L-2007-085, Alternative Source Term and Conforming Amendment

Alternative Source Term and Conforming Amendment

ML072000250
Person / Time
Site:	Saint Lucie
Issue date:	07/16/2007
From:	Johnston G Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2007-085
Download: ML072000250 (139)

Text

Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 0 July 16, 2007 FPL L-2007-085 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 RE: St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment Alternative Source Term and Conforming Amendment Pursuant to 10 CFR 50.90, Florida Power & Light Company (FPL) requests to amend Facility Operating License DPR-67 for St. Lucie Unit 1. FPL proposes to revise the St. Lucie Unit 1 licensing bases to adopt the alternative source term (AST) as allowed in 10 CFR 50.67.

Attachment 1 is a description of the proposed changes and the supporting justification including the Determination of No Significant Hazards and Environmental Considerations. Attachment 2 provides marked up copies of the proposed Technical Specification changes. Attachment 3 provides copies of the word processed TS pages. Attachment 4 provides information only copies of the marked up TS Bases pages. Enclosure 1 is Numerical Applications, Inc. (NAI) "AST Licensing Technical Report for St. Lucie Unit 1," NAI-1 101-043, Revision 2.

The proposed amendment has been reviewed in accordance with the FPL QATR requirements.

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

FPL submitted proposed license amendments via FPL letter L-2007-084 to implement TSTF-448, Rev. 3. This AST license amendment request needs to be approved by the NRC before or at the same time as the license amendment request for TSTF-448, Rev. 3. Therefore, FPL requests that the proposed license amendment be issued concurrent with the TSTF-448, Rev. 3. license amendments, with the amendments being implemented within 90 days.

Please contact Ken Frehafer at (772) 467-7748 if there are any questions about this submittal.

A00(

an FPL Group company .Lkeo-

L-2007-085 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on the / day of " 2007.

Sincerely, Gordon L. Johnston Site Vice President St. Lucie Plant 7! ý GLJ/KWF Attachments cc: Mr. William A. Passetti, Florida Department of Health

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 1 of 12 Alternative Source Term and Conforming Amendment Regulatory Assessment of the Proposed Implementation of the Alternative Radiological Source Term Methodology for the St. Lucie Plant, Unit No. 1 Introduction The current St. Lucie Plant, Unit No. 1, licensing basis for the radiological consequences analyses for accidents discussed in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR) is based on methodologies and assumptions that are primarily derived from Technical Information Document (TID) 14844.

Because of advances made in understanding the timing, magnitude, and chemical form of fission product releases from severe nuclear power plant accidents, 10 CFR 50.67 was issued to allow holders of operating licenses to voluntarily revise the traditional accident source term used in the design basis accident radiological consequence analyses with alternative source terms (ASTs).

Part 50.67 requires a licensee seeking to use an AST to apply for a license amendment and requires that the application contain an evaluation of the consequences of the affected design basis accidents. Regulatory guidance for the implementation of these ASTs is provided in Regulatory Guide (RG) 1.183.

As documented in NEI 99-03, several nuclear plants performed testing on control room unfiltered air inleakage that demonstrated leakage rates in excess of amounts assumed in the accident analyses. The AST methodology as established in RG 1.183 is being used to calculate the offsite and control room radiological consequences for St. Lucie Unit No. 1 to support the control room habitability program by addressing the radiological impact of potential increases in control room unfiltered air inleakage.

The following limiting UFSAR Chapter 15 accidents are analyzed:

0 Loss-of-Coolant Accident (LOCA)

. Fuel Handling Accident (FHA) 0 Main Steam Line Break (MSLB)

• Steam Generator Tube Rupture (SGTR)

• Reactor Coolant Pump Shaft Seizure (Locked Rotor)

" Control Element Assembly (CEA) Ejection

• Inadvertent Opening of a Main Steam Safety Valve (IOMSSV)

Each accident and the specific input assumptions are described in the Numerical Applications, Inc. (NAI) "AST Licensing Technical Report for St. Lucie Unit 1," NAI-110 1-043, Revision 2 (Enclosure 1). These analyses provide for a bounding allowable control room unfiltered air inleakage of 500 cfm. The use of 500 cfm as a design basis value is expected to be above the unfiltered inleakage value to be determined through testing or analysis consistent with the resolution of issues identified in NEI 99-03 and Generic Letter 2003-01.

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 2 of 12 Alternative Source Term and Conforming Amendment Description of Proposed Amendment Florida Power and Light (FPL),Company proposes to revise the St. Lucie Plant, Unit No. 1, licensing basis to implement the AST as allowed by 10 CFR 50.67 and described in RG 1.183, through reanalysis of the radiological consequences of the UFSAR Chapter 15 accidents listed above. As part of the full implementation of this AST, the following changes are assumed in the analysis:

• The total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11.

" 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.

• Increased values for control room unfiltered air inleakage are assumed.

• A steam generator tube leakage rate (for accident analysis) that exceeds the current Technical Specification operational leakage limit is utilized.

An SBVS bypass leakage value that is more restrictive than the current Technical Specification limit is utilized.

Accordingly, the following changes to the St. Lucie, Unit No. 1, Technical Specifications (TS) are proposed:

" The definition of Dose Equivalent 1-131 in Section 1:10 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,"

as the source of thyroid dose conversion factors.

• Tables 3.3-6 and 4.4-3 are revised to add the Control Room Isolation Area Radiation Monitors to the applicability of TS 3/4.3.3.1, Radiation Monitoring.

Surveillance Requirement 4.6.6.1 is revised to relocate the HEPA filter, charcoal adsorber, flow rate, and heater surveillance test acceptance criteria for the Shield Building Ventilation System to the Ventilation Filter Testing Program in TS Section 6.8.4.k.

• Surveillance Requirement 4.7.8 is revised to relocate the HEPA filter, charcoal adsorber, and flow rate surveillance test acceptance criteria for the ECCS Area Ventilation System to the Ventilation Filter Testing Program in TS Section 6.8.4.k.

The leakage rate acceptance criterion for secondary containment bypass leakage paths (i.e. Shield Building Bypass Leakage) stated in TS 6.8.4.h, "Containment Leakage Rate Testing Program," is reduced from 27% to 9.6%.

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 3 of 12 Alternative Source Term and Conforming Amendment

" The HEPA filter and charcoal adsorber test acceptance criteria for the ECCS Area and Shield Building Ventilation. Systems relocated to the VFTP are revised as follows:

o The filter efficiency test acceptance criteria are increased from 99% to 99.95%

o The inplace charcoal adsorber efficiency test acceptance criteria are increased from 99% to 99.95%

" In the VFTP (TS 6.8.4.k), reference to Regulatory Guide 1.52, Revision 2 is replaced with reference to Regulatory Guide 1.52, Revision 3.

" In the Surveillance Requirements for the ECCS Area Ventilation System and the Shield Building Ventilation System (SRs 4.7.8.1 and 4.6.6.1, respectively), as well as in the VFTP (TS 6.8.4.k), reference to ANSI N510-1975 is replaced with reference to ASME N510-1989.

" The accident induced leakage performance criteria of the Steam Generator (SG) Program described in TS Section 6.8.4.1 is changed from 1.0 gpm total through all SGs and 0.5 gpm through any one SG, to 0.5 gpm total through all SGs and 0.25 gpm through any one SG.

Justification of Proposed Technical Specification Changes Revision of the definition of Dose Equivalent 1-131 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," as the source of thyroid dose conversion factors is consistent with the guidance provided in RG 1.183. In the dose calculations, the dose conversion factors referenced in the definition of Dose Equivalent 1-131 are used to adjust the initial primary coolant iodine activities for use in the dose calculations. Use of thyroid dose conversion factors (versus effective dose conversion factors for inhalation or CEDE doses) results in slightly more conservative total iodine concentrations in the primary coolant and, therefore, slightly higher doses.

As described in the Enclosure 1 analyses, automatically actuated, redundant isolation valves are provided at each control room outside air intake and exhaust air path so that the control room envelope is isolated on receipt of an outside air intake high radiation signal. Addition of the Control Room Isolation Area Radiation Monitors to the applicability of TS 3/4.3.3.1 will provide the requisite controls to support crediting the high radiation automatic isolation function of the radiation monitors in the AST analyses.

The proposed alarm setpoint of< 2 times background is low enough to provide the required sensitivity, but high enough to avoid nuisance alarms. The proposed action statement ensures that, with less than the minimum number of channels operable, the isolation function is maintained by requiring that the emergency ventilation system be placed in the recirculation mode of operation. The proposed Surveillance Requirements

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 4 of 12 Alternative Source Term and Conforming Amendment are consistent with those for the radiation monitoring instrumentation currently included in the scope of the SR.

Relocation of the HEPA filter, charcoal adsorber, flow rate, and heater surveillance test acceptance criteria for the Shield Building Ventilation System and the ECCS Area Ventilation System, as applicable, to the Ventilation Filter Testing Program (VFTP) provides consistency with the existing format for the Control Room Emergency Ventilation System filter testing requirements which is modeled after the format of the VFTP in NUREG-1432, Standard Technical Specifications Combustion Engineering Plants.

With one exception discussed below, the current ventilation filter testing requirements are included in the proposed VFTP. As revised, the filter train operational-type surveillance tests remain in the LCO/Surveillance section, while the direction for the post maintenance or preventative maintenance tests are stated to be in accordance with the VFTP. The VFTP includes the applicable TS surveillance limits.

The testing methodology requirements are met by requiring that the tests be performed in accordance with ASME N510-1989 and ASTM D3803-1989, as applicable. The frequency requirements are met by describing the VFTP as a program that tests "at the frequencies specified in Regulatory Guide 1.52, Revision 3." These testing frequencies specify off-normal as well as normal (i.e., "scheduled") testing and align with the current TS testing frequencies. Off-normal testing requirements include HEPA and charcoal adsorber leak testing following filter or cell replacements, following filter train contact with foreign fumes and following train maintenance activities.

The proposed change to delete SRs 4.6.6.1 .b.4 and 4.7.8.1.b.4 is an administrative deletion only as these requirements are delineated in the RG 1.52 inplace HEPA and charcoal adsorber testing requirements.

The proposed change deletes SRs 4.6.6.1.c.[3].b and 4.7.8.1.c.[3].b for the Shield Building Ventilation System and the ECCS Area Ventilation System, respectively. These SRs require DOP testing of the HEPA filter banks subsequent to reinstalling the charcoal adsorber tray used for obtaining a carbon sample. There is no equivalent SR in RG 1.52, Revision 3. The HEPA filters are not affected by carbon sampling or subsequent reinstallation of the adsorber trays. As such, the SRs are not necessary to ensure operability of the HEPA filters.

Reduction of the acceptance criterion for secondary containment bypass leakage paths (i.e. Shield Building Bypass Leakage) from 27% to 9.6% improves (i.e., increases) the allowable control room unfiltered inleakage. The 9.6% value is supported by plant leakage test results.

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment I Proposed License Amendment Page 5 of 12 Alternative Source Term and Conforming Amendment The revised test acceptance criteria for the HEPA filters and the charcoal adsorbers in the ECCS Area and Shield Building Ventilation Systems ensure that the filters and adsorbers meet the filtration efficiencies assumed in the Enclosure 1 accident analyses. As described in RG 1.52, Revision 3, the efficiency assumptions allowed are dependent on the test acceptance criteria. The revised test acceptance criteria are consistent with the criteria provided in RG 1.52, Revision 3 and support the assumptions of the Enclosure 1 accident analyses.

" Replacing reference to Regulatory Guide 1.52, Revision 2 with reference to Regulatory Guide 1.52, Revision 3 reflects the adoption of the requirements of the most current revision of RG 1.52. As described above, the VFTP testing requirements are consistent with the requirements of Revision 3 of RG 1.52 and support the assumptions of the Attachment 3 accident analyses.

" Replacing reference to ANSI N510-1975 with reference to ASME N510-1989 is consistent with RG 1.52, Revision 3. Revision 3 of RG 1.52 states that engineered safety features (ESF) atmosphere cleanup systems tested to ASME N510-1989 (or its earlier v ersions) are considered adequate to protect public health and safety.

" Changing the accident induced leakage performance criteria of the Steam Generator (SG)

Program from 1.0 gpm total through all SGs and 0.5 gpm through any one SG, to 0.5 gpm total through all SGs and 0.25 gpm through any one SG, continues to maintain margin to the operational leakage limit specified in the Technical Specifications. TSTF-449, Steam Generator Tube Integrity, changed the SG tube leakage Technical Specification limit to 150 gpd per SG which is roughly equivalent to 0.1 gpm. The limit of 0.25 gpm per SG was chosen to provide additional margin above the 0.1 gpm TS limit.

The limit of 0.5 gpm total through all SGs is consistent with the limit of 0.25 gpm per SG and reflects the maximum total allowable leakage.

Accident Source Term The full core isotopic inventory for St. Lucie Unit 1 is determined in accordance with RG 1.183.

The inventory of fission products in the core and coolant systems that is 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. Event-specific isotopic source terms are developed using a bounding approach. The maximum core power of 2754 MWth is calculated as the current licensed rated thermal power of 2700 MWth plus the ECCS evaluation uncertainty of 2%. 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.

The core inventory release fractions for the gap release and early in-vessel damage phases for the design basis LOCAs utilized those release fractions provided inRG 1.183, Regulatory Position 3.2, Table 2, "PWR Core Inventory Fraction Released into Containment." For non-LOCA

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 6 of 12 Alternative Source Term and Conforming Amendment events, the fractions of the core inventory assumed to be in the gap are consistent with RG 1.183, Regulatory Position 3.2, Table 3, "Non-LOCA Fraction of Fission Product Inventory in Gap." In some cases, the gap fractions listed in Table 3 are modified as required by the event-specific source term requirements listed in the Appendices for RG 1.183.

The nominal primary coolant activity is based on 1% failed fuel. The iodine activities are adjusted to achieve the Technical Specification limit of 1.0 jtCi/gm dose equivalent 1-131 using the Technical Specification definition of Dose Equivalent 1-131 (DE 1-131) and dose conversion factors for individual isotopes from ICRP 30 (which are equivalent to the rounded values from FGR No. 11 for iodine isotopes). The remaining (non-iodine) isotopes are adjusted to achieve the Technical Specification limit of 1 00/E-bar microcuries per gram of gross activity.

Secondary coolant system activity is limited to .a value of < 0.10 VLCi/gm dose equivalent 1-131 in accordance with the Technical Specifications. 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 jiCi/gm. Thus, the secondary side iodine activity-.

is 1/10 of the primary coolant activity.

The fuel handling accident for St. Lucie Unit 1 assumes the failure of one assembly; therefore, the fuel handling accident source term is based on a single "bounding" fuel assembly.

Sensitivity studies were performed to assess the bounding fuel enrichment and bounding burnup values. The assembly source term is based on 102% of rated power (2754 MWth). For each nuclide, the bounding activity for the allowable range of enrichments and discharge exposure is determined.

The AST Licensing Technical Report for St. Lucie Unit 1 (NAI-110 1-043) provides the details of the LOCA and non-LOCA accident analyses performed according to the guidelines set forth in RG 1.183.

Dose Calculation The St. Lucie Unit No. 1 dose calculations using the AST methodology apply TEDE acceptance criteria. Dose calculations follow the guidelines of Regulatory Positions cited in RG 1.183.

Analyses consider the radionuclides listed in Table 5 of RG 1.183 and assume that fission products are released to containment in particulate form, except for elemental iodine, organic iodine, and noble gases. Radioiodine fractions released to containment in *apostulated accident are assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodine, including both gap releases and fuel pellet releases. In specific instances, transport models may affect radioiodine fractions.

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 7 of 12 Alternative Source Term and Conforming Amendment Assumptions and .Methodologies The AST analyses performed for St, Lucie Unit No. 1 .use assumptions and models defined in RG 1.183 to provide appropriate and prudent safety margins.

Except as otherwise stated, credit is taken for ESF and other appropriately qualified, accident mitigation features. Selected numeric input values are conservative to assure a conservative calculated dose. Except as otherwise required by regulatory guidance, analyses use current licensing basis values.

Meteorological data collected per the St. Lucie Unit 1 meteorological monitoring program described in the UFSAR is used in generating the accident atmospheric dispersion (X/Q) factors.

Dose Consequences Results Full implementation of the Alternative Source Term methodology, as defined in Regulatory Guide 1.183, into the design basis accident analysis is made to support control room habitability with increased 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), Reactor Coolant Pump Shaft Seizure (Locked Rotor), Control Element Assembly (CEA) Ejection, and Inadvertent Opening of a Main Steam Safety Valve (IOMSSV) are made using the RG 1.183 methodology. The analyses used assumptions consistent with proposed changes in the St. Lucie Unit No. 1 licensing basis and the calculated doses do not exceed the defined acceptance criteria.

Results of the St. Lucie Unit 1 radiological consequence analyses using the AST methodology and the corresponding allowable control room unfiltered inleakage are summarized in Table 1.

The analyses support a maximum allowable control room unfiltered air inleakage of 500 cfm.

NAI-1 101-043, "AST Licensing Technical Report for St. Lucie Unit 1," explains these results and acceptance criteria in more detail.

No Significant Hazards Determination The Commission has provided standards in 10CFR50.92(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 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 the margin of safety. FPL has reviewed this proposed license amendment for FPL's St. Lucie Unit No. 1 and determined that its adaptation would not involve a significant hazards determination. The bases for this determination are:

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 8 of 12 Alternative Source Term and Conforming Amendment This proposed change 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.

Alternative source term calculations have been performed for St. Lucie Unit No. 1 which demonstrate that the dose consequences remain below limits specified in NRC Regulatory Guide 1.183 and 10 CFR 50.67. The proposed changes do not modify the design or operation of the plant. The use of the AST only 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, demonstrate that the dose consequences of these limiting events are all within the regulatory limits.

With the exception of the deletion of SRs 4.6.6.1.c.[3].b and 4.7.8.1.c.[3].b, the proposed Technical Specification changes are consistent with, or more restrictive than, the current TS requirements. The proposed filter testing requirements continue to ensure that the associated filtration systems function as described in the UFSAR and as assumed in the accident analyses. None of the affected systems, components or programs are related to accident initiators.

Therefore, the proposed change does 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 change does not affect any plant structures, systems, or components. The operation of plant systems and equipment will not be affected by this proposed change.

Neither implementation of the alternative source term methodology, establishing more restrictive TS requirements, nor deleting SRs 4.6.6.1.c. [3].b and 4.7.8.1.c.[3].b have the capability to introduce any new failure mechanisms or cause any analyzed accident to progress in a different manner.

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

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 9 of 12 Alternative Source Term and Conforming Amendment

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

The proposed implementation of the alternative source term methodology is consistent with NRC Regulatory Guide 1.183. With the exception of the deletion of SRs 4.6.6.1.c.[3].b and 4.7.8.1 .c.[3].b, the proposed Technical Specification changes are consistent with, or more restrictive than, the current TS requirements. The proposed TS requirements support the AST revisions to the limiting design basis accidents. The proposed filter testing requirements continue to ensure that the associated filtration systems functionas described in the UFSAR and as assumed in the accident analyses. 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 use of the alternative source term methodology.

The proposed changes continue toensure 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 consequences of these accidents is considered to be that provided by meeting the applicable regulatory limits, which are 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, FP&L has determined that the proposed change does not involve a significant hazards consideration.

Environmental Considerations 10 CFR 51.22(c)(9) provides criterion 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 would 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, or (3) resulf in a significant increase in individual or cumulative occupational radiation exposure. FPL has reviewed this proposed license amendment request 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:

This change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 10 of 12 Alternative Source Term and Conforming Amendment

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 anynew 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 purely analytical and does not result in any physical plant changes or new surveillance that would significantly increase the .cumulative occupational radiation exposure. Therefore, the proposed amendment has no significant affect on either individual or cumulative occupational radiation exposure.

References

1. St. Lucie Unit No. 1 Updated Final Safety Analysis Report, through Amendment 21.

2. TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 23, 1.962.

3. Code of FederalRegulations, 10CFR50.67, "Accident Source Term," revised 12/03/02.

4. USNRC, Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants," July 2000.

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

6. NAI-110 1-043, "AST Licensing Technical Report for St. Lucie Unit 1," Revision 2, Numerical Applications, Inc., May 2007.

7. Code of Federal Regulations, 10CFR100. 11, "Determination of exclusion area, low population zone, and population distance center."

St. Lucie Unit 1 L-2007-085 Docket No. 50-335. Attachment 1 Proposed License Amendment Page 11 of 12 Alternative Source Term and Conforming Amendment

8. 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.

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

10. Florida Power & Light Company, St. Lucie Unit No. 1 Technical Specifications (through Amendment 200).

11. Code of Federal Regulations, 10CFR50.92, "Issuance of Amendment."

12. Code of Federal Regulations, 10CFR51.22, "Criterion for Categorical Exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review."

13. NUREG-1432, Standard Technical Specifications Combustion Engineering Plants, Volume 1, Rev. 3.0

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 12 of 12 Alternative Source Term and Conforming Amendment TABLE 1 St. Lucie Plant, Unit No. 1 Summary of Alternative Source Term Analysis Results Allowable ( Control Room 1 2 Case, Unfiltered CR EAB Dose( " LPZ Doset ) Dose(2)

Inleakage (rem TEDE) (rem TEDE) (rem TEDE)

" (cfm)

LOCA .500 1.08 2.53 4.69 MSLB - Outside of Containment 500 0.33 0.90 4.80 (1.8% DNB)

MSLB - Outside of Containment 500 0.36 0.97 4.97 (0.43% FCM)

MSLB - Inside of Containment 500 0.52 1.04 4.92 (29% DNB)

MSLB - Inside of Containment 500 0.76 1.43 4.91 (6.1% FCM)

SGTR Pre-accident Iodine Spike 500 0.31 0.30 3.03 Acceptance Criteria _<25(3) 25(3) < 5(4)

SGTRConcurrent Iodine Spike 500 0.08 0.08 0.60 Locked Rotor (13.7 % DNB) 500 0.25 0.54 2.53 IOMSSV 500 0.02 0.02 0.30 Acceptance Criteria <2.5 3 2.5 (3) < 5 (4)

FHA - Containment 500 0.53 0.52 1.23 FHA - Fuel Handling Building 500 0.53 0.52 3.02 CEA Ejection - Containment 500 0.26 0.50. 2.74 Release (9.5 % DNB, 0.5 % FCM) 5 CEA Ejection - Secondary Side Release (9.5 % DNB, 0.5 % FCM)

Acceptance Criteria 6.3 (3

_< < 6.3(3) < 5(3)

(')Worst 2-hour dose (2) Integrated 30-day dose (3)RG 1.183, Table 6 (4) 10CFR50.67

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 1 of 15 Alternative Source Term and Conforming Amendment Technical Specification Change Mark Ups Page 1-3 Page 3/4 3-22 Page 3/4 3-23 Page 3/4 3-24 Page 3/4 6-27 Page 3/4 6-28 Page 3/4 6-29 Page 3/4 7-24 Page 3/4 7-25 Page 3/4 7-26 Page 6-15b Page 6-15d Page 6-15e

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 2 of 15 Alternative Source Term and Conforming Amendment DEFINITIONS DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (jiCi/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 'CRP 39, Supplm.nnt to Pa-t 1, pagor. 102-212, Tabler. entitlod, "CoMmtoed_ Dcos Eguivalcnt In Target Organe. OF TiYe'ue par Itake Of Unit Ai'iy (SBV!P.q)."

- AVERAGE DISINTEGRATION ENERGY 1.11 Eshall be the average (weighted in prop rtion to the concentration of each radionuclide in the reactor coolant at the time of sam ling) of the sum of the average beta and gamma energies per disintegration (in EV) for isotopes, other than iodines, with half lives greater than 15 minutes, making p at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RES ONSE TIME 1.12 The ENGINEERED SAFETY FE U"RES REPONSE TIME shall be that time interval from when the monitored parame er exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment s capable of performing its safety function (i.e., the valves travel to their required p sitions, pump discharge pressures reach their required values, etc.). Times shall inclde diesel generator starting and sequence loading delays where applicable. The respo se time may be measured by means of any series of sequential, overlapping, or t al steps so that the entire response time is measured.

In lieu of measurement, res onse time may be verified for selected components provided that the compone ts and methodology for verification have been previously reviewed and approved b the NRC.

FREQUENCY NOTATION 1.13 The FREQUENCY NO ATION specified for the performance of Surveillance Requirements shall rrespond to the intervals defined in Table 1.1.

GASEOUS RADWASTE EATMENT SYSTEM 1.14 *AGASEOUS RA ASTE TREATMENT SYSTEM is any system designed and installed to reduce radioa ive gaseous effluents by collecting primary coolant system offgases from the primary ystem and providing for delay or holdup for the purpose of reducing the total radioa ivity prior to release to the environment.

ST. LUCIE - UNIT 1 1-3 Amendment No. 27, 69, 163,195

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 3 of 15 Alternative Source Term and Conforming Amendment 0

v~ CD !) :11! :tj C.~4 C*~I Z4 I--

z w E E E M.w CL CL C-O W z 0 0.

) CO (0- 0 (0 0 (0-0 0 0 -0 0 0 WU)0 M. 068 nI 0ý 0 0 Z EE

-fi v-(I 0 E 00 C

- 0 (U I

<V 02 (U 0 M

0.00.-

U, 0

5 0 a- I-C z 0 0 E E

ci) 0 E

C-)

Cd 0

C"i

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 4 of 15 Alternative Source Term and Conforming Amendment TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 12 - Wrth the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.12.

ACTION 13 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 14 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 15 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,;or:

1) Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and

2) Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 16 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, comply with the ACTION requirements of Specification 3.9.9.

ACTION 17 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation.

ST. LUCIE - UNIT 1 3/4 3-23 Amendment No. 59

St. Lucie Unit I L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 5 of 15 Alternative Source Term and Conforming Amendment Izl w wii Ca.. Ce)

CO..

uJ z 0) z0D- 6 z

Iz0 z

0n WF 04 z < wn(Y- w o Z)

C')

C..)

-j C..)

-j C)

ZW CO 0()to CO Ci) 4 3 0)

.20

-C 0

.00.

