wouild probably be about 204. The &exlusiono radionucl ides with half lives less than 30 minutes fromf these detem~fiinations has beeni made for- sever-al r.easons. The first consideration is the difficulty to ideni6' short lived radionulides in a sample that requires a significant timie to collect, tranIspot, and analyze The second considerationi is the pr-edictable delay time betw~een the postulated release of radioactivity from the r-eactor coolant to its r~elease to the enivir-onment and transport to the SITE BO"IDARY, which is relatable to at least 30 minuttes decaty time. The choice of 30 minutes for the half life cutoff was made beeause of the nuclearf chearateristics of the ty~pical r-eactor coolant r~adioacti-vity.

Based upont the above eonisiderations for &exluding ert~aint radionuelides from the sanmple anialysis, the allowable time of 2 hour-s betweent sample taking anld completing the inta anal'sis is based up on a ty~pical time necessar~y to perfof~f the sampling, tr-ansport the sample, and per-form the antalysis of about 90 minutes. After 90 minutes, the groess count sheuld be made in a r..eproducible geom.etr.y f sample and counter hav..ing reproducible beta or gammae self shielding properis The counter: should be reset to a r~eproeducible efficiency, ver-sus enier-gy. it is not necessar-y to ideniti' specific nuelides. The radio"hemical detennination* of nuelides should be based on multip.le:con sam.ple w4hith typical counting bai follwing sampling of less than 1 houra,about 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months sm, about 1 day, about 4 weeke, and about 1 month.

Reducing Tavg to less than 5009F prevents the release ofactivity should a steamf generator tube rcupture sine the saturation pressiue of he rettacr eoolant is below the oia preue to the atmospher-ic steam relief valves. The SurvNeillance Requirements provNide adequat-e asrncne that &exessive specific activity levels in the reactor coolant will be detectcdi sufficientt time to take eef-etive action. A r-eductioni in frequencey of isotopic anaflyses follwing po'wer- chaniges may be permissible ifjustified by the data obtained 3/4.4.9 Pressure/Temperature Limits All eomponents in the RCS are designed to withstand the effects of eyclic loads due to system temperature and pressure changes. These cyclic loads are induced by normal load transients, reactor trips and startup and shutdown operations. During RCS heatup and cooldown, the temperature and pressure changes must be limited to be consistent with design assumptions and to satisfy stress limits for brittle fracture.

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INSERT L-2009-133 Attachment 2 Page 10 of 20 3/4 4.8 Specific Activity The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 50.67. Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the dose consequences in the event of a steam line break (SLB), steam generator tube rupture (SGTR), rod cluster control assembly (RCCA) ejection, or locked rotor accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133. The allowable levels are intended to ensure that offsite and control room doses meet the appropriate SRP acceptance criteria.

The safety analyses assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator (SG) tube leakage rate of 0.20 gpm at room temperature exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 pCi/gm DOSE EQUIVALENT 1-131 from LCO 3.7.1.4, " Specific Activity."

The SLB and SGTR safety analyses consider two cases of reactor coolant iodine specific activity. One case assumes specific activity at 0.25 pCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB (by a factor of 500), or SGTR (by a factor of 335), respectively.

The second case assumes the initial reactor coolant iodine activity at 60.0 pCi/gm DOSE EQUIVALENT 1-131 due to an iodine spike caused by a reactor or an RCS transient prior to the accident. In both cases, the noble gas specific activity is assumed to be 447.7 pCi/gm DOSE EQUIVALENT XE-133.

The RCS Specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

INSERT Page 68a

INSERT L-2009-133 Attachment 2 Page 11 of 20 The iodine specific activity in the reactor coolant is limited to 0.25 pCi/gm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to 447.7 pCi/gm DOSE EQUIVALENT XE-1 33. The limits on specific activity ensure that the offsite and control room doses will meet the appropriate SRP acceptance criteria.

The SLB, SGTR, locked rotor, and RCCA ejection accident analyses show that the calculated doses are within limits. Violation of the LCO may result in reactor coolant activity levels that could, in the event of any one of these accidents, lead to doses that exceed the acceptance criteria.

The ACTIONS permit limited operation when DOSE EQUIVALENT 1-131 is greater than 0.25 pCi/gm and less than 60 pCi/gm. The Actions require sampling within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following to establish a trend.

One surveillance requires performing a gamma isotropic analysis as a measure of noble gas specific activity of the reactor coolant at least once per 7 days. This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This surveillance provides an indication of any increase in the noble gas specific activity.

A second surveillance is performed to ensure that iodine specific activity remains within the LCO limit once per 14 days during normal operation and following fast power changes when iodine spiking is more apt to occur. The frequency between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change of greater than15% RATED THERMAL POWER within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following iodine spike initiation.

INSERT Page 68b

L-2009-133 ATTACHMENT I Attachment 2 (Page 73 of 111) Page 12 of 20 TECHNICAL SPECIFICATION BASES 3/4.6 Containment Systems 3/4.6.1 Primary Containment 50.67 3/4.6.1.1 Containment Integrity Primary CONTAINMENT 1NTE RITY ensures that the release of radioactive materials from the containment atmosphere ill be restricted to those leakage paths and associated leak rates assumed in the safety a alyses. This restriction, in conjunction with the leakage rate limitation, will limit the E BOUNDARY radiation doses to within the dose guideline values of 10 CFR during accident conditions.

Note that some penetrations do not fall under Technical Specification 3.6.1.1. For example Penetration 38 is an electrical penetration only, closed by virtue of its seals, and therefore, nothing needs to happen to close the penetration during accident conditions; it is considered already closed. A passive failure would be required in order to get communication between the containment atmosphere and the outside atmosphere through this. penetration (Turkey Point's license does not require consideration of passive failures). Similarly, closed systems inside containment already satisfy the requirement for CONTAINMENT INTEGRITY, so Tech Spec 3.6.1.1 does not apply to them at all (unless the piping itself is breached, which would be a passive failure).

The primary CONTAINMENT INTEGRITY requirement of Technical Specification 3.6.1.1 is modified by a Note allowing containment penetrations to be unisolated by opening the associated valves and airlocks under Administrative Controls when necessary to perform surveillance, testing requirements, and/or corrective maintenance. The Note also enforces compliance with Specification 3.6.4, in conjunction with Specification 3.6.1.1. The Administrative Controls shall consist of a dedicated person that is assigned the responsibility to close the valve(s) in the event of an emergency or when operation is complete.

The activity which places the pressurizer steam space vent in service to remove noncondensable gases from the Reactor Coolant System (RCS) is considered a corrective maintenance activity for the purposes of Technical Specification 3.6.1.1 compliance.

Nitrogen, the primary noncondensable gas contributor, enters the RCS during refueling outages when it is used as a cover gas in the volume control tank and from exposure of the reactor coolant to atmospheric conditions during the refueling sequence. The pressurizer heaters drive these gases out of solution during RCS heatup and they collect in the pressurizer steam space. This causes two problems. First, the pressurizer does not respond to spray actuation, creating a "hard bubble." Second, the noncondensable gases migrate to the reference leg condensing pots of the pressurizer level instruments, preventing reference leg makeup and proper indication of pressurizer liquid level. The pressurizer steam space venting activity is necessary to correct these problems and achieve and maintain acceptable nitrogen levels in the RCS. Accordingly, opening the containment penetration isolation valves needed to conduct the pressurizer steam space venting maintenance activity is typically performed under Administrative Controls in accordance with the provisions of the noted exception to the CONTAINMENT INTEGRITY requirement of Technical Specification 3.6.1.1.

v/In W2003:DPS/In/nv ~/1n W2003:DPS/In/nvft

L-2009-133 ATTACHMENT 1 Attachment 2 (Page 78 of 111) Page 13 of 20 TECHNICAL SPECIFICATION BASES 3/4.6.2.1 (Cont'd)

When PC-600/-601 are calibrated, a test signal is supplied to each circuit to check operation of the relays and annunciators operated by subject controllers. This test signal will prevent MOVs 862A, 862B, 863A, 863B from opening. Therefore, it is appropriate to tag out the MOV breakers, and enter Technical Specification Action Statement 3.5.2.a. and 3.6.2.1 when calibrating PC-600/-60 1.

3/4.6.2.2 Emergency Containment Cooling System The OPERABILITY of the Emergency Containment Cooling (ECC) System ensures that the heat removal capacity is maintained with acceptable ranges following postulated design basis accidents. To support both containment integrity safety analyses and component cooling water thermal analysis, a maximum of two ECCs can receive an automatic start signal following generation of a safety injection (SI) signal (one ECC receives an A train SI signal and another ECC receives a B train SI signal). To support post-LOCA long-term containment pressure/temperature analyses, a maximum of two ECCs are required to operate. The third (swing) ECC is required to be OPERABLE to support manual starting following a postulated LOCA event for containment pressure/temperature suppression.

The allowable out-of-service time requirements for the Containment Cooling System have been maintained consistent with that assigned other inoperable ESF equipment and do not reflect the additional redundancy in cooling capability provided by the Containment Spray System.

The surveillance requirement for ECC flow is verified by correlating the test configuration value with the design basis assumptions for system configuration and flow. An 18-month surveillance interval is acceptable based on the use of water from the CCW system, which results in a low risk of heat exchanger tube fouling. IINSERT I The OPELABRAILT of the PfEmcrgeney Containment Filter-ing System ensures that suftieient iodine re-m*-val capability will be available in the event of a LOCA.

reductioen int eentainmffent iodinte inventoryj r-eduees the fesulting SITE BOU4NDARY r-adiatiefn doses asseeated wit eantainment leakage. gystem components are nat subject to rapid deterioration Visual inspeetiefn and epereating/lperfcnnanee tests after maintenanee, pro.longed ep*cati.., and at the .e.uir..df .egue... vide a.. ur.anees o..

system r-eliabilit-y and will prevent system -failufre. In; siuilter fefiafiee testsar eondutiecd int taccodance with the methodology an4neto AS 51 9:75. Chareoa samples are tested uzing ASTM D;"03 1989 in ath"" Gencie Le..er. 99 02.

The test eenditiefts (30'G and 954; relaive humfidity,) are as 'pefedi he Generie Lerter.

Table 1 of the ASTM standard pr-evides the ht utbhheaeemiet durfing the tesfr each test par-ameter. The speeified methyl iodide penietr~atin valuie isbased on the a.6umptief used in the LCnOA an.alysis with a I fato-r of 2. Teehnieal Spe.cifica 3.6.3 r-eguitte three ECFzs to be OPERAk~LE in Modes 1, 2, 3, and 1. vei44enee Requtiremfent 4.6.3.d.2) states that eaeh ECF be deffonitrated OPERABLE... at least oncee per 18 mottaths... by v~eriifing that the filter: eoolifg seleneids cant be opented by operator W2003:DPSI1n~nwI1n

INSERT A L-2009-133 Attachment 2 Page 14 of 20 3/4.6.2.3 RECIRCULATION PH CONTROL SYSTEM The Recirculation pH Control System is a passive safeguard consisting of 10 stainless steel wire mesh baskets (2 large and 8 small) containing sodium tetraborate decahydrate (NaTB) located in the containment basement (14' elevation). The initial containment spray will be boric acid solution from the refueling water storage tank. The recirculation pH control system adds NaTB to the containment sump when the level of boric acid solution from the containment spray and the coolant lost from the Reactor Coolant System rises above the bottom of the buffering agent baskets. As the sump level rises, the NaTB will begin to dissolve. The addition of NaTB from the buffering agent baskets ensures the containment sump pH will be greater than 7.0.

The resultant alkaline pH of the spray enhances the ability of the recirculated spray to scavenge fission products from the containment atmosphere. The alkaline pH in the recirculation sump water minimizes the evolution of iodine and minimizes the occurrence of chloride and caustic stress corrosion on stainless steel piping systems exposed to the solution.

The OPERABILITY of the recirculation pH control system ensures that there is sufficient NaTB available in the containment to ensure a sump pH greater than 7.0 during the recirculation phase of a postulated LOCA. The baskets will not interact with surrounding equipment during a seismic event.

To achieve this, the analysis considered the minimum and maximum quantities of boron and borated water, the time-dependent post-LOCA sump temperatures, and the minimum time to reach switchover conditions. In addition, the formation of acid from radiolysis of air and water, radiolysis of chloride bearing electrical cable insulation and jacketing, and spilled reactor coolant were considered. Since, not all of the NaTB will be dissolved at the onset of containment spray recirculation, the final long-term pH was also calculated. The table below provides a summary of the quantities of NaTB and resulting pH values at the onset of recirculation and long-term.

Case Purpose Small Basket Large Basket Initial NaTB Dissolved pH at Long-Term No. Min/Max Height Height Mass Mass prior Recirc pH pH (ft) (ft) (Ibm) to Recirc (Ibm) 1 Min 2.5 2.77 14264 4687 7.005 7.396 (min) 2 Min 1.9167 2.1867 11061 4637 7.000 7.241 (min) 3 Max 2.5 2.77 17034* 17034 8.105 8.108 (max)

This case conservatively assumed that an additional small basket is installed in containment.

As shown, to satisfy the surveillance requirement, the two large baskets and eight small baskets must contain a combined mass of greater than 11,061 Ibm of NaTB to ensure the minimum sump pH of 7.0 at the onset of spray recirculation. The large baskets have a length and width of 54 inches, and a height of 33.25 inches and elevated 3.5 inches above the containment floor.

The smaller baskets have a length and width of 36 inches and a height of 30 inches and INSERT Page 89a

L-2009-133 Attachment 2 Page 15 of 20 elevated 6.5 inches above the containment floor. Varying basket dimensions or elevation (e.g.

basket leg height) impacts the surface to volume ratio and changes the time the NaTB is in contact with containment sump water. For instance, shorter legs would allow the NaTB to contact containment sump water sooner, therefore increasing the pH at the onset of recirculation. Longer legs, however, would reduce the pH at the onset of recirculation. The level of NaTB in the baskets required to provide an equilibrium sump solution pH greater than 7.0 is seven inches from the top of the basket; 26.25 inches for the large baskets and 23 inches for the small baskets from the bottom of the basket. The 18 month frequency for Surveillance Requirement 4.6.2.3 is sufficient to ensure that the stainless steel buffering agent baskets are intact and contain the required quantity of NaTB.

INSERT Page 89a

Procedure No.: Procedure

Title:

Page:

90 Approval Date:

0-ADM-536 Technical Specification Bases Control Program 2/18/09 L-2009-133 ATTACHMENT 1 Attachment 2 (Page 79 of 111) Page 16 of 20 TECHNICAL SPECIFICATION BASES The Technical Specifice~atio does not r~equir~e that both inidependent trains of ECFdosn componcnts be OPERXkBLE to suppoi4 thc ECF-s. Disabling one tr-ain of ECF dousin comfpoenifts docs noet rcendcr- the associatod ECF inoeper~able.

Thc UFSAR statcs that thc dcsign rqic nt fcr: thc ECF systcm is to rceducc thc iodn conccntration in thc containment atopcr 'olowinga M4A to lcvcls cnsur-ing thatth off sitc dosc will noet exceed the guidelines of 1 F 100 at the site boundary. Details of the site bouindary dose acltosacgvni Section 114.3.5 of thc U4FSAR.

Following a loss of coolant aecidcnit, a safet injcctioni signal will automaticeally cnrgz motor- controel circuits to staA thc thrcc filtcr unit fains. if outsidc powcr or full cmcrgcncy!

poc savailablc, all thrcc filtcr- units arc stancid (only tA'o arc rcqfuirced). if clcctric powc is limitcd duc to thc failuirc of an cmcrgcncey dicscl gcncrator, two of thc thrcc unlits A borated water- spr-ay system is inistalled in each filter- unit to dissipate the radioactive deew' heat and intitiated by the loss of air flow, through the filter- unit, suchb as fatiluire of the fan. The Design Basis Docutment for- the ECF system states that radioacti-ve decay heat+

r-emoval by dousing the ECF charco~al bed with containment spr-ay Water- On ECF fain failure is a Quality Related ftnction. As suceh, single failure cr-iter-ia do noat apply to the ECF spr-ay system components because:

i/-I tnr thz Ft lmrln2 n nnt rzflhJir2ri S--- :w tn rzrtnrm it %rltcr; minted tunetien ci r--------------------------------

femvin radseaetive iedine and inethyI iadide tf em thie containment 4atmsphere-,

2) Dousing is not r-equir-ed to mfaintain offsite doses below 10CFRI 00 limaitsftftd

3) The ECF system cani pcrfcem its saf' related functions with anly single failure without r-equir-ing dousing.

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L-2009-133 ATTACHMENT 1 Attachment 2 (Page 88 of 111) Page 17 of 20 TECHNICAL SPECIFICATION BASES 3/4.7.5 (Cont'd)

The Control Room Emergency Ventilation System is considered to be OPERABLE (Ref: JPN-PTN-SENP-92-017) when 1) Three air handling units (AHUs) (one of each of the three air conditioning units) are operable, 2) Two condensing units (two out of three available condensers) are operable, 3) One recirculation filter unit is operable, 4) Two recirculation fans operable, and 5) Associated dampers are operable. The reason three AHUs are required is that in the event of a single failure, only two AHUs would be available to supply air to the suction of the recirculation filter and fan. This is the configuration tested to support Technical Specification operability for flow through the emergency charcoal filter. Taking one AHU out of service renders the system incapable of operating in accordance with the tested configuration assuming an accident and a single failure (i.e., only one air handling unit available instead of the two assumed by the analysis). Any one of the three condensing (air conditioning) units is capable of maintaining the control room equipment within its environmental limits for temperature and humidity. Thus, one condensing unit can be taken out of service without impacting the Add ability of the Control Room Emergency Ventilation System to accomplish its intended INSERT 1 function under single failure conditions.

next page System components are not subject to rapid deterioration, having lifetimes of many years, even under continuous flow conditions. Visual inspection and operating tests provide assurance of system reliability and will ensure early detection of conditions which could cause the system to fail or operate improperly. The filters performance tests prove that filters have been properly installed, that no deterioration or damage has occurred, and that all components and subsystems operate properly. The in-situ tests are performed in accordance with the methodology and intent of ANSI N510 (1975) and provide assurance that filter performance has not deteriorated below returned specification values due to aging, contamination, or other effects. Charcoal samples are tested using ASTM D3803-1989 in accordance with Generic Letter 99-02. The test conditions (30'C and 95%

relative humidity) are as specified in the Generic Letter. Table 1 of the ASTM standard provides the tolerances that must be met during the test for each test parameter. The T OGfA Analysis.

3/4.7.6 Snubbers All snubbers are required OPERABLE to ensure that the structural integrity of the R actor Coolant System and all other safety-related systems is maintained during and follo ing a seismic or other event initiating dynamic loads.

The visual inspection frequency is based upon maintaining a constant level of s ubber protection to each safety-related system during an earthquake or severe tra sient.

Therefore, the required inspection interval varies inversely with the observed s ubber failures and is determined by the number of inoperable snubbers found duri g an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of suc early inspections performed before the original required time interval has elapsed (nomin 1 time less 25%) may not be used to lengthen the required inspection interval. Any ins ection whose results require a shorter inspection interval will override the previous schedul The specified methyl iodide penetration value is based on theL

,^,)n, ,D<,,n,,l,,assumptions used in the LOCA analysis with a safety factor of 2 applied.

INSERT 1 L-2009-133 Attachment 2 Page 18 of 20 In addition, the Control Room Emergency Ventilation System also includes the emergency air intakes, located beyond the southeast and northeast corners of the Auxiliary Building. The Control Room Emergency Ventilation System emergency air intakes are considered OPERABLE when: 1) both flow paths are available, 2) have balanced intake flow rates and 3) flow path capable of drawing outside makeup air from only the analyzed intake locations. The alternate source term radiological analyses assume both emergency air intake flow paths are available with parallel dampers ensuring outside makeup air can be drawn through both intake locations during a design basis accident and a single active failure. These analyses rely on a provision in Regulatory Guide 1.194 Section 3.3.2 that allows a reduction in the atmospheric dispersion factors (X/Qs) for dual intake arrangements with balanced flow rates to one half the more limiting X/Q value provided the two intakes are not within the same wind direction window for each release / receptor location. Accordingly, any maintenance on the emergency intake dampers or associated duct work that would prevent the Control Room Emergency Ventilation system from accomplishing these functions would require entering the action statement.

INSERT Page 99a

Procedure No.: Procedure

Title:

Page:

101 Approval Date:

0-ADM-536 Technical Specification Bases Control Program 2/18/09 L-2009-133 ATTACHMENT 1 Attachment 2 (Page 90 of 11) Page 19 of 20 TECHNICAL SPECIFICATION BASES 3/4.7.8 Explosive Gas Mixture This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the GAS DECAY TANK SYSTEM (as measured in the inservice gas decay tank) is maintained below the flammability limits of hydrogen and oxygen.

Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4.7.9 Gas Decay Tanks The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification. Restricting the quantity of radioactivity contained in each Gas Decay Tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will not exceed .rem.

3/4.8 Electrical Power Systems° 0.1 3/4.8.1, 3/4.8.2,

& 3/4.8.3 A.C. Sources, D.C. Sources, and Onsite Power Distribution The OPERABILITY of the A.C. and D.C power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for (1) The safe shutdown of the facility, and (2) The mitigation and control of accident conditions within the facility.

The loss of an associated diesel generator for systems, subsystems, trains, components or devices does not result in the systems, subsystems, trains, components or devices being considered inoperable for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation for the affected unit provided (1) Its corresponding normal power source is OPERABLE; and (2) Its redundant systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generators as a source emergency power to meet all applicable LCOs are OPERABLE. This allows operation to be governed by the time limits of the ACTION statement associated with the inoperable diesel generator, not the individual ACTION statements for each system, subsystem, train, component or device. However, due to the existence of shared systems, there are certain conditions that require special provisions. These provisions are stipulated in the appropriate LCOs as needed.

W2003:DPS/ln/nw/In

Technical Specification Bases Control Program L-2009-133 ATTACHMENT 1 Attachment 2 (Page 109 of I 11) Page 20 of 20 TECHNICAL SPECIFICATION BASES 3/4.9.9 (Cont'd)

A normal refueling consists of 2 core alteration sequences: unloading the core, and reloading the core, typically with a suspension of core alterations in between. The core unload sequence begins with control rod unlatching, followed by removal of upper internals, followed by unloading fuel assemblies to the SFP. The core reload sequence consists of reloading fuel assemblies from the SFP, followed by upper internals installation, followed by latching control rods. Therefore, if the Containment Ventilation Isolation System is demonstrated OPERABLE at least once per 7 days following the specified testing within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of control rod unlatching, then Containment Ventilation Isolation System operability need not be demonstrated within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of core reload. Otherwise, the specified testing is required to be performed within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of core reload.

3/4.9.10 &

3/4.9.11 Water Level - Reactor Vessel And Storage Pool The restrictions on minimum water level ensure that sufficient shielding will be available during fuel movement and for removal of iodine in the event of a fuel handling accident.

The minimum water depth is consistent with the assumptions of the safety analysis.

3/4.9.12 Handling Of Spent Fuel Cask Limiting spent fuel decay time from last time critical to a minimum of 1,525 hours0.00608 days 0.146 hours 8.680556e-4 weeks 1.997625e-4 months prior to moving a spent fuel cask into the spent fuel pit will ensure that potential offsite doses are a fraction of limits should a dropped cask strike the stored fuel assemblies.

The restriction o allow only a single element cask to be moved into the spent fuel pit will ensure the mai tenance of water inventory in the unlikely event of an uncontrolled cask descent. Use a single element cask which nominally weighs about twenty-five tons will also increase c ane safety margins by about a factor of four.

Requiring tha spent fuel decay time from last time critical be at least 120 days prior to moving a fue assembly outside the fuel storage pit in a shipping cask will ensure that potential offs te doses are a fraction of I' "" ' " limits should a dropped cask and ruptured fuel ssembly release activity directly to atmosphere.

3/4.9.13 diation M onitorin a The OPE BILITY of the containment radiation monito ensures continuous monitoring of radiation evels to provide immediate indication of an uns e condition.

10 CFR 50.67 10-CFR 50.67 W2003:DPS/In/nw/ln

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 3 License Amendment Request 196 Alternative Source Term and Conforming Amendment LICENSE AMENDMENT REQUEST ALTERNATIVE SOURCE TERM AND CONFORMING AMENDMENT LAR 196 ATTACHMENT 3 NUMERICAL APPLICATIONS INC.

NAI REPORT RELEASE NAI REPORT NO. NAI-1396-045 AST LICENSING TECHNICAL REPORT FOR TURKEY POINT UNITS 3 AND 4 REVISION 1 DATED JUNE 18, 2009

Page I of 81 NAI Report Release Report Number: NAI- 1396-045 Revision Number: I

Title:

AST Licensing Technical Report for Turkey Point Units 3 and 4

==

Description:==

This report documents the results of the analyses and evaluations performed by Numerical Applications, Inc. in support of the Turkey Point licensing project to implement Alternative Source Term (AST). Design basis accidents and radiological consequences are evaluated using the AST methodology to support control room habitability in the event of increases in unfiltered inleakage. The analyses and evaluations performed by NAI are based on the guidance of Regulatory Guide 1.183.

-;ý ýP" Author (Mark Pope)-, -

6 Iog Date Ri*ie-wer (Jim Harrell) Date NAI Managemeant (Tom George) I Dat/

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 2 of 81

.* UIONSeN EN lN*ERING Page 2 of 81I*

Table of Contents 1.0 Radiological Consequences Utilizing the Alternative Source Term Methodology ............................. 4 1.1 Introduction ...................................................................................................................................... 4 1.2 Evaluation Overview and Objective ........................................................................................... 4 1.3 Proposed Changes to the Turkey Point Licensing Basis ............................................................. 5 1.4 Com pliance with Regulatory Guidelines .................................................................................... 6 1.5 Computer Codes ............................................................................................................................... 7 1.6 Radiological Evaluation M ethodology ......................................................................................... 8 1.6.1 Analysis Input Assumptions ............................................................................................... 8 1.6.2 Acceptance Criteria ................................................................................................................ 8 1.6.3 Control Room Ventilation System Description ................................................................. 8 1.6.3.1 Control Room Dose Calculation M odel ....................................................................... 9 1.6.4 Direct Shine Dose ................................................................................................................. 10 1.7 Radiation Source Term s ................................................................................................................. 10 1.7.1 Fission Product Inventory ................................................................................................ 10 1.7.2 Prim ary Coolant Source Term ......................................................................................... 11 1.7.3 Secondary Side Coolant Source Term .............................................................................. 12 1.7.4 Core Source Term ................................................................................................................. 12 1.7.5 Fuel Handling Accident Source Term .............................................................................. 12 1.7.6 Gap Release Fractions .......................................................................................................... 13 1.8 Atm ospheric D ispersion (X/Q) Factors ....................................................................................... 13 1.8.1 Onsite X/Q Determination .............................................................................................. 13 1.8.2 Offsite XIQ Determination ................................................................................................ 14 1.8.3 M eteorological Data ............................................................................................................. 15 1.9 Sum p pH Determ ination ................................................................................................................. 16 2.0 Radiological Consequences - Event Analyses ................................................................................... 17 2.1 Loss of Coolant Accident (LOCA) ........................................................................................... 17 2.2 Fuel Handling Accident (FHA) .................................................................................................. 24 2.3 M ain Steam line Break (M SLB) ................................................................................................. 26 2.4 Steam Generator Tube Rupture (SGTR) .................................................................................... 30 2.5 Locked Rotor .................................................................................................................................. 33 2.6 Rod Cluster Control Assembly (RCCA) Ejection ...................................................................... 36 2.7 W aste Gas Decay Tank (W GDT) Rupture ................................................................................. 40 2.8 Spent Fuel Cask Drop .................................................................................................................... 42 2.9 Environm ental Qualification (EQ) ............................................................................................. 44 3.0 Summ ary of Results .............................................................................................................................. 44 4.0 Conclusions ........................................................................................................................................... 44 5.0 References ............................................................................................................................................. 45

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 3 of 81 Figures and Tables FIGURE 1.8.1-1 ONSITE RELEASE-RECEPTOR LOCATION SKETCH ............................................................ 48 TABLE 1.6.3-1 CONTROL ROOM VENTILATION SYSTEM PARAMETERS ..................................................... 49 TABLE 1.6.3-2 LOCA DIRECT SHINE DOSE .............................................................................................. 49 TABLE 1.7.2-1 PRIMARY COOLANT SOURCE TERM ................................................................................. 50 TABLE 1.7.3-1 SECONDARY SIDE SOURCE TERM ...................................................................................... 51 TABLE 1.7.4-1 LOCA SOURCETERM ................................................................................................... 51 TABLE 1.7.5-1 FUEL HANDLING ACCIDENT SOURCE TERM ........................... .................................. 53 TABLE 1.8.1-1 RELEASE-RECEPTOR COMBINATION PARAMETERS FOR ANALYSIS EVENTS ...................... 55 TABLE 1.8.1-2 ONSITE ATMOSPHERIC DISPERSION (X/Q) FACTORS FOR ANALYSIS EVENTS ................... 56 TABLE 1.8.1-3 RELEASE-RECEPTOR POINT PAIRS ASSUMED FOR ANALYSIS EVENTS ............................. 58 TABLE 1.8.2-1 OFFSITE ATMOSPHERIC DISPERSION (X/Q) FACTORS FOR ANALYSIS EVENTS .................. 59 TABLE 1.9.-i POST-LOCA SUMP PH ANALYSIS RESULTS ........................................................................ 59 TABLE 2.1-1 LoSS OF COOLANT ACCIDENT (LOCA) - INPUTS AND ASSUMPTIONS ................................. 60 TABLE 2.1-2 LO CA RELEASE PHASES ......................................................................................................... 62 TABLE 2.1-3 LOCA TIME DEPENDENT RWST TOTAL IODINE CONCENTRATION ..................................... 63 TABLE 2.1-4 LOCA TIME DEPENDENT RWST ELEMENTAL IODINE FRACTION ........................................ 63 TABLE 2.1-5 ADJUSTED RELEASE RATE FROM THE RWST ...................................................................... 65 TABLE 2.1-6 LOCA DOSE CONSEQUENCES ............................................................................................. 64 TABLE 2.2-1 FUEL HANDLING ACCIDENT (FHA) - INPUTS AND ASSUMPTIONS ....................... 66 TABLE 2.2-2 FUEL HANDLING ACCIDENT DOSE CONSEQUENCES ............................................................ 66 TABLE 2.3-1 MAIN STEAM LINE BREAK (MSLB) - INPUTS AND ASSUMPTIONS .......................................... 67 TABLE 2.3-2 INTACT SGs STEAM RELEASE RATE ................................................................................... 68 TABLE 2.3-3 60 .tCi/gm D .E. 1- 131 ACTIVITIES ....................................................................................... 68 TABLE 2.3-4 IODINE EQUILIBRIUM APPEARANCE ASSUMPTIONS ............................................................. 68 TABLE 2.3-5 CONCURRENT (500 x) IODINE SPIKE APPEARANCE RATE ....................................................... 69 TABLE 2.3-6 M SLB DOSE CONSEQUENCES ............................................................................................. 69 TABLE 2.4-1 STEAM GENERATOR TUBE RUPTURE (SGTR) - INPUTS AND ASSUMPTIONS ........................ 70 TABLE 2.4-2 SG TR M ASS FLOW RATE S...................................................................................................... 71 TABLE 2.4-3 60 ICi/gm D.E. 1-131 ACTIVITIES ....................................................................................... 71 TABLE 2.4-4 IODINE EQUILIBRIUM APPEARANCE ASSUMPTIONS ............................................................. 72 TABLE 2.4-5 CONCURRENT (335 X) IODINE SPIKE APPEARANCE RATE .................................................... 72 TABLE 2.4-6 SGTR DOSE CONSEQUENCES ............................................................................................... 72 TABLE 2.5-1 LOCKED ROTOR - INPUTS AND ASSUMPTIONS .................................................................... 73 TABLE 2.5-2 LOCKED ROTOR STEAM RELEASE ....................................................................................... 74 TABLE 2.5-3 LOCKED ROTOR DOSE CONSEQUENCES ............................................................................... 74 TABLE 2.6-1 RCCA EJECTION - INPUTS AND ASSUMPTIONS .................................................................... 75 TABLE 2.6-2 RCCA EJECTION STEAM RELEASE ...................................................................................... 76 TABLE 2.6-3 RCCA EJECTION DOSE CONSEQUENCES .............................................................................. 77 TABLE 2.7-1 WASTE GAS DECAY TANK (WGDT) RUPTURE- INPUTS AND ASSUMPTIONS ...................... 77 TABLE 2.7-2 W G D T SOUR CE TERM ............................................................................................................ 77 TABLE 2.7-3 W GDT RUPTURE DOSE CONSEQUENCES ............................................................................... 78 TABLE 2.8-1 SPENT FUEL CASK DROP - INPUTS AND ASSUMPTIONS ...................................................... 79 TABLE 2.8-2 SPENT FUEL CASK DROP DOSE CONSEQUENCES ........................ ............................... 80 TABLE 3-1 TURKEY POINT UNITS NO. 3 AND 4

SUMMARY

OF AST ANALYSIS RESULTS ............................ 81

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 4 of 81 SOý4UTINSINENGINý.RING ANODSF'wAREP o 1.0 Radiological Consequences Utilizing the Alternative Source Term Methodology 1.1 Introduction The current Turkey Point licensing basis for radiological consequences analyses of accidents discussed in Chapter 14 of the Updated Final Safety Analysis Report (UFSAR) is based on methodologies and assumptions that are primarily derived from Technical Information Document (TID)-14844 and other early guidance.

Regulatory Guide 1.183 provides guidance on application of an Alternative Source Term (AST) in revising the accident source terms used in design basis radiological consequences analyses, as allowed by I OCFR50.67. Because of advances made in understanding the timing, magnitude, and chemical form of fission product releases from severe nuclear power plant accidents, I OCFR50.67 is issued to allow holders of operating licenses to voluntarily revise the traditional accident source term used in the design basis accident (DBA) radiological consequence analyses with an alternative source term.

1.2 Evaluation Overview and Objective As documented in NEI 99-03 and Generic Letter 2003-01, several nuclear plants performed testing on control room unfiltered air inleakage that demonstrated leakage rates in excess of amounts assumed in the current accident analyses. The AST methodology as established in Reg. Guide 1.183 is being used to calculate the offsite and control room radiological consequences for Turkey Point Units 3 and 4 to support the control room habitability program by addressing the radiological impact of potential increases in control room unfiltered air inleakage. In all cases, the most limiting configuration between Units 3 and 4 is utilized for the radiological analyses.

The following limiting UFSAR Chapter 14 accidents are analyzed:

" Loss-of-Coolant Accident (LOCA)

" Main Steam Line Break (MSLB)

" Steam Generator Tube Rupture (SGTR)

" Locked Rotor

" Rod Cluster Control Assembly (RCCA) Ejection

" Fuel Handling Accident (FHA)

" Waste Gas Decay Tank (WGDT) Rupture

" Spent Fuel Cask Drop Accident Note that although RG 1.183 does not include the WGDT or Spent Fuel Cask Drop accidents, these events were included in the AST analysis to incorporate new atmospheric dispersion factors and to evaluate the dose consequences using the TEDE criteria consistent with the other limiting UFSAR events.

Each accident and the specific input and assumptions are described in Section 2.0 of this report. These analyses provide for a bounding allowable control room unfiltered air inleakage of 115 cfm. The use of 115 cfm as a design basis value was established to be above the unfiltered inleakage value determined through testing or analysis consistent with the resolution of issues identified in NEI 99-03 and Generic Letter 2003-01. The control room inleakage testing performed at Turkey Point in 2003 has resulted in less than 10 cfm of unfiltered inleakage when all uncertainties and tolerances were considered. Therefore, there is significant margin between the bounding dose analysis inleakage value of 115 cfm and the measured control room unfiltered inleakage.

[JN NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 5 of 81 sOUNS INENGINEERING AD ,SOFTW.R 1.3 Proposed Changes to the Turkey Point Licensing Basis Florida Power and Light (FP&L) Company proposes to revise the Turkey Point licensing basis to implement the AST, described in Reg. Guide 1.183, through reanalysis of the radiological consequences of the UFSAR Chapter 14 accidents listed in Section 1.2 above. This is a full implementation of the AST as described in Section 1.1.3 of the Reg. Guide. Key elements of the revised analysis include:

" The total effective dose equivalent (TEDE) acceptance criterion of 10CFR50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of IOCFR100.l11.

" New onsite (Control Room) and offsite atmospheric dispersion factors are developed.

" Dose conversion factors for inhalation and submersion are from Federal Guidance Reports (FGR)

Nos. 11 and 12, respectively.

" A bounding value for control room unfiltered air inleakage was established by increasing the inleakage until the dose acceptance criteria for the limiting event (LOCA) was approached.

" A primary coolant specific activity that is more restrictive than the current Technical Specification limit is utilized.

" A steam generator tube leakage rate is used that is more restrictive than the current Technical Specification program limit for primary-to-secondary accident induced leakage

" A containment leakage value that is more restrictive than the current Technical Specification limit is utilized.

" Sump pH control is provided by sodium tetraborate decahydrate (NaTB) baskets.

" Containment ESF filter units are not being credited.

" The Containment Spray System and Recirculation pH Sump Control System are being credited for post-LOCA iodine removal.

For the Secondary RCCA Ejection, FHA with a Fuel Building release, and Spent Fuel Cask Drop events, credit is taken for the operators to manually isolate the Control Room Ventilation System within 30 minutes.

" The Control Room emergency intakes are being relocated beyond the southeast and northeast and corners of the Reactor Auxiliary Building.

" The Control Room Emergency Ventilation System methyl iodine penetration criterion is being reduced from 2.5% to 1.25%.

Accordingly, the following changes to the Turkey Point Units 3 and 4 Technical Specifications (TS) are proposed:

" The definition of Dose Equivalent 1-131 in Section 1.12 is revised to reference Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989, as the source of effective dose conversion factors.

" The definition of Average Disintegration Energy in Section 1.13 is replaced with the definition of Dose Equivalent Xe-133 based upon effective dose conversion factors from Table 111.1 of Federal Guidance Report No. 12 (FGR 12), "External Exposure to Radionuclides in Air, Water, and Soil,"

1983.

" The Reactor Coolant System (RCS) specific activity limit for dose equivalent Iodine-131 (DE I-13 1), stated in Limiting Condition for Operation (LCO) 3.4.8.a, is reduced from 1 microcurie per gram to 0.25 microcurie per gram.

" The Reactor Coolant System specific activity limit for gross radioactivity stated in Limiting Condition for Operation (LCO) 3.4.8.b, is changed from IOO/E-bar to dose equivalent Xenon-133.

The DE Xe-133 limit of 447.7 microcuries per gram is established at a value which maintains the current maximum allowable RCS activity level based upon E-bar.

" The methyl iodine penetration criteria for the Control Room Emergency Ventilation System charcoal adsorber in surveillance requirement 4.7.5.c.2 is reduced from 2.5% to 1.25%.

" The terminology used in Limiting Condition for Operation (LCO) 3.7.9 to set the maximum contents of the Waste Gas Decay Tank is clarified to read 'DOSE EQUIVALENT Xe-133.'

" The maximum allowable containment leakage rate acceptance criterion stated in TS 6.8.4.h, "Containment Leakage Rate Testing Program," is reduced from 0.25% to 0.20% of containment air weight per day.

" The primary-to-secondary steam generator leakage rate assumed in the accident analysis and specified in the accident induced leakage performance criterion of TS 6.8.4.j.b.2, "Steam Generator Program," is reduced from "1 gpm through all steam generators and 500 gpd through any one steam generator" [at accident conditions] to "0.6 gpm through all steam generators and 0.2 gpm through any one steam generator at room temperature conditions."

" A method for controlling the pH of the post-LOCA containment sump solution using sodium tetraborate decahydrate is being proposed. This passive system will consist of baskets of NaTB in the lower regions of the containment. Appropriate technical specifications and surveillance requirements are proposed for Section 3/4.6 "Containment Systems."

" Operability requirements for emergency containment filter units in Section 3/4.6.3 will be deleted.

1.4 Compliance with Regulatory Guidelines The revised Turkey Point accident analyses addressed in this report follow the guidance provided in Reg.

Guide 1.183. Assumptions and methods utilized in this analysis for which no specific guidance is provided in Reg. Guide 1.183, but for which a regulatory precedent has been established, are as follows:

Selection of the WGDT Rupture offsite dose consequences acceptance criteria for the EAB is based on the Branch Technical Position 11-5 of the Standard Review Plan, which gives the total body exposure acceptance criteria for an individual at the nearest exclusion area boundary following a waste gas system failure as 0.1 rem. Additional guidance related to the application of the AST methodology to the analysis of this event is given in Issue #11 of NRC Regulatory Issue Summary 2006-04. RIS 2006-04 sets the AST dose limit to a member of the public for this event at 100 mrem TEDE. Since RIS 2006-04 associates this acceptance criteria with the annual limit established in 10CFR20, the EAB dose will be evaluated over the duration of the event rather than for the worst two-hour period. Branch Technical Position 11-5 does not require the dose consequences to be evaluated at the LPZ or the control room. These locations will be evaluated for completeness. The LPZ dose will be evaluated against the EAB acceptance criteria of 0.1 rem TEDE, and the TEDE dose limit of 5 rem from 10CFR50.67 will be applied to the control room consistent with the other AST events.

Regulatory Guide 1.183 does not provide specific guidance for the Spent Fuel Cask Drop event.

However, the guidance for the Fuel Handling Accident in Appendix B of the Reg. Guide is judged

to be closely applicable to the conditions of this event. Therefore, the evaluation of this event will conform to the positions of Reg. Guide 1.183, Appendix B, and.the acceptance criteria for the Fuel Handling Event from Section 4.4 and Table 6 will be applied.

Use of the MicroShield code to develop direct shine doses to the Control Room. MicroShield is a point kernel integration code used for general-purpose gamma shielding analysis. It is qualified for this application and has been used to support licensing submittals that have been accepted by the NRC. Precedent for this use of MicroShield is established in the Duane Arnold Energy Center submittal dated October 19, 2000 and associated NRC Safety Evaluation dated July 31, 2001.

1.5 Computer Codes The following computer codes are used in performing the Alternative Source Term analyses:

Computer Code Version Reference Purpose ARCON96 June 1997 5.12 Atmospheric Dispersion Factors MicroShield 5.05 5.13 Direct Shine Dose Calculations ORIGEN 2.1 5.14 Core Fission Product Inventory PAVAN 2.0 5.15 Atmospheric Dispersion Factors GOTHIC 7.1 5.16 - 5.18 Containment Mixing RADTRAD-NAI 1.0p3 5.19 Radiological Dose Calculations 1.5.1 ARCON96 - used to calculate relative concentrations (X/Q factors) in plumes from nuclear power plants at control room intakes in the vicinity of the release point using plant meteorological data.

1.5.2 MicroShield - used to analyze shielding and estimate exposure from gamma radiation.

1.5.3 ORIGEN - used for calculating the buildup, decay, and processing of radioactive materials.

1.5.4 PAVAN - provides relative air concentration (XIQ) values as functions of direction for various time periods at the EAB and LPZ boundaries assuming ground-level releases or elevated releases from freestanding stacks.

1.5.5 GOTHIC - The GOTHIC containment analysis code is used to determine the containment mixing due to actuation of the containment sprays and operation of the Emergency Containment Coolers.

1.5.6 RADTRAD-NAI - estimates the radiological doses at offsite locations and in the control room of nuclear power plants as consequences of postulated accidents. The code considers the timing, physical form (i.e., vapor or aerosol) and chemical species of the radioactive material released into the environment.

RADTRAD-NAI began with versions 3.01 and 3.02 of the NRC's RADTRAD computer code, originally developed by Sandia National Laboratory (SNL). The code is was modified to compile on a UNIX system. Once compiled, an extensive designreview/verification and validation process was completed on the code and documentation. The subject of the review also included the source code for the solver, which was made available in a separate distribution from the NRC. RADTRAD-NAI validation was performed with three different types of tests:

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 8 of 81 SOtrno~S INeNcAIN-ERINA ANDSOFTWAnRE

" Comparison of selected Acceptance Test Case results with Excel spreadsheet solutions and hand solutions,

" Separate effects tests, and

" Industry examples.

The industry examples included prior AST submittals by BWRs and PWRs, as well as other plant examples.

In addition to reviewing the code and incorporating error corrections, several software revisions were made. One revision involved the consideration of noble gases generated by decay of isotopes on filters that are returned to the downstream compartment. Another revision involved the modification of the dose conversion and nuclide inventory files to account for 107 isotopes to assure that significant dose contributors were addressed. The dose conversion factors used by RADTRAD-NAI are from Federal Guidance Report Nos. 11 and 12 (FGR 1 land FGR 12).

RADTRAD-NAI was developed and is maintained under Numerical Applications' 10CFR50 Appendix B program.

1.6 Radiological Evaluation Methodology 1.6.1 Analysis Input Assumptions Common analysis input assumptions include those for the control room ventilation system and dose calculation model (Section 1.6.3), direct shine dose (Section 1.6.5), radiation source terms (Section 1.7), and atmospheric dispersion factors (Section 1.8). Event-specific assumptions are discussed in the event analyses in Section 2.0.

1.6.2 Acceptance Criteria Offsite and Control Room doses must meet the guidelines of Reg. Guide 1.183 and requirements of 10CFR50.67. The acceptance criteria for specific postulated accidents are provided in Table 6 of Reg.

Guide 1.183. For Waste Gas Decay Tank Rupture and Spent Fuel Cask Drop events, which are not addressed in Reg. Guide 1.183, the basis used to establish the acceptance criteria for the radiological consequences is provided in Section 1.4.

1.6.3 Control Room Ventilation System Description The Control Room Ventilation System is required to assure control room habitability. The design of the control room envelope and overall description of the Control Room Ventilation System are discussed in the Turkey Point UFSAR, Section 9.9.1.

The Control Room Ventilation System consists of three 100% capacity air handling units and a ducted air intake and air distribution system. Outside air is drawn into the air handling units through roughing filters and cooled as required. Conditioned air is then directed back to the rooms through a supply air duct system.

Under emergency conditions, the Control Room Ventilation System has the capability to go into the recirculation mode. In the recirculation mode, fresh and recirculated air is processed through high efficiency particulate (HEPA) filters and charcoal filters to maintain the control room environment at acceptable conditions. The recirculation mode is automatically entered on receipt of an outside normal air intake high radiation signal, a containment high radiation signal or a safety injection signal.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 9 of 81 Redundant isolation dampers at the normal outside air intake and exhaust paths are automatically closed, so that the control room envelope is isolated except for filtered fresh air makeup. The system is designed to perform its safety functions and maintain a habitable environment in the control room envelope during isolation.

The control room envelope is slightly pressurized relative to the surroundings during normal plant operation with outside air continuously introduced to the control room envelope at a rate of 1000 cfm.

In the recirculation mode, the control room is pressurized at the rate of 525 cfm to maintain a positive pressure differential. Makeup air for pressurization is filtered before entering the control room. The recirculated air flow rate of 375 cfm is filtered by the same filters as the makeup air.

The net volume of the control room envelope serviced by the Control Room Ventilation System is 47,786 cubic ft.

1.6.3.1 Control Room Dose Calculation Model The control room model includes a recirculation filter model along with filtered air intake, unfiltered air inleakage and an exhaust path. System performance, sequence, and timing of operational evolutions associated with the control room ventilation system are discussed below. Control room ventilation system parameters assumed in the analyses are provided in Table 1.6.3-1. The dispersion factors for use in modeling the control room during each mode of operation are provided in Tables 1.8.1-2 and 1.8.1-3. Control room occupancy factors and assumed breathing rates are those prescribed in Reg. Guide 1.183. Figure 1.8.1-1 provides a site sketch showing the Turkey Point plant layout, including the location of onsite potential radiological release points with respect to the control room air intakes. The elevations of release points and intakes used in the control room AST dose assessments are provided in Table 1.8.1-1.

The control room ventilation system contains a filtration system for removal of radioactive iodine and particulate material that may enter the control room during the course of the event. Calculation of the dose to operators in the control room requires modeling of various system configurations and operating evolutions of the control room ventilation system during the course of the accident. While in the short duration normal mode prior to control room normal air intake and exhaust isolation, a single inlet to the control room (on the southwest comer of the Control Building) with an unfiltered flow rate of 1000 cfm is modeled. When in the emergency/recirculation mode, the control room model will define two concurrent air inlet paths representing the defined control room ventilation system air intake and the unfiltered inleakage into the control room. In the emergency/recirculation mode, outside air enters the control room through the filtration/ventilation system from both intakes which are located near ground level off of the southeast and northeast corners of the Auxiliary Building. Based on the release point, the dispersion factors for the most conservative emergency intake are modeled. Since the control room envelope remains pressurized while in the emergency/recirculation mode, unfiltered inleakage is modeled as makeup air which bypasses the filtration system. Modeling of the control room conservatively addresses these factors as they apply to the various release locations for each analyzed event. Details of the control room modeling for each event are described in subsequent event analyses sections.

For all events, delays in switching to the emergency/recirculation mode from the normal mode are conservatively considered with respect to the time required for signal processing, relay actuation, time required for the dampers to move and the system to re-align and diesel generator start time.

1.6.4 Direct Shine Dose The total control room dose also requires the calculation of direct shine dose contributions from control room filters, from the radioactive plume in the environment, and from the containment building. The contribution to the total dose to the operators from direct radiation sources were calculated for the LBLOCA event. The LOCA shine dose contribution is assumed to be bounding for all other events.

Direct shine dose is determined from four different sources to the control room operator after a postulated LOCA event. These sources are the containment building, the control room recirculation filters, the external cloud that envelops the control room, and from the containment purge duct penetration. All other sources of direct shine dose are considered negligible. The MicroShield 5 code is used to determine direct shine exposure to a dose point located in the control room. Each source required a different MicroShield case structure including different geometries, sources, and materials.

The external cloud is assumed to have a length of 1000 meters in the MicroShield cases to approximate an infinite cloud. A series of cases is run with each structure to determine an exposure rate from the radiological source at given points in time. These sources were taken from RADTRAD-NAI runs which output the nuclide activity at selected points in time for the event. The RADTRAD-NAI output provides the time dependent results of the radioactivity retained in the control room filter components, as well as the activity inventory in the environment and the containment. A bounding control room filter inventory is established by increasing the unfiltered inleakage to produce a control room dose slightly in excess of the 5 rem TEDE dose limit and by maximizing the filter efficiency.

The RADTRAD-NAI sources were then input into the MicroShield case file where they are either used as is, or 'decayed' (once the release has stopped) in MicroShield to yield the source activity at a later point in time. The exposure results from the series of cases for each source term were then corrected for occupancy using the occupancy factors specified in Reg. Guide 1.183. The cumulative exposure and dose are subsequently calculated to yield the total 30-day direct shine dose from each source. The results of the Direct Shine Dose evaluation are presented in Table 1.6.3-2.

The applicable components of the LOCA shine dose are used as a conservative assessment of the direct shine dose contribution for all other accidents.

1.7 Radiation Source Terms 1.7.1 Fission Product Inventory The source term data to be used in performing alternative source term (AST) analyses for Turkey Point are summarized in the following tables:

Table 1.7.2 Primary Coolant Source Term Table 1.7.3 Secondary Side Source Term (non-LOCA)

Table 1.7.4 Core Source Term Table 1.7.5 Fuel Handling Accident Source Term Note that the source terms provided in the referenced tables do not include any decay before the start of the events. Decay time assumptions are applied in the RADTRAD cases for individual event analysis. For example, the RADTRAD case for the Fuel Handling Accident analysis would account for the required decay time before the movement of fuel is allowed (as determined by Technical Specifications).

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 11 of 81 The Turkey Point reactor core consists of 157 fuel assemblies. The full core isotopic inventory is determined in accordance with Reg. Guide 1.183, Regulatory Position 3.1, using the ORIGEN-2.1 isotope generation and depletion computer code (part of the SCALE-4.3 system of codes) to develop the isotopics for the specified burnup, enrichment, and burnup rates (power levels). The plant-specific isotopic source terms are developed using a bounding approach.

The assembly source term is based on 2644 MWth, with 0.3% calorimetric uncertainty, or 2652 MWth, which exceeds the current rated core thermal power of 2300 MWh. For non-LOCA events with fuel failures, a bounding radial peaking factor of 1.65 is then applied to conservatively simulate the effect of power level differences across the core that might affect the localized fuel failures for assemblies containing the peak fission product inventory.

The following assumptions are applied to the source term calculations:

1. A conservative maximum fuel assembly uranium loading (463 kilograms) is assumed to apply to all 157 fuel assemblies in the core.

2. Sensitivity studies were performed to assess the bounding fuel enrichment and bounding burnup values.

3. Radioactive decay of fission products during refueling outages is ignored in the source term calculation.

4. When adjusting the primary coolant isotopic concentrations to achieve Technical Specification limits, the relative concentrations of fission products in the primary coolant system are assumed to remain constant.

1.7.2 Primary Coolant Source Term The primary coolant source term for Turkey Point is calculated based upon maximum equilibrium concentrations from operation at 2652 MWth with small defects in I percent of the fuel rod cladding.

The equilibrium iodine activities were then adjusted to achieve the proposed Technical Specification 3.4.8 limit of 0.25 gtCi/gm dose equivalent 1-131 using the proposed Technical Specification definition of Dose Equivalent 1-131 (DE 1-131).

The non-iodine activities were determined by first developing a list of isotopes which satisfied the radionuclide requirements specified in Table 5 of Reg. Guide 1.183. Iodine nuclides and isotopes with half lives less than 30 days were deleted from this list in accordance with the current Technical Specification definition of E-bar. Equilibrium RCS activities based upon 1% fuel defects were combined with corrosion product activities from ANSI/ANS-1 8.1-1999 for Cr-5 1, Fe-55, F-59, and Mn-54 to calculate a total RCS specific activity. This value was found to be slightly greater than the Tech. Spec. limit of 100/E-bar. The activity for each isotope was then adjusted by a constant factor such that the sum of the adjusted activities was equal to the 100/E-bar limit. The adjusted primary coolant source term activities for the applicable Table 5 list of isotopes is presented in Table 1.7.2-1, "Primary Coolant Source Term."

A value for DE Xe-133 was calculated using the proposed Technical Specification definition and the equilibrium noble gas activities based upon 1% fuel defects. This value was corrected using the same adjustment factor needed to achieve a total specific activity equal to 100/E-bar described above. The resulting adjusted DE Xe-133 is equal to 447.7 tCi/gm. An RCS DE Xe-133 limit of 447.7 PCi/gm ensures that the maximum RCS activity will remain below the current Tech. Spec. limit based upon E-bar.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 12 of 81 SOLUSINeNGINeeRINGANDSOiNARP o 1.7.3 Secondary Side Coolant Source Term Secondary coolant system activity is limited to a value of < 0. 10 gCi/gm dose equivalent 1-131 in accordance with TS 3.7.1.4. Noble gases entering the secondary coolant system are assumed to be immediately released; thus the noble gas activity concentration in the secondary coolant system is assumed to be 0.0 gCi/gm. Thus, the secondary side iodine activity is 40% of the activity given in Table 1.7.2-1.

The secondary side source term is presented in Table 1.7.3-1, "Secondary Side Source Term (non-LOCA)."

1.7.4 Core Source Term Per Section 3.1 of Reg. Guide 1.183, the inventory of fission products in the Turkey Point reactor core and available for release to the containment is based on the maximum full power operation of the core and the current licensed values for fuel enrichment and fuel burnup. The period of irradiation is selected to be of sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. In addition, all fuel assemblies in the core are assumed to be affected and the core average inventory is used.

The core source term is based on an "average" assembly with a bounding core average burnup of 45,000 MWD/MTU and a bounding average assembly power* of 16.892 MWth. The minimum fuel enrichment is based on an historical minimum of 3.0 w/o and a bounding maximum assumed fuel enrichment of 5.0 w/o. The limiting isotopic concentration from either enrichment was used. It is conservatively assumed that a maximum assembly uranium mass of 463,000 gm applies to all of the.

fuel assemblies.

Average assembly power = (2652 MWh)(1 / 157 assemblies) = 16.892 MWdh / assembly The ORIGEN runs used cross section libraries that correspond to PWR extended burnup fuel. Decay time between cycles is conservatively ignored. For each nuclide, the bounding activity for the allowable range of enrichments is determined.

The core source term is presented in Table 1.7.4-1, "Core Source Term."

1.7.5 Fuel Handling Accident Source Term The fuel handling accident for Turkey Point assumes the failure of one assembly; therefore, the fuel handling accident source term is based on a single "bounding" fuel assembly.

Per Section 3.1 of Reg. Guide 1.183, the source term methodology for the Fuel Handling Accident is similar to that used for developing the LOCA source term, except that for DBA events that do not involve the entire core, the fission product inventory of each of the damaged fuel rods is determined by dividing the total core inventory by the number of fuel rods in the core. To account for differences in power level across the core, a radial peaking factor of 1.65 is applied in determining the inventory of the damaged rods.

The LOCA source term is based on the activity of 157 fuel assemblies and the radial peaking factor is 1.65. Thus, based on the methodology specified in Reg. Guide 1.183, the fuel handling accident source term is derived by applying a factor of 1.65/157 to the LOCA source term.

The FHA source term is presented in Table 1.7.5-1, "Fuel Handling Accident Source Term."

1.7.6 Gap Release Fractions The core inventory release fractions for the gap release and early in-vessel damage phases used for the design basis LOCA are provided in Reg. Guide 1.183, Regulatory Position 3.2, Table 2, "PWR Core Inventory Fraction Released into Containment." For the RCCA Ejection accident, the fraction of the core inventory assumed to be in the fuel rod gap are those event-specific source term requirements listed in Appendix H of Reg. Guide 1.183. The fraction of the core inventory assumed to be in the fuel rod gap for the Fuel Handing and Spent Fuel Cask Drop accidents are discussed below. For the other non-LOCA events, the fraction of the core inventory assumed to be in the gap are consistent with Reg.

Guide 1.183, Regulatory Position 3.2, Table 3, "Non-LOCA Fraction of Fission Product Inventory in Gap."

For the Fuel Handling and Spent Fuel Cask Drop events, the gap fractions specified in Table 3 of the Reg. guide are modified to account for high burnup fuel. Footnote 11 on Table 3 establishes burnup limits for the applicability of the gap inventory for non-LOCA events. Consideration is given to fuel with a current burnup greater than 54,000 MWD/MTU which may have exceeded the 6.3 kw/ft linear heat generation rate during a previous operating cycle. This is done using the guidance of NUREG/CR-5009, which endorses the gap release fractions for fuel handling events outlined in Reg.

Guide 1.25 with some modification for higher burnups. The following table lists the modified noble gas and iodine gap fractions from NUREG/CR-5009, which are approximately twice those of Reg.

Guide 1.183:

Group Fraction 1-131 0.12 Kr-85 0.30 Other Noble Gases 0.10 Other Halogens 0.10 Although only a few rods may have exceeded the burnup limits of Table 3 of Reg. Guide 1.183, these values are conservatively applied to the entire fuel assembly. This methodology was approved for use with high burnup rods in the fuel handling accident at Indian Point 3 in License Amendment No. 215.

1.8 Atmospheric Dispersion (X10) Factors 1.8.1 Onsite XIQ Determination New X/Q factors for onsite release-receptor combinations are developed using the ARCON96 computer code ("Atmospheric Relative Concentrations in Building Wakes," NUREG/CR-633 1, Rev. 1, May 1997, RSICC Computer Code Collection No. CCC-664). Additionally, NRC Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," June 2003, has been implemented. Reg. Guide 1.194 contains new guidance that supersedes the NUREG/CR-6331 recommendations for using certain default parameters as input. Therefore, the following changes from the default values are made:

" For surface roughness length, m, a value of 0.2 is used in lieu of the default value of 0.1, and

" For averaging sector width constant, a value of 4.3 is used in lieu of the default value of 4.0.

A number of various release-receptor combinations are considered for the onsite control room atmospheric dispersion factors. These different cases are considered to determine the limiting release-receptor combination for the events.

Figure 1.8.1-1 provides a sketch of the general layout of Turkey Point that has been annotated to highlight the release and receptor point locations described above, among others. All releases are taken as ground releases per guidance provided in Reg. Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Rev. 1, February 1983.

Table 1.8.1 -1; "Release-Receptor Combination Parameters for Analysis Events," provides information related to the relative elevations of the release-receptor combinations, the straight-line horizontal distance between the release point and the receptor location; and the direction (azimuth) from the receptor location to the release point. Angles are calculated based on trigonometric layout of release and receptor points in relation to the North-South and East-West axes. Plant North is aligned with True North.

Table 1.8.1-2, "Onsite Atmospheric Dispersion Factors (X/Q) for Analysis Events," provides the Control Room X/Q factors for the release-receptor combinations listed above. These factors are not corrected for occupancy. This table summarizes the X/Q factors for the control room intakes used in the various accident scenarios for onsite control room dose consequence analyses. Values are presented for the normal intake prior to control room isolation and for the unfavorable emergency intake during control room isolation.

Table 1.8.1-3, "Release-Receptor Point Pairs Assumed for Analysis Events," identifies the Release-Receptor pair and associated Control Room X/Q factors from Table 1.8.1-2 that are used in the event analyses during each of the modes of control room ventilation.

A building wake term is only applied to releases directly from the containment surface. The building area used for this wake term is 1,254 mi2 . This value is calculated to be conservatively small in that the height used in the area calculation is from the highest roof elevation of a nearby building to the elevation of the bottom of the containment dome.

Section 3.3.2.2 of Reg. Guide 1.194 allows for the use of an effective X/Q for dual intake arrangements if the two intakes are not located in the same wind direction. This credit allows for a reduction in the X/Qs to the more limiting intake in proportion to the relative flow rate through the intakes. The control room emergency intakes are being relocated into separate wind sectors for all release points and will be balanced to have equal flow rates. Thus, the dual intake dilution credit enables the X/Qs to the most limiting emergency intake to be reduced by a factor of two in the event analyses.

1.8.2 Offsite X/Q Determination For offsite receptor locations, the new atmospheric dispersion (X/Q) factors are developed using the PAVAN computer code ("PAVAN: An Atmospheric Dispersion Program for Evaluating Design Bases Accident Releases of Radioactive Material from Nuclear Power Stations," NUREG/CR-2858, November 1982, RSICC Computer Code Collection No. CCC-445). The offsite maximum X/Q factors for the EAB and LPZ are presented in Table 1.8.2-1, "Offsite Atmospheric Dispersion Factors (X/Q)."

In accordance with Regulatory Position 4 from NUREG/CR-2858, the maximum value from all downwind sectors for each time period are compared with the 5% overall site X/Q values for those boundaries, and the larger of the values are used in evaluations. Note that the 0-2 hour EAB atmospheric dispersion factor is applied to all time periods in the analyses.

All of the releases are considered ground level releases because the highest possible release elevation is 200 feet (from the plant stack). From Section 1.3.2 of Reg. Guide 1.145, a release is only considered a stack release if the release point is at a level higher than two and one-half times the height of adjacent solid structures. For the Turkey Point plant, the elevation of the top of the containment structures is given as 186 ft and 4-3/8 in. The highest possible release point is not 2.5 times higher than the adjacent containment building; therefore, all releases are considered ground level releases. As such, the release height is set equal to 10.0 meters as required by Table 3.1 of NUREG/CR-2858. The building area used for the building wake term is the same as for some of the ARCON96 onsite X/Q cases. The building height entered into PAVAN is the top elevation of the cylindrical portion of the containment building of 170.28 ft less the plant grade elevation of 18 ft.

1.8.3 Meteorological Data Meteorological data over a five-year period (2003 through 2007) is used in the development of the new onsite XIQ factors used in the analysis. The meteorological data is converted from the raw format into the proper formatting required to create the meteorological data files for the ARCON96 runs. Five years worth of meteorological data is used which meets the guidance set forth in Section 3.1 of Reg.

Guide 1.194. The raw data for 2003 through 2007 was provided in electronic format. The data from these files was manipulated within a spreadsheet for appropriate formatting for use with ARCON96.

ARCON96 analyzes the meteorological data file used and lists the total number of hours of data processed and the number of hours of missing data in the case output. A meteorological data recovery rate may be determined from this information. For the 2003 to 2007 data base, the meteorological data recovery rate is 97.5%. No regulatory guidance is provided in Reg. Guide 1.194 and NUREG/CR-6331 on the valid meteorological data recovery rate required for use in determining onsite X/Q values.

However, Regulatory Position C.5 of Reg. Guide 1.23 requires a 90% data recovery threshold for measuring and capturing meteorological data. Clearly, the 97.5% valid meteorological data rate for the cases in this analysis exceeds the 90% data recovery limit set forth by Reg. Guide 1.23. With a data recovery rate of 97.5% and a total of five years worth of data, the contents of the meteorological data file are representative of the long-term meteorological trends at the Turkey Point site.

The meteorological data were also provided in annual joint frequency distribution format for 2005 through 2007 for the offsite analysis. The joint frequency distribution file requires the annual meteorological data to be sorted into several classifications. This is accomplished by using three classifications that include wind direction, wind speed, and atmospheric stability class. The format for the file conforms to the format provided in Table 1 of Reg. Guide 1.23, with the exceptions of a category for the wind direction and that the wind directions are listed from NNE to N instead of N to NNW. These data are provided for each year in terms of the percent of hours of that year that fell into each classification category. The data for each category (i.e. wind speed, wind direction, and stability class unique combination) were converted from percent to number of hours.