.0 0) 75 E w 0 0 -j0 8.0 00 0 A . >. 0.0cL 0( 9 0 ýa I-Z0 LooL . ~ ~-C

- r z z

LM0 (,)

ei

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 6 of 15 Alternative Source Term and Conforming Amendment CONTAINMENT SYSTEMS 314.6.6 SECONDARY CONTAINMENT SHIELD BUILDING VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.1 Two independent shield building ventilation systems shall be OPERABLE.

APPLICABILITY: MODES 1, 21 3 and 4.

ACTION:

With one shield building ventilation system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.6.1 Each shield building ventilation system shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

b. , At-4east-ene -pe-8-meriths-er-(o+after ny-str-uturakmain-ILL I IP*A *E ..... L ..... I -- J------L.

ST. LUCIE - UNIT 1 *3/4 6-27 Amendment No. 27

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 7 of 15 Alternative Source Term and Conforming Amendment CONTAINMENT SYSTEMS

'SURVEILLANCE REQUIREMENTS (Continued)

a. Verifying that a labercter; analysis of a earbon sample demonstrates methyl ,+AA ..aom , .,..-.*aI efficiene of pW7.5% , . fOr, redloa tyiodide when s-ample -i emd In .... da..e wil..

AST-M 9380-3 1980 (300CG, 70% RH). The earbon samples not obtained frem test eandstero sh;ll be prepared by eithor:

a-)

. . :--adoo

..... b...

.. ... rr.---

.... .. :rbe a Emp ,, one e.

,ty"ing ............. .. ...

tray, mixing the adorbont thwroughly, and ebtlo*inig samples at least tWo inehes in diamoeter and with a length equal to the thitn "~s 4f;h" c.o tray, mixing t.he adsorbe-nttheroughly, and obtaining samples at least t4m inehes in diameter and vith a le..ngth eqal t the thi....ss of the bed.

4r Ve-fi a lorate-of 80100- VON%diuring sysem pertin -whentooted in AAAArdcno Mith ANSI After evefv;FRE) hours of svstem eBefafien bv either!

4-7 Vefifying that a laberatory analysis of a earban samploe obtained from a de~ntaie Sotcnso a prA ý1Meva the sample isatested in accordance with ASTM 03803 198-9

~C~-70-RH)~or Verifying that a laboratory analysis of at least twe carbon samples demonstrate a removal effiefieney of t 97.5% for radie witt-ASTM--8390(0C 0 ~am~es-~e-me-pared by either-8* t:Fnptyng one enioaata ooaasro samples at least two inches in diamoeter and Mih et length equal to the thioicness of the bed, er Emptying a longitudinal semple Ifrom an adserber tiny, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and wvith a iength eq~ual to,the. thiel~e. o..sf the bed.

ST. LUCIE - UNIT 1, 3/4 6-28 Amendment No. 107

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 8 of 15 Alternative Source Term and Conforming Amendment CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (ontinuedd Subsequent to reinstalling the adsorbcr tryudfr obtaining the ea - -ampl th mon straftd OPERALE by aAao a), Vcrifying that the chafee.. adorbcr. r.m... > 99% ef a halogenated hydrOc.rbon rcfrigcront toot g wiin wt they arc toeted in plas in aoFrdanco ,vith ANSI N51100975whillcpzt tentva~tion system-at a fleMO rate- tf6000 sFm 10%, and Verifying that the HEPA filter banks remoye > 99%lf the DOPhven they arc toestdpa . int eordl.no with ANSI N510 175 1ilo .poratingthe ventilation sysotem at a flow fate of M000 efm i 10%.

At least once per 18 months by:

1. Verifyingta h pr+ uo ro acos th*obndH fitero RAnd ehalre- a-1ad ASarbo r baln It-AS -. 6 inohe tOterGauge ,hilcoporatng thc v".,:"t ..on system at a flow rate of6000ofM+/- 10%. wj

2. Ve rifyin g -that th e air flow distjrWo~n- is u~ni'fo rm within 20% across HEPA filte* charcoal adsorbers when tested in accordance with ANt N510 3.

\ Verifying that the filtration system starts automatically on a Containment Isolation Signal (CIS).

4.

\ Verifying that the filter cooling makeup air and cross con-nection valves can be manually opened.

-- Verifying that each system produces a negative pressure of

> 2.0 inches W.G. in the annulus within 2 minutes after a Containment Isolation Signal (CIS).

6 Verifying h that th. m.in heater.s dis ipatO .. +/- 3 w. and th.

• aiicrheaters dissipate 1.5 +/- 0.25 kw vA~N toted in aocordarnoo Yith ANSI N640 1976.

eý7 A&ftoraoh oonplete Or Partial roplaeement of a HE:PA filtr bank bY veriying that the HEPA filter banks remove ;, 9% of the 0O

• ~ ho~~ r ettested in place in aGOcdancc With ANSI N51 0 1 975 poraingthofiltration -y-ecm at a floW ratO Of 6000

• whlo f- AftF meah oOMplete Or partial replaeemont of a chorcoal adsoober

• bank by verifying that the chareeal adsorbers rmove _ 00%of a +.

• halogenated hydrOoorbcn refrigerant test gas 1he th. arc tlotc in place in accordance Mih ANSI N51 0 1975 %Ohilo OPOrating-tho filtrationsstematle lw-rte-of-600"-m 1016.

ST, LUCIE UNIT 1 .31406-29 Amendment No ),27

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 9 of 15 Alternative Source Term and Conforming Amendment PLANT SYSTEMS 3/4.7.8 ECCS AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8.1 Two independent ECCS area exhaust air filter trains shall be OPERABLE.

APPLICABIUTY: MODES 1, 2, 3 and 4.

ACTION:

With one ECCS area exhaust air filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.8.1 Each ECCS area exhaust air filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiat-ing, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the train operates

-for at least 15 minutes.

b. At-least-- *,G .Per-4&ff*thS OF (1) afteF anV-StrUtral I'fRafntcnan~c on thc HEPA~filtef~er~ehasreeal ads~rbzr hcusinas.

Veiyr;tht th~ hr~lo~resrr~o_9%~

halogenated hydocarbon refrigerant test ga= w.hcn they are--tested-in-plaeer-o4coerdance-with-ANSi--544.04D7-5

, hile-epeati,.g the ventilation system at a flow rate-of Fa-30,000 + f 1%.

system filter testing in accordance with the Ventilation Filter Testing Program.

ST. LUCIE - UNIT 1 3/4 7-24

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 10 of 15 Alternative Source Term and Conforming Amendment PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (continued) verifyfrng ma~t a lanerater; anapysis of earazn sanm" ac-ns~eharceovdIfm n fth hroali-odsm b ers-demonsratestca remviroa efflieneyc of 9:7.5% for radic aetive melthyI iodide whe~n the sample ia tested irt

~rdz~wii;AGSTM D3800130 l9%3G2-,7G%0%R)i. T csrbsn mpL. fin bAiAzc ht ffz m test es nstcr shall be a*Emptpng one centire bed frem a Femeed- adzsreber rs. .. i.ing the adSOrborfl tho~roughly, and obtaining samples at least tm inehe i diametef and with-a Isrngth equal to the thietnessc of the bed, ar b) Emnptying a longitudinal sample from or aaýbe tray. mixing the adasorben~t thoroughly, and obtaining 1cigth equ.al to the thickness of the bed.

4 Ve Fifying a systefm flow rate ef 30,'000 em! 10% during systezm operation Mein tested in M--F-.scrdaiR-ee witl9#-AN61NS41&1--

i 97-5 0-1 After evf 720 hzx.rz of sys.t...,. epfll by e.,the...

vecrtyAng Mfat a iewmwyatr naiysis Of 8 csrbwn sacmp' iserncy

'f of +/- 7.5% fo radoaeie-ehyl-oddeWtthen thsSS...rk -..... ... eeedee wh ASTM DO80Z 195 0

(30 G, 70% RH); OF 2-7 Verifying that a Iaboratsr'; analysis of at least tWO Cerbcr-smlsdemontrate a reoa -- ffw~ene of t 9:7.5% fs.

seecrdanee wfth ASTM D3803-19-9 -G-OG. 70% RII) and the samples are prepefed b ... t~r a*~ ~ t;:dbet;,.Iffca 1-ir .-.msvc cdsIq-.. bz.

irs. mxln th sdsrb .thsrughly, and obtainin~g samls atlea- m iee ndrete f and with a length equal to h thicfness ef.th:P bed. er b) Emptying a lcngitudinel somplc from an adsorbcr

.r-ay, mixng adsorbent thorogh~y, and obtainirfg camlcsct zr~t t.m' *nehes ýndmameter and v.itha lzngth equal to the thickness of the bed.

ST. LUCIE - UNIT 1 3M47-25 Amendment No- 167

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page I Iof 15 Alternative Source Term and Conforming Amendment PLANT SYSTEMS SUVELLNE REQUIREMENTS (Continued)

Subsequent tO reinstalling the a' Ib~tr- I~ 1ft'F ob*-.i.n thc-a~rbon s. nplc, thc system sall be demo nstrat-ed OPERABLE b a) Ve'cifYing that the eharcoaI adSGorbcrzS remFA;-O' > 199%

ofa halegenatedh~rcrz rzrgerarA teel gas

'..lin they arc tested in plaee in aSOrdanee vith

'ANSI N510,497-5 whilc enilt prai~ h erifyinithct the .HEPAfilt*b*c,^ . 00i%

.F:movc'.

Of ther P th rtsi pl in as"",

system at :ziovfalcw aeof30,000 fmj-, 1094, 4 At least on~ce per,18'rmonths:'

V4r4fYing that theIpreSS*c.edr-pe a-cree the "mbied. 'HPA

4. flow ratc ---0d +40% A =ME Verifying that the air flovwdistribution is*-- om,.

within 20% across, HEPA filters an elarcoal adsorbers when tested in accordance with ANSI N510-497._

Verifying that the filter train starts on a Safety Injec-tion Actuation Signal.

e- After each o"rnp te OF Partial repla-ement of a HEPA filter' bank by rifyir-g that the HEPA filter banks remave ...... %e the DOP when they are tcsted in place in accord-ance-- vith ANSI N51Q 1975 while operating the vcntilaticn system at a flow Fate Of 30,000-Gfcfr - 10.

After cAh1 eampltc or patial replaeement of , ehaaroal adsoerF bank -byverifying that the charcoal adsorbers rrnvv.

--8% of a ha!egenated hydroc-irbon refrigerant test-gs--wh, th-eyp eztcd in plthe in steordanem wih ANSI N510,9 175 while opcirating !he ventilation system at a flew Fate ei

- 000-cfmn-~40W.%

ST. LUCIE - UNIT 1 314 7-26

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 12 of 15 Alternative Source Term and Conforming Amendment ADMINISTRATIVE CONTROLS (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters.in the ODCM.

2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and

3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

h. Containment Leakage Rate Testing Program A program 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, as modified by approved exemptions. This program is in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," as modified by the following exception(s):

a) Bechtel Topical Report, BN-TOP-1 or ANS 56.8-1994 (as recommended by R.G. 1.163) will be used for type A testing.

b) The first Type A test performed after the May 1993 Type A test shall be no later than May 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident Pa, is 39.6 psig. The containment design pressure is 44 psig.

The maximum allowed containment leakage rate, La, at Pa, shall be 0.50% of containment air weight per day.

Leakage rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests,

< 0.75 La for Type A tests, and < 0-4 La for secondary containment bypass leakage paths.

b. Air lock testing acceptance criteria a

1) Overall air lock leakage rate is < 0.05 La when tested at > Pa.

2) For the personnel air lock door seal, leakage rate is < 0.01 La when pressurized to _>1.0 Pa.

3) For the emergency air lock door seal, leakage rate is < 0.01 La when pressurized to > 10 psig.

ST. LUCIE - UNIT 1 6-15b Amendment No. 69, 86, 4-23, 449,

.487, 197

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 13 of 15 Alternative Source Term and Conforming Amendment ADMINISTRATIVE CONTROLS (continued) k.

A PF@~ai shall be "etablmrhcd to imp ment the WfolIowin roelUiFelrcdting of FeRginReeee Safe it': Feature (ESF) fi-ter v1nti.. ti -ste,,mr at the frec--enn*es oe'cifiedRn Ro.l-ator:

GuideN.52, R*(onin'ue 2.

emo for each of tr -9ESF systenms that an inptace test of the high efficiency Fs rhowG a penetration and ystem bpass f: 0.05% when rtesltedn accordEAnc filth RNSI N51041975 at the s,',tem f,'-rato A_,ecified be!o,.:

REPLACE VT=ISR1 SSFNeniStioT1se fIova-ate cnr~ral R-OAM tr-mracne'0 ýýrcnu4iaue.

Dec'monstrate for each of the ESF systeme that an inplace test Gf the charcoal adsorber 6hGW6.a penetratiOn and system bypass 0544 w.~hen tested-i ceroo with ANSt N40 1975 at the cyctom flo-rate specified below.

Control RoomR F!!rger~eRGY Veiantiabo 2999 o 299 MAf DOmonetrsate for each of the ESF cyrstcms that a laboraor teto pie of the charcoal ad~eitor, when obtained a6 desonbed in Reuat- Guid 1.52, Reviion 2, shows-the me#4y! iodide penet-atien less than the I:olue Abeowwh.

wpecifi n tested 'n a.ordan.c with AST-M D38, 3 1080 at a t*mpcratur. of 30.. and the relati:'e humbidilt

%I, * "iI~i,. ~Patratlon R4 Control Room Emnergency Ventilation 25

4. Demoocstrate for each of !he ESF systemso that the preceure drop acracGS the comfb'nod HEPA filters and charcoal adeerbers as leas than the value specified below when tested at the system flowrato-*,.pecifiedbelw.

DefaR Fkowrate ueoffroi KE~oom tLmorgorny venolation 2000 +200o-fmn The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the VFTP test frequencies.

1. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained, In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

ST. LUCIE - UNIT 1 6-15d Amendment No. 47-6, 44D7, 200

St. Lucie Unit I L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 14 of 15 Alternative Source Term and Conforming Amendment INSERT 1:

Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 3.

1. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass less than the value specified below when tested in accordance with ASME N510-1989 at the system flowrate specified below.

ESF Ventilation System Penetration Flowrate Control Room Emergency Ventilation < 0.05% 2000 + 200 cfm Shield Building Ventilation System < 0.05% 6000 + 600 cfm ECCS Area Ventilation System < 0.05% 30,000 + 3000 cfm

2. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass less than the value specified below when tested in accordance with ASME N510-1989 at the system flowrate specified below.

ESF Ventilation System Penetration Flowrate Control Room Emergency Ventilation < 0.05% 2000 + 200 cfm Shield Building Ventilation System. < 0.05% 6000 + 600 cfm ECCS Area Ventilation System < 0.05% 30,000 + 3000 cfm

3. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 3, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C and the relative humidity specified below.

ESF Ventilation System Penetration RH Control Room Emergency Ventilation < 2.5% 70%

Shield Building Ventilation System < 2.5% 70%

ECCS Area Ventilation System < 2.5% 70%

4. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below.

ESF Ventilation System Delta P Flowrate Control Room Emergency Ventilation < 4.15" W.G. 2000 + 200 cfm Shield Building Ventilation System < 6.15" W.G 6000 + 600 cfm ECCS Area Ventilation System < 4.15" W.G 30,000 + 3000 cfm

5. At least once per 18 months, demonstrate that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ASME N510-1989.

ESF Ventilation System Wattage Shield Building Ventilation System Main Heaters 30 + 3 kW Auxiliary Heaters 1.5 + 0.25 kW

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 15 of 15 Alternative Source Term and Conforming Amendment ADMINISTRATIVE CONTROLS (continued)

1. Steam Generator (SG) Program (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. 0.5

2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any desig basis accident, other than SG tube rupture, shall not exceed the leak e rate assumed in the accident analysis in terms of total leakage rate for SGs and leakage rate for an individual SG. Leakage is not to exceed -1 gpm total through all SGs and O-6 ge any one SG.

utrough .

3. The operational leakage performance criterion is specified in LCO 3.4.6.2.c,

'Reactor Coolant System Operational Leakage."

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

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.l, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

An assessment of degradation shall be performed to determine the type and location of flaws to which the tube may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

ST. LUCIE - UNIT 1 6-15e Amendment No. 200

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 1 of 15 Alternative Source Term and Conforming Amendment Word Processed Technical Specification Changes Page 1-3 Page 3/4 3-22 Page 3/4 3-23 Page 3/4 3-24 Page 3/4 6-27 Page 3/4 6-28 Page 3/4 6-29 Page 3/4 7-24 Page 3/4 7-25 Page 3/4 7-26 Page 6-15b Page 6-15d Page 6-15e Page 6-15f

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 2 of 15 Alternative Source Term and Conforming Amendment DEFINITIONS DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (p.Ci/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 Federal Guidance Report 11. "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

F - AVERAGE DISINTEGRATION ENERGY 1.11 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 (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES REPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

ST. LUCIE - UNIT 1 1-3 Amendment No. 27, 6D. 4W, 495,

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 3 of 15 Alternative Source Term and Conforming Amendment 0

C) (0 In i',-

z w E E E tU CL 0C IfO a-I I CL CD 0 ocoo 0

n2 0-4 w 0 0

z 0

0 0

(]0 z E E 0 LI- U) -a z 8 VOV VI Vl z

w

-j v~j 0

0(0 Ca) 40 Ca) w 2 Ca) 100 C.)~ (0.

0 0

2 =)WM 0 M0

- V) .- r CID 94W 00 C C 03 0

0 CO 00 Z0 0) -1 5 0- C w3 0 it Q( 0 0 U.

to 13 - -

St. Lucie Unit I L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 4 of 15 Alternative Source Term and Conforming Amendment TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 12 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.12.

ACTION 13 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 14 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 15 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

1) Initiate the preplanned alternate method of monitoring the appropriate parameter(s),and

2) Prepareand submit a Special Report to the Commission pursuant to Specification6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 16 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, comply with the ACTION requirements of Specification 3.9.9.

ACTION 17 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation.

ST. LUCIE - UNIT 1 3(4 3-23 Amendment No. 59,

St. Lucie Unit 1 " L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 5 of 15 Alternative Source Term and Conforming Amendment

'3* 'r z:.5 0 eli oli wWI 5f z

Wz E 20:)

U.

04 0

I?

Ci W* -j

-J Z 0 Zw CO 0I CO 0 00O 0I CO) 0)

0 0E 'S.0.

-C CM0 00 0 TM -

? z 7 j _.- -o!c 0

0- (U -O o) 0' V) 0, Cu 0 I-0.

z uJ WU 0 C) C- 0 c cCuC a.0 -,

E-Do 00 0 iý 2i

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 6 of 15 Alternative Source Term and Conformina Amendment CONTAINMENT SYSTEMS 3/4.6.6 SECONDARY CONTAINMENT SHIELD BUILDING VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.1 Two independent shield building ventilation systems shall be OPERABLE.

APPLICABILITY: MODES 1,2,3 and 4.

ACTION:

With one shield building ventilation system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.6.1 Each shield building ventilation system shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

b. By performing required shield building ventilation system filter testing in accordance with the Ventilation Filter Testing Program.

c. At least once per 18 months by:

1. Verifying that the air flow distribution is uniform within 20% across H EPA filters and charcoal adsorbers when tested in accordance with ASME N510-1989.

2. Verifying that the filtration system starts automatically on a Containment Isolation Signal (CIS).

3. Verifying that the filter cooling makeup air and cross connection valves can be manually opened.

4. Verifying that each system produces a negative pressure of

> 2.0 inches W.G. in the annulus within 2 minutes after a CZontainment Isolation Signal (CIS).

ST. LUCIE - UNIT 1 3/4 6-27 Amendment No. -27.

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 7 of 15 Alternative Source Term and Conforming Amendment Page Deleted ST. LUCIE - UNIT 1 314 6-28 Amendment No. 467,

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 8 of 15 Alternative Source Term and Conforming Amendment Page Deleted ST. LUCIE - UNIT 1 3/4 6-29 Amendment No. -W,

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 9 of 15 Alternative Source Term and Conforming Amendment PLANT SYSTEMS 314.7.8 ECCS AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8.1 Two independent ECCS area exhaust air filter trains shall be OPERABLE.

APPLICABILITY: MODES 1,2, 3 and 4.

ACTION:

With one ECCS area exhaust air filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.8.1 Each ECCS area exhaust air filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes.

b. By performing required ECCS area ventilation system filter testing in accordance with the Ventilation Filter Testing Program.

c. At least once per 18 months:

1. Verifying that the air flow distribution is uniform within 20% across HEPA filters and charcoal adsorbers when tested in accordance with ASME N510-1989.

2. Verifying that the filter train starts on a Safety Injection Actuation Signal.

ST. LUCIE - UNIT 1 314 7-24 Amendment No.

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 10 of 15 Alternative Source Term and Conforming Amendment Page Deleted

• ST. LUCIE ý UNIT I 3/4 7-25 Amendment No. 4*7,

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 11 of 15 Alternative Source Term and Conforming Amendnient Page Deleted ST. LUCIE - UNIT 1 3/4 7-26 Amendment No.

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 12 of 15 Alternative Source Term and Conforming Amendment ADMINISTRATIVE CONTROLS (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM.

2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and

3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

h. Containment Leakage Rate Testincq Proaram A program 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, as modified by approved exemptions. This program is in accordance with the guidelines contained in Regulatory Guide 1.163, "Performnance-Based Containment Leak-Test Program," as modified by the following exception(s):

a) Bechtel Topical Report, BN-TOP-1 or ANS 56.8-1994 (as recommended by R.G. 1.163) will be used for type A testing.

b) The first Type A test performed after the May 1993 Type A test shall be no later than May 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident Pa, is 39.6 psig. The containment design pressure is 44 psig.

The maximum allowed containment leakage rate, La, at Pa, shall be 0.50% of containment air weight per day.

Leakage rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests,

< 0.75 La for Type A tests, and < 0.096 La for secondary containment bypass leakage paths.

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is < 0.05 La when tested at > Pa.

2) For the personnel air lock door seal, leakage rate is < 0.01 La when pressurized to > 1.0 Pa.

3) For the emergency air lock door seal, leakage rate is < 0.01 La when pressurized to > 10 psig.

ST. LUCIE - UNIT I 6-15b Amendment No. 69, 86, 42 -449, 487, 497,

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 13 of 15 Alternative Source Term and Conforming Amendment ADMINISTRATIVE CONTROLS (continued)

k. Ventilation Filter Testing Program (VFTPI A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 3.

1. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass less than the value specified below when tested in accordance with ASME N510-1989 at the system flowrate specified below.

ESF Ventilation System Penetration Flowrate Control Room Emergency Ventilation < 0,05% 2000 + 200 cfm Shield Building Ventilation System < 0.05% 6000 + 600 cfm ECCS Area Ventilation System <0.05% 30,000 + 3000 cfm

2. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass less than the value specified below when tested in accordance with ASME N510-1989 at the system flowrate specified below.

ESF Ventilation System Penetration Flowrate Control Room Emergency Ventilation <-0.05% 2000 + 200 cfm Shield Building Ventilation System < 0.05% 6000 + 600 cfm ECCS Area Ventilation System <.0.05% 30,000 + 3000 cfm

3. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 3, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 301C and the relative humidity specified below.

ESF Ventilation System Penetration RH Control Room Emergency Ventilation *<2.5% 70%

Shield Building Ventilation System *2-5% 70%

ECCS Area Ventilation System *<2.5% 70%

4. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below.

ESF Ventilation System Delta P Flowrate Control Room Emergency Ventilation < 4.15" WG. 2000 + 200 cfm Shield Building Ventilation System < 6.15" W.G. 6000+- 600 cfm ECCS Area Ventilation System < 4.15"W.G, 30,000 + 3000 cfm

5. At least once per 18 months, demonstrate that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ASME N510-1989.

ESF Ventilation System Wattage Shield Building Ventilation System Main Heaters 30 + 3 14W Auxiliary Heaters 1.5 + 0.25 kW ST. LUCIE - UNIT 1 6-15d Amendment No. 476, *97, 200,

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 14 of 15 Alternative Source Term and Conforming Amendment ADMINISTRATIVE CONTROLS (continued)

The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the VFTP test frequencies.

I. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

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 ifthe 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 exceed 0.5 gpm total through all SGs and 0.25 gp through.5 any one SG. a

3. The operational leakage performance criterion is specified in LCO 3.4.6.2.c, "Reactor Coolant System Operational Leakage."

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

ST. LUCIE - UNIT I 6-15e Amendment No. 200,

St. Lucie Unit 1 L-2007-085 Docket No. 50-33 5 Attachment 3 Proposed License Amendment Page 15 of 15 Alternative Source Term and Conforming Amendment ADMINISTRATIVE CONTROLS fcontinued)

I. Steam Generator (SG) Program (continued)

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

An assessment of degradation shall be performed to determine the type and location of flaws to which the tube may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outages nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary-to-secondary leakage.

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 NRC.

STARTUP REPORT 6.91.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment of 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 plant.

ST. LUCIE - UNIT 1 6-1 5f Amendment No. 200,

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 1 of8 Alternative Source Term and Conforming Amendment TS Bases Changes (Information Only)

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 2 of 8 Alternative Source Term and Conforming Amendment SECTION NOM. rrLE TECHNICAL SPECIFICATIONS PAG2.1, 3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 7 REVISION NO.: REACTOR COOLANT SYSTEM 3 ST. LUCIE UNIT I -M 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314.4.5 STEAM GENERATORS (SG) TUBE INTEGRITY (continued)

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses the steam discharge to the atn h is

-.---.-baAed on the total primary-to-secondary leakage from all SGs nd

.- ... pm through any one SG as a result of accident induced condI lgns.

-..*----Foaccidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 Is assumed to be equal to the limits in LCO 3.4.8, 'Reactor Coolant System Specific Activity." For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or)e-a ppro C sis (e.g., a small fraction of these limi Stegei1~atoetube-* satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Limiting Condition for Operation (LCO)

The LCO requires that SG tube Integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

During a SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria Is removed from service by plugging. Ifa tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube..