The number of hours for each classification is then rounded to the nearest whole hour. The total values for each stability class are then transposed so that the rows correspond to the wind speed bins and the columns correspond to the wind directions. The wind directions are then ordered properly so that the first column corresponds to the north wind direction and the last column corresponds to the NNW direction as required by the PAVAN code. The final ordered numbers are used in the input file for PAVAN.

The tower height at which the wind speeds are measured is 11.58 meters above plant grade. There were 81 calm hours in the three year joint frequency data. This low number of calm hours is likely due to the positioning of the Turkey Point plant and its proximity to the Atlantic Ocean. The highest windspeed category is classified in both Reg. Guide 1.23, "Onsite Meteorological Programs," February 1972 and in the raw data as 'greater than 24 mph', however the PAVAN code requires that the maximum speed for each category be input. Therefore, a 65-mph value is chosen as the upper limit on

M N NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 16 of 81 the fastest windspeed category because the raw meteorological data showed that there were no hours with windspeeds faster than 65 mph for the period under consideration.

An additional process was performed using the available meteorological to determine the average air temperature swing over a 24-hour period. This was done by combining the yearly ARCON96 data files from 1997 to 2001 and calculating the average temperature swing over any 24-hour period and the median 24-hour temperature change. The average air temperature range over the five years of meteorological data was calculated to be 10.1 'F, with a median temperature swing of 9.5 'F. The higher value is used to support determining the leakage rate from the RWST.

Another process was performed on the meteorological data used for the ARCON96 runs to determine the 95th percentile wind speed at the limiting MSSV release height. The limiting release height is the one at which the calculated 95th percentile wind speed is the greatest. To determine the 95h percentile wind speed at the MSSV release height, the meteorological data used for the ARCON96 runs was evaluated. Hourly entries with bad data were neglected in the evaluation. Wind speed multipliers for the MSSV release were selected based upon the stability class for each hour of data. The wind speed multiplier and the 10 m wind speed were multiplied together to obtain the wind speed at the height of the release for each hour of data.

The wind speed multiplier is selected based upon the stability class, and is taken from ARCON96 case runs. The results of these case runs present the wind speed correction factors for the MSSV release location.

The valid hourly release height wind speeds were then utilized to determine the 95th percentile value; The 95h percentile wind speed for the Unit 4 MSSV release height is 24.6 miles per hour. That is, 95% of all of the hourly wind speeds at the MSSV release height are less than 24.6 miles per hour.

The minimum exit velocity for the MSSVs and ADVs at the opening setpoint are greater than 5 times the above listed 95th percentile wind speed for the duration of the steam release through the valves.

The MSSV exit velocity is greater than 125.7 ft/s and the nominal ADV exit velocity is 194.0 ft/s. For these conditions, the plume rise credit discussed in Section 6 of Reg. Guide 1.194 may be applied.

This factor of 5 reduction to the X/Q values for the MSSV and ADV releases was only credited prior to the beginning of RCS cooldown. This conservatively neglects any plume rise credit which may be applicable during the early portion of the cooldown.

1.9 Sump pH Determination Standard Review Plan Section 6.5.2 requires that the pH of the solution collected in the containment sump after completion of injection of containment spray, ECCS water, and all additives be maintained at a level sufficiently high to provide assurance that significant long-term iodine re-evolution does not occur. The SRP identifies that long-term iodine retention may be assumed only when the equilibrium sump solution pH, after mixing and dilution with the primary coolant and ECCS injection is above 7; and should be achieved by the onset of the spray recirculation mode.

A manual calculation was performed to determine the amount of sodium tetraborate decahydrate that must be located in the containment sump to ensure a pH of at least 7.0 by the time of the onset of containment spray. This calculation used conservative assumptions for sump fill rate, RWST drain rate, RCS volume, RWST transfer volume, sump liquid temperature, and fluid hold-up in containment.

Sodium tetraborate decahydrate dissolution rates were based upon information contained in WCAP-16596=NP (Reference 5.33). This analysis also considered the formation of nitric acid due to the irradiation of water in the sump and the generation of hydrochloric acid due to the radiolysis of electrical cable insulation per the guidelines of NUREG/CR-5950 (Reference 5.30). The NaTB will be placed into two large and eight small baskets. The large baskets will have a length and width of 4.5

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS,... . Turkey Point Units 3 and 4 Page 17 of 81 feet and a height of 2.77 feet. Small baskets will be 3 feet square at the base and are 2.5 feet high. A parametric study was performed to determine the minimum and maximum NaTB loading of the baskets. Lower NaTB levels reduce the total amount of buffering agent available for dissolution.

Higher NaTB levels require higher sump liquid levels before the top surface area of the NaTB in the baskets become exposed to water. The amount of buffering agent required and the resulting pH values are presented in Table 1.9-1.

2.0 Radiological Consequences - Event Analyses 2.1 Loss of Coolant Accident (LOCA) 2.1.1 Background This event is assumed to be caused by an abrupt failure of the main reactor coolant pipe and the ECCS fails to prevent the core from experiencing significant degradation (i.e., melting). This sequence cannot occur unless there are multiple failures, and thus goes beyond the typical design basis accident that considers a single active failure. Activity is released from the containment and from there, released to the environment by means of containment leakage and leakage from the ECCS. This event is described in the Section 14.3.5 of the UFSAR.

2.1.2 Compliance with Reg. Guide 1.183 Regulatory Positions The LOCA dose consequence analysis is consistent with the guidance provided in Reg. Guide 1.183, Appendix A, "Assumptions for Evaluating the Radiological Consequences of a LWR Loss-of-Coolant Accident," as discussed below:

I. Regulatory Position 1 - The total core inventory of the radionuclide groups utilized for determining the source term for this event is based on Reg. Guide 1.183, Regulatory Position 3.1. A conservative power level is used which exceeds 102% of the rated core thermal power. The resulting core source term is provided in Table 1.7.4-1. The core inventory release fractions for the gap release and early in-vessel damage phases of the LOCA are consistent with Regulatory Position 3.2 and Table 2 of Reg.

Guide 1.183.

2. Regulatory Position 2 - The sump pH is controlled at a value greater than 7.0 based on the addition of sodium tetraborate decahydrate baskets. Therefore, the chemical form of the radioiodine released to the containment is assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. With the exception of elemental and organic iodine and noble gases, fission products are assumed to be in particulate form.

3. Regulatory Position 3.1 - The activity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the containment. The release into the containment is assumed to terminate at the end of the early in-vessel phase.

4. Regulatory Position 3.2 - Reduction of the airborne radioactivity in the containment by natural deposition is credited. A natural deposition removal coefficient for elemental iodine is calculated per SRP 6.5.2 as 5.58 hr-1. This removal is credited in both the sprayed and unsprayed regions of containment.

A natural deposition removal coefficient of 0.1 hr 1 is assumed for all aerosols in the unsprayed region of containment as well as in the sprayed region when sprays are not operating. Industry Degraded

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 18 of 81 Sý1TINS INENGINEERNG ANDSOFW Pg 8 f8 Core Rulemaking (IDCOR) Program Technical Report 11.3, "Fission Product Transport in Degraded Core Accidents," December 1983, documents results from Containment Systems Experiment testing.

These tests show that settling of aerosols due to gravity is the dominant natural mechanism for fission product retention. This report finds that significant removal by sedimentation would be expected even at very low particulate concentrations. Figure 4-2 of IDCOR Program Technical Report 11.3 shows a ten-fold reduction in the airborne cesium concentration over a 7-hour period at relatively low concentrations. This represents an aerosol removal rate of 0.33 hr-1 . A more conservative value of 0.1 hr1 is used in the analyses based upon NRC approval of this value in the safety evaluations for the St.

Lucie Unit 2 License Amendment No. 152 in September 2008 (ADAMS Accession No. ML082060400) and the Palisades Nuclear Plant License Amendment No. 226 in September 2007 (ADAMS Accession No. ML72470667).

No removal of organic iodine by natural deposition is assumed.

5. Regulatory Position 3.3 - A single train of containment spray provides coverage to 34.5% of the containment volume. Therefore, the Turkey Point containment building atmosphere is not considered to be a single, well-mixed volume. The containment is divided into three regions: sprayed, an unsprayed region above the operating deck and an unsprayed region below the operating deck. The mixing rates for the containment sprayed and unsprayed regions are based on a GOTHIC analysis which produced results consistent with NUREG/CR-4102, "Air Currents Driven by Sprays in Reactor Containment Buildings". Precedent for similar containment mixing results is established in the Fort Calhoun Issuance of Amendment (IA) and Safety Evaluation (SE) for Amendment No. 198 to DPR-40 issued April 4, 2001.

The GOTHIC analysis utilized for Turkey Point to demonstrate the level of spray induced mixing in containment included both subdivided and lumped parameter models. The detailed subdivided models were used to calculate flow patterns produced by the containment sprays and the emergency containment coolers. Gas concentrations from the subdivided models were compared with concentrations in the lumped parameter model and used to determine equivalent mixing flow rates for the lumped model.

Based on the results of this analysis, the AST dose calculations were conducted using a three volume model similar to the lumped parameter GOTHIC model. The AST model includes separate volumes representing the unsprayed lower, unsprayed upper and sprayed upper regions of containment. Mixing flow rates up to 375,000 cfm between lower and upper unsprayed regions and 990,000 cfm between upper sprayed and unsprayed regions conservatively cover the possible combinations of sprays and emergency fans that may be available during an accident scenario.

The method used in the Turkey Point AST LOCA analysis for determining the time period required to reach an elemental iodine DF of 200 was based on a containment atmosphere peak iodine concentration equal to 40 percent of the core iodine inventory per Table 2 of Reg. Guide 1.183.

The SRP requires that the elemental iodine spray removal coefficient should be set to zero when a decontamination factor (DF) of 200 is reached for elemental iodine. In addition, the particulate spray removal coefficient should be reduced by a factor of 10 when a DF of 50 is reached for the aerosols.

As discussed in the SRP, the iodine decontamination factor (DF) is a function of the effective iodine partition coefficient between the sump and containment atmosphere. Thus, the loss of iodine due to other mechanisms (containment leakage, surface deposition, etc.), would not be included in the determination of the time required to reach a DF of 200. In addition, since the iodine in the containment atmosphere and sump are decaying at the same rate, decay should not be included in determining the time to reach a DF of 200. Additional RADTRAD-NAI cases were performed for determining the time to reach a decontamination factor of 200.

The first RADTRAD-NAI case was used to determine the peak containment atmosphere elemental iodine concentration and amount of aerosol in the containment atmosphere. This case included:

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 19 of 81

No containment spray

" No elemental iodine surface deposition

" No aerosol surface deposition

No decay

No containment leakage The second RADTRAD-NAI case determined the time required to reach a DF of 200 based on the peak elemental iodine concentration from the first RADTRAD-NAI case. The second RADTRAD-NAI case included:

" Containment sprays actuated at 0.018 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months

No surface deposition

No decay

" No containment leakage Due to the high mixing rate between the containment regions, the activity in all three containment regions was considered. The second RADTRAD-NAI case showed that a DF of 200 for elemental iodine was reached at a time greater than 2.305 hours0.00353 days 0.0847 hours 5.042989e-4 weeks 1.160525e-4 months .

A separate RADTRAD-NAI case was then used to determine the time required to reach a DF of 50 for aerosols based on the peak aerosol mass from the first RADTRAD-NAI case. This RADTRAD-NAI case included:

" Containment sprays actuated at 0.018 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months

Aerosol surface deposition credited

No decay

No containment leakage Due to the high mixing rate between the containment regions, the activity in all three containment regions is considered. The third RADTRAD-NAI case showed that a DF of 50 was reached at a time greater than 3.06 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Containment spray flow is assumed to be stopped for a period of five minutes to allow for manual re-alignment of the pump suction from the RWST to the recirculation sump. Termination of spray flow is considered in the determination of the iodine decontamination factors and is reflected in the mixing rates between the containment regions

6. Regulatory Position 3.4 - Reduction in airborne radioactivity in the containment by in-containment recirculation filter systems is not credited in this analysis.

7. Regulatory Position 3.5 - This position relates to suppression pool scrubbing in BWRs, which is not applicable to Turkey Point.

8. Regulatory Position 3.6 - This position relates to activity retention in ice condensers, which is not applicable to Turkey Point.

9. Regulatory Position 3.7 - A containment leak rate of 0.20% per day of the containment air is assumed for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months based on proposed Technical Specification 6.8.4h. After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the containment leak rate is reduced to 0.10% per day of the containment air. The containment leakage was applied to all three containment regions.

10. Regulatory Position 3.8 - Routine containment purge is considered in this analysis. The purge release evaluation assumes that 100% of the radionuclide inventory of the RCS is released instantaneously at the beginning of the event. The containment purge flow is 7000 cfm and is isolated after 8 seconds, which is before the onset of the gap release phase. No filters are credited.

11. Regulatory Position 4.1 through 4.6 provide guidance for the evaluation of the transport, reduction, and release of radioactive material through dual containment structures. These positions are not applicable to Turkey Point.

12. Regulatory Position 5.1 - Engineered Safety Feature (ESF) systems that recirculate water outside the primary containment are assumed to leak during their intended operation. With the exception of noble gases, all fission products released from the fuel to the containment are assumed to instantaneously and homogeneously mix in the containment sump water at the time of release from the core.

13. Regulatory Position 5.2 - Leakage from the ESF system is taken as two times the value from UFSAR Table 6.2-12. ECCS leakage is assumed to start at the earliest time the recirculation flow occurs in these systems and continue for the 30-day duration. Backleakage to the Refueling Water Storage Tank is also considered separately as 0.1 gph, which exceeds two times the expected leakage through the two sets of isolation valves between the RWST and recirculation flow. Backleakage to the RWST is assumed to begin at the start of recirculation and continue for the remainder of the 30-day duration.

14. Regulatory Position 5.3 - With the exception of iodine, all radioactive materials in the recirculating liquid are assumed to be retained in the liquid phase.

15. Regulatory Position 5.4 - A flashing fraction of 9.2% was calculated based on a conservative maximum sump liquid temperature and containment design pressure. However, consistent with Regulatory Position 5.5, the flashing fraction for ECCS leakage is assumed to be 10%.

For ECCS leakage back to the RWST, the analysis demonstrates that the temperature of the leaked fluid will cool below 212'F prior to release to the RWST tank.

16. Regulatory Position 5.5 - The amount of iodine that becomes airborne is conservatively assumed to be 10% of the total iodine activity in the leaked fluid for the ECCS leakage entering the Reactor Auxiliary Building. For the ECCS leakage back to the RWST, the sump and pH history and temperature are used to evaluate the amount of iodine that enters the RWST air space.

17. Regulatory Position 5.6 - For ECCS leakage into the auxiliary building, the form of the released iodine is 97% elemental and 3% organic. No credit for ESF filtration of the ECCS leakage nor holdup or dilution in the auxiliary building is taken. For releases from the RWST, the temperature and pH history of the sump and RWST are considered in determining the radioiodine available for release and the chemical form. Credit is taken for dilution of activity in the RWST.

18. Regulatory Position 6 - This position relates to MSSV leakage in BWRs, which is not applicable to Turkey Point.

19. Regulatory Position 7 - Containment purge is not considered as a means of combustible gas or pressure control in this analysis; however, the effect of routine containment purge before isolation is considered.

2.1.3 Methodology For this event, the Control Room ventilation system cycles through both modes of operation (the operational modes are summarized in Table 1.6.3- I). Inputs and assumptions fall into three main categories: Radionuclide Release Inputs, Radionuclide Transport Inputs, and Radionuclide Removal Inputs.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 21 of 81 SOJIONS INENGINEERING ANDsop'nwnAe For the purposes of the LOCA analyses, a major LOCA is defined as a rupture of the RCS piping, including the double-ended rupture of the largest piping in the RCS, or of any line connected to that system up to the first closed valve. Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. A reactor trip signal occurs when the pressurizer low-pressure trip setpoint is reached. A safety injection system signal is actuated when the appropriate setpoint (high containment pressure) is reached. The following measures will limit the consequences of the accident in two ways:

1. Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat, and

2. Injection of borated water provides heat transfer from the core and prevents excessive cladding temperatures.

Release Inputs The core inventory of the radionuclide groups utilized for this event is based on Reg. Guide 1.183, Regulatory Position 3.1, at 2652 MWth and is provided as Table 1.7.4-1. The source term represents end of cycle conditions assuming enveloping initial fuel enrichment and an average core burnup of 45,000 MWD/MTU.

Per TS 3.6.1.2 and proposed TS 6.8.4h, the leakage rate acceptance criteria for the containment is 0.20% of the containment air weight per day. Therefore, for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the containment is assumed to leak at a rate of 0.20% of the containment air per day. Per Reg. Guide 1.183, Regulatory Position 3.7, the primary containment leakage rate is reduced by 50% at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months into the LOCA to 0.10% /day based on the post-LOCA primary containment pressure history.

A containment purge is also assumed coincident with the beginning of the LOCA. Since the purge is isolated prior to the initial release of fission products from the core at 30 seconds, only the initial RCS activity is available for release via this pathway. The release of 7000 cfm is modeled for 8 seconds until isolation occurs.

The ECCS leakage to the auxiliary building is 4,650 cc/hr based upon two times the current licensing basis value of 2,325 cc/hr. The leakage is assumed to start at 15 minutes into the event and continue throughout the 30-day period. This portion of the analysis assumes that 10% of the total iodine is released from the leaked liquid. The form of the released iodine is 97% elemental and 3% organic.

The ECCS backleakage to the RWST is assumed to be 0.1 gph based upon doubling of the expected total seat leakage through both sets of motor operated valves which isolate the recirculation flow from the RWST. The leakage is assumed to start at 15 minutes into the event when recirculation begins and continue throughout the 30-day period. Note that based on the leakage rate and the size of the piping, the leakage would not reach the RWST for an extended period of time after recirculation begins. This time period is conservatively not credited for determining when the leakage reaches the RWST (i.e., the leakage is assumed to reach the RWST instantaneously allowing no time for radioactive decay); however, this time period is credited for determining the temperature of the leakage reaching the RWST. It should also be noted that based on the small leak rate and pipe length between the ECCS isolation valves and the RWST, the leaked ECCS fluid would probably not reach the RWST during the first 30 days of the event.

Based on sump pH history, the iodine in the sump solution is assumed to all be in nonvolatile form.

However, when introduced into the acidic solution of the RWST inventory, there is a potential for the particulate iodine to convert into the elemental form. The fraction of the total iodine in the RWST which becomes elemental is both a function of the RWST pH and the total iodine concentration. The amount of elemental iodine in the RWST fluid which then enters the RWST air space is a function of the temperature-dependent iodine partition coefficient.

The time-dependent concentration of the total iodine in the RWST, including stable iodine, was determined

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 22 of 81

ýýUIONS INENGINEERING ANDS0FT*A from the tank liquid volume and leak rate. This iodine concentration ranged from a minimum value of 0.0 at the beginning of the event to a maximum value of 8.03E-08 gm-atom/liter at 30 days (see Table 2.1-3).

Based on these results, a constant value of 1.OE-07 gm-atom/liter is applied in the analysis. Due to the small backleakage rate, the RWST pH remains at the conservative initial value of 3.0 for the duration of the event. Using this pH and the time-dependent total iodine concentration in the RWST liquid space, the amount of iodine which is converted to the elemental form was determined using guidance provided in NUREG/CR-5950. This RWST elemental iodine fraction ranged from 0.0 at the beginning of the event to a maximum of 0.0882 (see Table 2.1-4). Conservative application of the constant total iodine concentration of L.OE-07 gm-atom/liter resulted in an elemental iodine fraction of 0.1058, which was then used to calculate the iodine release rate from the RWST.

The elemental iodine in the liquid leaked into the RWST is assumed to become volatile and partition between the liquid and vapor space in the RWST based upon the temperature dependent partition coefficient for elemental iodine as presented in NUREG-5950. The RWST is a vented tank; therefore, there will be no pressure transient in the air region that would affect the partition coefficient. Since no boiling occurs in the RWST, the release of the activity from the vapor space within the RWST is calculated based upon the displacement of air by the incoming leakage and the expansion due to the daily heating and cooling cycle of the contents of the RWST. The average daily temperature swing of 10.1 'F is applied for every 24-hour period for 30 days and no credit is taken for cooling of the tank contents via conduction.

The iodine release is implemented via an adjustment to the vapor flow rate from the RWST. This adjustment accounts for the time-dependent relationship between the elemental iodine concentration in the RWST vapor space with respect to the sump iodine concentration. The average adjusted RWST vapor release rate is then applied to the entire iodine inventory in the containment sump.

This same approach is used with the organic iodine. An organic iodine fraction of 0.0015 is assumed in combination with a partition coefficient of 1.0. The particulate portion of the leakage is assumed to be retained in the liquid phase of the RWST. Therefore, the total iodine flow from the RWST represents the sum of the elemental and organic concentrations in the RWST vapor space. The average adjusted RWST release rate is presented in Table 2.1-5.

The release points for each of the above sources is provided in Table 1.8.1-3.

Transport Inputs During the LOCA event, the activity collected in containment is assumed to be released to the environment via a ground level release from the containment building. The containment purge activity is modeled as a ground level release via the plant stack with no filtration. The activity from ECCS components and from RWST leakage are modeled as an unfiltered ground level releases from the location of the RWST. For the ECCS leakage,*the X/Qs from the RWST to the emergency intakes are more limiting than from any of the Auxiliary Building penetrations.

The Control Room atmospheric dispersion factors (X/Qs) used for this event are based on the postulated release locations and the operational mode of the control room ventilation system. The release-receptor point locations are chosen to minimize the distance from the release point to the Control Room air intakes.

Onsite XIQ values for the various combinations of release points and receptor locations are presented in Table 1.8.1-2. Table 1.8.1-3 presents the Release-Receptor pairs applicable to the control room dose from the LOCA release points for the different modes of control room operation during the event.

For the EAB dose analysis, the X/Q factor corresponding to the 0-2 hour time period was used for the entire duration of the event. The LPZ dose is determined using the XJQ factors for the appropriate time intervals also. These X/Q factors are provided in Table 1.8.2-1.

For this event, the Control Room ventilation system cycles through two modes of operation:

Initially the ventilation system is assumed to be operating in normal mode. The air intake to the ControlRoom during this mode is 1000 cfm of unfiltered fresh air.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 23 of 81 Control Room isolation will occur from an SI actuation or high radiation in containment or at the normal intake. A 30-second delay is conservatively applied to account for the time to reach the signal, the diesel generator start time, load sequencing and damper actuation and positioning time.

After isolation of the Control Room normal air intake, the air flow distribution consists of 525 cfm of filtered makeup flow through the worst of the two emergency intakes, 115 cfm of unfiltered inleakage and 375 cfm of filtered recirculation flow.

The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 97.5% for particulates, elemental, and organic iodine.

Removal Inputs Reduction of the airborne radioactivity in the containment by natural deposition is credited. A natural deposition removal coefficient for elemental iodine is calculated per SRP 6.5.2 as 5.58 hr-'. This removal is credited in the sprayed and unsprayed regions. A natural deposition removal coefficient of 0.1 hr1 is assumed for all aerosols in the unsprayed regions. No natural deposition removal of aerosols is credited in the sprayed region. No removal of organic iodine by natural deposition is assumed.

Containment spray provides coverage to 534,442 ft3 of the total 1.55E6 ft3 containment volume. Therefore, the Turkey Point containment building atmosphere is not considered to be a single, well-mixed volume.

The containment is divided into three regions: sprayed and unsprayed regions above the operating deck and an unsprayed region below the operating deck. The mixing rates between the regions are based on a separate sensitivity study evaluating various combinations of containment fans and sprays to produce the most conservative mixing rates. The final conservative mixing rates are 990,000 cfm between the upper sprayed and upper unsprayed containment regions above the operating deck and 375,000 cfm between the lower unsprayed region below the operating deck and the upper unsprayed region above the operating deck.

According to SRP 6.5.2, the effectiveness of elemental iodine removal by the containment sprays is presumed to end when the decontamination factor (DF) reaches a maximum value of 200. The maximum initial airborne elemental iodine concentration is based on the release of 40 percent of the core iodine inventory. With the elemental iodine spray removal rate set to the SRP limit of 20 hr1, the decontamination factor for elemental iodine reaches 200 at just over 2.305 hours0.00353 days 0.0847 hours 5.042989e-4 weeks 1.160525e-4 months .

The spray aerosol removal rate is reduced by a factor of 10 when a DF of 50 is reached. Based upon the calculated aerosol iodine removal rate of 6.44 hr1, the time for containment spray to produce an aerosol decontamination factor of 50 is calculated to be greater than 3.06 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Filter removal in the Control Room Emergency Mode is simulated using conservative assumptions based on plant design data as listed in Table 1.6.3-1.

2.1.4 Radiological Consequences The radiological consequences of the design basis LOCA are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. In addition, the MicroShield code, is used to develop direct shine doses to the Control Room. The post accident doses are the result of four distinct activity releases:

1. Containment leakage.

2. ESF system leakage into the Auxiliary Building.

3. ESF system leakage into the RWST.

4. Containment Purge at event initiation.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS. INC. Turkey Point Units 3 and 4 Page 24 of 81 The dose to the Control Room occupants includes terms for:

1. Contamination of the Control Room atmosphere by intake and infiltration of radioactive material from the containment and from ECCS leakage.

2. External radioactive plume shine contribution from the containment and ECCS leakage releases.

This term takes credit for Control Room structural shielding.

3. A direct shine dose contribution from the Containment's contained accident activity. This term takes credit for both Containment and Control Room structural shielding.

4. A direct shine dose contribution from the activity collected on the Control Room ventilation filters.

As shown in Table 2.1-6, the sum of the results of all dose contributions for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

2.2 Fuel Handling Accident (FHA) 2.2.1 Background This event consists of the drop of a single fuel assembly either in the Fuel Handling Building (FHB) or inside of Containment. The FHA is described in Section 14.2.1.2 of the UFSAR. The UFSAR description of the FHA specifies a case that assumes all of the fuel rods in a single fuel assembly are damaged.

This analysis considers both a dropped fuel assembly inside the containment with the equipment hatch open, and an assembly dropped inside the FHB without credit for filtration of the Fuel Handling Building exhaust. The source term released from the overlying water pool is the same for both the FHB and the containment cases. Reg. Guide 1.183 imposes the same 2-hour criteria for the direct unfiltered release of the activity to the environment for either location.

A minimum water level of 23 feet is maintained above the damaged fuel assembly for both the containment and FHB release locations. This water level ensures an elemental iodine decontamination factor of 285 per the guidance provided in NRC Regulatory Issue Summary 2006-04.

2.2.2 Compliance with Reg. Guide 1.183 Regulatory Positions The FHA dose consequence analysis is consistent with the guidance provided in Reg. Guide 1.183 Appendix B, "Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident,"

as discussed below:

1. Regulatory Position 1.1 - The amount of fuel damage is assumed to be all of the fuel rods in a single fuel assembly per UFSAR Section 14.2.1.2.

2. Regulatory Position 1.2 - The fission product release from the breached fuel is based on Regulatory Positions 3.1 and 3.2 of Reg. Guide 1.183. Section 1.7.5 provides a discussion of how the FHA source term is developed. A listing of the FHA source term is provided in Table 1.7.5-1. The gap activity available for release is modified from that specified by Table 3 of Reg. Guide 1.183 to account for high burnup fuel as described in Section 1.7.6. This activity is assumed to be released from the fuel assembly instantaneously.

3. Regulatory Position 1.3 - The chemical form of radioiodine released from the damaged fuel into the spent fuel pool is assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. The cesium iodide is assumed to completely dissociate in the spent fuel pool resulting in a final iodine distribution of 99.85% elemental iodine and 0.15% organic iodine.

4. Regulatory Position 2 - A minimum water depth of 23 feet is maintained above the damaged fuel assembly. Therefore, a decontamination factor of 285 is applied to the elemental iodine and a decontamination factor of 1 is applied to the organic iodine. As a result, the breakdown of the iodine species above the surface of the water is 57% elemental and 43% organic. Guidance for the use of 285 for the elemental iodine decontamination factor is provided in NRC Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternate Source Terms."

5. Regulatory Position 3 - All of the noble gas released is assumed to exit the pool without mitigation.

All of the non-iodine particulate nuclides are assumed to be retained by the pool water.

6. Regulatory Position 4.1 - The analysis models the release to the environment over a 2-hour period.

7. Regulatory Position 4.2 - No credit is taken for filtration of the release.

8. Regulatory Position 4.3 - No credit is taken for dilution of the release.

9. Regulatory Position 5.1 - The containment equipment hatch is assumed to be open at the time of the fuel handling accident.

10. Regulatory Position 5.2 - No automatic isolation of the containment is assumed for the FHA.

11. Regulatory Position 5.3 - The release from the fuel pool is assumed to leak to the environment over a two-hour period.

12. Regulatory Position 5.4 - No ESF filtration of the containment release is credited.

13. Regulatory Position 5.5 - No credit is taken for dilution or mixing in the containment atmosphere.

2.2.3 Methodology The input assumptions used in the analysis of the FHA dose consequences are provided in Table 2.2-1. It is assumed that the fuel handling accident occurs at 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after shutdown of the reactor per TS LCO 3.9.3. 100% of the gap activity is assumed to be instantaneously released from a single fuel assembly into the fuel pool. A minimum water level of 23 feet is maintained above the damaged fuel for the duration of the event. All of the noble gas released from the damaged fuel assembly is assumed to escape from the pool. All of the non-iodine particulates released from the damaged fuel assembly are assumed to be retained by the pool liquid. The iodine released from the damaged fuel assembly is assumed to be composed of 99.85% elemental and 0.15% organic. The activity released from the pool is then assumed to leak to the environment over a two-hour period. No credit for dilution in the containment or FHB is taken.

The FHA source term meets the requirements of Regulatory Position 1 of Appendix B to Reg. Guide 1.183.

Gap release fractions have been increased to address fuel with rod average burnups greater than 54,000 MWD/MTU. The FHA source term is listed in Table 1.7.5-1 and the applicable gap release fractions are presented in Section 1.7.6.

For this event, the Control Room ventilation system cycles through two modes of operation:

Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 1000 cfm of unfiltered fresh air.

NUMERICAL AST Licensing Technical Report for NAI- 1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 26 of 81 SOUTIONSINENGINLERING ANDSOFWAREM For the FHA in containment, Control Room isolation occurs on high radiation on the containment radiation monitors. A 30 second delay time is assumed for signal processing and damper closure.

For the release from the Fuel Handling Building, the Control Room is assumed to be manually isolated by the operators 30 minutes after the beginning of the event. After isolation, the air flow distribution consists of 525 cfm of filtered makeup flow from the outside, 115 cfm of unfiltered inleakage, and 375 cfm of filtered recirculation flow.

The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 97.5% for particulate, elemental iodine, and organic iodine.

The atmospheric dispersion factors (XIQs) used for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These X/Qs are summarized in Tables 1.8.1-2 and 1.8.1-3. The Control Room atmospheric dispersion factors applied to the FHA in containment are based upon a release from the most limiting containment equipment/personnel hatch. For the Fuel Handling Building release, the most limiting X/Qs correspond to a release from the Unit 4 spent fuel pool.

The EAB and LPZ doses are determined using the X/Q factors for the appropriate time intervals. These XIQ factors are provided in Table 1.8.2-1.

2.2.4 Radiological Consequences The radiological consequences of the FHA are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. As shown in Table 2.2-2 the results for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

2.3 Main Steamline Break (MSLB) 2.3.1 Background This event consists of a double-ended break of one main steam line outside of containment. The radiological consequences of such an accident bound those of a MSLB inside containment. The faulted steam generator rapidly depressurizes and releases the initial contents of the steam generator secondary to the environment. Plant cool down is achieved via the remaining unaffected steam generators. This event is described in UFSAR Section 14.2.5.

2.3.2 Compliance with Reg. Guide 1.183 Regulatory Positions The analysis of the MSLB dose consequences is consistent with the guidance provided in Reg. Guide 1.183, Appendix E, "Assumptions for Evaluating the Radiological Consequences of a PWR Main Steam Line Break Accident," as discussed below:

1. Regulatory Position 1 - No fuel damage is postulated to occur for the Turkey Point MSLB event.

2. Regulatory Position 2 - No fuel damage is postulated to occur for the Turkey Point MSLB event.

Therefore, two cases of iodine spiking are evaluated.

3. Regulatory Position 2.1 - One iodine spiking case assumes a reactor transient prior to the postulated MSLB that raises the primary coolant iodine concentration to the maximum allowed by TS 3.4.8 Figure 3.4-1, which is a value of 60.0 gCi/gm DE 1-131. This is the pre-accident spike case.