A S tube has tube Integrity when it satisfies the SG performance criteria.

The SG performance criteria are defined in Specification 6.8.4.1, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. Failure to meet any one of these criteria is considered failure to meet the LCO.

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 3 of 8 Alternative Source Term and Conforming Amendment sEMc'oN No.: Tte TECHNICAL SPECIFICATIONS PA 314.4 BASES ATTACHMENT 6 OF ADM-25.04 8 4f.

REVISION NO.: REACTOR COOLANT SYSTEM 3 ST. LUCIE UNIT 1 M 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314.4.5 STEAM GENERATORS (SG) TUBE INTEGRITY (continued)

The structural Integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included In the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area Increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load verses displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant' is defined as "An accident loading condition other than differential pressure Is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube Integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

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).

The accident induced leakage performance criterion ensures that the primary-to-secondary leakage caused by a design basis accident, than a SGTR, is within the accident analysis assumptions. The a den analysis assumes tha-acciT.ttinduced leakage does not exceed 60bpm j total from all S~s ano-__.ffWough any one SG. The accident induýed leakage rate includes arly-p-dfary-to-secondary leakage existing prior to the accident in addition to primary-to-secondary leakage induced during the accident

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 4 of 8 Alternative Source Term and Conforming Amendment SECTIONNO-" TE: TECHNICAL SPECIFICATIONS 3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 12.

REVwsIoNNO.: REACTOR COOLANT SYSTEM 12 3 ST. LUCIE UNIT 1 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 3/44.5 STEAM GENERATORS (SG) TUBE INTEGRITY (continued)

SR 4.4.5.2 During a SG inspection any inspected tube that satisfies the Steam Generator Program repair criteria Is removed from service by plugging. The tube repair criteria delineated in Specification 6.8.4.1 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error In the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The frequency of prior to entering HOT SHUTDOWN following a SG tube inspection ensures that the Surveilance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary-to-secondary pressure differential.

References

1. NEI 97-06, 'Steam Generator Program Guidelines*

2. 10 CFR 50 Appendix A, GDC 19

3. 10CFR 100

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

5. Draft Regulatory Guide 1.121, 'Bases for Plugging Degraded PWR Steam Generator Tubes,* August 1976 Ed* Water Reactor Steam Generator Examination Guidelines'

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 5 of 8 Alternative Source Term and Conforming Amendment SECTION NO.. TITLE: TECHNICAL SPECIFICATIONS PAGE.

3/4.4 ........ BASES ATTACHMENT 6 OF ADM-25.04 14 - 9 REVISIONNO.:

3 REACTOR COOLANT ST. LUCIE UNITSYSTEM 1

314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 3/4A.6.2 REA&CTOR COOLANT SYSTEM OPERATIONAL LEAKAGE (continued)

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 reactor coolant pressure boundary (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 safety analysis for akneent resulting in steam discharge to the atmosphere assumes primary to secondary leakage as the Initial condition.

Primary to secondary leakage Is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary is leased via the safety valves or atmospheric dump valves. Th~l*j*Vrimary to secondary leakage is relatively inconsequeftiit~.

Primary-to-secondary leakage contaminates the secondary fluid. The safety analysis for an event resulting In steam discharge to the atmosphe assumeWa-to-secondary leakage from all steam generators /,o.5 gpm an~r0rgpmrough any one SG as a result of accident induced conditior odose consequences of these events are within the limits of GDC 19, 10 CFR 100,r the NRC approved licensing basis (e.g., a small fraction of these limits. The LCO requirement to limit primary-to-secondary leakage through any o0e steam generator to less than or equal to 150 gpd is ams

• Uoos assumed in the safety analysis.

The RCS operationaes Critenon 2 of 10 CFR 50.36(c)(2)(ii).

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 6 of 8 Alternative Source Term and Conforming Amendment SECTIONNO.: TtTE TECHNICAL SPECIFICATIONS PAGE 3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 REv~ISIONO:: REACTOR COOLANT SYSTEM T 3 ST. LUCIE UNIT 1 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Umits 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 limit 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.

"- 314.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed Ir Lfe- so.

.m following a steam genera or tuB-ruptureaccidant in conjuni th an assumed steady state primary-to-secondary steam generator leakage rate of* n a concurrent loss of offsite electrical power. The values for the limi on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the St Lucie site, such as site boundary location and meteorological conditions, were not considered In this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCVgram DOSE EQUIVALENT 1-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible Iodine spiking phenomenon whih m occur following changes in THERMA WER.

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 7 of 8 Alternative Source Term and Conforming Amendment 5050N NO. TrrTE' TECHNICAL SPECIFICATIONS PAGE:

3/4.6 BASES ATTACHMENT 8 OF ADM-25.04 10 of 10 VSION NO. CONTAINMENT SYSTEMS 4 ST. LUCIE UNIT 1 314.6 CONTAINMENT SYSTEMS (continued)

BASES (continued) 314.6.6 SECONDARY CONTAINMENT 3146.6.1 SHIELD BUILDING VENTILATION SYSTEM The OPERABILITY of the shield building ventilation systems ensures that containment vessel leakage occurring during LOCA conditions into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is necessary to meet the assumptions used in the accident analyses and limit the siteboundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions. e maWith respect to Surveillance 4.6o6.1n. a* to uil bere.5strIcte*edto thoseleakag paths sce 0and as .e Iacpci arssued Irn d analyses.Thti restr, In.

con iolyjucton wt teratioe shi*eldg ,

rbeui thaEoSF wi*ll*eliitt site *bo ndary ratio nde dit trena7 lmt o 1uring. aI*enovcods.

cdad 314.6.6.2 SHIELD BUILDING INTEGRITY SHIELD BUILDING INTEGRITY ensures that the release of radioactive materials from the primary containment atmosphere will be restricted to those leakage restriction, analyses. Thispaths and associated leak rates assumed In the accident in conjunction with operation of the shield building ventilation system, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

314.6.6.3 SHIELD BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural Integrity of the containment shield building will be maintained comparable to the original design standards for the life of the facility. Structural integrity Is required to provide 1) protection for the steel vessel from the external missiles,

2) radiation shielding in the event of a LOCA, and 3) an annulus surrounding the steel vessel that can be maintained at a negative pressure within two milnutes after a-LOCA.

St. Lucie Unit 1 L-2007-085 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 8 of 8 Alternative Source Term and Conforming Amendment

• O.

rn. TECHNICAL SPECIFICATIONS PAG,*

3/4.7 BASES ATTACHMENT 9 OF ADM-25.04 8 of 10 REVIS*ONNO. PLANT SYSTEMS 1A ST. LUCIE UNIT 1 314.7 PLANT SYSTEMS (continued)

BASES (continued) 314.7.8 ECCS AREA VENTILATION SYSTEM The OPERABILITY of the ECCS area ventilation system ensures that radio-active materials leaking from the ECCS equipment following a LOCA are filtered prior to reaching the environment The operation of this system and the resultant effect on offsite dosage calculations was assumed In the accident analyses.

4 With respc SurCellance4.N Te52e ulA SL n

,9m,9. 6.oJ,' sr too , *,rnaf 0o 1h aVin pr "Oh1 er la*n f t a° 0 dth ev 0 " s*

3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on sealed source removable contamination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the probable leakage from the source material. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. Quantities of interest to this specification which are exempt from the leakage testing are consistent with the criteria of 10 CFR Part 30.11-20 and 70.19. Leakage from sources excluded from the requirements of this specification is not likely to represent more than one maximum permissible body burden for total body irradiation If the source material is inhaled or ingested.

I

NAI Report Release Report Number: NAI- 1101-043 Revision Number: 2

Title:

AST Licensing Technical Report for St. Lucie Unit 1

Description:

This report documents the results of the analyses and evaluations performed by Numerical Applications, Inc. in support of the St. Lucie Unit 1 licensing project to implement alternative radiological source terms.. 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.

7/,~AA AP"J" ,-te 2 Author (Mark Pope) Date Dae 07 RViewer (Jim Harrell) Date

'Reviewer (Joe Sinodis) Date NAT Management (Tom George) Date

NUMERICAL AST Licensing Technical Report for NAI-l 101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit 1 Page 2 of 87 S. .. S.IN. N.INEPRINGAeD

... 2of87 Table of Contents 1.0 Radiological Consequences Utilizing the Alternative Source Term Methodology ............................. 4 1.1 Introd uction ...................................................................................................................................... 4 1.2 Evaluation Overview and Objective ............................................................................................ 4 1.3 Proposed Changes to the St. Lucie Unit No. 1 Licensing Basis .................................................. 4 1.4 Compliance with Regulatory Guidelines ..................................................................................... 6 1.5 C omputer Codes ............................................................................................................................... 6 1.6 Radiological Evaluation Methodology ......................................................................................... 7 1.6.1 Analysis Input Assumptions .............................................................................................. 7 1.6.2 Acceptance Criteria ........................................................................................................ 7 1.6.3 Control Room Ventilation System Description ................................................................ 7 1.6.3.1 Normal Operation ................................................................................................. 8 1.6.3.2 Emergency Operation .......................................................................................... 9 1.6.3.3 Control Room Dose Calculation Model .............................................................. 9 1.6.4 Control Room Inleakage Sensitivity Study .................................................................... 11 1.6.5 D irect Shine D ose ................................................................................................................ 11 1.7 Radiation Source Terms ................................................................................................................. 12 1.7.1 Fission Product Inventory ............................................ 12 1.7.2 Primary Coolant Source Term .......................................................................................... 13 1.7.3 Secondary Side Coolant Source Term ................................... 13 1.7.4 LOCA Containment Leakage Source Term ................................ 13 1.7.5 Fuel Handling Accident Source Term .............................................................................. 14 1.8 Atmospheric Dispersion (X/Q) Factors ....................................................................................... 14 1.8.1 Onsite X/QIDetermination .............................................................................................. 14 1.8.2 Offsite X/Q Determination ........................................... 15 1.8.3 Meteorological Data ...................................................................................................... 16 1.9 Consideration of High Burnup Fuel ............ .............................. 17 2.0 Radiological Consequences - Event Analyses ..................................................................... I............. 18 2.1 Loss of Coolant Accident (LOCA) .......................................... 18 2.2 Fuel Handling Accident (FHA) ................................................................................................. 26 2.3 Main Steamline Break (MSLB) ................................................................................................ 29 2.4 Steam Generator Tube Rupture (SGTR) ....................................................................................... 33 12.5 Reactor Coolant Pump Shaft Seizure (Locked Rotor) ................................................................ 37 2.6 Control Element Assembly Ejection (CEA) .............................................................................. 41 2.7 Inadvertent Opening of a Main Steam Safety Valve (IOMSSV) ............................................. 45 2.8 Environmental Qualification (EQ) ........................................................................................... 48 3.0 Summ ary of R esults .............................................................................................................................. 48 4 .0 Conclusion ............................................................................................................................................. 48 5 .0 R eferences ............................................................................................................................................. 48

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 3 of 87 SOUJITIONS INENGINEERING ANDSOFTWARE Figures and Tables Figure 1.8.1-1 Onsite Release-Receptor Location Sketch ............................................................................ 51 Table 1.6.3-1 Control Room Ventilation System Parameters .................................................................. 52 Table 1.6.3-2 LOCA Direct Shine Dose ............................................. 53 Table 1.7.2-1 Primary Coolant Source Term ................  ;......................... 53 Table 1.7.3-1 Secondary Side Source Term .......................................................................................... 54 Table 1.7.4-1 LOCA Containment Leakage Source Term ...................................................................... 54 Table 1.7.5-1 Fuel Handling Accident Source Term ............................................................................. 56 Table 1.8.1-1 Release-Receptor Combination Parameters for Analysis Events ...................................... 57 Table 1.8.1-2 Onsite Atmospheric Dispersion (X/Q) Factors for Analysis Events ................................. 60 Table 1.8.1-3 Release-Receptor Point Pairs Assumed for Analysis Events ........................................... 62 Table 1.8.2-1 Offsite Atmospheric Dispersion (XIQ) Factors for Analysis Events ................................ 63 Table 2.1-1 Loss of Coolant Accident (LOCA) - Inputs and Assumptions ............................................ 64 Table 2.1-2 LO CA Release Phases ......................................................................................................... 67 Table 2.1-3 LOCA Time Dependent RW T pH ........................................................................................ 67 Table 2.1-4 LOCA Time Dependent RWT Total Iodine Concentration* ............................................... 68 Table 2.1-5 LOCA Time Dependent RWT Liquid Temperature ............................................................ 69 Table 2.1-6 LOCA Time Dependent RWT Elemental Iodine Fraction .......................... "............................ 70 Table 2.1-7 LOCA Time Dependent RWT Partition Coefficient ........................................................... 71 Table 2.1-8 LOCA Release Rate from RWT .......................................................................................... 72 Table 2.1-9 LOCA Dose Consequences ................................................................................................ 72 Table 2.2-1 Fuel Handling Accident (FHA) - Inputs and Assumptions ................................................. 73 Table 2.2-2 Fuel Handling Accident Dose Consequences ...................................................................... 73 Table.2.3-1 Main Steam Line Break (MSLB) - Inputs and Assumptions .............................................. 74 Table 2.3-2 M SLB Steam Release Rate ................................................................................................... 75 Table 2.3-3 MSLB Steam Generator Tube Leakage .............................................................................. 76 Table 2.3-4 M SLB Dose Consequences ................................................................................................ 76 Table 2.4-1 Steam Generator Tube Rupture (SGTR) - Inputs and Assumptions .................................... 77 Table 2.4-2 SGTR Integrated M ass Releases ................................................ *..................... :....................... 78 Table 2.4-3 SGTR 60 gtCi/gm D.E. 1-131 Activities ............................................................................. 78 Table 2.4-4 SGTR Iodine Equilibrium Appearance Assumptions ........................................................ 78 Table 2.4-5 SGTR Concurrent Iodine Spike (335 x) Activity Appearance Rate .................. 79 Table 2.4-6 SGTR Dose Consequences .........  :.................................... 79 Table 2.5-1 Reactor Coolant Pump Shaft Seizure (Locked Rotor) - Inputs and Assumptions ............... 80 Table 2.5-2 Locked Rotor Steam Release Rate ................................................ ....................... 81 Table 2.5-3 Locked Rotor Steam Generator Tube Leakage ..................................................................... 81 Table 2.5-4 Locked Rotor Dose Consequences ....................................................................................... 81 Table 2.6-1 Control Element Assembly (CEA) Ejection - Inputs and Assumptions ............................. 82 Table 2.6-2 CEA Ejection Steam Release Rate ...................................................................................... 83 Table 2.6-3 CEA Ejection Steam Generator Tube Leakage ................................................................... 84 Table 2.6-4 CEA Ejection Dose Consequences ....................................................................................... 84 Table 2.7-1 IOM SSV - Inputs and Assumptions .................................................................................... 85 Table 2.7-2 IOMSSV Steam Generator Tube Leakage ................................... 86 Table 2.7-3 IOM SSV Dose Consequences ............................................................................................. 86 Table 3.1 St. Lucie Plant, Unit No. 1 Summary of Alternative Source Term Analysis Results .............. 87

NUMERICAL AST Licensing Technical Report for NAI-i1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 4 of 87 SO.L5tIONS ANDSOFTWARE INENGIN4EERING 1.0 Radiological Consequences Utilizing the Alternative Source Term Methodology 1.1 Introduction The current St. Lucie Plant, Unit No. 1, licensing basis for the radiological analyses for accidents discussed in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR) is basedon methodologies and assumptions that are primarily derived from Technical Information Document (TID)- 14844 and other early guidance.

Regulatory Guide (RG) 1.183 provides guidance on application of Alternative Source Terms (AST) in revising the accident source terms used in design basis radiological consequences analyses, as allowed by 10CFR50.67. Because of advances made in understanding the timing, magnitude, and chemical form of fission product releases from severe nuclear power plant accidents, IOCFR50.67 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 alternative source terms (ASTs).

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 RG 1.183, and supplemented by Regulatory Issue Summary 2006-04, is being used to calculate the offsite and control room radiological consequences for St. Lucie Unit No. 1 to support the control room habitability program by addressing the radiological impact of potential increases in control room unfiltered air inleakage.

The following limiting UFSAR Chapter 15 accidents are analyzed:

. Loss-of-Coolant Accident (LOCA)

&. Fuel Handling Accident (FHA)

• Main Steam Line Break (MSLB)

• Steam Generator Tube Rupture (SGTR)

• Reactor Coolant Pump Shaft Seizure (Locked Rotor)

• Control Element Assembly (CEA) Ejection ITnadvertent Opening of a Main Steam Safety Valve (IOMSSV)

Note that although RG 1.183 does not include the IOMSSV event, it was included in the AST analysis to provide consistency of methodology with the remaining limiting accident analyses.

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 500 cfm. The use of 500 cfm as a design basis value was established to be above the unfiltered inleakage value determined through testing and analysis consistent with the resolution of issues identified in NEI 99-03 and Generic Letter 2003-01.

1.3 Proposed Changes to the St. Lucie Unit No. 1 Licensing Basis Florida Power and Light (FP&L) Company proposes to revise the St. Lucie Plant, Unit No. 1, licensing basis to implement the AST, described in RG 1.183, through reanalysis of the radiological consequences of the UFSAR Chapter 15 accidents listed in Section 1.2 above. As part of the full implementation of this AST, the following changes are assumed in the analysis:

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 5 of 87 SOLUTONS IN ENGINEERING ANDSOF1WAR

• The total effective dose equivalent (TEDE) acceptance criterion of 10CFR50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10CFR100.11.

• 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.

• Increased values for control room unfiltered air inleakage are assumed (unfiltered inleakage increased until applicable dose limit is approached).

• A steam generator tube leakage rate for accident analysis which exceeds the current Technical Specification operational leakage limit is utilized.

• An SBVS bypass leakage value that is more restrictive than the current Technical Specification limit is utilized. Plant maintenance and surveillance history demonstrate that the proposed reduced containment leakage values have been met in the past (Reference 5.8).

Accordingly, the following changes to the St. Lucie, Unit No. 1, Technical Specifications (TS) are proposed:

• The definition of Dose Equivalent 1-131 in Section 1.10 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,"

as the source of thyroid dose conversion factors.

" Tables 3.3-6 and 4.4-3 are revised to add the Control Room Isolation Area Radiation Monitors to the applicability of TS 3/4.3.3.1, Radiation Monitoring.

0 Surveillance Requirement 4.6.6.1 is revised to relocate the HEPA filter, charcoal adsorber, flow rate and heater surveillance test acceptance criteria for the Shield Building Ventilation System to the Ventilation Filter Testing Program in TS Section 6.8.4.k.

• Surveillance Requirement 4.7.8 is revised to relocate the HEPA filter, charcoal adsorber and flow rate surveillance test acceptance criteria for the ECCS Area Ventilation System to the Ventilation Filter Testing Program in TS Section 6.8.4.k.

• The leakage rate acceptance criterion for secondary containment bypass leakage paths (i.e. Shield Building Bypass Leakage) stated in TS 6.8.4.h, "Containment Leakage Rate Testing Program," is reduced from 27% to 9.6%.

" The HEPA filter and charcoal adsorber test acceptance criteria for the ECCS Area and Shield Building Ventilation Systems relocated to the VFTP are revised as follows:

o The filter efficiency test acceptance criteria are increased from 99% to 99.95%

o The inplace charcoal adsorber efficiency test acceptance criteria are increased from 99% to 99.95%

" In the VFTP (TS 6.8.4.k), reference to Regulatory Guide 1.52, Revision 2 is replaced with reference to Regulatory Guide 1.52, Revision 3.

" In the Surveillance Requirements for the ECCS Area Ventilation System and the Shield Building Ventilation System (SRs 4.7.8.1 and 4.6.6.1, respectively), as well as in the VFTP (TS 6.8.4.k), reference to ANSI N510-1975 is replaced with reference to ASME N510-1989.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 6 of 87 SOU=fON5INCENTNEERNAIOSO4MAWAR The accident induced leakage performance criteria of the Steam Generator (SG) Program described in TS Section 6.8.4.1 is changed from 1.0 gpm total through all SGs and 0.5 gpm through any one SG, to 0.5 gpm total through all SGs and 0.25 gpm through any one SG.

1.4 Compliance with Regulatory Guidelines The revised St. Lucie Unit No. 1 accident analyses addressed in this report follow the guidance provided in RG 1.183. Assumptions and methods utilized in this analysis for which no specific guidance is provided in RG 1.183, but for which a regulatory precedent has been established, are as follows:

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 5.11 Atmospheric Dispersion Factors MicroShield 5.05 5.12 Direct Shine Dose Calculations ORIGEN 2.1 5.13 Core Fission Product Inventory PAVAN 2.0 5.14 Atmospheric Dispersion Factors RADTRAD-NAI 1.la 5.15 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 (X/Q) 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 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 initially modified to compile on a UNIX system. Once.compiled, an extensive design review/verification and

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 7 of 87

.. ... IN eNGINEERING..

ONS validation process began on the code and documentation. The subject of the review also included the source code for the solver, which is made available in a separate distribution from the NRC. RADTRAD-NAI validation is performed with three different types of tests:

0 Comparison of selected Acceptance Test Case results with Excel spreadsheet solutions and hand solutions, 0 Separate effects tests, and 0 Industry examples.

0 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 is 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), atmospheric dispersion factors (Section 1.8), and consideration of high bumup fuel (Section 1.9). 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 RG 1.183 and requirements of IOCFR50.67. The acceptance criteria for specific postulated accidents are provided in Table 6 of RG 1.183. For analyzed events not addressed in RG 1.183, the basis used to establish the acceptance criteria for the radiological consequences is provided in the discussion of the event in Section 2.0. For St. Lucie Unit No. 1, the event not specifically addressed in RG 1.183 is the Inadvertent Opening of a MSSV.

1.6.3 Control Room Ventilation System Description The Control Room Air Conditioning System' (CRACS) and Control Room Emergency Ventilation System (CREVS) are required to assure control room habitability. The design of the control room envelope and overall descriptions of the Control Room Air Conditioning and Emergency Ventilation Systems arediscussed in the St. Lucie Unit 1 UFSAR, Sections 6.4 and 9.4.1.

The control room envelope is pressurized relative to the surroundings at all times during normal plant operation with outside air continuously introduced to the control room envelope at a rate of 750 cfm (for dose analyses purposes, this value is conservatively increased to 920 cfm). Following a design basis accident the control room is pressurized at the rate of up to 450 cfm to maintain a positive pressure differential. Makeup air for pressurization is filtered before entering the control room.

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 8 of 87' S*.UTIONSINENGINEERING ANDSOFTWIAREI Automatically actuated, redundant isolation valves are provided at each outside air intake and exhaust air path so that the control room envelope is immediately isolated on receipt of a CIAS, or outside air intake high radiation signal. Unfiltered inleakage paths through the isolation valves are reduced by using low leakage butterfly valves. The isolation valves close on loss ofoperating power. The Control Room Air Conditioning System is capable of automatic actuation or manual transfer from its normal operating mode to the pressurized or isolated modes as necessary. The system is designed to perform its safety functions and maintain a habitable environment in the control room envelope during isolation.

The net volume of the control room envelope serviced by the Control Room Air Conditioning and Emergency Ventilation Systems is approximately 62,300 cubic ft.

The Control Room Air Conditioning System (CRACS) consists of three air conditioners and a diicted air intake and air distribution system. The system is zone isolated with filtered recirculated air, widely separated dual air inlets, and provisions for positive pressurization (Ž1/8 in. water gauge). Each air conditioner includes a cabinet type centrifugal fan, direct expansion refrigerant cooling coil, roughing filter, water-cooled refrigerant condenser and refrigerant compressor. Air conditioning unit capacity is 50 percent each during normal operation and 100 percent each during post LOCA operation. Under emergency conditions, only one out of three air conditioning units and one train of the Control Room Emergency Ventilation System are required to maintain the habitability of the control room envelope.

The habitability systems (air filtration and ventilation equipment with associated instrumentation, controls and radiation monitoring) are capable of performing their functions assuming a single active component failure coincident with a loss of offsite power. Redundant equipment which is essential to safety is powered from separate safety related buses such that loss of one bus does not prevent the Control Room Air Conditioning System from fulfilling its safety function.

The control room operator has the ability, through radiation monitors, to determine radiation levels in each of the outside air intake ducts. There are two radiation monitors located in a common skid.

Either monitor is capable of sampling either outside air intake duct.

The Control Room Emergency Ventilation System (CREVS) consists of a filter train with HEPA filters and charcoal adsorbers with two redundant booster centrifugal fans.

1.6.3.1 Normal Operation During normal operation, the control room is air conditioned by the air conditioning units. Two of the three air conditioning units in the control room are in the automatic mode of control. One or two units are normally running with the other unit(s) in a standby status, available for manual actuation in the event of a failure of an operating unit. Fresh air is taken in through either the northern or the southern outside air intakes by remote manual opening of the redundant motor operated isolation valves.

Control room air is drawn into the air handling section through a return air duct system and roughing filters (not credited for dose analyses), .and is cooled as required. Conditioned air is directed back to the control room through the supply air duct system. Outside air makeup is supplied through either of two outside air intakes located in the northern and southern walls of the Reactor Auxiliary Building at elevation 78 feet 9 inches. This makeup air replenishes the air exfiltrated to the outside in addition to that being exhausted by the toilet and kitchen exhaust fans. The return air flowrate is controlled automatically by the return dampers with their corresponding controller either in Auto or Manual control mode to maintain a constant positive pressure of 1/8 inch wg in response to the average pressure differential between the control room and its surroundings.

During normal operation the CREVS is isolated from the CRACS ducts by dampers.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 9 of 87 SOWOONSINENGINERINGMAD SOFTWARE 1.6.3.2 Emergency Operation The emergency modes of operation of the Control Room Air Conditioning System are:

a) automatic isolation and automatic recirculation with partial filtration of recirculated air, or b) automatic isolation with immediate manual and/or automatic filtered pressurization and recirculation with partial filtration.