4. Regulatory Position 2.2 - One case assumes the transient associated with the MSLB causes an iodine

NUMERICAL AST Licensing Technical Report for NAI- 1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 27 of 81 INENGINE*ERING SOL.UTIONS ANSI SOFITWARE*

spike. The spiking model assumes the primary coolant activity is initially at the proposed TS 3.4.8 value of 0.25 g.tCi/gm DE 1-131. Iodine is assumed to be released from the fuel into the RCS at a rate of 500 times the iodine equilibrium release rate for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This is the accident-induced spike case.

5. Regulatory Position 3 - The activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant.

6. Regulatory Position 4 - Iodine releases from the steam generators to the environment are assumed to be 97% elemental and 3% organic.

7. Regulatory Position 5.1 - The primary-to-secondary leak rate is equal to the value specified by proposed TS 6.8.4.j.b.2, which is 0.6 gpm through all steam generators and 0.2 gpm through any one steam generator at room temperature conditions.

8. Regulatory Position 5.2 - The density used in converting primary-to-secondary volumetric leak rates to mass leak rates is 62.4 lbm/ft3, which is consistent with the leakage limits at room temperature conditions.

9. Regulatory Position 5.3 - The primary-to-secondary leak rate is assumed to continue until the temperature of the leakage is less than 212'F. This is conservatively calculated to occur at 125.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The release of radioactivity from the unaffected steam generators is conservatively assumed to continue until RHR is capable of removing decay heat and for providing for any further cooldown, which occurs at 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months .

10. Regulatory Position 5.4 - All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.

11. Regulatory Position 5.5.1 - In the faulted steam generator, all of the primary-to-secondary leakage is assumed to flash to vapor and be released to the environment with no mitigation. For the unaffected steam generators used for plant cooldown, a portion of the leakage is assumed to flash to vapor based

on the thermodynamic conditions in the reactor and secondary immediately following plant trip when tube uncovery is postulated. The primary-to-secondary leakage is assumed to mix with the secondary water without flashing during periods of total tube submergence.

12. Regulatory Position 5.5.2 - Any postulated leakage that immediately flashes to vapor is assumed to rise through the bulk water of the steam generator into the steam space and is assumed to be immediately released to the environment with no mitigation; i.e., no reduction for scrubbing within the bulk water is credited.

13. Regulatory Position 5.5.3 - All leakage that does not immediately flash is assumed to mix with the bulk water.

14. Regulatory Position 5.5.4 - The radioactivity within the bulk water is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient of 100 is assumed for the iodine. The retention of particulate radionuclides in the unaffected steam generators is limited by the moisture carryover. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the steam generator carryover rate of less than 1%. No reduction in the release is assumed from the faulted steam generator.

15. Regulatory Position 5.6 - Steam generator tube bundle uncovery in the intact steam generators is postulated for up to 30 minutes following a reactor trip. During this period, the fraction of primary-to-secondary leakage which flashes to vapor is assumed to rise through the bulk water into the steam space and is assumed to be immediately released to the environment with no mitigation. The flashing

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 28 of 81 SOLUIONS INENGINEERING AD OF*NR fraction is based on the thermodynamic conditions in the reactor and secondary coolant. The leakage which does not flash is assumed to mix with the bulk water in the steam generator.

2.3.3 Other Assumptions

1. This analysis assumes that the equilibrium specific activity on the secondary side of the steam generators is equal to the TS 3.7.1.4 limit of 0.1 gtCi/gm Dose Equivalent 1-131.

2. The steam mass release rates for the intact steam generators are provided in Table 2.3-2.

3. Data used to calculate the iodine equilibrium appearance rate are provided in Table 2.3-4, "Iodine Equilibrium Appearance Assumptions."

4. This evaluation assumes that the RCS mass remains constant throughout the MSLB event. No change in the RCS mass is assumed as a result of the primary-to-secondary leakage or from the safety injection system.

5. Radionuclide concentrations in the secondary side fluid of the steam generators assume that Auxiliary Feedwater is provided to maintain a constant secondary mass during periods of steam release.

6. Releases from the faulted main steam line are postulated to occur from the main steam line associated with the most limiting atmospheric dispersion factors. Releases from the unaffected steam generators are postulated to occur from the MSSV or ADV with the most limiting atmospheric dispersion factors.

7. The steam generator partition factor for iodine and particulates is applied in the analysis by reducing the steam release rate from the steam generator compartment. This methodology conservatively allows the activity which is not released to remain in the steam generator compartment and contribute to the radionuclide concentration.

2.3.4 Methodology Input assumptions used in the dose consequence analysis of the MSLB are provided in Table 2.3-1. The postulated accident assumes a double-ended break of one main steam line outside containment. The radiological consequences of such an accident bound those of a MSLB inside of containment. Upon a MSLB, the faulted steam generator rapidly depressurizes and releases the initial contents to the environment. Plant cooldown is achieved via the remaining unaffected steam generators.

The analysis assumes that activity is released as reactor coolant enters the steam generators due to primary-to-secondary leakage. The source term for this activity is presented in Table 1.7.2-1. All noble gases associated with this leakage are assumed to be released directly to the environment. Primary-to-secondary leakage into the faulted steam generator is also assumed to directly enter the atmosphere. Leakage into the unaffected steam generators is partitioned by the secondary fluid and released via the MSSVs and ADVs.

All primary-to-secondary leakage is assumed to continue until the primary system is cooled to 212 'F at 125.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The release of the initial iodine content of the steam generator secondary is also considered.

The source term for this activity is presented in Table 1.7.3-1.

Fuel damage is not postulated for the MSLB event. Consistent with Regulatory Guide 1.183, Appendix E, Regulatory Position 2, if no or minimal fuel damage is postulated for the limiting event, the activity released is assumed as the maximum allowed by Technical Specifications for two cases of iodine spiking:

(1) maximum pre-accident iodine spike; and (2) maximum accident-induced or concurrent, iodine spike.

For the case of a pre-accident iodine spike, a reactor transient is assumed to have occurred prior to the postulated MSLB event. The primary coolant iodine concentration is increased to the maximum value of

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 29 of 81 60 gCi/gm DE 1-131 permitted by TS 3.4.8. The iodine activities for the pre-accident spike case are presented in Table 2.3-3.

For the case of the accident-induced spike, the postulated MSLB event induces an iodine spike. The RCS activity is initially assumed to be 0.25 g.tCi/gm DE 1-131 as allowed by proposed TS 3.4.8. Iodine is released from the fuel into the RCS at a rate of 500 times the iodine equilibrium release rate for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . With iodine activity at equilibrium, the iodine release rate is equal to the rate at which iodine is lost due to decay, purification, and primary system leakage. Parameters used in the determination of the iodine equilibrium release rate are provided in Table 2.3-4. The iodine activities for the accident-induced (concurrent) iodine spike case are presented in Table 2.3-5.

For this event, the Control Room ventilation system cycles through two modes of operation:

" Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 1000 cfm of unfiltered fresh air.

" Control Room is isolated following receipt of a safety injection signal. A 41.5-second delay is applied to account for the signal processing and damper closure time. After isolation, the air flow distribution consists of 525 cfm of filtered makeup flow from the outside, 115 cfm of unfiltered inleakage, and 375 cfm of filtered recirculation flow.

" The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 97.5% for particulate, elemental iodine, and organic iodine.

The atmospheric dispersion factors (XIQs) used for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These X/Qs are summarized in Tables 1.8.1-2 and 1.8.1-3. Releases from the intact steam generators are assumed to occur from the MSSV/ADV which produces the most limiting X/Q. Releases from the faulted steam generator are assumed to occur from the location on a steam line closest the in-service intake.

The EAB and LPZ doses are determined using the X/Q factors for the appropriate time intervals. These XIQ factors are provided in Table 1.8.2-1.

2.3.5 Radiological Consequences The radiological consequences of the MSLB Accident are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. Cases for MSLB pre-accident and concurrent iodine spikes are analyzed. As shown in Table 2.3-6, the results of both cases for EAB dose, LPZ dose, and Control Room dose are within the appropriate regulatory acceptance criteria.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 30 of 81 2.4 Steam Generator Tube Rupture (SGTR) 2.4.1 Background This event is assumed to be caused by the instantaneous rupture of a Steam Generator tube that relieves to the lower pressure secondary system. No melt or clad breach is postulated for the Turkey Point SGTR event. This event is described in UFSAR Section 14.2.4.

2.4.2 Compliance with Reg. Guide 1.183 Regulatory Positions The SGTR dose consequence analysis is consistent with the guidance provided in Reg. Guide 1.183, Appendix F, "Assumptions for Evaluating the Radiological Consequences of a PWR Steam Generator Tube Rupture Accident," as discussed below:

1. Regulatory Position 1 - No fuel damage is postulated to occur for the Turkey Point SGTR event.

2. Regulatory Position 2 - No fuel damage is postulated to occur for the Turkey Point SGTR event. Two cases of iodine spiking are assumed.

3. Regulatory Position 2.1 - One case assumes a reactor transient prior to the postulated SGTR that raises the primary coolant iodine concentration to the maximum allowed by TS 3.4.8 Figure 3.4-1, which is a value of 60.0 pCi/gm DE 1-131 for the analyzed conditions. This is the pre-accident spike case.

4. Regulatory Position 2.2 - One case assumes the transient associated with the SGTR causes an iodine spike. The spiking model assumes the primary coolant activity is initially at the proposed TS 3.4.8 value of 0.25 piCi/gm DE 1-131. Iodine is assumed to be released from the fuel into the RCS at a rate of 335 times the iodine equilibrium release rate for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This is the accident-induced spike case.

5. Regulatory Position 3 - The activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant.

6. Regulatory Position 4 - Iodine releases from the steam generators to the environment are assumed to be 97% elemental and 3% organic.

7. Regulatory Position 5.1 - The primary-to-secondary leak rate is equal to the value specified by proposed TS 6.8.4.j.b.2, which is 0.6 gpm through all steam generators and 0.2 gpm through any one steam generator at room temperature conditions.

8. Regulatory Position 5.2 - The density used in converting primary-to-secondary volumetric leak rates to mass leak rates is 62.4 Ibm/ft 3, which is consistent with the leakage limits at room temperature conditions.

9. Regulatory Position 5.3 - The primary-to-secondary leak rate is assumed to continue until the temperature of the leakage is less than 212'F. This is conservatively calculated to occur at 125.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The release of radioactivity from the unaffected steam generators is assumed to continue until RHR is capable of removing decay heat and for providing for any further cooldown, which occurs at 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months . Termination of the ruptured steam generator activity release is occurs when the ruptured steam generator is isolated at 30 minutes by operator action. While this isolation terminates releases from the ruptured steam generator, primary-to-secondary leakage continues to provide activity for release from the unaffected steam generators.

10. Regulatory Position 5.4- The release of fission products from the secondary system is evaluated with the assumption of a loss of offsite power coincident with reactor trip.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 31 of 8]

SaLTONS SN ENSI NEERINO ANDSOTARIE

11. Regulatory Position 5.5 - All noble gases released from the primary system are assumed to be released to the environment without reduction or mitigation.

12. Regulatory Position 5.6 - Regulatory Position 5.6 refers to Appendix E, Regulatory Positions 5.5 and 5.6. The iodine transport model for release from the steam generators is as follows:

Appendix E, Regulatory Position 5.5.1 - A portion of the primary-to-secondary ruptured tube flow following the SGTR is assumed to flash to vapor based on the thermodynamic conditions in the reactor and secondary. The flashed flow is released to the environment with no mitigation. For the unaffected steam generators, flashing is considered immediately following plant trip when tube uncovery is postulated. The primary-to-secondary leakage is assumed to mix with the secondary water without flashing during periods of total tube submergence.

" Appendix E, Regulatory Position 5.5.2 - The portion of leakage that immediately flashes to vapor is assumed to rise through the bulk water of the steam generator into the steam space and is assumed to be immediately released to the environment with no mitigation; i.e., no reduction for scrubbing within the bulk water is credited.

Appendix E, Regulatory Position 5.5.3 - All of the steam generator leakage and ruptured tube flow that does not immediately flash is assumed to mix with the bulk water.

Appendix E, Regulatory Position 5.5.4 - The radioactivity within the bulk water is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient of 100 is assumed for the iodine. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the steam generator carryover rate of less than 1%.

e Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery in the unaffected steam generators is postulated for up to 30 minutes following a reactor trip. During this period, the fraction of primary-to-secondary leakage which flashes to vapor is assumed to rise through the bulk water of the steam generator into the steam space and is assumed to be immediately released to the environment with no mitigation. The flashing fraction is based on the thermodynamic conditions in the reactor and secondary coolant. The leakage which does not flash is assumed to mix with the bulk water in the steam generator.

2.4.3 Other Assumptions

1. RCS and steam generator volumes are assumed to remain constant throughout both the pre-accident and the accident-induced iodine spike SGTR events.

2. During a SGTR event, from the onset of the tube rupture until the time of reactor trip, there will be no actual steam releases from through ADVs or MSSVs. Radionuclides will most likely enter the atmosphere through the condenser Steam Jet Air Ejector (SJAE). Due to relative proximity of the SJAE to the normal Control Room intake, this release-receptor pair produces a more limiting atmospheric dispersion factor than releases from the ADVs/MSSVs to the normal intake. In addition, the pre-trip flashing fraction of the primary-to-secondary leakage is substantially higher than the pos-trip flashing fraction. For these reasons, the analysis assumes full rated steam flow from the steam generators prior to reactor trip. In addition to the partition factor of 100 in the steam generator, a partition factor of 100 is also applied to iodines and particulates released from the SJAE prior to the reactor trip. This condenser partition factor is no longer used when the steam release from the ADVs/MSSVs begins. The value of 100 for the condenser partition factor is consistent with that approved for use by Kewaunee and South Texas.

3. The steam generator and condenser partition factors for iodine and particulates are applied in the analysis by reducing the steam release rate from the steam generator compartment. This methodology conservatively allows the activity which is not released to remain in the steam generator compartment and contribute to the radionuclide concentration.

4. This analysis assumes that the equilibrium specific activity on the secondary side of the steam generators is equal to the TS 3.7.1.4 limit of 0.1 gCi/gm Dose Equivalent 1-131.

5. Radionuclide concentrations in the secondary side fluid of the steam generators assume that Auxiliary Feedwater is provided to maintain a constant secondary mass during periods of steam release.

6. The steam release rates and ruptured tube flow rates are provided in Table 2.4-2.

7. Data used to calculate the iodine equilibrium appearance rate are provided in Table 2.4-4, "Iodine Equilibrium Appearance Assumptions."

2.4.4 Methodology Input assumptions used in the dose consequence analysis of the SGTR event are provided in Table 2.4-1.

This event is assumed to be caused by the instantaneous rupture of a steam generator tube releasing primary coolant to the lower pressure secondary system. Initial radionuclide releases occur through the condenser SJAE until the time of reactor trip, thereby causing steam relief directly to the atmosphere from the ADVs or MSSVs. This direct steam relief continues until the ruptured steam generator is isolated at 30 minutes.

A thermal-hydraulic analysis is performed to determine a conservative maximum break flow, break flashing flow, and steam release inventory through the ruptured steam generator relief valves. The analysis assumes that activity is released as reactor coolant enters the steam generators due to primary-to-secondary leakage. The source term for this activity is presented in Table 1.7.2-1. All noble gases associated with this leakage are assumed to be released directly to the environment. Primary coolant is released into the ruptured steam generator through the ruptured tube and from a fraction of the total proposed allowable primary-to-secondary leakage until the ruptured steam generator is isolated at 30 minutes. Additional activity, based on the proposed primary-to-secondary leakage limits, is released via the unaffected steam generators. All primary-to-secondary leakage is assumed to continue until the temperature of the leakage is less than 212'F, which is conservatively calculated to occur at 125.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Steam release from the unaffected steam generators is assumed to continue until RHR is capable of removing decay heat and for providing for any further cooldown, which occurs at 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months . The release of the initial iodine content of the steam generator secondary is also considered. The source term for this activity is presented in Table 1.7.3-1.

Per the Turkey Point UFSAR, Section 14.2.4, no fuel melt or clad breach is postulated for the SGTR event.

Consistent with Reg. Guide 1.183 Appendix F, Regulatory Position 2, if no or minimal fuel damage is postulated for the limiting event, the activity release is assumed as the maximum allowed by Technical Specifications for two cases of iodine spiking: (1) maximum pre-accident iodine spike, and (2) maximum accident-induced, or concurrent, iodine spike.

For the case of a pre-accident iodine spike, a reactor transient is assumed to have occurred prior to the postulated SGTR event. The primary coolant iodine concentration is increased to the maximum value of 60 gCi/gm DE 1-131 permitted by TS 3.4.8. The iodine activities for the pre-accident spike case are presented in Table 2.4-3.

For the case of the accident-induced spike, the postulated STGR event induces an iodine spike. The RCS activity is initially assumed to be 0.25 pCi/gm DE 1-131 as allowed by proposed TS 3.4.8. Iodine is

released from the fuel into the RCS at a rate of 335 times the iodine equilibrium release rate for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . With iodine activity at equilibrium, the iodine release rate is equal to the rate at which iodine is lost due to decay, purification, and primary system leakage. Parameters used in the determination of the iodine equilibrium release rate are provided in Table 2.4-4. The iodine activities for the accident-induced (concurrent) iodine spike case are presented in Table 2.4-5. All other release assumptions for this case are identical to those for the pre-accident spike case.

For this event, the Control Room ventilation system cycles through two modes of operation:

" Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 1000 cfm of unfiltered fresh air.

" Control Room is isolated due to a safety injection, which occurs at 291 seconds. A 30-second delay is applied.to account for the signal processing, diesel start, and damper closure time. After isolation, the air flow distribution consists of 525 cfm of filtered makeup flow from the outside, 115 cfm of unfiltered inleakage, and 375 cfm of filtered recirculation flow.

" The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 97.5% for particulate, elemental iodine, and organic iodine.

The atmospheric dispersion factors (X/Qs) used for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These X/Qs are summarized in Tables 1.8.1-2 and 1.8.1-3. Prior to the time of reactor trip, releases are from the condenser SJAE to the normal intake. Immediately following reactor trip, releases from the steam generators are assumed to occur from the MSSV/ADV which produces the most limiting X/Q. The receptor point shifts to the most liming emergency intake after control room isolation occurs.

The EAB and LPZ doses are determined using the XIQ factors for the appropriate time intervals. These X/Q factors are provided in Table 1.8.2-1.

2.4.5 Radiological Consequences The radiological consequences of the SGTR Accident are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. Two activity release cases corresponding to the RCS maximum pre-accident iodine spike and the accident-induced iodine spike, based on proposed TS 3.4.8 limits, are analyzed. As shown in Table 2.4-6, the radiological consequences of the Turkey Point SGTR event for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

2.5 Locked Rotor 2.5.1 Background This event is caused by an instantaneous seizure of a primary reactor coolant pump rotor. Flow through the affected loop is rapidly reduced, causing a reactor trip due to a low primary loop flow signal. Fuel damage may be predicted to occur as a result of this accident. Due to the pressure differential between the primary and secondary systems and assumed steam generator tube leakage, fission products are discharged from the primary into the secondary system. A portion of this radioactivity is released to the outside atmosphere from the secondary coolant system through the steam generator via the ADVs and MSSVs. In addition, radioactive iodine contained in the secondary inventory prior to the event is released to the atmosphere as a result of steaming from the steam generators following the accident. This event is described in Section 14.1.9 of the UFSAR.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 34 of 81 2.5.2 Compliance with Reg. Guide 1.183 Regulatory Positions The revised Locked Rotor dose consequence analysis is consistent with the guidance provided in RG 1.183, Appendix G, "Assumptions for Evaluating the Radiological Consequences of a PWR Locked Rotor Accident," as discussed below:

1. Regulatory Position 1 - The total core inventory of the radionuclide groups utilized for determining the source term for this event is based on Reg. Guide 1.183, Regulatory Position 3.1, and is provided in Table 1.7.4-1. The inventory provided in Table 1.7.4-1 is then adjusted for the fraction of fuel damaged and a radial peaking factor of 1.65 is applied. The fraction of fission product inventory in the gap available for release due to fuel breach is consistent with Table 3 of Reg. Guide 1.183.

2. Regulatory Position 2 - Fuel damage is assumed for this event.

3. Regulatory Position 3 - Activity released from the damaged fuel is assumed to mix instantaneously and homogeneously throughout the primary coolant.

4. Regulatory Position 4 - The chemical form of radioiodine released from the damaged fuel is assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. Iodine releases from the steam generators to the environment are assumed to be 97% elemental and 3% organic.

These fractions apply to iodine released as a result of fuel damage and to equilibrium iodine concentrations in the RCS and secondary system.

5. Regulatory Position 5.1 - The primary-to-secondary leak rate is equal to the value specified by proposed TS 6.8.4.j.b.2, which is 0.6 gpm through all steam generators and 0.2 gpm through any one steam generator at room temperature conditions.

6. Regulatory Position 5.2 - The density used in converting primary-to-secondary volumetric leak rates to mass leak rates is 62.4 lbm/ft3 , which is consistent with the leakage limits at room temperature conditions.

7. Regulatory Position 5.3 - The release of radioactivity is assumed to continue until RHR is capable of removing decay heat and for providing for any further cooldown, which occurs at 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months .

8. Regulatory Position 5.4 - The analysis assumes a coincident loss of offsite power in the evaluation of fission products released from the secondary system.

9. Regulatory Position 5.5 - All noble gas radionuclides released from the primary system are assumed released to the environment without reduction or mitigation.

10. Regulatory Position 5.6 - Regulatory Position 5.6 refers to Appendix E, Regulatory Positions 5.5 and 5.6. The iodine transport model for release from the steam generators is as follows:

Appendix E, Regulatory Position 5.5.1 - A portion of the primary-to-secondary leakage is assumed to flash to vapor based on the thermodynamic conditions in the reactor and secondary immediately following plant trip when tube uncovery is postulated. The flashed leakage is assumed to be released to the environment with no mitigation. The primary-to-secondary leakage is assumed to mix with the secondary water without flashing during periods of total tube submergence.

Appendix E, Regulatory Position 5.5.2 - The portion of leakage that immediately flashes to vapor is assumed to rise through the bulk water of the steam generator into the steam space and is assumed to be immediately released to the environment with no mitigation; i.e., no reduction for scrubbing within the bulk water is credited.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 35 of 81 SOLUTIONS ANDSOFWARE INENGINEERING Appendix E, Regulatory Position 5.5.3 - All of the steam generator leakage flow that does not immediately flash is assumed to mix with the bulk water.

Appendix E, Regulatory Position 5.5.4 - The radioactivity within the bulk water is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient of 100 is assumed for the iodine. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the steam generator carryover rate of less than 1%.

Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery in the unaffected steam generators is postulated for up to 30 minutes following a reactor trip. During this period, the fraction of primary-to-secondary leakage which flashes to vapor is assumed to rise through the bulk water of the steam generator into the steam space and is assumed to be immediately released to the environment with no mitigation. The flashing fraction is based on the thermodynamic conditions in the reactor and secondary coolant. The leakage which does not flash is assumed to mix with the bulk water in the steam generator.

2.5.3 Other Assumptions I. Reg. Guide 1.183, Section 3.6 - The assumed amount of fuel damage caused by the non-LOCA events is analyzed to determine the fraction of the fuel that reaches or exceeds the initiation temperature of fuel melt and to determine the fraction of fuel elements for which fuel clad is breached. This analysis assumes DNB as the fuel damage criterion for estimating fuel damage for the purpose of establishing radioactivity releases. For the Locked Rotor event, Table 3 of Reg. Guide 1.183 specifies noble gas, alkali metal, and iodine fuel gap release fractions for the breached fuel.

2. This analysis assumes that the DNB fuel damage is limited to 15% breached fuel assemblies.

3. Radionuclide concentrations in the secondary side fluid of the steam generators assume that Auxiliary Feedwater is provided to maintain a constant secondary mass during periods of steam release.

4. The steam mass release rates are provided in Table 2.5-2.

5. This analysis assumes that the equilibrium specific activity on the secondary side of the steam generators is equal to the TS 3.7.1.4 limit of 0.1 gtCi/gm Dose Equivalent 1-131.

2.5.4 Methodology Input assumptions used in the analysis of the dose consequences of the Locked Rotor event are provided in Table 2.5-1. This event is caused by an instantaneous seizure of a primary reactor coolant pump rotor.

Flow through the affected loop is rapidly reduced, causing a reactor trip due to a low primary loop flow signal. Following the reactor trip, the heat stored in the fuel rods continues to be transferred to the reactor coolant. Because of the reduced core flow, the coolant temperatures will rise. The rapid rise in primary system temperatures during the initial phase of the transient results in a reduction in the initial DNB margin and fuel damage.

For the purpose of this dose assessment, a total of 15% of the fuel assemblies are assumed to experience DNB. The activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant. The source term is based upon release fractions from Table 3 of RG 1.183

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS. INC. Turkey Point Units 3 and 4 Page 36 of 81 SýtUONS INENGINEERINO ANDSF ýAR which have been increased by the radial peaking factor of 1.65. The core source term used as the basis for this activity is presented in Table 1.7.4-1. The analysis assumes that activity is released as reactor coolant enters the steam generators due to primary-to-secondary leakage. All noble gases associated with this leakage are assumed to be released directly to the environment. All primary-to-secondary leakage is assumed to continue until the primary system is cooled to 212 'F at 125.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Activity is released to the atmosphere via steaming from the steam generator ADVs and MSSVs until RHR is capable of removing decay heat and for providing for any further cooldown, which occurs at 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months . The release of the initial iodine content of the steam generator secondary is also considered. The source term for this activity is presented in Table 1.7.3-1.

For this event, the Control Room ventilation system cycles through two modes of operation:

" Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 1000 cfm of unfiltered fresh air.

For the secondary release, the Control Room is isolated on a high radiation reading at the normal intake monitors. A 60 second delay is applied to account for the time to reach the setpoint, signal processing, and damper closure time. After isolation, the air flow distribution consists of 525 cfm of filtered makeup flow from the outside, 115 cfm of unfiltered inleakage, and 375 cfm of filtered recirculation flow.

" The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 97.5% for particulate, elemental iodine, and organic iodine.

The atmospheric dispersion factors (X/Qs) used for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These X/Qs are summarized in Tables 1.8.1-2 and 1.8.1-3. Releases from the steam generators are assumed to occur from the MSSV/ADV which produces the most limiting X/Q.

The EAB and LPZ doses are determined using the X/Q factors for the appropriate time intervals. These X/Q factors are provided in Table 1.8.2-1.

2.5.5 Radiological Consequences The radiological consequences of the Locked Rotor event are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. As shown in Table 2.5-3, the results for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

2.6 Rod Cluster Control Assembly (RCCA) Ejection 2.6.1 Background This event consists of the ejection of a single RCCA. This event is the same as the Rod Ejection event referred to in Reg. Guide 1.183. The RCCA Ejection results in a reactivity insertion that leads to a core power level increase and subsequent reactor trip. Two RCCA Ejection cases are considered. The first case assumes that 100% of the activity released from the damaged fuel is instantaneously and homogeneously mixed throughout the containment atmosphere. The second case assumes that 100% of the activity released from the damaged fuel is completely dissolved in the primary coolant and is available for release to the secondary system. This event is described in the UFSAR, Section 14.2.6.

r i NUMERICAL APPLICATIONS, INC.

N ANlO ENGINENNERING E AST Licensing Technical Report for Turkey Point Units 3 and 4 NAI-1396-045, Rev. 1 Page 37 of 81 Page 37 of 2.6.2 Compliance with Reg. Guide 1.183 Regulatory Positions The RCCA Ejection dose consequence analysis is consistent with the guidance provided in Reg. Guide 1.183 Appendix H, "Assumptions for Evaluating the Radiological Consequences of a PWR Rod Ejection Accident," as discussed below:

I. Regulatory Position 1 - The total core inventory of the radionuclide groups utilized for determining the source term for this event is based on Reg. Guide 1.183, Regulatory Position 3.1, and is provided in Table 1.7.4-1. The inventory provided in Table 1.7.4-1 is adjusted for the fraction of fuel damaged and a radial peaking factor of 1.65 is applied. The release fractions provided in Reg. Guide 1.183 Table 3 are adjusted to comply with the specific Reg. Guide 1.183 Appendix H release requirements. For both the containment and secondary release cases, the activity available for release from the fuel gap for fuel that experiences DNB is assumed to be 10% of the noble gas and iodine inventory in the DNB fuel. For the containment release case for fuel that experiences fuel centerline melt (FCM), 100% of the noble gas and 25% of the iodine inventory in the melted fuel is assumed to be released to the containment.

For the secondary release case for fuel that experiences FCM, 100% of the noble gas and 50% of the iodine inventory in the melted fuel is assumed to be released to the primary coolant.

2. Regulatory Position 2 - Fuel damage is assumed for this event.

3. Regulatory Position 3 - For the containment release case, 100% of the activity released from the damaged fuel is assumed to mix instantaneously and homogeneously in the containment atmosphere.

For the secondary release case, 100% of the activity released from the damaged fuel is assumed to mix instantaneously and homogeneously in the primary coolant and be available for leakage to the secondary side of the steam generators.

4. Regulatory Position 4 - The chemical form of radioiodine released from the damaged fuel to the containment is assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. Containment sump pH is controlled to 7.0 or higher.

5. Regulatory Position 5 - The chemical form of radioiodine released from the steam generators to the environment is assumed to be 97% elemental iodine, and 3% organic iodide.

6. Regulatory Position 6.1 - For the containment leakage case, natural deposition in the containment is credited. Containment sprays are not credited in the mitigation of this event.

7. Regulatory Position 6.2 - The containment is assumed to leak at the proposed TS maximum allowable rate of 0.20% for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and 0.10% for the remainder of the event.

8. Regulatory Position 7.1 - The primary-to-secondary leak rate is equal to the value specified by proposed TS 6.8.4.j.b.2, which is 0.6 gpm through all steam generators and 0.2 gpm through any one steam generator at room temperature conditions.

9. Regulatory Position 7.2 -. The density used in converting primary-to-secondary volumetric leak rates to mass leak rates is 62.4 Ibm/ft3, which is consistent with the leakage limits at room temperature conditions.

10. Regulatory Position 7.3 - All of the noble gas released to the secondary side is assumed to be released directly to the environment without reduction or mitigation.

11. Regulatory Position 7.4 - Regulatory Position 5.6 refers to Appendix E, Regulatory Positions 5,5 and 5.6. The iodine transport model for release from the steam generators is as follows:

Appendix E, Regulatory Position 5.5.1 - A portion of the primary-to-secondary leakage is assumed to flash to vapor based on the thermodynamic conditions in the reactor and secondary

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 38 of 81 immediately following plant trip when tube uncovery is postulated. The flashed leakage is assumed to be released to the environment with no mitigation. The primary-to-secondary leakage is assumed to mix with the secondary water without flashing during periods of total tube submergence.

Appendix E, Regulatory Position 5.5.2 - The portion of leakage that immediately flashes to vapor is assumed to rise through the bulk water of the steam generator into the steam space and is assumed to be immediately released to the environment with no mitigation; i.e., no reduction for scrubbing within the bulk water is credited.

Appendix E, Regulatory Position 5.5.3 - All of the steam generator leakage flow that does not immediately flash is assumed to mix with the bulk water.

Appendix E, Regulatory Position 5.5.4 - The radioactivity within the bulk water is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient of 100 is assumed for the iodine. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the steam generator carryover rate of less than 1%.

Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery in the unaffected steam generators is postulated for up to 30 minutes following a reactor trip. During this period, the fraction of primary-to-secondary leakage which flashes to vapor is assumed to rise through the bulk water of the steam generator into the steam space and is assumed to be immediately released to the environment with no mitigation. The flashing fraction is based on the thermodynamic conditions in the reactor and secondary coolant. The leakage which does not flash is assumed to mix with the bulk water in the steam generator.

2.6.3 Other Assumptions

1. This analysis assumed that the equilibrium specific activity on the secondary side of the steam generators is equal to the TS 3.7.1.4 limit of 0.1 [tCi/gm Dose Equivalent 1-131.

2. Radionuclide concentrations in the secondary side fluid of the steam generators assume that Auxiliary Feedwater is provided to maintain a constant secondary mass during periods of steam release.

3. The steam mass release rates for the secondary release are provided in Table 2.6-2.

4. It is assumed that 0.25% of the fuel is assumed to experience melting and 10% of the fuel is breached due to DNB.

2.6.4 Methodology Input assumptions used in the dose consequence analysis of the RCCA Ejection are provided in Table 2.6-1.

The postulated accident consists of two cases. One case assumes that 100% of the activity released from the damaged fuel is instantaneously and homogeneously mixed throughout the containment atmosphere, and the second case assumes that 100% of the activity released from the damaged fuel is completely dissolved in the primary coolant and is available for release to the secondary system. The RCCA Ejection is evaluated with the assumption that 0.25% of the fuel experiences FCM and 10% of the fuel experiences DNB. The release fractions from the damaged fuel correspond to the requirements set out in Regulatory-Position I of Appendix H to Reg. Guide 1.183.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 39 of 81 For the containment release case, 100% of the activity is released instantaneously to the containment.

Natural deposition of the released activity inside of containment is credited. Radionuclide removal by the Emergency Containment Filters and containment spray is not credited. The containment is assumed to leak at the proposed TS maximum allowable rate of 0.20% for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and 0.10% for the remainder of the event.

For the secondary release case, primary coolant activity is released into the steam generators by leakage across the steam generator tubes. The core source term used as the basis for this activity is presented in Table 1.7.4-1. Core activities are then increased by the radial peaking factor of 1.65 for this event. All noble gases associated with this leakage are assumed to be released directly to the environment. Secondary activity is then released to the atmosphere via steaming from the MSSVs/ADVs until the RHR system is capable of removing decay heat and for providing for any further cooldown. The release of the initial iodine content of the steam generator secondary is also considered. The source term for this activity is presented in Table 1.7.3-1.

For this event, the Control Room ventilation system cycles through two modes of operation:

" Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 1000 cfm of unfiltered fresh air.

" For the secondary release, the Control Room is isolated on a high radiation reading at the normal intake monitors. A 60 second delay is applied to account for the time to reach the setpoint, signal processing, and damper closure time. For the containment release, Control Room isolation occurs on high radiation on the containment radiation monitors. The 60 second delay time is conservatively applied to this release model. After isolation, the air flow distribution consists of; 525 cfm of filtered makeup flow from the outside, 115 cfm of unfiltered inleakage, and 375 cfm of filtered recirculation flow.

" The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 97.5% for particulate, elemental iodine, and organic iodine.

The atmospheric dispersion factors (X/Qs) used for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These X/Qs are summarized in Tables 1.8.1-2 and 1.8.1-3. Releases from the steam generators are assumed to occur from the MSSV/ADV which produces the most limiting X/Q. Atmospheric dispersion factors for the containment release correspond to the nearest containment penetration.

The EAB and LPZ doses are determined using the X/Q factors for the appropriate time intervals. These XIQ factors are provided in Table 1.8.2-1.

The secondary release scenario credits control room isolation from a high radiation signal on the control room intake monitor. The Technical Specification setpoint for this instrument is 2 mR/hr. In the RCCA Ejection analysis, an analytical setpoint of 5 mR/hr was used to account for measurement and test uncertainties and to apply additional conservatism. For the design basis fuel failure and core melt fractions, the calculated exposure rate at the detector exceeded the analytical setpoint by approximately 35%. It was recognized that with only 35% margin, a scenario could be postulated with fuel failure fractions less than the design values in which the analytical setpoint would not be reached and a delayed manual isolation must be assumed. While the offsite dose consequences would be lower in such a scenario, the relative impact of lower fuel failure fractions with a longer control room isolation time was not immediately obvious. Therefore, an additional case was performed which combined the reduced source term with a 30-minute control room isolation time.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 40 of 81 2.6.5 Radiological Consequences The radiological consequences of the RCCA Ejection are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. As shown in Table 2.6-3, the results of both cases for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

2.7 Waste Gas Decay Tank (WGDT) Rupture 2.7.1 Background This event involves a major rupture of one of the Waste Gas Decay Tanks as currently presented in Section 14.2.3 of the Turkey Point UFSAR. This analysis assumes that the ruptured WGDT contains an inventory equivalent to the equilibrium RCS noble gas activity from operation with 1% -fuel defects. The leak rate from the WGDT to the environment simulates a major tank rupture which instantaneously releases the entire contents of the tank. No credit is taken for hold-up, dilution, or filtration in the Reactor Auxiliary Building.

2.7.2 Compliance with Reg. Guide 1.183 Regulatory Positions Reg. Guide 1.183 does not provide direct guidance relative to the Waste Gas Decay Tank Rupture event.

Guidelines for the review of WGDT analyses are given in Branch Technical Position 11-5 of the Standard Review Plan, with additional instruction available from Regulatory Issue Summery 2006-04. Compliance of the WGDT rupture analysis with the positions of BTP 11-5 is discussed below.

1. Position B.I .A - The analysis demonstrates that the exposure to an individual at the exclusion area boundary following a WGDT rupture does not exceed 0.1 rem TEDE. This limit is consistent with the guidance provided in Issue #11 of RIS 2006-04.

2. Position B.1 .B - The analysis source term is equal to the RCS noble gas inventory resulting from extended full power operation with 1 % defective fuel, and the entire RCS inventory is transferred to the WGDT following a shutdown.

Exception: BTP 11-5 identifies the PWR-GALE code as an acceptable method for calculating the WGDT source term. However, the source term applied in the Turkey Point WGDT analysis is derived from ORIGEN 2.1 per the guidance of Section 3.1 of Reg. Guide 1.183. The WGDT source term is consistent with the RCS source term listed in Table 1.7.2-1 used in other applicable AST event analyses.

3. Position B.I.C - This analysis assumes an instantaneous failure of a batch-type waste gas system and a release through a pathway not normally used for planned releases. The release is from the building which houses the WGDT, and no effluent monitoring or isolation is assumed. The source term consists of only noble gases since particulates and iodines have been removed by other processes. The release is modeled as a ground level release without credit for a building wake factor. The EAB and LPZ X/Qs represent the maximum values from all downwind sectors and are conservative with respect to 5%

overall site values. No deposition is assumed to occur downwind of the release.

Exception: BTP 11-5 identifies an acceptable method for determining X/Qs outlined in SRP Section 2.3.4. However, the X/Qs used in the WGDT analysis were developed using the PAVAN code in accordance with Reg. Guide 1.145 per Section 5.3 of Reg. Guide 1.183. The release point and corresponding atmospheric dispersion factors are the same as those applied to releases from the Auxiliary Building in the analysis of the LOCA event in discussed in Section 2.1.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 41 of 81 2.7.3 Other Assumptions

1. The WGDT source term provided in Table 2.7-2 is based upon the plant operating at a power level of 2652 MWth with one percent failed fuel for an extended period of time sufficient to achieve equilibrium radioactive concentrations in the reactor coolant system. The entire noble gas inventory of the reactor coolant system is then assumed to be stripped and placed into a single WGDT. This inventory was calculated to be equal to 84,274.8 Curies Dose Equivalent Xe-1 33, which exceeds the Tech Spec LCO 3.7.9 limit of 70,000 Curies.

2. Since the control room recirculation filters do not remove noble gas isotopes, the sensitivity to control room isolation is primarily dependent upon the magnitude of the total flow rate into the control room and the relative size of the atmospheric dispersion factors. Since both of these parameters are larger when the control room ventilation system is in the normal alignment, it is assumed that the control room remains unisolated for the duration of this event.

2.7.4 Methodology The input assumptions used in the dose consequence analysis of the WGDT are provided in Table 2.7-1.

The dose assessment model releases the above-prescribed inventory from the tank at a high rate of release to simulate the tank rupture. The contents are released to the environment without any hold up, dilution or filtration.

The release-receptor point locations are chosen to minimize the distance from the release point to the Control Room intake. The WGDT tank area is served by the Auxiliary Ventilation Exhaust Fans.

However, since the exhaust fans are not required to be operable by the Technical Specifications, and since the Control Room is assumed to remain in the normal operating mode, the limiting X!Qs correspond a release from Auxiliary Building vent V-10 to the normal intake as shown in Tables 1.8.1-2 and 1.8.1-3.

The EAB and LPZ doses are determined using the XIQ factors for the appropriate time intervals. These X/Q factors are provided in Table 1.8.2-1.

2.7.5 Radiological Consequences Reg. Guide 1.183 does not provide specific requirements or dose limits for a WGDT failure. The TEDE limit of 0.1 rem at the EAB is established by Regulatory Issue Summary 2006-04. Since this limit is referenced to dose restrictions to the member of the public during normal operation from 10CFR Part 20, EAB dose will be evaluated over a 30-day period. Although no limit is given for the LPZ boundary, this location will also be evaluated against the EAB limit of 0.1 rem acceptance criteria. A TEDE limit 5.0 from 10CFR50.67 will be applied to the control room.

Area Dose Criteria EAB 0.1 rem TEDE (for 30 days)

LPZ 0.1 rem TEDE (for 30 days)

Control Room 5 rem TEDE (for 30 days)

As shown in Table 2.7-3, the radiological consequences of the Waste Gas Decay Tank Rupture are all within the appropriate acceptance criteria.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 42 of 81 2.8 Spent Fuel Cask Drop 2.8.1 Background This event considers the drop of a spent fuel transfer cask into the spent fuel pool as described in Section 14.2.1.3 of the Turkey Point UFSAR. Reg. Guide 1.183 does not provide any specific guidance for the Spent Fuel Cask Drop event; therefore, the requirements of the Fuel Handling Accident in Appendix B of the Reg. Guide are followed for the cask drop reanalysis.

Per Tech. Spec. 3.9.12, 1525 hours0.0177 days 0.424 hours 0.00252 weeks 5.802625e-4 months of decay is required before movement of the spent fuel cask is allowed.

The analysis assumes that all 157 assemblies of a recently discharged core are assumed to be damaged by the cask drop. Thus, the core source term presented in Table 1.7.4-1 is used in the analysis. High burnup gap release fractions are also applied. This source term is allowed to decay for 1525 hours0.0177 days 0.424 hours 0.00252 weeks 5.802625e-4 months prior to release.

A minimum water level of 23 feet is maintained above the damaged fuel. This water level ensures an elemental iodine decontamination factor of 285 per the guidance provided in NRC Regulatory Issue Summary 2006-04.

2.8.2 Compliance with RG 1.183 Regulatory Positions Regulatory Guide 1.183 does not provide specific guidance for the Spent Fuel Cask Drop event: However, the guidance for the Fuel Handling Accident in Appendix B of the Reg. Guide is judged to be closely applicable to the conditions of this event. Therefore the following discussion for refers to the Reg. guide 1.183 positions as stated in Appendix B for the Fuel Handling Accident:

1. Regulatory Position 1.1 - The amount of fuel damage is assumed to be equal to 157 assemblies.

2. Regulatory Position 1.2 - The fission product release from the breached fuel is based on Regulatory Positions 3.1 and 3.2 of Reg. Guide 1.183. The gap activity available for release is modified from that specified by Table 3 of Reg. Guide 1.183 to account for high bumup fuel as described in Section 1.7.6.

These modified gap releases are applied to the core source term provided in Table 1.7.4-1. This activity is assumed to be released from the fuel assembly instantaneously.

3. Regulatory Position 1.3 - The chemical form of radioiodine released from the damaged fuel into the spent fuel pool is assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. The cesium iodide is assumed to completely dissociate in the spent fuel pool resulting in a final iodine distribution of 99.85% elemental iodine and 0.15% organic iodine.

4. Regulatory Position 2 - A minimum water depth of 23 feet is maintained above the damaged fuel assembly. Therefore, a decontamination factor of 285 is applied to the elemental iodine and a decontamination factor of 1 is applied to the organic iodine. As a result, the breakdown of the iodine species above the surface of the water is 57% elemental and 43% organic. Guidance for the use of 285 for the elemental iodine decontamination factor is provided in NRC Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternate Source Terms."

5. Regulatory Position 3 - All of the noble gas released is assumed to exit the pool without mitigation.

All of the non-iodine particulate nuclides are assumed to be retained by the pool water.

6. Regulatory Position 4.1 - The radioactive material released from the fuel pool is assumed to be released from the building to the environment over a 2-hour period.

7. Regulatory Position 4.2 - No credit is taken for filtration of the release.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS. INC. Turkey Point Units 3 and 4 Page 43 of 81

8. Regulatory Position 4.3 - No credit is taken for dilution of the release.

9. Regulatory Position 5 - The event does not occur in the containment.

2.8.3 Other Assumptions The dose acceptance criteria for the Spent Fuel Cask Drop are assumed to be the same as those for the Fuel Handling Accident.

2.8.4 Methodology The input assumptions used in the dose consequence analysis of the Spent Fuel Cask Drop are provided in Table 2.8-1. 100% of the gap activity specified in Table 3 of RG 1.183 is assumed to be instantaneously released from 157 assemblies into the fuel pool and transported to the environment over 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . A minimum water level of 23 feet is maintained above the damaged fuel for the duration of the event. All of the noble gas released from the damaged fuel is assumed to escape from the pool. All of the non-iodine particulates released from the damaged fuel are assumed to be retained by the pool. The iodine released from the damaged fuel is assumed to be composed of 99.85% elemental and 0.15% organic.

The source term meets the requirements of Regulatory Position 1 of Appendix B to RG 1.183. Since a full core of 157 assemblies is assumed to be damaged by the cask drop, the source term for the Spent Fuel Cask Drop event is equal to the core source term listed in Table 1.7.4-I which is decayed by 1525 hours0.0177 days 0.424 hours 0.00252 weeks 5.802625e-4 months .

For this event, the Control Room ventilation system cycles through two modes of operation:

Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 1000 cfm of unfiltered fresh air.

. The Control Room is assumed to be manually isolated by the operators 30 minutes after the beginning of the event. After isolation, the air flow distribution consists of 525 cfm of filtered makeup flow from the outside, 115 cfm of unfiltered inleakage, and 375 cfm of filtered recirculation flow.

The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 97.5% for particulate, elemental iodine, and organic iodine.

The atmospheric dispersion factors (X/Qs) used for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These XIQs are summarized in Tables 1.8.1-2 and 1.8.1-3. The atmospheric dispersion factors applied to the Spent Fuel Cask Drop event are based upon a release from the Unit 4 spent fuel pool.

The EAB and LPZ doses are determined using the X/Q factors for the appropriate time intervals. These X/Q factors are provided in Table 1.8.2-1.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 44 of 81 2.8.5 Radiological Consequences The radiological consequences of the Spent Fuel Cask Drop are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. As shown in Table 2.8-2, the results for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

2.9 Other Dose Consequences The Turkey Point UFSAR, Appendix 8A, discusses equipment EQ due to a radiation environment. Reg.

Guide 1.183, Regulatory Position 6, allows the licensee to use either the AST or TID-14844 assumptions for performing the required EQ analyses until such time as a generic issue related to the effect of increased cesium releases on EQ doses is resolved. Analysis performed by Sandia National Labs in Reference 5.44 found that for equipment exposed to sump water, the integrated doses calculated with AST exceeded those calculated with TID-14844 only after 42 days for a typical PWR. These findings were cited in the closure of Generic Safety Issue 187 with no requirements to implement AST. Therefore, the Turkey Point EQ analyses will continue to be based on TID-14844 assumptions.

Positions 1.3 and 4.3 of Reg. Guide 1.183 identify various licensing requirements for which compliance may have been demonstrated, in part, by the evaluation of radiological consequences of design basis accidents. The Turkey Point responses to these licensing commitments are not being revised as a result of implementation of AST.

3.0 Summary of Results Results of the Turkey Point radiological consequence analyses using the AST methodology and the corresponding allowable control room unfiltered inleakage are summarized on Table 3-1.

4.0 Conclusions Full implementation of the Alternative Source Term methodology, as defined in Regulatory Guide 1.183, into the design basis accident analysis has been made to support control room habitability in the event of increases in control room unfiltered air inleakage. Analysis of the dose consequences of the Loss-of-Coolant Accident (LOCA), Fuel Handling Accident (FHA), Main Steam Line Break (MSLB), Steam Generator Tube Rupture (SGTR), Locked Rotor, Rod Cluster Control Assembly (RCCA) Ejection, Waste Gas Decay Tank (WGDT) Rupture, and Spent Fuel Cask Drop have been made using the Reg. Guide 1.183 methodology. The analyses used assumptions consistent with proposed changes in the Turkey Point licensing basis and the calculated doses do not exceed the defined acceptance criteria.

This report supports a maximum allowable control room unfiltered air inleakage of 115 cfm.

5.0 References 5.1 Turkey Point Units 3 and 4 Updated FSAR, Electronic Version, reviewed on 4/24/09.

5.2 TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 23, 1962.

5.3 USNRC, Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants," July 2000.

5.4 Code of Federal Regulations, 10CFR50.67, "Accident Source Term," revised 12/03/02.

5.5 NEI 99-03, "Control Room Habitability Guidance," Nuclear Energy Institute, Revision 0 dated June 2001 and Revision I dated March 2003.

5.6 NRC Generic Letter 2003-01, "Control Room Habitability," June 12, 2003.

5.7 Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

5.8 Federal Guidance Report No. 12 (FGR 12), "External Exposure to Radionuclides in Air, Water, and Soil," 1993.

5.9 Florida Power & Light Company Turkey Point Plant Units 3 and 4 Technical Specifications through Amendments 239.

5.10 Letter from G. Van Middlesworth to USNRC, "Duane Arnold Energy Center, Docket No: 50-331, Op. License No: DPR-49, Technical Specification Change Request (TSCR-037): "Alternative Source Term"," October 19, 2001.

5.11 Duane Arnold Issuance of Amendment (IA) and Safety Evaluation (SE) for Amendment No. 240 to DPR-49 issued July 31, 2001.

5.12 ARCON96 Computer Code ("Atmospheric Relative Concentrations in Building Wakes,"

NUREG/CR-6331, Rev. 1, May 1997, RSICC Computer Code Collection No. CCC-664 and July 1997 errata).

5.13 MicroShield Version 5 "User's Manual "and "Verification & Validation Report, Rev. 5," Grove Engineering, both dated October 1996.

5.14 Oak Ridge National Laboratory, CCC-371, "RSICC Computer Code Collection - ORIGEN 2. 1,"

May 1999.

5.15 "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Bases Accident Releases of Radioactive Material from Nuclear Power Stations," NUREG/CR-2858, November 1982, (RSICC Computer Code Collection No. CCC-445).

5.16 NAI 8907-02, Revision 17, GOTHIC Containment Analysis Package User Manual, Version 7.2a(QA), January 2006 5.17 NAI 8907-06, Revision 16, GOTHIC Containment Analysis Package Technical Manual, Version 7.2a(QA), January 2006

r NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 46 of 81

LTNN ENQfl*-RING Page 46 of 81 5.18 NAI 8907-09, Revision 9, GOTHIC Containment Analysis Package Qualification Report, Version 7.2a(QA), January 2006 5.19 Numerical Applications Inc., NAI-9912-04, Revision 2, "RADTRAD-NAI Version l.0p3(QA)

Documentation," July 2002.

5.20 Numerical Applications Inc., "Dose Methodology Quality Assurance Procedures," Revision 1, June 4, 2001.

5.21 NRC Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," June 2003.

5.22 USNRC, Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Rev. 1, February 1983.

5.23 USNRC, Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.

5.24 NUREG-0800, USNRC, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,".

5.25 Industry Degraded Core Rulemaking Program Technical Report 11.3, "Fission Product Transport in Degraded Core Accidents," Atomic Industrial Forum, December 1983.

5.26 NUREG/CR-4102, "Air Currents Driven by Sprays in Reactor Containment Buildings," May 1986.

5.27 Fort Calhoun Issuance of Amendment (IA) and Safety Evaluation (SE) for Amendment No. 198 to DPR-40 issued April 4, 2001.

5.28 USNRC, Regulatory Guide 1.52, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants," Rev. 3, June 2001.

5.29 NRC Information Notice 91-56, "Potential Radioactive Leakage to Tank Vented to Atmosphere,"

September 19, 1991.

5.30 NUREG/CR-5950, "Iodine Evolution and pH Control," December 1992.

5.31 USNRC, Regulatory Issue Summary 2006-04, Experience with Implementation of Alternate Source Terms, March 7, 2006.

5.32 ANSI/ANS- 18.l-1999, "Radioactive Source Term for Normal Operation of Light Water Reactors," approved September 21, 1999; including Errata dated December 1, 2005.

5.33 WCAP-16596-NP, "Evaluation of Alternative Emergency Core Cooling System Buffering Agents," Revision 0.

5.34 NUREG/CR-5009, Assessment of the Use of Extended Burnup Fuel in Light Water Reactors.

5.35 USNRC, Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors", March 1972.

5.36 Entergy Nuclear Northeast, Letter IPN-02-044,

Subject:

Proposed Changes to Technical Specifications: Selective Adoption of Alternative Source Term and Incorporation of Generic

NUMERICAL AST Licensing Technical Report for NAI-l1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 47 of 81 INE SOCUTIONS ANDSO ERING ae4Af8 Changes; TSTF-51, TSTF-68, and TSTF-312, June 5, 2002. (ADAMS Accession No. ML021840136).

5.37 CP&L, Serial HNP-01-120, "Shearon Harris Nuclear Power Plant Docket No. 50-400/License No.

NPF-63 Revision to Alternative Source Term Methodology Analyses in Support of the Steam Generator Replacement and Power Uprate License Amendment Applications", August 17, 2001.

5.38 NMC Letter NRC-02-024, "Revision to the Design Basis Radiological Analysis Accident Source Term", Docket 50-305, Operating License DPR-43, Kewaunee Nuclear Power Plant, March 19,2002.

5.39 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 166 to Facility Operating License No. DPR-43, Nuclear Management Company LLC, Kewaunee Nuclear Power Plant, Docket No. 50-305. (ADAMS Accession No. ML0302100620).

5.40 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Nos. 182 and 169 to Facility Operating License Nos. NPF-76 and NPF-80, STP Nuclear Operating Company, Et. Al., South Texas Project, Units I and 2, Docket Nos. 50-498 and 50-499. (ADAMS Accession No. ML072680192).

5.41 USNRC SER, Attachment to License Amendment 215, Facility Operating License DPR-64, Indian Point Nuclear Generating Unit 3 - Issuance of Amendment Re: Selective Adoption of Alternate Source Term (TAC No. MB5382), March 17, 2003. (ADAMS Accession No. ML030760135).

5.42 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 226 to Renewed Facility Operating License No. DPR-20, Entergy Nuclear Operations, Inc, Palisades Plant, Docket No. 50-255, September 28 2007. (ADAMS Accession No. ML72470667).

5.43 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 152 to Renewed Facility Operating License No. NPF- 16, Florida Power and Light Company, St. Lucie Plant, Unit No. 2, September 29 2008, Docket No. 50-389. (ADAMS Accession No. ML082060400).

5.44 Evaluation of Radiological Consequences of Design Basis Accidents at Operating Reactors Using The Revised Source Term, Sandia National Laboratories, September 28, 1998.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS. INC. Turkey Point Units 3 and 4 Page 48 of 81 Figure 1.8.1-1 Onsite Release-Receptor Location Sketch (Not to scale)

A - Unit 3 Main Steam Line Closest Penetration Q - Unit 4 Spent Fuel Building B - Unit 4 Main Steam Line Closest Penetration R - Unit 3 Equipment Hatch C - Unit 3 Purge Duct Outlet S - Unit 4 Equipment Hatch D - Unit 4 Purge Duct Outlet T - Unit 3 Personnel Hatch E - Plant Stack U - Unit 4 Personnel Hatch F- Unit 3 RWST V - Unit 3 Emergency Escape Lock G - Unit 4 RWST W - Unit 4 Emergency Escape Lock H - Unit 3 Closest MSSV X - Unit 3 SJAE I - Unit 4 Closest MSSV Y - Unit 4 SJAE J - Unit 3 ADV Silencer Z - Unit 3 Electrical Penetration K - Unit 4 ADV Silencer AA - Unit 4 Electrical Penetration L - Unit 3 Main Stm Line Closest Point (Normal Intake) BB - Aux. Bldg Vent V-10 M - Unit 3 Main Stm Line Closest Point (NE & SE Intakes) * - Normal Control Room Intake N - Unit 4 Main Stm Line Closest Point (Normal Intake)

NE - Northeast Emergency CR Intake O - Unit 4 Main Stm Line Closest Point (NE & SE Intakes)

SE - Southeast Emergency CR Intake P - Unit 3 Spent Fuel Building

Table 1.6.3-1 Control Room Ventilation System Parameters Parameter Value Control Room Volume 47,786 ft3 Normal Operation Filtered Make-up Flow Rate 0 cfm Filtered Recirculation Flow Rate 0 cfm Unfiltered Make-up Flow Rate and Inleakage 1000 cfm Emergency Operation Recirculation Mode:

Filtered Make-up Flow Rate 525 cfm Filtered Recirculation Flow Rate 375 cfm Unfiltered Make-up Flow Rate 0 cfm Unfiltered Inleakage 115 cfm Filter Efficiencies Elemental 97.5%

Organic 97.5%

Particulate 97.5%

Table 1.6.3-2 LOCA Direct Shine Dose Source Direct Shine Dose (rem)

Containment 0.059 Containment Purge Duct 0.333 CR Filters 0.270 External Cloud 0.061 Total 0.723

Table 1.7.2-1 Primary Coolant Source Term*

RCS Activity RCS Activity Nuclide (j+/-Ci/g) Nuclide (RiCi/g)

Co-58 3.507E-01 Xe-135 4.866E+00 Co-60 7.246E-02 Cs-134 1.023E+01 Kr-85 3.340E+01 Cs-136 5.439E+00 Kr-85m 1.138E+00 Cs-137 4.826E+00 Kr-87 6.856E-01 Ba-139 4.135E-04 Kr-88 2.024E+00 Ba-140 4.533E-03 Rb-86 8.063E-02 La- 140 6.415E-03 Sr-89 3.216E-03 La-141 2.711E-04 Sr-90 2.956E-04 La-142 6.996E-05 Sr-91 1.237E-03 Ce-141 2.761E-02 Sr-92 5.302E-04 Ce-143 1.423E-03 Y-90 4.83 1E-04 Ce-144 6.536E-02 Y-91 3.077E-02 Pr-143 1.344E-02 Y-92 6.489E-04 Nd-147 4.428E-03 Y-93 4.249E-04 Kr-83m 2.931E-01 Zr-95 4.348E-02 Xe-131m 2.729E+00 Zr-97 7.967E-04 Xe-133m 3.138E+00 Nb-95 6.103E-02 Cs-138 6.838E-01 Mo-99 4.395E+00 Cs- 134m 5.399E-02 Tc-99m 4.200E+00 Sb-124 4.464E-02 Ru-103 3.387E-02 Sb-125 7.716E-0I Ru- 105 1.748E-04 Sb-126 8.069E-03 Ru-106 4.187E-02 Te-134 2.072E-02 Rh-105 1.413E-03 Te- 125m 1.783E-01 Sb-127 2.078E-01 Te-133m 1.216E-02 Sb-129. 2.819E-02 Rh-103m 3.379E-02 Te- 127 5.701E-01 Nb-97 9.994E-05 Te-127m 3.888E-01 Nb-95m 3.216E-04 Te-129 3.988E-01 Pm-147 8.795E-03 Te-129m 5.687E-01 Pm-148 1.098E-03 Te-131m 8.316E-02 Pm-149 I.001E-03 Te-132 1.979E+00 Pm-151 1.844E-04 1-131 2.006E-0 1 Pm-148m 7.099E-04 1-132 1.422E-01 Y-91m 7.204E-04 1-133 2.431E-01 Br-82 2.147E-02 1-134 2.633E-02 Br-83 5.948E-02

Table 1.7.2-1 Primary Coolant Source Term*

RCS Activity RCS Activity Nuclide (RiCi/g) Nuclide (AtCi/g) 1-135 1.200E-01 Br-84 2.554E-02 Xe-133 2.260E+02

The iodine activities have been adjusted to equal 0.25 gCi/gm DE 1-131, and the non-iodine activities correspond to 447.7 VCi/gm DE Xe-133 based upon the proposed Technical Specification definition.

Table 1.7.3-1 Secondary Side Source Term Nuclide Activity (gCi/g) 1-131 8.022E-02 1-132 5.688E-02 1-133 9.725E-02 1-134 1.053E-02 1-135 4.801E-02 Table 1.7.4-1 Core Source Term Containment Containment Nuclide Leakage Source Nuclide Leakage Source (Curies) (Curies)

Co-58 O.OOE+00 Pu-239 3.021E+04 Co-60 0.00E+00 Pu-240 5.103E+04 Kr-85 9.615E+05 Pu-241 1.287E+07 Kr-85m 1.813E+07 Am-241 1.403E+04 Kr-87 3.468E+07 Cm-242 5.267E+06 Kr-88 4.878E+07 Cm-244 8.359E+05 Rb-86 2.251E+05 1-130 4.475E+06 Sr-89 6.702E+07 Kr-83m 8.629E+06 Sr-90 7.655E+06 Xe-138 1.1 70E+08 Sr-91 8.252E+07 Xe-131m 8.291E+05 Sr-92 8.958E+07 Xe-133m 4.595E+06 Y-90 7.990E+06 Xe-135m 2.970E+07

Table 1.7.4-1 Core Source Term Containment Containment Nuclide Leakage Source Nuclide Leakage Source (Curies) (Curies)

Y-91 8.684E+07 Cs-138 1.299E+08 Y-92 8.999E+07 Cs-134m 5.657E+06 Y-93 1.043E+08 Rb-88 4.963E+07 Zr-95 1.192E+08 Rb-89 6.351E+07 Zr-97 1.173E+08 Sb-124 2.030E+05 Nb-95 1.205E+08 Sb- 125 1.524E+06 Mo-99 1.355E+08 Sb-126 1.217E+05 Tc-99m 1.186E+08 Te- 131 6.530E+07 Ru- 103 1.270E+08 Te-133 8.486E+07 Ru- 105 9.731E+07 Te-134 1.170E+08 Ru- 106 5.914E+07 Te-125m 3.302E+05 Rh-105 8.808E+07 Te- 133m 5.184E+07 Sb-127 9.265E+06 Ba-141 1.152E+08 Sb- 129 2.58 1E+07 Ba-137m 9.693E+06 Te- 127 9.194E+06 Pd-109 3.639E+07 Te-127m 1.237E+06 Rh-106 6.552E+07 Te- 129 2.540E+07 Rh-103m 1.144E+08 Te-129m 3.785E+06 Tc-101 1.247E+08 Te-131m 1.11OE+07 Eu-154 1.284E+06 Te-132 1.036E+08 Eu-155 8.731E+05 1-131 7.41 OE+07 Eu- 156 2.598E+07 1-132 1.056E+08 La- 143 1.066E+08 1-133 1.433E+08 Nb-97 1.183E+08 1-134 1.563E+08 Nb-95m 8.547E+05 1-135 1.346E+08 Pm- 147 1.044E+07 Xe-133 1.437E+08 Pm-148 2.165E+07 Xe-135 3.779E+07 Pm-149 4.768E+07 Cs-134 2.184E+07 Pm-151 1.623E+07 Cs- 136 6.015E+06 Pm- 148m 2.573E+06 Cs-137 1:023E+07 Pr-144 9.855E+07 Ba-139 1.270E+08 Pr-144m 1.175E+06 Ba-140 1.231 E+08 Sm-153 4.971E+07 La-140 1.278E+08 Y-94 1.053E+08 La-141 1.158E+08 Y-95 1.134E+08 La-142 1.118E+08 Y-91m 4.790E+07

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 53 of 81 SOýUTiONS ANDS INENGINCRING MOFTARE Table 1.7.4-1 Core Source Term Containment Containment Nuclide Leakage Source Nuclide Leakage Source (Curies) (Curies)

Ce-141 1.179E+08 Br-82 6.087E+05 Ce-143 1.073E+08 Br-83 8.605E+06 Ce-144 9.786E+07 Br-84 1.487E+07 Pr-143 1.064E+08 Am-242 8.269E+06 Nd-147 4.650E+07 Np-238 5.195E+07 Np-239 1.897E+09 Pu-243 5.861E+07 Pu-238 4.201E+05 Table 1.7.5-1 Fuel Handling Accident Source Term Nuclide Curies Nuclide Curies Co-58 O.OOOE+00 Pu-239 3.175E+02 Co-60 O.OOOE+00 Pu-240 5.363E+02 Kr-85 1.010E+04 Pu-241 1.353E+05 Kr-85m 1.906E+05 Am-241 1.474E+02 Kr-87 3.645E+05 Cm-242 5.536E+04 Kr-88 5.127E+05 Cm-244 8.785E+03 Rb-86 2.366E+03 1-130 4.703E+04 Sr-89 7.044E+05 Kr-83m 9.068E+04 Sr-90 8.045E+04 Xe-138 1.230E+06 Sr-91 8.672E+05 Xe- 131 m 8.714E+03 Sr-92 9.415E+05 Xe-133m 4.830E+04 Y-90 8.397E+04 Xe-135m 3.122E+05 Y-91 9.126E+05 Cs-138 1.365E+06 Y-92 9.458E+05 Cs- 134m 5.945E+04 Y-93 1.096E+06 Rb-88 5.216E+05 Zr-95 1.253E+06 Rb-89 6.674E+05 Zr-97 1.233E+06 Sb-124 2.133E+03 Nb-95 1.267E+06 Sb-125 1.602E+04 Mo-99 1.424E+06 Sb-126 1.279E+03 Tc-99m 1.247E+06 Te-131 6.862E+05 Ru-103 1.335E+06 Te-133 8.918E+05 Ru-105 1.023E+06 Te-134 1.229E+06

NUMERICAL AST Licensing Technical Report for NAI- 1396-045, Rev. 1 APPLICATIONS. INC. Turkey Point Units 3 and 4 Page 54 of 81 Table 1.7.5-1 Fuel Handling Accident Source Term Nuclide Curies Nuclide Curies Ru-106 6.216E+05 Te-125m 3.470E+03 Rh-105 9.257E+05 Te-133m 5.448E+05 Sb-127 9.737E+04 Ba-141 1.211E+06 Sb-129 2.713E+05 Ba-137m 1.019E+05 Te-127 9.662E+04 Pd- 109 3.825E+05 Te-127m 1.300E+04 Rh-106 6.885E+05 Te-129 2.670E+05 Rh-103m 1.202E+06 Te-129m 3.978E+04 Tc-101 1.311 E+06 Te-131m 1.167E+05 Eu-154 1.349E+04 Te-132 1.088E+06 Eu-155 9.176E+03 1-131 7.788E+05 Eu-156 2.73 1E+05 1-132 1.11OE+06 La-143 1.120E+06 1-133 1.506E+06 Nb-97 1.244E+06 1-134 1.642E+06 Nb-95m 8.983E+03 1-135 1.414E+06 Pm-147 1.097E+05 Xe-133 1.51 OE+06 Pm-148 2.275E+05 Xe-135 3.972E+05 Pm-149 5.011E+05 Cs-134 2.295E+05 Pm-151 1.706E+05 Cs-136 6.321E+04 Pm-148m 2.704E+04 Cs-137 1.075E+05 Pr- 144 1.036E+06 Ba-139 1.335E+06 Pr-144m 1.235E+04 Ba-140 1.293E+06 Sm-153 5.224E+05 La- 140 1.344E+06 Y-94 1.107E+06 La-141 1.217E+06 Y-95 1.192E+06 La-142 1.175E+06 Y-91m 5.034E+05 Ce-141 1.239E+06 Br-82 6.397E+03 Ce-143 1.128E+06 Br-83 9.044E+04 Ce-144 1.028E+06 Br-84 1.563E+05 Pr-143 1.1 18E+06 Am-242 8.691E+04 Nd-147 4.887E+05 Np-238 5.460E+05 Np-239 1.993E+07 Pu-243 6.159E+05 Pu-238 4.415E+03

.5. 1 1

Table 1.8.1-1 Release-Receptor Combination Parameters for Analysis Events Release- Receptor Release Receptor Distance Direction Building Receptor Release Location Location Height Height (in) (deg) Area I Pair (in(m)in) (inm)

A Plant stack Normal 55.5 4.3 46.3 95 0.01 B Plant stack SE emergency 55.5 1.2 85.4 323 0.01 C Unit 4 RWST Normal 15.2 4.3 92.9 97 0.01 D Unit 4 RWST SE emergency 15.2 1.2 60.3 354 0.01 E Unit 4 Closest MSSV Normal 18.6 4.3 17.0 158 0.01 F Unit 4 Closest MSSV SE emergency 18.4 1.2 98.4 291 0.01 Unit 4 Main Steam G LnetosPint Line Closest Point Normal

__________________ 11.2 4.3 18.5 157 0.01 Unit 4 Main Steam H Line Closest Point SE emergency 11.2 1.2 92.7 294 1254 LinetClssPerointnel_________

I Unit4Personnel Normal 3.3 4.3 23.1 148 1254 Hatch I Unit 4 Emergency SE 11.1 1.2 63.8 307 1254 Escape Lock K Unit 4 Spent Fuel Normal 4.3 4.3 57.3 118 0.01 Building (NW comer)

L Unit 4 Spent Fuel L_______ Building (SE corner) SE emergency 1.2 1.2 43.0 319 0.01 M Unit 4 SJAE Normal 7.5 4.3 9.4 331 0.01 Unit 4 Westernmost N U Normal 4.0 4.3 22.7 113 1254 Electrical Penetration O Auxiliary Building Normal 4.9 4.3 52.4 86 0.01 Vent V- 10 1 1 1 1_1

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units3 and 4 Page 56 of 81 SOL-rTIONS ANDS N ENGINEERING ARE Table 1.8.1-2 Onsite Atmospheric Dispersion (X/Q) Factors for Analysis Events This table summarizes the results for XIQ factors for the control room intakes for the various accident scenarios.