Upon receipt of a containment isolation actuation signal (CIAS) or a high radiation signal, the redundant isolation valves on the outside air intake and exhaust ducts close automatically. The two CREVS booster fans start automatically while the air conditioning units remain running. The isolation time including the damper closure time is equal to a maximum of 35 seconds, assuming offsite power is available. If offsite power is not available, the isolation time is 45 seconds, which includes the 10-second diesel generator start time (for dose analyses purposes, this value is conservatively increased to 50 seconds). A portion of the control room air is recirculated through the HEPA filters and charcoal adsorbers for removal of radioactive particles and iodine, respectively.

Outside air intake dampers are adjusted to allow sufficient outside air makeup flow to maintain control room pressurization. By observing the radiation monitors located in the outside air intake ducts, the operator restores outside air makeup by selecting which set of isolation valves to open. After determining which outside air intake has the least, or zero, amount of radiation, the operator opens the isolation valves on that intake and adjusts the system dampers for proper flow. All outside air make-up and a portion of the control room return air is passed through a filter train for removal of radioactive particulates, iodides, carbon dioxide and other gaseous impurities before it enters the air conditioning units. Depending on the cooling required the operator may stop or start air conditioning units. The operator may stop one of the two CREVS filtration fans.

In the event of a CIAS or high radiation signal followed by a loss of offsite power, the outside air intake isolation valves are designed to fail as is and the CREVS fans stop. Outside air is not drawn into the control room because the control room is pressurized during normal operation and the coasting down fan is discharging against a positive pressure in addition to overcoming ductwork and damper frictional losses. When sequenced onto the diesel generator the valves automatically close and the CREVS fan is started.

The Control Room Emergency Ventilation System removes potentially radioactive particulates and iodine from the control room air during the post-LOCA operating mode. The filter train consists of a HEPA filter, charcoal adsorber, and two booster fans. The system operates post-LOCA to maintain a positive control room pressure. The flow control valves, installed in each air intake, control the flow of air being drawn into the control room. Post-LOCA makeup flow enters through one of these ducts and passes through the charcoal filters. Thus, all makeup air is filtered.

The air cleaning unit removes radioactivity from the control room envelope atmosphere. The HEPA filters remove 0.3-micron particles from atmospheric air at an efficiency greater than 99.9 percent.

The charcoal adsorbers have an elemental and organic iodine removal efficiency of 99.825 percent minimum.

1.6.3.3 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

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 10 of 87 ANDSFTAR S-tLLIONSINCE4GrNERING evolutions associated with the CR 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 RG 1.183: Figure 1.8.1-1 provides a site sketch showing the St. Lucie Unit 1 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 CR 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. The control room model will define two concurrent air intake paths representing the defined CR ventilation system air intake and the unfiltered inleakage into the CR. Outside air can enter the control room through the filtration/ventilation system from either or both of two ventilation intake locations that are located on opposite ends of the CR. Due to their diverse locations, these intakes are assigned different dispersion factors for calculating the concentration of radioactive isotopes in the air drawn in through that intake due to the activity released from various locations on the site during an accident. Unfiltered outside air can also enter the CR directly from various sources of unfiltered inleakage. Modeling of the Control Room will address these factors as they apply to the various release locations for each analyzed event.

Details of the CR modeling for each event is described in subsequent event analyses sections.

During normal operation, both of these control room ventilation intakes are open and the control room ventilation alraws in 750 cfm of fresh outside air through both of the vents in parallel and delivers it unfiltered.to the control room. For AST analyses this value is being conservatively increased to 920 cfm. In this configuration, the dispersion factor for air being drawn into the control room is assumed to be the more limiting of the dispersion factors for the two intake locations. These intakes are both automatically closed upon actuation of the CR isolation mode and no air is intentionally drawn into the control room ventilation. However, the control room ventilation system recirculates the air within the CR through the filtration system to remove contaminates that are already drawn into the system or have leaked.into the control room. During the course of the event, fresh air is required to be added to the CR in order to maintain positive pressure and air quality. The operator will selectively open the ventilation system intake location with the lower radioactive concentrations and draw up to 450 cfm of outside air through the filtration system and into the control room. Therefore, at this point, the model uses the dispersion factor for the more favorable air intake location for assessing the dose from this

-filtered makeup contribution. This filtered intake is assumed to continue throughout the rest of the 30-day duration of the, dose calculation.

During the entire course of the event it is also assumed that contaminated outside air can also enter the control room (unfiltered) via various leakage paths. This air may enter the control room through a number of different locations that may be defined by testing. In the absence of detailed testing results, some judgments are necessary in order to assign a single dispersion factor that is appropriate for the combined unfiltered inleakage from various diverse sources. At the beginning of the event, the dose calculation conservatively assigns an initial dispersion factor applicable to.the least favorable control room ventilation system intake location. Following CR isolation, when both CR ventilation intakes are closed, the dispersion factor for the CR unfiltered inleakage assumes a dispersion factor corresponding to a location that is at the midpoint of both of the CR intake locations. At the time when the operator unisolates the control room by opening'the favorable air intake, this analysis will apply the dispersion factor for the more favorable CR intake location to the unfiltered inleakage component.

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. The

1 J NUMERICAL APPLICATIONS, INC.

ANDSOFTWVARE SýtýONS IN ENGINEERING AST Licensing Technical Report for St. Lucie Unit 1 NAI-1101-043, Rev. 2 Page 11 of 87 model imposes a 50-second delay to allow the CR ventilation system to physically switch into isolation mode.

The time at which the operator will act to unisolate the control room and initiate filtered air makeup is a proceduralized operator decision during the course of the event. For St. Lucie Unit 1, the nominal time to unisolate the CR is assumed to be 90 minutes from the start of the event based on past experience and procedures.

1.6.4 Control Room Inleakage Sensitivity Study Control room inleakage testing identified a potential unfiltered leakage pathway into the control room envelope via the switchgear room through louver L-1 1. Separate atmospheric dispersion factors were developed for this leakage pathway as described in Section 1.8. To ensure that the most liming configuration is considered, all events were analyzed using both the intake methodology described in Section 1.6.3 as well as with unfiltered inleakage through the switchgear rooms. Results of the limiting cases are presented in the discussion of the event analyses in Section 2.0, and the applicable release-receptor pair is shown in Table 1.8.1-3.

1.6.5 Direct Shine Dose The total control room dose also requires the calculation of direct shine dose contributions from:

" the radioactive material on the control room filters,

" the radioactive plume in the environment, and

" the activity in the primary containment atmosphere.

The contribution to the total dose to the operators from direct radiation sources such as the control.

room filters, the containment atmosphere, and the released radioactive plume were calculated for the LOCA event. The LOCA shine dose contribution is assumed to be bounding for all other events. The 30-day direct shine dose to a person in the control room, considering occupancy, is provided in Table 1.6.3-2. Note that the shine dose for the LOCA event is conservatively assumed to bound the values for the remaining events.

Direct shine dose is determined from three different sources to the control room operator after a postulated LOCA event. These sources are the containment, the control room air filters, and the external cloud that envelops the control room. Per Table 6.4-2 of the UFSAR, 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 that output the nuclide activity at a given point 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 CR filter inventory is established using a case from the sensitivity study with unfiltered inleakage that produced a control room dose slightly in excess of the 5 rem TEDE dose limit to control room operators. The direct shine dose calculated due to the filter loading for this conservative unfiltered inleakage case is used as a conservative assessment of the direct shine dose contribution for all accidents.

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

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 12 of 87 ANDSOF'WARS SOLUTONSINENGINEERING for occupancy using the occupancy factors specified in RG 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.

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 St. Lucie Unit I are summarized in the following tables:

Table 1.7.2-1 - Primary Coolant Source Term Table 1.7.3-1 - Secondary Side Source Term (non-LOCA)

Table 1.7.4-1 - LOCA Containment Leakage Source Term Table 1.7.5-1. - 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).

The St. Lucie Unit No. 1 reactor core consists of 217 fuel assemblies. The full core isotopic inventory is determined in accordance with RG 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.

Sensitivity studies were performed to assess the bounding fuel enrichment and bounding burnup values. The assembly source term is based on 102% of rated power (2700 MWth x 1.02 = 2754 MWth).

For rod average burmups in excess of 54,000 MWD/MTU the heat generation rate is limited to 6.3 kw/ft. Exceptions to this burnup limit are discussed in Section 1.9. For non-LOCA events with fuel failures, a bounding radial peaking factor of 1.7 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 core inventory release fractions for the gap release and early in-vessel damage phases for the design basis LOCAs were obtained from RG 1.183, Regulatory Position 3.2, Table 2, "PWR Core Inventory Fraction Released into Containment." For non-LOCA events, the fractions of the core inventory assumed to be in the gap are consistent with RG 1.183, Regulatory Position 3.2, Table 3, "Non-LOCA Fraction of Fission Product Inventory in Gap." In some cases, the gap fractions listed in Table 3 are modified as required by the event-specific source term requirements listed in the Appendices for RG 1.183.

The following assumptions are applied to the source term calculations:

1. A conservative maximum fuel assembly uranium loading (424,160 grams) is assumed to apply to all 217 fuel assemblies in the core.

2. Radioactive decay of fission products during refueling outages is ignored.

3. 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.

NUMERICAL AST Licensing Technical Report for NA-i 101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit 1 Page 13 of 87 INEN6INCEERI4AND WSLOtUOS S ATARE Conservatisms used in the calculation of fission product inventories include the following:

• Use of ORIGEN 2.1 with revised data libraries for extended fuel bumup.

0 Use of a core thermal power corresponding to the plant design power plus 2% calorimetric uncertainty.

• Use of bounding maximum assembly and bounding core average equilibrium cycle maximum burnup.

• Use of a bounding range of average assembly enrichments.

• Use of a bounding maximum assembly uranium loading.

• Neglect of decay of fission products during refueling outages.

1.7.2 Primary Coolant Source Term The primary coolant source term for St. Lucie Unit 1 is derived from Table 11.1-1 of the UFSAR. Per the assumptions listed in Table 11.1-2 of the UFSAR, the activities given in Table 11.1-1 are based on 1% failed fuel. Table 11.1-1 of the UFSAR presents the activities in units of g.tCi/cc for 70°F water.

The density of 70°F water is 1.0 gm/cc; therefore, 1.0 gtCi/cc is equal to 1.0 gtCi/gm.

The iodine activities from UFSAR Table 11.1-1 are adjusted to achieve the Technical Specification limit of 1.0 gtCi/gm dose equivalent 1-131 using the Tech. Spec. definition of Dose Equivalent 1-131 (DE 1-131) and dose conversion factors for individual isotopes. from ICRP 30, which are equivalent to the rounded thyroid values from FGR 11 for iodine isotopes. Dose equivalent 1-131 calculated using the thyroid dose conversion factors results in higher primary coolant concentrations than the DE 1-131 determined from effective DCFs. The non-iodine species are adjusted to achieve the Technical Specification limit of 100/E-bar microcuries per gram of gross activity for non-iodine activities.

The dose conversion factors for inhalation and submersion are from Federal Guidance Reports Nos. 11 and 12 respectively. The final adjusted primary coolant source term is presented in Table 1.7.2-1, "Primary Coolant Source Term.'?

1.7.3 Secondary Side Coolant Source Term Secondary coolant system activity is limited to a value of < 0.10 ptCi/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 pCi/gm. Thus, the secondary side iodine activity is 1/10 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 LOCA Containment Leakage Source Term Per Section 3.1 of RG 1.183, the inventory of fission products in the St. Lucie Unit 1 reactor core available for release to the containment are 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, for the DBA LOCA, all fuel assemblies in the core are assumed to be affected and the core average inventory is used. -

NUMERICAL AST Licensing Technical Report for NAI-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 14 of 87 SOUTIONSINeNGINEERING ANiD SOR~WARE During a LOCA, all of the fuel assemblies are assumed to fail; therefore, the source term is based on an "average" assembly with a core average burnup of 45,000 MWD/MTU and an average assembly power* of 12.691 MWth. The minimum fuel enrichment is based on an historical minimum of 3.0 w/o and the maximum fuel enrichment is the Tech. Spec. maximum value of 4.5 w/o. It is conservatively assumed that a maximum assembly uranium mass of 424,160 gm applies to all of the fuel assemblies.

• Average assembly power = (2700 MWh)(1.02)(1 / 217 assemblies) = 12.691 MWh / assembly The ORIGENruns used cross section libraries that correspond to PWR extended burnup fuel. Decay time between cycles is conservatively ignored. For each nuclide, the boundingactivity for the allowable range of enrichments is determined.

The LOCA source term is presented in Table 1.7.4-1,"LOCA Containment Leakage Source Term."

1.7.5 Fuel Handling Accident Source Term The fuel handling accident for St. Lucie Unit I 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 RG 1.183, the source term methodology for the Fuel Handling Accident is similar to that used for developing the LOCA containment leakage 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 1.7 (total integrated peaking factor per St. Lucie Unit 1 COLR), is applied in determining the inventory of the damaged rods.

The LOCA containment leakage source term is based on the activity of 217 fuel assemblies at an average assembly power of 12.691 MWh. Thus, based on the methodology specified in RG 1.183, the fuel handling accident source term is derived by applying a factor of 1.7/217 to the LOCA containment leakage source term, where 1.7. is the radial peaking factor. To ensure that the "bounding" assembly is identified, the activity of a peak burnup assembly (62,000 MWD/MTU), at both 3.0 w/o and 4.5 w/o, is determined and compared to the source term derived from the LOCA data. For each nuclide, the bounding activity for the allowable range of enrichments and discharge exposure is determined.

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

1.8 Atmospheric Dispersion (X/Q) 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-6331, 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.

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 15 of 87 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 St. Lucie Unit 1 that has been annotated to highlight the release and receptor point locations described above. All releases are taken as ground releases per guidance provided in RG 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. Direction values are corrected for "plant North" offset from "true North" by 280 41' 56".

Table 1.8.1-2, "Onsite Atmospheric Dispersion Factors (XI/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 and for switchgear room louver L- 11 that apply to the various accident scenarios for onsite control room dose consequence analyses. For the intakes, values are presented for the unfavorable intake prior to control room isolation, the midpoint between the intakes during isolation, as well as values for the favorable intake following manual restoration of filtered control room makeup flow. These values include credit for dilution where allowed by Reg. Guide 1.194. Based on the layout of the site, the only cases that may take credit for dilution are when the releases are from the plant vent stack. However, dilution is not credited during the time period when the control room intakes are isolated for these cases.

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 three modes of control room ventilation.

.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 selected for use in the 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 height is 184 feet (from the plant stack). From Section 1.3.2 of RG 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 St. Lucie plant, the elevation of the top of the Unit 1 containment is given as 225.5'0 feet. 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 the ARCON96 onsite X/Q cases. This area of 1565 m2 is calculated to be conservatively small in that the height used in the area calculation is from the highest

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 16 of 87 SOLUTIONS INENGINEERING AND SOFTWARE roof elevation of a nearby building to the elevation of the bottom of the containment dome. The containment height used in the building wake term is the containment top elevation minus the bottom (grade) elevation of 19 feet.

The tower height at which the wind speeds are measured is 10.0 meters. There are zero hours of calms in the joint frequency distribution data. This low number of calm hours is due to the positioning of the St. Lucie plant and its proximity to the Atlantic Ocean. The highest windspeed category is classified in RG 1.23, "Onsite Meteorological Programs," February 1972, as greater than 24 mph, however the PAVAN code requires that the maximum speed for each category be input. Therefore, the 30-mph value is chosen as the upper limit on the fastest windspeed category because the raw meteorological data showed that there were no hours with windspeeds faster than 30 mph.

1.8.3 Meteorological Data Meteorological data over a five-year period (1996 through 2001) are used in the development of the new X/Q factors used in the analysis. The St. Lucie Plant, Unit No. 1, Meteorological Monitoring Program, complies with RG 1.23; "Onsite Meteorological Programs," February 1972. The Meteorological Monitoring Program is described in Section 2.3.3 of the St. Lucie Plant Unit No. 1 UFSAR.

For the onsite X/Q determinations, the five years include the last six months of 1996; all of 1997, 1998, and 1999; the first six months of 2000; and all of 2001. The last six months of 2000 data are not included because of the poor quality of the raw data (i.e. significant portions of time with unrecorded data). Since the poor data period occurred in the middle of the time period under consideration, and that 5 years' worth of data is desired, the last six months of 1996 data are included at the beginning of the meteorological data file. For the offsite X/Q determinations, the five years are from 1997 through 2001.

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. Since all of the St. Lucie Unit 1 cases use the same meteorological data file, all of the cases in this analysis have the same data recovery rate. The ARCON96 files present the number of hours of data processed as 43,454 and the number of missing data hours as 2,108. This yields a meteorological data recovery rate of 95. 1%. 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 RG 1.23 requires a 90% data recovery threshold for measuring and capturing meteorological data. Clearly, the 95.1% valid meteorological data rate for the cases in this analysis exceeds the 90% data recovery limit set forth by RG 1.23. With a data recovery rate of 95.1% 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 St. Lucie site.

The meteorological data were also provided in annual joint frequency distribution format for 1997 through 2001. 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 RG 1.23, with the exceptions of a category for the variable 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

NUMERICAL AST Licensing Technical Report for NA-i1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 17 of 87 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.

An additional process is performed on the met data used for the ARCON96 runs to'determine the average air temperature swing over a 24-hour period for the five years' worth of data. The yearly data is combined so that the dates match the data used for the ARCON96 met file. That is the last 6 months of 1996 are included and the last 6 months of 2000 are omitted as previously explained. Any data determined to be invalid is excluded. The average air temperature range over the five years of meteorological data is calculated to be a 9.60 F temperature swing over any 24-hour period. A median value is also calculated. The median 24-hour period temperature swing value is 8.7°F. The higher value is used to support determining the leakage rate from the RWST.

1.9 Consideration of High Burnup Fuel Footnote 11 of Reg. Guide 1.183 states that the gap fractions listed in Table 3 of the Reg. Guide are acceptable for fuel with a peak burnup of 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU. The St. Lucie Unit 1 core design allows for a maximum of 150 rods which may exceed these burnup limits. With 217 assemblies in the core, and 176 fuel rods per assembly, the number of high burnup rods represents only [(150)/(217)(176)] x 100 = 0.393% of the core.

Previous AST submittals (e.g. Byron and Braidwood, Fort Calhoun) addressed the high bumup issue by doubling the gap release fractions in combination with the peaking factor at the limiting value for affected rods. Although the number of rods exceeding the burnup/linear heat rate for St. Lucie Unit 1 will be limited to 150 rods, the AST analysis will assume a total of 1408 rods, equivalent of 8 fuel assemblies, to be exceeding the high burnup limit. The release fractions for all rods in 8 assemblies will thus be doubled and a peaking factor of 1.7 will be applied for the source term. This approach would impact [8/217] x 100 = 3.687% of the core, which is nearly 10 times larger than the number of affected rods. Doubling the gap release fraction of 3.687% of the core would yield a core-wide high bumup adjustment factor of:

High Bumup Adjustment = [(2)(0.03687)(Gap Frac.) + (I-0.03687)(Gap Frac.)] = (1.03687)(Gap Frac.)

This factor is applied to the release fractions for all events in which fuel damage causes the inventory of the fuel rod gaps to be released into the reactor coolant. For the Fuel Handling Accident, in which 100% of the rods in the dropped assembly are assumed to be damaged, high burnup is addressed by increasing the gap release fraction of the entire assembly by a factor of two.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit I Page 18 of 87 AND OF'iMARE SOLUnONSINENGINEERING 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 15.4.1 of the UFSAR.

2.1.2 Compliance with RG 1.183 Regulatory Positions The revised LOCA dose consequence analysis is consistent with the guidance provided in RG 1.183, Appendix A, "Assumptions for Evaluating the Radiological Consequences of a LWR Loss-of-Coolant 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 RG 1.183, Regulatory Position 3.1, at 102% of core thermal power and 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 RG 1.183.

2. Regulatory Position 2 - Per Section 6.2.6.1, the long term recirculation sump pH remains greater than 7.0. 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 iodine. 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-vessei 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 2.89 hr-1 . This removal is credited in both the sprayed and unsprayed regions of containment.

A natural deposition removal coefficient of 0.1 hr-' is assumed for all aerosols in the unsprayed region of containment as well as in the sprayed region after spray is terminated at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Industry Degraded Core Rulemaking 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 hr-1 is used in the analyses based upon NRC approval of this value in the safety evaluations for the Harris Nuclear Plant License Amendment No. 107 in October 2001 (ADAMS Accession No. ML012830516) and the

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit I Page 19 of 87 SMUnONSIN ENGINEERING ANDSOFTARE Kewaunee Nuclear Power Plant License Amendment No. 166 in March 2003 (ADAMS Accession No. ML030210062).

No removal of organic iodine by natural deposition is assumed.

5. Regulatory Position 3.3 - Containment spray provides coverage to 86% of the containment. Therefore, the St. Lucie Unit 1 containment building atmosphere is not considered to be a single, well-mixed volume. A mixing rate of two turnovers of the unsprayed region per hour is assumed.

Reg. Guide 1.183 and the SRP state that the elemental iodine spray removal coefficient should be set to zero when a decontamination factor (DF) of 200 is reached for elemental iodine. The particulate spray removal coefficient should be reduced by a factor of 10. when a DF of 50 is reached for the aerosol.

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:'

• No containment spray

• No 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 0.017917 hours (64.5 seconds)

• No surface deposition

• No decay

• No containment leakage The second RADTRAD-NAI case showed that a DF of 200 for elemental iodine was reached at a time greater than 3.02 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

A third RADTRAD-NAI case determined the time required to reach a DF of 50 for aerosol based on the peak aerosol mass from the first RADTRAD-NAI case. The third RADTRAD-NAI case included:

• Containment sprays actuated at 0.017917 hours (64.5 seconds)

• Aerosol surface deposition credited

• No decay

• No containment leakage The third RADTRAD-NAI case showed that a DF of 50 was reached at a time greater than 2.60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLI** TIONS, INC. St. Lucie Unit 1 Page 20 of 87 SOLUTONSIN ENINEERING ANDSOFTA

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

7. Regulatory Position 3.5 - This position relates to suppression pool scrubbing in BWRs, which is not applicable to St. Lucie Unit No. 1..

8. Regulatory Position 3.6 - This position relates to activity retention in ice condensers, which is not applicable to St. Lucie Unit No. 1.

9. Regulatory Position 3.7 - A containment leak rate of 0.5% per day of the containment air is assumed for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the containment leak rate is reduced to 0.25% per day of the containment air.

10. Regulatory Position 3.8 - Routine containment purge is considered in this analysis. 100% of the radionuclide inventory of the RCS is released instantaneously at the beginning of the event. The containment purge flow is 42000 cfm and is assumed to be isolated after 5 seconds. No filters are credited.

11. Regulatory Position 4.1 - Leakage from containment collected by the secondary containment is processed by ESF filters prior to release from the plant stack.

12. Regulatory Position 4.2 - Leakage into the secondary containment is assumed to be released dAirectly to the environment as a ground level release prior to drawdown of the secondary containment at 310 seconds.

13. Regulatory Position 4.3 - SBVS is credited as being capable of maintaining the Shield Building Annulus at a negative pressure with respect to the outside environment considering the effect of high windspeeds and LOCA heat effects on the annulus as described in UFSAR Section 6.2: No exfiltration through the concrete wall of the Shield Building is expected to occur.

14. Regulatory Position 4.4 - No credit is taken for dilution in the secondary containment volume.

15. Regulatory Position 4.5 - 9.6% of the primary containment leakage is assumed to bypass the secondary containment. This bypass leakage is released as a ground level release without credit for filtration.

16. Regulatory Position 4.6 - The SBVS is credited as meeting the requirements of RG 1.52 and Generic Letter 99-02. The filters in the SBVS ventilation system are credited at 99%

efficiency for particulates and 95% for both elemental and organic iodine.

17. 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.

18. Regulatory Position 5.2 - Leakage from the ESF system is taken as two times the maximum potential recirculation loop leakage outside containment provided in UFSAR Table 15.4.1-2.

The 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 RWT is also considered separately as two times 1 gpm, which is bounding value based upon RCS leakage monitoring documented in the Control Room database. RWT leakage is assumed to begin at the start of recirculation and continue for the remainder of the 30-day duration.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICTIONS, INC. St. Lucie Unit 1 Page 21 of 87

19. 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.

20. Regulatory Position 5.4 - A flashing fraction of 3.4% was calculated based upon peak pressure conditions in containment. However, consistent with Regulatory Position 5.5, the flashing fraction for ECCS leakage is assumed to be 10%. This ECCS leakage enters the Reactor Auxiliary Building. For ECCS leakage back to the RWT, the analysis demonstrates that the temperature of the leaked fluid will cool below 212'F prior to release into the tank.

21. 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 RWT, the sump and RWT pH history and temperature are used to evaluate the amount of iodine that enters the RWT air space.

22. Regulatory Position 5.6 - For ECCS leakage into the auxiliary building, the form of the released iodine is 97% elemental and 3% organic. An ECCS area ventilation system filter efficiency of 95% is assumed for both elemental and organic iodine. The ECCS area ventilation system meets the requirements of RG 1.52 and Generic Letter 99-02. There is no credit for hold-up or dilution in the Reactor Auxiliary Building.

The temperature and pH history of the sump and RWT are considered in determining the radioiodine available for release and the chemical form. Credit is taken for dilution of the activity in the RWT.

23. Regulatory Position 6 - This position relates to MSSV leakage in BWRs, which is not applicable to St.

Lucie Unit No. 1.

24. Regulatory Position 7 - Containment purge is not considered as a means of combustible gas or pressure control in this analysis.

2.1.3 Methodology For this event, the Control Room ventilation system cycles through three modes of operation (the operational modes are summarized in Table 1.6.3-1). Inputs and assumptions fall into three main categories: Radionuclide Release Inputs, Radionuclide Transport Inputs, and Radionuclide Removal Inputs.

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 SIS signal is actuated when the appropriate setpoint 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.

NUMERICAL. AST Licensing Technical Report for NAI-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 22 of 87 ANDSOFTAIRE 5OýMSI1NS ENGINeE~RNG Release Inputs The core inventory of the radionuclide groups utilized for this event is based on RG 1.183, Regulatory Position 3.1, at 102% of core thermal power 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.