Values are presented for the normal air intake prior to intake isolation and the least favorable emergency air intake after Control Room isolation. The same atmospheric dispersion factor is applied to both the makeup flow and unfiltered inleakage for each release-receptor pair. These values are not adjusted for Control Room Occupancy Factors. Note that the letters that indicate the release-receptor pairs do not necessarily correspond with the release identification letters on Figure 1.8.1-1.

Release- Receptor 0-2 hour 2-8 hour 8-24 hour 1-4 days 4-30 days Receptor Release Point Point XIQ XIQ XIQ XIQ XIQ Pair A Plant stack Normal intake 1.85E-03 B Plant stack SE emergency 9.02E-04 7.50E-04 2.76E-04 2.22E-04 1.53E-04 intake C Unit 4 RWST Normal intake 9.72E-04 D (1) Unit 4 RWST SE emergency 1.89E-03 1.49E-03 6.OOE-04 4.67E-04 3.49E-04 intake E Unit 4 Closest MSSV/ADV (2 Normal intake 1.33E-02 (3)

F (1) Unit 4 Closest SE emergency 6.79E-04 (3) 4.44E-04 1.84E-04 1.36E-04 8.47E-05 MSSV/ADV (2) intake Unit 4 Main Steam G U Normal intake 1.53E-02 Line Closest Point H( Unit 4 Main Steam SE emergency 7.32E-04 4.61E-04 1.94E-04 1.45E-04 9.21E-05 Line Closest Point intake I Unit 4 Personnel Hatch Normal intake 1.03E-02 j(1) Unit 4 Emergency SE emergency 1.44E-03 1.03E-03 3.89E-04 3.29E-04 2.27E-04 Escape Lock intake K Unit 4 Spent Fuel Normal intake 2.31E-03 Building (NW corner)

L (I) Unit 4 Spent Fuel SE emergency 3.27E-03 2.66E-03 L.OOE-03 8.15E-04 5.99E-04 Building (SE corner) intake M Unit 4 SJAE Normal intake 6.43E-02 (4)

Unit 4 Westernmost Unt4Netrnot Normal intake 1.1 4E-02 N Electrical Penetration 0 Auxiliary Building Normal intake 2.80E-03 Vent V-10

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 57 of 81 Table 1.8.1-2 Notes:

(1) This receptor location qualifies for the dual intake credit allowed by Section 3.3.2.2 of Reg. Guide 1.194. This credit is not applied to the values shown in this table; however, these values are reduced by a factor of 2 when applied in the event analyses.

(2) The atmospheric dispersion factor corresponding to the limiting MSSV or ADV is used for each time period. No distinction is made between automatic steam relief from the MSSVs and controlled releases from the ADVs for radiological purposes.

(3) This release location meets the requirements for the plume rise credit described in Section 6 of Reg. Guide 1.194. The 0-2 hour values shown in this table are reduced by a factor of 5 when used in the applicable event analyses.

(4) The distance from the Unit 4 SJAE to the normal intake is 9.4 meters as shown in Table 1.8.1-1:

Section 3.4 of Reg. Guide 1.194 that states ARCON96 should not be used to address situations with distances of less than about 10 m. Therefore, the value in this table was derived using a I/r 2 relationship referenced to an ARCON96-calculated value at 20 meters. The 1/r2 approach was demonstrated to calculate conservative atmospheric dispersion factors with respect to values determined directly from ARCON96 at the same distance. For example, the 10-meter X/Q value determined in this manner is 5.68E-02 sec/m 3 compared with the ARCON96 calculated value of 5.02E-02 sec/m3 , a difference of 11.6%. For shorter distances, this approach becomes more conservative. At 9.4 meters, the ARCON96 result is 5.57E-02 sec/m 3 , which is 13.4% less than the 6.43E-02 sec/m 3 value used in the analysis.

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS. INC. Turkey Point Units 3 and 4 Page 58 of 81 Table 1.8.1-3 Release-Receptor Point Pairs Assumed for Analysis Events (1)

Event Prior to CR Isolation During CR Recirculation LOCA:

- Containment Leakage N J

- ECCS Leakage n/a D

- RWST Backleakage n/a D

- Containment Purge A B FHA

- Containment Release I J

- FHB Release K L Spent Fuel Cask Drop K L MSLB:

- Break Release G H

- MSSV/ADV Release E F SGTR M, E (2) F Locked Rotor E F RCCA Ejection:

- Containment Leakage N J

- Secondary Side Release E F WGDT Rupture 0 n/a

( Letters correspond to Release-Receptor pairs listed in Table 1.8.1-2 (2) Prior to reactor trip, the release receptor pair is from the SJAE to the normal intake. The release point changes to the MSSV/ADVs immediately after reactor trip, and the receptor point shifts to the southeast emergency intake following control room isolation.

Table 1.8.2-1 Offsite Atmospheric Dispersion (X/Q) Factors for Analysis Events 3 3 Time Period EAB X/Q (sec/mr ) LPZ X/Q (sec/m )

0-2 hours 1.71E-04* 3.76E-05 0-8 hours 9.84E-05 1.64E-05 8-24 hours 7.47E-05 1.08E-05 1-4 days 4.11E-05 4.38E-06 4-30 days 1.74E-05 1.20E-06

With the exception of the WGDT Rupture, only the 0-2 hour EAB X/Q is used in the event analyses Table 1.9-1 Post-LOCA Sump pH Analysis Results Parameter Minimum Value Maximum Value NaTB Level in Large Baskets (ft) 2.187 2.77 NaTB Level in Small Baskets (ft) 1.917 2.5 Dissolved Mass of NaTB at 4637 4687 Switchover (Ibm)

Total NaTB Mass (lbm) 11061 14264 pH at Switchover 7.000 7.005 Long Term pH 7.241 7.396

NUMERICAL AST Licensing Technical Report for NAI- 1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 60 of 81 SOLUTIONS INENINEERING ANDSOFw E Table 2.1-1 Loss of Coolant Accident (LOCA) - Inputs and Assumptions Input/Assumption Value Release Inputs:

Core Power Level 2652 MW1 ,

Core Average Fuel Burnup 45,000 MWD/MTU Fuel Enrichment 3.0 -5.0 w/o Initial RCS Equilibrium Activity 0.25 gCi/gm DE 1-131 and 447.7 pCi/gm DE Xe-133 (Table 1.7.2-1)

RCS Mass (maximum) 397,544 Ibm Core Fission Product Inventory Table 1.7.4-1 Containment Leakage Rate 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.20% (by weight)/day after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.10% (by weight)/day LOCA release phase timing and duration Table 2.1-2 Core Inventory Release Fractions (gap release and early in- Reg. Guide 1.183, Sections 3.1, 3.2, and Table 2 vessel damage phases)

ECCS Systems Leakage (from 15 minutes to 30 days)

Sump Volume (minimum) 239,000 gallons (31,949.5 ft3)

ECCS Leakage (2 times allowed value) 4,650 cc/hr Flashing Fraction Calculated - 0.092 Used for dose determination - 0.10 Chemical form of the iodine released from the ECCS 97% elemental, 3% organic leakage No filtration or credit for building dilution, released directly to the environment

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS,. INC. Turkey Point Units 3 and 4 Page 61 of 81 Table 2.1-1 Loss of Coolant Accident (LOCA) - Inputs and Assumptions Input/Assumption Value RWST Back-leakage (from 15 minutes to 30 days)

Sump Volume (minimum) 239,000 gallons (31,949.5 ft.3)

ECCS Leakage to RWST (2 times allowed value) 0.1 gph Flashing Fraction 0 % based on temperature of fluid reaching the RWST.

Elemental iodine is released into tank space based upon partition factor.

RWST liquid/vapor elemental iodine partition factor 41.18 Elemental Iodine fraction in RWST Table 2.1-4 Initial RWST Liquid Inventory (minimum at time of 60,000 gallons recirculation)

Release from RWST Vapor Space Table 2.1-5 Containment Purge Release (unfiltered) 7,000 cfm for 8 seconds Removal Inputs:

Containment Aerosol/Particulate Natural Deposition (only 0.1/hour credited in unsprayed regions)

Total Containment Volume 1,550,000 ft3 Surface Area for Wall Deposition 537,903 ft2 Containment Elemental Iodine Wall Deposition 5.58/hour Containment Sprayed Region Volume 534,442 ft3

.Spray Fall Height 70 feet Volumetric Spray Flow Rate 2.986 ft 3/sec Containment Upper Unsprayed Region Volume 643,864 ft3 Containment Lower Unsprayed Region Volume (below 371,694 ft3 operating deck)

Flowrate between Sprayed and Upper Unsprayed Volumes 990,000 cfm Flowrate between Upper Unsprayed and Lower Unsprayed 375,000 cfm Volumes Spray Removal Rates:

Elemental Iodine 20 hr-'

Time to reach DF of 200 2.305 hours0.00353 days 0.0847 hours 5.042989e-4 weeks 1.160525e-4 months Aerosols 6.44 hr-1 (reduced to 0.644 at 3.061 hours7.060185e-4 days 0.0169 hours 1.008598e-4 weeks 2.32105e-5 months )

Time to reach DF of 50 Greater than 3.06 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

Table 2.1-1 Loss of Coolant Accident (LOCA) - Inputs and Assumptions Input/Assumption Value Spray Initiation Time 63.8 seconds Control Room Ventilation System Table 1.6.3-1 Isolation Signal High Containment Radiation Time of CR Isolation 30 seconds Unfiltered Inleakage 115 cfm Containment Purge Filtration 0 %

Transport Inputs:

Nearest containment penetration to CR ventilation Containment Leakage Release intakes ECCS Leakage RWST vent RWST Backleakage RWST vent Containment Purge Plant stack Personnel Dose Conversion Inputs:

Atmospheric Dispersion Factors Table 1.8.2-1 Offsite Tables 1.8.1-2 and 1.8.1-3 Onsite Breathing Rates Reg. Guide 1.183, Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor Reg. Guide 1.183, Section 4.2.6 Table 2.1-2 LOCA Release Phases *

From Reg. Guide 1.183, Table 4

Table 2.1-3 LOCA Time Dependent RWST Total Iodine Concentration*

Time RWST Iodine Concentration (hours) (gm-atom/titer) 0 0 0.25 O.OOOE+00 0.50 2.797E-1 I 1.0 8.384E-11 4.0 4.191E-10 8.0 8.660E- 10 12.0 1.313E-09 24.0 2.654E-09 48.0 5.335E-09 72.0 8.016E-09 96.0 1.070E-08 100.0 1.114E-08 150.0 1.673E-08 200.0 2.231E-08 250.0 2.790E-08 300.0 3.348E-08 350.0 3.906E-08 400.0 4.464E-08, 450.0 5.022E-08 500.0 5.580E-08 550.0 6.137E-08 600.0 6.695E-08 650.0 7.253E-08 700.0 7.8 1OE-08 720.0 8.033E-08 Includes radioactive and stable iodine isotopes

Table 2.1-4 LOCA Time Dependent RWST Elemental Iodine Fraction Time (hr) Elemental Iodine Fraction 0.00 0.OOOE+00 0.25 0.OOOE+00 0.50 3.698E-05 1.0 1.109E-04 4.0 5.536E-04 8.0 1.143E-03 12.0 1.730E-03 24.0 3.485E-03 48.0 6.958E-03 72.0 1.038E-02 96.0 1.376E-02 100.0 1.432E-02 150.0 2.119E-02 200.0 2.788E-02 250.0 3.439E-02 300.0 4.072E-02 350.0 4.690E-02 400.0 5.292E-02 450.0 5.879E-02 500.0 6.452E-02 550.0 7.012E-02 600.0 7.559E-02 650.0 8.093E-02 700.0 8.616E-02 720.0 8.822E-02

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS. INC. Turkey Point Units 3 and 4 Page 65 of 81

ýO ~ NSNPage 65 of 81 Table 2.1-5 Adjusted Release Rate from the RWST Time Adjusted Release Rate (hours) (cfm) 0.00 0 0.25 1.225E-07 12.0 8.437E-07 72.0 1.546E-06 100.0 3.603E-06 300.0 6.796E-06 500.0 9.977E-06 600.0 1.089E-05 700.0 1.148E-05 720.0 1.148E-05 Table 2.1-6 LOCA Dose Consequences 1 Control Room Dose(2)

EAB Dose° ) LPZ Dose(2)

(rem TEDE) (rem TEDE) (rem TEDE)

LOCA 5.85 1.58 4.87 Acceptance Criteria 25 25 5 (1)Worst 2-hour dose (2) Integrated 30-day dose

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 66 of 81 O.UTIONS ANDSOFTWARE INENSIhlERING Table 2.2-1 Fuel Handling Accident (FHA) - Inputs and Assumptions Input/Assumption Value Core Power Level Before Shutdown 2652 MW,,

Discharged Fuel Assembly Burnup 45,000 MWD/MTU Fuel Enrichment 3.0 - 5.0 w/o Radial Peaking Factor 1.65 Number of Fuel Assemblies Damaged 1 Release Fraction from Breached Fuel See Section 1.7.6 Delay Before Spent Fuel Movement 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Release Duration 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months FHA Source Term for a Single Assembly Table 1.7.5-1 Water Level Above Damaged Fuel Assembly 23 feet minimum Elemental Organ - - 285 Iodine Decontamination Factors Organic - I Noble Gas Decontamination Factor I Elemental- - 0.85%

99.85%

Chemical Form of Iodine In Pool Organ Organic - 0. 15 %

Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System Table 1.6.3-1 Time of CR Isolation - Containment Release 30 seconds - High Containment Radiation Time of CR Isolation - FHB Release 30 minutes from Manual Isolation Unfiltered Inleakage 115 cfm Breathing Rates Reg. Guide 1.183 Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor Reg. Guide 1.183 Section 4.2.6 Table 2.2-2 Fuel Handling Accident Dose Consequences EAB Dose(') LPZ Dose(2) Control Room Dose(2)

(rem TEDE) (rem TEDE) (rem TEDE)

FHA in Containment 0.91 0.20 1.32 FHA in Fuel Handling Building 0.91 0.20 3.80 Acceptance Criteria 6.3 6.3 5

('1 Worst 2-hour dose (2) Integrated 30-day dose

Table 2.3-1 Main Steam Line Break (MSLB) - Inputs and Assumptions Input/Assumption Value Core Power Level 2652 MWt, Initial RCS Equilibrium Activity 0.25 [tCi/gm DE 1-131 and 447.7 VCi/gm DE Xe-133 (Table 1.7.2-1)

Initial Secondary Side Equilibrium Iodine Activity 0.1 gCi/gm DE 1-131 (Table 1.7.3-1)

Maximum pre-accident spike iodine concentration 60 gCi/gm DE 1-131 Iodine Spike Appearance Rate 500 times Duration of accident-initiated spike 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Steam Generator Tube Leakage Rate 0.2 gpm/SG Time to establish shutdown cooling and terminate steam 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months release Time for RCS to reach 212'F and terminate SG tube 125.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months leakage RCS Mass (minimum) 366,086 Ibm SG Secondary Side'Mass Faulted SG - 131,516.5 lbm Intact SGs - 67,707 Ibm per SG Release from Faulted SG Instantaneous Steam Release from Intact SGs Table 2.3-2 Time to re-cover Intact SG Tubes 30 minutes Tube Uncovery Flashing Fraction 11%

Faulted SG - 100none Steam Generator Secondary Side Partition Coefficients Iact SG -

Intact SGs - 100 Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System Table 1.6.3-1 Isolation Signal Safety Injection Time of CR Isolation 41.5 seconds Unfiltered Inleakage 115 cfm Breathing Rates Offsite Reg. Guide 1.183, Section 4.1.3 Control Room Reg. Guide 1.183, Section 4.2.6 Control Room Occupancy Factors Reg. Guide 1.183 Section 4.2.6

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 68 of 81 Table 2.3-2 Intact SGs Steam Release Rate*

Time Intact SGs Steam Release Rate (hours) (ibm/min) 0.0 2622 2.0 2058 3.0 1931 4.0 1814 5.0 1694 8.0 1070 11.0 965 16.0 864 24.0 820 63.0 0.0

Stored energy above RHR entry conditions is released between 2 and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Table 2.3-3 60 gCi/gm DE 1-131 Activities Activity Isotope (PCi/gm)

Iodine-131 48.1440 Iodine-132 34.1280 Iodine-133 58.3440 Iodine-134 6.3192 Iodine-135 28.8000 Table 2.3-4 Iodine Equilibrium Appearance Assumptions Input Assumption Value Letdown Flow 132 gpm Identified RCS Leakage 10 gpm Unidentified RCS Leakage 1 gpm RCS Mass 397,544 Ibm 1-131 Decay Constant 6.OOOE-5 min' 1-132 Decay Constant 0.005023 min-1 1-133 Decay Constant 0.000555 min' 1-134 Decay Constant 0.013178 min-1-135 Decay Constant 0.001748 min1

Table 2.3-5 Concurrent (500 x) Iodine Spike Appearance Rate 8-hour Isotope Appearance Rate Production (Ci/min) (Ci)

Iodine- 131 53.90 25870 Iodine-132 101.83 48881 Iodine-133 76.17 36564 Iodine-134 38.22 18343 Iodine-135 50.50 24241 Table 2.3-6 MSLB Dose Consequences EAB Dose (1) LPZ Dose (2) Control Room Dose (2)

(rem TEDE) (rem TEDE) (rem TEDE)

MSLB pre-accident iodine spike 0.03 0.02 1.53 Acceptance Criteria (pre-accident iodine spike) 25 (3) 25 (3) 5 (4)

MSLB concurrent iodine spike 0.05 0.04 1.57 Acceptance Criteria (concurrent iodine spike) 2.5 (3) 2.5 (3) 5 (4)

(1)Worst 2-hour dose (2) Integrated 30-day (3)Reg. Guide 1.183, Table 6 (4) IOCFR50.67

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 70 of 81 INENGINEERING S0LUTIONS AMoSOl-nNeP e Table 2.4-1 Steam Generator Tube Rupture (SGTR) - Inputs and Assumptions Input/Assumption Value Core Power Level 2652 MWt, Initial RCS Equilibrium Activity 0.25 g.tCi/gm DE 1-131 and 447.7 pCi/gm DE Xe-133 Initial __RSEquilibriumActivity_ (Table 1.7.2-1)

Initial Secondary Side Equilibrium Iodine Activity 0.1 tCi/gm DE 1-131 (Table 1.7.3-1)

Maximum pre-accident spike iodine concentration 60 ftCi/gm DE 1-131 Iodine Spike Appearance Rate 335 times Duration of accident-initiated spike 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Integrated Break Flow and Steam Release Table 2.4-2 Prior to Reactor Trip - 21%

Break Flow Flashing Fraction Following Reactor Trip - 11%

Time of Reactor Trip 291 seconds Time to isolate ruptured SG 30 minutes Steam Generator Tube Leakage Rate 0.2 gpm/SG Time to establish shutdown cooling and terminate intact 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months steam release Time for RCS to reach 212'F and terminate SG tube leakage leakage125.4 hours RCS Mass (minimum) 366,086 Ibm SG Secondary Side Mass 67,707 Ibm per SG Time to re-cover Intact-SG Tubes 30 minutes Tube Uncovery Flashing Fraction 11%

SG (Flashed tube flow) - none Secondary Side Partition Coefficients SG (Non-flashed tube flow) - 100 Condenser- 100 Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System Table 1.6.3-1 Isolation Signal Safety Injection Time of CR Isolation 321 seconds Unfiltered Inleakage 115 cfm Breathing Rates Offsite Reg. Guide 1.183, Section 4.1.3 Control Room Reg. Guide 1.183, Section 4.2.6 Control Room Occupancy Factor Reg. Guide 1.183, Section 4.2.6

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 S......

APPLICATIONS, SO.UTYIONS INC.

ANDSOFT'WARE INENGINEERING Turkey Point Units 3 and 4 Page 71 of 81 Table 2.4-2 SGTR Mass Flow Rates ()

Break Flow into Steam Release from Steam Release from (hours) Ruptured SG Ruptured SG Unaffected SGs(2)

(Ibm/min) (Ibn/min) (Ibm/min) 0- 0.0808 6507 64,800 129,600 0.0808- 0.5 4161 3579 4033 0.5-2 0 0 4033 2-8 0 0 2833 8-24 0 0 1525 24-63 0 0 1270 (1)Flowrate is assumed to be constant within the time period (2) Stored energy above RHR entry conditions is released between 2 and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Table 2.4-3 60 tCi/gm D.E. 1-131 Activities Activity Isotope (laCi/gm)

Iodine- 131 48.1440 Iodine-132 34.1280 Iodine-133 58.3440 Iodine-134 6.3192 Iodine-135 28.8000

(i NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS. INC. Turkey Point Units 3 and 4 Page 72 of 81 SOWINS INENGINERING Pae 2 o 8 Table 2.4-4 Iodine Equilibrium Appearance Assumptions Input Assumption Value Letdown Flow 132 gpm Identified RCS Leakage 10 gpm Unidentified RCS Leakage I gpm RCS Mass 397,544 ibm 1-13 1 Decay Constant 6.OOOE-5 min-'

1-132 Decay Constant 0.005023 min-1-133 Decay Constant 0.000555 min' 1-134 Decay Constant 0.013178 rain' 1-135 Decay Constant 0.001748 min-'

Table 2.4-5 Concurrent (335 x) Iodine Spike Appearance Rate 8-hour 8hu Appearance Rate Isotope (CAemin) Production (Ci)

Iodine-131 36.11 17333 Iodine-132 68.23 32750 Iodine-133 51.04 24498 Iodine-134 25.60 12290 Iodine-135 33.84 16241 Table 2.4-6 SGTR Dose Consequences EAB Dose (1) LPZ Dose (2) Control Room Dose (2)

(rem TEDE) (rem TEDE) (rem TEDE)

SGTR pre-accident iodine spike 0.82 0.18 2.85 Acceptance Criteria (pre-accident iodine spike) 25 (3) 25 (3) 5 (4)

SGTR concurrent iodine spike 0.28 0.07 1.10 Acceptance Criteria (concurrent iodine spike) 2.5 (3) 2.5 (3) 5 (4)

(1)Worst 2-hour dose (2) Integrated 30-day (3)Reg. Guide 1.183, Table 6 (4) 1 OCFR50.67

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 73 of 81 Table 2.5-1 Locked Rotor - Inputs and Assumptions Input/Assumption Value Core Power Level 2652 MW, Core Fission Product Inventory Table 1.7.4-1 Initial Secondary Side Equilibrium Iodine Activity 0.1 gtCi/gm DE 1-131 (Table 1.7.3- 1)

Gap Release Fraction Reg. Guide 1.183, Section 3.2, Table 3 Core Average Fuel Burnup 45,000 MWD/MTU Fuel Enrichment 3.0 -5.0 w/o Radial Peaking Factor 1.65 Fuel Failure 15% DNB RCS Mass (minimum) 366,086 Ibm Steam Generator Tube Leakage Rate 0.2 gpm/SG Time to establish shutdown cooling and terminate steam 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months release Time for RCS to reach 212'F and terminate SG tube 125.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months leakage Time to re-cover SG Tubes 30 minutes Tube Uncovery Flashing Fraction 11%

SG Secondary Side Mass 67,707 Ibm per SG Secondary Side Mass Releases to environment Table 2.5ý2 Secondary Side Partition Coefficients SG (Flashed leakage) - none SG (Non-flashed leakage) - 100 Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System Table 1.6.3-1 Isolation Signal High Radiation on Control Room Intake Monitors Time of CR Isolation 60 seconds Unfiltered Inleakage 115 cfm Breathing Rates Offsite Reg. Guide 1.183, Section 4.1.3 Onsite Reg. Guide 1.183, Section 4.2.6 Control Room Occupancy Factor Reg. Guide 1.183, Section 4.2.6

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 74 of 81 Table 2.5-2 Locked Rotor Steam Release*

Time Intact SGs Steam Release Rate (hours) (Ibm/min) 0.0 2598 2.0 2143 3.0 2016 4.0 1900 5.0 1779 8.0 2598 11.0 965 16.0 864 24.0 820 63.0 0.0 Stored energy above RHR entry conditions is released between 2 and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Table 2.5-3 Locked Rotor Dose Consequences EAB Dose (1) LPZ Dose (2) Control Room Dose (2)

(rem TEDE) (rem TEDE) (rem TEDE)

Locked Rotor 0.58 0.62 1.33 Acceptance Criteria 2.5 (3) 2.5 (3) 5 (4)

(1)Worst 2-hour dose (2) Integrated 30-day dose (3)Reg. Guide 1.183, Table 6 (4) 10CFR50.67

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 75 of 81 Page 75 of 81 SO.llSIN ENSH NfRI NG Table 2.6-1 RCCA Ejection - Inputs and Assumptions Input/Assumption Value Core Power Level 2652 MW1, Core Average Fuel Burnup 45,000 MWD/MTU Fuel Enrichment 3.0 -5.0 w/o Radial Peaking Factor 1.65 Percent of Core in DNB Design Basis 10%

Manual CR Isolation Case 6.22%

Percent of Core with Centerline Melt Design Basis 0.25%

Manual CR Isolation Case 0.16%

Gap Release Fraction Reg. Guide 1.183, Appendix H, Position 1 Core Fission Product Inventory Table 1.7.4-1 Initial Secondary Side Equilibrium Iodine Activity 0.1 gaCi/gm DE 1-131 (Table 1.7.3-1)

Release From DNB Fuel Section 1 of Appendix H to Reg. Guide 1.183 Release From Fuel Centerline Melt Fuel Section 1 of Appendix H to RG 1.183 SG (Flashed leakage) - none Secondary Side Partition Coefficients SG (Non-flashed leakage) - 100 Steam Generator Tube Leakage Rate 0.2 gpm/SG Time to establish shutdown cooling and terminate steam 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months release Time to re-cover SG Tubes 30 minutes Tube Uncovery Flashing Fraction 11%

RCS Mass (minimum) 366,086 Ibm SG Secondary Side Mass 67,707 Ibm per SG Particulate - 95%

Chemical Form of Iodine Released to Containment Elemental - 4.85%

Organic - 0.15%

Particulate - 0%

Chemical Form of Iodine Released from SGs Elemental - 97 %

Organic - 3%

Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System Table 1.6.3-1 Time of CR Isolation - Containment Release 30 seconds - High Containment Radiation Time of CR Isolation - Secondary (Automatic) 60 seconds - High Radiation on CR Intake Monitors Time of CR Isolation - Secondary (Manual) 30 Minutes from Manual Isolation Unfiltered Inleakage 115 cfm

Table 2.6-1 RCCA Ejection - Inputs and Assumptions Input/Assumption Value Breathing Rates Reg. Guide 1.183 Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor Reg. Guide 1.183 Section 4.2.6 Containment Volume 1.55E+06 ft3 Containment Leakage Rate 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.20% (by weight)/day after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.10% (by weight)/day Aerosols - 0.1 hr 1 Containment Natural Deposition Coefficients Elemental Iodine - 5.58 hr-1 Organic Iodine - None Table 2.6-2 RCCA Ejection Steam Release*

Time Intact SGs Steam Release Rate (hours) (Ibm/min) 0.0 2598 2.0 2143 3.0 2016 4.0 1900 5.0 1779 8.0 2598 11.0 965 16.0 864 24.0 820 63.0 0.0

Stored energy above RHR entry conditions is released between 2 and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

Table 2.6-3 RCCA Ejection Dose Consequences 2

EAB Dose*') LPZ Dose(2) Control Room Dose( )

(rem TEDE) (rem TEDE) (rem TEDE)

RCCA Ejection - Containment Release 0.88 0.40 2.47 RCCA Ejection - Secondary Release 0.61 0.57 1.19 (Automatic CR Isolation)

RCCA Ejection - Secondary Release 0.38 0.36 3.46 (Manual CR Isolation)

Acceptance Criteria 6.3 (3) 6.3 (3) 5 (4)

() Worst 2-hour dose (2) Integrated 30-day dose (3)Reg. Guide 1.183, Table 6 (4) 1 0CFR50.67 Table 2.7-1 Waste Gas Decay Tank (WGDT) Rupture - Inputs and Assumptions Input/Assumption Value Core Power Level 2652 MWt, WGDT inventory RCS equilibrium with 1% fuel defects (Table 2.7-2).