From TS Surveillarice Requirement 3.6.1.1, the initial leakage rate from containment is 0.5% of the containment air per day. Per RG 1.183, Regulatory Position 3.7, the primary containment leakage rate is reduced by 50% at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the LOCA to 0.25% /day based on the post-LOCA primary containment pressure history.

The ESF leakage to the Auxiliary Building is assumed to be 4510 cc/hr, based upon two times the current licensing basis value of 2255 cc/hr. The leakage is conservatively assumed to start at 20 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. Dilution and holdup of the ECCS leakage in the Reactor Auxiliary Building are not credited.

The ECCS backleakage to the RWT is initially assumed to be 2 gpm based upon doubling the current bounding value of 1 gpm. This leakage is assumed to start at 20 minutes into the event when recirculation in the sump starts and continue throughout the 30-day period. Based on sump pH history, the iodine solution is assumed to all be nonvolatile. However, when introduced into the acidic solution of the RWT inventory, there is a potential for the particulate iodine to convert into the elemental form. The fraction of the total iodine in the RWT which becomes elemental is both a function of the RWT pH and the total iodine concentration. The amount of elemental iodine in the RWT fluid which then enters the RWT air space is a function of the temperature-dependent iodine partition coefficient.

The time-dependent concentration of the total iodine in the RWT (including stable iodine) was determined from the tank liquid volume and leak rate. This iodine concentration ranged from a minimum value of 0 at the beginning of the event to a maximum value of 4.22E-05gm-atom/liter at 30 days (see Table 2.1-4).

Based upon the backleakage of sump water, the RWT pH slowly increases from an initial value of 4.5 to a maximum pH of 5.01 at 30 days (see Table 2.1-3). Using the time-dependent RWT pH and the total iodine concentration in the RWT liquid space, the amount of iodine converted to the elemental form was determined using guidance provided in NUREG-5950. This RWT elemental iodine fraction ranged from 0 at the beginning of the event to a maximum of 0.17 (see Table 2.1-6).

The elemental iodine in the liquid region of the RWT is assumed to become volatile and to partition between the liquid and vapor space in the RWT based upon the partition coefficient for elemental iodine as presented in NUREG-5950. A GOTHIC model was used to determine the RWT temperature as a function of time (see Table 2.1-5) which was then used to calculate the partition coefficient shown in Table 2.1-7.

The RWT 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 RWT, the release of the activity from the vapor space within the RWT is calculated based upon the displacement of air by the incoming leakage. The elemental iodine flow rate from the RWT is equal to the air flow rate times the elemental iodine concentration in the RWT vapor space.

For the organic iodine flow, the same approach was used with an organic iodine fraction of 0.0015 from Reg. Guide 1.183 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 RWT. Therefore, the total iodine flow is the sum of the elemental and organic iodine flow rates.

The time dependent iodine release rate presented in Table 2.1-8 in then applied to the entire iodine inventory (particulate, elemental and organic) in the containment sump. The iodine released via the RWT air vent to the environment was effectively set to 100% elemental (the control room filters have the same efficiency for all forms of iodine).

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 23 of 87

.S......IN.ENGI ERINGMD

.. SOFaWARE 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 (at an assumed 1.0 microcuries per gram DE I-131and 100/E-bar gross activity) is available for release via this pathway. The release is conservatively modeled for 5 seconds at 42,000 cfm until isolation occurs.

The release point for each of the above sources is presented in Table 1.8.1-3.

Transport Inputs During the LOCA event, the initial containment purge is released through the plant stack with no filtration.

Leakage into the secondary containment is assumed to be released directly to the environment as a ground level release prior to drawdown of the secondary containment at 310 seconds. Activity subsequently collected by the SBVS is assumed to be a filtered release from the plant stack with a filter efficiency of 99% for particulates and 95% for both elemental and organic iodine. The activity that bypasses the SBVS is released unfiltered to the environment via a ground level release from containment. ECCS leakage into the Auxiliary Building is modeled as a release via the Auxiliary Building. For this release, the ECCS area ventilation system is credited with a particulate removal efficiency of 99% and elemental and organic iodine efficiencies of 95%. The activity from the RWT is modeled as an unfiltered ground level release from the RWT.

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

" Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 920 cfm of unfiltered fresh air and an assumed value of 500 cfm of unfiltered inleakage.

" After the start of the event, the Control Room is isolated due to a CIAS as a result of a high containment pressure signal. A 50-second delay is applied to account for the time to reach the signal, the diesel generator start time, damper actuation time. After isolation, the air flow distribution consists of 0 cfm of makeup flow from the outside, 500 cfm of unfiltered inleakage, and 2000 cfm of filtered recirculation flow.

" At 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event, the operators are assumed to initiate makeup flow from the outside to the control room. During this operational mode, the air flow distribution consists of up to 450 cfm of filtered makeup flow, 500 cfm of unfiltered inleakage, and 1550 cfm of filtered recirculation flow.

• The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulates, 95% for elemental iodine, and 95% for organic iodine.

LOCA Removal Inputs Reduction of the airborne radioactivity in the containment by natural deposition is credited. The natural deposition removal coefficient for elemental iodine is calculated per SRP 6.5.2 as 2.89 hr1. A natural deposition removal coefficient of 0.1 hr 1 is assumed for all aerosols in the unsprayed region and in the sprayed region after spray flow is secured at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. No removal of organic iodine by natural deposition is assumed.

Containment spray provides coverage to 86% of the containment. Therefore, the St. Lucie Unit 1 containment building atmosphere is not considered to be a single, well-mixed volume. A mixing rate of two turnovers of the unsprayed region per hour is assumed.

NUMERICAL AST Licensing Technical Report for NAI-i 101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit 1 Page 24 of 87 SOLrTIONS INENGINeERING ANDSOF'WARE The elemental spray coefficient is limited to 20 hr1 per SRP 6.5.2. This coefficient is reduced to 0 when an elemental decontamination factor (DF) of 200 is reached. Based upon the elemental iodine removal rate of 20 hr1, the DF of 200 is conservatively computed to occur at 3.02 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

The particulate iodineremoval rate is reduced by a factor of 10 when a DF of 50 is reached. Based upon the calculated iodine aerosol removal rate of 6.43 hr-1 , the DF of 50 is conservatively computed to occur at 2.60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />.

2.1.5 Radiological Consequences 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 intake.

When the Control Room Ventilation System is in normal mode, the most limiting X/Q corresponds to the worst air intake to the control room. When the ventilation system is isolated at 50 seconds, the limiting X/Q corresponds to the midpoint between the two control room air intakes. The operators are assumed to reopen the most favorable air intake at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The X/Q values for the various combinations of release points and receptor locations are presented in Table 1.8.1-2. Table 1.8.17 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.

The EAB and LPZ dose consequences are determined using the X/Q factors provided Table 1.8.2-1 for the appropriate time intervals. For the EAB dose calculation, the X/Q factor for the zero to two-hour time interval is assumed for all time periods. Using the zero to two-hour X/Q factor provides a more conservative determination of the EAB dose, because the X/Q factor for this time period is higher than for any other time period.

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, Version 5.05, Grove Engineering, is used 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 (for example, see Duane Arnold Energy Center submittal dated October 19, 2000 and associated NRC Safety Evaluation dated July 31,2001.)

The post accident doses are the result of five distinct activity releases:

1. Containment leakage via the secondary containment system.

2. Containment leakage bypassing the secondary containment

3. ESF system leakage into the Auxiliary Building.

4. ESF system leakage into the RWT.

5. Containment Purge at event initiation.

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 ESF.

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

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 25 of 87 SOUIOftS IN ENGINeERING ANO SOFTWARkE 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 2.1-4, 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.

NUMERICAL AST Licensing Technical Report for NA-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 26 of 87 SOLUMONSIN ENGINEERING NOSOflAR 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 15.4.3 of the UFSAR, which specifies that 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 maintenance hatch open, and an assembly drop 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. RG 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 RG 1.183 Regulatory Positions The FHA dose consequence analysis is consistent with the guidance provided in RG 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 15.4.3.3.

2. Regulatory Position 1.2 - The fission product release from the breached fuel is based on Regulatory Positions 3.1 and 3.2 of RG 1.183. Section 1.7 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 specified by Table 3 of RG 1.183. Gap release fractions are doubled to account for high burnup fuel rods as described in Section 1.9. This activity is assumed to be released 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 500 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. Due to a conflicting requirement that the overall iodine decontamination factor be equal to 200 (which results in an elemental iodine decontamination factor of 285), an additional set of FHA cases were run with an elemental iodine decontamination factor of 285.

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.

NUMERICAL AST Licensing Technical Report for NAI-i1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 27 of 87 ANOSOFIWARE S(.UTIONS INENGINEERING

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 maintenance 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 dose consequence analysis of the FHA are provided in Table 2.2-1. It is assumed that the fuel handling accident occurs at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown of the reactor per TS 3.9.3.

100% of the gap activity specified in Table 3 of RG 1.183 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. 100% 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. Iodine released from the damaged fuel assembly is assumed to be composed of 99.85% elemental and 0.15% organic. All activity released from the pool is 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 I of Appendix B to RG 1.183.

Section 1.7 discusses the development of the FHA source term, which is listed in Table 1.7.5-1. The analysis includes a decay time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before the beginning of fuel movement. Since the FHA source term presented in Table 1.7.5-1 does not include this decay time, it is accounted for in the RADTRAD-NAI model.

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

Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 920 cfm of unfiltered fresh air and an assumed value of 500 cfm of unfiltered inleakage.

After the start of the event, the Control Room is isolated due to a high radiation reading in the Control Room ventilation system. A 50-second delay is applied to account for diesel generator start time, damper actuation time, instrument delay, and detector response time. After isolation, the air flow distribution consists of 0 cfm of makeup flow from the outside, 500 cfm of unfiltered inleakage, and 2000 cfm of filtered recirculation flow.

At 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event, the operators are assumed to initiate makeup flow from the outside to the control room. During this operational mode, the air flow distribution consists of up to 450 cfm of filtered makeup flow, 500 cfm of unfiltered inleakage, and 1550 cfm of filtered recirculation flow.

The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, 95% for elemental iodine, and 95% for organic iodine.

1 NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATION,. INC. St. Lucie Unit I Page 28 of 87 2.2.4 Radiological Consequences 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 intake.

For the FHA event, the release from the FHB is the closest point to the control room. The containment release corresponds to the containment maintenance hatch.

When the Control Room Ventilation System is in normal mode, the most limiting X/Q corresponds to the worst air intake to the control room. When the ventilation system is isolated at 50 seconds, the limiting X/Q corresponds to the midpoint between the two control room air intakes. The operators are assumed to reopen the most favorable air intake at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The X/Q values for the various combinations of release points and receptor locations are presented in Tablel.8:1-2. Table 1.8.1-3 presents the Release-Receptor pairs applicable to the control room dose for the different modes of control room operation during this event.

The EAB and LPZ dose consequences are determined using the X/Q factors provided Table 1.8.2-1 for the appropriate time intervals. For the EAB dose calculation, the X/Q factor for the zero to two-hour time interval is assumed for all time periods. Using the zero to two-hour X/Q factor provides a more conservative determination of the EAB dose, because the X/Q factor for this time period is higher than for any other time period.

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.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 29 of 87 S5.UflONSINENGINEERING AND SOFTWARE 2.3 Main Steamline Break (MSLB) 2.3.1 Background This event consists of a double-ended break of one main steam line either inside or outside of containment.

Allowable fuel failure rates due to DNB and fuel centerline melt are determined for both break locations based upon the dose limits specified in Table 6 of RG 1.183. The affected steam generator (SG) rapidly depressurizes and releases the initial contents of the SG to either the environment or the containment. Plant cool down is achieved via the remaining unaffected SG. This event is described in the UFSAR, Section 15.4.6.

2.3.2 Compliance with RG 1.183 Regulatory Positions The MSLB dose consequence analysis followed the guidance provided in RG 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 - The total core inventory of the radionuclide groups utilized for determining the source term for this event is based on RG 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 1.7 is applied. The fraction of fission product inventory in the gap available for release due to DNB is consistent with Regulatory Position 3.2 and Table 3 of RG 1.183. Gap release fractions are increased by a factor of 1.03687 to account for high burnup fuel rods as described in Section 1.9. For fuel centerline melt, the guidance provided in RG 1.183, Appendix H, Regulatory Position 1 is used to determine the release.

3. Regulatory Position 2 - Fuel damage is assumed for this event. It was determined that the activity released from the damaged fuel will exceed that released by the two iodine spike cases; therefore, the two iodine spike cases were not analyzed.

4. Regulatory Position 3 - The activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant.

5. Regulatory Position 4 - Iodine releases from the faulted SG and the unaffected SG to the environment (or containment) are assumed to be 97% elemental and 3% organic. These fractions apply as a result of fuel damage.

6. Regulatory Position 5.1 -.The primary-to-secondary leak rate is apportioned between the SGs as specified by proposed TS 6.8.4.1 (0.5 gpm total, 0.25 gpm to any one SG). Thus, the tube leakage is apportioned equally between the two SGs.

7. Regulatory Position 5.2 - The density used in converting volumetric leak rates to mass leak rates is based upon RCS conditions, consistent with the plant design basis.

8. Regulatory Position 5.3 - The primary-to-secondary leakage is assumed to continue until after shutdown cooling has been placed in service and the temperature of the RCS is less than 212'F.

9. 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.

NUMERICAL AST Licensing Technical Report for NA-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 30 of 87 SO.UT1ONS INENGINEGRING ANDSORiWARE

10. Regulatory Position 5.5.1 - In the faulted SG, all of the primary-to-secondary leakage is assumed to flash to vapor and be released to the environment (MSLB outside of containment) or the containment (MSLB inside of containment) with no mitigation. For the unaffected steam generator 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.

11. Regulatory Position 5.5.2 - The postulated leakage that immediately flashes to vapor is assumed to rise through the bulk water of the SG 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 SG bulk water is credited.

12. Regulatory Position 5.5.3 - All leakage that does not immediately flash is assumed to mix with the bulk water.

13. 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 SG is limited by the moisture carryover from the SG. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the SG carryover rate of less than 1%. No reduction in the release is assumed from the faulted SG.

14. Regulatory Position 5.6 - Steam generator tube bundle uncovery in the intact SG is postulated for up to 45 minutes following a reactor trip for St. Lucie Unit 1. During this period, the fraction of primary-to-secondary leakage which flashes to vapor is assumed to rise through the bulk water of the SG 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.

2.3.3 Other Assumptions I. The initial RCS activity is assumed to be at the TS limit of 1.0 pCi/gm Dose Equivalent 1-13 1and 100/E-bar gross activity. The initial SG activity is assumed to be at the TS 3.7.1.4 limit of 0.1 pCi/gm Dose Equivalent 1-131.

2. The steam mass release rates for the intact SG are provided in Table 2.3-2. These values are

• based upon a cooldown rate of 100 °F/hr until the RCS temperature reaches 300 'F. This temperature is maintained until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when Shutdown Cooling is assumed to become available. The cooldown is then continued at a rate of 38 °F/hr until the RCS temperature is reduced to 212 'F at 10.32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />. No credit is taken for energy removal by the Shutdown cooling system.

3. The RCS fluid density used to convert the primary-to-secondary leakage from a volumetric

• flowrate to a mass flow rate is consistent with the RCS cooldown rate applied in the generation of the secondary steam releases. The high initial cooldown rate conservatively maximizes the fluid density. The SG tube leakage mass flow rate is provided in Table 2.3-3.

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 MSLB or from the safety injection system).

5. For the purposes of determining the iodine concentration of the SG secondary, the mass in the unaffected SG is assumed to remain constant throughout the event. However, it is also assumed that operator action is taken to restore water level above the top of the tubes in the unaffected steam generator within a conservative time of one hour following a reactor trip.

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 31 of 87 SOMflON5 IýN ENCNRING"AD SOFMVAM

6. All secondary releases are postulated to occur from the ADV with the most limiting atmospheric dispersion factors. Releases from containment through the SBVS are assumed to be released from the plant stack with a filter efficiency of 99% for particulates and 95% for '

both elemental and organic iodine. The activity that bypasses the SBVS is released unfiltered to the environment via a ground level release from containment.

7. The initial leakage rate from containment is 0.5% of the containment air per day. This leak rate is reduced by 50% after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.25% /day. 9.6% of the containment leakage is assumed to bypass the SBVS filters.

8. For the MSLB inside of containment, natural deposition of the radionuclides is credited consistent with the LOCA methodology presented in Section 2.1.3. Containment sprays are not credited.

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 consists of two cases; one case is based upon a double-ended break of one main steam line outside of containment, and the second case is based upon a double-ended break of one main steam line inside of containment. The primary difference between these two models is the transport of the primary-to-secondary leakage through the affected steam generator. Upon a MSLB, the affected SG rapidly depressurizes. The rapid secondary depressurization causes a reactor power transient, resulting in a reactor trip. Plant cooldown is achieved via the remaining unaffected SG.

The analysis for both cases 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 1.7.4-1 with adjustments for the fraction of damaged fuel, the non-LOCA fission product gap fractions from Table 3 of RG 1.183, high burnup fuel, a radial peaking factor of 1.7. All noble gases associated with this leakage are assumed to be released directly to the environment. For the break outside containment, primary-to-secondary leakage into the affected steam generator is also assumed to directly enter the atmosphere. For the break inside containment, the affected steam generator leakage is released into containment. All primary-to-secondary leakage is assumed to continue until the faulted steam generator is completely isolated at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Primary-to-secondary tube leakage is also postulated to occur in the unaffected SG for both cases. This activity is diluted by the contents of the steam generator and released via steaming from the ADVs until the RCS is cooled to 212'F. In addition, the analysis of both cases assumes that the initial iodine activity of both SGs is released directly to the environment. The entire contents of the faulted steam generator is released immediately, while the intact steam generator release occurs during the RCS cooldown. The secondary coolant iodine concentration is assumed to be the maximum value of 0.1 gtCi/gm DE 1-131 permitted by Tech. Specs. These release assumptions are consistent with the requirements of RG 1.183.

Allowable levels of fuel failure for DNB and fuel centerline melt are determined for both the MSLB outside of containment and the MSLB inside of containment. These allowable fractions are based on the dose limits specified in Table 6 of RG 1.183. The activity released from the fuel that is assumed to experience DNB is based on Regulatory Positions 3.1, 3.2, and Table 3 of RG 1.183. Theactivity released from the fuel that is assumed to experience fuel centerline melt is based on Regulatory Position 1 of Appendix H to RG 1.183.

For this event, the Control Room ventilation system cycles through three modes of operation (the operational modes are summarized in Table 1.6.3-1):

Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 920 cfm of unfiltered fresh air and an assumed value of 500 cfm of unfiltered inleakage.

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 32 of 87 ANDSOFMMAN SOfONS INENGIN~eRrNG~

After the start of the event, the Control Room is isolated due to a high radiation reading in the Control Room ventilation system. A 50-second delay is applied to account for diesel generator start time, damper actuation time, instrument delay, and detector response time. After isolation, the air flow distribution consists of 0 cfm of makeup flow from the outside, 500 cfm of unfiltered inleakage, and 2000 cfm of filtered recirculation flow.

At 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event, the operators are assumed to initiate makeup flow from the outside to the control room. During this operational mode, the air flow distribution consists of up to 450 cfm of filtered makeup flow, 500 cfm of unfiltered inleakage, and 1550 cfm of filtered recirculation flow.

• The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, 95% for elemental iodine, and 95% for organic iodine.

2.3.5 Radiological Consequences 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 intake.

For the MSLB event, all secondary releases are from the closest ADV. X/Qs for containment releases via the SBVS are from the plant stack, and containment releases which bypass the SBVS correspond to the nearest feedwater line which penetrates containment.

When the Control Room Ventilation System is in normal mode, the most limiting XIQ corresponds to the worst air intake to the control room. When the ventilation system is isolated at 50 seconds, the limiting X/Q corresponds to the midpoint between the two control room air intakes. The operators are assumed to reopen the most favorable air intake at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The X/Q 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 for the different modes of control room operation during this event.

The EAB and LPZ dose consequences are determined using the X/Q factors provided Table 1.8.2-1 for the appropriate time intervals. For the EAB dose calculation, the X/Q factor for the zero to two-hour time interval is assumed for all time periods. Using the zero to two-hour X/Q factor provides a more conservative determination of the EAB dose, because the X/Q factor for this time period is higher than for any other time period.

The radiological consequences of the MSLB Accident are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. Cases for MSLB inside and outside of containment with DNB and FCM fuel failure are analyzed. As shown in Table 2.3-4, the results of all four cases for EAB dose, LPZ dose, and Control Room dose are within the appropriate regulatory acceptance criteria.

NUMERICAL AST Licensing Technical Report for NAI- 1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 33 of 87 SOtUlOS ANDSOFMhARU eNGINCERIN6 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 St. Lucie Unit No. 1 SGTR event. This event is described in Section 15.4.4 of the UFSAR.

2.4.2 Compliance with RG 1.183 Regulatory Positions The SGTR dose consequence analysis followed the guidance provided in RG 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 - The total core inventory of the radionuclide groups utilized for determining the source term for this event is based on RG 1.183, Regulatory Position 3.1. No fuel damage is postulated to occur for the St. Lucie Unit No. 1 SGTR event.

2. Regulatory Position 2 - No fuel damage is postulated to occur for the St. Lucie Unit No. 1 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, Fig. 3.4-1 value of 60.0 tCi/gm DE 1-131. 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 TS 3.4.8 value of 1.0 gaCi/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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. 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 apportioned between the SGs as specified by proposed change to TS 6.8.4.1 (0.5 gpm total, 0.25 to any one SG). Thus, the tube leakage is apportioned equally between the two SGs.

8. Regulatory Position 5.2 - The density used in converting volumetric leak rates to mass leak rates is based upon RCS conditions, consistent with the plant design basis.

9. Regulatory Position 5.3 - The primary-to-secondary leakage is assumed to continue until after shutdown cooling has been placed in service and the temperature of the RCS is less than 212'F. The current Licensing Basis for the termination of the affected SG activity release states that the affected SG is isolated within 30 minutes by operator action. This isolation terminates releases from the affected SG, while primary-to-secondary leakage continues to provide activity for release from the unaffected SG.

10. Regulatory Position 5.4 - The release of fission products from the secondary system is evaluated with

NUMERICAL AST Licensing Technical Report" for NAI-i1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 34 of 87

.SOtLOTIONS AND S*OFTWNAR*E IN ENGINRERINO the assumption of a coincident loss of offsite power (LOOP).

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 through the SGTR is assumed to flash to vapor based on the thermodynamic conditions in the reactor and secondary. For the unaffected steam generator used for plant cooldown, 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 SG, enter the steam space, and be immediately released to the environment with no mitigation; i.e., no reduction for scrubbing within the SG bulk water is credited.

Appendix E, Regulatory Position 5.5.3 - All of the SG tube leakage and ruptured tube flow that does not flash is assumed to nix 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 SGs is limited by the moisture carryover from the SGs. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the SG carryover rate of less than 1%.

Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery in the intact SG is postulated for up to 45 minutes following a reactor trip for St. Lucie Unit 1. During this period, the fraction of primary-to-secondary leakage which flashes to vapor is assumed to rise through the bulk water of.the SG 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. This evaluatiOn assumes that the RCS mass remains constant throughout the event

2. For the purposes of determining the iodine concentrations, the SG mass is assumed to remain constant throughout the event. However, it is also assumed that operator-action is taken to restore water level above the top of the tubes in the unaffected steam generator within a conservative time of one hour followinga reactor trip.

3. Data used to calculate the iodine equilibrium appearance rate are provided in Table 2.4-4, "Iodine Equilibrium Appearance Assumptions." The iodine spike actiyity appearance rates are provided in Table 2.4-5.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 35 of 87 SO. . IN ..NGINERIN ONS SF 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. In the unlikely event of a concurrent loss of power, the loss of circulating water through the condenser would eventually result in the loss of condenser vacuum, thereby causing steam relief directly to the atmosphere from the ADVs. This direct steam relief continues until the faulted 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 faulted SG relief valves. Additional activity, based on the proposed primary-to-secondary leakage limits, is released via steaming from the ADVs until the RCS is cooled to 212-F.

Per the St. Lucie Unit 1 UFSAR, Section 15.4.4.5.1, no fuel failure is postulated for the SGTR event.

Consistent with RG 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 (see Table 2.4-3). Primary coolant is released into the ruptured SG by the tube rupture and by a fraction of the total proposed allowable primary-to-secondary leakage.

Activity is released to the environment from the ruptured SG via direct flashing of a fraction of the released primary coolant from the tube rupture and also via steaming from the ruptured SG ADVs until the ruptured steam generator is isolated at 30 minutes. The unaffected SG is used to cool down the plant during the SGTR event. Primary-to-secondary tube leakage is also postulated into the intact SG. Activity is released via steaming from the unaffected SG ADVs until the RCS is cooled below 212'F. These release assumptions are consistent with the requirements of RG 1.183.

For the case of the accident-induced spike, the postulated STGR event induces an iodine spike. The RCS activity is initially assumed to be 1.0 9xCi/gm DE 1-131 as allowed by 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Parameters used in the determination of the iodine equilibrium release rate are provided in Table 2.4-4.

The iodine activities and the appearance rates 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 three modes of operation:

" Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 920 cfm of unfiltered fresh air and an assumed value of 500 cfm of unfiltered inleakage.

" After the start of the event, the Control Room is isolated due to a high radiation reading in the Control Room ventilation system. For this event, it is conservatively assumed that the CR isolation signal is delayed'until the release from the ADVs is initiated at 379.2 seconds. An additional 50-second delay is applied to account for the diesel generator start time, fan start and damper actuation time. After isolation, the air flow distribution consists of 0 cfm of makeup flow from the outside, 500 cfm of unfiltered inleakage, and 2000 cfm of filtered recirculation flow.

NUMERICAL AST Licensing Technical Report for NA-M 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 36 of 87 IN eNGINEERING 501LJTI0NS ANDSOFFWARE

• At 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event, the operators are assumed to initiate makeup flow from the outside to the control room. During this operational mode, the air flow distribution consists of up to 450 cfm of filtered makeup flow, 500 cfm of unfiltered inleakage, and 1550 cfm of filtered recirculation flow.

The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, 95% for elemental iodine, and 95% for organic iodine.

2.4.5 Radiological Consequences 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 intake.

Prior to reactor trip, the release is assumed to originate from the condenser. Following the trip, the releases from both the intact and faulted SGs are assumed to occur from the closest ADV.

When the Control Room Ventilation System is in normal mode, the most limiting X/Q corresponds to the worst air intake to the control room. When the ventilation system is isolated at 50 seconds, the limiting X/Q corresponds to the midpoint between the two control room air intakes. The operators are assumed to reopen the most favorable air intake at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The X/Q 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 for the different modes of control room operation during this event.

The EAB and LPZ dose consequences are determined using the X/Q factors provided Table 1.8.2-1 for the appropriate time intervals. For the EAB dose calculation, the X/Q factor for the zero to two-hour time interval is assumed for all time periods. Using the zero to two-hour X/Q factor provides a more conservative determination of the EAB dose, because the X/Q factor for this time period is higher than for any other time period.

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 TS 3.4.8 limits, are analyzed. As shown in Table 2.4-6, the radiological consequences of the St. Lucie Unit 1 SGTR event for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

NUMERICAL AST Licensing Technical Report for NA-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 37 of 87 SOMnONSIN ENGSNAEERNGMAD SOF1WA 2.5 Reactor Coolant Pump Shaft Seizure (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, radioactivity is contained in the primary and secondary coolant before the accident and some of this activity is released to the atmosphere as a result of steaming from the steam generators following the accident. This event is described in Section 15.3.4 of the UFSAR.

2.5.2 Compliance with RG 1.183 Regulatory Positions The Locked Rotor dose consequence analysis followed 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 RG 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 1.7 is applied. The fraction of fission product inventory in the gap available for release due to DNB is consistent with Regulatory Position 3.2 and Table 3 of RG 1.183. Gap release fractions are increased by a factor of 1.03687 to account for high burnup fuel rods as described in Section 1.9.

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 SGs 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 apportioned between the SGs as specified by proposed TS 6.8.4.1 (0.5 gpm total, 0.25 gpm to any one SG). Thus, the tube leakage is apportioned equally between the two SGs.

6. Regulatory Position 5.2 - The density used in converting volumetric leak rates to mass leak rates is based upon RCS conditions, consistent the plant design basis.

7. Regulatory Position 5.3 - The primary-to-secondary leakage is assumed to continue until after shutdown cooling has been placed in service and the temperature of the RCS is less than 212'F.

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.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit 1 Page 38 of 87 SOLLIlIONS INENGINEERING AND SOýIARE

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 - Both steam generators are used for plant cooldown. 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 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 SG, enter the steam space, and be immediately released to the environment with no mitigation; i.e., no reduction for scrubbing within the SG bulk water is credited.

• Appendix E, Regulatory Position 5.5.3 - All of the SG tube leakage flow.that does not 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 SGs is limited by the moisture carryover from the SG. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the SG carryover rate of less than 1%.

Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery in the SGs is postulated for up to 45 minutes following a reactor trip for St. Lucie Unit 1. During this period, the fraction of primary-to-secondary leakage which flashes to vapor is assumed to rise through the bulk water of the SG 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

1. RG 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 RG 1.183 specifies noble gas, alkali metal, and iodine fuel gap release fractions for the breached fuel.

2. The initial RCS activity is assumed to be at the TS limit of 1.0 uCi/gm Dose Equivalent 1-131 and 100/E-bar gross activity. The initial SG activity is assumed to be at the TS 3.7.1.4 limit of 0.1 tCi/gm Dose Equivalent 1-131.

3. The steam mass release rates for the SGs are provided in Table 2.5-2. These values are based upon a cooldown rate of 100 °F/hr until the RCS temperature reaches 300 °F. This temperature is maintained until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when Shutdown Cooling is assumed to become available. The cooldown is then continued at a rate of 38 °F/hr until the RCS temperature is reduced to 212 OF at 10.32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />. No credit is taken for energy removal by the Shutdown cooling system.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 39 of 87 SOUITIONS INENGINEERING AND5OFIARE

4. The RCS fluid density used to convert the primary-to-secondary leakage from a volumetric flowrate to a mass flow rate is consistent with the RCS cooldown rate applied in the generation of the secondary steam releases. The high initial cooldown rate conservatively maximizes the fluid density. The SG tube leakage mass flow rate is provided in Table 2.5-3.

5. This evaluation assumes that the RCS mass remains constant throughout the event.

6. For the purposes of determining the iodine concentrations, the SG mass is assumed to remain constant throughout the event. However, it is also assumed that operator action is taken to restore secondary water level above the top of the tubes within a conservative time of one hour following a reactor trip.

7. This analysis assumes that the DNB fuel damage is limited to 13.7% breached fuel assemblies.

2.5.4 Methodology Input assumptions used in the dose consequence analysis 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.

Forthe purpose of this dose assessment, a total of 13.7% 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 Appendix G of RG 1.183 with an adjustment for high burnup fuel and a radial peaking factor of 1.7. Primary coolant is released to the SGs as a result of postulated primary-to-secondary leakage. Activity is released to the atmosphere via steaming from the steam generator ADVs until the RCS is cooled to 212'F. These release assumptions are consistent with the requirements of RG 1.183.

For this event, the Control Room ventilation system cycles through three modes of operation (the operational modes are summarized in Table 1.6.3-1):

0 Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 920 cfm of unfiltered fresh air and an assumed value of 500 cfm of unfiltered inleakage.

0 After the start of the event, the Control Room is isolated due to a high radiation reading in the Control Room ventilation system. A 50-second delay is applied to account for diesel generator start time, damper actuation time, instrument delay, and detector response time. After isolation, the air flow distribution consists of 0 cfm of makeup flow from the outside, 500 cfm of unfiltered inleakage, and 2000 cfm of filtered recirculation flow.

• At 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event, the operators are assumed to initiate makeup flow from the outside to the control room. During this operational mode, the air flow distribution consists of up to 450 cfm of filtered makeup flow, 500 cfm of unfiltered inleakage, and 1550 cfm of filtered recirculation flow.

0 The Control Room ventilation filter efficiencies that-are applied to the filtered makeup and recirculation flows are 99% for particulate, 95% for elemental iodine, and 95% for organic iodine.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 40 of 87 2.5.5 Radiological Consequences The Control Room atmospheric dispersion factors (XIQs) 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 intake.

For the Locked Rotor event, all releases are from the closest ADV.

When the Control Room Ventilation System is in normal mode, the most limiting X/Q corresponds to the worst air intake to the control room. When the ventilation system is isolated at 50 seconds, the limiting X/Q corresponds to the midpoint between the two control room air intakes. The operators are assumed to reopen the most favorable air intake at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The X/Q values for the various combinations of release points and receptor locations are presented in Tablel.8.1-2. Table 1.8.1-3 presents the Release-Receptor pairs applicable to the control room dose for the different modes of control room operation during this event.

The EAB and LPZ dose consequences are determined using the X/Q factors provided Table 1.8.2-1 for the appropriate time intervals. For the EAB dose calculation, the X/Q factor for the zero to two-hour time interval is assumed for all time periods. Using the zero to two-hour XIQ factor provides a more conservative determination of the EAB dose, because the X/Q factor for this time period is higher than for any other time period.

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 2.5-4, the results for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

NUMERICAL AST Licensing Technical Report for NAI-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 41 of 87 SMUTIWS IN ENGINERINGMAD SFTAR 2.6 Control Element Assembly Ejection (CEA) 2.6.1 Background This event consists of an uncontrolled withdrawal of a single control element assembly (CEA). This event is the same as the Rod Ejection event referred to in RG 1.183. The CEA Ejection results in a reactivity insertion that leads to a core power level increase and subsequent reactor trip. Following the reactor trip, plant cooldown is performed using steam release from the SG ADVs. Two CEA 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 15.4.5.

2.6.2 Compliance with RG 1.183 Regulatory Positions The CEA Ejection dose consequence analysis followed the guidance provided in RG 1.183 Appendix H, "Assumptions for Evaluating the Radiological Consequences of a PWR Rod Ejection Accident," as discussed below:

1. Regulatory Position I - The total core inventory of the radionuclide groups utilized for determining the source term for this event is based on RG 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 1.7 is applied. The release fractions provided in RG 1.183 Table 3 are adjusted to comply with the specific RG 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. Gap release fractions are increased by a factor of 1.03687 to account for high burnup fuel rods as described in Section 1.9.

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 SGs.

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 SGs 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. In addition, the shield building ventilation system (SBVS) is credited. Containment spray is not credited.

NUMERICAL AST Licensing Technical Report for NAIM 101-043, Rev. 2 APPLICATIONS, INC., St.

SO. ONS ..... . S Lucie Unit Page 42 of 87

7. Regulatory Position 6.2 - The containment is assumed to leak at the TS maximum allowable rate of 0.5% for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.25% for the remainder of the event.

8. Regulatory Position 7.1 - The primary-to-secondary leak rate is apportioned between the SGs as specified by proposed TS 6.8.4.1 (0.5 gpm total, 0.25 to any one SG).

9. Regulatory Position 7.2 - The density used in converting volumetric leak rates to mass leak rates is based upon RCS conditions, consistent with the plant design basis.

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 7.4 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 - For the secondary release case, both steam generators are used for plant cooldown. 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 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 SG, enter the steam space, and be immediately released to the environment with no mitigation; i.e., no reduction for scrubbing within the SG bulk water is credited.

" Appendix E, Regulatory Position 5.5.3 - All of the SG tube leakage that does not 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 theSGs is limited by the moisture carryover from the SG. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the SG carryover rate of less than 1%.

• Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery in the SGs is postulated for up to 45 minutes following a reactor trip for St. Lucie Unit 1. During this period, the fraction of primary-to-secondary leakage which flashes to vapor is assumed to rise through the bulk water of the SG 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. The initial RCS activity is assumed to be at the TS limit of 1.0 VCi/gm Dose Equivalent 1-131 and 100/E-bar gross activity. The initial SG activity is assumed to be at the TS 3.7.1.4 limit of 0.1 pCi/gm Dose Equivalent 1-13 1.

NUMERICAL AST Licensing Technical Report for NA-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 43 of 87 SVUJTONSINe4GINEERING ANDSO~TWARM

2. The steam mass release rates for the SGs are provided in Table 2.6-2. These values are based upon a cooldown rate of 100 0F/hr until the RCS temperature reaches 300 'F. This temperature is maintained until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when Shutdown Cooling is assumed to become available. The cooldown is then continued at a rate of 38 °F/hr until the RCS temperature is reduced to 212 'F at 10.32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />. No credit is taken for energy removal by the Shutdown cooling system.

3. The RCS fluid density used to convert the primary-to-secondary leakage from a volumetric flowrate to a mass flow rate is consistent with the RCS cooldown rate applied in the generation of the secondary steam releases. The high initial cooldown rate conservatively maximizes the fluid density. The SG tube leakage mass flow rate is provided in Table 2.6-3.

4. This evaluation assumes that the RCS mass remains constant throughout the event.

5. For the purposes of determining the iodine concentrations, the SG mass is assumed to remain constant throughout the event. However, it is also assumed that operator action is taken to restore secondary water level above the top of the tubes within a conservative time of one hour following a reactor trip.

6. Following the CEA Ejection event, 9.5% of the fuel is assumed to fail as a result of DNB and 0.05% of the fuel is assumed to experience fuel centerline melt.

7. All secondary releases are postulated to occur from the ADV with the most limiting atmospheric dispersion factors. Releases from containment through the SBVS are assumed to be released from the plant stack.with a filter efficiency of 99% for particulates and 95% for both elemental and organic iodine. The activity that bypasses the SBVS is released unfiltered to the environment via a ground level release from containment.

8. The initial leakage rate from containment is 0.5% of the containment air per day. This leak rate is reduced by 50% after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.25%/day. 9.6% of the containment leakage is assumed to bypass the SBVS filters.

9. For the release inside of containment, natural deposition of the radionuclides is credited consistent with the LOCA methodology presented in Section 2.1.3. Containment sprays are not credited.

2.6.4 Methodology Input assumptions used in the dose consequence analysis of the CEA 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.

For the containment release case, 100% of the activity is released instantaneously to the containment. The releases from the containment correspond to the same leakage points discussed for the LOCA in Section 2.1. Natural deposition of the released activity inside of containment is credited. In addition, the shield building ventilation system (SBVS) is credited. Removal of activity via containment spray is not credited.

/

For the secondary release case, primary coolant activity is released into the SGs by leakage across the SG tubes. The activity on the secondary side is then released via steaming from the ADVs until the RCS is cooled to 212'F. All noble gases associated with this leakage are assumed to be released directly to the environment. The primary-to-secondary leakage is assumed to continue until the faulted steam generator is completely isolated at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In addition, the analysis assumes that the initial iodine activity of both SGs is immediately released to the environment. The secondary coolant iodine concentration is assumed to be the maximum value of 0.1 g.tCi/gm DE 1-131 permitted by Tech. Specs. These release assumptions are

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 44 of 87 consistent with the requirements of RG 1.183.

The CEA Ejection is evaluated with the assumption that 0.5% of the fuel experiences FCM and 9.5% of the fuel experiences DNB. The activity released from the damaged fuel corresponds to the requirements set out in Regulatory Position I of Appendix H to RG 1.183 with an adjustment for high burnup fuel and a radial peaking factor of 1.70 applied in the development of the source terms.

For this event, the Control Room ventilation system cycles through three modes of operation (the operational modes are summarized in Table 1.6.3-1):

• Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 920 cfm of unfiltered fresh air and an assumed value of 500 cfm of unfiltered inleakage.

0 After the start of the event, the Control Room is isolated due to a high radiation reading in the Control Room ventilation system. A 50-second delay is applied to account for diesel generator start time, damper actuation time, instrument delay, and detector response time. After isolation, the air flow distribution consists of 0 cfm of makeup flow from the outside, 500 cfm of unfiltered inleakage, and 2000 cfm of filtered recirculation flow.

• At 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event, the operators are assumed to initiate makeup flow from the outside to the control room. During this operational mode, the air flow distribution consists of up to 450 cfm of filtered makeup flow, 500 cfm of unfiltered inleakage, and 1550 cfm of filtered recirculation flow.

• The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, 95% for elemental iodine, and 95% for organic iodine.

2.6.5 Radiological Consequences 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 intake.

• For the CEA Ejection event, all secondary releases are from the closest ADV. X/Qs for containment releases via the SBVS are from the plant stack, and containment releases which bypass the SBVS correspond to the nearest feedwater line which penetrates containment.

When the Control Room Ventilation System is in normal mode, the most limiting XIQ corresponds to the worst air intake to the control room. When the ventilation system is isolated at 50 seconds, the limiting X/Q corresponds to the midpoint between the two control room air intakes. The operators are assumed to reopen the most favorable air intake at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The X/Q 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 for the different modes of control room operation' during this event.

The EAB and LPZ dose consequences are determined using the X/Q factors provided Table 1.8.2-1 for the appropriate time intervals. For the EAB dose calculation, the X/Q factor for the zero to two-hour time interval is assumed for all time periods. Using the zero to two-hour X/Q factor provides a more conservative determination of the EAB dose, because the X/Q factor for this time period is higher than for any other time period.

The radiological consequences of the CEA Ejection are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. As shown in Table 2.6-4, the results of both cases for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

NUMERICAL AST Licensing Technical Report for NA-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 45 of 87 SOLUTION/S AND S:O[FTWhARE IN ENGINEERING 2.7 Inadvertent Opening of a Main Steam Safety Valve (IOMSSV):

2.7.1 Background This event is caused by an Inadvertent Opening of a Steam Generator MSSV. Due to the pressure differential between the primary and secondary systems and assumed steam generator tube leakage, fission products contained in the primary coolant before the accident are discharged from the primary into the secondary system. The analysis assumes that the SG tubes do not remain covered and therefore no credit is taken for scrubbing in the SG or any credit for a flashing fraction for the primary leakage into the SGs. As a result, all of the leaked RCS radioactivity is released to the outside atmosphere from the secondary coolant system through the steam generator via the MSSVs. In addition, all of the activity initially present in the SGs is assumed to be released to the environment over a 2-hour period. Radiological releases due to the opening of a power operated atmospheric dump valve are bounded by the inadvertent opening of a MSSV event. This IOMSSV event is described in Section 15.2.11 of the UFSAR.

2.7.2 Compliance with RG 1.183 Regulatory Positions Since Regulatory Guide 1.183 does not provide specific guidance for this event, the guidance of Appendix G for the RCP Shaft Seizure (Locked Rotor) event is judged to be closely applicable to the conditions of an inadvertent open MSSV. Therefore the following discussion for the IOMSSV refers to the RG 1.183 positions as stated in Appendix G for the Locked Rotor event.

I. Regulatory Position 1 - No fuel damage is postulated to occur. The source term for this event is due to the initial RCS and Secondary side activity present at the beginning of the event.

2. Regulatory Position 2 - No fuel damage is assumed for this event.

3. Regulatory Position 3 - The activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant.

4. Regulatory Position 4 - Iodine releases from the SGs to the environment are assumed to be 97%

elemental and 3% organic.

5. Regulatory Position 5.1 - The primary-to-secondary leak rate is apportioned between the SGs as specified by proposed TS 6.8.4.1 (0.5 gpm total, 0.25 gpm to any one SG). Thus, the tube leakage is apportioned equally between the two SGs.

-6. Regulatory Position 5.2 - The density used in converting volumetric leak rates to mass leak rates is based upon RCS conditions, consistent with the plant design basis.

7. Regulatory Position 5.3 - The primary-to-secondary leakage is assumed to continue until after shutdown cooling has been placed in service and the temperature of the RCS is less than 212'F.

8. Regulatory Position 5.4 - The analysis assumes a coincident loss of offsite power.

9. Regulatory Position 5.5 - All noble gas radionuclides released from the primary system are assumed released to the environment without reduction or mitigation.

NUMERICAL AST Licensing Technical Report for NAI-i 101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit 1 Page 46 of 87 S nN IN ENGINEERING ANDSOFIWARE

10. Regulatory Position 5.6 - The steam generator tubes are not assumed to remain covered throughout this event for St. Lucie Unit 1. Therefore, the iodine and transport model for release from the SGs is as follows:

Appendix E, Regulatory Position 5.5.1 - Both steam generators are assumed to "dryout."

Therefore, all of the primary-to-secondary leakage is assumed to flash to steam and be released to the environment with no mitigation.

Appendix E, Regulatory Position 5.5.2 - All of the SG tube leakage is assumed to flash for this event.

Appendix E, Regulatory Position 5.5.3 - All of the SG tube leakage is assumed to flash for this event.

Appendix E, Regulatory Position 5.5.4 - The radioactivity within the bulk water in the SGs is assumed to be released directly to the environment over a 2-hour period.

• Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery is postulated for this event for St. Lucie Unit 1.

2.7.3 Other Assumptions

1. The initial RCS activity is assumed to be at the TS limit of 1.0 VCi/gm Dose Equivalent 1-131 and 100/E-bar gross activity. The initial SG activity is assumed to be at the TS 3.7.1.4 limit of 0.1 p.Ci/gm Dose Equivalent 1- 131.

2. This evaluation assumes that the RCS mass remains constant throughout the event.

3. The entire contents of both steam generators are assumed to be released to the environment over a 2-hour period.

2.7.4 Methodology Input assumptions used in the dose consequence analysis of the Inadvertent Opening of a MSSV event are provided in Table 2.7-1. Primary coolant is released to the SGs as a result of postulated primary-to-secondary leakage. The activity in the RCS tube leakage is released directly to the environment until the terminated at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In addition, the entire secondary side activity is released to the environment over a 2-hour period.

For this event, the Control Room ventilation. system cycles through three modes of operation (the operational modes are summarized in Table 1.6.3-1):

" Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 920 cfm of unfiltered fresh air and an assumed value of 500 cfm of unfiltered inleakage.

" After the start of the event, the Control Room is isolated due to a high radiation reading in the Control Room ventilation system. A 50-second delay is applied to account for diesel generator start time, damper actuation time, instrument delay, and detector response time. After isolation, the air flow distribution consists of 0 cfm of makeup flow from the outside, 500 cfm of unfiltered inleakage, and 2000 cfm of filtered recirculation flow.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 47 of 87 S..U.ONS.INeNGIN.EERINGANDSOFA7E At 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event, the operators are assumed to initiate makeup flow from the outside to the control room. During this operational mode, the air flow distribution consists of up to 450 cfm of filtered makeup flow, 500 cfm of unfiltered inleakage, and 1550 cfm of filtered recirculation flow.

• The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, 95% elemental iodine, and 95% organic iodine.

2.7.5 Radiological Consequences 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 intake.

Values from the closest ADV are more limiting than those from the MSSV for all time periods and were used in the analysis of this event.

When the Control Room Ventilation System is in normal mode, the most limiting X/Q corresponds to the worst air intake to the control room. When the ventilation system is isolated at 50 seconds, the limiting X/Q corresponds to the midpoint between the two Control Room air intakes. The operators are assumed to reopen the most favorable air intake at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The X/Q 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 for the different modes of control room operation during this event.

The EAB and LPZ dose consequences are determined using the X/Q factors provided Table 1.8.2-1 for the appropriate time-intervals. For the EAB dose calculation, the XIQ factor for the zero to two-hour time interval is assumed for all time periods. Using the zero to two-hour X/Q factor provides a more conservative determination of the EAB dose, because the X/Q factor for this time period is higher than for any other time period.

RG 1.183 lists no specific acceptance criteria for this event; therefore, the most limiting dose limits from Section 4.4 and Table 6 of RG 1.183 for the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) are used:

Area Dose Criteria EAB 2.5 rem TEDE (for the worst two-hour period)

LPZ 2.5 rem TEDE (for 30 days)

Control Room* 5 rem TEDE (for 30 days)

• Control room dose limit is specified in 10CFR50.67 The radiological consequences of the Inadvertent Opening of the MSSV event are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. As shown in Table 2.7-3, the results of both cases for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

NUMERICAL AST Licensing Technical Report for NAI-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 48 of 87 SGýUTIDNSIN eNGTNeERIN3AND SORflNAR 2.8 Environmental Oualification (EO)

The St. Lucie Unit No. 1 UFSAR, Section 3.11, discusses equipment EQ due to the radiation environment.

RG 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. The St. Lucie Unit No. 1 EQ analyses will continue to be based on TID-14844 assumptions.

3.0 Summary of Results Results of the St. Lucie Unit 1 radiological consequence analyses using the AST methodology and the corresponding allowable control room unfiltered inleakage are summarized on Table 3-1.

4.0 Conclusion 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), Reactor Coolant Pump Shaft Seizure (Locked Rotor), Control Element Assembly (CEA) Ejection, and Inadvertent Opening of a Main Steam Safety Valve (IOMSSV) have been made using the RG 1.183 methodology. The analyses used assumptions consistent with proposed changes in the St. Lucie Unit No. 1 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 500 cfm.

5.0 References 5.1 St. Lucie Unit No. 1 Updated Final Safety Analysis Report(UFSAR), (through Amendment 21).

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

5.3 NRC Generic Letter 2003-01, "Control Room Habitability," June 12, 2003.

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

5.5 .Code of Federal Regulations, 10CFR50.67, "Accident Source Term," revised 12/03/02.

5.6 USNRC, Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants," July 2000.

5.7 Florida Power & Light Company, St. Lucie Unit No. 1 Technical Specifications (through Amendment 200).

5.8 PSL-ENG-SENS-03-022, "Engineering Evaluation, Containment Bypass Leakage History, St.

Lucie Nuclear Plant, Units 1 & 2," Revision 0, Florida Power & Light Company.

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 49 of 87 S ýOUýNSNENGINEERING ANOSOTWARE 5.9 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.

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

5.1.1 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).

5.12 MicroShield Version 5 "User's Manual" and "Verification & Validation Report, Rev. 5," Grove Engineering, both dated October 1996.

5.13 Oak Ridge National Laboratory, CCC-371, "RSICC Computer Code Collection - ORIGEN 2.1,"

May 1999.

5:14 "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).

5.15 Numerical Applications Inc., NAI-9912-04, Revision 2, "RADTRAD-NAI Version 1.1a(QA)

Documentation," October 2004.

5.16 PSL-ENG-SENS-03-001, "Engineering Evaluation, Alternate Source Term Design Inputs, St.

Lucie Nuclear Plant, Unit 1," Revision 3, Florida Power & Light Company.

5.17 Florida Power & Light Company, St. Lucie Unit No. 1 Core Operating Limits Report (COLR).

5.18 USNRC Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants", June 2003.

5.19 USNRC Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessment at Nuclear Power Plants," Rev. 1, February 1983.

5.20 Numerical Applications Inc., "Dose Methodology Quality Assurance Procedures," Revision 1, June 4, 2001.

5.21 NAI Calculation Number NAI-1 101-002 Rev. 0, "Qualification of ORIGEN2.1 for Florida Power

& Light AST Applications," August 8,'2002. Report (UFSAR), (through Amendment 21).

5.22 NUREG-0800, USNRC, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," September 1981 (or updates of specific sections).

5.23 Industry Degraded Core Rulemaking Program Technical Report 11.3, "Fission Product Transport in Degraded Core Accidents," Atomic Industrial Forum, December 1983.

5.24 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.

5.25 NRC Generic Letter 99-02, "Laboratory Testing of Nuclear-Grade Activated Charcoal," June 3, 1999.

5.26 NRC Information Notice 91-56, "Potential Radioactive Leakage to Tank Vented to Atmosphere,"

September 19, 1991.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 50 of 87 SOMTMMIONS4eNGITERINSANDSFTARe 5.27 USNRC, Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.

5.28 NUREG./CR-5950, "Iodine Evolution and pH Control," December 1992.

5.29 Duane Arnold Issuance of Amendment (IA) and Safety Evaluation (SE) for Amendment No. 240 to DPR-49 issued July 31, 2001.

5.30 Kewaunee Nuclear Plant - Issuance of Amendment Regarding Implementation of Alternate Source Term (TAC No. MB4596), March 17, 2003.