Equal to 84,274.8 Curies Dose Equivalent Xe-133.

Tank volume 525 ft3 Tank leak rate (arbitrarily high) I E+06 cfm Control Room Ventilation System Table 1.6.3-1 Time of CR Isolation Not isolated Unfiltered Inleakage 115 cfm Makeup Flow 1000 cfm Atmospheric Dispersion Factors Table 1.8.2-1 Offsite Tables 1.8.1-2 and 1.8.1-3 Onsite Breathing OffsiteRates Reg. Guide 1.183, Section 4.1.3 Reg. Guide 1.183, Section 4.2.6 Control Room Reg. Guide 1.183, Section 4.2.6 CR Occupancy Factors Reg. Guide 1.183, Section 4.2.6

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. I APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 78 of 81 OrrNSIN ENGINtING AND POae7 o 8 Table 2.7-2 WGDT Source Term (1)

Isotope Tank Inventory (Curies)

Kr-85m 214.22 Kr-85 6286.02 Kr-87 129.06 Kr-88 381.02 Xe-131m 513.74 Xe-133 42555.99 Xe-133m 590.55 Xe-135 916.04 Xe-135m 82.41 Xe-138 85.94

(')Tank activity equals total RCS equilibrium noble gas activity with 1% fuel defects Table 2.7-3 WGDT Rupture Dose Consequences EAB Dose (I) LPZ Dose () Control Room Dose (1)

(rem TEDE) (rem TEDE) (rem TEDE)

WGDT 0.08 0.02 0.33 Acceptance Criteria 0.1(3) 0.1(3) 5 (2)

(1)Integrated 30-day dose (2) 10CFR50.67 (3)SRP BTP 11-5

Table 2.8-1 Spent Fuel Cask Drop- Inputs and Assumptions Input/Assumption Value Core Power Level Before Shutdown 2652 MWh Core Average Fuel Burnup 45,000 MWD/MTU Fuel Enrichment 3.0 - 5.0 w/o Core Fission Product Inventory Table 1.7.4-1 Number of Fuel Assemblies Damaged 157 Gap Release Fraction FHA high burnup gap fractions from Section 1.7.6 Delay Before Spent Fuel Movement 1525 hours0.0177 days 0.424 hours 0.00252 weeks 5.802625e-4 months Release Duration 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Water Level Above Damaged Fuel Assemblies 23 feet minimum Iodine Decontamination Factors Elemental - 285 Organic- I Noble Gas Decontamination Factor I Chemical Form of Iodine In Pool Elemental - 99.85%

Organic - 0.15%

Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System Table 1.6.3-1 Time of CR Isolation 30 minutes from Manual Isolation Unfiltered Inleakage 115 cfm Breathing Rates RG 1.183 Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6

NUMERICAL AST Licensing Technical Report for NAI-1396-045, Rev. 1 APPLICATIONS, INC. Turkey Point Units 3 and 4 Page 80 of 81 Table 2.8-2 Spent Fuel Cask Drop Dose Consequences EAB Dose (1) LPZ Dose (2) Control Room Dose (2)

(rem TEDE) (rem TEDE) (rem TEDE)

Spent Fuel Cask Drop 0.40 0.09 2.05 Acceptance Criteria 6.3(3) 6.3(3) 5(4)

(1) Worst 2-hour dose (2) Integrated 30-day (3) FHA Criteria from Reg. Guide 1.183, Table 6 (4) 10CFR50.67

Table 3-1 Turkey Point Units No. 3 and 4 Summary of AST Analysis Results TEAB Dose() LPZ Dose (2) Control (

Case (rem TEDE) (rem TEDE) Room Dose(2 )

(rem TEDE)

LOCA 5.85 1.58 4.87 MSLB Pre-accident Iodine Spike 0.03 0.02 1.53 SGTR Pre-accident Iodine Spike 0.82 0.18 2.85 Acceptance Criteria < 25(3) 5 25") < 5(4)

MSLB Concurrent Iodine Spike 0.05 0.04 1.57 SGTR Concurrent Iodine Spike 0.28 0.07 1.10 Locked Rotor 0.58 0.62 1.33 Acceptance Criteria < 2.5 (3) < 2.5 (3) _5 (4)

FHA - Containment Release 0.91 0.20 1.32 FHA - Fuel Building Release 0.91 0.20 3.80 Spent Fuel Cask Drop 0.40 0.09 2.05 RCCA Ejection - Containment 0.88 0.40 2.47 RCCA Ejection - Secondary 0.61 0.57 1.19 (Automatic CR Isolation)

RCCA Ejection - Secondary 0.38 0.36 3.46 (Manual CR Isolation)

Acceptance Criteria <6.3 13) 6.313) <5(4)

WGDT 0.08(2) 0.02 0.33 Acceptance Criteria <0.1(51 0.15) < 5 (4)

(1)Worst 2-hour dose (2) Integrated 30-day dose (3)Reg. Guide 1.183, Table 6 (4) 10CFR50.67 (5)SRP BTP 11-5

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 4 License Amendment Request 196 Page 1 of 2 Alternative Source Term and Conforming Amendment LICENSE AMENDMENT REQUEST LAR 196 ALTERNATIVE SOURCE TERM AND CONFORMING AMENDMENT ATTACHMENT 4 LIST OF COMMITMENTS

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 4 License Amendment Request 196 Page 2 of 2 Alternative Source Term and Conforming Amendment Regulatory Commitments The following table identifies those actions committed to by FPL in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENTS DUE DATE/EVENT FPL will relocate the Control Room Ventilation System emergency air intakes prior to implementation of Alternative Source Term.

FPL will install stainless steel wire mesh baskets to maintain pH during the sump recirculation phase following a Design Basis LOCA.

FPL will supersede the commitment from PTN Letters L-2009-062 and L-2009-063 related to NaTB sampling with the proposed Technical Specification requirements for the Recirculation pH Control System.

FPL will replace the aluminum fins on the normal containment coolers with copper fins.

FPL will update the necessary procedures to implement the manual operator action for initiation of the emergency control room ventilation system.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 1 of 39 Alternative Source Term and Conforming Amendment LICENSE AMENDMENT REQUEST ALTERNATIVE SOURCE TERM AND CONFORMING AMENDMENT LAR 196 ATTACHMENT 5 INPUT/ASSUMPTIONS COMPARISON TABLES

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 2 of 39 Alternative Source Term and Conforming Amendment TABLE 5-1 LOSS OF COOLANT ACCIDENT (LOCA)

AST INPUTS AND ASSUMPTIONS Release Inputs INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Core Power Level 2346 MWt 2652 MWt AST core power models planned EPU.

Core Average Fuel Burnup 40,000 MWD/MTU 45,000 MWD/MTU AST core average burnup conservatively increased to accommodate a planned EPU.

Fuel Enrichment 3.0 - 4.5% U-235 3.0 - 5.0% U-235 AST maximum fuel enrichment conservatively increased to accommodate a planned EPU.

Initial RCS Equilibrium Activity 1.0 jiCi/gm DEI-131 0.25 RCi/gm DEI-131 and AST values are consistent with proposed 100/F_ gCi/gram of 447.7 iCi/gm DEXe-133 changes to more restrictive TS limits.

gross radioactivity Maximum RCS Mass 366,086 lbm 397,544 Ibm Current Nominal RCS no-load conditions AST - Maximum RCS solid conditions Since the initial RCS activity is assumed to be at TS limits, this maximizes the release to containment used in the calculation of the containment purge dose.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 3 of 39 Alternative Source Term and Conforming Amendment TABLE 5-1 (continued)

LOSS OF COOLANT ACCIDENT (LOCA)

AST INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Core Fission Product Inventory Attachment 3 Table 1.7.4-1 AST source term conservatively bounds 1-131 6.17E7 the planned EPU and current conditions.

1-132 8.19E7 1-133 1.26E8 1-134 1.39E8 1-135 1.18E8 Kr-85m 1.64E7 Kr-85 7.80E5 Kr-87 3.15E7 Kr-99 4.44E7 Xe-131m 6.87E5 Xe-133m 3.93E6 Xe-133 1.26E8 Xe-135m 2.46E7 Xe-135 3.62E7 Xe-138 1.05E8

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 4 of 39 Alternative Source Term and Conforming Amendment TABLE 5-1 (continued)

LOSS OF COOLANT ACCIDENT (LOCA)

AST INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Containment Leakage Rate 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.25% (by weight)/day 0.20% (by weight)/day AST values are consistent with proposed After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.125% (by weight)/day 0.10% (by weight)/day changes to more restrictive TS limits.

LOCA Release Phase Timing and Duration Gap Release Onset Instantaneous 30 seconds Duration Instantaneous 30 minutes Early In-Vessel Onset Instantaneous 30 minutes Duration Instantaneous 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Core Inventory Release Fractions (gap release and early in- N/A RG 1.183, Sections 3.1 and Current- Instantaneous release of 50% of vessel phases) 3.2 and Table 2 core iodine and 100% of noble gases.

ECCS Leakage Current - Dose consequences from ECCS leakage not part of the current licensing basis.

Sump Volume (minimum) N/A 239,000 gallons AST - Minimum sump volume 3 maximizes sump activity concentration.

(31, 949.5 ft )

ECCS Leakage Duration N/A 15 minutes to 30 days

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 5 of 39 Alternative Source Term and Conforming Amendment TABLE 5-1 (continued)

LOSS OF COOLANT ACCIDENT (LOCA)

AST INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS ECCS Leakage (2 times allowable value) N/A 4,650 cc/hr AST - The maximum allowable ECCS (2 times allowable value) leakage value is doubled in accordance with RG 1.183.

Flashing Fraction N/A 9.2%

Airborne Iodine Fraction N/A 10%

Chemical Form of Iodine Released from ECCS Leakage Elemental N/A 97%

Organic N/A 3%

Initial RWST Liquid Inventory (minimum at time of 60,000 gallons 60,000 gallons recirculation Release from RWST Vapor Space N/A Attachment 3 Table 2.1-5 Containment Purge Release (unfiltered) 7,000 cfmn for 8 seconds 7,000 cfm for 8 seconds RWST Back-leakage Sump Volume (minimum) N/A 239,000 gallons Current - Dose consequences from (31,949.5 ft3) RWST back-leakage are not part of the RWST Back Leakage Duration N/A 15 minutes to 30 days current licensing basis.

ECCS Leakage N/A 0.1 gph Flashing Fraction N/A 0% based on temperature of fluid reaching RWST.

RWST Liquid/Vapor Elemental Iodine Partition N/A 41.18 Factor Elemental Iodine Fraction in RWST N/A Attachment 3 Table 2.1-4

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 6 of 39 Alternative Source Term and Conformin2 Amendment TABLE 5-1 (continued)

LOSS OF COOLANT ACCIDENT (LOCA)

AST INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Removal Inputs Total Containment Volume 1,550,000 ft3 1,550,000 ft3 Containment Sprayed Region Volume Not credited 534,442 ft3 Current - Containment spray is not Containment Upper Unsprayed Region Volume Not credited 643,864 ft3 credited in the current licensing dose consequences analysis.

Not credited 371,694 ft3 Containment Lower Unsprayed Region Volume Containment Natural Deposition Coefficients Instantaneous Iodine Plateout at t=0 50% 0%

Aerosols None 0.1 hr-'

Elemental Iodine 5.94/hour for DF < 100 5.58 hr-1 0/hour for DF > 100

_ Organic Iodine None None Surface Area for Natural Deposition 569,219 ft2 537,903 ft2 AST - Total containment surface area participating in surface deposition was reduced by 25% for natural deposition of elemental iodine.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 7 of 39 Alternative Source Term and Conforming Amendment TABLE 5-1 (continued)

LOSS OF COOLANT ACCIDENT (LOCA)

AST INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Spray Fall Height Not credited 70 feet Current - Containment spray is not Volumetric Spray Flow Rate Not credited 2.986 v/sec credited in the current licensing dose consequences analysis.

Flowrate between Sprayed and Upper Unsprayed Volumes Not credited 990,000 cfm Flowrate between Upper/Lower Unsprayed Volumes Not credited 375,000 cfm Spray Removal Rates Elemental Iodine Not credited 20/hour Time to reach DF of 200 Not credited 2.305 hours0.00353 days 0.0847 hours 5.042989e-4 weeks 1.160525e-4 months Aerosol Not credited 6.44/hour (reduced to 0.644 at 3.053 hours6.134259e-4 days 0.0147 hours 8.763227e-5 weeks 2.01665e-5 months )

Time to reach DF of 50 Not credited 3.053 hours6.134259e-4 days 0.0147 hours 8.763227e-5 weeks 2.01665e-5 months Spray Initiation Time Not credited 63.8 seconds

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 8 of 39 Alternative Source Term and Conforming Amendment TABLE 5-1 (continued)

LOSS OF COOLANT ACCIDENT (LOCA)

AST INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION " CURRENT VALUE AST VALUE REMARKS Control Room Ventilation System Control Room Volume 47,786 ft3 47,786 ft3 Normal Operation Filtered Makeup Flow Rate 0 cfm 0 cfm Filtered Recirculation Flow Rate 0 cfm 0 cfm Unfiltered Makeup Flow Rate and 1,000 cfm 1,000 cfm Inleakage Recirculation Mode Filtered Makeup Flow Rate 525 cfm 525 cfm Filtered Recirculation Flow Rate 375 cfm 375 cfm Unfiltered Makeup Flow Rate 0 cfm 0 cfm Unfiltered Inleakage 10 cfm 115 cfm Filter Efficiencies AST - Filter efficiencies are consistent Elemental 95% 97.5% with proposed change to more restrictive TS.

Organic 95% 97.5%

Particulate 95% 97.5%

Time of Control Room Isolation Control room isolated 30 seconds prior to arrival of plume Control Room Unfiltered Inleakage 10 cfm 115 cfm

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 9 of 39 Alternative Source Term and Conforming Amendment TABLE 5-1 (continued)

LOSS OF COOLANT ACCIDENT (LOCA)

AST INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Emergency Containment Filter Efficiency 2 ECFs credited Not credited Particulate 95%

Elemental 90%

Methyl 30%

Containment Purge Filtration 0% 0%

Transport Inputs Containment Leakage Release Containment to control Nearest containment room penetration to CR ventilation intakes ECCS Leakage N/A Plant stack Current - Dose consequences from ECCS leakage not part of the current licensing basis.

RWST Backleakage N/A RWST vent Current - Dose consequences form RWST back-leakage are not part of the current licensing basis.

Containment Purge Containment to control Plant stack room

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 10 of 39 Alternative Source Term and Conforming Amendment TABLE 5-1 (continued)

LOSS OF COOLANT ACCIDENT (LOCA)

AST INPUTS AND ASSUMPTIONS Personnel Dose Conversion Inputs INPUT/ASSUMPTION CURRENT VALUE AST VALUE ] REMARKS Atmospheric Dispersion Factors AST - All Dispersion Factors recalculated Offsite 1.54 E-4 sec/m 3 Attachment 3 Table 1.8.2-1 using guidance from RG 1.194 and RG Low Population Zone 1.145.

3 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.50 E-5 sec/m 3

2 - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 6.50 E-6 sec/m 12 -720 hours 2.40 E-7 sec/m 3 S0-Control Room 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 9.58 E-4 sec/m 7.52 E-4 sec/m 3

3 Attachment 3 Tables 1.8.1-2 and 1.8.1-3 3

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 5.26,E-4 sec/m 3

96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 2.94 E-4 sec/m Breathing Rates Offsite 3.47 E-4 M3/sec 3.5 E-4 m3/sec 0-8 hours 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.75 E-4 m3/sec 1.8 E-4 m3/sec 3

24 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 2.32 E-4 m /sec 2.3 E-4 M3/sec Control Room Duration of Event Same as Offsite 3.5 E-4 m3/sec

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 11 of 39 Alternative Source Term and Conforming Amendment TABLE 5-1 (continued)

LOSS OF COOLANT ACCIDENT (LOCA)

AST INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Control Room Occupancy Factor 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.0 100%

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 0.6 60%

96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 0.4 40%

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 12 of 39 Alternative Source Term and Conforming Amendment TABLE 5-2 FUEL HANDLING ACCIDENT (FHA)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Core Power Level Before Shutdown 2346 MWt 2652 MWt AST core power models planned EPU.

Discharged Fuel Assembly Burnup Core Average Burnup 40,000 MWD/MTU 45,000 MWD/MTU AST core average burnup conservatively increased to accommodate a planned EPU.

Peak Average Burnup 62,000 MWD/MTU 62,000 MWD/MTU Current & AST - gap release fraction for breached fuel increased to accommodate high bumup fuel; see Enclosure 3, Section 1.7.6.

Fuel Enrichment 3.0 - 4.5 % U-235 3.0 - 5.0 % U-235 AST maximum fuel enrichment Core Average Assembly 4.5% U-235 conservatively increased to accommodate a planned EPU.

4.5% U-235 Peak Assembly Radial Peaking Factor 1.7 1.65 AST assumed radial peaking factor combined with the planned EPU core power value bounds the current radial peaking factor combined with the current power level.

Number of Fuel Assemblies Damaged 1 1 Gap Release Fraction for Breached Fuel 0.10, except 0.12 (1-131) 0.10, except 0.12 (1-13 1) & Current & AST - gap release fraction for

& 0.30 (Kr-85) 0.30 (Kr-85) breached fuel increased to accommodate high bumup fuel; see Enclosure 3, Section 1.7.6.

Delay Before Spent Fuel Movement 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Release Duration 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-25 1 Attachment 5 License Amendment Request 196 Page 13 of 39 Alternative Source Term and Conforming Amendment TABLE 5-2 (continued)

FUEL HANDLING ACCIDENT (FHA)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS FHA Source Term for Single Assembly UFSAR Table 14.2.1-1 Attachment 3 Table 1.7.5-1 AST source term conservatively bounds the planned EPU and current conditions.

Water Level Above Damaged Fuel Assembly 23 feet minimum 23 feet minimum Overall Iodine Decontamination Factor 200 200 Current & AST - An overall iodine contamination factor of 200 is used per RG 1.183, Appendix B. This is equivalent to a decontamination factor of 285 for elemental iodine and I for organic iodine per the guidance of RIS 2006-04.

Noble Gas Decontamination Factor 1 1 Chemical Form of Iodine in Pool Elemental 57% 57% Current & AST - pool chemical Organic 43% 43% composition based on RG 1.183, Appendix B.

Atmospheric Dispersion Factors AST - All Dispersion Factors recalculated Offsite (EAB) 1.54 E-04 sec/m 3 Attachment 3 Table 1.8.2-1 using guidance from RG 1.194 and RG 1.145.

Offsite (LPZ) NA Onsite (Control Room) 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (FHA Puff Release) 1.51 E-03 sec/m 3 Attachment 3 Tables 1.8.1-2 3

0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 9.58 E-04 sec/m and 1.8.1-3 3

8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7.52 E-04 sec/M 3

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 5.26 E-04 sec/m 3

96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 2.94 E-04 sec/m

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 14 of"39 Alternative Source Term and Conforming Amendment TABLE 5-2 (continued)

FUEL HANDLING ACCIDENT (FHA)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE [ REMARKS Control Room Ventilation System Control Room Volume 47,786 ft3 47,786 ft3 AST filter efficiency values are consistent Normal Operation with proposed changes to more restrictive TS limits.

Filtered Makeup Flow Rate 0 cfm 0 cfm Filtered Recirculation Flow Rate 0 cfm 0 cfm Unfiltered Makeup Flow Rate and 1,000 cfm 1,000 cfm Inleakage Recirculation Mode Filtered Makeup Flow Rate 525 cfm 525 cfm Filtered Recirculation Flow Rate 375 cfm 375 cfm Unfiltered Makeup Flow Rate 0 cfm 0 cfm Unfiltered Inleakage 500 cfm 115 cfm Filter Efficiencies Elemental 95% 97.5%

Organic 95% 97.5%

Particulate N/A 97.5%

Time of Control Room Isolation- Containment 30 seconds 30 seconds Current- Dose consequences of the FHB release Release not specifically modeled and assumed equivalent to the containment release dose consequences.

AST - FHB release specifically modeled in to address the recalculated Atmospheric Dispersion factors.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 15 of 39 Alternative Source Term and Conforming Amendment TABLE 5-2 (continued)

FUEL HANDLING ACCIDENT (FHA)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE] AST VALUE REMARKS Breathing Rates Offsite (LPZ) 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.47 E-04 m3/sec 3.5 E-4 m3/sec 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.8 E-4 m3/sec After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.3 E-4 m3/sec Offsite (EAB) 3.47 E-04 m3/sec 1.71 E-4 m3/sec Control Room Duration of the event 3.47 E-04 m3/sec 3.5 E-4 m3/sec Control Room Occupancy Factor 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 100% 100%

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 60% 60%

96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 40% 40%

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 16 of 39 Alternative Source Term and Conforming Amendment TABLE 5-3 MAIN STEAM LINE BREAK (MSLB)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Core Power Level 2346 MWt 2652 MWt AST core power models planned EPU.

Initial RCS Equilibrium Activity 1.0 ltCi/gm DEI-131 0.25 pCi/gm DEI-131 and AST values are consistent with proposed changes to 447.7 tCi/gm DEXe-133 more restrictive TS limits.

Initial Secondary Side Equilibrium 0.1 pCi/gm DEI-131 0.1 tCi/gm DEI- 131 Iodine Activity Maximum Pre-Accident Spike Iodine 60 tCi/gm DEI-131 60 pCi/gm DEI-131 Concentration Iodine Spike Appearance Rate 500 times 500 times Duration of Accident Initiated Spike 0.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months AST - The spike duration is consistent with the guidance in RG 1.183, Appendix E.

Steam Generator Tube Leakage Rate 500 gpd/SG at accident 0.2 gpm/SG at room AST - Accident leakage consistent with analysis conditions temperature conditions assumptions and proposed change to SG performance criterion TS.

Time to Establish Shutdown Cooling 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months AST - Cooldown times conservatively bound the current and Terminate Steam Release and planned EPU conditions.

Time for RCS to Reach 2127F and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 125.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Terminate SG Tube Leakage Minimum RCS Mass 366,086 Ibm 366,086 Ibm Current - Minimum RCS mass bounds the current and EPU conditions due to higher steam generator tube plugging assumption (20% vs. 10%).

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 17 of 39 Alternative Source Term and Conforming Amendment TABLE 5-3 (continued)

MAIN STEAM LINE BREAK (MSLB)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT[ AST REMARKS VALUE VALUE Steam Generator Secondary Side Current - Values are based on nominal mass at full power.

Mass AST - Conservatively assumes maximum SG mass for the faulted SG to maximize the pre-existing secondary activity release to the environment- and the minimum mass for the intact SGs to maximize the release due to RCS iodine spiking and bounds current and planned EPU conditions..

Faulted Steam Generator 84,128 Ibm 131,516.5 (Maximum) lbm Intact Steam Generators 84,128 Ibm 67,707 Ibm (Minimum) per SG per SG Release from Faulted SG AST - Steam releases conservatively bound current and planned EPU conditions 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 84,128 Ibm 131,516,5 Ibm; Instantaneous Steam Release from Intact SGs 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 269,700 Ibm Attachment 3 Table 2.3-2 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 369,300 lbm 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 984,700 Ibm

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 18 of 39 Alternative Source Term and Conforming Amendment TABLE 5-3 (continued)

MAIN STEAM LINE BREAK (MSLB)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Time to Re-Cover Intact SG Tubes N/A 30 minutes Current - Dose N/A 11% consequences from tube Tube Uncovery Flashing Fraction uncovery are not part of the current licensing basis.

Steam Generator Secondary Side Partition Coefficients Equivalent values used o1.0 None for both Current & AST Current - values are Intact Steam Generator 0.01 100 multipliers AST - value is a divisor Atmospheric Dispersion Factors AST - All Dispersion Offisite Attachment 3 Table 1.8.2-1 Factors recalculated 0 -2 hours (EAB) 1.54E-4 m3/sec M3 using 1.194 guidance from RG and RG 1.145.

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (LPZ) 1.5 E-5 m3/sec 3

2 - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (LPZ) 6.5 E-6 M /sec 3

>12 hours (LPZ) 2.4 E-7 m /sec

[Onsite N/A Attachment 3 Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System N/A Attachment 3 Table 1.6.3-1 Current - CR doses not calculated - bounded by Time of Control Room Isolation N/A 41.5 seconds calysis.

LOCA analysis.

Control Room Unfiltered Inleakage N/A 115 cfm

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 19 of 39 Alternative Source Term and Conforming Amendment TABLE 5-3 (continued)

MAIN STEAM LINE BREAK (MSLB)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Breathing Rates Offsite 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.47 E-4 m3/sec 3.5 E-4 m3/sec 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.75 E-4 m3/sec 1.8 E-4 m3/sec After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months N/A 2.3 E-4 m3/sec Control Room Duration of the event N/A 3.5 E-4 m3/sec Control Room Occupancy Factor 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months N/A 100%

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 60%

96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 40%

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 20 of 39 Alternative Source Term and Conforming Amendment TABLE 5-4 STEAM GENERATOR TUBE RUPTURE (SGTR)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT AST VALUE REMARKS VALUE Core Power Level 2300 MWt 2652 MWt AST core power models planned EPU.

Initial RCS Equilibrium Activity 1.0 gCi/gm DEI-131 0.25 pCi/gm DEI-131 and AST values are consistent with proposed changes to more accident initiated 447.7 pCi/gm DEXe-133 restrictive TS limits.

iodine spike case Initial Secondary Side Equilibrium 0.1 gCi/gm DEI-131 0.1 gCi/gm DEI-131 Iodine Activity Maximum Pre-Accident Spike Iodine 60 pCi/gm DEI-131 60 ptCi/gm DEI-131 Concentration Iodine Spike Appearance Rate 500 times 335 times AST - The iodine appearance rate is consistent with the

- _guidance in RG 1.183, Appendix F.

Duration of Accident Initiated Spike 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months AST - The spike duration is consistent with the guidance in RG 1.183, Appendix F.

Integrated Break Flow and Steam Attachment 3 Table 2.4-2 AST - Steam releases conservatively bound current and Release planned EPU conditions Ruptured SG Break Flow: 118,300 0 - 30 minutes ibm Steam Release:

63,000 30 minutes - 8.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 2160 Ibm Intact SGs 346,00 Ibm 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 hours - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1,936,000 Ibm

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 21 of 39 Alternative Source Term and Conforming Amendment TABLE 5-4 (continued)

STEAM GENERATOR TUBE RUPTURE (SGTR)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT AST VALUE REMARKS VALUE Break Flow Flashing Fraction AST- Break flow flashing fraction used that bounds current Prior to Reactor Trip N/A 21% and planned EPU conditions Following Reactor Trip 11% 11%

Time of Reactor Trip N/A 291 seconds Current - Loss of Offsite Power assumed at event initiation.

AST- Loss of Offsite Power assumed concurrent with reactor trip. This conservatively models releases from the condenser steam jet air ejector, which has higher Atmospheric Dispersion Factors than the ADVs or MSSVs.

Time to Isolate Ruptured SG 30 minutes 30 minutes Steam Generator Tube Leakage Rate 500 gpd/SG 0.2 gpm/SG AST - Accident leakage consistent with analysis assumptions and proposed change to SG performance criterion TS.

Time to Establish Shutdown Cooling and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months Cooldown times conservatively bound the current and planned Terminate Intact Steam Release EPU conditions.

Time for RCS to reach 212oF and Terminate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 125.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SG Tube Leakage Minimum RCS Mass 366,086 Ibm 366,086 Ibm Current - Minimum RCS mass bounds the current and EPU conditions due to higher steam generator tube plugging assumption (20% vs. 10%).

SG Secondary Side Mass 81,860 Ibm per SG 67,707 lbm per SG AST - The minimum mass for the intact SGs is assumed to maximize the release due to RCS iodine spiking and the value bounds current and planned EPU conditions..

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 22 of 39 Alternative Source Term and Conforming Amendment TABLE 5-4 (continued)

STEAM GENERATOR TUBE RUPTURE (SGTR)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Time to Re-Cover Intact SG Tubes N/A 30 minutes Current - Dose consequences Tube Uncovery Flashing Fraction N/A 11% from tube uncovery are not part of the current licensing basis.

Steam Generator Secondary Side Partition Coefficients SG (Flashed tube flow) 1.0 None Current - values are multipliers SG (Non-Flashed tube flow) 0.01 100 AST -value is a divisor Condenser N/A 100 Atmospheric Dispersion Factors AST - All Dispersion Factors S site Attachment 3 Table 1.8.2-1 recalculated using guidance from 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (EAB) 1.54E-4 m3/sec RG 1.194 and RG 1.145.

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (LPZ) 1.5 E-5 m3/sec 2 - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (LPZ) 6.5 E-6 m3/sec

>12 hours (LPZ) 2.4 E-7 m3/sec Onsite N/A Attachment 3 Tables 1.8.1-2 Current - CR doses not and 1.8.1-3 calculated - bounded by LOCA analysis Control Room Ventilation System N/A Attachment 3 Table 1.6.3-1 Current - CR doses bounded by N/A 321 seconds LOCA analysis.

Time of Control Room Isolation Control Room Unfiltered Inleakage N/A 115 cfm

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 23 of 39 Alternative Source Term and Conforming Amendment TABLE 5-4 (continued)

STEAM GENERATOR TUBE RUPTURE (SGTR)

INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Breathing Rates Offsite 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.47 E-4 m3/sec 3.5 E-4 m3/sec 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.75 E-4 m3/sec 1.8 E-4 m3/sec After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months N/A 2.3 E-4 m3/sec Control Room Duration of the event N/A 3.5 E-4 m3/sec Control Room Occupancy Factor 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months N/A 100%

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months N/A 60%

96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months N/A 40%

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 24 of 39 Alternative Source Term and Conforming Amendment TABLE 5-5 LOCKED ROTOR INPUTS AND ASSUM PTIONS REMARKS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Core Power Level 2346 MWt 2652 MWt AST core power models planned EPU.

Core Fission Product Inventory See Table 5-1 values Attachment 3 Table 1.7.4-1 The AST core fission product inventory bounds current and planned EPU conditions.

Iodine Secondary Side Equilibrium Iodine Activity 0.1 pCi/gm DEI-131 0.1 gCi/gm DEI-131 Release Fraction from Breached Fuel 1-131 0.10 0.08 AST -RG 1.183 Section 3.2 Table 3 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12 Core Average Fuel Burnup 40,000 MWD/MT 45,000 MWD/MTU AST core average burnup conservatively increased to accommodate a planned EPU.

Fuel Enrichment 3.0 - 4.5% U-235 3.0 - 5.0 % U-235 AST maximum fuel enrichment conservatively increased to accommodate a planned EPU.

Radial Peaking Factor 1.7 1.65 AST assumed radial peaking factor combined with the planned EPU core power value bounds the current radial peaking factor combined with the current power level.

Fuel Failure 10% 15% DNB AST - Assumed fuel failures conservatively increased to accommodate a planned EPU.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 25 of 39 Alternative Source Term and Conforming Amendment TABLE 5-5 (continued)

LOCKED ROTOR INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Minimum RCS Mass 366,086 Ibm 366,086 Ibm Current - Minimum RCS mass bounds the current and EPU conditions due to higher steam generator tube plugging assumption (20% vs. 10%).

Steam Generator Tube Leakage Rate 1.0 gpm at accident 0.2 gpm/SG at room AST values are consistent with proposed conditions temperature conditions changes to lower TS limits.

Time to Establish Shutdown Cooling and Terminate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months AST - Cooldown times conservatively Steam Release bound the current and planned EPU Time for RCS to reach 212'F and Terminate SG Tube 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 125.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months conditions.

Leakage Time to Re-Cover SG Tubes N/A 30 minutes Current - Dose consequences from tube Tube Uncovery Flashing Fraction N/A 11% uncovery are not part of the current licensing basis.

SG Secondary Side Mass 84,128 Ibm per SG 67,707 Ibm per SG AST - The minimum mass for the intact SGs is assumed to maximize the release due to RCS iodine spiking and the value bounds current and planned EPU conditions..

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 26 of 39 Alternative Source Term and Conforming Amendment TABLE 5-5 (continued)

LOCKED ROTOR INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Secondary Side Mass Releases to Environment Attachment 3 Table 2.5-2 AST - Steam releases conservatively bound 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 521,000 Ibm current and planned EPU conditions 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 448,400 Ibm 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1,196,000 Ibm Secondary Side Partition Coefficients SG (Flashed Leakage) N/A None Current - values are multipliers SG (Non-Flashed Leakage) N/A 100 AST - value is a divisor Iodine Partition Factor 0.01 N/A Atmospheric Dispersion Factors AST - All Dispersion Factors recalculated Offsite Attachment 3 Table 1.8.2-1 using guidance from RG 1.194 and RG 1.145.