5.31 Shearon Harris Nuclear Power Plant, Unitl - Issuance of Amendment Re: Steam Generator Replacement and Power Uprate (TAC Nos. MB0199 and MB0782).

5.32 USNRC, Regulatory Issue Summary 2006-04, Experience with Implementation of Alternate Source Terms, March 7, 2006.

5.33 RS-06-010, Letter from Exelon Nuclear Corp to USNRC,

Subject:

Response to NRC Request for License Amendment Related to Application of Alternate Radiological Source Term, Byron and Braidwood stations, January 27, 2006.

5.34 USNRC, TAC No. MB 1221, "Fort Calhoun Station, Unit No. 1 - Issuance of Amendment 201 to License No. DPR-40", December 5, 2001.

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATION. ,INC. St. Lucie Unit 1 Page 51 of 87 SOtLlONS IN ENINEERING AN SFTARE Figure 1.8.1-1 Onsite Release-Receptor Location Sketch t Plant North Fuel Handling Building Turbine GF Building F C

K O RWT

• ¢ Control Reactor D Room Auxiliary

.................... B uilding E I* II I (Not to scale)

• - Control Room Intakes / Receptor Point A- Plant Stack B- RWT C- FHB Closest Point D- Louver L-7B E- Louver L-7A F- Closest ADV G- Closest MSSV H- Closest Main Steam Line Point (containment penetration)

I - Closest Feedwater Line Point (containment penetration)

J - Containment Maintenance Hatch K- Condenser

NUMERICAL AST Licensing Technical Report for NAI-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 52 of 87 SOTUTONSIN ENGINEERINGMAD)SOFTARE Table 1.6.3-1 Control Room Ventilation System Parameters Parameter Value Control Room Volume 62,318 ft3 Normal Operation Filtered Make-up Flow Rate 0 cfm Filtered Recirculation Flow Rate 0 cfm Unfiltered Make-up Flow Rate 920 cfm Unfiltered Inleakage 500 cfm Emergency Operation Isolation Mode:

Filtered Make-up Flow Rate 0 cfm Filtered Recirculation Flow Rate 2000 cfm Unfiltered Make-up Flow Rate 0 cfm Unfiltered Inleakage 500 cfm Filtered Make-up Mode:

Filtered Make-up Flow Rate 450 cfm Filtered Recirculation Flow Rate 1550 cfm Unfiltered Make-up Flow Rate 0 cfm Unfiltered Inleakage 500 cfm Filter Efficiencies Particulate 99%

Elemental 95%

Organic 95%

NUMERICAL AST Licensing Technical Report for NA-i1101-043, Rev. 2 APPLIATION.S. INC. St. Lucie Unit 1 Page 53 of 87 Table 1.6.3-2 LOCA Direct Shine Dose Source Direct Shine Dose (rem)

Containment 0.03 Filters 0.07 External Cloud 0.07 Total 0.17 Table 1.7.2-1 Primary Coolant Source Term Nuclide gxCi/gm Nuclide j+/-Cilgm 1-131 0.7920 SR-90 4.551E-04 1-132 0.2175 CR-51 6.627E-03 1-133 1.1293 FE-59 3.715E-05 1-134 0.1237 CO-60, 9.051E-04 1-135 0.5387 SR-91 6.208E-03 H-3 2.302E-01 Y-90 1.779E-03 KR-85M 2.599E+00 Y-91 1.936E-01 KR-85 1.543E+00 ZR-95 1.630E-06 KR-87 1.413E+00 MO-99 3.541E+00 KR-88 4.534E+00 RU-103 7.203E-03 RB-88 4.447E+00 RU-106 4.326E-04 RB-89 1.116E-01 TE- 129 4.377E-02 XE-131M 2.582E+00 TE-132 5.755E-01 XE-133 3.156E+02 TE- 134 4.568E-02 XE-135 1.313E+01 BA-140 1.065E-02 BR-84 8.126E-02 LA-140 1.020E-02 CS-134 1.744E-01 CE-144 7.203E-03 CS-136 4.447E-02 PR-143 1.018E-02 CS-137 5.580E-01 MN-54 4.796E-05 CS-138 1.203E+00 CO-58 8.126E-03 SR-89 8.840E-03

NUMERICAL AST Licensing Technical Report for NA-i1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit

• Page 54 of 87 SýWTARR IN EGINEERINGMAD SOLMTONS Table 1.7.3-1 Secondary Side Source Term Isotope RCi/gm 1-131 0.07920 1-132 0.02175 1-133 0.11293 1-134 0.01237 1-135 0.05387 Table 1.7.4-1 LOCA Containment Leakage Source Term Nuclide Curies Nuclide Curies Co-58 0.OOOE+00 Pu-239 3.828E+04 Co-60 0.OOOE+00 Pu-240 6.456E+04 Kr-85 1.152E+06 Pu-241 1.626E+07 Kr-85m 1.784E+07 Am-241 2.152E+04 Kr-87 3.383E+07 Cm-242 6.998E+06 Kr-88 4.752E+07 Cm-244 1.053E+06 Rb-86 2.348E+05 1-130 4.626E+06 Sr-89 6.480E+07 Kr-83m 8.634E+06

.Sr-90 9.253E+06 Xe-138 1.198E+08 Sr-91 8.105E+07 Xe-131m 8.582E+05 Sr-92 8.882E+07 Xe-133m 4.765E+06 Y-90 9.615E+06 Xe-135m 3.081E+07 Y-91 8.483E+07 Cs-138 1.334E+08 Y-92 8.925E+07 Cs-134m 5.846E+06 Y-93 1.046E+08 Rb-88 4.841E+07 Zr-95 1.206E+08 Rb-89 6.176E+07 Zr-97 1.207E+08 Sb-124 2.157E+05 Nb-95 1.220E+08 Sb-125 1.797E+06 Mo-99 1.405E+08 Sb-126 1.244E+05 Tc-99m 1.230E+08 Te-131 6.773E+07 Ru-103 1.320E+08 Te-133 8.797E+07 Ru-105 1.010E+08 Te-134 1.188E+08 Ru-106 6.560E+07 Te-125m 3.947E+05

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 55 of 87 SOLUniONS INENGINEERING ANDSOFrWARE Nuclide Curies Nuclide Curies Rh-105 9.303E+07 Te-133m 5.267E+07 Sb-127 9.609E+06 Ba- 141 1.184E+08 Sb-129 2.678E+07 Ba-137m 1.216E+07 Te- 127 9.546E+06 Pd-109 3.771E+07 Te-127m 1.294E+06 Rh-106 7.109E+07 Te-129 2.637E+07 Rh-103m 1.189E+08 Te-129m 3.930E+06 Tc-101 1.293E+08 Te-131m 1.151E+07 Eu- 154 1.606E+06 Te-132 1.073E+08 Eu-155 1.088E+06 1-131 7.686E+07 Eu-156 2.847E+07 1-132 1.094E+08 La-143 1.086E+08 1-133 1.486E+08 Nb-97 1.218E+08 1-134 1.616E+08 Nb-95m 8.606E+05 1-135 1.396E+08 Pm-147 1.187E+07 Xe-133 1.492E+08 Pm-148 2.220E+07 Xe-135 4.333E+07 Pm-149 4.726E+07 Cs-134 2.606E+07 Pm- 151 1.686E+07 Cs-136 7.018E+06 Pm-148m 2.843E+06 Cs-137 1.284E+07 Pr-144 1.021E+08 Ba-139 1.307E+08 Pr- 144m 1.218E+06 Ba- 140 1.258E+08 Sm- 153 5.086E+07 La-140 1.310E+08 Y-94 1.062E+08 La-141 1.190E+08 Y-95 1.152E+08 La-142 1.146E+08 Y-91m 4.705E+07 Ce-141 1.208E+08 Br-82 6.291E+05 Ce-143 1.094E+08 Br-83 8.606E-06 Ce-144 1.014E+08 Br-84 1.470E+07 Pr-143 1.088E+08 Am-242 1.041E+07 Nd-147 4.809E+07 Np-238 5.399E+07 Np-239 1.960E+09 Pu-243 6.043E+07 Pu-238 5.475E+05

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 56 of 87 soufloNSINeNGIemRNGAN SFTAR Table 1.7.5-1 Fuel Handling Accident Source Term Bounding Bounding Bounding Nuclide Activities Nuclide Activities Nuclide Activities (Curies) (Curies) (Curies)

Co-58 0.OOOE+00 1-135 1.094E+06 Sb-126 9.746E+02 Co-60 0.OOOE+00 Xe-133 1. 169E+06 Te-131 5.306E+05 Kr-85 9.025E+03 Xe-135 3.395E+05 Te-133 6.892E+05 Kr-85m 1.398E+05 Cs-134 2.042E+05 Te-134 9.307E+05 Kr-87 2.650E+05 Cs-136 5.498E+04 Te-125m 3.092E+03 Kr-88 3.723E+05 Cs-137 1.006E+05 Te-133m 4.126E+05 Rb-86 1.839E+03 Ba-139 1.024E+06 Ba-141 9.276E+05 Sr-89 5.076E+05 Ba- 140 9.855E+05 Ba-137m 9.526E+04 Sr-90 7.249E+04 La-140 1.026E+06 Pd-109 2.954E+05 Sr-91 6.350E+05 La-141 9.323E+05 Rh-106 5.569E+05 Sr-92 6.958E+05 La-142 8.978E+05 Rh-103m 9.315E+05 Y-90 7.532E+04 Ce-141 9.464E+05 Tc-101 1.013E+06 Y-91. 6.646E+05 Ce-143 8.571E+05 Eu-154 1.258E+04 Y-92 6.992E+05 Ce-144 7.944E+05 Eu-155 8.524E+03 Y-93 8.194E+05 Pr-143ý 8.524E+05 Eu- 156 2.230E+05 Zr-95 9.448E+05 Nd- 147 3.767E+05 La-143 8.508E+05 Zr-97 9.456E+05 Np-239 1.535E+07 Nb-97 9.542E+05 Nb-95 9.558E+05 Pu-238 4.758E+03 Nb-95m 6.742E+03 Mo-99 1.101E+06 Pu-239 2.999E+02 Pm-147 9.299E+04 Tc-99m 9.636E+05 Pu-240 5.058E+02 Pm-148 1.739E+05 Ru-103 1.034E+06 Pu-241 1.274E+05 Pm-149 3.702E+05 Ru-105 7.912E+05 Am-241 1.686E+02 Pm- 151 1.321E+05 Ru-106 5.139E+05 Cm-242 5.482E+04 Pm-148m 2.227E+04 Rh-105 7.288E+05 Cm-244 1.516E+04 Pr-i144 7.999E+05 Sb-127 7.528E+04 1-130 3.624E+04 Pr-144m 9.542E+03 Sb-129 2.098E+05 Kr-83m 6-764E+04 Sm-153 3.984E+05 Te-127 7.478E+04 Xe-138 9.385E+05 Y-94 8.320E+05 Te-127m 1.014E+04 Xe-131m 6.723E+03 Y-95 9.025E+05 Te-129 2.066E+05 Xe-133m 3.733E+04 Y-91m 3.686E+05 Te-129m 3.079E+04 Xe-135m 2.414E+05 Br-82 4.928E+03 Te-131m 9.017E+04 Cs-138 1.045E+06 Br-83 6.742E+04 Te- 132 8.406E+05 Cs-134m 4.580E+04 Br-84 1.152E+05 1-131 6.02 1E+05 Rb-88 3.792E+05 Am-242 8.155E+04 1-132 8.571E+05 Rb-89 4.838E+05 Np-238 4.289E+05 1-133 1.164E+06 Sb- 124 1.690E+03 Pu-243 4.734E+05 1-134 1.266E+06 Sb-125 1.408E+04

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 57 of 87 S~tMfMtS IN ENiINEERING ANDSFWR Table 1.8.1-1 Release-Receptor Combination Parameters for Analysis Events Direction Release Release Receptor Receptor Distance Distance with Receptor Release Point Point Height Height Height Height (ft) (m) respect to (ft) (m) (ft) (m) true north Stack/Plant N CR intake 184 56.1 59.75 18.2 48.08 14.6 58 Vent Stack/Plant S CR intake 184 56.1 59.75 18.2 126.69 38.6 354 Vent RWT N CR intake 48.22 14.6 59.75 18.2 245.31 74.7 65 RWT S CR intake 48.22 14.6 59.75 18.2 263.64 80.3 39 FHB Closest N CR intake 43.25 13.2 59.75 18.2 120.6 36.7 48 Point FHB Closest S CR intake 43.25 13.2 59.75 18.2 184.26 56.1 11 Point Aux. Bldg. N CR intake 38.17 11.6 59.75 18.2 123.77 37.7 72 Louver L-7B Aux. Bldg. S CR intake 38.17 11.6 59.75 18.2 136.97 41.7 34 Louver L-7A Condenser N CR intake 5.25 1.6 59.75 18.2 153.24 46.7 245 Closest ADV N CR intake 53 16.1 59.75 18.2 105.68 *32.2 306 Closest ADV S CR intake 53 16.1 59.75 18.2 214.82 65.4 319 Closest Feedwater Line N CR intake 17 5.2 59.75 18.2 83.29 25.3 306 Point Closest Feedwater Line S CR intake 17 5.2 59.75 18.2 193.15 58.8 321 Point Containment Maintenance N CR intake 16 4.9 59.75 18.2 172.4 52.5 359 Hatch Containment Maintenance S CR intake 16 4.9 59.75 18.2 279.09 85.0 348 Hatch Stack/Plant Midpoint between 184 56.1 59.75 18.2 74.85 22.8 8 intakes Midpoint RWT between 48.22 14.6 59.75 18.2 244.91 74.6 52 intakes

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPPLICTIONS, INC. St. Lucie Unit 1 Page 58 of 87 Direction Release Release Receptor Receptor Distance Distance with Receptor Release Point Point Height Height Height Height (ft) (in) respect to (ft) (in) (ft) (in) true north Aux. Bldg. Midpoint Louver L-7A between inae 38.17 11.6 59.75 18.2 118.59 36.1 59 intakes Midpoint Closest ADV between 53 16.1 59.75 18.2 160.26 48.8 314 intakes Closest Midpoint Feedwater Line between 17 5.2 59.75 18.2 138.15 42.1 315 Point intakes Containment Midpoint Maintenance between 16 4.9 59.75 18.2 223.66 68.1 351 Hatch intakes FHB Closest Midpoint 43.25 13.2 59.75 18.2 142.19 43.3 25 Point between intakes Stack/Plant Vent Louver L-11 184 56.1 49.5 15.1 127.68 38.9 355 RWT Louver L- 11 48.22 14.6 49.5 15.1 267.32 81.4 40 Aux. Bldg. Louver L-11 38.17 11.6 49.5 15.1 140.21 42.7 35 Louver L-7A'-

Closest ADV Louver L-11 53 16.1 49.5 15.1 213.23 64.9 320 Closest Feedwater Line Louver L-11 17 5.2 49.5 15.1 191.68 58.4 322 Point Containment Maintenance Louver L- 11 16 4.9 49.5 15.1 279.61 85.2 349 Hatch FHB Closest Louver L-11 43.25 13.2 49.5 15.1 186.43 56.8 12 Point Notes:

1. Release heights are calculated as 19 feet less than the reference elevations to account for the plant grade elevation.

2. The FHB closest point release elevation is taken as the roof elevation since the SW comer of the roof is the closest building point to the intakes.

3. Release and receptor points are considered to be at the centerpoint or centerline of all openings.

NUMERICAL AST Licensing Technical Report for NAI- 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 59 of 87 SOLL0TIONS N ENGINEERING ANDSOFTWARE

4. The only release/receptor combination that does not have the intakes in the same wind direction window from the release point is for the'releases from the plant stack. All other release points analyzed result in both control room intakes being in the same wind direction window. Therefore, credit may be taken for intake dilution only for releases from the plant. stack.

5. The receptor point for the "midpoint between intakes" is taken as being on the outside of the control room (and H&V room) east wall. The receptor elevation is taken as the average of the receptor elevations for the two outside air intakes.

6. For events where the limiting unfiltered inleakage location is through the control room intakes, atmospheric dispersion factors corresponding to the midpoint between the control room intakes are to be used during the time period when the control room intakes are isolated.

7. The closest containment/shield building penetration to the intakes that is directly exposed to the atmosphere is the closest feedwater line penetration.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit I Page 60 of 87 Table 1.8.1-2 Onsite Atmospheric Dispersion (XIQ) Factors for Analysis Events This table summarizes the X/Q factors for the control room intakes and for switchgear room louver L-1 1 that apply to the various accident scenarios. For the intakes, values are presented for the unfavorable intake prior to control room isolation, the midpoint between the intakes during isolation, as well as values for the favorable intake following manual restoration of filtered control room make-up flow. These values are not corrected for Control Room Occupancy Factors but do include credit for dilution where allowed. Based on the layout of the site, the only cases that may take credit for dilution are when the releases are from the plant vent stack. However, dilution is not credited during the time period when the control room intakes are isolated for these cases.

• Indicates credit for dilution taken for this case.

1. The atmospheric dispersion factors corresponding to ADVs were determined to be more limiting than those from the MSSVs for all time periods. Therefore, the more limiting ADV values have been used throughout the analyses for all secondary releases. No distinction is made between automatic steam relief from the MSSVs and controlled releases from the ADVs for radiological purposes.

Release- 0-2 hour 2-8 hour 8-24 hour 1-4 days 4-30 days Receptor Release Point Receptor Point X/Q XIQ X/Q X/Q X/Q Pair A Stack/Plant N CR intake* 2.35E-03 Vent

• B* Stack/Plant S CR intake* 6.68E-04 4.55E-04 2.11 E-04 1.26E-04 9.25E-05 Vent

• C RWT N CR intake 1.38E-03

'D RWT S CR intake 1.1OE-03 9.30E-04 3.96E-04 2.94E-04 2.28E-04 E FHB Closest N CR intake 4.93E-03 Point F F-B Closest S CR intake 2.OOE-03 1.40E-03 6.36E-04 4.22E-04 3.09E-04 Point G Aux. Bldg. N CR intake 4.85E-03 Louver L-7B N H Aux. Bldg-Louver L-7A S CR intake 3.59E-03 2.94E-03 1.24E-03 8.84E-04 6.91E-04 I Condenser N CR intake 2.47E-03 J Closest ADV # N CR intake 6.24E-03 K Closest ADV # S CR intake 1.61E-03 1.26E-03 5.08E-04 3.60E-04 2.71E-04 Closest L Feedwater N CR intake 7.30E-03 Line Point Closest M Feedwater S CR intake 1.75E-03 1.35E-03 5.76E-04 3.94E-04 2.94E-04 Line Point

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 61 of 87 ANDSOFTWARE SOLUnOS INENGIREERING Release- 0-2 hour 2-8 hour 8-24 hour 1-4 days 4-30 days Receptor Release Point Receptor Point XQXQX/XQXQ Pair W X/Q X/Q X/Q X/Q Stack/Plant Midpoint N between 3.78E-03 intakes Midpoint 0 RWT between 1.33E-03 intakes Aux. Bldg. Midpoint P between 5.04E-03 Louver L-7A intakes Midpoint Q Closest ADV # between 2.82E-03 intakes Closest Midpoint R Feedwater between 3.17E-03 Line Point intakes Containment S Maintenance N CR intake 1.87E-03 Hatch Containment T Maintenance S CR intake 8.11E-04 6.11E-04 2.79E-04 1.72E-04 1.28E-04 Hatch Containment Midpoint U Maintenance between 1.19E-03 Hatch intakes Midpoint V FHB Closest between 3.26E-03 Point intakes W Stack/Plant Vent Louver L-1 1 2.47E-03 1.66E-03 7.69E-04 4.60E-04 3.40E-04 X RWT Louver L-1 1 1.06E-03 9.06E-04 3.90E-04 2.95E-04 2.25E-04 y Aux. Bldg.

• Louver L-7A Louver L-1 1 3.53E-03 2.88E-03 1.23E-03 8.78E-04 6.87E-04 Z Closest ADV # Louver L-1 1 1.63E-03 1.28E-03 5.19E-04 3.63E-04 2.74E-04 Closest AA Feedwater Louver L-1 1 1.81E-03 1.40E-03 5.98E-04 4.09E-04 3.OOE-04 Line Point Containment BB Maintenance Louver L-11 8.84E-04 6.85E-04 3.17E-04 1.91E-04 1.40E-04 Hatch FHB Closest CC Coset Point Louver L-11 198E-03 1.38E-03 6.18E-04 4.11E-04 2.98E-04

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit. Page 62 of 87 SOWT1ONS ANDSOFTARE INENGINEERING Table 1.8.1-3 Release-Receptor Point Pairs Assumed for Analysis Events Prior to Control During Control After Initiation of Event Room Isolation Room Isolation Filtered Air Make-up LOCA:,

- Containment Leakage (SBVS) A N B

- Containment (SBVS Bypass) L R M

- ECCS Leakage G P H

- RWT Backleakage C 0 D

- Cont. Purge/H 2 Purge A N B FHA:

Containment Release S U T FHB Release E V F MSLB:

- Outside Containment. I Q K

- Inside Containment (SBVS) A N B-

- Inside Containment (SBVS Bypass) L R M SGTR I (Prior to Turbine Trip) K J (After Turbine Trip) Q_ __K Locked Rotor JQ K CEA Ejection:

- Secondary Release J Q K

- Inside Containment (SBVS) A N B

- Inside Containment (SBVS Bypass) L R M IOMSSV J Q K TOMSSV I Q K

NUMERICAL AST Licensing Technical Report for NA-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 63 of 87 SOffOýSIN~ENINEERIHGANDSOFTWAR

. Table 1.8.2-1 Offsite Atmospheric Dispersion (XIQ) Factors for Analysis Events Time Period EAB X/Q (sec/m 3) LPZ XIQ (sec/mr3) 0-2 hours 1.03E-04 9.97E-05 0-8 hours 5.69E-05 5.47E-05 8-24 hours 4.22E-05 4.05E-05 1-4 days 2.22E-05 2.11E-05 4-30 days 8.79E-06 8.29E-06 The above table summarizes the maximum X/Q factors for the EAB and LPZ. Note that the 0-2 hour EAB X/Q factor was used for the entire event.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 64 of 87 SOUtlmOkSINCNA5IWERMNM ANDSOFTWARE Table 2.1-1 Loss of Coolant Accident (LOCA) - Inputs and Assumptions Input/Assumption Value" Release Inputs:

Core Power Level 2754 MWfh (2700 + 2%)

Core Average Fuel Burnup 45,000 MWD/MTU Fuel Enrichment 3.0 - 4.5 w/o Initial RCS Equilibrium Activity 1.0 tCi/gm DE 1-131 and 100/E-bar gross activity (Table 1.7.2-1)

Core Fission Product Inventory Table 1.7.4-1 Containment Leakage Rate 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.5% (by weight)/day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.25% (by weight)/day LOCA release phase timing and duration Table 2.1-2 Core Inventory Release Fractions (gap release and early in- RG 1.183, Sections 3.1 and 3.2 vessel damage phases)

ECCS Systems Leakage Sump Volume (minimum) 55,460 ft3 ECCS Leakage to RAB (2 times allowed value) 4510 cc/hr Flashing Fraction Calculated - 7.5%

Used for dose determination - 10%

Chemical form of the iodine in the sump water 0% aerosol, 97% elemental, and 3.0% organic Release ECCS Area Filtration Efficiency Elemental - 95%

Organic - 95%

Particulate - 99% (100% of the particulates are retained in the ECCS fluid)

NUMERICAL AST Licensing Technical Report for NAI-i 101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit 1 Page 65 of 87

ý eNINEERINGMAD NSOLfOISN SOFTARE Input/Assumption Value RWT Back-leakage Sump Volume (at time of recirculation) 57,140 ft3 ECCS Leakage to RWT (2 times allowed value) 2 gpm Flashing Fraction (elemental Iodine assumed to be released 0 % based on temperature of fluid reaching RWT into tank space based upon partition factor)

RWT liquid/vapor Elemental Iodine partition factor Table 2.1-7 Elemental Iodine fraction in RWT Table 2.1-6 Initial RWT Liquid Inventory (minimum) 38,842 gallons Release from RWT Vapor Space Table 2.1-8 Containment Purge Release 42,000 cfm for 5 seconds' Removal Inputs:

Containment Particulate/Aerosol Natural Deposition (only 0.1/hour credited in unsprayed regions)

Containment Elemental Iodine Natural/Wall Deposition 2.89/hour Containment Spray Region Volume 2,155,160 ft3 Containment Unsprayed Region Volume 350,840 ft3 Flowrate between sprayed and unsprayed volumes 11,695 cfm Spray Removal Rates:

Elemental Iodine 20/hour Time to reach DF of 200 3.02 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Particulate Iodine 6.43/hour Time to reach DF of 50 2.60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> Spray Initiation Time 64.5 seconds (0.017917 hours)

Spray Termination Time 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Control Room Ventilation System Table 1.6.3-1 Time of automatic control room isolation 50 seconds Time of manual control room unisolation 1.5 hrs Particulate - 99%

Secondary Containment Filter Efficiency Elemental - 95%

Organic - 95 %

Secondary Containment Drawdown Time 310 seconds Secondary Containment Bypass Fraction 9.6%

Containment Purge Filtration 0%

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 66 of 87 SOULýONS IN ENGINEERING ANDSOFTWARE Input/Assumption Value Transport Inputs:

Containment Release Nearest Containment penetration to CR ventilation Secondary Containment release prior to drawdown intake Containment Release Secondary Containment release after drawdown Plant stack Containment Release Nearest Containment penetration to CR ventilation Secondary Containment Bypass Leakage intake ECCS Leakage ECCS exhaust louver RWT Backleakage RWT 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 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 NA-i 101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit. Page 67 of 87 Table 2.1-2 LOCA Release Phases Phase Onset Duration Gap Release 30 seconds 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Early In-Vessel 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />

• From RG 1.183, Table 4 Table 2.1-3 LOCA Time Dependent RWT pH Time (hours) SIRWT pH 0.00 4.500 0.33 4.500 0.50 4.500 0.64 4.500 0.83 4.501 2.78 4.503 4.17 4.505 5.56 4.507 9.72 4.512 11.11 4.514 15.28 4.520 22.22 4.528 55.56 4.568 83.33 4.599 97.22 4.614 111.11 4.628 152.78 4.667 194.44 4.704 250.00 4.748 305.56 4.788 402.78 4.850 500.00 4.905 597.22 4.953 694.44 4.997 720.00 5.007

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 68 of 87 SM~0.UNS N eNGrNeCRrNG"MD S0ThARE Table 2.1-4 LOCA Time Dependent RWT Total Iodine Concentration*

Time RWT Iodine Concentration (hours) (gm-atom/liter) 0.00 O.00E+00 0.33 O.00E+00 0.50 3.15E-08 0.64 5.79E-08 0.83 9.38E-08 2.78 4.59E-07 4.17 7.17E-07 5.56 9.73E-07 9.72 1.73E-06 11.11 1.97E-06 15.28 2.70E-06 22.22 3.88E-06 55.56 8.92E-06 83.33 1.25E-05 97.22 1.41E-05 111.11 1.56E-05

.152.78 1.96E-05 194.44 2.29E-05 250.00 2.67E-05 305.56 2.97E-05 402.78 3.39E-05 500.00 3.71E-05 597.22 3.97E-05 694.44 4.17E-05 720.00 4.22E-05

• Includes radioactive and stable iodine isotopes

NUMERICAL 'AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit. Page 69 of 87 SOMLMONSIN ENGINEERINGMAD SFTAR Table 2.1-5 LOCA Time Dependent RWT Liquid Temperature Time (hr) Temperature (F) 0.00 100.0 0.33 100.0 0.50 100.0 0.64 100.0 0.83 100.0 2.78 100.0 4.17 100.0 5.56 100.0 9.72 100.0 11.11 100.0 15.28 100.1 22.22 100.6 55.56 103.3 83.33 104.6 97.22 105.0 111.11 105.3 152.78 105.8 194.44 105.9 250.00 105.8 305.56 105.8 402.78 105.7 500.00 105.7 597.22 105.7 694.44 105.6 720'00 105.6

NUMERICAL AST Licensing Technical Report for NAI-I1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 70 of 87 ANDSOFTWARE INEMNC-ERTNG

,.AJUTON*i4 Table 2.1-6 LOCA Time Dependent RWT Elemental Iodine Fraction Time (hr) Elemental Iodine Fraction 0.00 0.OOE+00 0.33 0.OOE+00 0.50 7.09E-04 0.64 1.30E-03 0.83 2.1OE-03 2.78 1.OOE-02 4.17 1.54E-02 5.56 2.06E-02 9.72 3.47E-02 11.11 3.91E-02 15.28 5.13E-02 22.22 6.88E-02 55.56 1.22E-01 83.33 1.46E-01 97.22 1.54E-01 111.11 1.59E-01 152.78 1.69E-01 194.44 1.73E-01 250.00 1.71E-01 305.56 1.66E-01 402.78 1.54E-01 500.00' 1.41E-01 597.22 1.28E-01 694.44 1.16E-01 720.00 1.13E-01

NUMERICAL AST Licensing Technical Report for NAIM 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 71 of 87 SLOT.UTONS AND SOFrWARE INENGINEERtNG Table 2.1-7 LOCA Time Dependent RWT Partition Coefficient Elemental Iodine Partition Time (hr) Coefficient 0.00 45.65 0.33 45.65 0.50 45.65 0.64 45.65 0.83 45.65 2.78 45.65 4.17 45.65 5.56 45.65 9.72 45.65 11.11 45.65 15.28 45.56 22.22 45.13 55.56 42.86 83.33 41.82 97.22 41.50 111.11 41.26 152.78 40.87 194.44 40.79 250.00 40.87 305.56 40.87 402.78 40.95 500.00 40.95 597.22 40.95 694.44 41.03 720.00 41.03

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 72 of 87 SOI.IPTOI4S IN ENGINERINS ANDSOFTWARE Table 2.1-8 LOCA Release Rate from RWT Time Adjusted Iodine Release Rate (hours) (cfm) 0.33 4.800E-07 4.17 3.700E-06 11.11 1.444E-05 22.22 1.300E-04 111.11 4.437E-04 305.56 6.429E-04 402.78 6.612E-04 500.00 6.568E-04 597.22 6.404E-04 694.44 5.615E-04 Table 2.1-9 LOCA Dose Consequences 1

EAB Dose( ) LPZ Dose(2) Control Room Dose(2)

Case (rem TEDE) (rem TEDE) (rem TEDE)

LOCA 1.08 2.53 4.69 Acceptance Criteria 25 25 5 (1)Worst 2-hour dose (2) Integrated 30-day dose

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 73 of 87 SOLUTION5N ENGINEERING AND501FWARE Table 2.2-1 Fuel Handling Accident (FHA) - Inputs and Assumptions Input/Assumption Value Core Power Level Before Shutdown 2754 MWth (2700 + 2%)

Core Average Fuel Burnup 45,000 MWD/MTU Discharged Fuel Assembly Burnup 45,000 - 62,000 MWD/MTU Fuel Enrichment 3.0 - 4.5 w/o Maximum Radial Peaking Factor 1.7 Number of Fuel Assemblies in the Core 217 Number of Fuel Assemblies Damaged 1 Delay Before Spent Fuel Movement 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> FHA Source Term for a Single Assembly Table 1.7.5-1 High Burnup Fuel Adjustment Factor 2.0 Water Level Above Damaged Fuel Assembly 23 feet minimum Elemental - 285 Iodine Decontamination Factors Organic -1 Noble Gas Decontamination Factor 1 Elemental- - 0.15%

99.85%

Chemical Form of Iodine In Pool Organ Organic - 0. 15%

Elemental- - 43%

57%

Chemical Form of Iodine Above Pool Organ Organic - 43%

Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3.

Control Room Ventilation System Time of Control Room Ventilation System Isolation 50 seconds Time of Control Room Filtered Makeup Flow 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Room Unfiltered Inleakage 500 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 Table 2.2-2 Fuel Handling Accident Dose Consequences EAB Dose(') LPZ Dose(2) Control Room Dose(2)

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

Containment Release 0.53 0.52 1.23 FHB Release 0.53 0.52 3.02 Acceptance Criteria 6.3 6.3 5

(')Worst 2-hour dose (2) Integrated 30-day dose

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 74 of 87 SOWJONSIN ENGINCE~RING ANDSOFTWAN Table 2.3-1 Main Steam Line Break (MSLB) Inputs and Assumptions Input/Assumption Value Core Power Level 2754 MWh (2700 + 2%)

Core Average Fuel Burnup 45,000 MWD/MTU Fuel Enrichment 3.0 - 4.5 w/o Maximum Radial Peaking Factor 1.7

% DNB for MSLB Outside of Containment 1.8%

% DNB for MSLB Inside of Containment 29%

% Fuel Centerline Melt for MSLB Outside of Containment 0.43%

% Fuel Centerline Melt for MSLB Inside of Containment 6.1%

Core Fission Product Inventory Table 1.7.4-1 Initial RCS Equilibrium Activity 1.0 gtCi/gm DE 1-131 and 100I/E-bar gross activity Initial ___CSEquilibriumActivity_(Table 1.7.2-1)

Initial Secondary Side Equilibrium Iodine Activity 0.1 gtCi/gm DE 1-131 (Table 1.7.3-1)

Release Fraction from DNB Fuel Failures RG 1.183, Section 3.2 Release Fraction from Centerline Melt Fuel Failures RG 1.183, Section 3.2, and Section 1 of Appendix H High Bumup Fuel Adjustment Factor 1.03687 Steam Generator Tube Leakage 0.25 gpm per SG (Table 2.3-3)

Time to Terminate SG Tube Leakage 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Steam Release from Intact SGs Table 2.3-2 Intact SG Tube Uncovery Following Reactor Trip Time to tube recovery 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Flashing Fraction 5%

Steam Generator Secondary Side Partition Coefficient Unaffected SG - 100 Faulted SG - None Time to Reach 212 °F and Terminate Steam Release 10.32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> Containment Volume 2.506E+06 ft3 Containment Leakage Rate 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.5% (by weight)/day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.25% (by weight)/day Particulate - 99%

Secondary Containment Filter Efficiency Elemental - 95%

Organic - 95%

Secondary Containment Drawdown Time 310 seconds Secondary Containment Bypass Fraction 9.6%

411,500 lbm RCS Mass Minimum mass used for fuel failure dose contribution to maximize SG tube leakage activity.

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 P g7TWARE APPLICATIONS, INC. St. Lucie Unit I Page 75 of 87 AND

.SOTMlNSIN ENHGNERING Input/Assumption Value minimum - 105,000 Ibm (per SG) maximum - 205,000 Ibm (per SG)

SG Secondary Side Mass Maximum mass used for faulted SG to maximize secondary side dose contribution. Minimum mass used for intact SG to maximize steam release nuclide concentration.

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 Time of Control Room Ventilation System Isolation 50 seconds Time of Control Room Filtered Makeup Flow 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Room Unfiltered Inleakage 500 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 Aerosols - 0.1 hr-'

Containment Natural Deposition Coefficients Elemental Iodine - 2.89 hr-I Organic Iodine - None Table 2.3-2 MSLB Steam Release Rate Time Intact SG Steam Release Rate (hours) (Ibm/min) 0-0.25 7900 0.25-0.50 4196 0.50- 0.75 4707 0.75- 1.0 5362 1.0-1.5 5028 1.5 - 2.25 4725 2.25 -4.0 3924 4.0- 8.0 2558 8.0-10.32 3094 10.32-720 0

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 76 of 87 ANDSOFTWARE INENGINEERING SOUJTIONS Table 2.3-3 MSLB Steam Generator Tube Leakage Time Tube Leakage per SG (hours) (Ibm/min) 0-0.50. 1.47 0.50- 1.0 1.52 1.0-1.5 1.62 1.5-2.0 1.71 2.0-2.5 1.78 2.5-3.0 1.85 3.0-3.5, 1.90 3.5-9.69 1.92 9.69-12 1.96 12-720 0 Table 2.3-3 MSLB Dose Consequences Fuel EAB Dose°') LPZ Dose(2) Control RoomDose(2)

Case Failure (rem TEDE) (rem TEDE) (rem TEDE)

MSLB - Outside of Containment 1.8% DNB 0.33 0.90 4.80 MSLB - Outside of Containment 0.43% FCM 0.36 .0.97 4.97 MSLB - Inside of Containment 29% DNB 0.52 1.04 4.92 MSLB - Inside of Containment 6.1% FCM 0.76 1.43. 4.91 Acceptance Criteria . 25 25 5

(') Worst 2-hour dose (2) Integrated 30-day dose

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 77 of 87 SaýONli INEWNGIEERING GNUSOFIWARE Table 2.4-1 Steam Generator Tube Rupture (SGTR) - Inputs and Assumptions Input/Assumption Value Core Power Level 2754 MWth (2700 + 2%)

Initial RCS Equilibrium Activity 1.0 gCi/gm DE 1-131 and 100/E-bar gross activity (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 60tCi/gm DE 1-131 Maximum Equilibrium Iodine Concentration 1.OgCi/gm DE 1-131 Iodine Spike Appearance Rate 335 times Duration of Accident-Initiated Spike 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Integrated Break Flow and Steam Release Table 2.4-2 Prior to Reactor Trip - 17%

Break Flow Flashing Fraction Following Reactor Trip - 5%

Time to Terminate Break Flow 30 minutes Steam Generator Tube Leakage Rate 0.25 gpm per SG Time to Terminate Tube Leakage 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Time to Re-cover Intact SG Tubes 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Steam Generator Secondary Side Partition Coefficients Flashed tube flow - none Non-flashed tube flow - 100 Time to Reach 212 °F and Terminate Steam Release 10.32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> RCS Mass Pre-accident Iodine spike - 411,500 Ibm Concurrent Iodine spike - 438,843 Ibm minimum - 105,000 lbm (per SG) maximum - 205,000 lbm (per SG)

SG Secondary Side Mass Minimum used for primary-to-secondary leakage to maximize secondary nuclide concentration. Maximum used for initial secondary inventory release to maximize secondary side dose contribution.

Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System Time of Control Room Ventilation System Isolation 429.2 seconds Time of Control Room Filtered Makeup Flow 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Room Unfiltered Inleakage 500 cfm Breathing Rates Offsite RG 1.183, Section 4.1.3 Control Room .RG 1.183, Section 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 78 of 87 SýUTONSIN ENINEERINGMADSOFTWARE Table 2.4-2 SGTR Integrated Mass Releases (1)

Break Flow in Steam Release from Steam Release from (hours) Ruptured SG Ruptured SG Unaffected SG (Ibm) (Ibm) (Ibm) 0- 0.1053 hrs : 661,842 (via Condenser) 656,568 (via Condenser) 0- 0.5 104,660 0.1053 - 0.5 hrs: 88,352 (via MSSV) 86,821 (via MSSVs) 0.5 -2.0 0 0 601,096 (via ADVs) 2-8 N/A N/A 876,233 8-10.32 N/A N/A 31.09 1bn/rain

Flowrate assumed to be constant within time period Table 2.4-3 SGTR 60 gCi/gm D.E. 1-131 Activities Activity Isotope (0tCi/gm)

Iodine-131 47.5 Iodine-132 13.1 Iodine-133 67.8 Iodine-134 7.42 Iodine-135 32.3 Table 2.4-4 SGTR Iodine Equilibrium Appearance Assumptions Input Assumption Value Maximum Letdown Flow 150 gpm at 120'F, 650 psia Maximum Identified RCS Leakage 10 gpm Maximum Unidentified RCS Leakage 1 gpm RCS Mass 438,843 Ibm 1-131 Decay Constant 0.003038 1-132 Decay Constant 0.008001 1-133 Decay Constant 0.003533 1-134 Decay Constant 0.016156 1-135 Decay Constant 0.004726

NUMERICAL AST Licensing Technical Report for NAI- 101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit 1 Page 79 of 87 SOLUTIONS SOFTWARN INENGINEERINGA1D Table 2.4-5 SGTR Concurrent Iodine Spike (335 x) Activity Appearance Rate Total 8-hour Isotope Activity Appearance Rate Production (CUmin) (Ci)

Iodine-131 160.4 77013 Iodine-132 116.0 55700 Iodine-133 266.0 127670 Iodine-134 133.3 63967 Iodine-135 169.8 81488 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 0.31 0.30 3.03 Acceptance Criteria (pre-accident iodine spike) 25 (3) 25 (3) 5 (4)

SGTR concurrent iodine spike 0.08 0.08 0.60 Acceptance Criteria (concurrent iodine spike) 2.5 (3) 2.5 (3) 5 (4)

(1) Worst 2-hour dose (2) Integrated 30-day dose

<3)RG 1.183, Table 6 (4) 10CFR50.67

NUMERICAL AST Licensing Technical Report for NAI-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 80 of 87 SOUTPI NENGIEERINGMAD SOFIWARE Table 2.5-1 Reactor Coolant Pump Shaft Seizure (Locked Rotor) - Inputs and Assumptions Input/Assumption Value Core Power Level 2754 MWth (2700 + 2%)

Core Average Fuel Burnup 45,000 MWD/MTU Fuel Enrichment 3.0 - 4.5w/o Maximum Radial Peaking Factor 1.7 Percent of Fuel Rods in DNB 13.7%

High Burnup Fuel Adjustment Factor 1.03687 Core Fission Product Inventory Table 1.7.4-1 Initial RCS Equilibrium Activity 1.0 gCi/gm DE 1-131 and 100/E-bar gross activity (Table 1.7.2-1)

Initial Secondary Side Equilibrium Iodine Activity 0.1 gtCi/gm DE 1-131 (Table 1.7.3-1)

Release Fraction from Breached Fuel RG 1.183, Section 3.2 Steam Generator Tube Leakage 0.5 gpm (Table'2.5-3)

Time to Terminate SG Tube Leakage 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Secondary Side Mass Releases to Environment Table 2.5-2 SG Tube Uncovery Following Reactor Trip Time to tube recovery 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Flashing Fraction 5%

Flashed tube flow - none Steam Generator Secondary Side Partition Coefficient No-lsetuefw 10 Non-flashed tube flow - 100 Time to Reach 212 °F and Terminate Steam Release 10.32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> 411,500 lbm RCS Mass Minimum mass used for fuel failure dose contribution to maximize SG tube leakage activity.

minimum - 105,000 Ibm (per SG) maximum - 205,000 Ibm (per SG) dMass Minimum used for primary-to-secondary leakage to SG Secondary Side Mmaximize secondary nuclide concentration. Maximum used for initial secondary inventory release to maximize secondary side dose contribution.

Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System Time of Control Room Ventilation System Isolation 50 seconds Time of Control Room Filtered Makeup Flow 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Room Unfiltered Inleakage 500 cfm Breathing Rates Offsite RG 1.183 Section 4.1.3 Onsite RG 1.183 Section 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6

NUMERICAL AST Licensing Technical Report for NA-i 101-043, Rev. 2 APPLICATIONS, INC. I St. Lucie Unit Page 81 of 87 SMUMNS~45IN ENINEERINMAD SFTAe Table 2.5-2 Locked Rotor Steam Release Rate Time SG Steam Release Rate (hours) (lbmmin) 0-0.25 7900 0.25 - 0.50 4196 0.50- 0.75 4707 0.75- 1.0 5362 1.0-1.5 5028 1.5 - 2.25 4725 2.25 -4.0 3924 4.0- 8.0 2558 8.0-10.32 3094 Table 2.5-3 Locked Rotor Steam Generator Tube Leakage Time SG Tube Leakage (hours) (lbjmin) 0-0.50 2.94 0.50-1.0 3.05 1.0-1.5 3.25 1.5-2.0 3.42 2.0-2.5 3.57 2.5-3.0 3.70 3.0-3.5 3.80 3.5-9.69 3.83 9.69-12 3.91 12-720 0 Table 2.5-4 Locked Rotor Dose Consequences EAB Dose (l) LPZ Dose (2) Control Room Dose (2)

Case (rem TEDE) (rem TEDE) (rem TEDE)

Locked Rotor 0.25 0.54 2.53 Acceptance Criteria 2.5 2.5 5 (1) Worst 2-hour dose (2) Integrated 30-day dose

NUMERICAL AST Licensing Technical Report for NAT-i 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit I Page 82 of 87 SOLIUiONS INENGINEERIN6 ANDSOFTWAM Table 2.6-1 Control Element Assembly (CEA) Ejection - Inputs and Assumptions Input/Assumption Value Core Power Level 2754 MWth (2700,+ 2%)

Core Average Fuel Burnup 45,000 MWD/MTU Fuel Enrichment 3.0 -4.5 w/o Maximum Radial Peaking Factor 1.7 Percent of Fuel Rods in DNB 9.5%

Percent of Fuel Rods with Centerline Melt 0.5%

Core Fission Product Inventory Table 1.7.4-1 Initial RCS Equilibrium Activity 1.0 gtCi/gm DE 1-131 and 100/E-bar gross activity Initial __RCSEquilibrium__Activity_(Table 1.7.2-1)

Initial Secondary Side Equilibrium Iodine Activity 0.1 gtCi/gm DE 1-131 (Table 1.7.3-1)

Release Fraction from DNB Fuel Failures Section 1 of Appendix H to RG 1.183 Release Fraction from Centerline Melt Fuel Failures Section 1 of Appendix H to RG 1.183 High Bumup Fuel Adjustment Factor 1.03687 Steam Generator Tube Leakage 0.5 gpm (Table 2.6-3)

Time to Terminate SG Tube Leakage 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Secondary Side Mass Releases to Environment Table 2.6-2 SG Tube Uncovery Following Reactor Trip Time to tube recovery 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Flashing Fraction 5%

Flashed tube flow - none Steam Generator Secondary Side Partition Coefficient Non-flashed tube flow - 100 Time to Reach 212 'F and Terminate Steam Release 10.32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> minimum - 411,500 lbm RCS Mass Minimum mass used for fuel failure dose contribution to maximum SG tube leakage activity.

minimum - 105,000 Ibm (per SG) maximum - 205,000 Ibm (per SG)

SG Secondary Side Mass Minimum used for primary-to-secondary leakage to maximize secondary nuclide concentration.. Maximum used for initial secondary inventory release to maximize secondary side dose contribution.

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%

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 83 of 87

• t J ~*ON UT NINERW FT AD5FARE Input/Assumption Value Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System Time of Control Room Ventilation System Isolation 50 seconds Time of Control Room Filtered Makeup Flow 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Room Unfiltered Inleakage 500 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 Containment Volume 2.506E+06 ft3 Containment Leakage Rate 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.5% (by weight)/day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.25% (by weight)/day Particulate - 99%

Secondary Containment Filter Efficiency Elemental - 95%

_________________________________________Organic - 95%

Secondary Containment Drawdown Time 310 seconds Secondary Containment Bypass Fraction 9.6%

Aerosols - 0.1 hrl Containment Natural Deposition Coefficients Elemental Iodine - 2.89 hr1 Organic Iodine - None Table 2.6-2 CEA Ejection Steam Release Rate Time SG Steam Release Rate (hours) (lbJmin) 0-0.25 7900 0.25-0.50 4196 0.50- 0.75 4707 0.75- 1.0 5362 1.0-1.5 5028 1.5 - 2.25 4725 2.25 - 4.0 3924 4.0- 8.0 2558 8.0-10.32 3094

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit 1 Page 84 of 87 SOLULiONS IN eNGIN6eRING ANDSOF1WARE Table 2.6-3 CEA Ejection Steam Generator Tube Leakage Time SG Tube Leakage (hours) (lbjmin) 0- 0.50 2.94 0.50-1.0 3.05 1.0-1.5 3.25 1.5-2.0 3.42 2.0-2.5 3.57 2.5-3.0 3.70 3.0-3.5 3.80 3.5-9.69 3.83 9.69-12 3.91 12-720 0 Table 2.6-4 CEA Ejection Dose Consequences EAB Dose(1 ) LPZ Dose(2 ) Control Room Dose(2)

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

CEA Ejection - Containment Release 0.26 0.50 2.74 CEA Ejection - Secondary Release 0.29 0.63 2.60 Acceptance Criteria 6.3 6.3 5

("Worst 2-hour dose (2) Integrated 30-day dose

NUMERICAL AST Licensing Technical Report for NAI-1 101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit. Page 85 of 87 SOL.U1ONS INENGINeERING ANDSOFIWARE Table 2.7-1 IOMSSV - Inputs and Assumptions Input/Assumption Value Core Power Level 2754 MWth (2700 + 2%)

1.0 gCi/gm DE 1-131 and 100/E-bar gross activity Initial RCS Equilibrium Activity (Table 1.7.2-1)

Initial Secondary Side Equilibrium Iodine Activity 0.1 gtCi/gm DE 1-131 (Table 1.7.3-1)

Steam Generator Tube Leakage 0.5 gpm (Table 2.7-2)

Time to Terminate SG Tube Leakage 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Secondary Side Mass Releases to Environment Entire inventory in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Steam Generator Secondary Side Partition Coefficient none Maximum - 205,000 lbm per SG SG Secondary Side Mass Maximum mass used for initial secondary inventory release to maximize secondary side dose contribution.

Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Tables 1.8.1-2 and 1.8.1-3 Control Room Ventilation System Time of Control Room Ventilation System Isolation 50 seconds Time of Control Room Filtered Makeup Flow 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Room Unfiltered Inleakage 500 cfm Breathing Rates:

Offsite RG 1.183 Section 4.1.3 Onsite RG 1.183 Section 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS, INC. St. Lucie Unit. Page 86 of 87 5~fOS fl eNGINCERINGMAD S4WTARG Table 2.7-2 IOMSSV Steam Generator Tube Leakage Time SG Tube Leakage (hours) (lbm/min) 0 - 0.50 2.94 0.50- 1.0 3.05 1.0-1.5 3.25 1.5-2.0- 3.42 2.0-2.5 3.57 2.5 -3.0 3.70 3.0-3.5 3.80 3.5-9.69 3.83 9.69-12 3.91 12-720 0.

Table 2.7-3 IOMSSV Dose Consequences EAB Dose (1) LPZ Dose (2) Control Room Dose (2)

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

Inadvertent Opening of a MSSV 0.02 0.02 0.30 Acceptance Criteria 2.5 2.5 5

) Worst 2-hour dose (2) Integrated 30-day dose

NUMERICAL AST Licensing Technical Report for NAI-1101-043, Rev. 2 APPLICATIONS. INC. St. Lucie Unit 1 Page 87 of 87 SOLUTIONS AND SOF'WARC INENGINEERING Table 3-1 St. Lucie Plant, Unit No. 1 Summary of Alternative Source Term Analysis Results LOCA 500 1.08 2.53 4.69 MSLB - Outside of Containment 500 0.33 0.90 4.80 (1.8% DNB)

MSLB - Outside of Containment 500 0.36 0.97 4.97 (0.43% FCM)

MSLB - Inside of Containment 500 0.52 1.04 4.92 (29% DNB)

MSLB - Inside of Containment 500 0.76 1.43 4.91 (6. 1%FCM)

SGTR Pre-accident Iodine Spike 500 0.31 0.30 3.03 Acceptance Criteria <25 3) 2513) < 5_(4 SGTR Concurrent Iodine Spike 500 0.08 0.08 0.60 Locked Rotor (13.7 % DNB) 500 0.25 0.54 2.53 IOMSSV

• 500 0.02 0.02 0.30 Acceptance Criteria <2.5 (3) _ 2.5 (3) <5 (4)

FHA - Containment 500 0.53 0.52 1.23 FHA - Fuel Handling Building 500 0.53 0.52 3.02 CEA Ejection - Containment 500 0.26 0.50 2.74 Release (9.5 % DNB, 0.5 % FCM)

CEA Ejection - Secondary Side 0.29 0.63 2.60 Release (9.5 % DNB, 0.5 % FCM)

Acceptance Criteria <6.313 6.3 (3) 5(3)

') Worst 2-hour dose (2) Integrated 30-day dose (3) RG 1.183, Table 6 (4) 10CFR50.67

• see appropriate event summary in Section 2.0 for basis of acceptance criteria