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (EAB) 1.54E-4 m3/sec 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (LPZ) 1.5 E-5 m3/sec 2 - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (LPZ) 6.5 E-6 m3/sec

>12 hours (LPZ) 2.4 E-7 m3/sec Onsite N/A Attachment 3 Tables 1.8.1-2 Current- CR doses bounded by LOCA and 1.8.1-3 analysis.

Control Room Ventilation System N/A Attachment 3 Table 1.6.3-1 Current - CR doses bounded by LOCA Time of Control Room Isolation N/A 60 seconds analysis.

Control Room Unfiltered Inleakage N/A 115 cfm

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 27 of 39 Alternative Source Term and Conforming Amendment TABLE 5-5 (continued)

LOCKED ROTOR INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Breathing Rates Offisite 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.47 E-4 m 3/sec 3.5 E-4 m3/sec 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.75 E-4 m 3/sec 1.8 E-4 m3/sec After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months N/A 2.3 E-4 m3/sec Control Room Duration of the event N/A 3.5 E-4 m3/sec Control Room Occupancy Factor 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months N/A 100%

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months N/A 60%

96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months N/A 40%

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 28 of 39 Alternative Source Term and Conforming Amendment TABLE 5-6 RCCA EJECTION INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Core Power Level 2346 MWt 2652 MWt AST core power models planned EPU.

Core Average Fuel Burnup 40, 000 MWD/MTU 45,000 MWD/MTU AST core average burnup conservatively increased to accommodate a planned EPU.

Fuel Enrichment 3.0 - 4.5% U-235 3.0 - 5.0 % U-235 AST maximum fuel enrichment conservatively increased to accommodate a planned EPU.

Radial Peaking Factor 1.7 1.65 AST assumed radial peaking factor combined with the planned EPU core power value bounds the current radial peaking factor combined with the current power level.

Fuel Failure 10% DNB 10% DNB / 6.37% DNB AST -The case with lower fuel failure, Percent of Core with Centerline Melt 0.25% 0.25% / 0.16% secondary release and manual CR isolation was found to have limiting CR dose consequences.

Gap Release Fraction 0.10 RG 1.183 Appendix H Position I Core Fission Product Inventory See Table 5-1 values - Attachment 3 Table 1.74-1 The AST core fission product inventory bounds current and planned EPU conditions.

Initial RCS Equilibrium Activity 60 jiCi/gm DEI-131 0.25 [tCi/gm DEI-131 and AST values are consistent with proposed 447.7 piCi/gm DEXe-133 changes to more restrictive TS limits.

Initial Secondary Side Equilibrium Iodine Activity 0.1 gtCi/gm DEI- 131 0.1 jtCi/gm DEI-131

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 29 of 39 Alternative Source Term and Conforming Amendment TABLE 5-6 (continued)

RCCA EJECTION INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE [ REMARKS Release from DNB Fuel RG 1.183 Appendix H 1-131 6.24 E5 Section 1 1-132 .8.96 E5 1-133 1.27 E6 1-134 1.39 E6 1-135 1.19 E6 Kr-85m 1.64 E7 Kr-85 9.63 E5 Kr-87 3.15 E7 Kr-88 4.44 E7 Xe-131 m 7.29 E5 Xe-133m 3.99 E6 Xe-133 1.30 E8 Xe-135m 2.46 E7 Xe-135 3.63 E7 Xe-138 1.05 E8

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 30 of 39 Alternative Source Term and Conforming Amendment TABLE 5-6 (continued)

RCCA EJECTION INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS

_______________________________________________ j _______________________ ,. ________________________ _____________________________________

Release from Fuel Centerline Melt Fuel RG 1.183 Appendix H Section 1 1-131 7.72 E4 1-132 1.11 E5 1-133 1.58 E5 1-134 1.73 E5 1-135 1.48 E5 Kr-85m 4.10 E4 Kr-85 1.95 E3 Kr-87 7.88 E4 Kr-88 1.11 E5 Xe-131rm 1.72 E3 Xe-133m 9.82 E3 Xe-133 3.16 E5 Xe-135m 6.15 E4 Xe-135 9.05 E4 Xe-138 2.61 E5 Secondary Side Partition Coefficients SG (Flashed Leakage) N/A None Current - values are multipliers SG (Non-Flashed Leakage) N/A 100 AST - value is a divisor Iodine Partition Factor 0.01 N/A

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 31 of 39 Alternative Source Term and Conforming Amendment TABLE 5-6 (continued)

RCCA EJECTION INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Steam Generator Tube Leakage Rate 1.0 gpm/SG 0.2 gpm/SG AST values are consistent with proposed changes to lower TS limits.

Time to Establish Shutdown Cooling and Terminate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 63 hours7.291667e-4 days 0.0175 hours 1.041667e-4 weeks 2.39715e-5 months AST - Cooldown times conservatively Steam Release bound the current and planned EPU conditions.

Time to Re-Cover SG Tubes N/A 30 minutes Current - Dose consequences from tube Tube Uncovery Flashing Fraction N/A 11% uncovery are not part of the current licensing basis.

RCS Mass 366,086 Ibm 366,086 Ibm Current - Minimum RCS mass bounds the current and EPU conditions due to higher steam generator tube plugging assumption (20% vs. 10%).

SG Secondary Side Mass 84,128 Ibm per SG 67,707 Ibm per SG AST - The minimum mass for the intact SGs is assumed to maximize the release due to RCS iodine spiking and the value bounds current and planned EPU conditions..

Chemical Form of Iodine Released to Containment AST - Values are consistent with guidance Particulate 91% 95% given by RG 1.183, Appendix H.

Elemental 4% 4.85%

Organic 5% 0.15%

Chemical Form of Iodine Released to Steam Generators Particulate 0% 0%

Elemental 100% 97%

Organic 0% 3%

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 32 of 39 Alternative Source Term and Conforming Amendment TABLE 5-6 (continued)

RCCA EJECTION INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Atmospheric Dispersion Factors AST - All Dispersion Factors recalculated Offsite Attachment 3 Table 1.8.2-1 using guidance from RG 1.194 and RG 1.145.

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (EAB) 1.54 E-4 m3/sec 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (LPZ) 1.5 E-5 m3/sec 2 - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (LPZ) 6.5 E-6 m3/sec

>12 hours (LPZ) 2.4 E-7 m3/sec Onsite N/A Attachment 3 Tables 1.8.1-2 Current- CR doses not calculated - LOCA and 1.8.1-3 doses are bounding Control Room Ventilation System N/A Attachment 3 Table 1.6.3-1 Current - CR doses bounded by LOCA Time of Control Room Isolation N/A Containment 30 seconds analysis.

Release AST - Two secondary side release models were analyzed; one case relies on automatic Secondary 60 seconds Control Room isolation and the other case Release relies on manual Control Room isolation Secondary 30 minutes (see Enclosure 3, Section 2.6.4).

Release Control Room Unfiltered Inleakage N/A 115 cfm Breathing Rates Offsite 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.47 E-4 m3/sec 3.5 E-4 m3/sec 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.75 E-4 m3/sec 1.8 E-4 m3/sec After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.32 E-4 m3/sec 2.3 E-4 m3/sec Control Room Duration of the event N/A 3.5 E-4 m3/sec Current - CR doses not calculated

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 33 of 39 Alternative Source Term and Conforming Amendment TABLE 5-6 (continued)

RCCA EJECTION INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Control Room Occupancy Factor 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months N/A 100%

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months N/A 60%

96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months N/A 40%

Containment Volume 1.55 E+06 ft 3 1.55 E+06 ft3 Containment Leakage Rate 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.25% per day 0.20% (by weight)/day AST values are consistent with proposed After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.125% per day 0.10% (by weight)/day changes with more restrictive TS limits.

Containment Natural Deposition Coefficients Instantaneous Iodine Plateout at t-0 50% None Aerosols None 0.1 hr' Elemental Iodine 5.94/hour for DF < 100 5.58 hr-'

0/hour for DF > 100 Organic Iodine None None Surface Area for Natural Deposition 569,219 ft2 537,903 ft2 AST- Total containment surface area participating in surface deposition was reduced by 25% for natural deposition of elemental iodine.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 34 of 39 Alternative Source Term and Conforming Amendment TABLE 5-6 (continued)

RCCA EJECTION INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Emergency Containment Filter Efficiency Start Time Delay 300 seconds Not credited Number of Units 2 Flow Rate per Unit 33, 750 cfm Elemental 90%

Methyl 30%

Particulate 95%

Operating Time 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 35 of 39 Alternative Source Term and Conforming Amendment TABLE 5-7 WASTE GAS DECAY TANK (WDGT) RUPTURE INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Core Power Level 2346 MWt 2652 MWt AST core power models planned EPU.

WGDT Inventory RCS equilibrium with 1% RCS equilibrium with 1% AST - Curie content value conservatively fuel defects equal to fuel defects equal to bounds the TS limit of 70,000 curies Dose 55,000 curies Dose 84,274.8 curies Dose Equivalent XE-133 (See Enclosure 3, Equivalent Xe-133 Equivalent Xe-133 Section 2.7.3).

Tank Volume 525 ft3 525 ft3 Tank Leak Rate (Arbitrarily high) Instantaneous rupture I E6 cfm Atmospheric Dispersion Factors Osite Attachment 3 Table 1.8.2-1 AST - All Dispersion Factors recalculated 0 -2 hours (EAB) 1.54 E-4 m3/sec using guidance from RG 1.194 and RG 1.145.

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (LPZ) 1.5 E-5 m3/sec Onsite N/A Attachment 3 Tables 1.8.1-2 Current - CR doses not calculated - LOCA and 1.8.1-3 doses are bounding Control Room Ventilation System N/A Attachment 3 Table 1.6.3-1 Current- CR doses bounded by LOCA Time of Control Room Isolation N/A Not isolated analysis.

Control Room Unfiltered Inleakage N/A 115 cfm Control Room Makeup Flow N/A 1,000 cfm

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 36 of 39 Alternative Source Term and Conforming Amendment TABLE 5-7 (continued)

WASTE GAS DECAY TANK (WDGT) RUPTURE INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Breathing Rates Offsite 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.47 E-4 m 3/sec 3.5 E-4 m3/sec 3

8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.75 E-4 m /sec 1.8 E-4 m3 /sec After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.32 E-4 m 3/sec 2.3 E-4 m3/sec Control Room Duration of the event N/A 3.5 E-4 m3/sec Control Room Occupancy Factor 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months N/A 100%

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months N/A 60%

96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months N/A 40%

Turkey Point Units 3 and 4 L-2009-1 33 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 37 of 39 Alternative Source Term and Conforming Amendment TABLE 5-8 SPENT FUEL CASK DROP INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Core Power Level Before Shutdown 2346 MWt 2652 MWt AST core power models planned EPU.

Core Average Fuel Burnup 45,000 MWD/MTU 45,000 MWD/MTU AST core average burnup conservatively increased to accommodate a planned EPU.

Fuel Enrichment 3.0 - 4.5% U-235 3.0 - 5.0 % U-235 AST maximum fuel enrichment conservatively increased to accommodate a planned EPU.

Core Fission Product Inventory See Table 5-1 values Attachment 3 Table 1.7.4-1 AST assumed radial peaking factor combined with the planned EPU core power value bounds the current radial peaking factor combined with the current power level.

Gap Release Fractions AST - See Attachment 3, Section 2.8.2.

1-131 0.12 0.12 Kr-85 0.30 0.30 Other Noble Gases 0.10 0.10 Number of Fuel Assemblies Damaged 157 157 Delay Before Spent Fuel Movement 1,525 hours0.00608 days 0.146 hours 8.680556e-4 weeks 1.997625e-4 months 1,525 hours0.00608 days 0.146 hours 8.680556e-4 weeks 1.997625e-4 months Release Duration 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Water Level Above Damaged Fuel Assembly 23. feet minimum 23 feet minimum

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 38 of 39 Alternative Source Term and Conforming Amendment TABLE 5-8 (continued)

SPENT FUEL CASK DROP INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Iodine Decontamination Factors AST - Iodine decontamination Elemental 133 285 factors consistent with RG 1.183, Appendix B were Methyl 1 N/A adapted for this event.

Organic N/A I Noble Gas Decontamination Factor 1 I Chemical Form of Iodine in Pool AST - The guidance from RG Elemental 99.75% 99.85% 1.183, Appendix B was adapted for this event.

Methyl 0.25% 0%

Organic 0% 0.15%

Control Room Ventilation System N/A Attachment 3 Table 1.6.3-1 Current- CR doses not Time of Control Room Isolation N/A 30 minutes calculated - bounded by LOCA analysis.

Control Room Unfiltered Inleakage N/A 115 cfm

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 5 License Amendment Request 196 Page 39 of 39 Alternative Source Term and Conforming Amendment TABLE 5-8 (continued)

SPENT FUEL CASK DROP INPUTS AND ASSUMPTIONS INPUT/ASSUMPTION CURRENT VALUE AST VALUE REMARKS Chemical Form of Iodine in Pool AST-The guidance from RG 1.183, Appendix B was adapted for this event.

Atmospheric Dispersion Factors AST - All Dispersion Factors recalculated Offsite Attachment 3 Table 1.8.2-1 using guidance from RG 1.194 and RG 1.145.

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (EAB) 1.54E-4 m3/sec 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (LPZ) 1.5 E-5 m3/sec Control Room N/A Attachment 3 Tables 1.8.1-2 and 1.8.1-3 Breathing Rates Offsite 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.47 E-4 m3/sec 3.5 E-4 m3/sec 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months N/A 1.8 E-4 m3/sec After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.3 E-4 m3/sec Control Room N/A Duration of the event 3.5 E-4 m3/sec Control Room Occupancy Factor 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months N/A 100%

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months N/A 60%

96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months N/A 40%

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 6 License Amendment Request 196 Page 1 of 8 Alternative Source Term and Conforming Amendment LICENSE AMENDMENT REQUEST ALTERNATIVE SOURCE TERM AND CONFORMING AMENDMENT LAR 196 ATTACHMENT 6 REGULATORY INFORMATION

SUMMARY

(RIS) 2006-04 EXPERIENCE WITH IMPLEMENTATION OF ALTERNATIVE SOURCE TERMS COMPLIANCE TABLE

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 6 License Amendment Request 196 Page 2 of 8 Alternative Source Term and Conforming Amendment RIS 2006-04 ISSUE ADDRESSED BY

1. Level of Detail Contained in LARs An AST amendment request should describe the licensee's The LAR provides a summary of the changes, analyses of the radiological and non-radiological impacts including the justification for each change.

and provide a justification for the proposed modification in Attachment 3 provides the AST Licensing sufficient detail to support review by the NRC staff. For Technical Report, including specific details on example, the AST amendment request should the individual radiological analyses in Section 2.0, including AST analysis input (1) provide justification for each individual proposed values. Attachment 5 provides a comparison of change to the technical specifications (TS), current Licensing basis analysis input parameter values with those used for the AST analyses.

(2) identify and justify each change to the licensing basis ARCON96 meteorological data files are being accident analyses, and submitted on CDs under separate cover letter (3) contain enough details (e.g., assumptions, computer analyses input and output) to allow the NRC staff to confirm the dose analyses results in independent calculations.

2. Main Steam Isolation Valve (MSIV) Leakage and Fission Product Deposition in Piping For calculation of aerosol settling velocity in the main steam This item is applicable only to boiling water line (MSL) piping of boiling water reactors, some LARs reactors.

reference Accident Evaluation Report (AEB) 98-03, "Assessment of Radiological Consequences for the Perry PTN Units 3 and 4 are pressurized water Pilot Plant Application Using the Revised (NUREG-1465) reactors, therefore this item is not applicable to Source Term" (Ref. 2). This is acceptable. However, it is PTN.

important to note that the report was written based on the parameters of a particular plant and, therefore, the removal rate constant is specific to that plant. Any licensee who chooses to reference these AEB 98-03 assumptions should provide appropriate justification that the assumptions are applicable to their particular design.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-25 1 Attachment 6 License Amendment Request 196 Page 3 of 8 Alternative Source Term and Conforming Amendment RIS 2006-04 ISSUE ADDRESSED BY

3. Control Room Habitability When implementing an AST, some licensees have proposed The Control Room Emergency Ventilation that certain engineered safety features (ESF) ventilation System is credited in the control room dose systems not be credited as a mitigation feature in response to analyses. The system is classified safety an accident. In some cases, the licensee's revised design related. The fans, dampers, and filter are tested basis analysis introduced the assumption that normal in accordance with Technical Specifications (non-ESF) ventilation systems are operating during all or 3/4 7.5.

part of an accident scenario. Such an assumption is inappropriate unless the non-ESF system meets certain The control room ventilation outside air intakes qualities, attributes, and performance criteria as described in are being relocated to new locations that fall RG 1.183, Regulatory Positions 4.2.4 and 5.1.2. For within diverse wind sectors for post-accident example, credit for the operation of non-ESF ventilation contaminants. The relocated intakes will be systems should not be assumed unless they have a source of designed to seismic criteria, protected from emergency power. In addition, the operation of ventilation environmental effects, and will meet the systems establishes certain building or area pressures based requirements of 10 CFR 50 Appendix A, upon their flow rates. These pressures affect leakage and GDC 19. This will reduce the potential dose to infiltration rates which ultimately affect operator dose. the personnel in the control room following a Therefore, to credit the use of these systems, licensees design basis event.

should incorporate the systems into the ventilation filter Attachment 3 Section 1.6.3 describes the testing program in Section 5 of the TS. In summary, use of non-ESF ventilation systems during a DBA should not be operation of the control room ventilation system.

assumed unless the systems have emergency power and are part of the ventilation filter testing program in Section 5 of No credit is taken for the Auxiliary Building the TS. Ventilation or Fuel Handling Building Ventilation Systems.

Generic Letter (GL) 2003-01, "Control Room Habitability" (Ref. 5) requested licensees to confirm the ability of their The response to GL 2003-01 confirmed that the facility's control room to meet applicable habitability PTN control room envelope and ventilation regulatory requirements. In addition, licensees were systems meet all applicable habitability requested to confirm that control room habitability systems regulations and are designed, constructed, were designed, constructed, configured, operated and configured, operated, and maintained in maintained in accordance with the facility's design and accordance with the current licensing basis as licensing bases. described in the Turkey Point UFSAR.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 6 License Amendment Request 196 Page 4 of 8 Alternative Source Term and Conforming Amendment RIS 2006-04 ISSUE ADDRESSED BY

4. Atmospheric Dispersion Licensees may continue to use atmospheric relative Meteorological data over a five-year period concentration (X/Q) values and methodologies from their (2003 through 2007) was used in the existing licensing-basis analyses when appropriate. development of the new onsite X/Q factors used Licensees also have the option to adopt the generally less in the AST analyses.

conservative (more realistic) updated NRC staff guidance on The meteorological data was provided in annual determining X/Q values in support of design basis control room radiological habitability assessments provided in joint frequency distribution format for 2005 through 2007 for the offsite analysis.

RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Attachment 3 Section 1.8 describes the Nuclear Power Plants" (Ref. 6). Regulatory positions on methodology used to develop the PTN X/Q X/Q values for offsite (i.e., exclusion area boundary and low values.

population zone) accident radiological consequence assessments are provided in RG 1.145, "Atmospheric The analysis of the atmospheric dispersion Dispersion Models for Potential Accident Consequence coefficients used in the AST analyses was Assessments at Nuclear Power Plants" (Ref. 7). Based on performed following the guidance in RG 1.145 submittal reviews, the NRC staff identified the following and RG 1.194.

areas of improvement for licensee submittals that propose revision of the design basis atmospheric dispersion analyses for implementing AST. They should include the following information:

" A site plan showing true north and indicating locations of A site plan indicating true north and locations of all potential accident release pathways and control room releases and receptor points is included in intake and unfiltered inleakage pathways (whether Attachment 3 as Figure 1.8.1-1.

assumed or identified during inleakage testing).

Justification for using control room intake X/Q values for Justification for X/Q values for unfiltered modeling the unfiltered inleakage, if applicable. Control Room inleakage are provided in Section 4.1.3.

A copy of the meteorological data inputs and program Meteorological data and release-receptor outputs along with a discussion of assumptions and distances are being provided under a separate potential deviations from staff guidelines. Meteorological cover letter. Assumptions and atmospheric data input files should be checked to ensure quality (e.g., relative concentrations are included in compared against historical or other data and against the Attachment 3 Section 1.8. The electronic data raw data to ensure that the electronic file has been has been verified to be properly converted and properly formatted, any unit conversions are correct, and formatted.

invalid data are properly identified).

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 6 License Amendment Request 196 Page 5 of 8 Alternative Source Term and Conforming Amendment RIS 2006-04 ISSUE ADDRESSED BY

5. Modeling of ESF Leakage ESF systems that recirculate sump water outside the primary The Emergency Core Cooling System (ECCS) containment may leak during their intended operation. This leakage to the auxiliary building is 4,650 cc/hr release source includes leakage through valve packing based doubling the current licensing basis value glands, pump shaft seals, flanged connections, and other of 2,325 cc/hr. The leakage is assumed to start similar components. This release source may also include at 15 minutes into the event and continue leakage through valves isolating interfacing systems (e.g., throughout the 30-day period. This portion of refueling water storage tank). Appendix A to RG 1.183, the analysis assumes that 10% of the total iodine Regulatory Position 5, states that "the radiological is released from the leaked liquid. The form of consequences from the postulated [ESF] leakage should be the released iodine is 97% elemental and 3%

analyzed and combined with consequences postulated for organic.

other fission product release paths to determine the total calculated radiological consequences from the [loss-of- The ECCS backleakage to the RWST is coolant accident] LOCA." assumed to be 0.1 gph based upon doubling the expected total seat leakage through both sets of motor operated valves which isolate the recirculation flow from the RWST. The leakage is assumed to start at 15 minutes into the event and continue throughout the 30-day period. The fraction of the total iodine in the RWST which becomes elemental is a function of the RWST pH and the total iodine concentration. The amount of elemental iodine in the RWST fluid which then enters the RWST air space is a function of the temperature dependent iodine partition coefficient.

Attachment 3 Section 2.1.3 discusses the ECCS leakage assumptions.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 6 License Amendment Request 196 Page 6 of 8 Alternative Source Term and Conforming Amendment RIS 2006-04 ISSUE ADDRESSED BY

6. Release Pathways Changes to the plant configuration associated with an LAR The fuel handling accident analysis considers (e.g., an "open" containment during refueling) may require a both a dropped fuel assembly inside the re-analysis of the design basis dose calculations. A request containment with the equipment hatch open and for TS modifications allowing containment penetrations an assembly drop inside the fuel handling (i.e., personnel air lock, equipment hatch) to be open during building without credit for filtration of the fuel refueling cannot rely on the current dose analysis if this handling building exhaust (Attachment 3 analysis has not already considered these release pathways. Section 2.2.1).

RG 1.194, Regulatory Position 3.2.4.2 supports review of penetration pathways, by stating that "leakage is more likely The spent fuel cask drop analysis considers the to occur at a penetration, [and that the] analysts must drop of a cask into the spent fuel pool as consider the potential impact of leakage from building described in UFSAR Section 14.2.1.3.

penetrations exposed to the environment." Therefore, RG 1.183 does not provide any specific releases from personnel air locks and equipment hatches guidance for this event; therefore, the exposed to the environment and containment purge releases requirements of the fuel handling accident in prior to containment isolation need to be addressed. Appendix B of RG 1.183 are followed (Attachment 3 Section 2.8.1)

The atmospheric dispersion coefficients using the unique event release-receptor pathways are calculated for all AST analyzed events consistent with the guidance in RG 1.145 and RG 1.194.

7. Primary to Secondary Leakage Some analysis parameters can be affected by density The primary to secondary leakage rate assumed changes that occur in the process steam. The NRC staff 0.60 gpm total through all SGs and 0.20 gpm for continues to find errors in LAR submittals concerning the any one SG at room temperature conditions for modeling of primary to secondary leakage during a all analyzed events with secondary release postulated accident. This issue is discussed in Information pathways.

Notice (IN) 88-31, "Steam Generator Tube Rupture Analysis Deficiency," (Ref. 11) and Item 3.f in RIS 2001-19. An The density used in converting volumetric leak acceptable methodology for modeling this leakage is rates to mass leak rates is consistent with the provided in Appendix F to RG 1.183, Regulatory basis of surveillance tests used to show Position 5.2. compliance with the steam generator leak rate TS. The value used in the analyses is 62.4 lbm/ft3 (Attachment 3 Sections 2.3.2 for MSLB, 2.4.2 for SGTR, and 2.5.2 for Locked Rotor.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 6 License Amendment Request 196 Page 7 of 8 Alternative Source Term and Conforming Amendment RIS 2006-04 ISSUE ADDRESSED BY

8. Elemental Iodine Decontamination Factor (DF)

Appendix B to RG 1.183 provides assumptions for The iodine decontamination factors for the PTN evaluating the radiological consequences of a fuel handling fuel handling accident and spent fuel cask drop accident. If the water depth above the damaged fuel is event are Elemental - 285 and Organic - 1 23 feet or greater, Regulatory Position 2 states that "the (Attachment 3 Table 2.2-1).

decontamination factors for the elemental and organic

[iodine] species are 500 and 1, respectively, giving an overall effective decontamination factor of 200." However, an overall DF of 200 is achieved when the DF for elemental iodine is 285, not 500.

9. Isotopes Used in Dose Assessments For some accidents (e.g., main steam line break and rod Noble gas and cesium isotopes were considered drop), licensees have excluded noble gas and cesium in the dose assessments for all events except the isotopes from the dose assessment. The inclusion of these WGDT accident which included only noble isotopes should be addressed in the dose assessments for gases consistent with the guidance of BTP 11-5 AST implementation. of the SRP. See Attachment 3, Section 2.0.

10. Definition of Dose Equivalent 131 In the conversion to an AST, licensees have proposed a DOSE EQUIVALENT 1-131 shall be that modification to the TS definition of dose equivalent 1-13 1. concentration of 1-131 (microcuries per gram)

Some have modified the definition to base it upon the that alone would produce the same dose when thyroid dose conversion factors of International Commission inhaled as the combined activities of iodine on Radiation Protection (ICRP) Publication 2, "Report of isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 Committee II on Permissible Dose for Internal Radiation" actually present. The determination of DOSE (Ref. 12) or ICRP Publication 30, "Limits for Intakes of EQUIVALENT 1-1 31 shall be performed using Radionuclides by Workers" (Ref. 13). Others have proposed thyroid dose conversion factors from Table 2.1 a definition which is a combination of different iodine dose of EPA Federal Guidance Report No. 11 conversion factors, (e.g., RG 1.109, Revision 1, "Calculation (FGR 11), 1988, "Limiting Values of of Annual Doses to Man from Routine Releases of Reactor Radionuclide Intake and Air Concentration and Effluents for the Purpose of Evaluating Compliance with Dose Conversion Factors for Inhalation, 10 CFR [Part] 50, Appendix I" (Ref. 14), ICRP Publication Submersion, and Ingestions." The RCS dose 2, Federal Guidance Report 11, "Limiting Values of equivalent iodine source term has been Radionuclide Intake and Air Concentration and Dose calculated for the AST analyses consistent with Conversion Factors for Inhalation, Submersion, and the definition.

Ingestion" (Ref. 15). Although different references are available for dose conversion factors, the TS definition should be based on the same dose conversion factors that are used in the determination of the reactor coolant dose equivalent iodine curie content for the main steam line break and steam generator tube rupture accident analyses.

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 6 License Amendment Request 196 Page 8 of 8 Alternative Source Term and Conforming Amendment RIS 2006-04 ISSUE ADDRESSED BY

11. Acceptance Criteria for Off-Gas or Waste Gas System Release As part of full AST implementation, some licensees have The dose criteria for the PTN waste gas decay included an accident involving a release from their off-gas or tank rupture event are 100 mrem TEDE at both waste gas system. For this accident, they have proposed the Exclusion Area Boundary and the Low acceptance criteria of 500 millirem (mrem) total effective Population Zone (Attachment 3 Section 2.7.5).

dose equivalent (TEDE). The acceptance criterion for this event is that associated with the dose to an individual member of the public as described in 10 CFR Part 20, "Standards for Protection Against Radiation."

When the NRC revised 10 CFR Part 20 to incorporate a TEDE dose, the offsite dose to an individual member of the public was changed from 500 mrem whole body to 100 mrem TEDE. Therefore, any licensee who chooses to implement AST for an off-gas or waste gas system release should base its acceptance criteria on 100 mrem TEDE.

Licensees may also choose not to implement AST for this accident and continue with their existing analysis and acceptance criteria of 500 mrem whole body.

12. Containment Spray Mixing Some plants with mechanical means for mixing containment Containment spray provides coverage to air have assumed that the containment fans intake air solely 534,442 ft3 of the total 1,550,000 ft3 from a sprayed area and discharge it solely to an unsprayed containment volume. Therefore the PTN region or vice versa. Without additional analysis, test containment atmosphere is not considered a measurements or further justification, it should be assumed single, well-mixed volume. The containment is that the intake of air by containment ventilation systems is divided into three regions: upper sprayed, supplied proportionally to the sprayed and unsprayed unsprayed above the operating deck, and volumes in containment. unsprayed below the operating deck. The mixing rates between the regions are based on a sensitivity study evaluating various combinations of containment fans and sprays to produce conservative mixing rates. The final conservative mixing rates are 990,000 cfm between the upper sprayed and upper unsprayed containment regions and 375,000 cfm between the lower unsprayed region and the upper unsprayed region (Attachment 3 Section 2.1.4).

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 7 License Amendment Request 196 Page 1 of 2 Alternative Source Term and Conforming Amendment LICENSE AMENDMENT REQUEST ALTERNATIVE SOURCE TERM AND CONFORMING AMENDMENT LAR 196 ATTACHMENT 7 LIST OF MODIFICATIONS

Turkey Point Units 3 and 4 L-2009-133 Docket Nos. 50-250 and 50-251 Attachment 7 License Amendment Request 196 Page 2 of 2 Alternative Source Term and Conforming Amendment LIST OF MODIFICATIONS

1. The Control Room Ventilation System intakes are being relocated. The relocated intakes will be designed to seismic criteria, protected from environmental effects, and will meet the requirements of 10 CFR 50 Appendix A, GDC 19. The new intakes will be located near the ground level of the southeast and northeast corners of the auxiliary building and will fall within diverse wind sectors for post-accident contaminants. Post-modification testing in accordance with the plant design modification procedures will ensure the Technical Specification pressurization flow remains constant and demonstrate the integrity of the relocated intakes.

2. Stainless steel wire-mesh baskets (two large and eight small baskets) containing sodium tetraborate decahydrate are being installed in the containment basement.

3. The Normal Containment Cooling system aluminum cooling fins are being replaced with copper cooling fins. The copper fins will create less debris than the aluminum fins, and thus reduce the possibility of debris buildup in the containment sump.

4. The Emergency Containment Filtration System is being removed from TS.

v t e Letter Sequence Request
TAC:ME1624, Administratively Control Containment Penetrations, TSTF-68, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance, Acceptance, Acceptance, Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance Meeting, Meeting, Meeting, Meeting, Meeting, Meeting Questions 17 December 2009 25 February 2010 24 March 2010 23 June 2011 Answers 26 August 2009 10 February 2010 15 March 2010 14 April 2010 15 September 2010 14 December 2010 7 January 2011 22 February 2011 3 March 2011 31 March 2011 28 April 2011 25 March 2011 11 May 2011 29 April 2011 22 April 2011 20 May 2011 19 May 2011 7 July 2011 5 August 2011 29 July 2011 29 August 2011 19 August 2011 25 August 2011 17 January 2012 13 July 2012 21 June 2011 21 June 2011 Results Approval, Approval, Approval, Approval, Approval, Approval Other: L-2011-540, Safety Evaluation Report Review Supporting Fuel Storage Criticality Amendments 246 and 242, L-2012-118, Response to NRC ITSB RAI Regarding Clean TS Changes for EPU LAR No. 205 and Incorporation of Amendments 247 and 243 TS Changes on Relocation of Cycle Specific Parameters to the Core Operating Limits Report, L-2012-263, CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems in Light Water Nuclear Power Reactors - 30 Day Special Report, L-2012-293, FPL Review Comments on Extended Power Uprate Safety Evaluation Report, L-2012-426, Extended Power Uprate Cycle 26 Startup Report, ML110960521, ML110960523, ML112280292, ML12074A248, ML12074A251, ML12122A162, NRC-2011-0094, Federal Register Notice, Notice of Consideration Regarding Order Imposing Procedures for Document Access to Sensitive Unclassified Non-Safeguards Information.
MONTHYEAR2010-06-25 [Table View]

L-2009-133, License Amendment Request 196 Alternative Source Term and Conforming Amendment

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